ML17244A598
| ML17244A598 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/29/1979 |
| From: | James Shea Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-10, TASK-RR NUDOCS 7907120494 | |
| Download: ML17244A598 (33) | |
Text
L.
'MEETING SUMtORY DISTRIBUTION:
cket 50-244 Local PDR ORB82 Reading NRR Reading H.
R. Denton E.
G.
Case V. Stello R. Vollmer kl. Russell B. Grimes.
T. Carter D. Eisenhut D. Davis De Zi8llann P.
Check G. Lainas A. Schwencer R. Reid T. Ippolito V. Noonan G. Knighton D. Brinkman J.
Shea OELD OIGE(3)
H. Smith R. Fraley, ACRS(16)
J.
Calvo M. Greenberg J.
R.
Buchanan TERA'2 9 OVXIOypp orrICEW DURNAME&
DATC3N'C EORM 318 (976) NRCM 0240 0 V,0, OOVCRNMCNT 0RINTINO ONRICC I I 07 ~
00 ~
7 ~ 0
pl/i J' n
I l
Docket No. 50-244-LICENSEE:
Rochester Gas
& Electric Corporation (RG&E)
FACILITY:
R.
E. Ginna Nuclear Power Plant
SUBJECT:
SUMMARY
OF MEETING WITH ROCHESTER GAS
& ELECTRIC CORPORATION (RG&E) r NRC and RG&E representatives met in Bethesda, Maryland, on May 10, 1979, to review the R.
E. Ginna Nuclear Plant Auxiliary Feedwater Systems and other related systems as part of the on-going review following the Three Mile Island Unit 2 accident.
A list.of attendees is included as Attachment 4.
To prepare for the meeting, the NRC telecopied requests for information to RG&E on May 4, 1979 ) and May 7, 1979 (Attachment 2).
The RG&E responses ) to the NRC requests for information provided the basis for the meeting 'discussion.
RG&E also provided the following drawings that were used by the NRC representatives to trace out systems and check for interactions.
'I 33013-544=1 Feedwater,Systems Flow Diagram 33013-.534 Piping Flow Diagram Main & Reheat System 33013-543 Condensate System Flow Diagram 33013-509A Service Hater System D-302-071E Standby Auxiliary Feedwater System James J.
- Shea, Project Manager Operating Reactors Branch 82 Division of Operating Reactors Attachments:
As stated I
I gO t page Se'e ne It OlrlrlCC~
SURNAM4&
DATE~
DSS:PSB------
DSS:ASB--
DOR-:ORB82 DOR:ORB(,
JJShea:
a
~ ~ /
MGreen6'e 5/23/79 DLZiemann JACalvo
~ ~ ~ ~ ~ ~ ~ ~ ~
~
~ r ~
~ ~ ~ ~ ~ ~
5P >/79 I
5/Q/79 5P'/79
,NEC RORbf 518 (976) NRCbf 0240
- V4 OOVRRNM4NT NRINTINO OtrtrICNI ~ O'I ~
1047 ~ 4 The results of the NRC review will be'included in the NRR staff report to the Commission currently scheduled for completion near the end of May 1979.
II
,V
~ X I
I
Mr. Leon D. White,'r. May 29, 1979 CC Lex K. Larson, Esquire
- LeBoevf, Lamb, Leiby & MacRae 1757 N Street, N.
W.
Washington, D.
C-20036 Mr. Michael Slade 1250 Crown Point Drive
- Webster, New York 14580 Rochester Committee for Scientific Information Rober:
E. Lee, Ph.D..
P. 0.
Box 5236 River Campus Station Rochester, New York 14627 Jeffrey Cohen Hew York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza
- Albany, New York 12223 Director, Technical Development Programs State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Mr. Leon D. White, Jr.
Vice President Electric and Steam Production Rochester Gas 5 Electric Corporation 89 East Avenue Rochester, New York 14649
1
ATTACHMENT 1 As part of its on-going review of the Three Mile Island Unit 2 accident, the staff finds that it needs additional information regarding the auxiliary febdwater systems (AFWS).
This information as outlined below, is required to evaluate AFWS reliability for Combustion Engineering (CE) and Westinghouse (W) designed pressurized water reactors.
The requested information is in addition to that requested in the IE Bulletins, and should be brought to the meeting scheduled with the staff on May 8 thru May 12, 1979.
Written system description (as built) including:
List of Support Systems for Auxiliary Feed System Operation (Both Electric and Steam)
'ater Supplies for AFWS (primary and backup)
Current operating procedures and test and maintenance requirements including:
All LCO's for AFWS, main FW system and related support systems.
Listing of operator actions (local and/or control room) and timing require-ments for such actions.
Procedures for reinitiating main feedwater flow.
As Built P8IDs with symbol keys including condensate and steam side Ledgible Equipment layouts drawings including:
Isometrics, if available Identification of inhibits preventing accessibility to AFWS components and related electrical equipment Relevant control systems description including:
Schematic or logic control diagrams Listing of actuation signals/logic and control MSIS logic for isolating'FWS, if installed electric power dependences All "readouts" available in control room for AFWS operation AC 8 DC Power One line diagrams (normal and emergency power supplies)
Divisional designation e.g., Train A, Train B, requirements on all AFWS components and support systems List of normal valve states and loss-of-actuation power failure, position Operatirig Experience, including Number of main feedwater interruptions per year experienced to date for each unit Number of demands on AFWS per year to date (test and actual) for each unit Summary of AFWS malfunctions, problems, failures Provide Available reliability analyses Steam Generator dry-out times (assuming loss of all feedwater flow, with 100$
initial power, with Reactor trip, no line breaks)
System design bases including:
Seismic and environmental qualification Code and guality, gA
~
0 May 3, 1979 Provide written responses to the following set of questions by 5/8/79 Describe backup systems available (to auxiliary feedwater) for providing feedwater to steam generators.
Discuss actions and time required to make tHese systems available.
Are procedures available?
If so, provide.
Provide the following procedures:
loss of offsite power loss of feedwater LOCA (small and large)
Steam Line Break Provide following information for PORV's:
Number capacity setpoints (open and close) manufacturer and model indications of position record of periods isolated (isolation valve shut) challenges during life of plant (from plant records) including performance of valve,'ause of challenge.
experience of two-phase or subcooled discharge of PORVs and safety valves with description of valve performance Provide indications of PORV isolation valve in the control room.
Provide the following information on ECCS:
initiation setpoints system description pump performance characteristics (head curves)
Provide reactor protection system trip setpoints.
Provide information on charging
- pumps, how they relate to ECCS including:
number flow vs. pressure power sources and backup water sources seismic qualification List all challenges (and cause) to ECCS as indicated on plant records.
List and discuss all instances during which your plant has undergone natural circulation.
Describe all automatic and manual features which canstop the reactor coolant pumps.
ATTACHMENT 1
tr' P
'J
ATTACHMENT 2
~3L 7k ~i g provide fog'ioeing ioformatioo for i dr'A+>g@ r~@~c CeL~o~g 4ovides; 1 ~ Number 2-capacity
- 3. setpoints (open and close) 0 manufacturer and model 5 indications of position 4< challenges during life of plant (from plant records)
'ncluding performance of valve,'ause of challenge.
7.'experience of two-phase or subcooled discharge of PORVs and safety valves with description of valve performance-iFhlcT'~(o~ ~WW-j9 Pye eM C "E ~~>
8: Sc ~AY <o, le79 vo iuec.
8Y titZc coo pic'r:
A. SelLtee aNcwV
lel
~ I
~ I
~
I I I
~
II s
II s
s
~
I
~
~
~
~
I
~
I
~
I I
I I
I I
I
~
I
~ I
~
I I
~ I
~
I III I
~
I I
I
~
~
I I
I
~
I
~
~
I I
~
~
I, I
I I
~
I I
~
I ~
c,.~.
C/g%(
ATTACHMENT 3 AE/l All LCO's for AWS, main FW system ard related support systems.
Listing of operator actions (local ard/or control rocm) and timing require-ments for such actions.
Procedures for reinitiating main feedwater flow.
All operating,
- test, and maintenance procedures that concern safety related systems are being reviewed and updated.
The limiting corditions for. ceration for the auxiliary feedwater system are those stated in Tech.
Specs.
Sections 3.4, 4.8.
The limiting conditions for operation for the main feedwater system is governed by administrative control ard covered in operating procedures.
The support systems for the auxiliary feedwater and main feedwater systems are condensate and AVT system, ard service water system.
Limiting conditions for operation of the condensate and AVC systems are covered in operating procedures.
The limiting corditions for operation of the service water system is covered in Tech.
Specs.
Section 3.3.4.
If the main feedwater pumps trip the auxiliary feedwater system is automatically initiated,, m cperator action is to verify pumps have started ard are pumping.
Control board irdications will verify that pumps are.running ard delivering water to the steam generators.
Our steam generators are U Tube type-and have about a
52%
level at full power, operator action, is to verify auxiliary pumps start and heat is being removed fran steam generators.
Emergency procedures are written to direct subrent operator actions.
On loss of main feedwater flow a steam flow/feed flow mismatch auld cause a
reactor trip so reinitiating feedwater flow would not occur without -subsequent investigation for cause.
RELEVtRZ CDPfQL SYSTEMS DESCRIPTION INCLUDING:
All "readouts" available in control rocrn for AVOWS operation The read-outs available in control rocm for the aux. feedwater system are as
"B" pump flow indicator (0-275 gpn)
Turbine pump flow indicator for "A" S/G and "8" S/G (0-500 gpn)
"A" pump discharge pressure (0-2000 psig)
"B" pump discharge pressure (0-2000 psig)
Turbine pump discharge pressure (0-200 psig)
"A" pump recirc valve "B" pump recirc valve Turbine Pump recirc valve "A" pump service water suction valve "B" pump service water suction valve Turbine pump service water suction valve "A" pump dischazge valve "B" pump discharge valve Turbine pump discharge valve to "A" S/G and "B" S/G (0-100%)
Turbine. pump steam admission valves Turbine pump AC oil pump "A" pump start/stop switch "B" pump start/stop switch Crossover valves Condensate storage tank level indicator Turbine pump DC oil pump Alarms on control board associated with the auxiliary feedwater system are:
Auxiliary feedwater pump cooling water filter high differential pressure.
. Auxiliary feedwater pump light load.
Turbine driven feedwater pump DC oil pump auto start.
Motor driven auxiliary feedwater pump oil pump off.
Turbine driven auxiliary S/G feedwater pump steam valve A-B blocked closed.
Turbine driven feedwater putnp low oil pressure.
Attachment 3
- Divisional designation e.g., Train A, Train B, requirements on all AVOWS caago-nents and support systems "A" motor driven pump is fed from Buss 14 and motor operated valves associated with the pump are fed from MCC 1C.
Air operated valves are fed frcm instrument air header.
Oil pump is fed from MCC 1C.
1 "B" motor driven pump is fed from Buss 16 and motor operated valve associated with the pump are fed from MCC 1D.
Air operated ves are fed from the instrument air header.
Oil pump is fed from
?
Mc.c.
i+i a
Ghe turbine driven pump has steam supply lines from the "A" and "B" main steam header before the main stean isolation valves.
AC oil pump is fed from MCC lC.
The service water supply valve mtor is fed from MCC 1D.
%he steam admission mtor operated valves are fed from DC control panels.
Attachment 3
AC & DC KMER List of normal valve status and loss-of actuation paver failure position
%he following are the valves associated with the Auxiliary Feedwater System:
%carbine Driven Pum MV NV MV MV MV MV MCV MOV MV AGV MV AOV MV MV MV MV AOV AOV 4070 locked 4071 Locked 4074 locked 4075 Locked 4015 locked 4098 locked 4013 Closed 3996 Open 3999 Open 4297 Closed 4000 Open 4298 Closed 4001 Open 4002 Open 4005 Open 4006 Open 4294 Closed 4291 Open Open Open Open Open Open Closed Fails As Is Fails As Is Fails Open Fails Open Fails Open Fails Open "A" Motor MV 4025 Open MV 4019 Locked Open MV 4357 Closed MOV 4000A ~C(chic.8 MOV 4007 Open MV 4011 Open AOV 4304 Closed Fails As Is Fails As is Fails Open "B" Mtor Pum MV 4026 Open MV 4018 locked Open MV 4356 Closed MOV 4000B @pea C.lo5~
MOV 4008 Open MV 4012 Open MV 4360 Closed MV 4359 Open Fails As Is Fails As Is Attachment 3
OPERATING EXPERIENCE g INCLUDING:
Number of main feedwater interruptions per year experienced to date for each unit 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 1979 N OF TIMES TRIPS OCCURRED) 'IO MAIN FW INTERRUPTIONS 3
5 4
2 1
1 2
A PPA 0 V $0'/o F'OWE<
4 OF TIMES MAINZENANCE NhS PERPORMED ON MAIN FNP (NON-RDUTINE)
B P 1
2 Attachment 3
" OPERATING EXPERIENCE, INCUJDING:
nunber of demands on AH% per year to date (test ard actual) for each unit.
The auxiliary feedwater pumps are used during all plant startups.AND sguTDocurvd Periodic test (PT-16) "Auxiliary Feedwater System Flmr Check" is performed monthly (except during scheduled refueling shutdowns) and after any pump maintenance which may alter pump parameters.
Actual initiation of the auxiliary feedwater system has occurred during two transient conditions.
July 1973 due, to a low low steam generator level and feedwater flow-steam flow mismatch associated with the "B" steam generator'hich was attributed to the separation of the "B" steam generator feedwater control valve "plug" fran the valve stem.
(Abnormal Occurrence 73-6 and 73-7 dated July 31, 1973)
, October 1973 due to a low low steam generator level condition in both steam generators as a result of a loss of offsite power.
(Abnormal Occurrence 73-9 dated October 23, 1973).
Attachment 3
I
OPERATING EXPERIENCE, INCLUDING
- Summary of AFWS malfunctions, problems, failures Unusual Event 74-01 6/7/74 During performance of PT 16 "Auxiliary Feedwater System Flmr Check" on May 9, 1974 - After APNP had started and valves opened normally, very little flow reduction occurred to indicate that the normal throttling sequence had been initiated.
IGC Department adjusted the setpoints of the flow bistables in the closing circuit for each valve.
Abnormal Occurrence 73-11 12/14/73 During performance of PT 16 When the "A" and "B" motor driven AEWP were started there.were no pressure or flow indications.
Pumps were stopped and vented and retested successfully.
Abnormal Occurrence 74-20 12/18/74 Test procedure PT 16 did not permit recognition of the fact that the administrative electrical limit on the open position of the discharge rrator operated valve of each~ of the two motor driven auxiliary feedwater pumps prevented these pumps from delivering required flow at postulated maximum pressure.
%he electrical limit was removed.
Abnormal Occurrence 75-1 1/17/75 During performance of PT 16 on the turbine driven auxiliary feedwater pump the turbine governor valve and associated trip ard throttle valve tripped to their closed position in approximately 15 seconds after the steam supply quick start valves were opened.
This was caused by a degraded lube oil supply pressure due to drift of the lube oil supply regulator.
Attachment 3
OPERATING EXPERIENCE, INCLUDING Summary of AFWS malfunctions, problems, failures Maintenance on Auxiliary Feechrater System (Information from annual and semi-annual reports to NRC)
Mar. 70 Dec. 71Mune 72 "B" MDAEWP Leaking oil seal Steam valve to turbine APWP seats cut July 73-Dec.
73 I
July 74-Dec.
74 "B" MDAEWP feechrater line vibration due to control valve failure "B" MDKPWP excessive wear of balance disc caused hydraulic imbalance Jan.
75 Mar. 75-May 75 June 75 July 75 July 75 Nov. 75*
April 75 K)AH%'
low oil pressure "1B" main steam admission valve to TDAPWP - valve disc and seat cut due to steam leakage Steam isolation valve to TDMWP valve valve leaking "A" AFWP recirculation valve cut valve plug and seat "A" AFWP discharge check valve undetermined parts worn "A" AMP motor operated discharge valve valve plug and seat cut "B" APWP leaking gasket TIRE%'
metal chip in relay valve "B" AFWP relief valve blcvn fuse Main steam to TIRFWP supply valves cut valve seats and plug Main steam to TDAEWP supply valves cut valve seats and plugs July 75 June 77 "B" AFWP leaking gasket TDAFWP metal chip in relay valve t
Main Steam to TDAFWP motor operated valve faul'ty clutch assembly Oct. 77 "B" AFWP relief valve cut seat Attachment 3
STEAM GENERMQR DRYMUT TIMES (ASSUMING TDSS OF ALL FEEIÃATER FICN, WITH 100%
INITIAL PCNER ~ WIIH REACKR TRIP g NO LINE BREQ(S )
Assume (for 1520 MWT) 1).
The water in the S/G's at 100% power is 1680 2).
The steam temperature is 513'F at 770 psi.
3).
The energy required to heat water fran 513'F dip) is 3.7 x 106 BTU or 1.05 Kl.
4).
(8.116 x 104 lbs.)
(970.3 BTU latent heat)
=
IE.
23.075 MÃ ft.3 or 8.116 x 104 lbs.
to 547'F (to initiate steam 7.875 x 107 B'lV's X KW 34T3 BIG 5).
Pump heat
.5%.= 7.23 x 103 PQJ or approximately 2 x 10-3 MN secs
~
secs.
6).
Total heat removal capacity of the secordary side water is item 3 + item 4 = 24.125 MW.
7).
From the attached table the time it would take decay heat and pump heat to consune the available heat sink following a reactor trip with ra feedwater flow to the S/G's is 33.23 minutes Attachment 3
(FRACTION OF F.P.)
x (1520 MW/hr.) x 1 hr./3600 secs x TIME (secs)
+ PUMP HEAT = LOZAL TIME
~SEC.
1 2
4 6
8 10 20 40 60 80 100 200 400 600 800 1000 2000 DECAY HEAT GENERATION RATE
. (FRACTION OF FULL POWER)
.0775
.0732
.0684
.0661
~0635
.062
.0558
.0491
.045
..043
.041
.0338
.0289
.0260
.024
'.0228
.0193 HEAT ENERGY (MW)
.0349
.0329
.0618
.0598
.0576
.0564
.2556
.4586
~ 42
.403
.386 1.63 2.84 2.59 2.43 2.33 10.15 (MW)
.0678
.1296
.1894
.2470
.3034
.5590 1.0176 1.4376 1.8406 2.2266 3-.8566 6.6966 9.2866 11.7166 14.0466 24.1966 24';125 MWT = 1994.1 sec. or 33.23 min.
Attachment 3
SYSTEM IZSIGN EASES INCLUDING:
Seismic arx3 environmental qualification Motor driven aux.
feed pump a). Seismic Gilberts RO-2267
.32 G vertical 6 horizontal (simultaneous) b). Environmental,
- 1).
Motor Westinghouse 676427 40' max. ambient 80' rise for windings
- 2).
Pump (2-NTF-87 Worthington) 104'mbient Attachment 3
PROVIDE 'lHE EOLLCNIHG INFORMATION K)R PORV'S:
- 1). number 2 (PCV-430
& PCV-431-C) 2). capacity 179,000 lbs./hr. each 3). setpoints (open and close) 2335 psig open, 2185 psig close 4). manufacturer
& rradel Copes-Vulcan D-100-160-2 1/2 operator 2" 1500 psi 3" inlet & outlet 5). Indications of position open/close (limit switches) 6). Record of pericds isolated MOV-515
& MOV-516 insufficent time for detailed study experience shows valves closed greater than 50% of time 7). Challenges during life of plant (including performance of valve, cause of challenge.
None 8). Experience of 2-phase or subcooled discharge of PORV's and safety valves with description of valve performance - None NOZE: EGRV's start to cpen with 695 air 991 air fully cpen.
Safety Valves setpoint 2485 psig manufacture
& mdel Crosby Nozzle type relief HBBP86 Type E, size 4K26 capacity 288,000 lbs./hr. saturated at 3% over pressurization nunber 2 (PCV-434
& PCV-435)
Attachment 3
PROVIDE INDICATIONS OF EORV ISOZATION VALVE IN THE CDNTBOL BQCN.
The indication on the control board for both PORV's position is an cpen or closed light indicator which ccmes frcm limit switches located on each valve.
Also associated with these valves is a temperature indicator (0-400'F) down stream of both valves, with alarm on control board.
These valves relieve into the pressurizer relief tank.
There are actor operated isolation valves associated with each valve, with control switches with light indication, on the control board, that indicate open or closed.
The pressurizer relief tank has read out indicators on the control board for high pressure, high temperature, hi level.
There are alarms associated with these indications.
Attachment 3
PROVIDE 'ZHE FOLDING INFORMATION CN ECCS:
initiation setpoints system description pump performance characteristics (head curves)
Initiation setpoints a).
Residual heat removal-c).
Safety injection-2/3 Hi Hi containment pressure (305) 2/2 manual pushbutton auto on safety injection (recirc) manual (breaker) 2/3 Zo Lo steam press.
"A" S/G (500N) 2/3 Lo Lo steam press.
"B" S/G (500N) 1/3 Za pressurizer press.
(17154)
+ lo pressurizer level (5%)
1/2 manual pushbutton 2/3 contairment pressure (60)
System description a).
Safety injection Pump performance characteristics (head curves) a).
Cont. spray - Ingersal-Rand (2 pumps) b).
R.H.R. Pacific (2 pumps) c).
S.I.worthington (3 pumps)
Provide reactor protection system trip setpoints.
1.
Manual 1/2 pushbutton 2.
Power Range High Range 2/4 108%
3.
Power Range Zow Range 2/4 24%
4.
Intermediate Range 1/2 current equivilent to 25%
5.,
Source Range High Flux 1/2 105 cps 6.
Overtemperature delta T 2/4 delta T (increasing Tavg, decreasing press.,
delta flux greater + 5 or -18) 7.
Overpmmr delta T 2/4 de1ta T (tavg. greater Tref, delta flux greater +5 or
-18) 8.
Zaw pressurizer press.
2/4 1865 psig.
9.
High pressurizer press.
2/3 2377 psig.
10.
High pressurizer pressurizer level 2/3 92%
ll.
Loss of coolant flow - 2/3 90% both loops or RCP breakers open (less than 50% F.P.)
12.
Loss of coolant flow 2/3 90% either loop or either RCP breaker cgen (greater than 50% F.P.)
13.
Undervoltage 1/2 + 1/2 70% of normal on busses llA arxl 11B 14.
Underfrequency -1/2 + 1/2 58 Hz on busses ll A and 11 B 15.
Steam/Feedwater mismatch 1/2 low feedwater
(.8 x 106 I/hr. less than steam flow) plus 1/2 low steam generator leve1 (30%)
16.
Lo Lo S/G 2/3 15% either steam generator 17.
Turbine Trip - 2/3 low autostop oil press.
reactor greater than 49% or blocked by low vacua or loss
'of circ. water pumps.
2/3 turbine stop valves closed 18.
Safety injection initiation Attachment 3
PROVIDE INFORMATION CN CHARGING PUMPS'N
'CHEZ RELATE 'IO ECCS INCLUDING:
nunber flow vs. pressure power sources and backup water sources seismic qualification Charging pump data:
NOIR: 'Ihe charging pumps at Ginna are not required for or part of the ECCS systems.
Nunber 3 Positive displacement triplex recipricating.
3000 psi 60 gpn rated at 2385 psi head at 130'F 15 psi NPSH at 127'F Power source 480 V ac vital bus alternate source Diesel (if required)
Water sources VCT (normal letdown
& seal water return)
, a).
normal makeup b).
BAST directly (2 paths) c).. EST (emergency low level makeup)
Charging puris a).
Seismic (Espic-676370)
- 52 G vertical 6 horizontal (simultaneous) b).
Environmental None Attachment 3
LIST ALL CHALLENGES (AND CAUSES)
TO ECCS AS INDICATED CN PLANT RECORDS.
Our plant records indicate one actuation of our ECCS system.
On October 21, 1973 at approximately 0552 hours0.00639 days <br />0.153 hours <br />9.126984e-4 weeks <br />2.10036e-4 months <br /> a loss of offsite power occurred resulting in an autanatic reactor trip fram 91% power.
An electrical disturbance on the instrument busses created the overpower delta temperature reactor trip signal to trip the reactor.
The loss of offsite power also generated a turbine trip signal and a "low low steam generator" condition in both steam generators.
This initiated the start of the steam driven auxiliary feedwater pump Mich injected anbient temperature water fran the condensate storage tanks into the steam generators.
This cold water injected into the steam generators cooled down the RCS to the point &ere the safety injection system was actuated by a ccmbination of low pressurizer pressure (1715 psi) and low-low pressurizer level (5%).
Safety injection pumps pumped about 700 gallons (causinp the B.A.S.T. level to change fran 60% to 50%) of 12%
boric acid into the primary system when a B.A.S.T. low-low level (10%) was
- received, switchiog pump suction fran the B.A.S.T. to the refuelirg water storage tank.
'Ibis false indication was due to the loss of power to Instrument Bus 1A which caused the B.A.S.T. level indication to fail downscale.
(Abnormal Occurrences 73-9 and 73-10 dated October 23, 1973)
Attachment 3
M LIST AND DISCUSS ALL INSTANCES DURING WHICH YOUR PLANT HAS UNDERGONE NATURAL CIRCULATION.
A natural circulation test was completed on January 18, 1970.
%he data was evaluated and was found to be in excellent agreement with the predictions in the FSAR, section 14.1.12.
On October 21, 1973 a loss of off site power occurred at approximately 0552 hours0.00639 days <br />0.153 hours <br />9.126984e-4 weeks <br />2.10036e-4 months <br /> resulting in a reactor trip from 91% power.
Upon loss of pcarer to the reactor coolant pumps, the coolant flew for core cooling and removal of residual heat was maintained by natural circulation'.,
~ ~
Attachment 3
DESCRIBE ALL AUIOMATIC AND MANUAL FEA'IURES WHICH CAN SIOP %K REACIOR COOLAPZ PUMPS.
Manual start/stop switch Autanatic those effects on the 4 KV busses that will cpen the breakers quch as:
Under voltage, Underfrequency and Overcurrent Attachment 3
What trips pumps and what make and rmdel are they?
Main Feedwater Pump (2)
Motor Auxiliary Feedwater Pump (2)
Condensate Pump (3)
Cordensate Booster Pump (3)
Turbine Auxiliary Feedwater Pump (1)
Worthington Worthington Byron Jackson Bingham Worthington 14WNCD 1311 2WTF-87 BJ 24KXH-7 14X14X16.5HSB 3PKr 87
'Ihe main feedwater pumps are fed from the 4160V Busses and would trip on loss of off site power.
They would also trip on a safety injection signal or can manually be tripped.
Overcurrent or thermal overload will cause a trip.
The auxiliary feedwater pumps can be tripped manually, or by undervoltage or overcurrent or thermal overload.
The mndensate pumps are tripped manually, or by loss of off site power, or by overcurrent, undervoltage or thermal overload.
'Ihe condensate booster pumps can be tripped manually or by low suction pressure of 100 lbs.
'Ihey will also trip on undervoltage, overcurrent, thermal overload or loss of off site power.
Pumps will also trip cn high pressure of 425 lbs.
'Ihe turbine driven auxiliary feedwater pump will trip on loss of steam, overspeed or low oil pressure.
Main Feechrater Mtor Auxiliary Feedwater Pump Condensate Pump Condensate Booster Pump Turbine Auxiliary Feedwater Pump Demineralizers (Cochrane
- 4) "=
Feedwater Heaters Bypass 7400 GEM 200 GPN 6600 GPM 4800 GFN 400 GPM Full Power Need 3
Yes 2150 ft. head 2850 ft. head 660 ft. head 210 ft. head 3000't.
head Can be bypassed Air operated valves fail
.As is Manual/Auto fails as is'.
'ttachment 3
ATTACHMENT 4 LIST OF ATTENDEES RGKE B. Mecredy J
~ Larizza F.
Schwind J. Huff Sandia Lab G. Kolb NRC J.
Shea J.
Calvo*
M. Greenberg
.r I