ML18085A845
| ML18085A845 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/06/1981 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18085A846 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8102200246 | |
| Download: ML18085A845 (92) | |
Text
....
PS~G Public Service Electric and Gas Company 80 Park Plaza Newark, N.J. 07101 Phone 201/430-7000 February 6, 1981 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Frank J. M_iraglia, Chief Licensing Branch 3 Division of Licensing Mr. S. A. Varga, Chief Operating Reactors Branch 1 Division of Licensing Gentlemen:
REACTOR VESSEL LEVEL INSTRUMENTATION NO. 1 AND 2 UNITS SALEM NUCLEAR GENERATING STATION DOCKET NOS. 50-272 AND 50-311 PSE&G hereby submits additional information required by NUREG-0737, concerning Item II.F.2, "Instrumentation for Detection of Inadequate Core Cooling." This information.
supplements our submittal of December 31, 1980 and addresses specifically the Reactor Vessel Level Indicating System.
As this submittal contains information proprietary to West-inghouse Electric Corporation, it is supported by an affida-vit signed by Westinghouse, the owners of the information.
The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in Paragraph (b) (4) of Section 2.790 of the Commission's Regulations.
Accordingly, it is respectfully requested that the informa-tion which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's Regulations. Correspondence with respect to the proprietary aspects of this application for withhold-ing or the supporting Westinghouse affidavit should refer-ence CAW-81-12, and should be addressed to R. A. Wiesemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P. o. Box 355, Pittsburgh, Pennsylvania 15230.
The Energy People Z/0~00~1~ 95-0942
2/6/81 This submittal consists of (5) copies each of the proprietary and non-proprietary versions of this report and (1) copy of the supporting affidavit.
Should you have any questions in this regard, do not hesitate to contact us.
- ~i;;;;r-R. L. Mittl General Manager -
Licensing and Environment CC: Mr. Leif Norrholm Senior Resident Inspector DD02 1/2
Westinghouse Water Reactor Nuclear Technology Division Electric Corporation Divisions Box355 Pittsburgh Pennsylvania 15230 February 3, 1981 Director of Nuclear Reactor Regulation CAW-81-12 U. S. Nuclear Regulatory Commission Phillips Building
. 7920 Norfolk Avenue Bethesda, Maryland 20014 ATTENTION: Mr. F. J. Miraglia, Chief License Branch 3 Mr. S. A. Varga, thief Operating Branch 1
SUBJECT:
Public Service Electric and Gas Company, Salem Units 1 and 2, Reactor Vessel Level Instrumentation REF: Application for Withholding, Mittl. to Miraglia and Varga, February 1981 *
Dear Messrs. Miraglia and Varga:
The proprietary materia-1 for which withholding is being requested by Public Service Electric and Gas Company is of the same technical type as that proprietary material recently provided by* Westinghouse in response to the .,
J~.
concern given in NUREG-0737 for vessel level instrumentation. The previous application for withholding, AW-77-18, was accompanied by an affidavit signed :,
\_,
_ by the owner of the proprietary information, Westinghouse Electric Corporation.
Further, the affidavit submitted to justify the prev1o-us mater"ial i's-equally - - -
applicable to the subject material. The subject proprietary.material is being submitted in support of Public Service Electric and Gas Company, Salem Units 1 and 2. Accordingly, this letter authorizes the utilization of the previously furnished affidavit in support of Public Service Electric and Gas Company. A copy of the affidavit, AW-77-18 dated April 20, 1977, is attached.
Correspondence with respect to the proprietary aspects of the application for _
withholding or the Westinghouse affidavit should reference CAW-81-12 and should be addressed to the undersigned.
Very truly yours,
/bek * -
- -£lJJlJJlli~<<-e!AL4J Robert A. Wiesemann, Manager Attachment Regulatory & Legislative Affairs cc: E. C. Shoma~er, Esq. /
Office of the Executive Legal Director, NRC
,~,, ,..,., ,., .
- ._* .., .:* * *** **..,**** .*:.* * * * -~ *_;_.'-~*~:*:_:* *~.:
A"**=-**..<:.-:: ..,*.:_: . ,?.:*'-.1i'.*i-'*>.*' :* -, ..J a
- * *. -* .,.*,*. **1..
- 1 *1?' ~ :.' , . . * ,,., t ** , a... f '-/ r* '/. *;, *1~*1 I; .:J
- J,1 I * **
AW-77-18 AFFID,'WIT COMMOmlEALTH OF PEWISYL vr\~HA:
SS COUNTY OF ALLEGHENY: ..... _
Before me, the undersigned authority,* personally appeared Robert A. Wiesemann, who, being by me duly sworn according to law, de-poses and says that he is authorized to execute this Affidavit on behalf of WestJngho.use Electric Corpora ti on ("Westinghouse") and that the aver-ments of fact set forth in this Affidavit are true and correct to .. the best of his knm*:ledge, information, and belief:
Robert A. Hi cse;;;ann, i*:an.!ger Licensing Programs Sworn to and subscribed before me thJs i.!!_ day of ([~-'fl f 1977.
~v.
}, ,((({/. /}t{;{.{l-( Notary Public
... . - :* * ::Ji?
j
- 'i*,),,., 0)1 1~.l ) J.,.,,._ ~:.!1,:.:.; ..... *' **:'\:*:*~ \"~* *.~*: .* -;. ~**. :*. . *..*,,.-;,*,.1 :Y
- ':') ~" I .-:. -~::, _'._. : ; :...-~..,. *-:,*,,,.. '} ,.,\' *,*,l}' 'a... I.,~'*' .. *,* ":1* i) *
.~
- ~ ~
i
- i 1,* w-
, * -., * * * ' I /
10
- yt***
.~
0
.,.ti
- , *** . (
I* * .
-2.:** """'
/ *..*
- 1 ,* *,*. ,!;1)'*J*1,,*,*;:******1 . . Jl;,r
- I.,, *1
- --'I AW-77-18
~,.. ;,.,*
(1) I am Manager, licensing Progra~&, 1n the Pressurized Water Reactor Systems Division, of Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary infonnation sought to be '°'ithheld from public dis-closure in connection with nuclear power plant licensing or rule-maki ng proceedings, and am authorized to apply for its wi thho 1ding on behalf of the Westinghouse Water Reacto~ Divisions.
(2) I am m~king this Affidavit in conformance with the provisions of 10 CFR Section 2.790 of the CoJTiiilission's regulations and in con-junction with the Westinghouse application for withholding ac-compa*nying this Affida.vit.
(3) I have personal knowledge of the criteria and procedures utilized
. by Westinghoc:.:- ..~iu::1:::***-i::-:-**.::: Systems in designating iniorr.iation as a trade secret, privileged or as confidential cor.;r.ercial or financial i nfonnati on. .
. . {4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Cmiss~ion's regulattons1 the follO'iling is furnished for c*o!'sideration by the Co.TJilission in detennining whether the in-formation sought to be withheld from public disclo~ure should be withheld.
(i) The infonnation sought to_be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
. I , .... ~ ....
-~ AW-77-lS (11) The infonnation is of a type customarily*
held in confidence by Westinghouse and not customarily disclosed to the public.
Westinghouse has a rational basis for determining the types of infonnation customarily held in confidence by it and, in that connection, uti 1i zes a sys ter.J to determine when and whether to hold certain types of i ~fornation in confidence. The ap-plication of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential com-petitive advantaqe. as ..'.IIfo 11 c~1s:
(a) The infonnation reveals the di_stinguishing aspects of a process {or comp~nent, structure, tool, method, etc.)
where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other co~panies.
(b) It consists of suppo~ting data, including test data, relative to a process (or component, structure, tool, method~ etc.), the application of which data secures a tompetitive economic advantage, e.g., by optimization or improved marketability *
, .. . .,_ *********. .-~***.*.**~***11****;,~'~* .* ,:*.*.. ;***
- ---~
'"**-**,-,11.:,**~*1.,
- e
.. ;,,.,,~;*1*,1'/**;,**~I**,
,*.i,**'"*'/ .,.; ; , , *., .* J. I 4- e _*. AW-77-13 (c) Its use by a cor.ipetitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d) It reveals cost or price inforT.lation, production cap-acities, budget levels, or commercial strategies of Westinghouse, its customers or supp1iers.
(e) It reveals aspects of past, present, or future West-inghouse or customer funded development plans and pro-grams of potential coi'imercial value to Westinghouse~
(f) It contains patentable ideas, for which patent pro-tection may oe aesirablea (g) It is not the property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.
There are sound policy reasons behind the Westinghouse system which include the following:*
(a) The use of such information by Westinghouse gives WQstinghouse a competitive advantage over* its com-petitors. It is, therefore, withheld from disclosure to protect.the Westinghouse competitive position *
" ., .. ..... , .... - --- ..e**
. . *...... ' . . . *r*, . . .,, . .. ... **9*
J. * * ,_, .,,
' " .. '-,,. \ . .
'\
- ~*#*, *-*~****
- ~
. ' * *.',
- I I
~
- :.... Iii.* I*1
- ':I: *
- \ \I 11 *\
' / *
,. '* I y ,' ',
~-* . , .. **
,*, , ~ ** 1 ..
1 I.*,,, 1 *.
I* 1
- : I!*; : *' . *
- 1 II * * , * *
. *' * *. ' I *
- _, I A!4-77-18 (b) It is information which is marketable in many ways.
The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the
- information.
(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing.hi~ expenditure
_..... of resources at our expense *
(d) Each component of pr.oprietary information pertinent to a particular ~ompetitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire cc~~onents of proprietary infor-mation, any one component ma~' be the key to the entire puzz-1 E, thereby depriving ~*les ti nghous e of a competitive advantage.
(e) Unrestricted disclosure would jeopardize the position of prominence of Wes~inghouse in the world market, and thereby give a market ad':antage to the co~petition in those countries.
(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantag~ *
. *- '.- ~,, ... .. .. ' ... - . .. '*'la ** ) ., t ' ,
AH-77-Je (iii) The infonnation is being transmitted to the Cormiission in confidence and, under the provisions of 10 r.r~ ~~ction 2.790, 1t fs to be received in confidence b;; the Cr;r.::ission.
(iv) The infonnation is not available in public ~o:.a*ces to the best of our knm*1l edge and be 1i ef.
(v) The proprietary information sought to be withheld in this submittal is that which is attached to Westi1:~!~'j*Jse letter Nuinber NS-CE-1403, Eicheldinger to Stolz, Gt:~*~:~ .:i..pril 6, 1977. The letter and attachment are being.s!.!~:*~tted in support of the Westinghouse emergency core c:rio~i~g system evaluation model.
Public disclosure of the information sought tt~ he \'lithheld is likely to cause substantial harm to the ccs~ctitive position of Westinghouse, taking into account the value of the fnfo~ation to Westinghouse, the ar.iount o*::* effort and ~ I I
I money expended by Westinghouse in developing the infonnation, and considering the ways in which the information could: be
( .
acquired or duplicated by others.*
Further the deponent s aye th not.
.i
- i* .
- I L:.::~ ..... ~~:~~-.*,,-~;- .-.::.;;_*."--**~ '.~,~--~*..:,.-_'~u ~ ... ~.t-. . ,~~~~r-.-. ***->:-'l.K...,,"',1- ,-.* ?*
---~~~- *
.,.*-~ .**
- __l ____
- -*****--*.*. *--- ---. .-*- ~.- ~-*-=-~*--._ ..,*:*:.* -~ ..- ;:-*" '" -*'*' * *= :~"':'.'.":-"<~-:-:* :o-*.:*.,_o,;,-*~--,,.*-1
oArE = ;i. - ~3 - eI
. NOTE TO NRC AND/OR LOCAL PUBLIC DOCUMENT ROOMS
. . -~-to- FJ I - .- .
The following item submitted with letter dated------------------------------------
. from --J?..~~--/:_G!_ ____________________ is being withheld from p~blic disclo.~~re
.. in accordance with Section 2.790.
PROPRIETARY INFORMATION 0/6 Distribution Services Branch
~~ WESTINGHOUSE, CLASS 3 l(a) Design Description of Reactor Vessel Level Instrumentation System 1.1 GENERAL DESCRIPTION The reactor vessel level instrumentation system (RVLIS) uses differential pressure (d/p) measuring devices to measure vessel level or relative void content of the circulating primary coolant syste:r.
fluid. Th~ system is redundant and includes auto-rnatic compensation for potential ternperatur~ var1a-tions of the irn~luse lines. Essential information is displayed in the main control room in a form dir~ctly us~able by th~ operator.
The functions p0rforrn~d by the RVLIS ar0:
- 1. Assist in detecting the presence of a gas bubble or void in the reactor vessel.
- 2. Assist in detecting the approach to ICC.
- 3. Indicate the formation of a void in the RCS during forced flow conditions.
M P81 33/l l
1.2 DETAILED SYSTEM DESCRIPTION 1.2.l HARDWARE DESCRIPTION 1.2.l.l Differential Pressure Measurements The RVLIS (Figure 1-1) utilizes two sets of three d/p cells. These cells measure the pressure drop from the bottom of the reactor vessel to the top of the vessel, and from the hot legs to the top of the vessel. This d/p measuring system utilizes cells of differing ranges to cover different flow behaviors with and without pump operation as discussed below:
- 1. Reactor Vessel - Upper Range (4Pa)
The d/p cell6 Pa shown in Figure 1-1 provid~s a measurement of reactor vessel level above the hot leg pipe when the reactor coolant pump (RCP) in the loop with the hot leg connection is not operating.
- 2. Reactor Vessel - Narrow Range (6 Pb)
The measurement provides an indication of reactor vessel level from the bottom of the M P81 33/l 2
EACTOR VESSEL HEAD VENT PIPE NARROW RANGE REACTOR CORE MOVEABLE DETECTOR CONDUIT -~_.__
TRAIN A TRAIN B Figure/-/ Reactor Vessel Level Instrument System
reactor vessel to the top of the reactor during natural circulation conditions.
- 3. Reactor Vessel - Wide Range ( 4S Pc)
This instrument provides an indication of re-actor core and internals pressure drop for any combination of operating RCPs. Comparison of the measured pressure drop with the normal, singlephase pressure drop will provide an approximate indication of the relative void content or density of the circulating fluid.
This instrument will monitor coolant condi-tions on a continuing basis during forced flow conditions.
To provide the required accuracy for level measure-ment, temperature measurements of the impulse lines are provided. These measurements, together with the existing reactor coolant temperature measurements and wide range RCS pressure, are employed to compensate the d/p transmitter outputs for differences in system density and reference leg density, particularly dur-ing the change in the environment inside the contain-ment structure following an accident.
M P81 33/l 3
The d/p cells are located outside of the containment to eliminate the large reduction (approximately 15 percent) of measurement accuracy associated with the change in the containment environment (temperature, pressure, radiation) during an accident. The cells are also located outside of containment so that sys-tern operation including calibration, cell replace-1 ment, reference leg checks, and filling is made easier.
1.2.l.2 System Layout A schematic of the system layout for the RVLIS is shown in Figure 1-2. There are four RCS penetra-tions, one connection in the reactor vessel head vent pipe, one connection to an incore instrument conduit at the seal table, and connections into the side of the two RCS hot leg pipes.
The pressure sensing lines extending from the RCS penetrations will be a combination of 3/4 inch Schedule 160 piping and 3/8 inch tubing and will include a 3/4 inch manual isolation valve as de-scribed in Sect~on 2.2. These lines connect to six
- M PBl 33/l 4
HYDRAULIC ISOLATOR (TYPICAL)
SEAL TABLE CONTAINMENT WALL Figure /-2. Process Connection Schematic, Train A
sealed capillary ~mpulse lines (two at the reactor head, two at the seal table and one at each hot leg) which transmit the pressure measurements to the d/p transmitters located outside the containment build-ing. Th~ capillary impulse lines are sealed at the RCS end with a sensor bellows which serves as a hydraulic coupling for the pressure measurement. The impulse lines extend from the sensor bellows through the containment wall to hydraulic isolators, which also provide hydraulic coupling as well as a seal and isolation of the lines. The capillary tubing extends from the hydraulic isolators to the d/p transmitters, where instrument valves are provided for isolation and bypass.
Figure 1-3 is an elevation plan of a ty~ical plant showing the routing of the impulse lines. The im-pulse lines from the vessel vent connection must be routed upward out of the refueling canal to the operating deck, then radially toward the seal table and then to the containment penetration. The con-nection to the bottom of the reactor vessel is made through an incore detector conduit which is tapped with a T connection at the seal table. The impulse M P81 33/l 5
'i so G.. so
+40FT r-+-
I
+
..J I "'~
I IA.I I >-
a:L I
I o~
~-
I
~
L- +-PROTECTED IA.I ENVIRONMENT a:
I ...... t II OFT I
-11FT Figure 1-3 Plant Arrangement for RVLIS
- line from this connection is routed axially and radially to join with the head connection line in routing to the penetrations. Similarly, the hot leg connection impulse lines are routed toward the seal table/penetration routing of the other two connections.
The impulse lines located inside the containment building will be exposed to the containment tempera-ture increase during a LOCA or HELB. Since the vertical runs of impulse lines form the reference leg for the d/p measurement, the change in density due to the accident temperature change must be taken into account in the vessel level determination. There-fore, a strap-on RTD is located on each vertical run of separately routed impulse lines to determine the impulse line temperature and correct the reference leg density contribution to the d/p measurement.
Temperature measurements are not required where all three impulse lines of an instrument train are routed together. Based on the studies of a number of repre-sentation plant arrangements, a maximum of 7 inde-pendent vertical runs must be measured to adequately compensate for density changes.
M P81 33/l 6
- 2. Following is the design analysis including the evaluation of various instruments employed in monitoring water level.
2.1 RESISTANCE TEMPERATURE DETECTORS (RTD)
The resistance. temperature detectors (RTD) associat~d with the RVLIS are utilized to obtain a temp~ratur~
signal for fluid filled instrument lines inside th~
containment during normal and post-accident opera-tioni. The temperature measurement for all vertical instrument lines is used to correct th~ vessel level indication for density -changes associated with the environmental temperature chang0.
Th~ RTU assembly is a totally enclos~d and hermeti-cally sealed strap-on device consisting of ther~a:
element, extension cable, and terminatio~ cable as indicated in Figure 2-1. The sensitive portion o:
the device is mounted in a removable adapter assembly which is designed to conform to the surface of the tubing or piping being monitored. The materials arc all selected to *be compatible with the normal anJ post-accident environment. Randomly selected s~mples M PBl 33/l 7
- a,b,c Figure ?-1 Surface Typ0 Clnmp-On R0sistance Temperature Detector
- from the controlled (material, manufacturing, etc.)
production lot will be qualified by type testing.
Qualification testing will consist of thermal aging, irradiation, seismic testing and testing und~r simu-lation high energy line break environmental condi-tions. The specific qualification requirem8nts for the RTDs are as follows:
- 1. Aging The thermal aging test will consist of operatin3 the detectors in a high temperature environment:
either 4000F for 528 hours0.00611 days <br />0.147 hours <br />8.730159e-4 weeks <br />2.00904e-4 months <br /> or per other similar Arrhenius temperature/time relationship.
- 2. Radiation The detectors shall be irradiated to a total integrated dose (TID) of 1.2 x 108 rads ga~~~
radiation using a co60 source at a minimum rate of 2.0 x 106 rads/hours and a maximum rate o: 2.S x 106 rads/ hour. Any externally exposed organic materials shall be evaluated or tested to 9 x 108 rads TIO beta radiation. The energy of M P81 33/l 8
the beta particle shall be 6 MEV for the first 10 MRad, 3 MEV for 340 MRad, and l MEV for 150 MRad.
- 3. Seismic The netectors will be tested using a biaxial seismic simulation. The detectors shall b~
mounted to simulate a plant installation and will be energized throughout the test.
- 4. High Energy Line Break Simulation The detectors shall be tested in a saturated steam environment using the temperature/pressur~
curve shown in Figure 2-2.
Caustic spray, consisting of 2500 ppm boric acid dissolved in water and adjusted to a pH 10.7 at 2s 0 c by sodium hydroxide, shall be applied during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The test units will be energized throughout the test.
The RTIJ device is designed to operate ove:- a t~mp~ratur~ rang~ of -ss 0 to 400°F (the normal tempcratur~ range is 50° to 130°F).
M P81 33/1 9
CAUSTIC SPRAV 1: 72 PSIA -----41*'""1'.-- SATURATED~
STEAM I
420 I I
I I
I 3-40 ------1 LL 305
-w 0
a:
t-
~
a:
w Q,,
~
205 U.I 120
- 6 0 10 3 6 20 24 18 0 10 3 MIN SEC MIN MIN MIN HOUR DAY SEC MIN TIME
- TIME BETWEEN TEMPERATURE TRANSIENTS MUST BE AT LEAST ONE HOUR OR UNTIL TEST UNITS RETURN TO A STEADY STATE OUTPUT. TIME ABOVE 34D°F MUST BE FIVE MINUTES OR LESS.
F, "..ire 2-2 HELB Stmulatton Prof tle
2.2 REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM VALVES Two types of valves are used for the RVLIS. The 3/4" root valves are ASME Class 1, stainless steel, glob~
valves. The basic function of the valve is to isolate the instrumentation from the RCS. The oth~r valves (3/8" And~rson-Greenwood), are an instrumr::nta-tion-type valvr::. It is a manually actuated globr::
valve used to provide isolation in the fully closed position.
2.3 TRANSMITTERS, HYDRAULIC ISOLATORS, AND SENSORS 2.3.l Differential Pressure Transmitter Th~ d/p transmitters are of a seismically qualified design. In the RVLIS applicatio~,
accuracy considerations dictate a protected environment, consequently transmitters ar~
rated for 40 to 130°F and 104 rad TID.
Tric: s:;,~c1al reqljirements for* these transmitters arc: as follows:
M P81 33/l 10
- l. Must withstand long term overloads of up to 300 percent with minimal effect on calibration.
- 2. High range and bi-directional units re-quired for pump head measurements.
- 3. Must displace minimal volumes of fluid in normal and overrange operating modes.
The first two requirements are related to the vernier characteristic of the pumps off level measurements and the wide range measurements, respectively. The third is related to the limited driving displacement of the hydraulic isolator when preserving margins for pressure and thermal expansion effects in the couplin~
fluid.
The d/p transmitters are rated 3000 ps1g working pressure and all units are tested to 4500 psig. Internal valving also provide~
overrange rating to full working pressure.
M P81 33/l 11
2.3.2 Hydraulic Isolators a,c M P81 33/l 12
a,b,c Figure 2-3 ITT Barton Hydraulic Isolator Internal Scheme
a,c High Volume Sensors
_I a,c M P81 33/1 13
_ a,c M P81 33/1 14
a,b,c i
i I
i I
I Figure 2-4 ITT Barton "High Volume" Sensor Bellows Check Valve J
- 3. The following is a description of the test programs to be and being conducted for evaluation, qualifi-cation, and calibration of the RVLIS.
3.1 TEST PROGRAMS A variety of test programs are in progress or will be carried out to study the static and dynamic per-formance of the RVLIS at two test facilities, and to calibrate the system over a range of normal operating conditions at each reactor plant where the system is installed. These programs will pro-vide the appropriate verification of the system response to accident conditions as well as th~
appropriate procedures for proper operation, mair.-
tenance and calibration of the equipment. A de-scription of these programs is presented in the following section:
3 .1 .1 Forest Hills Test Facility A breadboard installation consisting of one train of a RVLIS was installed and tested at the Westing-house Forest Hills, PA. Test Facility. The system
- M P81 88 01/l
consisted of a full single train of RVLIS hydraulic components (sensor assemblies, hydraulic isolators, isolation and bypass valves and d/p transmitters) connected to a simulated reactor vessel. Process connections were made to simulate the reactor head, hot leg and seal table connections. - Capillary tub-irlg which in one sensing line simulated the maximum expected length (400 feet) was used to connect the sensor assemblies to the hydraulic isolators and all joints were welded. Connections between the hydraulic isolators, valves and transmitters uti-l~zed compression fittings in most cases. Resist-ance temperature detectors, special large volume sensor bellows and volume displacers inside the hydraulic isolator assemblies which are normally part of a RVLIS installation were not included in the installation since elevated temperature testing was not included in the program.
The hydraulic isolator assemblies and transmitters were mounted at an elevation slightly below the simulated seal table elevation.
The objectives of the test were as follows:
M P81 88 01/2
- 1. Obtain installation, filling, and maintenance experience.
- 2. Prove and establish filling procedures for initial filling and system maintenance.
- 3. Establish calibration and fluid inventory maintenance procedures for shutdown and normal operation conditions.
- 4. Prove long term integrity of hydraulic components .
- 5. Verify and quantify fluid transfer and makeup requirements associated with instrument valve operations.
- 6. Verify leak test procedures for field use.
3.1.l~l Reactor Vessel Simulator The reactor vessel simulator consisted of a 40 foot long 2-inch diameter stainless steel pipe with taps M PBl BB 01/3
at the top, side and bottom to simulate the reactor head, hot leg, and incore detector thimble conduit penetration at the bottom of the vessel. Tubing (0.375 inch diameter) was used to c6nnect this lower tap to the sensor at the simulaten seal tablr:
elevation and the hot leg sensor to the head connection wai simulated by 1-inch t~bi~g which connected the sensor to the vessel.
The reactor vessel simulator was designed for a pressure rating of 1400 psig to comply with local stored energy and safety code considerations *
- 3.1.1.2 Installation The system was installed in the high bay test area of the Westinghouse Forest Hills Test Facility by Westinghouse personnel under the supervision of Forest Hills Test Engineering. All local safety codes were consfdered in the construction.
- M PBl 88 01/4
3.1.1.3 Filling Operation a,c M P81 88 01/5
a,c M P81 88 01/6
3 .1. 2 SEMISCALE TESTS In order to study the transient response of the RVLIS during a small-break LOCA and other accident conditions, the hydraulic components of the RVLIS have been installed at the Semiscale Test Facility in Idaho. Vessel level measurements will be ob-tained during the current semiscale test program series which runs from December 1980 to March 1982. The tests scheduled to be completed by July 1981 are expected to provide the desired transient response verification; additional data will be ob-tained from the tests scheduled for completion by November 19 81.
The Semiscale Test Facility is a model of a 4-Loop pressurized water reactor coolant system with ele-vation dimensions essentially equal to the dimen-sions* of a full-size system. The reactor vessel contains an electrically heated fuel assembly con-sisting of 25 fuel rods with a heated length of 12 feet. Two reactor coolant loops are provided, each having a pump and a steam generator with a full M P81 88 01/7
height tube bundle. One loo~ models the loop con-taining the pipe break, which can be located at any point in the loop. The other loop models the three I
intact loops. A blowdown tank collects and cools the fluid discharged from the pipe break during the simulated accident. Over 300 pressure, tempera-ture, flow, l~vel and fluid density instruments are installed in the reactor vessel and loo~s to record the fluid conditions throughout the test run. Test results are compared with* predictions for verifica-tion of computer code models of the transient performance *.
The Westinghouse level measurements obtained during a test run will be compared with data obtained frorr existing instrumentation installed on the semiscale reactor vessel. The semiscale facility has two methods of measuring the level or fluid density:
d/p measurements are obtained over 11 vertical spans on the reactor vessel to determine level within each span, and gamma densitometers are in-stalled at 12 elevations on the reactor vessel to d~termine the fluid density at each elevation.
M P81 88 01/8
This data establishes a fluid density profile within the vessel under any operating condition, and this information will be compared with the data obtained from the Westinghouse level instrumenta-tion. Other semiscale facility instruments (loop flows and fluid densities when pumps are operating, and pressure ahd temperatures for all caies) will provide supplemental information for interpretation of the test facility fluid conditions and the level measurement.
Specific tests included in the semiscale test pro-gram during which Westinghouse RVLIS measurements will be obtained are as follows:
- 1. Miscellaneous steady state and transient tests with pumps on and off, to calibrate test facility heat losses.
- 2. Small-break LOCA test with equivalent of a 4-inch pipe break.
M P81 88 01/9
- 3. Repeat of small-break LOCA test with test facility modified to simulate a plant with ()
upper head injection (UHI).
- 4. Several natural convection tests covering sub-cooled and saturated coolant conditions and various void contents.
- 5. Tests to simulate a station blackout with dis-charge through relief valves.
- 6. Simulation of the St. Lucie cooldown incident.
3 .1. 3 PLANT STARTUP CALIBP.ATION During the plant startup, subsequent to installing the RVLISi a test program will be carried out to confirm the system calibration. The program will cover normal operating conditions and will provide a reference for comparison with a potential acci-dent condition. The elements of the program are described below:
M P81 88 01/10
- l. During refilling and venting of the reactor vessel, measurements of all 6 d/p transmitters would be compared to confirm identical level indications.
- 2. During plant heatup with all reactor coolant pumps running, measurements would be obtained from the wide range d/p transmitters to confirm or correct the temperature compensation pro--
vided in the system electronics. The tempera-ture compensation, based on a best estimate of the flow and pressure drop variation during startup, corrects the transmitter output so that the control panel indication is maintained at 100 percent over the entire operating tem-perature range.
- 3. At hot standby, measurements would be obtained from all transmitters with different combina-tions of reactor coolant pumps operating, to provide the reference data for comparison with accident conditions. For any pump operating condition, the reference data, represents the M P81 88 01/11
normal condition, i.e., with a water-solid sys-tern. A reduced d/p during an accident would be
- an indication of voids in the reactor vessel.
- 4. At hot standby, measurements would be obtained from the reference leg RTDs, to confirm or cor-rect reference leg temperature compensation provided in the system electronics.
M PBl 88 01/12
4.1 OPERATING PERFORMANCE Each train of the RVLIS is capable of monitoring coolant mass in the vessel from normal operation to a condition of complete uncovery of the reactor core. This capability is provided by th~ thr~c d/? transmitters, each trans~1tter covering a specific range of operating conditions. Th~
three instrum~nt ranges provide. overlap so that the measure-ment can be obtained from mar~ than one display und~r most accident conditio~i::. Ca;;a~_,ilitir:c: of each of th<;- mr.:asJr":-
ments are descritJ'.:'l t:i*.:l:::;,;:
- 1. Reactor Vessel - UpfY.:r Rang*:
The transmitter span covers the distance fiom the hot leg piping connection to the top of the reactor vessel.
With the reactor coolant pump shut down in the loop with the hot leg connection, the transmitter output is an.
indication of the level in the upper plenum or upper M PBl 88 03/l
head of the reactor vessel. The measurement will also provide a confirmation that the level is above the hot leg nozzles.
When the pump in the loop with the hot leg conn~cti0n is operating, the d/p would be greater than the trans~itt~r span, and the transmitter output would be deleted fro~
the digital panel. An invalid status statement would be indicated.
- 2. Reactor Vessel - Narrow Rang0 Tne transmitter S?an covers the total height of the reactor vess~l. ~1th pumµs shut down, the transrnitt~r output is an indication of the collapsed water level, i.e., as if the steaffi bubbles had been separated fro~
the water volume. The actual water level is slightly higher than the indicated water level since there will be some quantity of steam bubbles in the water volu~0.
Therefore, the RVLIS provides a conservative indication of the level effective for adequate core cooling .
- M P81 88 03/2
- When reactor coolant pumps are operating, the d/p woulrl be greater than the transmitter s~an, and the transmit-ter output would be deleted from th~ digital dis~l~;
panel. An invalid status statement would b~ in~icate~.
- 3. Reactor Vessel - Wide Range The transmitter span covers the entire rang'= of inter-est, froffi all pum?s operating with a wat~r-solid syst'=~
to a completely empty reactor vessel and, therefor~,
covers the measurement ~pans of the other two instr~-
men ts. Any reduction in d/p compared to the norffial opr::rating c0nJ1tion is an in.J1cation of vr..Jios int:.'.:
vesse 1. Thr.: r~actor coolant pumps will c1rc-ilat1.: t:.r~
water and steam as an essentially homogeneoJs ~ixt~r~,
so there:: would bl: no distinct water lev*.:l in th~ vr.::?-
sel. When pumps are not operating, the transmitter O .... -
put is an additional indication of the level in the ves-sel, supplem~nting the indications from the other instruments .
M P81 88 03/3
The output of each transmitter is compensated for the d~n-sity difference between the fluid in the reactor vessel and the fluid in the nderencc leg at tht.: initial amr;i*.:nt t~r:.:,-
erature. The compensation is based ori a wide l <*- *;,
temperature measurement or a wide rang(: system pressJr~
measurement, whichever results in the highest value of wat~r density, and, therefore, the lowest value of indicated leve 1. Compensation based on temperature is applied when the syste~ is subcooled, and compensation oased on pressur~
(saturated conditions) is applierl if superheat exists at the hot le~ te~r~ratur~ mcasure~ent point.
d~ring an accid~nt with elevated temµeraturt in th~ co~t~:*.-
m~nt an~ th0 fluirl. in th~ reference leg at th~ initial a~~:-
ent tempr.:raturr:. The compensation is based on temperatur~
meas;Jrfo-m<::nts on tii'.: vertical sections of the referen.::.:: l'.:*j.
Th0 correcterl transmitter outputs are shown on a digital display installed in th*.: contrul room, one statement for each measurement in each train. A three-pen recorder is
- M P81 88 03/4
also provided in the control room to record the level or relative d/p and to display trends in the measurements. Th~
display would also indicate which reactor co0lant pum~s ar~
operating, and which level measurements are invalid du~ t0 pump operation.
During normal plant heatup or hot standby operation with all rea'.ctor coolant pumps operating, the wide range d/p displaJ would indicate 100 perc~nt on the display, an indicatio~
that the system is water-solid. If less than all pumps ar~
- operating, the display would indicate a lower d/p (deter-*
mined during the plant startup test program) that would also te a~ injicat1on of a water-solid system. ~ith pumps oper-ating, th~ narrow range and upper range displays would indicate off-scale.
If all pumps are shut down, at any temperature, the narro~
range and upp8r range displays would indicate 100 perce~t, an indication that the vessel is full. The wide range d/~
display would indicate about 33 percent of the span of th~
display, which would be the value (determi~ed during the test program) corresponding to a full vessel with pumps shut down.
- M P81 88 03/5
In the event of a LOCA where coolant pressure has decreased to a predetermined setpoint, existing emergency procedures would require shutdown of all reactor coolant pum2s. In these cases, a level will eventually be establishr::d in th<.:
reactor vessel and indicated on all of the displays. ..
'I'. I .*
plant operator would monitor the displays an~ the record'=r to determine the trend in fluid mass or lev~l in the vess~l, and confirm ~hat the ECCS is adequately compensatin3 for th~
accident conditions to prevent ICC.
Future procedures may require operation of one or morE- pu:-:-,_;:.,s for recovery from certain types of accidents. When pu~?S cir-*::: op~reitin? w~.11*: vo11s ar': devi::lopirq in th: syst'"=':*, t:.*.
pu~~s will circulat~ th~ weit~r and stea~ as an essential~;
homogeneous mixtur~. In these: cases, there will ~ no d:~-
cernible level in the r~actor vess~l. A decrease in th0 measured d/p compared to the normal operating value will bE:
an indication. of voids in the system, and a continuously de-creasing d/p will indicat~ that the void content is increas-ing, that mass is being lost from the system. An increasing d/p will )
indicate that the mass content is increasing, that the ECCS is effectively restoring the system mass content.
M PBl 88 03/6
4.2 RVLIS ANALYSIS In order to evaluate the usefulness of the RVLIS durinq tL*:
approach to ICC, it was decided to determine the respons~ of the RVLIS under a variety of fluid conditions. The RVLIS response was analytically determined for a numb~r of small break transients. The response was determined by calculat-ing the pressure difference between the upper head and lower plenum ani cohverting this to an equivalent vessel hea~ 1~
feet. (Note that RVLIS indications will actually be re?r~-
sented by percent of span). Saturation density at the fluirl te~perature in th~ upper plenum was used for this con-version. This approximat~s the calibration that will G0 u s <c: d f or tr: '.: k */ :... 1 :::; .
This indication corresponds to th0 RVLIS configuration us~
for non-UHI plants. The indication of the upper span (no:
leg to upperhead) is not included in this analysis.
When the reactor coolant pumps are not operating, the RVLIS reading will be indicated on the narrow range scale ranging from zero to the height of the vessel. A full scale reading (100 percent of span) is indicated when the vessel i~ f~ll M P81 88 03/7
of water. This reading r~presents the equivalent collaps~d liquid level in the vessel which is a conservative indica-tion of the approac~ to ICC. The RVLIS indication can al>:rt the operator that a condition of ICC is being approachr:::J and the existance of ICC can be vcrif ied by checking the car~
exit thermocouples. When the reactor coolant pum?s ar~
operating, the narrow rangE: RVLIS meter will be pE!gged a':
full scale.
When the reactor coolant pumps are op8r~ting, the PVLIS reading will b>: ind1c:it.::*1 on the wirh: range scall'.? wr.ich reads fro~ 0 to lUU percent. Th0 lOU percent rearlin~
c 0 r re s p 0 n d s tr_, C: t j l l v ': s ~; (_: l wi t ~. al l 0 f tr l *...: p .J ::-. ;:.. ?, 1 "*
operati0r..
With the :EJUIT<!JS running, the RvLIS reading is a:-, ind1cat10:.
of the void fraction of the vessel mixture. As the void content of th~ vessel mixture increases, the density decreases and the RVLIS reading will decrease due to th~
reduction in static head and frictional pressure drop. Th0 latter effect will be enhanced by degradation in reactor coolant pump performance. When this reading drops to approximately 33 percent, there will also be an indication M P81 BB 03/B
on the narrow range scale. This fraction approximately corresponds to a vessel mass which would just cover th~ car~
if the pumps were tripped.
A srnal 1 break transient ( l inch cold leg break - no hi')~.
head safety injection; NOTRUMP) for a 4 loop, non-UH I, 3411 M~T plant is discussed in the next section. This case Wci~
obtained from the ICC analysis using NOTRUMP. A descriptio~
o~ N~TH~~? can be found in References l through 6 in S~ctio~
A discussion of this transient is provided in the next se:cti0n. Figur~s 4-1 through 4-4 provide plots of v~ss~l t~~ phas~ mixtur~ level, RVLIS narrow range readin3, mixt~r~
cind v~ss~l void fraction.
The two-phas~ mixture level plotted is that which was predict~d t/ the codes for the mixture height below the upp~r supp0rt plate. Water in the upper head is not ref lect~rl in this plot. The RVLIS reading that would be seen is plott~d on the same figure for ease of comparison .
- M PBl 88 03/9
a,b,c Figure 4-1 Case D l Inch Cold Lr~ nrcak, ICC Cnse, RVLIS Rrading and Mixtur0 10v,,l.
a,b,c Figure 4-2 Case D 1 Inch Cola Lrg BrPak, ICC CnsP~ Mixture Level, RVLIS Readin~ ana Mrasurrd Inventory.
a,b,c Fip;ure 4-3 Case D 1 Inch Cnlcl Lr*g Br('ak, ICC CnsP, RVLIS Reading and Mixture Lcvrl.
0.9
~ t
- > LEGEND:
z UJ 0.8
- - TOTAL VOID FRACTION I "' *I CL
- - - MIXTURE VOID FRACTION ,I
}. I a:
UJ 0.7 I ,1 I CL I I CL
- > I :I I 0 I 11 I z<( 0.6 I II I w
I 11 I a: I '1 I
~
0 0.5
~ I I:
I :
z 0
... 0.4 0
<(
a:
n:
11 I I
- u. ,,_JI I I 0 0.3
,_ ~1,...,. _, __ ..,.________.,I ~I II I tJ 0
UJ
~
UJ 0.2
,,~ :1-----vi :*
~
~
0.1 II 0 .__.,__~~~---~~~~~---~~~~--~~~~~---~~~~_._~~~~~...__
0 2500 6000 7600 10000 12500 16000 TIME ISECONDS) ........
U>
w Ffrnirf'.' '/--'/- Ca~e 0 1 Inch .Cold q nre.1k, ICC Case, Void Fraction l.J
The void fraction plots are for the core and upper plenum fluid volumes~ The mixture void fraction includes th~
volume be low the two phasr! mix tun~ level w~1 i l ': th'= tr;t.:.i 1 void f rac tu re al so includes the steam spa C(.: a tJry;r: thr:
mixture level.
- 4. 2. l Investigated Transient This case (4LOO?, non-UHI, 3411 MWT plant, 1 inch cold l":'.:i break-no high head safety injection) is one of the trans-ients investigat~rl for the ICC study using NOTRUMP. A mor~
detailed discussion of this transient can be found in Thr:: R'JLIS rr::adinj is b'=lo*,.,i the vessel mixtur*: l*::*:r.:l throughoJt most of the transient and is therefor~ a conservativ~ indication. The RVLIS reading follows the same trend as the vess~l mixture level, except for early in th~
transient when the mixture void fraction is fluctuating.
M P.81 88 03/10
4.2.2 Observations Of The Study The RVLIS will provide useful information for breaks in the system ranging from small leaks to breaks in the limiting small-break range. For breaks in this range, the syst~~
conditions will change at a slow enough rate that th8 operator will be able to use the RVLIS information as a basis for some action.
For larger breaks, the response of the RVLIS will b~ m~re erratic, due to rapid pressure changes in the vesst:l, in the early portion of the blowdown. The RVLIS reading will be useful for monit~ring accident recovery, when oth~r corroborative indications of ICC could also be observ~J.
Very few instances have been identified where the RVLIS may give an ambiguous indication. These include a break in the upper head, accumulator injection into a highly voided downcomer, periods of time when the upper head behaves like a pressurizer, and periods of void redistribution.
A break in the upper head may cause a much lower pressure to exist in the upper head compared to the rest of the RCS.
M P81 88 03/11
Because of this, the pressure difference between the lower plenum and the upper head is much larger than is seen for an equivalent vessel level when the break is located elsewhere in the system. The reading, in fact, may never reach th r~
narrow range scale. If the narrow range reading remains at full scale and the wide range reading is greater than that reading which would indicate a full vess~l with the reactor coolant pumps triµp~J, a break in the upper head is indi-cated. This situation should not caus~ a proble~ in aetect-l n g I CC be ca u s e o f the pa r a l l e l l o g i c f or the " k i ck o u t " t r.J
- the proced u n: s. If the RVLIS indication is to a brea(. in the reactor vessel upper heac, operator will begi~ following the ICC procedJre if erroneo~;,
th*.~
th~
selected car~ exit thermocouples read 1200°~.
This situation only exists, however, when the break discharge is large enough to cause a large d/p through the flow paths connecting the upper head to the rest of the system. These flow paths becom0 th0 limiting factor in th~
depressurization rat0.
The time, when ambiguous indications due to accumulator injec-tion and upoer M PBl 88 03/12
head pressurizer behaviour is brief. The situation corrects itself and the RVLIS resumes giving a good indication of the trend in level. Both situations result in an indication of vessel level that is low. The operator must know that a brief period of erratic RVLIS indication may occur when accumulators are injecting. This effect is partially real in that the vessel level may depress for a moment when ac-cumulator injection occurs. Unlike accumulator injection, the operator will not know when the indicated vessel levE:l is being affected by the upper head pressurizer phenomena.
However, no premature indication of ICC will occur since the core exit thermocouples will still read saturation temperature.
During periods when the void distribution in the vessel is changing rapidly, there may be a large change in two-phase mixture level with very little change in mass inventory in the vessel. This could happen if the reactor coolant pumps (RCPs) were tripped when the mixture in the vessel was high-ly voided. This could cause the mixture level to drop fro~
the hot leg elevation to below* the top of the core. The operator would expect this to happen based on the fact that the RVLIS reading was within the narrow range indicatfon~
M P81 88 03/13
- The operator should know in general that, for a brief period of time after tripping the RCPs, transient RVLIS response will occur.
Flow blockage is not expected to decrease the us~fulness of the RVLIS indication. The increased d/p due to the f lo~
blockage will be small during natural circulation. The RVLIS will continue to follow the trend in vessel level.
When the reactor coolant pumps are operating, flow blockage is not expected to occur unless the pumps had previously been tripped and are being restarted after an ICC situation
- already exists. If flow blockage were present when the pumps were running, the RVLIS indication would still useful and, although the indication would be somewhat b~
higher, would continue to follow the trend in vess0l inventory.
4.2.3 Conclusions
- 1. With the RCPs tripped, the Wstinghouse RVLIS will result in an underpredicted indication of vessel level while providing an unambiguous indication of the mass in
=
- M P81 88 03/14
the vessel. The Westinghouse RVLIS will also measur~ th~
vessel level trend reasonably well.
- 2. With the RCPs tripped, it is feasible to determin0 a setpoint for the RVLIS to warn the operator that the system is approaching an uncovered core.
- 3. The RVLIS should be used along with the core exit thermocouples to detect ICC.
- 5. When the RCPs ar~ running, and the RVLIS reading drops to the narrow range seal~, there is significant voiding in the vess~l and the core would just be covered if th~
pumps were tripped.
- 6.
- A break of sufficient size in the upper head could cause the RVLIS to give an amuiguous indication of vessel mass. The core exit thermocouples, however, will provide an indication of ICC if appropriate.
M PBl 88 03/15
- 7. Accumulator injection when the downcomer is highly void~d could recult i!1 ~temporarily erratic in1ication.
- 8. The RVLIS may si~nific~ntly underpre1i~t t~,~ ve::~l ~~==
while fluid in the upper he~1 i: fla:hin~. However, u:~
of the core exit thermocouple: will preclude a prenatur~
entry to the ICC procedur~:.
- 9. reiictributio~: will nat te dete~te~ by ~te
- ~
=
- s. Following is a description of the computer functions associated with RVLIS.
5.1 MICROPROCESSOR FOR RVLIS The microprocessor RVLIS indication~ in-elude equivalent reactor vessel lev~l on redundant flat panels with alphanumeric displays provided for control room instal-lation in addition to having this informa-tion available for display at the micro-processor chassis. RVLIS is configured as two protection sets in se?araterl sections of a single instrument rack. The envelo~0 of an instrument rack occupi~s a the base of [
J. The block d1-'l-gram of the RVLIS using microprocessor equipment is shown in Fugure 5-1. This diagram shows that in addition to the reac-tor vess~l level (d/p) transmitter input, there are also temperature compensating signals, reactor pump running status in-puts, and RCS parameter inputs to each M P81 44 03/l
a,c Figure 5~1 Reactor Vessel Level Instrument System Block Diagram (One Set of Two Redundant Resets Shown)
- chassis of the two redundant sets. The output of each set will be to displays and to a recorder, as well as an output for a serial data link. A general display arrangement is shown in Figure 5-2.
Conformance with Regulatory Guid~ 1.97 fQr the processor display system is given in Table 5.1.
5 .1.1 RVLIS Inputs The microprocessor system inputs ar~ as follows:
5.1.l.l Differential Pressure Trans~1tt~rs The: three: d/p transmitters pc:r set are use8 to measure the d/ps between the three pres-sure tap points on the primary system, as discussed below:
l.
M P81 44 03/2
- - - - - r-- - - - -l. ._-_-_@
__-_ __ v_E_ss_E_L_L_E_v_E_L_MO_N_*_To_R_ _ -_-_-_-_-_~ - - - - - ~
I I CAUTION
~- ------J I~ER 16 L- - - - - - - - - - - - - - --- - - - - - - - - - - - - - -
IRES,:JJ::OP I G fu-*R1 ITREND ID
The direction of this transmitt~r's output is full seal!? (20 ma) with th<.:
vesst:.'l full ano zero seal~ (4 ma) witr-.
the vessel emptieo to the hot leg ta~.
M PBl 44 03/3
TABLE 5.1 CONFORMANCE WITH REGULATORY GUIDE 1.97, DRAFT 2 REV. 2 ( 6/4/80) FOR THE MICROPROCESSOR DISPLAY SYSTl::."1 Seismic qualification Yes Single failure criteria Environmental qualification
- [IEEE-323-1971 applicability]
Power Source Vital Quality Assu~ance Yes 10CFR50 Appendix B applicability Display type and method Vertical scali::
voltage processed in addition to a recorde:r Unique identification Yes Periodic Test.i::_: . -.
Y r**--
.)
M PBl 44 03/4
These endpoints are nominal and are for low coolant temperatures. If no pumps arc operating, .1Pa givc:s an inr:liceitior, of level in the region abav~ th~ h0t leg.
If the pump is running in the loop wit~
the hot leg connection, this indication will be invalid and most likely off-scale. The reading would be flagged as "invalid" under these conditions. Tho::
effect on the indication from the pu~~
not running in this loop, but runn1~~
in oth:::r loops, is less tha!i 10 percr:;'_
of thE: rany*..:.
2.
M PBl 44 03/5
- 4 Pb gives an indication of reactor vessel level when no pumps are run-ning. If one or more pumps arr:
running, ~Pb will be off-scale an'] th1:
reading invcilirl.
The sense of the a\ Pb output is suc:h that a 20 ma signal is a nominally full vessel and a 4 ma signal is for a nomi-nally empty vessel:
3*
J The sense of the 4 ?c outp:Jt l~
that 20 ma represents all pum?s runn1~1 and 4 ma is empty vessel. With all pumps running and no void fraction, th~
APc should read 100 percent at zero pow0r. The reading at full power is slightly higher.
M P81 44 03/6
5.1.l.2 Reference_Leg Temperature RTD The reference leg temperature RTDs ar~ US'.:'l to measure the temperature of th~ coolant in the capillary tub~ referenc~ legs. This is used to compute th~ density of th~
reference leg fluid.
The arrangement of the reference leg temperature RTDs is shown in Figure 5-3.
The conversion of RTD resistance to te~
perature shall cover the temperature rang~
of 32° to 450°F.
The RTDs are 100 ohm platinum four wire RTDs as shown in Figure 2-1.
5.1.l.3 Hot Leg Temperature Existing hot leg temperature sensors are used to measure the coolant temperaiure.
These sensors are being rep laced year 1 y.
This temperature is used to calculate coolant d~nsity.
M P81 44 03/7
- a,b,c 5.1.1.4 Wide Range Reactor Coolant Pressure Existing or new wide range pressure sensors will be used to measure reactor coolant pressure. The pressure is used to calculate reactor coolant density.
The block diagram of the compensation func-tions is shown in Figure 5-4.
5.1.l.5 Digital Inputs The reactor coolant pump status signals in-dicate whether or not pumps ar~ running.
Recognizing that hydraulic isolators ar~
provided on each impulse line for conta1~
ment isolation purposes, each hydraulic isolator has limit switches to indicat0 they have reached the limit of travel.
5.1.l.6 Density Compensation System To provide the required accuracy for vessel level measurement, temperature measurements of the impulse line are provided. These M P81 44 03/8
a,b,c Figure 5-4 Block Diagram of Compensation Function
- measurements, together with the existing reactor coolant temperature measurements and wide rang~ RCS pressure, are emplo:;r.:rJ to compensate the d/p transduc1..:r out1~:.Jts for differences in system density and reference leg density, particularly durin3 the change in the environment insid~ th~
containment structure following an acci-dent. A simplified schematic of the d~~
sity compensation system is shown in Figure 5-5. The ~/p cells are located outsid~ th~
containment.
T 111.:- r e f f..: r e n cc: l e g f 1 u i d d e n s i t y c a l c u l :: t i *'.) --,
covf..:rs a rang0 of 32° to 450°F. The fluid is assum~j to be compressed liquid water at 120U psia.
Each of the three d/p measurements will have density corrections from certain tem-peraturf:' measurements. Some of these will have a positive correction and some nega-tive depending on the orientation of the M P81 44 03/9
a,c
-1I Figure 5-5 Simplified Schematic 6f Density Compensaticn System
impulse line where the temperature is being measured, 5.1.1.7 Vessel Liquid Density Calculation a,c 5.1.1.8 Vessel Vapor Phase Density Calculation
- a, r.
M P81 44 03/10
a,c 5.1.1.9 Vessel Levr:l Calculation
- 8. ' c 5.1.1.10 Pump Flow d/p Calculation
- ~'
--I M P81 44 03/11
The lower of the two calculated d/p corr~c tions is divided into the m~asured d/?.
Tn~ result is the percent of expected d/p anJ should read 100 percent will all pu~~s operating and no circulating v01ds.
5.1.l.ll Scaling of Displayed Value-:-;
Each of the three d/p measurements aft~r the preceding calculations shall be scaled to read in percent. With the vessel full of water and no pumps running, the outputs of Al'a and 4)Pb should read 100 perce.nt.
M PBl 44 03/12
5.1.2 Fla~~ Operator Interface and Displays Information displayed to the operator for the RVLIS is intended to be unambiguous and reliable to minimize the potential for operator error or misinterpretation. The redundant control room displays provid~ the following information:
a,c l . 1 2.
M P81 44 03/13
3.
All signals are input to a microprocessor-based data analysis system. The contr~l roo!71 dis?lciy format utilizes an alph.J-numr.:'r i c d l splay loca t~d re:not~ l y fro:-:. t!.*:
computational syst~~.
Redundant displays are provided for th~ two sets. Lev~l information based on all thre~
d/~ measurements is presented. Correction for reference leg densities is automatic.
Any error conditions such as out-of-range sensors or hydraulic isolators are auto-matically displayed on the affected measurements.
M PBl 44 03/14
- There are two display formats for reactor vessel leve 1: the first is a summary for-rnat, and the second is a trending of the three vessel level indications.
"-; c.
5 .1. 3 Display Functions for Remote Control P~~*:
The prime display unit for the vessel level monitor is the 8 line, 32 character per line alphanumeric display which is locate1 in the control room.
5.1.3.l Vessel Level Monitor Summary Display Figures 5-2, 5-6, and 5-7 give_ example dis-plays. General arrangement is shown on Figure 5-2. The vessel level summary dis-play is shown on Figure 5-6. The following is a description of the display.
M P81 44 03/15
REACTOR VESSEL LEVEL
SUMMARY
NORMAL STATUS VALUE 100% ALARM 73%
PLENUM LEVEL 4 7%. 0% INVALID VESSEL LEVEL 100% 0 FF SC A FLOW HEAD PUMPS RUNNING: 41. 42, 43, 44
- ISOLATOR ALARMS: Ll3
- DISABLED: T3 TH1
REACTOR VESSEL LEVEL TREND VESSEL F'LOW TIME PLENUM LEVEL HEAD MIN LEVEL 47% I >110% OS 00 73%
78% 4 9" I 911"
-15 79% 52% I 9 7"
-30 5 6" I 98%
-45 82%
-60 97% 99% I
~ 1qure 5- 7 Vec,c.e 1 I rve 1 *rrnd Oi sp 1oy
- l. The first line gives the title of the display as shown. The use of the underbar feature delineates this line from the rest of th~ display.
- 2. The second line gives column headings as shown. Again, the use of the underbar clarifies the display.
- 3. The third line gives the measured and normally expected values from the 6 ?a measur~ment. The first field gives the title, the second gives the measured level, the third gives the normal valur:
for the current status, and the last field g1v~s the validity status and is blank under normal conditions.
- 4. The four th line gives the 4 Pb measure-ment results using the same format as in line J.
- 5. The fifth line gives the 6Pc measure-ment results using the same format as M P8l 44 03/16
in line 3. The use of underbar in line 5 delineates this line from the next.
- 6. The sixth line gives the status of th~
pumps as seen by the unit. The runn1n3 pumps are identif i~~.
7-8. The seventh line and eight lin~ ar~
normally left blank and are reserved for hydraulic isolator limit switch indicators, out of range sensors and operator disabled sensors.
5.1.3.2 Trend Display The trend display for th~ vessel l~vel monitor shall usr~ the format shown in Figure 5-7.
5.1.3.3 Displays on Main Processing Unit The one-line forty character alphanumeric display on the front panel of the main processing unit is used to display indi-M P81 44 03/17
vidual sensor inputs. The sensor is selected with a two digit thumbwheel switch.
The following information is given for each sensor:
- l. Sensor identification 2* Input signal level
- 3. Input signal converted to engineering units
- 4. Status of sensor input
- s. l. 4 Disabled Inputs Any inputs can be disabled by the oper-ator. This action is under the control of a keyswitch on the front panel of the main computational unit and causes the processor to disregard the analog input for that variable.
M P81 44 03/18
- 6. The current schedule for installing, testing, and cali-brating the R VLIS is during the first refueling outaqe for this unit.
M P81 55 04/1
- 7. Guidelines for the use of the RVLISA and analyses used to develop these procedures.
7.1 Reference Owners Group Procedures Based on the analyses defined in Section 7.1.1 below and Section 4.2 of this report, Westinghouse Owners Group have developed a Reference Emergency Operating Instruction to address recovery from ICC conditions caused by a small break LOCA without high head safety injection. This instruction has been transmitted to the NRC via Westinghouse Owners Group letter OG-44, dated 11/10/80.
7 .1.1 Conditions or Events which Describe the Approach to ICC The most obvious failure that would lead to ICC during a small break LOCA, although highly un-realistic since *multiple failures are required is the loss of all high pressure safety injection.
The approach to ICC conditions and the analyses of t~is event sequence are provided in Section A, References 1 and 2.
M P81 55 04/2
7.2 Sample Transient The response of the vessel level indications and system response during these ICC events and recovery actions are described in Section A, References 1 and 2.
M P81 55 04/3
A* REFERENCES
- 1. Thompson, C. M., et aL, *inadequate Core Cooling Studies af Scenarios with Feedwater Available, Using the NOTRUtf> Ccrnputer Code,* W::AP-9753 (Proprietary) and WCAP-9754 (Non-Proprietary), July 1980.
- 2. Mark, R. H., et al., *Inadequate Core Cooling Studies of Scenarios with Feedwater Available for UHI Plants, Using the NOTRUl'I' Canputer Code," W::AP-9762 (Proprietary) and WCAP-9763 (Non~Proprietary), June 1980.
- 3. *Report on Small Break Accidents for Westinghouse Nuclear Stear.;
Supply System: WCAP-9600 (Proprietary) and WCAP-9601 (Non-Pro-prietary), June 1~79.
- 4. Esposito, V. J., Kesavan, K., and Maul, B. A., *WFLASH - A FORTRAN-IV Canputer Program for Simulation of Transients in a Multi-Loop PWR,* W:AP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision l (Non-Proprietary), July 1974.
- 5. Skwarek, R., Johnson, W., and Meyer, P., *westinghouse Emergency Core Cooling System Small Break October 1975 Model,* WCAP-8970 (Pro-prietary) and WCAP-8971 (Non-Proprietary), April 1977.
- 6. *Analysis of Delayed Reactor Coolant Pump Trip During Small Loss of Coolant Accident for Westinghouse NSSs,* W.CAP-9584 (Proprietary) anj WCAP-9585 (Non-Proprietary), August 1979.