ML19206B315
| ML19206B315 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/26/1976 |
| From: | Arnold R Metropolitan Edison Co |
| To: | Kniel K US Atomic Energy Commission (AEC) |
| References | |
| NUDOCS 7905090113 | |
| Download: ML19206B315 (11) | |
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,.e METHOPOLil AN EDISON COMPANY PCST OFFICE SOX 542 RE AOING, PENNSYL'/ ant A 196C3 TELEPWCNE 2?3 - 929G01
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ATTACEMENT I SINGLE FAIIURE XUl.YSIS i
The TMI-2 ECCS network single failure analysis has been updated to assure that all of the considerations in your July 24, 1975 letter have been addressed:
A.
Evaluation Criteria The guidelines fer performing the Three Mile Island Unit 2 ECCS single f ailure analysis are:
1.
The emergency core cooling syste=s required are the Core Flood system, the Makeup (HPI) system, the Decay Heat Re= oval (LPI) system and the Safety Features Actuation system.
2.
The ECCS auxiliary supporting systems are the C'
.E electrical power system, the Decay Heat Closed Cooling Water system, the Nuclear Services Closed Cooling Water system, and t!.e Nuclear Services River Watr r system.
3.
Electrical syste=s are considered to mee t the single failure criterion if the inability of any electrical component to perform its intended function does not prevent accomplishment of the necessary ECCS functions:
a.
For relays, one failure code examined is where both the coil and contacts are considered to fail in either the energized or de-energized state.
b.
For relays, a second failure mede examined is where individual contacts are assumed to fail in either the closed or open state regardless of the status of the relay coil itself.
c.
For the preferred mad standby electrical systems, failure modes exa=ined include f ailure of any breaker, or f ailure of manual switches to perform their transfer function (the TMI-2 design does not utilize automatic transfer switches).
In considering the failure of transfer switches, the propagation of f aults to redundant syste=s has not been considered since credit is assured for breakers in limiting f aults.
Similarly, fuses are considered acceptable fault limiting devices.
i.
Mechanical syste=s are considered to meet the single failure criterien if the inability of any active rechanical component to perform its intended function does not prevent ac co=p lis hcen t of the neces ary ECCS functions in the short-term.
Similarly passive techanical failures were considered in evaluating the ability of the ECCS to perform its long-term cooling function:
O '3 kb c
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a.
For active cc=ponents, cne failure code examined is where the pu=p or valve is considered to fail in either of its two s table states; namely, running or stopped, or open or closed.
i b.
For electrically operated, =anually controlled velve s, cne f ailure mode examined is where the valve is considered to undergo an unanticipated change of state which wculd prevent the accc=plishment of a necessary ECCS function.
c.
For. passive cceponents, the change of state of =echanical components in the unsafe directions (e. g., spurious operation of a spring operated relief valve or of closure of a spring operated instru=ent check valve) have not been considered.
d.
For =echanical cceponents, ficw check valves and pressure relief valves are considered to assume their proper open or closed position when required.
e.
Hand operated =anual valves which are ad=inistratively locked in position are assumed to be in their proper position when required.
5.
The postulated failure of cc=ponent asse=blies, such as instrument racks or control roc = cabinets, has not been considered.
6.
The physical separation frc= postulated hazards, such as those resulting from whip or jet imp inge=ent, are assu=ed to have been adequately de=ons trated in the TMI-2 FSAR.
7.
The adequacy of electrical isolation of any ECCS circuit with respect to other redundant ECCS circuits or non-ECCS circuits has been considered at the circuit diagra= level; hewever, installation and coble routing details are assumed to have been de=enstrated as adequate.
3.
Seis=1c and envirencental equip =ent qualification has been assu=ed adecuate for both =echanical and electrical cceponents.
For each postulated component f ailure, the overall ECCS network has been cen-sidered in the determination of the consequences of that failure upon the plant.
The mini =u= ECCS require =ents (as identified in BAW 10103 "ECCS Analysis of 3&W's 177-F.A Lcwered Locp NSSS") for each of the four pcstulated accident conditions are:
1.
A pipe break >0.5 square feet requires that both Core Flooding sys te=s and cEe Decay Heat Re= oval sys te= operate ;
2.
A pipe break <0.5 square feet requires that both Core Flooding syste=s and one Makeup sys te= operate ;
3.
A Core Flooding syste= no::le break requires that the redundant Core Flooding sys te= and one Makeup syste= =ust operate; 23 157 I-2
4 For long term cooling, one Decay Heat Removal syste= nust operate.
3 Analysis Metheds The main elements of the random single failure analysis consist of:
1.
Success diagrams for each ECCS network.
2.
Failure mode and ef fects analyses.
3.
The evaluation criteria presented in (A) above.
4 Identification of the worst effect for the folicwing acciden:
considerations:
a.
A large break b.
A small break c.
A core flood systes no::le break d.
Long term cooling considering the ef fects of boron precipitation.
The consequences were then ranked according to severity in four classifications :
a.
Less than minimum ECCS capability for the accident considered b.
Minimum ECCS capability c.
More than minimum ECCS capability d.
No effect on ECCS capability.
D.
Analvsis Results The single random f ailure analysis confirms that the electrical and instrumentation portions of the Three Mile Island ECCS network are adequate to accccplish their protection functions under the stated evaluation criteria given that one single random failure exists at any point in the system.
All =anually controlled, electrically cperated valves located in the safety related fluid systems were reviewed. The valves listed on Table I-l eculd, due to a single electrical failure or operator error reduce the ECCS network to less than its mini =um capability.
The means by which these valves meet the single failure criterion is discussed.
The ability of the ECCS network to provide long term cooling considering the effects of boron precipitation is still being investigated. This report will be revised to reflect the results of this analysis by March 31, 1976.
jrn 30
)
t I-3
TABLE I-l VALVE MA R K th).
VALVE SERVICE SI!1GLE FAILURE COMPLIAtlCE WITil S!!1GLE FAILURE CRITERIO!1 CF/VIA/VlB Discharge isolation valve Failure of passive Breakers for these valves are tagged and for Core Flooding Tank component in the locked open by administrative control.
CF-T-1A/lB.
electrical system Redundant visual and audible. alar ms of can cause the valve valve positfon provided in control room.
to close.
SR-V 56 Isolation valve for 11uclear Same as above, Breaker for this valve is tagged and locked Services River Water (f3SRW) open by administrative control.
Emergency Water diccharge to Mechanical Draft Cooling Tower.
7 SR-V24 Isolation valve for hSRW Same as above during Ereaker f or this valve is tagged and locked discharge to Mechanical normal system line open by administrative control during normal Draft Cooling Tower.
The up (SR-V24 in open system line up (SR-V24 in open position).
valve is maintained normally position),
open.
Ilowever, dur ing Gooling Tower maintenance, the valve will be closed and the sluice gate SR-V55 located in parallel will be open, t
SR-VSS Sluice gate isolation for Failure of a passive Breaker for the sluice gate la tagged and tJSCW discharge to Mechani-component in the locked open by administrat ive contr ol when N
cal Draft Cooling Tower electrical system the parallel valve SR-V24 is c lomi (open W
Pit.
The gate is maintained during maintenance during Cooling Tower maint enance ot:ly),
normally closed and opem J on the Mechanical only during Cooling Tower Draft Cooling Tower maintenance.
can cause t im: gate to close.
.4.
6
AI~ ACE'Z'!T II JUEMIRGED 7AL7ES A review of all valve motors located inside containment was performed to deter =ine if any of these ectors could becoce subcerged following a LOCA.
The results of this review, presented in Table II-1, show that no valve motor will be submerged. There is a possibility that valve indication could be lost for several containment isolation valves. Valve indicatien would not be lost until af ter the valves had perfor:cd their function:
The maxi =um water level inside the reactor building is ccuservatively calculated to be 289'-1".
This number is based on the useable borated water storage tank volu=e (i.e., greater than the Technical Specification volume of the tank), chemical addition tank volu=es, and the reactor coolant system volume. The total volu=e of water introduced is 70,680 f t3, which results in 6'-7" of water to the 289 '-1" elevation. The next lowest elevation valves are CA7 4A/B at elevation 293 '-6" and RR7-26 A, 3, C, D, E a t elevation 293 '-11".
In order for the water level in contaic=ent to reach the 293'-6" elevation, 3 of additional vater vould have to be added to the contairrect.
abcut 47,400 f t This number is about 2/3 : ore than the volu=e which has conservatively been esti=ated to be available for addition to the containment.
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t i l l LT Ol' bilhMt.xLi t.CL l l i V A 614 f 4 t iltd.T il pl4 14 H A l'O!. l T i t.ti V Al.VL CONikul.
OF VAINt.
Ct 4 f?n t.T S titi VIA/h l e t ilown contor that s e spels ed to Open or flan.1 a c t ua t ril Valve re a s t.o n a nd limit =.wi t c h i t.
lio tifett un plant
/ H '#
- t" tiow iontent n.lttgate a 1.ot: A.
Closed f rom. cont e sel pon t f a l ly tut.mer ged, t.u t siot u s matety.
volve stom.
uperatot l es imi t.
I l'o e.a. il,1 = lues o f va l ver l a.i nt t ions.
plant I
till v / A / /it tontalunn nt C l o wd,
ope n ES.
Ila nd Valve tearl.on and limit =w i t t i. i.
tu effett un 1 tt9 '
6" l.ilatton ac t uat e.1 f rom partially e.uteennied, fut motor a.a t t s y.
tontsol room, ope at or in not.
valve.
Pommit.le lomo of valve s i.f it.a t ion.
18 V I A / l lt Ilow tonttol teo t s o <sul a rif t o Ope n u r llan1 attisated Valve Lear l.ox utkl limil t t,wl t c h i n the citett a.n plant i
g
.' h 'i '
d valve f or intes-ini t t) a t e allCA.
Clound f r ami C mt rol partially e.ul.p< r red. 1.u t aio t u s-s.dl o t y.
me.l f.e t e aloord s oon..
opetator l e. not.
3 ll
.. iling w. iter l'o u n.11,1 c loom ut valve ind i c a t ion,
tu l a t. lown i s ol e i.
I l-Ia n t Wlit V/
t..ntainment C l one.1 C l o s e.1 ES.
Bland Valve yearl.ox and l i a. i t ewitti. la f.o (tfitt ou o.en actuated from pa r t i a l l y a.ul.ine r ged, 1,u t imit u r a.a t t t y.
/ H 'e '
l lation on n
valve control room.
operator la i.o t.
t P oe. n l l. l e loss ait v.alve indtiativu.
plant Util V/
lu as lor tuulunt th a t t regui s ed t o C l n ht-d llausi ut.tuated f4un e.
taa atlett on.
.'H'**
/
ilialn poinp ley-ruf t ly u t e a loCA.
from control I
N p.o line inom.
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144 11.
1.
H.. n f i..on.. s e.l t i.i r wate level in61.te containment lu 269'-1".
ro !i l l. l e:
I...
..f v.e l v e t i t lic at ion wou l.1 inet oc a u r until aftsr valvem fia ve p= r f or mest t !.e lt i n t t nd est ma t e t y f usw t ia.ii.
y ATTACEMENT III MINIMUM CONTAINME';T BACKPRESSLTE i
The minbm containment backpressure was calculated for the 211-2 reactor building and heat re= oval systems using the cethods outlined in Section 4.3.6.1 of 3AW-10104. The analysis was perfor ed for an 8.55 f t.2 double ended break at the pu=p discharge; the break location and size which yielded the highest cladding peak te=perature as shown in Section 6 of BAW-10103.
Table III-l gives the results of the contain=ent analysis for TMI-2 compared to the reactor building used in the generic analysis. The comparison shows that the TMI-2 building yields a higher containnent pressure transient than that calculated by using the generic contai=ent
=odel described in Section 4.4 of BAW-10103 (i.e., the analysis of the generic plant in BAW-10103 is conservative for the TMI-2 reactor building.
Tables III-2 and III-3 provide the heat sinks and contaic=ent net free volumes for the TMI-2 plant. Building spray water temperature is assu=ed to be 40*F, which is a Technical Specification lower limit.
EAW-10103 describes the initiation ti=e f or contain=ent spray pu=p assu=ing of fsite power is available. Spray flow was =odeled as a linear ramp from 40 gym at 25 seconds to 1800 gpm at 70 seconds (i.e., while the spray headers are filling) and constant thereafter.
The TMI-2 contain=ent fan cooler heat re= oval rate was conservatively assu=ed to be the sa=e as the O{I-l heat removal rate (which was based on a 32*F cooling water inlet te=perature). TMI-l has a larger contain=ent fan cooler capacity than'TMI-2.
t v) 162 III-l L.
4
-s-s TABLE III-l CONTAINMENT PRESSURE COMPARISCN PRESSURE (PSIG)
TIME GENERIC (SEC)
TMI-2 B A'a'-10103
=
2.9 13.5 1 12.90 4.7 18.56 17.52 6.7 23.04 21.46 8.9 27.32 24.75 9.9 28.94 26.00 12 32.11 28.11 13 33.34 28.81 15 35.36 29.87 17 36.86 30.41 19 38.00 30.68 21 38.55 30.59 23 38.51 29.94 25 37.79 28.71 27 36.83 27.60 29 35.97 26.61 31 35.15 25.77 33 34.36
- 25. 11 35 33.66 24.53 37 33.07 24.08 39 32.46 23.69 42 31.63 23.22 46 30.54 22.77 48 29.99 22.50 52 28,95 22.10 56 28.20 21.84 58 27.85 21.78 62 27.21 21.66 66 26.62 21.59 68 26.37 21.53 72 25.88 21.41 76 25.42 21.31 78 25.2 2 21.28 82 24.82 21.21 88 24.30 21.14 92 23.98 21.07 94 23.8 3 21.06 96 23.69 21.03 98 23.5 4 20.99 105 23.09 20.90 115 22.58 20.78 1 25 22.08 20.65 135 21.64 20.52 145 2 1. 25 20.38 155 20.90 20.21 165 20.57 20.09 175 20.28 19.94 185 20.00 19.79 195 19.72 19.63 III-2
}
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TABLE III-2 REACTOR BUTI. DING EEAT SINK DATA Sink 1 - Building Cylinder Sink 8 - Painted Steel: Electrical Equip.
Area 64,200 ft.2 Area 26,000 ft.i Paint Thickness 0.0008333 ft.
Paint Thickness 0.000833 ft.
Steel Thickness 0.0418 f t.
Steel Thickness 0.006994 ft.
Concrete Thickness 4.0 ft.
Sink 9 - Painted Steel: Scall Pipinz Syst.
Sink 2 - Euilding Doce Area 3,074 ft.-
Ar ea 18,650 f t. 2 Paint Thickness 0.000833 ft.
Paint Thickness 0.000833 ft.
Steel Thickness 0.05452 f t.
Steel Thickness 0.0522 ft.
Concrete "tickness 3.5 Sink 10 - Painted Steel: Lee. Pipine syst.
Area 9,486 ft.Z Sink 3 - Building Floor Paint Thickness 0.000833 ft.
Area 10,800 ft.2 5 teel Thickness 0.031872 ft.
Paint Thickness 0.002333 ft.
Concrete Thickness 2.0 ft.
Sink 11 Stainless Steel: Piping Systen Steel Thickness 0.02083 ft.
Area 6,103 ft.2 Concrete Thickness 11.5 f t.
Stainless Steel Thickness 0.03788 ft.
Sink 4 - Internal Concrete Sink 12 - Stainless Steel: Insulation on Area 110,074 ft.2 Piping & Mirror Insulation Paint Thickness 0.002333 ft.
Area 31,386 ft.Z Concrete Thickness 2.87 ft.
Stainless Steel Thickness 0.00272 ft.
Sink 5 - Painted Structural Steel Sink 13 - Fuel Transf er Canal Lines Area 129,556 ft.Z Area 6,800 ft.2 Paint Thickness 0.000833 ft.
Stainless Steel Thickness 0.020833 ft.
Steel Thickness 0.04267 ft.
Concrete 4.142 ft.
Sink 6 - Painted Structural Steel Sink 14 - Stainless Steel: CRDM's Area 44,294 ft.'
Area 2,774 ft.2 Faint Thickness 0.000833 ft.
Stainless Steel Thickness 0.03125 ft.
Steel Thickness 0.03125 f t.
Sink 15 - Copper Cooling Coils and Sink 7 - Painted Steel: Heating & Vent.
Grounding Cables Area 101.050 ft.'
Area 11,330 ft.2 Paint Thickness 0.00833 ft.
Copper Thickness 0.007439 ft.
Steel Thickness 0.00625 f t.
{k III-3
~
TAELE III-3 NE" CO!TrAINME'.~r FREE VCLUME (See Attached Figure III-l for Identification of Nodes) i SODE 1:
Gross Volume 2,323,270 ft.3 Less:
(a) Concrete 299,285 ft.3 (b) Polar Crane 507 ft.3 (c) Large Tanks 4,240 ft.3 (d) Miscellaneous Piping 11,6?.6 ft.3 (e) Reactor Head 2,827 ft.3 NODE 1 Net Free Volume:
2,004,795 f t. 3 NODE 2:
East Steam Generator 74,756 fr.3 Compartment NCDE 3:
West Steam Generator 77,000 ft.3 Compartment NCDE 4:
Steam Tunnel 14,008 fc.3 SODE 5:
North Passage 7,580 ft.3 NODE 6:
East Letdown Cooler 3,980 ft.3 NCDE 7 :
West Letdevn Cooler 3,258 ft.3 NODE S:
Reactor Cavity 5,e89 ft.3 Net Containment Free Volume:
2,191,095 ft.3 III-4
\\
4
FIGURE III-l IDENTIFICATION OF REAC*tR EUILDING NODES Ni l o ' FT 3 2.00 x 8 N4
!+,0 08 FT 3 I
I I
N4 OR 5 A6T Si&A M
^
T G E N E R ATO R.
N 3 SVB COPAPARTMENT !
77,000 FT 5 74,75(o FT 3 1
1 I
i 1
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NS 7580 ET 3 i
NG Hy 3980 FT3 5288PT5 23 166 III-5 N
y U.s. NuC:.E A r. R E CU L A T O R Y C Ov' alON OOCK T NUM9E R l~
JRC=cnu 195
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NRC DIST318UTION som PART 50 DCCKET.'.1 ATERt AL l
8 FROM: ?etroccii~.an tCisCn CO.
DATE CF oCCbvE NT To.
A. Kniel l.
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2-20-in r
R.C. Arnold cATE RECEwE0 i
2-1-76 l
CLETTER C N OTO RIZ E D 887'
!NPUTTOfV NUM BE R C# COPIES RE CE a'. t o C OMICiN AL CUNCLA5318tEO CCCPY l
' CESCRIPTICN ENCLCSURE Ltr. re. our ltr. of 7-2J-7....
Request for additional infer.ation concerning
~
Emergency Core Ccoling Syste.m.... W/ Attach.ments,
Tabl es and S-igu res.......
6
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