ML20138M421
| ML20138M421 | |
| Person / Time | |
|---|---|
| Site: | Oconee, 05000000 |
| Issue date: | 09/30/1985 |
| From: | Pegram J, Tally C, Weimer J BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20138M420 | List: |
| References | |
| 77-1159095, 77-1159095-00, BAW-1890, NUDOCS 8511010240 | |
| Download: ML20138M421 (47) | |
Text
{{#Wiki_filter:. _ _ . . _ _ _ _ _ .. _- _ __ -_ _. . rs, ~ r 77 5 -00 3EPTEMBER 1985 h I"d, OWNERS GRObP g Transient Assessment g Committ,ee 8 Transient Assessment Program 1 (TAP) i f l 1 JUSTIFICATION FOR RAISING SETPOINT l FOR REACTOR TRIP ON HIGH PRESSURE 1 I I B&W 177-FA Operating Plants A85$ SSohgg9 {8A i j Babcock & Wilcox a McDermott company
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; p BAW-1890 1 f
' 77-1159095-00 ., s 4 SEPTEMBER 1985 l t ; i I I I i i JUSTIFICATION FOR RAISING SETPOINT I FOR 4 4 REACTOR TRIP ON HIGH PRESSURE I r i 4 i i BY = J. W. PEGRAM i i C. W. TALLY J. A. WEIMER l , 4 i 4 e 4 1 I l I l l l l
I BAW-1890 77-1159095-00 SEPTEMBER 1985
.I t
JUSTIFICATION FOR RAISING SETPOINT FOR REACTOR TRIP ON HIGH PRESSURE i f i j Babcock & Wilcox
. Nuclear Power Division
' P. O. Box 1260 Lynchburg, Virginia 24505 i j Babcock &WHcom a McDermott company
1 l i 1 CONTENTS r Page
- 1. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . 1-1
- 2. POTENTIAL BENEFITS . . . . . . . . . . . . . . . . . . 2-1 2.1. Impact of Post-TMI Changes . . . . . . . . . . . 2-1 2.2. Potential for Trip Reduction . . . . . . . . . . 2-2
- 3. IMPACT ON PORV OPENING FREQUENCY . . . . . . . . . . . 3-1
- 4. ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Overview . . . . . . . . . . . . . . . . . . . . 4-1 4.1.1. Objectives . . . . . . . . . . . . . . . 4-1 i 4.1.2. Technical Approach . . . . . . . . . . . 4-1 4.2. Plant Data Review . . . . . . . . . . . . . . . 4-2
~
4.2.1. Overview of Data . . . . . . . . . . . . 4-2 4.2.2. Data Evaluation . . . . . . . . . . . . 4-2 4.3. POWER TRAIN Analysis . . . . . . . . . . . . . . 4-8 4.3.1. Analysis Method . . . . . . . . . . . . 4-8 4.3.2. Results . . . . . . . . . . . . . . . . 4-9 4.3.3. Overshoot at 2355 psig vs 2300 psig . . 4-9 4.4. Statistical Evaluation of Data . . . . . . . . . 4-10 4.4.1. Method . . . . . . . . . . . . . . . . . 4-10 4.4.2. Results . . . . . . . . . . . . . . . . 4-11 4.4.3. Details of the Monte Carlo Analysis . . 4-11
- 5. NRC REQUIREMENTS / PREVIOUS ANALYSIS COMPARISONS . . . . 5-1 l
I 5.1. Guidelines . . . . . . . . . . . . . . . . . . . 5-1 5.2. Current Study Results . . . . . . . . . . . . . 5-1 5.3. Comparison of Current / Previous Analysis . . . . 5-2 5.4. Summary . . . . . . . . . . . . . . . . . . . . 5-3
- 6. REFERENCES . . . . . . . . . . . . . . . . . . . . . . 6-1 List of Tables Table 1-1. Successful Runback Actions (Pre 1979) . . . . . . . 1-2 2-1. Reactor Trips Having Good Potential for Avoidance with Increased Satpoint . . . . . . . . . 2-5 3-1. PORV Opening Frequency . . . . . . . . . . . . . . . 3-3 4-1. Plant RPS Settings for High Pressure Trip . . . . . 4-15
-i-a l
t ll d 1 3
; List of Tables (Cont'd)
Table Page j 4-2. Plant Data for High Pressure Trips . . . . . . . 4-16 4-3. Sammary of Major Inputs to Monte Carlo Simulation Program . . . . . . . . . . . . . . . 4-17 I List of Ficures i Figure l 2-1. Average Number of Hig,h Pressure Trips 3 Per Year . . . . . . . . . . . . . . . . . . . . . 2-6 2-2. Average Number of High Pressure and ARTS Trips Per Year . . . . . . . . . . . . . . . 2-7 2-3. Anticipatory Reactor Trips on Turbine Trip For B&W Plants (1980-1984) . . . . . . . . . . . . 2-8 4.3-1. POWER TRAIN vs. Plant Data . . . . . . . . . . . . 4-18 4.3-2. POWER TRAIN Results . . . . . . . . . . . . . . . 4-19 4.4-1. Monte Carlo Results . . . . . . . . . . . . . . . 4-20 4.4-2. Overshoot Distribution for Monte Carlo Input . . . 4-21 4.4-3. Model for Monte Carlo Analysis . . . . . . . . . . 4-22 9 d I I 5 i l l l 1 h
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-- -= - . - .- -_-_
t Executive Summary
. This report documents the results of an engineering evaluation performed to justify increasing the setpoint for reactor trip on ,
high pressure from its current value of 2300 psig to its pre-1979 value of 2355 psig. The potential benefit of the change is a reduction in the frequency of reactor trips. The conclusion of this study is that the high pressure trip setpoint can be raised to 2355 psig with negligible impact on
; the frequency of opening the PORV during anticipated overpressurization transients. The sum of the transient induced
- pressure overshoots
- and the total effective instrument string errors would result in the PORV (set at 2450 psig) opening one
- time out of 100,000 high pressure trip transients.** This is a sufficiently low probability to assure that a normal high pressure trip would virtually never open the PORV and, conse-
] quently, will satisfy the NRC small break LOCA (SBLOCA) criteria for the PORV. Also, since 2355 psig is the design high pressure trip setpoint, the original FSAR analyses remain applicable for this setpoint.
Of the 47 high pressure trip transients reviewed in detail for this study, twelve were judged to have high potential for l avoidance if more margin to reactor trip had been available. In addition, if the high pressure trip setpoint were raised to 2355 psig, B&W plants would be capable of surviving turbine trips i e
- Overshoot is the difference between the maximum indicated RC pressure and the actual trip pressure.
**In five years (1980 through 1984) all B&W NSS plants experi-enced a total of ~65 high pressure trip transients. This is f less than 2 trips per plant per year.
- 111 -
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I 2 [ without reactor trips at power levels well above the current 20% arming threshold for the anticipatory reactor trip (ART) on
! turbine trip. Raising the ART arming threshold to a level t
consistent with the runback capability at the 2355 psig trip g setpoint would have avoided twelve .of the reactor trips on turbine trip experienced during 1980-84. Based on 1980-84 l experience, the results of this study suggest that the potential ! benefit of raising the high pressure trip setpoint to 2355 psig 3 is a 10% reduction in the average reactor trip rate for B&W plants. The NRC requirements regarding the PORV are: (1) probability of SBLOCA due to stuck open PORV must be less than .001 per reactor year and (2) less than 5% of the high pressure trips are allowed to open the PORV. Both requirements are met with the high pressure trip setpoint at 2355 psig. The methods used for this analysis are similar to previous analyses on this subject (see References 2 and 3). A major I difference is that this study predicts overshoot based on actual plant high pressure trip transients in which the PORV did not open. The previous analyses calculated overshoots (for a closed PORV) based on plant data where the PORV opened prior to reactor trip. Compensating for the open PORV added more uncertainty to these past calculations. The data used for the current analysis provided a more realistic overshoot and probability distribution j of the overshoot. The comparable results of these two studies are 2.2 x 10 ~4 PORV opening per high pressure event (old) verses 1.0 x 10-5 PORV opening per high pressure event (new). l l' l
- IV -
1 l
-- - ,. .g.. . - - - - , - .
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- 1. BACKGROUND 1
The B&W NSSS, with its once Through Steam Generators (OTSG) and Integrated Control System (ICS), was designed with the capability to adjust to minor plant upsets and certain anticipated events without a reactor trip. The system was designed to initiate a plant runback, upon detection of an upset or equipment malfunc-tion, to a power level consistent with the limiting condition. The ef fectiveness of this runback capability is illustrated by Table 1-1. Table 1-1 was extracted from reference (1) and shows 47 successful runback actions for one B&W station for 5.5 years of reactor operation prior to 1979. Following the TMI-2 accident, several changes were made to B&W j plants for the purpose of reducing challenges to the Pilot Operated Relief Valva (PORV). These changes have been effective in reducing challenges to the PORV but have also resulted in a I substantial reduction in runback capability for B&W plants. The changes which have had the greatest impact on runback capability
/ are:
- Raising the PORV opening setpoint from 2255 psig to l 2450 psig
)
- Lowering the setpoint for reactor trip from 2355 psig to 2300 psig 1
- Implementation of an Anticipatory Reactor Trip (ART) on turbine trip for reactor power levels of 20% and above.
A consequence of this loss of runback capability is that events i' such as the loss of 1 of 2 main feedwater pumps or minor feed-water upset's now nearly always result in a reactor trip. The B&WOG is committed to reducing reactor trips to improve plant availability by keeping plants on line and to improve safety by h 1-1 1 1.
[ reducing challenges to safety systems. As part of a generic trip reduction program, the B&WOG has identified the restoration of runback capability as a prime candidate for achievement of trip f reduction. Accordingly, the B&WOG has undertaken various studies to: Identify the current constraints on runback capability Examine the feasibility of removing or relaxing these constraints Identify modifications which would enhance the prob-ability of successful runbacks. Included in this effort was a study to examine the potential benefits and feasibility'of restoring the setpoint for reactor trip on high pressure to its original value of 2355 psig while keeping the PORV setpoint at 2450 psig. The results of this study are reported herein. Table 1-1. Successful Runback Actions (Pre 1979) Initial power Final power Event h rance rance % h* Turbine trip / load rejection 14 15-100 2-22 Feed pump trips (1 of 2) 14 100 55-75 Rod drops 4 100 50-60 10% step load increases 4 17-90 27-100 10% step lead decreases n 37-100 15-90 Total 47 1 i l' g 1-2 1
1 I i
- 2. POTENTIAL BENEFITS 2.1. Incact of Post-TMI Chances In an effort to identify the potential benefits of restoring the
. high pressure trip setpoint to its original value, a detailed comparison of pre- and post-TMI trip frequency was made. A primary objective of this comparison was to determine the impact of the changes on high pressure trip frequency.
Figure 2-1 displays the average high pressure trip frequency for B&W plants for the pre-1979 and post-1979 periods. The pre-1979 data for each plant includes trips occurring from the date of commercial operation through the and of 1978. The post-1979 data includes trips from 1980 through 1984. The trips which occurred f in 1979 were not included in the comparison as the setpoint changes were implemented at various times during the year. Figure 2-1 shows that high pressure trip frequency increased for some plants and decreased for others. The average for B&W plants as a group shows no change pre-1979 versus post-1979. This simple comparison of pre- and post-1979 high trip rates does not, however, accurately reflect the impact of the changes. The , I reason for this is the effect of the addition of the Anticipatory i Reactor Trips which occurred in 1979. These trips, turbine trip (power > 204) and loss of both feedwater pumps, pre-empt high pressure trins in the post-1979 period. These trips, in the l post-1979 period, are in fact anticipatory high prassure trips. If the ART were not in place, a turbine trip or loss of both feed pumps event would result in a high pressure trip with the current setpoints. Thus, a more reasonable comparison, for assessing the impact of post-TMI changes on trip frequency, is to compare pre-l 2-1
1979 high pressure trips with post-1979 high pressure plus ARTS trips. 1 i When this comparison is made, a different picture emerges, as shown in Figure 2-2. This figure shows that, when trips due to the same causes (turbine trip, loss of both MFWPs, other) are compared, the average trip frequency due to these causes has { doubled for B&W plants as a group since the implementation of a post-TMI setpoint changes. Figure 2-2 also shows that some 1 plants have been affected more than others by the changes. The t reason for this is that individual B&W plants differed in pre-1979 runback capability. Those plants which had greater runback capability show the larger increases in trip frequency due to the loss of this capability. 2.2. Potential For Tric Reduction As noted previously, two key factors contributing to the reduc-tion of runback capability were the raising of the PORV opening setpoint and the lowering of the high pressure trip setpoint. It is recognized that, prior to TMI-2, the PORV played a significant role in keeping primary pressure below the trip setpoint during rurbacks. For this reason, restoration of runback capability to pra '.MI levels cannot be expected with the PORV set to open at
; pressures above the high pressure trip setpoint. However, a partial restoration could be achieved if the high pressure setpoint were restored to its original 2355 psig value. The ,
I basis for this statement is as follows: A. Increasing the high pressure setpoint to 2355 psig would l
! increase the operating margin to high pressure trip by 55 psi. For slow moving transients resulting in primary i
i pressure increase rates of 1 to 2 psi /second, the increased margin'would provide an additional 30 seconds to one minute I f bef' ore reaching the trip setpoint. This would provide additional time for an operator to determine the cause of an upset r .1 take manual actions to avoid a trip. Examples of such ' : ,ns are switching to an alternate input signal 2-2 . 1
I 8 f upon discovery of a failed input, initiation of pressurizer spray, or placing appropriate portions of the ICS in manual. It is not possible from a review of past transients to establish with certainty that a given trip could have been avoided by operator action if more margin were avail-able. A qualitative indication of trip reduction potential was developed, however, by categorizing previous trips according to their potential for avoidance by operator action. For example, a fast moving transient such as that initiated by a load rejection (CR-3, 7/5/82) was judged as having poor potential for avoidance. A slow moving transient which might be caused by a failed ICS input (CR-3, 7/29/83) was judged as having good potential for avoidance. The 47 high pressure trips for which sufficient data was available were categorized as described above. This review identified 12 trips as having good potential for avoidance through operator action if increased margin were available. These
; trips are identified in Table 2-1.
B. In a separate study, B&W demonstrated by analysis that if the ] high pressure satpoint were raised to 2355 psig, B&W plants I could survive a turbine trip from 40% power without a reactor g trip. Thus, if the high pressure setpoint were raised to I 2355 psig and the power level for arming the Anticipatory Reactor Trip on turbine trip were raised to a value above 40%, turbine trips occurring at powers below 40% would no longer result in reactor trips. Figure 2'3 indicates the potential reduction. Assuming successful runbacks for all l f turbine trips at initial power levels s 404, 12 of the 1980-1984 trips could have been avoided. These trips are also listed in Table 2-1. L C. Had the 24 trips listed in Table 2-1 been avoided, the average trip frequency for B&W plants for the 1980 through i 1984 period would have been 4.6 trips / year / plant instead of 5.3 trips / year / plant. This represents a 13% reduction in 2-3 1
i
, average trip frequency. Assuming a future mix of transients similar to that of the 1980-1984 period, it is reasonable to i project a reduction in average trip frequency of approxi-mately 10%.
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I i Table 2-1. Reactor Trips !?aving Good Potential for Avoidance with Increased Setooint i l Hich Pressure Tries Date Plant Initial Power Comment 3/19/80 CR-3 53 Maneuvering /FWP Trip ( 12/5/80 0-2 14 Manual Underfeed 9/5/81 0-3 30 FW Valve Sticking
. 6/23/81 RS 50 Manual FWP Operation, Flow ,t Oscillations 6/17/81 RS 80 MFWP Trip 8/8/82 CR-3 94 Loss of RC Flow Signal 7/15/82 CR-3 100 Loss of RC Flow Signal 11/4/82 0-2 100 MFW Pump Trip 5/21/82 0-1 90 Runback on Rod Drop 3/24/82 0-1 51 Turbine & FW A Control in Manual 7/256/83 CR-3 9 MFW Flow Reduction 12/3/83 RS 65 MFW Upset, Valves & Pumps in Manual Anticinat y Reactor Trics on Turbine Trie D.A13 hant Initial Power 11/6/80 DB-1 18 3/14/81 0-3 4 1
1/1/82 0-1 24 1/1/82 0-1 20 1/2/8p O-1 28 f 5/19f82 0-2 20 7/9/82 CR-3 11 - [ 7/9/82 CR-3 , 18 6/1/83 0-1 15 I 8/6/83 RS 20 8/10/83 RS 40 6/7/84 0-3 19 2-5 1
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m - - - - - - _ . . - .. . . . . - - - - . . .. _ Figure 2-3. Anticipatory Reactor Trips on Turbine Trip for B&W Plants (1980-1984) 30 - 25-- g 20-- b 5 y B , . ... ce g n Z* 10-5-- 3 3 o . 0-30 31-40 41-50 51-60 61-70 71-80 81-90 91-100 INITIAL POWER LEVEL (%) . i
I
- 3. IMPACT ON PORV OPENING FREQUENCY The high pressure trip setpoint was lowered to 2300 psig and the PORV opening setpoint raised to 2450 psig to reduce the prob-ability of opening the PORV for several categories of events.
l These are: A. Overpressure transients B. Transients with Delayed Auxiliary Feedwater C. Operator Action Under ATOG Guidelines D. Instrumentation and control faults E. Overcooling transients that initiate HPI Table 3-1 lists the PORV opening frequencies for these events as estimated in previous studies (References 2&3) and the current study. Two important points are illustrated by this table:
- 1. Only the PORV opening frequency for overpressure transients is af fected by the high pressure trip setpoint. Opening frequencies for the other events are independent of the high pressure trip setpoint.
- 2. The PORV opening frequency due to overpressure transients is a relatively small contributor to total PORV opening fre-quency, the total PORV cpening frequency is dominated by the frequency of opening due to operator actions under ATOG.
The current study concludes that the PORV opening frequency for overpressure transients is 1.86 x 10 -5 openings per reactor year with a high pressure trip setpoint of 2355 psig and a PORV opening setpoint of 2450 psig. With the current 2300 psig high prescure trip setpoint, the frequency would be less than 1.86 x
-5 The total PORV opening frequency for all other causes 10 .
(Reference 3) is 3.06 x 10 -2 If we assume the PORV opening frequency for overpressure transients to be zero at the 2300 3-1 1 (
I < f psig trip setpoint, the following comparison can be made: Total PORV = 8.06 x 10-2 + 0 = 8.06 x 10-2 I Opening Frequency ! at 2300 psig setpoint
~
Total PORV = 8.06 x 10" + 1.06 x 10 8.06 x 10 -2 Opening Frequency at 2355 psig setpoint Thus, raising the high pressure trip setpoint from the current f s value of 2300 psig to the original value of 2355 psig will have a 5 negligible impact on the frequency of PORV openings. Section 5.0 of this report discusses dif ferences between the current study results and the results of the previous study. e 1 I I I I . I 2-2 I
I Table 3-1. PORV Ocenina Frecuency Ocenina Per Reactor Year B&W '81 TER I (Ref. 2) (Ref. 3) Current Studv i Overpressure -5 l transients 3.9x0
-5 2.2xlO -3 2x10
, Transients with ,4 _4
\ delayed aux. FW 7.6x10 7.6x10 No change i
Operator actions 00-2 -2 under ATOG 1.58x10 7.7x10 No change
-3 ~
I&C faulcs 1.7x10 1.7x10 No change i Overcooling transients that
~4 ~4 initiate HPI 8.4x10 8.4x10 No change
-2 ~
-2 Total 1.9x10 8.28x10 8.06x10 t
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- 4. ANALYSIS ,
l 5 4.1. Overview 4.1.1. Obiectives i
'. The purpose of this analysis was to determine if the high i
4 pressure trip setpoint could'be raised to a value above the present 2300 psig setpoint without exceeding NRC guidelines for
. PORV opening frequency. The maximum allowable set point was based on the criterion that the probability of opening the i PORV will be sufficiently low so as not to exceed the maximum
! allowable small break LOCA limit of 10-3 events per reactor
. year. (This SBLOCA limit for the PORV was set by the NRC per Reference 4).
f 1 4.1.2. Technical Accroach The main thrust of this analysis was to determine the maximum
] reactor coolant (RC) pressure relative to opening the PORV during over pressure transients. High pressure transients referred to
{ ' in this document do not include cases with excessive HPI or total loss of main and auxiliary feedwater. These transients could open the PORV regardless of the high pressure trip setpoint although the PORV opening would occur well after reactor shut-down. This analysis included a statistical evaluation of instrumen-l tation errors (associated with the trip set point and PORV),
- and the amount of indicated RC pressure increase above the trip j set point (overshoot). The transmitter and instrument errors I were based on actual plant set point and calibration data
! and instrument design specifications. The maximum indicated system pressure (overshoot) was evaluated based on (1) plant l , 4-1 I _ . - _ _ - - _ _ - . _ _ - . - - _ _ _ _ . -_-_-- i.
l I j data from Transient Assessment Program reports (Tap Reports) , and (2) information derived from " Digital Power Train" computer
! code analyses.
1 The sequence of the analysis is outlined below.
- 1. First, plant transient data was reviewed to determine the expected range of overshoots during the anticipated over
, pressure transients. This data was all based on the current 2300 psig trip setpoint (section 4.2).
; 2. Secondly, Power Train analyses were performed to determine a whether the overshoot would change when the high pressure trip setpoint was increased (section 4.3).
- 3. Next, this overshoot data was statistically combined with instrumentation errors to predict the probabilities of the PORV opening from various high pressure trip setpoints (section 4.4).
- 4. Next, these results were compared to NRC limitations on the PORV and previous analyses on this subject (section 5).
4.2. Plant Data Review 4.2.1. Overview of Data Various plant data were reviewed. The data included high pressure trip setpoints, calibration tolerances and errors associated with these set points, and TAP reports involving over-pressure transients. The trip setpoints and calibration data were used to determine the pressure at which the reactors actually tripped. This data was then factored into the TAP f report evaluation to determine the overshoot. . 4.2.2. Data Evaluation I 4.2.2.1. Mich Pressure Setcoints As previously discussed, the existing nominal high pressure trip I setpoint is 2300 psig. The method used to insure that this satreint is not exceeded is outlined below. I 1. Pressure transmitters are calibrated during refueling. 4-2 I
1 1
- 2. Each of the 4 high pressure trip set point channels are initialized at a value that allows for drift and instrument string errors. (See Table 4-1)
- 3. RPS channels are tested periodically througnout the sub-sequent cycle by artificially inducing a ramping voltage to the RPS bistable and determining the trip voltage (i.e.,
pressure) for the channel. The set points and error limits based on this procedure are listed in Table 4-1. The data in this table are based on conversations with B&W site personnel. As can be seen by the data in Table 4-1, the actual trip set point varied from 2285 to 2298 and the' reactors consistently trip within 2 or 3 psi of the set points. Therefore, the initial RC pressure used to calculate overshoot was assumed to be the specific trip set point of each plant. However, a uniform distribution of zero to plus five psi error was assumed for the trip set point for predictions of maximum RC transient pressure 1 during the Monte Carlo simulation (see section 4.4).
, 4.2.2.2. Indicated System Pressures Durina Transients The maximum indicated RC pressure during an overpressure trans-isnt would include:
- 1. actual overshoot that occurred due to the nature of the transient,
- 2. any additional instrument string error downstream cf the RPS channel, f 3. uncertainties due to the print out device readability,
, 4. uncertainties due to the frequency at which the data were recorded.
g ! Items 2, 3, and 4 are function- of the device used to monitor pressure and are* discussed in Section 4.2.2.4. f It was not apparent what part of the indicated RC pressure during
,a transient was due to actual overshoot and what part to any f
4-2 1 . 1
1 instrument errors. Therefera, the indicated maximum RC pressure minus the trip set point was all assumed due to overshoot. This 1 1 ) is a conservative assumption because any instrument error that may be included in this overshoot is redundantly covered in ( o Section 4.4 (statistical evaluation of data). l
\ l 4.2.2.3. Overshoot O
Overshoot is the maximum RCS pressure in excess of the actual trip pressure. It is primarily a ' function of the relative rates l
} of heat transfer in the steam generator compared to heat transfer in the core region - up through the first few seconds after the trip. The relative rate of heat transfer determines the rate of RC pressurizations. The delay time between the initiation of the trip signal until the core thermal power is adequately decreasing dictates how long the heat up continues. This time delay is usually one to four seconds after the trip. The maximum pressure
] that occurs during this time is dependent on the pressuriza-l tion rate prior to the trip. This pressurization rate varies ! from 2 to 3' psi /second for a mild feedwater upset to 40 psi /sec-ond for some load rejection transients. j 4.2.2.4. Instrumentation Accuracy { i The accuracy of the indicated RC pressure is primarily dependent i on whether it is a narrow range (NR) or wide range (WR) signal. The NR signal is usually more accurate than WR. However,
; determining a relatively accurate peak RC pressure during a transient may be more dependent on the recording device. This is I
g becausa the readability of the device or the frequency of the s measr.rement may introduce more uncertainty than the signal itsalf. For example, the frequency of the plant computer s'anitoring (every 15 seconds) could completely miss the peak RC pressure during a load rejection since the peak pressure occurs in about 5 seconds into this event. Similarly, even though the l analog signal to an indicator or recorder may be very accurate, the scale of the readout device may make it difficult to read the peak pressure within 10 psi. Because of these types of problems,
- g 4-4
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the peak pressure from each high pressure trip transient was evaluated based on the particular data available. Narrow range signals were used where available. If more than one channel or device was available, the data was compared. Extrapolation or curve fitting was applied when the appropriate data were avail- ' I able. Using these methods for plant data evaluation, the following general observations were noted: 1 1. Peak pressure from extrapolated digital output often agreed , i well with analog strip chart data.
- 2. NR pressures from the A and B loops were usually within a few psi of each other.
e
- 3. Wida range pressure changes during a transient were about the same as narrow range pressure changes.
These results led to the conclusion that, in general, the overshoot predictions were relatively accurate and that the degree of overshoot was primarily due to the transient itself and not due to any large instrument errors. 4.2.2.r. Monitorine Device The plant data evaluation was based on five different monitoring devices. These included:
- 1. Reactimeter
{
- 2. Plant Computer
- 3. Special Transient Monitoring Equipment (i.e. Test Transient Monitor (TTM) at Oconce or the post trip review at CR-3)
- 4. Control Room Alarms Data
- 5. Strip Charts The following is a brief description of the data from each of i these devices:
?
l i l
\
l 1 4-5 i I e
- r. - - - - - - - - -,---r-- - - - - . , - , - - , - - , - , - - - - , . - - - , - - - , - - , - , , . + , , , ,,--n-,,.--..-,---ru ,.--,,v., -..,- - , - - _ _-en-. -
Reactimetez The reactimeter is a digital device that is typically set to track plant parameters at frequencies down to 2 seconds.* It takes a " snap shot" of the data at each time step. .The reacti-meter is usually connected to NR channels. Plant comeuter i , . The Plant Computer is functionally similar to the reactimeter except for the sampling frequency. The data is averaged and 1 l printed at intervals of 15 to 30 seconds. The usefulness of this data for this study depended on the plant, the circuntance, and the timing of the data. Test Transient Monitor The TTM at Oconee is similar to the plant computer except it monitors data at one second intervals. However, the output format is usually in a graphical form (in the TAP Reports) with data points often plotted every 6 to 12 seconds. The frequency j of plotting is sometimes too coarse to accurately monitor peak RC i pressures during many transients. However, some of the trans-y lents reviewed had listed or plotted RC pressure data on a one, two or three second interval. These cases provided accurate resolution of the time delays to peak pressure during the t transient. Alarms Data i I Control room alarms data is a snapshot taken at the exact time a
- g limit is reached or in some cases it prints an "on demand"
- a listing of specific data groups. The usefulness of this data for maximum RCS pressure predictions also depends on the plant, the timing, and the circumstances.
1 i I .....
*The reactimeter is capable of one second data monitoring but usually only sat at this fast frequency for test situations.
4-6 4
- I i
I
Strio Chart Data Strip Chart Data is an analog representa tion of the assigned plant parameters. Evon though it is a real time analog signal, its time scaling is so coarse that it does not give a good profile of the faster transients. Strip chart data is very good, however, for showing maximum and minimum values. Each plant has its own combination of devices and channels for monitoring transient data. As pointed out in the above discus-sion, each recording device has some disadvantage for determining peak RC pressure. However, a combination of data from two or more devices can give added confidence in predicting peak RC transient pressures.
, 4.2.2.6. Oceratina Plant Data Table 4-2 is a summary of the high pressure trips in une past few years for which the best data was available. This table is based on TAP report data. The maximun RC pressures were determined using the methods previously discussed.
{ The conclusions drawn from a review of the data in this table 1 are: I 1. The three primary categories of high pressure trip events are (a) total or partial loss of feedwater, (b) feedwater/ power mismatches during runbacks, and (c) load rej ections/MSIV closings. Many of these events were initiated by an errone-l ous ICS input signal. Also, the RCS pressurization rate of these transients varied from 2 psi /sec to 40 psi /sec. This corresponds to a maximum transient time
- of about 2.0 j minutes down to a minimum time of 5 seconds.
- 2. The maximum transient pressure overshoots usually occurred I during the loss of load transients (see transients 1/23/85, 4/11/85, 6/20/82, 1/19/82, 2/5/80).
l *The transient time is from the initiation of the event to the I time of maximum RC pressure. ( l 4-7 I
i
- 3. For the 47 overpressure transients listed, the overshoot I distribution is as follows:
Overshoot Number of 1 Rance (esi) Events 0 - 10 33 1 11 - 20 7 1 21 - 30 1 g 31 - 40 3
# 41 - 50 2 51 - 60 1 This data was used to construct the overshoot distribution l
3 for the Monte Carlo analysis.
- 4.3. POWER TRAIN Analysis i
The five years of plant daa discussed in Section 4.2 provide a sound basis for characterizing the overshoot distribution with a 2300 psig trip setpoint and no PORV opening. The objective of
. the POWER TRAIN analysis was to determine whether or not this overshoot distribution would change with a 2355 psig setpoint.
4.3.1. Analysis Method j The approach to this task was to: I o Select a representative transient for evaluation i o Develop a 177 FA plant model for evaluation l o Compare the model's predictions for a setpoint of 2300 psig against predictions for a 2355 psig setpoint I The plant data from the 47 trip events indicates that a turbine j trip event from full power with no anticipatory trip causes the largest RCS pressure overshoot. This transient represents a l severe primary to secondary power mismatch during those first few ,
- seconds of the event when the steam generator nucleate boiling region is suppressed. Therefore, it is a logical choice for
! evaluation. l The second step of the evaluation was to compare the model's ) results against plant data for a setpoint of 2300 psig. For all I cases that were run, the POWER TRAIN model overpredicted the peak RCS pressure by approximately 20 to 60 psi. A typical comparison
~
4-8 l l . _ _ .-- _- _ _ _ . _ _ _ _
i l of POWER TRAIN results with plant data is shown in Figure 4.3-1. Although duplication of the exact RCS pressure profile is not essential, the goodness of agreement between the model results l ] and the plant data add confidence to the results. The third part of the evaluation was to simulate the turbine trip event with a high pressure trip setpoint of 2355 psig and compare the overshoot to the overshoot generated at 2300 psig. The I difference between the overshoots for the two setpoints is the result of primary interest. 4.3.2. Results ] POWER TRAIN predicts an overshoot of 50 psi for the turbine trip I (and no ARTS trip) with the high pressure trip setpoint of 2300 psig. This compares with an observed 40 psi overshoot for similar plant transients. POWER TRAIN predicts an overshoot of 30 psi for the same event but with the trip setpoint at 2355 psig. A comparative plot of these results is depicted in Figure 4.3-2. ! The conclusion from this comparison is that the pressure over-
, shoot is a weak function of the high pressure trip setpoint in j the 2300 to 2355 psig setpoint range.
The results suggest that the overshoot actually decreases as the setpoint is raised. For purposes of this evaluation, however, l the overshoot distribution was assumed not to change. This conclusion can be explained in terms of the physical phenomena affecting the heat transfer rate in the steam generators, as discussed in the following paragraph. I 4.3.3. Overshoot at 2355 esia vs 2300 nsig The RCS pressure overshoot is a function of the steam generator heat transfer rate. Since the heat transfer is predominantly in the nucleate boiling region of the steam generator (SG), the relative height of this region throughout the transient is an indication of the rate of SG heat transfer and consequently RC pressure. The nucleate boiling (NB) region is in turn dependent 4-9 I
I g on the SG pressure and steam flow rate. As the SG pressure increases during the transient, the NB level tends to decrease I thereby decreasing heat transfer and increasing RC pressure, i However, as the various steam reliefs open, the flow increases and the NB level increases promoting more SG heat transfer. When l the RCS trips at 2355 psig, the steam line is already at or very 9 close to the setpoint for the first bank of main steam safety I valves. When the MSSVs open, the steam flow rate starts increas-ing sharply. This increased steam flow increases the NB level and quickly turns RC pressure around. The same transient with a 2300 psig trip setpoint requires more post trip time to reach the MSSV opening pressure resulting in a longer time to turn the RC pressure around. 4.4. Statistical Evaluation of Data The probability of opening the PORV was estimated using Monte Carlo simulation methods. This technique is described in reference 5. The procedure and assumptions used are discussed i generally in section 4.4.1. The details are covered in section 4.4.3. The conclusion of this activity is that there is very little chance that the PORV will be called upon to actuate at any high pressure reactor trip setpoint below 2375 psig for the transients covered by this study. The original plan was to start with the trip setpoint at 2300 psig and proceed to 2355 psig in 5 psi increments. It was noted, however, that there would be no PORV actions generated in the lower pressure ranges. Therefore, in doing the study, it was necessary to turn the problem around in order to find situations in which the PORV would be actuated. This was done by varying the reactor high pressure trip setpoint from 2355 to 2375 psig. 4.4.1. Method The Monte Carlo simulation accounts for the effects of errors on the pressure signals to the RPS and the control of the PORV. Instrumentation errors are a combination of signal processing I errors associated with the individual RPS channels and set points 4-10 l I
. e f for the decision to trip the reactor and analogous errors associated with the PORV. In addition to these errors, the study accounts for the pressure overshoot phenomenon.
1 The model is straightforward. It is assumed that there is a single true system pressure observed at each of the four narrow range pressure sensors. This true value is processed by
' each of the RPS channels and is subject to instrumentation errors t
in individual channels. This information is used to determine when, in terms of true pressure, the RPS trips. Of the four RPS {N i channels, two are available for NNI signal processing. (For the j Oconee plants, the signal to the NNI is selected from a fifth (inactive) RPS channel. As the errors for this channel are the same as for the active channels, the analysis results are
; applicable.) Of these two potential NNI signals only one is used by the NNI for processing to the PORV bistable. This study compares the output of these two channels and conservatively selects the larger indicated pressure to activate the PORV.
} 4.4.2. Results The results of the Monte Carlo simulation approach indicate a very low probability of opening the PORV. For a trip setpoint corresponding to a true system pressure of 2355 psig, one trial in 100,000 resulted in opening the PORV. The average margin for PORV actuation relative to true pressure was 66.6 psi, with a standard deviation of 13.9 psi. Additional runs at 2365 psi and 2375 psi are included in Figure 4.4-1 to exhibit behavior of the system. The results of these two cases show that 55 PORV openings occurred at the 2365 psi set point, and 1485 occurred at the 2375 psi setpoint.
4.4.3. Details of The Monte Carlo Analysis The direct Monte Carlo procedure used in this study will be detailed in this section. A simple drawing of the process to be modelled is shown in Figure 4.4-3. I 4-11 I l l .- _ -. . - - .
4 0 t . The objective of this study was to evaluate the likelihood of opening the PORV at various RPS setpoints larger than the current nominal value of 2300 psig. A computer program was designed to perform a large number of Monte Carlo experiments for 2355, 2365, and 2375 trip setpoints. Each individual Monte Carlo trial incorporates the error l contributions of the signals to the RPS and to the PORV. The largest error sources are the RPS, the NNI, and those associated
}
with the high pressure trip and the PORV set points. Another i significant contributor is the pressure overshoot, which is treated as a physical phenomenon by a probability model and not as an uncertainty. The various error contributions used for the analysis are summarized in Table 4-3. The RPS contributes the major portion of the instrument errors. These errors include (1) the sensor / transmitter through the t bistable with its internal setpoint, (2) the NNI string to the PORV and (3) an additional conservatively assumed error applied to both the RPS and PORV setpoints. The RPS errors were taken from reference 6. This study determined a bias component and a random component for the sensor / transmitter through the RPS high pressure setpoint under the B&W recommended method for string error treatment for operating units. The bias contribution was treated as a constant 4.83 psi and was applied in the conservative direction. The I random part of the RPS error was an assumed normally (Gaussian) distributed random variable with a zero mean and a standard deviation of 6.2 psi. The NNI buffer amplifier and the NNI bistable errors are both treated as normal random variables with means equal to zero and standard deviations of 1.4 psi and 1.3 psi respectively. l This represents an assumption regarding the type of distribu-tions for these errors. I 4-12 i
1 l l l i The conservatively assumed error associated with the PORV and the RPS trip set point is treated as a uniformly distributed random contribution in the range from zero to five psi. It l f inflates the pressure at which the reactor will trip and decreases the pressure at which the PORV will open. These errors are included for conservative contributions due to i set point variations. I The pressure overshoot is a physical phenomenon which is i associated with high pressure trips. The probability model used for the pressure oversh'oot is an exponential distribu-tion truncated at 10 psi and 60 psi. This distribution is based on plant data (Table 4.2) and is represented in Figure 4.4-2. While uncertainty of pressure overshoot is not directly treated, the distribution parameter is conservatively estimated as shown in Figure 4.4-2. Any case that predicted an overshoot less than 10 psi was conservatively redone and any case that predicted overshoots
! greater than 60 psi were truncated to a 60 psi value. This a
tended to generate higher overshoot cases than the actual distribution would predict. The program treats the single true system pressure input to each of the four RPS channels with individual channel processing errors. The result is four individual values of sensed pressure. Each of these is identified with a channel. In each I channel there is a bistable with an internal setpoint. Each of the channel set points is also treated with its own random setpoint error mentioned above. Trip logic is met if two or more of the sensed pressures exceed or equal the actuation values of RPS setpoint plus the O to 5 psi random portion. If this trip l condition is not satisfied on the first pass through, the true pressure is incremented by two psi and another iteration is made , on the trip logic. This continues until the trip criterion is met. I 4-13 i I
f l I e At this point in a particular trial, the reactor has tripped. The true pressure which yielded the trip was then incremented by the overshoot. This new value (original true pressure plus l overshoot) was then subjected to the RPS error treatment. As 4
) noted previously, one pressure signal can be selected from I
either of two RPS channels for transmittal to the NNI PORV i circuit. The Monte Carlo program was set to select the highest
, of the two RPS channels available to the NNI. Thus, the pressure signal selected for the NNI in a given trial was the highest ! value of (true pressure + overshoot) + RPS errors for the two channels available. This value of indicated pressure was then treated with NNI buffer amplifier errors and sent to the PORV bistable to determine if a PORV actuation was required. The comparison at the bistable incorporated the zero to minus five psi error previously mentioned.
1 The margins to PORV actuation for these cases were grouped in 5 psi ranges to provide estimates of the probability of opening, i or being near to opening, the PORV. This data is shown in Figure 4.4-1. I i 4 I l 1 1 . u 4-14 e II l
. . l I
c t'
, Table 4-1. Plant RPS Settinas for Hioh Pressure Trio
, Rancho Oconee g
Seco_ (1.2 & 3) ANO-1 CE-1 DB-1 Trip setpoint (psig) 2290 2290 2298 2296 2285 2 Max expected string error I 2.96 4.0 at histable 2.4 1.55 2.0 during weekly (RMS) (RMS) (RMS) (Algebraic
. checks (+/-psi) Sum) i l
1 I s I t s 4-15 I .
Table 4-2. Plant Data for Hiah Pressure Trips i AVERACE MAX RCs DELTA I Petssuet CVEtsM007 PetsSutt lufflAL RECOR0!aG Daft L117 (Ps!G) (Pst) (Pst/stC) POWER OtVICE CO M ets 2/05/80 08 1 2336 51 37.0 100 ma REACflMtitt fuse Govttmas VALyts ClostD 2/26/80 Ca 3 2320 24 "" 100 STRIP CMAAf Loss of mal X
! 5/11/80 as 2290 0 "" 11 at TAP Plot al CAL Wifu ROD != nam.
5/30/80 as 2300 10 10.0 95 m atACTIntite mAmuAL aussACE 6/26/80 Amo 1 2310 12 "" 100 m sitIP CnAaf LOAD REJ., MAN tum4ACK AfftmPT 8/12/80 Ca 3 2300 4 "" 53 na stalp CnAnf mN PWP TRIP 4/22/80 Amo 1 <2300 <2 .". 90 mR ALAams NN PWP TRIP 8/29/80 Ca 3 2300 4 "" 72 me,15 stC DATA 'es 00vtance VALyt Peosttus 10/28/80 Rs <2300 <10 "" 94 na sit!P CnAaf Att FAILuet, MN v&Lvts LOCKto UP 11/08/80 Os 1 2300 15 7.0 28 se REACfingite N 05CILLAflows 11/12/80 08 1 2334 49 10.6 40 me REACfintitt Loss of ICs luPuis Fim RPS 12/05/80 02 2291 1 4.2 14 us,ffm(1stC) nAmuAL umotartto 4/08/81 Am0 1 2300 2 6.0 98 ut FAILto si Pwie sicmAL 4/11/81 Ca 3 2305 9 10.0 100 he TAP PLOT L0ss Of POWER TO uni Y 6/17/81 as 2300 to 2.5 80 he TAP Plot MN PWP TRIP 6/23/81 RS 2305 15 .." 50 na TAP PL0f FW OSCILLAftoms 6/27/81 CA 3 2305 9 "' 55 ne STRIP CMAaf N 05CILLAflout 6/30/81 CR 3 23',0 4 2.5 50 at TAP PLOT MN ConfROL PeosttMS 7/08/81 Am0 1 2300 2 100 he TAP PLOT Mfw TRIP DUR!aG MAN RuusACK 9/01/81 Ami 9 *2300 <2 "" 75 he TAP PL01 Fault tu ? Avt stGNAL 9/05/81 03 <2300 <10 4.5 30 W ,ffm(1stC) N OSCILLATION 10/16/81 08 1 <2300 <15 .". 100 na atACTIMtigt meauAL umotartto 1/10/82 CR 3 <2300 4 "" 100 he STRIP CnAaf MN PWP TRIP 1/19/82 as 2334 46 15.0 100 he A stRIPCnAmi L0ss 0F TOTAL RC FLOW s!GNAL 3/24/82 01 <2300 <10 4.5 51 me, fin (1MC) fue8 & N A Conf aut tu mAm. 5/21s82 01 <2300 <10 2.0 90 me DutimG a0D De0P rumeACK l g 6/20/82 Ca 3 2300 4 21.0 82 se sfRIPCnAaf ImAcyttitut asty CL0 sung 7/05/82 Ca 3 2300 4 10.0 81 ne sit!PCMAaf LOAD atJtCTIon 7/15/82 Ca 3 <2300 <4 5.0 100 he Pim FAILED AC FL w input 8/08/82 Ca.3 1 9/11/82 01 2300 2301 11 4 "" 94 he stalP CnAaf 42 me, itM Loss CF #C FLOW $1CNAL N TAAmslief 11/06/82 02 <2300 <10 1.5 100 me, ifm(1stC) MN PUMP TRIP 3/19/83 Ca.3 <2300 4 "- 20 at stair CnAer N PW P ft P i 5/10/83 08 1 <2300 <15 4.0 90 m8 ICs testi 7/26/83 Ca 3 <2300 e4 5.0 16 un stalP Cnaaf Loss et N FLw Input to ICS CR 3 <2300 <4 1.0 9 mA STEIP CMAaf N FLOW REDUCiton I 8/22/83 CA 3 <2300 <4 2.5 65 se sfRIP CnAaf fun 8 covtance VALyts Ct05t0 8/26/83 CR 3 2300 e4 "" 75 se A pfW PUMP Pa0sttus 8/31/83 Amo 1 2310 12 4.5 100 he N Ptsus ie P ousIno numeAct I 11/12/83 Ca 3 12/03/83 as
<2300
<2300
<4
<10 5.0 4.0 75 se A N 8 LOCK vatyt CLosuet 65 na 8, sit!P CnAaf N testi, VALyts & Puurs In MAN.
2/16/84 03 <2300 <10 "" 100 W ,ffM FAILED SC FLOW NNiffta S/12/84 01 2322 32 14.0 100 me,ffm(1stC) FAILtc f nOf la its 1/22/85 01 2328 38 LC.0 100 W ,TTM(1MC) fur 8 luftectpf v4 Lyt CL0 suet 4/11/85 01 2293 3 3.8 17 me.ffn(IMC) N osCILLAftoms 01 2330 40 "" 100 at stAIP CMAaf fue8 tuttactPt VAtyt closuet 1 4/26/85 02 2300 to 18.0 75 me,ffp(1MC) 4-16
l' i ] I Table 4-3. Summary of Major Inputs to Monte Carlo simulation Procram ~ RANDOM ERRORS Probability Distribution Distribution l Source (Model) Parameters Reactor Normal Average 4.83 Protection psi ! ! System Standard Deviation 6.2175 psi s I Non-Nuclear i Instrumentation
-Buffer Amplifier Normal Average 0.0 l psi Standard Deviation 1.4 psi
-Bistable Normal Average 0.0 i Psi Standard Deviation
. 1.3 psi Satpoints t RPS Uniform Range O to
, + 5 psi I
NNI (PORV) Uniform Range -5 to j 0 psi PHYSICAL PHENOMINA l Pressure Exponential 16.4 psi overshoot (Truncated I s 60 psi p_ 10 psi) i I l1 ! l 4-17 1 I e
. . . _ _ _ _ _ _ _ _ _ _ _ . _ , . _ . _ _ _ . . . _ _ . .._,y _
_. _ , . - _ . . . _,.--,_m._-_ - _,. . . _ . - , .
_ _ _ _ _ _ . _ . . . _ . _ _ _ _ _ _ _ . _ _ . _ _ _ __ __ _ ._.._ _ __.___ _ _ _ __.._.m .. _ _ _ _ _ . _ _ _ _ _ _ . __.
._ .~ . . . ..._ .
~ ,
Figure 4.3-1. POWER TRAIN Vs. Plant Data j Turbine Trip Without ARTS ,
- 2300 psia Setpoint '
i i
! 2450--
)~ l l 1 t j RC PR PT
- 24o0- -
\ '
j
....... RC PR PLNT I
r
, ... . ... 2300 TRIP ,
j 2350- - a i A 1 t a, n " A 4 c_) .. . U) -'.--------
------------------------------------------.-[-------------------
w O. 2300- . t !
- u) :
l u) :
~.
c= to : i M : n 2250- -
/ . ,
l 0 x : i i: : .
- 22ao. . : .
~.
\
s
- i ~
.- 1 1
2150- -
- 2100 -
i j O .5 1 1.5 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 7.5 8 8.5 9 I i SECONDS AFTER TURBINE TRIP . l PLANT DATA FOR OCONEE-1,1/22/85 i
1 - - - - - - _ _ .. . . . . . .- - .. Floure 4.3-2. POWER TRAIN Results Turbine , j Trip Without ARTS 2450-- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - s l ! i l 4 2300 CASE t
] 2400- - ! -_ 2355 CASE a
p~ N, 2300 TRIP 2350--
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .............y./............ . ... . . . . . . . . . . .
1 p , 1 i '- .. . .. . 2355 TRIP i i m 5:2 .__ PORV m 2300- - ------------------------------- --------------------------- ------------------------ v Q. l i .
- m
= mn.a x
.; Q_ 22so- - j O a m 1 2200- - 4 i I 3 2150-- l : 4 i .a I 2ioo : :
- : : : . +-- - : . . : : . . .
i
- o .5 1 1.522.533.544.555.566.577.588.59 4
SECONDS AFTER TURBINE TRIP
a e i t s s p s p n 0 e 5 1 p _ o n n i i t
) )
3 5 ) Y h h n 9 6 5 R t t u 8 3 8 Oi i __ 2 2 4 PWW ( ( 1 ( 0 X e - - - 5 o x0 , 3 7 2 s t l u s e R o ) g
, l r 7 i s
a 7 C 0 ) p 1 5 e ( 5 , _ t n ( 5 t 6 n M o 0 ? 3 i 2 op t 1 S e 4 4 e r q u i F
)
_ 4! 4! (! 5 9 5 3 2
~
- ~ -
0 0 0 0 0 0 3 0 0 _ 3 2 1 _ :2.h 8o8 S O$n i yU
1 . 1 - - -- - t l ' Figure 4.4-2. Overshoot Distribution for Monte Carlo input I i i
.60 Observed (Plant Data)
- , n l ca o
- -c - -
3, E .50 ^ c E _
. - e n g ,8 .40 -
2 I Used o I
% E 3e 8 (Result of applied model)*
- g ~~ 8 3
go .30 -{g 8 4, y ao n i t- so _ _1 - i e R 3 .20 e" - I i i ~
- E~E
- a. 1
- v n 3,
e a __, -
.10 -
e L___ s c I [ s l I l 10 20 30 40 50 60 .l ] Overshoot, pst
- Approximated f
I i I i d I I
i - - - - I i I J t Figure 4.4-3. Model for Monte Carlo Analysis ) i-RPS "A" String P (Buffer Amplifier - Bistable High Press S.P. ,, String l _ PORV I # (Setpoint) k
- RPS "s" t '
? P i L l N RPS Errors " " String h+(4.83,6.2175) R l True 1 RC High Press U (0,5)
~
Pressure Setpoint - i l 1911 Errors String i RPS "c. s/A n(0 t.4) N(0,1.3) P _ Setpoint U(-5.0) i i 1
- . i RPS D" flote: N (a,b)s4>IIormal distribution with l
p mean a, standard deviation b.
)
i U (a,b)4> Uniform distribution on e I interval (a,b)
i i i 9 l i
- 5. NRC REQUIREMENTS / PREVIOUS ANALYSIS COMPARISONS Shortly after the PORV and high pressure trip set points were reset in 1979 the NRC requested an analysis showing the
' probability of a small break LOCA at the PORV. A response was submitted in January 1981 (reference 2). This study included a l statistical analysis of all the possible methods the PORV could i open, one of which was the high pressure trip transient. The NRC evaluation of this submittal was subsequently documented in a safety evaluation based on a Technical Evaluation Report (TER),
Reference 3. Table 3.1 summarizes the results of these previous studies. This section compares the results of this current study with the previous work and discusses both realtive to the NRC requirements on this subject. t 5.1. Guidelines Reference 4 describes the guidelines for the maximum alllowable SBLOCA porbability at the PORV. The overall requirement is that this type of SBLOCA occur less than .001 times per reactor year. The specific requirement for the high pressure trip transient is that no more than 5% of them are allowed to open the PORV. 5.2. Current Study Results The current study developed an estimated PORV opening frequency for overpressure transients of 1 x 10-5 openings per event. For the period 1980 through 1984, the seven B&W operating plants experienced 65 high pressure trip transientss. This is an average of 65/(7x5) = 1.86 events per reactor year. Thus, the estimated PORV opening frequency for overpressure transients is: 5-1 I i
- - - . , . - . - , - . - - - - - - - - - - - _ , - . _- _ - . - - - , . . . - - . - - . - . ----.._,_--m -
! -5 Openinas ~
'1.86 Events x 10 = 1.86 x 10 Reactor year Event Openinas Reactor year As stated previously, raising the high pressure trip satpoint affects only the PORV opening frequency for overpressure transients. Thus, the previous analyses (References 2 & 3) need i only be compared for the impact of the overpressure transient contribution to PORV openings. Reference 3 estimated a PORV
! opening frequency for overpressure transients of 2.2 x 10-3, Since the 2.2 x 10~3 PORV opening frequency yielded acceptable results, it follows that the 1.86 x 10-5 openings per reactor year frequency will also result in a probability of a small break LOCA due to a stuck open PORV of less than 1 x 10~3 The results of the current study are summarized as follows:
NRC Guidelines Current Study Resulig Probability of SBLOCA <1 x 10 ~3 2.53 x 10
~4 due to stuck-open PORV PORV Cpenings on <5 x 10 -2 1 x 10
-5
. Overpressure 9 Transients 5.3. Comearison of current Previous Analyses f The previous analyses addressed all aspects of PORV opening frequency and probability of a SBI4cA due to a stuck open PORV.
The current analysis focused on only the PORV opening frequency for overpressure transients since only this frequency is affected by a change in high pressure trip setpoint. The PORV opening frequency for overpressure transients developed in this study is l considerably lower than that estimated in reference 3 (1.86 x 10-5 vs 2.2 x 10~3). The reasons for this difference are
; discussed below.
The methodology and statistical components of the previous and the current analyses are similar with the exception of overshoot 5-2 II
i r
, I i
l l
; predictions. The previous analysis was based on transient plant l data where the PORV was open prior to the high pressure reactor trip. If the PORV had been closed, it was estimated that the pressure overshoot would be 17.4 psi higher. Therefore, without the PORV the mean pressure overshoot was calculatd to be 26.6 psi (9.2 + 17.4 from reference 2 page 4) on a normal distribution
. with a 27.52 standard deviation. The normal distribution was assumed due to the wide scatter of plant data used. On the other hand, the current analysis used plant data where the PORV did not open during the transient and consequently eliminated the need to estimate the ef fect of the PORV. Also, the data scatter in the overshoot prediction was reduced due to the methods discussed in section 4.2. This current data allowed for a more realistic distribution (see section 4.4) with a 16.4 mean overshoot. The effect of the inlet skewed overshoot distribution is reflected in j the lower probability calculation of (1.86 x 10-5) compared to
. 2.2 x 10-3 openingc per reactor year of the 1981 results.
q 5.4. Summary The previous analysis (Reference 3) calculated 2.2 x 10~4 PORV
; opening per high pressure event when the PORV was set 150 psi above the trip set point. The current analysis calculated 10-5 PORV opening per high pressure event when the PORV was 95 psi i above the trip set point. Based on the last 5 years of operating experience and data, the current analysis is considered the more realistic of the two. However, both analyses meet the NRC SBLOCA requirements.
I I I I 5-3 1
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O I 1
- 6. REFERENCES
- 1. BAW-1564, " Integrated Control System Reliability Analysis",
August 1979.
- 2. 12-1122779 Rev. 1, " Report on PORY Opening Probability and Justification for Present System ans Setpoints," Jan. 1981.
- 3. TER-C5506-410, " Operating Reactor PORV Reports (F-37),
Generic Report - Babcock Designed Units," July, 1983.
- 4. NUREG 0737, " Clarification of TMI Action Plan Requirements,"
Nov. 1980.
- 5. Hahn, G. J. and Shapiro, SS, " Statistical Models in Engineering, John Wiley & Sons, NY, Ny, 1967.
- 6. 32-1125056-02, " Statistical Errors - Rancho Seco," by H. T. Dass, NPD.
4 I i l 1 6-1 1
O ,#$ MEETING
SUMMARY
DISTRIBUTION Licensee: B&WOG Transient Assessmant and Availability Cocinittees
- Copies also sent to those people on service (cc) list for subject plant (s).
Docket File NRC PDR L PDR ORB 84 Rdg Project Manager - GVissing JStolz BGrimes (Emerg. Preparedness only) OELD EJordan, IE ACRS-10 PMorriette NRC Meeting
Participants:
WPaulson DCrutchfield CThomas AThadani RJones NLauben
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