TXX-4703, Forwards Response to Section IV of Generic Ltr 85-12, Automatic Trip of Reactor Coolant Pumps. Determination of Pump Trip Criteria,Pump Potential Problems & Operator Training & Procedures Discussed
| ML20153E049 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/17/1986 |
| From: | Counsil W TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | Noonan V NRC - COMANCHE PEAK PROJECT (TECHNICAL REVIEW TEAM), Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM GL-85-12, TXX-4703, NUDOCS 8602240352 | |
| Download: ML20153E049 (6) | |
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Log # TXX-4703 File # 10035 TEXAS UTILITIES GENERATING COMPAhT SETW AY TOWRs e ese NOSTtt OtJVE WTBEET.1.B. S 8
- DALLAS. TEX A4 789e B February 17, 1986
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Director of Nuclear Reactor Regulation Attention:
Mr. Vince S. Noonan, Director Comanche Peak Project Division of Licensing U. S. Nuclear Regulatory Commission Washingtor., D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 RESPONSE TO NRC GEhERIC LETTER 85-12, IMPLEMENTATION OF TMI ACTION PLAN ITEM II.K.3.5,
" AUTOMATIC TRIP 0F REACTOR COOLANT PUMPS" Ref:
NRR letter dated June 28, 1985.
Dear Mr. Noonan:
The attachment to this letter provides the CPSES response to Section IV of NRC Generic Letter 85-12, " Automatic Trip of Reactor Coolant Pumps (RCP)".
Specific items addressed are the determination of RCP trip criteria, reactor coolant pump potential problems, and operator training and proce-dures.
Very truly yours,
/'
lNh W. G. Counsil RWH/ arm Attachment c - (Original + 40 copies)
A. L. Vietti-Cook
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ATTACHMENT TO TXX-4703 February 17, 1986 IMPLEMENTATION OF NRC GENERIC LETTER 85-12 SECTION IV A.1 Reactor coolant pressure and steam generator pressure instrumentation are used to determine the Reactor Coolant Pump (RCP) trip setpoint at CPSES.
Redundancy of measurement and indication is accomplished for both reactor coolant pressure and steam generator pressure.
A.2 Instrumentation uncertainties for both normal and adverse containment conditions were addressed in the RCP trip setpoint calculation.
An evaluation of local conditions has shown that fluid jets from two postulated piping breaks would impinge upon the common tubing for one RCP instrument and a locally mounted test gauge (which is not used for plant operation). However, the resulting loads have been evaluated and are acceptable.
Review of the instrument locations and charac-teristics indicate that the concerns of Section IV.A.1 and IV.A.2 have been considered and addressed.
A.3 The plant specific RCP trip setpoint calculation includes con-sideration of computer code uncertainties associated with the WOG supplied analyses values.
B.1 A Phase B (Hi-3) Containment Isolation signal would have to be generated before the Component Cooling Water (CCW) cooling of the RCP seals, via the RCP Thermal Barrier, is terminated. RCP seal cooling is still maintained using the normal Chemical and Volume Control System (CVCS) seal injection pathway. With the RCP seal cooling mechanism available, the RCP seal performance is not expected to significantly degrade during the Steam Line Break (SLB) event sce-nario so as to prevent the safe shutdown of the affected unit.
Continued operation of the RCPs during a SLB is not required, and the operator has sufficient control capabilities to ensure that a RCP trip can be initiated in a timely manner.
The primary RCP damage mechanism arising from the isolation of CCW is discussed in CPSES FSAR Section 5.4.1.3.3.
This damage mechanism should not lead to significant RCP leakage and would not prevent the safe shutdown of the affected unit.
RCP protection circuitry is available to provide an indication and to minimize the impact of this potential damage mechanism.
A Phase B Containment Isolation signal can be reset, after a minimal time delay, and the CCW supply to the RCP thermal barrier can be quickly reestablished. The RCPs are not expected to be damaged while operating during this short period without CCW cooling.
B.2 The components required to trip the RCPs have been identified.
The primary instrumentation (used to insure an RCP trip) is located out-side of the primary containment.
Therefore, this equipment is not significantly affected by the in-containment environment.
'C.1 The operator training program for tripping the Reactor Coolant Pump is divided into two main areas. The first area of training is conducted during the initial phase of the licensed operator training program.
During the Transient and Accident Analysis lecture series, a lecture ic given solely on forcea coolant flow (which includes starting and stopping the Reactor Coolant Pumps). The second area of training is conducted during the licensed operator requalification program.
During the Emergency Response Guideline (ERG) lecture series, Reactor Coolant Pump starting and stopping criteria is defined, and the reasons for each is given. Also included in this training program is the tripping of these pumps during normal conditions, which is covered under the Integrated Operating Procedures lecture series and the Abnormal Operating Procedures lecture series.
The following is a brief description of the training program for the starting and stopping of the Reactor Coolant Pumps under normal and abnormal conditions:
Under condition 1 (Normal Operations and Operational Transients) and condition 2 (Faults of Moderate Fraquency) events, the operators are trained to recognize upper limits on selected parameters that establish conditions for tripping the Reactor Coolant Pumps.
By having a para-meter exceed its limit, the operator is instructed to trip the pump (ABN-101A).
The parameter would be indicative of an off normal condition and the Reactor Coolant Pump trip is performed to prevent potential damage to the pump. Therefore, during normal power opera-tion, Reactor Coolant Pump trip criteria is established largely to pro-vide protection for the pumps and does not result in plant damage or challenges to the protective and/or safety systems.
Under condition 3 (Infrequent Faults) and condition 4 (Limiting Faults) events, the operator is trained to trip the Reactor Coolant Pump using the same criteria as mentioned above.
In addition to this criteria, there are other considerations that must be observed when the Reactor Coolant Pumps should be tripped or should remain running.
During accident conditions, there are some situations which may warrant tripping the Reactor Coolant Pumps.
For example, during the initial stages of a Small Break Loss of Coolant Accident (SBLOCA), if selected system parameter setpoints are reached, the Reactor Coolant Pumps should be tripped to avoid more serious conseluences.
During the long term recovery from many accidents, it is desirable to trip some of the operating Reactor Coolant Pumps to make recovery operations more easily achievable.
These situations arise when the additional heat input to the RCS from the Reactor Coolant Pumps is large enough to hinder plant cooldown.
Under condition 3 or 4 events which require using the function Restoration Guidelines (FRG), the operators are taught that Reactor Coolent Pump trip is based on different considerations. In FRC-0.2, RESPONSE TO DEGRADED CORE COOLING, one Reactor Coolant Pump is tripped (if all pumps are running), to prevent possible pump damage by running under highly voided conditions.
This is done to save the pump for potential future use.
In FRH-0.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, all Reactor Coolant Pumps are tripped to minimize secondary side inventory depletion.
In FRC-0.1, RESPONSE TO INADEQUATE CORE COOLING, and FRC-0.2, RESPONSE TO DEGRADED CORE COOLING, when adequate core cooling has been established and plant conditions are stabilized at a low temperature, trip of all Reactor Coolant Pumps is permitted.
Operators are also taught that there are numerous situations when the RCPs should remain operating or should be restarted if the pumps have been remo.>ed from operation earlier in an accident sequence.
These situations arise from a desire to provide normal pressurizer spray and forced RCS flow. Also, RCP restart may be necessary in response to certain accident conditions beyond the plant design basis, such as to respond to an Inadequate Core Cooling condition.
C.2 The following list identifies those procedures which include RCP trip related operations.
(a) RCP trip using WOG alternate criteria (1) E0P-0.0,REACkORTRIPORSAFETYINJECTION (2) E0P-1.0, LOSS OF REACTOR OR SECONDARY COOLANT (3) EOS-1.2, POST-LOCA C00LDOWN AND OEPRESSURIZATION (4)
E0P-3.0, STEAM GENERATOR TUBE RUPTURE (5) EOS-3.1, POST-SGTR C00LDOWN USING BACKFILL (6) E05-3.2, POST-SGTR C00LD0WN USING BLOWDOWN (7) E05-3.3, POST-SGTR C00LDOWN USING STEAM DUMP (8) ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION (9)
ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS (10) ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUBC00 LED REC 0VERY DESIRED (11) ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT-SATURATED i
REC 0VERY DESIRED (12) ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL (13) FRC-0.1, RESPONSE TO INADEQUATE CORE COOLING (14) FRC-0.2, RESPONSE TO DEGRADED CORE COOLING (15) FRH-0.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK i
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(b) RCP restart (1) E05-0.1, REACTOR TRIP RESPONSE (2) EOS-0.2, NATURAL CIRCULATION C00LDOWN (3) E05-0.4, NATURAL CIRCULATION C00LD0WN WITH STEAM VOID IN VESSEL (WITHOUT RVLIS)
(4) E05-1.1, SI TERMINATION (5) EOS-1.2, POST LOCA C00LOOWN AND DEPRESSURIZATION (6)
E0P-3.0, STEAM GENERATOR TUBE RUPTURE (7) ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS (8) ECA-3.1, SGTR WITH LOSS OF REACTOR C0OLANT SUBC00 LED REC 0VERY DESIRED (9) ECA-3.2, SGTR WITH LOSS REACTOR COOLANT - SATURATED REC 0VERY DESIRED (10) ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL (11) FRC-0.1, RESPONSE TO INADEQUATE CORE COOLING (12) FRP-0.1, RESPONSE TO IMMINENT PRESSURIZE 0 THERMAL SH0CK CONDITIONS (13) FRI-0.3, RESPONSE TO VOIDS IN REACTOR VESSEL (c) Decay heat removal by natural circulation (1) EOS-0.2, NATURAL CIRCULATION C00LDOWN (2)
E0P-1.0', LOSS OF REACTOR OR SECONDARY COOLANT (3) EOS-1.1, SI TERMINATION (4) EOS-1.2, POST LOCA C00LD0WN AND DEPRESSURIZATION (5)
E0P-3.0, STEAM GENERATOR TUBE RUPTURE (6) E05-3.1, POST-SGTR C00LDOWN USING BACKFILL (7) E05-3.2, POST-SGTR C00LDOWN USING BLOWDOWN (8) EOS-3.3, POST-SGTR C00LDOWN USING STEAM DUMP (9) ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION (10) ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS (11) ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED REC 0VERf DESIRED (12) ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT-SATURATED REC 0VERY DESIRED
-(13) ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL (d) Primary system veid removal (1) FRI-0.3, RESPONSE TO VOIDS IN REACTOR VESSEL (e) Use of steam generators with and without RCPs operating (1)
E0P-0.0, REACTOR TRIP OR SAFETY INJECTION (2) E05-0.1, REACTOR TRIP RESPONSE (3) E05-0.2, NATURAL CIRCULATION C00LD0WN.
(4) EOS-0.4, NATURAL CIRCULATION C00LDOWN WITH STEAM VOID IN VESSEL (WITHOUT RVLIS)
(5) E05-1.1, SI TERMINATION (6) E05-1.2, POST LOCA C00LD0WN AND DEPRESSURIZATION (7)
E0P-3.0, STEAM GENERATOR TUBE RUPTURE (8) E05-3.1, POST-SGTR C00LDOWN USING BACKFILL (9) E05-3.2, POST-SGTR C00LDOWN USING 6 LOWDOWN (10) E05-3.3, POST-SGTR C00LD0WN USING STEAM DUMP (11) ECA-1.1, LOSS OF EMERGENCY COOL ANT RECIRCULATION (12) ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS (13) ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED REC 0VERY DESIRED (14) ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT-SATURATED REC 0VERY DESIRED (15) ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL (f) RCP trip for other reasons (1) ABN-101A, REACTOR COOLANT PUMP TRIP / MALFUNCTIONS