ML20078P113
| ML20078P113 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 10/28/1983 |
| From: | ILLINOIS POWER CO. |
| To: | |
| Shared Package | |
| ML20078P098 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8311030201 | |
| Download: ML20078P113 (70) | |
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3e CLINTON POWER STATION-SAFETY PARAMETER DISPLAY SYSTEM
' PARAMETER SET VALIDATION REPORT l
'B311030201 831028 PDR ADOCK 05000461 F
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CONTENTS Section Title 1
Introduction
'2
. Guidelines and Assumptions 3
' Safety Functions to Consider in~ Selecting a Safety Parameter Display System Parameter Set 4
' Application of Safety. Functions to the Clinton Power Station Safety' Parameter
' Display Parameter Set 4.1 System Overview 4.2 Containment of Radioactivity 4.3 Barrier Integrity.
4.3.1 Fuel Cladding Integrity 4.3.2 Reactor Coolant System Integrity 4.3.3 Primary Containment Integrity 4.3.4 Secondary. containment Integrity 4.4 Heat Transport 4.4.1 Fuel Clad Cooling 4.4.2.
Reactor Coolant System Cooling 4.4.3 Primary Containment Cooling 4.5 Reactivity Control 5
Validation of CPS SPDS Parameter Set 5.1 Validation' Procedure 5.2 Loss of Coolant Accidents 5.2.1 Large Breaks Inside Primary Containment 5.2.2 Instrument Line Pipe Break Inside i
Containment 5.2.3 Large Breaks Outside Primary Containment 5.2.4 Small Breaks Outside of Containment i
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CONTENTS (Cont'd)
Section.
Title 5.3 Reactor Pressurization' Events 5.4. Moderator Temperature Decrease Events 5.5 Reactivity and Power Distribution Anomalies 5.5.1 Control Rod Withdrawal Error - at Power 5.5.2 Control Rod Withdrawal Error - Refueling 5.5.3 Control Rod Drop Accident 5.5.4 Recirculation System Transients 5.5.5 Anticipated Transients Without Scram 5.6 Coolant Inventory Decrease 5.6.1 Inadvertent Safety / Relief Valve Opening 5.6.2 Loss of Feedwater Flow 5.6.3 Loss of Coolant Accidents 5.7 Coolant Inventory Increase 5.7.1 Inadvertent Operation of the High Pressure Core Spray System or Reactor Core Isolation Cooling System 5.7.2
' Failure of Feedwater Control to Maximum Demand 5.8 Decrease in Reactor Coolant Flow Rate 5.9-Failure to Remove Residual Core Heat 5.10 Loss of. Auxiliary System Power 5.10.1 Loss of Auxiliary Power Without Loss of an Emergency Bus 5.10.2 Loss of Auxiliary Power With Concurrent Loss of an Emergency Bus 5.10.3 Loss of Instrument Air 5.11 Radioactive Release from Subsystems and Components ii
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CONTENTS (Cont'd)
Section Title 5.11.1 Radioactive Gas Waste (Off-Gas) System Failure 5.11.2 Liquid Radwaste System Failure 5.12 Water Chemistry Transients 5.13 Results of-Transient and Accident Analysis 1 -
5.14 CPS SPDS Parameter Set Comparison With Other Parameter' Lists for BWR Accident Monitoring 5.14.1 Comparison With NSAC/21, " Fundamental Safety Parameter Set for Boiling Water Reactors" 5.14.2 Comparison With NUREG/CR-1440 BWR Accident Monitoring Variables 5.14.3 Comparison With Regulatory Guide 1.97 BWR Parameter List 5.14.4 Comparison With BWR Emergency Procedure Guideline Entry Condition Parameter Set 6
Conclusions and Recommendations.
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,4 Section 1 INTRODUCTION The Three Mile Island (TMI) Unit'2 accident has placed special emphasis upon the information available to the nuclear control room operators and the methods of displaying such information.~ The concept of a Safety Parameter Display System (SPDS) has been. developed as a partial response to deficiehcies inLthe man-machine interface.
The operational objective of the SPDS is to provide the control room operators with concise ond unambiguous data which will characterize the overall plant safety status under a' wide variety of plant operating conditions.
The objective of this report is to review the bases for the Clinton Power Station (CPS) SPDS Parameter Set and to apply methods of validation to the selected parameter set to ensure the appropriate parameters have been chosen.
This effort is part of the CPS SPDS Validation & Verification _ Program.
This report is based upon the methodology used in NSAC/21, " Fundamental Safety Parameter Set for Boiling Water Reactors", prepared by S. Levey Inc. for_the Nuclear Safety Analysis Center of_the Electric Power Research Institute. t
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.Section 2 GUIDELINES AND-ASSUMPTIONS A major-guide:for this report is'the-definition of the SPDS parameter; set-as a1 minimum set which is sufficient to' determine.
that the plantfis-infa safe condition.
A. minimum set'is. desired to: assist the user by limiting.the bulkiof information he must.
-assimilate to determinensafety-status, particularly during abnormal: events.
ReducingEthe number of parameters,oof course, g
lis counterbalanced by the need for sufficient information to make
.the determination of-safety status.-
An evaluation.of the specific equipment and displays which
'will. process these parameters is:not within the scope of this report. 'In this respect, it will be sufficient to show that a process for. determining. safety status exists.
The. discussion of p(
the process will indicate computer operations or~ displays, but this is onlyLintended for information purposes.
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- A second major guideline is that the SPDS Parameter Set must
!triggerithe usertinto reviewing the proper control' room displays, performing pertinent diagnosis.and initiating corrective actions.
.As:an> extreme illustration, consider a.SPDS Parameter Set which consists of-only1 one parameter'that designates the plant as safe
.The'usefulness of this-parameter is clearly-limited l'
orLunsafe.
because it gives the user-no suggestion lof where to begin in p
dealing with an unsafe condition.
Conversely, the degree to
'which the SPDS Parameter' Set will localize a problem must be limited to a few broad categories so that-the quantity of data presented'is tractable to-the user.
The primary users.will be l
-the control: room staff.'-
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. It:is assumed that'the readiness-to-serve-status of safety-
.related. systems will be monitored'by.other control room displays.
It is not reasonable.that the SPDS' Parameter Set should be u
expected to monitor-status of the many safety related devices and i
systems _in the plant.
To do-so would expand the set to an Lintractable' size.
Thus, the SPDS Parameter Set will not attempt 7
p-to monitor:the readiness of safety systems to respond to demands.
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- The'SPDS Parameter Set is primarily a. synoptic set of parameters.
That is, it is concerned with safety status at the
'present and does not Lattempt to diagnose how cuc why the current' status 1 developed or what the' status will be-later.
This is not to say that changes in status with time cannot be used to infer
' improvement or' deterioration, only that the SPDS Parameter Set is
.notlintended'to be used to make predictions or diagnoses.
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GUIDELINES AND ASSUMPTIONS Finally, the SPDS Parameter Set should-not'only permit detection of:the initial development of an unsafe condition, but it should also continue-to be~ effective in permitting continued monitoring of-the. safety status as conditions' change during the course.of an event.
Section 3 SAFETY. FUNCTIONS TO CONSIDER IN SELECTING A SAFETY PARAMETER' DISPLAY SYSTEM PARAMETER SET In selecting the SPDS Parameter Set, parameters should be identified which are sufficient to determine if the important
-safety functions are being performed.
Of these safety functions, the overall function (termed " Containment of Radioactivity,") is oto prevent the release of radioactivity from the plant.
Monitoring radioactivity in both gas and liquid discharges to the
-environs would permit evaluation of how well this function is being performed.
This radioactivity is being restrained from release by the fuel cladding 3 reactor coolant system and
' containment barriers.
It therefore becomes important to know if
'the reactor coolant system and containment barriers are being called upon to contain radioactivity because the information could' indicate that normally safe operations have become unsafe.
For example, primary containment purging would not be safe if high. containment radioactivity levels existed.
Success in containing radioactivity is dependent on the
-integrity-of the barriers provided to prevent release.
This
" Barrier Integrity" safety function should be monitored and can be divided into two parts.
First, it must be determined that integrity does or does not exist at a given time.
That is, does the barrier now have leak paths for radioactivity?
Second, are there significant threats to the integrity of the barriers even though they may currently have integrity?
Two threats of this type.have been identified which are of such major importance as to be considered separate safety functions.
They are " Heat
. Transport" and " Reactivity Control."
Heat Transport ~as a safety function is the recognition that failure to remove core power or even decay heat within a barrier will eventually and certainly destroy the barrier's integrity.
.This applies to the fuel cladding, reactor coolant system and primary containment.
Similarly, Reactivity Control as a safety function is the recognition that reactivity excursions or the failure to control steady state power level are intrinsic threats to barriers.
An excursion is characterized by an excess neutron production which rapidly raises power generation to the point that a destructive energy density is reached or approached in the fuel.
Failure to control steady state power level is characterized by a level of power production which exceeds the capability of the system to transport energy.
Failure to reduce power level by reactivity control leads to energy being released to the containment (via the safety relief valves) faster than it can be removed and eventual. failure of the containment as a barrier to radioactivity release. 1
Section 3 (Continued)
' SAFETY FUNCTIONS TO CONSIDER IN SELECTING A SAFETY PARAMETER DISPLAY SYSTEM PARAMETER SET Any. threat to barriers in addition to Heat Transport and Reactivity Control will be identified under Barrier Integrity.
The safety functions to be considered are summarized in Table 3-1.
As indicated in the prior description of these functions,-monitoring of their-performance is subdivided, according to-the barriers provided, as-an aid in a systematic and thorough development.
For example, the safety function of Barrier Integrity.is monitored by'considering, individually, the integrity of the Fuel-Cladding, Reactor Coolant System, Primary
. Containment and Secondary Containment.
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Safety Functions to be Considered in Selecting an SPDS Parameter Set
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' Containment of' Radioactivity 2.
Barrier. Integrity is Fuel Cladding Integrity Reactor Coolant System Integrity.
Primary Containment Integrity Secondary. Containment Integrity
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Heat Transport Fuel Clad Cooling Reactor Coolant System Cooling Primary Containment-Cooling 4.
Reactivity Control
i Section 4 APPLICATION OF SAFETY FUNCTIONS TO CPS SAFETY PARAMETER DISPLAY SYSTEM PARAMETER SET In this section, the CPS SPDS Parameter Set is discussed.
The selected parameters are identified by the safety functions that they monitor and are then entered in Table 4-1.
Selected parameters are underscored-in this and the remaining sections to clearly identify them as CPS selected parameters.
4.1 SYSTEM OVERVIEW The CPS SPDS is being implemented as a part of the Display Control and Performance Monitoring Systems (DCS/PMS) currently installed in the Main Control Room (MCR).
The Number 5 Cathode Ray Tube (CRT) in the NUCLENET (Principal Plant Console) has been designated as part of the SPDS display.
The SPDS Parameter Set Display will be accomplished through the use of a permanently displayed parameter set, a temporary Alarm Initiated Display (AID) and use of the Area and Process Radiation Monitoring (ARM /PRM) System.
The permanent display and the AID display will be provided on CRT #5 of the DCS.
Information necessary to monitor the radioactivity status is provided by a separate ARM /PRM System CRT in the MCR.
Figures 4-1 and 4-2 provide schematics of the SPDS dicplays.
4.2 CONTAINMENT OF RADIOACTIVITY As shown in Table 4-1, the following parameters are monitored by the CPS SPDS ARM /PRM displays:
Forty-six fixed digital Area Radiation Monitors (ARMS) are located throughout the plant to monitor gamma dose rate.
Twelve portable ARMS are available for connection to other parts in the plant.
Fourteen fixed digital Constant Air Monitors-(CAMS) measure airborne radioactivity within the station, with ten portable CAMS available.
One Process Radiation Monitor (PRM) samples the common station HVAC exhaust (with one in standby), one PRM monitors Standby Gas Treatment System (SGTS) (with one in B1 mdby),
one PRM cach monitors Pre and Post-treatment Air 7, ir Off-gas (with one in standby for Post-treatment), a,td one PRM samples the Liquid Radwaste Effluent discharge.
Six PRMs monitor various liquid streams to detect intersystem leakage of heat exchangers.
Finally, sixteen safety-related i
PRMs, with control functions to initiate SGTS, monitor HVAC ducts on Containment Building Exhaust, containment Building Fuel Transfer Vent Plenum, Fuel Building Exhaust, or Main Control Room air intake. __
The status of all 90 of the permanent monitors and as many of the 22 portable monitors as are connected to system communication ports shall be provided on the ARM /PRM Status Grid, as shown in Figure 4-2.
Monitoring of activity in gas discharges to the environs is performed by the CPS plant ventilation monitors and the main stack monitor.
The buildings monitored totally enclose all sources of airborne activity and the ventilation systems maintain these buildings at a negative pressure which assures that all discharges are through the exhaust ports where the activity monitors are located.
With the monitoring of the main stack, to which potentially radioactive gases are routed, the monitoring of gas discharges to the environs is complete.
Liquid discharges such as Radwaste Effluent and various liquid streams to detect intersystem leakage of heat exchangers are monitored by the licuid process monitors.
Containment of Radioactivity is completed by monitoring primary containment activity and a measure of reactor coolant system activity.
Reactor coolant s_ystem activity is not directly measured although a reasonably good representation is provided by the Air Eiector off-cas monitor which samples gases ejected from the main condenser.
This monitor is effectivc only when the condenser is connected to the reactor coolant system.
When the main steam isolation valves are closed, as is likely during the course of an accident, a measure of the reactor coolant activity level is not available on a continuous basis.
Direct and continuous monitoring of reactor coolant system activity, for purposes of SPDS, is not recommended.
The usefulness ' f this o
variable is discussed in the following paragraphs:
Regulatory Guide 1.97 specifies that the status of the fuel cladding be =enitored during and after an accident.
The specified variable to accomplish this monitoring is variable Cl--radioactivity concentration or radiation level in circulating primary coolant.
The range is given as "h Tech Spec limit to 100 times Tech Spec Limit, R/hr."
In Table 1 of RG 1.97, instrumentation for measuring variable Cl is designated as Category 1.
The purpose for monitoring this variable is given as " detection of breach," referring, in this case, to breach of fuel cladding.
The usefulness of the information obtained by monitoring the radioactivity concentration or radiation level in the circulating primary coolant, in terms of helping the operator in his efforts to prevent and mitigate accidents, has not been substantiated.
The critical actions that must be taken to prevent and mitigate a gross breach of fuel cladding are (1) shut down the reactor and (2) maintain water level.
Monitoring variable Cl, as directed in RG 1.97, will have no influence on either of these actions.
The purpose of this monitor falls in the category of "information that the barriers to release of radioactive l l
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.e material are being. challenged" and " identification of degraded conditions and their magnitude, so the operator can take actions that are available to mitigate the consequences."
Additional operator actions to mitigate the consequences of fuel barriers being challenged,-other than those based on Type A and B variables as defined in R.G.
1.97, have not been identified.
Regulatory Guide 1.97 specifies measurement of the radioactivity of the circulating primary coolant as the key variable in monitoring fuel cladding status during isolation of the NSSS.
The words " circulating primary coolant" are interpreted to mean coolant, or a representative sample of such coolant, that flows past the core.
A' basic criterion for a valid measurement of the specified variable is that the coolant being monitored is coolant that is in active contact with the fuel, that is, flowing past the failed fuel.
Monitoring the active coolant (or a sampla thereof) is the dominant consideration.
The post-accident sampling system (PASS) provides a representative sample which can be monitored.
The subject of concern in the RG 1.97 req'2irement is assumed to.be an isolated NSSS that is shutdown.
This assumption is justified as current monitors in the condenser off-gas and main steam lines provide reliable and accurate information on the status of fuel cladding when the plant is not ivolated.
Further, the post-accident sampling system (PASS) will provide an accurate status of coolant radioactivity, and hence cladding status, once the PASS is activated.
In the interim between NSSS isolation and operation of the PASS, monitoring of the primary containment radiation and containment hydrogen will provide information on the status of the fuel cladding.
4.3 BARRIER INTEGRITY 4.3.1 Fuel Cladding Integrity Failure of fuel cladding is accompanied by a release of radioactive gases.
Experience in operating plants has shown that the off gas activity monitor is capable of detecting the activity released by failure of a single fuel rod.
This sensitive indication of loss of fuel cladding integrity would be lost when the main steam line isolation valves are closed.
However, as noted above, direct and continuous monitoring of reactor coolant system-activity is not necessary.
4.3.2 Reactor Coolant System Integrity As shown in Table 4-1, the following parameters are monitored by the. CPS SPDS permanent and AID displays: i u
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,a Reactor Feed Flow and Total Core Flow, each in pounds per hour, shall be a permanent part of the SPDS Display.
Reactor Recirc Flow A and B, in thousands of gallons per minute, shall be a permanent part of the SPDS Display.
Wide Range Reactor Pressure and Narrow Range Drywell
. Pressure, in pounds per square inch gauge, shall be a permanent part of the SPDS Display.
Drywell Floor Drain Sump Flow, in gallons per minute, shall be a permanent part of the SPDS Display.
Because this flow is the most sensitive indicator of leakage it is a standard technical specification basis for shutdown on small breaks and shall also be displayed on the Alarm Initiated Display
. portion of the SPDS Display.
Drywell Equipment Drain Sump Flow, in gallons per minute, shall be indicated on the Alarm Initiated Display portion of the SPDS Display.
Monitoring of reactor coolant system integrity can be divided into the portion of the system within the primary containment and the portion outside the primary containment.
Within the primary containment, the fill and pump-out times of the drywell equipment drain sump and drywell floor drain sump are sensitive indicators of-relatively small leaks in the reactor coolant system.
They are particularly useful in detecting leaks which are within the cooling capacity of the drywell cooling system.
For larger leaks, the drywell will pressurize and indicate loss of reactor coolant system integrity.
Thus, the combination of monitoring' sumps and drywell pressure are prime indications of reactor coolant system integrity within the containment /drywell.
However, their limitations should be recognized.
When the reactor coolant system pressure and temperature are low, drywell pressure will not rise even during a major loss of integrity.
Thus, the reactor coolant system pressure is monitored to indicate whether_ the drywell-pressure is or is not valid as an indicator of integrity.
When it is not, reactor coolant system water level can be substituted as an integrity indicator.
Careful interpretation of water level and its trends might be required under some conditions to infer a leak and, indeed, cases can be postulated wherein a leak would not be detected.
In general, these would be leaks which are made up by inventory controls and, therefore, are of minor safety concern.
Also, the sump indicators would be effective in indicating leaks under such conditi-ns.
Thus, the combination of sump indicators-and water level should be effective when the reactor coolant system is at low pressure.
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In the case of high drywell pressure or low reactor coolant system water level, the discharge lines of the equipment drain sump and the drywell floor drain sump are automatically isolated.
These sump fill and pump-out times are, of course, not valid under these conditions and can be considered lost.
Loss of sump indications due to a low water level isolation in the absence of a high drywell pressure signal, requires the presumption that the reactor coolant system is leaking at a rate which can be cooled by the drywell cooling system.
This presumption is considered acceptable.
Loss of sump indications due to a low watar level isolation while at low pressure, so that high drywell pressure is not a valid leak detector, requires the presumption of a significant loss of reactor coolant system integrity.
Being on the conservative side, this presumption is also considered acceptable.
Finally, there can be an ambiguity in the interpretation of high drywell pressure.
Loss of drywell cooling will cause high drywell pressure as surely as a leak in the reactor coolant system.
The ambiguity can be resolved by recognizing that the pressurization rate due to loss of cooling will be slow and not accompanied by an increase in sump flows.
Therefore drywell pressure will be monitored.
The integrity of that portion of the reactor coolant system outside of the primary containment can be monitored by the plant ventilation radioactivity monitors, the status of which is provided by the SPDS ARM /PRM Status Grid display.
Because the buildings monitored completely enclose the extensions of the reactor coolant system, their ventilation monitors are expected to detect leaks by sensing radioactivity carried by leakage ficw which eventually finds its way into the plant ventilation systems.
Other parameters could_be monitored to fulfill the need for determining reactor coolant system integrity.
Specifically, reactor system discharge to the suppression pool is possible via the Safety / Relief Valves (SRVs).
The SRV positions, open or closed, could be monitored to fulfill this function.
The need for monitoring this parameter as part of SPDS will be discussed in more detail in Section 5.0.
Presently, the CPS SPDS Parameter Set does not include SRV position as a monitored variable.
4.3.3 Primary Containment Integrity As shown in Table 4-1, the following parameters are monitored by the CPS SPDS permanent and AID displays:
Containment pressure, in pounds per square inch gauge, shall be a permanent part of the SPDS display. l
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,o Containment Isolation, both inboard and outboard, for each of the 11 isolation groups specified in CPS Procedure No.
10N4001.02S, " Automatic Isolation", shall be a permanent part of the SPDS display.
The Alarm Initiated Display portion of the SPDS Display shall provide drywell, containment, and suppression pool temperatures, in degrees Farenheit, suppression pool level, in feet, containment and drywell pressures, in pounds per square inch gauge, and containment hydrogen concentration expressed as a percentage.
A primary containment leak below the suppression pool water level will be indicated by a decline in suppression pool level.
Since suppression pool level is a parameter that will be monitored by the AID display portion of SPDS, the operator will be provided with an immediate indication of such a leak in the CPS primary containment, llavi.ig selected suppression pool level as a parameter to be monitored, it will be necessary to include those limits on level which are important to the safety function of the suppression pool.
Without them, the user might infer erroneously that there are no limits on level.
One of these limits is dependent on the coolant system pressure; hence, it is needed as a parameter to permit generation of a level limit.
The suppression pool load limit curve is defined by these two parameters.
The suppression pool load limit curve is obtained by calculating, for SRV actuations at various RPV pressures, the suppression pool water level at which the stress in the limiting submerged structural component (including the containment boundary) equals the yield stress in that component.
Direct threats to the integrity of the containment can be indicated by high containment pressure and/or high drywell pressure.
Monitoring both of these parameters, both negative and positive values, will therefore detect an approach toward design limits for whatever cause.
In certain highly degraded events leading to a loss of RPV water level, a reaction between fuel cladding and steam may take place generating significant quantities of hydrogen.
An uncontrolled combustion with oxygen in the containment could threaten primary containment integrity.
For this reason, hydrogen concentration is monitored by the AID display of the CPS SPDS.
Presently, hydrogen concentration is monitored only in the containment by SPDS.
During a degraded core scenario, hydrogen can be released directly into the drywell and/or the containment wetwell.
Hydrogen concentration can be measured from several sample ports in both the drywell and containment using hardwired safety grade instrumentation already available in the CPS main control room. _ -
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'In' addition, due to steam inerting in the ldrywell, hydrogen burns in the'drywell are much less likely to occur, thus supporting the position that monitoring the drywell for hydrogen, via SPDS, is not required.
Therefore, hydrogen concentration-in the drywell need-not be monitored by SPDS.
Suppression pool temperature is a heat. transport parameter but is also important to-primary containment-integrity.
The CPS containment is:of the. Mark III design which is a " pressure-suppression" type,.the suppression pool functioning as an
. energy contained in the.RPV.
High suppression pool temperatures reduce-the available-heat. capacity, reduce'the net positive suction head of pumps drawing suction from the pool, increase the vapor. pressure in.the suppression chamber (i.e. containment),
affect containment loads during SRV actuation, and reduce operating margins.. High temperatures could result from a LOCA,
.SRV actuation, or RCIC. operation.
The heat capacity temperature limit.(HCTL) defines the set of initial reactor pressure and suppression pool temperature combinations for which continuous stable steam condensation is assured should an RPV blowdown be initiated from these conditions.
The heat capacity level limit (HCLL) defines'the required margin between actual suppression pool-temperature'and HCTL when the suppression pool water level
.s below:that assumed in calculating HCTL.
This adjustment is i.necessarylto accommodate the resulting higher temperature change during a blowdown without exceeding the pool temperature limits imposed by HCTL.
'Drywell temperature and containment temperature are monitored for similar reasons.
High drywell temperature could be caused by partial or complete loss of drywell cooling, or by a Jpipe break or. leak'in the.drywell.' Adverse effects include
.. inaccurate RPV water level' indications and possible equipment or structural damage.
In addition, drywell temperature also affects the drywell. atmospheric' pressure._ Deleterious effects of high containment temperature are similar to those of high drywell temperature..
' Containment isolation valve' positions status is provided by Such indication on
" group" on the CPS SPDS permanent display..
Monitoring of the SPDS-is.of a " readiness to serve" nature.
cdrywell pressure and/or primary containment activity would serve as triggers for-the operator to review isolation valve status.
However,.such indications on the SPDS will provide direct-status
.of primary containment integrity.
14.3.4 Secondary Containment Integrity "The_ secondary containment (gas control boundary for CPS) is designed to operate continuously at a small negative pressure which is an excellent indicator of integrity.
The amount of
.cinleakage is not of great concern as long as all outflow is to the designed exhaust points and this is assured if a negative pressure is maintained.
The need for monitoring'this parameter N -
as part of SPDS will be discussed in more detail in Section 5.0.
Presently, the CPS SPDS Parameter Set does not include secondary containment pressure as a monitored variable.
4.4 HEAT TRANSPORT As shown in Table 4-1, the following parameters are monitored by the CPS SPDS displays:
Reactor Wide Range Water Level, in inches, and Reactor Steam, Feed, and Total Core Flow, in millions of pounds per hour, are provided as a permanent part of the SPDS display.
Because of its importance, Reactor Water Level is also provided as a single value in the AID display portion of the CPS SPDS.
4.4.1 Fuel Clad Cooling When core power is at the decay heat level, adequate cooling of the fuel cladding is assured if the bundle is below the sensed water level.
Hence, reactor vessel water level is a fundamental safety parameter and is monitored by the CPS SPDS.
If water level cannot be maintained above the top of active fuel, it is possible to remove core decay heat via the core spray systems.
If the core spray flow of either of the two systems (Low Pressure Core Spray and High Pressure Core Spray) reaches rated, adequate core cooling is assured.
Although it is desirable that core spray flow be monitored on the SPDS, the monitoring of reactor vessel water level will provided adequate warning of tae possible development of adverse conditions requiring use of core sprays.
Therefore, it is believed that adequate time.ts available for the
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operator to use other means to monitor core spray flows.
At higher power levels, cooling flow is required..
Qualitatively, the flow required increases with power.
To determine if cooling is adequate, both core average power range monitor, and core flow are required, and they are included in the CPS SPDS Parameter Set.
As these are core average parameters, they do not preclude local or regional conditions in the core from being significantly different.
Under current practice, adequate local heat transport is assured by monitoring against established local core operational limits.
These limits are Critical Power Ratio (CPR), Maximum Planar Linear Heat Generation Rate (MAPLHGR) and Preconditioning Interim Operating Management Recommendations (PCIOMR).
Core performance monitoring with respect to these limits is based on measurements derived from the in-core neutron flux detectors.
The capability of thermo-hydraulic modeling of core performance is restricted, to date, to normal nucleate boiling flow and heat transfer modeling, and is limited to fuel rods and flow paths which are in as-designed configurations.
While such modeling methods have considerable merit, they are not recommended for inclusion in the CPS SPDS Parameter Set because the necessary assumptions concerning the physical configuration of fuel rods and flow passages make the results less than conclusive.
This, coupled with the large amount of data, periouic reinitializing and extensive calculations required, makes them inconsistent with the selection philosophy and the remainder of the fundamental safety parameter set.
Therefore the monitoring of core average parameters via SPDS is sufficient.
4.4.2 Reactor Coolant System Cooling If the core is adequately cooled as confirmed by the parameters noted in the prior discussion of " Fuel Clad Cooling",
it is only necessary to monitor reactor coolant system pressure to determine that heat transport from that system is adequate.
Clearly, if pressure significantly exceeds the safety valve setpoints, adequate heat transport is not being provided.
If pressure is below this level, energy transport is not required from the point of view of immediate concern for the integrity of the reactor coolant system.
Reactor pressure does affect the performance of emergency core cooling systems.
For example, reduced pressere is required for rated core spray to be reached.
However, without knowing the intended operating state, pressure cannot be deemed safe or unsafe over an extended range.
It is not recommended that signals and logic be developed to resolve this indeterminancy but that the Fuel Clad Cooling parameters be monitored and, if the cladding is not adequately cooled, diagnosis be employed to determine if the reason is incorrect reactor coolant system pressure.
Similarly, reactor coolant system pressure response is indicative of the performance of the residual heat removal (RHR) system when it is used to remove energy from the reactor coolant system.
It will not be necessary to monitor RHR parameters on the CPS SPDS for, if the pressure does not respond as intended, a diagnosis using other data in the control room is appropriate.
Finally, reactor coolant system pressure is not sufficient to indicate that reactor coolant system heat transport is being accomplished if the reactor coolant becomes superheated.
Superheating is only possible, although not certain, if the fuel cladding is inadequately cooled.
However, under this degraded condition, core cooling is the priority item and until it is re-established and the coolant returned to a non-superheated state, transport through the reactor coolant system is of little operational interest.
Therefore, monitoring for superheated conditions in the reactor coolant system is not recommended for the CPS SPDS.
4.4.3 Primary Containment Cooling The primary containment system has systems which removes varying amounts of core fission and decay power under different plant operating states both normal and abnormal.
Further, some of the operations are discontinuous and/or undergo long transients before a steady state is reached.
For instance,.
I following a steam line isolation valve closure, safety / relief valves transport decay heat to the suppression pool intermittently until the primary system is depressurized via the RHR system.
Extensive logic and data would be required to determine what is the intended heat transport path through the primary containment at a point in time.
To overcome these difficulties, it is proposed that the containment be monitored to j
determine if too much energy is stored or if energy is being l
stored at an excessive rate.
To this end, the suppression pool J
temperature, and level are monitored by the CPS SPDS.
Level is indicative of the pool mass while temperature measures energy per unit mass.
Their product indicates energy stored.
With regards to the rate at which energy is being stored, some indication of l
SRV Position Status would be useful.
This will be discussed in
]
more detail in Section 5.0.
To complete the primary containment heat transport monitoring, the drywell and wetwell airspace temperature are monitored by the AID display portion of SPDS.
This should be performed to preclude the oversight of energy being stored in either airspace region.
4.5 REACTIVITY CONTROL As shown in Table 4-1, the following parameters are monitored by the CPS SPDS displays:
The Average Power Range Monitor (APRM) channels monitor power during normal cperation and indirectly indicate if negative reactivity has been inserted upon receiving a r
Reactor Protection System (RPS) trip by displaying decreasing power.
The four APRM channels shall be averaged.
Any APRM found to be more than 10% from the average shall be ignored and the average of the remaining APRMs shall be a permanent part of the SPDS Display.
The APRM display shall be a percentage of full power indication.
The Source Range Monitors (SRMs) provide neutron flux indication during reactor startup/ shutdown or low flux level operation.
The SRMs allow the operator to confirm long term net negative reactivity by observation of power at steady i
state source levels.
This ensures safe shutdown and the ability to detect potential restart events.
The four SRM signals shall be averaged and indicated as a permanent part i
of the SPDS Display.
SRM information shall be presented as
[
counts per second with rate information provided as period (time for a power change by a factor of e) in seconds.
Scram Discharge Volume (SDV). is of interest to an operator during shutdown.
While neutron flux is,the actual indicator of reactivity, it is prudent for an operator to be alerted whenever the ability to insert control rods may be jeopardized.
An indication of SDVs A and B, in gallons, has been provided on the Alarm Initiated Display portion of the SPDS Display. _..
..o Monitoring of-Reactivity Control is accomplished on the CPS SPDS displays through the use of APRM and SRM indication.
This is considered acceptable.
However, note that when the flux level has extended the range of the SRMs, these monitors are withdrawn from the core, and hence cannot measure flux.. To preclude erroneous. interpretation of their signals when the SRMs are withdrawn, the control room operator should'be made procedurally aware that the SRM positions should be checked prior to interpretation of the indicated flux levels.
Finally, the ability of the SRM to detect criticality can be employed to detect inadvertent criticalities if the position of the reactor mode switch is known.. Although reactor mode switch position is not recommended as an SPDS parameter, its importance to the operator should be noted.
Table 4-1 CLINTON POWER STATION SAFETY PARAMETER DISPLAY SYSTEM PARAMETER SET PERMANENTLY DISPLAYED PARAMETERS Safety Function Parameters Reactivity Control APRM (Neutron Flux)
SRM (Neutron Flux)
Reactor Core Cooling /
Wide Range Reactor Water Level Heat Removal
~
Reactor Steam Flow Reactor Feed Flow Total Core Flow Reactor Coolant System Integrity Reactor Steam Flow Reactor Feed Flow Total Core Flow Reactor Recirculation Flow (A&B)
Reactor Pressure (wide range)
Drywell Floor Drain Sump Flow Drywell Pressure Radioactivity Control (ARM /PRM 46 Fixed ARMS (throughout Status Grid)
Plant) 12 Portable ARMS 14 Fixed CAMS (throughout Plant)
- 10. Portable CAMS 2 PRM (common station
^
HVAC exhaust) 2 PRM (Standby Gas Treatment System) 3 PRMs (one'in Pre-and two in Post-treatment Air Ejector Off-gas) 1 PRM (Liquid Radwaste Effluent Discharge) 6 PRMs (various liquid streams) 16 PRMs (monitor various building HVAC exhaucts)
Containment Integrity Containment Pressure (narrow range)
Containment Group Isolation (1-11) c.
p....
Table ' -1.(Continued) 4
~CLINTON POWER'. STATION SAFETY PARAMETER DISPLAY SYSTEM.
- PARAMETER-SET ALARM INITIATED DISPLAY'(AID) PARAMETERS Parameters
- Safety Function Reactivity Control Scram' Discharge Volume--
(levels A&B)
Reactor Core Cooling /'
Reactor Water Level
-Heat Removal
-Reactor Coolant Drywell Floor Drain Sump System Integrity-Flow
~
Drywell Equipment Drain Sump Flow
-. Radioactivity Control.
'N/A
.(ARM /PRM Status Grid)
Containment Integrity Drywell Pressure DrywelliTemp,erature Containment Pres'sure Containment Temperature Suppression-Pool Water Level Suppression Pool Temperature Containment-Hydrogen Concentration
f., -
1........
2........
3........
4........
5........
6........
7..
1 APRM V@1 WM
^
2
'M J
nnx xun
-,$p.12
^
.3
-,493 4
SRM ll ::::: ; ::::'::::: CPS PERIOD fpi?4 8En c -
- x E:t
-xxx SEC
+9dPS5
'.;?
. 5
' d)).6
. :. 6
.b?P7 7
'RX LEVEL (WR) l: ::':'::::::
':.:: INCH D33 8
~
EEE3
-xxx
-xxx
' INCH
. '.,,, 8
....e
%.9 9
10 RX STEAM FLOW anci xx.n u n. ::
/- 10 MLB/HR
' !11
, 11
- :::::'llll: : : ::::::: MLB/HR xn.n x:c. n
'.u! 12
.RX FEED FLOW EE:2 9::'.12 1
M L B / H R.. 9. p ll: : :::::::: lll': '!ll MLB/HR
'13 l,,n:'.; 14
-TOT CORE FLOW e=2
~ ff
.t4
., ' M5; I15
' *y ~,,. " '
xx.x -
xx.x
. 15 7'
4
..,~([16 u '16
.. >. A a=1
... ~
xx.x
.xx.n
- 17
'RX RECIRC FLOW
- llllllll: lllll: ::
KGPM.
KGPM OVd7
.B EE!:2 xx.n xx.x
- wp318 c.9 -.;j... y..
. W 18
,39
~
3439 20
- RX PRESSURE (WR)
- S$20 22
- W 21 21 szm:
nunn xxxx PSIG 4
22
.23 DW FLOOH DRAIN GPM
,'GPM
- 23 24 SUMP FLOW EZ:2 xx.n xx.n
.- ;,' 24 25 25 26 DW PRESSURE (NR)
- :::':::: l PSIG
'26 27 EE'3 n.x x.x PSIG 27 2e 28 29 CONTAINMENT
- :::..: :.::::: ': PSIG 29 30 PRESSURE sm:2 xx.x u n. ::
PSIG 30 31 31 32 32 33 CNMT ISOLil 1082 IO 3 10 4 IO 5 IO 6 IO 7 10 8 10 9 10 10 10 11 IO 33 34 34
~ 35 35 ALARM INITIATED DISPLAY 36 36
- 37 RX WTR LEVEL
-xxx.x IN SUPP POOL LVL xx.x FT s37 38 38 39 DW PRESS xx.x PSIG SUPP POOL TEMP xxn.x F
.~ 39 40
- 40 41 DW TEMP xxx.x F CNMT PRESS
- ca.x PSIG 41 42 42 l
43 DW EO SUMP FLOW xx.x GPM CNMT TEMP xxx.x F 43 44.
44 45 DW FL SUMP FLOW xx.x GPM CNMT H2 CONC xxx.x %
45 46 46 47 SDV A LEVEL xx GAL-SDV B LEVEL xx GAL' 47 48 48 1........
2........
3........
4........
5........
6........
7..
CIGURE E -L :
-19 A -
a s
STATUS GRID CLINTON POWER STATION UNIT I DAY XXX TIME XXXX:Xx FUEL BLDG coNTAWMENT AUX-~ TURsNE BLOG UMASEND' 3
1 2 31 32 33 34 35 81 98 93 94 96 238 4
5 6
82 96 39 loo 237 7 8 9
45 83 sol 802 las los 50 85 11 H1 11 5 24 5 253 55 57 58 59 145 18 lis 20 61 6z 63 64 65 87 az3 251 262.
67 26 7.9 30 95 77 7El139 90 90 lag 14 0 coutRoc SLD6 yow 4sTE eLos pysg gy k6 120 48 40 18 6 18 8 COLOR CODES 199 200 NOR81 uN-INir STIC-ey faist.
223 M4#wrte-ALM 229 236 ALTALM ALARM ACKNOWLEDGE Mon #;
7ABLE OF A. ARMS I
101004166 coZ 223 033 til OSI 045 l
Fieuge 4-2. : 5FDs ARM /PEM P ' 5fAY-
- 19 3 -
i e
1 l
't j
g-l
.Section 5 VALIDATION OF CPS.SPDS PA'RAMETER SET' 5.'l
. Validation Procedure The validation procedure consists of reviewing a-broad range of possible plant transients and accidents and determining.the pertinent parameters needed for early and continuous monitoring.
The pertinent parameters for each event are compared to the CPS SPDS parameter set discussed in Section 4 (identified in this section as Table 5-2) to see if all necessary parameters have
.been~ identified.
The transients and accidents to be reviewed are taken from the Chapter 15 of CPS Final Safety. Analysis Report (FSAR),, WASH-1400, and significant operating plant events as well as events-selected by'the SPDS V&V Team.
A list of the' events reviewed is presented in Table 5-1 and then discussed individually in the order shown.
The' CPS SPDS Parameter Set is also compared'to four other documents which contain pertinent parameter lists to be considered for. the SPDS in a Boiling Water Reactor (BWR) plant:
'l.
NSAC/21, " Fundamental Safety Parameter Set for Boiling Water Reactors". (Table 5-3)
-2.
'NUREG/CR-1440:
also entitled EGG-EA-5153 (8), is a
-study of light water reactor status monitoring during accident; conditions.and was performed for the NRC.
(Table 5-4)
~3.
Regulatory Guide 1.97:
these BWR signal lists are not necessarily intended to be SPnS signals but instrumentation to monitor these parameters are recommended to assess plant and environmental
-conditions,during and following an accident.
(Table 5-5) 4.
The BWR Emergency Procedure ~ Guidelines (EPG) - these EPG's contain generic lists of symptom-based entry conditions which should be considered in choosing an SPDS Parameter Set. (Table 5-6)
~ t.
5.2-LOSS OF COOLANT ACCIDENTS 5.2.1 Large' Breaks Inside Primary Containment Description of Event:
This event is any break within containment resulting in reactor coolant system pressure boundary leakage from a flow area greater than or equal to 0.1 square feet.
Description of Plant Transient:
A large break will result in the following predictable transients:
Rapid loss of reactor coolant system pressure Rapid increase in drywell/ containment pressure Drop in reactor coolant system water level A large suppression pool temperature rise Possibly some fuel clad failures resulting in an increase in reactor coolant system activity.
Automatic Actions:
The following automatic actions are expected to occur for this class of events:
. Scram and isolation of primary containment.
Start of diesel generators.
Initiation of LPCS and the LPCI mode of RHR.
Appropriate Operator Actions:
After observing low reactor coolant system water level and high drywell pressure, the operator should confirm all control rods full in, isolation of the reactor coolant system and primary containment, HPCS and RCIC flows initiated, and diesel generators running in standby.
Af ter ' observing low reactor coolant system water level and low reactor coolant system pressure, the operator should determine if LPCS and LPCI are operating and if not, take steps to place them
-in operation.
After observing high suppression pool temperature, the-operator should place the RHR in the suppression pool cooling mode, and monitor hydrogen concentration in the drywell.
The parameters to cue the operator actions under this event are:
Reactor coolant system level Reactor coolant system pressure 1
3 Drywell/ containment pressure Suppression pool temperature Containment isolation' groups It.is apparent.that the CPS SPDS parameter set provides the plant
. operators with sufficient information to trigger appropriate responses to this event.
5. 2.' 2 Instrument Line Pipe Break Inside Containment Description of Event:
This event is any break within containment, but outside the drywell, resulting in reactor.
cooltsnt system pressure boundary leakage from a small steam or liquid instrument line.
Description of Plant Transient:
This event results in the release of reactor system coolant to the containment resulting in:
Increase in area radiation level Increase in area temperature Increase in floor drain sump level Automatic Actions:
The following automatic actions are expected to occur for this class of events:
Initiation of appropriate containment isolation signals from the leak detection system Appropriate Operator Actions:
After observing the leak detection system initiation, the operator should confirm appropriate isolations have occurred.
After having~made the decision to shutdown the plant, the operator should proceed to in an orderly manner.
The parameters to cue the operator actions under this event are:
Leakage Isolation Demand Containment Isolation Groups Containment Area Radiation Levels Although the CPS SPDS Parameter Set does not include monitoring for a leakage isolation demand, the remaining parameters are monitored which are indicative that an excessive leak is in progress.
A leakage isolation demand signal is monitored by other indicators in the CPS Main Control Room and is not considered significant enough for SPDS..-
i 5.2.3 Large Breaks Outside Primary Containment Description of Event-This event is a large break in any of the following piping systems containing reactor coolant outside of the containment:
Main Steam Line Reactor Water Cleanup system Residual Heat Removal system during shutdown mode Reactor Core Isolation Cooling system steam line Main Feedwater j
Description of Plant Transient:
A large break will result in the following predictable transients:
Rapid loss of reactor coolant system pressure Drop in reactor coolant system water level Rise of secondary containment pressure Increase in Area radiation levels Increase in Area water levels Automatic Actions:
The following automatic actions are expected to occur for this class of events:
Isolation valve closure in the broken line.
Scram if there is a Main Steam Line Isolation Valve closure.
Operation of the safety / relief valves if there is a main steam line valve closure.
Initiation of RCIC and HPCS if low reactor coolant system water level occurs.
Appropriate Operator Actions:
At the initiation of leakage isolation demand, the operator should review the status of all isolation valves.
If the break is inside of the secondary containment, high secondary containment pressure would occur.
High signals from the plant ventilation monitors could also indicate the location of the break.
In addition, reduced reactor coolant system pressure and a loss of reactor coolant system water level when operating the Residual Heat Removal System would be the operators cue to manually isolate this system.
t The parameters to cue the operator actions under this event are:
Leakage isolation demand Secondary containment pressure Plant ventilation monitors Reactor coolant system pressure Reactor coolant system level As discussed previously, a leakage isolation demand signal is not deemed necessary for monitoring by the SPDS display.
Of the remaining parameters listed above, all are displayed by the CPS SPDS except Secondary Containment pressure.
The Secondary Containment is designed to operate continuously at a small negative pressure which is an excellent indicator of integrity.
In the event of a pipe break or excessive leakage into the Secondary Containment, the activity released can be properly treated by the SGTS.
The Secondary Containment outflow must pass through the design exhaust points and this is assured if negative pressure is maintained.
Therefore, Secondary Containment pressure should be monitored by the CPS SPDS.
More details regarding the recommended monitoring of this parameter are provided later in this section.
5.2.4 Small Breaks Outside of Containment Description of Event-This event is a small break or leak in any of the following piping systems outside of containment:
Main steam line branch lines Reactor Water Cleanup system Residual Heat Removal system Reactor Core Isolation Cooling system Description of Plant Transient:
If the break or leak is small enough, there will not be a significant plant transient.
During power operation, the feedwater flow rate will increase to offset a small leak.
During Residual Heat Removal system shutdown operation, a small leak will result in a decrease in reactor coolant system water level if the reactor pressure vessel head is not yet removed.,After heat removal and flooding of the refueling cavity, a leak will cause a decrease in cavity water level or a decrease in fuel storage pool water level if the refueling cavity and pool are connected.
Flooding in the RHR/RCIC cubicles would be evident from leak detection alarms and area water levels / temperatures.
-s
?
~'
If there is radioactivity in the reactor system coolant, a small break event can cause fission product release to the environs via the turbine building and secondary containment gas control boundary ventilation systems.
Automatic Actions:
There are no automatic plant actions that will occur for this event..
Appropriate Operator Actions:
Locate the leak and shutdoun if the leak is resulting in an unacceptable release of radiation or loss of reactor coolant system water level.
The operator will be warned of a small leak problem from radiation levels in the secondary ccntainment and turbine building ventilation exhausts as indicated by the plant ventilation monitors.
If there is little reactor coolant system activity, then there will be no significant offsite release.
Normal plant inspection and patrolling can be used to detect small non-radioactive leaks.
The parameters to cue the operator actions under this event are:
Plant ventilation monitor level Reactor coolant system water level These parameters are found in the CPS SPDS Parameter Set.
P <
a 5.3 REACTOR PRESSURIZATION EVENTS Description of Event:
This class of events results in an inability to deliver steam to the main turbine and condenser at rates equal to reactor operating power.
Typical causes of this event are:
Turbine pressure regulator failure - closed or partially closed.
Generator load rejection Turbine trip with or without turbine bypass Main steam isolation valve closure Loss of condenser vacuum Loss of auxiliary power Description of Plant Transient:
Inability to deliver steam to the turbine and condenser results in the following predictable transients:
Increase in reactor coolant system pressure due to the loss of the turbine / condenser heat sink.
Increase in average power range monitor level caused by positive reactivity insertion due to void collapse with rising pressure.
Reactor scram from high flux, loss of condenser vacuum, etc.
Operation of one or more safety / relief valves.
Loss of turbine driven feedpump flow.
Initiation of RCIC and HPCS by low reactor coolant system water level.
Increase in suppression pool temperature caused by steam flow through the safety / relief valves.
Automatic Actions:
The following automatic actions will occur for this event:
Turbine Trip Scram Initiation of RCIC and HPCS Opening of one or more safety / relief valves Start of diesel generators if auxiliary power is lost Appro?riate Operator Actions:
The operator actions which should be tacen are:
Determine cause of high reactor coolant system pressure.
Evaluate corrective action.
It may be possible to reopen the main steam line isolation valves, est' blish turbine a
bypass, etc.
Confirm scram on receipt of scram demand.
After observing low reactor coolant system water level, verify RCIC and HPCS initiation.
After verifying that no safety / relief valves are stuck open by reviewing the safety / relief valve positions, monitor the suppression pool temperature to indicate when suppression pool cooling is needed.
After stabilizing the reactor coolant system water level, pressure,.and heat transport, the plant can be taken to cold shutdown by remote manual depressurization and initiation of the shutdown cooling mode of RHR.
The parameters to cue the operator actions under this event are:
Scram demand Safety / relief valve positions Reactor coolant system pressure Reactor coolant system level Suppression pool temperature Of those parameters listed above, a scram demand signal and safety / relief valve positions are not monitored by SPDS displays.
Scram demand signals can originate from a number of independent scram sources..These scram signals can be generated ~by any one of a number of causes (e.g. high drywell pressure, MSIV closure, low reactor water level, etc.).
Indication that a scram demand has occurred, in and of itself (i.e. without additional detailed 4
information) is not considered significant enough to warrant displaying on SPDS (main control room annunciation that a scram demand signal has been generated is available).
However, as noted in Section 4.0, regarding reactor coolant system integrity, safety / relief valve (SRV) positions should be monitored.
Therefore, it is recommended that. particularly due to industry operating experience with SRV failures (e.a. TMI-2), SRV positions should be monitored by the SPDS.
More details regarding the recommended monitoring of this parameter are provided later in this section...
5.4 MODERATOR TEMPERATURE DECREASE EVENTS Description of Event-This is a class of transients that results in a decrease in reactor coolant temperature at the core inlet (e.g. a loss of feedwater heating event).
Typical of such events are:
Operation of Residual Heat Removal System in shutdown cooling mode while at low power and low pressure.
Loss of Feedwater heating while at power Feedwater controller failure - to maximum demand Pressure Regulator Failure - Open Inadvertent Safety / Relief Valve Opening Startup of Idle Recirculation Pump Description of Plant Transient:
A decrease in reactor coolant temperature at the core inlet will result in increased reactivity and an increase in power.
The turbine pressure regulator would be expected to respond by increasing steam flow and generator output.
Also, operation of the Residual Heat Removal system in its shutdown mode while the reactor is at low power will also result in a decrease in core inlet temperature and a rising
-neutron flux.
However, Chapttr 15 of the CPS FSAR, which considered all listed transients, shows these transients to be Yelatively mild and slow.
Automatic Actions:
If the plant is in the automatic flow control mode, it is expected that recirculation flow would automatically decrease to maintain constant power.
If the plant is in the manual flow control mcde, the power will increase due to increased subcooling at the core inlet.
There may or may not be a scram due to high neutron flux, depending on the severity of the transient.
Appropriate Operator Action:
Neutron flux data, as measured by the average power range monitors would signal the operator to take corrective actions such as manual scram, control rod insertion, or core flow rate reduction.
The parameter to cue the operator action under this event is the average power range monitor which is part of the CPS SPDS Parameter Set. -
1 4
5.5 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 5.5.1 Control Rod Withdrawal Error - At Power The Rod Withdrawal Error (RWE) transient results from a procedural error by the operator in which.a single control rod or a gang of control rods is withdrawn continuously until the Rod Withdrawal Limiter (RWL) mode of the Rod Control and Information System (RCIS) blocks further withdrawal.
Description of Plant Transient:
If the plant is at full power, the worst case withdrawal error results in 104% power, a peak heat flux of 139% and a minimum critical heat flux ratio of 0.75.
The transient will be less severe at lower power levels.
Automatic Actions:
There is no automatic correction of the conditions resulting from this event.
The local power range monitor (LPRM) will detect high flux in the vicinity of the withdrawn control rod.
However, the increase in average power will be insufficient to cause scram.
Appropriate Operator Actions:
Under most normal operating conditions, no operator action is required since the transient which would occur would be very mild.
Should the peak linear power design limits be exceeded, the nearest Local Power Range Monitor (LPRM) would detect this phenomenon and sound an alarm.
The operator would acknowledge this alarm and take appropriate action (such as scram, reinsert rod (s), etc.) to rectify the situation.
Since an increase in the nearest LPRM level would sound an alarm in the main control room, it is not considered necessary that the LPRMs be included in the CPS SPDS Parameter Set (also see discussion in Section 4.0 regarding monitoring local power
. changes on SPDS).
5.5.2 Control Rod Withdrawal Error - Refueling This event involves the withdrawal of a control rod (s) that results in a local criticality.
Description of-Plant Transients:
This transient is not reported in the CPS Final Safety Analysis Report as the refueling interlocks allow only one rod to be withdrawn for service or testing if the refueling bridge is over the core or there is a fuel bundle in the hoist.
These interlocks operate if the plant mode switch is in the REFUELING position.
Automatic Actions:
Scram via the intermediate range power monitors would result if flux level rose enough.
Appropriate Operator Actions:
The operator should respond to the intermediate range power monitor or source range monitor and, due to the fact that the plant is in a refueling mode, as indicated by the mode switch position, insert or scram the control rods.
The parameters to cue the operator action under this event is the following:
Mode Switch Source Range Monitor Intermediate Range Power Monitor Of these parameters, the source ringe monitor is included in the CPS SPDS Parameter Set.
Although the position of the reactor mode switch is not monitored by the CPS SPDS Parameter Set, this is not considered significant since in the CPS Control Room the location of this switch on the Primary Plant Console is very close in proximity to the SPDS display.
Finally, monitoring the intermediate power range monitors via the SPDS for this transient is not considered since a scram would most likely occur in this range before significant operator actions could be taken.
5.5.3 Control Rod Drop Accident The control rod drop accident is the result of a postulated event in which a highest worth control rod, within the constraints of the banked position RCIS, drops from the fully inserted or intermediate position in the core.
The highest worth rod becomes decoupled from its drive mechanism.
The mechanism is withdrawn
~but the decoupled control rod is assumed to be stuck in place.
At aclater moment, the control rod suddenly falls' free and drops to the control rod drive position.- This results in the rapid removal ofilarge negative reactivity from the core and results in a localized power excursion.
Description of Plant Transient:
The transient reported in the CPS Final Safety Analysis Report results in fuel failures which occur before the local neutron flux transient is turned around by the-Doppler and void coefficients.
The worst case transient is at about 10% power when the void coefficient is not strong and there is a leakage path from fuel to the environment via the main condenser.
Automatic Action:
There are no automatic actions that are intended to prevent a rod drop.
The design does include features to limit drop velocity, but not to prevent the drop after the postulated separation occurs.
Appropriate Operator Actions:
The transient that occurs is rapid.- The operator is limited to post-event actions to reduce the. consequences of the event, such as:
Scram and isolation Reviewing status of local neutron flux
~
= C,'
\\
Controlsubse)quentreleaseofradioactivitybycorrect t
operation'of the offgas and Reactor Water Cleanup systems.
For this transient, the CPS SPDS Parameter Set provides offgas radiation monitor, average power range monitor and containment isolation groups.
This is considered sufficient.
1 5.5.4 Recirculation System Transients There are several events that can insert reactivity by misoperation of the recirculation system:
Startup of a recirculation loop while at power when the loop has cold water in it.
Failure of the valve flow control system in the open position.
Slow upward drifts in recirculation flow due to a mal-function.
All of these events' result in the insertion of reactivity due to increased subcooling.
Description of Plant Transient:
Failure of the master controller or neutron flux controller can cause an increase in the core coolant flow rate.
Failure within a loop's flow controller can also cause an increase in core coolant flow rate.
Automatic Actions:
There may or may not be a reactor scram, depending on-initial conditions and the severity of the transient.
Appropriate Operator Actions:
Initial action by the operator should include:-
(1)
Transfer flow control to manual and stabilize flow.
(2), Identify cause of failure.
Reactor pressure will be controlled as required, depending on whether a restart or cooldown is planned.
In general, the corrective action would-be to hold reactor pressure and condenser vacuum for restart after the malfunctioning flow controller has been repaired.
The following is the sequence of operator actions expected during the course of the event, assuming restart.
The operator shou).d:
(1)
Observe that all rods are in.
(2)
Check the reactor water level and maintain above low level trip to prevent MSIVs from isolating.
< 2
~
.l.
1
(
\\s
~(3)
position.-
Switch the reactor mode switch to-the "startup"
-(4)
Continue to maintain _ condenser vacuum and turbine
-seals.
(5)
Tr'ansfer the; recirculation--flow controller to the Emanual position 1and reduce set point.to zero.
-(6)~
Survey maintenance requirements and complete the
~
. scram report.
~ (7).' Monitor the turbine coastdown and auxiliary systems.
(8)'. Establish' a restart' of the reactor per the normal 1-sprocedure.
Time required from first trouble alarm to
. NOTE:
restart would be approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Since there.-are no significant consequences,.rua operator action important~to' safety is required.
The operator will be able to control;this. event by monitoring:the average power range monitor, freactor coolant (system water level, and reactor coolant system pressure..
5.5'.5'.
Anticipated Transients Without Scram J There'are many plant conditions and anticipated events that can Lresult in the:need for scram.
These are:
Low reactor water level High; reactor pressure Main' Steam Line Isolation Valve closure
. Turbine' control valve fasticlosure Turbine:stop valve closure High neutron flux High primaryfcontainment pressure-Scram discharge volume high water level
. Main' steam.line high radiation
~
4
%j.,
Condenser" low vacuum.
by
-? This ' event 'is the -f ailure 'of the control rods to fully insert if
~
"yl Lone of~the above conditions-exists..
'Pu fjp
, Description of Plant Transient:
The_ worst case situation for 7.?
.this. event occurs-when it.is initiated by isolation of the
?E
^ reactor"from-.the main condenser heat sink:and trip of steam
$hi
. driven' reactor feedwater pumps.. This particular case is similar I L?
'to that-described in this section-on " Reactor Pressurization-
$7k'k. !h, _.Events",'except that thereLis a~ higher rate of boiloff and energy
~
fgnfes-trelease to the suppression pool through the safety / relief valves
, resultingi rom higher reactor coolant system pressure and faster f
/fk?
rate ~of reacto'r' coolant system water level decrease.
up Q
gi
~_ '
eo n, m
- s
+
Automatic Actions:
The automatic actions are the same as described in the section on " Reactor Pressurization Events" except'that scram has failed to take place.
Appropriate Operator Actions:
In addition to the operator
. actions identified for " Reactor Pressurization Events", the operator should take the following action:
Attempt manual scram and control rod insertion Initiate Standby Liquid Control system (if needed based.
upon suppression pool temperature).
These steps would be motivated by high average or intermediate power range monitors failing to decline to near zero following a scram demand.
The monitoring of those parameters discussed previously for
" Reactor Pressurization Events", plus monitoring of the average power range monitor and suppression pool temperature is considered adequate for the CPS SPDS Parameter Set (note previous discussions regarding intermediate power range monitor and scram demand).
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t l5.6i COOLANTIINVENTORY1 DECREASE:
t 5.6.l' Inadvertent safety / Relief Valve Opening
. Description of Event:. This is a unique event involving the
. inadvertent opening of a. safety relief valve:(or its failure to reclose),Jcausing' reactor steam to -flow to the -suppression pool.
Description of ' Plant-Transient: ' Inadvertent opening of a-
- f safety / relief valve will result in the following
~
Local" heating of the: suppression; pool by condensation of the x
r steam flowing through the safety / relief" valve.
i
'A small decrease in reactor pressure due to-the-inadvertent 4
_. valve opening.
A decrease in turbine-generator output as the pressure l'
Jregulator responds to the pressure. transient.
F The worst case consequences of this event are the localized a
1 overheating of the suppression pool leading to highfcontainment Eloads,'possible local 7 inability to: condense the steam flow and.
' inability to locally perform pressure suppression in a subsequent
. loss of-coolant event.
v I
- Automat'ic' Actions:
There are no automatic. actions ~that will.
terminate this event.
The event must be terminated by operator I
action.
i L
- Appropriate Operator Actions:
The-operator actions in response
. to this event are to initiate reactoroshutdown and
+
-depressurization when he'ob' serves that one er more safety / relief
' valve positionisignals indicate.open but're, cor coolant system pressure _is~not high enough to require open S/RV's.: In most
~
' cases,'a stuck open valve;will reclose after some degree of
~
reactor depressurization.- If reclosure doesinot occur and
-suppression p3ol' temperature continues to' increase, the operator should initiate the suppression pool cooling mode of RHR.
L As was recommended previously underL " reactor pressurization events", SRV positions should be monitored by the CPS SPDS.
All
~
other parameters to' cue operator action for this event are n
presentlypart of the CPS SPDS' Parameter Set.
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~
5.6;2. Loss of Feedwater Flow 1
Description of Event:
This event'is a partial or complete loss
..of feedwater flow during power; operation.
The event can be
,1" caused by a feedwater-pump trip,: control valve failure, failure Lofithe'feedwater line outside..of containment, or steam supply
! shutoff to the feedwater pump turbines.
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Description of Plant Transient:
A complete loss of feedwater while at rated power will result:in a reduction in reactor coolant system water level..
- Automatic Actions-The following automatic actions are expected to occur for this class of events:
Reactor scram due to low reactor coolant system water level Recirculation pump trip Main Steam Line Isolation valve closure Initiation of HPCS and RCIC systems Appropriate Operator Actions:
The operator checks that the high
_ pressure systems are supplying makeup to the reactor.
This_would
,lme instigated by the CPS SPDS parameter reactor cooling system water level trending' low.
Therefore, the CPS SPDS Parameter Set is adequate for monitoring this event.
5.6.3 Loss of Coolant Accidents
. These events are discussed under subsection 5.2.
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'5. 7 COOLANT INVENTORY INCREASE 5.7.1 Inadvertent Operation of the High Pressure Core Spray System or Reactor Core Isolation Cooling Syste3 Description of Event:
This class of event is the operation of a high pressure water makeup system without the occurrence of low reactor coolant system water level.
Description of Plant Transient:
Operation of the Reactor Core Isolation Cooling System (RCIC) or High Pressure Core Spray System'(HPCS) will result in the maximum addition of 600 or 6400 GPM, respectively, to the reactor coolant inventory.
In the event that the feedwater three element control is functioning correctly and the plant is at sufficient power, feedwater will decrease to compensate for the additional inflow and reactor water level will stay in its normal range.
Exhaust steam from the RCIC system turbine will be dumped to the suppression pool.
In the even'. that the feedwater controller does not compensate, water level will rise causing a turbine trip and reactor scram to avoid excessive moisture carry over to the turbine.
The ensuing plant transient is similar to reactor pressurization events discussed previously.
Automatic Action:
The automatic actions were discussed above:
feedwater controller compensation, or turbine trip and scram.
In the event that.the transient is not terminated by manually securing.the injection system or by a high reactor coolant system water level trip, water level will continue to rise until the steam lines are flooded.
Appropriate Operator Actions:
The operator in this event should remote manually trip'the injection system.
Operation of the RCIC system.will cause heatup of the suppression pool, and, if continued, will result in high local temperature, requiring the l
operator to remote manually initiate suppression pool cooling.
[
Hig., reactor coolant system water level and later high suppression l
pool temperature will initiate operator diagnosis and action.
These parameters are part of the CPS SPDS Parameter Set.
i 5.7.2 Failure of Feedwater Control to Maximum Demand Description of Event:
The event is a failure in the feedwater control system that results in maximum opening of the feedwater valves'and/or maximum feedwater pump turbine speed.
Description of Plant Transient:
The transient in reactor water level is similar to the transient described in the preceding event, " Inadvertent Operation of the High Pressure Core Spray System or Reactor Core Isolation Cooling System".
Failure to secure the feedwater pumps or regain control will result in ;
'o moisture carry over to the main steam line and water and two phase flow in the safety / relief valves if a pressurization transient subsequently occurs.
Automatic Actions:
Turbine trip and consequential scram along with feedwater pump trip can be expected as a result of high water level.
-Appropriate Operator Actions:.The operator should transfer to manual feedwater control or trip the feedwater pumps from the control-room.
The CPS SPDS parameter which instigates his action is high reactor: coolant system water level..
Therefore, the CPS SPDS Parameter Set is adequate for monitoring this event.
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[5 '.' 8 DECREASE.IN REACTOR COOLANT FLOW RATE Description of Event: :This is a-class of events characterized as
- a decrease.in corezinletLflow.
Typical _ initiating events are:
Recir'culation Pump Trip.
Recirculationiflow control failure - decreasing flow Recirculation pump! seizure
-Recirculation pumprshaft break Jet pump blockage Description of Plant Transient:
All of:the above events result
-in increased-core voids, a decrease in power ~and steam flow-to-the' turbine-generator.
The CPS Final Safety Analysis Report
, indicates no. fuel ~ failure! occurs'.
A blockage of up to one half.
~
- the jet pumps would-be no.more severe.than a recirculation pump
- seizure (jet pump blockage-is not specifically addressed in the
l Automatic Actions:
For a Recirculation Pump Trip (RPT) :
The trip ofsone-recirculation pump produces a milder transient than
.does'the-simultaneous trip'of.two' recirculation pumps.
A two pump trip.results inia high-water level trip of the: main
. turbine whichufurthericauses a stop valve closure and a subsequent SCRAM actuation. -Main steamline isolation soon occurs 1
n and is followed'by RCIC/HPCS systems initiation on low-water
. level. - Relief' valve actuation will follow.
For Recirculation flow control failure:
The number.and type ofnflow controller failure modes determine the protection' sequence for1the event.
For Recirculation' pump seizure and' shaft break:
A main turbine trip will: occur as-vessel water level swell exceeds the turbine
-trip-setpoint.
This~results'in-a trip scram and a RPT when the I
turbinefstop valves'close..-Relief valve opening will occur to-
- control. vessel' pressure and temperature.. RCIC.or HPCS. systems will maintain' vessel water level. -Prolonged' isolation will-
[
. require. core-and' containment' cooling and possibly some L
radiological effluent control.
l-
? Appropriate Operator Actions:
J Trip-of-One Recirculation Pump:-
Since;no'. scram occurs'for,the trip of one recirculation pump, no immediate~ operator action is required.
As1soon as possible, the operator should verify that no operating limits are being
~
exceeded,'and reduce flow of the. operating pump to conform to the single pump flow; criteria.
'Also, the operator-should determine
- the.cause of. failure prior to returning the system to normal and follow the' restart procedure.
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Trip of'Two Recirculation Pumps:-
.The. operator Eshould ascertain that. the reactor scrams with the
~
< turbine trip resulting from. reactor water level swell.. The operator should regain control of reactor water level through
-MDRFP or RCIC operation, monitoring reactor water level and-
. pressure control after shutdown.
When both reactor pressure and
' level are under control, the. operator should secure both IIPCS and RCIC'as;necessary.
The operator should also determine the cause of the1 trip prior toLreturning the system to normal.
Fast' Closure'of One-Main' Recirculation Valve:'
'Since no scram occurs, no immediate operator action is required.
As soon as possible, the operator should verify-that no operating
' limits are being exceeded..The operator should determine the c'ause of failure prior to. returning the system to normal.
Fast Closure'of Two Main Recirculation Valves:
As soon'as possible, the operator must verify that no operating limits are being exceeded.
If they are, corrective actions must bel initiated..- Also, the operator must determine the cause of the trip prior to returning the system to normal.
- Recirculation pump seizure and shaft break:
.The operator should ascertain that the. reactor scrams resulting from reactor water level swell.
The operator should regain control of1 reactor water level through RCIC operation or by
- restart of a feedwater pump; and he should monitor reactor water-level ~and pressure control after shutdown.
The operator would be cued by reduced core flow, reduced reactor
-vessel water level and average power range monitor signals.
The CPS SPDS Parameter Set is considered adequate for monitoring this event.
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.5.9 FAILURE TO' REMOVE RESIDUAL CORE HEAT Description'of Event:
This event is initiated by loss of'offsite power which results in a loss of main generator load, followed by a failure of all emergency diesel generators to activate the emergency buses.
Alternately, a faildre of all residual heat removal pumps or all service water (SX) pumps is postulated instead of failure of the emergency diesels.
Description of Plant Transient:
Loss of_offsite power will result in-loss of the power conversion system (main condenser) heat sink, reactor scram, containment isolation and opening of one or more safety / relief valves to discharge reactor steam to the. suppression pool.
The reactor core isolation cooling (RCIC) system will be initiated to' maintain water level as they require station battery power only.
Failure of all residual heat removal or all SX pumps or all i
Jemergency buses will result in the inability to remove heat from the suppression pool.
Over a period of hours, continued steam 4
discharge will raise suppression pool temperature and hence drywell pressure until over pressure failure of the primary containment occurs.
Transfer of suction to the suppression pool occurs when the initial source, RCIC/HPCS. storage tank, is exhausted.
Without high pressure makeup flow, reactor coolant system water level will fall uncovering the core and leading to fuel cladding failure as the core heats up.and ultimately melts.
Fission gas released would escape through the-safety / relief valves and through the failed-primary containment to the environment.
Automatic Actions:
There are.no automatic actions-which establish heat removal-from the suppression pool or reactor coolant. system.
l
' Appropriate Operator Actions:
Initially this event-is a L
" pressurization event" and " loss of feedwater event" and operator l
actions for these events are appropriate.
If pump failures are the reason, but cannot be recovered, rising suppression pool c
temperature and drywell pressure will advise the operator as to L
the time available for remedial actions.
'If. loss of all~ emergency bus AC power is postulated, the CPS SPDS will be powered.via an Uninterrupted Power Supply (UPS) system.
~
This UPS system can be maintained for ninety minutes.
Operator action should be~ directed toward re-establishing power to at least one emergency AC bus.
The CPS SPDS Parameter Set is considered sufficient to monitor and respond to this event, at least as long as UPS power is available. -
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5.10 -LOSS OF AUXILIARY SYSTEM POWER 5.10.1 ' Loss of Auxiliary Power Without-Loss of an Emergency Bus i
Description of Event:
This event is the loss of auxiliary power due to reasons internal or external to the plant.
The emergency i
power buses are normally powered from the plant auxiliary.
I transformer.
The emergency buses are alternately powered from the emergency diesel generators which automatically start on low bus voltage.
Description of Plant Transients:
Loss of auxiliary power will cause loss of power production components such as condensate pumps, turbine building auxiliaries, etc.
If a turbine-generator trip is not the initiating event, the loss of the auxiliary bus will result in turbine-generator trip, as well as loss of feedwater.
The transient becomes a " pressurization event" and a
" loss of feedwater event" as discussed before.
The loss of auxiliary power will result in loss of all AC power for approximately 10 seconds while the emergency diesels are started.
Automatic Actions:
In addition to the automatic actions resulting for a " pressurization event" and a " loss of feedwater event", the emergency diesels should automatically start on loss of emergency bus voltage.
Appropriate Operator Actions:
Operator actions are the same as those for a " pressurization event" and a " loss of feedwater event".
5.10.2 Loss of Auxiliary Power with Concurrent Loss of an Emergency Bus Description of Event:
This event is the loss of auxiliary power due to reasons internal or external to the plant.
The emergency power buses are normally powered from the plant auxiliary transformer and alternately powered from the emergency diesel generators which automatically start on low bus voltage.
Description of Plant Transient:
Loss of auxiliary power will cause loss of power production components such as condensate pumps, turbine building. auxiliaries, etc.
If turbine-generator trip is not the initiating event, the loss of the auxiliary bus will result in turbine-generator trip, as well as loss of feedwater.
The transient becomes a " pressurization event" and a
" loss of feedwater event".
The loss of auxiliary power will result in' loss of all AC power for approximately 10 seconds while the emergency diesels are started.
Failure of a diesel to start or-a circuit breaker to close results in loss of 1 or 2 emergency buses.
The loss of both emergency buses was addressed in the failure to remove residual core heat event. -
or.
e, Automatic Actions:. The automatic actions described in the
" pressurization-event" and the " loss-of feedwater event" will take place as well asithe start of the emergency diesel generators.. If an emergency bus' fails to-recover voltage, no further automatic action occurs.
'Aspropriate O?erator Action:
Operator actions.are-the same as in the case of the " pressurization event" and the " loss of feedwater
. event".
'The CPS SPDS Parameter Set is considered adequate for these
. events.
5.10.31_ Loss'of Instrument Air Description of Event:
Air is:provided for operation of instrumentation and certain valves.
With the exception of safety / relief valves in-the,BWR:6 class of plant, these are not considered essential to.the safety of the plant.
Description of Plant Transient:
Loss of instrument air will most
.likely. result in a1 plant upset-condition leading to reactor. scram and: loss of.feedwater flow.
For example, the plant employs. air operated feedwater control valves.
Also, loss of air pressure-will cause.the' main steam isolation valves to close, as they are
~
air-to-open, spring-to-close valves.. Thus, loss of instrument air will cause a " pressurization event" and could.cause a " loss of feedwater flow event"-as previously discussed.
-The following conditions will also result from loss of instrument air.-
Drywell and containment' cooling system valves and dampers williclose,, fuel pool; cooling and makeup valves will close'and HVAC dampers.will close for ECCS areas and control ~ room.
Automatic Actions:
Automatic actions will be similar to those of the pressurization event and the loss of feedwater event.
In
-addition, high~drywell temperature / pressure and subsequent reactor 1 scram would' occur due to loss of drywell cooling.
A)propriate Operator Action:
Operator actions are the same as those in the pressurization event and the loss of feedwater event.
,The CPS SPDS Parameter Set is adequate for these events. _ _ _
4
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, 4
?W 5.11 1 RADIOACTIVE RELEASE FROM' SUBSYSTEMS AND COMPONENTS 5.11.1_ Radioactive Gas Waste (Offgas) System Failure 1.
LThe' radioactive gas waste systemiconsists of components that remove air, hydrogen and: fission; gas from the main condenser, recombines the oxygen and hydrogen, and' delays' release of the fission products-by. holdup in charcoal beds at sub-ambient
- temperatures. 'This class of event is any system or component failure that1results in early fission gas. release or release from other-than the designated 1 release point, the common HVAC stack.
LPractically all of the system components are located in the turbine building.
Description of Plant' Transient:
A release of offgas prior to holdup in the charcoal can result in release of radioactivity into the turbine building and into the atmosphere through the
-turbine. building ventilation exhaust.
LBreakdown of.the retention ~ capability of the charcoal, for, example by moisture buildup, will result in increased release rate through the common HVAC stack.
A failurelin components downstream of the charcoal will change the release elevation but not the release rate.
. Automatic Actions:
The post-treatment air ejector offgas radiation. monitors upon detection of-an alert radiation level will-shut the' carbon absorber bed bypass ~ valves sending offgas to
'the. carbon absorber beds.
Upon detection of a high-radiation level-the air ejector offgas treatment. system is isolated from the' common HVAC stack. causing a loss of condenser vacuum, turbine itrip and subsequently reactor. scram.
Appropriate Operator Actions:- There are several potential causes of of fgas s'/Etem f ailure and increased. release and each will require different operator action..Some.of these causes are wet l
-charcoal, explosion of H,.and. loss of charcoal vault cooling AC power.--. The parameter-set to alert the operator to investigate and'take. corrective action are the: post-treatment-monitors, the HVAC' stack monitors, pre-treatment monitor, and plant continuous air monitors..
Since the CPS SPDS~ monitors these variables, this is considered
-sufficient.
Liquid Radwaste System Failure 5.11.2 1
The radwaste. system collects all potentially radioactive liquids from sources such as floor and. equipment drains, decontamination
_ pits,. laboratories and-laundries.
Also processed are resins and j
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' filter aid from the Reactor Water Cleanup System, Fuel Pool-Cooling and Cleanup. System, condensate and Radwaste demineralizers.
The radwaste system filters, evaporators, and demineralizers process the collected liquids and return the water to the
-condensate storage tank or discharge 'taa the environs after
. dilution with plant service water.
The radwaste system dewaters the wet solids and packages the residue in 55 gallon drums.in a mixture.of concrete.
Description of Plant Transient:
The radwaste system is located and designed so as to preclude the release of liquid waste being stored in tanks to areas outside of the radwaste area.
The 55 gallon drums are stored on site or shipped to a waste disposal site.
The waste stored in the drum is low level and immobile.
It is possible to release excessive amounts of activity to the environment by mixing of waste water into the. plant service water system.
Reaction transients do not directly affect plant discharges from radwaste since radwaste is a-batch release process.
Automatic Actions:
All control of the radwaste is'from the Radwaste Operations Center in'the Radwaste Building.
A monitor on the waste discharge line will alarm an excessive activity concentration and automatically stop the discharge.
~ The liquid process monitor will
-Appropriate Operator Actions:
alert the operator to take action.
Therefore, the CPS SPDS Parameter Set is adequate-for these types of events.
a 0
. 5.12 WATER CHEMISTRY TRANSIENTS
- During power operation, the BWR is continuously supplied water
. from the condensate /feedwater system at rates equal to-the steam generation rate.
The feed / boil-off process concentrates impurities such as chlorides, which can be in the feedwater.
The Reactor Water Cleanup System is provided to maintain water chemistry at acceptable levels.
Description of Plant Transients:
Large amounts of impurities, especially chlorides, can enter the reactor coolant in the event of a condenser tube failure, a failure of the condensate demineralizer_ system to remove normal impurities, or a failure of
. the Reactor Water Cleanup System to function properly.
Automatic Actions:
There are no automatic actions to prevent the buildup of reactor water impurities beyond those described above.
Appropriate Operator Actions:
The appropriate operator actions
- are to shut down for condenser repair, to correct operation of the condensate demineralizers or to correct operation of the Reactor Water Cleanup System, depending on the nature of the problem.
Because water chemistry transients are not directly safety related as they result in long term discharge, they are not considered necessary to be monitored by the CPS SPDS.
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- ?.'
5.13 RESULTS OF TRANSIENT AND ACCIDENT ANALYSIS As a result of the review of the transients and accidents identified in this report, it is recommended that some
' changes be made to the existing CPS SPDS Parameter Set.
The l
. recommended changes identified earlier in this report include the addition of Secondary Containment Pressure and SRV Position Status to the Parameter Set.
These proposed changes would enhance the ability of the SPDS to monitor the safety' functions of Reactor Coolant System Integrity and Secondary Containment Integrity.
These changes, along with other recommendations, are discussed in more detail in the following comparison with other BWR accident monitoring parameter lists and in Section 6.0 entitled " Conclusions and Recommendations."
7 s
5.14 CPS SPDS PARAMETER SET COMPARISON WITH OTHER PARAMETER LISTS FOR BWR ACCIDENT MONITORING 5.14.1 Comparison with NSAC/21, " Fundamental Safety Parameter Set for Boiling Water Reactors" NSAC/21 describes the selection and validation of a set of fundamental safety parameters that is considered to provide an overview of the status of plant safety for a boiling water reactor..The generic fundamental'BWR safety parameter set
' recommended by NSAC/21 is shown in Table 5-3.
A comparison of this table with the CPS SPDS Parameter Set shown in Table.5-2 indicates that the.following parameters recommended by NSAC/21 are not monitored by the CPS SPDS:
cl.
Primary Coolant System Activity 2.
Safety / Relief Valve Positions 3.
Leakage Isolation Demand 4.
Secondary Containment Pressure 5.
Source Range Monitor Position 6.
Scram Demand Signal 7.
Mode Switch Position For the reasons noted-earlier in Sections 4.0 and 5.0 of this report, it is recommended that of these seven parameters, provisions should be made to include Safety / Relief Valve Position and Secondary Containment Pressure in the CPS SPDS Parameter Set.
It might be noted that the TMI Action Plan (NUREG - 0737) requires the incorporation of SRV position indications in the control room.
Long-term SRV discharge would be evident from iincreasing suppression pool temperature which is in the CPS SPDS Parameter Set and would therefore trigger responses from operations personnel.
However, SRV Position indication via SPDS would provide for a more rapid operator assessment of such events.~ In addition, there is little action that can be taken to close a stuck-open SRV.
It therefore. appears that SRV position indication may lack sufficient safety significance for inclusion in the SPDS Parameter Set.
However, the primary consideration here is the assignment of top level significance to such.a breach of the reactor coolant system integrity.
For this reason, it is recommended that SRV positions be monitored by the SPDS.
It is not considered necessary that this parameter be monitored such that the positions of each of the 16 SRVs be displayed as open/ closed on the SPDS.
It is recommended that the 16 SRV position' signals be combined as a single "SRV Position Status" signal and included in the CPS SPDS Parameter Set as part of the Alarm Initiated Display.
" Secondary Containment Pressure" should be implemented into the SPDS Parameter Set as part of the Permanent SPDS Display.
. L
i s-From the comparison between Tables 5-2 and 5-3, it is also noted that the following parameters displayed by-the CPS SPDS Parameter Set ~are not included in the NSAC/21 fundamental set:
1.
Rx Steam Flow 2.
Rx Feed Flow 3.
Rx Recirculation Flows (A & B loops) 4.
Containment Isolation Groups 5.
Drywell Equipment Sump Flow 6.
Scram Discharge Volume Level (A & B)-
Of these parameters, it is recommended that Rx Feed Flow, Rx Recirculation Flow and Drywell Equipment Sump Flow be deleted from the CPS SPDS Parameter Set.
.With regards to reactor core cooling and heat removal safety functions, the monitoring of Total Core Flow is considered adequate and is also recommended by NSAC/21.
Monitoring of the additional variables Reactor Feed Flow and Reactor Recirculation Flow lacks sufficient safety significance to be included in the Permanent SPDS Displav.
Feed flow and recirculation flows are more pertinent to the monitoring of normal power operations and, as can be seen from reviewing the accident analyses performed in this report, the symptoms provided by these parameters are not significant enough to be included in the SPDS Parameter Set.
The drywell equipment drain sump has been provided in the plant as a means of separating small anticipated leaks in the reactor coolant system from unanticipated leaks which are detected by the drywell floor drain sump.
The expected leaks to the equipment drain sump come from recirculation pump seals and valve stem leak-off pipes.
These leaks are expected, normal, and therefore do not constitute a safety concern.
Further, these leaks are self-limiting by the physical configuration of the packing systems.
Thus, Drywell Equipment Sump Flow should be eliminated from the CPS SPDS Parameter Set.
A review of the accident analyses performed in this report also supports this conclusion.
5.14.2 Comparison with NUREG/CR-1440 BWR Accident Monitoring Variables In NUREG/CR-1440 a novel technical approach for systematically determining operator information needs during' reactor accidents
- ~
is proposed.
Various risk-significant accident sequences are analyzed to identify a set of light water reactor instrumentation needed to analyze the appropriate operator response to various l
plant states.
The BWR accident monitoring variables identified i
in this study are shown in Table 5-4.
A comparison of this table with the CPS SPDS Parameter Set shown-in Table 5-2 indicates that the following parameters recommended y NUREG/CR-1440 are not monitored by the CPS SPDS:
1.
Control Rod Position 2.
RCS Temperature 4
o t
3.
Main Steam Flow Isolation Valve Position 4.
Safety / Relief Valve Positions in Primary Systems (including ADS) 5.
Radiation Level in Coolant 6.
Boron Tank Level-7.
SLCS Flow or Pump _ Discharge Pressure 8.
Boron Concentration 9.
Feedwater Pump Discharge Pressure, Current to Pumps, or Controller Position 10.
Steam Flow to RCIC Turbine 11.
RCIC Flow or Pump Discharge Pressure 12.
HPCS Valve Positions 13.. HPCS Flow, Pump Discharge Pressure, or Current to Pumps 14. - RHR Valve Position 15.
RHR Heat Exchanger Inlet / Outlet Temperature 16.
HPSW Valve Position 17.
HPSW Flow or Pump Discharge Pressure It is not recommended that control rod positions be included in the CPS SPDS Parameter Set. -The primary reason for this recommendation is that, while an indication of all rods inserted
.would be strong assurance of subcriticality, an indication of partial insertion would be indeterminate.
Further, if there were no rods inserted, the standby liquid control system would be ultimately employed. -In both of these situations, the neutron flux.would be monitored to confirm criticality.
Thus, neutron flux, as measured by the APRMs, is the fundamental parameter not control rod position.
In addition, control rod positions are indicated on the Principal Plant Console just to the right of the
. Safety / Relief Valve position indication has been recommended previously for the SPDS.
Radiation Level in Coolant has been discussed previously and is not recommended.
The-remaining parameters listed above from NUREG/CR-1440 are not recommended for the CPS SPDS Parameter Set since, as can be seen from the accident analyses, they are not considered significant enough-to provide an overview of the plant safety status to the
. control room operators.
Many of these parameters are monitored by the EOP Support Displays of the plant process computer system as well as various analog indications in the Main Control Room.
5.14.3 Comparison with Regulatory Guide 1.97 BWR Parameter List Regulatory Guide 1.97 describes a method acceptable to the NRC
-Staff for complying with the Commission's regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant.
R.G.
1.97 states that
"... Indications of plant variables are required by the control room operating personnel during accident situations to (1) provide information required to permit the operator to_take preplanned manual actions to accomplish safe plant
- hutdown; (2) determine whether the reactor trip, engineered s
safety-feature systems, and manually initiated safety systems and other systems important to safety are performing
-their intended functions (i.e., reactivity control, core cooling, maintaining reactor. coolant system integrity, and maintaining containment integrity); and (3) provide information to the operators that will enable them to determine the potential for causing a gross breach of the barriers to radioactivity release (i.e., fuel cladding, reactor coolant pressure boundary, and containment) and to determine if a gross breach of a barrier has occurred.
In addition to the above, indications of plant variables that provide information on operation of plant safety systems and other systems important to safety are required by the control room operating personnel during an accident to (1) furnish data regarding the. operation of plant systems in order that the operator can make appropriate decisions as to their use and (2) provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat."
-A comparison of the R.G. 1.97 BWR Parameter List (Table 5-5) with
-the CPS SPDS Parameter. Set (Table 5-2) indicates that many of the parameters identified by R.G.
1.97 are not part of the CPS SPDS Parameter Set (see Table 5-5).
Those parameters not included in the CPS SPDS Parameter Set are monitored by.a number of other means in the Main Control Room (i.e. analog indication, EOP Support Displays) and are not considered of the " safety status overview" type to warrant further consideration.
5.14.4 Comparison with BWR Emergency Procedure Guideline Entry Condition Parameter Set As a result of the TMI accident and the lessons learned by analyzing the accident, emergency procedure guidelines have been under development.
These guidelines are intended to provide the centrol room staff with recommended procedures for dealing with emergency conditions (both design basis and beyond design basis events).
An important element in these procedures is the identification of those conditions which necessitate initiation of an emergency procedure.
These initiating conditions termed
" entry conditions," are, by their nature, indicators of the safety state of the plant. L
- o t
The guides which have been prepared are the RPV Control Guideline, Primary Containment Control Guideline, Secondary
-Containment Control Guideline, and the Radioactivity Release Control Guideline.
The entry conditions for each of these guides is shown in Table 5-6.
For the RPV Control Guideline:
The conditions requiring entry to the RPV Control Guideline are symptoms indicative of a loss of reactor coolant or a failure-to-scram.
Low water-level, high drywell pressure and an RPV isolation are indicative of an existing or impending coolant inventory loss.
Reactor power above the APRM downscale setpoint following reception of a scram trip is-indicative of a failure-to-scram.
The low RPV water level scram setpoint is below the levels normally attained during operational transients.
A lower water level therefore suggests that core cooling may be jeopardized if additional makeup is not provided.
Above this elevation, normal operating procedures specify appropriate methods for control of RPV water level.
High reactor pressure threatens the integrity of the reactor coolant pressure boundary.
Additionally, a pressure increase during power operation will collapse the steam voids and add positive reactivity to the core.. This reactivity addition causes increased core heat generation which can lead to fuel damage and/or system over-pressurization.
Vessel steam dome pressure is utilized and allowances are made for the fact that this is not the point of highest system pressure.
Scramming the reactor
~
under this condition counteracts the system pressure increase by reducing core heat generation.
The high reactor pressure scram setpoint'is. chosen high enough above normal operating pressure to preclude spurious scrams during normal plant operations, yet low enough to provide an adequate margin to the maximum allowable system' pressure.- This scram, in conjunction with the safety / relief valves, prevents system overpressurization.
It also provides a secondary level of protection to thermal hydraulic limits which-can be approached should significant pressure increases occur during reactor operation.
i.
A high drywell pressure may be caused by a loss of coolant through-a pipe break inside the containment, and is thus symptomatic of a level control problem.
The scram setpoint was
~
chosen-to:
(1) provide an easily recognized action level; (2)
- avoid unnecessary or premature entry to emergency procedures; (3) distinguish _between pipe breaks and small leaks; (4) be consistent with the Containment Control entry conditions.
' Actuation of containment isolation logic is also indicative of a coolant inventory loss, either inside or outside the containment.
The entry condition is not limited only to those isolations -
. ~.
4 4*
-directly1resultingfin a scram, but also includes ~isolations
" requiring"fafscram..A reactor scram may be desirable, for example,Lto terminate' containment pressurization or limit off-site release, even;if;not initiated by the RPS logic.
1
-The'APRM downscale trip'was selected as an action level because cit is an easily monitored parameter applicable in all scenarios andi.is:approximately equivalent to the normal decay heat rate.
b Rod-position cannot practically be used, since-a large number of i
rod patterns are possible and the reactivity effects of each
- cannot be easily predicted.. Additionally, incomplete scrams are expected:to cause complications onlyjif a significant power level
_ exists.,Below the AEEM'downscale setpoint, plant response will
'be similar'to that observed.during a normal shutdown and entry 1
Linto the emergency-procedures-is-unnecessary.
' For7the Primary Containment Control Guideline:
.' The. CPS-MK III containment is.offthe " pressure suppression" type, the suppression pool functioning as an emergency heat sink during
-a LOCA'or ADS operation to absorb the energy contained in the
' RPV.
High suppression gool temperatures reduce the available
- heat capacity,. reduce the.NPSH of pumps drawing suction from the pool,-Jincrease the vapor pressuretin the suppression. chamber,
- affect containment loads during SRV. actuation, and reduce
. operating margins.
High temperatures could result from a-LOCA, SRV actuntion, or RCIC operation.
l High drywell temperatures'could-be_ caused by partial or complete loss:of drywell cooling capability, by sustained SRV operation,
- or by'a-pipe br'ak or leak inside the drywell.
Adverse effects e
include-.inac' curate RPV water level ~ indications and possible i
equipment'or structural damage.
Drywell temperature'also affects i the~drywell: atmospheric pressure.. Deleterious _ effects of high containment-temperatures are similar-to those of high drywell-
+
temperatures.
High drywell pressure challenges containment integrity through direct failure,.possibly resulting-in release of radioactivity to:
-the environment.
The entry condition is selected consistent with the related-RPV Control entry condition.
Both guidelines would thus.be entered concurrently.
[I' Both high and low suppression pool water levels require entry to the2 Containment Control-Guideline.
Low levels reduce the 2
suppression pool heat-capacity and may result in exposure of drywell vents and pump suction strainers.
High levels reduce the suppression chamber. volume, thereby increasing the suppression
- chamber-pressure, increase static and dynamic-loads'in the containment.
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- The= buildup andtsubsequent burning of hydrogen inside the primary-
-containment.can threaten vital equipment and the integrity of the
-primary 1 containment.'-Therefore, hydrogen concentration is an
-_ entry _ condition to_the Primary Containment Control Guidel3'...
For the Secondary Containment Control Guideline:-
1These entry conditions are-chosenJto alert the operators to
~
Leonditions such that actions can be taken to protect equipment, limit radioactivity, release to secondary containment,-maintain secondary containment integrity, and limit the radioactivity
~ release from.the secondary containment.
.For the Radioactivity Release Control Guideline:
The high radioactivity release rate. (offsite) entry condition alerts.the" operators to take actions that will limit
' radioactivity releaselinto' areas outside the primary and
~
A comparison.of-the BWR EPG Entry Conditions (Table 5-6) with the CPS SPDS Parameter Set.(Table 5-2) indicates that.the following entryiconditions are not monitored by.the CPS.SPDS:
1.-
RPV' Control Guideline Condition requiring MSIV
_ isolation.
2._
Secondary Containment Control Guideline High Area Differential Pressure High Area Differential Temperature High HVAC Cooler Differential. Temperature High Floor-Drain Sump Water Level High Area Water Level
^A condition requiring an-MSIV. isolation will most likely lead-to a reactor scram.- Therefore,.this need not be monitored by SPDS.
1Withithe additional monitoring of Secondary Containment Pressure
. recommended earlier, the CPS SPDS Parameter. Set is considered adequate _for purposes of assuring Secondary Containment integrity.. The other parameters listed hereLare not of immediate safety 7 significance as supported by;the accident analyses.
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Table 5-1 Events Reviewed for Validation'of'the CPS SPDS Parameter Set-
-Event Source LLoss of Coolant Accidents Large Breaks Inside Primary Containment-CPS FSAR/ WASH 1400-Instrument Line-Pipe Breaks Inside CPS FSAR/ WASH 1400 Containment Large Breaks Outside Primary Contain-CPS FSAR ment Small Breaks Outside of Containment CPS FSAR Reactor Pressurization Events CPS FSAR Moderator Temperature Decrease. Events CPS FSAR Reactivity and Power Distribution Anamolies
. Control Rod Withdrawal Errors at Power.
CPS FSAR Control Rod Withdrawal Error - Refueling CPS FSAR Control Rod Drop Accident CPS.FSAR Recirculation System Transients CPS FSAR Anticipated Transient Without Scram
' CPS FSAR/ WASH 1400' Coolant' Inventory Decrease Inadvertent Safety / Relief Valve Opening CPS FSAR/ WASH 1400 Loss of Feedwater Flow CPS FSAR/ WASH 1400 Co'olant Inventory Increase Inadvertent Operation of the High AUTHORS Pressure Core Spray System or Reactor Core Isolation Cooling System Failure of Feedwater Control to-Max.
CPS FSAR Demand Decrease.in Reactor Coolant Flow Rate
. Decrease in Core Coolant Flow CPS FSAR Source Legend CPS FSAR - Clinton' Power Station, Final Safety Analysis Report WASH 1400 - NRC~ Reactor Safety Study Table 5-1 (Continued)
Events Reviewed for Validation of the CPS SPDS Parameter Set Event Source Failure to Remove Residual Core Heat CPS FSAR/ WASH 1400 Loss of' Auxiliary System Power Loss of Auxiliary Power Without Loss CPS FSAR of an Emergency Bus Loss of Auxiliary Power With Concurrent CPS FSAR Loss of an Emergency Bus Loss of Instrument Air CPS FSAR Radioactive Release from Subsystems and Components Radioactive gas waste (offgas) system CPS FSAR failure Liquid radwaste system failure CPS FSAR Water Chemistry Transients AUTHORS s
Source Legend CPS FSAR - Clinton Power Station, Final Safety Analysis Report
' WASH 1400 - NRC Reactor Safety Study -
L Table 5-2 CLINTON PO'WER STATION SAFETY. PARAMETER DISPLAY SYSTEM PARAMETER SET PERMANENTLY DISPLAYED PARAMETERS Safety Function Parameters Reactivity Control APRM (Neutron Flux) 9 SRM (Neutron Flux)
Reactor Core. Cooling /
Wide Range Reactor Water Heat Removal Level Reactor Steam Flow Reactor Feed Flow Total Core Flow Reactor Coolant System' Integrity Reactor Steam Flow Reactor Feed Flow Total Core Flow Reactor Recirculation Flow (A&B)
Reactor Pressure (wide range)
Drywell Floor Drain Sump Flow-Drywell Pressure
-Radioactivity Control (ARM /PRM 46 Fixed-ARMS (throughout Status Grid)
Plant) 12 Portable ARMS 14 Fixed CAMS (throughout Plant) 10 Portable CAMS 1 PRM (common station HVAC exhaust) 1 PRM (Standby _ Gas Treatment System) 2 PRMs (one each in Pre-and Post-treatment Air Ejector Of f-gas) 1 PRM (Liquid Radwaste Effluent Discharge) 6 PRMs (various liquid streams) 16 PRMs (monitor various building HVAC exhausts)
Containment Integrity Containment Pressure (narrow range)
Containment Group
(
Isolation (1-11) t--
}
t j
Table 5-2 (Continued)
CLINTON POWER STATION SAFETY PARAMETER DISPLAY SYSTEM PARAMETER SET ALARM INITIATED DISPLAY (AID) PARAMETERS Safety Function Parameters Reactivity Control Scram Discharge Volume (levels A&B)
Reactor Core Cooling /
Reactor Water Level Heat Removal Reactor Coolant.
Drywell Floor Drain Sump System Integrity Flow Drywell Equipment Drain Sump Flow Radioactivity Control N/A (ARM /PRM Status Grid)
Containment-Integrity Drywell Pressure Drywell Temperature Containment Pressure Containment Temperature Suppression Pool Water Level Suppression Pool Temperature Containment Hydrogen Concentration -
m-
t TABLE 5-3 NSAC GENERIC FUNDAMENTAL BWR SAFETY PARAMETER SET
- Plant Ventilation Monitors
- Main Stack Monitor Primary Coolant System Activity
- Primary Containment Activity
- Drywell Floor Drain Sump
- Drywell Pressure
- Primary Coolant System (Reactor Vessel) Pressure
- Primary Coolant System Water Level Safety Relief Valve Positions Leakage Isolation Demand
- Suppression Pool Level
- Hydrogen Concentration (if not inerted)
Secondary Containment Pressure
- Drywell Temperature
'* Average Power Range Monitor
- Core Flow
- Suppression Pool Temperature
- Source Range Monitor Source Range Monitor Position Scram Demand Signal Mode Switch Position
- e-l, TABLE 5-4 NUREG/CR-1440 BWR - ACCIDENT MONITORING VARIABLES Control. Rod Position
- Neutron Flux
- RCS Pressure
~RCS Temperature
- Vessel Water Level Main Steam Flow Isolation Valve Position
. Safety / Relief Valve Positions in Primary Systems (including ADS)
Radiation Level in Coolant
- Containment-Pressure
- Containment Temperature
- Containment Radiation Level
- Suppression Pool Level
~ Boron Concentration
- Feedwater Flow Feedwater Pump Discharge Pressure, Current to Pumps, or Controller Position Steam Flow to RCIC Turbine RCIC Flow or Pump Discharge Pressure HPCS Valve Positions HPCS Flow, Pump Discharge Pressure, or Current to Pumps RHR Valve Position RHR Heat-Exchanger Inlet / Outlet Temperature HPSW Valve Position HPSU Flow or Pump Discharge Pressure
- The " asterisk" means that the same parameter is included in the CPS.SPDS-Parameter Set.
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%4 Table 5-5 sM D..
REG GUIDE 1.97 BWR PARAMETER LIST.
Jh
~7
,, "* Neutron Flux' Control Rod' Positions
- RCS' Soluble Boron Concentration (Sample) y,7
. JCoolant Level in Reactor
' Q " i BWR Core Thermocouples
- RCS. Pressure d-
'C
- Drywell Pressure-
- Drywell: Sump Level-L %
'* Primary Containment Pressure ~
i
- Primary Containment Isolation Valve' Position Qya? ?
Primary-Coolant Radioactivity Concentration or Level g' l"/6 Analysis of. Primary-Coolant (Gamma Spectrum) fjf
- Primary.Conta'inment. Area Radiation Jos.
- Suppression Pool 1 Water Level'
,* Containment and Drywell Hydrogen Concentration Containment and Drywell Oxygen. Concentration ~(inerted
' containment)'
- Contaihnient; Effluent Noble Gases Radioactivity L,ut,,
' /1" -
.; Area ~ Building Interior Radiation Exposure Rates
- Mai'n Feedwater Flow.
Condensate: Storage Tank Level Suppression Chamber. Spray Flow
_
- Suppression-Pool Water Temperature
- Drywell Atmosphere Temperature-
-Drywell Spray Flow.
-MSIV Leakage Control System Pressure Primary' System SRV Positions,': incl. Flow Through or Pressure in-9?*
IValve-Lines
- Isolation. condenser
- System Shell-Side Water Level Isolation Condenser: System -Valve Position RCIC Flow HPCI Flow'
. Core. Spray System Flow e
LPCI' System Flow
-,j' SLCS-Flow.
2
,1 SLCS' Storage Tank Level
- RHR System Flow.
CRHR Heat Exchanger Outlet Temperature
. Cooling. Water Temperature to ESF System Components Cooling Water / Flow to ESF System Components Emergency' Ventilation Damper Position
' Status of Standby Power-and Other Energy' Sources Important to Safety Reactor Building or-Secondary Containment-Area Radia-tion
- Noble Gases'and. Vent Flow.Ra+e for:
DrywellFPurge, Standby GasLTreatment-System (Mark I'and II
... plants). or Secondary Containment Purge (Mark:III plants)
Secondary-Containment. Purge (Mark I, II and III plants)
- Secondary containment (reactor: shield building annulus, if in design).
A.-
- o Table 5-5 (Continued)
Auxiliary Building Common Plant Vent or Multipurpose Vent Discharging Any of Above.
Releases ~(if drywell or SGTS purge is included)
All Other Identified Release Points Particulates'and Halogens for All Identified Plant Release Points, Plant and Environs:
Radiation Exposure Meters
~
-Airborne Radiohalogens and Particulates
, Plant and Environs Radiation Plant'and Environs Radioactivity Wind Direction
-Wind Speed Estimation of Atmospheric Stability High Radioactivity Liquid Tank Level Accident-Sampling Capability - Primary Coolant and Sump Accident Sampling Capability - Containment Air The " asterisk".means that the same parameter is included in the
)
'o
.\\
I Table 5-6 BWR EPG ENTRY CONDITION PARAMETER SET l
l EPG ENTRY CONDITION RPV Control Guideline
- Low RPV Water Level
- High RPV Pressure
- High Drywell Pressure Condition Requiring MSIV Isolation
- Condition Requiring Reactor Scram and Power Above APRM Downscale Trip Primary Containment
- High Suppresr'on Pool Temperature j
Control Guideline
- High Drywell
'mperature
- High Containms.
Temperature
- High Drywell P. assure
- High Suppression Pool Water Level
- Low Suppression Pool Water Level
- High Primary Containment Hydrogen Concentration Secondary Containment High Area Differential Pressure Control Guideline High Area Temperature High HVAC Cooler Differential Temperature
- High HVAC Exhaust Radiation Level
- High Area Radiation Level High Floor Drain Sump Water Level High Area Water Level Radioactivity Release
- High Offsite Radioactivity Control Guideline Release Rate The " asterisk" means that this parameter is included in the CPS SPDS Parameter Set.
Under RPV Control Guideline, the CPS SPDS monitors APRMs..
o a
o Section 6 CONCLUSIONS AND RECOMMENDATIONS As a result of the reviews and validation procedures implemented by the CPS V&V Team, the CPS SPDS Parameter Set is considered acuepcable, provided adequate consideration is given to the recommended changes contained herein, as a means of providing the Clinton Power Station operators with a concise overview of the safety status of the plant.
The following recommendations were discussed within this report and are summarized below:
1.
Additions to the CPS SPDS Parameter Set (a)
Secondary Containment Pressure - suggested that this be continuously monitored by the Permanent SPDS Display.
(b)
SRV Position Status - suggested that the 16 SRV position signals be combined into a single
" status" signal and included in the SPDS AID display.
2.
Deletions from the CPS SPDS Parameter Set (a)
Reactor Feed Flow - delete from Permanent SPDS Display.
(b)
Reactor Recirculation Flows - delete from Permanent SPDS Display.
(c)
Drywell Eauipment Sump Flow - delete from SPDS AID display.
In addition to these recommendations, the following change in the present CPS SPDS display is recommended:
The accident analyses evaluated indicate a significant level of importance with respect to monitoring average suppression pool temperature.
This parameter was a significant cue for operator actions in such events as the large break LOCA, reactor pressurization scenarios, ATWS, inadvertent SRV opening, coolant inventory increase, and failure to remove residual core heat.
The present CPS SPDS Parameter Set provides for monitoring of the average suppression pool temperature via the AID display.
Due to the significance of average suppression pool temperature noted in this report, it is recommended that this parameter also be displayed continuously on the Permanent SPDS Display. L
lV,
- e.,s.
4 With these recommendations considered, the CPS SPDS V&V Team considers the Clinton SPDS Parameter Set _to provide a rapid-overall' assessment of those parameters determined to be of significance to the general safety status of the plant.
e 1 t.