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Summary of 900520 & 21 Meetings w/C-E in Windsor,Ct to Discuss Instrumentation & Control for Sys 80+.List of Attendees & Meeting Agenda Encl
	ML20055E519
Person / Time
Site:	05000470
Issue date:	06/29/1990
From:	Singh R; Office of Nuclear Reactor Regulation
To:	; Office of Nuclear Reactor Regulation
References
	PROJECT-675A NUDOCS 9007120091
	Download: ML20055E519 (133)
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CESSAR System 80+

APPLICANT:

. Combustion Engineering,.Inc. ~ (CE)

SUBJECT:

SUMMARY

OF MEETING WITH CE ON SYSTEM 80+

4 On May.2O and 21, 1990 r the NRR staff and contractors met with; representatives Lof CE at their offices _in Windsor, Connecticut to discuss the instrumentation and ' control for System 80+'. lists the meeting participants. provides the meeting. agenda..

Following a_brief overview of. System 80+, CE presented the design bases:and-t cdescription of NuplexL80+ Advanced Control Complex and related systems,i and hardware and software qualification. The presentations included a: mock-upt live. demonstration of Nuplex 80+.

Copies of the presentation slides used by CE are provided in Enclosure 3.

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'Overall," tt was _a productive meeting. The staff obtained-a better understanding

,of; System 80+- 1 istrumentation and control which will be helpful in the-review.,

Y of CESSAR-DC. During the meeting, CE agreed.to make the 13 volumes of_Nuplex 80+-

.: reference design documentation available to support the staff's ongoing review..

The' reference documentation is now available at CE'sNRockv111e offices.

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.1 June ~29,'1990-Project No. 675-FACILITY:-

.CES'SAR System 80+

APPLICANT:.

CombustionEngineering,Inc.(CE) j

SUBJECT:

SUMMARY

OF MEETING WITH CE ON SYSTEM 80+

i On~May 20 Land 21-1990, the NRR staff and contractors met with representatives' M

of CE'at their offices in Windsor, Connecticut to discuss the instrumentation-

.and control for System 80+. lists the meeting participants.

1 provides the meeting agenda.

l Following a brief overview of System 80+, CE presented the design bases and -

description of-Nuplex 80+ Advanced Control Complex and= related systems, and-hardware and software qualification. The presentations included a mock-up _

live demonstration of Nuplex 80+.

Copies of the presentation slides used by CE are provided in Enclosure 3'.

0verall, it was a productive meeting. The-staff obtained a better understanding--

of System 80+ instrumentation ~and control which will be helpful in the review.

of CESSAR-DC. During the meeting, CE agreed to make the 13 volumes of Huplex 80+

reference design docurentation available to support-the staff's' ongoing review.

l The refererce docu m. ation is now available at CE's Rockville offices.

Isl

.Rabinclra N. Singh, Project Manager-o Standardization Project Directorate:

Division of Reactor Projects

III, j

IV,~V and Special Projects

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Enclosures:

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'MEETINGiBETWEEN NRR AND CE.0N SYSTEM'80+

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MEETING'BETWEEN NRR AND CE'ON SYSTEM 80+

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WINDSOR, CONNECTICUT MAY 20 AND 21, 1990:

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System 80+ Overview S. Ritterbusch,

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NUPLEX 80+ OVERVIEW 0

NUPLEX 80+ IS AN' INTEGRATED NSSS AND BOP ADVANCED CONTROL COMPLEX THAT MEETS ALL POST-TMI AND CURRENT-LICENSING REQUIREMENTS.

q 0-NUPLEX 80+ PROVIDES MONITORING, CONTROL-AND PROTECT 10N FUNCTIONS'USING REDUNDANT AND DIVERSE

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h0PtFX 80+'DETA11FD DESIGN

-.i NUPLEX 80+ IS-BEING APPLIED TO THE FOLLOWING PROGRAMS:-

EACH IS CONTRIBUTING TO THE DETAILED DESIGN:.

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ADVANCED CONTROL COMPLEX FOR SYSTEN 80+ (EVOLUTIONARY ALWR)'- DOCUMENTATION HAS BEEN SUBMITTED TO THE U.S.

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ADVANCED-CONTROL COMPLEX FOR MHTGR - NEW PRODUCTION REarinR o

ADVANCED CONTROL COMPLEX FOR.EHWR:- NEW PRODUCTION REACTOR o-APPLICATION OF NUPLEX 80+ TECHNOLOGY TO SELECTED SYSTEMS FOR YGN 384 yypggggg+

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i NUP1FX 80+ DESIGN BASIS DEVELOPMENT l

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o THE NUPLEX 80+ DESIGN BASIS WAS ESTABLISHED THROUGH REQUIREMENTS ORIGINATING FROM:

CUSTOMERS - NUCLEAR POWER PLANT OWNERS AND OPERATORS REGULATORS - U.S. NRC, DOE INDUSTRY - EPRI, INPO, IEEE C-E EXPERIENCE - PAST PLANTS 0

THE NUPLEX 80+ DESIGN REPRESENTS A CAREFULLY EXECUTED' TECHNOLOGY EVOLUTION FROM NUPLEX 80 i

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TO IMPROVE PLANT SAFETY:

DIGITAL PROTECTION SYSTEMS WITH CONTINU0US AUTOMATIC TESTING FOUR TRAIN ESFAS IMPROVED MAN-MACHINE INTERFACE 3.

TO IMPROVE PLANT AVAILABILITY:

FAULT TOLERANT CONTROL SYSTEMS PRE-TRIP CONTROL ACTIONS POWER DEPENDENT PROTECTION LIMITS IMPROVED MAN-MACHINE INTERFACE 4.

'T0 lMPROVE THE COST EFFECTIVENESS OF NUCLEAR PO S GENERATION:

LOWER CONSTRUCTION COSTS SHORTER DESIGN AND CONSTRUCTION SCHEDULES LOWER OPERAT10N AND MAINTENANCE C0STS gypggxff+

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NUPLEX 80+ SAFETY IWROVEENTs 1

0 PLANT PROTECTION SYSTEM PROVIDES CONTINUOUS SOFTWARE EXECUTION TO VERIFY TRIP LOGIC FUNCTIONALITY

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0 PLANT PRCTECTION SYSTEM PROVIDES EVENT BASED SEGMENTATION WITHIN EACH CHANNEL 1

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MININIZE SEQUENCING TIE o

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HALDEN REACTOR PROJECT PWR SIMULATOR j

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E SLIDE 127. DOC 49'
-i l
NUPtfX 80+ SOFTWARE QUALIFICATION y
.j i
'o-VERIFICATION AND VALIDATION OF USER DEVELOPED APPLICATIONS SOFTWARE IN ACCORDANCE WITH:
1
-ANSI /IEEE - ANS!- 7.4.3.2 R
REGULATORY GUIDE 1.152 l
1 l
o HIGH-LEVEL LANGUAGES ARE USED TO THE MAXIMUM EXTENT j
L PRACTICAL:
1
,C' l
LADDER LOGIC i l '.
0-COMPILERS AND MACHINE OPERATING. SYSTEMS ARE 0UAL'IFIED-BASED ON PRIOR FIELD 0PERATING EXPERIENCE AND V8V AT THE APPLICATIONS LEVEL-
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SOFTWARE-QUALIFICATION i
i o
SOFTWARE REQUIREMENTS DOCUMENTATION IS WRITTEN,: THEN VERIFIED THROUGH INDEPENDENT REVIEW.
t
+
.O STANDARD-SOFTWARE' MODULES ARE WRITTEN AND DEBUGGED USING.
DEVELOPMENT SYSTEM' EQUIPMENT.
YFRIFICATION:IS THROUGH INDEPENDENT REVIEW 0F CODE FOR E6 d MODULE.
i o
STANDARD SOFTWARE MODULES ARE INTEGRATED AND DEBUGGED USING
. DEVELOPMENT SYSTEM.
o UPON COMPLETION, STANDARD SOFTWARE CAN BE SHIPPED TO THE' SITE TO-PERMIT EARLY FUNCTIONAL VERIFICATION OF INSTALLED HARDWARE.
o FINAL SOFTWARE MODULES ARE WRITTEN,. INTEGRATED AND DEBUGGED u
WITH INDEPENDENT VaV AS BEFORE.
INTEGRATED-TESTING-IS
=
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m IF--AVAILABLE.
l' 0:
AT SITE,-QUALIFIED SOFTWARE-IS-LOADED IN INSTALLED EQUIPMENT AND FINAL-ACCEPTANCE TESTING IS PERFORMED.
I o ' REVISION LEVELS OF APPLICATION SOFTWARE, OPERATING SYSTEM SOFTWARE AND HARDWARE ARE CONTROLLED THROUGHOUT-THE' LIFE OF.
THE EQUIPMENT THROUGH STRICT CONFIGURATION: CONTROLS.
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QUALIFICATION STANDARDS 1
Li HARDWARE
\\
IEEE-323-1983 IEEE STANDARD FOR QUALIFYING CLASS'1Ei EQUIPMENT ~FOR NUCLEAR POWER GENERATING STATIONS a
1
- IEEE-344-1987 RECO N NDED PRACTICES FOR SEISMIC QUALIFICATION OF CLASS 1E EQUIPMENT-
.l l
FOR NUCLEAR POWER GENERATING j
STATIONS' IEEE-472-1974 GUIDE FOR SURGE WITHSTAND-CAPABILITY FOR TESTS IEC-801-ELECTRICAL MAGNETIC COMPATIBILITY FOR INDUSTRIAL' PROCESS MEASUREMENT'
-AND CONTROL < EQUIPMENT-4 i
' MIL-STD-461 ELECTROMAGNETIC EMISSION AND A
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.FERENCE 1
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A REPRESENTATIVE DESIGN IS BUILT USING ALL HARDWARE MODULES.
L.
l 0-PERFORMANCE TEST-SOFlyARE IS LOADED-AND HARDWARE SEISMIC l
TEST IS PERFORMED.
Im o-ALL IN-SERVICE HARDWARE CONFIGURATIONS ARE ANALYZED USING.
3 L
< SEISMIC TEST RESULTS.
-it'-
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' EQUIPMENT MANUFACTURER'S SPECIFICATIONS AND: TEST DATA ARE-
' ANALYZED FOR QUALIFICATION:IN:0THER AREAS'(I.E.,.
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SUPPLEMENTAL TESTING IS PERFORMED, IF REQUIRED
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MIBE g
a In' general terms, the Assembly or System failure rate is the summation of the l
individual failure rates of'all the independent elements of the Assembly or-
'l System. The MTBF is the reciprocal of the.fallure rate, j
M in general terms, The Mean Time To Repair is the total Assembly or System -
W.
repair time divided by the total number of failures.
Availabilitv q
Availability = MTBF / (MTBF + MTTR) X 100 in E i
Unavailability = 1 A' ailability.
v
' System Calculations.
a.
Generally, all systems are reduced to simple: series or parallel elements and computations then performed. If circumstances warrant, special consideration can be.given to any system configuration. The following approaches, however,
{
are the general case.
Series Elements Series Availability (As) = (A1) (A2) (A3)......(An).
MTBF = 1/(1/MTBF1 + 1/MTBF2.+ + + 1/MTBFn).
MTTR = MTBF (1 As)/As, b,
Parallel Elements Unavailability = 1 Availability.
l Parallet Unavailability (Up) = (U1)-(U2) (U3)....... (Un).
Parallel Availability (Ap)~= 1 Up.
,p-MTTR = 1/(1/MTTR1 + 1/MTTR2 + + +.1/MTTRn).
~'
J MTBF = MTTR(Ap)/(1 Ap).
' Table 2.1 provides sample MTBF and MTTR-data' for several. basic (i.e.,
-. nucleus) computer systems. Both' single -(CPU) ~and dual (CPU /IPU) i
. configurations are included for the CONCEPT 32/67,32/87 and 32/97 product lines.
i Combustion Engineering, Inc.
22 l
Gould CSD 9
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arei a valuable tool in design, maintenance / spares planning, system-configuration and evaluation.
Field performance of Gould CSD products consistently demonstrate reliabil;ty which exceeds that estimated by a factor of 2:1 or greater. The use of MIL-HDBK 217D as a common denominator, coupled -
with conservative assumptions in the areas of duty cycle, quality levels ~ and; j
worst case configurations, provide the user with estimates that are both uniform-t and traceable.
J\\,
Assumptions.
D The following assumptions are utilized in system calculations.
l Assembly predications are generated considering all elements in a series configuration.
. System calculations are performed by developing reliability block diagrams which reflect the actual configuration of the system' (series, parallel, series / parallel, with/without repair, etc.).
IC failure rates are predicted by means of Section 5.1.2 of MIL-HDBK 217D. Bit and gate counts are obtained from MIL HOBK-
. 217D as well as from vendors and RADC Handbook MDR 18.
Failure data on all other components'is derived from MIL-STD-t 217D.
The Ground Benign environment ( '=1) is used throughout.
f
(
Calculations are based upon +35 degrees C ambient temperature.
.i H.
The duty cycle of all hardware, except.for memories, is assumed to be 100%. Memory devices have duty cycles which are defined by :
L:
the configuration of the memory assemblies.
In order to inject worst case conditions, peripheral devices are l
assumed to experience 100% duty cycles.' Failure rates are generally supplied by the vendor.
A quality level of B 2 is used throughout.
The effect of using Error Correcting Codes (ECC) are considered
+
in calculations of memory devices.
Combustion Engineering, Inc.
21 Gould CSD
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FUNCTIONAL GROUP CONTROL y
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Reliability Andysis Shows the CCS Functional Group Control to be More Reliable than Traditional Single Loop Control I
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ML20055E519

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