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by W. Kerr on 10/10/91
ACRS SUBCOMMITTEE MEETINGSUMMARY/MINUTES OF THE INSTRUMENTATION
AND CONTROL SYSTEMSAUGUST 29, 1991
BETHESDA, MARYLAND
INTRODUCTION
The ACRS Subcommittee on Instrumentation and Control Systems held
a meeting on August 29, 1991 in Bethesda, Maryland.
The entire subcommittee meeting was open to public attendance. The
documents submitted to the subcommittee Members for this meeting
are attached.
ATTENDEES NRCACRS
W. Kerr, ChairmanH. Lewis, MemberC. Wylie, MemberM. El-Zeftawy, Staff
OTHERS
K. Barrett, BechtelC. Tuley, KJ. Trotter, EPRIK. Klaube, NUSP. Ludgate, NUSC. Swanson, B&WG. Srikantiah, EPRIR. Krich, PE Co.G. Salamon, FP&LS. Palena, PE Co.J. Armstrong, PE Co.L. Hopkins, PE Co.G. Stewart, PE Co.
W.R.T.R.W.G.E.J.D.A.J.C.
Butler, NRRClark, NRRKenney, SRIVan Houton, SECYSwenson, NRRGalletti, NRRImbro, NRRJacobson, NRRNorkin, NRRGautam, NRRGallagher, NRRDoutt, NRR
I-ler I? -S-
9204130127 911010 -PDR ACRS2766 PDR I/I,
I
Inst & Cntl 2Mtg. (8/29/91)
Meeting Highlights. Actions. Agreements
1. Dr. Kerr, Subcommittee Chairman, convened the meeting and
stated that the purpose of the meeting is to discuss with theEPRI representatives, representatives of Philadelphia Electric
(PE) Company, and the NRC staff the following items,respectively.
* A set-point methodology developed by EPRI
* Transient Response Implementing Plan (TRIP) procedures
developed and implemented by PE Co.
* Electrical Distribution System Functional Inspections(EDSFIs) currently being performed by the staff.
2. Mr. Srikantiah, EPRI, briefed the subcommittee Members
regarding an EPRI procedure for determining the set-points forsafety systems. He described a generic process forestablishing appropriate and justifiable limits for the
reactor monitoring and protection systems set-points. This
process was designed to be applicable to utilities with either
PWRs or BWRs who plan to assume primary responsibility for
performing safety analysis and set-points determination.
Mr. Srikantiah stated that utilities are familiar with thegeneral aspects of set-point methodology, as well as theresults of its use in reactor operation. However, the methods
and assumptions used by vendors to determine set points are
not readily available for utility use. A plant's set points
depend on the operational goals of the licensee and are
derived from a consideration of safety analyses andperformance considerations. The performance evaluation takesaccount of the constraints on plant operation, given the
limitations imposed by the safety analysis. In turn, the
safety analysis verifies the acceptability of the assumed set
points. The safety analysis and the determination of set
points are interrelated processes.
Inst & Cntl 3Mtg. (8/29/91)
An important step in the determination of acceptable set-
points is the treatment of the uncertainties. These include
uncertainties inherent in the actual equipment settings and
those in the inputs to the license basis analysis and plant
performance evaluations. The estimated values have an
associated uncertainty distribution.
EPRI's set-point process is related to the process used in
the final safety analysis report (FSAR) design analyses. The
FSAR focuses on plant response to design basis events (DBEs)
while EPRI's set-point process focuses on monitoring and
protection system set-points used in the analyses of the
limiting cases in order to assure that acceptable consequences
result.
Verification of the monitoring and protection system limits
is accomplished through an analytical demonstration of
acceptable consequences of DBEs. The analysis plays a central
role in the determination of set-points. The results of the
event analyses are compared to the event acceptance limits.
If the results are acceptable then the set-points are
considered acceptable. If the results are not acceptable,
then a margin recovery process is undertaken which may include
changing set-points, changing the operating envelope, changing
model inputs, changing analysis assumptions and/or methods,
or changing the plant design. An alternate approach developed
by EPRI is based on what is described as a more realistic
treatment of uncertainties. The approach involves a
statistical combination of uncertainties (SCU). EPRI has
developed guidance for determining the appropriate event
analysis model inputs from the plant design and determining
the operating envelope sources.
Inst & Cntl 4Mtg. (8/29/91)
The following are EPRI's guidelines for LWR set-point
determination;
A. Complete the plant performance evaluation.
B. Identify the plant specific DBEs which must be analyzed
and the associated acceptance limits.
C. Identify the plant design model inputs and operating
envelope limits to be used in the event analyses.
a) Plant design model inputs
* From the plant design determine the numerical
ranges for the model inputs.
* Include in the ranges the appropriate
uncertainties.
b) Operating envelope limits
* From the monitoring system determine the
numerical ranges for the operating envelope
limits.
* Include in the ranges the appropriate
uncertainties.
* To these ranges apply any required maneuvering
to determine the analysis input ranges over
which the event analyses must be valid.
* Select axial power shapes.
D. Model the protection system in the event analyses.
* Identify trips to be used.
* Determine the protection system instrument set-
points.
* Apply appropriate uncertainty allowances to the
instrument set-points.
* Account for event effects through appropriate
allowances or explicit modeling to determine
the analytical set-point for the event
analyses.
E. Perform the event analyses insuring that limiting cases
have been analyzed for each event and all acceptance
Inst & Cntl 5Mtg. (8/29/91)
limits have been satisfied.
F. Implement margin recovery process if required.
* Change the protection system instrument set-points.
* Change the operating envelope.
* Modify the event analyses models.
* Change the plant design.
* Apply statistical treatment of uncertainties.
Mr. Srikantiah stated that EPRI's statistical combination of
uncertainties methodology can be used to reduce excessive
conservatism inherent in the existing deterministic analysis
process. Use of EPRI's method is expected to decrease the
number of scrams and to eliminate some actuations of safety
and relief valves, thus increasing operating flexibility. Mr.
Srikantiah cited some applications of the EPRI's set-point
methodology, such as Oyster Creek high-pressure set-points,
Vermont Yankee reload analysis, Diablo Canyon safety valve
tolerance, TMI-1 flux/flow set-points, and Palo Verde safety
valve tolerance.
Dr. Kerr commented that EPRI's approach appears to be
plausible but complicated. Further several steps employ
"engineering judgement" making the results unpredictable.
EPRI representatives agreed that given the same data, two
independent reviewers could arrive at two different
conclusions.
Dr. Kerr stated that there is no evidence of any systematic
consideration of whether use of EPRI method would have any
effect on risk. He commented that it is possible if the
method decreases the number of unneeded scrams, and at the
same time does not increase the likelihood of exceeding a
safety limit, risk may be decreased by its use. He
recommended that prior to the NRC approval of this method for
;
Inst & Cntl 6Mtg. (8/29/91)
widespread application, a systematic investigation be made of
its effect on risk. Dr. Lewis and Mr. Wylie agreed.
3. Mr. L. Hopkins, Limerick Units 1 & 2, presented an overview
of the TRIP program. He stated that following Three Mile
Island (TMI) accident, the NRC required licensees of operating
reactors to reanalyze transients and accidents and to upgrade
emergency operating procedures (EOPs). The NRC staff
presented guidelines to the licensees to upgrade EOPs and
described the use of a "Procedure Generation Package" (PGP)
to prepare EOPs. The PGP would include plant-specific
technical guidelines, a writers guide, a description of the
program to be used for the validation of EOPs, and a
description of the training program for the upgraded EOPs.
From this PGP, plant specific EOPs were to have been developed
that would provide the operator with the directions to
mitigate the consequences of a broad range of accidents and
multiple equipment failures. Some licensees such as Limerick
Generating Station (LGS) Units 1 and 2 and Peach Bottom have
developed and implemented their own EOPs, using guidance
formulated by the BWR Owners Group (BWROG), such as the
Transient Response Implementing Plan (TRIP). TRIP equals
EOPs. LGS is a BWR with a Mark III containment structure.
Mr. J. Armstrong, Philadelphia Electric Company, indicated
that the NRC is currently performing inspections on different
plants to evaluate the EOPs at the licensee facilities. The
NRC staff considers the objectives to be met if review of the
following areas were found to be adequate: comparison of the
TRIP procedures with the LGS plant specific technical
guidelines (PSTG) and the BWR owners group emergency procedure
guidelines (EPG); review of the technical adequacy of the
deviations from the EPG; control room and plant walkdowns of
Inst & Cntl 7Mtg. (8/29/91)
the TRIP procedures; real time evaluation of the TRIP
procedures on the plant simulator; evaluation of the licensee
program on continuing improvements of the TRIP procedures; and
performance of human factors analysis of the TRIP procedures.
Dr. Kerr discussed with PE representatives some of the
comments made in the staff's report transmitted by a letter
of July 3, 1990 to Mr. D. M. Smith (PE Co.). PE
representatives consider the comments made by the NRC staff
during the inspection as constructive and helpful.
Dr. Kerr commented that examination of some of the comments
and PE's responses might lead to the conclusions that PE
personnel consider staff suggestions tantamount to an order.
Dr. Kerr cited an example in which the staff commented that,
"the glare from the Plexiglass covering the flowcharts in the
control room is excessive." The PE personnel indicated that
they would have not been aware of this problem until the staff
called it to their attention and they corrected it
immediately.
4. Mr. E. Imbro, NRR, briefed the subcommittee Members regarding
the Electrical Distribution System Functional Inspections
(EDSFIs). He stated that during multidisciplinary
inspections, the NRC has identified many deficiencies related
to the electrical distribution system. To address these
deficiencies, the NRC has developed an inspection to
specifically evaluate the electrical distribution system. The
NRC completed 24 EDSFIs at the five NRC geographical regions.
During these inspections, the staff found several common
deficiencies in the licensees' programs and in the electrical
distribution systems as designed and configured at each plant.
These deficiencies included inadequate ac voltages at the 480
Vac and 120 Vac distribution levels, inadequate procedures to
Inst & Cntl 8Htg. (8/29/91)
test circuit breakers, and inadequate determinations and
evaluations of setpoints.
Mr. J. Jacobson, Mr. A. Gautam, and Mr. D. Norkin, NRR, stated
that for several nuclear power plants, the staff found that,
under certain conditions, the voltage available at the safety
buses would be inadequate to operate safety-related loads and
associated equipment. These conditions could occur when the
plant's electrical distribution systems were being supplied
from an offsite grid that had become degraded but that
continued to supply voltages that remained above the setpoints
at which the degraded grid relays would be activated.
The function of the degraded grid relays is to ensure that
adequate voltage is available to operate all Class 1E loads
at all distribution levels. In order to ensure that all
required Class 1E loads will remain operable during degraded
offsite voltage conditions, some licensees are currently
reanalyzing the basis for the degraded grid relay setpoints.
The staff identified repetitive deficiencies in licensees'
programs to test circuit breakers. These deficiencies
included inadequate procedures, inadequate test acceptance
criteria, inadequate test equipment, and inadequate control
of testing. At the Susquehanna plant, the staff found that
the licensee was testing dc molded case circuit breakers with
a procedure written for testing ac breakers. The licensee had
not established specific acceptance criteria for the dc
breakers.
Other findings identified during recent EDSFIs were related
to inadequate setpoint determinations. Some licensees have
operated equipment outside of acceptable limits because they
did not determine proper setpoints and did not evaluate and
Inst & Cntl 9Mtg. (8/29/91)
account for instrument drift. The staff has identified these
circumstances primarily for instruments for which the
setpoints were determined by the architect/engineer or the
nuclear safety system supplier. Those setpoints not contained
in the plant technical specifications were also more
frequently found to be deficient. During recent EDSFIs, the
staff identified deficiencies in setpoints for diesel day tank
level indicators, diesel air start compressor controllers and
alarms, invertor low voltage shutdown circuitry, degraded grid
relays, and diesel overcurrent relays.
The staff has published Information Notice No. 91-29 on April
15, 1991. The staff believes that its inspections have
resulted in an increased sensitivity to the issue throughout
the industry. It is planned to continue performing EDSFIs
until all plants have been inspected by January 1993.
In response to Dr. Kerr's question regarding the risk
reduction that is likely to be achieved by this program, the
NRC staff had no answers and indicated that no effort has been
made to calculate or estimate the possible risk reduction.
Dr. Kerr stated that he would have liked to see evidence from
some source (may be NUREG-1150), that electrical systems in
operating plants represent a significant risk contributor, and
the result of the EDSFI program will produce some risk
reduction. The subcommittee Members commented that it should
be a Commission Policy to make decisions on starting new
programs based (at least partially) on the need and
expectation of risk reduction.
Dr. Kerr asked whether, now that a significant number of
plants have been examined and the results of those
examinations are available, it would be more effective to make
these results available and to ask the remaining plants to
conduct their own examinations as part of their IPEs. The
Inst & Cntl 10Mtg. (8/29/91)
staff appeared to believe that NRC management would not
approve of this approach, and the resulting examinations would
not be as thorough nor as effective as they would be if
performed by the staff.
In conclusion, the subcommittee Members agreed that the plant
electrical system is important, and that its reliability and
performance are important to both the prevention and the
mitigation of accidents.
FUTURE ACTION
The Subcommittee Chairman is planning to brief the full Committee
at the September 5-7, 1991 ACRS meeting regarding the above three
topics. Pending the outcome of such briefing, the Committee may
decide on a course of action..
Attachments:1. EPRI Setpoints Projects/Methodology
and Applications - by G. Srikantiah
2. Philadelphia Electric Co. (slides)
3. EDSFI briefing to ACRS
KOTEs Additional meeting details can be obtained from atranscript of this meeting available in the INRC PublicDocument Room, 2120 L Street, NW, Washington, DC 20006,(202) 634-3273, or can be purchased from Ann Riley andAssociates, Ltd., 1612 K Street, NW, Suite 300,Washington, DC 20006, (202) 293-3950.
EPRI/NPD
EPRI Setpoints Projects
Methodology and Applications
Presented by
.G mSrikantiah
August 1991
. -S & R/ Reactor Performance ProgramACRS 8/27/91 1
EPRI/NPD
EPRI Setpoint ProjectsMethodology and Applications
IIIi
OVERVIEW- Background
-- ;.Setpoin .t-Optimization ObjectivesExisting Methodology
- Safety Analysis Process
- Current Treatment of MeasurementUncertainties
- Development of Safety AnalysisInputs
- Potential Problems
EPRI SETPOINT METHODOLOGY
EPRI SETPOINT PROJECTSk I
S & R/ Reactor Performance ProgramACRS 8/27/91 2
EPRI/NPD
EPFRI Setpoint Projects
Background
* Setpoints are Instrument Settings atwhich System Automatic orOperator Action Must Occur toPreserve the Assumption in thePlant Safety Analysis
* Setpoints Must Satisfy Plant Safetyand Licensing Requirements
* Plant Capacity and OperatingFlexibility Should be Optimized
S & R/ Reactor Performance ProgramACRS 8/27!91 3
EPRI/NPD
EPRI Setpoint Projects
Setpoint Optimization Objectives
To Improve Plant Performance throughOptimized Instrument Settings whileSatisfying Safety and LicensingRequirements
*'To Gain Operational Margin
* To Improve Operational Flexibility
S & R/ Reactor Performance ProgramACRS 8127191 4
EPRI/NPD
EPRI Setpoint Projects
Existing Methodology
* Existing Methods are Based onDeterministic or License Basis AnalysisAssumptions, Including the MostAdverse Plant Conditions.
* Highly Conservative Methodology-RestrictsOperational Flexibility andMay Result in Derates
* Major Improvements Achievable byQuantifying and CombiningUncertainties in Key Variables
S & R/ Reactor Performance ProgramACRS 8127191 5
EPRI/NPD
CURRENT SAFETY ANALYSIS PROCESS
(Deterministic Methodology)
EVENT LIMITS
ALLOWABLE RANGE
ANALYSIS RESULTS
r
INSTRUMENT TRIPS(ANALYTICAL LIMIT)
1 PLANT TRAUSIEWRESPONSE
MODEL INPUTS/OPERATING LIMITS
S & R/ Reactor Performance ProgramACRS 8/27/91 6
,-° EPRI/NPD
CUFMEASU
THRE. TYPICAL APPROACH.S
IRENT TREATMENT OFREMENT UNCERTAINTIES
1. HISTORICAL
PROCESSVAR;AB':
SAFETY ANALYSISRESULTS
NOMINALSET PO!NT
ANALYTICAL EVENTLIMIT LIMIT
2. ENGINEERING JUDGMENT
PROCESSVARIABLE
ALLOWANCE PORINSTRUMENT DRIFI 1
ALLOWANCE FOR INSTRUMENSYSTEM UNCERTAINTIES ANDCALIBRATION UNCERTAINTY
SAFETY ANALYSISRESULTS I
I I I I
NOMINALSETPOINT
ALLOWABLESETPOINT
ANALYTICALLIMIT
EVENTLIMIT
3. SIMPLIFIED STATISTICAL
PROCESSVARIABLE
MAXIMUM INSTRUMENIDRIFT (2d SAFETY ANALYSIS RESULTS
NOMINALSETPOINT
I I I
ALLOWABLE ANALYTICAL EVENTSETPOINT LIMIT LIMIT
S & R/ Reactor Performance ProgramACRS 8/27/91 7
EPRI/NPD
POTENTIAL PROBLEMS
EVENTLiMITS
ALLOWABLE RANGEANALYSISRESULTS
PLANT TRANSIENTRESPONSE
INSTRUMENT'TRIPS
MlODEL INPUTS& OPERATINGLIMITS
Deterministic AcceptableRESULTS ACCEPTABLE
Statistical AnalysisMay Be Solution
ANALYSISRESULTS
MVNTLUITS
PLANT TRANSIENTRESPONSE
INSTRUMENTTRIPS
MODEL INPUTS& OPERATINGLIMITS
RESULTS UNACCEPTABLE
ANALYSISRESULTS
EVENTLIMITS(
ANALYSISRESULTS
MODEL INPUTS& OPERATINGLIMITS
MODEL INPUTS& OPERATINGLIMITSINSTRUMENTTRIPS
ANALYSIS RESULTSUNACCEPTABLE
= S & R/
INCREASED OPERATING SETPOINT WVERLAMARGIN
Reactor Performance ProgramACRS 8/27/91 8
EPRI/NPD
EPRI Setpoint ProjectsMethodology and Applications
OVERVIEW
EPRI SETPOINT METHODOLOGY
* Safety Analysis and Setpoint Process
* Safety Analysis and MeasurementUncertainties
a Setpoint Modification Processa Methodology Options
-Project Plan Development
C Statistical Analysis Methodology
* Statistical Combination of Uncertainties* Network Considerations* Combined MethodologiesEPRI SETPOINT PROJECTS
S & R/ Reactor Performance ProgramACRS 8/27/91 9
EPRI/PPD
GENERALIZED SAFETY ANALYSISAND SETPOINT PROCESS
< _ S & R/ Reactor Performance ProgramACRS 8/27/91 10
EPRI/NPD
SAFE_SFTPOII
TY ANALYSIS RELATED4TS AND UNCERTAINTIES
Actual BarrierIntegrity
Defined BarrierIntegrity
Event AcceptanceLimits
Event AnalysisResults
AnalyticalInstrumentSetpoints
Not 4iADDITIONAL MARGIN Ouzntffitble
Marin 4
'4 ~~~~~~~~~~MARGINCORRELATION OFUNCERTAINTIES Quantifiable SAFETY
CORRELATION INPUT MarginUNCERTAINTIES 1-
ALLOWABLE RANGE
MODEL UNCERTAINTIES 4Event Analyses/ EVENTPlant Dynamic Responses ANALYSIS
MODEL INPUTUNCERTAINTIES
'4^CALIBRATIONUNCERTAINTYMEASUREMENTSYSTEMUNCERTAINTIES
INSTRUMENT DRIFT
Measurement
UncertaintiesINSTRUMENT
SETPOINTPROCESS
AllowableSetpoints
NominalSetpoints
OPERATING MARGINOperatingEnvelope
Limits
*PLANNED MANEUVERING 4 4ALLOWANCE Operating OPERATING- - - - - - - - - - - - - --- - Envelope ENVELOPEao a a * ,. _- _ _ lesZICAVI NIAIL Allowanlces FHVkwLo
Steady State OPERATING ALLOWANCEOperation '
~ S & R/ Reactor Performance ProgramACRS 8/27/91 11
EPRI/NPDSETPOINT MODIFICATION PROCESS
-
CURRENT PLANTSAFETY ANALYSIS
PLANT DESIGNINFORMATION
RASP GENERICSETPOINT
METHODOLOGY
_ - _
I
. i 4 IPLANT SPECIFIC
SETPOIN1METHODOLOGY
I- I
4 _DETERMINE EVENT
ACCEPTANCELIMITS
IDENTIFY EVENTSTOBE
ANALYZED
PERFORM BASELINEEVENT ANALYSIS
. _
I
i 4 IF��
QUANTIFYUNCERTAINTIES
-__-4DETERMINE ANDVERIFY THE NEW
SETPOINTS_
u ,._
PREPAREDOCUMENTATION J
S & R/ Reactor Performance ProgramACRS 8/27/91 12
EPRI/NPD
Description of Methodology Options
Deterministic* Conservative selection of analysis inputs* Analytical instrument setpoints at 95% probability* No additional model uncertainty required* Results very conservative
U Used in current safety analyses
Complete SCU* All analysis inputs statistically treated* Results determined at the 95% probability/95%
confidence level* Model uncertainty considered in the analysis* Minimum amount of conservatism in results* Only used to:determine ,DNBR/M1CPR safety limits
Mixed SCU* Limited number of analysis inputs statistically treated* Remainder of inputs same as deterministic process* Model uncertainty considered in the analysis* Results determined at the 95% probability level for the
statistically treated parameters* Intermediate amount of conservatism results
Increasing usage in the safety analysis process
S & R/ Reactor Performance ProgramACRS 8/27/91 13
EPRI/NPD
SEETPOINT MODIFICATIONJECT PLAN DEVELOPMENTPRO.
I
I
a
I
I
I
i
I
I
0
0
. . . . . .. . .. .. .. .. . . . . . . . .. ..... ................. . . . . .. . . . . . .. . .
Plannina
PROJECTPLAN
UNDERSTAND CAPABILITIESAND TRADEOFFS
AVAILABLEMETHODOLOGY
RESOLUTION OFPROBLEM
.x TREATMENT OFUNCERTAINTIES
ASSESSMENTOF IMPACT
SELECT OPTIMUM APPROACHCOSTiMARGIN TRADEOFF
............................ …I. . . . . . . . . . . . . . . . . . . . . . .
0
4
II
DETERMINISTICANALYSISPROCESS II
STATISTICALCOMBINATION OFUNCERTAINTIES
Available Methodologv
04
0
0
0
0
0
0
I
I. . . . . .--------------------- -- -------
ENHANCED UNDERSTANDINGAS PROJECT PROGRESSES
F Recycle asNecessary 1
S & R/ Reactor Performance ProgramACRS 8127/91 14
EPRI/NPD
{ cQnf -4�stical CombinationWJ LGLI L 1
of Uncertainties
* Methodology
* Data Manipulation
* Relationship to Screening Process
* Analysis Input Screening
* Response Surface Development
* STARS Methodology
* Benefits
S & R/ Reactor Performance ProgramACRS 8/27/91 15
<-Ak EPRI/NPD
-44"STATISTICAL COMBINATION OF UNCERTAINTYMETHODOLOGY
IDENTIFY EVENT LIMITS
DETERMINE TREATMENT OF INPUT PARAMETERS -SCREENING PROCESS
1. Identify Analysis Input Parameters to Be Treated Determinlistically
2. Identify Analysis Input Parameters to Be Treated Statistically
3. Characterize Statistical Inputs
DEVELOP UNCERTAINTY DISTRIBUTION FOR THE EVENTBEING ANALYZED - SCU PROCESS
1. Develop ResponseSurface
2. Simulate Event to Obtain Combined Effect of UncertaintiesIncluding the Uncertainties in the Response Surface Fit
DETERMINE EVENT PROBABILITY AS A FUNCTION OF THEEVENT LIMITS
COMPARISON TO THE EVENT ACCEPTANCE LIMITS
S & R/ Reactor Performance Program °ACRS 8/271/91 16
EPRI/NPD
SETFDATA MANIPULATION FOR
POINT MODIFICATION ANALYSES
Deterministic Statistical Combinationof Uncertainties
DATA FOR SPECIFICCOMPONENTS TO
BE CHANGED
DATA FORPOTENTIALLY. SENSITIVEPARAMETERS
DETERMINEUNCERTAINTIES
CONSERVATIVELYINCLUDE IN
ANALYSIS PROCESS
PERFORMANALYSES
. .4DETERMINE
UNCERTAINTIES
SCREEN
DEVELOPRESPONSESURFACE
I _.
Process Is Similar, but Number ofParameters and Their Treatment
Is More Extensive for SCU Process
S & R/ Reactor Performance ProgramACRS 8/27/91 17
EPRI/NPD
REL.ATIONSHIP OF SCREENING PROCESSTO SCU METHODOLOGY
STATISTICALDATA
I
DETERMINEtM EAN AND
-STANDARDDEVIATION
1 _~~~~~~~~~~~~~Fy
I SCREENING.-PROCESS
_ _ _
RESPONSESURFACE
DEVELOPMENT
STATISTICALDISTRIBUTIONNORMALUNIFORMOTHER
RESPONSESURFACE
FITTINGCOEFFICIENTS
MONTECARLO
ANALYSIS
I EVENTACCEPTANCE
LIMITS
S & R/ Reactor Performance ProgramACRS 6127/91 18
EPRI/NPD
ANALYSIS INPUT SCREENING PROCESS
r DesiredSetpointChanges
I1 Plant Model withAll Parameters at
License BasisValues - - �EVENTS
.-REQUIRINGANALYSIS --- A-4
VERIFY SETPOINT-RELATIONSHIPS'RETAItNED FOR
DESIRED CHANGES-
ELIMINATENOT APPLICABLE
PARAMETERS FOREVENTS TO BE
ANALYZED
4SCREEN FOR POTENTIALLYSIGNIFICANT PARAMETERS
SMALL UNCERTAINTIES ENGINEERING JUDGMENT- Physical Properties 1 Difficult- Geometric Inputs * Expensive
V
IDENTIFY NOMINALVALUES AND
VUNCERTAINTIESOF PARAMETERS
POTENTIALLYSENSITIVE
:''PARAMETERS.DENTIFIED
[ ~~SENSITIVITY STUDIES
Run Baseline Run Parameter Run ExtremeCases for Variation Cases Sensitive Cases
Each Event for Each Event as Necessary
DEVELOP CRITERIAPARAMETERS TO FOR SELECTION
BE USED IN F OR SELECTIONOF SENSITIVE ITHE SCU PROCESS PARAMETERS
S & R/ Reactor Performance ProgramACRS 8/27/91 19
EPRI/NPD
STATISTICAL L COMBINATION OF UNCERTAINTIES PROCESSESPONSE SURFACE DEVELOPMENT" I
or
STATISTICALLYTREATED INPUT
PARAMETERS
Ir
xi
i(x2)
X2
IAt(xi)
xi
Analysis Input Patameters Probability Density Functions_ -
I II.5
DETERMINISTICINPUT PARAMETERS H EVENT
ANALYSISMETHODOLOGY H RESPONSE
SURFACE DESIGN
(Type of Fit andCase Definition)
. .
:m;i-:4I-A L-
NN N
y~x)= BO+XB9x, +FxRixfoull i-I
rn
i(r)N N
+ X' TBqxjxi.1=1 i-"ij t..
- ~NC2at = NC K1 Y (Yk - YRSk 2
Lt- NC -K - 1 k-1 S
For OCCO
NC*2N+2N+1
K wN +) IN *Z2
;2 MONTE CARLOSIMULATION
_ .
Cases for a 3 Parameter OCCO (N-3)and Normally Distributed inputs
Par-± rl Par E2aLrAtj.-20 IA-20; p-2ag.+2a IA-2a IL-2cis-2c0 psa is-20JA+2a p.-2o i- 2oIL-2a p-2la is+2p+2a gL-2c IL+2p-2c i.+2a p+20p.+2a ipti2c p+20
IL-2Aa is )Aist2Aa A HL
,A js-2Aa p
H 9A p-AH H $L-2Aaas a
EVENTLIMITI EVENT 1~~~~~~~.0
i
, ._ A
S & R/
Figure of Merit
Reactor Performance Program 0ACRS 8/27/91 20
EPRI/NPD
C"TARS METHODOLOGY
-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ICODEINPUTS K
RESULTS OF STATISTICALSYSTEM PROPERTIES
TRANSIENT OF INPUTI CALCULATIONS PARAMETERSI III- - - - r - - - - - - - - -
III
RESPCS.URF
.CONSTR
(Least SEvalui
)NSEACEUCTION
iquaresation)
PolynomialCurve Fit
iTICAL,YSIS '
lo Sampleeier.Set)
Probability Distributionfor System Analysis I
STARS I
CODE…I-- - - - - - - - - - - - - -.
II
STATISANAL
(Monte Caiof Param;
V_
PROBABILITYOF EXCEEDING
EVENT LIMIT
S & R/ Reactor Performance ProgramACRS 8/27/91 21
EPRI/NPD
BEN
0.06
0.05 _
IEFITS OF STATISTICAL COMBINATIONOF UNCERTAINTIES METHODOLOGY
-OM0
0.04
0.03
0.02
0.01
0.00
FIGURE OF MERITProbability Density Function
75
0~
w
-J
7-)
1.0
0.9 -
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
I~~~~~~~~~~
-~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .4
/ ~~~~~~~~~~~~I/C Cd
/~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~I
IncreasedMargin
I I
I DetermninisticIAnalysis Results
Acceptabl(at 95% Pr(
Event? Value I Limit,bability I
I I . I
FIGURE OF MERIT
Cumulative Distribution Function
S & R/ Reactor Performance ProgramACRS 8/27/91 22
CDD5I /K1 fJr
Network Analysis
* Network Considerations
Typical Component Networks
v"Development of-Analysis -Inputs
* PLANETS Methodology
* Typical Individual ComponentPerformance
* Examples of Network Performance
* Event Analysis Inputs
S & R/ Reactor Performance ProgramACRS 8/27/91 23
EPRI/NPD
Network Considerations* Component networks are frequently
encountered in nuclear power plantdesign
- Single failure considerations- Avoidance of spurious trips or
system startups- Improved system reliability
* Safety Analyses Typically TreatComponent Networks as IndividualComponents
- Deterministic methodologyphilosophy
- Computer code model limitations* Component Networks
- Superior to individual components- Methodology available to quantify
benefitsI I
S & R/ Reactor Performance Program °ACRS 8127191 24
EPRI/NPD
TYPICAL COMPONENT NETWORKS
CHANNE-ATRIP
SYSTEM-- -- ACTUATiON
1 of 1 Logic
CHANNELATRIP
I - SYSTEMCHANNELB ACTUATIONIRiP
2 of 2 Logic
CHANNE_.ATRiP
CHANNEL3,RIP
CHANNELAIRIp
CHANN=E3T RIP
CHANNELCTRIP
CHANNEL3TRIP
CHANNELATRIP
CHANNEL3TRIP
CHANNELCTRIP
CHANNELNTRIP
CHANNELATRIP
CHANNELETRIP
CHANNELCTRIP
CHANNELNTRIP
SYST--tlZ oa - ACTUA-rON
1 of 2 Logic
SYSTEt.1ACTUATiON
CHANNELATRIP
CHANNELB SYSTEMTRIP ACTUATION
CHANNELCTRIP
2 of 3 Logic
CHANNELATRIP
CHANNELBTRIP
CHANNELCTRIP
CHANNELE0TRIP
El- SYSTEMD-~ACTUATION
1 of 2 X 2 Logic
2 of 4 Logic
1 of N Logic
CHANNELATRIP
CHANNELBTRIP _ Y
XINCHANNEL CTRIP
CHANNELN C
CHANNELA NTRIP A
CHANNELATRIP
CHANNELCTRIP V
CHANNELN
Xof N byYofMLogic
'YSTEMACTUATION
V nf NI I 0%lF^ -G .. ,-Vsp
S & R/ ReactorACAS 8/27/91 25
J
Performance Program $D
EPRI/NPD
Development Of Analysis InputsFor Component Networks
* Analyses Typically Simulate OneComponent - Even for ComponentNetwo rks
-* Component Performance can beStatistically Based (95% Probability)
* Network Performance is Superior toIndividual Components
* Equivalent Network Performance canbe Defined
* Network Inputs can be DevelopedConsistent with Deterministic orStatistical Combination ofUncertainties Methodology
S & R/ Reactor Performance ProgramACRS 8127/91 26
EPRI/NI mmr -
ru UPLANETS METHODOLOGY
r --------------------------- _ _ _ ICODE INPUTS
GROUP DISTRIBUTIONDEFINITIONS OF INITIATING
;Norninal Sepooint PARAMETERSI ana Numoer o0 (By Component
Componens Group
----- r - -- -- -- -- -- -- -- -- - -- r -
, Il
NETWORKl ~ANALYSIS
(Evaluate Distributionfor Component Logic)
ProbabilityDistribution forLogic Networkas a Function
PLANETS of InitiatingCODE Parameter
I~~~ I
PROBABILITYTHAT THE
COMPONENTNETWORK WILL
INITIATE
=S & R/ Reactor Performance ProgramACRS 8/27/91 27
EPRI/NPDTYPICj AL INDIVIDUAL COMPONENT PERFORMANCE
0.7
0.6
o-
coco0:
C.5
0.4
0.3
0.2
0.1
0.0.4 -3 -2 -1 0 t
STANDARD DEVIATION
Probabilitv Densitv Function
.93 +4
1.0
m-i
ICn
0a-w
-J
0
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0-4 -3 -2 -1 0 +1
STANDARD DEVIATION+2 +3 +4
Cumulative Distribution Function
S & R/ Reactor Performance ProgramACRS 8/27/91 28
EPRI/NPD
EX.AkMPLES OF NETWORK PERFORMANCE
m-
coJo
0
0.
a-wco
0-
C.
3ll
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
1.0
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
-4 -3 -2 -1 0 -,-1 +2STANDARD DEVIATION
Netwvork Probabilit% Densitv Function
+3 +4
-4
ACRS 8/27/91 29
-3 -2 -1 0 +1 +2 +3STANDARD DEVIATION
Network Cumulative Distribution Function
_ S & R/ Reactor Performance Program
+4
EPRI/NPD
DEVELOPMENT OF EVENT ANALYSIS INPUTSFOR SIMULATING COMPONENT NETWORKS
(Deterministic Methodology)
1.0 -
0.9
1-J
O0c:
P:-j
C2
3
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0-4 -3 -2 -1 0 +1 +2 +3 +4
STANDARD DEVIATIONNetwork Cumulative Distribution Function
S & R/ Reactor Performance ProgramACRS 8/27/91 30
EPRI/NPD
DEVELOPMENT OF EVENT ANALYSIS INPUTSFOR SIMULATING COMPONENT NETWORKS
(Statistical Combination of Uncertainties Methodology)
1.0
0.9
-
-J
F0
co
I--
M-
0.8
0.7
0.6
0.5
0.4
0.3
0.2
1 of 1I rinirt
2 of 4Logic
Equivalent Points/3 Parameter OCCD)
1 of I 2of 4
J.±9.2.43a ±+1.14ap- 2.00a .±+0.89a
}~~ ~ ~~~~~~ p -0.29jpi-2.00a p.-1.52ap.-2.43a 9-1.80a0.1
0.0-4 -3 -2 -1 0 +1 +2 +3 +4
I ~~~~~~STANDARD DEVIATIONNetwork Cumulative Distribution Function
S & R/ Reactor Performance ProgramACRS 8127191 31
EPRI/NPD
Combined Methodologies
* Overview
, Network Performance Overlap
* Overlap Example
* Code Relationships
* Evaluation of Statistical Limits
S & R/ Reactor Performance ProgramACRS 8/27/91 32
EPRI/NPD
COMBINED METHODOLOGIES OVERVIEW
COMPONENT COMPONENT PROBABIITY ROBABILTTYNETWORK 1 NETWORK 2 DISTRUBUTION DISTRIBUTION
PROBABILITY ROBABILITY OR F ShyDISTRIBUTION DISTUTION ANALYSIS UMIT
I r I~~~~~~ji~~~~~~~~~-
OVERLAP ANALYSISF___________. _________________________. __ ___,I I I ~~~~~~~~~~~~~MOONSCODE'
It $1~~~~~~~~~~~~~I
PROBABILITY OF PROBABILITY OF
NETWORK ORPERFORMANCE ORE LEINEVNT
OVERLAP UI
S & R/ Reactor Performance ProgramACRS 8/27/91 33
EPRI/NPD
Network Performance Overlap
System actuation overlaps can occur
- Nominal setpoints in close proximity
- Large measurement uncertainties
- Associated with deterministic treatmentof analytical limit
Network performance -may be acceptable
- Low probability of overlap acceptablefor many applications
- Statistical assessment of overlapfeasible
S & R/ Reactor Performance ProgramACRS 8/27191 34
EPRI/NPDNEETWORK PERFORMANCE OVERLAP
(\j
E
C
ej C U.0 .1)Z L%) ?ACL CC :.,t MQ zM 1; qr.z a
I
E
C
Nominal Selpoint 1- ..
-4 -3 -2 -1 0 -+1
STANDARD DEVIATIONDeterministic Setpoints
Nominal Setpoint 2
- +2 +3 +
0.8
0.7
0.6
-j
4:
0
0.5
0.4
0.3
0.2 -
0 .0 LC4 -3
ACRS 8/27191 35
-2 -1 0 +1 +2 +3STANDARD DEVIATION
Network Probability Density FunctionS & R/ Reactor Performance Program
EPRI/NPD
COMBINED METHODOLOGIES RELATIONSHIPS
r---------------------------------------…-________iI CODE INPUTSI
GROUP DISTRIBUTI1DEFINITIONS OF INITIATIN
(Nominal Setpoint PRESSURESana N.mber of (By ComooneCo.moonents) Group)
-- ---_--------- --________ _
l~~~G RESULTS OF STATI!C SYSTEM PROPMnt TRANSIENT OF II
CALCULATIONS PARAM
_~~~~~~~~~
STICAL IERTIES IINPUT IIETERS
- -I- ---
I-
II -NET1I ANA
(Evaluatefor Compc
: PLANETS£ CODE
.-- - -
MORKLYSIS
Distributiononent Logic)
IIIIIIIIIIIIIIIII
.ItIIII
.i
I-I
IIIII
RESPCSURF
COMM~f
(Least S. Evalw
STATISANAL
(Monte Carof Param
ProbabilityDistributionfor EventLimit
_-______________---r
)NSE tACEUCTION _
;quaresation)
PolynomialCurve Fit
;TICAL.YSIS
rlo Sampleeter Set)
Probability Distribution Ifor System Analysis
STARSCODE
__ __ _ __ _I
[LAP.YSIS
mit andknalysisIntegral) MOONS
CODE '
WO,
-Y
I- - - -
II
OF -"N II
PROBABILITYTHAT AN EVENT -4EXCEEDS LIMITS I
I
4.....~~~~~I
OVEFANAL
(Event LiSystem dOverlap I
I I
S & R/ Reactor Performance ProgramACRS 8/27/91 36
EPRI/NPD
[
0.06
0.05
0.04 -
< 0.030X 0.02
0.01
0.00
WALUATION OF STATISTICAL LIMITS
FIGURE OF MERIT
Probability Density Function
S & R/ Reactor Performance ProgramACRS 8/27/91 37
EPRI/NPD
(1 EPRI Setpoint ProjectsMethodology and Applications
OVERVIEW
EPRI Setpoint Methodology
.E4EPRI Setpoint Projects
Oyster Creek -Hich Pressure Setpoints* Application Overview* High Pressure Setpoint Relationships* Current and Proposed Setpoints* Analysis Considerations'Licensing Basis Events Selection Process* Plant Performance Events Selection
Process* Event Assessment* Analysis Results* Approach to Safety Valve Opening
Probability* Conclusion
S & R/ Reactor Performance ProgramACRS 8/27/91 38
EPRI/NPD
EPRI Setpoint ProjectsMethodology and Applications
Vermont Yankee Reload Analysis* Application Overview
Approach and Results
Diablo CanYon-Safety Valve Tolerance* Application Overview* Approach and Results
TMI- 1 Flux/Flow Setnoints- Application Overview* -Current Flux/Flow Setpoints* Proposed Flux/Flow Setpoints* Code Sequence* Project Status
Palo Verde Safetv Valve Tolerance* Application Overview* Project Approach
S & R/ Reactor Performance ProgramACRS 8/27/91 39
EPRI/NPD
Oyster Creek High PressureSetpoints Overview
* High pressure setpoints did notprovide sufficient margin forinstrument drift and calibrationtolerances
* Safety valve setpoint drift resulted infrequent and costly testing of valves
S & R/ Reactor Performance ProgramACRS 8/27/91 40
EPRI/ I IMI r4 1 L
OYSTER CREEK HIGH PRESSURE SETPOINT/TECHNICAL SPECIFICATION RELATIONSHIPS
1 090 _ EMRV (,CE) Tech. Spec.r-----i 1 _
<- - 080 - - - - - - - - - --- EMRV (B,C,E) Setpoint2 5 1080
-EMRV (AD) Tech. Spec.
_7 a' t l.
iiii
POENTIALRL
REGION
1020
ISU = Instrument System UncertaintiesD = Instrument Setpoint Drift
S & R/ Reactor Performance ProgramACRS 8/27191 41
EPRI/NPD
OYSTER CREEK HIGH PRESSURE SETPOINTS
1250 -
1240 -
1230 -
1220 - BANK
1210 -B BANK
1200 - i S.VS.V. -ANK
BANK Prga #4
1190 # 71 BANK
1180 - BANK#2
s.v.-~1170 BANK
#1
1100
1090 rI ~~~~~~EMRV
1080 GRP. #2
1070 -VR
1060 - GRP. #1
I I I ~~~~~~~~~C/RPT1050 ~~IC/RPT EMRV
GRP. #1 CA
1040 SCRAM
Current Setpoints Proposed Setpoints
S & RI Reactor Performance Program 00ACRS 8/27/91 42
EPRI/NPD
Analysis Considerations
License Basis Requirements
- Evaluation of recirculation pump trip andalternate rod insertion for high pressuresetpoi-nts for anticipated transientswithout scram
- Opening no more than one relief valvewithout scram
- Analysis to establish minimum criticalpower ratio operating limit
- Demonstration peak system pressureacceptable for proposed setpoints
S & R/ Reactor Performance ProgramACRS 8127/91 43
in n r IKI r% I*Op rI11111ru X
Analysis Considerations
Plant Performance Considerations
- Low probability of relief valve openingprior to isolation condenser initiation
- Low probability of alternate rodinsertion prior to isolation condenserinitiation
- Low probability or recirculation pumptrip prior -to scram
- Low probability of relief valve openingprior to scram
- Low probability of a safety valveopening as a result of an anticipatedoperational occurrence
S & R/ Reactor Performance Program °ACRS 8127/91 44
EPRI/NPD
LICENSE IBASIS EVENT SELECTION PROCESS
_
[ GENERIC LICENSEBASIS EVENTS
TO BE CONSIDEREDI
ELIMINATE ALL0 EVENTS TMAT ARE
NOT APPLICABLETO OYSTER CREEK
U.~
IDENTIFY ALL EVENTSTHAT ARE NOT LIMITING
FOR OYSTER CREEK
.4nLICENSE BASISEVENTS TO BE
ANALYZED
IDENTIFY ALL EVENTSREQUIRING FURTHER
EVALUATION INCLUDINGTHE POTENTIAL IMPACTOF SETPOINT CHANGESK 2
S & R/ Reactor Performance Program °ACRS 8/27/91 45
i
EPRI/NPD
PLANT PERFORMANCE EVENT SELECTION PROCESS
-
GENERIC PLANTPERFORMANCE EVENTS
TO BE CONSIDEREDIN
4ELIMINATE ALL
EVENTS THAT ARENOT APPLICABLE
TO OYSTER CREEK
ELIMINATE ALL LOWPROBABILITY EVENTSFOR OYSTER CREEK
I
IDENTIFY ALL EVENTSTHAT ARE NOT LIMITING
FOR OYSTER CREEK
.~~~
U PLANT PERFORMANCEEVENTS TO BE
ANALYZED
-
.
IDENTIFY ALL EVENTSREQUIRING FURTHER
CtlA I I IATldQJ ItfflI I lhIJV:l C 1L0S I Iva I %OI & I%
THE POTENTIAL IMPACTOF SETPOINT CHANGES
S & R/ Reactor Performance ProgramACRS 8/27/91 46
EPRI/NPDAssessment of BWR Events thatReach High Pressure Setpoints
AssessmentTransients Plant Performance
Inacvertent High Pressure MakeupSystem Start Not Applicable
Pressure Regulator Fail!-re-Closed Low ProbabilityGenerator Load Rejection with Bypass Not LimitingGenerator Load Rejection witnout Bypass Low ProbabilityTurbine Trip with Bypass AnalyzeTurbine Trip without Bypass Not LimitingMain Steamline Isolation Valve Closure AnalyzeLoss of Condenser Vacuum Not LimitingRecirculation Pump Trip Not LimitingRecirc. Flow Controller
Failure-Decreasing Flow Not LimitingInadvertent Opening of a
Safety or Relief Valve Not LimitingPressure Regulator Failure-Open Not LimitingLoss of Feedwater Flow Not LimitingLoss of Offsite Power Not LimitingLoss of Shutdown Cooling Not LimitingFeedwater Controller Failure Low Probability
AssessmentLicse Basis
Not ApplicableNot LimitingNot LimitingNot LimitingNot LimitingAnalyzeNot LimitingNot LimitingNot Limiting
Not Limiting
Not LimitingNot LimitingNot LimitingNot LimitingNot LimitingAnalyze
Accidents
Control Rod Drop AccidentMain Steamrine BreakSmall Loss of Coolant AccidentRecirculation Pump Seizure
LowLowLowLow
ProbabilityProbabilityProbabilityProbability
Not LimitingNot LimitingNot LimitingNot Limiting
SpecIal Events
Shutdown from Outside the Control RoomOverpressure Protection AnalysisAnticipated Transients without Scram
Low ProbabilityLow ProbabilityLow Probability
Not LimitingAnalyzeNot Limiting
S & R/ Reactor Performance ProgramACRS 8/27/91 47
EPRI/NPD
Analysis Results
License Basis (License Basis Methodoloav)|- Recirculation Pump Trip and Alternate Rod Insertion
Performance Acceptable- No Overlap of Relief Valve and Scram Setpoints- Current Minimum Critical Power Ratio Operating Limit
Acceptable- Peak System Pressure Below Safety Limit
Plant Performance (PLANETS Analysis)- Probability of Relief Valve Initiation Prior to Isolation
Condenser < 7X10-4 per Event- Probability of ARI Initiation .Prior to Scram < 10-6 per
Event- Probability of Recirculation Pump Trip Prior to Scram <
5X10-2 per Event- Probability of Isolation Condenser Initiation Prior to
Scram < 5X10-2 per Event- Demonstration of Acceptably Low Probability of Safety
Valve Opening for Anticipated OperationalOccurrences Most Difficult Task
S & R/ Reactor Performance ProgramACRS 8/27/91 48
EPRI/NPDANALYSIS PROCESS FOR SAFETY VALVE OPENING PROBABILITY
CODE INPUTS…__________I
-GERIOUP DISTRIBUTIONDEFINITIONSOF OPENING
(Nominal Se:ooin: PRESSURESa-c Numnoer of eSafe:y VaesBy Va ve Gr :
_. .. _
--- r-- -- -- -- -- -- -- -- - -- I -I I
I II
RESULTS OFSYSTEM
TRANSIENTCALCULATIONS
N f.
STATISTICAL IPROPERTIES
OF INPUT IPARAMETERS I
…~~~~~~~~~~~~~~~~~~~< _~~~~~~~~~~~~~~~~~
_____~~ -r - - -
I
RESPONSE* SURFACE
-CONSTRUCTION
. (Least SquaresEvaluation)SAFETY VALVE
ANALYSIS
(Evaluate PressureDistribution for First
Opening Valve)
PolynomialCurve Fit
IIIIIIIIIIIIIIIIIIIIIIIIII
ProbabilityDistributionfor Pressureof First OpeningSalety Valve
- - - - - - - - - --_ _
STATISTICALANALYSIS
(Monte Carlo Sampleof Parameter Set)
PLANETSCODE
-
Probability Distributionfor Peak SystemPressure
STARSCODE
-_- _____- __- __- __- __-
I;
II
IIII
I- - -
PROBABILITY ITHAT A TRANSIENT
__ I... .I
4 4 t 1OVER LAPANALYSIS_ . o_ . , .lI
I OPENS A ;tzatety valve ana II SAFETY VALVE I System Pressure I
I Overlap Integral) MOONS II CODE I
S & R/ Reactor Performance ProgramACRS 8/27/91 49
EPRI/NPD
Conclusions
* All License Basis Requirements Were Satisfied forthe Proposed Changes to the High PressureSetpoints for the Demonstration.
* No Other Changes to Any Other Current Plant.Operating Limits or Technical Specifications AreRequired.
* The Plant Performance Requirements Were Satisfiedfor the Proposed Changes to the High PressureSetpoints for the Demonstration.
* The EPRI RASP Setpoint Guidelines Can Be Used inEstablishing Complex and Interrelated Setpoints.
* Statistical Combination of Uncertainties MethodologyCan Be Used to Reduce the Excessive ConservatismInherent in the Deterministic Analysis Process.The Statistical Combination of UncertaintyMethodology is Relatively Simple to Apply.
* Analysis Capability for Complex Valve Networks Is anImportant Contribution Made by this Demonstration.
* The Setpoint Demonstration Program Identified aNumber of Important Lessons Learned.The Experience Gained Th'ough the Performance ofthis Demonstration Should Be Invaluable to OtherSetpoint Applications.
S & R/ Reactor Performance ProgramACRS 8/27!91 50
crlnfl% I t rll00- rml,^v~ru V
Vermont Yankee Reload AnalysisOverview
* Substantial Economic Benefits for HighEnergy Reload Core Design
* Planned High Energy ReloadConstrained by Minimum Critical PowerRation Operating Limit
S & R/ Reactor Performance ProgramACRS 8/27/91 51
EPRI/NPD
Vermont Yankee Reload AnalysisApproach and Results
* Reduced Critical Power Ratio MarginRequirements by.use of 1-D KineticsAnalysis and EPRI Guidelines
* Scram Speed Determined to be MostSensitive Analysis Parameter
.*. .Applied EPRI's Statistical Combinationof Uncertainties Methodology andCombined Scram Speed and ModelUncertainties to Justify SimplifiedAnalysis Approach
* NRC Approval - October 1989
S & R/ Reactor Performance ProgramACRS 8/27/91 52
0
__- [=n01/K1r3r,% -
Diablo Canyon Safety Valve ToleranceOverview
.ObjLectives
- Relaxation of Lift Tolerance on. Pressurizer and Steam Generator
Safety Valves
- Justify Increase of ToDIerance to ±3%
Benefits
- Fewer Safety Valve Surveillance TestFailures
- Reduction in Plant Down Time for ValveTesting and Maintenance
- IncreasEL in Operational Flexibility
S & R/ Reactor Performance ProgramACRS 8/27/91 53
EPRI/NPD
Diablo Canyon Safety Valve ToleranceApproach and Results
* Applied EPRI Reactor Analysis SupportPackage Guidelines to PerformAnalyses for Setpoint Modifi-cation
* Used PLANETS Codelto DetermineSafety Valve Opening Probability
* Analyses Demonstrated that VesselOverpressure Protection Requirements
-Satisfied for increased Safety ValveSetpoint Tolerance
* Statistical Analysis Demonstrated thatIncreased Tolerance Covered byPrevious Conservative Analyses
S & R/ Reactor Performance ProgramACRS 8/27,'91 54
EPRI/NPD
TM-1 Flux/Flow Setpoints
Overview
* Increase in Flux/Flow OperatingMargin Desired
* Reduction in Potential for SpuriousScrams
S & R/ Reactor Performance ProgramACRS 8/27/91 55
EPRI/NPD
CURRENT FLUX FLOW SETPOINTS FOR TMI-1
---------
1i20
-30. 5. 1 0 B (24.5.
At-50.0.70.0)
(-50.0.42.6)
(-50.0,15.1)
/ /
04 AAcceptable
i Operation 1IC 1
I II 1
1 ;-3 .80 60 (24.5.80.6)1
I Acceptable3&4 Pump
I Operation I
I I
-601 -30 C.53-l (24.5.953.1)1
108)
01II
II
IIIIII
Acceptable2 3. & 4 Pump
Operation C
40.'
L-
4I. I
) ~~~~~~~I
I I
I I I~
(45.0.70.0)
(45.0X42.6)
(45.0,15.1)
1. 6
so 60
I I-,
-60 -50 -40 -30 -20 -10 0 10 20
Axial Power Imbalance (%)30 40
S & R/ Reactor Performance ProgramACRS 8/27/91 56
EPRI/NPD
ll PROPOSED FLUXFLOW SETPOINTS FOR TMI-1
.~~~ -0 - - - - - - - - - - - - - - - - - - - - - - - - - - - -I AcceptaDle4 Pump
Operation
-50.0.85.01--…-…-…-…-…-I Accepiaole ~~~~~~Supponted by
3 I& 4 PUMp Analyses inOperation this Report
I 60
IIIIII11II
. IIIIIIIIII
I
(45.0.110)
(45.0.85.0)
(45.0,55.0)
II
(*50 .0.55.0)L …-----------------4
Acceptable2.3. & 4 Pump
OperationII 40-
I
II
1 20 -
I I I II
I l
IF
-
-
a
C0.-
J
a_
…~~~~~~~~~~~~~~~~~~I
D I
D I
D I
i I I. I I
-60 -50 -40 -30 -20 -10 0 10Axial Power Imbalance
20(%)
30 40 50 60
Requires Maneuvering-…-…-…-…-…- - fnrp rr.nmma
Justification
S & R/ Reactor Performance ProgramACRS 8/27/91 57
EPRI/NPD
TM I-1 CODE SEQUENCE FOR FLUXI FLOW ANALYSES
i ,.~~~~~~~..... 0.....I II I
M ANEAN'Ei'J'ER NG A,'ALY C-- S T. ,- v -Ad ------ - '-SIMULATE * 3P Ad - - -
-
a - aa a,
II0
I -. 1. - --- . I -
I II If I* - - - -e - -- ~~ 0
I~~~~~~~~CASMO
I II I
I I
-0A..
TWO-GROUP, I
-DNUCLEAFCONSTANTS
Analysis Process to Justify Setpointsfor 0 Axial Imbalance
Maneuvenng Analysis Required to- . . . Complete Justification of Flux/Flow
Setpoint Change
S & R/ Reactor Performance Program001
ACRS 8/27/91 58
EPRI/NPD
iI
TMI-1 Flux/Flow Setpoints
Project Status
Phase 1 _Comnete
* Safety Analysis for 0 Axial ImbalanceAcceptable
* Documentation Being Prepared
Phase 2- Beina Initiated
* Maneuvering Analysis Necessary
= S & R/ Reactor Performance ProgramACRS 8/27/91 59
EPRI/NPD
(f Dnlt%Verde Safety Valve Tolerance II "AW
Overview
* Test Data on Pressurizer and MainSte-am Safety Valves Indicate.Drift
* Technical Specification"Tolerance of±1 % too Restrictive
* Tolerance Relaxation to ±3% Desired
*'Requires "Demonstration of Adequacyof Overpressure Protection
* Requires Demonstration of LowProbability of Safety Valve OpeningPrior to Scram
S & R/ Reactor Performance ProgramACRS 8/27/91 60
EPRI/NPD
Palo Verde Safety Valve Tolerance
Project Approach
* Use EPRI Statistical Methodology to-Justify Larger Uncertainty in Valve"Opening Pressure
* Use PLANETS Code to Simulate theBehavior of the Valve Network
. Quantify the Overlap in the :ProbabilityDistributions of the Safety ValveOpening and Scram Occurrence
* Use License Basis AnalysisMethodology to Justify AcceptableOverpressure Protection
S & R/ Reactor Porformance ProgramACRS 8127/91 61
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ATTAhCMENT 3
EDSFI BRIEFING TO ACRSAUGUST 29. 1991
SCOPE OF EDSFI
- ADEQUACYSYSTEM
- EVALUATESUPPORT
OF THE ELECTRICAL DISTRIBUTION
ENGINEERINGCAPABILITY
AND TECHNICAL
* TEMPORARY INSTRUCTION IN USE BY REGIONS
* TO ACHIEVE CONSISTENCY OF TI IMPLEMENTATION
- FIVE PILOT INSPECTIONS CONDUCTED BY- TRAINING PROVIDED AT TTC- DRIS PARTICIPATION IN REGIONAL EDSls
DRIS
* STATUS OF TI IMPLEMENTATION
- 24 INSPECTIONS COMPLETED- TARGET COMPLETION OF TI -
AS O0 AUGUSTJANUARY 1993
30, 1991
PAGE 1
i
EDSFI BRIEFING TO ACRSAUGUST 29, 1991
* REGIONAL COUNTERPART MEETINGS HELD APRIL 24 ANDSEPTEMBER 5, 1991
* REGULATORY INFORMATION CONFERENCE EDSFI SESSIONMAY 7
* INFORMATION NOTICE 91-29 ISSUED
- ADDRESSES INITIAL FINDINGS FROM EDSFIs
* DEGRADED GRID VOLTAGE* INADEQUATE SETPOINTSE INADEQUATE TESTING OF
RELAY SETPOINTS
CIRCUIT BREAKERS
PAGE 2
...
r
ELECTRICAL DISTNRIBUTION FUNLTINAL INSPRECURRING PROBLE-MAllEAS
* DESIGN DEFICIENCIES
* PROCEDURAL WEAKNESSES
* CONFIGURATION MANAGEMENT
3
l
r
ELECTRICAL DISITRIBUTIN FUNCTIONAL INSPRE RRINa PRBLEM-AREA
* ELECTRICAL DESIGN IEFICIENCIES
- FAULT PROTECTION* UNDERSIZED AC 4* LACK OF SHORT
& DC CIRCUIT BREAKERSCIRCUIT ANALYSES FOR DC CIRCUITS
- VOLTAGE REGULATION* GRID UNDER VOLTAGE CONDITIONSa DIESEL GENERATOR TRANSIENT LOAbING* LOAD SEQUENCING
CONDITIONS
- BUS TRANSFER
- BREAKER COORDINAtION
- RELAY SETTINGS
- PHYSICAL SEPARATION
- RACEWAY FILL
4
i
.
ELECTRICAL DISTRIBUTION FUNCTIONAL INSPRECRRING PROBLEM ARE
MECHANICAL DESIGN DEFICIENCIES
- TEMPERATURE RELATED PROBLEMS* CONTROL OF MINIMUM BATTERY ROOM TEMP. EQUIPMENT QUALIFICATION
MINIMUM COOLING WATER TEMPERATUREMAXIMUM AMBIENT TEMPERATURE
- DIESEL GENERATOR FUEL OIL SYSTEM* CAPACITY OF FUEL OIL STORAGE TANK* FUEL SUPPLY AVAILABLE IN DAY TANK* TORNADO PROTECTION
- SEISMIC QUALIFICATION* FUEL OIL TRANSFER SYSTEM* DG LUBE OIL AND JACKET WATER HEATERS* FIRE PROTECTION tYSTEM IN DG ROOM|* RELIEF VALVES ON bG AIR START ACCUMULATORS
i
I
ELECTRICAL DIS~TRIBUTION FUNCTIONAL INSPRECURRIFING PROBLEM ARIEAS
* PROCEDURAL INADEQUACIES
- LACK OF ACCEPTANCE CRITERIA* MAINTENANCE PROCEDURES* BATTERY TESTING
- ERRORS OR OMISSIONS* INCONSISTENT WITH VENDOR RECOMMENDATIONS* EOP ELECTRICAL LOAD TABULATION
APPROPRIATE PROCEDURES DO NOT EXIST. FUSE CONTROL* MAINTENANCE
RELAY CALIBRATION
-FAILURE TO IMPLEMENT EXISTING PROCEDURES
i
o- Om
r
-~~~~~~~~~~~~CELECTRICAL DISTRIBUTION FUNcRIONAL INSPR.Eg PRIN PR BLEM AREA5
* CONFIGURATION MANAGEMENT
- LACK OF DESIGN BASES* BREAKER/RELAY SETTINGSa TRANSIENT ANALYSIS OF VOLTAGE
DURING LOAD SEQUENCINGPROFILE
- PLANT DOES NOTa BREAKER/RELAY
CONFORM TO DESIGN DOCUMENTSSETTINGS
7
j
I.
EDSFI BRIEFING TO ACRSAUGUST 29.1991
* TWO LEVELS OF UNDERVOLTAGE RELAYS
* PSB-1 REQUIREMENTS
- THE SELECTION OF UNDERVOLTAGE AND TIME DELAYSETPOINTS SHALL BE DETERMINED FROM AN ANALYSISOF THE VOLTAGE REQUIREMENtS O0 THE CLASS 1ELOADS AT ALL ONSITE SYSTEM DIST1IBUTION LEVELS
* GRID STABILITY
* 7 OF 21SECOND
PLANTS INSPECTED HAVE HADLEVEL DEGRADED GRID RELAY
INADEQUATESETPOINTS
PAGE 8
i
EDSFI BRIEFING TO ACRSAUGUST 29. 1991
* PROGRAM RESULTS
- LICENSEE'S CONDUCTING THEIR OWN EDSFIs- INCREASED ATTENTION TOWARDS EVALUATING
AND IMPROVING THE DESIGN BASIS OF THE EDS- IMPROVED THE OVERALL FUNCTIONAL CAPABILITY
OF THE EDS
PAG E 9
r
EDSFI BRIEFING TO ACRSAUGUST 29. 1991
* CONCLUSIONS
- EDSFI FOUND DEFICIENCIES THAT WERE:
* INTRODUCED* PRESENT IN
BYTHE
PLANT MODIFICATIONSORIGINAL DESIGN
- INDICATES NEED FOR BETTER
* ENGINEERING AND TECHNICAL SUPPORT* LICENSEE SELF-'ASSESSMENT PROGRAMS* UNDERSTANDING OF DESIGN BASES* AVAILABILITY OF bESIGN DOCUMENTATION
PAGE 10