ML20247H899
ML20247H899 | |
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Site: | Fermi |
Issue date: | 08/31/1989 |
From: | Bennett D, Fujitani J, Mccann B GENERAL ELECTRIC CO. |
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Text
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k CONTROL SYSTEMS COMMON SENSOR LINE FAI' LURE ANALYSIS EVALUATION REPORT 1
FOR ENRICO FERMI ATOMIC POWER PLANT UNIT 2 t'
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CONTROL SYSTEMS COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION REPORT
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FOR ENRICO FERMI ATOMIC POWER PLANT UNIT 2 AUGUST 1989
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CONTROL SYSTEMS COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION REPORT
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FOR U
ENRICO FERMI ATOMIC POWER PLANT UNIT 2
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AUGUST 1989 I-PREPARED FOR DETROIT EDISON COMPANY PREPARED BY
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B. R. McCANN REVIEWED BY J.Y.
FUJITANI
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GENERAL ELECTRIC COMPANY NUCLEAR ENERGY SAN JOSE, CALIFORNIA 95125
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l APPROVED BY:
b b.M' D. E.
Bennett, Technical Leader - Regulatory and Design Compli
- d_ Application Engineering - Nuclear Services Department cp
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A[j.Koslow, Manager - Regulatory and Design Compliance Application En in ering, Nuclear Services Department 4.
ru UJ gineering Analysis Services - Nuclear Services Depgte
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D.k.Sden, Licensing and Consulting Services - Nuclear Services Department EDE-25-0889
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CONTROL SYSTEMS COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION REPORT FOR ENRICO FERMI ATOMIC POWER PLANT - UNIT 2 The information contained herein, supplements the existing Fermi-2 Updated Safety Analysis Report (UFSAR) Chapter 15 transient analyses and documents an evaluation of the Enrico Fermi Atomic Power Plant - Unit 2 control systems interaction due to a common sensor line failure.
1.0 PURPOSE The general purpose of the Common Sensor Line Failure Analysis was to review the failure events of non-safety grade Enrico Fermi Atomic Power Plant - Unit 2 (Fermi-2) control systems which utilize common sensor lines or sensor signals. The specific purpose of the analysis and this report was to supplement the existing Ferm',-2 UFSAR Chapter 15 accident analyses and satisfy the facility operating license Item 2C, Condition 8 pertaining to NRC concerns of failure of a common sensor line and sensors.
2.0 CONCLUSION
Transient category events have been postulated as a result of this study; however, the net effects have been determined to be less severe than and bounded by the events in chapter 15.
The conclusion of this evaluation is that the limits of minimum critical power ratio (MCPR), peak vessel and main steamline pressures, and peak fuel cladding temperature for the expected operational occurrence category of the identified events would not be exceeded as a result of a common sensor line failure.
Although not a safety concern the analysis identifies a scenario where iiPCI and RCIC both start and the high level trip of HPCI and RCIC are inhibited (see 3.8).
3.0 ANALYSIS METHODOLOGY In conjunction with the Fermi-2 Control Systems Common Power Source Failure Analysis (CPA) portion of the overall Fermi-2 Control Systems Failure Analysis Program, a comprehensive approach was developed and implemented to address the general purpose of the analyses as well as specific NRC concerns.
The activity list, Table 1.1, and following descriptions highlight the methodology used to perform the analysis.
.)
I
It should be noted that this study used the event-consequence logic of the Fermi-2 'IFSAR Chapter 15 analysis, but it started the logic chain from a specific source (e.g., a single common sensor line failure) rather than a system condition (e.g., feedwater runout).
By approaching the study in this manner, a great deal of confidence can be placed in the study conclusions. The soundness of the total plant design is demonstrated by its being tolerant of these effects.
3.1 System Identification The scope of control systems to be analyzed was established by first compiling a complete list of the Fermi-2 systems and subsystems.
I Next, the list was reviewed to confine the analysis to only those non-safety grade control systems with the potential to affect reactor pressure vessel (RPV) pressure, water level, or power level changes.
3.1.1 All the Fermi-2 plant instrumentation and control systems were identified, listed, and agreed upon as complete by the two principal analyzing engineer groups, i.e., Gereral Electric Company (GE) and Detroit Edison Company (DECO).
3.1.2 System and component elimination criteria (see Table 1.2) were derived and agreed upon by the principals to delete non-electrical, non-operational, or non-control systems or components (included in some previously analyzed systems and components already addressed in UFSAR Chapter 15) from the systems identified in 3.1.1 above (see Table 1.3).
If there was any uncertainty as to whather or not a system met the criteria, it was retained for further analysis.
Those systems that met the criteria for elimination were so noted in the complete system list, leaving the remaining control systems to be analyzed.
3.2 Common Sensor Line or Sensor Identification The Common Sensor Failure Analysis portion of the Control Systems Failure Analyses then identified strategic reactor process sensor lines or sensors commonly shared by two or more plant systems, at least one system of which was a non-safety grade control system identified in Section 3.1.2 above.
3.3 Failure Type Determination Based on conservative assumptions, a complete and instantaneous sensor line break or plug during normal, full power reactor opera-tion was determined to be the bounding failure type for each sensor line analyzed. -.
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3.4 Definitions Common Instrument Line: A line providing a process pressure signal to two or more instrument sensors (pressure to electrical current instrument, P/I, transmitters) which serve two or more instrument systems, of which at least one system is a non-safety grade control
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system, e.g., Feedwater Control System (C32).
Common Instrument Sensor: An instrument sensor which provides inputs to two or more instrument systems, of which at least one system is a non-safety grade control system.
Line Failure Types:
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Broken: An instantaneous instrument line break (guillotine break) that vents to an ambient pressure (near atmospheric g
pressure) environment.
Plugged: An instantaneous instrument line plug (complete
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blockage, pinch) maintaining as-failed line pressure at the instrument sensor and essentially inhibtting any monitoring change, especially actual process, vessel, or line parameter changes.
Note:
In the case of differential pressure sensing
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instruments monitoring reactor pressure vessel water level, a plugged reference or variable line could result in a more complex response. The conservative response, however, is still an inhibited response, i.e., no actions would occur.
Primary Effects: The direct, instantaneous effects, if any, on the
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specifically identified sensor or component resulting from the failure. Usually, the sensed input signal to the sensor component goes to a minimum or maximum value or, in the case of a sensor line plug, remains relatively constant at an inaccurate (as-failed) value, insensitive to any actual process changes.
Secondary Effects: The indirect effect, instantaneous or delayed,
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if any, on the specifically identified sensor or prominent subsequent instrument loop components, i.e., indicators, trip units (trips, permissives, initiators), controls (controllers, valves),
or devices (relays or lights).
RPV Liquid Level, Pressure or Power Level Effects: Any actual or
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probable reactor pressure vessel liquid level, pressure or power level change directly or indirectly attributable to the identified failure and component actions or i'nactions.
Combined Effects: The systematic evaluation of the identified line sensors primary and secondary effects and the resulting
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actions (s), if any, which would most likely result as a direct accumulation of each, or all, sensor failures and RPV pressure, liquid, and power level change effects on plant performance.
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3.5 Line-Group and Component Tabulation and Failure Analysis The sensor instrumentation directly connected to and receiving an input signal from the identified sensing line or sensor were individually identified, grouped, listed, and analyzed to determine what, if any, action would result from the occurrence of each line/ sensor failure type described in Section 3.3 The primary, secondary, and RPV parameter change effects, if any, were then identified, analyzed, and tabulated.
Note:
Because signals from these common lines and sensors were frequently utilized by components previously deleted as part of step 3.1.2 above, for completeness, these and questionable, non-safety grade control components were retained in the analysis groups.
3.6 Combined Failure Effect Analysis The components and their failure effects identified in Section 3,5 were evaluated and reviewed for cumulative effects by the principal parties to identify the prime component and combined component failure event scenarios listed and discussed in the Common Sensor Line Failure Analysis and Evaluation Summary, Table 1.4, and more comprehensively in the Common Sensor Line Failure Analysis load sheets, Appendix A.
3.7 Comparison of Analysis Result to UFSAR Chapter 15 The consequences of the postulated failures and their associated process disturbances were compared to the consequence of the event analyses described in Fermi-2 UFSAR Chapter 15.
Where the Chapter 15 event description contained consequences for the postulated failure, the Chapter 15 ovent was considered to bound the postulated failure.
3.8 Analysis Exceptions The Line 5 and 6 (see Table 1.4) failure scenarios identified the start of HPCI and RCIC on a sensing line break.
Chapter 15 (15.5.1) analyzes inadvertent HPCI start but not the start of both HPCI and RCIC pumps.
However, since RCIC flow is a small fraction of HPCI flow (about 10 per cent) the net result will be approximately the same. With respect to cold water injection this event would be bounded by 15.1.1, Loss of Feedwater Heating.
Loss of high level trip of HPCI and RCIC is not a chapter 15 concern.
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TABLE 1.1 MAJOR' COMMON SENSOR LINE FAILURE ANALYSIS ACTIVITIES AND RESPONSIBILITIES
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ACTIVITY' LEAD SUPPORT A
- 1. SYSTEM IDENTIFICATION GE DECO
--2.
ELIMINATION CRITERIA GENERATION GE DECO y
- 3. ELIMINATION. CRITERIA APPLICATION GE-DEC0 t
4' COMMON SENSOR LINE/ SENSOR IDENTIFICATION GE DECO
- 5. FAILURE MODE APPLICATION TO INDIVIDUAL DEVICES DEC0 GE g
~6. TABULATION OF INDIVIDUAL FAILURE EFFECTS DEC0 GE
- 7. COMBINED FAILURE EFFECT ANALYSIS AND TABULATION GE
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- 10. MAJOR EVENT RESOLUTION AND CHAPTER-15 GE MODIFICATION RECOMMENDATIONS (IF REQUIRED)
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- 11. COMPILE DRAFT REPORT GE
- 12. REVIEW DRAFT REPORT AND COMMENT DECO
- 13. RESOLVE COMMENTS GE DECO
-14. ISSUE FINAL REPORT GE 15.= FOLLOW-UP; ANSWER QUESTIONS, ESTABLISH DRF GE DECO
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c TABLE 1.2 FERMI-2 CONTROL SYSTEM FAILURE ANALYSES CRITERIA.FOR ELIMINATION OF SYSTEMS AND COMPONENTS OF SYSTEMS FROM THE ANALYSES CODE ELIMINATION CRITERION
- N1 Non-electrical systems or components, i.e., sole mechanical or software systems or components. Examples: the reactor system, vessels, steam turbines. Note: Any associated electrical control
. components might be relevant and are to be reviewed.
Examples:
vessel liquid level, pressure and temperature controls, and turbine speed controls.
. N2 -
Non-control type electrical systems or components, i.e., systems or components having no direct or indirect controlling or controlled function, including permissive input and output signals (strictly passive systems and components).
Example: the nuclear boiler process instrun.entation sensors, transmitters lights, meters or recorders, which only provide information, measurement indications, and records. Note: For the control system failure ana~ lyses, such infonaation, although possible of interest or importance to reactor operation and operating personnel's manual control actions, is not' considered relevant to initiating or prohibiting any automatic electrical control actions.
N3 Non-operational type electrical control systems or components, i.e., systems or components not normally used or required to be usable during normal reactor power operation.
Examples: the refueling interlock control system, the startup range portion of the neutron monitoring system, the turbine generator turning gear controls.
N4 Operational electrical control systems or components which have no direct or indirect interaction with normal reactor operating control systems or components.
Examples: building heating, and air conditioning control systems, and lighting controls.
N5 Operational electrical control systems or components which do directly or indirectly interact with reactor operating control systems or components but which can in no way affect changes in the reactor vessel liquid, pressure, or power levels.
Examples:
.the radwaste control system, sump pump level controls.
- In some cases, more than one criterion may apply. _ _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _._ _ _ ______-_ _ - __ __ _ _ - _ - __ _ -
r:
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TABLE I.2 (continued)
FERMI-2
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l CONTROL SYSTEM FAILURE ANALYSES
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CRITERIA FOR ELIMINATION OF SYSTEMS AND COMPONENTS OF SYSTEMS FROM THE ANALYSES
[QQE ELIMINATION CRITERION *
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N6 Operational safety-related electrical control systems or components or portions of systems or components which perform direct plant safety control functions.
Examples: the reactor protection system, the mafn steam line radiation monitoring portion of the process radiation monitoring system, or the steam leak detection temperature elements and controls of the leak
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detection system.
Note: Any related response of these safety systems or components to conditions or actions brought about by non-safety related control system or component actions, resulting directly or indirectly from a non-safety control system failure, are to be.
)
identified and analyzed.
Example: a reactor vessel low water-level RPS trip and a subsequent reactor scram resulting from a loss of feedwater flow which was, in turn, directly or indirectly caused by non-safety power source or sensor failure, e.g., a feedwater pump motor power failure.
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N7 Electrical power systems or components involved in distribution, transformation, or interruption of electrical power.
- However, controls for these systems / components might need to be considered if the loss of such control power could lead to the failure of other systems and components. Example: the 125 Vdc control power for a condensate pump circuit breaker.
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2
- In some cases, more than one criterion may apply.
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7
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L TABLE 1.3 CONTROL SYSTEM FAILURE ANALYSES 1-IDENTIFICATION AND ELIMINATION OF SYSTEMS FOR THE COMMON POWER SOURCE AND COMMON SENSOR OR SENSOR LINE FAILURE ANALYSES i.
SYSTEM ID SYSTEM DESCRIPTION ELIMINATION CODE
- At A71 PRIMARY CONTAINMENT ISOLATION N6 B21 NUCLEAR BOILER PROCESS INSTRUMENTATION N3,N4 B21 JET PUMP N2
'B21 ADS N6 B21-06 MSIV LCS N6 B21-07 LOOSE PARTS MONITORING N2 B31 RECIRC NONE
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Cll RPIS N2 C11-00 REACTOR MANUAL CONTROL NONE C11-08 R0D WORTH MINIMIZER N3 C11-09 R0D SEQUENCE CONTROL N3 C11-50 CRD HYDRAULICS NONE C32 FEEPWATER CONTROL NONE C35-REMOTE SHUTDOWN N3 C36 DED!CATED SHUTDOWN N3 C41 STANDBY LIQUID CONTROL N6 C51 STARTUP RANGE NEUTRON MONITORING N3 C51 INTERMEDIATE RANGE NEUTRON MONITORING N3 CSI POWER RANGE NEUTRON MONITORING (RPS)
N6
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C51 R0D BLOCK MONITOR N6 C51 RECIRC FLOW BIAS N6 C51 TIP N2 C71 REACTOR PROTECTION SYSTEM N6 C71 RPS MG SET N6 C91 PROCESS COMPUTER N2,N5
)
C94 ERIS N2,N5 D11 PROCESS RADIATION MONITORING N2 D11 MAIN STEAM LINE RADIATION MONITORS N6 021 AREA RADIATION MONITORING N2 D23 ENVIRONMENTAL MONITORING N2
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- See Table 1.2 for code criterion explanation.
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8-
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F TABLE 1.3 (continued)
SYSTEM ID SYSTEM DESCRIPTION ELIMINATION CODE
- L D30 SEISMT.C MONITORING N2 D40 METEOROLOGICAL N2 E10
' LEAK DETECTION N6 Ell RHR / RESERVOIR & SERVICE WATER N6-E21 CORE SPRAY N6
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E41 HPCI N6 E51 RCIC N6 FIS REFUELING N3 Gil RADWASTE N5 G33 REACTOR WATER CLEANUP N5 G41 FUEL POOL COOLING AND CLEANUP N5
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G61 TORUS WATER MANAGEMENT N5 H30 ANNUNCIATOR N2,N5 H40 COMMUNICATIONS N2 H50 BEARING & MOTOR WINDING TEMPERATURE N2 MONITORING JXX FUEL N1
).
N11 MAIN STEAM 52" MANIFOLD NONE N11 THIRD MSIVs N6 N20 CONDENSATE NONE N20-02 POLISHING FILTER DEMINERALIZERS NONE N21 FEEDWATER NONE N21 STANDBY FEEDWATER N3 i
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N21-02 RFPT CONTROL NONE N22 FEEDWATER HEATER DRAINS NONE N30-11 MAIN TURBINE, VALVES, PIPING N1 N30-12 TURBINE SUPERVISORY N2 N30-12A TURBINE GOVERNOR (PRESSURE CONTROL)
NONE N30-12C TURBINE PROTECTION NONE
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N30-13 GLAND SEALING NONE N30-14 LUBE OIL NONE N30-16 EXTRACTION STEAM N3 N30-17 BYPASS PIPING N1 N30-1B REHEAT & MOISTURE EXTRACTION NONE N30-19 TURNING GEAR _
N3
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N30-20 LP HOOD SPRAY N3 N30-21 UNITIZED ACTUATORS NONE N30-22 FLANGE HEATING N3 N30-32 GENERATOR HYDROGEN SEAL OIL NONE N30-33 STATOR WATER COOLING NONE
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l oSee Table 1.2 for code criterion explanation.
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9_
l
)*
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TABLE 1.3 (continued)
SYSTEM 1
ID SYSTEM DESCRIPTION ELIMINATION CODE
- N30-34 MAIN GENERATOR EXCITOR NONE N30-35 GENERATOR HYDROGEN COOLING SYSTEM NONE N30-36/37 GENERATOR GAS SUPPLY PURGE N5 N30-39 TURBINE DRIPS & DRAINS N5 N61 RAIN CONDENSER & AUXILIARIES NONE-
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N62 0FFGAS NONE
'N71 CIRCULATING WATER NONE P11 CONDENSATE STORAGE & TRANSFER N5 4-P12 FAKE UP DIMINERLIZER N5 P21 F0 TABLE WATER-N4 P33 SAMPLING N2
)
P34 POST ACCIDENT SAMPLING N3 P41 GENERAL SERVICE WATER CHLORINATION N5 P41 GENERAL SERVICE WATER NONE P42 REACTOR BUILDING CLOSED COOLING WATER NONE P43 TURBINE BUILDING CLOSED COOLING WATER NONE P44 EECW N6
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P45 EESW N6
-P50 CONTROL AIR (SAFETY)
NS PES BREATHING AIR N1 F50-0)
STATION AIR NC1E P50-0E CONTROL AIR (BOP)
NGNE P61 AUXILIARY BOILER N4
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P70 WASTE OIL N4 P80 FIRE PROTECTION N5 P82 FIRE DETECTION N2 P90 CHEMICAL SYSTEMS N5 RXX ELECTRICAL DISTRIBUTION N7 S11 MAIN TRANSFORMER N7
)
S11-00 MAIH GENERATOR TRANSFORMER NONE S12 GENERATOR ISOLATED PHASE BUS NONE S13 GENERATOR METERING & SYNCHRONIZING NONE S14 TELEMETERING N2 S20 345 KV SWITCHYARD N7 T23-01/02 DRYWELL/ TORUS NI
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T23-03 VACUUM BREAKERS N3 T31 DVERHEAD CRANES N3 T41 REACTOR BUILDING VENTILATION N4
)
- See Table 1.2 for code criterion explanation.
)
_ 10
i f
TABLE 1.3 (continued)
SYSTEM ID SYSTEM DESCRIPTION ELIMINATION CODE *
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T41-02 CONTROL ROOM HVAC' N6 T45 REACTOR BUILDING SUMPS FLOOR AND N5 EQUIPMENT DRAINS T46 STANDBY GAS TREATMENT N6 T47 DRYWELL COOLING N6
)
T48-01 DRYWELL PRESSURE CONTROL N3 T48-02 NITROGEN INERTING N3 T48-03 PURGE PIPING N3 T48-04 HYDROGEN RECOMBINER N6 T49 DRYWELL PNEUMATICS N6 T50 PRIMARY CONTAINMENT ATMOSPHERE N6
)
MONITORING U41 TURBINE BUILDING VENTILATION N4 U45 RADWASTE BUILDING SUMPS FLOOR &
N5 EQUIPMENT DRAINS U45 TURBINE BUILDING SUMPS FLOOR AND N5 EQUIPMENT DRAINS
)
V41 RADWASTE BUILDING VENTILATION N4 W23 COOLING WATER CHLORINATION N5 W24/25 COOLING WATER COOLING TCWERS/ RESERVOIR N1 W41 COOLING WATER PUMPHOUSE VENTILATION N4 X41-01 CSB VENTILATION N4 X41-03 RHR CONPLEX VEN11LAT10N N6
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X41-04 GSW PUMPHOUSE VENTILATION N4
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1 I
- See Table 1.2 for code criterion explanation.
) -
4 F
TABLE I.4-COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
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LINE FAILURE B0UNDING
- FA TYPE EVENT DESCRIPTION FSAR SEC I
Broken High reactor pressure vessel (RPV) level indication 15.2.3 (not actual) trips the main turbine and reactor 15.2.7
)
feedpump turbines (RFPTs).
Reactor scrams on
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turbine stop valve closure and RPV level decreases due to loss of feedwater.
Lo-level (L-2) channel A & C signals to RCIC and HPCI are disabled but
. channel ~B & D (Div II) will initiate RCIC and HPCI on actual lo-level (L-2) signals.
)'
will each receive one channel hi-level (L-8) trip signal (half trip) but will not trip until actual L-8 trip signal from Div II, channel B for RCIC and channel D for HPCI.
Plugged.
If feedwater control is aligned to
'A', the level 15.2.3
).
input signal will be fixed. Feedwater flow will 15.2.7 increase or decrease as steam flow increases or decreases and RPV level will drift up or down.
If RPV level reaches hi-level (L-8), the main turbine and RTPTs will trip on channel B & D (Div II) level signals. Reactor will scram on turbine stop valve y
clusure.
If RPV level drops to lo-level (L-2),
RCIC and HPCI will start on actual lo-level signals from channel B & D (Div II). RCIC and HPCI will not trip at RPV hi-level (L-8) due to loss of both channel A & C level signals.
)
2 Broken High reactor pressure vessel (RPV) level indicatien 15.2.3 (not actual) trips the main turbine and reacter 15.2.7 feedpump turbines (RFPTs).
Reactor scrams on turbine stop valve closure and RPV level decreases due to loss of feedwater.
Lo-level (L-2) channel B & D signals to RCIC and HPCI are disabled but
).
channel A & C (Div I) will initiate RCIC and HPCI on actual lo-level (L-2) signals.
RCIC and HPCI will each receive one channel hi-level (L-8) trip signal (half trip) but will not trip until actual L-8 trip signal from Div I, channel A for RCIC and channel C for HPCI.
1 _
ij '
- y. -
h TABLE 1.4 (continued)
' COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
i LINE FAILURE.
. BOUNDING l
N_Q,__
TYPE EVENT DESCRIPTION FSAR SEC
]
2
. Plugged If feedwater contrcl is aligned to;'B', the level.
15.2.3 input signal will be fixed. Feedwater flow will 15.2.7 P
increase or decrease as steam flow increases or decreases and RPV. level will. drift up or down.
If RPV level reaches.hi-level (L-8), the main turbine W
and RFPTs will trip on channel A & C'(Div I) level signals. Reactor will scram on turbine stop valve closure.
If RPV level drops to lo-level (L-2),
)-
RCIC and HPCI will start on actual lo-level signals from channel A & C.(Div I), RCIC and HPCI will not trip at RPV hi-level (L-8) due to loss of both-channel B & D level signals.
3 Broken Low RPV. level indication (not actual) trips 15.2.7
)
. RPS Al & B1 logic channels causing a lo-level (L-3) reactor scram.
If feedwater control is aligned to
'B', the level input signal (not actual) will decrease causing feedwater flow to increase.
If RPV level reaches h,i-level (L-8), the main turbine and RFPTs will trip on hi-level signals
)
-from channel B & 0 (Div II). On loss of feedwater RCIC and HPCI will maintain RPV water-level.
Plugged If feedwater control is aligned to
'B', the levM 1L2,1 input signal (not actual) will be indeterminate.
RPV level could increase or decrease.
If the RPV
)'
level increased to hi-level (L-8), the main turbine and RFPTs will trip on hi-level signals from channel B & D (Div II).
If the RPV level decreasas to lo-level (L-3), reactor will scram on signals from channel A2 and B2.
If the RPV level decreases to lo-level (L-2), RCIC and HPCI will start and
)
maintain water level.
5 )
).
So>
TABLE 1.4 (continued)
COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
)-
LINE FAILURE BOUNDING L
TYPE EVENT DESCRIPTION FSAR SEC 4
Broken Low RPV level indication (not actual) trips 15.2.7
).-
RPS A2 & B2 logic channels causing a lo-level (L-3) reactor scram.
If feedwater control is aligned to
'B', the level input signal' (not actual) will decrease causing feedwater flow to increase.
If RPV level reaches hi-level (L-8), the main -
turbine and RFPTs will trip on hi-level signals
)
from channel A & C (Div I).
On loss of feedwater RCIC aqd HPCI will maintain RPV water level.
Plugged If feedwater control is aligned to
'B', the level 15.2.7 input signal (not actual) will be indeterminate.
RPV level could increase or decrease.
If the RPV
)
level increased to hi-level (L-8), the main turbine and RFPIs will trip on hi-level signals from channel A & C (Div I).
If the RPV level decreases to lo-level (L-3), reactor will scram on signals from channel Al and Bl.
If the RPV level decreases to lo-level (1.-2), RCIC and HPCI will start and 3
maintain water level.
i
)
)
{
)
)
i I
) _ _ _ _ _ _ __ -_ ____ _- __ ___ __ _
TABLE 1.4 (continued)
COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
LINE FAILURE B0UNDING NO.
TYPE EVENT DESCRIPTION FSAR SEC 5
Broken Low RPV level indication (not actual) trips 15.2.4 NS/4 (A71) Al & B1 logic channels causing closure 15.1.2 of main steam isolation valves (MSIVs).
Reactor will scram on MSIV closure. RCIC and HPCI will start on lo-level (not actual) L-2 trips in A & C logic channels.
RCIC and HPCI will not trip on hi-level (L-8) due to loss of level indication (actual) from logic channels A & C.
Plugged None 6
Broken Low RPV level indication (not actual) trips Primary 15.2.4 Containment Isolation (PCI) A2 & B2 logic channels 15.1.2 causing closure of main steam isolation valves (MSIVs).
Reactor will scram on MSIV closure.
RCIC and HPCI will start on lo-level (not actual) L-2 trips in B & D logic channels.
RCIC and HPCI will not trip on hi-level (L-8) due to loss of level indication (actual) from logic channels B & D.
Plugged None 7
Broken None Plugged None 8
Broken None N'ne Plugged o
9 Broken None Plugged None 10 Broken None Plugged None 11 Broken None Plugged None c
)
TABLE 1.4 (continued)
COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
)
LINE FAILURE BOUNDING N0, TYPE EVENT DESCRIPTION FSAR SEC 12 Broken Total steam flow signal to feedwater 3-element 15.2.7 control system is reduced by one fourth. _ RPV level
)
decreases to a new equilibrium point (above lo-level L-3).
.gf Plugged
_Feedwater control system response is reduced.
13 Broken MSIVs close on high flow signal. Reactor scram on 15.2.4
)
MSIV closure. RPV level increases to a new 15.1.2 equilibrium point (below hi-level L-8).
Plugged Feedwater control system response is reduced.
14 Broken Total steam flow signal to feedwater 3-element 15.2.7
)
control system is reduced by one fourth.
RPV level decreases to a new equilibrium point (above lo-level L-3).
Plugged Feedwater control system response is reduced.
)
15 Broken 14SIVs close on high flow signal.
Reettor scram on 15.2.4 MSIV closure. RPV level increases to a new 15,1.2 equilibrium poir.t (below hi-level L-8).
Plugged Feedwater control system response is reduced.
)
36 Broken Total steam flow signal to feedwater 3-el2 ment 15.2.7 control system is reduced by one fourth. RPV level decreases to a new equilibrium peint (above lo-level L-3).
Plugged Feedwater control system response is reduced.
)'
17 Broken MSIVs close on high flow signal.
Reactor scram on 15.2.4 MSIV closure.
RPV level increases to a new 15.1.2 equilibrium point (below hi-level L-8).
Plugged Feedwater control system response is reduced.
)
) '
1 TABLE 1.4 (continued)
COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
)-
LINE FAILURE BOUNDING NO.
TYPE EVENT DESCRIPTION FSAR SEC
- 18; Broken Total steam flow signal to feedwater 3-element 15.2.7
)
control system is reduced by one fourth.
RPV level decreases to a new equilibrium point (above lo-level L-3).
4-
~
Plugged Feedwater control system response is reduced.
)
19 Broken MSIVs close on high flow signal. Reactor scram on 15.2.4 MSIV closure.
RPV level increases to a new 15.1.2 equilibrium point (below hi-level L-8).
Plugged Feedwater control system response is reduced.
)
20 Broken Feedwater Flow signal to feedwater 3-element control 15.1.2 system is reduced by one half.
Feedwater turbines increase speed to compensate for apparent low flow.
RPV level increases to hi-level (L-8) where main turbine and RFPTs are tripped. Reactor scrams on turbine stop valve closure. RPV level decreases
)
due to loss of feedwater to lo-level (L-2), where RCIC and HPCI start and maintain RPV water level.
Plugged Feedwater control s3 stem response is reduced.
21 Broken Feedwater flow signal to feedwater 3-element control 15.2.7
)-
system is increased.
Feedwater turbines decrease speed to compensate for apparent high ficw.
RFV level decreases to lo-level (L-3) where reactor scrams. RPV level decreases to lo-level (L-2) where RCIC and HPCI start and maintain RPV water level.
)
Plugged Feedwater control system response is reduced.
1 l _ _ _ _ _ _ _ _ - _ _ _ _ - _.
,.4 Y
s 1
)
TABLE 1.4 (continued)
COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
)
' LINE FAILURE B0UNDING NO.
TYPE EVENT DESCRIPTION FSAR SEC 22 Broken Feedwater flow signal to feedwater 3-element control 15.1.2 system is reduced by.one half. Feedwater turbines
)
increase speed to compensate for apparent low flow.
RPV level increases to hi-level (L-8) where main turbine and RFPTs are tripped.
Reactor scrams on
+
turbine stop valve closure.
RPV level decreases
~
due to loss of feedwater to lo-level (L-2), where RCIC and HPCI start and maintain RPV water level.
)
Plugged Feedwater control system response is reduced.
~23 Broken Feedwater flow signal to feedwater 3-element control 15.2.7 system is increased. Feedwater turbines de >rease speed to compensate for apparent high flow.
)
RPV level decreases to lo-level (L-3) where reactor scrams.
RPV level decreases to lo-level (L-2) where RCIC and HPCI start and maintain RPV water level.
Plugged Feedwater control system response is reduced.
)
24 Broken Pressure signal to pressure regulator drops and 15.2.1 pressure regulater
'l' fails low. Reactor power increases.
If pressure regulator '1' was in control, pressure regulator '2' takes over.
Plugged Pressure signal to pressure regulator is indeterminant and control r,ystem response is
)
reduced.
25 Broken Pretsure signal to pressure regulator drops and 15.2.1 pressure regulator '2' fails low. Reactor power increr.ses.
If pressure regulator '2' was in control, pressure regulator
'l' takes over.
)
Plugged Pressure signal to pressure regulator is indeterminant and control system response is reduced.
)
) _ _ _ _. _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ - _
M i
1 e
)-
TABLE 1.4 (continued)
COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION
SUMMARY
)
LINE- ' FAILURE BOUNDING NQi_
TYPE EVENT DESCRIPTION FSAR SEC 26 Broken.
High back pressure signal (not actual) trips 15.2.4
>PCI (A71) logic channels Al & B1 causing
)
isolation of the MSIVs. Reactor scrams on closure of MSIVs.
4:
Plugged None 27 Broken High back pressure signal (not actual) trips 15.2.4
)
PCI (A71) logic channels A2 & B2 causing isolation of the MSIVs. Reactor scrams on closure of MSIVs.
Plugged None
)
)
)
)
)
)- - _ _ _ - _ _ _ - - - _ _ _ _ _
9 CONTROL SYSTEMS COMMON SENSOR LINE FAILURE ANALYSIS EVALUATION REPORT FOR ENRICO FERMI ATOMIC POWER PLANT UNIT 2 SUPPLEMENT ADDITIONAL SINGLE FAILURE IN MITIGATING SAFETY SYSTEM Subsequent to completing the Common Sensor Failure Analysis evaluations, each postulated failure was reviewed to determine if a conservatively selected sensor line failure which, in combination with an additional single component failure in a mitigating safety system, could result in a failure event not previously identified in Fermi-2 UFSAR Chapter 15.
Based on the results given in Appendix A of the report, it was postulated that a break in Line 1 in combination with the additional single component failure described below was the " worst case" identified failure.
The Line 1 failure event, described in Table 1.4, results in the RPV level sensors B21-N091A & C failing upscale causing a high level trip of the feedwater turbine and the main turbine, which would result in a reactor scram and loss of feedwater.
Also, as a result of the Line 1 failure event, Division I low level signals to initiate HPCI, RCIC, LPCI, Core Spray and ADS would be inhibited.
If there was an additional single failure of the RPV level sensor B21-N091D, the Division II level 2 and level 1 inputs to HPCI, LPCI, Core Spray and ADS would also be inhibited.
This additional failure would also cause a level 8 trip of HPCI which would inhibit manual start of HPCI.
However, RCIC would be available for manual initiation to maintain adequate RPV water level.
The additional single failure analysis results were evaluated and compared with the UFSAR Chapter 15 analyses.
No single completely boundinc failure was identified.
The above scenaric was satisfactorily addressed in DECO 3etters to NRC, EF2-65,624, dated September 23, 1983 and EF2-67,230, dated April 23, 1984, which were in response to the Michelson concern questions.
These responses were reviewed and accepted by the NRC as documented in the Fermi-2 Safety Evaluation Report Supplement 4 (SSER4) Section 7.3.2.
{
Note: The NRC has recently issued the Resolution of Generic Issue 101
" Boiling Water Reactor Water Level Redundancy" which addresses an instrument line break coupled with an additional independent failure in a protection system.
The concern was resolved for all BWR designs including Fermi-2 design.
,0
)
)
CONTROL SYSTEMS COMMON SENSOR LINE FAILURE ANALYSIS LOAD SHEETS PAGES 1 THRU 54 APPENDIX A
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