ML20246E160
| ML20246E160 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 07/27/1989 |
| From: | BECHTEL POWER CORP. |
| To: | |
| Shared Package | |
| ML20246E158 | List: |
| References | |
| PROC-890727, NUDOCS 8908290023 | |
| Download: ML20246E160 (110) | |
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SAMDA ESTIMATE PROCESS AND COST ESTIMATE BREAKDOWN LIMERICK GENERATING STATION FOR j PHILADELPHIA ELECTRIC COMPANY PREPARED BY l
BECHTEL POWER CORPORATION POTTSTOWN, PA JULY 27,1989
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I SAMDA' ESTIMATING PROCESS BECHTEL LICENSING GATHERED 7 NUS ESTABLISHED DESIGN BASES AVAILABLE SAMDA INFORMATION BECHTEL CONCEPTUAL DESIGN TEAMS ESTABLISHED DESIGN CONCEPTS USING NUS INPUT, RDA REPORT AND OTHER AVAILABLE INFORMATION.
DESIGN CONCEPTS WERE REVIEWED BY PECO AND NUS AND REVISED AS REQUIRED. l l
i BECHTEL CONCEPTUAL DESIGN TEAMS WORKED WITH DISCIPLINE ENGINEERS, CONSTRUCTION ENGINEERS AND ESTIMATORS TO DEVELOP SCHEDULES, MATERIAL' QUANTITIES AND MANHOUR ESTIMATES.
PECo STATION REVIEWED THE DESIGN CONCEPTS AND ;
PROVIDED OWNER COST INPUT.
1 COST ESTIMATES WERE GENERATED AND A DRAFT REPORT ISSUED. 1 1
THE DRAFT REPORT WAS REVIEWED BY BECHTEL, NUS AND PECo.
THE FINAL REPORT WAS ISSUED.
SAMDAEST. PRO I
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m a I
l SAMDA ESTIMATE - GENERAL NOTES i i
Response to Item 11, Part I:
- a. The SAMDA estimates were prepared at the commodity level, and not by structure. Cost details are not currently available by structure. However, new structures were ident'ified where applicable.
- b. -A detail list of all estimated commodities has been provided with associated quantities and costs.
- c. Manual labor wage rates are based on avert.ge composite wage rates for the various trades including the cost of overtime premiums and training.
Nonmanual labor includes costs for field office personnel including field engineering, supervision, management, quality control, quality assurance, cost / schedule, and other departments.
- d. PECo's Nuclear Engineering Department costs were based on a percentage of Bechtel's engineering costs (Hours were not evaluated for PECo NED).
- e. Bechtel QA costs were not estimated separately, but were included in the nonmanual labor estimate figures.
PECo's QA costs are identified in section j.
- f. Health Physics costs are identified in section j.
Exposure estimates were not evaluated in these studies.
- g. Procedural costs are included within the various PECo departments listed in section j. Cost details are not currently available by this category.
- h. Training costs are included separately for General Employee Training of Bechtel's personnel. PEco's training costs are included within section j., but are not broken out separately.
- i. Replacement Power costs are identified where applicable.
- j. Other costs include PECo departmental costs, subcontract costs, AFUDC, and contingency.
l Response to Item 11, Part II:
- a. Yearly maintenance costs are identified by PECo l
department in present day dollars.
- b. No other recurring costs were evaluated.
_ - - - - _ - - _ _ _ _ _ _ _ - _ _ _ ~
a . e. e OPTION A1 PART1 INITIAL INVESTMENT-
' COST
~($ X 1000)
- A NEW STRUCTURES -CONTAINMENT HEAT REMOVAL STRUCTURE ---
$ 2,980 B; i WARE AND MATERIALS - DIRECT MATERIAL
' hAR_)
$ 1,628
- INDIRECT MATERIAL .
TOTAL $ 4,608 $ 4,608 (SEE ATTACHED UST OF COMMODITIES, OUANTITIES AND COSTS)
C. LABOR COSTS HOURS LABOR RATE ' COSTS DIRECT LABOR 325,562 - $ 32.31/hr. $ 10,519
- INDIRECT LABOR - 81,391 $ 27.00/hr. $ 2,198 TOTAL MANUAL LABOR 406,953 $ 12,717 NONMANUAL LABOR 130.225 $ 25.50/hr. $ 3,321 TOTAL LABOR' 537,178 $,16,038 $ 16,038 '
D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL. ENGINEERING 40,000 $ 2,600 OTHER HOME OFFICE ' 20,000 1,200 PECo NUCLEAR ENGINEERING -
$ 1.520 TOTAL ENGINEERING $ 5,320 $ 5,320 E. ' OA COSTS - included in: -
. Section C. Bechtel Nonmanual Labor Section J. PECo Other Costs
' F. HEALTH PHYSICS COSTS - locluded in:
Section J. PECo Other Costs (Note: Exposure Estimates were not evaluated)
G. PROCEDURAL COSTS - See Section J. PECo Other Costs 3
,: 'e-l OPTION A1 COSTS (S X 1000)
H. - TRA\NING COSTS - Bechtel General Employee Training is --
~ included in Section C. Direct Manual Labor ($ 552,000) ,
- PECo training is included in various --
departments listed in Section J (not broken out as a separate line item)
L 1. REPLACEMENT ENERGY COSTS - DAYS = 0 $0 COSTS = 0 EJ. OTHER COSTS
- 1) PECo COSTS -FIELD ENGINEERING 84 1 & C- 184.
- QA AUDIT 200 HEALTH PHYSICS 284 RADWASTE 520 TEST ENGINEERING 216 CONSTRUCTION SUPPORT 4,360 PECo MATERIAL 126 REGULATORY 1.330 7,304 $ 7,304
- 2) SUBCONTRACT COSTS S 230
- 3) AFUDC COSTS $ 6,611
- 4) CONTINGENCY & ROUNDING S 6,124 i
TOTAL PART I $ 46.235 PART 11 RECURRING COSTS A. MAINTENANCE - I&C 18 RADWASTE 200 -
TEST ENGINEERING 4 MAINTENANCE 300 TOTAL 522/yr. $ 522/yr.
B. OTHER RECURRING COSTS - N/A --
TOTAL PART !! $ 522/yr.
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OPTION A2 PARTI !NITIAL INVESTMENT :
' COST l'
($ X 1000)
A." NEW STRUCTURES . CONTAINMENT HEAT REMOVAL STRUCTURE _ -
i' B. HARDWARE AND MATERIALS - DIRECT MATERIAL . $ ' 3,314 '
- INDIRECT MATER %L' $ 2,037 TOTAL $ . 5,351 'S 5,351
-(SEE ATTACHED LIST OF COMMODITIES, QUANTITIES AND COSTS)
C. LABOR CCGTS ' HOURS LABOR RATE COSTS-DIRECT LABOR 407,434 $ 32.71/hr. $ 13.329 -
E!NDIRECT LABOR 101,858 ~ $ 27.00/hr. ~ $ 2,750 TOTAL MANUAL L. ABOR 509,292 $ 16,079 NONMANUAL LABOR . 162.973 $ 25.50/hr. $ 4,156
- TOTAL LABOR. 672,265 $ 20,235 $ 20.235
. D. - ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL ENGINEERING 56,900 $ 3,699 OTHER HOME OFFICE 28,450 1,707 PECo NUCLEAR ENGINEERING --
$ 2.162 TOTAL ENGINEERING $ 7,568 $ 7,568 E. CA COSTS Includedin: --
Section C. Bechtel Nonmanual Labor Section J. PECo Other Costs F.. HEALTH PHYSICS COSTS -included in: --
Section J. PECO Other Costs (Note: Exposure Estimates were not evaluated)
. G. . PROCEDURAL COSTS - See Section J. PECo Other Costs ---
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H. TRAfNING COSTS - Bechtel General Employee Training is -
- included in Section C. Direct Manual Labor (? R32,000)
- PECo training is included in various - - -
departments listed in Section J (not broken out as a separate line item)
- 1. REPLACEMENT ENERGY COSTS - DAYS = 28 S 23,800 COSTS = 23,800 J. OTHER COSTS
- 1) PECo COSTS -FIELD ENGINEERING 84
.1 & C - 228 OA AUDIT 200 HEALTH PHYSICS 580 RADWASTE 1600 TEST ENGINEERING 264 CONSTRUCTION SUPPORT 4,688 PECo MATERIAL 144 REGULATORY 1.892 9,680 $ 9,680
- 2) SUBCONTRACT COSTS $ 230-
- 3) AFUDC COSTS $ 13,390
- 4) CONTINGENCY & ROUNDING $ 7,778 TOTAL PART I S 88,033 PART 11 RECURRING COSTS A. MAINTENANCE - I&C 22 RADWASTE 200 TEST ENGINEERING 8 MAINTENANCE 300 TOTAL 530/yr. $ 530/yr.
B. OTHER RECURRING COSTS - N/A --
TOTAL PART 11 $ 530/yr.
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r y;( <p OPTION A3 PART'l INITIAL INVESTMENT COST -
($ X 1000)
' A: - NEW STRUCTURES - CONTAINMENT HEAT REMOVAL- STRUCTURE ---
B. HARDWARE AND MATERIALS - DIRECT MATERIAL $ 3,082 '
- INDIRECT MATERIAL $ 1,665 TOTAL $ 4,747 5 4,747 (SEE ATTACHED LIST OF COMMODITIES, OUANTITIES AND COSTS)
C. . LABOR COSTS HOURS LABOR RATE COSTS DIRECT LABOR _ 332,902 $ 32.26/hr. $ 10,741 INDIRECT LABOR ' 83,225 $ 27.00/hr. $ 2,247
-TOTAL MANUAL LABOR . -416,127 $ 12,988 NONMANUAL LABOR 133.161 $ 25.50/hr. $ 3,396 TOTAL LABOR 549,288 $ 16,384 $ 16,364 D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL ENGINEERING 48,500_ $ 3,153 OTHER HOME OFFICE 24,250 1,455 PECo NUCLEAR ENGINEERING - $ 1,843 TOTAL ENGINEERING $ 6,451 $ 6,451 E. OA COSTS - Included in: -
Section C. Bechte! Nonmanual Labor Section J. PECo Other Costs F. HEALTH PHYSICS COSTS - included in: -
i Section J. PEco Other Costs (Note: Exposure Estimates were not evaluated) i G. PROCEDURAL COSTS - See Section J. PECo Other Costs -
l-l7 L__. _ _ . _ _ _ _ _ _ _ _ . ___u__
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OPTION - A3 COSTS
($ X 1000)-
'H. TPAINING COSTS - Bechtel General Employee Training is -
. included in Section C. Direct Manual Labor ($ 564,000)
- PECo training is included in various -
departments listed in Section J (not broken out as a separate line item)
- 1. REPLACEMENT ENERGY COSTS - DAYS = 0 $ 0 COSTS = 0 -
J. OTHER COSTS
- 1) PECo COSTS -FIELD ENGINEERING 84 I&C 202 OA AUDIT 200.
l HEALTH PHYSICS 290 RADWASTE 571 TEST ENGINEERING 246 i CONSTRUCTION SUPPORT- 4,688 PECo MATERIAL 134 REGULATORY 1.613 8,028 $ ' 8,028
- 2) SUBCONTRACT COSTS $ 226 l
- 3) AFUDC COSTS S 6,663 l
- 4) CONTINGENCY & ROUNDING $ 6,535 TOTAL PART l $ 49,034 PART ll RECURRING COSTS A. MAINTENANCE - 1&C 20 RADWASTE 200 TEST ENGINEERING 6 MAINTENANCE 300 TOTAL 526/yr. $ 526/yr.
l' l B. OTHER RECURRlNG COSTS - N/A --
TOTAL PART 11 S 526/yr.
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t )e ~ :--- Ol : } OPTION 81-L -PARTI: INITIAL INVESTMENT ' L COST
. ($ X 1000)
' A.- .NEW STRUCTURES NONE, --.
V L-B. HARDWARE AND MATERIALS - DIRECT MATERIAL ' $ 416 INDIRECT MATERIAL - $ 189 TOTAL $ 605 $ 605 J (SEE ATTACHED LIST OF COMMODITIES, QUANTITIES AND COSTS) C.. LABOR COSTS HOURS' LABOR RATE COSTS DIRECT LABOR 37,781 $ 33.38/hr. $ 1,261 INDIRECT LABOR 9,445 $ 27.00/hr. $ 255
- TOTAL MANUAL LABOR - 47.226 $ 1,516 NONMAN'UAL LABOR 15.112 $ 25.50/hr. $ 385-
. TOTAL LABOR 62,338 $ 1,901 5' 1,901 D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL. . ENGINEERING 11,600 $ 754 OTHER HOME OFFICE 5,800 348 PECo . NUCLEAR ENGINEERING - $ 441 TOTAL ENGINEERING $ 1,543 - S 1,543 E. CA' COSTS - Included in: -
Section C. Bechtel Nonmanual Labor Section J. PECO Other Costs F. HEALTH PHYSICS COSTS -includedin: - Section J. PECo Other Costs (Note: Exposure Estimates were not evaluated) G. PROCEDURAL COSTS See Section J. PECo Other Costs - _ .J
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~ OPTION -' B1 COSTS
($ X 1000) 1 H. ' TRA\NING COSTS - Bechtel General Employee Training is --
- . included in Section C. Direct
' Manual Labor ($ 76,000) .l
- PECo training is included in various ---
departments listed in Section J (not broken out as a separate line item)
- 1. REPLACEMENT ENERGY COSTS - DAYS = 0 'S 0 COSTS = 0 J. OTMER COSTS
- 1) PECo COSTS -FIELD ENGINEERING 14 i&C 56 QA AUDIT - 30 HEALTH PHYSICS 66 RADWASTE 64 TEST ENGINEERING 134-CONSTRUCTION SUPPORT 600 PECo MATERIAL - 50 REGULATORY 386 1,400 $ 1,400
- 2) SUBCONTRACT COSTS S O'
- 3) AFUDC CO?T9 $ 712
- 4) CONTINGENCY & ROUNDING S 892 TOTAL PART I $ 7,053 PART ll RECURRING COSTS A. ' MAlfRENANCE - I&C 6 RADWASTE 40 TEST ENGINEERING 8 MAINTENANCE __. 20 TOTAL 74/yr. S 74/yr.
B. OTHER RECURRING COSTS - N/A -- TOTAL PART 11 5 74/yr. i o i
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4, .-.. OPTION B2 PARTI INITIAL INVESTMENT ' COST ($ X 1000)
, Af NEW STRUCTURES GRAVEL BED FILTER STRUCTURE ---
l B. HARDWARE AND MATERIALS - DIRECT MATERIAL $ 2,045
- INDIRECT MATERIAL ' S- 705 TOTAL $ 2,750 .$ 2,750 (SEE ATTACHED LIST OF COMMODITIES, QUANTITIES AND COSTS)-
- C, LABOR COSTS HOURS LABOR RATE COSTS
~ DIRECT LABOR' 140,991 $ 32.36/hr. $ 4,563 INDIRECT LABOR - 35,248 $ 27.00/hr. $ 952-TOTAL MANUAL' LABOR 176,239 $ - 5,515
' NONMANUAL LABOR 56 396 $ 25.50/hr. $ 1.,438 TOTAL LABOR 232,635 $ 6,953 $ 6,953 -
D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL ENGINEERING 19,100 $ 1,242 - OTHER HOME OFFICE 9,550 573 PECo NUCLEAR ENGINEERING -
$ 726 TOTAL ENGINEERING 28,650 $ 2,541 $ 2.541
' E, CA COSTS -included in: -
Section C. Bechtel Nonmanual Labor Section J. PECo Other Costs F. HEALTH PHYSICS COSTS - locluded in: -- Section J. PECo Other Costs (Note: Exposure Estimates were not evaluated) G. PROCEDURAL COSTS - See Section J. PECo Other Costs -- 50
OPTION B2 COSTS ($ X 1000) H. TRAINING COSTS - Bechtel General Employee Training is -
~ included in Section C. Direct Manual Labor ($ 282,000)
- PECO training is included in various --
departments listed in Section J
- ' (not broken out as a separate line item)
- 1. REPLACEMENT ENERGY COSTS - DAYS = 0 S 0 COSTS = 0 J. OTHER COS,TS
- 1) PECo COSTS FIELD ENGINEERING 14 i&C 56 OA AUDIT 100 HEALTH PHYSICS 86 RADWASTE 260 TEST ENGINEERING 168 CONSTRUCTION SUPPORT 2,398 i PECo MATERIAL 36 REGULATORY 635 3,753 $ 3,753
- 2) SUBCONTRACT COSTS S 0
- 3) AFUDC COSTS S 2,917
- 4) CONTINGENCY & ROUNDING S 2,882 TOTAL PART I $ 21,796 PART ll . RECURRING COSTS 1
L. A. MAINTENANCE - I&C 6 l RADWASTE 40 TEST ENGINEERING 12 MAINTENANCE 30 l TOTAL 88/yr. S 88/yr. B. OTHER RECURRING COSTS - N/A - TOTAL PART 11 S 88/yr. 2 'i
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s.. 1 OPTION B3
.PART1 INITIAL INVESTMENT . j COST
($ X 1000)- ;
- A. , NEW STRUCTURES . MULTIPLE VENTURI SCRUBBER STRUCTURE- ---
J I L: . l s - B. HARDWARE AND MATERIALS - D'IRECT MATERIAL $ 1.340
- INDIRECT MATERIAL $ 257 l' _
- TOTAL $ 1,597 $ 1,597 (SEE ATTACHED LIST OF COMMODITIES, QUANTITIES AND COSTS)-
A . C. . LABOR COSTS HOURS LABOR RATE COSTS DIRECT LABOR 51,414. $ 32.89/hr. $ - 1,691 INDIRECT LABOR 12,854 $ 27,00/hr. $ 347 TOTAL MANUAL LABOR 64,268 $ - 2,038 NONMANUAL LABOR 20.566 $ 25.50/hr. $ 524 TOTAL' LABOR 84,834 $ - 2,562 $ 2,562 D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL ENGINEERING 14,500 $ 943 OTHER HOME OFFICE 7.250 435 PECo NUCLEAR ENGINEERING - S 551 TOTAL ENGINEERING 21,750 $ 1,929 $ 1,929 E.. CA COSTS - included in: --- Section C. Bechtel Nonmanual Labor Section J. PECo Other Costs F. ' HEALTH PHYSICS COSTS -included in: - Section J. PECo Other Costs (Note: Exposure Estimates were not evaluated) G. PROCEDURAL COSTS - See Section J. PECo Other Costs -
p
;- .3 7
I p OPTION 83
. . COSTS (S .X 1000) .
H. TRA\NING COSTS Bechtel General Employee Training is --
~ included in Section C. Direct Manual Labor ($ 103,000)
- PECO training is included in various --
i- departments listed in Section J (not broken out as a separate line item)
- 1. ' REPLACEMENT ENERGY COSTS . DAYS = 0 $ 0 COSTS = 0 J. OTHER COSTS
- 1) PECO COSTS -FIELD ENGINEERING 14 l&C 29 OA AUDF 50 HEALTH PHYSICS 44 RADWASTE 132 TEST ENGINEERING 168 CONSTRUCTION SUPPORT 700 PECo MATERIAL 36 REGULATORY 482 1,655 $ 1,655
- 2) SUBCONTRACT COSTS S 0
- 3) AFUDC COSTS S 1,445
- 4) CONTINGENCY & ROUNDING S 1,383 TOTAL PART l $ 10,571
- PART 11 RECURRING COSTS A. MAINTENANCE - l&C 3 RADWASTE 40 TEST ENGINEERING 12 MAINTENANCE 30 TOTAL 85/yr. S 85/yr.
B. OTHER RECURRING COSTS - N/A -- TOTAL PART ll $ 85/yr. l 5A _ = _ - - _ . _ - _ _ _ _
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PARTI- INITIAL INVESTMENT. COST ($ .X 1000)
' A. ; NEW STRUCTURES .. N0NE ,
' B .' HARDWARE AND MATERIALS - DIRECT MATERIAL -$ 180 INDIRECT MATERIAL $ 126 TOTAL $ 306 $ 306 (SEE ATTACHED LIST OF COMMODITIES, QUANTITIES AND COSTS)
C. LABOR COSTS HOURS LABOR RATE COSTS DIRECT LABOR 25,277 $ 33.83/hr. $ 855 INDIRECT LABOR 6,319 $ 27.00/hr. $ 171 TOTAL MANUAL LABOR 31,596 $ 1,026 NONMANUAL LABOR ' 10.111 $ 25.50/hr. $ 258 TOTAL LABOR - 41,707 $ 1,284 $ . 1,284 -
' O. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL ENGINEERING 10,180 $ 662 OTHER HOME OFFICE 5,090 $ 305 PECo. ' NUCLEAR ENGINEERING -- $ 387 TOTAL ENGINEERING $ 1,354 $ 1,354 E. OA COSTS Included in: -
'Section C. Bechtel Nonmanual Labor Section J. PECO Other Costs F. HEALTH PHYSICS COSTS - included sn: -
Section J. PECo Other Costs (Note: Exposure Estimates were not evaluated) G. PROCEDURAL COSTS '. See Section J. PECo Other Costs - 3r
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' OPTION B4 COSTS-($ X 1000)
H. TRAINING COSTS Bechtel General Employee Training is --
- jncluded in Section C. Direct Manual Labor ($ 51,000)
- PECo training is included in various --
l_ departrnents listed in Section J l (not broken out as a separate
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- 1. REPLACEMENT ENERGY COSTS - DAYS = 0 S 0-COSTS = 0 J. OTHER COSTS -
1)- PECo COSTS -FIELD ENGINEERING - 14 i&C 56 QA AUDR 50 HEALTH PHYSICS 45 RADWASTE 42 TEST ENGINEERING 160 CONSTRUCTION SUPPORT 550 PECo MATERIAL 41 REGULATORY E 1,296 S 1,296 l l
- 2) _- SUBCONTRACT COSTS S 0
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- 4) CONTINGENCY & ROUNDING $
TOTAL PART I $ 5,432 PART 11 RECURRING COSTS A. MAINTENANCE - I&C 6 RADWASTE 40 TERT ENGINEERING 12 MAINTENANCE 30 TOTAL 88/yr. S 88/yr. B. OTHER RECURRING COSTS - N/A --- TOTAL PART ll $ 88/yr. l' l 37 i 1
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..A. NEW STRUCTURES . DIESEL GENERATOR ENCLOSURE -
B. ' HARDWARE AND MATERIALS - DIRECT MATERIAL $ 454
- INDIRECT MATERIAL $ 243 TOTAL $ 697 $ 697-(SEE ATTACHED LIST OF COMMODITIES, QUANTITIES AND COSTS)
, C. LABOR COSTS HOURS LABOR RATE COSTS DIRECT LABOR . 48.672 $ 32.24/hr. $ 1,569 INDIRECT LABOR. 12,168 $ 27.00/hr. $ 329 TOTAL MANUAL LABOR 60,840 $ 1,898
.NONMANUAL LABOR 19.469 $ 25.50/hr. $ 496
-- TOTAL LABOR 80.309 $ 2,394 5 2,394-D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL: ' ENGINEERING 19,800 $ 1,287 OTHER HOME OFFICE 9,900 594 PECo- NUCLEAR ENGINEERING - $ 752 l- TOTAL ENGINEERING $ 2,633 $ 2,633 E. OA COSTS - Included in: -
Section C. Bechtel Nonmanual Labor
- Section J. PECo Other Costs F. . HEALTH PHYSICS COSTS - locluded in; -
Section J. PECo Other Costs (Note: Exposure Estimates were not evaluated)' G.' PROCEDURAL COSTS - See Section J. PECO Other Costs dE i ~ L - - - - - - - . - _ - - - - _ _ _ _ _ _ - _ _ _ _ _
i OPTION C1 COSTS (S X 1000) H. TRAINING COSTS - Bechtel Genera' Employee Training is -
~ Jncluded in Section C. Direct Manual Labor ($ 7C,000)
- PECo training is included in various --
departments listed in Section J (not broken out as a separate line item)
- 1. REPLACEMENT ENERGY COSTS - DAYS = 0 S 0 COSTS = 0 J. ' OTHER COSTS
- 1) PECo COSTS FIELD ENGINEERING 52
'I & C - 56
~ OA' AUDIT 0 l HEALTH PHYSICS ' 68 RADWASTE 94 TEST ENGINEERING 104
-- CONSTRUCTION SUPPORT 510 PECo MATERIAL 144 l REGULATORY 658 1,686 S 1,686 SUBCONTRACT COSTS S 16
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- 3) AFUDC COSTS S- 1,000
- 4) CONTINGENCY & ROUNDING S 1.212 l.
l TOTAL PART I S 9,639 PART 11 RECURRING COSTS A. MAINTENANCE - 1&C 6 RADWASTE 40 TEST ENGINEERING 16 MAINTENANCE 29 TOTAL 82/yr. S 82/yr. B. OTHER RECURRING COSTS N/A -- TOTAL PART 11 S 82/yr. 43
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h. r-p p OPTION _ D1 PART1- INITIAL INVESTMENT COST (5 X 1000) . A. NEW STRUCTUAES : DRY CAUCIBLE CORE DEBRIS RETENTION STRUCTURE ---
- UNDERGROUND COOLING STRUCTURE
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B. HARDWARE AND MATERIALS . DIRECT MATERIAL $ 10,094
- INDIRECT MATERIAL $ 1,807 TOTAL $ 11,901 $ 11,901
' (SEE ATTACHED UST OF COMMODITIES, OUANTITIES AND COSTS)
C. LABOR COSTS HOURS LABOR RATE COSTS DIRECT LABOR 361,393 S 34.16/hr. $ 12,345 INDIRECT LABOR 90.348 $ 27.00/hr. $ ' 2,439 TOTAL MANUAL LABOR 451,741 $ 14,784 NONMANUAL LABOR 144.557 $ 25.50/hr. $ 3,686
- TOTAL LABOR 596,298 $ 18,470 ' $ 18,470 D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL ENGINEERING 67,000 $ 4,355 OTHER HOME OFFICE - 33,500 2,010 PECo NUCLEAR ENGINEERING -
$ 2.546 TOTAL ENGINEERING $ 8,911 $ 8,911 E. OA COSTS - Included in:
Section C. Bechtel Nonmanual Labor Section J. PECo Ott,er Costs F. HEALTH PHYSICS COSTS -included in: Section J. PECO Other Costs (Note: Exposure Estimates were not evaluated) G. PROCEDURAL COSTS - See Section J. PECo Other Costs de,
2.. :.: F OPTION D1 COSTS ($ X 1000) H. TRAINING COSTS - Bechtel General Employee Training is ' ---
~ included in Saction C. Direct Manual Labor ($ 531,000)
- PECO training is included in various --
departments listed in Section J (rot broken out as a separate line item) L REPLACEMENT ENERGY COSTS - DAYS = 147 5 124,950 COSTS = 124,950
.J. OTHER COSTS
- 1) PECo COSTS -FIELD ENGINEERING 7 i&C 112 OA AUDIT 200 HEALTH PHYSICS 889 RADWASTE 2,400 TEST ENGINEERING 270 CONSTRUCTION SUPPORT 6,330 PECo MATERIAL - .126 REGULATORY gj@
12,562 S 12,562
- 2) SUBCONTRACT COSTS S 13,760 AFUDC COSTS $ 17,876 3)'
CONTINGENCY & ROUNDING S 25,204 4) TOTAL PART l $ 233,634
.PART11 RECURRING COSTS A. MAINTENANCE - I&C 11 RADWASTE 170 TEST ENGINEERING 6 MAINTENANCE - 220 TOTAL 407/yr. $ 407/yr.
B. OTHER RECURRING COSTS - N/A TOTAL PART 11 $ 407/yr. Mk 1
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PARTl' INITIAL INVESTMENT COST- .
)
($ X 1000). l A." : NEW STRUCTURES -NONE - B. : HARDWARE AND MATERIALS - DIRECT MATERIAL - $ 1,930
- INDIRECT MATERIAL $ .728-TOTAL $ 2,658 $ 2,658 (SEE ATTACHED LIST OF COMMODITIES, QUANTITIES AND COSTS)
C. LABOR COSTS HOURS LABOR RATE COSTS DIRECT LABOR : 145,662- $ 34.61/hr. $ 5,042 ' INDIRECT LABOR - 36,415 ' $ 27.00/hr. $. 983 TOTAL MANUAL LABOR ' 182,077 $ 6,025 NONMANUAL LABOR 58.265 $ 25.50/hr. $ 1,486 TOTAL LABOR 240,342 5 7,511 $ 7,511 D. ENGINEERING / DESIGN COSTS HOURS COSTS BECHTEL~~ ENGINEERING 23,100 $ 1,502 OTHER HOME OFFICE 11,550 693 PECo NUCLEAR ENGINEERING ' -
$ 878 TOTAL ENGINEERING - $ 3,073 $ 3,073 E. CA COSTS - Included in: -
Section C. Bechtel Nonmanual Labor Section J. PECo Other Costs
.F. ' HEALTH PHYSICS COSTS - included in: -
Section J. PECo Other Nsts (Note: Exposure Estimates were not evaluated) G. PROCEDURAL COSTS - See Section J. PECo Other Costs - 64-
OPTION D2 COSTS .
- ($ X 1000)
H. '. TRAINING COSTS - Bechtel General Employee Training is -
~ included in Section C. Direct Manual Labor ($ 219,000)
- PECo training is included in various -
departments listed in Section J (not broken out as a separate line item)
- 1. . REPLACEMENT ENERGY COSTS - DAYS = 56 ' $ 47,600 COSTS = 47,600 J. OTHER COSTS
- 1) PECo COSTS -FIELD ENGINEERING 0
-l&C 48 QA AUDIT 100 HEALTH PHYSICS 166 RADWASTE 2,320 TEST ENGINEERING 200=
CONSTRUCTION SUPPORT 2,164 PECo MATERIAL 126 REGULATORY 768 5,892 $ 5,892
- 2) SUBCONTRACT COSTS $ 900 3)L AFUDC COSTS $ 3,689 4)- CONTINGENCY & ROUNDING $ 4,636 -
TOTAL PART I $ 75,958 PART11 RECURRING COSTS A. MAINTENANCE - I&C 5 RADWASTE 40-TEST ENGINEERING 4 MAINTENANCE 30 TOTAL 79/yr. $ 79/yr. B. OTHER RECURRING COSTS - N/A - TOTAL PART 11 5 79/yr. l TC L __-- _ __ _ _ _ - - - - - - - _ __ _
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l: l ENCLOSURE 3 l l ACCIDENT CLASS FREQUENCY BY INITIATOR l (PER YEAR) 1
~
TOTAL CLASS INTERNAL FIRE FLOOD * (TABLE 2-2) l 1 4.44E-06 .4.2E-06 2E-07 8.8E-06 lc 2 1.42E-07 3.3E-08 - 1.7E-07 3 2.73E-07 - - 2.7E-07 4 1.05E-06 - - 1.1E-06 IS - - - S 1.OE-08 - - SUBTOTAL 5.91E-06 4.2E-06 2E-07 - l TOTAL 1.03E-05
- Includes Other Special Initiators 1-1
-o -
i
CORE DAMAGE FREQUENCY BY INITIATOR (INTERNAL INITIATORS ONLY) CORE DAMAGE INITIATOR FREOUENCY % CONTRIBUTION Transients 2.16E-06 36.5 Loss of Condenser Vacuum (cdf=1.03E-06) Turbine Trip (cdf=2.81E-07) MSIV Closure (cdf=4.74E-07) Manual Shutdown (cdf=1.95E-07) Loss of Feedwater (cdf=1.50E-07) IORV Event (cdf=2.48E-08) TE-Loss of Offsite Power 2.32E-06 39.3 Station Blackout (cdf=1.42E-06) Common cause Failure of (cdf=3.79E-07) Batteries Support Stato TE1 (cdf=2.71E-07) Support State TE4 (cdf=8.87E-08) Support State TE2 (cdf=8.21E-08) Support State TE3 (cdf=7.60E-08) ATWS Secuences 1.17E-06 19.8 Turbine Trip (cdf=3.77E-07) Loss of Condenser Vacuum (cdf=3.75E-07) MSIV Closure (cdf=2.40E-07) IORV (cdf=8.56E-08) Loss of Offsite Power (cdf=7.83E-08) Loss of Feedwater (cdf=1.75E-08) FW 1.58E-07 2.7 Medium LOCA (cdf=1.09E-07) Large LOCA (cdf=4.45E-08) Small IOCA (cdf=4.45E-09) Random Vessel Rupture 1.0E-07 1.7 5.91E-06 100.0
/-2 m
9-i
. COMPARISON OF CLASS FREQUENCIES INTERNAL INITIATORS FREQUENCY (PER YEAR)
CLASS PRA/ SARA CURRENT REASONS FOR CHANGES
'1 1.2E-05 4.44E-06 EOPs, Training, LOOP Modeling, Initiator Frequency, ADS Modifica-tion..
2 9.6E-07' 1.42E-07 Plant Performance and Data, Initiator Frequency, Venting 3 1.1E-06 Initiator Frequency, EOPs, 2.73E-07)4 Revised Modeling, Lowering 4 1.3E-07' IV Closure Set Point S 2.7E-08 1.0E-08
/(076-4Mean/ Median, Not Included in NUREG-1150.
i 1-3 l _ __ i
.$ 4 COMPARISON OF CLASS FREQUENCIES FIRE INITIATORS FREOUENCY'(PER YEAR)
CLASS PRA/ SARA CURRENT REASON FOR CHANGES l' 2.5E-06 4.2E-06' Plant Design, New Initiator and Suppression Data, New Plant Model
'2 9.3E 3.3E-08 As Class 1 Plus Venting 3 - -
. 4~ - -
g _ _ 1-4 a _ - -____- _ _ _ _ - - - _ _ _ _ _ - _ - _ _ _ _ n
a a t l l- 1 COMPARISON OF CLASS FREQUENCIES ) I FLOOD AND OTHER INITIATORS I F'REOUENCY (PER YEAR) CLASS PRA/ SARA CURRENT
- REASON FOR CHANGE
. FLOOD OTHER l 1 <5E-07 BE-08 9E-08 E1imination of Conservatism 2 <7E-08
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- 2E-07 Used 1-5 x - _ _ __ _ _-_____-_____ -__ ____
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UNCERTAINTY o NOT O,0NE FOR UPDATE o WAS DONE'IN SARA o ESTIMATED BY BNL IN NUREG/CR-3028 o NUREG-1150 PEACH BOTTOM ANALYSIS l 2-1 l _- _ _ - _ _ _ _ _ - _ _ _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ._ . _ _ _ .- _ _ - _ - _-______________x
*f.: ,
k ,i . UNCERTAINTYiESTIMATES f NUREG-1150 o SAB&' NUREG/CR-3028 '2ND DRAFT. INTERNAL 6.5/3.8 '8.9/5'.6 6.8/5.4 25 50- ' FIRE 8.'6/8.2 - li3/11
.70 58-
\ ;, KEY: Ratio 95%-to Median / Ratio Median to 5%'~ Ratio 95% to 5% 2-2 1. 1 'SU _
BNL AREAS OF CONCERN IN ACCIDENT SEQUENCE QUANTIFICATION
- 1. Deficiencies in Incorporation of Dependencies in the various Types of-Logic Trees
- 2. Disagreement With Some System Unavailabilities and Other Event TTwe Values l
- 3. Differences in Frequencies of Initiating Events
~
INCORPORATION OF DEPENDENCIES
- Impace of Dependencies' Introduced by. Support Systems Servicing Multiple Frontline Systems
- Impact of Dependencies Introduced by Hardware Shared
/ Among Frontline Systems b l 1
- Dependence Between Q and W Functions
- Dependence Between Q Function and MSIV Closure Initiator
- Dependence Between U and W Functions
- Vapor Suppression Function l
h
a- . FREQUENCY OF TRANSIENT INITIATING EVENTS ORIGI!J AL Hilk UPDATED Turbine Trip
~
3.98 8.17 5. 6
- MSIV Closure 1.78 L.23 0.23 Loss of Offsite Power 0.05 0.17 0.074 IORV- 0.07 0.25 0.07 1
Manual Shutdowns 3.2 3.2 3.2 Loss of Feedwater 0.19 Loss of Condenser Vacuum 0.38 TOTAL 9.08 13.02 9.75
- Revised June 1949 ** 2 ,5 5 4 . re.f les.+ L Gs eyrience
.n -_ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ - _ - - _ _ _ _ - - - _ _ _ _ - _ _ _ _ - - _ _ _ _ _ _ .
+ ,
-Table LI .
SUMMARY
OF THE BNL COMMENTS 'ON THE 1981 LGS PRA AND THEIR RESOLUTION IN THE UPDATED LGS PRA~ COMMENTS RESOLUTION e The success criteria used in the IES No Action; These, have. keen aka.3ed PRA represent reilistic requirements and ' g g j # ,# f,, ,, they do not conespond to safety analysis a = c ce,s f / .ma44.J e# report (SAR) criteria. The criteria were developed from analysis contamed in . con 4a6 men + .ke*+ -a f. NEDO-24708. o The success criteria for the This remams an open item. NEDO 24708 transient imnators are considered indicates RCIC is adequate for injection reasonable based on the content of until the RPV is depressunzed: however, this dommaat, - except for the the Ooen Item that renuins is that the assumption that RCIC ' is capable event trees do not require L.P. injection
. of supplying adequate _ vessel water following successful H.P. injection for the makeup to an isolate reactor with SORV.
an SORV. [pg. 2-5] o Additionally, the success c.seria No Action: This has been performed as for. ATWS scenarios . require part of the GE. design record file and-verification to detennine the have since been'used in other PRAs. adequacy of each system or function for mitigating these events. [pg. 2-5] The partitioning of transient initiators into four groups was reviewed and considered acceptable. Specifically, o The treatment of initiating events No Action: Loss of Eeedwaiv- **d in the LOS PRA was more g , ,, , # c, clea se,- vacuum realistic than in the RSS and 3.p.,,4 J 4.., g szy ./.m Grand Gulf RSSMAP. bec as. .O diFferu+ cAa llea3e.s fe a,,d re spon s e. . C + A a. pl a i+. o Some imtiators included in the Big No Action: These have been added in Rock PRA were not explicitly the updated PRA. treated, i.e.,
- Loss of instrument air
- Interfacing LOCAs
- Steam line break outside contamment
+ t4 e ac;+ en fo ,- +6;s 4a k le. Nec44 e.s ms.tu+ t en is c,,,,pl e+e a,.d so % + hey ae.45en is needed.
S-728510-012 I-3
)
032389D89F _____ - - _ \
m Table L1 (con't)
SUMMARY
OF THE BNL COMMENTS ON THE 1981 LOS PRA AND THEIR RESOLUTION IN THE UPDATED LGS PRA COMMENTS RESOLUTION' o ' Additionally, the. following
. initiators , developed by the y
reviewers were . not explicitly 1 addressed in the LOS PRA:
- ' Loss of DC power - No Action: Now included in-updated PRA
- RCP seal fahure following - No Action: PWR issue an SBO
- Pipe breaks 'in aux 211ary - No Action: Floodmg examined in buildings and instrument updated PRA tube LOCA
- Scarc discharge volume - No Action: Exammed by NRC LOCA and GE on Generic Basis; frequency much less than IE.6/yr.
- Loss of component cooling. - No Action: Incorponted in loss water of SW
- Loss of instrument and - No Action: Incorporated in loss control power of DC However, the reviewers stared that the No Action: Verified by updated PRA initiating events not treated in the LGS would not significantly affect the total core damage bequency. [Pg. 2-10]
PRA neglected potennally No Action: The linhng of fault nees The LGS the subsequent performance . of important We in the accident and Boolean manipulations of the resulting sequence quanrification process. 'I bis is due to the fact that the functional fault expanded use account for system and trees were used in isolation to quannfy functional dependencies in the Level I the probability of failure of the PRA. corre= pan a ng functions. [Pg. 3-10] I4 S-728510 012 l 032389D89F 3
- , . = _ . = . - ___ . . . . -..-..---..- .
,' Tcble L1 (con't).
SUMMARY
OF THE BNL COMMENTS ON THE 1981 LOS PRA AND THEIR RESOLUTION IN,THE UPDATED LGS PRA-COMMENTS RESOLUTION The impact of the amission of fault nees for the following systems was not evaluated, . but determmed to be potentially important due to the systen L interdependence with fit >ntline systenu: [Pg. 3-12]- RPS - No Action: RPS fault tree added.
- Plant air - No Action: Dismissed - on judgement that plant air is not a major support systems.
- Turbine enclosure cooling - No Action: Included in Loss of water SW for updated PRA.
Reactor enclosure cooling - No Action: Included in Loss of ) water SW for updated PRA. Generally, the fault trees appeared to the reviewers as being complete and accurate. However, BNL revised some models. These changes are described in Table L2. [Pg. 3-12] BNL disagreed with the value used for No Action: BNL has since changed this the probability associated with the estimate; simulator data by NRC/RMIEP cognitive human error involving failure to (NUREG/CR 4834) also supports the use depressunze the RPV (event 'X'). [p. 3-7] of the origmal PRA estimate and even lower values. The LOS PRA has used an HEP derived from a sophisticated HEP model and has performed sensitivity studies to confirm the contribution to uncertainty. i S-728510-012 15 032389D89F
) y 4
h p Table. L1 (con't)- SUMMiRY OF THE BNL COMMENTS ON THE 1981 LGS'PRA AND THEIR RESOLITUON IN THE UPDATED LGS PRA -
' COMMENTS - R' ESOLUTION-Particular cognitive . human error probabilities which were modeled in the LGS PRA' fault trees. adjusted by BNL include: [Pg. 3-17]-
FpmWATER
- Failure' of the operator to No Action: Duplicate event identified by reset ' and restan the .FW ^ ' BNL has been deleted.-
system
-- Failure of the operator to Qs.q: Values not changed close RFPT steam exhaust
- butterfly valves F
- Failure of the operator for Q2e.g: . Values not changed bypassing a . failed sealing steam pressure regulation .
. ADS.
- Failure of the ~ operator to - No' Action: Value conservatively set "to line up instrument air to the 0.1.
- ADS valves EIE l
- Failure of the operator to Qge_D: Not Included open common valves MOV-67A and MOV-67B.
BNL idenn5ed cognitive human errors leading to common mode failures which were not included in the system fault tree models: [Pg. 3-17] S-728510-012 I-6
' 032389D89F-
-- _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ _ _ - - _ _ _ - - _ - _ - _ _ _ _ _ _ _ ____-_--____-_--__--__-_____-_____x
._= _
7_ _ - - , ? Table I.1 (con't)
SUMMARY
OF THE BNL COMMENTS ON THE 1981 LOS PRA AND THEIR RESOLUTION IN THE UPDATED LGS PRA COMMENTS RESOLUTION
~
FEEDWATER Operator fails to start No Action: Failure to start mechanical mechanical vacuum pump vacuum pump is included with same given SJAEs fail to probability suggested by BNL (0.02). maintain condenser vacuum. AD.S.
- Miscalibration of core No Action: Instruments are different for sprays and RHR pump each set of valves and are not judged to discharge pressure sensors. have a substantia: common cause failure.
BNL change not incorporated because common cause is judged to be most applicable within the RHR system and within the CS system. This latter is accounted for. SkG
- Miscalibration of tank level No Action: Common mode miscalibation sensor of tank level sensors is included, probability of failure is lE-3.
System dependence between functions No Action: The updated LOS PRA has were not always addressed (e.g., functions used Imked fault trees to explicitly model Q and W both include the PCS system the commonalities between systems and and functions V and W include the LPCI functions. and the RHR systems, respectively, which share some hardware). [Pg. 3-21] l Some functional dependencies were o Even+ +re.e.s a*+ omuted from the LGS PRA model (e.g., e % pen: s d. dependence of the HPCI and RCIC systems on the suppression pool temperature). [Pg. 3-21] Dependencies of frontline systems on No Action: Common dependent failure support systems were not " carried over" modes affecting multiple systems are across functions included in the linked fault tree scheme of the updated LOS PRA. Se728510-012 I-7 032389D89F
1 . e l Table L1 (con't)
SUMMARY
OF THE BNL COMMENTS ON THE 1981 LOS PRA AND THEIR RESOLUTION IN THE UPDATED LGS PRA COMMENTS RESOLUTION System _ physical- dependencies were No Action: Considered by BNL to be ' covered only margmally in the LOS PRA. outside the scope of the PRA. [Pg. 3-22] Component physical dependencies were not included in the 1 PRA. [Pg. 3-23] Component functional dependencies were No Action: Lmked fault trees are being not included in the PRA. [Pg. 3-22] used in the updated LGS PRA - . to explicitly account for the component functional dependencies. The vapor suppression function as used in hio A c,+ e n: LO C. A even+ +re.as the RSS was not included in te LOCA chan3ed +o incJade, vapr sup-event trees. [Pg. 3-23] pre.ssion fuu+ ion. The emergency coolant function ability No Action: - WASH-1400 and subsequent was not included in the large LOCA BWR PRAs have concluded that this event tree. [Pg. 3-23] event is not appropriate. ) The . frequencies of the initiating events No Action: The initiating frequencies detennined by the BNL approach differ, have been updated using the latest as shown in Table 4.1 from those derived available data, but editing out the first in the PRA. [Pg. 4-5] year of commercial operation. This has been extensively discussed and used in current BWR and PWR PRAs. The probability of the common cause No Action: The failtue probabilities of failure of all four diesels used in the diesels have been reviewed and revised PRA was IE-3, whereas, BNL calculated 1.9E-3. [Pg. 4-7] The value of 2E-3 used for the "X" event No Action: See discussion of "X" values in the PRA was regarded by BNL to be supported by simulator data. optimistic. [Pg. 4-8] S-728510-012 I-8 032389D89F
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f p. RELEASE / CONSEQUENCE MODELS I. RELEASE FRACTION o Release fraction (source term) calculations in 1982 PRA.
-o One representative sequence per accident class.
o- In-Plant P-T conditions from INCOR (INCOR = BOIL + PVMELT + INTER + COMTEMPT-LT) o Release fractions from CORRAL (Wash-1400) using INCOR data, and various containment failure modes. 4-1
..A t. _ _. . _ _ . - - . _ _ _ _ _ _ _ . . _ _ . _ . - . _ - _ _ - _ _ _ - _ _ . - - . _ _ . _ . _ - - - _ _ _ _ _ _ _ - _ _ _ _ . - - - _ - - - - - - - - - - _ _ _ - - _ - - . . _ _ J
!. 4 .- 4 . l - RELEASE / CONSEQUENCE MODEL8
-II. CONSEQUENCES
.3.
! o Conifeguence calculations in 1987 SARA. o- Consequence results from CRAC2.
-PRA/ CORRAL release f.; actions 2 Containment- failure nodes based on 1987 PRA.
o Consequence characterized as accident . class occurence
' conditional.
-- For' example, given the occurrence of a-Class.IV accident; sequence,.the conditional 50 mile total is 2.7 x 10 person-rems (per occurrence) .
4-2
7 Y.,' _(
' RELEASE / CONSEQUENCE MODELS y ,
III. RISK . o Accident frequences in updated (1989) PRA results. Includes internal, fire, and flood initiators. For example, the sum'of the estimated.fregency'of occurrence of all Class IV accident sequences ~ is 1.05 x 10'I/ year. o SARA conditional accident class consequences, o Public risk estimated as: Accident class Conditional Class. Frequency Consequence For example, for Class IV accidents, offsite exposure = 1.05 x 10"I/yr x 2.7 x 10 7person-rems
= 3 person-rems / year (unmitigated) 1 1
4-3 _ - _ _ - . - _ _ ____---______________---_-_____-__m
- p. -
p .s: 2. . !..r RISK REDUCTION BENEFIT 1 I. METHOD o Risk' Reduction Evaluation-RR =' F , x [ P , x ( C ,, -Cg)) i Where: F, = accident sequence class frequency for class i P, = conditional probability of mitigating this sequence L with.a specific SAMDA C ,i = the conditional consequences (population dose in man-rem) for the unmitigated sequence in class i Cg = the conditional consequences for the mitigated sequence in class i
~
o For each SAMDA (j), estimate for each accident class (i): Risk Reduction, averted person-rems i,j as Accident Sequence Class Frequency / Year i times Probability of Mitigation by SAMDA i,j times (unmitigated-mitigated) population dose for sequences in clavs, person-rems i o Sum over all classes to obtain tota.3 risk reduction benefit for given SAMDA 5-1 e -- _I
o +-
~
RISK REDUCTION BENEFIT II. MITIGATION EVALUATION o Probability of Mitigation, Pm Based on engineering evaluation of SAMDA and accident progression. Numerical Probability assigned according to the following table: Qualitative Assessment Assioned Mitication Probability very likely to be affective .99 Highly likely to be effective .95 Likely to be effective .9 Indeterminate .5 Somewhat unlikely to be effective .25 Unlikely to be effective .1 Very unlikely to be effective .01 Impossible (or extremely unlikely) O. to be effective-o Consequence Mitigation Effectiveness Majority-of cases: SAMDA considered capable of complete mitigation that is, mitigated population dose = 0 Some cases: Assessment of actual mitigation process and fission product transport paths result in assigning an incomplete mitigation effectiveness, that is, mitigated population dose > 0 i 5-2 l _ _ _ _ _ _ _ _ _ - - _ _ _ _ - - - . - _ _ _ _ - - - - - m !
- q: .-
RISK REDUCTION BENEFITS l III. EXAMPLES EUBBLE BED CORE RETENTION SAMDA ,
- l CLASS 'Em . KEE R R. NOTES 11 .25 1.0 12 Some. Debris remains in DW 2 0 -
0 OP CF occurs 3 .25 1.0 1 Same as Class 1 4 0 - 0 OP CF occurs Total 13 Person-rems /yr. averted DRY CRUCIBLE CORE RETENTION SYSTEM SAMDA CLASS Em HEE B.,L. NOTES 1 .95 1.0 45 Prevents OP/OT CF 2 .9 1.0 1 Prevents OP CF 3 .95 1.0 1 Same as Class 1 4 .95 .37 10 OP/OT CF occurs, but DW sprays effective Total 57 Person-rems /yr. averted MEF = (Cumi - Cmi)/Cumi, all in person-rems is mitigation effectiveness factor for example, MEF=(2.7 x 10 -7 1. 7 X 10 )7 /2. 7 x 107 1 l l l 5-3 w____-_______-_-_-______ . _ .-
q -- - - - -
+ (. : ,
sD 4.0 SANDA EVALUATIONS
- 4.1 ~ Methodology The . Severe Accident Mitigation' Design Alternatives- (SAMDAs) proposed are listed in Table 4-1. An input into any decision on the need.to install any of these SAMDAs is an evaluation of the
.value and impact or. benefit and. cost of.the SAMDA. ' The . maj or - -
benefit of a SAMDA is the reduction in severe accident risk that the SAMDA ' provides. The usual measure ~of ' risk utilized. is the L mean population dose (i.e., person-rem) integrated out through 50 miles ~ of. the plant. This is consistent ~with past NRC' .value-impact analyses practices (Ref. 14). The population dose risk reduction - (person-rem was converted to'a dollar benefit using-
.$1000/ person-rem as the monetary equivalent of a unit dose.
(Refs. 14 and 15) Hence, the annual risk reduction. benefit was calculated as: Annual Benefit ($)= Annual Risk Reduction (man-rem / year) X
$1000/ man-rem The present worth of the annual benefit was calculated using the
'following formula-(Ref. 15):
(1 + r)t . 1 PW = Ca = 9.56 Ca r.(1 + r)E Where: ca = the annual benefit ($) r = the annual discount rate.(.1025 from PECo) t- = the remaining plant life (40. years) The risk reduction potential for each of the SAMDAs considered in
-this analysis was evaluated for each accident class and for each release category associated with that class. (Definitions of SARA- accident classes and release categories 'are contained - in Reference 1). The basic approach to evaluating the risk reduction potential for a SANDA was to estimate ' the - probability that an accident sequence in a given class and release category would be mitigated by a specific SAMDA and to assess what the population dose would be for the mitigated sequence. The risk reduction for a SAMDA was evaluated as follows:
RR = Fi x [Pi x (Cumi - Cai) 3 i where: Fi = accident sequence class frequency for class i Pi = conditional probability of mitigating this sequence with a specific SAMDA 46 1__ _ _ _ - - - - - _ _ _ _ _ _ _ _ _ _ - _ _ - - - - - - - - - - - - _ _ _n
'^ ' g: _',.
TABLE 4-1 SEVERE ACCIDENT MITIGATION DESIGN ALTERNATIVES EVALUATED
-o POOL HEAT REMOVAL SYSTEM A separate . independent dedicated. system- for transferring heat from the suppression pool to the spray pond utilizing-a diesel driven 3,200 gpm pump and heat exchanger without dependence on the. Station's present AC electrical power or other systems. The diesel is cooled with. water tapped off the spray pond suction line..
o DRYWELL SPRAY. A new dedicated . system for heat and fission product removal using the Pool Heat Removal System described above'to inject water into the drywell. o CORE DEBRIS CONTROL (" CORE CATCHERS") . Two techniques, either a basemat rubble bed, or using a dry crucible approach, to contain the debris in a.known stable condition in the containment. , o ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS) VENT A large wetwell vent line to an elevated release point to remove heat added to the pool in an ATWS event. o FILTERED VENT Drywell and Wetwell vents to a large filter (two types
- gravel or enhanced water pool) to remove heat and fission products.
o LARGE H2 RECOMBINER Independently powered recombiners to remove H2'from the , containment in the long-term after a severe accident. i o IARGE CONTAINMENT VACUUM BREAKER To restore containment pressure to atmospheric level j through 20" valves in certain severe accident cases where a vacuum has been produced. I 47 l I s - _ __- __ _ _-_____--__-_ - - - _ - _ _ .n .
I
.1 .
= the conditional consequences (population dose in Cumi man-rem) for the unmitigated sequence in class i
=
I Cmi the conditional consequences for the mitigated l sequence in class i { l The rational ~e for the selection of the mitigation probabilities and the mitigated consequences for the individual SAMDAs are presented in the following sections. Several broad generic assumptions were employed which impact all SAMDAs. These are listed below: a General Assumptions
- 1. The probability for mitigating steam explosion and hydrogen burn containment failure ' sequences (release category OXRE) i was assumed to be zero (for all accident classes where the SAMDA does not prevent core melt).
. 2. Seismic and large reactor vessel rupture sequences were assumed to be unmitigated.
- 3. The mitigation probabilities were assigned based on the following assessment strategy.
Qualitative Assessment Assioned Mitication Probability very likely to be effective .99 Highly likely to be effective .95 Likely to be effective .9 Indeterminate .5 Somewhat unlikely to be effective .25 Unlikely to be effective .1 Very unlikely to be effective .01 Impossible (or extremely unlikely) O. to be effective
- 4. Class 3 sequences characterized by failure to shutdown the I reactor with loss of core coolant injection are very similar l to the class 1 (loss of core coolant injection following a transient or IOCA initiator) sequences. In the SAMDA benefit analysis it was always assumed that the class 3 sequences were mitigated to the same extent as the class 1 sequences by a specific SAMDA.
- 5. All risk values (man-rem / year) are rounded to integer values.
l
- 6. Seismic population dose risk was not included as specified by NRC question 2.
1 48
4.2 Evaluation of Benefit of Each SAMDA 4.2.1 Dedicated Suppression Pool Cooling System (DSPCS) Class 1 Secuences The DSPCS is.unlikely to be effective in mitigating Class 1 loss of core coolant injection sequences since no mechanism is provided for preventing drywell overtemperature failure following vessel rupture. Furthermore, this SAMDA does not provide for any mitigation of radionuclides release to the environment. Mitigation Probability (P m) = 0.1 If the DSPCS is successful in preventing containment failure then the accident source term is very small. Mitigated Sequence Consequences (Cm = 0 man-rem) RRi = 8.84 x 10-6 x 0.1 (5.4 x 106 _ o)
= 5 man-rem / year Class 2 Secuences The DSPCS is likely to be effective in preventing steam overpressure failure and core melt for the Class 2 sequences.
Mitigation Probability (P m) = 0.9 If containment failure and core melt are prevented no consequences are expected. Mitigated Sequence Consequences (Cm) = 0 man-rem The Class 2 risk reduction is then approximately: RR2
=
1.75 x 10-7 x 0.9 (9.3 x 10 6_o)
= 1 man-rem / year Class 3 Secuences Class 3 esquences are similar to Class 1 sequences (mitigation probability and mitigated sequence consequences are the same as for class 1). Hence, the Class 3 risk reduction is approximately:
RR3
=
2.73 x10-7 x 0.1 (5.4 x 106 .o) ,
= 0 man-rem / year 49
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - __ _ n
Class 4 Secuences It is extremely unlikely or impossible that the 'DSPCS will . be ' effective.in mitigating ATWS sequences. The design heat removal capacity of ' this system (~ 45 MWt) is far below the heat production rate during an ATWS ( ~ 10% of full core power or 330Mwt). Hence, .this system' will not prevent containment-overpressure" failure or core melt. Furthermore, this system provides . no mitigation of the radionuclides released during the-accident.
- Mitigation Probability (P m) = 0.
Risk Reduction (RR 4) = 0 man-rem / year Summary-Dedicated Suppression Pool Cooling System Class Risk Reduction (man-rem /vear) 1 5 2 1 3 0 4 0 Total 6 4.2.2 Enhanced Drywell Spray System (EDSS) Class 1 Secuences The EDSS is likely to prevent both containment overpressure and overtemperature failure for Class 1 sequences since the drywell air space and the core debris are provided with a cooling spray of water. Mitigation Probability (Pm) = 0.9 If containment failure is prevented a very small or zero source term would be expected. Mitigated Sequences Consequences (Cm ) = 0 man-rem The Class 1 risk reduction is then approximately: RR1
=
8.84 x 10-6 x 0.9 (5.4 x 10 6_o) 1
= 43 man-rem / year Class 2 Secuences The EDSS is likely to prevent containment failure and core melt since it provides the containment heat removal function which has been lost for these sequences.
50
Mitigation Probability (P m) = 0.9 If containment failure and core melt are averted then the consequences will be zero. Mitigated Sequence Consequences (C m) = 0 man-rem The Class 2Jrisk reduction is approximately:- RR2
=
1.75 x 10~7 x 0.9 (9.3 x 10 6_o)
= 1 man-rem / year Class 3 Secuences Mitigation probability and the mitigated sequence consequences are similar to Class 1. The rish reduction is then approximately:
RR3
=
2.73 x 10~7 x 0.9 (5.4 x 106-0)
= 1 man-rem / year Class 4 Secuences This system has an insufficient design heat removal capacity to prevent suppression pool heatup, steam generation and containment overpressure failure for ATWS sequences with power levels near 10% of full core power. However, assuming that the EDSS system can survive containment failure it will provide some mitigation of the radionuclides release due to spraying of the drywell gas-space.
The probability of mitigating . the fission product release by spraying the drywell is: Mitigation Probability (P m) = 0.9 We assume that spraying of the drywell gas space will reduce the source term (and offsite consequence) for these sequences to that of the OPREL release category. This reduces the overall consequences by a factor of approximately 1/3 from their unmitigated values for Class 4 sequences. Mitigated Sequence Consequences = 1.7 x 107 man-rem The Class 4 risk reduction is approximately:
. RR4
=
1.05 x 10-6 x 0.9 (2.7 x 107 - 1.7 x 107)
= 9 man-rem / year 51 a_=____-____-____________-__. - _ _ _ - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ __-
4 . .
- Summary-Enhanced Drywell Spray System Class Risk Reduction (man-rem /veari v
1 43 2 1 3 ' 1 4~ - 9 Total 54 i 4.2.3 Rubble Bed Core Retention System Class 1 Secuences The floodable rubble bed system is judged to be somewhat unlikely in preventing overtemperature drywell failure since no cooling is provided for the debris that.does not relocate to the rubble bed from the drywell pedestal area. However, this system ' should reduce the probability of gross overpressure failure of containment by providing. cooling to the majority of the debris. Mitigation Probability (P m) = 0.25 If containment failure is prevented the source term will be very small. Mitigated Sequence Consequences (Cm) = 0 man-rem The values result in an approximate Class 1 risk reduction of: RRy = 8.84 x 10-6 x 0.25 (5.4 x 10 6_o)
= 12 man-rem / year Class 2 Secuences
- The rubble bed system does not prevent overpressure containment failure or core melt for loss of containment heat removal sequences and its mitigation potential for these sequences is very small.
Mitigation Probability (P m) =0 RR2 =.O man-rem / year Class 3 Secuences Class 3 Sequences are similar to Class i sequences (mitigation probability and mitigated sequence consequences are similar). Therefore, the Class 3 risk reduction potential is approximately: 52 l
l
- l i
l l RR3
=
2.73 x 10~7 x .25 (5.4 x 10 6_o)
= 1 man-rem / year Class 4 Secuences The rubble bed system does not provide any mechanism for removing the heat load generated by an ATWS event and will not prevent pool heatup, steam generation and overpressure failure of the containment. Hence, containment failure and core melt are not prevented in this class of sequences.
Mitigation Probability (P m) =0 RR4 = 0 man-rem / year Summary - Floodable Rubble Bed Core Retention System Class Risk Reduction (man-rem /vear) 1 12 2 0 3 1 4 0 Total 13 4.2.4 Dry Crucible Core Retention System Class 1 Secuences The drywell spray and independent heat removal portions of the dry crucible system can remove the heat generated by the debris (both debris relocated to the crucible itself and remaining in the drywell) and it is very likely that both overtemperature and overpressure failure of containment from steam generation or noncondensible gas generation from debris concrete attack can be prevented. Mitigation Probability (P m) = 0.95 If the system prevents containment failure then the source term will be very small. Mitigated Sequence Consequences (Cm) = 0 man-rem The risk reduction potential is approximately: RR1
= 8.84 x 10-6 x .95 (5.4 x 106 .o)
= 45 man-rem / year Class 2 Secuences If the system is activated early in the accident sequence then it is capable of removing the decay heat being injected into the 53
_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ ---. m
a suppression pool and can prevent containment overpressure failure and core melt. Mitigation Probability (Pm) = 0.9 If containment failure and core' melt are prevented then the source term ,is essentially zero. Mitigated Sequence Consequences = 0 man-rem The risk reduction is approximately: RR2
=
1.75 x 10-7 x 0.9 (9.3 x 10 6_o)
= 1 man-rem / year Class 3 Secuences Similar mitigation probability and mitigated sequence consequences apply to class 3 as to class 1 sequences.
Consequently, the risk reduction potential for Class 3 sequences is approximately: RR3
=
2.73 x 10-7 x .95 (5.4 x 10 6 0)
= 1 man-rem / year Class 4 Secuences The dry crucible retention system does not have the heat removal capacity to prevent containment overpressure failure from steam production during an ATWS sequence. Hence, this system will not prevent containment failure or core melt for Class 4 sequences.
-The ~ system can, however, mitigate the radionuclides release by spraying the drywell atmosphere and attenuating radionuclides in the drywell atmosphere. As for the enhanced drywell spray system it is assumed that the source term for Class 4 sequences can be reduced to the equivalent of the OPREL release category.
l Mitigation Probability (Pm) = 0.95 (for scrubbing radio-nuclides in drywell atmosphere) Mitigated Sequence Consequences (Cm ) = 1.7 x 10 7 man-rem The risk reduction for class 4 sequences is approximately: RR4
=
1.05 x 10-6 x .95 (2.7 x 107 -1.7 x 107)
= 10 man-rem / year Summary - Dry Crucible Retention System 54 L_s . . -_ .- - . . ._. _ _ _ _ _ _ _ _ . _ _ _ . - _ - _ _ _ _ _ . _ _ - - - _ _ _ _ - _ _ _ _ _ - _ .
3 .. ,
]
s
.C.lang-- Risk Reduction (man-rem /vear) l '- 45 t
'2' 1 3 .l.
4 1Q
. Total 57 4.2i5 A'IWS Vent -
g,lajss 1 Secuer:ces - Followingivessel failure the core debris will- drain from the vessel onto the lower drywell pedestal floor.. The core' debris'is-
.then' expected to attack the drywell pedestal c1 rain line plate and open 'a . pathway. between the drywell and wetwell . air . space; effectively hypassing the- suppression pool. Consequently, even
~if venting is employed to protect the ' containment against
. overpressure containment failure the post-vessel . failure radionuclides . releases. would be ' unmitigated by 'the. pool.
Furthermore, the ATWSLvent does not protect the drywell against overtemperature failure - due to residual . debris in. the drywell. Consequently,_the probability of successfully mitigating class 1 sequences with the ATWS vent is considered very unlikely. Mitigation Probability (P m) = 0.01 If pool bypass and drywell overtemperature failure are avoided. and the : vent :is employed to prevent containment overpressure failure then radionuclides will pass.through and be mitigated by the-suppression pool resulting in a fairly small source term.- It is: estimated that-the consequences would.be intermediate between' the SARA LEAK 1 and LEAK 2 release categories. Mitigated Sequence Consequences (C m ) = 7.6 x 105 man-rem The class 1 sequence risk reduction is then approximately: RR1
= 8.84 x 10-6 x .01 (5.4 x 106 - 7.6 x 105)
= 1 man-rea/ year Class 2 sacruaggg l-L ,The .'impaci of the existing watwell venting capability in t mitigating ' Class 2 sequences has been considered in the PRA analysis. It is indeterminate whether an independent, hardened, high-capacity vent system will. provide additional benefits.
Mitigation Probability (Pm) = 0.5 If used during Class 2 sequences the ATWS vent will prevent 55 7 a-
overpressure containment ' failure ~ and core melt and will . reduce - the consequences to effectively zero. Mitigated Sequence Consequences (Cm) =-O. _ The estimated risk reduction is approximately:
- 1.75 x 10-7 x' . 5 (9.3 x 106-0)
~
=
RR2'
= 1 man-rem / year Class 3 ' Secruences The mitigation pubability and mitigated sequence consequences for Class 3 are similar-to Class 1 sequence results.
The Class 3 risk reduction potential is approximately: i RR3
=
2.73 x 10-7 x .01 (5.4 x 106 -7.6 x 105)
= 0 man-rem / year 1
Class 4 Secuences The optimistic assumption is made that it is likely that the ATWS vent will prevent steam overpressure failure and core melt. THis presumes - that core coolant injection can be continued until reactor shutdown efforts are successful. Mitigation Probability (P m) = 0.9 If containment failure and core melt are prevented by he ATWS vent then the consequences from the mitigated ATWS sequences will be very small. Mitigated Sequences consequences (C u) = 0. man-rem / year The Class 4 risk reduction potential is approximately: RR4
= 1.05 x 10-6 x 0.9 (2.7 x 107-0)
= 25 man-rem / year Summary - ATWS vent
- l. Class Risk Reduction (man-rem /vear) l L 1 1 1
2 1 3 0 4 25. Total 27 56 ! _ _ - _ _ _ _ _ _ . _ _ _ -. - . - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ J
t 4.2.6 Filtered Vent System This section summarizes the benefits for both the gravel bed and multi-venturi scrubber filtered vent systems. Class 1 Secuences i The filtered vent system will prevent overpressure containment failure. However, it is indeterminate as to whether the filtered vent will protect against overtemperature drywell containment failure. Mitigation Probability (P m) = 0.5 If containment failure is prevented then the filtered release of non-noble gas radionuclides will be very low. The consequences of a successfully mitigated sequence can be assumed to be equivalent to release category LEAK 2. Mitigated Sequence Consequences (C m ) = 1.5 x 105 man-rem The potential risk reduction is approximately: RR1
=
8.84 x 10-6 x 0.5 (5.4 x 106 - 1.5 x 105)
= 23 man-rem / year Class 2 Secuences The filtered vent system is likely to prevent containment failure and core melt for Class 2 sequences (it is assumed that it is much more likely to be effective than the existing wetwell vent capability).
Mitigation Probability (P m) = 0.9 Since use of the filtered vent will prevent containment failure and core melt the consequences of the mitigated sequences are effectively zero. Mitigated Sequence consequences (Cm) = 0 man-rem The risk reduction potential is approximately: RR2
= 1.75 x 10~7 x 0.9 (9.3 x 106-0)
= 1 man-rem / year Class 3 Secuences Mitigation probability and consequences are similar to Class 1.
57 i C. .
f~~5 L h RR3
= 2.73 x 10-7 x .5 ,5.4 x 106 - 1.5 x 10 5)
= 1 man-rem / year Class 4 Secntences The filtered vent system have insufficient capacity to relieve the ' steam g'eneration rates from an ATWS cvent.at 10% full core power and will not prevent containment overpressure failure or core melt.
Mitigation Probability (P m) = 0. The risk reduction potential for class 4 sequences is then:
= 0 man-rem / year RR4 Summary - Filtered Vent Systems Class Risk Reduction (man-rem /vear) 1 23 2 1 3 1 4 0 Total 25 4.2.7 Large Hydrogen Recombiner This system does not prevent (early) containment failure or mitigate radionuclides release for any identified accident sequence. It is viewed as more of a long-term accident recovery system than a short-term accident mitigation system. It is judged that the risk reduction potential for this system is small.
4.2.8 Large Containment Vacuum Breakers A qualitative assessment by the Boiling Water Reactor Owner's Group (Ref. 16) of the conditions that would lead to large negative pressures concluded that such conditions are not expected following recovery of normal containment heat removal and termination of venting. Additionally, the reinforced concrete Mark II containments such as Limerick.are not expected to fail even for pressure differentials exceeding twice the design differential pressure of 5 psid (Ref. 16). Therefore, the vacuum breaker would not mitigate any accident sequences currently identified. 58
,V
.. 4.3' Summary of Cost Benefit Results The costs and benefits of the mitigation systems are summarized
'in Table 4-2.' TheLtable provides the following:
Benefit:' The estimated. risk reduction in dollars per year calculated from the. estimated man-rem per year averted by. the mitigation device. (see section 4.2) times $1000 per man-rem. Total Benefit: The .present worth in. - dollars . of the yearly benefit assuming a 40 year plant life and a 10.25% ' discount rate. Total Cost: The' total cost of the mitigation- device- including construction costs and the present worth of annual operating costs over a 40 year ' plant life. These results are from reference 17. In reference: 17, the costs were estimated for installation at 2 units and were divided by 2 to obtain a per-unit cost. Benefit / Cost Ratio: The ratio of the total benefits to total costs. A value creater than 1.0 would indicate a cost beneficial mitigation device. Cost / Man-rem Averted: The cost per man-rem averted. These .. values were calculated as the total cost times $1000/ man-rem divided by the total benefit. A cost less than
$1000/ man-ren would indicate a cost beneficial mitigation system.
The results presented in Table 4-2 show that none of the mitigation systems examined are cost beneficial. In fact, the
.results indicate that no mitigation system is within an order of magnitude (factor of 10) of being cost-beneficial.
I 1 f 59 l
. - _ . _ _ - - _ . - . _ _ _ _ _ - _ _ _ _ - _ - _ -__..-___-__-_.-___________-_-_-_rL_.
TABLE 4-2 COST / BENEFIT COMPARISON COST / _. TOTAL TOTAL BENEFIT / MAN-REM MITIGATING SYSTEM BENEFIT BENEFIT COST COST RATIO' Avrsfr.;u l- Dedicated-Suppression Pool Cooling $25,000/Yr $239K(1)$25,600K .009 $107,000 Enhanced Drywell $54,000/Yr $516K $46,500K(2) ,oli g 90,100 Sprays $27,000K(3) .019 $ 52,300 Rubble Bed Core $13,000/Yr $124K $38,400K .003 $310,000 Retention Dry Crucible Core $57,000/Yr $545K $119,000K .005 $218,000 Retention ATWS Vent $27,000/Yr $258K $ 3,900K .066 $ 15,100 Filtered Vent $24,000/Yr $229K $11,300K .020 $ 49,300 (Gravel Bed) Filtered Vent $24,000/Yr $229K $ 5,700K .040 $ 24,900 (MVSS) Large Hydrogen S 0/Yr $ 0 $ 5,200K .0 - Recombiner Large Vacuum Breakers $ 0/Yr $ 0 $ 0 .0 - 1 Denotes that the item is in thousands of dollars 2 New drywell spray nozzle distribution header 3 Use of existing drywell spray header 60 m - - - -. - - - - . - - - - - - - - - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _u
CONTAINMENT ANALYSIS GENERAL: o DISCUSSED IN SECTIONS 3.4.5 AND 3.5.4 OF LGS PRA.
.o CET FIGURE 3.5.6 I
6-1 _ _ _ - _ _ _ _____________ _____ _ ___ x
CONTAINMENT' ANALYSIS SPECIFIC QUESTIONS - A. 0.5 PROBABILITY OF LEAK TO PREVENT RUPTURE O.5.LARGE > 100%/HR. O.5'SMALL < 100%/HR. B. SUPPRESSION ' PO' O L BYPASS CONSIDERED ONLY AS RESULT OF CONTAINMENT RPUTURE IN PRA/ SARA EXPECT BYPASS AS A RESULT OF DRAIN FAILURE AT 6 MIN. CONSERVATIVE PRA/ SARA SOURCE TERM MEANS THAT IMPACT ON RISK IS SMALL. i IF- GAMMA PRIME (VAPOR SPACE) FAILURE MODE IS ASSIGNED A GAMMA (DRYWELL) ' SOURCE TERM THERE IS ONLY A 5% INCREASE IN POPULATION DOSE. POOL BYPASS WAS ACCOUNTED FOR IN DETERMINING EFFECTIVENESS OF SAMDAs. C. APPROXIMATELY 15% OF CDF IS FROM TQUX (HIGH PRESSURE) TYPE SEQUENCES. AN ADDITIONAL 14% OF CDF IS FROM SEQUENCES WHERE ALL DC IS LOST.
'. PRA UTILIZES 0.01 AS UNAVAILABILITY OF INERTING.
DE-INERTING FOLLOWING VENTING NOT SPECIFICALLY EVALUATED. 6-2
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... DOMINANT SEQUENCES EARLY FATALITY.
l- CONDITIONAL'.EARLY-SEOUENCE FREQUENCY' CLASS FATALITY-RISK TSRPV. .4.SE-07' 3 (1.6E-07) 0.58 g/ S'(3.2E-07) 599 TCP2 M 2.0E-07 4 173 s] ' TMP2I M 1.2E 4 173 gr .: TTPPP' 1.2E-07 4 173 V' TCP2P 7.1E-08 4 ~173 Q.I TTPPL W 5.SE-08 4- 173 7-1 1 e a I.b- - - - _ _ _ _ _ _ . . _ . _ _ - . _ _ _ _ _ _ _ . _ _ - . _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ _ _ _ h
-gm_, --r-n-,,
r e FIRE RISK ANALYSIS DOMINANT DOMINANT SEQUENCE AREA CONTRIBUTOR I
~
F44QUV- - SAFEGUARD SYS CABLE FIRES (FGS 2) ACCESS AREA F2QUV 13KV SWITCHGEAR CABLE FIRES (FGS 2) ROOM CABLE FIRES (FGS 3) PANEL l FIRES (FGS 2) F45QUV CRD HYDRAULIC EQUIP CABLE FIRES (FGS'3) AREA CABLE FIRES (FGS 2). F47QUV RWCU COMPARTMENTS CABLE FIRES _(FGS 3) AND GENERAL EQUIP CABLE FIRES-(FGS 2) F2QUWFWECC 13 KV SWITCHGEAR CABLE FIRES (FGS 2) PANEL FIRES (FGS 2) 8-1 L._ t.N_._ - - _ _ -______._-____m.__.__-_._m_____-__m___ . _ _ _ _ _ _ _ - - _
D
-CONSERVATISM 8 INCLUDED IN LG8 DOMINANT FIRE SEQUENCE 8 o' FIRES WERE ASSUMED TO DAMAGE ALL CABLES INITIALLY WHICH
-ARE ASSOCIATED WITH THE SDM IN WHICH THE FIRE STARTS.
o FIRES WERE ASSUMED TO DAMAGE ALL CABLES IN ALL UNPROTECTED SDMs'IMMEDIATELY IN FIRE AREA 2 AND IN 10 MINUTES FOR OTHER AREAS. o NO ATTEMPT WAS MADE TO IDENTIFY THE CRITICAL AREAS WHERE - MULTIPLE UNPROTECTED SDM CABLING RUNS WERE IN REASONABLY CLOSE PROXIMITY. o THE BASIS FOR THE 10 MINUTE PROPAGATION TIME AS COMPBRN' I CALCULATIONS ASSUMING THE MINIMUM CABLE-TRAY SEPARATION. o THE MAJOR CONTRIBUTORS TO THE DOMINANT FIDE SEQUENCES WERE, IN GENERAL, CABLE INITIATED FIRES. THERE IS A T ACK OF BACKGROUND INFORMATION ON THE HISTORICAL FIRE DATA REPORTED IN NUREG/CR-5088. THE THREE FIRES USED TO DETERMINE THE CABLE FIRE INITIATING FREQUENCY WERE
-QUESTIONABLE AS TO THEIR APPLICABILITY SINCE THE TYPE OF CABLE INVOLVED WAS NOT KNOWN. ALSO LGS USES IEEE-383-RATED CABLINC EXCLUSIVELY AND AS SUCH MAY NOT BE AS, SUSCEPTI3LE TO CABLE INITIATED FIRES.
l l f*$ a-r _t . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _ - _ _ - _ _ _ -
h l:: 1 o- MOST OF THE FIRE INITIATORS WERE ASSUMED TO RESULT IN AN MSIV CLOSURE AND THE MSIV CLOSURE EVENT TREE WAS QUANTIFIED FOR EVENTS D & F. THIS IS THE WORST CASE SCENARIO. o THE FIF.E SUPPRESSION PROBABILITY CURVE IN NUREG/LR-5088 IMPLIES TRAT THE TIME USED WAS THE TIME REQUIRED TO COMPLETELY EXTINGUISH THE FIRE. IN REALITY, AS SOON AS SUPPRESSION ACTIVITIES COMMENCE, THE TIME BEFORE CABLE DAMAGE IS EXTENDED DUE TO THE REDUCED HEAT FLUX. o THE DOMINANT FIRE SEQUENCES ARE, IN GENERAL, SEQUENCES-IN WHICH THE FIRE PROGRESSES TO FIRE GROWTH STAGE 2 (I.E. , DAMAGE TO ALL UNPROTECTED CABLES) . NO CREDIT FOR THE AUTOMATIC SUPPRESSION SYSTEMS WAS GIVEN IN THIS EARLY STAGE AND ONLY MINIMAL CREDIT WAS TAKEN FOR ANY SUPPRESSION DURING THIS TIME PERIOD. PORTIONS OF THE AREAS ARE PROTECTED BY AUTOMATIC SUPPRESSION' SYSTEMS
-WHICH COULD SUPPRESS THE FIRE EARLY.
3 -! an
* ,_a
[;. ,. [ . TRANSIENT. FREQUENCY
- o. SINCE, COMMERCIAL OPERATION BEGAN (FEBRURARY 86-THROUGH MAY 89) LGS HAD 8 TURBINE . TRIPS. (4 MANUAL /4. AUTO)~. -
. FREQUENCY OF TURBINE TRIP.IS 8.5*/3.33 OR 2.55/YR.
(
- 0. 5 - EFFECT OF BAYESIAN . UPDATE WITH NONINFORMATIVE PRIOR) o .NEW TRANSIENT FREQUENCY 6.69 VS 9.74 9-1
_ - . _ _ _ _ _ _ _ _ _ _____.________________________________n
E. a-
)
- i. 1 i
i _-~ TABLE 2-1 !
SUMMARY
OF THE FREQUENCY OF TRANSIENT INITIATORS I Frequency- } Initiator f Per _ R_gggj;;or Year) Nov 88 Uedate Present Uodate Turbine Trip. 5.6 2.55 Manual Shutdown 3.2 3.2 MSIV Closure 0.23 0.23
-Loss of Feedwater O.19 0.19 Loss of'Offsite Power 0.074 0.074 Inadvertent Open Safety Relief Valve 0.07- 0.07 Loss.of Condenser vacuum 0.38. 0.38 Total 9.74. 6.69 4 -.c'- - - _ _ _ - - - _ _ . _ _ - - - - - . - ----------.________________________J
m..._.. { jj ' 'e l A.6 . COMPLETE LOSS OF OFFSITE POWER L Complete loss of offsite power- to a generating station is an event which is influenced by ~ l oca.1 . factors such as type -of weather exposure, transmission system design, and operating procedures. Therefore, a. local or regional data base is more suitable .than a'ntsional data. base for predicting the frequency and duration of.such events at a specific plant. Limerick ' Generating Station is connected to the Pennsylvania-New Jersey-Maryland Interconnection (PJM) System and the remainder of the - PECo System via five transmission lines. Section A.6.1 reviews the PJM/PEC0' data base and analytical-techniques'used in this study to determine (1) the frequency of complete loss of offsite power and (2) the probability of recovery of offsite power as a function of time from interruption. The analyses show a' relatively high reliability for the PJM/PECo plants. Even so, the use ofLthese levels of reliability in this study. is probably conservative since the five.. transmission 1:ne design at Limerick exceeds the average level of . redundancy for the plants included in the data base.
~
Section A.6.2 discusses the specific case of loss of Offsite Power caused by trip of the Limerick turbine generator. A.6.1 PJM/PECo Experience Complete loss of offsite power experience for PJM nuclear plants is summarized in Tables A.6.1. In total, these plants have an experience of four occurrences in 53.71 plant years. The exposure for eacn site is calculated as the amount of_ time at least one unit at a site is operating at or near full power. Time in which all units at a site are shutdown is not included because the recovery time is so long that recovery of offsite power is essentially assured before core damage occurs. Additionally, the configuration of offsite connections for plants shutdown are sometimes significantly altered to the point where that configuration would be prchibited during power operation. For these reasons loss of offsite power occurrences at Hope Creek and Salem while shutdown for outages have not been included. A-S3
a {_. - l 'f.. [ I l i
, ... Table A.6.1 i.
L C05'PLETE LOSS OF 0FFS:TE POWER Pennsylvania-New Jersey-Maryland Interconnection'tPJM) _ Juclear Plant Experience;
~.from Commerical Operation through December:19S7*.
Average Finnt Exposure Occurrences Durntion iPlant-Years) (Minutes) Calvert C1iffs 11.23 1- 350 Calvert Cliffs ! 90
- 0yster. Creek ~ 11.16 1 IIS:
Hope Creek' '0.80 0 - Peach Settom- '11.51 0 Limeriek' 1.49 0 - Salem S.S 0 - Susquehanna- .3.02 1 11 Three Mile Island 5 . 7, 0, Total 53.71 4 142.25
- Exposure time for 195~ was estimated The annual frequency of complete !=oss of offsite power is 4/53.71 =
0.074. A-84 _1=___. .
4
'Another important factor is the probability of recovery of offsite power within specific times. The PJM/PECo data base was
~
again used in this assessment. The recovery times'for the four occurrences actually experienced were used to determine the mean recovery time and the variance .of recosery time. A gatta distribution was then constructed to fit the mean and variance. This dis,tribution is a'shown in 7able A.6.2. TABLE A.6.2 PROBABILITY DISTRIBUTION OF RECOVERY TIME Recovery Cumulative Time Density Density (Min.) Function Function 0.12 0.00932 0.001 0.57 0.00S57 0.005 1.16 0.00823 0.010 6.3S 0.00730 0.050 13.55 0.00670 0.100 29.64 0.00578 0.200 48.28 0.00498 0.300 70.07 0.00422 0.400 96.10 0.00349 0.500 128.18 0.00277 0.600 169.79 0.00207 0.700 228.75 0.00137 0.S00 330.09 0.00068 0.900 431.S4 0.00034 0.950 ' 668.99 0.00007 0.C90 Means = 142.25 STD.DEV = 145.7 alpha = 0.95 Beta = 149
+
A-S5
. . . _ _ - __ ._ - .- ____--__-_ -_ - -_--____-__-___- - t
v L The probability that recovery takes more than a given number of
- hours can be found from this distribution. Specif ically ,
4 P(Recovery of offsite power > 1 hour = 0.65
'P(Recovery of offsite power > 2 hours) = 0. .; 23
.P(Recovery of offsite. power > 5 hours) = 0.15 P(Recovery of effsite power > 10 hours) = O.01596 P(Recovery of offsite power > 20 hours) = 2.76 E-4 i
A-86
e ns . _, c- )
' i b.
U b .i 1 Loss of ' Of f site power Resulting f rom Turbine / Generator
~
'{
4 A.6.2/, j T r 1 E. A sudden' loss of al significant portion of grid generating capacity due;to the: lack of grid stability.may result from 'in-plant transient . events- that cause a turbine or generator trip.
~
If the' sudden loss of generator exceeds the transient stability limit ~ of- the local or ' regional grid system, then all offsite power to the plant could be lost. Eased upon information developed; for' WASH-1400, .the' probability for: complete loss et offsite power following a turbine or generator trip was estimated at:-approximately 1 x 10-8 per demand. This failure probability for any particular plant could be lower depending on the transmission systems,- the transient stability limit resulting from high installed capacity, the extent.of grid connections with other large utilities, and the number of transmission lines
'i connecting'the plant to the grid. It is judged that the conditional probability of such a' scenario is substantially less than that assumed in WASH-1400.
In order to support the judgment that a salue of 10-2 per reactor year is a conservative estimate, an evaluation is performed using
.he ~ nuclear plant experience- data base. Two cases need to be evaluated and summed to calculate a best estimate:
- 1. Offsite power loss due to load rejection at time of transient (Contribution 1).
- 2. Offsite power loss during the time immediately following a shutdown - any shutdown (Contribution 2) .
Contribution 1 The loss of offsite power frequency initiated by a transient within the plant can be developed from data which were not available during the WASH-1400- investigation. Using only the nuclear operating experience data, it is found that in more than 700 reactor years of nuclear experience there are no recorded cases of a loss of offsite power being induced by a nuclear plant trip. Based upon these data, an estimate can be made of the frequency of such postulated occurrences: 1 Probability (700 Rx years) (C sranstents per Reactor year) Probabi'lity 1.6 x 10-' per Reactor year Contribution 2 The loss of offsite power may also occur as a random independent failure at anytime during the year. If it occurs during the 10 hours immediately following a reacter shutdown, the result may be a test of the plant systems similar to a loss of offsite power A-57 ; & _ _ - _ _ _ ___- _ . _ _ _ _ _ _ m
.c (LOSP). Therefore,- the contribution fron such instances is eniculated belov., based upon PJM grid data.
LOSP f recuency (per Ex year) = 0.074 per Reactor Year
= S.4E-6 per Hour Thus, the conditional probabil2ty of the loss of offsite power due to ran'or d independent causes during the reactor safe shutdown 2s estimated using the failure frequency of .074/ year and a mission time of 10 hours follouang a shutdown:
S.4E-6/Hr x 20 Hr = S.4E-5/ shutdown Therefore, the total conditional probabilities of the loss of offsite power during, or as a direct result of, a transient or a manual shutdown are as follows: 2.4E-4 per transient (Contributions 1 and 2) S.4 x 10*5 per. manual shutdown (Contribution 2 only) 4 l A-SS l 1 i' - _- - _ - l
(/ 'o: L - l' 3.4.3.2 Reactor Pressure Vessel Failure Di srup't i v e' ' failure of the reactor pressure vessel is included in. the Limerick analysis at.10-t. 3.4.3.3 Interfacing LOCA L< Thus far,- the 'LOCA initiators, identified in the Limerick probabilistic evaluation are.within the spectrum of LOCAs which are typically considered. in ,the FSAR.D These LOCA initiators
- involve unisolatable primary system failures inside containment, as- such,. these breaks result in a transfer of primary system-fluid into the drywell and eventually to the suppression pool, and a requirement for coolant makeupy and containment heat removal.
In additiot, to these- sets of accidents, there is a class of postulated' events which could result in a loss of primary coolant into'the reactor building. The differences present in this class of events .from the -LOCAs inside containment include the following:
- 1. Isolation of the break is,possible in order to limit the release of. fluid to the reactor building.
- 2. In the event.of an unisolated break, there may be a high environmental stress produced on equipment. in the reactor building; therefore, equipment used for reaching
._' a safe stable state may be compromised.
The frequency of core damage associated with the following large LOCA events outside containment could contribute to the overall core damage frequency:
- 1. Steam line or main feedwater breaks outside containment (within the reactor building).
- 2. Breaks in the HPCI/RCIC steam supply or pump discharge lines.
- 3. Interfacing LOCAs in low pressure systems.
3.4.3.3.1 Approach The evaluation of the large LOCA outside containment in terms of potent.ial core damage frequency.can be evaluated by considering two separate categories of effects:
- 1. Prevention
- 2. Mitigation 3-10S QL----_____--- _-- _ _ _ _ _
c. t .1
- 1. , n Prevention of.a.LOCA outside containment has two aspects:
' l .' Prevention of a pipe or component rupture outside containment.
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- 2. Isolation of the failure from the primary system.
Mitigat4on is necessary for successful execution of the remaining key core' and containment functions if the event cannot be prevented including scram, coolant makeup, and containment heat l remov.al. 3.4.3.3.2 Limer:ck Unique Features There .are a number of Limerick unique features that minimize the
'importance of this initiator at Limerick. These features include the following:
- 1. Cycling of the interface valves (LPCI and core spray ,
injection valves are cycled on a longer test interval' ! than many other plants; i.e., each refueling outage rather than monthly during power operation).
- 2. 316 stainless steel minimizes the chance of stress corrosion cracking induced pipe failures in steam lines-and feedwater lines.
- 3. highly compartmentalized reactor building with steam.
relief panels located at precisely the location of possible interfacing LOCA minimizes the potential impact of such a LOCA on the reactor building equipment which can be used for safe shutdown.
- 4. . Check valves in the low pressure injection systems are either not air-operated testable check valves or methods of positively assuring they are seated when the reactor is pressurized are available.
3.4.3.3.3 Quantifiention There are two types of initiators that can be discussed as subgroups within the LOCA outside containment category. These two LOCA initiator types include:
- 1. Pipe rupt. es in high pressure lines attached to the primary system which are run outside containment.
- 2. Interfacing LOCAs induced in low pressure pipe connected to high pressure primary system pipe.
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'd' Large pipe ruptures .in high pressure pipe include main steam-lines, feeduater lines, and HpCI-lines. Other -smaller diameter lines are not considered ~ as significant challenges' to safe
. shutdown.- i f i The ' frequency of a pipe rupture in the high pressure primary system pipes external to containment is calculated to be a- very low frequency. In addition, at Limerick the isolation valves are specifically designed to close in the event of. such a rupture. Therefore, the combined frequency. of such a combination of f ailures (rupture plus. a double isolation valve failure) is calculated to be negligible relative to other potential core damage contributors and is not explicitly included in the Limerick model. The frequency for interfacing-LOCA is far below the more dominant core damage contributors. This judgement is based on evaluation of historical incidents. A number of incidents have occurred in. BER nuclear operating experience in which operator error, use of testable check valves, and on-line surveillance testing >of low pressure injection valves have exposed low pressure ECCS piping to high pressure and have demonstrated that ~ the highreal temperature water. capability ofTheselow incidents pressure systems is not exceeded. Because Limerick Technical Specifications do not require this on-line testing of .the interfacing salves these incidents are considered very unlikely and is not explicitly included in the Limerick model.
,' 3.4.3.3.4' Summary The potential initiating frequency of a LOCA outside containment due to the rupture of a high energy line or an interfacing LOCA is found to be negligible (less than 10-7 per year).
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S3ECIAL INIT ATORS Core Damage I initiator Freauency internal Flood 8E-8 ) 6.66E-8 i Reactor Water Level Reference Leg Leak or break Loss of Service Water 1.8E-8 l Loss of 1 DC Bus 2.74E-9 High Drywell Temperature 1E-8 1 l Loss of instrument Air 1E-8 to 1E-9 Loss of a Single AC Bus 1E-8 to 1E-9}}