ML20010H168
| ML20010H168 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 08/31/1981 |
| From: | Lyon R EG&G, INC. |
| To: | Israel S Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6442 EGG-EA-5533, NUDOCS 8109230778 | |
| Download: ML20010H168 (17) | |
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" o - / 5.6 EFFECT OF RECIRCUt.ATION PUMP TRIP FOLLOWING ANTICIPATED TRANSIENTS WITHOUT SCRAM AT sc A
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y This is an informal report intended for use as a preliminary or working document Prepared for the U.S. Nuclear Regulatory Comission Under 00E Contract No. DE-ALO7-761001570 FIN No. A6442 U EGnG w.n.
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roaw exam ptev 11 m INTERIM REPORT Accession No.
Report No.
EGG-EA-5533 Contract Program or Project
Title:
Review of Risk Study at Big Rock Point Subject of this Document:
Effect of Recirculation Pump Trip Following Anticipated Transients Without Scram at Big Rock Point Type of Document:
Informal Report Author (s):
NRC Researci anc' 7ecinica R. E. Lyon N ASSIS:ance 'lep0TF-Date cf Document:
August 1981 Responsible NRC Individual and NRC Office or Division:
S. Israel, Division of Safety Technology This document was prepared primariiy for preliminary or internal use. it has not received full review and approval. Since there may be substantivo changes, this document should not be considered final.
EG&G Idaho, Inc.
Idaho Falls. Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE AC07 76lD01570 NRC FIN No. A6442 INTERIM REPORT
0464J EFFECT OF RECIRCULATION PUMP TRIP FOLLOWING ANTICIPATED TRANSIENTS WITHOUT SCRAM AT BIG ROCK POINT August 1981 R. E. Lyon Reliability and Statistics Branen Engineering Analysis Division EG&G Idaho, Inc.
1 l
CONTENTS
1.0 INTRODUCTION
I 2.0 TECHNICAL DISCUSSION............................................
1 2.1 Transient Description.....................................
1
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2.2 Analysis Results..........................................
3 2.3 Comparison With Previous Analysis.........................
6 13.0
SUMMARY
AND RECOMMENDATIONS.....................................
8
4.0 REFERENCES
9 APPENDIX A--DISCUSSION OF ITEMS REQUESTED BY NRC PERSONNEL INVOLVED IN BIG ROCK POINT ATWS RISK REVIEW.....................
12 TABLE 1.
Operator Response Times for ATWS Events.........................
11 m
I 11
EFFECT OF RECIRCULATION PUMP TRIP FOLLOWING ANTICIPATED TRANSIENTS WITHOUT SCRAM AT BIG ROCK POINT 1.
INTRODUCTION As requested by the U.S. Atomic Energy Commission (now U.S. Nuclear Regulatory Conrnission) in the;r Technical Report on Anticipated Transients Without Scram (ATWS) for Water-Cooled Reactors (WASH-1270).- Consumers Power Company has submitted analyses which describe the response of their Big Rock Point (BRP) Plant to an ATWS. The original analyses were submitted on I
February 21, 1975 ' and the results indicated that a recirculation pump trip (RPT) was very effective in limiting the consequences of an ATWS. The response of BRP to an ATWS was reanalyzed'as a part of the Big Rock Point Probabilistic Risk Assessment (PRA). Results of the analysis were 2
submitted on February 26, 1981 with the conclusion that automatic RPT provides little safety improvement at BRP. The purpose of this report is to evaluate the submitted analyses to determine the effectiveness of Recirculation Pump Trip in ATWS recovery.
2.
TECHNICAL DISCUSSION 2.1 Transient Description For the purguse of the ATWS analysis, it is assumed that tne reactor is operating at rated power when a transient occurrs.
The transient 1
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g:nerates a reactor protection system trip signal however the control rods fail to insert.
Following detection of the failure to scram, the operator will attempt a manual scram followed by a trip of one or more recirculation pumps.
-Other methods of achieving shutoown will be attempted, culminating in actuation of the liquid poison system. The basic premise throughout the analyses is that shutdown must be achieved prior to actuation of the Reactor Depressurization System (RDS). Failure to do so is postulated to result in probable core damage as well as poss451e containment damage.
Specific response of the reactor to an ATWS depends on the initiating event and the availability of supporting systems such as the main condenser and the feedwater system. The resulting transients can be grouped into f our categories:
1)
Infinite Feedwater Transients 2)
Low Level Transients 3)
High Pressure Transients with Limited Feedwater 4)
High Pressure Transients with No Feedwater.
The first category, infinite feedwater, assumes that the feedwater system and main condenser. remain operational. The reactor continues to produce power which is dissipated in the condenser until shutdown is 2
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. achieved. An infinite feedwater transient will only lead to core damage if it degrades into one of the other categories or if long term cooling f ails after shutdown of the reactor.
Thus this category of transient has not been analyzed in detail.
The low level transient which has been analyzed is a loss of feedwater. The turbine bypass is assumed to operate normally to maintain normal system pressure until the turbine trips on low steam drum level, following which the dump valve opens and continues to deplete reactor inventory until the reactor is shutdown or RDS actuation occurs.
The high pressure transient with limited feedwater which has been analyzed is a turbine trip without bypass. Feedwater is assumed to be available until the hotwell level drops to the point where the condensate pump trips after which the transient is similar to the loss of feedwater transient.
The high pressure transient without feedwater which has been considered is the loss of station power. This transient has not been specifically analyzed but has been shown to be limited by the two analyzed Cases.
2.2 Analysis Results Reactor response to an ATWS is calculated using a RETRAff 3) model of the BRP primary coolant system. The review status of this code ar.d its acceptability to NRC as a licensing ccde is not known.
It should be noted 3
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that this code apparently gives different results than the original ATWS i analysis for a similar event. This is discussed in more detail in SGction 2.3.
Containment response is calculated based on an approximation derived
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from earlier calculations and is given by the equation i
P = (3.6 x 10-4) Msteam where P is the containment pressure (psig) and M is the mass of steam i
steam dumped to the containment during blowdown.
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A comparison of this approximation with Figure VII.1 of Reference 4 l
indicates that the approximation is probably valid for the area of interest here.
It should be noted that the actual response as shown on the Figure l
l is nonlinear at higher steam inputs and will tend to underpredict the pressure. Thus the approximation should not be used in the event that the analysis is ever extended to higher steam inputs.
L There are two assumptions used in the analysis which may not be l
- onservative, the initial volumes of the hot well and the steam drum.
Only L
a nominal value is given for each.
If the initial volumes were less than noniinal, the time for operator
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action prior to RDS would be decreased. Conversely, if they were higher r
than nominal, more makeup to the reactor would be available, the operator
.may delay' initiation of.the LPS, and the resulting steam release and
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corresponding containment pressure would be greater.
It is recommended that CPC be asked to provide the minimum and maximum values of these two volumes and to discuss the effects of these variations on.the analysis results.
The results of the analysis show that, if reactor shutdown can be achieved prior to RDS actuation, the consequences of the transient are acceptable. The reactor power stabilizes at an acceptable level with energy being removed by the cordenser, if available, or the safety valves.
Sizing of the safety valves is adequate to prevent system overpressur ization. Based on a limited amount of feedwater, containment a
pressure will be maintained below design pressure.
2 In the event that RDS actuation occurs prior to shutdown, core damage
-is predicted to occur. The BRP PRA also predicts that containment a
overpressurization failure could occur between 16 and 49 minutes after RDS, but further says that the understanding of the processes by which liquid poison mixes in the core following RDS actuation is inadequate to predict nuclear shutdown prior to 50 minutes after RDS actuation. Therefore, failure to achieve shutdown prior to RDS actuation is predicted to cause containment failure.
Regulatory Guide 1.70 identifies seven transients to be considered as initiating events for an ATWS. The analyses presented by CPC adequately cover all but the inadvertent control rod withdrawal. The BRP FHS report infers that the bypass valve does not have sufficient capacity to handle the excess flow without reactor trip, so some flow through the safety 5
e valves will be necessary. Assuming that the turbine and bypass system will
- continue to function and feedwater is available, inventory loss and rCquired operator response will be slower than for the other analyzed events..It is recommended, however..that CPC be requested to address this particular transient.
i The limiting transient from an operator response standpoint is the loss of feedwater. As shown in Table 1, the time to actuation of RDS is 145 seconds with no RPT. With a time allowance of 41 seconds to mix LPS r
the operator has 104 seconds to initiate poison injection. Actually, the I
time may be even less than this. From the data available it appears that a normal reactor trip on low level would not be expected to ocur until 27 seconds.
It is at this point that the operator would become aware of a problem with the scram system. Subtracting this time gives a time of o
77 seconds for the operator to initiate LPS or RPT.
If he actuates RPT in this time frame then he will have '-2 minutes additional to actuate LPS.
The high' pressure transients are not effected as greatly by this factor as an early trip is expected on high pressure or high power at about I to 2 seconds into the transient so the operator would have an earlier T
indication that a problem exists.
o 4
2.3 Comparison With Previous Analyses A comparison of the results of this analysis with those obtained in
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NEDE-21065 shows a significant difference, particularly in the area of maximum containment pressure. The main reason for the difference lies in 6
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l the assumption-of the availability of feedwater for. the duration of the transient in NEDE-21065. The additional mass input into the containment results in significantly-increased containment pressures.
In the current analysis the feedwater is limited and the limiting factor is the time before the level drops to the point.at which RDS is actuated.
In the L
original analysis, which was time limited, the reduced power caused by RPT resalts in a significant reduction in containment pressure whereas in the current analysis, which is mass limited, all available mass is assumed dumped to_the containment whether or not RPT occurs, resulting in the same p ressure'. RPT just delays the time for which operator action must be taken to prevent RDS actuation.
A comparison of the current analysis of a turbine trip with a previous i
analysis having approximately the same LPS initiation time, both with and j
without recirculation pump trip, was made. For the case without RPT, the original analycis resulted in a significantly higher peak pressure l
(42.3 psig v 22 psig). There are several factors which contribute to this difference.
In the original analysis, steady state power during the transient is projected to be about 110% and to remain there until LPS begins to take effect. A time of 300 seconds is assumed from inis point until shutdown is achieved. The current analysis assumes an initial steady
-state power of approximately 80% until feedwater is lost at which time the power drops to 60% and remains there until the LPS takes effect. From this point on a time of 11 seconds is assumed until shutdown is achieved.
This combination of lower power levels and f aster shutdown both contribute to a 4
lower steam release and subsequent con',ainment pressure. There is one area of concern here.
It is n u apparent why the initial power level is D
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idifferentlfor the previous and current' analysis (110% vs 80%).- It is
.rtcomrnended that CPC be requested to clarify this difference.
The concern
.h:re'is.that, if the power level and resulting steam flow were greater than that assumed in'the current analysis,'the time for operator response to prevent RDS would be even' shorter.
3.
SUMMARY
AND RECOMMENDATIONS The analyses presented by CPC indicate that the BRP plant can recover safely from an ATWS event, however, with the present design, operator action.is required in a very short time frame. Current analysis assumes
.that if reactor shutdown cannot be achieved prior to RDS actuation, core failure and containment failure are likely to occur. Operator initiation of LPS injection is required in as short a time as approximately one minute for the limiting transient. This time can be extended by two minutes or more by early initiation of RPT. Assuming that shutdown can be achieved i
prior to RDS actuation, RPT has no significant effect on the consequences 1
of the transient, but only on the timing.
3 t
.There are several areas which need additional clarification as a
[
result of.this review.
a)
The status of the RCTRAN code needs to be established.
If NRC nas not approveu this code, CPC 3hould be requested to provide confirmatory data. Of particular concern is the fact that the original analyses, which were based on another code which has received some ve-ification for BRP, predicted ci init ial steady 4
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c state power level of 110% following the turbine trip ATWS, whereas the current analysis, using RETRAN predicts 80%
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l b)
The allowable maximum and minimum volumes in the condensate hot well and in the steam drum need to be specified by CPC and the effect of these variations on the analysis results need to be i
determined c)
-The LPS has been indicated to not be environmentally qualified.
~
This point needs to be addressed further, specifically the effect of steam releases during the early stages of an ATWS event on the components inside containment needs to be established d)
CPC should be requested to address the plant response to an ATWS initiated by an inadvertent control rod withdrawal.
%I le e)
CPC should be requested to address the delay in operator response time for the loss of feedwater transient due to the delayed normal reactor trip.
4.
REFERENCES L
1..
P. M. Gururaj, " Anticipated Transients Without Scram Study for Big
' Rock Point Power Plant", NEDE-21065,.0ctober 1975.
2.
Letter, G. C. Withrow, Consumer Power Company to Director, Nuclear Rcactor Regulation, February 26, 1981.
9
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3.
"RETRAN--A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Compicx Fluid Flow Systems," EPRI CM-5, December 1978.
-4.
Consumers Power Company, Probabilistic Risk Assessment, Big Rock Point Plant.
s 9
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l TABLE 1.
OPERATOR RESPGhSE TIME FOR AT'WS EVENTS' Time to Time to Operator Time RDS Mix LPS to inject Poison Transient Seconds Seconds Seconds Low Level No RPT 145 41 104 RPT @ 80 seconds 258 75 193 RPT @ 60 seconds 287 75 212 RPI @ 35 seconds 312 75 237 High Pressure with Feedwater No RPT 267 41 226 RPT @ 60 seconds 430 75 355 RPT @ 8 seconds 530 75 4"'
High Pressure No Feedwater RPT @ 60 seconds 309 75 234 RPT @ 0 seconds 350 75 275
- From Reference 2.
This data does not incicde potential delays in operator action due to expected delays in normal reactor trip.
0 4
4 11
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APPENDIX A
~ DISCUSSION OF ITEMS REQUESTED BY NRC PERSONNEL INVOLVED IN BIG. ROCK POINT ATWS RISK REVIEW L
Lc
- 1..
Are the sequences that lead to core uncovery and containment failure correctly identified by CPC?-
The ATWS event tree presented by CPC addresses only core damage and not containment failure. Core damage is assumed for all sequences not
' achieving LPS injection and mixing either before or after RDS actua-tion. Delay of LPS injection until after RDS actuation is assumed to cause only_ limited core damage, which is not considered in the risk determination. The releases resulting from these sequences are stated to be similar to TMI.
i
.As previously noted, the event tree does not address containment fail-ure. Page VII-32.of the Big Rock Point Probabilistic Risk Assessment states that ATWS sequences which lead to RDS actuation are predicted to produce containment overpressure failure.
It would seem that this f ailure, coupled with a TMI level release, would provide a significant risk contribution, which has not been considered in the risk calcula-tion.
As noted in the draft report, the transient response to a rod with-I drawal initiated ATWS is not addressed.
It is net clear whether or 12 t
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'notithis.has been included in the event tree quantification.- It may-be' included under-miscellaneous scrams.
22.
Are the assertions about~RPT not being effective correct?
n I This assertion is based on-the quantification of the event tree wnich indicates that the~ probability of core' damage is not significantly reduced by automatic RPT. The validity of this assertion is dependent upon the accuracy of the data input to the quantification. RPT, either automatic or manual, is effective in delaying the time at which the operator.is required to take action to initiate LPS injection. Also, it has been-stated that the LPS is not environmentally qualified. RPT will reduce the rate of steam input into the containment, thus increas-ing-the probability that the LPS will operate when the operator does
'd take action to initiate it.
o-3.
Identify ary sequences where RPT would help.
RPT,feither manual or automatic, will improve the chances of recovery from any sequence which results in loss of system inventory or blow-down to the containment, as discussed in item 2.
4.
Are the predicted system transient responses reasonable conservative?
,)
' Concerns with the conservatism of the analysis are discussed in the j*'
draft report.
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