ML18144A046
| ML18144A046 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/06/1985 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Rubenstein L Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM 85-510B, GL-85-12, NUDOCS 8512120233 | |
| Download: ML18144A046 (10) | |
Text
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 W. L. STEWART VICE PRESIDENT NUCLEAR OPERATIONS December 6, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. Lester S. Rubenstein, Director PWR Proj~ct Directorate #2 Division of PWR Licensing-A U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 RESPONSE TO GENERIC LETTER 85-12 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS Serial No.
85-510B NO/ETS:acm Docket Nos.
50-280 50-281 License Nos. DPR-32 DPR-37 Generic Letter 85~12, dated June 28, 1985, requested submittal of the information identified in section IV of your Safety Evaluation (SE) for Westinghouse Owners Group Reactor Coolant Pump Trip. This letter provides the requested information in the enclosed attachment.
An upgraded incore thermocouple system is to be installed at Surry Units 1 and 2.
Upon its completion, we will choose between (1) continuing to utilize input from the wide range hot leg RTDs or (2) utilizing input from the incore thermocouple system as an input to the Subcooling Monitoring System.
If we choose to utilize the incore thermocouple system~ we will re-evaluate this response and inform you of any changes.
If you have any question or require additional information please contact us.
~rr~**
W. L. Stewart
Attachment:
8512120233 851206 PDR ADOCK 05000280 p
e VIRGINIA ELECTRIC AND PoWER COMPANY TO cc:
Dr. J. Nelson Grace Regional Administrator NRC Region I I Mr. Donald J. Burke NRC Resident Inspector Surry Power Station Mr. Terence L. Chan Mr. Harold R. Denton NRC Surry Project Manager Operating Reactors Branch No. 1 Division of Licensing e
e RESPONSE TO SECTIONS IV.A, IV.Band IV.C OF NRC GENERIC LETTER 85-12 IMPLEMENTATION OF TMI ACTION ITEM II.K.3.5 AUTOMATIC RCP TRIP SURRY POWER STATION
e e
Section IV.A - Determination of RCP Trip Criteria IV.A.l - Identify Required Instrumentation For the Surry plant, Virginia Electric and Power Company has chosen RCS subcooling based on wide range hot leg RTDs as the criterion for manually tripping reactor coolant pumps (RCPs).
Use of the subcooling criterion requires the following instrumentation:
- 1) wide range RCS pressure transmitters, 2) wide range hot leg RTDs and 3) analog core subcooling meter.
Backup temperature measurement is available from the core exit thermocouples.
The channel mark numbers and instrument model numbers are provided below.
Each of the above instruments provides an input to the existing Subcooling Monitoring System (SMS).
The required subcooling value, for RCP trip, will be read directly from the installed subcooling meter.
Parameter Wide Range RCS Pressure Wide Range Hot Leg Temp.
RCS Subcooling Meter Mark No.
PT-402, 402-1 TE-413,423,433 SCI-RClOOA, lOOB Instrument Make/
Model No.
Rosemount/1153GD9 Weed/N90071D2B Westinghouse The SMS is divided into two redundant channels, each receiving several inputs including two wide range pressure and two wide range RTD inputs.
Each channel also receives input from 8 core exit thermocouples.
Each channel of subcooling is processed in a separate microprocessor, with input signals isolated from the reactor protection instrumentation. Each channel has an analog meter displaying margin to saturation in degrees F on the control board.
The calculated setpoints do not take credit for any existing redundancy of instrumentation.
Therefore, the setpoints are applicable as long as at least one instrumentation channel is operating for each of the SMS input parameters.
At the Surry plants, Channel A of the SMS receives these inputs:
PT-402, PT-402-1, TE-413, TE-423 and signals from 8
thermocouples.
The inputs to Channel B are: PT-402, PT-402-1, TE-423, TE-433 and 8 thermocouple signals.
IV.A.2 - Identify Instrumentation Uncertainties Listed below are the specific normal and adverse instrument uncertainties for the portion of the instrument loop which is input to the SMS.
The overall channel accuracy for subcooling includes these individual instrument uncertainties and uncertainties associated with processing and reading the subcooling indication.
The adverse containment parameters were selected from the plant Environmental Zone Descriptions for the areas in which required instrumentation was located.
An evaluation of effects on instrumentation from local conditions such as fluid jets or pipe whip was performed.
Instrument and instrument piping locations were reviewed for exposure to effects from large and small RCS piping breaks, main steamline and main feedline breaks.
This evaluation concluded that the redundant instrumentation and physical separation of instrument piping
e lines ensures at least one channel of each subcooling parameter is maintained following the events for which a RCP trip decision is needed.
Parameter Wide Range RCS Pressure Wide Range Hot Leg Temp.
Instrument Uncertainty for These Containment Conditions Normal Adverse
+/- 52 psia
+/- 382 psia
+/- 16.7 deg F +/- 16.7 deg F IV.A.3 - Consider WOG Generic Analysis Uncertainties The LOFTRAN computer code was used to perform the alternate RCP trip criteria analyses.
Both Steam Generator Tube Rupture (SGTR) and Non-LOCA events were simulated in these analyses. For all but three of the cases analyzed, results from the SGTR analyses were used to obtain the RCP trip parameter valves.
LOFTRAN is a Westinghouse licensed code used for FSAR SGTR and Non-LOCA analyses.
The code has been validated against the January 1982 SG'.fR event at the Ginna plant.
The results of this validation show that LOFTRAN can accurately predict RCS pressure, RCS temperatures and secondary pressures especially in the first ten minutes of the transient. This is the critical time period when minimum pressure and subcooling is determined.
The major causes of uncertainties and conservatism in the computer program results, assuming no changes in the initial plant conditions (i.e. full power, pressurizer level, full SI train and AFW pump operation) are due to either models or inputs to LOFTRAN.
The significant inputs in determination of the RCP trip criteria are:
- 1. Break flow
- 2. SI flow
- 3. Decay heat
- 4. Auxiliary feedwater flow (AFW)
The following sections provide an evaluation of the uncertainties associated with each of these items.
To conservatively simulate a double ended tube rupture in safety analyses, the break flow model used in LOFTRAN includes a substantial amount of conservatism (i.e. predict higher break flow than actually expected).
Westinghouse has performed analyses and developed a more realistic break flow model that has been validated against the G.:inna SGTR data. The break flow model used in the WOG analyses has been shown to be approximately 30%
conservative when the effect of the higher predicted break flow is compared to the more realistic model.
The consequence of the higher predicted break flow is a lower than expected predicted minimum pressure.
e The SI flow inputs used were derived from best estimate calculations, assuming both SI trains operating.
An evaluation of the calculational methodology shows that these inputs have a maximum uncertainty of+/- 10%.
The decay heat model used in the WOG analyses was based on the 1971 ANS 5.1 standard.
When compared with the more recent 1979 ANS 5.1 decay heat inputs, the values used in the WOG analyses are higher by about 5%.
To determine the effect of the uncertainty due to the decay heat model, a sensitivity study was conducted for SGTR.
The results of this study show that a 20% decrease in decay heat resulted in only a 1% decrease in RCS pressure for the first 10 minutes of the transient. Since RCS temperature is controlled by the steam dump, it is not affected by the decay heat model uncertainty.
The AFW flow rate input used in the WOG analyses are best estimate values, assuming that all three auxiliary feed pumps are running, minimum pump start delay, and no throttling.
To evaluate the uncertainties with AFW flow rate, a sensitivity study was performed.
Results from the 3 loop plant study show that, a 27% increase in AFW flow resulted in only a 3%
decrease in minimum RCS pressure, a 2% decrease in minimum RCS subcooling, and a 2% decrease in pressure differential.
The effects of all these uncertainties with the models and input parameters were evaluated and it was concluded that the contributions from
- the break flow conservatism and the SI uncertainty dominate.
The calculated overall uncertainty in the WOG analyses as a result of these considerations for the Surry units is -3 to +20 degrees F for the RCS subcooling RCP trip setpoint.
Due to the minimal effects from the decay heat model and AFW input, these results include only the effects associated with the break flow model and SI flow inputs.
e Section IV.B - Potential Reactor Coolant Pump ProblemsSection IV.B.1 - Containment Isolation for Non-LOCA Events RCP seal degradation is not expected within approximately 30 minutes of loss of both seal injection/return and component cooling water supply/return to the thermal barrier cooler.
This conclusion is based on WCAP 10541 which states that for "... a non-design basis accident, such as a station blackout,it is expected that seal integrity, under loss of all cooling due to a loss of all AC, will be maintained for many hours". Seal water injection is supplied by the charging/SI pumps through valves which do not receive a containment isolation signal.
Also, the valves through which seal injection is supplied will fail open on a loss of power.
As stated in UFSAR Section 9.1.3.1, the charging/SI pumps are cross-connectable between Units 1 and 2.
This ensures continued seal water supply upon a total loss of charging/SI pumps for one unit. The cross-connection is a local manual operation which can be performed in less than 30 minutes.
Seal water return from the RCPs to the charging/SI pump suction line passes through valve MOV-1381.
The valve is closed automatically by Phase I (SI) containment.isolation. However, seal return flow is maintained by diversion to the pressurizer relief tank via a full flow relief valve.
The other valves through which seal water return passes are the three Number 1 seal leak-off control valves (one for each RCP).
These valves are closed only upon a loss of seal water injection to the RCP, as required by the manufacturer. They fail open upon a loss of power. It can be concluded that seal water will not be disrupted by either an automatic containment isolation signal or isolation which may possibly result from misdiagnosis of a LOCA event.
Component cooling water (CCW) for the RCP thermal barrier coolers and the other RCP motor cooling is supplied via containment isolation check valves CC-1, CC-58 and CC-59.
There are no valves in the CCW supply piping which shut on containment isolation signals.
Surry has these valves in CCW return piping which isolate on a Phase III (Hi-Hi pressure) signal: CCW return from the thermal barriers (TV-CC107) and remaining CCW from each RCP (TV-CC105A, 105B, 105C).
Automatic valves terminate CCW flow to RCPs following a Hi-Hi pressure signal. Operators are instructed by emergency procedures to stop any RCP which has lost CCW and to stop the RCPs when CCW is lost following a Phase III isolation signal.
Because of these instructions, it is not expected that RCPs will continue running after any loss of CCW.
Since seal injection is maintained following containment isolation, it alone is expected to provide adequate RCP seal cooHng.
This will maintain seal integrity, but restoration of RCP motor component cooling is necessary prior to restarting RCPs.
WCAP-10541 (Westinghouse Owners Group Report on RCP Seal Performance Following Loss of All AC) states that RCP operation is allowed for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with either the loss of seal injection or cooling water to the thermal barrier, but not both.
e e
The above has shown that Phase I or Phase II containment isolation will allow continued RCP operation.
These signals operate only one valve associated with RCP water services. As mentioned above, these valves shut on Phase I isolation, diverting seal water return flow to the pressurizer relief tank.
RCP water services will thus be maintained.
Events generating a Phase III containment isolation (either LOCA or Non-LOCA) will lead to CCW termination and tripping of RCPs in accordance with station procedures.
This will prevent* RCP damage and allow for potential restart upon confirming that RCP CCW has been restored.
Section IV.B.2 - Identify Components Required for RCP Trip The RCP trip components are powered from 125V DC at the 4160V switchgear.
The circuit required to actuate manual RCP trip contains the components listed below.
No components are subject to adverse containment conditions, as defined in the plant Environmental Zone Descriptions.
- 1. Relays None needed to achieve manual RCP trip
- 2. Power Supplies DC Bus/Batteries lA, lB, 2A, 2B 30A DC Trip Fuses 50A DC Breaker in DC Distribution Panel
- 3. Breakers Brown Boveri 1200A ITE SHK-250
- 4. Control Switches Electroswitch
Section IV.C - Operator Training and Procedures (RCP)
IV.C.1 Describe the operator training program for RCP trip.
Include the general philosophy regarding the need to trip pumps versus the desire to keep pumps running.
The operator training program for RCP trip is currently under development.
The philosophy and information of paras A and B above are being utilized to formulate this program.
The actual training will include lectures and simulator training.
The training program will include the philosophy of the need to trip pumps versus the desire to keep pumps running.
Operations personnel will receive training on the RCP Trip Criteria as part of their licensed operator requalification program by July 31, 1986.
IV.C.2 Identify those procedures which include RCP trip related operations:
Operating procedures for reactor coolant pump trip will be revised by May 31, 1986.
IV.C.2(a)
RCP Trip Using WOG Alternate Criteria EP-1.00 EP-2.00 EP-4.00 Reactor Trip/Safety Injection Loss Of Reactor Or Secondary Coolant Stearn Generator Tube Rupture IV.C.2(b)
RCP Restart EP-1.01 EP-1. 02A EP-1.02B EP-1.02C Reactor Trip Recovery Natural Circulation Cooldown Natural Circulation Cooldown With Void With RVLIS Natural Circulation Cooldown With Void Without RVLIS SI Termination Following Spurious SI SI Termination Post LOCA Cooldown And Depressurization In Reactor In Reactor Vessel Vessel EP-1.03 EP-2.01 EP-2.02 EP-3.01 EP-4.00 ECA-4.01 ECA-4.02 FRP-C.1 FRP-I.3A FRP-I. 3B FRP-I. 3C Uncontrolled Depressurization Of All Stearn Generators Stearn Generator Tube Rupture SGTR With Loss of Reactor Coolant Subcooled Recovery SGTR With Loss Of Reactor Coolant Saturated Recovery Response To Inadequate Core Coolant Response To Void In Reactor Vessel Response To Void In Reactor Coolant System Alternate Response To Void In Reactor Coolant System IV.C.2(c)
Decay Heat Removal by Natural Circulation EP-1.01 Reactor Trip Recovery EP-l.02A Natural Circulation Cooldown EP-1.02B Natural Circulation Cooldown With Void In Reactor Vessel
e With RVLIS EP-10.2C Natural Circulation Cooldown With Void In Reactor Vessel Without RVLIS EP-1.03 SI Termination Following Spurious SI EP-2.01 SI Termination EP-2.02 Post LOCA Cooldown And Depressurization EP-4.00 Steam Generator Tube Rupture ECA-1.01 Loss Of All AC Power Recovery Without SI Required ECA-3.01 Uncontrolled Depressurization Of All Steam Generators ECA-4.01 SGTR With Loss Of Reactor Coolant Subcooled Recovery ECA-4.02 SGTR With Loss Of Reactor Coolant Saturated Recovery IV.C.2(d)
Primary System Void Removal FRP-I.3A Response To Void In Reactor Vessel RVLIS Avaliable FRP-I.3B Response To Void in Reactor Coolant System No RVLIS FRP-I.3C Alternate Response To Void In Reactor Coolant System IV.C.2(e)
Use of Steam Generators With and Without RCP's Operating FRP-C.1 FRP-C.2 ECA-1.00 ECA-2.01 EP-4.00 Response To Inadequate Core Cooling Response To Degraded Core Cooling Loss Of All AC Power Loss Of Emergency Coolant Recirculation Steam Generator Tube Rupture IV.C.2(f)
RCP Trip for Other Reasons Emergency procedure fold out page requires trip of RCP on loss of component cooling services to the RCP Motor.
(attachment to EP-1.00, 2.00 and 4.00)
FRP-H.1 Response To Loss Of Secondary Heat Sink