ML20236P745
| ML20236P745 | |
| Person / Time | |
|---|---|
| Site: | Monticello, Kewaunee, Point Beach, Prairie Island, Callaway, Cook, 05000000 |
| Issue date: | 11/13/1987 |
| From: | Lynch D Office of Nuclear Reactor Regulation |
| To: | Holahan G Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8711180144 | |
| Download: ML20236P745 (17) | |
Text
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a November'13,:1987- i 1
MEMORANDUM'FOR: Gary M..Holahan, Assistant' Director' ' DISTRIBUTION:
for Regions III and V'- Docket Tf1Mit i' Division of. Reactor' Projects - III,
-NRC & Local'PDRs IV, V and Speciall Projects PDIII-3 r/f 4
'l PDIII-3 PM's THRU: David L. Wigginton,~ Acting Director GHolahan Project Directorate. III-3 MDLynch ,)
Division'of Reactor Projects - III, PKreutzer i IV, V and Special. Projects KPerkins- ,
1 FROM: Dave Lynch, Senior Project' Engineer Project Directorate III.-3 ] j Division of Reactor Projects - III, i
-IV, V:and Special' Projects
SUBJECT:
DETERMINATION OF QUALITY PERFORMANCE USING PERFORMANCE INDICATORS
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- 1. Introduction and Conclusions L I have reviewed and' analyzed the latest set"of performance indicators =
for the six utilities in PDIII-3 with the intent of determining whether these performance indicators can be,used to ascertain quality performance.
To place the performance of these six utilities.in-perspective, I have also reviewed the performance indicators of a number of other utilities.
and compared them with those in PDIII-3..
The performance indicators I reviewed are: (1) scrams from power levels. <
greater tions; (3)thar 15 percent significant per (4) events; 1000' critical safety hours; system -(2) safety failures; 5 the (s stem actua-forced outage rate (FOR); (6) equipment cutages per 1000 critical hours;.and (7) total radiation exposures. The' definition of these terms is contained.
in the Performance Indicator Report and is presented in Enclosure.l. In using these performance indicators, due note has been taken of the.precau-tions to be considered in using this data. However, since this data is being used only to determine whether quality performance is being achieved -
by the utilities in PDIII-3 and not for any ranking, I believe.that the conclusions reached in this review are valid.
The primary conclusion of this study is that five of.the six plants in PDIII-3 are achieving very good quality performance (i.e., all but-Callaway), and that Callaway.is achieving good quality performance', j especially in light of its being rela'tively new. Additionally, upon j reviewing the performance indicators for those plants with the poorest) i history of radiation exposure, I conclude that there is a strong correla- !
tion between some' of the performance indicators (e.g., the forced outage ~ .j rate and equipment outages per 1000 critical hours) with radiation ;
exposure. This latter conclusion may be significant inasmuch as'it may _/ ,i i
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indicate that a plant with a poor history of radiation exposure may a'lso -)
be expected to experience a high incidence of. equipment failure. . Specifically, this correlation may mean that a plant with poorly maintained equipment .. ;
must subject its personnel to high exposures in order to repair equipment !
ofter each breakdown. It may also indicate that a plant management which ,
takes minimum precautions to limit radiation exposure ~also takes little care with plant equipment. INPO has suggested that the latter situation. .
is the more probable. l
- 2. Radiation Exposure Table 1 contains a summary of the radiation exposure by year for the six ,
plants in PDIII-3 from 1980 to 1986. It also contains the yearly exposure p of those five plants which have the highest average exposure over this 1 7-year period. For those plants in PDIII-3, this. exposure ranges from a' i low of 101 for Kewaunee (1982) to' a high of.2462 for. Monticello (1984).
The comparable range of radiation exposures for the five plants with the highest average exposures varies from a' low of 192 for Indian Point (1985) to a high of 4082 for Pilgrim (1984). The more significant charac- .
teristic for those plants in PD111-3 is that, with one exception (Monticello 1 in 1984). the radiation exposures do not vary significantly from year .to: -l year. In contrast, the five' plants with the highest radiation exposures 1 vary dramatically from year to. year. This large year to year variation for these five plants could be explained by a one-time operation (e.g...
replacement of steam generators). However Pilgrim, Oyster Creek, ,
Indian Point-2 and Robinson-2 have exhibited consistent dramatic yearly l variations in the amount of radiation exposure. ;
The radiation exposure data of Table 1 has been evaluated as multi-year l averages to place the matter in perspective.. These multi-year averages are presented in Table 2 as 7-year, 3-year and 2-year averages, including the latest yearly data, for the six plants in PDIII-3 and the five plants ";
with the worst multi-year averages. . The five plants in PDIII-3 with multi-year experience demonstrate consistent behavior in their multi-year averages as would be expected from Table 1. -l
- 3. Correlation of Performance Indicators with Radiation Exposures I extracted six of the perfomance indicators from the latest Performance ,
Indicator Report for the six. plants in PDIII-3 and for the'five plants with the highest average radiation exposure. This data is presented in-Table 3 and represents the four quarters ending in December 1986. This data was then plotted against the average radiation. exposure over the . i latest 2 years ending in December-1986. (Callaway was plotted using.the CY 1986 radiation exposure.) Figure 1 shows the relationship between ,
the equipment outage rate per 1000 critical hours-and the latest average ,
2-year radiation exposure. Except for Brunswick 1 & 2, there seems to- !
be a strong correlation between these two performance indicators for Pilgrim, Oyster Creek, Indian Point-2 and Robinson 2. . The five' plants in PDIII-3 with multi-year operating history plus Callaway, are at the extreme lower end of this correlation indicating excellent performance
4 g 3 ,
in both parameters. The average value for radiation exposure and the normal-ized equipment outage for the six plants in.PDIII-3 are plotted as an open triangle in Figure 1 and is designated as III-3. The average values for the five plants having the highest average radiation exposure are plotted as a darkened triangle designated as HiR fnr high radiation. A dotted line is drawn to indicate a trend. If a larger data base were used, a least squares t fit (i.e., a regression analysis) might be appropriate.
Figure 2 shows the forced outage rate plotted against radiation exposure for the same eleven plants. While there is considerably more scatter in this plot than in the previous one, there is still a discernable trend As before, the six plants in PD 111-3 demonstrate excellent performance. :
In Figure 3, safety system failures are plotted against radiation exposure 1 for the same eleven plants. Again, there is a discernable trend despite the inherent mechanism for scattering of the data. Specifically, failures are recorded as individual events but tabulated in the Performance Indicator Report on a quarterly basis. Therefore, the tabulated values for one failure per year and two failures per year would be 0.25 in the first instance and 0.50 in the second. Accordingly, the basic data is inherently quantized in increments of 0.25, thereby tending to scatter the data.
Figures 4 and 5 plot safety system actuations and significant events against radiation exposure. As before, while there is a scattering of the data, there is a discernable trend. As discussed above, safety system actuations and significant events are quantized in units of 0.25.
Finally, in Fioure 6 the perfomance indicator, scrams at power levels greater than I k per,1000 critical hours, against radiation exposure, indicates a general trend but with significantly more scatter than in the previous plots.
This is not unexpected since scrams can be caused either by inadequately main- .
tained equipment (and hence potentially higher levels of radiation exposure), l or by inadvertent personnel errors not associated directly with exposure of j operating perr.onnel to radiation.
I conclude from these various plots that there is a significant correlation between the level of radiation exposure at a nuclear power plant and the operating reliability and availability of equipment (e.g., the forced outage rate) as well cs the incidence of challenges to the safety related equipment (e.g., safety system actuations).
- 4. Evaluation of Quality Performance for Plants in PDIII-3 l All five of the plants in PDIII-3 with multi-year operating experience (e.g., Cook, Kewaunee Monticello, Point Beach 1 & 2 and Prairie Island 1 & 2) demonstrate excellent performance as measured by the seven performance indicators considered in this study. Callaway's performance i
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', is only.slightly belcw that of the other five plants which is' understand-able considering that it is~a relatively new plant. 'I3 conclude,Loverall. ..
that-the six plants in PDIII-3 are achieving. quality' performance as measured; by their performance. indicators.
b ,
-)
. Dave Lynch ..-.1
< U Senior: Project Engineer _-
cc: D. Crutchfield
, S. Varga.
I F. Miraglia T. Marley. ->
J. Cunningham ,
J. Roe ,
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]i I Office: LA/PDI: 1-3 SPE/PDIII-3 PD P III-3 Surname: PKregtzer MDLynch/r1 inton-Date: 11/ [f 87. 11/tdP87' DWigj/87 11/ !5
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. 3. ' CEFlh!T10h5 Of INDICATORS
.? s-: ; n;.w.u.L The definitions of the seven indicators currently in the program are: d.&%;iLL&pp'
'otal scrams are the.nunter of unp'lanned automatic scrans while' critical.
This indicator is the same.as the corresponding INP0 indicator. Examples of the types.of scrams-counted are. those-that result from unplanned transients, equipment failures, spurious signals, or human error. Also included in the total scrams are those that occur during the execution of procedures in which:
there is a high chance of a scram occurring, but the occurrence of the scram is not planned. Scram data are primarily derived from LER infomation and' supple-mented as necessary from 50.72 reports. The . reactor is critical if so stated in the reports. Otherwise it is determined from the review of the infonnation.
(28) Scrams Above 15 Percent Power per 1000 Critical Hours This subset of total scrams includes the automatic scrams occurring above 15 percent reactor power per 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of critical operation.
(IC) Scrams at or Below 15 Percent Power This subset of-total scrams includes the automatic scrams occurring while the.
reactor is at or below 15 percent power.
Safety System Actuations
[
Safety system actuations are actuations of the emergency core cooling system. -
(ECC5; (actual or spurious) and the emer response to low voltage on a safety bus)gency .
ac power This indicator systems is the same as(actual, the in^
cc'rresponding INPO indicator. Input for this indicator is derived from LERs and supplemented by 50.72 reports.
In determining what items should be included in the data for this indicator, the following conventions are used:
Only actuations of the high pressure injection systes, low pressure injection' system or safety injection tanks are counted for PWRs. - For SWRs, only actuations of the high pressure coolant injection system, the.
low pressure coolant injection system the high pressure core spray system, or the low pressure core spray, system are counted. 'No-actuations of the reactor core isolation cooling system are counted.
Actuations of emergency ac power system due to loss of power to a safe-guards bus are captured primarily based on indications of low voltage j signals- in the emergency power system.
j Actuations of. any of the equipment associated with the specific ECCS or emergency ac power system are considered necessary and sufficient to.
constitute a data count. For exampIe, if only a valve in a system is
( commanded to move to its emergency operational position, this is counted-n.___ _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ __ _ . _ _ _ _ _ _ _ _
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as an actuation. A pump does not have to be comanded l gency mode of operation, fluid does not need t04f(tif,"6r iD. to agogenera-toJts.g.y_;,h..p,4; tor does not need to be loaded for an occurrence to be counted.
(( Only one ECCS actuation is counted in any one occurrence, even if multiple l ECCSs actuate during the occurrence. For example, actuation of both the !
high pressure injection and the low pressure injection systems at a PWR- 1 during the same occurrence counts as only a single ECCS actuation for that !
occurrence. l Only one EDG actuation is counted in any one occurrence, even if multiple !
EDGs actuate during the occurrence. For example, actuation of all four ~
EDGs at a unit counts as only a single EDG actuation for that occurrence.
Occurrences involving actuations of both an EDG on a dead bus and an ECCS are given a count of two, one for the EDG actuation and one for the ECCS actuation.
At multi-unit sites that share equipment' (e.g., swing EDG or shared buses), actuations are counted and assigned to only one unit, even if.
multiple units are involved. This count is assigned to the unit where the '
actuation signal or loss of power originated. . If the signal source cannot be detennined to be associated with one unit,- the actuation is assigned to !
the unit with the lowest unit number unless the licensee has specifically I assigned the reported occurrence to a higher number unit.
3.3 Significant Events j
.( ~
ua tior, of cperating experience by the. NRC The staff.ificart events are those even screening process '
ircludes the caily review and discussion of all reported operating reactor events and operational cata such as special tests being conducted or construc- i tion activity. -
An event identified from the screening as a candidate significant event is forther evaluated to determine if there is an actual or potential threat to the health and safety of the public involved. Specific examples of the types of criteria are sunmarized as follows.
(2) Dec radation of important safety equipment. Events that will be considered unter this classification include situations where either there existed the potential for or there was an, actual reduction in the operational capability of equipment. For example, identification of a cosmon cause failure mechanism which could cause failure of redundant components or multiple independent component failures in response to a test or actual demand signal. This category would not ir.clude such items as a missed surveillance test where the equipment was subsequently tested and deter-mined to be operable.
(2) Unexpected plant response to a transient er a major transient. Events that will be considered under this classification include situations where changes in reactor parameters represent ur, anticipated reductions in margins of safety. For example, a rapid plant cooldown following a
( reactor trip exacerbated by a balance of plant malfunction or an
l
.< , -4 i undesirable systems interaction, such as a tsastoewessdievel4N #
bation and ECCS initiaticn following a standby-liquid control system actuation. This category would not include minor differences in predicted 1
, and observed conditions that can be reasonably explained by instrument i
( errors or modeline techniques and simplifying assumptions;
)
\
Degradation of fuel integrity, primary coolant pressLre boundary, or q 1mportant associated stri.ctures. Events considered under this category '
would include those of smilar character to.those identified in Item 1
~
related to the fuel, RCS, containment, or important plant structures.:
)
(O Scram with complication. A " scram with complication" is an RPS actuation, l when critical, followel by an equipment failure or malfunction or operator j e rror. The failure, malfunction, or error is generally not to include. '
those that cause the transient which leads to the'RPS actuation, or those that directly cause the scram. Failures that both cause.the scram and reduce the capability of the mitigating system (e.g., electric power, instrument air, or other auxiliary support functions) will be' counted.
Examples of equipment failure / malfunctions include:
- a. Mitigating system failures - loss of redundancy due to single fail- -
ure, reduced capacity or margin. This includes components or trains out of service for maintenance.
- b. Failure adding to complexity of event - erroneous control' system responses, electrical switching difficulties, mitigating. system and key plant parameter instrumentation malfunctions / failures.
- c. Additieral event ir,itiators - stuck open pritrary or secondary relief /
safety valves, pipe breaks.
Exartples of operator errors include:
(
- a. improper contrcl or tennination of mitigating system. )
i
- b. Misdiagnosis of the event or failure to follow procedures. 1 i
In addition to the situations described in items 1-4 other broad categories considered for significant events are as follows: !
(5) Unplanned release of radioactivity. Events considered under this category include an unplanned release of radioactivity that had the potential for ,
or actually exceeded the limits of the Technical Specifications or the Regulations. 1 ~
i (6) Operation outside the limits of the Technical specifications. Events that will be considered under this classification include situations where plant operation was conducted inconsistent with the license requirements.
This category would apply to risk significant deviations and most likely not include an incident involving a missed surveillance, ses11 errors in setpoints, or other administrative 1y inoperable conditions.
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(7' Other. .For example, a series or events. or recurrin considered collectively represent ineffective ~cWrW,c ,ircidents that, wlye.D...y'j*.,. . ,.. * .-
Mre1rdt%W69Y"
'ideficiency in the plant hardware or administrative' programs.
R ;(
- 4 Safety SygerJfilures
/; ;, <
Safety system failurps are any events or conditions that, by themselves, could crevent fulfillunt of ffquipty fr.$nction for. structures or systems. Where a system consist'af trultiple redundant subsystem'or trains, failure of all trains constitutes,a safety sydem failure. Failure of ene of two or more trains is not counted as a safet:y system fiil.ure . The definition for the indi-cator parallels NRC rr.rteting requirsents in 10 CFR 50.72 and 10 CFR 50.73.
Tre following list gives the major systems and subsystems which are monitored fcr inclusion in this indicator:
l Reactor Trip System and Instrumentation Engineered Safet) Features ' Instrumental'lon I
Recirculatico Pump Trip Actuation Instrumentation (BWR)
Mccident Monitoring instrumentation
,; , Radiation Monitoring Instrumentation. "
Reactor Co01 ant System <
i
'sSafety Yalves q i
' Eurgency Core Cooling Systems H
Auxfifery (and Ed,gency) Feedwater System (PWR)
ReactM Core Isolation Cooling System (BWR)
Isolttion Condenser (BWR) l Stan@y Liquid Centrol System (BWR) y Main Steam Line4 1sslation Valves !
' ,di Component Cooling Water System 6 Esser.tial or Erergency Service Water i U1titrMe heet Sink ContqY :Fwimergency Ventilation System j Onsite-Degency' AC and DC Power and Associated Distribution '
Contmiment and Cc'ntainment Isolation Containment Coolant Systems Residual Heal Removal Systems Combustible Gas Control i
Fire Detection and Suppression Systems Low Temperature Overpressure Protection (PWR)
Spent Fuel Systems Essential Compressed Air Systems 3.5 Forced Outage Rate Forced outages are those reauired by the end of the weekend following the discovery of an off-nonnal cc*dition. The forced outage rate is the number of forced outage hours divided by :.N sum Of unit service hours (i.e.. generator online hours) and forced outage hour:. Yhis indicator is the same as that of INPO and the NRC monthly operating report. The data are generally obtained from the monthly operating reports. In some cases when the reports are not available, the data are obtained directly from the licensees.
i l ,
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, 3. 6' Equipment Forced Outages per 1000 Critical houw:wMMA~25NMAN' This is the nurr.ber of forced outages caused by equipment failures per 1000 !
'tical hours. It is the inverse of the mean time between forced outages.
i{. sed by equipment failures. The inverse number was adopted to facilitate 41culation and display. The source of this data is the same as that'for the fcrced outage rate.
3.7 Cc11ective Radiation Exposure This is the total radiation dose at the site for a given period. The site total is divided by the number of units at the site contributing to the radia- )
tion exposure to obtain unit values. This indicator is the same as that of .
INPO. .
l 3.8 Additional Notes Although certain NRC perforinance indicators are the same as those of INPO overall performance indicators, the criteria for including the data in the calculations for industry average are not the same in all cases. For example, INPO does not include scram values for a plant with cumulative capacity factor of less than 25 percent during the time period being considered in calculating the industry average. The NRC does not exclude such plants. Therefore, the industry average values of the comon indicatort are likely to be different.
There are three additional categories of information that are included in the program for collection. monitoring, and future development, but are not inclu- l
, d in this report. They are systems involved in scrams, safety systems i ations, significant events, and safety systems failures; causes associated I s
.a scrarrs, safety system actuations, significant events, and safety system i failures such as personnel error, maintenance problems, ecuipment failures and I desicn/ fabrication / construction error; and the number of forced outages. l In addition, the staff is working on additional or improved indicators. The !
l highest priority areas for development are maintenance and training. Other J 1mportant areas include limiting conditions for operation (LCO) action state-i ments, number of items out of service, causes of events, and operator licensing '
examination results as well as risk-based and programmatic indicators.
- 4. PRECAUTIONS r
The data for this report were obtained from reliable NRC sources as discussed-earlier and were reviewed by NRC personnel in headquarters and the regions for completeness and accuracy. The data for the second quarter of 1987 will be reviewed again in preparation for the next quarterly report in order to ensure that late information, if any, is accounted for.
For scrams above 15 percent power per 1000 critical hours, the results for plants with less than 1000 critical hours in a quarter can be distorted. For this report the degree of distortion has been reduced by using at least a cinimum value of 200 critical hours in the calculations for any given quarter.
Th2 results for equipment forced outages per 1000 critical hours can be distor-tv in a similar manner. This distortion has also been reduced by using at 14 200 critical hours in the calculations.
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