ML20079R810
| ML20079R810 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 02/07/1994 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9402140325 | |
| Download: ML20079R810 (33) | |
Text
1,
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GE Nuclear Energy Gened Elecinc Camps 175 Chitm henne Sr Jxt CA 95125 L
February 7,1994 Docket No.52-001 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactr
'mlation
Subject:
Submittal 5 ing' Accelerated ABWR Schedule-
')
Response to open Item F19.2.3.3.7-1 j
-1
Dear Chet:
j Enclosed is a SSAR markup addressing the subject open item pertaining to equipment survivability.
q Please provide a copy of this transmittal to John Monninger.
Sincerely, a
J
- k Fox Advanced Reactor Programs cc:
Joe Quirk GE) i Alan Beard GE 1
Carol Buchholz GE Jack Duncan GE Norman Fletcher DOE) t 100002 JM94414 Fl
. ) C (O -
9402140325 940207 PDR ADOCK 05200001 Q
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23A6100 Rsv. s ABWR Standard Safety Anslysis Report from the remote shutdown panel or even by local control at the motor control centers and switchgear. Following restoration of power and initiation of the reactor cooling water system, the ECCS areas of secondary containment will be cooled by their safety grade room coolers so normal operation of the safe shutdown systems could be restored. The turbine building electrical systems and the non-safety-related secondary j
cooling system provide a backup means of restoring cooling to the ECCS equipment areas within secondary containment.
19E.2.1.2.2.6 Conclusions The ABWR plant is being designed to be capable of maintaining core cooling and containment integrity for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the loss of offsite and onsite AC electrical power including the combustible gas turbine. This capability assessment follows the general criteria of:
(1) Assuming no additional single failures (2) Realistic analytical methods and procedures A summary of the key plant parameters, design basis values and capability assessment is '-
shown in Table 19E.2-2. Note that the response of the ABWR containment to this event would be successful even if the design basis values were exceeded, as long as the ultimate capability were not exceeded.
19E.2.1.2.3 Equipment Survivability JL "In order to ensure the operability of equipment during a severe accident as modeled in
{ the ABWR PRA, a review of the severe accident analysis was undertaken to identify all
) equipment considered in the analysis. In addition, the information required by Regulatory Guide 1.97, contained in Section 7.5, was reviewed to ensure that all l
essential instruinentation requirements for beyond design basis accidents were identified.The environmental conditions for each pi: ce of essential equipment were
( then identified. These conditions were then com standards. Thefe~q~uipmenJ/ qualification standardsgere not used as a strict measure.
Rather, they were used to provide a measure of cohfidence that the equipment would D'd survive the expected conditions.
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[ Two types of accidents were considered in developing the environmental conditio j
the equipment. The first, and generally more limiting accident is a severe accidentin I
which the core melts and the vessel fails.The emironmental conditions for this event are based on the insights obtained by the accident sequence aralysis considered in Subsection 19E.2.2. As the fuel melts the gas in the vessel heats up. The containment response is characterized by containment pressurization due to steam and non-condensable gas generation. If the vessel fails, then there may be high j
temperatures in the drywell for a short period of time due to the introduction of core i
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23A6100 Rsv. 3
'ABWR standudsafety Analysis Report r.j W
g debris in the lower drywell. It is important to note that the ABWR containment is
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I inerted. Therefore, there is no containment challenge due to hydrogen burning or
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detonation.
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cat T: ur. *_8*N.. corresponds q
g:5d h 10CFR50.34(f)(ix)ge egtgicguld result in the con Tc.his
_ e cem, core cooling is degraded sufficiently to j
] ]pW result inge generation of 100% oxidation of the active claddingle n:f, c, ore cooling isgecovered before the vessel failsffherefore, during this type of accident there. p o
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eillbe a sh6Yt penod in wnicn tne gas temperature in the vessel rises. However, after t ecovery of core cooling, the vessel atmosphere will again become saturated /This PRA I
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has confirmed the results of previop studies which have-show/that i: m "!dy #
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accidegadnitiated by transients. Therefore, a transient initiated event is sp y
for A.CO2i(f, evaluation [For a transient initiated event, the only effect on the
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'kq (containment is pressurization due to non-condensable gas generation.The analysis of.
. L l-* O o 19E.2.3.2 indicates that the peak containment pressure for this event is about 0.6 MPa, 1
M (which is well below service level C for the containment.
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19&iK.2.3.1 5;!mt "ht "e@ed *~ AMd-t "'t!;;2'c,,r.
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[The PRA was reviewed to determine the minimum set of equipment require l
accident mitigation. In this review [the critical functions of reactivity control, vessel inventory control, containment isolation and containment integrity were considered.
u MThe functions of reactivity control and containment isolation are required in the very early stages of an accident, during which all parameters are well within their design basis values. Therefore, since the survival of equipment to support these functions is assured, -
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this equi mentis not cor. side d here, alhw@l.
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6 co ns**c W Ce (The instrumentation required by Regulatory Guide 1.97 is specified in Table 7.5-2. In that table, the instrumentation is assigned a type according to its use during an accident.
Actions, either automatic or operator initiated, are only specified for Type A variables.
Qerefore, all other variable types are excluded from this evaluation.
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- Not all of the type A variables are considered essential for accident mitigation. Th fundamental reason for the exclusion of some Type Avariables is thafthe action wh f
would be specified by the variableg{. algo required if die oper status of the variable. "a am4r the neutron flux measurement indicate the CI - ' '
reactor has not been scrammed during an accident, the operator is required to scram.
thej;ractor._hsame action is specified if the operator cannot detegnine the neutron flux. 'imilarly, if the operator cannot determine the wetwell pressure, the drywell pressure or the suppression pool water temperature, the operator would initiate the RHR system. Furthermore, the COPS system provides an assurance that the containment will not fail, thus no venting strategies are required. In a similar manner, l-the EPGs require that the vessel be depressurized, either via the ADS system or' p
Deterministic Ans!ysis of Plant Performance - Amendment 33 19E1 12 i
F 23A6100 Msv. 3 ABWR StandantSafety Analysis Report s
manually, if the water level falls, therefore, the vessel pressure indication would merely demonstrate that successful depressurization occurred.
The drywell water level and the wetwell atmesphere temperature are not included in s
this evaluation because they are notjudged to be critical. It is of course important that any debris in the vessel is covered with water; however, this can be accomplished by the operation of the passive flooder which requires no instrumentation. Further, the operator is instructed to terminate injection into the containment based on the water level in the wetwell. The wetwell atmosphere temperature does not provide information which is essential to the management of a severe accident, and no accident management strategies have been employed in the ABWR design based on this measurement.
The final Type A instrumentation which has not been included in this evaluation are-the drywell/wetwell hydrogen and oxygen concentration measurement systems. During a severe accident, whether or not it is terminated in the vessel, the hydrogen concentration will rise. However, the' oxygen content will remain nearly constant, remaining below the lean flammability limit for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If there is any steam present, the potential for burning will be further suppressed. During the later stages of the accident, it will be possible to sample the containment atmosphere and perform a i
remote analysis of the oxygen content. Further, if a severe accident has occurred,'the need to operate the recombiners will clearly be indicated. Therefore, the normal instrumentation for containment atmosphetic monitoring is not critical to accident management operations.
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19E.2.1.2.3 Equipme$t Requiredor Accident Mitigation N
Based on a review of the PRA and the above discussion, the following equipment ha)s been identified as being important to accident mitigation)The basis for equipment sunivability haralsobeerr provided for each piece of critical equipment.
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(1) Depressurization System f
During a core damage event, the SRVs must be able to remain open during the in. vessel phase of the accident to ensure that any potential vessel failure
. occurs at low pressure. After vessel failure, SRV operability is not required.j The depressurization capability will not be degraded due to radiation
)
exposure because the design basis radiation exposure for the SRVs is higher than that predicted using best estimate severe accident methods. In addition the thermalloads on the elastomers in the valve actua:or, which limit the temperature capability of the valve, will be similar to that used for equipment :
qualification of the SRVs. Therefore, the operability of the valves will not be degraded by the conditions in the containment.
19E 213 Deterministic Analysis of Plant Performance - Arr,endment 33
23A6100 Rsv. 3 J
ABWR standantsatery Aratysis Report y bwtN The SRVs are held open wi h a nitrogen actuator. The nitrogen supply is t
located outside of containment. As discussed in Subsection 19E.2.1.2.2, the nitrogen supply will be adequate to assure SRV operability over a full range of y TM hypothetical accidents.
5 gx (2) Residual Heat Removal System (RHR)
The RHR system may be called upon to remove decay heat from the containment dudng a severe accident. Either shutdown cooling mode or suppression pool cooling mode may be used. The pressure and temperature of the suppression pool are not expected to exceed the system design pressure and temperature during any accident sequence. Shutdown cooling will only Jmf4 be used when the RPV is at relatively low pressures which are below the j
capability of the RHR system, approximately 1.0 MPa (135 psig).
The integrated radiation exposure to the RHR equipment will not reach the qualification limits for several days after system initiation during a severe accident. The RHR control system is outside of the containment and will not be significantly affected by a severe accident. The integrity of the system piping, spray headers and injection headers within the containment will not
( be adversely affected dudng an accident.
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(3) Firewater System und@
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The firewater system may be called upon to injec ater into the vessel or m % of M
d through the dqwell sprays during a severe acciden. The system is manually initiated. All flow in the system is from outside the containment. Thus,
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accumulation of radioactive materialin the firewater pumping system will not g/cd occur. All components of the firewater system are outside of the containment re accident. Inside the and will not be significantly affected during g@ ping and spra containment, the firewater system utilizes RHI whicheremendwod effccted-by+(eore-meltevene:<
i toen discussed.h O.
(4) Passive Flooder The passive flooder may be needed to provide a wa:er flow path from the suppression pool to the lower dqwell after vessel failure. The flow path is opened as a direct result of high temperatures in the lower dqwell which occur after debds relocation from the vessel. This system does not contain any active systems, instrumentation or controls. Additionally, the system components are not hindered from performing their functions due to high -
radiation levels which might exist in the lower dgwell after debds relocation from the vessel. Therefore, the system is expected to operate under the required conditions.
(5) Containment Overpressure Protection System (COPS) 19E.2-14 Deterministic Analysis of Plant Performance - Amendment 33
23A6100 Rev. 3 ABWR standard satory An:tysis soport X )do endt o (dcan The COPS may be needed during a severe accident to relieve high 6 % POPS si M containment pressure?The system contains piping, a rupture disk and two Q. L [co"fo elal-mlves which are normally open and fail open. To relieve containment
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p3,dct renckon pressure, the rupture dig must burst. Activation will not be adversely affected
.cAlced.
by the *...c._..
M radiation in the wetwell airspace during a severe accident. The sensitivity of nipture disk activation to wetwell temperature is discussed in Subsection 19F_2.8.1.2.
4 a "S*4 **M far oDeld"1 (6) Vacuum Breahers The vacuum breaker may need to open during a) accident to relieve high differential pr ssures between the wetwell and d,rywell. Vacuum breakers o
are passive in natu e and have no instrumentation and-control other than J
position indicatio jSince the vacuum breakers are located on stub tubes high 3 I
A m the wetwell airspace, they will not be subjected to pool swellloads.There is no direct means for debris to reach the vacuum breakers. Therefore, they are g _.
not expected to be adverselv affected_Aring a severe accidsnt.
(7) RIP Vertical Restraints dul 4L, ;,s - ves: stb The vertical restraints on the RIPS prevent the pumps from being ejected if the g,
of./L, addeh RIP attachment welds are destroyed during a core melt event. The restraints S.hce.144 mkaml3 5 re attached to the oWiressel smf and do not experience the severe midsdl sc, Ot.
}c nditions within the vessel during core me ng. Therefore, the integrity of art he vertical restraints is notjeopardized4- -
g}so 10 $ Hof (8) Containment Isolation Valves k
}[ vcsal MU" fe containment isolation valveNe expected to remain closed during a f
[
evere accident. The pressure capability of the isolation valves does not limit occ w
the containment response to overpressurization events. The radiation exposure to which containment isolation valves are designed to are higher than that which is expected to occur during a severe accident. Thus, the (isolation valvm will na f2il durine a severe accident E' nsu+
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(,td tih VesselWater LevelInstrumentation 2-n
&+we A
/Xsevere accident can be terminated in-vessel if water injection to the vessel is recovered prior to vessel failure. For these sequences, the water levelinside the
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_TfvnW b vessel should be monitored to maintain core coverage.The ABWR contains FI "Y
RPV water level instrumentation which provides indication of conditions 1cading to inadequate core cooling, Identification ofand Recoveryfrom Conditions (Leading to inadequate Core Cooling ((I.F.2] (Subsection 1 A.2.16).
19L 2 15 Deterministic Analysis of Plant Pedormance - Amendment 33
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23A6100 lisv. 3 ABWR Standard Safety Analysis Report (10) Containment Water Level Precise water level determination in the containment is not necessary.
j However, the operator must monitor the approximate water level in order to avoid overfilling the containment and potentially damaging the COPS system.
Level sensors in the wetwell will provide adequate indication of containment water level, The level sensors in the wetwell are expected to be operable during the entire course of a severe accident because the benign conditions in the suppression pool. The sensors located in the lower dqwell above the floor may also be operable. Thus, the operator will be provided with indication to allow termination of a severe accident involving debris relocation from the-vessel.
(11) BefwelfTemperature Instrumer.tation If the drywell temperature exceeds the degradation temperature of the elastomers used in the containment penetrations, excessive containment leakage could occur. This can be prevented by use of the dqwell sprays.
Dnwell sprays are manually initiated upon indication of high dowell temperature. Control of the dqwell temperature so as to prevent excessive leakage should ensure temperature sensor operability. Further, the radiation levels in the containment will be below the design basis for the instmmentation. If the sensors may go off scale, dqwell spray initiation will be indicated.
19E.2.1.3 Phenomenological Assumptions This subsectior contains a summary of those phenomena which are not considered in an integral fashion using MAAP. These phenomena fall into two categories: those which are ruled out as being incredible for the ABWR and others which are neglected because they produce an insignificant change to the overall performance of the ABWR under severe accident conditions. A more detailed explanation' of some of these phenomena l
is given in Subrection 19E.2.3.
19E.2.1.3.1 Stoam Explosions Iarge scale steam explosions are deemed incredible. The geometry of the ABWR will prevent a sufIiciently large contiguous mass of corium from falling into water in either the vessel or lower drywell regions. A more detailed description of this phenomenon as l
well as the justification for its neglect is provided in Subsection 19E.2.3.1. Small steam explosions which do not in themsehes threaten the integrity of the vessel or containment are calculated by MAAP. Additionally, a scoping calculation is performed in Subsection 19E.2.6.7 to determine the mass of core material which could participate in a steam explosion without damaging the containment.
Deterministic Analysis of Plant Performance - Amendment 33 19E.216
e-4 Insert A-A The requirements for equipment sunivabilty are derived from two sources.10 CFR 50.34 (f) specifies the conditions required for an analysis in which the 100% of the active fuel cladding is oxidized. Addidonal requirements for demonstrating the survivability of equipment needed to trJtigate a severe accident are specified in SECY-90-016. In order to meet these requirements, three categories of events were considered. The first category consists of one event which responds to the requirements of 10CFR50.34(f) paragraphs (2)(i.x)(C) and (3)(v). A non-mechanistic scenario is modelled which results in the requisite oxidation but which follows the rules of design basis analysis. The other two categories respond to the requirements of SECY-90-016. The second category consists' of events representing the risk dominant events ending in in-vessel recovery. Similarly, category three is made up of '
events representing the risk' dominant events ending in ex-vessel recovery. Together the events in categories two and three represent 98% of the core damage frequency.
The list of required instrumentation and equipment was derived from reviews of the safe shutdown equipment, the EPGs, the PRA, and the severe accident analysis. The list of required equipment varies for the three categories of events described above.
The capability of each piece ofidentified equipment was then compared to the environmental conditions for the appropriate category of events. In reviewing the equipment capability, the environmental Insert A 19E.2.1.2.3.1 Definition of Sursability Profiles For each of the three categories of events, a set of curves representing the bounding environmental conditions for that category were developed for use in evaluating the equipment and instrumentation survivability. These conditions were then compared to the equipment capabilities to provide a measure of confidence that the necessary equipment would survive the expected conditions. It is important to note that the ABWR containment is inerted for all of the events described below. Therefore, there is no containment challenge due to hydrogen burning or detonation.
The basis for each category of events is provided below along'with a brief summary of -
the event progression.
19E.2.1.2.3.1.1 10CFR50.34(f) case Insert B Best estimate analyses do not result in oxidation of 100% of the active cladding. In order to simulate the hypothetical event, MAAP-ABWR was run using a multiplier to non-mechanistically generate 100% oxidation of the active cladding. The ' event -
progresses as follows:
1 1
An isolation event occurs.
All core injection is assumed to fail.
Drywell and wetwell sprays are initiated 30 minutes after the initiation of the accident, water flow is directed through the RHR heat exchanger.
The core begins to heat up and zirconium begins to oxidize.
ECCS is recovered.
Additional hydrogen is generated as the core is quenched.
Vessel water level is recovered, terminating the event.
Curves representing the environmental conditions during this event are shown in Figures g2Sa %qh /96.z,2Sf 19E.2.1.2.3.1.2 Severe Accidents Recovered In-Vessel
}.
This category is designed to represent the dominant in-vessel recovery sequences.
1 There are four sequences of this type that have a core damage frequency of greater than IE-9 as shown in Figure 19D.5.3. The events are LCHP-IV-N-N, LCLP-IV-N-N, LCLP-IV-R-N, and SBRC-IV-N-N.
The S13RC-IV-N-N sequence received a very conservative treatment in the PRA. In the SBRC-IV-N-N sequence the RCIC operates for several hours before it fails, due to the loss of sufficient battery power for RCIC controls. As discussed in Subsection 19E.2.2.3(1), the firewater system can be used to prevent core damage in this instance.
The probability associated di the successful use of the firewater addition system in the development of the com.inment event trees is consistant with prevention of core damage. However, this possibility was not modeled in the core damage event trees.
Therefore, for consistancy, no credit was taken for the prevention of core damage.
Nonetheless, the sequence indicated in Figure 19D.5.3 for SBRC-IV-N-N would not be expected to have core damage. Thus, it is excluded from further consideration for the purpose of assessing equipment survivability.
All of the re naining events in this category are initiated by transients with a presumed. oss of core cooling at initiation. The core gradually uncovers and heats up. Some core damage occurs, but core cooling is recovered and the vessel does not fail. In two of the sequences (those in which the s venth character is N),
containment cooling is recovered before the tu tre disk opens, while in one (with the seventh character R) the rupture disk opens to prevent the potential for containment failure. The curves shown in Figures Prepresent the bounding environmental conditions for this category of events.
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196.2 M 19E.2.1.2.3.1.2 Severe Accidents which Progress Ex-Vessel This category is designed to represent the dominant ex-vessel sequences. There are -
six sequences of this type with a potential core damage frequency greater than lE-9 in Figure 19D.5.3. The events are LCHP-FS-N-N, LCHP-FS-R-N, LCLP-FS-N-N, LCLP-FS-R-i
N, LCLP-PF-N-N, and LCLP-PF-R-N. For the high pressure melt sequences (LCHP), it is known that the drywell spray system is available since the sequence does not result in a penetration overtemperature failure (i.e. the seventh character in the sequence is not P). For the low pressure scenarios, the use of the firewater addition system cannot be distinguished from sequences with the passive flooder. Therefore, both methods of mitigation are considered.
The details of the core melt progression are discussed in Subsections 19E.2.2.1 and 19E.2.2.2. In general the accident progression is as follows:
A transient results in scram and containment isolation.
All core cooling is lost and the vessel water level fails, resulting in core uncovery.
The core melts and vessel breach occurs.
For the high pressure scenario, debris may be entrained from the lower drywell, so drywell sprays are used to cool the containment and quench the core debris.
For the low pressure scenario, either the firewater addition system or the passive flooder may be used to cool the molten core debris.
This category is characterized by core melt and vessel failure. As the fuel melts the gas in the vessel heats up. The containment response is characterized by pressurization due to steam and non-condensible gas generation. When the vessel fails, high temperatures are generated in the drywell for a short period of time due to the introduction of core debris in the lower drywell. The curves shown in Figures Mt 14Ef'%7A represent the-bounding emironmental condition for this category of events.
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/9c.1.-274 19E.2.1.2.3.2 Identification of Required Equipment and Instrumentation Three primary sources were used to identify the equipment and instrumentation required for the mitigation of either the 10CFR50.34(f) event or a severe accident.
10CFR50.34(f) requires that the equipment required for safe shutdown and containment isolation be considered, while the equipment and instrumentation -
required to survive severe accident conditions may be extracted from the discussion of the accident sequences in Subsection 19E.2.2. Additionally, all instrumentation which monitor plant variables required for operator actions were reviewed.
19E.2.1.2.3.2.1 Requirments for 10CFR50.34(f)
Safe shutdown is def'med in 10CFR50.2 for non-DBA events as hot shutdown. In addition,10CFR50.34(f) requires that containment integrity be demonstrated. Thus,
Insert C The 10CFR50.34(0 event does not impact the secondary containment in excess of the impact of design basis events.Therefore, equipment located iri the secondary containment is not considered in this review.
The core cooling function can be performed by the HPCF, the RCIC, or, following depressurization of the vessel, the LPCI mode of RHR or the firewater addition system (ACIWA). The operability of both LPCI and ACIWA will be demonstrated to satisfy equipment survivability for severe accident. Therefore, the survivability of HPCF and RCIC will not be considered.
Maintenance of containment integrity requires that. olation valves remain closed, and that excessive leakage does not occur through the containment penetrations.
For the 10CFR50.34(0 event, the RHR system is used to prevent containment overpressurization.
The required instrumentation was developed from Table 7.5-8 which contains a list of all variables required for manual actions. These are obtained from a review of the events included in Chapter 15 as well as the EPGs, as discused in Subsection 7.5.2.1. In one Case, Insert D
[ Same paragraph]
Thus, the instrumentation to determine neutron flux is not included as required for survivability.
Ihtsed on the above discussion, the equipment and instrumentation list contained in Tablepwill be used in assessing the survivability for the 10CFR50.34(O event.
6/9t,1-29 19E.2.1.2.3.2.2 Requirements for Severe Accidents As discussed above a review of the PRA and severe accident analysis was done to determine the set of equipment required for ' accident mitigation. Both in-vessel and ex-vessel scenarios were considered. The survivability of all equipment which is used in the development of the containment event trees or in the severe accident analysis is addressed. It is noted for clarification that, although the RCIC system is discussed in the developement of the severe accident analysis, it is only used before core damage occurs. This ensures the proper initial conditions for the accident.
Therefore, the survivability of RCIC is not addressed.
In-vessel recovery sequences occur when ECCS fails. Since the mean time to recoveg for ECCS is approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> and core cooling must be recovered within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the initiation of the accident, the in-vessel recoveg sequences are dominated by cases in which the reactor is blown down and the firewater addition system is used to provide core cooling. In the long term, the,RMR St'
~.
A system must also be recovered to provide containment heat removal. Therefore, only these systems are considered for equipment sunivability.
The instrumentation considered for equipment survivability for severe accidents was derived form the list required for the 10 CFR 50.34(0 instrumentation list described in Subsection 19E.2.1.2.3.2.1. This ensures that allinstrumentation considered in the severe accident analysis is accounted for, since all operator actions for severe accidents have been included in the Emergency Procedure Guidelines. The list contains more instruments than are actually considered in the severe accident analysis. For example, no actions in the PRA or severe accident analysis are based on -
the Exhaust Vent radiation level. As in the case of 10CFR50.34(0 instrumentation, the neutron monitoring function is not required to survive the event for either in-vessel or ex-vessel accidents.
For ex-vessel accidents, it is not necessary for the SRVs or the in-vessel instrumentation to sunive past the time of vesesl failure. Thus, although very high temperatures persist in the vessel for the duration of an ex-vessel accident, the.
depressurization function, RPV water level instrumentation and RPV temperature instrumentaiton must only survive approximately one hour after core uncovery.
The exhaust fan radiation monitor is used during normal operation. Upon sensing high radiaiton levels, the normal exhaust path is isolated anf flow is directed through the Standby Gas Treatment System. Isolation also occurs if the. monitor fails.
For the classes of accidents considered here, the containment will be isolated.
Therefore, this instrument uill not be affected by the event. Further, the monitor is not a post accident monitoring device, so its sunivability is not an issue.
Insert E
/962.-M For each required sysicm identified in Table ?@, the components of the system which are located inside the containment are identified in the discussion which follows.
Components which are located outside of containment and are not exposed to containment process iluid, are excluded from the discussion'since neither the 10CFR 50.34(0 event nor a severe accident will cause significant changes outside of the containment itself.
hisert F t
Stainless steel components such as piping, spargers and quenchers will not be threatened by the conditions in the containment. Therefore no further consideration of those components will be given in this discusion.
The valve actuation cabling within the primary containment is composed of concentric-lay coated copper. All of the cabling inside containment is housed in insulation which is a flame retardant cross-linked polyethylene. Additionally', the insulated cable is housed within a thennoplastic chlorinated polyethylene jacket which provides protection from severe radiation environments. Tests performed by ORNL (NUREG/CR-5950) shcw that the inslulation andjacket can withstand tempratures exceeding 1033 K-(1400 F) Finally, eighty percent of the actuation cabling located inside containment is enclosed within metal conduit which further shields the cabling from severe environments. Therefore sunival of the cabling for the environment < considered is not a ennecrn
---_--_-_--_,._.___--.__--._--_-_-_-_--_---_-______-__._._$.-____--_-._A____--..
s Insert II luside of the primary containment the depressurization system consists of the following equipment and instrumentation:
nitrogen supply nitrogen supply line
+
valve actuation cabling piping and quenchers safety relief valves SRV solenoid temperature and position monitoring instrumentation -
For the 10CFR50.34(f) and core melt scenarios with in-vessel recovery, the safety relief valves must survive or fail in the open position for the duration of the event. For the ex-vessel cases, the safe +y-relief valves must survive only until the vessel fails. The ex-vessel vessel temperature, pressure, and radiation profiles fall below those for the in-vessel cases. Hence, this discussion will consider the in-vessel cases to be bounding.
Insert I The nitrogen supply hoe consists of piping, valves and condensation tanks, none of which will be threatened by environment within the containment as discussed above.
Also, the integrity of the va!ve actuation cabling, system piping and quenchers within the containment will not be adverely affected during the accident as discussed above.
The vessel pressure does not pose a problem because it remains within design limits.
Comparison to radiation qualification limits are based on two day integrated dose rates.
The equipment integrated radiation doses are below the equipment qualification integrated dose rates of 2E8 R and 2E9 R for gamma and beta radiation, respectively, as set forth in Table 31-16 of the SSAR.
During the early part of the event, the temperature of the process gas exceeds the SRV
- design limit. This will not pose a problem for several. reasons. First, organic material is located only in the solenoids which are far enough removul from the process fluid that overheating from that source is not a problem. The remaindeiof the valve is steel. Any deformation that might occur will tend to stretch the valves outward because of the internal pressures. This type of deformation will not limit the valves ability to relieve -
pressure. Valve closure is not required. Finally, the SRVs pres.sure relief capacity is substantially oversized for this event so only one valve must remain open to bring the vessel to low pressure. Deformation of all 18 valves.in a manner which would prevent the.
vessel depressurization function is not credible.
The drywell temperature and pressure are below the design basis for the SRVs (Figure 5.1-3) up to the time of vessel failure for the ex-vessel scenario and throughout the duration of the event for the in-vessel and 10CFR50.34(f) scenarios.
w q
Temperature and position monitoring instrumentation are not needed for accident mitigation therefore their survival is not important.
Insert J In modeling the 100% metal-water reaction scenario (10CFR50.35(f)), the RHR removes heat from the containment through drywell and wetwell sprays. This decreases the wetwell airspace pressure enough to avoid COPS activation, eliminating the potential for breaching containment for the duration of this transient. Inside of primary containment -
the RIIR system consists of piping, spargers, valves, and valve actuation cabling. The integrity of the _ system piping, spray spargers and valve actuation cabling within the -
containment will not be adversely affected during the accident as discussed above. The valves for the drywell sprays are located in the upper drywell.. The maximum ambient drywell temperature is 370K (207'F) and is reached at approximately.1800 seconds into the transient. This temperature is below the valve design condition of 444K (340 F)
(Table 6.3-9). Similarly, the valves for the wetwell sprays are located in the wetwell. The maximum wetwell temperature is 440K (330 F), which is above the "alve design condition of 377K (219 F). This exceedence does not pose a problem, however, as the valve does not contain organic material which could degrade upon exposure to a high temperature environment and the metallic valve will not be damaged by this temperature.
For the in-vessel core melt scenarios, the LPCF mode of RHR provides low pressure in-vessel core injection which eventually quenches the molten core. Core injection in the low pressure core flooder mode utilizes piping from the suppression pool, the RHR heat exchanger, and the injection valve. None of these components are located inside the
' i vessel. The heat exchanger is located outside of primary containment. The injection valve sees the ambient drywell conditions which reach a maximum of 400K (260"F) which is below the valve thennal qualification limit of 575K (576 F).
The RHR system is needed to remove decay heat from the containment during an ex -
vessel transient to avoid COPS activation as is the case in three of the six ex-vessel events identified. As with the 10CFR50.34(f) case, the RHR functions in the containment cooling mode which may involve drywell sprays. - As above, the drywell spray mode.
consists of piping,'spargers, valves and valve actuation cabling. The ambient drywell temperature exceeds the qualification limit of 444K (340*F) by approximately 90K. This exceedence does not pose a problem, however, as the valve does not contain organic material which could degrade upon exposure to a high temperature environment and the metallic valve will not be damaged by this temperature. As discussed above, the valve actuation cabling used inside containment is protected by insulation and thermoplastic
. Jackets which will not fail in these accident conditions. The pressure in the suppression pool exceeds the system qualification pressure of 0.31 MPa. However, the piping is
~
nominally capable of withstanding pressures of more than 2.5 times that based on
~
previous work. The supression pool temperature slightly exceeds the sytem qualification temperature of 377 K but this is not a concern as discussed above.
The process fluids that are used in the RHR system come from either the suppression pool er RPV. The suppression pool is limited to a maximum pressure of 0.62 MPa by the -
a w
k COPS system and the suppression pool temperature never rises above 425 K for the ex-vessel accident scenarios. The RHR piping and valves used for suppression pool cooling and sprays are rated to a pressure of 2.8MPa or greater and a temperature of 455 K or greater. Therefore the p. icess fluids will not pose a threat to the RHR loop. Prior to receiving a low pressure signal the RHR loop is isolated from the RPV by piping and valves rated to a pressure of 8.6MPa and a temperature of 575 K. Once the valves isolating the RHR from the RPV are opened piping is that same as for suppression pool cooling and is rated to a pressure of 2.8MPa or greater and a temperature of 455 K or above as before. The reactor pressure does not go up again after the RHR system has been activated so overpressurization will not occur.
Insert K (6) Vacuum Breakers The vacuum breakers may need to open during an accident to rel: eve high differential pressures between the wetwell and drywell. Vacuum breakers are passive in nature and have no instrumentation or control other than position indication which is not essential for operations. The vacuum breakers are located on stub tubes high in the wetwell and will be subject to the temperature loads in the wetwell airspace. The differential pressures between the wetwell and the drywell during the three scenarios do not exceed the design -
basis. The temperature in the wetwell exceeds the equipment qualification limit of 377 K by approximately 60 K. The valves are composed of steel and organic seals. There is no concern that the temperature in this area will exceed the capabilities of the steel. Also, tests at Sandia have shown that organic materials are capable of withstanding temperatures of up to 606 K (630 F). The seats for these valves will be selected to meet the temperatures of the accident environment. Per subsection 6.2.1.1.4.1, the COL :
applicant will be responsible for ensuring that the vacuum breakers are_ shielded from pool swell loads. There is no direct means for debris to reach the vacuum breakers. Therefore, they are not expected to be adversely affected during any of the three accident scenarios considered.
Insert L The containment isolation valves will close very early in the event when all of the parameters are well within the design basis values. The valves will remain closed during the remainder of the event. All of the valves attached to the primary system are rated to a pressure of 8.6MPa. Therefore they will not be threatened by the pressure environment during an accident. The remaining containment isolation valves are rated to pressures above the COPS pressure limit of 0.62MPa so they will not be threatened by the pressure environment.
The air supply to the air actuated valves is automatically closed on the containment isolation signal. Therefore the valve can not activate and reopen even if the~ elastomers in the solenoid fail due to high temperature. The power to the motor operated valves is -
shutofToutside of the containment so self-actuation of the MOVs cannot occur if shorting is caused by the high temperature environment inside the containment. In addition, all of the motor operated valves are self-locking so they will not relax and allow leakage. Metal
~.
~
seats are specified for the check valves so they will not be affected by the high
- temperatures in the containment.
(9) Containment Structure -
Extensive ~ work has been done to demonstrate containment survivability. All three scenarios considered here have a very high probability of containmentintegrity as -
discussed in Appendix 19F.
(10) Containment Penetrations The survivability of mechanical and electrical fixed penetrations as well as operable a
penetrations is discussed in Subsection 19F.3.2.2. The fixed penetrations will maintain their integrity beyond the containment qualification limit of 0.770MPag (97 psig) at 533K (500 F). The radiation loads on the penetrations are below the TID limits.
(11) Recombiners The recombiner system is needed in the long-term (order of days) in an accident to ensure that the oxygen concentration does not reach flammability limits. The recombiners are located outside of the primary containment. Piping is used to remove and return fluid to the primary containment. Therefore the only impact on these sensors comes from the -
sampled process fluid. The supply and return lines are isolated during the early part of the event so the recombiners are not subjected to the primary containment pres' ure and s
temperature until days later, after accident recovery when the environment is not as '
severe. Additionally, the integrated radiation doses will be~well below the design basis values. Therefore the recombiners will not be threatened by these accident scenarios.
Insert M The pressure sensors used to measure both water level and pressure are located outside of containment. The conditions in the vessel and containment are monitored via pressure -
taps. The pressure sensors will not see the higher vessel or primary containment--
temperatures and radiation doses due to the high length to diameter ratio of the piping.
- The. integrated radiation does for the pressure sensors are slight _ly over the equipment qualification limits' for gamma radiation. The limits set for design basis events are.j conservative howev'er and survival of the sensors should not be a problem. The sensors are capable of withstanding very high overpressure events, on the order of 13.8MPa (2000 psi), so there is no possibility of damage froin high contaimnent pressure.
(13) Temperature Instrumentation The GE standard practice is to use thermocouples that are rated to 575K (575 F) and -
H 13.8MPa (2000 psi). This is well above the loads seen during a severe accident for both.
the drywell and wetwell. Therefore operation of the thermocouples should not be' adversely affected. Comparison to radiation qualification limits are based on two day -
integrated dose rates. The equipment integrated radiation doses are below the equipment q
- )
-l
~
qualification integrated dose rates of 2E8 R and 2E9 R for gamma and beta radiation, respectively, as set forth in Table 31-16 of the SSAR.'
(14) Ilydrogen and Oxygen Concentration Sensors These sensors are located outside of the primary containment and monitor containment gas concentrations via taps located within the containment. Therefore the only impact on.
these sensors comes from the conditions of the sampled steam. Because of the long pipe nms to the sensors, temperature will not be a problem. The pressure in the sensed gas.
will be approximately that of the primary containment. The sensors will be selected to_
1 I
survive these pressures. The sensors are subjected to the radiation environment of the -
process fluid, however, the integrated dose will be well below the design basis values.
l.
{
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--__.________L-__s_
99 14C,2-2 9 Table ????' Equipment and Instrumentation list 10CFR50.34(f)
In-Vessel Ex-Vessel Severe
' Severe Accident Accident SAFE SHUTDOWN EO_UIP.
+
+
+
-ADS
+
+
+
ACIWA
+
+
+
CONTAINMENT EOUIP.
containment structure
+
+
+
CIV's - inboard
+
+
+
CIV's - outboard
+
+
+
clect. penetrations'
+
+
+
mech. penetrations
+
+
+
hatches
+
+
+-
sealing mechanisms (welds, bellows,...)
Passive Flooder
+
COPS
+
+
Vacuum Breakers
+
+
+
RIP Vert. restraints
+
+
+.
INSTRUMENTATION RPV water level
+
+
RPV pressure
+
+
+
sup pool water temp
+
+
+
DW/WW H2 conc -
+-
+
-+
DW/WW O2 conc
+
+
+
DW temp
+
+
+
DW pres -
+
+
+-
WW pres
+-
+
'+
DW water level
+
+
+
3
_ WW water level
+
+
+
1
- + Indicates that the equipment / instrumentation is required for the event, 1
Indicates that the equipment /instrumentaiton is not required for the event.
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