Text

Forwards Rept on Browns Ferry 3 Partial Failure to Scram on 800628. Executive Summary Encl
	ML20027A622
Person / Time
Site:	05000000, Browns Ferry
Issue date:	08/01/1980
From:	Michelson C; NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:	Harold Denton; Office of Nuclear Reactor Regulation
Shared Package
ML20027A608	List: ML20027A621; ML20027A622; ML20027A664; ML20027A670; ML20027A693; ML20027A694; ML20027A696; ML20027A722; ML20027A761; ML20027A788;
References
	FOIA-82-393 NUDOCS 8010210736
	Download: ML20027A622 (7)
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ik .....' / AUG 1 Ic80 4 3 L I n 4 MEMORANDUM FOR: Harold R. Denton, Director j Office of Nuclear Reactor Regulation nW FROM: Carlyle Michelson, Director P Office for Analysis and Evaluation of Operational Data T P:

SUBJECT:

AE00 REPORT ON THE BROWNS FERRY 3 PARTIAL. K FAILURE TO SCRAM ON JUNE 28, 1980 b p! AE00 has.recently completed its analysis of the June 28, 1980 Browns Ferry 3 F: event involving a partial failure to scram. The. principal findings, conclu-t siens, and recommendations from this study are highlighted in the Executive Sumary (Enclosure 1) and the full report is forwarded (Enclosure 2) for your y], . information and appropriate acticn.. ,a ~~ j' As discussed specifically in the report, we believe that: ('l) there are several ?- .p-credible ways which water can accumulate undetected in the scram discharge volume Fi ~. 't.-' ' (, providing a potential for unreliable scram capability; and (2) there are scra events that can result in an unisolatable reactor coolant blowdown outside of. ? ]pprimary containment i_f the sinole isolation valve should fall. We bilieve that reasonably prompt corrective action is required to rectify these design deficien-cies. Specific recommendations in this regard have been identified in the report. Because of the risk presented by these deficiencies, we further believe that t appropriate systems modifications should be completed before the end of 1981 on all applicable plants. Should your staff have informal questions or wish ' clarification en the report, 4 pleasa feel free to centact the authors directly. If we can provide additional assistance, please contact me. S 4 P 9 Carlyle Michelson, Director Office for Analysis anc Evaluation of Operational Data o +

Enclosures:

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} ki s v Prepared by,: Stuart Rubin, lead George Lanik A NOTE: This report documents results of studies completed to j date by the Office for Analysis and Evaluation of Opera-i tional Data with regard to a particular operating event. F-The findings and recommendations contained in this report are provided in suppor.t of other ongoing NRC activities / concerning this event. Since the studies are ongoing, the report is not necessarily final, and the findings and recor:::endations do not represent the position or 5 requirements of the responsible program office of the Nuclear Regulatory Ccgission. W f golo%S75 n g * ,,, w. \\ p m..m==mp ,e

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. g u. +, n s: m.p a u. m y g g...;.- a 3 g g gy3 g y ,p;h{ h (', I* if 3 PREFACE M [y The findings, recommendations, and conclusions contained in this report it are, for reasons of timeliness, based on infor=ati'n gathered through g o h informal channels between the Tennessee Valley Authority, the General og Electric Company, and the US NRC Headquarters and Re'gional offices. To [ the extent possible, the information used in the report has been verified (fS by cross checking with other sources. The findings contained in this y report, including the underlying causes of the partial scram failure which k occurred at Browns Ferry Unit No. 3 (BF-3) on June 28, 1980, relate most directly to the Browns Ferry reactor. However, similarities among boiling 34 water reactor facilities leads us to believe that the findings and rec-I.d ommendations may be broadly and generically applied to most if not all wif operating BWRs. To this end, we reco=end that a plant-by-plant review, N not possible in this investigation, be undertaken by others, to assess the t applicability of these findings and recomendations to other BWRs and to }; provide analysis and evaluation of plant-unique design problems not un-y; covered in this investigation. Additionally, the scope of our investi-O gation and recommendations was intentionally limited so as to address'enly a$ the specific, direct and underlying causes of the partial scram failure m g.] at BF-3. 'We have not, therefore, taken the broader view, as could be g taken by those most directly involved in the ATWS issue. We do believe, [ however, that some of the information presented in the report can be useful to those involved in this important generic concern. Finally, this in-9 j vestigation was not able to pinpoint a single precise root cause(s) which [.4 led to the BF-3 partial scram failure event, beyond to say it was caused h by water in the scram discharge volume. However, we believe that, in [ totality, the various possible cause mechanisms discussed in this report b include the actual, albeit, indeterminable root cause(s) of the event. a h $f As a footnote, the writers wish to acknowledge the invaluable and timely x fg information provided by the BF-3. resident inspectors, James Chase and s Robert Sullivan, without whose cooperation, timely issuance of this report }f would not have been possible. t e 1 I[ i 3 m. .,._,=

F q q q,R M N m ..+ c - .:.sj; I l:.:A %S Od@M g 4 (... ..d, 1 Q., r . w. k L.- M TABLE OF CONTENTS Page l5 PREFACE..........................................'....................... i h 1 INTRODUCTION........................................................ I k 2 EVENT SEQUENCE 3 J 3 DESIGN AND OPERATION OF THE BROWNS FERRY UNIT 3 SCRAM SYSTEM........ 5 ?. 4 CAUSES INVESTIGATED................................................. 11 ) 5 EVENT SEQUENCE ANALYSIS............................................. 14 / 6 SCRAM DISCHARGE VOLUME / SCRAM INSTRUMENT VOLUME INSPECTIONS m ps AND TESTS.......................................................... 19 l 7 PREVIOUS BWR EXPERIENCE OF. FAILURE TO FULLY INSERT................... 23 ( 8 FINDINGS...,........................................................ '24' !i 9 RECOMMEtiDATIONS..............*................,.......;.............. 35 10 CONCLUSIONS..................................~....................... 40 KY M LIST OF FIGURES 4 a;i Ficure m 2-1 Control Rod Positions Before First Manual Scram................... 43 - 1

a 2-2 Control Rod Positions After First Scram...........................

44: D)J 2-3 Centrol Rod Positions After Second Scram.......................... 45 T, 2-4 Control Rod Position After Third Scrai. 46 f 3-1 Control Rod Drive................................................. 47 3-2 Scram Electrical Diagram.......................................... 48 49 3-3 Centrol Rod Scram Group Assignment................................ a 1 3-4 S cr am V a l ve A rr an g eme n t........................................... 50 h 3-5 Scram Volume Orain Arrangement.................................... 51 ~ 6 d. LIST OF TA3LES 9 d Table a W 2-1 Event Sequence Recorder Printout.................................. 41 y M 5-1 Scram Discharge Volume Orain Time and Total Positions Inserted.... 52 1 IV O m 4 + i ]+ m m'.._~_ _ mm i-_ _. _ __ __ _ _ M ..i -. e-WWE' h d r 2

2 .v, 3.:. a ;u % g v :.; ::; c g.4;v W di; W gyu Q Q:.1.s.Q ..a _a.s '[ (,i (? t* p j /. i 1 INTRODUCTION i j On June 28, 1980, the Browns Ferry 3 reactor experienced a partial failure of the scram system, while shutting down for a scheduled maintenance of the feedwater system. The reactor had been brought down to approximately 35% n o] power by reducing recirculation flow and by manual insertion of control rods. The subject event occurred when the control room operator initiated a manual scram to make the reactor subcritical which was the next step in the normal shutdown evolution. After manual scram actuation, the control rods on the West side of the core were observed to be fully inserted. How-ever, the control rods on the East side of the core did not fully insert. j Most of the East side rods came to rest in notch positions ranging between 1 ~ 6 00 and 46 after all East side rod motion had ended. Three additional

crans ud about 14 minutes were required to achieve full insertion of the partially withdrawn East side control rods. After all rods were com-a pletely inserted, the operators resumed nomal shutdown operations.

l b On July 2,1980, a team of NRC Headquarters representatives from IE, NRR, M and AE00 went to the Browns Ferry site to gather det. ailed information on yM the event, the scram system design and operation, and the results of scram vf) system tests which already had been perfomed by TVA personnel. kith this I.], initial direct contact at the plant, an independent investigation of the y event cause and the recommended corrective actions was begun by the Office [ for Analysis and Evaluation of Operational Data (AE00). Over the next d several days, additional equipment testing was performed on the SF-3 scram j system. Testing and analysis was also being conducted during this time by ,j General Electric in San Jose, California to support TVA activities at the plant. During this period, AE00 continued to obtain, analyze, and evaluate information as it evolved frca these and other sources to continue its investigation. A 1 ihe purpose of this report is to. provide the analysis, evaluation, findings, 'h fi J i-0. 3 4' a t ~ ~ ~ ~ 2a'n x.,,.~,,,...;~~Q T Q ~~~^~* Q } ~~'q';;,

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ff by the AE00, US NRC. Section 2 of the report contains an event sequence.

Section 3 provides a description of the design and operation ef the BF-3 a scram system. Section 4 discusses the possible causes of the event which '( c,) were investigated and the conclusions in each case. *Section 5 provides g an event sequence analysis. Section 6 provides a sur: nary of the tests and 7., inspections performed at BF-3 which support the event sequence analysis Q and some of the findings. Previous operating experience and investigation ,[ findinas are contained in Sections 7 and 8, res;;ectively. Specific 3. li recomendations to correct the deficiencies discussed in the findings are A provided in Section 9. 'The conclusions of this inv'estigation are given in s IV, Section 10! J I& 4

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~.# ,ww ' MF1 M7%gggw,:.g... ..a.'..c;4eig33 s x U.'. 2 EVE T SEQUENCE On June 28, 1980, power was being reduced by the control room operator at the Browns Ferry Unit 3 nuclear reactor in preparation for a scheduled shut-down for feedwater system maintenance. By 0131. hours, the reactor power had been brought to 390 MWe via decreased recirculation flow a~nd manual con-trol rod insertion. The operating personnel then initiated a manual reactor scram to complete insertion of the remaining control rods' (which were at q the positions shown in Figure 2-1 at the time) and thereby bring the reactor .j to a subcritical state, k. Ir.ediately after depressing the manual scram buttons, the operators placed 2 the reacton mode switch in the SHUTOOWN mode. Control room personnel ob- / served that the blue scram li hts for all control rod drive scram inlet and L outlet valves were illuminated, indicating th'at all scram valves were open, u Control rod position indication also showed that all of the rods on the West T side of the core were fully inserted (except for one which had stopped at [ position "02"). However, position indication showed that 75 rods on the East h, side of the core were not inserted fully. The East side control rods came to. j rest at positions ranging from 46 to 00 withdrawn with an average of cbout 23 fq positions ' withdrawn (position 48 corresponds to fully withdrawn). Rod position a ] indications following the first manual scram are shown in Figure 2-2. At 3 this time 18 rods on the East side were fully inserted. As estimated by the Y LPRM readings, power level on the East side of the core following the first J.j scram appeared to be less than two percent. 4 Following scram, the Scram Instrumant Volume began to fill and the Scram In-strument Volume Hi Level Scram (Tevel switches) actuated. This occurred some-what sooner than expected at about 19 seconds. The Hi Level scram condition was subsequently bypassed by the operator (as allowed fa SHUTDdWN mode), to a. i permit reactor protection system reset which occurred 4 minutes and 31 seconds l following the first scram. One minute and 33 seconds later a second manual scram was initiated by the i ~ 1 4 e e 3. s-e-. 't ~~'~*% cr- % 7.r. ' 7gg,q J.h.hyggg ~r:t.w.,. a :.m t- .; s -, -.. - - ~ - _

9 g.:w:. pn-uc: nr ww > anw w.suma.-++m. xur.,6 43&+@%4.:+-.w u@ ? :s; I n :'. a 0-O t operator. The time f'ollowing reset allowed partial drainage of the East and West Scram Discharge Volumes. Rod positions following the second scram are shown in Figure 2-3. After this scram, 33 rods'on the East side fully inserted. The second manual scram was reset after 59 seconds,and the scram S discharge volume was allowed to drain for 53 seconds' at which time a third j manual scram was actuated. Upon completion of this scram, 47 rods on the East side were fully inserted. Rod positions following the third manual N,. scram are shown in Figure 2-4 The third scram was reset after 3 minutes and safj 26 seconds. The scram discharge level bypass skitch was returned to normal fj 2 minutes and 40 seconds.later. This action initiated a fourth, automatic [ scram due to a Scram I?fstrument Volume Hi Level scr'am condition which had fj not cleared, At this time al,1 rods on the East side were fully inserted. 4 A detailed sequence of events as provided by the event sequence recorder is shown in Table 2-1. The total elapsed time between the initial scram and e 'a final insertion of all rods was 14 minutes 2 seconds. At this time the operators continued normal shutdown operations. 3 3 V

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ym..ww wa=ngn+.=~ w.a waaww. i.x. R O O / 3 OESIGN AND GPERATION OF THE BROWNS FERRY UNIT 3 SCRAM SYSTEM m { Mechanical and Hydraulic Desicn of_the Scram System On a GE BWR, such as Browns Ferry Unit No. 3, the Control Rod Drive (CRO) e J j and its associated Hydraulic Control Unit (HCU) provide the me'ans by which, 3 each individual control rod can be rapidly inserted upward into the core during a reactor scram. A simplified drawing of the CR0 mechan' ism is shewn y(g in Figure 3-1. During periods of no rod motion, the collet fingers are spring loaded into a groove en the index tube to hold the drive stationary against the force of w4 gravity. High pressure cooling water is applied below the drive piston and equalized without CRD motion v'ia controlled inleakage past the CR0 seals and h into the reactor. A CR0 temperature probe is' provided internally to monitor ' each CRD to detect CRD heat-up should cooling water flow be interrupted er k should excess leakage of high temperature RCS water flow out through the' drive, drive insert,line and scram cutlet valve. Scram cutlet valve leakage into the g j scram discharge volume on the order of 0.1 cpm wou'1d raise the probe temper- [ ature to the alarm setpoint of about 3Sb0F. .;j At SF-3, water exhausted from the CRDs is routed to either an East or West o a header scram discharge volume. The scram discharge volume (SDV) is sized to G g provida a volume of approximately 3.3 gallens per CR0 (approximately 600 gallons { total). The SDV volume is sized to limit the total amount of hot reactor water y leakage past the seals during a reactor scram (maximum volume requirement) while F{ providing enough free space at atmospheric pressure so that back pressure en the s CRDs does not increase so rapidly as to impede red insertion speed (minimum h volume requirement). In particular, the system design results in a pressure d in the SDV irmediately following full insertien rod motions of'less than 65 psig. Low pressure in the SDV is necessary to assure that technical specification scram ~ h speeds and full-in rod motion are achieved. The volume of water exhausted o y through the scram cutlet valve of a single normal drive for a full stroke is about 0.75 gallens, not including seal leakage and bypass flew. The leakage. au + .~. t . nQ:':" %;~ 4p:7: g,g ._,;-~.-----,-..- y

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(.I h k *.'*g;(, ~ v ),, p E c.- t4 and bypass flow for a single drive can be in excess of 5' gallons per minute. h Normal scram time; frcm full out to 90 percent insertion is less than 3 seconds. C. N Although the SDV is sized for a volume of approximately 3.3 gallons per drive and f the drive stroke (without bypass) is cnly approximately 0.75 gallons, cnly a W sinole reactor scram is normally possible with respect to the scram discharge p E volume. Leakage of reactor water past the seals fills the SDV rapidly as long I as the scram outlet valves are open which would be the case without an RPS ff reset. This leakage occurs even on rods that are fully inserted. The leakage is p an average of 2 gpm to 3 gpm per CRD. Thus, from this sour'ce alone, the 3.3 j gallons per drive of free volume available in the SDV is filled and pressurized h within two m'inutes. Thus, more than one scram would be possible only if the n E operator were able. to reset the scram (closing the scram cutlet valves) well m ('g within this time period. Without an early reset, the SDV would be filled and f the SDV wculd have to be drained to attempt a rescram if rod motion is to be s produced. h [j The East and West SDV headers are each provided with a vent line and vent valve. Each header drains via a separate drain line into a scram instruce'nt volume ~ d (SIV) where level monitoring instruments are' located. The SIV, in' turn, has qF drain piping and a drain valve. y 4 During normal operation, the vent valves of tne East and West SDV headers and h the drain valve of the SIV are open. These valves are kept open to allow the 'j leakage past the scram outlet valves entering the SDV to drain continuously [; into the SIV so that no build-up of water in the SDV occurs which could prevent k a reactor scram. These valves close during centrol rod scram insertion to p g centain and limit the reactor water released thrcugh the scram gutlet valves. .{ During a scram, inflew of water to the SDV normally centinues af ter centrol rod [.y inser,tien is completed due to leakage past the CRD seals. Leakage continues } until the scram is reset or until the SDV prosure ecuilibrates with reactor pressure. F; ~ t 1 = t

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m ,.',.. '} (-) ( y li A pressure difference of at least 550 psi must be applied between the CR0 insert and withdraw lines to drive the rods in during a scram. The' pressure difference applied at the beginning of a scram is provided by the 1500 psia scram accumulator and atmospheric pressure in the empty 50V. As CR0 scram L insertion progresses, pressure losses in the driving fluid due to line losses j reduce the insert line pressure to below reactor coolant system pressure, g At that time, the ball check valve, integral to the CRO, allows reactor coolant [ system water to come in under the piston to complete the scram, before any b significant build-up in scram discharge volume pressure due to filling from leakage and bypass flow. 3 { RpS Electrical Desion h A simplified schematic of the electrical components of the Reactor protection h System (RPS) is shown in Figure 3-2., It is divided into two independent trip [ channels A and S. Each.of the channels can be tripped (de-energized) by eithe'r bk the manual scram relays or the two subchannel relays. The subchannel relays f are de-energized and opened whenever any one of a variety of trip conditions ty exist in the reactor or associated equipment. Tne automatic logic can be j described as. "one-out-of-two taken twice." For purposes of analyssis of the M Browns Ferry event, the automatic trip logic will not be discussed because I :-E this event occurred first with a manual scram. k n With reference to Figure 3-2, both scram solenoid valves A and S must change $h positien to provide a scram. Electrically, this requires a trip of both channel A and channel B. De-energizing the two scram solenoids changes the air flow from the centrol air supply to the vent path. For manual scrams, a separate h scram butten is provided on the control panel for each channel. A manual scram 4 is initiated by depressing both the channel A scram butten and the channel 3 ,3 scram button. Because of the power requirements of 185 3eparate scram solenoid h valves en each channel, each channel is divided electrically into 4 separate scram groups. Centrol rods associated with the HCOs frcm the four groups are f distributed randomly throughout the core as shown in Figure 3-3. (. L< p Q.

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q :.. p g,.. w.m._. y,-..:_..;._.:; n : .z.,g .g._ g .. v... ' ). C. O 7 3 Q .;) Scram Oceration The Reactor Protection System performs its design function by de-energizing the 370 sc:am solenoid air supply valves (2 for each control rod drive HCU), de-energizing the two scram discharge volume (SDV) air supply solenoid valves, h and energizing the four backup scram solenoid valves in the air supply lines a j as shown in Figure 3-4. k) Scram insertion is achieved for each individual control rod by opening the M scram inlet and scram cutlet valves. This applies 1500 psi accumulator pressure p.-3 to the " insert" side of the control red drive piston and vents the " withdraw" N side of the piston to the SDV which is at atmospheric pressure.- u, M p For normal unscrammed conditions, the scram inlet and outlet valves are held b shut by control air pressure ' applied through the energized scram air supply - ( valves (539A and 5398 in Figure 3-4). The SDV vent and SIV drain valves are held h open by air pressure applied through the energized discharge volume air supply $) valves (S37A and S378). The air header which supplies control air to all of N the 372 air. supply valves (370 scram, 2 vent / drain) is pressurized'through de-va * .j energized backup scram valves (535A and 5358, S70A and S708). The SDV vent and SIV drain valves can be operated manually from the control room. g A scram signal de-energizes both air supply valves for each rod, de-energizes I the scram discharge volume air supply valves, and energizes the back-up scram valves, thus venting air pressure frem the scram inlet and outlet valves and Y the SDV and SIV valves. This causes the scram valves to,open and the SDV vent h and SIV drain valves to close. In the event the individual control rod p air supply valves shculd fail to change positicn (i.e., mechanical bind-up, etc.), sy the back-up scram valves which were energized and vented air to depressurize k the air supply header assure opening of the scram valves. Thus, even if an air w@ supply valve failed to shift, that rod would still scram. A check valve is a f provided around the downstream back-up scran valve in the air supply line so } the upstream valve can assist in the, air header venting or assume venting in. + ' case the dcwnstream valve f ails. afh ~ 3-t P T MO F

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.r.v:.<e ; w..,,, 5 mm: - 3.gy._ 3, p g. g g y -;.g;q;g y. 3 3 c.w4..,-4.. : w ..e .e ..",,gh h { t 1, Physical 1.ayout of the Scram System Hydraulic Comoonents at Browns Ferry Unit 3 At Browns Ferry Unit No. 3, the HCUs-for all of the CR0s are physically arranged ) in rows on the " East" and " West" sides of the reactor vessel, outside the drywell p and inside the reactor building. The CRDs on the West side of t'he core are j controlled by the West side HCUs and the CRDs on the East side of the core are ,ll controlled by the East side HCOs. Drives along the interface cen'terline, T between the East and West sides of the core, are alternately routed to the East 4 and West headers. A simplified diagram of the physical arrangement of the-hi HCUs, scram discharge volume, and vent and drain system is shown in Figure 3-5. tL) 2 The HCUs on ea'ch side of the reactor are arranged in 4 rows. Immediately above 9% the a rows of HCus are two cross connected " race track" shaped headers fabri-cated with 6" piping into which the discharge from each scram outlet valve is v g.~ piped. The two connected 6" headers on the East side ccmprise the East bank [ scram discharge volume (SOV) and th'e two connected 6" headers on the West side p comprise the West bank scram discharge vol"me. Each HCU insert and withdraw n line is connected to the CR0s below the reactor vessel with '3/4" Schedule 80 Jj piping through which the scram inlet and scram outlet water flows (and water r for normal rod drive motion). These lines average over 50 feet in length. J The lines from the HCU scram outlet valve to tne SDV are fabricated with 3/4" Schedule 80 piping and are approximately 10' in length. The Scram Instrument Volume (SIV) is located on the West side of the reactor at one end of the West side HCUs (and SDV). It is configured as a 12" diameter 10' hi'gh vertical cylinder. Single float-type level switches are installed to monitor the 3 gallen and 25 gallon levels. Four float-type level switches are k provided at the 50 gallon level for the purpose of initiating a reacter scram (SIV Hi Level Scram) before the SDV begins to fill beycnc tne point where complete control rod insertion would be prevented. f At Browns Ferry Unit 3, the East bank and West bank SDV each drain via 2" ? schedule 160 pipe to a single SIV located on the West side. The drain line d for the West bank is approximately 18' icng while that from the E'ast bank is k approximately 150' long. In each run, the total elevation fall in the line r 4 ' I h %i@v k ;W i--hfi-a m-%s[-G G 740Q rp

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.-.u. n (', b (,: e 1 4 CAUSES NVESTIGATED Immediately folicwing the event at Browns Ferry 3, all aspects of the scram system were investigated in an effort to find the cause. The Reactor Pro-tection System (RPS), the air system, mechanical aspects of the CRD and '/ various valves, the CRD and HCU hydraulics, and the possibility of air in i the hydraulic system were considered. Finally, attentien was focused on the East bank Scram Discharge Volume. j a b ~ Electrical Investications a Fj Following the first manual scram, the operators verified that,the blue scram lights were illuminated for all control reds.. Both the scram inlet j and the out.let valve stem position switches must show a:$ open valve position h to illuminate these lights. Illuminatbn of these lights for all CRDs would ' J. indicate that the electrical portion of the RPS had successfully generated D, a scram signal to cpen all scram solenoid valves and that all scram valves il had actually opened for all control rods. M a M The Reacter Manual Control System (RMCS) which is designed to centrol only 9 M. cne control rod at a time was reviewed to determine if there could have 7 been possible interference with the scram function. It was determined that l# postulated gross failure of the RMCS and initiation of multiple centrol rod ) drive withdrawal signals would not prevent insertier. during scram since E upward scram forces are more than three times the nagnitude of the with-a drawal forces under these conditions. y By use of reference drawings, hydraulic control unics frca each of the four .I ] red scram groups were verified to be randemly positiened en both sides of 3 the core as shown in Figure 3-3. Control rod electrical signals to a group k 1 rod on the East side of the core and a group 1 rod en the West side of 0 the core would be identical. The rod iasertion pattern during the event shows that en the East side a number of rods frem each electrical group did h, not ecmoletely insert while en the West side, reds from all electrical groups did ccepletely insert. [ o V 0 Aq i, -r-- 3-.,. _n_ E,a ki' _3 [j

y a,;4;.s g \\ . g. y y m. w.p. m a.. mn,;..

.; e

[..*(,}$. hi. (3 r 9 9 Basei on this analysis it was concluded that the failure of rods to

p fully insert only on the East side was not caused by any electrical mal-i function in the RPS trip logic, s

aj TVA test (entitled SMI 150) was performed to verify that the response times j for the scram actuating relays to fully de-energize were within technical i$ specifications. Verification that they were, eliminated the concern that y an electrical problem delayed opening of the East side scram valves which h in turn resulted in partial insertion. b g (k A test of the holtage on all chancel A and channel-S scram grcups showed $[f. that -all went to zero follow!ng a manual scram and all returned to 125 VAC f;; when reset. This test was run to verify the requirements of US NRC IE h Bulletin 80-17. A visual and electrical search of the scram circuitry caoinets for spurious volta.ge sources and loose wires (that might have i provided a path for electrical pcwer) to prevent a drepout of the scram My relays was performed by TVA. None was found. ha ? CR0 Tests .j Various tests were run on the CRDs on the East side to verify that CRD 33 seal intecrity, friction and scram times were within allowable limits. r; U, CR0 seal integrity was measured via a stall test. Results of these tests j did not indicate any unusual amount of flow during stall conditions and, '7[ consequently, the CR0 seals were judged to be intact. Stall tests could [i,; also have provided a means of detecting scram cutlet valve leakage. How-f ever, this test was not done. friction tests and single red scram tests 1 also shcwed no anomalies. p 6 ,s j Non-Cendensible Gas in the Hydraulic System 3 The effects of air or nitrogen in the CR0 Hydraulic system were considered. [] Upon questioning, GE CRD experts stated that non-condensible gas in the fj;; hydraulic system would only cause problems with normal insert and withdraw y motions but wculd not cause problems with scram' insertion. This is because stepping the rods requires intricate timing of rod motion and Tatching h. o 3 t 'l$:2 Q Q.r2.C. W W ~'~.L y 4.,2:.4 5 X = T'~ ;:...;,. M 7 7 Xl m.- - - - m-L -.. ::s

w..

.m t

,,, 2. ", x w,

. n q p a g e ; p ;< :. h.s & 4 1y> % :Sqa9& \\

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h.,

() 3 3 [ whereas scra=ning is a single motion. During a stepping functicn any non-condensible gas would undergo compressions and expansions much different from the behavior of the non-compressible liquid. The presence of nitrogen gas in the 50V prior to scram would tie no different than the presence of air which is there routinely. Following initiation of the scram, the vent valve closes sicwly enough to allow a good portion of the non-condensible gas in the 50V to be vented. e \\ 9 D e g ? L. 4N 7-D

Q;g.

~.

x WL ..M u.nj w C n sm isf.y abp::;9kk;&y:LQgy 3 O. O ,V 5 EVENT SEQUENCE ANALYSIS a As discussed previously, the control rod drive HCU exhausts are partitioned 1 into East and' West bank scram discharge headers. Control Rod Drives which j discharge into the East header are located on the Eas,t half of the core while CR0s which exhaust into the West headers are positioned on the West side of rs the core. The most notable observation of the control rod, posit' ions after the first h( manual scram was that all of the control rod drives exhausting into the West header inserted full-in (dxcept for the CRD at posit.icn 30-23 which inserted to within on,e notch of full-in) while the control rods exhausting into the East bank header inserted an average of only 20 positions. This CRD insertion pattern provides strong evidence that the fundamental cause* of the extensive ($ f ailure-to-fully-insert of the CR0s on the East side of the core was hydraulic g o in nature. More specifically, the rod pattern resulted from an inability of q 9., the East header CR0s to exhaust properly for seme reason. f: With respect to possible multiple scram outlet valve failures, all of the East h header scram discharge valves were observed.by the control room operators to .n Q have opened upon manual scram actua* ion. Additionally, all of the manual 3* isolation valves on the scram discharge lines of the East header HCUs were k inspected by the licensee imediately upcn shutdown, with each found to be 'S fully open. Accordingly, the remaining possible hydraulic cause; could have h been blockages in most of the CR0 scram exhaust discharge lines or inadequate free volume (or high back pressure) in the East header 50V. Subsequent scram n k testing of numerous East header' CR0s which failed to fully insert demonstrated, k however, that no blockages existed in the CR0 exhaust lines. Excessively h rapid buildup of back pressure in the East bank 50V, due to multiple CR0 seal \\ failures, cculd also be postulated as a mechanism which could inhibit full-in control rod motion. However, stall tests performed on the East bank CR0s, 9 W 'See Section 4 for a discussion of other possible causes investigated. 12 1 ~ t 3 p: CA p 9 : p ;7-y T'g. 7 -":] { ~r J. ,,p . c; gv , y. w-

em.s-w - w g u o< w ;. e wmw.woA.s w.-~ n ca hm.. ',. ?.- [. (:; i. s f together with individual rod and full core scram tests performed prior to j restart, demonstrated that an excessively rapid increase in SDV back pressure resulting frem multiple CRD seal failures was not the cause of the partial p. Accordingly, it was concluded that, for some reasen, the fast i scram failure. f bank SDV had inadequate free volume available to accept the full scram discharge from all East bank CRDs exhausting into the East header. Thus, the observed East bank centrol rod insertion behavior can best be exclained on the basis h that the East header SDV was at least partially filled with water when the ooerator manually scrammed the reactor. r As discussed in Section 3, adequate free volume cust be available in both the ,f M East and West headers to acccc$:odate water exhausted during control rod scram / insertion. Furthemore, water must be exhausted into the SDV with low back

I

.?: pressure on the drive piston to assure that technical specificaticn scram q speeds are met. A reduction in the free volume in the SDV could tend to in-a crease back pressure en th.e drive pistons tco fast which could then increase j the time required to complete scram insertien. Complete rod insertien would l still be possible, however, even for significant reductions in the available free volumi.in the SDV as demonstrated in recent single CRD scram test simula- [ j tions perfomed by GE. The GE tests showed that for a 40% decrease in the available SDV, a control rod can still fully insert over a broad range of seal lj leakage values. For a 70% reduction (i.e.,1.0 gal / drive remaining) in available d scram discharge header free volume, the rods could still fully insert if seal } leakage rates were small enough. For a reducticn in SDV of.this magnitude, however, increasing seal. leakage rates can cause the CRD travel (number of f b positions inserted) to decrease. The tests sbcw that drive travel decreases to only 36 positions (out of t.8) when a 70% reduction is ceupled with a seal [ 1eakage rate of 8.9 gpm. The GE test cases run for 85% reduction in free { volume (.5 gal / drive remaining) showed that even with no seal leakage, the drive would insert only 28 posi.tions and decreased to 22 positions for D tij 5.2 spm and 18 positions for 8.9 gpm leakage. Finally, as expected, the tests showed that the CRDs would not insert at all if there were nu free d , volume in which to exhaust (0.0 gat / drive) regardless of seal "kage.

Thus,

,l ( N I rmwmmw

mg. g g g 4wi-uh w on a W nh.. Q 2,.u cr/4 W:m eAQ.q. a.a M A % v 1t- -g j r l....3 o. these tests clearly demonstrate that CR0 travel during scram insertions can j be sharply reduced if the amount of available exhaust volume is reduced suffi-ciently. ~ ja. Since the uanual rescrams (scrams !2 and 3) occurred with East bank scram dis-j charge volume almost full of water en each occasion, these later scrams can h be used as models for back-checking the cause of the observed East bank centrol IT j red insertion behavior during the first scram. That is, the fullness of the ( East bank SDV during the first scram can be qualitatively and somewhat quan-J titatively inferred by comparing it with rod motions, during the later scrams. The available free volumes in the East bank SDV for each of the later scrams H can be calculated by multiplying the drain times discussed in Section 2 by the C East bank scram discharge volume drain rates discussed in Secticn 6. The amount ~ of free volume which would tiave had to have been available d0 ring these later scrams ~ ] can also be calculated frcm the observed rod motions during these scrams F.j together with the GE test results. Comparing the volumes calculated both ways ] can then be used to show whether or not the observed rod motions during each [I] scram were censistent with the amcunt of discharge volume made avkilable by Q the drains between scrams. Once these are shown to be.censistent, cne can ti 1 infer the limited amount of free volume which must have been present in the 3.] East bank SDV during the first scram. The East bank drain times, total number h of positions inserted, and average number of positions inserted per rod used h in this analysis are shown in Table 5-1. Mb The drain times between the first and second manual scram was 93 sec:nds and i between the.second and third manual scrams the drain time was 53 seconds. [ Tests at Srewns Ferry show that the normal drain rate for the Eest SDV is j about 11.6 gpm when East ano West scram discharge volu=es are draining j simuluneously. Thus, by multiplying this normal drain rate times the drain [ time between scrams, we can calculate approximately how much water cculd fi have drained out (free volume made available) of the East bank header during @4 k i [ }. s '(( I? y W,. -- w % mrm-&7m,7.--g, W ; 9.:.~- ,3 3

w,- -.,f...., # _ _ y a _,,w,,,g .;, a,. j ,3 h ( i.,: the periods between scrams. Multiplying, one finds that about 18 gallens would have been made available during the first drain (between scrams 1 and H f

2) for the second scram while abou't 10.2 gallons we'uld have been made avail-able during the seccnd drain (between scrams 2 and 3) for the third scram.

n On the other hand, from the GE tests and the average rod motion given in

g. ;

Table 5-1 to a first approximation and assuming no CRD seal leakage, an C average of.18 gallons per drive was available for the second scram while j about.07 gallons per drive was available on average for the third scram. 4 Thus, for 93 drives, to a first approximation and given no saal leakage, d a total of about 17 gallons of free volume was available in the East SDV 'S I. h. for the second scram while abo,ut 7 gallons was available for the third scram, s pA However, if every East bank CR0 were assumed 1;o hav,e a seal leakage of. 5 gpm,* q frem the GE test results the required volume per drive would have had to have H been no more than about 20 percent more than the above values. That is, about / 20.5 gallons of free volume would have had to have been available during the d Jj second, scram and about 9 gallons for the third scram.

1 Comparing the results of the above calculaticns, it could be concluded that J

the East SDV was draining normally between scrams one and two, and' two and three, ,r M and that the average rod insertien during the second and third scrams was the O. amount which one would expect for the amount of volume made available by the f drain. Thus, the insertion behavior of the East bank centrol rods logically h could be explained en the basis of a virtually filled SDV during the second T M and third scrams. b W{ This same. approach can now be used to infer the cause of limited centrol rod (' (( motion during the first manual scram. From Figures 2-2 and 2-3, the average ontrol rod insertion during the first scram was 20 positions.' From this value we would infer (using the GE test results) that there was an average of only.35 gallons per drive'available (or about 33 gallens total) in the C F] t Q = g

Conservative based en CR0 maintenance recommendations.

r ?; w - ( 9.. 5 )

m. =--,..

m ..m g y.w...m. m.m.,....

. ::2 - ~ =

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"27:&'% ku;d,&M:sb (' 4 t*.s s East scram volume during the first manual scrm. This assumes no seal W leakage. With a 5 spm seal leakage, we would infer that only about.45 h gallons per drive (or about 42 gallons total) had b'een available. {1? The above calculaticnal results show that the partiaT scram failure during the first scram can most easily be explained by having an initially partly , ?< O filled East bank SUV. Similarly, the subsequent CR0 f ailures-to-fully-insert h are explainable based en a partly fillee scram discharge volume. P_ tf4 It should be pointed out, however, that there was considerable spread accng 3 f the control rods in the number of notches inserted after the first scram. The variaticn from red-to-rod,cculd be explained by CRO-to-CR0 differences 4 ~ in such parameters as seal leakage (which sign,ificantly effects number of notches inserted), centrol rod drive fricticn, nitrogen accumulator pressures, M m etc. N Id Finally, evidence that the East bank scram discharge volume was initially q1 Ri partly filled with water can be fcund in the elapsed time to activate the ID SIV Hi Level scram switches folicwing the first manual scram. Reacter scrams t. l at BF-3 prior to the June 23, 1980 event resulted in time delays 'frem reactor scram actuation to SIV Hi Level scram actuatien ranging frem 42 to 54 seconds. o The first =anual scram frem the June 28 event had a delay of only 19 seconds. <k For a normally empty 50V and SIV, the time delay would represent the time i it takes for water exhausted frca the CR0s to enter and begin to fill the SDV, travel down the SDV-to-SIV drain lines, and fill the SIV to the 50 gallen level. n f If water were already in the East SDV, water exhausted from the CR0s wculd 2 almost immediately start to push water cut of the East SDV and into the crain [ line. This wculd cause the SIV to fill more rapidly. Thus, an elacsed time of only 19 seconds to actuate the SIV Hi Level scram switches provides important evidence that the East 50V was already almost c:mpletely filled with water h at the time of the first manual scram. w

W

. j; ~ r.' 'p 7 t y- ~

mpm

.-.n ._.m

- = _. 1.:. c. C. C 6 SCRAM OISCHARGE VOLUME / SCRAM INSTRUMENT VOLUME INSPECIIOli5 AfiD TE5T5 h Following the partial scram failure event at 8F-3, TVA, with the assistance i of GE, embarked on an extensive in'specticn and test program. These inspections 2, ji and tests were performed to try to pinpoint what cau. sed substantial water g to be present in the East bank scram discharge volums on June 23, 1980, y while the scram instrument volume level switches were indicating both headers were empty. The inspection program included physical examinations of the m j drain and vent piping, the scram discharge and instrument volumes, as well as the drain and vent valves. These inspections were performed in an attempt A to determine if a vent or. drain line blockage had caused the East bank scram a ~ tj discharge volume to not drain properly. Additionally, drain tests were per- [$ formed on b6th the East and West headers to establish the drain characteristics g of these components. The following paragraphs summarize the results of these inspections and tests. By Inscections 9.I The 2" drain line between the East bank scram discharge volume and the scram

e(j instrument volume were checked for blockages. The drain piping was cut at fj several locations. Metal tape was then inserted through the draih piping seg-M ments. These inspections uncovered no obstructions in the piping'between the lh SDV and SIV which could have impeded nor=al draining o'f the SDV. A fiber h

optics inspection of the inside of the S'DV at the low point of the 6" diameter f;{ SDV (where the 2" drain line connects to the 6" SOV) revealed no foreign objects which could have blocked water from draining out of the 6" 50V into ~ .s}; the 2" drain line. The vent piping which cross-connects the high points of h the' East bank scram' discharge header was also cut, flus,hed, and inspected. h No obstruction was found in these vent l'ines which could have impeded or y prevented nomal draining of the East SDV. I!d L { Following the event, the vent valve on the East header was removed and a vacuum pump connected to the Clean Radioactive Waste (CRW) side of the vent [ line. Eight (3) inches of mercury was indicated by the 1.35 CFM vacuum pump b M S ? E yc t KE. u.L ' ~' Y y,;, f.',,1 Z n . :w. : w~:" ~

yn gg u e f..e,;av. gn ~ -..a....... g ' - ggg c.

- g

-} } e. w a, i t' g ? 9f: In view of these design deficiencies, AE00 believes it is necessary that w J, modifications be made to the SDV/SIV arrangement and isolation features. t P Specific recommendations for changes in the SDY/SIV. design are provided in a C; the report. These recommendations should be considered along with those ay from oth'ers who are also reviewing the BF-3 event. AE00 believes, however, 'i, that the changes described in its recommencations, which result from the ( j findings provided in its report, are necessary to adequately reduce the risks g[ associated with unreliability of the BWR scram system which can stem from O the undetected accumulation of water in the scram discharge volume. I;l ?! e j 1 9 ); ,Y.j h)y s;. i$,, s / 41 M A> k t.E a i [ 9 'i, L;s s y L w - -Y %9 -3 g .*T*

I 9,--

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3. :... -. :-

3,. 3. 3 z,. ,g.,mmzg;aq,4, ( },; C-(: s c k p u [ gauge after a few minutes of pumping but fell off sharply to 2 inches shortly thereafter. Neither the validity of this vacuum reading nor the O [ reasen for the apparent and brief vacuum pull could be determined by TVA. h Several days later when the 1.35 CFM vacuum pump was reconnected to the ~ T same East header commen vent pipe, no vacuum could be drawn after the vent e j line was flushed. A test of the East header vent valve itself showed that s.1 it was operable. y m f.f] The scram instrument volume was also visually examined with a boroscope by 9 inserting it through the vent and drain line penetrations. No cbstructions [a were found %hich could have pr.evented draining into or cut of.the instrument H volume. m hj An inspection of the 6" East bank SDV and drain line showed that they sloped 1 continuously downward toward the instrument volume, with the exception of a j] localized 3/4" rise in drain line at the expansion loop in the steam vault. This might have been a loop sea'1'of greater depth when the steam vault was [j hot during normal reactor pcwer operation. The overall drop in the drain line d between th'e East SDV and instrument volume was determined to be l' 7" over its MW 150 ft. length. From the inspections discussed above, TVA was not able to m locate a blockage, loop seal, valve maloperation, or other impediment to ] I draining which could be described as the root cause for holding water in ng East SDV. TNy Scram Discharce Volume Vent and Drain Tests !,J TVA ::erformed a series of drain cests en both East and West SDV headers over [ a period of several days icnediately following the partial scram failure even't. 5 The purpose of these tests was to determine the effects of a restricted vent j path en East and West bank SDV drain capabilities and to quantify the ncmal d drain characteristics of the SDV. Special test procedures were written for r b. these tests. hg h5 T h ! x e E \\ w n, m aerr =.= u m m._;p, c _m.w rur=m m. _y

ge,,.2. g.:, c -... e , - 9p g g., m.y.%., t i. .O O f Typically, these tests involved initially filling the East and West 50V g i discharge headers and scram instrument volume tank.with room temperature demineralized water. During filling operations, the East and West header vent' valves were kept open and the scram instrument volume dra,in valve' was y: J kept closed. Nomal drain times and drain rates for'the East and West SDV Al headers and scram instrument volume were then detemined by recording the- ] elapsed time necessary to empty these volumes with the vent and drain valves .m Vacuum hold tests (simulating vent line blockages) were performed to 77 open. C, determine the drain capabilities of the headers with the vent valves closed. Water level in the SIV and SDV was monitored by ultrasenic equipment and verified by a clear tygon (manometer) tube attached' to the scram discharge volume headers. Clearing times of the 50, 25, and 3 gallon level switches L[j-attached to the SIV tank were also recorded during the tests.

i.,

h Summary of Test Results n f} Scram Discharce Volume Vacuum Hold Tests j% ~ East Header ap j With the West header drained to empty, the East header was allcwed to drain 1 into the SIV with the East header vent valve and SIV drain valve closed and U) the West header vent valve open. For this conditicn (which simul.ated a blocked East header vent), water drainec from the East SDV into the SIV tank [, at a rate of only 0.6 gpm. g 72$ West Header a M> For this test, the East header was first drained to empty by open'ing its p associated vent valve together with the SIV drain valve. The West header ( was then allcwed to drain into the SIV with the West header vent valve and ,~ [t SIV drain valve closed. For this condition (which simulated a blocked West ag header vent), water drained frem the West SDV into the SIV tank at a rate h of about 3.2 gpm. p (( R },; y q. [? ii L' ?w t ~ [ -'W.

4P h[* bk 1B

r.A

m..w - m c u.a g y u :, n,,; gg,,, 3 .. ~. '.,. ] (,'. Q ~ East and West Headers [ For this test, both the East and West headers were allowed to drain simultan- ) eously into the SIV tank with their respective vent valves closed and the k SIV tank drain valve closed. After an initial water surge, the ccmbined L' drain rates of the two headers into the SIV tank was 0.6 gpm. I lW Scram Discharce Volume Drain Tests W. These tests were performed to detennine the dra-in times and drain rates of U the SDV and SIV during normal draining (open vent and drain) ccnditions. s i Drain tests were performed for both East and West headers draining at the N same time. - The system was first filled with the SDV vent valves open anc wy the SIV drain valve closed. kt time zero the drain valve was cpened. Ilf Ultrasonics indicated that the West header e5ptied after abcut g% minutes while the East header emptied censiderably later at about 25 minutes. Jd Additionally, the 50 and 25 gallen switches in the scram instrument volume cleared at about gb minutes and~ 101/4 minutes, respectively. The SIV T' 3 gallen switch cleared after 11 minutes and 20 seconds had elapsed. Based f en the volumes associated with the SDV headers, these tests showed the average drain rate (with both SDV headers draining together) of the East

f,

SDV header to be 11.6 gpm while the average drain rate of the West SDV H. header was shown to be about 35 gpm. The average drain rate for the SIV f based on clearing of the SIV level switches was 24.5 gpn. However, this drain rate was with the East SDV header still draining into the SIV at an $e average rate of 11.5 gpm. That is, the SIV drained 24.5 gpm faster than the East SOV drained. ~ W e N t w ,g a:7._m,rT = -~=:gq ~ -~;7'~

,;;g 7 7 3 p gn q a;;m..,; ayn.n.7, u.__

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ycp.~.... mr.re.wwpm ve%g.N. nc ' -ggs.

a. - Mce m O MW

. S.,, h - @r; j z 7 PREVIOUS BWR EXPERIENCE OF c FAILURE iu FULLY iliSERT et s A review of previous BWR experience was performed with respect to failure $v to fully insert control rods and problems with the,50V. The sources of in-E formation used were NUREG-0640 and computer searefie.t of LERs.. Computer ( searches via the NRC LER system and thi Oak Ridge.' Nuclear Safety'Information 4 Center data base revealed no later events'more significant than those re-s f ported in NUREG-0640. w i eg j Most instances of failure of rods to fully insert resulted in a number of rods latching in position 02 (position 00 is fully inserted). Up to the time of publication of NUREG-0640 in April of 1978,]12 scram events where y$

However, some rods failed to fully insert were tabulated. These events in general involved a relatively small number of CR0s, between 2.and 15.

j one edent at Dresden 2 in November of 1974 inYolhed 96 rods. Ninety-three .);j stopped at position 02, one at position 04, and two at position 06. The $,/ only cause reported for the failure'of rods to fully insert was damaged /f stop piston seals. Stop piston seal damage can cause excessive leakage h past these seals during a scram which could be large enough to fill (and j pressurize)the discharge volume in advance of the control rods reaching 3 their ful.1-in position. kq - hi M N li; r.. 't kA0 lb

M M.

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s

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8 FINDINGS r) ? l 1. The eartial failure to scram at BF-3 on June 28, 1980, was accarently due I to the cresence of water in the East scram discharce volume header. As supported by the tests, inspections and analyses discussed in Sections 4 and 5 of this report, the apparent cause of the extensive f'ailure of con-g trol rods to fully insert on the East side of the core was the presence of h water in the East scram discharge volume. header. 1 1 l G h A %N N., Q% r ~ (nh ka v p)> b4MmW v;' m

Q it k

~ q i';y 4 P,.t ne i< i d. l @I. y h, t. y, 4. <

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m== ~ .c klg. a.; o O / y t k P 2. The BF-3 scram instrument volume Hi level scram function did not and does not provide orotection against the accu ulatien of water in the East scram discharce volume header (with attendant loss of East bank scram function) even for normal venting and draining ccaditions. I ~ Orain rate tests performed at SF-3 show that water drains out of the scram instrument volume tank considerably faster than water drains into it from d theEastbankscramdischargeholumeheaderehenfornormal, free, unobstructed

r

3. 2 venting and draining. Based on the tests, the averace drain rate of the b

SIV is approximately 35 spm while the averace drain rate of the East bank >+g, Q scram discharge volume header is approximately 11.6 spm. For these drain y j characteristics, water.wil'1 drain out of the SIV leaving it Eirtually empty (( while water may still be preshnt in the East bank SOV. This actually occurred 1 in the East header drain tests. During the test, the SIV emptied about~ h 20 minutes before the East header fully drained. h3 Withtheserelatihedrainingcharacteristics,ifwate.rweretoleakinto l(1 the 50V faster than 11.6 gpm, water would accumulate in and fill the East t header (since water is being added faster than it can drain out). At the n same time, the water draining out of the East header (i.e., at 11.6 gpm) [ will not accumulate 'in the SIV since the SIV drains at a fast'er rate (i.e., p 25gpm). This process would result in water filling the East header with-P out an automatic SIV Hi Level scram ever occurring. We have also found that water drains out of the SIV so rapidly that the SIV Not Drained alarm would not alann in the control, rocm. Thus, there would be neither control room indication that water is filling an East SDV ner automatic reactor scram ) actuation to provide protection against partial loss of. scram capability._. J In view of the above, with regard to the SIV Hi Level automatic scram function, we have found that continuous automatic protecticn against filling the 4; East bank SDV (with subsequent partial loss of scram function) never did and still dees not appear to exist at SF-3. Furthermore, any BWR with a SIV normal drain rate significantly faster than its SDV normal drain rate also would be without automatic protection'against fillitg of the SBV. Although . J t 1 _,..m&.j suw w w paw m.xz=xm_ w. <I,3 f- .l-j.

j.

?

_Q -)O

+...;; ~.:s.m. g . y: ...am. :7 3:4....... y.,._..+:,; q 3, 3 , i. O. o i,q. not verifiec by test, it is likely that the BF-3 West 50V header also would be in -his category. j The loss of automatic scram function can be explained in hydraulic head terms. The SIV is a high cylindrical tank and the 50 gallon'SIV Hi Level .c scram is located over 8' above the bottom of the tank. Thus, it is 4 necessary to build up a head of 8' in the SIV before the Hi level Trip i switches can actuate. If the drain line from the SIV to CRW is a relatively short line (as is the case at BF-3) an 8' driving head, would result in a fairly rapid drain rate. On the other hand, the SDV header is a horizontal ? pipe with a small slope. Even when filled, the maximum head of water that k can be developed above the SDV drain (at BF-3) is approximately 24'.

Thus, o

~ even with a relatively short drain line between the 50V and the SIV, the flow rate in this line would normally be low' because of the low head. 1 . g L! Actually the SDV header drain and the SIV drain are the same size for BF-3, l but the 50V drain is considerably longer. As a result,the lower available } hydrostatic head in, combination with the higher fluid flow resistance re-A sults in 'a much slower drain rate for the East 50V header than for the SIV. d Such an arrangement can never detect accumulation of Water in the SDV. .1 y M R

)

%) e p E .g .y 'f 9 i M2.;=;gn=r==;r.=v=q yqq,9 g #. a==,g._ %,5=cer g c. g g 3.,

u i ara.,n w na g.yi? m c

. ~ u _,,gg

.e.

m.. _ _c rre y m.. gg,,4;,, ._ s s o c . L' ( I, 3. A sincle blockace in the West header vent or drain line could comaletely

s..

T disable the automatic reactor orotecticn function installed to orotect against a loss of scra:n cacability for all control rods. E.1, ~ 3: For plants like BF-3 which have one 50V which normal'1y drains significantly sloce than the SIV, it is possible to comoletely disable the protection [ provided by the SIV Hi Level scram for both the East and West SDV by i postulating a blockage on the f aster draining 50V. Reduced flow from a e-Q blockage on this faster draining header SDV, when combined with the normally j slower draining header flow, may total less than the scram instrument volume drain rate which would then result in the SIV empty'ing with both SOVs still y 5 containing ' water. This would,be a serious and undetectable condition if +.,{ water inleak' age were to subsequently develop into both 50V headers such as 3 to keep the headers full at all times. For such a situation, there would lj be no automatic scram to protect against a total loss of scram function due to CRD water inleakage since the SIV water level would never rise to actuate 2 the SIV Hi Level scram switches. o ~ L 4 'i .S9 ll r .L %w Yl fN d c lC i% ?G {N f ~ l t e 4 i T

._ 3 3 ;. m y y 7A w.u.s. n-6:a-~,. g.i a= ~ ~+ + - % g - w A 4u=a +-,%QihHa i, {

. *, 3 G-O p

t-B u a 4. With the current scram discharge volume / scram instrument volume design, e f,. a blockage in the SDV vent or drain oath can cause a cartial loss of. I scram caoability and disa'ble the protection function installed to prevent [ it. b: %p As discussed in the previous sections, a blockage in the SDV header vent ,y or drain path will drastically reduce the drain rate of the scram discharge II: volume. Water leaking past the scram outlet va,1ves (or fron other sources) h could then cause the scram discharge volume to fill. ~ Since the CR0 temperature 2 probes would allow about'.1 gpm of undetected leakage, as much as 9 gpa could %fj leak into the SDV header undetected from all CRDs. Thus, given' a partially blocked West header drain, for example, the West header could' easily start j to fill with water, leaking in undetected through the West side CR0 scram a P cutlet valves. At the same time, since the drain rate of the West header -a )j with a drain line blocked could now be substantially less than the SIV u" drain rate, water would not accumulate in the SIV. Therefore, the JIV Hi J;j Level scram skitches would not actuate to prevent filling of the header, p) Thus, with the present SDV/SIV and Hi Level scram arrangement, a single fj failure such as a blockage of a SDV' drain or vent can help initiate a partial loss of scram capability and disable the protective function designed to l9 prevent the loss. t a m es f[,d s,v -= ae 4 n e d k d 'oH . 28 - ~ ki8 - r N $.d D 4 g l:v; p:r.~,y-v~c -:..v+= u = "Y====~---~+-f T*=='- ~~=== W== '* W" T * " m

Uggg+nM+g 4sw zock w

.p_: a. 4.p j g.; g g g, g g,

[ p'$ h h '~ ~ ,g h y 5. There are numerous actual and cotential mechanisms for introducing and (g retainino water in the SDV with no accumulation in the SIV. ~ Review of the vent and drain paths for the scram discharge volume and the mg scram instrument volume has shown that there are numerous actual and potential mechanisms which could slow or even stop SOV drainage into the SIV. Since the SIV would still maintain a high drain rate, it would be h possible for the 50V to retain water while SIV instrumentation indicates ,ef empty. N p R Possible sources of water are: water from the previous scram; multiple fj scram outlet valve leakage; or injection frcm SDV flush lines. s ? Mechanisms which retard free draining of water out of the SDV include: 4 a blockage in the' vent piping; a plugged SOV-to-SIV drain line; a closed u Vi 50V vent valve; a vacuum held in the 50V by a loop seal somewhere in the vent line; vent line siphon effects from water in the 50V vent line; 3 1g venting to the closed CRW system in the Reactor Building Drain Sump below water without vacuum breakers; vacuum effects from fluid flows through the a') CRW piping system; Vacuum effects frcm condensing hot water in 50V from the-d) l previous scram. g Venting of the SOV to atmospheric pressure while the SIV drains into the [ closed CRW drain system (which could be pressurized above atmospheric pressure) a} could also inhibit draining of the 50V headers if there is insufficient y downward slope in the SDV drain line. Since the CRW exhausts under watar f in the Reactor Building Drain Sump and non-condensible gases are present in the fluids draining through the CRW drain system, there is a-sg k possibility for pressure to build up in the CRW drain system..This [ pressure, in conjunction with a small loop seal in the drain Tine from the h 50V to the SIV, could hold up water in the SDV even if the SDV were vented h directly to atmosphere. N W 3 / e J) 5 pa I.) k g=gw.gg%-gg;=p: =r._g a_= .;m_qqwg .y

g ,~r,-.~ a. w w 3:3; @ w A M i-Q ~_ - _ _ _ - _ _ - _ - - _ _ _ - _ _ - _ - - - _ _ - - - - - - -, - - - - - - - - - - - - - - - - - -.iw& Q i1 l* v *.,:. G. o 1 / I 6. The current scram discharce volume / scram instrument volume desien results h in the automatic Hi level scram (safety) function beino directly decendent en the nonsafety-related reactor buildina Clean Radioactive Waste drain system. I For the scram instrument volume fii Level scram switches to activate, water } must accumulate in the scram instrument volume. For water to be able to accumulate in the SIV, it must be able to drain at an adequate rate from the SDV into the SIV. However, from the drain rate tests performed at l SF-3, improper venting of the SDV can sharply or totally prevent water [ from draining out of the SDV. Proper draining of 'the SDV. is directly dependent on the venting function provided by the reactor building Clean c O Radioactive Waste drain system (a required systems' interaction). Accordingly, k we would conclude that operab,ility of the SIV Hi Level scram function is r. dependent on the venting provided by the nonsafety-related reactor building 2 CRW system. Unanticipated adverse venting behavior of the CRW system, ~ which results in reduced venting of air back into the SDVs, can result in H the holdup of water in the SDV with little or no accumulation of water in j the SIV. This dependancy appears to be particularly inappropriate if not unacceptable for a reactor protection function which is intended to prevent the loss of reactor scram capability. J e 9 4 9 y t 'l e e t N'2 1 T 3

.. ~., ~,,. .a m.:. y.c

a. a +..

.s. w.. ,w.a,._._, w - x .,... _.;. e.... y

a....

a-a_m.u.a.. .-).. ~ j, h. h .t n w ,i j[ 7. The BF-3 cartial scra:n failure event, tocether with recent events at

i' other BWRs, have shewn tham ficat-tyce water level monitorinc instrument _s, s

5d have a sienificant degree of unreliability. g C The BF-3 partial scram fai. lure event demonstrated en several occasiens m Q a significant unreliability of float-type level switches. As.shown en h.5 the event sequence recorder printout (Table 2-1), several of the 50 gallon level instru=ents failed to activate on different occasions. Furthermore, El

}

during calibration testing of the SIV leve.1 switches following plant shut- ? g down, both the 3 gallon and 25 gallon switches were found to be inoperable. 5j After the instrument taps.were flushed of residue, 'the switches operated satisfactorily. During drain rate testing of the BF-3 SDV, two of the four b =0 gallon switches failed to activate twice.in two drain tests. Additionally, m@ inspections at Brunswick Unit No.1, following a reactor scram on November 14, u }) 1979, revealed inoperable alarm and' rod block level switches due to bent +, float rods. Other surveillances and inspections at Hatch L' nit 1 cn June 13, 1979, found two SIV Hi Level switches inoperable due to bent floats binding i ] against the inside of the float chamber. These recent experienc'es indicate a a significant degree,of unreliability of float-type level switches resulting ~ 13 from varicus causes. b L.e f e kM $q >a

u S

p t< s dj y A h$) 31 - n n $i f"j t .) y: -= = a w g,, ~;~ Q Tq;;rg ~ ~ ; ~~^ _ 47'-~~~. n: my ;;g;n.. . :n : .+ n -, m_ x

. r... -, -,. ,,_..c.., e jg .,.g,, g. s

= \\ 4 '

3 w 8. With the current BWR' Reactor Protection System looic the cresence of certain automatic scram conditions creclude 50V drainino (scram reset) to cermit a rescram. 0(- In order to drain the 50V for rescram following a scram actuation, it is u necessary to reopen the SDV vent and drain valves and to recle,e the scram o h inlet and outlet valves (RPS reset). This requires the following steps:

1) place the reactor made switch in SHUTOOWN or REFUEL; 2) actuate the

] DISCHARGE VOLUME HI WATER LEVEL BYPASS switch; 3) the reactor trip signal Q must clear or be bypassed in SHUTDOWN or REFUEE modes; and finally 4) ? reset the RPS. However,. the following reactor trip functions cannot be 4 T. bypassed by the operator in the SHUTDOWN or REFUEL mode: } y 1 Orywell High Pressure, Reactor Vessel Low Level g;. 0 Main. Steam Line Hi Radiation g! Neutron Monitor System Trip Reactor Vessel High Pressure Condensor low Vacuum

a a

Main Steam Line Isolation Valve Closure * - j $} .Thus,if any of the above trip conditions are pretent,' resetting the RPS [ would not be possible. ~ For example, if a spurious MSIV closure event should occur at power with the 50V initially full of water, a reactor scram would occur with the control

1 rods failing to fully insert.

If the MSIV closure trip (or Reactor Vessel-HighPressure)conditionpersis'ted,thenarescramatt'emhtwouldnotbe g possible since it cannot be bypassed in SHUTDOWN or REFUEL modes. Thus, the trip condition itself would prevent the possibility of rescram. However, we do not consider that any modification is required in the RPS trip / reset circuitry to enable the operator to reset the RPS in the presence of any j automatic scram condition, since the capability to reset and rescram has f, not been defined as a required protective action. O

Depends on Reactor System Pressure Interlock setpoint.

+ 3 f L s . ~.: g.==.= w.,,.,... -

" 'L D" W5YYYSW"WYNb

^l.f.0.h ~

h h
) e s

9 s 9. If a scram condition exists which cannot be byoassed in SHUTDOWN or REFUEL } mode, then fai10re (to close) c'f a SDV vent or SIV drain valve can result in an unisolatable blowdown of reactor coolant outside crimary containment. /. With the reactor in an unscranTned state, the' scram outlet valves provide 1 both a reactor coolant pressure boundary function and a primary containment [ isolation function. Deringareactorscram,theic'ramoutlet,va10esopen N3 (onepercontrolroddrihe)andtheSDVventandSIidrainvalvesclose. sj, Reactor coolant pressure boundary integrity and primary containment iso- ] lation functions are then transferred to the scram discharge volume vent i and SIV' drain valves which seal the SDV. We' have found that there are no a

j redundant isolation valves in tlie vent or drain lines to. provide these

] isolation functions for a scram condition. The failure of any one of G these valves in the open' posi~ tion, therefore, could result in an uncontrolled h blowdown oY reactor water outside primary containment and into the CRW n The blowdown would. g drain lines if the operator could not reset.the scram. { ultimately discharge to a drain sump which is not designed to handle the ii heat load or pressure buildup. With the present SWR RPS design, the operator N would be able to reestablish pr'imary containment isolation (with scram out-h , let halhe closure) only if the RPS could be reset. However, if a reactor h scram condition persists and it cannot be bypassed in SHUTDOWN cor REFUEL { mode (i.e., any of those listed in Finding !8) it would be impossible to [G reset the RPS to tenninate the blowdown. ,4 Thus, for example, a scram caused by spurious closure of the MSIVs with a (,16 failed open scram' instrument volume drain valve would result in an uncon- +Q trolled blowdown of reactor coolant cutside primary containment and into k the drain sump room which contains the engineered safeguard pumps which are required for mitigation. Slowdown would continue as. long as the MSIV e closure ic' a5 cond'itio'n existed (MSIVsiot reopened') since this ' trip cannot [ r be bypassed in SHUTDOWN or REFUEL mode. That is, the scram outlet valves could not be reclosed to isolate tne blewdewn until the MSIVs could be a N reopened. For events which result in scrams caused by conditions which (;]f cannot readily be cleared, uncontrolled blowcown into the reactor building y (secondary containment) could be sustained for an indefinite period of time M with possible envircnmental impact on the required mitigating, features. M nM h, m [4 6 n e a Y'

- *N *

,l f5b ' b ** s ~ ~, & E5k;-l-ff-QM y j l 2

.c-

- :.w. - ,m N. b g. I.' h. h 1 3

10. The emeroency coeratino instructions at BF-3 did not include'a'erocedure or cuidance for the coerator to follow in the event of a cartial or corolete scram failure.

The Browns Ferry plants, as perhaps do most (if not all) other SWP.s (and ~ probably all other LWRs), do not have emergency procedures for the operator t. f to follow in the event of a partial or complete scram failure. We have 4,

)

found that, although control room operators are trained to verify that the + M rods have fully inserted upon a scram actuation, procedures do not exist

4 j

for the operator's ir:inediate or subsequent actions if full control rod insertion does not occur. Moreover,althoughoperatorsarefuliyknowledge-y,; able of the function and operation of the standby liquid control (poison) L. system, the plant does not have specific procedures which state when the a. i alternate shutdown system mu t be actuated. W s Nd

M

,1 ya M Y,b &n ($.. ~ g-m r-h. m ff O. Ik i.: N O '1 4; ,.o il e 0 l 3 h s - si - '4 ] t c. _ _ _m... _ _,,_. ___,. 3rpy.y;~. p.~gwwwrm

u. e,.,;m[.,-~,-+-; + n %,%.:,-+ + ~.~~;FC k_Qiyw-%

.x+ m. m.,,7, ~ .c.

.e b.. :......- O O 4 s, k 9 RECOMMENDATIONS

q 1.

The ocerability of the Scram Instrument Volume Hi level Scram p function should be indeoendent of the Scram Discharge Volume -I venting and draining recuirements. 6g The current BWR scram discharge volume / scram instrument volume design hj configuration requires proper venting of the SDV and proper SDV-to-SIY .h draining to assure operability of the scram instrument voluce Mi Level y scram function. We recom=end that the operability of the Hi Level scram h be made indeoendent of SDV venting or draining requiremnts. We make ~ aj this recomendation because of Finding Nos. I through 6 discussed S in the previous sectio'n. That is, the hydraulic factors which control water level in the SDV an'd SIV should not be able to negate the response ~ fi of the Hi Level orotection function. We believe the acceptable configura-h tion would be to place the SIV tcnk directly under the low and of the 6" g SDV header and to connect the top of the SIV tank to t'. bottom of' the V low end of the SDV header by a short vertical 6" diamecer pipe (rather than 5 the current 2" diameter horizontal pipe). This arrangement should assure Dy water soillage from the SDV directly down to the tank containing the level k monitoring instruments. Furthermore, it would not depend on ' venting or f draining phenomena 'which are sensitive to blockages. We also, recommend two separate scram instrument volume tanks, one on each SDV header bank. ( Separate instrument volumes, in immediate proximity to their respective } headers, should assure proper water spillage into the SIVs and provide j adequate redundancy for protection against a total loss of scram capability. { It is. our fim belief that modifications which simply improve the venting ( of the SDV/SIV volume arrangement to assure operability of,the SIV Hi { Level scram functicn are not adequate. We recom:end that this uniquely p important safety functicn be made completely indeoendent of any vent or drain arrangements, thereby separating the water accumulation control and protection functions. We further recommend that in situ fill tests 3 be perfomed to demonstrate th,at the operability of the prgtective Hi j level scram function is insensitive to the vent or drain arrangement t for the design configuration inally installed. P.@% h - S n t p =59;c 55 w m ml_~g 9:== c F n - '7 =rmv==w e er e jf __ Q &n ~~ -Q - i a 'u s v : L _ : - z = J u g_L - 3

w-gy.A.<,pg<;c _-;w.o qps;.. 7 n:. p - o g g.m a a m u..; 9 g.

g.

2.,_ :..c.. ~ + . I'. '.. 'l 1 C) 6.. r 1 f .s i 9 2. Scram instrument volume water level monitoring instruments for the SIV Hi level scram function should be both redundant and diverse. It is recomended that diversity be added to the, redundancy of SIV level monitoring instruments for the SIV Hi Level scram function. Currently, there are redundant float-type level switches for each RPS' channel for the Hi Level scram function. On several occasions recently, as discussed in Finding No. 7, more than one float-type, level switch was observed to be inoperable at once. During and imediately following the BF-3 partial scram failure event, several float-type level' switches in the, instrument volume failed to actuate. In view of these experiences, we recommend that diversi*cy be included in the level monitoring function for the SIV Hi Level scram function. The important and unique protection g provided by this trip function requires that the presence of water in d the SIV be monitored continuously with extremely high reliability. We i g are recommending that diversity be added in order to assure this reliability. g Mcnitoring techniques, such as differential pressure cells, ultrasonic 's detection or conductivity probes, may be considered along with others for [ this purpose. 8 y Wy V3 u 41 ,i l 4 ~ e, ~ g 0N r; K i s. R'UY - Jf * **T **?. .'**kQ"f T*- s

' M % O M -OF-k D O W W i f k W~ % h m U D ?t % $L@ $ Q.%2bu ~ L'&._ si O O f u. (i 3. All hent and drain paths frmi the scram discharce volume and scram

j.}

0 instrument volume should have redundant automatic isolation valves. t. 'l As discussed in Finding No. 9, scrams which cccur as a result of automatic reactor trip conditions which cannot be cleared or bypassed in REFUEL or SHUTDOWN modes can result in unisolatable reactor system blowdowns out-side of primary containment if the 50V vent or SIY drain valYe fails to n!r' close. To protect against such occurrences, we recomend that redundant p f valves be placed on all vent and drain lines cor.nected to these volumes. 6 f., Redundant valves would also protect against equipment damage which might [J otherwise occur as a' result of excessively slow closure or delayed closure [ of one 'of' the isolation helves. These valves must be qualified and 3 capable of closing against full reactor pressure, flow, and temperature 4 7 cenditions in case the lines are not isolated within normally.specified a 4, p1 time limits. The vent and drain lines and drain supports must also be designed for the hydraulic loads and instabilities associated with the jy,. blowdown of the high pressure / temperature reactor coolant to the drain dj system. Prolonged blowdown may be ruled out as a design basis with * ~ %[. appropriatediverseiso'laticnorotheracceptableprehisions, Blowdown ~ instability due to isolatien valve time delay'is believed to be the cause {} of failure of the float-type level swit'ches at Brunswick Unit'No.1. l-. .4 , ?/ $b b t ,g Y m ( S h b 7 i m t.; ' { ..v .I.I E s '7 ,.g.,.. _QA 5J.7., w 7 ~ '] [ ~ ' ~ ~ ' I a,,.444.,.4 .TZ

gj

. r, jg ;yA .. =. -. p

jg. .3,5 R N ETT T,sO<EEM @@: sg...:. p,. _.y.g.s,_..+ @.

Ew.

b;j. r 5 G C ,j 4 i I L 4. Emercency coeratinc oracedures and coerator traininc should be k provided for comolete and partial scram failure conditions. I u; In view of Finding No.10, we recx. mend that emergency operating procedures and training be provided to control room operators to k respond to partial or complete scram failure conditions. These a procedures should include explicit statements regarding the conditions '4) for which the standby liquid control system must be used. The procedures ( should incluce cautions regarding operator actions which should not-f be taken which could. result.in a severe transient conditien (e.g., k main turbine trip) being created. The procedures should provide guidance I to the operator for start,irg up safety systems for standby readiness T (e.g., HPCI on minimum flow) or for tripping other systems (e.g., re-circulation pumps). The order of operator actions (i.e., immediate,- y 1j subsequent) should be considered, as well as when the operator should i begin attempting to insert rods manually. ,4 .k We believe that such operations (human factors) aspects can and shoufd be implemented in the near term. Such procedures and training wcold assure, ,q in the near tem, the most appropriate ' control room operator' action during ~ ). a scram failure event and well in advance of any ATWS modificatic'ns which [; may be required in the long term. ' '3 hl ,4 1l' k ~ 4 ,W f \\ r, 5 '.i l{ N; - 38 '. h .i lt ( T R t l

_ _tm_.__Me..,E%Q.74 [CI,"

t,.!A'._.:Q..O W3 f.i'I5 D12 +, - g l .UC171-c ,.y ..f.,.. , _l, _ r L ..~s-a

' NN@#'W4 ' d~ M " h e-AA d -h A , ]ti'f 'i / 6., e j (3 g i ? a Fi. -L 5. Consider modifyinc the SDV vent and SIV drain' arrancement to imoreve u. 7 scram discharce volume drain reliability.* p? As discussed previously, scram discharge volume draining currently ( depends on uncertain vent and drain functions provided by the reactor building Clean Radioactive Waste drain piping, along with relatively 7 small diameter, nenredundant vent and drain piping, which are susceptable p to blockage. This current, relatively unreliable SDV v2nting arrangement could be improved by increasing the vent line size and by adding an 7 alternate, reliable, and isolatable vent path. The alternate vent { path could be installed with a' check valve and air operated isolation valve to provide an alternate.and isolatable path for air inleakage into thescramdischargeholume. The alternate vent path could be vented .l either directly to the Reactor Building atmosphere or to a gas treatment y d system with a vacuum breaker. The check valve would provide automatic 'N isolation of this redundant line upon pressuri:aticn of the scram dis-charg. volume during a reactor scram. The' drain function could also ? be improved by providing a second drain li.ne fran the SIV to the CRW .q ~ 'h. floor-drain. .q a ? We believe that modificatiens, such. as these describec above,' would help ] improve SIV drain reliability. Improvements such as these wculd thus help [- to further reduce the number of challenges to.the SIV Hi level protective tp scram function. Y y o O F O

Althcugh this recommendaticn is only for censideration, we do believe s<

3 that it would further reduce the risks associated with loss of scram [ capability arising from water accumulatien in the SDV. s ? i - 39 _ 3 s On,rr; 7ZZ32;=g r===== .g._a- .t. m. m.,.. --. m. ;-. % )

.n uw = a#~ L y:_ I.. * '., {} y 7 10 CONCLUSIONS The Browns Ferry Unit 3 partial scram failure eYent which occurred on ,I June 28,1980, demonstrated that the present BWR scram system can be vul. j. nerable to loss of scram capability while operating at power,. Furthermore, the event showed that the loss of scram capability can occur in a way which goes undetected by the operator and unprotected by the reactor protection h system. M h f~ The information and analysis of the BF-3 partial scram failure, which is ];; provided in this report, concludes that the cause of the loss of scram l^1 L capability was the presence of water in the East scram discharge header. O g Furthermore, our analysis of the scram discharge volume / scram instrument Q volume design configuration, together with,its vent and drain characteristics, o leads us to cenclude that numercus actual and post'ulated mechanisms exist s p which can cause the scram. discharge volume to fill undetected and without j protecticn against such filling. Our analyses also show that certain scram 1 events can result in an unisolated reactor coolant blowdown outside of 3 primary containment following a single isolation valhe failure. i ] In view of these design deficiencies, we believe it necessary t' hat modifica-t tions be made to the scram discharge volume / scram instrument volume arrange-E ment and isolation features. Our spec'ific recommendatiens for change in the g SDV/3IV design which flow frem our findings have been provided in this report. E We believe that these reccmendations should be corsidered along with those of others who are also reviewing the BF-3 event. We do believe, however, that the desigri changes described in the recc:mendations are necessary to - j adecuately reduce the risks associated with the present unreliability of the BWR scram system arising from undetected accumulation of water in the k scram discharge volume. k . f e 4 t YQQ~ (Q)?}_.Q&K.CCQC'-EST5%dm~d;;5N "!" WIN % W'& rn e ~ ~. _~

' ^hw

_ _ 2. h =: % "

m .t _.r., ~#m,. _._. a . 7.. # _._ a g c.ig,s,,m s,.. ._w,a _.. CjventSequenceRecorderPrintoug Table 2-1 01 31 16 A034 Reactor Scram Manual B CY 5 34 A034 c CY 6 44 A033 Reactor Scram Manual A 01 31 24 A035 Reactor Trip Actuator Al or A2 CY 3 38 A035 CY 3 39 A021 Reactor low Water Level A CY 3 42 A023 Reactor Low Water Level C I CY347Ab21 Reactoi Low Water Level 0 CY 3 47 A036 Reactor Trip Actuator 81 or 82 f j CY 3 56 A022 Reactor Low Water Level B Ig CY 5 11 A076 REPT C Tripoed h 01 31 34 A003 Discharge Volume High Water Level C sn f:S CY 3 42 A003 g ~ 01 31 37 A002 Discharge Volume High Water level B CY.5 01 A004 Discharge Volume High Water Level O tg ? CY 6 58 A002 ik 01 31 40 A001 Discharge Volume High Water Level A i CY 0 03 A001 3 CY 0 18 A106 Malfunction Bus Energized- ?ej CY 0 33 A038 Turb. Stop Valve Closure Scram Trip A R, CY 0 33 A040 Turb. Stoo Valve Closure Scram Trio C t M. CY 0 33 A041 Turb. Stop Valve Closure Scram Trip D ?q CY 0 34 A079 Turb. Stop Valve Closure Scram Trip B l 0 CY 0 47 A043 Turb. Gen. Load Rejection Scram Trip B J.s CY 0 47 A045 Turb. Gen. Load Rejection Scram Trip D p CY 0 48 A042 Turb. Gen. Load Rejection Scram Trip A d; Turb. Gen. Load Rejection Scram Trip C CY 0 48 A044 .A084 Turb. Tripped - Loss of Hydr. Trip Pressure n 01 32 01 NO21 Reactor Low Water level A. C CY l 12 NO21 m [( CY l 57 NO23 Reactor Low Water Level C CY 2 04 NO24 Reactor low Water Level 0 k CY 3 35 NO22 Reactor Low Water Level B A:Q 01 34 45 A058 IRM Upscafe Trip on Level F NN CY 4 30 A058 F 01 34 48 A057 IRM Upscale. Trip on Level 0 CY 7 36 A057 Il CY 0 07 M57 CY S 13 A057 l 41 - t P esse

===amg t e * +**e

- - Am*

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Table h E, vent Sequence Recorder PrintCl p e v CY026Nb58 IRM Upscale Trip on Level F CY 0 49 A056 IRM Upscale Trip on Level B CY 0 55 N056 CY l 01 A056 CYl16Nb56 CY l '41 A056 [i CY 1 48 N056 ~ CY l 56 A056 y ~, CY ? 21 N057 IRM Uoscale Trip on Level 0 h CY 4 14 N056 IRM Upscale Trip on Level B 01 42 00 A035 Reactor Trip Actuator Al or A2 n 2 CY 9 23 A035 F? 01 2 37 NO35 o 9 CY 6 35 NO35 h (c CY 6 36 NO36 Reactor Trip Actuator 31 or 82 '9 CY 8 05 NO34 Reactor Scram Manual B I; CY8b6NO33 Reactor Scram Manual A y 01 4517 A035 Reactor Trip Actuator Al or A2 3.~ i CY 6 47 A035 s ~ CY 47 A036 Reactor Trip Actuator 31 or B2 h bl4536N002 Discharge Volume High Water Level B ~ r j] CY 6 09 N002 [ CY 6 16 A002 h 0146 30 A031 Reactor Scram Manual B 1 CY 9 48 A034 y CY 9 48 A033 Reactor Scram Manual A 01 47 43 NO35 Reactor Trip Actuator Al or A2 CY 2 37 NO35 CY 2 38 NO33 Reacter Trip Actuator 31 or 32 CY 3 05 NO33 Reactor Scram Manual A CY 3 05 NO24 Reactor Scram Manual 3 01 57 04 N002 Discharge Volume High Water Level 3 9 L CY 3 22 NCO2 CY 4 03 N001 Discharge' Volume High Water Level A l 01 57 34 N004 Discharge Volu.te High Water Level 0 ~ ~ ( .- CY 4 07 N004 Y CY 4 19 N003 Discharge Volume High Water Level C 02 28 06 0 80 TVA BFUP J 42 - J t \\ '~ '..~'i ~_

4; _;

.T _, , rvm,_

p. 7 3 7 9 3 _i;...w a a i..<, A ~'. + x -. i, s " ^.

s -o ; h, s EAST

WEST.

( T a. ~ 59 48 43 43 48 43 43 43 55 48 48 48 42 48 42 48 48 48 s [ 51 48 48 48 48 48 48 48 48 48 48 48 U 7 48 12 48 0 48 8 48 8 48' O 48 12 48 43 g 48 48 48 48 48 48 48 48 48 48 48 48 48 48 48 h 39 38 48 48 48 0 4.8 48 48 48 48 12 48 48 48 33 y si 35 48 48 48 48 48 1 48 48 48 48 48 48 48 48 48 48 fd t s-tj 31 48 48 0 48 48 48 8 48 8 48 48 48 0. 48 48

a l2 27 43' 48 48 48 48

48 48 48 48 48 48 48 48 48 48 3

4i 23 38 48 48 48 0 48 48 48 48' 48 0 48 48 48 38 19 48 48 48 48 48 48 48 48 48 48 48 48 48 48 48 15 48 12 48 0 48 8 48 8 48 0 48 12 48 2 11 . )' 48 48 48-48 48 ' 48 48 48 48 48 48 d. 07 g 48 48 48 42 48 42 48 48 48 q ,j 03 1 48 48 48 48 48 48 48

)

4 o il /,1 02 06 10 14 18 22 26 30 34 33 42 46 50 54 58 .t e 1800 .?j a Figure 2-1 Coltrol P.cd Positions Before Fir.st Manual Scrar 4 o ,E. pgy, g, I mre

e. m e

M 3 te. s** w y*y "%*= PI

9-

'I"' I 1 L_ m._

m,, 9 9 u. -. c.n : e_~ R c .- o.. .a.,e a w; c.. .m. w a.. .u.U p%:+2.J4 Mis- - _. -. _m. -u t. '.. g., ', ' 'h. (,) 00 EAST WEST. n

~

==. 59 40 4 36 42 0 0 l0 O Si 36 46 38 2 b 0 0 0 b W s l e { 51 36 34 0 36 42 10 0 0 0-0 0 O g 42 0 20 0 24 0 0, 0 0 0 0 0 0 47 c, d 43 12 4b 44 0 30 28 40 38 0 0 0 0 0 0 0 39 24 0 40 0 0 34 32 0 0 0 0 0 0 0 0 W ce k d 35 30 4 38 b 18 8 34 30 0 0 0 0 0 0 0 ..i 31 22 10 0 34 23 2 0 0 0 0 0 0 0 0 0 t W 27 42 36 30 14 26 "28 36 24 0 0 0 0 0 0 0 i 55 8 b 38 32 b 12' 36 2-b 0-0 0 0 0 0 W 19 34 42 34 20 l'0 28 22 26 0 0 0 0 0 0 0 4 15 40 0 36 0 u 2.=E= ~38 0 0* 0 0 0 0 0 0 1.; 11 26 6 14 40 40 24 0 0 0 0 0 07 O 40 2 0 0 'O O O O 4 W. C3 3 28 42 42 30 0 0 0 I ~ 3.,. 419 E P 02 06 to 14 18 22 26 30 34 33 42 45.... '. 50 54 ~53 ..J a 1800 a. I^ 4 h:- Fi.gure 2 2 tontrol Rod Positions Aft'er 'First Scram IT 1$ f 44 e I 4 i ' J(#'4,.". "t ~

- S vg n:.>g m e#% s:is>g+gtS Q4;ig 49 %.E +, w ;~W'fy. & s y2nrv:.p~+'- 2 y Ca 5 EAST WEST x: .= " 59 30 0 14 18 0 0 0 .g V. 55 25 42 28 0 0 0 0 0 0 s n" =. ..i; 51 12 22 b 28 34 b 0 0 0 0 0 47 3b 0 0 0 12 0 0 0 0 0 0 0 0 43 0 30 38 0 14 4 30 25 0 0 0 0 0 0 0 A i 39 8 0 30 0 0 25 14 0 0 0 0 0 0 0 0 35 14 ' 0 30 0 4 0 22 20 0 0 0 0 0 0 0 31 g 6 6 14 12 b b 0 0 0 0 0 0 0 0 .u mi 6 27 25 22 20 2 12' 12 ~ 15 2 0 0 0 0 0 0 0 s f 3 g 0 0 30 20 0 0 16 0 0 0 0 0 0 0 0 l0 h 19 10 26 24 0 0 S 6 4 0 0 0 0 0 0 y .. 15 18 b 18 b 28 b 0 0 0 0 0 0 0 = 'l ,I 1 H 11 12 0 0 32 30 14 0 0 0 0 0- .e L' C7 b 3b b b 0, 0 0 0 0 l?J ,1 w / 03 2 34-22 4 0 0 0 4 -i ,i s 02 06 10 14 18 22 26 30 31 33 42 46 50 ' 54 53 N 1800 ee e T { Figure 2-3 . Centrol Rod Positions After Second Scra: 8

m ^

m n - =,. _ n = -. n =,

s '_....c.. o.. w,..-- --- -....- m ~ M#in; - 4/*nMiV?WAS4Wel&MusdMES ! , m..: i ? D. C; \\ l EAST WEST --E 59 24 b 'b b 0 0 0 4 ~55 26 40 22 0 0 0 0 0 0 t SI b 16 b 24 30 b b 0 b 0 0 47 26 b b b 6 b 0 0 0 0 0 0 0 .gu l!!) 43 b 24 36 0 2 0 24 18 0 b b 0 0 0 0 f>4 .39 .2 b 24 0 b 20 0 0 0 0 0 0 0 0 0 gj N 35 6 0 26 0 0 0 14 16 0 0 0 0 0 0 0 H: e: 8 0 31 b b b 2 4 h 0 b b b 0 0 0 0 0 p 'fk.. 27 18 16 14 2 4 6 2 b 0 0 0 0 'O O O s b gj 23 0 0 26 12 0 0 2 0 0 0 0 0 0 0 0 ....r 19 0 16 18 0 0 0 2 2iO O O O O O 0, q". 9 I k.; 15 4 6 16 o o' 6 6 6 o 6 c. o .o 6 11 6 0 b 28 25 8 0 0 0 0 0 s a 07 Gl} 0 24 0 0 0 0 0 0-0 .J 03 b 30 lb .0 0 0 0 m D kb

ki

N:@ 02 06 10 14 18 22 2S 30 34 38 42 46 50 54 58 '. jjj 4V D fj 180 .Wp ~ Qn a h1 t L. }J.- - Figure 2-4 Centrol Rod Positions After Third Scra.m k L,i i-J.. (

r. ' '

G,s ).ej %LV t u,. ~,, . r,.. .w - 3 g :.,;- -- 1,-+;r 7.1r**3 &

W *"7?T

,.x; r fr..c1.W, .%.p., n,.:-------.-... - p r ~ -- - - ~~x**. ~ ~. ~~ n ;W".-~.= w - w.- =-cf h,& p c*T 2 t'.7.e' > -- =

. ;+,, w e. r :., a m;pmng.y riy+ppy ny;q z.,y,,y. .g.,,,;,,w g., y; , c x. gg x_p. - ~. - - - u. w 6 t -.,

p.,' ] *

.] h ~ ~ R I' hm,e e I '"( COUPUN G y $ PUD g 's SOTTCMOF b / } REACTOR VE13EL [I i'p:I [ ) ( je T1 I I ( 'h.f m A ?} l 6 7 J { (p([ s HOU$1NG ^ FIXED PISTON - g{ f ($ TOP Pl3 TON), j g i r,- !?n : e 'l l CCLLET FINGERS RETURN $PRING G 4i FC !9i )I ( i E / p Pt3TONTUSE [i CCt.LET PtSTON M j'y 3 i3 I e o4 4 p[. y INCEX TUBE j $) f / .) q r i5 CYUNCER N e

w_h If 1

y (j f[. .j t.' , ontVE FtSTon i: I / l' CRIVE IN!ERT UNg ( 3 "5 ( f ORIVE wlTHORAir s i. . uME e I .C ) i / k.1 ~ Y'E a = REACTOR REssuRE M ~ Figure 3-1 CCntrol Rod Drive t M m gg , m,;.,_,.,, 7

W 4 :; 7.5 p y g m i r q y p;25 igg.y g y -;.g y

3 3

34y 4,3, sym:3ggg,ggy.ggy_q., _g. 6.jp8 P. q ,7;l 7 48u VAC REACTOR 480 VAC RE ACIOR 4 i AtOV BO ARO 1 B a AtOV 00ARD 1 A e

p RPS A NOnMAL SUPPLY RPS 8 NOR..1AL SUPPLY s;,

a 'j ' T X1 M. kk g f.I Ql: Note: All relays and solenords G shown energized position (non-scram poltition) 4 Q flPS DUS *A* 320VAC 120 VAC RPS BUS *B* 1 -Manual Scram Relays b;[ D Subchannel One Relays c.-- 3 p p p CiiannelA [.' 3 ] j T j j .[.,l} Channg e .,y. : i<l. y, Subchannel Two Relays 5t u s.t p% a .n. 1 46 m +- - %g ri sNST CRDs\\ 3 d,, in 46 ^8a l /d j

CRDs r-Oi e

+ x u u 1-s A a .J N tr l ,,r .- tj uA*. l-4-- 6-4-

t

-n ( 4 3 2 1 veul 3CA Ahl 50LENOIO VALVE 1 2 3 4 .g m. Scram 3 4 Scram l Groups t . l' Groups m . l l 1 8 Scram l>4 Discharge.4 g -[ q I Inside dashed line e Volume - lga,y,,g scag,i typical of all 105 llCus I [ . z. c utaion OulLET E F SCRAM l ~fg W A T E R. Z ]) S ~ 4,.~'*:. ~ I CNLET .l l . u _._/,, ~ .a _._._ _ a. 1 i G t 'RCS Water t -t L ' s,} i.- r c h) ], 3 sl 6 m .: e 1

as-~- + l :[ h h ,.,,s y '.b v East West p. L" "9 2 1 3 4 2 1 3 .n Y1 63 55 3 4 2 1' 3 4 2 1 3 s EQ 2., 51 1 3 4 2 1 3 4 2 1 3 *4 $q

4. -

47 1 3 4 2 1 3 4 2 1 3 4' .2 1 h;1 d A 43 2 1 3 4 2 1 3 4 2 1 3 4 2 1 3 .+>4 .C ' .k' 39 4 2 1 3 4 2 1 3 4 2 1 3 4. 2 '1

b. '..

D 35 4 2 1 3 4 l 2 1 3 4 2. 1 3 4 2 1 'S* 31 7 1 3- .4 2, 1 3 4 2 1 3 4 2 1 3 'Lf %u 27 2 1 3 4 2 1 3 4 2 1 3 4 2 1 3 y A.g 23 4 2 1 3 4 2 1 3 4 2 1 3 4 2 1 -y ,Sgj 19 4 2 1 3 4 2 1' 3 4 2 1 3 4 2 1 4 h 5.- 1 3 4 2 1 3 4 2 1 3 4 2 1 m . 15 .Nw . 11 3 4 2 1 3 4 2 1 3 4 2 g b7 3 4 2 '1 3 4-2 1 3 g -4 it 03 g 4 2 1 3 4 2 1

i. :.0

.C r. 02 06 10 14 18 22 26-- 30 34 38 42 46 50 54 58. y I ed.s

d

+ 1800 m 'M e[] ~ q a& a s Figure 3-3 Centrol Rod Scr'am Group Assignment 4 i e J., c. 4 4 .N Q q %c~

~

m. sa.. - g e 9 M*

D h

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e. er g

~ s y; *..p.);

pM M TrASMSf.dMil5THiMIMMh75fFgky/53A7,g g g g g;ig.jff gjg gg g g ggy g g y:p-4 e y. a.: ,.. t y. ..s.

f

.n s 35A s Ma wa s.t - 9 DC DC . PA Q e, e d' - 33 P$sG e , CONI. + gy d aoo 04

j t 1}}

_ co i -y. - j. g g -y,{- s.na 5 338 DC " 33pA L, yg r$3G g.E']. + {% D _ ff{*I Cons. q e Vlael APSA AP54 ( a .3 g fi Ins (de daslied line 7 3 ,,,f,hy,,>I. 3. = > s uA i typical of all 185 llCus- -. --. - -.- ,,u ~ ,l l ~l"I m '.50.tr sC.aau sest as Ano ourtst Nos Also sVatvtsIF. O f AN4 A.41o Clolt.tH.M4 A8so I 34& 5 ?-1. 3 tfsCAAu AC AC ie y]. l ona.es vaavas.b.lf VtNI A40 o.5 CHARGE Vot nese arsA r n A. s. C ~j

fq.

g .c. Ana A.a soorts sog guoiog-*OV p1 - .n vAtvas S ^ - f (*'~.- .-f 50V l Caaac. a g vs=r ve=r waian .g o s.C on. n - f-398 SCTSSR b c... s."= m"s.mo ' = l e ss= . t o.sCuanGL woguus jf.'

4 A.as j

r '.u l eg

=.,

. n. ......n. ....u. ..A m_ , N.j , %) se. Scram .e ,nA ,,...n. ......n. ouiter .a e O e e g .assinuk.f MI 3 Cno Votua.( 5-Me l .m....as. .....s. ' dj ; un f90 .4 SC AA.4 01HE A e. s 34 .. s a....u. ... ma a..n. 8 882 H30 )' n .g oultti a. J SCna ?{

1 g

vALwas sna .e........ } unne .=taf .;l, g, p v g I p.~. in4 i.e....u. l i y g 99. :-@ 7-I suA i. ...u. s c m .a .i c I ,^ = => A 'M ? ..n. ...u. I 3* l CH ANGtNG Q)' * * * * * * *

p'.3ts I g..s..

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$1

,a l-----,- o i s.UC or( N C.oll o j i s RCS O 74 a ...=e=.. q tM '8 .3 'e" q .n.a n. i-. a n.* ' A. Water a e..~.......... u... u ~ ~ d N -T WQ. l,j 8 g fjy I .j es ' :,8 l s y n,, 9 3 p() s. u h l *'b*. j f:,'1 3 g .c' k e 3 n en N.k

."_. _ _b.b hfdih ' _ N-_.',' m ' "_..._. _. _-kQ.1 _.Y a $i.,_^ a_3.'2N ^ .m ' s,.J ;. 1,.i.. L..L.: i wm' _u. ' 1 m_ .,m. - LS - ,m 1 .s ~ guaca5uraay ulEJO a=ntoA n aos S-t aanny o, s .a ~ l ?l

.?

m ) 8 g-v m e t HCU L. Ka v F.' N f s .2 c-m _._ i m HOU = w ( e D ET Ik N "L r U o P HOU HCU N y F' x J \\ b .c 1 .( e c cn g,, s I F ( \\ ee ne= E'.- t 3,- O t N c3 m ,s. t E ,o c 1 e2 o o, Es c ow, g <m th i e Level Switches 2 onn 4 l g 4 p-

. J.

r. s d..a w o 5 a Su f,h e .A C A mm eo e f,a .O O ~5 r= n Da :m-cD n oe ss ,e r f:. o ,c= n bC w li r. 3 --* G m cw q 4* T =- z e .t 3, .~ a r pq <-Other Drains n ..a =g .~ 7-HCU Discharge ? i f HCU 'A - v Line* HCU n a m. Y w' ,$.Y i C d{ =O P m 6 s}

==m

Q E.

MCU c. 3 D 2 *C a .-i f,D o-HCU m w -: m [ M E

ty e, o f*

3% &* 2 - C.

~1 c.d;,

C .a ..q-a u.. ' ' * ' ~ - 4 rp -, .,r .,n ~ ~Yd"- g z

x- ~ - -. - - - ,_.._-_m. . y m..m - 0.;., ., o .,: t. T y; o, G . g: n vi,+ .3, .g N-y r Table 5-1 Scrarn Discharge Volume Drain Time and Total Positions Inserted [ Drain Tima from.- Total Positions Positions Scram No. Previous Scrap, Sec. Inserted Per CR0 2 93 956 ~ 10.2 3 53 402 4.3 i r ./ t 4 160 300 ~ 8.6 4 p 3 g 4.+ y l '+1 3 k n/g h

. /

e l-1

1

f

+ h4 1 i

1 b"

a k' u' 6 a D t,. u 'Il

's 3

i 4 b i .p 1 .V M. j[ 52 4

/.g p..

.ur h- ,.n. N i -k 'ed g ,e t. - e ew M = ' m gl>5***=ep,

.,e* eP -y%g e me w y w*
f

,,: *~ ~.v ;-e n g rc p ~:->,.. s r e g _M:m 7 r- .e. 3.m(nv<,wn,$ *sA ^k yU >p' * > [ ' n-,' _!_ l ?. ,- mr or:- ~~w : M eMiM5'lh i j A_f&# .O .~,

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ML20027A622

Text

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