ML20008E734

Inadequate Core Cooling Detection Sys Summary Status Rept.

ML20008E734
Person / Time
Site:	Fort Calhoun
Issue date:	12/31/1980
From:	OMAHA PUBLIC POWER DISTRICT
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ML20008E733	List: ML20008E732 ML20008E734
References
NUDOCS 8103090541
Download: ML20008E734 (38)

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l 9

INADEQUATE C0P.E C00LIrlG DETECTION SYSTEM

SUMMARY

STATUS REPORT December, 1980 50 3 00 0 {Ljj

I TABLE OF CONTENTS Section Ti tl e Page

1.0 INTRODUCTION

1-1

1. l' Summary of Activities 1-1 1.2 Definition of ICC 1-1 1.3 Description of Event Progression 1-3 1.4 Summary of' Sensor Evaluations 1-3 2.0 SYSTEM FUNCTIONAL DESCRIPTION 2-1 2.1 Subcooling and Saturation 2-1 2.2 Ccolant Level Measurement in Reactor Vessel 2-1 2.3 Fuel Cladding Heatup 2-2 3.0 SYSTEM CONCEPTUAL DESIGN DESCRIPTION 3-1 3.1 Sensors Design 3-1 3.2 Sig,nal Processing Equipment Design 3-4 3.3 Display Design 3-7 4.0 SYSTEM VERIFICATION TESTING 4-1 4.1 RTD and Pressurizer Pressure Sensors 4-1 4.2 HJTC Sys tem Sensors and Processing 4-1 4.3 Core Exit Thermocouples 4-4 4.4 Processing and Displays 4-4 5.0 SYSTEM QUALIFICATION 5-1 6.0 OPERATING INSTRUCTIONS

• 6-1 l

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1.0 INTRODUCTION

I 1.1

SUMMARY

OF ACTIVITIES This report responds to the requirements in Section II.F.2 of NUREG-0737 (Ref.1).  ;

The report describes the status of design and development activities being conducted by the C-E Owners Group to define a system of instrumentation to be used to de-tect inadequate core cooling (ICC). The report also provides information specific i to Fort Calhoun Station in order to demonstrate the applicability of the generic activity to Fort Calhoun Station.

Results of initial studies by the C-E Owners Group are documented in reports CEN-ll7 [

(Ref. 2) and CEN-125 '(Ref. 3) . All studies have been based on the requirement to indicate the approach to, the existence of and the recovery from ICC.

A three step process is being used to define the ICC Detection System. First, ,

t a definition for the state of ICC has been selected. Second, typical accident '

events which progress toward the defined state of ICC have been analyzed. Thi rd , f instruments which indicate the progression of these events have been selected l and evaluated.

. Based on initial evaluations of a variety'of instruments, an ICC Detection System f I has been defined. This system is judged to be technically sufficient for ICC f detection. However, this system is not uniquely necessary, and functions of its  !

various components may be performed by alternative components. This system is I described in Section 3. Further developments are necessary before the system j can be implemented and these are planned as described throughout this report.

i l 1.2 DEFINITION OF ICC The definition of ICC and the functional requirements for the ICC Detection ,

System have been established within the bounds of the following core conditions:  ;

l -1

- -~ - - - - - .- ,,, . , , , -

l l

i

, 1. The reactor is tripped so only decay power is considered. ' l i i

2. The coolant level falls below the top of the core, which can occur  !

I only with a loss of coolant mass from the Reactor Coolant System (RCS). '

i

3. The event proceeds slowly enough so that the operator has time to observe [

l and to make use of the instrument displays.  !

t These conditions provide the boundaries for a range of sizes of small break {

loss of coolant accident (LOCA) caused by either RCS ruptures or primary -

coolant expansion. '

l i The following definitions of ICC have been considered:

1. First occurrence of saturation.

2. Core uncovery. l

3. Fuel clad temperature of 9000F (below which return to normal operation may  !

be permissible).

4. Fuel clad temperature of 11000F (below which clad rupture in not expected  !

to occur). f i

5. Fuel clad temperature of 22000F (which is the licensing limit for design l basis' events using approved analytical models),

f It has been concluded that the events can progress too rapidly for the instrumen- l tation to reliably display the approach of ICC if one of the first four defini-tions is used. Therefore, it is concluded that definition 5, a fuel clad .

temperature of 22000 F, should be selected as the criterion for existence of ICC. -

v 1

t P

f i

2 1-2

1.3 DESCRIPTION

OF EVENT PROGRESSION i

A typical small break LOCA illustrates the progression of an event which causes [

the approach to ICC. Figure 1-1 shows a representative behavior for the two phase mixture level and the RCS pressure vs. time for the event. The event l progression is divided into four intervals which are shown in Figure 1-1 and i are defined in Table 1-1.

1.4

SUMMARY

OF SENSOR EVALUATIONS Several sensors have been evaluated for use in an ICC Detection System. The instruments considered are listed in Table 1-2, where their capabilities are i summarized. Significant conclusions about each instrument are given below. ,

1.4.1 Subcooled Margin Monitor The Subcooled Margin Monitor (SMM), using input from existing Resistance Temperature Detectors (RTD) in the hot and cold legs and from the pressurizer pressure sensors, is adequate to detect the initial occurrence of saturation during LOCA events and during loss of heat sink events.

The usefulness of the SMM can be significantly increased by also feeding into l it the signals from the fluid temperature measurements from the Reactar Vessel Level Monitoring System (RVLMS) and the signals from selected core exit themo-couples and by modifying the SMM to calculate and display degrees superheat

~

(up to about 18000F) in addition to degrees subcooling. The signals from the RVLMS temperature measurements provide infomation about possible local differen.ces in temperature between the reactor vessel upper head / upper plenum (location of the RVLMS) and the hat or cold legs (location of the RTDs). The core exit themoccuples respond to the coolant temperature at the core exit and their signal indicates superheat after the coolant level drops belcw the top of the core .nd, thus , provide an approximate indication l of the depth of core uncovery.

l  ?

l-3

With. tnese modifications, the SMM can be used for detection. of the approach to ICC, r.amely Interval 1 (loss of subcooling), Interval 3 (core uncovery) l and Interval 4 (core recovery). Even with the modifications, the SMM will $

not be capable of indicating the existence of Interval 2 when the coolant [

. is at saturation conditions and the level is between the top of the vessel and the top of the core.

t The recovery interval may occur at low system pressure and temperature.  ;

Since the errors in the existing SMM calculations increase with lower j temperature and pressure, required subcooling margins need to be revised I for this situation.  !

t

1. . / . 2^ Resistance Temcerature Dete'ctors (RTO) f The RTD are adequate for sensing the initial occurrence of saturation. The  !

hot leg RTD range is sufficient to sense saturation for events initiated i at power. The cold leg RTD, which have a wider' range, are sufficient to [

sense saturation for events initiated from zero power or shutdown conditions.  !

t The RTD range is not adequate for ICC indications during core uncovery. For

• depressuri:ation LOCA events, the core may uncover at low pressure, when the i saturation temperature is below the lower limit of the hot leg RTD. Ini tial f superneat of the steam will therefore not be detected'by the hot leg RTD. As f

the uncovery proceeds, the superheated' steam temperature may quickly exceed l the upper limit of the RTD range. In order to be useful during the core uncovery interval, the range of the RTD needs to be increased to cover a temperature range from 100*F to 1800*F. '

1.4.3 Reactor Vessel Level Monitoring . System The Reactor Vessel Level Monitoring System (RVLMS) is being designed to show the liquid inventory of the mixture of liquid and vapor coolant above the core. It is an instrument which shows the approach to ICC and is the only one which functions in Interval 2, namely the period from the initial occurrence of saturation conditions until the start of core uncovery.

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V l-4 >

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1.4.4 Core Exit Themoccuoles The core exit themoccuples are adequate to show the approach to ICC after core uncovery for the events analyzed provided that the signal processing and display does not add substantial time delay to the themal delay at the themoccuple junction . As mentioned above, the core exit thermocouples respond to the coolant temperature at the core exit and indicate superheat after the core is no longer completely covered by coolant. Except for a time delay of about 200 to 400 sec, depending on event, the trend of the change in superheat corresponds to the trend of core uncovery as well' as to the accompanying trend of the change in cladding temperature.

4 l

1.4.5 Self Powered tieutron Detectors (SPND) '

The SPND yield a signal caused by high temperature as the two-phase ievel falls belca the elevation of the SPND. However, testing is required. to identify the phenomena responsible for the anomalous behavior of the SPND at TMI-2. At the present, their use is limited to low temperature events (less than 1000*F clad temperature) or to only the initial unco <ery portion of an event.

• 1.4.6 Ex-Core Neutron Detectors Existing source range neutron detectors are sensitive enough to respond to the fomation of coolant voids within the vessel during the events analy:ed.

However, the signal magnitude is ambiguous because of the effects of varying baron concentration and deuterium concentration in the reactor coolant.  :

i A stack of ex-core detectors gives less ambiguous information on voids and level in the vessel. The relative shape of the axial distribution of signals from a stack of five detectors shows promise as an ICC indicator, but additional development is needed.

1-5

i l

1.4.7 In-Core Thermoccuoles It appears in general feasible that in-core thermocouples may be added to or substituted for some SPND in the in-core instrument string. They respond more quickly to core uncovery than the' core-exit thermocouples. Also, due to thermal radiation frcm the fuel rods they see temperatures closer to the cladding tem-peratures than to the steam temperature seen by the core exit thermocouples.

For top mounted in-core instrumentation, the core exit thermocouples may survive longer for deep uncovery events because the thermoccuples and their leads see only core exit steam temperature which is less than the fuel clad temperature.

For bottom mounted in-core instrumentation, those in-core thermocouples which are located below the two-phase level will survive longer than the core exit thermocouples because the core exit thermocouple leads past down through the

~

high temperature region ,during core uncovery.

Jsing a synthesis approach, it is expected that the in-core thermocouple temper-l ature can be related directly to the adjacent fuel clad temperature. However, additional work is required to study the temperature response of in-core thermo-couples as well as to develop the mechanical design for incorporating the thermocouples into the in-core instrument string and to develop a synthesis method for calculating fuel cladding temperatures.

1-6

N Table 1 -1_

Definition of Intervals in ICC Event Progression Bounding Parameter Description Interval No. ICC Phase 1 Approach to Reduction in RCS subcooling Depressurization of RCS to satura-until saturation occurs, tion pressure at hot leg temperature or heatup to saturation temperature at safety valve pressure.

2 Approach to Falling twophase mixture level in Het loss of coolant mass from RCS upper plenum, down to top of active accompanied by boiling from continued fuel. depressurization and/oi decay power.

3 Approach to Two phase level falls from top Two phase level drops in core causing 1

and/or of active fuel until minimum clad heatup and producing superheated Existence of level during event progression steam at core exit.

occurs or until 2200 F clad teruperature occurs.

4 Recovery from Two phase level rises above top Coolant addition by ECCS raises level of core, and quenches fuel. ICC progression is defined to terminate when vessel is full or when stable, controllable conditions exist.

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FisuRE l-1 DEFIflITI0f1 0F IilTERVALS Ifl EVEili PROGRESS!0ft

I 2.0 SYSTEM FUNCTIONAL DESCRIPTICN In the following sections a functional description of the instruments of PA ICC Detection Sys. tem is given and the function of the instruments is related to the ICC intervals which were described in Section 1.0.

2.1 SUSC00 LING AND SATLRATION The parameters measured to detect subcooling and saturation are the RCS coolant temperature and the pressurizer pressure. Temperature is measured in the hot legs for typical LOCA type events and is measured in the vessel upper head region for cooldown events. The measurement range extends frcm the shutdown ccoling conditions up to saturation conditions at the pressurizer safety valve setpoint.

The response time needs to be such that the operator obtains cdcouate information during those events which proceed slowly enough for him to observe and to act '

71ay. Plant specific analyses for upon the information from th,e instrument Fort Calhoun Station will be performed for 11ection of small break LOCA events in ,

order to establish the required response times. Generic analyses done to date ,

1 show that existing or planned instruments have adequate range and response.  !

.l The information wk.ich is derived frem the reactor vessel tenperature and pressure measurements is the amount of subcooling during the initial approach to saturation conditions and the occurrence of saturation during Interval 1. During Interval 4, the reestablishment of subcooled conditions is obtained.

2.2 COOLANT LEVEL MEASUREMENT IN REACTOR VESSEL r

The Reactor Coolant System is at saturation conditions until sufficient coolant is lost to lower the two-phase lev 3 to the top of the active core.

During this interval there are no existing instruments which would mea 3ure directly One coolant inventory loss. A Reactor Vessel level Monitoring System provides ;

direct measurement during this period. The parameter which is measured is the collapsed liquid level above the fuel alignment plate. The collacsed level represents the amount of liquid mass which is in the reactor vessel above the core. Measurement of the collapsed water level was selected in preference to 2-1

measuring two-phase level, because it is a direct indication of the water inventory wnile the two phase level is determined by water inventory and void fraction.

The collapsed level is obtained over the same temperature and pressure range as the saturation measurements, thereby encompassing all operating and accident conditions where it must function. Also, it is intended to function during j Interval 4, the recovery interval. Therefore it must survive the high steam temperature which may occur during the preceeding ccre ur.covery interval. l The level range extends from the top of the vessel down to the top of the fuel alignment plate. The response time is short enough to track the level during small break LOCA events. The resolution is sufficient to show the initial level .

drop, the key locations near the hot leg elevation and the lowest levels just above the alignment olate. This provides the operator with adequate indication to track the progression during Intervals 2 and 4 and to detect the consequences of his mitigating actions or the functionability of automatic equipment.

2.3 ~ FUEL CLADDING HEATUP ,

The overall intent of ICC detection is understood to be the detection of the potential for fission product release from the reactor fuel. The parameter which is most directly related to the potential for fission product release is the cladding temperature rather than the uncovery of the core by cool, ant.  ;

l l

Since clad temperature is not directly measured,- a parameter to which cladding "

temperature may be related is measured. This parameter is the fluid temperature at the core exit. ffter the core becomes uncovered; the fluid leaving the core is superheated steam and the amount of superheat is related to the fuel length l exposed and to the claddi~gn temperature.

The amount of superheat of the steam leaving the core will be measured by the core exit thermocouples. The time behavior of the superheat temperature is, with the exception of an acceptably small time delay, similar to the time -

2-2

behavior of the cladding temperature. Thus, from the observation of the steam superneat, the behavior of the cladding temperature can be inferred. Observation of the cladding temperature trer.ds during an accident is considered to be of more value to the operator than information on the absolute value of the cladding temperature.

The core exit steam temperature is measured with the themoccuples included in the In-Core Instrument (ICI) string. They are located inside the ICI support tube, at an elevation a few inches above the fuel alignment plate. Generic calculations of a similar installation for representative uncovery events show that the thermocouples respond sufficiently fast to the increasing steam tem-perature. Plant saecific calculations nn the Fort Calhoun Station confiouration will be made to verify this response.

The required temperature range of the thermoccupies extends frcm the lowest satura-tion temperaturerat which uncovery may occur up to the maximum core average exit temperature which occurs when the peak clad temperature reaches 2200UF. The required themoccuple range is therefore 200U F to about 18000 F, which is the approximate upper service temperature limit. Thermocouples are expected to function with reduced accuracy at even higher temperatures, so the range for processing the thermocouple output extends to about 2300 F.

i l

i 2-3

3.0 SYSTEM CONCEPTUAL DESIGN DESCRIPTION The following sensors have been selected as the basic instruments to meet the functional requ'.rements described in Section 2. I i

1. the Subcooled Margin Monitor (SMM) (Ref.1),

2. the Heatad Junction Thermocouple (HJTC) System (Ref. 2), and

3. the Core Exit Thermocouple (CET) System.

The conceptual design of each ICC instrument is described in this section which addresses:

1. sensors design

2. signa'. processing design

3. display design.

Figure 3-1 is a functionai block diagram for the ICC instrument systems. Each instrument system consists of two safety grade channels # rom sensors through signal processing equipment. The outputs of processina equipment systems feeding the primary display are isolated to separate safety g.ade and non safety grade systems. Channelized safety grade backup displays are included for each instrument system. The following sections present details of the conceptual design.

3.1. SENSORS DESIGN 3.1.1 Subcooled Margin Monitoring System The subcooled margin monitor requires reactor coolant system temperature and pressure inputs to determine saturation margin. Each of the two SMM channels are expected to use the following sensors:

1. two cold leg resistance temperature detectors (RTDs)

2. one hot let RTD

3. one gr.3surizer instrument 3-1 l

l

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4. one HJTC System thermocouple (maximum HJTC System unheated junction temperature frem one channel)  ;

5. one core exit thermocouple (maximum CET temperature of one channel) [

b The RTDs are the wide range element of a dual RTD design. The RTDs are located within wells which protrude through the pipe wall into the coolant flow. l The RTD output is connected to a transmitter located outside containment. I F

The pressurizer pressure sensors are of a force balance design obtaining the pressure input from taps or. the pressurizer in the ' steam volume, The sensor and transmitter functions are performed in one piece of equiument located i inside containment. The p'ressure range is from U psia to '4b00 psia. These  ;

sensors also provide the process inputs for the Reactor Protection System ,

low pressurizer pressure trip.

t 3.1.2 Heated Junction Thermocouple (HJTCl System The HJTC System measures reactor coolant liquid inventory with discrete HJTC sensors located at different levels within a separator tube ranging from the top of the core to the reactor vessel head. The basic principle of system operation is the detection of a temperature difference between adjacent heated i and unheated thermocouples.

As pictured in Figure 3-2, the HJTC sensor consists of a Chromel-Alumel thermocouple near a heater (or heated junction) and another Chromel-Alumel thermocouple positioned away from the heater (or unheated junction). In a fluid with relatively good heat transfer properties, the temperature difference between the adjacent thermocouples is very small. In a fluid with relatively poor heat transfer properties, the temperature difference between the thermocouples is large.

I I

Two design f aatures ensure proper operation under all thermal-hydraulic conditions.

First, each hITC is shielded to avoid overcooling due to direct water contact during two phase 'luid conditions. The HJTC with the splash shield is referred  ;

3-2

to as the HJTC sensor (See Figure 3-2). Second, a string of HJTC sensors is enclosed in a tube that separates the liquid and gas phases that surround it.

Tne separator tube creates a collapsed liquid level that the HJTC sensors measure. This collapsed liquid level is directly related to the average liquid traction of the fluid in the reactor head volume above the fuel alignment pl, ate. This moce of direct in-vessel sensing reduces spurious effects due to pressure, fluid properties, and non-homogencities of the fluid medium. The string of HJTC sensors and the separator tube is referred to as the HJTC instrument.

The HJTC System is composed of two channels of HJTC instru-ments. Each HJTC instrument is manufactured into a probe assembly. The probe assembly includes eight (8)-HJTC sensors, a seal plug, and electrical connectors (Figure 3-3). The eight (8) HJTC sensors are electrically independent and located at eight levels from.the reactor vessel head to the fuel alignment plate.

The probe assembly is housed in a stainless steel structure that protects the sensors from flow loads and serves as the guide path for the sensors. Installation arrangements have been developed for ecch C-E reactor vessel including Fort Calhoun Station. In-stallation details will be provideo in tuture documentation if Omaha Public Power tDi trict decides to install the HJTC system.

3.1.3 Core Exit Thermoccuoles (CET) System The Fort Calhoun Station reactor contains 28 thermocouples that are

~ l top mounted and placed above the fuel assemblies above the fuel alignment plate. Figure 3-4 shows the CET locations. The thermocouples are Type K (Chrcmel-Alumel) and are connected in the same Incore Instrumentation (ICI) cabling as the fixed incore neutron detectors. The thermocouples monitor the temperature l

3 - - -

of the reactor coolant as it exits the fuel assemblies. f The junction of each thermocouple is located above the fuel assembly inside a structure which supports and shields the instrument string frcm flow forces in the outlet plenum j region.

i The basic design of the CETs will not change for the ICC Detection System. However, design ' modifications must be made to meet the qualification requirements (See Sectiori 5.0).

The CETs have a maximum usable temperature range from  ;

700F to approximately 23000F (Reference b).

3.2 SIGNAL PROCESSING EQUIPMENT DESIGN s

The processing equipment of the ICC instruments is presently being developed. The processing equipment portion will be composed ,

of a combination of new and existing equipment. The design objective j for the equi,pment is to ~ address ~the NUREG-0737 criteria, including the  !

criteria of Attachment I to II.F.2 and Appendix A. The following description present functional and general hardware design '

criteria in terms of the three instrument systems described in

• Section 3.1.

l The prowessing hardware will be conficured to orovide '

information to the displays described in Section 3.3. The processing equipment includes operator interfaces for equipment testing, setup, and maintenance. The descriptions are for each of the two separate channels.

All three ICC instrument systems will have similar sensor input processing. The outputs of t' a sensors will be transmitted to the processor, all of which 13 outside of containment, using qualified cable systems. >

t 3-4 l

The processing for the ICC instrumentation will have' surveillance  :

testing and diagnostic capabilities. Automatic on-line surveil-lance tests will continuously check for specified hardware and software malfunctions. The on-line automatic surveillance tests  ;

I as a minimum will indicate inputs that are out of range and [

ccmputer hardware mal #anctions. The malfunctions will be indicated through the operator interface. f 3.2.1 Subcooled Margin Monitoring System i 1

The SMM processing equipment will perform the following functions- I i

f

1. Calculate the subcooled margin.

~

l i

The saturation temperature is calculated from the minimum pressure input and the saturation pressure is  !

calculated from the maximum temperature input (See f Section 3.1). The temperature subcooled margin is the t

[ difference between saturation temperature and the maximum temperature input. The pressure subcooled j j margin is the difference between saturation pressure  !

i and the minimum pressure input. The SMM wil"1 indicate l d superheated conditions. ,

1 l 2. Process all outputs for display.

[

3. Provide an alarm output wi.en subcooled margin reaches a j preselected setpoint.

The SMM will accept the temperature and pressure inputs over the input range of the sensors -- CET from 100 F to 2300 F, the HJTC from 100 F to 1800 F, the RTDs from 0 F to 710 F, i t

i e

b 3-5

- - - - -. - -~ +-m--

'and the pressure from 0 psia to 2500 psia. The saturation i

temperature and pressure are calculated from a saturation curve derived from the 1967 ASME steam tables and an inter-polation routine.

3.2.2 HJTC System The processing equipment for the HJTC performs the following functions:

i

1. Determine if liquid inventory exists at the HJTC positions.

I The heated and unheated thermocouples in the HJTC are connected in such a way that absolute and differential temperature signals are available. This is shown in Figure 3-5. When water surrounds the thermocoUples, their temperature and voltage output are approximately equal. V n Figure 3-5 is, therefore, approximately M-C) zero. In the absence of liquid, the te toccup.e temperatures and output voltages become unequal, causing V(A-C) *U #iS*' Wh""'#(A-C) f the individual HJTC rises above a predetermined setpoint, liquid l inventory does not exist at this HJTC position.

t

2. Determine the maximum upper plenum / head fluid temperature from the unheated thennoccupies for use as an ii put to the SMM. (The temperature processing range is from 100*F to 1800 F.)

3. Process all inputs and calculated outputs for display.

4 Provide an alarm output to the plant annunciator system when any of the HJTC detects the absence of liquid l level.

l 3-6 l _ _ _ - - _ _ __ _ _ _ __

5. Provide control of heater power for proper.HJTC output signal level. Fiqure 3-6 shows a single channel conceptual design which includes the heater power controller.

3.2.3 Core Exit Thermocouple System The processing equipment for the CET will perform the following functions:

1. Process all core exit thermocouple inputs for display.

Half of the available CET inputs will be processed in each channel.

2. Provide an alarm output to the plant annunciator system when the temperature from any of the CET's exceeds a preselected setpoint.

3. Deter [ninethemaximumCETtemperaturetobesuppliedto the St@t. The processed temperature range will be from 100 F to 2300 F.

These functions are intended to meet the design requirements of NUREG-0737, II.r'.2 Attachment 1. The display section will describe the display design for the CET system.

3.3 . DISPLAY DESIGN The ICC instrurer$t outputs will be displayed through a human engineered catho.de ray tube (CRT) based primary disclav and separate backup displays. The ' Critical Function Monitor (CFM) System (Ref. 7) is being considered as the primary display for the ICC instrument outputs. As shown in Figure 3-1, each channel of the ICC instrument system will also have safety grade backup displays.

Both primary and backup displays are intended to be designed consistent with the criteria in NUREG-0737 Action Item II.F.2, II.F.2 Attachment 1, and Appendix A. The following description presents the conceptual design for display.

l 3-7

The CFM System is a dedicated, computer based display system that monitors critical plant functions: ,

1. Core reactivity control

2. Core heat removal control

3. RCS inventory control

4. RCS pressure control

5. RCS heat removal control

6. Containment pressure / temperature control

7. Containment isolation If any of the critical functions are violated, (by exceeding logic setpoints) a Critical Function Alarm (CFA) is initiated.

The ICC instruments outputs will be incorporated in this critical function alarm logic. ,

The CFM displays data on four cathode ray tubes. The data has three levels of information: "

Critical functions status (very general)

. Level 1 Level 2 System overview (general, on system)

Level 3 System detail (specific information)

This hierarchy allows the operator to progress from an overall system view to a detailed diagnostic view. The ICC instrument outputs will be incorporated in all three levels of display. The

, detailed ICC information is anticipated to be displayed on the Level 3 display. Trending displays are also available with the CFM.

1 I

3-8  ;

1

~

l I

l l

t Each channel of backup display will present the most reliable i basic information for each of the ICC instrument systems. These displays will be human engineered to give the operator clear unambiguous indications. The backup displays are designed:

1. to give primary instrument indications in the remote chance  !

that the primary display beccmes inoperable. ,

to provice confirmatory indications to the primary display.

2.

3. to aid in surveillance tests and diagnostics.

The following sections present details on the display for each of the instrument systems as presently conceived.

3.3.1 Subcooled Margin Monitor Display The following information is anticipated to be presented on the primary display:

1. Pressure margin to saturation. .

4

2. Temperature margin to saturation.

3. Maximum temperature and source (i.e., HJTC, RTD, or  ;

CET)

4. Minimum pressure ,

i The following information is anticipated to be presented on the backup displays:

t

1. Pressure margin to saturation

! 2. Temperature margin to saturation

3. Temperature inputs

! 4. Pressure inputs c

I t.

1 I

3-9

3.3.2 Heated Junction Thermocouple System Display -

The following information is anticipated to be displayed on the primary display:

1. Two channels of 8 discrete HJTC positions indicating liquid inventory above the fuel alignment plate.

2. Maximum unheated junction temperature of each of the twc channels which is provided to the SMM.

The folicwing information is anticipated to be displayed on the backup displays:

1. Liquid inventory level above the fuel alignment plate derived from the 8 discrete HJTC por.itions

2. Unheated junction temperature at the 8 positions

3. Heated junction temperature at the 8 positions 3.~3.3 Core Exit Thermocouple System Display The following information is anticipated to be displ6,eed on the primary display:

1. A spatially oriented core map indicating the'tempe ature at each of the CET locations.

2. A selective reading of CET temperature i

At least the maximum CET temperature of each of the two channels which is provided to the SMM will be presented.

The backup displays are anticipated to display at least four CET from each quadrant with an identification l

, number for each CET temperature. At least 16 CET I

temperature will be displayed within 6 minutes. '

3-10 l

CilAtitlEL A

^ "P SEilSORS PROCESSIllG DISPLAY 0

PRIMARY ISOLATI0tl DISPLAY Y

O

^CKUP SErlSORS PROCESSIllG DISP [Ay L

CilAtitlEL B FIGURE 3-1 ICC IllSTRUMEtlTS FUtiCTI0tlAL BLOCK DIAGRNi

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Si A

L P

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A D E L H

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SEPARATOR TUBE ELECT'RICAL C0fillECTORS.

(ONE PER SEfiSOR)

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L J Y

SEAL PLUG li,lTC SENSOR 8

FIGilRE.3-3 l ilEATED JUNCTI0ft TilERl10 COUPLE PROBI. ASSEMBLY 9

. CALLED NORTH -

t I;'?

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FIGURE 3-4 FORT CALHOUN STATION CORE EXIT THERMOC0UPLE ARRANGENENT A B C D EFGHJKLMN P R .S T I . I 2- - -

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3 4 - O O @ O -

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20 19 id 12 - - 12 13 - O22 O21 -!O i4 _

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CHROMEL ALUMEL A

• t A MEL ,

. B ,

~

CHROMEL - . .

f COPPER _

(.) --

. j i

i V (A . B) - ACTUAL TEMPERATURE, UNHEATE.D JUNCTION ,

V (C .8) = ACTUAL TEMPERATURE, HEATED JUNCTION V (A . C) - DIFFERENTIAL TEMPERATURE

. t r

. l FIGURE 3-5 ,

E LECTRIC A L DIAGRAM OF H . J . T .. C .

i 3-15

l l

I

. r I

r I

i SENSOR 1 c. =  !

THROUGH i SIGNAL PROCESSOR + PRIMARY DISPLAY LOGIC AND CHANNEL BACKUP SENSOR 8 c' CONTROLS DISPLAY

=

• ALARM k

PO'.'lE R FEEDBACK SIGNAL

' POWER

, , POWER CONTROL -

SIGNAL l V V

. HEATER POWER CONTROLLER V

POWER TO HEATERS FIGURE 3-6 HJTC SYSTEM PROCESSING CONFIGURATION (ONE CHANNEL SHOWN) l l

3-16 l .

l ..

4.0 SYSTEM VERIFICATION TESTI'IG ,

This section describes tests and operational experience with ICC system instruments.

i 4.1 RTD AND PRESSURIZER PRESSURE SENSORS

! The hot and cold leg RTD temperature sensors and the pressurizer 2

pressure sensors are standard NSSS instruments which have well l known responses. No special verification tests have been per-formed nor are planned for the future. These sensors provide j basic, reliable temperature and pressure inputs which are con-4 sidered adequate for use in the SMM and other additional display

. functions.

(

l 4.2 HJTC SYSTEM SENSORS AND PROCESSING 1

'The HJTC System is a new system developed to indicate liquid in'ventary above the core. Since it is a new system, extensive

testing has been performed and further tests are planned to assure that the HJTC System will operate to unambiguously indicate liquid inventory above the core. '

The testing is divided into three phases:

Phase 1 - Proof of Principle Testing Phase 2 - Design Development Testing Phase 3 - Prototype Testing j The first phase consisted of a series of five tests, which have been completed. The testing demonstrated the capability of the 4-1

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

I

.. t HJTC instrument design to measure liquid level in simulated reactor vessel thermal-hydraulic conditions (including accident condi tions) . )

l Test 1 Autoclave test to show HJTC (thermoccuples only) response to water or steam, t

a i In April 1980, a conceptual test was performed with two thermocouples j

in one sheath with one thermocouple as a heater and the other  ;

I thermocouple as the inventory sensor. This configuration was  !

t placed in an autoclave (pressure vessel with the capabilities to j

> adjust temperature and pressure). The thermocouples were exposed l 1 ,

to water and then steam environments. The results demonstrated a p 4

l significant output difference between steam and water conditions i for a given heater power level.

Test 2 Two phase flow test to show bare HJTC sensitivity to ,

voids.  ;

r In June 1980, a HJTC (of the present differential thermocouple  !

design) was placed into the Advanced Instrumentation for Reflood l Studies (AIRS) test facility, a low pressure two phase flow test l facility at Oak Ridge National Laboratory (ORNL). The HJTC was exposed to void fractions at various heater power levels. The  :

results demonstrated that the bare HJTC output was virtually the same in two phase liquid as in subcooled liquid. The HJTC did generate a significant output in 100% quality steam.

i Test 3 Atmospheric air-water test to show the effect of a splash sh'. eld A spiash shield was designed to increase the sensitivity to voids. The splash shield prevents direct contact with the liquid in the two phase fluid. The HJTC output changed at intermediate i

i 4 - 2' i

void fraction two phase fluid. The results demonstrated that the HJTC sensor (heated junction thermocouple with the splash shield) i l I sensed intermediate void fraction fluid _ conditions.

i Test 4 High pressure boil-off test to show HJTC sensor response to reactor thermal-hydraulic conditio'ns.

In September 1980, a C-E HJTC sensor (HJTC with splash shield) was installed and tested at the ORNL Therm'l-Hydraulics a Test Facility (THTF). The device is still installed and available for further tests at ORNL. The HJTC senscr was subjected to various two phase fluid conditions at reactor temperatures and pressures.

The results verified that the HJTC sensor is a device that can sense liquid inventory under norm'al 'and accident reactor vessel high pressure and temperature two phase conditions.

Test 5 Atmosphe,ric air-water test to show the effect of a separator tube ,

A separator tube was added to the HJTC design to form a collapsed liquid level so that the HJTC sensor directly measures liquid inventory under all simulated two phase conditions. In October, 1980, atmospheric air-water tests were performed with HJTC sensor and the separator tube. The results demonstrated th:t the separator tube did form a collapsed liquid level.and the HJTC ou.tput did accurately indicate liquid inventory. This test verified that the HJTC instrument, which includes the HJTC, the splash shield, and the separator tube, is a viable measuring device for 1iquid inventory.

The Phase 2 test program will consist of high pressure and tempera-  ;

ture tests on the HJTC instrument. These tests will provide input for the C-E HJTC instrument design and manufacturing ,

effort. The Phase 2 test program is expected to be completed in early 1981. ,

4-3

. .=. -. - . - ... .---

l The Phase 3 test program will consist of high temperature and -

pressure testing of the manufactured prototype system HJTC probe assembly and processing electronics. Verification of the HJTC system prototype will be the goal of this test program. The Phase 3 test program is expected to be completed by the end of 1981.

4.3 CORE EXIT THERMOCOUPLES 4 No verification testing of the CETs is planned. A study at ORNL was performed to test the response of CETs under simulated acci-

, dent conditions (Reference 6). This te'st showed that the instru-ments remained functional up to 2300'F. This test along with previous resctor operating experience verify the response df CETs.

l the response of CETs.

i 4.4 PROCESSING AND DISPLAYS The final processing and display design for the ICC detection system has not been completed. As the design effort proceeds, design evaluations will be performed prior to installation.

Correct implementation of the software and hardware will be included and documented as part of the design effort.

e l

1 4-4 1

- .- - - . , - , - , a.,, . - - - -

5.0 SYSTEM OUALXFICATIOPI [

The qualification program for the ICC Detection System instrumentation ,

has not been completely defined. However, plans are being developed based on the following three categories of ICC instrumentation: ,

1. Sensor instrumentation within the pressure vessel. ,

i

2. Instrumentation ccmponents and systems which extend from the  ;

primary pressure boundary up to and including the primary display  ;

isolator and including the backup displays. j f

3. Instrumentation systems which ccmprise the primary display l i equipment. i A preliminary outline of a qualification program for each classification 7

is given below. .

The in-vessel sensors will meet the NUREG-0737, Appendix A guide to install the best equipment available consistent with qualification and schedular requirements. Design of the equipment j will be consistent with the guidelines of Appendix A as well as the clarification and Attachmeat 1 to Item II.F.2 in NUREG-0737. Specifically, l instrumentation will be designed such that they meet appropriate stress criteria when subjected to nonnal and design basis accident loadings.

Verification testing will be conducted to confirm operation at DBA (as defined by C-E) pressure and . temperature conditions., Seismic testing to safe shutdown conditions will verify function .after being subjected to the seismic loadings.

The out-of-vessel instrumentation system, up to and including the primary display isolator, and the backup displays will be environmentally I qualified in accordance with IEEE-323-1974 l

l i

4 5 -1

a Plant-specific containment temperature and pressure design profiles will be utilized where appropriate in these tests. This equipment will also be seismically qualified.

The primary display will not be designed as a Class lE system, but will be designed for high reliability; thus it will not be qualified environmentally or seismically to Class lE requirements nor will it meet the single failure criteria of Appendix A, Item 2. Post-accident i

maintenance accessibility will be included in the design. The quality assurance provisions of Appendix A, Item 5 do not apply to the primary display according to NUREG-0737. However, the ccmputer driven primary display system will be separated from the Class 1E senrors, processing

]

and backup display equipment by means of an isolation device which will be qualified'to Class lE criteria.

I l

r i

5-2

I i

, 6.0 OPERATING INSTRUCTIONS ,

Guidelines for reactor operators to use to detect ICC and take . I corrective action has been developed by the C-E Owners Group and ,

submitted to NRC for review (Ref. 8). These guidelines have been used to review and revise the plant emergency procedures for Fort Calhoun Station. In addition, the C-E Owners Group has {

developed reactor operator training materials concerning ICC.

The Fort Calhoun Station training staff attended a training seminar conducted by C-E in November,1979, to initiate the Fort Calhoun Station ICC training program.

The C-E Owners Group is defining a program for development of further emergency procedure guidelines and operator training materials associated with the ICC Detection System described in Section 3.

This program is expected to provide these guidelines and training materials during 1981. A more specific schedule is subject to finalization of the ICC Detection System design, specifically the instrument displays.

i t

I 6 , _ _

I i-REFERENCES n

i

1. NUREG-0737, " Clarification of TMI Action Plan Requirements,"

] U. S. Nuclear Regulatory Ccmmission, November,1980.

. 2. CEN-117 " Inadequate Core Cooling - A Response to NRC I E Bulletin 79-06C, Item 5 for Combustion Engineering Nuclear Steam Supply Systems," Combustion Engineering, October,1979.

i ,

3. CEN-125, " Input for Response to NRC Lessons Learned Requirements for Combustion Engineering Nuclear Steam Supply Systems," Combustion

. Engineering, December,1979.

4. C-E Proposal No.1579 SP, "C-E PWR Subcooled Margin Monitor," September, 1979.

1

! 5. C-E Proposal No. 2580 SP, " Heated Junction Thermocouple System,"

l Septembe r, 1980. ,

) 6. Anderson, R. L. , Sanda , L. A. , Cain, D. G. , "Incor e Thermocouple i Performance Under Simulated Accident Conditions", presented at IEEE Symposium, November, 1980.

l l

7. C-E Paper TIS-6649, " Operational Aids to Improve the Man-Machine Interaction in a Nuclear Power Plant," Presented at American Nuclear Society Annual Meeting, Las Vegas, Nevada, June 8-12, 1980.

8. Letter C-E Owners Group to NRC, "C-E Generic Emergency Procedure Guidelines," December 10, 1980.

i a

}

t'

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