U-600066, Forwards Responses to NRC Comments on Revised SPDS (TMI Action Plan Item I.D.2) for Review.Scheduling of Design Validation Audit in mid-Oct to Support Final Closure Requested
| ML20126B367 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 06/12/1985 |
| From: | Spangenberg F ILLINOIS POWER CO. |
| To: | Butler W Office of Nuclear Reactor Regulation |
| References | |
| TASK-1.D.2, TASK-TM U-600066, NUDOCS 8506140082 | |
| Download: ML20126B367 (15) | |
Text
.
U 600066 L30-85 (06-12 )-L P04-85(06-12)-L
0 ILLINDIS POWER COMPANY IP CLINTON POWER STATION P.o. BOX 678. CLINTON ILLINOIS 61727 June 12, 1985 Docket No. 50-461 Director of Nuclear Reactor Regulation Attention:
Mr. W. R. Butler, Chief Licensing Branch No. 2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Clinton Power Station Unit 1 CPS Responses to NRC Staff Comments on the Revised Safety Parameter Display System (TMI Action Plan Item I.D.2)
Dear Mr. Butler:
Illinois Power (IP) Company provided the results of the Clinton Power Station (CPS) Safety Parameter Display System (SPDS) Corrective Action Plan implementation in a meeting with the Staff and your consultants (Science Applications International Inc. - SAIC) on February 20, 1985 in Bethesda, Md.
Recently IP received several additional NRC/SAIC comments related to this revised SPDS design. Attached are the formal CPS responses to these comments for your review. This material, in conjunction with the detailed information provided in IP letter U-0830, dated April 11, 1985, represents a comprehensive IP response to the NRC/SAIC concerns identified during the SPDS Design Verification Audit conducted on December 12 and 13, 1984 at CPS. As such, IP considers these concerns resolved and the Design Verification Audit closed.
IP is proceeding with the implementation of the revised SPDS design.
IP understands that final NRC closure of TMI Action Plan Item I.D.2 is contingent upon successful completion of the Post-Implementation Audit (Design Validation Audit) yet to be performed by the NRC.
In addition, the Staff has recently indicated that issuance of the CPS Operating License will not be contingent upon closure of SPDS issues.
In light of the significant changes recently made to the CPS SPDS, IP considers the revised design to be in full compliance with the applicable requirements of NUREG-0737 Supplement #1 (per Generic Letter 82-22).
IP believes the results of the SPDS Dynamic Simulation Test (scheduled for late August 1985) will confirm the design adequacy. As such, IP will continue to strive for closure of the SPDS issues, and thus TMI Action Plan Item I.D.2, prior to issuance of the Operating License (OL). Therefore, IP requests that the Staff consider scheduling the Design Validation Audit in the mid-October timeframe in support of j
final closure in the last Supplement to the CPS Safety Evaluation Report (NUREC-0853) prior to issuance of the OL.
8506140002 850612 iO\\
PDR ADOCK 05000461 F
PDR 0
1 ho i
U 600066 L30 85(06-12)-L
- P04 85(06-12)-L 1A.120 Please confirm that the timeframe identified above for the Staff's audit is acceptable so that a firm date can be established and planned for. If you should have any questions on the attached information, please contact us.
Sin erel yours, i$.N.
1k,
F. A. Sppngey erg j
Directo1 - Nt clear Licensing and Configuration Nuclear Station Engineering TLR/ lab Attachment i
cc:
B.L. Siegel, NRC Clinton Licensing Project Manager NRC Resident Office Regional Administrator, Region III USNRC Illinois Department of Nuclear Safety I
r l.
Attachment to U-600066 CPS RESPONSES TO NRC COMMENTS ON THE REVISED SPDS DESIGN NRC Audit Report Item # 10 It is unclear how Reactor Scram and MSIV Isolation feed into the Critical Safety Function (CSF) for "LVL".
It would seem that drywell temperature or humidity would be better inputs to this CSF since (1) these inputs would indicate a break in the system and (2)
Reactor Scram and MISV Isolation would alarm too frequently, leading to operator desensitization of these alarms.,
CPS RESPONSE The Primary basis for parameter selection is for support of the Clinton Emergency Operating Procedures (EOPs) which are, in turn, based on NRC approved BWR Emergency Procedure Guidelines (EPG).
The extensive research, evaluation, and analysis used to develop these EPG (and E0Ps) led to the conclusion that high drywell pressure, low reactor water level, high reactor pressure, reactor scram (without power downscale), and MSIV isolation are the five symptoms that should lead the operator into the Level Control Emergency E0P. Therefore, these parameters were selected for the LVL CSF. Other parameters were not considered essential by the EPG development groups.
NRC Audit Report Item #18 A review of the color coding used in the revised SPDS design identified the following problems despite IPC's establishment of a color coding convention:
The Clinton SPDS color coding convention for displaying static elements is cyan and for dynamic elements is green, red, or white. However, the following color usages appear to violate this convention:
cyan is used for the dynamic display of the time of day (hours, minutes, and seconds).
green and red are used to indicate the static display of the CSFs.
If red must be used as the color for the CSF abbreviation as well as for the CSF box in an alarm condition, then the abbreviation should not blink in " reverse video", as is used elsewhere.
Red alphanumerics are not always displayed in " reverse video" as indicated in the color coding convention. Not only are the CSF abbreviations in violation of this convention for red, but alarms such as " neutron trip" or " scram" are also in violation.
l-Attacha:nt to U-600066 Although yellow has been eliminated from the SPDS display and color coding convention and thus eliminates previous inconsistencies in the use of this color, the operator no longer has an indication of a trend towards an alarm state.
Yellow, consistently applied could be used to provide the operator with a warning or advanced notice of a parameter trending towards an alarm state.
. CPS RESPONSE The display of the system time of day will be green.
The Critical Safety Function boxes have been changed to avoid the conflicts mentioned in the NRC comments of 4/26/85.
The CSF names are static elements and therefore displayed in cyan. They are located above the CSF boxes.
The CSF boxes are dynamic elements and are displayed in green " normal", and red for " alarm".
All numerical data in alarm is displayed in red reverse video. To avoid an overuse of reverse video the alphanu-meric and graphical elements are green " normal" and red for the " alarm" condition.
It is necessary for the SPDS display to be consistent with the rest of the Nuclenet system in the use of the color. yellow. The GE supplied generic Nuclenet system operates with a'"two color /two state" data base and alarm system. Therefore, a " cautionary alarm" capability does not exist in the Nuclenet design and is not
- considered necessary to meet the CPS SPDS design objectives.
- NRC Audit Report Item #26 In response to the NRC concern for the lack of input for a " failure to scram" condition in-the " Power Control" status box, IPC has replaced this box with a " Reactivity" status box.
It is unclear whether IPC retained the high power trip input.
CPS RESPONSE
~
The reactor scram signal used for the reactivity control (RCTY) CSF includes the trip inputs from the neutron monitoring system. This trip signal includes APRM fixed upscale trips, APRM flow biased
' high-power trip, APRM inoperable trip, IRM upscale (based on run switch position), and Source Range Upscale trip in addition to the j
ATWS entry condition (power above 3% after 6 seconds).
4 y
w w
,,.w-
,-v-er
,,w-e n--,,*r---<
---r,y v----
--e--..r+-..,-y
- -, ~ - -, - - --
T:
Attachm:nt to U-600066 NRC Audit Report Item #31 With.the present'SPDS design, operators rely _soley on the time clock to signify functioning of the computer and display systems.
An indication'of a failed system by " freezing" the time of day indication is not immediately obvious to.the operator. This particular type of system failure indication has been observed in the past to go unnoticed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after system failure.
A means to clear the screen and display a_" SYSTEM FAILURE" message to. provide a readily noticeable indication of. failure should be investigated.
CPS RESPONSE In response to the NRC's concerns with a failure in the Display Control System which would result in a " static display" for the control room operator, the following is offered:
A failure of the "in control" Display Control Processor (DCP) is detected.by.the other processors and a system reconfiguration occurs (via the Test and Reconfiguration Unit) in which the
" standby".DCP assumes the "in control" role. An annunciator and status lights (on the Standby Information Panel (H13-P678)) alert
~
the operator to the reconfiguration. This failure will not result in a " static display". The failure modes of the video monitor are such that the screen will either be blank or. rolling. This will alert the operator of the problem.
The remaining failure that could result in " static display" is a failure of the display generator.
The fc11owing indications are available for the control room operators to detect the failure of a display generator:
1.-
The "DCS Component Failure" annunciator will be activated.
2.
Status indicating lights on the Standby Information-Panel (H13-P678) identify the failure (and any associated reconfiguration).
3.
In the-case of a failure to CRT #5, CRT #4 will display the SPDS display.
4.
All "non-failed". video monitors will indicate the failure-of the failed display generator channel on the bottom line of.the display.
5.
Time on the " failed" video monitor does not change.
l l
i
Attacharnt to U-600066 SAIC Audit Report Item #1 As stated.by IPC during and in its response to--this item of the SAIC audit report,-a checklist review of the SPDS in addition to a dynamic validation program will be conducted and integrated with
'the DCRDR.- A brief evaluation of the revised SPDS design by SAIC human factors engineers identified discrepancies with established human factors display design principles. These discrepancies should be considered and integrated with those identified in the SPDS evaluations as part of the DCRDR. The following discrepancies were found:
a.
The bargraphs do not conform to established human factors display.
design principles as follows:
The number of minor or intermediate marks between scale graduations in the primary display is 24. The number should not exceed 9.
(NUREG-0700, 6.5.1.5.a.1)
Only one size of graduation marks between numerals is used.
Established human factors principles for display design recommend the use of major, intermediate, and minor or
~
graduations marks if there are 5 or more graduations between numbers.
(NUREG-0700, 6.5.1.5.a.3)
The SRM bargraph graduation marks are not numbered at all.
(NUREG-0700, 6.5.1.2)
The Containment Pressure scale in the primary display increases by 3 units of measure and the Reactor Water Level -
Narrow Range scale increases by increments of 2.4.
Established human factors principles for display design recommend that scale graduations / internal values progress by 1,-2, or 5 units or decimal multiples thereof.
(NUREG-0700, 6.5.1.5.c) b.
The display titles (e.g., LEVEL CONTROL /RCTY) employ characters of the same. height as lower level information (e.g., parameter names).
Established human factors principles for display design recommend that.such summary labels be larger than labels used to identify individual parameters.
(NUREG-0700, 6.6.1.2)
'c.-
Vertical bargraphs on the CSF displays do not have the lower and upper limits identified or labeled.. (NUREG-0700, 6.5.1.1.b and 6.5.1.2)
-d.
Pictorial representatives of " Containment" and " Stack" are not standardized across.CSF displays (NUREG-0700, 6.6.6.4.c.2) e.
The MSIV is not labeled as such.
(NUREG-0700, 6.6.6.4.b.6)
~-
Attech::nt to U-600066 f.
With the exception of H - GAS, the CSFs are labeled on their 2
addedtotheexistingtitle"ContainmentContro$."GASshouldbe related CSF display formats. It appears that H -
(NUREG-0700, 6.5.1.1.b) g.
The-CSF characters in addition to the boxes or demarcated borders blink,' making the reading of the characters difficult.
(NUREG-0700, 6.7.2.1.a)
CPS RESPONSE The SAIC evaluation was done on the Prototype Display developed for the Feb. 20, 1985 NRC presentation. They were mock-up formats
' meant only to provide the overall approach CPS was taking to the SPDS. Design of the SPDS will meet all the stated SAIC concerns.
The guidance of NUREG-0700 will be used as follows:
a.
For bar graphs the number of minor or intermediate marks between scale graduations in the primary display does not exceed 9.
Major, minor and intermediate graduation marks are used on all bar graphs SRM bar graph graduation marks are numbered.
Scale graduation progresses by 1, 2 or 5 units or decimal multiples thereof.
b.
Display titles are larger than other lower level information.
c.
Vertical bar graphs have lower and upper limits identified and labeled.
d.
Pictorial representations of " Containment" and " Stack" have been standardized per the NRC concerns from the Feb 20, IP presentation, e.
MSIV is labeled.
f.
H -GAS has been added to title.
2 g.
CSF characters do not blink, per the NRC concerns from the Feb 20, IP presentation.
SAIC Audit Report Item #35 1
In response to the concern that primary containment pressure does not trigger the containment integrity CSF alarm, IPC states that drywell pressure is included in the "CNMT" and "LVL" CSFs and that since there are four redundant vacuum breakers between the drywell l
l
Attechm:nt to U-600066 and containment, the probability is low that containment pressure will be higher than the drywell pressure. The concern here is for leakage in the other direction, i.e., pressurizing primary containment which has a lower design pressure than the drywell.
CPS RESPONSE The basis for selection of inputs to.the CSF alarm boxes relates to entry conditions for the CPS Emergency Operating Procedures (EOPs).
Containment pressure does not provide an entry into any CPS E0P.
The containment integrity CSF alarm is triggered at a drywell-pressure of 2 psig. This is well below the design pressure of the containment; and as previously discussed, the containment pressure will not exceed drywell pressure.
It should be noted that the "CNMT" CSF Alarm box will be triggered by high Drywell pressure long before any threat to Containment integrity can exist.
In addition, containment pressure indication, from -5 to +45 psig, is provided on the secondary SPDS displays related to containment integrity and radioactivity control.
SAIC Audit Report Item #46 In response to the concern regarding the adequacy of only the wide range reactor vessel water level being supplied to the SPDS, IPC responded that a review was conducted of the adequacy of the level instrumentation with respect to operation during elevated drywell temperatures and while controlling level to control power during the ATWS event.
IPC states that the parameter set has been
. expanded to include additional instrumentation signals. However, IPC did not state what the new instrumentation signals were.
CPS RESPONSE The SPDS displays provide an indication of the following Reactor Water Level signals:
Wide Range (WR):
-160" to +60" (referenced to instrument zero
- 15" above bottom of dryer skirt)
Narrow Range (NR):
0" to +60 (referenced to instrument zero)
Shutdown Range (SR):
0" to +400" (referenced to instrument zero)
Fuel Zone Range (FZR):
-150" to +50" (referenced to Top of Active Fuel)
"NR" is provided on the SPDS Summary Display.
"WR" is provided on all three SPDS Seconday Displays.
"SR" and "FZR" are provided on the SPDS Level Control /RCTY Secondary Display.
Attachment to U-600066 i
The water level indication signals provided on SPDS cover the entire range monitored from Bottom of Active Fuel to above the Main Steam Lines (see attached figure).
"FZR" and "WR" provide the indication needed for monitoring conditions during an ATWS. The NRC-approved CPS Emergency Procedure Guidelines direct the operator to lower Reactor Water Level to the Flow Stagnation Water Level (near Top of Active Fuel for CPS) under ATWS conditions so as to minimize power production and subsequent Suppression Pool heatup.
The review referred to with respect to the water level instrumentation accuracy during elevated drywell temperature condit. ns was conducted by IP in response to Generic Letter 84-23 and TMI Action Plan Item II.F.2, entitled " Inadequate Core Cooling Detection" (see IP letter U-0766, dated December 5, 1984). This analysis concluded that the CPS RPV Water Level instrumentation provides an accurate indication of conditions leading to and the existence of inadequate core cooling. The NRC Staff accepted IP'a position and these issues were closed in SSER #4, Section 4.4.2.
CPS RESPONSE TO NRC/SAIC
GENERAL COMMENT
S General Comment #1 The documentation provided by IPC in the February 20, 1985 meeting with the NRC Staff does not discuss the details of the plans for performing the verification and validation of the SPDS. The adequacy of IPC's SPDS design heavily depends on this process. The V&V process should be documented in enough detail to allow a comprehensive evaluation of the methodology, criteria, and results to be performed by the Staff.
CPS Response #1 The Clinton Power Station Safety Parameter Display System Verification &
Validation (SPUS V&V) Plan was originally provided to the NRC Statf, as required by NUREG-0737 Supplement #1, in the CPS SPDS Pre-Implementation Package (submitted via IP Letter U-0676, dated October 28, 1983). The staff identified several concerns during the December 1984 Design Verification Audit, related to the results being achieved by the V&V effort. Due to these concerns, the Staff emphasized the need for the V6V Program to include a dynamic simulation test of the CPS SPDS. The CPS SPDS Corrective Action Plan, developed and submitted to the NRC in IP Letter U-0771, dated December 21, 1984, provided a comprehensive
" top-down" approach to resolving these concerns as well as the methodology and criteria to be used in this process. The V&V Program review team has been enhanced with additional plant personnel and a consultant from S. Levy Inc. (has experience on the Susquehanna SPDS
-development). Also, the V&V Team has documented their comments produced from the reviews of the Corrective Action Plan efforts and has
Attachm2nt to U-600066
. determined that the revised SPDS design meets the required functional obj ectives.. In addition, IP has committed to develop and perform a udynamic simulation test of the SPDS following design completion..The SPDS Corrective Action Plan was completed by February 1985, with the exception of the dynamic simulation test. A meeting was held with the NRC Staff and their consultants (from Science Application International Inc.)Lon February 20,~1985, to provide the Staff with the results of the Corrective Action Plan efforts. As a result of this meeting, the Staff concluded that IP's general approach to SPDS development was acceptable
, pending the final review to be conducted by the Staff during the SPDS Design _ Validation Audit..The Staff reemphasized the need for the dynamic simulation test of the SPDS at Clinton.
IP reaffirmed that this
. test would be. performed and noted that the schedule and details for this test would be provided to the Staff as soon as possible.
The CPS SPDS Dynamic Simulation Test will be performed as a V&V Program activity. The following schedule milestones relate to the performance of this test:
1.
Submit SPDS Dynamic Simulation Test Scope Document to the NRC Staff - July 1, 1985; 2.
Perform SPDS Dynamic Simulation Test - late August, 1985; and 3.
Submit SPDS Dynamic Simulation Test Results and Evaluation-Report to the NRC Staff - October 4, 1985.
The purpose of the SPDS Dynamic Simulation Test is to evaluate the adequacy of the CPS SPDS displays and the chosen parameter set. The test will be performed on the CPS Simulator utilizing certified Control
- Room operators. Dynamic simulation will involve the interface between the operators, the SPDS display information, the CPS Emergency Operating Procedures, and the remaining Control Room equipment controls and instruments. This test will involve running several transient / accident scenarios (such events as Loss of Coolant' Accidents, Anticipated Transient Without Scram and Loss of Decay Heat Removal are under consideration) on the CPS Simulator. The V&V Team will include members from Plant Technical Staff, Licensing, the CPS V&V consultants, Nuclear Training, and Human Factors. The evaluations to be performed will
-consider the following:
-1.
Adequacy of the SPDS Parameter Set to provide the necessary operator information required to take appropriate response actions; 2.
Acceptability of the SPDS Display Formats and color coding schemes - e.g.,
questions like "Is the information provided in a clear and concise manner, so as to minimize the potential for operator error?" will be evaluated; 3.
The time-dependent response of the operator during each scenario will be considered as part the of evaluation of display / parameter design; and t
4 v's
=
n-.w,-
,-r-w--
w e
r-
--w,.--
+--ww-me-,nr-,,e-v--,,
,---s
-,e..
.a m-,-v.-ww
--n
-,e- - -, -. - - -- -
Attachm;nt to U-600066 4.
The operator shift team-dependent actions and the required interface with other Control Room instruments will be evaluated as they relate to the SPDS information provided.
The scenarios selected for the SPDS Dynamic Simulation Test will be chosen based on the following:
1.
Exercise each of the SPDS Critical Safety Functions; 2.
Exercise each of CPS Emergency Operating Procedures; and 3.
Time-sequence of events chosen should cover a range of-developing plant conditions requiring slow-to-rapid operator responses.
Additional information related to the CPS SPDS Dynamic Simulation Test will be provided to the NRC Staff as described in the above schedule.
General Comment #2 It is unclear how the SPDS and operator training will coordinated.
CPS Response #2 The Nuclear Training Department will start the Licensed Operator Requalification Program on Sept. 4, 1985.
In this program, Control Room Operators, Shift Supervisors, and Shift Technical Advisors are cycled through a training week as they rotate from shift to shift. Every 5 weeks each license holder and STA undergo upgrade training.
The Operator Requalification Program Description (Draf t) describes two program facets which apply to the SPDS implementation:
1.
Pre-Planned Lecture Series - Operational Proficiency 2.
Skills Training The Operational Proficiency Lecture Series is the mechanism by which changes to facility design are presented to license holders. A lesson plan will be developed from the SPDS design document and a lecture presented to all license holders and STA's prior to fuel load. This lecture will be presented at the appropriate level for individuals required to manipulate the displays and interpret the data.- IP Skills Training utilizes the CPS plant specific simulator to drill license holders in abnormal and emergency plant evolutions. The SPDS Displays will be installed in the simulator when the Licensed Operator Requalification Program begins. Each license holder and STA will be required to participate in evolutions which utilize the SPDS displays prior to fuel load.
AttachmInt to U-600066 Ceneral Comment #3 Although IPC states that it has conducted a " top-down" approach to designing the SPDS, its methodology does not consist of or is based upon
- an analysis of tasks involved in emergency operations which determines the information needed by the operator to accomplish tasks. The applicant should describe how the task analysis that is being/was used to generate Emergency Operating Procedures will be used to validate the SPDS design.
CPS Response #3 See CPS response to NRC/SAIC General Comment #1.
The information needed by the CPS Control Room Operators to perform emergency operations is currently under. evaluation as part of the_ CPS Detailed Control Room Design Review (DCRDR) System Function and Task Analysis (SFTA). The results of the SFTA will be available by the end of May'1985. When this effort is complete, the results will be compared L
. to the SPDS parameter set to determine the adequacy of the information provided for monitoring the CPS Plant Criticaf Safety Functions. The
- SPDS Verification and Validation Team will utilize the list of operator information needs generated from this task analysis in performing-this validation review. This review will consist of a comparison ~of the SFTA information needs-to the parameters presented on the SPDS.
The current CPS SPDS parameter set has been developed from the j
following:
1.
The original SPDS Design Parameter Set was developed by CPS l
Operations staff personnel using the Emergency Operating Procedures and industry guidance documents available at that.
time; 2.
The original SPDS Parameter Set was subsequently reviewed by the CPS SPDS V&V Team. The V&V review was documented in the "SPDS Parameter Set Validation Report" provided to the NRC as part of the Pre-Implementation Package. -This V&V evaluation was performed using NUREG-0737 Supplement #1 requirements for the associated Critical Safety Functions, and anticipated operator information needs identified from the CPS Final Safety. Analysis Report (FSAR) Chapter 15 Transient / Accident Analysis, the " Reactor. Safety Study" (WASH-1400), and the generic Boiling Water Reactor (BWR) SPDS parameter set discussed in NSAC/21. As a result of this review, several recommendations were made regarding the parameter set, including deleting various parameters and adding others, which were later incorporated into the design; and
[
3.
A reevaluation of the CPS SPDS parameter set was considered 1
appropriate as a result of the December 1984 Design Verification Audit performed by the NRC Staff and SAIC. The 1
l
,,._,m.
-_-..,e.,
Attachm:nt to U-600066 reevaluation took place in January / February 1985 as'part of the SPDS Corrective Action Plan. This reevaluation included the following:
(a) Parameter Set Task Force - this-task force reviewed the existing SPDS parameter. set for consistency with the CPS Emergency Procedure Guidelines and Emergency Operating Procedures (EOPs), NSAC/21, Regulatory Guide 1.97, Rev.
3, and the CPS Emergency Plan Emergency Action Level Initiating Conditions. This effort resulted in a revision to the' parameter set which has been implemented.
The methodology used and the results were documented in a report entitled "SPDS Recommended Parameter Set", as provided to the'NRC via IP Letter U-0830, dated April 11, 1985; (b) Operator Integrated SPDS/EOP Walkthroughs - the purpose of the walkthroughs of selected accident scenarios using static displays on the plant simulator was to evaluate the_understandability and compatibility of the SPDS displays to assist the. operator in monitoring the Critical Safety Function parameters. The displays v verified to have met the established system objectives.
In addition, operator recommendations for improving the display. arrangement, use of color, priority of the parameters, and any additional information needed were found most.useful early in the design phase. The procedures, methodology, and results of these walk-throughs were documented in the report. entitled
" Evaluation of SPDS using the Emergency Operating Procedures in Selected Accident Scenario Walkthroughs";
(c) Operator Questionaires - Questionaires were used to provide operator feedback in the design development. The first questionaire was given to several operators without-making any preference to the preliminary SPDS design.
The purpose of the first questionaire was to identify which parameters the operators considered most important/
-would use most often under a wide range of accident conditions. The results were used in assigning priority to the various parameters. The second questionaire was used to evaluate the preliminary SPDS displays. After completion of the integrated walkthroughs the operators were asked to complete this questionaire. The results of these questionaires were documented in the "SPDS Questionaire 1 and 2 Analysis Report";
'(d) The SPDS displays were subjected to several Human Factors reviews using NUREG-0700 criteria and operator feedback; (e) The V&V Team reviewed the results of the Corrective Action Plan efforts and found that the evaluations performed confirmed the adequacy of the SPDS display / parameter set; and l
l
(.
Attachmint to U-600066 (f) An independent' design review was performed to determine if the overall SPDS design objectives were met by the results of the SPDS Corrective Action Plan. This design review was performed by a committee composed of.IP representatives from Nuclear Station Engineering-Department (NSED)-/ Control & Instrumentation, NSED'-/
Technical Assessment, CPS Plant Technical Staff, Human Factors,.NSED /-Licensing, NSED / Computer Engineering, and Nuclear Support / Emergency Planning. This committee concluded that the design objectives had been met and thus reaffirmed.the adequacy of the results obtained.
This committee review is documented in the "SPDS Design Review Team Report".
As a result of these efforts, the Staff (Mr. Mike McCoy, PSRB) has indicated that the current SPDS parameter. set fulfills-the requirements of NUREG-0737 Supplement #1. The SPDS Dynamic Simulation Test to be performed by IP, as identified in the CPS response to General Comment
- 1, is expected to cenfirm by real-time simulation the adequacy of the CPS SPDS design. Given the extremely extensive reviews performed to date regarding the required information for the operators via SPDS, the' acceptable verification performed.by the V&V process, and the commitment to actual simulation tests, IP considers the CPS SPDS parameter set to fully met the applicable NRC requirements.
l l
l UPSET / SHUTDOWN RANGE G REFERENCE CCf1-f g DENSATE CHAMBER
(+) 400 in l
l I
t
(+) 180in.
U R
o R s RECORD ( TYP! CAL) k I: INDICATE (TYPICAL)
NARRCW/ WIDE / FUEL ZCNE RANGE REFERENCE 2
CONCENSATE CHAMBER J
W 3
NARROW RANGE WIDE RANGE E
O
+ 60 in.
+60 in.
m e
b TYP 4_.
g f
R R
F c:
TA PS-M 4 1
m 1
m E
O WATER E
I LEVEL-G O- -
0---
O lNST IERO ff NARROW RANGE e
p TRANSMITTER O
y U
---(-) ll2.06in e
z e
R I
UPSET / SHUTDOWN RANGE i
i s
TRANSMITTER g
g g
W FUE* (TAF)
(-1862.06 in 7 _ g--y
(-)l60. Coin c
l TYP. 4 TAPS E
WIDE RANGE C
TRANSMITTER e
o 3
k, 3
$_ W s
e m
1 1
S BOTTOM OF ACTIVE FUEL
- -f-1312.061n i
TYP. OF 2 TAP S or 8
FUELZONE RANGE TRANSMITTER I
LEVEL INSTRUMENT RANGES
__,