ML20247K101

Summary of 890517 Meeting W/Inpo in Bethesda,Md Re Similarities & Differences Between INPO & NRC Performance Indicators

ML20247K101
Person / Time
Site:	Beaver Valley, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Palisades, Perry, Indian Point, Oyster Creek, Hope Creek, Grand Gulf, Cooper, Sequoyah, Arkansas Nuclear, North Anna, Crystal River, Diablo Canyon, Duane Arnold, Farley, Quad Cities, Rancho Seco, Fort Calhoun, McGuire, 05000000
Issue date:	05/16/1989
From:	William Burton NRC
To:	Novak T, Mike Williams, Wolf T NRC
Shared Package
ML20247K094	List: ML20247K090 ML20247K101
References
NUDOCS 8906010210
Download: ML20247K101 (26)

Text

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• k. UNITED STATES )

g 3 p, NUCLEAR REGULATORY COMMISSION

.5 WASHINGTON, D. C. 20555 l ,

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May 16, 1989

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Note to: Distribution :

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Through: S. M. Stern

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From: W. F. Burton /

Subject:

MEETING WITH'INPO ON PERFORMANCE INDICATORS Place: Bethesda Maryland; Maryland Nat. Bank Building; Third Floor.

Time: 1:00 PM; May 17, 1989 Topics: - Differences and Similarities between INPO and NRC Performance Indicators.

- INPO Safety System Performance Indicator.

Backaround on Differences between INPO and NRC Indicators. i INPO and NRC both track four indicators in common: Automatic Scrams while Critical (Scrams), Safety System Actuations (SSA),

Forced Outage Rates (FOR) and Collective Radiation Exposure

-(CRE). NRC will use INPO's CRE data beginning with the Performance Indicator (PI) Report for the First Quarter of 1989.

Part of the differences (amount unknown) between the industry 4 averages developed by INPO and NRC for Scrams can be explained by the fact that INPO does not count plants that have less than 25% 4 capacity factor during a year and those events for new plants and INPO does not consider new plants until Jan 1 of the second full year following full power licensing. Similarly, INPO does not reflect new plant SSAs, FORs and CREs in their industry averages.

NRC excludes plants in long term shut down from its industry averages. Attachment 1 is a discussion of the differences between the INPO and NRC PIs.

A major difference between the INPO and NRC indicators is in the treatment of SSAs. While both INPO and NRC base SSAs on similar systems, INPO has a higher threshold than NRC for classifying

. events as SSAs, INEL has compared the Safety System Actuation counts for the Third and Fourth Quarters of 1988 as delineated by INPO and NRC.

Attachment 2 is a summary comparison of the those plants where differences between the SSA counts were identified. Attachment 3 compliments Attachment 2 and details those SSAs identified by NRC but not INPO and a brief description of the reason for Our classifying that event as an SSA.

Attachment 4 is a copy of the 1988 INPO PIs published in March 1989. Attachment 5 are the detailed definitions of the INPO indicators, as issued by INPO in February 1988.

8906010210 090523 NEXD PDR ORQ p w_-____-______-_-_-____-____________. _. __ _

.s Distribution:~

M. Williams S. Pettijohn T. Novak R. Dennig T. Wolf J. Crooks V. Benaroya l

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ATTACHMENT 1 DIFFERENCES BETWEEN THE INPO AND NRC INDICATORS ,

The NRC is attempting to better understand the performance indicators used by INPO and how they may differ from NRC PIs. The INPO and NRC PIs that are common are Unplanned Automatic Scrams While Critical, Safety System Actuations, Forced Outage Rate, and Collective Radiation Exposure. The items below are highlights of the differences, the actual definitions of the INPO indicators, l

which contain many clarifying notes, are in Attachment 5.

UNPIANNED AUTOMATIC SCRAMS WHILE CRITICAL:

The INPO indicator is defined as the number of unplanned automatic scrams (RPS logic actuations) that occur while the reactor is critical. This is similar to the NRC PI definition for Automatic Scrams While Critical, with the following differences:

- INPO collects scram data on all plants beginning with January 1 of the first calendar year following full power licensing. The NRC scram counts for the PI report includes all critical scrams.

- INPO does not count manual turbine trips which lead to reactor scrams which were effected to protect important equipment or to minimize the effects of transients . The NRC PIs do count such events.

- INPO considers short-term transient conditions in its determination of whether a unit was critical or not. The NRC determines the actual plant condition at the time of the event.

- INPO industry averages exclude data prior to January 1 of the second full year following commercial operation, and those years where the capacity factor is less than 25 percent, or data element were not provided for the full period. NRC industry averages exclude plants in long term shut down.

UNPLANNED SAFETY SYSTEM ACTUATIONS This INPO indicator is defined as the sum of a.) the number of unplcaned ECCS actuations that result from reaching an ECCS actuation setpoint or from a spurious / inadvertent ECCS signal and b.) the number of unplanned emergency AC power system actuations that result from a loss of power to a safeguards bus. This is the same as the NRC PI definition of Safety System Actuations.

Although the INPO and NRC definitions essentially are the same, the body of data to which the definition is applied is different:

- Although it appears that the same ECCS systems are  !

included in the definitions, what is considered a valid actuation seems to be different. INPO SSA definition requires the actuation of a " major" system, whereas the NRC interprets this actuation as either a valid or spurious signal (whether the equipment starts or not). For both INPO and NRC, an undervoltage signal on a safeguards bus is counted as a diesel start.

- INPO industry averages exclude plant data prior to January 1 of the second full year following commercial operation. NRC industry averages exclude plants in long term shut down.

- Since INPO has a higher threshold than NRC for classifying SSAs, the NRC indicator would have a higher value than the INPO indicator.

FORCED OUTAGE RATE:

The formula for computing the FOR used by INPO and NRC are the same.

In computing industry averages, INPO uses data for units beginning January 1 of the second full calendar year following full power licensing, and has a requirement that data elements be provided for at least 50% of the time period to be included in the industry average. NRC excludes plants in long term shut down from its industry averages.

COLLECTIVE RADIATION EXPOSURE:

NRC uses the INPO supplied data.

'4 9- ,.

O, f ~ ATTACHMENT 2 PLANTS FOR WHICH INPO AND NRC HAD DIFFERENT COUNTS FOR SSAs FOR THE THIRD AND. FOURTH QUARTERS OF 1988 PLANT YR-QTR INPO NRC ARKANSAS ~1 88-4 1 0 ARKANSAS 2 88-3 0 1 BROWNS FERRY-1 88-4 1 3 BROWNS FERRY 2 88-4 0 2 COOPER 88-3 1 2 CRYSTAL RIVER 3 88-4 0 2 DAVIS BESSE 88-3 1 0 DAVIS BESSE 88-4 1 0 DIABLO CANYON 1 88-3 0 1 DIABLO CANYON 2 88 0 2 DUANE ARNOLD 88-4 1 2 FORT CALHOUN 88-4 0 ,1 GRAND GULF 88-4 0 1 HOPE CREEK 88-3 0 2 INDIAN POINT 3 88-4 0 1 MCGUIRE 1 88-4 0 1 NORTH ANNA 1 3 0 1 NORTH ANNA 2 88-3 0 1 OCONEE 3 88-3 1 0 OYSTER CREEK 88-4 0 1 PEACH BOTTOM 3 88-3 1 2 PERRY 88-4 0 1 QUAD CITIES 2 88-4 0 1 RANCHO SECO 88-4 0 1 SALEM 2 88-3 1 0 SEQUOYAH 1 88-3 2 0 SEQUOYAH-2 88-3 0 1

_ _ _ _ _ _ - _ _ _ _ - - _ . _ _ J

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- .y ' + t ATTACHMENT.3 e

AUTO SCRAMS WilILE CRITICAL

• GREE ON ALL SCRAMS Safety System Actuations (ESFs)

Diesel Start Sign.nl ~

Any valid low volt signal on a safety bus which should cause the diesel generator to start.

ECCS Any $1 signal which causes the following system to actuate.

HPCS, HPCI, LPCS, LPCI, LPSI, HPSI (not accumulator injection), no Si if no major equipment operated (i.e. valves did not move or pumps did not start).

Arkansas 1 88 4 Can find no event in the fourth quarter where an ESF occurred.

Arkansas 2 88-3, 368880011, 08/01/88

@M HPSI manually started to control PZR level af ter the stram.

Beaver Valley 1 88-4, RE-14095

- Beaver Valley 1 cancelled the event. We now agree with INP0.

4

-- -_________m__ ____.___-.m__.. . _ _ __ _ _ . - _ _ . _ _ - .__m__ _ _ _ _ _ . _ .

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Browns Ferry-1 88-4, 259880044, 11/01/88, 1311 4 Operator did not hold switch long enough to ensure (M ' breaker shut. causing a low voit signal on 4160 shutdown board 'B'.

88-4,'259880045,11/01/88,1654 INPO counted Breaker failed to close while shifting power one event.- . supplies causing undervoltage for 30 seconds and a diesel start. ,

88-4, 259880045, 11/01/89, 1700 Breaker failed to close while shifting power supplies causing undervoltage for 30 seconds and a diesel start.

Browns Ferry 2 88-4,260880017,12/18/88,0435

'2C' core spray pump started. Discharge valve

' 5',p tagged out.

l-88-4, 260880016, 12/09/88, 1344 L

'2D' RHR pump started in normal standby. LPCI lineup manual injection valve shut.

88-4. 259880049, 12/17/88, 1721 RHR pump start not in LPCI mode as stated in LER.

Cooper 88-3, 298880021, 08/25/88, 0040 I

cg Auto start HPCI and RCIC (no PI) after scram.

)

88-3, 298880026, 09/30/88, 1257 Undervoltage on emergency transformer for about 2 seconds caused diesel start.

W y, , ,

Crystal River 3 88-4,302880021,10/14/88,0949 ESFAS actuation' injected 1000 gallons borated water by LPCI.

-y F

!1 88-4, 3028800024, RE-13838, 10/28/88, 0327 S1 initiated for a tiine to maintain pressurizer level in RE.

Davis Besse- 88-3 Can find no LER for this time frame that was an ESF.

88-4 Can find no LER for this time frame that was an ESF.

Diablo Canyon 1 88-3, 323880008, 07/17/88, 0746 f Loss of startup power to Unit I and 2 caused Unit 1 f/ diesel to start.

Diablo Canyon 2 88-3, 323880008, 07/17/88, 0746 Loss of startup power to Unit 1 and 2 caused Unit 2 (2 counts) diesel to start and RPC1 on high steamline gc differential pressure.

Duane Arnold 88-4,331880016,10/17/88,2334 Low volts on bus 1A3. 'A' diesel tagged out.

88-4, 3318800'14, 10/26/88, 0353 go" N Moisture in reactor level switches caused low level l signal and HPCS start.

88-4, RE-14295 Plant cancelled event.

., . s, 1'

1 Farley 1 88-4, 348880024 LER states only RHR pump started, no high head SI pump as stated in RE.

i Fort Calhoun- 88-4, 285880024, 10/03/88, 1243 Momentary low volts on IA4, 4160V bus caused diesel shut.

fo, 88-4,285880038,12/31/88,2156 No SIAS components actuated. -

Hope Creek 88-3, 354880019, 07/28/88, 1029

. 'C' core spray. pump started due to a human factor (o' testing design deficiency.

88-3, 354880022, 08/26/88, 1825 HPCI and RCIC (no PI) started on low reactor level after scram.

Indian Point 3 88-4,286880006,10/09/88,1852 Breaker 52/SA opened. Diesel started and loaded 40 bus.

McGuire 1 88-4,369880038,11/29/88,1050 g6 Switch placed incorrectly caused loss of power to Y 1/2 of unit. Diesel started and loaded.

North Anna 1 88-3,338880020,08/06/88,2257 90% undervoltage on bus caused diesel to start and Go' D @ load bus.

North Anna 2 88-3, 339880002, 07/26/88, 1130 go, W1 High head SI pump started during testing.

Oconee 3 88-3 Can find no ESF events.

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Oyster Creek 88-4, 219880022, 10/02/88, 1357-t'l Fault on 'B side electrical distribution. DG did not start due to a problem in cable to DG.

Palisades 88-4, RE-13956 Plant cancelled the event due to the event not being an ESF.

Peach Bottom 3 88-3,277880020,07/29/88,1858  :

JdB Lost off-site power due to a transformer shorting.

Diesel started and loaded bus.

88-3,278880009,08/31/88,2145 l Startup feeder breaker opened and other leg of i off-site power unavailable. Diesel started and  !

loaded bus.

Perry 88-4, 440880043, 10/30/88, 2259 h,4q0 'HPCS start signal when during HPCS breaker. ]

maintenance HPCS room cooler started.

Quad Cities 2 88-4,265880027,11/14/88,1650 l j

g Technician shorted leads causing HPCI start.

b Operator secured pump before injection could occur.  ;

Rancho Seco 88-4, 312880018, 12/09/88, 1826 '

p'9% 'B' HPI pump manually started to ;intain RCS l inventory.  ;

j Salem 2 88-3 i No ESF's - two scrams, but no indication of any ESF starts, j

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.Sequoyah'l 88-3,.(2)L327880029, 08/04/88 and 08/05/88 Undervoltage signal in starting circuit caused load y1 4 shedding and diesel starts. No actual' low volts on.

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foT bus.

Sequoyah 2 '88-3,'328880034. 08/15/88, 1651

. 60 ' g Lost 'lA' start' bus which powers 'ISB' shutdown-board. All four diesel started.

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T .1 1988 -

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PERFORMANCE INDICATORS .

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FOR THE U.S.

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. - NdCLEAR UTILITY INDUSTRY +

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INSTm)TE OF NUCIEAR POWER OPERATIONS - . .

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-[E. R F @ R M A N C E I N D I C A T O R S 7 9P uclear plants with few The performance indicator pro- This foldout section provides per-unplanned scrams, few gram, now six vears old. was reimed in formance data from 1980 to 1968 in Ww -icnificant events, low 1985 when three special review groups selected areas. Graphs are included far personnel radiation exposure and high joined the Institute in developing a set seven oithe 10 overall pericirmance indi-equivalent availabihtv are cene ally ofoverall performance mdicators cators. InJustrvwide coals thr 1990 are recogm:ed as well managed overall. designed to promote long-term industrv included for these seven indicators. The Such plants are more rehable and can improvement. Senior nuclear utility 1990 goals were determined by averacine be expected to have Ngher marems managers, a setuor nuclear executive the individual unit coal 3 furnished to ofsaferv. from each of the U.S. nuclear steun INPO by each utihtv.

In recocninon et this and m keep. system suppliers and a group oioutside Data collection for the remainme ing with its coal to promote excellence experts contnbuted to this effort. three indicators becan in 1987 and and the hichest margin visaiety, the The groups agreed on 10 overall 1988. Sufficient data is act vet available Institute collects mdustrv data on key indicators as an important manacement to show meaninciul trends tbr these performance mdicators and shares this tool for goal settmg and for moniconng indicators. These indicators are safety data with its members and participants. plant performance. system performance, thermal renor-Overall plant performance graphs, By Arnl 1966, each U.S. utihty mance and fuel rehabihtv.

such as those provided in this report, with an operating unit had set challeng-summart:e the industry's performance ing short term and long-term 1990 goals through the en d eithe vear. !mproving for most of the overallindicators.

trends are evident in all areas. Many oilNPO's international part -

cipants use the performance indicators.

Several have also estabhshed long-term performance goals.

l l

, . INPO and the industry are continuing to review the perf:rmance indicator t * *

, program. Three relatively new indicators-safety syst:m performance.

thermal performance and fuel reliability-are now being tracked each year.

l Safety system performance Thermalperformance Fuelreliability Safety system penormance is Thermalperformance is deimed Fuel rehability is measured bv the defmed separatelyic each oithree as the ratio of corrected design gross amount oinssion products re' tased into boiling water reactorj.,d pressun:ed heat rate to the ad;usted actual gross reactor coolant. More reliable fuel water reactor saferv syt;ms.The heat rate. The desten gross heat rate is releases fewer 6ssion products. Hich fuel indicator is based on ine hours that corrected to retlect plant modifications reliability reduces radiological impact components in the safety system and operating deviations from the ini- on plant operations and mamtenance are unavailable to perform their tial thermal design. The actual gross activities.

intended functions. A low value heat rate is adjusted for circularme water Fuel rehabihty is measured differ-indicates a greater marem cisaiety temperature and the effect of feedwater entiv for boihng water reactors and in preventing reactor core damage, pump efficiency. pressun:ed water reactors due to desicn and suegests a reduced chance of Thermal performance rerlects differences. Data collecnon for the fuel extended plant shutdown due to emphasis on thermal efficiency and rehabihty mdicator began in 1987.

safety system iatlure dunng an opera- maintenance cibalance-of-plant tional event. Data collection for this systems, as does gross heat rate. How-indicator began in 1958. ever, this mdicator provides a more meaningful basis for unit-to-unit '

performhmce compansons than gross heat rate. Data collection for the thermal performance indicator began m 1987.

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Institute of Nuclear Power Operations

- Suite 1500 1100 Cacle 75 Parkway Atlanta. Georgia 30339-3064 Telennone 404 953-3600

3899M

+. *  !

W Equhalettt availability factor 80 76 2 Equivalent availabihty factor is the 63.5 643 ratio oithe total power a unit could have produced.considerme equipment and 60 59.8 60.9 60.5 61.8 63.7 60.3 61.8 regulatory limits, to its rated capacitv, expressed as a percentage. A high y, 40 equivalent availability factor indicates j effective plant procrams and practices to maximi:e electncal generation. 20 The industrywide average of 64.9 percent is an improvement over 1987, 1 but is impacted by eight units that were shut down for most oil 988. Excluding u ,, ,

I these plants, industrywide equivalent ,

I availability nses to 70.7 percent.

Unplanned automatic scrams i I

The graph shows the average num- 8 I4 ber of unplanned automatic scrams while the reactor is entical that 6.2 6.1 occurred at nuclear plants operatmg 6 d8 with a 1988 capacitv iactor oi25 g 4.5 4'3 percent or greater. A low number is E 3.9 desirable, since scrams result from equip- g 2.7 ment failures or human enor.1980 83 "

5 2.1 scram numbers were estimated from the 2 1.5 number of automatic scrams while the units were synchroni:ed to the power grid. For years after 1983, the data was n g

expanded to include unplanned 1980 1981 1982 1983 1984 1985 1986 1987 1988 1990 0"'

automatic scrams that cccurred anytime the reactor was cntical.

The 1988 industry average repre-sents more than a threefold improve-ment over the 1950 data.

1.4 1'3 Unplanned safety system i,2 1.2 actuations Unplanned safety system actuations I compnse unplanned emergency core LO u cooline system actuations and emer- E O.8 0.8 0.8 gency hC power svstem actuanons due 5 to loss of power to a safeguards bus. $ 0.6 Fewer actuations indicate greater care in plant operation. which contnbutes to a

! g.4 0.2 higher maretn cisafety.

The industry as a whole reached the gg 1990 goalin this area m 1988. 1985 1986 1987 1988 1990 Gul

e, o Gross heat rate t0.5s0 '"" sued us2 t s.3 i i 0-Low gross heat rate, or Btu per kilo. 3 lull ID.315 10'278 ' " 31 '""

watt-hour. retlects emphasis on thermal efficiency and at:ention to detail in the 3 10 000 mamtenance cibalance-of plant sys- g tems. Efficient, well-tuned plants enable 2-operators to detect abnormal trends and E correct them eativ. The minimum heat j 9.500 rates attainable are a function ciplant g design. is The industry as a whole surpassed the 1990 goalin this area in 1988. 9.000 1980 1981 1982 1983 1984 1985 1986 1987 1988 1990 Goal Collective radiation exposure per unit 1.400 This indicator examines the averace 1.200 collective radiatton exposure in man- 1.061 f.Dif I 137 1.0c3 rem per unit for both boiling water reac- g 1.000 tors and pressun:ed water reacters since 800 1980. Low exposure indicates good

{ 800

! 622 management controls and attention and  ;; BOO 521 G 8" that radiological protection procrams 4sg 400 are effective.

The 1958 industry averaces for boil' 200 ing water and pressuri:ed water reactors represent improvements oi38 percent 0 1980 1981 1982 1983 1984 1985 1986 1987 1988 1990 and 42 percent, respectively, over 1980 scii averages. BWR PLANTS 800 707 600 591 l

597I633 1980 1981 1982 1983 1984 1985 1986 1987 1988 PWR PLANTS 199

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2.00 14st-time accident rate 1 ost time accident rate is the num- I ber of worker infunes involving davs I away from work for eve.tv 200.000 man- j I.50 I 38 1.26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> (100 man-sears) worked. This indi-cator rerlects the industry's progress in f i gg improvmg industnal safety for workers E 0.71 0.66 0.64 over the past several years. E 0.58 E O.50 0.40

" 0.0 1980 1981 1982 1983 1984 1985 1986 1987 1988 1990 Goal low-level, solid radioactive waste 1.200 3.ii3 per unit The average volume of radioactive waste per unit for both bolling water and 888 852 863 pressun:ed water reactors since 1980 is 1 800 shown on these charts. Minimi:ing the E productton oiradioactive wastes reduces 483 466 459 storage, bunal and transportation needs and thereby reduces the environmental impact of nuclear power. Management j 400 312 attentton and good controlover many plant activities are required to achieve 0

this. 1980 1981 1982 1983 1984 1985 1986 1987 1985 1990 in 1988 the industry as a whole Gasi surpassed the 1990 goals in this category 8wnPLANTs for both types of units-a 72 percent reduction since 1980.

600 586 575 459

= 407 5 400 h 334 ll48I 5

E 211  !!3 II4 ng

$200 1980 1981 1982 1983 1984 1985 1986 1987 1988 1990 Goni PWR PLANTS

r ,. . ATTACHMENT 5 2/88-

,. .o ,

UNPLANNED AUT0NATIC SCRANS WHILE CRITICAL runPost The purpose of the unplanned automatic scrams while critical indicator is to monitor industry performance in reducing the number of unplanned automatic reactor shutdowns. It provides an indication of how well a plant is operated I. and maintained because scrams result from equipment failures or personnel errors that cause undesirable and unplanned thermal-hydraulic or reactivity transients. Manual scrans and automatic scrams as a result of manual turbine trips to protect equipment or mitigate consequences of transient are not 1.

counted because operator-initiated scrass and actions tb protect equipment l should not be discouraged.

DEFINITION The indicator is defined as the number of unplanned automatic scrams (reactor protection system logic actuations) that occur while the reactor is critical.

The indicator is further defined as follows: '

o Unplanned means that the scram was not part of a planned test or evolution. . 4 o

y<y/ #9 Scram means the automatic shutdown of the reactor by a rapid insertion P of all control rods that is caused by actuation of the reactor protection system. The scram signal may have resulted from exceeding a setpoint or may have been spurious.

o Automatic means that the initial signal that caused actuation of the reactor protection system logic was provided from one of the sensors monitoring plant parameters and conditions, rather than the manual scram switches or, in certain cases described below, manual turbine trip switches (or pushbuttoms) in the main control room.

o Critical means that during the steady-state condition of the reactor prior to one.to the scram, the effective multiplication factor (k,ff) was equal DATA ELEMENTS The basic data element required to determine each unit value for this indicator is the number of unplanned automatic scrams while critical. In the U.S., the number of scrams is determined by INPO from an analysis of licensee event reports that are submitted by the utilities to the U.S. Nuclear Regulatory Comission, In addition to the number of scrams, sufficient data must be available for calculating the cumulative capacity factor for the period (see Calculations and Data Qualification Requirements below). The data elements necessary for calculating cumulative capacity factor are as follows:

o gross electrical generation (defined under the thermal performance indicator) 2/88 o gross maximum capacity (defined under the equivalent availability factor indicator) o period hours (defined under the equivalent availability factor (indicator)

CALCULATIONS The unit and industry values for this indicator are determined as follows:

o value for a unit = sum of unplanned automatic scrans while critical o value for the industry - average (mean) of the unit values The cumulative capacity factor (CCF) for the period is calculated as follows:

o CCF = Ltotal oross electrical oeneration) x 100%

(period hours) x (gross maximum capacity)

DATA QUALIFICATION REQUIREMENTS Data col'lection begins January I of the first full calendar year following full power licensing for U.S. units (following commercial operation for international units). Data for U.S. units are included in the calculation of industry values beginning January 1 of the second full calendar year following full power licensing. However, in order to be included in the industry values, data elements must be provided for the entire time period, and the unit must have a cumulative capacity factor of at least 25 percent for the time period. Requiring this capacity factor minimizes the effects of plants that are shut down for long periods of time and whose limited data may not be statistically valid.

CLARIFYING NOTES o

Scrams that are planned to occur as part of a test (e.g., a reactor protection system actuation test), or scrams that are part of a normal operation or evolution and are covered by controlled procedures, are not included.

o Reactor protection . system actuations that occur while all control rods are inserted are not counted because no control rod movement occurred as a result of the actuation.

o Each scram caused by intentional manual tripping of the turbine will be analyzed. Those scrams which clearly involve a conscious decision by the operator to manually trip the turbine to protect important equipment or to minimize the effects of a transient will be counted as manual scrams.

, 2/88

- \

o During a startup, shutdown, or changing power condition, the reactivity transients may cause the reactor to go subcritical or super-critical for a short period of time. However, the plant is considered critical for i purposes of this indicator if the reactor was critical prior to the 3 j

reactivity transient and may be assumed to return to a critical i condition after the transient is completed (e.g., a plant is considered to remain critical after initial criticality is declared on a reactor startup, and to be critical until taken permanently subcritical on a ,

i reactorshutdown).

i

12/87 UNPLANNED SAFETY SYSTEN ACTUATIONS

~

PURPOSE The purpose of the unplanned safety. system actuation indicator is to monitor progress in reducing the number of occurrences of significant off-normal plant conditions. Emergency core cooling system (ECCS) actuations indicate events that are severe from a themal-hydraulic perspective, while emergency AC power system system.

actuations indicate a significant degradation of a vital support Limiting the number of unplanned safety system actuations indicates that a larger margin of nuclear safety is being maintained.

DEFINITION This indicator is defined as the sum of the following safety system actuations:

o the number of unplanned ECCS actuations that result from reaching an ECCS actuation setpoint or from a spurious / inadvertent ECCS signal o

the number of unplanned emergency AC power system actuations that result from a loss of power to a safeguards bus An unplanned safety system actuation occurs when an actuation setpoint for a safety system is reached or when a spurfous or inadvertent signal is generated (ECCS only), and major equipment in the system is actuated. Unplanned means that the system actuation was not part of a planned test or evolution.

For PWRs, the ECCS actuations to be counted are actuations of the high  !

pressure injection system, the low pressure injection system, or the safety injection tanks (accumulators, core flood tanks). For BWRs, the ECCS actuations to be counted are 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. Safety systems that may be used during normal plant operations (e.g., startup, shutdown) have not been included. ,,.

DATA ELEENTS The following data are required to determine each unit's value for this indicator:

o the number of unplanned ECCS actuations that result from reaching an ECCS actuation setpoint or from a spurious / inadvertent ECCS signal-o the number of unplanned actuations of emergency AC power systems that result from a loss of power to a safeguards bus

v, .

.. o

• 12/87 .

CALCULATIONS-The unit and industry values for this indicator are determined as follows: -

o value for a unit = (number of ECCS actuations) + (number of emergency AC power system actuations) o value for the industry = average (mean) of the unit values DATA QUALIFICATION REQUIREMENTS Data collection begins January 1 of the first full calendar year following full power licensing for U.S. units (following commercial operation for international units). Data.for U.S. units are included in the calculation of industry full power values beginning January 1 of the second full calendar year following licensing. In addition, data elements must be provided for the antire time period in order to be included in the industry values.

CLARIFYING NOTES o

Reactor protection system actuations are not included, because unplanned automatic reactor scrams are a separate performance indicator.

Actuations of other safety-related systems such as auxiliary feedwater, reactor core isolation cooling, or residual heat removal, are not included since they are often actuated during normal operations (e.g.,

startup, shutdown) that do not represent significant off-normal plant conditions.

o Only one ECCS actuation is counted for each event that actuates one or more ECCS systeus. For example, actuation of both the high pressure injection and the low pressure infection systems during the same event would events,count as one ECCS not individual systemactuation.

actuations. The intent is to count actuation o

For actuations to be counted, major equipment (e.g., pumps, diesels) must be actuated. For example, the spurious opening of one motor-operated valve in the high pressure injection system would not count as an ECCS actuation.

o Emergency AC power system actuations due to spurious or inadvertent-starts of the emergency AC power source are not counted because these actuations represent no degradation in plant safety, and the inadvertent start does not cause a plant transient. ECCS system actuations due to spurious or inadvertent starts are counted because these actuations can result in a plant transient or equipment damage.

o When power is lost to one or more safeguards buses at a unit, only one emergency AC power system actuation is counted.

o Safety system actuations (as defined above) are counted during all plant conditions (e.g. operating, shutdown).

I

_ _ . _ _ _ _ _ _ - = - - - - - - - -- - - - - - - ~ - ~ ~ ~ ~ - - ~ '

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12/87

, g. *

' FORCED OUTAGE' RATE-PURPOSE .

The. purpose of the forced outage rate indicator is to monitor industry progress in minimizing unplanned outages that are forced as a result of equipment failure or other conditions.

This indicator reflects the effectiveness of plant programs and practices (e.g., preventive. maintenance and the correction of design problems).in maintaining systems available for safe electrical generation. Experience has shown that units with high equivalent availability factors and low forced outage rates are often well maintained, follow good operating practices, and can be expected to have a higher margin of safety.

OEFINITION This indicator is defined as the percentage of time that the unit was unavailable due to forced events compared to the time planned for electrical generation. Forced events are failures or other unplanned conditions that require removing the unit from service before the end of the next weekend.-

Forced events include startup failures and events initiated while the unit is

.in reserve shutdown (i.e., the unit is available but not in service).

DATA ELEENTS The following data is required to determine each unit's value for this indicator:

o forced outage hours: the time attributable to unit startup failures and unscheduled outages required before the end of the next weekend -- Forced outage hours include the time from opening the output breaker or declaring the unit unavailable for synchronizing to the grid, until the output breaker is closed or the unit is declared available in reserve shutdown, o service hours: the time during which the unit is' synchronized to the

-system CALCULATIONS The unit and industry values for this indicator are determined as follows:

o .value for a unit = (forced outage hours for the time period) x 100%

(sum of the forced outage hours and service hours for the time period) o value for the industry = average (mean) of the unit values

l 12/87 R.

• DATA QUALIFICATION REQUIRDENTS Data collection begins January 1 of the first full calendar year following -

full power licensing for U.S. units (following commercial operation for international units). Data for U.S. units are included in the calculation of 4 1

industry values beginning January 1 of the second full calendar year following full power licensing. In addition, data elements must be provided for at least 50 percent of the time period in order to be included in the industry values.

CLARIFYING NOTES '

o If a unit is in an unplanned outage that was (or could be) deferred past the next weekend after the problem was identified, but could not have been deferred until the next planned outage, then the unit is in a  ;

" maintenance outage" rather than a forced outage. Also, a unit is in a

" planned outage" rather than a forced outage if it is unavailable due to inspection, maintenance, testing, overhaul, or refueling which has been scheduled "well in advance." This usually means at the start of the current fuel cycle.

o In some cases, the opportunity exists during forced outages to perform some maintenance that would have been performed during the next planned outage. If the additional work extends the outage beyond that required for the forced outage, the remaining outage time is considered a planned or maintenance outage.

o If the duration of a " planned outage (basic)," i.e., the initially scheduled outage period, is extended to complete planned and scheduled work that was originally defined as a part of the planned outage, but could not be completed as scheduled, then the period of extension is called a " planned outage (extended)." Any condition identified during the planned outage (basic) that was not initially scheduled, requires j corrective action to make the unit available, cannot be completed during the planned outage (extended) period, and cannot be deferred, should be considered a forced outage. The forced outage hours are counted from the time a planned outage (basic) was terminated until the unit is made available. .

o The forced outage rate definition and calculation are consistent with that used by the Generating Availability Data System (GADS) of the North American Electric Reliability Council (NERC).

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12/87 L

COLLECTIVE RADIATION EXPOSURE PL5tPOSE The purpose of the collective radiation exposure indicator is to monitor efforts to minimize total radiation exposure at each facility and in the industry as a whole. Radiation exposure has been demonstrated to be related to health risks. This parameter is a measure of the effectiveness of radiological protection programs in minimizing this health risk to plant L workers.

1 DEFINITION Collective radiation exposure is the total external whole-body dose received by all.on-site personnel (including contractors and visitors) during a time period, as measured by the primary dosimeter, thermoluminescent dosimeter (TLD) or film badge. Exposure measured by direct reading dosimeters should be included only for those periods when more accurate data are not available.

Collective radiation exposure is reported in units of man-rem.

i DATA ELEMENTS The total collective radiation exposure for the station is the only data required to determine each unit's value for this indicator. This indicator value is normally based on data obtained quarterly. However, since TLD or film badge information may not be available for the current quarter, the following data is reported:

o total man-rem for the current quarter (TLD, film badge, or direct reading dosimeter) o total man-rem for the previous quarter (TLD or film badge only) -- This replaces the data for the previous quarter which may have used direct reading dosimeter values.

The conversion from Sieverts to rem is 1 Stevert = 100 res.

CALCULATIONS The unit and industry values for this indicator are determined as follows:

o value for a unit = total unit collective radiation exposure during the '

(forperiodsof period one year or less)

To allow more meaningful comparison of unit performance, collective radiation exposure is presented for a three-year period to minimize the impact of annual variations due to refueling and planned maintenance outages. The unit values for the three-year period are determined as follows:

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, ., j 12/87 three-year unit value = average (mean) of the annual unit values o value for the industry = average (mean) of the unit values 1 DATA QUALIFICATION REQUIREMENTS Data collection begins January 1 of the first full calendar year following )

full power licensing for U.S. units (following connercial operation for international units). Data for U.S. units are included in the calculation of industry values beginning January 1 of the second full enlendar year fc11owing i full power licensing. In addition, data elements must be provided for at least 50 percent of the time period in order to be included in the industry values.

CLARIFYING NOTES o For multi-unit stations that do not track collective radiation exposure separately for each unit, unit values are estimated by dividing the sta-tion data by the number of operating units at the station. This allows for more meaningful comparisons among single and multi-unit stations.

o Due to design differences, this indicator is presented by reactor type (e.g., BWR or PWR).

o This indicator measures the total exposure received on-site by all personnel and therefore includes contractors and visitors.

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