ML20215H509
| ML20215H509 | |
| Person / Time | |
|---|---|
| Site: | Surry, 05000000 |
| Issue date: | 08/16/1972 |
| From: | Ragone S VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Giambusso A US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20215H426 | List: |
| References | |
| FOIA-87-20 4545, NUDOCS 8706240130 | |
| Download: ML20215H509 (32) | |
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y _j' Vepcoa gWl.*~#' ** VI - 4 RfCT Asc AND AQWim CcMPANY. AICHMONO. ve4GINIA H209 E . ....
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// August 16, 1972 g
(a Mr. A. Giambusso Deputy Director for Reactor Projects Directorate of Licensing U. S. Atomic Energy Commission Docket No. 50-280 Washington, D. C. 20545
Dear Mr. Giambusso:
In response to your letter of August 4,1972 fequesting information , concerning an incident which occurred at the Surry Power Station on July 27, - 1972, I am enclosing ten copies of a report entitled " Virginia Electric and Power Company, Surry Power Station, Unit 1, Operating Incident of July 27, 1972," l which fully describes the incident and subsequent deter,ninations which have ] been made.
- As a result of the publicity the incident has received, I would like j to review the actions of the Company f ollowing the incident. The primary concerns 1
of the ccmpany immediately following the incident werc: 1) to ascertain that the l incident did not create a hazard to the health *M aafety of station personnel
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1 or to the general public, 2) to ensure that the injured men received prompt I medical attent. ion, and 3) to see to the we.1f are of families of the injured men. Inunediately af ter the inciden': described in detail in the Report, the f Station operating personnel evaluated the status of Unit 1 and determined that it was in a safe condition. As a precaution they initiated a safety injection to ensure that the safe condition was maintained. At the same time, the injured men were being cared for and medical assietance and ambulances were sunnoned. These initial ef forts, in additiot. to the normal operation of the Station, occupied the operating staff and it wa.s not until later in the day of , the incident that an investigation to determine ti.s causes of the incident could be begun. Because there had been no release of radioactivity and no AEC regulations or Technical Specifications had been violated, there did not appear to be any requirement to insnediately inform the Atomic Eigrgy Commission. Nevertheless n.
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the Office of the Directorate of Reghlatorf-Op'e'rttids-Reg'id II was-nottfied-by telephone later in the day of t'ho, in.cident and given the. fliJr th l
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At approximately noon on the day of the incident the Company received j ' an inquiry f rom United Press International (UPI) requesting information. Our policy is to release detailed information only after a complete investigation and report has been made. In the interest of providing accurate and reliable information to the general public concerning this incident, a brief news release containing the then-available information was furnished to UPI about seven (7) hours af ter the incident. It should be noted that the news release (a copy of _ f0TM-??.10 8706240130 870619 4545 %. PDR FOIA ZWELLI NG87-20 PDR
vincua nscraic ano powin courany to U. S. Atomic Energy Ccemiss ion saattwo.2 which is attached hereto), although brief, was accurate and has been confirmed by the investigation and Report. We have made a conscientious of fort to handle an unfortunate incident in a responsible manner. At the outset, we alleviated the concern for the safety ( of station personnel and the general public and cared for the injured men and i I their f amilf es. Once these immediate needs were satisfied we focused our attention on the cause of the incident and means to prevent recurrence. These have been l documented in the enclosed Report which has been reviewed by our Station and I System Nuclear Safety and Operating Committees. The Report is responsive to those I issues outlined in your letter of August 4,1972. )
-1 I would like to reiterate the finding of the esport that there was no j release of radioactivity as a result of the incident and at no time was there - '
any hazard to the health and safety of the public. "l If you require any additional information or have any comments on the. attached report, please do not hesitate to contact us. Very truly yours, . Stanley ne Vice President Attachments O l
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g g . .. ._ PRE $3 INQUIRY . PUBLIC RELATIONS DEPARTMENT tacW oaf ( Fred McNeese 4 7/27/72 United Press International ,,
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644-0701 12:55 p.m. l o s v a.s s Questions: Understand men involved in accident at Surry. Was it nuclear-related? . any fatalities? identification? what happened? P'roposed reply: Two Vepco employees were injured Thursday when a steam vent , system malfunctioned near the area where they were working at Surry Power . Station. Both men sustained burns from steam escaping from the vent.
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They were identified as William L. Van Duyn, 30, of Carrollton, and Roger J. - i Wood, 28, of Newport News. They are being treated at Medical College of Virginia. I l Early on Wednesday evening, the power station had been taken out of service l . I after operating at the 35 per. cent p.over level during the recent peak-load l period for standard pre-operational tests at zero power level. The steam system was slovly being cooled down. The two instrument technicians were working in the area when the accident occurred at about 7:15 a.m. Thursday. 4 1 (In conversation, I stressed the men wer outside the reactor containment and no radiation was involved.) 4 o c a o bl4: A.S.A.P. Aillo N ED
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VIRGINI A ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT I ' OPERATING INC10ENT OF JULY 27, 1972 b ( - l t 2
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- l. In t roduc t ion The Virginia Electric and Power Company was issued Facility Operat ing 1.icense No. OPR-32 on M.y 25,1972, by the Atomic Energy Commission nermi t ting opera t ion of Surry Uni t I which is designed for a warranted ouput of 2441 megawatts thermal (MVC) wi th an equlvalent warranted gross output of $22.6 megawatts electric (MWe).
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Unit I has been producing electrical energy since July 4, 1972, ~ when i t was firs t synchronized to the Vepco electrical transmission sys tem. Startup operations hans proceeded up to approximately 25% of warranted power (220 MWe) in accordance wi th approved operaeing and testing procedures and the Technical Specifications. ll. Description of the incident
- A norma l shu t down o f Un i t I began on the evening of July 26, 1972,
. In preparation for zero power physics tes ting and maintenance. At 8:48 p.m. on July 26, 1972, the reactor was subcritical with reactor coolant loop temperature and pressure being maintained between 530*F and 340*F and 2235 psig. The main steam temperature was being maintained in this .
same range and the pressure was apprd'ximarei $hh -
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~*W coolant pumps were in operat ton anV deE4y.ticat was being removed ~br us tngT two (2) condenser s teae dump valves , TCV-MS-1078 6 1088. Six (6) of the eight (8) conder ser s team dump valves were out of service (Fig. 1). Bank "A" consisting of fouc (4) s tear. Jung val ves TCV-F.5-105A, 100A, 107A i
and .as ma sally isolated due to 5:ea- leakage f rom t9e valve flange e on TCV MS-1334 of one of the four (4) valves. (Condesser steam dump
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i l 2 i valves cannot be individually isolated. The manua isolat io va ? ve. In the condenser steam dump lines are located such thet a l l fou r- (6 valveA in a bank must be isolated.) The two (2) o;4rable va!<es. l TCV-MS-1058 and 1068, which were scheduled for male:enance, were { stuck closed at the exis ting temperatures and press.res associatend .s a ; i hot shutdown condition. Powe r ou t put from the electrical generator prior to the shutdown was li -i ted to 220 MWe pending maiatena,ce r e:r, -
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to these condenser durnp valves. ~
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At 6:07 a.m. , July 27, 1972, reactor coolant pcp 8 was sto .cee m . l rede:e the amount of purnp heat input to Reactor Coolant Systen atre pms l I A and C were lef t in operation /to maintain ' hot shutoown conditforts. .c l approximately 6:40 a.m., the four (4) bank "8" unisolated condenser smain dump valves were belns tagged out of service in preparation for in.ainfance j and the cont rol room opera tor, Mr. R. l.. Heads, ar texted to use t-ne atmospheric s team dump valves RV-MS-10l A, 1018 and 1:lC, to renove de: 2y heat. None of the three atmospheric steam dump valves would operar:e i R. J. Wood, and W. I., Van Ouyn , ins t rueen t techn i cia-s, we re avi s,ed :- the Inoperability of these valves by the shi f t supervisor, Mr. C. wt. Brooking, and they proceer'ed to determine why the .atzspheric stear:n - . _ ____
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. ..:: 7 _. ';- 5 dump valves would not operate. Initially, Mr. R. 3. Vood checked che ' ~
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controllers in the relay room and found no' r4al func t ic s there I-dica't ; ' that the problem was at the valves in the upper leve' ef the sa'egt 4 r. building. 1 Mr. W. L. Van Du, aos ac:cmpanie:
- to tl' e s a 'e;.a rds b.s i ':i ;:
n by J. l.. Greenwood, the as s is tant cont rol room ope ra t: . Mr. G "e e m-o:.:
p kd had beer' irs;rheted by Mr. Brooking, the shi f t supez vipr. throu;9 s.' 1 control roo ope rator, Mr. R. L. Meads , to c lose the ra1ua l i so la t ic-valve 1.~5- i for "A" a tmosphe ric s team dur p valve for the ins t rument techniel ns ark., and to open the manual isolation valve t-MS-Ild in the de:ay heat release line so tha t the decay heat release salve HCV-MS-;C4 y [ could be use:. Fig. Ii & III. While this work was in :ros ess, R. J. Wood, the second 1 stes.cnent technician had reached the upper level rf the. h s $4feguards b; Iding.
'" A f te r comp l e t i .19 the va l v i ng, -t'r . 3 ree nwood in fo rme .
'Mr. Heads fecr the lower ,
level safeguards building inter:orr that the valve Iineup hac bem compteted. At aporo, ma tely 7:
- 00. a.m. , Mr. Heads , the cont ro l race. opera tor, 3
L et tempted to c:en 7.he-decayc heat .celease valve by impcsir.; a 10% dem
} signal on the ontrol room benenboard controller, % At this t inne the assistant con 1 room operator was still on the intercoe alth the control room opera tor . ;
s The assis tant control room opera tor could 90% i determine f frorn the noise 'evel in the safeguards building whethe o nc.: the valve had opened so 9e taid the control room operator that he going war outside the building te ascertain whether the decay heat release ulve was releasing s team h'< ots.arv49 the 9ent stack. Mr. Green.c:d wnt outside ,the building and noted tha: h '
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sm steam coHns ow of the vent . He started to re enter the, pfeguards-j ~~.' ~ (~
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building and wh'~e he was stYll outside'h'e heard a rear, Ic.:se :: up, at:d saw stea c: 'ng out of the vent. He re-enterec the sa eg.ards i 1 i t m
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4 building and he tried to climb the stairs to the urper level but was - unable to do so because of the s team. He lef t the safeguards building and called Mr. Meads from the chemical addition tani intercom, near the refueling water storage tank and informed him of the steam in j l the safeguards area, and requested ass is tance because he knew Mr. Wood I and Mr. Van Duyn were still in the safeguards building and may have been injured. , l
-l Mr. Heads, the control room operator, set the decay heat release j valve controller to zero percent open. The shif t supervisor lef t the i
control room and made his way to the safeguards building. As Mr. Brooking l opened the door to the safeguards building he heard someone outside the I building call out that Averyone was~out. He then went into the j i administration building and founo that the two instrument technicians j l were being cared for there by operators f rom the oncoming shi f t. Mr. { Brooking then returned to the control room to direct control room activities following the incident. The Isle of Wig t Rescue Squad and i j l l a nurse from the Uni t 2 construction forces were called for assistance. l l The nurse arrived in the administration building to render medical assistance at approximately 7: 10 a.m.
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x..~~ ? _ .. .Q- - - The Reactor Coolant System wa,s c:c,* ling down fron the. release . .. - of . . ~.=Q)' ;l steam. The operators on duty were of the coinion that there appeared l to be a steam line break in tne safeg.a-ds building, and it was decided that the Safe ty in,*ection S / stem shc.'c be ini t ia tet in accordance wito approved e+ergency precedures. T9e Safety injec .ic- System f-1ctionec _ properly. Two of the three charging / safety injectio* pumps were stop ed
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af ter 5 mlnytes ane the one remaining pump was allowed to operate in the safety injectio, mode for an cJditional 10 minutes. 1 i The reactor was shutdown a..J in a safe condition at all times I prior to, during, or following the incident, and there was no release of radioactive material. An initial investigation was made and the indication was that tne _ ,. decay beat release salve exhaus t pipe had backed out of its vent s leeve I l (Fig. I V 6 V) in the ceiling of the safeguards building, releasing steam "l i into the building instead of out through the vent and that there was in fact no steam line pipe rupture. The two injured tiEhnicisns we're individually transported approxitnately 70 mlles by ambulance to the south energency room of the Medical College l l of Virginia In Richnond. Mr. Van Duyn was admi tted to the hospi tal at approximately 8: 15 a.m. and Mr. Wood was admi tted shortly therca f ter. Both patients were taken immediately to the emergency room for medical treatment. Mr. Wood succumbed four days later on July 31s t , and Mr. Van Duyn succumbed five days af ter the accident on August ist, bo th f rom s team burn related causes.
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Ill. Identification Taken of the Causes of .the Malfunctions and Correc;lve Measures , . ._ - _"J e
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A. Condenser Steam Dump Valves (Figure I)
- 1. Va l s e *:V-MS-1CS A wh ic' aas leuk i r g s tea- f ro., t he t
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. 1 6-i flange has been inspec ted, and t he cause c# :*e 'eakage ., !
was found to t;e a f aul ty gaske t w9ich has : replacem by the stat ion ma i n ter.ance grcuo .
- 2. The two (2) valves TCV-MS-1058 6 1068 o F t e .;n i s,ol a t e d -
bank which did not operste at sys t:em preess. t haw-e been. i inspected. The angle of' tne valve: seats o# re four (411 valves which enter the lower level of the c: :anser ,
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TCV-MS-105A and 105B, TCV-MS-106A & 106E hat:ecn ") modi fied to reduce binding, and valve sperirg : ens ion has been readjusted on all eight valves to re current i reconynendations of, the valve manu f acturerr. . 1
- 8. Atmospheric Steam Dump Valves j
- 1. The "A" atmospheric steam dump va l ve RV-P'.5-1 A wa 5 last isolated on July 1 7, 1972, due to mivo r.ean ;
I leakage past the valve seat and had been ve:. ed to f 4 service at approximately 6:30 a.m. on July ;*-. This ] valve did not open when activated from the .. : ol roora. ,
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( Subsequent invest igations ' revealed" thit t ne r *:t~rit c/.. . ~ .~.,. @@ ) l l pneumatic converter did not hold I t5 orig"na l i t:, f a t fo# *
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which was las t pe rforr.ed on Februa r y 11 " $'. Tne l atmosphe ric s team dump . a he rar.u f a c turer 52 :<! . ed tea.- l l temperatures in excess cf i30'r for exten e. oc. raf i effect the operation of the atr espheric s .ea .. p la;ve electro-pneunatic converter. I l 1
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- 2. The "B" anc "r" atmospheric s team dump valves RV-MS-1018 l
and 10lc failed to opera:e because their feedback linkage to the positioners had become disengaged in the (1me j interval between July 21 1972, and the time of the ) { incident on July 27. 1972 The valve positioners have I been relocated and IInkages extended to provide more positive contact with the valve stem guide. All three i l 1 valves have been recalibrated in accordance with established __- ) l Ins trumentation procedures and linkage t8tainer clips _ have been installed. P j C. Decay Heat Release Sys tem i I 1. Decay heit release vatve 4 5 k i' ' The decay heat release valve was installed in accordance ( with approved engineering drawings. Following hot functional operations in February 1972, this valve was inspected internally. There nad been indications that the stem was binding. The valve was installed with the stem in a horizontal position without a support of the I valve actuator. ( I t was dete rmined _j jat. h3 tJme. .that g. _;. ,;'j7 } V-I
.. .u p Ve stem was stick.ing in C9e bushing during.the hot -
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# W functional tests as evidence $ by wear on the stem. The stem was replaced following rot functionals and tests were cordacted v,hich demoas t ated ;;rocer valve coerat ica I cold shvtdonn cc.ditions. -
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8 I l During the incident on July 27th, the valve e.ay not have 1 l responded to the 10% control signal as the signal was manually applied by the control room operator. It has not been established if smooth operation was evicent at the time of the incident. However, in light of this previous his tory of required maintenance on this valve, f and the manufacturer's current suggestion that supports be added to the valve actuator to prevent stem _ binding, it is concluded that th em operation of the valve may have ~ been erratic and that s team flow may have exceeded that normally permitted by a 10% control signal. Tne resulting force to the decay heat release line may have exceeded the. design force of 7780 lbs calculated for maximum steady state steam flow of 330,000 lb/ hour at 1085 psis and 55 P F. The control system of the decay heat release valve is designed so that if a remote manual signal is inposed on the electro pneumatic converter, the posi tioner output pressure will continue to increase until the unbalanced control signal is satisfied by feedback, from actual
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valve movement,
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n to' provide '
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Inspections conducted on this valve (cIlowing the incident j
. under cold shutdown condi tions re vealea proper operat ion with no indication of erratic valve movement. The decay heat i release valve was completely dissassembled and inspected
{ on August II, 1972, with no indications of s tem wear.
- 2. The decay heat release line The decay heat release valve exhaus t pipe was designed ~
to be Installed 4 inches into the s leeve. During the i incident it backed out of the vent sleeve as shown in ! Fig. 'l V and V. Fig. VI shows the configuration of the decay heat release valve i!xhaust pipe and sleeve prior to the incident. The decay heat release system had been operated on approximately 20 occasions prior j 1 to the incident without separation of the decay heat release line from the vent sleeve, i The post incident investigation revealed the release line l { extended into the vent sleeve approximately 4 1/8 inches l af ter the line had been returned to its operating position.
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lt has been caIcuiated. thai pipc de5,4p. . scMei,;eads'e'd by .thi
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force exerted from,ma' imum. des ign s te a dy s ta t e 's't eam. H ow.
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J._ 3r W r.onditions, would be in the downward direction i I/2 inches.
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4 This, deflection combined with the 2 3/8 inch movement
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permltted by the decay heat release line support hanger, enabled the line to move downward a total of 3 7/8 inches to a position where the pipe extendad into the vent sleeve approximately 1/8 to l/4 of an inch. The design of this line conforms to the accepted piping
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code (ASME USA Standard Code USAS 631.1 (1955)) for pressure piping and was installed in accordance with
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the decay neat release valve, piping, or structural failure ' of its restraints. The piping support system for the decay heat release line was designed to accorrrnodate deflections resulting from controlled steady state q l discharge of steam flow permitted by the decay heat ; i release valve. The s teady s tate thrus t resul ting from the discharge of steam at the maximum design conditio9s specified (330,000 lbs/hr at 1,085 psig and 556*F) is approximately 7,780 lb. This desig, force would result in a depression of the pipe suoport hanger to the limit of its travel and a deflection of the piping at the discharge end of the decay-heat celeasNkrEdf'J-7/87 '
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inches, but such total.. movement weuhl not cause dis.. W;
- - .. _.J _ *3'W engagement of the exhaust pipe f rom the vent s leeve.
In order for the exhaust pipe to be disengaged from the exhaust sleeve a force of 8,iOO 35. must be im osed on the pipe. _ 4 1 4 l
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A thrust loading exceeding 8,900 lbs. may have been imposed on the pipe as a result of the dy amic load 1 i imposed by s team flow resulting from rapid opening of of the decay heat celease valve to 1 t s full open pos i t lon. 1 { This. force Wuld result in a depression of the pipe ! support hanger to the linit of its travel and a deflection 1 of the piping at the discharge end of the decay heat release line far enough to cause disengagenent. 3 Modi ficat ions to the decay hea t release system The decay heat . release system has been modified as shown in Figures Vil and Vf 1.L. i Four (4) 1/2 inch thick steel gusset plates have been welded at 90* angles connecting i l the decay heat release exhaust pipe to the vent sleeve l ( to prevent separation of this pipe from the vent s leeve. k Additional supports for the decay heat release valve { actuator have been installed using a clevis spring hanger supperted by a 1/2 inch rod connection to the safeguards
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I building ceiling by a 6 inch square steel plate. A stress and flexibility analysis has been cone cted on
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the inodified decay beat release _ sys tem which shows
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i that the modified' des ign wil~i ~withs tand the steady state T. _. ~ %-e thrust resulting from the discharge of steam at maximsm
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* :fes'; c and'tio-'e 6 f ed wi th comb ined dynamic loaoings
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' n s u * :a +e o. t. c@ .>e l* f the valve to the full open pos': ion, D. A::es s :: 54feeguardS2 -
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- JSnc s tal fwell exis ted for JCces s ..
t: anc: f-.r te+e tg.e r' e - cof the safeguarcs building. An ac:i t C ori doc e has Stet .veed to the oppos ite end of tne utae r leat! of the Wafe r.J s building, in addition, an as t r ont ener ger<v. . le: nxi t has been installed from ear. i.evel cf c,re saflee:4.,.: building. Fig. 111. E. Verti lait i:- Ter: era:u ts up :o 1722* H *ee upper levei safeguards buiIding ! Mac teer, enerve::: wrir N:s e range operat ion wi th the safeguar-ds :wilclieg eva c. covers in place resulting in rmin'rua l ve .ilar' an in 1,.sreer levels of the building. SecaAse the safe'caa c5
* :wNg roof vest. covers. were removed . . ~
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W i%. ,. . crn .taly 10- :o i : : cf. - .
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. ion and because the uni t had
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..c- prfor to the incident, it i s uC i ka li : a : t :.:* a * - .
the safeguards building at t ~e :'me of : e : :r - ' 's . nve 150*F. 1 l' t t
The design of the safeguards building ventilation system has
'been ravlewed and modifications wili be made to improve the ventilat ion for be t ter terpera ture cont rol. This will consist of replacement of the existing 13,000 cubic feet per minute propeller fan with a 16,000 cubic feet per minute axlai fan
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equipped with ducts to enaus t air from the upper levels of I the safeguards building. Until such time as this permanent j rrodification can be made temporary ventilation equipment has been placed in service to reduce temperature levels during ~
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power operation by increasing air flow circulation usin.g two additional 17,000 cubic feet per minute fans. IV. Consequences of injection of Cold Water into the Reactor Coolant System i i A. Descriptlon i Because the operators recognized that the reactor coolant system was cooling down due to the release of steem during the incident and because it then appeared that there might ( have been a steam line break in the safeguards building, j the safety injection system was manually initiated at the
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direction of the shif t-supervJsor,Mf htt. 37' o oking. dn ~ ]'
'* g accordance with estab'l'ished gera ting procedure's ' automat te . . . 7J._. ...m
-M.M Initiation had been Intentionally blocked prior to the incident to preclude the possibility of inadvertent safety injection ini tia;icn as the 'eactor was beir.g maintained in a hot shu td ,<. condi t ion to tha t low power physics tes t -
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could be condu::ec. As the safety injection sys tem came : into opera t ior, a l l three charging / safety injection pumps started, their fuction transferred to the refueling water s torage tank a-2 beir discharge path Opened through the boron injec t io- ta v. to the three cold legs of the reactor coolant system. l The inject ion f'ow to the reactor coolant system with three a injection pumps in operation was calculated to be 390 gallons per minute (g pe. based on the pump characteristics and the sys tem res is tarcs. At this flow rate the boron inject ion I tank contents e swept out in two to three minutes and resulted in an iterease in the reactor coolant boron concentration o8 about 350 ppm. Tw of the three injection pumps were stop:vd af ter five minutes and the one remaining pump operated i' the safety injection mode for an additional ten minutes. V:1 cme pump in operation the flow was i calculated to be 215 gpin from the refueling water storage l tank to the rear:3r coolant system and the refueling water resul ted in addi:fonal boration of the reactor coolant.
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. .. p The total quant *., of flul~d pumped to the reactor' "coolarrt .- c. _ __ -M,b system during t e time the injection pumps were in operation was apprcxir ,a te' 4100 gallona. Tnis quantity of coolant is consis teat w : th e volume of 4300 galions recuired to make t
up for the s ri , Of the reactor coolant due to the temperature
i decrease plus the volume of coolant forced into the pressurizer which caused the le ve l te rise from about i ( 22% to 74 indica ted. l I The reactor coolent system transient which resulted fran the steam withdrawal is described by figare IX. The temperature decreased rapidly from 535'F to Si2'F where it was stabilized by operator action and remained at Sil'F ~ for one hour. The pressure was held at t5e normal operating - pressure due to the action of the injection pcmps in maung up for the coolant shrinkage with temperature and causirs the pressurizer level to rise. The pressurizer spray was In automatic operation and maintained the proper operating e . .. pressure as the pressurlier steam bubble was compressed. The injection flow from the safety injection pumps entered the reactor coolant system through three 6 inch connectio}; one on each cold leg. The flow was uniformly dis tributed to the three connectlons and was therefore 130 gpm per loop. This injection water was then mixed with approximately 97,003 gpm of reactor coolant at about 535'F in each of the two active loops befora passing to the'feac3ar;vessek and yltt:
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about 22,000 gpm of tiackflpw ir the one inac t ive loop, . , . 7 -. .Q M Therefore,15e tempe rature of the injection water had a negligible effect on the temperature of tne reactor coolar.t 1 1
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co u r'* g t he t ra n*.-s ie 1.9mpa re d t o t he coo l down o f t he coo l a n t I Fron 535's to E .::' a: sed by the re lease of s ee.im f rom the s teart gere ratorT- + .
)
i T ic iajeccion ocX.zze 65 inch, Sch, 160) on the reactor I cala t cold leg;c 5 c iinitially at 535'F. Immed ia te ly i ucson starc'ing of tg4hety injection system the fluid in l t r.e :*;ing and he-or >.tection tank was swept through the ' -i -) nc.22:es .
'Th i s f' J u n e at a temperature of about 165'F, t tne it:'ma l temper %a.* Jff this hea t t raced sys tem. After I
ttne c:r' ten =s of erwaram injection tank were swept through the Is.~ec3.T.on no2=ez a,_he temperature of the fluid decreased to atst 7CF whic:cte trhe ambient ternperature of tne piping j anc c:etent.s in crw:a.dling. This transient is described on l Figure X. ' Note C N -t linjection flow was terminated before the cer*en ts of C.-We =~4] system was completely supt out anc the efo"e the ~; con awa ter from the refuellng wa ter s torage tame. (M'F) die ncEt W :the reactor coolant sys tem. This a cooler =ater was I-e:ef us nant in the piping system af ter the l sys cem flow ms suteam. , l
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i steam generators, pressurizer, reactor coolant pumps, reac tor colant ciping, and safety injection nozzles has been evaluated. ' The s tress and change in cumulative fatigue life was .da;termined to be negligible for the reactor vessel, steam generators, pressurizer, reactor coolant pumps, and reactor coolant piping. The safety injes: tion transient experienced is less severe than - several transients used in the design ~ analyses of these - conponen ts . The safety inject ion transient had no adverse affect on the s tructural Integri ty of these components and did not irupair their future use. i
. The thirmal ihock e~xperienced by the safety injection nozzles Is less severe than that experienced by the charging nozzle at the charging Initiation. During charging initiation the i water temperature is assumed to drop 490'F in 1.0 seconds t
whereas the temperature dropped 370*F in approximately 12 i i seconds during this safety injection. i f Evaluation of the safety injection transient has shown that at i
)
i least thirty safety. -injectionsipf,. hdmilar nature ptt be'
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., . ,e j experienced before the effect of this transient ori.(otal, fatigde li fe becomes ignificant$
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-IS-l V. ConclusLon
. it is concluded that tne occay heat release piping deflected beyond I the design lisilts of the piping support system. The force which caused this deflection was crea:ed by stearn flow passing through the decay heat release line. This line normally handles s teady s tate s team flow in i
the range of 181,000 lb/hr, hoi.e ve r , it is designed to handle the maximum --l Steady state steam flow of 330,000 l b/hr a t 1085psigand556'F. The - thrust loading resulting from fc-ces that were imparted to the pipe l has been calculated to be 7780 its. for this maximum design s teady s tate l flow conditions, \ it has been de:ernined that a thrust of approximately 8900 lbs. Imparted to the' decay heat release pipe as originally designed w . .. would cause the pipe to disengage from the vent sleeve. A force of this magnitude or greater may have res41ted from the sudden opening of the decay heat release valve caus ing the pipe to deflect and back out of the vent sleeve thereby releasing stean into the safeguards building. This decay heat release piping support system was not designed for dynamic loads that may have been imparted to this pipe as a result of the operation of the decay heat release valve.
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g-At no t ime was the heal,th .act safe ty. of the public 'e'ndangered as, a ,__ -w. ;, .. ,4 result of this incident. There wu no release of radioactivi ty prior to, during, or following the incident aid the reac tor was at all times in a safe condit:en. ii 1
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O FIGURE IV DECAY EAT RELEASE CONTROL VALVE PIPIN3 VENT N DURING SWT00VN CONDITIONS FOLLOWING INCIDENT 4 IO
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