ML20009G479

Forwards Response to NUREG-0737,Item II.F.2 Re Design Details of Instrumentation for Detection of Inadequate Core Cooling.Nonproprietary Version of Westinghouse Summary Rept Encl

ML20009G479
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	07/30/1981
From:	Delgeorge L COMMONWEALTH EDISON CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8108040315
Download: ML20009G479 (4)

Text

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Commonwealth Edison

) One First National Plaza, Cticago. lilinois O

O' ] Address Reply to: Post Office Box 767 j Ct.;cago, Illinois 60690 July 30, 1981 Oh Mr. Darrell G. Eisenhut, Director f

1 Division of Licensing 9

e r U.S. Nuclear Regulatory Commission la l-s Washington, DC 20555 AUG 0 31981 =-

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Subject:

Zion Station, Units 1 and 2 q

Response to NUREG-0737 Item II.F.2, Outstanding Design Co n

Information NRC Docket Nos. 50-295/304 Reference (a):

Cordell Reed letter to C. G. Eisenhut dated October 18, 1979.

(b):

Cordell Reed lettter to H. R. Denton dated November 21, 1979.

(c):

D. L. Peoples letter to H. R. Denton dated January 1, 1980.

(d):

D. L. Peoples letter to H. R. Denten dated February 22, 1980.

(e):

A. Schwencer letter to D. L. Peoples dated February 29, 1980, (f):

J. S. Abel letter to D. G. Eisenhut dated April 1, 1981.

i (g):

L. O. DelGeorge letter to D. G. Eisenhut dated July 2, 1981.

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Dear Mr. Eisenhut:

Reference (g) committed the Commonwealth Edison Ccmpany to provide by August 1, 1981, the design information for NUREG-0737 Item II.F.2, that we believe to be outstanding for our Zion Station.

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The enclosure to this letter fulfills this commitment by j

providing these outstanding design details of the Zion Station l

instrumentation for detection of inadequate core cooling.

l 8108040315j % h 4s j

PDR ADOCK DR

'_._.~.P,_

I D. G. Eisenhut July 30, 1981 l

To the best of m>

knowledge and belief, the statements contained herein and in the enclosure are true and correct.

In some respects these statements are not based on my personnel knowledge j

but upon information furnished by other Commonwealth Edison employees.

Such information has been reviewed in accordance with Company practice and I believe it to be reliable.

Please address any further questions that you may have concerning this matter _to this office, i

One (1) signed original and thirty-nine (39) copies of this letter are provided for your use.

Due to the voluminous nature of the enclosed Summary Report, only seven (7) copies of this document i

are being provided.

Very truly yours,

/

L. O. DelGeorge j

Director of Nuclear Licensing i

j Enclosure cc:

Messrs. T. A.

Ippolito J. G. Keppler j

Region III Inspector - Zion 1m i

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ENCLOSURE ZION STATION UNITS 1 and 2 RESPONSE TO NUREG 0737 ITEM II.F.2 II.F.2 Instrumentation for Detection of Inadequate Core Cooling (ICC)

Zion S' ation Response The Commonwealth Edison Company has previously addressed this item in References (a), (b), (c), (d), (f) and (g).

The following design information and discussions are provided in order to either fulfill commitments made in the above references or provide an updated status.

1.

Reactor Vessel Level Indication System (RVLIS)

Attached is a non-proprietary version of the Westinghouse l

Summary Report entitled " Westinghouse Reactor Vessel Level Instrumentation System for Monitoring Inadequate Core Cooling".

This document has been edited by the Commonwealth Edison Company i

to reflect the specific design of the system currently installed at Zion Station.

However, some additional differences unique to Zion Statioi are not identified in this Summary Report.

A number of components were not available at the time of system installation nor could they be obtained in order to meet the NRC deadline of January 1, 1980.

These differences are as follows:

1 a)

Since the hydraulic isolators were not avail 9ble at the time of installation, differential pressurt oransmitters were installed inside of containment.

b)

Hiah volume sensors were not installed on Unit 2, and c)

Instrumentation valves were not Installed on Unit 1 nor Unit 2.

This system has been installed and made fully operational in the control room.

It is our understanding that this system will not

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be used as a basis for operator decisions until completion of the NRC pre-implementation review.

No modifications will be made to this system until the NRC has completed its review.

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ll ENCLOSURE,

2.

Subcooling Meters i

In our judgement, all required information has been submitted to the NRC.

Reference (e) contained an NRC staff report titled,

" Evaluation of Licensee's Compliance with Category "A" Items of i

NRC Recommentation Resulting From TMI-2 Lessons Learned".

NRC 1

concluded that......

"The subcooling meter is acceptable in the interim provided that Commonwealth Ediran provides redundant qualified wide-range pressure inputs as soon as possible, but no later than January J, 1981.....

Zion is in compliance with this requirement".

These redundant, qualified wide range RCS pressure transmitters were installed.

Therefore, it is our understanding that this item is complete.

1 3.

Incore Thermocouple System d

Reference (c) provided design details of the existing incore thermocouple system.

This reference also indicated that the environmental qualification of the thermocouples was being investigated and modifications would be made to the extent practicable.

A review of the entire incore thermocouple system is currently underway, both in-house and by Westinghouse on a generic basis.

At this time, no conclusions have been reached.

However, if modifications are considered necessary, it is likely that the implementation date of January 1, 1982 may not be met.

t 4.

Procedures Some of the inadequate core cooling procedures requiring a pre-implementation review have been developed by the Westinghouse Owners Group and have been submitted to the NRC for review.

A schedule has also been submittej to the NRC by the Owners Group covering the submittal of the remaining procedures j

still yet to be developed.

It is our understanding that the NRC does not want procedures in place directing the operator to use this system as a basis for operator action until the pre-implementation review is complete, i

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THE ENCLOSED DOCUMENT IS A NON-PROPRIETARY VERSION OF THE WESTINGHOUSE

SUMMARY

REPORT ENTITLED "WESTINGH0USE REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM FOR MONITORING INADEQUATE CORE COOLING".

THIS DOCUMENT HAS BEEN EDITED BY THE COMMONWEALTH EDIS0N COMPANY TO REFLECT THE SPECIFIC DESIGN OF THE SYSTEM CURRENTLY INSTALLED AT ZION STATION.

2357N

THE ENCLOSED DOCUMENT IS A NON-PROPRIETARY VERSION OF THE WESTINGHOUSE

SUMMARY

REPORT ENTITLED

" WESTINGHOUSE REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM FOR MONITORI::S INADEQUATE CORE CJ0 LING".

THIS DOCUMENT HAS BEEN EDITED BY THE COMMONWEALTH EDISON COMPANY TO REFLECT THE SPECIFIC DESIGN OF THE SYSTEM CURRENTLY INSTALLED AT ZION STATION, i

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SUMMARY

REPORT l

WESTINGHOUSE REACTOR VESSEL LEVEL INSTRUMENTATION l

SYSTEM FOR MONITORING INADEQUATE CORE COOLING (7300 SYSTEM) i i

i Decencer,1980 I

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TABLE OF CONTENTS

1.0 INTRODUCTION

1.1 NRC Requirements i

1.2 Definition of ICC 1.3 Condition or Events Which Describe the Approach to ICC 2.0 FUNCTIONAL REQUIREMENTS 4

2.1 Parameters 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

4.0 RVLIS - SYSTEM DESCRIPTION 4.1 General Description 4.2 Detailed System Description 4.2.1 Hardware Description 4

4.2.1.1 Differential Pressure Measurements 4.2.1.2 System Layout 4.2.2 7300 Series RVLIS 4.2.2.1 RVLIS Inputs 4.2.2.2 RVLIS Outputs 4.2.2.3 Additional 7300 System RVLIS Features 4.2.3 Resistance Temperature Detectors 4.2.4 RVLIS Valves j

4.2.5 Transmitters, Hydraulic Isolators, and Sensors l

4.3 Test Programs 4.3.1 Forest Hills i

4.3.2 Semiscale Tests 4.3.3 Plant Startup Calibration i

7581A l

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TABLE OF CONTENTS (Continued) 4.4 Operating Perfonnance

.J 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 Westinghouse Owners Group Procedure 5.2 Sample Transient

6.0 REFERENCES

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LIST OF FIGURES Figure 4-1 Reactor Vessel Level Instrument System Figure 4-2 Process Connection Scherratic, Train A Figure 4-3 Typical Plant Arrangement for RVLIS Figure 4-4 7300 Series RVLIS Processing Equipment - Block Diagram Figure 4-5 7300 Serieis RVLIS Displaj Figure 4-6 Typical Plant Arrangement For RVLIS Figure 4-7 Block Diagram of Compensation Function 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 Voltne" Sensor Bellows Check Valve Figure 4-12 Case A 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip with Reactor Trip, RVLIS Reading and Vessel Mixture Level i

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 7581A

LIST OF FIGURES (Continued)

Figure 4-16 Case B 3-Loop Plant, 3 Incn 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 75 Seconds, Cold Leg Mass Flowrate (LB/Sec)

Figure 4-18 Case C 2.5 Inch Pressurizer Break, No. SI, RVLIS Reading and Mixture Level.

Figure 4-19 Case C 2.5 Inch Pressurizer Break, No. SI, Void Fraction Figure 4-20 Case D 1 Incn Cold Leg Break, ICC Case, RVLIS Reading ana Mixture Level.

Figure 4-21 Case D 1 Inch Cold Leg Break, ICC Case, Mixture Level, RVLIS Reading and Measured Inventory.

Figure 4-22 Case D 1 Inch Cold Leg Break, ICC Case, RVLIS Reading and Mixture Level.

Figure 4-23 Case D 1 Inca Cold Leg Break, ICC Case, Void Fraction 7581A

. _ - _. = _ -.

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1 LIST OF TABLES i

Table 3.1 Information Required on the Core Subcooling Monitor i

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Table 4.1 Compliance with Regulatory Guide 1.97 Draft 2, Rev. 2 6/4/80

I Table 4.2 Transients Investigated i

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7581A

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1.0 INTRODUCTION

1.1 NRC REOUIREMEN'~

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 of inadequate core cooling (ICC).

Under certain plant accident conditions, the potential exists for the formation of voids in the reactor coolant system (RCS). Under 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 level or relative void content of the circulating primary coolant system fluid.

1.2 DEFINITION 0F 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 occur.

1-1 7581A

l 1.3 CONDITIONS OR EVENTS WHICH DESCRIBE THE APPROACH TO ICC The most obvious failure that would lead to ICC during a small-break LOCA, although highly 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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WESTINGHOUSE PROPRIETARY CLASS 2 d

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

)

j To briefly sunnarize those parameters, ICC is detected by either high core exit thericocouple te.Tperatures or by a icw reactor vessel lesel

]

indication (core uncovery) in conjunction with core exit therraoccuple indicaticns. Mitigation actions consist of depressurizing the reactor j

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.

j Critical parameters at this point are steam gene'rator pressures and wide i

range RCS loop temperatures. Once icw head safety injection flow is established, trcnsfer oe', of the ICC procedure can be made when core exit thermocouple temperatures are reduced and the reactor vessel level i

gauge indicates a level above the top of the core.

With the exception of reactor vessel level, all param2ters are n'onitcred l

by currently existing instrumentation, f

2.2 'IfiSTRUMENTATION ACCURACIES;-RANGES; AND TIME RESPONSE Accuracy l

l An accuracy of 6 percent is required on bo4h types of reactor i vessel level instruments. This should be a statistical combination of i

l all uncertainties including those due to envirer.:i.entcl effects (if any) l on instrumentation.

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l the narrow range instrument this correspcnds to an allowable deviation 2-1 7581A l

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IESTIfMHOUSE PROPRIETARY CLASS 2 i

f of about + 2.5 feet elevation head. This is required to: 1) provide l

adequate margin against inadvertant use of the.ICC operating guideline (E 01-1, see Section 5.1), 2) assure that'the vessel level reading can 2

l be reasonably used to aid in the detectio.1 of the onset of ICC condi-tions, 3) derive useful information reguarding vessel level behavior during the vessel refiil perlod of a LOCA transicot.

Range i

The wide range instrument will cover the full range of expected differ-j The maximum ential pressures with all reactor coolant pumps running.

j span of the wide range instrument will change with the number of purps i

operating. The operator must be aware of the maximum span for a given l

number of operating pucps.

The narrow range l

instrument indication should be sec to indicate that the vessel is full l

with the purrps tripped.

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Time Response 1

This time The d/p instrument response time shall not exceed 10 seconds.

1 delay is defined as the time required for the display instrument to i

reach the midpoint of a 50 percent step input d/p change.

j 2.3 QUAllFICATION REQUIREMENTS Envircnmental qualification of the RVLIS shall verify that the system equipment will meet, on a continuing basis, the performance requirements j

determined to be necessary for achieving the system requirements as Verification must include confirmation that those presented above.

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 requirement for electrical equipment inside containment is 120 days 2-2

~ ' ' " - ' " - ~

7

r following certain postulated events. The electrical equipment that is installed outside of containment need not meet a qualified life for an extended period of time providing replacement or calibration checks can be made in snort enough time commensurate with the reliability goals of the redundant system. For the resistance temperature detectors (RTDs) environmental requirements for service within the containment, refer to Section 4.2.3.

Electrical equipment inside containment snail be instal-led such that it is removed from areas where high energy pipe breaks or pipe wnip could cause failure. 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 located 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 snail be netween 60 and 80*F and 0 to 50 percent relative humidity. The instrumentation snall be qualified to assure that it continues to operate and read within the 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 witn the following Codes and Standaros:

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

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e 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 nd 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, "IEEE Trial-Use Standard: General Guide for Qualifying Class 1 Electric Equipment for Nuclear Power Generating Stations"*

IEEE-338-1971, "IEEE Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems" i

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 Equipment and Circuits" ASME BPVC,Section III, Class 2 Nuclear Power Plant Components l

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

For certain specific plants, IEEE-344-1975 is applicable.

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ANSI B31.1.0, 1967 including addenda through and including 6/30/71,

" Code for Pressure Piping", including nuclear code cases where applicable i

Regulatory Guides R.G. 1.11 Instrument Lines Penetrating Primary Reactor Containment R.G. 1.22 Periodic Testing of Protection System Actuation Functions R.G. 1.75 Physical Independence of Electric Systems 2-5 7581A

3.0 ICC INSTRUMENTATION IDENTIFICATION Adequate instrumentation is necessary to diagnose the approach to ICC and 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 derivable from the addition of new instrumentation. The following is a list of existing instrumentation considered (refer to tne 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 i

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 l

will generally lie outside the ranges of these instruments l

during an ICC event in a Westingnouse PWR. Pressurizer pres-sure and level are not used for determining mitigation actions to be taken during ICC.

c.

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

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d.

WIDE RANGE RESISTANCE TEMPERATURE DETECTCRS - present instru-l mentation is available in determining trends of recovery actions but may not be available in determining the onset of ICC conditions for all break 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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CORE SUBC00 LING - does not provioe useaole information during an ICC condition. Will indicate superneat conditions in core cool ant. Will help indicate the approacn to ICC by showing saturation conditions. Since the core suocooling monitors may not be described in the FSAR, refer to Table 3.1 for informa-tion.

9 STEAMLINE PRESSURE - present instrumentation is availaole for determining heat sink availabilty and heat removal capability during ICC mitigation actions.

h.

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

2.

New Instrumentation a.

REACTOR VESSEL LEVEL - provices an indication of the approacn to ICC and confirms the achievement of adequate core cooling wnen level in the reactor vessel is restoreo3 To summarize the above considerations, current plant instrumentation is j

adequate to oetermine heat sink availability, to detect the onset of ICC, and to detect the effectiveness of mitigation actions following the onset of an ICC event. To permit a more continuous indication of tne approach to ICC, the RVLIS is required.

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CORE SUBC00 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 subcooling monitors may not be described in the FSAR, refer to Table 3.1 for informa-tion.

g.

STEAMLINE*PREtSURE - 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 summarize the above considerations, current plant instrumentation is adequate to determine heat sink availability, to detect the onset of ICC, and to detect the effectivaness 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.

32 7581A

TABLE 3.1 INFORMATION REQUIRED ON THE CORE SUBC00 LING MONITOR Disolay Infomation Displayed (T-Tsat, Tsat, P-Psat subcooled press,etc.)

T-Tsat superheat Display Type (analog, digital, CRT)

Analog and digital Continuous or on Demand Analog - continuous Digital - on demar.d Single or Redundant Oisplay Redundant Location of Display User supplied Alarms Caution - 250F subcooled for RTD Alarm - 00F subco31ed (include setpoints) 15 F subcooled for T/C for RTD and T/C Overall Uncertainty (oF, psi)

Digital - 4ef for T/C; 3 F for RTO Analog - 5*F for T/C; 5 F for RTD Range of Calibrated region - 1000 psi subcooled to 20000F t,uperheat Display Overall - never offscale Qualifications None at present*

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

If process computer is used, specify availability N/A (percent of time)

Single or Redundant Calculators Redundant Selection Logic (highest T., lowest press)

Highest T for RTD or T/C; Lowest P Qualifications None at present Calculational Technique (steam tables, Functional fit -

functional fit, ranges) ambient to critical point

• The display is currently undergoing seismic qualification testing by Westinghouse which will conform to IEEE-344-1971. This infonnation will only be provided at the specific request of the customer and after appropriate installation checks have been made to verify the applicability of this qualification.

3-3

TABLE 3.1 (Continued)

Inout Temperature (RTDs or T/Cs)

RTO, T/C and Tref Temperature (number of sensors and locations)

RTO - 2 hot and 2 cold leg per channel T/C - 8 per channel Range of Temperature Sensors RTD 700 F T/C 1650 F (calibration unit range 0 - 2300 F)

Uncertainty

• of.Tenperature 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 Availability of Temp and Press Availability of Steam Tables etc.

Procedures Uncertainties must address conditions of forced flow and natural circulation 3 -4

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WESTINbliOUSE Pli0f Rll'l ARY CLAD 5 2 4.0 REACTOR VESSEL LEVEL INSTRU:tENTATON SYSTEM - SYSTEM DESCRIPTION 4.1 GENERAL DESCRIPTION The reactor vessel level instrum2ntation system (RVLIS) uses dif feren-tial pressure (d/p) measuring devices to measure vessel level or rela-tive void content of the circulating primary coolant system fluid. The system is redundant and includes autcmatic ccmpensation for potential temperature variations of the impulse lines. Essential information is displayed in the main control room in a form directly useable by the operator.

The functions perform 2d by the RVLIS are:

1.

Assist in detecting the approach to ICC 2,

Indicate the formation of a void in the RCS during forced flow conditions.

4.2 Df. AILED SYSTEtt DESCRIPTION

4. 2

• HARD'..%RE DESCRIPTION

4. 2.1.1 Differential Pressure Measuremants l

The RVLIS (Figure 4-1) utilizes two sets of 4wo d/p cells. These cells m235ure the pressure drop from the bottom of the reactor vessel to the top of the vessel, This d/p measuring system utilizes cells of differing ranges to cover differcnt flow behaviors with and without pump operation as discussed below:

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MESTINGHOUSE PROPRIETARY CLASS 2 1.

Reactor Vessel - Narrow Range (aP )

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

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Reactor Vessel - Wide Range (AP )

c This instrument provides an indication of reactor core and inter-nals pressure drop for any ccmbination of operating RCPs. Com-parison of the measured pressure drop with the normal, single-phase pressure drop will provide an approximate indication of the I

relative void content or density of the circulating fluid.

This instrument will monitor' coolant conditions on a continuing basis during forced fica conditions.

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To provide the required accuracy for level measurement, terparature 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 structure following an accident.

The d/p cells are located outside of the containment to eliminate the l'rge reduction (approximataly 15 percent) of measurement accuracy asso-a ciated with the change in the containment environment (temperature, pressure, radiation) during an accident. The cells are also located outside of containm2nt so that system operation including calibration, cell replacement, reference leg checks, and filling is made easier.

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WESTifiGHOUSE PROPRIETARY CLASS 2 4.2.1.2 System Layout A schematic of the system layout for the RVLIS is shewn in Figure 4-2.

RCS penetrations for the cell reference lines; one reac-There are uo

- tor head connection at a spare penetration near the center of the head or the reactor vessel head vent pipe,o.done cennection to an incore x

instrument conduit at the seal table, The pressure sensing lines extending from the RCS penetrations will be a ccabination of 3/4 inch Schedule 160 piping and 3/8 inch tubing and will include a 3/4 inch manual isolation valve as described in Section 4.2.4.

These lines connect to fos sealed caoillary imDulse lines (two atthereactorhead,twoatthesealtable) which transmit the ' pressure measurements to the d/p transmitters located outside the containment building.

The capillary irpulse lines are sealed at the RCS end with a sensor bellows which serves as a hydraulic coupling f or the pressure measurement. The icpulse lines extend from the sensor bellows through the centainment wall to hydraulic isolators, which also provide hydraulic coupling as well as a seal and isolation of the lines. The capillary tubing extends fron the hydraulic isolators to the d/p transmitters, where instrument valves are provided for isolation and bypass.

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 rust be routed upward out of the refueling cmal to the operating deck, 1

I then radially toward the seal table =nd then to the containment penetra-tion. The cor.nection to the bottom of the reactor vessel is made through an incore dctector conduit which is tapped with a T connection I

at the seal table. The impulse line from this conrection is routed

' av.iolly and radially to join with the head connection line in routing to the penetrations.

I 4-3 7581A

WESTINGliOUSF PROPRIETARY CLASS 2 The impulse lines located inside the containment ouilding will be exposed to the containnent temperature increase during a LOCA or llELB.

Since the vertical runs of impulse lines form the reference leg for the d/p m2asurem2nt, the change is density due to the accident temperature change must be taken into account in the vessel level determination.

Therefore, a strap-on RTD is located on each vertical run of separately routed impulse lines to determine the impulse line te.nperature and cor-rect the reference leg d2nsity contribution to the d/p measurer.2nt.

b ot h irpulse lines Tc perature measurements are not required where of an instr'umant train are routed together.

Based on the studies of a number of representative plant arrangements, a maximun of 7 independant vertical runs rust be masured to adcquately compcosate for density changes.

4.2.2 7300 SERIES RVLIS The 7300 series RVLIS is configured as two trains (protection sets) in the outer bays of a standard three-bay cabinet.

The system uses the sc.e components and cabinet that is used in the 7300 series nuclear protection and control systems. The block diagram of the process equip-ment is shaan in Figure 4-4.

For displayed information, see Figure 4-5.

Conf ormance with Regulatory Guide 1.97 for tne 7300 display systen is given in Table 4.1.

4.2.2.1 RVLIS Inputs The 7300 series process equipment inputs are as follows.

If existing unqualified inputs are used, isolation as required will be provided by I

the caner.

Hot leg Wide Range Ter.perature Each RVLIS train receives two hot leg wide range tcmperature signals derived from existing channels in the NSSS Process Protection System.

4-4

WESTING'100SE PROPRIETARY CLASS 2 The hot leg temperature signal is used to comper. sate the measured reac-tor vessel d/p to produce an indicated liquid level value during condi-tions wher. the liquid is subcooled.

Wide Range RCS Pressure Each RVI.IS train receives one wide range RCS pressure signal derived from existing channels in the NSSS Process Protection System. The RCS pressure signal is used to compensate the ;.'easured reactor vessel d/p to produce a liquid level value during conditions when the coolant is satu-rated. The selection between temperature and pressure compensation is automatic.

4-$

Reactor Vessel Narrow Range Differential Pressure Each train receives one narrow range d/p measurement. This signal is provided from a new transmitter and when compensated, yields the level indication spanning the entire reactor vessel during periods when the reactor coolant pumps are not running.

a,c APb gives an indication of reactor vessel level when no pumps are run-ning.

If one or more pumps are running, AP w'i.1 be off-scale and the b

reading invalid.

The sense of the S output is such that a 20 ma signal is a nomir;sily b

full vessel and a 4 ma signal is for a nominally empty vessel.

Reactor Vessel Wide Rang ( Differential Pressuree Each train receives one wide range d/p measurement. This signal is provided from a new transmitter and when compenseted, would yield the relative void content of the circulating primary coolant system fluid during periods when any reactor coolant pumps are running.

a,c

]ThesenseoftheAP output is that 20 m3 represents c

all pumps running and 4 ma is empty vessel. With all pumps running and no void fraction, the AP should read 100 percent at zero power. The c

reading at full power is slightly higher.

4-6 7581A

Capillary and Conduit Temperature Each train receives up to 7 temperature measurements from new RTDs.

These RTDs provide compensation signals used to cancel out temperature induced d/p effects on the instrumentation system.

A typical arrangement of the reference leg temperature RTDs is shown in Figure 4-6.

The conversion of RTD resistance to temperature shall cover the tempera-ture range of 32 to 450 F.

The RTCs are '.00 ohm platinum four wire RTDs.

Density Calculation Each of the b e d/p measurements will have density corrections from certain temperature measurements. Some of these will have a positive correction and some negative depending on the orientation of the impulse line where the temperature is being measured.

Vessel Liquid Density Calculation a,c Vessel Vaoor Phase Density Calculation a,c d_7

l Vessel Level Calculation

_. a,c Pumo Flow d/o Calculation

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4.2.2.2 RVLIS Outputs Plant Operator Interface and Displays l

Information displayed to the operator for the RVLIS is intended to be unambiguous and reliable to minimize the potential for operator error or misinterpretation.

4-8 7581A

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i WESTItG110VSE PROPRIETARY CLASS-2 i

Level Indication Narrow range and wide range level signals are available i

from each train for display on standard VX-252 t3pe vertical scale volt-j age meters. Thus, the indication is compatible with existing control board layouts. The indication signals are electrically isolated from the protection set and are suitable to serve as either a standard con-trol grade or post-accident monitoring output.

i 4

The control board displays provide the following information:

1 1.

An indication of reactor vessel level (narrow range) for each instrumented set displaying vessel level in percent from 0 to 100 percent after compensation for the effects of the reactor coolant and capillary line temper ature and density, when reactor coolant i

pumps are not operating.

2.

An indication of reactor d/p (wide range) from each instrumented j

I set displaying d/p in percent from 0 to 100 percent, after compen-i sation for the effects of the reactor coolant and capillary line l

temperature and density effects, v. hen reactor coolant pumps are i

operating.

i i

i I

Redundant displays are provided for the two sets. Level information 1

based on both d/p teasurcments is presented.

'orrection for refer-ence leg densities is automatic.

l t

I 4_9-7581A

WESTINGliOUSE PROPRIETARY CLASS 2 Level Recording k' arrow range and wide range level signals are available on one of the two trains for trending on a chart recorder. These signals are standard 0 to 10 volt range and are electrically isolated from the protection set. Thus, they are suitable for either control grade or post-accident monitoring applications.

Corputer Outputs Analog inputs to the plant computer are available to monitor each of the RTD and transmitter inputs plus each of the compensated level outputs.

These outputs are standard 0 to 10 volt ranges and may be used by the plant co,puter to do an independent compensation calculation of the reactor vessel d/p.

4.2.2.3 Additional 7300 Series RVLIS Features 1.

The 7300 series RVLIS features full systems testing capability without having to lif t wires at the termination area. Test injec-l tion points and test measurement points are available throughout the system to f acilitate ease of calibration and maintenance.

RVLIS channels are designed to permit maintenance on one channel during pc.;er operation. During such operation the active parts of the system need not themselves continue to meet the single f ailure criterion. As such, monitoring systems comprised of two redundant chtnnels are permitted to violate the single f ailure criterion during maintenance provided that acceptable reliability of opera-tion f or the channel not under maintenance can be demonstrated.

l TFe time interval a'llowed for a maintenance operation will be specified in the plant Technical Specifications. Bypess indicaticn may be applied administratively.

I r

4-10 7581A

1 1

2.

The 7300 series RVLIS has card edge adjustments and settings for ease in scaling in modifications due to changes in the installation layout. All systems set-up may be performed by a field technician rather than requiring offsite calibration by a specialist.

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 vertical instrument lines is used to correct the vessel level indication for density changes associated with j

the environmental temperature change.

The RTD assemb;/ 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 surface of the tubing or piping being monitored. The materials are all selected to be compatible with the normal and post-accident environment. Randomly selected samples frun the controlled (material, manufacturing, etc.) production lot will be qualified by type testing.

Qualification testing will consist of thermal aging, irradiation, seis-mic testing and testing under simulation high energy line break environ-mental conditions. For the qualified life requirements, see Section l

2.3.

The specific qualification requirements for the RTDs are as fol-lows:

1.

M The thermal aging test will consist of operating the detectors in a high tenverature 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.

4-11 7581A w.

l 2.

Radiation The detectors shall be irradiated to a total integrated dose (TID) of 1.2 x 108 rads gamma 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 blaxial seismic simulation.

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

4.

High-Er.ergy Line-Break -Simulation The detectors shall be tested in a saturated steam environment using the temperature / pressure curve shown in Figure 4s9, 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.

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 from 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 7581A r--.

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.

3.

Must displace minimal volumes of fluid in normal and overrange oper-ating modes.

The first two requirements are related to the vernier characteristic of the pumps off level measurements and the wide range measurements, respectively. The third is related to the limited driving displacement of the hydraulic isolator when preserving margins for pressure and 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 ratings to full working pressure.

4-13 4

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 cf a RVLIS was instal-led and tested at the Westinghouse Forest Hills Test: Facility. The system consisted of a full single train of RVLIS hyJraulic cog onents (sensor assemblies, hydraulic isolators, isolation and bypass valves and d/p transmitters) connected to a simulate

• reactor vessel. Process i:ennections 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 assen611es 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 tem erature detec-tors, special large volume sensor bellows and volume displacers inside 4-15

the hydraulic 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 program.

The hydraulic isolator assemblies and transmitters were mounted at an elevation slightly below the simulated seal table elevation.

The objectives of the test were as follows:

1.

Obtain installation, filling and maintenance experience 2.

Prove and establish filling procedures for initial filling and system maintenance.

3.

Establish calibration and fluid inventory maintenance procedures for shutdown and normal operation conditions.

4 Prove long term integrity of hydraulic components 5.

Verify and quantify riu;d 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, side and bottom to simulate the reactor head, hot leg and incore detector thimble conduit penetra-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 elevhticr4 and the hot leg sensor to the head connection was simulated by 1 inc5 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 7531A

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 a,c a

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_ a,c 4.3.2 SEMISCALE TESTS In order to study the transient response of the RVLIS during a f

small-break LOCA and other accident conditions, the hydraulic components of the RVLIS have been installed at the Semiscale Test Facility in l

Idaho. Vessel level measurements will be obtained during the current l

semiscale test prot. ram series which runs from December 1980 to March 1982. Tne 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 camp'etion by November 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 heated length of 12 feet. Two reactor coolant loops are provided, each having a pump and a steam generator with a full Height 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

(

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 l

versel and loops to record the fluid conditions throughout a test run.

Test results are compared with predictions for verification of computer l

code models of the transient performance.

l l

4-18 7581A

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 methods of measuring the level or fluid density: d/p measurements are obtained j

over 11 vertical spans on the reactor vessel to determine level within l

each span, and garsna 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 from the 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 octained 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 r odified 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.

l l

S.

Tests to simulate a station blackout with discharge through relief l

valves.

6.

Simulation of the St. Lucie cooldown incident.

l 4-19 7581A

i 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-l l

ments of the program are described below:

1.

During refilling and venting of the reactor vessel, measurements of all4 d/p transmitters would be compared to confirm identical level 1

indications.

2.

During plant heatup with all reactor coolant pumps running, measure-ments would be cbtained from 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 variatior during startup, corrects the transmitter output so that the control board indication is maintained at 100 percent over the entire operating temperature l

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 l

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.

I

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4-20 l

7581A

WESTlHGHOUSE PROPRIETARY CLASS 2 4

w i

j 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 4u-d/p transmit-ters, each transmitter covering a specific range of operating condi-tions. The tuxa instrument ranges orovide overlap so that the j

measurement can be obtained from more than one meter under most accident conditions. Capabilities of each of the measurements are described below:

i l

i 1

8 l

1.

Reactor Vessel - Narrow Range i

The transmitter sp'an covers the total height of the reactor vessel.

With pumps shut down, the transmitter output is an indication of the I

collapsed water level, i.e.,

as if the steam bubbles had been separ-l ated frcm 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 I

provides a cons trvative indication of the level effective ;for ade-

]

quate core cooling.

1 I

i 4-21 75SIA k

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WESTINGliOUSE PROPRIETARY CLASS 2 When reactor coolant pumps are operating, the d/p would be greater than the transmitter span, and the transmitter output would be dis-regarded.

2. Reactor Vessel - Wide Range The transmitter span covers the entire range of interest, from all puTps operating with a water-solid system to a completely empty reactor vessel and therefore, covers the measurement spans of the other tvlo 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 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 roeasurement, whichever results in the highest value of water den-sity, and, therefore, the lowest value of indicated level. Ccmpensation based on temperature is applied when the system is subcooled, and com-i 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 ir the reference leg during an accident with elevated temperature in the containment and the fluid in the reference leg at the initial ambient temperature. The compensation is based on temperature measurements on the vertical sections of the reference leg.

- 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 1581A

~......

UtSTIIGif00SE PROPRIETARY CLASS 2 level or relative d/p and to display trends in the measurements.

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 meter would indicate of f-scale.

If all punus are shut down, at any temperature, the narrow range meter vould 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) corresponding 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 eperator would r.onitor the meters and the reccrder to i

determine the trend in fluid mass or level in the vessel, and confirm that the ECCS is adequately compensating for the accident conditions to i

prevent ICC.

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 decrea'se in the 4-23 l

7581A

l i

I t

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.

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

ICC, it was decided to determine the response of the RVLIS under a variety of fluid conditions. The RVLIS response was analytically deter-j j

mined for a number of small break transients. The response was deter-mined by calculating the pressure difference between the upper head and lower plenum and converting this to an equivalent vessel head in feet.

(Note that RVLIS indications will actually be represented by percent of span) Saturation density at the fluid temperature in the upper plenum l

was used for this conversion. This approximates the calibration that will be used for the RVLIS.

I l

This indication corresponds to the RVLIS configuration used for non-UHI pl ants. The conclusions of the study are expected to be the same for 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 thermocouples. When the reactor coolant pumps are operating the narrow range RVLIS meter will be pegged at full scale.

4-24 en s

1 l

l When the reactor coolant pumps are operating, tne RVLIS reading will De 1

inoicated on the wide range scale whicn reads from 0 to 100 percent.

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

l With the pumps running the RVLIS reading is an indication of the voio l

fraction of the vessel mixture. As the void content of the vessel mix-(

ture increases, the density decreases and the RVLIS reading will l

decrease due to the reduction in static head and frictional pressure l

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 approxim&tely corresponds to a vessel mass at which would just cover the core if the pumps were tripped.

Four small-break trans;ents under a variety of conditions are discussed in the next section. Three of these cases were obtained from WFLASH analyses and the other was obtained from the ICC analysis using NOTRUMP. A description of these codes can De found in References 1 through 6 in Section 6.0.

The transients included in this report are listed Table 4.2 whicn gives a brief description of the transient, the plant type, and tnc model used i

for the analysis. A discussion of each transient is provided in the next section. Figures 4-12 through 4-23 provide plots of vessel two-phase mixture level, RVLIS narrow range reading, mixture and vessel void l

fraction, and for Case B with pumps running, RVLIS wide range reading and cold leg mass flowrate..

1 The two-phase mixture level plotted is that which was preoicted by 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 comparison.

Th. void fraction plots are for the core and upper plenum fluid l

volumes. The mixture void fraction includes the volume below the two phase mixture level while the total void fraction also incluoes tne steam space above the mixture level.

i 4-25 l

4.5.1 Transients Investigated Case A The initiating event for this transient is a 3 inch break in the cold l eg. Af ter the break opens, the system depressurizes rapidly to the steam generator secondary safety valve setpoint. Consistent 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 setp.oint until the loop real 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 when 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 accumulators 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 snall-break did occur. The response of the RVLIS for typical con-ditions for which it would be used can be investigated with this tran-sient.

After the reactor coolant pumps trip the RVLIS reading drops rapidly to the narrow range 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

-e-.,

~

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 thermocouples. During this phenomenon, the core exit thermocouples would read saturation temperature. Therefore, this early phenomena in the upper head will not cause a false indication of ICC.

When the vessel begins to drain during the loop seal uncovery the RVLIS l

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 transient is due to the system being at a lower pressure which allows more bubbles to exist in the mixture.

Case B This case is the same as case A except it was assumed that the reactor coolant purps continued to operate until 750 seconds. If the reactor coolant pump trip criteria is followed the pumps would be tripped much earlier in the transient. This case is, however, instructive in deter-mining the RVLIS response when the pumps are running.

After the break opens, the system depressurizes rapidly t0 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 vessel remains at the hot leg elevation, although the void fraction of the mixture continues to increase.

4-27 7581A

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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 accmulators 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 vessel and the decreasing pressure drop as the pump performance 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 oy 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 r.ontinued 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 mixture void fraction although the vessel void fraction continues to rise. The f act 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 accmulators come on.

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

4-28 7581A

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 pnenomenon has only been observed when the accumu-lators inject wnen the downcomer is nighly voided. There is no apparent discrepancy during accumulator injection when tnere is a significant l

amount of water in the downcomer. It is believed that this effect is exaggerated by the modeling techniques used in WFLASH (which utilize a f

homogenous equilibrium assumptions at the accunulator 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 tne pressur-izer power operated relief valves (PORVs). The reactor coolant ptsnps and the reactor trip early in the transient on a low pressurizer pras-sure signal consistent with FSAR assumptions. Auxiliary feedwater is availaDie 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-sient. The flw througn the PORVs alternates between steam and twopnase mixture while the pressure in the sy.em 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 begins to decrease and continues until the core completely uncovcrs.

The RVLIS reading drops rapidly to the narrow range span after the reac-tor coolant pumps are tripped. When the vessel level reaches 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 the vessel by point contact connections. This modelling technique causes the hot leg fim to alternate between steam and two pnase flow. The oscillitory behavior of the calculated RVLIS reading continues wnile tne 4-29 e-,.

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 D 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 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 Ob ervations Of The Study The RVLIS will provide useful information for breaks in the system ranging from small leaks to breaks in the limiting small-break range.

For breaks in this range, the 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

I 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 other corroborative indications of ICC could also be observed.

Very few instances have been identified where the RVLIS may give an amibiguous indication. These include a break in the upper head, accumu-later 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 readir.g remains at full scale and the wide range reading is greater 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 parallel logic for the " kick-l out" to the ICC procedures. If the RVLIS indication is erroneous due to a break in the reactor vessel upper head, the operator will begin fol-lowing the ICC procedure if the selected core exit thermccouples 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 plenun 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

/

WESTINGHOUSE PROPRIETARY CLASS 2

'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 mix with the steam in the upper plenum

/

causing condensation to accumulate. This condensation will form faster than the system response. The system will equilibriate after a short period of time. Upon equilibrating, the system wil'. continue to accu-rately trend reactor vessel level. For the vast majority of small-breaks, the condition of upper plenum injection does not cause a signi-ficant impact.

For the remainder, the impact is very small and within tolerable limits.

As discussed elsewhere in this section, the time when ambiguous indica-tions due to accumulator injection and upper head pressurizer behaviour is br9f. The situation corrects itself and the RVLIS resumes giving a good indication of the trend in level.

Both situations result in an indication of vessel level that is low. The operator must know that a brief period of erratic RVLIS indication may occur when accumulators are injecting. This effect is partially real in that the vessel level me; depress for a moment when accumulator injection occurs.

Unlike accumu-lator injection, the operator will not know when the indicated vessel level is being affected by the upper head pressurizer phenomena.

However, no prematore indication of ICC will occur since the core exit thermocouples will still read saturation terrperature.

During periods when the void dis'.ribution in the vessel is changing rapidly, there may be a large change in two-phase mixture level with l

very little change in mass inventory in the vessel. This could happen if the reactor coolant purrps (RCPs) were tripped when the mixture in the l

vessel was highly voided. This could cause the mixture level to drop l

from the hot leg elevation to below the top of the core.

The operator l

would expect this to happen based on the fact that the RVLIS reading was within the narrow range indication. The operator should know in general that, for a brief period of time after tripping the RCPs, transient RVLIS response will occur.

l 4-32 s

=>

i 7FPM

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 ptsnps 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.

l 4.5.3 Conclusiens 1.

With the RCPs tripped, d c 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 trend reasonably 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 l

vesscl.

l S.

When the RCPs are running, and the RVLIS reading d'rops to the narrow l

range scale, there is significant voiding in the vessel ar.d 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 ther:noccuples, however, will provide an indication of ICC_if appro-priate.

l 4-33 7581A

7.

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

8.

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

)

9.

Rapid void redistributions will not be detected by the RVLIS.

i 4-34 7581A n

g,~,rwysw -

<n, r - --

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

FOR THE 7300 DISPLAY SYSTEM Seismic Qualification Yes Single failure criteria Yes Environmental quailification Yes

• [IEEE-323-1971 applicability]

Power Source Class 1E Quality Assurance Yes 10CFR50 Appendix B applicability Display type and method Vertical scale voltage processed in addition to a recorder Unique identification Yes Periodic Testing Yes

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

4 35 7581A

~

TABLE 4.2 a

TRANSIENTS INVESTIGATED CASE PLANT DESCRIPTION A

3 loop 3 inch ccid leg break - FSAR assumptions *; WFLASH 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 - no UHI type pumped safety injection - pumps not running:

3411 MWt WFLASH D

4 loop 1 inch cold leg break - no high head safety s

Non-UHI injection; NOTRUMP 3411 MWt TRCPs tripped at reactor trip, minimum pumped safety injection is available, minimum auxiliary feedwater is available.

4-36 7581A

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5:0 GUIDELINES 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, Wettinghouse and the Westinghouse Owners Group have developed a Refer-ence Emergency Operating Instruction to address recovery from ICC condi-tions caused by a small-break LOCA without high head safety injection.

This instruction has been transmitters to the NRC via Westinghouse Owners Group Letter 0G-44, dated November 10, 1980. It should be noted that this instruction was developed on a generic basis as a technical reference 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 actions are described in References 1 and 2.

~

5-1 7581A s

l

6.0 REFERENCES

1.

Thompson, C. M., et al., " Inadequate Core Cooling Studies of Scenarios witn 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," WCAP-9762 (Proprietary) and WCAP-9763 (Non-Proprietary), June 1980.

3.

" Report on Small Break Accidents for Westinghouse Nuclear Steam Supply System," 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 Program for Simulation of Transients in a Multi-Loop PWR," WCAP-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-prietary) and WCAP-8971 (Non-Proprietary), April 1977.

6.

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

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