ML20237H173
| ML20237H173 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/31/1985 |
| From: | INDUSTRY DEGRADED CORE RULEMAKING PROGRAM, TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20237H133 | List:
|
| References | |
| 23.1S, NUDOCS 8708170092 | |
| Download: ML20237H173 (411) | |
Text
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.' 8708170092 870811 PDR ADOCK 05000327 P PDR , ] The Industry Degraded Core Rulemaking Program, Sponsored By the Nuclear Industry 1*
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New %rk Power Authority Arizona Public Serivce C6?np iny The Babcock & Wilcox Company Niagara Mohawk Power Corporation Baltimore Gas and Electric Company Northeast Utilities Service Company
' Bechtel Puwer Corpora
- ion Northern Indiana Public Service Company Black & Veatch, Consulting Engineers Northern States Power Company n Boston Edison Company ,
hacife Gas and Electric Company ,
' , CFBraun & Co Pennsylvania Pdwer & Light Company
'u ,* The Cincinnati Gas & Electric Company Philadelphia Electric Company .
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The Cleveland Electric illuminating Company Purtland GeneralElectric Company Combustion Engineering,Inc. ~ Public Service Company ofOklahoma ,,
. Commonweahh Edison Company Public Service Electric 2nd Gas Company ,
Con <olidated Edison Company ofNew %rk, Inc. ..- Public ServiceIndiana
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. . l, Consumers Power Company Puget Sound Power & Light Company
. ' Daniel Construction Company Rochester Gas and Electric Corporation w The Detroit Edison Company 'Sargent & Lundy
., y ; l, Duke Power Company , ,,. , , j South Carolina Electric and Gas Company
; , , Duquesne Light Company [,
- Southern Cahfornia Edison Company G
. Q.'Ebasco Services incorporated : Southern Company Services, Inc. , ,
1 : Exxon Nuclear Company, Inc. Stone & Webster Engineering Corporation Swedish State Puwer Board
%l Florida Edwer & Light Company
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Fluor Power Services, Inc. Taiwan Power Company m i} GeneralElectric Company , Technical Research Centre ofEinland
. Gibbs & Hill, Inc. Tennessee Valley Authority
, ' Gilbert / Commonwealth Companies Texas Utilities Generating Company
,;. ' GulfStates Utilities Company The Toledo Edison Company w l Houston Lighting & Power Company Union Electric Company -.
Illinois Pdwer Company United Engineers & Constructors Inc.
. Japan Atomic Industrial Forum, Inc. Virginia Electric and Power Company -
Kansas Gas and Electric Company Washington Public Power Supply System . Long Island Lighting Company Westinghouse Electric Corporation Middle South Services, Inc. Wisconsin Electric Power Company , Nebraska Public Power District %nkee Atomic Electric Company - The IDCOR program is a large, independent technical effort sponsored by the nuclear industry. The Program is directed by a Policy Group comprised of representatives of the sponsoring organizations and operates under the corporate auspices of the Atomic Industrial Forum. The Program's purpose is to develop in an expeditious manner a comprehensive, in- - tegrated technically sound position to assist in determining whether changes in regulation are needed to reflect degraded core and core melt accidents. Further information on the Program can be obtained by contactingJohn R. Siegel, Special Licensing Projects Manager /IDCOR, Atomic Industrial Forum,7101 Wisconsin Avenue, Bethesda, MD 20814-4805, , (301) 654-9260. U (
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i [ - L - Technical Report 23.15 i: Sequoyah Nuclear Plant L Integrated Containment n Anal;ysis j .l May 1985 sf* N by: ) {j, Tennessee Valley Authority Knoxville, Tennessee
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l I The Industry Degraded Core Rulemaking Program, Sponsored by the Nuclear Industry l 1 - - - - - -
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'w' TABLE OF CONTENTS Pm ' ..'f , Section 1.0 Introduction . . . . . . . . . . . . . ... . . . . . . . . ' l.2-1 L3 -
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- 1.1 . Statement of the Problem . . . . . . . . . . ., . . . 1.2-1 s
1.2 Relationship to Other Tasks . . . . . . . . . . . . .- 1.2 l1[ l s 1.3 References . . . . ........ . . . . . . . . . 1.3-1 , j ~t' 2.0 S tra te gy and Me thodolo gy . . . . . . . . . . . . . . . . 2.0-1 '{ 1 2.1 References . . . . ........... . . . . . . 2.1-1 , 3.0 Descriptions ~ of Models and Major Assumptions . . . . . . . 3.0-1 3.1 Plant Description . . . . . . . . . . . . . . . . . 3.1-1 3 .' 1.1 Reactor Coolant System Description . . . . . 3.1-1
.- 3.1.2 Reactor Core . . . .. . . . . . . . . . . . . 3.1-3 3.1.3 Reactor Vessel . . . . . . . . . . . . . . . 3.1-4 ,
l 3.1.4 Steam Generator . . < . . . . . . . . . . . . . 3.1-4 3.1.5 Reactor Coolant Pumps . . . . . . . . . . . . 3.1-6 3.1.6 Pressurizer . . . . . . . . . . . . . . . . 3.1-6 l 3.1.7 Containment Description . . . . . . . . . . . 3.1-7' 3.1.8 Containment Heat Removal System . .. . . . . . . 3.1-15
' 3.1.9 Emergency Core Cooling Systen . . . . . . . . 3.1-17
- I[. . 3.1.10 Auxiliary Feedwater System . . . . . . . . . . 3.1-20
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- 3.2 Modular Accident Analysis Program (MAAP) . . . . . . 3.2 ! l 1
3." . MAAP Nodalization . . . . . . . . . . . . . . 3.2-1
, 3.2.2 Safety Systems Modeled in MAAP . . . . . . . . 3.2-4 3.2.3 Fission Product Release from Fuel . . . . . . 3.2-4 l 3.2.4 Fission Product Release from Aerosol Generation Resulting from Core-Concrete Attack . . . . 3.2-9 3.2.5 Description of the Natural Circulation Model . 3.2-10 3.2.6 Fission Product Deposition . . . . . . . . . . 3.2-12 3.3 References . . . ................. .
3.3-1 4.0 Sequences Analyzed . . ................ . . 4.0-1 4.1 Sequence No. 1-SD . . . . . . . . . . . . . . . . 4.1-1 1 2 e 4.1.1 Accident Sequence Description . . . . . . . . 4.1-3 4.1.2 Reactor Coolant System Response . . . . . . . 4.1-1 1 j 4.1.3 Containment Response . . . . . . . . . . . . . 4.1-2 l l I o{e 1 4 i f 1 e t WI
I - TABI.E OF CONTENTS (Continued)
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4.2' Sequence No. 2 - S 2H ....... . .. .. . . . 4.2-1 s 4.2.1 t.ccident Sequence Description . ...... .,4.2-1 ' 4.2.2 Reactor Coolant System Response . . . . . .. 4.2-1 4 4.2.3 containment Response . . . . . . . . . . . . 4.2-3 , 4.3 Sequence No. 3 - S HF . ... .... ...... . 4.3-1 2 4.3.1 Accident Sequence Description . .. . . . . 4.3-1 y 4.3.2 Reector Coolant System Responsa 4.3-1 j (Drains Blocked) ...... . . ... . . . 4.3-1 4.3.3 Containment Response (Drains Blocked) . . . 4.3-2 4.3.4 Reactor Coolant System Resp nse (Drains Open) 4.3-4.3.5 Containment Response (Drains Open) . . . . . 4.3-4 4.4 Sequence No. 4 - TEB' ........ ... ... . 4.4-1 , 4.4.1 Accident Sequence Description . .... . . 4.4-1 4.4.2 Reactor Coelant System Response . ... . . 4.4-1 4.4.3 Containment Response . .. .. . . . . . . . 4.4-2 4.5 Sequence No. 5 - T # 23
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4.5.1 Accident Sequence Desedption . ..... .,. 4.5-1 a 4.5.2 Reactor Coolant System Response . .... . 4.5-1 - 4.5.3 Containment Response . ... . . . . . . . . 4.5-2 4.6 Sequence No. 6 - AD . ...... . .... .... . 4.6-1 4.6.1 Accident Seqursnce Description . . . . . . . 4.6-1 4.6.2 Reactor Co61 ant System Response . . .. . . 4.6-1 4.6.3 containment Response . . . .. . . . . . . . 4.6 M 5.0 Plant Response with Recovery Actions . . . . . . . . . . 5.0-1
. 5.1 S D Sequences . ... ..... . . . . . . . . . . 5.1-1 2
5.1.1 Minimum Safeguards . ... . .. .. . ..
. 5.1-1 5.1.2 Full Restoration and Injection .' . . . . . . 5.1 -I 5.1.3 Secondary Side (Steam Generator)
Depressurization . .... . .... . . . . 5.1-2 5.2 S H2 Sequences . ...... .. . . . . . . . . . . 5.2-1 5.2.1 Minimum safeguards . . . .. . . . . . . . . 5.2-1 5.2.2 Partial Restoration of Recirculation . . . . 5.2-2
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i i k TABLE OF CONTEFTS (Continued) 4 5.3 TEB' seguences . . ... ...... . . . . . . . 5.3-1
' 5.3.1 Complete Power Restoration (Pre-Vessel Failure) . . . . . . . . . . . . . . .-. . 5.3-1 5.3.2
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Complete Power Restoration (Post-Vessel Failure) . . . . . . . . . . . . . . . . . 5.3-2 5.4 T23 E Se q uenc e s . . . . . . . . . .' . . . . . . . . 5. 4-1 5.4.1 Bleed and Feed - Ta3E.* . . . . . . . . . . . 5.4-1 5.4.2 * * * * * * * * * * *
- Feed and Bleed - T'23 5.5 AD Sequences . .. . .... . . 4. . . . . . . . . . $ 5-1 5.5.1 Minimum Safeguards . ..... . . . . . . . . 5.5-1 5.5.2 Full Injection Restoration of Injection . . . 5.5-1
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6.0 Fission Product Relesse, Transport, and Deposition . . . . 6.2-1 6.1 Introduction . . .. .... . . . . . . . . . . . . 6.2-1 f.2 Modeling Approach . . ........... . . . . . . 6.2-1 ; 6.3 Sequences Analyzed
, v' 6.3.1 S HF (Drains Blocked) . . . . . . . . . . . . 6.3-1 !
6.3.2 HF (Drains Open) . .. . . . . . . . . . . . 6.3-9 6.3.3 B' With a Seal LOCA . . . . . . . . . . . . 6.3-15 6.4 References . .. . . ... ... . . . . . . . . . . 6.4 " 7.0' Summary of Results . . .". ..... . . . . . . . . . . . 7.0-1 7.1 Base lasse . ... . . . ... . . . . . . . . . . . 7.1-1 7.2 Opert..:ir Action Cases . ... . . . . . . . . . . . . 7.2-1 j 7.3 Fission Product Transport . ... . . . . . . . . . . 7.3-1 I 8.0 Conclusions ( . .. ... . .. ... ... . . . . . . . . 8.0-1 . Appendices j Appendix A - Computer Code Inputs A.1 MAAP Parameter Fila . .. . . . . . . . . . . . . . . A.1-1 A.2 MAAP Input Listings . . .. . .. . . . . . . .. . . A.2-1 _ Appendix B - Accident Sequence Description . . . . . . . . B.1-1 ! _ Appendix C - Accident Signatures . . . . . . . . . . . . C.1 -iii-
Il.LUSTRATIONS Figure Page 3.1-1 Cross Section of Reactor Coolant System . . . .. . . 3.1-2 3.1-2 Steam Generator Cutawa'y. ... .... .. .. . . 3.1-5 l 3.1-3 containment Cross Section . . . ..... . . .. . 3.1-9 4 3.1-4 Containme'nt Cross Section . .. .... . . . . .. . 3.1-10 3.'-5 Containment Cross Section . . . . . .. . . .. . . 3.1-11 3.1-6 Containment Cross Section . . . .. . . . .. . . . 3.1-12 l 3.1-7 Ice Condens.ar Cutaway . . . . . . . . . . . . . . . 3.1-13 3.1-8 Reactor Cav:,ty Cutaway .. . .
. . . . . . . . . . 3.1-16 3.1-9 Emergency Cc<re Cooling System Flow Diagram . . . . . 3.1-18 1 1
3.2-1 Ice Condenser Containment N5dalization . . . ... . 3.2-2 j 3.2-2 MAAP Primary System Nodalization. ...... .. . 3.2-3 f 3.2-3 Ice Condenser PWR Safety and Other Systems . ... . 3.2-5 i
, 5.1-1 Minimum Safeguards - S 2D .. . . . . . . . . . . . 5.1-4 5.1-2 Full Restoration of Injection - S 2D . . . . . . . . 5.1-5 5.1-3 Full Restoration of Injection - S 2D . . . . . . . . 5.1-5 5.1-4 Full Restoration of Injection - S 2D . . . . . . . . 5.1-6 '{
5.1-5 Full Restoration of Injection - S 2D . . . . . . . . 5.1-6 5.1-6 Full Restoration of Injection - S 2D . . . . . . . . 5.1-7 l 5.1-7 secondary Side Depressurization - S2D . . . . ... . 5.1-8 5.1-6 Secondary Side Depressurization - S2D - * * * * * *
- S*I-0 5.1-9 Secondary Side Depressurization - S2D . . . . .. . 5.1-9 5.2-1 Minimum Safeguards - S 2H . . . . . . . . . . .. . 5.2-4
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m. i ILLUSTRATIONS (Continusd) l 5.2-2 Partial Restoration of Recirculation - S 2B -. . . . . 5.2-5
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. 5.2-3 Partial Restoration of Recirculation - S 2H . . .. . 5.2-5
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'( .[j 5.2-4 Partial Restoration of Recirculation - 5 2H . . .. . 5.2-6
- .1 j . r3 5.2-5 Partial Restoration of Recirculation - S 2H . . . . . 5 . 2 .6
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," 5.3-1 Coc:plete Power Restoration at 2.5 Hours - TEB' . . . 5.3-3
. 5.3-2 Conxplete Power Restoration at 2.5 Hours - TEB' . . . 5.3-3 5.3-3 Cocplete Power Restoration at 2.5 Hours - TEB' . . . 5.3-4 j 5.3-4 Cocrplete Power Restoration at 2.5 Hours - TEB' . . . 5.3-4 a ~5.3-5 Coc:plete Power Restoration at 2.5 Hours - TEB' . . . 5.3-5
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5.3-6 Cociplete Power Restoration at 5.0 Hours - TEB' . . . 5.3-6 l[ 5.3-7 Cocplete Power Restoration at 5.0 Hours - TEB' . . . 5.3-6 5.3-8 Couplete Power Restoration-at 5.0 Hours - TEB' . . . 5.3-7 ry [j 5.3-9 Couplete Power Restoration at 5.0 Hours - TEB' . . 5.3-7 r- 5.3-10 Couplete Power Restoration at 5.0 Hours - TEB' . . . 5. 3-8' W, V 5 . 4,-1 Bleed and Feed - T 23E
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- I'4-4 M 5.4-2 Bleed and Feed - T 23E . . . . .. . .. . .. 5.4_4 g
5.4-3 Bleed and Feed - T23E . . . . . .... .. .. . 5.4-5
'[' 5.4-4 Bleed and Feed - T23E , . . . .. ... . ... . 5.4-5 i
'1 5.4-5 Bleed and Feed - T23E . . . . . . . ... . . . . 5.4-6 i O 5.4-6 Bleed and Feed - T 23E . . . . . . .... .. . . 5.4-6
,i 5.4-7 Feed and Bleed - T 23E . . . . ... . ... ... . 5.4-7 5.4-8 Feed and Bleed - T 23E . . . . . . . . .. . . . . 5.4-7
, ,) 5.4-9 Feed and Bleed - T 23E . . . .. . . . . . . . . . 5.4-8
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IL2,USTRATIONS (Coneinued) {:t 5.'4-10 Feed and Bleed - T 23HL . . . - . . . . . . . .- . . . 5.4-8 j 5.4-11 Feed and Bleed - T 23ML' . . ... . . . . . . . . . . 5.4-9 -
- q 1 5.4-12 Feed and Bleed - T23hL . . . . . . . . . t . . . 5.4-9 ,
*k 'm 5.5-1 Minimum Safeguards - AD . . . . . . . . . . . . . . 5.5-3 .
5.5-2 FullRestorationofInjection-kD . . . . . . . . . 5.5-4
. 5.5-3 Full Restoration of Injection - AD . . . . . . . . . 5.5-4 5 . 5-4 Full Restoration of Injection - AD . . . . . . . . 5.5-5~
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5.5-5 Full Restoration of ' Injection - AD . . . . . . . . 5.5-5
', 6.3.1-1 S2HF (Drains Blocked) . . ... . . . . . . . . . . . . 6.3-5 6.3.1-2 S2 HF (Drains Blocked) . . . . . . . . . . . .. . . 6.3 . 6.3.1-3 S2 HF (Drains Blocked) . . .,,. . . . . . . . . . . . 6.3-6 !
4 6.3.1-4, S2 HF (Drains Blocke'd) . . . . . . . . . . . . . . . 6.3-6 6.3.1-5 S2 HF (Drains Blocked) . . . . . . . . . . . . . . . 6.3-7 6.3.1 S2 HF (Drains Blocked) . 6.3-7
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, 6.3.1-7 S2 HF (Drains Blocked) . . . . . . . . . . . . . . . 6.3-8 .
6.3.1-8 S2 RF (Drains Blocked) . . . . . . . . . . . . . . . 6.3-8 6.3.2-1 S2HF (Drains' Open) . . . . . . . . . . . . . . . 6.3-11 6.3.2-2 S2 HF (Drains Open) . . . . . . . . . . . . . . . . 6.3-11
- f. 6.3.2-3 S2 HF (Drains Open) . . . . . . . . . . . . . . . 6.3-12 ]
6.3.2-4 S2 RF (Drains Open) . . . . . . . . . . . . . .. . . 6.3-12 . 6.3.2-5 S2 HF (Drains Open) . . . . . . . . . . . . . . . . 6.3-13 6.3.2-6 S2 HF (Drains Open) . . . . . . . . . . . . . . . . 6.3-13 l 6.3.2-7 S2HF (Drains Open) . . . . . . . . . . . . . . . . 6.3-14 1 6.3.2-8 S2 RF (Drains Open) . . . . . . . . . . . . . . . . 6.3-14
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ILttiSTRNSIONS (ConCinu0d) 6.3.3-1 TMLB' ... . . . . . . . . . . . . . . . . . . . . . . 6.3-18
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6.3.3-2 TMLB' . . . . . . . . . . .. . . . . . . . . . . .- . 6.3-18
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,B 6.3.3-3 TMLB' . . . .. . . .. . ... . . . . . . . . . . . 6.3-19' I
[3 6.3.3-4 'TMLB' . .. . . . . . . . . . . . . . . . .. . . . . . 6.3-19
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6.3.3-5 'TMLB' . . ... . . . . . .. . . . . - . . . . . . . . 6.3 .
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.6- -6.3.3-6 TMLB' . .. . . . . . . . . . . . . . . . . . . . . . ,6.3-20
, .x 6.3.3-7 TMLB' . .. . . .. . . . ... . . . . . . . . . . . .. 6.3-21
. f, ' 4, d 6e3.3-8 TMLB' '. .. . . . . . . . . . . . . . . . . . .- . .- . . 6.3-21 i j t
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l, TABLES
- Table Page
'. 7 3.2-1 Initial Inventories of Fission Products and .
Structural Materials . . . .. . . . . . . . .. . . . 3.2-7
- 1 4.0-1 Primary and Containment System Status . . . . . . . . 4 . 0 -i.
4.1-1 S2D Event Summary . . . . . . . . . . . . . . . . . . 4.1-4 - 4.2-1 S2 H Event Summary . . . . . . . . . . . . . . . . . 4.2-5 4.3-1 S2 HF Event Sum ary (Drains Blocked) . . . . . . . . . 4.3-6 . 4.3-1 S2HF Svent Summary (Drains Open) . . . . . . . . . . 4.3-8 4.4-1 TEB' Event Summary . . . . ...... . . . . . . . 4.4-4 4.5-1 T 23 E Event Summary . . . . . . . . . . . . . . . . 4.5-4 4.6-1 AD Event Summary . . . . . . . . . . . . . . . . . . 4.6-A 6.3.1-1 Fission Product Release - S RF (Drains Blocked) . . 6.3-3 6.3.1-2 Distribution of Cs1 in Plant and Environment . . . . 6.3-4 6.3.2-1 Fission Product Release - S HF 2 (Drains Open) . . . 6.3-10
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" 6.3.3-1 Fission Product Release - TE3' . . . . . . . . . . . 6.3-17 7.1-1 Summary of Sequences . . . ... . . . . . . . . . . 7.1-3 7.3-1 Sequoyah Release Fractions Comparison with WASH-1400 7.3-2
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- Disclaimar This work was prepared by the Tennessee Valley Authority, in part, for I
,t use by the Atomic Industrial Forum. Neither'the Tennessee. Valley j.}a_
. Authority, the Atomic Industrial Forum, nor any of their employees, y makes any warranty, expressed or implied, or assumes legal liability
} g; or responsibility for the accuracy, or completeness of any information-contained in this report.
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Acknowledgement ( '[;
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.fj, . The authors and~ contributors . wish to acknowledge the efforts of several individuals in the preparation of-this report: '
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;j Messrs L. W. Lau, R. F. Christie, . and D. G. Renfro for manuscript it review; Messrs K. - D. Keith and 'M. L. Miller for input' d'ata review; and .
! the NEB secretarial staff for manuscript typing. In addition, our
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', ef forts and tribulations have benefited from discussion with ' '
i Dr. E. Fuller the IDCOR phenomenology manager. -
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1 d o LIST OF ACRONYMS 33 l .i e ft
.j' BWR Boiling Water Reactor 1~[U
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CRDM Control Rod' Drive Mechanism
...l..g, CST. Condensate Storage Tank .
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CVCS Chemical-and Volume Control System
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ECCS. Emergency Core Cooling System .'. ,. s i ' FSAR Final Safety Analysis Report
, , LOCA Loss of Coolant Accident Modular Accident Analysis Program E.l i} ,
MAAP Ed MSIV Main Steam Isolation Valve n [3 MSLB Main Steam Line Break PORV Power-OperatedeRelief Valve-
'v-n;l, PWR Pressurized Water Reactor s -
t ,, RCS Reactor Coolant System
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, 'I RHR Residual Heat Removal ae
,_ r). RWST Refueling Water Storage Tank UHI Upper Head Injection a
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! 1.0!. Introduction- * !I'q GJ 1.1 Statement of the Problem n-4:'
The'sain objective of this investigation is to calculate the thermal - d
}.
Q}'an : hydraulic and radiological. response of the Tennessee Valley Aut ority's
;jh 3
Sequoyah Nuclear Plant, p'rimary system and containment - for postulated
^'
ir severe accident sequences, i.e., those which have been identified as i et.-
' ts' - potentially leading to core degradation and melting. These sequences E- will be addressed ' on a realistic basis and .will include assessments' of
~
, .. ~
the results of operator intervention in these sequences. Similar r
];; studies have been performed for three other reference plants: Zion,
- r. Grand Gulf, and Peach Bottom.
.1 n .:
j7] The results of the containment analysis are incorporated into an
. c. .
I assessment of the fission product release and deposition within the (G
.,, ] various regions of the containment building. For sequences in which
- j. -
- p. ,
containment integrity is violated, the release of fission products to
"' the surrounding environment is calculated for inclusion in a separate
([ evaluation of the potential health effects associated with those L3 4 specific accident sequences. R I4 Ls
,, 1.2 Relationship to Other Tasks I (1 -
Ul The containment, analyses of IDCOR Subtask 23 are dependent upon the ] 1
; ". primary system and containment response models developed in Subtask
,s 1 16.2 and 16.3, " Executive Analysis Program," (Reference 1.1) and the i
., ;j fission product release, transport, and retention models developed in i I l
IDCOR Subtask 11, " Fission Product Behavior" (Reference 1.2). The j l'a-1 jy
-3 1.2-1 7
n kJ i
' * " " .,,-n -
-e, w. . - . . . _ , , . . ,
i dominant accident sequences used for the analyses along with the operator interventions were developed by considering the relevant or i : j key accident sequences presented in Subtask 3.2 (Reference 1.3). :.
.yl The ultimate structural capabilities of the reference plant I containments .md other typical designs were assessed in IDCOR Subtask .
I 10.1 (Reference 1.4). These analyses define the containment failure pressure and ft.ilure mode assumed in this analysis. For the Sequoyah containment this failure was identified as a breach at the containment spring line. Calculations of the rate and amount of fission products released from the containment, for those sequences which result in containment failure, were supplied to IDCOR Subtask }8.1 (Reference 1.5) to . formulate assessments of the health' consequences associated with these postulated accident saquences. These health consequence analyses were then supplied to IDCOR Subtask 21.1 (Reference 1.6) to evaluate the risk reduction potential for possible additional mitigating devices I considered for the Sequoyah Nuclear Plant. i Potential operator interventions were developed and applied to the specific accident sequences in the Sequoyah analysis to determine those potential actions which could terminate the accident sequence and result in a safe stable state. This was considered as part of IDCOR Subtask 22.1, (Reference 1.7), " Safe Stable States," which discusses potential means of terminating the various core damage sequences f
~
considered for the Sequoyah Nuclear Plant. l.2-2 1 1 -
.ii-
' L) g Finally, it should ba noted that th2 cnnlyses dsveloped as pcrt of LJ IbCOR Subtasks 16.2 and 16.3 involve the detailed consideration of many r.m
, ; ('(j different phenomena which are themsalves considered in separate IDCOR - t {. . subtasks. .These include: hydrogen generation, distribution and j.a ~
,,3 combustion (subtasks 12.1, 12.2, and 12.3), steam generation (subcask 14.1), core heatup (subcask 15.1), debris behavior (subcask 15.2), and
' ~
;.q correconcrete interactions (subtask 15.3) as discussed in Reference 1~ '
1.1. Detailed discussions of these topics can be found in the final 3 - s reports submitted for that specific task. Individual issues will be
- .t b- addressed only as required to understand the specific behavior obtained
,., for the accident sequences considered and the specific design "d
characteristics of the Sequoyah Nuclear Plant.
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4 1.3 References /;
.f_
1 1.1 "MAAP, Modular Accident Analysis Program," Technical Report on'
'j.
j IDCOR Subtaska 16.2 and 16.3, June 1983. , i 1.2 " Fission Product Transport in Degraded Core Accidents," Technical Report on IDCOR Subtask 11.3, December 1983. . s.
~
1.3 Technical Report on IDCOR Subtask 3.2. ,.
+
1.4 Technical Report on IDCOR Subtask 10.1. 1.5 Technical Report on IDCOR Subtask 18.1. . e . i 1.f 1,
- 1.6. Technical Report on IDCOR Subtask 21.1.
1.7 Technical Report on IDCOR Subtask 22.1.
}
0 i l- ! t-1.3-1 i. L---- _ _ . . . . _ . . _
i. 2.0 Stre. tory end Mithodolorv l
,1 j,
The basic strategy is to analyze some of the relevant or key accident
], sequences leading to a degraded core state. These analyses will first I i
I [*)I consider whether such sequences lead to core uncovery and damage and l ^. r ,, m then determine the progression of the accident for those sequences in " 4 which core degradation and melting is calculated. This analysis Tr includes the performance of the ECCS and the containment engineered
.1 f . safety systems, such as the UHI, ice condenser, containment sprays,
. :1 hydrogen igniters, RER system, etc.
;j t
, ,3, 13 The principal tool used te perform the containment thermal-hydraulic I, response analyses is the MAAP code (reference 2.1). This code
' [L considers the major physical processes associated with an accident F progression, including hydrogen generation, steam formation, debris
,d 4-g coolability, debris dispersal, core-concrete interactions, and hydrogen L- combustion. The FPRAT module for HAAP was used to evaluate the fission product release from the fuel. Natural and forced circulation within t #
1
; the primary system is modeled both before and af ter vessel failure and h
[j is integrated with the fission product release model to determine the transport of vapors and aerosols throughout the primary system and t .t N containment. Fission product deposition processe.c modeled include i vapor condensation, steam condensation, and sedimentation.
; f, ,;i, ' With the defined accident sequences, analyses were carried out for the f,} best estimate path of the accident progression including the fission i i-d product transport before and af ter reactor vessel failure and also 1 .
$ 2.0-1 1
*\
b i
l ' 3 after containment failure. Flows between the primary system and l j' containment and natural circulation flows within the primary sytem are l3 - Ij. included in this analysis. The primary system response following l3 '
/ vessel failure including heatup of the reactor vessel and its ,
A structures, is evaluated through the natural circulation models 'for 1 both primary system and containment. Fission product transport of both vapors and aerosols is datermined by these density driven flows. Included in this evaluation is the containment pressurization which l would be imposed upon the primary system, and would determine the l' magnitude and direction of flows between the primary system and containment. -
-In addition to the containment analyses discussed in this report, j uncertainty and sensitivity analyses have been performed on several key parameters associated with the accident response. These results are
~
reported in reference 2.2. Evaluations performed relating to containment bypass and failure to isolate scenarios are reported under IDCOR Task 23.5 (reference 2.3). - For each of the accident scenarios selected for analysis in this report, thermal-hydraulic calculations were performed both with and without operator intervention during the accident. The " base case" analyses, which assume only minimal operator reponse during the j accident, establish a reference system response during each of the ) i accident scenarios. The " operator action" analyses are branch l calculations of the base cases. These operator intervention cases 1 demonstrate the effect of a realistic operator response on the I ii 2.0-2 I
-i
_ _ _ _ _ . - I _ ________.__.________________o
lL t
. progroseica of en cecidsnt and provide e messure of tha time available to the operator for such actic.t G1 al.
1Q In the analysis of the containment response for the ice condenser
, ;;M -
,; containment design two features have been observed to provide 7, substantial accommodation for energy deposition and fission product '
source teres for a wide range of accident scenarios. These are the igniter system for hydrogen combustion at low concentration levels and
. s the ice condenser which condenses steam released from the RCS. The igniter system is modeled in terms of the number of igniters and their location throughout the containment compartments.
%O 9
I] For the Sequoyah Nuclear Plant, the ice condenser has a dominant u) - influence on the accident progression from several different response l=j
- q. ;
characterizations. First, overpressurization of the containment by ,. ~ r, steam can oniv occur if the ice bed has completely melted, which
-e -
requires substantial energy deposition and insufficient heat removal by the RHR system and containment sprays. Second, the ice condenser {
. runoff, when added to other sources of water in the lower compartment,
- jl i can result in a mass sufficient to flood the cavity and quench the core ma debris. Finally, the ice bed can retain substantial , quantities of V,
.J fission product material, specifically cesium and iodine, which would g
be released from the fuel during a core melt-down event. All gases
- evolved from the vessel would be forced through the ice bed to the i ,
t,, upper compartment either by differential pressures or by the air return
.a fans.
<3 i -
2 2.0-3 ; 9 C
C
~
These features: are included in the MAAP analyses carried out for the- : Sequoyah Nuclear Plant. These will'be presented in the following .d: sections starting with'the description of the plant.and its systems, ' I
.I - - the accident analysis models and the major assumptions associated'with' ~
the models, fo11cwed by the plant response,' recovery actions, and'the
- influence of selected mitigating features.
6 4
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i t e 1 I l' . j i r i 2.0-4 4 4 wemoge - _ h ---____m__-_-._-.---.____._..________.,_..m_ _
L; U-n N- 2.1 ' : Re ferences
..l- ',q ibij ll 2.1 "MAAP, Modular Accident' Analysis Program, User's Manual," Technical 4 ,-- -
j j ,?. Report on IDCOR Tasks 16.2 and 16.3, May 1983. g
'V t 5 e
- r. . L..
2.2 IDCOR Technical Report on Task 23.4, " Uncertainty and Sensitivity Analyses for the IDCOR Reference Plants," to be published. i, ,
. ( .,; 2.3 IDCOR Technical Report on Task ' 23.5, " Evaluations of Containment Bypass i L.. ,
. and Failure to Isolate for the IDCOR Reference Plants," to be i 7' published.
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3.0 Description cf Modnis end Major Assumptions d This section of the report describes the plant medel and major l O assumptions used in the IDCOR Task 23.1 analysis of the Sequcyah plant 1q 4 %
] using the MAAP computer code.
[h q 3.1 Plant Description
- 1 9 The Sequoyah Nuclear Plant is a two unit plant consisting of Westinghouse-designed reactor coolant systems with a rated thermal power
' b _'
- of 3423 MWe. An ice condenser pressure suppression containment is n
' !; employed along with several other uniit ue plant systems and features that
.. determine the overall thermal hydraulic and fission product response t.j J characteristics to degraded core events.. As a basis for understanding g the results presented later in this report, a description of the important geometric and system details is given in the following a
'G,: section. A review of the salient features of the MAAP code is then
~
presented'in conjunction with a discussion of input parameter
., determinations.
3.1.1 Reactor Coolant System Description \ ,.. The RCS consists of fours' imilar heat transfer loops connected in l parallel to the reactor pressure vessel. Each loop contains a reactor l, . i coolant pump, steam generator, and associated piping. In addition, i' the system includes a pressurizer, a prer,surizer relief tank, and interconnecting piping. All the above components are located in the (., ij containment building. Figure 3.1-1 indicates a typical reactor
,. , coolant loop cross section. The high elevation and U-tube design of a ,. 3 jd the steam generator creates the pote'ntial for condensation refluxing 1
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.. t I CROSS SECTION OF REACTOR COOLANT SYSTEM FIGURE 3.1 1
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b' j and countercurrent hot leg flow'during sequences with inadeq'uate l q ,] . primary system makeup. f - i, jq 3.1.2 Reactor Core , -- i C Two-hundred sixty-four rods are mechanically joined in a square 9 array to form a fuel assembly. One-hundred ninety-three assemblies make up the Sequoyah core. The fuel rods are supported at intervals along their length by grid' assemblies which maintain the lateral
.. spacing between the rods. The grid assembly consists of an " egg-
- i
- crate" arrangement of interlocked straps. T-straps contain spring
'a fingers and dimples for fuel rod support as well as coolant mixing C vanes. The fuel rods consist of slightly enriched uranium dioxide p]
i ceramic cylindrical pellets ce r,ained in Zircaloy-4 tubing which is .
, plugged and seal welded at the ends to encapsulate the fuel. A total p>
l mass of 222,645 lbm of uranium dioxide is used in a typical fuel .c '
""i loading. The approximate Zircaloy weight of the fuel. assemblies is 47,000 lbm. Potentially, complete oxidation of this zirconium could y
I ,' result in the release of over 2000 lbm hydrogen. All "cel rods are pressurized wit.:. helium during fabrication. ( I:,, d 4 p The core is cooled and moderated by light water at .a pressure of 2250 l3 lb/in2a. The coolant contains boron as a neutron poison. Boron concentration in the coolant is varied as required to control relatively slow reactivity changes including the ef fects of fuel
'd burnup. The CRDM are of the magnetic latch type such that upon a loss ,7 e
of power to the coils, the rod cluster control assembly is released
;J and falls by giavity to shutdown the reactor.
l ~' 3.1 - 3 nN L
h 3.1.3 Reactor Vessel , t
; The reactor vessel is cylindrical with a welded hemispherical bottom -
t, - , j head and a removable hemispherical upper head. The reactor vessel
] closure region is sealed by two hollow metallic 0-rings. The. vessel ,
contains the core, core support structures, control rods, and other parts directly associated with the core. The reactor vessel closure . head containo CRDM and UHI adapters. The bottom head of the vessel containss penetrations for connection and entry of the nuclear in-core instrumentation. Each in-core instrumentation tube is attached to the inside of the bottom head bf a partial penetration weld. It is this weld that is projected to fail under corium attack for the Sequoyah
! vessel. ,,
t 3.1.4 Steam Generator i The steam generator is a vertical shell and U-tube evaporator with
. f" integral moisture separating equipment as shown in Figure 3.1-2. The reactor coolant flows through the inverted U-tubes, entering and leaving through the nozzles located in the hemispherical bottom head of the steam generator.
, Feedwater at approximately 4300F flows directly into the annulus formed by the, shell and tube bundle wrapper before entering the boiler section of the steam generator. Subsequently, water-steam mixture flows upward through the tube bundle and into the steam drum section.
A set of centrifugal moisture separators, located above the tube bundle, removes most of the entrained water from the steam. Steam dryers are employed to increase the steam quality to a minimum of
! 99.75 percent (0.25 percent moisture). Recirculating flow from the 3.1-4 4
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.1 3.1-5
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h C moisture separators mixes with feedwater as it passes through the annulus forned by the shell and tube bundle vrapper. Steam exits the e generator at 857 lb/in2a with a flowrate of 3,730,000 lbm/hr per ~ steam generator. -
~
Each steam generator also has 5 safety valves with a total capacity of . 3,917,000 lbm/hr per steam generator. The set points for these valves range from 1064-1117 lb/in2 g An atmospheric relief valve is also provided on each steam generator with a capacity of 890,000 lbm/hr per steam generator at 1085 lb/in2 ,g ] 3.1.5 Reactor Coolant Pumps The , reactor coolant pumps are identical single-speed centrifugal units t i driven by three phase induction motors. The shaft is vertical with J the motor mounted above the pumps. A flruheel on the shaf t above the motor provides additional inertia to extend pump coastdown. The inlet ia at the bottom of the pump; discharge is on the side. The reactor i coolant pumps impart a total heat input of 12 MWt to the RCS. ' i j 3.1.6 Pressurizer The pressurizer is a vertical, cylindrical vessel with hemispherical top and bottom heads that is connected to the RCS on one of the hot
, legs of a reactor coolant loop. Electrical heaters are installed
'.'nrough the pressurizer bettom head while the spray no: , relief, and safety valve connections are located in the pressort upper -
head. The spray system condenses steam to prevent the pressurizer pressure from reaching the set point of the power operated relief 1 j 3.1-6 l
~!
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. . . - . . - - + - ..._.m. _ _ . .
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H ,. I hh. 1 valves' during a step reduction in powe.r level of- ten percent' of load. 'j
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f b)! The pressurizer is equipped with 2 power-operated relief valves which- { limit system pressure an'd thus prevent actuation of the fixed high
~
lp , pressure rwactor trip. The capacity of each of these valves is
203,600- les/hr at 2350 lb/in2 g The relief valves are operated automatically and can be opened by remote manual control to initiate !
, L once-through cooling.in degraded events. Operation of these valves
-, -aise limits the undesirable opening of the 3 spring-loaded safety <
G valves.- Remotely operated block valves are provided to isolate the [', power-operated relief valves if excessive leakage occurs. The safety C valves each have a capacity otr420,000'1bm/hr at 2485'lb/in28 rn
. p;-
The pressurizer reliaf tank is a horizontal,. cylindrical vessel with
... ci - i J
elliptical ends. Steam from the pressurizer safety and relief valves
.e P" is discharged into the pressurizer relief tank through a sparger pipe wJ under the water level. This condenses and cools the steam by mixing it with water that is near containment ambient temperature. Two 18' r
inch diameter rupture disks are provided on the tank for overpressure l-la protection. The disks fail at a pressure of 104.7 lb/in2d and
,- ry discharge into the lower compartment.
?
3.1.7 Containment Description The primary containment uses the ice condenser pressure suppression et i .d design. The containment, which has a n'et free volume of about ; 1,192,000 cubic feet, is divided into three major subvolumes, r]
!# including a 289,000 cubic foot lower compartment enclosing the reactor i ~,
and RCS, a 158,000 cubic foot ice condenser compartment enclosing.the
' .i EJ 3.1-7 e
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energy-sbsorbing ice bed in which steam is condensed, and a 651,000
]'
u
; cubic foot upper compartment which accommodates the air displaced from J
j the other volumes during postulated LOCA and MSLB accident. Figures ' ( i ,j 3.1-3'through 3.1-6 show' typical cross sectiors. y n The primary containment vessel is a free-standing, welded steel i structure consisting of a vertical cylinder, a hemispherical dome, and a concrete basemat with steel membrane. It has a design pressure of 12 lb/in 2g. IDCOR Task 10.1 (Reference 3.1) reviewed the ultimate pressure capacity of the Sequoyah containment shell and estimated a failu?s pressure greater than 50 lb/in2 g This value was used in these analyses. Design basis J.eakage is 0.25 percent per day at 12 lb/in2 g The shield building is a medium-leakage concrete structure ,
, enclosing the containment vessel and is designed to provide the .,
collection, mixing, holdup, and controlled release of containment k
- vessel fission product leakage following an accident. The annular !
region . Jean the primary containment and the shield building has a l free a..- space of 375,000 cubic feet. The ice condenser, Figure 3.1-7, is the primary pressure suppression . component. During .sormal plant operation, the ice bed (approximately 2.1 x 106 lbm of ice? is maintained ac about 15 degrees Fahrenheit by a redundant refrigeration system. Refrigeration ducts and insulation on the ice condenser val 1; serve to mimimize heat losses y , from the ice. The insulation within the ice condenser is sufficient
^
to prevent theicefrommeltingfor{aminimumperiodofsevendays folicwing a cuaplete los! of the refrigeration system. Inlet and
~
I out le t doors' are provided at the bottom and top of the ice condenser j 3.1-8 t 1 7 l _
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=
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. C- \
"a n=
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US wz < c-ns . == a < we z s Ex
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FIGURE 3.1-5 CONTAINMENT CROSS SECTION t lt 3.1 12 i
- {
e e
.e,.*-*er******"-'**'*'--' * ' * ~ .,m... .#....e. wee ==
u m 1 .j a i i
'n.
t s p" anonums J - commmara urramacas 6iJ L* H I. ,.. : p - CRAMEWALL
'y _
5 u errartExus asa
$f /
o g<h;. .(~_ l'd Y
"r a.
-- ; /
.. :s.
..e p)f .
e _ y ,-
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- 1. .
u : ...: a .. r-b (O1 x, -
* ' k:' h ,..
unEDancer.x ocoRs
\
l -- ,*. ., , ' w ~. .. ,j
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f .... lf h r.
-t g
i4 !cy l j>, ggs. " I
- g. ,' EE BASKET j ICE CONUEN5a f -
!D i.6d**=====#!! di
'"i ' '
} Q ;
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' hr, .'
's .. .l k
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IA
- FIGURE 3.17 lCE CONDENSER CUTAWAY
.. c, I
L 3.1 13 n s* i _ _ - - . ----.-.-_.____m_
j compartment. In the event of a LOCA, the lower inlet doors will open due 1
-j to the pressure rise in the lo' erv compartment caused by the release of the '.i j reactor coolant to the lower compartment. The differential pressure will 'i 4 then cause air, entrained water, and steam to flow from the lower -
compartment into the ice condenser. An operating deck separates'the upper and lower compartments and ensures that stear and air flow resulting from a LOCA is directed through the ice condenser to the upper compartment rather than through uncontrolled bypass paths. The resulting pressure l l rise, due principally to the increased air mass in the ice condenser at the start of an accident will cause the doors at the top of the ice
, condenser to open and allow the air to flow from the ice condenser to the upper compartment. Steam will'be condensed as it contacts the ice contained in the baskets in the ice condenser compartment and therefore i
does not appear in the upper compartment until the ice is depleted.
, . Virtually complete steam condensation is assured because of the ica mass and geometrical arrangement of the ice , columns. It is anticipated that
- substantial fission product retention will occur in the ice condenser.
A hydrogen igniter system consisting of electrically operated heaters is used in the reactor building containment to control hydrogen accum'21ation following severe accidents. A total of 68 igniters are currently used in the upper, lower, annular compartments and ice condenser upper plenum for this function (64 were conservatively l , assumed in this analysis based on an earlier plant configuration). l Design basis accident hydrogen concentration is controlled by two safety grade permanent hydrogen recombiners. Each recombiner i 3.1-14 I . I k . . . . . . . . _
l 7 processe s 100 sefm of containment atmosphere. The recombiners are located in the upper con:partment. 1
, m l
The reactor cavity, illustrated in Figure 3.1-8, is divided into a region
,)f!
)
! !;j directly below the reactor vessel and a region between the vessel and the j l
r- instrument tunnel. The former region is approximately 15 feet in diamefer l t- l and 20 feet high. The latter region is 35 feet in length and 23 feet in I i width. This unique design has important consequences in the behavior of I
. la t Sequoyah for degraded core accidents in that the geometric configuration j ,, precludes corium dispersal into the lower compartment (reference 3.10).
,a Fortunately, the cavity has a relatively large floor area for debri.=
4 cooling. The in-core instrumentation passes through an instrument tunnel
] starting at the seal table and intersecting the rectangular region at an
. t .) -
angle of approximately 60 degrees and 5 feet above the cavity floor. A p personnel access hatch is located at the upper end of the instrument
, g, ~
~
e
;n tunnel opening into the lower compartment There are two pathways for water to spill over into the cavity from the lower compartment. The first I pathway is through the reactor vessel nozzle penetrations in the reactor i
c shield wall. The second pathway is for water to accumulate above the M i!. u., personnel access hatch flooding the cavity via the instrument tunnel. 1%
\ :'
L 3.1.8 Containment Heat Removal System i The energy released to the containment following an accident is absorbed by the ice condenser. However, after the ice bed has =elted,
,{ mass and energy will continue to be released to the containment. The t..
concernment spray systems are designed to maintain the containment
]
-]. pressure, in the long term, below the containment design pressure, and
] 3.1-15
.J o
. e A
w 6 I + SEAL .,
'2
*AaLE e L
e ,
%? **
, , g .
- ACOL55 mafCM f wtL.w493 mau.
0 (
** . ,A EL s79 75 l
* !,' a (atACTom WsstL l 6 EL 636.0 *.
\ * {i t
- TYPCAL F S *
\.j',Q' ., .
. t e* - i
/ .
\
w . 8
/.
i
,/. * .
N M......... - - - e INaCORC
, MSTRUMENTA TugE TYPCAL & 54 N
Eb M (FLOOR m ., ,
. Lueg n- - - -
; s
,e
> *, j -
*e
. e
~
FIGURE 3.1-8 REACTOR CAVITY CUTAWAY 3.1 16 p 4
d r-
.p. eventually reduce the containment pressure to about atmospheric .q ]A pressure. !; Q
!12 g The containment' spray for the Sequoyah Nuclear P'lant is provided by p t*J two redundant. spray trains, each designed to provide the cooling capacity required to maintain the peak pressure at less than design f.e ; pressure for the full spectrum of design basis events. Each of the n . redundant containment spray train pumps delivers 4750' gallons per lA n minute to the containment. ' Additionally, 2000 gallons per minute may
'M
, LJ
. be diverted from one RHR pump and heat exchanger through a RER spray I' header. The containment spray pump is started by a containment
}'" pressure signal set at 2.81 lbfin2 ,g and containment spray' starts at Ib
*13 about 30 seconds after a large LOCA. Containment spray from the RER j ,_
pump may be manually initiated. a 2 = The containment' is equipped with a redundant air return fan. system. P c.j
- t. ,
Each of the two air return fan systems uses a 40,000 cubic feet per j minute fan to force air from the upper compartment back to the lower
'i , compartment. The air return fans are started by the containment 3 .
2 isolation signal, but the fan startup is delayed for 10 minutes to
~! p provide increased backpressure during the large LOCA core reflood.
i ls t 3.1.9 Emergency Core Cooling System a The ECCS is designed to provide core cooling as well as additional s ' .: a . shutdown capability for accidents that result in significant loss of
. ,rr water inventory from the reactor coolant system. The design basis is 1
's to limit clad damage due to excessive temperatures and cladding metal-7 water reactions. Important systems are diagrammed in Figure 3.1-9.
d 3.1-17 rn l 14d IC __u.____-._---.___.__m._.___ ___-_ _- - _
., 1 n5 '
. ngt. -{:
Eas ,j , a 4
; 4 .
j
. , ~
4 .
.r, o . ; i . , i -
c' ; I 9 Q_f i -!
" 1
. g
=_, 1 m: e:
g EEN l t zis m: x --- ' J f D E Q3, .I ,0 !!
< =e-
"ist, nabut "
- =
g
=
m: e b:t , ; ..S 5 _ .. 1, .4y. . . =. J 5:eJ., w s
.. . . . . . . . .p .. .. . . . . . . . . g . .g.yj;.g.g . Am . 4. . . . . g . .e. . . .q'r..
., E ... g g..
0"
) E
~
C' i l l E at !a s ne p \. - WWIM Mi S- M I if 1P IP m M e M _b. W M - g _ . -
-. . M __
g q ___ . - g g g _ r o ., M E ., , 4 o., E N c= t,o -,
=
w
# v
- 5* i h N g. I Q b W<
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_1_53 EX: 1 T
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- FIGURE 3.19 EMERGENCY CORE COOLING SYSTEM FLOW DIAGRAM i
. 3.1 12
____________a
9P "4,
'~
y Th2 ECCS consist of both passiva cud cctiva systams. The UHI and. low
.N a pressure accumulator tanks are passive s'ystems that are actuated when the reactor coolant pressure falls below 1255 lb/in 2a and 415 lb/in 2,,
p . respectively. The activ.e components of the ECCS_are high head (charging), medium (safety injection), and low pressure (RHR)~ pumps that m i j are actuated by a safety injection signal. Following a postulated 4 l i, accident, the passive and active injection systems may be called to -
- p. operate, and af ter the water inventory 'in the RW3T has been depleted; the a long-terS recirculation mode will be activated. The ECCS incorporates
- p.
~
,[ two subsystems which serve other Jfunctions. The RER system provides for v.,
- decay heat removal during resceor shutdown. At other times the RER
]q
. system is ' aligned for emergency core cooling operation. The centrifugal
,q charging pumps are utilized during normal operation for maintaining the
>M required volume of primary fluid in the RCS. Given an ECCS actuation' i~
j~ signal, the system is aligned to emergency core cooling operation and the
~~
CVCS. function is isolated. 1, s-.,
.a m The URI system consists of a borated water-filled tank connected to a
~
nitrogen tank that is pressurized. When the RCS pressure falls below fi[ 1255 lb/in2a , water will be injected into the top of the reactor
.w vessel. This system provides potential for top down quenching and upper D plenum cooling during degraded core events. Nominally,1839 cubic feet 9
f ., of 1200F water is available for injection into the upper head region
,'" using this passive system.
l . l . ;l i1 Each of the four low pressure accumulator tanks contains approximately [ ! f4 l !g 1000 cubic feet of borated water pressurized with nitrogen gas to I i S 3.1-19 L. J j l! ' l 4 e m n
-i j approximately 415 lb/in2a. When the RCS pressure falls below that in the accumulator tanks, water is forced into the four cold legs.
4
.j i
The EPI mode consists of the operation of two high head centrifugal 7-
.i t pumps, rated'for 150 gpm at 2500 lb/in2g , which provide high p'ressure injection of borie icid solution into .the reactor coolant system, upon actuation by a safety injection signal. Also part of the
'i high pressure injection mede are two safety injection pumps, rated for j c 425-gpm at 1100 lb/in23, which take suction from the RWST.
Low pressure injection consists of two RER pumps which take suction s from the RWST. The pump performance is 4500 gpm at 125 lb/in2g, j 1 Switshover from the injection to recirculation phase is accomplished manually with' automatic backup. i.e., automatic switching of RHR pump
.,.e ~*
suction from the RWST to the containment sump at a level 40,000 gallons below the lov level set points in the RWST. (Approximately 350,000 gal are injected from the RWST.) 3.1.10 Auxiliary Feedwater System The' auxiliary feedwater system is designed to supply unheated water to the steam generators for RCS sensible and decay heat removal. This need would occur when the normal feedwater system is not available. Therefore, the auxiliary feedwater system will be utilized during certain periods of normal startup and shutdown, in
^
j the event of malfunction such as loss of offsite power, and also, in the event of accidents. l l ' i 4 l 3.1-20 1 1 4
. w.s.me==-M-
Lf y- Th3 cuxiliary faadwatsr system contcins two motor-driven pumps end it one turbine-driven pump. Each motor-driven pump has a capacity of
.j-
[!9 ' 440 gallons per minute, at 2900 feet. head, which is sufficient for
! U$
; safe cooldown. The motor-driven pumps are connected to separate .
1' O? l L y emergency power buses. The turbine-driven pump has a capacity of 880 m, gallons per minute at 2600 feet head. '
,-f Steam supply to the auxiliary feedwater turbine is taken from one of
..i e two main steam lines at a point upstream of the MSIVs. Separate f~
g remote operated isolation valves are provided for these connections. c ',
. ?
a Normally, the auxiliary feedwater pumps take suction from two CSTs. [.] Each tank has a capacity of 397,700 gallons of which 190,000 gallons L is reserved for the auxiliary feedwater system by means of a
- l. standpipe in the tank. The CSTs are not designed to seismic Category
^#~
1 requirements; however, the essential raw cooling water system l-i provides an alternate source of water. All three auxiliary feedwater
, pumps will start automatically in the event of a safety injection f ,
signal, loss of offsite power, tripping of both main feedwater pumps,
' (f ' or tripping of one main feedwater pump if plant load is greater than v
80 percent. In addition, the motor driven pump starts automatically
- V)
() in the event of a two-out-of-three low-low water level signal in any 1 i steam generator. The turbine-driven pump also starts automatically i. in the event of a two-out-of-three low-low water level signal in any
.{l steam generator. Auxiliary feedwater flow will be adjusted by ,
c2 ? remote-operated flow control valves. l1
' if I
1 ma,. 3.1-21 l 4 o _._______o
_ _ = _ _ _ _ _ __ - - _
.j .
Ttis vslvas aceocisted' with the turbine-driven pump are served by both
-W L electric and control air subsystems. The turbine-driven pump A 'j.
receives control' power from a third direct current electrical channel .- 4 [ that is distinct from the channel serving the. electric pumps. Except f a - vi for the ommon supply line from the CSTs, the two ' reactor units have . separate auxiliary feedwater systems. ' 1 l
#9
.i: . -l l
9 i .s 1 1 e 4
.i I.
! 3.1-22
.t w___..-___-_--. - _ _ __ A
d 3.2 Moduler Accient Annivsis Program (MAAP)
.q d
W Within the IDCOR Program, the phenomological models developed in Tasks 11, jQ 12, 14, and 15 have been incorporated into an integrated analysis code a i3 j (MAAP) (reference 3.2) to analyze the major degraded core accident scenarios
- ')
b' for both PWRs and BWRs. MAAP is designed to provide realistic assessments -
.u for severe core damage accident sequences, including fission product ,
release, transport, and deposition, using first principle models for the r- major phenomena that govern the accident progression. The following
~
sections de' scribe the primary system nodalization and containment
. nodalization, the safety systems modeled in the MAAP-PWR code as applied to J .4 o the Sequoyah ice condenser containment design, the fission product release ;
p i,
;, model and the fission product deposition models. A complete Sequoyah i ; r p, parametep file is given in Appendix A.l.
l iAa .
! 3.2.1 MAAP Nodalization The MAAP plant model for a ice condenser containment is divided into O
L. several nodes as shown in Figure 3.2-1. Nodes exist for the upper c-, compartment (compartment A), lower compartment (compartment 3), il" ant.iular compartment (compart. ment D), reaccor cavity (compartment C), ice condenser, ice condenser upper plenum, quench tank (pressurizer iu . relief tank), and primary system. This nodalization provides detailed ! (? 1 ,, tracking of containment gas temperature, wall temperatures, and ; i steam / hydrogen concentrations as shown in Figure 3.2-1.
, ,1 !
'S The primary system is divided into ten nodes as shown in Figure 3.2-2.
')
> Nodes exist for the core region, upper plenum, downcomer, broken loop l jD) i
- .j cold leg, broken loop hot leg, unbroken loop cold leg, unbroken loop l ' 3.2-1 Id i >
i Q 3
4 i~ l < 4 , - '
/j u es l, .
. l l
.- caw.ermore i.. ; -
r
<a - . ., .
l l . es ee,..a= ; ,- .
,=q ,
, se
, r ,
' r
- m es ._;,..
, O '
; l ,
..e. m - '
co ** === - 11 , L o
- t. e l l 0 -:' 2 ;
==: ; p u c==**ataa - -
e r
- l e== [r
/
a m aa r ' ' f y e f'
, ', , "l comesarusaY;""l c 8' '
,* cuence fans d ra _ _ k' / 's "" L h
1 t
#e "
='
r y m i FM ,
.re.m a e t V
V h (i y' :i' iy' . v
- p. ,,< '
s
?
,4 7 .,
, trettu
- g'/, ^
< " * .f V Y r -
,!/, , :
. , .i, l UPPE A COesPAATMUs? (A) 8/M/0l e e
we e .. ; -
- ComtanssWENT F AM.umE
' up MAaL tu , e
- e a e I
* , SJM/S
'f5 e
e ~} esmoe 3 s sa to 1 i
. lass =Aaasei
,- 3 [cosessess l . e,,,,, gemee+== . - -
,,# ,e 31 1 - - .
. . +
w qi n ' i'3 '.e
! 's/ win- "
_ 808f4#8 048K i eye.cm
, seg I fama av venTy,
,towsA coupAAtusut tal " ' ' "
. , , , sewsM jg a ese ev ers" esses ae*
eMeessustask aswam t i .m .- a
- 3 l !
a,,e 3 g a m mises
.4,3
. . {3
- CAVffY (c)
KST: i 8 - er:Au W = WATRA M* MYO4048N 4=GA888 (ms onco. co s) c - comAms l 1
; Figure 3 2-1 Sequoyah ice condenser containment.
t i 3 2-2 . L e d I e 1- -- - -
- ,--,- - ,,-,,-------,-------,-------------------------,----,-,----,--,-,e- , - - - - - - - - - - - - - - - - . - - - - , - - - - , - . -
ga
- J W4 2O 1
. h. -
<o l 1
i h=t/)= 'd i
;Ej- ooW
., y3
)
e.a H xO , i -=4 C (
. -a N w d e:: .
4 1 x _; z
. ;1 i_ O<2
- - e - - - - --- - ;
D ,J CO
,ce;
.,. 'j = .
,. u . > c
/ =*
e p N i ** T. . n -
*. C g
'j W W N
"e C h D
.h) V3 >
6
- 4. U3 w .% E C .
'n a 4 lg 9 t> -
C W.. D WW l
- 3 ,
l z - l z~ l I s c W: ;._ u= i. r-
.' 3 C ;l c
' [j '
.- =
C l .d .
~.
r i C i.. . O
. ~
Q b W r ^., W :<, .: .' I A~Q g WC sw q O
.w 5
L -W=W Oww: ~d n C
~~'
.; I V) C W o n
, .I -
g C y) $
- u. -
x - g im - 5 c; (
- 7 C,
~
4
). , ---4.-.-- ~
A Z
. s i W
-) y x
, o
" ') m
.. 2 J C.
C Z e o 3 1 w & C ' c :: a 2 <C J .J c M
- OW c < W -.1 oa
? ,d U JW=W C >- W =
O' O M C f4 I . l rm sJ. 3.2-3 n
\- ... - _ _ _ _ _ - . - - _ - . _ _ - _ _ -
I- i J hot leg, pressurizer, and both the broken and unbroken loop steam
- j. -
I generator secondary side. This primary system nodalization permits a . b 5 detailed accounting of the water which is available for cooling the core and for reacting with the Zirealoy fuel cladding. In addition, i
. this scheme follows the user to track hydrogen and fission product concentrations through the primary system and thereby calculate release rates to the containment. The core is further divided into a user selected number of subnodes; a 7 radi*al x 10 axial nodalization is used for the Sequoyah analysis.
. 3.2.2 Safety Systems Modeled in MAAP The safety systema consideredin the Sequoyah reference analysis include the charging pumps, safety injection pumps, low pressure injection (or RER) pumps, UHI and cold leg accumulators, auxiliary
,, , feedwater, containment' sprays, ice and containment fans. These are shown in Figure 3.2-3 along with other systems important to accident progression such as the pressurizer and steam generator safety and ]
I power operated relief valves. All of these systems can be enabled or j disabled by the use of " event codes" in MAAP at the discretion of the user. The MAAP User's Manual (reference 3.2) gives a complete description of the use of MAAP and also compares the physical models with pertinent experiments. 3.2.3 Fission Product Release from Fuel The FPRAT module for MAAP, as adapted from reference 3.3 was used to calculate the release rates of fission products from the fuel matrix. These rates are dependent upon the fuel temperature history during 3.2-4
~
.l' I L
. . _ _ . . . -_ _ _ _ _ _ . _ _ m_____- _ _ - _ - - - - - - _ - _ - - - - - - - -
.d .
u%. t
. s/ . . .t .
6 .~~~
. e .
s__ .. .
~ .~.- e.. .
) .
= . .=. .~
. . .~ =s..
y e . . r = . 5 -:.~~. g.. g
- :~ =~:.
- ~.=.a.-
- :. e
- e
.:. G
~.
3: . s E ..': ..=.. p .
" . . ,t
) .
. s n .
m
~ .
.m t.
s ' .
. e t
e . s
. .. y s
T 8 r e g-h
.- w .
- - a e
. % t o
a .
/
g' / _, /
/ d
- g
. K.
, / . . n
. . - ,. / ,
- a g
, =
, /
,. //
. g ,
j , y
-s
,o r
.T
/ ,
t r
/
/ f e
~ .
i
. g
_ v
/
/ a
- y. .
/
/
s 7,/ Y/ . i ;
.. f R
- W r
1 r . P
~
1 1 n . c Sl. 1 . 1 g r 1 .
. e 9,
g . s n p,/i' g . g ' . e E g g_.
- w. .
r' d n
. .. I .. . .. .. .
o c T.
, ,,.,.yN5 t e f .
Y. 5 c g - .
. J .
7' I
]
,r .,, -
, < < , $ ,,i
,e ij / ,. ; ,i , ' .;
- r i t 3
C~ -
- 2
~
J
- _ n .
. ( . - . .. .. 3
-Il ,ti,,,.i;,t . .
. e
- e. .. - .
r l;
. u g
- a. . .. . i e' .
. F
- m- -
. s 4
.~ L 2 L L A Z E wnm o S JL . q ,1
. .s
-heatup an upon. characteristics of the atmosphere within the vessel which affect saturation of the chemical species'as discussed in IDCOR'
'l task 11.1 (reference 3.4). Fuel temperature histories for the 70 i d regions in the core were tracked to determine the release .-
r
, characteristics for the fission products and inert maerials. . The-initial inventories of the various fission products lwere obtained from reference 3.5 and are given in Table 3.2-1.
\.
The FPRAT calculation considers evaporation and condensation characteristics of various chemical species. Several key assumptions, consistent with the re*, commendations of IDCOR Task 11.1 were made regarding the physical form of' released fission products. These are: O e f t I l l . I l
?
t 4 lt 3.2-6 - i.. . >- _ . _ _ _ _ _ . . _ . . . . _ . . _
INITIAL EVE 1CCRIES CF FISSICN FRCOUCS
' A2D STRUCURAL M/LTIALS Fl S Sl o N PRODUCTS I N ITI AL I N VENTCRY (XG) {
h t 3- Ke . 17.0
.l H vs
- j .,, X. 330
- Cs 1 66 1 8
' ~i l 15.2 i !t ,ie 31.7 t
[', Sr- 60.9 i LJ m Ru 132 Le - 79.2 o) :
.y
% +197
~
i
. 'd ~
Sn 332 x (" t [.;' kk") 1
'l Ac 2257 l
="O $
l
..s h $ /M.
1 P, I'
"6
~
, ~ . ~ . ~
t (_ s
'I.8 l5.5
,I _, 3.2-7
!d I]
tj.'.
.]
d 1. Cesium and iodine combine to form Cs! ;:s entry to the fission I i product release pathway. The excess cesium forms Cs0H. Both , t ]s chemical species exhibit similar physical behavior, hence the source I rate for the Cs, I fission product group is assumed to be the sum of the Cs and I release rates. The form of this source is assumed to, be vapor.
- 2. Tellurium is assumed to enter the release pathway as vaporized Te0 2
- 3. Inert aerosol generation rate is the combined release rates for volatile structure materi* dis (Cd, In, Ag, Sn, and Mn).
1
- 4. Cesium, iodine, and tellurium are completely released during fuel
_ ,, ... heatup.
- f. Strontium and ruthenium are assumed to represent their respective nonvolatile fission product groups as defined in WASH-1400. Both were assumed to enter the release pathway in aerosol form. The melt l release for strontium and ruthenium was assumed to cease upon vessel failure beccuse the portion of the fuel hot enough to release these species would drop to the lower cavity. Releases in the cavity are calculated separately (see next section).
I 3.2-8 l 1 I _______________J
,; +
U 3.2.4 Fission Product ' Release and Aerosol Generation Resulting from Co e-1 Concrete Attack 4' .{,. a j .f'3-1 .
.i j.y The release of aerosols due to core-concrete attack was Getermined using
's a model based on the concrete ablation rates from MAAP. The mass of low
. g,
,.,'- volatility fission products and inert aerosols released from core debris i 2 is based upon a vapor stripping model assuming the melt constituents z.-
(.l, follow Raoult's law. This calculation is dependent upon the amount of gas sparging through the core debris, the molar concentration of fission
..1 11
products in the core debris, the vapor pressure of the chemical species
' . i3 of interest, and the temperature of the core debris.
\.3 7,-
gj The key assumptions are: ll i
-**j l . .) 1. The masses of CO2 and water vapor released per cubic meter
' [.7 ablated for the limestone concrete used at Sequoyah are 484 kg and l-
~'
108 kg, respectively. l .' b
- 2. Stripping only occurs when the corium is molten.
, . m. o r) .
- 3. The gases released by the downward attack pass through the molten pool and cause stripping. Cases generated by sidewall attack are b assumed to bypass the pool.
r1
- 4. The predominant form of Sr is St e- of Te is Te02,
- l ;,,I , f Ru is
,g elemental Ru, and of La is La 0 .
23
; .'d t
t., t LJ 3.2-9
.i t '.
m' I a
- . _ _ _ _ _ _. _ _ . _ _ _ _ . _ _ _ _ _ _ ___________d
. . . ._-)
?
k a
- 5. Inert aerosols of cao may be generated during core-concrete
?
attack. This chemical form is used as a surrogate for the various , s., concrete melt constituents that could be added to the corium pool. C I 9
- 9. ,j l 3.2.5 , <
Description of the Natural Circulation Model l MAAP models the primary system thermal-hydraulics, prior to and after vessel failure, including the effects of volatile fission product , . release. If large amounts of volatile fission products are retained .
, in the primary system after vessel failure, which is generally the case, the ' feedback mechanisms between fission product behavior and the {
thermal-hydraulics must be modeled. s The natural circulation model calculates the primary system fission '
- t. ,
product transport and thermal-hydraulics af ter reactor vessel failure, ,j
, ..and includes models for the following phenomena:
- a. Natural circulation flows due to temperature and density ,
differences around the primary system. - , I
. i
- b. Heat transfer between gas and structures in the primary system.. -
- c. Heat transfer between the primary system and the steam generator shells. ,
J
~
- d. Heat trans fer to containment through reflective insulation. This y' .1 treatment includes degradation of insulation performance due to .j long-term oxidation of the sr.ainless steel sheets in the
..l insulation. J 3.2-10 6
y suv + 't _.__________m. . _ _ _ -_--_
id I
- e. Fission product transport due to re-volatilization and subsequent lq
- condensation and sedimentation in cooler nodes.
, l 'i a O The chemical state of the fission products represents an uncertainty J in the calculations. IDCOR Subtask 11.1 (reference 3.4) identified the dominant chemical species for cesium and iodine to be cesium iodide and ensium hydroxide. Recent experiments (reference 3.7) show this may characterize much of the material, but significant quantities have also been observed to be irreversibly plated-out on steel 6., surfaces above the fuel region. For these analyses, the cesium iodine and cesium hydroxide fission products are assumed to have a vapor g pressurecharacteristicofcesbmhydroxide. Because of this d assumption, and the molar dilution of Cs1 by Cs0H, releases of Cs and -
': I to the environment quoted in this report should be regarded as
.a consisting mainly of Cs0H. If surface reactions between fission products and steel surfaces occurred, the vapor pressure could be greatly reduced. Experience with MAAP indicates that even large vapor b pressure reductions delay but do not ultimately effect the mobility of j cesium and iodine in the primary system. Thus, the basic behavior is the same as for the analyses reported here, but the releases to the
( ', environment are lower since less material is gaseous at the time of containment failure. If, on the other hand, vapor pressure reductions were so large as to prevent the movement of cesium and iodine before j' the melting temperature of steel was reached, the steel on which the
'! 'i fission products were deposited would melt, and the fission products
' ;l would drop to the reactor cavity and be added to the melt there. This would also be expected to result in a net reduction in the ultimate
- 0 3.2-11 )
l" I
- i
*a ,
i I
release to the environment since less material would be in a gaseous j state at the time of containment failure. Other analyses reported in
, _ reference 3.8 investigate the sensitivity of th'e release frac _tions.to the details of the representation of the Cs and 1 compounds and'
.}
l
., -i l include the use of recent experimental data for the vapor pressure of
{ l Cs0H. These recent data indicate that the Cs0H vapor pressure used in the calculations reported here is the best available.
~ , A 1
3.2.6 Fission Product Deposition IDCOR Task 11.3 applied state-of-the-art fission product behavior models to produce the RETAIN 3 ode, which describes the aerosol' agglomeration and deposition processes for both vapor and aerosol - forms of fisson products (reference 3.6). , These removal processes
- reduce the magnitude of radionuclides release to the environment. The
.c .
corresponding MAAP models depict physical mechanisms for vapor condensation on structures and aerosol retention due to steam condensation and gravitational settling. The agglomeration and sedimentation are represented as a removal rate that can be correlated as a function of the aerosol cloud density (reference 3.9). This
$' formulation is consistent with the available large, scale experimental resulta. Vapor retention is gove'rned by vapor condensation / evaporation on aerosol surfaces and valls. Mechanisms considered for aerosol retention are steam condensation and sedimentation. The MAAP nodalization scheme for fission product transport is identical to that . used for the thermal-hydraulic models in MAAP.
1 A h 3.2-12 1
I .j 3.3 References '1 l
'i I;Y 3.1 '" Containment Structural Capability of Light Water Nuclear Power J. )@ Plants," Technical Report IDCOR Subtask 10.1, July 1983. -
1 tj
.~~ .
{j 3.2 "MAAP, Modular Accident Analysis Program Users' Manual," Technical
; Report on IDCOR Tasks 16.2 an 16.3, May.1983.
tw ;: t Ij 3.3 " Analysis of In-Vessel Core Melt Progression," Technical Report on U .
~
{ IDCOR Subtask 15.13, September 1983.
, r.' . t.:
1
, g 3.4 EPRI/NSAC, " Technical Report 11.1, 11.4, and 11.5, Estimation of~
j' i Ia; Fission Product and Core-Material Source Characteristic's," October l "! 1982. iJ j- ~
.m i' 3.5 J. A. Gieseke, et al., " Radionuclides Release Jnder Specific 1WR , u:
Accident Conditions, PWR Ice Condenser Containment," Draft Report fd
- BMI-2104, July 1983.
- r.;
, ,3. 6 IDCOR Technic 1 Reprt on Task 11.3, " Fission Product Transport in f'
;i Degraded Core Accidents," December 1983.
c.,
, J 3.7 Richard K. McCardell, " Severe Fuel Damage Test 1-1 Quiet Look
,p' Report," EC&G Idaho, October 1983.
O I .m i 3.8 " Uncertainty and Sensitivity Analyses for the IDCOR Reference Plants," IDCOR Technical Report on task 23.4, to be published.
.j L) 3.3-1 m
) 3.9 " Fission' Product Deposition Models in MAAP," FAI Report, to b'e t
,i . published. i
;a. .
[ ,
}' 3.10 " Debris Coolability, Vessel Penetration, and Debris Dispersal,"
{ Techni, cal Report IDCOR Subtask 15.2B, August 1983. ! j s !
< \s f
61 s e$
*(.
4 l i . ,
,4*
e i
)
's I
e t t 9 6 1 i 4 9 1 4 1 1 3.3-2 '
~
m ,, pao e.o -%.- *-'
-%s*e -
, ,. ,. q,. . ... _
l ,~L b i< l3,, 4.0 Segunnees Anelyz2d l li Considerations of the dominant accident sampling sequene.es leading to 1.3 CE - potential core damage as given in the report of IDCOR Task 3.2, C[b 1 /1
- j "5 ~
resulted in six small LOCAs and two transient initiators, comprising 94.4 1 j h percent of the likely core damage initiators. These. sequences sere sa
-developed by reviewing the Sequoyah RSSMAP study with some regrouping of..
. !;,. oj. sequences. The AD accident sequence was added to determine the plant response to a 10 inch diameter LOCA. Translation of these sequences into s
the Sequoyah reference plant input model include the following assumptions:
~: ; ~.
ni. a. . J 1. All LOCA sequences incorporate manual reactor coolant pump trip via
! 7
~
j operator action subsequent to reactor scram.
! rn
' 1
- 2. Credit is taken for the full complement of emergency safeguards for'
..i ~
, 3 accident sequences where they are available unless otherwise specified.
i6
,Lk "
Table 4.0-1 illustrates the status of both primary and containment
'l
; .j systems for each accident sequence used in the analysis. The sequences analyzed are:
'1 :.
3
- 1. SD - Small LOCA with loss of ECCS injection, i
{i 2
' j L,
- 2. S2H - Small LOCA with loss of ECCS recirculation, 3 ui
'pj 3. S2HF - Smali LOCA with loss of ECCS and containment sprays in
, the recirculation mode,
- 4 4. TMLB' - Loss of all AC power and auxiliary feedwater, 4
4 -, 1
- 5. T 23HL - Transient with loss of auxiliary feedwater and loss of '
1 e.s i charging pumps, and
.] m 1 a, g 6. AD - Large LOCA with loss of ECCS injection.
i I 1 4.0-1 1 l ! 4 '.J 1 1
*, l
. % l
- l 9
7 _ _ PRIMARY SYSTEMS STATUS 2 j; EVENT. $2H 520 S2HF TML9' T23?.L AD. ,, j RCP-COASTDOWN X X X X X lX h 1 UPPER HEAD X- X X X X . X -
; INJ ECTI ON ,
$, CHARGI N G -
~
X X Pt.MPS . - SAFE.1f INJ X X _ PLMPS RHR PLMPS X X ,
-COLD LEG X X X ,
X X X ACCUMULATORS , ECCS REClR C - ECCS HT XCHNG N%IN FEEDWATER l
+
. AUX FEEDWATER X X X X CONTAINMENT SYSTEMS STATUS .
EVENT S2H S2D S2HF TMLB' T23ML AD i I AI R Re. i URN X X X X X FANS SPRAY X X lX X lX SPRAY REC 1R C X X X X SPRAY HT. XCHNG X X- X X I GNITORS . X X X X X
~
l
> l 1 ..
1, 4.0+2 .
L 4.1 Saguance No. 1 - S 2 D W (, 4.1.1 Accident Sequence Description 1 S2 D consists of a small LOCA initiator with subsequent failure of the i,..y]
'" ECCS in the injection mode. The ICCS continues to be unavailable in the i
p recirculation mode. Containment aafeguards systems (ice condenser, L: sprays, air return fans, and igniters) are available throughout the ,, m
, accident.
4.1.2 Reactor Coolant System Response i i f. Upon initiation of a 0.0218 ft2 cold leg break, the reactor is f^' scrammed, followed by reactor coolant pu=p coastdown and auxiliary ss j feedwater startup at five seconds. Figure's C.1-1 through c.1-5 illustrate ' the variables of interest. Immediately following break initiation, the
~
primary system pressure decrea$es to approximately 1250 lb/in2a . At-U- this time (approximately 0.2 hours) the UHI rupture disk fails and
~ * 'l relatively cool water injection is ini'tiated. The rate of inventory loss -
k._'. e
- out of the break is partially offset by the injection of UHI water. The
( ., primary system depressurization continues as decay heat is being { transferred to the steam generators and lost through the break. This b gradual depressurization continues until 0.8 hours at which time the core
'n- p begins to uncover. As the water level in the core continues to drop, the cladding temperature begins to increase. Approximately 0.2 hours af ter core uncovery the metal-water reaction initiates hydrogen generation.
- .s p
l., The primary system pressure continues to decrease as the remaining water from the UHI is injected (UHI water depletes at 2.02 hours). At ' { approximately 2.2 hours, the primary system pressure has dropped below r.. the 415 lb/in2a set point for the cold leg accumulators and cool water injection begins. At the time of injection initiation the reactor vessel 4.1-1
'Q'l 1o J -
i tra'ter 10vs1 is abrut .10 feet which indicates the bottom of the active
. core is' uncovered. The effect of this " bottom to top " reflood is to
- i. :[ - initially quench We lower nodes of the core. However, this quenching is j -
not maintained and the heat up of the injected water supplies steam to the cladding-water reaction and hydrogen production is restarted. As core nodes reach the melting temperature, the mass of molten core collecting on the core support increases until about 110,000 lbm (40 percent of the original core mass) have accumulated at 2.80 hours. At this time, the lower core support plate fails and the molten core material falls into the lower plenum of the reactor vessel.
, Approximately one minute later (2.81 hours), the molten core material fails one of the penetrations in the bottom of the vessel and the melt is discharged through the hole into the reactor cavity. Following the molten core, the remaining hydrogen, steam, and water is discharged into the cavity along with the remaining accumulator water. The core nodes remaining in the vessel continue heating adiabatically. As each node reaches $1440F it then falls into the cavity. The corium discharge rate after vessel failure decreases with the final core node reaching the melting temperature at 7.7 hours. Total hydrogen production from in-vessel.Zircaloy oxidation is 660 lbs. The average rate is 0.10 lb/see and the reaction is equivalent to a total core average clad oxidation of d
32 percent. 4.1.3 Containment Response Immediately following the accident initiation, the lower compartment pressuri:es as RCS inventory is discharged. At 64 seconds the i containment spray pressure set point is reached. The containment sprays take suction from the RWST until recirculation realignment occurs at 0.4 4.1-2 l 4 l l
--- -e==a- ** --
. , , , . _ , p.. -e, .
_____.-_-_-._____ - a
t At' 2.81 hsurs the vassel fcils casing a pressure spike' to .about
., hture. .., 3 ..
1 : d 21.0 lb/in2.a .The available air return fans, ice, and containment-
]. ] Qi= sprays rapidly decrease the pressure to appror.imately 18 lb/in2 a '
d:Id j .Since the ice has not been depleted' at this time, the temperature f,%.j ' ' :ys ', response 'in the upper compartment remains relatively constant. Pressure
~
- i. suppression is effective as anticipated. As.the ice continues to melt.,
- h.1 and RCS inventory is lost from the break, the water level in the lower j compartment exceeds the necessary curb height required for spilling water
. .,. 1
}M into the' cavity at approximately 0.8 hours. Therefore, by the time reactor vessel failure occurs, the cavity is flooded. This flooded s ,
j condition limJ.ts core-concrete ablation to the " jet" attack resulting in 9O a 0.14 ft penetration depth. The flooded cavity results in immediate h quenching of the corium. 1 I;j . i m l The remaining ice mass at time of vessel failure is approximately J ' 7 9.1x105 lbs (about 57 percent melted). At 4.92 hours all of the ice m
; fj64 has melted and containment pressurization begins. Following ice g depletion, the ice condenser and ice condenser upper plenum temperatures
- {": immediately increase to approximately the lower compartment temperature.
- a
] The containment sprays continue to remove heat from the containment i
. atmosphere with the continued molten corium discharge from the vessel and the decay heat from quenched debris generating steam. This heat removal
. . . , rate matches the decay heat se approximately 7.5 hours when the j
~I containment pressure reaches about 20.5 lb/in2a. Afterward, the l : Il containment spray heat removal rate exceeds that of decay heat and the
- Li
.! containment pressure continues to decrease, thus precluding containment failure.
g n, 1
' ,1 4.1-3 i l .J l
I ~
.j
TABLE'h.1-1 j S2D U1MAAP l 1 ) 1 SEC HR EVENT DESCRIPTION CODE i
! 0.0 0.00 REACTOR SCRAM 0.0 0.00 13 )
LETDOWN FLOW OFF - 46 ' O.O 0.00 AUX FEEDWATER ON 154 0.0 0.00 MSIV CLOSED 156" 0.0 0.00 PS BREAK FAILED 209 1 0.0 0.00 HPl FORCED OFF 216 l 0.0 0.00 LPI FORCED OFF 217 0.0 0.00 MANUAL SCRAM 227
; 0.0 0.00 CHARGING PWPS FORCED OFF 232 0.0 0.00 MAKEUP SWITCH OFF 242 0.0 0.00 LETDOWN SWITCH OFF 243 60.5 .02 MAIN COOLANT PLMPS OFF 4 60.5 .02 MCP SWITCH OFF OR HI-VIBR TRIP 215 63.7 .02 CONTMT SPRAYS ON 103 1466.3 .41 REClRC SYSTEM IN OPERATION 181 4
1466.*3 .41 RECIRC SWITCH: MAN ON 220
, 1483.9 .41 CH PLMPS INSUFF NPSH 183
- 1483.9 .41 HPl PWPS lNSUFF NPSH 185 t 2945.5 .82 FP RELEASE ENA8 LED -
14 4597.5 1.28 BURN IN PROGRESS IN 1/C UPPER PLENW 141 i 5122.7 1.42 SURN IN PROGRESS IN UPPER CMPT 102 5186.7 1.44 BURN IN PROGRESS IN ANNULAR CMPT 122 5545.2 1.54 BURN IN PROGRESS IN LOWER CMPT 75 5629.1 1.56 NO BURN IN LOWER CMPT 75 7281.5 2.02 UH1 ACCLM EMPTY 190 7774.8 2.16 BURN IN PROGRESS IN LOWER CMPT 75 8441.8 2.34 NO BURN IN LOWER CMPT 75 10037.3 2.79 SUPPORT PLATE FAILED 2 10100.4 2.81 RV FAILED 3 10115.5 2.81 BURN IN PROGRESS IN LCWER CMPT 75 10191.8 2.83 NO BURN IN 1/C UPPER PLENLM 141 10195.2 2.83 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 10199.3 2.83 ACCUMULATOR WATER DEPLETED 188
$ 10207.0 2.84 NO BURN IN LOWER CMPT 75 10210.8 2.84 NO BURN IN I/C UPPER PLENW 141 5 10213.8 2.84 BURN IN PROGRESS IN I/C UPPER PLENW 141 I
l i k.1 k
TABIZ k.1-1 aB 1"1 S2D U1MAAP CONT.
.1 d [id jt SEC HR EVENT DESCRIPTION CODE 10245.0 2.85 NO BURN lN 1/C UPPER PLENLM 141
' I? 10247.3 2.85 NO BURN IN' UPPER CMPT 102 2 10251.5 2.85 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 n 10261.4 2.85 BURN IN PROGRESS IN UPPER CMPT 102 j 11514.8 3.20 NO BURN IN UPPER CMPT 102 11554.3 3.21 BURN IN PROGRESS IN UPPER CMPT 102
. ii 11594.0 3.22 NO BURN IN UPPER CMPT 102
+* 11768.8 3.27 BURN IN PROGRESS IN UPPER CMPT 102 7, 11813.5 3.28 NO BURN IN UPPER CMPT 102 j 11956.2 3.32 BURN IN PROGRESS IN UPPER CMPT 102 11975.1 3.33 NO BURN IN UPPER CMPT 1.02
, 12026.6 3.34 NO BURN IN ANNULAR CMPT 122 12042.3 3.35 BURN IN PROGRESS IN ANNULAR CMPT 122 4
g 12120.6 3.37 NO BURN IN ANNUL,AR CMPT 122 tj 12126.2 3.37 BURN IN PROGRESS IN ANNULAR CMPT 122 12213.8 3.39 NO BURN IN ANNULAR CMPT 122
.p 12232.8 3.40 BURN IN PROGRESS IN ANNULAR CMPT 122 iE 12406.2 3.45 NO BURN IN ANNULAR CMPT 122 r, 12452.3 3.45 BURN IN PROGRESS IN ANNULAR CMPT 122 ,N ~
12500.7 3.47 NO BURN IN ANNULAR CMPT 122 12505.2 3.47 BURN IN PROGRESS IN ANNULAR CMPT 122 12582.2 3.50 NO BURN IN ANNULAR CMPT 122 [0 12593.2 3.50 BURN IN PROGRESS 1N ANNULAR CMPT 122 12593.2 3.50 NO BURN IN l/C UPPER PLENLM 141
) [,j 1'2604.2 3.50 BURN IN PROGRESS IN I/C UPPER PLENLM 141 12611.9 3.50 NO BURN 1N ANNULAR CMPT 122 P 12619.7 3.51 SURN IN PROGRESS IN ANNULAR CMPT 122 12624.1 3.51 NO BURN 1N I/C UPPER PLENLM 141
._ 12626.8 3.51 ~ BURN 1N PROGRESS 1N I/C UPPER PLENLM 141 l" 12711.2 3.53 NO BURN IN ANNULAR CMPT 122 12711.2 3.53 NO BURN IN 1/C UPPER PLENLM 141
;O 12764.4 3.55 BURN IN PROGRESS IN ANNULAR CMPT 122
!:) 12793.1 3.55 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 12800.1 3.56 NO BURN IN I/C UPPER PLENLM 141 l
- h 12807.0 3.56 BURN IN PROGRESS IN 1/C UPPER PLENLM 141
+
12842.0 3.57 NO BURN IN l/C UPPER PLENLM 141 :
!F lI' R l
l h.1-5
"J' GI2 k.1-1 ) S2D U1MAAP CONT.
I SEC HR EVENT DESCRIPTION CODEl 12856.0 3.57 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 12881.9
~
3.58 NO SURN IN I/C UPPER PLENUM 141 12916.8 3.59 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 12930.9 3.59 BURN IN PROGRESS IN UPPER CMPT 102" 12951.3 3.60 NO SURN IN UPPER CMPT 102 12975.2 3.60 NO BURN IN 1/C UPPER PLENUM 141 12988.9 3.61 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 13002.5 3.61 NO BURN IN ANNULAR CMPT 122 13002.5 3.61 NO BURN IN 1/C UPPER PLENUM 141 13032.5 3.62 BURN IN PROGRESS IN ANNULAR CMPT 122 13108.5 3.64 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 13117.4 3.64 NO BURN IN 1/C UPPER PLENUM 141 13135.3 3.65 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 13153.2 3.65 NO EURN IN 1/C UPPER PLENUM 141 13199.8 3.,67 NO BURN IN ANNULAR CMPT 122 13211.5 3.67 BURN IN PROGRESS IN ANNULAR CMPT 122 13226.8 3.67 NO BURN IN ANNULAR CMPT 122 13242.0 3.68 BURN IN PROGRESS IN ANNULAR OuPT 122 13257.3 3.68 NO BURN IN ANNULAR CMPT 122 13272.1 3.69 BURN IN PROGRESS IN ANNULAR Ch?T 122 13367.5 3.71 NO BURN IN ANNULAR CMPT 122 13398.9 3.72 BURN IN PROGRESS IN ANNULAR CMPT 122 13498.6 3.75 NO BURN IN ANNULAR CMPT 122 13513.5 3.75 BURN IN PROGRESS IN ANNULAR CMPT 122 13526.8 3.76 NO BURN IN ANNULAR CMPT 122 13533.6 3.76 BURN IN PROGRESS IN ANNULAR OWPT 122 13603.3 3.78 NO BURN IN ANNULAR CMPT 122 13620.2 3.78 BURN IN PROGRESS IN ANNULAR CMPT 122 13632.2 3.79 NO BURN IN ANNULAR CMPT 122 13646.8 3.79 BURN IN PROGRESS IN ANNULAR CMPT 122 13796.0 3.83 NO BURN IN ANNULAR CMPT 122 13805.0 3.83 BURN IN PROGRESS IN ANNULAR CMPT 122 13821.2 3.84 NO SURN IN ANNULAR CMPT 122
; 13828.6 3.84 BURN IN PROGRESS IN ANNULAR CMPT 122 q 13863.5 3.85 NO BURN IN ANNULAR CMPT 122 ]
13876.7 3.85 BURN IN PROGRESS IN ANNULAR CMPT 122 j h.1-6 ' i .
~~~ '
l
TA3r2 h'.i.1 _L 9, S2D U1MAAP CONT. !l SEC HR EVENT DESCRIPTION CODE b]'d 13892.5 3.86_ NO SURN,IN-ANNULAR CMPT _
-122 1A 13919.9 3.87 BURN IN PROGRESS 1N ANNULAR CMPT --
122 [$'_, 13994.5 14020.9 3.89 3.89 NO SURN-lN ANNULAR CMPT BURN IN PROG 9ESS IN ANNULAR CMPT 122
>122 .
>J 14040.9 3.90 NO BURN IN ANNULAR CMPT 122 14064.9 3.91 BURN IN PROGRESS IN ANNULAR CMPT 122 '
-r' 14166.3 3.94 NO BURN IN ANNULAR CMPT- 122 f 14193.6 3.94 BURN 1N PROGRESS IN ANNULAR CMPT 122 r 14340.4 3.98 NO BURN IN ANNULAR CMPT 122 l.i>
14397,.8 4.00' BURN IN PROGRESS IN ANNULAR CMPT 122-
;- 14449.6 44.01 NO BURN IN ANNULAR CMPT 122- -
q 14470.3 4.02 BURN lN PROGRESS IN ANNULAR CMPT 122
~ i :.' 14573.3 4.05' NO BURN 1N ANNULAR CMPT 122 14593.3 4.05 BURN IN PROGRESS IN ANNULAR CMPT 122 .3" 14607.7 -4.06 NO BURN IN ANNULAR CMPT 122 14626.3 4.06 BURN IN PROGRESS IN ANNULAR CMPT ,
122~
,r 14711.4 4.09 NO SURN IN ANNULAR CMPT 122 - .b; 14718.4 '4.09 BURN IN PROGRESS IN ANNULAR CMPT 122:
{F'- 14735.7 4.09 BURN IN PROGRESS IN UPPER CMPT 102 14757.2 4.10 NO BURN IN UPPER CMPT 102 14797.2 4.11 -NO SURN IN ANNULAR CMPT 122
+
14851.3 4.13 BURN IN PROGRESS IN ANNULAR CMPT 122 U 14912.4 4.14 NO BURN IN ANNULAR CMPT 122 14925.8 4.15 BURN IN PROGRESS IN ANNULAR CMPT 122 14939.1 4.15 NO BURN IN ANNULAR CMPT 122 f"]' 14957.8 4.15 BURN IN PROGRESS IN ANNULAR CMPT 122
;- 15002.7 4.17 NO BURN IN ANNULAR CMPT 122 i 15061.9 4.18 BURN IN PROGRESS IN ANNULAR CMPT 122 15134.4 4.20 NO BURN IN ANNULAR CMPT 122 0 15177.0 4.22 BURN 1N PROGRESS IN ANNULAR CMPT 122 1 15218.4 4.23 .NO BURN IN ANNULAR CMPT 122 n 15314.1 4.25 BURN IN PROGRESS 1N ANNULAR CMPT 122 'j i:. 15375.7 4.27 NO BURN IN ANNULAR CMPT 122 15390.8 4.28 BURN IN PROGRESS IN ANNULAR CMPT 122 !S U 15439.0 4.29 NO BURN IN ANNULAR CMPT 122 i 15488.2 4.30 BURN IN PROGRESS IN ANNULAR CMPT 122 .q
[a
~
e U , k.1-7
~
TABIf N.1-1 1 '
.l
' !' S2D U1MAAP-2 CONT."l ., SEC HR EVENT DESCRIPTION CODE / f 15554.1 4.32 NO BURN iN ANNULAR CMPT - 122 , 15641'.8 4.34 BURN IN PROGRESS IN ANNULAR CMPT
~
122 15722.0 4.37~ NO BURN IN ANNULAR CMPT 122
, 15759.5 4.38 BURN IN PROGRESS IN ANNULAR CVPT 122^ l 13825.5 4'.40 NO BURN IN ANNULAR CMPT '
122 I 15861.9 4.41 BURN IN PROGRESS IN ANNULAR OAPT 122 I 15941.9 4.43 NO BURN IN ANNULAR CMPT ~ 122. j 16995.2 4.44 BURN IN PROGRESS IN ANNULAR CMPT 122 16007.2 4.45 NO BURN iN ANNULAR CMPT 122 16062.6 4.46 BURN IN' PROGRESS IN ANNULAR ~CMPT. 122 16108.0 4.47 NO BURN IN ANNULAR CMPT 122 16172.7 4.49 BURN IN PROGRESS IN ANNULAR CMPT 122 16213.3 4.50 NO BURN IN ANNULAR CMPT
~
122 16267.6 4.52 BURN IN PROdRESS IN ANNULAR CMPT 122 16344.4 4.54 NO BURN IN ANNULAR CMPT 122 16355.5 ~4.54 BURN IN PROGRESS IN ANNULAR CMPT 122 16433'.5 4.56 NO BURN IN ANNULAR CMPT 122. 16484.7 4.58 BURN IN PROGRESS IN ANNULAR CVPT 122 16 39.9 4.59 NO' BURN IN' ANNULAR CMPT 122 16598.3 4.61 BURN IN PROGRESS IN ANNULAR CVPT 122 16680.0 4.63 NO BURN IN ANNULAR CMPT 122 16739.6 4'.65 8 URN IN PROGRESS IN ANNULAR CMPT 122 16750.9 4.65 NO BURN IN ANNULAR CMPT 122 16773.6 4.66 8 URN iN PROGRESS IN ANNULAR CMPT -122 16784.9 4.66 NO BURN IN ANNULAR CMPT 122 16816.2 4.67 BURN IN PROGRESS IN ANNULAR CMPT 122 16907.2 4.70 NO BURN IN ANNULAR CMPT 122 ' 16926.8 4.70 BURN IN PROGRESS IN ANNULAR CMPT 122 - 16994.8 4.72- NO BURN IN ANNULAR CMPT 122 17010.8 4.73 BURN IN PROGRESS IN ANNULAR CMPT 122 17018.1
~
4,73 NO BURN IN ANNULAR CMPT 122 17025.4 4.73 BURN IN PROGRESS IN ANNULAR CMPT 122 17052,9 4,74 NO BURN IN ANNULAR CMPT 122 17078.6 4.74 BURN IN PROGRESS IN ANNULAR CMPT 122 l
- - 17085.2 4.75 NO BURN IN ANNULAR CMPT 122 1 17091.8 4.75 BURN IN PROGRESS IN ANNULAR CMPT 122 i
f 4.1-8 ['
2 !!1 S2D U1MAAP CONT.
'T ; ~
SEC HR EVENT DESCRIPTION CCDE 17131.1 4.76 NO SURN IN ANNULAR CMPT , 122 g 17144.0 4.76 BURN IN PROGRESS IN-ANNULAR CMPT _ 122 [j 17157.7 4."77 NO BURN IN. ANNULAR CMPT 122 17198.5 4.78 BURN IN PROGRESS IN ANNULAR OAPT 122 .
? 17215.6 4.78 NO BURN IN ANNULAR CMPT 122 17251.9 4.79 BURN IN PROGRESS IN ANNULAR CMPT 122 g 17317.5 4.81 NO BURN IN ANNULAR CNFT 122
. 17375.9 4.83, BURN 1N PROGRESS IN ANNULAR OAPT 122 17434.4 4.84 NO BURN IN ANNULAR CMPT 122
.ri 17448.5 4,85 BURN IN PROGRESS IN ANNULAR CMPT- 122 U 17510.7 4.86 NO BURN IN ANNULAR CMPT 122
,, 17537.2 4.87 BURN 1N PROGRESS IN ANNULAR CMPT 122
'y 17597.8 4.89 NO BURN IN ANNULAR CMPT 122
, 17614.4 4.89 BURN IN PROGRESS IN ANNULAR CMPT 122 f~; 17650.4 4.90 NO BURN IN ANNULAR CMPT 122 17654.6 4.90 BURN IN PROGRESS IN ANNULAR CMPT 122
.., . 17707.3 4.92 NO BURN IN ANNULAR CMPT 122 17729.6 ICE DEPLtitD i 4.92 132 e f 17731.8 4.93 BURN IN' PROGRESS IN ANNULAR CMPT 122 m 17792.0 4.94 NO BURN IN ANNULAR CMPT 122-
,. -)
..b s
!a
(, , l% : 9
,aq n
r
. i ;J f' (q*= h.1-9
f I 4.2 S'qu:nce Ns. 2-$ 2H
. 4.2.1 Accident Sequence Description i
j S2 H consists of a small LOCA initiator with subsequent failure of the i
.] ECCS in the recirculation mode. Emergency core cooling in the injection ,
mode is successful and the containment safeguards systems (ice condenser, sprays, air return fans, and igniters) are available throughout the accident. 4.2.2 Reactor Coolant System Response Upon initiation of a 0.0218 ft2 cold leg break, the reactor is scrammed, followed by reactor pump coastdown, and auxiliary feedwater startup at five seconds. Figures C.2-1 through C.2-5 illustrate the variables of intere.st. Immediately following break initiation, the primary system pressure decreases to approximately 1250 lb/in2a During this depressurization I period (0.0-0.1 hours) high pressure injection charging pumps and safety injection pumps start and UHI initiates injection at 1255 lb/in2a. This introduction of cool water into the reactor vessel results in initially cooling the primary system water. The primary system water mass continues to increase until 0.37 hours when the recirculation switchover point is reached. This increase in primary system ir.ventory and cooling results in decreasing the secondary side temperature and preseure. Since the primary system pressure is continually decreasing after unsuccessful recirculation switchover, the UHI continues to inject past 0.37 hours. This continued injection cools the primary and secondary side until a minimum pressure of about 1000 lb/in2a is reached in the primary system. At this point, the primary side temperature and pressure begins to increase because of the l t decreased heat removal capability of the steam generators, due to secondary
, side heat.ing. The primary side pressure increase results in termination o:
UHI injection. 4.2-1 i
. . _ . . . . _ . . - - . - . . . - ~ _ - -
, 4 J
Both primary and ascond ry prassuriza to tha szcondary sida relisf. f5 [ valve set point of approximately 1100 lb/in2 a Nich no more water available for injection, reactor coolant inventory starts 4a M
.i;- decreasing within the primary system. The primary system pressure
.i .
. remains somewhat constant until about 1.3 hours. At this time, the
[:} 13 reactor vessel water level falls below the top of the core and , i i ,~ superheated steam bcgins to exit the core. As the water lu el in the core continues to decrease, the cladding temperature increases. 2 Approximately 0.2 hours after core uncovery, the cladding metal-water [~i. reaction initiates significant hydrogen generation. The increasing void o m s.- in the primary. system coupled with the increased flow out of the break j'
.,4 causes a depressurization at a relatively constant rate until 1.5 hours.
i At this time, the pressure has decreased enough for UHI initiation. UHI
!O i j j continues to inject until depletion occurs at about 2.3 hours, after
)
. '2 which the injected water is quickly heated to reactor vessel conditions.
.)
~ During the period 1.5-2.0 hours, the UHI is insufficient to quench the l n; l
; fuel resulting in continued hydrogen production.
; 4
, p l h j
'j At approximately 2.45 hours, the primary system pressure has decreased to j l
3 the cold leg accumulator set poir.c (415 lb/in2 a) and bottom-to-top reflood is initiated. This results in providing additional' water for ( p steam production and further oxidation of the cladding as indicated by the re-initiation of hydrogen production. Continued accumulator f' d discharge causes the vessel water level and mass to increase as the m. pressure decreases to approximately 350 lb/in2a. As the core concinues i d to heat up, the first node reaches the melting temperature of 5144or at j il 1.9 hours. Increased heating and node melting results in the molten core d collecting on the core support plate until about 110,000 pounds have l q 4.2-2 l Ul l l
'"1 n
J l
.h_.____-_-__________
accumulated at 3.30 hours. At this time, the lower core support plate q fails and the molten ccre material falls into the lower plenum of the I reactor vessel. Within one minute, the molten core material fails one of
~j the penetrations in the bottom head of the vess'el and the molten core l
material is discharged through the hole into the reactor cavity. Following the molten core, the remaining hydrogen, steam, and water is discharged into the cavity along with the remaining accumulator water. The core nodes remaining in the vessel continue heating adiabatically
, with each node draining into the reactor cavity when it reaches 5144cF.
The corium discharge rate af ter vessel failure decreases, with the final core node reaching the melting temperature at 8.4 hours. A total U hydrogen mass of 680 lbs is g3nerated with an average hydrogen production t rate of 0.09 lb/sec. This corresponds to an overall Zircaloy clad oxidation of 33 percent. I _h
- 4.2.3 Containment Response Immediately following the accident initiatica, the lower compartment pressurizes as RCS inventory is discharged. At 65 seconds, the pressure set point for the containment spray is reached. The containment sprays ,
take suction from the RWST until the recirculation alignment occurs at 0.37 hours. At this point the sprays recirculate , vater from the containment sump. At 3.31 hours when the vessel fails the lower compartment pressure increases to about 21.5 lb/in2 a However, the air return fans, containment sprays, and available ice reduce this pressure [ to approximately 18 lb/in2 a The water level in th'e lower compartment i 1 exceeds the necessary curb height required for spilling water into the i j cavity at approximately 0.8 hours. The r e fore , by the time the reactor a i l vessel failure occurs, the cavity is flooded. This flooded condition l 4.2-3 )
)'
~
l L_____-_____. . _ _ - .1
l 3 LJ limits c:ro-cenereto ablation to tha jet attack only resulting in a 0.15 I?
', 15 ft penetration depth. The flooded cavity results in the.immediate l 4
l
- quenching of the corium.
o .I . - l ? 1 *1 . i +' ' i
'i -
4 ms
.:!d The ice remaining at the time of vessel failure is approximately l i
ti ! 5.75x105 lbs. At 4.55 hours all the ice has been melted and the . 1 s;
;; containment pressure rapidly increases due to loss-of the passive ice j t
3 heat sink. The containment sprays continue to remove heat from the l C containment atmosphere, but lag the input decay heat energy until.7.0 a. { hours,'at which time the containment pressure of about 20 lb/in2 a ; is reached. Afterward, the-containment s' pray heat removal rate l
, q
\7 exceeds that of decay heat and the containment pressure continues to j 1
decrease, thus precluding containment failure.
!. i i
. v: =
1.
-g n
, I . '!
p. yf*
,~,
S ' b- .
~
. i il
.U g.
p e l
. [)
- ,1
, s, G4*
n n l' iyv N ; 3 . _} . ,
*) 4.2-4 d I 9
~1 T.A3LE k.2-1 O-4 j S2H.U2MAAP :-
j[ 1 SEC. HR EVENT DESCRIPTION CODE .. I 0.0 0.00- REACTOR SCRAM 13 _ I ~ 0.0 0.00 LETDOWN FLOW OFF 46 {
~
0.0 0.00 AUX FEEDWATER ON . 154 y 0.0 0.00 MSIV CLOSED 156 {
, 0.0 'O.00 PS BREAK FAILED 209 {
0.O O.00 MANUAL' SCRAM 227 { 0.0 0.00 MAKEUP SWITCH OFF 242 { O.0 0.00 LETDOWN SWITCH OFF 243
.47.6 .01 CHARGING PUWPS'ON 11 61.1 .02 NRI N COOLANT PLMPS . OFF . 4 61.1 .02 MCP SWITCH OFF OR Hi-VIBR TRIP 215 64.9 .02 CONTMT- SPRAYS ON 103 160.3 .04 HPI ON , 5 1342.3 .37 HP! OFF -
5 1342.3 .37 CHARGING PUMPS OFF 11. ,
,1342.3 .37 RECIRC SYSTEM IN OPERATION 181
~
1342.3 .37 HPl FORCED OFF 216 l 1342.3 .37 LPI FORCED OFF 217 l 1342.3 . 37 REClRC SWITCH: MAN ON 220 - ! 1342.3 .37 CHARGING PUMPS FORCED OFF 232 l, 1354.9 .38 CH PUNPS INSUFF NPSH 183 L 1354.9 .38 'HPl PUMPS INSUFF NPSH 185 l 4442.3 1.23 FP RELEASE ENABLED 14
-5818.5 1.62 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 6294.7 1.75 -BURN IN PROGRESS IN LOWER CMPT 75 ,
6426.6 1.79 BURN IN PROGRESS IN UPPER CMPT 102 l, 6493.3 1.80 BURN IN PROGRESS IN ANNULAR CMPT 122 ) '; 6601.6 1.83 NO SURN IN LOWER CMPT 75 l 7213.5 2.00 SURN IN PROGRESS IN LOWER CMPT 75 7280.2 2.02 NO BURN IN LOWER CMPT 75 8400.1 2.33 UHI'ACCLM EMPTY 190
; 9449.7 2.62 BURN IN PROGRESS IN LOWER CMPT 75
- 9895.2 2.75 NO BURN IN LOWER CMPT 75 1 10234.8 2.84 BURN IN PROGRESS IN LCWER CMPT 75 I 10458.2 2.91 NO SURN IN LOWER CMPT 75 11771.0 3.27 BURN IN PROGRESS IN LOWER CMPT 75 l
j . 1 -, 1 1 L& .2-5 -
~ ~
~
':'ABI2 40 2-1 iu i
lG !* S2H U2MAAP CONT.
)i SEC HR EVENT DESCRIPTION CODE l 11868.9 3.30 SUPPORT PLATE FAILED 2 9 11915.4 3.31 NO BURN IN LOWER CMPT -
75 11927.3 3.31 RV FAILED 3 11936.7 3.32 BURN IN PROGRESS IN LOWER CAPT 75 12021.4 3.34 ACCUMULATOR WATER DEPLETED 188
~
12051.6 3.35 NO BURN IN UPPER CMPT 102 12058.4 3.35 NO BURN IN LOWER CMPT 75
- 12062.9 3.35' NO BURN IN 1/C UPPER PLENLM 141 j r, 12063.8 3.35 BURN IN PROGRESS IN 1/C UPPER PLENLM 141
[3 ' 12079.5 3.36 NO BURN IN 1/C UPPER PLENLM 141 12084.9 3.36 BURN IN PROGRESS IN l'/C UPPER PLENLM 141 12105.7 3.36 BURN IN PROGRESS IN UPPER CMPT 102 13429.'2 3.73 NO BURN I UPPER CMPT 102 I g 13523.7 3.76 BURN IN PROG #tSS IN UPPER CMPT 102
.j 13550.7 3.76 NO BURN IN UPPER CMPT 102 ;
13581.0 3.77 BURN IN PROGRESS IN UPPER CMPT 102 l n 13597.7 3.78 NO BURN IN UPPER CMPT 102 IJ 13599.0 3.78 BURN IN PROGRESS IN UPPER CMPT 102
,. 3 13647.7 3.79 NO BURN IN UPPER CMPT 102
,l 13685.1 3.80 NO BURN IN ANNULAR CMPT 122 13722.7 3.81 BURN IN PROGRESS IN ANNULAR CMPT 122 l,
13784.0 3.83 NO BURN IN ANNULAR CMPT 122
- 13853.5 3.85 BURN IN PROGRESS IN ANNULAR CMPT 122 13943,6 3.87 NO BURN IN ANNULAR CMPT 122 13951.0 3.88 BURN IN PROGRESS IN ANNULAR CMPT 122 14097.5 3 92 NO BURN IN ANNULAR CMPT 122
.3 3.93 NO U N N CM 22
, 14182.3 3.94 BURN IN PROGRESS IN ANNULAR CMPT 122 14260.0 3.96 NO BURN IN ANNULAR CMPT 122 14294.6 3.97 BURN IN PROGRESS IN ANNULAR CMPT 122 0 14301.6 3.97 NO BURN 1N ANNULAR CMPT 122 u 14307,1 3.97 BURN IN PROGRESS IN ANNULAR CMPT 122
,.., 14380.4 3.99 NO BURN IN ANNULAR CMPT 122
' (~ j 14388.2 4.00 BURN IN PROGRESS IN ANNULAR CMPT 122 i~ 14488.5 4.02 NO BURN IN ANNULAR CMPT 122
." 1 l0 f,a3 L.2-6 l j
TAs u h.2-1 j . I 7 1 S2HLU2MAAP .
' CONT.
SEC. HR EVENT DESCRIPT1ON CODE I 14521.7 4.03 . BURN IN PROGRESS IN ANNULAR CVPT 122, . 14638.4 4.07.
~
1 BURN IN PROGRESS IN UPPER CMPT 102 . 14665.2 4.07 NO BURN IN UPPER CMPT 102 1 14703.1 4,08 NO BURN'iN ANNULd8 CMPT 122' 14775.6 4.10 BURN IN PROGRESS IN ANNULAR CMPT 122 14783.8 4.11 NO BURN-IN ANNULAR CMPT 122 14792.0 4.11 BURN lN PROGRESS IN ANNULAR CMPT 122 14808.4 4.'11 -NO SURN IN ANNULAR CVPT' 122 14840.5 4.12 EURN IN PROGRESS IN~ ANNULAR CMPT 122
, 14892.9 4.-14 NO BURN 1N ANNULAR CMPT -122 14926.2 4.15 NO BURN IN 1/C UPPER PLENLM' 141' 14939.6 4.15 BURN' IN PROGRESS IN 1/C UPPER PLENW 141 14959.6 4.16 NO BURN IN J/C UPPER PLENLM 141 14981.1 4.16 BURN IN PROGRESS.IN ANNULAR CMPT 122 g 14994.4 4.17' BURN IN-PROGRESS IN I/C UPPER PLENLM 141 15075.8 4.19 NO BURN IN I/C UPPER PLENUM 141- .
15084.4 4.19 BURN IN PROGRESS lN 1/C UPPER PLENLM 141 15133.8 4.20 NO BURN IN I/CLUPPER PLENUM 141 15152.7 4.21 NO BURN IN ANNULAR CMPT 122 15200.3 4.22 BURN IN PROGRESS IN' ANNULAR CVPT 122 15288.9 4.25 NO BURN IN ANNULAR CMPT 122 15324.8 4'.26' BURN . IN PROGRESS iN ANNULAR CVPT 122~ 15392.6 4.28 NO BURN'IN ANNULAR CMPT 122-15481.1 4.30 BURN IN PROGRESS IN ANNULAR CMPT 122 ! 15558.2 4.32 NO BURN IN ANNULAR CMPT 122 15575.6 4.33 BURN IN PROGRESS IN ANNULAR CMPT 122 15672.1 4.35 NO BURN IN ANNULAR CMPT 122 15680.5 4.36 8 URN IN PROGRESS IN ANNULAR CMPT 122 15765.2 4.38 NO BURN IN ANNULAR CMPT 122 15779.4 4.38 BURN IN PRCGRESS IN ANNULAR CMPT 122 15991.1 4.44 NO BURN IN ANNULAR CMPT 122
. 16007.5 4.45 BURN IN PROGRESS IN' ANNULAR CVPT 122 16020.2 4.45 NO BURN IN ANNULAR CMPT 122 ,
16032.8 4.45. BURN IN PROGRESS 1N ANNULAR CMPT 122 ! l 16230.2 4.51 NO BURN IN ANNULAR CMPT 122 I 16241.9 4.51 BURN IN PROGRESS IN ANNULAR CVPT 122 l I-
'1 '
4.2-7 l
7-------- , , , , , _ . . _ ,
-)
b
] S2H U2MAAP CONT.
i lg SEC HR EVENT DESCRIPTION l CODE j ul 16256.1 4.52 BURN IN PROGRESS IN UPPER CMPT 102 1 16276.2 4.52 NO BURN IN UPPER CVPT - 102
'F 16325.3 4.53 NO BURN IN ANNULAR CMPT
~
122 16329.2 4.54 BURN IN PROGRESS IN ANNULAR CMPT 122 16373.9 4.55 ICE DEPLETED 132 16387.7 4.55 BURN IN PROGRESS IN UPPER CMPT 102 16407.7 4.56 NO BURN IN UPPER CMPT 102
,I 16440.1 4.57 NO BURN IN ANNULAR CMPT 122
..a v t
'j 414 0
.b a -
t" s .: I
+
u {. ' . o j'.l a J 7 h.2-8
.a e
4.3 S: quince Ns. 3 - S 2E j 4.3.1 Accident Sequence Description i j
.i S
2 E consists of a small LOCA initiator with subsequent failure of the-d ECCS and containment spray system in the recirculation mode. Emergency core cooling and containment sprays are available during the injection phase only and the containment safeguards systems (ice condenser, air return fans, and igniters) are available throughout the accident. The following sections will present two scenarios for this accident sequence. The first sequence (4.3.2, 4.3.3) postulates that the drains between the upper and lower compartments are either closed or blocked resulting in the spray water' accumulating in the refueling pool thus preventing the normal flowback from the upper compartment to the lower compartment sump. The second sequence (4.3.4, 4.3.5) presented postulates an equipment failure preventing the accumulated water in the lower
~
d' compartment sump from being recirculated back into the upper compartment. 4.3.2 Reactor Coolant System Response (Drains Blocked) Upon iniciacion of a 0.0218 ft2 cold leg break, the reactor is scram =ed, followed by reactor pump coastdown and auxiliary feedwater startup at five seconds. Figures C.3-6 through C.3-10 illustrate the variables of
; interest. Immediately following break initiation, the primary system pressure drops 'to sa'turation pressure followed by the initiation of ECCS injection at 0.01 hours to replace the mass of primary coolant lost out of
;, the break. In this analysis, ECCS and spray system operation were terminated at the time of switchover to the recirculation mode. In actuality, a few additional minutes of operation of these systems would be l
l achieved before the RWST water is depleted. These few minutes of operatior !
, have negligible impact on the subsequent plant behavior. The ECCS supplies 4.3-1 i t
, I i
**- ** 1
, lj p ",
. , ; e ,,,
L ;( . . 1 4 water ta' the ESIbardman the ~ timehf O'.01 and 0.37 houra,. <During this timi! ' 8 4
' l l 'a 0
' period, the;'ACS pressure decreases at s' slower rate. The UHI begins to
~
q 2
. . .o p
A- inject water when the primary system preseure drops below"1255.lb/in2,, a .i . 4' Lj . ' This addition of cool water' depresses the primary system pressure to's '
~
F
}h " sinimum of. about 1000 lb/in2a se about 0.4 hours'atter'which the reactor coolant pressura and temperature increase because of decreased' heat removal .
b;g through the steam generators due'to hea' ting of the sedondary side water. g" Continued' loss of primary system inventory leads r.o core uncovery at 1.2' I 16 s hours accompanied by initiation of the, cladding metal-wacer reaction hm producing hydrogen at a significant rate around 1.5 hours. Total hydrogen production is 680 lbs at an average rate of 0.10 lbs/sec. This corresponds r lj . to an average clad oxidation of '34 percent. At approximately 2.5 hours the~i n, .. i q- primaryp system pressure decreases below 415 lb/in2a and the cold leg ~ 4 p '
," accumulators begin to dump water into the reactor vessel. The core
.e, . .
].j .
continues to heat up until aufficient molten fuel accumulates. leading to
"" I failure of the core support plate. The molten corium fails the support n
E
> plage at ,approximately 3.34 hours. At 3.35 hours, the vessel fails and the p remaining ~wate.r, hydrogen, accumulator water, and molten corium is '
- r. $
N discharged into the cavity region.
)
,.J 4.3.3 Containment Response (Drains Blocked) Immediately following the accident
~
initiation, the lower compartment pressurizes as the RCS inventory is 3" . discharged. At 65 second's the pressure set point for the containment spray , is reached. The containment spray takes suction frou the RWST until !
~
recirculation switchover is attempted unsuccessfully at 0.38 hours. At U 3.35 hours the vessel fails and the containment pressure increases to about
, LJ l i
25 lb/in2a. The forced circulation of the air return fans and remaining
- 1. j ,_,
ice' reduce the pressure to approximately 21 lb/in2a. At the time of l L' q vessel failure, the water level in the lower compartment is approximately 9-4.3-2
]
'f 4 n ..
_ - _ - _ - - _ - - _ . -e. -_. - . _ _ . .
1 1 feet,;which.is lessthan the 10 feet necessary for spillover into'the 4
- cavity. Although the containment sprays have delivered all the RWST water ;
I ti J prior to recirculation svitehover at 0.38 hours, all of this inventory is - 1 trapped in the upper compartment duetothefaiiuretoremove.upperto
- s ;,4 >
,; g . ( lower compartment drain plugt. Therefore, the molten corium is released s ,
'into a dry cavity
. , . Immediate concrete ablation acurs. due to " jet" attack.
a' $' M'Y during the corium blowdown, resulting in an initial penetration depth of ._ i ( ' about 0.3 feet.
., j Following reactor vessel failure,'the water level in the lower compartment 1
increases due to accumulation from the melted . ice but never reaches ,the. l )' , l necessary 10 foot spillover height. Therefore, once the water discharged ' during vessel blowdown (cold leg accumulators and remaining vessel inventory) is evaporate', by decay heat, the corium in the reactor cavity reheats and thermally attacks the concrete basemat generating s noncondensible gases. The mass of ice remaining at the time of vessel
, , s l
failure is approximately 5.75x105 lbm. The air return fans in
- n. 3 conjunction with the remaining ice provide containment pressure suppressic until 3.81 hours, at which time all the ice has melted. With no method of.
removing decay heat from the containment, and the continued generation of noncondensible gases from the csre-concrete attack, the containment failuz, pressure of 65 lb/in2a is reached at 25.93 hours. At this time, the containti.ent depresturizes through the assumed 0.02 f t2 containment 1 failure hole. The hole size was. chosen to preclude further containment pressurization ba' sed on a leak-before-break hypothesis (reference 6.7). 4.3.4 Reactor Coolant System Response (Drains Open) i Upon initiation of a 0.0218 ft 2 cold leg break, the reactor is
; scrammed, followed by reactor pump coastdown and auxiliary feedwater j 4.3-3 - '
Os een am .mamme w - ww. . p . es, w inemen M m +e omm e ...e mi,s-.om,7# e s .w
L l .: startup at 2ive seconds. Figures C.3-1 through C.3-5 illustrate the 1: If
, variables of interest. Immediately following break initiation, the pri:rary 3 system pressure drops to saturation pressure followed by the initiation o'f -
i I
}_
;M'-
ECCS injection at 'O.01 hours to replace the mass of ' primary coolant lost
~
f, out of the break. The ECCS system supplies water to the RCS between the q time of 0.01 and 0.38 hours. During this time period, .the RCS pressure decreases at a-slower rate. The UHI begins to inject water when the
- primary system pressure drops below 1255 lb/in2 a This addition 'of cool a w water depresses the primary system pressure to a minimum of about 1000 r-
[2 lb/in2a at about 0.4 hours after which the reactor coolant pressure and
.g temperature increases due to the heat transferred from the secondary side.
p, 4 Continued loss of primary system i,nventory leads to core uncovery at 1.2
] hours accompanied by 'initiatien of the cladding metal-water reaction
. .sk producing hydrogen.at a significant rate around 1.5 hours. Total hydrogen F.1
.; [? "
production is 700 pounds with an average rate of 0.10 lbs/sec, which
,. y g corresponds to an average clad oxidation of 35 percent. At approximately u
d 2.5 hours the primacy system pressure decreases below 415 lb/in2a and the cold leg accumulators begin to dump water into the reactor vessel. The I].:e' j , - - core continues to heat up until suf ficient molten fuel accumulates to cause h, failure of the core support plate with molten corium flowing into the lower plenum at approximately 3.30 hours. Vessel failure occurs about one minute
' fg.
'-- later and the remaining water, hydrogen, accumulator water, and molten
[ corium is discharged into the reactor cavity region, u 4.3.5 Cone'ainment Response (Drains Open) j- Immediately following the accident initiation, the lower compartment I; I.J pressurizes as the RCS inventory is discharged. At 65 seconds the q pressure set point for the containment spray is reached. The containment
, tj 4.3-4 t
l- _ - - - - - - , - - . - . . ~ _ - - - - - - - - - - - - - - - - - -
spray takes suction from the RWST until recircualtion switchover is i. attempted unsuccessfully at 0.37 hours. At 3.31' hours the vessel fails l, causing a containment pressure increase to 28 lb/in2 a The forced h circulation of the air return fans and the remaining ice reduce the .
) .
pressure to approximately 18 lb/in2 a The water level in the lower , j compartment' has equaled the height required for spillover into the cavity at 0.8 hours. There fore , the molten corium is release into a flooded 1 cavity. Immediate concrete ablation occurs due to " jet" attack during j the vorium blowdown, resulting in an initial penetration depth of 0.15 feet. However, the debris is immediately quenched, halting.any more j
.1 concrete attack and ' the containment pressure remains low until the ice j '- melts at 4.36 hours. Subsequently, with no method for remoiring decay heat, the containment pressurizesdue' to steam formation and fails at .o .
.l 9.54 hours. At this time, the' containment depressurizes through the assumed
- 0.10 ft2 failure hole. The hole size was chosen to preclude I . f' .
further containment pressurization based on a leak-before-break hypothesis (reference 6.7). t I p
- \
4.3-5 - r 2.
~ _ _ . _ . _ _ . _ . . -
^j (
.___________a
l J l,E S2HF U7MAAP (DRAINS BLOCKED) l1 I;$ SEC HR EVENT DESCRIPTION CODE
. 0.0 0.00 REACTOR SCRAM 13 ,
l m 0.0 0.00 LETDOWN FLOW OFF _ 46 [ 0.0 0.00 AUX FEEDWATER ON 154 g 0.0 0.00 MSIV CLOSED 156 -
,- 0.0 0.00 PS BREAK FAILED 209 0.0 0.00 k%NUAL SCRAM 227 l2 0.0 0.00 MAKEUP SWITCH OFF 242 t 0.0 0.00- LETDOWN SWITCH OFF 243 47.6 .01 CHARGING PUMPS ON 11 l E 61.1 .02 NAIN COOLANT PUMPS OFF 4 61.1 .02 MCP SWITCH OFF OR Hi-VIBR TRIP 215
. :n 64.9 .02 CONDWT SPRAYS ON 103
.m 160.3 .04 HPI ON 5 1356.8 .38 HPI OFF "'
5 i 1356.8 .38 CMARGING PUMPS OFF 11 1356.8 .38 CONDWT SPRAYS OFF 103 r, 1356.8 .38 HPl FORCED OFF 216 U 1356.8 .38 LPI FORCED OFF 217
<- 1356.8 .38 SPRAYS FORCED OFF 222 P 1356.8 .38 CHARGING PUMPS FORCED OFF 232 4435.5 1.23 FP RELEASE ENASLED 14 -
)
5916.7 1.64 8 URN IN PROGRESS IN 1/C UPPER PLENLM 141 if 6312.1 1.75 8 URN IN PROGRESS IN LOWER CMPT 75 6430.8 1.79 BURN IN PROGRESS IN UPPER CuPT 102 {- 6491.2 1.80 BURN IN PROGRESS IN ANNULAR CuPT 122 6592.4 1.83 NO BURN IN LOWER CuPT 75 8442.7 2.35 UHI ACCUM OuP/Y 190 BURN IN PROGRESS IN LCWER CuPT 75
,, 9250.4 2.57 9858.8 2.74 TK) BURN IN LOWER CMPT 75
! 11413.3 3.17 8 URN IN PROGRESS IN LOWER CuPT 75 11697.7 3.25 NO SURN IN LOWER CuPT 75 r, 11805.4 3.28 BURN IN PROGRESS IN LOWER CuPT 75 g 11970.6 3.33 NO SURN IN LOWER CuPT 75 12014.3 3.34 SUPPORT PLATE FAILED 2 12077.0 3.35 RV FAILED 3 I] 75 i" 12129.4 3.37 BURN IN PROGRESS IN LCWER CuPT l O' l
17 k 3-6 u
~
_J
; S2HF U7MAAP(DRAINS BLO.CKED) CONT. ;'
i
! SEC HR EVENT DESCRIPTlON CODE .
I 12169.8 3.38 ACCUMULATOR WATER DEPLETED 188
- 12261.4 3.41 NO BURN .lN 1/C UPPER PLENLM 141 'I 12272.7 3.41 BURN IN PROGRESS IN 1/C UPPER PLENLM- 141 5 12366.3 3.44 NO BURN IN LOWER CMPT 75
, 13730.4 3.81 ICE DEPLETED 132 i 13739.1 3.82 NO BURN 1N I/C UPPER PLENLM 141 13762.0 3.82 NO BURN IN UPPER CMPT 102 13790.1 3.83 NO BURN IN ANNULAR CVPT 122 31512.0 8.75 BURN IN PROGRESS IN LOWER CMPT 75 31594.5 8.78 BURN IN PROGRESS IN 1/C UPPER PLENLM 141
; 31996.0 8.89 BURN IN PROGRESS IN UPPER CMPT 102 32079.5 8.91 BURN IN PROGRESS IN ANNULAR CMPT 122 j 34084.8 9.47 NO BURN IN LOWER CMPT 75 )
34104.8 9.47 BURN IN PROGRESS IN LOWER CMPT 75 ,
, 34124.8 9.48 NO BURN IN LOWER CMPT 75 34164.8 9.49 BURN IN PROGRESS IN LOWER CMPT 75 i
~
34164.8 9.49 NO BURN IN 1/C UPPER PLENLM 141 34184.8 9.50 NO SURN IN LOWER CMPT 75 34184.8 9.50 BURN IN PROGRESS IN I/C UPPER PLENLM 141 j
' 34204.8 9.50 NO SURN IN I/C UPPER PLENLM 141 -
34244.8 9.51 NO BURN IN UPPER CMPT 102 i 34244.8 9.51 NO BURN IN ANNULAR CMPT 122 93364.4 25.93 CONTMT FAILED 104 4
. e i
h.3-7 -
] TA3LI: M3-1 b S2HF U3MAAP (DRAINS OPEN)
IC 'l :j SEC HR EVENT DESCRIPTION CODE l 1 0.0 0.00 REACTOR SCRAM 13 d@ 0.0 0.00 LETDOWN FLOW OFF
~
46 "" 0.0 0.00 AUX FEEDWATER ON 154
% 0.0 0.00 MSIV CLOSED 156 i.I . 0.0 0.00 PS BREAK FAILED 209 0.0 0.00 MANUAL SCRAM 227 242
~
O.O 0.00 MAKEUP SWITCH OFF
" 0.0 0.00 LETDOWN SWITCH OFF 243 , r- 47.6 .01 CHARGING PUMPS ON 11
(- 61.1 .02 MAIN COOLANT PUMPS OFF 4 61.1 .02 MCP SWITCH OFF OR Hi-VIBR TRIP 215 I. 64.9 .02 CONTMT SPRAYS ON 103 4 160.3 .04 HPI ON , 5
; r. 1342.3 .37 HPl OFF 5 d 1342.3 .37 CHARGING PUMPS OFF 11, ,' 1342.3 .37 CONTMT SPRAYS OFF 103 l
[l 1f 1342.3 1342.3 1342.3
.37
.37
.37 HPI FORCED OFF LPI FORCED OFF SPRAYS FORCED OFF 216 217 222
[, 1342.3 .37 CHARGING PLMPS FORCED OFF 232 4441.5 1.23 FP RELEASE ENABLED 14 P 5798.9 1.61 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 C 6281.4 1.74 BURN IN PROGRESS IN LOWER CMPT 75 n 6420.4 1.78 BURN IN PROGRESS IN UPPER CMPT 102 [g, , 6486.4 1.80 BURN IN PROGRESS IN ANNULAR CMPT 122 1.83 75 6601.9 NO BURN IN LOWER CMPT iP! 8316.7 2.31 UH1 ACCLM EMPTY 190 2 9346.3 2.60 BURN IN PROGRESS IN LOWER CMPT 75
,.., 9819.3 2.73 NO BURN IN LOWER CMPT 75 75 ,.] 9925.4 2.76 8 URN IN PROGRESS IN LOWER CMPT 75 10112.0 2.81 NO BURN IN LOWER CMPT 'Pi 11764.7 3.27 BURN IN PROGRESS IN LOWER CMPT 75 ,U 11868.8 3.30 SUPPORT PLATE FAILED 2 11923.9 3.31 NO BURN IN LOWER CMPT 75
{ 7.; 11932.7 3.31 RV FAILED 3
- J 11942.1 3.32 DURN IN PROGRESS IN LOWER CMPT 75
- l1 d
7 u.3 8
- -_.. -__e.- - - _ . _ _ _ _ _ . _ _ - - . . . _ - _ - _ _
uBII k.3-1 1 S2HF U3MAAP (DRAlNS OPEN) CONT. 1 - i SEC HR EVENT DESCRIPTION , CODE j 11951.6 3.32 NO BURN IN LOWER CMPT 75
" ^
11965.7 3.32 BURN IN PROGRESS IN LOWER CMPT 75 11989.9 3.33 NO-BURN IN 1/C' UPPER PLENLM 141 11991.9 3.33 BURN IN PROGRESS IN 1/C UPPER PLENUM 141" 12028.0 3.34 ACCUMULATOR WATER DEPLETED 188 12073.1 3.35 NO SURN IN LOWER O/PT 75 12077.9 3.35 NO BURN IN UPPER CMPT
~
102
! 12123.2 3.37 NO SURN IN ANNULAR CMPT 122 l 12148.2 3.37 NO BUWN IN I/C UPPER PLENUM 141 12158.8 3.38 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 . 12230.0 3.40 BURN IN PROGRESS IN UPPER CMPT 102 ; 12286.5 3.41 BURN IN PROGRESS IN ANNULAR CMPT 122 . 13694.1 3.80 NO SURN IN ANNULAR O/PT 122 1
13761.5 3.82 NO BURN IN PPER CMPT 102 13801.5 3.83 BURN IN PROGRESS IN UPPER CMPT 102 13821.0 3.84 NO BURN IN UPPER O/PT 102 l 13849.8 3.85 BURN IN PROGRESS IN UPPER CMPT 102 13864.8 3.85 NO BURN IN UPPER CMPT 102-13903.9 3.86 BURN IN PROGRESS IN UPPER CMPT 102 13923.9 3.87 NO BURN IN UPPER O/PT 102 14006.0 3.89 BURN IN PROGRESS IN UPPER CMPT 102 14022.1 3.90 NO BURN IN UPPER OAPT 102 l 14132.4 3.93 NO BURN IN 1/C UPPER PLENLM 141 14183.9 3.94 BURN IN PROGRESS IN UPPER CAPT 102 i 14206.6 3.95 NO BURN IN UPPER CMPT 102 15707.8 4.36 ICE DEPLETED 132
- 34346.5 9.54 CONTMT FA1 LED 104 i
1 4 9 l- l
-l u.3-9 .!
.ma J
4.4 S qt onca No. 4 - TMLB' 4.4.1 Accident Sequence Description
, TML3' consists of a transient sequence initiated by loss of off-site i- ,8 AC power with subsequent loss of on-site AC power. Due to lack of
~
- 9. cooling, the reactor coolant pump seals fail resulting in a small LOCA s,
c. (50 gpm/ pump). In this sequence, several potential sequences are ,
~
i} lumped together. These include immediate failure of main' and 73 auxiliary feedwater as well as sequences involving no interruption of i' a main feedwater but subsequent failure of the power conversion system and failure of the auxiliary feedwater. For the base case analysis, L]j . both main and auxiliary feedwater are both as'sumed lost at the time of
;a the initiating event. Emergency core cooling, containment sprays, air j n, return f ans, and hydrogen ignitersare not available due to loss of i <J .
.; B ,
all AC power. ,
!F] 4.4.2 Reactor Coolant System Resuonse LJ ,, This sequence is initiated by loss of off-site AC power with ,
i I subsequent loss of on-site AC power, reactor trip, reactor pump J coastdown, and loss of both main and auxiliary feedwater. Figures
, 's i.$ C.4-1 through C.4-5 illustrate the variable s of interest. Due to lack of in.jection and cooling, the reactor coolant pump seals fail at i lqi 0.75 hours resulting in a total 200 gal / min leck. The RCS water mass hi continues to decrease as RCS inventory is depleted through the pump u
seals. The primary system maintains a relatively constant pressure of
, 2000 lb/in2a as the steam generators provides a heat sink. However,
'q .,
the steam generators are losing mass through the secondary side relief 1 J.;
, valves with no make-up from feedwater.
;m j !-
'a
' "1 4.4-1 j
Ib _ ________________________________m.____.m~
- t. The primary sys tem pressure starts to rapidly increase between 1.4 and 1
l 1.7 hours due to the loss of the secondary side steam generator heat i ! sink. The pressure continues to increase to the set point of the I pressurizer relief valves. Continued blowdown to the quench tank results in failure of the tank rupture disk at 1.66 hours. S team generator dryout also occurs at 1.66 hours. During this time of high pressure RCS blowdown, the water level in the reactor vessel rapidly decreases with core uncovery around 2.10 hours and initiation of hydrogen production occurring at approximately 2.2 hours. The total hydrogen production is 590 lbs. at an average rate of 0.14 lbs/sec. This corresponds to an overall oxidation of 30 percent. The primary system continues to remain at high pressure and suf ficient molten corium is accumulated to fail the core support place at approximately 3.33 hours as evidenced in the vessel. pressure spike and slight level swell in the vessel. About one minute
,la ter , the vessel fails and the remaining water, hydrogen, and corium are discharged from the vessel into the cavity at high pressure. Due to the elevated RCS pressure, no water is injected by either UHI or cold leg accumulators until the time of vessel failure.
4.4.3 Containment Resconse The containment pressure increases to 17 lb/in2a following failure of the pump seals and then increases further to approximately 21 lb/in2a following quench tank rupture disk failure. At 3.35 hours the vessel fails, increasing the containment pressure to approximately 30 lb/in2 a At the time of vessel failure the water level in the lower compartment is approximately 2.8 feet which is less than the 10 feet necessary for
' spillover into the cavity. Therefore, the molten corium is released into j 4.4-2 1
** a
l l a dry cavity. Imm:dicte conersta cbiction occurs dua to " jet" attack jr iV during the corium blowdown, resulting in an initial penetration depth of f. ((]
; he about 0.20 ' feet.
i i
..] Following reactor vessel failure, the water level in the lower tj compartment never reaches the necessary 10 foot spi 11over height. ,
5 Therefore, once the water discharged during vessel bloudown (cold leg
- accumulators and UHI) is evaporated by decay heat, the corium in the m s reactor cavity reheats and decomposes the concrete, thus generating m-noncondensible gases. The mass of ice remaining at time of vessel hj f ailure is approximately 1.25x106 lb s . , bu t this has melted by 5.84
, h. l,
'i hours. With no method of removing decay heat from . the containment,' and the continued generation of noncondensible gases from the corium-concrete
#]
attack, the containment failure pressure of 65 lb/in2a is reached at
' approximately 27.1 hours. At this time, the containment depressurizes
. The hole size was
~,, through the assumed 0.02 f t2 containment failure hole.
-- chosen to preclude further containment pressurization based on a leak-fl before-break hypothesis (reference 6.7).
i t.-
.~-
..a e:
fk s, t
.s e-
- L{i i
! 4.5-3 [] i O ic u
TABLE b.k-1 , i TMLB' U4MAAP : SEC HR EVENT DESCRIPTION CODE O.O O.00 MAIN COOLANT PLMPS OFF - 4 '
~
O.0 0.00 REACTOR SCRAM 13 O.0 0.00 LETDOWN FLOW OFF 46 O.0 0.00 MSIV CLOSED 156 . 0.0 0.00 POWER NOT AVAILABLE 205 0.0 0.00 MAKEUP SWITCH OFF 242
~
0.0 0.00 LETDOWN SWITCH OFF 243 l 2717.1 .75 PS BREAK FAILED 209 5970.3 1,66 Q/T RUPTURE DISK FAILED 92 5991.3 1.66 UN8KN S/G DRY 161 5992.3 1.66 BROKEN S/G DRY 151 6789.9 1.89 MCP SWITCH OFF OR Hi-VIBR TRIP 215 l 7546.8 2.10 FP RELEASE , ENABLED 14 11984.2 3.33 SUPPORT PLATE FAILED 2 12046.4 3.35 RV FAILED 3 12064.2 3.35 8 URN IN PROGRESS IN 1/C UPPER PLENUM 141 12071.9 3.35 NO BURN IN 1/C UPPER PLENLM 141 12980.5 3.36 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 12177.0 3.38 ACCUMULATOR WATER DEPLETED 188 12190.8 3.39 NO BURN IN 1/C UPPER PLENLM '141 12192.2 3.39 BURN IN PROGRESS IN I/C UPPER PLENLM 141 12208.1 3.39 NO BURN IN 1/C UPPER PLENLM 141 12215.9 3.39 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 12218.0 3.39 NO SURN IN 1/C UPPER PLENLM 141 12219.6 3.39 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 12235.9 3.40 NO SURN IN 1/C UPPER PLENud 141 12237.O 3.40 BURN IN PROGRESS IN I/C UPPER PLENLM 141 12252.7 3.40 NO BURN IN I/C UPPER PLENLM 141 12259.5 3.41 BURN IN PROGRESS IN 1/C UPPER PLENLM 1A1 12262.9 3.41 NO SURN IN l/C UPPER PLENUM 141 12269.4 3.41 BURN IN PROGRESS IN I/C UPPER'PLENud .141 12271.5 3.41 NO SURN IN 1/C UFPER PLENUM 141 12280.2 3.41 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 12281.4 3.41 NO SURN IN 1/C UPPER PLENLM 141
; 12284.7 3.41 UHI ACCLM EMPTY 190 i 12286.5 3.41 BURN IN PROGRESS IN I/C UPPER PLENLM 141
~
1, i I h .h h _
a g TML3U4MAAP CONT.. [ SEC l HR EVENT DESCRIPTION l CODE 4g 12293.9 3.41 NO SURN IN I/C UPPER PLENUA 141 11 $ 12300.7 3.42 8. URN'IN PROGRESS IN 1/C UPPER PLENLM 141 12302.8 3.42 NO BURN IN I/C UPPER PLENUM 141 j{
. 'y' :y 12310.6 3.42 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 12312.3 3.42 NO BURN IN I/C UPPER PLENLM 14 L
}'. 12319.5 3.42 BURN IN PROGRESS IN I/C UPPER PLENUA 141
>- 12321.0 '3.42 NO BURN IN I/C UPPER PLENUA 141
.i 12329.3 3'.42: BURN IN PROGRESS IN l/C UPPER PLENLM 141 12330.6 3 43 NO SURN IN 1/C UPPER PLENUA 141 1 12341.2 3.43 BURN IN. PROGRESS IN 1/C UPPER PLENUA 141 12342.6 3.43 NO BURN IN 1/C UPPER PLENUM 141
N .J.3 .
12352.9 3.43 BURN IN PROGRESS IN 1/C UPPER PLENud 141 I ,.c 12354.4 3.43 NO SURN IN 1/C UPPER PLENLM 141
.' y 12361.1 3.43 EURN IN' PROGRESS IN.1/C UPPER PLENLM 141 3
12363.0 3.43 NO BURN IN l/C UPPER PLENud 141 3 9 12370.7 3.44 BURN IN PROGRESS IN 1/C UPPER PLENUA 141
, 12372.5 3.44 NO BURN IN I/C UPPER PLENLM 14'1
- 12379.0 3.44 BURN IN PROGRESS IN 1/C UPPER PLENLM 141-12381.1 3.44 NO SURN IN 1/C UPPER PLENLM 141 lel 12388.2 3.44 BURN IN PROGRESS IN 1/C UPPER PLENLM 141
*[ 12390.4 3.44 NO SURN IN I/C UPPER PLENLM 141 ., 12399.1 3.44 BURN IN PROGRESS lN 1/C UPPER PLENLM 141 12401.8 3.44 NO BURN IN 1/C UPPER PLENLM 141 3
L tj 12410.0 3.45 BURN IN PROGRESS IN I/C UPPER PLENLM 141 l' 12413.6 3.45 NO BURN IN 1/C UPPER PLENUM 141
~'"'
' ij' 12420.2 3.45 8 URN IN PROGRESS IN 1/C UPPER PLENUd 141 12424.8 3.45 NO BURN IN I/C UPPER PLENLM 141 1 -
12433.0 3.45 -SURN IN PROGRESS IN I/C UPPER PLENLM 141 1245.6 3.46 NO BURN IN 1/C UPPER PLENUA 141 12453.2 3.46 5 URN IN PROGRESS 1N I/C UPPER PLENLM 14i I;- 12469.4 3.46 NO BURN IN I/C UPPER PLENLM 141 12473.6 3.46 BURN IN PROGRESS IN 1/C UPPER PLENLM 141
'q 12489.2 3.47 NO BURN IN 1/C UPPER PLENLM 141 b 12492.0 3.47 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 l ,,, 12507.5 3.47 NO BURN IN I/C UPPER PLENUM 141
{,lj 12511.9 3.48 BURN IN PROGRESS IN I/C UPPER PLENLM 141
! '}
i g y B .
! TMLB' U4MAAP CONT.5 t
1 SEC HR EVENT DESCRIPTION CODE 12527.3 3.48 NO BURN IN 1/C UPPER PLENUM 141 .i 12530.3 3.48 BURN IN PROGRESS IN I/C UPPER PLENLM 141 - 12545.9 3.48 NO BURN IN 1/C UPPER PLENUM 141 12549.7 3.49 BURN IN PROGRESS IN 1/C UPPER PLENLM 141 12565.6 3.49 NO BURN IN 1/C UPPER PLENUM 141 12571.4 3.49 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 13258.9 3.68 NO BURN IN 1/C UPPER PLENLM 141 21018.1 5.84 ICE DEPLETED 132
, 38157.3 10.60 BURN IN PROGRESS IN LOWER CMPT 75 38183.7 10.61 NO BURN IN LOWER CMPT 75 38528.8 10.70 BURN IN PROGRESS IN LOWER CuPT 75 38555.2 10.71 NO BURN IN LOWER CMPT 75
- 39111.6 10.86 BURN IN PROGRESS IN LOWER OMPT 75 39137.5 10.87 NO BURN IN TOWER CMPT 75 41372.8 11.49 BURN IN PROGRESS IN LOWER CuPT 75 41378.9 11.49 BURN IN PROGRESS IN UPPER OuPT 102 41378.9 11.49 BURN IN PROGRESS IN 1/C UPPER PLENLM 141
. 41394.9 11.50 NO BURN IN 1/C UPPER PLENLM 141 e -
41398.7 11.50 NO BURN IN LOWER CMPT 75 41400.1 11.50 NO BURN IN UPPER CMPT 102 42362.9 '1.77 BURN IN PROGRESS IN LOWER OuPT 75 42389.4 11.77 NO SURN IN LOWER CMPT 75 97695.6 27.14 CONTMT FAILED 104 l I i k.L-6
?%
t J3 4.5 Snqutnca No. 5 - T23NL 7 q
>. [ 4.5.1 Accident Secuence Description
'l
) 77 T23ML consists of a transient initiator other than loss of off-site
' ' , (M ~
i power with automatic reactor trip and loss of main and auxiliary
.2 F} feedwater. AC power is available and, therefore, emergency core cooling
'. h i and containment safeguards are available throughout the accident. ,
i p
, .g Although suf ficient time exists for operator action, the base case g.. assumes human or equipment failures prevent proper charging and safety I
!d system operation. The assumption of these additional failures makes . this f
I a very low probability event. Higher probability sequences are discussed I {}) r. in section 5.0. j
, in
, E!
u.
., 4.5.2 Reactor Coolant System Response f
This sequence is initiated by loss of both main and auxiliary feedwater, followed by reactor trip and reactor pump coastdown.
. 'J] ,, Pigures C.5-1 through C.5-5 illustrate the variables of interest.
h] Following loss of all feedwater and reactor scram, the primary system f
, pressure decreases momentarily followed by the actuation of the
'h pressurizer heaters which maintain the pressure at approximately 2250
)
3 lb/in2 The water level in the pressurizer increases during heat up
' {i' .
and volumetric expansion causing the pressurizer to go solid around
.r-
[,l 1.0 hour af ter accident initiation. t [:
,t: The primary system pressure starts to increase af ter 0.95 hours due to
{' the loss of the secondary side steam generator heat sink. The pressure
'\d' continues to rise to the set point of the pressurizer safety valves.
i
; 1d However, blowdown through these valves decreases primary system inventory 1 1D i .. and with no makeup available, both the primary system pressure and level F' 4.5-1
\l
,.s
begin to de. crease. Therefore, the primary system pressure stabilizes at
) the PORV set point of 2350 lb/in2a with continued inventory deple tion I ] and core uncovery occurring at 1.7 hours. As the water level in the core 11
]; continues to drop, the cladding temperature begins to increase. .At approximately 1.9 hours, the metal-water reaction initiates significant hydrogeti generation and further core melting. Total hydrogen production from in-vessel Zircaloy oxidation is 520 lbs. The average production 1 rate is 0.14 lbm/see and the reaction is equivalent to a total core I average clad oxidation of 26 percent. 'The primary system continues to remain at high pressure and sufficient molten corium is accumulated to fail the core support plate at 2.90 hours. At 2.91 hours the vessel fails and the remaining water, hydrogen, and corium are discharged from the vessel into the cavity at high pressure. 4.5.3 Containment Response . r-The containment pressure remains at about 15 lb/in2a until quench tank rupture disk failure at 1.20 hours. The containment pressure rapidly increases to 19.5 lb/in2a but is suppresssed as the containment sprays (actuated at 1.21 hours), air return fans, and ice are available. The
, containment sprays take suction from the RWST until successful recirculation realignment occurs at 1.60 hours. This pressure suppression reduces the pressure to about 17.0 lb/in2a until vessel , failure occurs at 2.91 hours with a corresponding pressure increase to 23 lb/in2a which is quickly suppressed. As the ice continues to melt and RCS inventor *.s lost from the pressurizer relief valves, the water level ) in the lower compartment exceed = the necessary curb beight required for i
i spilling water into the cavity at approximately 1.2 ho'na. Therefore, when the vessel fails the cavity is flooded. This flooded condition 4.5-2 4
- -- - ]-
l
. !.i ' )
licits . cora-conerste chlation to ths " jet" attack resulting in a 0.14 j f.'m foot penetration depth. The flooded cavity results in immediate 7 quenching of the corium.' I
. j 'A 'j i
- P',1 j The remaining ice at time of vessel -failure .is approximately .1.3x106 Iba. At 5.36 hours, all of the ice has melted and containment .,
e im
- i. pressurization begins.. Following ice deple tion, the containment pressure rapidly rises to about 20 lb/in2 a However, the containment sprays continue to ' emove r heat from the containment atmosphere. This' heat
[ . removal . rate matches the heat decay at approximately 7.5 hours. Therefore, the containment spray heat removal rate is more than adequate
. (A to remove decay heat and the containment pressure continues ' to decrease,
~
,eg thus precluding containment failur .
iq
.e ,
r:3
!i.i em
,gl LJ
, l9 -
ii y i
?- i
. rm
),
J
;i a
id - l 4 f]w]- a i I[7 4.5-3 l A 1 % - - __-_x-______ - . - _ . _ ._.-_
nat. a.:-i i -
)
1 T23ML USMAAP ' i SEC l HR l- EVENT DESCR I PTI ON l CODE]
) 0.0 0.00 REACTOR SCRAM 13 0.0 0.00 LETDOWN FLOW OFF 46 0.0 0.00 MSIV CLOSED 156 0.0 0.00 HPl FORCED OFF 216 0.0 0.00 LPI FORCED OFF 217.
! 0.0 0.00 AUX FEED WATER FORCED OFF 224
. 0.0 0.00 MANUAL SCRAM 227 0.0 0.00 MAIN F# ShVT OFF 228 0.0 0.00 CHARGING PLMPS FORCED OFF 232 0.0 0.00 MAKEUP SWITCH OFF -
242 0.0 0.00 LETDOWN SWITCH OFF 243 l 3404.4. .95 BROKEN S/G DRY 151 3404.4 .95 UN8KN S/G DRY 161 l 4310.3 1.20 Q/T RUPTURE DISK FAILED 92 4348.8 1.21 MAIN COOLA W PUAPS OFF 4 4348.8 1.21 MCP SWITCH OFF OR HI-VIBR TRIP 215 4350.9 1.21 CONTMT SPRAYS ON 103 5753.7 1.60 RECIRC SYSTEM IN OPERATION 181 5753.7 1.60 RECIRC SWITCH: MAN ON 220 y 5765.8 1.60 CH PLMPS INSUFF NPSH 183 5765.8 1.60 HPi PUAPS INSUFF NPSH 185 6163.6 1.71 FP RELEASE ENABLED 14 10429.2 2.90, SUPPORT PLATE FAILED 2 10457.2 2.90 BURN IN PROGRESS IN LOWER CMPT 75 10489.6 2.91 RV FAILED 3
, 10490.5 2.91 SURN IN PROGRESS IN I/C UPPER PLENUd 141 10493.0 2.91 NO SURN IN LOWER CMPT 75 10538.5 2.93 BURN IN PROGRESS IN LOWER CMPT 75 10615.0 2.95 ACCUMULATOR WATER DEPLETED 188 10636.8 2.95 NO BURN IN LOWER CMPT 75 10702.3 2.97 NO BURN IN 1/C UPPER PLENud 141 10709.0 2.97 BURN IN PROGRESS IN 1/C UPPER PLENud 141 10727.5 2.98 UHI ACCud EMPTY 190 , 10922.4 3.03 NO BURN IN I/C UPPER PLENLM 141 , 19284.8 5.36 ICE DEPLETED 132 l
l 1 l 4.5-4 ______ _ . _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - " - - - - - - ' - - - - - ~ - - - - - - - - - - - - ' - - - - - ~ ' - - - ' ' ~
C 4.6 Sequence No. 6 - AD g w 4.6.1 Accident Seouence Description i en l- W(f! gj AD consists of a large LOCA (10" diameter) initiator with subsequent
, failure of the ECCS in the injection mode. The ECCS' continues to be
+d
'J
.. unavailable in the recirculation mode. Containment safeguards systems
[, are available throughout the accident. - r
- j. , 4.6.2 Reactor Coolant System Response Upon initiation of a 0.5454 f t2 cold leg break, the reactor is
,9,
'., scrammed, followed by reactor pump coastdown. and auxiliary feedwater go startup at five seconds. Figures C.6-1 through C.6-5 illustrate the i i
>"~'
; variables of interest. Immediately following break initiation, the ri
}1 primary system pressure rapidly decreases to containment pressure.
9 - l The decrease in reactor vessel water level results in core uncovery . (.] tj about 0.5 hours and initiation of hydrogen production at 0.7 hours. r , pq . The core continues to heat up until fuel melting occurs leading to ( failure of the core support plate at 1.50 hours as evidenced in the n . vessel pressure increase and level swell in the vessel. The molten i.J
'. corium falls into the lower plenum and fails the reactor vessel at T?j.
b approximately 1.52 hours and the remaining water, hydrogen, and molten B corium is discharged into the cavity region. L; f* i." Total hydrogen production from in-vessel Zirealoy oxidation is 840 Ibs. at an average rate of 0.25 lbs/sec. This corresponds to an I'l (j average clad oxidation of 42 percent. The hydrogen production during
; bo ildown for th AD sequence (prior to two phase vessel level swell) is iil L>
slightly less (approximately 40 lbm) than for the S D2 sequence.
- 'l (This is expected since the boildown is more rapid for the AD 4.6-1 I'L3 e
n a
sequence.) The total hydrogen production for the AD segunence' exceeds that of all. the base cases due to the large amount of hydrogen 9 . produced during the two phase vessel level swell (nearly 325 lbm). l The AD case produces more hydrogen during the two-phase vessel level .
,{ swell than the S 2D case because the intact core nodes are at a higher average temperature. This is due .to the greater decay heat
- ' generation at 1.5 hours compared to 2.75 hours.
l Hydrogen production can occur in re-covered, unmolten nodes in MAAP due to the severely restricted , heat trans fer which ~ is allowed ' to take ! place between the cladding and the water. The model used during ' two-phase vessel . level swell for ref1 pod) assumes an intact geometry; in reality, the unmelted core nodes would be largely disrupted .by the time of slump. This assumption of intact geometry was made to conservatively represent the maximum hydrogen production that could ! 3 occur as a core is reflooded, and is judged to greatly overpredict the rate of hydrogen generation (reference 6.7). 4.6.3 Containment Response i Immediately following break initiation, the lower compartment rapidly l pressurizes as the RCS inventory is discharged. This immediate pressure increase leads to actuation of containment sprays. The ; containment spray takes suction from the RWST until 0.39 hours at l which time successful spray recirculation switchover is achieved. At 1.52 hours, the reactor vessel fails and the containment pressure i [i. incr2ases to about 22 lb/in2 The air return fans, containment sprays, and remaining ice reduce the containment pressure. The water level in the lower compartment reaches the spillover curb height at
- l. 4.6-2 I -
i-
Tharsford, at tha tims vassal f ailure
~, :
cpproximatoly 0.25 hours.
,))
.cccurs the cavity is floodsd.- This floodsd condition limits cors-
>9 concrete ablacion to the " jet" attack only resulting in about 0.13 f t
}
,. .F 4
i ,, penetration depth. The flooded cavity results in the immediate
..,) . ]$
j3 quenching of the corium. The ice remaining at time of vessel failure is approximately 9.5x105
,) lbs (about 60 percent mel'ted) . At 3.15 hours, all the ice has melted and containment pressurization begins. Following ice deple tion, the
'r I
'. containment sprays continue to remove heat from the containment a- l
[7 atmosphere. This heat removal rate matches the decay heat at 9 .. approximately 6 hours when the containment pressure reaches 21 f, lI lb/in2 a Afterward, the containment spray heat removal race exceeds that of decay heat and the containdsnt pressure decreases, thus.
'.i precluding containment failure. .
,a..
f o. y
* .-g hjI
.; l i
.9 (A
**i t .;
(.1
._)
,f.-
l 'a ? . J
! -r 1 ;y L.;
4.6-3 r] j LJ m i'
TABLE 4.b-L -- AD U6MAAP "
?
I SEC HR EVENT DESCRIPTION l CODE - O.0 0.00 REACTOR SCRAM 13 ] 0.0 0.00 LETDOWN FLOW OFF 46 ( l- .0. 0 0.00 -AUX FEED. WATER ON 154. l~ 0.0 0.00 MSIV CLOSED 156 - I
.O.0 0.00 PS BREAX FAILED 209 . - 0.0, 0.00 HPl FORCED OFF 216 -
0.0 0.00 LPI FORCED OFF 217 e 0 '. 0 0.00 MANUAL SCRAM 227 j 0.0 0.'00 CHARGING PuuPS FORCED OFF 232 0.0 0.00- l MAXEUP SWITCH OFF . 242 l 0.0 0.00 LETDOWN SWITCH OFF 243 I 2.1 .00 CONTMT SPRAYS ON 103 f 60.3 .02 MAIN COOLANT PLMPS OFF 4 { 60.3 .02 MCP SWITCH OFF OR HI-VIBR TRIP 215' l 335.5 .09 UHI ACCUV, EMPTY 190 791.9 .22 ACCUMULATOR WATER DEPLETED 188 1407.2 .39 RECIRC SYSTEM IN OPERATION 181. 1407.2 .39 RECIRC SWITCH: MAN ON 220 1411.0 .39 CH PUMPS INSUFF NPSH 183
~
1411.0 .39 HPI PUMPS INSUFF NPSH 185 1745.9 .48 FP RELEASE ENASLED 14 - 2664.9 .74 BURN IN PROGRESS IN LOWER CMPT 75 2745.0 .76 BURN IN PROGRESS IN 1/C UPPER PLENUM 141 3531.7 .96 BURN IN PROGRESS IN UPPER CMPT 102 3581.9 .99 BURN IN PROGRESS IN ANNULAR CMPT 122. 4796.4 1.33 NO SURN IN LOWER CuPT 75 - 4916.4 1.37 SURN IN PROGRESS IN LOWER CMPT 75 5315.6 1.48 NO BURN IN LOWER CMPT 75 5415.5 1.50 SUPPORT PLATE FAILED 2 5424.1 1.51 SURN IN PROGRESS IN LOWER CMPT 75
'5473.1 1.52 RV FAILED 3 5479.4 1.52 NO SURN IN LOWER OuPT 75 5484.6 1.52 SURN IN PROGRESS IN LOWER OuPT 75 5503.5 1.53 NO BURN IN LOWER CuPT 75 6782.9 1.88 NO BURN IN UPPER CuPT 102
~
, 6828.1 1.90 SURN IN PROGRESS IN UPPER CMPT 102
- e I
4.6-4 -, t _ - _ _ ____-_ l_
yY, CONT #
.AJ U6MAAP .
rm [ SEC , HR l EVENT. DESCRIPTION l CODE 6909.5 1.92 NO SURN IN UPPER OAPT 102 6925.3 1.92- BURN IN PROGRESS IN UPPER CMPT 102
. , q%
6945.8 1.93 NO BURN IN UPPER CMPT 102 ! 17, 6972.0 .1.94 BURN IN PROGRESS IN UPPER OAPT 102
~
I
- (j 7001.6 1.94 NO-BURN IN UPPER CMPT 102 4 7089.7 1.97. BURN IN PROGRESS IN UPPER CMPT 102
,% .7132.4 1.98 NO BURN IN UPPER CMcT 102 .
2 7228.9 2.01 BURN IN PROGRESS IN UPPER OAPT 102 7274.1 2.02 NO~ BURN IN UPPER CMPT 102 l 3l;7 7517.9 2.O9s BURN IN PROGRESS IN UPPER CMPT 102 7554.5 2.10 NO SURN IN UPPER CMPT 102 C 7684.3 2.13 BURN IN PROGRESS IN UPPER CMPT 102
- 'i .7702.6 2.14 NO BURN 1N UPPER CMPT 102 '
- r. 7800.6 2.17 BURN IN PROGRESS IN UPPER CMPT 102 NO BURN IN UPPER CMPT 102
*C -
7829.8 2.17 I 7921.2 2.20 NO BURN 'IN ANNULAR OACT 122
,7929.8 2.20 BURN IN PROGRESS IN ANNULAR.CMPT 122-lf
8024.9 2.23 BURN IN PROGRESS IN UPPER OAPT 102 8052.9 2.24 NO BURN IN UPPER CMPT 102.
;q
,j 8291.6 2.30 SURN IN PROGRESS IN UPPER CMPT 102 83'16.4 2.31 NO SURN IN UPPER CMPT 102-8687.3 2.41 BURN IN PROGRESS IN UPPER OAPT 102
~
8725.5 2.42; NO BURN IN UPPER OAPT 102 7 8931.8 2.48 BURN IN PROGRESS IN UPPER CMPT 102 id 8954.0 2.49 NO BURN IN UPPER CMPT 102
. 9075.7 2.52 BURN IN PROGRESS IN UPPER CMPT 102
, .(! 9118.3 2.53 NO BURN IN UPPER OAPT 102
'" 122 9184.5 2.55 NO BURN IN ANNULAR CMPT ,
9204.5 2.56 8 URN IN PROGRESS IN ANNULAR CMPT 122 ]
,3 J 9219.0 2.56 NO BURN IN ANNULAR CMPT 122 1 9233.5 '2.56 8 URN IN PROGRESS IN ANNULAR CMPT 122 n 102
'9395.8 2.61 BURN IN PROGRESS IN UPPER CAPT ii l" 9414.2 2.62 NO SURN IN UPPER CMPT 102 l 102 1 9740.5 2.71 BURN IN PROGRESS IN UPPER CMPT L'd 9759.4 2.71 NO BURN IN UPPER OAPT 102 9827.2 2.73 BURN IN PROGRESS IN UPPER CMPT 102 1
. y-5$1 n
'b 4.6-3
'd; .
~ ^
e - _ - _ _ _
i m u 4.6-1 .
. AD U6MAAP ..
CONT.." 1 i SEC HR l EVENT DESCRIPTION l ~ l CODE . l
. 9847.0 2.74 NO BURN IN UPPER CuPT 102 i 9891.6 2.75 NO BURN IN ANNULAR CMPT 122 9896.5 2.75 BURN IN PROGRESS IN ANNULAR CMPT -
122 10084.0 2.80 BURN IN PROGRESS IN UPPER CuPT 102 10106.8 2.81 NO BURN IN UPPER CuPT 102 l 10167.3 2.82 NO BURN iN ANNULAR CuPT 122 l 10176.1 2.83 BURN IN PROGRESS IN ANNULAR CMPT 122 1 10274.5 2.85 EURN IN PROGRESS IN UPPER CuPT 102 10294.2 2.86 NO BURN IN UPPER CMPT 102 10339.4 2.87 NO BURN IN ANNULAR CMPT 122 l 10359.9 2.88 BURN IN PROGRESS IN ANNULAR CuPT 122 l 10379.0 2.88 NO BURN I N ANNUL AR CMPT 122 l 10418.1 2.89 BURN IN PROGRESS IN ANNULAR CuPT 122 10527.3 2.92 . BURN IN PROGRESS IN UPPER CuPT 102 10552.5 2.93 NO BURN IN UEPER CuPT 102 10685.8 2.97 NO BURN IN ANNULAR CuPT 122
~
10739.8 2.98 BURN IN PROGRESS IN ANNULAR CUPT 122 10834.3 3.01 NO BURN IN ANNULAR CuPT 122 10857.2 3.02 ,8 URN IN PROGRESS IN ANNULAR CMPT 122 10867.0 3.02 NO BURN IN ANNULAR CuPT 122 10876.9 3.02 BURN IN PROGRESS IN ANNULAR CMPT 122 10941.7 3.04 BURN IN PROGRESS IN UPPER CMPT 102 10975.2 3.05: NO BURN IN UPPER CuPT 102 10994.3 3.05 NO BURN IN ANNULAR CuPT 122 11062.1 3.07 BURN IN PROGRESS IN ANNULAR CMPT 122 11116.6 3.09 NO SURN IN ANNULAR CuPT 122 11132.5 3.09 BURN IN PROGRESS IN ANNULAR CuPT 122 11144.3 3.10 NO SURN IN ANNULAR CuPT 122 11154.7 3.10 BURN IN PROGRESS IN ANNULAR CMPT 122 11314.3 3.14 NO BURN IN ANNULAR CuPT '22
, 11323.4 3.15 BURN IN PROGRESS IN ANNULAR CuPT 122 11348.7 3.15 ICE DEPLETED' 132 11351.4 3.15 BURN IN PROGRESS IN UPPER CuPT 102
, 11356.0 3.15 NO BURN IN 1/C UPPER PLENLM 141 5
11386.2 3.16 NO SURN IN UPPER CuPT 102 i j 11422.7 3.17 NO BURN IN ANNULAR CuPT 122 l s 4.6-6 ,
a ' t- : 5.0. Pinne Rispense'with Rtcovary Actions g A series of parametric studies was performed to determine the ef fects of ' w( i :) ' ~ ' 41 . . the number or ' amount of emergency core cooling' system options available
!\ .
d ;)M) i and possible beneficia1' as - well as detrimental operator . actions. The 4 },7- - l followi'ng cases were selected for this study:. -- y' .gr, SgD - Minimum safeguards ,
' (.
- :s -
Full restora't ion of injection (pre-core melt) Secondary-side (steam generatort blowdown ra' '
'l
,; '.U - SH2 - Mininem safeguards
< pr Partial restoration of. recirculation (pre-core melt)
L j ' *' TMLB'- - Comple te power restoration (pre-core' melt) l, j:t:4
.' [l Complete power restoration (post-vessel failure) ni T - Bleed and feed mode 23ML
-iC g Il.Ij Feed and bleed ,
; AD - Minimum safeguards a
,g . 5
,jy ~ Full restoration of injection (pre-core melt) q.
'i
, a
.t
[J-
. 4 s
7 AJ a
'3 1
..r" 15 .
" !.d .
> r
,1,.4.
j :L 1
,; .2 ^1- A ia 1
i'i j .L e 5.0-1 g sa l 1
.I
-----,__~_.s. . - . - - - - - - - - - - - - -
L ' 5.1 S,D S*cuenens
~ .
5.1.1 Minimum safeguards - S 2D
.1 The minimum safeguards case assumes that only one air return fan and e .t > } one containment spray pump are available during the. accident. Since i # ~
the ECCS is unavailable during the accident, the primary system response is identical for both cases. ' The predicted lower compartment pressures for the base case and the 3 minimum r:afeguards case are compared in Figure 5.1-1. Prior to ice a melt, the somewhat higher (less than 1.0 lbf/in2 )alower compartment pressure during th'e accident for the minimum safeguards case 'is due to the reduced air flow through the ice condenser and the reduced heat removal capability of the single containment spray. The rather rapid increase in pressure at approximately 5.2 hours (for the minimum . safeguards case) is coincident with the depletion of the ice. At
~
i <e 'Approximately 7.5 hours the heat removal capability of the single c ontainmen t spray exceeds the heat production rate of the quenched debris bed and the lower compartment pressure starts to decrease. The maximum containment pressure reached in the minimum safeguards case is 22.5 lbf/in2a at approximately 7.5 hours, as compared to 21.1 lbf/in2a for the base case, both of wt are well below the containment failure pressure of 65 lbf/in2 ,, 5.1.2 Full Restoration of Injection - S D 2 The purpose of this modified base case is to determine the ef fect of regaining full ECCS injection capability prior to core support plate
] f ailu re . Full ECCS injection is defined as regaining all charging pumps, all safety injection pumps, and .11 RHR pumps. Restoration of
; 5.1-1 .
i i i EI
~
l
L. , l injtetion occurs at 1.5 hcurs. At th2 time of injsetion, tha raector-vessel water level is,approximately 10.0 feet (see Figure 5.1-2),- corresponding to the bcetom of the active fuel. The total hydrogen {? ~ mass produced in the core is plotted in Figure 5.1-3 for the two lj.
.. cases. Note ~that the quenching of the core produces 97'S lbm of _
l- hydrogen about (325 lbm more hydrogen than produced in the base case). 1 , l' ' The greater hydrogen production is due to the ECCS injection water l providing additional steam for the Zr, H 2O reaction. 1- - 1 i., % Figures 5.1-4 and 5.1-5 show comparison of the two cases of the l primary . system corium temperature and upper compartment hydrogen mass, respectively. The maximuin nydrogen taass in the upper compartment for the full restoration of injection case and for the base case ('see Figure 5.1-5).is 225 lbm. The lower compartment pressure for the full. restoration of injection case does not exceed 20.8 lbf/in2,, ,,11
,1 below the containment failure pressure of 65 lbf/in2a (see Figure
..c , -
i* 5.1-6). l I l 5.1.3 secondarv side (Steam Generator) Depressurization - S2 D The purpose of this case is to determine the effectiveness of depressurizing the steam generator to cool and reduce primary system i pressure, given a small break scenario with no high pressure injection I but with low pressure injection (RHR) available.
~
'I 1
J j 5.1-2 a.
.m
Tho plots of . the primary system pressure and core . water temperature . for the secondary side blowdown esse are shown in Figures 5.1-7 and 5.1-8, respectively, where they are. compared to the respective plots .
~
l i d for the base case. Due to .the rather rapid decrease in primary side 1.
-l pressure (i.e., the primary pressure decreases from 2250 lbf/in2, ,e 0.0 hours to less than 200 lbf/in2 at 0.4 hours), it is possible for the RHR pumps to start injecting water into the primary system at 0.4 hours. . Thus, the reactor vessel water level is maintained throughout.
the accident as shown in Figure 5.1-9. Since the core is never i uncovered, there is no hydrogen production. i t W e 0 e 9 9 r
. s l-l.*
I . 1 l 5.1-3
-[
9
_ eI c *~- W -- _ .3l. s E h
~
g' i. f .
. s P N'
-L
~ A , 's A
,i.
-- M ,
i'I
.d 4
9 . . U .. W
+
/
P ,,, x
,L A ~ , .
A s t. M 1 , U , D c 2 ~, 5 S k ' w
.s i
- r. . .
v n .
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_. ,L Q.m )r K(n _ n30u D d 3: O;_ t s.per 1
S2D U1MAAPAJSMAAP N. i
*: i D. /
I N i , ,
; a: '. i s
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n o: '. l
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EP, o-a, sie ,' i d, .
- o. :
I o.oo o.:S c.5o c.75 too t:5 tso t75 2.00 1:5 2.50 175 3.00-TIME (hr) o FIGURE 5 1-2 8 r 9: .... o- l e l
. \'
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N*: l 08 Uo- . Eo: ! i o- '
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. 3-l ,
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, 2< , i o:
o l j o.co chs cho o.7s t'ao Us t'sc Us :.'co Us :.so :.7s 2.c o
- TIME Cnr) 1 FIGURE 5 1-3 l
l-5.1-5 l l - l
'3 S2D U1MAAPAJ8MAAP p .
r.1 l'd g
- p. ,
2w i9 i:
- a. ,-
/
.' p.1 * -
Chl
* '} - ,
s e m- l e 4 eq *l ! - j 'j- 8 o
?
3 Co' l I i'. hf gn- ( l M' g ) {g. l j o. S I. . 8- l l....*.. . . . , . . . . . . . ...................
',.. . o, d
l' .;j 0.0 d.5 50 iS 50
. 55 io 3.0 3.5 A5 ~
TIME Otr)
.p o FIGURE 5 1 h e
( ,*
, h'
,-- o;
** N N< ,
. e. .
' - 9 '
., A (
R; v.. . - s....h N i
, , ...... .........3.........,
C. v mo' Md,
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;lj T'ME Chr) e
,i FIGUP.E 5 1-5 l '$
1 '* l l F. , 5 1-6 e
!Ii!Illl
, l> .
T
,- ,!Nn E E ye& ~ 0 5
2 m m
,~
5 _ 2 2 r A 0 0 2 P
% 5
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t A ~ A M 8 Q t-0 5 1 U g
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- 5. (
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S2D UiMAAPAJ10MAAP . e n l l . ...
- 4. ,u e.
l n l .* . l I ,.1 e- ,
-i
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=g g .,y
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r ~. I e. , [ *.: M'. n '... ........ ,,,..............
. ..+...f,.. ...... . . . . . . . . . ...........: ...........
. -. .; 1 C. 3, d' . .
c.0 0.5 to t$ 10 15 10 15 Ao AS 5.0 I' d. . # ~ TME 0e)
.,. . FIGURE 5 1-7 u
.1 s
*l t.
. N U E; \ \ i
, ~: 1:
F.: i: . ( : i g'
. .s, .
I t
'a, r- b<
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. 8 -
n ., 1,- ...,... I" 9. q, y* *"* ... ,-
-....................,'s'.
- o a.,
h' l.- c.o is i.o U lo is io 2.s de As 5.o
' ld -
TME (br) l FIGURE 5 1-8 e Li n 518 3 .
~ _
z, s . r _ u _ P _ A _ A _ M 0 1 U . y
/
P
'h A #
A _ M 1 U D _ f 1 2 . S _ 1 \
. O F 1 iN w 9oN kW 99 Oiu o6 kI NJ g t
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e J 5.2 s.H saquaneos
- a. .
! 5.2.1 Minimum safeguards - S H 2 The minimum safeguards case assumes that one air return fan, one
- i. ,?j -
-!,-(d
, safety injection pump, one charging pump, . one RHR _ pump, and one containment spray pump. are available. For this casa, partial
~
- r-{'})
injection is available until the recirculation switchover point is . G l' reached, then only the one containment spray is available. 1 The reactor vessel fails at 3.57 hours for the minimum safeguards
,r case which is 0.26 hour af ter the reactor vessel fails for the base t-case. This is due to the fact that only one spray spump is operating
- so the time to recirculation switchover is extended, thus delaying the
~
D,- time to vessel failure due' to an i$ creased mass of water available for i . . .
, .; injection. -
b!
~
jf The lower compartment pressures for the
- mini:m2m safeguards case and r.
{- the base (S2 H) case are plotted in Figure 5.2-1. Note the m . similarity in the pressure spike at vessel failure for both cases. I ,! i
- i. Prior to ice melt, the somewhat higher (less than 1 lbf/in2 a) lower i compartment pressure for the minimum safeguards case is due to the
' ~
' reduced air flow through the ice condenser and the reduced heat rq l f,1 removal capability of the containment spray. The rapid increase in pressure at approximately 4.5 hours is nearly coincident with the il !
i,,j depletion of the ice at 4.65 hours. At approximately 7.5 hours, the
]
heat removal capability of the single containment spray exceeds the
' t. j heat production rate of the quenched debris bed and the lower i
d' { I L3 5.2-1 7 1 4
~
L
~
compartment prassure secres co- dacrassa. Tha maximum containment. pressure reached in the minimum safeguards case is 23.8 lbf/in2a,
..l.
which is well below the containment failure pressure of 65 lbf/in2,, _ h lyT .+ 's 5.2.2 Partial Restoration of Recirculation - S 2H - This case assumes minimum injection is available with one charging - pump, one safety injection pump, and one RHR pump operating. Once the
- recirculation switchover point is reached, the ECCS fails as in the base case, and minimum ECCS injection capabilities are not restored until 2.5 hours.
The plot of the reactor vessel water level for the partial restoration
, of recirculation (at 2.5 hours) case is compared to the plot for the I base case in Figure 5.2-2. The reactor vessel weter level for the minimum ECCS injection case decreases at a slightly greater rate (see av.
F,igure 5.2-2) due to the reduced ECCS flow rate into the reactor
~
vessel. When injection is restored at 2.5 hours, the water level in the reactor vessel recovers within 0.5 hours. 4 The plots of the total hydrogen generated in core for the partial injection case and base case are shown in Figure 5.2-3. Note that, as expected, the total hydrogen production for the partial injection case is greater than for the base case (by approximately 300 lbm). The greater hydrogen production is due to increased Zr, H O 2reactions due to the addition of ECCS injec tion water to the hot, semimolten Core. j ; l 1 i i
~'
-l
! 5.2-2 1
i f s
~ + - - - - .__ . . . . . . . . . . . --w- -
a .l Th3 plots of tha primary systsm corium temparcture for the two cases 1 [,, are shown in Figure 5.2-4. The rapid reduction'in corium temperature
,. . g ij- for the partial injection case can be seen at 2.6 hours due to the j\
quenching of the core. ' i n. l,
!N . Figure 5.2-5.shows the upper compartment hydrogen mass of the two j 9 cases. Note that the maximum amount 'of hydrogen in the. upper I. '
compartment is nearly the same for both cases (at approximately 225 l Ibm). This iss due to the ef festiveness of the hydrogen igniters which
.a .-
maintain the hydrogen concentration at the lower limit of flammability i 'i
; i. throughou t the accident. The containment pressure never exceeds 20.4
.t Ibf/in2 a, which is well below the 65 lbf/in a2 containment failure I-
. u pressure. .,
R
,N '
- sg .
,9 ,
'M g #
L g. s e 6 l *, tr.i
,F
' .i
-u
, 9 s
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T.1 i'.*, i !.J i l 5.2-3 l t.a
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.._._-__--._-.__._____.___J
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' []
5.3~ TMLB' sequ2neas rn 5.3.1. Complete Power Restoration at 2.5 Hours - TMLB' { This case compares the base case described in section 4.4 with the jfd assumption of complete power restoration at 2.5 hours. The purpose' of
. rm , this modified base case is to determine the overall plant response to 6
g. restoration of all plant functions before vessel failure. Just prior I to power restoration, the pressurizer is empty, and the vessel water level is below the bottom of the active fuel. At 2.5 hours, all ECCS
.p ,tj pumps, containment sprays, air return fans, igniters, and auxiliary ,n feedwater pumps are restored. Referring to Figures 5.3-1 through 5.3-5 the primary system pressure shows an immediate decrease (at 2.5
[l} hours) as feedwater refills the steam generators providing an a ef fective heat sink. The primary dystem pressure reaches a cdnicum at q
!1 C'
about 20 minutes af ter power restoration. At this time, the.ECCS ~ pumps quickly refill the primary system, quench the fuel, and ' drive r1 jy the pressure up to the shut off head of the charging pumps. The
, . f] containment response shows a maximum pressure of 21 lb/in2, IJ .
corresponding to the failure of the quench tank rupture disk due to f .] j the pressurizer relief valves relieving the primary system pressure at
-. approximately 1.66 hours. Containment pressure is rapidly suppressed d
iL to about 18 lb/in2a af ter power restoration and remains relatively Ti constant. With power restoration at. 2.5 hours, the integrity of c.) vessel and containment is never challenged. r] i r.,, I l l i J j p 1j h. u lj L '~1 5.3-1 [3 ! e 1 l 7 a a
l 5.3.2 Complete Power Restoration at 5.0 Hours - TML3'
; This case compares the base case described in section 4.4 with the 1 '
l assumption of complete power restoration at 5.0 hours. The purpose of ~ 1. lj this modified base case is to determine the overall plant response to j - restoration of all plant functions af ter reactor vessel failure. - Referring to Figures 5.3-6 through 5.3-11, at 5.0 hours the reactor vessel has already failed and all the invessel hydrogen has been released and the hydrogen mass in all the compartments is at a maximum. The containment experiences a maximum pressure of about 30 Ib/in2a at approximately 3.5 hours due to reactor vessel failure. The upper compartment gas temperature and hydrogen mass are 1800y and 140 lbm, respectively at 5.0 hours. Immediately after power 1 restoration, the ECCS system and containment sprays rapidly add water to the containment resulting in flooding the cavity thus providing suf ficient water to keep the corium quenched. There fore , no concrete i attack can occur and the vigorous concrete ablation rate with noncondensible gas generation as seen in the base case is precluded. When the containment sprays are actuated the temperature and pressure of the lower compartment are quickly suppressed. The containment pressure decreases to about 18.5 lb/in2 a Once the injection phase is complete, successful realignment to recirculation is achieved. Af ter 'about 6.5 hours, containment heat removal continues to decrease l the containment temperature and pressure. Therefore, c on t ainment l l integrity is not cha llenged. t t i 1 j 5.3-2 l O
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O 5.4 T..,ML Sequenees
- L]- . . ,
l :# 5.4.1 Bleed and Feed - T 23E
] This case compares the base case dese-ibed in section 4.5 with the - .,d -
bleed and feed sequence. The purpose of this case is to determine the
] plant response to an accident sequence in which the recovery actions
,_ require bleeding the primary system and using the,ECCS pumps to remove I
~~
i. energy via injection water. The assumptions are as follows (1) the
'- ~
, PORV manually opened at 2500 seconds and (2) ECCS injeceion (two t s
. charging, two safety injection, two RHR pumps) available at 2500 seconds. At time zero, the reactor is scrammed and no auxiliary feedwater is available. Referring to Figures 5.4-1 through 5.4-6, the
-t
!. ; pressurizer water level and pressure initially decrease but quickly ry stabilits themselves in response to the actuation of the pressurizer
- l"A -
heaters. The steam generators pressurize to the relief valve set j point and rapidly lose inventory due to the loss of auxiliary
'j-* ,. .
feedwater resulting in roughly 85 percent of the inventory being lost t by 0.70 hours. At approximately the same time, 0.70 hours, the PORV is opened resulting in a rapid depressurization of the pressurizer (primary system) pressure to approximately 1150 lb/in2,, ge [ about 0.72 hours, the ECCS pumps start to refill the primary system. L. The injection of cool water immediately causes primary system pressure i-and temperature suppression. The ECCS injection and loss of secondary pc side cooling drives the pressurizer solid at 0.8 hours and the quench
^"
tank rupture disk fails at 0.82 hours. The PORV discharges liquid water as the charging pumps circulate cooling water into the cold leg, w t*1 m
..i 5.4-1 n
) I L______--.
~
L lI i ij . i 133tv2cn.cbsut.1.5' cud 3.0 hours, th2 pressurizer (primary systsm) l pressure decreases from about 1550 to 1450 lb/in2 a The core - !} [ (primary system) water temperature decreases. from about 565 F at 1.0 ' - j hour to about 400 F at 3.0 hours.. ' ' Beyond 3.0 hours, the primary
; system pressure \
slowly decreases due to the equilibration between the " h pung injection rate and the flow out of the PORV. Although the . pressure remains somewhat constant, the primary sys tem water : temperature continues to decrease as the ECCS heat exchangers and
~
containmen t spray continue to remove decay heat from the primary system and the containment. Referring to Figure 5.4-6, the ' I containment maximum pressure'of'approximately 18.6 lb/in2 a occurs
; around 6.0 hours due to the depletion of ice in the ice condenser, bde
~
l it- is suppressed as the decay heat relieved from the primary system is exceeded by the heat removal rate of the ECCS heat exchanger and . containment sprays. s 5.4.2 Feed and Bleed -T23ML This case compares the base case described in section 4.5 with the feed and bleed sequence. The purpose of this case is to determine the
?
plant response to an accident sequence in which the recovery actions require using the ECCS pumps to remove energy via injection water. The assumptions are as follows: at 3000 seconds, the charging pumps, safety injection,' and RHR pump are turned on forcing water into the primary sys tem. At time zero, the reactor is scrammed and no auxiliary feedwater is available. Referring to Figures 5.4-7 through !
, 5.4-12, the pressurizer water level and pressure initially decrease 6
I i- 5.4-2 - 1, 1 5
-)
x_ . . _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _;
but quickly stabiliza chmsalvas in respsnsa to tha actuation of tha j pressurizer heaters. The steam generators pressurize to the relief valve . set pc. int and rapidly lose inventory due to the loss of
~
auxiliary feedwater resulting in steam generator dryout at 0.98 hours. ( At approximately 0.83 hours,' the ECCS- pumps are turned on forcing cool 1 1 r M s. l d water into the primary system. The injection of cool water
- immediately causes primary system pressure and temperature
.[
a. suppression. Soon af ter initiation of the ECCS injection, the pressurizer is solid at approximately 1.0 hour and the quench tank
-, rupture disk fails at 1.14 hours. The PORV discharges liquid water as p-t/ the charging pumps circulate cooling water into the cold leg. After lj 1.0 hour the - pressurizer safety relief valves are automatically ,
.=J j opened to relieve excessive pressure of primary system, the
- g) :
,. {d pressurizer is maintained at approximately 2350 lb/in2a The
'~]
primary system pressure remains constant due to the equilibration uj
- 6etween the pump injection rate and the flow out of the PORV.
m lc Although the pressure remains constant, the primary system water u , temperature continues to decrease as the ECCS heat exchangers and [.) containment spray continue to remove decay heat from the primary p' system and the containment. The containment pressure is maintained at M about 17.4 lb/in2a until the ice is depleted at 7.5 hours and then f increases to about 19.5 lb/in2a at 10 hours. After 10 hours, the decay heat relieved from the primary system is exceeded by the heat removal rate of the ECCS heat exchangers and containment sprays p causing the containment pressure to decrease.
.)
i p i
} II.
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=
L l 5.5 AD Secuences .. 1 5.5.1 Minimum Safeguards - AD 5 The minimum safeguards case assumes that only one air return fan and [ i one containment spray pump are available for operation. Since both 1 - , ECCS injection and recirculation are inoperable, the primary response is identical for both cases. ' The predicted lower compartment pressure for the minimum safeguards case and the base case are compared in Figure 5.5-1. Pr'ior to ice melt, the somewhat higher (less than 2.5 lbf/in2 )alower compartment pressure during the accident for the minimum safeguards case is due to the reduced air flow through the ,,ce 1 condenser and che reduced heat removal capability of the single containment spray. The rather rapid P increase in pressure at approximately 3.25 hours is nearly coincident with deple tion of the ice in the ice condenser at 3.23 hours. At approximately 6.0 hours, the heat removal capability of the single containment spray exceeds the heat production race of the quenched debris bed and the lower compartment pressure starts to decrease. The maximum containment pressure reached in the minimum safeguard case is 23.8 lbf/in2a at approximately 6 hours, which is well below the containment failure pressure of 65 lbf/in23, . 5.5.2 Full Restoration of Injection - AD The purpose of this modified base case is to determine the system response to regaining full injection capability prior to core support plate failure. Complete restoration of injection is assumed l 5.5-1 -
!t -
b g to occur at 1.1 hour which is when the water level in the reactor l
- , : vessel is at 10.5 feet (see Figure 5.5-2), approximately the bottom of 1
]@[- the active fuel.
, w i- . ?:- -
f] When'ECCS operation is restored at 1.1 hour, the water level in the reactor vessel is restored to its normal height within 0.1 hour. As 61 il was noted in section 5.1.2, the quenching of the core produces more
. [] hydrogen than the base case (approximately 900 lbm). As noted in F
'l.-
section 4.6, hydrogen production can occur in re-covered, unmolten D f.j nodes in MAAP due to the severly restricted heat transfer which is
,, allowed to take place between the cladding and the water. The model q . - ;J ' used during reflood assumes an intact geometry which conservatively
[ bounds hydrogen production (reference 6.7). L]I . la.
.tj.I Die primary system corium temperature and upper compartment hydrogen
'y mass for the full restoration of injection case are plotted in Figures
- i. 5.5-4 and 5.5-5, respectively, where they are compared to the base
;) case plots. Note that the maximum amount of hydrogen in the upper compartment for the full restoration of injection case is 280 lbm p,
t" compared to 225 lbm of hydrogen base case (see Figure 5.5-5).
.a
. FT.
,e .
!~j
. The containment pressure never exceeds 22.8 lbf/in2a for this case,
[7 thus containment integrity is never challenged. 7 k.i y h l . i il ta 1
' [:; -
ig
- i. 5.5-2
; j
" 1
p
- o s
~_
n m-4 P A A M 9 N ,~ 1 J A P ' A A
.~
. .g w M _ 6 U g r.,, e
.y-D 's, a
A '
+
?'
L a V
' V
$; i i'. L
. I
_ s I ' '% . J
. n
_f l g _ g s-e:: I.a: ,!
, ,* k N
_ r, . g syoJo. , : 8' 3 a . - . -
' =
o4m 0IF o4N c ow 'o h 9e o!d 0,w od o b' o sO a m g OUet30I a .
- m yu ,
d AD 'U6MAAPAJ20MAAP 1 l i, 3 0 n, 'i "y i l M. N l -g :
.. 3 i !) a: l u N. ,
r o_
~"'o. :
o k
!. 7. gg: .
., g, o, l
- g< 5f .
; /
- a. '
! l i
- e. :
\ -- l l.
l I
\
h8'
., t
\
tu , : c: s - ol 6 m,
- }- .:
en O# d
- v. - o' . .
O.0 0.2 0.4 0.6 0.8 to t2 to t6
'a -
TIME Oy)
. FIGUEE 5 5-2 g.
L .,
..~ e*
- b ,eoeeeoea eeeoee*********************
- e. n ,
. 9 i i yo' *
' d$'
O ** e u o' .' aa< t' e . Ua 8 74 l
' 1 ; <o' .!
{i gd-
- s. :
6 4 p o, ', no rt [. Ec: O i-
' ..; 6- 7
}
- %.! o b<
f a
"t
; 0.0 0.2 d.4 d.6 d.8 (0 (2 te (6 te 2.0 2.2 j;3 T1ME Oy)
FIGL7.I 5 5-3 e J l
'i ! 55h i i f7 1b 1
______.____.___Q
AD U6MAAPAr20MAAP '
'} N -
j< t ~ 1.- n. t.. 8 *
.* T
/ -
1., 3 . i j . G 4
... l l, co. j
' I .
q 4. ; O. I o a 8 - n
$0 M
( !
?
- a. ..
s t g- . l l
, a......................... *-
a. 1 6 0.0 dJ to t$ 10 2.5 3.0 $.5 de AS T M E G r) ,
*. FIGUF2 5 5 h
. o. .
.'.?* .
- o. \ .
6 ' S. I. 3;
;n.
i y f % ,... W_ ---. s - 0' 0
-8 I
U o. 6 . 9 e' 2 ,
- )' l l
l o' ? 0.0 0.5 to LS 2.0 2.$ 10 3.5 Ao A5 TME (br)
- FIGURE 5 5-5 i
5 5-5 .. 4 e
. - + * + -
__ _____.___________m_._ . _ . _ . - - - _ _ . - - . _
e t i 6.0 Fission Product Release. Transoort, and Deposition a l 6.1 Introduction The phenomena of fission product . release from the fuel matrix, their
$a A,
transport within the primary system, their release fro'm the primary -
.I system into the containment, their deposition within the containment and 7 the subsequent release of some fission products from the containment are treated through the use of FNJ (reference 6.1). , Release of' fission
~
products from the fuel matrix and their transport to the top of the core are treated by a subroutine in MAAP which is based on the FPRAT code
,' (reference 6.2). Transport of fission products outside the core r- boundaries is treated by fission product models in MAAP described in v.
1 re ference 6.1. Fission product behayior is considered for the best
] estimate transport, d eposi t i.on, and relocation processes as discussed in ,
ij . re ferences 6.1 and 6.4. The influence of surface reactions between
.s.
.] chemically active substances like cesium hydroxide and other
~
p- uncertainties are considered in Subtask 23.4 (reference 6.3). The best h estimate calculation, assuming cesium iodids and cesium hydroxide are
~
the chemical state of cesium and iodine, is discussed below.
- a 6.2 Modeling Aoproach g
}j Evaluatio'n of the dominant chemical species in reference 6.5 show the i
_, states of the radionuclides (excluding noble gases.) which dominate the U public health risk to be cesium iodide, cesium hydroxide, tellurium, and [', strontium oxide. These and others are considered in the code when calculating the release of fission pro d ucts from the fuel matrix. p dw Vapors of these dominant species form dense aerosol clouds in the upper . 1 plenum, in some cases approaching 100 g/m3 for a very short time, which agglomerate and settle onto surfaces. Depending upon the chemical [1 compound and gas temperature, these deposited aerosols can be either O 6.2-1
PARAMETER FlLE CONTE i m
! == ,
l alNITIAL CONDITIONS e 01 578.2 NOMINAL FULL POWER PRIMARY SYS WTR TEMP
. 02 2250. NOMINAL FULL POWER PRIMARY SYS PRESSURE
.; 03 31.12 PRESSURIZER WTR LEVEL i 7
04 15.0 CONTMT BUILDING PRESSURE l' 05 100.0 LOWER CONTMT COMPTS TEMP 06 60. ICE CONDENSER GAS TEMP 07 0.30 LOWER CONLfT COMPTS RELATIVE HLMIDITY.(0-1) , OS 2.106 INITIAL ICE MASS " 09 94877.0 INITIAL MASS ON SECONDARY SIDE OF EACH S/G
. 10 93.0 INITIAL TEMP OF CONC AND METAL STRUCTURES 11 857.0 INITIAL PRESSURE ON SEC SIDE OF S/G'S l 12 .300 UPPER CCMPT REL HLMIDITY j 13 85. UPPER CCMPT TEMP 1
** 14 I N ITI AL PR IMARY SYS WTR TENF ~f J
- ** 15 INITI AL PRIN%RY SYS PRESSURE
- PRIMARY SYSTEM 01 4 NO. OF CLD LGS 02 2.459 DIAMETER OF A. HOT LG PIPE -
03 6.958200 RADIUS OF RX HEMISPHERICAL HD 04 9.805 ELEV OF SUPPORT PLATE 05 56.07 FLOW AREA OF CORE + BYPASS AREA 06 104.12 VOL OF THE HOR!Z LG OF PIPE THAT RUNS
** FROM THE CLD LG NZL TO THE PMP BOWL
^
07 6.4170-2 RADIUS OF VSL PENETRATIONS 08 1.023907 ENERGY INPUT FRCM A PRIA%RY SYS PMP
** (WHEN RUNNING) -
09 39500 PRIMARY SYS A%KEUP FLOW--NORAMLLY SHOULD
** EQUAL LETDOWN FLOW BELOW 10 130.Do TEMP OF MAXEUP WTR 11 2.2917 DIAMETER OF A CLD LG PlPE 12 27.328 ELEV OF SURGE LINE ABOVE SOTTCM OF RV 13 4 ENTER BREAK LOCATION KEY:
** 1 CLD LG NODE (PMP BOWL OR VERT SECT LEADING
** TO S/G OR S/G CLD LG SIDE (SEE USER'S h%NUAL)
, ** 2-ALL CLD LGS (EG PMP SEAL LOCA-ENTER TOT SK AREA f
A.1-lk , l -
* ' * *~ e m- me s e e -.p m y hJ t
j ,^r e PARAMETER FlLE CONT. I' ** FOR NO. 14) .
== 3--DOWNCCMER NODE (lNCL HOR 12 PART OF CLD LGS ra ** RUNNING OUT TO THE PMP BOWL r.- ** 4---HOT LG NODE 14 .0218 PRIMARY SYS BREAK AREA (HOT OR CLD LG)
,$ 15 27.328 ELEV OF BREAK ABOVE BOTTOM OF RV 16 ????? VOL LEFT IN ONE COOLANT LOOP IF THE CL PIPE
** GOING TO THE VSL JUST EMPTIED (PROP)
, 17 ????? MAX VOL OF WTR IN A CLD LG FOR GAS TRANS
** TO OCCUR ("PMP BOWL VOL") (PROP)
I 18 ????? VOL OF A CLD LG (PROP) 19 ????? VOL OF A HOT LG (PROP) 20 ????? FLUID VOL IN THE RX VSL (PROP) { ** 21 GAS FLOWRATE OF THE RX HI-POINT VENTS AT NCM
~
** SYS PRESS DOWNCOMER IS MODELLED AS ENDING AT THE POINT
** WHERE THE LWR HD OF THE RV MEETS THE CYLLINORICAL SECT -
22 '????? VOL OF THE DOWNCOMER (PROP) c, 23 ????? VOL OF THE DWNCMR BELOW THE CLD LG NZL (PROP)
- 24 1 ENTER A 1 FOR PZR TO BE IN BKN LOOP; O TO
{. ** BE IN UNBKN LOOP ,
'.' 25 4 NO. OF HOT LGS 26 .500 VOlD FRAC AT WHICH MCP'S TRIP OR FAIL
. 27 1815 LOW PRIMARY SYS PRESSURE SCRhi SETPOINT I' 28 2.400D3 HIGH PRIMARY SYS PRESSURE SCRAM SETPOINT 29 75.D0 HIGH LOOP DELTA-T SCRAM SETPOINT 30 12.51 LOW PZR WTR LEVEL SCRAM SETPOINT
' F",
4 .. 31 32 44.97 5.556D-4 HIGH PZR WTR LEVEL SCRAM SETPOINT SCRAM DELAY TIAE l 33 -1.DO LOW S/G WTR LEVEL SCRAM SETPOINT 34 5.DO NO. OF ENTRIES IN MCP PMP COAST-DOWN CRVS
** (5 MAX)
[' 35 3.450D7 FIRST MASS FLOW IN TABLE
** (MUST BE THE ONE-PMP FULL FLOW VALUE) lf. 36 2.415D7 NEXT FLOW IN TABLE
- 37 1.72507 NEXT FLOW IN TABLE
- i.j 38 8.625D7 NEXT FLOW IN TABLE L 39 5.17506 NEXT FLOW IN TABLE
. 40 0.00 FIRST TIME IN PMP COAST-DOWN TABLE
** (USUALLY ZERO) i e
l
, ! ,1 A.1-15 1 -- - - - - - - - - -
a - J PARAMETER FILE CONT., j'
', u
, 41 1.111E-3 NEXT TIME IN TABLE 42 43
.0025
.007222 NEXT TIME IN TABLE NEXT TIME IN. TABLE
[ 44 1.667E-3 NEXT TIME IN TABLE m ' 45 35.13500 ELEV OF TUBESHEET ABOVE BOTTOM OF RV HD , 46 .4792 THICKNESS OF RV LOWER HD 47 26,151 HGT OF 87M OF NZLS A80VE 80TTCM OF RPV
~'
48 10.458 DIST FRCM BTM OF CLD LG TO THE STM OF THE
** NZLS 49 ????? VOL OF THE HORIZ RUN OF HOT LG IN ONE LOOP
** (PROP) 50 40000 LETDOWN FLOW 51 3.46D1 NORML DELTA-P FRCM CORE INLET TO HOT LG
** S1DE OF OUTLET (NEW)
- PRESSURIZER 01 1.803 VOLLME 02 ????? CROSS-SECTIONAL AREA (PROP) 03 2.264703 PRESSURIZER HTR PRESSURE SETPOINT 04 2.2747D3 PRESSURIZER SPRAY PRESSURE SETPOINT 05 12.51 WTR LVL (ABOVE STM OF PZR) AT WHICH
** HTRS TRIP -
06 6.14306 NCMINAL HEAT INPUT FRCM HTRS 07 2.959E5 NCMINAL SPRAY SYS FLOWRATE 08 420000. NOMINAL FLOWRATE OF A SAFtif VLV AT ITS
** SETPOINT ,
. 09 2499.7 LOWEST SET POINT OF SAFETY VLV 10 2499.7 HIGHEST SET POINT OF SAFtiY VLVS 11 .932 DIAVETER CF SURGE LINE 12 52.833 ELEV OF SPRAY HD A80VE 80TTCM OF PZR 13 59.12 LENGTH OF SURGE LINE 14 3 NO. OF SAFETY VLVS 15 2.297D-3 NOMINAL PZR SPRAY DROPLET DIAVETER ;
16 2350. LOWEST SET POINT OF.PORV J 17 2350. HIGHEST SET POINT OF PORV 18 2 NO. OF PORV'S l 19 2.03605 NCMINAL FLOWRATE OF A PORV AT ITS SETPOINT
. 20 1.9551D5 EMPTY MASS OF PZR A.1-16 i
.( q b
N' PARAMETER FILE CONT. jQ 21 0.00 ENTER A 1 IF THE SURGE LINE HAS A LOOP SEAL l
,4 22 40. SEDIMENTATION AREA y,,
-1 I **
*STM GEN - (UNLESS SPEC lFIED, VALUES REFER TO ONE UNIT)
M- 01 5.86803 SEC SIDE VOLW E 02 ????? DOWNCCNER CROSS-SECTIONAL AREA (PROP) ]
,. 03- ????? SECONDARY SIDE GUNDLE FLOW AREA (PROP)
- 1. 04' O.00 8 AND W ONLY--HGT OF AUX FEED SPRAY - HD
** ABOVE LOWER TJBE SHEET 05 1.D8 INITIAL CST MASS-OR A LARGE NO. 1F NO LIMIT Ii ** ON AFWS SUPPLY 06 ????? NORWL OPERATION 2-PHASE BUNDLE HGT (PROP)
[ 07 4.34602 MA1N FEEDWTR TEMP
- 08. 1079.0 LOWEST SETPOINT OF SEC SAFETY VLVS .
09 1132.0 HIGHEST SETPOINT OF SEC SAFETY VLVS 1 10 5 NO. OF SAFETY VLVS PER S/G . 11 7.83405 NOMINAL ELOWRATE OF A SAFETY VLV L 12 1105.0 SETPO I NT OF ,SEC REL I EF ' VLV ( ASSLhED SAVE
** FOR ALL RELIEFS)
F' 13 1.DO NO. OF RELIEF VLVS PER S/G 14 8.9005 NCMINAL FLOWRATE OF A RELIEF VLV 15 3.73E6 MAX WIN FEEDWTR FLOWRATE PER S/G J. 16 320.3 VOL OF S/G PRIMARY HD
^
17 1.667D-2 TIME DELAY FOR ACTIVATION OF AUX FEED
) 18 1.944E-3 TIME DELAY TO SHUT MSlV (5 SEC)
[$. 19 1080.0 TOT PRl W RY SIDE VOLWE 20 2.178405 MAX AUX FEEDWTR FLOWRATE PER S/G h- 21 120.0 AUX FEEDWTR TEMP 22 3388 NO. . OF BJ8ES IN A S/G
- 23 4.167E-3 TH!CKNESS OF S/G TUBES
.; 24 6.458E-2 ID OF S/G TUBES 25 10.631 THER!#L CONDUCTIVITY OF S/G TUBES 26 1.E10 THROTTLED AFW FLOW PER S/G (SEE DISC / AT
]"
** END OF ESF SECT) 27 1. FRAC AREA USED FOR STM DUMPS IN BKN LOCP S/G
;3 b 28 1. FRAC AREA USED FOR Shi DLMPS IN UNSKN
** LOOP S/GS
; l.i 29 30 DOWNCOMER PROGRAM WTR LVL FOR SGWLC SYS
; a .
I' A.la17 L1
. - _ - _ _ - - - - _ - - _ - _ - - . - - _ _ _ _ _ . - - _ _ _ - - _ _ - - _ - _ _ _ - _ _ _ - - - - . . - . _ .--_-_--J
m= -- ---- . x . -
. . a . -. . _
p- ~ [u. PARAMETER FlLE . CONT. .. p 1 ** IN BKN LOOP S/G [ 30 30 DOWNCMR PROG WTR LVL FOR SG#LC SYS IN '
- i. ** .UNSKN LOOP S/GS
**. j'
- TIMING DATA (ENTER ALWAYS IN. SECS) ,
' 03 ~ 20.00 MAX TIME. STEP (SECS) 04 .005 MINIMLM TIME STEP (SECS) ~
- 05. .05- RELATIVE MASS CHANGE USED TO SELECT. TIME
** STEP 06 .02 RELATl'd OX!DATlON FILM THlCKNESS USED TO
** SELECT TIME STEP 07 .04 RELATIVE GAS TEMP CHANGE USED TO SELECT
** TIME STEP 081 .1 RELATIVE MASS CHANGE OF FlSSION PRCDS-
** USED TO PICK FP TIME STEP
*= ,
- QUENCH TANK ~
01 1800 VOLLME 02 55800 INITIAL WTR MASS 03 104.7 FAILURE DP OF RUPTURE DISK 04 13.2 HGT OF RUPTURE DISK ABOVE BCCMPT.FLR
** 05 SEDIMENTATION AREA
*MODEL PARAMS 01 .005 CORIUM FRICTlON COEFF 02 .4- FRAC OF CORE MASS THAT MELTS BEFORE
** CORIW PILE SLLMPS TO THE CORE SUPPORT PLATE '
03 .016 TIME TO Fall VSL PEN. WELDS AFTER CONTACT
** WITH CORlW -)
04 .033 TIME TO FAIL SUP. PLATE AFTER CM P1LE { 4
** HAS REACHED IT 05 2.0 MULT OF NORMAL CLAD SURF AREA TO ACCOUNT
** FOR POTENTIAL CLAD RUPTURE (MJST BE STNN 1 AND 2) 06 1300.DO CRIT 1 CAL FLAME TEMP AT ZERO STM MOLE FRAC
** USED IF NO. I GN SOURCES ; TH I S I S MJLT I PL I ED BY THE
** WESTINGHOUSE FLAME TEMP MULT CORRELATlON
** FOR HIGHER STM MOLE FRACTIONS 07 3700 ZlRCALLOY OXlDATION CUT-OFF AND CHANNEL A.1-18
.l
W A. A. PARAMETER FlLE CONT. j }{1- ** BLOCKING TEMP , oj" 08 50 NON-RADIATIVE FILM SOIL. HT TRANS CCEFF
**. FROM CORILM TO POOL N, , 09 150. NAT. CIRC. (MCP'S OFF) S/G PRIMARY SIDE
** FILM HEAT XFER COEFF P 10 .3 REFERENCE THERMAL BOUNDARY LAYER
** THICKNESS IN CONC
, y, 11 .100 S AND W ONLY: FRAC OF S/G TUBES
,' ** STRUCK BY AFW r 12 175 HT XFER COEFF STWN MOLTEN CM AND FROZEN CRST 13 0.00 ENTER A 0 FOR ENTRAINMENT FRCM C TO B;
. []
;L' ** 1 FOR C TO A
, . . 14 0.00 IF 19 IS NONZERO, FRAC MASS ENTRAINED i
** WHICH STRIKES THE MISSLE SHIELD WlTHOUT INTERACTING
** WiTH ACCMPT GAS 15 1.DO DRAG COEFF OF RISING PLLME DURING BURNS f ia ** IN A COMPT g .. 16 1.00 SAME FOR S COMPT
! 17 1.D0 SAME FOR C CCMPT
~
1.00 18 SAME FOR D COMPT 19 1.00 SAME FOR U CCMPT { > 20 1.53 CHURN-R)RBULENT CRIT l CAL VELOClTY CCEFF 21 3.7 DROPLET FLOW CRITICAL VELOCITY COEFF ['" 22 1. SPARGED POOL VOID FRAC COEFF 23 2. VOLUMETRIC STM GENERATlON VOID FRAC COEFF
; ], 24 .00014 ENTRAINMENT TIME CONSTANT
.; 4{ ; 25 1. EMISSIVITY OF WTR 26 .75 EMISSIVITY OF WALLS l 27 1. EMISSIVITY OF EQUIPMENT b' ~
- 1. EMISSIVITY OF CORit.M SURF 28
", 29 .75 EMI SS IVITY OF GAS
+. .' 30 .3 CORE HYDRODYNAMIC LIMIT KUTATELADZE NO.
** FOR REFLOOD1NG HT AND OXlDATION CALCULATIONS 31 100.DO NO. TO MULTIPLY KUTATELADZE CRITERION BY
** TO REPRESENT DIFFICULTY (GT 1.00) OR EASE (LT 1.00) FOR
! ** MATERIAL TO GET OUT OF CAVITY
.,U : 32 3.O FLOODING CRIT l CAL VELOCITY COEFF 33 .14 FLATE PLATE CHF CRITICAL VELOCITY COEFF i
l0J 34 1. NO. OF VSL PENETRATIONS THAT FA1L l t fj A.1-19
p PARAMETER FlLE CONT.1 35 .75 DISCHARGE COEFF FOR PRIMARY SYS BREAK i- 36 1.DO SCALE FACTOR FOR BURN VELOC1TY CCRRELATICN j 37 1.00 SCALE FACTOR FOR HEAT XFER COEFFI TO , PASSIVE HT SINKS '
** 38-42 ARE PARAMS IN THE GRAV AGGLOMERATION AND FALLOUT
** CORRELATIONS IN SUBROUTINE FPTRANS 38 .02200 MULT OF EXPRESSICN FOR H1-DENS AEROSCLS
** (NCM .022) 39- .001600 MULT OF EXPRESSION FOR LO-DENS AEROSOLS- '
** (NOM .0016) 40 .5900 DENSITY EXPONENT FOR Hi-DENS AEROSOLS
** (NOM .59) 41 .3300 DENSITf EXPONENT FOR LO-DENS AEROSOLS ~
** (NCN .33)
** 42. DENSITY AT WHlCH.WE SWlTCH FRCM THE H1-
** DENS EXPRESSION TO THE LOW (1.0-5 IN KG/M**3) 43 1.00 ABS (MULT) OF CSI AND CSCH VAPOR PRESSURE-
** ENTER A NEG NO. TO SELECT JANAF CSCH VP; POS FOR' SAND 1A 44 .1DO FRAC OF CLAD OXlDIZED WHICH CAUSES CORE TO
** COLLAPSE ON REFLOOD (GIVES SMALLER KU FOR HEAT XFER
** THAN INTACT MODEL) 45 0.00 FOR S AND W UNITS ONLY, FRAC OF PERFECT
** CONDENSATION OF STM ENTERING DOWNCCMER THRCUGH
- ** FLAPPER VLV3 (
** SI UNITS BELOW
*Si )
*FlSSION PRODUCTS
, ** FISSION PRCOUCT GROUPlNG SCHEME:
** GROUP 1: NOELE GASES AND " INERT" (NON-RADIOACTIVE)
** AEROSOLS
** GROUP 2: C3I
** GROUP 3: TEC2 AND TEM j
** GROUP 4: SRO (BA LLMPED IN) j L. ** GROUP 5: RU (MO LLMPED IN)
.- ** GROUP 6: CSCH I
A.1-20 f L - - - -
!.l
!O PARAMETER FILE CONT. I h ~~
** STRUC~NRAL MATER I AL GROUP 1 NG SCHEME
- ** USED IN CORE NODES (TRACKED IN CONTMT AS GRCUP ;
m ** 1 AEROSOLS) . E
** GROUP 1: CD
*= GROUP 2: IN l
I
** GROUP 3: AG l
** GROUP 4: SN
** GROUP 5: MN 01 .028 FRAC OF FlSSION PROD POWER IN GROUP 1 j 02 .151 SAME FOR GROUP 2 03 .0194 SWE FOR GROUP 3 !
- p. ** NOTE: CALCULAT1ONS USUALLY SHOW VERY LIiiLE SR OR
'- ** RU RELEASED-TO ENSURE CONSERVATION OF ENERGY, IT IS
** RECCt.NENDED AT PRESENT TO ASSIGN NO DECAY ENERGY TO
** THESE GROUPS-THE END RESULT WlLL BE TO DO THE
** CALCULATIONS AS 1F THE SR AND RU WERE LEr7 iN THE
*= DESR1S; THE CODE WILL CONT lNUE TO CALC THEIR MASS I,j ** TRANS AS IF THEY COULD BE RELEASED, HOWEVER,
** SO THAT THE ACCURACY OF THIS APPROXIMATION
[ ** CAN SE CHECKED. 04 0. SNE FOR GROUP 4 (ACT .062) 05 O. SWE FOR GROUP 5 (ACT .0547)
! 06 .011 SWE FOR GROUP 6 07 347. INITIAL MASS OF FISSION PRODS IN GROUP 1
( ** (NOBELS ONLY) 08 30. INITIAL MASS IN GROUP 2 o 09 31.7 GROUP 3 P 10 '139. GROUP 4 11 329. GROUP 5 12 151 GROUP 6 13 144. INITIAL MASS OF CD IN CCRE 14 421. INITIAL MASS OF IN IN CORE ll 15 2287. INITIAL MASS OF AG IN CORE 16 332. INITIAL MASS OF SN IN CORE
- jj INITIAL MASS OF MN IN CORE
~
17 202. lD 18 6. NO. OF MAXILM GROUP YOU WISH TO MODEL
'. *= (NORMALLY 6)
[ == [] 4.1-n 1
PARAMETER FlLE CONT.[
*. i
- CIRC
** ALL ENTRIES 1N THE ClRC SECT ARE CONSIDERED PROPRIETARY
** THIS SECT IS FOR INPl.TTS TO THE DETAILED PR IN%RY SYS T/H
]!
I'
** MODULE SUBROtJTINE C1RC AND 1TS ANC1LLARY SUSRCUTINES
** TO DEVELOP INPUTS FOR THIS SECT, CONSULT THE APPROPRIATE ii
** FIGURES FOR YOUR PLANT TO DELINEATE NODAL SOUNDAR!ES
** INSERT NO. IN THE SPACES SHCWN; THE CODE CCMPUTES THE "
l,
** MiSS1NG NO. FRCM THESE AND OTHER INPUTS
** ****************************************================ q
** NOTE SPECIAL DEF FOR SCME ENTRIES FOR B AND W l
** NCDALIZATION .
** CIRC ALLOWS EACH NODE TO HAVE 1 OR TNO MASSES; EACH
** NCDE HAS A " STEEL" MASS AND MAY IN SCME CASES ALSO
== HAVE A " HEAT SINK" MASS; THE HEAT SINK IS DISTINGUISHED 1
** FRCM THE STEEL MAINLY BY TNO DIFFERENCES: 1. MAY BE AT
** A DIFFERENT TEMP; THIS MAY BE DUE I N P A R T T O T H E H E-A T J
l
** SINK HAVING LOSSES TO COtriMT WHEN THE STEEL DOESN'T
*= (EG S/G FHELLS VS TUEES)
** 2,.' HEAT SINKS ARE ASSLMED NOT TO HAVE FlSSION PRCDS ;
** PLATED ON THEM I
- AT PRESENT, THE ENERGY EXCHS IN A NODE MAY INCL ONE
** OR MORE OF THE FOLLOWlNG, DEPENDING ON THE INPUT -
** PARAMS SUPPLIED:
** 1. HEAT SINK AND STEEL EXCH ENERGY RADiATIVELY (GAS
** ASSLMED TRANSPARENT) l '
J
** 2. STEEL EXCHS ENERGY WITH PR IN%RY SYS GAS VI A INTER-
** NODAL OR INTRA-NODAL NATL'RAL C1RCULATION _
*= 3. HEAT SINK MAY EXCH ENERGY WITH PRIMARY SYS GAS VIA l
** INTER- OR INTRA-NODAL NAT1JRAL C1RCULATION '
** ITEMS 2-4 ARE HT AREAS COUPLING THE 2 NCDAL HEAT
*= SINK MASSES (STEEL AND HEAT SINK) .
02 ????? LIMITING AREA FOR RADIATIVE HEAT EXCH ETNN UPPER ,
** PLENLM EQUIPMENT AND THE ADJACENT RV SHELL (PROP) 03 ????? ENTER HALF THE AREA CF THE S/G SHELL
** (8 AND W: ENTER C) 'I 1
A.1-22
===
1
-= . - t J
,a
.7 PARAMETER FlLE CONT.
'l 04 ????? ENTER HALF THE AREA OF THE S/G SHELL
[I.[ '
** (8 AND W: TOT AREA)
'q **
,P ** ITEMS 11-15 ARE " STEEL" (INTERNAL MASS) N%SSES 11 ????? STEEL NASS FOR CORE NODE: ENTER THE SLM OF THE
** N%SSES OF THE INSIDE HALF OF THE CORE BARREL (OTHER PART .
** IS IN DOWNCCNER NODE) AND THE SUPPORT PLATE 12' ????? MASS OF RV UPPER PLENLM INTERNAL STRUCTURE- '
- b. 13 ????? SUM OF HALF THE N%SS OF ONE S/G'S TUBES PLUS
. ** HALF THE LOWER HD PLUS THE N%SS OF THE HOT LG PIPE
}.*.
**-(8 AND W: ENTER N%SS OF HOT LG ONLY) 14 ????? SLM OF HALF THE MASS OF ONE S/G'S TUSES PLUS 3
** HALF THE LOWER HD PLUS THE MASS OF THE CLD LG BTNN THE
** S/G AND THE PMP SCWL (CE: INCL BOTH CLD LGS IN THE LOOP)
** (8 AND W: ENTER TOT S/G MASS ASSOCIATED WITH PRIM SIDE,
~
** IE TUBES +TUBESHEET+ PRIMARY HEADS, PLUS THE MASSES OF THE
** 2 CLD LGS UP TO THE PMP BOWL) !
,, 15 ????? ENTER THE TOT MASS OF THE HORIZ PART OF ALL CLD I
** LGS (IE THAT PART STWN THE RV AND THE PMPS) PLUS
- ** THE N%SS OF THE RV SHELL CONTAINED WITHIN THE DOWNCCMER
[ ** (lE THE MASS WHICH LIES BELOW THE NZL ELEV)
** PLUS THE MASS OF THE CORE EARREL NOT INCL IN NO11
** (DOWNCCMER NODE STEEL MASS)
** ITEMS 22-24 ARE THE " HEAT SINK" N%SSES
{l 22 ????? ENTER THE MASS OF THE RV SHELL NOT INCL AS PART
** OF THE DOWNCCMER (ITEM 15) IE APPROX. THE N%SS ABOVE THE l
.,- ** NZL ELEV (UPPER PLENLM NODE HEAT SINK MASS) 0 23 ????? STEEL MASS OF HALF THE S/G SHELL
** (8 AND W: ENTER 0) 1 24 ????? STEEL MASS OF HALF THE S/G SHELL (8 AND W:
** ENTER TOT MASS OF ONE S/G SHELL) ,
** ITEMS 33 AND 35 ARE THE AREAS ASSCCI ATED WITH ENERGY -l
-l
** EXCH BTNN STEEL N%SSES AND CONTMT Li 33 ????? ENTER 0 (8 AND W: ENTER SURF AREA OF
- ,J ** ONE HOT LEG)
., 35 ????? ENTER THE SURF AREA OF THE RV SHELL INCL AS
;d ** PART OF THE DOWNCCMER (SURF AREA OF RV SELOW THE ELEV i
jq
]' A.1-23 L- ____ _ __
I
PARAMETER FlLE CONT.5 ' ** WHERE THE DCWNCCMER STARTS, iE APPROX THAT SURF AREA
** WHICH IS BELCW THE NZL ELEV) ,
** ITEMS 42-44 ARE THE AREAS ASSOCIATED WITH HEAT
** SINK LOSSES TO CONTMT 42 ????? SURF AREA CF THE RV NOT INCL IN THE DCWNCCMER
** (ITEM 35) (IE APPROX THAT SURF AREA ABOVE THE NZL ELEV) 43 ????? SURF AREA CF HALF A S/G SHELL
** (8 AND W: ENTER 0) 44 ????? SURF AREA CF HALF A S/G SHELL (8 AND W:
** ENTER TOT AREA CF CNE SHELL)
** ITEMS 51-55 ARE THE AREAS ASSOCIATED WITH EXCH CF
** ENERGY STNN STEEL NCDES MD PRIMARY SYS GAS 51 ????? INT SURF AREA CF MASS IN ITEM 11 (RV NCDE:
** IE CCRE BARREL PLUS THE SUPPORT PLATE) 52 ????? INT SURF AREA CF MASS IN ITEM 12
** (UPPER PLENLM ECPT AREA) 53 ????? INT SURF AREA CF MASS IN ITEM 13
** (SRCKEN HOT LEG NODE) 54 ????? INT SURF AREA CF MASS IN ITEM 14
** (BROKEN COLD LEG NCDE) 55 ????? INT SURF AREA OF MASS IN ITEM 15
** (DOWNCCVER NODE)
** ITEM 82 iS USED TO CCMPUTE GAS VELOCITY IN THE
** UPPER PLENLM 82 ????? GAS FLOW AREA IN UPPER PLENLM
** ITEM 92 iS USED TO CCMPLrTE HEAT XFER COEFFS IN THE
** UPPER PLENLM 92 ????? HYDRAULIC DIAMETER CF UPPER PLENLM STRUCTVRE
** )
** SEDIMENTATlON AREA ASSCC1ATED WlTH STEEL MASSES IN
** EACH NCDE FOR GRAV St:. l i t. l NG l 101 ????? CORE NODE SEDIMENTATION AREA (IE APPROX f
; ** SUPPORT PLATE AREA) l 102 ????? UPPER PLENLM SEDIMENTATION AREA '
103 ????? BROKEN HOT LG NOCE SEDIMENTATION AREA ; i ! A.1-2k
.i
=
'd a
- b PARAMETER FILE CONT.
[ 104 ????? BROKEN CLD LG NODE SEDIMENTATION AREA
! 105 ????? DOWNCCMER SEDIMENTATION AREA (APPROX LOWER A ** HD PROJECTED AREA) 3
-. ** HEAT LOSS INPUTS:
- i. 111 ????? NOMlNAL OPERATION CONVECT 1VE (NOT THROUGH
** RADiACT!VE 1NSULATlON)
** HEAT LOSSES 112-????? No. OF PLATES IN RADlATIVE INSULATION
~. 7
** (16 MAX) ON PRI SYS
;, 113 ?????- No. OF PLATES IN RADIATIVE INSULATION
** (16. MAX) ON S/G'S
.l 114 ????? TOT THICKNESS OF RADIATIVE INSULATION
!$ ** IF DESIRE TABULAR CUTPUT AND OPERATOR INTERVENTIONS
.. ** IN BRITISH UNITS INSERT A *8R HERE (EG FOR A PARAM DLMP
. [j' ** IN BRITISH, OPERATOR INTERVENT1ONS, AND TABULAR CUTPLrr) '
*BR r-f e
'i 9
^e
[.I .
,- 1 ,
h i; 3.. 1
- !1 l L:
-l}
L., A.1-25
y I S2D U1MAAP I 1, PARAMETER, FILE 5C - 3-i . SEQUOYAH S2D SFN = U1MAAP 1 . 25-1 1 2,13,.1 / BREAK KEY: COLD LEG 2,14,.0- / COLD LEG BREAK 2" DIA 2,15,19 /BRK ELEV FRCM BOTTCM OF RV (FT) 7,38,0. . / CONTAINMENT FAILURE HOLE SIZE (FT**2) . 7,39,5. / DESIGN LEAKAGE'O.10" DIA . 0,0,0 . 1 0-O. 25.00 / RUNNING TO 25.00 HOURS 0.25 / RESTART AND TABULAR OUTPUT EVERY .25 HCURS 202 '/ BREAK IN COLD LEG 1 209 /PS BREAK FAILED i 1 216 /HPI PLMPS FORCED OFF 1 217 /LP1 PLMPS FORCED OFF 1 232 / CHARGING PLMPS FORCED OFF 1' 242 / MAKEUP SWITCH OFF 1 243 /LETDCWN SWlTCH OFF 1 227 / MANUAL SCRAM 1 0 8 /CPERATOR INTERVENTlCN
.0167 /AT 1. 0 MI NtJTE i A.2-1 i .
-- J
1 y . CON __I. e~1 - 32D U1MAAP
! [9 j i;J O i 215 AACP F#1TCH OFF
. ,a 1 s .V ,
o
' .r , 5 .,
' [f 4.51 / WATER LEVEL OF 4.51 TRIGGERS RECIRC 0
220 /RECIRC SNITCH: MAN ON
, {' ..
3
,.,- 0 a.
p i.
+
,",i l ( [j
. l_ ,
i ,
)
.l
\ ',
t, a 1 r :s l 1 1
- 6 vi: l
~
.=. ,
A.2-2 4 i
(
- S2H U2MAAP I L ;
PARAMETER,F1LE5C ,- E i SEQUOYAH S2H SFN=U2MAAP 1 _,. 25' ; 1 1 2,13,1 /8REAX KEY: COLD LEG 2,14,0.0218 /CCLD LEG BREAK 2" DIA 2,15,19. /BRK ELEV FORM BOTTCM OF RV (FT) 7,38,0.02 / CONTAINMENT FA1 LURE HOLE SIZE (FT=*2) i 7,39,5.114E-05 / DESIGN LEAKeGE 0.10" DIA 0,0,0 . o ' 0. 25.00 / FINAL TIME OF 25.00 HOURS
.25 / RESTART AND TABULAR OUTPLrf EVERY .25 HRS 242 / MAKEUP SWITCH OFF 1
243 / LETDOWN SWITCH OFF 1 202 /8REAK IN COLD LEG 1 209 /PS BREAK FAILED 1 227 / MANUAL SCRAM 1 0 8 /CPERATCR INTERVENTION
.O167 /AT 1.O MINUTE O
215 /MCP SWITCH OFF 1 0 5 4.51 / WATER LEVEL OF 4.51 FT TRIGGERS RECIRC 1 I A.2-3
- ! \
i .,I
-- - . = - . . . . . .
i d' f S2H U2MAAP CONT.
!1..l'T) '
O
- 216 /HPl FORCED OFF
.n 1 0 217 /LPl FORCED OFF 1
,F 232 / CHARGING PLMPS FORCED OFF 1
220 / REC lRC ON t 1 0 T* I r; i A' l 4 6 *
. i s .:
t.s a
" ,e f
1. a '. l l1
! l._I, 4
'q !.3 A.2 h I
/
S2HF U7MAAP(DRAlNS BLOCKED) . j PARAMETER,FILESC j 1 i SFN=U7MAAP: SEQUOYAH S2HF (DRAINS BLOCKED) j- 1 25 1 1 2,14,.0218 /PS BRK SIZE (2.0" DIA) 7,38,.02 /CONTMT FAILURE HOLE. SIZE (FT*=2) 7,39,5.114E-5 /NCMlNAL CONTMT LEAK HOLE SIZE (FT==2) 7,6,100. / CURS HEIGHT (FT); DRAINS BLOCKED 2,15,19 / BREAK ELEV FRCM BOTTOM.0F RV (FT) 2,13,1 / BREAK KEY; COLD LEG NODE O,o,0 /END OF PARAMETER CHANGE O 0. 35.00 / FINAL RUN TIME ( HRS } 1.5 / EDIT FREQUENCY ( HRS } , 242 A%KEUP SWlTCH OFF 1 243 / LETDOWN SWITCH OFF 1 202 / BREAK IN COLD L5G 1 209 /PS BREAK 1 , j 227 A%NUAL SCRAM l 1 0 8 / OPERATOR INTERVENTlON i
.0167 /AT 1.0 MIN 0 1 215 AACP TRIP AT .0167 HOURS l
'. 1 j O
j 5 / INTERVENTION WITH RWST WATER LVL I 1 s
*~
,":0,
{\
, 1
\
A.2-5
\ *
.i i i U'
CONT. l.5 S2HF U7MAAP(DRAINS BLOCKED) 4.51 /OF 4 51 FEET (RECIRC FAILS) i' [$f
,~ 0
' g7 222 / SPRAYS FORCED OFF
-: 3 216 /HPI FORCED OFF il 1 217 /LPI FORCED OFF
; 1 i .:! 232 / CHARGING PUMPS FORCED OFF 1
220 / REC 1RC SWITCH N%NUALLY TURNED OFF o] ' O ,
,, 0
..t 0 4
1; .
,{l i J il
- i. -'
.u ,
u I " e e
, eo
'h
- q u
i J A.2-6
']
l . . 1 S2HF U3MAAP (DRALN OPEN) .
-l ,! PARANIETER,F1LE5C ' .! 1 - ; SFN=U3MAAP: SEOUOYAH S2HF (DRAIN OPEN) j j 1 ! 25
- 1' -
2,13,1 /BREAX KEY; CLOD LEG NODE 2,14,.0218 /PS BRK SIZE (2.0" DIA) 2,15,19 /BREAX ELEV FROM BOTTCM OF RV (FT) 7,6,0. / CURB HEIGHT (FT); DRAIN'OPEN 7,38,.10 /CONTMT FAILURE HOLE SIZE . (FT=*2) 7,39,5.114E-5 /NCMINAL CONTMT LEAK HOLE S IZE (FT**2) - 0,0,O /END OF PARAMETER CHANGE O . O.
.,. 20.00 / FINAL RUN TI AE ( HRS )
1.5 / EDIT FREQUENCY ( HRS ) 242 / MAKEUP SWlTCH 6FF 1 243 / LETDOWN SWlTCH OFF 1 202 /5RK IN COLD LEG 1 209 /PS BREAK 1 227 / MANUAL SCRAM 1 ! O -
/ OPERATOR INTERVENTION S
.0167 /AT 1.0 MIN O
215 /MCP TRIP AT .0167 HOURS 1 0 5 /1 INTERVENTION WlTH RWST WATER LVL O I e I A.2-7 j
~ * * ' * -*--e . .-~as .s v w - ,-.~a,m. e- m -en ~ ~*>~~~eem-e_ .-~.m ., ,,
u f c. CONT. !
!W S2HF U3MAAP (DRAIN OPEN) l1 4.51 /OF 4.51 FEET (RECIRC FAILS) +
,d -
0 222 / SPRAYS FORCED OFF g,
! . .' 1 216 /HPI FORCED OFF l
1 217 /LPI FORCED OFF , 1 l .. 232 / CHARGING PLMPS FORCED OFF
~ . 1 7 220 / REC 1RC SWITCH MANUALLY RJRNED OFF 1 0 O
l ', . 0
. u
.)
H t i
.J r;.'
( l1 g d i I t
!J j A.2-8
- - - - - - - -- ---___.m____,___.___ ___ _ _
j TML.B ' U4MAAP a PARAMETER, FILE 5C 1 SFN=U4MMP : SEOUOYAH TM_B' (NON-ADIABATIC) 1
- 25. q
- J -
3 2,13,2 / BREAK KEY: ALL COLD LEGS 3" 2,14,.00158 / TOTAL SEAL LOCA AREA-(FT**2) 2,15,25. / BREAK ELEV FRCM BOTTOM OF RV (FT) } 7,3B,.02E0 /CONTMT HOLE SIZE (FT**2) .l 7,39,5.114E-5 / NOMINAL CONTMT LEAKAGE AREA (FT**2) , 18,35,1. /DISCH COEFF=1 SINCE WE WANT A GIVEN FLOW : O,0,O /END OF PARAMETER CHANGE m O O. / START TIME (HR) .. } - 30.0 /END TIME (HR) 1.0 / EDIT FREQUENCY (HR) " l' 205 /AC AND DC PCWER,AFW OFF AT T=0 1 3
/k%KEUP SWITCH OFF -j 242 1 ,
l 243 / LETDOWN SWITCH OFF J' ( 1 1 0 [. 8
.75 /1 INTERVENTION AT 0.75 HR ,
0 . J 209 /PS BREAKS; PUMP SEAL LOCA 1 O
, O
^l J
7 l I 6 l' A.2-9
'i W
L
a' jd' T23ML U5MAAP
,I- f] PARNAETER,FlLE5C 1 ; 1 i.. SEQUOYAH T23h4. SFN = V5hMAP d 1 l
25 ,
., 1
., j 7,38,0.02 / CON NT FAILURE HOLE SIZE (FT**2)
!. 7,39,5.114E-05 / DESIGN LEAKAGE 0.10" DIA 3,16,2364.7 / CHANGE PORV SETPOlNT 3,17,2364.7 / CHANGE PORV SETPOINT f.
i 0,0,0 0 . l... O. 20.0 / FINAL TIME OF 20.0 HRS
, .'; .25 / RESTART AND TABULAR OUTPUT EVERY .25 HRS
.i 224 / AUX FEED FORCED OFF ,
1 228 AAA!N FEEDWATER SHt/T OFF
- h. .
[t 216 /HP1 FORCED OFF
\ 1 217 /LPI FORCED OFF
- t. , 1 232 / CHARGING PLMPS FORCED OFF
-; 1 U 242 NAKEUP SWITCH OFF 1
['l 243 / LETDOWN SWlTCH CFF 1
- e. ; 227 / MANUAL SCRAM
- 1 0
l[j 2 /lNTERRUPT AND TRIP MCP'S 17.5 /WHEN CONTMT PRESS EXCEEDS 17.5 PSIA
'p 0 12
,2
~
5,
~
1 A.2-10 e
- . T23ML.U5MAAP .. CONT. 3, 215 / TRIP MCP'S J
1- - - 0 , u 5 4.51 / WATER LEVEL OF 4.51 FEET TRIGGERS RECIRC [ I! - ; 0 220 / REC 1RC FN1TCH MANUALLY ON 0 - ( 8 l
, i l
t I t 7 I
=e l
l .
~i A.2-11 !
. g
'~~~ ' ' " - ' -
'd n
D AD U6MAAP ) i i-jh PARAMETER, FILE 5C 1 'f g' SEQUOYAH AD SFN = U6MAAP
- 1 .
3 n 25
!; 1 1
y 2,13,1 /8REAK KEY: COLD LEG
.'i 2,14,.5454 / COLD LEG BREAK 10" DIA 2,15,19. /8REAK ELEV FRCM BOTTOM OF RV (FT)
~j 7,38,0.02 / CONTAINMENT FAILURE HOLE S12E (FT*=2}
7 ,39 ,5 .1 14E--05 / DESIGN LEAKAGE O.10" DIA o,0,O p':
. 0 0.
! 25.0 / FINAL TIME OF 25.0 HRS 0.25 / RESTART AND TABULAR CtJTPUT EVERY .25 HRS -
9, 202 / BREAK IN COLD LEG J 1 - 209 /PS BREAK FAILED
.p , 216 / SAFETY INJECTION FORCED OFF 1 3 1 'd 217 /RHR FORCED OFF 1 .
4.3- 232 /CHARG1NG PLMPS FORCED OFF
; .d 'n .
242 / MAKEUP SWlTCH OFF L 1 243 / LETDOWN SWlTCH OFF
!? 1 227 / MANUAL SCRAM I
Ii t, J 0 8 / OPERATOR INTERVENTION
.0167 /AT 1.0 MINUTE lh j ..,. !J c.
b') , A.2-12
1 AD.U6MAAP CONT.- v 55 j o 3 l l .i 215.- AACP SWITCH OFF Tl 1 ! 4 ! O i ./
.]
5 3 4.51 / WATER LEVEL CF 4.51 FT TRIGGERS RECIRC ' '
.', 0 i ^
220 / REC 1RC SWlTCH: MAN ON l 1 i f i 1* e l$ e l l l 4 4 4 I 4 I P l I i I l A.2-13 .! 1
,9 L I 1
L S2D U8MAAP in ij PARAMETER,FILESC . 1 1 ! S2D-FULL INJ RESTORATION @1.5 HRS FN=USMAAP ljq SEQUOYAH ]'
. 1 ,
25 , i' o 1 4,13,1 /8REAK KEY: COLD LEG 2,14,.0218 / COLD LEG BREAK 2.0 INCH DIA. l 2,15,19. /8RK ELEV FROM BOTTOM OF RV (FT) j ..-' 7,38,0.02 / CONTAINMENT FAILURE HOLE SIZE (FT**2) 7,39,5.114E-05 / DESIGN LEAKAGE O 10" D1A 7 0,0,0 2 0
- 0. / START TIME IS 0.0 HR fl 5.00 /END TIME IS 5.0 HRS 0.25 / PRINT OUTPUT AND RESTART F1LE EVERY .25 HR
.i 202 / BREAK IN COLD LEG 1
i 209 /EREAK IN PRIMARY SYSTEM j [ 1
/HP1 PUMPS FORCED OFF
~
216 1
*i 217 /LPI PLMPS FORCED OFF
[ 1 l l ' 232 / CHARGING PLMPS FORCED OFF i
$ l
- 242 / MAKEUP SWITCH OFF '
j .- , 3 243 /Ltiv0WN SWlTCH OFF 1
~"
227 / REACTOR MANUALLY SCRAMED e, 1
';. . ' O 8 s i] .0167 / OPERATOR INTERVENTION AT 1.O MINUTE ,] .
a 1 I Lc A.2-lh
==-.;=-- -- - ,.:._ .:. . . - - . . ,
4
-S2D'U8MAAP CONT._'!
p
.{ 0 l 215 AtCP TURNED OFF !
1 C 0 _f 5 ' 4.51 /RWST WTR LVL OF 4.51 FEET TRIGGERS RECIRC ; O 220 /RECIRC ON 1 0 8
'1.50 / OPERATOR INTERVENTION AT 1.50 HRS O
213 /LP1 SWlTCH MANUALLY ON , 1 216 /HP1 PLMPS FORCED JN O
- 217 /LP1 Pt.MPS FORCED ON .
0 ' 232 / CHARGING PluPS FORCED ON O 0 k 4 a I
?
I I, A.2-15
. j
-1
---------_ __ j
]
K S2D U9MAAP
!4 PARAMETER, FILE 5C -
, ta 3
n SEQUOYAH S2D-MINIMLM SAFEGUARDS FN=U9MAAP Cf
~
j 25
.] 1 1
. 2,13,'1 /8REAK KEY: COLD LEG
- ) 2,14,.0218 / COLD LEG BREAK 2.0 INCH DiA.
2,15,19. /8RK ELEV FRCM 80TTCM OF RV (FT)
/1 AIR RETURN FAN AVAILABLE FOR OPERATION
- 6,13,1.
6,18,1. ./1 HPl PUMP AVAILABLE FOR OPERATION 6,19,1. /1 LPl PLMP AVAILABLE FOR OPERATION
; 6,43,1. /1 CHARGING PUMP AVAILABLE FOR OPERATION 6,88,1. /1 SPRAY PLMP AVAILABLE FOR OPERATION E 7,38,O.02 / CONTAINMENT FAILURE HOLE SIZE (FT**2) 9 7,39,5.114E-05 / DESIGN LEAKAGE 0.10" Dl-A ,
a 0,0,0 E: 0
/ START TIME IS 0.0 HR 0.
20.0 /END TIME IS 20.0 HRS O.25 / PRINT OUTPUT AND RESTART FILE EVERY .25 HR 202 / BREAK IN COLD LEG
, 1 209 / BREAK IN PRIMARY SYSTEM 1
+ 216 /HPI PLMP FORCED OFF I
n [, 217 /LP1 PLMP FORCED OFF 1 232 / CHARGING PLMP FORCED OFF l 1 t 242 A%KEUP SWlTCH OFF 1 243 / LETDOWN SWITCH OFF 1
-l:
U i ia, A.2-16 )
. l
< - - -- - ------_---____-__a
u S2D U9MAAP CON _I._ 227 / REACTOR MANUALLY SCRAED 1, 0 B-
.0167 / OPERATOR INTERVENTION AT 1 O MINUTE. ];)
0 215 AtCP TURNED OFF 1 0 5 4 51
. /RWST WTR LVL OF 4.51 FEET TRIGGERS RECIRC l
0 220 /RECIRC ON 1 , 0 _ iP l
.Ii 1
*M 8 R' j
.a 1 i
-]
*1 A.2-17 l
I
9 L - 1
'[ S2D U10MAAP- i PARA ETER, FILE 5C l
! [1 -l
- ll.)
1 SEQUOYAH S2D-SECONDARY SIDE DEPRESS RHR ONLY FN=U1CWAP l q .. 3 25
. 0 1
l 2,13,1 / BREAK KEY: COLD LEG f
*- 2,14,.0218 / COLD LEG BREAK 2" DIA l 2,15,19. /BRK ELEV FROM BOTTOM OF RV (FT) 7,38,0.02 /CONTA1NMENT FAILURE HOLE SIZE (FT**2) 7,39,5.114E-05 / DESIGN LEAKAGE 0.10" DIA '
0,0,0 O O. / START TIME IS 0.0 HR [ 5.00 /END TIME IS 5.00 HRS H 0.25 / PRINT OLfrPUT AND RESTART FILE EVERY .25 HR 202 / BREAK IN COLD LEG Il! ' 1 209 / BREAK IN PRIN%RY SYSTEM i 1 216 /HPI PLMPS FORCED OFF 1 F 213 /LP1 SWITCH MANUALLY ON
)
" 1 217 /LPI PLMPS FORCED ON s 0 232 . / CHARGING Pt.MPS FORCED OFF
'l 1 242 /h%KEUP SWITCH OFF 1
, 243 / LETDOWN SWITCH OFF 1
o 227 / REACTOR MANUALLY SCRAED
a 3 233 /S.G. PORV OPENED MANUALLY I3) i
.)
.J
\, A.2-18 j] ,
----------_-._______________d
-=m===. - - - - - - S2D U10MAAP CONT.- " J:
.l. . 1 O .
i 8 -
.0167 / OPERATOR ! INTERVENTION AT 1.O MINUT 0
215 /)ACP RJRNED OFF 1 . O 5 4.51 /RWST WTR LVL OF 4.51 FEET TRIGGERS RECIRC 0 220 /RECIRC ON , 4 0 i 1 e a 1 f l i 4 1
, c j
A.2-19 i
- ----_a
- =
y - i= S2H U11MAAP ii I j i3 PARAMETER,FILESC , d-3
'a SEQUOYAH S2H-MiNIMJM SAFEGUARDS FN=U11MAAP j 1 25
- 0 1
e- 2,13,1 / BREAK KEY: COLD LEG
. '.j 2,14,0.0218 / COLD LEG BREAK 2" DIA
~
2,15,19. /BRK ELEV ABOVE BOTTCM OF RV (FT)
,? 6,13,1. /1 AIR RETURN FAN AVAILABLE FOR OPERATION 6,18,1. /1 HPl PLMP AVAILABLE FOR OPERATION
..; 6,19,1. /1 LP1 PUMP AVAILABLE FOR OPERATION 6,43,1. /1 CHARGING PLMP AVAILABLE FOR OPERATION 6,88,1. /1 SPRAY PUMP AVA1LABLE FOR OPERATION ,
1 7,38,0.02 / CONTAINMENT FAILURE HOLE SIZE (FT==2) 7,39,5.114E-05
/ DESIGN LEAKAGE O.10" DIA
. 0,0,0 . .
.)
0 C. - l
': 25.0 /ONLY RUNNING TO 25.'O HOURS .:
.25 / RESTART AND TABULAR OUTPUT EVERY .25 HRS 242 A%KEUP SWITCH OFF !
F! 1 .
. 243 / LETDOWN SWlTCH OFF l T 1 J 202 / BREAK IN COLD LEG 1
p. l 209 /PS BREAK FAILED 1 l 227 A%NUAL SCRAM 1 i
,., O
~
8 / OPERATOR INTERVENTION l
.0167 /AT 1.0 MINUTF l n 0 1 j j i
5 A.2-20 {.]
a S2H U11MAAP. CONT. ,_ 1 215 /MCP SWITCH OFF
~
1 . 0 .t. : 5 # 4.51 / WATER. LEVEL OF 4.51 FEET TRIGGERS REClRC .i n 216 /HPI FORCED OFF ' 1 217 /LPI FORCED OFF , 1. 232 / CHARGING PLMPS . FORCED OFF 220 /RECIRC ON 1 . 0 l
' A.2-21
~
1 k -
c- --
.. - z- - - - -
=
L j[ S2H U12MAAP
!] '2 -
PARAMETER, FILE 5C 1
.- SE0VOYAH S2H PARTI AL REC lRC RESTORATION @2.5 HRS FN=U12N'AAP n
25 0 . 1
., 2,13,1 /8REAK KEY: COLD LEG 2,14,0.0218 / COLD LEG BREAK 2.0" DIA.
i 2,15,19. /8RK ELEV ABOVE BOTTCM OF RV (FT) f,- 6,18,1. /1 HPI AVAlLABLE J 6,19,1. /1 LPI AVAILABLE 6,43,1. /1 CHARGING AVAILABLE 7,38,0.02 /CONTA1NMENT FAILURE HOLE SIZE (FT*=2)
. 7,39,5.114E-05 / DESIGN LEAKAGE 0.10" DIA
' O,0,0 n
- 0. / START TI AE OF 0.0 HR
'I "
5.00
.25
/END TINE OF 5.0 HRS
/ RESTART AND TABULAR OUTPUT EVERY .25 HRS
' 242 /MA)<EUP F# ITCH OFF i: 1 243 / LETDOWN F#1TCH OFF 1
202 / BREAK IN COLD LEG r, j
,b 209 /PS BREAK FAILED 1
,. l 227 A%NUAL SCRAM 1 )
1 0
.8 -
j
.0167 /CPERATOR lbriERVENTION AT 1.0 MINUTE )
. 0
~
215 ACP FNITCH OFF
;) 1 J
i, A.2-22 1
CONT. - j S2H U12MAAP -
\ .
. O.
5 ;-l i- 4.51 / WATER LEVEL OF 4.51 FEET TRIGGERS RECIRC 0 216 /HP1. FORCED OFF 1 217 /LPI FORCED OFF 1 232 / CHARGING Pt.MPS FORCED OFF 1
/ REC 1RC SWITCH MANUALLY ON 220 1 ,
0 8 2.5 /RECIRC REGAINED @ 2.5 HRS 0 213 /LPI SWlTCH MANUALLY ON 1 216 /HPI FORCED ON 0 217 /LPI FORCED ON 0 232 / CHARGING FORCED ON 0 O l
/
9 4 j A.2-23 f i l l 1 ________________-______a
-t c t t"
T
- TMLB' U21HMAP ,
!il "** PARAETER ,F iLE5C ,
1 ; c) .
', SEQUOYAH TMLB-POWER RESTORATION AT 2.5 HOURS FN"J21HMAP 1
.' 25 0
1
~2,13,2 /8REAK KEY: ALL COLD LEGS 2,14,.00158 / TOTAL SEAL LOCA AREA (FT*=2) 2,15,25. /8REAK ELEV FRCM BOTTOM OF RV (FT) 7,38,.02EO /CONWT HOLE SIZE (FT**2) 7,39,f 114E-5 /NCMINAL CONTMT LEAKAGE AREA (FT*=2) j 18,35,1. /DISCH COEFF=1 SINCE WE WANT A GIVEN. FLOW ]
O,0,0 /END OF PARAETER CHANGE i l
. 0 5 0, / START TIME IS 0.0 HR 15.00 /END TIME IS 15.0 HRS I.
J 0.250 /PP1MT OUTPUT AND RESTART F1LE EVERY .25 HR 205 / POWER NOT AVAILABLE 1 242 /MAXEUP SWITCH OFF . 1 243 / LETDOWN SWlTCH OFF 1
': O d 8
.75
/ OPERATOR INTERVENTION AT .75 HR 0
200 /PLMP SEAL LOCA IN COLD LEG ) 1 209 /8REAK IN PRIMARY SYSTEM
- 1
.j.s 0
g 1 2.50 / OPERATOR INTERVENTION AT 2.50 HRS I I'.J 0 H u
*i A.2-24
L TMLB' U21HMAP CONT." ii ~ 'i 205 / POWER RESTORED AT 2.30 HRS O
! 213 /LPl SWITCH TURNED ON 'j - 1 _.
0 1 5 ,. 4.51 /RWST WTR LVL OF 4.51 FEET TRIGGERS-RECIRC 0
- l. 216 /HP1 PUMPS FORCED ON O
217 /LP1 PLMP' FORCED ON O - 232 / CHARGING PLMP FORCED ON O j 220 / REC lRC ON 1
, 213 /LPI SWlTCH MANUALLY ON 1 ,
0 t S J I j A.2-25 .
~ - ..
L l
! c! TMLB' U22HMAP i
\
I bf PANAETER ,F I LESC .
,a' 3
SEQUOYAH TELS-POWER RESTORATION 5.0 HOURS FN=F22HMAP i '. p- <
- . 1 j
l .. 25 0 4 1 l - 2,13,2- / BREAK KEY: ALL COLD LEGS ) i 2,14,.00158 /TOTAf. SEAL LOCA AREA (FT**2) 2,15,25. /8REAK ELEV FROM 80TTOM OF RV (FT) 7,38,.02E0 /CONTMT HOLE S12E (FT**2) l~~ 7,39,5.114E-5 / NOM 1NAL CONTMT LEAKAGE AREA (FT=*2) . l
' 18,35,1. /D1SCH COEFF=1 SINCE WE WANT A G1VEN FLOW
. .I 0,0,0 /END OF PARAMETER CHANGE y 0
- 0. / START TIE IS 0.0 HR f l; -
15.00 /END TIE IS 15.0 HRS 0.250 / PRINT OUTPUT AND RESTART FILE EVERY .25 HR
* !j 205 / POWER NOT AVAILABLE 1
l 242 / MAKEUP SWITCH OFF 1 243 / LETDOWN SWlTCH OFF 1 0
' . l l; 8 -
" /0PERATOR INTERVENTION AT .75 HR
.75
- ~ 0 200 /PWF SEAL LOCA IN COLD LEG 1
209 / BREAK IN PRIMARY SYSTEM 1 . 0
.; 8 5.0 /0PERATOR INTERVENTION AT 5.0 HRS
- 1
'3 ..
0
~ !] A.2-26
___..m -
r J l TML8' U22HMAP CONT . - l di 205 / POWER RESTORED AT 5.0 HRS _, j 0 E
~~
213 API S# ITCH TURNED ON 1 ,. O ;..; >
- i. 5 4.51 /RWST WTR LVL OF 4.51 FEET TRIGGERS RECiRC 0
216 /HP1 PaiPS FORCED ON 0 217 API PaFS FORCED ON _, 232 / CHARGING Pa9 FORCED ON O -- 220 /REClRC ON 1 213 api 5# ITCH AMNUALLY ON ' 1 0 - - - i
~
e l y,,,,. , ec e # g
. . A.2-27 -.
~
s., i ll_ m# T23ML U16MAAP' u- - R PARAMETER,FlLE5C FN=U16MAAP [,'2 EQUOYAH T23Nt.-BLEED AND FEED PER WESTINGHOUSE 1 25
.j 1 1
4 7,38,0.02 / CONTAINMENT FAILURE HOLE-SIZE (FT**2) 7,39,5.114E-05 / DESIGN LEAKAGE O.10" DlA , [, 3,16,2364.7 / CHANGE PORV SETPOINT
+
. 3,17,2364.7- / CHANGE-PORV SETPOINT 0,0,0 !
i
,, O O. / START TIME IS 0.0 HR e 12. /END TIME IS 12.0 HRS <
.25 / PRINT OUTPUT AND RESTART FlLE EVERY .25-HR 224'
/ AUX FEED WATER' FORCED OFF l
i 1 228 / MAIN FEED WATER SHUT OFF 1 216 /HPl PLMPS FORCED OFF 1
, 217 /LPl PLMPS FORCED OFF
- . j l
232 / CHARGING PLMPS FORCED OFF lj 1 242 / MAKEUP SWITCH OFF I 1 . 243 / LETDOWN SWITCH OFF
., 1 j 227 / REACTOR SCRAM l- .
0
' S / OPERATOR INTERVENTlCN
., O.6944 /AT 0.6944 HR(2500 SEC)
! O i!
, i -> lX
" A.2-28 i _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ _ _ _ _ _
i m
! T23ML U16MAAP CONT.;
l 210 /PZ PORV OPENED -' l
! 1
^
232 / CHARGING PLMPS FORCED ON 0 ., 216 /HPl PUAPS FORCED ON O 213 /LPl SWlTCH WNUALLY ON 1 217 /LPI PLMPS FORCED ON , 0 ' O 2 /INTERVENT1ON WITH CONTA1NMENT PRESSURE 17.5 /OF 17.5 PSIA - 0 215 ACP TURNED OFF 1 0 5 / INTERVENTION WlTH RWST WATER LEVEL
.. 4.51 / WATER LEVEL OF 4.51 FEET TRIGGERS RECIRC 0
220 /RECIRC ON 1 0 ee k A.2-29 2 l ,
_ . = - - - - l l"- l T23ML 'U17MAAP
$P
[ I. PARAMETER,F1LE5C 1
! SECUOYAH T23ML-FEED AND BLEED PER WESTINGHOUSE FN=U17NRAP r, 25
.., 1 1 l 7,38,0.02 /CONTA1NMENT FAILURE HOLE S1ZE (FT**2)
~
7,39,5.114E-05 -/ DES.IGN LEAKAGE O 10" DIA n 3,16,2364.7 / CHANGE PORV SETPOINT 3,17,2364.7 / CHANGE PORV SETPOINT 6,42,2382. / CHANGE CHARGING PLMP SET POINT 0,0,0 O c- 0. / START TIME IS 0.0 HR b 20.0 /END TIME IS 20.0 HRS
.25 / PRINT OUTPUT AND RESTART EVERY .25 HR
'I 224 / AUX FEED WATER FORCED OFF
.J $
<- 228 AAAIN FEED WATER SHUT OFF
.. 1 216 /HPI PLMPS FORCED OFF i i U 217 /LPI PLMPS FORCED OFF 1
q: 232 / CHARGING PLMPS FORCED OFF 1 fl - 242 /h%KEUP SWITCH OFF 1
- 243 / LETDOWN SWlTCH OFF
; 1 227 / REACTOR SCRAM l 1
' oi 0
;- 8 / OPERATOR INTERVENTION
] 0.8333 /AT 0.8333 HR(3000 SEC)
; n d
m 0
. A.2 -30
i-T23ML U17MAAP CONT.m r' o 7; 231 /CHARGlNG PLMP SWlTCH: MANUALLY ON
. !'.) '.;! 232 / CHARGING PUMPS FORCED ON -:
0 , 213 /LP1 SWITCH MANUALLY ON 1 217 /LP1 PUMPS FORCED O!4 0 0 2 / INTERVENTION WlTH CONTAINMENT PRESSURE 17.5 /OF 17.5 PSIA 0 215 AACP WRNED OFF 1 i 0 . 5 /1 INTERVENTION WITH RW5T WATER LEVEL. 4.51 / WATER LEVEL CF 4.51 FEET TRIGGERS RECIRC 0 216 /HPI PUMPS FCRCED ON , 0 217 /LPl PLMPS FORCED ON O 232 / CHARGING PLMPS FORCED ON ! O I 220 / RECIRCULATION ON 1 i 0 1 . I' 5 ! b
- A.2-31 I .
~ ' ^ ~ ~ ' ~ ~
I]
. }.
AD U19MAAP TU ~ tu PARAMETER, FILE 5C i 1 SEQUOYAH 10" AD-MINIhAM SAFEGAURDS FN=U19N%AP {3.. 3 7 25
! .> 0 1
i .- 2,13,1 / BREAK KEY: COLD LEG
. 2,14.,.5454 / COLD LEG BREAK 10" DIA r, 2,15,19. / BREAK ELEV FROM BOTTCM OF RV (FT)
W 7,38,0.02 / CONTAINMENT FAILURE HOLE SIZE (FT**2) 7 ,39,5.1 14E--05 / DESIGN LEAKAGE 0.10" DIA [ 6,13,1. /1 AIR RERJRN FAN AVAILABLE FOR OPERATION 6,18,1. /1 HPl PUMP AVAILABLE FOR OPERATION m 6,19,1. /1 LP1 PUMP AVAILABLE FOR OPERATION [] 6,43,1. /1 CHARGING PLMP AVA1LABLE FOR OPERATION l 6,88,1. /1 SPRAY PLMP AVAILABLE FOR OPERATION
'il 0,0,0 a
o q 0. '
.j 10.0 / FINAL TIME OF 10.0 HRS l 0.25 / RESTART AND TABULAR-OUTPLTT EVERY .25 HRS
[] 202 / BREAK IN COLD LEG
- t. .
3
,p 209 '
/PS BREAK FAILED
'j 1 216 / SAFETY INJECTION FORCED OFF l'"i 1 217 /RHR FORCED OFF
;;., 1
; 232 / CHARGING PLMPS FORCED OFF 1
R 242 /hAKEUP SWlTCH OFF 4 .- 243 / LETDOWN SWITCH OFF , L 1 c: 4 l' A.2-32 i
i .[ f AD U19MAAP CONT.s
.j , ~; 227 / MANUAL SCRAM ll 1 l 0 ) '
8 / OPERATOR INTERVENT1ON .
.0167 /AT 1.0 MINUTE -
. 0 215 AACP SWITCH OFF
. 1 0
5 4.51 AVATER LEVEL OF 4.51 FEET TRIGGERS RECIRC 0 220 / REC lRC SWITCH MAN ON
.1 0 [-
O g 4 I i
/
k k i
~
A.2-33 l ii _ e
~ ~ ~ ' ' ~ ~ ' ' ' ~ ~ " ' ' ' ~ ' ~ ~ ~ ~ ' ' ' ~ ~ ~ ~ - ~ ~ ' ~ ~ ~ ~ ~~
,.ra . n- .n, 1
[_ G . N AD U20MAAP p ; d . PARAMETER, FILE 5C 1 1 SEQUOYAH 10"'AD-FULL INJ RESTORATION @1.1 HRS' FN=U2CMAAP ;
' I 3
-25 1 0 1
2,13,1 / BREAK KEY: COLD LEG 2,14,.5454 / COLD LEG BREAK 10" DIA 2,15,19. / BREAK ELEV FROM BOTTCM OF RV (FT)
,, 7,38,0.02 '/ CONTAINMENT FAILURE HOLE SIZE (FT**2) 7,39,5.114E-05' / DESIGN LEAKAGE 0.10" DIA i 0,0,0
- 0
-. O.
,3 10.0 / FINAL TIME OF 10.0 HRS O.25 / RESTART AND TABULAR. OUTPUT EVERY 0.25-HRS
'? 202 / BREAK IN COLD LEG j
'.'a
,. 209 /PS BREAK FAILED 1
216 / SAFETY INJ FORCED OFF i 1 217 /RHR FORCED OFF 1 i 232 / CHARGING PUMPS FORCED OFF l 1
. [3, 242 /hMKEUP SWlTCH OFF 1
i.. 243 /Lt.IvOWN SWlTCH OFF 1
~
227 / MANUAL SCRAM (' 1
.I
-- 0 j I
B / OPERATOR INTERVENTION l ! .0167 /AT 1.0 MINUTE a 3 "e .
. i b) i :n M3 A.2-3h i
._______________________________u
j ji
- AD U20MAAP i CONT.a '"
! O l 215 /MCP SWlTCHED OFF ;~ 1 '.a , 0 5 1,; 4.51 / WATER LEVEL OF 4.51 FEET TRIGGERS RECIRC O .si 220 /RECiRC SW!TCH: MAN ON , 1 0 8 1.10 / OPERATOR INTERVENTION AT 1.10 HRS 0 213 1 216 /HPI FORCED ON 0 - 217 /LP1 FORCED ON O 232 / CHARGING FCRCED ON O O - l e A.2-35 ~ i t- _.
'a__
l
..)
1 il
' *'l e ,rt .
L, 8' eT. ! "Ti APPENDIX B l I ( .' $ 1 w
.)
I i 4 l l e s I l , 1 1 l
...A l1 l'.
q*e S 9
~
1 4
.h
< . s .
.J i
4
- i. .i
.9 t
i t 4 a I ! g"J l
, .I .
-..I m ,
l o.; B.1 Accident Secuence Description Ij g. -The selection of' accident sequences to be analyzed for Sequoyah core
; pj <
1 . damage assessments have been made from a list of accidents generated in i , v.
' :j IDCOR Subtask 3.2. A description of the perceived dominant core damage $ accidents is provided in this section. Not all of these sequences have J
been analyzed in ' detail for our report. Notably, the "V" sequence has.
;- not been examined. This is due, in part, to the reduction of the r probability of this initiator via increased plant surveillance and testing following the initial probability contribution determination.
The V-sequence is analyzed in reference 2.3. Other sequences which are
~
l
' '1, merely variations in break size have been studied qualitatively as
,,- reported in Chapter 8. It is not anticipated that these sequences will 1
~'
result in more severe containment challenge than those studied in
, j' , detail. .' u p
L The following tables are excerpted from the draf t report for IDCOR subcask 3.2 (table 3-1 of that report). l' I l l l# l
!- i ; (2 I
q ' U
' i t, e= 4
'"1 .
t l*
) 3' 4 i
4 j j , i
+ B.1-1 b
H, 1. . . e A _ _ _ _ _ . _ _ _ _ _ _ _ - _ . _ _ _ _ _ . .
7
- .1
- g. !
(- Sequoyah Key Sequences - { $ to I
, Sequence: 5D2 a
Core Damage Contribution: 11% (RS'SMAP) Ti , l Sequence
Description:
S Initiated by small LOCA (S 2 ) (1/2" to 2"), followed by failure of ECCS in i injection (D). All containment EST would operate as designed but failure of , l
. ECCS would lead to boil-off of water; uncovering core; resulting in core melt. If containment doc s not fail due to vessel steam explosion, the pressure is expected to exceed the failure pressure due to gases generated during accident and combustion of hydrogen produced in the accident. The ECCS has 5 subsystems. The three which are required for S2 LOCA ares ,
- 1) upper head injection accumulators (UHI), 2) safety injection pumps, .and
- 3) centrifugal charging pumps. The other two ECCS subsys tems are assumed available but not ef f active for an S2 LOCA. They are cold leg injection accumulators and LPIS (using RHR pumps) to inject borated water. .;
1
.Svstem States:-
o Small LOCA (1/2" to 2") o ECCS unavailable. o ECCS Failure Criteria
- failure of UHI, or
- failure to deliver flow equivalent to one centrifugal charging pump (1/2); or
- failure to deliver flow equivalent to one safety injection pump (1/2) [
o Containment systems available at core melt e B.1-2 N u
c - .== :: w, :: -
. ..N .
] -
1 F , Sequoyah Way Sequences . _ d' p.
,; u.
1 Sequence: 52D- (, Q- . .' a Core Damage Contribution':' 11% ' (RS'SMAP) o .T y._ Sequence
Description:
f Initiated by small LOCA.($2 ) (1/2" to 2"),- followed by f ailure of ECCS'in
*i
-injection (D). All containment ESF would operate as ' designed but failure of
. gj ECCS would lead to boil-off of water; uncovering core; resulting in core melt. If . containment does not fail due to vessel steam explosion, the pressure is expected te exceed the failure pressure fue to ' gases. generated' ,
during accident and combustion of hydrogen produced in the accident. The
~
ECCS has 5 subsystems. The three which are . required for. 52 LOCA are: ,, --) il
- 1) . upper head injection' accumulators (UHI), 2) safety injection pumps, and
~
. 3) centrifugal charging gumps. The ot.her two ECCS subsys tems are assumed Il
.Ji available but not effective for an 52 LOCA. They are cold leg injection
, _7 t secunulators and LPIS (using RHR pumps) to inject borated water. -l 1
.i
- System States:-
i o Small LOCA (1/2" to 2") o ECCS unavailable. o ECCS Failure Criteria .j
- failure of UHI, or
- failure. to deliver flow equivalent to one centrifugal charging pump (1/2);.or 4-
- failure to deliver flow equivalent to one safety injection pump U/2) 9 1 o Containment systems available at core melt e
B.1-2 M
< . l:
T t,. I 5H
]k)J Sequence: 2 lN Core Damage Contribution: 30% (RSSMAP)
_j ! Sequence
Description:
Initiated by small LOCA (S2 ) (1/2" to ~2"), followed by failure of ECRS
> (emergency coolant recirculation system) (H). High pressure recirculation r; system (HPRS)' provides for recirculation of coolant in RCS following small-
'~
; rupture. The HPRCS' circulates water from the sump to the core, out of pipe break, back to the suu . Output of RHR pumps is fed to safety injection and s ..
charging pumps for teactor vessel entry at the high , pressure needed1 for this
,_' type of accident. . The ECR keeps core covered and removes heat from sump
,--' water through RRR-heat exchangers.. If containment does not fail due to vessel steam explosion, the pressure is expected to exceed the failure ,
pressure due to gas generated dur ng ' accident and combustion of hydrogen i produced in accident. System States: i o Small LOCA (1/2" to 2") d.' o RER pumps /high pressure injection pumps not available.
, V; o Assumes ECI success ful. ,
10 F, o Containment systems available at time of core melt. )
;p , ,-
a l
/ 1 l
,in h m.
a
- p , j
-j B.1-3 l ts I I. -
- - _ - _ . __ - -. I
- - w = = .;. = .-.::;; ---
m
; J l
!- Ssqueness Sggy Core Damage Contribution: 9.5% (RSSMAP)
, M Sequence Description Initiated by small LOCA (S 'Y 2 ) (1/2" to 2"), followed. by failure of emergency core cooling recirculation system (ECRS) (H) and containment spray Yi recirculation system (CSRS) (F). Failures of ECRS and CSRS dominated by common-mode contributor. Between upper and lower containment compartments .
.I are two drains which must be closed during refueling operations. Failure to- Vi reopen these drains would result in failure of both ECRS and CSRS which draw }'
water from sump in lower containment. The closure of the drains would cause all water to be retained in the upper containment compartment. Results in , core being uncovered and pump cavitation of CSRS. Containment is postulated to. fail due . to overpressurization or hydrogen burning. -
~
System States: o ECRS unavailable o CSRS unavailable
- Failure assumed to be flowrate less than the equivalent of one . . containment spray pump and one RHR spray pump for ' first 24 hours or normal output of one RHR spray pump thereaf ter.
o All other systems assumed available B.1-4 . i _ _ - _ - - - - _ - - _ - - _ _ - _ _ _ i
l l l._ V
-)
O] Sequences T133 MLB13' (TMLB' equivalent)
'f lT. Core Damage Contribution:- 0.7% (RSSMAP) , , '. R l , O Sequence
Description:
Initiated by loss of offsite power or loss of network load (T1 ), followed by loss of emergency AC power 3(3 ), which results in PCS failurt (M),
- failure of auxiliary feedwater system (AFWS), and failure to restore offsite Power or emergency AC power within a period of about I hour to 3 hours. Core I
-' boils dry at high pressure.
- 1 System States:
1 m o Of fsite power unavailable o Onsite AC power unavailable o PCS unavailable o AFWS unavailable
,.s
'o DC power assumed available r
.h" l'*,
-) ]
P f
\e*e i In a
3.1-5 l f' .
, r1 ...
j, m _ __-_m - _ - _ _ _ _ ._ _ _ _ __m___m__m___ _
S:qu:nca: T 23ML 7_ Core Damage Contribution: 5.3% (RSSMAP) t J.
]- Sequence
Description:
! Initiated by transient except loss of offsite power (LOOP) involving afstomatic trip plus main feedwater interruption (T2) r transient involving automatic trip but no main feedwater interruption (T 3). Followed by failure to recover PCS or PCS hardware faults and other system failures for I
; T2 or only PCS hardware and other system failures for T3 (both are
. referred to as M); followed by failure of secondary system relief (SSR) and auxiliary feedwater system ( AFWS) (L). These result in loss of normal and emergency means of supplying water to the steam generators, leading to their l
boiling dry. The resultant increase in RCS pressure wou'd lead to pressure relief of steam through the safety and relief valves. Water would boil of f l 1 from core leading to eventual core melt. Containment failure predicted to be combustion of hydrogsn generated during accident. r .
- t i
Svstem States: o PCS unavailable . 1
; o AFWS unavailable
- Dominated by DC power failure 1
o AC power assumed available ) i 1 B.1-6
. ..,.._a.. 2 2._,._ _ .
".t
! 1..;,
..A l
l ) E0 s Sequence: V 1 i- : li i 3. Core Damage Contribution: 8% (RSSMAP) l .1u i.; i3 l
< r',
i.j Sequence
Description:
l _. Initiated by rupture of two check valves (in series) in one of LPIS lines, 1 allowing high pressure reactor coolant to enter and rupture the lo > pressure piping outside containment. The containment ESF would be ineffective for t this accident, and LPIS would ' also fail due to LOCA. As a result, coee melt would occur.
?
- w. Svstem States:
' e .*
o Containment bypassed 1 7 o LPIS and LPRS unavailable ,
, fI$ o All other systems assumed available S
s' 9 p.
. L'
;. y e s )
. ~
i:
'h
'.b l
, . J* .
s 'N B.1-7 d1 pF 3 I
u l L S:quencos: SRI , Core Damage Contribution: 23% (RSSMAP) [ 4 f
- 4
]n '
j Sequence
Description:
lnitiated by a small' LOCA (SI )' (2" to 6") which is followed by f ailure of
- the.high pressure or low pressure recirculation system (H) causes loss of core make-up capability in the recirculation phase. All power. is available.
Containment ' systems available. Containment can fail due to a steam explosion (early) or overpressure (later). 70% of the estimated frequency represents operator error. 67% of the operator error frequency derives from events occuring at 24 hours. 33% of the operator error frequency derives from L' events occuring at the end of injection. l. I 1 System States: o HPR/LPR - unavailable l l o Containment systems available at time of core melt t 4 4 1 f' B.1-8 l l
lO
}' IS .
Sequen'ce s s SD t I il
., Core Damage Contribution: 7% (RSSMAP) !($
- 1 q{
'ij Sequence
Description:
,, n Initiated by a small LOCA (SI ) (2" to 6") followed by failure of ECCS in injection-(D). ECCS criteria for injection require one safety injection pump and one charging pump. Failure to provide make-up results in rapid ,i uncovering of the core. RSSMAP assumed core melt would start in 5 minutes.
j Containment failure is predicted by vessel steam explosion (20 minutes), hydrogen burning (1 hr) or overpressure (6 hours). Containment systems are i available at the start of core melt. i '. t.
' ~
- System States:
- o HPI unavailable o LPI available, but ine f fective 4
o AC power available o all containment systems available
. (.
h. e 4 _8
> =
6 e 6
!J l
N
,l$.a B.1-9 i f1 -
1JJ . w-
4 4 3 3 3 5 5 0 0 , y 1 1 6 c - n x x 0 % e 4 1 % 7 u 5 3 - 2 9 q 9 9 x- 3 e r 1 1 6 F y r o 5 g 3 5 5 - e - - - 0
. S t 0 0 0 1 E a 1 1 1 % %
C C 7 x 8 N x x x 8 8 E e 3 3 U s 1 3 3 . O a 1 s E e 2 2 2 e S l u e .l H R 7 a A 1 v Y 5 5 5 - t0 - - - 8P t 0 0 0 % A Q 1 1 1 % 0 eH E 0 0 gS S 2 x x x 0 1 aS pR 1 1 1 1 e 1 7 8 yh
- 9 - - d t .
1 - 0 0 u e
. 7 7 0 1 1 t f d B - - 1 s oa 0 0 x x % m E.
1 1 x % 0 P t l 2 3 4 8 A ne B x x 4 H eb A 1 1 3 S m T 2 2 3 S es R s n r o os
. di 5 6 3 nr y 5 5 5 - - 1 eap c - - - 0 0 -
n 0 0 0 % 1 1 7 em ee 1 1 1 5 t o gu x x % euc ea q m x x 4 3 l t r me 6 3 4 9 bi e oar 7 6 7 at v T si CDF 1 3 nt 5 5 3 d oa ncl a e t r 5 o
- ny l 8 l a s n
) t eeo P o s l o A ) T e bd t M t s c s a a S e ef n e T ah S S c os e cl l s t t R n e u n a a e ma
( l e% c q e nt c odd a u n e u oon r e s i qs e S qi t e f sd e t eau et u i n s* c i S q y Si% q dh e t s en n ) e e d e et m l e r e yI yt S K yd sS t m uc ou e( e e eA a af o
- s n C q K KS y l r y l oc e e e s e a gs e u e n D c er K
- 2. i n eK l
R u q t S n a e 2 e 3 n c 2. al 3iP i o i S g S g 3 d nP c l a us s P e c A S ng e t A t ueA i kiM kl uH e
.- M S y i a mm oa k u s q a e s nS aI S i
d d s cqS a n eS sh t ATI S e DD TS A TI S R
- RK T( R u_ .
; i l
'9 .** wee >w, me -...e. - -- ,,_ _
- )w.
.h .,
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e
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e 4 o og
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. . a a
. APPEND X C 0 e
4 0
-e, I
, 'y a
S 6 O e 5 e al 9
*=d a
n
.1 T
t 8 i b j (.
; e-
.a
;a J C.0 Accident Signatures r]t a
Accident signatures are presented for each of the base cases l analyzed in this report. These signatures are generated directly from the MAAP plot files using programs developed at TVA. Figures are arranged according to case number as described below. In all figures, the left axis is used in conjunction with the I. solid curve whereas the right axis, if present, is used with the dashed curve. An etcempt has been made to group multiple plots on each plate to show transient interlationships between i variables of interest. Cases are identified as described in the 3 report body. t Case 1 - SD 2 (U1MAAP) Case 2 - SH 2 (U2MAAP) Case 3 -- S HF 2 (U3MAPP)* S2HF (U7MAAP)** Case 4 - TMLB' (U4MAAP) { '; Case 5 - T 23 E (USMAAF) is Case 6 -- AD (U6MAAP)
- Drains open
** Drains blocked
*k 2
Ij C.0-1 c3 161 i li (*
' c1
- - - , = -. = = = - . = = - - - - - - - - _ _ _ _ _ _ - - - - - - y S2D U1MAAP q 1 t 4 _4
-[:. .
~
- r l< J f
I *;
- h. .
2. e t o: I' o- N % N E.i . \ m , s l , d R)- ,
- e. :
a - g gj gg ($ 10 b 10 M TIME OW , O C?O O5 G .
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,e o {
o 4 0.0 d.3 (0 b M } TIME (hd 9 3 FIGLRE C.1-1 i, l s I;
\
- l' l .
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solid or liquid. At the time of ' reactor vessel failure, some material [,
,- . aj j remains suspended as airborne aerosol or vapor and would be discharged ') .g -q from the primary system into the containment. The rate of discharge is a
determined by the gaseous flo'w between the primary system and 1 containment which is sequence specific. (It should be noted that some s j fission products can be discharged into the containment before vessel ! 1 failure through relief valves or through breaks in the primary system. This is also sequence specific.) This set of inter-related processes . are treated in MAAP and essentially result in a release of all airborne aerosol and vapor from the primary system into containment immediately following vessel failure. j 1 As a result of the . dense aerosols formed when fission products are released from the fuel, considerable deposition occurs within the primary system prior to vessel failure. For some accident sequences, the primary system may be at an elevated pressure at the time of core slump and reactor vessel failure. Resuspension of these aerosol deposits during the primary system blowdown is assessed in reference 6.6 in terms of the available experimental results and basic models. It is concluded that resuspension immediately following reactor vessel failure would not be significant (less than 1 percent of the deposited materials) even for depressurizations initiated from the nominal operating pressure. For delayed containment failure, this small fraction of material is depleted by in-containment mechanisms. Therefore, a major fraction of the volatile fission products are retained within the primary system following vessel failure, the i distribution being determined by the MAAP calculations prior to vessel i
. 6.2-2 -
t 1A rew e = ____.___m__ _ _ _ _ _ _ _ _ _ .__-_
m =:=.- . .._ ~_ . . . - . . . - - . - b fcilura.- N-tural circulation through th2 primary systsm after vassal failure is analyzed using MAAP which allows for heat and mass transport
. <l .]i.
l La in various oodes of the reactor vessel and the steam generators l, t[A including heat losses from the primary system as dictated by the 8 .a i
',,, reflective insulation. Material transport as aerosols and vapors af ter '5L vessel failure is governed by the heatup of structures due to pt radioactive decay of deposited fission products. This heacup is principally determined by the transport of cesium idiode and cesium hydroxide by the nat aral circulation flows. In this regard, the vapor i . ,_ , pressure of cesium iodide is applied to both the cesium iodide and 1'
1 cesium hydroxide chemical species. In carrying out these calculations, 4 J- the pressurization of the primary system is dependent upon the pressurization of the containment and the heating within the primary
!1 system. These determine the in- and out-flows between the primary
.L system and containment. . .
Deposition within the containment is calculated using thermal-hydraulic conditions determined by MAAP. The major aerosol sources are the releases prior to vessel failure (sequence specific), the airborne aerosols and vapors transferred from the primary sys tem at the time of
', J[] vessel failure, the subsequent releases from the primary system due to long-term heacup, and concrete attack.' At the time of containment b ,;
f ailu re, the remaining airborne aerosol and vapor can be released to the [I; environment. Assessments of the potential for resuspension of deposited l
<.a ; aerosols following containment failure (reference 6.5) show this
- 31
;j phenomenon is negligible. j k
ia f 6.2-3 1 7e
A i , l 6;3 Sicu ness An,1vz-d
- 1. .
s 6.3.1 S,HF (Drains Blocked) - Following core uncovery, the fuel becomes overheated and the release < of fission products is initiated. Referring to Table 6.3.1-2, approximately 90 percent of the volatile fission products have been
~
released from the core at the time of vessel failhre. At this time, about 83 percent of the CsI inventory has been deposited on the cooler surfaces of the upper plenum (see Figures 6.3.1-1 through 6.3.1-3). Following vessel failure, the CsI and CsOH deposited in the upper plenum region heats and revaporizes but is redeposited on the cooler regions of the coolant loops and steam generators. The majority of the Te remains deposited in. the upper plenum. Since the heat lo sse s through the reflective insulation and steam generator cooling by the auxiliary feedwater provide decay heat removal in these regions, the . volatile fission products remain deposited throughout the remainder of the scenarios. The distribution of fission products in the containment are shown in Figures 6.3.1-4 through 6.3.1-8. Prior to vessel failure, only about 7 percent of the volatile materials is released through the break and ) into the lower compartment. Once the water in the cavity has I vaporized (about 5 hours), the corium reheats. Although not modeled in the version of MAAP used in this analysis, release of nonvolatile from the debris would now take place. Subsequent analysis (with a version of MAAP which incorporates a model for release for nonvolatile from the debris) shows that this release is negligible due to suppression of the nonvolatile vapor pressure by dilution ef fects (reference 6.7). As the fission products are released into 6.3-1 I t . 1 . I -
1 J l the containment, the aerosols agglomerate and are subject to the 4t I a depletion mechanism modeled in MAAP. This results in the deposition
. m ; ,;j of nest of the fission products released in the containment. ,
2
' Referring to Table 6.3.1-1, the releases to the environment following
] containment failure are very small. The containment will depressurize following failure which will lead to an extended release with the possibility of slightly increased releases from the primary system as deposited material heats up and revaporizes. However, the majority of i';
o this is redeposited in the steam generators (which remain cool due to a the availability of auxiliary feedwater). The remainder is swept from i the primary system by depressurization induced flows. This ef fect is m of fset by continuind in-containment deple tion mechanisms. Long-term releases subsequent to containment depressurization will occur but at extremely slow rates. The amount of released material will also be a very low since the depletion mechanisms inside the depressurized I containment will continue to be ef fective. l 1 I J
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- !g 1NE h)
"4 I FIGW2 6.3 1 k
,-q l
6.3 6 h.]u j L l
S2HF (J7MAAo ORANS BLOCXED) -
. - ~
6 ' 7..... . . . . . . . r e< l \ si
, p
'o
% , p 6 : % Pow Ccw ,
Ie'
. : 4 f
E
~
2
.- i 4 .
W ! N' ..
" i l.-
- J' P
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. TME Or) I FIGL'FI 6 3 1-5
- er 3.; -j .
ik'PPER COW }
.. : i ,
'Bi O , .
. .i 3-l n-I E8 .
g-Wg: ! 3- !
- o. ;
1 p-o, . g] uP ~ fCWER CO MTY jCE NO o.o is s.o 7.s do t2.s is.o ri.s so.o 22.s 25.o 27.s so.o 22.5 25.o TME Ow) FIGL'PI 6 3 1 6 e es l 6 3-7 ._ l ( _ _ _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ __w
b S2HF U7MAAP (DRAINS BLOCXED) E q 4 .- _ . _ ib j 3 e< .
'n OND i 4
! 10 ,d ; ,-[.
,o .--.3hipon t ruc
- 4 -
1 ;, : ,i j g ,. ,l
': < l
- h d' .- t Uy l" f Ed; ; 8 f0
.' hl. 4 ;
I m l r. S' ? x eweacoup
.: : 7 i
- l d- b j
..- g. , j d< l gl
- s. l
>pVTT Y
. c.o 2J 5.0 7J 2.0 12.5 is.o d.s 20.0 22.5 25.0 NIJ 20.0 32.5 25.0
. "; TIME Oy)
^'
FIGWI 6.3 1-7 . t
, ,
- 6
- g. y. ,
. g' ,
./ CEi:CND
~
- 9. [
d*e, j
. f L.,
4-o El N.
+'I .
,O y,, COMP g *., .s....----.. . - . . . . . ..../. UPPER -.--.. . . . . . . - - - - - . - - - . . . . . . - - -- - - - - - - - - - -
, I k;
f *g - ,, i
,- O '. /
h '
* ' OWER CONP !
} ',
l l... a' i . ./ . I. !
> /~A o * - _
r o.o 2.5 S.o 7.5 vo.o d.5 15.0 d.S 20.0 22.5 25.0 2).5 20.0 32.! 25.0
; Ij T.ME Or)
FIGWI 6 31-8 i n . 1 4
'd 4
6.3-8
}
f I fl, 6.3.2 S HT (Drains open) Referring to Table 6.3.1-2, the fission product distribution within ._ the primary system is essentially the same as the previous case. The I amounts of material entering the containment and the environment are ~ nearly the same for both cases. Figures 6.3.2-1 through 6.3.2-8 indicate the fission product distributions in the primary system and containment. In this case, the corium is discharged into a flooded cavi ty. This flooded condition results in maintaining a quenched debris bed preceeding and following containment failure. Because the steam generators remain cool (due to the availability of auxiliary feedwater) the majority of any deposited fission products that may reheat and vaporize eventually redeposit in the steam generators. Fission products cannot leave the primary system through the breached botton head because of the flooded cavity. Consequently, there is virtually no release of any volatiles following vessel failure and the
~
quenched debris in the cavity prevents the late release of nonvolatile. Referring to Table 6.3.2-1, the released fission products to the enviroment are very small. The processes described in seccion 6.3.1 are very effective in, removing the fission products in the containment. This results in similar release fractions as the S HF 2 (drains blocked) case described in section 6.3.1 even though the containment fails almost 18 hours earlier. l 1
)
f 6.3-9 i .'
.I I
-l
__________-_______w
3 U3MAAP-S2Hf- (DRAINS OPEN) 7
- FlSSION PRODUCT RELEASE l
.g
, FlSSION PRODUCT RELEASE FRACTION GROUP TO ENVIRONMENT
~l -
CSI <10-5 a TE02 1.74X10-5 i m SRO <10-5 j RU <10-5 [, CSOH 2.92X10-5
,s TIME OF RELEASE 9.59 HOURS i- DURATION OF RELEASE 10.33 HOURS -U CONTAINMENT HOLE SIZE 0.10 FT 2
m k I"j . J
. t*1 i
IJ B
'~
.!' f
, 6 3-10 d
l' S2HF U3MAAP ORAINS OPOO .J I O g< . . l ( g
- I . ?4
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w, . ; 1 _l 0 i .
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't
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M ; .
- I NDUM LEG
{ ![ ))NBRKN CCLD , i i j, _ - - .. _
/ - - - -_
t
. J >CV5h'URf7G y' ? - . 'i f. t 0.0 h.D A0 k.0 8.0 10.0 kl.0 i ko 100 18.0 20.0 TIME Or)
FIGUP2 6 3 2-1 1.
~
- e. l .... .
I. j '
... ; li
$a.'
g
.- i, e
i ol l 1 , e ' l l
' ' ' ~ ""'
. f.
l':. o N. 39: : I A e t S. ,
$l l l
*l ! I N: I i e: $
UNBRKN f f0LD LEG W ; - -- ;
! Dos e kiQ M !
8
- e . - l / ! 8 l
4 . 4 i . -
+
0.0 io d0 LO LO 10.0 12.0 1A0 100 1&O .0.0 TIME Oy)
. FIGURZ 6 3 2-2
~
, 6.3-11
, , Se <--w = emee pm # N"**
_ame_ _ _ _ - - - - - _ _ - _ -
b 1m .*w*4 W.mgr epi ..g g
& ++
11_ m S2HF U3MAAP ORAINS OP90 I
. 1 .
h . M' e: -,...._
- q g! ,
.'i'~'.b usuu '
j.19 g.
}2: <
j d u 'm.
.g O< e h
J l f
# : . I p
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< 8 9
4, <
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-l I d 4
< 0
+. . ;
~
,{
6 _f / ORE f NBRKN CDt.D LF.G fRES3URtZER 0.0 2.0 40 de de do. d.0 de de de 20.0 T.ME k)
~'
FIGlaE 6.3 2-3 .
. 0, 4
g. o ei , , n 0 .$ g 8
' ! !_ UP904 COMP 3 ...... f..
g . 3 g- ! i.
)
g 1 gs. l' < i 0 g< > a*, 1
.n .i m
U : , ; 6 , g 9 p. a ,, # l e 8 l @
. e- . 3, l .
e tCOMP CE COND d : 'r JCWD i -g 0.0 d.0 A0 do 3.0 d3 12.0 de 16.0 18.0 20.0
" T.ME Oy) j FIG 12E 6 3 2 h
~
y 6 3-12 ul em I- ....-m. _ . _ . . . - - . . _ . . . . _ . m.___._ .. r - - - - - -
- ..q l .
F 1 w I S2HF U3MAAP ORAINS OPEN) < l l~ e :. st ,
.! r W. , ,
e-
/j lt _ UPPER I.'C MP
*. ~.....g,,, y
[.: . j . e- - '- - ... .> l '* l$, e
$ 8 e- t e '
i I I e: : Ad. . l I L .. . l 4 ' e
. e ke: d.
1-
. O y e' ! # '#
,, .**="**'"" Q '
s ' j
.uus 0.0 d.0 40 UC 8.0 10.0 . ~ 15.0 de 14.0 '1a.0 20.0 l
. TIME Or)
FIGURE 6 3 2-5 8 -
- n. \
g*k 3- / : N. ,' ; P 4
'~'
%g- ... _
- f. 9..'
-- COMP' -
5 ' , Ig: S: F 3'3 . S: i Wn: 6' . 4 t 0 ~ , ,
=-""
m.==""" l ,,. 3'. . _
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,. /
- p. ; ,,.
e, ; ,
- e 0.0 2.0 AO LJ 8.0 10.0 Q.0 WO 100 18.0 20.0 TIME Or)
FIGUPI 6 3 2-6
~
l.
'l 6 3-13 1)
)
- - - ... . .. _.. . _ . . . . . . . __ _ - . _ , . . , . . . . __1
' '1 S2HF U3MAAP ORANS OP90
- m. .,
?.
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s 9
;), .m. , .,
r
, ' s.$
w,
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EN. . J
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/. ~----- . . _ _
- N. j CWER COM3 === -
5 e
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g.,.===~"~~~~,,,,,,,.
, '.s .e6 . ,
\ '"- h t'y 1 / '~
- I / _
w
. g. ./ _ -y
- y 0.0 2.0 do Lo a.o so.c do wo to is.o 20.0 TIME Or)
I.~ FIGLE 6 3 2-7
., l 4;
J $ q r# 4# ,f 1ND fALiO 4 f,
/
*! So$.. '
ia 56- l.
- l l.-
4- l
.C. ,
* ' LCJ l p,;
e- / CE CC40
.K ! --
/.-
l -. , gg' . ;
/
- e. , . .
g'* g 3
......./ j f_ ' CWER COMF d' l T'! n,;
. ll / ~~
- ! / '~~ ~ NAvrry l, .l _*i '
g !_ j. b - ~~~'~~~
~
id o.o 2.o 4,o s.o a.o do do no te.o is.o ' 2o.o I TIME Chr) I
.-- FIGLE 6.3 2 8
.!I i -
-t
.k . 6.3.a
o j 6.3.3 TMLB' With A Seal LOCA Following core uncovery, the fuel becomes over-heated' and the release of k [ fission products is iniitated. Referring to Table 6.3.1-2, approximately , 99 percent of the volatile fission products have been released from the . core at the ci ne of vessel failure. At this time, approximately 99 percent of the CsI inventory has been deposited on the cooler surfaces within the reactor vessel (see Figures 6.3.3-1 through 6.3.3-3). Following vessal f ailure, the CsI and Cs0H deposited in the cooler regions of the reactor vessel reheats .ad vaporizes but is redeposited on the . cooler regions of the coolant loops and steam generators. The volatile fission products
~
remain deposited in the primary system for several hours. However, the heat loss from the reflective insulation and lack of auxiliary feedwater to the steam generators is insufficient to prevent reheating, vaporization, and transport of these deposited fission products ' from the primary sys tem Into the containment. The distribution of fission products in the containment is shown in Figures 6.3.3-4 through 6.3.3-8. Prior to vessel failure, only a very small percent of the volatile materials is released through the seal LOCA and i into the lower compartment. Once the water in the cavity has vaporized (about 6.2 hours), t h<r corium reheats and the relase of nonvolatile is ! initiated. As noted in Section 6.3.1, the version of VAAp used in this n analysis does not model the release of nonvolatile from the debris, but analyses using a version of MAAP which accounts for this release show that ] this release is negligible due to suppression of the nonvolatile vapor pressures by dilution effects (reference 6.7). Around 17 hours, the volatile Cs1 and Cs0H start vaporizing and moving out of the primary sys tem i i into the containment. As the fission products are released into the l l' 6.3-15 _
!g ig containent, the aerosols agglomerate end are subject to the deple tion . 1
(- _ mechanisms modeled in MAAP. This results in the deposition of most of the i id ' i Y:1 fission products released into the containment. t As the volatiles are released from the primary system into the containment I late in the scenario, the fission product depletion mechanisms modeled in MAAP are very effective in depositing almost all of the fission products
, released to the containment. Referring to Table 6.3.3-1, the released in fission products to t,he environment are therefore very small. The containment will depressurize following f ailure which will lead to an extended release with the possibility of slightly increased releases from the primary system as deposited material heats up and is swept from the primary system by depressurization-induced flows. This effect is offset by
; continuing in-containment depletion mechanisms. Long-term releases subsequent to containment depressurization will occur but at extremely slow
~
rates. The amoung of released material will also be very low since the
.2 depletion mechanisms inside the depressurized containment will continue to 4
be effective. g.t es t t 4d t wt I
-4
'1 i . .)
< _1 I t g I
l 6.3-16
. r L __
TABI2 6.3 3-1 U4MAAP-TMLB [!
~ ! FlSSION PRODUCT RELEASE t y~'
m FISSION PRODUCT RELEASE FRACTION GROUP TO ENVIRONMENT CSI '50.5X10-5 I t.02 2.57X10-5 SRO <10-5 RU <10-5 . CSOH 64.0X10-5 TIME OF RELEASE 9.59 HOURS DURATION OF RELEASE 10.33 HOURS 2 CONTAINMENT HOLE SIZE 0.02 FT i t I
^
.{
i i 6 3-17 ! 1; [.
- - =a, i
U TMLB' U4MAAP
] 8
's ,...,jpeopurs uu l
. : s i .. : ;
e.3
,.. N. , #
3 ' 6 '
., y
. ao-
' 3 n re i '
b O, S i e
! QN8RMN CO LEG 3
[o'# i
- / ./%,% '
i
. /
/
/
H.-wi l' / 0
, 64.
0.0 2.5 5.0 7.5 10.0 12.5 15.0 17.5 20.0 22.5 25.0 27.5 30.0 22.5 25.0
~: TIME Cnr)
, FIo mE 6 3 3-1
- d
'e 8 I l
. UPPER R.fNt.W.
a' o--l l s . .
' 3 l ,
%: ! i sg- l :
so, t a
- . i g: 'Nenx3 cets usa i.' <a. ')..
E$ ;
/
N E: : 3.. i, St . <
.i g- l o.
- I f w , e /
l ! ' l
*- ! ! --e l
- a. ; :
i y / aessuacen i
- y. '
r * - - - - r4 0.0 2.5 5.0 7.5 10.0 c.5 15.0 17.5 20.0 22.5 25.0 27.5 30.0 22.5 25.0
j' TIME (br)
FIGGE 6.3 3-2 11 i l l l 6 3-18 . b' l
3 TMLB' U4MAAP
.. .. .7 W 'i H i
*i e a . .
f.l. l' r .' .! ' . ......
- PPERFLENUh
. o,
~ /-- --
6.. yN,.
.- I
)
1
.. a 1
.. q iN '
Es q 1 8 -!
#.-s
-l l'
d i l P j, fNilRKN gLD LE; opg g _ /Resuntzp _ o.o 2.s so 7.5 do c.s do ri.s me.o m.s so n.s so.o 22.s 25.o t TIME Cnr) FIGURE 6.3 3-3 .
= .
l e, , g:
. ./
- a. ,
[ g: :
- o. j
^eig : i
#*: 1 g: . . .).
~
s.
...- 1 g: ...- I
*: PfER COAP . . . . . - ...- -
~~~ \ !
s: .. _____. . . . . . . . ... .----- -,,.... L 3 #: er !
! / 4 C a- .. .
/ l o.
)
g.; 1 5: / />/ .'I 'CE COND l [ f -
~
c.o ' 2.5 5.o 7J 10.0 12.s ts.o r1.s 20.o 22.5 25.o 77.s 20.o 32.s 35.0 TIME 03r) FIGURE 6.3 3 4 i i . 2 i 6.3-19 ., ( .. . .
Dd MW n o, ,
-a
.j
/
,d .
j j -ol .
./
j
^ '
-G g;- I
- R n.) e' .
l
- 6
....--y~.
,. ....* (
e ~~...-- j g JJPPG-----"--"--' COMP ... . {
...... f...-- .j v r! ' l....
U* ..,o,
' a /
(I M' f.
^
i j i
~~
. ,t l j/ ./
,' ! /
.. ;1, l DW /A .- QCE COND o
O / ___ y , i I , ., l C.0 2.5 5'.0 7.5 15.0 th.5 15.0 N.5 20.0 22.5 25.0 27.5 30.0 't2.5 25.0 p' TME Oy)
. '. 3' FIGURE 6.3 3-5 so
, o.
i i
~
g.
})pFpe CcyP
.....r---- r- ---
- 6. *
.. ............a--
,o ......*- . . ,
, SQ . ,
56' w ,o
. 6 i, 9 ,J o
' d, s I
! I w< .
..1' 2e' s
*yhi g 8. o<
' e l
! E l f r= b. I 7
):!{.'
d
/
k+' Hg-I l .-
/
f6
~
j f w
* /
.s I /
l
/bCCND -'
'. LOWG COM 3 i
' - j
- d. q. , ---- f
. .; g- /,, -
l
O.
o _ _ _ _ k_ f 35.0 i 4 0.0 2.5 5.0 7.5 15.0 d.5 15. 0 N.5 20.0 22.5 25.0 27.5 30.0 22.5 l
.'d TME Oy) l J FIGURE 6.3 3-6 !
i J d.~ i l
~
a 6.3-20 a
mLB U4MAAP d I j t o a .".
. j o' ,. !
" ' -1 A i :
e
,* l'PAE"R COMP ; i
- s. ,
o- ,' =J V. ; I 6 : ..j e< !
*6 I l o- ! ]
. j e<
U,, 34 o < Ea: l k; I o'
- n. .
o-i CE CONQ
; J.CWER CCMA SAVTY I' 3
0.0 2.5 IA . 5.0 1% 7.5 10.0
/! ,
t2.5 11 0 f7.5 20.0 22.5
!/
25.0 27.5 I 30.0 32.5 35.0 TIME (hr) - FIG EE 6.3 3-7 o o. o a r* o
- 6. ,
..q.....
l IPPER CONP Se , - 686, ' l l s
!h',
o j o l Uc. . E5o ? h
- g. .
o o, .' I H CE CND I l' *
/" i' h' ! ; l 6
/
l l 8-o- 4 I ll\
\ -
fyER CyMP
?
ggy7py s.----.-- - - _ --- - ..-.-- - . l 1 0.0 2.5 5.0 7.5 10.0 12.5 11 0 17.5 20.0 22.5 25.0 77.5 30.0 22.5 35.0 ) TIME (br)
, FIGEE 6.3 3 8 .h 5
6 3-21 I I
===.=--. -
L.. I i i
~]
6.4 References ll9 ljq 6.1 "MAAP, Modular Accident Analysis Program User's Manual," Technical
, - l: Report on IDCOR Tasks 16.2 and 16.3, May 1983.
6.2 FPRAT Users Manual. L 6.3 " Uncertainty and Sensitivity Analyses for the IDCOR Reference l Plants," IDCOR Technical Report on Task 23.4, to be published. l , 6.4 "Fisison Product Transport in Degraded Core Accidents," Technical Report on IDCOR Subt ask 11.3, December 1983.
'. 6.5 EPRI/NSAC, " Technical Report 11.1, 11.4, and 11.5, Estimation of Fission Product and Core-Material Source Characteristics," October l 1982. \ -.
6.6 "Resuspension of Deposited Aerosols Following Pricary System or Containment Failure," IDCOR Technical Report on Task 11.6, July 1984. 6.7 Memorandum from M. A. Kenton, Fauske and Associates, Inc., to William Z. 7, Mims, Tennessee Valley Authority, January 2, 1985. t we L. i e l \ t.
%M
( ., r m_ J
'. }
4
, s
)
!J l 6.4-1 C
t, l
m
' ' kJ!
7.0 ' Summary of Results The TVA Sequoyah Nuclear Plant has been successfully modeled using the IDCOR integrated thermal hydraulic and fission product transport
,.s computer code MAAP. Six base cases and several modified base cases were 7 ' studied. The base cases included: several small break loss-of-coolant
-(SLOCA) accidents, S 2D, S2 E' S2HF; two transient initiated events
- , T 23ML, TMLB'; and one large loss-of-coolant accident, AD. These
-initiators encompass - a range of break sizes and a range of complicating failures. The MAAP code was able to model each of these events.-
- i Alternatives to the base cases included examination of operator actions -
as well as reductions in safeguard capacity. Base case sequences that resulted in containment failure at the ultimate i S failure pressure were studied for fission product transport and i retention in containment. Information developed in this study was then forwarded to a subsequent IDCOR task 18.1, for determination of environmental transport and health effects. 4 (.1 ' d
.u p
f.
-w
{j.;
. i , 1 i
ae l 7.0-1 J-I t
i - 4 H 1 7.1 Base cases t The six base cases were analyzed assuming the availability of maximum d
- safeguards (two ECCS trains) except for systems failed as 'part of the
- T.
sequence initiator. Several interesting results can be observed in
'l table 7.1-1 and in the accident signatures of Appendix C. .,
J.
- 1. For each base case where containment safeguards were available (8 2D, S 2H, AD, T23ML) the containment was not predicted to reach the estimated failure pressure of,240 pereene of the design basis (50 lb/in2g ). In fact, the containment desien basis pressure, 12 lb/in2 g, was not exceeded for these cases.
- 2. Containment insensitivity to break size was noted. This is due to corium quenching in .the reactor cavity. Peak containment pressures for the S D 2 (2-inch SLOCA) and for the AD (10-inch LOCA) events differed only slightly.
- 3. Discharge of the upper head injection (UHI) system was calculated to q I
result in topdown quenching of the core. - Penetration of the j
-)
emergency core cooling system water was not precluded by steaming rates predicted in MAAP. Discharge of the cold leg accumulators was calculated to g'enerate bottom up quenching. Hydrogen production via zirconium oxidation is therefore, more accurately portrayed in the MAAP code than other degraded core models that omit this Sequoyah
. feature.
i 9 l
=l ; 7.1-1 ?
1 w + - = ~ wee mom. + ee-,e,.m-e m r** * *** *
- e <
1J q u 4. Igniters initiated burns at low hydrogen volume fractions and
-; q maintained deflagrations over a long time frame thereby dissipating '
3i K.J . the burn energy at rates that are within the capabilities of the f[l heat sinks. Lower containment temperatures than predicted by other t -a . codes were a direct result.
- 5. The two base cases that ended with containment failure-(S HF, 2 TMLB') did so because of a failure to remove heat from the !
l containment. In the S 2PY sequence, no sprays were available to
, mitigate the long term steam generation of the corium in che ;
i::. cavity or the long term concrete attack depending on the mechanism r' assumed for spray recirculation failure. This failure is, o;
~~
there fore, attributed to steam or noncondensible gas I; overpre s suriza tion. In the TMLB' sequence, the egneainment failed a due to noncondensible gas generation. The corium pile was'not
-)
continually quenched due to the loss of active emergency core
, cooling injection. The concrete attack had progressed only 1/4 of the way through the basemat at the time of containment failure. A liquid pathway release is not like ly.
J f a 6
,e 3
s
.N
,]
. . s
.J .
t
*l l' ; 'I 7.1-2 .
a ! I
Tabis 7.1-1 -] 1. i
SUMMARY
OF SEQUENCES - a ..
~
C. SEQUENCES ANALYZED EVENT S2D S2H S2HF . (DRAINS OPEN)
~
CORE 0.80 HRS 1.20 1.20 UNCOVERY' 48.0 MIN 72.0 72.0 AVG H2 360 LS/HR 324 360 PRODUCTION 6.0 LBAAIN 5.4 6.0 _ RATE 0.10 LS/SEC 0.09 0.10 i INTEGRATED H2 GENERATED 660 LBS. 680 700 IN CORE . FRACTION CLAD REACTED 0.32 0.33 0.35 VESSEL 2.81 HR 3.31 3.31 - FAILURE 168. MIN 199. 199. ICE 4.92 HR 4.55 4.36 DEPLETION 295. MIN 273. 262. CONTAINMENT FAILURE TIME N/A N/A 9.54 HRS CONTAINMENT STM OVRPRES FA! LURE CAUSE N/A N/A PEAK 21.0 PSIA 21.4 65.0 COtGAINMENT @ @ @ PRESSURE 2.8 HRS 3.3 9.54 f 7.1-3 i i b
-l~
' TABLE 7.1-1 n
J d;i
- SUMviARY OF SEQUENCES U!
. ; v. $. . - ".i *h SEQUENCES ANALYZED "'c EVENT S2HF TM B ' T23ML AD .. o (DRAINS BLOCKED)
CORE 1.20 HRS 2.00 1.70 0'.40 j !:; UNCOVERY 72.0 MIN 120.0. 102. 24.0
. ' ry AVG H2 360 LB/HR 504 504 900 L: PRODUCTION 6.0 LBAdlN B.4 8.4 15.0-RATE 0.10 LB/SEC- O.14 0.14 0.25 m
;j INTEGRATED H2. -
GENERATED 680 LBS. 590 520 840
~
IN CORE FRACTION CLAD
- REACTED 0.34 0.30 0.26 0.42 VESSEL 3.35 HR 3.35 2.91 1.52
; FAILURE 201. MIN 201. 175. 96.
ICE 3.81 HR 5.84 5.36 3.15
; DEPLETION 229. MIN 350, 321. 189.
CONTAINW NT
" ,,' FAILURE TIME 25.9 27.1 N/A N/A J CONTAINMENT STM OVRPRES/ SThi OVRPRES/
,. FAILURE CAUSE. NON-COND NON-COND N/A N/A i PEAK 65.0 PSIA 65.0 23.0 22.0
,, CONTAINMENT @ @ @ -@
j PRESSURE 25.9 HRS 27,1 2.9 5.3
;h .J -4 ' []" .
714 { c
H i s e j j 7.2 Operator Action Cases These base case variations assessed various system repairs and , j operator actions. Individual studies are summarized below.
- 1. S2D, S2 H, and AD with minimum safeguards. Reduction of.
emergency core cooling availability to the standard design basis ; assumptions (one train) did not fail containment for cases where
- containment cooling was present. Containment design basis pressure for these sequences was not exceeded.
- 2. SD 2 full restoration of injection, pre-melt; AD with full restoration of injection, pre-melt. These events demonstrated that a safe stable state can be achieved in the vessel if injection is restored prior to core support plate f ailure. In both cases, hydrogen generation was terminated and molten core material was quenched. Differences in hydrogen released to the containment were the principal variables of interest in these cases. Quenching the overheated core resulted in production of more hydrogen than the base case but operation of the igniters precluded any threat to the containment.
- 3. S2 H with partial restoration of recirculation, pre-me l t. This variation of the loss-of-recirculation base case illustrated in-i vessel coolability via recirculation restoration. Hydrogen generation was terminated and the core nodes were quenched.
l
, 7.2-1
r7 12.. .. u ._,- ,
,__ . . . . . . . . - - - - - _ _ -- 7__-----------
n, AM L
. 4. S2 D with secondary depressurization, low pressure injection q systems are available. Rapid depressurization of the primary dl - , system' via augmented heat transferral to the s'econdary system (7' becomes an effective means of allowing ECCS to inject and refill L:
the. primary system.
~
- 5. T 23ML with bleed and feed cooling -- T23ML with feed and bleed cooling. Results demonstrated bleed and feed or feed and bleed cooling is a viable mechanism for decay heat removal when-normal steam generator heat removal is unavailable.
- 6. TMLB' with restoration of power pre- and post-vessel failure. -
Restoration of plant power prior to core support plate failure terminated this accident without vessel failure or containment L challenge. Restoration of power post vessel failure precluded containment failure by quenching core debris in the cavity and by ~ reducing steam overpressurization of the containment. e 1 , d r,
- I $
e.. l il
;7 I '_ ' ! 7.2-2 l In t re -
1 .- I i
-- ---.===-:- a: _ =. .
L r
} '
8 7.3 Fission Product Transport
~
( l Fission product transport was studied using the MAAP code for sequences leading to containment failure. These sequences include _ S2HF and TMLB . Best estimate releases in both events are t several orders at magnitude less than those typically predicted by WASH-1400, as shown in table 7.3-1. i e 6 i 1 . 7.3-1 . 1 I
l 'J J 1 TABLE 7.3-1 I7
' 'M SEQUOYAH RELEASE FRACTIONS COMPARISON WITH WASH-1400 RESULTS-
. m. : .
,N ' Fission Product S2HF S2HF Category (Drains Ooen) (Drains Elocked) TMLB' WASE-1400*
CsI ( 10-5 2.1x10-5 5.1x10-' 6.0x10-1 __ Te 1.7x10-5 g 10-5 2.6x10-5 3,oxio-1 n-
*I1 Sr ( 10-5 g 10-5 4 10-5 6.0x10-2 l
Ru (10-5 (10-5 (10-5 2.0x10-2 Cs0H (10-5 6.9x10-5 6.4x10~'
,.a
- Release category PWR 2 l ':
(., m. W, M h i ", d T* R I il 11.3 t;
.g t
r; 7.3-2
r . . . _ _ _ . _ . . . . _ . . . . . _ l' - i i 1 ' 1- :j
~-
i 8.0 Conclusions Q Based on the various accident sequence analyses, several conclusions can ,
- 2 .
, be drawn. These conclusions may be Jiv' fed into two categories: (1) {s
(;i phenomenological observations; (2) the ability of the MAAP code to model. l l .. plant systems. C. - The results of the Sequoyah analysis indicate the importance of having water available ~ in the cavity subsequent to the time of vessel failure. In cases where the water level exceeds the necessary curb s_ . height for spillover . from the lower compartment to the cavity before
? '
or immediately following vessel failure (S 2 D,' S2 H, AD, S2HF)' 4 L f concrete decomposition is minimal thereby reducing the contribution of noncondensible gases to containment pressurization. Also, the ra . containment spray recirculation system quenches and cools the cavity
- . debris demonstrating its heat removal ability by matching and
' i -- exceeding the decay heat generation rate.
P. l.
- The minimal hydrogen generation rates and low deflagration rates
.. calculated by MAAP are beneficial for the Sequoyah containment design,
,L
,, - particularly for postulated sequences where ice melt preceeds the time 1 of deflagration. ,
i
.n
. u
- For accident sequences which have auxiliary feedwater available, j
. r-the steam generators remain cool enough to ensure that essentially all the volatile fission products deposited in these
( 1,1 - ,
'd a
' .);
>g' l ., 8.0-1 1, (:_ ,
1 i i l1 i-1 1 regions remain deposited. For accident sequences which do not have any long-term secondary side cooling, almost all of the volatile {
-[ materials initially are deposited in the core - and primary sys tem.
Long-term heating of these structures results in a slow release from
, the primary system to the containment late in the accident. However,- ,,
these releases are small and effectively removed within the containment. Thus, the release to the environment is much less than one percent of the core inventory.
- Fission product release and transport calculations including annulus and emergency gas treatment system models would likely predict even lower environmental releases than determined in this study. -
- Resuspension of deposited aerosols at the time of containment failure is not likely.
- The Modular Accident Analysis Program (MAAP) is a useful tool- for predicting plant response to class 9 accidents, particularly.for
/
evaluating the effects of operator actions on plant response. 6 l l 8.0-2 l- - 4 4
. j
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qJ 9iE PARAMETER FlLE II] E
** BRITISH UNITS
*8R R ** ICE CONDENSER PARAMETER VALUES d ** VALUES BASED ON SEQUOYAH 9
, ** j
- UPPER COMPARTMENT (ACCMPT)
F' 01' 6.51005 VOLWE .J 6 02 558.5 AREA OF REFUELING POOL
,., 03 93.86 -HGT OF SPRAY HD ABOVE BOTTOM OF. COMPT L, 04 11.8 FLOW AREA CONNECTING UPPER COMPT TO DEAD
** END ( ANNULAR) COMPT-TH I S IS FAN DUCTS; TOT AREA AVAIL fl ** lS ABOUT 14.5 FT**2 AND THIS iS CORRECTED FOR ASSLMED
** ACTUAL LOSS COEFF=1.5 VS. 1.O ASSWED 1N MAAP g., 05 1.0387D4 CHARACTERISTIC CROSS-SEC AREA OF COMPT FOR I
** BURNS
=[
** DECK REFERS TO THE-FLR (AND VERT WALLS IN ICE CONDENSER r ** PLANTS) THAT SEPARATES THE UPPER AND LOWER COMPTS' Y 06 0.00 CORB HGT USED IN REFUELING POOL TO GOVERN-FLOWRATE OUT
!'" 07- 20774 AREA OF OUTER WALLS 08 0.040700 LINER THICKNESS ON OUTER WALL l
~
09- 0.DO OUTER WALL L1NER GAP RESISTANCE i
;d 10 3 OUTER WALL THICKNESS 11 .840 THERMAL CONDUCT 1VITY OF OUTER WALL I,i 12 .204 SPECIFIC HEAT OF OUTER WALL ^ "
13 148. DENSITY OF OUTER WALL 1 14- 0 ENTER A 1 IF THE OUTER WALL IS SOLID STEEL 3 15 0.00 HALF AREA OF INTERNAL WALLS
** INNER WALLS ARE DEF AS WALLS TOT CONTAINED IN A GIVEN r.
'l ** COMPT PROPERTIES OF INNER WALLS ARE ASSLMED TO BE THE.
** SAME IN 8 AND A AND ARE ENTERED IN THE BCOMPT SECT
- 7. 16 0.00 LINER THICKNESS ON INNER WALL
- d. 17 0.DO LINER GAP RESISTANCE ON INNER WALLS 18 3.000 THICKNESS OF INTERNAL WALLS O. 19 1.97D4 DECK AREA ;
20 0.00 LINER THICKNESS ON DECK l
,r 21 0.DO LINER GAP RESISTANCE ON DECK L 22 2.5 DECK THICKNESS l 1-e !?
t- A.1-1
} PARAMETER F LE CONT.I. +
23 0.840 THERMAL CONDUCTIV1TY OF DECK g 24 0.204 SPEC 1FIC HEAT OF DECK I 25 148.- DENSITY OF DECK c 26 0 ENTER A 1 1F THE DECK 1S MADE OF SQLlD STEEL 27 2.319605 METAL EQPT MASS 28 11465 EQPT AREA 29 14 NO. OF.lGNTRS IN UPPER COMPT - 30 -52.9 AVG DIST OF IGNTRS FROM CELLING .),
** ITEM 31 AND THE LIKE ARE USED TO DETERMINE HOW MANY IGN
- l
** SOURCES CAN PROPAGATE BURNS 1NTO ANOTHER COMPT-lE
** WHICH HAVE GRATlNGS OR OPEN SPACE OVER THEM 'I-31 0.00 NO. OF IGNTRS/lGN SOURCES IN.B WHICH CAN BE II
**. SEEN FROM A 3 32 0.00 NO. OF IGNTRS/lGN SOURCES IN D WHICH CAN BE .i
** SEEN FROM A 33 75.1 DIST FROM THE TOP OF A TO THE DECK l' 34 1.D0 FRAC OF SPRAY WTR THAT RUNS INTO REFUELING
. ** POOL (VS. CONT 1NUiNG ON INTO THE LOWER COMPT DIRECTLY) ,
35 1.Do FRAC OF THE REFUEL 1NG POOL THAT DRAINS INTO :
- ** 8 (VERSUS DRAINING INTO C) 36 65. CONTMT FAILURE PRESSURE I 37 1. ENTER A 1.FOR CONAfT FAILURE IN A; O FOR l
** FAILURE IN.D ,
38 1. EQUlVALENT AREA OF CONAfr FAILURE HOLE 1 39 0. EQUIVALENT AREA TO GIVE NOM CONTMT LEAKAGE 40 0. MASS'OF WTR IN NEUTRON SHIELD BAGS 41 1.04 SEDIMENTATION AREA ..
. ,, e
- LOWER COMPARTMENT (BCOMPT) 01 5.135D1 DIST FROM FLR TO TOP OF CCMPT 5410.6 02 AREA OF COR1%i POOL i
03 9.85 HGT OF CURS OVER WHICH WTR FLOWS TO CCCuPT l 04 5410.6 CHARACTERISTlC CROSS-SEC AREA OF CCMPT FOR
** BURNS 05 2.89005 VOLUME l
- t. 06 36.7 VERT DIST FROM THE RX VSL NZLS TO THE 1 I
** TUNNEL'S ENTRANCE-INTO THE CAVITY ~
07 16.38 DIST FRCM THE FLR OF A TO THE CENTER
-i A.1-2 .
-l
e l . E. PARAMETER FlLE CONT. 35 ** FLOWPATH INTO D -
- . "f **
$g ** NOTE 1T lS ASSLMED THAT ANY CORILM lN 8 RADIATES TO ONE
. IJ ** FACE OF THE INNER WALLS IN 8 ONLY
, 08 5.D0 FOR CASES WHERE THE OUTER BOUNDARY OF CONTMT S ** IS A STEEL SHELL SEPARATED FROM A CONC SHIELD WALL,
** ENTER DIST BTNN THE TWO AND TREAT THE STEEL SHELL AS A p ** LINER (ACCMPT AND DCOMPT OUTER WALLS)
J 09 10357.1 AREA OF OUTER WALL (COVERS ICE CONDENSER)
- 10 0.000 OUTER' WALL LINER THICKNESS L 11 0.DO GAP RESISTANCE OF BUILD 1NG OUTER WALL L1NER 12 3.00 THICKNESS OF OLTTER WALL 9 13 0.84 THERMAL CONDUCTIVITY OF OUTER WALL
.U. 14 0.204 SPEC 1F1C HEAT OF OUTER WALL i
15 148. DENSITY OF OUTER WALL
'3 16 0 ENTER A 1 IF THE OUTER WALL IS SOLID STEEL
- 17 3561.8 HALF SURF AREA OF INTERIOR WALL
.p 18 0.DO INNER WALL LINER THICKNESS -
' s.;
, 19 0.DO GAP RESISTANCE OF BU1LDiNG 1NNER WALL LINER l ..,
20 5.96 THICKNESS OF INTERIOR WALLS l'- 21 0.84 THERMAL CONDUCTIVITY OF INNER WALLS 22 0.204 SPEC 1FiC HEAT OF lNNER WALLS 23 148. DENSITY OF 1NNER WALLS ij 24 5410.6 AREA OF FLR (USE WTR POOL AREA IF LESS) 25 0.000 FLR LINER THICKNESS
.[j 26 0.00 GAP RESISTANCE OF FLR LINER 27 12 THICKNESS OF FLR 4 c:
28 0.84 THERMAL CONDUCTIVITY OF FLR C 29 0.204 SPECIFIC HEAT OF FLR 30 :148. DENS 1TY OF FLR I' 31 3.5292E6 MASS OF EOPT I I- 32 29925.0 AREA OF EQPT , L, ** 33-36 ARE USED IN ALL COMPTS WHERE APPLICABLE jd 33 .617 HEAT XFER COEFF ON OUTSIDE OF ALL EXT. WALLS
} 34 100.00- TEMP ON OUTSIDE OF WALLS (j 35 1.D-3 FRAC AREA AVAILABLE FOR REVERSE FLOW ON S-1
" ** FLOWPATH (EG DUE TO ICE CONDENSER DOOR (S) SHUTT I NG)- !
! ** THIS NO. MUST BE NONZERO AND POSITIVE -
[d, 36 .002 FRAC AREA AVAILABLE FOR REVERSE FLOW ON A-D A.1-3 4
c - L. i PARAMETER: FILE CONT.T
! ** FLOWPATH -(EG A1R RETURN FMi FU5IE D5MPERS)-TH I5 - 1S y
** . BASED ON ABOLTT 2, SQUARE INCH PER - DUCT -
37L 298-- .Fi.OW AREALFROM B TO'D , 38 2.4 - FLOW AREA ~ FRCM. B TO A-TH1S REPRESENTS FLR j
. **. DRAINS'(2.2FT**2 WITH LOSS COEFF OF 1.5 AND 2.8 FT**2'OF
.**' LEAKAGE WITH LOSS COEFF OF 2.5) REFERENCED TO MAAP USAGE
** (LC=1.0))
39 .10 NO. OF~lGNTRS OR 1GN SOURCES IN 8 40 22.5 AVG D1ST OF-1GNTRS OR-IGN SOURCES TO B'S
** CEILI,NG 41 28.'1 HGT OF-THE'FLR OF B ABOVE C .
42 5400.- SFOIMENTATION AREA l l **
- CAVITY (CCOMPT)-INCL ALL THE VOL BELOW THE RX NZLS INSIDE
** THE BIOLOGICAL SHIELD AND ALL THE VOL- OUT TO.WHERE THE
** TUNNEL SLOPES UP -
01- 11.1 FLOW AREA THAT BYPASSES TUNNEL (EG AT THE
** NZLS NEAR THE RX VSL FLANGE) 02 649 ' AREA OF CAVITY POOL (lNCL FLR AND KEYNAY-
** ALSO USED FOR WTR AND CORIt.M POOL AREA) ,
03 649 'CHARACTERIST1C CROSS-SEC' AREA OF CCMPT FOR
** BURNS
-04 -1.5501 HGT OF VSL ABOVE BOTTOM OF CAVITY :
05 5.33D1 TOT INSTRLMENT TUNNEL FLOW AREA '!
'06 400 LARGEST CHARAC CROSS-SECTNL AREA THAT CM j
** MUST TRANSVERSE ON ITS WAY TO THE OPENING WHERE IT MAY
** BE ENTRAINED' OR FLOODED TO COMPTS A OR B-IN PLANTS l
** WlTH STM HD PENETRATIONS, THIS WlLL TYP1CALLY BE THE
** " KEYWAY" AREA (USED TO CALC ENTRAINMENT VELOCITY)
. 07 14800.DO CAVlTY,VOLLME 08 2.3101 HGT OF TOP OF TUNNEL ABOVE CAVITY FLR
** MEASURED AT CAVITY SIDE OF TUNNEL 09 2528 AREA OF CAVITY OUTER WALLS
; 10 0.0 LINER THICKNESS 11 0.0 LINER GAP RESISTANCE I 12 5 THICKNESS OF WALL (CR DEPTH CREDITED) )
13 .84 THERMAL CONDUCTIV1TY !
- 14 .204 SPECIFIC HEAT 1
. l
.. A.1-4 E___._-_.L.________-__-________-_.-_-__ _-
t
-]
~
3 I r ._ PARAMETER FlLE CONT. {
'.;.'1 ,
l
! ' ^* 15 148. DENSITY ]
m 16 0. NO. OF IGNTRS OR IGN SOURCES IN C 1
- j 17 0. AVG DIST OF THESE FRCM THE CEILING 18 650. SEDIMENTATION AREA '
- CONCRETE U
** FOLLOWING QUANTITIES USED FOR ALL CONC DECOMP CALCS p 01 56.6 MOLECULAR WEIGHT OF CONC
_ 02 2240.3 MELTING TEMP OF CONC 03 3.8618 A%SS/VOLLME OF FREE WTR RELEASED WHEN CONC
** MELTS 04 2.8592 MASS /VOL CHEM BOUND WTR RELEASED WHEN CONC
** MELTS I
~
05 30.278 MASS /VOL OF CO2 RELEASED WHEN CONC VELTS 06 429.93 ENERGY ABSORBED 1.N ENDOTHERMIC CHEMICAL.
"C ** REACTIONS WHEN CONC HEATED TO IT'S MELTING POINT J 07 343.94 LATENT HEAT OF MELTING
,. * * * =
i ,s
.* CONTROL CARDS J' 01 i ENTER A 0 TO USE CHEAP STM TABLES IN PS
" ** WHEN POSSIBLE
, 02 1 ENTER A O TO USE CHEAP Shi TABLES IN CONT
, ,t ** WHEN POSSIBLE 03 1 RUNGE-KUTTA ORDER (1 OR 2)
,T ( 04 31 UNIT NO. TO WRITE ON FOR RESTART FILES
"? -
** (MAIN PROG)
. 06 32 UNIT NO. TO WRITE ON FOR RESTART FILES i .
** (HEARJP) 07 100 MAX NO . OF RETA I N F i LE OUTPUT PO I NTS-S EST
** FOR ITEM 14 TO BE AN INTEGER MULTIPLE OF THIS NO.
08 06 UNIT NO. TO PUT PRI SYS OUTPUT ON
., 09 06 UNIT NO. TO PUT CONThff OUPUT ON
! : 10 26 UNIT NO. FOR THE FIRST PLOT FILE Lj~ ** (OTHERS SEQUENTIAL)
!O 11 02 UNIT NO. TO PUT SCENAR1O FlLE OUT ON jS 12 150 NON-PEAK NO. OF POINTS CONTROL NO.
i, ** (SEE USER'S MANUAL) I 13 10 P5AK NO. OF POINTS CONTROL NO. (OR ZERO TO
!j ** DE/.CTIVATE PLOT OPTIM12ER)
Q 1m A 1-5
i 1 PARAMETER FILE CONT.;
- l. .,
14 600 MAX NO. OF PLOT POINTS '
. 15 2 ESF PMP LINEUP IN RECIRC (1 FOR ZION, 2 FOR ,
** SEOUOYAH)
. 16 1 ESF PMP/ACCLM DISCHARGE SETUP (1 FOR SI TO
** CLD LGS) 17 0 ENTER A 1 1F B AND W; O OTHERWlSE 18 -2 ENTER A 2 FOR F1SSION PROD RELEASE TO SE
** CCMPUTED BY THE IDCOR/EPRI STM OX1DATION MODEL; 1 FOR
** NUREG-0772 ACDEL; NEGATIVE NOS. ACTIVATE THE SAVE ACDEL
** AS POSITIVE NO. BUT ALSO TURN ON A BLOCKAGE MODEL
** WHICH REDUCES THE RELEASE OF NONVOLATILE FISS1ON PRODS
** WHEN THE NODE IS BLOCKED FCR GAS TRANS ,
==
- CORE 01 3.117D-2 FUEL P1N D1AMtac.R 02 46993 INITIAL ZlRCALLOY MASS 03 50952 NO. OF FUEL PINS :
04 222645 INITIAL UO2 MASS 05 10.109 ELEV OF BOTTOM OF FUEL 06 22.109 ELEV OF TOP OF FUEL 07 8760 TIME OF IRRADIATION 08 1.1641D10 FULL POWER
** THE CORE NODAL 1ZATION ADMITS UP TO 70 NODES; IN
** ADDITION, NO MORE THAN 20 ROWS MAY BE USED AND NO NORE
** THAN 7 RINGS OR COLLMNS WHATEVER NODALIZATION IS USED, ,
** 1NSERT PEAKING FACTORS INTO APPROPRIATE ENTRY NO.
~
** (EG. SECOND RING FROM INSIDE RAD 1AL PEAKING FACTOR IS
** ALWAYS ITEM 32 NO MATTER HOW N%NY AX1 AL NCDES) 09 7 NO. OF RAD lAL NODES 10 10 NO. OF AXlAL NODES l , 11 0.412 AXlAL PEAKlNG FACTOR BOTTCM ROW j 12 0.818 AXlAL PEAKING FACTOR
! 13 0.994 AX1AL PEAKING FACTOR 14 1.075 AXlAL PEAKING FACTOR i 15 1.236 AXIAL PEAKING FACTOR ROW 5 16 1.412 AXlAL PEAKlNG FACTOR 17 1.437 AXlAL PEAK 1NG FACTOR
- 18 1.249 AXlAL PEAKING FACTOR i
i-A.1-6 ' L - - - - - - - - - - - - - ------_-.__-------____o
}. ., 1 d i CONT.
PARAMETER FILE j l n'W jj i l lj 19 0.903 AXlAL PEAKING FACTOR ]
,7 20 0.463 AXIAL PEAKING FACTOR ROW 10 1 I
3 ** 21-30 AXlAL PEAKING FACTORS NOT USED 31 1.034 RADIAL PEAKING FACTOR INSIDE COLLMN 32 1.087 RADIAL PEAXiNG FACTOR , 33 1.098 RADIAL PEAXlNG FACTOR !
. 34 1.120 RADIAL PEAKING FACTOR I
35' 1.082 RAD l'AL PEAK I NG FACTOR 36 0.893 RAD lAL PEAKING FACTOR
" 0,617 RADIAL PEAKING FACTOR COLLMN 7 37 I
38 0.099 AREA /VOLLME FRACTlONS INSIDE COLLMN 39 0.096 AREA /VOLLME FRACTIONS [.. 40 0.128 AREA /VOLWE FRACT1ONS 41 0.158 AREAA/OLLME FRACTlONS r: 42 0.189 AREA /VOLLME FRACTIONS
! 43 0.248 AREA /VOLLME FRACTIONS 44 0.083 AREA /VOLLME FRACTIONS CCLu el 7 I;. 45 30000 FUEL EXPOSURE AT SCRAM (MEGAWATT-DAYS / l
~
** METRIC TON) !
.; 46 .3 FUEL " ALPHA" AT SHUTDOWN (F1SSILE iSOTCPE !
's ** CAPTURES /FlSSIONS)
_ 47 .032 INITIAL ENRICHMENT OF FUEL iN ATCM FRAC ! j , 48 .6DO CONVERSION RATIO (PRODN RATE OF U-239/ )
** ABSOR8 TION RATE IN FISSILE ISOTOPES) AT SHUTDOWN (y 49 .5D0 FRAC OF F i SS I ON POWER MADE DUE TO F i SS I ONS l u ** IN U-235 AND PU-241 AT SHUTDOWN 50 .42 SNK AS 43 FOR PU-239 j' I. 51 .08 SAVE AS 43 FOR U-238 (FAST FISSIONS) 52 1.D-4 FRAC OF CLAD' MASS THAT IS ZRO2 AT TIME O ;, (MJST BE GT 0)
L ., 53 .0151 FUEL PELLET RADIUS i ** n, ;
- ANNULAR COMPARTMENT (DCCMPT-DEAD END CCMPTS IN ICE CONDS)
U 01 9.400D4 VOLLME O2 4809 AREA OF VfrR POOL
.I 03 0.00 DIST THE FLR OF D IS ABOVE THE FLR CF 8 04 4809 CHARACTERISTIC CROSS-SEC AREA OF CCMPT I.3 A.1-7
) -)-
i: PARAMETER FILE CONT.
** FOR BURNS v 05 11056 AREA OF_ EXTERIOR WALLS 06 0.097200 WALL LlNER THICKNESS 7
'I 07 0.Do GAP RESISTANCE OF WALL LINER 08 3 THICKNESS OF WALL ,
09 .840 THERMAL CONDUCT 1VITY OF. WALL ] 10 .204 SPEC 1FiC HEAT OF WALL 11 148 DENSITY OF WALL 12 0 ENTER A 1 1F THE O(JTER WALL iS MADE OF STEEL d
- 13 13.200 HGT OF CURS SEPARATING D AND B MEASURED ,
d
** FROM B'S FLR 14 16. NO. OF IGNTRS OR 1GN SOURCES IN D 15 10.5 AVG D1ST OF THESE FRCM THE CEIL 1NG 16 9600. SED 1 MENTATION AREA. >
*UPLENLM (UPPER PLENUM OF ICE CONDENSER) .
01 47000. VOLLME O2 2003.1 CHARACTERISTlC CROSS-SEC AREA OF~CCMPT
** FOR BURNS ,
03 16.6 HGT OF U ' 04 2003.1 FLOW AREA INTO A 05 16. NO. OF 1GNTRS 06 4.5 AVG DIST OF IGNTRS FROM THE CEILING OF U 07 21.5 AVG DIST FROM THE TOP OF U TO THE CEILING
** OF A JUST OVER U 08 6000. SEDIMENTATION AREA
- ICE CONDENSER 01 110500. VOLLME 02 ????? EXIT GAS TEMP (PROP) '
03 15.00 TEMP OF THE ICE 04 0.016022 SPECIFIC VOL OF ICE 05 3471.0 AREA OF WTR SLMP f 06 0.0417 HGT OF SLMP J 07 48. HGT OF ICE 80X 08 1088. FLOW AREA BTNN B AND 1 . 09 7.D4 SEDIMENTATION AREA (GIVES EFFECTIVE Z OF _ A.1-8 L a_____________________-_-__. -
p
!J FILE CONT.
i PARAMETER
; r3 ' 'd ** .5 METERS) 'n **
l J 9J
- ENGINEERED SAFEGUARDS
^f~ ** IN ENGLISH UNITS l
' ** VOLUMETRIC FLOWRATES SHOULD BE IN GPM; FLOWRATES NOT t
** SPEC 1F1ED TO BE VOL. SHOULD BE IN LBM/HR.; HEADS !
7,
** SHOULD BE IN FEET, PRESSURES IN PSIA; IN METRIC
] ** UNITS THESE ARE RESPECTIVELY M**3/SEC,KG/SEC,M,PA O1 .48 ACCUMULATOR PlPE DIAMETER 02 1.67D2 PRESSURE SETPOINT FOR LPI 03 1.489D3 PRESSURE SETPOINT FOR HPI
~
04 4.15D2 INITIAL PRESSURE OF ACCUMULATORS 05 105 TEMP OF RWST
~
06 1.002 TEMP OF ACCUMULATORS 07 2.90906 I N ITI AL h%SS IN RWST 08 6.51D4 INITIAL MASS PER ACCUMULATOR b, 09 1.48103 AREA OF BASE OF RWST
~
10 60. LENGTH OF AN ACCUMULATOR PIPE 11 1.751D1 PRESSURE SETPOINT OF BUILDING SPRAYS 12 0.D0 PRESSURE SETPOINT OF BUILDING FANS 13 2.00 NO. OF OPERATING FANS [ 14 2.992E5 VOL/SEC CAPACITY OF FANS (NOTE MUST BE GPM
** IN ENG UNITS) ;
j7 15 2.297D-3 NOMINAL DlAMETER OF COMPT SPRAY DROPLETS
~
3 16 1350. TOT VOL OF ONE ACCUMULATOR 17 4 NO. OF OPERATlONAL ACCUMULATORS N 18 2 NO. OF OPERATIONAL HPI PMPS 19 2.DO NO. OF OPERATIONAL LP1 PMPS g 20 5 NO. OF ENTRIES USED iN HP1 PMP-HD CRV TABLE s ** (5 A%X. ) 21 3438. HIGHEST HD IN TABLE (REMAINDER IN
** DESCENDING ORDER)
[ L , "" 22 3188. NEXT LOWEST HD IN HPI PMP-HEAD CRV TABLE tn 23 3187. NEXT LOWEST HD IN HPl PMP-HEAD CRV TABLE lN 24 1531. NEXT LOWEST HD IN HPl PMP-HEAD CRV TABLE ll 25 1406. NEXT LOWEST HD IN HPI PMP-HEAD CRV TABLE l ' [}R 26 0.000 VOL. FLOWRATE CORRESPONDING TO FIRST ENTRY j ** IN HD TABLE l IQ A.1-9
H i PARAMETER FILE CONT.c 27 137.5 NEXT FLOWRATE J 28 196.9 NEXT FLOWRATE - 29 650. NEXT FLOWRATE . 30 681.3 NEXT FLOWRATE , 31 5 NO. OF ENTRIES USED IN LPI TABLE l 32 396. HIGHEST HD iN LPl TABLE 33 369. NEXT HD ; 34 313. NEXT HD , 35 241. NEXT HD' 36' 180. NEXT HD 37 0,000 FiRST VOL. FLOWRATE IN TABLE 38 2437.5 NEXT FLOWRATE 39 4078,1 NEXT FLOWRATE , 40 5343.8 NEXT FLOWRATE 41 6062.5 NEXT FLOWRATE 42 1884.7 CHARGING PMP PRESSURE SETPOINT 43 2.000 NO. OF WORKING CHARGING PhPS 44 5.0D0 NO OF ENTRIES IN CHARGING PMP HD CRV TABLE 45 5500. FIRST HD 46' 5250., NEXT HD 47- 4500. NEXT HD , 48 3333. NEXT HD 49 938. NEXT HD 50 0.000 FIRST VOL. FLOWRATE 51 194. NEXT FLOWRATE 52 291. NEXT FLOWRATE ' 53 384. NEXT FLOWRATE 54- 563. NEXT FLOWRATE 55 29.7 AREA OF BASE OF CONTMT SLhP 56 12.8 DEPTH OF CONTMT SLMP 57 5.000 NO. OF USED ENTRIES IN SPRAY PMP HD CRVS
** (5 MAX) 58 435. FIRST ENTRY IN SPRAY PMP HD TABLE 59 415. NEXT ENTRY 60 280. NEXT ENTRY 61 335. NEXT ENTRY 62 280, NEXT ENTRY 63 0.000 FIRST VOL. FLOW ENTRY IN SPRAY PMP HD CRVS
, 64 2000 NEXT ENTRY A.1-lO
. _ - _ . _ _ _ _ _ - _ _ _ _ - _ _ _ _ _A
a U q . PARAMETER FILE CONT.
'q 65 40C0 6000 NEXT ENTRY NEXT ENTRY
;1 66 ..
i 67 7455 NEXT ENTRY lf' ** FOR NPSH TABLES, THE SAME FLOWS AS WERE GIVEN FOR HD
** CRVS ARE ASSLMED TO CORRESPOND TO THE NPSH HEADS GIVEN
- 68 5.000 NPSH REQ'O FOR CHARGING PMP AT FIRST FLOW
** IN TABLE 69 1.1301 NEXT NPSH ENTRY FOR CHARGING PMPS 70 1.44D1 NEXT NPSH ENTRY FOR CHARGING PMPS 71 1.6601 NEXT NPSH ENTRY FOR CHARGING PMPS
~ 72 2.1901 NEXT NPSH ENTRY FOR CHARGING PMPS
' 73 8.800 FIRST NPSH ENTRY FOR LPI 74 1.19D1 NEXT ENTRY FOR LPI
, 75 1.4001 NEXT ENTRY FOR LPI 76 2.1301 NEXT ENTRY FOR LP1
-, 77 3.59D1 NEXT ENTRY FOR LPI 78 8.800 FIRST NPSH ENTRY FOR HPl '
79 1.13D1 NEXT ENTRY FOR HPI 80 1.1601 NEXT ENTRY FOR HP1 81 2.3801 NEXT ENTRY FOR HPl
.. . 82 3.53D1 NEXT ENTRY FOR HPI 83 1.45D1 FIRST NPSH ENTRY FOR SPRAY PMPS f 84 1.6001 NEXT ENTRY FOR SPRAY PMPS
. 85 1.7001 NEXT ENTRY FOR SPRAY PMPS 12 86 1.8501 NEXT ENTRY FOR SPRAY PMPS
,, 87 2.7001 hEXT ENTRY FOR SPRAY PMPS
'H 88 2.000 NO. OF OPERATING SPRAY PMPS FOR UPPER CCMPT j 89 0.000 NO. OF OPERATING SPRAY PMPS FOR LOWER CCMPT 90 36.5 HGT 0F BTM OF RWST ASOVE THE ENG SAFE PMPS i l(~
u 91 - 2. HGT OF BTM OF CONTMNT SLMP ASOVE THE ENG
== SAF PPS 92 26.0 ELEV OF THE RV INJECTION NZLS ABOVE THE
** S1 PPS n 93 2.354D6 MASS FLOW THROUGH ONE SPRAY PMP WHEN (TEM 94 MEASURED 1
l~ ** 94 4.001 DIFFERENTIAL PRESSURE ACROSS THE SPRAY NZLS
'; j, 95 0.0 FLOWRATE OF EXT RWST REPLACEMENT WTR, IF ANY
!' 96 1.6670-3 TIME DELAY FOR HPI
! 97 2.7780-3 TIME DELAY FOR LPI
! j) l 1
l
! A.1-n fk C________._
. - ,_ = - - _ - _ _ - ______ ___ _- _ _- . . - _ . _. _ .
. ]
- PARAMETER FILE ' CONT.]
98 1.667D-3 TIME DELAY FOR CHARGING'PMPS "l
~
o 99 1.6110-2 TIME DELAY FOR UPPER COMPT SPRAYS ' 100 0.0 TlhE DELAY FOR LOWER COMPT SPRAYS ,
'101 .167 TIME-DELAY FOR FANS / FAN COOLERS '}}