ML20069H457
| ML20069H457 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/31/1983 |
| From: | Baxter T METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | Buck J, Edles G, Gotchy R NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8304060099 | |
| Download: ML20069H457 (106) | |
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SHAW, PITTMAN, PoTTs & TROWBRI A PARTNERSHIP OF PROFESSIONAL CORPORATIONS kg isoo - srRecr,~...
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g RAMSAY D. POTTS. P C JOMN A. McCutLouGM. P C (2o2) e22.sooo 0a n
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CT4 PMEN D POTTS. P C.
J THOMAS LENMART, P C TELECOpeER JCMN L M
WE N DE LIN ITE GUMLD CHANNOFF. P c STEVEN L MELTZER, P C.
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(202)822-1099 & 822-1199 4S @tESU ROBERTM QR R. TIMOTHY MANLON. P C.
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WINTMROp N SROWN, P C.
ThER NE P CHEEK t
t RuCE w CMURCHILL P C.
JAMES S HAMUN, P C-(202)822-6072 l
heuCELEM RE R-STEIN
^U LE7UE A. N#CMOLSON, JR., P C, RANDAL5 MELL PC Ang gg JWAVIS T. SROWN, JR MANTIN O MRALL. P C Rose RT E. ZAMLER, P c CAMPRELL M8LLEFER mlCHARD M. MMONTMAL LICHARO J. RENDALL P C.
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SE M. MOOGASIAN STEPHEN s. MEIM ANN JAY E S8LSE RG. P C.
STEVEN M LUCAS, P C.
TELEX SM E ILA MCC MARVEY sARsARA M. ROSSOTTI. P C.
DAvlO M RU m EN STEIN. P C.
DEUSSA A. RIDGWAY SANDRA E. BRUSCA*
GIORGE V ALLEN. JR., p C RBCHARD E GALEN 8 9-2693 (S M AWLAW WSM)
MENNETH J. MAUTMAN g,L EN L SROWNELL FAED ORASNER. P C.
LYNN WMITTLESEY WILSON
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OAVID LAWRENCE MILLER PAMELA M ANDERSON R. RE NLY WESSTER, P C.
MATfAS F. TRAVIE SO-OsA2 FREDERICR L. MLEIN ALEXANDER o. TOMAS2CZUM NATHANIEL p BREED, JR., p C.
VICTORfA J. PERMINS STEVEN P PeTLER*
PMlup O. PORTER MARMAUGENSUCR.PC.
JOHN M O NEILL JR.
R6 CHARD J PARRINO MICHAEL A. SWIGER ERNZSTL ELARE, J R, P C.
JAY A. EPSTIEN ELLEN A. FREDEL#
ELLEN SMERIFF CAftETONS. JONES.PC.
RANDL. ALLEN THOMAS A SANTE R, P C.
TIMOTMY 5. McSRIDE JOHN F DEALYe MMMEM UEBERMAN WM J NM SANDRA E. FOLSOM EILEEN M GLEIMER JAMES M BURGER.pC.
EUSASETM M. PENDLETON COUNSEL JUDITH A. SANDLE R CAVID R SAMR EMILDON J. WEISEL P C MARRY M GLA SS P8 EG EL EDWARD O. YOUNG, til C BOWDotN TRAIN
- enOT Aceer??to see O C WRITER S DIRECT D4AL NUMSER March 31, 1983 822-1090 Gary J. Edles, Esquire Dr. John H. Buck Chairman Atomic Safety and Licensing Appeal Atomic Safety and Licensing Appeal Board Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Washington, D.C.
20555 Dr. Reginald L. Gotchy Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 In the Matter of Metropolitan Edison Company (Three Mile Island Nuclear Station, Unit No. 1)
Docket No. 50-289 (Restart)
Administrative Judges Edles, Buck and Gotchy:
i This follows up on my letter of January 25, 1983, which provided new information with respect to Licensee's Exception No. 1 on the implementation of NUREG-0737 Item II.F.2 (inade-quate core cooling instrumentation).
Please find enclosed a copy of a letter dated March 10, 1983 from GPU Nuclear Corporation (Hukill) to the NRC Staff nn 8304060099 830331
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b PDR ADOCK 05000289
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f SHAw, PITTMAN, PoTTs & TROWBRIDGE A PARTNERSMtp CF pmOFESStONAL COmpomanoNS Gary J. Edles, Esquire Dr. John H. Buck Dr. Reginald L. Gotchy March 31, 1983 Page Two (Eisenhut) describing the additional instrumentation proposed for installation at TMI-1.
Respectfully submitted, Thomas A.
Baxter Counsel for Licensee TAB:jah Enclosure cc:
Service List attached 1
UNITED STATES OF AMERICA NUCLEAR REGJLATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289
)
(Restart)
(Three Mile Island Nuclear
)
Station, Unit No. 1)
)
SERVICE LIST Gary J. Edles, Esqu.tre James M. Cutchin, IV, Esquire Chai man Office of the Executive Legal Director Atmic Safety and Licensing Appeal U.S. Nuclear Regulatory cmmission Board Washington, D.C.
20555 U.S. Nuclear Regulatory Ce miasion Washington, D.C.
20555 Docketing and Service Section Office of the Secretary Dr. John H. Buck U.S. Nuclear Regulatory cntmiasion Atmic Safety and Licensing Appeal Washington, D.C.
20555 Board U.S. Nuclear Regulatory cerrmimsion John A. Ievin, Esquire Washington, D.C.
20555 Assistant Cbunsel Pennsylvania Public Utility cmmimsion Dr. Reginald L. Gotchy P.O. Box 3265 Atmic Safety and Licensing Appeal Harrisburg, Pennsylvania 17120 Board U.S. Nuclear Regulatory crrrmimsion Ibbert Adler, Esquire Washington, D.C.
20555 Assistant Attorney General 505 Executive House Ivan W. Smith, Esquire P.O. Box 2357 Chai man Harrisburg, Pennsylvania 17120 Atcmic Safety and Licensing Board U.S. Nuclear Reculatory Comnission Ms. Iculse Bradford Washington, D.C.
20555
'IMI ALERT 1011 Green Street Dr. Walter H. Jordan Harrisburg, Pennsylvan.ta 17102 Atmic Safety and Licensing Board Panel Ellyn R. Weiss, Esquire 881 West Outer Drive Harnon & Weiss Oak Ridge, Tennessee 37830 1725 Eye Street, N.W., Suite 506 Washington, D.C.
20006 Dr. Linda W. Little Atmic Safety and Li nsing Board Steven C. Sholly Panel Union of Concerned Scientists 5000 Hermitage Drive 1346 Connecticut Avenue, N.W., Suite 1101 Raleigh, North Carolina 27612 Washington, D.C.
20036
. Jordan D. Cunningham, Esquire 2320 North Second Street Harrisburg, Pennsylvania 17110 ANGRY /IMI PIRC 1037 Maclay Street Harrisburg, Pennsylvania 17103 William S. Jordan, III, Esquire Harmon & Weiss 1725 Eye Street, N.W., Suite 506 Washington, D.C.
20006 01auncey Kepford Judith H. Johnsrud Envisummit.al Cbalition on Nuclear Power 433 Orlando Avenue State College, Pennsylvania 16801 Marjorie M. Aamodt R. D. 5 Coatesville, Pennsylvania 19320
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GPU Nuclear Corporation f@ WUClear Me 25'Souin'S S
Middletcwn, Pennsylvania 17057 717 944-7621 TELEX 84-23S6 Writer"s Direct Dial Number:
March 10 1983 5211-83-671
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Office of Nuclear Reactor Regulation p C' p v
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D. G. Eisenhut, Director i'
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Dear Sir:
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V Three Mile Island Nuclear Station, Unit 1 (TMI-1)
N Operating License No. DPR-50 Docket No. 50-289 Inadequate Core Cooling Instrumentation (hTREG 0737 II.F.2)
In response-to the " Order for Modification of License" dated December 10, 1982, GPUN has developed an Inadequate Core Cooling (ICC) system using the guidelines of NUREG 0737, Item II.F.2.
The system is composed of the following major ele-ments:
Subcooling Margin Monitor - which indicates the approach to a.
ICC by showing saturation conditions and superheat conditions.
b.
Core Exit Thermocouples - which is available for determining both the existence of ICC and the trends of recovery action.
RCS Inventory Trending System (RITS) - which indicates trend c.of inventory in the RCS above the core in the quiescent state and the trend in void fraction with RC pumps on.
d.
Other Instrumentation - such as wide range RC pressure, pressurizer pressure and level, emergency feedwater flow, wide range Ig, secondary steam pressure and stea= generator level transfer which provides additional information conceming heat from the core to the secondary side under condition of TCC.
Enclosures 1 and 2 to this letter describe the RCS Inventory Trending System which we now propose to install during Cycle 6 Refueling.
The RITS is a differential pressure system design which is capable of tracking RCS inventory from both the top of the hot leg and from the top of the Reactor Vessel head to the low point of the hot leg with RC pu=ps of f.
By using RC pu=p motor power and cerrelating it GPU Nuclear Corporation is a sutsic:ary of the General Fac U:mt es Ccrporat:en
,b Mr. D. G. Eisenhut '5211-83-071 to void fraction, GPUN vill be able to trend voiding in the RCS with RC pumps i
on (Enclosure 2). describes the conformance of existing systems (by reference or information supplied) to the requirements of NUREG 0737, Item II.F.2.
The conceptual design and the schedule for submission of engineering details are discussed in Enclosure 4.
For those ite=s for which information is not currently available, schedules for obtaining it are provided.
In order to proceed expeditiously on the detailed engineering design and long lead item procurement, it is necessary that we receive your concurrence on this conceptual design of the RITS prior to May 15, 1983.
The design of the inventory tracking system described in Enclosures 1 and 2 uses an incore instrument tube as the reference tap for the A-P system.
It is our understanding that some of the B&W Owners propose to use the decay heat drop line for this purpose. We believe that the decay heat drop line may have some advantages over the instrument tube, particularly with regard to accuracy of the water inventory measurement. We also understand that there may be concern with the asymmetry of the decay heat drop line.
In view of the complicated issues involved in choosing between these locations, our evaluation is continuing.
If the evaluation results in any change to the proposed design which utilizes the incore instru=ent tubes, we will submit information on this change prior to April 15, 1983.
Any change in our submittal would not alter our commitment to install a RITS during the Cycle 6 refueling.
Sincerely,
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i H. D.
utill Director, TMI-l HDH:LWH:jrg Enclosure cc:
R. C. Haynes J. Van Vliet i
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e INADEQUATE CORE COOLING SYSTDI EVALUATION FOR TMI-l RCS HOT LEG
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TABLE OF CONTENTS I.
INTRODUCTION II.
SUMMARY
OF USES OF RITS III.
DESCRIPTION OF RITS IV.
INADEQUATE CORE COOLING DETECTION SYSTDI - OVERVIEW V.
TRANSIENTS ANALYZED VI.
EFFECTS OF HPI VII.
OPERATOR GUIDANCE VIII.
CONCLUSIONS IX.
REFERENCES f
I.
INTRODUCTION History Following the accident at TMI-2 in March,1979, a great deal of attention was drawn to the subj ec t of inadequate core cooling (ICC) and additional instrumentation which might be added to monitor the approach to and recovery from ICC.
Soon af ter the accident, GPUN committed to install a subcooling margin monitor and wire the installed core exit thermocouples to the plant computer (June 28, 1979 letter). With the issuance of the TMI-1 Restart Report further details of these systems were provided as well as the expansion of the range of the hot leg RTD (Thot wide range).
In the fall of 1980 NRC issued NUREG 0737 which required an investigation of additional ICC instrumentation which should be considered for installation.
During the ASLB hearing in the spring o f 1981 the issue of inadequate core cooling was litigated e.::censively and in the December 14, 1981 Partial Initial Decision, the Board recommended that a water level instrument be installed at the Cycle 6 refueling outage.
In the fall of 1981, GPUN completed its evaluation of additional ICC instrumentation which employed the service of Dr. Dhir of UCLA and concluded i
that a Hot Leg Level Instrumentation System (HLLIS) was appropriate.
In the spring of 1982 the NRC indicated that the HLLIS was not sufficient to meet their criteria for measuring the approach to inadequate core cooling. On December 10, 1982, NRC issued an " Order of Modification of License" for TMI-1 that required installation of an ICC Instrumentation sys tem that conforms to the design 1
parameters specified in NUREG 0737, Item II.F.2.
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REOUIREMENTS The threc principal recuirc:r.ents for f r.stra.n..nt.1 tion to cetect inadequate core ecoling (ICC), as most recently stateil in Himt.c,.0737, are that the instrumentation syste r ust be unambiguous, pnivlite advance w:.rning, and cover the full range f operations.
The ICr..t. i..e tion, system herein evaluated roets these principal requirement..
It consists of three instrumentatien components:
the saturation me t or, RCS Inventory Trending System (RITS)
' and core exi t thennin.uuples. These three components when considered together are the tr.C eletection instrumentation.
The installation of a RCS Iriventory TrendinS system ( RITS ) in the S&'n' NSS system has bcen proposed as une of the in.truments to satisfy 1*c >
II.F.2 of HUREG-0737, Instrumentation for ui tii tion of Ir.cJequate Core Cooling. The purpose of this report is to evaluate the RITS, and provide an engineering assessment of the po..lble benefits of such a device to a plant operator in the event of plant upsets.
A summary of system uses and an outline of the background, comprise Section 2 of this report.
Section 3 is a brie r l.:scription of the RIrs including the physical cporating assutgtion. anil characteristics.
An overvicw of inadequate Core Cooling (ICC) is prcsonted in Section '.. The specific LOCA and non-l.0CA transients evalucted are contained in Section 5.
A necessary part of this evaluation involw. an analysis of various HP1 flow rates and the effects on a small breat (.li) t.0CA transient.
This analysis is covered in Section 6 and leads attr.etly into suggested operator guidance for using the RITS. Section /.
Overall report conclusions can be found in Section 8.
Section 9 lists the references used.
n....
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II.
SUS'MW OF USES OF
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Item II.F.2 of NUREG-0737 specifies tne reccirement for an unactig;cus indicatien of ina.decuate core. cooling (ICC).
A part of this ICC t
indication ceuld involve w. iter icvel instrumentatien.
Th e E5'a' NSS system's unique design allows for =casurc:ent of reacter ecolant level in the riser section of the hot leg.
Otr.cr NSS system designs do not permit measurement of icvel at the high point and must utili:e a measurement cf reactor vessel icvel (refer to Figure 1). The acvantage of the ncs ;nven, Trending system (RITS) is that het leg and head voiding v111 Recede a possibic ICC situation.
the R3TS will provide khe cperator with an l
indication of voided conditions in the reactor coolant system (RCS) anc alert the operator to the potential for core uncovery.
f k'ith the impic entation of the RITS, tne cperator will have at his disposal an additional cevice for early warning of possible ICC events.
j Based on the icvel indications, the operator can take confir:atory
~
actions such as assuring =axicc= HPI fica, the stea ;cnerat:rs are i
depressuri:ing, anc the c crgency fcccsator level is being raised to g5' of the operating range. These ections, as discussed in this report, will provide a cass ac:cculation witnin tne reactor vessel to insure I
adequate core cooling.
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1 III.
Design Description The Reactor coolant Inventery Trending System (RITS) is a system designed to provide a means for the operator to monitor in the control room, o
The water inventory in the het leg of each primary loep and in the 4
reactor vessel when the reactor coolant pu=ps (RCP) are not operating, and o the veid content of the RCS when the RCP's are operating.
The RITS is composed of two subsystems to provide the above functions.
The water inventory trending subsystem is a differential pressure (d/p)
.l measurerent system censisting of two identical independent ins tru=ent loops to measure het leg inventory and two identical instru=ent loops to measure reacter vessel inv e n: cry.
The "A" Iceps will =easure the pri=ary coolant inventory in the "A" het leg and in the reae:cr vessel while the "3" loeps will measure the' inventery in :he "3" het leg and reacter vessel.
The differential pressure transmitters will be located in the Reacter Building outside of the D-rings. The het leg inventeries will be =easured by connecting the trans=itters to the high point vent ccnnection a: the tep of each het leg and to incere instru=ent guidetube via instru=en: tub ing.
The reactor vessel inventory will be =easured by connecting the transmitters to the thermoccuple vent valve cennec:ing in the vessel head and to incere guidetube Each transmit ter will have a wet reference leg.
Te = p e r a ture elements will be installed en the reference water eclu=ns.
Process ins tru=en t and signal condi:icning vil be previded to for water density as a function of water te=perature and to provide correct analeg =eter displays in the contrcl rec =.
(See Figures 2a.2b,2 c) s The differential pressure trans=it:ers will be pcwered from, and have their a
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signals conditioned by, the existing A2 and B2 Signal Conditioning Cabinets located in the Control Building El 338'6" for A2 and El 322'0" for B2.
The Signal Conditioning Cabinets are lE qualified Foxboro Spec 200 equipment powered from the vital AC buses A and B, respectively.
At the Signal Conditioning Cabinets the input signals from the transmitters will be converted from current to voltage signals, compensated based upon primary system reference leg temperature process variables and reference leg temperature, and routed to the Control Roo= and computer via qualified isolation devices.
The output signals will be displayed in the Control Room.
A " caution" sign will be mounted by the inventory indication stating that readings are valid only when the Reactor Coolant Pumps are idle.
a single The water inventory trending subsystem will be designed such that active failure will not prevent indication of the RCS inventory in at'least one of the hot legs and in the reactor vessel above the fuel. The range of the hot leg inventory measurement will account for approximately 51 vertical feet of water above the fuel while the range o f the vessel measurement will account for approximately 17 vertical feet of water above the fuel.
Connections to the primary coolant system will be made via 1/2" tubing so that a tubing break is within the capacity of the make-up pu=p.
The void fraction trending subsystem consists of two identicial independent instrument loops to measure RCP motor power.
One of the instrument loops will be associated with an "A" loop reactor coolant pump and one with a "B loop pump.
The monitored signals will be input to the plant computer where they can be displayed as RCS void fraction via an empirical correlation.
Since the RCP's are powered from non-lE sources of electrical power, this subsystem of the RITS will also be non-lE.
However, the void
_7_
fraction trending subsystem will be designed such that a single failure will not prevent ' indication of the RCS void fraction at all times when the RCP's are in operation. The range of the RCS void fraction indication will be from 0% to100%.
Void fraction indicators will be provided in the control room. A " caution" sign will be mounted by these indicators srsting that the readings are valid only when the RCP's are in operation.
Figure 2c shows instrumentation for the inventory trending using RCP power.
The pump power inputs ccee from existing power transformers. The Tc inputs come from the cold leg RTD's.
In the void calculator portion in the plant computer two operations would be performed:
a) The saturated liquid and vapor densities ({f and pg) corresponding to the Te input would be determined.
b) The densitites determined in a) would be input to the void fraction algorithm with the appropirate pump power and the void fraction c omputed.
The selector /averager permits the selection for output of void fraction for output of the void fraction for each pump or the loop average. The design concept assumes that all portions of the system are classified non-lE.
An estimate of the expected accuracy of the void fraction indication based on pump current has not been established.
The temperature ele =ents and current transformers have well defined accuracies. The pump, however, is not an instrument, and assignment of an accuracy to its performance as a density transducer cannot be expected to be made with the same certainty as for the other inputs.
The accuracy of the indication at the time of calibration will be determined by:
a) The accuracies of the temperature and current measurements.
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b) The ability of the signal processing circuit to match the pump
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current vs. density curve.
The subsequent accuracy of the indication will change to the extent that certain conditions deviate from the conditions at calibration.
The most
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significant of these are:
pump motor efficiency l
- motor power supply voltage
- motor power factor
- pump speed pump hydraulic torque In view of the proposed use of the void fraction measurement as a trending indication, we believe that the absolute accuracy is not as important as other characteristics of the measurement.
It is more important, in this application, that the measurement reliably follow trends in coolant density and void fraction. We believe that the important information to be derived from this indication is not the value of void fraction at a given moment but the trend of void fraction with time. Our assessment of the scale pump test data indicates that the pump current decreases with increasing void fraction in a manner which can provide clear and adequate trending signal to the operator. The data indicates that the usable range of the signal will be limited to between 15% and 40% void fraction at the pump suction.
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The process parameters applicable to the RITS are:
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Pressure:
0-2500 psig Temperature:
50*F-650*F Boron Cone.:
2270-0 ppm Water Level Range:
approx. 612" from top of Hot Leg.
approx. 200" from top of RV.
The environmental requirements for the D/P transmitter and associat d tubing inside the reactor containment building are:
Normal Conditions 40 year Base Temperature 130*F Relative ilumidity 1001 Pressure Atmos.
j Radiation MR/hr 0- 100 R/40 yr 3.5 x.0' Accident Conditions Temperature (Max.)
275'F Relative Humidity 100:
Pressure (Max) 50.6 P:IA Radiation Mr/hr at T=0 0-100 6 mon. accident exp 2 x 107 Chemical Spray PH E to Il The environmental conditions for the equipment in the Control' Builcing for both normal and accident conditions are:
40 year Base Temperature 75*F Relative Humidity 65%
Pressure Atmos.
Radiation MR/hr 0-10 R/40 yr Neglig: ole
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IV.
INADECUATE CORE CCOLING DETECTION SYSTEM - OVERVIE'a' The condition of ICC can develop only af ter a well defined set of events occur. This section is intended as an overview of ICC and an ICC detection system. A more detailed discussion can be found in Reference 7, " Evaluation of Instrumentation to Detect ICC".
First, there must be a loss of RCS inventory which leads to the saturation of the RCS. Then the inventory loss must continue such that the core actually starts to become uncovered.
This will result in the heat generated in the core not being removed since there will be portions not covered with liquid coolant.
This failure to remove the heat can then lead to high temperatures in the fuel. The ICC condition occurs when the heat removal capability of the coolant is exceeded to the extent that fuel temperatures rise to the point of fuel damage. This does not occur until the core becomes at least partially uncovered because when it is surrounded by liquid coolant, even at saturation temperature, there is adequate cooling of the core to prevent fuel damage.
The ICC detection system eenitors the prerequisite stages of ICC and the occurrence of ICC. The saturation meter will indicate when saturation conditions occur and thus provides advance warning that ICC is a possibility.
However, the onset of saturation conditions coes not in itself indicate that ICC will occur. There is a spectrum of LOCA events which will cause saturation conditions to occur but will not proceed to core uncovery or ICC. There are also non-LOCA events which will proceed to saturation conditions but do not have the potential for ICC.
Thus the saturation meter provides advance warning of ICC but is not an unambiguous' indication that ICC conditions could occur..
e In a small break LOCA of the si:e which could proceed to the ICC condition, the inventory loss continuing through saturation conditions will result in the formation of voids in the RCS. These voids will contain steam and gases released frem the coolant. The gas content will remain very small until the onset of ICC. Since there is no forced RCS flow at this point in the 53 LOCA scenario, steam and gas will collect at the system high points. The high points that will become void of liquid coolant first in the B&W plant will be the upper RV head and the upper U bend of the hot leg piping (Figure 1). The formation of steam and gas bubbles in the hot leg will cause a liquid level to develop which 'can be inferred from the differential pressure measured by a differential pressure transmitter connected to the top of the hot leg and to an incore instrumentation tube.
This instrument would indicate the level and thus provide additional information to the operator.
The specific nature of the information would be that the inventory loss is continuing (decreasing levtl) or that the inventory loss rate has been matched or exceeded by ECC injection rate (steady or increasing level). Thus the RITS would be additional advanced warning that conditions are progressing toward the ICC condition.
The RITS. not only supplements the advanced warning of the saturation meter but it also increases the range of the monitoring of the approach to ICC. About 63
).
This t
represents about100; of the total RCS inventory available to cool tne core.
4 The RITS would also indicate that a level existed in the hot leg for some non-LOCA transients which do not have the potential for ICC.
e e
However, the non-ambiguity of the RITS is a time related function.
Early in transient time, the LOCA and non-LOCA behave similar regarding hot leg level. But as the transient progresses, the level continues to decrease for the LOCA while, for the non-LOCA, the level will recover.
An incorrect indication of level when a level does not exist could be caused by instrument failure or a break in the RITS instryment sensory lines. However, this would not lead to incorrect actions by the operator because the action to be taken based only on RITs indication would be to ensure that the actions taken at saturation conditions were effected as required.
If a RITS instrument malfunction caused an incorrect level indication, the operator would have additional information on RCS conditions provided by the other loop 'RITS. the saturation meter and other normally monitored parameters. Even if the operator ignored this additional information and initiated HPI based on an incorrect belief that inventory has been lost in one hot leg, this would not be an unsafe action.
In the other case where instrument malfunction caused the RITS to indicate full when in fact a level existed, the operater would not be directed to secure HPI based on RITS indication. Rather, securing of HPI would be based on subcooling conditions. Thus, although the RITS could provide ambiguous or incorrect indication, it would not cause unsafe actions to be initiated.
As the SB LOCA scenario continues, the inventory loss is such that the top of the core becomes uncovered.
The core exit thermocouples will have indicated saturation temperatures since saturation conditons were reached much earlier in the transient.
As inventory continues to be depleted, the core fyel region becomes uncovered and the temperature of the steam
, 13-s l -
l
at the core exit begins to increase above saturation temperature. This superheated temperature sensed by the care exit thermocouples is additional advance indication that conditions are progressing toward ICC conditions. Ultimately the conditions of heat transfer would be such that core exit thermocouple temperatures would increase to the thresholds for further operator actions.
If for some reason, these actions were not affected, then ICC would occur and would be indicated by the core exit thermocouples.
e e
D a
t Y.
TRANSIENTS ANALYZED l
The transients to be reviewed for this evaluation required certain necessary modeling characteristics. Of primary importance was an adequate nodalization scheme to allow for void formation in the hot leg i
during the transient.
Also of importance was to be reasonably assured that a spectrum of small break LOCA's was represented. The.01 f t.2 break in the cold leg at the pump discharge is significant in this regard.
Break sizes of larger than. this.01' f t.2 (Reference 6) tend to continuously depressurize the RCS as the HPI and break size are adequate to remove produced energy.
The continuous depressuri:ation is also a result of continuous loss of inventory as one HPI pump is insufficient to overcome break flow.
However, the depressurization allows for other ECCS to be actuated (CFT and LPI) within adequate time to assure core covery and safe conditions.
On the other hand, for break sizes of snaller than.01 f t.2 the KCS can be kept at operating pressure and in a subcooled condition througn use of
~
the flakeup and HPI systems. Therefore, the.01 f t.2 break, being between these two regions, can result in an initial system depressurization but eventual repressuri:ation as decay heat energy production becomes too large for the HP1 and break size cooling combination. This repressurization will continue until such time that a condensation surface is established in the steam generators which will then dominate system response with a depressurization trend.
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Of the SB LOCA transients reviewed, the B&W " Blue Book" Repressuri:ation Analyses had the desired hot leg configuration (Reference 3).
Additional
' nodes were added (nodes 24 and 25, Figure 3) to more accurately represent formation of a steam bubble between the auxiliary feedwater injection point and the 180* bend in the hot leg. By decoupling these volumes (nodes 24 and 25) from the steam generator, a steam bubble can form within the upper portion of the RCS without being condensed through steam generator interaction.
Some points of interest concerning this SB LOCA transient (.01 f t.2 break at pump discharge) are as fo11cws:
Secuence of Events Time (Sec.)
Break occurs (.01 f t.2 at pump discharge) 0 Reactor trip, turbine trip, and RC pump 50 coast down occurs Main feedwater coastdown ends 65 Auxiliary feedwater flow to both steam 90 generators Hot leg voiding begins 100 HPI starts 190 Loss of Nat. Circ. in intact loop 340 Loss of Nat. Circ. in broken loop 650 Maximum repressurization reached 1500 Minimum mixture level at 5 ft. above the top 1700 of the core Long term cooling established about 4900 Peak cladding temperature 720F (initial value) d 1
e.
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~
The RCS depressurizes rapidly over the first 100 seconds to saturation pressure of about 1400 psia (Figure 4).
At this point, hot leg and upper plenum steam formation slow the rate of depressurization.
RCS depressurization continues until about 650 seconds when natural circulation ceases as hot leg level continues to decrease. The loss of steam generator heat removal causes the primary system pressure to begin increasing. At 1500 seconds, the maximum system pressure is reached
( 1750 psia) and begins to slowly decrease because steam condensation by the steam generator is established. This additional energy removal results in a decreasing RCS pressure transient thereafter. The hot leg mixture heights are shown in Figure 5.
As can be seen from the figure, natural circulation is lost at approximately 340 seconds in the intact loop and at approximately 650 seconds in the broken loop.
The cases reviewed for. non-LOCA events (Reference 1) have a detailed l
noding scheme arrangement in the upper regions of the hot leg so that the effect of void formation could be observed (NODE 32 and 33, Figure.6).
The specific case reviewed is a study of the impact of RC Pump Trip on the ESFAS Low RC Pressure signal.
Results (Figures 6-11) show that natural circulation flow was'te=porarily reduced in the PZR loop to 45 to 100 lb/sec. from 4 to 6 minutes (refer to Figure 9), with flow steadily increasing after this time period. The flow in the non-PIR loop remained relatively unchanged at about 1000 lb/sec. (refer to Figure 10). The l
steam bubble was collapsed, natural circulation was fully restored, and a i
greater than 50*F subcooled margin was achieved in the PZR loop.
I a
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l 17
The other non-LOCA cases reviewed, Reference 2, involved overcooling events analyzed in reply to Denton's 50.54(f) Show Cause, Parts A-B (177 FA Plants).
In this report, more than one transient type is, analyzed to address different frequency of occurrence classifications and' to assure the most severe cases are included in 'the evaluation.
Included are overcooling accidents, such as main and small steam line breaks, and overcooling transients, such as pressure regulator malfunction and main feedwater overfill. All these transients retained adequate core cooling through natural circulation even when analyzed with no operator action before 10 minutes and with only safety systems used to mitigate the event. A typical event is a small steam line break (SLB) case (Figures 12 thru 14). Following are the results of this particular event:
Secuence of Events Time (sec.)
Steam line break occurs 0.0 l
High flux trip setpoint 4.67 Rod movement starts 5.07 Low RC pressure ESFAS, RC Pump trip, LOOP 16.25 MSIV closes 23.75 Pressurizer empties 25.0 MFWIV closes 31.25 HPI starts to flow 46.25 Auxiliary FW initiation to steam 85.7 l
generator Hot leg voiding begins 100 Hot leg solid 600 e
~
A general overvied of hot leg level response for those LOCA and non-LOCA
'transients reviewed is given in Figure 15. Not shown on Figure 15 is the level response for other SB LOCA's. However, as stated previously, larger break sizes result in continuous loss of inventory, and hence a decreasing hot leg level, until such time as the depressuri:ation actuates other ECCS that assures long term recovery.
Smaller break sizes result in little or no inventory loss. The important point to notice is that within minutes, hot leg level is completely regained in the non-LOCA events due to the RC system regaining the subcooling margin and refilling. Herein lies the fundamental conclusion that the hot leg level can be a determining parameter for distinguishing between the LOCA and non-LOCA events.
This will then lead to further differentiation of the appropriate operator actions necessary to mitigate the particular events.
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--- SMALL STEAM LINE BREAK, PUW TRIP AT ESFAS, NO DECAY llEAT, STUCK RELIEF VALVE O
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100 200 300 400 500 600 700 000 900 1000 1100 1200 1300 1400 Time, seconds
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EFFECTS OF HPI An evaluation of the effects of changes in HPI flow rates has been performed. The transient used as a basis for this analysis is the
.01 ft.2 pump discharge repressurization break from Reference 3.
Calculations were used to show what effects HPI flow rate changes would have on the transient. Flow rates were altered to correspond with the flows expected from 1, 2, and 3 HPI pump operation, and at different times in the transient. Figures 16 and 17 provide a graphical representation of the results.
The calculational technique used for this analysis assumed a semi-steady state system response. The conservation of mass and energy equations p.rovide an iterative process through which HP1 flow rate changes alter the system pressure response. Referring to Figures 16 and 17, it is seen that these changes in HPI flow rates effect the system repressurization and refill rates. For this.01 f t.2 break, and the HP1 flow rates used, the repressurization rates indicate the inability of the primary system to relieve decay energy.
As HP1 flow increases, the repressurization rate is less severe. This is due primarily to the cold water injection decreasing core steam production. The trade-off between break flow, ECC injection,- steam production and condensation results in the particular repressurization rates shown in Figure 16. Figure 17 illustrates the system refilling response as a function of HP1 flow rate.
It is noted here that tne increase in mixture level at 650 to SCO sec.
associated 'with the one HPI transient is not due to refill. But, rather, the loss of natural circulation causes a swell to occur due to the increase in void fraction from steam buildup within the hot leg.
After 800 sec., the' steam is being released from the liquid via the bubble velocity, and a decrease in mixture level results.
The same general e
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trend may be seen with 2 and 3 HPI operation, but should not be as severe. Additional model and technique modification would be necessary to accurately predict this phenomenon.
The 2 and 3 HPI results shown on Figure 17 are the trends associated with system refill as the flow associated with these HPI quantities are enough to overcome the break fl ow.
As is expected, more HPI flow results in a faster system refill rate. As ECC injection continues, a mass accumulation occurs within the RCS, and should be indicated by the RITS.
At some point in time, system recovery will be dominated by a depressurization due to the establisnment of single phase natural circulation or the boiler-condensor mode.
From this analysis, the RI[S is seen as.very useful.
As the hot leg level decreases, the operator has a confirmation of a LOCA.
The operator can then confirm that actions taken upon loss of saturation conditions, i.e. HPI maximization, will ensure that RC5 stabilization, refill, and long term cooling will occur within a shortened time frame.
e
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20.
l Figure is RC PRESSURE VS TIME FOR VARYING HPI FLOW RATES 2
. (.01 FT PUMP DISCHARGE BREAK) 2200 2000 1 HPI 1800
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1600 E
l 0
500 1000 1500 2000 2500 3000 Time, Sec d
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2 Figure 17.
HOT LEG NIXTURE LEVEL RESPONSE.01 FT BREAK, CLPD, VARYING llPI FLOW 50
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VII. OPERATOR GUIDANCE A.
Inadequate Core Cooling The present NSS instrumentation provides adequate information to allow the operator to respond to transients. Under cost transient conditions, the primary system would remain subcooled, as measured by the hot leg RTD's, and a measure of the primary systen inventory would be provided by the pressurizer level instrumentation. However, under accident conditions, such as a small break LOCA, the primary system could progress to saturated conditions.
Under saturated RC fluid conditions, the pressurizer level ceases to be a reliaole indication of the primary system inventory. For a S3 LOCA, saturated conditions would occur prior to any possible core uncovery and would last until an actual inadequate core cooling condition developed.
The core exit thermoccuple indications are used to determine the temperature at the exit of the core region. These thermoccupies will indicate above saturation temperatures if the core liquid inventory is insufficient to provide adequate core cooling. Thus, the core exit then: occupies are a direct indication of inadequate core cooling and an indirect indication that the water inventory is below the top of the core region.
Under current operator guidelines described in " Abnormal Transient Operating Guidelines", ATOG, (Reference 4) tripping the RC pumps is the first preventive action required for a loss of subcooling margin. The operator then confirms two actuated HPIs at maximun capacity and a balanced HPI flow. Next the operator raises the steam generator water level to 95t on the operating range.
The water level is raised using EF4 in e
2L
e
> u_.
a continuous and controlled manner as described in the "Best Methods For Equipment Operation" chapter of Reference 4.
The last action, which is to attempt to locate and isolate the break is suggested whenever a SB LOCA is suspected.
The RITS would provide the operator with additional information relating to the primary system inventory under saturated fluid conditions.
As illustrated in Figure 1, the hot leg piping is the high point of the RC system. Thus, steam or non-condensible gas generated within the primary system will ultimately collect at this high point.
The proportion of gas to steam will be very small until such time that excessive fuel temperatures are reached.
Steam or gas generated in the RV will vent itself to the hot legs before the core is in danger of uncovery because the RY hot leg no::les are about 3.5 feet above the core region.
A level measured in the not leg will provide the operator with indication of the primary system inventory trends prior to any indication of the loss of inventory by the core exit thermocouples.
Based on the review of SB LOCA and non-LOCA transients, the initial system responses are similar. For both these types of events, hot leg voiding and possible loss of natural circulation will occur, accompanied by a decrease in the RITS indications. As the RITS instrument decreases off scale low, the operator will be alerted to appreciable inventory loss associated witn a LOCA and the potential for core uncovery.
l The present instrumentation will indicate that the plant is in an abnorma'l configuration (saturated conditions). However, there is no indication of potential core uncovery beyond the fact that the RCS is at saturation conditions. When the core exit thermocouples are
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indicating above saturation temperatures, the operator is alerted that inadequate core cooling is in progress.
The RITS will provide warning and trending information, and will aid the operator in measuring both the approach to core uncovery and the effectiveness of the systems being used for inventory makeup. Using this infonaation, the operator can then take confirmatory action regarding inventory, replacement and steam generator cooling.
Presently, the ICC guidelines instruct the operator to take these confirmatory actions, but only after the core exit thermocouples indicate superheated conditions. With the RITS available, the operator could begin these confirmatory actions (discussed below) with the RCS still at saturation, and prior to the core exit thermocouplcs indicating superheated conditions.
Upon loss of level indication on the RITS instrument, the operator should examine the performance of the ECCS. The HPI flow actually being delivered should be compared to the perfonnance curves utilized in the ECCS evaluations to assure that adequate inventory makeup is being provided to the RCS.
If the operator finds that inadequate injection is being provided, he then would know that the probability of an ICC situation developing is significant.
The operator could then maximize ECC flow and decrease the rate of inventory loss by depressurizing the SG at the maxiaum rate allowed by the tube-to-shell cooldewn limits. The success of tnese actions could be monitored by the RITS instrument. The inventory trend of the hot leg, as indicated by the RITS'
, wc ul d supply feedback to the operator on the effect of these actions to in_creas4 ECC flow and decrease inventory loss. Under SB LOCA situations with the RC pumps tripped, inadequate core cooling will
~
not occur prior to the RCS' Inventory Trending System,
- 2 3-h
going off-Scale 10a.
't may tnus be possible to avoid superheat2d and ICC conditions altogether cy confirming these actions c IEs cs /c be and not waiting until when the RITS is the core exit thennocouples indicate superheated conditions.
It is not intended that RIlb indications be used as a basis for action once severe supernesting concitions develep within the RCS which would involve such operator actions as starting the RC pumps with the RCS in a highly voided condition, or rapid st'can generator depressurization. The requirement for such actions will continue to be based on the core exit thermocouple indications.
The real value of the RITS will be to provide the operator with a better opportunity to avoid these circumstances.
Thus, the RITS provides the operator with earlier warning of the potential for ICC.
The 81TS would indicate inventory trends allowing the operator to m nitor the effectiveness of actions taken to avoid ICC.
At an appropriate time, these RITS guidelines, as discussed herein, could be integratec into the ATOG procedures on a plant speci fic basis.
VIII. CONCLUSIONS RITS will De a determining parameter in the LCCA vs. non-LCCA decision process.
Given evidence of a LOCA transient, the operater will be forewarned of, and will then be able to take confirmatory actions to prevent, a possible ICC event.
Particular operator actions involve HP1 maximization and steem generator cepressurization.
The RCS Inventdry Trending System would provide valuable feedback to the operator concerning inventory conditions and the effectiveness of ECC injection to mitigate the transient. The d/p Instrumentation
- 2Le
~ ~~
~ ~ ~ ~ " ~
e
. e e
useful while the RC pumps or the high point vents are in is not opera tion.
A reasonable response time is necessary to inform the operator of inventory conditions. As response time becomes large (i.e.
minutes) the primary function of the RITS, that of forewarning the operator of ICC potential, is compromised. However, the complete perspective is that the RCS Inventory' Trending ' System is to be used as an additional device that will confirm existing plant information.
The RITS is designed to aid in the determination of RCS conditions by being coupled with information from other plant indications such as primary and secondary temperatures and pressures.
It is not to be used as a sole indicator of RCS conditions'.
The saturation meter, the >RITS and core exit thermocouples (exceeding saturation temperature) satisfy the requirement for advanced warning of the potential for ICC at various stages of the accident.
The, full range requirement is provided for by the RITS (top of hot leg to top of coreT
~
and the core exit thermocouples (top of the core' to complete core uncovery). The requirement that the ICC detection instrument system be unambiguous is satisfied by the core exit thermocouples, since the actual high temperature in the fuel region is the only direct measure of whether or not inadequate core cooling is occurring.
The principal requirements of NUREG-0737 and how the suggested ICC detection system fulfills these requirements is summarized in the following table.
NR 0737 Saturation Hot Leg,,
Core Exit Head Recuirements Meter RITS.
RITS.
Thermocouoles Advance 'Jarning X
X X
X Full Range X
X X
Unambiguous x
X G
h
e Operator actions are required based on indications from the saturation meter and core exit thermocouple temperatures to preclude the onset of ICC. There are no direct actions that the operator would take based solely on RITS.
However, if the inventory trend is down (level decreasing) the operator would verify that actions to be taken at saturation conditions were taken and th'at these actions had the expected results (e.g. actual HPI flow indicated after starting an HPI pump). The role of the RITS is to provide the operator information on the progress of the transient and confirm that the actions taken to preclude ICC are actually initiated.
Taken in conjunction, the saturation meter. RITS, and core exit thermocouples provide advanced warning and an unambiguous full range ICC detection system.
i I
d
.21
' t ",' h. -
IX. REFERENCES 1.
Calculation File, "Effect of RC Pump Trip on non-LOCA Overcooling Events", B&W Occument #32-1103341-02, 9/19/79.
2.
Design Memo, " Appendix A - Overcooling Results/ Reply to Denton's 50.54(f) Show Cause, Parts A-B (177 FA Plants)",
B&W Document #86-1106752-03, 3/18/80.
3.
B&W " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant", Volume 1, Section 6, May 7, 1979.
' 4.
" Abnormal Transient Operating Guidelines" - ATCG, Part II, Volume 2,
" Discussion of Selected Transients" (DRAFT).
5.
Deleted 6.
Letter, J. H. Taylor to S. A. Varga, dated July 18, 1978, Subject, SB LOCA Spectrum Analyses.
7.
" Evaluation of Instrumentation to Detect Inadequate Core Cooling",
prepared for 177 Owners Group, August 15, 1980, B&W Document No.
86-1120838-00.
6 O
9 ti.
ENCLOSURE 2 INADEQUATE CORE COOLING SYSTEM EVALUATION FOR TMI-l RC FUMP MOTOR CURRENT
1.0 YNTRODUCTf0N 1.
Backoround Following the TMI-2 accident, the capability to monitor the primary i
system water inventory was identified as a potentially useful accident management tool.
Review of TMI-2 data and subsequent small break loss 1
of coolant accident (SB LOCA) analyses,2 has revealed that continuous operation of the RC pumps during the transient resulted in a highly i
voided primary system.
When the RC pumps were tripped in this condition, the liquid that was previously dispersed throughout the system via pumping action, collapsed to the low points of the primary system, such as the bottom of the reactor vessel and steam generators.
Consequently,
'[
the low water inventory at the time of the pump trip could result i.n an insufficient level for adequate core cooling.
N
/
In the fall of 1979, the NRC made a generic assessment of delayed RC pump trips during a SB LOCA.3 They concluded that due to the uncertainties involved in SB LOCA analysis the prudent course of action would be to trip the RC pumps immediately following an indication that a LOCA had occurred (RC pressure dropping below the HPI setpoint).
It was also a
recognized that the immediate pump trip approach was less than optimum.
h
_ For example, in overcooling events (i.e., steam line break) which cause shrinkage of the primary system and loss of RC pressure, early RC pump trip (and subsequent loss of pressurizer spray) can aggravate these transients and extend the time required to bring the plant into a controlled shutdown conditon.
However, since these transients did not lead to unacceptable consequences, early pump trip was adopted as a course of action..-
2.0 PUMP CURRENT Two mathematical models can be developed which relate pump current to RCS voiding.
Each model is famed using an expression for the energy transference from the pump to the coolant.
The input three-phase power, P, required to ' drive a constant speed (constant m
frequency) squirrel cage induction motor can be expressed as follows:
= [ IV (PF)
P s
where
[ = accounts for the 3-phase input of power I
= line current (RMS Amps)
V
= line voltage (RMS volts)
PF = power factor (accounts for energy lost in setting up the magnetic field)
Similarly, the power, P, required to drive a pump can be expressed by considering p
the development of head:
e0H p
P n
P 3
where p
= fluid density (lb/ft )
Q
= volumetric flow (ft"/sec)
H
= head generated by the pump (ft)
~' '
= overall pump efficiency (accounts for the mechanical np friction at the seals and bearings, hydraulic friction at the impeller vanes and the diffuser-vanes, and various other hydraulic losses due
- w eddy femation.
2-
If one accounts for the windage and ' mechanical friction losses of the motor by considering the motor efficiency, g, the transfer of power to the pump can be stated.
P
%,=P p
or, inserting the previous definitions, gkIV(PF)=fH P
ine cnange in current relative to varing fluid density can be addressed by denoting a reference point (0 superscript) condition.
UU IV (PF) Q H n m OO p
n o n,o I V (PF)0 QH p
Assuming that the. motor efficiency, voltage, power ~ factor, capacity, and head remain constant, a simple relationship results.
- (h) { a) o p
Inserting an expression relating void to density, o = of - c (of - o )
g will yield the desired relationship between voiding and current..
3 Of-0o ( y) (U a)
~
I p
n 9 a=
- f ~ *g A second codel can be developed by considering the transference of torque within the pump.
In this situation, P = Ta p
'I 6
where T = hydraulic torque transferred from the impeller to the fluid (ft-lb )
f n = pump speed (rpm)
'I = Pump efficiency accounting for the mechanical friction at the seals and bearings, hydraulic friction and eddy losses within the impeller.
Hydraulic friction and eddy losses within the diffuser vanes are not included since only power transfer at the impeller is being considered.
.4-
'i
~
Coupling equation with t;fte homologous or.nomalized relation for torque, 1
T
" 8 ( #PR) -
- yields, 1
P
= ST PU/"I R f P
R p
Where S = normalized torque I
T=
ated torque.fft-lb )
R f
3 P = density corresponding to TR (lb/ft )
R i
Assuming that the motor effigiency, voltage, power factor, and operating point of the pump (s,n) remaio constant, a nomalized relationship results.
- ( b)
I "
(ny );c o'o I o o
As before,
( iu -) ( r a) of -o a=
PfPg This completes the developmen.t of the mathematical models.
The four
.,. relationships involving genti.lle tap and void fraction, and pump current and void fraction are shown in-Table 1.
TABLE 1:
VOIDING MODELS
-' ~~
Pumo Current of - p (I/I ) (n /n o) p p Model 1:
a=
s
)
(I/I ) (nio pf -p Model 2:
a=
of _ og l
l
.s l
l?
'i l
l i 1
In summary, these models were based on the followino assumotions:
Pu:co Current, Model 1 1.
The motor efficiency, voltage, and power factor remain constant.
2.
The pump capacity and head do not degrade.
Pumo Current, Model 2 1.
The motor efficiency, voltage, and power factor remain constant.
2.
Pump speed remains constant.
3.
The hydraulic torque does not degrade.
The validity of these assumptions will be discussed further in Section 3.0 when the supporting data is presented.
i f
l l
. {
~'
3.0 SUPPOP, TING DATA To utilize the models developed in the last section, supporting data must be found to substantiate the assumptions used to develop the models.
Consequently, a literature survey was conducted.
This section summarizes the findings of this study.
Figure 3 displays the operating characteristics of a typical pump drive motor. Considering the density variation associated with voiding to be on the order of 0 to 40%, the operating range of interest lies between 6000 and 10000 HP.
In this range, when a constant vol-tage is supplied (6600 volts in this case), motor efficiency and power factor vary by less than ' %.
As a result, current varies g
linearly with power in this range as shown in figure 3.
The variations in pump speed are'small as ?oad changes. Typi cally,,
l the incuction motor can generate 1000 HP for every rpm off synchronous.
l Therefore, the pump speed can be assumed to be essentially constan:
when compared to the 1200 rpm synchronous motor speed.
l The assumptions of constant motor efficiency and power factor imply constant supply voltage.
It is recognized that off normal voltages I
may occur due to bus transfers, pump starts or grid disturbances.
These perturbations are expected to be of short duration and would i
i not have a significent impact on the use of pump current for inventory trending indication.
Application of this measure' ment for alarm or pump trip circuits would recuire provisions for delaying the alarm or trip signal to avoid spurious actuations.
~
Operating characteristics of the pump drive motor in the range of interest can be summarized as follows:
1.
The, motor efficiency remains constant.
2.
The power factor remains constant.
3.
The pump speed remains constant.
Experimental data has been generated concerning the performance of the pumps under two-phase conditions. These experimental studies are summarized below.
Combustion Engineering, Inc./EPRI In an effort to refine the analytical model of the reactor coolant pumps under hypothetical iarge break LOCA concitions, Combustion Engineering constructed a test system which utilized a 1/5 scale pump.
Steady state tests were conducted near the c;:erating point (v /'n = 1.0) and the results are presented in Figures 4 and 5.
n (Note v is defined as the ratio of the actual flow rate to the n
rated flow rate and e is the ratio of the actual speed to rated n
speed.)
For low void fraction (0-0.1), the pump head (Figure 4) remained at, or slightly above the water or non-degraded value.
At 0.15 void, the head degrades almost linearly until the worst t
I case (20% of the water value) is reached at.75 void.
I*.- -
j Head then recovers as the single phase steam region is reached.
!a A similar behavior is seen concerning torque, (see Figure 5),
l j !
but the degradation is less severe. The degradation in head and torque at low pressures is even more pronounced due to the 4
i 1arger density differences between the steam and the liquid.
l Losses associated with these two-phase mechanisms are three 4
I times greater at 500 psia than at 1000 psia.
Babcock & Wilcox Comoanv/EPRI
,i i,
With the same program objectives, B&W constructed a test apparatus 5
to analyze the performance of a 1/3 scale pump using air / water e
mixtures to simulate voiding. The results of these experiments are presented in Figures 6 and 7.
Restricting our attention to the re-gion around the operating point (v /*n = 1.0), the same degradation n
behavior is seen relative to the Combustion Engineering data. Hewever, the degradation effects are exaggerated due to two-phase losses re-sulting from the large density difference between air and water.
CREARE Inc./EPRI CREARE Inc. worked in parallel with the two studies mentioned above.
In their test rig, a 1/20 scale pump was installed to address the effects of scaling.
Figures 8 and 9 show the results of the tests conducted near the operating point for low pressure water / air mix-tures.
These results compare quite favorably with the B&W data presented earlier.
LOFT Research As part of the Loss-of-Fluid Test (LOFT) Program, RC pump motor pcwer and current meesurements, and their utiTity as indicators of RCS l.
. ~
inventory is being explored.
Current research shows a large head degradation at approximately 20% void (,See Figure 10). This phenomena is consistent with the other data presented thus far for scaled pumps.
It is interesting to note, that power and current measurements made during this transient (See Figures 11 and 12, respectively) do not exhibit the same discontinuity.
I The following can be summarized concerning two-phase degradation of pumps:
1.
In all cases, torque did not degrade as appreciably as develop-ed head. This implies that larger losses occur in the diffuser section of the pump than in the impeller section. The LOFT data presented in Figures 10 through 12 shows that this is tne case, with the impeller efficiently imparting hydraulic torque to a two-phase fluid at a homogeneous density, while the head abruptly degrades due to two-phase losses in the diffuser sec-tion.
2.
All degradation effects are minimized at pressures above 1000 psia. This is important since the application of the void measurement will usually be made at elevated pressures (> 1000 psia).
3.
Review of the RCS flow for the TMI-2 event reveals that little if any pump capacity degradation occurred.
Therefore, it must be concluded that scaling effects have not been adequately adcressed j
by the pump testing summarized in this study. The probable cause for this discrepancy deals with the inability of these tests to scale the bubbles such that the relationship bet'. keen the si:e of the bubbles with respect to the vane spacing on the impeller and diffuser is preserved.
For example, if ping-pong balls were -. _ _ - -
suspended in a fluid system, their. presence would choke a small pump, whereas a large pump would pass the balls virtually unde-tected.
4.
In developing the technical approach for the pump current meas-urement, model 2 (based on torque transfer) is a stronger approach l
since only impeller dynamics are involved. Model 1 (based en head and capacity) would involve resolving the energy losses in the pump casing, thus making this approach less desirable. There-I fore, model 1 will not be considered further.
1 I
e --. - -.
],
__FZGURE 3: [10 TOR CHARACTER 15 TICS 800
- ': I 700
+i '
Efficiency g4
/
,i 92 ij, 600 Po er Factor u
90 B
UO 88 500 I
uo Curren*
n r
c.
86
=-
g 3
84 400 "g
x E
s o
=
i G
82 O
C 1
80
'j
-- 300 u
Cv.
v 1
s.
78 i
c 76 200 1
1 03 0
2000 4000 6000 8000 10000 12000 Load (HP) 9
..-r--e,_
n.--
a i
i 1
Y i w-- A,1 i
i l
N-
'N,,
N N
N N
X N.
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\\.
i.
1 m
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c.
iiii.
o.
=
G i
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~
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=
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=
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=
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,- 6 m E
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=
.z z
=
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~:
=
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=
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e e e m e <
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ll
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m
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d 6
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.1
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==
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<=
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=
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g o e e e e e, y p
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s w.
e o e =
3
e g
c
~
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.=
Q O < M k
m?
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=
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l c
m u
l o/
1 2
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/
3
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+
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ci 6
a ci 6
6 m
c 6
3313180 E3d A11SN30 30VE3AY ($D / " 6 )
E E
011YE 300E0150000'lo.40H I
n I.e 1-FIGURE 6:
Hemologous Head at Varicus P~~u ha e tions for First-Quadrant Pump Cpe 1.5 i
i 1.4 N
i i
%*=%'w 4
weae Fracuoas j
%N h
i II N ~ ---
(c.n i
N w
(3 1C) i f
0
- A %
i
, ' %.e(20 15) l (0 3) 9 l
8
,e, 3 10 (20 20) 0.8
- e"~
j (n.=>.
n e.-
I l
e.*"-
sm
.a4 La
'""p. 30 ;
20 (20 3*)
AZ i
k (30 40s
" = = ssw,r.
~b
- ~
.#40 5:3 (30 50) jg 7.0
// ~
~
m-.-----
5 150 6 :
m C
j C1 C2 C3 04 c.5 60 ? )
(50 60) s (607C)\\
e
.c,2 h (ro..e.
j
\\
s I
\\
-C 4 e-s
-Q 6 (40 50)
(30 40)
- 0. 4 ',.
(3 10) f (20. sci 1C (0-1)
(10 20)
.l.2L 14 Pump-Speed Flow F. ate, a fy or y,73
T I
I I
FIGURE 7:
Homologous Tc.g e at Various Pep Inlet Void Fractions for First-Quadrant Pu=p Cperation 1.4 1.2 voto Fractaans t 3 to 13 3 410 to 1333 1.0
(%/ Alena)
"M
,T
.0 to 20% )
- " " ~ ~ ~ "" _ _,,,, - ""
- """", e T 3 to 10s 3 O to 33
.""'"~-.~#
0.8 to to 33) e QS to 20s) n. -- - - -
0.6 g
t, m, 1
N 2:
,.a
'O 18 C ~
f 0.4 460 to 70%)
/
\\
4s o tSC to 60%1
~
6 C
l n 2" 0.2.
-- " ~ ~ -
^IO "* E '
k e ~~, o #
c, t30 to SCE)
(50 to 701) i 2
- /
0
=
C 0.1 0.2
- 0. 3,
.. S 0.6 07 C.3 0.3 1.0
/
[
(50 to 60%)
\\
(60 to 2 3 "2
0'2 r N
2, t70 to 9C1) 5 t Alena/N) 0aw i
O=
(3 to lui
-0.6 (a0 to $0s) 0.8 (20 to 33 )
i to to I:1 1.0 l
(30 to act)
- 1. 2,
(10 to 23 )
l 1.4 Pump-Speed Flow Rate, 3 /v Or v /c i
i 1
I
t Z
I
~
O j
1-i O
l CE ok i
i o
~
O
(
G o
o>
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mo>
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r i
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H O C O N q.Ja = 4< g La CD C.::
Li.
i i.
e
1 1
I i
i i
i 79 k Saturation (2-phase ficw at pump inlet)
=
a (D
6 2-Phase bubbly flow across pump
'40=
Al
"~
ew; e
5 rg a
li d h }
Transitien, bubbly-to-churn
/
turbul ent il p:l #
i.
q 3e 3
V Churn turbulent flow
_ Bubbly ficw c
~
- Pu D
2a 5
' N' C
- Uncertainty ; 0._3 ps1
- - - ~
c gg l
I i
i t
I r
9 a
tee 22a 3m e
Sm 523 7c3 Time (s) i Figure 10: Differential Pressure Across Primary Ceclan: Pumps O
1 e
~
~ ~ - ^ -
sa l
1 1
I i
i i
I toy Cata Uncertain y - 515m/f:* _
'l ld.
R a
- 7
.5 as Data
?%
3a hi %
I 5
- Prh,
./"%b.9g y7 9
s-Calculation 1 N== - *.
3 1
I I
I I
I i
is 8
S3, 123 1'a 222 22 232 Za G
Time (s)
Figure 1). Measured Density Upstream of Puma Compared with Predicticn Using Pump Mo cr Pcwer Assuming Ccnstan: Volumetric Ficw
~
and Pump Head 53 I
I I
i l
i as b.,
Data uncertaioty M.
t 5 lbm/f 4
4 3J1 a.
se Ng A
-4 25
.s.
c-Cata m
V c
E
-Q
'M Aj 1
Calculation Q A g t 9 g< k,
s v t a_
% *1
&P' I
I I
I I
i 1
Es a
- a im
- a 233 Zsa
- na za e
Time (s)
FIGUTtE 1;,
Measured Censity U: stream of Dumo Compared With Predictica Using Pumo Mctor Curren Assuming Constan:
Volumetric Flow and Pump Head
- - -* ~ ~: --
~ ~,,
j..
. * - '.. - ~~ _
4.0 p0TEllTIAL USE OF INVENTORY MEASUREMENT TECHf1IOUES IN PUliP TRIP CIRCUITRY As explained in Section 1.0, the level of RCS voiding must be i
known to determine an optimum pump trip time in the event that the primary system pressure drops below the HPI setpoint. To use the proposed models and their associated voiding signals in y
pump trip circuitry, the following criterion were used:
1.
The trip should be delayed sufficiently to allow the HPI to make up the inventory lost for the most probable a
small breaks.
2.
The trip should be based on a void fraction sufficiently large to be outside of the normal noise band associated with a = 0 line.
3.
The trip void fraction should be sufficiently small such that the trip will occur before pump degradation becomes signi ficant.
Pump current / power tests at LOFT have shown that a setpoint of.15 (15%) void will meet the above criterion.
I This higher 1
setpoint can be justified since SBLOCA analysis indicates that a system can proceed to approximately 0.4 void (40%) before the pumps must be tripped.
In any event, either measurement can provide l
a conservative indication of void which will yield acceptable pump I
2.
- 1. -
i trip times for small breaks in shich the HPI cannot keep up.
When breaks in which the HPI can makeup the inventory being lost or when overcooling events occur, an unnecessary pump trip will be avoided and plant shutdcen will be made more manageable.
l l
A conceptual design for a void measurement system using pump current would require two inputs:
1.
Pump current measurement 2.
RCS pressure from the wide range taps located on the hot legs l
The current measurement will be used as the time-varying I input and the RCS pressure measurement will be used to estimate the suction pressure at the RC pumps, such that a calculating module can determine the saturation properties cf and o.. The calibration constants I and
^
o will be recorded e -.
REFERENCES j
1.
" Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant," Babcock & Wilcox Co.,
liay 7,1979.
I, I
2.
" Analysis Summary in Support of an Early RC Pump Trip," and
" Supplemental Small Break Analysis," submitted as Attachment B to i
B&W plant Owners Letters in response to ilRC Bulletin 79-05C.
l I
3.
" Generic Assessment of Delayed Reactor Coolant Pump Trip During Small Break Loss of Coolant Accidents in Pressurized Water Reactors",
NUREG-0623, flovember,1979.
Y.
" Reactor Coolant Pump Motor Power or Current Criteria for Reactor Coolant System Inventory Management in Comt.ercial PWR's During Accidents," LOFT Research llemorandum, ilarch,1982.
i G
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APPENDIX 1
E.
F
{l o give the reader a qualitative " feel" for the model, a simple computer code, F
10IOCCM, was developed.
VOIDCOM model the RCS as a single homogeneous thermo-r dynamic node.
Within this node, conservation of mass is determined by considering the addition of mass due to the HPI system and the loss of mass out of a small treak.
Conservation of energy is maintained by summing the energy contributions I,
to the RCS via the core, steam generators, HPI system, and the break.
By considering the RCS as a fixed volume system, the pressure of this large node can then be determined using the conservation of mass and energy principles.
A i
somewhat pseudo primary loop containing a RC pump is then sucerimposed within the RCS thermodynamic node.
The mathematical modeling of the the RC pump is of sufficient detail to allow the data presented in Section 3 to be utilized. The output of these component models is a simulated signal of current for the RC pump. The voiding model presented in Section 2 is then applied to yield an j
estimate of system voiding (via pump current) which can be compared j
to the thermodynamic voiding of the system which is known.
a f 1 i t
I e
l e
l t
B e
e.;
i A TMI-2 type scenario was used to create a voiding transient upon which y
the models and data could be applied. The chain of events are as l
follows: (See Figures 13-18)
I 1.
At time zero, the reactor is at 100". power.
2.
Loss of all feedwater begins just after time zero.
3.
As the steam generators provide less of a heat sink to the primary system, the RCS pressure rises until the PORV lifts. (t = 15 seconds)
)
4.
The FDRV sticks open and water continues to flow out of the break.
i' 5.
As the RCS pressure drops below the HPI setpoint, the HPI flow is throttled off. (This is done in this example to create significant voiding in a small period of time.)
[
6.
Auxiliary feedwater is valved in and reaches full capacity (t = 4 minutes) 7.
At.20 void, HPI ficw is reinstated at full capacity (t = 40 minutes')
': l 8.
System inventory then recovers and the RCS returns to a z i subcooled condition. (t = 56.5 minutes)
(
l The transient mentioned was analyzed twice, once assuming non-decraded pump performance, and then a second time assuming that the pump would experience scme degradation.
In the second ar.alysis, a pump head and torque degradation multiplier was applied which approximates the data l
presented in Section 3 for the Combustion Engineerina data at pressures around 1000 psia.
The expressions which were used for the head degradation multiplier, R, are as follows:
d l
K = 1.0 0.0 5a5 0.14 d
R = i.4242 - 3.03 a 0.14 < a < 0.25 g -,.
- y L.
?
I Similarly, for the torque degradation multiplier, l'.T the expressions
- are, M = 1.0 0.0 5 a 5 0.10 i
T M = 1.2 - 2.0 a 0.10 < a < 0.25 T
Note the expressions for the multipliers were limited to the range 0
- a < 0.25 due to the dynamics of the problem analyzed.
Review of figures 13 through 10 gives the raader a quantitative feel for the model presented,and the impact that the pump degradation can have on the resulting estimate of system voiding.
With these figures, comparisons can be made which allow the uncer-l tainties associated with each measurement technique to be estimated.
1 f
In figure 13, the transient behavior of 'the RCS system pressure is shown for both the degraded and non-degraded cases.
Due to the i
simplistic nature of the VOIDCOM thermodynamic model, the system pressure is identical for both cases. When a loss of all feedwater
- occurs, the pressure in the RCS increases from the steady state value until the PORV lifts and sticks open.
The pressure then drcps until it settles out to a saturation pressure determined by the heat sink provided by the steam generators. At approximately 56.5 minutes, the HPI flow, being greater than the leak flow, has brought the RCS back into a solid condition. The pressure then rises until a pressure level is reached where the leak flow becomes equal to the HPI ca:acity
,).j; (p : 2050 psia).
-1 r,-
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h$
- , l...,
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The loop flow rate versus time is shown in figure 14.
In the non-degraded case, the reduction in flow is due to changes in system density only (equation 1 Section 2). By comparison, when the system void exceeds
.14, the flow rate is further reduced due to the now degraded pump head (Figure 15). The degradation in flow rate is not large relative to the loss in developed pump head because it is assumed that the flow rate varies as the' square root of the t,P across the pump.
, I.
I In figure 16, the transient behavior of the hydraulic torque is shown.
Wnen non-degraded behavior is assumed, the torque varies as a function of density only. Under degrr.ded conditions ( a >.10), the transfer of torque becomes less efficient as void increases.
i a The void estimates predicted using pump current models assuming non-degraded pump performance are ccmpared to the actual thermodynamic void in figure 17.
It can be seen that both models produce good estimates of system voiding.
Figure 13 depicts the same estimates assuming pump degradation Occurs.
For the pump current method, the degraded pump performances cause the predicted voiding to be overstated.
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e LIST OF.SYb30LS SYMBOL DEFINITION A
Area g
gravitation constant c
H pump head I
current K
flow coefficient t.P pressure differential PF power factor P
motor power m
p pump power Q
volumetric capacity T
pump torque
?
velocity V
voltage specific volume v
W mass flow rate quality x
Y net expansion factor void coefficient a
normalized pump speed a
n normalized pump torque S
pump impeller efficiency ny motor efficiency nm overall puma efficiency np a
pump speed density o
normalized pump flow v
Subscripts or Superscripts i
G gas phase L
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saturated gas 9
reference or calibrated state r,o
5 6
4 ENCLOSURE 3 NUREG 0737 EVALUATION OF ICC SYSTDi FOR TMI-1
Checklist e
i
'fer Plant Spscific Review cf Inadequate Core Cooling (ICC) Instrumentation System For Three Mile Island Nuclear Station Unit 1 Docket No.
50-289 Operated by:
GPU Nuclear
\\
Reference Deviations Schedule 1.
Description of the proposed final system including:
- a. a final design description of i
additional instrumentation and dispisys; Encl. 3 Encl. 4 (RITS)
- b. detailed description of existing l
instrumentation systems.
Encl. 3
- c. description of completed or planned modifications.
Encl. 1, 2, & 3 2.
A design analysia and evaluation of inventory trend instrumentation, and test data to support design in iten 1.
Encl. 1 & 2 3.
Description of tests planned and results of tests completed for evaluation,' qualification, and calibration of additional instru-mentation.
Encl. 3 Enc);,4 4.
Provide a table or description covering the evaluation of con-formance with NUREG-0737: II.P.2, 2nd Appendix 3 (to be reviewed on a plant specific Encl. 3 basis) 5.
Describe computer, software and j
l display functions associated with Encl. 1 & 2 See Encl. 4 l
ICC monitoring in the plant.-
I' 6.
Provida a proposed schedule for installation, testing and cali-bration and i=plementation of any l
proposed new instru=entation or See Encl. 4 info:. arion displays.
7.
Describe guidelines for use of reactor coolant inventory tracking system, and analyses used to develop procedures.
See Encl. 1 See Encl. 4 8.
Operator instructions in emergency operating procedures for ICC and how these procedures vill be modified when final monitoring system is i=plemented.
See Encl. 4 9.
Provide a schedtle for additional
- b=ittals requirad**
See Encl. 4.
l Appendix 3 (of NInuMH)737, 'II.y.2)
Confirm explicitly the conformance to the Appendiz 3 ite=a listed below for the ICC instrumentation, i.e., the SMM, the reactor coolant inventory racking system, the core exit thernocouples and the display systems.
Reference Deviations 1.
Environmental qualification Encl. 3 RCPMotorPower/SMM(Seisnic) 2.
Single failure analysis Encl. 3 RITS Taps
, 3.
Class IE power source Encl. 3 RCP Motor Power 4.
Availability prior to an accident Encl. 3 NONE 5.
Quality Assuranca Encl. 3 NONE 6.
Continuous indications Encl. 3 NONE 7.
Recording of instrument outputs Encl. 3 NONE 8.
Identification of instruments Encl. 3 NONE 9
Isolation Encl. 3 NONE
t Inadequate Core Cooling Instrumentation Report Item (1)
A description of the proposed final system incluaing:
(a) a final design description of additional instrumentation and displays;
Response
The final ICC system is composed of the following elements:
1.
The core exit thermocouple system which is described in the TMI-1 Restart Report Sections 2.1.1.6.3.1, 2.1.1.6.4.1 and question 95 of supplement 1 part 2.
Further information is supplied in GPUN letter dated February 2,198 2.
2.
The saturation margin meter which is described in the TMI-l Restart Report Sections 2.1.1.6.3.3, 2.1.1.6.4.3 and question 20 of Supplement 1 part 1.
3.
The vide range Ih meter which is described in th TMI-1 Restart Report Sections 2.1.1.6.3.2 and 2.1.1.6.4.3.
4 Other instrumentation which would be useful in monitoring the approach to ICC include Tave, RCS flow meters, source range instrumentation, RCS pressure and pressurizer level which are all described in Chapter 8 of the TMI-l FSAR.
5.
See Enclosure 4 schedule of additional instrumentation.
Item (b) a detailed description of existing instrumenta tion systems (e.g.,
subcooling meters and incore thermocouples), including parameter ranges and displays, which pr ov id e operating information pertinent to ICC considerations; and
Response
See response to 1(a)1 and 2 above.
a Item (c) a description of any planned modifications to the instrumentation systems in item 1.b above.
Respon e:
See Enclosures 1, 2 and 4.
Item (2)
The necessary design-analysis, including evaluation of various instruments to monitor water level, and available test data to support the design,
described in item 1 above.
Response
Our evaluation of various instruments to monitor water level was described in the Restart Report Supplement 1 Part 2 Question 95 and to Mr. R.
Jacobs dated December 18, 1981 (Dhir Report), August 26, 1981 and November 13, 1981.
See Enclosures 1 and 2 for further information.
Item (3)
A description of additional test programs to be conducted for evaluation, qualification, and calibration of additional instrumentation.
Response
See Enclosures 1, 2 and 4.
Item (4)
An evaluation, including proposed actions, on the conformance of the ICC instrument system to this document, including Attachment I and Appendix A.
Any deviations should be justified.
Response
a.
Core Exit Thermocouples instrumentation was decribed in our letter of February 2, 1982.
b.
Saturation Margin Monitor - see Attachment 1.
c.
Additional Instrumentation - see Attachment 2.
Item (5)
A description of the computer functions associated with ICC monitoring and functional specifications for relevant sof tware in the process computer and other pertinent calculators.
The reliability of nonredundant computers used in the system should be addressed.
Response
The P/T plot and associated computer functions are described in attach =ent 2 of our letter of February 2, 1982 (82-007).
The reliability of this computer system is also discussed in that letter.
Further information on computer function display will be provided in accordance with the schedule of enclosure 4.
Item (6)
A current schedule, including contingencies, for installation, testing and calibration, and implementation of any proposed new instrumentation or information displays.
Response
Ins trumentation described in la have been installed. For the RITS see the schedule in Enclosure 4.
Item (7[
Guidelines for use of the additional instrumentation, and analyses used to develop these procedures.
Response
See schedule in Enclosure 4.
Item (8)
A summary.of key operator action instructions in the current emergency procedures for ICC and a description of how these procedures will be modified when the final monitoring system is implemented.
Response
Emergency Procedure 1202-6B and 1202-39 have been provided to the NRC for review. Copies of these procedures are also available ensite.
These procedures will be modified as necessary pending outeeme of negetiations with NRC on additional instrumentation. provides the schedule.
Item (9) r A description and schedule com=itment fer any additienal submittals which are needed to support the acceptability of the proposed final instrumentatica system and emergency procedures for ICC.
Respense:
Additional submittals are shown in Enclosure 4 Item (10)
Changes to the Technical Specification.
Response
Changes concerning the saturatien margin concier were included with Tech.
Spec. Amendment 78.
Changes concerning incere thermocouple Technical Specifications were discussed in our letter of February 2,1982.
Changes concerning additional instrumentatien are pending the outec=e of negotiations with NRC and are scheduled fer sub=issien in accordance with.
e..
s Attachment I l
DESIGN AND QUALIFICATION OF THE SATURATION MARGIN MONITOR Environmental / Seismic Qualification - The pressure measurement is obtained from 2 seismic and environmentally qualified sensors.
The temperature measurement is control grade, and all signals are seismic Category I and separated for use as redundant signals. The qualification of the meter was recently discussed in our letter of February 14, 1983.
Single Failure - Redundant channels which are electrically independent are used.
Power Sources - The two saturation margin monitors use separate power supplies (115 Vac 60 Hz) powered from safety grade inverters.
Availability - The operability / surveillance requirements of the saturation
, margin monitor are provided in Amendment dated QA Requirements - The quality level of all equipment covered for the saturation margin monitor is designated as Nuclear Safety Related, Class lE and meets the requirements of the OQA plan Rev. 9.
The temperature sensors, plant computer and announciation system are nonsafety related components.
Continuous Operation - The saturation margin menitor will continuously display the margin between actual primary coolant temperature and the saturation temperature.
Recording Instrumentation - Outputs of saturation margin are provided fer trending and alarm announciation by the plant computer..
Display Instrumentation
, Digital display of the margin between actual RCS l
temperature and saturation temperature fer the existing RCS pressure is provided in the control room on the back panel (PCL).
Isolation - The Tsat computation equipment provides isolation to the pressure and temperature signals through the use of isolation devices at the signal inputs. The Tsat outputs to the announciation system and the computer utilize isolation devices to minimize potential hazardous effects from these system.
Testing - Test signals may be substituted for normal RC pressure and temperature signals to verify operation on the Tsat Margin Monitor equipment. Operating checks can be performed by reading RC pressure and temperature and with calculations obtain the Tsat margin, i
Surveillance - See item 4.
i _ _
c Removal From Service - The Tsat Margin Monitor is designed such that all necessary functional tests can be performed en line without af fecting other reactor systems. Any testing that is required to be performed offline shall not be required to be performed at less than 15 month intervals.
Access for Adjustment - The Tsat Margin Montiors are rack mounted in signal processing channel A and B equipment cabinets. Accessibility for these cabinets is the same as for normal maintenance.
Anomalous Reading - Anomalous readings are reduced to a minimum by items 2, 3 and 9.
Ease of Repair - See item 13.
Directiv Measured Variable Sensors - RCS temperature and pressure sensors provide direct inputs to the saturation margin calculation.
Normal / Accident Ranges - The Tsat Margin Monitor operates over the range of
-100
- to 400*F which is suitable for nomal and accident conditions.
Periodic Testing - Testing is described in item 10 and surveillance will be provided as discusseo in item 11.
S m
e
i l
l l
Design and Qualification of The RCS Inventory Trending System l
l 1.
Environmental / Seismic Qualification The RITS will be environmentally qualified for the respective DBE environments and will be displayed in the Control Room via the plant computer (non seismic beyond isolator) to provide the necessary trending information for d/p, The installation will be seismically qualified and the ranges of the d/p instruments will be commensurate with the measure of the. cop of the RC hot leg /RV head to the top of the active fuel. For RCP motor power, see deviation 1 attached.
2.
Single Failure The RITS is designed such that no single failure of the instrument or auxiliaries prevents the Operator from determining the safety status of the unit.
Redundancy is provided by dual ins trument trains.
However, the two instrument trains have taps in common at the penetration in the RV head and bottom.
The c ommo nal tities do not jeopardize the redundancy of the two instrument trains since it is highly unlikely that the tap will fail either frec plugging or breaking.
Plugging is prevented by using corrosive resistant materials and the absence of flow of concentrated boric acid. Even in the event protection sys te= but rather a monitoring of tap failure, the RITS is not a system with adequate backup from the core exit thermocouples.
3.
Power Supplies Class IE power source will be provided for the d/p portion of the RITS.
The RCP motor power monitoring is non IE as is the RCP's.
Display beyond the isolation device is non class lE for the computer.
4.
Availability The operability requirements for the RITS will be address in a Tech Spec change request to be provided as discussed in Enclosure 4.
5.
QA Recuirements The QA requirements will be those imposed by the OQA plan Rev 9 for Nuclear Safety Systems which include the appropriate reg guide requirements.
6.
Continuous Operation Tite RITS will continuously display RCS trending information (pumps on/ pumps off).
7.
Recording Instrumentation Outputs of the RITS will provide for trending and alarm in the Control Room via the plant computer.
8.
Display Instrumentation into the plant computer. Location and type Display information will be broughtof display will be part of the Control Room progress.
9.
Isolation Isolation of signal channels will be provided up to the plant computer.
10.
Testing Test signals may be substituted f or normal RC pressure / level and temperature Further information will be provided.
signals to verify operation of the RITS.
11.
Servicing Testing and Calibration been developed at this stage of the Service, Test and Calibration have not design. They will be provided as shown in the enclosed schedule (Enclosure 4).
~
Removal From Service 12.
located inside containment and will be 2
The d/p portion of the system will be instrumentati~on is RC pump motor power accessible only during plant shutdown.
Because of the conceptual nature of the design located outside of containment.
no further details are available at this time.
13.
Acce:.ss to Adiustment Same as Item 11.
~
14 Minimize Anomalous Reading in general masked by the During RC pump operations anomalous readings are available ICC instrumentation.
2 trending nature of the inf ormation and otheroperations anomalies are exp 1
During quiescent
~
can be checked against other ICC instrumentation.
i 15.
Repair repair will be identified during surveillance and for system, problem
~_
Location of components During cperation, because of the simplicity of the identificatien will be simplified.
For the d/p portion of the sys tem the I
calibration.
and ar e inaccessible during power i
transmitters are located inside containment operations.
h 16.
Desired Variables Inventory is not measured directly but is inferred f rem hydros tatic head.
of actual level but generally Voiding will result in some error in measurementThis differential pressure measurement in the conservative direction (lower).
does indicate inventory.
i give information about cooling capacity buttrending when RC Pump does not results using For void fraction together with analytical results of Enclosure 2 show acceptable
~
1 e
_9_
"-"* w
RCP motor power, 17.
Normal / Accident Ranges The RIT.C is capable of measuring about 100% of the inventory above the top of the active fuel to the top of the candy canes in the quiescent state.
For RC pumps running the RITS measures motor current which is correlated to veid i
fraction from 0-100%,
18.
Periodic Testing Same as item 11, 9
9 6
Deviations 1.
Reactor Coolant Pu=p Motor Power a.
Deviation The reactor coolant pump motor power is not presently classified 1E or environmentally qualified (EQ).
Further, it is not intended that this instrumentation will be upgraded to 1E EQ for the final ICC system.
b.
Justification The reactor coolant pumps in B&W plants are not classified as lE or EQ (including Seismic) nor are the associated RCP motor power circuitry.
Experience with the RCP and motor current circuit indica t es that they are highly reliable and relocation of the pumps circuitry is not cost justified. Present procedures require RCP trip at 1600 psig which is long before voided conditions would exist in the RCS.
c.
Cost / Benefit Analysis The RITS is strictly a monitoring system and not a protection system.
Relocation of 4 RCP's switchgear would cost hundreds of thousands o f dollars large costs in man rem exposure and is therefere not justifiable when evaluated against the small benefit of monitoring in an intermediate phase of the ICC scenario where other ICC instrumentation would also provide indication (SMM).
2.
RITS taps (RV head and single Incore Tube) a.
Deviation The RITS taps on the RV head and incere tube are single tap points, b.
Justification The common tap points which minimize penetration of the RCS and particularly the reactor vessel have a very low probability cf f ailing frem either plugging or breaking.
Plugging is prevented by using corrosive resistant materials and the absence of flow cf concentrated boric acid.
Even in the event of tap failure the RITS is not a protection sys tem but a monitoring system with adequate backup from core exit thermecouples, c.
Cost / Benefit Analysis Additional taps represent additional penetration into the RCS at the reactor vessel.
Added taps would also provide greater man re= exposure.
The cost of the additional tap would be on the order of $100,000 which does not appear to be justified as indicated above.
3.
Saturation Margin Monitor a.
Deviation
o The saturation margin monitor indicator in the Control Room is not seismically qualified.
b.
Justification See GPUN letter dated February,18, 1983 c.
Cost Benefit Analysis Since no digital seismically qualified instrument is presently available no cost benefit information is supplied.
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ENCLOSURE 4 SCHEDULES FOR RITS I
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t ICC Instrumentation Schedule (RITS)
Proposed Final Engineering Design Description 12/1983 Including Display for Hot Leg d/p RV Head d/p and RC Pump Motor current Control Room Design Review of ICC Instrumentation 12/83**
Start Procurement of RITS Equipment 12/83 Operator Guidelines for RITS Including Analysis 4/84 Used in Developing Procedures Complete Procurement of RITS Equipment 6/84 Commence Installation of RITS Cycle 6 Refueling 7/84 Operating and Emergency Procedure Mods 7/84*
Commence Installation of Safety Grade (S/G) 7/84 Saturation Margin Monitor (SMM) Cycle 6 Refueling Submit Technical Specifications for ICC 7/84 Instrumentation Commence Training of RO/SRO 7/84 Complete Installation of RITS and S/G SMM 9/84 Testing and Calibration of RITS Cycle 6 Refueling 9/84*
Complete training of RO/SRO 9/84 Operation Date of RITS Based on approval of NRC#
l Environmental Qualification of RITS and Incore 3/85 Thermocouples 18Following NRC review of asterisked (*) items.
- Based on a May 1, 1983 concurrence by the NRC on the conceptual design.
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