ML20214C337
| ML20214C337 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/14/1986 |
| 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 85-510AA, GL-85-12, NUDOCS 8602210102 | |
| Download: ML20214C337 (10) | |
Text
m s
VauGINEA ELucTunc Axn Powr.n CoMPAxY Hicunoxo,V wnex A 2:sunt W,L.Stuwant V8CB I'WBelDBWT NocLBam Oramattone February 14, 1986 Mr. Harold R. Denton, Director Serial No.
85-510AA Office of Nuclear Reactor Regulation NAPS /JHL/ ace Atta: Mr. Lester S. Rubenstein. Director Docket Nos. 50-338 PWR Project Directorate #2 50-339 Division of PWR Licensing-A License Nos. NPF-4 U. S.
Nuclear Regulatory Commission NPF-7 Washington, D.
C.
20555 i
Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 RESPONSE _ TO GENERIC LE'!TER 85-12 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS 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 lector provides the requested information in the enclosed attachment.
l If you have any further questions regarding this matter please contact us.
Very truly yours, l
I l L.
W. L. Stewart Attachment i
l o!"AM @y s
\\
Woo wtA Etacimic Awo Powse COMPANY TO Ilarold R. Denton l
l 1
l c
l ec:
Dr. J. Nelson Grace Regional Administrator j
Region II l
Mr. Leon B. Engle NRC North Anna Project Manager PWR Project Directorate #2 Division of PWR Lie.ensing-A i
I Mr. M. W. Branch l
NRC Senior Resident Inspector l
North Anna Power Station l
I I
f b
t i
t l
c l
s I
?
t L
t I
i 7
I L
I E
+
i l
APPENDIX A i
RESPONSE TO SECTIONS IV.A. IV.B AND IV.C OF l
NRC GENERIC LETTER 85-12 l
IMPLEMENTATION OF TMI ACTION ITEM II.K.3.5 l
AUTOMATIC RCP TRIP NORTil ANNA POWER STATION t
i i
l t
l i-4 e
1 i
= _ - _.
Section IV.A - Determination of RCP Trip Criteria IV.A.1 - Identify Required Instrumentation For the North Anna Power Station we have 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.
Instrument Make/
Parameter Mark No.
Afodel No.
Wide Range RCS Pressure PT-402, 403 Rosemount/l153CD9PA Wide Range Hot Leg Temp.
TE-413,423,433 Weed /N9003D2B RCS Subcooling SCI-RC100A,100B 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 crit 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 ou-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 I
one instrumentation channel is operating for each of the SMS input parameters.
At NAPS, Channel A of the SMS receives these inputs:
PT-402, PT-403. TE-413, TE-423 and signals from 8 thermocouples.
The inputs to Channel B ares i
PT-402, PT-403 TE-423 TE-433 and 8 thermocouple signals.
l 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 I
channel accuracy for subcooling includes thene individual instrument uncertainties and uncertainties associated with processing and reading the l
subcooling indication.
The adverse containment parameters were selected f rom t
l the plant Environmental Zone Descriptions for the areas in which required instrumentatien 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 inntrumentation and physical separation of instrument piping linas ensuren at least one channel of each l
subcooling parameter in maintained following the eventn for which a RCP trip decision is needed.
1 l
l
Instrument Uncertainty for These Containment Conditions Parameter Normal Adverse Wide Range RCS Pressure
+/- 50 psi
+/- 333 pst Wide Range Hot Leg Temp.
+/- 16.2 deg. F +/- 16.2 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 values.
LOFTRAN is a Westinghouse licensed code used for FSAR SGTR and Non-LOCA analyses. The code has been validated against the January 19F2 SGTR 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, prescurizer level, full SI train and AFV pump operation) are due to either models or inputs to LOFTRAN.
The significant inputs in determination of the RCp trip criteria ares 1.
Break flo.i 2.
SI flow 3.
Decay heat 4.
Auxiliary feedwater flow 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 includen a substantial amount of connervatism (i.e. predict higher break flow than actually expected).
Wentinghouse han performed analysen and developed a more realistic break flow model that han been validated against the Ginna SGTR data.
The break flow model used in the WOG analyses has been shown to be approximately 30% conservative when the ef fect of the higher predicted break flow in compared to the more realistic model.
The consequence of the higher predicted break flow in a lower than expected predicted minimum pressure.
The SI flow inputs used were derived from bent estimate calculations, assuming both SI trains operating.
An evaluation of the calculational methodology shows that there inputs have a maximum uncertainty of +/- 10%.
The decay heat n.odel used in the WOC analyses van based on the 1971 ANS 5.1 rtandard. When cempared with the more recent 1979 ANS 5.1 decay heat inputn, the values used in the WOG analynen are higher by about 5%.
To determine the ef fect of the uncertainty due to the decay heat model, a nennitivity study w.u; conducted for SGTR.
The results of thin study nhow that a 20% decrease in decay heat resulted in only a 1% decrease in RCS prennure 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 the auxiliary feed pumps are running, minimum pump sta-t delay, and no throttling.
To evaluate the uncertainties with AFW flow rate, a sensitivity study was performed. Results from the three 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 dif ferential.
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 an a result of these considerations for the North Anna 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.
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 (Westinghouse Owners Group Report on RCP Seal Performance Following Loss of All AC) 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 loan of all AC, will be maintained for many hours".
Seal water injection is supplied by the charging /S1 pumps through valves which do not receive a containment isolation signal.
Also, the valven through which seal injection is supplied are normally open and will fail in this position on a loss of power.
Seal water return from the RCPs to the charging /SI pump suction line passes through valves MOV-1380 and HOV-1381.
These valves are closed automatically by Phase A (SI) containment inolation.
Iloweve r, 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 pannen are the three Number 1 seal leak-of f control valves (one for each RCP).
These valves are manually operated and are closed upon a lone of sent water injection to the RCP, an required by the manufacturer.
They fail open upon a loss of power.
It can be concluded that non1 water will not be dinrupted by either an automatic containment inolation nignal or manunt isolation resulting from a misdiagnosis of a LOCA event.
Component cooling water (CCW) for the RCP thermal barrier coolers and the other RCP motor cooling functions is supplied via containment isolation check valves CC-84, CC-Il9 and CC-154.
The following North Anna CCW valves are isolated on a Phase B signal (Hi-Ili pressure): CCW inlet for RCP cooling (TV-CC104A,
- 104B, 104C),
thermal barrier outlet (TV-CC101A, 101B) and remaining RCP cooling (TV-CC102A thru 102F).
Automatic valves terminate CCW flow to RCPs following a Phase B (Hi-Hi pressure) isolation signal. Operators are instructed by emergency procedures to stop all RCPs that have lost CCW cooling.
decause of these instructions, the RCPs will not continue to run af ter any loss of CCW.
WCAP-10541 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 /> wil5 the loss of either seal injection or cooling water to the thermal barrier, but not both.
Since seal injection is maintained following containment isolation, it alone will provide adequate RCP seal cooling.
This will maintain seal integrity, but restoration of RCP motor component cooling is necessary prior to restarting RCPs.
The above has shown that a Phase A containment isolation will allow continued RCP operation. This signal operates only two in-series valves associated with RCP water services.
As mentioned above, these valves are shut on Phase A isolation, diverting seal water return flow to the pressurizer relief tank.
RCP water services will thus be maintained.
Events generating a Phase B containment isolation (either LOCA or Non-l.0CA) will lead to CCW termination and tripping of RCPn in accordance with station procedures. This will prevent RCP damage and allow for rentart upon restoration of normal RCP cooling functions.
Section IV.B.2 - Identify Iomponents Required for RCP Trip The RCP trip components are powered from 125V DC at the 4160V switchgear. The circuit required to actuate a 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 Nonc needed to achieve manual RCP trip 2.
Power Supplien DC Bus / Batteries 1-1, I-II, 1-!!!, 2-1, 2-11, 2-!!!
35A DC Trip Funes 50A DC Breaker in DC Distribution Panel 3.
Breakers Brown Boveri 1200A ITE 5HK-250 4.
Control Switchen G.E. SBM
- - _ - -. - - - ~ _
Section IV.C - Operator Training and Procedures (RCP) l IV.C.1 Describe the operator training program for RCP trip.
Include the general philosophy regarding the need to trip pumps versus the i
desire to keep pumps running.
The operator training program is currently under development, with l
training to commence February 10, 1986. The philosophy and informa-tion of paras. A & B above are being used to formulate this l
program.
The actual training will include lectures and simulator sessions.
The training will include the need to trip pumps versus the desire to keep RCPs running. Operations personnel will receive t r u ning on the RCP trip criteria during the Rev. 1 Emergency Ope.. ting Procedure (Rev. 1, E0P) training, which is part of their annual Licensed Operator Requalification Program (LORP).
This training will be completed by June 30, 1986.
IV.C.2 Identify those procedures which include RCP trip related operations:
The Rev. 1 E0Ps, including those for Reactor Coolant Pump trip will be revised and implemented by June 30, 1986.
IV.C.2(a) RCP Trip Using WOG Alternate Criteria i
EP-0 Ecactor Trip / Safety Injection EP-1 Loss of Reactor or Secondary Coolant EP-3 Steam Generator Tube Rupture IV.C.2(b) RCP Restart ES-0.1 Reactor Trip Fesponse ES-0.2 SI Termination for Spurious SI ES-0.3 Natural Circulation Cooldown r
i ES-0.5A Natural Circulation Cooldown with potential for steam void in upper vennel head (without RVLIS)
ES-0.5B Natural Circulation Cooldown with potential for steam void in upper vennel head (with RVLIS) and no accident in progrenn ES-1.1 SI termination following loam of reactor coolant ES-l.2 Pont LOCA cooldown nnd depressurization ES-2.1 Si termination following lunn of nocondary coolant ES-2.2 Si termination following excennive RCS cooldown l
l
s' ES-3.3 SGTR with. secondary depressurization p
ES-3.4 SI~terminacion following SGTR FRP C-1 Response to' inadequate core cooling
- ~'
FRP-I.3A Response to Vold in Reactor Vessel 4
FRP-I-3B Response to Void in Reactor Coolant System FRP-I.3C Alternate Response to Void in Reactor Coolant System IV.C.2(c)
Docay liest hemoval by Natural Circulation 7
. F,9-0.1, Reactor Trip Recovery s
ES-0.3 ' Natural Circulation Cooldown 1
ES-0.5B Natural Circulhtion Cooldown with void in Reactor Vessel with RVLIS ES-0.5A Natural Circulation Cooldown with void in Reactor Vessel without RVLIS 4
ES-0.2 SI Termination Following Spurious SI l
l ES-1.1 SI Termination ES-1.2 Post LOCA Cooldown and Depressurization EP-3, Steam Generator Tube Rupture ECA-0.1 Loss of All AC Power Recovery without SI Required
~
ECA-3 Uncbntrolled Depressurization of all Steam Generators l
ES-3.3 SGTR with secondary depressurization IV.C.2(d) Primary System Void Removal FRP-I.3A Response to Void in Reactor Vessel RVLIS available FRP-I.3B Response to Void in Reactor ' Coolant System no RVLIS i
FRP-13C Alternate Responne tn Void in Reactor Coolant System l
l
,r
-0 IV.C.2(e) Use of Steam Generators With and Without RCP's Operating FRP-C.1 Response to Inadequate Core Cooling FRP-C.2 Response to Degraded Core CoolinF ECA-0.0 Loss of All AC Power EP-3 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-0, I and 3)
FRP-H.1 Response to Loss of Secondary Heat Sink