ML20140G744
| ML20140G744 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Shoreham |
| Issue date: | 06/18/1985 |
| From: | Hanan N, Ilberg D BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20140B832 | List:
|
| References | |
| CON-FIN-A-3740, FOIA-85-772 NUDOCS 8509100432 | |
| Download: ML20140G744 (60) | |
Text
{{#Wiki_filter:___ - .... .- - _ . . . -.__ _ . _- - - -. ENCLOSURE.3
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I :HNICAL REPORT A- 740 6-18-85 i s AN EVALUATION OF UNIS0 LATED LOCA OUTSIDE THE DRYWELL IN THE SHOREHAM NUCLEAR POWER STATION D. ILBERG AND N. HANAN RISK EVALUATION GROUP - I i
- DEPARTMENT OF NUCLEAR ENERGY, BROOKHAVEN NATIONAL LABORATORY UFTON, NEW YORK 11973 9
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; Prepared for the U.S. Nucieer Regulatory Commission Office of Nuclear Reactor Regulation '
'J _J 3 id Contract No. DE ACO2-76CH00016 l
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W'. - k A AN EVALUATION OF UNISOLATED LOCA OUTSIDE THE ORYWELL IN THE SHOREHAM NUCLEAR POWER STATION D. 11 berg and N. Hanan Risk Evaluation Group Department of Nuclear Energy
. Brookhaven National Laboratory .
Upton, New York 11973 June 1985 ' Prepared for U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract No. DE-AC02-76CH00016 FIN A-3740 l l
W. ~ ' ^ s . . ABSTRACT A sensitivity study was performed to test the impact on core damage fre- l quency stenuning from the assungtion of the failure of isolation valves to l close following a high energy line break outside drywell. The pipes connect-ing the reactor pressure vessel to the reactor building were identified, and
.. their rupture frequency was evaluated. The time available for the operator to respond before all equipment located in the reactor building fails was esti-
. mated. Event trees for large, medium, and small pipe breaks outside drywell l
.e were prepared in evaluating the core damage frequency. Comparison of the results with the case of successful operation of isolation valves is giveq.
The study concludes that the main contributors in the esse of successful . operation of isolation valves is the large LOCA outside drywell, wheroes, when failure of isolation valves is postulated, both med'um and la ge LOCAs become important. 9 9 e 8 i 6 e9 iii
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e 1 J CONTENTS l Page J ABSTRACT............................................................... iii LIST OF FIGURES........................................................ vi l L I S T OF TA8 L E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v i i l
., PREFACE................................................................viii l ACKNOWLEDGMENTS........................................................ ix
., 1. INTR 000CTION....................................................... 1 1.1 Background.................................................... 1 1.2 Obj ect i ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.3 Scope......................................................... 2 1.4 General Description of tne Problem Evaluated.................. 2
- 2. EVALUATION OF PIPE BREAK FREQUENCIES....................'........... 7
- 3. ASSESSMENT OF MITIGATION CAPABILITY................................ 16 l 3.1 Reactor Building Information.................................. 16 3.1.1 Ins trumentati on for Di agnos ti cs . . . . . . . . . . . . . . . . . . . . . . . . 16 3.1.2 Sug Pumps an Fl oodi ng Bu i l dup Vol umes . . . . . . . . . . . . . . . . 17 3.1.3 ContainmentA{t~osphere.................................
m 18 3.1.4 Procedures............................................. 18 3.2 A Small LOCA Outside Drywell (< 1-1/2" Break Size)............ 20 3.2.1 Acci dent Condi ti ons and A1 a rms . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.2.2 Reactor Bui ldi ng Envi ronment. . ; . . . . . . . . . . . . . . . . . . . . . . . .' 21 3.2.3 Operator Response...................................... 21 3.2.4 Estimation of Core Damage Frequency.................... 22 3.3 A Large Loca Outs ide Orywell () 6" Break Si ze). . . . . . . . . . . . . . . . 24 3.4 A Medi um LOCA Outs ide Drywel l (2" < 6 < 4.3") . . . . . . . . . . . . . . . . . 26 3.4.1 Accident Conditions Alarms and Uperator Response....... 26 3.4.2 Es timation of Core Damage Frequenci es . . . . . . . . . . . . . . . . . . 27
- 4.
SUMMARY
............................................................ 30 i
- 5. REFERENCES......................................................... 31 1
APPEN0!X A: PIPES AND VALVES FAILURE RATES............................ 32
. A.1 Pipe Rupture.................................................. 32 .
A.2 Va l v e Fa i l u re Ra t es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 ' A.3 Compa ri s on wi th LOCA Frequenci es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 APPENDIX B: LINES CONNECTING REACTOR PRESSURE VESSEL TO REACTOR BUILDING.......................................... 38 APPENDIX C: IDENTIFICATION OF PIPE SECTIONS AND DISCONTINUITIES FOR BREAK FREQUENCY ESTIMAT!0N............................ 49 y _ . _ - . . - . _ _ . - - - - - - - , . _ ~ _ - - _ , . , - - . . . . _ _ _ _ _ _ -
l i ! LIST OF FIGURES . Figure , Page , 1 General Description of SNPS Reacto;* Building Elevations (with Emphasis on HPCI Steam line Routing)................. 4 2a Lines- from Reactor Pressure Vessel to Reactor Building....... 5
. 2b TIP Drive Guide Tubes Connections to Reactor Pressure Vessel. 6 -
3 Event Tree Diagram for Sequences Following Small LOCA Outside Drywel1............................................ 23 4 Event Tree Diagram for Sequences Following Large LOCA - Outs i de Drywe l l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 l 5 Event Tree Diagram for Sequences Following Medium LOCA Out s i de D rywe l l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 i i j . I ' l p J l I I v1 y ,, --ew-w- iwp,-- i, -wa--e- y------w-y,my --
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LIST OF TABLES l Table Page I 1 Sunmary of Failure and Unavailability Data for Pipes and Va1ves................................................. 8 2 Estimated Frequencies of Breaks Outside Containment.......... 9
,, 3 Summa ry of Frequenci es of LOCA Outs i de D rywe11. . . . . . . . . . . . . . . 15 4 Reactor Building Temperatures at Several Elevations
- Res ul ti ng from a 40,000 l b. Di s cha. ge. . . . . . . . . . . . . . . . . . . . . . 19 5 Core Damage Frequencies for Unisolated LOCA Outside Drywell.. 30 A.1 Pipe Rupture Rates........................................... 33 A.2 Val ve Ruptu re or Excess i ve Leaka ge Rates . . . . . . . . . . . . . . . . . . . . . 34 A.3 Motor Operated Val ves Fai l u re Rates . . . . . . . . . . . . . . . . . . . . . . . . . . 36 A.4 A Cogarison of Frequencies of Loss of Coolant' Accidents..... 37 B.1 ,
Process Pipelines Penetrating Primary Containment............ 39 4 e e i i 4 i e e O i l i e vii l
, ~ o PREFACE This work was prepared for the NRC which requested it within a one man month time frame. This dictated the use of all readily available information and refraining from physical analyses. Some of the phenomenological assump-tions are approximate; hence, more accurate analysis may result in a.somewhat different contribution to the core damage frequency for the medium LOCA out- .
side the drywell (the major contribution to core damage frequency. in this study). Nevertheless, the identification of the relative hierarchy of con-tributors is believed to be reasonable. , . t I i I l l . l . l - 1 l l t l I l l: v111 lt
g * %N , ACKNOWLEDGMENTS The authors wish to thank Kenneth Perkins, Kelvin Shiu, and Robert Young-blood for their helpful comments. Ed Chow of the NRC (RRAB) is acknowledged
' for his useful conenents. Cheryl Conrad is much appreciated for typing this document to meet a tightly imposed deadline.
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- 1. INTRODUCTION
< 1.1 Background
- The SNPS-PRAI considered LOCA outside the drywell (LOCA in the Reactor Building) in two ways:
i1 .
? a) Interfacing System LOCAs: Appendix F of the SNPS-PRA estimates the initi-d.
G, ator frequency and the core damage frequency for this case. review 2 of the Shoreham PRA re-evaluated the initiator frequency as well The BNL
.i ., as the core damage frequency, and found an increase of about a factor of 3* five in the core damage frequency. The results from the BNL review are included in the present study without elaboration, and more details can be obtained from Appendix C of Ref. 2.
l b) High energy line breaks inside the Reector Building: The SNPS-PRA included in its analysis only pipes larger than 6 in. in diameter, on the premise that, If not automatically isolated ample time is available to isolate breaks in ,these smaller lines before adverse containment condi-
?) tions are generated. The frequency of unisolated line breaks downstream of the outboard isolation valve was calculated to be relatively small.
, The BNL review of this part agreed with the SNPS-PRA, as discussed in
< Appendix C of Ref. 2. In the SNPS-PRA and the BNL review, all the isola-
/J tion valves were assumed to be capable of operating under a postulated
- Q break and the resulting break-flow conditions; random failure of valves to operate was used in both studies,
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i - It is shown 'in Ref. 2 that interfacing system LOCAs are the major con-4 tributor to LOCAs outside the drywell (see Table 5). , 1.2 Objectives - This study is a special consideration of case (b) above steming from the
- assugtion of the failure of the corresponding isolation valves to close fol-lowing a line break outside drywell. NRC requested BNL to re-evaluate the core damage frequency from high energy line-b.eaks outside the drywell (same as case [b] above) under the assumption that most of the isolation valves may j not be qualified to close' under break-fisw conditions, i.e., assuming the failure of the isolation valves. Under this assugtion, there is a need to examine the rupture of any pipe (regardless of diameter) opening a
, path that leads from the Reactor Pressure Vessel (RPV) to the. Reactor Build-i ing, for potential adverse environment or flood effects. This assumption obviously increases the contribution of the high energy ! line breaks to core damage frequency and requires consideration of other lines I , connected to the RPV of diameter < 6 in. This study considers the following questions: (a) What would be the increase in core damage frequency due to the assumption stated before, i.e., the failure of isolation valves to perform their function?
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(b) What would be the contribution to core damage frecuency from each pipe connecting the RPV and the Reactor Building? ' (c) What isolation valves would be important for mitigating the outside dry-well LOCAs?
- (d) What is the characteristic time available for operator action?
1.3 Scope [ The scope of the BNL study was defined to cover the following: - 1
, (a) To identify any significant* high energy lines leading from the RPV to j the Reactor Building with a potential for affecting safety systems, if an unisolated break were postulated.
(b) To estimate the change in SNPS core damage frequency relative to the SNPS-PRAl and BNL review 2 due to the following assumptions on the opera-tion of isolation valves following the occurrance of a line break: (1) The Main Steam Isolation Valves (Inboard and Outboard) on all four main steam lines will isolate in all the cases considered, having the failure rates shown in Table 1 (discussed in Appendix A). (2) All check valves will close on reverse flow as designed with the failure rates shown in Table 1 (discussed in Appendix A). (3) A-11 other isolation valves will fail to close when receiving their j signal to close. No partial closure is assumed for these valves. I l (4) Manual valves are assumed to be available for isolation if accesss I l ible by the operator. (5) Remote operated valves that do not receive automatic closure signals upon sensing break conditions are identified. However, no credit is
, given for them in this study.
(c) To provide the list of the more important isolation valves from the standpoint of reducing the core damage frequency.
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(d) To provide some crude insights on the time available for the operator to respond to such accidents. ' 1.4 General Description of the Problem Evaluated The Shoreham. Reactor Building surrounds the MARK II containment structure " (the drywell). At its lowest elevation (referred to here as Elevation 8), the ' i building is an open cylindrical compartment, i.e., there are no barriers in : 1 .. Elevation 8 compartments. This open area presents the possibility that exces-sive water released into the compartment may adversely affect the ECCS l *The contribution from downstream moderate energy lines of a system was i Ii neglected if it was estimated to be smaller than the contribution of the lines upstream. ;
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equipment in Elevation 8. The SNPS Reactor Building has openings between its
. floors, and a .line break at a high elevation will affect the entire reactor
. building (see section 3.1 for more details). Figure 1 provides a general description of .the SNPS Reactor Building Elevations.
Figures 2a and 2b show lines that connect the RPV to the Reactor Building and provide a potential path from the RPV to the volume of the Reactor Build-ing in the event of a break with a failure of the pertinent isolation valves , to close. These figures .do not show all isolation valves, but only those that l
. are designated as containment isolation valves. In some cases, the most igortant being the RWCU, other valves are available to the operator for i remote line isolation from the control room; these valves are not shown in
',- Figures 2a and 2b.
A list of the lines emerging from the RPV and some additional information
. , associated with these lines (size, type of isolation valves, and process or 1 stand y line) is given in Table B.1 of Appendix B (reproduced from the SNPS-d FSAR3 4
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p; :,.- i- t . .e. Fig. 1 General Description of SNPS Reactor Building
==r '.7,'",,, Elevations (with Emphasis on HPCI Steam Line Routing)
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Fig. 2a Lines from Reactor Pressure Vessel to Reactor Building. s W __ _ _ _ _ _ - _ - - - - - - - _ _
n 8MB REACT 0A PREssolIE VESSEL x22A..RSCLCW-TO A 0To. CooLE : R.ECIACULAfl0sf PutsP R - 7 ... 4 , L E AO RSCL x.228.0TO.CW-C.LE T0 s f.tECIRCul AT808t os xt3A-RSCLCW-F A000 RECIACuLATl0se Puesp ' _ R h d M x238-fesCLCW-FROM RECIACULATloss Putse A800 800 TOR C00 Leas Also as0 TOR COOLER: J s uLATion 134 NITR04EN/A14 PURGE FoA TIP A f. x3TA TIP oAlvE Gulot TUGE: : ; h $
&I x3Te TiPonivE eueE Tuus == h ~ , x37C TiPonivt evioE TueEs =: % 4I l x3ro.Ts. on vE soeE Tvets == % & I
__g ORTWELL Q suPPHEsS4000 CHAtfSER x4s-onTWEtt rLOOa SEAL PatssuaizATi0ss = E ,m tea (Ev^ g 8'act E : x44-onTwfLL Ftoon SEAL PatssunizATiose suPPatseson Po0L - Fig. 2b TIP Drive Guide Tubes Connections to Reactor Pressure Vessel O k # .
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- 2. EVALUATION OF PIPE BREAK FREQUENCIES i This section covers the evaluation of the frequencies of high and mod-erate energy pipe breaks excluding interfacing LOCAs. The interfacing LOCAS are addressed in Appendix C of Ref. 2 and the results are included in Tables 2 and 3.
The pipes considered in this BNL study are listed in Appendix 8. All lines which are associated with General Design Criterion (GDC) 55 are analyzed
, in this 8NL study.* In addition, the Transversing Incore Probe (TIP) Drive Guide Tubes (GDC-57) are considered. All other lines referred to in Table 8-1 as 60C-56 or 57 are not connected to the RPV; they are mainly connected to the Suppression Pool (the routing was rechecked).
The SNPS-FSAR8 was the main source for determining the number of pipe - sections and valves or other discontinuities on each line. The isometric
, drawings of pipe routing in the Reactor Building shown in Appendix 3C of the i
SNPS-FSAR were used. They were compared with the system-specific drawings given in the other FSAR chapters. The summary of this task is presented in Appendix C of this report. The evaluation of pipe break frequencies was made with the failure and unavailability data summarized in Table 1. The bases for the values shown in this table are further discussed in Appendix A. The failure and unava11abil-
, ity data were used with the number of sections and valves or discontinuities identified for each line, to compute the frequency of line breaks. The sum-mary of this task is presented in Table 2. An example of this computation is shown in Appendix C.
1 The results of Table 2 were next grouped into seven different cases: f - (a) Large Interfacing LOCAs (Liquid discharge through, break) ~ 1 (b) Large LOCAs outside Drywell: (1) steam and (2) liquid discharge j (c) Medium LOCAs outside Drywell: (1) steam and (2) Ifquid discharge j (d) Small LOCAs outside Drywell: (1) steam and (2) liquid discharge. . l i I The combined frequency in each group is shown,in Table 3. Note that the j LOCA frequencies of the large and medium breaks groups are dominated by the l line breaks of a single system. For the liquid breaks, it is the RWCU, and ) j for the steam breaks, it is HPCI and MSL drain system In the latter case, j the 10-in. HPCI line break has a frequency of 3.5x10 g., while all other line ! i breaks which contribute to the large LOCA steam line break have a frequency of
. 3% of that of HPCI. Similarly, in the case of the Main Steam Line (MSL) drain break, its frequency is 92% while the RCIC break frequency is only about 8%. '
Therefore, in the rest of this study, when discussing large or medium breaks, only the line breaks of the dominating systems are included; namely, the HPCI
; 10-in. line break, the RWCU 6-in. and 3-in. line breaks, and the MSL drain 3-in. line break.
j *In References 1 and 2 consideration was given to large break LOCA outside the drywell, i.e., if nes which are 6 in. in diameter or more. l 7 i
Table 1 Summary of Failure and Unavailability Data for Pipes and Valves ' Failure Rate (Mean)
- Break Non-Break Component Failure Mode Exclusion
- Exclusion -
Pipes > 3" Rupture 8.6x10 11/hr 8.6x10-18/hr i (per section) -
, P1 pes < 3" Rupture 8.6x10 18/hr 8.6x10 8/hr (per section)
Check Valves Severe Internal -- 3.3x10-3/yr Leakage Rupture 1.5x10 10/hr 1.5x10-8/hr Motor Operated Failure to -- 8x10-3/d Valves (MOV) Operate (w/ comand faults ) l Failure to -- 6x10 3/d Operate (w/o comand)- faults) f Two MOVs (CMF)
-- 2x10 3/d l Rupture 1.5x10-10/hr 1.5x10 '/hr l
- Break Exclusion pipes and valves are those which are designed to criteria provided in Appendix 3C of SNPS-FSAR. The criteria specifies higher design margins and quality control than for the standard pipes and valves.
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l . i l . . . , , i l Table 2 Estleated Fre,sencies of Breats outside Centainment l ' 4 IngesER OF: INITIAL l gaEAK L 5 V ISOLATI0s VAttES BREAK ESTif64TES BESCRIPTIGIl 0F TIIE CASE ANALVZED f taEAK LOCATION SIZE I E A A55101EB FLOW: FREllutaCV N C L UALVES FAILUAE STEAII 0F BREAK { . CASE E T V BESIGNATORS Pn0BASILITV 08 Llqul0 GCCURAEIICE j 5 I E
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a ! 5 (*) Main Steam 1821-A05081 Bred esclusten settles and valve hetueen Reacter Line 1 24* 4 1 1 Inteard II5iv 6.0E-3 steme 5.0E-S Building penetratten and the outboard M5tv. i (Elevatten 78). Il 24" 4 1 e Inheard and 2.0E-3 steam 6.0E-9 Bred enclesten sectlen from autheard IISIV up to estheard the Jet-lepingement Barrier. (Elevatten 78). j le5IV 1821-
- A04082 1
Main Feed. Bred esclusten sectlen and testele check valve water Line I 18" 2 1 1 Check Valve hetueen reacter hullding penetratten and the F002 A/B 3.3E-3 steam 1.4E-4 test ele chectualve. (Elevatten 78). e Test ele C.V. Sred esclusion sectlens and It21-MOUO35A/S free
!! 18* 2 3 1 1821-A05036 [3.3E-3f stese 7.8E-Il testele check valve u A/B and C.V. Barrier (Elevatten 78)p . to the Jet-lapingement l F002 A/S Nigh Pressure Bred esclesien sectlen and valve between Reactor Coolant lajec- 10* 1 1 1 IE41400WO41 1.0 steae 2.lE-6 tullding penetratten and the autheard Iselatten tien (:#CI) valve IE41-400gO42. (Elevatten 66).
Steam Line t II 10* 1 6 6 IE41410v041 1.0 steam 1.4E-6 Iten break esclusten sectlens amit valves free j and IE41- autheard Iselatten valve up te l#CI turbine. Four . q NOUO42 epanings (24 hrs each) per year of valve It0V-042 - ' j 1" are assumed. (Elevatten 66 deun to elevatten IF). - 4 III 1 17 17 IE41-400WO48 1.0 steam 1.0E-3 Iten tred esclesien sectlens and valves fres
- and i Reacter Building penetrations up to the 1-1/2*
IE41-400V047 IIPCI/ACIC drain line to candenser. Normally open path. (Elevatten 66 denn to elevatten 11).
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- Tais incledes all discontimulties, i.e.: valves, pumps, reescers and heat eschangers (see Appendia A).
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, Table 2 Estimated Frequencies of Breaks htside Containment (Continued)
NUMBER OF: INITIAL BREAK L 5 V ISOLATION VALVES BREAK ESTIMATED DESCRIPTION OF THE CASE ANALVZED SREAK LOCATION SIZE I E A Assuma FLOW: FREQUENCY N C L VALVE 5 FAILURE STEAM cf SREAK CASE E T U DESIGNATORS PROBASILITT OR LIqulD OCCURRENCE
$ I E O $
N (*) 5 Interfacing LOCA:
- RNR Shutdeun 1 20' I - 2 - --
Liquid Coollag
- RHR Nead Spray 4* All four laterfaclog LOCA cases estimated on the 3
II 1 - 2 - - Liquid basis of 0.02 for testable check valve unavall-Line 3.0E-7
- RNR/LPCI Injec. 24" 2 ability times 10-3 for spurious MOV opening an-III - 2 -- -- Liquid other 0.1 for probability of laterlock failure and i Line to Rectrc.
i Lines IV 10* 2 2 0.1 for probability of lee pressure piping to fall Lfgald before isolation. See detall la reference 2
- LPCS lajection (Elevation - 8 g to elevation - 87).
LO-F008 and Standby Liquid I N l-1/2 1 1 1 Inboard C.V. 3.3E-3 Liquid 1. K-8 ! Control (SLC) 11 1-1/2 1 1 1 F007 [3.3E-3] Liquid 1.0E-9 Break exclusion section of the SLC l' Elevation 112) The above Non break esclusion section of SLC [ Elevation 112). and Gutboard C.V. F006
- 1' Control Bad Drive (CR0) I 1- 1.0 Liquid 1.0E-4 1
1-1/2 Scras Discharge Volume (50V) header rupture. (Non 1 break esclusion). The pipe break frequency is taken free NUREG-0803. (Elevation 78.63and40).l i e W 4 k e e e .
.e
l r . .. . . l l 1 Table 2 Estlested Fregnencies of Breds Gutside Centainment (Continued) InseER OF: INITIAL BREM L 5 V IS0tATION WALWS SAEAK ESTIIIATED BaE4K LOCATION I KSCRIPTION OF itE CASE ANALVZED SIZE B E C A A5 C Flew: FREQUENCY L VAL M S FAILWRE STEAst GF BREAK CASE E T V K SIGhATORS PA0tAsILITY OR Liqu!D OCCustfieCE 5 I E 8 5 m (*) 5 - tectrc. Pump . i Seal Injection I 3/4 2 2 ,2 1.8 Liguld 2.0E-7 l l Other 3/4" lines I 3/4 1 20 20 Valves of the 1.9 stese 1.E-3 Scanches from verleus sys- and system shown in tem shown la i Lipsid (All elevatten) l this table this table . Sanple Coolant free BPV I 3/4 1 2 2 1.8 Ligste 1.8E-4 Beacter Post Accident Sampl- I 3/4 1 2 2 1.4 Ligsid W tag system (PAS $) 1.0E-4 TIP Brtwe Guten j Tubes I 3/8 4 2 2 Sell valve T.4 Lfgsid 1.K-5 and sheer (Elevatten60). valve " l l l O 4 k
The small steam line bre5ks are mainly due to HPCI and RCIC bypass line breaks (it is the case of a blowdown limited by the 1-in. bypass line). This will be referred to as the 1-in. line break even though the lines are larger in diameter. The small liquid line breaks are represented in this BNL study by the RWCU 3/4-in. branches, and by the CRD SDV header piping rupture (repro-duced from NUREG-0803") which are about 1-1/2 in. equivalent diameter. Table 3. also includes, for each of the LOCA-outside-drywell groups, the - liquid or steam break discharge flow rate at two different times: (1) Initially, when the break occurs and flow rates are at their peak values. - and (2) At about 30 minutes later after coolant injection is established. Appar-
- ently, these estimates do not include special operator action taken to depressurize the RPV and to control the injection flow rate according to
! procedures, for keeping the core covered at level 3.
- These flow rates values should be taken as crude estimates. They were obtained from NEDO-24708s for the purpose of providing some indication of the
! time available for operator diagnosis and response. The NEDO-24708 report provides this information for the entire spectrum of break size under consid-j eration in this study.
j . e O 6 O i 5 14
Table 3 Summary of Frequencies of LOCA Qutside Drywell Break Flow Conditions
- Initial After 30 Minutes . Initiator Break Location Frequency
- Initiator Sta/ Liq Ib/sec Sta/ Liq lb/sec (Main Contributor) (Event /yr)
Large Size Breaks Steam 1400 Liquid <!200 HPCI** 3.6E-6 Steam ( 300 (elevation 8')
+ J_ 6" Liquid 1200 Liquid < 700 RWCU g.6E-6 (elevation 112')
Total 6 > 6 _ l.3E-5 Lcrge Interfacing Liquid 1200 Liquid < 700 LPCI/LPCS 3.0E-7 - LOCAs elevations 87' down to 4 ,
+>6" 8' -
Medium Size Breaks Steam 120 Steam < 60 MSL Drain 1.0E-4 2" < + < 4.3" Liquid 400 Liquid < 250 RWCU l.5E-3 (elevations 112'-126') . Total 2 < + < 4.3" 1.6E-3 Small Size Breaks Steam 10 Steam < 5 HPCI/RCIC** -3.0E-3 .. (elevation 8') 6 < 2" Liquid 25 Liquid < 12 RWCU Branches -1.5E-3 (elevations 112'-150') Total + < 2" -4.5E-3 cApproximate crude estimates of steam or 11guld discharge through break from NE00-24708. The break flow estimates for 30 minutes seem to correspond to a case in which the core is flooded above level 8. If the operator takes control of the in.lection flow rate. RPV pressure, and level (keeping level 3), then the flow rates are expected to be lauer.
- Break can occur between elevatten 66 and 8, but the other break locations discharge through a pipe chase to elevation 8. -
r
I ! I ! 3. ASSESSMENT OF MITIGATION CAPA8ILITY i The effects of LOCA outside the drywell are discussed in this section , according to the three different groups: small, medium, and large pipe breaks
! (see Table 3). Based on these effects, some insight on the time available for mitigation is presented. The first subsection provides general information on
- alarms available for diagnostics, containment sumps capacity and flooding data, and some crude information on the containment atmosphere temperature increase due to steam or saturated liquid discharges. The next subsections
- - describe the mitigation conditions for small, large, and medium LOCAs outside j the drywell. -
! 3.1 Reactor Building Information 3.1.1 I'nstrumentation for Diagnostics I' The following instrumentation and alarms are available to alert the operator in the case of a pipe break in the Reactor Building:
-- Reactor Building ventilation isolation alarm
) -- Reactor Building equipment sump level alarm in the vicinity of the break
-- Reactor Building floor drain sump level alarm l -- Reactor Building flooding alarm at elevation 8 (see additional descrip-
- tion below)
-- Area radiation monitor alarms , ,
! -- Reactor Building Standby Ventilation Exhaust high-radiation alarms
- -- Area high-temperature alarms on elevation 8 and on the floor where the l break occurs
!' -- Specific systems have their own break detection instrumentation such as
.the AMCU, M5L drain, HPCI, and RCIC. '
j ,
-- Reactor Building low differential pressure alarms.
j - Most of these alarms are also sensitive to a small break LOCA 'of about ! 3/4-in. diameter but some set points will only be reached after about half an . hour. l The Reactor Building (R8) water level at elevation 8 is detected by two . j R8 level monitors installed on the R8 floor. The flood alarms are activated ., j by the monitors when the water level is more than 0.5 inch above the floor. The sump alarms will be activated when the water level reaches the sung alarm setpoints installed at a level just below the level that activates the R8 I flood alarms. Sump alarm sensors are installed at various locations in the ' i R8. [ The area high temperature alarms include the following: l 16
-r=-.--t.---w=+-+,--eeww. _ _=------w-ur ,w,-*---ac+- ,y,mm., .,_*-~,-,--m--*etee----***N+-v~~ r'--*'+'
RCIC and HPCI turbine steam line space high tagerature (7 sensors i each). Isolation signal setpoint at 155'F (elevation 8) i RH.R space high temperature alarm (6 sensors) with setpoint at 175'F (ele-vation 8) RWCU space h1gh tegerature (18 sensors) isolation signal at 155'F (ele-
; vation 112)
Main Steam line space high temperature (4 sensors per line) isolation signal at 200*F (elevation 78). J. , i -- Main steam tunnel containment penetration area high temperature (4 sen-J sors) located in the area of MSL drain lines. Isolation signal at l 140*F. 3.1.2 Sumo Pues and Flooding Su11 duo Volumes The open area of the elevation 8 floor is approximately 5,500 sq. ft. l This area is the total floor area minus the area occupied by equipment founda-l tions , columns, drain tanks , etc. Based on this area, flood buildup on eleva. l tion 8 is 3400 gal /in. 1 I The drainage capabilities at SNPS are: ' i -- Reactor Building Floor Sumps - 2490-gal capacity t i Reactor Building Equipment Sugs - 1660-gal capacity r I ' l Reactor Building Porous Concrete Suaps - BOO-gal capacity. l These systems have a total sug c'apacity of 4650 gallons. The total sump pump capacity is 640 gym, as follows: l l 1 Four 50 gpm equipment drain sug pumps (elevation
- 9 ft)
{ . l Six 50 gpm floor drain sump pugs (elevation
- 9 ft) i Two 20 gpm porous concrete sug pumps (elevation
- 9 ft) l --
One 100 gpm leakage return pug (elevation
- 12 f t). .
The leakage return pump is designed to process radioactive water. If the { floor drain sump pump indicators register radioactive material, all sump pumps will isolate. The leakage return pug can then be manually activated by the ope.ator. In addition, only the leakage return pump is powered from onsite AC ,. ,
. power.
!' It can be inferred that if flooding is not arrested before it reaches the 1 ft level above the elevation 8 floor (elevation 9), the sump pump capacity l may drop from 600 gpa to 100 gpe. This level corresponds to accumulation of j
*If water reaches this elevation, the pug is assumed to fail. I 17 l t
~ , about 42,000 gallons. Furthermore, since this study considers primary water i release, it is assumed that only the leakage return pump would be operating (other sump pungs would be isolated).
RCIC, HPCI, LPCI/RHR, and LPCS are all located at elevation 8. It is assumed that they become disabled when water reaches 4 ft (equivalent of about
- 160,000 gallons) as. stated in SNPS-PRA.1 l 3.1.3 Containment Atmosphere i
!: The SNPS-FSAR3 includes in Appendix 3C a few calculations of Reactor l! Building temperatures for water and steam line breaks. Table 4 shows the re- , sults of one calculation for the discharge of 40,000 lb of saturated water at it RPV normal power conditions out of a 4-in. line break at elevation 112 ft of the Reactor Building. In this deterministic analysis, the break was assumed
- to be isolated by an RWCU isolation. signal at 40 sec after initiation of the t* break. This break results in the accumulation of less than 5,000 gal at ele-j, vation 8.that is equivalent to a water level of 1-1/2-in above the floor. It
! is seen from Table 4 that a break of this size is rapidly affecting Reactor Building atmosphere conditions. l l The other calculations reported in Appendix 3C of SNPS-FSAR8 are similar j and lead to the assumption that conditions of 212*F in the Reactor Building l elevation 8 will occur under the following circumstances: ) (1) A RWCU line break discharging more than 500,000 lb. of saturated water. This is approximately the amount discharged from a RWCU 3-in. line break l in 15 to 30 minutes (10 to 15 minutes for a 6-in. line). . l (2) A MSL drain line discharging more than 100,000 lb of steam at RP.V normal l power conditions. For a 3-in. MSL drain line break *.his will occur in j approximately 15 to 30 minutes.
, (3) A RCIC/HPCI 1-in. Ifne discharging more than 15,000 tb of steam at RPV
- normal power conditions, in a very short time, directl
! 8*. A calculation given in NUREG-803 (see Section 3.f.1)showsy to elevation that such i a rapid blowdown is not a typical case for a 1 in. Ifne break, and there-fore 212*F conditions at elevation 8 from these line breaks are not expected to occur **. However, temperatures hit.or than 140*F in eleva-tion 8 can result when steam is discharged directly to this elevation j from a 1-in. RCIC er HPCI line continuously (see Section 3.2.1). 3.1.4 Procedures - I, Given a LOCA outside drywell, the SNPS procedures dictate rapid manual depressurization of the RPV by the A05. This a: tion substantially reduces the - ! flow rate through the break. If low pressure injection is provided at '
'RCIC and HPCI steam linas are enclosed in piping chase which protects higher elevation against a steam ifne break in these systems.. Therefore, for most steam line breaks at higher elevations, steam will exit at elevation 8.
**The 15,000 lb discharge would cause the saturation conditions only if dis-charged during a very short time, which is not the case here.
18
- u . . '
Table 4 Reactor Building Temperatures at Several Elevations Resulting from a 40,000 lb. Discharge Equilibrium Initial Maximum
. Reactor Building Temperature
- Temperatures Elevation ['F] [*F] Comments 8'-0" 104 < 140
. 40'-0" "
148 63'-0" " 183 78'-7" " 194 112'-9" " 217 Break location at 112'. Outside the pump room tenp is 177'F 150'-9" " 148 175'-9" "
< 132
- Reactor Building humidity changed from 50% initially to 100%.
s s I f I f I 19
~
about 200 psi, break flow may become only about one-half of the initial break flow. If the operator controls the RPV pressure to below 50 psi, he may reduce the flow rates to about 10% of the initial flow. Given an R8 flooding alarm, the operator is required to:
-- Monitor R8 level to determine the approximate leak rate, and to ascertain the approximate location of the break (using additional sump alarms and -
high area temperature alarm). Monitor parameters such as line pressures and flow rate of the safety - systems, as a leak may affect these system parameters.
-- If required and plant conditions permit, dispatch an operator to the RS floor to visually locate the source of leakage.
Isolate the break using the appropriate system procedure (HPCI, RCIC, RHR, others). 3.2 A Small LOCA Outside Drywell (< 1-1/2" Break Size) 3.2.1 Accident Conditions and Alarms The description that follows is based on an analysis by NRC staff of a pipe break equivalent to a 1.2-in. line break. This is discussed in detail in
- NUREG-0803." The description in this section applies to small line breaks, in general, and applies to the SNPS. It does not, in particular, apply to SDV header pipe breaks to which the original discussion refers.
The break described is a water line break discharging 550 gpm (-70 lb/s) initially This is equivalent to a 1.2-in. line break discharging from the RPY at 1032 psi conditions. Several alarms are available to the operator as described in section j 3.1.1 above. The most expected early alarms are from the Reactor Building i radiation monitors and from local area high temperature alarms. ! NUREG-0803 cites a calculation for a typical SWR Reactor Buiiding 'that ! shows a tenperature rise to 110*F in 10 minutes and 140*F in 30 minutes for a-discharge of 550 (This amounts to about 130.000 lb
- l. over 30 minutes.) gpm at RPV conditions.It may activate high temperature alarms i>
is I20*F, but it will not isolate NPCI or RCIC systems. . l The SNPS sumps and flooding setpoints are low (see Section 3.1.1), i.e., at 1/E-in. above floor level which corresponds to 2000 or 4000 gallons of l' water accumulation. Therefore, the water accumulation at the 550 gpm flow rate will cause Reactor Building sump and flood alarms to actuate within 5 to I 10 minutts (assuming 35% flashing into steam, travel time through stairwells ! and floort, and partial accumulation in equipment sumps of up to 2,000 l gallons). i e b 20
-m--.- e'-- - --w e e- - - + +
- d '
I i t 3.2.2 Reactor Building Environment i The water released from the break will exceed the local drain sump capac-
, ity, and some will flow to lower elevations through stairwells. Assuming that l
! only the leakage return pump is available,* the accumulation of water at ele- l
" - vation 8 would be less than 0.13 in./ min, i.e., it would take six hours to
~ i j reach the level that threatens ECCS equipment availability. Thus, ample time
! is available for the operator to recognize the need to depressurire the reac- ;
- tor and reduce break flow. Note that Appendix 3C in the SNPS-FSAR states that l
equipment along stairwells is protected against dripping of 212*F water. ' ! During the initial blowdown, temperatures in the nearest area to the ! break can reach 212*F. The Reactor Building tesserature is expected to rise i significar.tly as shown in Table 4 for a discharge of 40,000 lbs of saturated ! } water at elevation 112 ft. This requires a 10 minute discharge from the ) 1.2-in. line break described here. While it may result in high Reactor Build- - ing tegeratures when discharged over a short period of time, it results in i 110*F in the Reactor Su11 ding if discharged during about 10 minutes (see sec-1 tion 3.2.1). However, the tagerature in the containment will continue to rise due to the continued discharge through the break and may reach the 155'F
! RCIC/HPCI isolation tegerature after about one hour. The Reactor Building
- Standby Ventilation System (R85VS) of the SNPS has a heat removal capability ;
, equivalent to less than 55 of the heat discharged by a 1.2-in. If ne break, '
- . before reactor is depressurized and the flow out of the break is reduced.
3.2.3 Operator Response l . At Shoreham, the operator will have a flooding alarm and high Reactor > Building radiation alarm at about 10 minutes as discussed in the previous section. i ' For a small LOCA outside drywell, with the feedwater system operating i when the LOCA occurs, scram may not always occur 1sumediately. Following the !
, scram, the operator will try to keep the normal feedwater injection and there- !
fore keep MSIV open. If the MSIV remains open (which is the more probable ! case), it may take a while before the operator will notice the abnormally high i feedwater flow rate. It appears that the flooding and high reactor building ; L , radiation alarms will indicate that a small LOCA have occurred, and the i j increased feceseter injection flow may be used for verification. l t ' i Therefore, it is expected that the operator will recognize a small break ! } , LOCA in the reactor butiding within about 30 minutes after scram. Unless the i i operator perceives a LOCA, he will depressurire the reactor at a rate of only ( } 100*F per hour. In such a case it will take 4 hours to depressurize the reac-i tor to 100 psi and reduce break flow by about a factor of 10. As seen in sec. l tion 3.2.2, four hours are available at SNPS, before flooding level reaches to , . elevation 12. However, in this case, the tegerature in the Reactor Building may reach 155'F or higher ** between 1 and 2 hours, isolate HPCI and RCIC, and
*Radweste system tanks capacity allows for about one day accumulations of '
untreated water at a 100 gpm pumping rate.
**A GE analysis estimates that the maximum bulk temperature in the reactor <
building would reach about 140*F (see NUREG-0803"). t I i j 21
i t i
;- most probably require depressurization for low pressure injection. These f )
events would lead the operator to recognize that a small LOCA outside !- containment has occurred with high probability, if he failed to recognize it j during the first half hour. l' Note that unlike the generic analysis on NUREG-0803, the authors believe ! that recognition of a small break LOCA outside drywell at SNPS would be a high L ! probability event. This is mainly because of the improved arrangement for - i flooding detection at elevation 8 (relative to the arrangement assumed in NUREG-0803). High radiation and hiS5 temperature conditions in the reactor , building will enhance the probability of recognition. This BNL study assumed . l that it is most probable that manual depressurization of RPV to reduce flow l
- and enthalpy discharge through the break would take place after about 30 ;
i minutes to 1 hour into the accident. I' l The depressurization of the RPV may reduce flow rate and enthalpy of the i 4 water discharged through the break to a level accommodated by the sump pumps, l !- and may reverse the conditions in reactor building, i.e., conditions may start . t to improve. It is indicated in NUREG-0803 that rupture of blowdown panels may
- be required to establish a path for leakage of hot humid air to outside con-
! tainment (which is larger than the " natural" 1005 per day leakage rate from . ) reactor building), in order to improve the reactor building atmosphere condi-l tions and to allow safe operator entry. As shown in NUREG-0803, depressuriza-i tion reduces significantly the dose received by an operator entering the reac-l tor building. I If an operator is required to enter the reactor building to isolate a break, it can be done for a 1.2-in. line break with early depressurization , I (and low primary water activity). It would be possible to stay for an hour. '
- and this seems to be sufficient for isolation purposes. Appendix 3C 'of SNPS- j i
FSAR considers 30 minutes to be sufficient time to walk through all SNPS ele- ' l vations, locate a break, and isolate it.' i l 3.2.4 Estimation of Core Damage Frequency i: , The description of the event and the reactor building conditions follow- ! ing a small break LOCA outside drywell were discussed in the previous sec- ' ! tions. These are now susuarized in the form of an event tree in Figure 3, and quantified. Feedwater and high pressure coelent injection are in general available under the circumstances of small LOCA. AOS, LPCI and LPCS have very low unava11 abilities. The values for their quantification are taken from Ref. 2. The events that are differently quantified are: (1) the probability that at 30-60 minutes the operators take actions and couplete rapid meeval depressurization. (XH ), and i sate flow if reguired (V ' ' ' The )(.2)XH the
=0.01probability of controlling is taken basically from NUREG-the condo ,
i - 0803 where 5x10- is used. The difference between NUREG-0803 and 8NL values ., is due to the SNPS improved early flooding alarms which increase the probabil-ity that the operator recognizes the LOCA outside the drywell and follows the required depressurization procedure. The V=0.1 is the common value used by BNL in Ref. 2 for controlling condensate injection if sufficient time is available to the operator (in our case 30 to 60 minutes). The Y=0.02 includes a factor of 0.2 for the possi-bility that no damage to LPCI/LPCS will occur even under the circumstances 22 __ 1. __
1 Small LOCA . W rator Condensate Core Damage ,
; Outside Fee &sater NPCI/RCIC Follow LPCI/LPCS Puey
! Drywell Recovered Available Timely ADS Procedures Available Injection Frequency 1 - t A,,g Q W X P.W' F Class V l t ) 6x10-4 1 0.g 0.1 1 2.4x10 7 1 0.01
! 4.5x10-3 0.02 i
i 8.tx10 7 1 1 j r4 6x10-* i 0.1 j 2.7x10 e : 1
- 0.01 >
I - 0.02 ;
- 0.1 . 9.0x10 s ,
1 I
. 6x10-4 i l 0.01 i 0.1 g 2.7x10 te j
i 3.6x10 ' i Total = 1.1x10-8 Figure 3 Event Tree Diagram for Sequences Following Small LOCA Outside Drywell. t ) 1 I . 4 j
-- w - - _ ~
__ - _ _ . , _ . - - g ._g .- - - - - - . - - - - . _ _ , , _,
i
- that the operator does not rapidly depressurize the reactor at an early time,
; but rather depressurizes it at the 100*F per hour rate, for 4 hours or more.
1 In such a case NUREG-0803 indicates that entry to the reactor building may be delayed for up to 20 hours. The LPCI/LPCS may survive the adverse environment in the reactor building for such a period, because they are qualified to sus-tain these conditions for at least for several hours. Th - 1.1x10 g per event tree forquantification a small LOCAyields a core damagewhenfrequency of about l year outside the drywell, it is assumed l that the motor operated isolation valves fail to close. t Note that no distinction was made between steam and liquid breaks in the ! case of the small LOCA. The calculated core damage frequency would not change ! much if a distinction between liquid and steam break were made and apparently the flow out of a steam line break would be smaller after depressurization. i 3.3 A large LOCA Outside Drywell D6" Break Size) l This case was treated in the BNL SNPS-review2. However, the assumption
; in the present study is that HPCI and RWCU isolation valves would fail to close.
Only HPCI lines were treated in Ref. 2, and a LOCA frequency .of 2.7x10-8/ year was obtained. If we postulate that the isolation valves fail upon demand, a LOCA frequency of 3.5x10-8/ year is obtained for the 10-in. HPCI line
- j. break (see Table 2).
l
- The 6-in. diameter RWCU line has three isolation valves inside the dry-i well. Only one of them closes automatically on sensing line break conditions in the RWCU lines.
frequency of 9.6x10 jn /yrTable 3 whenas is obtained noderived credit is in given to these Appendix C valves of this areport. break j In Ref. 2, the three valves were given credit (having different isolation sig-
- nals and one of them is of' a different design), and, it was estimated that
- their failure upon demand would be less than 2x10-"/d, and the frequency of l the 6-in. RWCU line break would be about 10-'/ year. It was not further con-sidered in Ref. 2 bec interfacing system LOCAs , was l
calculated to be 3x10 fuse / yearthe frequency which of is two orders of magnitude higher. The I interfacing LOCA frequency estimated in Ref. 2 does not change under the - specific assumptions of this report. The total frequency of large LOCA outside the drywell assuming isolation failure, and including interfacing LOCA becomes 1.3x10-s/ year. When this fre-quency is used with the larger LOCA event tree from Ref. 2, a core damage fre-quency.of 6.8x10-8/yr is found (see Fig. 4). The V' ' ' = 0. 5 is due to ( probability of operator failure to control the condensate system pumps,1) flow the to the RPV in the short time available (about 10-15 minutes), (2) the proba- ' bility that 1000 gpm makeup flow to the hotwell would be insufficient to cogensate for the flow out of the large break and for decay heat removal. In the car,e of a large LOCA outside drywell, the discharge to containment is about 1200 lb/s for liquid discharge or about 300 lb/s for steam discharge, so that saturaticn conditions in the bulk atmosphere of the reactor building are reached within 5 to 10 minutes. The ECCS equipment at elevation 8 would be flooded in about 15 to 25 minutes (the latter number corresponds to 35% 24
~ -
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, e i j NT .
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,-W<e
. . r i n b l ; flashing). Thus, entrance to the Reactor Building is prohibited and no manual
- 1 solation is possible, as it was also assumed in the SNPS-PRA and the BNL
{ review. j , This core damage rf'equency of 6.8x10-s/yr is more than an order of magni-
- tude larger than that given in Ref. 2,' where credit for isolation valve was i given. It is due mainly to breaks in the RWCU. ,
i 3.4 A Medium LOCA Outsids Drywell (2"4 9 4 4.3") *
- 3.4.1 Accident Conditions Alarms and Operator Response ,
i i ! The most dominant case of the medium LOCA is the 3-in. REU line break as ! I shown in Table 2. The frequency of a RCIC 4-in. line break is small compared
! to the total medium LOCA frequency of 1.6x10-8/yr; the RWCU 4-in. line break l frequency is significant but the sections considered are relatively downstream
- and estimated to be 1/4 of the total REU break frequency, whereas the other .
! 3/4 are for 3-in. line break or less. Thus, our discussion in this section ; i refers to a 3-in. REU line break. l The RWCU is located at elevation 112 ft to 150 ft. At 150 ft the domin-eralizers are located, which process water at low pressure and at about 125'F and, therefore, they are not considered. Thus, the break location of signifi- ! cance can occur at the 112 ft or 126 ft elevations. On these elevations, the ! i line are enclosed within concrete shields providing physical separation from ! l all safety related equipment (see Appendix 3C of the SNPS-FSAR). ! Table 4 present the approximate tenperatures in the reactor building fol- ! j lowing a RWCU 4-in. line break at elevation 112 ft in the RWCU pumps room. It is estimated that about 10 times the amount. discharged in that case, i.e. , l . 500,000 lb, would result in saturation conditions in the reactor building. ' i . This will take about 20 minutes if the flow rate of Table 3 (400 lb/s) is , ! assumed. It apparently will take longer because of the decrease expected in ' the break flow due to depressurization after a few minutes (no longer than 10 l minutes). l It is expected that the blowdown from the break will cause immediate MSIV [ r closure and loss of the feedwater system. In about 10 minutes or less, the , temperature at elevation 8 will reach 155'F and trip the RCIC and HPCI, which i ' started a few minutes before that on low level (L2). Therefore, in this case, it is inuieterial whether the operator depressurizes the RPV, because early > automatic AOS actuation is expected for this case. , t The water discharged during the first 10 minutes would flash (-35%) and ' the remainder (about 20,000 gallons) will cascade through the stairwells to elevation 8. Appendix 3C of the SNPS-FSAR considers this effsets and states that no safety system would be affected. This accumulation is equivalent to ' O.5 ft and will result in flooding alarm in the control room. The radiation and temperature alarms are expected to be on in many areas ! of the reactor building. Therefore, it is believed that the situation of LOCA outside drywell and the reactor building adverse conditions would be recog-nized with a high probability within the first 10 minutes. Earlier recogni- i tion of the LOCA and depressurization of the RPV would not change much of the ; 26 f
"s , P
. . . s. ll l a. l progress of this accident sequence.- However, if operators fail to' recognize the event and fail to follow the procedures (which call for keeping RPV at low pressure and controlling the injection flow), then the reactor building condi-tions may severely deteriorate.
. . . The depressurization would apparently happen at about 10 minutes. Then the LPCI, LPCS and condensate pumps, may all inject water to the RPV, and dis-
.- charge a large amount of hot water through the break. While this hot water would have less enthalpy than the saturated water discharged during the first 10 minutes, it has flooding potential because of its high flow rate. Flooding may occur in an additional 30 minutes if the flow rate to the RPV is not reduced by keeping the RPV at the lowest possible pressure without uncovering the core. This is an operator action spccifically required for the case of medium LOCA outside the drywell. If it is successfully performed. BNL esti-mated that LPC!/LPCS may maintain core cooling for a long period and the con-densate system would not be needed until several hours into the accident.
3.4.2 Estimation of Core Damage Frequencies The estimation of core damage frequency for the case of a medium LOCA outside drywell is shown in the event tree in Figure 5. The initiating event does not distinguish between water or steam line
, breaks. They are considered similar because even though the steam discharge through the break is smaller, the igact on containment atmosphere tagerature and pressure is about 5 times higher for a steam line break than for the case of a similar size water line break (see Section 3.1.3).
l In the long run, after the RPV is depressurized, the flow out of.a steam break may be significantly smaller if the RPV is not flooded so that water is discharged through the break. If the water level is kept below level 8 (L8), then the steam flow out of the break is expected to be relatively small. Thus, it may not be sufficient to create a flooding that can damage the ECCS equipment. The liquid line break is therefore the dominating case. Thus, the event
. tree starts with the medium LOCA frequency from Table 3. The feedwater and RCIC/HPCI are assumed to be unavailable. Depressurization by A05 is censidered to occur at about 10 minutes into the sequence. The low pressure injection systems will start to flood the core. Therefore, operator action to control the injection flow rate is needed to reduce the tapact on the reactor
, building and gain time before the condensate system would be required. If the operator recognizes the need to control the injection, then the condensate system pumps may be needed at a later time, and it will be controlled at a later time with a higher probability. If the operator fails to control the low pressure injection, less time will be available to control the condensate pugs injection because they may be needed as early as about 30 minutes into '
the accident.
- The values used for the probability of successful operator action are i thought to be on the conservative side. given the time estimated to be avail-
, able. Therefore, the core damage frequency for medium LOCA outside drywell ) may be smaller than 1.4x10-5 for the case that no credit is given for RWCU j isolation valves. On the other hand, the phenomenological assumptions used ! 27
1 i 4 1
; Medium LOCA Operator Condensate Core Damage Outside Feedwater M I/RCIC LPCI/LPCS Follows Pusy i Drywell Recovered Available Timely ADS Available Procedures Injection Frequency i
A 0 *Y Y H V' Class V cut ) l i 0.95 O.0 l 0.003 ! 0.0 4.6x10-' i 1.6x10 3 0.05
-eo
" 0.1
! 1.0 8.0x10 '
i .
! 6x10 4 J 1.0 5 0.2 1.9x10-7 ; 8x10
- 1.3x10 '
i j Total = 1.4x10 5 t Figure 5 Event Tree Diagram for Sequences. Following Medium LOCA Outside Drywell. l - t 4
- s e
. o.'.v .. .
may not be realistic and my underestimate the break-flow and Reactor Building conditions, so that less time will be available for operator corrective action than assumed above. . O
?
i . 1 f 1 G l . < J l i 29
- 4.
SUMMARY
The BNL review 2 of SNPS-PRA estimated a core damage frequency of 2x10 7 for LOCA outside the drywell in the SNPS; this is mainly due to interfacing system LOCAs. In this study, an additional assumption was introduced at NRC request: namely, that isolation valves would be treated as failing to close upon demand. The only exceptions to this assumption are the MSIVs and check valves. The effect of this assumption is shown in Table 5. It is seen that the core damage frequency increased by a factor of about 100. The leading contribution comes from medium LOCA outside the drywell; in particular, the RWCU 3-in. line break is seen to be of great inportance (see Table 3). - Table 5 Core Damage Frequencies for Unisolated LOCA Outside Drywell Class V Core Damage Frequency , Isolation Valves Isolation Valves Assumed Assumed to Fail to to C1ose on Demand Close on Demand Initiator (from BNL Reference 2) (from this analysis) Interfacing LOCA 1.5E-7 1.5E-7 Large LOCA Outside Drywell 5.0E-8 6.8E-6 Medium LOCA Outside Drywell -- 1.4E-5 Small Loca Outside Drywell -- 1.1E-6 Total 2.0E-7 2.2E-5
~
Table 2 provides the information on the most important isolation valves whose failures contribute to the results of Table 5. RWCU isolation valves are the most important. Next, but by far less important, are HPCI and MSL drain isolation valves. L Tables 3 and 5 show that under the assumptions used in this study, the core damage frequency from LOCA outside drywell is dominated by the RWCU i medium LOCA breaks. Also, the large LOCA contribution comes mainly from the l RWCU system. Therefore, it should be noted that beside the inboard and out- . l board containment isolation valves, the RWCU also has two additional isolation l valves that do not receive an automat'ic signal to close when a line break l occurs and are available for timely remote closure. This action can take up to half an hour after initiation of the accident before the loss of low pres- , , sure injection if the reactor is depressurized early and rapidly. e l 30 i l
n . , , l
. . . . . . s.. . . . . . ;
l l 1 l l
- 5. REFERENCES l
- 1. "Probabilistic Risk Assessment Shoreham Nuclear Power Station Long Island Lighting Company, Final Report," Science Application. Inc., June 24, 1983.
D. Ilberg, K. Shiu, N. Manan, and E. Anavis, "A Review of the Shoreham 2. Nuclear Power Station Probabilistic Risk Ass essment ," NUREG/CR-4050, 4 , BNL-NUREG-51836, Dec aber 1984 (Draft), May 1985 (Final). 5
- 3. " Final Safety Analysis Report Shoreham Nuclear Power Station Long Island
, Lighting Company," SNPS-1 FSAR (Revision 31. August 1983).
4 " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," NUREG-0803, August,1981.
- 5. " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," GE Report NED0-24708, December 1980.
- 6. Reactor Safety Study- "An Assessment of Accident Risks in U.S. Commerical Nuclear Power Plants," WASH-1400, NUREG/74-014. October 1975.
- 7. S.L. Basin and E. T. Burns, " Characteristics of Pipe System Failures in Light Water Reactors," EPRI-NP-438, August 1977.
- 8. W. H. Hubble and C. F. Miller, " Data Summaries of LERs on Valves at U.S.
Commercial Nuclear Power Plants," NUREG/CR-1363, EGG-EA-5125, May 1980. 4 8 O e e O I l l l 31 I
APPENDIX A l PIPES AND VALVES FAILURE RATES A.1 Pipe Rupture The main data sources used for probability of pipe ruptures were the - Reactor Safety Study 8 (RSS) and the EPRI-NP-438 report7 . In the Reactor Safety Study, pipe rupture rates are based on the large amount of data prior to 1973. The EPRI report includes data for an additional two years. Even though it does . not change the RSS results on pipe break rates, it provides more insights on the failure mechanisms leading to pipe breaks, mainly vibrations and pressure surges. It also points out that expansion joints and reducers may be at loca-tions more susceptible to breaks. In the BNL study, reducers and valves were considered as rupture locations, in addition to pipe sections. The SNPS-PRAI uses the RSS data for pipe breaks. However, it distin-guishes between pipe sections which are " Break Exclusion," 1.e., are designed 3 to criteria provided in Appendix 3C of SNPS-FSAR , which basically allow speci-fies larger design margins and higher quality control of these sections. These increased margins are assumed by SNPS to reduce the failure rate of these sec-tions by a factor of 10. BNL accepted this assumption, and the basic values used in the study are similar to the SNPS-PRA and are summarized in Table A.1 below. The pipe rupture data of the RSS is applied section by section, where a section is defined (RSS, page III-41) as follows: A section is an average length between. major discontinuities such as valves, pumps, etc. (approximately 10 to 100 ft). Each section can include several welds, elbows, and flanges. - In this study, piping was also divided into sections where discontinuities were considered to be:
-- Valves
-- Reducers .
-- Pumps
-- Heat Exchangers
! Appendix C presents the details of the pipings and their division into l sections. ' Y A.2 Valve Failure Rates I The main sources used for valve rupture or excessive leakage failure rates ' were the Reactor Safety Study and s NUREG/CR-1363 reports. The values of the NUREG/CR-1363 evaluation are about a factor of three higher than those in the RSS (see Table A.2 for comparison). However, the NUREG evaluation includes also small leakages such as from packing failure. Similarly, the internal leakage rate of check valves given in the NUREG evaluation includes many small ' leakages which are just violations of the Technical Specifications limits, and l too small to be considered in this study. l 32
?
i Table A.1 j Pipe Rupture Rates i Computational Mean ) Assessed Range
- (non break-exclusion Computational Break Non-Break ,
j Ccq onent pipas) , Median Exclusion Exclusion i i Pipes > 3" dia. - 3x10 3x10 '/hr lx10-38/hr 8.6x10-31/hr 8.6x10-ts/hr J per section f
~
i Pipes 4 3" dia. 3x10 3x10-s/hr lx10-8/hr 8.6x10-18/hr 8.6x10-'/hr
! per section
- 1 1
y M
- I l
j .
- 1 i
% e
Table A.2 Valve Rupture or Excessive Leakage Rates Computational Mean Assessed Break Non-Break Component Source Failure Mode Range Exclusion Exclusion - [hr-1] [hr-1] [hr-1]
~
Check RSS Internal Leak- 10 10-6 --- 3.8x10 7 Valves age (Severe) NUREG/CR- Internal Leak- --- --- 1x10-8 1363 age (all sizes) Check RSS Rupture 10 10-7 2.7x10-8 Valves 2.7x10 8 and Motor NUREG/CR- External --- 7x10-8 Operated 7x10-8 1363 Leakage / Valves Rupture , D l l e9 b ( 34
m a e- o The NUREG/CR-1363 evaluation reports about 130 LERs under the title of
" External Leakage / Rupture." However, no case of valve external rupture has occurred. SNPS-PRA estimated from this list that a value of 1/18 may be used to modify the RSS rupture rate to better represent severe rupture of valves.
; This value of 1/18 is also used in this study.
- Based on the above, the BNL study essentially adopted the SNPS-PRA approach, i .e. :
, (1) Use of RSS failure rates for valves.
(2) Apply a modifying factor of 1/18 to the RSS valve upture data. (3) Distinguish between valves which are in the break exclusion zone and those which are not. A factor of 1/10 is applied to the rupture rates of the break-exclusion valves, similarly to the factor applied to the pipe section they are located on. To summarize, the value used for valve failure rates were: check valve internal leakage: 3.8x10 7 x 8760 = 3.3x10-3/ year valve rupture (break exclusion): 2.7x10-8 x 8760 x 1/18 = 1.3x10-8/ year (non-break exclusion): 2.7x10-s x 8760 x 1/18 = 1.3x10-s/ year. For simplification of the analysis, the valve rupture rates were also used with ,
, other discontinuities between pipe sections, such as reducers or pumps; this may be a conservative assumption.
In addition to valve rupture and internal leakage, other failure modes of motor-operated valves were needed in this study. The additional failure modes and failure rates used are summarized in Table A.3. - A.3 Comparison with LOCA Frequencies The analysis in the main part of this report involves a large number of pipe sections and valves. In general, more pipe sections and valves are located outside the drywell. Thus, the frequency of. LOCA outside drywell should be a large fraction of the plant's LOCA frequencies. Table A-4 compares the results of the LOCA frequencies frp this BNL study with the RSS results
, (Table III-6-9 of RSS), the EPRI-NP-438 results, and those of the SNPS-PRA.
S l 35 i
- . - . _ . .-_ . . _ . . _ - . _ . - - - - _ . _ - - _ _ _ . - , _ . . - . . . - _ - - . . - - - _ _ - . . _ . . . , _ _ _ . ~ , - , _ , - - . , _ _ _ , . _ _ - . . _ _ .
Table A.3 Motor Operated Valves Failure Rates
~
i Value Used in Component Source Failure Mode Assessed Range Mean Value BNL Study
, RSS Failure to 3x10 4 - 3x10 3/d 1.3x10 3/d ---
eperate ~
! Motor (include
~
comunand) Operated
, NUREG/CR- Failure to ---
8x10 3/d 8x10 3/d i Valves 1363 operate
- (for BWRs) (include J -
(MOV) } NUREG/CR- Failure to --- 6x10-3/d 6x10-3/d 1 1363 operate I w (for BWRs) (w/o comunand) , Comunand Failure --- 2x10 3/d 2x10 3/d I Failure of of Inboard and , i Both MOVs Outboard MOVs i (Inboard and i Outboard) SNPS-PRA MOV --- 1.6x10 7/hr 1.4x10 3/y j App. A.2* Spurious ! Opening l
- Based on GE evaluation. .
i l .
o i Table A.4 A Comparison of Frequencies of Loss of Coolant Accidents RSS CPRI-NP-438 SNPS-PRA LOCA Pipe Break Mesa All Sensitive Mean This Study: Diameter 90% LOCA Pipes P pes (*) LOCA LOCA Dutside (Inch) Range Frequencies (Mean) y n Frequencies Drywell Small LOCA lx10 " - 1x10 2 2.7x10 3 -10 2 8x10 3 8x10-3 5x10-3 1/2" - 2" Medium LOCA 3x10 5 - 3x10 3 8x10 ' '
--- 3x10 3 3x10-3 1.6x10-3 2" - 6" Large LOCA lx10 1x10-3 2.7x10-" -lx10 3 7x10-" 7x10-4 3.5x105I**)
> 6" f
I*Ilt is assumed that 10% of plant piping are LOCA sensitive pipes.(Ref l) (**)The large diameter pipes are " break-exclusion" and are assumed to have 1/10 of the RSS rupture rate. - 6 l i
O e *
*e 9
0 9 6 e APPENDIX B LINES CONNECTING REACTOR PRESSURE VESSEL TO REACTOR BUILDING I . e 0 O D 38
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- Table B.1 (continued)
PROCESS PIPEl.INES PENETRATING PRIMARY CONTAINHENT lAtW tOfAlles 3AlW Assh/08 . telMA8f thatf A WA4 W 5 hortl4AL MtAllW IS OfikAIOR PMS POW [t Ct0%Ih4 N00sWL untums=1 af Pts alts sin talm ar lies lo orgs lactou IsatAlina ernlarilous u w s Ism Alto (r2) sat almas slut (Is.) coulAiuval p.61 lint (src) stAlus
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' ' l Tables.1(continued) l
- PROCESS PIPEl.INES PENETRATING PRIMARY CONTAINMENT I
- m.s. = = a,e s.ree .,n e., a _ ton.ser isa,eun ..
- 1. %1e steam iseIatten weIsee reeutre met tota seloneie pilets to 4eenergitoe te close valves.
Accianslater air pressure plus seeing set togetper close valves enen sets ptlets are tegnergtged. voltage failure at only one silet will not cause valve closiere. The velves are set te fully tiese to less taan 5 ;ecomes. .
- 2. Cantatement serer to erymell ane suservesten cnassee and iter test If ne return to suseresaten 3
saammer tseletten velves e411 neve tan casestitty to se enmusily ressense after autements . closure, this seche .411 serett contatament serer fee itgn erymell areasure cometstens ame/ec sussreseten meter emeltag. 'enen asesmette signals are au present, snese valves may se seemse , me test er eserettag essventense. ! 3. Testaele eneca valves are eseipeed for remote seeming utta sore ettforential pressure across ' spe .elve seet. The velves etit close en reverse fles even tasugn tne test setteses any se geststemos for esen. The velvee et11 seen men sume etsenerge pressure eassess router 3ressure even thougn see test setten may be posittenes fee tisse. 4 fhts itse is only nessed eartag estatsnance. Service air suegly is etsceanected disetag Slant oseretten by aestatstrettve control. S. AC ester eseretes velves reestree for taeletten fkastiene are esmerge from sne emergency AC poner tuses. OC escretse f eelettes selves are generes from the statten tattertes.
- 4. All essee escretse laslatten velves ell 1 remote i tne last sootttte usen fattere of velve semer. All air.eeeretoe feeletten selves will close uses air fat 16re.
- 7. Si p al 8 esame, s1 pel E everrtees to t!ese.
- 8. 8emer essreted velse saa to esense 'er closes by remote asauel sutts.t fee eserettag convenience eartag any sees of reester eseretten essest amen autements steel Is present (see sete 21
- 9. m
. f=m areal statu.sr.posttten of selve (some er eleseel is the posttten during noruel somer eserattes
- 10. The seestftee stesere rates am as reestree for contatsuent f eelettan only.
11. 4 Sessial air testaele casca velves vita a poettive elestag feature an eastped for remote teettag l . eartog morsel esoreates to assure noenenteel eserestitty of tne valg else. The remote testtag e meere .eu cause air a metici ans of tas ette ines u ric. stmen, otta wir a einse
. effect en fles, usen resetet of as isoletten signal, the estaater t.ortog fores esil ettnee cause
- a silgnt reamstten le flew . men ene foemeter system is avettaale se cause tae velve to c!sse. ,
[ 3reetetag a sostt1ve sleeure 4tfferestsel spessure se tne seeses etsa. emon tse Icemeter f1om te not avelleele. . .
- 12. *nis velve 3011 seen een toen a les rees.or spessure vessel preessai ase es aesteent signal are 3 resent.
- 12. De motor essestee of tats velse is toy lesass esen enring neeeel eseroting emnettiens.
- 14. freverstag to-Core Psees (71P) Sys ems teien ene f!P systes caele is insertes, the nell volve of the salested tune esens automatically se tant tae prese ame caele any aevemen. 4 eentenes of four velves rey to esense as any one stas te comeost the callarettee, aos any one gutes tune is uses, at seat, a fee neurs see year.
If sleeure of the Itae is reentree eartag celleretten, as instestes by a contatament 1seletten signal, tme caele is auteesttsally retrosted ane tse telt valve closes automatically after can. sletten of emelo ettnernmel. fe ensure eseletten tassellity. If a *1y caele falls to attneres er a nell velve falls to close as aseteolve. snese velve is lastaltee in eeen Itae. Usen reestes of a remote ensuel stpel tais eastestue velse util sneer tae flP caele ama seel tne
$stes thee.
LS. All unuses penetrettens(eastpetes *1eere*! are casses ame seel weises. .
- 16. 74tve =411 close en system nign fles.
- 17. Isolation signals A se F .411 (nettate tae routee tuttatag stansey venttlatten syntes =nten te cure isolates tae purge air toelasten velves. *
- 14. This velse will coen seen team a les atfferential pressurg across tre velve ame an accteent signel #
are present. c I 19. Pretsure tentert and senstag stees Ifne spessure are uses for f ater!ect Centrol 13 3revent inaevertent l velse seestag at afgn steen llAe pressure $ (aesse l$ pstg), l* i I i l 46
9 Table B.1 (continued) PROCESS PIPEl.INES PENETRATING PRIt%RY CONTAINMENT Notes (Continued)
. l
- 20. Control Aed Crive (CM) Insert and 'dttnerew Linest Criterta 55 concerns tnese Itnes of :ne reacter coolant pressure teundary penetrating tne scimary reactor contatament. The CSD lasert and witneraw lines are not part of tne reacter caelant pressure teundary. The classiffcatten of tne f asert and uttadraw ifnes fs cuality Grous 8. and therefore designed to accareance vita A5M 5ection !!!. Class 2. The tests ta anten tne CAD lines are costgned ts commensurate alta tne safety fusertance of fselettag these lines. $ lace taese Ifnes are vital to tne scree.functten, taetr oportellity is of uomest concern.
In tne oesign of tats systes, it has toen accosted crectice to cett outsestic valves for tselation purseses as tats introeuces a posstate fatture mecentsm. As a means of providing posttive actuetton, manuel sautoff valves are used. In tne event of a treet on tnese Ifnes. tne annual valves may se closee to ensure isolatten. In addition, a tell enect valve located in tne insert line Instoe tne CR0 ts destgme to automatically seal tats line in the event of a treak.
- 21. This M stas cect valve is normally in a closed positten due to its cect valve feature, but its NO is in the eoen sosition. The M provtoes a tactus' to close tne valve ta provice additional sign lean tigat integrtty.
- 22. Abertviations used f a taale:
Ad . Air CDertted . M . Meter CDersted VTO . Pwumatic Testable Checa Valve
- 4W = Residual Heat Removal System RPT = Aeacter Pressure vessel RCIC = 2eactor Core Isolation Cooling System aCl . Reacter Water Cleanus MPCI . Mfgn Pressure Caelaat Injection GCC . General Design Cetterion
- RBC.CW . Reactor Sullding Closed Lose Caeltag ' dater TIP - Transverstag Incore Prose CAO = Control And Drive MEIV Main Steam !selatten Valve e
b
.P 47
Table B 1 (continued) PROCESS PIPEl.INES PENETRATING PRIMARY CONTAINMENT ISOLATION SIGNAL NOTES . 112!& OESCH!PTTCN A* Assetse vessel low meter level 3 . (A scram will occur at tais level)
- I' Anactor vessel low meter level 2 - (The reactor core tselattan coelf ag syste and tnis nign pressure coolant injection system util be tattiates at tats level, ane rectreulatten pumos are trippeel C' Mfgn radf atiest . main steam line O' Line treet . sein steam Ifne (high steam flow)
(* Line treek . safn steam ifne.(staan Ifne tunnel hign temperature) F* Mtp drywell pigssure G Reactor vessel tow water lent 1 (The core saray systes and tne low pressure core injeciten mese of 80 tit systems will be tattiates at tats
._ level)
J' Line breet la meter meter cleanus system hign space tamperature, hign ef fferenttal flee, 41gn differential tasserature t' Line treet in itsee Ifne to/from tureine (hign steam Ifne seats tercerature, hign steam fleie. Iow steam line pressure or niga turetne canandt disonrega pres:sure) L aesctor Dutidtang standby ventilation systas initiation M nie restattoe signal domstream of primer / costatment purge filter trata . l 0 Ni p ambient tinsperature in meta steam tummel genetrasten are (M57?A) P* Law sein steen line pressure at inlet to curetne (llun mees only) i I Law consenser *scuum l T M1p tasserstus e la Turtine Sutiding U M1 p reacts? vessel pressure if* Nip terwretus e at outlet of cleanus system nonregenerative heat eacnanger I ' Low stems ;messwe T Stanety I! quid strel system actuated Z Low level fn AICJ"J need taas i AM* Remote manual sneitta from meta control rees a
= These are tne f se14tten A,rictions of the primary contaf anent and reactor vessel tselation control systet etner functions are gtwo for taformation only. '
l l l l 48 -
e ' s'* APPENDIX C IDENTIFICATION OF PIPE SECTIONS AND DISCONTINUITIES FOR BREAX FREQUENCY ESTIMATION Main Steam Lines
., All sections of the four lines in the Reactor Building are break exclusion. Two sections are considered: one to the outboard MSIV; one from the outboard MSIV.
Main Feedwater Lines All sections of the two lines in the Reactor Building are break exclusion. They include check valve inbnard and testable check valve outboard. Their failure rate is assumed to be similar. High Pressure Coolant Injection (HPCI)
Reference:
FSAR and LILCO drawings no. M10121-17 and M10122-14.
Description:
10 in.: one section and valve to the outboard valve (MOV-041). Break exclusion. Under normal RPV pressure conditions i because inboard valve is open. 10 in.: six nonbreak exclusion sections (4 challenges per year of 24 hrs each are assumed in these sections): To reducer Branch SHP-171 + valve MOV-049 . Reducer / valve F001 '
- To steam turbine stop valve
- To turbine admission valve The turbine assumed to be equivalent to one section 1in.: two bypass sections and a valve. Six sections downstream to the RCIC/HPCI drain line. TWo branches. All nonbreak exclusion. Normally open.
RCIC
Reference:
. FSAR and LILCO Drawings No. M10116-16 and M10117-13 Oescription: 4 in.: open MOV inside drywell to the outboard MOV. It has a bypass line of 1 in., normally open. Break-exclusion.
six sections and discontinuities: 3 in.:
- to 3x6 reducer
- to drain pot and 3x6 reducer
- to steam turbine stop-valve to steam admission valve
- to steam turbine governing valve the turbine treated as one section.
Following the turbine, low energy assumed. 49
l 1 in.:
- Bypass is 2 sections
- Drain lines from drain pot to RGIC/HPCI drain line are con- '
sidered six sections. l Branches: two or more 3/4 in. bran'ches. Quantification: . 4 in.: [8.6(-11) + 1.5(-10)]
- 8760 = 2.1(-6)
- 3 in.
- ;8.6(-9) + 1.5(-9)]
- 6
- 4 (times per year) x 24 (hrs) = 5.8(-6)
; 1 in.: l6 + 6 + 2] * [8.6(-9) + 1.5(-9)]
- 8760 = 1.2(-3).
!' Reactor Water Cleanup System (RWCU) Supply Line
Reference:
FSAR Figures 3C-4-15A,8,C and Figure 5.5.8-1,2,3 1
Description:
6 in.: One break exclusion section and valve l; 6 in.: One section nonbreak exclusion to reducer 3 in.: Two lines (having three sections each), two valves each and one pump each. I-2 in.: Two lines with section and reducer / check valve. . 3 in.: Two line with section, valve, section, reducer 4 in.: One section and two valves. One of these valves is i normally closed. Another line with section, HX, section ! HX. The heat exchanger (HX) considered as one section in ! ! our approximation. I i
- Beyond the second heat exchanger, temperature is less than 125'F and not considered to be high energy, and will not result in a large environmental
- effect. The high energy part of the RWCU on the return line from the regenera-tive HX to the feedwater line is not considered a significant additional con- :
i tributor, compared to the part already included. , I-Standby Liquid Control (SLC) i l
Reference:
Figure 4.2.3-11 of FSAR and LILCO Drawing M10115-16
Description:
1-1/2 in.-line; 2 check vlaves one inside and the othe routside drywell designated F006 and F007 respectively. Sections : up to CV-F006 is break exclusion section; from F006 to the two norn. ally closed explosive valves is nonbreak exclusion section. Branches: four 3/4 in.-branches from the main 1-1/2 in.-line. Quantification: ,,' [8.6(-10) [8.6(-9) +2+ 1.5(-10)] x 1.5(-9)] * (8760/2) * (8760/2)
- 3.3(-3)
* [3.3(-3)]2 = 1.0 = 1.5(-8) (-9 i
50
e .- .
,=e Control Rod Drive
Reference:
NUREG-0803 The contribution comes from the Scram Discharge Her. der rupture as explained in NUREG-0803. The value of the rupture frequency o" 10 " is derived
- from that report.
Recirculation Pump Seal Injection f
Reference:
FSAR Descriptio .: Two 3/4 in. -i f nes ; 2 check valves one inside and .the other outside drywell. Apparently, it is not break exclusion pipe. Quantification: Similar to SLC but not break exclusion -- 2.0(-7) l Sample Coolant From RPV
Reference:
FSAR i
Description:
3/4 in.-line; one normally open inboard ai r-operated globe valve. One normally open outboard air operated globe valve. Assumed to have one line, two sections, and two valves in reactor building. Nonbreak exclusion. 1 Quantification: 2 * [8.6(-9) + 1.5(-9)]
- 8760 = 1.8(-4)
Reactor Post Accident Sampling System (PASS) ,
Reference:
FSAR ,
Description:
3/4-in. line. One manually operated globe valve outboard, normally open. Two solenoid operated globe valves, normally closed, downstream. Quantification: same as above. TIP Drive Guide Tubes
Reference:
FSAR 6
Description:
four lines of 3/8 in. The tubes are normally with nitrogen. In i order to cause LOCA, all the following must occur: r - One tube rupture inside RPV Nitrogen system alarm fail to alert the operator - Operator error in using the system, failing to operate the shear valve. (The TIP is assumed to be used 4 times per year.) Quantification: 4
- 4x10 2 x 10-1 x 2.5 x 10-3
- 10 4 x 10-8 51
~
e* , Other 3/4-in. Lines It is estimated that there are about 20 sections of 3/4 in., test lines, and other lines branching from the systems listed in this table. Many of them are in the RWCU and are potential " liquid" break location. Other branch out of HPCI, RCIC, and other steam lines, and are potential " steam" break location. O e I G . O 9 I e i
.e 52
s 4 . _ g. 1 w
~ ~ ~ " "
gEd'@ LONG ISLAND LIGHTING COMPANY SHOREHAM NUCLEAR POWER STATION P.O. BOX 4518. NORTH CC UNTRY RO AD . WADING RIVER. N.Y.11792 , JOHN D. LEONARD. JR. vice PatsicEmf. NUCLEAR optmATioms June 28, 1985 SNRC-1185 f Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Unisolated LOCA Outside Drywell Shoreham Nuclear Power Station Docket No. 50-322
Reference:
USNRC letter (A. Schwencer) to LILCO (J. D. Leonard, Jr.) dated May 6, 1985
Dear Mr. Denton:
This letter is in response to the referenced letter from Mr. Schwencer regarding a scoping study currently being performed by the NRC staff. The study is concerned with unisolated LOCAs outside the drywell for the Shoreham reactor building. The isolation valves of concern, as identified in the referenced letter, are in the High Pressure Coolant Injection System (HPCI), Reactor Core Isolation Cooling System (RCIC), Reactor Water . Cleanup System (RWCU), and the Main Steam Line (MSL) drain line. LILCO was requested to provide documentation demonstrating the capability of the valves to isolate a pipe break downstream of the valve under blowdown conditions. To demonstrate isolation capability under these conditions, LILCO has performed an evaluation to assure that all required documen-tation concerning procurement and testing of the valves..is in place and that it is sufficient to demonstrate, by design and test,.that the valves have the capability to isolate as stated a_. above. - 9507100181 85062B PDR NADOCK 05000322 PDR P N
- a. I d l
t
. . . . - - . - . . , - . - . - . - . - . . . - ~ - - - - ~ - - - . - -- -
, - . , - - --. - , - , y __,. - _ , , _
- ~ ~- -- - - .; . . - . . -
- . SNRC-1185
- Page 2
- l The evaluation performed by LILCO shows the following:
- 1. The correct design criteria from the purchage specification
~
were used by the valve vendor for sizing calculations for the original selection of motor operators. The motor . operator sizing calculations indicate that the valves have the capability of closing against the anticipated i differential pressure in a guillotine line break. In the case of the listed valves, the maximum thrust capacity of
- the actuator exceeds the total stem thrust required. This is demonstrated in the documentation provided. (See Paragraph 2 below).
i
- 2. Enclosed is a copy of the calculations for the actuator design as received from the vendor. It is noted'that the
- , date of the calculation sheets is 1985. This was necessary
.. to clarify a complicated format that had originally been 1-used in 1975. The vendor indicated that the information provided in the new format is an accurate representation of the original calculations. The original documentation is
!' available for audit. I
; 3. Review of the original vendor records shows that all valves identified in the reference letter were tested for opening and closing under appropriate pressure differential condi- , tions. The tests provide evidence of the satisfactory performance of each valve against the differential pressure.
In addition, an isolated discrepancy in vendor records for valve 1E51*MOV-042 has been corrected and verified as to validity. A revised copy of the test data is enclosed.
- 4. Verification of ide'ntification numbers of the tested valve
, actuators and the operator size as shown on the test reports was accomplished during field walkthroughs to support the environmental qualification effort. This activity served to verify the design records. No motor data were found to exceed the limitorque design rating. In addition, actual
! voltage type and rating given in the test reports were ! verified against the original design records. . 4 The above discussion confirms the adequacy of the containment. isolation capability at Shoreham. The design parameters and test ! results demonstrate that the capability to isolate under blowdown ! conditions exists. Although LILCO has verified the appropriate
- data and included them in the above discussion, it should be noted that the three valves (1G33*MOV-F100, F106, and F102) '
included in the RNCU line are not classified as isolation valves. Isolation of the line containing these three valves is performed ! by RWCU isolation valves 1G33*MOV33 and MOV34. Although the * ! additional information supports their ability to operate, it is i not necessary to take credit for these three valves for isolation l t j capability. i i l
, , 1
....e-.. -. ...,...,e .- -w. ,. . , . . ~ . . . . . , .
-_ _ - _ . _ . . . - _ , . ~ . _ , , , . _ _ , . - , . _ , , . . . . . - . . . _ . _ , . _ _ , , _ , ~ . . . . - . - . _ _ , , , . . - , , . _ , . _ , . , - _ , . _ _ . _ . - .. ,
SNRC-1185 Page 3
- The information submitted herewith is believed to be fully responsive to the NRC concerns of the referenced letter. Should. ~
you have any questions, please contact this offiqe. Very truly yours, c / . /
,J
% O&b5- f n D. Leonard, r'.
Vi'e President - N,uclear Operations JVW:ck Enclosure - cc: J. Berry 1 e e e O e Y O ,
== *-%,-+=- - -
9
, . - - - . . ~ - - , - - . - - - , _ - , ., --n,. - , -. , , .,, , - - . , ,
= - ._
.z--_:--.--__ = -_-..x_____.. .
I
....tr.~---..e.,__, ,,, ,__,..,7.-= _ , , , , . . . -
, . ... . . . . . ..' .. . . . ; i ( * . ( a. i . O p l HDTOR OP13ATOR CALCULATTONS P.O. NO. OR PROJIc:: LILCO
- RIF.: 72-3287-NC/01 ,
1 G3 Tole 1 NAME: . STONE & WEBSTER - TELAN NO.: 72-3287 N UZMS: 1 (Tact 1521*MOV031) TELAN.,DUC. NO.: ' i VALTI DESC: 3 " 900 L35. B.B. CATE i LINI FRI53! 1337 ?$1 01:7. DIA.: 2.625 QRIF. AzzA: 5.409 AP 1165 PSI 4 TIMP. us 'T. sm DIA. : 1.125 STru AgrA: 0.994 TID:1/5 y 2/5' t Lm: 3 1/8 = STDi TIROST: 0.A. x AP x SEAT TACT.: 5.409 x 1165 x 0.3 = 1890 ! . LDfE PRESS. x 3.A. = 1'337 x 0.994 = 1329 , Pack.1xs Trie:1oa Load = 2454 Total' Stem T1msst = ' 5673 # j! . 0.01215 69 '# S m TCRQUI = STIM'TERTIT x STZM TACT. 5673 x = ,f . 0/A CR UN:: RATIO s MOTOR DISIGN R.F.M. = 1700 = = . ST. 5?IZD IN . / MDI . 12.186 . 55.8
. TERIAD LZAD 2 /5 .- -
. MOTCR CALO. TCRQUE = . . s m TOR 0tT < = 69 = 3.435 *# 7:1' 00: E77. x AFFL. FACT. x O/A AATIQ 0.4 x 0.9 x 55.,5 m- 3.435 = 3.435 s 7.01 's-MOTOR N.3. IT CC CALC. SU??LT. TORQUI DO NOT e REDUCID 50. 2 7 Tc' . -(;y,gg,)2 . 0.49
- 5**"" TCRQUI = MDT. STALZ. TORQUE z*ST. ETT. % x O/A 3ATTO a .
e nCT TOLTAc.t 11 x 0:5 x 55.8 x 1.21 s 3 71 ..s' x S m TOR UE = 2x 69 , 33 g [ 1/v FU*""" = R/W RATIO z UNIT E77. x 1/W DIA. 4.38 x0.95 x1 MA.I. TCRQ. 57. SIT;3tG z MDT. TORQ. x 7/0 IT7. x AFT. TACTCR x C/A RAUC: (g1=m. nLTACz) 10 x 0.49 x 0.A x 0.9 x 55.8 a 98 '# MAZ. E/*= E TORQUI = MAZ. VAL 7T TORf'ICT 1 . E/W AATIO x ETT. *
- 375 .
4.35 x 0.75 x 90 'f , . . OFIRATING T:ME = (60 x LITT) + STZH KFIZD = 16 SICCNDS. /APPROX.) i SMB - 00 QFIIATOR WITE 10 77.7 HOTOR. MAZ. TIRU57:'14000 i MA.I. STIM TCRQgg = 250 'i O/A RAno RANGg a 23 - 109 R/W RATIC = 11 ADD CIAA -
- 1 MA.I. STEM DIA. : 1 3/4 i C KRINT.SUFFLT 460 AC .FDLTS 3 PH 60 CY MUST 0 PIRATE AT 70 T70LTAGE
, .i 1 0 117. l 1 l 2 l3 1 4 l 5 I 6 -l l
CCMF" " 3T:gM 4.gs t i l I l l , A.FFRC7ED ET: INa4.f./rT l 1 i i l l IND. RI7.3T: lh).lf f W l l - 1 i i I l l l .
- #72-11-76. RIT. 1 ,
< i l \ . : -
~
.- l w
' ~ . - - - -_.. --.. . . - - . _ _-.. _
.~. . . _
i l l _ _ . _ _ _ _ _ _ . _ . . _ . _ _ . ~ _ _ _ _ _ . , . _ _ _ , _ _ _ . _ _
- - . _ _ . _ - - . . . .__ _ -. - - . - _ _ _ , .- ~ -.
M"rTOR OPERATOR CALCULATTONS 7.0, NO. OR PROJECT: LIT 40 . RET.: P2-3287-E/02 , CT3TOMEE NAME: . STONE ai WEBSTER . 7" AN No. : 72-3287-N m: 2 (Tact 1521*HOV032) vE:.AN.DWC. No.: , B.B. {; YALTE DESC: La 900 L33. GATE - LINI FRESSi 1337 PSI CR.T. DIA.i 2.625 CRIT. ARIA: 5.409 AP 1165 PSI O TZMF. 583 'y. STZi1 DIA. : 1.125 STEM AREA: 0.99 4 TED: 1/5p 2/5 g, g, y7: 3 1/8 = STEM TERUST: 0.A. x AP x SIAT TACT.: 5.409 x 1165 x 0.3 = 189,0
. LITE FRESS. x 5.A. = 1337 x 0.994 = 1329 ,
. Packing Tnesion Load = 2454 Total' Stem Thrust = 5673 #
- STZ2', TcRQTE = STIM "TEh.UST x STEM TACT. 5673 x 0.01215 = 69 e#
jl '
. 0/A CR UNIT RATIO = MOTOR DISTCN R.P.M. = 1900 = s STIM SPED IN . / MIN . 12.06'. .
63.0 i
. TERIAD LIAD 2/5 . .
MCTOR CALC. TORQUE * * . STEM TORCUE '
= 69 , 3.04 .,
PUL:. QUT ITT. x APFL. TACT. x O/A AATIC 0.4 x 0.9 x 63 . NDTot CALC. TORQUE 4 RZDUCED vcLTAct - 3.d4 . 3.04 s 3.80 '# N.3. IT DC SU??LT. DO NOT 50. I V. .ggy,gg,)2 . 0.8 STALLED TCRQCZ = MOT. STALL TORQUE x*3T. ETT. x O/A RATIO s g lic! votTACg 6 .x 0.5 x 63 x 1.1- 208 ,, . i
- gfg yg.g. , 2 x STIM ToROUT = 2x 69 . 16'5 g .
1/W RATIO - z CNIT ETT. x R/W DIA. 1 x 1 x0.834 i l MAZ. TORQ. SW. SETTTNG s ICT. TORQ. x F/0 EIT. x AF7. TACTOR x C/A RATIO: (fRD. ?cLTAc!) 5 s 0.8 x O.4 x 0.9 x 63 s 91 's i ! . MAZ. I/WEEfL TORQUE = MAX. VALVE TORftUT < i R/W RATIO x ETT. ! 375 . 1 x g x 375 's
. OPERATuc TIME = (60 : L:rT) + STzx syEED . 16 SIconDs.(prRox.)
S 000 Or nATog VITE 5 p
- SMB - TT.i MOTOR. MAZ. TERUST
- 8000 MAZ. STEM TORQUE = _40 'd 0/A RATIO RANGg s 33.5.- 136 I/W RATIO = 1 31 ADD GIAR -
I1 MAX. STEM DIA.: 1 3/8 CURRENT.SUFFLT 125 Dc #0LTS - C MDST CPERATE AT 80 :votTACE 0 RZv. l 4 1 2 l 3 1 & l 5 l 6 ., 4- , i , ccMPILID AT / A s 1s l I ., , . i e AFFROVID BYt' #4./4 4rT I I i u l . te. REv.3Y: ev+ 2rrtiI . . I l ' I j
#72-11-76 RZv. 1
- - . . . 2.
-----,tv- ----g,e--*e--ww -
_-v-e---w.me--+= .ve, --m---ee-m-w~-e-- e --- - * - "
._ . . . 'b
!CTOR OPERATOR CALCULATIONS .
i
; F.C. NO. 01 FROJECT: LILCO REF.: P2-3287-MG/24 -
CUSTOMER NAME: . STONE & WEBSTER , TELAN NO.: P2-3287-N ITEMS: 24 (Taa: 1E41*MOV041) VILAN DUC. NO.
~
YALVE DESC: 1 " R L35. Rg .CATE LINE PRESS: 1337 PSI ORIF. DIA.: 7.875 01:F. ARIA: 48.682 AF 1135 PSI 6 TIMP. 583 'F.
; STEM DIA.: 21/2 STEM AREA: 4.906 TID: 1/3F 2/3 L LIFT: 9.0 "
STIM TERUST: 0.A. x AF x SEAT TACT.: 48.682 x .i135 x 0.3 = 16576
. LINE PRESS. x S.A. = 1117 x 4.906 = 6559 , l 4
Packing Triction Lead = 2500 Total'Stes Thruas = 25635 # , STIM 7014UE = STIM TI1UST z STZM FACT. 25635 x Q,02424 = 621 '8 0/A CR UNIT RA!!O = ?c CR DISTON R.F.M. = unn = . l STIM 571ZD IN./ MIN. 11 og , 7n q.t l TEREA3 LIAD pj 3 , NOT01 CALC. TORQUE = . STEM TOROUE = 621 PCLL CUT EFF. x APPL. TACT. x O/A AA!!Q Q.4 x U.y x /U.yJ
- ICTOR CALC. TORQUE e REDUCID TULTAct - 24.34 = 24.34 s 49.66 's N.B. IF OC SUYYLT. DO NOT 50. T 7 -(;7,gg,)2 0.49 STALLED TORQUE = HOT. STALL TORQUE x'ST. EFF. I x O/A RATIO a e 110% 70LTAct 91 x Q.-50 x 70.93 x 1:21 a 3090 . '#
R/W PULI,= 2 x STEM TOR UE = 2x 621 '
. 73 g R/W RA%TO x UNIT EFF. x R/W DIA. 28.J/ x U.J x 4 MAZ. TORQ. SW. SETTING s !CT. TORQ. x F/0 EFF.
- AFF. FACTOR x O/A RATIO:
(ftED. 70LTAct) 80 'x 0.49 x O'.4 x 0.9 x 70.93 a 1001 ep i I MAZ. 3/ m TORQUE = MAE. VAL 7E TennUI R/W AATIO x ETT. M
~
- 28. 7
- 0.3 8 '8 OPERATNG TIME = (60 x LIFT) + STIM SPEED = 17.0 seconds. (17 Sec Max; SMB - 3 OPERATOR WITE 80 FT.# MOTOR. MAE. TIRUST: 140000 #
t- '# MAZ. STEM TORQUE = 4?nn 0/A RATIC RANCE = :3.9-95.5 E/W RATIO 8 22.17 81 ADD GIAR - ::, MAE. STEM DIA.: 5.0" CURRENT.SUFFLY 460 AC 70LTs 3 PH 60Cy MUST CPERATE AT 70 tvCLTAGE , _ . 1 0 REY. l 1 H 2 l 3 l A 5 l 6 j COMFILID BY: h. y n l . ' l
. AFFROVED 3'f: /Kk.. w hF i L IND. RZ7.3Y: dMJac V i l i l l 1 j 472-11-76, RZv. 1
{
.~\
w .- a, m.w., -.em . -ve.. em >=meno -----==om- -==-===s = ~ ~ ~ ~ * * = - * = * = ~ ~ ~ ~
-.v-----.----_.--.m . ,- , - - - - - . _ . _ _ _ - . . . . ,,.,.,,._m.,____-r. ._.-w...-y_ .% -,,.._,,.,w,- % .,_._,,w.-,,_,. - , . .
M:rTOR CPTRATOR CALC %TTCNS _ F.C. NO. OR PRcJECT: LRCO - RIT.: 92-3287-MC/25 . STONF A WFR W CTSTCHE1 NAME: * , 7E:AN NO.: P2-3287-N _I ZMS: 2E (TAG 1F41**MVndpi VELAN. DUG. NO.: YAL71 DESC: 1 " E LES. na CATE i LINE PRESS: 1117 ?$1 0R:7. DIA.: 7 A74 01ll:T. ARIA: dA FA> AP 1114 PS: 9 T23F. cA1 *F. STEM DIA.: 2.5 STIM AREA: 4 006 TED:RP 1 L LITT: 9.0 STIM TIRTST: 0.A. x AF x SIAT TACT.:48.682 x l.135 x 0.3 = 16576
. LINE FRESS. x 3.A. = 1337 x 4.906 = 6559 ,
Facking Tractica Load = 2500 Total'5tes Thrust = 25635 # ST *., TORQUI = STIM'TIKT37 x STZM TACT. 25635 x 0.02893 = 742 '#
. 0/A CR UNTT RA =0
- MD cR Dts:CN R.F.M. = 1900 = =
S IM SPIID 3. /MD. ,g,gq , q1 a
- TEEIAD LEAD 1 /1 . .
MC"CR CALC. TCRQUE = . STEM TURCUE = 742 ' FUL:. CUT ITT. x APPL. TACT. x O/A RAHO Q.4 x 0.9 x 53.4 McTOR CALC. TORCUE 4 RC UCID VCLTACI = 38.'60 . 38.60 48.25 '# N.B. IT OC SUFFLT. DO NOT $0. Z V. -(g9,gg,)4 g,g . t STALLID TORQUE = HCT. SULL TORQUE x*ST. ETT. Z x O/A RATIO = 0 110T TCLTACE 80 x 07 5 x .LLL.,.x 1.1 2350 ._ ' s 2 x STIM tor 0er . 2x 742 130 ,. af, m , R/w 24=0 x UNIT ETr. x g/w OIA. . 25.3x 0.3 x 1.5 f MAz. ToRQ. Sv. SET =NC
- NJr. ToRQ. x F/0 E77. x AFF. TACTOR x O/A RAUO:
(gRzD. vcL. ACE) 60 x. 0.8 : 'O.4 x 0.9 x 53.4 923 ., MAz. R/vEzz:, TORQUE = MAZ. VAL 7E TORMUE R/w RAno x LTT.
, 3100 .
25.3 x 0.3 8 40a '# OFIRA=No T:ME = (60 m LITT) + STEN SFEID . 16.0 SECONDS.(17 SEC MAX) SMB - 1 OPERAZUR VTTI 60 FT.# MOTOR. MAE. TERUST ' 45000 # MAZ. sT2n TeaQUE = 850 'i 0/A RATIC RANGE = 27.2.- 171.6 I/W RATIC s pg 131 ADD GEAR - :L MAX. STEM DIA.: 2 7/8 CTRRENT 57FFLT 17E vnf' #ULTS - C MUST OPERATT AT 90 - 270tTAGE O REY. 1 l 2 l- 3 i & l 5 6 ,_d ~ ccMFILID ET: Ms.. ., Jr i l i AFFROVID BY: #fE4. y (j '{ l
- l l IND. RIV.BT: W o .* M l l .
l 1 I i 972-11-76. RIV. 1
.. .ei 1_...-.___._. _ ._
l [ .
icTOR OPERA *OR CA:.CU'ATTONS
. y.0. NC. On ygC :c : LILCO -
RZI.: P2-3287-MC/32 , CT37CIE1 NAME: - STONE & WBSTER
, v!LAN No.: P2 17A7 N CIMS: 32 (TAGt 1E51*MOV041) vt:.AN., DUG. NO. :
VALYZ TESC: 1 " onn L33. RR CATI ' - LINE FRZ33i
. 1337 PS2 CRIF. DIA.I 7 s?q CRIF. ARIA: 5.409 AF 1135 PsT e TZMP. 563 + 'r.
ITI:it DIA.: 1 1/8 STZM ARIA: 0.994 TID 1/5 F 2/5 L LIrT:3.12. = 4 STEM TIRTIT: 0.A. x a? x SIAT TACT.: 5.409 x .1135 x 0.3 = 1842 .
. L:NZ PRZ33. x 5.A. = 1337 x 0.994 = 1128 ,
Facking Tnc n on Load = 2454 Total'5tas Thrust . = 5424 # STZMTORQUI= STIM"TIhtUSTxSTZMFACT. capa x n n1715 = 65 '8
. 0/A CR UNTT RATIO = McTUR ttst:N R.P.M. = 1700 = =
STIM 57EID Dt. MCN. 17_ igg , gg g
- TIZIAD LIAD 2/5 -
McTOR CALC. TcRQUE = . STIM TURCUT i
= 66 FULL CUT UF. x AFFL. FACT. x O/A RATIO
= 3.29 9 0.4 x 0.9 x 55.8 g
McTct CALC. TcRQUE e RZDUcID TcLTAst - 3'.29 = .3.29 = 6.71 '#
,2 N.3. T? DC M T. DO NO" 50. T 7. g;7,gg,)4 ,g,49 <
l ST.41LZD TORQUE = MDT. ST.A1.L TDRQUE x*3T. E77. I z C/A RATIO a ! f 1102 vetm 11.0 x 0.-5 x 55.8 _x '1. 21- 3 371 .# ' E/W FC:.L = xSUM m M = 2x Es = 158 # R/W 3A UD x UN C E27. x R/W DIA. }x } z 0,g34 MAZ7 TORQ. 57. IITTTNC a MDT. TORQ. x F/0 ETT. x AFF. FACTDR z C/A RATIO:
~ '
(FRED. TCLT. AGE) 10 x Q z O.4 x 0.9 x 55.8 s 98 't MAz. 1/uIZZZ. TORQUE = MAZ.' TALTE TUR"UT E/W RATIQ x s.a s . 17q
"_.- jx j z 375 '#
OPERAT2NG T2ME = (60 x LIFU + STIN SPED = In SECONDS. (APPROX).
- SMR - 00 0F13A201 UTTI 10 FT.# Mirrot. .w. TIRUST: 14000 #
. MAZ. STEM TOEfgt = 98tn '# 0/A RATIO RApog a pq ino I/W RAT 2D = 1 11 ADD GEAR - :, MAX. STEM DIA.: 1 3/4 -
CURRENT SUFF*.T * ' 480 AE #ULTS 3PH 60 C Y MUST OPERATE AT 70 270LTAct 0 REY. 1 2 l 3 1 4 5 l 6 . GDl'TILID BT: % ,.o y l l l . ,,..
. /Ifi.?'.se V AFFROYED BY: l l l j UtD. RZY.BTr (FQ"I M * , l l l l .
- l l . ,
n2-u-7s, Rzv.1 I
' ? .
1= '. v r. L
- c. -
=t.
h-2 .-
-~
-. ..-. . ~._ - . . - . . . . . . . . . . .
" ~ "
NtricR OFDA*0R CALC"" A*!ONS F.C. No. OR FROJECT: LRCO - RIT.: P2-3287-MC/59. . 1 C 3TCIC2 NAME:
- STnNF' t lifFR9TFR '
VELAN No.: p?.17A7 N N *s 44 (tan 1F41*MnVnd? VEL 1N. DUG. No.: l 7ALTE DESC: _T 900 L35. BB GATI i *
. LINI FRE.33: 1337 PSI ORIT. D u.7 2.625 01:r. ARIA:5.409 AF 1135 F57. G TZMF. 563 'r.
sTzu D u.: 1 1/8 sTzti ARIA: 0.994 TID: 1/5P 2/5 L Lzrr: 3.12 "
! ST:!i TIRTST: 0.A. x 47 x SIAT TACT.: 5.409 x .1135 x 0.3 = 1947 l
. LINE FRISS. x 5.A. = 1.117 x 0 QQA = 1128 !
Facking Tractice Lead - 2454 l i Total Stan T.. ust . = 5479 8 l ST221 TORQTI = STIM TIRT3; x ST22t FACT. 5424 x 0.01215 = e:s: '8 i l
. 0/A CR UNTT RAC O s NCTOR DISIGN R.F.M. = 1900 - =
STIM s7 TID m. /M::N. 12.064 . 63
- II13AD LEAD 2/ 5 .
MCTOR CALC. TORQUE = .
. i 5TTM TCRCUT i e 66 FULL QUT E77. x AFFL. TACT. x C/A 3An0 o,4 x 0.9 x 63
= 2.91 'l
( HDTC1' CALC. TORQUE f EZDUCID TCLT. ACE - ? 61 = 2.91 s 3.64 '# N.B. IT DC 5"JF?LT. DO NCT so. I v. .g;9,gg,)4 , g,g < su LLED TORQ:Z = MDT. suLL TORQUE z*ST. ETT. I x O/A RAC O s e n0T TULuCE 8.25 x 0.5 x 63 x 1.21 s *286 's gju ye::, . 2 x sn M ToROUT = 2x nn = 32 # 4 R/W RAIIC z UNC ETT. x R/W DIA. 4.37x 0.95 x 1 MAZ7 TORQ. SW. SIT =NC 8 McT. TORg. x F/0 ETT. x AFF. TACTDR x O/A RA= 0:
~
(fRZD. TUL* ACE) 73 z' M x 0.4 x 0.9 x 63 s 136 'f l + MAE. R/WEEII. TORQUE = MAX. TALTT TDarset ) R/W RATIO z E77. i . -
. 17R 4.37 x 0,95 x on '#
OPERATING T2ug a (6C x LIFT) + STZM sFEm = 16 SICDNDS.,( APPROX 1 i4 # 9e - 00 OPERATOR WITI 7.5 TT.# MCTOR. MAZ. TIRUST: 14000 # ls ' '
, M&Z. 3333 Togggt = 250 '# 0/A RATIO RANCg a 23 - 109 , R/W RATIO a 4 37 21 ADD CEAR - ::. MAZ. STEM DIA.: 1 3/4 CURAINT.SUFFLT
- f 25 Vfr POLTS - C MUST OPERATI AT 20 170LTACE ,.
I i O RIT. l 1 2 l 3 1 ' l $ 6 - C0!TILID 37: pf,a at l l l . AFFROVED BTs /Vs. le. v .fsh- *l 'l l l IMD. RZv.3T: WiUl 1- . I l l l l . i . ,
#72-11-76, Arv. I t
l . t . .me.
un , -a .o......
- "y.VELAN ENrNEERING COMP ~ Niigg 0 0 41 '3M
[ 9CC -#'#* i CERTIFICATE OF NOT APPROVAL l [2-3 2 ci 2- X/ 7 d g g < /* e8 . ' " REVISED EDITIO4 - FEB.4/85" vouso a uaomstic i PcNcTRANT l P49ffiCLc ! PART SERI Al. NQ: CR HCAT CQoC
- I Acc cq" e I
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l 0 Y/ / $ ?i$ soor surtwcLo { , p l' so==cr a sac.mscar(coven) ,hEv# /
) .I l ac/ 7 0 W..* /O I /979 x ' , Ifit i misc (wcoes) [g / .
sTcM svuos a r wcLo scars ' M"' 7 amo ouses .
/f7f nanovsce
- l Kf 7 scAf .
S [f/f
. ....c c o se. Acp.
s w cooci Sfo z
. gle, li Leak-off Pip;e Welds to Bonnet. o ee.u .
/ h -- W I
CERTIFICATE OF PRODUCTION (HYDROSTATIC) TEST . .' rync or vrst l su ct.t. l ...=^r....,,, l eacx scar l paculNG
. ,~s.l s'o nj.uS i & m).us. ! So wM l f.nk.d
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..c.! c fe,c,4.o;c ! / u- f -2 c c- ic en/Lceio & k e TEST DATA FOR MOTOR - OPER ATED V ALVES l cars or Tsst +
.openatomtree gm g ,, g g narco volts Q{: 12s J N 2 71975 z' semea6 wo. /795 7y F.'.'c's'."n" /, / Sf VELAS
. / e' [/ ffyg g/ ov2 vu aa'== *a'tscio% aa'== va're 4%M r=='==
H.T i j
' vou s to onc= (sces)
, vi c to c6oss (secs)
/g 7m , ,' /[ j Af., , h .
/g g A4, ! j/g, 9 g, , ,.
' eran svantino cumacar (apes) /j g g,,,g, i / / c sc 4 ,,, y .
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,, # #M. ,
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.--.,_ .e... -- . ~ ~ . - ~~~~- - ~ ~ - * ~ = ~ ~ * ' * * * ~ ~ ~ '
1
~
.. :5~
- o .. . . . . .. ..
)CTOR OPERATOR CALCOLATICNS - '
~
F.C. No. OR PROJIOT: LILCO - RI7.: P2-3287-MC/38 . CTSTCMER NAME: STON8*Iltr N TFW VILAN No.: P2-3287-N TTEMS: 38 (TAG 1G33*MOV030A/8 d VI:.AN.CWC. No.: 900 L33. BS CATI '. VAL 7E DESC: 1"
. L 3I FRESS: 1175 PSI C1:7. DIA.I. 1 da 0137. ARIA: 4 79 AF 1010 'FST f TZMP. 563 *T.
STES D:A.:' 1.175 STEM ARIA: 1.484 TID:R F 1/2 L LIFT: 4.0 = l
$ TIM TIRCIT: 0.A. x 4F x SIA: TACT.: 9,29 x1030 x 0.3 . 2871 LINI PRISS. x 5.A. = 1.174 x 1 4R4 . 204' Facking Fnction Load = 272' Total Sten Th mst = 7611 #
STIM TORQUI = STIM TIRUST x STIM TACT. rc 71 x 0.01404 = ild '# 0/A CR UNT! RACO = MD 02 DESIGN R.F.M. = 1700 = = l , STIM 57tc u. / c . 11.806 . 72 j
- TERIAD LZAD 1/2 .
McTOR CALC. TORQUI = . STIM TcRect = 114 FU:.:. cc; 577. x APFL. TACT. x O/A RACO 0,4 20.9 x 72
*
- MCTOR CALC. TORCUZ f RZDUCC VCLTAGE = = s g,gg 't N.3. T7 DC 507FLT. DO NOT 50. I 7. - ( .,.7, gg , ) 4 g,49 STALLD TcRQUI = MCT. STALL TORQUI x*ST. ETT. : x O/A RAMO a 6 11C: VCLTAGI 17 3 x 0 5 x 77 x 1 21 s 762 '#
' 1 E/W F'~"* '~ = 2 w STEM TCROUT = 2x 114 '
. gz g R/d RAno x UNIT 177. x R/W DIA. 4.37 x0.95 x 1 l l
MAZ. TURQ. 37. SITUNC x ICT. TCRQ. x F/0 EFT. x AFF. TACTOR x O/A RAUQ I (fRD. 7CLTACZ) 15s M x O.4 x 0.9 x 72 x 191 's I i MAZ. n/unzzI, TORQUE = MAX. VALVT TORnUE R/W RA U C x LTT.
. 19n
- l
' 4.37
- 0.95 x 84 *#
OPERAT3 G TZME a (60 x LITT) + STEM SPEED = 21 SECONDS. . APPROX. Um - *nn CPERATOR VITI 14 TT.# ICTOR. MAZ. TERUST 14000 # MAZ. STIM TORQUE = 740 _f' C/A RAU D RANCE a 23 - 109- l 1/V RATIC a 4.17 :1 ADD GEAA - :L MAZ. STIM DIA.: 1 1/4 CURRZNT.SUFFLT 460 AC #CLT3 3PM 60Gj MUST CPERATE AT 70 "2TCLTACT l 0 RET. l 1 2 l 3 1 & l 5 l 6 '
- COMFILID BY: ' ff y,.4,.gt l l l l . :
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NDTOR OPERATOR CALCUIATIONS i P.O. NO. OR PROJECT: LILCO REF.: P2-3287-MC/39 . CUSTOMER NAME: STONE & WEBSTER
, VIIAN No.: P2-32A7 N ITEMS: 30 (TAG
- 1n_11**40V0111* VELAN DWG. NO.:
.l .
VALVE DESC: 6 " 900 LRS. EB GLOBE LINE PRESS: 1774 PSI
- i ORIF. DIA.: 4 3/4 ORIF. AREA: 15.074 AP 1cin PSI 9 TDIP. 563 F.
! STEM DIA.: 1 3/4 STDi AREA: 2.404 TED:_ lij,P E L LIFT: 2 1/2 "
STEM THRUST: 0.A. x 4P z' SEAT FACT.: 15.025x 1030 x 1.1 = 17024
~
- k. Pacic ng Frict1on Lead * ' =
- 43 3 0' ' ' "
Total Stem Thrust = 21354 i STEM TORQUE = STDi THRUST x STEM FACT. 7119d x, 0 0111A = 7PN 'I 0/A OR UNIT RATIO = MOTOR DESIGN R.P.M. = 1700= = i STEM SPEED IN./ MIN. 4.05 84 i TEREAD LEAD ]/3 MOTOR CALC. TORQUE = STEM TORCUE = 286 " 0 PU'LL QUT EIT. x APPL. FACT. x O/A RATIO Q,4x 0,9 x 84 .I MO OR CALC. TORQUE 9 REDUCED VOLTAGE - N.B. IF DC SUPPLY, DO NOT SO. I V. 9.46 = 9.46 = 19.31 '# (gyngg,)2 4 "' 0.49 I 1.LI.D TORQUE = MOT. STA1.L TORQUE = ST. EFF. I x O/t. RA!!O =
, G 110% VOLTAGE 4R x 0.5 x R4 x 1.21 = 2439 ' #.
} E/V PULL . 2 NM = 2x 286 . .gg g H/W RATIO x UNIT EFF. x H/W DIA.
1 21 .1
- 0.3
- 1.17 ,
f 12I. .TORQ. S*J. SETTING = MOT. TORQ. x P/O EFF. x APP. FACTOR x O/A RATIO: j (fRD. VOLTAGE) 40 x 0.49 x 0.4 x 0.9. x 84 = 593
.i is l!
MAI. H/ WHEEL TORQUE = MAI. VALVE'TOROUE H/W RATIO x IFF. 2700
,i
- 21.1 x 0.3 x 427 't
-I 0 PIRATING TIME = (60.x LIFT) + STDi SPEED = 3R J SECONDS. ,
(APPROX) i SMB - 0 OPERATOR WITH 40 PT.i MOTOR. MAI. THRUST: 24000 # Uf - MAI. STDs 704QUE = 500 '# 0/A RATIO RANGE 8 26,4 - 150.8 j *IV RATIO a ' 11:1 ADD GEAR - 31 MAX. STDI DIA.: 2 3/8
"{ CURRENT SUPPLT' 4Kn Ac VOLTS 3PM 60 CY MUST OPERATE AT 70 IVOLTAGE h 0 REY. J 1 2 3 4 5 6 - -
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,- F.C. NO. 01 FROJICT: LTLCO RIT.: F2-1287-#C/41 .
COSTCHER NAME: - tTnNT t UFRCTFD v!:AN NO.: P2-3287-N ITIMS: 41 (Tag: 1G33*MOV033 VI:.AN DUG. No. : , CATE 7ALTE DESC: L " E L35. RR - i LINE FRISS: 1175 FSI CRIF. DIA.I (,, E 01:F. ARIA: 21.14 AF 1,165 PST G TZMP. 563 'F. STEM DIA.: 1 3/4 STIM AREA: 2.404 TED:,143,P XL LITT: 5.75 5 TEM TIR 3 : 0.A. m ar x stat TACT. : 21.14 x .1765 x 0.3 = 7388. f
. LINE FRESS. x 3.A. = 13.75 x 2.404 =- 3306
- Facking Fri.-tion Load = 2b00 i Total'5taa Thrust' = 13194 #
STIM 701 QUI = STIM'TER s x STIM FACT. 13194 x, 0.01733 = 229 '8 l . 0/A CR UN O RA O D s MDTOR DESICN R.F.M. . 1700 = = STZh 5? TID IN . nCN ._ 11.806 . 72
. TIRZAD LZAD 1/2 . .
MDTC1 CALC. 701 QUI = . j STTM TOROUT e 220 FIT;:. OUT ETT. x APPL. FACT. x O/A RAT *0
= 8.85 ,f 0.4 x 0.9x 72 HCTOR CALC. TORQUE 4 REDUCID VOLTAGE - R Rh a 8.85 3 18.05 '#
N.B. IT DC StTF?tT. DO NOT $0. 7. -(;y,gg,)4 0.49
- sTA
- .LZD TORQUE = McT. ST.ALL TC1 QUI z's!. ETT. I x 0/A RAno = ,
j e 110; 70LTACI 29.Q x *0.5x 72 x 1.21 = 1263 . '# I/V Ft "' = 2 x m Tom m a 2r 229 . 110' s I i R/W RAno x UNC ET7. x E/W DIA. 4,37 x Q,g,$x ) : MAZ. TCRQ. SV. SITCNG s tcT. ;01Q. x 'F/0 E27. x AFF. TJ.CTOR x 0/A RATIO: ,, (FRED. 70LTACZ) 25 xQx 04 x 0.9 x,72.0 a 318 's !' MAZ. 1/UIZZL TORQUE = MAX. VALTE TORMUT i R/W RAno x TJT .
= 1" *
~- 319 '8 4.37 5 0.95 j - OPERA =NG TZMg e (60 x LITT) + STEM 37EZD = 29.5 _st:DNDs.(APPROX).
SMR - nn CPERATOR UITI 75 FT.# ICTOR. MA :. TERUST 1&nno # NAZ. STIM TORQUE = 9Cn 'i O/A RATIO RANGE a 23 . _10q l R/W RATIC a 4.37 :1 ADO GEAR - :L MAZ. STEM DIA.: 1 3/4 CURRENT.SUFFLT 460 AC PULTS 3PH. 6(X:y MUST OPERATE AT '70' 2TOLTAGE i 0 REY. 1 2 l 3 1 4 l 5 l 6 ,* i COMFILID BT: #5. .ts ' I I I I ,, , AFFROVID BY: t/1 A.,. #r *1 l l l ! IND. AIV.BT: kF9 # n. . l l l l l i i 972-11-76. RZv. 1 l l I e b l 5 - ( l y 4
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j , .. .. i ICTOR CPGAT01t CALCOLAUCNS . F.o. no. oR FRcJEcT: 1. 11.0 0 RIT . ." P2-3287-MC/42 . CU3TQHER hME: . sTnNr A M A RT M
, TELAN No.: P2-3287-N CEMS: 42 (TAG? 1G31*MOV0341 7tLAN.,DUC. NO.: ,
VALTZ DESC: 1 " g ,.RS. L RR CATI - LINE FRZ33: _lut ,FSI CRIF. DIA.: 5 1/16 01 7. ARZA: 21.14 AF 1155 PSI f TEMP. 563 'F. 5TZM DIA.: 1 1/4 STIM AREA: ? and TID g F L L LIFT: E.75
, STEM TERUIT: 0.A. x AP z SEAT FACT.: 21.14 x 165 x 0.3 = 7388 ; . L:2ft FRZ33. x $.A. = 175 x 2.404 = 330fi , !
Facking Friction Lead = 2500 Total'5 ten Thrust = 13' 9L # S m TCRQUI = STEM'TIROST x STZM FACT. 13104 x 0 01738 = ppq ' s , i . . c/A CR UN C RA CC a ND*OR DESIGN R.F.M. = 1700 = 1700 = sun s71;p :24./c. 11.806 23,o14 72
- TIRIAD LIAD 1/2 . .
tcTC1 CALC. TCRQUE = . STEM Totect = 229 * .I l FULL CUT m . x AFFL. TACT. x O/A AAUO 0.4 x.0,9 x72
,- , !cTOR CALC. TORQUE f R2DUCID TULTAGE = A' A4 = 8.85 = 11.06 '#
l N.B. IF JC st??LT. Do McT so. 2 T. -(gy,gg,)2 0,8 STALLZD TCRQUE = McT. STALL TotqUZ s'5T. EF7. I z c/A RAno =
~
f 11CT YULTAct 20 x 0".5 x 72 x 1.10 s 792 . 's j , gjy n . 2 x $~EM TORCUE = 2x 229 . 110 , 1
. 1/W RATIC z UNC EFF. x R/W DIA. 4.J/ x 0.95 x. )
; MAZ. TURQ. SW. SETT:2tc a ICT. TCRQ. x F/0 EF7. x AFF. TACTOR x O/A RATTO:
(GRID. 70LTAGE) 15 x* g x 0.4 x 0.9 x 72 a 311 ' s MAZ.1/2 TORQUE * ' MAZ. VALTT TURfME l R/W EAno x LTT. ,
= 1"C .
-; ~ 319 '
~#
~ 4.37
- Q.95
-j - .
cFERATING TIME = (60 x LIFT) + STEM SFEID = 20.5 SECONDS. (APPROX)
-t SMs - 00 oFERArom VITE 15' FT.# MOTOR. MAX. TIRUST: lannn #
MAZ. STEM TORQUE = 9Cn '# C/A RATIO RANGE 8 91
- tna I/W RATIO = 4.37 1 ADD GEAR - :1 MAZ. STEM DIA.: 1 114
-f CURRZNT.IUFFLT 125 VDC #ULTS C yISST CPERATE AT Rfr TTOLTAGE O RET. I 1 2 1 3 1
& 5 ] 6 -
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g' fgf@ ~" LONG ISLAND LIGHTING COMPANY , sHOREHAM NUCLEAR POWER STATION *l p.o. sox eie, wonvn couwvnv nomo . waoewe neven m.v. mos 8 Joe.= o. Leo =4ao.Ja. watt Pet 900m? emWCLI44 Ort AATIOne t July 2, 1985 SNRC-1187 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulati6n U. S. Nuclear Regulatory Commission Weshington, D. C. 20555 LILCO Corments on Prelir.'. nary Review of Shoreham PRA ' Study Shoreham Nuclear Power Station Docket No. 50-322
References:
- 1. Letter from W. R. Butler (NRC) to J. D.
Leonard, 1985 Jr. (LILCO), PRA-SNPS, dated May 8,
- 2. Telephone memorandum R. Caruso (NRC) and R. W. Grunseich (LILCO): extension to July 3,
1985; dated June 13, 1985 3. SNRC-1149 letter, dated February 25, 1985, LILCO Comments on Preliminary Review of Shoreham PRA Study; J. D. Leonard, Jr. (LILCO) to H. R. Denton (NRC)
- 4. Letter from A. Schwencer (NRC) to J. D. Leonard, Jr. (LILCO) dated January 24, 1985 Dear Mr. Dentont Enclosed Referenceplease 1. find responses to comments as requested in We trust this letter addresses the areas in question relative to Brookhaven's review of the Probabilistic Risk Assessment (PRA) Study. As requested, a meeting has been arranged 1985 for July 17, 1985 (structural analysis) and July 18, (CET development), Stone & Webster and Science Appl 3 ':ations l will support these meetings. , a
}
. Seg . %
= -
- 8t f- ,
\ -
l SNRC-1187 1 Page 2 If you require office. additional information, please contact this
't very truly yours,
/
~ jf
[ s r // nE7 Leonard,J[ ~ Vi e President - Nbj: lear Operations ( BK/ f Attachment cc: J. A. Berry, Resident Inspector T. Pratt, K. Perkins, BNLBNL, Department of Nuclear Energy e i o e e d' e 1
- - , . _ - _ . - . . - _ _ . . , _ _ - . - . - - ~ . - - _ . . _ _ - . - _ - . . , _ , , , _ , , _ ,. . _ - _ _ - _ . - , . - . , _ _ , _ _ . . . _ _ , _
i Attachment SNRC-1187 . Page 1 , I l 4 Request No. 1 , Table II of Appendix M gives different pressure limits for the longitudinal reinforcement bars at the base of the containment and in the wetwell region. However, the longitudinal bars appear to be continuous and should therefore have the same stress. : Please explain the basis for the different results. Response No. 1 The longitudinal reinforcing bars t the base of the containment are continuous with the longitudinal reinforcing bars in the wetwell region. At the base of the containment, discontinuity ? moments and shears are developed as a result of fixity between the containment wall and the base mat. These discontinuity moments and shears reduce rapidly with height above the mat as the affect of joint fixity diminishes. At a point where the wall moments and shears become insignificant, the membrane zone within the cylinderical portion of the containment wall is reached. The wetwell region of the containment is within the cylinders membrane zone, and therefore, experiences significantly different loading response than the containment base detail when subjected to internal pressure. Therefore, although the longitudinal bars are continuous, different pressure limits (i.e., different limiting rebar stresses) exist for the two areas of the containment due to the. variation in applied loads. t F O 1 I l
l'
. o Attachment SNRC-1187 Page 2 Request No. 2 i*. .
Table II of Appendix M indicates that the shear' bars at the base
- i and drywell head have the lowest pressure holding capability (121 psi and 120 psi, respectively) but the discussion indicates that the additional reinforcement will preclude this failure mode.
Since the containment failure mode is a key ingredient of the release estimates, please provide a quantitative estimate of the additional shear strength provided by the non-shear reinforcement bars. Response No. 2 - At the base of the~ containment, the following values have been developed:
- 1. Shear capacity (considering dowel action) 168 psi
- 2. Flexural capacity (longitudinal bars) 134 psi At the base of the containment, the discontinuity bending moment is not required to maintain equilibrium. Since the base has a sufficiently larger shear capacity, the section will rotate without failure beyond 134 pri.
The drywell head ring beam'han been re-evaluated with
- consideration for the actual concrete strength of the section.
Based is a minimum on a value of 145ofps f'f.= 5500 psi, the ring beam shear capacity At approximately el 43 ft-0 in, the wetwell region is fully cracked in the hoop direction at 130 psi. Since the hoop bars are required for overall equilibrium of the containment, this area represents the critical section of the containment wall. The referenced attached Table II has been revised to reflect the above values. 9 i
i i a Attachment 1 SNRC-1187 , Page 4 l- *EN ! TABLE II h M" *
- wh&
k Limiting Pressure at Various Locations on the Containment .5 . I: "E i Locationa Failure M e Limiting Pressure ) poi At Base of Containment Yielding of Shear Bars a Yielding of Longitudinal Bars 168 (1) * - Yielding of Hoop Rebers 134 (2)
- a 283.0 s 1 Wetwell Region Yielding of Longitudinal Bars Yialding of Neop Rebers 149.0 130.0 (3) ,
" Cone to Cylinder Junction ,
Yielding of Longitudinal Bars Yielding of Noop Rebars 270.0
- 174.0 Drywell Region 3
1 - Yielding of Longitudinal Bars Yielding of Hoop Rebars 200.0 174.0 Drywell Head Ring Beam j Yielding of Shear Bars Yielding of Hoop Rebars 145 (4)
- 140.0 -
i l l I (1) ] (2) When dowel action is considered ' j (3) When Limiting bars are allowed pressure at el to 43yield ft-0 due in to flexure, and considering f' s j (4) Considering c = 5500 psi i f', = 5500 psi
- j
- Revised Valves l 'I
i t i Attachment SNRC-1187 Page 4 Request No. 3 If shear failure is precluded as discussed in section 3.2 of Appendix M, "it appears that the ultimate capacity is controlled by the yield of the longitudinal and the hoop bars at about 123 psi." These two failure modes appear to be very important to subsequent fission product release (particularly for Class IV ATWS) since they will both occur in the wetwell region. Please provide an estimate of the size, location and direction (vertical or horizontal) containment failures for each of the three possible failure niodes. Response No. 3 As described in the response to Request No. 2, the critical section of the containment wall is at approximately el 43 ft-0 in, with failure the result of hoop bar yielding. Hoop bar yielding would indicate that the probable containment wall failure would consist of a vertically oriented containment wall crack and liner tear. 4 1 s I o
. ~ .
- i
. . . O i
L . . . Attachment SNRC-1187 Page 5 Request No. 4 Section 3.6 of the PRA takes credit for containment leakage which will prevent gross containment failure for all pressurization rates except the very rapid pressurization associated with large treaks. However, the structural analysis by Stone and Webster (Appendix M) did not identify any significant source of leakage. The basis for the expected leakage source and the leakage rate as a function of pressure should be provided. Response No. 4 The Stone and Webster structural analysis (Appendix M) concludes that the ultimate pressure capacity is limited by the concrete containment. The study also concludes that prior to containment failure, ther.a may be small leakage through penetrations, valves and hatches. Although these leakages cannot be quantified, uncertaintier allow a range of possible leakage sizes that would relinve the current gas and steam generation such that the containment pressure would no longer increase. This assumption formed the basis for the lower limit or expected leakage rate as a function of pressure (or accident sequence) at Shoreham. For example, forClaes2accidentsequences,at30hoursintothe transient the total decay heat is about 4.8 x 10 Btu /hr. If all this is used to produce steam at the current containment pres.sure of 80 psia, the steam produced would be 14.7 lbs./sec. The hole
-size that would vent this amount of steam agd preclude containment over-pressure is about 0.147 ft . This corresponds to a circle with a diameter of 5 inches.
For Class 1 accident sequences, the containment pressurization beyond design basis occurs after the core degradation and vessel breach. Some of the debris would be involved in core-concrete
- interaction, hence, the amount of gasses and steam generated could be less because much of the heat is absorbed by the concrete. So a smaller hole would be adequate.
As the time past shutdown is extended and the decay heat decreases, the lower limit of the leakage rate required to prevent gross containment failure would also decrease In the Shoreham analysis, it was judged that the required hole size was not excessively large such that leakage around penetration sealants and hatches would be adequate. However, the exact leakage source and size were not quantified. - 4 I r e L
4 .
- 6 .
}'. . . .. . o i . Attachment SNRC-1187 Page 6 , l Request No. 5 . The basis for the partitioning between release category 10 and 11
- i (no pool bypass vs. partial pool bypass) should be provided. The phenomenological basis for the estimate of only 104 bypass should h provided. Preliminary results from ZDCOR indicate that for j
some BWR sequences the vessel will fail with only 20% of the core molten. Presumably 80% of the melt release would bypass the SRV's and be released into the drywell. j Response No. 5 1 suppression pool bypass is an explicitly featured event in the Containment Event Tree (CET) and is defined in Appendix N as "The j core meltdown proceeds in a manner such that radionuclide 4 releases (in-vessel) are directed to the suppression pool." i
' Success (no pool bypass) implies reactor coolant system (RCS)
' failure occurs only after the entire core melt release component
' of the fission products has been directed to the suppression pool. Failure (pool bypass) implies that all or part of the core melt release inventory bypasses the suppression pool.
1 Pool bypass was considered in the CET on the basis of either of the following being trues (1) There exists a break in the RCS such that the coolant and radionuclides are discharged to the drywell. This coul.d be either a large or medium break involving a significant pool bypass (LOCA) or a small break, which, with successful SRV j actuation results in competing flow paths leading to a , partial pool bypass. (2) The core melt progression is non-coherent and leads to RPV bottom head failure prior to total core meltdown. This i l allows a bypass flow path for the radionuclides remaining t within the vessel at the time of vessel breach. (3) Some radionuclides are airborne or resuspended in the l ' primary system at the tima of vessel failure which are then released into the drywell bypassing the suppression pool. , j The conditional probability of pool bypass was estimated using ' the Boolean combination of two factors: the probability that the core meltdown proceeds non-coherently and the probability that the primary system fails prior to total core melt. The first factor was used chosen to be .90/ demand (similar to the conditional
! probability in the Zion PRA). The second value is chosen to M 0.40/ demand for cases where the RCS and primary containment is '
initially isolated and 1.0/ demand for cases involving a LOCA. These result in conditional failure probabilities of 0.40 for SPB ! L in non-LOCA and small LOCA sequences and 1.0 for LOCA (medium and
- large) sequences.
'l i l 1 t l ,
.: - - . _ - . - - - - - - - J
- .. . , ~
. . . .- . o Attachment
- SNRC-1187 Page 7 .
) It was determined that the significant contributor for non-LOCA cases was the early failure of the versel bottom head prior to , total core meltdown. The amount of fission products remaining in the BCS (airborne or.as yet to be released from the fuel) was i estimated from an evaluation of the extent of core melting by the i time the melt front of the core came within the proximity of the l BAF. (From the MARCH predictions, this occurred at about 40-50% core melt.) t '. Appendix D discusses the core melt release model used in the ! Shoreham analysis. A time dependence function for the melt releases was based on ORNL tests (Reference D-12). These indicated that the volatile materials release probably take place , j before fuel melting (liquefaction) propagates through the test bundle. On this basis, the time interval used to represent the period of melt release was determined to reach completion at the i, 50% core melt time predicted by MARCH. For those sequences involving early vessel head breach, 40% core melt would translate to 804 melt release. In other words a maximum of 20% of the volati3es would potentially bypass the l suppression pool. A 10% pool bypass assumption was chosen in the shoreham analysis since further examination of the core power
- distribution showed that the remaining fuel bundles tended to be 4
those with lower peaking factors (less than 0.50) . This assumption does not appear to be inconsistent with the IDCOR I analysis in terms of the amounts of volatile fission products remaining in the core at the time of vessel breach. The Peach Bottom integrated containment analysis (IDCOR Subtask 23.1) shows , that 14% of the volatile fission products (Cesium and Iodine) are i still remaining in the core at the time of vessel breach when 20% of the core is molten for ATWS or Class 4 sequences. i i s I I I l I e
~
~
i Attachment SNRC-1187 ' Page8 i Request No. 6 The basis for binning into release categories is poorly described
- and the transfer from Tables R4-8 etc. into the 16 release i categories is difficult to interpret. A table listing the l specific sequences which are binned into each category should be provided.
] Response No. 6 A perspective of the CET binning process may be obtained from Figure 1. This provides a graphical description of the CET development into sub-trees and the procedure used to collapse the sequences which were judged to have similar characteristics, in 1 terms of the release potential. As the figure indicates, the j "binning" occurs at critical pinch points of the CET. , It was necessary to perfcra the CET evaluation in this manner to limit the propagation of hundreds of possible end states which in the end would have to be grouped in order to make the 4 quantitative analysis moria manageable. i 4 i A sequence, therefore, may be designated by a series of intermediate plant states and end states, the progression of which can be determined by examining the subtrees of the CET j separately. Those sequence designators assigned to - characteristic accident class release events, for example, C g4g R T -DW for class 1, are shown in Figure H.4-0. - -
! The attached Table 1, (Sumnary Table of Sequence Designators and Associated Release Categories), illustrates specifica1.*y those j sequences, associated with a particular accident class containment event tree, that are binned into the respective i
release categories. It is important to note that the sequences (or end states) are unique to their respective accident class, and that each accident class does not necessarily contribute to all of the sixteen Release Categories. The sequences in this table follow the same format as previously used. The sequence is defined by the CET's initiator followed by the transfer state and finally the end state. Where only a single designator is found, , it would correspond to the and state for that sequence. e e # 1 l
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,a
*s =reweiste Ptamt j
starts 1 a 4 seveer . sweese a s , arts ei n e=st t,ane v,1 sem east i 33 it w e w ,,3 we w rue sessasssa trees a w s, := stan s, v wwe untums a sucnic uts w swee usuus . is weernwee er >= see uws := w car saints, aw amannne av marinire ss = start cowinema.
==essestatus, o.a.. -
ets w si.e.: . ,,8 e ec., e r,8 - FIGURE 1 Conceptual Representation of CET Sinning Process 9
. . . . . . . .. 6 .
- Attcchmert . . ,
SNRC-1187 , Page 10
- CONTRIBUTIIG ACCIDENT' CLASS END STATES I II III IV V A2/038 31/046 Of A)/02
- i n%;
A2/044 nm 87/048 --
!!* ;;g.=0 ti:ga,,
' ' 011/02 in- u u A3/042 810/036 84(A)/D2 2 A3/044 Cl A3/037 84/033 4 84(Al/04 A7/048 C4/024 A3/038 84/034 84 A7/052 C7 012/03 A3/039 ' 86/046 A10/DS4 812/02
- A3/040 CS/024 2 86/048 - - - 012/03 14 A3/041 88/046 812/04 A3/042 88/048 81 05/08 A3/043 311/D51 5 83 B11/06 C1 C3 - . - _ . - -
A3/044 811/052 85/06 g CP/03 C4/03 81/C5 812/ml gg C2/06 C4/DS 81/C7 312/D$2 g, wCs Sn/en ,87/0,9
,, 10 8,9,/01,1
, 01 .. .. ..
10 84 2 82/C7 813/054 86,0, 6 88 A)/09 810 A /011 812/06
/ 2 0, N 1/010 glA/D12 83/033 9 h/05 /D5 C6/026 C2/09 C4/09 83/034 C6/028 C2/010 gg 84/031
. C4/010 .g3 C8/026 1
84/032 C8/028 33 ggfgg B6 A3/037 A!!/054 -- A3/039 . A!!/D56 811/06 AJ/041 811/D32 82(A) A3/043 811/034 2 84(4) A7/046 812/012 812(A)/D6 gg A7/050 812/034 87/09 A8/D46 C5 87/0,10 m II/Cl 82/C3 AS/048 C6/026 ggfg w g 51/C3 C2/DS 38/010 g 82/C1 C4/08 B2 ggi'a
'Al 88(A)/D10 gg A8/D$D A8/052 CE/D20 C7 12 ggfggg 8
89/012 w g 10 8 61 A I/06 010(A)/0!! 86LAf/08 910(A)/012 86 BID ff I hj#ffg 810/D11 810/D12 u st/Cl C2/DID u(Al m 812(Al m w SI/C3 C4/D3
* '**~~~~*""~
C1 e 12/C1 C3 811/01 C4/05 82(A/01 15 6 a 4 82/C3 C4/09 82(A /02 84 84(A)/D4 15 C2/02 C2/D4 C4/D2 C4/04 Bil/03 C2/D9 82(A;1/D3 B12 fA /01 g C4/010 gg 82(A;l/04 812 LA;'/02 . 82 512!'AJ/03 16 B2/C1/01 44 4 /01 0121.A;i/D4 82/C3 'DI A3/024 030/049 84 A /02 012 82/C$ BID /CSD B4 A /D3 82/C7 C8/024 I4/029 C8/D24 B4/03D 81/02 311/01 10 86 81/04 811/D2 88 82(A)/01 Bil/03 82(A)/03 811/04
~ ~ 15 82 'A)/01 83/02 012i'A)/02 0121 83/04 512A)/03 gg ns' 01/C7 D
09/049
- g:i !!'A)<=
g4 83/D29 89/D$D 03/030
~~ .
81/01 81/03 A!/037 08/045 16 B3/01 A2/D39 88/047 83/03 A2/ Del C2/02
- gg' A2/043 C2/04 B2/Cl C4/02 .,
B2/C3 C4/04 84/04$ 84/047 1 II/Cl 05/047 Ol/C3 1 87/045 01/C1/01 87/047 i 16 Ol/CJ/DI C2/02 e 82/C1/01 C2/04 B2/C3/01 C4/02 10 l
, as/04s c4/04 1 l _
1 e
o l
. 4 1 Attachment SNRC-1187 l 4 Page 11 1
, l q Request No. 7 The lack of R$ sequences in the release categories makes it i y apparent that these releases have been binned " downward" into the lesser release category R The basis for this " downward"
- binning and any other seq 8e.nces that are moved to less severe categories should be provided.
I Response No. 7 i The CET end states are defined on the basis of the various ,l attributes which impact the potential consequences of a release ! event, of which R and represent only a portion of a number of i-possible combinations. Rfor example, SNP-3 is a release category 7 represented by C1R T,- 7', a class 1 accident sequence end state
- involving a total 8ote melt (R with long term containment failure (T 4) occurringinthe8r)ywell.
] l j This end state was chosen as the representative release event
! sequence'for this category because of its higher conditional risk than a C1R,T,- 7' sequence. The end state screening process j indicated that the higher conditional probability of an R type I release did not offset the small incremental release frackion of the oxidation release. Therefore, R 5 type releases were binned ; with R 4.
In a similar manner, T and sequences (for classes 1 and 3) i were lumped with T , b3caus.Taf the higher risk impact of T ! Although T estimated $sc(and adme3 T ) sequences were found to have a lohe. ape fraction than'T , it was jud ed that the r i' potential consequences of a T thpeeventwoudnotoffsetthe j higherconditionalprobabilithofT2and T3 . Tables 1 and 2 (see Question 9) summarize the estimated escape fractions used for the , l t importance ranking for sequences, respectively. The source l } factors used represent the melt (Iodine), amidation (Ruthenium) and vaporization (Tellurium) release co'mponents. While the noble 1
- gases (Kr-Xe) are released in both R and
! not included in the importance ranki8g sin % sequences, M this radionnelideit was l t
i group is non-removable, hence would not impact differences in the ' escape fractions between the various sequence and states. ' i ! I l \
+
i.
)
i l I
- + sit- - - . .--- - - - -- ,.,r,w.g-- --,-,--------,.4- ,,-cu,.,,---m ,.mr-.w-..ev, m-,---- wwrm.= = , -
~
4 A d Attachment SNRC-1187 Page *.12 Request No. 8
\
Table H.4-25 appears to be incomplete in that it does r.ot include
~
sequences D6 and DS. The completed table should be provided. Response No. 8 The completed Table H.4-25 " Sequence Designators for Class V Release Event Sequences" is attached with the sequences D6 and D8 in the appropriate location. t / 3 P O
-. , 4 4
*
- Attachment .
SNRC-1187 ) Page 13 i l e
- e p%
p. l l
.h,4 a taas ists sa s its a I
\, j C;
jl - Ir tsj k
. 4 e 5 40 U
= .
1 us 8 - S s -
.M 1 %
a i W
$ I .
5 d a:
- n N I t
r 5 u fp af I rg o 6 ( W'l
, 5 -
E 4 3 en I w Q . .= f- . 4 E - - rg x I r. i i_ i 5 -
- gl=lt a
1 g . s I I 3l?! l1
!! ,,si! .66i=e6Lsil,reia ,!! !i r ri- '!f -
l' If 1
!Ii I
p I 1 l
. l 6
4 Attachment SNRC-1187 Page 14 1 l
?
Request No. 9
' The source escape fractions used for and state screening (Table
- 3.6-10) appears to be quite arbitrary yet it greatly influences the importance ranking. In particulars the use of I as the surrogate for melt release ignores the fact that there are noble gases in the melt release which will not be scrubbed at alls the use of a large scrubbing factor '(500) for C transients is inappropriate since most of the melt releas8 will be released dire:tly to a failed containment; the reduction factor of 0.01 for d'" failures is indefensible since the event tree precludes everything but large ruptures where the pool will be blown out into the reactor building at high pressures.
L Table 3.6-10 should be replaced by a table with defensible reduction factors. As a minimum the table should include a separate category for C, transients, which recognizes the defined sequence of events (containment failure before core melt). In addition, a detailed justification for each reduction factor should be provided along with the numerical results of the ranking prccess. This revised table will provide the basis for our independent importance ranking based on revised estimates of
- accident frequency and reduction factors.
Response No. 9 The functions, systems and phenomena treated explicitly in the containment event trees were defined on the basis of their potential risk impact, i.e., the product of the probability and potential consequences of the release event. The probabilities are estimated for each accident sequence progression leading to a release producing event (end state). The consequences of the release event may be expressed in terms of the public health impact or radionuclide source terms. In the CET development, it was apparent that a very large number of end stnces would be possihin. It was also recognized that to analyze each end state for consequences would be extremely expensive, if not unmanageable. The ultimate goal of the inplant accident analysis is to summarise this spectrum of release events into a relatively small group of release categories that can be used to estimate public risk. To achieve this goal, the accident sequences are binned into core meltdown release categories. In the shoreham analysis, . 16 such release categories were defined. These categories were defined through the use of an iterative procedure by which the potential consequences are estimated for various sequences, remaining sequences are conservatively grouped based on the similarity of sequence progression and release paths frequency estimates are compared to identify high frequency seq,uences in each group and source term calculations are carried out only for 1 the high frequency (and potentially high risk impact) sequences. I
. . . - . . . o Attachment SNRC-1187 Page 15 ,
The consequence measure used to define these categories is the estimated radionuclides release magnitudes calculated on the basis of source' escape fractions shown in Table 3.6-10 The
- I attenuation along a release fractions associated with the path phenomena. is represented by the escape The relationship between event treethe in various Figure 1.removal mechanisms The calculated is fractions escape shown inare the form of an e shown in Table 1 for Rg release type. For example, transient ,
. events have a characteristic release pathway involving an intact j RCS with the steam boil-off through the SRVs. The fission .
. products not captured by the suppression pool would become ! airborne within the containment. maintained, natural removal processes could further reduce theIf the containme radionuclide concentrations. Depending on the containment failure time after core melt, and failure mode, reduction factors for a release transient pathway can then be estimated within a group of events. For class 1 sequences, the containment failure time may range from the start of the transient event (T to long after core concrete interactions (T,). Ontheotheb) hand, class 4 sequences are characterized by a time phase T containment overpressure failure. boiloff to the suppression pool.BothsequenceclassesinvbiveanintactRCSwith Therefore, a CIR T F and state could possibly have a similar release magnit8dA as C4R T
'T end state Nowever, since thewasHence, estimated freq'aengy both may of C4Rbe T g grouped within a single evaluated to be higher than C1R T -Y , the re,presentative sequence T chosen for this releas8 dategory was Class 4, and C4R,he, In t Y final was then binned release with SNP-10 category sourc(or 11) release category.
e term calculations, the class 4 accident SNP-10 or sequence
- 11. was used in the MARCH / CORRAL calculations for It is important to note that the values in Table 3.6-10 are used only initially as a means of ranking the various and states for purposes of binning into release categories. The exact reduction factors for the representative sequences are calculated in Appendixby D.CORRAL on the basis of input information provided 6
I
i a Q :rg"n g EwQwo " 7*8 8
. I M *o 'o t
t E t I ss tt e t tea es t f 9 ne aas s ces ( lt c' o c
) , e ,
t *
- t aee 'o e e e e e e a e *a *o a a e'e*e s e e e a t a
s t t t ssss saeeeeesssssaaa e a e e 'o *o e e o 'e =e e e a *e'o se t e e e *e s e e e t emt e e ettr n a meae naanm on ttttttttttttttt aaannn,.aa tt e s s o o a s i s o o s o a a s. eaan ennr tetetttttttss ssssseeeeesssaseaaaaae ss nnenn wssaapsamaausm 0 ssos r ttttnttwntnt ceo eess en l l eecc ccecccececcecccecc ecccceccccc sssuasaizaa ecc ececcc-e o s , t e
.t aa em
*o -
s _ n . e r t sr e aa c wo eL _ .i - ,a ,n s
- ,i - -_ ,n - ,i
" ,n -" ,n e
t s e e s u ae s e ea r t na es r gI gg lI
. e g1 t g1 yl g1 g1 s
e s t t t s e a t
- c. s u
e 1l
- l} s w
n s ae tn i r m na a o ** = 1 . cr ! t
! s t
e i u g n : 1 u e - F s t : se 5 t em ns a n av
- u -
c u
. c a o s u n nt t e et
<a m
a _ o* n _ w s m a a a t e n a t e s t u
- w
= g a
e t
.m m e
e
'a e w
ee _ ss s i e - ew w e *
= tt u s ea o c e a t
. u c o s a r e s r e a n.r ut t r a
' tIl
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' a 2 ' l1
TABLE 1 *o en -> e :: r.
* %Mg an
~ :r ESTIMRTED ESCAPE FRRCTIONS FOR R5 #El.ERSES ae*0 '
#CS P004.
USED IN THE CET RANICING PROCESS $$r* SOURCE F8CTOR FRCTOR ESCAPE ING SCRues-RGE LERK- OVER-PRES CNT FRILURE TIPE PH BREAK BREAC LOCATION CET END
- SURROGRTEe RG E.*W TOTRt.
S WCS P SIZE COMPONENT *,'
- t. 73 T2 STRTE ESCAPE ca=============zzz.S .__.z:====================; T3 =====2- T9 BS DW WW FRACTIC -
= Ru T 1.00
.03
.20 .002 .05 a ----============z==W=W='=====sz==zzaser=1=zzzzzzz:==zz sezzzzzza: -
1.00 1.000 .05 1.00 CERT 1 d
.09 1.00 8.000 .05 1.00 2e-5 1.5e-3 9.Se-3 6.02e-t 1.00 1.00 .20 .002 .05
.03 1.00 1.000 .05 .50 C1RT1 &
.09 8.00 1.000 .05 .50 le-B 7.5e-4 2.2Se-3 3.01e-
.50 .
1.00 .20 .002 *
.03 1.00 -
3.00 8.000 1.00 .90 1.00 - C1RT1 g
.09 1.00 3.000 1.00 .90 1.00 3.se-4 2.7e-2 9.te-2 1.08e- .
.90 1.00 .
I.co .20 002 ' 1.00 .
.03 8.00 1.000 1.00 .90 .50 C1RT1 g'
; .09 1.00 1.000 .90 .50 2.Se-4 1.3Se-2 4.0Se-2 8.42e-1.00
.90 i-
.50 1 '
3.00 .20 .002 1.00 .
.03 3.00 1.000 .05 1.00 1.00 CIRT1 d 2e-S 1.Se-3 '
.09 1.00 1.00o 1.00
.05 1.00 . 4.5e-3 6. 02e=
. .05 1.00 i 3.00 .20 .002
.03 1.00 1.00 3.000 .05 .50
.09 1.00
*1.00
.05 C1RT1 & to-5 7.5e-4 2.2Se-3 3. 01'e-l 8.000 1.00 .50
.05 .50 j I.00 .20 .002
.03 .50 .90 I.00 1.000 3.00 1.00 C1RT2 g 1.Re-4
.09 1.00 8.000 1.00
.90 1.00 2.Te-2 S.te-2 1.0Se-
.90 1.00 1.00 .20 .002
.03 1.00 .50 .90 . '.,'0 1.000 1.00 CIRT2 e'
! .09 1.00 3.000 1.00
.90 .50 9e-5 1.3Se-2 4.05e-2 5.41e-
.90 .50
- 1.00 .20 .002 .50
.03 3.00 3.300 .90 .et CIRf2 g*
.09 3.00 1.000 1.00 .90 .08 1.se-4 2.7e-4 5.te-4 1.0ee- * .
1.00 .90 *
.01 3.00 .20 .002
.03 .50 .05 1.00 1.00 1.000 1.00 , CIRT2 d to-6 F.So-3
.09 3.00 8.000 .05 1.00 4.5e-3 6. ele- -
1.00 .05 1.00 e Reisese _ - s, - -e esente w t.,e r. ~ e.ie c..d e., Fe.es onlea since the notale gases would ultimately be the some eufoe"
.e i elI end states involwirig loss ee" containment intege1ty, e
'h
* *8 ,!; , ,
>n a3r:aa$ -
mz=:9w m** * - e * * ' o ea* mm N a 3 2 2 4 3 3 3 3 5 4 4 4 O s - - - - - - - - - - - -
.EI s e e 3
e 5 e 5 e, e e e e e e e t PT e 1 5 3 7 9 7 5 3 0 RAC s 4 7 4 2 6 3 1 3 6 3 2 TCA s 0. " OSR e 3 TEF s 9 4 9 5 2 e 4 S 4 2 1 s a 3 2 2 4 3 3 3 3 5 4 4 5 z - - - - e e ez e e e e e e e e e e 5 5 E Ts 5 s 2 t. 0 l.e 5 2 t. 5 0 t 5 S' 2 9 S R u s 2 3 4 4 2 s 4 8 4 2 E s . LE e ET e 4 2 3 4 4 4 4 4 4 S 4 5 RN a s - - - - - - - - - - - -
. E t a e e e e e e e e e e e e EN Rm 5 5 5 5 5 5 7 5 , 7 5 8 5 3 TO e 3 7 3 7 3 RP s 7 7 2 2 1 7-SM s 1 6 1 7 1 OO s RC s .
R U s u - o 5 5 7 4 6 S 4 S 7 7 7 S r o e e e e e e. e e e e e 1a S 6 6 2 l 6 2 l 4 m s. s. s. a 3 1 3 3 1 3 z s
=
= )
D s *
- s N sk g g g d .
6 g 'g "g d 6 ( EE a T z 2 3 3 3 3 S 4 4 4 4 4 4 TR z T T T T T T T T T T T T ET z R R R R R R R R R R R R CS z 1 I I 1 1 1 1 1 I I 1 1 s C C C C C C C C C C C C s N '4 u O s 111 111 I A= 8 000 000 T = A z C z O = 000 000 000 000 000 L d= a z 555 555 555 555 555 l CI z _ R z E = 000 000 000 000 I R B W= O= 000 000 000 000 s 111 111 111 111 E = L CI s AE E0Sus 000 555 999000 999 000 999 000 000 555 000 555 999 000 999 000 999 000 000555 000 555 B A T R1B= BS s s m H m P = 4s 000 000 000 000 000 111 E T= 110 110 110 110 110 000 H s 001 .001 001 001 . 001 000 1 = . . T* = z E = R 3s U T= 000 000 000 000 000 L = 150 150. 150 150 150 I z . . . . R z 1 .1 1 1 1 F z z . T 2= . N Tz 000 C z 500 z S n 11 E a R m P 1= -
- Tr R a E s V s D s a .
- z C
I z z 5 RE.a EG1 LD s s
- u e
e S= = 200 000 200000 000 200 000 200 000 200000 200000 200000 200000 200 000200 000200 000 200 Lu P= 000 ORO s OCN = 11 C. 0 0 000 000 000 000 000 000 000 000 11 11 11 11 11 11 11 11 000 000 11 11 11 PSI z z ER POS=r 000 200000 200000 200000 200000 200000 200000 200000 200000 200000 200000 200000 200 ATC= SCCR= 11 11 11 11 11 11 11 11 11 11 11 11 CSR z REF =
=
ER z 039 039 039 039 039 039 039 039 039 039 039 039 CO RT5== 000 000 000 000 000 000 000 000 000 000 000 000 0C = 1 1 1 1 1 1 1 1 1 1 1 1 0is = 1* : . V _ ,' !l I;
TABLE 1 om
'e z >n e an:o n.
00 RCS POOL SURROGMTE RELERSE TOTRt. 3URCE ESCAPE SCRUGO- t.ERK- OVER-PRES CNT FR1L1JPE TIME PH BRERIC DREft< LOCRTION CET END cot 1PONENTE EScr.PE CS V. TOR FRCTOR ING RGE SI2E STRTE FRrtCT10F "S . 5 RCS PS L T1 T2 T3 T4 DS OW 684 444' 1 Ru To
=c= ss=a===== er==== ===sssssammass=mszzzz=s=s ss==zzz=3ssm am m= == =ssz== =s z= ====r===s=sz=rarzs ssz=rarsrssssezzassensszzassezzasss:
1.00 .20 1.000 .01 (C1RT1 g") 2e-3 Se-4 9e-4 3.2e=3
.C3 1.00 1.000 .01
.09 1.00 1.000 .01 1.00 C3RT1 d 4.5e-2 1.5e-3 4.5e-3 5.le-;
' 1.00 .90 .05 -
.C3 1.00 .05 1.00 09 1.00 .05 1.00 .
1,00 .90 .05 .50 C3RT1 & 2.25e-2 7.5e-9 2.25e-3 2.55e-2 , 8
.C3 1.00 .05 .50 09 1.00 .05 .50 l 1.00 .90 1.00 .90 1.00 (C5tT13 S. l e=1' 2.Pe-2 S. t er.2 9.' lee-1
.C3 1.00 1.00 .90 1.00 09 1.00 1.00 .90 1.00 1.00 .90 1.00 .90 .50 C3RT1 g' 4.05e-1 1.3Se-2 4.05e-2 4.59e=1 ,
.C3 1.00 1.00 90 .50
.90 .50 ;
09 1.00 1.00 . 1.00 .90 1.00 . .90 01 C3RT1 g* 8.le-3 . 2.7e-4 9.te-4 9.18e-! . 63 1.00 1.00 .90 .01 09 1.00 1.00 .90 .01 05 1.00 C3RT1 d 4.5e-2 1.5e-3 4.So-3 B.le-; 1.00 .90 1.00 05 1.00
.C3 1.00 1.00
.C9 1.00 1.00 05 1.00 1.00 . .05 .50 CSRT1 & 2.25e-2 7.5e-4 2.25e-3 2.55e-:
1.00 .90 .50 1.00 05
.C3 1.00
.05 .50 l .09 1.00 1.00 .
.50 90 1.00 CSRT2 g 4.05e-1 2.7e-2 0.te-2 5.13e-1 1.c3 .90
.03 1.00 1.00 .90 1.00 1.00 .90 1.00 ..,,
09 1.00 . 1.00 .90 .50 .90 .50 C3RT2 g' 2.03 -1 ).35e-2 4.05e-2 2.57e-I j 1.00 .90 .50 4 .03 1.00 .90 .50 8
.09 1.00 1.00 i
.50 90 .01 C3RT2 g" 4.05e-3 2.Fe-4 0.te-4 S.13e=2 '
1.00 .90 . 90 .01 03 1.00 1.00 1.00 .90 .01 09 1.00 05 1.00 C3RT2 d 2.25e-2 1.5e-3 4.5e-3 2.95e-1 - 1.00 .90 .50 1.00 05 1.00
.03 1.00 .05 1.00
.09 1.00 1.00 .
f 1 I
l 1 . av m > ri :4 rt 00 lus r t-a n u ,'. E TABLE 1 og m, RCS POOL. , re SOURCEFRCTOR FRCTOR ESCAPEING Sceues. LERK= OUER-PRES CNT FR1LtRE T1NE PH GREmc BREMC LOCRTicN CET END SURROGRTE RELERSE TOTM. ROE S1?E CORPONENTS ESCRPE S RCS PS L T1 T2 STRTE T3 T4 es D 184 nM' FRACTIOP' cas== ====zz=mzemuzzazzazzzmummem=rus3===zzamazzanzsz=z==zzsanzass=W 1 Ru To 1.00 .90 s s as= esssss s uzzsara zzassenssssssssn essamssressam masarmeness
.50 .05 .50 1 .03 1.00 1.00 CSRT2 4 1.13e=2 F.Se-4 2.25e-3 1.43e-1 l .09 1.00 .05 .50 l 1.00 .05 .50
- 1.00 .90 .10
.03 .90 1.00 C3RT3 g
.09 1.00 1.00 .90 1.00 8.te-2 2.Fe-2 9.le-2 1.99e-I 1.00 1.00 .90 1.00 1.00 .90 .10
.03 1.00 .90 .50 C3RT3 g' i
.09 1.00 1.00 .90 .50 4.0Co-2 1.35e-2 4.05e-2 9. 45e-1 1.00 . .90 .50 ,
1.00 .90 .
.03 .10 .90 .01 C3RT3 g* n 1.00 3.te-4 2.Fe-4
.09 1.00 1.OC .90 .01 3 le-4 1.39e-!
1.00 .90 .01 t 1.00 .90 l
.10 .05 1.00
.03 1.00 1.00 C3RT3 d 4.5e-3 1.5e-3 4.5e-3 1. 05e-1
.' .09 1.CO .05 1.00 1.00 .05 1.00 s 1.00 .90 .10
. .05 .50 C3RT3k
.03
.09 1.00 1.00 1.00 .05 .50 2.25e-3 . F.5e-4 2.25e-3 5.2Se-:
1.00 .05 .50 . 1.00 .90
.03 .010 .90 1.00 C3RT4 g 1.00 .100 8.te-3 2.Fe-3 e.le-2 9.19e ~
.09 1.00 .90 1.00 1.000 .90 1.00 1.00 .90 .010 .90 .50 C3RT4 g'
.03
.09 1.00 .100 .90 .50 4.05e-3 1.35e-3 4.05e-2 4. 59e
- 1.00 1.000 .93 .50 1.00 .90 .010 .9)
.C3 1.00 .01 C3RT4 g* 8.te-5 2.Fe-5 8.le-4 9.19e-*
.100 93 .01
.09 1.00 1.000 .90 .01 6 1.00 .90 .010 .05 1.00 C3RT4 d
.03 1.00 .100 .05 1.00 4.5e-4 1.5e-4 4.5e-3 5. l e-!
.C9 1.00 1.000 .05 1.00
' 8 1.00 .90 .010
.03 1.00 .05 .50 C5RT4 6 2.25e-4
.100 .05 .50 F.5e-5 2.25e-3 2.55e-:
.09 1.00 1.000 .05 .50 1.00 .90 .001
.C3 1.00 C3RT4 Cs) 9e-4 3e-5 9e-5 1.02e *
.001
.09 1.00 .001 l
- . .a . ,
t Attachment SNRC-1187 Page 21 Request No. 10 Sheet 1 of Figure H.4.2 has been reduced so that is illegible. A full-size legible copy should be provided. Response No. 10 Attached is a full size legible copy of Figure H.4.2. sheet I the Containment Event Tree for Class I Time phases T y and T2 ' 9 L I t i
^
s 8
' ~' '
Attechssnt
,SNRC-1187 .- .
Page 22 *
.. se ==aw: mz .
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. I Attachment SNRC-1187 Page 23 I
Request No. 11 Appendix L provides a detailed discussion of the disposition of
- the corium (904 is expected to go down the vent pipes) based on the L.3-2.revised reactor pedestal geometry illustrated in Figure However, this figure is inconsistent with other descriptions of the geometry (e.g., Figure 2.3-2) and provides .
inadequate.information corium disposition. for an independent assessment of the Please provide detailed (as built) of the vent pipes reactor pedestal region. and their covers within and external to thedrawings Include a description of whether the air ducts and during operation.manways in the reactor support wall will be blocked 4 Response No. 11
; Figure 2.3-2 was included within the body of the PRA to provide a general arrangement of the Shoreham Primary Containment.
scale at which this figure is drawn does not lend itself toThe providing any level of detail that would support an assessment of corium disposition. Figure L.3-2 however, is drawn to a scale that provides a level of detail that could not be included on a small scale drawing such as Figure 2.3-1. as-designed dimensions and elevations for a section thru theFigure L.3-2 provides reactor support wall at one of the four air vents. Variation from nominal very these dimensions construction andtolerances. elevations would only be affected by I For clarity and as an aid to assessing corium disposition, additional figures.have been developed. Figures PRA-1 and PRA-2 provide details through the support wall manway openings. Corium disposition through this opening is blocked by the continuous steel ring around the outside of the support wall. i Additionally, drawings ll600.02-FP-4B, 18C, ISD and 18L are provided. l These drawings show the final arrangement of the l downcomers support wall. and their covers both inside and outside of the outside the pedestal region. Drawing FP-48 depicts the downcomer geometry
- Drawing FP-18L (detail 28-28) i l shows the configuration inside the pedestal with the raised floor almost flush with the downcomer lip.
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.be made continuous at the air vents and removable concrete wa ara planned to be placed in the manway region.
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- SNRC-1187 Page 26 Request No. 12 Provide the estimate of the fraction of the molten corium which -
is expected to spread out of the pedestal area through the open aanways and air ducts in the reactor support wall. Response No. 12 The CRD room of the Shoreham containment is provided with confining barriers around the pedestal wall which would effectively hinder transport of the molten debris to the drywell region outside the pedestal. Within the CRD room, the downcomer pipes would provide a funneling effect directing most of the fluid to flow into the suppression pool. The PRA depicts a proposed modification to partially block the air vents. While some of the debris may be dispersed onto the drywell floor outside the pedestal area, transport calculations of the possible competing flow paths (i.e. downcomer vent pipes, versus the airducts; and manway sills, once the fluid level exceeds the proposed concrete block height) indicates that almost all (984) of the fluid would flow to the pool. This function reduces to. approximately 80% without the proposed modification. It was conservatively estimated in the Shoreham analysis however, that only 90% of the core melt would flow to the pool to account for some dispersion, particularly for sequences involving a pressurized discharge from the reactor vessel. o O O e o
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