ML20069Q575
| ML20069Q575 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 01/11/1991 |
| From: | Stirn R GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC, NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| EEN-9105, MFN-004-91, MFN-4-91, NUDOCS 9101160191 | |
| Download: ML20069Q575 (16) | |
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GE NucIcar Energy t u in n.; san . >
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January 11,1991 l Docket No. STN 50 605 EEN 9105 .
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l Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C, 20555 Attention: Charles L. Miller, Director Utandardization and Non Power Reactor Project Directorate
Subject:
GE Responses to GE/NRC Conference Calls on Reactor Systems, SER Input for ABWR SSAR Chapters 4 5,6,9 and 15
Reference:
GE Responses (Proprietary Information) to GE/NRC Conference Calls on Reactor Systems, SER Input for ABWR SSAR Chapters 4 5,6,9 and 15, MFN No. 005 91, dated January .11,1991 Enclosed are thirty four (34) copies of the GE responses to the discussion items of the subject conference calls made on December 4,1990 and December 6,1990.
Responses to discussion items 1,5, and 6 contain-information that is designated as General Electric Company proprietary information and is being submitted under separate cover (Reference).
It is intended that GE will amend the SSAR, where appropriate, with these responses in a future amendment.
Sincerely, l R.C.Stirn, Acting Manager Regulatory and Analysis Services M/C 382, (408) 925 6948 cc: F. A. Ross DOE)
D. C.Scalett(i (NRC)
G. Thomas (NRC) ,
D. R. Wilkins (GE)
J. F. Quirk (GE) 9101160191 910111 ,
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1 RESPONSES TO DISCUSSION ITEMS REQUIRING GE' ACTION QL@RMEElGLN_LTEM-; 1 Markup new ABWR SSAR Appendix.4B to be consistent with GESTAR II Amendment 22. Also, provide a statement indicating that Appendix AB is to be used if analysis.is required.
BESPONEFC{,
Proprietary in f ormat i on providec' :under separat e cover.
j PISQMSSION EEM lt Revi se Sect ion 4. 4 t o include a loose- sart s-- monit oring syst em to be suppl.ied by the applicant referencing'the. ABWR designj. This l revitaun is to include a statement why -it is advantageous .to delsy design of the loose-parts monitor ng system-_unti1 detailed design or the reactor system has bean corulated and equipment vendors selected.
RESPONSgj, See atteched markup.
DJ1G1LSSION _ ITEM S
-Revise ABWR SSAR Subsection 15.2 to address'the-failure mo'de and ,
effects of-slow air system failure or-dirty.' air: in the-instrument- I air system..
RESPONSE 5 Proprietar.y information provided under r.e parat e - co ver.
1 DlSCME.1LO.N_ ITEM 6 Submit detailed information' including ~ drawings-of i the.HCU.
RESPONEF ^
Proprie. information provided under; separate cover.
D I SCy_@E,LQN ITEM'7 ]
Revise Subsection.5.4.7.;i.2 to reflect compliance with RSB BTP 5-1.
4 RC9PONSE 7.
l See attached markup. I a
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DIS _QUSSION ITEM 11 Provide calculations to substantiate the adequacy of the 100'.Dpm borated solution in meeting 10CFR50. 62 ( a) ( 4 )..-. .
REPONSE 11 As approved and documented in'the Safety Evaluation of Topical Report (NEDE-31096-P-A) " Anticipated Transient Without Scram;-
Response to ATWS Rule, 10CFR50.62" (October 1986)', the equivalency requirement can be-demonstrated-if the following relationship is satisified Q x M251 g_C g E >; 1 86 M 13 19.8 where Q = expected SLCS flow rate (gpm)
M =' mass of water in the reactor vessel and recirculation system a hot rated condi-tion (lbs)
C = sodium pentaborate solution concentration _(weight-per cent)
E = B'* isotope enrichment (19.8% for natural boron),. atom percent Values of M251 may vary somewhat depending on the design, (e.g., M251 for BWR/3/4 = 628,300 lbs, M251 for BWR/5 = 614,300 lbs, and M251 for'BWR/6 = 615,100 lbs), cr Q3 86 g M .g 13 .g 19.8 M251 C E For ABWR, l-M = 674,100 lbs L '
C= 1~3 E a 19.8 and using.M251 = 615,100 lbs, and obtains !
Q h 94.2 gpm.
Therefore, the 100 gpm capacity selected for the ABWR satisifies j the requirment.
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, . s DI SCQBSID_bl ITEM'14 3 Review and update'the GE Feburary 21, .-1980 .-re s pon s e s t'o ' t h e Michelson concerns for applicabi'lity to'ABWR.4 -
NESPONEE 14 }
See attached markup. ,
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. 4.4.$ Interfaces 4.4 .1 Power Flow Operating Map i
The speelfic power flow operating map to be used at the plant will be provided by the utility l to the USNRC for information.
. 4 4.4J.2 Thermal Limits The thermal limits for the core loading-at the plant will be provided by the utility to the USNRC for information, 8 +
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lNGERTS Fb F. 0\ S CO 3 5 \od tT5 M A 4. 4. 3 Loose-Parts Monitoring System The applicaat referencing the ABWR design shall provide a I loose-parts monitoring system on the reactor pressure vessel, and-
! i mplE ment a loose-parts detection program which conforms to the
, guidelines of the regulatory position contained in Regulatory l Guide 1.133. The design of the loose-parts monitoring system is I deferred so th6t it may be defined utilizing commerically l available components at the time of construction in a system and it can reflect a pp l i c ati; preference and perhaps experiened may be best satisified. See Subsection 4.4.4.3 for interface requirements.
B 4.4.4.3 L uose-Part s Monit orinD System The applicant referencing the ABWR design will provide a loose-parts monitoring system and' implement a loose-parts detection program (See Subsection 4.4.3).
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directly into the reactor pressure vessel to signal closes the RHR con'tinment isolation i tbe drywell conditions spray when header AC power degraded is not plant valves available from suction. that are provided Subsection for the 5.2.5 provides shutdown cooling }
an explanation either onsite or offsite sources. The RHR of the leak detectier. system and the isolation.
provides the piping and valves which connect the signals; see Subsection 5.2.5.2.1 (12) and Table ,
FPS piping with the RHR loop C pump discharge 5.2 6.
piping. The manual valves in this line permit ;
adding water from the FPS to the RHR system if The RHR pumps are protected against damage the RHR is not operable. The primary means for from a closed discharge valve by means of +
supplying water through this connection is by use automatic minimum flow valves which open on low of the diesel. driven pump in the FPS, A backup mainline flow and close on high mainline flow, to this purnp is provided by a connection on the ;
outside of the reactor building wbich allows 5.4.7.1.3 Design Basis for Pressurt Relief .
hookup of the FPS to a fire truck pump. Capacity ;
The vessel injection mode is intended to The relief valves in tige RHR s prevent core damage during station blackout after on -- ^' "--- beeira l rh e lo* ystem 3 i s bfare sized ,
RCIC has stopped operating, and to provide an in. vessel core melt prevention mechanism during a T1')) thermal reldfpa4 e d .,
severe accident condition. If the AC independent water addition mode is not actuated in time to (2) valve bypass leakage only, prevent core damage, core melting and vessel __ _
.;. lure, thec i ; overs the corium in the lower (3) control valve failure and the subsequent drywell wbtn initiated and adds water to k
uncontrolled flow which results, y containment, thereby slowing the pressure rise.
The drywell spray mode prevents high gas temperatures in the upper drywell and adds additional water to the containment, which .
increases the containmeu thermal mass and slows '
the pressurization rate. Additionally, the drywell spray provides fission product scrubbing l
to reduce fission product release h the event of failure of the drywell head. ,
Operation of the AC independent water addition mode is entirely manual. All of the valves which ruust be opened or closed durir:g fire water .
addition are located within the same ECCS valve room. The connection to add water using a fire truck pyrnp is located outside the reactor building at grade level.
5.4.7,1.2 Design Ilasls for Isolation of RilR y System from Reactor Coolant System ,
The low pressure portions of the RHR system are isolated from full reactor pressure whenever the' primary system pressure is above the RHR system design pressure. (See Subsection 5.4.7.13 for further -details.) In addition, autornatic Isolation occurs for reasons of maintaining water inventory which are unrelated to line pressure ratins. A low water level .5 Amnemni ts SMs.1 l 1
ABM 2mtewa Standard Plant RN C I
- . Redundant interlocks prevent opening valves discussed in Sections 3.5, 3.6, 3.7, a n d to the low pressure suction piping when the Subsection 9.5.1 reactor pressure is above the shutdown range.
These same interlocks initiate valve closure on 5.4.7.2 Systems Design increasing reactor pressure, l l g ) 5.4.7.2.1 System Diagrams l In addition, a high pressure check valve will t
close to prevent reverse flow if the pressure All of the components of the RHR system are i should increase. Relief valves in the discharge shown in the P&lD (Figure 5.410). A piping are sized to account for leakage past the description of the controls and instrumentation
, check valve. is presented in Subsection 7.3.1.1.1 cmcrgency L
core _ cooling systems control and instrumen.
5.4.7.1.4 Design Basis With Respect to General tation.
De.ngn Criterion 5 Figure 5.411 is the RHR procer diagrani and '
The RHR systern for this unit does not share data, All of the sizmg modes c' tb :ystem are equipment or structures with any other nuclear shown in the process data. Th. uterlock block unit. diagram (IBD) for the RHR system is providrJ in S:ction 7.3.
3.4.7.1.5 Design Basis for Reliabl!lty ar4 Operability Interlocks are provided to prevent: (1) drawing vessel water to the suppressinn pol, The design basis for the shutd .wn cooling (2) opening vessel suction valves abov- the mode of the RHR System is that this mode is suction lines or the discharge !ine ceaign controlled by the operator from the control pressure, (3) inadvertent opening of drywell room. The only operations performed outside of spray valves during RHR operation where the l the control room for a normal shutdown are manual injection valve to the reactor is open and when l operation of local flushing water admission drywell pressure is net high enough to require valves, which are the means of providing clean the drywell sriray for pressure reduction, and water to the shutdown portions of the RHR system. (4) pump start when suction valve (s) are not open. A description of the RHR system logic-Three separate shutdown cooling loops are (i.e., interlocks, permissives) is presented in provided; and although the three loops are Table 5.4 3.
required for shutdown uc der 'tormal circumstances, l the reactor coolant can be brought to 1000C in 5.4.7.2.2 Equipment and Component Description less than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> with only two loops in operation. The RHR system is part of the ECCS (1). System Main pumps and therefore is required to be designed with redundancy, piping protection, power separation, The following are system performance require, i etc., as required-of such systems. (See Section ments the main pumps must satisfy. The pump 6.3 for an explanation of the design bases for equipment performance requirements include ECCS Systems.)
-additional margins so that the system perfor.
mance requirements can be achieved. These Shutdown suction and discharge valves are margins are standard GE equipment specifica.
required to be powered from both offsite and tion practice and are included in procure-i standby emer; .cy power for purposes of isolation ment specifications for flow and pressure i
and Antdon following a loss of offsite power. measuring accuracy and for power source frc.
l quency variation.
5.4.7.1.6 Design Basis for Protection from Physical Damage Number of Pumps 3 The design basis for protection from physical Pump type Centrifugal damage, such as internally generated missiles, pipe break, seismic effects, and fires, are Drive unit type Motor Amendment 13 5A 19
I N s s 9.T FO R D\ttusstad twM 7 4
l Overpressure protection is achieved during system operation when
! the system is not isolated from the reactor coolant pressure.
The RHR system is operational and not isolated from the reactor coolant system only when the reactor is depressurized. Two modes i of operation are applicable; the flooder move and the shutdown cooling mode. For the flooder mode, the injection valve opens through interlocks only for reactor pressures less than I approximately 500 psig. For.the shutdown cooling mode, the l suction valves can be opened throuDh intericet.s only for reactor I pressures less than approximately 135 psig, Once the system is ,
operating in these lower pressure modes, events are not expected l that would cause the pressure to increase. If for some unlikely !
event the pressure would increase, the pressure interlocks that allowed the valves tu initially open would cause the valves to '
close on increasinD pressure. The RHR system piping would then be protected from overpressure. The valves close at low pressure, and the rate of pressure increase would-be low. During the time period while the valves are closing at these low pressure conditions, the RHR system desi Dn and marD i ns that satisfy the interfacing system LOCA provide ample operpressure protection.
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. Standard Plant uvc single failure, the reactor core remains covered with Response were eevs<we4 %J vgdabd water until stable conditions are achieved. Further-more, even with more degraded conditions invoking All f the generie February 21,1980 GE re-a stuck open relief valve in addition to the worst sponse ass-applisable404he AE'"PAesign-end-nre transient (loss of feedwater) and worst single failure edguaWn4wmocf :. rc; pense 4e4he-Mieheime (of high pressure core spray), studies show (NEDO 24708, March 31,1980) that the core re- "eeneerns 5 f '" 8 " *" P'm for dthe* d ABWR in rw e t A. Standard 2-t Plant. Th e mains covered and adequate core cooling is available 1A.2.34 Primary Coolant Sources Outside during the whole course of the transient. The con- Containment Structure (111,D 1.l(1)]
clusion is applicable to the ABWR, Since the ABWR has more high pressure make up systems NRC Position (2ilPCFs and 1 RCIC), the core covering is further assured. Applicants shallimplement a program to reduce leakage from systems outside containment that Other discussions of transients with single fail- would or could contain highly radioactive fluids dur-ute is presented in the respouse to NRC Ouestion ing a serious transient or accident to as low as prac.
440.111. tical levels. This program shall include the following:
l 1A.2.33.2 Evaluate Depressurization other (1) Immediate leak reduction than Full ADS (ll.K.3 (45))
(a) Implement all practicalleak *cduction mea-NRC Position sures fo( all systems that could carry radioac-tive fluid outside of containment.
Provide an evaluation of depressurization meth-ods other than by full actuation of the automatic de- (b) Measure actual leakage rates with systems pressurization system, that would reduce the possi. in operation and report them to the NRC.
bility of exceeding vessel integrity limits during rapid cooldown. (Applicable to BWR's only) (2) Continuing Leak Reduction establish and im.
plement a program of preurtive maintenance to Response educe leakage to as low as practicallevels. This This response is presided in Subsection 19A.2.11 1 A.2.33.3 R e s p o n di n g t o M i c h eI s o n l Concerns [lI.K 3 (46)]
NRC Position General Electric should provide a response to the Michelson concerns as they relate to boiling water reactors.
Clarincation General Electric provided a response to the Michelson concerns as they relate to boiling water reactors by letter lated February 21,1980. Licens-ces and applicants shouM assess applicability and ad-equacy of this response to their plants.
Amc ,dment t3 tA.219a
l TABLE 1A.2 1
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l RESPONSES TO QUESTIONS POSED BY HR C. HICHELSON (II.K.3(46)]
OUESTION 1 Pressurizer level is an incorrect measure of primary coolant inventory.
RESPONSE 1 BP s do not have pressurizers. BWRs measure primary coolant inventory directly using differential pressure sensors attached to the reactor vessel. Thus, this concern does not apply to ABWR, OUESTION 2 The isolation of small breaks (e.g., letdown line; PORV) "ot addressed or analyzed.
RESPONSE 2 Automatic. isolation-only occurs for breaks outside the containment. Such breaks are addressed in Section 3.1.1.1.2 of NED0 24708. It was shown that if the high pressure systems are available no operator actions are required. If it is assumed that all high pressure systems fail, the op-erator must manually depressurize to allow the low pressure systems to inj ec t and maintain vessel water level. Analyses in Section'3.5.2.1 of NEDO 24708 show that the operator has sufficient information and time to perform these manual actions. The necessary manual actions have been in-cluded in the operator guidelines for small break accidents.
QUESTION 3, Pressure boundary damage due to loadings from 1) bubble collapse in subcooled liquid and 2) injection of ECC vater in steam filled _ pipes.
RESP 0liSE 3 The BWR has no geometry equivalent to thatl identified in Michelson's report on B&W reactors relative to bubble-collapse (steam bubbling upward through the pressurizer surge line and pressurizer). Thus'the first con-cern in not applicable to ABWR, ECC injection in the ABWR at high pressure is either directly into the reactor vessel through water filled lines (RHR-B+C;HPCF-B+C) or into the feedwater lines (PIlR-A;RCIC) . The feedwater lines are normally filled with relatively cold liquid (420 F or less). ECCS injection in the - ABWR at ' low pressure is either directly into the reactor vessel or into - the feedwater lines. Thus the second concern is not applicable to the ABWR, OUESTION 4 In determining need for team generators to remove-decay heat, consider.
that break flow enthalpy is not core exit enthalpy.
RESPONSE 4 i BWRs do not use steam generators to remove decay heat, so this concern-does not apply to ABWR, !
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1 OUESTION 5 Are sources of auxiliary feedwater- adequate in the event of a-delay in cooldown subsequent to a small LOCA?
R do n t need foodwater to remove heat from the reector following a LOCA, whether the subsequent cooldown is delayed or.not. Therefore,-this concern is not applicable to ABWR. BWRs have a closed- cooling systein in which vessel water flows out the postulated break to=the suppression pool. The_ suppression pool is cooled and water is pumped back to the vessel with ECCS pumps. '
OUESTION 6 Is the recirculation-mode of_ operation of the llPCI pumps at high pressure an established design requirement?
RESPONSE 6 The high pressure iujection systems utilized in_the ABWR are the Reactor Core Isolation Cooling (RCIC) and fligh Pressure Core Flooder (HPCF). .
The RCIC and llPCP systems normally take suction from the condensato.stor-age tank and have an alternate suction source from the suppression pool, A recirculation mode of operation of these systems is established when .
the system suction is from the suppression pool. Following a LOCA when system suction is from the suppression pool, water injected into the re-actor is discharged threigh the break and flows back to the suppression pool forming a closed retirculation loop.
Other recirculation modes include test modes (e.g., suction from and_ dis-charge to the suppression pool) and system operation on low flow bypass with discharge to the suppression pool, All of these modes are established design requirements.
OUESTION 7 Are the HPCI pumps and RilR pumps run simultaneously? _. Do they share common piping?/ suction? If so, is the system properly designed to accom-modate this mode of operation-(i.e., are any NPSil requirements : violated, etc...?)
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! RESPONSE 7 l For ABWR the high pressure injection systems (RCIC/HPCF) do not share any common suction piping with the low pressure Rl!R and they can operate si-multaneously with this low pressure system.
The RCIC and llPCF systems share a-common suction line from the conden-l sate storage tank. The- REIR shutdown cooling ' operating mode does not share any common suction piping with the RCIC or !!PCF systems. It is an established design requirement to size the. suction piping, including l shared piping, such that adequete NPSil is available to RCIC-and llPCF l pumps for all simultaneous operating modes of these systems, l
! Pro operational and/or start up tests are conducted that demonstrate that l
the NPSit requirements are met, j ,
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OUESTION 8 Mechanical effects of slug flow on steam generator tubes needs to be ad-dressed (transitioning from solid natural circulation *a reflux boiling and back to solid natural circulation may cause slug flow in the hot leg pipes,)
RESPONSE 8 ;
BWRs do not have steam generators so this concern does not apply to ABWR.
OUESTION 9 I Is there minimum flow protection for the HPCI pumps during the recirculation mode of operation? ,
t RESPONSE 9 For the ABWR, the RCIC, RHR, and HPCF pumps all contain valves, piping, and automatic logic that bypasses flow to the suppression pool as re-quired to provide minimum flow protection for all design basis operating-modes of the systems.
OUESTION 10 The effect of the accumulators dumping during small break- LOCAa is not taken into account.
RESPONSE 10 BWRs do not use accumulators-to mitigate LOCAs. Therefore, this concern does not apply to ABWR, OUESTION 11 What is the impact of continued running of the RC pumps during a small LOCA?
l RESPONSE 11 .
l The impact of continued running of the recirculation pumps has been ad-I dressed in Sectiora 3.3.2.2, 3.3.2.3, and Section 3.5.2.1.5.1 of NEDO-24708. The conclusions were that continued. running of'the recirculation pumps results in little change in the time available for operator actions and does not significantly change the overall system response.
OUESTION 12 During a small break LOCA in which offsice power is lost,- the~ possibility and impact of pump seal damage and leakage has not been evaluated.
RESPONSE 12 The RCIC, HPCF, and RHR pumps are provided with mechanical scals. These seals are cooled by the pamp primary process water. No external _ cooling from auxillary support systems, such as site service water or room air i coolers, is required for RCIC pump seals. The HPCF and RHR seals are cooled by connections to the three separate divisions of the Reactor Building Cooling Water (RCW) System to protect against' single _ failures.
RHR Divisions A, B and C, and HPCF B and C are connected to their corre-sponding RCW divisions. If offsite power is lost, on site diesel genera-tions maintain the RCW three divisional function. These types of seals have-demonstrated (in nuclear and other_ applications)-their capability to operate for extended period of time at ter.peratures-in-excess of those expected following a LOCA.
a Should seal failure occur it can be detected-by room. sump. high level -l alarms, The RCIC,-'HPCF, and'RHR. individual pumps are arranged, and motor -
operated valves provided, so that a pump'with_a failed seal-can be shut-down and isolated without affecting the proper operation of_thelotherfre- _
dundant pumps / systems.
Considering the low probability of seal failure 'during' a LOCA,- the fact ;'
that a mump.with a failed seal can-'be isolated without affecting other redundant equipment, and the substantial redundanco provided in-the'BWR r
emergency cooling systems, pump l seal failure is not considered a signifi-cant concern.
OUESTION 13 . ;
During transitioning from solid natural circulation to reflux boiling and- ;
back again.-the vessel-level'will be unknown:to the operators, and emer-t gency procedures and operator training may be inadequate. _ This needs-to i be addressed and evaluated. !
RESPONSE 13 There is no similar-transition in the BWR case. . In addition, the1BWR has=
vater leva measurement-within the; vessel and the indication of the water
. level is _ incorporated into the operator _ guidelines. -Consequently this j concern does not apply to ABWR, OUESTION 14 The effect of non-condensable. gas accumulation ~ in the steam generators; i and its possible disruption of decay heat;removalfby, natural circulation .
ne'eds to be addressed.
- RESPONSE 14 The ef feet .'of non condensable , gas accumulation' is addressed in- Section 3.3.1.8.2 of NED0-24708._ For a BWR, vapor is present.in the core'.during~
both normal operation and natural circul'ation conditions.
Non condensables may change the composition of the. vapor . but would .have -
l an insignificant effect on tho' natural or forced circulation itself, since the non-condensables would rise with the steam :to the top of .the vessel after leaving'the steam separators.
- CONCERN 11 _
t Delayed cooldown following a small break LOCA could raid the containment pressure'and activate i.he containment spray system. -
RESPONSE 15-The ABWR containment spray system is manually initiated.1 _'A'1 essential equipment-inside'tha containment is required to be. qualified for the en-vironmental conditions- resulting. fromLthe initiation of the containment-spray system.
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OUESTION 16* ,
This concern relates to the possibility that an operator may be includea j and perhaps even trained to isolate, where possible, a pipe break LOCA l without realizing that it might be an unsafe action leading to-high pres-sure, and short term core bakeout, For example, if a BWR should experi- i ence a LOCA from a pressure boundary failure somewhere between the pump suction and discharge valve for either roactor recirculation; pump, it would be possible for the operator to closa these valves following the
, reactor blowdown to repressurize the reactor coolant system, Before such isolation should be permitted, it is first necessary to show by an appro- l priate analysis that the high pressure ECCS is adequate to reflood the uncovered core without assistance from the low pressure ECCS which can no longer delivery flow because of the repressurization, Otherwise, such isolation action should be c:glicitly forbidden in the emergency operat-ing instructions, RESPONSE 16 The ABWR does not have recirculation lines. However, thera are other systems where it is possible for the operator to close these valves fol-lowing the reactor blowdown to low pressure and thereby isolate the break. An example-of this would be a break in the Reactor Water Cleanup piping between the shutdown suction line valve and the containment bound-ary, In Reference 2, the NRC concluded based on information provided by CE that break isolation is not a problem.
In order for the reactor vessel to repressurize following isolation of a line break, the isolation would have to occur ~before initiation of ADS due to a high drywell pressure in concurrence with low water level'1 con-dition.
Isolation of a line break prior to obtainir. ahighdrywegl pressure signal might occur for very small breaks ( a r e a << 10. 01 f t )
which may require several hundred seconds following tho' break to reach the high drywell pressure set point. In this case'it has boen-shown in Reference 3 that the high pressure systems (RCIC, HPCF and feedwater) are sufficiert to maintain the water level above the top of the core.
If isolation of the break were to occur prior to *eaching level 1 but af ter the hign drywell pressure signal, the vest . would pressurize.to.
the SRV set point following isolation of the main steam lines and then oscillate as the SRVs cycle open and clostd. If no high pressure systems were available, the loss of mass out the SRVs-would cause the level to continue dropping and result in automatic ADS actuation shortly after reaching level 1. This would depressurize the vessel and allow the low pressure systems to begin injecting- This capability was demonstrated in NEDO 24708, in addition, explicitly provide for manual depressurization in the event of low reactor water level with high pressure systems unable to maintain level for any reason.
- Excerpt from Reference 1.
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l In summary, in order to repressurize the vessel following break isola-tion, the isolation would have to occur prior to ADS blowdown. For these cases, high pressure systems would snaintain inventory. If no high pres-sure system was available, the low pressure systems would control the vessel water level following automatic er manual vessel depressurization.
REFERENCES:
- 1. Memo, c. M'chaelson to D, Okrent, "Possible Incorrect Operator Action Such as Pipe Break Isolation," June 4,1979.
- 2. Letter, D. G. Eisenhut to R. L. Gridley, " Potential for Break Isolation and Resulting GE-Recommended BVR/3 ' ECCS Hodifica-tions," June 14, 1978.
- 3. " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," NEDO 24708, August 1979.
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