Text

Forwards Generic Background Info on Natural Circulation Cooling Mode
	ML19294C157
Person / Time
Issue date:	05/15/1979
From:	Bickel J; Advisory Committee on Reactor Safeguards
To:	; Advisory Committee on Reactor Safeguards
References
	ACRS-SM-0152, ACRS-SM-152, NUDOCS 8003070234
	Download: ML19294C157 (85)
	v • d • e

{{#Wiki_filter:h,. 6 pessse - 6 /%L .p na !( m[o, UNITED STATES gg Jgg j NUCLEAR REGULATORY COMMISSION , s., l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o, W ASHINGTON, D. C. 20555 g-May 15, 1979 ACRS MEMBERS TRANSMITTAL OF GENERIC BACKGROUND INFORMATION ON THE NATURAL CIR COOLING MODE The enclosed material on the natural circulation cooling mode is provided for your information. Included is a simplified thermal-hydraulic analysis describing the mechanisms which drive the natural circulation process and their sensitivities. The effects of raised steam generator water level (higher thermal center) on natural circulation in straight tube steam generators is also calculated. The primary limitation on establishing and maintaining natural circulation is noted to be the maintaining of subcooled thermal conditions in the Reactor Coolant System and continu-ously available secondary heat removal capability (e.g., normal or auxiliary feedwater). This package additionally includes infomation highlighting the types of assumptions made in the Safety Analysis regarding natural circulation and how these assumptions are verified via the start-up Test procedures.

SUMMARY

OF FINDINGS (1) The simplified thermal-hydraulic analysis indicates that during natural circulation, coolant mass flow may be expressed as: k :K dTcor,( N j 54 % ; N9.$ e,('Tg g 7 - for straight tube steam generator NSSSs Wo

E O lcore ( Hs3b-Heh) 3'T V N M b

3 l ~N 2 j - for U-tube steam generator NSSSs i Oc is the coolant mass flow rate in 1bs./hr. K is a proportionality constant (e< 3erimentally measurable) AT core is the core temperature di 'ferential in 0F Hsgt is the relative height of the top of the steam generator in ft. Hsgb is the relative height of the bottom of the steam generator in ft. 8008070 2.3 f

is the relative height of the top of the reactor flow path in ft. Hct is the relative height of the bottom of the reactor flow path in ft. Mcb

r[Tpyis the specific volume of water at ToF and Pres 1bs/sq. in.

expressed in ft.3/lb is the flow resistance coefficient (from: A%s:(Khef) and is ct experimentally measurable. The range is typically between 2.0 and 3.0 (2) If subcooling is maintained and there are no non-condensable gasses coming out of solutign, when pressure is lowered, there will be very little reduction in We due to,the relatively incompressible nature of the. reactor coolant. Hence dWe is very small.

dpy, The effects of Prcs on the fuel to coolant heat conductance was not evaluated at this time.

(3) The effects of raising the thermal center in a straight tube steam generator was evaluated and it was found that this produces a linear increase in the steam generatorop term. For further information on this subject, please contact the author at (202) 634-3283. .N John H. Bickel ACRS Fellow

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i r I Table 4. Compressed Liquid Temperaturer-Degrees Fahrenheit 890* 7J0* 108-(* l 82* 100* 200* 800* 400* 800* 800* 820' 840' 860* 880* p 0.08854 0.9492 11.526 67.013 24731 680 8 1542.9 1786.6 2059.7 2365.4 2708.1 2895.1 3093.7 3206.2 l Saturated vs 0.016022 0.016132 0.016634 0.017449 0.018639 0.020432 0.023629 0 02466 0 02598 0m777 0.03054 0.03277 0.03492 0.05030 IJguld he 0 67.97 167.99 269.59 374.97 487.82 617.0 64 6.7 678 6 714.2 757 3 784.4 823 3 902.7 Ab h. se 0 0.12948 0.29382 0.43694 0.56638 0.68871 0.8131 0.8398 0.8679 0J987 0.9351 0.9578 0!905 14520 8A./ha. Ia. t (s.t. Tm=gl (v-v ) 105 - 1.1 - 1.1 - 1.1 - 1.1 e 200 (h-he) 4 0.61 4 0.54 4 0.41 4 0.23 (381.79) (e-es) 10' 4 0.03 - 0.05 - 0.21 - 0.21 (v-v ) 10' -23 -2.1 -2.2 - 2.8 -2.1 n 1 400 (h - h.) 4 1.21 4 1.09 4 0.88 4061 + 0.16 (e-e ) 10s 4 044 - 0.16 - 0.47 - 0.56 - 0.40 (444.59) s (v-v ) 10' -3.5 -3.2 -3.4 -43 -4.4 e 800 (h -he) + 180 +1.67 +131 4 0.97 4039 (s-e ) los 4 0.07 - 0.27 - 0.74 - 0.94 - 0.96 (48631) s (,-v ) 10' -46 - 4.0 -4.4 -5.6 ~ 6.5 -1.7 800 (h -hs) +239 4 2.17 +1.78 +135 4 0.61 - 0 05 t (518.23) (e-ss) 10' 4 0.10 - 0.40 - 0.97 -1.27 - 1.48 - 0.53 -6.9 -8.7 -64 3 (v-v ) 105 - 5.7 - 5.1 -5.4 . +1.75 +084 - 0.14 1000 (h - he) 4 2.99 + 2.70 + 2.21 l (s-e ) los 4015 - 0.53 -120 -1.64 - 2.00 -1.41 (544.61) s (v-v ) 10' - 8.4 -7.5 -8.1 - 10.4 - 14.1 - 17.3 e 1 1600 (h -he) +4 48 + 3.99 +336 + 2.70 4 1.44 - 0.29 (596.23) (a-en) 10' 4 0.20 - 0.86 -1.79 - 2.53 -332 - 3.56 (v-v ) 10' - 11.0 -9.9 - 10.8 - 13.8 - 19.5 -27.8 - 32.6 - 24 e 2000 (h -hs) 4 5.57 +531 + 4.51 +3.64 +2.03 - 038 - 2.5 -1.8 (635 82) (s-s,) 108 4 0.22 -1.18 -239 - 3.42 -4.57 - 5.58 -43 -2.6 I (v-v,) 105 - 13.7 - 123 - 13.4 - 17.2 - 24.8 - 37.7 - 61.9 - 65 - 67 -48 t 2600 (h-hs) 4 7.49 4 6.58 + 5.63 + 4.55 +2.66 - 0.41 - 4.0 -5.6 -5.4 -3.1 (668.13) (o-os) 108 4 0.25 -1.48 - 2 97 - 4.25 -5.79 - 7.54 -85 - 8.2 6.9 -3.4 (v-v,) 105 -16.3 -14.7 -16.0 - 20.7 - 30.0 - 47.1 - 87.9 - 101 - 122 - 146 -172 - 166 8000 (h-he) 4 9.00 + 7.88 4 6.76 + 5.49 + 333 - 0.41 - 6.9 -8.7 -103 - 11.8 - 12.2 - 8.9 (69536) (e-es) 108 4 0.28 -1.79 - 3.56 - 5.12 - 7.03 - 9.42 - 12.4 -13 A - 13.4 - 13.3 - 12 4 - 8.2 (v-v ) 10' - 17.5 - 15.7 - 17.1 - 22.2 - 32.1 - 51.0 - 98 0 - 114 - 139 - 177 - 240 - 299 - 354 0 e 8206.2 (h-hs) 4961 4845 + 7.25 + 5.90 + 3.62 - 0.40 -7.6 -9.8 - 12.1 - 14.6 - 17.6 - 19.4 - 21.6 0 (s-e ) 108 4 0.29 -1.93 - 3.80 -5.50 - 7.54 - 10.19 - 14.0 - 154 - 16.0 - 16.8 - 17.8 - 18.4 - 19 3 0 l (705.40) s D (v-s ) 10' - 19.0 - 16.9 -18.5 - 24.2 - 35.0 - 56.1 - 111.1 - 133 - 166 - 215 - 312 - 407 - 634 - 1815 g-s Nt 8600 (h - hs) + 10 44 4 9.17 4-7.90 4 6.44 4 4.01 -034 - 8.6 - 11.2 - 14 3 - 17.8 - 24.2 - 29 7 - 42.8 - 104.1 te-s,).108 4030 - 2.08 - 4.14 -5.97 -831 - 11.24 - 16.0 - 17.4 - 19.1 - 20.9 - 24.9 - 28.7 - 39 0 - 91.0 to (g - J3 o !'on ~

i l (v-v ) 105 - 21.5 - 19.2 - 21.0 ~ 27.5 - 40.0 - 64.5 - 132.2 - 160 - 202 -270 -400 - 528 - 821 - 2079 e 4000 (h -he) + 11.88 + 10.49 4 9.03 4 7.41 4 4.71 - 0.16 - 10.0 - 13 3 - 17.4 - 22.5 - 32.4 - 41.2 - 59.5 - 126 6 4 I (e-a ) los 4 0.29 - 2.42 - 4.74 - 6.77 - 9.40 - 13.03 - 193 - 21.4 - 24.0 - 27.2 -34J -41.0 - 55.8 - 112 7 s (v-v ) 105 - 24.1 - 21.4 - 23.5 - 30.7 - 44.9 - 72.5 - 151.5 - 184 - 234 - 313 - 464 - 611 - 937 - 2219 e j d500 (h-hs) 4 13.32 4 11.80 + 10.15 4 8.40 15.40 4 0.02 - 11.1 - 15.1 - 19.9 - 26.4 - 38.4 -49.0 - 69 8 - 138.9 (s-s ) 108 -{ 0.26 - 2.74 -533 - 7.60 -20 58 - 14.80 - 22.4 - 25.1 - 28.3 - 32.7 - 41.8 - 50.0 - 67.0 - 125.6 e (v-v ) 10' - 26.7 - 23.6 - 26.0 -31.0 - 49.6 - 80.5 - 169 3 - 206 - 262 - 350 - 518. - 677 - 1017 - 2309 e 8000 (h -he) + 14.75 + 13.08 4 11.30 4 9.36 4 6.08 4 0.25 - 12.1 - 16.7 - 22.0 - 29.5 - 43.0 - 54.7 - 76.9 - 146.7 (s-se) 108 4 0.22 - 3.07 ~ 5.92 - 8.40 - 11.74 - 16.47 -25J - 28.5 - 32.2 - 37.5 - 47.9 - 57.2 - 75 3 - 134.6 (v-v ) 105 - 29 2 - 25.7 - 28.4 - 37.2 - 54.2 - 88 3 - 186.1 - 227 - 288 -J84 -562 - 727 - 1076 - 2375 e 8500 (h - he) 4 16.18 4 14.41 4 12.47 41036 4 6.78 8 9.52 - 13.1 - 18.0 - 23.7 - 31.8 - 46.4 - 58.7 - 82.0 - 152.6 (s - e ) 10' -{ 0.20 -339 - 6.47 - 9.20 - 12.86 - 12.10 -28.0 - 31.5 - 35.6 - 41.5 - 52.7 - 62.4 - 81.5 - 141.4 r (v-v ) 105 - 31.7 - 27.8 - 30.8 - 40 5 - 58.7 - 96.1 - 202.9 e 6000 (h - ha) + 17.60 1 15.72 -l 13.62 -11839 -l 7.50 4 0.77 - 14.0 (e-e ) 10' -1 0.10 - 3.72 - 7.06 - 10 00 - 13.96 - 19.57 - 30.6 a r w t g -- M -- g o - -h 3 M 20 19 --_'k f--- 'l - Mq. 39 1 .<=

D g I~ 1 i8 is J

gy T3 p } - { {- it ~ y,q Q__.__( n S g s is ~ 130 7 2 I ~ I ~ I I y 34 gg g -r po Il gy gy I / ~ ~ ~ l ?."_ s 8 d [- __~ ~- ~ %q s 8 y ~ ~ ~~ ~ ~ ~ ~ ~ $ M 7 5,' [,' s 5- ; _7 - 2 s s S 4 4 s 3 ~ -s ~ 3 ) s ~~ - -~ s - 2 N s r 2 2 2 - "[' s[] 2[ _.~ l ~_. f if y [ l g " I I 2 1 _ l 5 ~ ~ I O 04 02 03 04 05 06 07 08 09 10 10 g (ntropy-Stu. per Ib. deg F $. g Fiacas 3. Enthalpy changes for constant entropy. N *g Ikparture of enebster froan Lt esloe inr esturated liquid.t the sam, ao er. ROPY. The week of Ise.tropie steady Sow eomprenslom le the &serence twaween eENT awe e.i.e. e .t ..m. e.t.e.,. W %- t a ..-.e.,_ .__y,.

teterJ,eAb # lid Lies 6 Sssu.ree. Core Cool;% _ I.+ o To ecbslisk a dJa h tJamd Ccculd,% Gd y p n h i 9.< b J E d 4s egie 6 unmi h ecs rea g @ N a %.Ud Cleml ' Ta y)> % t y Sr h uptcs,%h INkW Y ALCu O rdL A&i s W& L = L.3 + ATue aQia THoT =r l coLb h C. Q LA gg MN'*p a _ qiw fo a()sw 60*F y pd dQ. O C]d @ g. N h lldwi g lik',4 brfeau s h L. h y T a J W a Tg tp) = Eo* F + Tgg + '4 3 he Se. gijk.b[elbdq: Ssiviu a a,z c,(Ta(foi)-T jc Tomo) 3.g$4f W' j PA66 A2 o F 88 s4s edaddk ans e.o.naJ nk uk (pl W J ~h %s.Lciu )Jg 4* M@"% jme., I.4. IMhd % edda four a r ~ v s Fre z a s o u u m. a m a s w,

d) bs pus t ' 1600 Fk Ted(M - s'yy. ss 'F (100*F,Poo(>s'a) D* N {o edcMf-h Wt, E A cu g so.o,u)= a o.2 rug e 4 (soot, lonop5k): r71.34 (Uk/4, , 4 (too*F,loaap,,) - f (toDT,100psh) Ep*F Alr. n co = F 1(3cov, icoopi)- d(vo f, zooagrW = 1.ong /n g c1's-noo*F y, ap . y 5 3 7 ' o L t -i.e 3 3 s g.g a _p w a~a 8, Arom : ), so r + T o + _ ^S SulstAR i un dfa'>w - 7-th. cp 44 +. c s e = T~.wa + oG '~ o. cr m & lu =W'* q (qq q.6S*F - L. ) bQL p' = i,oo3 E3. I yq4. Es *F -7m,'; .t.& *F \\ a f@'y Cd a wh5 6 A> K '. p.ce,t s vio ' # " /.l*: g yw* - 3 ,3, c,5q s x io 6 S*4. dH.5/M m 4 " *'* _

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m s a F 6. 32.74 x 10 6 kk. 2 M. M m-

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6' fas " 8 r5;* Tra (p,) = srz.2., Tsd - 55F* 4b8 2I'f &,4M c 4(1001,soops;r) = 70.lY B1k/h. ) { zoo *F, 200psk) =

161.77 61k//h.

4 [ Tot *F, 200 psin)

27* 'I4 IET"/)f-h,=IlT'77~70IN

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12,639 2/3,/4,

-1

0 43,IT6 U/An.

3eo ,g3 g jg_ = So o

Ioo,246 Nnlla-y9,699 Abs.lh Voo*

v /s,417 &&- l oo

e i

ggo* lLrs,3 89 lbt/4. tu.s ad '7%F.25 of 88

' G) hr-pus = 600 gck ~Tsd (pu,) = 486.2 *F $ jQg. y a ds M; $ (too*F, Goog,) - Gs.64 89/4. J (2.ao*F,(pe psk): lbl.30 ANb-A (yo*F, b00pck)20.$6 BTitf&. g, /64.I0- H '4 .99a arg,7 = Ioo f 2.70.54 -Ibi 10 1.0/24 87kk op ~ ?3-tbo 99f/4 Y h 4' /.00Y 5 ,p -9_ b% (q36 2 'F -Teo66) oG.L. (l.oo9hF)(436.2*F-7eog) 5 MWs l M Wi-1 Ye. Tcous IAlc luus ~ n, 97Y Al'/L- ,,,9g &./A. o. o, 100-50,s** Abs./ L iso- ,, o,,, 2 16./k. '11, 9 7 5 Alr'- l0 ' u> o ' ,y,3qgjhs/Jan.. 100* sw-at sz.o Abs./L zy, gsq 193./&. 2co-4**" 4s,92.5 lbs. ldtA. 400

29. vvu ek/A.

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AMO-l FSAK 14.1.2.5 3 Methods of Analysis O A dete ued digite1 s1=u1ation of the v1ent was used to eve 1uete the trens1ent response to this accident. The model includes point kinetics and a multi-region. fuel pin model connected through ti=e delays to a pressuriser =odel. A steam generator model was included. It was assu=ed that the plant was operating with two pu=ps at 60% of rated power when the re=aining two pu=ps were started. The startup of the idle pu=ps caused the system flow to increase frcm 49% to 100% of design flow in 9 seconds. It was found that the maximum temperature decrease for this case occurs in less than one loop time and a=ounts to about 6 F. Conservative values of the moderator coefficient and the Doppler coefficient were assu=ed to exist at the time of the accident. The mini =u= tripped rod worth was used. The conditions used in this analysis are shown in Table 14-9 14.1.2 5.4 Results of Analysis The results are shown in Figure 14-16. It is seen that the =aximum neutron power is reached several seconds after the pu=ps are started, and the pressure peaks well below its trip point several seconds later. The mi=atch between the heat re= oval in the steam generator and the power generation causes this pressure rise. The ther=al power lags the neutron power and reaches its maxi-mum value several seconds after initiation of the accident. The results of this analysis are shown in Table 14-10. Based on the analysis it is concluded that the pu=p control circuitry is ade-quate to prevent adverse transients in that it prevents starting an idel pu=p O if power is above 22% of rated power. Additionally, the integrated control sys-tem would serve to limit the imbalance between the reactor and stea= generator powers and reduce the severity of the accident. Since the themal power does not exceed about 75% of rated power at full flow, the DNBR will be greater than 13 and since the pressure does not exceed the trip setpoint, the protection criteria are met. 14.1.2.6 Loss of Coolant Flow 14.1.2.6.1 Identification of Casue The reactor coolant flow rate is reduced if one or more of the reactor coolant pumps fail. A pumping failure can occur from mechanical failures or frem a loss of electric power. With four, independent pu=ps available, a mechanical failure in one pu=p will not affect operation of the others. Each reactor coolant pu=p receives electric power frcs one of the two elec-trically separate buses, and, as discussed in Section 8, one pu=p in each loop is connected to each bus. Loss of the unit auxiliary transformer to which the 6,900-V bus is normally connected will initiate a rapid transfer to the second transformer source without loss of coolant flow. Faults in an individual pu=p motor or its power supply could cause a reduction in flow, but a ecmplete loss of forced flow is extre=ely unlikely and would occur only if all off-site power were. lost simultaneously with loss of the unit auxiliary transformer. Power O 14-8 pyf % 0 f $l

loss vould cause i==ediate reactor trip independent of protection system ac-tuation. In spite of the low probability of this event, the nuclear unit has been designed so that such a failure would not lead to core damage. () 14.1.2.6.2 Reactor Protection Criterien The criterien for reactor protection for the less-of-Flev acciden; is that the minimum DNB ratio shall not be less than 1.3. For the locked rotor accident, the minimus DNB ratio shall not be less than 1.0. 14.1.2.6.3 Methods of Analysis The loss-of-ecolant flew accident is analyzed using a ecmbination of analog and digital cc=puter progra=s. Analog simulation is used to determine the re-actor ficv rate folleving a loss of pumping pcVer. Reactor pcVer, coclant flew, and inlet temperature are input data to a digital program, which deter-mines the ccre thermal characteristics during the ficv coastdevn. The B&W digital ecmputer model used to deter =ine the neutron pcVer folleving reactor trip includes six delayed neutren groups, centrol rod worth and rod in-sertion characteristics, and trip delay time. The analog model used to deter-mine flow coastdown characteristics includes descriptions of flev-pressure drop relations in the reactor coolant loops. Pu=p flev characteristics are determined frcm manufacturers' :one maps. Flev-speed, flow-torque, and flow-heal relationships are solved by affinity laws. B&W has developed a co=puter code to calculate fuel temperature, cladding temperature, and DNB ratio as a functicn of time for reactor system transients. Input to the code censists of flew, power, inlet temperature, and system pres-The flow resulting from a coastdevn is calculated O. sure as a function of time. by a standard B&W analog model that includes a simulation of the pu=p and its associated inertia along with all pressure drops around the loop. The power, inlet temperature, and system pressure are calculated by a standard B&W digi-tal code that includes point kinetics with a closed-loop simulation. The Icop simulation includes a pr asurizer model and a steam generator =odel, the whole of which is connected through time delays to the kinetics calculations. The core transient analysis code simulates the reactor core through the use of a two-channel model. Each channel consists of one fuel rod with its as-sociated flow area and spacer grid geometry. Given the necessary input as stated above, the code vill calculate a pressure drop across a typical reactor channel (average channel) as a function of time. This pressure drop is then i= posed on the second channel (usually a hot channel) to determine hot chan-nel flov and DNB ratio in addition to fuel and cladding temperatures. Cem-pared to the average channel, the hot channel has greater heat generation and reduced flew area, as well as statistical hot channel factors. The analytical fuel pin model contains a transient response calculation, while the hydraulic model considers the steady-state solutions of energy, = ass, and comentum bal-ances at each time step. The transient response is obtained by applying the changing flev, pow r, in-e let temperature, and system pressure to the initial conditions of the average channel This calculation yields the average channel pressure drop as a func-tien of time, along with the hot channel power, inlet temperature, and system O 14-9 P4GE 37 6F Si

pressure. This yields the response of the hot channel in terms of cladding temperature, fuel temperature, and DUB ratio as a functicn of time. The loss-of-coolant flov analy:is has been carrie 1 out in the power rung, f..r coastdevn from power levels between rated and the design overpower conditicn. .l) The conditions used in the analysis are given in Taole 1k-11. I The conditions of Table 1k-12 were used in thermal calculations to deter:nine the minimum value that the DNB ratio would reach during the locked rotor nc-

cident, lb.l.2.6.h Results of Analysis The results of this analysis shev that the reactor can sustain a less-ef-ccol-ant flew accident withcut da= age to the fuel. The results of the evaluation are presented in Figures 1L-17 and 1h-18.

Figure ik-17 shows the percent re-actor flev as a function of ti=e after loss of all pu=p pcVer. Figure 1L-18 shows the minimu= DNBR vhich occurs during coastdown from various initial power levels using the minimum tripped rod worth (assuming 1% ok/k hot, shutdown margin). The degree of core protecticn during coastdown is indicated by com- ~ paring the mini =um DNBR for the coastdown with the criterion value of 1.3 Under normal conditions, the maxi =um indicated reactor pcVer level frc which a loss-of-coolant-flow accident could occur is 102 percent rated power. This power level provides an allevance of plus 2 percent rated pcVer for heat bal-ance error. Even with this error, Figure lh-18 and Table IL-13 show that an acceptable minimus DNBR exists. ~ The reactor coolant system is capable of providina natural circulation flev l after the pu=ps have stopped. The natural circulation characteristics of the reactor coolant system have been calculated with conservative values for all resistance and form less factors.. No voids are assumed to exist in the core or the reactor outlet piping. Table lk-lk shows the natural circulation flev capability as a function of the decay heat generation. These flevs provide more than adequate heat transfer capsbility for core cooling and decay heat removal by the reactor coolant system. ,s The reactor is protected frem the consequences of reactor coolant pump fail-ure(s) by the reactor protection system and the integrated control sys tem. The integrated control system initiates a pcVer reduction upon pu=p failure to prevent reactor pcVer from exceeding that permissible for the available flov. The reactor is tripped if insufficient reactor coolant flev exists for the power level. Thermal calculations perfor=ed to determine the mini =um value of DNBR that would be reached during the locked rotor accident are shown in Figure 14-19 As the figure shevs, the DNBR initially decreases very rapidly from its ini-tial value. After the flow transient ends, however, the DNBR decreases much less rapidly and reaches a new steady state value. The transient is termi-nated by the flux-flov trip, and as the figure shovs, the DNBR increases in response to the decreasing pcVer, going above 1.3 and continues to rise there-aft er. At no time during the transient does the DNRR go below 1.0; therefore, no severe fuel rod or cladding te=perature excursiona are expected to occur as a result of this transient. IL-10 en<a: 3s or ei

One additional loss-of-coolant flow mechanism has been analyzed in the re-actor design evaluation. This involves possible flow or leakage past the seat of a reactor internals vent valve. These valves are designed to be closed dur-t,) ing all normal operations and during all normal and accident transients except 7 for those accident transients involving reverse flow through the core. The design provides for positive closure even with no flev, and several rotational clearances are provided to ensure free motion and to prevent any tendency to stick. The valves also have a self-alignment feature to prevent reactor cool-ant leakage. Hydrostatic and vibrational tests have been made to demonstrate that the valves vill operate as designed. However, the case of a reactor in-ternals vent valve remaining open has been analyzed. This malfunction reduces the effective core flow and results in a reduced DNBR at the steady-state de-sign overpower. The steady-state design overpover DNBR for these conditiens is listed in Table lk-ll. The minimum transient DNBR for these conditions is listed in Table 14-13. lk.1.2.7 Stuck-Out, Stuck-In. or Droceed Control Rod Accident 14.1.2.7.1 Identification of Ce.use In the event that a control rod cannot be moved, localized power peaking and shutdown margin must be considered. If a control rod is dropped into the core while operating, the resulting transient must be examined. Adequate hot shutdevn margin is assured by requiring a suberiticality of 1 percent ak/k with the control red of greatest worth fully withdrawn frc= the The nuclear analysis reported in Section 3.2.2 demonstrates that this core. criterion can be satisfied. This criterion has been analyzed in terms of the minimum tripped rod vorth available in the loss-of-coolant-flow, rod ejection, O startup, rod withdrawal, And stea=-line-failure accidents. In all cases the available rod worth is sufficient to provide =argins belov any de= age thresh-old. If a control rod deviates from its group reference position by more than an 27 indicated 9 inches, the Control Rod Drive System and Integrated Control Sy-stem vill inhibit all rod-out motion, and the steam generator load de=and is run back to 60% of rated load. The details of these actions, which occur for both a dropped or stuck rod, are described in Sections 7.2.2 and 7.2.3 14.1.2.7.2 Reactor Protection Criteria The criteria for reactor protection for this accident are: The minimum DNB ratio shall not be less than 1.3 a. b. The reactor coolant system pressure shall not exceed code pressure limits. 14.1.2.7 3 Methods of Analysis A detailed B&W digital model has been used to analyze the transient response to a dropped control rod. This program includes fuel pin, point kinetics, pr.essurizer, and loop models, including the steam generators. O Amendment No. 27 14-11 August 4, 1972 PAGE 31 of si

The reactor is assumed to be operating at rated power when the control O rod is dropped. To achieve the most adverse response, the most negative values of the moderator and Doppler coefficients were used along with the maximum calculated rod vorth for rated power operation. In addition, no Control Rod Drive System or Integrated Control System action was assumed to occur. The parameters used in this analysis are shown in Table lk 15. 14.1.2.7.4 Results of Analysis The results are presented in Figure ih-20. The neutron power decreases causing a rapid decrease in both the core moderator temperature and the fuel temperature. These temperature decreases overcompensate for the vorth of the control rod, and the neutron pcver rises slightly above the initial neutron power level. The neutron power then decreases to below the initial power level and eventually levels out at the initial power level. The ther=al power response is similar to the neutron.pever; however, the ther=al power level never exceeds the initial value. Both the core moderator temperature and pressurizer pressure decrease during the transient and level out at a value lover than the initial value. Since the ther=al power never exceeds the initial value and the pressure decreases during the transient the protection criteria are met. Cases have been run for rod drops at beginning-of-life conditions and lover rod worths. These transients are not included in this discussion because they represent less severe conditions than the end-of-life conditions and the =axi=um calculated rod worth. 1h.1.2.8 Loss of Electric Power 1k.1.2.8.1 Identification of Cause 'Ihe unit is designed to withstand the effects of loss of electric load or electric power. Emergency power systems are described in Section 8. Two types of power losses are considered: a. A loss of load condition caused by separation of the unit from the transmission system. b. A hypothetical condition resulting in a complete loss of all system and unit power except the unit batteries. 1h.1.2.8.2 Reactor Protection Criteria The criteria for reactor portection for this accident are: a. Fuel damage must not occur. b. Reactor coolant system pressure shall not exceed code pressure limits. c. (1) Resultant doses for loss of all AC power shall not exceed h 10 CFR 100 limits. 2h-lk.h (2) Resultant doses for loss of load shall not exceed 10 CFR 20 limits. Ih-12 Amenfaent No. 14 February 29, 1972 pu,E Lto oF 8l

~~ 1k.l.2.8.3 Results of Loss-of-Load Conditions Analysis The unit has been designed to accommodate a loss-of-load conditien without a C reactor or turbine trip. Under circumstances where the external system dete-j riorates, as indicated by system frequency deviation, the unit vill auto-matica11y disecnnect from the transmission system. 'a' hen this occurs, a run-back signal causes an automatic pcVer reduction to 15 percent reacter pcVer. Other actions that occur include: All vital electrical leads, including the reactor coolant pu=ps, a. condenser circulating water pu=ps, condensate pu=ps, and other auxiliary equipment, vill centinue to obtain power fro = the unit generator. Feedvater is supplied to the steam generators by the steam-driven feedvater pu=ps. b. As the electric load is dr pped, the electro hydraulic syste= closes the governor valves. The unit frequency vill change =cmentarily, but the gsverner vill' rapidly restere the set frequency. c. During closure of the turbine governor valves, steam pressure in-creases to the turbine bypass valve setpoint and =ay increase to the steam system safety valve setpoint. Steam is relieved to the condenser and to the atmosphere. Steam venting to the atmosphere occurs for a brief period following loss of load from 100 percent initial power until the turbine bypass can handle all excess steam generated. The a=ount of steam relieved to the atmosphere is shewn in Table ik-16. Stear relief permits energy re= oval from the re-actor coolant system to prevent a high-pressure reactor trip. The initial power runback is to 15 percent reactor pover, which is a O higher pcuer level than needed for the unit auxiliary load. This allows sufficient steam flow for regulating turbine speed centrol. Excess steam above unit auxiliary load requirements is rejected to the condenser by the turbine bypass valves. d. During the short interval while the turbine speed is high, the vital electrical loads ccnnected to the unit generator vill undergo speed increases in proportion to the generator's frequency increase. All motors and electrical gear so connected vill withstand the increased frequency. e. After the turbine generator has been stabilized at auxiliary load and set frequency, the station operator =ay reduce reactor pcVer to the auxiliary load as desired. The loss-of-load accident does not result in fuel damage or excessive pressures on the reactor coolant system. There is no resultant radiological hazard to station operating personnel or to the public from this accident, since only secondary syste= stea= is discharged to the atmosphere. Unit operation with 1 percent defective fuel and a 1-gpm primary-to-secondary tube leak has also been evaluated for this transient. The stea= relief ac-companying a less-of-load accident would not change the whole-body dose. The whole-body dose is primarily due to the release of Xe and Kr and is considered to be negligible. Release of these gases is not increased by the steam relief 14-13 PA66 4l Of Fl

~ e C. because, even without relief, all of these gases are assumed to be released to the atmosphere through the condenser vacuum pun.ps. The rate of release of iodine during relief would increase because the iodine is released in steam vented directly to the atmosphere rather than through the condenser and unit vent. The iodine contained in the 1 gpm primary coolant leak e is assumed to be carried off in the secondary steam flow of 5.56 x loplbs/hr at the rate at which it enters the secondary system. Table ik-16 gives the quantity of steam released, the activity of the iodine contained in the steam, and the resulting site boundary thyroid des e. The relative concentration factor from Appendix 2.A is based on mixing of the discharge in the wake of the reactor building and a vind speed averaged over the height of the reactor building. 14.1.2.8.h Besults of Ccmplete Loss of All Unit a-c Power The second power loss considered is the hypothetical case where all unit power except the unit batteries is lost. The sequence of events and the evaluation of consequences for this accident are: a. A loss of power results in gravity insertion of the control rods and trip of the turbine valves. b. After the turbine stop valves trip, excessive te=peratures and pressures in the reactor coolant system are prevented by excess steam relief through the main steam line safety valves sad the atmospheric dump valves (turbine bypass valve steam relief is lost due to loss of power to the condenser cooling water circulating pumps). Excess steam is relieved until the reactor ecolant system pressure is below the pressure corresponding to the setpoint of the atmospheric dump valves. Thereafter, the atmospherie dump valves are used to remove decay heat. c. The reactor coolant system flow decays without the occurrence of fuel damage. Decay heat removal after coastdown of the reactor coolant pumps is provided by the natural circulation characteristics of the system. This capability is discussed in the loss-of-coolant-flov evslustion (Section 14.1.2.6). d. The condensate storage tank provides emergency feedvater to the steam generators. The condensate storage inventory is 2 5-1h. 3 200,000 gallons. e. The turbine-driven emergency feed pump normally takes suction from the condensate storage tank and is driven by steam from either or both steam generators. The emergency feedvater system is discussed in Section 10.2.1. The controls and auxiliary systems for the emergency feed pump operate on d-c power supplied from the station d-c buses. O 1h.14 Amendment No. 25 March 31, 1972 PA6E y2 o F Si

In' view of the foregoing sequence, the loss of all unit power does not result in fuel damage or excessive pressure in the reactor coolant system. There is no resultant radiological hazard to plant operating personnel or to the public from this accident, since only secondary syste= steam is discharged to the j atmosphere. This transient has been evaluat'ed further under conditions where the plant is assumed to have been operating with both 15 failed fuel and a 1-gpm tube leak-age in one steam generator. This operation continues until decay heat can be removed by the steam generator with no tube leakage, and the atmospheric du=p valve associated with the leaking generator is closed. This results in the folleving sequence of events: It is assu=ed that the operator further opens the atmospheric du=p a. valves 10 minutes after the loss of power. b. Cooling down at the maximum available rate requires an additional h5 minutes to reach a temperature belev the saturation te=perature corresponding to the setpoint pressure for the steam safety valve having the lovest setting. c. The steam generator with tube leakage is then completely isolated by closing its atmospheric dt p valve, and the other steam generator is used to re=ove decay heat. As in the loss of load transient evaluation, the whole body dose does not change because of ' steam relief. The total integrated thyroid dose is shown in Table lk-17. The activity of the iodine contained in the steam was calculated by the sa=e method used in the loss-of-load accident above. 14.1.2.9 Turbine Overspeed 14.1.2.9.1

Background

Westinghouse turbines have never experienced a major structural failure of a rotating part that resulted in missile like pieces leaving the turbine casing. (References 1 through 13 apply). Three important criteria have contributed to this record: a. Present manufacturing techniques - factory inspection and test procedures ensure sound discs with mechanical properties equal to or exceeding the specifie'd levels. b. Redundaat control system - the main speed governing system vill nor= ally hold the turbine speed within set li=its. An overspeed trip device backed by a redundant overspeed trip device provides three lines of protection in all. c. Routine testing - testing of the main steam valves and the overspeed trip devices while the unit is carrying load. O ik-15 PME H3 of $l

Tabic lb-11 l) Loss-of-Coolant-Flow Accident Parameters _ I Initial Pcver and D:aER Conditions: Steady-State Design Overpower, 5 11h rated power g DNER at Steady-State Design Overpower 1.55 At full flov At 95% flow corresponding to an 1.h0 open internals vent valve Maximus Indicated Power, % 100 Maximum Real Fover, % 102 System Characteristics: The initial core inlet te=perature for each power level is assu=ed to be plus 2 F in error. The initial syste= pressure is assu=ed to be minus 65 psi in error. 22 The trip delay time is 620 ms. The percent of rate 1 neutron power (at beginning of life) j[) as a function of time after trip is shown in Figure 3-13. l 22 A trip delay of 620 =s was used in this analysis rather than the 300 =s trip delay used in Figure 3-13 This figure also contains the shutdown curve using the minimum tripped rod worth, which is the vorth required to ensure a minimu= hot shutdown margin of 1% Ak/k with a stuck rod. The pu=p inertia is 70,000 lb-ft2, C 1h-82 Amendment No. 22 December 1h,1971

6 Table 1h-10 Locked Rotor AccidOnt Parameters ) Initial Power, % rated power 102 Initial Flov, 5 design flov 100 Flux-Flow Trip Delay Time, ms 650 System Characteristics: The initial pressure and inlet te=perature are nominal values minus 65 psi and plus 2 F, respectively. Maximum design conditions were assumed for the thermal conditions. The flow decreases from 100% of its steady state value to 75% of the steady state value in 100 =s. Film boiling is assu=ed to occur at DNBR = 1.0. The reactor is tripped by the flux-flow monitor. Table ik-13 Su-a7 of Loss-of-Coolant-Flow Accident Analysis Minimus DNER during coastdown for loss of all four pumps and the O -locked rotor ~ case: Situation _ Criterion Result Coastdown from 100% power and 100% flov 1.3 1.82 -Coastdown from 102% power and 100% flov 1.3 1 75 Coastdown from 100% power and 95% flov 1.3 1.68 Locked rotor at 102% power and 100% flov 1.0 1.15 llh 1h-83 PA6C H5 0F 8l

Table 14-14 Natural Circulation C,,p bility Time After Decay Heat Natural Circulation Flow Required for () Loss of Core Power, Core Flow Available, Decay Heat Bemoval, Power, a 5 full flow % full flow 3.6 x 101 5 k.1 2.3 2.2 x 102 3 33 1.2 1.2 x 10h 1 1.8 0.36 1.3 x 105 05 1.2 0.20 Table 14-15 Dropped Rod Accident Parameters Moderator Coefficient, (ak/k)/F -4.0 x 10-k Doppler Coefficient, (ak/k)/F -1.3 x 10-5 Control Rod Worth at Rated Pover, % ak/k 0.65 Control Rod Drop Time to 2/3 Insertion, a 1.40 0 f e e O 14-84 PA6E Q& of Si

~ O 100 N \\ 80 e L .2 [ I - 70,000 lb-ft2 E 60 u a w 5 \\ E 40 ' N 20 0 0 4 8 12 16. Time, see ARKANSAS POWER & LIGIIT CO. PERCHIT RFACICR C001 ANT FIDW AS FIG. No. ARKANSAS NUCLEAR ONE-UNIT 1 A FUNCTION OF TIME AFTER LOSS OF RJMP POWER 14-17 fab 6 4] Of 8l

Je 2.0 i i i i i 1 18 A N E I .6 'x N N [a o e x x I, g m m 55 Flow Loss E -(Vent Valve Stuck Open) g a N E N N 'E 1.2 N E g% 1.0 100 102 104 106 108 110 112 114 Overpower at_fnien Coastoorn Begins 5 O ARKANSAS POWER & LIGHT CO. MDIIFUM DNBR WHICH OCCURS DUREIG FIo, 8;o, 11+-1 ARKANSAS NUCLEAR ONE-UNIT 1 COASTDOWN FROM VARIOUS INITIAL POWER LEVELS t'Abl MS oF 8I

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Lou %+ 2. FSM z l 14.1.12 LOSS OF ALL A.C. POWER TO Tile STATION AUXILIARIES = l In the unlikely event of a coc:plete loss of all off-site a-c power and I j turbine trip while the reactor plant is at power, the reactor is tripped. The first few seconds of the transient would be almost identical to the four pump loss of flow case presented in Section 14.1.6, that is, the pu=p i co.astdown inertia and reactor trip would result in a DNBR > l.3. After the j trip, decay heat will be accommodated by the emergency feedwater system. j This portion of the transient would be similar to that presented in Section f 14.1.9 loss of normal feedwater. I Plant vital instruments are supplied by the emergency power sources. j a. (See Section 8) l b. As the steam system pressure subsequently increases, the steam system power relief valves are automatically opened to the atmosphere. Steam f bypass to the condenser is not available because of loss of the circulating water pumps. i As the steam flow rate through the power relief valves may not be c. sufficient, the steam generator self-actuated safety valves may temporarily 8 lift to augment the steam flow until the rate of heat dissipation is sufficient to carry away the sensible heat of the fuel and coolant i above no-load temperature plus the residual heat produced in the reactor. i d. As the no-load temperature is reached, the steam system power relief f valves are used to dissipate the residual heat and to maintain the i plant at the hot shutdown condition. E j The loss of normal feedwater supply signals the start of the auxiliary feedvater } purps. The turbine driven pump utilizes steam from the secondary system f to drive the feedwater pump to deliver nakeup water to the steam generators. j The turbine driver exhausts the secondary steam to the atmosphere. The electric motor driven auxiliary feedwater pumps are supplied power by the diesel generators. The pumps take suction from the condensate storage tank for delivery to the steam generators. I = 14.1.12-1 PA6E so oF fl 5

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rollowing the turbine trip, there is a rapid reduction of steam generator water level. This is due to the reduction of steam generator void fraction on the secondary side and because steam flow continues after normal feedwater stops. By one minute, flow is established from at least one auxiliary feedwater pump and further reduccion of water level is slow. The capacity of the auxiliary feedwater pump is selected to prevent the water level in the steam generators being fed from receding below the lowest level within the indicator range during the transient. This prevents the tube sheet from beco:ning uncovered at any time during the transient. The reactor operator in the control room monitors the steam generator water level and controls the feedwater addition with remote operated auxiliary feedwater control valves. The steam driven feedwater pump can be tested at any time by admitting steam to the turbine driver. The electrically driven auxiliary feedwater pumps also can be tested at any time. The auxiliary feedwater control valves and power relief valves can be operationally tested whenever the plant is at hot shutdown and the remaining valves in the system are operationally tested when the turbine driver and pump are tested. Upon the loss of power to the reactor coolant pumps, coolant flow necessary for core cooling and the removal of residual heat is maintained by natural circulation in the reactor coolant loops. The natural circulation capability for the unit has been calculated for the conditions of equilibrium flow and maximum loop flow impedence. The analytical model used to calculate the natural circulation flow has given results within 15% of the measured flow values obtained during natural circulation tests conducted at the Yankee-Rowe plant and has also been confirmed at San Onofre and Connecticut Yankee. The natural circulation flow ratio as a function of reactor power is given in Table 14.1.12-1. The average temperature, pressurizer water volume, steam generator water volume, and steam generator level assuming the most conservative initial plant conditions and equipment availability are shown in Figure 14.1.9-1 of "im. of Normal Fecdwater Accident." It is shown in that discussion that a loss of all off-site a-c power to the station auxiliaries and a loss of normal feedwater does not result in water relief from the pressurizer relief or safety valves. 34.1.12-2 fab & SI O F %I

TAl:LC 14.1.12-1 NATURAL CIRCULATION REACTOR COOLANT FLOW VS REACTOR POWER Reactor Power Reactor Coolant Flow % Nominal 0.5 2.3 1.0 2.9 1.5 3.3 2.0 3.7 5.0 5.0 10.0 6.2 ) PA(4 62 OF SI

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NATURAL CIRCUIATION TEC':'

REFERENCE:

1. CPL-SU C-8.4 t CONTE.'~0 A. DESCRIPTION B. TESTIIiG C. TABLE - TEST DATA 1. Table 1 - Natural Circulation Data [Wb b ~gf g %f , # 4 W"A /4tdi#2, <xudy,/972 $ hW ppy % of 9

NATUhAL CIRCU!.ATICN TECT A. LEOCRIPTIOn This test was performed to verify the ability of natural circulation of the reactor coolant system to remove decay heat u:ing the auxiliary feedwater system and dumping steam from the steam generators. Also demen trated was the ability to cool down ucing stea:a dump from onc ctea~. generator. Although thi is not a require-ment in this plant (3 - loops) where two steam generator would be available for heat removal with tube leak and isolation of a cinclu eteam generator. This tect was performed before any appreciable decay heat was present in the core, therefore nuclear heat wa: used to provide the her.t cource by having the reactor critical at a low power level, then securing the RC pumps. Limitations were placed on reactor power (/c measured by detector current; and limited by P-7 permissive circuit)AT and reactor coolant temperature as measured by the incore thermocouple. Also precuutionary mea:ure: were taken not to make sudden changes irr feedwater flow or to start a RC pur:p while the reactor was critical. B. TESTUG With the reactor critical and at a steady state condition of %2% the reactor coolant pumps were tripped. Plant temperatures were monitored to determine when steady state conditions were reached. After stable conditions were reached data was recorded (Refer to Table 1). After completing data taking, power was increased in steps until a maxir:um of % 8% was reached. At each increment of power the plant was allowed to stabalize and data was taken prior to increasing, power to the next higher level. After data was taken at," 8% power the reactor was brou6ht subcritical, and the control rods inserted sufficiently to pernit a cooldown of at least 50 F. At this point steam generators 3 and C were isolated by c. losing the steam isolation valves. Cooldown from 547 F to 5000F was then demonstrated on steam generator "A". After etable at C 500 F data was taken reactor coolant pump rectarted and hot chut-down conditicn entabliched. (ypf 5') Cf $\\

C. TEST RESULTS 0 During the test a mvim ~ AT of 55 F (avg) was established at a power level as measured by detector currents of 587,. These results verify the ability of natural circulation to remove decay heat in a three loop plant. Also the ability to cooldown with a single steam generator was verified. (Refer to Table 1 for cooldown data). 9 gg ss oF Ifl

f* / TABLE 1 II.B. ROBN.30H tir:IT NO. 2 k-* i I l NATURAL ClliCUIATION Revision l' \\. July 9, 1970~ ~ .I CPL-SU-C-8.4 8 i DATA SHEET (Continued) 1 Record Data For Conditions Specified Plant Parameter Initial Al ter Equil.. a After Equil'. After Equil. Af ter EMI. + After a 50'E~ 't 4% Pov'er atha

Power 20 Hias. at Cooldown at to be Recorded Conditions at 2% Power a

s L' i} ~ 10% power Suberitical I

3. Temperature f

J)) (Cold Leg)

y o

S '0 540 k70 - I (a) Loop A 545 .545I 5P, .530 1.80 5f0 (b) Loop B 530 530' 530 f 5'+0. 8*90 (c) Loop c ' 545; .545 540. .5 % 5 i g-B) (Hat Leg) Gq 'r EC ~ (a) Loop A 545. 565 575-596 596

>c5 -

[Q (b) Loop B $45.. 565 575 596 596 $O5 (c) Loop c, 540 - 560 570 $94 594 ',505 i .t

4. Feedwater FIow '

,p, (Cen) 7,, g l fs.V 2 31 1 (a) Flovmeters '..'O .O.0

0. 0.'

b({ .O.0 0.0. C5 J q =- .n.. - e d c;)

1. Steam

{ l6 t Driven Aux.* / l CC~ q p; e EM (b) Flovmeters, .O.O., . O.,0 , 0i '- h Loop,,a,,. 7 O . G.O. .O.0. 0.0 l p -1. S ;eam 3. o. Driven Aur.

s.; 7

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fi': n..

5. ..e v m i.. 4

l l..) L3 6* A3!E 1 (COi.T'D) T ,cvi,g,n 3 July 9, 1970 CP1.-SU-C-8. 4 4 t DATA SHEET (Continued) Record Data For Conditions Specified llant Af ter.Equil. Af ter Equil. Af ter Equil. + After a SO'F arameter Initial Af ter Equil. at 4% j'ower ' acMax Power 20 P.ir>. at Cooldova at _ s be Recorded Conditions at 2% Power 10Z power suberitical ~~ Change in Cond. i Tank tevel.

5. T bypass s.

hot i LYnos 539.097 537.10 538.15 536.00 ,457.83 toop "A"- Sk6.le8 3 toop "a". . ',.. Sk7.k1 * $3S.49 536.073 53d.49 535.1+9 i505.097

513.29: too, C" -

Si.7.49 560.66 538.61 536.88 536.097 g

r;a.. -

(#@ f,. T cold g,, i lines 547.17 54c.20 533.049 537.025 ' 536.63 158.78 M Loop "A". 539.325 I 536.625 537.05 535.37 LS3.83 too p "a" - 5;.50 I (... i.- 3, 536.25 535.097 l1.85.195 f 537.65 S*6.1.4 539.95 i A@ toop "c" (treader) I l 1000 1010 970 !k70 l f5)$ 7. steao Pressure 1010 3 1010 d o.. (... d.. i 8. reedvater ', s 1 j F Tea:perature ' s-e Loop "A" 137.6 137.6 149 1 133.6 135.5 11L1 9 j% Loep "a" 137 5 138.2 149 3 132.3 '133.8 '139.3 13 9 135.3 ! 144.2 131.1 '133 3 3138.g 6 too;, "* C". I 0.o 0" u-i v.., <.. v g R "r I l 4 i oI 'O g w ..~ s M I

TAB @l (CO!.T'D) ^ \\ i CPL-SU-C-8.4 DATA SHEET (CO!ffINUED) PIANT PARAMETER FICORD DATA FOR CO!!DITIONS SPECIFIED TO BE INITIAL APIER EQJIL. AFTER EQJIL. APIER EQUIL APIER EQJIL AT 6% PCf4ER AT IMX PO4ER + 20 MINS.AS 1 RECORDED CONDITIONS AT 2% P0h'ER MAX. PO4ER 9 Detector Current N 11 Upper .005 .0 016 .0092 .0190 .olB9 4 Ia.rer .0065 .006 .0132 .0268 .0262 N l>2 Upper .0044 .006 .0090 .0180 .c175 1 loyer .006 .0C6 .0128 .02h8 .0248 N-43 upper .0044 .004 .0090 .0175 .0172 Iower .0055 .005 .0118 .0230 .0230 N Ih Upper .0045 .004 .0098 .0192 .0190 t Iower .006 .006 .0132 .0268 .0262

10. Corrected Power Level Per Calorimetric Date, d

N 11 19 1.8 h.o 8.o 7.9 4 N 12 19 1.8 39 79 7.8 4 o 1.8 39 7.7 7.7 N l3 19 i o 1.8 1.8 4.1 8.0 79 N 1144 0

J.- s;ati 1 -(arr D).. N..' ,,(..; Y,, l ~ CPL-W.C 4 },, .[. i ?, DATA.S K (,!...1..,.C W;;..) J' i- - ' '.,,* j ,.,,,) s.: RECORD DATA FOR C01.. TIONs SPECIFIED 0 !.'I ~~ " . TO' BE INITI,[.[y AFTER EQUIL. AFTER 1*lIL. APfB EQUIL.

.Ps..., EQUlL.

AFTER A 50 F ANT PARAMETER ~* F5' CORDED CONDl'.'.' 4. AT 2% POJER ATh% PG.j AT l'AX POWER + EC W'.3S AT C00LDOWN AT IAf r ""R SUBCRITICAL l .'t.$Y ' ?*'. {'4l's,..' _ ;t a-;- -.b.*:.. '.'..~%. %,.c4: %n .d~

1

^

11. '36.QE RTD l

.s. l ILOOP 1 TE illD 119 21s 1415 83 3415 02 414,63 414.57 382.49 4 516.48 '539 097 ~537.10 '536.15 536.00 457.83 OlNS j 1 TEMP l Olces 418.46 /415.63 414.75 ,414,34 414.19 382.30 TE 412C '510.20 ~538.049 537.025 $36.63 1458.78 I 1 , TEMP 547.17 't 414.69 414.28 , 401.86 l LOOP 2 TE421D 1 Olms is19 14 ',1s15 50 414.52 TEMP 547.41 538.49 536.073 536'.!:9 -535.49 505 097 i TEh22C ~, 41, 51 ,1414 66 413 96 392.88 l 4 0104S 118.51 ',415.59 7 546.50 539 325 536.625 537 05 '535.37, 483.83 1 TEMP '( 1 16.37 415.53 1614.83 414.51 405.19 l LOOP 3 TE431D 119 15 6 Olms 4 5 7.49 540.66 538.61 536.88 536.097 513.29 1 TEMP l TE432C 118.67 416.C3 415 17 414.52 414.05 393.62 rEMP 546.!'4 539 95 537.85 536.25 535 097 485.195 ! OS.!S 4 g b (Bm i e D i to On b a

.,1 i l I TAD 1 (C0fff'D)' A CPL-SU-C-8.4 Revision 1 July 9,' 1970 NATURAL CIRCUI.ATION TEST DATA SilEET Record Data For conditions Specified t P t Parameter Initial After Equ11. Af ter Equil. Af ter Equil. After Equil. + After a SO Y j to be Recorded Conditions at 21 Power at 4% Pover at M.ax power 20 Mins. at Cooldown at Knx power Suberitical 1. Core outlet Tcoperatures .y TE-1E- (A-8)-35 550.2 571 5 584.2 605.0 604.0 ' 492.0 564.0 577.5 : 595 5 596.o 487.o I..- 515.5 TE-2E- (B-5)-55 i 4 f TE-3E- (B-10)-57 547.0 l 566.7 ' ' 582.2 i-601.7.', 602.2 433.2 5+7.0

569 0 583.2.

605.0 604.5 487 5 ~ TE r.E- (E-4)-55 I TE-5E- (E-7)-55 546.5 . 567 7 : 583.7 b-. 606.0 - 605.0 487.0 I I 1 87.0 l' TE-6E- (E-12)-57 546.5 566.2, $82.7 601.7 602.2. 4 517.0 ,I 488.2 579.0 600.2 $99 7 { 566.7,.. TE-7 E- (E-14)-57 4 i 567., 582 7 !603.2 602.7 486.5 8 TE-8 E- (F-3)-53 51.6. 0 605.0 605.0 1.86.5. TE-9E- (F-5)-51 546.5 \\ 567.2 - 583.7 ,' 603.0 - g 607.o 487.5 _ ^* TE-10E (r-9)-55 %7.0 .569.0 584.2 i 187 5' 317 0 568.2 584.2 \\ 606.5 605.5 T E-11 E-(F-11)-57 4 6 T E-12 E-(C-1)- 55 545.0 /564.0 g, 578.0 (- 608.0 6C's.5 596.5 596.5. 485.5 188.2 517.0 / 569.5 - 'yS3.7 TE-13 E-(C-6)-51 4 g 6 239 5,,, 239.5 239.5, 649.3 649.3 TE-14 E- (II-8) -5 3 - 239.5 I.86.5 566.7 $81.7 601 7 601.2 TE-15E-(It-15)-51, 546.0 'k. 517.5 570.0 l 583.7 605.5 605.0 6 487.5 TE-16E-(J-2)-53 TE-17E-(J-10)-57 51+6.5 - 567.74 584.2: - - 606.5 - 606.o'. ". l.8 7.o~ 565.7 '- 581.7,' 602.7 602.2.1 6 486.5 u TE-18E-(J-12)-57 545.5 567.2 581.7 ' '603.'2 602.7 1.86.5 3 TE-19E-(K-3)-53 546.5 5 81. 2 606.o 605 5 1.67.0, 517.0 - 568.2 6 6 TE-20E-(K-5)-53 5i6.0 566.2 581.7 N TE-21E-(K-3)-57 t 604.0 603.2 1.87.5 TE-22E-(K-11)-51 547.0 568.2 583.2 606.5 606.0 487.0 P TE-23E-(M-9)-51 546.5 566.7 582.2 605.0 603 2 487.0 G TE-24E-(N-6)-53 545.5 566.2 582.E 602.7 602.2 485.0 516.5 567.2 582.2 604.5 602.7 487 0 6 o TE-2*- 'P-8)-31 600.2 598.o' 1.87.5 547.5 566.7 581.0, 'r1 TE-:- . :-7)-53 TE-L (C-8)r55 547 0 567 2 582.2* 603.2 602 7 488.2 j

~. ACRSR!:!RENCEDDOCllMEN"-RETAT F0 RT E.!FE 0FTliCCI.EliTEi-R!F.ND~. li.!Z.':~'.ZES l DATE ?$ w& ZIGN NUCLEAR POWER STATICN UNIT 1 k STARTUP T2ST REPORT DLCKET I'C. 50 - 295 LICD:32 NO. DPR -39 NOVD3ER 1974 COMMOfD.'2ALTH EDISCN CCMPANY

an undervoltage condition on the RCS pumps to the initiation of control rod cotion for shutting the reactor down. The Underfrequency Trip Delay is essentially the maximum measured time from the initiation of an underfrequency condition for the power to RCS pumps to the initiation of control rod motion for shutting the reactor down. Thus, flow coastdowns are adequate for pump trips and the reactor trips are quick enouSh to prevent DNB. 5.C NATURAL CIRCULATICN Natural circulation flow was de=onstrated to be adequate for decay heat removal in the event that a reactor trip is accompanied by a trip of all four reactor coolant pumps. It was also demonstrated that the reactor coolant system temperatures did not become too high during natural circulation for the anticipated amount of decay heat removal. l The natural circulation test was started with the ' reactor at about 2% power when all the reactor coolant system pumps were manually tripped. The steam generator level and pressure were maintained constant by controlling feedwater flow and dumping steam. The most important parameters mea-sured during the test were reactor power, RC loop hot leg temperature, RC loop cold leg te=perature, and incore thermo-couple temperatures. The measured reactor power as a function of flow shown in Figure 5-3 was determined from measurements of the temperature increase across the reactor at three power levels. Figure 5-3 shows that there is more flow than that required PA66 6G cF 8l 5.3 m

.g by the FSAR. Thus, the maximum anticipated decay heat removal equivalent to 5% full power can be removed by natural circu-lation. In addition, it was demonstrated that the follow-ing temperature conditicns were not exceeded for natural cir-culation at 5.3% power: i No core outlet thermocouple exceeded 620*F. 1. 2. No loop tsT exceeded 55.46"F. 3. No loop average temperaturr, exceeded 574.73'F. $,D RTD BYPASS FLOW The RTD bypass flow is demonstrated to be large enough to ensure that there is less than a 6 second time de-lay from the initiation of high temperature in the reactor coolant loops to the actuation of an overtemperature ZLT reactor trip or an overpower /ST reactor trip. Thus, the ?SAR, Section 14.1 requirement is satisfied. The RTD bypass flows and transport times were measured (see Table 5-1) with the reactor core installed, the reactor coolant pumps running and at normal operating (no load) temperature and pressure. The value of the measured bypass loop transport times in Table 5-1 are larger than anti-cipated; however, the delay times for the trip logic were cuch less than anticipated for the overtemperature lit and over-power tsT reactor protection. Therefore, the results for the RTD bypass flow transport times are acceptable since the delay i time for overte=perature lit and overpower ZLT reactor trip is still less than 6 seconds. PM i~ *F 8l 5.4

I r --i : 1 .i l ---.--: H-----.--..--;...._..__ g.,.. ~. _ _ i 3

q -

6.

4 a

._:i : -i-F /t. 2 -0 ..i .;:._ = _ _ if:+- ..r. -- r v .. ~. .ex 3 0 /. C. .., / 'j ~. r4 . : i - _ _,. ____ ;._ - .'I r.. M r_..

.j7 -., y=;=t=,: j4 e

q.g.. y g --- _3 4 .a;.- e _._.% - i f..i / _ . _j...g.q5-i..g ;; -pg.. y C- ...g.. .,._..-g_._. I / .l.. _. p c . :-L .. j-./ i f .- -,,.m.,f i. p, --Measured. a-g .. = - .,..,.. =. _ _. e ~+ FSAR ..._.. ::. = = = :

. ' ". i.: --- d =_+ u",. -. _...' 3 a

... :J _". ' ' ~ ' " ~ ,a'Me'n i 12M----' ~-" -

r..

t g !. :L..-L2 - i - - !.i : / j.m f:. , e-.

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.ag g.{q r

3

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__ _' ~~ 0 2 4 d FLOk; percent of full flow Figure 5-3 Natural Circulation Flow For i Various Reactor Power Levels N #' # 5.9

STARTUP TGsT RE/DIT (c,- 99 MAME YAugEG 6.10 NATURAL CIRCULATIO.1 FLOh' !~ s' 6.

10.1 INTRODUCTION

I Natural circulation flow would be required in the unlikely event that pouer was lost simultaneously to all three reactor coolant pumps. Should this occur, the Reactor Proteettvc bystcc flow sensors would trip the reactor when total systen ficw Ge c re ased to 937. of normal flou. The turbine and generator irmediately trip and reactor coolant syst2m parameters would stabilize at escentially hot shutdown coaditions (see section 6.8. 3). During a rcactor trip in which oc least pcrtia2 reacto: coolant flow is retained, within a short period of time, inlet and outlet core temperatures essenti ally bacone equal at theac high flow rates source. Ecuever, through the core, relative to the vcry scall h- : if all flow is lost, this core c,T drops rapidly en first, ~out as the RC purps coastdown to zero speed the core exit temperatures begin to rise in proportion to the availabla decay heat. The decay heat of course, is a function of the previous operating power history and is governcd by recent powcr level and core burnup. Should the RC pumps reach cro speed and all ; low stop, core exit temperatures vould rise rapidly, eventually reaching saturation tcuperatures, while systen pressure vould also rapidly increase. However, the A T across the core provides the driving head for natural cos eve ctive flow, such that as the RC pumps coastdown, continued heat O' removal is available. Flou direction up through the core is encouraged by the direction of flow of the RC punps thenceives prior to stopping, as well as the differential density-gravity ef fect. Core exit temperatures will continue to risc however, at a relatively constant cora decay heat input, until equilibriun convecti"2 flow is reached. At this point the exit tempcratures will s cabilize. Early in core life this is a difficult test to perform becausc core decay heat is small, and thus the magnitudca of the critical paraneters are small. However, a carefully conducted and monitored test can yield satisfactory and meaningful results. Although many primary and secondary param'eters were monitored throughout this test, loop cold and hot leg tempcratures, loop A Tf. power, and. staan generator levul contribute the most significart data. Since heat inputs and losses were known f rom Post Core Loading Heat Loss Tests (section 3. 8), all normal primary systems were operated with the exception of the RC pumps. 6.10.2 On Decembe r 3,19 72, at a powet level of 3S%, all reactor coolant pumps were turned off, resulting in a reactor trip. Following the trip 3 core inlet and outlet temperatures were recorded providing data necessary to determine tne natural circulatioa flow. 23 minutes after the trip the feedwater flow to the steam generator waa s topped v so the steaming rate would provide the magnitude of the decay heat being generated at that time. From the decay heat being generated, and the coolant temperature rise across the core, th. core flow and power /flov ratio were determined. g4g g gg op fl ) 2,

6-60 The measured water level decreases in steam generators #1, 2 and 3 were 23.8, 23.9, and 23.0%/3 hours, respectively, with no feedwater flow. The heat required for this level change vas

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Adding 3.0 L' required to heat the charging flow and pu p seal flow 'and 2.1 IG system ther:al losses, results in a core decay heat of 11. 3 >2 (or 0.47% of full power) at cne hour aftar the trip. The coolant tc ?crature rise across the core at the saae ti..e was 10*F. This results in a calculuccd natural circulacu n flow rat; of 3.17 x 10" lbc/hr. (or 2.17% cf the full fic.; rate with all th ru reactc:. coolmat pumps operating). Taerafora tne pcwer/fic. ratio at ona hour af ter the trip vus 0.22. Figurc 6.10.2-1 shows the a T power calculated value as a function cf tire af ter the trip. This para eter represents a raasone.bly ;ood measure of the variation'of the core pwer/ fl.~ ratio during natural circulacion flow conditions. Core exit ec peratures were about 539-541*F just prior to the reactor coolant pump trip and dropped to 525-526*F right af ter the RC pump and subscquent rccctor trips. Cocidotta continued to 524-525'F until 16 minutes after the trips, when the exit ther occuples sicwly began to rise and then stabilized at 526.5-328'F, 25 =inutes later. Thereafter tc=peratures stabill:c or slowly decreased, as decay heat rate decreased, until the first RC punp was restarted 63 minutes af ter the trip. 6.10.3 Sci O.RY 't'ne natural circulation ficw teasurement verified the initictica of natural convective flow through the core as well as its adequacy in removin; core decoy heat and transferring it to a secondary heat re= oval sys tem. J 2 b I m O f%E (A oF tl

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Revision 7 ~. 06/21/78 20 nD _ _D COPY NRW Mid ISMND NUCLEAR STATION UNIT #2 EMERGENCY PROCEDURE 2202-2.1 V ASTER COPY sTATi0n stACx0ur Table of Effective Pages ]- } }{ { ~ Pg Date i Revision Pace Date Revision Pace Date R e' vision 1.0 12/30/77 3 2.0 12/30/77 3 3.0 06/21/78 7 3.1 03/30/78 6 4.0 02/27/78 5 5.0 06/17/77 1 6.0 12/30/77 3 7.0 01/29/76 0 8.0 03/30/78 6 Unit 2 Staff Recommends Approval Unit 1 Staff Recommends Approval Date Date Approval Approval Cognizant Dept. Head Cognizant Dept. Head Unit 2ePORC Recommends Approval Unit 1 PORC Recommends Approvai b'00 N b'" W Date k Date g 'Cha,dman of PORC Chairman of PORC Unit 2 S ri tende.nt proval ~ Unit 1 Superin'tendent proval d8 _ Date d //8 / Date V~' A' / / / Date Manager Generat on Quality Assurance Approval i T ui SS A Row 8/7 7 PAGE 92 0F D

Revision 3' 12/30/77 THREE MILE ISLAND NUCLEAR STATION UNIT #2 EMERGENCY PROCEDURE #2202-2.1 STATION BLACK 0UT 1.0 SYMPTOMS Sepa ation from the 230KV system and Turbine Generator trip as l.1 indication by: Zero volts on 230KV BUS voltmeters on the Electric Control 1.1.1 Console Panel No. 6A. Generator breakers open as indicated by alarm on Control Room 1.1. 2 Panel No.18. 2.0 It'NEDI ATE ACTION 2.1 Automatic Action Reactor trips due to loss of voltage to the Control Rod Drive 2.1.1 System. 2.1.2 Turbine trips. 2.1.3 Control Room DC lighting is energized. Turbine Driven Emergency Feedpump starts and commences to feed 2.1.4 the OTSG's (Once Through Steam Generators). Atmospheric steam dur.ip valves, tC-V3A and 3B, control main 2.1.5 steam pressure 1010 at psig. If a primary to secondary leak is detected, immediately CAUTION: close the affbeted OTSG's atmospheric dump valve isolation valve (MS-V1 A or IB) to eliminate the potential of leakage and an unmonitored release path. The Instrument Air Compressors will not restart NOTE: automatically. Therefore, the atmospheric steam dump valves, emergency feedwater regulation valves, pressurizer level control valve, and seal injection flow control valve will only function as long as the 1.0 PA6G 73 of fl

~;.. Revision 3 12/30/77 reserve air in the air receivers is available. The Instrumer.t Air Compressors should be restarted manually as soon as possible. T'brbine Generator DC Emergency Seal Oil Pump starts. 2.1. 6 ? 1 Turbine Generator DC Turning Gear Oil Pump starts. ~ 2.1.7 DC Emergency Bearing Oil Pump for the Main Feedpump Turbine 2.1. 8 starts. The DC Oil Lift pumps for the RC pumps start. 2.1.9 2.1.10 The Emergency Diesel Generators start. DH-VSA and SB automatically open. Both Motor Driven Emergency Feedpumps start. 2.1.11 Ventilation fans in the Control Room, Cable Room, Mechanical 2.1.12 Equipment Room, River Water Pump House and Diesel Generator Building start. 2.1.13 Make-up pumps (MU-P-1A and IC) start. (MU-P-1B may service i as Backup to either MU-P-1A or IC). 2.1.14 Nuclear Services River Water pumps start. Nuclear Services Closed Cooling Water pumps start. 2.1.15 The Intermediate Closed Cooling Water pump starts. 2.1.16 2.2 Manual Action

2.2.1 Verify

2.2.1.1 The Turbine Driven Emergency Feedpump starts. Atmospheric Steam Dump valves open when steam pressure reaches 2.2.1.2 1010 psig. k'hefollowingDCpumpsstart. 2.2.1.3 2.2.1.3.1 Turbine Generator DC Emergency Seal Oil Pump. 2.2.1.3.2 Turbine Generator DC Turning Gear Oil Pump. 2.2.1.3.3 DC Emergency Bearing Oil Pu...; for the Main Feedpump Turbine. I M l'$*Y SI 2.G

teur-c.: Revision 7 06/21/78 2.2.1.3.4 DC Oil Lift pumps for the RC pumps. Both Diesel Generators start and the generator voltage is 4160 2.2.1.4 volts and the frequency is 59-61 Hertz. T$e voltage available lights Unit Substations 2-llE, 2-12E, 2- { 2.2.1.5 21E, 2-22E, 2-31E and 2-41E energize. Verify the following equipment has started and the associated 2.2.1.6 systems are operating properly: 2.2.1.6.1 Nuclear Services Closed Cooling Water Pumps. 2.2.1.6.2 Motor Driven Emergency Feedwater Pumps. 2.2.1.6.3 Nuclear Services River Water Pumps. 2.2.1.6.4 Make-Up Pumps and DH-VSA and B open. 2104-2.3A. If the 2.2.2 " START" both Instrument Air Compressors per Instrument Air Compressors do not have sufficient capacity to maintain the air pressure above 85 psia, "CLOSE" the Turbine Building Instrument Air Isolation valve, IA-V29. 2.2.3 Verify the Intermediate Closed Cooling Water pumps start. Verify the Emergency Feedwater Regulation valves, EF-Vil A and 2.2.4 If EF-B, are functioning properly and "CLOSE" EF-V32A and B. VilA and B are not operable maintain OTSG level by manually throttling EF-V32A and B. - Verify the Turbine Driven Emergency Feedpump is functioning 2.2.5 properly and "STOP" both Motor Driven Emergency Feedpumps. Verify RCP seal temperatures did not exceed 185 F, and that 2.2.6 seal leakage plus seal ceturn flow does not exceed 1.9 gpm. If either condition exists, the RCP seals should be examined bhfore further RCP operation. If any of the following ICS Stations are ir Hand, Steam Generator 2.2.7 Reactor Demand, either Feedwater Demand, Main or Startup Feedwater Valve Demand, Feedpump Speed, Reactor Master and/or Diamond Station, runback the applicable ICS Station PAfsf 76 of 8) an

2202-2.1 Revision 6 03/30/78 3.0 FOLLOW-UP ACTION Attempt to restore one 230KV line and BUS based on direction 3.1.1 If at least one 230KV is re-energized, ,from the Dispatcher. ~ . re-energize plant 6900V, 4160V and 480V switchgear through the ~ auxiliary transformer to enable start of the Reactor Coolant Pumps and energization of the pressurizer heaters to enable a normal cooldown per 2102-3.2. 4 VA< Ib #f $l 3.1

Revision 5 02/27/78 If neither 230KV BUS can be energized, continue with cooldown by 3.1.2 natural circulation using the emergency feedpumps, the OTSG's and Remain the atmospheric dump valves to control the cooldown rate. ~ [ within the limits of the curve for cooldown by natural circulation (Figure #1). Pressurizer spray and heaters are not available for NOTE: pretsure control due to power loss. Primary system pressure decrease will be dictated by Pressurizer heat losses to ambient which will cause approximately 100 to 200 psi /hr decrease in primary system pressure. 3.2.1 If (2) MU pumps are running. Close DH-V5A/B as required to maintain MU Tank level. Trip a MU 3.2.2 pump as required to maintain pressurizer level. Borate the Reactor Coolant System using the Borated Water Storage 3.3 Tank as follows: Neither the Boric Acid Mix Tank, the Concentrated Waste NOTE: Tank nor the Reclaimed Boric Acid Tank are available as a source of borated water during a loss of Normal Power situation. 3.3.1 "0 PEN" DH-VSA and 5B. Manually "0 PEN" the RC Letdown to Bleed Hold-up Tank Isolation 3.3.2 Valve WDL-V46. Determine to which R.C. Bleed Hold-up Tank the RC Letdown will 3.3.3 flow and manually "0 PEN" the RC Hold-up Tank Inlet Valve, either WDL-V963, WDL-V964 or WDL-V965. CAUTION: These valves are located in a 3000 mr/hr radiation zone and exposure time must be minimized. 4.0 ff62 77 of g)

2202-2.1 Revision 1 06/17/77 Manually " POSITION" MU-y8 to the " BLEED" position. 3.3.4 3.3.5 "CLOSE" MU-V10. 3.3.6 h"CLOSE"MU-V12. -c - Monitor RC Bleed Hold-up Tank and Borated Water Storage Tank 3.3.7 levels. Following the addition of boric acid, sample the Reactor Coolant l 3.4 2103-1.9. for boron concentration and verify shutdown margin per During cooldown by natural circulation observe the following limits 3.5 Upper OTSG downcomer to cold leg maximum AT=25 F. 3.5.1 U The maximum cooldown rate is 100 F/hr. 3.5.2 Pressure to temperature limits are dictated by the curve for 3.5.3 l cooldown by natural circulation, figure fl. Manually " JACK" the Main Feedpump Turbines when they have stopp 3.6 The turbines should be jacked 180 every 20 minutes. The Main Manually " JACK" the Main Turbine when it has stopped. 3.7 l U Turbine should be turned 180 every 30 minutes. " START" one Control Building Area Fan-coil unit, AH-C-50A or B. 3.8 Do not exceed a load of 3 %' on either Diesel Generator. CAUTION: If the Reactor Building temperature exceeds 112 F " START" one RB 3.9 2104-5.1. ' Normal Cooling Unit and three RB Cooling Fans per U 3.10 If the RB penetration cooling air outlet temperatures exceed 140 " START" one RB Penetration Fcrced Air Cooling Unit, AH-C-48A or B, per 2104-1.7. 3.11 "STpRT" one Control Building River Water Booster pump and o Control Building Chiller Unit if required. ~ 3.12 Energize portions of the Heat Tracing System as required. 3.13 Between 1820 psig and 1600 psig RC System pressure, bypass channels for Safety injection. PME 78 or 81 5.0

2 Revision 3 12/30/77 3.14 At 700 psig RC System pressure "CLOSE" CF-VI A and B, and "0 PEN" and tag the power supplies to these valves. 3.15 At 500 psig RC System pressure rack out RB Spray pumps 85-P-1A and 1B, heakers to prevent unnecessary actuation and cut in low pressure inst umentation. (RC-3A-PI2). 3.16 At 275 F RC System temperature place the NDTT "AUT0-OFF" control 0 switch for the electromatic relief valve RC-R2 to "AUT0" to actuate the 500 psig relief set point. 3.17 When the RC System temperature is below 250 F shift cooldown from the atmospheric dump valves to the Decay Heat Removal System as follows: CAUTION: Do not exceed 3 MW load on either Diesel Generator. Line up the Nuclear Services River Water System for Decay Heat 3.17.1 Service Cooler Operation per 2104-3.1F. Do not start a second NSRW pump in the operating _C A. U. Ti ON : header to prevent overloading a Diesel Generator. " START-UP" the Decay Heat Closed Cooling Water lpg associated 3.17.2 with the Decay Heat Servi:e Cooler placed in operation per step 17a above. Refer 2104-3.3. " START-UP" the Decay Heat Removal Loop associated with the 3.17.3, Refer to operating Decay Heat C1' sed Cooling Water Loop. o 2104-3.3. 3.18 When the Decay Heat Removal System is operating properly and is capable of removing of all the decay heat generated in the Reactor, terminate heat removal with the OT5G's as follows: "CLOSE" the Atmospheric Dump valves, MS-V3A and B. 3.18.1 6.0 PAM 77 oF El

2202-2.1 Revision 0 01/29/76 Fill the OTSG's to the proper shutdown level. 3.18.2 "STOP" ;he Turbine Driven Emergency Feedpump and "CLOSE" EF-3.18.3 I -VilA and B or EF-V32A and B. e e 1 .~ 7.0 PA6E 8b 0F El

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ML19294C157

Contents

Text

SUMMARY

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Background

REFERENCE:

10.1 INTRODUCTION

2.2.1 Verify

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Navigation menu

ML19294C157

Text

SUMMARY

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REFERENCE:

10.1 INTRODUCTION

2.2.1 Verify

:

a --. / -__. _; :. _- c--J. _et: =r_-_{:.

Navigation menu

Search