ML20084F592
| ML20084F592 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/30/1984 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20084F570 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 GL-82-33, NUDOCS 8405040246 | |
| Download: ML20084F592 (27) | |
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ARKANSAS POWER S LIGHT COMPANY ARKANSAS NUCLEAR ONE UNIT E SAFETY PARAMETER DISPLAY SYSTEM SAFETY ANALYSIS APRIL 30,1984 ANO-2 PRESS-TEMP LIMITS '
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'l SAFETY ANALYSIS REPORT FOR THE ANO-2 SAFETY PARAMETER DISPLAY SYSTEM m
INDEX Page INTRODUCTION........................................... 1 1
BACKGROUND. ........................................... 2 1
l DESIGN 0VERVIEW................................... .... 3 BASIS FOR DISPLAYS AND PARAMETER SELECTION............. 5 I
CONCLUSION............................................. 12 I
1 1
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5 INTRODUCTION P
The purpose of this report is to provide a written safety analysis for the Arkansas Nuclear One, Unit Two (ANO-2) Safety Parameter Display System (SPDS). This document is also intended to satisfy the requirement for a safety analysis specified in Section 4 of NUREG 0737, Supplemert I.
In accordance with 10 CFR 50.59, implementation of the SPDS is a facility
- change and thus requires a written safety evaluation to provide the bases for the determination that the change does not involve an unreviewed safety question.
In addition, NUREG 0737, Supplement I, specifically requires a
- written safety analysis describing the bases on whicn the selected $POS parameters are sufficient.
The SPDS parameters should be adequate to access the safety status of each critical safety function identified in NUREG 0737,
" Supplement 1, for a widc range of events, incltding symptoms of severe accidents.
i 5 The hardware changes to ANO-2 associated with the installation of the SPDS
= were Thest implemented under a number of separate design change packages (DCPs).
k changes were controlled by various engineering and administrative procedures.
L The design control proceaure specifically addresses preparation of a safety determination for each DCP. This provides the basis for a pre-implementation determination of whether an unreviewed safety question axists with respect to the specific plant change.
Ihe safety determinations for the SPDS DCPs address the speci fic modifications c~,tained in the g
package and provide the required safety evaluations. This procedure.
therefore, ensures that proposed changes are adequately reviewea in g
~ accordance with the ANO Technical Specifications to determine whether the cha.ges involve an unreviewed saf ety question or a change to the Technical
- Specifications, per 10 CFR 50.59 requirements.
The 10 CFR 50.59 safety evaluations have been documented as part of each DCP and none of the changes involved an unreviewed safety question.
Furthermore, these DCPs have been reviewed by the Plant Safety Committee and the Safety Review Committee.
L As a result, this report focuses on the safety analysis for the selection of parameters.
V The SPDS for ANO-2 has not yet been declared operational with existing pa rame te rs ; however, certain capabilities currently exist. Discussions in this analysis are of these existing capabilities as well as currently proposed modifications.
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t BACKGROUND In response to the THI-2 accident in 1979, the HRC began issuing " Lessons Learned" NUREGs and Regulatory Guides addressing improved emergency response capabilities for nuclear power plants. The accident and subsequent investigations demonstrated the need for improving the presentation of plant and process information to reactor operators, especially during major transients. TMI-2 highlighted man / machine interface deficiencies because reactor operators were required to monitor and process such large amounts of data in order to ascertain the operating and safety status of the plant and take necessary actions. The available information in the control room was adequate, but was not presented in the most useful form, especially under stressful conditions.
To comply with the NRC's recommendations, various nuclear power industry organizations developed the concept of a safety console or panel to display plant safety information to the operator without subjecting them to "information overload". This concept was eventually refined into a safety parameter display system (SPDS) that would facilitate the assessment of plant safety by providing a set of predetermined color graphic displays which would be continuously updated with real time data to yield relevant, timely, accurate and unambiguous information.
The SPDS concept was to display a small but critical subset of the information already present in control room instrumentation in order to minimize information overload. Critical safety functions were to be carefully identified and parameters for their display selected in order to provide the operators with a concise set of data to aid them in rapidly and reliably determining the safety status of the plant.
AP&L began the development of an SPDS in 1979 as part of an in-house initiated E0P upgrade program and expanded the development early in 1980, in response to the " Lessons Learned" NUREGs 0578 and 0585. Two Systems Engineering Laboratory (SEL) 32/77 computer systems were ordered in June 1980 to provide the capability to perform the required functions. NUREG 0737, issued in October 1980, required the implementation of the plant SPDS.
NUREG 0737 referenced NUREG 0696 for use as the criteria for design of the SPDS and Technical Support Center (TSC)/ Energy Response Facility (ERF) instrumentation systems. However, NUREG 0696 was not issued until March 1981, so several modifications were required to the original computer system design as a result of the new criteria. Supplement I of NUREG 0737 was issued in December 1982 to provide additional clarification to certain NUREG 0737 requirements. Supplement I also promoted an integrated approach for the implementation of the SPDS, upgraded Emergency Operating Procedure (EOP), control room design reviews, emergency response facilities and Regulatory Guide 1.97 instrumentation reviews. The present SPDS computer .
system is designed to meet the objectives of the above referenced NRC documents.
2
I DESIGN OVERVIEW Prior to TMI-2, emergency response was event-oriented. Each pre-defined event (plant transient) was covered by an abnormal operating procedure. The i approach was straightforward: diagnose and recover, according to procedure.
TMI-2 showed that this method was not always adequate.
In complex systems like nuclear power plants, the operator must interpret _
and integrate a vast amount of information supplied by instrumentation and relate it to rule-based and knowledge-based training to determine the best recovery action.
Because of information overload, however, the operator may not always be directing his attention to the parameters which are the most relevant to the disturbance.
Integration of key plant process information from different subsystems is difficult under normal conditions and even more difficult under stress.
Control boards are often designed for control at the subsystem level, and are not necessarily conducive to overall plant response monitoring. The SPDS serves to integrate the appropriate information in a concise fcrm by means of an on-line computer monitoring system with human-factored CRT displays.
In accordance with NUREG 0737, Supplement 1, the SPDS displays are being integrated with the development of the upgraded ANO-2 Emergency Operating Procedure (EOP) to ensure compatibility. These new E0Ps will be __
symptom oriented. Instead of requiring diagnosis before action as the earlier emergency procedures, the new E0Ps will have the operator respond to the symptoms of the event.
The SPDS was designed to assist the operator in implementing the upgraded ~ --
E0P.
The ANO-2 E0P is being developed to achieve timely and accurate safety status assessment either with or without the SPDS. The design of the specific SPDS graphic displays correspond to specific sections of the upgraded E0P, so the SPDS will complement the use of the upgraded E0P.
The SPDS design also includes consoles and displays in the on-site Technical Support Center (TSC) and off site Emergency Operations Facility (E0F). This improves the operational aids available to the plant technical staff to 2 assist them in evaluating transient conditions and providing guidance and direction to the operations staff. The TSC and EOF both have access to the --
large-scale data storage and retrieval capabilities of the SPDS to assist in event diagnosis and historical documentation.
The computer system selected by AP&L was chosen primarily because of its
' flexibility, reliability, and maintainability. Flexibility .s needed to permit the incorporation of future modifications without unwarranted .
difficulty. The computer hardware selected is similar to existing hardware already in use at ANO. This hardware has been proven to be reliable.
[ Furthermore, AP&L personnel have considerable experience in maintaining this
) equipment which should improve the overall reliability of the system.
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N' The basic configuratio.n G.' the SPDS consists of redundant data acquisition, processing and display devices. The SPOS computers access the necessary input parameters fron senso'rs in ANO-1 and AND-?, s process these signals, and provide displays to each control room as wellt as to the TSC and EOF. The SPDS performs no plant cont'rol action, but serves ae, a human-engineered data - -
display system to aid the operator,i.7 rapidly and re'liably determining plant safety status.
" Touch screen" contreis 'ero u'.ilized on the color graphic CRTs to allow for rapid access'to the'information necessary to determine safety status of the plant. The,SPDS design'should provide enhanced capabilities for responding properly 'to both cnticipated and unanticipated plant conditions.
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.- . r ,g rr BASIS FOR DISPLAYS AND PARAMETER SELECTION The Emergency Operating Procedures (EOPs) for ANO-1 and 2 both have their origin in the Babcock & Wilcox Abnormal Transient Operating Guideline (ATOG) program. From this program is was determined, following a reactor trip and verification of shutdown, that their are three symptoms of primary interest to a pressurized water reactor operator to prevent core and reactor coolant
, system damage: 1) inadequate subcooling of the primary system inventory, 2) inadequate primary to secondary heat transfer, and 3) excessive primary to secondary heat transfer.
These symptoms are important for the following reasons:
- 1. Inadequate primary inventory subcooling: If the operator knows the primary fluids is in a liquid state, he is assured that it is available __
and capable of removing heat from the core. If subcooling is lost, these issues are in doubt, and he is therefore directed to make every effort to regain subcooling.
- 2. Inadequate primary to secondary heat transfer. This symptom addresses the heat transfer coupling across the steam generator. It describes i
the ability of the system to keep the flow of energy moving from the reactor coolant system to the ultimate heat sink.
l 3. Excessive primary to secondary heat transfer: In this case, the :
l symptom is indicative of a secondary side malfunction (e.g., loss of steam pressure control or steam generator overfill). The heat transfer _.
is again unbalanced and the operator's attention is directed toward generic actions to restore this balance.
( ,
I The information required to identify and track these symptoms is available in all nuclear power plant control rooms and simply consists of reactor coolant system temperature, reactor coolant system pressure, steam generator pressure and access to a set of steam tables.
[
( The AT0G pressure-temperature diagram (P-T Diagram) was developed to provide the above described information to the plant operator in a timely fashion with little or no effort on his part. The P-T Diagram is the top level _
display for the ANO-2 SPDS for this reason. -
, Pressure-Temperature (P-T)
The P-T diagram basic features are displayed in Figure 1. The static -
portion is a grid of RCS pressure versus RCS temperature with fixed J curves showing the saturation line, a 50 F margin to saturation line
( and the RCS NDT limits for normal operation. During power operation the Reactor Protection System (RPS) pressure trip limits are shown along with a normal transient window showing the minimum and maximum ._
pressures and temperatures expected immediately following a trip. A small box will appear inside this window showing the expected
' pressure-temperature relationship for normal hot shutdown conditions.
If no forced RCS flow is indicated, the single small box will be replaced by two boxes showing the expected T and T conditions during natural circulation at hot shutdown. coM M 5
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'The dynamic elements of the display include a bar graph representation of steam generator levels, digital . values of-selected pareneters displayed below the P-T p rd ' and T versus pressure. 'FollowYng a, and points idcatifying Treactor t' rip, the past pT8}.Ne temperature yersuss pressure remain on the screen showing the trajectory which they are follNing.; If no forced RCS flow exists, the average of k
the core exit temperatures is subst(tuted for J ' The values shown at the bottom of the screen are reactor. bMldinh ke.mperature and pressure, A and 8 stoasr generator pressures :and.the avsrage of the core, exit thermocouple t 6 peratures. ,,.
AtypicalplantresponsetoareactortripisshowninIigure2.
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T should merge with T as the decay heat rapidly drops. Both cold h
temperatures should Nen move toward normal hot shutdown pressure and temperature conditions. If they.do not, a departure from normal is indicated. If they move outside a larger normal transient limit' area, the definite need for operator action is indi pted.
Each of th'e'three basic symptoss' discusse e$'lierleavetheirunique signature on thp P-T diagram ay, displayed "n Figures 3,4, and 5. Use of this tool eubles an, operator's prior'.?/, to be fixed on controlling the plotted paraaeters'vithin t9rget bounds. If successful, he will be able to bring the reactor to a' safe condition. This will be the case
, regardless of why.ther or not-he' has properly diagnosed (or diagnosed at all) the event whi'ch has-occurrec. However, use of the SPDS in conjunction with the E0[ does act discourage an operator from^
diagnosing the cause.of the transient. The SPDS and E0P are based on directing the cpertturito take proper actions without diagnosis or wii.h misdiagaosis. '
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, 'c To further enhance the operators' ability to es.sess the plants' response to transients and to more precisely monitor specific safety functions,.
additional displays Gere de.el6 ped for the, ANO-2 SPDS based on subsw:.tial operator and AP&L engineering. input. Thes/aaditional displays were carefully created to b'e used iri conjunctioii with the ANO-2 EOP and will aid the operator in the implementWion of thi# procedure as well .ntsome select abnormal operating procedures. A description of these additi6nal displays is provided belog . '
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Primary to' Secondary Heat Tran'sfer (PSHT) .
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The primary to secondary heat transfer display was[cesigned f to provide more detailed historical data on some of the paramejers,shown on the-P-T diagram. Pages 1 and 2 of this display are shown in' Figures 6 and 7, respectively. 'g ..
Page one has trends for average1 ccrej, exit temperature, loop average. hot leg temperature, loop average cold leg temperature and saturation temperature for steam generator pressure.'- , ,
7 Page two displays r trend of feedwater flovgsteam generator level and steam generator pressure for both ' loops.
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This display will be particularly valuable in natural circulation conditions. Verification of natural circulation can be readi!y accomplished by noting that core exit and hot leg temperatures are tracking and trending down. Also cold leg temperatures will be tracking the saturation temperature for that loop's steam generator pressure.
Reactivity Control (RHO)
The condition for entry into the E0P is a reactor trip, either manual or automatic. The reactivity control display will be provided to aid the operator in immediate verification that the reactor is indeed tripped and remains shut down. Wide range neutron flux will be shown on this display along with indication of control element assembly status. A picture of this display is not provided as it is still under development.
Steam Generator Tube Rupture (SGTR)
A steam generator tube rupture is treated as a unique event in both the Abnormal Transient Operating Guidelines and the E0P. This event is unique in that it has the potential for a direct release of radiation to the environment. Also, if the reactor does not trip initially, every ef fort is made to shut the plant down without a reactor trip and subsequent release of steam to the atmosphere through the main steam safeties. The steam generator tube rupture display, as shown in Figure 8, is designed to aid the operator in diagnosing a tube rupture, determining the affected generator, and cooling the plant down to a condition where the primary to secondary leakage can be terminated.
This display shows trends of condenser off gas and main steam line radiation, steam generator levels and RCS average temperature.
RCS Inventory (RCSI)
The RCS inventory display, as shown in figure 9, is provided to aid the
{ operator in assessment and maintenance of RCS inventory. Volume control tank level, pressurizer level, RCS temperature and pressurizer pressure are trended on the top half of this display. On the lower trend are charging flow, letJown flow, quench tank pressure and narrow range containment sump level.
Trends shown on this display may also aid in the diagnosis of the initiating event. For example, the behavior of RCS temperature while pressurizer level and pressure are decreasing is the key to differentiating between a small loss of coolant accident and an i
I overcooling event. Both events will show decreasing pressure and ,
pressurizer level; however, rapidly decreasing temperature would indicate an overcooling event while a constant or very slight decrease in temperature would indicate a loss of coolant.
I 7
Containment Conditions (CONT)
The containment conditions display is being developed to show containment parameters which may be useful in assessing and maintaining containment integrity. Containment hydrogen and high range radiation will be trended on the upper half of this display. On the lower half, as shown in figure 10, will be displayed containment temperature, pressure and wide range containment water level.
Auxiliary Displays In addition to the dedicated selector buttons on the screen for the primary SPDS displays described above, there is a dedicated button for a menu display from which auxiliary displays may be selected. Touching the menu button calls up a display showing a list of auxiliary displays ,
with their associated selector buttons. When an " item" from the menu is selected, that display will appear and its label will be displayed in a blue backlit button at the bottom of the screen. Auxiliary displays include heatup and cooldown displays, the core exit thermocouple map and the graphic trend displays. A one-line diagram of the Engineered Safeguards (ES) electrical distribution system is planned when inputs are made available. Historical pressure versus temperature data from the previous 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> is saved and displayed on the heatup and cooldown displays when these displays are requested.
i These displays are not shown in this submittal.
As discussed above, the basis for the ANO-2 SPDS was the B&W ATOG program.
Implicit in this program was the consideration of the five critical safety function identified in NUREG 0737 Supplement 1. Correlation between the previously described SPDS displays and the five critical safety functions are provided below:
- 1. Critical Safety Function - Reactivity Control SPDS Display on which - Reactivity Control function is
{ assessed Parameters available - Reactivity Control
- Wide range neutron flux
- Control element assembly status (full in or not full in) 8
- 2. Critical Safety Function - Reactor core cooling and heat removal from the primary system SPDS displays on which - P-T Diagram function is assessed Primary to Secondary Heat Transfer Parameters available - P-T Diagram
- RCS pressure
- Thot (1 ps average)
- Tcold (1 ps average)
- Core Exit Thermocouple (average)
- Containment temperature (average)
- Containment pressure
- A/B steam generator pressures
- A/B steam generator levels
- Indication of saturation, 50 subcooled, and NDT limit for given pressure and temperatures Primary to Secondary Heat Transfer
- Core Exit Thermocouple (average)
- Hot leg temperatures
- Cold leg temperatures (loop average)
- Steam generator saturation temperature (calculated from pressure) ..
- Steam generator levels
- Steam generator pressures
- Feedwater flows _to each generator ,
1 9
- 3. Critical Safety Function - Reactor coolant system integrity SPDS Displays on which - P-T Diagram function is assessed RCS Inventory Parameters available - P-T Diagram (See List Above)
RCS Inventory
- Pressurizer level
- Pressurizer pressure
- RCS average temperature
- Volume control tank level
- Quench tank level
- Containment water level (wide range)
- Charging flow
- Letdown flow-
- 4. Critical Safety Function - Containment condition SPDS displays on which - P-T Disgram function is assessed Containment' Conditions Parameter available - 'P-T Diagram (See List Above)
Containment Conditions
- Containment radiation r
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- Containment hydrogen concentration
- Containment temperature (average)
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- Containment pressure
- Containment water level (wide range)
.10
- 5. Critical Safety Function - Radioactivity Control Comments: The four previous critical safety functions directly relate to one or more of the plant's barriers to the release of radioactivity (i.e., fuel clad, RCS pressure boundary and containment building). The monitoring of these four critical functions actually satisfies the goal of the fifth safety function, radioactivity control. However, as an additional level of useful information to the operator concerning radioactivity control, offsite gaseous radioactivity release informatin has been included as desirable to be concisely displayed. This information may be used by control room personnel to quickly assess the release data and make appropriate recommendations to offsite officials.
Effluent vent release data along with real time meteorological data is input to the Gaseous Effluent Radiation Monitoring System (GERMS). The GERM system correlates this data to provide real time dose projections. GERMS output CRTs are independent of the SPDS but are located inthe same facilities. This separate display technique is appropriate since the individuals who will actually i
be performing the dose assessment function are not the reactor operators who will be involved with the recovery of the plant. A detailed description of the GERM system and its capability was provided in out letter to Mr. John T. Collins dated October 21, 1984 (0CAN108213).
The displays described above were designed to monitor both emergency and abnormal operations. Many of the parameters displayed are provided so that the operator can better understand the plant status, but are not essential for the monitoring of the five critical safety functions identified in NUREG 0737 Supplement 1. For these reasons the list of parameters described to monitor each critical safety function exceed the minimum set required. We believe the added information provided onthe displays will enhance the operator's overall ability to understand the accident through the use of the SPDS and are thus justified.
Changes to the displays may be necessary in the future to further enhance the systems' canability. Changes may also be necessary to incorporate modifications which are identified as a result of the further development of the ANO-2 E0Ps. these modifications will be properly reviwed prior to being implemented.
11
CONCLUSION Based on the above discussion, AP&L has concluded that the ANO-2 SPDS design provides the control room operators with sufficient information to enable them to rapidly and reliably ascertain the safety status of the plant for a wide range of abnormal and emergency conditions. In addition, it has been concluded that the ANO-2 SPDS design provides sufficient information to be used with the upgraded emergency operating procedures to allow operator detection and mitigation of plant transients and accidents in a timely and accurate manner. Furthermore, the use of the SPDS will not mislead the operator and will not direct the operator to take improper actions.
The installation of the ANO-2 SPDS does not represent an unreviewed safety question or a change to the ANO-2 Technical Specification. In accordance with 10 CFR 10.59, the installation of the SPDS is being finalized and the safety evaluations are being documented as part of each Design Change Package.
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ANO-2 PRESS-TEMP LIMITS 2800
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O 200 300 400 500 600 700 RCS TEMPERATURE DEG F Figure 3. Inadequate Subcooling Margin: Th 18 n t Progressing toward its target value; in fact, ithasrapidlydroppedEftroughthesubcooledmarginline. This condition is diagnosed as loss of adequate primary inventory subcooling, or simply " inadequate subcooling margin," and the procedure is written with directions to take care of inadequate subcooling margin.
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ANO-2 PRESS-TEMP LIMITS 2800
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Figure 4. Loss of Primary-to-Secondary Heat Transfer: Thot is increasing as steam generator Tsat is decreasing. A A T between the two is growing larger. The secondary is no longer removing heat and has lost coupling with the primary. This condition is diagnosed and treated as loss of (inadequate) primary-to-secondary heat transfer.
SG Tsat is displayed as a digital readout below the P-T diagram.
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RCS TEMPERATURE DEG F Figure 5. Excessive Primary-to-Secondary Heat Transfer: Steam generator Tsat has decreased l below its established limit. Thot and Tcold have reached equal values but both have gone out of the post-trip window following steam generator T sat. This condition is diagnosed and treated as excessive primary-to-secondary heat transfer. SG Tsat is displayed as a digital readout below the P-T diagram.
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ARKANSAS POWER & LIGHT COMPANY I POST OFHCE BOX 551 LITTLE ROCK. ARKANSAS 72203 (501) 371-4000 l April 30, 1984 2CAN048402 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No. NPF-6 NUREG 0737 Supplement 1 -
g0-2 SPDS Safety Analysis Gentlemen:
We are pleased to submit the attached ANO-2 Safety Parameter Display System (SPDS) Safety Analysis Report in response to Section 4.2 of NUREG 0737 Supplement 1. Supplement 1 was transmitted to AP&L as an attachment to Generic Letter 82-33 (0CNA128226).
Details pertaining to the ANO-2 SPDS design and its implementation schedule were provided in our initial response to Supplement 1 dated April 15, 1983 (0CAN048312). The ANO-2 SPDS has not yet been declared operational with its existing parameters; however, certain capabilities currently exist on the system. Discussion in the Safety Analysis are of these existing capabilities as well as currently proposed modifications. The basis for the parameters selected and the display techniques are the same as those used for the development of the ANO-2 Emergency Operating Procedure (EOP).
Although the SPDS may be modified in the future, as the result of further E0P development as weil as other analysis, we believe that the SPDS displays -
described herein fully address the five critical safety functions identified in NUREG 0737 Supplement 1.
MEMOEA MOOLE SOUTH UTiuflES SYSTEM
4 Mr. Darrall G. Eisenhut April 30, 1984 I
l The safety analysis contains photographic reproductions of SPDS displays.
Due to the' cost associated with photographic reproduction, only three originals are attached. Should you require additional copies please contact
- me.
l.
Very truly urs,
{
n chn R. Marsha 1 anager, Licensing JRM:DEJ:ac l
Attachment i
i 1
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