ML20002C546
| ML20002C546 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/31/1975 |
| From: | Eckert E, Gururaj P, Holland L GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20002C544 | List: |
| References | |
| NEDE-21065, NUDOCS 8101100452 | |
| Download: ML20002C546 (24) | |
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NEDE 21065 Class 11 October 1975 GENERAL ELECTRIC COMPANY AND CONSUMERS POWER COMPANY USE ONLY ANTICIPATED TRANSIENTS WITHOUT SCRAM STUDY FOR BIG ROCK POINT POWER PLANT P. M. Gururaj Reviewed.
Approved:
E. C. Eckert, Me.oager L. K. Hctland, Manager Transient Performence Systems Plant Performance Analysis i
and Systems Specifcations BOILING WATER RE ACTOR SYSTEMS DEPARTIAENT e GENERAL ELECTRIC COMP 4NY f
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SAN JOSE, CALIF ORNI A 95125 GEN ER AL h ELECTRIC 8101i096S
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M TABLE OF CONTENTS Pege INT RODU CTION...........
.... 1 1 1.
11 1.1 Subject of Report..
.11 1.2 Previous Studies..
1-1 1.3 Approach Used in this Study..
SU M M ARY OF R ESU LT S........
.. ~..............-
-.........~. 2 1 2.
...............~..- --..... ~.... 3 1 3.
.. 31 3.1 Analytical Methods..
.3-1 32 Initial Operating Conditions..
.32 3.3 Equipment Characteristics.,
AN A LY SIS GUl0E S.................................
. -.~ ~....... 4 1 4.
.41 4.1 General..
.. 4 1 4.2 Reactor Coolant System Pressure..
.41 4.3 Fuel Thermal and Hydraulic Performance..
.. 4-1 4.4 Containment Conditions.
DEFINIT 10 N O F EV ENT S..................-
.. 5-1 5.
........~..............6-1 6.
ANALYTICAL RESULTS.....
.. 6 1 6.1 Description of the Worst isolation Event (Load Rejection Neglecting Bypass)--
.6-2 6.2 Short. Term Conditions..
6-3 i
6.3 Long-Term Conditions...
.6-6 6.4 Comparison to WASH 1270-R EF E R EN C ES....................
..... 7 1 7.
APPENDIX PARAMETRIC AND SENSITIVITY STUDIES OF THE CONTAINMENT RESPONSE....
- A-1 A.
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I LIST OF ILLUSTRATIONS Figure Title Page 1.
Fission Power and Fuel Center Temperature Transients Used in the Fuel Response 6-2 Estimate in an ATWS Load Rejection Without Bypass.......
2 Vessel Steam Flow Transient Used in t*e Containment Response Analysis in an 6-3 ATWS Load Rejection Without Bypass..
3 Vessel Steam Flow. Emergency Condenser Flow, and Steam Flow to Containment Used 6-4 in the Containment Rc3ponse Analysis in ATWS Load Rejection Without Bypass -
4
' Containment Pressure Response in an ATWS Load Rejection Without Bypass..
.6-5 5
Containment Temperature and Containment Wall Temperature in an ATWS Load 5 Rejection Without Bypass..
6 Containment Wall Surface Heat Transfer Coefficient as a Function of 6-6 Air-to Water. Vapor Ratio..
A1 Containment Pressure Response in an ATWS Load Rejection Without A2 Bypass, With Recirculation Pump Trip at 6 seconds.
A-2 Containment Temperature Transient in an ATWS Load Rejection Without A-3 Bypass. with Recirculation Pump Tnp at 6 seconds..
A.3 Multiplication Factor Used in Equation A.1 for the Effect of Recirculation Pump Tnp on Vessel Steam Flow.-
..A-3
'.vd vi.
+
i LIST OF TABLTS Toble Title Page
-21 2-1 Worst Reactor Isolation Without Scram...............
.32 31 init al C onditions........................................
........... =
.~~~.....~~...~3-2 32 Equipraent Performance Characteristics.
6-6 61 Full Load Rejection Without Eypass and Without Scram.
..A-1 A-1 Containment Response to ATWS: Worst Reactor isolation.....
-A2 A-2 Containment Response in ATWS: Worst Reactor Isolation......
..A 4 A3 Containment Response to ATWS: Worst Reactor Isola:.,n--
A4 A-4 Containment Response to ATWS: Worst Reactor Isolation -
.A 5 A5 Containment Response in an ATWS: Worst Reactor Isolation......
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INTRODUCTION 1,1 SUBJECT OF REPORT in September 1973, the NRC issued their Technical Report on Anticipated Transients Without Scram (ATWS) for Water Cooled Reactors (WASH-1270). That report defined three categones of plants. The first or A category requires improved reactor shutdown systems and is aoplicab!c to plants for which construction perrmt applications are made after October 1,1976. The second or B category requires provision to make the consequences of ATWS acceptable and is applicable for plants for which the need for provision for ATWS is noted in the NRC Saf ety Evaluation Report or the ACRS Report and for plants for which the construction permit applications are or havt, Men made poor to October 1,1976, and NRC SER is not yet issued.The remainder of the plants f allinto the third or C Categsry which requires an anafysis of ATWS consequences so that the NRC stati can evaluate whether there is need for plant changes to resolve the ATWS issue.
The Big Rock Point Power Plant falls in the C Category as defined in the Appendix B of WASH-1270. This report describes the studies performed for the Consumers Power Company in response to the re.ouirement of an ATWS consequences transient analysis for the Big Rock Point Plant.
1.2 PREVIOUS STUDIES The subject of transient evaluation without scram protection has been an N RC concern for some time. This comes out cf a desire to better understand and protect aga 1st common mode types of failures (CMF). General Electnc has responded to this concern with topicaireports NEDO 10189, ent tied "An Analysis of Common Mode Failures in GE BWR Protecton and ControlInstrumentation," July 1970, and NEDO 10349," Analysis of Anticipated Transients Without Scram,' March 1971. In addition, General Electric has issued NEDO-20626, "Otudies of BWR Designs for Mitigation of Anticipated Transients Without Scram," October 1974, which responds to the b category of SWR WASH-1270 requir ements for an ATWS mitigation i
design.
Both NEDO 20626 and this r eport are based on the premise that a failure to seram occurs coincident with an abnoimat operating transient. This premise is assumed in th's report to satisfy the requirement of WASH-1270 for analysis of the consequences of anticipated transients in the event of a postulated f ailure to scram. W ASH 1270 states that the likefihood of a severe ATWS is considered to be acceptably smallin view of the limited number of plants now in operation, the reliability of the current protective system designs, and the expected occurrence rate of anticipated transients of potential safety significance. WASH-1270 thus does not justify the imposit on of any new requirement for hardware changes on exist.ng plants. The General Electric Company is in complete agreement with this position. This report should not, therefore, be construed as a recommendation for plant modsfecations for the purpose of mitigat:ng the consequences of a postulated ATWS event. It is an exploratory engineering study to evaluate ATWS as required by WASH-1270.
1.3 APPROACH USED IN THIS STUDY This report provides a representative treatment of the main aspec's of WASH-1270 requirements. Smce the primary aspects of NEDO-10349 have been completely reconfirmed by further calculations,it was used as a guide for identficatica of the most limiting type of transients relative to each of the analysis guides when failure to scram was considered.
Attenton was dirocted loward the transients which have the highest probability of occurrence. This is consistent with the stated objective of WASH 1270. Very infrequent events are not considered as they make no significant contribution to public safety considerations when combined with the low probability of failure to scram.
Previous analyses have shown that among the events reasonably expected to occur,the ones which cause the most severe ATWS conditions are those which initiate f ast shutoff of the steam flow f rom the reactor,such as turbine-generator trip or the closure of the main steam isolation valves. For the Big Rock Point plant the MSIV closure time is much slower than in the later plants. This would make the MStV closure event relatively much less severe. Also, the large bypass system on the l
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Big Rock Point plant makes the turbine. generator trips relatively mild. The analysis given in this report, in which a ~
- turbine generator load rejection is assumed to cause reactor isolation by taking no credit for the action of the bypass valves, constitutes an entremely conservative bounding case.
in analysis, ATWS behavior can be separated into the bohavior of the nuclear boiler and that of the contelnment. In this eeport the nuclear boiler dynamico is not calculated since previous studies done for FHSR input. APED-4093
,c, (Roference 1), conteln safety valve siting analyses which are pertinent to the present work. Specifically they analyze the transient of turbine Iced rejection with safety valve operation, without bypass and without scram. This analysis, therefore, represents a bounding ATWS event. Comparison of reactor peak pressure and fuel response with WASH 1270 criteria shown in this report are based on these safety valve sizing analyses.
The main concern here is to calculate the response of the containment in an ATWS event. As far as the containment response is concerned, the difference between diff erent reactor isolation events without scram is insignificant when the initial power levelis the same. As the reactor response for an ATWS event of isolation caused by load rejection without bypass is evallable from Reference 1, containment response is studied for this tra'isient.
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SUMMARY
OF CSULTS Table 21 WORST REACTOR ISOLATION WITHOUT SCRAM (BOUNDED BY TURBINE LOAD REJECTION NEGLECTING BYPASS).-
Functional General Comparison Electric Value Parameter Guide Analysis Reactor Vessel Pressure (psg) 2700 1587*
Fuel Enthatpy (cal /gm) 280
<165 Cladding Oxidation (%)
17.
<1 Containment Pressure (psig) 54 57.7
' Peak reactor pressure increase taken from Reference 1, Fgure 15, Run 3
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ANALYTICAL BASES 3.1 ANALYTICAL METHODS The first step in the analysis of ATWS behavior is to calculate the dynamics of the nuclear boiler.This calculation is already bounded by the safety valve sizing analysis reported in APED-4093. That analysis used an analog model of the
. reactor and the details are documented there.
The response of the containment is evaluated using the techniques described in the General Electric Licensing Topicat Report. NEDO-20533, dated June 1974. entitled "The General Electnc Mark lli Pressure Suppression Containment System Analytical Model." The engineering methods used in this report are appropriate for use in all containment desig The specific methods used are described in Section 3 of NEDO-20533.The following assumptions were used in analyzing the containment response to an ATWS ovent.
The steam release from the reactor to the containment sphere as a funchon of bme is taken from the safety 1.
valve sizing analysis reported in Reference 1.
2.
The steam entenng the containment sphere has an enthalpy of 1190 Bfo/lbm.
3.
The air-water vapor system (excluding the containreient spray water) is in thermodynamic equihbrium. The equilibrium cond. tion is dependent on the in6tial conditions and the not steam and enthafpy dump into the containment.
4, No heat is lost to the outside by heat transfer through the containment w all. However, the cooling eff ect by heat transfer into the containment wall is taken into ecount The containment wallis assumed to be at a singte temperature and heat transfer between the air water vapor system inside the containment and its waff is assumed to take place until thermal equihbrium between the two is achieved. Similar cooling effect of other structures in the containment is neglected.
The heat transfer cocMicient between the containment air-water vapor system and the containment wallis 5.
assumed to be a function of the air-to-water-vapor weight ratio inside the containment.
6.
The effect of the containment spray is limited to abstraction of heat frnm the air. water vapor system.
T he etfectiveness of the containment spray is assumed to be 70% (TSo cffectiveness is defined as the ratio of 7.
the temperature rise in the spray water to the difference betweer the containment air water vapor system temperature and the initial temperature of the spray.)
Effects of the tuol surface heat transfer are estimated using the
- General Electnc BWR Thermal Analysis Basis (GETAB): Data. Correlation, and Design Application" code as desenbed in NEDO-10958. dated November 1973. Further evaluations of fuel and cladding temperatures are derived using methods basic to the LOCA analysis also described in NEDO-10958. However. In this calculation the nuclear power generation and fuel heat flux data needed as input are taken from the safety valve sizing analysis of Reference 1.
Other areas which are not ngorously treated n the anatytical bases of this report include nonhomogeneous mixing of hquid poison.
3.2 INITIAL OPERATING CONDITIONS
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Table 3.t hsts the initial conditions for the moslimportant plant parameters. T hese are chosen as tepresentatsve of the
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conditions of the reactor at the beginning of an e/ent for which ATWS impact is <atuated.
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t, Table 3-1 INITIAL CONDITIONS Initial Parameter Condition Reactor Operating Pressure (psig) 1335*
Power (MWt) 240*
Steam / Feed Flow (Ibm /sec) 278' Containment Sphere Free Volume (tt')
940.000*
Containment Sphere Thickness (in.)
0.702 (min)'
O 875 (max)*
Initial Pressure of Containment (psig) 0.08 Initial Temperature of Containment ('F) 100d Relative Humidity (%)
1008
' From Ret wtwice 4 t
- Vrom neierence 2
- Basedon240Mwipower endfeeometerentheipyof 347 iBtu/em
- From Re'erence 3 3.3 EQUIPMENT CHARACTERISTICS The characteristics of the important pieces of equipment used to mitigate the consequences of failure to *,(tam ure listed in Tablo 3-2.
d Table 3-2 e
EQUIPMENT PERFORMANCE CHARACTERISTICS Parameter Characteristic Safety Valves 6*
Number Capacity (Ibm /sec/ valve) 63.9' Setpoint Range (psig) 1535-1585*
Emergency Condenser Number of Tube tsuiid'es 2*
16x10" Capacity (Blu/hr/ bundle) initiation Signal (Reactor Pressure Rise, psig) 1108 4e Delay Time (sec)
Containment Spray Number of Sets 2*
400*
Flow Capacity (gpm/ set)
Initial Temperature (*F) 80' 70c Effectiveness (%)
Delay Time from Liquid Control Start to Beginning of 30' Shutdown (rec)
Time Required to Complete the injection of 306" Control Liquid (sec) 32 r
'i Table 3-2 EQUlf MENT PERFORMANCE CHARACTERISTICS (Continued)
Persmeter Characteristic Initiatot Timo 300' Standby Liquid Control (sec) 300c Containment Spray 37 Containment Spray Delay None' Reactor Recirculation Pump Trip
- rrom ne erence 4 v
- From Retstence 2
- Assumed
- From Reference 3
' No # etaculaten pump tro is essumed to conservatsely evabate m0 calculated oNecis of the ATWS in
' me analyvs Appencts A calculates me eMect of assuming recwcutscen pump try ruteted at varove eme instants.
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ANALYSIS GUIDES 4.1 GENERAL WASH 1270 asks for comparison of the ATWS transient ic show whether:
Calculated reactor coolant system transient pressure exceeds a value such that tne maximum primary stress a.
in the system boundary is equalto that of the emergency conditions" as defined in the ASME Nuclear Power Plant Components Code. Section Ill, or the b.
Effects of the ATWS event result in significant fuel cladding derfadation or significant fuel melting, or the -
c.
Calculated containment pressure exceeds the design pressure of the containment structure.
Since these guides are apphed to ali reactor types, and consequently are rather general, it is necessary to interpret the guides with respect to the Big Rock Point BWR design. The guide interpretations are discussed in the following paragraphs and compar ed to the guides proposed in NEDO-10349. General Electric feels that, f or events as improbable as those associated with failure to scram, the limiting cntena in NEDO-10349 are sutficient for maintenance of public safety.
4.2 REACTOR COOLANT SYSTEM PRESSURE 4.2.1 Reactor Coolant System Boundary Pressure On consideration of this guide and examination of the system, the WASH-1270 guide translates to a vessel pressure of 2040 psig NEDO-10349 secommends 2700 psig as the vessel pressure that can be accommodated without structural tailure.
4.3 FUEL THERMAL AND HYDRAULIC PERFORMANCE These subjects including pertinent tailure mechanisms were discussed at length in N EDO-10349, Section 5.1.3. The apphcation of the guide does not change from that made in the previous report. With respect to prompt failures, an energy deposition gude of 280 cal /gm has been selected. It has been shown that f ragmentation is avoided at raidation levels of less than 17% by volume.
4.4 CONTAINMENT CONDITIONS The containment sphere design pressure of the Big Rock Point Plant is 27 psig. NEDO-10349 recommends the enembrane yield limit of the primary containment which would be 54 psig.
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5.
DEFINITION OF EVENTS
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The occurrence of a common mode f >ture which completely disables the reactor scram function is a very low probability event, Theretore, no significant risk to pubhc safety is presented by the combination of an infrequent event and a
BWR analyses separate reactor transient duty into three main areas: normal plant maneuvers, abnormal transients,
.md design basis accidents. The middle group spans a very wide range of occurrence, f rom more than once per year to less than once per 40 years Other nuclear reactor suppi'ers have for some time been defining this diverse group of abnormal transients intn two categories of more and less-feuent ever.ts. (ANSt.N18.2, Nuclear Safety Cnteria for the Design of Piessurized Water Reactor Plants) All ves sets and related components are also designed with strong considerations for the frequency as w ?ll as the magnitude of the thermal cycles caused by the transients. Oper ational experience from the growmg hst of plants on br:e provides valuable input to the understanding and estabbshment of reasonable rates of occurrence for events of irnimr(Mce in this study.
T he fo lowing transients, derived for ATWS consideration on the basis of operational experience have the potential of a frequency of occurrence of at least once in 4 years of reactor operation at power cond tions such that a signif cant transient results and scram is called upon to shut down the reactor, 1.
Turbme/ generator trip 2.
Closure of all main steam isolation valves 3.
Pressure regulator failure in open direction 4
Feedwater ces miner failure to maximum demand g
5.
Loss of feedwater l'aw 6.
Loss c' Liiary power.
Further descriptions of these events are contained in NEDO 10349. These events cover the spectrum of transients which might seguito shutdown and are representative of the experience record of operating BWR's.
The results of previous analyses reported in NEDO-10349 and NEDO-20626 show that the most severe transient in the above group of ATWS events is the closure of the MSIV. These two reports mainiy address BWR plants constructed subsequent to Big Rock Point For Big Rock Point the MSIV closure time is relatively much longer. The worst ATWS event for Big Rock Point would then be the reactor isolation caused by turbine-generator trip when bypass valve action is neglected. It is this evert that is presented as the bounding analysis in this report. The analyses of N EDO-10349 and N EDO-20626 f urther show that as far as the dump of scram mass and energy into the containmentis concerned, reactor isolation ATWS events caused by turbine trip neglecting bypass and MSIV closure are nearly the same As steam release to containment data is already available from previous analyses (Reference 1) for the former ATWS event and as our main additional concem here is the containment response, containment response analyzed here is for this bounding case. However, it should be noted that the expected frequency of occurrence of the transient of load rejection with bypass f ailure is far below the value of once per four reactor years that is considered significant from the ATWS risk viewpoint. Fult load rejection with normal bypass oporahon is not expecter* 10 call for a reactor scram.
Previous ATWS analyses of other BWR plants have shown that recirculation pump tr.p considerably reduces the seventy of ATWS transients. The present analysis of the bound:ng case neglects this beneficial effect of recirculation pump trip which is the usual operator action on loss of turbine load. (Parametne studies of the etfects of the recirculation pump trip
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are given in Appendix A.) For these reasons the turbme trip tranMnt, even in the event of a postulated f ailure to scra much less severe than the calculations presrnted here.
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6.
AsiAl.YTICAL RESULTS
6.1 DESCRIPTION
OF THE WORST ISOLATION E/ENT (LOAD REJECTION NEGLECTING BYPASS) 8.1.1 Overview of Responte Without Scram The desenption of this event given here is based on the preseni analysis and that of Reference 1. The reactor behavior can be separated into a short-term condition -involving reactor pressure nse and fuel r esponse (0-48 seconds) -
and a longer term condition (to several hundred seconds) -involving coolant and containment conditions -until the reactor is shut down.
As the turbme valve closes, due to the assumed lack of bypass, the reactor pressure rises. This causes the collapse of the voids in the reactor core resulting in a nuclear power spike. The pressure nse is limited by the safety valve operation. The reactor then assumes a new state (at a shghtly higher thermal power) dictated by the safety valve setpoint and capacity. The safety valve steam flow released into the containment sphere pressurizes it, in tho absence of the scram function the long. term aspects of the ATWS event are terminated by insertng negative reactivity into the core by means of the standby liquid control system. The analysis shows the plant response for the case m which the liquid controt is initiated at 5 minutes. As the steam flow to the containment sphere is terminated by reactor shutdown. the containment spray system reCools the sphere, thereby bringing down its temperature and pressure. The emergency condenser which is initiated by high reactor pressure, reduces the steam output into the containment sphere. The metal containment wall absorbs heat from the air-water vapor mixture, thereby reducing the seventy of tha containment pressure and temperature conuitions.
Present operating procedures at Big Rock Point call for the immediate trip of a reactor recirculation pump on certain plant transients.
In the case of turbine trip without scram and without bypass, such a trip of a recirculacion pump would reduce core flow which increases the core voids.
This reduces the power and steam output of the reactor and reduces the severity of the ATWS event, especially in the containment sphere.
The analysis given here neglects the beneficial effects of operator-initiated recirculation pump trip.
The re-suits presented here are therefore conservatin.
The effect of tripping the recirculation pumps at various times af ter the occurrence of an ATWS transient is given in Appendix A.
6.1.2 Sequence of Events for the Worst ATWS leolation Event (Load Rejection Transient Neglecting Bypass)
Time Event (sec) 1.
Stop Valve Tnps 0
2.
Reactor Pressure Begins to Rise 0
3.
Emergency Condenser initiated
~3*
4.
Safety Valves Open
-3" Operator Initiated Recirculation Pump Top 5.
Liquid Control Reaches Core 330 6.
Containment Sprays Begin Operation 330 7.
Hot Shutdown is Achieved 630
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Containment Temperature and Pressure Peak
~600
- Flow begins afier 4-second deioy
" Neglected in the analysis for conservatem h the calaalated oftects 6-1 e
i 6.2 SHORT TERM CONDITIONS 6.2.1 Primsey System Pressure As stated tmfor e, the reactor pressure response is covered by the earlier safety valve sizing analyses of Ref erence 1 (Figure 15) Of the five different runs shown in Figure 15 of Reference 1, the one numbered 3 is for a setpoint of 200 psi above lho operating pressure for the first safety valve. This value is the same in the present case (see Tables 3 1 and 3 2),
and therefore Run 3 of Figure 15. Reference 1,is most appropriate for the present discussion. However, the capacity of the safety valves used in Reference 1 (Figure 15) was 200% of rated reactor flow. In the present case it is (63.9 x 6)/278 w 138% This ditf er ence must be kept in mindin the present discussion of reactor pressure response Taking the peak pressure rise of 252 pse from Reference 1 Fgure 15, Run 3, the peak reactor pressure in the present ATWS discussion would be (1335 4 252) = 1587 psig, which is well below the design pressure of 1700 psig, and therefore is clearly acceptable.
6.2.2 Power and Fuel Response An estimate was made for the maximum cladding oxidation and peak enthalpy experienced by the fuel for the worst reactor isolation case of fullload rejection without bypass. The ne atron flux (fission power) and fuel heat flux transients used in this analysis are shown ta Figure 1. These are again taken from Ref erence 1 Figure 15. Run 3, with the fuel heat flux be:ng assumed to be proportional to the fuel center temperature. The results obtained fcr the fuel response are similar to those in f4EDO-20626 for the BWR/4. In this case the value of the cladding oxidation is far less than 17*. by volume (<1%) and the peak tuel enthalpy is less than 280 cal /gm (< 165 cal /gm). These va!ues clearly demonstrate satisf actory 3:el performance.
300 FIS$10N POWER 200 FUEL CENTER TEMPERATURE o
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100 0
e to 20 30 40 TIME (seci Figure f.
Fossion Pow er and Fuct Center Temperature Transients Used on the Fuel Response Estimate in an A TWS Load
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Rejection Wsthout Bypass. Data Taken from the Safety Valve Sizing Analysos of Referency 1, Figuo a 15, Run 3.
P00R ORIGINAL p
6.3 LONG-TERM CONDITIONS 6.3.1 Power Shutdown Steu o WASif 1270 im;unem tho ussinnplion of a simthmus common-modo failure of all control rod drivo stuit.hnnisun, en crodit for normal en emergerty (vakul rod mobon can be taken in the transient analysis. C noother method must be used to bring the reactor core to shutdown. The alternative method available at plant is the uso of the standby liquid control system which injects a sodium pentaborate solution purpose of analysis. the shutdown using the standby liquid control system was assumed to occur as has 5 minutes to evaluate the situation and decide toinitiate the injection of the liquid poison. Thirty seco the transpor1 time of the liquid f rom the storage tank to the vossol and to become effective in the core.
then assumed to be inserted I:nearly until hot shutdown is achieved.
Having brought the reactor to the hot shutdown condition, time is now avaitable for the operator be done next. If the postulated ATWS event has really occurred, he must take the necessary action to shutdown.
6.3.2 Containment For the purpose >s of tht3 iepott,the term Containment ts used f or allthe enclosed spaces affected by the Conta:nment sphere. Containment pressure and temperature refer to the condetron of the air water the containment spher e. Containment walltemperature refers to the temperature of the metal conten in its role as a heat source or sink with respect to thermalinteraction with the air-water vapor system inside response refers to the behavior of the pressure and temperature of the air water vapor mixture inside t sphere.
The reactor vessel steam flow used for the calculation of containment response is shown in Fegures 2 and 3 n plotted from Reference 1. Figure 15. Run 3. Here the vessel steam flow f rom 0 to 40 seconds is shown 306 7
b e
500 --
3<
i l
1 I
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30
'40 O
10 20 TIME insc)
VesselSteam Flow Transient Used on the Containment Response Analysts in en A TV/S Load Rej t
Fogure 2.
Bypass. Data Taken from the Safety Valve Sizong Analysis of Reference 1 Figure 15. Run 3.
P00R ORIGINAL
r t
en o
attains a steady value at about 40 seconds. This is assumed to remain so until the liquid control becomes etfective in the core and thereaker is assumed to linearly drop to the flow corresponding to oecay heat. This is shown in Figure 3. The steam release into the containment sphere is the vessel flow minus the emergency condenser flow, as shown in Figure 3.
200 STE AM F LOW TO CONTAINMENT VESSEL FLOW 3
Y
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too 3
6t; EMERGENCY CONDENSER FLOW r-STEAM PLOW DUE TO DECAY HEAT.
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100 200 300 400 500 000 700 300 900 1000 TIME (set)
Figure 3.
Vessel Steam Flow Emergency Condenser Flow, and Steam Flow to Containment Usedin the Containment Response Analysis on ATWS Load R0jerction Without Bypass. (Initial Part of Vessel Flow Transient Replotted from Fogure 2.)
T he steam release into tt e containment sphere causes the containment temperature and pressure to rise. This rise is part'ially offset by the containment wall which absorbs heat from the air-water vapor system inside. The containment wall temperatur e also rises by heat absorption. The containment sprays are assumed to be initiated by the operator at 5 minutes.
After the steam release inte.ae containment is terminated, the cooling effect of the spray recools the containment. The containment wall also retools later on losing heat as the containment sprays cool down the air-water vapor system to a temperature below that of the wall.
The containment pressure ar d temperature responses are shown, respectively,in Fsgures 4 and 5. The containment w.ill temperature transient is also s'iown in Figure 5. The peak values of the containment pressure and temperatures are, im>pectively; 57.7 psig and 286 F.
The he.it transter coeffcien between the containment air-water-vapor system and the wallis assumed to be a I
function of the air-to-water vapor wiight ratio. The functional relationship that is used in the analysis is shown in Figure 6, which also shows the experimental c ata from Reference 5. The relationship used in the calculations is a simple modification of the data from Reference 5.
f-
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i P00R ORIGINAL
t
.=
GE CRITERION FOR CONTAINMENT PRESSURE 1
cr h
CONTAINMENT PRESSURE
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _E SSU R E CONTAINMENT DESIGN PR 8
0 0
500 1000
-1500 3000 TIME (sec)
Fogure l.
Containment Pressure Response un an ATWS Load Rejection Without Bypass _
300
/p#~
~_--
/
/
/'
/
CONT AINMENT TEMPERATURE (SEE DEFINITION IN TEXTl
-u
/
B
/
3
/
e
/
e 1oc
'*MPERATURE OF CONTAINMENT WALL (ACTING AS A >8 EAT SOURCE / SINK. SEE TEXT FOR DEFINITION) 9 I
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0
'2000 e
Soo 1000 1500
[
f TIM *, leec)
Figure S.
Corstainment Temperature and Containment Wall Temperature in an ATWS Load Rejection Wothout Bypass 6-5
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140 l
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=
8 cc~
1 E
l DAT A FROM REF ERENCE 5, FIGURE 20 l
.0 f
k' WALL HEAT TRANSF ER COEFFICIENT USED IN THIS REPORT
\\
'i t/- - - _ _
l 1
I l
\\
l l
1 r ~ --I - - 1 0
2 4
8 s
10 12 14 16 TS
'Je
- 72 24 o
AIR / WATER VAPOR RATIO Fogure 6.
Contamment Watt Surface Heat Transfer Coefficient as a function of Air-to-Water-Vapor Ratio G.4 COMPARISON TO WASH-1270 Appondix A. Paragraph 11 C.1 of WASH-1270 requests comparison of three specified functions to specified analytical go. des This compareson is shown in Table 6-1.
Table 6-1 FULL LOAD REJECTION WITHOUT BYPASS AND WITHOUT SCRAM Functional W ASH-1270 General Comparison Comparison Electric Value By Parameter Value Suggested Guide Analysis Vessel Pressure (psig) 2040 2700 1%7 Foot Enthalpy (cal /gm) 280 280
<165 Cladding Oxidation (%)
17 17
<1 Containment Pressure (psig) 27 54 57.7 i
P00R BRIGINAL 1
N r
t,
- g s
7.
REFERENCES
.1
\\ \\ J l k unn s and F. C. f Lkot t. lo;enssont Analysos Consurners Powor Company Beg Rock Pount Plant. General Electrec Cornpain,0.toler 1962 (Al'LD-4093) 2.'
Consruners Power Company, Bog Rock Poont Plant, Technical Specifications, (Revised to February 24,1975).
3.
Fonal Hazmds Summary Report, Big Rock Poont Powcr Plant, Consumers Power Company (November 1961).
4.
Letter from James S. Rang of Consumers Power Compaay to Ray Fairfield of General Electric Company, dated September 11,1975.
S.
lL Uctuda, A. Oyama, and Y. Togo, Galuation of Post-Incident Cooling Systems of Light Weter Power Reactors, Third international Conferonce on the Peaceful Uses of Nuclear Energy, pp. 93103 (in particular, Figure 20), New York (1965).
f 71/72
l.-
i
\\
' APPENDIX A PARAMETRIC AND SENSITli'TY STUDIES OF THE CON TAINMENT RESPONSE llwe gu ni ao il thr. p.m iniaeite *.tu.f r is tu m.ike the onal sm ns lotlosive eis possitilo over the life of tho plant 11 a40 i
helpn to e mer tulain uncoitatntitra in the input data used for the evaluation of the containment tesponso.
A.1 RECIRCULATsON PUMP TRIP Recirculation pump trip is an effective method of reducing the severity of an ATWS event that is being considered for many of the other BWR plants. Trip
.of the recirculation pumps reduces the core flow and, therefore, the core power (by increased core volds) and steam output into the containment. According to present plant operating procedures, the operator must immediately trip one re-circulation pump on certain plant transients. Similarly trip of both the recir-culation pumps can be initiated by the operator upon the occurrence of an ATWS.
The containment peak pressure and temperature for recirculation pump trip initiated at various times after the ATWS event were calculated and are shown in Table A-1.
The time ewiution of the containment pressure and temperature responses are shown in Figures A-1 and A-2, respectively Table A 1 CONTAINMENT RESPONSE TO ATWS: WORST REACTOR ISOLATION (LOAD REJECTION WITHOUT BYPASS)
Initiation Containment Containment Time for Peak Pressure Peak Temperature RPT (sec)
(psig)
(' F)
G 25 0 234 60 27.5 240 120 31 0 247 150 33 0 251 3
180 351 2 54.5 300 43.9 269 No RPT 57.7 286 The vessel steam flow used f.n the containment response e' aluation with recirculation pump trip was derived as f ollows.
Iiw. etfee t of tripiung the recuculation pumps on the steady, tate eperation of the roactor was studied in Reference 1 anel r.hown in l~n pu e 19 of that eefes ence 1bero the of fect of rec culation pump top on the steam flow is to reduce it to 40%
ovos atu nat t 7.ett,nds after a delay of abouV secn. ids, and th.n 1o ins case #1 to a steady-state value of about 50%. In the pu sent case, the of fet;t of the rocirculation pump inp is assumed to decrease the vessel stc'im flow to 50% in 12 seconds aller a delay of 2 seconds The base for this assumption is Figure 19 of Referenco 1 and the fact that the contain:r.9nt sesponso es not very &nsitive to the details of the steam tiow over short intervals of time (< 10 seconds). Thus the vessel steam flow witn recirculation pump inp is given by
~
~Vossel Steam Flow'
~ Vessel Steam Flow ~
iMultiplication Factor 4
Shown in Figure A-3)
(A1)
Without Rocirculation x
With Recirculation Pump Trip
_ Pump Tnp A.2 STEAMINO RATE
/
t The long-term response of !he nuclear boiler in ATWS dopends, among other thin {,,s on the cparaCteristics of the fuel which change w t1 soloads For antainment considerations, this restets in a change in the steam dump before reactor shutuown. To cover this aspect of ATWS transents, containment response was calculated with the vessel steam flow (shown in Fsgure 3) multiplied by a f.ictor of 0 8 and 1.2. The resulting peak containment pressures and temperatures are shown 6n Table A 2.
A1 r
Table A.2 CONTAINMENT RESPONSE TO ATWS: WORST REACTOR ISOLATION (t.OAD REJECTION WITHOUT BYPASS)
Containment Containment Containment Steam Peak Prassure Peak Temperature Flow Multipilcation Factor (psig)
(*F) 00 43*
19.7' 267*
222' 1.0 57.7*
25 06 286*
A34*
1.2 73.0*
30.5*
302*
246' 300 300 CONTAINMENT TEMPERATURE W
(SEE DEFINITION IN TEXT)
=
w S
100 4
e
.i l
I I
a O
S00 1003 1508 Entr TIME (soc)
Fogure A t.
Containment Pressure Response on an ATWS Load Rejection Wothout Oypass, Wuth Recirculation Pump Trip at 6 seconds P00R ORIGINAL i
(
'A2 v
~
b
.=
bbNUN DUlbdb
-- =-
I
'S 1
e
_ $$$$$$$b$
- - -- - - - ~~ ~ ~
ss CONTAINMENT PRESSURE 0
0 500 9000 1500 M
TIME (sec)
Fogure A 2.
Containment Temperature Transientin an ATWS Load Rejection Without Bypass, With Recirculation Pump inp at 6 seconds
> ?
1.5
'(IME AT WHICH RECIRCULATION PUMP TRIP IS INITIATED e
7 o
1.0
=
9-5 k
DELAY TIME
'2 -i 5
g u
I I
I I
I I
I I
I I
I I
I I
I I
I
'l i
TIME isec) l Fogure A-3.
Multiplication Factor Used on Equation A-1 for the Eflect of Recirculatron Pump Trip on Vessel Steam Flow.
This Plot is Based on Fogure 19 of Reference 1.
P00R Bhitilhat
~
m L
O A.3 WECT OF EMERGENCY CONDENSER The r eduction in thermal load on the containment caused by the emergency condenser depends on the number of 6ts tube bundles in operation. The effect of this on the containment peak conditions is shown in Table A 3.
Tabte A 3 CONTAINMENT RESPONSE TO ATWS: WORST REACTOR ISOLATION (LOAD REJECTION WITHOUT RYPASS)
No. of Emergency Containment Containment Condenser Tube Peak Pressure Peak Temperature Sundles in Service (psig)
(*F) 2 57.7*
25.0*
286*
234*
1 61.7*
27.3*
291*
240*
None 66.0' 30.56 295*
246*
- Wdh no bp of recrculaton pump
- With recteuiaton pump tro et 6 sec.
A.4 SLC INITIATION TIME Histoncally, a delay time of 10 minutes has been assumed, following an unforeseen event, to take cred.t for operator
(
action initiation or control of an emergency core cooling system. For the postulated ATWS event, however, the lack of scram woutd be evident within seconds. For this reason it would seem thatinitiation of the SLC system at less than 10 minutes could be justified. The base case results (shown in Figures 4 and 5) use a 5-minute initiation time. To show the effect of SLC initiation time. Tabte A-4 shows the peak containment pressures and temperatt..es for three different values of the initiation time.
Table A-4 CONTAINMENT RESPONSE IN ATWS: WORST REACTOR ISOLATION (LOAD REJECTION WITHOUT BYPASS)
SLC Containment Containmer.
Initiation Time Pressure Peak Temperature (sec)
(psig)
('F) 600 96.2*
37.6*
322*
259' 300 57.7*
25.0*
286*
234*
180 42.3*
20.5*
266*
233*
- W4n no to of rectoaletion pumps.
- With recrculaton pump to et 6 seconds.
u r
]
,s-4, A.5" SENSITIVITY OF CONTAINMENT RESPONSE TO WALL HEAT TRANSFER COEFFICIENT The heat transfer coeffcient between the air water vapor mixture in the containment andthe containment wallusedin the base case results of Figures 4 and 5 was based on the experimental riata from Rc!erence 5. However,the experimental data of Reference 5 were fof a vertical surf ace of 14 cm width and 30-cm he',ht. The geometry of the containment sphere is thus ditforont f rom that in Reference 5. For this and other reasons, the heat transf er characteristics in the two cases may be dif foront. To provide a feeling for the etfect of this on the containment response, analyses were made using half and twicethe value of the heat transfer coefficient shown in Figure 6. The results are shown in Table A-5.
i Table A-5 CONTAINMENT RESPONSE IN AN ATWS: WORST REACTOR ISOLATION (LOAD REJECTION WITHOUT BYPASS)
Containment Peak Containment Peak Pressure Temperature (psig)
(*F)
Containment Wall Heat Transfer Coefficient from Figure 6 57.78 256 286*
234' Containment Watt Heat Transfer Coefficient Equal to Half the Value from Figure 6
- 60. t
- 27.1*
289*
239'
(
Containment Wal! Heat Transfer Coefficiert Eque, to Twice the Value from Fgure 6 57.0 22.5 285 229 d Wim na trip of rettculaton punWS NFCuletM pump try at 6 sec.
A 5/A-6
[