Text

Forwards Rev 2 to Sequence of Events Re 851226 Reactor Trip. Related Info Encl
	ML20155H518
Person / Time
Site:	Rancho Seco, 05000000
Issue date:	01/04/1986
From:	Wichert R; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Perez G; NRC
Shared Package
ML20155H523	List: ML20155H518; ML20155H562; ML20155H585; ML20155H596; ML20155H625; ML20155H627; ML20155H636; ML20155H645; ML20155H654; ML20155H665; ML20155H686; ML20155H706; ML20155H720; ML20155H726; ML20155H742; ML20155H774; ML20155H798; ML20155H841; ML20155H894; ML20155H911;
References
	RTR-NUREG-1195 TS-86-006, TS-86-6, NUDOCS 8605160259
	Download: ML20155H518 (31)
	v • d • e

{{#Wiki_filter:- _ ___._~. _ _ ,.3-a SACRAMENTO MUNICIPAL UTILITY DISTRICT OFFICE MEMORANDUM Januahy4,1986 TO: Distribution DATE: TS 86-006 FROM: R. P. Wichert _ / susJEcn REVISION 2 to 12/26/85 TRANSIENT SEQUENCE OF EVENTS i The Sequence of Events relative to Reactor Trip #75 has been revised and is attached. This revision provides additional information as to plant conditions concurrent with specific events while clarifying certain events. It is anticipated that this Sequence of Events will be further revised whenever sufficient or pertinent new information becomes available. Revisions have been indicated with Revision marks. cc: Trip Report D.istribution ' Glen Perez, NRC Resident Inspector Control Room George Coward Steve Redeker Bill Spencer Ron Lawrence Charlie Linkhart Fred Kellie Jim Jurkovich-f Val Lewis I l i 8605160259 860306 PDR ADOCK 05000312 S PDR l

j u. 3._=a _x 01/04/85 ~g :t _.- 1600 Hours .r ' Res. 2 s l ii. ? CHRONOLOGICAL SEQUENCE OF EVENTS ~ 12-26-85 s, Initial Conditions t j Unit operating at steady state power of 76%, 710 MW(e) l Integrated control system in full automatic ? j Bailey Computer out of service (one of the plant's two main computer systems in the Control Room) f Transient Initiator Time Source Event / Action j ij 04:13:47 IDADS Loss of ICS power. All lights j on the ICS stations go out. All i demands fall to 50%. " Loss of ICS Power" annunciator alarms. I "I a 'h i 'I ~. l e

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SEQUENCE OF EVENTS j

n.- ) ~ ~ Following developed fron IDADS Print Out. Alarm Typer and Operator Interview. 1 Control.oom clock.was found to be R Time adjusted to IDADS Print Out time. i about (1) minute ahead of IDADS time. l TIE DESCRIPTION OF EVENT DATA SOURCE 14 04:13:47 " Loss of ICS Pcwer" Annunciator Alarms. IDADS Print Out Upon a loss of ICS power all ICS demands go to midscale, corresponding to zero volts. (The ICS works on i 10 volt scale, zero volts being 50% demand.) The startup and main F.W valves closed to 50% demand because of this i decrease in demand signal. The MFW pump j speeds were reduced to low speed. Deenergizing the ICS also causes the Main Feedwater block valves to close. This expected response of the feedwater system is illustrated by graphs of Main Feedwater pumps speed versus time, main feed pump discharge pressure versus time and i Main Feedwater flow rates, all.of which j indicate a decrease just prior to reactor trip. With the plant initially at about 76% full power, this reduction in feedwater flow is e.xpected to increase RCS pressure. j The loss of ICS power also sends a demand to j the Bailey AFW control valves and ADVs and ~ TBVs to open to 50% demand. The effect of the l ADVs and TBVs opening to about 50% is indicated on the plot of OTSG pressure versus time. The IDADS plot clearly shows SG pressure decreasing prior to the reactor trip. 04:13:7 Operators notice MFW flow decreasing rapidly. Operator Personal J Also noticed RCS pressure increasing. Operators Statement j opened pressurizer spray valve in an attempt to p stop RCS pressure increase. I Due to the overheating of the RCS by the reduction in MFW flow, the actuation of pressurizer spray is not sufficient to reverse RCS pressure, as shown on the IDADS Print Out of RCS pressure. t K d

.-._-.-....a..... .. a =.w y z- .,e TIE DESCRIPTION OF EVENT DATA SOURCE 04:14:01 The low main feedwater pump discharge pressure IDADS Print Out of less than 850 psig automatically started the electric driven AFW pump, P-319. h 04:14:03 Reactor trip on high RCS pressure. The turbine IDADS Print Out j trip is also initiated by the reactor trip. Operator closes pressurizer spray valve. 04:14:04 Peak. RCS pressure at reactor tr7p of 2298 psig IDADS Print Out 04:14:06 AFW dual drive pump, P-318, autostkr't's on low IDADS Print Out feedpump discharge pressure (850 psig). 4 j AFW pump,.P-318 was steam driven throughout i this transient. 04:14:06 Peak RCS hot leg temperature of 606.5'F IDADS Print Out 1 04:14:7 Immediately upon reactor trip, many fire alarms, Operator Personal TSC spray actuation alarm, seismic trouble alarm, Statement and SFP temperature high alarms were received. The operators performed the actions of the 1 Emergency Procedure Section E.01. This included reducfng letdown flow. l Operators then proceeded with Emergency j Procedures Section E.02. i 04:14:11 AFW flow began to both OTSGs. IDADS Print Out 04:14:25 0'perators performing procedure E.02, Vital Operator Personal System Verification due to pressurizer level Statement /IDADS decreasing, fully. opened "A" inject valve Print Out for more makeup addition to RCS. 04:14:30 Operators sent to close TBVs, ADVs, and AFW Operator Personal flow valves. Statement 04:14:48 Makeup Tank level decreasing rapidly due to Operator Personal high rate of makeup to RCS. Operators opened Statement BWST suction valve on "A" side (SFV-25003). 04:1$:04 Operators started "B" HPI pump to increase IDADS Print Out reactor coolant inventory from SWST. i 04:16:02 Operators tripped both MFW pumps. IDADS Print Out/ ) Operator Personal i The operators recognized the beginning of Statement f-* an overcooling transient due to the failed open startup and MFW valves, the half open 3 l TBVs and ADVs, along with main feedpump J speed remaining at around 2500 RPM. The I operators tripped MFW pumps per step 4.0 of Emergency Procedures Section E.02. The operators may have also noticed the feed-water flow indication remaining high on the control room strip charts at about 3.5 million pounds per hour.

However, 1

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= y._, 'ESCRIPTION OF EVENT DATA SOURCE .. [ TIE D ~ ~ this feedwater flow indication is passed through modules powered by ICS power and f it had also failed to midscale. The w.. actual main feedwater flow rat'e indicated i by the IDADS Print Out decreased to zero l upon reactor trip and had not begun increasing again before the reactor operators tripped the MFW pumps. The-actua'l main feedwater flow rate remained, at zero due to the increased pressure in both SGs and the low speed demand to both MFW pumps. At about this time, two (2) of the operators verified that AFW flow was high indicating greater than 800 gpm~to each SG. This is verified by the IDADS Print Out of auxiliary feedwater flow rates which show that.AFW flow to both steam generators was greater than 1000 gpm at the time of the MFW trip. 04:16 :1 The operator secured pegging steam to ensure / Operator Personal that pegging steam did not contribute to Statement overcooling the RCS.- Pegging steam had contributed to recent overcooling at Rancho Seco. i 04: 16:57 RCSpressurehasdecreasedto1605psig. Pressurizer level is 20 inches. SFAS auto-i matically initiates. "A," "B," "C," "D" HPI injection. valves travel to pre-throttle position. Selected SFAS equipment;. including motor-driven AFW pump P-319, trips and block loading of SFAS equipment begins'. AFW SFAS valves travel fullopen. "A" and "B" DHR pumps autostart. Diesel generators autostart, but do not close into vital busses as there has been no loss of power to the vital busses. 04:16:59 "A" HPI pump autostarts from SFAS signal. IDADS Print Out 04:17:10? Operator closed AFW SFAS valves which were Operator Personal fully opened by the SFAS actuation. These valves are in parallel with the already open ICS controlled AFW valves. Closure of these AFW SFAS valves is shown on the auxiliary feedwater flow rate versus time IDADS plot. Within about 10 seconds after the first AFW flow rate decrease, due to SFAS load sheddincj of the motor-driven AFW pump, a second smaller decrease is indicated which is attributed to closing of the AFW SFAS valves. 04:17:15 "A" & "B" CR/TSC Essential HVAC unit start from IDADS Print Out-the SFAS signal. '04:17:27 Motor-driven AFW pump, P-319 is loaded back on IDADS Print Out on its vital bus and immediately restarts. The dual drive AFW pump has been runni.ng continuously .and. powered by steam since it started a few moments after the loss of ICS power due to de-l creasing f1FW discharge pressure of less than l 850 psig.

~ a w -.=.:.=.=-.=u.- b N F ' TIME DESCRIPTION OF EVENT DATA SOURCE { ':.' The restart of the. motor drive AFW pump is con- ~ C ... I. - /. k firmed by the. plot of aux. feedwater flow rate / 1 S_..- ~ versus time. This plot shows the "A" aux. feed- + I. ater flow lat full. scale shortly after thepump'is restarted,at 1 aux. ~ ~ w i ';~ feedwater flow continues to increase to about .n 1250 gpm. = 04:18:58 . RCS Temperature goes below 500*F. One RCP should have been stopped by procedure to avoid core lift concerns. IDADS Print Out. j ] .h 04:19:15 Operators secured A-CR/TSC essential HVAC. IDADS Print Out' e 04:20:00. Pressurizer level offscale low. Subcooling margin is 85*F and increasing. IDADS Print Out- ?- p The IDADS pressurizer level point is non tem-erature compensated. However, the control room strip chart of pressurizer level does show temperature compensated level. A comparison of these two show that the IDADS pressurizer level correlates very clogely with the control room strip chart once pressurizer level re-turns on scale. i g ? Operator sent technician to look at the ICS power. Operator Person. Statement The technician reported that all four ICS j power supplies were de-energized. The ABT had not transferred and was still on the "C" i bus, and the "J" bus power was still avail- [! able. It was later determined that the ICS "J" bus loads had earlier been connected to [ the "F" power bus. The "F" power bus also supplies power to the TSC Fire System (which

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alarmed upon reactnr trip) and portions of l the security lighting (which was indicated by security personnel to be momentarily out 5 about the time of reactor trip). ~ IO SG pressures have decreased to 500 psig. IDADS Print Out 04:20: r At this pressure the running condensate pumps l' ~ ~.. began to supply FW to the SGs. This is in-4 dicated on the main feedwater flow versus time IDADS plot. This, added approximately 1000 gpm feedwater to each SG for c coup'e ahWe n l,~;,k. I i

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"TIlk DESCRIPTION OF EVENT DATA SOURCE e

r l- 04:21:25 Minimum RCS Pressure of 1064 psig. IDADS Print Out 04:22:00 Violated B&W interirn brittle fracture curve. IDADS Print Out ) 04:22:50 SG pressures have decreased to 435 psig. IDADS Print Out Main Steam line failure logic closes the startup and main feed valves. FW flow l - from condensate pumps is stopped. } f 04:23 ADVs and TBVs isolated by operators. Operator Personal ~ Statement 04:23:10 "B"'AFW control valve partially closed by IDADS Print Out/ operator #1. Operator Personal Statement i> The operator though he had completely 0 closed the vlave at this point. Feed-l water flow to the "B" SG, however has decreased by about 60%. . 04i25:30 Operator unisolated HPI pump SFAS recirc. .IDADS Print Out i valves. Opening recire, path to Makeup Tank. 04:26:15 CR/TSC ESS HVAC train "B" is secured. IDADS Print Out Operator notes that signal to start it in the high temperature /high rad' level mode was present. l '+ 04:26: 7 Operater attempts to close "A" AFW control Operator Personal l valve. Statement. {04:26:22 "A" AFW valve closed. Operator believes it Follow up Interviews. is only 80% closed. Leaves to locate valve wrench. l04:26:47 Pressurizer level back on scale and increasing. IDADS Print Out i' Subcooling margin is 170*F. Started to throttle HPI injection valves to minimize repressurization and further cooldown. 04:28:00 Makeup Tank level off-scale high. IDADS Print Out 04:28:00 ManuallystoppedRCP-Cpercorelhtrequirements. IDADS Print Out Excessive RCS flow may give an excessive lifting force to the core components (fuel assemblies). ! 04:28:59 HPI Pump "A" was manually stopped. IDADS Print Out I I J

g ._m 3 1 4,._f- ' ' TIIE DESCRIPTION OF EVENT DATA SOURdE l l I 04:29:40 Operator uses valve wrench on "A" AFW valve. Operator Personal . Manual operator is damaged. Valve re-opens. Statement Operator calls Control Room and is told to l close manual isolation valve, FWS-063. t I 04:29:45 Operator closed "C" & "D" HPI infection valves. IDADS Print Out l Reducing flow to stop the increase of subcooling margin. SCM=190*F. > 04:30 M/U Tank level was high, so closed SFV-25003; Operator Personal "A" Loop BWST Suction Valve (which had been Statement /IDADS f opened earlier). This isolated the suction Print Out of the M/U pump and the "A" HPI pump and the "A" Decay Heat pump. 04:30 Declared Unusual Event. Notified State and Operator Personal County Agencies. Statement 04:30:30 Started depressurizing RCS to return to condition Operator Personal outside PTS region using normal pressurizer spray. Statement & RC Pressure Plot l 04:33:20 Operator #2 arrives at "B" AFW valve and finds it Operator Personal partially open. He closes it all the way. Statement /IDADS Feedwater to the "B" OTSG has been s, topped. Print Out 04:33:40 "A" OTSG is full up to the top of the steam IDADS Print Out shroud and begins to spill water into the steam annulus. 04:36 Operator has attempted to close FWS-063, but it Operator Personal will not move, even with valve wrench. Statement 04:39:00 RCS subcooling margin reaches peak of 201*F IDADS Print Out and begins to decline. 04:40 ICS power restored. Operators closed breakers Operator Personal l S1 and S2 in ICS cabinet 3. All ICS demands Statement /IDADS [ went to 100%. Operators took manual control Print Out of demands to regain control. Operators regained control of ADV, TBVs l and AFW valves (Bailey). All these valves were immediately closed manually. All AFW flow to both OTSGs had ceased. RCS begins F to heat up. Lowest RCS temperature of 386*F was reached and at this time RC pressure is (1413 psig) being reduced to achieve conditions outside the PTS region. j \\ 1

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-7 -. =,. l. ' TIME DESCRIPTION OF EVENT DATA SOURCE i l ..w......... l 04:41:00 7 Operator calls. Control Room and informs them that Operator Personi- ,FWS-063 is stuck open. Told to reopen AFW yalves, Sta.tement told other operators to unisolate TBVs. f I 04:41 Minimum OTSG pressures reached."A" : 221 psig, IDADS Print Out, "B": 202 psig. t 04:41:10 A-0TSG level goes below steam shroud. IDADS Print Out 04:42:42 Shutdown B-HPI pump. M/U pump continuing to run. IDADS Print Out t 04:42:56 ' Closed A & B HPI injection valves. No makeup to IDADS Print Out RCS except RCP seal water. 04:43:50 Operator noted loss of RCP seal injection flow. Operator Person: Statement ) l 04:43:54 B-HPI pump restarted. IDADS Print Out G4:50:19 B-HPI pump shutdown. IDADS Print Out 04:50:30 Operator again notices loss of seal flow. Operator Persant j Restarts B-HPI pump. Statement j 04:52 SRO collapses in front of control panel. Operator Personc Statement l l I e f e l l l 1, I l i h ll b

f ' T.' .TIhE DESCRIPTION OF EVENT DATA SOURCE ). ~ ~~ .05:00 Operator in Control Room hears loud noise. Looks Operator' Persona ~ ~ JEM. down at makeup pump anineter and notes it is read-Statement ing about 1/3 of normal running current. He ~ " ' - realizes makeup pump has been damaged due.to lack of suction. 05:00:10 Operator trips makeup pump End opens makeup tank Operator Persona ~ outlet valve. Water spills out cf the makeup Statement' pump seals onto. the pump room floor. The valve. Approximately_450 gallons is spilled. is reclosed. 7 05:05 Cro:: sed out of B&W recommended Interim Brittle Operator Persona Fracture Limit region; 3-hour soak is in progress. t Statement / i IDADS Print out 05:05 Ambulance is called for SRO. Operator Persona Statement h 05:09 Both AFW pumps are manually secured while OTSG level IDADS Print out is reduced to reestablish normal main feedwater. - 05:27 Makeup pump isolated. Operator Persona Statement 05:29:04 Operators stopped the "A" RC pump. IDADS Print out I l 05:33 Smoke detector locks out Rad Waste Area Exhaust Operator Persona Fans. Operator repeatedly tries to reset these Statement .~ breakers. Later, lock out clears and exhaust fans are started. 05:40 Main steam line failure logic is inhibited. This Operator Persona i permits the normal feed flow pathway to be used. Statement ' 06:06:00 Bypassed Safety Features. Operator Persona Statement 06:11 Momentary ICS power supply alarm. Operator Persona-Statement 06:14 Loss of ICS power. Operators immediately reset Operator Persona breakers S1 and S2, in cabinet 3 to restore ICS Statement power. 07:0d SRO released from hospital. Operator Persona-Statement 08:41. Terminated Unusual Event. Emergency Coordi ator Log l I .i l

Y r .a ]. SACRAMENTO MUNICIPAL UTILITY DISTRICT i OFFICE MEMORANDUM Januaby4,1986 TO: Distribution DATE: TS 86-006 FROM: R. P. Wichert / M, i susJECT! REVISION 2 to 12/26/85 TRANSIENT SEQUENCE OF EVENTS The Sequence of Events relative to Reactor Trip #75 has been revised h and is attached. This revision provides additional information as to plant conditions concurrent with specific events while clarifying certain events. It is anticipated that this Sequence of Events will be further revised whenever sufficient or pertinent new information becomes available. Revisions have been indicated with Revision marks. cc: Trip Report Distribution Glen Perez, NRC Resident Inspector Control Room George Coward Steve Redeker Bill Spencer Ron Lawrence Charlie Linkhart Fred Kellie Jim Jurkovich f' Val Lewis ? e. i I 1 ~ 2j 1 1 i \\

r w _.- 01/04/85 -', ^i ' 1600 Hours Rev. 2 i CHRONOLOGICAL SEQUENCE OF EVENTS e N 12-26-85 s, Initial Conditions Unit operating at steady state power of 76%, 710 MW(e) Integrated control system in full automatic Bailey Computer out of service (one of the plant's two main computer systems in the Control Room) Transient Initiator Time Source Event / Action 04:13:47 IDADS Loss-of ICS power. All lights on the ICS stations go out. All l l demands fall to 50%. " Loss of ICS Power" annunciator alarms. G 9 l 9 3 e

. ~ s

SEQUENCE OF EVENTS

~ '-lFollowingdevelopedfromIDADSPrintOut. Alarm Typer and Operator Interview. Control Room clock.1was found to be Time adjusted to IDADS Print Out time. about (1) minute ahead of IDADS time. TIE DESCRIPTION OF EVENT DATA SOURCE D 04:13:47 "koss of ICS Power" Annunciator Alarms. IDADS Print Out Upon a loss of ICS power all ICS demands go ~ to midscale, corresponding to zero volts. (The ICS works on i 10 volt scale, zero volts being 50% demand.) The startup and main F.W valves closed to 50% demand because of this decrease in demand signal. The MFW pump speeds were reduced to low speed. Deenergizing the ICS also causes the Main Feedwater block valves to close. This expected response of the feedwater system is illustrated by graphs of Main Feedwater pumps speed versus time, main feed pump discharge pressure versus time and Main Feedwater flow rates, all.of which indicate a decrease just prior to reactor trip. With the plant initially at about 76% full-power, this reduction in feedwater flow is expected to increase RCS pressure. The loss of ICS power also sends a demand to the Bailey AFW control valves and ADVs and TBVs to open to 50% demand. The effect of the ADVs and TBVs opening to about 50% is indicated on the plot of OTSG pressure versus time. The IDADS plot clearly shows SG pressure decreasing prior to the reactor trip. 04:13:7 Operators notice MFW flow decreasing rapidly. Operator Personal Also noticed RCS pressure increasing. Operators Statement opened pressurizer spr'ay valve in an attempt to stop RCS pressure increase. Due to the overheating of the'RCS by the reduction in MFW flow, the actuation of pressurizer spray is not sufficient to reverse RCS pressure, as shown on the IDADS Print Out of RCS pressure. e 9 0 0 e 5 6 -,w w r y v T'

y, P I f .c TIME / DESCRIPTION OF EVENT DATA SOURCE / 04:14:01 The low main feedwater pump discharge pressure IDADS Print Out of less than 850 psig automatically started the electric driven AFW pump, P-319. 04:14:03 Reactor trip on high RCS pressure. The turbine IDADS Print Out trip is also initiated by the reactor trip. Operator closes pressurizer spray valve. 04:14:04 Peak RCS pressure at reactor t Ap of 2298 psig IDADS Print Out 04:14:06 AFW dual drive pump, P-318, autostkr'ts on low IDADS Print Gut feedpump discharge pressure (850 psig). AFW pump,.P-318 was steam driven throughout this transient. 04:14:06 Peak RCS hot leg temperature of 606.5'F IDADS Print Out i 04:14:7 Imediately upon reactor trip, many fire alarms, Operator Personal TSC spray actuation alarm, seismic trouble alarm, Statement l and SFP temperature high alarms were received. The operators performed the actions of the Emergency Procedure Section E.01. This included reducing letdown flow. Operators then proceeded with Emergency Procedures Section E.02. 04:14:11 AFW flow began to both OTSGs. IDADS Print Out 04:14:25 Operators performing procedure E.02, Vital Operator Personal System Verification due to pressurizer level Statement /IDADS decreasing, fully opened "A" inject valve Print Out for.more makeup addition to RCS. 04:14:30 Operators sent to close TBVs, ADVs, and AFW Operator Personal flow valves. Statement 04:14:48 Makeup Tank level decreasing rapidly due to Operator Personal high rate of makeup to RCS. Operators opened Statement BWST suction valve on "A" side (SFV-25003). 04:15:04 Operators started "B" HPI-pump to increase IDADS Print Out reactor coolant inventory from SWST. 04:16:02 Operators tripped both MFW pumps. IDADS Print Out/ Operator Personal The operators recognized the beginning of Statement h an overcooling transient due to the failed open startup and MFW valves, the half open TBVs and ADVs, along with main feedpump speed remaining at around 2500 RPM. The . operators tripped MFW pumps per step 4.0 of Emergency Procedures Section E.02. The-operators may have also noticed the feed-water flow indication remaining high on the control room strip charts at about 3.5 million pounds per hour.

However, i

~ a.,, 2- _.;_=-.=. ' TDE DESCRIPTION OF EVENT -DATA SOURCE t _'. - this fc<dwater flow indication is passed ' ~ through modules powered by ICS power and it had also failed to midscale..The actual main feedwater flow rate indicated by the IDADS Print Out decreased to zero upon reactor trip.and had not begun ) increasing again before the reactor i op.erators tripped the MFW pumps. The actua'l main.feedwater flow rate remained at zero due to the increased pressure in both SGs and the low speed demand to both MFW pumps. At about this time, two (2) of the operators verified that AFW flow was high indicating greater than 800 gpm to each SG. This is verified by the IDADS Print Out of auxiliary feedwater flow rates which show that.AFW flow to both steam generators was greater than 1.000 gpm at the time of the MFW trip. 04:16 :1 The operator secured pegging steam to ensure ! Operator Personal that pegging steam did not contribute to Statement overcooling the RCS. Pegging steam had contributed to recent overcooling at Rancho Seco. 04:16:57 RCSpressurehasdecreasedto160bpsig. ~ Pressurizer level is 20 inches. SFAS auto-matically initiates. "A," "B," "C," "D" HPI injection. valves travel to pre-throttle position. Selected SFAS equipment;. including motor-driven AFW pump P-319, trips and block loading of SFAS equipment begins. AFW SFAS valves travel fullopen. "A" and "B" DHR pumps autostart. Diesel generators autostart, but do not close into vital busses as there has been no loss of power to the vital busses. 04:16:59 "A" HPI pump autostarts from SFAS signal. IDADS Print Out 04:17:107 Operator closed AFW SFAS valves which were Operator Personal fully opened by the SFAS actuation. These valves are in parallel with the already open ICS controlled AFW valves. Closure of these AFW SFAS valves is shown on the auxiliary feedwater flow rate versus time IDADS plot. Within about 10 seconds after the first AFW flow rate decrease, due to SFAS load shedding of the motor-driven AFW pump, a second smaller decrease is indicated which is attributed to closing of the AFW SFAS valves. 04:17:15 "A" & "B" CR/TSC Essential HVAC unit start from IDADS Print Out j the SFAS signal. ~ 04:17:27 Motor-driven AFW pump', P-319 is loaded back on IDADS Print'Out l on its vital bus and immediately restarts. The dual drive AFH pump has been running continuously and powered by steam since it started a few moments after the loss of ICS power due to de- . creasing f1FW discharge pressure of less than / / 850 psig.. _ _ __

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~ N.. . ~ DESCRIPTION OF EVENT. DATA SOURCE . TIME '..~.E.. i The restart of the motor drive AFW pump is con-f . 5. !. firmed by the plot of aux, feedwater flow rate ._s_ versus time. This plot shows the "A" aux. feed- - 5j -- - water, flow l at full. scale shortly after the (f ~ I pump is restarted, at 1300 gpm and the "B" aux. feedwater flow continues to increase to about i 1250 gpm.- 04:18:58 . RCS Temperature goes below 500'F. One RCP should have been stoppe'd by procedure to avoid core lift 5 concerns.. IDADS Print Out i 04:19:15 0perators secured A-CR/TSC essential HVAC. IDADS Print Out 04:20:00. Pressurizer level offscale low. Subcooling margin is 85*F and increasing. IDADS Print Out 3 The IDADS pressurizer level point is non tem-erature compensated. However, the control room strip chart of pressurizer level does show temperature compensated level. A comparison of these two show that the IDADS pressurizer level correlates very clogely with the control room strip chart once pressurizer level re-turns on scale. ~ ~ I

5.. -

r ? Operator sent technician to look at the'ICS po<er. Operator Personi Statement i The technician reported that all four ICS power supplies were de-energized. The ABT had not transferred and was still on the ";" ). bus, and the "J" bus power was still avai7-able. It was later determined that the I,5 "J" bus loads had earlier been connected to the "F" power bus. The "F" power bue also supplies power to the TSC Fire Syr'.em (which alarmed upon reactor trip) and pr etions of the security lighting (which was indicated by security personnel to be momentarily out about the time of reactor trip). '4:20:20 SG pressures have decreas'ed to 500 psig. IDADS Print Out 0 At this pressure the running condensate pumps began to supply FW to the SGs. This is in-dicated on the main feedwater flow versus time IDADS plot. This,added approximately 1000 gpm feedwater to each SG for c cc @ e aht/fc n k ~;b. c t lj.

4, m. . '#N D'SCRIPTION OF EVENT DATA SOURCE E ~ TI \\ 04:21:25 Minimum RCS Pressure of 1064 psig. .IDADS Print Out 04:22:00 Violated B&W interir6 brittle fracture curve. IDADS Print Out ) IDADS Print Out 04:22:50 SG pressures have decreased to 435 psig. Main Steam line failure logic closes the startup and main feed valves. FW flow from condensate pumps is stopped. 04:23 ADVs and TBVs isolated by operators. Operator Personal Statement 04:23:10 "B"'AFW control valve partially closed by IDADS Print Out/ operator #1. Operator Personal Statement The operator though he had completely l l closed tne vlave at this point. Feed- / water flow to the "B" SG, however has decreased by about 60%. 04:25:30 - Operator unisolated HPI pump SFAS recirc. IDADS Print Out valves. Opening recirc. path to Makeup Tank., 4.04:26:15 CR/TSC ESS HVAC train "B" is secured. IDADS Print Out i Operator notes that signal to start it in the high temperature /high radlevel mode was present. h 04:26:7 Operator attempts to close "A" AFW control Operator Personal valve. Statement. 04:26:22 "A" AFW valve closed. Operator believes it Follow up Interviews - is only 80% closed. Leaves to locate valve wrench. 04:26:47 Pressurizer level back on scale and increasing. IDADS Print Out Subcooling margin is 170*F. Started to throttle HPI injection valves to minimize repressurization ~ and further cooldown. l 04:28:00 Makeup Tank level off-scale high. IDADS Print Out ~ 04:28:00 Manually stopped RCP-C per core lift requirements. IDADS Print Out Excessive RCS flow may give an excessive lifting force to the core components (fuel assemblies). k04:28:59 HPI Pump "A" was manually stopped. IDADS Print Out i = [ i I 1 .i--.------,,._,- ,7,-, ,m,,,,-~.,+-v,, ,,,,,.,,,,,,y- ,3,-,#w-, +-, ---,. -. -w,,3,- y-

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TIE DESCRIPTION OF EVENT DATA SOURCE 04:29:40 Operator uses valve wrench on "A" AFW valve.

Operator Personal . Manual operator is damaged. Valve re-opens. Statement Operator calls Control Room and is told to close manual isolation valve, FWS-063. Operator cl' sed "C" & "D" HPI inftetion valves. IDADS Print Out 04:29:45 o Reducing flow to stop the increase of subcooling margin. SCM=190*F. % 04:30 M/U Tank level was high, so closed SFV-25003; Operator Personal "A" Loop BWST Suction Valve (which had been Statement /IDADS opened earlier). This isolated the suction Print Out of the M/U pump and the "A" HPI pump and the "A" Decay Heat pump. 04:30 Declared Unusual Event. Notified State and Operator Personal County Agencies. Statement 04:30:30 Started depressurizing RCS to return to condition Operator Personal outside PTS region using normal pressurizer spray. Statement & RC Pressure Plot 04:33:20 Operator [2arrivesat"B"AFWvalveandfindsit Operator Personal partially open. He closes it all the way. Statement /IDADS Feedwater to the "B" OTSG has been s, topped. Print Out 04:33:40 "A" OTSG is full up to the top of the steam IDADS Print Out shroud and begins to spill water into the steam annulus. 04:36 Operator has attempted to close FWS-063, but it Operator Personal will not move, even with valve wrench. Statement 04:39:00 RCS subcooling margin reaches peak of 201*F IDADS Print Out and begins to decline. 04:40 ICS power restored. Operators closed breakers Operator Personal 51 and 52 in ICS cabinet 3. All ICS demands Statement /IDADS went to 100%. Operators took manual control Print Out of demands to regain control. Operators regained control of ADV, TBVs I and AFW valves (Bailey). All these valves were immediately closed manually. All AFW ' flow to both OTSGs had ceased. RCS begins to heat up. Lowest RCS temperature of 386*F was reached and at this time RC pressure is (1413 psig) being reduced to achieve conditions outside the PTS region. ~ A n e

m .= .g, l a - - -7 bESCRIPTIONOFEVENT DATA SOURCE TIME a.. Operator calls'. Control Room and informs them that Operator Personi ~04:41:00 ? j ,FWS-063 is stuck open. ToldtoreopenAFWyalves, ~ Sta.tement told other operators to unisolate TBVs. 04:41 Minimum OTSG pressures reached."A" : 221 psig, IDADS Print Out "B": 202 psig. 04:41:10 A-0TSG 1evel goes below steam shroud. IDADS Print Out 04:42:42 Shutdown B-HPI pump. M/U pump continuing to run. IDADS Print Out [ 04:42:56 ' Closed A & B HPI injection valves. No makeup to IDADS Print Out [ RCS except itCP seal water. 04:43:50 Operator noted loss of RCP seal injection flow. Operator Persont Statement i: 04:43:54 B-HPI pump restarted. IDADS Print Out 04:50:19 B-HPI pump shutdown. IDADS Print Out 04:50:30 Operator again notices loss of. seal flow. Operator Person,: i Res. tarts B-HPI pump. Statement 04:52 SRO collapses in front of control panel. Operator Person.- Statement f* e l r 1 p I i ~ [. i ? l i i t l

s 53: .~ l . TIME DESCRIPTION OF EVENT / DATA SOURCE ~ bperato in Control. Room hears loud noise. Looks Operator' Persona- ' ~'d500 .{. t.' " - down at makeup pump arrrneter and notes it is read-Statement ing about 1/3 of normal running current. He - ~ realizes makeup pump.has been damaged due.to 1ack of suction. 05:00:10 Operat'er trips makeup pump *and opens makeup tank Operator Persona outlet valve. Water spills out-Of the makeup Statement pump seals onto. the pump room floor. The valve. .,q,_ Approximately,450 gallons is spilled. is reclosed. i 05:05 Crossed out of B&W recommended Interim Brittle ' Operator Persona Fracture Lirnit region; 3-hour soak is in progress, t Statement / IDADS Print cut 05:05 Ambulance is called'for SRO. Operator Persona Statement 05:09 Both AFW pu'mps are manually secured while OTSG level IDADS Print out is reduced to reestablish normal mai.' feedwater. 05:27 Makeup pump isolated. Operator Persona Statement 4 05:29:04 Operators stopped the "A" RC pump. IDADS Print out i 05:33 Smoke detector locks out Rad Waste Area Exhaust Operator Persona Fans. Operator repeatedly trids to reset these Statement breakers. Later, lock out clears and exhaust fans are started. 05:40 Main steam line failure logic is inhibited., This Operator Persone permits the normal feed flow pathway to be uSEd. Statement , 06:06:00 Bypassed Safety Features. Operator Persona Statement 06:11 Momentary ICS power supply alarm. Operator Persona-Statement 06:14 Loss of ICS power. Operators immediately reset Operator Persona breakers S1 and S2 in cabinet 3 to restore ICS Statement power. 0706 SRO released from hospital. Operator Persona e Statement 08:41. Terminated Unusual Event. Emergency Coordi ator Log I , A_....-. I

REV 3 1 ] RANCHO SECO TRANSIENT 12-26-85 l PRELIMINARY SEQUENCE OF EVENTS ,. a i u.. J. d. I othd 04: 14:49

Loss of ICS Power 04:15:03
Reactor Trip on High Pressure 04:15:08
AFW Pump Autostart; AFW Flowrate at >500 gpm to each OTSG i

04:17:04

Manually tripped both main feedpumps f

04:17:57

SFAS initiation at 1600 psig, Automatic 04:22
Lowest RCS pressure = 1047 psig.. RCS Temperature = 456*F j

04:25

Isolation of Atmospheric Dump and Turbine Bypass Valves completed

{

Secured "A" HPI Pump 0430
Declared Unusual Event e stopped "C" RCP

[ 0435

Makeup Tank level high; closed SFV-25003 ("A" Loop BWST Suction Valve)
Started depressurizing RCS

"" ^ " "" f F - ' 0440

Lowest RCS Temperature = 386*F.

RCS Pressure = 1413 psig, o +{ t,

Regained ICS Power 0444
Secured "ts" HPI Pump 0445
Losing RCP Seal Injection, restarted "B", HPI Pump; manually tripped Makeup Pump 0450
Received Auxiliary Building Stack Rad Monitor High Alarm due to water released from damaged Makeup Punip 0500-
Tarminated RCS pressure decrease; started threc-hour hold at RCS j

pressure 716 psig,. Temperature 433*F 0505

Makeup Pump isolated 0507
Ambulance c.a..lled for SRO, Exhaustion eses

,s u...... c 1 0530

Sto d "A" RCP j

esso .s e 0547

Reactor Bldg. Rad Monitor (R15001 A and 8) 6.kl.

j 0614

Momentary loss of ICS power

. cem ntuw-w 0700

SRO released from h s ital, Satisfactory condition o no s a.)n u u...1

s. w a u'. a,),,,,,,,

0841

Terminated Unusual Event NOTE:

Minimum subcooling n$argin was not lost. e e

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i, m... y R rAC-O R ~~ R 3 J rC = V B r R 26, 985 AUX FEEDWATER FLOWRATES 1.4 0FF SCALE IllGli ~ 1.3 - w f" Q 1.2 - RxTRIPO 1 9 i 1.1 - o p' s i 1-O.9 - 6 h1 a.,A m c4 < u 7 a so O.8 - ot-s a t i ey dv$ open g E m O.7 - La 'Nl % i 3:g ( l Sg O.6 - f g 'f y %) // y s3 p. uv 0.5 - 4I w O 4 ga n Nt '1 J 1 0.4 - L f Ob [t k liv ";>voitt-h eh> C 'yes 4 , of,,jgv O.3 - (u y e' pf:/ce a l" O.2 - cl# [ // # r POSSIBLY INVALID DATA x -i 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME - El-A AUX FEED FLOW --G-- B AUX FEED FLOW ---=---s- - -,- n -n~~-----
,77 , ~ ' , vlT AL BUS IC 1 1 NON-VITAL BUS IJ ,i \\ 'I ) .../ s, ABT 185 VAC 185 VAC _a = r- - - - < i -1 l SI----- S2-------------7 [ g l L_______ l _____________q H l l 1 m o m L____ i l i I O 2 l l l k
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p u H Y Y X X l mh { + 24 -24 +24 -24 C 21 l l CAB # I VOC VOC VDC VDC 8 POWER MONITOR Os O A l l g p _ +_ + 750 ALARM 8 TRIP D' _d2 -_________________j ~ CAB # 3 = I o na, o o n a 08 *U H H ~Z 8 ] l O CAB
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y l g gg g }{ AUCTIONE E RING l -g DIODES 2 4 g M 1 y i COMMON BUS m l l l - 24 VDC BUS l J l 7 + 24 VOC BUS POWER AUCTIONEER CABINE T # 2 y p p p p p p U p d J._ > d._S J d.--S bu CAB *1 CAB # 3 CAB # 5 e M !i I
Tha + 24 VDC power supplies require a 115 VAC power source to produce the required + 24 VDC, which is produced through three basic functional sections. The three sections, shown on Figure 33-1-D1 are: n l(( AC/DC transtormation e Regulato;/0VP circuitry 5 e Pass transistor assembly The AC/DC transformation circ ui try incorporates a step-down transformer to provide the proper voltage and amperage for the rectifier / filter used in this section. The rectifier / filter takes this lower AC voltage and converts and filters it into a clean DC square wave output of approximately 24 VDC. l The 24 VDC output then passes through a regulator /0VP circuit where the voltage is regulated by means ot an integrated ci rc ui t. This circuit tunctions to maintain a constant + 24 volts at the output of the power supply, i The over-voltage protection circuit (OVP) senses the output voltage ,~{' of the power supply, and, if it exceeds a preset level, utilizes a silicon controlled rectifier (SCR) to short-circ uit or " crowbar" the circui t. This protects the downs tream components and the power 1I supply by blowing the fuse on high output voltage of the power supply, i The pass transister assembly is provided to limit the current provided to limit the current provided by the power supply. The current output is dependent on the number and total dissipation (set internally) of the pass transisters employed. l
2. Power Supply Monitor The power supply monitor (PSM) is a single * ' prin ted circuit board module located within ICS cabinet #2.
Shown on Figure 32-I-C2, the PSM monitors all four power supplies (X + 24 VDC and i + 24 VDC) and receives its inputs through the power actioneer circuit. l. I 32-17 i
\\: { 5 l The power supply monitor module monitors the output voltage levels of each + 24 VDC power supply and each system power bus, it provides ~. ~ ~ - j relay contact outputs indicating a normal (relay energized) or fault (relay de-energized) to switches 81 and S2 for: The positive power supplies The negative power supplies e lhe positive power bus [ e The negative power bus r e L Indicator lights (inside the ICS cabinet) associated with each relay g are installed to provide the operator with information regarding the f status of the relay. The lights are on during normal conditions and off during fault et 7ditions. f The power supply monitor provides both trip (S1 & S2) and alarm functions. The relay switch setpoints, and thus the trip and alarm setpoints, are independently adjustable. The internal construction of the power monitor consists of a printed circuit board with various components, including integrated circuits and relays. An integr ted c ircui t comparator circui t compares the input voltage to the circuit to a fixed reference voltage. The comparator will then function as a switch to apply a voltage to a relay driven transister based on wheather the input is above or below the reference voltage level. The ap propriate relays energize to close contacts energizing trip coils or energizing the alarm circuits ( associated with power failures. The bus monitor relays in the power supply monitor have two sets of normally closed contacts. These contacts are normally connected in series with the shunt trip coils of switches S1'and SZ on the power auctioneer panel. If either the positive or negative bus voltage falls below the trip setpoint, either the positive bus relay or the negative bus relay will de-energize, completeing the shunt trip coil circuits on both switches S1 and S2 If bus voltage falls and does not recover within U.5 seconds, Si and S2 will open removing 115 VAC power from all four power supplies. Switches S1 and S2 must bc manually reset to restore power. TVo other relays associated with 32-19
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f0hOf g l l l l. i i t1AtlAGEI) I NUCLEAR OPEf1ATIONS i s George Coward l .v'O ASST. f1GR. 10 CLEAR G'ERATICNS I s (Acting).ktn ikColligen I I I I I I ~ AGGNISTRATICN CatFLIANE SCEDIA.ING PUNT SLPT. 1ECmICAL SUPPORT l i. Wende Wells Ron Colosto Jim Shetler (Acting) Steve Redeker Jim Field ~ ASST.1D PLANT - gupy, ~ w - -(Actire) Den Whitney I I I l l I&C ELECTRICAL OPERATIONS
1ECHANICAL RAD /Det StPT.
Norm Brock D urlie Linkhart Bill Spencer Ron Lawrence Fred Kellie j i i N e ? e 4 0 'E' -~ _...... -,.. -
. -. -. ~..... - - ---....- ~ ( . fj, igi.Et2i -:...g,:j y,- sg. h M..,w.;.~'y,, % e. .'..i, U. S. NUCLEAR REGU!ATORY COMMISSION O' 3 5 OFFICE OF INSI'ECTION AND ENTORCEMF. T REGION V Report No. 50-312/80-13 Docket No. 50-312 t.icense tio. OPR-54 Safeguards Group Licensee: Sacramento Municipal Utility District P. O. Box 15830 ..J1 ' Sac'ramento, California 95813 h ~ Facility Name: Rancho Seco a / .Inspectkonat: Clay Station, California Inspection conducted:. lay 5 to 9,1980 i Inspectors: 1d ) / (.. 2, / hh GQp.' {@hig, @!adtor Inspector (} Ste' Signed Date Signed ^ Date Signed ApprovedByb.bB. H. Faulkenberry,k n M_ 7/2/ft) 'fQ Date Signed Reactor-Project Sec ion 2, Reactor Operations and Nuclear Sumary: Support Branch Inspection on '4ay 5 to 9.1980 (Report No. 50-312/80-13) Areas Inspected: Routine, unannounced inspection of the Document Control Program; Onsite Review Comittee Operations; the Reactor Operator Requalification Program; and followup on an IE Bulletin. hours onsite by one inspector. The inspection involved 36 inspector-Resul ts: Ho items of'noncomplianen or deviations were found. 1. i I RV Form 219 (2) gNi"[hL I 'l , /$
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~" ,.i 's DETAILS. 1. Persons Contacted P. Borchers, Site Decument Center Supervisor. ,',' m ' R. Colombo, Technical Assistant
G. Coward, Maintenance Supervisor

J. Mau, Training Supervisor
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H. Heckert, fluclear Engineering Technician,
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T. Tucker, Shift Supervisor
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H. Carter Outage Scheduler'
- *J.. Sullivan,. Senior Quality Assurance Engineer
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R. Miller, Chemistry and Radiation Control Supervisor.
- J. King, Shift Supervisor- ,o y, .c The inspector also interviewed a number of other licensee employees including a control room operator, an: instructor and document clerks.
Denotes those present at exit interview.
2.. Occument control program The inspector reviewed licensee procedun! ECP-1, " Rancho Seco Configuration Control," and verified that~ controls had been established for distribution,,, of updated drawings to the plant. site. For completed work, these updated drawings would consist either of marked-up drawingc printed on yellow paper or (later) aperture cards for the drawing as revised by the drafting i department. In addition, when changes. were outstanding against a drawing 7;- these.were noted on, the. aperture card. 5 The Site Document Center maintains controlled stick files of drawings at eleven locations onsite. The licensee'.s representatives at the Site ., Document Center stated that first priority in' distribution of updated '. drawings was given to the files in the Control Room,. - The inspector did.not identify formal procedures for removal of obsolete drawings from the controlled files. The licensee's representative stated. however, that personnel who deliver the updated d a ir w ngs to the controlled - stick files are instructed to remove the obsolete drawings when new drawings-are inserted. The inspector's examination of the controlled stick files indicated that although there were a few. instances where obsolete drawings had not been removed when the latest drawings were inserted, the instructions for-removal of obsolete drawings had been followed in most instances. The inspector recomended that an increased effort should be made to assure., that all obsolete drawings. arel removed f. rom the controlled stick files. The inspector verified that controls were provided for correcting identified discrepancies between plant drawings and the as-built condition of the i l plant. The principal mechanism for effectinq such corrections is Quality..,.: , Assurance Procedure (OAP) No.17. "flonconforming Materia.1 Control." Thi s ' 'c' \\ c, i ,,r y '., ;f.,, fs [ [' ~ ' ' y t{k. ' <~f 4 i I ne ..c. o- [ f, ' e, . ; Q,.[ 7 '- t ,['f,
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...w= 'q . ;I. d. ..?. [ N. "I,, '"'.L5.N'h'ollnwuponIEBulletin [ .The inspector examined the licensee's actions with respect to the following bulletin: IE Bulletin 79-27 (closed) a. 4 The licensee initially' responded.to this bulletin by letter dated . February 22, 1980. This response was' supplemented by the licensee's S letter of March 12, 1980. By letter. dated April 14, 1980 the NRC ' issued c order confinning the licensee's comitments to implement the actions ' described in the letter of March 12,'1980 prior to resumption. of operation following refueling. Impleraentation'of these commitments". .was verified by the Resident Inspectors and..is documented in Inspecti,on Reports 50-312/80-16.and.50.-312/80-17. 6. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncom-pliance, or deviations. An' unresolved. item disclosed during the inspection is discussed in Paragraph 3. 7. Exit Interview The inspector met with licensee representatives (denoted. in paragraph 1) at the conclusion of the inspection on flay 9,1980. The inspector .surrarized. the purpose and the scope of the inspection and the findings. 'The findings were acknowledged 'by the, licensee..- [ s + s -u-s e t 9 a O e.. .y= ,w ;,1,. t I t 5 t , ' k,;,"
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Q i , SACRAMENTO Y.UNCIPAL ufluTY DISTRICT C 6201 S street. Bos 15830. sacramento cahfornia 95813; 1916) 452 3211 r March 21,1980
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+ Director of Nuclear Reactor Regulation Attention: Mr. Robert W.'Reid, Chief 1 r-N Operating Reactors, Branch 4 e* .o... U. S. Nuclear Regulatory Comissi_on. m. Washington, 0."C.- 20555 i' t i + t- . Docket'50-312 '^ - (? ' 1 Rancho Seco Nuclear Generating ii Station, Unit 1 ~ i.
Dear Mr. Reid:
In response to the three telecopied questions submitted by Mr. M. Fairtile (NRR) to Mr. D. Raasch (SMUD), your attention is directed to the Wm. C. Walbridge letter with attached report to Harold Denton dated March 12, 1980. Each of these questions is addressed in the March 12, 1990 Report as indicated below: Question 1: Actions which will allow the operator to cope with various combinations of loss of instrumentation and control functions. This includes changes in (A) equipment and control systems to give clear indications of functions which are lost or unreliable; (B) procedures and training to assure positive '~ and safe manual response by the operator in the event that competent instruments are unavailable. ,g Respcnse: Refer to Paragraph 31, 3,j,' 4 and 5 of the March 12th Report.' s Question 2: Verification of the effects of various combinations of loss of instrumentation and control' functions by design review i analysis and by test. - ~ Response: ! Refer to Paragraphs 31, 4, and 5 of the March 12th Report. Question 3: _ Correction of electrical deficiencies which may allow the Jpower operated relief valve and pressurizer spray valve to open on non-nuclear instrumentation power failures, such as, the event.which occurred at Crystal River 3 on 2/26/80.
Response
Refer to Paragraphs 6b and.6c of. the March' 12th Report. h(tCO/ s ~ /lO i H-00326096-app. c 8
rc m ___s_. ~: t Mr. Robert W. Reid March 21,1980 The actions provided in our responses to Questions 1, 2 and 3 above will all be completed prior to return to power. If you have any further questions concerning this reply, please contact us. Sincerely, ohn J. Mattimoe Assistant General' Manager and Chief Engineer f g 1 ~t. i p 8 ( g o 8 8 9 9 ~ 1 ( \\ r JI- $e
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/ >), l UNITED STATES 8 " 7. NUCLEAR REGULATORY COMMISSION i y *, i wass m oron.o.c.mossa
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.n s E a / April 14, 1980
...*3 Docket.No. 50-312 f
T. a. f- '. h, '. : r S " Mr'. J. ' J. Ma t t imoe ' '
A.ssistant General Manager and Chief Engineer Sacramento Municipal Utility District 6201 S Street
~ P. O. Box 15830 Sacramento, California. 95813
Dear Mr. Mattimoe:
le The Connission has issued the enclosed Confirmatory Order for the Rancho Seco l Nuclear Generating Station. This Order confirms your connitments, as stated In your letter dated March 21, 1980, to implement certain actions prior to restart from your current outage. } .,This Order f s required as a result of the experience gained from the CR-3 . incident of February 26, 1980, wherein a non-nuclear instrumentation power loss resulted in a series-of unexpected events Your staff and the NRC mutually agreed on the actions provided in the Order. These actions should reduce the probability of a similar future power loss causing unexpected plant responses and allow the plant operator to better cope with losses of instru-mentation and control functions. Further actions may be r.ecessary. A copy of this Order is being filed with the Office of the Federal Register for pubilcation.' incerely,. 4 ]- l T 1j = kWractor .Da e E1 enh Division of ( erating Reactors - Office of Nuclear Reactor. Regulation
Enclosure:
Confirmatory Order cc w/ enclosure: See-next page
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(. 2;4fWU90$dWl'[ F (. o UNITED STATES OF AMERICA' 'fF t f NUCLEAR REGULATORY COMMISSION e i i ~In the Matter of ) ) Sacramento Municipal Utility ) Docket No. 50-312 District ) ) (Rancho 5eco Nuclear ) GeneratingStation) )' CONflRMMORY ORDER
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i .i ir .The Sacramento Municipal Utility District (the Licensee),is the holder of .F acility Operating License No. DPR-54 which authorizes the Licensee to operate the Rancho Seco Nuclear Generating Station (the facility) at power levels not in excess of 2,7/2 megawatts thermal. The f acility is a pressurized water reactor located at the Licensee's site in Sacraniento County, California.. .gg Following the incident of, February 26,1980,'at the Crystal River facility, the NRC staff held meetings with the Licensee, other operating licensees with Babcock and Wilcox (RAW) reactor systems,. and B&W. The meetings were held in Bethesda, Maryland on March 4,17 and 18,1980. These meetings resulted in the development of three. licensee connitments. ' , 1. Actions which will allow the operator to cope with various combinations of loss of instrumentation and control functions. This includes changes in (A) equipment and control systems to give clear indications of functions j which are lost or unreliable; (B) procedures and training to assure post-Live and safe manual response by',the operator in the event that coupetent instruments are unavailable. t s 1 &%g v vm -,W =w wm 3 y w ,--~v m- - - -, 3 yv v O
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'e t a ' j ....,,.'2., Determination of the ef fects of various combinations of Icss of instru-j 'P mentation and control functions by design review analysis and verification,,' F ) ^ by, tes t. , y.,. i :..,e, y /, .3.. Correction of electrical deficiencies which may allow the puwer operated. ' relief valve 'and pressurizer spray. valve to open on non-nuclear instra- ~ ,. men'ation power f ailures, such.as, the event whi_ch occurred at Crystal c River. Unit 3 on Iebruary 26, 1980. 1 The Licensee confirmed by letter dated March 21, 1980, that it'would iglement all ' three,act,.i.uns at its f acility prior..to the restart of Rancho Se.co which is currently.. shutdown for maintenance.and refueling.. The March 21 letter included by reference, i g. tM li,ce,nsee's submittal of. March 12,.1980; the March 12 letter.provided the detailed ,listin[;ofl various combinations..of Lloss of in!trumentation which constitutes,the comit-ment.in. Item 1 and the letter also provided the. specific list of tests which constitutes fthe comi.tment in Item 2. I have concluded that timely implementation of these three ishort t,erm. actions, at operating B&W system nuclear. power plants is necessary to provide i continued assurance of public health and safety. 111 .j-g... .q,, , In. view of the Igortance.of.this matter _I have determined that these comitments be, formalized by order.and that the public health, safety and interest require ttyat., i j t.his,Urder, be made immediately. ef fect ive.. Accordingly, pursuant to the Atomic Energy Act of.1954, as amended, and the Commission's regulations in 10 CFR Parts 2 and 50, If 15 lilREBY ORDERED, EFFECTIVE _IMMEDIATELY, THAT: The Licensee. prior to restart af ter'the, current outage will implement s .A .. ~ all'.three actions provided 'in Part.I t '.of this Order. '!V i a + ' Any person who has an, interest affected by this Order may request a hearing within i f trenty-five days of the date of the Order. Any request for a hearing will not stay the effectiveness of this Order. Any request for a hearing shall.be submitted to the Director, Of fice _of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Comission, .]. "i t =
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2 S.4,7 o.,.,3.. r my s -s Washington, D. C.,'20555.,~ with a. copy.jo the, Execut.t ve Legal Director at the above address. If a. hearing.is requested by a person who has an interest affected by this Order, the Coi.inission will issue an Order designating the time and place of any such hearing. i, In,the event.any. person who has.an interest affecte.d by this Order , requests' a: hearing as. prov.ided above and a hearing lis held,' the issues to be .t. considered.at such a hearing shall.be: 4 ' ' :i n,. ) : f; ~J .... :' r, < ~- (1). Whether the. facts ~ set forth in.Part II.of'this Order provide v. .. c, an adequate, basis for the actfans ordered,'.and e
f..(2), Whether.the Licensee should perf onn the actions required by Part 111 o
of this Order in accordance with the schedule stated therein. Operation of the facility on tern [s consistent with this Order is not stayed by the pendency of any. proceedings on the Order. FOR THE. NUCLEAR REGULATORY COPfliSSION l h/4th$ _ +:_- - Harold R. Denton, Director r Office of Nuclear Reactor - Regulation Dated at Bethesda, Maryland- . this 14thday of April 1980. [
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~/ 'l C ^ .v. 9. w. :xq-= } - ~ -Y . !se ne *\\ UNITED STAT 88 -E' ... NUCLEAR REGULATORY COMMISSION i .5 MEolON V 09 , ' lb,T ' :. O.,' 1990 N. CALIFORNIA BOULEVARD e,,e /, WALNUT Cit t E M.CALlP oRNI A 94696, SulTE 303. WALNUT Cnf tM PLAZA' g. e
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. :s ~ M i'; ?.: :.. ',. -, e,$u.. t..w l ' - ., n % V ". l;:;'I' [f*y., :q.,.... , sl -.,. ;.'< p;yJ, ' y rp i - ej,;. m m-,h.' .,T.6 ki,'. E.. Nf, : in.. f. d...el.iy.Ib,... Sacramento Milnirical,titility, Distr,1ct. m,;Q@,V,. '.2, a p Q..,,,. 3 P. O. Box 15830 ,7:.. l 4'.; Sacramento. Californi.a. 95813 . ' s... ; - m 4 Attention: Mr'. John J.' Mattimoe r,. s Assistant General Manager. and Chief Engineer Gentlemen: h. . ;. <,,.{ "i.[ : x ,y.
Subject:
i ' hAC_ Inspection of Rancho Seco This refers to the inspection conducted by Messrs. H. Canter J. O'Brien. G. Zwetzig and J. Carlson of this office between April 1 and 30.1980 of activities authorized by NRC License No. DPR-54, and to the discussion of our findings held by Mr. H.' Canter with.Mr. R.~ Rodriguez and other members of your staff.at.the conclusion of.the inspection. Areas examir.ed during this inspection are described in the enclosed inspection report. Within these areas, the inspection consisted of selective examinations of procedeces and representative records, inter-views with personnal.; and observations'.by.the 'inspectorf flo items of noncompliance with.NRC requirements were identified within the scope of this. inspection. ~ ' ) In accordance with Section 2.790 of the NRC's " Rules of Practice." i ~ Part 2. Title 10. Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in'the NRC's Public . Document' Room. If this report contains any.infomation that you believe to be proprietary, it is necessary that'ycu submit a ' written application to this office, within 20 days of the.date of this letter, requesting that such infonnation be withheld ~from public ~ disclosure. The applica-tion must include a full statement of the reasons why it is claimed that the information is proprietary. The appl.ication should be prepared so 3-
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.c \\ t that any proprietary information identified is contained'in an enclosure to the application, since the application without the enclosure will also be placed in the Public Document Room. If we'do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.- Shouidyouhaveanyquestionscloncerningthis. inspection,'wewillbe glad to. discuss them,.with.you.' . q + e . y., ; 3 s i.i* ...t,/. Sincerely.: ~ ^ G,;,,; ; L . ' j3.or
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Enclosure:
IE Inspection Report No. 50-312/80-16 Figure 4.1980 Power Supply Modification.. j 4-cc w/o enclosure: e R. J. Rodriguez. SMUD~ i L. G. Schwieger SMUD ,.7 , ie x../.,n. - .. : L. ' i;1 .,c ,..c. .;r a h O :.,'. -., _ -{. " (. L ,u' 4r I ') .. i.-.. t- , - H;c. 9.. T.6. '. j. ;;. c ,,,2 . i..
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^ (!U,_ q >l. l Doe:< *t No. j0-31? License No. OpR-54 Safeguards Group Licenses:.'9arramento Municipal Utility District ,i. P. D. Box 15330 l v.,P ' ' .4.. g g !. .y S cramento. Califoinia 94313 I j n., - Rancho Seco Generatino' Station .m ' ;. m 'i : "' ' ' j Facility'Hame:-. t inspection at: Herald. Califnrnia (Rancho Seco Site) l l I Inspection conducted: - Jpril'l-30,1980 IN[.E.,., N' Inspectors:</, I. f . 'Date Signed } ' i Harvey L. 04hter Senior Resident. Inspection s .I k ~/[f .-( ',m, 6 f S [A l ate Signed John P. O'Brien. Unit Resident Inspection, l D R,,'.. ,_s. 'i/(.s f w ,( 7, 7 L r,...c i d l(April 20-30only)JohnC.crlson.ReatorInspection. Date Signed f jf I .d. W, e t-f Y' j (April 28-30 only) Gerald Zwetzig, Reactor Inspection 'Date Signed t.M Y,v Approved By: c/u/t. ' i- ' u B. H. Faulkenberry. Chief, Reactor Operations.and Date 51gne1 Huclear Support' Branch "1 l t Suninary: ' Inspection between April 1 and 30,1980 (Report No. 50-312/80-16) 1 Areas inspected:' Routine inspections of'long-term shutdown activities; T6Tlowup on a Pentonal Reouest'itemt and, indeoendent inspection effort. l i
The inspection involved 106-inspector hours by the Senior, Resident i
. Inspector and.54. inspector! hours by other NRC. Inspectors. i l .,- Pesults: Of the three areas inspected, no -items o.noncompilance f or devTattons were. identified. ,4 s.:.; - 4 .t e, y, { -- c,. .n .g 4 j .] f j 6.
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Persons Contacted ~ ' ' t
R.' Rodgriguez, : Manager,' Nuclear. Operations
,.h,p P. Oubre'. Plant ~ Superintendent v !J.,,,. D.: Blachly,' Mechanical Engineer: y
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Coleman, Quality' Assurance' Engineer.*M.. Brock, Electrical /I&C Mainte g . Jy"Mk.:d. ~ .t.J ,9 3."
R.c. Colombo Technical-Assistant G.' Coward, Maintenance Supervisor'. :: d '
' g ?W. Ford,' Operating Supervisor ', 't.gi j. D.9,7['W "?,a'.} w,, - ; e i;H..Heckart, Engineering Technician, '. 2.' A g..q.. 3 ; c.,. <. a' . '.,*J.yJ.r.m t t. Scoinr Quali ty he.erance 0.gir. eor t @,4.;s.,.. y ,tJ. . (7 ...McColligan,' Mechanical Engineering Supervisor;.; W , c;;..,,p ~. o,*R. Medina, Quality-Assurance Engineer, 1 ji . R.' Miller, Chemistry / Radiological Supervisor ; q^Nc;.:,.. ' 1
L. Schwieger, Quality Assurance Director-
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J.:Sullivan, Quality Assurance Supervisor,. O,'_
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tD. Whitney,: Nuclear Engineer-J,,. m., - m./; i
} 9, i.< c .,y i B..Wichert, Nechanical. Engineer % a ',1 i.: c v, i The inspectors also talked with and' interviewed several other licensee . employees,. including members of the engineering.. maintenance, operations,. l .c and,. quality assurance (QA) organizations.j " -;s.L- .a , :. :. w .Denotesthoseattending.theExit'Interviewon;. April 30, 1980.; .... l
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Long-Term Shutdown' Activities' y .. The Rancho'Seco plant has been shut-down since January 12,-1980, for t ' Cycle 4 refueling 'and plant modifications." Daily,"as appropriate, the f
inspectors observed control room instrumentation, manning, at.d procedural compliance by operators. Logs.~and operating records were -
examined and the. clea.nliness of.radiationlcontrolled area access points l was observed..</. ..,,.c., 1.,.. ' '.+., ., ; 3 i ..c Several surveillance tests were observed. including' portions of J Non-Nuclear Instrumentation (NNI) Tests'and Diesel. Generator Sequencing i ' Tests..The inspector. queried various l_icensed operators to determine their knowledge of~ recent changes. to' procedures., facility configuration'. 'M and plant conditions.1
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'. : ;. 9 3 G. g.l a f.'h .,9,. ... i 7 .3 i Tours of accessible areas were taken 'to'make independent' assessments ~ ' ' of equipment conditions, plant conditions, radiological controls, security, safety'and adherence to regulatory requirements. Plant housekeepinq conditions were nbserved. especially in the areas of ' '.a,c "., . potential ignition sources..
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..On April 14 1980, the Conritssion issued;a Confirmatory Order concerning 2 j - Rancho Seco's commitments'to implement certain' actions prior'to restart i from the current refueling. outage. The Order was issued as a result of i the experience. gained from the Crystal River 3: incident of February 26, ) ?l980, wherein a non-nuclear instrumentation power loss resulted in a ~ ~ i series of unexpected events. The NRC staff and the licensee developed three comitments, which were.the subject of the Confirmatory Order: .c l (1)'; Actions which will allow the operator to. cope with various l
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combinations. of. loss ofi instrumentation. and control ' functions. v. " n' (a) Equipment and control systems to give clear indications of t
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~ functions which are; lost or;are unreliable. .u, (b) Procedures'and training to assure positive and safe manual f ',a.,
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.m. unavail.iale.. , i; j,.{.> > gp,. i l(2)~ Determination of the effects of arious combinations of loss of - ~ ,v 2. . Instrumentation and control de L ' [....and verifications by test.._, functions.by j ; sign review analysis .JE ", i .L 6 -m z c j ..JJ.(3):. Corrections of electrical deficiencies which~ may allow the Power + 1
",,. Operated Relief Valve (PORV) and pressurizer. spray valves to open on non-nuclear instrumention power: failures, such as, the event i
,, i'.which occurred at Crystal River,' Unit 3, on February 26, 1980. t-::1.v..i . c. m { The following'actionsLwere taken.by[the' licensee.to satisfy the above listed comitnents: r ~ " e ' ? :,(., ~# .(1). ! The Bailey Meter Company '(vendor for NN!'s) supplied installation ~ instructions on.the proper method.of NNI buffer card replacement. l (improper buffer card replacement may have initiated the Crystal River 3 event). All buffer. cards were inspected for proper l 'i installation and,the Bailey instructions were added to applicable procedures. 9 -
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j-l (2) - The Nuclear. operations g*oup developed. rectangular foam rubber: plugs to be'used to insert'in or. cover 4the open.backlighted push %tton;(BLPB) modules whenever the lamp bulb section of the ble is-lifted out..Also, the upper bulb fixture portion of odule is removed to a remote location for bulb replacement. 4 fn g (3).STFb. m perfonned on the original NN!-Y power supply system ~ to m ir/ various design conditions which were used.to supply .inforta.tio.n for. the present modified NNI s.yst.em., i
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'..v',. - e n. . v.. ',(4).Themajority(of.the5.ampfuses=intheNN!!systemwerereplaced. c ,3 M 'with 0.75: amp. fuses.: This should. allow.for faster clearing of ,fx faults and prevent: tripping of the 120 volt'AC: input circuit break,ers r. . ', In.other words, protective device.: coordination.should occur. C .. f.q. ,. q 'M(5). *N!'instrinet power supp1" mndifica'.icns:wcre st'rforrori.designfusesfall W na.- ~ u- ? AC circuits.' Enclosure 1 to:this" report is excerpted from a L,,; 5 S c' March.12. 1980 letter from W.lC2 Walb' ridge, General Manager, to ' x f H. R. Denton, Director of NRR. ^This figure ts a one-line diagram b L-.of the 1980 NNI power supply modifications.. -The key change that s was made to the previous system is that, drawing (NNI-Z)y and monitor-the~ power suppl arrangement on'the right portion of the' is now ~. available upstream of 51 and S2 such that1 he instrument selector t switches.findicating lamps and auxiliary l. relays are capable of receiving power from more than.one source:through an ABT.(Automatic BusTransferSwitch).' M : mig.o a ~ 3 ;;., - .,(6) - New instrumentation has been installed with readout capability on s .the plant computer in the control room. The following is a list. of this ~ instrumentation. All. listed instruments except the uncompensated pressurizer level;are completely independent of both. j ,,NNI-X and NNI.Y. power.*
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,{f-l 4-f. . t Range. ' /. g g.: Uncompensated pressurizer. levels.N 0-320 inches water I f Wide range RCS pressure C;, '0-2500 psig 4 .., r t..., Wide. range' RCS. Loop. A :Te L:
50-650 degrees F
',,'T.. . Wide range RCS Loop B I ' ".' '50-650 degrees F T l ci #. '. RCS Loop. A T ~ - " Yi "h, k.52n-620 degrees F. h@9WT 9..gn W e.>. RCS Loop'B T 520-G20 degrees ~F" - " * * " tr - I OTSGA;5tartUplevel:<. 0-600 inches water
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OTSG B Startup' level .-.0-600 inches water j j:4.,:. OTSG A pressu're .?- 17-af:;!j0-1200psig C' l P OTSG'8 pressure Mc',..., '.' 0-1200 psig. - "i/0-]00*6 .b Makeup. Tank level ,,, Source range nucleat. instruments 10 -10 cp3 y:.. 4
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Available on the computer,' on a spare panel ~ in the computer room adjacent.to the control. room proper.1and on the sh.utdown panel.
Incore thermocouple values with 'a range of 0-2000 degrees F. are also available on the computer in the control room.
The uncompensated pressurizer level inputs,' although not completely independent.of.both NNI-X and NNI-Y are designed so that there will
" always be two ' uncompensated pressurizer levels available for loss ' ~ of NN!-X or NNI-Y..That is, out of four uncompensated pressurizer ,C* 11evel inputs on the control room computer, one is ~ totally independent g,, of NNI-X or?.Y'.:itwo are supplied by NNI-X.and one..is supplied by NNI-Y. .,n. o .-. r.. (7) Ofrectinns.have been placed'on the computer console and in casualty ' ..-procedures directing ~ operators i.o plate ahe'foliewing' points un 'he ' six computer trend: recorders during lo.s,s'of.NNI events:- ,.m a - ..-. r.,. r Uncompensated pressurizer level (independent of flNI-Y and X), Wide range RCS Pressure e ..p.. Wide range RCS Loop A TC pp' . Wide range RCS Loop B T '. m p.. !. m., ; C ~ OTSG A startup level OTSG B startup level; In additica, group 15 on the computer has'been set aside for the na following printouts at one minute intervals (by procedure). t p'. i.,., 7 _..: '.9 - RCS Loop B Hot leg Temp. ~ / 9 -{ ;:q r .Incore thermocouples ~ ,. 4!.;,i,. RCS SFAS (A) Wide Range Pressure, .- ) c i., Hake-up Tank level ,..itt,. s OTSG A&B Outlet Steam Pressure.. e .. s. i... .Vi),!' I" -1., . Source Range N!'s l'
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Pressurizer level (uncompensated) ~. U'[',' .g,; p ~ D-y. 1 ~ .These items print out on the line printer and function independently ]f.. M from the alarm printer, therefore an overload on the line printer
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)lhdue to alarms,on.the alann prirter. is;not,possi le.. ~ 1 'O.(8): Casualty procedures C43, 44, and 45,"were written to handle.the... .i-i loss of NNI-X, NHI-Y, and all NN!'s respectively. Licensed operators ere trained in Parch 1980, on these changes to. procedures and w ..'.the modified plant NNI system.' ', j '(9).. STP-616 was performed to prove independence' 4, '. ' 4-t of>the above listed instrumentation on NN.! power upsets.' ,e ,9. i ; a (10).The PORV.ctrcuitry was modified to close the. valve on NNI failure. (11).The pressurizer spray' valves circuitry was modified such that an NNI ~ power supply. failure should.not cause either of the two spray valves, ?)'..to open. , i.;~ .y
No.1tems of.. noncompliance or. deviations were.identifled.
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U . 3.x. i. .'J... ..A-JUN 181980 ' ' Y U 4..-'.. c y' ll .'..,,.. N. I '0 Docket No.,50-312 i
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'p-4- Attention:/ Mr. John J.~ Mattimoe N m Assistant General Manager;., and. Chief. Engineer, ~ ' ca f.;. s-3.... c... v, c s Ge.ntlemen: ,.x; e ~ .i : n. 1 a
Subject:
NRC Inspection of Rancho Seco < ' m - c, e,: ,1 r. i 1 'q This refers to the: inspection conducted by Mr.' H. Canter of this office between May 1 and 30,:1980,' of activities authorized by HRC License No. DPR-54, .l and to the discussion of our findings held by Mr. H. Canter with Mr. P. Oubre. and. other menters of. your staff.at,the, conclusion of the inspection.. ,e r> Areas examined during this inspection are described in the enclosed inspection report. Within these areas, the. Inspection consisted of selective examinations l of procedures and representative records, interviews with personnel, and . observations by the.inspe.ctor. ' ~ e .c qy p. s, No items of' noncompliance with NRC.' requirements'were identified within the scope of.this inspection. O,
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.In accordance with Section 2.790 of the NRC's " Rules'of Practice," Part 2 . Title'10 Code of Federal Regulations,' a copy of this letter,and the enclosed 'l ,,. J. inspection report will be placed 'in the NRC's Public Document Room.- If this. .reonrt contains any -information that you believe to be proprietary.. It is . l,H.. necessary that you submit a written application to this office, within 20 days. ' of the date of. this,iletter, requesting.that such"information be withheld from. pubile disclosure.l The application must include' a' full statement of the reasons why it is claimed that,the.information is proprietary.1 The. application should be prepared 50 that any proprietary information identifled is contained, in an enclosure to the application,'since the application without the enclosure' ~ will also be placed in.the Public: Document' Room. "If we do not hear from you -in this regard within_the specified. period, the report.will be placed in the Pubile Docunent Room.,gg if ,, ; g "y ', 7' '.4 . w. /..c..; ;..
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o. ,.. v . m. - r. n. Should you have any questions concerning this. inspection, we wi,ll be clad to discuss them with you. 3 .i, p,..
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No. 50-312/80-17 2 s e a cc w/o
Enclosure:
' '. ' e U' ./ ; ' t C R. J. Rodriguez, SMUD -. l. 9... L.- G. Schwieger,' SMUD s W F. M.6, l 3 ' n-~
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. n". c. .4, ... o, s .r ~ . h"'IY.,'. .*l N.,Y.n.. !. :, ';l'.'.l, h..hh. ...? E .O\\=. ' ' -- n ; e - .. - ^ ..,... a A. y i. F.. '. ... - O,.no t;u{1-}"p,;$Me iU' S7. NUCLEAR REGULATORY: C0!Ci1SSION 'a *.F f. $d. r.u. :... h NO$@NS.%j.ca.j:f OFFICE 0F;INSPECTIONj AND[ ENFORCEMENT l'F' N. v., myn,,13,.- -ig=..:L ug;.= , : ;. m.;..a. :., n. c. e J l nit,J,cJ.?.p.t. '?.ilt f:r;';, j.,n:gg.9g y..,j; ;g, ! RECION VlNNW(p ' , i ", ' :" t " I, '~ ' .,., !,' t, e,- : >, ;, i d,. ' ',,, sq c q,4/.$.p., ~ 50-312/80 4 R2 port No. 1 DPR-54' Safeguards croup Dscket No; 50-312 License No. Sacramento Municipal Utility District. ., 7 t.icensee: P.'O. Box'15830 ' / ;,.: c. i r Sacramento,' Cali fornia :J95813 ' R ' '"*"' - U t' ' d' ' Rancho Seco Unit 1 ~ ricility Name: j,, Herald, California (Rancho Seco Site) Inspection at: a cpi'" + May 1-30' 1980-Inspection conducted: Mb gJbe - s /fjo d //7/fb Inspectors: I Date.S igned, f ' ident. nspector. 3, n HarveyT.. Canter, Se ' t. e i l Date Signed i .4, l s t { 3, Date Signed l Approved By: e u /[- thM/0
8. H. Faulkenberry, Ch' ac'tur Projects Section 2, Date Signed Reactor Operations.a d fluclear Support' Branch Summary:
f Inspection between May 1 and 30, 1980 (Report No. 50-312/80-17) s Areas Inspectedi ~ Routine inspections of' operations; plant trips; follow-up-l on items of noncompliance; follow-up on Headquarters requests; and, indepen-dent inspection effort. The~ inspection involved.76 inspector-hours by the . '... Senior Resident' inspector..- e .i i_' ' ',,,. Resul ts : 0f'the five areas. inspected,*no.: items of noncompliance or devia tions.'were. identified.
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o.:.; y j,'.,,, g.i,, /,y : q,t,J 7i t j ' ' s y %"', .n .. n. ." l','.'.,,,l m'c.'in ?.. f ':f; DETAILS is .c .c' m 1.. Persons Contacted _ I i j', '3. Rodriguez, Manager,. Nuclear Operat ons 3.:, Mec ,g, Q~,;p,,cp,.t..,,','. ~ ,7 R .l: 2 Oubre' Plant Superintendent P.f ': 'D..Blachly,t 7 M @pervisor, M, p i ','- ,}' .d:jf%;,' Su . ~ ~N i . $;.2.3. Brock, Eletrical/I&C Ma ntenance3 1 Co .w..in f.W. 0.iColeman, QA Engineer '. -f. - 1 R ' ? G.lCoward, Maintenance Supervisorp., n'.M... /Pp,:: ,t ;;,o ,a. i 2'3. Ford, Operating Supervisor- 'lifj t ,e :J '3 W )
H. Heckart, Engineering. Technician i; Engineer,i. A.. g'.k
.3 4 ug it. ', r. - J. Jewett, Senior Quality Assurance- .i 3. Lawrence.' Site Project. Engineer-3. McColligan, Mechanical Engineering Supervis.org:- .'.f R J' : J .j pc 3. Medina, Quality Assurance Enoineer' R 2. Miller, Chemistry / Radiological: Supervisor. ;'. R ,. 2..Schwieger, Quality Assurance - Director ' '. ' L J. Sullivan,' Quality Assurance Super. visor. .y D. Whitney,. Nuclear Engineer-1M v...;.:, y -i .) B. Wichert, Mechanical Engineer:'i... ( W g y . pg. The inspector also talked with and interviewed several other. licensee ~ employees, including members of the engineering, maintenance, operations, Landqualityl assurance,(QA-) organizations.;' ,y.,w pi.. i . < 0enotes those' attending the Exit Interview on.May. 16, 1980. 20enotes'those attending the Exit Interview on May'19,~1980. 3 Denotes those attending the Ext.t Interview on May 30, 1980. .s.. 30, 1980. Exit ..l . :The following Region V personnel also.' attended.the May 'n ) .",.g: + q Interview: ' I B. Faulkenberry, Chief, Reactor Projects Section 2 A. Johnson, ' Reactor! Inspector / Enforcement Coordinator. tg. m y d.,w ".... e t, ~~#,' ' 2.
Operational Safety verification? 7.ih..' '! - '
', T n ;;. ' < - The inspector observed controlf room operatinns, reviewed applicable loos ."and conducted, discussions.with contro.1 room operators durin Lof May 1980.Tours of the auxiliary 'and turbine buildinas were conducted to, 1 systems. observe plant equipment. conditions, including potential -fire hazards, fluid leaks, and excessive _ vibrations and.to: verify that maintenance re- . quests had been initiated-for equipment!.in. need of maintenance. The b '8' 6 p. ,[!. h ~ .f i taj,. ,i, ,j ['
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inspector by observation and direct' interview verified'that,the physical security. plan was being, implemented in accordance with the station security
, lan., ,; g, p 9 The inspector observed plant nousekeeping/ cleanliness conditions and . verified implementation of radiation protection controls. Durino the month. . of May 1980, the inspector walked the.accessibl.e portions of the Auxiliary ~ ,.,Feedwater System, t{ verify ope,rabi.Iity. jl,.,,;gg. S', p, . l. g, ..i 4 o ,,3, o ~.These reviews and observations:were conducted to verify that facility . operations were"in conformance with the requirements established under .i- '.. technical specifications,' 10 CFR, and administrative ! procedures. ..,5..a '.;;l : g. t. :.l 3 j.q,g. 3. ly, ~ M 3 Observations., u:..g ;, ; ., y 4 :' g g. t,. 3. i.:.. -,' f j! t.,. u a. Diesel Generator Time Delavs 4 e Time Delays'(TDI) in the Engine Control Panels of the two eneroency. l diesel generators were found by the inspector to be of different p.; i types. TDI in the."A" diesel was a 2412 PN Agastat, whereas TDI in the "tl" diesel was a 7012.PC Agastat. During recent maintenance. work performed on the A". diesel, TDI.was replaced with a 7012 PC Agastat, but a record review by the inspector did not indicate that 't there should.have been a difference between the time delays in the two. diesels. ..,r 7 The inspector asked the licensee for further information of this-issue. This. item will be followed-up at a later date (80-17-01). b. Seismic ~ System Response. .: s : 7, J -. During three recent earthquakes of naceitude 6.0 or greater centered I near Mammoth Lakes in the East Central part of California, no seismic 'L systen alarms.were received at Rancho Seco. All three earthquakes ~ .were folt by~ personnel at the, site. " '.. The seismic systems at Rancho'Seco have a minimum sensitivity of .~ O.0lg in two directions., All ' instruments have~recently been calibrated. 1,',. "f'. and made operational. A passive scratch-pad type of peak-recording accelooraph was' analyzed withJ.no; indication noted'areater than the minimum readability of. 005q. The inspector asked the-licensee if they had contacted the USGS to see'if their instrument'(on the Rancho Seco property) had been read and analyzed. A licensee representative stated that no such dis-cussions with.USG5 had been made, but.that.they would look into the USGS findings, e . 0 .'.t' ' i ', ' s .i.
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1 ~ [,1, .. < ru pp, tg... ;,.g.f. ;,,:.vQ.h,g :Ni s ;ip: j ' M[ f ;yJ,'i} } ;: a . s UM M'n,.p V& E NW..i : 1!' M< t. c c. PORV Glock Valve. .,.,, q. f,,..i. The inspector noted that on May 28, 1980, the PORV block valve was . closed. ~ When asked why it had been open since plant startup follow-ing refueling outage, a licensee representative stated that since _. the PORV -was reworked-duringt the outage and did not leak, it was t.'3 g: felt to be all right to o mrate with. the valve open.. Standino ,f'.. a Order 3-80 (dated Februar.i.15. 1980)'. allowed operation with the ?i:! $1 [ valve!open,but.on'May' 28 :1980, the operators. decided that by ~ f
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i. Or 1 N' The' following anunciatorjalarms occur'during ortafter the use ofEthe 'f, ' containment fan coolers. (No ' safety concern exists except for the fact that if these alarms-are always energized, the operators may". become complacent and. neglect.to. res, pond to real problems. .. - 1. Cora. Flood Tank Lo Press Alarms t n,. .~,; , i.i.,;. Reactor Building Emergency Flow Differential High on . Reactor. Building Emergency Vent. Coolers . i;f D
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..h ,w .o t,. ' ? A' licensee representative stated that they will pursue the problem 7 . ;'( . j:.with.the. Generation Engineering Department. (80-17-02) , b r. : t No items of. noncompliance or.. deviations were. identi fied.., ;. pg
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.Following the plan.t trip on May 30,:1980,"at 2:12.PM, due to an electrical . /.,. fault on buss 2E2, the-inspector ascertained the status of the reactor ' p " and. safety systems by observation of control room indicators and discus- .sions with licensee personnel'concerninti plant parameters,. emergency system status and reactor coolant chemistry..The inspector verified the establish- ' ' 1 ment of proper. communications and reviewed.the corrective actions taken by the licensee., y "' 7,Yll ' All systems responded' as expected, and. the' plant [was. returried to ' operation q,j' g.on the evening '.of May. 30,.1980. . :nF ,_:;*,. 's.: v;' p r t, .Q r .' ;. r.;' QE;(No, items of.noncompl f ance or dev,iations, were..i,dentifled j",., ~ ~ ,., 3
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Follow-up on ! tons of Nnncompliance ' %'d i.. The response to the items of noncompliance which led to the imposition of a $25,000 civil. penalty were examined to..ascerta.in that the corrective measures were completed..' 'y c t y, e N I ) F..
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i: j. By letter dated April 23, 1980, the licensee responded to three citations in the Notice of Violations attached to IE Inspection Report No. 50-312/80-06. Five items of noncompliance (Items 80-06-01 throuch 80-06-05) were discussed in the referenced report, but SlRC Headquarters combined the five items into the three items described in the Notice of Violations. All five items are thereby closed (80-06-01 to 80-06-05), as part of this inspection activity. All corrective actions were verified. Standino Order 5-80 is still in effect which requires dual verification of many procedures included in the response. The guidelines for dual verification for all procedures listed in the response state that the lineups must be performed by two operators / technicians sequentially, each indepesdent of the other. Further informa-tion on this item 15 discussed in the exit, interview portion of this report. ~ No items of noncompliance or deviations were identified. 5. Follow-up on Headqu1rters Requests a. Helical' Spring Inspection On May 16, 1980, the inspector was shown a B & W site bulletin discussing a' fuel assembly holddown spring problem. Due to broken [ spring problems at other B & W plants, Rancho Seco instituted an inspection program to see if the problem existed at the site. The helical ' spring'is located in the fuel assembly upper end fittino. It transmits a force from the upper reactor internals to the fuel assembly to counteract normal hydraulic lif t, assuring that the fuel assembly stays firmly seated against the lower reactor internals. Rancho Seco and onsite B & W personnel carefully reviewed core verification video tapes for evi,dence of broken sprinas. No broken springs were found. There were two fuel assemblies of which the video tapes were not clear enough 'to verify 'the spring's condition. Sixty-nine discharged fuel assemblies in the spent fuel pool were examined with a video apparatus. No indications of broken sprinos were noted in this' review. Based.on the above infonnation, the licensee reported to B & W that there is no.a. ppa, rent broken holddown spring problem at Rancho Seco. No items of noncompliance or; deviations were identified, b. .Ca tegn_ry,,"A" Requi rement Veri fica tinn ' By letter dated May 1,1980, the NRC informed the licensee of the staf f's evaluation for the Rancho Seco Nuclear Generatino Station g ^' a
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g 'o ". 73.' !? e); t :0 7 tpmzw - M. ^? ' /,,.. n 7..j.;i : :. n. y o.. 4 .'. ;, ;,.,' : l,, '. .a..' i j, b i'. ~ f 4 l,. ' actions taken to satisfy the Category, f'A". items of NUREG-0578, ,o '"TMI-2 Leassons Learned. Task-Force Status Report and Short-Tern . Recommendations." I. ',' .,l n; The referenced letter requires the Office of: Inspection and Enforcement to verify.many actions taken by che licensee and to document' the. verifications in an' appropriate; inspection. report. n .,Y ~ ~,.; ' .e,.(,. ;.,. a. n ' ~ a J. ~ 'p Of imediate concern was the status of four items that >were to, be + j,- substantially compl.ete.by,the' b.eginning. o(June 1980. . <.? As of May ~ u.- e.: e. , d.. E" . 30, 1980, the 'following.four titems 'were substantially ' 'y. ' compl ete.- Due to a reactor trip. on the afternoon of May. 30,1980,. c.' n a slight delay in the finallinstallation' of tne hiah range effluent,1), ^~ P monitors'is'likely, but the monitors.should be installed and tested, during the first week'in Jun'.' Following cis a.1 Jst.of the four e ' items by NUREG-0578 paragraph. number: 1 g 1,. . Item 2.1.3.a . $, Direct Indication of Power Operated Relief . >..e ' ", Valve and Safety. Valve ~ Position..(System 1,, - is. operational.) ~ ~ Item 2il.6.a System Integrityi (A report on the required 'c W system leakage was submitted to NRR on N ay:22,'1980.) M 7 '. g,.. * 'i _
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(A report on the., design and capabilities of the lonq-term h. post accident sampling facility has been .l ,. m 1.i. ,,. submitted.).,.,y.,. .m, ,.r t f8. 4 '.c Item 2.1.8.b . High' Range Effluent Monitors. (Not complete 4., t .4 8 as.of.this writing.-). No items of noncompliance or dev,iations were identified. J .y. 6. Independent inspection Effort ', ~ 1,.t i, 1 ..J,, ' Discussions were held between the Senior Resident inspector and operations', 5 {. security and maintenance-personnel in an attempt to better understand' 4 problems' they,may have which. are related to nuclear,. safety. These dis 'cussions will continue as a standard practice. 1 \\ .u, On numerous occasions, during the month of May, the Senior Resident " ; e9 '. Inspector attended operations status meetings. 'lhese meetings are held i i i. - by the Operations Supervisor toLprovide all disciplines onsite with an update on the plant status'and ongoing, naintenance wnrk. In addition to the above, independent inspection.effor..t was performed on I. i J i' the following' items: " ' 6 : ' jy.g i.[A;+f.g, '. , ', ;./ - - f, k..h-)[l;d.,:f' [ ~y. ' [. gg, -- .g =1 [.h, . '. N.,.' ,,, 4 i h.j,.'el {. ; . e, ' .. w. ', l',c, ].D,),;y, * ',7 [#"I O q]3,,'l. j,m ~. S(,:.., n,b"'. / .y l,'. W , [ 'b ' [$',C
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s n s . ' s.. a. Containment. isolation valve operations.- Safety; grade, anticipatory reactorltrip.T ' ~,,' l b ^ c- ~- b. s. c. Diesel generator operability.and the related technical specification requirements. ' W, e,., y - t f ,n .r. '~.6 No items of noncompliance or deviations were..Jdentified.. . n 4 %r. ;.y p,9 g y.~ gy 4.,, en g.p., ... q, ; -)...q . i. g,, a. ; i t y, (; ,j 7.. Exit Interview., ,,,. j, w%. The inspector met'with licensee' representatives (denoted in Pa , g t w e ,w 7,,.c throughout,the month ~and at the conclusion of the inspection on May 30,4,Jt. '1980, and summarized the scope andi findinos of. the;' inspection activities.','. ~ ...p .As p.,e's
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c Due to questions raised by the Performance Appraisal Team during a recent, . inspection at Rancho Seco'(April 14-25 and May 5-8, l'380),.the licensee d has asked for a clarification from Region.V on the meaning of Technical ~ c. Specification.6' 5.1.6.d. :;This. technical specification states: e The Plant Review Committee shall be responsible for: ...d. Review 'of all proposed changes or modif,1 cations.,to, plant systems ;or equip- ,t.. ment. that affect nuclear. safety... ? wa, n 1 En 4 By.a letter dated May 6,'1980,:from the IE Headquarters, the following interpretation.was rt:ceivea: ', 1 y The interpretation of this T/S allows for the position of a-i,,, f reviewer (screening engineer) and does not mean that the. Plant ' ~.' Re analysis.of proposed chanqes.. Design chances as used here, means. ' ? . t 3 3 ..those as defined in the IE Manual pertainino to 10 CFR 50.59. ~ They aref" responsible" to see that! such design review is accom- ' l '.' ~ plished, i.e., program /proceduras exist to. require the detailed . analysis and safety evaluation'.(if. required) be conducted and y. the results transmitted to the' PRC.! This~does not apply to yd,T. . routine maintenance. performed.. Jhe,PRC;can, then pass judoement ;~, 'd ~'on ths ' proposed change.. In the case-of LRancho' Seco, the 'PRC woul.d handle.al1 items specifically referred to.it by the " screen-i- ing enqioeer'."l The PRC would,'also review'the judaements made by, 5 ~ the "screeninq enqineer" on ' proposed channes to safety related systems'and proposed changes ithat affect. nuclear ' safety prior to'. ~ implementinq the chanqe.:.,q',r ..' b;, H. . t in accordance with_this interpretation,'the licensee Intended to channelj " procedures'to remove the screening enoincer's function and require the "c,. PRC to-review all. documents that cause chance or modification of Class I systems, however, Reqinn V has'not. received a. final readina from NRC Headquarters on this issue..Until~..that: time thisSitem'will be pursued (.!..'., a5 a fol1ow' up i tem.. (80-l 7-03) !.P.;M...:s' " :% 9 W ... ?.i'. M. ' ' ' ~ 1/ . m p.. y.. w' N(. w. n.. :V g. w..,.... '.e::.; by, ...,..: v.;..n.; s..... v, s ./ .k [ .N 'M
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,i, .s ,.c,,,,), c ..t. 4 q Finally, during a public meeting on May 2 1980, that was held in Sacramento to give the licensee a public forum to respond to the three . items of noncompliance and attendent civil' penalty issued in April 1980. ', (see Paragraph 4 in this report and IE Report 50-312/80-06), Mr. R. C. DeYoung. Deputy Director of the Office of Inspection and Enforcement 3 queried'the licensee. He asked if there should be other safety related systems or. components which should receive mandatory dual verifications. when removing from or placing the, systems i,nto service. 3, it E Specifically, Mr..DeYoung mentioned the PORV system valvino and the hydrogen regeneration system. The licensee's position is that, at present,' a'- the list of systems requiring dual verification is adequate. The licensee'. 'is not adverse, however, to adding systems to' the list in the future if . deemed appropriate. For example, wnen the' hydrogen receneration system, , and emergency.hioh level samplina' systems are completed in the future, i.. these systems, are likely candidates for dual verification. ../. m i 1 e r m w ,F ic .4 r 6 rg- +=
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-m_ ~5 c_ g ~ 2.]170 SACRAMENTO MuMICIPAI. UTit.lTY OfSTRICT O mi s street, ses :wm. Sacraniento, cas fornw 95a11 (9161 J l February 22, 1980 ,g i / ,sy:.V.Q*', z.. Nuclear Regulatory Corsnission ' '*/ ?.'f).. l Attention: Mr. R. H. Engelken, Of rector Reofon V Office of Inspection & En fort.emen t 1990 North California Boulevard 'lalnut Creek plaza. Suite 202 Walnut Creek, California 94596 'i.: Wih', Docket No. 50-312 Rancho Seco Nuclear Generatino Station, Unit No. 1 IE Ru11etin 79-27
Dear Mr. Encelken:
The Sacramento Municipal Utility District has reviewed IE Bulletin 79-27 concerning the loss of non-class-1-E instrumentation and t control power system bus during operation. The following information j is provided in response to the items in this Bulletin. 1. Review the class 1-E and non-class 1-E buses supplying power to safety and non-safety related instruentation and control systems which could affect the ability to achieve a cold shutdown condition using existing procedures developed under item 2 below, for each bus. I a. Identify and review the alarm and/or indication I provided in the control roor to alert the operator i to the loss of power to the bus. t 1.a. Answer Rancho Seco has seven buses supplyino instrumentation and control systems for the plant. Of the seven, four are class 1-E, and the remaininn three are non-class 1-E systems. The identification of 6 each power source is: 1. Vital power sourco 1-A 2. Vital power source 1-0 3. Vital power source 1-C 4 Vital power sourc9 1-D - NNI-X84 NNI~Y 5. Power source 1E 6. Power source 1-F 7. Power source 1-J - NN.r-x 4 NNI-Y At TYJRNM15' 50MW 2-;T cnH EC Su!NlGD l/ti;b.)hYCV yF.ste9 6ITHtK W5 3~W ?g.), op w m v e D urr M v A*
{ Mr. Eng]1 ken February 22, 1980 Each of the above' systems has annunciator indication of either trouble or failure within the control mom. The identification of each follows: I 1. Vital power bus 1 A trouble; 125v DC bus A tmuble (two separate points) 2. Vital power bus 18 trouble; 125v DC bus B trouble (two separate points) 3. Vital power bus 1C trouble; 125v DC bus C trouble (two seoarate points) 4. Vital power bus ID trouble; 125v DC bus D trouble (two separate points) 5. Turbine plant 120v AC bus IE and 1F trouble; 6. 125v DC bus E trouble; 125v DC bus F trouble (two separate points) / 7. Channel "A" power failure (for indication of loss of Safety Features "A" digital power) 8. Channel "B" power failure (for indication of' loss of Safety Featums "B" digital power) 9. NNI "X" power fail - Note: This power source is supplied from vital bus 1-D. 10. NNI "Y" or "Z" power failure - Note supplied from 1-J 11. ICS or " fan" power failum - Note: This power source is supplied from vital bus 1-C. 12. ICS or NNI 120v AC power transfer. - Note: The source of power when transfer occurs is taken from the 1-J bus, Therefore, indication of trouble or bus loss for either the class 1-E or non-class 1-E is adequately indicated to the operator. b. Identify the instrument and control system loads connected to the bus and evaluate the effects of loss of power to those loads including the ability to achieve a cold shutdown condition. I 1.b. Answer The instrument and control loads connected to each bus is as follows: w
---_-._n_ - ~.. ~' Mr. Engelken -3 February 22, 1980 1. Vital bus "A" (Class lE) Safety Features actuation Channel "A" ~ analog Safety Features actuation Channel "A" digital Reactor Protection system Channel "A" Control rod trip breaker "A" 2. Vital bus "B" (Class lE) l Safety Features actuation Channel "B" analoo Safety Features actuation Channel "B" digital Reactor protection system Channel "B" Control rod trip breaker "B" 3. Vital bus "C" (Class 1E) Safety Features actuation Channel "C" analog Reactor protection system Channel "C" Inteorated control system "X" A "Y" i Control rod drive systen logic 4. Vital bus "D" (Class lE) Reactor protection system Channel "D" Non-nuclear instrumentation system "X" Non-nuclear instrumentation system "Y" Control red drive system logic 5. Bus IE (non Class 1-E) Polishing demin control Console and vertical board instruments Plant instrunent power Plant miscellaneous control Reactor building instrument power 6. Bus IF (non dlass 1-E) Make-up de
control Pressurize, control Reactor instrument power Pressurizer level. control 7.
Bus IJ (non Class 1-E) Control rod drive system logic Integrated control system alternate Non-nuclear instrumentation alternate "X" Non-nuclear instrumentation alternate "Y" Auxiliary hollers E-360 and E-365 control Ib Continued The effects of a loss of Class lE power to the loads. 1. Safety Features actuation Channel "A" analoo a. The effect of a loss of power to this channel results in a channel trip on the analog subsystem.
~ Mr. Engelken February 22, 1980 i Ib Continued 2. Safe y Features Actuation Channel "A" digital J l a. The effect of a' loss of power to this channel 4 will result in a subsystem trip, however, no l ) actuation to the end devices connected to the i system will result. This system requires power to actuate the output relays within the system. However, if either of the other two active channels sense a trip requirement, the "P" subsystem devices will be actuated. 3. Reactor Protection System Channel "A" 4. The effect of a loss of power to this channel will result in a channel trip. 4 Control Rod Trip Breaker "A" ( The effect of a loss of power to this system a. will cause a breaker to trip open. 5. Safety Features Actuation Channel "B" Analoo ( a. The offeet of a loss of power to this channel is the same as. item "1" above. 6. Safety Features Actuation Channel "8" Digital The effect of a loss of power to this channel a. is the same as item "2" above. ("A" devices will actuate) 7. Reactor Protection System Channel "P" i The effect of a loss of power to this channel a. will result in a channel trip. 8. Control Rod Trip Breaker "R" The effect of a loss of power to this channel a. will result in the breaker tripping open. 9. Safety Features Actuation Channel "C" Analoo a. The effect of a power l'oss to this subsystem will result in a chanr.el trip, thus resulting in a 1 out of 2 of the remaining channels to actuate all end devices. 10. Reactor Protection System Channel "C" 4. The effect of a power loss to this channel will result in a channel trip. l
- - ~ ~ ~ ~ h e? Mr. Engelken Fdruary 22,1980 1 11. Integrated Control System (Supplied from "C") a. The effect of a loss of power to the ICS will l result in a power transfer of the ICS via a automatic transfer to a non-Class 1E bus. Therefore, a loss of power on this channel will have no affect on the operation of the ICS. However, if the assunption is taken j that a non IE bus is not available then the ICS failure mode is that all controlled devices will revert to their 50% position. 1
12. Control Rod Drive System Logic (Suoplied from "C")
a. The effect of a loss of power from this source would not affect the operation of the control rod drive system. Redundant power supplies are fed from a separate Class IE source which would take over the load independently.
13. Reactor Protection System Channel "D" t
The effect of a power loss to this channel a. will result in a channel trip.
14. Non-nuclear Instrumentation System "X" Power and "Y" Power (Supplied from "D")
The effect of a loss of power from this source a. would cause the following: i
1) Each source of power, one to the "X" supply and one to the "Y", is backed with an automatic transfer switch.
If the entire source was lost, both would be transfered to a non-Class IE bus. Therefore, no adverse affects would be noted.
15. Control Rod Drive System Logic (Supplied from "D")
a. The effect would be the same as 12 above. The redundant power supplies will assume all load for tho system. The followino lists the effect of power loss to non-Class 1E buses and addresses only those items essential to blant operation as far as instrumentation and control is concerned.
16. Polishing Deminerlizer Control (Supplied from E)
A loss of power to this subsystem would effectively a. cause a loss of feedwater which would result in trans-fering to the auxillary feedwater system. In addition, l
.. ~. - .clI,,- Mr. Engelk:n February 22,1980 i 16.a Continued Rancho Seco has the capability of manually bypassing the polishing system which would allow returning to the normal feedwater systen if desired. The bypass could be initiated within 10 minutes if this mode was required.
17. Console and Vertical Board Instruments a.
Tha instruments and indicators fed from bus "E" are not essential to plant operation, therefore, loss of power would have no adverse affect.
18. Plant Instrument Power, Plant Miscell'areous Control, and Reactor Puf1 ding Instrument Power
)! a. The systems listed here are not essential to plant operations, therefore, loss of power to this system would have no adverse affect.
19. Make-up Domin Control Fed From the "F" Bus a.
The make-up domin system is not essential to obtaining a cold shutdown, therefore ' loss of power to this system would have no adverse affect. In addition, a manual bypass of the system can be initiated to maintain flow in the system for letdown i and make-up. I b. l Rancho Seco has the capability through cross ties between this bus (F) and the power system (J) to maintain both buses on the Ifne on a failure of either. To initiate this cross tie would take approximately 5 minutes to restore the power to the lost system. ' Operating procedures are available to cover this requirement.
20. Any other system, fed by either the "F" bus or the "J" bus, necessary for a controlled shutdown of the plant, can be handled in the same manner as stated in item 19.b.,
\\ t l
/.
Mr. Engelken February 22, 1980 i c. Describe any proposed design modifications resulting from these reviews and evaluations, and your proposad schedule for implementing those modifications. 1.c. Answer Based on the results of this review, no design modifications are proposed. 2. Prepare emergency procedures or review existing ones that will be used by control room operators, including procedures required to achieve a cold shutdown condition, upon loss of power to each class 1-E and non-class 1-E bus supplying power to safety and non-safety related instrument and control systems. The emargency procedures should include: The diagnostics /alanns/ indicators' symptom resulting from / a. j the review and evaluation conducted per item 1 above. b. the use of alternate indication and/or control circuits which may be powered from ot:.er non-class 1-E or class 1-E instnamentation and control buses. c. methods for restoring power to the bus. Describe any proposed design modification or administrative controls to be implemented resulting from these procedures, and your proposed schedule for irrplementing the changes. Answer As described in response to Question 1.b upon loss of power to each class 1-E and non-class 1-E bus supplying power to safety and non-safety related instrument and control systems that may be required to achieve a cold shutdown condition there is an automatic transfer to another power source. Therefore, no additional emergency procedures are required. 3. Re-review IE Circular No. 79-02, Failure of 120 Volt Vital AC Power Supplies, dated January 11,1979, to include both ' class 1-E and non-class 1-E safety related power supply inverters. Based on a review of operating experience and your re-review of IE Circular flo. 79-02, describe any proposed design modification or administrative controls to be imple-mented as a result of the re-review. Ans' r As a result of the District's re-review of IE Circular flo. 79-02 Failure of 120 Volt Vital AC Power Supplies, it has been determine that no additional changes are required in either design or ad. mini-strative control.
.. c-t* Mr. Engelken February 22,1980 Please advise if we can provide any additional infonnations however, we consider this response to complete the requirements fo the subject bulletin and will take no further action unless so advised. Sincerely yours, i /. 3 3. 2..,.. m w.. W. S. Bossermaier Acting General Manager i i cc: Office of Inspection and Enforcement Division of Reactor Operations Inspection I i e t i I I I t i i i, i ..}}

ML20155H518

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