ML20046C702
| ML20046C702 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/06/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9308110388 | |
| Download: ML20046C702 (13) | |
Text
{{#Wiki_filter:) GENuclearEnergy - GewasEtenacCompany 175 Cumet Avenue. 5an kse, CA 95125 August 6,1993 Docke: No. S1N 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate , Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation i
Subject:
Submittal Supporting Accelerated ABWR Schedule - Requanification or - PRA (Early Venting) Item #06
Dear Chet:
Enclosed is a draft analysis that demonstrates that the containment pressure can be controlled by venting early in the sequence preventing the pressure increasing to the high drywell pressure containment isolation set point. In addition, this analysis demonstrates that core cooling can be maintained by RCIC or 11PCF for eight hours. i Please provide copies of this transmittal to Bob Palla and Glenn Kelly. Sincerely, SC*f Jack Fox Advanced Reactor Programs cc: Alan Beard (GE) Jack Duncan (GE) ' Norman Fletcher (DOE) Sid Smith (GE) ; 4 I i P JtWJU 1100C7 pY i 9308110388 930806 PDR ADOCK 03200001 D 3 ll I , j A PDR kp/ ig { l b
9 08/06/93 : > EARLY CONTA'NMENT VENTING i Introduction Following reactor isolation and residual heat removal (RHR) failure, early containment venting is an action whereby the contah. ment is vented during a period of increasing drywell pressure prior to reaching the containment isolation setpoint. By venting the containment and preventing it from isolating and continuing to increase in pressure, the time to recover the reactor decay heat removal (DHR) systems is extended. i The essential elements to be considered in the analysis of the capability for early containment venting are:
. Capabilities and operation of the Containment Vent and Purge system; ,
. Operator actions as defined by the Emergency Procedure Guidelines (EPG);
. The definition of the postulated early containment venting sequence;
. Discussion of the event recovery conditions; p . The analysis of the postulated event sequence 6 demonstrate systems capability; Source term analysis; Release analysis.
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Based on the analysis of these eie'menks, i i(s conclu\ ded that under the postulated I conditions, the containment-can he ve$ted ly'si' the sequence preventing containment pressure from increasing \to tlie dr$vell INgh pressure containment isolation setpoint. Further, reactor core isolathn co'olik (RDIC) or high pressure core flooder (HPCF) can be used for core cooling for least eisht hours. !
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Containment Vent and Purge System ! The contaimnent vent and purge system has two 22 inch (550mm) purge lines from the i
! containment; one from the wetwell and one from the drywell (see Figure 1). The ;
containment vent and purge lines exhaust through a common line to the reactor area (R/A) ! heating ventilation and air conditioning (HVAC) system which subsequently exhausts to ! the atmosphere through the plant main exhaust stack. These lines are normally used only for inerting and deinerting the containment at the beginning and end of a normal operating i cycle. These lines m normally isolated during the operating cycle but the isolation valves can be remote manually opened from the control room providing the isolation valves have not received an isolation signal from high drywell pressure, reactor pressure vessel (RPV) level 3, high radiation in the reactor building HVAC air exhaust, or high radiation in the fuel handling area exhaust. There is no provision for manually overriding the isolation . signals. However, the purge and vent valves may be opened remote manually subsequent i to isolation by resetting the isolation logic after the isolation signal has been cleared. Operator Actions The EPG action levels are essential parameters for defining the proper operator actions to ! be taken for early containment venting. The plant operator will take action that includes
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venting the containment based on the Emergency Procedures Guidelines (EPG). The EPG that addresses containment venting is 18A.5, " Primary Containment Control Guidelines." This EPG defines six entry conditions for the prescribed actions. The six entry conditions are:
. Suppression pool temperature above [43.30C (most limiting suppression pool temperature LCO)]
. Drywell temperature above [57.20C (drywell temperature LCO or maximum normal operating temperature, whichever is higher)]
. Drywell pressure above [0.12kg/cm2g (high drywell pressure scram setpoint)]
. Suppression pool water level above [7.1 m (maximum suppression pool water level LCO)]
. Suppression pool water level below [7.0 m (minimum suppression pool water level LCO)]
. Primary containment hydrogen concer.tration above [Hi Alarm (high hydrogen ,
alarm setpoint)] , If any one of these entry conditions occur, the operptof is instructed to execute actions to monitor and control suppressio, pool temperature,'yrywell temperature, containment ! pressure, suppression pool level, and hydrogen,and oxygen concentrations. i X Early Con'ainment It is postulated Ventine that an eventvecurs t olst thSecue%jat reactor by closingis\ the main steam line ' isolation valves (MSIVsf\and ke' reactor s' crams due to the closing of the MSIVs. It is assumed the evera is repr'esentat{ve%f aDteam line break outside the containment (i.e., l high main steam line flow, high ternperature in the main steam tunnel, etc), which does not result in containment isolatiorhdiowever, the event analyzed is not a station blackout or failure of suppon systems such as reactor service water (RSW). A station blackout or - failure of RSW would also fail the drywell coolers and the drywell pressure would rise to ; the high drywell pressure setpoint due to loss of drywell cooling before the operator could : take action to vent the containment. The reactor relieves steam to the suppression pool through the safety relief valves (SRVs). All divisions of the suppression pool cooling mode of the RHR system are assumed to fail. Therefore, there is no suppression pool cooling. Since the RHR is designed to only function at low reactor pressure and the reactor is at high pressure, the RHR and the low pressure flooding flooding (LPFL) mode of the RHR are not available. The RHR and the LPFL may not be available even at low reactor pressure (normally following ADS) because the failure of the suppression pool cooling mode may be the result of an RHR , system failure. However, assuming one of the high pressure injection systems is available, l core cooling can be provided with the reactor at high pressure and discharging steam through through the SRVs to the suppression pool. The reactor water level would normally be maintained by feedwater. If feedwater is not available, reactor water level will drop, the purge and vent system will isolate on level 3, , i f 2
4 and RCIC will initiate when RPV water level drops to level 2. The RCIC takes suction from the condensate storage tank. Although the operator is instructed to initiate RCIC on loss of feedwater (this probably would before the RPV water level decreases to level 3), for the purposes of this analysis it is assumed the operator does not take the prescribed ; actions prior to the automatic initiation of RCIC. ; The continuing SRV discharge sesults in a rise in suppression pool level. The temperature of the suppression pool and containment pressure and temperature also continue to rise. The drywell and wetwell pressure will tend to equalize through the vacuum breakers. In - order to prevent the containment pressure continuing to increase to the high drywell ! pressure isolation setpoint, the operator must open the valves on the vent line from the wetwell at a pressure below the high drywell pressure isolation setpoint (0.12kg/cm2), When the RPV level decreases to level 3, the RPV L3 signal will isolate the containment , vent and purge system and the R/A HVAC air ducts. The level 3 signal will also initiate i the standby gas treatment system (SGTS). When the reactor water level has recovered above level 3, the operator will reset the reactor water level 3 logic and proceed to take action to vent the containment. The operator will, isolate the SGTS, open the isolation valves on the HVAC, start the HVAC fans, and4 pen the isolation valves on the vent line from the wetwell. Since there has been no core damdge, there is no significant source term. However, if the containment ressu'ta sexc'eeds the 0.12kg/cm 2 setpoint or if there is high radiation in the reactor buildmg VAC'exhliust line,lhe vent and purge system will isolate and the operator cannot ohrri e thdXoihtijn signals.
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Sequence Analysis l The sequence description ab .ve does not consider the time at which each of the relevant limits or action levels occur dunng the event. The point in the sequence at which the ' respective EPG action level is reached is very important because it determines what actions the operator is directed to take and the time available to take those actions. This event was evaluated using the SHEX computer program. The results of this analysis are ' given below. The analysis assumes that RHR and suppression cooling has failed, RPV water level is l maintained by RCIC taking suction from the condensate storage tank (CST), the reactor is discharging through the SRVs to the suppression pool, and the drywell coolers are functioning. If the drywell coolers also fail, the drywell pressure increases at a rate that ; will not allow sufficient time for the operator to take action before the high drywell ; pressure containment isolation setpoint is reached. However, if the event sequence occurs approximately as described above, the EPG entry condition that will occur first is : suppression pool high level. The suppression pool level will continue to rise as the SRVs continue to discharge into the pool. The suppression pool level will increase to the EPG action level in about four minutes (see Figure 2). In addition, the suppression pool - temperature will rise and this EPG entry condition will occur in about ten minutes (see Figure 3). Based on either of these EPG entry conditions, the operator is instructed to monitor and control the primary containment pressure below 0.12kg/cm2 If no action is 3
taken, the containment pressure will increase to the containment isolation setpoint in about l one hour (see Figure 4). Therefore, the operator will have appropriate information as early as four to ten minutes to er.ecute the actions described in the EPG and can vent the
- wetwell before the containment isolation setpoint is reached.
i Assuming feedwater is not available, RCIC will initiate on low reactor water level (level 2) I and will maintain RPV level. If RCIC is not available, HPCF will initiate when RPV water level drops to level 1.5. Assuming RCIC is available, it takes preferred suction from the condensate storage tank with the suppression pool as a secondary source. As the RPV l steam continues to discharge to the suppression pool, the pool level increases to the high , level setpoint in about 40 minutes and the RCIC suction automatically switches from the . condensate storage tank to the suppression pool. However, the operator is instructed by ; the reactor level control EPG to override the RCIC automatic switch from CST to suppression pool and manually control RPV level between level 3 and level 8. This action avoids the cycling of the RCIC on at level 2 and ofrat level 8 and the operator having to reset the containment isolation logic and reopen the valves on the wetwell vent line each : time the level 3 isolation setpoint is reached and subsequently cleared. Over a period of , eight hours, the reactor will discharge 1.08 x 106,lhfii of water to the suppression pool. The condensate storage tank has a water volume 'of 1.26 x 106 lbm. Therefore, the condensate storage tank has adequate vojume for more than eight hours. In addition, since .
; the RCIC is taking suction only from the} CST,ihn temperature of the water for the entire ,
eight hours is less than the RCIdesign te rahre of 1700F. The limit of the postulated sequence is determined byth ic chde'nsate storage tank volume is exhausted. j
- If HPCF with suction fron; the CST was used in this analysis instead of the RCIC, the results would be essentiallkthe}same. The temperatures, pressures, and levels of the j essential parameters at the end of eight hours would not be significantly different whether ,
HPCF or RCIC are used. However, the HPCF is designed to 340oF and may be used to maintain core cooling taking suction from the suppression pool regardless of the suppression pool temperature or availability of the CST. If HPCF is used with suction , from the suppression pool at some time during the event, the suppression pool level would . not continue to increase. Event Recoverv ; Recovery from the postulated event is controlled by the EPGs. As indicated above, the ! event is assumed to continue no longer than eight hours. The conditions at the end of the
- event are expected to be as follows:
l
. RHR recovered; !
. Suppression pool at 2120F and steammg at approximately 25 lbm / sec.; ;
. Suppression pool level at 8.26 meters; !
3 . 0.18 x 106 lbm of water remaining in the CST;
. Wetwell vented with the wetwell airspace filled with steam from the .
I suppression pool; l 4
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. Drywell isolated and filled with original inventory of N 2and some steam that was accumulated prior to the venting of the wetwell; ,
. Drywell temperature and pressure within normal limits;
. Reactor pressure at 1065 psi; f
. Reactor water level between L3 and L8.
Assuming the above conditions reasonably represent the conditions at the end of the event, there are no adverse conditions identified that would prevent recovery in accordance with j EPGs. Generally the EPGs require RPV depressurization, establish cold shutdown and . return the suppression pool to within normal limits. There are no identified conditions that would require wetwell or drywell spray initiation. It is expected that the containment (both wetwell and drywell) will be purged of N 2and residual steam as part of the recovery , process. , I Soureg Term Analysis A brief analysis has been conducted to evaluate the potential for isolation of the vent and purge system due to the source term discharged through the SRVs to the suppression pool , and wetwell airspace. g . f s In a typical operating BWR, the primary source of vola' tile radioactive isotopes is the N16 produced by in core neutron irradiaticrhf obgen iso' topes in the reactor water and ; pea of rypton and' Xenon) which are a byproduct i radioactive of the fission process.noble 01 gases ra (primaril@dioisotop ' Ugh present in small quantities in the j reactor water and steamkare t\IIf5ciently vo atile to present a significant source of airborne radioisotopes in the wetivekairs' pace. Due to the 7.7 second halflife of N16, it i
-l will decay to negligible qu itie'p within a minute after scram. Since the event being analyzed does not result in an environmental release for at least several minutes, the N16 produced is ignored for this analysis. ]
The radioactive noble gasses considered in this analysis originates from two sources. The first source is the noble gasses produced within the fuel rods which have been released to < and remain present in the reactor steam. The second source is the additional noble gasses l produced by the fuel rods after scram. The radioactive noble gases present in the steam in the pressure vessel water are being i released at a rate of approximately 5,000 pCi/second (adjusted to a 30 minute decay time for reference purposes - see Note 1). Typically, the steam in the reactor water contains about three seconds production of noble gas inventory. It is assumed that the entire three l second inventory (15,000 pCi) is disch; d to the wetwell airspace Adjusting the l' assumed noble gas inventory back to zero e results in a total release of 0.155 curies of noble gases to the wetwell airspace from this source. { The additional noble gases produced in the co.. Jter reactor scram originates from two ! sources, recoil and plenum. The total release rate from these sources is approximately l 5000 pCi/second. The recoil source results from fissioning of transuranics contaminating 1 5
the surfaces of the fuel rods. It is estimated that 40% of the normally expected noble gas release rate of 5,000 pCi/second is from this source. Upon reactor scram this source becomes zero and has been neglected for this analysis. The remaining 60% of the normal release rate originates with the build up of fission products in the fuel rod plenum from the fuel matrix. This source continues to purge noble gases to the vessel water tnrough microscopic cracks in the walls of the fuel rods as long as there exists a sufficient temperature differential between the rod wall and the water. It was estimated for purposes of calculation that this condition will continue for approximately 24 seconds after reactor scram. Minor leakage will occur after this time but is neglected as a small fraction of prior t releases. Therefore the plenum release for 24 seconds after scram at a of rate of 3000 Ci/second (60% of 5,000 pCi/second), which when adjusted back to time zero, results in a total release of 0.674 curies of 15 isotopes of xenon and krypton to the wetwell airspace. ! It has been assumed the 0.829 curies (0.155 plus 0.674 curies) are discharged to the wetwell airspace and are homogeneously mixed over the entire volume. The volume of the wetwell is approximately 6000 m3 This results in isotopic concentrations shortly after SRV operation of 10-5 to 10-9 Ci/ml in the wetwell airspace depending on the specific isotope. f \ Release Analysis s \
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The release analysis is based the sys{emSonfigu(ation illustrated on Figure 1. When the , water level in the reactor vesselfrop4 to lev \3hthe containment vent and purge system will receive an isolation pgnals R/A AGfanswhl terminate operation and the isolation valves will close, and the\SGTSg shrt. 'Therefore, in order to vent the wetwell, when the reactor water level rechvers to a ovelevel 3, the operator will clear the level 3 logic, terminate SGTS operation,\initikte the operation of the R/A HVAC fans, open the ; isolation valves in the reactobtIuilding HVAC and then open the isolation valves in the wetwell vent line. It is conservative to assume the wetwell airspace is at a pressure of 1.7 psig. If the : pressure were higher, the vent line could not be opened. Approximately 700 3m [((Pi / P2 ) x V 3) - V ] iof the wetwell volume will be discharged as the wetwell airspace pressure , decays to zero psig after opening the wetwell vent line isolation valves. Conservatively assuming the wetwell is vented approximately four minutes after scram with isotopic , concentrations of 10-5 to.10-9 pCi/ml, this will result in a release ofless than 0.03 Curies of noble gases to the environment. This total will decrease rapidly with hold up in the wetwell. Assuming a conservative annual average meteorological dispersion constant for a stack release of 2 x 104 sec /m3 , the resultant isotopic concentrations at the site boundary based on isotopic concentrations at the initiation of the event would be in the range of 10-11 to 10-14 pCi/ml, depending on the isotope, all of which are less than one percent of unrestricted concentration limits of 10 CFR 20. The R/A HVAC system includes locally installed radiation elements to detect radiation in the duct and isolate the system upon detection of radiation levels exceeding setpoint levels. A typical installation is a beta and gamma GM detector installed in each of several 6
places outside the R/A HVAC duct. A typical setpoint value is in the range of 50 to 100 mr/ hour. Conservatively assuming the radionuclide concentrations at the initiation of the event of 10-5 to 10-9 Ci/mi flow past the R/A HVAC radiation elements in an undiluted 700 m3 volume at the normal R/A HVAC flowrate, the detected dose rate would be approximately 0.1 mr/ hour. This is significantly below typical R/A HVAC high radiation setpoint levels. The radiation level in the R/A HVAC duct was estimated using the following equation: R 2A S= - 14F ;r where: S e detector rew.>e in mr/hr R s source term passing the detector in pCi/sec F s total flow rate in cfm A a total 110w area in cm2
/g 1 This equation provides a convenient method tmg dose f*ora source rates assuming estima\
at hinutes,'the average is 0.904 MeV/ dis.), consisting and an uncollidedof 1(reasonabfy flux MeV/ disintegration go'od
' aks gammas (dmptkn for th'e dis
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here)into the detector. Assuming a release time of 0 sec ,ndgto a low for the containment vent valve to open, R becomes 3000 pCi/sec. The t al flow rate (F) is approximately 200,000 cfm. Assuming a HVAC duct radius of approximidely 1 meter, A is 31,400 cm 2. Therefore the calculated dose rate is 0.1 mr/hr, which is significantly less than the typical 50mr/hr setpoint used for HVAC radiation elements . Summary The analysis of this event sequence demonstrates that the containment pressure can be controlled by venting early in the sequence preventing the pressure increasing to the high drywell pressure containment isolation setpoint. In addition, core cooling can be maintained by RCIC or HPCF for eight hours. Although this postulated sequence involves . the venting of the containment during an event that is beyond the design basis of the plant systems, the health and safety of the public is protected in two ways. By venting the containment and preventing the containment pressure to continue to rise, the time to recover reactor DHR systems is extended. However, if the event further degrades and results in core damage or containment pressurization, the containment will isolate to prevent any significant source term release to the environment. 1 Note 1: When considering the production of noble gases in the core of a BWR, the quantities of l noble gases released is typically referenced to a 30 minute decay time. That is, the total 7
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quantity of noble gases produced in the core is decayed for a 30 minute period and the number so produced is the referenced quantity. The reason for this is primarily one of operational capability to measure the quantities of noble gases produced. In-vessel and , steam line measurements are dominated by the N16 radiation signature so that j measurements of nobles gases cannot be accurately taken. Typically, measurements of l noble gases are made in the Steam Jet Air Ejector or down stream from this system where other short lived isotopes and N16 have decreased to negligible quantities. This runs from i 6 minutes to 15 minutes after evolution from the core and depends on both plant design and power level. Therefore, historically all measurements have then been referenced to 30 minutes as a single point for comparison purposes.
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