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Summary of ACRS Subcommittee on Advanced PWRs (Wapwr SP/90) 890928 Meeting in Bethesda,Md
ML20006F567
Person / Time
Site: 05000601
Issue date: 10/20/1989
From:
Advisory Committee on Reactor Safeguards
To:
Advisory Committee on Reactor Safeguards
References
ACRS-2668, NUDOCS 9002280173
Download: ML20006F567 (18)
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Text

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i DATE ISSUED:

1_0/20/89

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ACRS SUBCOMMITTEE MEETING ON THE ADVANCED PRESSURIZED WATER REACTORS (~WAPWR SP/90)

SEPTEMBER 28, 1989 BETHESDA, MARYLAND q

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Purpose The purpose of this Subcommittee meeting was to discuss and review the.

Westinghouse APWR (RESAR SP190) design.

.l Attendees ACRS NRC J. Carroll, Chairman

= T. Kenyon,.NRR I. Catton, Member

' L. Donatell.-NRR C. Michelson, Member T. Hsia, NRR D. Ward, Member C. Y.-Li, NRR M. El-Zeftawy, Staff D. Notley, NRR L

- D. Perskino, NRR--

Others

. H. Vandermolen, RES g,

M. Shannon, W

' P. Niyogi, RES T.' Van de Venne, W E.:Chelliah..RES D. Sharp.W J. Monninger, NRR' i

l S. Stahl,;.!

D. Shum, NRk W. Shirley,y P. Trayers, W R.Lutz,y D. Noonan,-Bechtel T. Chu, BNL

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A. Tingle, bt:L T. Pratt, Bt:L L'. Rib, AECL.

['Og PESIGNATED ORIGINAL 9002280173 891020' h

Cortified Dy PDC N

PDR ACRS 2668.

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1 Advanced FkRs Meeting Minutes September'28, 1989 l

i Meeting Hichlights, Agreements, and Requests 1.

Mr. Carroll, Subcommittee Chairman,' stated the' purpose of the Subcommittee meeting and-introduced trie other ACRS members, l

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fir. L. Donate 11, NRC/NRR Project Manager, prcrented the current reviewstatusfortheMAPWRSP/90 design.'Ht

.ited-that t'he staff' is expecting the Cornission to establish a new approved priority.

for the SP/90 preliminary design approval (PDA).

So far, the staff:

hascompletedonedraftSERregardingthePRA-analysis.(front-end l

cnly,_ March 1988)'and two draft SERs (SRP on June 1988 and March 1989). Currently, there are 107 open items-that have to be-re-O solved before the PDA is issued. ~There are additional 53 open

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items that have to be resolved before the final oesign-approval (FDA) is issued.

In addition, there are 99 open items'that have to be resolved before the FDA is issued and/or plant specific-epplica-tion.

E Mr. Donate 11 indicated that there will be two more DSERs-that have i

to be issued. The first one is for the FRA (Back'end portion) and expected to be issued in November 1989.

The seccnd:SER is. regard-ing the USIs/GSIs and severe accidents, and-is expected to-be-issued in April 1900. To finalize the staff's review of the'SP/90 l

design, the staff is requesting three additional ACRS Subcommittee meetings to be hcid in November 1989, May 1990,-and August 1990.'

The PDA decision will be made approximately in October 1990..

3.

Mr. M. Shannon, y/ Licensing Manacer, described the review status'of' the NRC safety evaluation of RESAR-SP/90 particularly.with respect to the severe accident issues that are currently.being discussed within the industry and the NRC. Hr. Shannon-indicated that sin::e j

H_isinterestedinaPDA-atthecurrenttime,theSP/90'designis-not incorporating any additional design _ features that are being developed with the Japanese until a FDA submittal.

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Advanced PWRs Meeting Minutes September 28, 1989 i

Mr. Shannon indicated that }! is not reviewing the SP/90 design.

'I versus the EPRI requirer.1ents docunient at the PDA stage. W has responded to the 107 open items in the DSER, and currently finalizing its submittels of Module 2, which deals with USIs, and

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GSIs.

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Mr. Van de Venne, W/ Engineering Maneger, indicated that the SP/90 design could meet up to 98% of the EPRI requirements document.

Mr.-

Van de Verne outlined the plant design features and, system re-i liabilities. The primary systems consist of the-following:.

with creater internal volume than standard W design.

The increased volume of water above the core provides a longer time before core uncovery (e.g.: in case cf small LOCA and a loss of secondary cooling).

Core eflood tanks - four tanks with. low pressure nitrogen coverage that inject into the RCS vessel assist the HHSI.in reficoding the core following a LOCA.

These tanks eliminate i

the need for active low head SI pumps.

' Integrated Sefeguards System'(ISS) - there are four.high head pumps that infect through their'own RCS vessel connection to provide crergency core cooling for the LOCA events, and.

provide RCS makeup and boration for all non-LOCA' events. Only one of these four pumps is required for small LOCA and: feed--

and-bleed coolino.

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  • EmergencyWaterStorageTank(EWST)-thewatersupply'is located in the basement of the containment. The EWST provides.

a means to reduce the containment cleanup resulting from discharge-from the pressurizer relief tank rupture-disc and.

the hot leg vent path, or the steam generator overfill paths.'

1 Aovenced PWRs Foeting Minutes September'28, 1989 u.

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  • Charging pumps - have substantial RCS~ makeup capability and

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are autwatically. loaded on the. energency diesels in the-case of 1 css of offsite power.

However, they are not used to -

i mitigate design basis LOCAs.

-l Back-up Seal injection

-.the CVCS contt. ins a back-up seal injection pump which automatically provides RCP. seal; cooling in the event of loss of normal seal injection.' This; pump has--

its own self-contained diesel generator set.

  • Alternate Core Cooling'Means

,in addition to nomal alternate core cooling means (SFWS, EFWS) and their back-up, there are.

i other possibilities.

For instance, RCS feed'and bleed with charging pumps, RCS depressurization and feed and bleed with:

l RHR pumps, and SG feed by main _feedwater or condensate.~ pumps.

i The Secondary Systems consist of the following:

4 Emergency Feedwater system - _ contains 'four-pumps (two _ electric and two turbine driven). Any one ofJ the pumps is sufficient' te remove decay heat through S.G.

' Start-up Feedwater System - a single-'non-safety class pump-driven by a 1E motor _ takes suction from:the condenser hot well l

provides the nomal feedwater function following reactor trip.

a Steam Generator overfill protection'- each S.Gl 1s providedi with an automatic-drain system.to prevent high S.G. level.

The drain path is into -the~ EWST.

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A'dvance'd PWRs Meeting Hinutes

-S-September 28, 1989 q

The Auxiliary Systems consist of the following:

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q Corrponent_ Cocling Water System (CCWS) and Service Water System (SWS) are not interconnected.

Therefore.for events such as CCWS or SWS pipe breaks, only one subsystem can be affected.

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1 Mr. Van de Venne stated that SP/90 is the U.S. Version of of APWR for Japan. Most changes address U.S. licensing issues such as RV level instrumentation, technical support center, four Class IE i

t battery sets, redundant EFW storage ' tanks end' single main steam"

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isolation valves. Some' changes reflectx U.S. siting requirements, j

U Mr. Van de Venne outlined the PRA (core melt frequency) issues for j

the SP/90 design. He indicated that the. transients initiating:

frequency is assumed to be 10 per year.

Most operating plants -

approach 3 per year. SP/90 design has certain design features to reduce trips such as full load rejection capability, main generator t

breaker, enhanced I&C test capability, and no reactor trip follow-ing loss of main feedwater pump.

For specific events such as station blackout, the SP/90 has the following design-features:

  • Chemical ano volume control system includes backup; seal 1 in.!ection pump powered from small. de'dicated diesel generator..
  • Class 1E' batteries a" c4 e/ 4" #our hours of. operation' under 7

station blackout c'

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0-September 28, 1989~

l Advanced PkRs Meeting Minutes

  • Connections are provided between the backup seal injection.

L purp power source and the Class IE batteries.

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  • Station blackout coping time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. At that time, the e'nergency feedveter storage tank and spent fuel l pit need to be f

replenished.

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  • Current SP/90 design exceeds the requirements of Regulatory Guice 1.155.

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Mr. Van de Venne indicated that although risk,due to station blackout is substantia 11y' reduced' relative to current' plants, it'is' still the single largest contributor to SP/90' core melt and con-tairment failure frequencies.

For'the ATWS event, the SP/90 design-incorporates,the following features:

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  • The' integrated protection syst'em is highly reliable:

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- Two-eut-of-four logic 1

- Continuous on-line-testing

- Fail-safe principles.-

  • Reactor trip switchgear consists of eight breakers arranged -in two. separate cabinets which can be tested on-line without-

- g bypass.

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Aovanced PWRs Meeting Minutes September 28, 1989

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. protection system.

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  • Moderator temperature coefficient-is significantly more regative than for current plants..
  • ATWS censideratiunt v'ill be factored into the design of-pressurizer safety and relief valves, s
  • Detailed analyses of ATWS transients will be included in the-FDA application'to demonstrate compliance with ATWS acceptance-criteria.

The ATWS contribution to'SP/90 core melt frequency is less;than

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10~7/yr., and the containment failure following ATWS core meltiis-i highly unlikely.

For the intersysten LOCA, some of the SP/90 design features that'

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prevent erd mitigate such an event are:

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  • RHR-suction line is most credible path intersystem LOCA (10-6/yr.).
  • RhR isolation valves are -included in ISS: test header and will.

be leak tested.~during startup.

  • P.HR suction piping design pressure'has been increased such that gross failure would not occur even when subjected. to RCS operating' pressure.

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Advanced PWRs liceting Minutes September 28.-1989-

  • RHR~ suction piping is in open connection with the in-conteinment EWST such that pressure is relieved-following failure of RHR. isolation valves.

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  • RHR pumps and piping are arranged to. assure sufficient EWSTE inventory to allow continued core cooling;with non-affected; a

ISS subsystems.

kr. Van de Venne stated that the SP/90.PRA indicate the risk from intersystcni LOCA is very low.

t Other events such as vessel rupture loss of cooling, steam genera->

3 tors tube rupture and other internal events have been analyzed in'a conservative manner.

The total core melt frequency has been reduced to 1.5 x 10-6/yr. by systematic-application of PRA tech-r.iques throughout the design process..

No external events PRA has-been performed.

However, external events have been addressed in the design,;e.g.:-

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  • Improved separation _inside' containment.

i Flood

  • General arrangement limits consequences of' storage l

tank leakage,or failure.

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gated with drains to the outside.-

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favancid FKRs Meeting Minutes September 28, 1989 i

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  • Turbine plant. flooding cannot impact nuclear island.

Seismic

' Integral nuclear _ island basemat.

  • Water storage tanks inside building.

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  • Sep6 ration between' redundant divisions.
  • Separation sefety/non-safety.
  • Single point access. control.

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Mr. T. Chu, Brockhaven National' Laboratory (BNL),' outlined the status of ENL review of SP/90 probabilistic. safety study.(PSS); core damage frequency evaluation.

BNL performedian independent assess-ment of the front-end part of the PRA for'WSP/90. The scope-of'the revier included:

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  • Internal events only.

' Shutdown risk not included.

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  • Requantification or core damage frequency.

Mr. Chu stated that generally the methodology used by ][to model.

plant systems is typical of. current PRA practice.

BNL review generally fir.ds' that core melt'frequen(y for 'SP/90 was relatively:

low, with _several-open1 issues remain to be. addressed. Specifical-ly:

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7 Advanced' PWRs Meetilig Minutes

- September 28, 1989

  • Interfacins system LOCA through accumulators and reflood tanks needs to be modeled.
  • Partial or total loss of DC power.-
  • Loss of instruirent air.
  • Partial or total loss of vital AC.

Preliminary BNL assesscent of core detoage frequency is-.

5.98 x 10-6/yr, compared to y PS$ estimete of_1.50f x 10-6/yr?

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Nr. T. Pratt, BNL, presented the probabilistic safnty study- (PES) for the SP/90 containment and offsite risk evaluation which was performeo by BNL. The scope oflthe BNL review was to provide an-independent evaluation of containment perfomance, fission' product

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release, and offsite consequences for.the SP/90' design 'under severe' accident conditions. The BNL" evaluation-Lincluded containmont' event?

tree quantification (MARCH and CONTAIN codes), fission-product release (STCP and CONTAIN codes) and offsite consequence (CRAC2 End q

MACCScodes). The BNL revie9 vas also based on NUREG-1150 met.noc Ll ol.rjy and 59 TOP evert questions and 19 release categories.;

For the effsite risk evaluation, BNL concluded that-if direct-1 containment heating is neglected,- SP/90 risk was predicted'to bel low and if direct centainment heating occurs, predicted risk-is increased:

  • Latent effects and population dose increase slightl'y.

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  • Increase in early health effects'more_ noticeable.

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4 Advenced PWRs Meeting Minutes ScMember 28, 1989

  • Early health effects only slightly effected by evacuation.

(becauseofshortwarning.' time-ifDCHoccurs).-

  • Latent effects ana pcpulation dose not effected by evacuation assumptions.

Mr. Pratt stated that generally, the aethodology used by Westing-house to model cortainment performance, fission product relesse and offsite consequences is consistent with current PRA prac,tice.

i The El,L review gerertily confirmed SP/90 PSS results:

  • Longer times to core damage.

' Siewer conteir. ment pressurization.

  • Iower Risk.

Contairar.ent Performance uncertain -if core mel'ts with primary-

.systemt at high pressure; however:

' High pressure sequences relatively low frequency.-

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  • Long times available for operator actions (e.g., depressuri-zation).

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Mr. Van de Venne described.the mid-loop operation for the SP/90 design. The design includes features such as:

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  • Water level during mid-loop. operation is at least 9: inches.

above actual mid-plane elevation.-

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Alivance'd ?WRs Meeting Minutes September 28, 1989

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  • kith vortex breaker, air entrainment starts to occur at 7

approximately 3 inches below mid-plane elevation, but11s.

limited to less than 10%.

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  • RHR suctinn lines are sloped continuously downwards towards-I RHP pumps and are, therefore, self-venting.
  • RHR pump suction lires provide acequate purp NPSH_at full flow.

assuring saturation in the hot leg.

HHSI pump will be evailable during mid-loop operation for-emergency makeup, if req'uired.

All the concerns raired in Generic Letter 80-17 are being adequate-l ly addressed.

4 For the fire protection issue, the design features (outside con-tainment) incorporates the following:

  • Redundant civisions of safety related equipment are-located in dedicated areas which are seperated from each other and from 4

other areas of.the plant by three hour fire barriers.

  • Each safety aret is provided with its own ventilatinn systems, andpipingandcabling-interconnectionsareminihl zed.-
  • The main control room and main steam tunnel are exceptions to'

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,the above separation arrangement.-

D For inside containment:-

  • Containment constitutes a single fire area.

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l Advanceo PkRs Meeting Minutes September 20, 1989

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  • This single fire area will be subdivided -into several fire zones such that loss of one fire zone will not jeopardize the-t capability to achieve cold shutdown.
  • Separation between fire zones will preferentially be based on.

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existing structural walls. Where this is not possible, other nethods will be used (e.g., no line of sight evoosure, large

' l distancewithnointerveningcombustibles).

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Mr. R. J. Lutz, Jr., W, briefed the' Subcontaittee on severe accident 7

centainment performance issues for the SP/90 design..

fer accident sequences wnich progress to core melting, depressuria zation of the reactor coclant system' prior to reactor vessel

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failure can reduce the dynamic challenges to containment integrity-l tt reactor vessel failure.

Specifically:~

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  • The conteircent pressure transient due to-RCS blowdown is:less; severe.
  • The potential for and/or the.ragnitude?of direct containment-I heating is-reduced.

The SP/90 design includes pressurizer PORVs of sufficient' capacity to depressurize the reactor coolant syster toLless than 200, psi prior to reactor vessel failure for1 severe-accidtnt sequences.'

Depressurization is a backup to the cav'ity des'igr, features to

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preclude direct containment heating.

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Source Terms _- The methodology used:for theiprediction of! severe-j 4

accident source terms (i.e., releasesito the environment) is an-area of. disagreement between the U.S. NRC and the industry.-

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Abvance'd PWRs Meeting Minutes September 28, 1989 The SP/00 severe accident source terms were calculated using the MAAP 2.0B computer code. The source terms predicted by MAAP 2.0B are considered to be reclistic, based on best estimate inethodology.

Hydrogen Contrcl - For Accident sequences which progress to core melting, the accumulation of hyd'rogen in the containment may lead to flamability-conditions which can challenge the containment integrity.

Specifically; Mr. Lutz indicated that:

  • Per 10 CFR 50.34(f), the containment design shall be capable of withstanding a-hydrogen burn which involves a hydrogen inventory from reaction of 75% of the core active cladding zirconium inventory.
  • The hydrogen concentration in the-containment should remain below the limits at which transition to detonation would--

occur.

The containment design in the SP/90 is capable of withstanding a hydrogen burn resulting from a hydrogen inventory equivalent to the reaction of 75%; of the core zirconium inventory.

Tbc containment volume is large enough so that even in the event of hydrogen generatior: eouivalont to the reaction' of -75% of the core.

i zircenium inventory, the containment hydrogen would'not exceed 15 V/0.

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.o Although not required by the analysis results, hydrogen igniters; have been inclxed in the SP/90 design spect ?ications, Containment Venting - For the SP/90 design, the analyses indicate.-

that no containment failure is predicted during the first two daysi

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following a ~ severe accident; this is considered to be 'a' sufficient'

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September 28, 1989 Advanced PWRs Meeting Minutes time v ind to initiate accident management strategies to prevent ultihnte containment failure. Thus, the need for containment L

venting is precluded.

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Containnent Performance - For the SP/90 design, best estimate j

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analyses predict that the containment integrity will be maintained for at least 2 days for all severe accident sequences. This is a sufficient time period to iritiate accident management strategies l

to prevent containtrent failure for any severe accident sequence.

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l The capability to maintain containment integrity for 2 days follow-100 a severe accident is not sensitive to reasonable uncertainties

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in the dominant severe accident phenomena.

Core / Concrete Interaction - For the SP/90' design, the reactor i.

cevity is designed to preclude core / concrete interactions for all I

i severe accidents in which a water layer can be maintained in.the reactor cavity.

l The plant is designed to preclude cavity dryout by providing an alternate water supply from the EWST.

Mr. Van de Venne sunnarized at the end of Mr Lutz's presentation by stating that the:

  • Current SP/90 design is. expected to meet nuclear industry safety goals stated in EPRI ALWR requirements document; 1 5 per year core celt frequency.

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) 6 per year severe release-frequency..

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Advance'd PliRs Mee?.ing Minutes September 28. 1989

' If cetailed PFA analyses to be performed at the FDA stage indicate that the SP/90 design does not meet the industry goals, additionti features will be evaluated and incorporated to the extent-required to meet the goals.

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As a result of the Subcommittee discussion, some of the.Subcomit'-

tee's trembers expressed sorre. concerns in regard to the following:-

  • Dr. Catton expressed some concern that neither the NRC staff nor E has carefully-analyzed the flow instabilities'and vibration in the steam generators to investigate any~ fluid

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structural interaction problems.

  • Mr. Michelson comented that enother subcomittee' meeting is needed to discuss the safety-related systems before-finalizing l

the'PRA analysis.

Itr. Carroll indicated that the interaction.between y and EPRI j

recarding the compliance With the EPRI requirements document H

is not very clee.r at the PDA stage.

Kr'. Michelson agreed.

  • Mr. Carroll questioned the significance of the PDA-approach andaskedhrepresentativesif-thereisany'fundedplansfor final design approval that exist at the present time. ~W's response was no, and especially in the U.S.-

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  • Mr. Michelson advised the APWR Subcomittee to review y Modules (1 thrcuch 16) chapter by chapter..

t Dr. Catton indicated that throughout the~SP/90 design; the j

check valves perforinance, piping and location was not analyzed; carefully.

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/.dvanced Phh.weting 111rutes September 28, 1989.

Dr. Latton questioned the-depressurization process for the SP/90 design ar.d advised H to study the work that was per-formed by Mr. Aubrangeli.

Dr. Catton indicated that W may benefit from depressurizing scener and faster.

' Dr. Catton questioned the'use of the ill-documented MAAP computer code for ca.lculation-in-the severe accident source term.

  • Pr. Carroll inoicated that there is an error in the W's-presentation of.the hydrogen control.. According to' 10 CFR

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y Part50.34(f);H concentra+4n should not exceed 10% during 2

and following an accident that releases an equivalent' amount of H as would be generated from a 100% fuel-clad metal water 3

reaction /ardnot'75%asclaimed-byH.

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  • hr. h'ard questioned y philoso'phy for the hydrogen contro1 ~

issue to install the igniters..

m Mr. Michelscn expressed some concern regarding the lack of data base to support the final results'of the PRA analysis, especially for accident conditions.

Mr. Michelson expressed some concern.regarding the ' definition V

of.PDA and FDA in regard to the design and what kind:of. study-O and analysis will be included under each definition; e

Future-Action h

The Subcommittee Chairman and Members decided-to conduct another Subcom-mittee meeting on November 3, 1989, to continue' discussion of the y

subject matter.

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Advt.nced PWRs Meeting Minutes September 28.-1989 NOTE:

Additional meeting details cer be obtained f rom a transcript of this meetino available in the NRC Public Document Room, 2120 L Street, N.W..' Washington, D.C. 20006,(202)634-3273, or can be purchased from Heritage Reporting Corporation.1220 L Street, N.W., Suite 600, Washington.-D.C. 20005,'(202)

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628-4886.

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