ML19291A421
| ML19291A421 | |
| Person / Time | |
|---|---|
| Issue date: | 04/13/1979 |
| From: | Farmer W NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Mattson R, Murley T, Stello V Office of Nuclear Reactor Regulation, NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| NUDOCS 7905080405 | |
| Download: ML19291A421 (11) | |
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,J APR ! 9 1979 MEMORANDUM FOR: Those on Attached List FROM:
William S. Farmer Research Support Branch
SUBJECT:
SAFETY-RELATED OPERATOR ACTION STUDY Enclosed are the handouts from the midyear review (February 8, 1979) and subsequent program review (February 9,1979) meetings held in Silver Spring and Bethesda on the ORNL safety-related operator action study.
This program was set up at ORNL in response to an SD request to establish a data base which can be used to assess proposed criteria or develop criteria for determining whether or not required safety-related operator actions should be automated and the time requirements for operator action.
The handouts include an overview of the preliminary results of the study concerning the concept of an operator time margin which may be useful in developing interim criteria.
William S. Farmer Research Support Branch Division of Reactor Safety Research
Enclosures:
as stated cc:
F. R. Mynatt, ORNL T. F. Bott, ORNL P. M. Haas, ORNL l_
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e ADDRESSEES FOR MEMORANDUM DATED "7
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T. E. Murley R. J. Mattson, DSS V. Stello, 00R G. Arlotto, SD H. Thornburg, IE S. Hanauer, DSS R. Vollmer, DSE V. Benaroya, DSS R. Savio, ACRS L. S. Rubenstein, NRR T. Scarbrough, SD H. Balukjian, DSS D. H. Beckham, NRR R. Wright, ACRS T. Cintula, OMPA G. Constable, IE H. Li, DSS J. Milhoan, SD D. Tondi, PSYB W. E. Vesely, RES G. Chipman, DSE S. Newberry, DSS PDR (2)
e a
SAFETY-RELATED OPERATOR ACTION STUDY MIDYEAR REVIEW February 8, 1979 ATTENDEES Name Organization Gary Bennett NRC/RES Bill Farmer NRC/RES Tom Scarbrough NRC/SD Vic Benaroya NRC/ DSS Harry Balukjian NRC/ DSS Richard Savio NRC/ACRS Fred Mynatt ORNL L. S. Tong NRC/RES Don H. Beckham NRC/NRR Terry Bott ORNL Paul Haas ORNL February 9,1979 Gary L. Bennett NRC/RES William Farmer NRC/RES John Olshinski NRC/ DSS N. R. Anderson NRC/SEPB Hans Schierling NRC/DSE Gordon Chipman NRC/DSE Paul Collins NRC/0LB Kevin Mahan NRC/0LB B. A. Boger NRC/0LB Don Skovholt NRC/NRR Scott Newberry NRC/ DSS Don Beckham NRC/DPM Tom Scarbrough NRC/SD Matthew Chiramal NRC/PSYB
i e
O SAFETY-RELATED OPERATOR ACTI0ilS A SURVEY OF OPERATOR OPINION CONDUCTED BY OAK RIDGE NATIONAL LABORATORY O
= w
e Oper. No.
(ORNL Use Only)
I.
Survey Purpose, Goals and Procedures This survey is being performed as part of a research program which will help supply data necessary to establish industry (American Nuclear Society) design criteria for automation of safety-related actions.
The central question addressed by the criteria is whether certain key safety-related actions should be performed by the licensed operator or whether they should be performed by automatic action, with verification by the operator.
It is also hoped that the research will ultimately help identify and suggest ways to improve any conditions that tend to decrease the likelihood of reliable operator performance.
II.
Operator Personal Data (NOTE:
All Information (Raw Data) Collected in this Survey is Confidential)
AGE:
CURRENT POSITION:
TIME IN CURRENT POSITION (Years):
PREVIOUS EXPERIENCE (Directly Applicable to Nuclear Power Plant Operation):
Employment:
Position Years Exoerience Training:
Education:
TIME SINCE MOST RECENT TRAINING /REQUALIFICATION EXAM:
1
i a
Oper. No.
(ORNL Use Only)
III. General Model of Operator Response We are assuming that response of an operator following annunciation of an emergency / abnormal event can be generally categorized into four phases as follows:
a.
Initial Shock - a very brief period immediately following the event alarm during which the operator is surprised or alerted and is essentially inactivated (except perhaps for acknowledging the alarm).
b.
Diagnosis - the operator evaluates infonnation from alarms, annunciators and indicators, identifies what event has occurred, plans his action and initiates required immediate action.
c.
Immediate Action - operator carries out immediate actions, both verification of automatic responses and manual actions required to bring the plant to a safe condition.
d.
Subsequent Action - operator carries out over a longer time period subsequent actions necessary to maintain the plant in a safe condition, prevent further damage or release of radioactivity, etc.
In your opinion, is this a reasonably accurate general description of operator response?
[] Yes
[] No Please ccmment if desired:
IV.
There are twenty potential emergency / abnormal events listed below.
Please circle the numbers _ of the events you have experienced during plant operation.
Then answer the questions A, B, C and D regarding these twenty events.
1.
Loss of Feedwater Flow
- 11. Main Steam Line Break 2.
Loss of Condensor Vacuum 12.
Steam Generator Tube Failure 3.
Emergency Baration 13.
Excessive Primary Plant Leakage 4.
Loss All AC Power 14.
Reactor Coolant Pump Trip 5.
LOCA 15.
Loss of Component Cooling 6.
- 16. Continuous Rod Withdrawal 7.
Loss of Condensate /Cond.
17.
Loss of RHR System Booster Pump 18.
Safety Injection 8.
Fuel Handling Emergency
- 19. Rad. Mon. System-High Activity 9.
Reactor Coolant Pump Vibration Alarm 10.
High Coolant Activity 20.
Loss of Service Water 2
e Oper. No.
(ORNL Use Only)
IV.A.
Shock (Psychological Stress) 1.
Of the events listed, select thr2e which in your opinion are most likely to result in psychological stress (tension), as evidenced by tightening of stomach muscles, excessive sweating, dryness of mouth, pounding of the heart, or other typical anxiety symptoms.
Event Numbers:
Please explain briefly what factors (for example, potential harm to yourself or to the public, potential damage to equipment) make these events more stressful than others.
2.
Of the events listed, select the three which in your opinion are the least likely to result in stress, and explain why they are not as stressful.
Event Numbers:
Explanation:
IV.B.
Diagnosis 1.
Of the events listed, select three which in your opinion are the most difficult to diagnose.
Event Numbers:
Please explain briefly what causes the difficulty; for example, symptoms are similar to other events, symptoms are not annunciated quickly or precisely enough, etc.
Explanation:
NOTE: Use Reverse Side as Desired for Additional Comments / Explanation.
3
8 e
Oper. No.
s (ORNL Use Only)
IV.B.
Diagnosis (Continued) 2.
Of the events listed, select the three which in your opinion are the least difficult to diagnose, and explain what factors make them easier to diagnose.
Event Numbers:
Explanation:
IV.C.
Time Response 1.
Considering both stress level and difficulty of diagnosis, select three of the events listed hich you feel would require the most time for yo J to recover from initial shock, correctly diagnose the event, plan your action and initiate required immediate action.
Indicate the level of stress (moderate, high or severe) you would anticipate for these events.
(NOTE: These three events may or may not be three that were selected in Items A and B.'
Event Numbers:
Stress Levels:
Estimate the time you think would be adequate for you to respond to one of these events.
2.
Considering stress level and difficulty, select three of the events listed which you feel would require the least time for you to recover, diagnose, plan and initiate immediate action.
Indicate the stress level you would anticipate for these events.
Event Numbers:
Stress Levels:
Estimate the time you think would be adequate for you to respond to one of these events.
NOTE:
Use Reverse Side as Desired for Additional cot.ments/ Explanation.
4
e Oper. No.
(ORNL Use Only)
IV.C.
Time Response (Continued) 3.
Consider the three events you have selected as requiring the most time for response. Suppose that a large number of licensed operators experienced these events independently.
For each time that is listed, estimate the percentage (%) of operators you think would have completed
" response" (that is, recovery from shock, diagnosis, planning, and initiation of required action) by that time or before. Assume that all symptoms are indicated as described in emergency / abnormal procedures and that all safety equipment performs as designed.
(NOTE:
Each entry should be a cumulative percentage, that is, it should include,-
the total percentage you think would have responded by that time.)
Make similar estimates for the events you selected as requiring the least time for response.
Number of Operators Responding Correctly in Specified Time Time "Most Time" Events "Least Time" Events 10 sec.
30 sec.
1 min.
2 min.
5 min.
10 min.
20 min.
30 min.
V.
Specific Events.
There are five specific events listed on the following page.
If each of these events occurred many times at different plants and with different operators, there would be a " spread" or distribution in response times because of the variability in operators and specific circumstances of the event.
For each of the events, estimate the "mean" and " maximum" times to respond. The "mean time" is the time within which you would expect the response to be completed in 50% or more of the occurrences.
The
" maximum time" is the time in which you feel the response would be completed "9 times out of 10",
i.e., in 90% of the occurrences.
Recall that " response time" refers to the time required to recover from initial shock, diagnose the event, plan action and initiate required immediate action.
5
e Oper. No.
(ORNL Use Only)
Mean Time Maximum Time Event (50%)
(90t) 1.
Steam-generator tube rupture 2.
Continuous rod with-drawal in auto 3.
Loss of all AC power 4.
Automatic safety injection 5.
High coolant leak VI.
Discreet Manual Actions For most of the events described in the emergency / abnormal operating procedures the operator is required to perform one or more manual actions.
In our model of operator response we are attempting to separate these discreet manual actions from other tasks such as diagnosis, verification of correct operation of automatic systems, etc.
These actions are best explained by example.
The actions taken after a LOCA may include tripping the reactor coolant pump for the affected loop or loops. Tripping the pump for one loop is one " discreet manual action". Manual reactor or turbine trip, energizing an electric relief valve, activating a stop valve, or closing a breaker are other examples. As for question V, estimate a "mean" and " maximum" time for completing a discreet manual acticn such as these.
Mean Time (50%):
Maximum Time (90%):
~6
s Further Discussion In the future, it may be desirable to discuss in more detail with station operators specific experiences with abnormal events that have occurred.
If duty time were available would you like to participate in such interviews?
~.J Yes O No If "Yes", please print your name below:
(Name)
D m
4 SAFETY-RELATED OPERATOR ACTIONS (N660)
T. F. Bott P. M. Haas Engineering Physics Division Oak Ridge National Laboratory Oak Ridge, Tennessee 37830 Presentation Made To:
U. S. Nuclear Regulatory Commission Washington, D. C.
February 8,1979
--6
SAFETY-RELATED OPERATOR ACTIONS OBJECTIVE ESTABLISH A DATA BASE WHICH CAN BE USED TO ASSESS PROPOSED CRITERIA OR DEVELOP CRITERIA FOR DETERMINING WHETHER OR NOT REQUIRED SAFETY-RELATED OPERATOR ACTIONS MUST BE AUTOMATED.
APPROACH GATHER APPLICABLE DATA FROM OPERATING EXPERIENCE.
CORRELATE OPERATOR EXPERIENCE TO SIMULATOR RESULTS.
PERFORM SIMULATOR EXPERIMENTS.
DATA BASE WILL CONSIST OF CALIBRATED SIMULATOR RESULTS.
I B
PROGRAM PLAN PRELIMINARY ASSESSMENT - APPR0XIMATELY SIX MONTHS (0.6 PY)
BACKGROUND STUDY - N660 CRITERIA, EVENTS, ACTIONS ASSESS AVAILABILITY OF APPLICABLE DATA SUGGEST PROCEDURES FOR DATA COLLECTION COLLECT INITIAL DATA SUMMARIZE FINDINGS / RECOMMENDATIONS FIELD DATA COLLECTION - ONE YEAR (2.5 PY)
COLLECT AND ANALYZE DATA FROM OPERATING EXPERIENCE SIMULATOR CALIBRATION EXPERIMENTS - 18 MONTHS (3 PY)
DEVELOP / ADAPT METHODS FOR FIELD CALIBRATION PERFORM SELECTED SIMULATOR EXPERIMENTS FOR CALIBRATION ANALYZE RESULTS TO DEVELOP CALIBRATION SIMULATOR EXPERIMENT PROGRAM - 18 MONTHS (3 PY) e
a s
Preliminary i
Search hf Events and Sites oJ Interes F
\\[
Site Work I
Consultant r
Support
/
\\ I I f Documentation of Literature Search Specific Events PAQ Survey j
Operator Interviews!
Analysis of Results PAQ Survey i
PHASE - 1 PRODUCT Evaluation of Data Availability Outline Methods for Collection Preliminary Results for N660 Document Events l
Operator Estimates Non-Nuclear Data
s WORK ACCOMPLISHED 1.
STUDY OF N660 CRITERIA, THEIR HISTORY AND BACKGROUND.
2.
IDENTIFICATION OF ACCIDENT EVENTS OF INTEREST AND SPECIFIC OPERATOR ACTIONS.
3.
SEARCH NSIC DOCKET FILES FOR SPECIFIC EVENTS AT FIVE SELECTED SITES (10 UNITS).
4.
PREPARE OPERATOR SURVEY TO COLLECT QUALITATIVE AND QUANTITATIVE DATA (OPINION).
5.
SITE VISITS - SURVEYS, SITE RECORDS, INTERVIEWS.
6.
DOCKET SEARCHES, KEY EVENTS, ALL BWR'S AND PWR'S.
7.
ANALYSIS OF DATA (90% COMPLETE).
8.
LITERATURE SEARCH FOR NON-NUCLEAR DATA.
WORK TO BE DONE 1.
COMPLETE ITEMS #7 AND #8 ABOVE.
2.
PREPARE
SUMMARY
OF RESULTS AND CONCLUSIONS OF PHASE I.
N660 CRITERIA ANS CC'4MITTEE - VENDORS, UTILITIES, A/E, NRC HISTORY 1973 - INITIATED (FIRST DRAFT STANDARD) 4/74 - AEC " NEGATIVE WITH COMMENT" VOTE 2/75 - STANDARD REWRITTEN - ANS-50 BALLOT COMMENTS " YEA" AND "NAY" 11/76 - AFTER MANY REVISIONS, DRAFT RELEASED FOR TUC 6/77 - COMMITTEE REORGANIZED, NEW CHAIRMAN CURRENT - REVIEWING COMMENTS AND MODIFYING REQUIRED OPERATOR ACTION - MANUAL ACTIONS USED TO PREVENT VIOLATION 0F DESIGN REQUIREMENTS (DESIGN BASIS EVENTS, CHAP.15)
BASIS IS TIME REQUIREMENT - EXTENSION OF " TEN-MINUTE RULE" LONGER TIMES FOR - MORE SEVERE EVENTS LESS FREQUENT EVENTS LESS FAMILIAR EVENTS I.E., TIME SHOULD INCREASE WITH STRESS AND DIFFICULTY OF DIAGNOSIS
EQUIPMENT AND TIME OPERATOR ACTION PROCESS MARGIN DELAY TIME DELAY TIME A
A A
/
\\ /
\\/
\\
T T
T T
I E
A j
C l
TO l
l l
T = EVENT INITIATION O
T = EVENT ALARM E
T = OPERATOR ACTION ALARM A
T = INITIATE OPERATOR ACTION y
T = COMPLETE PROTECTIVE ACTION C
T = COMPLETE PROTECTIVE FUNCTION: REACll DESIGN REQUIREMENT LIMIT 1
I
PWR INCIDENTS BORON DILUTION ACCIDENT LOSS OF FEEDWATER LOSS OF ALL A.C. POWER FIRE IN THE PLANT LOCA MAIN STEAM LINE RUPTURE S/G TUBE RUPTURE LOSS OF CONTROL ROOM OVER PRESSURIZATION INADVERTENT SAFETY INJECTION
BWR INCIDENTS LOAD REJECTION TURBINE TRIP MSIV CLOSURE RECIRCULATION PUMP TRIP LOCA LOSS OF FEEDWATER INSTRUMENT LINE RUPTURE FAILURE OF MAIN CONDENSOR OFF-GAS SYSTEM PLANT FIRE LOSS OF CONTROL ROOM RELIEF VALVE STUCK OPEN LOSS OF A.C. POWER
POTENTIAL COOPERATING SITES (ANS 51.4 MEMBERS)
PLANT TYPE NSSS VENDOR SIZE (MWe)
COMMERCIAL OPEPATION ZION 1 PWR WESTINGHOUSE 1040 10/2/73 ZI0tt 2 PWR WESTINGHOUSE 1040 11/74 QUAD CITIES 1 BWR GENERAL ELECTRIC 789 8/72 QUAD CITIES 2 BWR GENERAL ELECTRIC 789 8/72 DRESDEN 1 BWR GENERAL ELECTRIC 207 8/1/60 DRESDEN 2 BWR GENERAL ELECTRIC 794 7/70 DRESDEN 3 BWR GENERAL ELECTRIC 794 10/71 PEACH BOTTOM 2
BWR GENERAL ELECTRIC 1065 7/74 PEACH BOTTOM 3
BWR GENERAL ELECTRIC 1065 12/74 CONN. YANKEE PWR WESTINGHOUSE 575 1/1/68 MILLSTONE 1 BWR GENERAL ELECTRIC 1100 12/28/70 MILLSTONE 2 PWR COMBUSTION ENG 828 12/75 BROWNS FERRY 1
BWR GENERAL ELECTRIC 1067 8/1/74 BROWNS FERRY 2
BWR GENERAL ELECTRIC 1067 3/1/75 BROWNS FERRY 3
BWR GENERAL ELECTRIC 1067 3/1/77
NO. OF NO. OF OPERATOR SPECIAL REPORTS Pj. ANT EVENTS LOGS CHECKED SURVEYS AVAILABLE Zion 1,2 25 Reactor Operator, Shift 21 Supervisor, Control Room, Computer Printout Connecticut Yankee 8
Reactor Operator, Computer 9
Plant Information Printout Report Peach Bottom 2,3 11 Reactor Operator, Shif t 4
Plant Upset Report Supervisor, Computer Printout Dresden 1,2,3 9
Reactor Operator, Shift 6
Deviation Report Engineer, Control Room Quad Cities 1,2 8
Reactor Operator, Shift 4
Deviation Report Engineer, Control Room
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' * ' ; operator observed.an' abrupt 70 MWe decrease in plant electrical outpuc'.
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Coinciding.with thief an. increase.in the "L" relief valve discharge line.
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'te=peratur. e a,s'. indicated?o.n[TR2-.02,103.wa s ob's erved. '. '.. v: -s,
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., ' '. " ',.. personnel. took : he,following.,-Surmisin'g,thatl the.'?L"'rel'ief valf.e'.had lif te
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HPSW m.os wer.e placed into operation in the: ~~.. ' ".
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Ia an attempt to reseat.t th.w,se valve,? reactor - pressure was reduced'from e. 9 adr
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atelyj3:45 a.uf atgabout 200 pst. reacter.7
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1 l A F 6 F 9 7 4 0 1D 0 0 0 1 0 A 8 TAD H T WI S S 0 D T 6 R N O E C V E R E F N T \\ N O \\ A T \\ L \\ W P Y N 0 l \\ I D L 4 E N N \\ A O C \\ \\ V R T T E \\ \\ E E P W N D K K C C \\ \\ O O \\ W \\ Y D D W l \\ l D l \\ I C C g \\ \\ g \\ R R Y D D N N \\ h \\ \\\\ D Y Y \\ \\ \\ D \\ \\ D Dd \\\\ N \\ i \\ \\ G W N O I R L T L S O F E N R U P O T 0 O M C C H l U C U A R S T S P E I E C O S R C E A L U C N V R S C R R N I E L R B A O S T O I W R E V I C O P T T E O U R A N M T R D T O F R E E G E R T P P T P A T A I U I S IL T T R H R E E O I A S T S T R R D b. c.. a. 1. 2 3 4 4 S. 7 8 i
OPERATOR SURVEY FORMS BACKGROUND PERSONAL DATA OPINION ON GENERAL MODEL CRITICAL INCIDENT APPROACH (20 EVENTS) MOST (LEAST) SEVERE EVENTS - WHY? MOST (LEAST) DIFFICULT TO DIAGNOSE - WHY? GREATEST (LEAST) TIME TO RESPOND TIME RESPONSE ESTIMATES SPECIFY DISTRIBUTION ON MOST/LEAST EVENTS SPECIFY 50% AND 90% CONFIDENCE ON FIVE SPECIFIC EVENTS
RESULTS PWR " CRITICAL INCIDENT" QUESTIONS NOTE: 30 RESPONDENTS AT TWO SITES EVENT CHECKS PRIMARY REASON (S) MOST STRESSFUL
- 1. LOCA 25
- 1 - Consequences
- 2. Main Steam Line Break 21
- 1 - Consequences
- 3. Station Blackout 13 None ( #3 Uncertainty)
LEAST STRESSFUL
- 1. High Activity Alarm 13 None el Consequences Rad. Mon.
- 2 Control
- 4 Demands
- 2. High Coolant Activity 10 Mone (d2,#4) 3a. Loss Condensate /Cond.
9 None (e2,d4) Booster
- b. Reactor Trip 9
Frequeit Occurrence, Experience MOST DIFFICULT
- 1. S/G Tube Failure 20 None #2 - Similar
- 1 - Inadequate
- 2. Excessive Primary System 16 None (#2,#1)
Leakage
- 3. Main Steam Line Break 11 None (#2)
LEAST DIFFICULT
- 1. Reactor Trip 22
- 1 - Adequate (First Out)
- 2. Loss FW Pump 15 None (#1)
- 3. RCP Trip 13 None (#1)
" REASON" CODES - STRESS MOST STRESSFUL LEAST STRESSFUL 1. Consequence Potentially Major; Hazard to No Major Consequences Self, Plant Personnel, Public; (7) Potential Major Plant Damage (18) 2. Plant Control Fear of " Losing Control of Plant Under Control, Plant";Related to Both Not Likely to Get Out Consequences and Performance of Control (5) and Uncertainty (3) 3. Uncertainty Not Clear What is Happening, Event Readily Identified, What Action to Take or What Subsequent Action Events Sequence of Events Will Follow Clearly Understood and (6) Anticipated (0) 4. Performance Fear of Performing Poorly, Demands on Operator Demands e.g., Allowing Plant to Scram Performance Not Severe, when Prompt, Accurate Action No Fear of Failure (13) Could Prevent Scram (4) 5. Other Inexperience with Event; Experience with Event; Event Cannot be Affected By Backup Systems (4) Operator Action (5)
REASON CODES - DIAGNOSIS MOST DIFFICULT LEAST DIFFICULT
- 1. Annunciation Adequacy Inadequate for Prompt Detection Immediate, Direct (i.e., No Immediate Alarm or Annunciation (14)
Alarm Not Directly Related to Event) (4)
- 2. Discrimination of Symptoms Well Annunciated, but Symptoms Clearly Symptoms More Than One Event has Similar Delineate which Symptoms; Cannot Distinguish Event; Multiple which is Occurring (12)
Indications of which Event (6)
- 3. Information Precision Adequate Annunciate; Symptoms Detail of Information Clearly Indicate Event, but Either Adequate or Details not Sufficient to Totally Unnecessary to Indicate Proper Action (1)
(5)
- 4. Event Time Event Time is Long; Symptoms Event Time Short, Take Time to Develop After Diagnosis either Alarm; Tension Builds (2)
Immediate or Not Required (0)
- 5. Other Inexperience with Event; Experience and Training Rely on Someone Else (2)
Give Great Confidence Can Recognize (1) +
RESULTS PWR " CRITICAL INCIDENT" OUESTIONS (CONTINUED) EVENTS REQUIRING THE MOST TIME FOR RESPGNSE EVENT CHECKS STRESS LEVEL AVERAGE TIME (SEC.)
- 1. LOCA 21 High (2.05) 122
- 2. Main Steam Line Break 20 High (1.89) 107
- 3. S/G Tube Failure 10 High (1.78) 98 Average of All Time Estimates = 111 Sec.
Correlation: 45 Out of 84 Corresponded to "Most Stressful". 31 Out of 84 Corresponded to "Most Difficult". EVENTS REQUIRING THE LEAST TIME FOR RESPONSE EVENT CHECKS STRESS LEVEL AVERAGE TIME (SEC.)
- 1. Reactor Trip 18 Moderate (1.38) 32
- 2. Loss FW Pump 14 Moderate (1.58) 24
- 3. Continuous Rod With-drawal 8
Moderate (1.25) 36
- 4. Loss Candensate/Cond.
Booster Pump 8 Moderate (1.0) 67 Average of All Time Estimates = 35 Correlation: 31 Out of 81 Corresponded to "Least Stressful". 35 Out of 81 Corresponded to "Least Difficult".
t O _d = 6 I i l i i i l I 4 .AW Wa 1, -- g } O 3 e m z .r - ( e c I I i 1 i t I i i I 6 e o o o c GQ w N DNICN0dS33 IN3]B3d 031VWIIS3 .e 9 e
1 i I i 2.0 DisTra:suitord $3sso ca prverage of E shmoder, Pu]2. "Kost Time" Events O i.c O q) C.c L e E R g Medran = o. 37 min. s 90% = 1.9 mn. ,,,e 9 5% =
- 3. I min.
qq1, 1 7. 6 m in. 99.47o : 19 w n. - 2.0 t i i i i 20 50 90 95 99 99.9 99.99 Perc ent Respondmg
PWR OPERATOR SURVEY RESULTS th th 50 PERCENTILE 90 PERCENTILE EVENT i (sec) s (sec) i (sec). s (sec) Steam Generator Tube Rupture 67 71 287 670 Automatic Control Rod Withdrawal 21 17 49 63 Station Blackout 42 84 93 137 Automatic Safety Injection 34 48 75 109 liigh Primary Coolant Leak Rate ( < 100 gpm) 125 190 413 774
KEY EVENTS EVENT ACTION TIME REQUIREMENTS PWR S/G Tube Rupture Initiate RCS Cooldown and 9 Minutes after Low Depressurization Pressure Trip Baron Dilution Terminate Manually s 45 Minutes Loss of A/C Power Place DHRS in Service (May None Specified Be Required after Diesels Come On Line) Loss of Service Water Reduce Power and Remove RCP At least 10 Mir.utes from Service Inadvertent Safety Assure S.I. is Inadvertent ASAP Injection and Stop It BWR Relief Valve Opens Try to Reclose Valve; If Not ASAP Inadvertently and Successful, Shutdown the Reactor, Will Not Reclose Initiate Torus Cooling Rupture of Primary Shutdown and Isolate Leak -< 10 Minutes Instrument Li"e Rupture of Off-Gas Clear Area of Personnel, Initiate by 1 Minute System Isolate Affected System Loss of A/C Power Maintain Pressure and Water As Required Levels by Manual Operation of Relief Valves and RCIC
SUMMARY
OF KEY EVENT SEARCil EVENT f!0. IDENTIFIED NO. FOUND NO. WITH DATA N0. WITH SITE DATA Off-Gas Explosion 17 16 7 6 BWR Relief Valve Stuck Open 42 45 8 4 Loss of AC Power 12 9 0 0 ' Inadvertent SI 56 49 44 18 PWR [ Loss of Service Water 3 3 1 0 Loss of AC Power 16 14 2 2 'S/G Tube Leak 20 17 1 0
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0 m m l m m e m m cm e o m e 'N oo e mo e x o >- es N 3 _m e o ,2 wm o0 e m n Q. O w g zo OS W i- '. - e m c e - m 0 - o l l - No O o 6 (u!w) 31NI.L 3SN0dS3B e O m* ,e. O 4
h SuffiARY OF DATA COLLECTED Ofl KEY EVErlT ACTI0flS DATA SURVEY th th EVENT ACTION REQUIRED FSAR MEDIAtl 90 PERCENTILE MEDIAN 90 PERCENTILE Inadvertent safety injection Determine cause and secure ASAP 1.7 min. 6.0 min. .? min. 1.25 min. S.I. (2.0 min.) Off-gas system rupture Clear area of personnel < 2 min. 6.3 min. NA NA NA' Isolate affected SJAE ASAP 13.6 min. NA NA NA Relief valve sticks open Recognize problem and ASAP 5.1 min. NA .35 min. 1.0 min. attempt to close valve Initiate torus cooling ASAP 9.2 min. NA NA NA
- Mean-time assuming log-normal response time distribution.
CONCLUSIONS AND RECOMMENDATIONS ON N660 CRITERIA CONCLUSIONS e THE CONCEPTUAL MODEL OF OPERATCR RESPONSE IS REASONABLE FOR SOME EVENTS, INAPPLICABLE FOR OTHERS. THE USE OF A TIME MARGIN FOR OPERATOR RESPONSE IS A REASONABLE e APPROACH FOR INTERIM CRITERIA, I.E., UNTIL A THOROUGH HUMAN FACTORS STUDY IS COMPLETED. INCREASING THE TIME MARGIN WITH STRESS AND DIFFICULTY OF DIAGNOSIS e IS REASONABLE; QUANTIFICATION OF STRESS AND DIFFICULTY IS EXTREMELY DIFFICULT. OPERATOR INPUT IS VALUABLE. WE DO NOT HAVE EN0 UGH DATA TO MAKE A JUDGEMENT ON THE CURRENTLY e RECOMMENDED TIME MARGINS. HOWEVER, DATA TO DATE 00 NOT SUPPORT SUGGESTION OF TIMES SIGNIFICANTLY GREATER THAN CURRENTLY PROPOSED (E.G.,ONE-HOUR). RECOMMENDATIONS PROCEED WITH DEVELOPMENT OF INTERIM CRITERIA BASED ON BEST AVAILABLE e INFORMATION. MODIFY TIME MARGINS AS DATA FROM THIS AND/OR OTHER PROGRAMS ACCUMULATES. INVESTIGATE POSSIBLE WAYS TO ACCOMMODATE " SYMPTOMATIC RESPONSE". e
e s CONCLUSIONS AND RECOMMENDATIONS ON DATA AVAILABILITY EXPENSIVE CONCLUSION - DATA COLLECTION IS NOT IMPOSSIBLE, JUST VERY TTTf93M PROBLEMS POTENTIAL FOR " PURE" DATA BASE IS SMALL - JUDGEMENT, EXTRAPOLATION e NECESSARY. e DEFINING EVENTS / ACTIONS - N0 " STANDARD" EVENTS OR PROCEDURES. DEFINING WHAT TO MEASURE - JUDGEMENT REQUIRED; CASE BY CASE VARIATION. e DOCKET SEARCHING NON-TRIVIAL - NO SINGLE SOURCE; LOGGING PROBLEMS. e CURRENT LER FORM HAS LESS DATA THAN OLDER FORMS OR SPCCIAL REPORTS. e INTERPRETATION OF SITE DATA, ESPECIALLY COMPUTER OUTPUT, REQUIRES e SKILLED SITE PERSONNEL. SITE ACCESS AND AVAILABILITY OF SITE PERSONNEL LIMITED. e e OPERATOR SURVEY BY SITE PERSONNEL, NOT EFFECTIVE. e SITE RECORDS KEEPING PLANS GOOD; EXECUTION ONLY FAIR. RECOMMENDATIONS DECISION ON VALUE OF DATA BASE FOR INTERIM CRITERIA VS. COMPLETE STUDY. e IF DECISION IS IN FAVOR OF CONTINUATION - PROCEED WITH DATA COLLECTION e AS OUTLINED FROM THIS STUDY - DOCKET SEARCH SUPPLEMENTED BY SITE RECORDS FOR QUANTITATIVE OPERATOR RESPONSE DATA. USE CAREFULLY CONSTRUCTED, CAREFULLY ADMINISTERED OPERATOR SURVEYS e (POSSIBLY PAIRED COMPARISONS) FOR QUALITATIVE INFORMATION AND QUANTIFICATION OF STRESS AND DIAGNOSIS DIFFICULTY. ULTIMATE DATA BANK IS CORRELATED SIMULATOR STUDIES; THIS DATA COLLECTION e IS FOR PURPOSE OF CORRELATION.
INPUT FROM MEASURABLE SIMULATOR OPERATOR EXPERTS ACTIONS DETAILED DOCKET SEARCH 0 SITE RECORDS INPUT FROM SEARCH SITE PERSONNEL I ITERATE DATA COMPILATION AS NECESSARY STATISTICAL ANALYSIS DATA BASE FOR SIMULATOR CALIBRATION
4
SUMMARY
/ CONCLUSIONS INITIAL DATA 1.
GENERAL MODEL SEEN AS REASONABLE BY OPERATORS, BUT MANY (USUALLY VERBAL) RESPONSES INDICATE CONTRADICTIONS. 2. SURVEY RESPONSES SUPPORT IDEA THAT POTENTIAL SEVERITY OF CONSEQUENCES IS MAJOR FACTOR IN PERCEIVED STRESS. 3. SURVEY RESPONSES INDICATE DIFFICULTY OF DIAGNOSIS DEPENDS ON ANNUNCIATION. 4. OPERATOR PREDICTION OF RESPONSE TIMES IS OUALITATIVELY SIMILAR TO LIMITED PERFORMANCE DATA COMPILED TO DATE - DIFFUSE LOG-NORMAL DISTRIBUTION. 5. OPERATOR PREDICTION OF RESPONSE TIMES IS QUANTITATIVELY MORE OPTIMISTIC THAN PERFORMANCE DATA - BUT RANGE IS SIMILAR. 6. WE HAVE ONE EXAMPLE (INADVERTENT SACETY INJECTION) 0F AN EVENT FOR WHICH WE HAVE BEEN ABLE TO: (A) DEFINE A SPECIFIC ACTION TO BE MEASURED IN AN EVENT OF REASONABLY DIRECT APPLICABILITY TO N660; (B) COLLECT A SIGNIFICANT NUMBER OF DATA POINTS. THE RESULTS SUGGEST A LOG-NORMAL DISTRIBUTION WITH A RANGE THAT EXTENDS TO TIME VALUE COMPARABLE TO THE N660 CRITERIA (95% = 7.4 MINUTES).
$5 FINAL STATEMENT CONTINUATION OF THIS WORK REQUIRES PERSONS WITH EXPERIENCE IN: NUCLEAR POWER PLANT OPERATIONS PSYCHOLOGY (HUMAN FACTORS) STATISTICAL ANALYSIS OF DATA SIMULATOR EXPERIMENTS DURING THE STUDY WE HAVE CONTACTED A SINGLE GROUP THAT HAS EXPERIENCE IN ALL 0F THESE AREAS - MEMPHIS STATE UNIVERSITY, CENTER FOR NUCLEAR STUDIES. IT IS SUGGESTED THAT NRC CONSIDER THIS ORGANIZATION OR ANOTHER THAT IS SIMILARLY QUALIFIED TO PURSUE LATER PHASES OF THE STUDY. MEMPHIS STATE UNIVERSITY HAS EXPRESSED AN INTEREST IN THIS STUDY AREA AND AN INTEREST IN PERFORMING THE WORK UNDER SUBCONTRACT TO ORNL.}}