ML20235U339

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Revised Emergency Plan Implementing Procedures,Including Rev 1.00 to 1-ES-3.1, Post-STGR Cooldown Using Backfill (W/No Attachments) & Rev 1.00 to 1-EP-0, Reactor Trip or Safety Injection (W/No Attachments). Related Info Encl
ML20235U339
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 06/12/1987
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20235U321 List:
References
FOIA-87-461 PROC-870612, NUDOCS 8710140042
Download: ML20235U339 (49)
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{{#Wiki_filter:- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ U-Revision No: 0 Date: 7/20/87 1 1 I* CHRON0!ACY STEAM GENERATOR TURE RUPTURE EVENT NORTE ANNA POWER STATION WIT 1 _ JULY 15. 1987 l Approved By: / M. L. Bowling / Assistant Statio ager Approved By: E. W. Harre11 V I 4 Station Manager _egh Approved Byt W. L. Stewart Recovery Manager 87101i40042 871000 Fo7_n-g7-94f hgA -461 PDR dG M.'t P[ I jrey MP Eeyca"of",*"'w dl b e i" ' r f * **+) ,w guwca I. p ) difs Q

j - n. e s Lf T ] q. s j w @? CHR M STEAN GENERATOR TUBE RUPTURE ETErr NORTE AM A POWER STATION UNIT 1 JULY 15, 1987 ^ ' To ' support the; investigation of the ' Steam Generator Tube Rupture the - followingschronology.was-. reconstructed fros the. print outs'of the alarm typewriter attached to the Control Room P-250 process computer, the Sequence of Events Recorder (Dranets) driven' by. the Hathaway annunciator system, the data- - printouts. extracted f rom the record _ kept by the ERF-Computer in the Technical' Support Center,.10 and SRO loss and interviews, and-strip; charts free Control 5 - Room' recorders. p,, Selected data was transmitted from the1various records based on the significance 'of each datus as it identified a sub-event or demonstrated, sxplicitly or implicitly, a sub-event in the sequence. The intent is that this chronology lcan be-integrated with other analysis l co determine the timaliness, accuracy and effectiveness of the measures applied to mitigate the accident. .Once the data was transcribed, s' review was performed to identify the-synchronism for time ofLehe various data sources. The principal item selected forisynchronism was-the Automatic Pzr Lo-Lo SI. The SI action. incorporates several actions including feedwater isolation and normal charging isolation that make it readily comparable;over all records. The Sequence of twents Recorder logged ~SI at 06:35:24:805; the alarm typewriter on the P-250 logged SI ' at 0639. However the earlier Reactor Manual Trip.has caused the P-250 to: alter its scan' rates.. The P-250 Post Trip review logged SI at 06:35:24 plus 1012 cycles, which equates to.06:35:40.86. The ERFC data set collected at 06:34: 14 ) records full normal charging flow and; full power feed flow to tha' steam ] . generators, approximately 16 seconds after the reactor trip had been manually 1 initiated.. By 06:34:21, the ERFC data set charging flow is reduced to 82.568 gpm'and feed flows are about 600KLBH to 800KLBH. By 06:34:34, all flows had reached a stable but low level. It appears that SI occurred at or slightly before 06:34:14.. This chronology will use 06:34:14. 1 For automatic initiation of Safety Injection, the clock compadsons are as follows: RECORDEP; TDE Sequence of Events Recorder (SER) 06:35:24:805 P-250 Computer (Alarm Typewriter) 06:35:41 ERF Computer (ERFC) 06:34:14 For Reactor Manual Trip the clock comparisons are as follows: RECORDER TIME Sequence of Events Recorder (SER) 06:35:04:548 P-250 Computer (Alarm Typewriter) 06:35:24 ERF Computer (ERFC) 06:33:56 l

e 1 i i' h. E; s '1 It is concludid that.the'P-250 led, the SER was within.20 seconds of the .P-250 and"the-ERFC was about one minute behind the P-250. j L .. The chronology: that follows is annotated by clock. time. based 'on the P-250. P ' All the~ events that occurred within each minute are listed in order of. occurrence as'could best be determined. 1 e. 'l I h 5 B l

> 14 A t m ~ STEAlf GENERATOR TURE RUPTURE EVENT NORTE ANNA POWER STATION UNIT 1 JULY 15, 1987 . July 14. 1987~ . p-2238.. Air Ejector Radiation Monitor (RM-RMS-121) was declared inoperable due' to erratic operation. l {.ur< L w m w r.:. A <.-!Ih Pi% A h-n E c-.A M 2- -M -L..u. t M \\ July 15,L1987. (.* b e. M - h * <. WM M'e d he e V ' 1 % no.rk k.w b.- r .;; w L e w 3 .

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.0630 5 An alarm was receiven w the Unit 1 annunciator panel for Main Steam (High Range) Radiation >.citor. When checked by backboard operator, ').7 -, -mM / "A"' and "B" monitors O.a in " Alert" and "C" monitot saa in "High" YGMN alarm..( $ d M c 'le, r - %p A.E -e cLu~ r ' e' 5 A- .-~ -

, g y y 'peJ.H s tw*1u~ %'n.[- '42.

i s 0631 ~ Unit 1 CR0 observad the pressurizer level decreasing rapidly. - L. 4 W

  • fe.~t #'8632 U-2 SR0 recal".ed the Shif t Supervisor and U-1 SR0 to the Control Room,

..g gg U-? CRO' to6 manual control' of charging and set FCV-JIM to full open. j Received 4'ressuriser low pressure alarm at 2135 pois (Alarm i Typewr'ter). ( wh de ~t Mudd Qs.dc{wy 2[**A 9 ) ('0 5 0633 Sh!tt Supervisor entered the Control Room and directed Jetdown. isflation. CRO initiated realignment of charging pump suction to RWST anc a 2% per minute turbine ramp p wn. Third CR0 (Backboards j Operator). assumed BOP duties on Unit 1~C5ntrol Board to assist Unit 1 CRO. (Alarm Typewriter: Make-up commenced. VCT low level alarm 20.3%.)-[.~ L 4 a ' A c d c W > M1 0634-STA arrived in the Contrel Room. Alarm Typewriter: Pressurizer Pressure 2109 psig. mammmam Superintendent of Operations was notified and g the Unit to be manually tripped. 0635 At direction of Shift Supervisor, U-1. CR0 manually tripped the reactor and turbine and initiated EF-0. CR0 observed pressurizer level at approximately 451 and pressurizer pressure at approximately.2100.psig at the time o M manual trip. 06:(35:24b, b t.;;, c,.g,4-y p c p.y ' j; POST TRIP REVIEV Initial Event Rx Manual trip (2) 0.00 sec. g '.gy-g c.,,,., ~ U Turbine Trip and P7 0.16 High Plux Rate Trip 0.33 } Rx Manual trip (1) 0.60 Pzr - Lo Press Trip 2.80 Sta Gen B Lo Lo Trip 4.2 Sta Gen C Lo-Lo Trip 4.5 l PAGE 1

d %. Sem-Gen A Lo-Lo TripManual'SITraFn"A"(1))43.3 4.5 Pzr Lo-Lo SI(Dee 16.86 Manual SI Train "B" (2) 44.97 seconds , y4 g, Alarm Typewriter: Auxiliary Feed Water Pumps Start. ;VCT level 22.1%l l increasing (indicates that charging pump suction shift to RWST is ~ ' completed.) SER: Main Feedwater Pumps Trip (06:35:25) 0636 Unit 1 CR0 noted pressurizer pressure less than 1700 psig and pressurizer level less_than 5%. Alarm Typewriter: Pressurizer level 2.7%. Main. Feed pump Breakers ~ 3 tripped. "B" Charging Pump Start. "A" and "B" LHSI pumps start. 0637 Alarm Typewriter: 0-12 breaker open. ~ 0639 A Notification of Unust:a1 Event was declared. Unit 2 SRO assumed ,)['l r/,' (Step 21 of EP-0) duties ss Interin Station Emergency Manager and initiated the EPIP's. i h * /2 DI 7. 40* Entered EP-3 from Step 23 of EP-0. ..a A1.ars / Q r,Tpewriteri, T 0641 h.AYf*"1ssthan543'F.P-12interlockset. -b .du,% 0642 Alarm Typewriter: "C" Steam Generator lovel increasing above 18% narrow range. /l t f 4. > m L Lu W ' w:.eC t% { crs T :.kum,.. "A" and "B" -C rm Typewriter: SIIndPhaseAresel.LH 0644 A ps T~ shutdown. (Steps 9,10 and 13, respectively of EP-3) 0645 Alars Typewriter: "C" Steam Generator at 25% narrow range and increasing. 0646* Auxiliary Feed Water to "C" Steam Cenerator isolated. (Shift Supervisor confirmed Steam Generator Tube Rupture in "C" Steam Generator based on "C" Steam Generator level continuing to rise.) 0646 ERFC: "C" Main Steam Trip Valve closed. 0647 Alarm Typewriter: "A" Steam Generator at 23% (Narrow Range) and increasing. Q6fd* Steam supply from "C" Steam Generator to 1-FW-P-2 (Terry Turbine) ~ ~ ~ isolated. (Step 4 of EP-3) 0648 ERFC: Graphs of pressurizer level and RCS pressure reveal increasing level and pressure. (Also noted on the strip chart in the Control Room.) 0649 Alarm Typewriter: Pressurizer Lo Press / Steam Line High Flow SI circuit blocked. (The Note prior to Step 15 of EP-3) Commenced rapid cooldown on "A" and "B" steam dump valves. (Step 15 of EP-3) Alarm Typewriter: "B" Steam Generator level at 25% (Narrow Range) and increasing. Alaru Typewriter: Pressurizer level 8.5% and decreasing. PAGE 2

~ t '! '. 4 e 1 ~ I 0650* .' Unit-I CR0 noted pressurizer level off scale low. -0651 ' Initial' notifications made to the State / Local Governments (EPIP 2.01) 'and NRC (EPIP 2.02). '0652-Alarm Typewriter: "A" T - 509. 5'F "B" T - 509.5*F "C" T C -523.5'F 0654 Interim Station. Emergency. Manager upgraded event classification to " ALERT" (Stjeg_.40.of EP-3)

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'I Alarm Typewriter: "B" Main Feed Pump. breakers rocked to test 'and . closed-(To provida a flow path for condensate pumps to, feed "A" and ~ "B": Steam Generators). Steam Generator "A".and "B" pressures at 589 pois. 0655 . Initiated EPIP-3.01, 5.03, and 5.04 (Call out,. Accountability, and Access Control)'. 0657 Strip Chart: RCS Temperature being maintained at 480*F'(Step 15 of EP-3). : Alarm Typewriter: Pressurizer Spray control at 100% demand. Valves""A" and "B" open (step 18 of EP-3). 0658* Unit 1 CR0 noted pressurizer level on scale and increasing. 0659l Alarm Typewriter: Source Range-Nuclear Instrumsats manually re-energized. (Intermediate Range Nuclear Instruments were undercompensated.) 0700 Alara Typewriter: ' Pressurizer low level heater cut off cleared (lavel l at 15% and increasing). Pressurizer heaters energized (835 KW). 0701 Alarm Typewriter: Unit 1 CR0 manually-de-energizes pressurizer heaters. 0702 Notifications made to the State / Local Governments and NRC of upgraded. l alert classification. % 0704 Alarm Typewriter: Opened one Rzaguriser PORV to tuduce pressure (Step 19 of EP-3). SRO observed prisieurs reduction of approximately 9,' 40 pois and instructed CR0 to close PORV and spray valves (Steps 18 and 19 of EP-3). Alara Typewriter: Pressurizer Relief Tank pressure j 15 psig. SR0 noted "C" Steam Generator level increase stopped. SI reduction criteria met (Step 21 of EP-3). "B" Charging Pump secured (Step 22 of EP-3). i Initiated the isolation of BIT flowpath (Step 24 of EP-3) and established the normal charging flowpath (Step 25 of EP-3). ] '0706** "A" and "B" pressurizer spray valves closed. 0709 Alaru Typewriter: Non Regenerative Heat Exchanger outlet flow 47 gpm. (Evaluation of this entry indicates that normal letdown had been j l restored in accordance with Step 29 of EP-3. PAGE 3

T: ; ;.e p 0710: Alarm Typewriter: ' Pressurizer heater breakers' closed (Step 31 of ' EP-3).. "l0711 Alarm Typewriters; Pressurizer heater breakers closed (Step'31 of t-EP-3). 07137 AlaraT[pewriter:. Secured "C" and "B" RCP's.(Step 38 of EP-3). '071A-Alarm Typewriter: Spray.d====E 76%.- (From this time forward RCS'

pressure is maintained by manual control of' spray;and heaters.)

0715 Superintendent of' Operation and SRO-On-Call arrived in the' Control -Room. .0718' Transitioned to ES 3.1 " POST-STEAM GENERATOR TUBE RUPTURE C00LDOWN USING BACKFILL" (Step 42 of.EP-3). 0720 . Station Manager arrived in the~ Control Room. 0721 Alarm Typewriter:.AFW Feed' Pump 3A to "C" Steam Generator secured. 0722 -Alarm Typewriter: Pressurizer level 73% and decreasing. 0723 ' Alarm. Typewriter:. AFW Feed Pump 3B to "B" Steam Generator secured. (Subsequently AFW pumps are run intermittently to support Steam i Generator feed requirements.) 0725 Alara Typewriter: Started "B" condensate pump (Both "A" and "B" Condensate pumps now running). 0727** Began RCS cooldown in accordance with ES 3.1. 0730 ' Assistant Station Manager arrives in Control Room and initiates . transition of EPIPs and communications from Control Room to TSC. .0739 Station Manager assumes Station Emergency Manager position. ( Turbine on the turning gear [. 0745' Alarm Typewriter: o I 0756 Condenser Air Ejector manually diverted to containment. 0757 Technical Support Center activated. 0810 Alarm Typewriter: Secured "B" condensate pump. 0820 Corporate Emergency Response Center activated. 0845 Started "B" RHR pump for system warm-up (Step 9 of ES 3.1). 0853 Alars Typewriter: Closed "A" MFP breakers. (Breakers in test to permit opening of pump discharge valve to use Condensate Pumps for feed to Steam Generators.) 0857 Alarm Typewriter: Open "B" MFP Breakers (To permit isolation of "B" MFP to stop spraying from "B" pump suction relief valve. PAGE 4 )

+ ), - 0900* Loose Parts Monitoring System Alarm on "C" Steam Generator. 0915 Local Emergency Offsite Facility activated. Commenced using auxiliary spray to supplement RCS depressurization. 0949. Containment partial pressure exceeded allowable set point due to Air Ejector exhaust diversion to containment. l 1049 Pressurizer PORY Key Switches to "AUT0" for NDTI protection. (SRO Log) 1108 Entered Mode 4. l 1153 Cycled reactor trip breaker to re-enable automatic Safety Injection. 1200 Placed "A" and "B" charging pumps and "B" LESI pump in " Pull-to-Lock" in accordance with 1-OP-3.3. 1219 Placed RHR System in service to continue RCS cooldown (Step 9 of ES 3.1). 1221 Secured "A" Reactor Coolant Pump. (SRO Log) 1254 Main Steam System secured in accordance with 1-OP-28.1. 1312 Restored Air Ejector exhaust to normal alignment. 1330 Entered Mode 5 1335 Station Emergency Manager terminated the emergency. 1336 Notified Nuclear Regulatory Commission, State and Local Governments of termination of emergency status. 1336 Implemented Recovery Organization.

  • Approximate time based on CRO, SRO, and/or STA observation.
    • Approximate time based on computer or strip chart data.

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~ _ Sur2 voar 1 ~ JULT .5 198E ~ fo11oving chronology was reconstructed fTo support m Cenerator Tube tapture the typewriter attached to the Control Rorom the print outs of the alarm twents Recorder (Dranets) driven by the R om P-250 process computer. ' the Sequenc printouts estracted from the record kept b athaway assenciator eyeten, the data Support Center Room recorders., 30 and St0 logs and interviews, and strty ch y the ERP Ceaputer ta the Technical ree Control Selected data was transmitted from the oxylicitly or implicitly, a sub-event in thesignificance sub-evanc or demonstrated, chronology can be int 33 rated with other an l sequence. accuracy and effectiveness of the measures a a ysis to determine the timeliness,T pplied to mitigate the accident. synchronism for time of the various data souonce th for synchronism was the Automatic Par Lo L o identify the rces. The SI action incorporatesThe several actions including feedwater isolatio - o.51. that maka it readily couperable over Recorder logged SI at all recorde.n and normal charging toelation 06:35:24:80$ et 063p. Iowever the earlier Resceo$ the alars typewriter on the F 1 its scan rates. The F-250 Post Trip review logged 81 at 06 3$r Man cycles, which equates to ogged SI 06:35:40.86. records in11 normal charging flow and full p ganarators. approximately 16 seconds after thThe ERFC da t24 plus 1012 initiated. By 06tS4 21 ower feed flow to the steam 14 before 06:34: reached a stable but low level. ppa and feed By 06: redsced to 82.$68 N o chronology will useIt appears that SI occurre:d at or s 14 06:34:14. y For automatic initiation of Safety Inj are as fo11over ection, the clock comparteous 23CQRDB _ TIE P-250 Computer (Alara Typewriter) Sequence of 06:35:24:405 ERF Computer (ERFC) 06:35:41 06234:14 For Reactor Manual Trip the clock compa i RECORDE_R r sons are as follsws: M P-250 Computer (Alarm Typewriter) Sequence of 06:35:04:548 DF Computer (ERFC) 06t35s24 06:33:56 FQOM NO ANNA TO REG 2 07/20/87 07 17 P. 3 _ - - - - -~ ~

4 4 b -It is coni:1oded that the F-250 led, the stR was within -20 seconds ~of the d F-250 and the ERFC was about one minute behind the P-250. The chronology that follove is annotated by elock time based on the F-250. .) I All the' events that occurred within each minute are listed in order of .ocestrence se could best he' determined. 1 1 I i i l l t l l

7 ~- y 1 + m anu - . stim ammaWir 9d" :. svar \\ r== men r= asur ur i i J45.Y 5 193: ~ t July 14, 1987 .2233 i - Ali tjeetor Radiation Monitor (RM-RMs-121) was declare to erratic operation. perable dus-July 15, 1987 1 0630 .l "A"sh Ramse) Radiation Monitor.An alarm wee rec (Wi a $ teen sad "B" aonitors were in " Alert" and "C" somiter wa stars. 0631-Unit 1 CR0 observed the pressurizer level decreasing rap 0632 y. ) U-2 SRO recalled the Shift Supervisor and D-1 330 to th Received Pressurizer low pressure alarm e control Room. o is11 open. \\ Typewriter). m 0633 Shitt supervisor entered the Control Room and directed let i isolation. and a 21 per minute rurhine ramp down.Ca0 initiated realig n CRO. - (Alarm TypewriterOperator) assumed 80F duttee on Un! 20 3%.~) Make-up coamenced. TCT low level alare assist Unit 1 0634 STA arrived in the Control Reca. Alatu Typewriter Pressurizer Pressure 2109 pets. 1 Superintendent of Operations was notified and directed I manually tripped. the Unit to be 1 0635 At direction of thift Supervisor, U-1 C10 annually tri i and turbine end initiated gP-0. pped the resetor C10 observed pressurizer level at approntastely 455 sad pressurizer pressure at approxima at the tien of the amasal trip. POST TRIP EEVIEN, s Initial Event 06:3$224 As Nasual trip (2) Turbine Trip and P7 0.00 sec. 0.16 Nigh Flux Rare Trip 0.33 1x Manual trip (1) 0.60 Par - Le Press Trip 2.30 Sta Gen 5 to-Lo Trip t 4.2 Sta con C Lo-Lo Trip 4.5 PACE 1 FROM NO ANNA TO REG 2 07/20/97 07:16 P. 5

x. m pu St3 Sea A Le-Lo Trip 6.5 Par 14-Le 81 16.86 Nasus 1. 81 Train "A" (1) 43.38 e ._ Manual 81 Train "3" (2) ~44.97 seconds Alars Typewriters Auxiliary Feed Water Pumps-Start. increasing (indicates that charging pump suction shift to RW8 YCT 1evel 22.1% O completed.)- is. 8th Mais Feedwater Pumps Trip.(04:35:25) 0634 Untt'.1:CR0 noted pressurizer pressure less than 1700 p pressurizer level lass.than SE. Alarm Typewriter Pressurizer level' 2.71. tripped.

  1. 3" Charging Pump Start.

Main' Food pump Breakers "A" and "5" LM81 pumps. start. 0637' Alarm Typewriter 0-12 breaker open. '0639 ~ A Notification of Unusual twest was declared. duties as Interim Station Emergency Manator and initiated th Unit 2 8R0 assumed. (Step 21 'of EP-0) s. s 0640* 1 Entered EP-3 from Step 23 of EP-0, 0641 Alara Typewriter T less than 543*F. P-12 interlock set. gg 0642 Alars Typewriters "C" Steam Cenerator level increasing above 142 narrow range. ) j 0643* 1 Cgsgd pin Steam Trip Valves (8tep 4 of EF-3). cpne i%4c4**5 0644 and'13, respectively of EF-3)Si and Phase A reset. LH8I pu "A" and "B" shutdown. 50 C 68"M (Steps'9, 10 0645 Alaru Typewriter: -i ' increasing. "C" Steam Generator at 252 narrow range and i 0646* Auxiliary Feed Water to "C" Steen Generator isolated. Generator based on "C" SteSupervisor confirsed 8 team Genera{ (Shift 0644 EUC, 3 "C " hhh Wp VcAve C(nud.Generater level continuing to r i 0647-Alarm Typewriter: i incrassing. "A" Steam Cesarator at 232 (Narrow Range) and j 0648 EtFC: Craphs of pressurizer level and RCS pressure revea level and pressure. (Also noted on the strip chart in the Control Room.) F '0649 Alare Typewriter blocked. (The Note prior to. Step 15 of EP-3) Pressurizer Lo on "A" and "B" steam dump valves. Commenced rapid cooldown i (Step 15 of Ep-3) Alarm Typewriter: "B" Steam Generater level at 252 (Warrow Range) and increasing. Alsen Typewritart Pressurizer level 4.55 and decreasins. I. 0650 I Unit 1 CEO noted pressurizer level off scale low. i PACE 2 FROM NO ANNA TO REG 2 07/20/97 07:19 P. 6

r. y ~f L* ' 0651 !aitial notification made to the State / Local Governments (EP and IRC (DIP 2.02). 1 0452 .Alaru typewriter: "A" ?*'- 509.5'F "B" T". - 309. 5 'F "C" T 523.3'F 0654 Interia Station Energency Manager upgraded event classification to l "ALBEr".. j lAlera Typewriter: "B" Main Peed' Pump breakers racked to test and closed (To provide a flow path for condensate pq to feed "A" and "B" Steam Generators) Steam Generator "A" and"B" pressures at 539 pois. '0655 Initiated WIP-3.01, 5.03, and 5.04 (call Out. Accountability, and l Access Centr'ol). m 0657 Strip Chart RCs Temperature being maintained at 440'F (8tep 15 of f IP-3). Alara p iter Prosauriser Spray control at 1002 demand. Yalves "A" g 3" opea (step 13 of I?-3). Unit. l' CSD noted pressurizer level on scale and tacreasing, l 0653 C659 Alare Typewriter: Source Range Nuclear Instruments manually re-energised. (Intermediate Reage Nuclear Instrumaats vers undereenpensated.) \\ 0700 'Alare Typewriter: Pressurizer low level heater cut off clotted (level at 151 and increasing). Praevuriser heaters energised (835 kW). 0701 Unit 1 C10 mar.ually de-enstgisse praeeuriser heaters. 0702 Wocifications made to the State / Local Governments and NRC of upgraded -{ alert classification. 070!. Opened one Pressurizer PORY. to reduce pressure (Step 19 of EF-3). SRO observed pressure reduction of approximately 40 pois and tastructed C10 to close PotY and spray valves-(Steps 18 and 19 of EF-3). Alara Typewriter Pressurizer Relief Tank pressure 15 pets. 4 8R0 metod "C" Steam Generator level increase stepped. $1 reduction criteria set (Step 21 of EP-3). "B" Charging Pump secured (Step 22 of IP-3). Initiated the isolation of SIT flowpath (Step 24 of EP-5) and established the normal charging flowpath (Step 25 of EP-3). ) l 0706 Shut "A" and "B" pressurizer spray valves. 0709 Alara Typewriter Non Regenerative Nea) Exchanger outlet flow 47 spa. I (Evaluatise of this entry indicates that normal latdown had been restored in accordanes with Step 29 of D-3. 0710 Alara Typewriter: Pressurizer heater breakers closed (Step 31 of EF-3).

.g 4 '*j EP-3). 0711 Alara Typewriter Pressurizer heater breakers closed (Step 31'of 0713 Alaru Typewriters. ~ Secured "C" and "S" RCF's - (Step 38 of D-3). - 0714 Alers Typewriter Spray demand 761. (From. this time' forward SCS pressure is maintained by manual control of spray and heaters.)' ..0715 Superintendent of Operation and SRO-On-Call arrived -in the Centrol-Rose. 0718 Transitioned to 38 3.1 " POST-STEAM GENERATOR TUBE RUPTURE C USINC BACHILL" (Step 42 of EP-3). 0720 Stetion Manager arrived in the Control Roos.. .i 0721'- Alaru Typewriter . AFW Feed Pump 3A to "C" Steam Generator-secured. 0722. Alarm Typewriter: Pressertser level 731 and decreasing. 0723 Alarm Typewriter: AFW F6ed Pump 38 to "3" Steam Generator escured. (subsequently AFW pumps are run intermittently to support 8 team Canarator feed requirements.) 0723 Alarm Typewriters - Started "B" Condensete pump (Soch "A" sad "B" Condensate pumps now running). 0727 Began PCS cooldova in accordance with 58 3.1. 0730 Assistant.$tation Manager arrives in' Control Room and initiates transition of W IPs and communications from Control Room to TSC. 0739 Station Mana8er assumes Station Emergency Manager position. l 0745 Alarn Typewriter Turbine on the turnin8 gear. 0756 Condenser Air Ejector maanally diverted to containment. 0757 Technical Support Center activated. ) i 0410 Alarm typewriters secured "3" condensate pump. l 0820 Corporate Emetgency Response Center activated. 0845 Started "B" RHR pump for systen vars-up (Step 9 of E8 3.1). 0853 Alarm Typewriters closed "A" MFF breakers. (Breakers in test to permit opening of pump dischar8e valve to use Condensate Pumps for feed to Steam Generators.) 0857 Alara Typewriter: Open "B" NFP Breakere (To permit isolation of "8" MPP to stop spraying from "S" pump suction relief valve. 0900* 1.cose Parts Monitoring System Alars on "C" Steam Generator. 0913 1.ccal Emergency offeite Facility activated. Commenced using auxiliary spray to supplement RC8 depressurisation. t .OMED A

L, I,-, 1, 0949 'tjsetor exhaust diversion to containment.Centeinm {" o Air .{ 1049 Preasuriser PotY Key Switches to "AUf0" for WDTT protectio j l 1100 Batered Mode4. -f 1153 , Cykled reactor trip breaker to re-enable estomatic Sofet .c '1200 is secordance with 1-0F-3.3.Placed "A" and " charging pumps and "B" LaSI pump in " Pull-to Lock 4 1215 ' Secured "A" Reactor Coolant Pump. 1219 Placed RER System in service to continue RC8 cooldown ( 3.1). 1254 Main Steen System secured in accordance with 1-CP-28.1. 1312 Restored Air Ejector exhavet to normel elignment. 1330 Entered Mode 5 1 1335-Stetton Emergency Manager terminated the energency. 1 1336-Botified Nuclear Regulatory Commission, State and 1.ocal G tesisinttien of emergency status. s of 1336 Isplaneated Recovery organization. '*Approxinete Time ha,m.1 e Cf.0; 520 M [py i ST4 0b5w MlON 5 l 1 FROM NO ANNA To REG 2 PAGE 5 l 07/20/87 07:21 P. 9 TOTAL P. 9

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me tmEs28 - 'euty 20.1987 I ~r Licencee/ Facility. Notification /Subjset VEPCO/ North Anna 1 2 R esident inspector,7/20/87. Steam Generater7ube Rupture .~.__.u.~..p_.. .a. y _a ~. J.- W Event Event No. N/A c The incident Response Team was upgraded to an Augmented Inspection Team on Ju '17,1987, because the steam generator tube (s) leak was determined to be 600 gpm (the FSAR evaluation of a single tube. failure indicated a leak of up to 785'gpm) and the f act that the licensee had conducted 100% tube inspection during the last outage. Total activity released for the steam generator tube leak was calculated to be 0.27 curies (primarily Xenon 135,138; Argon 41: Krypton 87,.88)(0.62% lechnical specification limit). By July 20,1987,8:00 a.m., the C steam generator had been isolated and drained. The leak appears to be on the hot side of the steam generator near the seventh support plate (at the point where the tubes begin to loop). Removal of the steam generator manway and diaphram are scheduled today Westinghouse personnel are onsite for detailed inspection of the internals. 'Nrr ~ and Region li specialists will be present during the examination and participate in the evaluation of results. Regional Action: PRIORIT TTENTION REQUIRED PAGE 3 July 20,1987 Ucensee/Faciht' Notification /Sebiect VEPCO/Surry 1,2 Resident insoector Congressional Tour r EVENT EVENT NO. N/A l { Senator C. Hecht (R), Neva a, hill be leading a congressional tdur with six staf f ( members of the fuelhandling will provide fuel handling demo strations.ullding and dry fuel cash storage area. The utility Regional Action: The resident ins ectors willinterface with'the utility and congression party. Ucensee/ Facility Notifica n/ Subject Nuclear Fuel Services Telephon Resident inspector, 7/1/87 Ra)) road car derall nt/ fire Event. Event No. N/A On Friday, July 17,1987, at 1:00 p.m., a train detallm one-hall mile north of the Nuclear Fuel Services Plant. t occurred approximately orty-one cars were -ederailed including tour tankers containing methanol. One ruptured and the escaping methanoli f the tankers was wars e,va nc..e se-s,. -o - --~ gnited. Residents in--- e near vicinity pp.g7 g "j < - :- o v ' '~ D7

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.J M p g [f~ 8,,2 7 3.a~ SECONDARY WATER CHEMISTRY PROGRAM Staff Recourended Action Licensees and* applicants should have a secondary water chemistry program (SWCP) to aini,aise steam generator tube degradation. 6 S The-specific, plant progran,should incorporate the~ secondary water chemistry guidelines in SGOG Special Report EPRI-NP-2704 "PWR Secondary ~ Water Chemistry Guidelines," October 1982, and. should address measures taken to minimise steam generator corrosion including materials selection, chemistry J limits..and control methods.- In addition, the specific _ ~ plant procedures should include progressively more stringent - I' correctiva actions for out-of-specification water che.aistry conditions. These corrective actions should include power reductions and shutdowns, as appropriate. when ~ excessively corrosive coaditions exist. Specific functional individuals should be identified as having the responsibility / authority to interpret plant water chemistry information, and initiate appropriate plant actions to adjust chemistry, as necessary. The referenced SGOG guidelines above were prepared by the Steam Generator Owners Group. Water Chemistry Guidelines Committee and represent and consensus-opinion of. a significant portion of the industry for state-of-the-art secondary water chemistry control. TReference' Section 2.5 of NUREG-0844. 1

Response

In order to provide maximum chemical protection to the secondary system components and to minimize steam generator tube degradation, a Virginia Power Nuclear Operations Department Policy has been established that incorporates the "PWR Secondary Water Chemistry Guidelines" 'into the appropriate station procedures at the operating. plants. Specifications have been adopted from the Steam Generator Owners Group Guidelines (SCOG) that address four unit conditions: 1. Wet Layup: The steam generators will be placed in wet chemical layup when maintenance schedules allow. The cheatstry specifica-tions are listed in Table I. 2. Unit Startup Prior.to heatup, the steam generator should meet the {. values listed in Table II. A chemistry hold will be implemented at 55% reactor power and again at s30% reactor power to ensure that the specifications are set prior to power escalation. 3. Unit Operations: Table III lists the control parameters for power operation with associated Action Levels. ,4. Unit Shutdown A chemistry hold at 350'F' will be implemented at each planned unit shutdown when the plant is being brought to cold shutdown condition to enhance impurity hideout return / steam generator cleanup. JWO/jla/8/0GREN k b/n

S *, For each of the above unit conditions, the SGOG Guidelines have been implemented.". _To ensure system integrity, Action Levels have been established. for at power out-of-specification conditions. Action Level . responses are summarized as follows: j Action Level 1: Return parameter to within scrual. value range within one week or Action Level 2 is initiated. Action Level 2: Reduce power to 30% or less within six hours of initiation of Action Level 2. Return parameter to within normal value range within 100 hours or go to Action Level 3. Action Level 3: Shutdown within six hours after entering Action Level 3. Station ~ procedures' that implement this policy, delineate the management responsibilities and the necessary corrective actions for out-of-specification parameters have been developed. i 1 JWO/jlm/8/0GREN l

a r L 'c. '. L - y., i '~ TAELE I g COLD SHUTDOWN / WET LAYUP LIMITS { STEAM GENERATORS 2 l Parameter Normal Value Initiate Action pH 025*C 9.8-10.5 <9.8 Hydrazine, ppa 75-200 <75 Sodium, ppa. 11.000 >1.000 Chloride, ppa _ 10.500 >0.500-Sulfate, ppa (1.000 >1.000 s TABLE II l START-UP LIMITS _ STEAM CENERATORS { Parameter Value Prior go value Prior to Value Prior to Leaving.200 F Power >5% Power >30% pH'925'c 8.5-9.051) Cation Con-ductivity, US <2.0 <0.80 Sodium, ppa' <0.100 <0.100 <0.020 Chloride, ppa -(0.100 <0.10'O - <0.020 ' Sulfate, ppa (0.100(2) <0.100(2) <0.020(2) Gilica, ppa ( 5) _0.300 U) I) pH value of >7.0 at North Anna Diagnostic values for North Anna only at this time; the Action Level time clock begins when the unit is escalated above 30% power. i l. i l l

r* TABLE III POWER OPERATION STEAM GENERATORS Parameter Normal Value Action Level 1 2 3 pH $25*C 8.5-9.0(1) <8.5,>9.0(1) Cation Con-l T ductivity, pS 10.80 >0.80 >2.0 >7.0 Sodium,' ppa <0.020 >0.020 >0.100 >0.500 Chloride, ppa <0.020 >0.020 >0.100 Sulfate, ppm (2) 10.020 >0.020 >0.100 Silica, ppa-10.300 >0.300 l { (1) North Anna 7.0-9.2 t (2) The Action Level clock for North Anna on sulfates applies only when the unit is >30% reactor power. i o 6 i l l 1 l 0 J 10 CHEM

4 ? ,.,9 y 73.b CORDENSER INSERVICE INSPECTICE PROGRAM Staff Recommended Action I Licensees should implement a condenser inservice inspection program. ' The pregram should be defined in plant specific safety-related procedures and include: j 1. Procedures to. implement a condenser inservice inspection program that will be initiated if condenser leakage is of such a magnitude that a power reduction corrective action is required more than once j per three month period; and is ' required more than once per three month period; and 2. Identification and location of leakage source (s), either water or l air; i 3. Methods of repair of leakage; l t 4. Methodology for determining the cause(s) of leakage; J 5.. A preventive maintenance program. Reference Section 2.6 of NUREC-0844 I +

Response

Virginia Power has implemented a secondary side inservice inspection program which includes inspection of the condensers. This program includes addy current examination of a. number of condenser tubes each refueling outage, with the selection of tubes to be examined based on I historical records of condenser performance. As part of the Virginia Power secondary water chemistry program, secondary chemistry is closely monitored and indications of condenser Isakage will result in initiating procedures to identify and correct the source of either water or air leakage. The procedures used t.o identify, locate and repair condenser leaks I are approved plant specific procedures, however, our condensers are not . classified safety.related. At present. Virginia Power has no specific procedures detailing the methodology for determining the cause of condenser leakage. However, as a matter of practice, the occurrence of condenser leakage of such a I magnitude that power reduction corrective action is required will initiate an investigatica to determine the cause of the leakage. As an saample, Virginia Power has recently completed an investigation of the causes of an increased incidence of condenser tube leaks at the North Anna Power Station during January and February 1985. Based on the results of this investigation, actions (e.g., evaluation and initial engineering) are presently in progress to correct the causes of the tuba 1eaks. ) jwr/LM6-os069

kh$;,,,.- f. p, JULY 21,1987 - 0900 MEETING l1 ', p 1. Plant' Status n ~ RCS at 112*F and 40 paig with bubble in pressurizer (277'F). RHR system in service. A and B Steam Generators in wet lay-up. C S/G drainir.g, present.ly at about 8% draining to containment p;? sump for processing. - Containment on purge. l' C RCS loop drained and S/G primary manways and diaphrages-removed. s f Initial video inspection of S/G tube sheet showed tube R9-C51 as leaker (T side). H Attempted ~ to insert fiber optics probe into ' R9-C51 with no. ' success. The probe appeared to be binding due to the water L~ - still dripping through the tube. s I Work on the MSR Reheat Stop and the Intercept Valves will start today. Unit.2 is at 77% poser Tav 574*F on'a coastdown. 2. Technical When failed tube identified, Eddy Current tape. history was examined. During recent outage, R9-C51 tube was examined up to and slightly past the 7th TSP ' on the hot leg side. No significant indications were identified during this inspection. Due to the location of the tube and the preliminary location of the break, visual examination from the secondary side is unlikely. In addition, the removal of the tube for further examination (destructive testing) may be restricted if the break is in the radius section of the tube. Next step will be to furtiar examine the tube using Eddy current and fiber optics techi.iques to gather all possible information. C S/G 1evel will be adjusted (filled and drained) to further reduce the boron concentration to less than 10 ppe for opening of the secondary side. 4 EC examination will include the 8 tubes surrounding the defective tube and then proceed to the T.S. inspection program. The possibility of a loose part causing the tube failures is diminished based on the location of the failure. However, the loose part logic has been prepared for incorporation into the inspection plan. Y'Z.kl'7*Vh( D/93 et,%. i

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( 37 s. i Page 2 W ,.g 07/21/87 {. /; 3. NS&L i b h^ NRC presentation today following the morning meeting to provide l [ background information along with a status of technical l investigations. f NRC thought that the loose part alarm had been a precursor to the event. It.was pointed out that the loose part alarm did V not come in until approximately 0900 (2.5 hours into the 1 event). ?-f Information on the loose part alarm has been turned over to the E-technical group for evaluation. A draft of the event report will be completed by tomor" row. NRC requested a demonstration of the S/G tube rupture on the simulator. ~ A meeting is scheduled for 1300 with NRC to do a detailed review of the Sequence of Events and associated documentation. f 4. Health Physics g Interviews with HP response personnel will be # completed this. 3 morning. The response report draft will be completed ? thereafter. 6[ Reviewed origin of environmental samples with NRC. These included: TLD's from 5 sectors, soil and vegetation samples taken by monitoring teams, and samples from the air particulate and radioactive Iodine sampling stations. 5. Industry Interf ace The sequence of events will be placed on Nuclear Network following the NRC presentation. Additional distribution will [ include DES and SCC, ODEC, EPRI, and several Foreign utilities. i Details of tomorrow's press conference presentation should be completed today.

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RIORITY ATTENTION REQUIRED MORNING REPORT - REGION il JULY 81,1987 . Licensee / Facility Notification / Subject VEPCO/ North Anna 1 Resident inspector, 7/21/87. I Update of S/G tube rupture event ~ Event Event No. N/A l Reference daily report 7/20/87 North Anna 1 tube rupture event. On 7/20/87 the licensee determined the leak in the C Steam Generator to be coming from tube 09-51. This determination was made based on visualinspection of the hot-leg side tube sheet. Based on a previous draining of the C S/G the leak appears to be around the seventh (top) support plate; A review of the inspection tapes from the ) previous outage did not reveal any reportable indications. This tube was inspected I on the hot side only through the seventh support plate. The licensee is preparing to perform a visualinspection of the inside of the tube in an attempt to determine some information on the type of the break. A press conference will be conducted on site on 7/22/87 at 11:00 c.m. l Regional Action: For inf ormation only. Licensee / Facility Notification / Subject i GPC/Vogtle Duty Officer Via Red Phone I A an Followup Telecon to Region 11 "UN CURED VITAL AREA DOOR" l Event Event No. N/A by the clarm station. A dispatched secur\\,ty o ficer found a vital area door At approximately 1:50 a.m., July 19,1987 vital area door alarm was received i i unsecured. Search of the area revealed that a uthorized plant eaulpment operator was conducting a routine check of the e ipment in the room. The door access key had previously been signed at 10:50 a.m., July 18,1987, i Regional Action: Licensee inv'estigation continuing and ro e RsJion 11 followup. I l 4 ) i i A 1 L Fern-F7-441 b/9y t. -m____________ I

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w P IORITY ATTENTION REQUIRED : MORNING REPORT - REGION ll July 22,1987 i Licensee / ility Notification / Subject ! APC/Farley' 1 & Resident inspector Non-Qualified 125 Vdc Splices - l' 'l ' Event. vent No. N/A. -The licensee informed the Resid t inspector on' July 21 that during the 1982-83 l environmental equipment (EQ) upgr e, approximately.84 (125 Vde) solenoids were replaced on each unit using sus eted non-qualified bolted and taped I ' wire splices instead of qualified heat sh kable wire splices., Bechtel. . and the licenses have identified, evaluated d prepared justifications for continued operation.' Both units are at 100% wer. Rework of accessible solenoid splices is in progress. The suspect so olds are associated with valves in all major systems. Raychem splices are being flown in overnight.. The licensee stated their review was initia af ter learning. of similar problems at other f acilities. <i Regional Action:. Resident and Regional followup in progress. Licensee / Facility Notification / Subject 'VEPCO/ North Anna 1 Resident inspector, 7/22/ 87 H Update of S/G Tube Rupture ( l Event Event No, N/A Reference Dalty Report of July 21 - North Anna 1 tube rupture event. On July 21; the licensee identified the type and location of the rupture on tube 09-51.' The break was circumferential and located above the seventh (top) support plate on the coldleg. and the separation between the two broken ends was found to be 11/4"..The licensee is prepar:ng to perform the eddy current test for all tubes on the cold leg and expects to complete the testing in 12 days. A press conference will be conducted on site today approximately at 11:00 a.m. 1 Regional Action: A PN will be issued. For Information Only. l [ .) ( REGION 11 MORNING REPORT PAGE 2 JULY 22,1987 Licensee / Facility Notification / Subject General Electric Nuclear Fuel Telecon from Licensee j Bomb Threat i Event Event No. At 10 30 p.m. on July 21, a bomb threat call was received in the Fuel Manuf acturing Foz4-tr-w 1 N m

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) ~. NORTH ANNA UNIT I STEAM CENERATOR TUBE RUPTURE EVENT I. EXECUTIVE

SUMMARY

i A. Brief Description of North Anna Power Station B. Overview of Event C.. Virginia Electric and Power Company and NRC Response Organizations i II.- DESCRIPTION OF EVENT .i A. Detailed Sequence of Events B. Equipment Response C. Operator Response D.. Lessons Learned l l III. RADIOLOGICAL EFFECTS OF. STEAM CENERATOR TUBE RUPTURE EVENT A. Summary B. Evaluation of Radiological Release 1. Offsite v 2. Onsite 'IV. EMERGENCY PLAN IMPLEMENTATION A. Chronology of Rey Events B. Listing of Emergency Response Personnel by Facility C. Key Elements of Emergency Plan Demonstrated By Event D. Miscellaneous Reports E. Lessons Learned V. SAFETT EVALUATION A. Primary-to-Secondary Leakage Calculations i B. 4 Comparison of North Anna Steam Generator Tube Rupture Event to UFSAR k Accident Analysis 1 1 1 Fern-27-yh ( Dht

.} Lh,N ~. l .g g ~ NORTH ANNA UNIT 1 STEAM CENERATOR TUBE RUPTURE EVENT l l VI..- STEAM GINERATOR TUBE EVALUATION .) A. Examinations .j B. Failure Analysis C. Tube Integrity Evaluation l D.. Repair Criteria E. Lessons Learned VII.

SUMMARY

OF LESSONS ' LEARNED A. Operations: B. Training C. Stem Generators

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