ML20197E121
| ML20197E121 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/08/1986 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Noonan V Office of Nuclear Reactor Regulation |
| References | |
| SBN-1047, NUDOCS 8605150172 | |
| Download: ML20197E121 (33) | |
Text
-
George S. Thomas VICe Pres;derit-Nuclect Prodtetosi Putsc Service of New Hampshire May 8, 1986 N:w Hampshire Yankee Division SBN-1047 T.F. B7.1.2 1
United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Vincent S. Noonan, Project Director PWR Project Directorate #5 l
Reference:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) Letter from G. S. Thomas (New Hampshire Yankee) to G. W. Knighton (NRC) dated August 23, 1985, " Supporting Analyses for Seabrook Station Technical Specifications" (c) Letter from G. S. Thomas (New Hampshire Yankee) to V. S. Noonan (NRC) dated December 17, 1985, " Table of Risk-Based Changes Included in the Proposed Seabrook Station Technical Specifications" (d) Letter from V. Nerses (NRC) to R. J. Harrison (PSNH) dated January 7,1986, " Technical Specification, Request for Additional Information" (e) Letter from C. S. Thomas (New Hampshire Yankee) to
! V. S. Noonan (NRC) dated January 31, 1986, " Response i
to Request for Additional Information Regarding Risk-Based Technical Specification Changes" l Subject : Response to Request for Additional Information Regarding Risk-Based Technical Specification Changes l
Dear Sir:
Enclosed are responses to your Request for Additional Information (Reference d), questions Q2, Q4, and Q7. Responses to the other questions were submitted by letter dated January 31, 1986 (Reference e).
i j
' ' The responses enclosed are in support of risk-based changes to Technical l Specification 3.5.5 RWST, 3.7.4 Service Water, 3.7.5 Ultimate Heat Sink, and 3.8.1.1 Electrical Power Systems (onsite). A summary of the results is l' provided in Enclosure 1.
1 8605150172 860508 l PDR ADOCK 05000443 A PDR g L
P.O. Box 300 + Seabrook,NHO3874
- Telephone (603)474-9521
United States Nuclear Regulatory Commission May 8, 1986 Attention: Mr. Vincent S. Noonan Page 2 The detailed analyses supporting these results are given in Enclosure 2 to this letter. The upper bound delta risks given for Q2 and Q4 are judged to be very conservative - i.e. . the best estimate of the delta risk is much smaller than the upper bound values. The primary source of conservation was the assumption that the mean repair time is equal to the Allowed Outage Time (A0T). This and other conservatisms are discussed in more detail in the Enclosure. In addition, there are additional risk benefits gained by making the above changes that have not been quantified. These benefits have been outlined in Reference e, in the column labeled " Basis for Changes." In total, these changes are considered to have a very small and insignificant effect on the risk of core melt and an even smaller effect on public risk. Also, a description of the Seabrook Station Diesel Generator Reliability Program is included in response to Q7.
We trust that the enclosed information provides adequate response to your questions. If you require additional information, please contact Mr. Kenneth Kiper at (603) 474-9574, extension 4049.
Very truly yours,
& ,=O George S. Thomas GST:KLK:cj b Enclosure ,
i
Enclorurs 1 Summary of Results Upper Bound NRC Proposed Estimate Best Estimate Question System Technical Specification of Delta Risk of Delta Risk Q2 RWST Two level TS: 4.6 E-6 4.6 E-8 o 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> A0T for out-of-spec less than 10%.
o I hour A0T for
, out-of-spec greater than 10%.
Q4 Cooling o One Cooling Tower 2.1 E-6 2.0 E-7 Tower train inoperable -
14 day A0T.
o Both Cooling Tower trains inoperable -
7 day A0T.
't o Cooling Tower basin -
7 day A0T.
l l
1 4
e 4
2 J
Enclosure 2 Responses to Questions Q2, Q4 and Q7
_7 Q.2 NRC REQUEST -
3.5.5 RWST: There is no supporting documentation on RWST outages. We need .the following to evaluate this request:
- 1. An analysis which shows that the NPSH is _ adequate with 431,000 gallons in the RWST when there is a switchover from injection to recirculation with all pumps running following a large LOCA.
- 2. Reference to an analysis that shows that 1800 ppm boron is adequate to cover the spectrum of DBA's analyzed in the-FSAR.
RESPONSE
The proposed Technical Specification (see Attachment Q2-1) would change the RWST inoperability action from a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowed outage time (A0T) to a two stage A0T: 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the RWST volume, boron concentration, or temperature less than 10 percent out-of-spec and I hour for out-of-spec conditions greater than this. This change would allow for greater flexibility in operating the plant while having a. negligible change in the core melt risk. Justification for this proposed change is provided in the following paragraphs:
(1) In response to the NRC Request, if an analysis were available that demonstrated adequate NPSH and boron concentration with a ten percent reduction in Technical Specification parameters, then the parameter limits would have been changed on a deterministic basis and no risk assessment would be needed.
(2) As stated in our submittal of December 17, 1985 (Reference Q2-1),
the proposed change would allow up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to ~ restore volume or boron concentration' if they were out-of-spec by 10% or less. This 10% deviation would have~ a minimal effect on the ability of the RWST to function in realistic accident considerations.
(3) The period of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> would allow time for reverification of level or concentration, if necessary, and restoration without requiring a plant shutdown. This change in A0T better reflects the importance of an out-of-spec condition and the priority for restoration.
(4) A conservative estimate of the " delta core melt" risk due to the proposed Technical Specification change is provided below in order to provide quantitative support to the qualitative justifications offered above. A number of conservative assumptions were made in order to simplify the analysis because of the low level of delta risk involved.
Maintenance Data The RWST unavailability due to tank failure (for small LOCA initiated sequences) is 1.60 E-7 (Reference Q2-2, SSPSA p. D.8-48). The main-tenance contribution was not considered in the SSPSA because of the low f requency and short duration of maintenance. To estimate a maintenance term, the mean duration of maintenance is assumed to be equal to the present A0T, I hour and the proposed A0T, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The frequency of maintenance (i.e., the frequency of being in the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> A0T) is assumed to be -roughly the same as the accumulator frequency of maintenance '(see Enclosure 2 to Reference Q2-3, 'page 12) due to similarities in monitored parameters (volume or level and boron concentration).
Q2-1 s
l l
l Tha spprcpriate mean valua is 5.36. E-5 per hour which is' a " par accumulator tank";rather than "per' accumulator system" value. Note that the maintenance term being calculated is not the contribution to RWST unavailability but to RWST "out of analyzed design basis space." Thus,
- the maintenance terms are
t (maintenance frequency) * (maintenance duration - A0T)
= 5.36 E-5 for I hour A0T
= 3.22 E-4 for 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> A0T i
i The highest frequency initiating event to demand the RWST is SLOCA, at a 1
frequency of 0.0173 events per year. The frequencies of challenges to the design basis due to RWST out-of-spec less than 10% are:
i j SLOCA
- RWST (Maintenance) j = 9.27 E-7 for I hour A0T i
J
= 5.56 E-6 for 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> A0T 4.63 E-6 delta I
i It should be noted that the likelihood of a 10% out-of-spec condition
{ causing failure of the RWST to function is judged to be very small (less j than one chance in 100).
4
- Core Melt Sequences i
1 The dominate core melt accident sequence initiated by SLOCA is Sequence 1 8A-1 (from Reference Q2-4, page 5-23):
1 8A-1: SLOCA
- L13
- L2C = 9.40 E-6 per RY t
where L13 =
failure of the RRR train A in the "miniflow recirculation" mode = 1.54 E-2, and '
i
! L2C =
failure of RRR train B in the "miniflow l recirculation"' mode = 3.53 E-2.
i The SLOCA risk is dominated by the failure of the RRR in miniflow
} recirculation. (See the SSPSA pp. D.8-6 and 49 for details of the j miniflow mode.) The unavailability of RWST due to maintenance (the 10%
i out-of-spec condition) is estimated to be less than 3.22 E-6 .(= 3.22 E-4 j
- 0.01). This clearly will have no effect on the frequency. of this-sequence. Thus, the delta risk associated with the change in A0T from I hour to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is judged to be very low (less than 4.63 E-6
- 0.01 =
- 4.63 E-8) and negligible.
4 i
The initiating event of concern for out-of-spec boron concentration, l steam line break inside containment, is more than 30 times less frequent l than SLOCA. Thus, the change in A0T is also judged to be negligible for j this initiating event.' Any other; initiating. events which demand RWST are
{ lower in frequency and the associated core melt sequences involve-
- failures that are much more likely to occur than. RWST " failure" due to a j 10% out-of-spec condition.
l Q2-2 4
,, _ , . ,m. _ - . - - _ , _ - . - ____ .._ _ _ __ _ _ _ . _ _ _ . . _ _ _ _ _ _ , . _ . _ _ _ , -
Thus, tha upper bound (licsnting) estimate cf dalte cora melt fraquincy is estimated to be 4.63 E-6 and the best estimate is 4.63 E-8.
~
Therefore, the total delta core melt frequency is judged to be very low and is negligible.
REFERENCES
'Q2-1. Thomas, George S., Table of Risk-Based Changes Included in the Proposed Seabrook Station Technical Specifications, New Hampshire Yankee, letter to V. S. Noonan, U. S. Nuclear Regulatory Commission, December 17, 1985.
Q2-2. Pickard, Lowe and Garrick, Inc., "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, PLG-0300, December 1983.
Q2-3. Thomas, George S., Response to Request for Additional Information Regarding Risk-Based Technical Specification Changes, New Hampshire Yankee, letter to V. S. Noonan, U. S. Nuclear Regulatory Commission, January 31, 1986.
Q2-4. Pickard, Lowe and Garrick, Inc., " Risk-Based Evaluation of Technical Specifications for Seabrook Station," prepared for Public Service Company of New Hampshire, PLG-0431, August 1985.
4 i
Q2-3
- w. -- . _
/ .tachment Q2-1 EORONINJECTION5YSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION
- /9 T DCC a.
A minimum contained barated water volume of 4M;jM gallons,
- b. A minimum boren concentration of 2000 ppm of baron, A minimum solution t'emperature of *F, and c.
- d. A maximum solution temperature of *F. .
APPLICAEILITY: MODES 1, 2, 3, and 4. -
ACTION: to, ,, p g ,, ; .- 7 Gj
-~ .
With the RWST inoperable, restore the tank to OPERABLE status within,f1four:or be in at least HOT STANO3Y within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />'s and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. _-_ - -. n ---- ~ ~ ~ . __
A A k4. to OPERAGLF Mur g L . (w n Lepr. A la aq ( .te M11CT.sTAN06'/ A , a$% v:4:s . (, 6MA
(.0l 0 -
J w'. Q,,. I k.or 44 bc 6 a.t- u. s t- .
~
S'itCT Ctt44 wi A A fede r 30 houn' ,
u SURVEILLANCEREOUIkE.ENTS 4.5.4 The RWST shall be demonstrated CPERABLE:
- a. At least once per 7 days by: -
- 1) Verifying the contained barated water volume in the tank, and
- 2) Verifying the baron concentration of the water.
- b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature.
- I - - -~ . . ,
.w ,. ,
j ,' : I- i i e MlI L A'i --- d I.: O d :i' 2:1. .,,d,"i f4~
SEA 8RCOX - UNIT 1 3/4 5-10 i y' w ..
1Q.4 NRC_ REQUEST-3.7.5 Ultimate Heat Sink: The request for relaxation of the allowed outage time for the cooling tower is not adequately supported. To
- evaluate the probabilistic significance of this change, we need:
- 1. A list of sequences that are affected by service water and cooling
- tower outages. Each sequence should identify the events that make up the sequence and the quantification of each event.
, At least two events in each sequence would cover the service water 1
and cooling tower outages and/or failure probabilities so that the staff can modify the quantification as deemed necessary.
- 2. Industry data on the frequency (per year) and duration of services water and/or cooling tower outages broken down by plant, year, and A0T of 7 days.
RESPONSE
i In order to put this response in perspective, it was necessary to first clearly define the proposed changes to the Cooling Tower Technical
, Specification. Then, the qualitative justifications for the changes are i summarized, followed by a detailed quantitative analysis of the effect of i the changes to the system and plant models.
1 I
Technical Specification Change The proposed Technical Specifications which affect the Cooling Tower are
! included with this response (Attachment Q4-1). These Technical l Specifications, 3/4.7.4 " Service Water System" and 3/4.7.5 " Ultimate Heat i
Sink," have been modified in order to make explicit the allowed outage times (A0Ts) for various combinations of Cooling Tower pumps, fans and j basin. The following summarizes the A0Ts for the Cooling Tower:
Allowed Inoperability State ' Outage Time TS l o 1 Cooling Tower train [ pump and/or fan (s)} 14 days 3.7.4 i o both Cooling Tower trains (pumps and/or fans) 7 days 3.7.4 j o mechanical draft Cooling Tower (basin) 7 days 3.7.5 I
In the same Technical Specification, the values for Service Water pump train A0T are much shorter (I required SW pump inoperable - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, 2 i
required SW pumps inoperable - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,.SW pumphouse inoperable - 24 l hours). These values were chosen in order to emphasize the relative risk importance of the Service Water pumphouse and associated pumps over the cooling tower. The analysis that follows provides the justification for j the proposed Cooling Tower Technical Specifications..
f Cooling Tower i
j The qualitative bases for the relatively low risk importance of the Cooling Tower compared to the Service Water;pumphouse and, thus, the justification for the proposed Cooling Tower Technical Specifications,
- were provided in an earlier submittal (Reference Q4-1). These are j summarized below
e
- Q4-1
(1) Tha licansing $saign basis -for tha Cosling Tcwer is that the Service Water intake and . discharge tunnels cannot be seismically qualified. However, the seismic analysis performed in connection with the;SSPSA Section 9.2_(Reference Q4-2) concluded that the tunnels would actually survive a much larger seismic event than the cooling tower. The median acceleration capacity for the tunnels and transition structures is 4.6g and for the cooling tower structure, 2.4g. . Based on these values, the Service Water tunnels and pumphouse are clearly more reliable' than the Cooling Tower with regard to seismic events. Thus, the proposed A0T's (1 day for Service Water pumphouse,- 7 days for Cooling Tower basin) reflect the relative risk-importance of the structures.
(2) The Service Water pumphouse provides suction for 4 pumps while' the Cooling Tower basin provides suction for 2 pumps and also requires fan operation. Based on this observation, the pumphouse and tunnels would appear to be more risk important than the Cooling Tower. This observation is supported by systems analysis in the SSPSA and the Technical Specification Study (Reference Q4-3) and is summarized in this response.
(3) Functionally, the Cooling Tower is used as a backup to the Service Water pumphouse. Due to operational constraints and design basis limitations, _ the Cooling Tower is not designed for normal plant modes of operation and would be used for plant cooling only for unusual and infrequent events. The only planned event when the Cooling Tower will be used is tunnel heat treatment which occurs over a short duration and during which the Service Water pumphouse and pumps are still available.
Thus, the extended A0T for the Cooling Tower also reflects its operational functions as well as risk importance.
Additional quantitative risk analysis is provided below to further -
support these changes to Technical Specifications. First, the values used for Cooling Tower maintenance frequency and duration are presented and the basis for the values explained. Then, the Cooling Tower system model is described with axplicit maintenance terms to examine the sensitivity of the proposed Technical Specification changes at the system level. Finally, important accident sequences which include Service Water system failure are evaluated to determine the sensitivity of the proposed change to core melt frequency and public risk.
Maintenance Outage Data The -Cooling Tower A0T values were modeled through maintenance unavailability terms, i.e., maintenance duration times maintenance frequency. The maintenance duration term is discussed first.
In order to examine sensitivities to the change in A0Ts, it was necessary to define a current (OLD) and proposed (NEW) set of A0T values. These maintenance duration values were further divided into " licensing" values and " realistic" values. For the " licensing" calculation, it was conservatively assumed that the maintenance duration (the mean time to restore) was equal to the A0T. For the ~" realistic" calculation, the mean value of a distribution for duration of maintenance for similar components was used. These values are discussed further and are summarized as follows:
Q4-2 4
M; int:nInca '
Lic naing Riolictic Outage Duration OLD NEW OLD NEW One CT Train 72 hrs 336 hrs 20.9 hrs 116.4 hrs Two CT Trains 168 hrs - - -
40.4 hrs CT Basin l 72 hrs 168 hrs The OLD " licensing" values are based on the interpretation of the Cooling Tower Technical Specification prior to the proposed modification. It allowed one Cooling Tower train or the Cooling Tower basin to be out for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> but had no allowance for two Cooling Tower trains to be down.
The NEW " licensing" values are the proposed A0T values. The " realistic" values for outage duration were taken from the SSPSA Section 6.4 (Reference Q4-2). In that section, four component maintenance mean duration distributions were developed based on applicable industry maintenance data. Three of those distributions were used for the
" realistic" values as follows:
Type B: 72-hour LCO -
mean = 20.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Type C: 7-day LCO - mean = 40.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Type D: No operability time limit (14 day LCO) - mean = 116.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> These distributions based on the generic industry data were judged to be the most appropriate for Seabrook prior to obtaining its own operational data . It was judged that additional generic data would not serve to reduce the levels of uncertainty for these distributions.
It is evident that assuming the " licensing" values for maintenance duration is very conservative when compared with the " realistic" values. These conservative assumptions were made because of the low sensitivity of the Cooling Tower to risk and because of the licensing nature of this analysis. However, in a realistic PRA model, such assumptions are considered inappropriate and the " realistic" values will be assigned in future uses of this system modeling.
The other term in the maintenance unavailability term, the maintenance frequency, was assumed to be equal to the Type 1 Distribution, as described in SSPSA Section 6.4. The Type 1 Distribution was developed for standby pumps, tested monthly and has a mean of 8.42 E-5 events per hour (or once per 495 days). This value was applied as follows:
Component Outage Frequency Licensing Realistic (Events Per Hour) OLD NEW OLD NEW One CT Train 8.42 E-5 8.42 E-5 8.42 E-5 8.42 E-5 Two CT Train --------
8.42 E-5 --------
8.42 E-5 CT Basin 8.42 E-5 8.42 E-5 i No outage frequency was included for the "two CT Train - OLD" case because, as discussed previously, the previous Cooling Tower TS did not provide any A0T for this case. For the "CT Basin - Realistic" case, the outage frequency is assumed to be so small as to be negligible. This is based on the very low likelihood of the Cooling Tower basin being actually unavailable to perform its function.
Q4-3
Cooling Tower Systen The Cooling Tower system model is given in Attachment Q4-2 to this response. It :1.s consistent with the analysis in the SSPSA (Appendix D.3, pp. D.3-17, 18, 23,24, 29-31) with several changes. The common cause model was expanded to be more' complete. 'Also, the maintenance unavail-ability model was expanded to include the "two train outage" and "CT basin" terms in addition to the '" single train outage" term, as described in the previous section.
The results of the changes in A0T are summarized as follows:
Cooling Tower A0T Licensing Realistic Outage Duration OLD NEW OLD NEW One CT Train 72 hrs 336 hrs 20.9 hrs 116.4 hrs Two CT Trains 168 hrs --------
40.4 hrs CT Basin 72 hrs 168 hrs -- --------
Unavailability Results CT Total: l 8.42 E-3 20.60 E-3 1.80 E-3 2.92 E-3 o CT Hardware 0.72 E-3 0.72 E-3 0.72 E-3 0.72 E-3 o CT Common Cause 0.85 E-3 0.85 E-3 0.85 E-3 0.85 E-3 o CT Maintenance: 6.86 E-3 19.03 E-3 1 0.23 E-3 1.35 E-3 o CT Single Train 0.79 E-3 3.69 E-3 0.23 E-3 1.28 E-3 o CT Two Train '
1.20 E-3 '
0.07 E-3 o CT Basin 6.07 E-3 14.15 E-3 . - -
Based on the " licensing" results, the changes in A0T's have a significant effect on system unavailability. The "CT Basin" clearly dominates the maintenance term. As discussed in the previous section, the modeling of "CT Basin" A0T was very conservative. The only conceivable means of entering the "CT Basin" A0T is due to high basin water temperature which will not likely fail the Cooling Tower function for most sequences.
For the " realistic" case, the "CT total" is much lower than for the
" licensing" case (by a factor of 5 to 7) indicating the extent of the conservatism in the licensing case. The "CT total" increases by about 62% from the "old" to "new" modeling. This increase is obviously driven by the increase in the maintenance term.
Core Melt Sequences The risk from Seabrook Station can be fairly accurately described by a group of 222 dominant accident sequences. These sequences account for more than 85 percent of the total core melt frequency and includes all sequences down to about 1.0 E-7 per RY in frequency. . These sequences were developed in the SSPSA and are clearly displayed in the Technical Specification Study, Table 5.5 (Reference Q4-3).
Independent failure of the Service Water system is included in 31 out of the 222 core melt sequences. These sequences can be grouped into four sets of sequences, as follows:
(1) Loss of station power (LOSP) initiating event, failure of both trains of Service Water; ,
Q4-4
4 (2) LOCA initiating sysnt, failure of both traina of Ssrvics Watsr; l (3) Transient initiating event, failure of both trains of Service Water; and (4) Loss of Service Water initiating event (LOSW).
i The Cooling Tower is modeled as a Service Water recovery action (SWR-1). This is appropriate because most losses of service water may not generate an automatic Tower Actuation (TA) signal. Thus, an operator action (OPACT) is modeled to diagnose the loss of service water and manually initiate the TA signal. This operator action was judged to be a i relatively reliable action because of the indications available of loss of Service Water and the simplicity of initiating the tower actuation signal. The operator failure rate was judged to be in the range of 1 in 100 to 1 in 10,000 failures per demand. While the "I in.100" value is conservative overall, it also has the effect of swamping out any change in the Technical Specifications. The value of "I in 10,000" may not be appropriate in all sequences because of the high stress condition of the ope rators. Thus, a value of 1 in 1000 was judged to be a reasonable estimate for OPACT. Thus, the Service Water recovery is: '
4 SWR-1 =
OPACT + CT Where OPACT = 0.001 as described above, and CT =
failure of the Cooling Tower system This recovery action is only credited in a limited number of sequences, j as discussed below.
4 (1) The first set of sequences are initiated by loss of station power (LOSP) and include 7D-2, 3, 4, 11, 17, 18, 20, 21; 4A-14, 15 ; 3D-3, 8, 10, 11, 17, 18 ; and 8D-17, 18. t The highest frequency sequence representative of this set is:
- WA3
- NEF2
- ER9 = 6.06 E-6 per RY 4
Where LOSP = frequency of loss of station power initiating event = 0.135 events per RY WA3 = failure of SW train A given LOSP = 1.80 E-2 WBC = failure of SW train B given failure of train A and LOSP = 5.90 E-2 NEF2 = not (N) failure of turbine driven EFW pump
{ = 0.952 ER9 = failure to recover offsite electric power i
before core melt = 0.0444 I,
t Numbers refer to sequence designators from Table 5.5 of Reference Q4-3.
Q4-5
This in a ststien bisckout czquinea - the diessla ctart on LOSP and then fail by overheating due to loss of Service Water cooling, assuming no- operator intervention. The operators fail to restore the offsite grid (diesels are assumed unrecoverable) before core melt. Core uncovery and melt result from the RCP Seal LOCA with no primary system makeup. Successful turbine-driven EFW pump operation delays, but does not prevent, core melt and thus affects' the time available to restore electric power-(ER9).
All but two of the sequences in this set are station blackout sequences, similar to 7D-2, but involve one train of Service Water train failure with the opposite train of electric power (diesel generator) or DC upower failure.- For station blackout sequences caused by loss of Service Water, operator action would be necessary to manually actuate the TA signal initiating tower operation. However, the operator action time is very short (less than 15 minutes) in which to diagnose the loss of Service Water and to manually initiate TA before the diesels overheat I and fail. Thus, no credit can reasonably be taken for the Cooling Tower in station blackout sequences with failure of Service Water.
The other two sequences in this set (4A-13 and 14) consist of LOSP sequences with only one train of service water failed. The failure has also caused failure of the train-related diesel due to overheating. Thus, Cooling Tower operation is not important because one train of Service Water has not failed and the other train has no electric power.
Therefore, for the first set of sequences, Cooling Tower recovery action has not been credited because of the limited time for operator action for the station blackout sequences or because it was not necessary for the sequences with only one Service Water train failed.
(2) The second set of sequences are initiated by a LOCA and includes sequences 2A-7, 8 (Medium LOCAs), 8A-9, 10, 8D-14, 14A, 23 (Small LOCAs) and 8D-24 (SGTR). The highest frequency sequence in this set is:
8D-23: SLOCA
- WAl
- WBA
- NEF2 = 3.79 E-6 per RY Where SLOCA = frequency of small LOCA = 0.0173 events per RY WAl = failure of SW train A, given SI signal
= 5.31 E-3 WBA = failure of SW train B, given failure of
- SW train A and given SI signal = 4.33 E-2 NEF2 = 0.952 For this sequence, Service Water failure is dominated (about 90%
of total system unavailability) by failure to automatically' iso-late the non-essential loads (i.e., failure of MOVs SW-V4 and VS).
Q4-6
This failura moda (indsp;ndant and c:mmon cIurs frilura cf SW-V4
, and V5) also fails the Cooling Tower function. Thus, Cooling Tower recovery is not available in the large proportion of the frequency. Also, the more appropriate recovery model is operator action to locally, manually close SW-V4 and V5 which restores service water in the large majority of cases (recovery action SWR-2). This operator action is realistic because of the time available before Primary Component Cooling Water (PCCW) heats up. Thus, the frequency of the sequence for which the Cooling Tower recovery can be credited is about 0.10 times the frequency of 8D-23, or about 3.79 E-7 per RY. Similar considerations can be made for the rest of the sequences in this set. The total frequency of the sequences in this set (1.25 E-3 x 0.1) times the cooling tower unavailability (8.40 E-3 for the OLD licensing . case) reduces these to about 1.0 E-8. At this level, small deltas are not significant. The frequency is even lower for the realistic calculation of Cooling Tower unavailability. Therefore, for this set of sequences, Cooling Tower recovery sensitivity is not significant.
l (3) The tnird set of sequences are initiated by a transient event and includes sequences 8D-19, 20, 21, and 25. The highest frequency sequence of this set is:
- WA2
- WBB
- NEF2
- SRI = 1.70 E-7 per RY where RT = frequency of reactor trip initiating event = 3.13 per RY WA2 = failure of SW train A = 3.54 E-4 WBB = failure of SW train B given failure of train A =-5.75 E-2 I
NEF2 = 0.952 i
where CT = 1.80 E-3 for the OLD realistic case, and OPACT = 1.0 E-3 The loss of all Service Water causes loss of PCCW which leads to failure of charging pumps due to overheating and failure to cool the RCP thermal barrier heat exchangers. This results in a RCP seal LOCA with no primary system makeup. Success of the turbine-1 driven EFW pump extends the time to, but does not prevent, core melt. Without recovery of cooling to the seals, eventual core uncovery and core melt occurs. For this set of sequences, the Cooling Tower has been credited. The value listed above for SRI includes the OLD A0T using a " realistic" value for the duration of maintenance (as discussed in the pravious section). The results of applying the NEW A0T and using " realistic" and
- licensing" assumptions for this sequence are given below.
i f
4 I
Q4-7
Liesnsing R:011stic OLD NEW OLD NEW CT Total 8.42 E-3 2.06 E-2 1.80 E-3 2.92 E-3 SWR 1 9.42 E-3 2.16 E-2 2.80 E-3 3.92 E-3 Sequence 8D-19 ! 5.71 E-7 1.31 E-6 I 1.70 E-7 2.38 E-7 All Set 3 Sequences i 1.64 E-6 3.76 E-6 l 4.88 E-7 6.84 E-7 Thus the dalta sequence frequency for 8D-19 licensing analysir is 7.39 E-7. The total " licensing" delta for all four nequences~in this set is 2.12 E-6. Similarly, the total " realistic" delta for all sequences is 1.96 E-7.
(4) The fourth set of sequences is initiated by a loss of service water (LOSW) event and includes only sequence 8D-7.
8D-7: LOSW
- NEF2 = 1.93 E-6 per RY where LOSW = loss of Service Water initiating event
= 2.03 E-6 NEF2 = 0.952 This sequence is similar to the above sequences except that Service Water is lost at the initiation of the sequence rather than subsequent to an initiating event. The plant response in the same as discussed previously. The LOSW frequency was calculated in the SSPSA (pp. 6.6-7 to 9) and is summarized as follows:
LOSW = (SW Hardware) (Cooling Tower) + ~ +
(SW Fire) (Cooling Tower) + (Block E)
= 7.62 E-9 + 1.50 E-6 + 5.24 E-7 = 2.03 E-6 per RY where SW Hardware = frequency of SW unavailability
= 4.28 E-6 per RY Cooling Tower = failure of Cooling Tower (assuming OLD A0T) = 1.78 E-3 SW Fire = frequency of fire causing loss of SW pumphouse ventilation = 8.4 E-4 per RY Block E = frequency of failure of common elements between SW and CT = 5.24 E-7 per RY '
Operator action to actuate the Cooling' Tower is not applicable for the last term (Block E) - components that are used in common for Service Water and Cooling Tower, The second term, with SW fire, (1.50 E-6) clearly dominates over the firat term-(7.62 E-9). This SW fire is a fire in one of -several areas in,tbe PAB or SW RVAC equipment room that causes faf. lure of the Service Water air hand-ling system. This ventilation failure causes long term failure of SW pumps if there is no operator action. Realistically, with- such a fire in a safety related area, the plant would begin an orderly ;
shutdown. Also, room temperature alarms in the SW pumphouse would 04-8 , '
i
- giva indic:tien to tha cparcters tha nesd for tetien - i.e.
opening doors, portable fans, etc. Thus, it is very unlikely that this fire would result in a loss of' service water initiating event. Fires that directly fail all service water pumps are much s lower' frequency.
, 1 With the second and third terms dismissed from this evaluation of LOSW sensitivity to Cooling Tower, only the first term remains, but at a very low frequency (less than 1.0 E-8). Thus, sequence ,
8D-7 is not important to Cooling Tower sensitivity.
- Therefore, of all the sequences contributing to core melt frequency and 1
public risk, only transient-initiated sequences (Set 3) include modeling the Cooling Tower and are affected by changes to the Cooling Tower Allowed Outage Times.
CONCLUSION 1
The change in Allowed Outage Times for the Cooling Tower has a significant
- ef fect on the system availability, increasing from 8.42 E-3 to 2.06 E-2 for the conservative " licensing" case. However, the increase in core melt frequency due to this change is insignificant, f rom 1.64 E-6 to 3.76 E-6 per RY. This is due to the fact that the Cooling Tower is a standby system and cannot be credited in most sequences with loss of Service Water.
For the realistic case, the changes are even less significant. At the system level, the Cooling Tower unavailability increases from 1.80 E-3 to 2.92 E-3.
This system level change causes a change in core melt frequency from 4.88 E-7 to 6.84 E-7 per RY. This is judged to be the best estimate of the change in core melt frequency. This change is insignificant because of the levels of uncertainty surrounding each of the data variables. Also, this analysis has not factored in addition benefits of extending the allowed outage times for the Cooling Tower, such as better reflecting the priority for restoring the Cooling Tower components. These benefits are difficult to quantify but contribute to the judgment that these changes do not significantly change the risk from Seabrook Station.
REFERENCES Q4-1. Thomas, George S., Table of Risk-Based Changes Included in the Proposed Seabrook Station Technical Specifications, New Hampshire Yankee, letter to V. S. Noonan, U. S. Nuclear Regulatory Commission, December 17, 1985.
Q4-2. Pickard, Lowe and Garrick, Inc., "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hampshire and 1
Yankee Atomic Electric Company, PLG-0300, December 1983.
Q4-3. Pickard, Lowe and Garrick, Inc., " Risk-Based Evaluation of Technical Specifications for Seabrook Station," prepared for Public Service Company of New Hampshire, PLC-0431, August 1985.
f
.Q4-9
e a -.
Attachment Q4-1 PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION Stbje. <d EMv+ h ,}o rf5 ~~
shall be OPERA 3LE.
.4
~
Att Teast two independent service watersy:t=:
APPLICABILITY _: H00E5 1, 2, 3, 4.
ACTION: ,
With only one service water loop OPERABLE restore two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOBY within the,next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD _5liUECWN within the followinc 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .
SURVEILLANCE REOUIREMENTS h*
4.7.4 At least two {t: tier Service Water 4 l:'::- shall be demonstrat
~
OPERAELE: ,
- a. At least once per 31 days by verifying that each valve (manual, power-cperated, or aut =atic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and- ' .
. ~
- g. At 1 East once 'per 18 months during shutdown, by verifying that:
c.
- 1) Each automatic valve servicing safety-related equipment actuatas to its correct position on its associated Engineered Safety Fea-ture actuation test signal,, $nd
( o d T. w Advdew %+ %d
- 2) Each -Station Service Water System- pump starts automatically upon loss of cr-f !hr:-t; :t:rt :f t.t n '.u cu. pu=p wi niu i;.c ~
r-:> f S i} % f , a d k TE: f e w V n esa cmuy h p, m.% %% a cJ13 e ~ re -
aw;~. w ap t.
~
esa (b. Qt w+ - p 9 a$s g s w g A & e .t is- w .. w .
cxg T .
A ca opa3 c. A % o a . w
, . . . . . . .Mjg )$ l@88
., . . s
- -~ y, l l .
\ i. .
)nJ....
-K . s. ;
'2 / ' 7i7 s
) , ..s
i
@ 3.7.4 Two independent Service Water System loops shall be OPERABLE with each loop comprised of:
a) one OPERABLE Service Water pump, 2
b) one OPERABLE Cooling Tower pump, t
c) one OPERABLE Cooling Tower fan *, and d) one OPERABLE flow path capable of taking suction ,
from the related Ultimate Heat Sink.
j i APPLICABILITY: Modes 1, 2, 3 and 4.
ACTION:
i With one Service Water System loop inoperable in the state listed below, restore the inoperable loop to OPERABLE status within the ALLOWED time or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> j and in HOT SHUTDOWN with the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Inoperability State Allowed Outage Time 1
1 Cooling Tower train [ pump 33; fan (s)] 14 days i
Both Cooling Tower trains (pumps 33; fans) 7 days I required Service Water pump 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> '4 1 required Service Water pump and the opposite train of Cooling Tower 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Both required Service Water pumps 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4
- For train B, the cooling tower operability requires both center cell cooling fans.
I
... __, ~_4 _ . - _ _ . _ _ ,
PLANT SYSTEMS 3/4.7.5 ULTIMATE- HEAT SINK LIMITING COODITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with:
- a. A service water pu=phouse water level at dr above minus 37'0" Mean
- Sea Level
- USGS datum, and jevsl
- b. A mechanical draft ecoling tower / comprised of -two-cooHng-tower-f ans-
'I of equal to or greater than 3fd #, %and-a contained basin water wci ::G10" biiuos 'at a -
to.*}C*F, and 9,jg n.2 - 1"~.GO LE A s .iv cays
-c. " p:?t:ble teuer 2 i , p ...g -5.7 s tem - a ts . .d -tw i 4:H :u'.gn m-5 fc Chutd ur. 5:rthquah.
APPLICASILITY: MODES 1, 2, 3, and 4. .
M hm.,.,s ,
ACTION: .
- a. With the se'rvice water pumpheuse inoperable,festore the service water pe=phouse to OPERAELE status within -72' hours, or be in at least HOT STANOSY within the next 6 hcurs and in COLD SHUTDOWN within the folicwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With the mechanical draft ecoling tower incperab e, restere the f7 d*f l b.
cooling tower to' 0PERABLE status within'Whou'rs, or be in at least -
HOT STANDBY within the next 6 hcurs and in COLD SHUTDCWN within the folicwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Mith -th: pert 251 e trwer --'2krup ep..,. 3,piS;;7 ,4, a -inuwai .M c, surtnut -
A ;;;;ga m .~.th_the- I ep :tfc - ; s-,,,*ee7+% ucc m l
-p . . :du re c'-! n - C 2 -S C . 72 -a f ,s . i v ua -u r-un 52nic21-dr:f W opnc i e s--co -ensure- -an--
t-cooling adaqu2te Mupp'y -d d =>p '?te r -te -the -:::
-tou:r 'c. e air.imur. Of-;C afs.
SUR[EILLANCEREQUIREMENTS T-4.7.5 The ultimate heat sink shall be determined OPERABLE:a4-k . wouc r o.. M leut once f er M hats by ;
.A,.* /).2 * -5curr hy Yerifying the water level in the service water pu=pheuse to be greater-than:or equal-to 37'- 0" Mean Sea Level,
. a.c or a.ca n e.iAu.s T
M. d 24.-L a b Yerifying f the water in the mechanical draf t cooling 8 E
- tower basin to be greater than or equal to a Nidie of.4X10 35.5
. gallons -and a t -a te=pe ra ture _less .than.oraequal to-70*F-
- k. At Ja.st once. pe n d.tys From Twn< s lo y S.n,t n bee wkR-fflggy wa2.,,e us.m au.s ssxp a.n, e,by venu.a n,,,y e 22m qu qc t:y q-
- g n is p s .c2I4_7-13 u ,n ._
3, , ._
u _j Q
.g 1\ gi ' [p ,
Y6 ._
PLANT SYSTEMS _
e LIMITING ' CONDITION FOR OPERATION CONTINUED 5c2"1, days by .1litartin'Everf 31 daevs bvpg frem the control room each UHS coolin Tf an that
" is required to be operable and cperating each of those fans for at least 15 minutes, i
-d.M1"da7s by'Te~rifying-that-the portable-tower makeup epump-system- s -
.s tore di n -its design cperational-readiness-state,-
i e
ar? . c ,
d;
'e g418 months by verifying automatic actuatien of each cooling tower-L'; fan on a Tcwer Actuation test signal. .
Su -
e e
9 e
9 e
M O
ae amm e e e
-e e **
g e g e se e-4
- p e
i e
e e
e O
m D //Tl).
n*, ,3, Q 'T;
?-
0
. =- <
E a
, Attachment Q4-2 4
- COOLING TOWER MODEL The Cooling Tower model used in this analysis is based on the Cooling Tower i
model used in the SSPSA (Ref. Q4-2). For reference purposes, two figures from the SSPSA are repeated here: Figure 1 is a simplified P&I diagram for the
[ Seabrook Service Water System (including the Cooling Tower). Figure 2 is a diagram showing the Cooling Tower ventilation system.
I j Figure 3 presents the Cooling Tower reliability block diagram used in this analysis. The following changes can be noted from the Cooling Tower reliability block diagram in Figure D.3-6 of the SSPSA:
- 1) The Tower Actuation blocks (IAA and TAB) are deleted since tower operation is assumed to be initiated by operator action.
- 2) Additional blocks (BFA and BFB) have been added to account for possible
' failure of the Cooling Tower due to backflow through the preferred Service Water pump in each train. This failure mode requires that the SW discharge MOV and check valve fail to close when Cooling Tower operation is initiated.
Table 1 identifies the specific components modeled in each block of the Figure 3 block diagram and identifies the possible failure mode (s) of each component. The component identification numbers are those used in Figures 1-3 and 2.
i The system failure equations are constructed to describe failures of three distinct types - independent hardware failures, common cause failures, and maintenance failures. - The total failure rate for the Cooling Tower system is therefore expressed as the sum of three terms:
CT = CTHW + CICC + CTM Each of these terms is discussed below:
Independent Hardware Failures
?
The equations describing independent failures of Cooling Tower system components are based on the development in the -SSPSA (Ref. Q4-2, p. D.3-29ff)
- and the block diagram in Figure 3. The equations are _ as follows
CTHW =
TRAINA*TRAINB + Rl*R2 + L (Total frequency ofJ independent hardware failure) .f,4pr j L =
VL P*T24 + VF P*T24 (Block L louver and fan)
R1 =
FN S + FN R*T24.+ BD 0 + BD T*T24 (Block .R1 fan and damper)
, R2 = R1
- TRAINB = TRAINA + K (Train B equivalent to Train 3 A~with additional fan) d Page 1 J
, ,. - . - , . - . ---e, r --r-, -e1- r+ -
+ .-e, n - -. - . . . . - , n .r. -- -v-,
, - +n, e-
TRAINA = W1 + K + M + VI + BFA (Sum of Blocks in Train A)
=
W1 FN_,S + FN_R*T24 + (Block W1 fan.
POD _,0 + POD _T*24 + FD__IA*T24 + fan damper and fire damper, BD 0 + BD T*T24 + VL P*T24 relief damper and louver)
K = CTFN_,S + CTFN_R*T24 (Block K Cooling Tower fan)
M =
(1-BP S)*P S + (1-BP R)*P R*T24 + (Block M Cooling Tower pump, CV O + CV T*T24 + check valve, MOV1 + 3*MOV,_T*T24 three MOV's)
MOV1 = (1-BM 0)*MOV O (MOV indep. failure on demand)
VI =
2*MV_T*T24 + (Block V1 manual valves, 3*MOV1 + 4*MOV_T*T24 four MOV's)
BFA =
MOVl*CV_p (Block BFA MOV and check valve on preferred A Train SW pump)
Note that consistent with the SSPSA (page D.3-28) failure of the strainers ir not included in blocks VI and V2. This is because it is considered to be negligible in comparison to other hardware failures due to the time available during plugging to effect recovery.
Also note that the terms for pump failure to start, pump failure to run, and MOV failure to open or close reflect use of the beta-factor model to distinguish between independent hardware failures and common cause failures.
The common cause model is described in the next section, following a listing of the data base variables used in the equations above.
Data Base Variables The data used in this system analysis is based on the data analysis in Section 6 of the SSPSA. The mean values are summarized below.
P_,S 3.29E-3 Standby pump - failure on demand (ZIPMSS)
P,_R 3.42E-5 Standby pump - failure to run (ZIPMSR)
BP_,S 1.llE-1 Beta factor for pump failure to start (ZBPSWS)
BP,_R 7.62E-2 Beta factor for pump failure to run (ZBPSWR)
MOV_p 4.30E-3 MOV - failure to open/close on demand (ZIVMOD)
MOV_T 9.27E-8 MOV - transfer open or closed during (ZIVMOT) operation.
BM_p 4.23E-2 Beta factor for MOV failure to open/ (ZBVMOD) close CV_p 2.69E-4 Check valve - failure to operate on (ZIVCOD)
., demand CV_T ' l.04E-8 Check valve - transfer closed / plugged (ZIVCOP)
MV_,T 4.20E-8 Manual valve - transfer open/ closed (ZIVHOT) during operation FN_,S 4.84E-4 Ventilation. fan - failure to start on (ZIFNIS) demand Page 2
FN_,R 7.89E-6 Ventilation fan - failure during (ZIFNIR) operation CTFN_S 2.93E-3 Cooling Tower fan - failure to start (ZIFN2S) on demand CTFN_R 7.89E-6 Cooling Tower fan - failure during operation (ZIFN2R)
POD _0 2.66E-4 Pneumatic damper - failure to transfer to (ZIDA0F) fall position POD _T 2.67E-7 Pneumatic damper - transfer open or shut (ZIDA0T) during operation BD_O 2.69E-4 Backdraft damper - failure to open on demand (ZIDBDD)
BD_,T 1.04E-8 Backdraft damper - transfer closed (ZIDBDT)
FD_,IA 4.20E-8 Fire damper - inadvertent actuation (ZIDFRI)
V F,_P 1.07E-6 Ventilation filter - blockage (ZIFLIP)
VL_,P 1.07E-7 Ventilation louver plugged (ZIFL2P)
T24 24.0 Hours of required operation DRATN1 (input) Maintenance outage duration - one train in maintenance DRATN2 (input) Maintenance outage duration - two trains in maintenance DRATN3 (input) Maintenance outage duration - basin outage Common Cause Failures The common cause model for this system analysis is based on the common cause model developed in the SSPSA (p. D.3-37ff) and has been expanded to be more complete. The common cause model includes Cooling Tower pump failure to start and failure to run and MOV failure to operate on demand.
Common cause failure of the Cooling Tower pump to start and run is expressed as:
PS2 = BP_S
- P_S + BP_R
- P_R
- T24 Where
=
BP_S Beta factor for pump failure to start (fraction of the total failure-to-start failure rate attributable to common cause)
P,_S =
Frequency of pump failure to start on demand
=
BP_,R Beta factor for pump failure to run P_,R =
Frequency of pump failure during operation T24 =
Required time of operation (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
Common cause failure of MOV's is based on considering those MOV's which are required to operate (change position) for successful operation of the Cooling Tower. These MOV's are listed below:
Train A Train B Valve Description and Valve Valve Required Operation
{
V54 V25 Tower Discharge MOV i
j Required to open on demand
.{
l V34 V23 Tower Return MOV Required to open on demand i
i Page 3
Train A Train B Valve Description and Valve Valve Required Operation V20 V19 Service Water Return MOV Required to close on demand V4 V5 SCC Isolation MOV Required to close on demand The following three common cause failure modes are included in the Cooling.
Tower model used in this analysis:
- 1) CCI = 16
- MOV2 (a single common cause valve failure results in loss of both Trains) where:
16 = Number of MOV pairs with one valve in each Train MOV2 =
MOV common cause failure frequency (discussed below)
- 2) CC2 =
6
- MOV2 * (TRAINA + TRAINB)
(a common cause valve failure occurs in one train and an independent hardware failure occurs in the other) where:
6 =
Number of MOV pairs in a single Train TRAINA =
Total failure rate of Train A TRAINB =
Total failure rate of Train B MOV2 =
MOV common cause ' failure frequency
- 3) CC3 =
36
- MOV2
- MOV2 (two common cause valve failures result in loss of both Trains) where:
36 =
Product of 6 possible pairs in each Train ( 6
- 6 )
MOV2 =
MOV common cause failure frequency The frequency of an MOV common cause event is:
MOV2 =
(1/7) * (BM 0) * (MOV 0) where:
BM_p =
Beta factor for MOV failure to operate on demand MOV_p =
Frequency of MOV failure to operate on demand (1/7) =
Factor accounting for the fraction of common cause failure any one valve shares with any other valve (for this set of 8 valves any single valve shares an equal fraction of common cause failure with 7 other valves)
The overall common cause term for the Cooling Tower model is the sum of the -
above expressions for pump and MOV common cause failures. Thus:
CTCC =
PS2 + CCI + CC2 + CC3 Page 4
Maintenance Outages The Cooling Tower maintenance term is comprised of three parts:
CTM =
ONETRN + TWOTRN + BASIN where:
ONETRN = unavailability of tower due to one train in maintenance and failure of the other train TWOTRN = unavailability of tower due to both trains in maintenance BASIN =
unavailability of tower due to tower basin outage The expression for ONETRN is:
ONETRN =
MA_l*TRAINB + MB_l*TRAINA where:
= unavailability of train A due to maintenance MA_,1 (with a single train in maintenance)
=
MB_,1 unavailability of train B due to maintenance (with a single train in maintenance)
The expressions for MA 1 and MA 2 are the product of:
(1) the frequency of maintenance outage for a single component (MFSR),
(2) the duration of the maintenance outage (DRATNI), and (3) the number of components in the train which require maintenance.
i In train A there are assumed to be two components which require maintenance (one cooling tower pump and one cooling tower fan). In train B there are assumed to be three components which require maintenance (one pump and two fans). The equations for MA 1 and MA 2 are:
MA_1
=
MFSR
- DRATN1
- 2 MB_1
=
MFSR
- DRATN1
- 3 The expression for TWOTRN is:
TWOTRN =
MA_2
- MB_2 where:
=
MA_2 unavailability of train A due to maintenance (with both cooling tower trains in maintenance)
=
MB,_2 unavailability of train B due to maintenance (with both cooling tower trains in maintenance)
The expressions for MA 2 and MB 2 are the product of (1) the frequency of maintenance for a single component (MFSR),
(2) the duration of the maintenance outage (DRATN2), and (3) the number of components in the train which require maintenance.
The equations are:
MA_2 = MFSR
- DRATN2
- 2 MB_2 = MFSR
- DRATN2
- 3 Note that unique outage duration variables (DRATNI, DRATN2) are defined in ONETRN and TWOTRN to permit use of different outage times for one train in maintenance and two trains in maintenance.
The expression for BASIN is:
BASIN =
MFSR
- DRATN3 which is the product of the outage frequency and outage duration. Note that a third outage duration variable (DRATN3) is defined to describe the basin outage time.
Page 5
4 Table 1 Components and Component Failure Modes for the Cooling Tower Component Identification Failure Block Component Number Mode (s)
L Intake Louver L-26 Blockage Intake Air Filter F Blockage R1 Exhaust Fan FN-71 Fail to start /run Exhaust Fan DP-68 Fail to open or transfer closed R2 Exhaust Fan FN-70 Fail to start /run 3
Exhaust Fan DP-67 Fail to open or Damper transfer closed W1 Supply Fan FN-64 Fail to start /run 1
Supply Fan DP-66 Fall to open or Damper transfer closed Fire Damper DP-190 Transfer closed Relief Damper DP-64B Fail to open or transfer closed Exhaust Damper L-28 Transfer closed W2 Supply Fan FN-63 Fail to start /run Supply Fan DP-65 Fail to open or Damper transfer closed Fire Damper DP-189 Transfer closed Relief Damper DP-64A Fail to open or transfer closed 4
Exhaust Damper L-27 Transfer closed K1 Tower Fan FN-51A Fail to start /run
- j. K2 Tower Fan FN-51B Fail to start /run K3 Tower Fan 2-FN-51B Fail to start /run
-,, ..w - . - , . ,
4
. Component -
Identification Failure Block Component Number Mode (s)
M Tower Pump P-110A Fail to start /run Check Valve V53 Fail to open or i
transfer closed Tower Discharge V54 Fail to open or MOV transfer closed Tower Bypass V55 Transfer open MOV i-Tower Test Line V56 Transfer open MOV 4
N Tower Pump P-110B Fail to start /run Check Valve V24 Fail to open or transfer closed Tower Discharge V25 Fail to open or MOV transfer closed Tower Bypass V26 Transfer open
. MOV Tower Test Line V27 Transfer open i
MV i
i VI Gate Valve V68 Transfer closed
! Gate Valve V70 Transfer closed Strainer S-10 Blockage i
Service Water V20 Fail to close or Return MOV transfer open Tower Return V34 Fail to open or MOV transfer closed SCC to Tower MOV V74 ' Transfer open i
i J _. , -, .- - _ - ,
i l
l Component i Identification Failure j Block Component Number Mode (s)
V2 Cate Valve V65 Transfer closed Gate Valve V67 Transfer closed Strainer S-Il Blockage Service Water Vl9 Fail to open or Return MOV transfer closed Tower Return V23 Fail to open or MOV transfer closed SCC to Tower MOV V76 Transfer open SCC Isolation V5 Fail to close or MOV transfer open BFA SW Pump Discharge V2 Failure to close j MOV - Train A SW Pump Discharge VI Failure to close Check Valve BFB SW Pump Discharge V29 Failure to close MOV - Train B SW Pump Discharge V28 Failure to close Check Valve i
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'l Figure 3 Reliability Block Diagram for the Cooling Tower i
Train A
-- R1 __
WI K1 M VI BFB L
W2 K2 K3 N V2 BFB -
R2 -
h Train B L Intake Air Louver R1 Pump Area Ventilation R2 Pump Area Ventilation
, WI Tower Switchgear Ventilation W2 Tower Switchgear Ventilation M Train A Pump Section N Train B Pump Section VI Train A Discharge Valves V2 Train B Discharge Valves K1 Train A Cooling Tower Fan K2 Train B Cooling Tower Fan 1 K3 Train B Cooling Tower Fan 2 BFA Train A SW Pump Backflow (Failure of SW Pump Discharge Valves VI and V2)
BFB Train B SW Pump Backflow (Failure of SW Pump Discharge Valves V28 and V29) 7 .
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-ee
Q.7 NRC REQUEST 3.8.1.1 Electric Power Systems: The request for relaxed testing schedule for diesel generators refers to Generic Letter as a basis for the change.
Also cited in this Generic Letter is a need for a reliability assurance program to maintain and improve the reliability of DGo. Provide a description of the reliability program you will implement in consideration of Generic Letter 84-15.
RESPONSE
i The Seabrook -Station Diesel Generator Reliability Program is being i developed and will be available for inspection when it is finalized. It l will include the following elements:
- 1. Periodic surveillance tests, including routine tests and cold start
, tests, i 2. Performance monitoring, including root. cause failure analysis and trending and analysis of failures and key performance parameters,
- 3. Maintenance, minor and major overhaul,
- 4. Review of relevant industry experience,
- 5. Data and program review, at least once per refueling outage,
- 6. Reporting requirements -
o failures to NPRDS o data and program review to Station Manager
- o annual report of diesel performance to NRC, and l 7. Performance goals.
! This program is being based on recommendations from the Vendor Manual as '
well as relevant information from Generic Letter 84-15.
J L
i i
Q7-1
- - .- . - _ .