ML18052A438
| ML18052A438 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/19/1986 |
| From: | Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 GL-82-33, NUDOCS 8605220084 | |
| Download: ML18052A438 (31) | |
Text
consumers Power l'OWERIN&
/llllCHlliAN"S l'ROliRESS General Offices:
1945 West Parnell Road, Jackson, Ml 49201 * (517) 788-1636 May 19, 1986
- Director, Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
Kenneth W Berry Director Nuclear Licensing RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - PALISADES PLANT SAFETY PARAMETER DISPLAY SYSTEM By letter dated August 21, i985, Con~umers Power Company submitted a revised preliminary Safety Analysis Report for the Palisades Plant Safety Parameter Display System (SPDS) to the NRC.
Subsequently, by letter dated March 17, 1986, the NRC requested that Consumers Power Company provide additional information regarding the Palisades SPDS in order for the NRC to complete its review.
That letter requested that Consumers Power Company respond by May 19, 1986.
The additional information requested concerned the identification of the parameters required by the Plant operators to evaluate the Critical Safety Functions identified in section 4.l(F) of the NRC Generic Letter 82-33, dated December 17, 1982, the basis for parameter selection, and where those parame-ters are displayed by the SPDS.
Information regarding the range of the displayed neutron flux, reactor vessel level, and radiation monitoring of an isolated steam generator was also requested.
In addition to the items identi-fied above, information was requested concerning the human factors guidelines used to design the display formats, the Verification and Validation Plan used to confirm the design and operation of the system, and information regarding the processing of trend data by the SPDS.
The attachment to this letter provides the additional information requested by the NRC with the exception of Consumers Power Company's final response to items 2.1, "Display Clutter" and 2.2, "Validation and Verification."
/" 8605,220084 860509 I I PDR ADOCK 050002~51
- p.
, PDFt I OC0586-0019S-NL01
Director, Nuclear Reactor Regulation Palisades Plant Response to Request for Additional Information May 19, 1986 2
Consumers Power Company letter dated May 6, 1986, "Notification of Delay in Responding to Certain NRC Concerns Regarding Palisades Safety Parameter Display System," identified that in order to adequately address the NRC concerns with regard to display clutter and validation and verification, Consumers Power Company was delaying its response to these items until August 29, 1986.
Although Consumers Power Company's final response to items 2.1, "Display Clutter" and 2.2, "Validation and Verification" has been delayed, a preliminary discus's ion of Consumers Power Company's position with regard to these items is provided in the attachment to this letter.
Consumers Power Company has not previously requested the NRC to perform a pr~implementatiori review of the Palisades SPDS.
Therefore, the delay in providing the information for the two items identified above, nor completion of NRC review of the SPDS, will result in delaying the implementation of the SPDS at the Palisades Plant.
The SPDS will be operational and plant operators will be trained by the end of December, 1986, as stated in the NRC "Order Modifying License to Confirm Additional Licensee Commitments on Emergency Response Capability "(Supplement 1 to NUREG-0737)," dated July 1, 1985.
In accordance with Section 4.2 of NRC Generic Letter 82-33, the safety evaluation performed by our offsite safety review committee for the SPDS is summarized below.
The design and function of the Palisades SPDS has been evaluated against the criteria of 10CFR50.59.
Based on that evaluation Consumers Power Company has determined that the Palisades SPDS does not invo*lve an unreviewed safety question.
The Palisades SPDS is designed to provide the Control Room Opera-tors with a concise display of the parameters to aid them in rapidly and reliably dete.rniining the safety status of the Plant.
These parameters have been determined based on the operator informational needs identified by the.
Function and Tasks Analysis and the upgraded Emergency Operating Procedures.
However, the Palisides SPDS does not control any Plant equipment and, no control actions will be taken by the Plant operators based solely on the information provided by the SPDS.
Existing Plant**sensors are used by the SPDS to generate the display.
The signals from these sensors are provided to the Central Processing Unit (CPU) of the SPDS through three separate input multiplexer cabinets.
One cabinet is provided for left safety channel input signals, one for right safety channel input signals and one for non-safety input signals.
The use of three separate cabinets provides for separation between safety*and non-safety signals.
Isolation between multiplexers and between redundant divisions within the multiplexers is 'provided by fiber optic cable.
(Detailed information regard-ing isolation was provided in Consumers Power Company letter dated August 21, 1985.)
Communication lines between the CPU and the multiplexers are terminat-ed in the non-safety multiplexer.
This isolation scheme has been installed to ensure that an electrical fault within the CPU or the multiplexers does not propagate into a safety related sensor channel.
Based on the above informa-tion, Consumers Power Company has determined that the probability of occur-rence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report has not been increased.
OC0586-0019S-NL01
Director, Nuclear Reactor Regulation Palisades Plant.
Response to Request for Additional Information May 19, 1986 3
As described above, the Palisades SPDS has been designed to provide a concise, readily accessible display of.critical Plant parameters.
However, since the SPDS is electrically isolated from any safety related Plant equipment and, since no operator action will be taken based solely on the SPDS display, Consumers Power Company has determined that the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report has not been created.
In addition, since the SPDS is not addressed by the Palisades Technical Specification, nor is the information displayed by the SPDS used in the determination to take action required by the Palisades Technical Specifications, the margin of safety as defined in the basis for any Technical Specification is not reduced.
72/11~~
Ken~
-B~~r;.
Director, Nuclear Licensing.
CC Administrator, Region III, USNRC NRC Resident Inspector ~ Palisades Attachments OC0586-0019S-NL01
ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT
. DOCKET 50-255 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION SAFETY PARAMETER DISPLAY SYSTEM MAY 1986.
MI0586-0984A-TC01-NL04
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PALISADES SAFETY PARAMETER DISPLAY SYSTEM 1.0 PARAMETER SELECTION
- i. l Functions and. Parameters NRC Position NUREG-0737, Supplement 1, requires that the data displayed by the SPDS shall be sufficient to provide information to plant operators about:
- 1.
Reactivity Control
- 2.
Reactor Core,Cooling and Heat Removal from.the Primary System 3.~
Reactor Coolant Sys_tem Integrity
- 4.
Radioactivity Control
- 5.
Containment Conditions For review purposes, these five items _have been designated *as Criti-cal Safety Functions (CSF's).
Our review of 'the licensee Is safety analysis was unable to make a
.direct.correlation of parameters in the SPDS with the above CSFs.
To
~ontinue our revi~w. the staff requests the l{censee to:
- 1) identify the minimum set of parameters required by operators to evaluate each of the above CSFs, 2) provide the.basis for each parameter's selec-tion~* _and 3) identify where in* the SPDS the parameter is displayed.
MI0586~0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant 2
Response to Request for Additional Information -. SPDS - Attachment 1 May 19, 1986 CPCo Response As discussed in our August 21, 1985 submittal, the Palisades CFM/SPDS is designed to monitor the critical safety functions as defined in*
the Combustion Engineering Emergency Procedure Guidelines (CEN-152).
The nomenclature used for the critical functions in the EPG's devel-oped by Combustion Engineering differs somewhat from that of Supple-ment 1 to NUREG.,.0737, however, i~formation on all critical functions identified in NUREG-0737, Supplemeht 1, is provided.
The nomencla-ture of each of th~ critical safety fun~tions monitored by the Palisades CFM/SPDS an.d the equivalent Supplement 1 CSF. is as follows:
CFM/SPDS Supplement 1 CSF
'i.
Reactivity Control
- Reactivity Control
- 2.
PCS Inventory Control
+
- 3.
PCS Pressure Control
- 4.
Core Heat Removal
+
- 5.
~cs Heat Removal
- 6.
Containment Atmosphere
+*
- 7.
Containment Isolation*
8 *. Environmental Control MI0586-0984A-TC01-NL04 Control Reactor Coolant System Integrity Reactor Core Cooling and Heat.
Removal from the P.rimary System Containment Conditions Radibactivity Control
Director, Nuclar Reactor Regulation Palisades Plant Response to. Request for Additional Information -
SPDS - Attachment 1 May* 19, 1986 3
PCS Inveritori and Pressure Contr61 taken together correspond to Reactor Coolant System Integrity id~ntifi~d in Supplement 1 to
PCS Heat Removal and Core Heat Removal taken together correspond to Reactor Core Cooling and Heat Removal from the Primary Coolant System identified in Supplement 1 to NUREG-0737.
Containment Atmosphere Control and Containment Isolation taken together corre-spond to Containment C.onditions while Environmental Con.trol corre-sponds to Radioactivity Control.
Thus, there is a direct relationship between the criti_cal safety functions defined in the Palisades CFM/SPDS and those functions described in Supplement 1 to NUREG-0737.
Each of the CFM/SPDS critical safety functions has been reviewed against the information requirements of the Combustion.Engineering Emergency Procedure Guidelines to determine the variables to be monitored in developing the CFM/SPDS Critical Function Alarm Algo.;..
rithms.
These variables comprise the minimum set of parameters required by operators to evaluate each of the CE EPG critical safety functions.
As described in our August 21, 1985 submittal, a top level display page provides the operator with a concise display of the status of each of th-e critical safety functions.
Various parame-ters are monitored for each critical safety function and an alarm is provided if the parameter is outside of pre-established limits.
This alarm directs the operator to appropriate second level display pages where the magnitude of the variable in question can be obtained.
This alarm further directs the operator to the appropriate Functional Recovery Procedure _containing instructions necessary for restoration of the critical fonction.
The following paragraphs identify parame-ters monitored for each of the criticai*safety functions, describes the basis for each.parameter, describes the*alarm associated with
- each parameter,. and* identifies the CFM/SPDS page where the informa-tion is displayed.
MI0586-0984A-TC01-NL04
/
Director, Nuclar Reactor Regulation Palisades Plant 4
Response to Request for Additional Information -
SPDS - Attachment 1 May 19, 1986 1.1.l.
1.1.2.
Reactivity Control The core reactivity control critical function is an indication of the state of reactivity in the reactor core following reactor trip.
The critical function *alarm logic is based on losing control of reactiv-ity, such as might be dueto an uncontrolled positive reactivity insertion, rather than an absolute* bound such as criticality.
Information on neutron flux is provided to monitor reactivity con-trol.
Both startup count rate and wide range log power are monitored to provide coverage from the lowest expected courit rate to greater than 100% power.
The reactivity control alarm logic uses a filtered value of the startup count rate or wide range power, depending on which parameter is in range.
For modes other than *startup and. power operation (mode is determined by. the trip breaker status) an alarm actuates for the case of a continually increasing neutron count rate or when power is not decaying as expec~ed after reactor trip. _An increasing count rate is indicative of positive* reactivity *insertion which is undesir-able during hot standby and hot or cold shutdown.. A second set of alarm logic uses the* startup count rate to annunciate. a high count rate when at cold shutdown conditions which would indicate an unde-sired reactivity addition.
Indication of the reactivity control parameters is provided on CFM/SPDS display page 211 (Figure Ill).
This display page has been slightly modified since our previous submittal to provide indication of power range neutron flux~ boron concentration and loop subcooled margin.
PCS Inventory Control The objective of the PCS Inventory Control critical function is to keep the core covered with an effective coolant medium.
Parameters
- MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant Response to Request for Additional Information -
SPDS - Attachment 1 May 19, 1986 5
included in the CFM/SPDS to monitor this function incl,.ude pressurizer level, reactor vessel level, and subcooled margin (as determined from T-hot, r~cold, core exit temperature arid pressurizer pressure)'.*
The PCS inventory control function display will alarm when normal control over either the coolant volume*or mass in the PCS has been lost.
For example, control over the mass is lost immediately follow~
ing a pipe break and control over the volume is lost immediately following a rapid cooldown~ Both of these events would lower pres-
- surizer level; therefore, a low pressurizer level is used for alarm.
High pr~ssurizer level initiates a second alarm to alert the operator of excessive inventory.
A third alarm is provided to indicate formation of a bubble in the PCS by monitoring for abrupt changes in
. pressurizer level.
Continued loss of inventory control will be alarmed by a subcooled margin alarm in the hot and cold legs and at the core exit..
Emergen-cy Proce.dure Guidelines (CEN-152) recommend monitoring the lowest of either hot or cold leg subcooled margin, or core exit subcooled margin for void detection in the PCS (with or without primary coolant pumps operating).
For this reason both low hot/cold leg subcooled margin and low core exit.subcooled margin provide an alarm.
Indica-tion of coolant level above the core is provided by the Reactor Vessel Level Indicating System (RVLIS).
Indication of the trend of cooling level below the top of the core is provided by core exit subcooled margin.
Reactor vessel level instrumentation and core exit theremocouple inputs will not be available to the CFM/SPDS computer until after the 1987 refueling outage~ No display currently exists for indicating these parameters, however, a display is being developed to sho~
Inadequate Core Cooling.and Inventory Control parameters.
This display will contain, at a minimum, the information shown on.
Figure 2.
In the interim, parameters indicative of PCS*inventory MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant 6
Response to Request for Additional Information -
SPDS - Attachment 1
. May 19, 1986
- 1. 1.3.
control, including pressurizer level and loop subcool~d margin, are displayed on CFM/SPDS display page 211 (Figure 1).
PCS Pressure Control The objective of the PCS Pressure Control critical function.is to maintain primary coolant system pressure between upper and lower bounds (i.e., within the.limits of the Post Accident Pressure-Temperature curves).
Parameters monitored to determine PCS pressure control include *pressurizer pressure and subcooled margin (as deter-mined from T-hot, T-cold, core exit temperature and pressurizer pressure).
The PCS pressure cont~ol critical function monitors the primary coolant system pressure in the range between the upper bound, which is the high pressure ~eactor trip setpoint, and the lowe~ bound, which is the pressure necessary to maintain.the PCS adequately subcooled.
A loss of pressure control will be alarmed by the CFM/SPDS when these boupds are exceeded or when pressure is changing so rapidly that the normal lnechanisms*of maintaining pressure control will not be able to prevent exceeding these bounds.
This alarm
- algorithm also monitors the maximum subcooled margin of the Primary Coolant System (PCS), to ensure that PCS stresses are not excessive following potential pressurized thermal shock events.
As.noted in Item 1.1.2 above, core exit thermocouple~ will not be available as inputs to the CFM/SPDS computer until after t.he 1987 refueling outage.
Thus, subcooled margin based on core exit tempera~
ture will not be available until then.
Following the 1987 outage, all parameters comprising PCS pressur-e control will be available on the proposed Inadequaie Core Cooling display (Figure 2).
In the interim, pressurizer pressure and loop subcooling are displa~ed on CFM/SPDS page 211 (Figure 1).
MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant 7
Response to Request for Additional Information - SPDS.- Attachment 1 May 19,_ 1986 1.1.4.
Core Heat Removal The objective of the Core Heat Removal critical function is to transfer heat from the reactor core to the primary coolant system.
Parameters monitored *to determine core heat removal capability include core tiT (T-hot - T-cold), subcooled margin (based on T-hot, T~cold, core exit temperature and pressurizer pressure) and core exit temperature.
Core heat removal may be accomplished by any one of the following methods:
- 1.
Forced circulation by primary coolant pumps
- 2.
Natural circulation
- 3.
Forced circulation by shutdown cooling system There are four *areas monitored in the core heat removal algorithm these are:
the temperature difference across the reactor, the
-subcooled margin at the core exit, subcooled margin at the coolant loops and core exit thermocouple temperature.
The temperature difference (tiT) across the reactor _indicates suffi:-
cient PCS cooling.
Cooling is maintained by forced circulation, natural circulation *or shutdown cooling.
The temperature difference is to he.below the full power value for natural circulation,.and below ~10°F for forced circulation with the reactor tripped.
The subcooled margin in.the coolant loops measures the saturation temperature, which is calculated as a function of pressurizer pres-sure, and the highest hot or cold leg temperature.
Maintaining subcooled margin assures that heat transfer occurs in the liquid
- phase.
MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation
- Palisades Plant
- 8.
Response to Request for. Additional Information -
SPDS - Attachment 1 May 19, 1986
- 1. 1. 5.*
Core exit temperature and core exit subcooled margin can indicate the formation of voids in the PCS and the trend of two phase water level in the core following a loss of coolant accident.
Core exit sub-cooled margin is calculated using the average core exit temperature and pressurize~ pressure~
Parameters associated with the core heat removal critical function, with the exception of reactor. core 6T, will be provided as shown on Figure 2 when this display is finalized following the 1987 refueling outage.
The reactor core 6T is available from the primary coolant system hot and.cold leg* temperature readings indicated on CFM/SPDS diaplay page 211 (Figure 2).
Until the Inadequate Core Cooling display.page is_implemented, loop subcooled margin is available on CFM/SPDS display page 211 (Figure 2).
PCS Heat Removal The objectiv~ of the PCS Heat Removal critical function is to trans-fer. stored and decay heat out of the primary coolant system.
Parame-.
ters provided to monitor PCS heat removal. capability include steam generator level, feedwater flow (main and auxiliary), shutdown
- cooling flow and shutdown heat exchanger inlet temperatures.
Foi proper heat removal by th~ steam g~nerators, adequate steam generator levels must exist and adequate feedwater flow must be available to maintain or restore steam generator levels.
Steam generator levels and feedwater flows are monitored and an alarm is provided. if these parameters a~e not within acceptable bands.**
When the primary system is being cooled by the shutdown cooling system (SDC), as indicated by the itatus of the SDC ~solation valves, shutdown cooling flow and shutdown cooling heat exchanger inlet temperature are monitored to assure proper operation.
Alarms are provided if SDC flow is less than required or if the rate of change MI0586"".'0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant 9
Response to Request for Additional Information -
SPDS - Attachment 1 May 19. 1986 1.1.6.
1.1.7:
of heat exchanger inlet temperature is positive, indicating inade-quate cooling.
Steam generator level and f eedwater flows are indicated on CFM/SPDS display page 233 (Figure 3).
Indication of shutdown cooling flow and heat exchanger inlet temperature is provided on display CFM/SPDS page 326 (Figure 4).
Containment Atmosphere Control The objective of -the Containment Atmosphere Control critical. function is to transfer energy from the containment to prevent*exceeding
.design pressure limits.
Parameters provided to monitor containment atmosphere control include containment pressure, temperature and containment hydrogen.concentration.
An alarm is provided on low.containment pressure to-notify the operator ~f the appro&ch to the negative design pressure.
Alarms on high containment pres.sure and temperature are provided to indicate the need to actuate spray and/or fan coolers to reduce the energy content of* the containment.
An alarm on hi-hi containment pressure indicates that.des_ign pressure is being approached.
The alarm for high hydrogen concentration provides notification of the need to
- actuate the hydrogen recombiners prior to obtaining a potentially explosive mixture.
Parameters associate\\i with the containment atmosphere control criti-cal function are provided on CFM/SPDS display page 244 (Figure 5).
Containment *Isolation The objective of the Containment Isolation critical function is to assure that containment isolation valves are closed when required to prevent the spread of* radioactive material to the environment.
MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades.Plant
- 10 Response to Request for Additional Information -
SPDS - Attachment 1 May 19, 1986 1.1.8 Parameters provided to monitor the containment isolation critical function 'include containment pressure and containment radiation.
Alarms are provided oh high-containment pressure and radiation to indicate the need to manualiy isolate the containment or to verify that automatic isolation has been accomplished.
The position of each of the containment isolation valves is also monitored as part of this safety function; an alarm is provide'd if one or more of the isolation valves in a flow path is open when isolation is required.
Containment pressure and radiation are displayed on CFM/SPDS page 244 (Figure 5).
Status of the containment isolation valves is displayed on CFM/SPDS display pages 344 and 345.(Figures 6 and 7).
The con-tainment isolation valve display page has been modified since our earlier submittal to pro~ide for additional ~alves and to better indicate the initiating parameters for closure.
Environmental Control The objective of the Environmental Control critical function is to limit ~nd/or quantify the release of radioactive.material from known release 'points.
Parameters provided to monitor the* environmental control safety-function includes condenser off gas radiation, stack radiation~ steam line radiation and containment radiation.
Alarm levels are set for each of these radiation parameters to indicate the need to i~olate various release paths or to indicate that a potential for release of radioactive material to the environment exists.
Paramete.rs associated with the environmental controf safety function are displayed on.CFM/SPDS display page 352 (Figure 8).
This display.
page has been modified since our previous submittal to incorporate.
the output of additional radiation monitors.
MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant Response to Request for Additional Information -
SPDS - Attachment 1 May 19, 1986
_l.2 Range of. Displayed Neutron Flux NRC Position 11 Neutron flux is a fundamental variable for monitoring the status of the Reactivity Control CSF and should be monitored for all power ranges.
However, from the information provided in Reference 1, it is not clear that neutron flux is_ monitored over the full operational range, from source to greater than design power.
The licensee should clarify that th~ neutron flux level, for ~11 flux ranges, is dis-played on the SPDS, or justify that the curren~ arrangement pr6vides adequate indic'ation for use unde-r all conditions.
CPCo Response Neutron flux level is monitored for all power ranges fr6m source range to greater-than design power.
As noted in the response to Item 1.1.1, parameters provided *to monitor reactivity control include startup count rate and wide range log power.
The range of the startup count rate indication is from 10° to 105 coun-ts per second. The wide range log power indication. covers the range from 10 to 125% power.
Power range indication covering the range from 0 -
125%
power is also indicated on CFM/SPDS display page 211, however~ this parameter is not included in the critical. function monitoring alarm algorithms.
1.1 Reactor Vessel Level NRC Position While the licensee has specified a "Primary Coolant System Inventory Control'i safety function in the submittal~ and has selected the pressurizer level as the variable to monitor, adequate justification that a rapid ind_ication of RCS inventory would be available for-all MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant 12 Response to Request for Additional Information -
SPDS - Attachment 1 May 19. 1986
. 1.4 conditions was not provided.
The licensee should provide justifica-tion for the sole use of pressurizer water level, or add reactor water level to the SPDS.
CPCo Response A reactor vessel water level monitoring system is being incorporated into the Palisades Plant design during the next refueling outage currently scheduled for mid-1987.
This level indication will be provided to the CFM/SPDS computer and will be included in the Primary Coolant System Iriventory Control safety function as identified in the response to Item 1.1.2 above.
A proposed CFM/SPDS display page for indicating reactor vessel level and core exit thermocouple informa-tion is shown in Figure 2.
Isolated Steam Generator and Radiation NRC Position From the information provided in the safety analysis, the staff was unable to determine how *radiation in the secondary system (steam generator and steamline) is monitored by the SPDS when the steam generator and/or their steamline are isolated.
The licensee should provide information on how radiation in the steam generator is monitored when the generator is isolated.
CPCo Response A main steam relief radiation monitoring system is installed to monitor releases in the event the atmospheric dump or steam generator safety valves lift.
Two monitors, one viewing each s te_am line, continuously record the activity present in the secondary steam.
The monitors are located outside of containment on the upstream side of the Main Steam Isolation Valves (MSIV's) and thus monitor main stea~
MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant Response to Request for Adqitional Information -
SPDS - Attachment 1 May 19, 1986 13 radiation levels even when the MSIV's are closed (if flow exists past the mon'itors).
The output of the main steam line radiation monitors are included in'the SPDS* and are displayed on CFM/SPDS displciy page 352 (Figure 8) *
. In. the event a steam generator is isolated and there is no flow
- through the atmospheric dumps or steam relief valves, the location of the main steam radiation monitors precludes monitori~g radiation levels in the steam generator as there is no flow past the monitors.
For this situation, the capability exists to sample the contents of-each steam generator.
This sa_mple can then be analyzed to determine steam generator radiation levels.
2. 0.
HUMAN FACTORS 2.1 Display Clutter.
NRC Position In reviewing the.display formats provided.in the safety analysis, the staff noted several displays that appeared very dense to the point of being *cluttered (e.g.; display page 222 and 233). *Many of. the, graphic based displays contained a large amount of text.
We also noted that many *of. the process*variabies were displayed with a one character identifi~r, followed by a numerical value, with no units.
A -0n~ character ide~tifier and the lack of tinits for process vari-ables makes it difficult to identify and use these data during emergencies.
The staff requests the licensee to submit for reiiew the human factors guidelines used to design the display formats and to provide test results on operator use of the dense display formats.
MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades. Plant.
Response to Request for Additional Information -
SPDS - Attachment 1 May 19 J 1986 CPCo Response 14.
Consumers Power Company considers the design of the current.CFM/SPn.s**
- display pages to be adequate for operators to obtain the plant status information that they require.
Operators who use.these displays are extensively trained in plant* operation and plant systems.
This training.includes an extensive review of P&ID's and one line diagrams of plant systems.
The mimics used as.a display format. reinforces this training and presents the information in a manner which is familiar to the operator.
Although there is a large amount of information presented on some CFM/SPDS display pages, the use of mimics provides for rapid identification of required information by allowing the operator to quickly focus on the component or.subsystem of interest.
The one character identifier utilized on the CFM/SPDS displays to characterize parameter type is limited to t~mperature, pressure, flow and level.
The one letter code (ie, T, P, F and L). relates dir-ectly to the type of parameter being displayed and is easily interpreted by users of the displays.
Spelling out the identification would in-crease the number of characters displayed on the CRT screen, thus increasing display clutter.
Further, there has been no negative feedback from users of the CRT displays concerning the use of the single character identifier.
Based on these considerations, Consum-ers Power concludes that use of the one character identifier does not make it difficult to identify and use this data during emergencies.
Units of measure were left off of the parameter values to minimize the number of characters displayed and, thus, reduce display clutter.
For some parameters, such as temperature and pressure, this approach is reasonable as the units are consistent for every application (ie,
°F and PSIG).
For other variables, such as flow, the units are not always c~nsistent ~nd indicatiori of units for these cases may be required.
Consumers Power Company is evaluating the* use of units on Ml0586:..0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant Response to Request for Additional Information -
SPDS - Attachment. 1 May 19, 1986 15 the CRT displays.
Unit designations will be added to those process parameters where required to allow for rapid identification during emergency conditions.
Due to the knowledge level of the users of the CFM/SPDS concerning the. systems and parameters displayed, the number of variables to which units will be appended will be minimal.
The vendor of our Critical Function Monitoring system (CFMS) devel-oped human factors guidelines prior to designing the CRT displays for Palisades.
We have requested that the vendor provide us with addi-tional information on the CFMS design philosophy as well as informa-tion on the human factors guidelines used in developing the CRT displays.
Following r~ceipt and review of this information, Consum-ers Power Company will provide a further response to this question.
This further response will be provided by August 29, 1986.
2.2 Verification and Validation NRC Position The safety analysis contained a few general statements that indicated
- an extensive verification and validation effort was ~erformed for the design and development of the display system.
- The staff requests the licensee to submit for review the Verification and V~lidation Plan used in.the design.
CPCo Response
. Consumers Power Company b.egan procu.rement of the Critical Function Monitoring System (CFMS). in 1980 prior to formal requirements.for verification and vat'idation being developed.
Therefore, no forma:l Verification and Validation Plan was used in the design of the CFMS, however, design and testing of the CFMS was controlled during the process to provide assurance that the final product would perform as specified.
Controls.utilized in designing the CFMS included the MI0586-0984A-TC01-NL04
Director, Nuclar Reactor. Regulation Palisades Plant Response to Request for Additional Information -
SPDS - Attachment 1 May 19,
- 1986 16 development of functional design specifications for each aspect of the design ~nd acceptance testing to assure proper functioning of both hardware and software.
The CFMS has been installed and func-tioning at the plant for approximately two years.
During this period, only a few minor software problems have been identified and corrected *. Based on operating experience with the.system, Consumers Power Company has concluded that backfitting a* formal verification and validation program would not be cost effective.
The vendor of the Palisades CFMS system ha~ informed us that exten-sive man-in-the-loop validation testing of the CFMS design philosophy was performed at the Loviisa Power Plant in Finland.
Consumers Power Company has requested our vendor to provide us with the relevant validation data collected during this testing.
Following receipt of this information, we will provide a further response to this question by August 29, 1986.
2.3 Oscillatio~ of Process Variables and Displayed Data NRC Position.
- our review o~ the safety analysis noted that trend graphs of process variables* are available from the SPDS.
To complete our review, we need *information on how* trend data is processed by the SPDS.
Fur-:
thermore, the licensee should describe the features of the SPDS design that transmit and display oscillating process variables or limit cycling process variables, which may be symptoms of a severe
- accident, CPCo Response The capability to trend input process parameters is provided in the SPDS.
Two display pages of. trend information can be accessed by the*
operator:.
Each trend page can display up to four parameters MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation
- Palisades Plant Response to Request for Additional Information -
SPDS - Attachment 1 May 19, 1986 17 simultaneously.
The parameters to be trended and the trend update rate are selectable by.the operator.
The maximum update rate select-able is once per second.
Due to constraints imposed by the trend display, only 256 points can be displayed at a time.
Thus, for a once per second update rate, slightly over four minutes (256 seconds) of data will be ~ispla~ed.-
Data to be displayed on the trend~graphs is processed by the computer and front end equipment prior to display.
The front end equipmerit consists of multiplexers which perform the analog to digital conver-sion or provide digital status indication.
The front end equipment also includes a communication station which continuously interrogates the multiplexers as to the current value of data and provides the current data upon request to the central processing unit (CPU).* The CPU requests. data from the communicaticinsstatiori: once per second.
The CPU converts the input data to engineering units and outputs* the converted data to the CRT display generators for use in updating the trend graphs.
No filtering of the raw input signal which could mask process variable oscillations is performed prior.to display.
The scan time of the front end equipment is such that at a minimum, the communications station is updated twice per.second.
Taken with the once per second.update rate of the CPU means that new data being displayed on the trend graph is, at most, 1.5 seconds old.
- Thus, timing considerations associated with processing the data for display on trend graphs does not r~sult in a significant delay in the presen-tation. of information to the operator.
As noted earlier, the maximum update rate of the trend graphs is once per second.
At.this update rate, oscillations with periods on the*
orde~ of a.few seconds or less could not be ~dequately evaluated.
However, oscillations or cycling associ~ted with reactor plant process parameters (e.g *. levels, pressures.* temperatures) would occur with a period o.n the.order of minutes.
Thus, the once per second MI0586-0984A-TC01-NL04
Director, Nuclar Reactor Regulation Palisades Plant Response to Request for Additional Information -
SPDS - Attachment 1 May 19, 1986
. update rate is more than adequate for displaying oscillating trends of plant parameters.
18 Based on the above, Consumers Power Company concluded that the speed of the processing system and the update rate of the trend graph displays is adequate to provide the operator with correct and current information.
Further, the update rate of the trend graph displays is adequate to provide information to the operator on oscillating or cycling process variables which may be the symptoms of a severe accident.
MI0586-09S4A-TC01-NL04
FIGURES.
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