ML20238C234
| ML20238C234 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/31/1987 |
| From: | Hunsader S COMMONWEALTH EDISON CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-1002 3520K, NUDOCS 8709090718 | |
| Download: ML20238C234 (23) | |
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'N Commonwealth Edison o
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/ One First N:tional Plaza. Chicago, Illinois T
(j} Address Reply to: Post OffKa Box 767
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Chicago, Illinois 60690 0767 August 31, 1987 Mr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Braidwood Station Units 1 and 2 Detailed Control Room Design Review (DCRDR)
Safety Parameter Display System (SPDS)
NRC Docket Nos. 50-456 and 50-457 Reference (a): Jatne 23, 1987 S.C. Hunsader letter to T.E. Murley (b): NUREG-1002, Safety Evaluation Report, Supplement No. 4, dated July, 1987
Dear Mr. Murley:
Reference (a) summarized specific DCRDR and SPDS items and provided Commonwealth Edison's committment to submit a response addressing each item. Reference (b) included the NRC Staff's review and acceptance of reference (a).
Attached to this letter you will find the report " Commonwealth Edison Company, Braidwood Station, Safety Evaluation Report (NUREG-1002, Supplement 4) Section 18, Response". This report addresses the items presented in references (a) and (b).
Please address any questions concerning this matter to this office.
Very truly yours,
-( f S. C. Hunsader Nuclear Licensing Administrator cc:
S. Sands Braidwood Resident Inspector 3520K B709090710 070831 ADOCK0500g6 DR l
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE.
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I' COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE INTRODUCTION Commonwealth Edison Company (CECO) submitted the Final Summary Report for Byron /Braidwood Stations Units 1 and 2 Detailed Control Room Design Review (DCRDR) to the. Nuclear Regulatory Commission (NRC) in December 1986, as required by Supplement 1 to NUREG-0737.
The NRC conducted a DCRDR and SPDS Site Audit of the Byron and Braidwood Stations, at Braidwood, on March 10th and lith, 1987.
A June 23, 1987 conference call between the NRC and Ceco is documented in a letter to T.E.
Murley from S. C.
Hunsader.
The Braidwood Safety Evaluation Report (SER)
Supplement 4 dated July, 1987 was subsequently transmitted by the Nuclear Licensing Administrator's letter S.C.
Hunsader NL-87-0935.
Section 18 (Human Factors Engineering) of the above (SER) identifies the remaining open items with respect to the Main Control Room and Remote Shutdown Panel (Section 18.2) and the Safety Parameter Display System (Section 18.3).
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COMMONWEALTH EDISON COMPANY'S RESPONSE Section 18.2 of the Braidwood Station SER statest "On the basis of a preimplementation site audit conducted on March 10-11, 1987, the staff concluded that all requirements of Supplement 1 to NUREG-0737 had been satisfactorily completed except for :
- 1) the evaluation of the long term engineering solution to the problem of radio transmitters that can inadvertently activate safeguard syrtems; 2) the evaluation of the lack of lamp test capability; 3) the determination of a long-term engineering solution to the indicator light bulb burnout problem; 4) the documentation of a detailed color versus use matrix for control room color coding and a commitment to make color coding in all control room contexts consistent with the green board concept; and 5) the documentation of a clarification of the licensee's response to HED No. 0165 regarding computer printer speed, in order to provide assurance that data will not be lost during a major event.
In its letter dated June 23, 1987, the licensee committed to submitting proposed resolutions to these five items by August 31, 1987.
The staff finds this schedule acceptable.
This is confirmatory Issue A(10)."
Section 18.3 of the Braidwood Station SER states:
"On the basis of a preimplementation site audit conducted on March 10-11, 1987, the staff concluded that all requirements of Supplement 1 to NUREG-0737 had been satisfactorily completed except for several operational problems with the SPDS.
On the wide-range iconic screen, four of the eight parameters varied from the expected value enough to detract from the display.
These were as follows:
(1)
The reactor vessel level indication system (RVLIS) was reading properly but the value spiked to the alarm level periodically.
(2)
The radiation level remained at the alarm setpoint.
l (3)
The steam generator level was reading far enough from the l
setpoint to be distracting.
(4)
The containment pressure was indicating far enough from the setpoint to skew the iconic display.
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i On the narrow-range screen, the following three readings were j
' off the setpoints:
l (1) high radiation alarm l
(2) containment temperature i
L (3) steam generator level j
In its letter dated June 23, 1987, the licensee committed to submitting proposed resolutions to these items by August 31, I
1987.
The staff finds this schedule acceptable.
In its l
1etter dated October 10, 1986, the licensee committed to l'
submitting the verification / validation report by August'1, 1987.
Confirmation that the SPDS operational problems have been corrected and submittal of the verification / validation L
report constitute Confirmatory Issue A(11).'
l CECO's response to these Confirmatory Issues are contained in 1
l the following 18.2 and 18.3 sections and are arranged in the same sequence as the issues were presented in the SER.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4) d SECTION 18 RESPONSE
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18.2 Main Control' Room and Remote Shutdown Panel Responses 4
Open Item 1:
"The evaluation of the long' term engineering
. solution to the problem of radio,, transmitters l
that can inadverten'tly activate wnfeguard jp systems" CECO RESPONSE 3 b
(4 RFI History h
s Radio frequency interference problems first became apparent
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+during hot functional testing of' Byron Unit 1.
This'was the first time that most.of the instrumentation had been placed in y
i service under simulated operating conditions.
A= number of malfunctions were observed which were related to RFI.
g Meetings were immediately scheduled to discuss these problems and field testing began.
The illusive nature of these RFI probleme quickly became apparent as test.results proved to be inconsistent.
Malfunctions caused by walkie-talkies at a f
distance of 25 feet"during one test could not be duplicated o
during an identical test two weeks later with the walkie-
, talkies held G inches from affected instrumentation.
Furthermore, what seemed at first to be obvious solutions, such.as adding screen shields around the affected instruments, proved by test to be ineffective and in sor.ae cases actually enhanced the RFI. problem.
At this point it was clear that quick solutions to these RFI problems would not be possible.
Discussions with representatives from other operating stations (Zion, LaSalle) confirmed the unique nature of these problems.
While certain m
aspects of RFI problems are similar in nature, resolutions must involve a detailed analysis of all factors for each' case.
Slight differences in the physical arrangement, frequency, instrumentation, etc., can mean the difference between a problem and no problem.
Therefore, it is difficult to draw on the experience of other stations, or even'other problems in the same station, as a basis for a particulan solution.
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WithLthe above:in mind, it~was determined that the.only way'to prevent serious problems'was to ban walkie-talkies in the-vicinity of critical instrumentation, while continuing efforts to solve the RFI problems.
Schematic drawings were reviewed to determine which instrumentation could-have an adverse effect on the Unit.
Drawings were quickly; prepared locating this critical instrumentation, and these drawings.were used as a basia to mark areas where. radio transmissions were not allowed for: Byron Unit 1.
Concurrently, CECO's System Operations Analysis Department-(SOAD) was preparing to conduct various. tests using hardware from Byron in an attempt to find a better solution.
The test results of this program were very informative.
One surprising result _was the fact that the PVC coated flexible conduit in use at Byron provides very little shielding, and in some cases can actually amplify the problem.
The tests also indicate that the Barton transmitters with unshielded pigtails
.were particularly susceptible to RFI.
Conclusions Since it was determined that the unshielded pigtails on certain-Barton transmitters and the PVC coated flexible conduit were a major weakness relative to RFI, or for that matter, any electro-magnetic disturbance,.it was decided to replace-the unshielded pigtails with shielded instrumentation cable and to replace the PVC coated flexible conduit with BOA stainless steel flexible conduit.
These changes were' implemented on Byron Unit 2 and Braidwood Unit 1 and 2.
However, since Byron Unit 1 was in the startup mode, it was not possible to completely incorporate these changes.
Additional Field Testino RFI testing has since been performed for Byron Unit 2 and Braidwood Units 1 and 2 during hot functional testing of each unit.
.In general, these additional field tests seem to indicate that the replacement of the unshielded pigtails and the use of BOA flexible conduit has greatly reduced the sensitivity of these instruments to RFI.
However, some inconsistencies in the test data still remain, thus confirming the unpredictability of RFI, particularly above 100 megahertz.
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Lena-Term Solution It has.been concluded.that while the modifications already en
-~ incorporated have greatly improved the RFI prob 1em,,there is
}J no. prar:tical method available that will completely,. eliminate this t,oncern.
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Therefore, administrative'qontrol will continue to be applied...
to prohibit radio transmission in sensitive steam.
'}2 Administrative control is a valid mechanism for addressing this concern.
The success of this program to date is y
evidenced by the lack of RFI related problems during operation at the Byron station, and is expected to be Just as successful at Braidwood.
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Accept as is.
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's COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE 18.2 Main Control Room and Remote Shutdown Panel Responses Open Item 2:
"The evaluation of the lack of lamp test capability" CECO RESPONSE t
At the March 10-11th Site Audit the NRC remarked that Susquehanna had identified a similar problem during the conduct of its DCRDR and that it ha6 developed a new indicator
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bulb that employed Light Emitting Diodes (LEDs) that had a longer life expectancy than incandescent light bulbs.
Preliminary testing by the plant and the bulb's manufacturer
-indicated t.he new LED bulb would last in excess of eleven 3
years.
Vith such a long life span the necessity of a bulb test capability becomes moot because the bulbs can systematical y be replaced at a time interval shorter than their postujated maximum thereby ensuring bulb operability.
Susquehanna provided the name of the manufacturer who was contected.
LED bulbs suitable for use at Byron and Braidwood Stativos can be obtained in a recsonable time frame.
A sample of bulbs will be purchased ar.d the feasibility of use in the control rooms will be evaluated.
Eactors such as luminance levels, contrast levels, power consumption, heat generation,
, fire code adherence, and seismic qualifications will be "avaluated.
Should the new bulbs prove suitable, they will be q.gstalled in the control room where applicable.
IMPLEMENTATION By the completion of the First Refueling Outage.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION ~
SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE 18.2 Main Control Room and Remote Shutdown Panel Responses Open Item 3:
"The determination of a long-term engineering solution to the indicator light bulb burnout problem" CECO RESPONSE At the March 10-11th Site Audit the NRC remarked that Susquehanna identified a similar problem during.the conduct of its DCRDR and.that it had developei a new indicator bulb that employed Light Emitting Diodes (LEDs) that had a longer life expectancy-than incandescent light bulbs.
Preliminary testing by the plant and the bulb's manufacturer indicated the new LED bulb would last in excess of eleven years.
Susquehanna provided the name of the manufceturer who was contacted.
LED bulbs suitable for use at Byron and Braidwood statior4s can be obtained in a reasonable time frcme.
A sample of bulbs will be purchased and the feasibility of their use in tne control rooms will be evaluated.
Factors such as luminance levels, contrast levels, power consumption, heat generation, fire code adherence, and seismic qualifications will be evaluated.
Should the new bulbs prove suitable, they will be installed in the control room where applicable.
IMPLEMENTATION By the completion of the First Refueling Outage.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE
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18.2 Main Control Room and Remote Shutdown Panel Responses Open Item 4:
"The documentation of a detailed color versus use matrix for control room color coding and a l
commitment to make color coding in all control room contexts consistent with the green board concept" CECO RESPONSE The current use of color at Byron and Braidwood for the Main Control Board (MCB), Safety Parameter Display System (SPDS),
Prodigy Graphics (PROD), and Rad Monitor (RM-11) is documented below.
BYRON /BRAIDWOOD COLOR USE MATRIX COLOR CODE LOCATION
- RED Abnormal MCB High Alarn.
SPDS High-Active Tic Mark SPDS Invalid SPDS High Alarm-Analog PROD On, Open, Energized PROD Open-Valves PROD; BWD only On-Pumps PROD; BWD only Abnormal-Pumps, Valves PROD; BY cnly and Digital Status Downwind-Met. Tower PROD Alarm SER PRINTER GREEN Normal MCB Highlight-Static Objects PROD Off, Closed, Not Energized PROD No Flow PROD Closed-Valves PROD; BWD only PROD; BY only Off-Pumps PROD; BY only Normal-Pumps, Valves and Digital Status Normal Operations RM-11 Not Downwind-Met. Tower PROD 6
BYRON /BRAIDWOOD COLOR USE MATRIX l
COLOR CODE LOCATION
- AMBER High Alarm-Rad SPDS Alert Alarm RM-11 i
I Info-Labels RM-11 1
BLUE Not Running MCB Water Flow PROD Operate Failure RM-11 WHITE Tripped MCB Default SPDS Normal Values SPDS Clock SPDS Loop, SG, and Other SPDS Designations Updates-Lettering PROD Normal-Analog PROD Poll Status RM-11 CYAN Spokes SPDS Low-Tic Mark SPDS High-Tic Mark SPDS Reference Values SPDS Field Overflow SPDS Suspect Datu SPDS Poor Data - PPP SPDS Bad Data - XXX SPDS Invalid Data SPDS Value too Large SPDS Lettering PROD Static Objects PROD Suspect Values PROD Bad Values - XXX PROD Sensor 0.0.S.
PROD Valves & Pumps Equipment Failure RM-11 Component Labels RM-11 1/2 CYAN Control Functions RM-11 MAGENTA Low-Alarm SPDS Low-Active Tic Mark SPDS Low Alarm-Analog PROD Communications RM-11 7
BYRON /BRAIDWOOD COLOR USE MATRIX COLOR CODE LOCATION
- YELLOW Caution-Zone Band MCB Highlights-Lettering PROD and Static Objects Iconic SPDS RED / GREEN Intermediate Valve MCB Positions ORANGE Open Solenoid Valve MCB NIS Conditions Energized MCB -PNL PMO7J BLACK Return from Alarm SER PRINTER Channel Numbers RM-11
- Location at both Byron and Braidwood, unless otherwise specified.
The Byron /Braidwood main control boards use green board color coding conventions whereas the Ceco computer display conventions currently do not use green board coding.
The feasibility of using green board coding for the computer graphics has been explored at both Byron and Braidwood.
At Byron, the plant status, pressurizer, and reactor coolant graphic displays have been modified to the green board coding consistent with the main control board.
However, the green board coding does not provide for indication of equipment status beyond normal / abnormal.
The CECO computer display conventions will be modified to green board conventions for Byron and Braidwood and these conventions will be implemented on the graphic displays.
The use of shape coding to provide equipment status information in addition to color coding will also be explored.
IMPLEMENTATION By the completion of the Second Refueling Outage.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (HUREG-1002, Supplement 4)
SECTION 18 RESPONSE l
18.2 Main Control Room and Remote Shutdown Panel Responses Open Item 5:
"The documentation of a clarification of the licensee's response to HED No. 0165 regarding computer printer speed, in order to provide I
assurance that data will not be lost during a l
major event" l
l CECO RESPONSE 1
The control room printers originally installed for the process computer, sequence of events recorder, and radiation monitoring computer have been upgraded from a 60 line per minute printer to a 150 line per minute printer.
The increased speed of the printers ensures that all pertinent data is printed soon after a transient.
Reactor trip data parameters saved by the process computer are printed at a 300 line per minute printer located adjacent to the control room.
Buffering provided by all computers within the control room ensures that no alarms are lost.
Based on current printer operation the first 150 alarms occuring after a transient will be printed within one minute of the actual occurence of the transient.
IMPLEMENTATION Accept as is.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE 18.3 Safety Parameter Display System Responses Wide Range Operational Problem 1:
"The reactor vessel level indication system (RVL2S) was reading properly but the value spiked to the alarm level periodically" CECO RESPONSE Evaluation and analysis of the SPDS vs. control room data subsequent to the NRC site visit of March 10-11th revealed that the " spiking" observed on the RVLIS spoke was not due to a fault in the SPDS program but rather to the graphics computer driving the system.
The CECO process computer graphics system was designed to display a suspect value on the screen when the graphics system had not received a "recent" value for a particular dynamic item in the display after 180 seconds.
However, in general, the process computer only sends dynamic data that has changed since the previous exchange in order to decrease the amount of information exchanged between j
the process and graphics computers.
The exception to this
" rule" is that the process computer will send a value for a dynamic data point when that point has remained static for 20 update cycles.
Value updates are requested to be sent every 5 seconds, so the maximum time between value transmissions to the computer is approximately 100 seconds ( 20 update cycles x 5 seconds between cycles = 100 seconds).
Because the process j
computer tasks have been prioritized, actual value updates could be more than 5 seconds apart, with the actual time a function of the demand being placed upon the process computer.
The " spiking" observed by the NRC on the RVLIS active line was
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caused by the graphics system " timing out" on the active value for the spoke with the process computer sending an updated value Just after the graphics system had marked the value as suspect.
Though thin anomaly occurs infrequently, the i
graphics system will be modified so that the " timing out" l
I problem is minimized.
IMPLEMENTATION By the completion of the First Refueling Outage.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE 18.3 Safety Parameter Display System Responses Wide Range Operational Problem 2:
"The radiation level remained at the alarm setpoint" CECO RESPONSE At the time of the March 10-11 audit, the radiation spoke was deflected to its outer limit; the tic mark was red and the text ("B/D") was cyan.
SPDS is programmed to only display validated data.
Suspect data is i.ot used in SPDS calculations.
The color cyan is used to indicate to operators a suspect value which has failed a validation test.
The suspect values are deflected to the high alarm li-it on the spokes, and the active value is displayed in cyan.
Thus, the spoke deflection at the high alarm setpoint for the radiation spoke indicated to operators that the computer had received a suspect steam generator blowdown radiation value.
Thus, the spoke deflection provided information to operators as intended by the design of the SPDS.
The steam generator blowdown radiation detector failure which resulted in the suspect values will be remedied.
IMPLEMENTATION By completion of the First Refueling Outage.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE 18.3 Safety Parameter Display System Responses Wide Range Operational. Problem 3:
"The steam generator level was reading far enough from the setpoint to be distracting" CECO RESPONSE This issue is a non problem.
When the NRC conducted its site visit at Braidwood, Unit 1 was in the throes of final construction turn-over testing, pre-operability testing and startup testing.
Some systems were out of service while parameters in other systems were deliberately
" abnormal" as part of the ongoing tests.
A review of the SPDS and the Unit's status during the site visit indicates that the SPDS was performing as programmed and designed.
The reference value for the steam generator level spoke can only be one of two values depending upon the average TAVE of the four reactor coolant loops (RCL).
At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on the day of the NRC site visit RCL average TAVE was 557.1 degrees fahrenheit.
At this value of average TAVE SPDS uses 66% as the reference level for the Wide Range steam generator level.
At 0911 hours0.0105 days <br />0.253 hours <br />0.00151 weeks <br />3.466355e-4 months <br /> the RCL average TAVE fell below 557 degrees fahrenheit and remained at that level for the duration of the NRC's review.
With average RCL temperature below 557 degrees fahrenheit the SPDS uses 68% as the reference for the Wide Range steam generator level.
Our records indicate that because of support system testing the level in steam generator 1A was below normal, or referent, level during most of the NRC's visit.
Thus, the iconic reading was deflected from the
" normal" setpoint.
The SPDS program on the process computer was performing as designed in that it indicated an " abnormal" steam generator condition, which at the time was the case.
No remedial action is anticipated.
IMPLEMENTATION Accept as is.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION
' SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSE
' 18.3 Safety' Parameter Display System Responses Wid.e Range Operational Problem 4:
'The containment pressure was indicating.far enough from the metooint'to skew.the iconic display" t
CECO RESPONSE When'the NRC conducted its site visit at Braidwood,- Unit 1 was in the process of final construction turn-over testing, pre-operability testing and startup' testing.
Some systems were out of service while parameters.in other systems were deliberately " abnormal" as part of the ongoing tests.
A review of the SPDS and the Unit's status during the site visit indicates that-the SPDS was performing as programmed and designed.
The containment pressure at the time of the visit was 0.6 PSIG, well within the. normal range..Since the referent'for this spoke is 0. 0, the value for the parameter.was accurate and correctly displayed.
The SPDS program functioned as designed and no remedial.
action is required.
IMPLEMENTATION Accept as is.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSES 18.3 Safety Parameter Display System Responses Harrow Range Operational Problem 1:
"The h1gh radiation alarm readings were off the setpoint" CECO RESPONSE At the time of the March 10-11 audit, the radiation spoke was deflected to its outer limit; the tic mark was red and the text ("B/D") was cyan.
SPDS is programmed to only displey validated data.
Suspect data is not used in SPDS calculations.
The color cyan is used to indicate to operators a st r;pect value which has f ailed a validation test.
The suspect values are deflected to the high alarm limit on the spokes, and the active value is displayed in cyan.
Thus, the spoke deflection at the high alarm setpoint for the radiation spoke indicated to operators that the computer had received a suspect steam generator blowdown radiation value.
Thus, the j
spoke deflection provided information to operators as intended
'j by the design of the SPDS.
The steam generator blowdown radiation detector failure which resulted in the suspect values will be remedied.
IMPLEMENTATION j
By completion of the First Refueling Outage.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSES 18.3 Safety Parameter Display System Responses Harrow Range Operational Problem 2:
"The containment temperature readings were off the setpoint" CECO RESPONSE On the Narrow Range SPDS display the Containment spoke displays two parameters, Containment Temperature and Containment Sump Level.
A review and analysis of the SPDS and the Unit's status during the NRC's site visit indicates that containment temperature was 110 degrees
. fahrenheit, within the normal range for that parameter for the unit's operating mode.
However, containment sump level was low at 21 inches, off the normal level of 41 inches.
The SPDS program on the process computer was performing as designed in that it indicated a low containment sump level.
The low level was due to final construction turn-over testing, pre-operability testing and startup testing that was ongoing at the time of the NRC audit.
No remedial action is required.
IMPLEMENTATION Accept as is.
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I COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT (NUREG-1002, Supplement 4)
SECTION 18 RESPONSES 18.3 Safety Parameter Display System Responses Narrow-Range Operational Problem 3:
"The steam generator level readings were off the setpoint" CECO RESPONSE At the time of'the March 10-11 audit Braidwood Unit 1 was in the Cold Shutdown Mode and in the startup testing process.
The steam generator 1A level reading was 51%.
The steam generator reference value was 66X.
The SPDS graphic display uses the most deviant steam generator for the Steam Generator Level spoke.
The graphic display at the time of audit reflected the deviant level value for Steam Generator 1A, as intended by the design of the SPDS.
The steam generator level readings conform to the octagon shape when steam generator level readings do not deviate from the reference values.
IMPLEMENTATION Accept as is.
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COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION SAFETY EVALUATION REPORT-(NUREG-1002, Supplement 4)
SECTION 18 RESPONSE 18.3 Safety Parameter Display System Responses Submittal of (1) Verification &' Validation (V&V) Report and
.(2) photographs of.Braidwood SPDS displays at a power level of 30% or greater.
CECO RESPONSE Item 1:
During the March '87 ons1te audit-it'was stated that the Safety Parameter Display System V&V program was expected to have been completed by August 1, 1987.
That program is an iterative process which is near completion at this time.
A report will document the V&Y program when that process is complete.
Item 2:
During the March '87 onsite audit the NRC indicated that photographs of the Braidwood SPDS displays under normal conditions of power would be adequate to document closing of those open items.
Since Braidwood Unit 1 is presently in the startup testing stage,-we do not have those photographs at this time.
However, the Section 18.3 CECO Responses confirm that most of the SPDS " operational problems" have been corrected.
The remaining problems will have been corrected by the First Refueling Outage.
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