ML20087J884
| ML20087J884 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/29/1984 |
| From: | Papazoglou I, Youngblood R BROOKHAVEN NATIONAL LABORATORY |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| BNL-NUREG-51721, NUREG-CR-3529, NUDOCS 8403230133 | |
| Download: ML20087J884 (102) | |
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NUREGICR-3529 BNL-NUREG-51721 Review of the Arkansas l\uclear One Generating Station Unit No.1 Emergency Feedwater System Reliability Analysis Prepared by R. Youngblood, I. A. Papazoglou Brookhaven National Laboratory Regulatory i
88 885 J 888sfa P PDR
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NUREG/CR-3529 BNL-NUREG-51721 Review of the Arkansas Nuclear One Generating Station Unit No.1 Emergency Feedwater System Reliability Analysis Manuscript Completed: September 1983 Date Published: February 1984 Y ngblood, I. A. Papazoglou Brookhaven National Laboratory Upton, NY 11973 Prepared for Division of Safety Technology Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
. Washington, D.C. 20555 NRC FIN A3393
ABSTRACT The purposes of this report are: (1) to review the Emergency Feedwater System Upgrade Reliability Analysis for the Arkansas Nuclear One Nuclear Generating Station Unit No. 1, and (2) to estimate the probability that the Emergency Feedwater System will not perform its mission for each of three different initiators: (1) loss of main feedwater with offsite power available, (2) loss of offsite power, (3) loss of all 4160 VAC power. The scope, methodology, and failure data are prescribed by NUREG-0611, Appendix III.
1 iii
TABLE OF CONTENTS Page
- ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii LIST OF FIGURES ............................ vi LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi
SUMMARY
................................ vii l 1. INTRODUCTION ............................ 1
- 2. SCOPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
- 3. SYSTEM MISSION ........................... 4 3.1 Mission and Success Criteria . . . . . . . . . . . . . . . . . . 4
- 4. SYSTEM DESCRIPTION ......................... 6 4.1 Pumps ............................. 6 4.2 Support Systems ........................ 6 4.3 Suction ............................ 6 4.4 Discharge Paths ........................ 6 4.5 Actuation and Control Logic .................. 6 4.6 Steam Supply to the STD Pump Turbine . . . . . . . . . . . . . . 7 4.7 Reci rcul ati on and Test Lines . . . . . . . . . . . . . . . . . . 7
- 5. QUALITATIVE RELIABILITY ANALYSIS .................. 9 5.1 Single Point Failures ..................... 9 5.2 Common Cause Failures ..................... 9 5.2.1 Degraded Failure .................... 10 5.2.2 Common Cause Maintenance Errors ............. 11
- 6. QUANTITATIVE RELIABILITY ANALYSIS . . . . . . . . . . . . . . . . . . 12 6.1 Limi tati ons of Methodol ogy . . . . . . . . . . . . . . . . . . . 12 6.2 Approach of the AP&L Study vs. Approach of the BNL Review ... 12 6.3 Assumptions .......................... 13 6.4 Dominant Failure Modes . . . . . . . . . . . . . . . . . . . . . 14 6.5 Comments on Failure Probabilities Used in the BNL Review . . . . 14
- 7.
SUMMARY
AND CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . 17 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 APPENDIX A: IMPACT OF DESIGN CHANGES ON SYSTEM AVAILABILITY . . . . . . 29 APPENDIX B: SUBMITTAL FROM AP&L TO NRC - AUGUST 11, 1983 ....... 41 l
l l
v
LIST OF FIGURES
__ Figure #_ Title Page
- 1. Arkansas Nuclear One Unit 1 Emergency Feedwater System ......................... 19 LIST OF TABLES Table # Title Page 6 la Dominant Contributors to Unavailability of Train 1 . . . . . . . . . . . . . . . . . . . . . . . . . 20 6.lb Dominant Contributors to Unavailability of Train 2 . . . . . . . . . . . . . . . . . . . . . . . . . 23 6.2 Results for Q1 and Q2.................. 26 6.3 Other Contributions to System Unavailability ...... 27 7.1 EFWS Unavailability . . . . . . . . . . . . . . . . . . . 28 A.la Changes to Table 6.la in New Design . . . . . . . . . . . 31 A.lb Changes to Table 6.lb in New Design . . . . . . . . . . . 35 A.2 Changes to lable 6.2 in New Design ........... 38 A.3 Changes to Table 6.3 in New Design ........... 39 A.4 Changes to Table 7.1 in New Design ........... 40 vi
SUMMARY
The probability of failure of the Arkansas tluclear One Nuclear Generating Station Unit No.1 Emergency Feedwater System (l) on demand, under different loss of main feedwater conditions and different success critgr a, has been evaluated using methodology and data put forth in NUP.EG-061152 The re-i sults are as follows:
I Initiator Probability of Failure to:
Avoid Supply EFW Dryout within 20 minutes
- 2. Loss of Offsite Power (LOOP) 1.6x10-3 1.5x10-3
- 3. Loss of all AC(LOAC) 2.8x10-2 2.7x10-2 The column " Avoid Dryout" gives the probability per demand of failure to prevent dryout, given each of the three initiators. These results have been obtained under the assumption that there is no time for any recovery action; thus, the calculations do not include any credit for operator action to recover from malfunctions or maintenance errors. The column " Supply EFW within 20 minutes" gives the probability of failure to deliver emergency feedwater within 20 minutes, given each of the initiators. This mission suc-cess requirement is less restrictive as far as the available time for recovery is concerned and, therefore, credit h.as been given for operator actions.
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- 1. INTRODUCTION The rposes of this study are: 1) to review and evaluate a Reliability Analysis ( of the Emergency Feedwater System (EFW) of the Arkansas Nuclear One Nuclear Generating Station Unit No. 1 (ANO-1) prepared by Babcock & Wilcox (B&W) for Arkansas Power & Light (AP&L) and submitted to the Nuclear Regulatory Commission (NRC); and 2) to perform an independent reliability analysis of he EFW system using methodology and data put forth in NUREG-0611(2 Af ter the accident at Three Mile Island, a study was performed of the Auxiliary Feedwater Systems (AFWS) of all then-operating plants. The results obtained for NUREG-0611(2) operating Westinghouse-designed plants were presented in At that time, the objective was to compare Auxiliary Feedwater System (AFWS) designs ; accordingly, generic failure probabilities were used in the analysis, rather than plant-specific data. Some of these generic data were presented in NUREG-0611. The probability that the AFWS would fail to perform its mission on demand was estimated for three initiating events: LMFW, LOOP, and LOAC.
Since then, each applicant for an operating license has been required (3) to submit a reliability analysis of the plant's AFWS, carried out in a manner similar to that employed in the NUREG-0611 study. In addition, some operating plants have also submitted reliability analyses of upgraded Auxiliary Feedwater Systems. Rec has been defined by NRC ly, a quantitative criterion for AFWS reliability in the New Standard Review Flan (SRP).
"...AnaccegtableAFWSshouldhaveanunreliabilityintherangeof 10-4 to 10- per demand based on an analysis using methods and data presented in NUREG-0611 and NUREG-0635. Compensating factors such as other methods of accomplishing the safety functions of the AFWS or other reliable methods for cooling the reactor core during abnormal conditions may be considered to justify a larger un-availability of the AFWS."
The objective of the present study is, therefore, to analyze the reliability of the ANO-1 EFWS, using methodology and data presented in NUREG-0611, in order to facilitate and supplement the qualitative review of the system design.
The report is organized as follows: Section 2 presents the scope of the present study.
Section 3 discusses the mission success criteria for the EFWS and highlights the important differences between the definition of EFWS success for B & W plants and Westinghouse plants.
of the analysis in NUREG-0611. Section 4 describesThe latter were the subject the basic configuration and characteristics of the EFWS. Section 5 discusses the qualitative aspects of the reliability of the system and presents the dominant contribution to tne unavailability. Section 6 presents the quantitative analysis and compares the results and approaches used in the AP&L study with those of this study.
Finally, Section 7 summarizes the results.
The original draft of the present report was based on Reference 1. After the original draf t of the present report was completed, substantial desi9n changes were made. These are summarized here in Appendix B, which contains AP&L responses to BNL questions. No new analysis was submitted to BNL for review. In keeping with the original objective of reviewing Ref.1, Sections 2 through 6 of this report continue to reflect the design of Ref.1; in Appendix A of this report, the effect of the changes on the analysis is presented, and in the Summary of this report (Section 7), results are presented for both the old and new designs.
- 2. SCOPE The scope of 111 of NUREG-0611(the1). reliability analysis of the AFWS is defined in APPENDIX In the present study, the probability that the EFWS will not perform its mission on demand is calculated for two mission definitions and three types of demands. The two mission success definitions were necessary because of the substantial dif ferences between the B & W plants and the Westinghouse plants.
The success criteria stipulated by NUREG-0611 were based on the Westinghouse design. For the B & W plants the following two success criteria are con-sidered:
- 1) Avoid dryout of the steam generators; and
- 2) Supply the steam generators with EFW within 20 minutes of the demand.
The failure to supply water to make up for a loss of main feedwater is es-timated for different conditions, namely,
- 1) Loss of Main Feedwater without Loss of Offsite Power (LMFW).
Since one purpose of this analysis is to assess the important characteristics of the EFWS design configuration, detailed modeling of such support systems as electrical power (both AC and DC), service water, instrument air etc. was not performed. Such an undertaking is beyond the stated scope of the BNL reviews. The goal is a rudimentary understanding of the properties of the design of the EFWS. To this end, standardized data are used wherever they are available (emergency AC is schematically represented by the diesel generators) or assumed to be available (e.g., DC power is assumed available), although in some cases where such a system is seen to introduce a common cause mechanism, this is pointed out. Moreover, such phenomena as earthquakes, fires, and floods are excluded, along with high energy line breaks. The emphasis in NUREG-0611 is on independent hardware failures within the AFWS itself; dependent failures are analyzed if they arise from shared ha rdwa re. Depth and resolution of the studies are set by NUREG-0611.
The quantity calculated here is the unavailability of the EFW system due to fluid system failures, maintenance acts, human errors, and failure to initiate, with mission success defined below (Section 3). " Unavailability" means (as in NUREG-0611) "the probability per demand that the system will fail to perform its mission". The AP&L report deals with a number of ad-ditional failure modes, including steam generator overfill due to EFW control failures, spurious isolation of the steam generators, etc. These failures lie beyond the scope of NUREG-0611 and, therefore, are beyond the scope of this report.
- 3. SYSTEM MISSION 3.1 Mission and Success Criteria In the AP&L reliability study (l), mission success is defined as attainment of adequate flow from at least one pump to at least one steam generator. In reference ll, minimum acceptable flow is given as 500 gpm at a SG pressure of 1050 psig, and the maximum acceptable time for achievement of this flow is given as 50 seconds from the time an initiation signal is given.
There are two time scales to consider. One is a 20-minute deadline for EFWS initiation, which refers to a point up to which EFWS initiation can prev-ent core damage. According to NUREG-0667(5), "For plants with reactors de-signed by B&W, analyses prepared by B&W concluded that a period of approx-imately 20 minutes is available after loss of all feedwater for the operator to either (1) restore feedwater (either auxiliary or main), or (2) start the HPI pumps and enter into a " feed-and-bleed" method of core cooling. This available time is consistent with independent staff analyses". [page 5-32 of Ref. 5]
On the other hand, it is acknowledged that "...B&W-designed 177-FA plants show some unique levels of sensitivity in their response and recovery from anticipated transients involving overcooling and undercooling events as well as small-break loss of coolant accidents. The recovery from such events has of ten led to undesirable challenges to engineered safety features (ESF) sys-tems. This sensitivity stems mainly from the small heat sink resulting from the operation of the once-through steam generator (OTSG), which is an inherent design feature of the B&W reactor plants ... [page 2-1 of Ref. 5] ... B&W plants place a premium on the reliability with which the auxiliary feedwater starts are properly timed. The penalty for late starts is an increased likelihood of transient-induced LOCA". [p. 7-8]
Therefore, for each initiator (LMFW, LOOP, and LOAC), failure probabilities are calculated for each of two mission success criteria:
- 1) Failure to deliver flow from at least one pump to at least one steam generator, with no credit for operator action;
- 2) Failure to deliver flow from at least one pump to at least one steam generator within 20 minutes, allowing credit for some operator actions within this time.
The top event for NUREG-0611 purposes is steam generator dryout. It is considered that dryout occurs on a sufficiently short time scale that no credit for operator actions is warranted. Thus, for NUREG-0611 purposes, mission criterion (1) above is applied.
Loss of main feedwater is assumed by 'the utility to lead occasionally to momentary loss of SG level, i. e., the SG level becomes zero although EFW flow is entering from the top of each SG and being vaporized. This is discussed in the response to Comment 1.b in Enclosure 2 of a transmittal from D. C. Trimble
l (AP&L) to R. W. Reid (NRC) dated March 12,1981(6). Since heat is being
! carried away, it is clearly inappropriate to consider this in the same i category as a boil-dry resulting from no EFW flow. For purposes of this analysis, then, this has been considered a success state of the system, since this is its design basis.
The utility's report raises the question of whether excessive recircu-lation capacity can degrade the flow of one pump sufficiently that such an event should be considered system failure. In Ref. 6, it is stated that the i
EFW pumps are sized for a decay heat load corresponding to approximately 3 1/2% of full power. If degradation of the specified flow (500gpm) by the amount corresponding to excessive recirculation (78 gpm) can be interpreted as reducing the capacity to correspond to 3% of full power, which is the naive
] conclusion suggested by comparison of the flow rates, then the time elapsed after trip before decay heat drops below pump capacity is almost doubled, i going from approximately 5 minutes to approximately 9 minutes. It is not j immediately clear whether this leads to a PORV challenge, which (cf. p. III-10 i
of FUREG-0611) for purposes of this analysis is considered to imply the top event. Comments on this scenario are given in Section 5.2.1.
i 4. SYSTEM DESCRIPTION Summarized below is pertinent information contained in Refs. (1), (6),
and (7).
4.1 Pumps There are two pumps, one steam turbine driven (STO) and one electric motor driven (EMD). Each is capable of providing the minimum required flow of 500 gpm at 1050 psig within 50 seconds (see Figure 1).
i 4.2 Support Systems Apart from power, neither pump directly requires support systems. Both pumps and the turbine are lubricated by slinging oil from reservoirs near the bearings. Cooling is accomplished by the transfer of heat to the sur-roundings; the character of the surroundings is not spelled out in the report, so that it is not clear whether a hostile environment is likely to occur.
4.3 Suction The two pumps take suction from the condensate storage tank through a common line in which there are a locked open manual valve (CS-19) and two check valves (C5-98 and CS-99), all in series. The tank is required by Technical Specifications to contain 107,000 gallons for EFWS use. The backup source is nuclear service water; each pump can be connected to one train of service water by manual action. Low suction pressure on the common header is alarmed; the operator then operates MOVs to switch suction for each pump. The valves that control suction of the STD pump are DC powered from battery-backed busses.
4.4 Discharge Paths Each pump is connected to each steam generator by an independent path.
The paths branch immediately beyond a check valve downstream of each pump.
There is no pump isolation valve between the pump and the branch point, so that isolating the pump requires isolating both paths. Each path contains two MOVs and a check valve; one MOV provides flow control, while the other is associated with the logic that isolates a faulted steam generator. The valves that control flow from the STO pump are powered by DC from battery-backed busses. The valves that control flow from the EMD pump require AC.
4.5 Actuation and Control Logic The EFWS is initiated by the " Initiate Logic" portion of the " Emergency Feedwater Initiation and Control" (EFIC) system when any of the following occurs: a) all 4 RC pumps are f *ipped, b) both main feedwater pumps trip, c) t low level occurs in either SG, f j low pressure occurs in either SG, e) flux to l
feedwater trip is present.
There are two channels of initiate logic (A and B). At ANO-1, both channels are connected to both pumps. (The system diagram shows only channel l
i A connected to the EMD pump, but the fault tree shows both channels connected to both pumps. It is assumed here that the fault tree is correct.) Thus, failure of one channel of logic alone has no effect on the system; the other channel actuates both pumps. This contributes to system reliability in this analysis (the possibility of interaction degrading both channels is beyond the scope of this report). Failure of both channels fails automatic initiation of both pumps; this event is listed separately from contributors to Q1 and Q2 (Table 6.3).
- 4.6 Steam Supply to the STD Pump Turbine I Steam passes from both SG's through check valves and normally-open MOV's into a common line. Either of two normally-closed M0V's (CVY-1 and CVY-2) in parallel can open to admit steam to the turbine. These M0V's are each i connected to one of the two channels of EFIC actuation logic. Steam then ~
)
passes through a pressure reducing station consisting of two pressure reducing i valves in parallel (CVY-3 and CVY-4). Downstream of this station, there are relief valves to protect the turbine in the event of failure of a pressure
- reducing valve in the open position.
l The fault tree assumes that failure of a pressure reducing valve will not fail the system unless one of the relief valves subsequently sticks open. This is equivalent to assuming that the relief valves can open and close in a manner that operates the turbine successfully without either leading to an overspeed trip or a lack of steam. It appears that this is a highly non-conservative assumption. Here, failure of either pressure reducing valve is assumed to lead to failure of the STD pump. This is not given a very high probability, however, since the closest compqrable component tabulated in l NUREG-0611 is an air-operated valve (3 x 10 '+/D). Miscalibration of the pressure would be given a 10-3 probability by NUREG-0611, and it is probably appropriate to include in this figure the effects of diaphragm degradation.
The redundancy of the steam admission system leads to low probability of failure but a relatively high probability of maintenance, at least on the as-i sumptions used here. There is also considerable scope for maintenance errors.
Events have occurred at ANO-1 in which the STD pump failed for reasons ap-parently associated with the buildup of condensate in the steam line because of errors in isolating the condensate traps.
4.7 Recirculation and Test Lines For each pump, there is a full flow test line which is normally closed and a minimum flow line which is normally open (15 gpm). In addition, there is a common recirculation path which is closed under conditions of high flow to the SG's and open under conditions of low flow to the SG's (78 gpm). This is discussed more fully under " Degraded Failure" (Sec. 5.2.1).
A total loss of pump recirculation flow could be caused by blockag' of either FW14, FW15, or the three-way valve immediately upstream of these. If total loss of recirculation can fail both pumps, then these are single failure i
i v--,-- , , , , _ - - ,--,,,-.----,r e-w-----+ - - - -
- - - - , - y --- -
-we - - - - - - ~ - - - - - - - - - - - - -
points of the system. It is not clear that these will be conditions under which minimum flow requirements will not be satisfied by system demand; but from the given description of EFIC control logic, such an event is at least a strong possibility. Actual data on the time dependence of auxiliary feedwater demand are sparse; but the event at TMI-2, which is documented in some detail, is suggestive. EFW control valves did not open for some 35 seconds after trip, because this was the elapsed interval before the SG 1evel setpoint was
.eached. The block valves, of course, remained closed for some extra time, which also would certainly contribute to a need for minimum flow protection.
A delay on the order of minutes before EFW flow is demanded would appear to be credible, ard minimum flow protection would appear to be necessary for such conditions. 10-4 has therefore been assessed for each of the three single-point recirculation-blockage events. This valve is probably conservative for the manual valve and the three-way valve, but not for the check valve.
Flow diversion through open test lines is considered to be a contributor to pump failure. Flow diversion through the spurious opening of the extra-capacity recirculation lines is considered in Section 5.2.1.
l 1
l
i i
- 5. QUALITATIVE RELIABILITY ANALYSIS t
! There are two trains in the EFWS (see Figure 1). Each train consists of j one pump and a pair of discharge paths, one to each steam generator. Mission
- success is defined as adequate flow from one pump to one steam generator within the appropriate time (see Section 3). Therefore, failure of the EFWS
! is:
i Failure of both trains = (Q 1 + 12 A 8 ) * (92+AB) 21 l = Q129 + Q A B121+QAB2i2+AA12 BBi2 i
where Qi = inadequate flow out of FW-10B (STD train)
Q2 = inadequate flow out of FW-10A (EMD train)
! A1 = flow path from FW-10B to SGA blocked I A2 = flow path from FW-10B to SGB blocked f B1 = flow path from FW-10A to SGA blocked B2 = flow path from FW-10A to SGB blocked The dominant system cut sets are contained in Q 12Q . Most of these events are double failures, products of large single contributors to Q}
multiplying large single contributors to Q 2. These will be discussed in Section 6.
5.1 Single Point Failures Event Q Q1 2 contains single failures: blockage of the valves in the lines to and from the CST. These valves are implicitly tested every time either of the pumps is tested (bi-weekly). Nevertheless, 10-4 per demand is assessed for blo::kage. For some of these valves, this is arguably a ccmservatism; on the other hand, were these not single failure points, their blockage would not be an important issue.
- The valves on the suction side that are considered single failure points are CS98, CS99, and CS19. On the recirculation side, FW14, FW15, and the 1 upstream three-way valve have been quantified as single-failure points assuming that loss of recirculation can fail the system. See Section 4.7 for a fuller discussion of this.
j 5.2 Common Cause Failures 4
Event A A 8 81 2 1 2 contains at least one double event; both paths entering SGA converge at a single check valve, and similarly for SGB, so that a dcuble check valve blockage fails the systen. In the absence of common interaction, this is not a significant contributor. Apart from this, the seening redundancy of the flowpaths suggests that they contribute only to higher-order cut sets (3 or more failures, e.g., Q i21 A B ). This is true i
4 1
-_,y u-y - - - , . - - , +
,--.y e , -.
r._.r_ - -.,,,-_-.4
i only if there are no commonalities. There are examples in the LER files of events which bear on this question. At Arkansas Nuclear One, a maintenanca error disabled two flowpaths on one occasion (4/6/80), and on another occasion 5
(5/22/79) the unexplained lifting of cable leads disabled two paths. Thus, system failures involving flow paths may not be wholly negligible, especially
' those involving coupled maintenance errors. However, these errors are unlikely i to dominate the other failures assessed here. ,
In a broader review, it would be appropriate to inquire whether common cause f ailures (e.g., miscalibrations) in the control logic itself could cause i the loss of multiple flow paths. This is considered to be beyond the scope of the present reviews.
1 The four control valves depend on electric power (DC or AC), and fail as is.
Loss of DC is beyond the scope of this report. Loss of AC fails the EMD pump, so that its valve positions do not matter except insofar as these influence the " degraded failure" (Sec. 5.2.1).
5.2.1 Degraded Failures A minimum flow line is always open from each pump, allowing 15 gpm to recirculate back to the condensate storage tank. When no flow is called for by the control system, an additional recirculation path opens up. This is controlled by valves CV2815 and CV2816, which are DC operated MOVs. These valves respond to the prevailing conditions on the four discharge paths:
either the states of valves CVX-1, CVX-2, CVX-3, CVX-4, CV2620, CV2626, CV2627, and CV2670, if the system diagram is correct, or low flow in these lines as sensed by the flow elements upstream of CVX-1, CVX-2, CVX-3, and CVX-4, if the system description is correct (page D-3). The concern expressed in the report is that given loss of AC, CVX-2 and CVX-3 will not open; the system, thinking that the EMD pump needs more recirculation capacity, will open CV2815 and CV2816; this will degrade flow from the STD pump, which is the j only pump operating given LOAC. There is some question whether this should be considered to lead to the top event (e.g., a PORV challenge). The report states that the matter should be investigated further.
Deciding whether this event implies the top event is beyond the scope of this report. However, comments are offered below on its probability, which is considered separately from the probability of the other (unambiguous) failures.
As indicated above, there is ambiguity concerning which physical parameter controls the operation of CV2815 and CV2816. Let us assume at first that flow is the controlling parameter. Then no flow through either discharge line of a given pump causes recirculation to open up. This means that, in the language introduced in section. 5, the event leading to this degraded failure is Q1 + Q2 + A1A2 + B1B2 . Maintenance on P7B, for example, degrades the flow l from P7A, and vice versa.
if valve positions are the controlling parameter, the conclusion changes somewhat. Failure of P7A does not automatically cause recirculation to open l
1 I
_ _=
I
'l a
up if P7A's discharge paths are open. However, maintenance on P7A will still presumably cause the problem, because the discharge paths are presumably isolated for maintenance.
i If flow is the controlling parameter, then from Tables 6.1-6.2, one arrives at the conclusion that the probability of a degraded failure is 7x10-2 per LMFW. If valve positions are the controlling parameter, then the probability of a degraded failure is a factor of two to three less. l 5.2.2 Common Cause Maintenance Errors If it were necessary to isolate CV2802 for maintenance, P7B would probably be deprived of its normal suction source. In such an event, P78 would need to be aligned to service water. An error in performing this alignment would leave the entire system unavailable for the duration of the maintenance act. Depending on the details of the procedure, the probability of leaving P7B without suction for any of several reasons could well be of order 10-2; if maintenance on CV2802 is 2x10-3 system unavailability would be of order 2x10-5., then Suchthe contribution contributions to be would regarded as significant; however, they have not been included here because the information available does not support their. inclusion. Moreover, the figure being used for valve maintenance is probably conservative (see Table 6.la).
Similar remarks apply to leaving the wrong pair of flow paths disabled for pump maintenance. Depending on the actual procedures and on the actual plant layout, such an event could contribute, but more information would be required to justify its assessment.
An event qualitatively resembling the above occurred at Connecticut Yankee on 6/16/81. While one pump was out for maintenance, an attempt was made to isolate it by closing a valve on the recirculation line of the other pump.
i i
l l
- 6. QUANTITATIVE RELIABILITY ANALYSIS 6.1 Limitations of the Reliability Analysis The significarce of the point estimates obtained in this review is best illuminated by the following quotation from NUREG-0611 (page III-19):
"The data was applied to the various identified faults in the fault logic structure and a point value es-timate was determined for the top fault event (i.e.,
AFW System unavailability). Such an approach is con-sidered adequate to gain those engineering and re-liability based insights sought for this AFW System reassessment. As noted, no attempt was made to intro-doce the somewhat time consuming, calculational ele-gance, associated with the process of error propaga-tion into this assessment (e.g., Monte Carlo). Prior experience with such a calculational process has re-vealed a somewhat predictable outcome that, even with i the very redundant systems, could be slightly higher than the point value solution (e.g., factor of ap-proximately three times higher than the point value and usually less). Should there exist a clearly over-whelming fault in a systems design, then the process of error propagation would be expected to be merely one of higher elegance and it would yield no important change to the quantitative solution".
Clear cut dependencies or commonalities have been sought in the analysis, but parametric modeling of common cause failures (e.g., beta factor treatments) has been considered to be beyond the scope of this report.
6.2 Approach of the AP&L Stuay vs. Approach of the BNL Review
< Data Base: The AP&L study does not give details of basic event probabilities. According to Paragraph 2.3 of the study, generic data wer.e obtained and then made plant-specific to AN0-1 by incorporating AN0-1 experience. The approach taken in the provided in NUREG-0611 and WASH-1400(7)present analysis wherever these exist, and supplement is to use them as necessary from IEEE sources (9).
Unavailability for Different Initiators: The FPC study does not present unavaTlability given LMFW, unavailability given LOOP, and unavailability given LOAC; it presents unavailability averaged over the initiators. In this re-view, unavailability is calculated for the three different initiators.
Level of Detail in the Fault Tree- The AP&L fault trees goNUREG-0611, into con-on siderable detail in treating the actua ion and control logic.
the other hand, simply assigns 7 x 10- as the failure probability of each i
channel. Although some qualitative information is provided about the details of the actuttion and control logic, its relative newness and our lack of
l t
information concerning its details prevent us from offering a meaningful independent estimate of its failure probability within the scope of this re-view. Accordingly, we adopt the 7 x 10-3 figure given in NUREG-0611.
Mission Success Criterion: The AP&L study provides conclusions for un-availability given no operator intervention for 20 minutes, and unavailability given credit for operator intervention. This is a useful distinction, which will be observed here (see Section 3).
, Failure of normal suction: Failure of locked open manual valve CS-19 to remain open does not appear on the AP8L f ault tree. Flow blockage of check valves CS-98 and CS-99 does appear on the AP&L fault tree. All three events are included here.
6.3 Assumptions Valve Maintenance: NUREG-0611 indicates that valve maintenance should be assessed. In some studies, this has been done (notably the RSSMAP study (10) of Oconee). WASH-1400 indicates (Page 111-40) that maintenance on valves should be (m essed, but the only important contributor showing up in Table II 5-9 of WASH-1400 is maintenance on the steam admission valve. WASH-1400 acknowledges that maintenance is performed on the MOV's in the AFWS, but in that system the multiplicity of flow paths is such that these contributions to system unavailability are negligible. Here, the difficulty of isolating certain valves causes valve maintenance to contribute in spite of the seeming redundancy of flowpaths. Consider maintenance on CVX-1 (see Figure 1). In order to isolate this valve from the high-pressure discharge of P7A, it is necessary to disable P7A. From an unavailability point of view, then, maintenance on CVX-1 is logically equivalent to maintenance on P7A. A similar remark applies to CVX-4 for P7A, and CVX-3, CVX-2 for P7B. These contributions are tabulated in Q1M and Q2M (Table 6.1).
Maintenance on the MOVs controlling suction to each pump entails dis-abling one pump, and possibly realigning the other pump to service water. If these acts occur, they give rise to cut sets in which one train is out for maintenance and the other train is inadvertently disabled (e.g., incorrectly aligned to its alternate source). These somewhat hypothetical events are dis-cussed separately in Section 5.2.2.
Fault Duration of Maintenance Errors: Since testing is performed monthly while maintenance is perfonned every 4.5 months, it can be asked whether a given maintenance error (e.g., leaving a valve unrestored after maintenance) should be assumed to persist over the entire period between maintenance acts.
For example, if leaving a certain valve unrestored after maintenance may be '
assigned a probability of 5x10-3, but this error is discovered at the next monthly test then the average unavailability of the flow path due to this er-ror is 5x10-$/4.5. However, the argument becomes complicated if maintenance of other components is factored in: e.g., pump maintenance might occasion a given error in January, May,'etc., while maintenance on an associated downstream valve might occasion the same error in February, June, etc. In other words, uncorrelated maintenance acts on different components tend 'to wash out the effect of recovery, by decreasing-the period between op
?
portunities to coramit the error. As more such opportunities are factored in, the unavailability due to the error approaches the value obtained with no credit for recovery at all. Here, the basic error rate is used with no credit for discovery by subsequent testing. A guaranteed operability test after ev.
l ery maintenance act would have a substantial impact in reducing some errors,
- but in this system, the benefits of testing are limited by the fact that the operating configuration is substantially different from that used for tes-l ting.
6.4 Dominant Failure Modes I Singles: For scenarios in which no operator intervention is considered, i
the failure of CS-19 to renain open (1 x 10-4) is a dominant failure. This does not appear on the AP&L fault tree. The value of 10-4 is dictated by NUREG-0611. Some would consider this conservative; on the other hand, were this not a single failure point, it would not contribute much to the system unavailability. The point is that suction switchover is manual and relatively
- involved, and failure to diagnose loss of suction can quickly lead to damage of the pumps. Similar remarks apply to CS98 and CS99. An analogous situation ;
is created by blockages of the recirculation path. These are discussed in i
- Section 4.7.
In Table 6.3, single failures contribute 6 x 10-4, which is 10-4 each
- for CS-19,CS-98, and CS-99, FW14, FW15, and the upstream three-way valve (see Section 4.7).
Doubles: The important doubles are contributions to Q1 multiplying contributions to Q2, where Q1 = unavailability of the EMD train from CV2800 to FW-10A, inclusive, and Q2 = unavailability of the STD train from CV2802 to FW-10B, inclusive.
(There may be doubles arising from commonalities between pairs of flow paths l which are isolable by a given logic channel.)
I Contributions to Q1 and Q2 are given in Table 6.1.
6.5 Comments on Failure Probabilities Used in the BNL Review The data used by AP&L were not made available to BNL. Evidently, they I were plant-specific data. The data used here have been derived from l NUREG-0611 wherever practicable.
Maintenance and Test unavailability: NUREG-0611 effectively prescribes these numbers for pumps and valves. They have been assessed here wherever they can be assessed consistently with reasonable operating practice. An effort has been made to account for components whose states are altered by maintenance acts on other components (e.g., valve closures or disablings performed to isolate the component being maintained).
l l
l . -- --- . - . . . - - ..
i Human Error Probabilities: NUREG-0611 gives substantial credit for valve position indication in the control room, which ANO-1 has for many of the i
valves in the EFWS. Thus, the probability of leaviag a suction valve in the l wrong position after maintenance is 5x10-4 from Table III-2 of NUREG-0611.
There is no pump isolation valve downstream of the pumps. Thus, during maintenance, pumps must be isolated from the steam generators by closure of l the flow control valves. The error of leaving a block valve closed in another plant corresponds here to leaving the flow control valves in the discharge paths closed and disabled.
Failure to restore operability of steam to the turbine after maintenance was also included here . An event apparently of this type occurred at Farley on 3/25/78. It would appear that this failure should be considered on a par with failure of suction valve restoration. Since the steam admission valves are normally closed anyhow, credit for their position indication is superfluous; the value adopted here is that corresponding to failure to restore a suction valve without position indication.
This is not a substitute for a detailed human error analysis; it is simply an attempt to be consistent with the scope and methodology of the rest of the analysis without ignoring previously untabulated failure modes. It is not clear that this is a conservatism; several failures of the turbine pump have occurred at Arkansas Nuclear One which, though not precisely of this type, involve degradation of the steam supply because of maintenance errors.
Recovery Factors: Where recovery of a failure is practicable within 20 minutes, substantial credit for such recovery gffectively removes that failure from the list of contributors. Example: 5x10-'+ for an unrestored discharge valve drops to the 10-6 to 10-5 range when 20-minute recovery is taken into account, which makes it relatively insignifica contributorstoQ1HandQ2H,whichareoforder10gtcomparedwithother
. Substantial credit for recovery is appropriate on a 20-minute time scale. However, some events must be recovered much more quickly. For example, the pumps do not trip on loss of suction, and pump damage is expected within a few minutes if
'; suction is lost. Restoration of suction at 19 minutes is therefore superfluous (the pumps are presumed damaged).
It should be borne in mind that the recovery factor being considered is not simply " failure to diagnose within 2 minutes and promptly correct a closed suction valve"; rather, it is this failure given that the other train of the EFWS has also failed, and that the initiating event might have been, for example, a loss of offsite power. In other words, there are many claims on the operator's attention. Swain and Guttmann (NUREG/CR-1278, page 17-24)[11]
suggest that for the first 5 minutes into a transient, it should be assumed that the operator is alone in the control room. Finally, given all this, stress is understandably moderate to high, so that even if the operator gets around to this particular problem, his error rate is somewhat elevated.
Recovery of flow from the EFW turbine-driven pump after dryout is dub-ious, because it is not known whether there is enough steam in the system to l
l
i operate the turbine long enough to generate more steam [Cf. NUREG-0611, j 111.4.3., bottom of p.Ill-10] There is an argument for disallowing recovery l 1
1 of the turbine-driven train, and this conservatism is adopted here.
4 Actuation Logic: NUREG-0611 prescribes 7x10-3 per channel for actuation logic failure probability. One should ask whether the assessment of 7x10-3 per channel is reasonable. In Westinghouse plants, this value tends ,
not to dominate the system unavailability, because NUREG-0611 prescribes substantial credit for operator actuation within the "available" time. Here, for some purposes, we (and AP&L) are giving no credit for operator actions within the first few minutes, so that the conclusions are correspondingly ,
sensitive to this parameter. For example, system fa actuationchannelswithoutoperatorbackupis(7x10glgrebyfailgreofboth) = 4.9x10 ,
which virtually exhausts the unavailability contemplated by the new SRP for '
auxiliary feedwater systems. B&W plants arguably need, and may have, actuation systems which are more reliable than this. But AP&L has not j explicitly documented this by providing failure data, and in any case, this level of detail is beyond the scope of this analysis.
khile a more complete analysis might substantiate a lower actuation failure probability, this would be partially offset by the inclusion of failures of automatic control, which have not been addressed here, Evidently, it is possible to place one channel of the Emergency Feedwater Initiation and Control Logic (EFIC) in " maintenance bypass". Since failure of
, a single logic channel has no effect, this has been neglected here; its 3
probability is dominated by the failure probability prescribed by NUREG-0611.
There are other bypasses associated with startup, which automatically remove
! themselves, and still others associated with shut down. These latter bypasses do not enter this analysis. ,
Steam Supply to STD Pump: Although not mentioned in NUREG-0611, the probability of blowing down both steam generators through stuck-open relief
. valves has been included here. The value adopted was used by Florida Power in
! their analysis of the CR-3 EFWS (12). This is a small contributor to the 1 unavailability of this train; it is included for reasons of completeness and
{ consistency with other analyses.
1 d
- 7.
SUMMARY
AND CONCLUSIONS i
Since work on the main body of this report was completed, design changes have been made by AP&L. The main body of the report (Sections 2 through 6) continues to reflect properties of the old design. The new design is discus-l
' sed in Appendix A, which provides details of the impacts of the changes on the system unavailability. Table 7.1 provides a comparison of results calculated here for both the old and new designs, and AP&L results for the old design, i
which are obtained from Appendix B. It should be noted that the figure sup-i plied in the AP&L submittal states that " comparison with NUREG-0611 results is inappropriate." AP&L results for the new design are not available. ,
- Major contributors to the BNL results include the following
- 1. Single valves in the common lines to and from the CST.
- 2. Maintenance on valves, calculated using NUREG-0611 assumptions. This is a major contributor because isolating the valve under maintefance loses more than the flowpath actually being maintained.
- 3. Actuation logic failure, calculated according to NUREG-0611 as-sumptions. This is a large contributor because NUREG-0611 mandates 7x10-3 per train, and it is assumed here that operator corrective action cannot recover this in time to prevent dryout. At least some of the time, corrective action takes place outside the control room, and some time is necessary for diagnosis.
- 4. Maintenance errors. These have been assessed by analogy with failure of valve restoration, which is quantified in NUREG-0611.
f' 1
i i
t
, . _ _ . _ _ . _ . , ~ , _ . . ,
4 REFERENCES <
- 1. " Emergency Feedwater System Upgrade Reliability Analysis for the Arkansas Nuclear One Nuclear Generating Station Unit No.1", prepared by B&W Plant Performance Engineering, submitted to NRC by AP&L in October 1981; herein called "The AP&L report."
- 2. " Generic Evaluation of Feedwater Transient and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants", NUREG-0611, U. S.
j Nuclear Regulatory Commission (January 1980).
- 3. Letter from D. F. Ross, Jr. (NRC) to "All Pending Operating License Ap-plicants of Nuclear Steam Supply Systems Designed by Westinghouse and Com-bution Engineering", dated March 10, 1980.
- 4. USNRC Standard Review Plan, Sec.10.4.9 (NUREG-0800), Revised July 1981.
- 5. " Transient Response of Babcock & Wilcox-Designed Reactors", NUREG-0667, U.
! S. Nuclear Regulatory Commission (May 1980).
l
- 6. " Request For Information, Arkansas Nuclear One, Unit 1, Auxiliary Feedwater System (AFWS) Requirements, Docket No. 50-313", submitted to NRC by letter from D. C. Trimble to R. W. Reid dated March 12, 1981.
- 7. "ANO-1 Emergency Feedwater Upgrade Preliminary Design", submitted to NRC by letter from U.C. Trimble to R. W. Reid (NRC) dated October 15, 1980.
i 1 8. " Reactor Safety Study, An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants, WASH-1400", NUREG-75(014), U. S. Nuclear Regulatory Commission (October 1975).
- 9. "IEEE Nuclear Reliability Data Manual", IEEE STD 500 (1977),
i
- 10. " Reactor Safety Study Methodology Applications Program: Oconee #3 PWR 1 Power Plant", NUREG/CR-1659, U. S. Nuclear Regulatory Commission (January.
1981), (Rev. May 1981).
- 11. " Handbook of Human Reliability Analysis With Emphasis on Nuclear Power Plant Applications", NUREG/CR-1278, U. S. Nuclear Regulatory Commission (October 1980).
- 12. See " Review of the Crystal River Nuclear Generating Station Unit No. 3 Emergency Feedwater System Reliability Analysis", NUREG/CR.... , U.S.
Nuclear Regulatory Commission.
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Table 6.la Contributors to Q1H (LMfW)
Note: C G 10-4 Value Description of Event 1x10-3 Failure of P7A to start 2.4x10-4 Failure of P7A to run 5.1x10-5 Failure of Steam Admission:
CVY-2 Fail to open ((3x10-3)gither CVY-1 and
'), or failure of one channel ' "* ""of**actuation
- P*rlogic coupleg w*ith
- C *7*
failuS*)
3x10-1.6x10-3 Either of the two pressure reducing vgives (CVY-3 and CVY-4) failing high (3x10-'+ e or miscalibration of their settings (10 gch), )
5x10-3 CVY-1 and CVY-2 left disabled after pump maintenance (these are normally closed, so position indication is superfluous) 5x10-4 CV2667, CV2666, CV2617 left closed after maintenance (control room indication) 1x10-4 FW-10B Flow Blockage 1x10-4 CV2802 Flow Blockage 4.4x10-5 CV2802 Spurious Closure: A=1.24x10-7/hr, 1/2 T = 1/2 month (Ref. 8) 5x10-4 CV2802 left closed 5x10-3 Test line inadvertently left open 5x10-4 CVX-1 and CVX-4 left closed and disabled after maintenance 3.4x10-4 Relief valve failure on both steam generators (Loss of steam to turbine)
= 1.5x10-2
- Q1H(LMFW), NO OPERATOR CREDIT
= 1.5x10-2 = Q1H(LMFW), Failure to initiate within 20 min.
OPERATOR CREDIT NOT GIVEN PAST DRYOUT FOR RECOVERY OF STD Train
Table 6.la (cont.)
ADDITIONAL CONTRIBUTORS TO Q1H GIVEN LOOP:
None.
{=1.5x10-2 = Q1H(LOOP), NO OPERATOR CREDIT
= 1.5x10-2
- Q1H(LOOP), Failure to initiate within 20 min.
OPERATOR CREDIT NOT GIVEN PAST DRYOUT FOR RECOVERY OF STD Train ADDITIONAL CONTIRBUTORS TO Q1H GIVEN LOAC:
None. l
{=1.5x10-2 = Q1H(LOAC), NO OPERATOR CREDIT
{=1.5x10-2 = Q1H(LOAC), Failure to initiate within 20 min.
OPERATOR CREDIT NOT GIVEN PAST DRYOUT FOR RECOVERY OF STO Train
Table 6.la (cont.)
Contributors to Q M1 (LNFW)
Note: C <<10-4 Value Description of Event 2.1x10-3 Pump maintenance (24-hour outage allowed) 2x10-3 Pump testing 2.1x10-3 CVY-1 maintenance 2.1x10-3 CVY-2 maintenance 2.1x10-3 CVY-3 maintenance 2.1x10-3 CVY-4 maintenance 2.1x10-3 CVX-1 maintenance 2.1x10-3 CVX-4 maintenance 2.1x10-3 CV2802 maintenance 2.1x10-3 CV2806 maintenance
[ = 2.1x10-2 g1M(LMFW), NO OPERATOR CREDIT
[ = 2.1x10-2 g1M(LMFW), Failure to initiate within 20 min. OPERATOR CREDIT NOT GIVEN FOR RECOVERY OF STD TRAIN AFTER DRYOUT ADDITIONAL CONTRIBUTORS T0 Qi g GIVEN LOOP:
None
[ = 2.1x10-2 Qi g(LOOP), NO OPERATOR CREDIT
{ 2.1x10-2 Q1M(LOOP), Failure to initiate within 20 min. OPERATOR CREDIT NOT GIVEN FOR REC 0VERY OF STD TRAIN AFTER l DRYOUT
)
l ADDITIONAL CONTRIBUTORS TO Q1M GIVEN LOAC: i
! None
[ = 2.1x10-2 Q1M(LOAC), NO OPERATOR CREDIT l
[ = 2.1x10-2 Q1M(LOAC), Failure to initiate within 20 min. OPERATOR CREDIT NOT GIVEN FOR REC 0VERY OF STD TRAIN AFTER DRYOUT
i i Table 6.lb Contributors to Q H2 (LMFW)
Note: C <<10-4 Value Description of Event 5x10-3 Failure of P7B: 1x10-3 pump, 4x10-3 control circuit 2.4x10-4 Failure of P7B to run: 3x10-5/hr. x 8 hrs.
5x10-3 ' , Control circuit left disabled after maintenance or ,
or l
C Control circuit left disabled after maintenance and failure to recover 1x10-4 FW10A Flow Blockage 1x10-4 CV2800 F1ow Blockage
- 5x10-4 CV2800 Left closed 4.4x10-5 CV2800 Spurious Closure: A =1.2x10-7/hr, 1/2 T=1/2 month (Ref. 8) 5x10-3 ' Test line inadvertently left open or ( or l
C Test line inadvertently left open and operator fails to recover 5x10-4 ' CVX-2 and CVX-3 left closed and disabled after maintenance i
or [ or C CVX-2 and CVX-3 left closed and disabled and failure to recover in 20 min.
[ = 1.6x10-2 92H(LMFW), NO OPERATOR CREDIT
[ = 5.9x10-3 Q2H(LMFW), Failure to initiate within 20 min.
. . _ . - - - - . - . _ - - _ -. . .. - . - . . . . = - - . - . . -- . . - . . . . . ~ . . . - . . -
l 4
k l
4 4
! Table 6.lb (cont. )
! Additional Contributors to Q2H given LOOP I
3.6x10-2 Diesel Generator Unavailability
= 5.2x10-2 02H(LOOP),NoOperatorCredit Q2H(LOOP), Failure to initiate within 20 l f=4.2x10-2 min.
i i
THE EMD TRAIN IS UNAVAILABLE GIVEN LOAC Q2H(LOAC), with or without operator. credit f=1.0 i
\
I l
l
!- ._24 -
l l
l l
Table 6.lb (cont.) l Contributors to Q M 2 (LMfW)
_4 Value Description of Event 2.1x10-3 Pump maintenance (24-hour outage allowed) 2x10-3 Pump testing i or or C
Pump testing and operator fails to realign 2.1x10-3 CV2800 maintenance 2.1x10-3 CV2803 maintenance 2.1x10-3 CV-X2 maintenance 2.1x10-3 CV-X3 maintenance
[= 1.3x10-2 02M(LMFW), no operator credit 1
[ = 1.1x10-2 g2M(LMFW), Failure to initiate in 20 min.
ADDITIONAL CONTRIBUTORS TO Q2M GIVEN LOOP:
None (Diesel maintenance included in Q2H)
{ = 1.3x10-2 Q2M(LOOP), no operator credit
[= 1.1x10-2 02 M(LOOP), failure to initiate in 20 min.
ADDITIONAL CONTRIBUTORS TO Q2M GIVEN LOAC:
THIS TRAIN IS UNAVAILABLE GIVEN LOAC Q2M(LOAC) = 0.
TABLE 6.2 RESULTS FOR Q1 AND Q2 41H Q1M Q2H 02M 1
LMFW, 1.5 x 10-2 2.1 x 10-2 5.9 x 10-3 1.1 x 10-2 operator credit LMFW, 1.5 x 10-2 2.1 x 10-2 1.6 x 10-2 1.3 x 10-2 no operator credit LOOP, 1.5 x 10-2 2.1 x 10-2 4.2 x 10-2 1.1 x 10-2 operator credit '
LOOP, 1.5 x 10-2 2.1 x 10-2 5.2 x 10-2 1.3 x 10-2 no operator credit LOAC, 1.5 x 10-2 2.1 x 10-2 1 0 operator credit LOAC, 1.5 x 10-2 2.1 x 10-2 1 0 no operator credit I
l l
Table 6.3 '
i Single Point & Common Failures Note: C <<10-4 :
I Value Description of Event 1x10-4 Failure of CS-98 to remain open
! 1x10-4 Failure of CS-99 to remain open l 1x10-4 Failure of CS-19 to remain open 1x10-4 Failure of FW14 to remain open 1x10-4 Failure of FW15 to remain open 1x10-4 Failure of three-way. valve to remain open 4.9x10-5 '
Failureofbgtgchanelsofinitiate logic _(7x10-)
or ,
or C Failure of both logic channels and failure to initiate manually J
= 6.5x10-4 Other, no operator credit f=6x10-4 Other, failure to initiate in 20 minutes J
d
.1 4
-l
j Table 7.1 EFWS Unavailability Values given here are calculated from Q = Q1H 02H + Q1H 02M + 4 1M 92H + Other, where Q H = Hardware failures associated with P7A Q H = Hardware failures associated with P7B Q M = Maintenance unavailability associated with P7A Q M = Maintenance unavailability associated with P78 For the old design, contributions to Q H,3 Q H, 2 Q M, 3
and Q 2M are given !
in Table 6.2 and sumarized in Table 6.3. Contributions to "other" are given in Table 6.3.
Changes to the tables resulting from the design changes are detailed in Appendix A, where new tables are provided.
in the sumary at the beginning of the present report, the results quoted
- are "New Design, This Work." -
Old Design, Old Design, New Design, This Work AP&L Report This Work LMFW, Operator Credit 9.8 x 10-4 1.4x10-5 6.3x10-4 LMFW, No Operator Credit (Dryout) 1.4 x 10-3 3.6x10-4 8x10-4 4 LOOP, Operator Credit 2.3 x 10-3 9.1x10-5 1.5x10-3 LOOP, No Operator Credit (Dryout) 2.7 x 10-3 5.2x10-4 1.6x10-3 LOAC, Operator Credit 3.6 x 10-2 5.7x10-3 2.7x10-2 LOAC, No Operator Credit (Dryout) 3.6 x 10-2 1.4x10-2 2.8x10 2 l
1 l
. - ... ..... . . . . . . . . . . . - . - . , n --
APPENDIX A:
IMPACT OF DESIGN CHANGES ON SYSTEM AVAILABILITY 1
Af ter work on the main body of this report was completed, new information was received from AP&L by way of NRC, which is enclosed here as Appendix B.
This is in the form of responses to written questions. Since those questions were raised, significant design changes have been made, which are covered in the AP&L responses. This appendix summarizes important changes and describes i their impact on system availability. AP&L has not provided a revised estimate of system unavailability reflecting the impact of the changes.
A.1 Revisions in Suction Fonnerly, the line from the CST had one manual valve (CS19) and two check valves (CS98 and CS99) in series, any one of which could block flow to both pumps.
Now, there are redundant check valves in paralled with CS98 and CS99; these markedly reduce the probability of blockage due to check valves.
However, manual valve CS19 can still block flow to both pumps. NUREG-0611 prescribes 10-4 for this; considered as a hardware failure, this is a conservatism (by comparison with LER rates), but on the other hand, the scope for inadvertent manipulation is unknown (see Table A.3 for effect of these changes).
A.2 Revisions in Steam Supply to Turbine Driven Pump Previously, the STD pump required a pressure reducing station, and was equipped which with could failoverpressure open. In the relief valves (upstream of the governor valve) new design, there are no valves between the steam admission valves and the governor valve, so there are fewer things which can fail (see Table A.la for the effect of these changes).
I A.3 Changes in Pump Recirculation Control 1
l Previously, there was a minimum flow path common to both pumps whose i capacity was varied by control signals which depended on pump flow. This arrangement lef t the system vulnerable to certain common cause failures, which were discussed in the AP&L report and are reviewed in the main body of the present report.
In the new design, minimum flow is controlled independently for each pump, so that common cause flow diversion has been eliminated as a failure mode. The recirculation path back to the CST still contains a check valve and a manual valve in series; if total loss of recirculation can still fail the system (e.g., in periods of low flow demand), then these are still single failure points.
Minimum flow control is accomplished by a mechanical device which opens up the main flowpath or the minimum flowpath depending on flow demand. At one point, this is called " interlocking check valves" in the AP&L response. There are also devices called " recirculation valves," which may be the same thing.
It is not clear what failure probability to assign to this arrangement (e.g.,
for the probability of its sticking in position to block flow to the SGs).
. __ . - . . - _ - - - - . _ = . . - - - . _ _
1 4
i l The result of the calculation is sensitive to this only if it is much worse (say, ten times) than a check valve failure probability. Here, a check valve j
failure probability is used, in the expectation that this device is unlikely j to fail more often than the pumps unless there is something fundamentally wrong with its design. Design problems with new devices are not within the i
scope of this report.
The modification reduces unavailability due to pump testing. When a test is performed, a downstream H0V is commanded closed by the operator. In this
- configuration, with the pump running, the recirculation valve automatically provides minimum flow. In the event of a system demand during a pump test, the logic automatically opens the downstream MOV, and the recirculation valve will automatically direct flow to the SGs. No operator action is necessary.
Previously, the pumps were tested at high flow rates, sufficient to constitute flow diversion. The new testing flowrate is not known. (The present analysis is insensitive to this, in any case.) Leaving test lines open is no longer a contributor.
Tables A.la and A.lb show the effects of these changes in several places.
! The common cause flow diversion was not originally reflected in these tables.
1 i
I f
4
1 Table A.la Changes to Table 6.la in New Design Contributors to Q H1 (LMfW)
_4 Deleted by Value ,0esign Change Description of Event 1x10-3 Failure of P7A to start 2.4x10-4 Failure of P7A to run 5.1x10-5 Failure of Steam Admission:
CVY-2 Fail to open ((3x10-3)gither CVY-1 and
), or failure of one channel of actuation logic coupled with failu x 3x10geofonevalvetooperate(2x7x10-3
)
1.6x10-3 ***
Either of the two pressure reducing valves (CVY-3 and CVY-4) failing high (3x10-4 each or miscalibration of their settings (10-3) ),
5x10-3 CVY-1 and CVY-2 left disabled after pump maintenance (these are normally closed, so position indication is superfluous) 5x10-4 CV2667, CV2666, CV2617 left closed after maintenance (control room indication) 1x10-4 FW-10B Flow Blockage (now a recirculation valve) 1x10-4 CV2802 Flow Blockage j 4.4x10-5 CV2802 Spurious Closure: A=1.24x10-7/hr, 1/2 T = 1/2 month (Ref. 8) 5x10-4 CV2802 left closed 5x10-3 ***
Test line inadvertently left open 5x10-4 CVX-1 and CVX-4 left closed and disabled after maintenance 3.4x10-4 Relief valve failure on both steam generators (Loss of steam to turbine)
= 1.5x10-2 = 41H(LMFW), NO OPERATOR CREDIT, BEFORE DESIGN CHANGE
=
1.5x10-2 =Q1H(LMFW), failure to initiate within 20 min., BEFORE DESIGN CHANGE. OPERATOR CREDIT NOT GIVEN PAST ;
DRYOUT FOR RECOVERY OF STD Train
= 8.4x10-3 = Q1H(LMFW), CREDIT FOR DESIGN CHANGES (Asterisks denote failure modes removed by design change.)
1 Table A.la (cont.)
ADDITIONAL CONTRIBUTORS T0 Q1H GIVEN LOOP:
None. 1
{=1.5x10-2 =Q1H(LOOP), N0 OPERATOR CREDIT, BEFORE DESIGN CHANGE
= 1.5x10-2 = Qig(LOOP), Failure to initiate within 20 min., BEFORE DESIGN CHANGE. OPERATOR CREDIT NOT GIVEN PAST DRYOUT FOR REC 0VERY OF STD Train
{=8.4x10-3 . Q1H(LOOP), CREDIT FOR DESIGN CHANGE ADDITIONAL CONTIRBUTORS TO Q1 H GIVEN LOAC:
None.
= 1.5x10-2 = Q H(LOAC),
1 NO OPERATOR CREDIT, BEFORE DESIGN CHANGE
=Q1H(LOAC), Failure to initiate within 20 min., BEFORE
{=1.5x10-2 DESIGN CHANGE. OPEPATOR CREDIT NOT GIVEN PAST DRYOUT FOR RECOVERY OF STD Train
= 8.4x10-3
- Q H (LOAC),
1 CREDIT FOR DESIGN CHANGE Table A.la (cont.)
r Contributors to Q gi (LMFW)
I Note: C <<10-4 Deleted by Value Design Change Description of Event t
2.1x10-3 Pump maintenance (24-hour outage allowed) 2x10-3 ***
Pump testing 2.1x10-3 CVY-1 maintenance 2.1x10-3 CVY-2 maintenance l 2.1x10-3 ***
CVY-3 maintenance 2.1x10-3 ***
CVY-4 maintenance 2.1x10-3 CVX-1 maintenance 2.1x10-3 CVX-4 maintenance l
2.1x10-3 CV2802 maintenance 2.1x10-3 CV2806 maintenance
- =
2.1x10-2 Qig(LMf'W), NO- OPERATOR CREDIT, BEFORE DESIGN CHANGE
=
2.1x10-2 Q ig(LMFW), Failure. to initiate within 20 min., BEFORE 4
DESIGN CHANGE. OPERATOR CREDIT NOT GIVEN FOR RECOVERY 0F STD TRAIN AFTER DRYOUT
{=1.5x10-2=QM(LNFW),CRDEITFORDESIGNCHANGE 1
(Asterisks denote failure m@ cem ed by design change.)
i s
l W
F- v y .+
, c. -
/ =
, i
,,r'^ v)
. . , ,' ' +
,A.
, ,_.--e . , -
L -33 , f :y;: > .
. i+~ ' ~ ! ~~ ' 9
- ) ,,
l d ' , j j{ w g.+ . .* '
.. . ,.7 ., ,
___, _ , , ....... -## I ,-
Table A.la (cont.)
Contributors to Q1M ADDITIONAL CONTRIBUTORS T0 Q1M GIVEN LOOP:
None
= 2.1x10-2 Q1M(LOOP), NO OPERATOR CREDIT, BEFORE DESIGN CHANGE
= 2.1x10-2 Q1M(LOOP), Failure to initiate within 20 min., BEFORE DESIGN CHANGE OPERATOR CREDIT NOT GIVEN FOR REC 0VERY OF STD TRAIN AFTER DRYOUT
= 1.5x10-2 = Q1M(LOOP), CREDIT FOR DESIGN ADDITIONAL CONTRIBUTORS TO Q1M GIVEN LOAC:
None
= 2.1x10-2 Q1M(LOAC), NO OPERATOR CREDIT, BEFORE DESIGN CHANGE
= 2.1x10-2 Q1M(LOAC), Failure to initiate within 20 min., BEFORE DESIGN CHANGE OPERATOR CREDIT NOT GIVEN FOR REC 0VERY OF STD TRAIN AFTER DRYOUT
= 1.5x10-2 = Q1M (LOAC), CREDIT FOR DESIGN CHANGE l
Table A.lb Changes to Table 6.lb in New Design Contributors to Q H2 (LMfW)
Deleted by Note: C <<10-4 Value Design Change Description of Event 5x10-3 Failure of P78: 1x10-3 pump, 4x10-3 control circuit 2.4x10-4 Failure of P78 to run: 3x10-5/hr. x 8 hrs.
5x10-3' Control circuit left disabled after maintenance or or c Control circuit left disabled after maintenance and j
failure to recover 1x10-4 FW10A Flow Blockage (Now a recirculation valve) 1x10-4 CV2800 Flow Blockage
) 5x10-4 CV2800 Left closed 4.4x10-5 CV2800 Spurious Closure: A=1.2x10-7/hr, l 1/2 T=1/2 month (Ref. 8) 5x10-3 ***
4 Test line inadvertently left open i or p or C ,
Test line inadvertently lef t open and operator fails to recover 5x10-4' CVX-2 and CVX-3 left closed and disabled af ter
- I maintenance or / or C CVX-2 and CVX-3 left closed and disabled and j f ailure to recover in 20 min.
= 1.6x10-2 g2H(LMFW), N0 OPERATOR CREDIT, BEFORE DESIGN CHANGE
=
5.9x10-3 Q2H(LMFW), Failure to initiate within 20 min., BEFORE DESIGN CHANGE
=
1.2x10-2
- 02H(LMFW), NO OPERATOR CREDIT, CREDIT FOR DESIGN CHANGE 1
5.9x10-3
- Q2H(LMFW), Failure to initiate within 20 min., CREDIT FOR DESIGN CHANGE (Asterisks denote f ailure modes deleted by design change.)
. . -. . - . _. -- . . = .
- Table A.lb (cont.)
Additional Contributors to Q H2 given LOOP 3.6x10-2 Diesel Generator Unavailability f=5.2x10-2 Q2H(LOOP), No Operator Credit, Before Design Change f=4.2x10-2 Q2H(LOOP), Failure to initiate within 20 min., Before Design Change l l
> 1 f=4.8x10-2 Q2H(LOOP), No operator credit, credit for design change I i
f=4.2x10-2 Q2H(LOOP), failure to initiate within 20 min., credit for design change J
l i
I 1
j 1 l
Table A.lb (cont.)
Contributors to Q M2 (LMFW)
Note: C <<10-4 Deleted by Value Design Change Description of Event
! 2.1x10-3 Pump maintenance (24-hour outage allowed) 2x10-3 ***
Pump testing or or C ' ***
Pump testing and operator fails to realign 2.1x10-3 CV2800 maintenance 2.1x10-3 CV2803 maintenance 2.1x10-3 CV-X2 maintenance 2.1x10-3 CV-X3 maintenance i
E = 1.3x10-2 02M(LMFW), no operator credit E = 1.1x10-2 92M(LMFW), Failure to initiate in 20 min.
E = 1.1x10-2 92M(LMFW), CREDIT FOR DESIGN CHANGE ADDITIONAL CONTRIBUTORS TO Q 2M GIVEN LOOP:
None (Diesel maintenance included in Q2H)
E = 1.3x10-2 Q2M(LOOP),nooperatorcredit E = 1.1x10-2 02M(LOOP), Failure to initiate in 20 min.
- E = 1.1x10-2 Q2M(LOOP), CREDIT FOR DESIGN CHANGE ADDITIONAL CONTRIBUTORS T0 2 QM GIVEN LOAC
THIS TRAIN IS UNAVAILABLE GIVEN LOAC Q2M(LOAC) = 0.
(Asterisks denote failures deleted by design change.)
e -
~~~ '
-- . - - ~ .. .
TABLE A.2 RESULTS FORt Q AND2 Q , Credit for Design Change Q1H Q1M QH 2 Q2M LMFW, 8.4 x 10-3 1.5 x 10-2 6 x 10-3 1.1 x 10-2 operator credit LMFW, 8.4 x 10-3 1.5 x 10-2 1.2 x 10-2 1.1 x 10-2 no operator credit LOOP, 8.4 x 10-3 1.5 x 10-2 4.2'x 10-2 1.1 x 10-2 operator credit 5
i LOOP, 8.4 x 10-3 1.5 x 10-2 4.8 x 10-2 1.1 x 10 no operator
- credit LOAC, 8.4 x 10-3 1.5 x 10-2 1 0
- operator credit J
LOAC, 8.4 x 10-3 1.5 x 10-2 1 0
! no operator credit i
i l
l l
Table A.3 Changes to Table 6.3 in New Design Single Point and Common Failures Note: C <<10-4 i Deleted by -
Value Design Change Description of Event 1x10-4 ***
Failure of CS-98 to remain open lx10-4 ***
Failure of CS-99 to remain open 1x10-4 Failure of CS-19 to remain open 1x10-4 Failure of FW14 to remain open 1x10-4 Failure of FW15 to remain open 1x10-4 Failure of three-way valve to remain open 4.9x10-5 '
Failureofbgtgchannelsofinitiate logic (7x10- )
or ,
C Failure of both logic channels and failure to initiate manually
[=6.5x10-4 Other, no operator credit, Before Design Change r .
[=6x10-4 Other, failure to initiate in 20 minutes, Before Design Change I
b=4.5x10-4 Other, no operator credit, credit for design change
[=4x10-4 Other, failure to initiate in 20 minutes, credit for design change (Astrisks denote failures deleted by design change.)
Table A.4 Changes to Table 7.1 in New Design EFWS Unavailability Before and After Design Change Values given here are calculated from Q = Q1H Q2H + Q1H Q2M + Q1M Q2H + Other, where Q H = Hardware failures associated with P7A Q H = Hardware failures associated with P78 Q M = Maintenance unavailability associated with P7A Q M = Maintenance unavailability associated with P78 Contributions to Q 4H and Q2H are given in Table A.1. Contributions to "other" are given in Table A.3.
Before Design After Design Change Change LMFW, Operator Credit 9.8 x 10-4 6.3x10-4 LMFW, No Operator Credit (Dryout) 1.4 x 10-3 8x10-4 LOOP, Operator Credit 2.3 x 10-3 1.5x10-3 LOOP, No Operator Credit (Dryout) 2.7 x 10-3 1.6x10-3 LOAC, Operator Credit 3.6 x 10-2 2.7x10-2 LOAC, No Operator Credit (Dryout) 3.6 x 10-2 2.8x10-2
1 1
[
APPENDIX B: SUBMITTAL FROM AP&L TO NRC - AUGUST 11, 1982 1
I Fj. 4
!J h NN ARKANSAS POWER & LIGHT COMPANY PCST CFRCE BOX 551 UTTLE ROCK, ARKANSAS 72203 (501)371-4000 August 11, 1982 1CAN088202 Director of Nuclear Reactor Regulation ATTN: Mr. J. F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555
SUBJECT:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. OPR-51 Questions on the Emergency Feedwater System Upgrade Reliability Analysis Gentlemen:
Your letter of April 21,1982,(1CNA048212) requested additional information pertaining to the report " Emergency Feedwater (EFW) System Upgrade Reliability Analysis for Arkansas Nuclear One, Unit No.1". In AP&L's letter of May 26,1982,(1CAN058211) we proposed to submit the requested information by August 15, 1982. Therefore, attached is our response.
Very truly yours, n
' ol m . f;o John R. Marshall Manager, Licensing I l JRM:MCysc , g y-Attachments 8208170129 820811 gDR ADOCK 05000333 PDR
ADDITIONAL INFORMATION CONCERNING EMERGENCY FEEDWATER UPDRAGE RELIABILITY ANALYSIS l
FOR ARKANSAS. NUCLEAR ONE, UNIT NO. 1 DOCKET N0. 50-313 Question 1 "The report's results are not presented in a manner that lends itself to a "NUREG-0611-type" comparison. For example, unavailability given LOOP is not tabulated. LOOP appears as a basic event on the fault tree; was this entered as the probability of LOOP given LMFW? How was this handled?"
Response 1 The presentation of these results in the format of NUREG-0611 does not imply that comparisons to NUREG-0611 are appropriate. Because the scope and assumptions of the analysis are different than for the NUREG-0611 analysis; there is no common basis for comparison. For example, the assumptions used for power availability illustrate the differences.
NUREG-0611 assumed perfect DC power (i.e., failure probability of 0.0) for all cases, and one of two perfect diesel generators for this LOOP case. The analysis of ANO-1 EFW reliability evaluated power reliability in detail and assumed realistic probabilities for DC power and diesel generator failure.
The attached figure 1.1 " Reliability of ANO-1 EFW System" supplies the results of the report " Emergency Feedwater System Upgrade Reliability Analysis for the Arkansas Nuclear One Nuclear Generating Station Unit No.1, April 1981" in the format of NUREG-0611. The attached table (Figure 1.2) "Unavailabilities for ANO-1 EFW" supplies the numerical results broken down into the three cases; loss of main feedwater (LMFW),
loss of offsite power (LOOP), and loss of AC power (LOAC).
Question 2 "A narrative description of cut sets was provided, but no quantitative details were given. What were the contributions from the dominant cut sets?"
Response 2 The attached pages A-1 through A-3 explain the code used for the cut sets. The attached pages B-1 through B-36 supply the computer listing of
dominant cut sets for all of the analyzed cases. Each 8 letter code describes a basic event which can be found on the fault trees in Appendix B of the April 1981 Reliability Report.
Question 3
! " Failure data were not given. What are they?"
Response 3 The attached pages C-1 through C-4 provide the basic event failure data !
- used in the computer analysis of ANO-1 EFW Reliability. !
l Question 4 i
, "The following" questions pertain to the scenario discussed in point 4 of Section 3.2.1.
- a. "At the bottom of p. 3-1, it is stated that ". .. AC powered valves CV-X2 and CV-X3 will not open." (Fig. D-7 seems to indicate that CV-X3 is DC powered. Is this correct?) In paragraph D.2.3 (p. D-3), it is stated that "The flow of EFW to each SG is controlled by redundant normally-open modulating solenoid motor operated control valves in parallel paths." The fact that loss of AC fails the control valves closed suggests that they are normally closed, but the subsequent discussion says that they are normally
!. open. What are the normal positions of all valves in the four discharge paths? In this scenario, why were they presund closed?"
- b. "Section 3.2.1.4 also talks about the opening of recirculation valves CV-2815 and CV-2816 during the loss-of-all AC power event which causes portion of turbine driven pump flow bypassing to the condensate storage tank. The ANO report recommends that an analysis be done to detemine if adequate flow will be available to the SGs with this bypass flow. Has the analysis.been performed? How serious is the bypass in degrading the total flow to the SGs?"
Response 4a Attached figure 4 is the latest P&ID for the EFW System, this f.igure will be referenced in response to the remaining questions. Figure 4 indicates that four modulating control valves are normally open and fail open on loss of control signal or motive power. The four motor operated isolation valves are also normally.open and fail "as is".
Response 4b The recirculation piping will be changed to that shown on figure'4. This will eliminate the common recirculation flow path and resolve the concerns expressed in this question. The analysis will therefore not be done. The recirculation flow will not significantly degrade flow to the l
SG. This is due to the automatic nature of the recirculation valves to l
be installed versus the constant flow nature of the existing design.
Pump recirculation will occur only during periods of low total flow, and was considered in pump capacity calculations.
Question 5 l "What physical measurement (s) actually regulate the recirculation valves:
i
' (CV-2815 and CV-2816)? According to the text (p. D-3, paragraph D.2.5),
flow elements upstream of th.e control valves are used to decide whether flow is being demanded from each pump; however, the diagram suggests that the positions of the control valves (CV-X2 and CV-X3) are sensed."
Response 5 The recirculation flow piping will be changed according to figure 4. All valves in the recirculation flow path will be normally open and recirculation flow controlled by FW10A and FW108. FW10A & B use pump flow to regulate recirculation flow. These valves consist of a set of interlocked check valves. One check valve is in the discharge path to the OTSG, the other is in the recirculation path. When the discharge path check valves closes due to reduced flow, the interlock will open the check valve in the recirculation line. This method does not require electric measurement and interlocks to establish recirculation and by design eliminates common (P7A and B) recirculation flow control.
Question 6 "If actual flow is the parameter to determine the recirculation then a simple failure of the EMD pump will also cause the recirculation in the TD pump flow (to increase) during the loss of main feed pump event. This failure mode should be included in the fault tree."
Response 6 This question no longer applies because the new recirculation control valves and piping design make pump recirculation flow independent of each other.
Question 7
" Pump Trip: Under what conditions do the pumps trip? Are there trips which are supposed to operate during a test but not during an emergency?"
"The fault tree takes credit for operator recovery of suction source in the event that it had been inappropriately left valved off after maintenance? How much time was assumed to be available for this before pump damage occurred?"
Response 7 The electric pump trips on overcurrent.
The turbine pump trips on overspeed.
There are no trip functions which are bypassed during an emergency.
Recovery of the suction source was assumed to occur before pump damage occurred.
Question 8 "What valves are closed for pump maintenance? Are any of the discharge valves disabled for this purpose?"
"What action is taken to isolate steam from the turbine driven pump during maintenance."
Response 8 During maintenance activities all suction and discharge valves are closed and tagged out of service per ANO procedures. In addition the steam inlet valves to the turbine pump are closed and the circuit breaker for the electric pump is opened and tagged out of service.
Question 9 "Providing steam to the turbine pump: One " steam unavailable" scenario involves the following: One of the valves in the pressure reducing station fails open, and one or the other of the relief valves fails to reseat. According to the fault tree, the relief capacity is such that there is insufficient pressure to drive the turbine if a relief valve is stuck open. It seems likely that given a wide-open control valve, the relief valve will cycle open and closed, so that even a small cyclic failure probability will lead to a substantial overall failure probability as the valve is repetitively cycled. Doesn't this scenario therefore have essentially the probability of either of the control valves failing high?"
Response 9 As part of the EFW upgrade project, the existing turbine is being replaced with a high pressure model. This eliminates the need for pressure reducing and safety valves.
Question 10
- a. " Isolation of the discharge paths: Are there single failures in the vector logic that can isolate both discharge paths from a given pump? Example: Channel'D logic can isolate both discharge paths from the turbine driven pump. Are there failures in Channel D (e.g., power failures) that isolate both paths?"
- b. "Are these valves commanded open by vector logic given a simple LMFW, or does the logic assume that they are already open as they are supposed to be?"
l
. . . .. \
Response 10a It is possible for a channel, in this case Channel C, to fail. It is conceivable that such a failure could isolate both discharge paths from the turbine driven EFW pump. However, the single failure criterion as applied to the EFW system requires that at least one pump be able to supply EFW to at least one steam generator. Isolation of the turbine driven pump would require the following additional failures before the EFW system would be inoperable:
- 1. Loss of offsite power (LOOP)
- 2. Loss of both Diesel Generators (LOAC)
Thus the single failure of channel C would not render EFW inoperable.
Response 10b These valves are normally open and commanded open by the vector logic.
This is due to the system requirement that these valves must be closed during pump testing. An EFW initiation during a test will open these valves and close the test isolation valves, thus terminating the test and providing proper EFW system alignment.
The EFW vector logic commands the valves closed in the event of steam generator overfill or main steam isolation.
2 f.D 4 2 O Cl3 4 O w C C W e-- W a M 2 VB *
\ w ::3 W M s,3 e- C W w W4 2 M CEll V3 " 4 e
L OM M w 2L M O g >= Q w a M M
> 2A ==A
>== 4h
- 24 W J 2
>==. 2- >=
M C >=
W 4 M
>"=
V3 a Zw .M. Cl3 2 - Cz: w
> W ( W C#9 M -
> w >= >=
2 4 WA 4 E 2 -@ 2
- O us >=
W -
- l3 W
L 2 Wg E#D Q E 4 e O C3 4 .=
O w W 2 N a
M w ==.
4 g 4#s e M M .=O 4 h O.=
p t#9 W W E#9 3m e
M O M .J W W 2 s
- M W 4E 2 4 2 ll:3 t== 6
" O M 42 M
4
>=
" w to 2 O
=
4 Q
m
=>
2_
2 M
[
~
- >= M a w -
LL W O =2 E g CO >=
2 Q w 4#9 2 U
w WE >
2 A .4: W qy Of If GM 4#9
 O =
Q 2 C"3 >=
- O 20 W 42 K
~' M 4 O s#9 w 2*
- h C3. =W -aO Wb M >= 4ll3 4O O I b e C.3 a
==A EG 4#3 N OW W
 m W6
 O 2M 42 >= d K2 *
- OO O g p - 2 2
-2 4 >- ,
>== O M C.3 4- W4 M >= A O4 OW
> = - ll>
O >=* EO "
4 == W >=
2 O C.3
> = - llll3 W J aE aM QM ll"3 W 2O ww =r3
UNAVAILABILITIES FOR ANO-1 EFW Case 1 Case 2 Case 3 LMFW LOOP LOAC EFW Initiate includes failure to-initiate EFW due to fluid system failure, spurious isolation by FOGG or overfill protection, and EFIC initiation failure.
Fully automatic initiation 3.6x10 -4 5.2x10
-4 1.4x10
-2 Includes operator corrective action within 1.4x10 -5 9.1x10 -5 5.7x10 -3 20 minutes.
EFIC Control
-3 ~
Fully Automatic control 8.0x10 8.0x10 9.7x10 -3 includes operator corrective action within 1.3x10 -6 1.3x10 -6 2.6x10
-6 20 minutes.
FOGG A> 600 psi B < 600 psi (feed only A) 3.8x10 -4 3.8x10 ~4 3.8x10 ~4 A < 600 psi B > 600 psi (feed only B) 3.8x10 ~4 3.8x10 -4 3.8x10 ~4 A < 600 psi B < 600 psi A 150 psi > B 1.5x10 -3 1.5x10 -3 1.5x10 -3 (feed only A)
A < 600 psi B < 600 psi B 150 psi > A 1.5x10 -3 1.5x10 ~3 1.5x10 ~3 (feed only B)
EFW Overfill Protection 6.2x10 -4 6.2x10 -4 3.1x10 -3 I
710q l . 7.
I 1
1 Babcock &Wilcox ' " ' ' " ' " "
Nuclear Power Generation Division GENERAL CALCULATIONS BASIC EVENT IDENTIFICATION CODE XXYYYYFM XX - Compone.,t Identification Code YYYY - Unique Component Identifier FM - Failure Mode PREPARED BY Daft DOC. NO.
l ttvitwt0 SY -59xt: ,,og wo, N-l
l COMPONENT IDENTIFICATION CODE l
COMPONENT CODE Buffler amp or battery BA (batteries have numbers, e.g. , BAD 07) l Breaker BR Cabinet main circuit breaker CB Charger CE Controller CN Pressure sensor, delta P DP FOGG module PC FOGG module PG Invertar IN Laval sensor LS Level transmitter LT Valve or htunes opcrator OF Proportioner P1 Rasistor EE Rata follower RF Relay II Setpoint signal SP Sensor power supply transformer SPT Subtractor SU Trip cabinet TC Timer circuit TI Valve stem VS 4
I i
COMPONENT FAILURE MDDES IDENTIFICATION CODE l
FATT-URE MODE CODE All Modes AE Catastrophic Failura CF Erroneously Trips ET Flow Blockaga FB Fails Closed FC Fails to Energiza FE Fails to h4"4"4814'* FT Fails Loir FL Fails Open 70 Fans to Run Fa Fails to Start FS Fails to Switch FW T=A4entas Trip IT Left Closed LC Lefe Open LO M.chan4em1 Failure MF Miscalibrated High ME Miscalibrated Low ML l
No Signal NS i
In Preventive Maintananea PM Fails to Essent RS
. Spuriously Closes SC Shorts SE Spurious Signal SS
( Fails to Close TC Fails to Detect Trip TD I Fails to Open TO In Test TS Unable to Trip UT k-b
I Babcock &Wilcox PDs mu.i v..ii Nuclear Power Generation Division GENERAL CALCULATIONS DOMINANT CUT SETS FOR ANO-1 EFW UNAVAILABILITY S. OE.
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BIBLIOGRAPHIC DATA SHEET BNL-NUREG,-51721 4 f tTLE AND SUBTITLE (Add vokmr No.of apprweraart 2. (Leave blmk Review of the Arkansa Nuclear One Generating e/S ACCESSION NO.
Stg ion Unit No ljme ency Feedwater System aRECieis/
[ AlJ kEsDR('Sh
- 5. D ATF[EPORT COMPLE TED R. Youngblood, I.A. Papaz lou IlW
- [5tember
- 9. PERFORMING ORGANIZATION NAME -D M AILING ADDRE SS (inclu* I,a Co*J
[ ATE REPORT ISSUED Department of Nuclear Engine ing IlM Brookhaven National Laborator [ NBruary Upton. NY 11973 8- (Leave Nank)
- 12. SPONSORING ORGANIZATION NAME AND mat NG ADOHESS (Include I,a Co*/
10 PRCiJE CT/ TASK / WORK UNIT NO.
Division of Safety Technology Office of Nuclear Reactor Regulatio it. FIN No.
U.S. Nuclear Regulatory Commission A-3933 Washington, DC 2055S 13 TYPE OF REPORT r soo cove RED (Inclus,ve daars)
Technical Report 15 SUPPLEMENTARY NOTES 14. (Leave otankJ
- 16. ABSTR ACT a00 words or lesst The purposes of this report are: (1) to rev w the Emergency Feedwater System Upgrade Reliability Analysis for the Arkansas Nuclea One Nuclear Generating Station Unit No.1, and (2) to estimate the probability that the Em r ncy Feedwater System will not perform its mission for each of three different initia rs (1) loss of main feedwater with offsite
- power available (2) loss of offsite powe , (3) loss of all 4160 VAC power. The scope, methodology, and failure data are prescri ed by 'UREG-0611, Appendix III.
- 17. KE Y WORDS AND DOCUMENT ANALYSIS 17a DESCRIPTO
- Reliability Analysis Auxiliary Feedwater System Arkansas Nuclear One Nuclea enerating Station Unit No. 1 Pump and Valve Failure Rates 17tt IDENTIFIE RS!OPEN ENDE D TERMS 18 AVAILABILITY STATEMENT 19. U Y ( his report) 21. NO. OF PAGES Unlimited 2a SECuRiTv CLATMTh,s pg) 22. PRICE Unclassified S NRC FORM 335 11141)
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