ML20040C266

Application for Withholding Proprietary Summary Rept: Westinghouse Reactor Vessel Level Instrumentation Sys for Monitoring Inadequate Core Cooling (Microprocessor Sys).

ML20040C266
Person / Time
Site:	North Anna
Issue date:	12/23/1980
From:	Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML20040C263	List: ML20040C262 ML20040C266
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM AW-77-18, CAW-80-76, NUDOCS 8201270471
Download: ML20040C266 (7)

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Westinghouse Water Reactor wer 'ecmcicar ca 5:ca Electric Corporation Divisions e ,33

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mosmsome two December 23, 1980 CAW-80-76 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Phillips Building 7920 Norfolk Avenue Bethesda, Maryland 20014 .

ATTN: Lawrence E. Phillips Core Performance Branch, DSI APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

SUBJECT:

Summary Report, Westinghouse Reactor Vessel Level Instrumentation System for Monitoring Inadequate Cor,e Cooling (Microprocessor System)

REF: NURIG-0737PartII.F.2,InstrumentationforInadequateCoreCooling

Dear Mr. Eisenhut:

The proprietary material transmitted by the referenced letter supplements the proprietary material previously submitted concerning the Westinghouse development of ECCS models. Further, the affidavit submitted to justify the material previously submitted, AU-77-18, was approved by the Commis ion on Octotser 28, 1977, and is equally applicable to this material.

Accordingly, withholding the subject information from public disclosure is requested in accordance with the previously submitted affidavit and appli-cation for withholding, AW-77-18, dated April 6,1977, a copy of which is attached.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference CAW-80-76, and should be addressed to the undersigned. -

Very truly yours,

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'Io Rot 't diesemann, Manager Attacht.nt D Regulatory & Legislative Aff. irs cc: E. C. %omaker, Esq.

Office of the Executi : L3 11 Director, NRC 8201270471 811106 DR ADOCK 05000338 PDR

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AW-77-18 AFFIDAVIT COPN0!! WEALTH OF PEililSYLVA!!IA:

ss COUliTY OF ALLEGHEllY:

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 Westingho'use Electric Corporation (" Westinghouse") and that the aver-

. ments of fact set forth in this Affidavit are true are terrect to.the best of his knowledge, inf6rmation, and belief:

L W 0 b2'utl8%:L-Robert A. Wiesemann, innager Licensing Programs-Sworn to and subscribed before me this el 0 day of M "!cI 1977. .

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AW-77-18 ,

(1) I am Manager, Licensing Programs, in the Pressurized Water Reactor Systems Division, of Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public dis-closure in connection with nuclear power plant licensing or rule-making proceedings, and an authorized to apply for its withholding on behalf of the Westinghouse Water Reactor Divisions. .

I (2) I am making this Affidavit in conformance with the provisions of ,

10 CFR Section 2.790 of the Commission's regulations and in cen-junction with the Westinghouse application for withholding ac- ,

, companying this Affidavit. ,  ;

(3) I have personal knowledge of the criteria and procedures utilized

. by Westinghou: !!u.;1:.-"-7--- y Systems in designating information as a trade secret, privileged or as confidential commercial er ,-

financial information. -

(4) pursuant to the provisions of paragraph (b)(4) of Section 2.790

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of the Commission's regulations, the follcuing is furnished for ,

consideration by the Commission in determining whether the in-

,. formation sought to be withheld from public disclosure should be withheld. '

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(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. .

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AW-77-13 (ii) The information 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 information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The ap-plication of that syst'em and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

. Under that systeh, 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 ccm-petitive advantace, as, follcus:

.s (a) The information reveals the dif tinguishing aspects of a process (or component, structure, tool, method, etc.)

where prevention of its use by any of Westinghouse's

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competitors without license from Westinghouse constitutes i

a competitive economic advantage over other companies, (b) It consists of suppor. ting data, including test data, relative to a process (or component, structure, tool, methe f, etc.), the application of which data secures a l

competitive economic advantage, e.g., by optimization or improved marketability. .

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• AW-77-lO (c) Its use by a competitor 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 information, production cap-acities, budget levels, or commercial strategies o'f Westinghouse, its customers or suppliers. .

(e) It reveals aspects of past, present, or future West-inghouse or customer funded development plans and pro-grams of potential comr.ercial value to Westinghouse.

(f) It contains patentable ideas, for which patent pro-tection may oe oesirable.

(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 Westinghouse a -competitive advantage over its com-petitors. It is, therefore, withheld from disclosure to protect.the Westinghouse competitive position.

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AW-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 compet'itor would put Westinghouse at a competitive disadvantage by reducing his expenditure

.. of resources at our expense.

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(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If m '

competitors acquire components of proprietary infor-mation, any one component may be the key to the entire puzzle,- thereby depriving Westingnouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the po ition of prominence of Westinghouse in the world market,

. and thereby give a market advantage to the competition 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 advantage.

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• AW-77-18 (iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 C7D Section 2.790, it is to be received in confidence by the Ccr:.T.ission.

(iv) The information is not available in public sources to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is attached to Westingh:use Letter Number NS-CE-1403, Eiche1dinger to Stolz, etted April 6, 1977. The letter and attachment are being ss!t-itted in -

support of the Westinghouse emergency core cooling system ,

evaluation model. ,

Public disclosure of the inforcation sought tc be withheld is likely to cause substantial harm to the ccrgetitive

. position of Westinghouse, taking into account the value of the information to Westinghouse, the amount of effort and_

. . money expended by Westinghouse in developing the information, and considering the ways in which the information could be acquired or duplicated by others.

Further the deponent sayeth not. ,

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RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION ON THE WESTINGH0l;SE R.V.L.I.S.

SUMMARY

REPORT (u Processor) l l

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l Item 1 j

i Justify that the single upper head penetration meets the single failure requirement of NUREG-0737 and show that it does not negate the redun-dancy of the two instrument trains.

Response

1. Redundancy is not compromised by having a shared tap since it is not conceivable that the tap will fail either from plugging or break-ing. Freedom from plugging is enhanced by,1) use of stainless steel connections which preclude corrosion products and, 2) absence of mechanisms, such as, flow for concentrating boric acid. It is also inconceivable that the tap will break because it is in a pro-tected area. It should also be pointed out that in other cases where sharing of a tap occurs in the RCS, we know of no prior experience reporting deleterious malfunctions of the shared tap. .

Also, even if the shared tap does fail, it should be recognized that RVLIS is not a Protection System initiating automatic action, but a monitoring system with adequate backup monitoring'such as by core exit thermocouples for operator correlation. .

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Item 2 Describe the location of the level system displays in the Control Room with respect to other plant instrument displays related to ICC monitoring, in particular, the saturation meter display and the core exit thermocouple display.

Response (

2. The RVLIS displays in the Control Room are located in the upper portion of Vertical Board PAMC-1 for North Anna Unit I and PAMC-2 for North Anna Unit 2. This is a 35 inch wide panel containing other Post Accident Monitoring instrumentation. This location was chosen so that the RVLIS displays can be op? rated in conjunction with the Re. actor Vessel Head Vent System, which is controlled at this panel.

The Inadequate Core Cooling Saturation Meters are located on. Vertical Board Section 1-1 for florth Anna Unit 1 and Section 1-1 for North Anna Unit 2, directly below the Boric Acid Tank Level Indicators. 'This' location was chosen based upon the proximity of Reactor Coolant System pressure and temperature data and the available panel space.

The Incore Exit Thermocouple Meter is located on Incore Instrument Panel 1-EI-CB-95D for North Anna Unit 1. This panel is located in the Control Room to the !!ortheast side of V::rtical Panel 1-1 along the Computer Room wall. ~ The Incore Exit Thercoccuple Meter for North Anna Unit 2 is located on Incore Instrument Panel 2-EI-CB-960 which is located to the Southwest side of Vertical Board 2-1 along the Computer Room wall. The location of these panels was chosen during the original design of the plant.

The location of each device with respect to each other is nearly in line (see attached figures). The RVLIS displays are located to the right of the Inadequate Core Cooling Saturation Meters. The Incore Exit Thermocouple Meter is located to the lef t of the Inadequate Core Cooling Saturation P.eters for florth Anna Unit 1 and to the right or -

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. e the RVLIS displays for florth Anna Unit 2.

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Item 3

' l Describe the provisions ad procedures for on-line verification, cali- l bration and maintenance. ,

Response '

3. In general, the system electronics are verified, maintained and calibrated on-line by placing one of the redundant trains into a test and calibrate mode while leaving the other train in operation to monitor inadequate core cooling.

A general verification is performed before shipment, but plant specific data is not used. The capability exists for the operator to verify the operation of the system. This would involve discon-necting the sensors at the RVLIS electronics, providing an arti-ficial input, and observing the response of the system on the front panel and remote display.

On-line calibration of the system is made possible by the controls available on the main processing unit. The calibration consists of entering constants into the non-volatile RAM along with adjusting the potentiometers on the analog to digital conversion cards. The initial calibration is done when the system is installed, but subse-quent calibrations can be perfomed as described in the Technical Manual to maintain system accuracy.

The RVLIS system requires the nomal maintenance given to other control and protection systems within the plant. On-line mainte-nance is accomplished by placing only one of the two redundant trains into maintenance at a time this will allow continued moni-toring of inadequate core cooling.

In addition, software programs are provided so that the front panel controls and display can be used to perfom a functional test, serial data link tests, calibration tests and deadman timer tests.

These tests are considered part of.the operator maintenance proce-dures and should be performed monthly. For additional details of procedures see " Attachment A".

ATTACHMENT A SYSTEMS OPERATING PROCEDURES 2-1. PURPOSE The objectives of these instructions are to establish the requirements for the use of the Reactor Vessel Level Instrumentation System (RVLIS) for various plant conditions and to specify the maintainability require-ments of the system equipment.

2-2. PREREQUISITES o The capillary lines have been vacuum filled, per the instruc-tions of section 4.

o Ensure that the hydraulic isolators are zerced (within plus or minus 0.1 in.3),

o Calibrate the d/p cells per instructions of ITT Barton Manual '

for Model 752, Level 8, transmitters. '

o The process equipment must be scaled using the appropriate scaling document.

o Determine the height of the upper top piping above the inside top of the vessel, t

i 2-3. INITI ALIZATION With the plant less than 2000F and less than 430 psig, obtain the following data for trains A and B:

(1) With an automatic data 'soggcr, record the following:

o T hot o RCS pressure o d/p transmitter output

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o Signal to the remote display (2) Manually record:

o Level indication readings o Hydraulic isolator diai readings o Reference leg RTD output (3) Record the above data for the following reactor coolant pump operations:

NOTE i

The various configurations should be obtained through the normal startup if possible. .

NOTE Upper plenum will read offscale if pump is running in the instrumented loop; narrow range will read offscale with one or more pumps running.

o No pumps running NOTE l An indication of 100 percent reading repre-sents a level to the inside top 'of the t

vessel. The height of the upper top piping above the inside top of the vessel will result in a reading greater than 100 per-cent. This added height is plant specific and must be determined prior to adjusting the process equipment (upper plerum and narrow range) for full scale indication.

o One non{nstrumentej loop pump running o Two noninstrumented loop pumps running o Two noninstrumented loop pug s and one instrumented loop pug running o All pumps running -- Adjust process equipment so that wide range indication reads 100 percent.

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(4) With all pumps running, increase RCS pressure - temperature to T ayg no-load and record data refer to step (1) every 500F increment. Data of step (2) should be recorded at 3500F and at T avg no-load. Adjust process electronics for density cog ensation at T avg no-load. Verify that wide range indication reads 100 percent.

(5) Trip all pumps and record data per steps (1) and (2).

Verify that upper plenum and narrow range indication is in agreement with the reading of step (3) "No pumps running".

(6) Restart pumps in sequence and record wide range readings '

tor both trains for each pump combination.

(7) Enter into the equipment programing the expected percent level for the various pug combinations per the micro-processor instruction manual.

2-4 NORMAL PLANT OPERATION l

With the plant at power, the level readings should be as follows:

Wide range #110 percent (wide range reading will increase from 100 percent to approximately 110 percent with all pumps running, as reactor power is increased from zero to 100 percent)

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Narrow range Off Scale - High Upper plenum Off Scale - Low (RCP status light on main control board is off)

Any reduction in wide range expected readings (with all pumps running) can only be caused by the presence of voids in the circulating water.

Voids will not exist without reduced pressure which could trip the reactor, so all accident conditions will proceed from a condition of zero power (100 percent reading on the wide range). Check that the pressure has decreased or that subcooling meter confirms saturation conditions exist; then readings below 100 percent are an indication of voids in the coolant.

If the actual readings differ from the expected readings by 3 percent "or a single train, refer to troubleshocting (paragraph 2-10).

If the indication for both trains differs from the expected readings, refer to the emergency operating instructions for imediate and subse-quent action.

2-5. REFUELING ,

Af ter depressurization and prior to lif ting the reactor vessel head, perform the following steps to prepare the RVLIS:

(1) Close reactor vessel level head connection isolation valve.

i (2) Disconnect piping between the isolation valve and the sensors.

NOTE Contaminated water residue may be in the l pipe.

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. o (3) Provide temporary plugs for the pipe ends of the remoeable secticn and stationary sections.

Restore the RVLIS af ter reactor vessel head installation as follows:

(1) Remove pipe end plugs and reconnect piping section."

(2) With the isolation valve open, backfill the piping from sensors by attaching a water source to the sensor vent.

(3) Disconnect waterfill apparatus.

(4) At startup (450 psig, <2000F), visually inspect piping / coupling of the reinstalled piping for leakage.

(5) At full system pressure, repeat inspection.

2-6. PERIODIC TESTING 2-7. Plant at Power Perform monthly calibration checks of the process electronics in accor- '

dance with the process equipment instruction manual.

2-8. Refueling Outages (1) For tne d/p transmitters, perform zero check of each d/p trans .itter by closing the respective isolation valves and opening the bypass valve. If zero reading differs from the last recorded reading by percent, then recalibrate d/p tr ans.aiter using instructions of Barton Instruction Manual (Model 752) and the instructions contained in the RVLIS system manual and the appropriate equipment instruction manuals.

(2) Record the appropriate hydraulic isolator dial readings and compare results with previous cold shutdown readings.

Readings should be within plus or minus 0.1 in.3 (3) Perform the calibration check of the process electronics in accordance with the equipment technical manual.

(4) Verify the operability of the RVLIS System during the startup/heatup of plant following a refueling or major plant outage by tracking the displays of the two trains.

Readings should be within percent of the previous recorded readings.

2-9. Every Other Refueling Outage In addition to the steps of paragraph 2-8, perform the following every other refueling outage:

(1) At the proce'ss equipment cabinets, read the impulse line RTD resistances.

- NOTE Take the ambient temperature reading near the RTD and adjust the measured resistance accordingly. Compare the adjusted resis-tance to the original results or the previous recorded data.

(2) Employing a pneumatics calibration, per instructions of section 4 at the sensor vent ports, check the calibration of the transmitters and perform a time response check of the system. The calibration results should be within plus or minus percent of instrument span of the previous recor-ded data. The time response of the system should be within 10 seconds. This is the time required for the display instrument to reach the midpoint of a 50 percent step input variable change.

2-10. TROUBLESHOOTING, PLANT AT POLTR If single indication varies from the expected value, check the following:

, (1) Call for the sensor status display for any abnormalities.

(2) Compare hydraulic isolator dial reading with reading taken from diverse train and those taken at T ayg no-load condi-tions. Dial readings deviating by more than plus or minus 0.1 in.3 may be indicative of potential capillary line leakage; however, it may not be the reason for the devia-tion in the display reading until the isolator reached the valve-off point.

(3) Perform a calibration check of the process equipment, per the appropriate instruction manual.

(4) Perform a zero check of the appropiate d/p transmitter.

If more than one indicator / display deviates from the diverse train or ~

from T avg no-load readings, check the following:

o Comon isolator dial readings versus previous reading o d/p transmitter valve lineup o Process equipment power supplies If repairs are required to the capillary lines, the system must be vacuum-filled and calibrated per the instructions contained in section 4 and the appropriate equipment irstruction manuals.

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Item 4 Describe the diagnostic techniques and criteria to be used to identify malfunctioning components.

Response ~

The microprocessor based RVLIS performs internal diagnostic' checks of the non-volatile RAM, non'-volatile PROM and other microprocessor compo-nents. No operator interface is required for these internal checks which are performed in each cycle. A " deadman" circuit is provided to detect microprocessor failure. This circuit will indicate a processor problem on the front panel of the unit and automatically reset the CPU to restart the microprocessor. The remote display unit of the RVLIS indicates the status of the input sensors. If any sensor _ is out of range or disabled a symbol will follow the affected level reading on the summary display page. In addition, sof tware programs are provided so that the front panel controls and display can be used to perform a func-tional test, serial data link tests, calibration tests and deadman timer test. These tests are considered part of the operator maintenance pro-cedures and should be performed monthly.

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12em 5 Estimate the in-service life under conditions of normal plant operations and describe the methods used to make the estimate, and the data and sources used.

Response

The 'in-service life of the RVLIS Microprocessor based electronics is dependent upon proper maintenance, iacluding the replacement of individual component parts when necessary. The provisions for this maintenance are included in the technical manual. Based on the assumption of normal conditions and proper maintenance of the components, the only limitation to the in-service life will be the availability of replacement parts. It is estimated that in 20 years, some of the components will be technically obsolete and no longer produced. Consequently, the cards may have to be modified in the future to accommodate the current technology. Thus, any individual component f ailures are regarded as maintenance considerations and their replacement is necessary to prolong in-service life.

In-Service life which is different than Design Life and Qualified Life is dependent upon implementing a scheduled preventative maintenance program including periodic overhaul of the equipment. In this manner, the equipment is restored to a level that continual operatibility fe ensured. In developing the maintenance program, repair costs may necessitate replacement of the equipment.

If the maintenance program is followed there is no apparent reason that-operation of the equipment cannot be extended.

Some of the equipment is similar to equipment installed in present Westinghouse plants that have been operating for 10-15 years.

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~h: failcwing valva: STva beeq s :pplic; hj '/e :ingnause f:- the ~ ::::ar Vessel Level Instrumentation System for iiorth Anna Units ; and .

1 W Cesign Code

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W Valve ID Oty Manuf acturer Soecification Acolicability 3/4 T 78 4 Rockwell G-952855; Rev 0 ASME B&PV Class II 1/4 X 28I -

10 Autoclave Engineers G-955230; Rev 2 N&S 1/4 N 281* 6 Autoclave Engineers G-955230; Rev 2 N&S

• Shut off valve which is part of the transmitter access assembly.

The 3/4 T78 valve is a stainless steel, manually operated globe valve whose basic function is to isolate the flow of fluid. The valve is designed for a cycle life of 4000 cycles over the 40 year design life, which satisfies the normal plant operating requirements established in above referenced specification. The vave is a hermetically sealed valve, designed to be maintenance free with no consumable materials making a pressure boundary seal.

The instrumentation valves (W Valve ID's 1/4 x 281 and 1/4 N281) stainless steel, manually operated valves, designed to meet the requirements of the above referenced specification, which calls for zero leakage (environmentally and across the seats), minimai fluid displacement during stoke and a 1000 cycle life. .For normal plant operating conditions, the metallic parts are designed for a 40 year service life. The consumable itens, where applicable, are identified in the appropriate drawings and instruction manua'ls, with recantended maintenance schedules. .

Item 6

- Explain how the value of the system accuracy (given as +/- 6% was derived. How were the uncertainties from the individual components of the system combined? What were the random and systematic errors, assumed for each component? What were the sources of these estimates?

Response

6. The system ac:uracy of 16% water level was a target value established during the conceptual design and was related to the dimensions of the reactor vessel (12% from nozzles to top of core) and core (30%), and the usefulness of the measurement during an accident. Suosequent analyses have established a system accuracy based on the uncertainties introduced by each component in the instrument system. The individual uncertainties, resulting from random effects, were combined statistically to obtain the overall instrument system accuracy. Some of the individual uncertainties vary with conditions such as system pressure. The following table identifies the individual uncertainties for the narrow range ,

measurement while at a system pressure of 1200 psia.

Uncertainty Comoonent and Uncertainty Definition  % Level

a. Differential pressure transmitter 1 2.1 calibration and drift allowance, (11.5% of span) multiplied by the ratio of ambient to operating water density.

b. Differential pressure transmitter 1 0.7 allowance for change in calibration due to ambient temperature change (10.5% cf span for 1500F) multiplied by the density ratio.

c. Dif t erential pressure transmitter 1 0.34 allowance for change in calibration due to change in system pressure (10.2% of span per 1000 psi change) multiplied by the density ratio, ,

d. Differential pressure transmitter 1 0.7 allowance for change in calibration due to exposure to long-tenn overrange (10.5% of span) multiplied by the density ratio.

e. Reference leg temperature instrument 1 0.64 (RTD) uncertainty of 1 50F and or allowance of 1 50F for the difference between the measurement and the true ,

average temperature of the reference leg, applied to each vertical section of the reference leg where a measurement is made. Stated uncertainty is based on a maximum containment temperature of 4200F, and a typical reference leg installation.

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f. Reactor coolant density based on auc- 1 2.3 tioneering for highest water density obtained from hot leg temperature j (1 60F) or system pressure (+ 60 psi).

l Magnitude of uncertainty varies with system pressure and water level, with l

l largest uncertainty occurring when the reactor vessel is full.

g. Sensor and hydraulic isolator bellows j;1.46 displacements due to system pressure _

changes or reference leg temperature changes will introduce minor errors in the level measurement due to the small volumes and snall bellows spring constants. The changes, such as pressure or temperature, tend to cancel, i.e., the bellows associated with each measurement move in the same direction. Maximum expected error due to differences in capillary line volume and incal tempera-tures is equivalent to a level change of about 5 inches, multiplied by the density ratio.

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h. Density function generator output mis- f;0.50 match with ASME Steam Tables limited to a maximum of:

i. Electronics syste'm calibratien, overall f; 1.0 uncertainty limited to less than:

j. Control board indicator resolution; f; 0.5

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microprocessor digital readout to nearest percent of level span.

The statistical combination of (square root of the sum of the squares) of the individual uncertainties described above results in an overall system instrumentation uncertainty of f; 3.9% of the level span. For the narrow range indication of approximately 40 feet, or f; 1.5 feet, at a system pressure of 1200 psia. Examples of the uncertainty at other system pressures are:

Uncertainty = j; 3.6% at 400 psia Uncertainty = j; 4.2% at 2000 psia

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' Assume a range of sizes for "small break" LOCA's. What are the relative times ava.ilable for each size break for the operator to initiate action to recover the plant from the accident and prevent damage to the. core?

What is the dividing line between a "small break" and a "large break"?

Response

7.

Inadequate core coolant (ICC) was defined in WCAP-9754, "Indequate Core C,ooling Studies of Scenario With Feedwater Available Using the NOTRUMP Computer Code", as a high temperature condition in the core such that the operator is required to take action to cool the core before significant damage occurs. During the design basis small loss of coolant accident, the operator is not required to take any action to recover the plant Other than to verify the operable status of the safeguards equipment, trip the reactor coolant pump (RCPs) when the primary side pressure has decreased to a specific point, and initiate cold and hot leg recirculation procedures as required.

In the design basis small LOCA, a period of cladding heatup may

, ' occur prior to automatic core recovery by the safeguards equipment.

The heat up period is dependent upon the break size and ECCS perfor-mance.

An ICC condition may arise if there is a failure of the safeguards equipment beyond the design basis. In that case, adequate instru-mentation exists in the North Anna plant to diagnose the onset of ICC and to determine the effectiveness of the mitigation actions taken. The instrumentation which may be used to determine the ade-quacy of core cooling consists of a subcooling meter, Core Exit I

Thermocouples (T/Cs), and the Reactor Vessel Level Instrumentation System (RVLIS).

For a LOCA of an equivalent size equal to approximately six inches or less, an ICC condition can only occur if two or more failures occur in the ECCS. As indicated in WCAP-9754, an ICC condition can be calculated by hypothesizing the failure of all high head safety 9

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injection (HPSI) for LOCAs of approximately one inch in size. For a 4 inch equivalent size LOCA one can' hypothesize an ICC condition by assuming the failure of all HPSI as well as the failure of the passive accumulator system (a truly incredible sequence of events).

For LOCAs of sizes of six inches or less, the approach to ICC is unambiguous to the reactor operators. The first indication of a possible ICC situation is the indication that some of the ECCS pumps have failed to start or are not delivering flow. The second indica-tion of a possible ICC situation is the occurrence of a saturation condition in the primary coolant system as indicated on the subcool-ing monitor. Shortly after the second indication, the RVLIS would start to indicate the presence of steam voids in the vessel. At some point in time the RVLIS will indicate a collapsed liquid level below the top of the core. The core exit thermocouples will begin to indicate superheated steam conditions. If appropriate the RVLIS and core exit T/C behavior will provide unambiguous indications to operator to follow the ICC mitigation procedure.

WCAP-9754 indicates that the selected core exit T/Cs will read 12000F at approximately 11000 seconds af ter the initiation of a 1-inch LOCA with the lors of all HPSI. The Generic Westinghouse E0P Guideline instruct operator to pursue ICC mitigation procedures when these conditions are reached. The 4-inch LOCA will indicate 12000F at about 1350 seconds. By following the Westinghouse reconnended Emergency Operating Procedures (EOPs), the operators will have earlier indication of a possible ICC situation. Recovery procedures to depressurize the primary below the core pressure safety injection shutoff head may be followed. These procedures include correction of the HPSI failure, opening steam dump, or open-ing pressurizer PORVs. The RCPs may be restarted to provide addi-tional steam cooling flow.

Large break LOCAs consist of LOCAs in which the fluid behr.vior is l inertially dominated. Small break LOCAs, on the other hand, have

the fluic cabre er dominated by gravitatianai ef acts. Fe- LC:As which are significantly larger than an equivalent 6-inch break, the ECCS has the maximum potential for flow delivery since the primary coolant system is at low pressure.

No early manual action is useful in recovering from ICC. Ana' lyses for LOCAs in this range indicate ambiguous behavior of the core exit T/Cs and RVLIS early in the accident due to dynamic blowdown effects. This behavior is temporary and the core exit T/Cs and the RVLIS will indicate the progress being made by the ECCS in recover-ing the core. When the core exit T/Cs and RVLIS may be temporarily providing ambiguous indications, no manual action is needed or use-ful. Later in the accident when manual action may be useful, the core exit T/Cs and RVLIS will provide an unambiguous indication of ICC if it exists. This unambiguous indication may be present as early as 30 seconds af ter the ' initiation of the LOCA for a double ended guillotine rupture or a main coolant pipe.

The limiting small break size is normally found to be between 3 and 4

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inches. Westinghouse typically envelopes

• this situation with calculations up to 6 inches. Breaks between 6 inches and 18 inches 'are typically not limiting. Large breaks become a concern around 18 inches.

Item 8 Describe how the system response time was estimated. Explain how the response times of the various components (differential pressure trans-ducers, connecting lines and isolators) affect the response time. _

Resoonse

8. The microprocessor reads all the inputs every five seconds and updates the digital display and analog outputs within four seconds af ter the inputs are read. Thus, a worst case time from analog input change to display and analog output change is 9 seconds. Any analog delays due to the front end electronics, sensor electronics, sensor mechanics, impulse lines, hydraulic isolato'rs, etc., have five seconds to settle out. Thus, analog delays only add to the 9 second worst case response time if they are longer than 5 seconds.

The front end electronics of the microprocessor system has a time constant less than 0.5 seconds, and the total analog delays due to the sensor electronics, mechinics, impulse lines and hydraulic iso-lators are less than 3 seconds. Therefore, the worst case response

- time is 9 seconds for the system.

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Item 9 The-e are indications that the TMI-2 core may be up to 95% blocked.

Estimate the effect of partial blockage in the core on the differential oressure measurements for a range of values from 0 to 95% blockage.

Resoonse ,

9 Blockage in the core will increase the frictional pressure drop and increase the total differential pressure across the vessel. This will be reflected as a higher RVLIS indication. The increase in the RVLIS will be most significant under forced flow conditions when the reactor coolant pumps are operating.

In order for blockage to be pre'sent, the core would have to have been uncovered for a prolonged period of time. A low RVLIS indication along with a high core exit thermocouple indication would have been indicated during this time. If the RCP's had been

• operating throughout the transient, there would have been sufficient cooling to prevent significant core damage. Therefore, for l

significant blockage t'o' exist during pump operation, the operator would have restarted the pumps after an ICC condition had existed for a period of time. Based on the history of the transient, the operator muld know that the RVLIS would read higher than expected.

Although the RVLIS would read high, it would still follow the trend in vessel inventory. The operator would be able to monitor the recovery with the RVLIS.

l Under natural circulation conditions, the impact of core blockage is not expected to be large. Although the RVLIS indication will read slightly higher than normal, the RVLIS will still trend with the vessel inventory and provide useful infonnation for monitoring the recovery from ICC. ICC will have been indicated at an earlier time; I before a significant amount of core blockage has occurred. The coerator will know that the RVLIS could read slightly high, based on the history of the transient.

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Item 10 Describe the eff ects of reverse ficws within the reactor vessel on the

. indicated level. .

Resoonse

10. Reverse flows in the vessel will tend to decrease the DP across the vessel which would cause the RVLIS to indicate a lower collapsed level than actually exists. The low indication would not cause the operator to take unne'cessary actions, since the RVLIS would be used along with the core exit thermocouples to indicate the approach to ICC. It is important to note that large revec e flows are not expected to occur for breaks smaller than 6" in diameter during the time that the core is uncovered. Large reverse flow rates may occur

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early in the blowco'.<n transient for large diameter breaks but, as is discussed in the response to Item 7, it is not necessary to use the RVLIS as a basis for operator action for breaks in this range.

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Item 11 What is the experience, if any, of maintaining O/p cells at 300% over-range for long periods of time?

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Response

11. Experience in overranging of D/p Instruments has been obtained in previous applications of D/p capsules similar to those used in RVLIS. In Dual Range Flow (D/p) Applications the " Low Flow" trans-mitter (and/or gages) are overranged to 300% or greater by normal flow rates yet provide reliable metering when required for startup.

Also, test data exists on the basic transmitter design showing about 0.5% effect on calibration with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exposure to 3000 psig over-range. All units are similarly exposed to this overrange for 5 min-utes in both directions as a part of factory testing.

There have been instances involving accidental overrange of these instruments (including RVLIS) as the result of leakage or' operator errors where full line pressure overranges have occurred for up to several weeks with minimal effect on instrument accuracy.

Based upon this experience and test data we expect to prove statis-tically that reliable measurements can be made by the selected over-4 ranged instrument designs used for RVLIS. On line calibration capa-bility is provided if needed to support gathering of statistical data.

Iten 12 Five ccnditions were identified which could cause the DP level system to pive ambiguous indications. Discuss the natura of the ambiguities for

1. accumulator in.iection into a highly voided downcomer, 2. when the uoper head behaves like a pressurizer, 3. upper plenum injection, and

4. periods of void redistribution. ,

Resoonse

12. 1. When the downcomer is highly voided and the accumlators inject, the cold accumulator water condenses some of the steam in the downcomer which causes a local depressurization. The local depressurization will lower the pressure at the bottom of the vessel which will lower the DP across the vessel, causing an apparent decrease in level indication. The lower pressure in the downcomer also causes the mixture in the core to flow to the lower plenum, causing an actual decrease in level. The period of time when the RVLIS indication is lower than the actual collapsed liquid level will be brief.

An example of when this phenomenon may occur is when the reactor coolant pumps are running for a long period of time in a small break transient. After the RCS loops have drained and the pumps are circulating mostly steam, the level in the downcomer will be depressed. A large volume of steam will be present in the downcomer, above the low' mixture level, which allows a large amount of condensation to occur. For most small break transients, the reactor coolant pumps will be tripped early in the transient and the downcomer mixture level will remain high, even in cases where ICC occurs. When the downcomer level is high the effect of accumulator injection on the RVLIS indication will be minor.

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2. When the uccer head begins to drain, the pressure in the upper head decreases at a slower rate than the pressure in the rest of the RCS. This is due to the upper head region behaving much like the pressurizer. The higher resistance across the upper suoport plate relative to the rest of the RCS prevents the upper head f rom draining quickly. This situation only exists until the mixture level in the upper head falls below the top of the guide tubes. At this time, steam is illowed to flow from the upper plenum to the upper head and the pressure equilibrates.

While the upper head is behaving like a pressurizer, the vessel differential pressure is reduced and the RVLIS indicates a lower than actual collapsed liquid level.

This phenomenon is discussed in the sumary report on the RVLIS*

relative to the three inch cold leg break. Since that time, the upper head modeling has been investigated in more detail. It was found that the modeling used at that time assumed a flow resistance that was too high for the guide tub'es. Subsequent analyses have shown that the pressurizer effect has less impact on the vessel dp than was originally shown. There is very little impact on the results af ter the level drains below the i

top of the guide tubes. The pressurizer effect is still l believed to exist and it becomes more significant as break size increases. The interval of time when the upper head behaves like a pressurizer is brief and the RVLIS will resume trending with the vessel level after the top of the guide tubes uncover.

The reduced RVLIS indication w'ill not cause the operator to take any. unnecessary . action, even if a level below the top of the core is indicated since the core exit thermocouples are used as i

a corroborative indication of the approach to ICC.

• Westinghouse Electric Corporation, " Westinghouse Reactcr Vessel Level Instrumentation System for Monitoring Inadequate Core Cooling," December 1980. -

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3. The normal cCndition for CCntinuous uccer' Dienum injection (UPI) occurs only with the oceration of the low head safety injection oumos, which does not occur until a cressure of under 200 psi is realized. The RVLIS may not accurately trend with vessel level during the initial start of UPI. During this short period of time, the cold water being injected will nix with the steam in the upper plenum causing condensation. This condensation will occur f aster than the system response.' The system will equilibrate after a short period of time. Upon equilibrating, the system will continue to accurately trend with the vessel level.

In the rar.ge of break sizes where RVLIS is most useful in detecting the approach to ICC, the system pressure will equilibrate at a level above the pressure where UPI will normally occur. It is important to note that the flow from the low head pumps is sufficient to recover the core and no operator action based on the RVLIS indication will be necessary.

For the vast majority of small breaks, the condition of upper plenum injection does not cause a significant impact. For the remainder, the impact is very small and within tolerable limits.

4. During the time when the distribution of voids in tha vessel is changing rapidly, there can be a large change in the two-phase mixture level with very little change in collapsed mixture level. The use of the RV'LIS, in conjunction with the core exit themocouples, is still valid f or this situation, however. The only event that has been identified which could cause a large void redistribuition is when the reactor coolant pumps are tripped when the vessel mixture is highly voided. Af ter the l

pump perf omance has degraded enough that the flow pressure drop O

cont-ioution to the vessel differential p' essure is small, the change in P.VLIS indication will be small when the our.cs are tripped. As dis:Ussed in the summary report, the accroach to ICC would be indicated when the wide range indication read 33 percent. If the pumos were tripped at this time, the core would still be covered. The operator would know that the core may uncover if the pumps were tripped with a wide range indication lower than 33 percent. Prior to pump trip, the core will remain adequately cooled due to forced circulation of the mixture.

When the pumps trip the two phase level may equillibrate at a level below the top of tha core. The narrow range indication will provide an indication of core coolability at this time.

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• ftem 13

. No recontendations are made as to the uncertainties of the pressure or temperature transducers to be used, but the choice appears to be lef t to the owner or AE. What is the upper limit of uncertainties that should be allowed? Describe the effect of these uncertainties on the measure-ment of level. What would be the effect on the level measurement should these uncertainties be exceeded?

Response

13. The reactor coolant pressure and temperature signals originate from the existing wide range pressure and hot leg RTD's already installed in the plant, and the uncertainties for these instruments are understood. As indicated in the response to question 6, the pressure uncertainty is, + 60. psi and the temperatur<: uncertainty is

+ 60F, resulting in a maximum level uncertainty contribution of

+ 2.3% when the vessel is full. This uncertainty is smaller when the level is at the elevation of the reactor core. This contribution to the total uncertainty would increase roughly in proportion to an increase in the pressure or temperature measurement uncertainty.

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, Only single RTD sensors on each vertical run are indicated to determine the temperatures of the impulse lines. Where are they to be located?

What are the expected temperature gradients along each line under normal operating conditions and unde'r a design basis accident? What is the worst case error that could result from only determining the, temperature at a single point on each line?

Response

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14. RTD sensors are installed on every independently run vertical sec-tion of impulse line, to provide a measurement for density co'mpensa-tion of the reference leg. .lf the vertical section of impulse line runs through two compartments separated by a solid floor, an RTD sensor is installed in each compartment.

The RTD is installed at the midpoint of each vertical section, based on the assumption that the temperature in the compartment is uniform or that the temperature distribution is linear in the vicinity of , .

the impulse line. As stated in the response to question 6, an allowance for the true average impulse line temperature to differ

- from the RTD measurement by 50F is included in the measurement uncertainty analysis. This allowance permits a significant devia- -

tion from a linear gradient, e.g., 20% of the impulse line could be up to 250F different from a linear gradient without exceeding the allowance. During nomal operation, forced circulation from cooling fans is expected to maintain compartment temperatures reasonably uniform. During the LOCA, turbulence within a compartment due to release of steam would also produce a reasonably uniform tempera-ture. Note that the impulse lines subject to direct jet impingement are protected by meta +t instrument tubing channels.

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Item 15 What is the source of the tables or relationships used to calculate density corrections for the level system?

Resoonse The relationships used in the microprocessor based RVLIS system to cal-culate density corrections are used on the ASME Steam Tables dated 1967 These relationships are implemented in the system using two fourth order polynomials, end to end, fit to approximate the tables above.

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,em 15 The microprocessor system is stated to display the status of the sensor input. Describe how is this indicated and what this actually means with respect to the status of the sensor itself and the reliability of the indication.

Response

The remote display unit of RVLIS indicates the status of the input sensors. If any sensors are out of range, regardless of the reason, a symbol shows the affected level reading on the summary display page.

The particular sensor that is out of range is identified at the bottom of the summary display page. Due to the redundant sensors and trains it is possible for the operator to disable some of the sensors without affecting the system reliability. The display indicates which level readings are affected. The disabled sensors are also displayed at the bottom of the summary page. A separate sensor status page can be dis-played showing all sensors which are disabled or out of range and their affected level readings. -

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Item 17

- Describe the provisions for preventing the draining of either the upper head or hot leg impulse lines during an accident. What would be the resultant errors in the level indications should such draining occur?

Response

17. The layout of the impulse lines from the upper head and hot leg are arranged to prevent or minimize the impact of drainage during an accident. In general, however, the water in the impulse lines will be cooler than the water in the reactor or hot leg, and there will be sufficient subcooling overpressure in the lines so that very little, if any, of the water would flash to steam during a depres-surization or containment heatup. Heat conduction along the small diameter piping and tubing would be insufficient to result in flash-ing in a significant length of piping.

The connection to the upper head from a spare control rod drive mechanism pcrt or vessel vent line drops or slopes down from the highest point of the vessel connection to the sensor bellows mounted on the refueling canal wall, so water would be retained in this piping. Draining of the vertical section imediately above the reactor vessel has no effect on the level measurement, since this section is included in the operating range of the instrument.

Draining of the horizontal portion of vessel vent piping above the vessel also has no effect on the measurement since no elevation head is involved.

The connection from the hot leg to the sensor bellows is a horizon-tal run of tubing, so draining of this tubing has no effect on the measurement since no elevation head is invol'v ed.

The majcrity of the impulse line length is in capillary tubing sealed at both ends with a bellows (sensor bellows at the reactor end, hydraulic isolator at the containment penetration end), so

1 water would be retained in this system at all times. The water will be pressurized by reactor pressure, and since the reactor tempera-ture will be higher than containment temperature during an accident, the water in the sealed capillary lines cannot flash.

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Item 18 Discuss the effect on the level measurement of the release of dissolved, noncondensible gases in the impulse lines in the event of a depressuri-zation.

Response

18. The majority of the impulse lines are sealed capillary tubes vacuum filled with demineralized, deaerated water. The lines contain no noncondensible gases and are not in a radiation environment suffi-cient for the disassociation of water.

The short runs of impulse .line connected directly to the primary system will behave as described in the response to question 17.

There would be no error due to gases in the hot leg line since the line is horizontal . Since there is no mechanism for concentration of gases at the top of the reactor vessel during normal operation, the connection to the top of the vessel would contain, at most, the normal quantity of dissolved gases in the coolant, and the subcool ~

ing pressure during an accident would maintain this quantity of gas in solution.

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Item 19 In some tests at Semi-scale, voiding was observed in the core while the upper head was still filled with water. Discuss the possibility of cooling the core-exit thermocouples by water draining down out of the upper head during or af ter core voiding with a solid upper head.

Response

19. One of the indicators of an approach to an Inadequate Core Cooling (ICC) situation is the response of the core exit thermocouples (T/Cs) to the presence of super-heated steam. The core exit thermo-couples will not provide an indication of the amount of core void-ing. Response of the core exit T/Cs provides a direct indication of the existence of ICC, the effectiveness of ICC recovery actions, and restoration of adequate core c'ooling. The core is adequately cooled whenever the vessel mixture level is above the top of the core and the core may have a significant void fraction and still be ade-quately cooled.

Realistically, an indication of an ICC condition would not occur until the primary coolant system has drained sufficiently for the

. reactor vessel mixture level to fall below the top of the core.

Westinghouse has perfomed analyses which indicate that the upper head will drain below the top of the guide tubes before ICC condi-tions exist. The guide tubes are the only flow path from the upper head to the upper plenum. In WCAP-9754, "Inadequa+e Core Cooling Studies of Scenarios With Feedwater Available, Using the NOTRUMP Computer Code", it was found that inadequate core cooling situations woul not result for LOCAs of an equivalent size or equal to approxi-mately 6 inches or less without two or more f ailures in the ECCS.

In both specific scenarios examined in WCAP-9759, a 1-inch and 4-inch small LOCA, the upper head and upper plenum had completely drained before the onset of an ICC condition.

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from the bottom of the support columns (see attached Figure). In this location, they measure the temperature of the fluid leaving the core region through the flow passages in the upper core plate. Fled from the upper head must enter the upper plenum via the guide tube before being able to enter the upper core plate flow passages. In addition, the LOCA blowdown depressurization behavior must be such that there is a flow reversal for the core exit T/Cs to detect the upper head fluid temperature. The upper head fluid is expected to mix with the upper plenum fluid as it drains from the upper head.

5 The potential for core exit T/C cooling from colder upper head fluid, while the core has an appreciable void fraction is not viewed

. as a potential problem for the detection of an inadequate core ccoi-ing situation. Although some Semi-scale tests indicated core void-ing while the upper head was 1-iquid solid that does not imply that the core exit T/Cs would give an ambigious indication of ICC.

Calculations for a Westinghouse PWR and consideration of the core exit T/C design would not result in ambigious ICC indications.

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. . Item 20 Describe the behavior cf tne level measurement system when the upper head is full, but the lower vessel is not.

Resoonse

20. During the course of a LOCA transient, the upper plenum will experience voiding before the upper head. The voids in the upper plenum will be indicated by a lower RVLIS reading. The RVLIS will not indicate where the voiding is occurring, but at this point in the transient, it is not necessary to know where the region of voiding is. In the early part of the transient when the mixture level is above the top of the guide tube in the upper head, it is sufficient for the operator to know that the vessel inventory is

- decreasing, irrespective of the region where voiding is occurring.

As discussed in the response to Item 21, the fluid in the upper head does not affect the RVLIS indication after the upper head has drained to below the top of the guide tubes. As discussed in the response,to Item 19, the uipper head will drain befor.e the onset of ICC and there will not be an ambiguous indication during the period of time when RVLIS will be used.

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TM m 21 One discussion of the microprocesor system states that water in the upper head is not reflected in the plot. Does this mean that there is no water in the upper head or that the system is indifferent to water in the upper head under these conditions?

Response

21. The discussion in the system description is contained in' the section describing the analysis of the system performance. The statement in question is referring to the WFLASH code calculation of mixture level, rather than how the RVLIS will respond to water in the upper head. The computer code includes calculation of water mass and pressure in the upper head, but this water mass is not included in the calculation of mixture level; hence, the mixture level is indi-cated only below the elevation of the upper support plate.

The RVLIS measurement from top to bottom of the vessel will measure

• the level in the following regions: top of vessel tc, top of guide tube; inside, guide tube from top to upper support plate; upper plenum; reactor core; lower plenum. During a LOCA, the RVLIS will measure the water level in the upper head only until the level drops to the top of the guide tubes; RVLIS would then meascre level reduc-tion in the guide tubes and upper plenum. The water remaining in the upper head below the top of the guide tubes would not be mea-sured by RVLIS. This water would eventually drain through small holes into the guide tubes and downcomer, and this draining would be accomplished within a few minutes, depending on the a:cident. In i any case, the water temporarily retained in the upper head would have no effect on the RVLIS indication. ,

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Item 22 Describe the details of the pump flow /Dp calculation. Discuss the pos-sible errors.

Response

22. Calculations are performed to obtain an estimate of the differential pressure that the wide range instrument will measure with all pumps operating, from ambient temperature to operating temperature. The calculations employ the same methods used to estimate reactor cool-ant flow for plant design and safety analysis. These calculations are used primarily to define the instrument span and to provide an estimate for the function that compensates the differential pressure signal over the full temperature range, i.e., that results in the wide range display indicating 100% over the full temperature range with all pumps operating, pumping subcooled coolant. During the initial plant startup following installation of the instrumentation, wide range differential pressure data would be obtained and used to confirm or revise the compensation function so that a 100% output is obtained at all temperatures. Since the calculated compensation function is verified by plant operating data, any uncertainties in the flow and differential pressure estimates are eliminated.

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Item 23

. Have tests been run with voids in the vessel? Describe the results of these tests.

Response

23. At present a Westinghouse RVLIS is installed at the Semiscale Test Facility in Idaho. Small break loss-of-coolant experiments are being conducted at this facility by EG&G for the NRC. The results of these tests are used to compare the RVLIS measurements with Semi-scale differential pressure measurements, gamma densitometer data and core cladding surf ace thermocouple indications. To date..after correcting for differences between PWR reactor vessel internals and Semiscale modeling, good correlation between Semiscale level indica-tions and RVLIS measurements has been observed. In cooperation with the NRC, EG&G and ORNL, Westinghouse is preparing a report summariz-ing the RVLIS performance during selected Semiscale tests.

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Item 24 Estimate the expected accuracy of the system af ter an ICC event.

Response

24. The accuracy of the system as described in the response to question 6 would be the'same for any LOCA-type incidibi',~ including an'ICC event, causing a temperature increase within the reactor contain-ment. Uncertainties due to reference leg temperature measurements and sensor and hydraulic isolator displacements are included in the accuracy analysis.

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Item 25 l

, Describe how the conversion of RTD resistance to temperature made in the analog level system.

Response

The RTD is connected such that an analog voltage which is proportional to RTD temperature, is input to the microprocessor system. This analog voltage is converted to ter..perature by using a curve stored in memory which relates voltage to RTD temperature.

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WES"ING:iOUSE C ASS 3 1

SUMMARY

REPORT WESTINGHOUSE REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM FOR MONITORING INADEQUATE CORE COOLING (MICROPROCESSOR SYSTEM)

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TABLE OF CONTENTS

1.0 INTRODUCTION

1.1 NRC Requirements 1.2 Definition of ICC 1.3 Condition or Events Which Describe the Approach to ICC 2.0 FUNCTIONAL REQUIREMENTS 2.1 Parameter Critical to ICC 2.2 Instrumentation Accuracies, Ranges, and Time Response 2.3 Qualification Requirements 2.4 Codes and Standards 3.0 ICC INSTRUMENTATION IDENTIFICATION 4.0 RVLIS - SYSTEM DESCRIPTION 4.1 General Description 4.2 Detailed System Description 4.2.1 Hardware Description

4. 2.1.1 Differential Pressure Measurements 4.2.1.2 System Layout 4.2.2 Microprocessor System 4.2.2.1 Inputs

{ 4.2.2.2 Density Compensation System l 4.2.2.3 Plant Operator Interface and Displays l

4.2.2.3.1 Display Functions for Remote Control Board 4.2.3 Resistance Temperature Detectors 4.2.4 RVLIS Valves 4.2.5 Transmitters, Hydraulic Isolators, and Sensors 4.3 Test Programs 4.3.1 Forest Hills 4.3.2 Semiscale Tests 4.3.3 Plant Startup Calibration 7683A MP

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I TABLE OF CONTENTS (Continued) 1 4.4 Operating Performance 4.5 RVLIS Analysis 4.5.1 Transients Investigated 4.5.2 Observations of the Study 4.5.3 Conclusions 5.0 GUIDELINES FOR THE USE OF ICC INSTRUMENTATION 5.1 Reference Westinthouse Owners Group Procedure 5.2 Sample Transient -

6.0 REFERENCES

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1 LIST OF TABLES ,

Table 3.1 Infomation Required on the Core Subcooling Monitor Table 4.1 Compliance with Regulatory Guide 1.97 Oraf t 2, Rev. 2 6/4/80 l

Taole 4.2 Transients Investigated i

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LIST OF FIGURES Figure 4-1 Reactor Vessel Level Instrument System Figure 4-2 Process Connection Schematic, Train A Figure 4-3 Typical Plant Arrangement for RVLIS Figure 4-4 Reactor Vessel Level Instrument System Block Diagram (One Set of Two Redundant Resets Shown)

Figure 4-5 Renote Display Module (Control Board)

Figure 4-Sa Vessel Level Summary Display Figure 4-5b Vessel Level Trend Display Figure 4-6 Typical Plant Arrangement for RVLIS Figure 4-7 Block Diagram of Compensation Function Figure 4-7a Simplified Schematic of Density Compensation System Figure 4-8 Surface Type Clamp-On Resistance Temperature Detector Figure 4-9 HELB Simulation Profile Figure 4-10 ITT Barton Hydraulic Isolator Figure 4-11 ITT Barton "High Volume" Sensor Bellows Check Valve 1

l Figure 4-12 Case A 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip with Reactor Trip, RVLIS Reading and Vessel Mixture Level 7683A MP i

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l Figure 4-13 Case A 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip with Reactor Trip, Void Fraction Figure 4-14 Case B 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip at 750 Seconds, Wide Range Reading Figure 4-15 Case B 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip at 750 Seconds, RVLIS Reading and Mixture Level Figurs 4-16 Case B 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip at 750 Seconds, Void Fraction.

Figure 4-17 Case B 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip at 750 Seconds, Cold Leg Mass Flowrate (LB/Sec)

Figure 4-18 Case C 2.5 Inch Pressurizer Break, No. SI, RVLIS Reading and Mixture Levels Figure 4-19 Case C 2.5 Inch Pressurizer Break, No. SI, Void Fraction Figure 4-20 Case 01 Inch Cold Leg Break, ICC Case, RVLIS Reading and Mixture Level.

Figure 4-21 Case 01 Inch Cold Leg Break, ICC Case, Mixture Level, RVLIS Reading and Measured Inventory.

Figure 4-22 Case 01 Inch Cold Leg Break, ICC Case, RVLIS Reading and Mixture Level.

Figure 4-23 Case 01 Inch Cold Leg Break, ICC Case, Void Fraction -

7683A MP i

4 1.0 -INTRODUCTION 1.1 1RC REOUIREMENTS The NRC has established requirements (items I.C.1 and II.F.2 of NUREG-0737,

• Clarification of TMI Action Plan Requirements") to provide the reactor operator with instrumentation, procedures, and training neces-sary to readily recognize and implement actions to correct or avoid conditions cf inadequate core cooling (ICC).

Under certain plant accident conditions, the potential exists for the formation of voids in the reactor coolant system (RCS). U. der these conditions, it would be advantageous for the reactor operator to monitor the water level in the reactor vessel or the approximate void content during forced circulation conditions in order to assist him in subse-quent, actions. Therefore, a reactor vessel level instrumentation system (RVLIS) has been incorporated to provide readings of vessel level which can be used by the operator. Vessel level as measured by the RVLIS is the collapsed liquid level in the vessel.

The RVLIS provides a relatively simple and straight-forward means to monitor the vessel level. This instrumentation system neither replaces any existing system nor couples with any safety system; however, it does act to provide additional information to the operator during accident conditions. The RVLIS utilizes differ'ential pressure (d/p) measuring devices to indicate relative vessel le'el v or relative void content of the circulating primary coolant system fluid.

1.2 DEFINITION OF ICC ICC as defined in References 1 and 2, is a high temperature condition in the core such that operator action is required to cool the core before damage occurs.

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l 1.3 CONDITIONS OR EVENTS WHICH DESCRIBE THE APPROACH TO ICC The most obvious failure that woulo lead to ICC during a small-break LOCA, althougn highly unrealistic since multiple failures are required, is the loss of all high pressure safety injection. The approach to ICC conditions and the analyses for this event sequence are provided in References 1 and 2.

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2:0 FUNCTIONAL REQUIREMENTS 2.1 PARAMETERS CRITICAL T0 Irc The analysis provided in References 1 and 2 delineates those parameters critical for the detection of and the necessary mitigation actions for the recovery from an ICC condition.

To briefly sununarize those parameters, ICC is detected by either high core exit thermoccuple temperatures or by a low reactor vessel level indication (core uncovery) in conjunction with core exit thermocouple indications. Mitigation actions consist of depressurizing the reactor coolant system (RCS) to permit injection of accumulator water and/or to establish low head safety injection flow. The RCS is itself depressurized by depressurizing the steam generator secondary side.

Critical parameters at this point are steam generator pressures and wide range RCS loop temperatures. Once low head safety injection flow is established, transfer out of the ICC procedure can be made when core exit thermocouple temperatures are reduced and the reactor vessel level gauge indicates a level above the ta of the core.

With the exception of reactor veswl legal, all parameters are monitored by currently existing instrtsnentation.

2.2 INSTRUMENTATION ACCURACIES; RAN6ES; AND TIME RESPONSE Accuracy An accuracy of 6 percert is required on all three types of reactor

. vessel level instruments. This should be a statistical combination of all uncertainties including those due to environmental effects (if any) on instrumentation. For the upper range instrument, this corresponds to an allowable deviation of about + 1 foot elevation head. This will give the operator a good estimate of the steam or gas volume in the upper head during a situation in which the head vent would be employed. For the narrow range instrument this corresponds to an allowable deviation 2-1 7581A ,,

of about 2.5 feet elevation head. This is required to: 1) provide adecuate margin against inadvertant use of the ICC operating guideline (E2 01-1, see Section 5.1), 2) assure that the vessel level reading can be reasonably used to aid in the detection of the onset of ICC condi-tions, 3) derive useful information reguarding vessel level behavior during the vessel refill period of a LOCA transient.

Rance The wide range instrument will cover the full range of expected differ-ential pressures with all reactor coolant pumps running. The maximum span of the wide range instrument will change with the number of pumps operating. The operator must be aware of the maximum span for a given number of operating pumps. Bbth the narrow range and the upper range instrument indications should be set to indicate that the vessel is full with the pumps tripped.

Time Response The d/p instrument response time shall not exceed 10 seconds. This time delay is defined as the time required for the display instrument to reach the midpoint of a 50 percent step input d/p change.

2.3 OUALIFICATION REQUIREMENTS Environmental qualification of the RVLIS shall verify that the system equipment will meet, on a continuing basis, the performance requirements determined to be necessary for achieving the system requirements as presented above. Verification must include confirmation that those portions of RVLIS equipment which are within the containment will oper-ate during and subsequent to the conditions and events for which the system is required to be operational. Verification will include deter-mination that the system is sufficiently accurate during this time to meet its design basis. The system post-accident environment qualified life requirenent for electrical equipment inside containment is 120 days 2-2

1 following certain postulated events. The electrical equipment that is installed outsice of containment need not meet a qualified life for an extended period of time providing replacement or calibration cnecks can be made in snort enougn time commensurate witn tne reliability goals of the recundant system. For the resistance temperature detectors (RTDs) environmental requirements for service within the containment, refer to Section 4.2.3. Electrical equipsent inside containment snall be instal-kd such that it is removed from sreas where high energy pipe breaks or pipe whip could cause f ailure. The d/p transmitters and electronic processing equipment shall be located in a low amoient radiation area.

The RVLIS sensing transmitters and associated electronic processing equipment shall be locatec; in an area whose temperature range is between 40 and 120*F with 0 to 95 percent ambient relative humidity. Normal operating environment for transmitter locations snall be netween 60 and 80*F and 0 to 50 percent relative humidity. The instrumestation shall be qualified to assure diat it continues to operate and reaa within tne required accuracy following but not necessarily during a safe shutdown earthquake. Qualification of the electronic equipment and reactor ves-sel level sensing transmitters . applies to and includes the channel iso-lation device or where interface with a computer is involved, the input buffer. The location of the electronic isolation device or input buffer should be such that it is accessible for maintenance during accident conditions.

2.4 CODES AND STANDARDS The RVLIS is in conformance with the following Codes and Standaros:

Regulations GDC 1 Quality Standards and Records GDC 2 Design Bases for Protection Against Natural Phenomena GPC 4 Environmental and Missile Design Bases 2-3 meme e l

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GDC 13 Instrumentation and Control GDC 16 Containment Design GDC 18 Inspection and Testing of Electric Power Systems GDC 19 Control Room GDC 24 Separation of Protection and Control Systems GDC 30 Quality of Reactor Coolant Pressure Boundary GDC 31 Fracture Prevention of Reactor Coolant Pressure Boundary GDC 32 Inspection of Reactor Coolant Pressure Boundary GDC 50 Containment Design Basis GDC 55 Reactor Coolant Pressure Boundary Penetrating Containment GDC 56 Primary Containment Isolation 10CFR50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" Industry Standards IEEE-308-1971, "IEEE Standard Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations" IEEE-323-1971, "!EEE Trial-Use Standard: General Guide for Qualifying Class 1 Electric Equipment for Nuclear Power Generating Stations"*

IEEE-338-1971, "IEEE Staridard Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems" IEEE-344-1971, " Guide for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations"**

IEEE-384-1977, "IEEE Standard Criteria for Independence of Class 1E Ec.1pment and Circuits" ASME BPVC,Section III, Class 2 Nuclear Power Plant Components

• For certain specific plants, IEEE-323-1974 is applicable.

l " For certain specific plants IEEE-344-1975 is applicable.

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ANSI 331.1.0, 1967 including addenda through and including 6/30/71,

" Code for Pressure Piping", including nuclear code cases where applicable Regulatory Guider R.G. 1.11 Instrument Lines Penetrating Primary Reactor Containment i

R.G. 1.22 Periodic Testing of Protection System Actuation Functions R.G. 1.75 Physical Independence of Electric Systems 2-5 75SIA

I 3.0 ICC INSTRUMENTATION IDENTIFICATION Adequate instrumentation is necessary to diagnose the approach to ICC ano to determine the effectiveness of the mitigation actions taken.

During the preparation of the ICC operating instructions, consiceration was given to the adequacy of current instrumentation and tne benefits derivaDie from the addition of new instrumentation. The following is a list of existing instrumentation considered (refer to the FSAR for details) and conclusions derived:

1. Current Instrumentation

a. WIDE RANGE REACTOR COOLANT PRESSURE '- present instrumentation is available for determining general RCS pressure trenos during the ICC event. The expected accuracy following ICC events is such that this instrument cannot be used for precise determina-tions of the pressure required to assure onset of low head safety injection flow to the RCS.

b. PRESSURIZER PRESSURE AND LEVEL - conditions in the pressurizer will generally lie outside the ranges of these instruments during an ICC event in a Westingnouse PWR. Pressurizer pres-sure and levol are not used for determining mitigation actions to be taken during ICC.

c. AUXILIARY FEE 0 WATER FLOW - present instrumentation is available for assuring the sufficiency of makeup water flow to the steam generators during an ICC event.

d. WIDE RANGE RESISTANCE TEMPERATURE DETECTORS - present instru-mentation is available in determining trends of recovery actions but may not ne available in determining the onset of ICC conditions for all breat sizes.

e. CORE EXIT THERMOCOUPLES - present instrumentation is available in determining both the existence of ICC and the trends of recovery actions.

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f. CORE SUBC00 LING - does not provide useable infonnation during an ICC condition. Will indicate superheat conditions in core coolant. Will help indicate the approach to ICC by showing saturation conditions. Since the core subcooling monitors may not be described in the FSAR, refer to Table 3.1 for informa-tien.

g. STEAMLINE PRESSURE - present instrumentation is available for determining heat sink availability and heat removal capability during ICC mitigation actions,

h. STEAM GENERATOR LEVEL - present instrumentation is available for determining the avdilability of a heat sink for the RCS during an ICC condition.

2. New Instrumentation

a. REACTOR VESSEL LEVEL - provides an indication of the approach to ICC and confirms the achievement of adequate core cooling when level in the reactor vessel is restored.

To sunenarize the above considerations, current plant instrumentation is adequate to determine heat sink availability, to detect the enset of ICC, and to detect the effectiveness of mitigation actions following the onset of an ICC event. The RVLIS is provided to permit a more continuous indication of the approach to ICC.

• 3-2 7581A i

f. CORE SUSC00 LING - does not provide useable information during an ICC condition. Will indicate superheat conditions in core coolant. Will help indicate the approach to ICC by showing saturation conditions. Since the core subccoling monitors may not be described in the FSAR, refer to Table 3.1 for informa-tion.

g. STEAMLINE* PRESSURE - present instrumentation is available for determining heat sink availabilty and heat removal capability during ICC mitigation actions.

h. STEAM GENERATOR LEVEL - present instrumentation is available for determining the availability of a heat sink for the RCS during an ICC condition.

2. New Instrumentation

a. REACTOR VESSEL LEVEL - provides an indication of the approach to ICC- and confirms the achievement of adequate core cooling when level in the reactor vessel is restored.

To swanarize the above considerations, current plant instrumentation is adequate to determine heat sink availability, to detect the onset of ICC, and to detect the effectiveness of mitigation actions following the onset of an ICC event. The RVLIS is provided to permit a more l

l continuous indication of the approach to ICC.

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TABLE 3.1 INFORMATION REQUIRED ON THE CORE SUBC00 LING MONITOR 01solay Information Displayed (T-Tsat, Tsat, P-Psat succooled press,etc.) T-Tsat superheat Display Type (analog, digital, CRT) Analog and digital Continuous or on Demand Analog - continuous Digital - on demand Single or Redundant Display Redundant Location of Display User supplied Al arms Caution - 250F subcooled for RTO Alam - 00F subcooled (include setpoints) 150 F subcooled for T/C for RTO and T/C Overall Uncertainty ('F, psi) Digital - 4cF for T/C; 3'F for RTO Analog - 5'F for T/C; 5'F for RTO Range of Calibrated region - 1000 psi subcooled to 2000'F superheat Display Overall - never offscale Qualifications None at present*

Calculator Type (process computer, dedicated digital Dedicated digital or analog calc.)

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(percent of time)

Single or Redundant Calculators Redundant Selection Logic (highest T., lowest press) Highest T for RTO or T/C; Lowest P Qualifications None at present Calculational Technique (steam tables, Functional fit -

I functional fit, ranges) ancient to critical point

• The display is currently undergoing seismic qualification testing by Westinghouse which will conform to IEEE-344-1971. This infomation will only be provided at the specific request of the custczner and af ter aopropriate installation enecks have been made to verify the applicability of this qualification.

3-3

l l

l l

TABLE 3.1 (Continued)

Input Temperature (RTOs or T/Cs) RTO, T/C and Tref Temperature (number of sensors and locations) RTD - 2 not and 2 cold leg per cnannel T/C - 8 per cnannel Range of Temperature Sensors RTD 700*F T/C 1650 F (calibration unit range 0 - 2300*F)

Uncertainty

• of Temperature Sensors ( F at le) User supplied Qualifications User supplied Pressure (specify instrument used) User supplied Pressure (number of sensors and locations) 2 wide range - Loop 1 narrow range -

Pressurizer Range of Pressure Sensors Wide range 3000 psi Narrow range - 1700 - 2500 psi Uncertainty ** of Pressure Sensors (psi at le) User supplied Qualifications User supplied Backup Capability l

Availability of Temp and Press l

l Availability of Steam Tables etc.

Procedures

• • Uncertainties must address conditions of forced flow and natural circulation i

3-4

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4.0 REACTOR VESSEL LEVEL INSTRUMENTATON SYSTEM - SYSTEM DESCRIPTION 4.1 GENERAL DESCRIPTION ine *eactor vessel level instrumentation system (RVLIS) uses differen-tial pressure (d/p) messuring devices to measure vessel level or rela-tive void content of the circulating primary coolant system fluid. The system is redundant and includes automatic compensation for potential temperature variations of the impulse lines. Essential information is displayed in the main control room in a form directly useable by the oper ator.

The functions performed by the RyLIS are.

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.

4.2 DETAILED SYSTEM DESCRIPTION 4.2.1 HARDWARE DESCRIPTION 4.2.1.1 Differential Pressure Measurements l

The RVLIS (Figure 4-1) utilizes two sets of three d/p cells. These l 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:

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1. Reactor Vessel - Upper Range (AP3 )

The d/p cell AP, snown in Figure 4-1 provides a measurement of reactor vessel level above the hot leg pipe when the reactor cool-ant pump (RCP) in the loop with the hot leg connection is not operating.

2. Reactor Vessel - Narrow Range (aPb )

This measurement provides an indication of reactor vessel level from the bottom of the reactor vessel to the top of the reactor during natural circulation conditions.

3. Reactor Vessel - Wide Range (aPg )

This instrument provides an indication of reactor core and inter-nals pressure drop for any combination of operating RCPs. Com-parison of the measured pressure drop with the normal, single-phase pressure drop will provide an approximate indication of the relative void content or density of the circulating fluid. This instrument will monitor coolant conditions on a continuing basis during forced flow conditions.

l To provide the required accuracy for level measurement, 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 during the change in the environment inside the containment l

structure following an accident.

The d/p cells are located outside of the containment to eliminate the large reduction (approximately 15 percent) of measurement accuracy asso-ciated with the change in the containment environment (temperature, pressure, radiation) during an accident. The cells are also located outside of containment so that system operation including calibration, cell replacement, reference leg checks, and filling is made easier.

4-2

4.2.1.2 System Layout a schematic of the system layout for the RVLIS is shown in Figure 4-2.

There are four RCS penetrations for the cell reference Ifnes; one reac-tor head connection at a spare penetration near the center of the head or the reactor vessel head vent pipe, one connection to an incore instrument conduit at the seal table, and connections into the side of 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 describ.ed in Section 4.2.4. These lines connect to six sealed capillary impulse 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 building. The 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.

l Figure 4-3 is an elevation plan of a typical plant showing the routing of the impulse lines. The impulse lines from the vessel head connection must be routad upward out of the refueling canal to the operating deck, then radially toward the seal table and then to the containment penetra-tion. The connection 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 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 con-nections.

- 4-3 7581A

The impulse lines located inside the containment building will be 4

exposed to the containment temperature increase during a LOCA or HELB.

Since the vertical runs of impulse lines form the reference leg for the o/p measurement, the change is density due to the accident temperature change must be taken into account in the vessel level determination. ,

Therefore, a strap-on RTO is located on each vertical run of s.narately routed impulse lines to determine the impulse line temperature and cor-rect 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 nurter of representative plant arrangements, a maximum of 7 independent vertical runs must be measured to adequately compensate for density ch anges.

4.2.2 MICR0 PROCESSOR RVLIS The microprocessor RVLIS includes equivalent reactor vessel level indications on redundant flat panels with alphanumeric displays provided for control room installation in addition to having this information available for display at the microprocessor chassis. RVLIS is configured as two protection sets, in certain installations in separated sections of a single instrument rack and in other installations in two separated instrument racks. The envelope of an instrument rack occupies a,c a space at the base of The block diagram of the RVLIS using microprocessor equipment is shown in Figure 4-4. This diagram shows that in addition to the reactor vessel level (d/p) transmitter input, there are also temperature compensating signals, reactor pump running status inputs, and RCS parameter inputs to each 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 to Figure 4-5.

Conformance with Regulatory Guide 1.97 for the processor display system is given in Table 4.1.

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4*

4.2.2.1 RVLIS Inputs The microprocessor system inputs are as follows. If existing unquali-fled inputs are used, isolation as required will be provided by the owner.

Differential Pressure Transmitters The three d/p transmitters per set are used to measure the d/ps between the three pressure tap points on the primary system, as discussed below:

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The direction of this transmitter's output is full scale (20 ma) with the vessel full and zero scale (4 ma) with the vessel emptied to the hot leg tap. These endpoints are nominal and are for low coolant temperatures. If no pumps are operating, aP3 gives an indication of level in the region above the hot leg.

If the pump is running in the loop with the hot leg connection, this indication will be invalid and most likely off-scale. The reading would be flagged as " invalid" under these conditions. The effect on the indication from the pump not running in this loop, but running in other loops, is less than 10 percent of th? cange.

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APb gives an 19:':ation of reac::r vesse: 's/ei unen no p;mps are running. ~.' ane or more pumps a-e ca-aing, AP 3 uil; ce off-scale ans : s reading invali:.

The sense of t,s AP b output is s cn :qa a 20 ma signal is a nominally full isssel and a 4 ma signal is for a nominally empty vessel.

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. 3.

J The sense of tes 1,- oatout is tqat 20 3

ma represents all pumps running ana a ma is emoty vessel. With all pumps running and no void fraction, ,e APc snoul2 read 100

, , ,ercent p at zero power. The reacing at f;il power is sligntly higher.

Reference Leg Temoerstare RTD The reference leg temperature RTDs are used to messare tne temoerature of the coolant in the Capillary tube reference legs. This is used to compute thi density of the reference leg fluto.

s A typical arrangement of the reference leg temce-st;re RTDs is snown in Figure.4-6.

The conversion of RT3 resistance to temperat;-e small cover :ne tempera-ture range of 32 to 250*F.

c The RTDs are 100 ohm olatinum four wire RT3s as snown 'n Figure 1-8.

Hot Leg Temper &ture Either existing or -sw wide range hot leg ti :n-*:.es sensoes a-i used to measure the coolar,: temperature. qis e :t-a;.rt is .ss: :: Calcu-late coolant density.

4-6 7683A MP

A'ae Range Reactor Coolant Pressure i E'-her existing wide range pressure sensors or new pressure sensors will ce used to measure reactor coolant pressure. The pressure is used to calculate reactor coolant density.

Tne block diagram of the compensation functions is shown in Figure 4-7.

Dioital Inputs Tne reactor coolant pump status signals indicate whether or not pumps are running. Recognizing that hydraulic isolators are provided on each impulse line for containment isolation purposes, each hydraulic isolator has limit switches to indicate they have reached the limit of travel.

4.2.2.2 Density Compensation System

• To provide the required accuracy for vessel level measurement, tem-

erature measurements of the impulse lines are provided. These Teasurements, together with the existing reactor coolant temperature reasurements and wide range RCS pressure, are employed to compensate the

/p transducer outputs for differences in system density and reference lag density, particularly during the change in the environment inside the containment structure following an occident. A simplified schematic of the density compensation system is shown in Figure 4-7a. The d/p cells are located outsida the containment.

e reference leg fluid density calculation must cover a range of 32 to 4500F. The fluid is assumed to be compressed liquid water at 1200

sta.

I:ch of the three d/p measurements will have density corrections from certain temperature measurements. Some of these will have a positive

rrection and some negative depending on the orientation of the impulse

'ine where the temperature is being measured.

1-7

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Vessel Liquid Density Calculation a,c

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Vessel Vapor Phase Density Calculation a,c Vessel Level Calculation

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Pump Flow d/p Calculation

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4-8 7683A MP

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l The lower of the two calculated d/p corrections is divided into the measured d/p. The result is the percent of expected d/p and should read 100 percent with all pumps operating and no circulating voids.

Scaling of Displayed Values Each of the three d/p measurements after the preceding calculations shall be scaled to read in percent. With the vessel full of water and no pumps running, the outputs of aPa and aPb should read 100 percent.

4.2.2.3 Plant 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 board displays provide the

• following information:

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d.0 WD

All signals are input to a microprocessor-based data analysis system.

The control room display format utilizes an alphanumeric display located remotely from the computational system.

Redundant displays are provided for the two sets. Level information based on all three d/p measurements is presented. Correction for refer-ence leg densities is automatic. Any error conditions such as out-of-range sensors or hydraulic isolators are automatically displayed on the affected measurements.

There are two display sheets for reactor vessel level: the first is a summary sheet, and the second is a trending of the three vessel level indications.

- . a,C 4.2.2.3.1 Display Functions for Remote Control Board The prime display unit for the vessel level monitor is the 8 line, 32 character per. line alphanumeric display which is located in the control board remote from the main pro, cessing unit.

Vessel Level Monitor Sumary Display Figures 4--5, 4-Sa and 4-5b give example displays. General arrangement is shown on Figure 4-5. The vessel level sumary display is shown on Figure 4-Sa. The following is a description of the display.

1. The first line gives the title of the display as shown. The use of the underbar feature delineates this line from the rest of the display.

2. The second line gives column headings as shown. Again, the use of the underbar clarifies' the display.

7683A MP

3. The third line gives the measured and normally expected values from the aPa measurament. The first field gives the title, the second gives the measured level, the third gives the normal value for the current status, and the last field gives the validity status and is blank under normal conditions.

4. The fourth line gives the aP measurement b results using the same format as in line 3.

5. The fifth line gives the AP measurement C results using the same format as in line 3. The use of underbar in line 5 delineates this line from the next.

6. The sixth line gives the status of the pumps as seen by the unit.

The running pumps are identified.

7-8. The seventh line and eighth line are normally left blank and are reserved for hydraulic isolator limit switch indicators, out of range sensors and operator disabled sensors.

Trend Display The trend display for the vessel level monitor shall use the format shown in Figure 4-5b.

Disolays on Main Processing Unit The one-line forty character alphanumeric display on the front panel of the main processing unit is used to display individual sensor inputs.

The sensor is selected with a two digit thumbwheel switch.

The following information is to be given for each sensor:

1. Sensor identification

2. Input signal level

3. Input signal converted to engineering units

4. Status of sensor input MP 4-10a

l 1

Disablec !nouts Any inputs can be disabled by the operator. This action is under the control of a keyswiten on the front panel of the main computational unit and causes the processor to disregard the analog input for that variable. I 4.2.3 RESISTANCE TEMPERATURE DETECTORS (RTD)

The resistance temperature detectors (RTD) associated with the RVLIS are utilized to obtain a temperature signal for fluid filled instrument lines inside containment during normal and post-accident operation. The temperature measurement for all yertical instrument lines is used to correct the ves:el level indication for density changes associated with the environmental temperature change.

The RTD assembly is a totally enclosed and hermetically sealed strap-on device consisting of a thermal element, extension cable and termination cable as indicated in Figure 4-8. The sensitive portion of the device is mounted in a removable adapter assembly which is designed to conform to the surf ach of the tubing or piping being monitored. The mater,ials are all selected to be compatible with the normal and post-accident environment. Randomly selected samples from the controlled (material,

manufacturing, etc.) production lot will be qualified by type testing. '

Qualification testing will consist of thermal aging, irradiation, scis-mic testing and testing under simulation high energy line break environ-mental conditions. For the qualified life requirements, see Section 2.3. The specific qualification requirements for the RTDs are as fol-lows:

l. Ag The thermal aging test will consist of operating the detectors in a high temperature environment: either 400*F 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.

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2. Radiation The detectors shall be irradiated to a total integrated dose (TID) of 1.2 x 108 rads gama radiation using a Co60 source at a minimum rate of 2.0 x 106 rads / hour and a maximum rate of 2.5 x 106 rads / hour. Any externally exposed organic materials shall be evaluated or tested to 9 x 108 rads TID beta radiation. The energy of the beta particle shall be 6 MEV for the first 10 Mrad, 3 MEV for 340 Mrad and 1 MEV for 150 Mrad.

3. Seismic The detectors will be tested using a biaxial seismic simulation.

The detectors shall be mounted to simulate a plant installation and will be energized throughout the test.

4 Mieh Ener gy Line Break Simalation The detectors shall be tested in a saturated steam environment using the temperature / pressure curve shown in Figure 439, Caustic spray, consisting of 2500 ppm boric acid dissolved in water and adjusted to a pH 10.7 at 25'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.

I The RTD device is designed to operate over a temperature range of

-58 to 500*F (the normal temperature range is 50 to 130*F).

4.2.4 REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM VALVES Two types of valves are supplied for the RVLIS. The root valve (3/4 T78) is an ASME Class 2, stainless steel, globe valve. The basic func-tion of the valve is to isolate the instrumentation frar the RCS.The other valve (1/4 x 28 ID), is an instrumentation-type valve. It is a manually actuated ball valve used to provide isolation in the fully 4-12 l

7581A

closed position. The valve is hermetically sealed and utilizes a pack-less design to eliminate the possibility of fluid leakage past the stem to the atmosphere.

4.2.5 TRANSMITTERS, HYDRAULIC ISOLATORS, AND SENSORS Differential Pressure Transmitters The d/p transmitters are a seismically qualified design as used in numerous other plant applications. In the RVLIS application, accuracy considerations dictate a protected environment, consequently trans-mitters are rated for 40 to 1300F and 104 rad TID.

Several special requirements for these transmitters are as follows:

1. Must withstand long term overloads of up to 300 percent with minimal effect on calibration.

2. High range and bi-directional units required for pump head measure-ments.

j 3. Must displace minimal volumes of fluid in normal and overrange oper-

! ating modes.

l 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 ther-mal expansion effects in the coupling fluids.

The d/p transmitters are rated 3000 psig working pressure and all units are tested to 4500 psig. Internal valving also provides overrange l ratings to full working pressure.

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4-13 7581A l

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4 Hydraulic Isolator

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Migh-Volume Sensor

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4 14 7581A

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4.3 TEST PROGRAMS A variety of test programs are in progress or will be carried out to study the static and dynamic ' performance of the RVLIS at two test facil-ities, and to calibrate the system over a range of normal operating conditions at each reactor plant where the system is installed. These programs, which supplement the vendors' tests of hydraulic and electrical components, will provide the appropriate verification of the system response to accident conditions as well as the appropriate procedures for proper operation, maintenance and calibration of the equipment. A description of these programs is presented in the following section:

4.3.1 Forest Hills A breadboard installation consisting of one train of a RVLIS was instal-led and tested at the Westinghouse Forest Hills Test Facility. The system 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 tubing 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. Con-nections between the hydraulic isolators, valves and transmitters util-ized compression fittings in most cases. Resistance temperature detec-tors, special large volume sensor bellows and volume displacers inside 4-15

the 9ydraulic isolator assemblies which are normally part of a RVLIS installation were not included in the installation since elevated tem-perature testing was not included in the proram.

• he cycraulic isolator assemblies and transmitters were mounted at an elevation slightly below the simulated seal table elevation.

1 The objectives of the test were as follows:

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 proce<bres for shutdown and normal operation conditions.

4. Prove long term integrity of hydraulic components

5. Verify and quantify fluid transfer and makeup requirements asso-ciated with instrument valve operation.

6. Verify leak test procedures for field use Reactor Vessel Simulator The reactor vessel simulator consisted of a 40 foot long 2 inch diameter stainless steel pipe with taps at the top, -'ie and bottom to simulate

! the reactor head, hot leg and incore detector thimble conduit penetra-l tion at the bottom of the vessel. Tubing (0.375 inch diameter) was used to connect this lower tap to the sensor at the simulated seal table elevation and the hot leg sensor to the head connection was simulated by

! 1 inch tubing 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.

{

4-16 7581A

(

Installation The system was installed in the high bay test area of the Westinghouse Forest Hills Test Facility by Westinghouse personnel under the supervi-sion of Forest Hills Test Engineering. All local safety codes were considered in the construction.

Filling Operation

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4-17 7581A .

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. a,c 4.3.2 SEMISCALE TESTS In order to study the transient response of the RVLIS during a small-break LOCA and other accident conditions, the hydraulic canponents of the RVLIS have been installed at the Semiscale Test Facility in Idaho. Vessel level measurements will be obtained during the carrent semiscale test program series which runs from December 1980 to March 1982. The test scheduled to be completed by July 1981 are expected to provide the desired transient response verification; additional data will be obtained from the tests scheduled for c'ompletion by Noveneer 1981.

The Semiscale Test Facility is a model of a 4-Loop pressurized water reactor coolant system with elevation dimensions essentially equal to the dimensions of a full-size system. The reactor vessel contains an electrically heated fuel assembly consisting of 25 fuel rods with a l

heated length of 12 feet. Two reactor coolant loops are provided, each having a pump and a steam generator with a full Neight tube bundle. One loop models the loop containing the pipe break, which can be located at any point in the loop. The other loop models the three intact loops. A i blowdown tank collects and cools the fluid discharged from the pipe break during the simulated accident. Over 300 pressure, temperature, flow, level and fluid density instruments are installed in the reactor vessel and loops to record the fluid conditions throughout a test run.

! Test results are compared with predictions for verification of computer code models of the transient performance.

4-18 7581A

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l The Westinghouse level measurements obtained during a test run will be compared with data obtained from existing instrumentation installed on the semiscale reactor vessel. The semiscale facility has two metnods 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 installed at 12 elevations on the reactor vessel to determine the fluid density at each elevation. This data establishes a fluid density profile within the vessel under any operating condition, and this information will be compared with the data obtained frc.n ti.e Westinghouse level instrumentation. Other semiscale facility instruments (loop flows and fluid densities when pumps are operating, and pressure and temperatures for all cases) will provide supplemental information for interpretation of the test facility fluid conditions and the level measurement.

Specific tests included in the semiscale test program 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.

3. Repeat of small-break LOCA test with test facility modified to simu-late a plant with upper head injection (UHI).

4 Several natural convection tests covering subcooled and saturated coolant conditions and various void contents.

5. Tests to simulate a station blackout with discharge through relief valves.

6. Simulation of the St. Lucie cooldown incident.

4-19 7581A

4.3.3 PLANT STARTUP CALIBRATION During the plant startup, subsequent to installing the RVLIS, 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 accident condition. The ele-ments of the program are described below:

1. 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, measure-ments would be obtained fr'om the wide range d/p transmitters to confirm or correct the temperature compensation provided in the system electronics. The temperature compensation, based on a best estimate of the flow and pressure drop variation during startup, corrects the transmitter output so that the control board indication is maintained at 100 percent over the entire operating temperature range.

3. At hot standby, measurements would be obtained from all transmitters with different combinations 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 normal condition, i.e., with a water-solid system. 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 correct reference leg temperature compensa-tion provided in the system electronics.

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4.4 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 the three d/p transmit-ters, each transmitter covering a specific range of operating condi-tions. The three instrument ranges provide overlap so that the measurement can be obtained from more than one meter under most accident

~

conditions. Capabilities of each of the measurements are described below:

1. Reactor Vessel - Upper Range The transmitter span covers the distance from the hot leg piping connection to the top of the reactor vessel. With the reactor cool-ant 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 head of the reactor vessel. The measurement will provide ,

an accurate indication for guidance when operating the reactor ves-sel head vent. 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 connection is operating, the d/p would be greater than the transmitter span, and the trans-mitter output would be disregardec.

( 2. Reactor Vessel - Narrow Range

~

The transmitter span covers the total height of the reactor vessel.

With pumps shut down, the transmitter output is an indication of the collapsed water level, i.e., as if the steam bubbles had been separ-.

ated from the water volume. The actual water level is slightly higher than the indicated water level since there will be some quan-tity of steam bubbles in the water volume. Therefore, the RVLIS provides a conservative indication of the level effective for ade-( quate core cooling.

4-21 l

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When reactor coolant pumps are operating, the d/p would be greater than the transmitter span, and the transmitter output would be dis-regarded.

3. Reactor Vessel - Wide Range The transmitter span covers the entire range of interest, from all pumps operating with a water-solid system to a completely empty reactor vessel and therefore, covers the measurement spans of the other two instruments. Any reduction in d/p compared to the normal operating condition is an indication of voids in the vessel. The reactor coolant pumps will circulate the water and steam as an essentially homogeneous mixture, so there would be no distinct water level in the vessel. When pumps are not operating, the transmitter output is an additional indication of the level in the vessel, sup-plementing the indications from the other instruments.

The output of each transmitter is compensated for the density difference between the fluid in the reactor vessel and the fluid in the reference ,

I leg at the initial ambient temperature. The compensation is based on a wide range hot leg temperature measurement or a wide range system pres-sure measurement, whichever results in the highest value of water den-sity, and, therefore, the lowest value of indicated level. Compensation based on temperature is applied when the system is subcooled, and com-pensation based on pressure (saturated conditions) is applied if super-heat exists at the hot leg temperature measurement point.

The output of each transmitter is also compensated for the density dif-ference between the fluid in the reference leg during an accident with elevated temperature in the containment and the fluid in the reference l leg at the initial ambient temperature. The compensation is based on temperature measurements on the vertical sections of the reference leg.

l The corrected transmitter outputs are displayed on meters installed on the control board, one meter for each measurement in each train. A.

( three-pen recorder is also provided on the control board to record the 4-22 7581A

level or relative d/p and to display trends in the measurements. An indicator lignt installed under the upper range level meter would pro-vide an indication if the pump in the loop with the hot leg connection is coerating, and therefore an indication that the off-scale reading on the eter should be disregarded.

During normal plant heatup or hot standy operation with all reactor coolant pumps operating, the wide range d/p meter would indicate 100 percent on the meter, an indication that the system is water-solid. If less than all pumps are operating, the meter would indicate a lower d/p (determined during the plant startup test program) that would also be an indication of a water-solid system. With pumps operating, the narrow range and upper range meters would indicate off-scale.

If all pumps are shut down, at any temperature, the narrow range and upper range meters would indicate 100 percent, an indication that the vessel is full. The wide range d/p meter would indicate about 33 per-cent of the span of the meter, which would be the value (determined during the test program) co'rresponding to a full vessel with pumps shut down. ,

In the event of a LOCA where coolant pressure has decreased to a prede-termined setpoint, existing emergency procedures would require shutdown of all reactor coolant pumps. In these cases, a level will eventually be established in the reactor vessel and indicated on all of the meters. The plant operator would monitor the meters and the recorder to determine the trend in fluid mass or level in the vessel, and confirm that the ECCS is adequately compensating for the accident conditions to prevent ICC.

l Future procedures may require operation of one or more pumps for recov-ery from certain types of accidents. When pumps are operating while, voids are developing in the system, the pumps will circulate the water and steam as an essentially homogeneous mixture. In these cases, there will be no discernible level in the reactor vessel. A decrease in the 4-23 7581A

measured d/p compared to the nomal operating value will be an indica-tion of voids in the system, and a continuously decreasing d/p will indicate that the void content is increasing, that mass is being lost fran the system. An increasing d/p will indicate that the mass content is increasing, that the ECCS is effectively restoring the system mass content.

4.5 RVLIS ANALYSIS In order to evaluate the usefulness of the RVLIS during the approach to l

ICC, it was decided to detemine the response of the RVLIS under a variety of fluid conditions. The RVLIS response was analytically deter-mined for a number of small break transients. The response was deter-mined by calculating the pressure difference between the upper head and

( lower pianum and converting this to an equivalent vessel head in feet.

l (Note that RVLIS indications will actually be represented by percent of i span) Saturation density at the fluid temperature in the upper plenum was used for this conversion. This approximates the calibration that will be used for the RVLIS.

This indication corresponds to the RVLIS configuration used for non-UHI plants. The conclusions of the study are expected to be the same for l

the UHI configuration. The indication of the upper span (hot leg to upperhead) is not included in this analysis. The upper span indication will be used for head venting operations and will not be used to indi-cate the approach to ICC.

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 is full of water. This reading represents the equiva-lent collapsed liquid level in the vessel which is a conservative indi-cation of the approach to ICC. The RVLIS indication can alert the operator that a condition of ICC is being approached and the existance of ICC can be verified by checking the core exit themoccuples. When the reactor coolant pumps are cperatinn the narrow range RVLIS meter will be pegged at full scale. .

4-24 n..

When the reactor coolant pumps are operating, tne RVLIS reading will oe inaicated on the wide range scale whicn reaas from 0 to 100 percent.

The 100 percent reading corresponds to a full vessel witn all of tne pumps in operation.

Witn tne pumps running the RVLIS reading is an indication of tne voic fraction of the vessel mixture. As the void content of the vessel mix-ture increases, the density decreases and the RVLI5 reading will decrease due to the reduction in static head and frictional pressure drop. The 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 on the narrow range scale.

This fraction approximately corresponds to a vessel mass at wnich would just cover the core if the pump's were tripped.

Four small-break transients under a variety of conditions are discussed in the next section. Three of these cases were obtained from WFLASH

' analyses and the other was ootained from the ICC analysis using NOTRUMP. A description of these codes can ce found in References 1 through 6 in Section 6.0.

The transients included in this report are listed Taole 4.2 whicn alves a brief description of the transient, the plant type, and the model used for the analysis. A discussion of each transient is provided in tne next section. Figures 4-12 through 4-23 provide plots of vessel two-

. phase mixture level, RVLIS narrow range reading, mixture and vessel void fraction, and for Case B with pumps running, RVLIS wide range reading and cold leg mass flowrate..

The two-phase mixture level plotted is that wnich was precicted oy tne codes for the mixture height below the upper support plate. Water in the upper head is not reflected in this plot. The RVLIS reading that would be seen is plotted on the same figure for ease of canparison.

The void fraction plots are for the core and upper plenum fluid volumes. The mixture void fraction includes the volume oelow tne two pnase mixture level . nile tne total void fraction also incluces tne steam space above the mixture level.

4-25 f

4.5.1 Transients Investicated Case A The initiating event for this transient is a 3 inch break in the cold leg. Af ter the break opens, the system depressurizes rapidly to the steam generator secondary safety valve setpoint. Consistant with the FSAR assumptions, the reactor coolant pumps are assumed to trip early in the transient when the reactor trips.

The system pressure hangs up at the secondary setpoint until the loop seal unplugs at approximately 550 seconds, allowing steam to flow out the break and the depressurization continues. The core uncovers while the loop seal is draining then recovers een the loop seal unplugs. The core then begins to uncover again as more mass is being lost through the break than is being replaced by safety injection. The core begins to recover at about 1500 seconds when the acctmulators begin to inject.

This transient does not represent a condition that would lead to ICC but it does represent a break size in the range that would be most probable if a small-break did occur. The response of the RVLIS for typical con-ditions for which it would be used can be investigated with this tran-i sient.

After the reactor coolant pumps trip the RVLIS reading drops rapidly to the narrow rte.ge scale. It f alls until the pressure drop due to flow becomes insignificant compared to the static head of the fluid in the vessel. The first dip in the RVLIS reading is due to the behavior of the upper head.

When the upper head starts to drain it behaves like a pressurizer. The pressure in the upper head remains high until the mixture level drops to below the top of th'e guide tube where steam is allowed to flow from the 4-26

upper head to the upper plenum. When this occurs the upper head pres-sure decreases - thereby increasing the vessel d/p - and the RVLIS reading again more accurately reflects the vessel inventory. This phenomenon is more prevalent for large-break sizes and the effect will be of brief duration for breaks in this range. Furthermore, the ICC guidelines recuire verification of the RVLIS reading through the use of the core exit thermoccuoles. During this pher.cmenon, the core exit thermocouples would read saturation temperature. Therefore, this early phenomena in the upper heao will not cause a false indication of ICC.

When the vessel begins to drain during the loop seal uncovery the RVLIS reading trends in the same direction as the vessel level. The RVLIS reading remains below the vessel mixture level and is therefore a con-servative indication.

When the vessel mixture level increases after the loop seal unplugs the RVLIS reading follows it. Then, RVLIS readings continue to follow the vessel mixture level throughout the transient while underpredicting the actual two-phase level. The wider difference between the RVLIS level and the two-phase level later in the transiert is due to the system j being at a lower pressure which allows mere bubbles to exist in the mixture.

t Case B l

1 This case is the same as case A except it was assumed that the reactor coolant pumps continued to operate unti,1750 seconds. If the reactor coolant pump trip criteria is followed the pumps would be tripped much l earlier in the transient. This case is, however, instructive in deter-l mining the RVLIS response when the pumps are running.

l After the break opens, the system depressurizes rapidly to the secondary safety valve setpoint, and then begins a period of very slow depressuri-zation. During this time the upper portions of the system drain. Due to the reactor coolant pump operation, the two-phase mixture in the l vessel remains at the hot leg elevation, although the void fraction of the mixture continues to increase.

4-27 75alA l

At 750 seconds the system has drained to the point that steam can be vented through the break and the system begins to depressurize more rapidly. The pumps are also tripped at this time resulting in a col-lapse of the mixture in the vessel and the core uncovers.

The vessel continues to drain until the accumulators inject at about 1000 seconds to recover the core. There is a subsequent uncovery which will be ended when the pressure is low enough for the safety injection to make up for mass lost through the break.

During the early portion of the transient the wide range RVLIS reading drops fairly smoothly from 100 percent to about 20 percent, which is due to the decreasing mass in the ve'ssel and the decreasing pressure drop as the pump perfcrmance is degraded. The plot of cold leg mass flowrate is indicative of the pump degradation. The oscillations in this plot are due to alternate steam and two-phase flow predicated by WFLASH. When the flow through the pump becomes mostly steam, the increasing void fraction of the vessel mixture becomes the predominant factor in the decreasing RVLIS reading.

RCP operation keeps the steam and water mixed enough that the mixture level does not f all below the hot legs, although the mixture void frac-tion is increasing during this time. This loss of inventory is indi-cated by the continued drop in the RVLIS reading. When the pumps trip, the steam and water in the mixture separate and there is a rapid decrease in the core mixture level and mix ture void fraction although the vessel void fraction continues to rise. The fact that mass is being redistributed rather than lost is'seen in the RVLIS reading - there is little change in the reading (compared to the change in level) from 750 seconds to the time that the acctsnulators come on.

The prolonged reactor coolant pump operation has caused the downcomer to drain so that when the accumulators come on the cold accumulator water condenses steam in the downcomer causing a local depressurization. The downecrner pressure 'is then temporarily lower than the upper head pres-sure due to inertia and the RVLIS reading becomes temporarily negative.

4-28 75EIA

l l

1 This period of erratic indication is brief (one or two minutes). The pressure will equilibrate and the RVLIS will resume following the vessel mixture level. This phenomenon nas only been observed wnen the accumu-lators inject wnen the oowncomer is nighly voicea. There is no apparent discrepancy during accumulator injection when Enere is a significant amount of water in the downcomer. It is believed tnat this effect is exaggerated by the modeling techniques used in WFLASH (which utilize a homogenous equilibrium assumptions at the accumulator injection loca-tion). For the remainder of the transient the RVLIS reading follows the vessel level closely.

Case C The initiating event for this transient is the opening of the pressur-izer power operated relief valves (PORVs). The reactor coolant pumps and the reactor trip early in the transient on a low pressurizer pres-sure signal consistent with FSAR assumptions. Auxiliary feedwater is availaole in this case but, no pumped safety injection is assumed.

The pressurizer mixture level rises to the top of the pressurizer early in the transient and stays at this level throughout most of the tran-

! s ient. The flow througn the PORVs alternates between steam and twopnase mixture while the pressure in the system drops rapidly to the steam generator secondary safety valve setpoint. The pressure hangs up at this value until the upper portion of the system has drained and then continues to decrease. When the upper portions of the primary system (excluding the pressurizer) have drained the vessel mixture level oegins to decrease and continues until the core canpletely uncovers.

The RVLIS reading drops rapidly to the narrow range span after the reac-tor coolant pumps are tripped. When the vessel level reacnes the hot leg elevation the calculated RVLIS readings begin to oscilate due to the modelling used in WFLASH. In WFLASH, the hot legs are connected to tne

! vessel by point contact connections. This modelling tecnnique causes the hot leg flow to alternate between steam and two pnase flow. The l

oscillitary benavior of the calculated RVLIS reading continues wnile tne l

4 29

level remains at the hot legs. The average calculated value during this period of time shows that the RVLIS reading is a conservative indication of the mixture level.

When the vessel mixture begins to decrease, the RVLIS reading decreases as well. - The RVLIS continues to underpredict the two-phase mixture level and to follow the trend.

Case 0 This case is one of the transients investigated for the ICC study using NOTRUMP. A more detailed discussion of this transient can be found in Reference 1.

The RVLIS reading is below the vessel mixture level throughout most of the transient and is therefore a conservative indication. The RVLIS reading follows the same trend as the vessel mixture level except for early in the transient when the mixture void fraction is fluctuating.

Included in the plots for this case is a ' comparison of the mass inven-tory in the core and upper plenum regions to the RVLIS reading. This ,

comparison shows that the RVLIS reading also corresponds very well with the relative vessel mass inventory. Also included is a comparison for the UHI and non-UHI RVLIS configurations. For the UHI RVLIS configura-tion, the pressure difference is measured from the hot leg to the lower l plenum rather than the upper head to lower plenum. This plot shows a very good comparison between the two systems, indicating that either will give a useful indication.

4.5.2 Observations Of The Study l

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 system conditions will change at a slow enough rate that the operator will be able to use the RVLIS information as a basis for some action.

4-30 7581A

For larger breaks, the response of the RVLIS will be more erratic, due to rapid pressure changes in the vessel, in the early portion of the blowdown. The RVLIS reading will be useful for monitoring accident recovery, when othe- corroborative indications of ICC could also be observed.

Very few instances have been identified where the RVLIS may give ar.

amibiguous indication. These include a break in the upper head, accumu-lator injection into a highly voided downcomer, periods of time when the upper head behaves like a pressurizer, upper plenum injection, and peri-ods 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. 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 the narrow range scale. If the narrow range reading remains at full scale and the wide range reading is grea'ter than that reading which would indicate a full vessel with the reactor coolant pumps tripped, a break in the upper head is indicated. This situation should not cause a problem in detecting ICC because of the para 11e1 logic for the-" kick-out" to the ICC procedures. If the RVLIS indication is erroneous due to a break in the reactor vessel upper head, the operator wi' > begin foi-lowing the ICC procedure if the selected core exit thermocouples read 1200'F.

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 become the limiting factor in the depressurization rate.

This analysis is applicable to all Westinghouse PWR plants, including those plants with upper plenum injection (UPI). The normal condition for continuous UPI occurs only with the operation of the low head safety injection pumps, which does not occur until a pressure of under 200 psi 4-31 7581A

Flow blockage is not expected to decrease the usefulness of the RVLIS indication. The increased d/p due to the flow 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 be useful and, although the indication would be somewhat higher, would continue to follow the trend in vessel inventory.

4.5.3 conclasions

1. With the RCPs tripped, the ' Westinghouse RVLIS will result in an underpredicted indication of vessel level while providing an unambi-guous indication of the mass in the vessel. The Westinghouse RVLIS will also measure the vessel level tr' sonably well.

2. With the RCPs tripped, it is feasible to determine 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.

4. With the RCPs running, the RVLIS is an indication of the mass in the

~

vessel.

S. When the RCPs are running, and the RVLIS reading drops to the narrow range scale, there is significant voiding in the vessel and the core would just be covered if the pumps were tripped.

6. A break of sufficient size in the upper head could cause the RVLIS to give an ambiguous indication of vessel mass. The core exit thermocouples, however, will provide an indication of ICC if appro-priate.

4-33 7551A

7. Accumulator injection when the downcomer is highly voided could result in a temocrarily erratic indication.

S. The RVLIS may significantly underpredict the vessel mass wnile the fluid in the upper head is flashing. However, use of the core exit thermocoupies will preclude a premature entry to the ICC procedures.

9. Rapid void redistributions will not be detected by the RVLIS.

I i

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4-34 1 1581A i

.- . . - . .-_. . _= ..

TABLE 4.1 CONFORMANCE WITH REGULATORY GUIDE 1.97, ORAFT 2. REV. 2 (6/4/80)

FOR THE MICROPROCESSOR OISPLAY SYSTEM Seismic qualification Yes Single failure criteria Yes -

Environmental quailification Yes

• [IEEE-323-1971 applicability]

) Power Source . Class IE Quality Assurance Yes 10CFR50 Appendix B applicability

^

, 01 splay type and method Vertical scale i -

-voltage processed in addition to a recorder i

Unique identification Yes Periodic Testing Yes i

• In some cases IEEE-323-1974 is applicable.

i s.

j 4 35 l 663A MP l

1

. . . I TABLE 4.2 TRANSIENTS INVESTIGATED CASE PLANT DESCRIPTION A 3 loop 3 inch cold leg break - FSAR assumptions *; LFLASH 2775 MWt B 3 loop 3 inch cold leg break - RCPs trip at 750 seconds -

2775 MWt otherwise, FSAR assumptions; WFLASH C 4 loop 2.5 inch break in top of pressurizer - no UHI - ro UHI type pumped safety injection - pumps not running; 3411 MWt WFLASH 0 4 loop 1 inch cold leg break - no high head safety Non-UHI injection; NOTRUMP 3411 MWt

• RCPs tripped at reactor trip, minimum pumped safety injection is available, minimum a>xiliary feedwater is available.

4-36 7581A

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0 5.0 GUIDEt!NES FOR THE USE 0F-ICC INSTRUMENTATION 5.1 REFERENCE 0WNERS GROUP PROCEDURES Based on the analyses defined in Sections 1.3 and 4.5 of this report, Westinghouse and the Westinghouse Owners Group have developed a Refer-ence Emergency Operating Instruction to address recovery from ICC condi-tions ccused by a small-break LOCA without high head safety injection.

This instruction has been transmitters to the NRC via Westinghouse Owpers Group Letter, DG-44, dated November 10, 1980. It should be noted that this instruction was developed on a generic basis as a technical eference for implementing plant specific procedures, and must be tailored to meet plant specific needs.

5.2 SAMPLE TRANSIENT The response of the vessel level indications, other ICC instrumentation and system response during these ICC events and recovery acticns are described in References 1 and 2.

5Y 7581A

6.0 REFERENCES

1. Thompson, C. M., et al., " Inadequate Core Cooling Studies of Scenarios with Feedwater Available, Using the NOTRUW Computer Code," WCAP-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 NOTRUW Computer Code," EAP-9752 (Proprietary) and WCAP-9763 (Non-Proprietary), June 1980.

1

" Report on Small Break Accidents for Westinghouse Nuclear Steam  !

3.

Supply Systen," WCAP-9600 (Proprietary) and WCAP-9601 (Non-Pro-prietary), June 1979.

4. Esposito, V. J., Kesavan, K., and Maul, B. A., "WLASH - A FORTRAN-IV Computer Progran for Simulation of Transients in a Multi-Loop PWR," EAP-8200, Revision 2 (Proprietary) and WCAP-8261, ,

Revision 1 (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-j l prietary) and WCAP-8971 (Non-Proprietary), April 1977.

l l 6. " Analysis of Delayed Reactor Coolant Pump Trip During Small Loss cf Coolant Accident for Westinghouse NSSS," WCAP-9584 (Proprietary) and WCAP-9585 (Non-Proprietary), August 1979, i

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