ML20214R959
| ML20214R959 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/20/1986 |
| From: | Glynn J NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML20214R948 | List: |
| References | |
| NUDOCS 8609290336 | |
| Download: ML20214R959 (28) | |
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A PRA BASED INTERACTIVE SYSTEM FOR PLANNING REACTOR INSPECTION ACTIVITIES Dr. James C. Glynn*
Division of Risk Analysis and Operations Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Comission Washington, D.C.
20555 A.
INTRODUCTION In 1975, the publication of the U.S. Reactor Safety Study (WASH-1400) provided the nuclear industry with a mechanism to logically assess a plant risk (core melt frequency and off-site consequences) by combining infomation on faults associated with those systems, components and human-actions that are required Eager to capitalize on the inherent usefulness of the for safe operation.
PRA's logic structures, e.g., fault and event trees, and its quantitative results, e.g., system failure probabilities, several U.S. studies were then initiated to see how this risk infomation could be integrated into the NRC's Regulatory process, in particular its inspection and enforcement (IE) procram.
The results of these initial studies however, met with little success in devel-op.ing a methodology that would provide an inspector with the PRA data that could make his decisionmating less subjective. As time went on however, and the number of U.S. operating plants of siultiple designs began to increase significantly, the NRC's Division of Research (RES) chose to re-visit this
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matter by funding its own project. yhich again was to see if and how PRA data This renewed interest by RES in could be melded into the inspection process.
investigating PRA application for inspection purposes was further enchanced at that time by the large number of plant specific PRA being published by industry and the NRC e.g., IREP program, whose risk insights seemed to many to offer considerable promise,in such areas as plant reliability and safety.
Accordingly, in 1983 RES 1r.itiated a program entitled, " Risk Assessment Appli-cation to NRC Inspection" whose initial focus was to thoroughly study the IE inspectior program e.g., responsibilities and actions of resident inspectors, and then g to establish the relationships between plant risk and i
decisions Mmatting large volumes of PRA infomation into a system for use by an inspec-l This particular approach to the project, that is--requiring the PRA analyst to first understand and witness those risk relevant situations and wr.
decisions that are a daily part of an inspectors job, is now seen to perhaps have been the most important programatic decision leading to the
- Dr. James Jenkins presented paper at PRA workshop in Brighton, U.K.,
j May 20-23, 1986.
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2 For example, it was successful development of a PRA based inspection system.
found that the information to support an inspection decision needed to be:
(1) readily accessible, (2) free of PRA jargon, (3) readily recognizable to inspectors with varied technical backgrounds as to its usefulness, and (4) re-sponsive to the types of decisions inherent in an inspector's IE program e.g.,
mandatory inspection procedures.
This research study has now proceeded to where a very promising methodology has been developed, using the Arkansas Nuclear Power Plant PRA as its test case, that can supply reactor inspectors with a significantly useable data base for inspection decisionmaking activities through the use of a desk top PC-computer.
The methodology specifically provides information to the inspector for daily inspection planning based on the risk relevance of the immediate plant status, or, for longer range inspection planning, information that would address his The responsibilities for completing specifically required inspection tasks.
methodology is now being incorporated into a system called PRISIM (Plant Risk Status Information Management) that will contain an inspection oriented data base in conjunction with an interactive capability to enable decisionmakers such as e.g., inspectors, reliability engineers etc. to gain access to essen-tial information, put it into meaningful form, and use it to improve their productivity.
B, PRISIM SYSTEM DESCRIPTION:
The Plant Risk Status Information Management System (PRISIM) is a deci-sion-oriented, user-friendly, menu-driven program that contains data base man-agement and interactive routines to aid inspectors in allocating their efforts p
L towards those areas of greatest impact on plant safety. The program was written for an IBM XT personal computer with special high resolution graphics and a 20 megt-byte hard disc. PRISIM's controlling feature is a data base (1) it manager that essentially performs two main functions for the user:
selects screen images from the PRISIM data base and displays them on a monitor, and (2) controls the PRISIM interactive routine--the portion of the program that calculates the risk status of the plant at a particular time--and then The PRISIM data displays the results of these calculations on the monitor.
t base contains pre-processed or " canned" information that is relevant regard-less of the plant's status, whereas the interactive routine provides the user with PRA data that has been updated to reflect the status of the plant at the The following are brief summary descriptions of the types of informa-moment.
tion presented on PRISIM data base screens and the types of information obtained from PRISIM's plant status interactive routine.
B.1 PRISIM DATA BASE
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The information contained in PRISIM's data base consists of:
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t Dominant Accident Sequences The accident sequences that make the largest contributions to a plant's An inspector can coninand PRISIM to display risk are listed in PRISIM.
He will then see a infomation on a particular accident sequence.
description of the accident sequence, a listing of the most important causes of the sequence, and a description of the relevant recovery actions.
Safety-Related System Importances The PRISIM data base provides four types of risk importance measures for (1) safety assurance importance, (2) risk safety-related systems: reduction importance, (3) risk sensitivity importance, and (4 Table I contains a qualitative definition of these four importance measures from which specific inspection activities can be nificance importance.
correlated unambiguously.
Safety-Related Subsystems PRISIM provides the same four types of risk importance measures forIt also lists the safety-related subsystems that it provides for systems. surveilla If a test is not integral, the components that are not an integral test.
tested are identified.
F Safety-Related Components L
In addition to the same four importance measures, PRISIM provides modified This information infomation if a particular component is of service.
includes lists of single component failures which would disable the system These failure modes are if a specified component is taken out of service.those that are covered by of two categories:
the plant's technical specifications and those that are not.
Support System Interfaces ystems and support To identify dependencies among front-line safety 1
systems (e.g., electrical power and service water PRISIM provides, for each system, a table that shows the support services required by compo-nents in that system.
Component Failure Data _
Two types of information on component failure data are incorpo@ted in th First, PRISIM includes sunnaries of licensee event PRISIM data base.
Second, there are reports (LERs) by component type for the plant.
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comparisons of plant-specific failure data with industry-averaged failure These comparisons highlight plant equipment data for plant equipment.that is more or less reliable than the industry average f the same type.
Fire Zones PRISIM provides a ranking of fire zones at the plant with respect to their An assumption that is inherent in these rankings is importance to risk.
that a fire will fail all equipment in the zone where it occurs.
B.2 PRISIM INTERACTIVE ROUTINE 1re interactive routine allows the inspector to manually input a plant status (40 the program, i.e., components that are currently out of service, and Tw(he instantaneously the recomputed core melt frequency and the dominantS faHur.t scenarios for the new plant configuration.
tive fr.utine has the capability to generate the following information for a given plant state:
The factor by which the instantaneous core melt frequency increases when the specified set of components is out of service, o
The most important. failure scenarios for core melt ranked according to c.
their expected frequencies of occurrence.
A ranking of the unfailed equipment according to their relative contribu-tions to the instantaneous core melt frequency (risk reduction),
o L
A ranking of the failed equipment according to the benefit of restoring o
each to service (risk response).
A ranking of the unfailed equipment according to their relative contribu-tion to instantaneous core melt frequency if the equipment were to be o
taken out of service (safety assurance).
f The factor by which core melt frequency increases due to components being out-of-service is not intended to be literally interpreted, but rather to serve as an index on which an inspector can assess the risk implications of a plant state and then decide whether he should focus his innediate attention on seein to it that he makes those decisions that might prevent any further increase in The interactive routine for the ANO-1 plant contains only the dominant minimal cutsets which are sufficient to allow the program to retain plant risk.
the failure modes that represent about 85 percent of the total expected co melt frequency for the ANO-1 plant.
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5 (1) a routine can be found in its capability to provide an inspe: tor with:
strong case as to whether he should focus his attention on the risk asso with a particular plant status (2) a ranking of the impor the complex interaction effects arising from potential the capability of an inspector, have been properly accounted for in his j
decisionmaking process.
APPLICATION OF PRISIM TO INSPECTION C.
The PRISIM program contains an architecture that was develcped to facilitate l
d the decisionmaking process that an inspector constant must first make the decision, perhaps unwittingly, to either focus his atten-in.
tion in response to existing plant conditions or to ins The following are contained in the U.S. NRC's inspection program (IE Manual).
brief descriptions as to how PRISIM responds to whichever decision type th inspector chooses to follow and a demonstration of exactly what inform provided the inspector to assist him in implementing this decision.
C.1 SCHEDULING DECISIONS At the present time, the PRISIM program contains inspection infomation e
of the most frequently used inspection procedures as defired in the IE M These procedures describe various types of inspections to be perfomed allow a great deal of latitude in terms of where the inspector should focu
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Accordingly, when an inspector is faced with the decision as to how to schedule which components should be inspected most often PRISIM attention.
him by(providing the results of systematic analyses i) the specific inspection decisions associated with the procedure, ing:) the relevant procedural decision categories, i.e., witness corrective maintenance, that can then be directly correlated to th (ii l to Other types of information provided in PRISIM that are relevant to inspection scheduling decisions are trends in component failures and how him.
All of nent reliability performance at the plant compares to ot Figures 1-8 (schedule) which components should be inspected most often.
For provide exact replicas of the information PRISIM can preside an inspe example, if he were to choose to implement procedures 71707, we ca Figures 3-8 those decisions facing the inspector e.g., vich safety-re systems should be emphasized, and the relevant PRA inferr.ation him to schedule what systems he should most frequ methodology that will be described in a soon to be of an inspection program's procedures from a risk perspective.
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6 PLANT RESPONSE DECISIONS C.2 l
The PRISIM system further provides an inspector with an extremely usef l t to assist him in judging whether the risk associated with a current p anThis ca status warrants his immediate attention. rapidly after he has detem The inspector would then call on components known to be out of service.
PRISIM's interactive routine, input this specific plant infomation and alm instanteously be infomed of the effect these conditions have on the plan Assuming the overall core melt frequency e.g., change in plant risk index.
tinue risk index registers a significance increase, the inspector can either con l
to use the interactive routine in deciding where to foc The specific to assist him in deciding where he should direct his atten h
management portions has been discussed previously in this paper; how following example illustrates very clearly how the interactive system wo provide an inspector important information to:
i his immediate attention on the plant status, and (2) decide where his atte should be focused based on prioritization rankin d
from service.
Assume the inspector has visited the control roor and found that a valve in the Eriergency Feedwater System (EFW) and a valve in the EXAMPLE:
Battery and Switchgear Emergency Cooling Syster service.
g plant status.
L The inspector is first presented with the optier of obtaining safety-related information through a direct access path or anIf he inspection procedure path.in this example, the user is first presen For this case, he tion categories addressed in PRISIM (Figure 9). se He is then presented with two options that allow him to sDecify theIf the inspect out-of-service components (Figure 10).
" schematics" option, as Figure 10 indicates he does, he will be Having presented with a list of safety-related systets (Figure 11).
selected the schematic for the EFW, the inspector will next see screen appearing in Figure 12.
on the schematic, the inspector specifies the appropriate va menu and sclects the ECS to specify the valve in this system th the EFW is out of service.
know to be out of service (Figure 13 and 14).
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7 After the inspector has specified the out-of-service components for the EFW and ECS, he selects the "END OF INPUT" option and is presented (1) the factor o'f with a screen (Figure 15) that show three things:
increase in core melt frequency. (2) the componer.ts specified as being out-of-service, and (3) the inspector's options for additional As indicated in the figure, the core melt index is information.
increased by a factor of 70 when the two valves are out-of-service.
To obtain infomation that will help him decide how to focus his efforts on the components not known to be out of service, the user selects the " ranking of safety-related equipment
- option at the He is then presented with a screen that pro-bottom of the screen.
vides a ranking of equipment (Figure 16). For this example, the EFW Train A-to-Train B crossover line is one of the listed components and the inspector knows that maintenance is to be perfomed on a valve in To assess the impact of performing this maintenance, the the line.
user specifies the crossover line as an additional out-of-service He is then presented with an updated increase condition (Figure 17).
in core melt frequency (Figure 18), which he can now use as an aid in deciding whether to respond to the plant status.
The above example was chosen to demonstrate the ease by which the PRISIM interactive routine can inform an inspector that his imediate attention should be to see that plant conditions that could further exacerbate the risk are detemined and possibly therefore avoided.
In this regard, it is worth noting that the equipment ranking, Figure 16 is a direct consequence of the plant status and could only be obtained from PRISIM's interactive system and not the PRA itself.
i The data base management system could be used also by the inspector L
i as a aid in assisting him to effectively focus on the risk of the plant status.
Conclusions The initial development of the PRISIM system for use at the Arkansas Nuclear A
Unit-1 plant by the NRC's resident inspector will be completed in July 1986.
period of testing and evaluating the system will be perfomed over a three l
month period so that information ecn be obtained in tems of how, when andDuring this i
where this system can be most helpful in inspection planring.
period, work will begin on the development of similar systems for the Peach This multi-plant Bottom, Surry, Sequoyah, Grand Gulf, and Zion nuclear plants.
effort will produce inspection systems identical to the ANO-1 system; however, several of the NRC's research programs will be integrated into the process in an effort to produce a methodology for rapidly translatir; PRA results into useable regulatory tools.
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8 In parallel with the development of the decisionmaking syster.s for inspec ility a PC based prototype system will be available this su t
human error probabilities, allowed outage times and surveill Other available to NRC's regulatory engineers for their eva intervals.
i td are training and plant risk evaluations.
Risk Assessnent Application to NRC Inspection, Technical
Reference:
(1)
JBF, Associates.
Progress Report, July 1984:
Risk Assessment Application to NRC Inspection, interim JBF, Associates.
(2)
Progress Report, December 1985:
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f Definitions of Measures of Importance Table 1:
Measure of Definition Importance The factor by which risk increases when the Safety Assurance equipment'is out of service.
The factor by which risk decreases when the Risk Response out-of-service equipment is returned to service.
The decrease in risk when the equipment is (When assumed to be perfectly reliability.
Risk Reduction normalized to the average risk, these results represent the likelihood that the equipmentwo were to occur.)
The rate at which the risk changes with changes Risk Sensitivity in the equipment failure probability (or frequency).
Combined risk reduction importance and risk
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(Equipment is grouped Risk Significance sensitivity importance.
Equip-according to risk reduction importance.
ment with a high risk sensitivity importance is then moved to the next higher group.)
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PROCEDURE IN9PECTION PROCEDURE TITLE NUMBER Operational Safety Verification
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71797 Monthly Survelliance Observation 61726 Monthly Maintenance Observation 627e3 t
ESF System Walkdown Onsite Followup of Events at Operating Reactors 7171s Onsite Followup of Written Reports of Nonroutine Events 92799 IE Bulletin /Immediate Action Letter Followup 92792 92793 Review of Plant Operations 73711 Review of Plant OperationsInformation Type Bulletin 72793 92717 Followup IE Circular, Surveillance - Refueling 61781 Maintenance - Refueling' 62793 Independent inspection Effort 92796 Followup - Headquarters Requests 92794 Followup - Regional Requests 92795 j
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DECISIONS ASSOCIATED WITH INSPECTION PROCEDURE 71767--
VERIFICATIDN--FOR WHICH PRA INFDRMATION IS AVAILABLE OPERATIONAL SAFETY i
DAILY INSPECTION 1.
What emphasis should be gl,ven to the owleting plant status?
(Interactive Routine) i WEEKLY INSPECTION Which safety-related subsystems should be emphastred?
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BlWEEKLY INSPECT!DN What safety-related tag-outs should be emphastred?
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which fire zones should be emphasized?
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When a plant tour is performed, tour, what information collected during the plant 3.
Based upon the the ewisting plant status? (Interactive emphasis should be given to Routinel i
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i IMPORTANCE SUBSYSTEMS GROUPED BY RISK SIGNIFICANCE HIGH RISK SIGNIFICANCE Train A and Switchgear Emergency Cooling System--Chill Water Battery Train B Battery and Switchgear Emergehcy Cooling System--Chill Water Emergency Even DC Power Initiation and Control System--Initiation Emergency Feedwater Channel A Initiation and Control System--Vector C Emergency Feedwater ITurbine Drivent Emergency Feedwater System--Train B l
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SUBSYSTEMS GROUPED BY HIGH RISK SIGNIFICANCE
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Service Water System--Loop System--Loop 2 Service Water (failure to close)
Safety / Relief Valves MODERATE RISK SIGNIFICANCE Emergency Even AC Power System--Inttiation Chan.9e1 9 InttIatlon and Contrel Emergency Feedwater
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SUBSYSTEMS GROUPED BY RISK SIGNIFICANCE IMPDR i
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Channel 3 Engineered Safeguards Actuation System--Ana og Injection System--Trains A and B High Pressure Low Pressure Recirculation System -Train A B
Low Pressure Recirculation System--Train Safety / Relief Valves (fallure to open)
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Engineered Safeguards Actuation System--Digi tal Channel i
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I SUBSYSTEMS GROUPED BY RISK SIGNIFICANCE IMPORTANCE (CON LOW RISK SIGNIFICANCE Injection System--Train A Low Pressure Injection System--Train B Low Pressure A
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Battery and Switchgear Rooms Emergency Cooling System North System I.
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SAFETY-RELATED SYSTEMS FDR WHICH SCHEMATICS ARE AVAILABLE Rooms Emergency Cooling System and Switchgear
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South Battery 3.
DC Power System Emergency AC Power System
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Initiation and Control System 5.
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High Pressure High Pressure Recirculation System i
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IS THE RISK FACTOR WITH TSE F0t. LOWING EQUIPMENT DUT OF SERVICE l
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Improvement from repair Ranking of safety-related equipment 2.
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RANKING OF EDUIPMENT NOT KNOWN TO BE OUT OF SERVICE and Switchgear Room Cooling System--CW Train B fails Battery I.
Blockage of EFW Train A-to-Train B Crossover Line 2.
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Auwiliary Cooling Water System Isolation Valve CV3643 falle Isolation Valve CV3928 falls Intermediate Cooling Water System 5.
ID-32D falls Initiation and Control--Vector Signal Path Initiation and Control--Vector Signal Pathe 2D-22D and 3D-22D fall 6.
EFW EFW 7.
to reclose Both safety / relief valves open and fall B.
High Pressure Injection System Pump P36C falls 18.
EFW Initiation Signal Paths Acel-ace 4 and B001-9D94 fall 9
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RISK IS TE RISK FACTOR WITH TE FOLLOWING EQUIPMENT GUT OF 178 Emergency Feedwater System--Train A fallsBattery and Swltchgear Roo Blockage of EFW Train A-to-Train B Crossover Line i
1 E Nil FOR ADDITIONAR. IW ORMATION Improvement from repair 3.
Ranking of safety-related equipment Return to Master Menu 4,
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,p REQUEST FOR PHILADELHIA ELECTRIC COMPANY'S ASSISTANCE IN PROVIDING CERTAIN INFORMATION PERTAINING TO PRISIM To assist our efforts in developing the Peach Bottom PRISIM system, our contractors (the Idaho Nuclear Engineering Laboratory and JBF Associates) require the follcwing infonnation:
1.
Transient Response Implementation Plan (Emergency Procedures) 2.
Index for Test and Maintenance Procedures 3.
P&ID diagrams 4.
Index of Functional Control Diagrams 5.
One-line diagrams of AC & DC systems with circuit breaker designations 6.
Index of lay-out diagrams pertaining to Reactor Bldg (including Control Room) and Aux. Bldg 7.
Technical Specifications 8.
Updated FSAR