ML20070U618

Forwards Responses to Discussion Items of 910219 Ge/Nrc Reactor Sys Branch Conference Call.Ge Will Amend Ssar,Where Appropriate W/Response in Future

ML20070U618
Person / Time
Site:	05000605
Issue date:	04/01/1991
From:	Marriott P GENERAL ELECTRIC CO.
To:	Chris Miller NRC
References
EEN-9121, MFN-033-91, MFN-33-91, NUDOCS 9104090196
Download: ML20070U618 (103)

Text

GE Nuclear Energy

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April 1,1991 MFN No.033 91 Docket No. STN 50 605 EEN 9121 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:

Charles L Miller, Director Standardization and Non Power Reactor Project Directorate

Subject:

GE Responses to Discussion items of February 19, 1991, GE/NRC Reactor Systems llranch Conference Call Enclosed are thirty four (34) copies of the GE responses to the subject discussion items.

Responses to Enclosure 1 and 2 discussion items are provided in Attachments I and 2, s uspectivey.

It is intende( that GE will amend the SSAR, where appropriate, with this response in a future amendment.

Sincerely, (M b i

/,

P. W. Marriott, Manager l

Regulatory and Analysis Services M/C 382, (408) 925-6948 l

cc: F. A. Ross (DOE) l D. C. Scaletti (NRC)

G. Thomas (NRC)

D. R. Wilkins (GE) l l

J. F. Quirk (GE) i n

{S4 %2$$e $$8$f'[$5 02'g e

ATTACHMENT 1

(

RESPOtlSE T_O EllCLOSURE 1 ADVAtlCED BOILIllG WATER REACTOR (ABWR)

SSAR SECTIOtt 15 Even though GE reanalyzed most of the transients in Amendment 15, four events; cressure regulator downscale failure, trio of af f reactor intwnal pumos, recireviation flow control f ailure and abnormal startup of an idle reactor internal ovuo at Mgh power are all still analyzed as limiting fault incidents. This is a signifi-cant deviation from the SRP where they are categorized as incidents of moderate frequency events.

The staff requires sufficient justi-fication to support the deviation from the SRP.

GE has submittad 3robability analyses of pressure regulator failure and of reactor internal pumo trips to justify the recategorization of the events. However, the staff doas not have enough actual plant operating experience of the equipment to recategorize the events at this time.

It is the staff Dosition that the above events eust be categorized and analyzed as incidents of moderate frequency until sufficient operating experience is gained to change them. We require that GE reanalyze all the events as incidents of moderate frequency according to the SRP.

All the above events are categorized as limiting fault events in AppendixISA(NSOA). Revise the Appendix to show the events as moderate frequency events and apply the appropriate acceptance criteria.

RESPO!!SE j, Chapter 15 and Appendix 15A (NSOA) hos been revised to reflect the comments except for pressure regulator downscale failure and trip of all reactor internal pumps. For these two events, c b " J" g reclassification is requested. (see o tto ch e d NOSA p a ge s) ( Also sae cAcW papa i s ts 3 and + 4-s)

L.

In Section 15.1.1.3.2 results for loss of FW heating, it is stated that the 3-D core simulator has been used to evaluate the event for the equilibrium cycle.

Identify the 3-0 core simulator used for the analyses. Confire that the staff approved model was used for the analyses.

RESPOtJSE 1 The 3-D core simulator used in the analysis is the PANACEA code, which has been approved by the NRC for this application, 3.

Inadvertent RHR shutdown cooling operation is classified as a limiting fault. The EPRI document NSAC-88 ' Residual Heat Removal Experience Review and Safety Analysis - Boiling Water Reactors,"

cresented 480 BWR events involving RHR systems.

It seems that inadvertent RHR shutdown cooling operations have occurred in operating BWRs. Therefore, the event should be classified as an incidant of moderate frequency with application of proper acceptance criteria.

RESPO?lSE 1 The EPRI docunent NASC-88 presented events in which loss or failure of RHR' shutdown cooling occurred when needed. This type of event is evaluated in Section of 15.2.9 of ABWR SSAR. Section 15.1.6 evaluates a different type of event in which inadvertent cperation of RHR shutdown cooling occurs when not needed. This event is considered unlikely to occur.

L M.

In Section 15.5.I inadvertent HPCF startup is categortred as en infrequent event.

But in the suianary Table 15.0-2 this event is categorized as a moderate frequency event. Correct Section 15.5.1 to indicate the event as a. moderate frequency event with application of proper acceptance criteria.

RESPONSE 1 Section 15. 5.1 has hee HvsI'd(f*4^UcchEd P 9 " '5'5"

b-Revise Table 15.0-2 to include subsections 15.4.6 " Rod Qection tecident" and 15.4.9 ' Rod Drop Accident" results.

RESPotiSE ;

Table 15.0-2 bc=5 beew nw u d A S 40a & d.

O l

l 0

ABM 2146iaaso Rrv c Standard Plant Table 15.0 2 RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS (Cont.)

MuCon No.

M u.

Mu.

Anrigt of Du rsuor.

Mu.

Mu Vesel Steam Surface Valves of Sub Neutron Dome Donoen lans flut %

A Fnq Mrsi ikmsom15 Secuon ripre Dux Prusurt Pnsaun Pnuv-t

(% of in Cats. Blom-Occonds) 2 2

d 1D.

De m u*

1.lil$

(1.1/Qt g g2g gggg g,)

gg,g) gg,, g;;3, g 15412 RWE-Stanup Sif TLYT 15 4.2 RWE at Psatt SEE TErf 1543 Control Rrd SEE TT.XT Mgrauon 1544 Abnormal Stanvp STI TE\\T of One Ructor leiernal Pump 1543 15 4-2 Fut Runout r.

71 1 72.3 10.6 1161 a

0 0

of One Ruetor laternal Pump 1543 15 4-3 Fut Runout 115 0 723 74 7 713 1683 a+

0 0

of AllRuetor laternal Pump

-@E TEX]T 1547 Mieplaced Bundle t

Accident 15 3 Inettue in Ructor Coolant lawntory 1511 1551 laadnnent 102.0 73 1 75 6 71 6 100 0 a

0 0

llPCF Stanup

's r.4. W Re EJ SEg TEA T goM >

ce. des l Aset JEG Tfx T*

wf at..' bet A

Frequency definition is discussed in Subsection 153 4.1 Not limiting (See Subsection 150.4.5)

Transient.s initiatedfrom low power.

a bloderate Frequency b

Infrequent c

Limiting Fault This event should be classified as a limitingfault. tiowever, enteria for moderate frequent incidents are

+

conservatively applied.

1561D Amendment 15

ABWR man Standard Plant RPV C

'i Figure 15.3 2 graphically shows this event control system,instead of an r.nalog system as O

with the minimum specified rotating inertia for used in BWR 2 through BWR 6. The RFCS controls the RIPa. The vessel water level swell due to all ten reactor intern'al pumps (RIPS) at the rapid flow coastdown is expected to reach the same speed. As presented in Subsection high level trip, thereby tripping the main d 1.2.1.1, no credible single failure in the

'l turbine and feed pump. Subsequent events, such cortrol system will result in a minimum demand as initiation of the RCIC system occurring late to.11 RIPS. A voter or actuator failure may in this event, have no significant_effect on the result in an inadvertent runback of one RIP at resultsi The peak clad temperature during (his its maximum drive speed (~40%/sec.), in this event is calculated to be less than 59)C.

C, case, the RFCS will sense the core flow change which is below the applicable limit of 1204 and command the remaining RIPS to increase speeds and thereby automatically mitigate the-15.3.1.4 Barrier Performance transient and maintain the core flow.

15.3.1.4.1 Trip of'Itree Reactor Internal Pumps As presented in Subseetion 15.1.2.1.1, multiple failures in the control system might The results shown in Figure 15.31 indicate cause the RFCS to erroneously issue a minimum that pgnk pressures stay well below the 96.7 demand to all RIPS. Should this occur, all RIPS kg/cm*g limit allowed by the applicable code, could reduce speed simultaneously. Each RIP Therefore, the barrier pressure boundary is not drive has a speed limiter which limits the i

threatened.

maximum speed change rate to 5%/sec. However, the probability of this event occurring is low 15J.1.4.2 Trip of All Reactor Internal Pumps (less tban 7 x 10 5 f allures per reactor as a limiting fault. Aw8N crM h M4l year); and hence, the event should be considered The results shown in Figure 15.3.2 indicate j

that peak pressures stay wel! below the limit

.[d# AstelND s.4 sN g /'e4,

allowed by the applicable code. Therefore, the 15J.2.1.2 Frequency Classification O

barrier pressure boundary is.not threatened.

15 3.2.1.2.1 Fast Runback of One Reactor

!!J.l.5 Radiological Consequences Internal Pump Trip of all 10 internal pumps due to a loss of The failure rate of a voter or an actuator is power supply is considered extremely unlikely to about 0.0088 f ailures per reactor year.

result in perforation of fuel under conditions of Ilowever, it is analyzed as an incident of

. boiling transition. The release of fission moderate frequency, products would be however much less than that assumed in the Loss of Coolant Accident for an _153.2.1.2.2 Fast Runback of All Reactor event of equal probability. Therefore, the internal Pumps l radiological exposures-not in Subsection.15.6.5 cover the consequences of this event.

This event should be classified as a limiting fault event. However, criteria for moderate 15.3.2 Recirculation Flow Control frequent incidents are conservatively applied.

Failure DecreasingFlow 153.2.2 Sequence of Events and Systems

-15.3.2.1 Identification of Causes and Operation

' Frequency Classification 15J.2.2.1 Sequence of Etents 15 3.2.1.1 Identification of Causes 15 3.2.2.1.1 Fast Runback of One Reactor The recirculation flow control system (RFCS)

Internal Pump uses a triplicated, f ault tolerant digital Table 15.3 3 lists the sequence of events for Figure 15J.3.

Amendment 13 1533

_ -. - ~ ~

ABWR 2micorn Standard Plant uvA the overcurrent protection logic of the controls. No protection systems action is

(

electrical bus which supplies the power to the anticipated. No ESF action occurs as a result of idle RIP, This electric bus is tripped by the the event.

x protection logic. Consequently, the other RIPS powered by this electrical bus are also tripped.

15.4.4.3 Core and System Perfonnance Therefore, an abnormal restart of the idle RIP becomes a trip of one or two RIPS, which is An abnormal restart of a idle RIP becom:s a presented in Subsection 15.3.1.

trip of one or two RIPS event which is presented in Subsection 15.3.1.

15.4.4.1.1.1 Normal Restart of Reactor Internal Pump 15.4.4.4 Barrier Perfor-nce This transient is categorized as an incident No evaluation of barrier performance is of moderate frequency.

required for this event because no significant pressure increases are incurred during this 15.4.4.1.1.2 Abnormal Startup of idle Reactor transient (see Subsection 15.3.1).

Internal Pump at High Power tsde.JJ k sw i M j 15.4.4.5 Radiological Consequences This transient (-y-gas a limiting fault. Ho we,,g g,f g.g4%AA An evaluation of the radiological fra t in s.'.I4,J. w a W 4. d n 4 consequences is not required for this event, 15.4. 2 Sequence of Events and System %.// ad, because no radioact v: materialis released from Operation the fuel.

15.4.42.1 Seqrence of Escots 15A.5 Recirculation Flow Control Failure With Increasing Flow Table 15.4 3 lists the sequence of events for i

an abnormal startup of an idle RIP.

15.4.5.1 Identincation of Causes and Frequency Classification 15.4.4 2.1.1 Operator Actions l

15.4.5.1.1 Identification of Causes The normal sequence of operator actions expected in starting the idle loop is as The ABWR recirculation flow control system follows. The operator should:

(RFCS) uses a triplicated, fault. tolerant digital control system. The RFCS controls all ten (1) adjust rod pattern, as necessary, for new reactor internal pumps (RIPS) at the same speed.

power level following idle RIP start; As presented in Subsection 15.1.2.1.1, n o L

credible single failure in the control system (2) reduce the speed of the running RIPS to results in a maximum demand to all RIPS. A voter their minimum speeds; _

or actuator failure may result in an inadvertent -

runout of one RIP at its maximum drive speed (3) start the idle loop pump and adjust speed to (-40%/sec). In this case, the RFCS senses the match the running RIPS (monitor reactor core flow change and commands the remaining RIPS l

power); and to decrease spaed and thereby automatically mitigate the transient and maintains the core

.(4) readjust power, as necessary, to satisfy flow.

plant requirements per standard procedure.

As presented in Subsection 15.1.2.1.1, 15.4.4 2.2 Systems Operation tuultiple failures in the control system might l

cause the RFCS to erroneously issue a maximum This event assumes and takes credit for normal demand to all RIPS. Should this occur, all RIPS functioning of plant instrumentation and could increase speed simultaneously. Each RFCS t

' i 1545

~

4 W

I MNN 2M610MB Standard Plant prv n 15.5 INCREASE IN REACTOR COOIANT INVENTORY 15.5.13.1 laput Parameter and laitial Conditloos 1 15.5.1 Inadverten! HC' Startup i

Tbc water temperature of tbc HPCF system is l 15.5.1.1 Identincation of Causes and Frequency assumed to be 400F with an enthalpy of 11 ClasslGention Btu /lb.

15.5.1.1.1 Identincation of Causes Inadvertent startup of the HPCF system is l chosen to be analyzed, because it provides the l

Manual startup of the HPCF system is greatest aux.iliary source of cold water into the postulated for this analysis (i.e., operator vessel.

error).

15.5.1 3.2 Results 15 3.1.1.2 Frequency Classifiestion Figure 15.51 shows tbc simulated transient This transient disturbance is categorized as event for the manual flow control snode, it an k'.g.:.... lum *4 of %d begins with the introduction of cold water into 49 tbc upper core plenum. Within 1 sec, the full 15.5.1 Sequence of Events and System HPCF flow is established at approximately 3.2% l Operstlon of rated feedwater flow rate. This flow is nearly 138% of the HPCF flow at rated pressure. l f

15.5.1.2.1 Sequence of Esents No delays are considered because they are not

\\

relevant to the analysis.

l Table 15.51 lists the sequence of events for Figure 15.51.

Addition of cooler water to the upper plenum

)

causes a reduction in steam flow, which results 15.5.1.2.1.1 Identification of 0perator Actions in some depressurization as the pressure

~

regulator responds to the event. Tbc flux lesel Relatively small changes are be experienced settles out slightly below operating level.

in plant conditions. Tbc operator should, after Pressure and thermal variations are relatively i-I hearing the alarm that the HPCF has commenced small and no significant consequences are t

l operation, check reactor water level and drywell experienced. MCPR remains above the safety pressure. If conditiom m 9ermal, the operator limit and, therefore, fuel therrnal margins are shuts down the syr., m.

maintained. Therefore, this event does not base to be reanalyzed for specific core configura.

15.5.1.2.2 Sptem Operation tions.

To properly simulate the expected sequence of 15.5.133 Consideration of Uncertainties l

events, the analysis of this event assumes normal l

functioning of plant instrumentation and important analytical factors, including l

controls specifically, the pressure regulation reactivity coefficient and feedwater temperature and the vessel level control which respond change, are assumed to be at the worst directly to this event.

conditions so that any deviations in the actual plant parameters will produce a less severe Required-operation of engineered safeguards transient.

other than what is described is not expected for this transient event.

15.5.1.4 Barrier Performance The system is assumed to be in the manual flow Figure 15.51 shows a slight pressure control mode of operation.

reduction from initial conditions; therefore, no further evaluation is required as RCPB pressure 15.5.1J Core and System Performance margins are maintained.

uu

1 i

i I

HSOA CHANGED PAGES (Response 1)

O O

ABWR u-n Standard Plant nrv1 APPENDIX 15A PLANT NUCLEAR SAFE'lY OPERATIONAL ANAINSIS (NSOA) 15A.1 OBJECTIVES or Single Operator Error (SOE) criteria. Each The objectives of the Nuclear Saf ety protective sequence entry is evaiated relative Operational Analysis (NSOA) are cited below.

to SACF or SOE criteria. Safety classification aspects and interrelationships between systems 15A.I.1 Essential Protective Sequences are also considered.

Identify and demonstrate that essential 15A.I.4 NSOA Criteria Relatise to Plant protection sequences needed to accommodate the Safety Analysis plant normal operations, moderate frequency incidents (anticipated operational Transients),

Identify the sy stems, equipme nt, or infrequent incidents (abnormal operational components' operational conditions and transients), and limiting f aults (design basis requirements essential to satisfv the nuclear accidents) are available and adequate. In safety operational criteria utilized in thc addition, each event considered in the plant Chapter 15 plant events.

safety analysis (Chapter 15) is further examined and analyzed. Specific essential protective 15A.l.f Technical Specification Operational acquences are identified. The appropriate liasis sequence is discussed for all BWR operating modes.

Will e st ablish limiting oper ating conditions, testing and surveillance bases

!5A.I.2 Design Basis Adequacy relatise to plant technical specification.

Identify and demonstrate that the safety design basis of the various structures, systems or components, needed to satisfy the plant essential protection sequences are appropriate, available and adequate. Each protective sequence identifies the specific structures, systems or components performing safety or power generation functions. The interrelationships between primary systems and secondary (or auxiliary equipoent) in providing these functions are shown. The individual design bases (identified throughout the SSAR for each structure, system, or cort aonent) are brought together by the analysis in this scetion. In addition to the individual equipment design basis analysis, the plant wide design bases are examined and presented here.

15A.I.3 System Level /Qualitatise. Type FMEA identify a system level / qualitative type Failure blodes and Effects Analysis (FMEA) of essential protective sequences to show compliance with the Single Actisc Component Failure (SACF)

IM l l

.me ndme nt

ABWR ursix^n Standard Plant nry c Table 15A.21 UNACCEPTABLE CONSEQUENCES CRITERIA PLANT EVENT CATEGORY: NORMAL OPERATION Unnecentable Conscauences 11 Release of radioanive material to the environs that exceed the limits of either 10CFR20 or 10CFR50.

l 12 Fuel failure to such an extent that were the freed fission products released to the emirons via the normal discharge paths for radioactive material, the limits of 10CFR20 would be exceeded.

13 Nuclear system stress in excess of that allowed for planned operation by applicable industry codes.

14 Existence of a plant condition not considered by plant safety analyses.

Table 15A.2 2 UNACCEPTABLE CONSEQUENCES CRITERIA PLANT EVENT CATEGORY: MODERATE FREQUENCY INCIDENTS l-(ANTICIPATED OPERATIONAL TRANSIENTS)

Unaccentable Consecuentes 21 Release of radioactive rnaterial to the emirons that exceed the limits of 10CFR20.

22 Reactor operation induced fel cladding failure, l

l 23 Nuclear system stress exceeding that allowed for transients by applicable industry codes.

l' 2-4 Containment stresses exceeding that allowed for transients by applicable industry codes.

Table 15A.2 3 UNACCEPTABLE' CONSEQUENCES CRITERIA PLANT EVENT CATEGORY: INFREQUENT INCIDENTS (ABNORMAL OPERATIONAL TRANSIENTS)

Unaccentable Constguences 3 Radioactive material release exceeding of a small fraction of 10GR100.

l 3-2 Fuel damage that would preclude resumption of normal operation after a normal restart.

3-3 Generation of a condition that results in consequentialleu of fur.ction of the reac:or coolant system.

3-4 Generation of a condition that results in a consequentialloss of function of a necessary l

containment barrier.

t$A 2-6 Amendment

1 ABWR u4 aman Standard Plant ni'v c i

SECTION 15A.6 CONTENTS (Continued)

Section Iills Pan 15A.63 Moderate Frecuenes incidents ( Anticloated Operational Transients) 15A.6 5 15A.63.1 General 15A.6-5 15A.63.2 Required Safety Actions /Related Unacceptable Consequences 15A.6-5 l

15A.633 Event Definitions and Operational Safety Evaluations 15A.6-6 15A.63.4 Other Event Definitions and Operational Safety Evaluations 15A.6 8.1 15 A.6.4 Infrecuent incidents ( Abnormal Oneratio. mil Transients) 15A.6 9 15A.6.4.1 General l5A.6-9 15A.6.4.2 Required Safety Actions /Related Unacceptable Consequences 15A.6 9 15 A.6.43 Event Definition and Operational Safety Evaluation 15A.6-9 15 A.6.5 1ltnitine Faults (DesMn Basis Accidents) 15 A.6-10 15A.6.5.1 General 15A.6-10 15A.6.5.2 Required Safety Actions / Unacceptable Consequences 15A.6-10 15A.6.53 Event Definition and Operatioril Safety Evaluations 15Ah 11 15 A.6.6 Special Events 15A.616 15A 6.6.1 General 15 A.6-16 15A 6.6.2 Required Safety Act:.on/ Unacceptable Consequences 15A.6 16 15A.6.63 Event Definitions and Operational Safety Evaluation 15A.6-16 15A.6-iii Amendment

ABWR mman Standard Plant nrv c operation in State C are as follows:

15A.6.3.1 General V)

Safetv Actions The safety requirements and protection sequences for moderate frequency incidents Radioactive material telease control (anticipated operational transients) are Reactor vessel pressure control described in the following subsections for Reactor vessel water level control Events 7 through 22. The protection sequence Nuclear system temperature control block diagrams show the sequence of frontline Nuclear sptern water quality control safety systems. (Figures 15A.6 7 tbrough Nuclear systcm leakage control 15 A.6 22). The auxiliaries for the frontline Core reacthity eontrol safety systems are presented in the auxiliary Containment building pressure and diagrams (Figures 15A.61 and 15A.6-2) and the temperature control commonality of auxiliary diagrams (Figures Spent fuel shielding, cooling and 15A.6 58 tbrough 15A.6 63).

reactivity control 15A.6.3.2 Required Safety Actions /Related State D Unacceptable Consequences In State D, the reactor vessel head is on, and The following list presents the safety the reactor is not shutdown. Applicable planned actions for anticipated operational transients operations are achieving criticality, heatup, to mitigate or prevent the unacceptable safety power operation and achieving shutdown (Events 2, consequences. Refer to Table 15A.2 2 for the 3,4, and 5, respectively).

uneceptable consequences criteria.

Figure 15A.6-6 presents the necessary safety actions for planned operations, corresponding Related plant systems and events for which the safety Unaccentable O

actions are necessary. The required safety Safety Consecuences Reason Action

(/

actions for planned operations in State D are as Action Criteria Requirni follows:

Safety Actions Scram 22 To prevent fuel damage and/or 23 and to limit RPV system Radioactive material release control trip of pressure rise.

l Core cooling flow rate :ontrol 4 RIPS Core power level control Core neutron flux distributie control Pressure 23 To prevent excessive Reactor vessel water level control relief RPV pressure rise Rcactor vessel pressure control Nuclear system temperature control Core and 21,22 To prevent fuel and Nuclear system water quality control Contain.

2-4 containment damage in Nuclear system leakage control ment the event that normal Core reactivity control cooling cooling is interrupted.

Rod worth control Containment and reactor buildingpretsure Reactor 22 To prevent fuel damage and temperature control vessel by reducing the outflow Stored fuel shielding, cooling and isolation of steara and water from reactivity control the reactor vessel, thereby limiting the 15A.6.3 Moderate Frequency Incidents decrease in reactor (Anticipated Operational Transients) vessel water level.

e 15A t,5 Amendment

ABWR man Standard Plant RIN. C farnt 12.lbisolation of One or All Main restore and maintain water level For long term

(

Sp a mlin e s shutdown and extended core coolings, containment / suppression pool cooling systems are isolation of the main steamlines can result in manually or automatically initiated.

a transient for which some degree of protection is required only in operating States C and D. In Event t5-Loss of a Feedwrer lleater operating Stater A and B, the main steamlines are continuously isolated.

Loss of a fsedwater heater must be considered with regard to the nuclear safety operational isolation of all main steamlines is most criteria only in operating State D because severe and rapid in operating State D during significat.. feedwater heating does not ocsur in power operation.

any other operating stage.

Figure 15A.612 shows how scram is accomp-A loss of more the 30 F of feedwater lished by main steamline isolation through the heating causes an 2iarm to be initiated by actions of the reactor protection and the CRD feedwater control system (FWCS). Therefore, the systems. The nuclear system pressure relief most severe case is a loss of 30"F of system provides pressure relief. Pressure relief feedwater heating, just below the alarm l

combined with loss of feedwater flow causes initiation. This 30 F lo.s in feedwater reactor vessel water level to fall and the RCIC heating results in a minimal 4% power increar system supplies water to maintain water level and ar.d no scram is expected. The operator can to protect the core until ncrmal steam flow (or control inis minimal increase in power. The othei planned operation) is established.

protection sequence for this event is shown on Figure 15A.615.

Isolation of one main steamline causes a significant transient only in State D during high Event 16 Feedwater Controller Failure Runout power operation. Scram, if it occurs, is the n( One Feedwater Pumn Q

only unique action required to avoid fuel damage nd nuclear system overpressure. Because the A feedwater controller failure, causing feedwater system and main condenser remain in runout of one feedpump is possible in all operation following the event, no unique operating states, in operating States A and B, requirement arises for core cooling.

no safety actions are required, because the vessel head is removed and the moderator temper-As shown in Figure 15A.613, the scram safety ature is low. In operating State D, feedwater action is accomplished though the combined control system (FWCS) reduces flow from the actions of the neutron monitoring, reactor other feedpump to maintain constant feed flow.

protection and CRD systems.

Steady state operation may continue as no scram or turbine trip is expected as shown on Figure Event 101 ess of All Feedwater Flow 15 A.6 16.

A loss of feedwater flow results in a net Event 17. Pressure Reculator Failure-One Bvnats decrease in the coolant inventory available for Valve Failed Onen core cooling. A loss of feedwater flow can occur in States C and D. Appropriate responses to this A premre regulator failure in the open transient include a reactor scram on low water direction, causing the opening of one turbine level aud restora, ion of reactor water level by control or bypass valve applies only in the RCIC system.

operating States C and D as in other states the pretsure regulator is not in operation. An As shown in Figurc 15A.6-14, the reactor opening of a bypass valve is more severe than protection and CRD systems effect a scram on low opening of a control valve. In either case, the water level. The RCIC system maintains adequate pressure regulator slightiv closes the remaining water level for initial core cooling and to V

15A M Amendment I

l

ABWR mcima Standard Plant REV. C control valves to maintain set prenure. Steady SCRAM) and a trip of four RIPS and also l

state operation may continue as shown in Figure initiates isolation, pressure relief valve and 15 A.6 17.

RCIC actuation. Below 40% power (State D) scram is initiated by a high neutron flux or high Event 1% Pressure Reculator Failure-One Control vessel pressure signal. Figure 15A.6 20 shows Valve Failed the protection sequences. Decay heat necessitates extendeo core and suppressic n pool A pressure regulator f ailure in the closed cooling. When the RPV depressuri cs direction (or downscale), causing the closing of sufficiently, the operation of RilRS shutdown a turbine control valve, applies only in cooling is achieved, operating States C and D because in other states ti pressure regulator is not in operation.

Event 21 Generator Load Reiection. Bvnass On The pressure regulator slightly opens the A main generator load rejection with bypas, remaining control valves or bypass valves to system operation can occur only in operating maintain,et pressure. This action may not be State D (during heatup or power operation),

fast enough to mitigate the event. A high Fast closure of the main turbine control valves neutron flux scram due to the increasing pressure is initiated whenever an electrical grid is expected for initial rated power operation.

disturbance occurs, which results in significant The protection sequence is shown in Figure less of electrical load on the generator. Tne 15 A.6-18.

turbine control valves are required to close as rapidly as possible to prevent excessive Event to Main Turbine Trins (With Bvnass System oserspeed of the main turbine generator rotor.

Oneration Closure of the turbine contro; valses causes a sudden redection in steam flow, which elts in A main t arbine trip can occur only in a increase in system pressure. Above 40% power, operating State D (during heatup or power scram occurs as a result of fast control valve l

operation). A turbine trip during heatup is not closure, as will a trip of four RIPS. A as severe as a trip at full power because the generator load rejection during heatup (<40%) is initial power level is less than 40G, thus not severe because the turbine bypass system can minimizing the effects of the transient and accommodate the decoupling of the reactor and enabling return to planned operations via the the turbine generator unit, thus minimizing the bypass system operation. For a turbine trip effects of the transient and enabling return to above 40% power, a scram occurs via turbine stop planned operations. Figure 15A.6 21 presents l

valve closure as will a trip of four RIPS. the protection sequences required for a Subsequent relief valve actuation occurs. Figure generator lo.d rejection Main generator load 15A.619 presents the protection sequences rejection event and main turbine trip are required for main turbine trips, Main turbine similar events having the same required safety trip and load rejection events are similar

actions, anticipated operational transients having the same required safety actims.

Event 22-Loss of Unit Auxiliary Transformer Event 20- Loss of Main Condenser Vacuum The loss of unit auxiliary transformer causes a generator trip, a scram, a trip of four RIPS, l

A loss of vacuum in the mein turbine condenser a loss of feedwater flow and a loss of condenser can occur any time steam pressure is available

vacuum, and the condenser is in use; it is applicable to operating States C and D.

However, scram Figure 15A.6 22 shows the protection sequence protection in State C is not needed, because the for this event including a scram, a trip of four l

reactor is not coupled to the turbine system.

RIPS, a vessel isolation, pressure relief, and core and containment cooling. This event is For State D above 40% power, loss of condenser applicable only in States C and D, because vacuum initiates a turbine trip with its normal AC power in states A and B is supplied attendant stop valve closures (which leads to from the grid.

15A 6 M Amendment

tMN mm Ali Sin; 3ard Pltml km Event 23. Inndvertent ilPCf Pumo Start vahe closure as will a trip of four RIPS.

l (Coolant / Moderator Temocrature Decrease)

Subsequent relief vahe actuation occurs.

Figure 15 A.6 26 presents the protection An inadscrtent pump start (temperature sequences required for main turbine trip with a decrease) is defined as an unintentional start of failure of one bypass vahe. The response of any nuclear System pump that adds sufficient cold the plant to a turbine trip or a generator load water to the reactor coolant insentory to cause a rejection with a f ailure of one bypass sabe is measurable decrease in moderator temperature.

similar to that with a full bypass operation, 1his esent is considered in all operating states protection sequences for these cases are the because it can potentially occur under any

same, operating condition. Since the llPCF pump operates over nearly the entire range of the Event 2LGenerator 1 cad Reiection with Failuti operating states and delbers the greatest done ftvoass Whc amount of cold water to the sessci, the following analysis will describe its inadscrtent operation A main generator load rejection with f ailure rather than other NSSS puty (e g., RCICS. RilRS).

of one bypass salve can occur only in operating State D (during heatup or power operation).

While all the safet) criteria apply, no unique Fast closure of the main turbine control sahes safety actions are required to control the is initiated w henese r an electrical grid effects of such a pump start. In operating disturbance occurs, which results in significant states A and C, the safety criteria are met loss of electrical load on the generator. The through the basic design of the plant ',rstems, turbine control sahes are required to close as and no safety action is specif.:!. in States B r apidly as possible to pres ent excessis e and D, where the reactor ;s not shutdown, the oserspeed of the main turbine generator rotor.

pressure end temperate c will decrease. The Closure of the turbine control sahes causes a onera'or of the plant r ormal control systern can sudden reduction in steam flow, which results in control any power chant s in the normal manner of an increase in system pressure. Abose4W1 power control, power, scram occurs as a result of fast control Figure 15A.6 23 illustrates the protection sequence for the subject esent.

Prolonged shutdown of the turbine generator unit necessitates extended core and containment 15A.U.4 0ther Esent Definitions and cooling. Figure 15 A.6 27 pr e se nt s t h e Operational Safet) 1:5aluations protection sequences requited for a main generator load rejection. Main generator load The following esents should be classified as rejection with a failure of one bypass vahe is cither infrequent or limiting faults. Ilowe s e r, similar to a load rejcetion whh a full bypass criteria for mode' ate frequency incidents are operation. Therefore, the required safety consers athcly applied.

actions for both are the same sequence.

Esent 2LMain Turbine Trios with Failure of One he n t W A bn o r m a l S t ant.e_.d O n e R e a c t o r thpau Vahr Internal Pumr (RlP)

A main turbine trip can occur only in The abnormal startup of a reactor internal operating State D (during heatup or power pump (RIP) can occur in any state and is most operation). A turbine trip during heatup is not severe and rapid for those operating states in as sesere t.s a trip at full power tucau:.e the which the reactor may be critical (States B and initial power level is les than 40G, thus D).

minimiting the effects of the transient and enablin return to planned operations sia the Occurrence of this event is prevented by a bypass system operation. For a turbine trip recirculation flow control system (RFCS) abose 40% power, a scram occurs via turbine stop interlock that presents a pump start unless all remaining pumps are at their minimum speeds.

For this case of multiple failures and operator t$A 641 Amendment

ABM

wimAii Signdard Plapt

_1wv. c errors, the large flow reversal and associated system. Initial restoration of the core water starting pump inverter oscrcurrent activates a leselis by the RCIC or ilPCP systems.

protectise logic thai trips the two or three RIPS on the bus, in that case, the event is covered bent 45-Prenure Rmtlater I ailutt 12ptaitul by Event 11. Figure 15A.6 38 shows the All Turbine Contol and ihnass Yahes protectise sequence for this esent.

A pressure regulator f ailure in the open Esent 3hRecirculation Flow Control Failure direction, causing the opening of all turbine (Intreasing Flowb Runout of All RIPS control and bypass vahes, applies only in operating States C and D because in other states A recirculation flow control failure causing the pressure regulator is not in operation. A runout of all RIPS is,ipplicable in States C an61 pressure regulator f ailure is most sescre and D. In State D, the resulting increase in power rapid in operating State D at low power.

level is limited by a reactor scram. As shown in Figure 15A.6 39, the scram safety action is i u sarious protection sequences giving the accomplished through the combined actions of the safety actions are shown in Figure 15A.6 45.

neutron monitoring, reactor protection and FMCRD Depending on plant conditions existing prior to the esent, scram is initiated either on main

systems, steamline holation, main turbine trip or reactor Event 40 Recirculat<on Ilow Control Failure vessel low water lesel. The sequence resulting (Decreasine Flowb Runback of All RIPS in reactor u ssel isolation also depends on initial conditions. With the mode switch in RUN, This recirculation flow control malfunction isolation is initiated when main ste amline causes a decrease in core coolant flow. This pressure dectenes to 750 psig. After isolation event h not applicable to States A and 11 because is completed, decay heat r uses teuctor sessel the reactor sessel head is o f and the reactor pressure to increase until limited by the internal pumps normally would not be in use, operation of the relief vahes. Core cooling Figure 15 A.6 40 shows that no protection following isolation is provided by the RCICS or sequences are required for this esent.

IIPCF. Shortly af ter reactor vessel isolation, normal core cooling is re established sia the bent 44 -Feeiuter Contreller Failure-H eneut d main condenser and feedwater systems or, if Two feedwater DLmE5 prolonged isolation is necessary, ertended core and containment cooling will be manually A feedwater controller failure, causing an actuated.

excess of coolant insentory in the reactor vessel, is possible in all operating states.

Event 48 h11dn Turbine Trio (Without thpan Feedwater controller failures considered are System Operation) those that would gise failures of automatic flow control, manual flow control, or feedwner bypass A main turbine trip without bypass can occer valve control. in operating States A and II, no only in operating State D (during heatup of power safety actions are required, since tne vessel operation:

Figure 15A.6 48 presents ihe head is remosed and the moderator temperature is protection sequences required for main turbine low, in operating State D, any positive trips. Plant operation with bypass system reactivity effects of the reactor caused by operation above or below 4% power, due to bypau cooling of the moderator can be mitigated by a system f ailure, results in the same transient scram. As shown in Figure 15 A.6 44, the effects: a scram, a trip of four RIPS, and l

accomplishment of the scram safety action is subsequent relief sabe actuation. Af ter initial satisfied through the combined actions of the shutdown, extended core and cont.;nment cooling neutron monitoring reactor protection and FMCRD is required as nated previously in Event 19 systems. Due to the increasing water level and the resulting L 8 turbine trip, pressure relief Turbine trips without bypass system operation is required in States C and D and is achiesed results in more severe thermohydraulic impacts on through the operation of the RPV pressure relief 15A (4 2 Amendment

AllWR m,, u,,4 n Standnrd I'Innt say c e

the reactor core than with bypass system (q'~)

operation. The allowable limit or acceptable calculational techniques for this event is less restrictive, because the event is of lower probability of occurrence than the tubine trip with a bypaks operation esent.

[ yent 4% Generator Lead Reimien with I'ailu1I (d.AILibrai Yahes A main generator trip without bypass system operation can occur only in operating State D (durir.g heatup or power operation). A generator trip during heatup without bypass operation results in the same situation as the power operation case. l'igure 15A.6 49 presents the protection sequences required for a generator load rejection with f ailure of all bypass valves. The event is basically the same as described in Event 21 at power levels ahose l

40 % A scram, trip of four RIPS, and relief valve operation immediately results in prolonged shutdown, which follows the same pattern as Eve nt 21.

.The thermohydraulic and therrnodynamic effects on the core, of course, are more severe than m \\

(~d with the bypass cperating. liccause the event is of lower probability than Event 21, tne unacceptable consequences are less limiting, l

l l

!"h v]

IM r4 I l

Amendment l

ABWR MMWAH Standard Plant my c (p) 15A.6A Infrequent incidents (Abnorntal Operational Transients) 15A.6.4.1 General The safety requirements and protection sequences for infrequent incidents (abnormal operational transients) are described in the Esent 24..Innhtutet Openire of A Safetv/Rehrf following paragraphs for Events 23 through 27

,Yahr The protection sequence block diagrams show the sequence of frontline safety systems (Tigure Tbc inadvertent opening of a saf ety/ relief 15A.6 23 through 15A.6 27). The auxiliaries for valve is possible in any operating state. The the frontline safety systems are indicated in the protection sequences are shown in Figure auxiliary diagrams (Tigures 1$A.61 and 15A.6 2) 16 A.6 24. In States A and D, the water !cvel and the commor.ality of auxiliary diagrams cannot be lowered far enough to thicaten fuel (Figures 15A.6 58 through 15A.6-63).

damage; hence, no safety actions are requhed.

15 A.6.4.2 Required Safety Actinns/Related in States C and D, there is a slight decrease Unacceptable Consequences in reactor pressure following the event. The pressure regulator closes the tuain turbine Table 15A.6 6 relates the safety actions for control vahes enough to stabihre pressure at a infrequent incidents to mitigate or prevent the level slightly below the initial solue. There unacceptable safcty consequences cited in Table are no unique safety system requements for 15 A.2 3.

this esent.

15A.6.4.3 Esent Definition and Operational if the esent occurs when the IcedwMer systern

/

Safety Evaluation is not active, a scram is initiated by a low b

water level signal and core coming is accomplished by the RCIC system, wbh are automatically initiated by the nuclear bailer instrumentation system (NBIS). The n'.m matic depressurization system (ADS) or the in nual relief salve system tt main as the btu lup depressurization syster-if needed. Afte r de vessel has depressurized, long. term core cochng

!s accomplished by the RilR. Containmerd end suppression pool cooling are automaticady or manually initiated.

Esent 25.. Control Rod Withdrawal Error Dung Refueline and Startuo Operations Because a control rod withdrawal error resulting in an increase of positise reactait) can occur under any operating condition, it must be considered in all operating states.

RefueUnc No unique safety action is required in operating State A for the withdrawal of nne control rod because the core is more than voc p

control rod subcritical. Withdrawal of more V

15^fd Amendment

ABWR

. nn Standard Plant

,,.g;,,c

(]

than one control rod is precluded by the protec-(/

tion sequence shown in Figure 15A.6 25. During core alterations, the mode switch is normally in the REFUEL position, which allows the refueling

)

equipment to be positioned over the core and also inhibits more than one control rod withdrawal.

htoreover, mechanical design of the control rod assembly prevents physical removal of the control rod blade from the top without tenioving the adjacent fuel assemblies.

Startup During startup, while pulling control rods in States C, the reactor is subcritical by more than one rod. Therefore, no protection sequence is needed for this condition.

During low power operation (States B and D),

the RPS initiates SCRAh! on short period or high neutron flux in addition to a short period rod block as shown on Figure 15A.6 25.

ISA.6.5 Limiting Faults (Design Basis Accidents)

/

(

!! A.6.5.1 General u

The safety requirements and protection sequences for limiting faults (accidents) are described in the following paragraphs for Events 28 through 52. The protection sequen6e block diagrams show the safety actions and the sequence of frontline safety systems used for the accidents (Figures 15 A.6 28 through 15 A.6 52). The auxiliaries for the frontline safety systems are presented in the auxiliary diagrams (Figures 15A.61 and 15A 6 2) and the commonality of auxiliary diagrams (Figures 15 A.6 38 through 15 A.6 63).

15A.6.5.2 Required Safety Actions / Unacceptable Consequences Table 15A.6 7 presents the safety actions for design basis accident to mitigate or prevent the unacceptable consequences cited in Table 15 A.2 4 I

t l

13 % 10 AmenJment 1

I 1

ABWR namAn Standard Plant Mu O

15Ai!.3 then Definition and Operational 15Ah31. Containment and/or reactor building V

Safety Datuations isolation and standby gas treatment operation are automatically initiated by the respective IMnt 2S-Control Rod Fjection Accident building, pool and/or ventilation radiation monitoring systems.

A control rod ejection accident for the fine motion control rod drise design is not a etcdibir inent 32 Loss-of Coolant Accidents (1 OCA) event. Therefore, no protection sequence is Resultinc from Postulated Pirine Breaks Within required.

RPCB Inside Primary Containment pipe breaks inside the primary containment are considered only when the nuclear system is significantly pressurized (States C and D). The result is a release of steam and water into the containment. Consistent with NSOA criteria, the protection requirements consider all slic line breaks, including liquid pipe breaks down to small steam instrument line breakt 'T he most sesere cases are the circumferential break of the high pressure core flooder (liquid) system injection line and the circumferential break of the largest (steam) main steamline.

As shown in Figure 15A.6 32, in operating Event NControl Rod Dron Accident (CRDA)

State C (reactor shut down, but prenurized), a pipe break accident up to the largest pipe break p

A control rod drop accident for the fine can be accommodated within the nuclear safety motion control rod drive design is not a credible operational criteria through the sarious event. Therefore, no protection sequence i. operations of the h151%, emergency core cooling required.

systems (llpCF, ADS, RH R.LPFL, RCIC), leak detection and isolation system, standby pas Esent 30 Control Rod Withdrawal Ftror (Durine treatment system, main control room heating, Power Oneration) cooling and sentilation system, plant protection system (RilR heat exchangers) and the nuclear During power operation in State D, the boiler instrumentation system. For small pipe automated rod block monitoring system (ARBhi) of breaks inside the containment, pressure relief rod control and information system prevents is effected by the nuclear system pressure control rod withdrawals ths.t ould result in relief system which transfers decay beat to the thermal limit violations. Therefore, this event suppression pool. For large breaks, is not a credible event and no protection depressurization takes place through the break sequence is required as shown in Figure 15A.6 3^.

itself. In State D (reactor not shut down, but pressurized), the same equipment is required as Event 31 - Fuel Handlinc Accident in State C but, in addition, the reactor protection system and the PhiCRD system must Because a fuel-handling accident can operate to scram the reactor. The limiting potentially occur any time when fuel assemblies items, on which the operation of the above are being manipulated, either over the reactor equipment is based, are the allowable fuel core or in a spent fuel pool, this accident is cladding temperature nad the containment considered in all operating states. Con-prenure capability. The Fh1CRD housing supports siderations include mechanical fuel damage caused are considered neceuary whenever the system is by drop impact and a subsequent release of pressurized to prevent excessive control rod fission products The protection sequences mosement through the bottom of the reactor p

pertinent to this accident are shown in Figure pressure sessel following the postulated rupture k

ArnerrJment 15A M t

AHWR

mrami, Slandard Plani Rt'v. c Table 15A.6 2 MODERATE FREQUENCY ACCIDENTS (ANTICIPATED OPERATIONAL TRANSIENTS) (Continued)

SAft:TY llWR NSO4 ANAINSIS OPERATING i

I:VI:NT NSOA 1:YF.NT SI:CTION STATE 13 TNT DESCRIlil0N FIGURE NO.

E i l! C D 23 Inadvertent Startup of IIPCF 15A.6 23 15.5.1 X X X X Pump 26' Main Turbine Trip with One 15A.6 26 15.2 3 X

Bypau Valve Failure

.27' Generator lead Rejection 15A.6 27 15.2.2 X

with One Uypau Valve railure 38' Abnormal Startup System 15A.6-38

!$.4.4 X X X X Reactor Internal Putnp 39' Recirculation Flow Control 15A.649 15.4.5 X X Failure All RIPS Runout 40' Recirwlation flow Control 15A.6-40 15 3.2 X X Failute All RIPS RunbacL 44' Feedwater Controller Failure 15A.6 44 15.1.2 X X X X Runout of Two Feedwater Pumps 45' Pressure Regulator Failure.

15A 6 45 15.1 3 X X Opening of all Bypass and Control Valves 48' Main Turbine Trip with 15A.6 43 15.2 3 X

Bypass Failure l~

49' Generator lead Rejection 15A,6-49 15.2.2 X

l-with Bypass Failure r

I This event should be classified as an infrequent event or a limiting fault. Ilowever, criteria for moderate frequent incidents are conservatively applied.

O l$A rel? 1 Amendment l

.. ~,.. _.

ABWR mimo Sandard Phint mvr Event WLiould Radwaste leak or l'aihtts Releases which could occur inside and outside of the containment, not covered by Estnts 28,29, 30, 33, 35 and 36, include small spills and equipment leaks of radioactise materials inside structures housing the subject process equipment. Consersatise values for leakage have been essumed and c>aluated in the plant under routine releases. The offsite dose that results from any small snill which could occur outside containment is negligible in comparison to the dose resulting from the accountable (expected) plant leakages. The protective sequences for this event are presented in figure 15A.6 36.

Event 37 Liogid Radwaste System Storagt, Ink Lliluti An unspecified event causes the complete release of the average radioactivity inventory in the storage tank containing the largest quantities of significant radionuclides from the liquid radwaste system, This is assumed to be one of the concentrator waste tanks in the radweste building. The airborne radioactivity released during the accident passes directly to the environment via the radwaste building v;r.t.

The postulated events that could cause release of the radioactive inventory of the concentrator waste tank include cracks in the vessels and operator error. The possibility of small cracks and consequent low level release rates receives primary consideration in system and component design. The conceritrator waste tank is designed to opera i at atmospheric pressure and 200"F maximum iemperature so the possibility of f ailure is considered small. A liquid radwaste release caused by operator error is also considered a remote possibility. Operating techniques and administrative procedures emphasire detailed system and equipment operating instruction. A positive action interlock system is provided to Event 41 Trio of All Reactor laternal Pumru present inadscrtent opening of a draia valse. GIP 11 Should a release of liquid radioactive wastes occur, floor drain Sump pumps in the floor of the This esent is not applicable in States A and radwaste building will receive a high water level 11 because the reactor vessel head is off and the alarm, activate automatically and remove the RIPS normally would not be in use. The trip spilled liquid to a contained storage tank. The could occur in States C and D. A trip of all protective sequences for this evert are presented RIPS results in a scram and may cause a high in l'igure 15A.6 37.

15A rel)

Amendmcnt

ABWR

- mn Standard Plant nyc (N

water level trip of the rr.ain turbine and the

()

feedpump turbines. Figure 15A.6 41 provides the protection sequence for this event. While a simultaneous trip of all RIPS may cause some fuel j

cladding heatup due to momentary transition boiling. The cladding heatup is insignificant.

its temperature is below 2200 F, the fuel entahlpy is lower than 280 cal /gm and event consequences are acceptable.

Event 42-Loss of Shutdown Coqling Loss of shutdown cooling is applicable in States A, D, C and D, during normal shutdown and cooldown. Because each of the three RilRS loops may be lined up independently in the shutdown cooling mode. A simultaneous loss of all three loops is not a credible event and therefore no protection sequence is required as shown in Figure 15 A.tr42.

Event 43 RIIR Shutdown Coolinc.hndmf Cooling An RilR shutdown cooling malfunction causing a moderator temperature decrease must be considered in all operating states, floweser, this esent is e

not considered in States C and D if RPV system

(-

pressure is too high to permit operation of the shutdown cooling (RilRS) (Figure 15A,6 43). No unique safety acticus are tequired to asoid the unacceptable safety consequences for transients as a result of a reactor coolant temperature decease mduced by misoperation of the shutdown

coolin,

.at exchangers.

In Stat ss B and D, where the reactor is at or near critical, the slow power increase resu! tint.

frora the cooler moderator temperature would be controlled by the operator in the same manner normally used to control power in the source er intermediate power ranges.

Event 46. Pressure Reculator Failure Closure of All Turbine Control and Evoass Valves A pressure regulator failure in the close direction (or downscale), causing the closing of all turbine control and bypass valves, applies only in operating States C and D because in other states the pressure regulator is not in 7

i

(

Amendment 15A fr!4 l

ABWR nunu Sinndard Plant iuv_ c

(')

op:tation. The protection sequence shown on

(_)

Figure 15A 6 46 includes a high neutron flux scram by neutron monitoring, reactor protection and FMCRD systems, a high pressure trip of five RIPS, pressure relief snd core and containment conling.

Event 47 -Deleted Event 50- Misplaced., Fuel Bundle Accident Operation with a fuel assembly in the improper position is shown in Figure 15A=6 50 and can occur in all operating states. No protection sequences are necessary relative to this eve nt.

Calculated results of wor st fuel handling loading error does not cause fuel cladding integrity damage, it requires three n

independent equipment / operator errors to allow

(]

this situation to develop.

Event $1 -Reactor Internal Pumn (RIP) Seizure A RIP scirure event considers the instantaneous stoppage of the pump motor shaft of one RIP. The case insolves operation at design power in State D.

Because a seizure of one out of ten RIPS produces a flow disturbance of less than 10%, consequences of a RIP seizure are mild and no scram occurs. Therefore, normal operation rnay continue and no protection sequence is required as shown in Figure 15 A.6 51.

Ennt $2 Reactar Internal Pumn (RIP) Shaft Break A RIP shaft break event considers the degraded, delayed stoppage of the pump motor shaft of one RIP. The case involves operation at design power in State D. The consequences of i

I this event are bounded by Event 51 RIP l

Seiture. Normal operation may continue and no protection sequence is required as shown in Figure 15 A.6 52.

(n) i

%./

15^ 415 Amendment

ABWR muun Standard Plant nry c O.

Table 15A.6 3 INFREQUENT ACCIDENTS (ABNORMAL OPERATIONAL TRANSIENTS)

SAFLTY BWR NSOA ANALYSIS OPERATING EVENT NSOA EVENT SECTION STATE SIL DTNT DESCRIPT!ON FIGURE NO.

NO..

AB CD 24 Inadvertent Opening of a 15A.6 24 15.1,4 X X Safety Relief Valve 25 Control Rod Withdrawal Error -

15A 6 25 15.4-1 X X X X Startup and Refueling Operations O

Table 15A.6 4 LIMITING FAULTS (DESIGN HASIS ACCIDENTS)

SAFE 1T llWR NSOA ANALYSIS OPERATING DThT NSOA EVENT SECTION STATE QTNT DESCRIITION FIGURE NO. NO.

AB C D 28 Control Rod Ejection Accident 15A.6 28 15.4.8 X X X X 29 Control Rod Drop Accident 15A.6-29 15.4.9 X X X X 30 Control Rod Withdrawal Error 15A.6 30 15.4.2 X

Power Operation 31 Fuel Handling Accident 15A.6 31 15.7.4 X X X X O

15A (,20 Amendment

,v w.-w.,

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ABWR ms-n Standard Plant RIV C Table 15A.6 4 i

i LIMITING FAULTS (DESIGN llASIS ACCIDENTS)

(Continued) l SAFETY BWR NSOA ANALYSIS OPERATING EVENT NSOA EVENT SECTION STATE EVENT DESCRIITION FIGURE NO.

E(L,.,,

6. ]l - C D 32 less of Coolant Accident 1FA.6-32 15.6.5' X X Resulting from Spectrum of Postulated Piping Breaks Within the RCPB Inside Containment-33 Small, Large. Steam and 15A.6 33 15.6,4 X X Liquid Piping Breaks 5 -

Outside Containment 1

34 Gascous Radwaste System 15A.644 15.7.1 X X X X-Leak or failure 35 -

Augmented Offgas Treatment.

15A.6 35 15.7.1 X X X X System railure 36 Liquid Radwaste System 15A.6 36-15.7.2 X =X X X Leak or railure 37 Liquid Radwaste System 15A.6 37 15,7 3 X X X X o

Storage Tank Failure T

41 Trip of All Rifs 15A.6-41 15 3.1

.X X

42 1.oss of RilRS Shutdown Coolin3 15A.6-42 15.2,9 X X.X X 43 RHRS Shutdown Cooling 15A.6-43 15.1.6 X X X X Increased Cooling l

15A (41 Amendment

__2..___.;_-___.__.._-._._,._

,,. -.. _ _.. _ _ _ _. -,. _. _... _,. _ _.... _ _, ~ _ _ _ _ _., _... _, -,, _ - - - - _, _, _ _ _. _... _., _,... _., _,,

ABWR nu=n Slandard Plant urr Table 15A.6-4 LIMITING FAU11TS (DESIGN ilASIS ACCIDENTS)

(Continued)

SAIE1Y ltWR NSO4 ANALYSIS OPERATING EVENT NSOA EVENT SECTION STATE E

EVENT DFSCRIPTION FIGt.'RE NO.

NO.

A. 11 C 1]

46 Pressure Regulator Failure -

15Ah46 15.2.1 X

Closure of all Dypau and Control Valves 47 (Deleted)

O SC Misplaced Fuel bundle Accident 15A h50 15 4.7 X X X X 51 Reactor Internal Pump Seizure 15A h51 15,3.3 X

52 Reactor Internal Pump Shaft 15Ah52 15.3 4 X

11reak O

Amendment 15A 622

ABWR mr.mu Slaridard.PlanL Rtv c

( 's Table 15A.6-6 i

SAFETY ACTIONS FOR INFREQUENT INCIDENTS REIATED UNACCEl'TA1:LE pre 1Y ACTION CONSEOUENCES REASON ACTION REQLIELD Seram and/or Trip 32 To limit gross core wide fuel damage and to l

of four RIPS 33 limit nuclear sptem pressure rise.

Pressure relief 3-3 To prevent excessive nuclear system pressure ris Core, Suppression 32 To limit further fuel and containment damage pool and contain.

3-4 in the esent that normal cooling is interrupted.

ment cooling Reactor vessel 32 To limit further fuel darnage by reducing the isolation outflow of steam and water from the reactor vessel, thereby limiting the decrease in reactor vessel water level.

Restore AC power 32 To limit initial fuel damage by restoring AC power to systems essential to otber safety actions.

Containment 31 To limit radiological e(fects.

isolation l

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Amendment l

l

ABWR zwimAn Standard Plant Riv ^

O EVENT 16 LOSS OF A FEEDWATER HEATER STATE D 0

AT 6 30 r AT > 30 F V

ACCEPTABLE FEEDWATER ALARM IF STE ADY ST ATE CONTROL AT >30.F

=

OPE R ATION SYST E M (MINIM AL POWE R INCRE ASE-TO BE S

p CONTROLLED BY OPE R ATC %

CONTROL ROD DRIVE SYSTEM S

F CORE R E ACTIVITY CONTROL I

ACCEPTABLE STE ADY STATE OPE R ATION 812% A2 Figure 15A,6-15 PROTECTION SEQUENCE FOR A LOSS OF A FEEDWATER HEATER O

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MAIN CONTINUED ON TURBINE FIGURE i

MAIN STEAM TRIP ISA.6-57 PRESSURE

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40% POWER EELOW TR ANSF E R CLOSE ON LOW 40% POWER DECAY HEAT TO CONDENSER SUPPRESSION VACUUM i

NEUTRON HIGH TURB!NE MONITORING

= NEUTRON STOP VALVE 5 85%

j SYSTEM FLUX CLOSURE OPEN OPEN ON STEAM SlF TURBINE =

BYPASS S

F l

TRIP SYSTEM SCRAM SIGNAL ON REACTOR NEUTRON MONITOR S

F PROTECTION

SYSTEM TRIP OR SYSTEM TURBINE STOP VALVE CLOSURE SlF CLOSE ON LOW STEAM CONSENSER

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CONTROL INSERT f

ROD DRIVE

= CONTROL TRIP OF VACUUM SYSTEM l

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Figure 15A.6-20 PROTECTION SEQUENCES FOR LOSS OF MAIN CONDENSER VACUUM i

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REJECTION WITH BYPASS 9

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Figure 15A.6-26 PROTECTION SEQUENCES FOR MAIN TURBINE

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Figure 15A.6-49 PROTECTION SEQUENCES MAIN GENERATOR LOAD REJECTION -WITH BYPASS FAILURE

1. =

ATTACHMENT 2 EESPO!!SE TO EllCLOSURE _2 ADVAtlCED BOILIl1G WATER REACTOR (ABWR)

SSAR SECTION 15 1

1, Provide information (or a reference) for how containment integrity limits were defined.

In Table 15E.2-1 perfonunce requirements, auxtmum oool temoeratureisgivenas97,2'c(207'F).

If ABWR design uses ECCS oumps of the same design as operating plants, there may not be sufficient NPSH for ECCS oumos with a maximum temperature of 207'F. Justify the maximum pool tesoerature of 207'F for ECCS pumos or provide additional justification on NPSH require-ments for these cuans.

RE S Pol.(S E :

The containment functional design is described in Section 6.2.1 of the SSAR. Specifically, the suppression pool design pressure O

is identified as 3.16 kg/cm2 g (45 psig), and design temperature V

as 104*C (219'F). For conservatism, the criterion for ATWS is selected as 97.2'C (207' F).

In the past, lower temperature limit was specified in ?!UREG-0783 because of concerns about unstable condensation causing large containment dynamic loads at high pool temperature. Recent evaluations of test data show that the pool temperature limit specified in NUREG-0783 is not necessary and may be climinated as documented in llEDO-30832, " Elimination of Limit on BWR Suppression Pool Temperature for 3RV Discharge wi n Quenchers,"

(1989). The pool temperature up to saturation is acceptabic.

Therefore, the criterion for ABWR is acceptable.

For ECCS pumps, a maximum temperature of 97.2*C (207'F) is used in the calculations of flPSH available to ECCS pumps. The tosults are documented in Table 6.2-2b for RHR pumps, and in Table 6.2-2c for HPCF pumps. The calculations show that adequate NPSH is available.

For the RCIC, the NPSH cale.ation is based on 77' C (170*F) as shown in Table 5.4-la. Since the preferred water source for the RCIC is the condensate storage tank, which has a volume to supply about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of RCIC flow. After that time period, the suppression pool temperature is expected to be below 17' C during any ATWS event. Therefore, the pool temperature limit for ATWS is

[3 acceptable.

L.}

i 7.

In Table M.3-2, nominal closure time of MSIV is given as 1.2 seconds. Typically MSIV closure time is 3 to 5 seconds.

Justify the 1.2 seconds closure time of the RSIVs.

Also HPCF start time is given as 20 seconds.

But in SSAR Section ~/.3, HPCF starting tist is given as 36 seconds. Which is correct?

RE S POrlS E :

(1)

During the design stage. the ABWR design censidered to use a different type of liSIVs, which have a closure time of abou.r 1.2 seconds. Although that kind of MSIVs was not adopted in the ABWR design at this time, the ATWS analysis, which was performed before this decisien, used 1.2 seconds closure time to bound all types cf itSIVs. The results are considered more severe than the case with 3 seconds closure time.

(2)

Regarding the HPCF start time, the 20 seconds value is the nominal value, while the 36 seconds value is the maximum allowable design limit, In accordance with tiUREG-0460, nominal operating conditiens, not bounding conditions, are used in the ATWS evaluation. A 20 seconds start time for HPCS/HPCI was alsc used in the evaluations of ?!EDE-24222.

3.

Which events in Appendix iSE are calculated using ODYN and which events using REDY?

R ES pot 1S E :

All events in Appendix 15E are calculated with REDY 4,.

Explain any significant differences between the methodology (including code modeling) used for the analysis in Appendix 15E and thct used for the analysis reported in NE02-24222.

RESPCUSE:

The sar..e methodology used in tIEDE-24222 is used for ABWR analysis, except the REDY/ODYt1 codes are replaced with REDYA/ODY!!A codes, which simulate ABWR design features. The mode ~

modifications are documented in Appendix 20A.1 for ODY!1A and Appendix 20A.2 for R E ; ' ' /-

AJ 6.-

sue 1y the o (Tabie isE.31) cury. used for Aar/rncno r.etivity.

RESPONSE

-The D curve, which is also used in NEDE-24222, is tabulated in the following:

Fraction of Core Control Rod Fully Controlled Reactivity 151 0.0 0.0 0.05 0.102 0.211 0.1 0.450 0.2 0.820 0.3 0.4 1.403 2.344 0.5 4.207 0.6 8.410 0.7 0.8

- 17.602 0.9

- 33.811 1.0

- 39.0 (s.

_Why-is the Doppler' reactivity coefficient (Table 15E.31) alsiost twice as large as the coefficient used for the WEDE-24222 analysis? -

RESPONSE

-The-unit of the Doppler coefficient in Table 15E.3-1 is in error.

'It should be-b/*C, instead of t/'F. The' value used is -0.504 $/* C

'(-0.28 4/?F),-which is the same value'used in NEDF.-24222.

O

Oh 1.

What is the mechanism which would insert control rods in 25 seconds when ARI is used? (4 seconds are assumed in REDE-24222.)

R ES Pot 1S E L The 25 seconds shown in Table 15E 3-2 includes 15 seconds delaf before start of ARI control rod insertion and 10 seconds control rod insertion time. This value is accepted by the flRC in the review of NEDE-31096-P-A entitled " Anticipated Transients without Scram: Response to NRC ATWS Rule, 10CFR50.62" (1987). The use of 10 seconds red insertion is expected to be more conservative than the case with 4 seconds.

S.

In view of the fact that the ARWR has several systems that are different from current BWR designs, what is the justification for using results from NEDE-24222 to reach conclusions as to which events are limiting in the ABWR?

M_SJONSE:

l l

Although the actual hardware designs for ABWR are different fror l

operating BWRs, the functions and characteristics of trese hardwares are similar to those in the operating BWRs. Therefere, it is expected that the ABWR ATWS response will be similar to those for operating BWRt. This can be evident from the comparison of transient responses analyzed in Chapter 15 for both AEWR and operating BWRs. The transient responses and relative t ent severity shown in Chapter 15 of ABWR SSAR are similar to those for operating BWRs. Therefore, it is expected the conclusions drawn from flEDE-24222 for operating BWRs are applicable for ABWR.

l 9.

Calculata plant rasoonse to additional ATW5 events including those initiated by a turbine trio and a bypass of high pressure feedwater crehesters. Consider the effect of ARI, FMCRD rod l-insertion, and SLCS for these initiating events.

R_e_sponse:

As discussed in the rasponse to Question 8, the calculations are not necessary, since these events are less severe than those already analyzed.

1

b 10.

What is the maximum amount of time that the operator would have v

before SLCS must be actuated in order to have acceptable consequences for each ATWS7 KESPONSEl Since the probability of an ATWS requiring the initiation of SLCS, which is caused by a failure of both ARI and FMCRD run-in at the same time, is very low, this type of ATWS events has no pool temperature limit as shown in Table 15E.2-1. The limit is the containment pressure. It is estimated that the operator has more than 10 minutes to initiate the SLCS before the containment "ressure is exceeded. (See Response to Question 17.)

Fuathermore, the preferrd water source for the RCIC and HPCFs is the condensate storage tank, which has a volume to supply about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of RCIC flow. Therefore the pool temperature has no significant effect on the system response to an ATWS. Thus, the o.perator has more than 10 minutes to initiate the injection of boron, if needed.

en 11.

Since the ATWS pool temperature limit is not satisfied with U

only one SLCS oumo operation and two puso operation is-required to meet the ATWS Aule, why is this case considered in Appendix 15E?

RESPONSE

Since the probability of an ATWS requiring the initiation of SLCS, which is caused by a failure of both ARI and FMCRD run-in at the same time, is very low, this type of ATWS events has no pool temperature limit as shown in Table 15E.2-1. The limit is the containment pressure. The case with one SLCS pump operation still meets the requirements. This case is to show the capability of the ABWR design.

l l

-pd

l 2,,

What is the estimated frequency of an event with FMCR0 run-in (i.e., failure of hydraulic scres only)? What signals will actuate electHc run in only (e.g., low hydraulic pressure in scrassystem)?

RESPONSE

The estimated frequency of an event with FMCRD run-in (i.e.,

hydraulic scram failure) is less than 10-' / year. There are many signals to initiate FMCRD run-in:

(1) all scram signals, (2) manual scram signal, (3) diverse high pressure signal, which also initiates ARI.

(4) diverso low water level signal, which also initiates ARI, and (5) manual ARI signal, which also initiate FMCRD run-in.

There are no signals which initiates FMCRD run-in only.

l'$.

What is the justification for taking credit for operator actuation of the SLCS in less than 2 min as done for the MSlY

,- 3 (j

closure cases shown in Appendix 15E?

RESPONSE

This is an estimate based on the assumption that the operator follows the instruction of the EPG to initiate the injection of boren. The BWROG considers this is a resonable time period for operator action. This is also the value used in the ATWS analysis as documented in NEDE-24222.

W.

What is the effect on thermal limits in the too of the core where there will be severe power peaking when motor driven scram-is used?

R E S PotiS E :

The results.of peak cladding temperature (PCT) calculations are summarized as follows:

MSIV closure ATWS with ARI:

793'C.

MSIV closure ATWS with FMCRD run-in:

867'C, and MSIV closure ATWS with boron injection: 1025'C.

All cases meet the requirement.

15.

Supply a copy of clots of calculated results that is more readable than found in Appendix 15E (one to a page rather than fourtoapage).

RESPONSE

Copies of plots that are more readable ore anochtel.

Plots with rne to a page are not available.

e

O SSr4R SI Z ED F IGURE S O

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Figure 15ES-8 LOSS OF FEEDWATER FLOW AT' ;

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90-215-34 xS 28

)SS OF FEEDWATER FLOW ATWS (FMCRD RUN-IN/RPT) hB O

1 ID.

In order for our consultants te perfons audit calculations we require the following infonnation:

(a) Flow areas and volumes for the following vessel downconer, lower plenum, core, riser, steam separators and done; steen lines including bynass; and feedwater lines inciudir,g preheatars; (b) For the feedwater spargers, the number of nozzles and the flow areas, and for the feedwater line the water temperature at different points (including before and afterpreheeters);

(c) The characteristics for the RIPS and feedwater pumost (d) Loss coeffic" ts for the inlet orificing and for the fuel bundles and the spacer grid locaticas.

RESPONSE

(a)

Ay2 Flow Area (ft3)

Free Volure (ft) )

Downcomer 125.S 4696 Lcwer Plenum 209.8 3605 Core 172.S 2102 Riser 214.4 1726 Separators

- Inside 117.1 1559

- Outside 227.2 2137 Dome 411 7878 Steam Lines 13.81 4000 i

i Feedwater Lines 4.04 1500 (b)

The total flow area of feedwater spargers is 0.903 ft8 per line. For water temperature at different points, see Figure 10.1-2 of the SSAR.

(c)

For RIP eharacteristics, see Figure 5.4-3 of the SSAR. The inertia time constant of the RIP is 0.7 seconds. After a trip of feedwater pumps, it is assumed that the flow drops to zero in 5 seconds.

(d)

The total core pressure drop at the rated conditions is 24.4 pri, in which about 45 % is single phase pressure drop.

_ _ _. _.. _ _ _ _. = _ _ _

_ _ _ ~

G.

In 15E.6 Conclusion Section, it is concluded that the operator has about 10 minutes to inject the boron into the vessel.

Identify the case for which 10 minutes are available. Also submit pool tamparature verns time curves for all cases analyzed.

P.E S P O!!S E :

)

The results for the case with the operater manual initiation cf the SLCS at 10 minutes are shewn in the attached plots. The results are also sun.marized in the fo110 wing:

Maximum fleutron Flux 626 % (at 1.6 see)

Maximum Vessel Pressure 1335 psig (at 5

.0 see)

Maximum Heat Flux 144 % (at 3.0 see)

Maximum Suppression Temperature 209

• F (at 37 min)

Maximum Containrent Pressure 23.; poig (at 37 min)

\\

It is concluded that all performance requirenents are net fer this case. In fact, some margins are available to accept a longer operator delay tinc.

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