ML20140E685
| ML20140E685 | |
| Person / Time | |
|---|---|
| Site: | Limerick, 05000000 |
| Issue date: | 02/12/1985 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Fraley R Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML20140B832 | List:
|
| References | |
| FOIA-85-772, REF-GTECI-A-17, REF-GTECI-SY, TASK-A-17, TASK-OR NUDOCS 8502190009 | |
| Download: ML20140E685 (2) | |
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DISTRIBUTION: .
FEg 1,2 5 Central F1les RMirfogue GEdison NRR Division Director! -
TOSB R/F JHeltames RHernan .
W0ircks GCunningham. EDO R/F I* - JRoe HDenton ACRS R/F.
TRehm DEisenhut PBrandenburg (EDO 000165)
"VStello JFunches K8ownan (EDO 000165) a.. .
gm.' - . .. i C. '. MEMORANDUM FOR: ' Raymond F. Fraley, Exec'utive Director ,
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Advisory Committee on Reactor Safeguards ,
William J. Dircks I FROM:
' . Executive Director for Operations .
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SUBJECT:
295THACRSMEETING(N0VEM8ER1-3,1984)'
L' - FOLLOW-Up ITEMS ,
REFERENCEi . Memo from R. Fraley to W. Dircks, subject as -
"'Nbove, dated December 4,1984
~
The following information is provided in response to those specific
- items in the referenced memorandum that pertain to NRR.
. 1. ,A.,CP.5 Report on Limerick Generating Station. Unit i .
The November 6,1984 ACRS report stated that the NRC and industry should continue to work to develop methods which can be used to quantify seismic risk and to ' identify any seismic outliers which
. might exist.- The NRC staff currently has established an Ad Hoc
- comittee, composed of members from NRR and RES, to carry out a seismic design margin program. One of the purposes of this program
, is to develop more precise means to detemine seismic margins. This
' ' effort is. expected to be completed early in FY 86. The staff is tentatively scheduled to give the appropriate ACR$ subcomittees a status report on this program on March 26, 1985. Additional reports -
,' will be given thereafter, as appropriate, and the ACRS will have an l . opportunity to coment prior to finalizing the resolution to this
- issue.
! The ACRS, in its report, also recomended that 1.imerick receive special attention in the NRC staff's resolution of USI A-17 (Systems Interactions in Nuclear Power Plants). The staff has reviewed the Philadelphia Electric Company's effort to identify systems interaction .
problem areas and has concluded that additional interaction studies ,
at Limerick would not be likely to yield improvements that would- I result in a significant improvement in risk. The staff has considered ., l all systems interaction
- studies performed to date, including the one
- Jor Limerick, as part awf developing the resolution of U5! A-17. Should 3
the generic' resolution indicate the need for plant-specific actions.
the staff will provide specific criterfa and guidance as part of the resolution. .
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Raymond F. Fraley . ,
.. 5. Mitigation Studies Applicable to the Limerick Generating Station g' % .
The Research De'velopment Association report on containment venting -
cited in your letter has been given to the appropriate ACRS members by th.e staff. . .' .
AIDtG Milaa J. Diggy, . .
William J. Dircks Executive Director for Operations L
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- 8 o NUCLEAR REGULATORY COMMISSION a wAsmnoron.o.c.20sss S 8 s.,
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- y MAR 2 21985 MEMORANDUM FOR: Hugh' L. Thompson, Of rector Division of Licensing James P. Knight, Acting Director, Division of Engineering FROM: Themis P. Spets, Ofrector Division of Safety Technology
SUBJECT:
UNISOLATED LOCA OUTSIDE ORWELL IN SHOREHAM The enclosed draf t report on unisolated LOCAs outside of the drywell in the Shoreham reactor building is a scoping study to identify high-energy line breaks that are import' ant with respect to isolation requirements. It identified breaks in the RWCU, HPCI, and MSL drain lines as important.
This study used an upper bound assumption that the isolation valves in these lines do not work. The preliminary results of the analysis indicated that the estimate of core-damage frequency for unisolated LOCA outside the drywell at Shoreham assuming that the ist.lation valves failed to close upon demand is about 2x10.s/ reactor year. If the isolation valves were assumed to close upon demand, the estimate of the core-damage frequency would be about 4x10 7/ reactor year. These frequencies of core damage are predicated upon the assumption that the condensate system can be used to mitigate the ,
consequences of an unisolable large LOCA, with an 80% success rate.
In order for RRAB to complete its review of this issue, it is necessary that OL obtain adequate information from the applicant to support the operability of the valves in the HPCI, RWCU and MSL drain lines under pipe break conditions.
This information will need to be reviewed by DE to verify adequacy of the valves' operability. The operability of the isolation valves is important for putting these line breaks in the proper safety perspective.
For further information, contact E. Chow, RRAB, x24727.
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M Themis P. Spets, Director -
Division of Safety Technology
Enclosure:
As stated cc: T. Novak hH. Caruso V. Noonan R.. Wright '
R. Bernero
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l An Evaluation of Unisolated LOCA Outside Drywell in the Siioreham Nuclear Power Station
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D. IIberg N. Hanan Draft (for comments only) 1 y .
l Risk Evaluation Group
- 1 Department of Nuclear Energy I Brookhaven f;ational Laboratory I '
Upton, tiew York 11973 -
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February 1985 l l '
, Prepared for .,
U.S. fluclear Regulatory Commission j >
Washington, D.C. 20555 1
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!> CO,NTENTS I Page ,
y L I S T 0F F I GU R E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .'. . . i i i LIST OF TA8LES......................................................... iv PREFACE.............................................................'... v
- , a KN0KLEDGMENTS........................
- .....................,......... vi I
! 1. INTR 000CTION....................................................... 1-1 1.1 B a c k g r ou n d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2 0ojec-tvas..........g......................................... 1-1 1.3 Sc op e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '. . . . . . . 1 - 2
- 1.4 General Description of the Problem Evaluated................. 1-3 '
1 .
. 2. EVALUATION OF PI PE BRE AK FREQUENCIES. . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . 2-1 4
1 3. ASSESSMENT OF MITIGATION C APA8 ILITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1 Reactor Building Information................................. 3-1 i
3.1.1 Ins t rumentation f or 0t agnos ti cs . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1.2 Sump Pumps a nd Fl oed i ng Bu i l dup Vo l umes . . . . . . . . . . . . . . . 3-2 ,
3.1.3 Containment Atmosphere................................ 3-3 l
3.1.4 P r o c edu r es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -6 3.2 A Small LOCA Dutside Drywell (< 1-1/2" Break 5
t Size)...........
3-6 l I ,
3.2.1 Acci dent Condi ti ons and Al a rms . . . . . . . . . . . . . . . . . . . . . . . . 3-6
)
1
. 3.2.2 Rea ctor Buil di ng Envi ronment . . . . . . . . . . . . . . . . . . . . . . . . . . 3-7 3.2.3 Op e ra t or R es p ons e . . . . . .*. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-8
- 3.'2.4 Es timati on of Core Damage Frequency. . . . . . . . . . . . . . . . . . . 3-9 i !
l 3.3 A large Loca Outside Drywell > 6" Break Size)............... 3-11
[, 3.4 A Medi um LOCA Duts ide Drywell (T2" < ( < 4") . . . . . . .3-12 ........... '
i-i 3.4.1 Accident Conditions Alarms and Opertor Response....... 3-12 '
!- 3.4.2 Es timation of Core Damage Frequencies . . . . . . . . . . . . . . . . . 3-15 r
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- 4.
SUMMARY
........................................................... 4-1 I l .i 5. REFERENCES........................................................ 5-1
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)J APPENDIX A: PIPES AND VALVES FAILURE RATES............................ A-1 r A.1 P i p e R u p tu r e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A - 1 I i
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i A.2 Va l ve Fa i l u r e R a t es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 2 A.3 Comp a ris on wi th LOCA Frequenci es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-3 4
A?;ENDIX 3: LINES COG"ECTING REACTOR PRESSURE VESSEL TO . i
,. R E A C T OR B u l L*) 1 H G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B - 1, C* K!DrMI j APPEnnfX FOR BdtA FREqu!FICATION OF PIPE SECTIONS.AND DISCONT!HUITIESEnCY EST1 HAT 1011..................................... C-1 1
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a LIST OF FIGURES
- .Ej gure Page 4
l4 1 General Description of SNPS Reactor Building Elevations (wi th Emphas i s on HPCI Steam Li ne Routi ng ) . . . . . . . . . . . . . . . . .
2a Lines from Reactor Pressure, Vessel to Reactor Building.......
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Tip Drive Guide Tubes Connections to Reactor Pressure Vessel. .* .
Eviat Tree Diagram for Sequences Following Small LOCA i, Outside Drywe11............................................
4 Event Tree Diagra:.: for Sequences Following large LOCA Ou ts i de D rywe 1 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 5 Event Tree Diagram for Sequences Following Medium LOCA
- Ou ts i de D rywe 1 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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4 LIST OF TABLES Table Page -
1 Summary of Failure and Unavailabil.ity Data for Pipes -
and Yalves.................................................
2 Estinated Frequencies of Breaks Outside Containment........'..
3 Sure:a ry of Frequenci es of LOCA Outs i de Drywel l . . . . . . . . . . . . . . .
4 P.ea:ter Building Ter.peratures at Several Elevations '
Res ul ting f rom a 40,000 l b. Dis cha rge. . . . . . . . . . . . . . . . . . . . . 3-4 5 Core Damage Frequencies for Unisolated LOCA Outside Drywell. 4-1 e
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P
t PREFACE This work was prepared for the NRC which requested it within a one man
- I l month time frame. This dictated the use of all readily available information and refraining from physical analyses. Some of' the phenomenological assump- l tions a,re approx,imate; hence, more accurate analysis may result in a somewhat
! cifferent contribution to the core damage frequency for the medi'.m LOCA cut- *
- l side the drywell (the major contribution to core damage frequency 'is this l study). Nevertheless, the identification of the relative hierarchy of con- '
j tributors is believed to be reasonable.
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- ACKNOWLEDGEMENT The authors wish to thank Kenneth Perkins, Kelvin Shiu, and Robert Young * ,
[
blood for their helpful comments. Cheryl Conrad is much appreciated for l
- typing this document to meet a tightly imposed dealine.
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l 1. INTRODUCTION
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1.1 Background
The SNPS-PRA(1) considered LOCA outside the Drywell (LOCA in the Fes: tor Building) in 'two ways: .
a) Interfacing System LOCAs: Appendix F of the SNPS-PRA estimates the initi-ator frequency and the core damage frequency for this case. The BNL review (2) of the Shoreham PRA re-evaluated the initiator frequency as well as the core damage frequency, and found an inciase about an order of .
magnitude of the core damage frequency. This result is included in the present study, and for more details see Appendix C of Reference 2.
b) High energy line . breaks inside the Reactor 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 smaller lines before they cause adverse containment conditions. 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 Reference 2. In the SNPS-PRA and the BNL review, all the isolation '
valves were assumed to be capable of operating under a postulated break and the resulting break-flow conditions; random failure to operate was
, used in both studies. ,
It is shown in Reference 2 that interfacing system LOCAs are the major l contributor to LOCAs outside the drywell.
1.2 Ob.iectives This study is a special consideration of case (b) abovs stemming from the assumption of the failure of the corresponding isolation valves in the case of a line break outside drywell. HRC requested BNL to re-evaluate the core damage frequency from high energy line-breaks inside the Reactor Building ,
(same as case (b) above) under the assumption that most of the isolation vsives are not qualified to close under break-flow conditions, i.e., assuming
- e failure of most of the isolation valves. Under this assu..iption, there is t
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s 1-2 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-ing, for potential adverse environment or flood effects.
This asstaption obviously increases the contribution of the high energy Itr.a creaks to core damage frequency and requires considera:to'n of other lines connected to th'e 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?
4 (b) .<
What would be the contribution to core damage frequency from each pipe connecting the RPV and the Reactor Building? '
l.
{c) ;; hat isolation valves would De important for mitigating the outside drywell LOCAs?
(d) What is tne cnaracteristic time available for operator action? .
j 1.3 Scope The scope of the BNL study was defined to cover the following:
'(a) To identify any significant(*) high energy lines leading from the RPV
[
to the Reactor Building with a potential for affecting safety systems, if an unisolated creak were postulated.
(b)
' To estimate the change in SNPS core damage frequency relative to the . '
l SNPS-PRA(1) and BN( review (2) due to the following assumptions on the 3
operation of isolation valves following the occurrence of a lin,e break:
.l
$*)The contribution from downstream moderate energy lines of a system wa '
neglected if it was smaller than the contribution of the lines upstream.
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(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 2 (discussed in Appendix A).
(2) All eneck valves will close .on reverse flow as designed with the l 'styre rates shown in Table 2 (discussed in Appendix A).
(3) All other isolation valves will fail to close when receiving their ,
signal to close. No partial closure is assumed for these valves.
(4) Manual valves are assuned to be available for isolation if access- .
ible by the operator.
(5) Remote operated valves that do not receive automatic c'losure signals upon sensing break conditions are identified. However, no credit is given for theci in this study. '
(c) To provice the list of the more important isolation . valves from the standpoint of reducing the core damage frequency.
(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 Shorenam Reactor Building surrounds the MARK !! containment structure -
(the drywell).' At its lowest elevation (referred to here as Elevation 8), ~the '
butiding is an open cylindrical compartment, i.e., there are no barriers in Elevation 8 compartments. This open area presents the possibility that excessive water released into the compartment may adversely affect the ECCS l
equipment in Elevation 8. The $NPS 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 fo'r more details). Figure 1 provides a general description of the SNPS Reactor Building Elevations. .
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F.ig. 1 General Description of SNPS Reactor Cuilding
+ .= = fl,'.". 7.s. Elevations (with Emphasis on HPCI Steam Line .
Routing)
From: Shorcham Nuclear Power Station - Unit 1 '
, Final Safety Analysis Report i
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,: I Figures 2 Wand &hhow lines that connect the RPV to the Reactor Building.
j and provide a potential path from the RPV to the volume of the Reactor Build-i t
, ing in the event of a break with a failure of the pertinent isolation valves j .
.to close. These figures do not show all isolation valves, but only those that are designated as contairment isolation valves. In some. cases, the mos: .
impar: ant cei.ng the RWCU, other valves are available to the operator for ,
s i rer.ote line isolation from the control room; these valves are not shown in i Figures 2 and 3. ,
A list of the lines emerging from the HPV and some additional information -
f associated with these lines (size, type of isolation valves, and process or standby line) is given in Table 51 of Appendix B (reproduced 'from the SNPS-I FSAR(3)). .
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4 d 9 AEACTOR PSE9euRE wtSSEL 888 ese , i - l es 888 n 22 a- nectCw- ro arcseculateem pussp $_ m I si . . 1 I n X 22 h -nect ce -Toercsecutar som puesP l1-ame anorce coo. Ens 1
~ -
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~
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~
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aL g _
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~ -
ano asOT04 COCLERS -
A8sts 280 TOR CooLEss AEAC10R m34 assTRCGtsstaast Pum6C F9st TtP 1
- M AECdeC4A.ATIDII PUIIPs -
X374 Tap DetwE SIMOE TuSEs T 0 m316 TIPsetwE Ges9E TesNS b b i X3FC Tap enewt Sweet Th& -
--- b l s3re tse answE semeE Tuots : : I; b l n SNTWELL o y
e-swPatse -a y *f
... . . .m m , . .< P._ . .
= ,,,,,,L,= - .. . . . .m . .L , . .. .I l sometsseos pect .
Fig. AjTip Brive Guide Tubes Connections to lieactor Pressure Vessel
. se
. O G 4 e e
e
2-1
- 2. EVALUATION OF PIPE BREAK FREQUENCIES ' .
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 Reference 2 and the results are includeo in j Taeles 2 and 3. ,
The pipes considered in this BNL study are listed in Appendix B. All I
lines whien are associated with. General Design Criterion (GDC) 55 are analyzed j in this BNL' study.(*) In add'ition, the Transversing Incore Probe (TIP) Drive Guide Tubes (GDC-57) are considered. All other lines referred to in Table B-1 '
. as GDC-56 or 57 are not connected to the RPV; they are mainly cohnected to the I Suppression Pool (the routing was rechecked).
The SNPS-FSAR(3) w'as the main source for determining' the number of pipe sections and valves or other discontinuities on each line. Tne isometric drawings of pipe routing in the Reactor Building from Appendix 3C of the SNAPS-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
- j
. unavailability data sunmarized in Table 1. The bases for the values ;shown in '
this table are further discussed in Appendix A. The fail'ure and unavailabil-ity data were used with the nunber of sections and valves or discontinuities identified for each line, to compute the frequency of line breaks. The sunmary of this task is presented in Table 2. An example of this computation is shown in' Appendix C.
The results of Table 2 were next grouped into seven different cases:
(a) Large Interfacing LOCAs (Liquid, discharge through brea'k) ~
~-
(b) Large LOCAs outside Drywell: (1) steam and (2) liquid discharge (c) Medium. LOCAs outside Drywell: (1) steam and (2) liquid discharge '
(d) Small LOCAs outside Drywell: (1) steam and (2) liquid discharge.
('j;.. uierencas 1 and 2 consideration was given to large creak LOCA cutside tne drywell . .i.e., lines which are 6 in. in diameter or.mo're.
O e
e
- Table 1 Sumary of. Failure and Unavailability Data for Pipes and Valves
. Failure Rate (Mean)
Break Non-Break
. Component Failure Mode , Exclusion Exclusion N;ss > 3" Rupture (per sectidn) 8.6x10-ll/hr 8.6x10-10/hr Pipes < 3" Rupture 8.6x10-10/hr (persection) -
8.6x10 9/hr Check Valves Severe Internal --
3.3x10-3/yr:
Leakage Rupture 1.5x10-10/hr 1.5x10-9/hr Motor Operated Failure to --
Valves (MOV) Operate 8x10 3/d (w/ comand faults)
Failure to --
6x10-3/d Operate (w/o comand) -
faults) .
'Two MOVs (CMF) --
2x10-3'd /
Rupture 1.5x10-10/hr' 1.5x10-8/hr 9
e l _.. _ _.. _ _ -_ , _ _ -:_ _. _ - _ _ ._, _ _ , .__ . _ _ _ . _ , _ . , .
s t
't -
'L -
J b
i i
TABLE 2 Estimated Frnluencies of Breaks Outside, Containment
- NtHRER Or: INITIAL RRfAK L 5 V 150LATl0N VAIVIS MkEAK ESTIMAftp OESCRIPTION OF THE CASE ANALVl[D ;
BREAK LOCA110N SilE I E A A551sito FLN: FREsplENCY j N C L VALVE 5 IAILUNE . STEAM OF BREAK CASE E i V OESIGNATOR$ l'R00ABillif ON Lil)Ul0 OCCUNRENCE 5 I E o 5 11 5 (*)
Main Steae '
1521-A0V001 Break exclusion section and valve between Reactor -
Line I 24" 4 1 1 Inboard M51V 6.0E-3 steam 5.0E-8 Building penetration and the outboard !!$lV.
24",
(Elevation 78).
- 11 4 I U Inhoard and 2.0E-3 steae 6.0E-9
- Break exclusion section from outboard MSIV up to Outboard the Jet-lepingement Barrier. (Elevation 73).
M51V 1821-
,A0V082
, Main Feed- Break esclusion section and testable check valve -
utter Line I 1s* 2 I I Check Valve between reacter hullding penetration and the F002 A/B 3.
X-3 steam 1.4E-8 testable checkwalve. (Elevation 73).
Testable C.V. Break exclusion sections and 1821-MOV035A/B f ruts II 18* 2 3 1 IB21-A0v036 [3.3E-3f steae 7.8E-Il testable check valve op to the Jet-Impingesmnt '
, A/B and C.V. Barrier (Elevatlee 78).
1 F002 A/B High Pressure Break exclusion settlon and valve between Reactor
, Coolant ,lajec- I 10" l' I I IE41-MOV041 1.0 steam 2.lE-6 Building penetration and the outboard Isolation Lion (HPCI) valve IE41 fleve42. (Elevattain 66).
5tsan Line
- j 11 10* I 6 6 IE41-M0V041 1.0 steam! 1.4E-6 Hon break esclusten sectlens and valves from
, and IE41- outboard tseletten valve up to HPCI turbine. f .n r
! MOVO42 openings (24 hrs each) per year of valve 110V-01/
- i are assumed. (Elevatten es down to elevation li). ,
III 1* 1 11 17 IE41-M0v048 1.0 steam 1.00-3 Ison Break enclesten sectlens and valves trum and Reactor Building penetrations' up to the 1-1/2"
, IE41-MOV047 HPCl/RCIC drain llee to condenser. Normally ep n *
! j .
path. (Elevatten 66 down to elevation ll).
I --
I i '
, a This incluees all discontinuities, i.e.: valves, pumps, reducers and. heat enchangers (see Appendia A). .
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J TA8tE 2 Estimated Frequencies of Breaks Outside Containment Cont'd.
hlNSER OF: INITIAL DREAK L 5 V ISOLATION VALVES BREAK ESTINATED DESCRIPTION OF THE CASE AllALYZED BREAK LOCAil0f4 SIZE I E A ASSUMt) -- FLON: FRE0utieCr
- N C L VALVES FAILilRE STEAN OF BREAK
( A'.I E T U DESIGilATORS PROP. ABILITY OR LIQUID OCCultREIICE i 5 I E 0 5 .
N (a) 5 Interfacing LOCA:
j - Rel4 Shutdiun' I 20" 1 - 2 -- -- Liquid Cooling . All four laterfacing LOCA cases estlested on the
. - RitR llead Spray 11 4* 1 - 2 -- -- Liquid basis of 0.02 for testable check valve unavall-Lise ' 2.0E-6 ability times 10-8 for spurious MDV apening and
- - ReiR/LPCI Injec.
III 24* 2 - 2 - - Liquid 0.1 for probability of low pressure piping to fall Line to Recirc.
- before isolation. See detall in rescrence 2' 4
Lires IV 10" 2 - 2 -- -- Liquid (Elevation - 8 up to elevation - 87).
- LPCS Injection i LO-F008 a'ad
' 5tasdby Liteld I l-1/2 1 1 1 Inboard C.V. 3.3E-3 Liquid 1.5E-8 Break exclusion section of the SLC (Elevation 112)
Control (5tc) II I-l/2 1 1 1 Fool [7.X-3]r Liquid 1.0E-9 Non break esclusion section of SLC (Elevation .
The above 112).
and thstboard j C.V. F006 I Contrst Rod -
Drtre (rRn) I 1 1.0 Liquid 1.0E-4 Scram Discharge Volume ($nV) header rupture. (Ison 1-1/2 hreak exclusion). The pipe break fre peency is
, taken free fluREG-0003. (Elevation 7tl. 61 and 40). .
i
]
I .
1 s., .
4 .
i i
i
/. . .
TABLE 2 Estimated Frequencies of Breaks outside Containment Cont'd.
N't'BER OF: INITIAL I BREAK L 5 Y ISOLATION VALVES BREAK ESTIMATED DESCRIPTION of THE CASE ANALYZED BREAr, InCATiori SIZE I E A A55(HED FLOW: FREQUCHCY N C L VALVES FAILilRE STEAM OF BREAK CASE E T V DESIGIIATORS PROSABILITY OR LIQUID OCCURRENCE -
S I E 0 5 .
. N (*) ,
5 Recirc. Purp
. Seal injection I 3/4 2 2 2 1.0 Liquid 2.0E-7
- Other 3/4" lines I 3/4 1 20 20 Valves of the 1.0 steam 1.8E-3 Bracches fron- various sys- and systen shown in tem shown in _ Liquid (Allelevation) this t(ble this table ,
Sag le Coolant from RPY I 3/4 1 2 2 1.0 Liquid 1.8E-4 Reactor Pnst
- Accident Sarpl- I 3/4 1 2 2 1.0 Liquid 1.8C-4 Ing systro (PASS) -
Tir Drive Colde -
Tubes I 3/8 4 2 2 Rail valve 1.0 Liquid 1.0E-5 ,(Elevation 60). .
and shear valve ,
9 s . .
2-3
- The combined frequency in each group is shown in Table 3. Note that the-LOCA frequencies of the large and medium breaks groups are dominated by tne line breaks of a single . system. For the liquid breaks, it is the RWCU, anc for the steam breaks, it is HPCI and MSL drain systems. In the latter case, tne 10-in. HPCI'line break has a frequ'ency of 3.5x10-6, while all other line -
breaks wnich 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 927. while the RCIC break frequency is only about 8%.
Therefore, in the rest of this study, when discussing large or medium breaks, i
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. .
The small steam line breaks are mainly due to HPCI and RCIC' bypass line creaks (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' may be larger in diameter. The small liquid line breaks are represented in this BNL study by the RWCU 3/4-in. branches, and by the CR0 SDV header piping rupture -
, (reproduced from NUREG-0803(4)) which are about 1-1/2 in. . equivalent dia-meter. ,
Table 3 also includes, for each of the LOCA-outside-Drywell groups, the 1 liquid or stea'm break discharge flow rate at two different times:
) (1) Initially, when the break occurs and flow rates are at their peak values, 2
and .
- (2) At about 30 minutes later after coolant injection is established, depres-surization of the RPV is c,ompleted and operator takes control of the injection according to procedures, keeping.the core covered.
, These flow rates values should be taken as crude estimates. ' They were obtained from NED0-24708(5) for the purpose of providing some indication of the time available for operator diagnosis and response. The NE00-24708 report provides this information for the entire spectrum of break size under con-sideration in this study, i
i e
e I
6
-<.o 4 .. >se 4 ., ,- e - e v. , ,-n. . -+,,,n ,,-,---n-,..e- - - - , ~ , - - - , - -
~ . -.
3-1
~
- 3. ASSESSMENT OF MITIGATION CAPABILITY In this section, the effects of LOCA outside the drywell are; discussed according to the three different groups: small, medium, and large plipe breaks (see Table 3). Based on these effects, some insight on the time available f:r mitigation is presented'. The first subs'ection provides gener51 information cr.
' alarms 'available for diagnostics, containment sumps capacity and flooding data, and some crude information on the containment atmosphere temperature e
increase due to steam or saturated liquid discharges. The next subsections describe the mitigation conditions for small, large, and medium LOCAs outside the drywell.
3.1 Reactor Building Information i
3.~ 1.1 Instrenentation for Diagnostics The following instrumentation and al arms 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 vicin.ity of the break
, 5 Reactor Building floor drain sump level alarm . .
~
Reactor Building flooding alarm at elevation 8,(see additional descrip.
tion below)
{
Area radiation monitor alarms Reactor Building Standby Ventilat' ion Exhaust high-radiation alarms .
Area high-temperature alarms on elevation 8 and on the floor where the f break occurs Specific systems have their own break detection instrumentation such as .
, the RWCU, MSL drain, HPCI, and RCIC.
Reactor Building low differential pressure alarms. ,
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 a.fter about half an i nour. ,
! t e
l t
3-2
- The Reactor Building (RS) water level at elevation 8 is detected.by two-RB level monitors installed on the RB floor. The flood alarms are activated by the monitors when the water level is more than 0.5 inch ahave the floor.
The sump ' alarms will .be activated when the water level reaches the sump alam setpoints instailed at a level just tielow the level that . activates the R3
- flood alarms. - Sump alarm sensors are installed at various locations in the
!' RB.
The area high temperature alarms include the following:
RCIC and HPCI turbine steam line space high temperature (7 sensors
. . each). Isolation signal setpoint at 155'F (elevation 8)
RHR space high temperature alarm (6 sensors) with setpoint at 175'F (e'le-
~
- , vation 8)
RWCU space high temperature (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). -
~
Main steam tunnel containment penetration area high temperature.,(4 ken- ,
sors) located in the area of MSL ' drain lines. ' Isolation signal at 140*F.
N -
. 3.1.'2 Sunp Punos and Flooding Build'up Volumes .
,'! The open area of the elevation. 8 floor is approximately 5,500 sq. ft.
. This area is the total floor area minus the area occupied by equipment founda-tions, columns, drain tanks, etc. Based on this area, flood buildup on eleva- l tion 8 is 3400 gal /in. -
l The drainage capabilities at SNPS are:
Reactor Building Floor Sumps - 2490-gal capacity Reactor Butiding Equipment Sumps - 1660-ga'l .capac'ity Reactor Building Porous Concrete Sumps - 500-gal capacity. j i !
. . l
w _.
,.__._.- _ . . _ _ . _. . . _ _ _ _ .. . . ~ . . .
i .
. Table 3 i
Summary of Frequencies of LOCA Outside Drywell Break' Flow Conditions (*) Initiator '
Initial After 30 Kinutes Break Location Frequency Initiator Stm/ Liq lb/sec Stm/ Liq Ib/sec (Main Contributor) (Event /yr) ,
I Larg'e Size llreaks Steam 1400 Liquid 1200 HPCI(**) 3.6E-6 l (elevation 8')
! $ >_ 6" Liquid 1200 Liquid 700 RWCU 9.6E-6 j .
, (elevation 112')
- Total 4 >_ 6 1.3E-5 Large Interfacing Liquid 1200 Liquid 700 LPCI/LPCS
~
2.0E-6 LOCAs ;
elevations 87' down to 4 >_ 6" : 8' j Hedium Size Breaks Steam '120 Steam 60' MSL Drain 1.0E-4 2" < $ < 4.3" Liquid 400 Liquid 250 RWCU 1.5E-3 l i (elevations 112'-126')
i l Total 2 < + < 4.3" _
1.6E-3 l Small Size Breaks Steam 10 Steam ,
5 HPCI/RCIC(**) -3.0E-3 3 i .. (elevation 8')
I
+<2" Liquid 25 Liquid 12; RWCU Branches -1.5E-3 (elevations 112'-150')
i Total.4 < 2" -4.5E-3 I'I Approximate crude estimates of steam or liquid discharge through break from NED0-24708. ..
I")Breal: can occur between elevation 66 and 8, but the other break locations discharge thrmitsh a pipe chase to elevation 8.
t' I %
.- ._m-2 ... _ . _ - -, __
3-3
pump capacity is 640 gpm, as follows:
~
Four 50 gpm equipment drain sump pumps (elevation (*) 9 ft) 1 Stx 50 spa floor drain sump pumps . (elevation (*) 9 ft) ,
i --
Two 20 gpm porous concrete sump pumps (elevation (*) 9 ft)
One 100 gpm leakage return pump (elevation (*) 12 ft).
, The leakage return pump is designed to process radioactive water. If the l
,, floor drain sunp pump indicators register radioactive material, all sump: pumps '
, will isolate. The leakage return pump can then be manually activated by the
- operator. In addition, only the leakage return pump is powered from onsite AC ,
I 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._, _ .
may drop fres 600 gpm to 100 gpm. This corresponds to accumulation of about 42,000 gallons. Furthennore, since this study considers primary water release, it is assumed that only the leakage' return pump would be operating'
. (other sump pumps would be isolated). i -
l RCIC, HPCI, LPCI/RHR, and LPCS are all located at elevation 8. It is assumed that xthey become disabled when water reaches 4 ft (about 160,000'
! . gallons) as stated in SNPS-PRA.(1) 3.1.3 Containment Atmosphere l The SNPS-FSAR(3) includes in Appendix 3C a few calculations of Reacto'r Building temperatures for water and steam line breaks. Table 4 shows the re-suits of one calculation for the discharge of 40,000 lb of saturated water at ,
RPV normal power conditions out of a 4-in. line break at elevation 112 ft of I i
(*)If water reaches this elevation, the pump is assumed to fail. ,
I' s
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.h I
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, ~!
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s
. , _ . . . . , . . _ - . . - _ . , . . _ _ _ . . _ _ . . . . - . ,.___,_ .__~,,___ ,_. .. _ - _.
,._,.,..l._-....--.._. v _ _ . .. . __,. ._-.
3-4 TaDie 4 Reactor Building Tdmperatures at Several Elevations .
Resulting from a 40,000,1b. Discharge Equilibria -
. Initial Maximum Reactor Building Temperature (*) Temperatures Elevation ['F) [*F) Comments 8'-0" 104 < 140 .,
40'-0" "
148' 6 3' -0" 183 78'-7" "
194 112'-9" "
217 Break location at 1
112'. Outside the pump room temp is 177'F
, 150.' - 9" - -
148 4
175'-9" "
< 132
(*) Reactor Building humidity changed from 50% initiall'y to 100%.
N e
e r
9 e
6 e
, . 4 a
, ,-- , . . , , . - - . . , , - - - , . , , , e-,.-, ww-.- --
3-5 ,
/
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 break. This break results in less than 5,000 gal at elevation 9 or less than 1-1/2-in. water accumulation on that floor. It is seen from Table 4 that a
^
i cruk of tnis size is rapidly affecting Reactor Building atmosphere condi- ,
-i: .s. -
. The other calcul ations reported in Appendix 3C of SNPS-FSAR(3) are similar and lead to the assumption that conditions of 212*F in the Reactor Building elevation 8 will occur under the following circumstances: .
, , (1) A RWCU line break discharging more than 500,000 lb. This is approximate-ly the amount discharged from a RWCU 3-in. line break in 15 minutes (5 minutes for a 6-in'. line).
( 2) A MSL drain line discharging more than 100,000 lb of steam at RPV normal power conditions. For a 3-in. MSL drain line break this will occur in approximately 10 minutes.
(3) A 'RCIC/HPCI 1-in. line discharging more than 15,000 *1b of steam at RPV ..
~
normal power conditions directly to elevation 8(*). For a 1" line .
break, this will occur in more than 25 minutes, and therefore 212*F con-ditions at elevation 8 from these line breaks are not e,xpected to occur (**).
f Temperatures higher than 140*F in elevation 8 can result when steam is discharged directly to this elevation from a 1-in. RCIC or HPCI line continu-ously.
N RCIC and HPCI steam lines are enclosed in piping chase which protects higher elevation against a steam line break in these. systems. However, l for most steam line breaks in higher elevations, steam will exit at eleva-tion 8. -
I
(**)The 15,000 lb discharge would cause the saturation conditions only if dis- ' I cnarged during a very short time, which is not the case here.
l 3-6 ,
3.1.4 Procedures ,
f Given a LOCA outside containment, the SNPS procedures dictate rapid i
manual depressurization of the RPV by th'e ADS.' Dis action substantially ,
4 l reduces the flow rate through the break. If low pressure injection is pro-I~ ticed 3 about 200 psi, break flow may beccme only about~ one-half of :nc initial break "ficw.
l Given an RB flooding alarm, the operator is required to:
Monitor RB level to determine the approximate leak rate, a'nd to ascertain
- the approximate location of the break (using additional . sump alarms and high area temperature alarm) mnitor parameters such as line pressures and flow rate of the safety ,
systems, as a leak may affect these system parameters If requireo and plant conditions permit, dispatch an operator to the RB
. floor.to visually locate, the source of leakage.
J Isolate the break using the appropriate system procedure (HPC1, RCIC, .
RHR,others).
~
3.2 A Small LOCA Outside Drywell (< 1-1/2" Break Size) '
- 3.2.1 Accident Condittoris and Alarms .
The description that follows is bas"ed on an analysis by NRC staff of a pipe break equivalent to a 1.2-in. line break. This is discussed in detail in l - NUREG-0803.(4) 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.
I 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 RPV at 1032 psi conditions.
, Several alarms are available to the operator as described in section 3.1.1 above. The most expected early alarms are from the Reactor Building
] radiation monitors and from local area high temperature alarms.
9 e
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a-NUREG-0803 cites a calculation for a typical BWR Reactor Building that.
shows a temperature rise to 110*F in 10 minutes and 140*F in 30 minutes for a i
discharge of 550 gpm at RPV conditions. (This amounts to about 130,000 lo 1
over 30 minutes.) It may activate high temperature alarms if the set points j is 120'F, but if will not isolate HPCI 'or RCIC systems. , .
I The SNPS ' sumps and flooding setpoints are low (see section 3.1.1), i.e.,
at 1/2-in. above floor level which corresponds to 2000 or 4000 gallons of 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 j 10 minutes (assuming 35f. flashing' into steam, travel time through stairwells l'* and floors, and partial accumulation in equipment sumps (up to 2,000 gallons).
3.2.2 Reactor Buildina Environment
- The water released from the break will exceed the local drain sump capac-I ity, and some will flow to lower elevations through stairwells. Assuming that j only the leakage return pump is availabl_e,(*) the accumulation of water at l elevation 8 would be less than 0.13 in./ min i.e., it would take six hours to reach the level that threatens ECCS equipment" availability. Thus, ample time is available for the operator to recognize the need to depressurize i the reactor and reduce brean flow. Note that Appendix 3C in the SNPS-FSAR states thit 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 temperature 'is expected to rise significantly as shown in Table 4 for a discharge of 40,000 lbs of saturated water at elevation 112 ft. This is a 10 minute discharge from the 1.2-in. -
) line break described here. While it may result in high Reactor Building
! temperatures when discharged over a short period of time, .it results in 110*F
- in the Reactor Building if discharged during. about 10 minutes (see section j., 3.2.1). However, the temperature in containment will continue to r'ise'due to
- the continued discharge through the break and may reach the 155'F 'RCIC/HPCI -
i isolation temperature after about one hour. The Reactor Building Standby Ventilation System (RBSYS) of StiPS has a heat removal capability of less than PMacwaste system tanks capacity allows for about one day accumulations of I untreated water at a 100 gpa pumping rate. l 4 -
-l t l
. I
3-8 ,
5% of the heat discharged by a '1.2-in. line break, before reactor is depressurized and the break flow is reduced.
3.2.3 Operator Response
- At Shoreham, the operator will have a flooding alarm and high Reactor 3.ailding radiation alarm at about 10 minutes as discussed in the previous section.
For a small LOCA outside,drywell, with feedwater still operating when the LOCA occurs, scram may not always occur immediately. Following the scram, the operator will try to keep the normal feedwater injection and therefore 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 feedwater flow rate. It appears that the flooding and high r.eactor building radiation alarms will indicate that a small LOCA have occurred, and the increased feedwater injection flow mayb ' e used for verification.
Therefore, it is expected that~ the operator will recognize a small break LOCA~in the reactor building within about 30 minutes after scram. Unless the .
operator . perceives a LOCA, he will depressurize the reactor at a rate of only .
100*F per hour. In such a case it will take 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />' to depressurize the *
. reactor to 100 psi and reduce break flow by about a factor of 10. As seen in section 3.2.2,. four hours are available at SNPS, without flooding to elevation
- 12. However, in this case, the temperature in Reactor Building may reach 155'F or higher (*) between 1 and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and trip HPCI and RCIC, and most probably require depressurization for low pressure injection. These events .
would lead the operator to recognize the small LOCA outside containment with very high probability, if he failed to recognize it during the first half hour.
- It should be noted 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 probability event mainly because of the improved arrangement for flooding detection at elevation 8 (relative to the arrangement assumed in NUREG-0803). High radiation and high temperature conditions in the reactor M A GE analysis estimates that the maximum bulk tem building;would reach about 140*F (See NUREG-0803(4)perature
). in the reactor I
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3-9 ,
building will enhance the probability of recognition. This BNL study assumed..
that it is most probable that manual depressurization of RPV to reduce flow and enthalpy discharge through the break would take place after about 30 ,
minutes to I hour into the accident. ,
Tne cepress rization of 'the RPY may reduce flow rate and enthalpy of the
- water discharged through t'he break to a level accommodated by the sump pumps, and may reverse the conditions in reactor building i.e., conditions may start 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-tai nment (which is larger than the " natural" 100% per day leakage rate from
. reactor building), in order to improve the reactor building atm'osphere condi-tions and to a.llow safe. operator entry. As shown in NUREG-0803, depressuriza-tion reduces significantly the dose received by an operator entering the reac-tor building.
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 (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-
- FSAR considers 30 minutes to be sufficient time to walk through all SNPS ele-vations, locate a break, and isolate it.-
3.2.4 Estimation of Core Damage Frequency 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 strnmarized in the form of an event tree in Figure 3, and quantified. Feedwater and high pressure coolant injection are in general available under the circumstances of small LOCA. ADS, LPCI.and LPCS have very low unavailabilities. The values for their quantification are taken from
.. Reference 2. The events that are differently quantified are: (1)'the'proba-bility th'at at 30-60 minutes the operators take actions and complete rapid '
manual depressurization, (XH ), and (2) the probability of controlling the condensate flow if required (V). The Xg=0.01 is taken basically from NUREG-0803 where 5x10'-2 is used. The cif ference between NUREG-0303 and B"L values is due to the SNPS improved early flooding alarms which increase the probability that the operator recognizes the LOCA outside the drywell and follows the required depressurization procedure.
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The V=0.1 is the common value used by BNL in Reference 2 foc.
controlling condansate injection if sufficient time is available to the operator (in our case 30 to 60 minutes). The V=0.02 includes a factor of ,
0.2 for the possibility that no damage to LPC1/LPCS will occur even under the cir:umstances tnat the operator does n'ot depressurize the reactor early, but .
ratner depressurizes it at 100'F per hour rate, for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or more. In sucn a case NUREG-0803 indicates that entry to the reactor building may be delayed for up to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. The LPCI/LPCS may survive the adverse environment in the reactor building for such a period, because they are qualified to sustain these conditions for at least several hours. A factor of 0.2 (compare Vlli on Fig. 3) for the LPCI/LPCS availability apparently underestimate their availability.
The event tree quantification yields a core damage frequency of about 1.1x10-6 per year for small LOCA o'utside the .drywell, when it is assuned that the motor operated isolation valves are failed.
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.
3.3 A large LOCA Outside Drywell (>,6" Break Size) '
This case was treated in the BNL SNPS-review (2). However, the assump-tion.in the present study is that HPCI and RWCU isolation valves would fail to close.
Only HPCI lines were treated in Reference 2, and a LOCA frequency of 2.7x10-s/ year was obtained. If we postulate that the isolation valves fail upon demand, a LOCA frequency of 3.5x10-'/ year is obtained 'for the 10-in. HPCI line break (see Table 2).
The 6-in. diameter RWCU line has three isolation valves inside the ,
drywell . Only one of them close automatically on sensing line break conditions in the RWCU lines. In Table 3 when no credit is given to these 1
val ees a break frequency of 9.6x10-6/yr is obtained as derived in Appendix C i of tnis report. I, Reference 2, the three valves were considerea (having l
3-12 .
different isolation signals ar.d 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 the 61n. RWCU line break would be about 10-8/ year. Thus, in ,
Ref. 2 it was not further considered because the frequency of interfacing system LOCAs, was calculated to be 2xi0 6/ year (see Tables . 2 and 3 of this report for resalts and Ref. 2 for more details).
The interfacing LOCA frequency is also estimated in Reference 2. The results are reproduced in Tables 2 and 3. This LOCA frequency '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 1.5x10-5/ yea r. When this is "
used with the event tree of Ref. 2 (see Fig. 4), a core damage frequency of 3.0x10-8/yr is found. The 0.2 factor is the probability of operator failure to control the condensate system pumps' flow to the RPV in the short time ,
available (about 10-15 minutes).
In the case of a large LOCA'outside drywell, the discharge to containment -
is about 1200 lb/s and saturation 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% flashing). Thus, it is obvious that no isolation is possible, as it was also assumed in the SNPS-PRA and the BNL review for large LOCA discharging saturated water or steam into the reactor building.
This core damage frequency of 3x10-6/yr is 7 times larger than that given in Reference 2. This is because in Reference 2 the interfacing LOCAs were the dominant contributors . They' are dominant when credit to isolation valve closure is considered.
3.4 A Medium LOCA Outside Drywell (2"< 0 < 4")
3.4.1 Accident Conditions Alarns and Operator Response The most dominant case of the medium LOCA is the 3-in. RWCU line break as shcan in Table 2. The frequency of a RCIC 4-in. line break. is small cccpared to the total nedium LOCA frequency of 1.6x10 3/yr; the RWCU 4-in. line break 6
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frequency is significant but the sections considered are relatively downstream and estimated to be 1/4 of the total RWCU break frequency, whereas the other 3/4 are for 3-in. line break or less. Thus, our discussion in this section ,
refers to a 3-in. RWCU line break.
Tne RWCU is located at elevation 152 ft to 150 ft. At 150 ft the demin-eralizers are located, which process water at low pressure and at about 125'F 1
and, therefore, not considered. Thus, the break location of significance can occur at the 112 ft or 126 ft elevations. On these elevations, the line are enclosed within concrete shields providing physical separation from all safety
- related equipment (see App. 3C of the SNPS-FSAR).
Table 4 present the approximate temperatures in the reactor building fol-lowing a RWCU 4-in. line break at elevation 112 ft in the RWCU pumps room. It is estimated tnat about 10 times the amount discharged in that case, i.e.
E33,000 lb, would result in saturation conditions in the reactor building.
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 (up to 10 minutes). -
It is expected that the blowdown from the break will. cause immedia.te MSIV ,
. 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 started a few minutes before that on low level (L2). Therefore, in this case, I
it is immaterial whether the operator depressurizes the RPV, because early automatic ADS actuation is expected for this case.
The water discharged during the first 10 minutes would fiash (~35%) and ;
the remainder (about 20,000 gallons) will cascade through the stairwells to !
elevation 8. Appendix 3C of t'he SNPS-FSAR considers this effects and states that no safety system would be affected. This' accumulation is equivalent to ,
i 0.5 ft and will result in flooding alarm in the control room.
l '
The radiation and temperature alarms are expected to be on in many areas s
of the reactor building. Therefore, it is believed that the situation of LOCA i :
outside drywell and the reactor building adverse conditions and flooding would l ce recognized w.ith high probability within the' first 10 minutes. Earlier I
recognition of the LOCA and depressurization of the RPV would not change l l
3-15 r-l much the progress of this accident sequence. However, if operators fail to, j recognize the event and follow the procedures (which call for keeping'RPV at
{ low pressure and controlling the injection flow), then the reactor building ,
1 conditions may severely deteriorate.
, one depressurization would apparently happen at about 10 minutes. Then * '
tne LPCI, LPCS' and condensate pumps, may all inject water to the PRY, and dis-charge a larga amount of hot water through the break. While this hot water -
would have less enthalpy than the saturated water discharged during the first j
j 10 minutes, it has flooding potential because of its high flow rate. Flooding ,
1 may occur in an additional 30 minutes if the flow rate to the RPV is not
- reduced by keeping it at the lowest possible pressure without ' uncovering the
)- core. This is the operator action specifically required for the case of '
J medium LOCA outside the drywell. In such a case LPCI/LPCS may maintain core cooling for long period and condensate 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
~
l, outside drywell is shown in the event tree in Figure 5. ,
i e . ;
The initiating event does not distinguish between water or steam line 4
1 ,
breaks. They are considered similar because even though the steam discharge .
l! through the break is smaller, the impact on containment atmosphere temperature
! and pressure is about 5 times higher for a steam line break than for the case l
of a similar size water line break. _
In the long run, after the RPV is depressurized, the flow out of a steam i, break may be significantly smaller if the core is not flooded so that water is discharged through the break. If the water level is kept ,below level 8 (L8), i l then the steam flow out of the break is expected to be relatively, small.- ,
Thus, it may not be sufficient to create flooding sufficient to damage the ECCS equipment.
The liquid line break is therefore the dominating case. Thus, the event j, tree starts with the radium LOCA frequency from Table 3. The feedwater and RCIC/HPCI are ass c.ed to be unavailable. Depressurization by ADS is l l
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considered to occur at about 10 minutes into the sequence. The low pressure-injection systems will start to flood the core. Therefore, operator action to ,
centrol the injection flow rate is needed to reduce the impact on the reactor ,
e building and gain time before the condensate system would be required. If the operator recognizes the need to control the injection, then the condensate . -
1 system pumps will also be controlled at a later time with a higher reliabil-ity. If the operator fails to control the injection, less time will be avail- .
able to control the condensat,e pumps injection because they may be needed at about 10 minutes into the accident.
j , The values used for the probability of successful operator action are j thought to be on the conservative side given the time estimated to be avail-
- able. Therefore, the core damage frequency for mediJn LOCA outside drywell 4
may be smaller than 1.4x10-5 for the case that no credit is given for RWCU isolation valves. On the other hand, the phenomenological assumptions used may not be realistic and may underestimate the break-flow and Reactor Building conditions, so that less time will be available for operator corrective action than a,ssumed above. .-
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- 4.
SUMMARY
The BNL review (2) of SNPS-PRA estimated a core damage frequency of 4.2x10-7 for LOCA outside the' drywell in the 'SNPS; this is mainly due to *
. interfacing system LOCAs. In this study, an additional assumption was' ir.t-Maced at fdC request: namely, thaf.1 solation valves would be treated as '
, failing to close upon demand. The only exceptions to this assumption are the
! MSIVs and eneck valves. The effect of this assumption is shown in Table 5.
j; It is seen that the core damage frequency increased by a factor of almost 50. !
I The leading contribution comes from medium LOCA outside the drywell; in. par-ticular, the RWCU 3-in. line break is seen to be the most important (see Table 3). '
i*
l Tasle 5 Core Damage Frequencies for Unisolated LOCA Outside"Drywell Class V Core Damage Frequency Isolation Valves Isolation Yalves, Assumed
'to Close on Demand Assisned to Fail to Close on Demand Initiator (from BNL Reference 2) (from this analysis) .
Interfacing LOCA , 4.0 E-7 4.0 E-7 - ,
large LOCA Outside Drywell 2.0 E-8 2.5 E-6
- Medium LOCA Outside Drywell -- -
1.4 E-5
- Small Loca Outside Drywell --
. 1.1 E-6 i
Total 4.2 E-7 1.8 E-5 l Table 2 provides the information on the most important isolation valves 3
1 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 '
a I*' drain isolation valves. .
2
-,-n- ~ - ,-- ,,,_,,,-,,,--------,,-p-------,---+v----,
,-+-----,-,---r ,w .- -,-m. ,--n .--,-n---,-e,,
--r-- -------,----c~~g ,-,-w-,vv,-
4-2 .
Tables 3 and 5 show that under the assunptions used in this study, the core damage frequency from LOCA outside drywell is dominated by the RWCU ' '
medium LOCA breaks. Also, the large LOCA contribution comes mainl'y from the - ,
RWCU system. Therefore, it should be noted that beside the inboard and out-
- sted containment isolation valves, the RWCU also has two additional isolation
. va'.ves that co not receive an automatic signal to close when a. line break occurs and are available for timely remote closure. This action can be taken half an hour after initiatior) of the accident when the reactor is depressu-
, rized, and before the loss of low pressure injection.
t I
1 -
8 9
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-_ __________.__m___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
- .. \ .
- \ e 5-1
- 5. References l 1. "Probabilistic Risk Assessment Shoreham Nuclear Power Station Long Island Lighting Company, Final Report," Science Application, Inc., June 24,19835 ,
~ 2. D. .Ilberg, K. Shiu, N. Hanan, and .E. Anavim, "A Review of the Shore 5ar
! f;uclear Power Station Probabilistic Risk Ass essment," NUREG/CR *C50,
- BNL-NURE9-51236, December 1984.
- 3. " Final Safety Analysis Report Shoreham Nuclear Power Station Long Island Lighting Company," SNPS-l'FSAR (Revision 31, Auonet 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 NE00-24708, 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 b.S.
Commercial Nuclear Power Plants," NUREG/CR-1363, EGG-EA-5125. May 1980. -
9.
} Oconee PRA', A Probabilistic Risk Assessment of Oconee Unit 3, NSAC/60,
, June 1984 .
n$
e
)
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d e
e o
A-A -
Appendix A Pipes and Valves Failure Rates Y
A.1 Pipe Ruoture .
The main data sources used for probacility of pipe ruptures were the Reactor Safety Study (6) (RSS) and the EPRI-NP-438 report (7). 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 locations more susceptible to breaks.. In the BNL study, reducers and valves were considered as rupture locations, in addition to pipe sections.
The SNPS-PRA(1) uses the RSS data for pipe breaks. However, it distin-guishes between pipe se~ctions which are " Break Exclusion," i.e., are designed
~
to criteria provided in ' Appendix 3C of SNPS-FSAR(3), which basically allow for 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 cssumption, and the basic values used in the study are similar to the SNPS-PRA and are summarized in Table' A.1 below.
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Table A.1 Pipe Rupture Rates '
Computational lle.in hsessed Rg T
(non break-exclusion Computational Break thm-Ilresk
. Component pipes) Median Exclusion Excluston.
Pipes > 3" dia. 3x10 3x10 8/hr lx10-18/hr, 8.6x10-II/hr 8.6x10 18/hr
- per section Pipes 4 3" dia. 3x10 Il - 3x10 s/hr lx10 8/hr 8.6x10-18/hr 8.6x10 8/hr per section i
I 1
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A-2 .
l The pipe rupture data of the RSS is applied section by'section, where a section is defined (RSS, page III-41) as follows:
l A section is an average length between major discontinuities such as -
valves, pumps, etc. (approximateJy 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 II -- Heat Exchangers, Appendix C presents the details of the pipings and their division into sections.
- A.2 Valve Failure Rates k The main sources used for valve rupture'or excessive leakage failure rates ,
l were the Reactor Safety Study (6) and NUREG/CR-1363 report (8). 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 i
includes also l small leakages such as' from packing failure. Similarly, the
, internal leakage rate of check valves given in the NUREG evaluation includes many small leakages whicn are just violations of the Technical Specifications limits, and too small to be considered in.this . study.
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-FRA(1) 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.
l -
valves. This value of 1/18 is also used in this study.
Based on the above, the BNL study esse.ntially adopted the SNPS-PRA approach, i.e.:
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. . i Table A.2 Valve Rupture or Excessive Leakage Rates Computational Mean Assessed Break Non-Break Cor::ponent Source Failure Mode . Range Exclusion Exclusion
[hr-13 [hr-1] [hr-1] -
Check RSS Internal Leak- 10 10-6 Valves 3.8x10 7 age (Severe)
NUREG/CR- Internal Leak- --- ---
, 1363 age (all sizes) 9 Check RSS Rupture Valves 10 10 7 2.7x10-8 2.7x10 s and Motor NUREG/CR- External 0;;erated 7x10-8 7x10 s 1363 Leakage /
Valves Rupture d
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O I t O
____________-______________A___
. . ~ . . - . .a .. ... . . , . . .. ..- . . . . - . . - - > -
1 l' Table A.3 Motor Operated Valves rallure Rates
)
- Value
~
i liscil in
', Coi.ponent Source Failure Mode Assessed Range Mean Value I,Illt Study -
1 l RSS Failure to 3x10 4 - 3x10 3/d 1.3x10 3/d ---
- operate Hotor (include connand)
Operated -
! Valves NUREG/CR- Failure to ---
8x10 3/d 8x10-3/d .;
- 1363 operate '
i
, (Il0V) (for 8WRs)' (include ,
command) l NijREG/CR ' Failure to ---
! 1363 operate l, (for BWRs) (w/ocommand) 3.
il Command Failure ---
2x10-3/d 2x10-3/d of Inboard.and.
' ~
Failure of
. ~
Both MOVs Outboard POVs (Inboard and i Outboard) .
J, ' ..
i Oconee MOV --- 10-3/y PRA III- N"#I'"5 '
Opening j __
i 1: .
s ,
o
Table A-4 A Comparison of Frequencies of Loss of Coolant Accidents RSS EPRI-NP-438 SNPS-PRA
~
LOCA Pipe lireak Mean- All Sensitive Mean . This Study:
Diameter 90% LOCA Pipes LOCA LOCA Outside
. (Inch) Range Frequencies (Mean) y[5(*) Frequencies Drywell Small LOCA 1x10 * - 1x10 2 2.7x10 3 -102 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 1x10 5 - 1x10-3 2.7x10-" -1x10-3 7x10 " 7x10-4 3.5x10-5I**)
> 6"
(*)It is assumed that 10% of plant piping are LOCA sensitive pipes.(Ref 1) -- --
(**)The large diameter pipes are " break-exclusion" and are assumed to have 1/10 of the RSS rupture rate.
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A-3 .
, (1) Use of RSS failure rates for vaives. .
~
(2) Apply a modifying factor of 1/18 to the RSS valve rupture 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 val ve s , similarly to the factor app 1'iea to the pipe I 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-sfy,,p, For simplification of the analysis, the valve rupture rates were a.lso used with 1 .
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 stamarized in Table A.3. .
A.3 Comparison with LOCA Frequencies .
The analysis in- the main part of this report involves a large ' number of 1
pipe sections and valves. In general, more pipe sections and valves lare
. located outside the drywell. Thus, the frequency of LOCA o'utside containment
$i should be a large fraction of the plant's LOCA frequencies. Table A-4 compares
, the results of the LOCA frequencies from this BNL study with the RSS results
'(table III-6-9 of RSS), the EPRI-NP-438(7) results, and those of the SNPS-PRA.
1
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Appendix B: Lines Connecting Reactor Pressure V;ssel to Reactor Building -
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, Tabic B-1 (continued) .
PROCESS PIPELINES PEllE1 RATING PRIHARY CONTAINHENT i i se w lerAnas sew mees ' teir.t., news vee ws mainnat- an Anw w art naien ress rewa (i sins anent
.senissa as -
et Pt a fitt litt tettAat lift le erga le nag 158841141 I th4 4101Alla liti teostaltaval Iles (sic) SIAgy$
; s l as. I..b4 t ims)
. . . _ _ .--__5 ChC llwl llut fle.)
(6,27) (S.6) (5.6) Stianas (se) (a,3) elemegg s-los ese tot a see ses.se l6 I M Swiside 4 le &lebe AC M to 5.etansl*e (ha der. 8 . 6 .151 79 (leted (2)
*g.ssies reel (le.e.p metere, 2 6 b eside t IIS Sete AC K A.F.Ist see steen tenJenslag 9tssberge. I 4 beside IG Este 38 (Med ha ns.t Flee. I l AC E f .G. Int M- (lesed
] I 4 beside 018 Sale M M . i ( ee sy..y test t Ime, and I I le beside se Eiehe ass M Spee (86) AC AC t e is.ey nennee II., I I tatside* f .E.8BS 67 (lesed 3, te Sete AC E ist M apan (M) - t tee mee ene atee peewe to $6 I I 96 1646Bde le Siehe AC AC F.E.et
- i
%,.nsl.a (keet.er. 79 (lated (2) stle n:.t 81 . 4 I 2 beside se Elabe K st est i natsl naeta.e 11 . 8 I 4 8.eside le Elshe K le (leted (M) see itse. ( Anslag Dessherge. I K- -
Ist 29 (lesed i h a pt.o I 4 beside te Sete AC ' AC t .5.IBI 28 (Ml fle=, 5 I 4 beside le Sete K (lesed j ( Seear lest flae. I beside
. AC aus 20 Spes (16) te.e Sr.e, estalawn Ile , eed 4 88 le State AC AC 8.G 48
- 67 (Wed i
i I 3 6.tside IG Gate M AC Ise } Selle t valve 80samarse free et I I 2 beside belief telee Nigh 04ffer- Spring 86 Spee (M) , ] httly to E8C r.my 5.cties e/A h/A (lesed i . ential re essere a-al men - pred Speer time te Srv && ' 1 1 4 lastde se Sete M E A.f.s. set 1 4 beside DIB Globe 7e flesed DC DC 4.1 W.Iul Il flosed
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see Este pC 1 teaside IG Elete at s.ase 43 (leted i 7 i K SC E. sus B2 Spee 1 7)1 i 4 3-ll ur(I I.ebisie taheest 56 I I BS beside j 8E8 Sete K DC en le? Spee 2 le beside thest g lem neuerte flee ureerse Ilse h/A (lated 1 ..,4 ,,eee . . . . - . . -
. iisi a-In saril rem,5.stles 96 I I 96 beside - se Gene BC SC, s.est , PI (lesed i ,
i j 3 56 mt 8( Insb8.s Steam lolet Elst 96 I 3 le Sete M I loside AC n.Ist E apes l I teslee se slebe AC AC K.es 33 apse Ii l I eesside se sete K st E. Int u flosed I i 1 i beside - IS Elate K at L ast 3r opse t I i t
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, Table B-1 (cont.inued) 1 l
PROCESS PIPfl.lNES PENETRA!!NG PHlHARY CONTAlHMENT i - 3 rziemos tesala veu laranan unw amt a amitt= = =4 sanus nement allallel la ersnaien rana , er emas sinues elat issium et a rett site tenant Ivrt le wi e seen 1
- I:=s Isot Ane (n) s.= lies Isot tie.) (misseems (6.nl (s.61 le cieg esataleau sus (.sq sieses (s.s) swaal gleg go,,3
. 8 i ~& C bl=e taheerst 9 3.,5 t 8 8 beside le Sete hc d
3 8 8*ts348 thest fles et ne erse glem h proerse Item
}g m/A h~
(leged (llg i
- a-la a(IC Wes==e remy Sasaberse M I & I blside le Stay thest Itestet tre. fleu/9C
, 9 2 beside (hett f lew Gre. Ileistest Il gleged I -
Severse fles arvesse flew a/A cleged lit.28) 1 3-19 st3C r.=se sectlee M l 5 6 beside se Sete pc et est >
- il glosed i' s-2ns.s ( e screr r==p Sitcharge le M I ary I 98 l= side Vlt Item 8 2 leside Aeoerse fleu pesesse flew N/A fut Elete AC AC (lesed (l) i 3 le entside le Sete AC AC at get la flesed i
45 gjesed (geg
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l 36 ep.o 8.na.e tuttu se tecirc. rump and 57 8 suser teelers . 8 4 beside M Este AC AC WI 25 gyse . 8 234.8 pa(s fu #1== tectrc. riep and 57 3
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gy,e 3 244 se as Destiu to buy = ell tentt feelers M i S l 3 lastdo thest flem Reserse flew severse fles 8 2 Setside fe Sete AC n/A syse
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Table B-1 (colitinued) PROCESS PIPEl.INES PENETRATING PRIMARY CONTAINHENT
~
l lasW te(Als84 taevi Aus/es rstatteT IBeett Vat W1 merJinst NAlist le talatsea rensa touta
. tesisiwa si ete%las Imeegl.
er et e elet sail rewaas llra le sets lerieu tseistles i n u su eltar. Bluts gle.) alma istCI siAlas
. slims Isetalso t u ) seC tsu cenialme ns (s,n3 is s3 (s s) sismat (len go.9) sonness a-aa Scese . . . . . . . . . . . .
. . . . (85)
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- - o e - - - 1- 1-I i : 2 2 I 1 1 1 1 ' ! e i i s : a
- 2 2 5
5 j i 7 ; 8 3 2 ] ]
.m I
j }'9 'I ,'
. v= =1 si ,
1 I:
.> . . ,: . . .. j4l ! Iar: l! ! d5:
.gi. ....
J a i l, 1: 1 J : i } l ,L' i i i '. J! ;!J
..}. .
I ,:l
-l' 1;..1; il j t oi l2 ! .is j .l
.o !
g
.- 1_ : + -
s
$ 2. f' 2,,2 'f 2 E U O % Cg4 2;2 2 l l-E j I,i d'did1d'd d d Id d d E did,2 ed! d I l7 *4 3 'I s
l
, TableB-1(continued)
! , PROCESS PIPE!.INES PENETRATING PRDtARY CONTAINMENT i ) s ee=m m .evasrn.se,a umso= an.uninur ima. .
- : t. =ei. in. intat
- ac :usur air mu.co nivac..emire, um intenu n3u,mone u.,se,ete voeuoitou
.n nuseu.tu.orguw som m n cucerem.
u .ai.u. l niuse faii.n n mir one son .iii not u.u nive on n. n eaiva m ut a feir ,, em in :m ma i sunu. .
, l:. :n uir.-
eme, et smr u cry.ett s.urus= sa
, , turamavav .m we n.ve m useentt,mme am.easu tus n, i.ine
,s - en.m a f te, u essmuin s.u~ m .
en.re. mt uc.e .ili u-tt untumt un, for sto 4 poeti mu.n unutm mne. j . sussressten .eter taaltag. onen sesseette signals are act present. taese velves aior se seense Per test er esersting convenience, i ' l
- 2. Tottaste asett valves are test;*H fse evante seeging alta teae etfrereattal sress.re agress
- tse velve test. *>e velves etit close en avverse fle. even neven the test s.itanes ner se .
a pesttienes for seen. ne velves etlt seen maen suas essenerte pressere esseees evettae ' j i 3ressere eve 9 mewgn Ine test lettSS ser se 3esttteate for (Itse. I 4 I This Itne is salt nesse esPing astatenesse. $ervjte air sueely is etscenmMtes eurtag $1M4 4 estretten er eeBlatstrettve sentrol. '
. $. AC assee eserease glees reestree fee isolettee fumettens an esmerge free the seerguer AC
] pe.ee suses. CC eserete tseletten velves are 9 emerge free See stallet settersee. 1 8. All seter eserstes toelation valves .411 remete in tne last soststen veen fatture of velve . . . j- so.e a. All ate.eserstee tselatten velves will slese seen air fattues. j t g.f Signal 8 esees signal E everrises to slese. I 4. so e* sae'stes valve saa se ocease er aleses er resste menval settas ter eeeesting uneentense j suring any esse of rosesse eserstien esseet onen autemette stpel is present (see Sete II. i ' s. u,a out mitten of weive c en e, c.s.: 18 tne uc.n e., tag n.ree: e, .ane. 3 ef me e m u r. to. The imtftee on.,e nu m a no.irw fee unut.en tuiattu mir. . i
- 11. $settel sie testacle taest velves vita a seettive elestag feetare are settpos for remote testing
- eartag normel eeeratten to asswee aetsenteel eserestitty of tme velve else. The rosess testing feetwee will essse only a partial sevement of tre etse into tse fles stroes. sita emir a einee
.effect en rtou. men recotet of an toelatten signal. tne octuator seeing force sell ettner uwee *
- a silent reewetten in flow .aon tae fees.eter systes is avesisele er tasse tae velve to elete. ,
3revistag a testtive closere esf*erential pressere en tne seates site .non tae.feeemeter flom , te not avetteele. *
- 12.
- hts velg *.411 eeen unen tota a few reactee pressure vessel mssure and m usteent signal are *
$ resent.
- \
, 12. The aster estreser of tais valve is key leches open eart#g normel operating esasjttene. 14 frewestag in Cere Prose (TIP) Srstees 4 j
' When ene T17 systes caele is insectos. the tall velve of the selectee tune seems aetamattully
; se saat tae scese saa uele mer sevence. A ausses of fome velas mer se seense at ear one time
' to genauet the salteretten, ans any one ptee toes is oses. at sost. a tem neurs see peer.
4 If stesore of tse Ifne is reewsree asetag talleretten. as fasteated by a sentetament testatten signal. tse nele is autasetteelly retreetse one tne sell velve closes autenstically after coe. '
. pletten of taele ottaen el. fe eneert tsetetten saaest19ty, if a itP taele falls te we taerem I er a sell velve falls to elete an eseleetw. sneer volve is installes in seen Itne. Veen recetat of a renote asavel sipal. tnis eselesive velve allt sneer tae ftP uele and seal see gutes ta.se.
j 18 . All mauses penetrottensteestratee 'Seere*) are capped me seet elsee. - ' t 16. Talve .411 slese en systen ntta flee. - i . 17. Iselattan steals A se f .t11 inettate tSe reester tvilstag sunser venttetten syttes =atta to tare tieletes tse earte att tselasten valves. i 18. j r This valve .41) egen eren se u a few stfemgttel sressure ag=sts the valve one an uttaent sipal are sus e t. . l ;9. Pnst.re senters one seastat stene ttae spessure see vue for fateelets tantrol ta see.nt inae.ortest valve ste9 tag at up steam line artss.res ' stew !! stig). I e i 1 1 1
}
e
9 TableB-1(continued) PROCESS PIPELINES PENE RATI?!G PRIMARY CottTAIMMENT Notes (Centinuedi
- lC:. C:ntrol to:
- rive (CX3) lasert and Withers = Lines .
C?tterta 11 s:ncarns tasse 1 tees of tre rea:t:r :: lant pressure bouncary genetrattag the artmary react:r c:ntaineest. De CRO fniert ane itscrs= lines are not eart of me react:r ::alant :ressure scuncary. ne classification of tw f asert and wiucraw lines is Cuality Grous 8. anc Derefere
, cesignet in accorcance =tti a!?! fe:: ton !!!. Class 2. The 34sts ts =nten ce CA3 Ifnes are :est$ned ts c:measurate .tu ce safety tesortance of Isolating uese Itnes. Stace taese lines are vital to tae scras fuiction. taene acerantl.ity is of womest cancern.
In tae cesign of tais systes, it has been accostes crsctice to sett automatic valves *ar tsalatten surcoses as uts intreauces a sessiste fatture escrantsa. As a reans of providing scsttive actuotton, annual snutsff valves are uses. In the event of a sreak on tRese Itnes. the manual valves ney te closee to ensure isolation. In asettion, a ball casca valve located in the f asert Ifne testse ue CAO
, is designed to automatically seal this itse in the event of a break.
~
!!. This 5 stas caett valve is nor:411y in a closed =csttien due to its check valve feature but 1:s PO is in the ooen :osttton. The m provtces a tactuo ta close the valve to provice addittenal n1ga less signt intagetty. -
- 22. Atreviations used in table '
M . Air C:erated M . Fotar C;erated VTC . Pneumatic festante Chect Valve
- RMit ~ . Aestewal Peat Aereval Systes RPY . Aeactor Pressure Vessel
, RCIC . Reactor Cars !selation Cooling Systes LC'J
. Reac*ar Water C1esaus hPCI . Mf gn Pressure Caolant Injection i
GCC . General Cesign Criterton - RSC:.C'J - Reactor tutiding Closed Leos Cooling Water
- 717 . frsasverstag Incare prece l CRO . Control Red Crive .
M5!V Main Steam !selation Valve-
- s h
i 1 k'
- Table B-1 (continued) .
PROCESS P!PE!.INES PENETRATING PRIMARY CCNTAIMOT
!!CLCION SIGNAL. ' NOTES .
!* 0m. . Ct!:e!P'!ce A'
Reactar vessel low mater level 3. (A scram =t11 oc:ur at this level) l' teneter vessel low water level 2 . (The reactor core tsalatten c:c1f ag systaa ane tne nign cryssure csolast injeciten systes will te intitated at tats level, and recirculation sens are trtaced) C' Hign radiation . asin steam line
, O' Line break
- main $*.eam line (high steam flem) '
(* Line break matn steas Itne.(steam itne tunnel nign tesserature) F' Mfgn drymell pressure 3 2eactar vessel fo. =ater level 1 - (The core scray systens and tse few aressure core injecitan esos af AMR systees allt to tattiates at tats
, level)
J' '.ine treat in rtactar =ater cleanus systee . ntti s: ace tescerature, nign differential fles, hign differential tesserature c' Line treat In ' state Ifee te/ free turnine (hign steam Ifne teace temeerature.
... Mign steem flow. Iow steam Itne pressure or nign turstne ennaust disparage pressure) -
L Aeactar bullef ag s'.anity ventilatten systen lattf atten , M Mign radiation signal deanstrene of primary containment purge filter trata .
. O Mfgn asetent tesserature in esta steam suaael penetrasten arca (MSTM)
P' Law asin steam 1tne 3ressure at inlet to turstne (tuu mese only) A Lew csadenser vacuus l 7 Mfgn temcersture fa furstne lullding U Mign reacter vessel pressure id* Mign toscerature at outlet of cleanus system nenregenerative heat escnanger
- I Lee stone pressure .
f Stanear 114v14 control system actuated 1 Lew level in ASCLC'J head tant RM'
- emote manual swite.1 fr:e main c:ntrol roca *
* !Rese are tse tsolatten fur.cttens of the prieary containment and reactor vessel tse14 tion c:ntrol systemt ot.*er f6nctions are given for infor atten caly. '
e _1_____________.____'..______
.. . . _ - . _. -. - . . ~_. - . . . - - - - - . - - - - . .-. . . -
=.
d 9m n4 --s ! C-1 -
' .: l 1
Appendix C l Identification of Pipe Sections and Discontinuities for i
- Break Frequency Estimation ,
!, Main Steam Lines .
- f 11 secti,ons of the four lines in the Reactor Bu'ilding are break'
! exclusion. Two sections are considered: one to the outboard MSIV; one from !' the outboard MSIV. , l Main Feedwater Lines a ]~ All sections of the two lines in the Reactor Building are break ) ~ exclusion. They include check valve inboard and testab1'e check valve [ outboard. Their failure rate is assumed to be similar. I
- Hich Pressure Coolant Injection (HPCI) l
Reference:
FSAR and LILCO drawings no. M10121-17 and M10122-14. ' j
Description:
10 in.: one section and valve to the outboard valve (M0V-041). i Break exc)usion. Under normal RPV pressure conditions '.. because inboard valve ls' open. , ,, q 10 in.: six nonbreak exclusion sections (4 challenges per' year , j i of 24 hrs each are assumed in these sections):
- To reducer l '. -
Branch SHP-171 + valve M0Y-049 ] Reducer / valve F001 {- - To steam turbine stop valve l. To turb1ne admission valve -
- The turbine assumed to be ' equivalent to one section i
1 in.: two bypass sections and a valve. Six sections downstream - l to the RCIC/HPCI drain line. Two branches. All nonbreak l ) , , exclusion. Normally open. : l i ' ] t l i . t I ,i
- i ' '
! L1 .
_.7 - C-2 , RCIC .
Reference:
FSAR and LILC0 Drawings No. M10116-16 and M10117-13 ., !
Description:
4 in.: open MOV inside drywell to the outboard M0V. It has a i l bypass line of 1 in., normally open. Break-exclusion. ,
. six sections and discontinuities-3 in.:
I - 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. 1 in.:
- Bypass is 2 sections
- Drain lines from drain pot. to RCIC/HPCI drain line are con- -
! ', sidered six sections. , f Branches : two or more 3/4 in branches. ,
l Quantification: ~ 4 in.: [8.6(-11)+1.5(-10)3*8760=2.1(-6) P 3 in.: [8.6(-9) + 1.5(-9)]
- 6
- 4 (times per year) x 24 (hrs) = 5.8(-6) l 1 in.: [6 + 6 + 2) * [8.6(-9) + 1.5(-9)]
- 8760 = 1.2(-3).
Reactor Water Cleanup System (RWCU) Supply Line . . P
Reference:
FSAR Figures 3C-4-15A,B,C and Figure 5.5.8 1,2,3 I ,
Description:
6 in.: One break exclusion section and valve . 6 in.: One section nonbreak exclusion to reducer .
. 3 in.: Two If nes (having three sections each), two valves ' each
', and one pump each.
~
l 2 in.: Two lines with section and' reducer / check valve. 3 in.: Two line with section, valve, section, reducer ; f s . i 1 -
~
2 C-3 - -
/
4 in.: One section and two valves. One of these valves is, normally closed. Another line with section, HX, section HX. The heat exchanger (HX) considered as one section in , our approximation. Seyond the second heat exchanger, temperature is less than 125'F and not
- ll cer.sidered 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 regenerative HX to the feedwater line is not considered a significant additional contributor, compared to the part already included. .
. Standby Liquid Control (SLC)
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 normally closed explosive valves is nonbreak exclusion section. Branches: four 3/4 in.-branches from the main' 1-1/2 in.-line. i ," Quantification: [8.6(-10) + 1.5(-10)] * (8'760/2)
- 3.3(-3) = 1.5(-8) . .
[8.6(-9) + 2 x 1.5(-9)] * (8760/2) * [3.3(-3)]2 = 1.0(-9) Control Rod Dr'ive
Reference:
NUREG-0803 The contribution comes from the Scram Discharge Header rupture as ' explained in NUREG-0803. The .value of the rupture frequency of 10 " is derived
, , from that report.
l l t . J e
. ._7 , ,
o . . C-4 . 1 Recirculation Pump Seal Injection . 1
- l.
Reference:
FSAR , (i
Description:
Two 3/4 in.-lines ; 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)'
j Sarole Coolant From RPV [
Reference:
FSAR i 4 <
Description:
3/4 in.-line; one normally open inboard , air-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. .. Quantification: 2 * [8.6(-9) + 1.5(-9)3
- 8760 = 1.8(-4) i.
Reactor Post Accident Sampling System (PASS) i
Reference:
FSAR '-
Description:
3/4-in. line. One manually operated ' globe valve outboard, . j normally open. Two solenoid operated gjobe valves, -normally i closed, downstream. , , i s , Quantification: same as above. TIP Orive Guide Tubes
Reference:
FSAR i -
, i
- -Description: four lines of 3/8 in. The tubes are normally with nitrogen. In 3, order to cause LOCA, all the following must occur: ,
l i i
- e !
{ a i -
--.7 .
C-5 -
- 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 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 systens 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.
+
. f e
I
- t I
e e p A $ 6 4 0 ______._._..________..__.________.____.________.._______._._______._____.__a-m.__}}