ML20050B162
| ML20050B162 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/29/1982 |
| From: | Vincent R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-06-10.A, TASK-6-10.A, TASK-RR NUDOCS 8204050059 | |
| Download: ML20050B162 (82) | |
Text
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o O Consumers Power Company Generna othees. 212 West Michigan Avenue. Jaca son, MI 49201 *1517) 788-0550 March 29, 1982 C,
A fiECEn g 9
APR 021gg y 'Z Dennis M Crutchfield, Chief yDp,k7 Operating Reactors Branch No 5 g
ttrm 3 10 Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 1G DOCKET 50-155 - LICENSE DPR -
BIG ROCK POINT PLANT - SEP TOPIC VI-10.A ELECTRICAL, INSTRUMENTATION AND CONTROL PORTIONS OF THE TESTING OF RTS AND ESF Attached is the Consumers Power Company Evaluation of SEP Topic VI-10. A for the Big Rock Point Plant.
As you will note, this evaluation describes in detail the testing performed on the Big Rock reactor trip and engineered safety feature instrumentation.
We trust that this information will satisfy the staf fs' needs.
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Robert A Vincent Staff Licensing Engineer CC Administrator, Region III, USNRC NRC Resident Inspector-Big Rock Point 1 pages 0?ye L
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SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY TOPIC VI-10.A rp0182-0856a140-42
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SEP TOPIC VI-10.A Abstract This report documents the technical review of SEP Topic VI-10.A associated
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with the electrical, instrumentation and control portions of the testing of the Reactor Protection System and Engineered Safety Features for the Big Rock
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Point Nuclear Power Plant using current licensing criteria.
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SCP TOPIC VI-10.A Table of Contents SEction Page 1.
INTR 3 DUCTION.
1 1-3 f2.
CURRENT LICENSING CRITERIA >.
.T 2.1 LICENSING CRITERIA FOR THE REACTOR PROTECTION 1-2 SYSTEM (RPS) 2.2 LICENSING CRITERIA FOR THE ENGINEERED SAFETY.
3 FEATURES (ESP)
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'3.
SYSTEM DESCRIPTION.
3 - 20
' 3.1 bESCRIPTIONOFTHEREACTORPROTECTION, SYSTEM.
3 - 18 3.1.1 Automatic Actions 4
4-5 3.1.2
' Power Supply.
3.1.3
~ Alternate-Power Supply 5
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5 - 11 3.1.4 RPS Component Description 3.1.4.1 Sensors 5-6 6-7 3.1.4.2 Logic Unit 3.1.4.3 Power Switches.
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7-9 9 - 11 3.1.4.4 Annunciator Control Units 11 3.1.4.5 Operations Recorder 3.1.5 RPS Trip Function Description 11 - 17 3.1. 5.1 Manual Scram Actuation.
11 - 12 12 3.1.5.2 RPS Bus Undervoltage.
12 3.1.5.3 High Neutron Flux j
3.1.5.4 Protection Against Picoatmeter Circuit Failure.
13 3.1.5.5 Short Reactor Period.
13 3.1.5.6 High Reactor Pressure 13 - 14 rp0182-0856bl40-42
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s-S ction Page i
3.1.5.7 Low Reactor Water Level 14 3.1.5.8 Low Steam Drum Water Level................
14 3.1.5.9 Closure of Main Steam Line Backup isolation Valve 14 - 15 3.1.5.10 Simultaneous Closure of Recirculation Line Valves 15 3.1.5.11 High Condenser Pressure 16 3.1.5.12 High Reactor Building Pressure.
16 - 17 3.1.5.13 High Scram Dump Tank Level................
17 3.1.6 System Response Time 18
3.2 DESCRIPTION
OF THE ENr.INEERED SAFETY FEKIURES SYSTEM.
18 - 20 4,
. EVALUATION AND CONCLUSTONS.
20 - 31 s
4.1 REACTOR PROTECTION SYSTEMS.
20 4.1.1 Testing During Refueling 20 - 24 24 - 25 4.1.2 Testing During operation.
4.1.3 Test Overlap for Response and Timing Verifications 25 - 28 4.1.3.1 Overlap in Response Verifications 25 - 26 26 - 28 4.1.3.2 Overlap in Timing Verifications 4.1.4 Reactor Protection System Testing Program Conclusions 28 4.2 ENGINEERED SAFETY FEATURES (CONTAINMENT SPRAY SYSTEM) 28 - 31 4.2.1 Description of the Containment - Spray System (CSS) 28 - 31 Test Program 4.2.2 Entineered Safety Features System (Containment Spray.
31 System) Test Program Conclusions Attachments
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CP Co Dwg 0740F30760, Sh 2 5
CP Co Dwg 0740G40122 6
CP Co Dwg 0740G30107 7
CP Co Dwg 0740G30114, Sh 2 8
CP Co Dwg 0740G40125, Sh 1 9
CP Co Dwg 0740G30112, Sh 1 10 CP Co Dwg 0740G40107, Sh 1 11 CP Co Dwg 0740G30734, Sh 1 12 CP Co Dwg 0740G30112, Sh 2 13 CP Co Dwg 0740G30111, Sh 1 14 CP Co Dwg 0740G40121, Sh 1 15 CP Co Dwg 0740G30103 16 CP Co Dwg 0740G30731, Sh 1 17 CP Co Dwg 0740G40123 18 Instrument Data Sheet 75 19 Instrument Data Sheet 78 20 Response Time Testing - LER Search 21 Big Rock Point Technical Specifications References 1
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1 TONR 3-82 SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY TOPIC VI-10.A 1.
INTRODUCTION SEP Topic VI-10.A " Testing of the Reactor Protection System (RPS) and 2ngineered Safety Features (EST) System Including Response Time Testing" deals with the design, testability and operability of the RPS and the ESF system. The RPS and ESF designs should permit periodic testing and such testing should be performed such that a high degree of availability of these systems can be demonstrated. The testing should also verify ade-quate response times of these systers.
The objective of this review is to evaluate the Plant design and testing program to ensure that all RPS components are included in the component and system tests, that the scope of the periodic testing is adequate and that the test program conforms to (or meets the intent of) the current licensing criteria as described in Section 2.
This review considers the containment spray system as an example that is typical of all ESF systems. A review of the Plant design and testing program is conducted to ensure that all containment spray system portions of the ESF components (including pumps and valves) are included in the component and system tests, that the scope of the periodic testing is adequate and that the test program conforms to (or meets the intent of) the current licensing criteria as described in Section 2.
The frequencies for checks, calibrations and testing of the RPS and the containment spray system are identified in the Plant's Technical Specifi-cations Items 6.1.5 and 11.4.3.4, respectively. As in the case of the NRC review of this SEP Topic for Palisades, this review will not conclude whether the Plant complies or does not comply with test frequency cri-teria.
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2.
CURRENT LICENSING CRITERIA 2.1 LICENSING CRITERIA FOR THE REACTOR PROTECTION SYSTEM (RPS)
GDC 21, entitled " Protection System Reliability and Testability" states in part that:
The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capa-bility to test channels independently to determine failures and losses of redundancy that may have occurred.
Regulatory Guide 1.22, entitled " Periodic Testing of the Protection System Actuation Functions" states in Section D.1.a that:
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The periodic tests should duplicate as closely as practicable the performance that is required of the actuation devices in the event of an accident.
Regulatory Guide 1.118, entitled " Periodic Testing of Electric Power and Protection Systems" states in part in Section C-12 that:
Safety system response time measurements shall be made periodically to verify the overall response time (assumed in the safety analysis of the Plant) of all portions of the system from and including the sensor to operation of the actuator.
The response time test shall include as much of each safety system, from sensor input to actuated equipment, as possible in a single test. Where the entire set of equipment from sensor to actuated equipment cannot be tested at once, verification of system response time may be accomplished by measuring the response times of discrete portions of the system and showing that the sum of the response times of all portions is equal to or less than the overall system requirement.
IEEE Std-338-1975, entitled " Periodic Testing of Nuclear Power Gener-ating Station Class 1E Power and Protections Systems" states in Section 3 that:
Overlap testing consists of channel, train or load group verification by performing individual tests on the various components and subsys-tems of the channel, train or load group. The individual component and subsystem tests shall check parts of adjacent subsystems, such that the entire channel, train or load group will be verified by testing of individual components or subsystems.
Regulatory Guide 1.22 states in Section D.4 that:
Where actuated equi.pment is not tested during reactor operation, it should be shown that:
a.
There is no practicable system design that would permit operation of the actuated equipment without adversely affecting the safety or operability of the Plant; b.
The probability that the protection system will fail to initiate the operation of the actuated equipment is, and can be main-tained, acceptably low without testing the actuated equipment during reactor operation, and c.
The actuated equipment can be routinely tested when the reactor is shut down.
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2.2 1.ICENSING CRITERIA FOR THE ENGINEERED SAFETY FEATURES (ESF)
All criteria listed in Section 2 of this report are applicable to the ESFs.
In addition, the following criteria are also applicable.
GDC 40, entitled." Testing of Containment Heat Removal System" states the containment heat removal system shall be designed to permit sp-propriate periodic pressure and functional testing to assure:
a.
The structural and leak-tight integrity of its components.
b.
The operability and performance of the active components of the system.
c.
The operability of the system as a whole and under conditions as close to the design as prattical the performance of the full operational sequence that 1 rings the system into operation, in-cluding operation of applicable portions of the protection systems, the transfer between normal and emergency power sources and the operation of the associated cooling water system.
Standard Review Plan, Section 7.3, Appendix A, entitled "Use of IEEE-Std-279 in the Review of the ESFAS and Instrumentation and Controls of Essential Auxiliary Supporting Systems" states in Section ll.b that:
Periodic testing should duplicate, as closely as practical, the inte-grated performance required from the ESFAS, ESF systems and their essential auxiliary supporting systems.
If such a " system level" test can be performed only during shutdown, the testing done during power operation must be reviewed in detail. Check that " overlapping" tests do, in fact, overlap from one test segment to another. For example, closing a circuit breaker with the manual breaker control switch may not be adequate to test the ability of the ESFAS to close the breaker.
3.
SYSTEM DESCRIFTION
3.1 DESCRIPTION
OF THE REACTOR PROTECTION SYSTEM The function of the reactor protec. ion system (RPS) is to initiate rapid control rod insertion (scram) for reactor shutdown in the event that certain undesirable conditions develop in the nuclear steam sup-ply system or certain auxiliary systems. Depending on the initiating cause for shutdown, other secondary protective functions such as containment isolation, actuation of the scram dump tank valves and turbine trip are also initiated by the reactor protection system.
As can be seen in Attachments 1 and 2, the RPS features two parallel safety channels. Each Channel features its own power supplies (see ), chains of sensor trip centacts and logic circuitry to effect reactor protection and certain other protective functions.
rp0182-0856a140-42 J
4 The channels are designed on the fail-safe principal such that de-energizing the channels will initiate the protective functions.
In addition, failure of a single major component or power supply does not prevent a "real" scram nor causes a spurious scram.
Generally (exceptions are described in the following paragraphs),
there are four independent sensor switch inputs for each trip func-tion; all operating at the same setpoint. Two of the sensor switches are connected to each protection channel. Generally, trip operation of only one of the two sensor switches is required to de-energize and thereby operate its associated protective channel. To initiate the protective functions, however, both RPS channels must operate.
3.1.1 Automatic Actions As can be seen in Attachment 1, the following actions occur on any scram initiation:
a.
All control rods will be driven into the core at high speed.
b.
The scram and cause of scram will be alarmed in the main control room.
c.
The reactor containment intake and exhaust ventilation ducts will close.
d.
The turbine will trip.
e.
The scram dump tank vent and drain isolation valves will close.
f.
The scram dump tank equalization vent valves will open.
When a scram is caused by high enclosure pressure, low water level in the reactor or loss of station auxiliary power the above actions will be initiated as well as closure of all reactor containment penetrations which are not considered as integral parts of the containment. Due to fail-safe design, loss of instrument air will operate the control rod inlet and outlet valves to shut the reactor down.
3.1.2 Power Supply Each reactor protection Channel is powered by a separate motor generator set (MG Set No 1 or MG Set No 2; see Attachment 2).
Each MG set has a self-contained exciter and output voltage regulator (see Attachment 2a). The motor input is three phase 480 volts from Bus lA or 2A.
Each generator is rated at 6.25 Kva and supplies 120-volt ac single-phase power to each protection system Channel as well as to various critical instrumentation systems. Each MG set is mechanically coupled rp0182-0856a140-42 t
5 to an inertia flywheel enabling it to ride out minor system voltage disturbances. Loss of 480-volt power to the motors is alarmed immediately in the control room (see Attachment 11).
Generator output is restored (after motor is up to speed) by a manual reset.
3.1.3 Alternate Power Supply Loss of power supplied by one of the MG sets described above would cause only one reactor protection system Channel to operate. By manual actuation of the alternate power control-1er in the main control room, an alternate 120-volt supply from Panel 1Y can be switched to either of the two protection-buses or to the Neutron Monitoring Bus No 3 (which is normally supplied by the 125-volt, de station battery through an inverter). This alternate power supply control is interlocked so that only one of these three buses can be supplied at any one time from Panel 1Y.
This alternate supply (Panel 1Y) is normally fed from the 30 Kva Instrument and Control Transformer 1A on Bus 1A unless this source of power is lost, in which case an automatic throwover operates to supply power from the 30 Kva Instrument and Control Transformer 2B on Emergency Bus 2B.
The control rod position indication is normally fed through Panel 1Y; however, a third MG set can supply power to the control rod position indication system in the event of total loss of power to Panel 1Y.
This MG set, which starts automatically upon loss of power to Panel 1Y, is powered by the station battery at 125 volts de and has a single-phase 115 volts ac output.
With the power selection switch in the " Pull For Bus 3" posi-tion, this MG set is interlocked off and Instrument and Control Transformer 1A supplies power to the control rod position indication system through Panel 1Y via alternate contacts (see Attachments 2 and 3).
4 3.1.4 RPS Component Description 3.1.4.1 Sensors The scram sensor switches are of two general types, either contact type or voltage level type. The contact type is used with pressure, water level, undervoltage, manual switch and valve position sensors. As previcumly stated, there are generally four independent switch in-puts for each trip function. Two of the sensor switches are connected to each protection channel. Trip operation of one of two of the contact type sensor switches is required to op-erate its associated protection channel.
rp0182-0856a140-42
6 Exceptions to the general case are the manual scram, recirculation line valves closed and the ac undervoltage trip function inputs. All trip functions are separately described in the " Trip Function Description" section.
The voltage level type of inputs are derived from transistorized trip units in the neutron monitoring system (refer to Attachments 3 and 4).
A total of five inputs of this type are connected directly to the logic unit in each protection channel. Three inputs are derived from the power range picoammeters and two from the intermediate range Log-N period amplifiers.
These RPS inputs and their associated logic are also described in the " Trip Function Description" section.
The contact type inputs are normally in the closed condition keeping the OR gates conduc-ting, and thereby maintaining a 26 volt, de input to the logic unit. This logic unit voltage input drops to zero whenever one contact input opens. The 26 volt, de circuit is sup-plied by solid state power supplies located in each logic unit.
The voltage level type circuits normally main-tain a
- 11-volt, de input to the logic unit.
This voltage drops to less than one volt on trip level signals. Power is supplied by the individual trip unit power supply located in the neutron monitoring amplifier units.
3.1.4.2 Logic Unit l
The logic unit is a transistorized unit that performs a rapid low-power switching function upon receipt of a trip signal from a sensor or combination of sensors. RE03A serves Protection i
Channel 1 and RE03B serves Protection Channel 2.
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The logic unit contains circuit breakers and j
push-button swit ches which can be operated anu-ally to simulate sensor operation (contact opening) for testing purposes. The contacts designated as " period bypass" (1K6) are closed to bypass the period trip function whenever any two out of three range switches of the power l
range channels are placed in a range of 4% power or greater (refer to Attachments 1 and 3).
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7 The self-contained logic unit power supply is a solid-state, 26 volt, de supply which powers the contact-type sensor circuits and also supplies power to the logic circuitry at a level of 16 volts dc.
The normal output of the logic unit is a low-voltage de signal to the control rod scram circuitry and the reactor containment penetration closure circuitry of the power switch. This output drops to less than one volt upon receipt of appropriate trip signal inputs to the logic unit.
3.1.4.3 Power Switches The power switches (RE04A and RE17A for Chan-nel 1 and RE04B and RE17B for Channel 2) perform a rapid electrical switching function through the use of a combination of five relays. These power switch coils (K1 through K5) are all nor-mally energized. Upon loss of input from the logic unit, the coils will de-energize to initi-ate the following actions:
a.
Scram valve pilot solenoids S0/NC27A-A2 through S0/NC27A-F5 will de-energize as a result of contacts from relays K3, K4 and K5 of RPS Channel 1 opening. When the second set of scram valve pilot solenoids are de-energized by RPS Channel 2 (such that both Channel 1 and Channel 2 solenoids are con-currently de-energized), the contrrl rods will be scrammed. Control rod scram occurs when the scram valve pilot soler.oids de-energize to vent the air supply from both the 32 scram inlet and 32 scram outlet valves CV/NCG9 and CV/NC10. Upon venting the air supply, these valves will automatic-ally open to allow the control rod drive accumulators to discharge into the control rod drive c linders to force the rods into the core and also to relieve the pressure on l
top of the rod piston to the scram dump tank l
(see Attachment 5).
b.
Master scram pilot valve solenoids SO/NC22A (for RPS Channel 1) and S0/NC22B (for RPS Channel 2) also de-energize whenever power l
l switch unit relay K3 or K5 de-energizes (see Attachments 1 and 5).
This serves as a backup means of cperating the 32 air-supplied scram inlet (CV/NC09) and 32 scram rp0182-0856a140-42 l
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l
g outlet (CV/NC10)' valves which open on air failure to cause control rod scram. As can be seen in Attachment 5, de-energizing both solenoid valves (SO/NC22A and B) results in venting the air supply to the scram inlet and outlet valves. Both channels of the RPS must operate to vent the scrua valves.
c.
Scram dump tank vent and drain isolation valves CV/NC11 and CV/NC12 (see Attachment 5) close whenever both solenoid valves S0/NC22C (RPS Channel 1) and S0/NC22D (RPS Channel 2) de-energize. These solenoids de-energize whenever power switch unit relay K3 or K5 de-energizes.
In addition, solenoid valves S0/NC22F and S0/NC22G (for RPS Channel 1) and S0/NC22H and S0/NC22J (for RPS Channel 2) de-energize approx-imately 60 seconds after relays K3 and K5 de-energize. De-energizing these solenoid valves allows the dump tank equalization vent valves CV/NC14 and CV/NCIS to open on loss of supply air. As shown, both channels of RPS must oper-ate before valve CV/NC14 or CV/NCIS will open.
d.
A turbine trip is effected whenever power switch relay K3 or K5 de-energizes and the mode switch S4 is not in the " shutdown" position ("run" is the S4 position during power operation). As can be seen in Attachments 1 and 6, relays IKSE (RPS Channel 1) and 2K5E (RPS Channel 2) drop out to energize turbine trip solenoid HTS (Scheme 9604) whenever K3 or K5 de-energizes. As shown, both RPS channels must operate to effect the turbine trip.
e.
Containment ventilation isolation is also effected whenever power switch relay K3 or K5 de-energizes and the S4 mode switch is not in the " shutdown" position.
(In
" shutdown" position, ventilation valve closure is transferred to isolation valve control relays K1 and K2.) As shown in Attachments 1, 7 and 8, relays IK5A, 1K5B (for RPS Channel 1), 2K5A and 2K5B (for RPS Channel 2) de-energize whenever relay K3 or K5 de-energizes. This results in relays SVX1 and SVX6 de-energizing (see Attachment 7, Scheme 8501) thereby dropping out Sole-noids SV/9151, SV/9152, SV/9153 and SV/9154.
De-energizing these solenoids results in the
' closure of containment Ventilation Valves CV/4094, CV/4095, CV/4096 and CV/4097 rp0182-0856a140-42
9 (see Attachment 8).
As shown, both RPS channels must operate to effect closure of the supply and exhaust ventilation lines that penetrate the containment.
f.
The balance of the containment isolation valves are closed upon de-energizing power switch unit relay K1 or K2.
As shown in Attachments 1, 9 and 10, relays IK4A, 1K4B (for RPS Channel 1), 2K4A and 2K4B (for RPS Channel 2) de-energize upon contacts from either relay K1 or K2 opening. This results in Solenoid Valves SV/4922 and SV/4876 (these are typical containment isolation valves, see Attachment 9, Scheme 6505) de-energizing to permit containment Isolation Valves CV/4117 and CV/4027 to close (see Attachment 10).
As shown, both RPS channels must operate to permit valves CV/4117 and CV/4027 to close upon loss of supply air.
The operation of Power Switches A and B is monitored by seven amber lights per RPS Channel (see Attachment 1, Units RE04A and RE17A for Channel 1).
These lights are all normally energized. The lights will de-energize whenever the appropriate power switch relays (K1 through K5) de-energize.
These lights are used to indicate which protective functions have been initiated by the RPS.
Combined Power Switches A and B per RPS Channel (ie, RE04A and RE17A for Channel 1 and RE04B and RE17B for Channel 2) contain 32 external switches to facilitate testing of control rods, one at a time.
Each switch controls 115-volt, ac power to one control rod scram pilot valve solenoid. Hence, manual operation of Switch A2 on Power Switch A of each protection Channel will scram Rod A2.
Five other switches facili-tate separate testing of isolation trip circuits, ventilation and turbine trip circuits, dump tank valve trip circuits and master scram valve trip circuits.
3.1.4.4
' Annunciator Control Units Each protection Channel contains two annunciator control units (RE02A and RE02C for Channel 1 and rp0182-0856a140-42
10 RE02B and RE02D for Channel 2) which contain 26-volt, de relays and 115-volt, ac relays. These relays perform annunciator functions as well as trip bypass control functions associated with Operation Selector Switch S4 located on the con-trol console (see Attachment 1).
With reference to Protection Channel 1, the RE02A unit contains Relays K1 through K16 which are normally energized by the 26-volt logic sen-sor contact circuits such that, upon loss of power (as is the case for a trip input), the relay corresponding to the trip function will de-energize closing a contact in the 125-volt, de circuit to the Station Annunciator (see 1). When either sensor input operates in the other protection channel, an identical sequence occurs and closing of its series contact in the 125-volt, de circuit to the Station Annunciator will then alarm the scram and indicate the cause of scram. Opera-tion of only one RPS Channel will alarm a Channel scram (see Relays K19 and K20, RE02C, and Annunciators 2 and 3, Attach-ment 11) but will not alarm the cause. The specific sensor or sensors which did cause the Channel scram can be identified only on the Operations Recorder (a description of this re-corder is provided in Section 3.1.4.5).
Annunciator Control Units RE02C, serving Protec-tion Channel 1, and RE02D, serving Protection Channel 2, each contain Relays K11 through K17 associated with the bypassing functions of the Reactor Operations Switch S4.
These normally de-energized relays (note that switch S4 is normally in the "run" position) are energized by appropriate contacts on Switch S4.
Relay contacts then close to bypass the trip sensor contacts and thus normal input voltage to the Logic Unit is maintained even though the trip sensor operates open.
In this manner, the permissive circuits for control rod withdrawal can be kept intact for various operations which are necessary under refuel, resetting, testing or start-up conditions.
Annunciator Control Unit RE02C also contains relays K1 through K6.
These normally energized relays provide control room annunciation for neutron flux high scram (K1 through K3), short period scram (K4 and KS) and manual scram (h6).
rp0182-0856a140-42
11 Attachments 1 and 11 show these annunciator circuits.
Time Delay Relay K10, controlling the duration of the Operations Recorder high-speed chart travel, is also located in Units RE02C and RE02D. This relay is adjustable from 15-second to 30-minute operation and is presently set for a 5-minute operation.
3.1.4.5 Operations Recorder All protection system sensor circuits are con-tinuously monitored so that an operation or failure is recorded for later reference or for identifying spurious single-channel trips which would not be annunciated on the Station Annunci-ator. The monitoring is provided by two, Thirty-Channel Operations Recorders RE01A and RE01B, one for each protection channel (see ).
These recorders normally operate with a chart speed of 1-1/2" per hour. When the first trip signal is received, the chart speed is increased by a factor of 60 to 1-1/2" per minute.
Upon application of the trip signal, the Opera-tions Recorder pen relay, corresponding to the particular sensor, de-energizes, initiating a 1/10" pen travel offset. Subsequent sensor operations initiate offsets also and thus trip sensor action can be more accurately timed at.
the faster chart speed.
3.1.5 RPS Trip Funct' ion Description 3.1.5.1 Manual Scram Actuation One manual scram switch, S3, is located on the center section of the control console. Depres-sing this switch opens contacts to both protec-tion channels causing a control rod scram.
In addition, it de-energizes the two protection bus undervoltage relays (RE11A for Channel 1 and RE11B for Channel 2) which by themselves would cause a control rod scram.
When the reactor operation switch, S4, is in the " shutdown" position, Contacts 1 and 1C (Channel 1, contacts 10 and 10C for Channel 2),
l t
rp0182-0856a140-42 l
12 series with the Manual Scram Switch contacts are open, simulating a continuous manual scram.
3.1.5.2 RPS Bus Undervoltage Each protection bus has an undervoltage relay which opens contacts in the 115-volt, ac circuit to the scram pilot valve solenoids, the master scram solenoid valve, the scram dump tank solenoid vent, drain and equalizing vent (see Section 3.1.4.3c) valves and the turbine trip and sphere ventilation trip relays. These undervoltage relays (RE11A and REllB) operate at a RPS bus voltage of 52 1 20 volts and require manual reset upon tripping.
It should also be noted that the balance of the automatically actuated containment isolation valves will close upon a sustained loss of power to both protection channels by de-energization of the isolation valve control relays 1K4A,1K4B, 2K4A and 2K4B (see and Reference 1).
3.1.5.3 High Neutron Flux The occurrence of high neutron flux would indi-cate a reactor output in excess of the safe level for continuous operation. Protection against such a condition is provided by the Out-of-Core Reactor Neutron Monitoring System.
A contact (which is normally closed) in each power range picoammeter opens if its associated indicator shows 105% of power on any range position resulting in de-energizing auxiliary alarm relay 3K1 (see ).
De-energizing 3K1, in turn, p'rovides the " neutron flux high" alarm (see Annunciator 5,
! 1).
If two of the three picoacmeters l
indicate 120% ( 5%) on the "0-125%" scale or 38%
l
( 2%) on the 40% scale on any range, the trip i
output voltages drop to near zero volts and both j
protection channels operate to scram the recctor (see Reference 1 and Attachment 1).
i l
A downscale interlock is also provided on each picoammeter, which requires two picoammeter indicators to be at least 5% upscale on the 0-125% scale (or 2% upscale on the 0-40% scale) l for the withdrawal of control rods. This keeps the high-flux scram point within about one decade of the operating power. On start-up, this interlock can be bypassed when all three power range picoammeter range switches are set on the minimum operable range (see Reference 2).
rp0182-0856a140-42 l
13 3.1.5.4 Protection Against Picoammeter Circuit Failure As described in Reference 1, protection against picoammeter circuit failure is afforded by the downscale trip of one of the three trip contacts in each RPS channel and the concurrent upscale trip of one of the remaining two contacts. The scram setting for this condition is 5% for the low trip and 120% on the high trip. This feature is bypassed at power ranges of less than 40 x 10 % to prevent spurious reactor trips during start-up (see Attachment 3).
3.1.5.5 Short Reactor Period The occurrence of a short reactor period would indicate an excessive rate of rise of reactor power during start-up conditions. Protection against such a condition is provided by the Out-of-Core Reactor Neutron Monitoring System.
Intermediate Channels 4 and 5 (see Attachment 4) have a 15-second period alarm and if the period indication on either channel reaches 10 (i 2) seconds, the trip output drops to near zero volts and both protection channels operate to scram the reactor (see Reference 1).
Period scram signals are bypassed when the range switch of any two powet range picoammeters are in the 4% full-scale (or higher) power range positions (see the Auxiliary Switch S4 development table in and the period bypass circuitry which energizes relays IK6 and 2K6 in Attachment 1).
This bypass feature is used to avoid spurious scrams by fluctuating period measurements caused by steam voids when boiling occurs in the core.
3.1.5.6 High Reactor Pressure The occurence of high pressure in the reactor could indicate trouble in the nuclear steam supply system. As can be seen in Attachment 1 and Reference 1, four pressure detectors, two (PS/RE07A and PS/RE07C) connected to Protection Channel 1 and two (PS/RE07B and PS/RE07D) connected to Protection Channel 2, are set to trip at 50 5 psi above reactor operating pres-sure. Upon high reactor pressure, these open contact signals are fed to the logic units initiating a control rod scram.
If the reactor pressure should continue to rise above the scram trip setting, additional rp0182-0856a140-42
14 contacts on the pressure detectors close to initiate emergency condenser operation when the pressure has increased to 100 psi above reactor operating presure (see Reference 3 and the con-trols for the emergency condenser inlet and outlet valves as described in Attach-ments 10 and 12).
3.1.5.7 Low Reactor Water Level A low water level in the reactor could indicate a loss of water such that the reactor core is in danger of being uncovered. As can be seen in and in Reference 1, four level switches (LS/RE09A and LS/RE09C on Protection Channel 1 and LS/RE09B and LS/RE09D on Protec-tion Channel 2) detect this condition when the water level drops to a level of
- 2'9" above the reactor core (with a tolerance limit of -1").
If a trip signal is received, a control rod scram is initiated and the containment penetra-tions are automatically closed. This trip also energizes the controls of the core spray cooling system initiating core spray if the reactor pressure drops below 200 psi (see Reference 1).
3.1.5.8 Low Steam Drum Water Level A trip signal from this condition anticipates a loss of water which could lead to a low reactor water level. As shown in Attachment 1, four level switches (LS/RE06A and LS/RE20A connected to Protection Channel 1 and LS/RE06B and LS/RE20B connected to Protection Channel 2) detect this condition when the~ steam drum water level drops to a level above or equal to 8" (with a tolerance limit of -0.5") below the cen-ter line of the drum, resulting in a control rod scram (see Reference 1).
These trip signals are bypassed by contacts of Auxiliary Relays K11 and K12 (in the annunciator control units) which are energized by closure of Contacts 2, 2C (RPS Channel 1) and 9, 9C (RPS Channel 2) on the Reactor Operation Switch S4 when it is in the Shutdown, Refuel or Bypass Dump Tank position.
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15 3.1.5.9 Closure Of Main Steam Line Backup Isolation Valve Closure of Isolation Valve MO-7050 to 50 i 5% of full closure (either by operation of Remote-Manual Control RMC-5500, by malfunction, or by isolation signals from the RPS) initiates a con-trol rod scram by the opening of limit switch contacts (9,10,15 and 16) on the valve drive unit (see Attachments 1 and 9).
These trip signals are bypassed by contacts of Auxiliary Relays K11 and K13 in the annunciator control units which are energized by closure of Contacts 2, 2C (RPS Channel 1) and 9, 9C (RPS Channel 2) on the Reactor Operation Switch S4 when it is in the Shutdown, Refuel or Bypass Dump Tank position.
3.1.5.10 Simultaneous Closure of Recirculation Line Valves The closure of these valves prevents coolant circulation to the core and a scram is initiated to prevent the development of excessive fuel temperatures.
A trip signal is received at approximately 10%
of full simultaneous closure of both discharge valves (M0/N001 - A and B), or both suction valves (M0/N003 - A and B), or 60% of full sim-ultaneous closure of both butterfly valves (MO-N006 - A and B), or any combination of one of these valves in each loop. Reference 1 and Attachments 13 and 14 serve to describe the trip operation. It should be noted, however, that the butterfly valves are locked open (with their RPS limit switches closed) prior to startup. This is required in Master Checkoff Sheet 0-TGS-1, A-1 Rev 10.
The sensors are limit switch contacts on each of the drive units of the valves.
These trip signals are bypassed by contacts of Auxiliary Relays K12 and K13, located in the annunciator control units, which are energized by closure of Contacts 2, 2C (RPS Channel 1) and 9, 9C (RPS Channel 2) on the Reactor Operation Switch S4 when it is in t'he Shutdown, Refuel or Bypass Dump Tank position.
rp0182-0856a140-42
16 3.1.5.11 High Condenser Pressure High condenser pressure is used as an indication that the main condenser is no longer available as a heat sink for the reactor output, thereby making necessary a control rod scram.
Pressure Switches PS/654 and PS/655 connected to Protection Channel 1, and Pressure Switches PS/652 and PS/653 connected to Protection Chan-nel 2 detect this condition when condenser pres-sure reaches 8" (
0.5") Hg absolute (see Reference 1 and Attachment 1).
These trip signals are bypassed by contacts of Auxiliary Relays K14 and K16, located in the annunciator control units. These relays are energized by closure of Contacts 3, 3C (RPS Channel 1) and 8, 8C (RPS Channel 2) on the Reactor Operation Switch S4 when it is in the Refuel or Bypass Dump Tank position. The high condenser pressure reactor trip is automatically bypassed any time steam drum pressure is below a set point maximum of 500 psig (see Reference 13).
3.1.5.12 High Reactor Building Pressure A differential pressure between the inside and the outside of the reactor building (contain-ment) could indicate a major rupture within the containment. To monitor such a condition, four differential pressure sensing detectors, PS/664 and PS/665 connected to Protection Channel 1 and PS/666 and PS/667 connected to Protection Channel 2, are strategically located outside the reactor containment under the cable tray area.
When a differential pressure of 1.0 psi above atmospheric exists, a trip signal initiates a control rod scram and a closure of the contain-ment isolation valves (see Reference 1 and ).
As shown in scheme D01 (see Attachment 15),
independent pressure switches PS/636, PS/637, PS/7064A and PS/7064B automatically initiate containment spray by closing contacts in the opening circuit of emergency spray valve MO-7064. These contacts close to initiate spray at a containment pressure of 2.2 psig (see Reference 4).
Manual actuation of containment spray is also provided via backup valve M0/7068.
rp0182-0856a140-42
17 In order to prevent the possibility of external pressure on the containment sphere due to atmos-pheric changes, two additional, independent pressure switches are used to open the contain-ment sphere ventilation valves. As can be seen in Attachment 7, pressure indicating switches PIS 173 (scheme 8501) and PIS 187 (scheme 8512) feature coilacts in the control circuits of the sphere ventilation supply isolation valves and the exhaust isolation valves, respectively. As shown in the attachment, PIS 187 operates to open the exhaust valves (CV/4094 and CV/4095, see Attachment 8) by energizing solenoid valves SV/9153 and SV/9154 above (-) 0.85 psig on in-creasing vacuum. Above (-) 1.00 psig, PIS 173 operates to open the supply valves (CV/4096 and CV/4097) by energizing solenoid valves SV/9151 and SV/9152. As is shown, the vacuum relief function will override a standing isolation sig-nal to open the valves, thereby preventing damage to the containment sphere due to possible external pressures.
3.1.5.13 High Scram Dump Tank Level A high-water level in the scram dump tank would prevent high-speed insertion of the control rods. Four level switches, LS/RD08A and LS/RD08C connected to Protection Channel 1 and LS/RD08B and LS/RD08D connected to Protection Channel 2, detect this condition when a level of 5/16" i 1/2" below the tank center line is reached, initiating a control red scram (see Reference 1 and Attachment 1).
These trip signals are bypassed by contacts of Auxiliary Relays KIS and K17, located in the annunciator control units, which are energized by closure of Contacts 4, 4C (RPS Channel 1) and 7, 7C (RPS Channel 2) of Reactor Operation Switch S4 when it is in the Bypass Dump Tank l
position, provided all control rods are fully inserted. Full insertion of control rods ener-gizes Auxiliary Relay 4K30A (for coil of 4K30A, see Attachment 16), the contacts of which close thus enabling relays K15 and K17 to be ener-gized. Bypass of this trip function is neces-sary to enable emptying the dump tank after a scram in the event the tank level is above the scram trip setting.
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18 3.1.6 System Response Time According to the Plant's Technical Specifications (as referen-ced below) the following RPS respcase times have been designed into the system:
Test Procedure Section
Response
Technical Used To Describing Description Time Spec Reference
- Verify Test a.
Closure of Main 60 s 5
TR-32 4.1.3.2 Steam Isolation Valve b.
Maximum Scram 2.5 s for 7
TR-01 4.1.1 Insertion Time 90* insertion c.
RPS Response Time 100 ms 8
TR-33 4.1.1 d.
Closure of Sphere 16s 8
TR-52 4.1.1 Ventilation Valves
- See Attachment 21
3.2 DESCRIPTION
OF THE ENGINEERED SAFETY FEATURES SYSTEM As in the NRC's SEP Topic VI-10.A report for Palisades, this report addresses the Containment Spray System (CSS) as an example typical of the engineered safety features (ESF) systems.
As described in the Bases section of Reference 9 and in the Plant's Final Hazards Safety Report (Item 3.7.3), the CSS is provided to reduce pressure and temperature as well as the leakage of airborne radioactivity following a loss-of-coolant accident (LOCA). Operation of only one of the two redundant l
containment spray (CS) sets is sufficient to provide the re-quired CS flow for containment temperature and pressure sup-i pression following a LOCA. The primary CS set is automati-cally initiated. The secondary CS set is manually initiated l
upon failure of the primary set.
To provide a more detailed description of the CSS, Attach-ment 17 is used. As shown, branch lines from the core spray system lead to the two sets of CSS nozzles in the upper part l
of containment. The primary CS line (rated at 146 gpm by the capacity of the nozzles, see Reference 10) is supplied by water from the normally pressurized fire system header through a 6" connecting line. The secondary (backup) CS line (rated
(
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t l
l I
o 19 at.233 gpm by the capacity of the nozzles, see Reference 10) is supplied from the pressurized fire system through a 4" con-necting line. The fire system is maintained pressurized by a fire jockey pump. Upon an increased demand for fire water (as in the case of core spray and CS during a LOCA), both the electric and the diesel (backup) fire pumps will automatically start when the fire header pressure decreases to the proper setpoint. Each f're pump is rated at 1000 gpm (see Refer-ence 12).
Upon an increasing containment pressure of 2.2 psig (see References 4 and 11 and Attachments 18 and 19), pressure switches PS/636, 637, 7064A and 7064B will operate to auto-matically open primary spray valve M0/7064 to transfer water from the pressurized fire system to the spray nozzles. As shown in Attachment 15, a tormally-open contact from each of the aforementioned pressure switches is wired into the M0/764 open circuitry. These contacts close whenever the containment pressure increases to +.he pressure switch setpoint. The wiring logic is such that at least one contact from each pair of pressure switches (ie, PS/636 and PS/637 or PS/7064A and PS/7064B) is required to close to energize the valve's open starter coil. Also shown in Attachment 15 is that the second-arj (backup) spray valve, M0/7068, can only be manually operated.
Should the CSS operate for a significant length of time, the accumulated water in the bottom of the containment may in-crease to a level of 587 feet. As described in a letter from RA Vincent to DM Crutchfield (USNRC) " Big Rock Point Plant-SEP Tcpic VI-7.B, ESF Switchover From Injection to Recirculation Mode", dated October 21, 1981, the operator is required by procedure to manually initiate the switchover from the injec-tion to the recirculation mode of the Post Incident Spray System. This has the effect of discontinuing the accumulation of water in the bottom of the containment since fire water system isolation valves VFP-29 and 30 are closed. At this point in time, the backup spray set may be operated intermit-tently to either cool the core or to conserve sphere capacity to accumulate additional water (refer to the Final Hazards Safety Report, Item 3.7.6).
It should be noted that a loss of offsite power will not prohibit the CSS from performing its intended function. A diesel-driven fire pump, which automatically starts upon decaying fire system pressure serves to backup the electri-cally-driven fire pump. The diesel-driven fire pump can sup-ply either the primary or secondary spray sets. However, should the electric fire pump be required it can be supplied by the emergency diesel generator.
In addition, the primary spray valve M0/7064 is powered by the 125 VDC station battery.
rp0182-0856a140-42 l
e a
20 4
M0/7064 is the only electrically operated valve in the primary spray piping path between the fire pump and the spray nozzles.
There are no contain. ment spray system timing requirements specified in the Plant's Technical Specifications.
4.
EVALUATION AND CONCLUSIONS 4.1 REACTOR PROTECTION SYSTEM As has been described in Section 3.1 of this report, the RPS has been designed to permit periodic testing of its functional capability dur-ing conditions of either reactor shutdown or power operation. The RPS has also been designed to allow channels (and specific inputs) to be tested idependently to determine failures and losses of redundancy that may have occurred. As will be described in the paragraphs that follow, established periodic testing serves to verify RPS capability from the sensor input to the output of the control rod drives. Dur-ing the conduct of these tests, the protective function of the RPS is not prohibited.
In the following paragraphs, the RPS testing program is described.
Throughout the descriptions, the High Reactor Building Pressure (high containment pressure) input to the RPS is considered as typical of all of the other RPS inputs. Therefore, the High Reactor Building Pressure (HRBP) input and subsequent RPS output actions are used for illustrative purposes whenever specific detail is necessary. Select-ing the HRBP input as typical was based on the fact that this is one of the two RPS inputs (the other being Low Reactor Water Level) that when tripped result in complete containment isolation as well as reactor scram.
4.1.1 ' Testing During Refueling During every refueling outage, each sensor that provides an input to the RPS is calibration tested. During these tests, the process variable which inputs to the sensor (eg, pressure for a pressure switch or level for a level' switch, etc) is varied and the sensor's output is monitored and adjusted if necessary. The procedures that require this testing are listed below.
IRPS-1, Rev 4 " Calibration and Testing of Reactor Pressure Sensors PS/RE07 A, B, C and D" IRPS-2, Rev 6 " Calibration and Testing of Reactor Water Level Sensors" (LS/RE09A through H)
IRPS-3, Rev 4 " Calibration and Testing of the Reactor Enclo-sure High-Pressure Scram Sensors" (PS/664, PS/665, PS/666 and PS/667)
I rp0182-0856a140-42
21 IRPS-4, Rev 5 " Calibration and Testing of the Reactor Steam Drum Water Level Sensors" (LS/RE06A, LS/RE06B, LS/RE20A and LS/RE20B)
IRPS-5, Rev 5 " Calibration and Testing of High Condensor Pressure Scram Sensors" (PS/652, PS/653, PS/654 and PS/655)
IRPS-6, Rev 4 " Calibration and Testing of High Condenser Pressure Scram Bypass Pressure Switches" (PS/RE15A, PS/RE15B, PS/RE15C and PS/RE15D)
IRPS-7, Rev 3 "leactor Protection System Undervoltage Breaker Check" (REllA and RE11B)
IRPS-8, Rev 3 " Calibration and Testing of the Reactor Recircu-lation Pump Valve Position Scram Switches" (It should be noted that this procedure tests the limit switches of recirculation pump valves M0/N001A, M0/N001B, M0/N003A and M0/N003B only.
The limit switches in valves M0/N006A and M0/N006B are not tested since these valves are required to be locked open prior to start-up per Master Checkoff Sheet 0-TGS-1 A-1, Rev 10.
This results in maintaining closed the valv'.e limit switches that input to the RPS.)
IRPS-11, Rev 3 " Cleaning and Inspection of Scram Dump Tank High Level Scram Sensors (Magnetrol LS RD08A through E (5))"
TR-32, Rev 11 " Reactor Protection System Scram Sensors Test."
This procedure tests the following RPS sensor inputs: closure of the main steam isolation valve M0/7050, ope, ration of the power and intermediate range Nuclear Measurement Systems (NMS)
Channels 1 through 5 and the manual scram pushbutton.
In the case of the NMS channel tests, however, the actual process variable (flux) is not varied to verify correct channel opera-tion.
(In the case of all the other sensor inputs to the RPS the process parameter is varied.) Rather, a calibration sig-nal is applied using a test potentiometer to the amplifier input to verify trip setpoints. The actual flux signal, like the calibrated test signa 1, also inputs to the amplifier.
Certain operating procedures, however, provide verification of the ability of the NME channels to process actual flux inputs.
System Operating Procedure SOP 31, Rev 2 " Nuclear Instrumenta-tion System" requires that Channels 4 and 5 provide logarith-mic neutron flux level and period information from approxi-
-5 mately 10 to rated power with an 84-bundle core. S0P 31 also requires that Channels 1, 2 and 3 provide linear neutron
~
flux level information from approximately 40 x 10 % to 125%
rated power with an 84-bundle core. These requirements serve to verify that the NMS channels are correctly processing actual flux input. Furthermore, General Operating Procedure GOP 5, Rev 7 " Power Operation" indicates that the power range
22 picoammeter readings should be compared to the calculated thermal power level and adjusted to 97* or greater of sne calculated value.
TR-32 also requires that a reactor scram be verified when one power range channel is placed in an upscale trip condition and a second power range channel is placed in a downscale trip condition. TR-32 requires that the above verification be made for each combination of power range channels. This require-ment verifies the operability of the RPS trip input described in Section 3.1.5.4.
As an example of the' type of tests performed in the above procedures, IRPS-3 requires that the technician isolate and pressurize each HRBP scram switch separately. During the pressurization, the RPS setpoint of 1.5 psi i 0.2 psi is verified.
(It should be noted that Reference 1 states that a new setpoint of i 1.0 psi above atmospheric will be incorpor-ated at a convenient outage but not later than the refueling outage scheduled for the first quarter of 1982.)
As previously described, Procedure TR-32, Rev 11 " Reactor Pro-tection System Scram Sensors Test" is also performed during refueling. This test overlaps the aforementioned calibration tests by either varying the sensor's input process parameter or by mechanically tripping the sensor (note, sensor response to varying the process input variable is verified in the cali-bration tests previously described) to effect a RPS channel scram. All sensors are tested and each is tested separately.
For example, TR-32 requires that HRBP scram sensor PS/664 be repressurized to approximately 2.0 psi (to ensure that the RPS trip setpoint is surpassed). At this pressure, the Channel 1 scram and an of fset of Channel 1 Operational Recorder Pen 1 is verified at the RPS (see Attachment 1).
While the pressure to PS/664 is maintained, switch PS/666 is pressurized to approxi-mately 2.0 psi.
At this point, the following verifications are made:
a.
Received Channel 2 scram (at the RPS).
b.
Received full isolation scram (note, both channels of RPS are required to operate before reactor scram or full isolation is effected).
c.
Received Chsnnel 2. Operational Recorder Pen 1 offset.
d.
Received annunciation (at the station annunciator, see 1).
HRBP switches PS/665 (Channel 1) and PS/667 (Channel 2) are then tested in a similar fashion.
rp0182-0856a140-42
23 TR-33, Rev 5 " Reactor Protection System Response Time" is also performed at every refueling. This test requires that the reactor protection system response be timed and be verified to be less than 100 milliseconds (see Section 3.1.6).
The actual interval measured is from the time a sensor trip contact test circuit breaker is opened until the scram solenoids or RPS output relays are de-energized. More specifically, the re-sponse time is measured by using a high-speed chart recorder.
After the recorder is placed on high-speed, the HRBP test circuit breaker for the channel under test is opened and the recorder pen deflects. The recorder then monitors and records the time until the isolation valve control relays (eg, IK4A and IK4B for Channel 1) de-energize. At this point, a second recorder pen deflects. The above test is then repeated until each group of RPS output relays or solenoids are timed to de-energize (ie, isolation output, K1 through K2; Group 1 solenoids, K3 through K4; Group 2 solenoids, K4 through K5; ventilation trip, K3 through K5) as a result of opening of the same HRBP test breaker. This sequential series of tests are then repeated for the other HRBP test breaker. RPS Channel 2 is then tested and timed in a similar fashion.
Other refueling test procedures are conducted to verify the RPS outputs. During the conduct of these tests, the RPS actu-ated components are verified to operate correctly whenever simulated sensor actuations are initiated utilizing the test circuit brealers. These tests overlap with TR-32 and are described below:
TR-01, Rev 8 " Control Rod Drive Performance Testing." This test verifies the proper insertion of each control rod (as controlled by its appropriate solenoids) when a simulated sensor trip is initiated. This test also times the insertion from the initiation of the simulated trip to 90% full rod insertion. The insertion time of i 2.5 seconds is verified (see Section 3.1.6).
It should be noted that Reference 6 states that the scram insertion time for the initial 10% of rod stroke shall be 0.6 seconds maximum. Although TR-01 does not specifically require this verification, it should be noted that the first 10% of rod insertion is recorded as part of the 90% verification. Should the rod's insertion rate be greater than 0.6 seconds per 10% insertion, the 90% verification acceptance criteria could not be met.
TR-37, Rev 6 " Control Rod Drive Scram Dump Tank Vent Delay."
i This test requires that the scram dump tank equalizing vent valves (CV/NC14 and CV/NCIS) be verified to open (as control-led by their associated solenoid valves) upon initiating a scram using the manual scram push button.
In addition, the time from the initiation of the manual scram to the initiation of valve opening is recorded. This test serves to check the reliability of the time delay relay (TD) in the RPS output rp0182-0856a140-42 l
24 circuit (see Attachment 1).
This timing measurement is not required by the Technical Specifications.
TR-52, Rev 11 " Sphere Isolation Test." This test requires that all containment sphere isolation valves (including the sphere ventilation valves) are verified to close upon a simu-lated HRBP scram as initiated by opening the appropriate Chan-nel 1 and Channel 2 test circuit breakers in the RPS. This test also requires that all of the isolation valves close upon inserting an isolation signal by placing the penetration closure Switch SS in the " Isolate" position. In addition, this procedure requires that each isolation valve be timed from the initiation of a closure signal to the closure of the valve. The procedure requires that the sphere ventilation isolation valves close in less than 6 seconds (see Sec-tion 3.1.6).
TR-80, Rev 1 " Master Scram Valve Operability Test."
This test requires that the correct operation of the master scram sole-noid valve and the associated scram dump tank vent and drain valves be verified. The correct operation of the valves is verified upon simulating sensor trip by opening the master solenoid valve test circuit breaker. The time from the initi-ation of the simulated trip until the vent and drain valves close is recorded. This test is completed on each RPS channel independently.
4.1.2 Testing During Operation During power operation, the operability of the RPS is verified utilizing two test procedures:
T7-04, Rev 7 " Weekly Reactor Protection Logic System Test" and T30-01, Rev 5 " Monthly Reac-tor Protection System Test at Power."
Procedure T7-04 requires that all of the RPS outputs occur whenever a simulated HRBP sensor trip is initiated utilizing one test circuit breaker. The output verification is made by observing that all seven of the amber lights located in power switches A and B (RE04 and RE17) de-energize. After reset, the test is repeated utilizing the other test circuit breaker in the same RPS channel. Upon ensuring that Channel I has been reset, the entire procedure is then repeated in similar fashion in Channel 2.
Test procedure T30-01, Rev 5 " Monthly Reactor Protection Sys-tem at Power" is also conducted at power. This test requires that each RPS channel be independently tested by verifying proper RPS output for each input sensor trip as simulated by opening the sensor's test breaker. During each sensor test, the procedure requires that the proper amber lights on the A and B power switches (RE04 and RE17) de-energize when the sensor's test circuit breaker is opened. For example, during rp0182-0856a140-42
25 the Ch:nnel 1 test of HRBP, it is verified that all seven of the amber lights in power switches A and B de-energize after either HRBP test circuit breaker is opened. Also verified is that the proper Operational Recorder pen deflects for all tests except the Nuclear Measurements Sensor tests.
4.1.3 Test Overlap for Response and Timing Verifications In an attempt to describe the degree of overlap designed into the testing program for both response and timing verifica-tions, the HRBP function will again be regarded as typical.
4.1.3.1 Overlap in Response Verifications As described in Section 4.1.1, procedure IRPS-3 provides a calibration of the HRBP pressure switches.
During the conduct of the procedure, a test pressure input is applied to one of the pressure switches.
The switch setpoint is verified utilizing a voltmeter to detect approximately 26 VDC when the switch con-tacts open. As found and as left setpoint data is recorded.
IRPS-3 overlaps with procedure TR-32.
Proce-dure TR-32 requires that a Channel 1 HRBP switch be repressurized above the trip setpoint until the Channel 1 scram is alarmed at the RPS. Maintaining this condition, a Channel 2 pressure switch is pres-surized and held until the Channel 2 scram is alarmel and a full isolation scram is received at the RPS.
TR-32 also overlaps with procedure T30-01.
Procedure T30-01 requires that the HRBP inputs to the RPS be simulated by opening the appropriate test circuit breaker. As described in Section 4.1.2, T30-01 re-quires that all of the RPS power switch lights for a particular channel be verified as de-energizing when a HRBP sensor's test breaker is opened.
In turn, procedure T30-01 overlaps with procedures TR-01, TR-37, TR-52 and TR-80.
In each of these four procedures, the full response of the component: which are operated by the RPS is verified whenever a RPS trip is initiated. As previously stated, T30-01 re-quires that each RPS sensor's test circuit breaker (with the exception of reactor short period which is performed in Procedure TR-32) be opened to effect a channel scram as indicated by the RPS power switches.
Overlap with TR-01 (for example) exists since TR-01 requires that each control rod be fully inserted into the core whenever a HRBP sensor's test circuit breaker in one channel is opened concurrent with an rp0182-0856a140-42
26 open condition (in the other channel) of-the control rod's test circuit oreaker (of the drive to be tested).
Another example of overlap would,be TR-52.
TR-52 requires that all of the toolacisn valves be verified to close'whenever a HR.?? sensor's test breaker is opened in each RPS channel.
In an attempt to verify tha: all of the RPS actuated component, respond properly to the RPS initiation signal, it was observed t'.at no procedure exists which requires the verification that a turbine trip is effected (ie, turbine trip relay HTS energizes to close the emergency stop valve) whenever the RPS is tripped. Although IEEE 338-1975 is not strictly adhered te in this case, it should be noted that ver-ifications are presently conducted per Proce-dure TR-52 which require that the closure of the sphere venc valves occurs as a result'of a HRBP simu-lated trip. The relay coils that de-energize to close the sphere vent valves are connected directly across the relay that initiates turbine trip.
In addition, a procedure was not identified which verifies that the scram dump tank vent and drain valves (CV/SCll and CV/NCl2) close as function de-energizing solenoid valves S0/NC22C and SO/NC22D.
In this instance, however, the redundant method of RPS initiated valve closure is tested. TR-80 requires that the valves' closure be verified as a result of solenoids S0/NC22A and SO/NC22B de-energizing when the Channel 1 and Channel 2 master scram valve test breakers are opened.
4.1.3.2 Overlap in Timing Verifications Although RPS sensor response to varying process para-meter inputs which exceed the sensor's setpoint is not specifically timed, the RPS's response to the sensor's trip output is timed.
In addition, the RPS actuated components are also timed. As previously described in Section 4.1.1, procedure TR-33 requires that the RPS response be recorded from the time that a HRBP trip is simulated until each group of RPS out-put scram solenoids or relays are de-energized.
TR-33 overlaps with three of the four refueling pro-cedures described in Section 4.1.1 that are conducted to verify the RPS outputs. These four procedures are TR-01, TR-37, TR-52 and TR-30.
As an example of the overlap, procedure TR-01 requires that each rod be timed to 90% full insertion from a simulated HRBP rp0182-0856al40-42
4 27 trip initiated by opening that sensor's test circuit breaker. (Refer to Section 4.1.1 for a description of the TR-01 test).
Timing overlap between procedures TR-33 and TR-52; however, does not exist in the strict definition.
TR-35 requires that the RPS response be timed from the moment a simulated HRBP trip is initiated until the RPS output isolation relays de-energize. TR-52 does not require that the RPS output isolation relays be de-energized to time the response of the isolation valves. Rather, TR-52 requires that the isolation valve closure be timed from closing of the valve's manual hand switch in the control room. Although strict overlap in the timing verification is not re-quired, it should be noted that strict overlap is t
required for the response verification in procedures TR-01 and T30-1 as described in Section 4.1.3.1.
Also, the timing requirements of TR-33 and TR-52 provide a timing verification of total RPS and RPS actuated component response for all practical pur-poses. As stated, TR-33 requires that the time be measured to de-energize the RPS output isolation relay coils. These coils have their contacts wired in series with the hand switch contact that initiates isolation valve closure for timing verification in procedure TR-52.
As mentioned in the first paragraph of this section, the RPS sensor response to inputs which exceed the sensor's setpoint is not timed. Although the test program does not conform to Regulatory Guide 1.118, it is the opinion of the Consumers Power Company that this verification will not provide useful informa-tion. Attachment 20 describes a study performed by
~
the Consumers Power Company with the assistance of the Institute of Nuclear Power Operations. The study shows that electronic equipment failures are not normally detected by response time testing. The Con-sumers Power Company considers the sensor calibration and functional testing requirements to be adequate in and of themsc1ves.
Although specific time response testing of the RPS sensor's response to an input in excess of its set-point is not required by the Technical Specifica-tions, nor by any of the aforementioned procedures, it should be noted that procedure TR-32 does require that timing tests be conducted for the valves that input to the RPS. TR-32 requires that the time be measured from the initiation of closure of the main steam isolation valve (M0/7050) to RPS scram as well rp0182-0856a140-42
28 as to the valve's full closure. Similarly, TR-32 requires that the time be measured from the initi-ation of closure of the recirculation line valves to s
4.1.4 Reactor Protection System Testing Program Conclusions The RPS design and surveillance testing program at the Big Rock Point Plant conforms to, or meets the intent of, the cur-rent licensing criteria as described in Section 2.0 of this report. One exception, however, is that of the turbine trip initiation from a RPS scram. Presently, the turbine trip from a RPS scram is not verified.
IEEE 338-1975 states that "the entire channel, train or load group will be verified by test-ing of individual components or subsystems."
4.2 ENGINEERED SAFETY FEAWRES (CONTAINMENT SPRAY SYSTEM) 4.2.1 Description of the Containment Spray System (CSS) Test Program Every refueling, the containment spray (CS) initiation pressure sensors PS/636. PS/637 PS/7064A and PS/7064B (see 7) are calibrated. To perform this calibration, I
procedure IPIS-2, Rev 8 " Calibration and Testing of the Reactor Enclosure High Pressure Sensors Used for Enclosure Spray Actuation" requires that the technician isolate and I
pressurize each switch individually. The procedure then
{
requires that the setpoint of i 2.2 psig be verified by moni-toring sensor contact closure with a continuity tester.
IPIS-2 also requires that the trip system logic (ie, one switch out of two taken twice) be verified.
To verify the logic, the technician is required to actuate one pressure switch per pair (ie, pair PS/636 and PS/637 and pair PS/7064A and PS/7064B) by raising the input pressure above the setpoint of i 2.2 psig. Continuity across the series-parallel string of pressure switch contacts in the opening circuits of valve M0/7064 (see Scheme D01, Attachment 15) is verified using the continuity tester. The test is then repeated using the other two pressure switches.
During each refueling outage, Procedure T180-15, Rev 18 " Core Spray and Enclosure Spray Valve Initiation and Operability Test" requires that the proper operation of the primary CSS valve (M0/7064) and the backup CSS valve (M0/7068) be verified. T180-15 also requires that timing of the valves (both opening and closing times) be documented per ASME Boiler and Pressure Vessel Code Section XI, IWV Testing Requirements.
T180-15 overlaps with Procedure IPIS-2 in that T180-15 re-quires that one pressure switch in each pressure switch pair fle, pair PS/636 and PS/'37 and pair PS/7064A and PS/7064B) be rp0182-0856a140-42
29 repressurized to approximately 2.5 psig (to ensure that the switch setpoint of i 2.2 psig is surpassed) and held. T180-15 then requires that Valve M0/7064 be verified to open. The pressure switches are then depressurized and the valve is closed. The other two pressure switches are then pressurized above their setpoints and the valve is verified to reopen.
T180-15 also requires that the manually actuated Spray Valve M0/7068 be verified to open when a manually initiated signal from the control room hand switch is applied. Finally, T180-15 requires that each valve's (M0/7064 and M0/7068) open-ing time be measured and recorded when the valve is opened using the control room hand switch.
Although Regulatory Guide 1.118 states that "the response time test shall include as much of each safety system, from sensor input to actuated equipment, as possible.
. ", the Con-sumers Power Company considers the aforementioned valve timing test alone adequate. As described in Section 4.1.3.2 and in 0, it is the opinion of the Consumers Power Com-pany that time response tests of the sensor's response to a high-pressure input will not provide useful information.
Procedure TR-50, Rev 4 " Calibration and Testing of Core Spray Flow, Enclosure Spray Flow and Fire System Strainer Differ-ential Pressure Instrumentation" is performed on a refueling basis. This procedure requires that the primary CS set in-strumentation (FT/2164, FI/2333 and FS/2520, all located immediately downstream of valve M0/7064) be calibrated and tested. The tests require that the flow alarm of
- 100 gpm (as annunciated in the control room) be verified. The proce-dure also requires that the secondary (back up) CS set instru-mentation (FT/2161, FI/2334 and FS/2521, all located immedi-ately downstream of valve M0/7068) be similarly calibrated and tested. The secondary CS set flow alarm (as annunciated in the control room at i 100 gpm) is also verified.
In addition, TR-50 requires that the differential pressure indicating switches (PDIS/7814, 7815 and 7816) that monitor the pressure drop across the three strainers in the fire water system supply lines to the containment sprays and the core spray heat exchanger be calibrated. During this calibraion, the "High dF" alarm ( 2 O psid) and the " Strainer Plugged" alarm (i 5.0 psid) is verified for each switch. These alarms are annunciated in the control room.
Regarding the ability of the fire water system to serve as a reliable supply of water for the containment spray system, refueling Procedure TR-70 " Fire Suppression System Functional Test and Pump Capacity Test" is performed every refueling out-age.
During this test, each fire pump (both the electric pump and the diesel pump) is verified to automatically start up on decaying fire header pressure. During each pump's test, the rp0182-0856a140-42
.~ _-
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I 30 procedure requires that a test gauge be installed at the pump's discharge to monitor the pressure at which the pump automatically starts. The procedure requires that this pres-sure be recorded. The procedure also requires that once the pump has started a flow condition be established such that a pump flow rate of 1000 gpm is verified while the pump devel-
'i opes a discharge pressure of 110 psis. Each pump is tested independently.
In addition to Procedure TR-70, two other procedures are per-formed regarding the fire water supply to the post incident cooling system. Procedures T30-22, Rev 16 " Emergency Core Cooling Valve Tests" (performed monthly) and T180-15, Rev 18
" Core Spray and Enclosure Spray Valve Initiation and Oper-ability Test" require that the fire water supply valves to the core spray ring and the core spray heat exchanger (which is used during the recirculation phase of core and containment cooling) be stroked and timed. Procedure T30-22 requires that i
the fire water supply valve M0/7066 to the core spray heat exchanger be stroked open and closed and both the opening and closing times be recorded. Procedure T180-15 requires the same for the alternate fire water supply valve to the core spray ring M0/7072.
As described in Section 3.2, the post incident cooling system is manually switched from the injection to the recirculation i
mode whenever the accumulation of water at the bottom of the containment reaches an elevation of 587 feet. The accumu-lation of water may be caused, in part, by the actuation of the containment sprays. A number of surveillance procedures are performed to verify the capability of the postincident spray systqm to serve in the recirculation mode. These proce-dures are kriefly described below:
Procedure TR-05, Rev 12 " Core Spray Pump Run and Test Loop Operation" requires that the core spray pumps be lined up for recirculation flow testing through the core spray heat exchanger and core spray test tank. During the test, one of the pumps is started and adequate pump suction is verified.
The core spray heat exchanger is also inspected for tube leaks.
(Note, the shell side does not pass flow during this test). The procedure then requires that the other core spray pump be tested in a similar fashion. During every third refueling outage, the procedure requires that the pump be ISI performance tested per ASME Pressure and Vessel Code,Section XI.
Regarding the shell side of the core spray heat exchanger, Procedure TR-09, Rev 9 " Core Spray Heat Exchanger Shell Side Flow" requires that the fire water system be lined up through the heat exchanger shell to verify proper system flow require-ments. During the flow test, (with the electric fire pump rp0182-0856a140-42
31 passing flow through valve M0/7066) the discharge pressure of the electric fire pump as well as the pressure at the dis-charge of the shell is recorded. Upon closing valve M0/7066 and using a 2-1/2" fire hose installed between a fire system manifold and the core spray heat exchanger, tie-in at valve VPI-10, the test is repeated.
Procedure T30-14, Rev 9 " Monthly Core Spray Heat Exchanger Leak Test" is performed monthly to verify the integrity of the core spray heat exchanger tube bundles by filling the shell and measuring tube leakage.
I 4.2.2 Engineered Safety Features System (Containment Spray System)
Test Program Conclusions Since the CSS design requires that the fire water system be isolated from the core spray and containment spray systems prior to the test and the fire water system serves as the sup-ply to each of these systems during an accident, CSS surveil-lance testing is performed during shutdown only. However, since the containment spray system's protective response is relatively simple by design (ie, pressure sensor's detect the high containment pressure condition and automatically open one motor-operated Valve, M0/7064, to align a pre-pressurized fire water system to the primary, sprays) and a manually-operated redundant spray system also exists, it is the opinion of the Consumers Power Company that the nature and frequency of the shutdown surveillance test program is adequate to ensure that the CSS will operate as intended during power operation.
Furthermore, the Consumers Power Company considers that the surveillance test program serves to verify the operability of the system as a whole and under conditions as close to the design as practical, the full operational sequence that brings the system into operation.
It is concluded that the test pro-gram complies with, or meets the intent of, the current licensing criteria as given in Section 2.0.
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TONR 3-82 SEP Topic VI-10.A Page 1 of 6 RESPONSE TIME TESTING - LER SEARCH 0 A Licensee Event Report record search for those reports which identifed defective instrumentation via response time testing was completed by the Consumers Power Con.;any on September 25, 1981. The objective of this data search was to determine if modern day response time testing does indeed identify instrumentation defects that are not detected through normal accuracy, hysteresis and linearity checks. The search began by contacting the Event Analysis section of the Institute of Nuclear Power Operations (INPOJ. INPO performed a series of computer searches on the 22,000 plus LERs which have been accumulated. The search was a three step process. The first query was " component." This rendered a large subset of'the data base. INPO would then query by " method of discovery" using Rou-tine Test / Inspection or Special Test / Inspection. This subset would then receive a third query for the work " time" within the event description. This three step process was applied to Instruments, Valves, Diesel Generators, Con-trol Rods and Pumps. Final result was a computer listing of 632 events. The list of 632 events was reviewed by the Consumers Power Company using the following criteria: was the event discovered while performing a time respo se surveillance and would the event have been identifed without time being re-corded (ie, a complete failure to operate does not require a timing device for detection)? The results of this review, listed in Table 1, indicate that 165 time response failures occurred. The most common (47) event involved main steam isolation valves failing to meet the maximum or minimum closing time. The second (34) was diesel generators failing to start. Then came valves (33), control rods (16), emer-gency power sequencer (9) and timing devices (9). The events of interest are all included within the "others" category. They are one pressure switch, three resistance temperature devices and no electronic devices other than timers. One must conclude from this data search that electronic equipment failures are not normally detected by response time testing, RTD response time beir.g a possible exception. rp0182-0856c140-157
s TONR 3-82 SE P Trp ac VI-10. A Atttchment 20 Page 2 or 6 TABLE 1 - RESPOMSE TIME TESTING IER SEARCD MSiv & Em9 Timer Cont isol-Turbine D/G Pwr & Time & ESF Cont Sea rch t ELNo yaives_ Slars les pe_ lay _ Vaives Rod OJW No 282-77000 x Pump accelleration too slow PM-8 325-79010 x IN-BT 331-76000 x IN-BT 328s-19018 x Prtessure switch response time IN-BT 259-18000 x Turbino driver governor too slow IN-BT ' 4 333-79032 x IN-BT 333-8001:4 x IN-BT 271-73000 x IN-BT 328-78051 x Turbine driver governor IN-BT 4 368-81017 x RID response IN-BT 333-80069 x IN-BT 331-77000 x IN-BT 388-78075 x D/G rallure or field flashing IN-BT 4 336-77000 x RID response IN-Bi 245-15000 x IN-BT 336-75000 x IN-BT 277-76000 LS contact bounce IN-BT 289-76000 x PCP Bkr operating speed FSAR in error IN-BT V-3 263-7t000 x 4 V-3 333-80012 x 254-78000 x V-3 4 V-3 289-76000 x 4 V-3 219-73000 x V-3 237-8000's x 155-78000 x V-3 4 298-16000 x V-3 331-18000 x V-3 4 V-3 321-19026-1 x 293-18s000 x V-3 259-78000 x V-3 4 366-19050 x V-3 V-3 258 -19020 x 4 2:a1-81004 x PORY V-3 301-78000 x V-3 4 kWB 35-81 rp3182-Ot473a 157
9 TONR 3-82 SEP Troic VI-13.A Atttchment 20 e Page 3 of 6 l TA8LE 1 - RESPONSE TIME TESTING LER SEARCH MSIV & Eng Timer Cont isol Turbine D/G Pwr & Time & ESF Cont Sea rch _L{ R_pto yelves Elart leg Delev Valves Rod Other No 2a45-78010 x V-3 265-80010 x V-3 298-80013 x V-3 259-79012 x Turbine driver supply stop valve V-3 321-79062 x V-3 2849-81011 x V-3 219-80030 x V-3 4 313-80025 x Purge Exh ran failed duration test V-3 331-76000 x MSiv control sys heater V-3 316-18052 x V-3 298-80021 x V-3 289-T2000 x V-3 4 265-72000 x V-3 155-78083 x V-3 4 271-79029 x V-3 271-78032 x V-3 265-80022 x V-3 315-75000 x V-3 258-7908a2 x V-3 4 338-80081 x V-3 4 219-71000 x V-3 263-F1000 x V-3 308 -780184 x V-3 4 220-17000 x V-3 261-78030 x V-3 28 9-8008a6 x V-3 4 293-80093 x V-3 298-800t9 x V-3 4 321-801284 x CR Vent isolation Valve V-3 237-75000 x V-3 315-75000 x V-3 331-75000 x V-3 2ta5-70000 x V-3 ^ 331-76000 x V-3 kWB 35-81 d-l rp0182-08473a 157
9 TONR 3-82 SEP Topic VI-13.A Atttchment 20 e Page 4 of 6 TABLE 1 - R U PONSE TIME TESilNG l_ER SEARCH MSlv & Eng Timer Cont Isol Turbine D/G Pwr & Time & ESF Cont Sea rch _[IR No yaive5 llaIt Sea DeIav Vaives Rod Other No 218-16000 x v-3 32t-16000 x V-3 4 V-3 335-16000 x 263-73000 x PORV V-5 259-73000 x PORV V-5 220-80013 x V-5 V-5 263-80026-1 x 321-80003 x VP3 266-71014 x VP3 VP3 278-76000 x 302-18001-1 x VP3 258e-79006 x VP3 VP3 269-74000 x ~333-19016 x VP3 VP3 321-16000 x 316-1802F x VP3 VP3 293-79023 x 331-80021 x Pump ralled to reach rated speed VP3 384-16000 x VP3 4 311-18030 x VP3 31 T-81020 x VP3 263-78014 x VP3 266-75000 x VP3 VP3 293-75000 x 218-15000 x VP3 VP3 277-77000 x 331-80047 x VP3 VP3 293-77000 x VP3 336-77000 x VP3 278-79038 x 237-75051 x VP3 298-80036 x VP3 VP3 245-71000 x VP3 324-75000 x KWS 35-81 rp0182-0473a157
t TONR 3-82 SEP Tcpic VI-10 A 0 s Page 5 of 6 TABLE 1 - RESPONSE TIME TESTING LER SEARCH MSIV k Ing Timer cont Isol Turbine D/G Pwr & Time k ESF Cont Sea rch JRfio VaIves Syra }eg DeIav VaIves Rod Q1her No 386-E0093 x VP3 4 321-71000 x VP3 331-76000 x VP3 321-16000 x VP3 324-76000 x VP3 293-75000 x VP5 2514-79041 x VP5 331-77083 x VP5 155-75000 x CONROO-BT 2f9-15000 x CONROO-BT 4 220-69000 x CRDRVE-CT 321-78:000 x CRDRVE-BT 155-74000 x CRDRVE-BT 111-12000 x CRDRVE-BT 249-75000 x CRDRVE-BT x CRORVE-BT 237-15000 23T-70000 x ChDRVE-BT 237-70000 x CRDRVE-BT 336-190084 x PCS RID IN-CT 320-76000 x x Cont bldg fans IN-CT 388-79001 x DG-41 4 213-80001 x DC-41 321-78000 x Battery rechar le rate DC-41 4 321-77086 x DC-41 155-77000 x DC-41 155-76000 x DG-41 155-19008 x DC-41 3f6-79011 x DG-41 4 348-79016 x DC-41 155-7600f4 x DG-41 285-78000 x DC-41 4 155-74000 x DG-41 312-80024 x DG-41 255-74000 x DG-41 kWB 35-81 rp0182-0473a157-153
- ~ -. t TOMR 3-82 SEP Tcpic VI-10.A Attcchment 20 >p Page 6 of 6 TA8tf 1 - RESPONSE TIME TESTING t.ER SEARCH MSly & Eng Timer Cont isol Turbine D/G Pw r & Time & ESF Cont Sea rch _j,[ R_8lo Valves }ta_r_t jeg Delay Valves Rod Othe r No DC-41 3f8-78020 x i DC-41 261-79009 x DC-41 3 t:3-800 386 x DC-41 155-76000 x DC-41 155-76000 x DG-41 155-76000 x DG-41 155-01005 x DC-41 2",5-78017 x DG-41 219-78011 x DG-41 155-76000 x DG-41 155-16000 x DC-41 1 155-81007 x DG-41 409-76000 x DG-41 211-18035 x DC-41 348-80051 DC-41 348-80052 x DC-41 366-78060 x DG-41 318-71000 x DG-41 3fs8-E0062 x x DG-41 320-18068 x DG-41 155-7F000 x DG-41 155-F1000 x DC-41 321-77000 x DC-41 250-77000 x DG-41 217-18050-1 x DG-41 318-77000 x DG-41 293-75000 x DC-41 155-77000 x DG-41 155-17000 x TOTAL 47 34 9 9 33 16 19 kW8 35-81 rp0182-0473a157
c r TONR 3-82 SEP Topic VI-10.A 1 Page 1 of 1 BIG ROCK POINT TECHNICAL SPECIFICATIONS References Reference No Reference Identification 1 Table 6.1.2 2 Item 6.2.1(b) 3 Item 6.1.4(d) 4 Table 11.4.3.4 5 Item 3.4.3(c) 6 Item 5.1.3 7 Item 5.2.2(a) 8 Item 6.1.1 9 Item 11.3.3.4 10 Item 3.5.1(b) 11 Item 3.5.1(e) 12 Item 4.2.6 13 Item 6.1.3(e) t I rp0182-0856c140-157 l - _ -.}}