ML20058E249
| ML20058E249 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/19/1990 |
| From: | Greeg R, Gregg R EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | Ruth L Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6553, RTR-NUREG-1150 NUDOCS 9011070094 | |
| Download: ML20058E249 (7) | |
Text
'
a
~Q (NMk f*'
5
- if"' ^
- V Q
idaho Natronal ingrneereng laboratory i
F i
d_
)
p October 19, 1990 m
?
L. C. Ruth i
?.
Office of Nuclear Reactor Regulation ci U. S. Nuclear Regulatory Commission 1
E; OWFN Mailstop 12 H26
(
Washington, DC 20555
(
TRANSMITTAL OF LETTER REPORT " SUPPLEMENT 1 TO THE RISK-BASED INSPECTION yT GUIDE FOR THE SEQUOYAH NUCLEAR STATION", REG-18 90 (FIN A6553) g.
L
Reference:
NRC form 189, "PRA Application Program for Inspection at q
Nuclear Power Plants" (FIN A6553 Rev. 4), January 1988 m
.Ma eM F
DearMr. Ruth:
l:
NLhb n hiW W Theattachedre$0?$h?$2kgD;bbA K
.. % portlis transmitteddnifulfillment of node 126-18 of the p
NRC Technical Assistance Division milestone chart for the referenced 189.
This report provides supplemental information for the Sequoyah risk-based inspection guide (RIG).
The information is being provided as requested by the RIG review comments received from Region II.
It will allow h
inspectors to compare actual system and component failure rates with G
those assumed in NUREG-Il50. Thus allowing the inspector to assess the g
importance of equipment / system failures and the length of their i
associated maintenance downtime.
A-y Please direct any comments or questions you might have to me.
I may be y;
contacted on FTS 583-0504.
- 19 F
Sincerely.
4.
h
- c R. E. Gregg
[
NRC Risk Analysis Unit i
reg y
310111 Attachments p-As Stated p
a cc:
S. F. Armour. DOE-ID, MS 1134
?
S. H. Long, NRC-NRR (5 Copies)
Nefe c
d1 S eeJT
? b R-Pe2.
, gg
'2)o Ri5 Moo 8'E l'M-9f
- l, P
it.
l}
Q hEGcG..
sI p.o. sox 1s2s idaho rett 83415 k
D l ' OT I
\\
w 4
m- _
g-;
W.
Supplement 1 to the
('
Risk-Based inspection Guide for the O
Sequoyah Nuclear Station 23 f
This document will allow inspectors to compare actual plant performance a
values with the estimated risk values used in the Sequoyah NUREG-1150 P
probabilistic risk assessment (PRA).
The information is divided into three categories: initiating event frequencies, system failure rates, and component failure rates.
p Initiating Event Frequencies (Table 1)
L' C
Initiating events are those off-normal occurrences that begin potential accident sequences. The more frequently these events occur, the more often the plant's Engineered Safety Features (ESF) are required to function to prevent core damage. Thus, unusually high occurrence rates for these events will increase risk, even though the ESF equipment is operating with the reliability levels assumed in the PRA.
Table 1 lists the initiating events used in the Sequoyah PRA and provides their assumed frequencies of occurrence. The table also indicated the fraction of the total core damage prot' ability that is expected to result c
from each of these initiating events, given the occurrence frequencies indicated.
2 This information can be used to estimate the risk significance for an actual k
frequency of an initiating event that is different form the value assumed in 5
the PRA. For instance, if the actual frequency for loss of offsite power (LOSP) is once every two years, this is 0.5/0.091 = 5.5 times the frequency assumed in the PRA. Since LOSP is the initiator for 26% of the core melt frequency in the PRA, the actual LOSP rate is producing a risk level of 5.5 x 26% = 143% of the total risk level estimated in the PRA, or a risk increase of 142% - 26% = 117%.
System failure Rates (Table 2) l
,w. M h k ' f., 2, 7 r
System reliability information;is provideddin; Table 2. 'Because it is unusual for a whole' system to'be' inoperable, the table also provides P
information about train reliabilities. Two types of train-reliability information are provided: the probability that a normally running f{
system / train will not continue to function as required,-and the probability that a standby system / train will not start and then continue to run as required. For standby trains, maintenance outage time contributes to the 1
probability that the system will not be functional when required.
Therefore, the number of hours per year when a train is out of service for li maintenance can have a large impact on train reliability. Maintenance is N
of ten the major contributor to train unavailability.
Because it is a e
convenient way for an inspector to gauge the potential increase in risk due P.
to frequent or extensive maintenance on a system, assumed PRA train maintenance times are provided in Table 2.
s Y
I 1
s w
b hQlJE N AMIM NWA~ ~
.a:
~
ZZZ
tep p,
When using Table 2, note that the failure probability for a whole system is 7[-
greater than the product of the failure probabilities for its individual trains. The reason is that an allowance has been included for common cause failures simultaneously affecting two (or more) trains of the system.
Maintenance errors, design flaws, and other things affecting all trains in the same way are examples of the events that have contributed to the system failure rate beyond the value that results from independent concurrent failures of all trains.
In addition to the system failure rates, Table 2 indicated the fraction of the core melt probability that involves failure of each system. This O
information is intended to indicate the relative importance of each system, E
so there is a context for assessing the importance of abnormal failure m
rates.
If an abnormal failure rate is found for an entire system, the effect on core damage probability can be estimated by the same method illustrat'ed above for initiating events.
However, if the actual failure rate is determined only at the train level, calculation of the ef fect on core damage probability is not so easily explained.
Still, the importance of the system and the magnitude of the failure rate increase should provide an inspector with an adequate indication of the significance of such a finding.
Component Failure Rates (Table 3)
Table 3 provides the failure rates assumed in the PRA for some of the major components and classes of components.. In the case of major components, such as the emergency diesel generators, failure rates can be tracked individually, and the, significance to reliability of a system is obvious.
AL In the case offaiclassieficomporients;isuchtas' motor operated valves, the f1 relationshipVof,the' fallhe: rate'to system'reliabilities or core damage frequency'is not so obvious, even to an experienced PRA analyst. Still, it is useful to know if the failure rate for a particular type of component is substantially exceeding the expected values.
If so, then attention can be directed toward determining the root cause and appropriate corrective measures.
h i
s
{
f h
2 y
d b:.;w.
4,
rr. -
l Table 1 Frequency.of initiating events
- '.r J:
l Initiating event PRA Annual Frequency
% Contribution I
~
Description Abbreviation (Events / Year) to Core Damage Loss of Offsite Power T1 9.1E-2 26 Station Blackouta SB0 3.1E-4 25b Loss of Main Feedwater T2 7.2E-1 4
Turbine Trip with MFW T3 6.3 1
and Power Conversion System Initially Available' Transient Requiring T
5.3 3
Scram Loss of 125V DC Bus TDC 5.0E-3 1
Steam Generator Tube Tg 1.0E-2 3
3 Rupture Large loss of Coolant A
5.0E-4 5
Accident (LOCA)
Medium LOCA Si 1.0E-3 11 g,.:+-
(:
Small LOCA' S.
1.0E-3 11 2
v.
Very Small LOCA S3 1.3E-2 34 Interfacing LOCA V
6.5E-7 1
a.
SB0 is a significant initiator during high power operation because of the safety equipment that must function to remove decay heat. However, an SB0 is more likely to occur during shutdown periods when the station's turbine generator is not available as a source of power.
b.
5B0 is a subset of loss of Offsite Power that occurs when the lf emergency diesel generators fail to provide backup power.
p E
1
.r.
t i
i G
3 s
zt ei:
a
~*
1 t;, *
- 6..
I Table 2 Probability of system failures Probability
- Contribution System (or Train)/ Functional Requirement of Failure to Core Damage Accumulators (Flow from 3/3 intact trains) 2.6E-3 2
l Auxiliary Feedwater System 18 j
1/3 pump trains to 2/4 SGs 8.9E-5 Li.
2/2 MOP trains or TDP train to 3/4 SGs 2.6E-3 Turbine driven pump train 1.6E-1 Motor driven pump train 7.6E-3 AFW MDP train test /maint. unavailability 2.0E-3 (18 Hr/Yr)
AFW TDP train test /maint. unavailability 1.0E-2 (88 Hr/Yr) h Charging System 5
4 Injection from 1/2 trains 1.4E-4 Recirculation from 1/2 trains 5.1E-3 Charging pump train 5.3E-3 Charging train test /maint unavailability 2.0E-3 (18 Hr/Yr)
Component Cooling Water System 7.0E-5
<1 Running pump train 7.2E-4 Standby ;.;mp train 5.9E-3 CCW train test /maint. unavailability 2.0E-3 (18 Hr/Yr)
L Electric Power 27 Power from 1/2 diesel generators (LOSP) 3.4E-3
- 4. )
Supply emergency power from other unit 2.8E-1 Supply DC power from other unit 5.5E-1 DC power unavailable from a vital bus 9.0E-5
{
Diesel generator test /maint. unavail.
6.0E-3 (54 Hr/Yr)
L Essential Raw Cooling Water
<1 Running pump train 7.2E-4 Standby pump train 7.6E-3 ERCW train test /maint. unavailability 2.0E-3 (18 Hr/Yr)
P, Low Pressure Injection 16 C
Low pressure injection 6.4E-4 b
Low pressure recirculation 2.8E-2 8 -
Lp! pump train 6.1E-3 LPI train test /maint. unavailability 2.0E-3 (18 Hr/Yr) k High Pressure injection 41 f
High pressure injection 1.5E-3 f[t.
High pressure recirculation 2.9E-2 S1 pump train 8.3E-3
[.
51 train test /maint. unavailability 2.0E-3 (18 Hr/Yr) i 4
L h
T'
{
. s.w
+.
.a m
e a
i
~...
Table 2 (Continued) i 3
Probability
- Contribution Syst (or Train)/ Functional Recuirement of Failure to Core Damage Pressure Relief 1
2/2 PORVs and block valves 3.6E-2 3/3 SRVs or 2/3 SRVs and 2/2 PORVs 1.7E-4 Any SRV or PORV to reclose 1.4E-1 Reactor Protection (Auto Scram) 6.0E-5 3
Y 1
i l
I I
4 5
l 4
x tp..
m u-
l
- 4 g
i j J Table 3 Component failure rates i
I.
(L +
Failure rate per t'
Comoonent type Failure mode hour or demand Actuation train Fails to actuate 1.6E-3/D Air operated valve Fails to open 1.3E-3/D I
['
Plugged 1.0E-7/Hr l.
Automatic scram system Fails to initiate scram 6.0E-5/D Block valve Valve shut 2.0E-la
)
Falls to open 3.0E-3/D Check valve Fails to open 1.0E-4/D 1
Fails to close 1.0E-3/D
}
l Diesel generator Falls to start 3.0E-2/D i
Fails to run 2.0E-3/Hr j
Test /maint. unavail.
6.0E-3 (54 Hr/Yr) i DC bus Loss of bus 9.0E-5/Hr Motor driven pump Fails to start 3.3E-3/D Fails to run 3.5E-5/Hr l.
Test /maint. unavail.
2.0E-3 (18 Hr/Yr) i i
Motor operated valve Fails to open 3.3E-3/D l
Plugged 1.0E-7/Hr
{
Test /maint, unavail.
2.0E-4 (2 Hr/Yr) i 1
PORV Fails to open 6.3E-3/D l!
Fails to close 3.0E-2/D i
P Pump Actuation Standby pump fails 1.6E-3/D
{
t to start ll Strainer Plugged 3.0E-5/Hr l
t i
Sump strainer Plugged 5.0E-5/D 1
Throttle / trip valve Fails to open 3.3E-3/D Turbine driven pump Fails to start 3.3E-3/D Fails ta run 3.5E-5/Hr Test /maint, unavail.
1.0E-2 (88 Hr/Yr) 1 a.
This value is the percent of time a block valve is left in the shut position.
6 p
sa v
_. -.....