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July 8, 1982 MEMORANDUM FOR: Dr. David Okrent, ACRS Member , p a} Jh40u FROM: Mr. John MacEvoy, ACRS Senior Fellow SUBJECT : Effects of Non-Seismic Category I System Failures During Seismic Events Introduction i'
This report is my response to your question of a few months .ago as to whether the control room operator would have sufficient controls and indications to cope with the failure of any combination of non-seismic category I system failures resulting from an earthquake. To answer that question, I looked specifically at the San Onofre Nuclear Generating Station, Units 2 83 (SONGS 2 & 3) as described by the FSAR, SER, and Responses to Questions (References 1, 2 & 4). The SONGS 2 & 3. Nuclear Steam Supply System was designed by Combustion Engineering and is similar to Arkansas Nuclear One - Unit _2. The reactor coolant systems are similar to Calvert Cliffs Units 1 & 2.
Since the FSAR is a very broad document which contains almost no specific information on system controls, interlocks and operating characteristics, I confined ray study to the effects of whole system failures, and whether the operator _would be able to recognize and deal with those failures from the control room. It was also not possible to address the response of individual control systems to a change in input parameters since the FSAR does not present this information.
Unlike a Probabilistic Risk Assessment which calculates probabilities of occurrence of earthquakes of any size and the probabilities of equipment failure resulting from those earthquakes, I assume an earthquake equivalent to the SSE (called DBE at SONGS) which causes no seismic category I system failures with any combination of non-seismic category I system f ailures.
Discussion The Equipment Classification List (Table 3.2-1 of Reference 1) from the San Onofre 2 & 3 FSAR was used as the source list of non-seismic category I components and systems scrutinized in this report. Using the FSAR system 8508230153 850722 PDR FOIA DELLD5-363 PDR g/J2 Enf-L M
description as my major source of operating information, I looked at the effect of the loss of each system assuming it failed to function or it failed in an active unsafe mannar, such as could happen if the turbine bypass control system failed in a " dump 45% steam" mode. The list of sys-tems and components, along with my notes as to the effect of their f ailure is provided in Attachment 1. Attachment 2 lists the Safety Related Instru-mentation available in the control room.
Possible Problems from Multisystem Failures The source range nuclear instrument is not seismic category 1, nor is it powered from a class 1E electrical bus. In a recent letter to the NRC (Reference 3) SCE committed to upgrading the detectors and preamplifiers to seismic category 1, but no mention is made regarding the power source, therefore I assume it remains non class 1E. During a Design Basis Earth-quake it is possible for the following equipment to fail because of vibra-tion or loss of power:
- 1. Source Range Nuclear Instruments (note, the ex core nuclear instruments are also subject to flooding during a LOCA)
- 2. Individual Rod Position Indication (IRPI)
- 3. Boronometer (and letdown portion of CVCS)
- 4. Normal Primary Sampling System
- 5. Post Accident Sampling System Items 2 through 5 above deprive the operator of any indication of reactor shutdown margin. One or more control elements may have jammed due to the DBE but because of the loss of IRPI their positions would be unknown.
Because of the loss of the Boronometer and sampling systems the Reactor Coolant System (RCS) boron concentration would have to be estimated based on the makeup water sources being used and probable dilution rates.
Additionally if the source range nuclear instruments failed, a restart accident of a type not analyzed in the FSAR could occur since the operator has no indication of reactor power level. (The FSAR assumed tne operator was alerted by a high source range counts alarm, a low boron concentration alarm and a high Volume Control Tank level alarm.- It's reasonable to assume none of these alarms wculd occur following a.DBE).
While looking at the Process Sampling System design I noticed another problem not directly related to seismic problems: .the primary sample heat exchangers were cooled by normal chilled water, however there does not appear to be any means for detecting primary to chilled water system leakage. Possibly an administrative procedure exists for sampling the chilled water system at an appropriate interval. I could not find any reference to any procedure, and it could be of minor concern.
Most of the systems lost either by tnemselves or in combination with others during a DBE result in a loss of steam load or loss of feedwater to the steam generators. These systems are:
Feedwater Control System Main Turbine Overspeed Control Offsite Power Onsite non-Class 1E Power Main Circulating Water Intake and Offshore Outfall Turbine Plant Cooling Water Turbine and Generator Main Steam Main Condenser Steam Jet Air Ejectors Mechanical Vacuum Pumps Gland Seal System Condensate and Feed Systems Attachment I contains brief notes regarding the assumed effect of failures of each of these systems. Loss of condenser vacuum and loss of feedwater are addressed in the FSAR and it is shown that the operator is capable of shutting down the plant. I haven't thought of credible situations more restrictive than those analyses.
Failure of the Pressurizer Level Control System (PLCS) and the Pressurizer Pressure Control System (PPCS) can lead to a potentially confusing situation for the operator which is possibly more severe than that studied in the FSAR, but not terribly serious either. Following a DBE, assume the PPCS fails high, but the PLCS fails as is, which I will assume to be in a condition calling for enough charging flow to slowly raise pressurizer level. Pressurizer level would go up while spray holds the pressure down.
The charging pump should trip on nigh pressurizer level, but the FSAR does not take credit for this trip in the accident analysis. From this I infer the trip is non-seismic Category 1, therefore I will also assume the charging pump fails to trip. The high pressurizer pressure SCRAM, assumed by the FSAR to protect tne reactor coolant system from overpressurization due to excessive charging, will possibly not be activated during this scenario. Tne operator also may not be alerted by a high level alarm since this function is usually generated downstream of the level instrument isolation amplifiers which means the alarm circuit is'probably non-seismic category 1. (This is true of Westinghouse plants. I'm not sure in the case of SONGS. The information immediately available does not address this subject). Regardless of whether the alarm fails or not, with this scenario the reactor operator must take manual action to avoid filling the pressurizer solid. Once solid the high pressure scram should trip the reactor, causing rapid pressure transients, which the operator could eventually control by charging, but probably the letdown system would isolate due to loss of instrument air. Thus overpressure control must be ef fected by temperature control. (These plants do not have primary Power Operated Relief valves. ) In any event the primary safety valves would lift if necessary, and the operator would have sufficient controls and indica-tions to shutdown and cooldown the plant.
Although SONGS 2 & 3 are designed for shutdown and cooldown from all anticipated accident scenarios without the need to use letdown, the lack of a PORV or any other means for reducing pressure Other than the primary safety valves (which are not under operator control), seems to me to severely restrict the operator's repertoire of responses to unexpected transients. ' Not exactly a safety problem, this is rather a question of operating flexibility. Likewise there is a question regarding autemtic pressure control while shutdown and solid: How will the operator control primary system pressure without the letdown system and Volume Control Tank?
The best I can tell from the FSAR is that RCS pressure can only be increased by running charging pumps intermittently to raise pressure but not decreased.
There are two considerations here:
- 1. Following a D8E, will the operator be expected to devote his undivided attention to controlling solid plant pressure, and
- 2. Will he be expected 'to manually control pressure in a solid plant with no means to control overpresure transients (other than by relying on shutdown overpressure relief valves)?
Again, these are not plant safety questions as much as they are concerns about operating flexibility.
Conclusions All but one of the non-seismic category I system failure scenarios I could hypothesize based on my review of the SONGS 2 & 3 FSAR I eventually dis-covered had been evaluated via the question and response process (Reference
- 4) conducted by the NRC Staff. The one remaining scenario, loss of proper shutdown margin indication, has been pursued by the Staff from a Source Range Instrument seismic qualification viewpoint, but no mention has been made of providing a Class 1E power supply, or what action to take should detector flooding occur. Except for this one scenario, I found no other situation whereby failure of any combination of non-seismic category I system and components would put the operator in a situation where he had insufficient controls and indications to shutdown and cooldown the plant.
There were also a few minor concerns wnicn may require some additional research:
- 1. Simultaneous f ailure of pressurizer pressure and level control systems taking pressurizer solid,
- 2. The lack of operating flexibility resulting from no primary PORV and no seismic category I letdown path, and
- 3. . The unmonitored cnilled water system supplying a primary sample system cooler.
A stated in the introduction, this analysis was kept simple using the references immediately available, i.e., the FSAR, SER and Responses to Questions. A more detailed and time consuming review using the resources available to the NRC Staff is possible which could take into account control and electric power system trips, interlocks, permissives, control signal inputs, setpoints, outputs, and interactions. The F5AR has very sparse information along these lines.
Some Personal Views on Power Plant Design I am convinced that nuclear power plants, like people, have their own per-sonalities resulting from individual system design, equipment manufacturers, A/E construction practices, etc. Most of the systems interactions will be predictable from design reviews, and from what I saw of the SONGS 2 & 3 review, the NRC Staff is doing a good job spotting problems. However, the idiosyncrasies of individual stations are difficult, if not impossible to detect until the proper initiating event occurs. I'm thinking of events like tne dropped light bulb incident at Rancho Seco. To predict events of this nature would require an incredible effort by reviewers with at least an SR0 level knowledge of plant operation. And even then, attempting to predict the possible results of any combination of short circuits, open circuits, overloads and jumpered contacts for tens of thousands of switches, relays control circuits, fuses, indicators, sensors, and terminal blocks that could be subject to f alling debris (f orgotten pens, wrenches, screwdrivers, bolts, etc.) during a D8E or the slip of a technician's wrist, is well nigh impossible. On submarines we tested the adequacy of our cleanup QA program with " angles & dangles" (subjecting the ship to extreme up and down angles) designed to show up loose equipment. This being impossible for a commercial power plant, we must wait for a good earthquake to test tne adequacy of their QA programs.
The point I'm making is that analysis of all the possible combinations of events during an earthquake is impossible. The big events are analyzed in great detail, and then redundancy, flexibility and containment systems are added to cope with uncertainty. Obviously design problems are being o ve rlooked. Not the big tnings fortunately (so far), but the little,
" unforeseeable" tnings, which when amplified by the complexity of the plants, assume major proportions. Given the huge numbers of "little things" that can happen and the practical limits on our ability to review plant designs, the best solution is to design for simplicity.
If I may stand on a soapbox for a few paragraphs, it appears that our solutions to design problems invariably involve the addition of active components as a "fix". An analogy to automobile design illustrate this philosophy. When the steering wheel of a turning automobile is released, the car returns to a straight or almost straight path as a result of the location of the axle pivot points with respect to the tire. Good design produces a dynamically stable system. If nuclear power plant design
techniques were applied to correct a problem and produce the same effect, instead of redesigning the mechanism, we'd probably install the following system:
- 1. Add pressure sensors to the steering wheel (to sense the driver releasing the wheel)
- 2. Use a microprocessor control system to compare the " wheel released" signal to a rate of turn indicator (to see if the car is turning but should be straightened)
- 3. Provide a feedback signal from the front wheels (verify the wheels are indeed turned) and
- 4. Provide a PID control signal to a " wheel straightening motor" to return tne car to a straight path.
The owner's manual would probably state, "Due to the safety significance of this system, redundant sensors and motors will be utilized. A manual override is provided in the event of system failure." Sound familiar?
As we continue to build and modify our power plants this way, we continue adding more complexities, increasing the possibility of individual com-pcnent failure, and increasing the chance that systems will interact with each otner in subtle unforseen ways. Unless we find practical methods for identifying these interactions, we will be building hazards into our plants in the name of safety.
References
- 1. San Onof re Nuclear Generating Station Units 2 &3 Final Safety Analyis Report, including Amendments 1 through 24.
- 2. Safety Evaluation Report related to the opertion of San Onofre Nuclear Generating Station Units 2 & 3 (NUREG-0712) dated February 2,1981.
- 3. K.P. Baskin (SCE) letter to F. Miraglia (NRC) dated May 13,1982. ACN 8205170186.
- 4. Responses to NRC Questions, San Onofre Nuclear Generating Station Units 2 & 3, including Amendments 1 tnrough 22.
Attachment 2 Non-Seismic Category I Systems and Components and the Effect(s) of Their Failure System / Component Effect of Failure
- 1. Seawall (II)* Southern Cal Edison considers failure of the seawall an incredible event, even though not designed to seismic category I criteria.
Failure could restrict sea water inlet flow to the screenwells
- 2. Probable Maximum Flood Although not seismic category I, the Berm was Berm (II) constructed to remain operational during the 50 year runof f plus " seismic event"
- 3. Reactor Neutron Source (II) No major effect
- 4. Reactor Coolant Pump (RCP) a. Loss of Forced Circulation Flow Motors (II) b. Reactor Trip
- c. Loss of automatic primary pressure control
- b. Same as a,b,c, for 4 above
- 6. RCP Motor Heat Exchanger (II) No major effect. Additional containment heat load
- 7. Pressurizer Heaters (II) Loss of automatic presure control. System must eventually be taken solid resulting from loss of steam bubble
- 8. Pressurizer Relief Quench Tank more susceptible to' overpressurization Tank (II) and rupture disc blowout if primary reliefs lift
- 9. Quench tank valves (II) Gas vent failure may lead to tank rupture s
disc blowout
, 10. Pressurizer Safety Valve Small LOCA if safety valves lift Discharge Piping (II)
- This Roman Numeral is the Seismic Category of the named system or component.
System / Component Effect of Failure
- 11. Safety Injection System Minor primary coolant leakage through pump Drains (II) seals
- 12. Control Element Assembly No indication of the failure of one or Position Indication (II) more CEAs to drop following a reactor trip. Indication failure may initiate a reactor trip
- 13. Critical Function Monitoring No major effect (Backup, computer formatted System (III) plant parameter display)
- 14. Health Physics Computer (III) No major effect
- 15. Reactor Regulating Not required when shutdown System (II)
- 16. Pressurizer Pressure Control Pressure must be maintained by manual System (II) control of heaters and spray. May initiate a reactor trip due to high or low pressure
- 17. Pressurizer Level Control Level must be controlled by manual adjust-(II) ment of charging flow and letdown flow.
Level alarms warn of abnormal level
- 18. Plant Computer System (III) Operator must rely on panel indications for shutdown plant control. Lose core operating limit supervisory system, in core flux monitoring, and core exit thermocouples automatic readout
- 19. Core Operating Limit (III) Not required when shutdown Supervisory System
- 20. In Core Instrumentation In core neutron monitoring not required wnen Indication (III) shutdown. Use local manual monitoring of core exit thermocouples
- 21. Steam Bypass Control Shutdown cooling during first phase must System (II) be shif ted to steam dumps. If system failure causes excessive steam load, reactor will trip due to high (swell) or low steam generator level or hign neutron flux
System / Component Effect of Failure
- 23. Ex Core Instrumentation, Control functions not required when shut-Startup and Control down. Loss of startup power level indication Channels (II) (to be rectified prior to S/U)
- 24. Feedwater Control System Not required when shutdown. Possible over-(II) cooling if system reacts to seismic load by opening feed regulating valve. Feed pumps do not trip on high steam generator level or reactor trip
- 25. Main Turbine Overspeed Backed up by a mechanical overspeed device Control (II) which would probably trip due to earthquake stress or an overspeed condition.
- 26. Essential Plant Parameter For use only if fire deactivates control Monitoring System (III) room or emergency shutdown panel. Should not be needed following DBE
- 27. Offsite Power (II) If normal onsite supplies are available, loss of offsite power has no effect
- 28. On Site, Nonclass 1E Forces reliance on IE supplies, all but Power (II) selected safety related loads are deenergized
- 29. Fire Detection Equipment Loss of all plant fire detection capability.
(III) SER states applicant will do a fire inspection within two hours following an earthquake
- 30. Fire Protection Water & Per FSAR "not required to safety shutdown Gaseous Systems (II) the plant for any credible fires." Have subsequently installed seismic Cat I stand-pipe systems at strategic locations (per SER).
- 31. Nonessential Portions o'f No major effect.
Fuel Pool Cooling &
Cleanup (II)
- 32. Fuel Handling (II) No major effect
- 33. Closed Cooling Water No major effect Chemical Feeders '(II)
- 34. Demineralized Water Makeup No major effect (II)
System / Component Effect of Failure
- 35. Potable Water (II) No major effect
- 36. Ultimate Heat Sink Main Backed up by auxiliary intake to supply Intake (II) the essential salt water system '
- 37. Offshore Outfall Conduit (II) Operator must shift to component cooling water emergency discharge lines
- 38. Condensate Storage Tank (II) Reduces AFW supply from 24 hr to 4 hr.
(Balance of Plant) & Pumps Allows cooldown to RHR operation point
- 39. Nuclear Service Water (II) Euphemism for " primary grade make-up water." No major short term effects.
Lose makeup water for:
- b. Component Cooling Water
- c. Primary Makeup Storage Tanks
- d. Charging Pump Reservoir
- 40. Turbine Plant Cooling No major ef fects. Loss of cooling to:
Water (II)
- a. Feed & Condensate Pumps
- b. Turbine & Generator
- c. Plant Air Compressors
- d. Aux Building Chillers
- e. Containment Normal Coolers
- f. Sample ' Coolers & Chillers Systens "a-f" are addressed separately
- 41. Process Sampling System (II) No major effects. Sampling of primary will be done via the high radiation sampling system which cools samples with CCW
- 42. Non Radioactive Floor No major effect Drains (II/III)
- 43. Radioactive Floor Drains (II) Minor effects: Uncollected leakage from radioactive system pumps and valves
- 44. Chemical & Volume Control System (I/II)
- a. Volume Control Tank (II) Loss of automatic pressure control during cooldown while solid
- - - - . _ . -, - --- -n , - ,-.,
System / Component Effect of Failure
- b. Boric Acid Batch Tank No najor effect (II) .
- c. Chemical Addition No major effect Tank (II)
- d. Letdown System f ran Back- Loss of normal letdown, loss of auto Pressure Regulator to pressure control while solid Radwaste Diverter Valve (II)
- e. Ion Exchangers & Filters No major effect (II)
- f. Boronometer (II) No major effect. Requires increased boron sampling
- 45. Normal HVAC (II) Emergency systems take over in all rooms containing ESF or other safety equipment.
- 46. Hydrogen Purge Supply & H Recombiners Provide Backup 2
Exhaust (II)
- 47. Communications Only alarms are seismic Cat I, however repeaters for hand held radios may be manually switched to class 1E power supplies
- 48. Lighting Systems (II/III) Backed up by emergency & essential lighting
- 49. Turbine & Generator (II) a. Loss of steam load, then reactor trips
- b. Loss of onsite power, fast transfer to'offsite power or class 1E busses transfer to EGD supplies
- c. Loss of turbine 1st stage pressure signal to:
- 1. Rod Blocks (not required when shutdown)
- 2. Steam Dumps (can cause dumps to open or fail to open)
- 3. Rod Control System (not re-quired when shutdown)
- 4. Steam . Generator Level Program (not required when shutdown)
- 5. Feedback to Electrohydraulic Control (not required when shutdown)
System / Component Effect of Failure
- 50. Main Steam, Downstream of a. Steam line break automatically isolated MSIV!s (II) by ESFAS
- b. Loss of downstream system results in loss of. heat sink
- 51. Main Condenser (II): Loss of heat sink, turbine trip, loss of steam dump capability
, 52. Steam Jet Air Ejectors (II) Loss of heat sink, turbine trip, loss of steam dump capability. Mechanical vacuum pumps may be used as backup
- 53. Mechanical Vacuum Pumps (II) No effect
- 54. Turbine Gland Seal System Loss of heat sink (II)
- 55. Turbine Bypass (Steam Dump) Loss of steam dump capability if turbine (II) trips or excessive steam load if system fails to full (45%) load l
56.. Circulating Water & a. Loss of RCS heat sink Traveling Screens (II)
- b. Flooding f rom condenser expansion joints
- 57. Condensate & Feedwater (II) a. Loss of RCS heat sink if system stops operating or a major leak develops.
- b. Excessive cooldown by failure of the system to shutdown can be prevented manually by shutting feed isolation valves (which also shut on ESFAS).
Feed. reg valves fail as is on loss.
of air. Feed pumps do not trip on high steam generator level
- 58. Steam Generator Blowdown No major ef fect. Seismic Category II (I/II) section isolates on AFW initiation
- 59. Turbine Plant Chemical Not required when shutdown Addition (II)
- 60. Coolant Radwaste (II) No major effect
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System / Component Effect of Failure
- 61. Misc. Liquid Radwaste (II) No effect
- 62. Boric Acid Recycle (II) No major ef fect
- 63. Waste Gas System Possible venting of H from reactor
- a. Surge Tank Drain Pump coolant drain tank 2 (II)
- b. Compressor Motor (II)
- 64. Radwaste Solidification No effect (II/III)
- 65. Non-Safety Process Radiation No effect Monitor (II)
- 66. Non-Safety Related Area No effect Radiation Monitor (II)
- 67. Plant Air (II) at Air powered valves fail safe
- b. Atmospheric steam dumps (Seismic Cat I) are automatically powered from a backup nitrogen supply
- c. Some shutdown cooling valves must be reopened manually. NRC Staff is investigating
- d. Loss (shutting) of most sample system containment penetrations, steam generator blodown lines, and contain-ment normal A/C chilled water
- 68. Post Accident Sampling Loss of sampling capability System (II)
- 69. EDG Air Start System (up Receivers contain enough air for 5 starts to Receivers) (II)
- 70. Component Cooling hater Loss of cooling to RCP seals' and Spent (1/II) Fuel Pool Cooling Water heat exchangers.
Both events have been analyzed by SCE and found to be not serious I
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Mr Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 6 of 32) peresster me. of AccarecyI #I e penet poser System ItseeeredibI cheeeele Seege(e) (1) seedeetf e) leesttee(f) pe. RFSOI asFO) A310)
It see PPDIOI ESPIO) FellO) StetOIIII apris eeld teek level 2 0-100% 5 1 centrol Case see a rose sorte sold tenet level 2 0-100% 5 l RSP 1A2 pe I I asete eeld teek 4 30-200F S 1/C local - De I temperature votese oestret teek l 0-1001 S 1 Castrol CRSS Its I level rose voleos esotrol teak l 0-100% S I asF lAt me R R level tn Cherstes lies flee 1 0-150 S 1 Castrol CRSS pe E (borte eeld) gel /ete reve $
Telene oestrol tank 1 = = L ESP IA2 IIe I E
- J es
- t. e to ,e. weive peeitime g
>g f
h Telumme eentrol teak 1 = L Control CRM pe E gg g
t.seste - -i.
poettlee
- g ,
y i
h 1stesse heet 1 0 200F S 1 Centrol CRM pp 3 esehenger settet I M rose M temperature h
letdess heet 1 0 200F S 1 R$F lAt its R I eschenger settet l temperatore c
>4 Seres esmeestretten 1 0-5000 S 1 Centrol CRSI see I En FFM rose N Seven eensentrattes 1 6-5000 S 1 RSP 1A2 No I
- E ><
FFM Spree esmeestrettes 6-2500 M 1 3 8 Centrol CRM Its I F,tt Intesse test 1 9-400 $ 1/C Centrol Case se x 4Stf escheaeor settet 16/te.Ig roen $
Preseers g, latemen best t 9-000 $ 1 33F lAl go I g >
esehenger outlet Ib/te.Ig N W '*
O za
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Table 7.5-1 P
SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 7 of 32) e TFra psemester me. of Accuracy (d) of penet power Systee SteeseredIb) Chamete Range (c) ggy g 4,,g (e) tacetton(f) no. It See RPS@I ESFOI ASIOI PPDIOI RSPIOI PasttOI ptSIONII Refueltas seter tenh 1 0-8001 5 8 centret CR57 no X I IT905) level roen Refueling unter tank 4 0-1001 5 t Centret CR57 fee I x 1 (T006) levet geen Defueltog ester tenh I 0-1001 3 a Centret CR59 Tes I I I (T006) level enom Centetsument summy level 2 0-8001 5 t Centret CR57 fee I reen NF safety tejection 4 0-8004 5 L/I Centrol Ca57 Tee I M E
pump on/off room LP safety te}ection 2 0-8004 5 L/t Central CR57 fee I I pump on/off team
, nP safety tejectten 2 0-2500 5 I Centrol Cm57 pe 1 y header pressure 2 iblin roce gg I
H g LP safety tejeetten 1 0-600 ' 5 i Centret Ca50 Me I N u ****** " '" ' -*
O LP safety injeetten 2 0-400F 5 I Centrol Ca57 fee X X *ef heeder toeperature room M C LP safety tejection 1 0-400F 3 R Control Ct39 fee I I 3 heeder temperature reen Centelament eump velve 4 - - L Centret CR57 fee I pee tt ten reem I y w
Safety injectlee pumpe M 4 - -
L Centrol CR57 fee X- N 3
etelmum restrewletten room valve positten h g
LP header velve 4 0-1001 Centret peeitten 5 t teen Cn57 no a a g g
RP heeder selve 8 0-1901 5 I Centret CR57 Tee I I poeitten race E tvl RF heeder het les 2 0-100% 5 I contret CR57 na I I I
tejectise, selve room poeitten WP heedet to DC dreta task, vetve puettien 2 - -
L Centret roen Cm57 Tea R y%
gg s
e
- b. _
7
S:n Onofro 2&3 FSAR SAFETY-RELATED DISPLAY INSTRUMENTATION E
-E = =
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E-
= = = m = = = m u E
O E m a w
n m u u n u H H
= 5 a -
E e z 5 5
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= ~ _ _ _ _
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.xe $ 2 1 2 .
e : .
t . :. : : -
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5 j=i -
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4 Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 9 of 32) paeameter so. of Accuracy (4) eyeeen Itseeeredit) toege(8) of remet reser his (1) heedeut (e) toesttee(f) pe. 15 Bee RM O) BerOI AtlO) FF9tOI ESPIOI PWIOI StBIOI(II safety telectise task 4 - -
L Centre t l drete & 8811 selve Cn37 no X 1
poettlee rose safety tejeettee test 4 -
L N Btll **lv* y'*ttt'e Centret Cast no a 2 r*
- Safety tejeettee task 4 - -
L Centret
*t **t'* *itI'* CR37 Yes X X N2 "*"
Caetateneet omat heedet 1 - - L Centrol setwo pseittem CR57 Tee 2 rees I Safety tejeettee eteek 4 - - L Centrol CES7 Yee 2 tti wetwo leaksee esecret rose en volve peeltsom -D N . shetema oesttag best 2 6-330 3 1 Centret
, e-wr. toi.t an.2 re.m CR37 Tee 1 o
I ,ree-H ;
w E shutdese ecolles meet 2 o-seer s i ,,
t Cente.t Cast v.. : i t".
M encheeser, settet temperetare rose i
w
' 3 j 4,3 Shutdeme eestres itse 1 0-1991 i
( centret selve poettlee S I Centret room (RSS po I m cn shutdeen easttog 1 -
L brymee velve peettlee Centret CRSS We l I reen Sheadeem cooltag time t 0-10,000 $
ties t Centrol Cait no n I D gottets reos y
t tn E
- .ii.v-t M
S
- \ M
\ as E i;l j ze D
_ ~-
e* .
% % \ m -
\,
\
i i
t Table 7.5-1 -
SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 10 of 32) a poemeneer me. .f Aeem ory 14) '."f poet reser i
syneen snomeered3) chamete assen te) se, (t) seedeut (e) toesttem(f) se see areM) esp 3) Ast#) pret#) nort#) seest#1 est#)(I) steen esmerater 1 4/s.C. e-test S Centret cast levet Tee I I E 1 reen steen semerater 1 t/s.C. e-test S att IA2 tevea t Tee a I I st4 m senerecer i t/S.C. e-test 3 a Centret Cast teve t Tee I I r reen stone preeur ye eter 4/s.C. S 2 Centret 9-12es,a th/te.
cm32 Tee x a i reen steen seescoter 1/s.C. e-1230, S 1 asy 1A2 preeeere Tee a 1 2 tb/te. . to se steen esmoteter 1/S.C. e-1200 D 3 a Centret CRS9 fee pressere Ib/te.In I I I toen p steam emy to etmeeshere volve 1/5.C. e-test 5 t/c Centret Ct32 Tee 1 U peettles reen Q se staan deep to t/S .C. e-tesE S t/C RSr LAI Tee e.
et e .. esi e p ettlee
- I h
- eg Amatitary feedester 1/3.0. 0-sSe to i
5 t Centret CRS2 Tee I ttee se!/ete rees I h
Austitary feeduster 1/S.C. e-see 3 a Centret Cn59 Tee E flee set /ste reen 1 Amittery feedseter 2/S.C. .
t$
L Centret CR32 fee I H vetwo poettlee En reen N
Aaertitary feedseter 2/S.C. =
L velve poettien
=
Centret 'CR57 Yes I rose E g Aemittery feedseter t/S.C. 9-test he estee-detwee ym p S t Centret Ca52 Tee X discheese valve reen I
[Mk poettlee tvl A m sttery feedsetee 1/S.C. 9-test tarttee-detwee peep S I Cartret CRS2 fee I 8 toen 1 discheroe velve poettlee
.t. et.a. t. . team- it. .C . .
t.
driven pugt (7140).
velve peettles a
C t.
e.32 ,ee i g@
e
i Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 11 of 32) paremmeer aerwecy(d)
System Itseeered O) no. .f D.c re.ei resee Chasesele Reage(CI (!) Reedoet (e) gecettee(I) Wo. IE Due APSDI ESFOI ASIO) PpDtOI RSPIO) pdsetO) StStOIIII Male steam to steam- t/S .C. - -
L driven piemp (Pt40). Concret CR32 Tee I vetse posttlee reen Matp steam toeletten 1/S .C. -
L we1ve peeittem Centret Ca57 feo I E rsom Mata steam testatten t/S.C. - -
L Centret vetve puettlee CR32 Tee. I a reen Feedeoter footetten t /S.C . -
L velve peelttom Centret CR57 Tee I I reae Feedeoter teetetten 1/S .C. -
L vaive peeitise Centrol C352 Yee I I rese forbies-detwee :s 1 -
%J easillery Fw pump.
L Centret CRS2 Tee I turhtee stop vetve room h o
f 0e positten d forttee-drivee ! -
L y
meettiary tw peamp.
Centret CR37 Teo 2 room 1 w turbim step vetve peettles IP=
W Electrie motor-delven q I 1 - -
L Centret CR57 eustilary tw pienp room fee I I
to
, (Ptet) en/off h PU tieetrie meter-delsen 1 0-1504 5 L/t Centret eustitary Fw pemip CR52 Yee E reen (Pt48) en/ett U Steam dump to 1/S.C. - -
L y
RSP 1A2 Yes
.t.o. e .e i.e X I peeItten E
d too steen generator 4 - -
L RSP 1A2 fee 2
,re.eere oo ,ei.t e.eet 2 5 gn m
Coudeemete pemmy 4 -
g ,e en/ett L RS P IA2 De R R h
rn I Condemoete eterese 2 0-100L S 1 Centrol tesh level CBS3 he K 2 reen Ceedonoste eterage 2 0-100L yh g
$ I BSP EA! We R H e4 tendt levet I a O t's ZD b -
- h
S:n Onsfra 2&3 FSAR SAFETY-RELATED
.(
O DISPLAY INSTRUMENTATIOR i
E E
I N N N I
E**
N N N I
O 2
5 8
- I
% 2 LO 5
y .
-e ez la s s- s
, b, j
oz m
n m
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a em
>* b g f4 E _
W g- . . . '
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>* O E E 1
1 :s :s -
I i
~
, , t E I i
y l' l' 1 Y I i ,a
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( g 7.5-15
Table 7.5-1 SAFETY-RELATED DISP'.eY INSTRUMENTATION (Sheet 13 of 32)
TFF" paremme., ~..f asterocy(d) .e om.ei reser System Itseeereath) Chesmete toegel') (I) Reaseut(*) tecet tee (f) pe. It mee Ryt OI ESPOI pp9tOI R$ptO) pettO)
AS t@) 913tOIIII Chemicet eterage teak 2 0-800L 5 I Centret CR37 Yee I leest room Chesteet addittom llee 2 0-25 S I Centret CILS7 fee R flew get/ete rose Chesteet adottles 2 . -
L Centret CR57 Tee E remp. en/off I rose Chesteet addities, 2 . . L Centret CR37 Tee X teetetten estee X reen poettlee
. - e teei ntes 2 .
eentret estee
. L- C.etret reen C.37 Tes , ;
peettles g3
- Sprey peep en/off 2 e-SSO 5 L/1 Centret p ib/to.2, ,,,,
CRS7 Tea R I O,
g W H e e. sprey lies flew 2 0-100L
- 1 Centret CR37 Yee I fD k
M 3 prey time 1 pressere 0-See 3 a Centret CR39 Yes I e
Ib/te.2, 3 ba 1 Sprey itse 1 0-400F 3 R Centret CR59 Tee 1 I temperatere rose Spray header selee 2 0-100L 3 Centret 1 CR57 Yes R I poeittem rose Sprey header estee 2 . . L Centret C357 O
fee I H poeittoo resa y
C tesi eter e to . .-2s , C.etrei Cas? -
preseere Ib/te. ,g i
rose E
M Sprey lies pressere 2 0-550 Centret Ib/te.2 g S 1 team CR57 fee I g $m tn Sprey Stes temperature >4 2 0-40er 5 t Centrol CR37 Tes X W P8 sia
%C
-e o pt ZU b ,
w
"r & %
Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 14 of 32) 5 Parameter lee. of Accurac y e penet power
. System SteeeeredOI Channele aange(C) (t) Seedeut(e) Iscetion(f) me. It hoe RPSOI ESPOI ASINI PPDIOI R$PROI P8510) 919tOIIII Circulottas water pump 4 . . L RSP IA2 No I I e
i .
t#9 D
D e" F Y Po, t: :
E w e.
$ W m
in u
O tn m
e<
H
- 2 tt>
=A ei m hC
-e
, O tM ZO
o- Table 7.5-1 D
w SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 15 of 32) type seeem ear me. .f ae orec ,t*I .e ree.i emner System lesseuredOI Oesoao t e Range (c) (13 g 4 gfe) tecetteett) oe. It hee G PSOI ESFOI 45tOI PPetOI GSPIOI PANIOI St$tOItII Comtetemmet sitteren 2 5 I Centret L103 Tee redistsee reen I I I 1
- 1. Casesse 4 30 to cetteet d l 8'**
to act/re
- 2. Portlestate IS*' to e t0*'eCI/ceI
- 3. Iedles to to Comtatammet staterne 10'*ect /en' 1 3 8 Centret LLO) Tee I I redtettom toen I 1 same es testeater cetteet eree Centelmusst pressure 4 4 to 30 $ g Control Ca57 fee I E E I it/te,2g ,,,,
o Comtetsmaet puesoure 4 4 to 95 Ut 3 I Centret G37 fee E I It/te.2g E I reen H
" Comtatammet preneuro 1 4 to 95 3 8 Centret CR39 Tee It/te.2g geen I E I a e-=_- : stenophees 2 S 400F 3 I Centret GSF Tee temperatore E l O ttled unter to 1 .
L team I
sernet eesties emite Centret G57 fee I I E 12 reen Chitted unter from 1 . -
L reeltag unite Centret G37 fee I a team R t3 Ca.te- -
diocharge testettaa
- 1. C t,.i Gs, ,e. .
H I
eetwo poettles rose E a
g l>
Centelemmet etasophore 1 0-400F 3 8 'd:
temperature Centret G59 fee 2 team a g
Coutodemmat dams 4 .
gg g eierelater en/edt L Centret CD40 fee I 3 en >
reen H *w
,, Dtf tw
. 3 3
er M
-a Ob b ,
.. M
/
^r %s %
n s
o>
Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 16 of 32)
Peremeter ite. of Acceracy(d) of Penel Peeer I Seedeet I'I teestgen III System Itessueedibl Chemmele Sense (t) us. It see RPS ESF ASI PPet PAftt( BISt I IIII Centelement emergemey 4 -
.- L Centrol C360 Tes I I teoltag sett en/ oft rene COf free emergency 4 - - L Centrol CB60 Tee I coettes entre team b free energeery 4 - - L Centrol C957 Tes E I cooltag entte reen CDs to emergency 4 - - L Centrol Cs60 Tes R emelleg sette race COf to emergency 4 - - L Centrol CR57 fee I E V8 coolEng enite race Smergency redtattee meettertas systee 1 9 ge 10 30 1 Centrol reem L-103 Yes I g M O stem /h %
toergency redletion I 9 go 30 R Centrol L-IO3 Yes I 23 N
-e senttoring eyetee to tone y 90 oree/h p.
W Caetainment high 2 s go 20 t Control L-405 Tee I q
range redlettee 10 reen gf3 es.it.ro .. del, g
Caetatement high 1 I go 20 t Centrol L-405 Tee I reage vedtatten SO room eenItere Rede(h Centstament Pressere 2 b200 7
5 I Centrol CE57 Tee I lb/in a reen t/3 Caetatement Preocure 1 0-200 2
lb/te a 5 R Ccettet rose CR59 Tes I h
e4 Centelament sternal o-2 10*-2 9...,"-
te 22 l Cente t t
0857 Tee I yg u,
W a
H
- =
r3 3
" 8 E CE
~
w ad ze
$D nLOM3 Mh lh e w
- t t>t'M l
/* I t>HM1C
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ae ase me l g g er lt lee l it l t ao eeo me ae es t ey eg t e rp a
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Table 7.5-1 SAFETT-RELATED DISPLAY INSTRUMENTATION (Sheet 17 of 32)
IM poe m eter as. .e ace-acy(d) .4 re et eswr System semesuredW) Obeameto Reage(8) (1) seedeett e) RecetteelII Wo. It See RPSOI ESF&I ASINI PP9tOI 99PIOI F#EtOI SISI 8( }
Centeel rose 2 0-850F S I Cent rol LIM Tee I E temperstwo ee e cobtert erre alsh edpartty telet 1 - - L Centret Ll% We 2 I damper ieenho) reum poeit See cabinet area tanke enheeet damper 1 - - L Centret LIM No R I poettlee team rettnet (#3 are. g Suske chmuseeser 1 - -
L Centret Lt% pe I E o dumper puettles seem :3 M cettnet O
, .re.
i .
M Centrol reen redtetten 2 S I Centrol L104 Yes R I O $ t caseene 80-6 ,, ,,,, N E to-t cohteet P pct /cm3 ere. 'd I .. to.t.e,p.rtteet.t. 10-, t.
10-4
~
M pct /cm3 in Castret team toeletten 2 - - L Centret C360 Yee R 8 damper smettlee reen merent enhamot 4 - - L Cent ret Cate fee I I D toelettee deper teen pooittem g
pseet air esmettiener 4 - - L Centret Cate Yes R I seetetten demper team Centret team esbiset eres damper puettles 4 - - L Centret rame CR60 fee I I $m tn c et teet H etee WM Terttee teneretary, t - - L Centret Lt% Yes I g emergeery eehemet damper posities room cableet Vg
.re. ny BN ze b _ _
O %, '
Table 7.5-1 SAFETT-RELATED DISPLAT INSTRUMENTATION (Sheet 18 of 32) im semesseer m. .e accurery(e8 .c eum.: sees, Syeen EmesuredOI theemste taugeft) (1) seedeut(e) tecettes(f) pe. Is ama aps@) gw@) Agg@) pystO) asygO) pegg@) 333g9)(1)
W ett 2 0-105L 3 t/c Centrol LtM Tee I esed4ttemor fles roen esttest eree tmmegmary maksop air 2 e-ten 5 t/c Centret LtM Tee I fl* rose cettest stee Centret rose echteet 2 *
- L Centret (340 as I aree restreetettee rose fem tes entforeettet TA pressere genegummy ett tetehm 2 *
- L Centret LtM Yes I y att heettag telle reen o 15 e cetteet O
- v. eroe :;-
4
~ a
, u 0
I a m
e Y
m YSn l'i 4 l"3 d
t hC
-e o 19 2O
S:n On:fra 263 FSAR SAFETY-RELATED DISPLAY INSTRUMENTATION
- }-
e ~$ \
I e
i E
I E**
M M M M M M M M M M E
~ e M u = w w = = = = = m g
e E
E I t E 6 E h .
., I I I I I I I I 2 2
- . h li !$5 $ 5 $ 5 $ $$
25 E 5
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a a as: at all :aII au1 ast :a. t =
1 ,. M 3
7.5-22
r Dm .
Table 7.5-1 SAFETT-REI.ATED DISPLAT INSTRUMENTATION (Sheet 20 of 32) w id) Tne wereinter so. .e Aceweer .I . esset pseer System lesesored(b) Cheemels mangeI*I tt) seedestI *I toesttee(f) se, t g he ePSOI t1FOI &$10I PF9tOI RSPINI PettOI StttOIIII suottletten toelettes & = . L Centret CBee Tee I I desper peeIttem rose Air steemmy emit 2 . - L Centrol LI% Tee I heater emiert team
, cetteet eroe Air steemmy unit 2 3 E Centrol Lt% no I ear ties roem cetteet oree Air staanse sett 2 3 e Control Lt% no I En differsettet pressere roen D eenteet U
~~~
M
- g wg air eteesmo emit dieshorge pressere 2 3 I costret Lt% no I O rose M
W y retteet Q g
g eroe M
I autIdles enhmust 2 19 to 3 3 punt tecat tee g g
- y redtattee 10'Istiles' headitag
- 1. eseesse -9 totidtes M E 2. nedteeivers testeea ,j-Ay,1 ta 8.e tuttetes enhamot 2 10,9 te q S I Centret L101 Tee I I R
.l redtetten 30 eC1/cm toen d I. r gg-9 ,, cetteet
- r. t tami,ortiesi.te ,,-.,,,,,.. .e g
E puso some temperature t m
. . L Centret Cato no I I ei.e. - g'tf 4
5en ElRi am hC 5M ze
S:n On:fra 2&3 FSAR SAFETY-RELATED DISPLAY INSTRUMENTATION
- l M
e a
i a
1 e- =w w w a w w as a w w w w a e
R n
2 u u u u u u u u u u u u u u e
n ,
t E m
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-l 4
s i t 1 i 1 1 1 1
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e b .e . .a a .a .a .a a .a =
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En . . . . . . . .
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8 91 81 a,
i 7.5-24
y s m s.
s
% Table 7.5-1 w
SAFETY-REIATED DISPLAY INSTRLHENTATION (Sheet 22 of 32) seementer e, fr**
me. .e accareer .e pas.1 p-ee 3,eten lesenewediti Cheauete anagele) (t) seessett e) tecette (f) BPSOI tsFAI AstAI os, it see PP9 TAI BSPINI PdeslOI StstOIIII EI38 supply header 2 0-200 5 t Centret Cau no I a pressere Ib/te.?g rose CCar heet eachooger 2 0-299F 5 t Centrol Ceu me I I totet touperature th/te.Ig toen
. CDr heet eerhauser 2 0-818F S I Centret CDM We 1 E j g outlet temperature room s
g CDB emersteleet 2 - - L Centrol CD17 Yee I E loue, eestatesset rose 9 teetetSee estee
$ poettlam b CCtf eaurettleet loup. 2 L m
- - Castret Ceu Tee I E 8
{g eestateueet testattee votes poettles rose U' u Q q CDs senarteleet leep 2 L D
- - Centrol Ca12 fee E 3 O e toelettee estee room M nde peeatsee M
I y Mar euerettleet toep 2 - - L Centret Can Yes I E pa I
feelettee estee room P l peettles W
i - --
feet peeteetten 2 - - L Centret teel ve, s room )
- m. - , - -
L C.ete.I C, ,ee ,
rese H
- Perellet ande 2 - -
L Centret coal ve, a 3- rees t'3 fet ,ette est reert ee 2 - - L Centret (341 Tee I t;
- rose g stas eserride# 2 - -
L rentret (se ) Tee a 5
, ..,~.2 re 1 ' ,W E
=
sme dsee 2 - - L Ceneret Caos see s g e r.ee Mm
,, 3 ad) et sette. 2 - -
L Cent rol Co.1 Tee E 3 5 terreseeldecreeee rene ,
Q. Se ameset ande ,
- .d}eet soit.. 2 - - t -t re. C. ,ee ,
, secreemetesereese rese I to este modo p.4 >
>4 d
e Table 7.5-1 C
w SAFETY-RELATED DISPI.AY INSTRUMENTATION (Sheet 23 of 32)
%8 Tm eeremeter me. .r Arrerer, tot .I ,-I ser y 99eten steeeeredfhI theemete penge ft ) (13 seedeetfel tacettenft) pe. It see EPs("I E5FOI A110I PFDIOI ESPIOI PapilOI $19tONII ATR sottrh est to 2 - - L Cent rol Cat) vee t
.et. rose s neist - e ende 2 - - 'ev :eemed
-i.- Ce=_t re. rei (n., vee t l 5-040.3 et teme .tt ,se,on 2 - - u Cee.rei Can vee
- room Start 2 - - L rentrel Cen) Tee I rene Step 2 -- - L Ce=tret Coal Tee I reen Stort fattere 2 - - L Centret C:53 Tee a r=== g Ceepeeeeer reentes 2 - - L Centret Casi vee a race q ! Lehe ett temp ht 2 - - L teatret re13 y,e 3 0
. - r_ m tse 'J *t 8 E Adjeet goverser 2 - - L Caetret Cn13 Yee I ID M terresee/derresee rose M
(> h, Poet eterage 2 0-12* 3 I Centret Cs60 Yee I P test tewel rose 4*8 o
Feel tremerer 2 - - L Centrol Ce60 Tee t M IO h peepe states rose h these 4 empo - 0-1000A 3 I Centrol Cat) no I g rece
- thee asee - 0-tonoa i I Contr.1 C=o me E toom rhese C ewe - e-10non 3 : Central Cass no n en rees Wettege - 0-50nD 1 I Control Ct61 pe I t'l war. race 9485 - -840 3 I Centret Cr63 no x
+1100 N avan *Z M En >
Fre W 7 - 15-6182 3 8 Centret CW61 no I d t rene N f
4& last t 0-7000 Centrol no
- 3 I Cet) 3 0 . he race i B
- > t*
D H>
r' HH b
D-
San Onofra 2&3 FSAR SAFETY-RELATED DISPLAY INSTRUMENTATION E
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i Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 29 of 32)
Systen rero-eer Moseered O) no. .e acewecy(4) '.*c r-ei co-r Che-ele Reage(*) (1) needest(e) leeettee(r) no. It mee RFSO) 33F0) AS10) PPDIO) 33710) PertO) BigtO)(I)
Meeel generster 2 . . L Centrol De X I lose of 430r-se reso amo eyetes temperable 2 . . L Control No I I ross Mosel gesareter 2 . . L Centrol Do I I estatemenee rose Diesel gemoreter 2 . . L Coetre! Cm63 po I stater ground rose 123 W-de bee 4 .-1509 3 7 Centret Cp63 ,ee I to rene g
us. e w . .. im , i Co.trei ,s.
,~ r- g o
w a N
N SP*
.W fn a
m I
e 5
.e 5
- 2 th M r.,
1M, 21 E 8M ze b _ _
e D O, Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 30 of 32)
TFPe '
pereenter ps. of Accurecy(a) og penet pe ,g SFotem 90essered(b) Channele Range ('I (1) seedeut(e) gaestion(f) No. It bus RPSOI ESP @I ASIOI PP9tOI RSPIOI PalttOI StSIOIIII Spent feel pool pump 2 - - L Centrol CR64 Yee 2 ee/off room Pump discherse 2 - - L Centrol CR61 no R pressere los eters reen Peep emettee t = = campeter Centrol IIA feo I temperature team e
D D
M Y
. o N
I el W 4 DJ 4
P w
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tn N
4
- H Z (#5
- c.
Ml 9
4E 5M ze b
p R Table 7.5-1 SAFETY-RELATED DISPIAY INSTRUMENTATION (Sheet 31 of 32)
- b. PPS - Plant Protection System
- DNBR -
Departure from Nucleate Boiling Ratio RCP - Reactor Coolant Pump LPSI -
Imw-Pressure Safety Injection
, HP - High Pressure LP - Iow Pressure SG , Steam Generator '
Component Cooling Water
- c. Safety-related equipment supplied by a 4.16-kV bus is provided with an ammeter integral with the hand switch module. g
- d. The accuracy given is that of the complete channel and includes for example the measuring element, @
transmitter and indicator. The figures quoted are representative for the type of measurement and &
{
e-are based on average values. 2 w
. e. I - Indicator (analog) C.
L - Light (on/off) .g R - Recorder "
I/C . Indicating Controller t
- f. Control room includes the back panels not directly visible to the operator. The control panels that e are in the operator's console area are denoted by CR and the panels that are in the control room U cabinet area or at the remote locations are denoted by L.
RSP - Remote shutdown panel used'when evacuating control room.
f.<
i 5E
- g. Prefixes ZL and HS when referring to a piece of equipment, denote two lights, open/close for valves, UM start /stop for motors, and on/off for heaters.
' g{
5N tie 5M xe
[pKfGy 'o UNITED STATES P' ' ,g NUCLEAR REGULATORY COMMISSION
,,,,-p ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 20555 g
e% ....* /
July 8, 1982 Dr. David Okrent, ACRS Member j .:
MEMORANDUM FOR: ,
I FROM:
Mr. John MacEvoy, ACRS Senior Fellow d ;
J
SUBJECT:
Effects of Non-Seismic Category I System Failures During Seismic Events Introduction This report is my response to your question of a few months ago as to whether the control room operatur would have sufficient controls and indications to cope with the f ailure of any combination of non-seismic category I system f ailures resulting f rom an earthquake. To answer tnat question, I looked specifically at the San Unofre Nuclear Generating Station, Units 2 83 (SUNuS 2 & 3) as described by the FSAR, SER, and Responses to Questions (References 1, 2 & 4). Tne SONGS 2 & 3 Nuclear Steam Supply System was designed by Combustion Engineering and is similar to Arkansas Nuclear Une - Unit 2. The reactor coolant systems are similar to Calvert Cliffs Units 1 & 2.
Since the FSAR is a very broad document which contains almost no specific information on system controls, interlocks and operating characteristics, I confined my study-to the ef fects of whole system failures, and whether the operator would be able to recognize and deal with those failures from the control room. It was also not possible to address the response of individual control systems to a change in input parameters since the FSAR does not present tnis information.
Unlike a Probabilistic Risk Assessment which calculates probabilities of occurrence of earthquakes of any size and the probabilities of equipment failure resulting from those earthquakes I assume an earthquake equivalent to the SSE (called DBE at SONGS) whicn causes no seismic category I system f ailures witn any combination of non-seismic category I system failures.
Discussion The Equipment Classification List (Table 3.2-1 of Reference 1) from the San Onofre 2 & 3 FSAR was used as the source list of non-seismic category I components and systems scrutinized in this report. Using the FSAR system l
l
description as my major source of operating information, I looked at the effect of the loss of each system assuming it failed to function or it 4
failed in an active unsafe manner, such as could happen if the turbine bypass control system f ailed in a " dump 45% steam" mode. The list of sys-tems and components, along with my notes as to the effect of their f ailure is provided in Attachment 1. Attachment 2 lists the Safety Related Instru-mentation available in the control room.
Possible Problems from Multisystem Failures The source range nuclear instrument is not seismic category 1, nor is it powered from a class 1E electrical bus. In a recent letter to the NRC (Reference 3)-SCE committed to upgrading.the detectors and preamplifiers to seismic category 1, but no mention is made regarding the power source, therefore I assume it remains non class 1E. During a Design 8 asis Earth-quake it is possible for the following equipment to f ail because of viDra-tion or loss of power:
- 1. Source Range Nuclear Instruments (note, the ex core nuclear instruments are also subject to flooding during a LOCA) l 2. Individual Rod Position Indication (IRPI)
- 3. Boronometer (and letdown portion of CVCS)
- 4. Normal Primary Sampling System
- 5. Post Accident Sampling System Items 2 tnrough 5 above deprive the operator of any indication of reactor shutdown margin. One or more control elements may have jammed due to the DBE but because of tne loss of IRPI tneir positions would be unknown.
Because of the loss of the Boronometer and sampling systems tne Reactor Coolant System (RCS) boron concentration would have to be estimated based on the makeup water sources Deing used and probable dilution rates.
Additionally if the source range nuclear instruments feiled, a restart accident of a type not analyzed in the FSAR could occur since.the operator has no indication of reactur power level. (The FSAR assumed the operator was alerted by a high source range counts alarm, a low boron concentration alarm and a high Volume Control Tank level alann. It's reasonable to assume none of these alarms would occur following a DBE).
blhile looking at the Process Sampling System design I noticed another problem not directly related to seismic problems: the primary sample heat exchangers were cooled by normal chilled water, however there does' not appear to be any means for detecting primary to chilled water system leakage. Possibly an administrative procedure exists for sampling the chilled water system at an appropriate interval. I could not find any reference to any procedure, and it could be of minor concern.
- t
+- - -- - .,__-,____.._m-. .. ,
l Most of the systems lost either by tnemselves or in combination with others during a DBE result in a loss of steam load or loss of feedwater to the steam generators. These systems are: ;
Feedwater Control System Main Turbine Overspeed Control Offsite Power Onsite non-Class 1E Power Main Circulating Water Intake and Offshore Outfall Turbine Plant Cooling Water Turbine and Generator Main Steam Main Condenser Steam Jet Air Ejectors Mechanical Vacuum Pumps Gland Seal System Condensate and Feed Systems Attachment I contains brief notes regarding the assumed ef fect of failures of each of these systems. Loss of condenser vacuum and loss of feedwater are aadressed in the FSAF and it is shown that the operator is capable of shutting down the plant. I haven't thought of credible situations more restrictive than those analyses.
Failure of the Pressurizer Level Control System (PLCS) and the Pressurizer Pressure Control System (PPCS) can lead to a potentially confusing situation for the operator which is possibly more severe than that studied in the FSAR, but not terribly serious either. Following a DBE, assume the PPCS fails high, but the PLCS fails as is, which I will assume to be in a condition calling for enough charging flow to slowly raise pressurizer level. Pressurizer level would go up while spray holds the pressure down.
The charging pump should trip on nigh pressurizer level, but the FSAR does not take credit for this trip in the accident analysis. From this 1 infer.
the trip is non-seismic Category 1, therefore I will also assume the charging pump fails to trip. The high pressurizer pressure SCRAM, assumed by the FSAR to protect tne reactor coolant system from overpressurization due to excessive charging, will possibly not be activated during this scenario. Tne operator also may not be alerted by a hign level alarm since this function is usually generated downstream of the level instrument isolation amplifiers which means the alarm circuit is probably non-seismic category 1. (This is true of Westinghouse plants. I'm not sure in the case of SONGS. The information immediately available does not address tnis subject). Regardless of wnether the alarm fails or not, with this scenario the ' reactor operator must take manual action to avoid tilling the pressurizer solid. Once solid the high pressure scram should trip the reactor, causing rapid pressure transients, which the operator could eventually control by charging, but probably the letdown system would isolate due to loss of instrument air. Thus overpressure control must be effected by temperature control. (These plants do not have. primary Power Operated Relief valves. ) In any event the primary safety valves would lift if necessary, and tne operator would have sufficient controls and indica-tions to shutdown and cooldown the plant.
l
Although SONGS 2 & 3 are designed for shutdown and cooldown from all anticipated accident scenarios without the need to use letdown, the lack of a PORV or any other means for reducing pressure other than the primary safety valves (which are not under operator control), seems to me to severely restrict the operator's repertoire of responses to unexpected t ransients. Not exactly a safety problem, this is rather a question of operating flexibility. Likewise there is a question regarding automatic pressure control while snutdown and solid: How will the operator control primary system pressure without the letdown system and Volume Control Tank?
The best I can tell from the FSAR is tnat RCS pressure can only be increased by running charging pumps intermittently to raise pressure but not decreased.
Tnere are two considerations here:
- 1. Following a 08E, will the operator be expected to devote his undivided attention to controlling solid plant pressure, and
- 2. Will he be expected to manually control pressure in a solid plant with no means to control overpresure transients (other than by relying on shutdown overpressure relief valves)?
Again, these are not plant safety questions as much as they are concerns about operating flexibility.
Conclusions All but one of the non-seismic category I system failure scenarios I could hypothesize based on my review of the SONGS 2 & 3 FSAR I eventually dis-covered had been evaluated via the question and response process (Reference
- 4) conducted by the NRC Staff. The one remaining scenario, loss of proper shutdown margin indication, has been pursued by the Staff from a Source Range Instrument seismic qualification viewpoint, but no mention has been made of providing a Class 1E power supply, or wnat action to take snould detector flooding occur. Except for this one scenario, I found no otner situation whereby failure of any combination of non-seismic category I system and components would put the operator in a situation where he had insufficient controls and indications to shutdown and cooldown the plant.
There were also a few minor concerns wnich may require some additional research:
- 1. Simultaneous failure of pressurizer pressure and level control systems taking pressurizer solid,
- 2. The lack of operating flexibility resulting from no primary PORV and no seismic category I letdown path, ana
- 3. The unmonitored cnilled water system supplying a primary sample system cooler.
I
A stated in the introduction, this analysis was kept simple using the references immediately available, i.e., the FSAR, SER and Responses to Questions. A more detsiled and time consuming review using the resources available to the NRC Staff is possible which could take into account control and electric power system trips, interlocks, permissives, control signal inputs, setpoints, outputs, and interactions. The FSAR has very sparse information along these lines.
References
- 1. San Unofre Nuclear Generating Station Units 2 &3 Final Safety Analyis Report, including Anendments 1 through 24.
- 2. Safety Evaluation Report related to the opertion of San Onofre Nuclear Generating Station Units 2 & 3 (NUREG-0712) dated February 2,1981.
- 3. K.P. Baskin (SCE) letter to F. Miraglia (NRC) dated May 13,1982. ACN 8205170186.
- 4. Responses to NRC Questions, San Onofre Nuclear Generating Station Units 2 & 3, including Amendments 1 tnrougn 22.
4 6 Attachment 2 Non-Seismic Category I Systems and Components and the Effect(s) of Their Fail,ure System / Component Effect of Failure I. Seawall (II)* Southern Cal Edison considers failure of the seawall an incredible event, even though not designed to seismic category I criteria.
Failure could restrict sea water inlet flow to the screenwells
- 2. Probable Maximum Flood Although not seismic category I, the Berm was Berm (II) constructed to remain operational during the 50 year runof f plus " seismic event"
- 3. Reactor Neutron Source (II) No major effect 4 Reactor Coolant Pump (RCP) a. Loss of Forced Circulation Flow Motors (11). b. Reactor Trip Loss of automatic primary pressure c.
control
- b. Same as a,b,c, for 4 above
- 6. RCP Motor Heat Exchanger (II) No major effect. Additional containment heat load
- 7. Pressurizer Heaters (II) Loss of automatic presure control. System must eventually be taken solid resulting from loss of steam bubble
- 8. Pressurizer Relief Quench Tank more susceptible to overpr.essurization Tank (II) and rupture disc blowout if primary reliefs lift
- 9. Quench tank valves (II) Gas vent failure may leaa to tank rupture disc blowout
- 10. Pressurizer Safety Valve Small LOCA if safety valves lift Discharge Piping (II)
- This Roman Numeral is the Seismic Category of the named system or component.
System / Component Effect of Failure
- 11. Safety Injection System Minor primary coolant leakage through pump Drains (II) seals
- 12. Control Element Assembly No indication of the failure of one or Position Indication (II) more CEAs to drop following a reactor trip. Indication f ailure may initiate a reactor trip
- 13. Critical Function Monitoring No major effect (Backup, computer formatted System (III) plant parameter display)
- 14. Health Pnysics Computer (III) No najor effect
- 15. Reactor Regulating Not required wnen shutdown System (II)- -
- 16. Pressurizer Pressure Control Pressure must be maintained by manual System (II) control of heaters and sprdy. May initiate a reactor trip due to high or low pressure
- 17. Pressurizer Level Control Level must be controlled by manual adjust-(II) ment of. charging flow and letdown flow.
Level alarms warn of abnormal level
- 18. Plant Computer System (III) Operator must rely on panel indications for shutdown plant control. Lose core operating limit supervisory system, in core flux monitoring, and core exit thermocouples automatic readout
- 19. Core Operating Limit (III) Not required when shutdown Supervisory System
- 20. In Core Inz* mmentation In core neutron nonitoring not required wnen .
Indication (III) shutdown. Use local manual monitoring of core exit thermocouples
- 21. Steam Bypass Control Shutdown cooling during first phase must System (II) be shif ted to steam dumps. If system failure causes excessive steam load, reactor will trip due to high (swell) or low steam generator level or nign neutron flux
. -. . .- . _ _ = _
System / Component Effect of Failure
- 23. Ex Core Instrumentation, Control functions not required when shut-Startup and Control down. Loss of startup power level indication Channels (II). (to be rectified prior to S/U)
- 24. Feedwater Control System Not required when shutdown. Possible over-(II) cooling if system reacts to seismic load by opening feed regulating valve. Feed pumps do not trip on high steam generator level or reactor trip
- 25. Main Turbine Overspeed Backed up by a mechanical overspeed device Control (II) which would probably trip due to earthquake stress or an overspeed condition.
- 26. Essential. Plant Parameter For use only if fire deactivates control Monitoring System (III) room or emergency shutdown panel. Should not be needed following UBE
- 27. Offsite Power (II) If normal onsite supplies are available, loss of offsite power has no effect
- 28. On Site' Nonclass 1E Forces reliance on IE supplies, all but Power (II) 'selecteo safety related loads are deenergized
- 29. Fire Detection Equipment Loss of all plant fire detection capability.
(III) SER states applicant will do a fire inspection witnin two hours following an earthquake
- 30. Fire Protection Water & Per FSAR "not required to safety shutdown l Gaseous Systems (II) the plant for any creaible fires." Have subsequently installed seismic Cat 1 stand-pipe systems at strategic locations (per j SER).
e 1
l 31. Nonessential Portions of No major effect l Fuel Pool Cooling &
Cleanup (II)
- 32. Fuel Handling (II) No major effect
- 33. Closed Cooling Water No major effect Chemical Feeders (II)
- 34. Demineralized Water Makeup No major effect
-(II)
l System / Component Effect of Failure ;
- 35. Potable Water (II), No major ef fect
- 36. Ultimate Heat Sink Main Backed up by auxiliary intake to supply Intake (II) the. essential salt water system
- 37. Offshore Outfall Conduit (II) Operator must snift to component cooling water emergency discharge lines
- 38. Condensate Storage Tank (II) Reduces AFW supply from 24 hr to 4 hr.
(Balance of Plant) & Pumps Allows cooldown to RHH operation point
- 39. Nuclear Service Water (II) Euphemism for " primary grade nake-up water." ho major snort term effects.
Lose makeup water for:
-b. Corponent Cooling Water
- c. Primary Makeup Storage Tanks
- d. Charging Pump Reservoir
- 40. Turbine Plant Cooling No major ef fects. Loss of cooling to:
Water.(II)
- a. Feed & Condensate Pumps
- b. Turbine & Generator
. c. Plant Air Compressors
- d. Aux Building Chillers
- e. Containrent Normal Coolers
- f. Sample Coolers & Cnillers Systems "a-f" are addressed separately
- 41. Process Sampling System (II) No major ef fects. Sampling of primary -
will oe done via the high radiation sampling system which cools samples with CCW
- 42. Non Radioactive Floor No major effect Drains (II/III)
- 43. Radioactive Floor Drains (II) Minor effects: Uncollected leakage from radioactive system pumps and valves
- 44. Chemical & Volume Control System (I/II)
- a. Volume Control Tank (II) Loss of automatic pressure control during cooldown while solid
System / Component Effect of Failure
- b. Boric Acid Batch Tank No major effect (II)
- c. Chemical Addition No enajor ef fect Tank (11)
- d. Letdown System from Back- Loss of normal letdown, loss of auto Pressure Regulator to _ pressure control while solid Radwaste Diverter Valve (11)
- e. lon Exchangers a Filters No major effect (11)
- f. Boronometer (II) No major effect. Requires increased boron sampling
- 45. Normal HVAC (11) Emergency systems take over in all rooms containing ESF or other safety equipment.
- 46. Hydrogen Purge Supply & H Recombiners Provide Backup 2
Exhaust (II)
- 47. Communications Only alarms are seismic Cat I, however repeaters for hand held radios may be manually switched to class 1E power supplies
- 48. Lighting Systems (11/111) Backed up by emergercy & essential lighting
- 49. Turbine & Generatur (11) a. Loss of steam load, then reactor trips D. Loss of onsite power, fast transfer to offsite power or class 1E busses transfer to EGD supplies
- c. Loss of turbine 1st stage pressure signal to:
- 1. Rod Blocks (not required when
[
snutdown)
- 2. Steam Dumps (can cause dumps to open or fail to open)
- 3. Rod Control System (not re-quiredwnenshutdown)
- 4. Steam Generator Level Program (not required when shutdown)
- 5. Feedback to Electrohydraulic Control (not required when shutdown)
l
.. l l
I System / Component Effect of Failure
- b. Loss of downstream system results in loss of heat sink
- 51. Main Condenser (II)
- Loss of heat sink, turbine trip, loss of steam dump capability
- 52. Steam Jet Air Ejectors (II) Loss of heat sink, turbine trip, loss of steam dump capability. Mechanical vacuum pumps may be used as backup
- 53. Mecnanical Vacuum Pumps (II) No effect
- 54. Turbine Gland Seal System Loss of neat sink
'(II)
- 55. Turbine Bypass (Steam Dump) Loss of steam dump capability if turbine (II) trips or excessive steam load if system fails to full (45%) load
- 56. Circulating Water & a.. Loss of RCS heat sink Traveling Screens (II)
D. Flooding from condenser expansion joints
- 57. Concensate & Feedwater (II) a. Loss of RCS heat sink if systen stops operating or a major leak develops.
- b. Excessive cooldown by failure of the system to shutdown can be prevented manually by shutting feed isolation valves (which also shut on ESFAS).
Feed reg valves fail as is on loss of ai r. Feed pumps do not trip on high steam generator level No major ef fect. Seismic Category II
'58. Steam Generator Blowdown (1/II) section isolates on AFW initiation Not required when shutdown
- 59. . Turbine Plant Chemical Addition (II)
- 60. Coolant Radwaste (II) No major effect
System / Component Effect of failure
- 61. Misc. Liquid Radwaste (II) No effect
- 62. Boric Acid Recycle (II) No major ef fect
- 63. Waste Gas System Possible venting of H2 from reactor
- a. Surge Tank Drain Pump coolant drain tank (II)
- b. Compressor Motor (II)
- 64. Radwaste Solidification No effect (II/III)
- 65. Non-Safety Process Radiation No effect Monitor (II)
- 66. Non-Safety Related' Area No effect Radiation Monitor (II)
- 67. Plant Air (II) a. Air powered valves fail safe
- b. Atmospheric steam dumps (Seismic Cat I)'are automatically powered f rom a backup nitrogen supply
- c. Some shutdown cooling valves must be reopened manually. NRC Staff is investigating
- d. Loss (shutting) of most sample system containment penetrations, steam generator blodown lines, and~contain-ment normal A/C chilled water
- 68. Post Accident Sampling Loss of sampling capability Systen (II) -
- 69. EDG Air Start System (up Receivers contain enough air for 5 starts to Receivers) (II)
- 70. ~ Component Cooling Water Loss of cooling to RCP seals and Spent Fuel Pool Cooling Water heat exchangers.
(I/II) Both events have been analyzed by SCE and found to be not serious
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Table 7.5-1 SAFETY-REl.ATED DISPLAY INSTRUMENTATION (Sheet 7 of 32)
Type Peremotee No. of Accuracy (d) of panel peuer System Stoseured(b) Chosene t e Renee(C) (t) seedeut(8) Recettee(f) No. It Due RPSOI ESF(hI ASIO) FF9tOI BSPIOI PettO) StSIM II Refueling water ta d 1 0 8001 5 I Centrol CR57 se X X (T005) level see.
Refueling meter tank 4 0-1001 5 t Centret CB57 Yes I I I (T006) levet reon Defueling water tank l 0-1001 3 a control CD59 fee R I I (1906) Leve? rene Centstament og level 2 0-1001 5 I Cent rol CR57 Tee I room NP safety tejectten 4 0-800A 5 L/I Control C357 Yes E E M pimap on/off toen j Centrol LP safety injeccles pimp on/off 2 0-1004 5 t/l room CB57 Tee X X g
e" ut oefety injection 2 0-25002 5 I Centrol CR57 Me I u header pressure Ib/in reon fD 1
M g LP safety injectten 1 5 l Cent rol CR58 see x M 0-600 '
e * " " * "" - O LP safety injection 2 0-400F 3 I Centrol C357 Tes I x **f I eder t - r.t.re r -
C LP safety injectlen 1 0-400F 5 R Centrol Ca59 fee X X 3 header temperature room Centaineemt sump velve 4 - - L Centret Ca57 fee x x g pasItten teen H t/l Safety injeetten pompe 4 - - L Centrol CR57 fes X X M etetsum sec treetat ten room volve positten LF lieeder selve 4 0-100L 5 I Centrol Cm57 se x x H peelttee room hM H
nr header selve .S 0-1001 5 l Control CR57 fee I I E W til poettien reem nr heeder het leg 2 0-t001 5 I Control CR57 na I a injectien, weise voce positten ur heeder to SC drein teak, satse poettles 2 - - L Centrol reon CR57 Tee I NN o
g t-p b
. S:n On:fra 2&3 FSAR SAFETY-RELATED DISPLAY INSTRUMENTATION
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V Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 9 of 32) eyeess paressear lesseured(b) n.. .f Chesente anage(e) accuracy (d)
(1)
D.
asedeut(*) Recettem(II remot pe.
power It Bee RFSOI ESPOI aSIOI PF9tOI RSPIOI FellOI SEBIOIIII Safety injoettee task 4 - - L Centret CR37 De I drate & fill velse room peeatten Safety injecttee tant 4 - - L Centret CR37 po I pg JBil velve poettles rose Safety tejeettee task 4 - - L Centret CR57 Tee I I m2 *" ** 'III'* *"
Centatesset weet sneedei 1 - - L Centrol CR57 fee 1 I velve poettles roam Safety injeettee elmet 4 - - L Centrol CR37 Yes I vetve teoksee costret room 12 velve peettles N Shutdown easitag heet 2 0-550 5 Centrol (137 Tee
. I I o enchanger, tulet Ib/te.2 s reen m Le ,oseere W
M l Sheedene esottog heat 2 0-400F S I costret CR57 Yes I w g eschenger, outlet rose fr*
G temperature L*8 sheadsen emelles flee 1 0-100L 3 t teatret CRSS no I m
, ..ete.t t.e - itt r.o.
Shutdese emelleg 1 - - L Control CR30 No I bypees weive peeittoo rsom Shuteese esottag llee 1 0-10,000 S I Centret CBS8 We 2 D 1 flow get/ete reen y E
m.
w H
W t
hE aM ze f
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Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 10 of 32) ser-mtw so. .f Ace ==,tdi '.7'"
f puest pesar e,eten nesserod m) -i. nosente) sti no.doette) toestles(f) so. I .me e ,em) espm) Ast#) root #) aset#1 res#1 onst#W)
Steen generator 1 4 / S .C . 0-805L 5 1 Coettet C152 Tee I 1 1 E levet reen Steam esmerater 1 1/ S .C . 9-195L 5 I RSP IA2 fee I I E toget Ste'n e geenreter 1 1/S.C. S-195L 3 R Centret CR59 fee E E I.
Levet roam Steen reter 4/S .C . 9-1200 5 t Centret CR32 Tee I 1 I pressere Ib/te.2 a room Steen somerator 1/8.C. 1289 5 t RSP IA2 Tee E I I to
.. e 0./t..re i
g Steen genereter t/S.C. 9-1290 3 R Castret CR59 Tee q pressere Ib/te.3 e team I I I g p
O Ut Steam du e to t/S.C. 0-190L 5 t/C Centret CR32 Tes a
,i, et.o.~re -i. re.m Q y poettlee DJ Steam emp to t/8.C. A-105L et.o. r.. -i.
- 5 1/C RSP RA2 Yee E poeaesee sus in j Anettlery feedseter 1/ S .C. 0-800 5 I Centret CR52 Tee I 3 fles est/ete team %
X Anettlery feedseter I/S.C. 9-000 3 S Centret CR59 Tee E I flev get/ete toen g
Aesittlery feedenter 2/3.C. M
- - L Centret CR52 Tee 1 tn vetse poettles reen 'd Aesittlery feedseter 2/S.C. - - L Centret 'CES2 Yee I E s weive poeittem rose ~
N Aemillery feedester 1/S.C. 0-Icet 5 t Centret CRS2 fee I I motor-drives pep room %80 discherte velve h tw$
poettles Anettlery feedeeter 1/S .C. O-tost 5 t Caetret CR52 Tee I I turbtes-desven peep team y
discharge volve tvl peeitten >
.te steem t. et.em. iis.C.
drives p g (F140),
velve peettien
- - i. C t.ei seem CR52 Tee x a e@
Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 11 of 32)
Type perense.r me. of Acevreer(d) .f ten.1 re-er System meneeredib) Chennele penge(*I (t) Seedeut(*) location (f) pe. It Bue G PSO) R$fSI ASISI PPDIO) R9FIOI PettO) St910NII nota steem to steme- t /S .C . - - L Cent re t CR32 Tee I delven peep (F140), room valve peettien netp steam teetetten 1/ S .C . - - L Centrol CS57 Tee 2 I veive poeItten esom .
,emin eteen testatten t/3.C. - - L Centrol CRS2 Tes I I valve poettien team Feedweter testatten 1/ S .C . - - L Centrol CRSF Tee 3 I valve poettien room Feedweter toeletten valve poettien 1/3.C. - -
L Centret re==
CR32 Tee I 1 h
- 2 M
Tvrtime-delves aunttlery Fw p e p, 1 - - L Centret room CRS2 fee I h
o
- turbine stop velve m f
6e poettien e Turbine-driven 1 - - L Centret CR$7 Yee I I w austilary Fw pump, reen to turbine stop valve W
,. Poettien ,,g
(#3 tiectric motor-driven 1 = . L Cent res CR57 fee R I g 1
g eeni1tery FW pump (Ptet) en/off reen W Eteetrie notor-delven 1 0-1504 5 L/t Centret CRS2 Tee a semittery tw pump seen O (F141) en/off Steen desar to 1/S .C . - - L RSP IA2 Yes R I et.os, er., vaive E 4
peelt ten taw steen generator 4 - - L RSP L42 fee R R tal gg
,re.. ore set - int t,,
re.et Condeneste pump m/ett 4 - - L RS P 1A2 No R I h M I Condeneste storese 2 0-10R 5 I Centret CR55 No X X y
tank levet roce e,.g h Camdeneste storage 2 0-1001 5 1 RSP L42 no I x x
-s OM tank teett ZD I
b ~
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Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 13 of 32)
Tr**
Pare nter me. of Acewecy td> of Famel reser s,eten sneeser.dt *) a- t e nee (4 (2) s.edeut ta) toesti.ett) .. is ese mesm) Esem) astm) proimi aset#) ren #1 sat #m)
Chesteet storees task 2 0-1001 5 t Centret CR57 Tee I tesel rose Chesteet odottles flee 2 0-23 5 t Castret CR57 fee R Item get/ete team Chesteel adottles 2 . - L Control CR57 fee X E p g ,se/eff reen Cheetset addities, 2 . . L Centret CR57 Yes R R testatten velve reen peettlee Sprey chanteel fles 2 . . L Centrol CR57 Tea R R [
e.mtrei . .
peettlee too.
o N
e sprey peep en/off 2 0-550 5 L/1 Centret Cm37 Tee R p Ib/te.2, ,,,,
R
,o,,
- n. Sprey lies flem I' s 0-1001 5 t Centret CR57 fee I !
l reos y
- e. Spray itse 1 pressere 3 e.
0,-5001 ,t. 2 R Centret CR59 Tes X X w g r
,e, Spray lies 1 0-400P 5 R Centret CR59 Tee X I temperatore reen Sprey header velve 2 0-100L $ t Control CR57 fee X E poeittoo room t:7 Sprey header volve 2 - . L Centrol CR37 Teo R H poeittoo resa y Chesteet eteroge tash pressere 1 0-25 Rb/te,y g S I contret reae C337 ile I h M
.,re, it- ,ree- 2 0-550 Ib/te.2,
$ t Co.trei CR57 Yes R R g ,,
gn >
H *T1 sprey lies temperatore 2 0-400F S I Centrol CR57 Tee I E N I'3 R? -
ein hE
-H o tus 2: t1 9
Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 14 of 32)
Peteenter me. of Act or ec y I pen,g pg ,er Systee neeeer,4(b) Channels Range (c) (2) g,%g (e ) tecationif) No. IE Rue RPS @) ESF AI ASI @) Pret@) espt @) 74539) Sigt@)(I)
Circulottog water pump & . . L 3$P t,4 2 No X 3 9
e D
D e k Y' a t: : :
I V E w E ~3
.s t 30 C
N
- 2. tJ5 (n
64 In8 aa
%C
. OM 2$ 0
. Table 7.5-1 Dw SAFETY-RELATED DI3 PLAY INSTRUMENTATION (Sheet 15 of 32)
Tr,e Foreunter De. of Accuracy III of Penel Pueer s,.to. ses o.edt *) Che.neio asse.(e) m s.edest (*) tacert.afft no. it w. nesel estml assm) rectal aseim) emen e) ensim>(8)
Comtatom et earner.e 2 $ i Centr.: tio) Tee a redtottoo
- 1. Geeesee -4 10 to cabinet a n l1 l *'**
10'Iott/ce
- 2. Portgestate 10 to e
l@eCstreI
- 3. gadine 80 t o Centstament etrborne 10' vClice 1 3 R Cent re t LIO) Tes a R I 1 radiettee team same se tedienter retenet eree l
Comtsameset pressere 4 -4 to 20 $ I Centrol CRS7 Yes R I E ib/fe.2g r, R
. Comtatammet preesere 4 -4 to SS 3 Un I Centret G37 Tee I E 3 th/in.Ig team i
N
" Centstammet pressere 1 -4 to SS 3 8 Coetrol CRS9 ' fee I I E It/te.23 roen Centenmuset stameghere 2 0-40er S Centret temperature I MST Tee I I team Chitted unter to 1 - - L Centret MS7 fee I2 E I
==-met asettes emite roen R Chilled enter fren 1 - -
L Centret CRS? Tee coeltog emite E I I team tll$
C.mte-.t - - - t H Co.tr.i C.S, f.e . .
diocharge teetettee sg rose votes postgles Centes w atmooghere 1 0-40er 3 9 Centret CBS' Yes I M
teesetatore I toen g g yy Cameetament dame 4 - -
L Ceintrol G40 fee in etreetator en/off room I I H
- A8
=,
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M >4 E9 D - .
- ^ %, %
w s Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 16 of 32)
Type Fereneter Iso. of Acevracy(49 of Penel Pcwer g g g g,y g g System fleasured(b) Channelo Range (1) Reedoet Location No. It Bus RPS F5F ASI PPDI PAffl BISI Containment emergency 4 - - L Centrol CR60 Tee R E cooling unit on/ofI toom CCW free emergency 4 - - L Control Cs60 Tes I coeling unite tone CCW free emergency 4 - - L Control CR57 Tee X I cooling waite enom CCW to emergency 4 - - L Control C960 Tee R tooling enita ro.e CCW to emergency 4 - - L Control CR57 Tes I E M cooling unite race j geersency radiation 1 Igo 10 I Control L-103 Tee I monitorlag system to room O N eree/h %
- et Emergency radiation 23 *
[ a monitoring system 5 1go 10 10 R Control tone L-101 Tee a y
@ aree/h g, W
Containment high 2 Igo 20 I Control L-405 Tee E q range radiation 10 room gn monitore Sade/h Containment high I tgo 20 R Control L-405 Tee I tense vedtatten 10 raor monitore Rade/h O
Containment Freesere 2 0-200, 5 I Control r957 Tee I N lbfin g ruom CA Contalement Freesere 1 0-200 Ib/la g 5 R Control room C259 Tes I h
M Containment Enreal seep 2 10*-2* to th' 7" 12 l Contral race C957 Teo I
$m gh Fe o %
3 cn.
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Type Peremeter De. of securacy(d) of ~
panel power Systee lesseured(b) Cheestels Gene,(t ) (t) geodout('I lacettee(I) No. IE Due R PSDI ESPOI A310) PPDIDI RSPIOI PasttO) 3{stOIIII feettletion testatten 4 - - L Centrol CR60 Tee I I damper poettles rose Att cleas' amit 2 - - L Centrol LI% Tee I heater en/off toen
, cableet eres ,
Air cleanup unit 2 3 R Centrol LI% No R ett fleur room cabinet eres Air elemmep sett 2 3 R Centrol Ll% IIe I (4 differential pressere rose D cabteet U ere.
g g Air eleseep unit 2 5 I control LtM ue I 0 u
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p.re -t.e no. .I accareer,,, .e remet power System steeeereath) Channele sange(c) (t) needeetf eI tacettee(f) No. It eve RPS@I r$FOI ASINI Pf0lOI RSFINI PafE10I SISIOIIII COf sepply header 2 0-200 S I Centrol Cau pe x I pressere tblin.Is rose Car heat eschenner 2 0-200F S I Centret Cau Ice 2 E telet temperatore Ib/in.Is rose as COf heat eschemeer 2 0-150F S I control Ceu tee I E g settet temperatore rose n .
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D mr E e . . . - 3 S T e5 r1 A a(
c L c E A -
R ) 4
- t 9 0 Y f 0 0 5 2 e .
T g = . - 1 1
E a s 6 0 F a A
S e I.le m 2 2 2 2 4 4 e
am t
e 6
e r
) r e r rb o p e e(
td e
aw t
rt em t
t a
r A.ma su e s r e0 m8 n ea m e b b u
e a t r-a-
p-o6 sf t o t
e.
y t g_ .s t r -
d e
4 s e et ee W T ee ss st es Mte g s
a M.o i.
ea it Ds 5
2 1
5 2
1 b se t
e ii l
y g
M. w
Table 7.5-1 SAFETY-RELATED DISPLAY INSTRUMENTATION (Sheet 30 of 32)
I '
F*N' ster flo. of Acre, '
System gemeeered(b) Chennele Reage(C) ( ) racy *Ih t(*) W ettm(f) e mS) m@) mN gN yN gg yy
,me, wi -i - > - - i. C.etc., C. . ,ee ,
re W dier k go 2 * - L Control Cs41 IIe g pressere low eleve ,.
888 I - - Campeter ret sin gas g 9
O k
O w .
9%3 5
0
' m W8 k
E 5
!a 5 .
m o,
bl
!sf 2,
am NE BM xa b
l r O
]
i Table 7.5-1 SAPETY-REIATED DISPIAT INSTRUMENTATION (Sheet 31 of 32) i b. PPS - Plant Protection System '
DNBR - Departure from Nucleate Boiling Ratio l RCP - Reactor Coolant Pump
- IPSI -
Inw-Pressure Safety Injection HP - High Pressure
! LP - Iow Pressure l SG , - Steam Generator /
1 FW - Peedwater -
CCW - Component Cooling Water {-
f c. Safety-related equipment supplied by a 4.16-kV bus is provided with an asumeter integral with the hand l switch module. g
- d. The accuracy given is that of the complete channel and includes for example the measuring element, S 1 ." transmitter and indicator. The figures quoted are representative for the type of measurement and 2.
! T' are based on average values. 2 4 w k , H I e. I - Indicator (analog) C,
! L - Light (on/off) .g 4
R - Recorder
- I/C - Indicating Controller i f. Control room includes the back panels not directly visible to the operator. The control panels that e are in the operator's console area are denoted by CR and the panels that are in the control roon U cabinet area or at the remote locations are denoted by L. **
l
! g<
l RSP - Remote shutdown panel used when evacuating control room. s i
55 3 Prefixes ZL and MS when referring to a piece of equipment, denote two lights, open/close for valves, @@
j start /stop for motors, and on/off for heaters. gQ
)
j aa hC 5N ze I
I a
_ _ _ _ . _ _ _