ML19206A441

In Response to 781220 Telcon,Forwards Addl Safety Evaluation Info.Analyzes Transient Resulting from Operator Not Controlling Steam Generator Level

ML19206A441
Person / Time
Site:	Crane, Davis Besse
Issue date:	12/22/1978
From:	Roe L TOLEDO EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
Shared Package
ML19206A398	List: ML19206A396 ML19206A413 ML19206A426 ML19206A441
References
TASK-TF, TASK-TMR TAC-11050, NUDOCS 7904200034
Download: ML19206A441 (19)

Text

{{#Wiki_filter:~~ \\ \\ g -) m TOLEDO M '= a Ef210 G DJ hl - Dock-: No. 50-346 -r LowELL E. ROE %ca Pets.eens License No. MPF-3 / L WS3 259 5242 Serial No. 475 Dece=b er 22, 1978 Director of Nuclee.: Reactor Regulation Attention: Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors U.S. Nuclear Regulatory Co-4 ssion Washington, D.C. 20555

Dear Mr. Reid:

In response to the December 20, 1978, telephone conversation between your Mr. Cuy Vissing and our Mr. E. C. Novak, and the Deccabar 20, 1978 telephone conversation between NRC Region III personnel (G. Fiorelli, R. Knop, T. Ta:abling and J. Streeter) and our Mr E. C. Novak, attached is an additional safety evaluation supporting continued operation of Davis-Besse Nuclear Power Station Unit 1. This additional safety evalu-atica supple =ents the analysis we provided to.you by our letter dated December 11, 1978, Serial No. 471. The atte.ched safety ev11uation analyzes the transient resulting from the operator not controlling steam generator level at 35 inches in accordance with current operating proce-durcs. Yours very truly, W l (/ LER:CRD Enclosure * -a b3 e/7 ggo42,000k ~ w. m..sb V h

Decket No. 50-3'6 License No. !??-3 Scrial No. 475 Decer.her 22, 1976 Additional Safety Evaluation of Transient Resulting from Inability of Operator to Control Stea= Generator Level at 35 Inches ,'q I. INTRODUCTION Tw The Davis-Basse Unit 1 Steam and Feedwater Line Rupture Control System NE (SFRCS) design objectives are to prevent the release of high energy steam, to automatically start auxiliary feedvater (AFW), and to provide adequate AFW, via essential. steam generator level control, to remove decay heat during anticipated and design basis events when AFW is required. conditions for Table I correlates the station variables and accident which AFV actuation is required. For all actuation signals, the STRCS ) initiates and controls AFW addition autoaatically to maintain a 120" level I (96" indicated on the startup range instrumentation) in the steam generator i' The recent natural circulation test at Davis-Besse 1 (IP800.04) demonstrate 1 that a 35-inch (indicated) steam generator level of AFW provides adequate nataral circulation for decay heat removal. The auto essential SG 1evel control setpoint of 120-inches (96-inch-indiuat is thus in excess of minimum SG level requirements. Operating procedures requiring manual control cf steam generator level at 35-inches on the startup range level indicators following non-LOCA events were developed and used at Davis-3 esse Unit 1 pending installation of L permanent design changes to the SFRCC Margin in maintenance cf indict.ted pressurizer level and assurance of adequate natural circulation capability vill exist through operator intervention during conditions where AEW is required. . Inability of the operator to comply with the present operating procedura-will possibly result in a comentary loss of pressurizer level and/c J.<' indication under certain conditinns, but will not produce consequi are non-reversible or detrimental to safe operation of Davis-Besse II. DISCUSSION The following section is divided into three segments: Relationship with Events Presented in the Davis-Besse Unit 1 FSAR, Loss of Of f site Power, and Loss of Feedwater. i l A. Relationship with events presented in the FSAR Addition of auxiliary feedwater at rates considerably greater than the decay heat generation rate will result in overcooling of the reactor coolant, contraction and a reduction of pressurizer level. This sequ- ,.O _q. 0 of events is typical or several transients presented in the isAs unte. U have been subnitted to the NRC and approved as a part of the licensin. l

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Docket No. 50-346 License No. N?F-3 Serial No. 475 Dece=ber 22, 1978 Page Two t operator interactions. From a practical viewpoint each single dis-coverable possible transient cannot be analy:ed and presented as a j part of the FSAR analysis, but a broad variety of transients have been selected. This specific transient fits within that broad category-Each of the FSAR transients has been de=onstrated to produce acceptable results. I I Overcooling transients resulting from a variety of causes are described in Section 15.2.10 " Excessive Heat Removal due to Feedwater l' lfunctions a This section describes a transient resulting from excessive =ain feed-vater. addition, which is stailar to the specific transient of increased level addition by auxiliary fcedwater. Further infor=ation is presented in response to questions 15.2.15 and 15.2.16. The stea: line break (see FSAR sections 15.4.4, ' 15.4. 8, 15.4.1) is the i most seve-e overcooling transient,in that the reactor coolant syste= is decreased 50 F in average core te=perature over a 30 second time pe '.od. This is cenpared with the cooldown in question, which takes a =uch longc time to achieve a clailar te=perature drop.and system conditions. During the steam line break, RC system pressure is reduced fro 2200 psi to about 900 psi as system temperature is driven toward equilibriu: witF the unaffected (pressurize'd) stec= generator attaining saturation te=per ture of about 5300F. The pressurizer is near e=pty at about 20 seconds and thereafter loses its influente on the system, thus permitting the upper elevations of the reactor coolant loop to approach saturation as cooldown continuas toward 330 F.. Eigh pressure injection (HPI) pu=ps a: actuated on low RC pressure such that pressurizer level vill be restorac As shown in Figures 15.4.4-1 and 15.4.4-2 of the Davis-Besse Unit 1 FSA? the rapid cooldown of RCS after reactor crip is limited by the pressure maintained in the pressuri:cd steam generator in much the sa=e f ashion as anticipated for events such as the event of concern. As the RCS approaches saturation, core cooling is not i=peded. Mini =u= DN3Ryl. 3 occurs just before reactor trip and subsequencly increases with substan: margin throughout the re=ainder of the cooldown. The close relationship of the an: ciliary feedwater level increase as an overcooling trans'ient with these similar overcooling transients allows us to drce the conclusica that no unreviewed safety question exists. To show a co=parison to the detailed analyses reported in the FSAR, we have perforced conservative bounding analyses of two representative case. E. Loss of Feedwater and Loss of Offsite Power We have analyzed two transients resulting from auxiliary feedwater addi: and establish =ent of SC level above the operating procedure 35" limit. The two transients exa=ined are a loss of offsite power (reactor coolan-pumps stop, makeup stops, main feedwater stops) and a loss of f eedwater M ',- (reactor coolant pu=ps centinue,.::akeup continues).- me-

Docket No. 50-346 License No. NPF-3 Serial No. 475 Dece:ber 22, 1978 Page Three r I Of these two transients the loss of feedwater results in the greater volumetric coolant contraction, because the forced coolant flow (RC Pu=ps operating) causes a faster rate of heat rejection to the steam generator. 1. Loss of offsite Pouer Preliminarv calce'ntions for a reactor trip following a loss of offsite power show that the pressurizer loses indication but does g not e=pty. The assu=ptions used to derive this result included full runout auxiliary feedvater flow 002400 gpm) resulting in a fill time to 120" of about 4 minutes. No net mars change to the .4w primary coolant (no makeup, ne letdown) was considered, even thoug I the makeup controls would respond to decreasing pressurizer level by increasing the net input to above 200.gpa. At the termination of the transient the pressurizer level is slightly above the outle* into the surge line. Reactor coolant. pressure reaches about 1600 and high pressure injection may be automatically initiated. Altbaugh the net makeup was not coasidered, it would in fact cause the pressurizer to refill to the normal level. At the same time c a=pression of the steam would cause a partial repressurization of tn2 system ensuring that the coolant remains subcooled. This tran presents no safety concerns. 2. Less of Feedwater ~ This transient has a greater reactor coolant contraction then the loss of offsite power case, resulting in e.mptying of the pressuri: Consequently it vill be described in greater detail. A brief su==ary of the events is: Reactor t. rip Time = 0 e Makeup control valve opens vide aloitting i = 0+ e full makeup to reactor coolant system e AFW initiated Time lhf 40 sec Pressurizer e=pties; RC system pressure e slightly greater than 1800 psi Time lbf2 min HPI initiated by SFAS; makeup isolated Time l3l2+ min e Steam generator level = 10 ft; voids e exist in reactor coolant Tite ((j4 min t,q -7gp3 HPI inflow replaces volume occupied by e voids; pressurizer level begins to be restored Timefy7-Snin The major concerns that evolve frca this transient ~cre the y,fAc nnd rhn annrnach to DSB. Roth a<--,-:.<- .ec .sn e.,,- NMM E ' 1" 47TTMFfr"fdMMME****,h gb4 *J+ fjME?p3biQ'}MJ,%p-g-me.c VG m' y c.

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w. cc:',;c: No. 50-3b ~ -- ~ I -.. License No. S?F-3 f Serir.1 So. 475 ucce=ber 22, 197S -[ Page Four -Y Steam voids vill not collect in reactor coolant piping and no flow 1 blockage vill c ecur because of dispersal and =ixing by the forced D:,*3 acceptance criterion ' licit vill be =et because the power i-. flow. M ou:put of the core is at the decay heat level and all reactor pumps h$. are operating, =aintaining core hea re= oval. We conclude that no safe:y proble= exis:s. %= yt

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e' TA3LE 1: dr. (SFRCS) ACTUATION PAR.AMETERS Acciden: Actuation Parz=eter Station Variables Secocint I2 Stea= Line Break 1. Low Stez= Line 'd 591.6 psig j i Feedwater Line Brea Pressure I i[ 2. h w SC Level 4 17 inches Loss of F/W 1' > 197.6 psi FWLB, LOMTW ~( 3. SG Pressure Minus Main Feedwater Line Pressure 3 Loss of Off-Site Pc 4. Loss of All RC Pu=ps !;0TES: 1. Vnen actuated, SF?.CS clases both =ain s' ana isolation valves, closes bo:1- =ain W control and stop valves, initiates AW and controls AL' to =ainta: a 120 inch level in the SGs. Align =ent of A W to a pressurized SG is provided for stea= and feedwater 2. line breaks. 3. AW initiation but stea= and feedvater line isolation does no: occur. 8 e 5

Page Five t III. Ecundin: Analysis of Loss of F'cedwater Event Vith Failure of Osera:or to Control Feedwater Level at 35" Introductien: s The following bounding analysis conservatively predicts the events occurring within the pr4->ry reactor coolant syste= and reactor ( following a loss of =ain-feedwater fro 100% power for the Davis-Besse Unit 1. Auxiliary feedwater control has been assu=ed at 10 feet within both stea: generators. Results: [Because of the conservative, bounding, nature of this calculation, ~ the overcooling of the primary syste= due to auxiliary feedwater injection causes a con:: action of ecolant voluna suf ficient to g (create stea= wi:hin the pri=ary system. The stea= is shown to be unifer=ly distributed virhin the RCS and the void frac: ion is 4%. The reactor coolant pe=ps =aintain full capability. The DN3 ratio is shown to exceed 2.0 and no return to criticality potential ~ exists. Thus, durin; the course of the incident, no core problems 5 -develop. Further, following the ti=e of =axi=== contraction, the syste: recovers to full pressure, pressur'izer function is regained and the reactor coolant returns to a subcooled water configuration withou: operator action. E Analysis: The following assu=ptions have been =ade to assure the bounding nature of the resul:s: h Reactor Power: 100% until boiling stops in the stea= generators; 0% af ter that tina. This assunption is conservative as core heat would co=pensate for the cooling caused by the auxiliary feedwater. Initial Coolant Inventories Water: RCS = 11290 ft Pressurizer = 864 ft These assunprions ar,e no=inal operating values. Initial Te=peratures: The whole syste= is taken to be at T = 5S2 F. average This assu=ption is a reasonable average. g, c, g,, Initial Syste= Mass: % 500,000 lb= The n2ss is figured frc= the :c=perature and volu=es above. e

Page Six Makeup Syste=: No credi: is taken for additional =akeup flow which will occur as the pressurizer loses level. (In all likelihood, the =2keup syste= vill contribute approxi=ately 100 f # ex:ra liquid volume). ~ Local Power (kv/f t): 18.4 kv/ft This value is taken as the =axi=u= allowed by Technical Specifica-tions. Secondary Side Volu=e At 10 Foo: Level 3 711 ft per generator, actual volume. Auxiliary Feedvater Flow: 166.5 ft /=in. per generator actual value. ' Auxiliary Feedvater Enthalpy: 8 Btu /lbs lower bound for =aximum cooling. k'ith the initiating event, loss of =ain feedvater, the reactor coolan systes pressute vill sta rt to rise. Reactor trip will occur on high RCS pressure. Following tr.p, the RCS pressure will fall because ' core power has been reduced and be.iling of residual =ain feedvater or auxiliary feedvarer is occurring in the steam generators. These events are.1=ost identical to those which occur in a main feed line break and are analyzed in detail in Sectior 15.2.8 of the FSAR. In short order, the system vill return to its initial configuration but, ~.c4f because the auxiliary feedvater heat absorption rate exceeds the decay heat generation ra te, the RCS continues to depressurize. During this phase, residual =ain feedvater and inj ected auxiliary f eedvat.cr will be boiled and vented through the steam generator safety relief ralves. The T

1. pri=ary syste: averagt ta=perature vill fall to the saturatian te=perature) of_ vater at,the safety valve set pressure., At this clne, plicary and ~ ~

secondary conditions are expected to be approxi=ately as follows: w. Pri=ary Secendary Pressure 1800 psia 980 psia Tenperature 542 F 542 F Ibss 503344 lb= 0 lb= Liquid Volune in Press. 400 f t N.A. Tine into Transient .s 2 min. _N_2 min. - - d m --x mm_,3

Page Seven It is conservative to assume co=plete boiling of the secondary side vater and co=plete equilibrium between pri=ary and secondary sides, as these assu=ptions lead to the maxi =u= follow on injection of auxiliary ,feeduater and therefore, =axi=u contraction. RCS pressure is held up by the stea bubble in the pressurizer. The ti=e has been esti=sted by calculating the necessary energy 1 css by the primary syste fro: i:s.nitial conditions, the = ass of auxiliary feedwater required to re=ov ' his energy and then dividing by the auxiliary feedwater flow rate. (SS6 - 542) 503344 ti=e 2; (1194-S) 3S3 62 3 54 sec. Six seconds was used to esti= ate the initial pressuri=ation portion of Ga the transient. ,p In perfor=ing the re=ainder of' the evaluation 10 feet of cooled (40 F) auxiliary feedwater is placed in each stea= generator and :he _the; pal _eguilibriu: condition calcula cd. Because after a 10 foot level is ob:ained this auxiliary f eedwater flow stops, this condition represents the,=aximu= contraction possible. The state variables.resulting are: Pri=ary Secondary 560 psia Pressure '560 psia Tc perature 478 F 478 F 462 Bru/lb 462 Stu/lb: Enthalpy of Water s 3 Specific Volu=e .020 ft /lb= .020 f t /lb: Fro = the specific volu=e, the pri=ary liquid volu=e can be calculated: 3 Vol = MV = 10052 ft f As 10052 is s= aller than the RCS =inus pressurizer volt =e, the re=aining volu=e must be filled with stea=. 3 3 V = 10426 10552 = 374 ft ; :400 ft s 3 400 ft corresponds to a syste= void fraction of 3.8% 3 4%, and as will be shown 1; er, is of no consequence as far as core heating or systen perfor=ance is concerned. This stea= volu=e is larger than actually expected for two reasons:

1) so=e te=perature difference would always exist between the primary and secondary sys:c=s, and 2) the effect of core decay heat has been ignored.

Both of these would increase the pri=ary side liquid te=perature, thus increasing its volu=e and reducing the steam volu=c. Following this state of maxi =u: contraction, no further hea is recoved fro = the RCS via the secondary side until the RCS rises in te=perature due to decay heating; this will expand the liquid volu=e, compress the {gg _c {sgea: and repressurize,the RCS. As no r. ass can be lost fro the secondary

~. Page Eight sys:c= prior to achieving 930 psia the first reheating stage vill end at and liquid volume of 980 psia, a primary systeg pressure, temperature, 542 F, 10332 ft Subtracting 10426 fro = 10S32 shows that abou: 400 f:3 Pressuri=c function . of fluid has been forced back into the pressurizer. vould then be restored (if not directly, then, by either the makeup or EPI system), the RCS subcooled and the transient ended. Several ques: ions. exist about the transient: I. How is the 400 ft dispersed within the pri=ary system and can that j volu=e collec: in one locatica? From the auxiliary feedwater flou y rate, over 4 =inutes age required to fill the genera: ors. As thg pressurizer has 400 ft in it at 950 psia and the RCS has 400 ft in it at =aximu= contraction, approxi=stely 2 minutes are used to 9Fs ejecc steam from the pressurizer to the RCS. Because this steam y vill be superheated when it enters the RCS.it vill first desuperheat and then condense at a rate governed by its expanding pressure cc= pared to the contraction of the liquid coolant. In the time of 2 minutes the reactor coolant vill have tiada about 8 complete circles of the pri=ary syste= and the stea= can be considered well mixed. As the flow velocity in the RCS will remain nor=al, about 25 ft/sec, steam water separation vill tend not to occur. Some limited stes: accumulation cay occur in the upper head of the reactor vessel as in that specific location of the RCS, velocity is I lov. II. How well vill the pu=ps work? Experiments perferned on stez= carry over capability show that for void fractions up to 10% no 1 css of pump capability is observed. This is dccumented in Figure 5-47 of BAW-10104, "EEW's ECCS Evaluation Leport With Specific Application to 177 FA Class Plants With Lower Loop Arrangement." Actually pump capability increases for the first 5% of void introduced into the sys:cm. III. Will any return to power be encountered because of the lou RCS te=perature? A return to power can occur for a non-borated core at 490F. This te=perature' includes the assumption of the most reactive rod stuck out of the core; if that rod were taken as inserred*the critical temperature would fall to at or below 400F. Although no credit was taken for HPI in calculating the RC steam veluse below 1600 psia, the HPI will be injecting borated water and, therefore, preventing any return to power condition. If the pri.cary systen were to stabilize at 1600 psia and thus prevent the HPI from,roviding baron the RCS temperature would be at least 511F and, there re, no return to power would be expected. IV. kill DN3 be encountered in the ccre? The maximum contraction condition is again: P = 560 psia T = 478F ij -{[h a = 4%, .f e

Page Nine ? and occurs at l' east 5 minutes af ter paver shutdown (trip occurs very early within 10 seconds of =ain feedvater loss). At this

i=e, the decay hea: -?te is less than 3.2% using ANS + 207. (the LOCA evalua: ion cur.

As low pressure and high void and high l power are conservai-. sounds a DN3 evaluation was perfor=ed at: P = 500 psia . = corresponding saturated value a = ST. power = 10% ) W = full volu=etric flow. { ,The resultan: DN3R was >15 in the ho:tes: channel with taxi =u= design conditions assu=ed and well within acceptable values. V. Will any. stes= re=ain trapped in the pri=ary syste=? Some may be trapped for a shor: period of :i=a in the upper head of the reactor vessel but this vill be of no consequence and will eventually be condensed by ther=al conduction through the interf acing va:er. Conclusion The taxi=== contraction of the RCS vater has been e.alculafed taking no credit for =itiga:ing syste=s (=akeup fles, EPI) and no credit for decay heating. No adverse consequences of the transient have been shown and, therefore, this tra=sien: poses no concerns to the safe operatica of the plant. a IV. CONCLUSIONS For SFRCS actuation and fill of the stea= generators tc the auto-essential level control point of 120" without operator action: No unreviewed safety question _ exists e The loss of'offsite power transient will not cause the pressurirer e to drain although a loss of pressurizer indicated level vill occur. .Tne 1 css of feedvater transient =ay result in pressuri=e e=ptying, however acceptance criteria for DNB will be =et. Stea= bubbles which exist in the reactor coolant for a short time vill bc collapsed by E?I inj ection. Pressurizer refilling by E?I will occur. No return to power will result in the l'ong ter=. e r t},} " (. ' L ]) l ' Pm i

~ Babcock &Wilcox Power Generat.:n Gre. ;> E_'"' k P.O. Box 1260, Lyn:r.::er;;. Va. 24 Tele : hone: (304) 354 5111 June 12, 1978 SOM #382 620-0 123h3 T3.3 ~ SIP #1k/2S9 I 4 Pr. T. D. M= ray, Station Superi= ten lent Davis-3 esse Nuclea:- Pcver Station 5501 Herth State Route #2 Osh Harbor, Ohio h3kk9 I CEDCS Trip 3reaker Maintenance i Subject :

Dear. Tern :

In the past, sc=e of our plants have experienced proble=s.vith CEDCS Trip Breakers. The prcble._s have been traced to lack of preventive raintenance. 3W suggests that a planned, carefully executed, =sintenance pro, a= be e established using the raintenance prog-a= cutlined in the Diacond Power ~ 22., CEDC Syste= Vendor Fa=ual. Particular attention should be directed to M-proper cycling, cleaning, and lubricatics of the breakers. 3W further recc=nends that this progra= be scheduled at a =in"~ Nequency of every re:Neling cycle and =cre frequently for plants du-ing startup, vben g the equip =ent is subject to adverse e=viren= ental conditions. Our concern is that if proper =aintenance is not acco=plished, adf.itional failures will occur resulting in an NEC de=and for diverse qualified trip bre akers. Also, we need to prevent =D failures ve can to reduce the nu=- ber of lost capacity deys. If ve can be of further assistance, please advise. Yours t uly, %u% shw fc t_ F. R. Faist FPJ:IDG:nif Site Operations Manager cc: W. E. Spacgler R. C. Luken R. L. Pitt=an D. A. L,e t ' J* S* G# "'* TECo R. E. Blanchong, TECo g E. C. Novak, TEC J. C. Buch, TEco g3 L. R. Do=eck, TECo V* J. G. Eva=s, TECo B. R. Beyer, TECo r t; -(3{y J. 7. Lenard:en, TECo

t - y f.. .f.- c .~ y s ' z Babcock &Wilcox %er cemmu= cecup b P.O. Box 1250. Lyn:ncurg. Va. 2 Telephone: (804) 3S '-5111 August 9, 1978 SOM !h03 620-001h /12322 T3.3.1 SIP #1L/295 l L l I Mr. T. D. Murray, Station Superintendent Davis-3 esse Nuclear Pover Stati-on C 5501 North State Route #2 Oah Ecrbor, Ohio h3hh9 Sub j ect: SM!1D Rapid Cooldovn Transient Dec.~ Terr;; On March 20, 1978, Eancho Seco experienced a devere cer-a transient initic; 17 the loss of electrical power to a substential portion of the Neo-Nuclear Instr--antation OCE). Tne loss cf power directly caused the loss of Contro- @g Roc indice. tion of =any plant pa eneters, the less of input of these para-to the plant ec=puter, cad erroneous input signals (=idrange, zero, ceter: or othe: vise incorrect} to the Integrated Control Systen (ICS). The plant rupense ves L, the usud transient in that the ICS responded to the erroneous input signals rather than actual plant conditions, and resulte in a Reactor Protection Syste= (EPS) trip on high pressure. Subsequent to the Beactor Trip, the errenecus signcle te the ICS contributed to the rapid cooli cs of the 30S. Plant operators hel extrene difficulty in deter =ining thu true st tus of se=e of the plant parc=eters and in centrolling the plar.t o 'Uccanse of the erroneous indications in the Centrol Roc =. An investige. tion of the events folloving this loss of power points out a nee for e. : lose lock at operator training and e=ergency operating procedures for for po-tion thereof). The following reew__ endations av len of ICE power are cade to essist your steff in a reviev of training and procedures to ess-proper operator action. for events of this nature. 1. Operators should be trained to recognize a loss of po.ier to all or a rajority of their ICE (e.g. indicators fail to nid-range, autenatic or =s..aal transfer to elte:. ate inst-unent strings brings no respctse). The loss of power is enphariced here rather than the failure of any one instrunent or control signal which are adequately covered in current si=ulator training courses. Lth.E # T 3 h

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620-oolk e "= -B. ab c kEWilcox August 9, 1978 7 Page 2 / .s-2. Given that the enerator can deter =ine that electrical power has been ~ lost to all or pen of the ME, he should kncv the location of the power supply breakers, and have a procedure available to ou.'.chiv re-gain pover. 3. If the f. ult ca. net be cleared (i.e. the breakers to the power supplies reopen), the operator should have a' list of' alt'ernate instrunentation available to hi=, and he should be thoroughly trained in its use. Ex- .f j [ a=ples are: a. ESTAS panels [ b. ??S penels I.CI (Essential Centrols and Instru=entation) g SRCI (Safety Felated Controls and Instru:entation} c. d. pl e. Recote shutdr: panels ~ f. Local gages Plant ec=puter ) g., ,s h. Recognizing that no procedure can cover a" possible co=binations of EE fe.ilu-es, the operator's respense should be keyed to certain va-iables. If the operater realizes that he has an instru=entation proble= (as op-posed to a LOCA or stec= line break, for exe=ple), he can 14 it the transient by controlling a few critical variables: Pressuriser level (via EFI or nor=al Makeup Pu=psl l e.. l b. RCS pressure (via Pressurf:er heaters, spray, E/M relief valves, etc. ) Steen Generator level (via feed flov, feedvater valves, etc.) c. } d. Stea= Generator pressure (via turbine bypass syste=) The pressurizer level and RCS pressure assure that the Reactor Coolant Syste= is filled; the Stes= Generator level and pressure assu-e adequate decay heat re= oval. l Attach =ents 1 and 2 are provided to give a brief description of the events l folleving this loss of MTI power at Rancho Seco. As can be seen by this transient, pro =pt precise operator action and the ability to recegnize a loss of miI pover are critical factors in li=iting the severity of a trans-ient such as this. If,you have any questier,s or cu _.ents, pleese advise. Yours truly, w-Ivan D. Green Site Operations Manager IDG: TIS:nir encl. __ c r ty t t '^ '"I ec:,See attached sheet. -mm_

ATTACIIMEtiT 1 .M. .20, 1978 }, SEcurNCE OF EVENTS - SMUD 04 :25 to 05 :34 - MARCH (Revision 1, 5/25/78) t t EVENT (2 & 7 - Lost NNI power supply cabinets 5,6, 5:35 i - This caused a loss of valid signals j ~ BTU limits ran back to the ICS. E resulting in a partial loss feedwater, (actual Rx power was 72%). F of feedwater turbine bypass - Probable-ocening of "B" 6alves to the condenser (timing uncertain). [ turbine trip t Reactor trip on hich cressure, 25:44 on interlock. - Pressurizer code relief setting was known E' rl The ~ to be' low (approximately 2225 psig). [ electromatic relief was isolated due p to previous leakage problems. The data ~ indicates, primary pressure went =2400 i psig,=>.. code relief valye. lifted. - ICS closes main control and start-up I feed valves and drive main feed pumps to minimum speed following trip. ~ l - Decay heat and RC pumps energy' removal accomplished through generators by inventory t-boil off and the addition or main feedwater. - Pressurizer code relief-valve reseats at

26:1-5 approximately 2100 psig.

"B". - Op'erator starts HPI pump "B". .:28:23 - Operator stops HPI pump pressure reaches 435 psig set-point - OTSG "B" i:30 of Steam Line Failure Logic. - OTSG "B" goes dry. . dc.. _n, O

- Operator incrcases spec m '. )f4 "A" OTSG. This starts RCS on pressure , and temperature decrease. 'p- .v - RC pressure =1900 psi 34:25 ,bFAS actuation at 1600 psig

37:16 This starts HPI, LPI and initit.es

' emergency feed. The emergency FW pum.o r is started and the bypass emergency IN valves are ocened to full ocen cosition. The system mikes no autcmatic attempt { to, control steam ge.nerator water level. It starts - RC pressure at 1475 psig. ,i40 'to ^ recover frcm this point due to HPI. T = 5280F. ~ ave 543:56 "A" EPI pump secured. - LPI secured. 15:09 "A" HPI inf._ated. From this point on, the

49:54 operator started and stopped EPI pumps as necessary to maintain pressurizer level.

j g - Steam Line Fai' lure Logic closes ICS-control .-50 start-up feed valves to each OTSG when the corresponding OTSG pressure falls below 435 psig. I:51: 25 - Secured RCP-D (T =435 F) ve This reduced ?RCP's tb three - OTSG "A" water level - 599.7"

57: 27 of tubes are not

~ Speculate that =2 f t. flooded (at top) due to steam line ~ arrangement. - Hourly computer log prin,t-out f:00:00 ^ Stcam tc=p. 380 F (OTSG _"B") 0 Sl', -e f 1 Steam pressure 171 psig (OTSG "B") o Assun.ing T =T => T = 380 F ~ 'N M E N 5h W EST-lP5 h h $ @y gfgggg ^

..u M -T,* . &y. 3 ..s % 3:~47 OTSG "B" level - 599.1" .3

• 7

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' Powcr restored"to NNI cabinch.s 5,6,c7 ~ T = 285 e ave 0 RCS Pressure =2000.usi9 Both OTSG full level ranges pegged high Operator begins to reduce RC pressure using pressurizer spray. ICS closes turbine bypass valves to condense Operator stops emergency W flow. Operator stops main W pumps. e e e 0

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