ML20087P801

SPDS Parameter Selection/Safety Analysis Study

ML20087P801
Person / Time
Site:	Oyster Creek
Issue date:	04/02/1984
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20087P797	List: ML20087P801
References
RTR-NUREG-0737, RTR-NUREG-737 TR-016, TR-16, NUDOCS 8404090380
Download: ML20087P801 (38)

Text

TR # 016 e

OYSTER CREEK SPDS Parameter Selection / Safety Analysis Study i.

8404090380 840402 PDR ADOCK G5000219 F

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oc, TR # 016 Page 1 TABLE 0F CONTENTS PAGE 1.0 Introduction 5

2.0 Literature Review 7

3.0 -SPDS Parameter Identification 11 3.1 Critical Safety Functions 11 3.2 Parameter Selection 12 3.3 CSFS and E0PS-21 3.4-Comparison with Other Parameter Sets 25 4.0 SPOS Response to Transients / Accidents 28 4.1 Large/ Intermediate Break Below the Core 28 Inside the Containment

-4.2 SPDS Detection 31 15.0- References 37 6

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LIST OF FIGURES PAGE FIGURE 4-1 Event -Sequence Diagram Symbols 35 FIGURE--4-2 LOCA Event Sequence Diagram L1 36 b

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TR # 016 Page 3 LIST OF TABLES PAGE TABLE 2.1 Recommended Fundamental Safety Parameter 9

Set for Boiling Water Reactor (S. Levy)

TABLE 2.2 Hypothetical BWR CSF Status Parameters 10 (NUTAC)

TABLE 3.1 SPDS CSFS'- Parameters Matrix 23 TABLE 3.2 Area Radiation Monitors 24 TABLE 3.3 Oyster Creek Safety Parameter Set 27 TABLE 4.1 Transients / Accidents Scenarios-33 TABLE 4.2

-SPDS Response to L1 34 4

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TR # 016 Page 4

SUMMARY

~A Critical Safety Functions (CSF) approach was used as the basis for a Safety Parameter Display System and five CSFs were chosen that correlate with the basic objectives of Emergency Operating Procedures. A parameter set was then selected largely based on E0Ps parameters and the Emergency Plan radiation monitoring parameters (ARMS). Groups of parameters were then assigned to the different CSFs based on plant operating and emergency procedures logic. These safety functions were then activated to warning or alarm modes using different setpoints that are used by the different procedures. The.CSFs were then tested on a broad range of transient scenarios to check their response. The final parameter list (Table 3.1) responded to every manual or automatic action that was required to be taken during these scenarios.

It is therefore believed that the CSFs and the parameter set chosen form a complete set that should respond to plant conditions during power operation and shutdown modes.

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This represents the system's conceptual design which will be updated to reflect evolution of actual design when instruments to be used have been determined.

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Page 5 i

1.0 INTRODUCTION

NUREG-0737 Supplement 1, Section 4.0 requires the incorporation in the control room of a Safety Parameter Display System to aid the user in

. assessing plant safety status. A spectrum of opinion exists as to what plant safety status means. The BWROG position is to integrate SPDS with Emergency Operating Procedures (EOPS) resulting in a com-puterized E0PS system where E0PS steps are displayed for the user to follow, with certain parameter trends, and systems status shown as an aid for proper E0PS implementationIII. This approach would limit the usefulness of SPDS to emergency situations and would not add any new information which is not immediately available on the control panels or in the procedures and easily accessible. Another opinion that was discussed was to use SPDS in addition to E0PS, as a display terminal where important plant conditions are shown or trended.

In this manner f

it will only be used by control room supervisors as an overview function. This approach would lead to a computerized control room display with the user still having to resolve the interdependence of plant parameters and make judgment as to the safety status of the plant. Another opinion was to limit SPDS to supply information which is not. easily accessible to the user, like level rate of change, torus limit curves etc. After careful review, we have adopted the concept of Critical Safety Functions (CSFs) where a CSF is a measure of the safety status of a group of plant parameters which together convey a coherent meaning with regard to plant safety. These CSFs are also supported by a set of displays that would convey a concise picture of plant

I-TR # 016 Page 6 conditions where the affected safety function is concerned. The CSF approach is followed here because it focuses attention on that aspect of plant safety where a response is most urgently needed.

The following requirements were followed in this study:

1.

A minimum set of parameters must be chosen that would uniquely describe a certain CSF under all plant conditions (during power, emergency and shutdown modes).

2.

The SPDS will assist in E0Ps implementation by identifying E0Ps entrance criteria and displaying parameters, trends and limits which are included in the plant emergency procedures.

3.

Displays to be generated would have to concisely reflect plant status showing without ambiguity the interactions between the different parameters that form each CSF.

4.

. Users' guidelines which would supplement existing procedures must be generated.

The basis SPDS is an overview tool to be used by the GSS, GOS, or STA.

It should be noted that control room operating personnel may have additional displays tailored to their needs to assist them in following the symptom based E0Ps. These additional' displays will be supportive of the critical safety functions. The SPDS is to be provided as an aid for use at the discretion of the GSS, G0S or STA.

In no cases is its usa mandatory for safe operation.

TR # 016 n

Page 7 2.0 LITERATURE REVIEW The available literature on this subject is related to the time period before the release of NUREG-0737 Supplement 1 on December 17, 1982.

The purpose of this review is to highlight the basic methodologies used and how they compare with the approach used here.

S. Levy Inc. carried out a parameter selection study for a generic BWR(2)

Their recommended critical safety functions and the parameter set are shown in Table 2.1.

The basic arguments used in this selection were concerned with:

Parameter duplication; Loss of the parameter during an accident, e.g. steam flow is lost upon vessel isolation; Degree of interaction between the different parameters, i.e.

whether a safety function status can be inferred from a para-meter used for another safety function.

NUTAC(3) presented a different set of CSFs and parameters for a hypothetical BWR (Table 2.2). Here the reactiv'ity CSF has been included as part of core cooling and heat production CSF. These two references were the only ones found in the literature that present a set of CSFs and their respective parameters for a BWR.

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TR # 016 Page 8 NUREG-0737 Supplement 1, Section 4.0 requires information to be pro-vided to the user on:

Reactivity Control Reactor Core Cooling and Heat Removal from the Primary System-Reactor. Coolant System Integrity Radioactivity Control Containment Conditions The above NRC functions were use'd as guidelines in generating Oyster Creek CSFs.

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Page 9 TABLE 2-1 4

RECOMMENDED FUNDAMENTAL SAFETY PARAMETER SET FOR A BOILING WATER REACTOR (S. Levy)

Safety Function Monitored P arameter Containment of Radioactivity Plant Ventilation Monitor

-(Radioactivity Release Control)

Main Stack Monitor Primary Coolant System Activity Primary Containment Activity Fission Product Ba' rier Integrity Plant Ventilation Monitor

• r Primary Coolant System Activity

• Drywell Floor Drain Sump Drywell Pressure Primary Coolant System Pressure Primary Coolant System Water Level Safety / Relief Valve Positions Leakage Isolation Demand Suppression Pool Level Hydrogen Concentration Secondary Containment Pressure Primary Containment Activity

• Heat Transport Primary Coolant System Pressure

• Primary Coolant System Water Level

• Suppression pool Level

• Drywell Temperature Average Power Range Monitor Core Flow Suppression Pool Temperature Reactivity Control Average Power Range Monitor

• Source Range Monitor Source Range Monitor Position Scram Demand Signal Mode Switch Position Repeated Parameter s

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TABLE 2.2 HYP0THETICAL BWR CSF STATUS PARAMETERS ' NUTAC)

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Critical Safety Function Associated SPDS Parameters Containment.of Radioactivity Plant Ventilation Monitors

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Main Stack Monitor Primary Coolant System Integrity Drywell Floor Drain Sump Drywell Pressure Primary Coolant System Pressure Primary Coolant System Water Level

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Safety Relief Valve Positions Containment Integrity Drywell Pressure Primary Coolant System Pressure Suppression Pool Level Secondary Containment Pressure Drywell Temperature Suppression Pool Temperature -

Core Cooling and Heat Production Primary Coolant System P~ressure Primary Coolant System Water Level Core Spray Flow Core Flow Average Power Range Monitor Intermediate Range Monitor Source Range Monitor s

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TR # 016 Page 11 3.0 SPDS PARAMETER IDENTIFICATION The first step in the identification process was to define a set of Critical Safety Functions-(CSFS) that will convey to the user the plant safety status and at the same time satisfy NUREG-0737 Supplement 1 Section 4.0, main safety categories. The second step was to select plant parameters that are consistent with E0Ps and that will reflect the status of'each critical safety function. The third step will be to generate a users' guideline that would aid in evaluating and dealing with abnormal conditions. These guidelines would then serve as the basis for developing displays and for training users in the use of SPDS. The first two steps are discussed herein.

3.1 Critical Safety Functions The CSFS chosen for Oyster Creek and their sub-divisions are:

1.

Reactivity./ Power Distribution 2.

Heat Removal Fuel clad cooling Primary system cooling

'3.

Reactor Coolant System Integrity Fuel integrity Primary system-integrity 4.

Radioactivity Control' 5.

Containment Conditions r

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.3.2 Parameter Selection

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Each CSF will.be discussed separately. Emergency Operating Procedures parameters will be selected that best reflect the status of each'CSF.

This approach would make sure,that E0PS parameters are tied into the

-SPDS. The.overall SPDS logic will switch from 'AT POWER' to ' SHUT 00WN'

' logic.upon a scram demand.

3.2.1 Reactivity / Power Distribution When the plant is.at steady' state power operation the Reactivity Control CSF will monitor for power peaking and core thermal limits.

Since power peaking and thermal limits are not directly measurable, they will be monitored by using other parameters that will then bound the puwer peaking and thermal limits concerns.

~During all modes, the core startup rate will be monitored to determine gross reactivity concerns.

Reactor power will also be monitored during power operation and post trip to determine if an ATWS or power excursion has

- occurred..

-The requiredLparameters are:

• Neutron Flux.

• - Scram Demand

• -: Recirculation Flow J

.9 TR 0 016 Page 13 3.2.2 heat Removal The basic parameters of heat transfer are mass flow, power, pressure and temperature. Once heat balance is violated, this CSF will provide warning and alarm.

3.2.2.1 Fuel Clad Cooling Fuel clad temperature is the desired parameter but since it.is not available, RPV level may be used as an alternative parameter under certain conditions. Other candidate parameters are core mass flow (recirculation flow)'and power (Neutron Flux).

During normal operation core flow and power are the required parameters while during shutdown vessel level (downcomer and fuel zone levels) and core spray discharge pressure are the required parameters.

Total Parameters are:

RPV level Recirculation Flow

• - Neutron Flux Scram Demand Core Spray Discharge Pressure Core spray flow may be used instead of discharge pressure. This will be decided during design evolution when instruments to be used are determined.

TR # 016 Page 14

'3.2.2.2 Primary System Cooling

-This index will measure the adequacy of the primaty system as a heat removal medium. At shutdown, if fuel clad cooling is assured, then RPV level and pressure are the main ~ parameters required.

If RPV pressure is above relief valves or emergency condenser setpoints (on RPV isolation), or if RPV level is at low-low (Emergency Condenser setpoint), then adequate heat removal may not be provided. While at power, steam flow and feed flow are the required parameters for proper heat removal.

RPV isolation which might be caused by a number of signals is represented by an isolation demand

. parameter. _This parameter together with steam flow are required to monitor the status of the main condenser as a heat sink, i.e. if isolation demand is present while steam flow is still available then user action is required to prevent using the condenser as a heat sink.

Steam flow at power may be used as a backup parameter for heat balance representation.

Required Parameters are:

RPV Level RPV Pressure Steam Flow Feed Flow Isolation Demand Neutron Flux

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TR # 016 Page 15 3.2.3 ' Reactor Coolant System Integrity 3.2.3.1 Fuel Integrity If fuel clad integrity is violated during normal operation, the steam line radiation monitor will detect it. Such monitoring may induce the user to take action before further breach takes place since this monitor is quite sensitive to fuel failure.

If steamline radiation monitors fail to detect it, the stack offgas monitors will provide detection.

If fuel failure happens after vessel _ isolation, the radioactive fission products (and hydrogen) will find

- their way to the drywell where they will be detected by drywell area radiation monitors (ARMS).

Therefore, required parameters are:

Steam line radiation monitor Drywell act1vity monitor Offgas monitor (stack)

Combustible gases 6

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TR # 016 Page 16 3.2.3.2 Primary System Integrity This will be divided into that portion of the system within the primary containment and the portion outside.

Within primary containment: For very small breaks-inside the drywell, the uncontrolled sump pump out rate, steam flow and feed flow will be required. For larger breaks, drywell pressure, torus pressure, RPV level and pressure are the required parameters.

The addition of a

'drywell_ level indicator is being evaluated in conjunction with RG 1.97 requirements.

If in the future a drywell level indicator is made available, it will be added.and used as an indicating parameter.

Outside the primary containment:

Breaks outside the containment will be detected by isolation demand and ventilation radiation monitors where leakage of radioactivity will be detected.

Therefore, required parameters are:

Scram demand Drywell pressure RPV level Steam flow Torus pressure

TR # 016

.Page 17 Uncontrolled sump pump out rate

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i Isolation demand Turbine / Reactor buildings ventilation monitors Feed Flow Steam flow is required to detect main steam line breaks

-(causing vessel isolation) while feed flow is required since it is the first inventory make up source used.

These parameters may also be useful in detecting small breaks, i.e. relief valve leaks to the torus.

Isolation demand is needed to set up the algorithms that would differentiate and prioritize monitoring the different CSFs, i.e. if isolation demand and high drywell pressure are present then' containment isolation has taken place

-and the containment integrity CSF has to be monitored on a priority basis, in conjunction with other challenged CSFs.

3.2.4

. Radioactivity Control In general, radioactivity would come about~in two forms:

Gaseous activity

. Area activity (contamination)

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TR # 016 Page 18 Gaseous activity will show up through the ventilation systems, therefore, the required parameters are:

Stack vent monitor Turbine / Reactor buildings ventilation systems radiation monitors e

Area contaminatiin is measured by the area radiation monitors. The relevant monitors have been chosen and are provided in Table 3.2.

3.2.5 Containment Conditions The conditions of concern are containment cooling and containment integrity.

3.2.5.1

~ Containment-The heat paths into and out of the containment are:

Torus 1.

EMRVs discharge 2.

Heat storage in pool inventory 3.

Heat removal through containment spray in test mode 4.

Drywell/ Torus vents Drywell 1.

SVs discharge 2.

Normal heating from vessel structure 3.

Heat removal through containment spray 4.

Heat removal through drywell coolers 5.

Leaks / rupture

TR # 016 Page 19 Torus conditions above can be monitored through torus water level and temperature. RPV pressure is also required to guard against the onset of unstable steam condensation in the pool. Torus temperature and level limit curves are required to initiate alarms and warnings once-these limits are reached. Drywell conditions above are monitored through drywell pressure and temperature.

Required parameters are:

o Torus Level o

RPV Pressure o

Drywell Pressure o

Drywell Temperature (bulk and axial distribution) o Torus Temperature 3.2.5.2 Primary Containment Integrity Drywell: Monitoring drywell pressure, both positive and negative and isolation demand will detect an approach toward design limits. Combustible gases concentration is also required since hydrogen oxygen combustion may result in drywell integrity violation. Drywell pressure setpoint should be a function of drywell design limits.

Torus: Torus water level is the required parameter for leak detection below the operating level. However, l-torus water temperature is also required since a leak during torus heat up may be masked by water expansion.

TR 0 016 Page 20 In such cases, area radiation monitors in the reactor building will.be required for such detection since water activity during EMRVs discharge will increase. Torus air space pressure is required to monitor the status of the vacuum breakers and hence, the integrity of the ring header / vent pipes. For cases where drywell pressure equals torus pressure, then torus design pressure becomes limiting.

The required parameters are:

Drywell pressure (function of structural limits)

Isolation Demand Combustible gases Torus water level Torus temperature Torus air space pressure Radiation monitors (Table.3.2)

Table 3.1 shows the Critical Safety Functions -

Parameters Matrix. -The setpoints required to initiate various CSFs and the logic to be used will be investigated later. User actions as a result of CSF initiation will be discussed when CSFs initiation logic isLfully investigated.

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.Page 21

.3.3 CSFS and E0PS

~The CSFS should be. compatible with E0PS objectives. E0PS objectives are to control power, level, pressure and containment conditions.

PowerLis correlated.with the Reactivity CSF while level is correlated with the Heat Removal CSF, which is.further divided into Fuel Clad Cooling and Primary System Cooling. The two subdivisions were used to

. highlight the.importance-of fuel clad cooling requirements and primary system cooling requirements. Maintaining the latter will avoid getting into the -former. Containment Conditions is a separate CSF which includes Containment Cooling and Containment Integrity.

A Separate Radioactivity Control Safety Function was chosen to supplement E0PS. Control of radioactivity is not dealt with explicitly.in the E0PS, but CSFS guidelines for radioactivity will be

. ritten to be consistent with E0PS. E0PS figures and limits will also w

be considered in development of SPDS displays and setpoints.

It is therefore concluded that the chosen CSFS encompass the basic objectives of the E0PS in maintaining proper plant conditions.

All. chosen parameters are E0P parameters except:

- -Drywell Activity Monitors

= Reactor / Turbine buildings ventilation monitors Combustible gases' monitors Area radiation monitors (ARMS) shown in Table 3.2

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Uncontrolled Drywel1 Sump level monitor Drywell Axial temperature monitors e

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a TR # 016 Page 22 The list of ARMS was chosen in accordance with Emergency Plan

- requirements. A number of EOPS parameters have not been included in

- the list since those parameters are used to check the status of various systems, e.g. valves and pumps. status for systems line ups, etc. None of those parameters are the primary indicators of plant

- safety.

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. Parameter 1.

NEUTRON FLUX X

X X

-2.

SCRAM DEMAND X

X X

-3.

RPV LEVEL X

X X

4.

RPV PRESSURE-X X

5.

RECIRCULATION FLOW X-X 6.

DRYWELL PRESSURE X

X X

7.

-DRYWELL TEMPERATURE (BULK & AXIAL)

X 8.

TORUS WATER LEVEL X

X 9.

TORUS TEMPERATURE X

X 10.

TORUS AIR SPACE

~ PRESSURE X

X l

11.

STEAM LINE RADIATION

~ ONITOR X

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DRYWELL ACTIVITY X

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STACK ACTIVITY X

X 14.

TURBINE / REACTOR BUILDINGS VENTILL MONITORS X

X 15.

DRYWELL SUMP (UNCONTROLLED)

X 16.

COMBUSTIBLE GASES X

X 17.

AREA RADIATION MONITORS X

X 18.

STEAM FLOW X

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FEED FLOW X

X 20.

ISOLATION DEMAND X

X X

21.

CORE SPRAY DISCHARGE PRESSURE X

TABLE 3.1: SPDS CSFS - Parameters Matrix

TR # 016 Page 24 TABLE 3.2

. AREA RADIATION MONITORS l.-

Admin. Bldg. to Turb. Bldg. N. Entrance 2.

Turb. Oper. Fl.

3.

Feed Pump Area (T.B.)

4.

Feed Pump area (T.B.)

5.

Cond. Pump (above pumps, T.B.)

6.

Cond. Demin. Valve (T.B. Bsmt)

7..

Regeneration Area 8.

Make up Demin. (T.B. Bsmt)

9. -

Air Compressors (T.B. Bsmt) 10.

T.I.P. Equip. Area (R.B.)

11.

Personnel Lock 12.

RX. Oper. Floor (Eq. Hatch area)

'13.

RB Eq. Drain tank 14.

Cleanup pump' area (RB) 15.

Iso. Cond. Area 16.

Shutdown Cooling Area 17.-

Spent Fuel Pool (RB. 95'-3")

18.-

Liq. Poison Syst. (95'-3")

19.

CRD modules area (RB, 23'-6")

20.

Air Ejectors (T.B. 3'-6")

21.

Fuel Pool Area (119')

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TR # 016 Page 25 3.4-Comparison with Other Parameter Sets, In order to compare Oyster Creek CSFs paramete s set with those of S. Levy and NUTAC,. Table 3.1 has been re-arranged in terms of the main four CSFs and put in the same format as Tables 2.1, 2.2 (Table 3.3).

1.

Containment of Radioactivity: The first two parameters are common to all lists. The area radiation monitors have been added as part of the ERF coordination with SPDS as stated in NUREG-0737 Supplement 1, Section 3.4.d.

The ARMS list was taken from the Emergency Plan. Primary coolant activity monitors (not available and will not be installed in accordance with BWROG position on RG 1.97) and drywell activity monitors (not available but will be installed as part of RG 1.97 compliance) were not included here as compared to Table 2.1 (S. Levy) because this CSF has been limited to controlling radioactive release outside the primary containment.

2.

RCS Integrity / Containment Conditions: The parameter list of these CSFs is more comprehensive-than either of those in Tables 2.1, 2.2.

The main differences are the inclusion of torus air space pressure,-steam line_ monitor and ARMS.

The first two were 1

included since they are,important E0Ps parameters while the last one is a useful parameter for area contamination detection due to pipe breaks. The safety / relief valve position are missing from the OC list since it is felt that the operation of these valves can-be inferred from RPV pressure, drywell pressure and torus temperature. Secondary containment pressure was not regarded as a v;eful parameter since the ARMS are better indicators of the secondary containment integrity.

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Heat Removal: Again the Gyster Creek list is more comprehensive

,than either of the other lists. The inclusion of steam flow and

. feed flow was thought to-be important since a wide range of scenarios exist where those parameters are the best--indicators of heat' removal from the vessel. Core spray discharge pressure was included as compared to core spray flow (this may change to core flow when-instruments are chosen).

Reactivity / Power Distribution: The same parameters are used as in

' 4.-

~ Table 2 1 except for recirculation flow instead of the mode switch since the operation._ modes for Oyster Creek SPDS are AT POWER ~and

- SHUTDOWN.

Using the SRMs for-post shutdown reactivity monitoring was not

- considered to be beneficial since once the APRMs are below 2% after

a scram demand-(EOPs-criterion _for shutdown) enough negative reactivity has.been inserted to prevent a criticality concern during any acceptable time frame. Also, plant procedures require

- control rod positionsJt40.be checked'after a scram for partially inserted rods.

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Page 27 TABLE 3.3 0YSTER CREEK SAFETY PARAMETER SET Safety Function.

Parameter Radioactivity Control Stack Monitor Plant Ventilation Monitors Area Radiation Monitors

. Reactor Coolant System Integrity Steam Line Radiation Monitor Drywell Activity Monitors Stack Monitor Combustible Gases RPV Level Drywell Pressure Plant Ventilation Monitors Drywell Sump (Uncontrolled)

Steam Flow Isolation Demand Torus Air Space Pressure Feed Flow Scram Demand Heat Removal Neutron Flux Scram Demand RPV Level Recirculation Flow RPV Pressure Steam Flow Feed Flow Isolation Demand Core Spray Discharge Pressure (or core spray flow)

Reactivity / Power Distribution Neutron Flux Scram Demand Recirculation Flow Containment Conditions Drywell Pressure Drywell Temperature Torus Level Torus Temperature Isolation Demand Combustible Gases RPV Pressure Torus Air Space Pressure Area Radiation Monitors l-

TR 0 016 Page 28 4.0 RESPONSE TO TRANSIENTS / ACCIDENTS A broad range of transients / accidents were analyzed and tested against the SPDS parameter list. The purpose of these tests was to see if each event in a transient can be adequately monitored and evaluated by the SPDS parameter list-(Table 3.1).

The Oyster Creek Probabilistic Safety Analysis (4) (OPSA) transients and others were used for this purpose.

The list of transient /accidetit scenarios that were analyzed is shown in Table 4.1..An example of SPDS response to a LOCA is presented below.

This was chosen for its simplicity as an example but much more elaborate event sequences were analyzed (5),

4.1 Large/ Intermediate Break Below the Core Inside Containment An example of this type of break would be a recirculation line break or a break in any connected piping to the recirculation system below the core. This break is large enough to result in rapid depressurization of the reactor vessel which u. lows core spray injection without the assistance of the automatic depressurization system. Core spray injection is the only viable means of core cooling for this type of break since the reactor core cannot be reflooded due to the break size below the core.

In addition, a reactor scram is not required to bring the core to a subtritical state because the reactor core is not reflooded.

Break sizes with an effective total leak area of 0.5 square feet or more fall into this category. All transfers to the core damage state assume there is water on the reactor cavity floor. All

-breaks that meet the above criteria are categorized as initiating event Ll.

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TR 0 016 Page 29 4.1.1 Event Sequence Diagram L1 The"Large/IhtermediateBreakBelowtheCoreInsideCon-tainment" event sequence diagram is shown in Figure 4.2 and described below.

4.1.2 Reactor Vessel Internals Remain Coolable (Event VC)

Event VC is the core fuel and other reactor vessel internals remaining in a coolable geometry due to'the hydrodynamic and differential pressure loadings of the blowdown. Failure of event VC is assumed to result in an early, low pressure core damage state. SPDS monitoring of this event is done through RPV level, area and drywell radioactivity monitoring, vessel

-pressure, containment temperature and pressure.

4.1.3 Suppression Pool Integrity Maintained (Event SP)

' Event SP is maintaining suppression pool structural integrity as a result of.the hydrodynamic loads from the-blowdown so that it can be used as a water source for core spray and containment spray' operation. Failure of event SP is assumed to result in'an early, low pressure core damage state since the water source is lost and the containment is failed. Suppression pool air-space pressure, pool temperature, drywell pressure and temperature are SPDS parameters used to monitor SP integrity.

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TR # 016 Page 30 4.1.4 Core Spray Injection (Event CS)'

Event CS is actuation of at least one train of the core spray system on low-low reactor vessel water level or high drywell pressure and injection of torus water into the reactor vessel after the vessel-pressure drops below 285 psig.

Failure of event CS is assumed to result in an early, low pressure core damage state.

Suppression pool temperature, vessel level and core spray discharge pressure are used by SPDS for this event.

4.1.5 Containment Sprays & Suppression Pool Cooling (Event DS)

Event DS is actuation of the containment spray and emergency service water systems on high drywell pressure and low-low reactor water level to provide drywell sprays and suppression pool cooling for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Success is one containment spray pump train, one associated containment spray heat exchanger, and one associated emergency service water pump train providing actuation, injection, and cooling of drywell sprays using torus water.

Failure of event DS for 30 minutes or longer results in

^ high suppression pool temperatures which would result in questionable core and containment spray pump performance as a result of the probable loss of net positive suction head (NPSH). This is assumed to result in containment failure as a result of drywell overpressure. Failure of event DS demands-other measures such as operator action in aligning the fire protection water system to the core spray system (event FP) to provide core cooling without using core spray pumps even with the containment assumed to be failed. This event is monitored by SPDS through drywell and suppression pool pressure and temperature.

TR # 016 Page 31 e

4.1.6 Long Term Core Spray Injection (Event LT)

Event LT is continued operation of core spray injection for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Success is achieved by the operator taking the actions necessary to utilize redundant core spray system components to maintain injection. Success of event LT leads to stable decay heat removal using core spray, containment spray, and sup-pression pool cooling (success state S1). Failure of event LT demands other measures to provide core cooling without using core spray pumps such as operator action in aligning the fire protection system to the core spray system (event FP). This event is monitored by SPDS through RPV level, suppression pool temperature and level.

4.2 SPDS Detection A break inside the containment that results in high drywell pressure

-and low RPV pressure will trigger the Primary System Integrity CSF to the alarm mode.

If drywell pressure reaches drywell structural limits then the Containment Integrity CSF will change to the warning or alarm modes depending on setpoints used. The Containment Integrity CSF will also be triggered by suppression pool structural limits setpoints as monitored by pool level, temperature and air space pressure.

It is to be emphasized here that the SPDS contains all vessel, drywell and suppression pool available level, temperature and pressure parameters, which would completely monitor their integrity and heat removal status. The fuel clad integrity and containment of radioactivity CSFs may also be activated if relevant parameters setpoints are reached.

The overall SPDS response to L1 is shown on Table 4.2.

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- TR # 016 Page 32 7,

The final SPDS list shown in Table 3.1 was arrived at only after going through the OPSA and the other scenarios listed in Table 4.1.

A number of' parameters were deleted and added in this process before the final

. list was arrived at.

O T

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w

TR # 016 Page 33 o

TABLE 4.1 TRANSIENTS / ACCIDENTS SCENARIOS Transient Accident Initiators Tl Reactor Trip T2 Partial Pressurization T3 Full Pressurization T4 Loss of Offsite Power T5 Loss of Feedwater Flow T6 Excessive Feedwater Flow Loss of Coolant Accident (LOCA) Initiators

.L1 Large/ Intermediate Break, Below Core, Inside Containment L2 Small Break, Below Core, Inside Containment L3 Large/ Intermediate Break, Above Core, Inside Containment L4 Small Break, Above Core, Inside Containment L5 Large/ Intermediate Break, Below Core, Outside Containment L6 Small Break, Below Core, Outside Containment L7 Large/ Intermediate Break, Above Core, Outside Containment L8 Small Break, Above Core, Outside Containment L9 Inadvertent Opening of One Relief Valve L10 Inadvertent Automatic Depressurization System ( ADS)

Actuation Others Reactivity Transients (Rod Withdrawal Errors and Rod Drop)

Failure of the Off-gas System Oyster Creek May 2, 1979 Event N

i; TR # 016 Page 34 c

ETABLE 4.2.

SPDS RESPONSE T0 L1 CSFs' MODE REASON 1.

RCS Integrity / Primary System Integrity:

Alarm.

High Drywell Pressure & Low RPV Pressure 2.

Heat Removal / Primary System Cooling Alarm Level below Low-Low 3.

Heat Removal / Fuel

. Clad Cooling Alarm Level below Low-Low-Low (or on core spray operation) 4.

Containment Conditions /

Containment Cooling Alarm High Orywell Pressure (2 psig)/high torus tempera-

-ture/ limit curves violation

5..

Containment Conditions /

Containment' Integrity Alarm If drywell pressure / torus air space pressure / torus level reach structural limits

'6.

RCS. Integrity / Fuel Clad Integrity Alarm If drywell radiation monitors / combustible gases monitors reach alarm setpoint 7.-

Containment of Radioactivity.

Alarm If area radiation monitor.s in reactor building around isolation condensers, clean-up system piping, etc. reach alarm setpoint.

e I'

L

TR # 016 Page 35

(

REACTOR TRIP l

INITIATING EVENT BLOCK j

(T= TRANSIENT, L=LOCA)

T1 f

SYSTEM FUNCTION OR EVENT BLOCK REACTOR SCRAM (A= AUTOMATIC INITIATION.

SUCCESS M= MANUAL INITIATION)

U FAILURE STA8LE DECAY HEAT REMOVAL WITH CORE AND CONTAIN-SUCCESS OR STABLE STATE SLOCK l

MENTSPRAY l

CORE DAMAGE STATE TRANSFER BLOCK l

l FIGURE 4-1:

EVENT SEQUENCE DIAGRAM SYMBOLS o

s r

9

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.m, y,.--

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TR 0 016 Page 37

5.0 REFERENCES

l.

BWROG Committee'on Display / Procedures Integration, Project

Plant. September 14, 1983-2.

NSAC/21, " Fundamental Safety Parameter Set for Boiling Water

-Reactors," prepared by S. Levy Inc.

December 1980 3.

' Guidelines for an Effective Safety Parameter Display System

' Implementation Program. NUTAC, January 1983.

4.

OPSA, Oyster Creek Probabilistic Safety Analysis, Plant Analysis Update PLG-0253 (unpublished).

December 1982

' 5.

Oyster Creek SPDS Parameter Selection and Safety Analysis Study.

TDR 509 - Draft.

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