ML20056C376

Summary of 930426 Meeting W/Ge in Rockville,Md to Discuss Resolution of Severe Accident Issues for Advanced Bwr. Attendees Listed in Encl 1

ML20056C376
Person / Time
Site:	05200001
Issue date:	05/10/1993
From:	Son Ninh Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9305200347
Download: ML20056C376 (54)

Text

. - - - - - . - . _.

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g kB REG l

ye . 4 UNITED STATES g ,

j .j NUCLEAR REGULATORY COMMISSION l

g WASHINGTON, D.C. 20555-0001  ;

%, sf

• • • • • May 10, 1993  !

Docket No.52-001 l

APPLICANT: GE Nuclear Energy (GE)

PROJECT: Advanced Boiling Water Reactor (ABWR) ,

SUMMARY

OF MEETING BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION i (NRC) STAFF AND REPRESENTATIVES OF GE (APRIL 26,1993) {

i i

A public meeting was held between the NRC staff and GE at m NRC offices j in Rockville, Maryland, on April 26, 1993. The purpose of this meeting was l l

primarily to discuss the resolution of severe accident issues for the ABWR. -

Enclosure 1 is a list of attendees, and Enclosure 2 is the meeting agenda.  ;

Enclosure 3 is a copy of t L handouts provided by GE during the meeting. l Enclosure 4 is a copy of the staff's comments on the severe accident analysis i input to Tier 1 and 2.

l As result of this meeting, both the NRC and GE staff have agreed on the paths to technical resolution of all outstanding severe accident issues. Agreement l was also made regarding GE and staff actions needed to address issues identi- 1 fied by the Advisory Committee on Reactor Safeguards in the recent meeting. l The highlights of the discussion items, the staff actions, and GE's commit-ments are summarized below- l i

Containment Sump Desian - The staff informed GE that Sandia National Laborato- i l

ries would evaluate the sump shield design provided by GE. GE indicated that the conceptual design would now be the final design. The staff will provide GE with the results of the review of the sump design when it is completed.

Containment Bypass (Desian Bases Accident) - GE agreed to provide to the staff by May 21, 1993, the following information: (a) an evaluation of the. capability of the ABWR containment relative to Mark II containments, (b) a i'

clarification of A/k zti , and (c) an analysis of bypass

• specified by Standard Review Plan Section 6.2. Further interaction between GE and the staff is necessary to reach closure for this issue.

Containment Ultimate Pressure - The staff commited to provide GE with a preliminary safety evaluation report by April 30, 1993. A conference call to further discuss this issue was scheduled for April 28, 1993.  !

4 130010 9305200347 930510 DR ADOCK052OOg1 . { } } {. _{gy.,. . - y. , h- -

May 10, 1993 Sucoression Pool oH Control - GE agreed to add a section to the standard safety analysis report (SSAR) to address this issue. The preliminary calcula-tions provided by GE indicated that the suppression pool pH would not become acidic within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a severe accident. The staff concluded that this was sufficient information to close out this issue.

l Hydrocen Detonation (Containment Overoressure System (COPS) and Bypass) - GE l stated that no credit is given to the stack for elevated releases, and that the reactor building is designed to withstand the collapse of the stack. GE indi:ated that the risk of a hydrogen detonation following a bypass scenario is sufficiently low. GE agreed to check on the reclosure capability of the isolation valves in the COPS line following a detonation. GE committed to <

provide this information to the staff for review by May 21, 1993. The staff indicated that this issue would be closed following receipt of the information relative to the isolation valves.

Gratino (Concrete Core Interaction (CCI) and Fuel Core Interaction (FCI)) - GE indicated that the grating was not expected to have a significant impact on containment performance. The staff concluded that GE had provided sufficient qualitative information during this meeting to close out this issue.

Eauipment Survivability - GE agreed to conduct a review of all equipment for which credit is taken during a severe accident to ensure that it will operate.

l' GE indicated that at least the following systems would be discussed:

(1) automatic depressurization system (2) residual heat removal system -

(3) fire water system (4) passive flooder system and

. (5) containaent overpressure protection system GE also agreed to address degraded core melt scenarios in which the core remains in-vessel. As part of this, GE would evaluate the need for instru-mentation. GE agreed to provide this evaluation to the staff for review by May 21, 1993.

Containment Bypass (Severe Accident) - GE provided a discussion of the aerosol plugging model and agreed that a discussion would be added to the SSAR. GE also agreed to provide : road map of the impact of bypass on key parameters, to discuss qualitative' the impact on risk of the plugging model, to provide the data on the knife-steam and non-soluabb. ae,e rosolorifice

s. GEinto the SSAR, committed to and to discuss provide the impacttoof this information the staff for review by May 21, 1993. The staff indicated that this would provide sufficient basis to close discussion on this issue.

Staff MELCOR Analysis - The staff informed GE that Sandia National Laborato-ries would perform MELCOR analysis of accident sequences similar to that performed by GE using the MAAP code. The staff would provide GE with prelimi-nary results by June 15, 1993.

l

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May 10, 1993 Closure of FCI Analysis - The staff informed GE that closure of the FCI issue l was based on the low probability of a pre-flooded lower drywell.  !

Desion Control Document - Significant discussions ensued between the staff and ,

GE on the following issues: probabilistic risk assessment (PRA) and severe  :

accident insights, development of Tier 1 and 2 information, and modifications to Chapter 19 in the design control document. GE agreed to provide a com-pleted list of the design insights based on PRA and severe accident closure by May 31, 1993. GE also agreed to provide the "strawman" Tier 1 and 2 and inspections, tests, analyses, acceptance and criterie on COPS by May 31, 1993.

(Original signed by)  ;

Son Q. Ninh, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page DISTRIBUTION w/ enclosures:  !

Docket File PDST R/F DCrutchfield SNinh PDR PShea DISTRIBUTION w/o enclosures:

TMurley/FMiraglia JKudrick, 8D1 RBarrett, 8D1 BPalla,10E4 JMonninger, 8D1 TBoyce CPoslusny GKelly, 10E4  ;

EJordan, MNBB3701 RBorchardt ACRS (11) GGrant, 17G21 l JMoore, 15B18 J0'Brien, RES LShao, RES bha , RES ,, l OFC: LA:PDSTJ: DA PM T:ADAR PM:PDST:AD/>R

' PDST:ADAR

/ JJ ilson h C56 S NAME: PShea g - SNi@nhitz CPoslusny t$ 'ET j DATE: 05/// /9 05/q/93 05/9 /93 5/t/93 o e l' G3 l

0FFICIAL RECORD COPY: MSUM426.SQN l

l 1

1 GE Nuclear Energy Docket No.52-001  ;

l cc: Mr. Patrick W. Marriott, Manager Mr. Joseph Quirk Licensing & Consulting Services GE Nuclear Energy .

GE Nuclear Energy General Electric Company i 175 Curtner Avenue 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 San Jose, California 95125 Mr. Robert Mitchell General Electric Company  :

175 Curtner Avenue San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs GE Nuclear Energy '

12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division i Office of Radiation Programs U. S. Environmental Protection Agency  !

401 M Street, S.W.

Washington, D.C. 20460 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C. 20585 i Mr. Steve Goldberg i Budget Examiner 725 17th Street, N.W. I Room 8002 Washington, D.C. 20503 l i

Mr. Frank A. Ross l U.S. Department of Energy, NE-42 l Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 Mr. Raymond Ng 1776 Eye Street, N.W.

Suite 300 ,

Washington, D.C. 20006  ;

i Marcus A. Rowden, Esq.

Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.

Suite 800 Washington, D.C. 20004 ,

Jay M. Gutierrez, Esq.  ;

Newman & Holtzinger, P.C.

1615 L Street, N.W.

Suite 1000  ;

Washington, D.C. 20036

l l

GE ABWR SEVERE ACCIDENT MEETING April 26, 1993 l

NAME ORGANIZATION Son Ninh NRR/ADAR/PDST Norman Fletcher DOE /ALWR Stephen Additon TENERA/ARSAP Jack Kudrick NRR/DSSA/SCSB Richard Barrett NRR/DSSA/SCSB John Monninger NRR/DSSA/SCSB Bob Palla NRR/DSSA/SPSB 4 Allan Beard GE Jack Duncan GE Carol Buchholz GE Tom Boyce NRR/ADAR/PDST Chet Poslusny NRR/ADAR/PDST Sterling Franks DOE /ALWR Glenn Kelly NRR/DSSA/SPSB Bill Borchardt NRR/ADAR/PDST.

I Enclosure 1 l

l

NRC/GE SEVERE ACCIDENT MEETING Date: April 26, 1993 Location: OWFN 4-B-13 l

Participants:

Beard, Buchholz, Duncan, GE i Russell, SCSB, SPSB l

AGENDA OPEN ITEMS FROM MARCH 30, 1993 SENIOR MANAGEMENT MEETING

- Containment Sump Design

- Containment Bypass (DBA)

- Containment Ultimate Pressure

- Suppression Pool pH Control

- Hydrogen Detonation (COPS and Bypass)

- Grating (CCI and FCI)

- Equipment Survivability

- Containment Bypass (Sev. Acc.)

RESOLUTION OF OTHER ITEMS

- Staff MELCOR Analysis

- Closure of FCI Analysis DESIGN CONTROL DOCUMENT

- Insights

- Submittal Schedule

- Partitioning of Chapter 19 for Tier 2

- Development of Certified Design Description

- ITAACs for Severe Accident Aealysis BRIEFING FOR SENIOR MANAGEMENT l

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1 l

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Enclosure 2

I h GENuclearEnergy 1 Resolution of Severe Accident Issues for ABWR i

i Meeting with the NRC Staff i

l CarolE. uchholz, PrincipalEngineer ABWR Engineering 1

April 26,1993 i 2 2

w

Topics to be addressed

• Aerosolplugging creditin bypass analysis

• DBA containment bypass

• Level 3 submittal and impact of bypass on risk

• Hydrogen detonation in the COPS or in the reactor building due to bypass

• Suppression poolpH

• Impact ofgrating on core debris coolability

• Equipmentsurvivability

• Areas where GE requires further clarification

• Severe accidentinsights

• Minorclarification issues e

l F

CE8 M 19 D-2

1 AerosolPlugging issue:

Could the plugged leak paths through the partially opened vacuum breakers be porous enough to allow significant gas flow?

History:

This issue was raised in the ACRS meeting of 3/18.

Reference:

" Leakage of Aerosols from Containment Buildings", H. A. Morewitz, Health Physics Vol. 42, No. 2 (February), pp 195-207,1982.

Response

A discussion will be added to the SSAR 19EE.2.3 as outlined on the d following page.

1 I

Porosity of AerosolPlug

• A small, concrete, tilt-up-panel building was tested at Atomics internationalin the early 1960's Building was overpressurized and cracked so that it leaked badly To plug the leaks, a sodium fire was lit inside the building

- Observers were stationed around the building No smoke was seen issuing from the building Upon pressure testing the building, no gas leaks could be detected

'

• Some experiments indicate a porous plug was generated for large pressure differences (30-1000 psia) l

• Maximum pressure difference across the plug in the AP*i vR will be limited to the head of water above the first row of horizontal vents (14 psid)

• Complete blockage is expected i,

g

i

DBA Bypass Leakage "k"vaka of 3.0 Issue:

GE should provide justification for a 'k" value of 3.0? Why is "k" defined? The unidentified flow path value is ANk. Therefore, by assigning a value to "k", GEis defining the flowpath characteristics.

l The staff believes it is inappropriate to assign a value to an unidentified flowpath.

History:

This question is #8 on the Monninger List. The response was discussed with the staff on 11/17/92.

i l

h cranzenas l

! O DBA Bypass Leakage "k"value of 3.0

Response

• Methods used for ABWR are standard practice, the same as Mark 11 andMarkIII

• Unidentified leak paths will typically be cracks in walls or floors

• Assumed contributors to loss coefficient are:

Entrance - 0.5 i -

Exit - 1.5 Through flowin orifice - 1.2

• Total value of3.0is used

• Typical value for a vacuum breaker is approximately 2.7 to 2,8 Assumed k value is appropriate and consistent with current practice CEB 4/16/17336

4 Fullopen vacuum breaker Issue: ,

The staff requested GE to submit the analyses that demonstrates that the i design can accommodate bypass assuming that one breaker fails full-open. Per a June 12,1992 meeting, GE made reference to an

}

analysis assuming one vacuum breaker valve failed full open which j corresponds to an area of 1.26 ft2 and a containment spray flow rate i of 500 gpm. GE should also address the role of containment sprays ,

j with regard to bypass during a S80.

l History:

l This is issue #6'on the Monninger List i .

Response: _

i The following analysis indicates that the ABWR can accommodate the i case of a feedwaterline break with the late failure of a vacuum

l. breakerto close.

A i

CEB4R6n93N l-

)

1 Late Failure of a Vacuum Breaker - the scenario

• A feedwaterline break initiates the accident l

• Water level falls and initiates ECCS injection 1

i

• Cold ECCS water flows through the break

• DryweIIpressure drops causing the vacuum breakers to open l

• As pressure continues to rise the vacuum breakers should close but one ispresumedto stick fullopen

• Sprays are initiated i

i i

I i

3 4

CEBM1993-0

BYPS-93.XLS TAHLEI KBWFJ~STEKM BYPASS CEXKXGE SENSITIVITY STUDY I I EVENT: FEEDWA~ TER LTNE IIRETKCU0'M BRE~AKERTATL TO CEOSE-~AFTER7T FIRST OPENED I I

SUMMARY

RESULTS CASE SNUMB #, Date Peak Pros RHR HPCF RCIC Spg s REMARKS

1. psla) 1Available (Available)_(Available) (WW & DW)

(peia) 1 4958T, 4/19/93 106 1LPFL 1 HPCF RCIC NO LPFL Throttled to Maintain LB 1

2 5038, 4/19/93 __

99 1 LPFL 1HPCF RCIC YES 1 WW Spray,500 ppm (Current Capacity) @ 250 sec LPFL Throttled to Maintain L8 ,

2A 5799T 105 1LPFL 1 HPCF RCIC YES SAME AS CASE 2. Except Spray Actuation @ 1800 sec 3 4932T, 4/19/93 96 1 LPFL 1HPCF RCIC YES 1 WW Spray,1140 gpm (Operator Control)

LPFL Throttled to Maintain L8 4 4789T, 4/18/93 94 1 LPFL 1HPCF RCIC YES 1 WW Spray,500 gpm AT 250 seconds 1 DW Spray,3700 gpm (Current Capacity)

LPFL Throttled to Maintain LS 5 4786T, 4/18/93 92 2 LPFL 2 HPCF RCIC NO 1 LPFL Throttled to Maintain L8 Page 1

Impact oflucreased WetweII Spray Flow Rate on Bypass Capability Issue:

• The staff will review GE's analysis on the maximum allowable leakage path area. GEshouldprovide the analyses varying the containment spray flow rate to determine the maximum allowable leakage area.

• GEshouldprovide the analyses showing that an increase in wetweII spray flow rate does not improve bypass capability. If the ABWR design cannot accommodate one vacuum breaker failed full-open, can the spray flow splits between containment areas be changed with i

orifices to limit bypass challenges on the containment? GEshould 1 provide the basis for the selected design flow rates.

History:

l These are issues 6 and 7 from the Monninger list.

i. Response:

l The following graph indicates the bypass capability assuming different times and maximum wetwell spray flow rate capability. Also given

. are some design challenges for the increased capability of wetweII

'- sprays.

CE84/2W193310 ,

l. #

Design challenges associated with increased WWspray flow rate

• Wetwell spray flow rate could be increased from 500 gpm to about 1150 gpm without changing the spraypipe size

• This increase would result in 35 ft/sec fluid velocity in spray pipe for pump runaut conditions - the same value used in previous BWR designs

• If drywell spray flow rate is retained, the combined operation of wetwell spray and dryweII spray is not possible

• The rated flow of 4200 gpm must be retained for all modes to ensure l adequate decayheatremoval

• A multi-step initiation would be required:

Initiate pool cooling in normal rated flow (4200 gpm)

Throttle back pool cooling with the heat exchanger throttle valve to abouthalf the ratedflow(2000gpm)

Initiate wetweII spray, adding about 1000 gym to the flow rate Bring the combined wetwell spray and pool cooling to the full rated

. flowusing the 4200gpm l CTB4/2tV199311 4

e o

BYP4 93.XLS Chart 13 ABWR STEAM BYPASS LEAKAGE CAPABILITY

[W SPRAY: 500 gpm (current design) NO DW SPRAY l 0.40 O

0.30 x

w /

0.20 A P T9

>C W . Enadei achu W.

b=.a .a cvas A

• m 0.10 "

m A

8 II Desg.g

>. No co 0.00 0.00 200.00 400.00 600.00 800.00 1000.00 1200.00 1400.00 1600.00 1800.00 2000.00 TIME WW SPRAY ACTUATED (sec)

Page 1

4 Use of cutoff frequency in Level 3 analysis Question:

l Table L2-1 (190.5-7) does not account for all cases run in level 3. A

complete accounting of allSTCs which contribute to each Level 3 l case is needed.

interpretation:

I believe GE is being asked to include all branches of the STC trees

(Figure 190.5-3) in Table 190.5-7 and the level 3 analysis).

l History:

Bob and I exchange l1 voice mail several times but we have not yet actually discusskd the response. Informal fax sent on April 1.

l Response:

i l AII cases with a frequency of IE-11 were included in the analysis. The V remainder of the sequences (22) have a frequency ofless than IE-11

and therefore will not contribute to containment performance i measurements. No change is necessary.

2 l CEB tRW199112 l

1

Level 3 analysis 1 Question:

Provide a level 3 submittalin SSAR format with LOCA outside containment factored into the risk profile.

i History: .

! The affected tables and figures were submitted in June (without LOCA

outside containment). The request to include LOCA outside

containment in the risk profile was discussed at the end of March.

Response

l l The text associated with the Level 3 submittal has been prepared.

l Including the consequences associated with LOCA outside containment in a manner which would not overstate the risk may require substantial i effort.

l G$ willinvestigate the possibility of a simplified approach to include these sequencesin the riskprofile.

l =

l CEBof26/1993-14

['

i

Hydrogen Detonation in the Reactor Building via Bypass issue:

If a severe accident occurs with an unisolated containment, can hydrogen detonation occur which could threaten the reactor building integrity.

History:

l This issue was raised by Michelson at the March 18 ACRS meeting in l relationship to an unisolated RWCU line break.

Response

l

\

• The RWCU line break was not evaluated in the backend analysis since l the frequency of this event leading to a severe accident is very small

{ (on the order of E-11). There will be no impact on risk.

i

• A LOCA outside containment is not expected to have any impact of risk b since at most one division covId be affected due to the divisional

( separation and the high pressure capability of the structures.

Therefore, any damage caused by a hydrogen detonation should not propagate to other rooms or divisions.

l

(

• The impact of this question will be considered in the development of a

\ means for inclusion of bypass in the Level 3 analysis ,, ,, ,_,,

Hydrogen Burning or Detonation in the stack Issue:

Is it possible for hydrogen burning or detonation in the stack to lead to an additional failure which could cause an increase in risk?

History:

i This issue was raised in the ACRS meeting on March 18. GEindicated we thought we had analyzed the impact of detonation in the stack.

I Response:

• We have been unable to locate anyprevious calculations on this

. subject. However we do not believe there will be any impact on risk.

1

• No credit was taken for the presence of the stack in lofting the fission product release from the containment.

l

• Since the stack is a non-seismic structure, the reactor building is designed to withstand the collapse of the stack without sustaining l damage to the reactorbuilding.

• Piping leading to the stack will be designed to withstand the burn / detonation. A pressure rating of 350psiis used in the piping of thi offgas system for a similar challenge.

[ a g ,, ,_,,

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1

- - -____________________.__________..__m_____ __. .. . - _--.. - - .. ,. _ . - . . - . - _ _ _ _ . . . _.

Suppression PoolpH Control Question:

Research performed at Oak Ridge indicates that a substantial increase in airborne iodine can occur if the suppression pool becomes acidic.

The radiolytic formation of nitric acid has been identified as a mechanism for the pH of the suppression pool to decrease. What impact does this mechanism have on the fission product behaviorin the ABWR?

. History:

This issue was raised by Dr. Kress of the ACRS in the March 18 meeting.

Reference:

NUREG/CR-5732, " lodine Chemical Forms in LWR Severe Accidents", Final Report, January 1992.

Response

A preliminary calculation has been outlined. These calculations, which should be considered preliminary follow.

A section will be added to the SSAR to address this issue.

y asamu

Suppression PoolpH - Scoping Calculation l .

• Perthe reference:

"If the pH is controIIed so that it stays above 7, a reasonable value for the fraction ofI- converted to 12 is 3E-4. ... Table 3.6 indicates a small

\ production of volatiles for PWRs but virtually none for BWRs."

• Results of preliminary analysis indicate pool pH will NOT become acidic i

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> forphysically realistic accidents due to prpsence of of Cs0Hin the pool.

i Time (hr) pH' pH 2 pg 3

O 7 9.65 10.56 1 5.6 9.65 10.56 l 10 4.6 9.63 10.53 t

i 24 4.3 9.59 10.49 i

~

Case 1 - Initially neutral pool, 80% of Csl and Cs0H energy - not physical l Case 2 - 10% initial Cs0H in pool,10% of Csl and Cs0H energy

Case 3 - 80% initial Cs0H in pool, 80% of Csl and Cs0H energy

. cremvim.ra e

___________.___.J

FMCR0 platform grating I '

Issue:

What is the impact of the FMCR0 handling machine refueling platform in the lower dryweII on fuel coolant interactions and core debris coolability. .

5 History:

This issue was raised in the March 18 ACRS meeting.

Response: .

i The grating is not expected to have a significant impact on containment l performanc'e for the reasons described in the following slides.

i

' 084f26/1933.I9

l i

FMCR0 platform grating i

• The rotating platform is circular and mounted on the rotating rail under the reactor vessel. There will be an opening area at the center and will be provided with a traveling rail for he CR0 handling devise. Gratings will be installed on both sides of the rails.

1

• In previous designs the grating consists of 1"by 3/8" metal stats mounted edge-wise to form a grid with a grid size on the order of 1"by 2"

• Early fragmentation of debris will tend to increase the voiding of any pre-existing waterpool. This will tend w reduce the loading from an FCI decreasing the potential for early containment failure from this mechanism.

• The grating will quickly ablate due to the flow of debris in a manner i similar to the ablation of the vessel bottom head. Therefore, there will be i ^

virtually no effect of the grating after the initial debris pour.

[

• The late debris pour would be a slow, drip-like relocation which would

? fall straight through the ablated region of the platform.

i b

i .

i. creenenmx U

Equipment Survivability Issue:

Can the equipment called upon to operate during a severe accident perform the required function under severe accident conditions i History: -

This issue was raised during the March 18 ACRS meeting.

Response: .

GE will conduct a review of all equipment for which credit is taken '

during a severe accident to ensure it can operate This evaluation will not use equipment qualification standards as a strict measure A few examples are provided on the following page 1

.I h

0 8 4/26/1933-21

l l .

l Equipment Survivability- examples ADS

• These valves must be able to remain open during the melt process to ensure that any potential vessel failure occurs at low pressure

• The radiation environment used for design basis is more severe than that in a physically-realistic core melt accident

• It is expected that the thermalloads on the elastomers in the actuator, which limit the temperature capacity of the valve will be very similar to

that used for equipment qualification RHR

• The RHR system may be called upon to remove decay heat from the l containment l

• The integrated dose rate during a severe accident will not reach design basis values forseveraldays I

• The temperature profile for the system will be similar to the design basis Firewater

,

• AII components of the firewater addition system will be outside of

{ contiinment and will not be significantly affected by the accident a_,,

j

Areas where GE require further clarification

• Containment Ultimate Pressure

- A conversation was held clarifying the January submittal between Gary Ehlertand Gutam Bagchi

- The staff has indicated it has further questions about the analysis submittedby GEin late January

• Containmentsump design

- The staffindicated that they would teII GE what further information is required i

1 TOAOWlW-E0 r

i - _. - ._.

Overflow ofsuppression pool Question:

Resolve the amount of water required to result in the overflow of water from the suppression pool to the lower dryweII (See CEB-92-X-26, MAAP runs with overflow @ 4.2E6 kg and EPGs) A problem with the pool heights indicated in Figure 19E.2-3c.

History: .

(

issue was raised in regard to steam explosions. Questions about the overflow conditions were resolved in CE892-60#2. Additional informalfaxsent on April 1.

Response

Further clarification of the question is needed.

094/2M993-73

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4 i(' .

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Drywell/Wetwell Vacuum Breakers l l '

l s

f.

! Staff: Reliable wetwell/drywell vacuum breaker operation .

1

is very important to reducing the consequences of

! an accident. This is because suppression pool

function (severe accident progression and fission  !

j product removal) can be compromised in the ABWR

design by a single failure of a wetwell/drywell l vacuum breaker (i.e. , a stuck open vacuum breaker),  ;

] or by excessive leakage of one or more vacuum '

i breakers. In the ABWR design there is only a i single vacuum breaker in each path in operating BWRs. Periodic surveillance and testing of the l vacuum breaker valves and their associated position j indicators will be crucial to minimizing this risk.  ;

The vacuum breakers and their position indication 3

i switches are very important in reducing the chances j of suppression pool bypass. GE has taken credit

! (reduction of suppression pool bypass from 18% to 2% for cases in which releases are into the

drywell) for aerosol plugging to reduce the

significance of vacuum breaker leakage. The PRA assumes that (1) each vacuum breaker will be l equipped with a position indication switch that 3

j will indicate the valve to be open should the gap i i between the disk and seating surface exceed 0.9 cm, and (2) normal operating procedures will call for

the operators to periodically confirm that all vacuum breakers are closed.

y!

i i GE: The ABWR contains eight 20-inch diameter vacuum i breakers which provide positive position indication I

< in the control room. They have also been located  !

' high in the wetwell to reduce potential loads l occurring during pool swell. The result of the  !

vacuum breaker design in the,ABWR is to reduce the potential for suppression pool bypass. I j

Conclusion:

GE/ Staff agree on what is important. Staff's items  !

(1) and (2) could be included. Reference to PRA l results not necessary.

l l

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..,-,,,e. =- .,--r..-

, . , , - - - - ,-,--,-ew -

---w.-- ---- w-,

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.- i l

Passive Flooder Staff: The passive flooder system provides a passive means of adding water to the lower drywell following reactor vessel breach. This water would cover the core debris, thereby enhancing debris coolability, cooling the drywell, and providing fission product scrubbing. The passive flooder is a backup to other means of lower drywell water addition in the ABWR. FRA-based sensitivity studies indicate that the incremental risk reduction offered by the passive flooder is system is minimal. This is because of credit taken in the ABWR for water addition using the ACIWA mode of RHR.

GE: The lower drywell flooder system has been included in the ABWR design to provide alternate cavity flooding in the event of core debris discharge from the reactor vessel and failure of the firewater addition system. This system is actuated from the ,

melting of a fusible plug. The temperature set j point for the plug is 533 K. The system consists of ten 4 inch diametet lines located about 4 m below the normal suppression pool water level.

discharging into the lower drywell about 1 m above the floor. Assuming only 9 of the 10 flooders open, the total flooder flow would be 97 kg/s. By flooding after the introduction of core material, i the potential for energetic core-water interactions during debris discharge is minimized. The flooder I will cover the core debris with water providing for  !

I debris cooling and scrubbing any fission products released from the debris due to core-concrete

. interactions.

Conclusion:

GE text provides more details of system. Staff's mention of PRA sensitivity studies not pertinent to the discussion.

O

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COPS Staff: The pressure at which the COPS system vents is important to safety. The COPS system provides for a scrubbed release path in the event that pressure in the containment cannot be maintained below the structural limit of the containment. COPS provides significant benefit to Class II sequences and seismic events. Two competing goals were considered in establishing the pressure relief setpoint, specifically, minimizing the probability of drywell failure prior to COPS actuatica, while maximizing the time before fission product release to the environment. As such, the tolerance in the pressure at which the system actuates is an important aspect of the design. The ABWR PRA assumes the following system par & meters: (1) rupture disk actuation at a pressure of 90 psig, with atmospheric pressure on the downstream side of the disk, (2) reliability of the system actuation within +/- 5% of the mean value, (3) upstream and i dowristream isolation valves maintained normally '

open.  ;

I GE: The COPS is part of the atmospheric control system and consists of two 8-inch diameter overpressure relief rupture disks mounted in series on a 14-inch line which connects the wetwell airspace to the ,

stack. This system will provide for a scrubbed  ;

release path in the event that pressure in the l containment cannot be maintained below the structural limit. This controlled release will occur at a containment pressure of 0.72 MPa (90 psig). This system is beneficial for several of the severe accident issues. In cases with continued core-concrete attack, or those with no containment heat removal operational, the i containr.2nt will pressurize. The COPS provides a controlled release path which will mitigate the l fission product releases. This is an example of I how uncertainties in severe accident behavior, i.e.

debris coolability, are addressed by the ABWR design.

Conclusion:

GE/ Staff agree on what is important. Could include competing goals discussion and (2) and (3) from above.

_ ~ _ __

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i Sprays i

Staff: Wetwell/drywell sprays provide significant benefit l for mitigation of severe accidents. In the event l of a stuck open vacuum breaker, excessive vacuum l breaker laakage, or other bypass leakage to the ,

wetwell airspace, the wetwell and/or drywell sprays l provide a backup means of containment pressure and fission product control. .

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GE: This system not only can play an important role in i preventing core damage, it is the primary source of )

water for flooding the lower drywell should the i core become damaged and relocate into the containment. The drywell spray mode of this system i not only provides ~ for debris cooling, but it is capable of directly cooling the upper dryvell atmosphere and scrubbing airborne fission products. i This system has sufficient capacity to cover the core' debris ex-vessel and provide debris cooling and scrub fission products released as a result of ,

continued core-concrete interactions.  !

The firewater addition system operating in the drywell spray mode will also reduce the consequences of a suppression pool bypass or containment isolation failure. This is due to the fission product removal function performed by this mode of operation. Fission products will be l scrubbed by the sprays prior to leaving the  ;

containment. _

The firewater addition system has been sized to l optimize the containment pressure response. The l system is capable of delivering whter to the containment up to the setpoint pressure of the COPS system. The flow rate, nominally 0.055 m'/sec at runout and 0.044 m*/sec at the COPS setpoint, is, sufficient to allow cooling of the core debris, while maximizing the time until the water level reaches the bottom of the vessel, at which point it is turned off.

Conclusion:

.GE/ Staff agree on what is important. GE discussion is more comprehensive. l l

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. Inerted Containment  ;

Staff: Because the ABWR containment is inerted during normal operation, containment failure as a result l 4 of hydrogen combustion is not considered important l for power operation, and was not modelled in the  !

PRA.  !

i GE: One of the important severe accident consequences f is the generation of combustible gasses.  ;

Combustion of these gasses could increase the 3 1 containment temperature and pressure. . The ABWR  ;

containment will be operated inerted to minimize the impact from the generation of these gasses.

~

Conclusion:

GE/ Staff agree on what is important.

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Vessel Depressurization Staff: ADS plays a major role in reducing the frequency of containment failure from direct containment heating (DCH). As a result of ADS, the fraction of reactor vessel failures at high pressure are reduced to

! about 30 percent. The staff has estimated the containment failure probability for DCH to be about 5%, conditional upon reactor vessel failure at high pressure. This results in a very low absolute frequency of a core damage event leading to containment failure from DCH (2.3E-9 per year).

GE: The ABWR reactor vessel is designed with a highly reliable depressurization system. This system plays a major role in preventing core damage, l however, even in the event of a severe accident,

! the RPV depressurization system can minimize the affects of high pressure melt ejection. If the reactor vessel would fail at an elevated pressure, fragmented core debris could be transported into the upper drywell. The resulting heatup of the upper drywell could pressurize and fail the drywell. Parametric analyses performed in Section 19AE of the ABWR SSAR indicate that even in the . '

event of direct containment heating, the probability of early drywell failure is quite low.

The.RPV depressurization system further decreases the probability of this failure mechanism.

Conclusion:

GE does not agree that PRA results are important here.

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Lower Drywell Design (CCI)

Staff: The contribution of core concrete interactions (CCI) to risk has been minimized in the ABWR design by the inclusion of the following design features:

(1) a m thick reactor pedestal capable of withstanding approximately n of erosion from l CCI without loss of structural integrity, and (2) {

the use of basaltic concrete in the floor of the l lower drywell.

GE: The details of the lower drywell design are i important in'the response of the ABWR containment to a severe accident. Six key features are described below.

Sacrificial Concrete The floor and walls of the ABWR lower drywell include a 1.5 meter layer of concrete above the containment liner. This is to insure that debris will not come in direct contact with the containment boundary upon discharge from the reactor vessel. This added layer of concrete will protect the containment from possible early l failure.

Basaltic Concrete The sacrificial concrete in the lower drywell of 4 l the ABWR has been constructed of low gas content i l concrete. The selection of concrete type is yet l another example of how the ABWR design has striven not only to provide severe accident mitigation, but to also address potential uncertainties in severe accident phenomenon. Here, the uncertainty is whether or not the core can be cooled by flooding l

the lower drywell. For scenarios in which the l

lower drywell flooder is unable to cool the core debris, the concrete type selected will result in a very low gas generation rate. This translates into a long time to pressurize the containment. This is important because time is one of the key factors in aerosol removal.

Conclusion:

GE/ Staff agree on what is important. GE agrees to address pedestal erosion capability.

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Lower Drywell Design  !

, Staff: It is important that the as-built plant have (1) a

~

1ack of any direct pathways by which water from the upper drywell (e.g. , from drywell sprays) can drain to the lower drywell, other than by overflow of the l suppression pool, (2) negligible probability of premature or spurious actuation of the passive flooder valves at temperatures less than 500 F or under differential pressures associated with 3

reactor blowdown and pool hydrodynamic loads, (3) a capability to accommodate approximately cubic 4

meters of water in the suppression pool before the pool overflows into the lower drywell, and (4) a reactor pedestal capable of withstanding an impulse loading of without loss of integrity. This '

is important because the contribution of fuel-coolant interactions or steam explosions is considered negligible in the ABWR design due to two factors: (1) a very low probability that the lower drywell will be flooded at the time of reactor vessel failure (0.3%), and (2) the capability of the ABWR reactor pedestal to withstand the loads associated with an ex-vessel steam explosion.

GE: Developed based on " features" that are important to" severe accident response:

(1) Solid vessel skirt

~

(2) Flooder design conclusions: GE could include discussion of overflow volume. GE I does not agree that discussion of maximum pedestal loading is important. Resolution of FCI issue is needed before insights can be developed. FRA results not pertinent to discussion.

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( .... . . . . - . . _ _ . . . - . . . - .- . . . . - -

%< l Lower Drywell Design (Sump)

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Staff

As part of the resolution of the issue of CCI, GE  !

l proposed a sump shield to prevent core debris from entering the lower drywell sump. The impact of  !'

lower drywell sunps or sump shields on severe accident progression was not considered in the PRA.  ;

. i l

GE: The lower drywell sumps.are protected such that a l

substantial amount of core debris will not enter. l This ma'ximizes the upper surface area between-the- r debris and the water and maximizes the potential to l quench the core. debris.  : l l

Conclusion:

GE could include that the failure of the sump j shield was not considered in the PRA. However; we  ;

believe that.the. shift in focus from the features i to PRA is not desired. j l

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Time of Fission ProductRelease issue:

What is included in the frequency for <8 hours and 16-24 hours in the time of release table (CEB92-41, Table COPS 2.1 ska Table 19.3-6)?

History:

Forwardedto GEinformally on April 1 Response: -

• The following table gives the grouping of the categories described above.

• Note that in reviewing the calculations I determined I double counted STC#5. The correct value for 16-24 hours, RO release is 7.7E-11 (not 1.1E-10).

!

• SSARmarkuppages attached i

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CEB4/26/17333

e 4

Time of Release for Dominant Sequences - Grouping of sequences I

STC# Case Name Release Mode Timing 8 LCHPPBR RD <8 hours 10 LCHPPBD DW <8 hours 12 LCHP00E DW <8 hours 30 SBRCPFR RD 16-24 hours 5 LCHPPFP DW 16-24 hours 18A* LCLPPFR RD 16-24 hours i STC#18 contains cases with both FA and PF, so the details of the CETs l were examined to determine the fraction of sequences which could fallinto this timing bin.

08of26/199377

ABWR m ei m s u4 i

Standard Plant l l

1 TABLE 19.3-6  !

i FREQUENCY OF FISSION PRODUCT RELEASE l

)

Time of Release

- Release Frequency I

l No Release 134E-07 i

t l

Release da Rupture Disk Release da Drpell

> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.1E-08 3.9E-10 l l

16 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.1E 7.~7 E . g1 3.6E-11  ;

i i 8 to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 0.0 0.0 l

< 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 73E-11 33E-10 l

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i 19.3 23a Amendment l

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i 23A6100AS Standard Plant __ an. 4 i

describes early containment failure and Subsec-tion 19E.2.4.4 discusses the associated release).

Subsection 19E.2.8.1.4 examines the benefits and risks associated with the inclusion of the containment overpressure protection system in the design. The

, analysis indicates a tremendous reduction in the fraction of volatile fission products released from the containment. Typically, the Csl release fracdon drops from on the order of 1% to a release fraction of IE.-7. On the other hand, there is a slight decrease in the time of release. However, the effect of the lower fluion product release dominates the net impact. This leads to a substantial decrease in risk.

' I 19.5.3 Alternate Definition of Cohtainment l j

Failure

{

In this PRA, containment failure has been inter-preted to mean failure of the containment function.

For calculational convenience, this has been taken te

mean doses greater than 25 Rem at 1/2 mile. It has been shown that the ABWR can meet the goal of 0.1 conditional containment failure prob' ability (CCFP) using this definition (Subsection 19.6.83).

The NRC staff has proposed an alternate defini-tion of containment failure, one independent of source term:

• Containment failure occurs when integrity as a pressure boundary can no longer be con-

. trolled.

This definition recognizes the containment function by permitting normal leakage, as well as acknowledging credit for suppression pool scrubbing in conjunction with a *last resort" controlled release path, while properly accounting for postulated gross structural failure. .

I Based on this pressure integrity definition, a new l 1

conditional containment failure probability, desig-nated CCI P PI, can be found. Examining the data in Table 19.3-6, the CCFP-PI is 91)05- Therefore, the ABWR meets the containment performance goal regardless of the definition of contaihn.ent failure.

L -- O.CC3 19.5-2 Amendment l

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Bypass DETSummary Question:

There appears to be a typo in Response 3 on page CEB92-41-13.Under pool bypass the 2% should read 20%.

History:

Bob and I exchanged voice mail several times but we have not yet actually discussed the response. Informal fax sent on April 1.

Response: ,

GE concurs. The text will be fixed in the SSAR (19E.2.1.5.3.3) as attached.

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J 0 84/26/173348 i

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ABWR 234aooxs  !

Standard Plant w,  !

l

2. Even for those low frequency cases with The sum of the frequency of pool bypass l significant CCt. radial erosion remains below the sequences with no drywell spray available is 7.4E-11; i l  !

' structural %dt of the pedestal. After 0.05% of all core damage events. Since this value is considerathn of uncertainties only 1.5% of the extremely low there is no impact on offsite dose.  !

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l sequences with significant CCI wil! suffer l

l pedestal fdure. Combining this conclusion with l

the first, only 0.15% of all core melt sequences i

with vessel failure willlead to additional drywell failures as a result of CCI.

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. 3. The time of fission product release is not l l

tipificantly affected by continued CCI. t 1  !

! 4. The fission product release is dominated by the I noble gasses when the containment. overpressure l j

protection system operates.This conclusion is  !

unaffected by assumptions on debris coolability.

Therefore, the offsite dose for sequences with rupture disk operation is not impacted by core i concrete attack.

l These conclusions would indicate that the uncertainties associated with CCI have an insignificant influence on the containment failure  ;

j probability and risk. .

r 19E.2.1.533 Pool Bypass  ;

9 i

Analyses performed in subsection 19E.2333(4) 9- indicate that the only significant mode of suppression  ;

M pool bypass occurs via the vacuum breakers.

73 Uncertainty analyses and sensitivity studies were ,

" [

performed to assess the effect of pool bypass on risk.

Some of the key conclusions of these studies are l summarized below.

i L The probability of a large leakage path between the wetwell and drywellis approximately 0.4%.-

p 2.c'/s

2. There is aia% probability that there is a small leakage path between the drywell and wetwell.

Based on the Morowitz plugging model,90% of these sequences are expected to plug before the rupture disk setpoint is reached. In sequences with plugging, there is no significant increase in the time of fission product release or in offsite dose.

t

3. Use of the firewster spray system can prevent early opening of!be rupture disk for a bypass path of any size.

I 19E.2 9 4 Amendmeat

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. f CET submittal Question:

^

The complete CETsubmittalis needed in SSAR form History:

The CETsubmittal was sent in draft form in June 1992.

Response

A complete copy of the CETsection for the SSAR was sent to the staff on April 23 (formal Amendment will be submitted later) i 9

TB4fi6/199329

Q 1

STC#15 grouping Question:

Why is DWHEAD release (STC#15) grouped in Case 1 along with the rupture disk.

History:

Bob and I exchanged voice mail several times but we have not yet actually discussed the response. Informal fax sent on April 1.

, Response:

• The basis for grouping STC#15 with Case 1 is provided in the note to Table 19E.5-7.

• STC#15 is a transient induced core melt with in-vessel recovery, therefore all of the release will be scrubbed as it flows through the SRVs to the suppressionpool.

i

• Further clarification will be added as shown on the attachedpage.

crewsmna l

. ABWR m .i m , j Standard Plant mA a

Table 19D.5-7 54 i BINNING OF CONTAINMENT EVENT TREE RESULTS E j-4 .- ,

i Ste # Deterministic Bin Consequence Bin  ;

I NCL NCL  ;

4 NCL ~NCL 5 LCHPPFP Case 7 6 LCHPFSR Case 1 g LCHPPBR Case 8 See Notes 10 LCHPPBD Case 7 See Notes  ;

' ~

12 LCHP00E Case 8 ]

NCL NCL i 13 2

14 LCLPFSR Case 1 See Notes 15 LCLPFSR Case 1 See Notes l I

16 NCL NCL 18 LCLPFSR Case 1 See Notes l

)

19 LCLPFSD Case 7 l 1

d 21 LCLPFSD Case 7 25 NCL NCL ,

l 26 LCLPFSR Case 1 See Notes  :

28 NCL NCL 30 SBRCPFR Case 1

)'

37 NCL NCL 38 LBLCFSR Case 1 See Notes 40 NCL NCL 1

Notes-Sequences 8 and 10: Releases taken from suppression polbypass study in Attachment 19EE.

Sequence 14,26 and 38: Sequence is arrested in vessel indicating high probability of the use of the firewater addition systern. e.ssa.1 a, ..dar a.f.

Sequence 15: This sequence is binned with those which have releases through the rupture disk since any fission products Mreleased from the vessel will be scrubbed through the suppression pool.

Sequence 30: No credit taken for firewater system since a long time was available to prevent core damage but the operator failed to do so.

19D.5-20 wn I

Consequence Bin 7 Question:

.i Case 7seems to be a mixed bag. Is Table L2-1 (Table 190.5-7) correct?

History: ,

Bob and i exchanged voice mail several times but we have not yet actually discussed the response. Informal fax sent on April 1.

Response

• The binning for the consequence analysis was performed based on the characteristics of the release.

• Any impact of the in-plant response is considered before this final

binning of the deterministic sequences for consequence analysis.

• Therefore, considerable variation in the deterministic sequences is l possible.

i CEBV26/1993-31 i

_ _ __ _ y Source term ba.nning frequencies r

Question:

There appears to be a discrepancy in the CETbinning frequencies for -

Cases 7 and 8 in Careway Table 1-1 (Table 19E.3-6) and the end states l in Figure 2 o ' CER92-39 (Figure 190.5-3) using the binning described in

> Table L2-1 of CE892-41-23 (Table 190.5-7).

History:

Bob and I exchanged voice mail several times but we have not yet actually discussed the response. Informal fax sent on April 1.

Response

it appears that there may be a discrepancy in the total frequencies for these categories in the range of E-11. GE will resolve any error.

l CEBofiSM73342 E_---_-_-.__--_---_-______-_---____-_-_--- -_ _ - _ _ _ _ - _ _ _ _ - - - _ _ _ _ _ _ . .

.. .. - ... - . . --- .= _ . = .. . - - - - . - .

Use of firewateraddition system Question:

What is the finalphilosophy regarding the use of Firewater for spray as opposed to vesselinjection (CEB92-54-10) and changes being made to EPGsin Appendix 18A. .

History:

Transmitted by iisformal fax on April 1

Response

The philosophy stated in CEB92-54 remains unchanged. The EPGs will be

}

modified to state that the containment flooding will be accomplished

via vesselinjection unless high dryweII temperatures are indicated.

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. . .-. .. - - - . - . -. .- -_. - - . . .-= . - -__ _

.l E0Pissue Issue:

• PRA assumes RCIC runs for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If this occurs, the suppression pool may heat up to the point where the EPGs call for reactor vessel depressurization, which would cause the loss of RCIC.

• PRA andEPGs need to be consistent.

History:

This issue was raised in a 3/12 letter from NRC.

Approach:

• GE is reviewing bases for EPG containment related limits to confirm that long term RCIC operation is justified.

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Cl84RW1993-34 4

TWissue issue: -

AII ECCS pumps are not designed to pump water at suppression pool temperatures which may occurin a TWsequence (core cooling OK, containment cooling not OK.)

History:

Issue raised in 3/12 letter from NRC. Discussed in 3/31/93 meeting.

Response

GE opinion is that manual containment venting and normal cold suction source make the ECCS capability OK.

GE is considering increasing the HPCF design temperature capability.

i CEBofi6/173335

.. s, l

SEVERE ACCIDENT ANALYSIS INPUT TO TIER 1 AND 2 P:ITAAC i Recommendation: GE should review, as a minimum, the following systems and  !

indicate the corresponding sections of the SSAR which will be included into Tier 4

2 of the Design Control Document and which sections of the SSAR addressing these '

systems will be deleted. In addition, GE should provide an appropriate "roadmap" as to where the important aspects of the systems discussed within the Tier 2 information, have been incorporated into Tier 1.  ;

SYSTEM SSAR ONLY TIER 2 TIER 1 l j i ACIWA  ? 5.4.7.1 -injection to vessel l (19.2) 5.4.7.1.1.10 -containment sprays DW & WW

-deliv ry to Ser. Lev.C

• .055 /sec at runout

.044 /sec at Ser. Lev.C ADS  ?  ? -reliable for HPME ,

(90-016) -battery supply  !

-Nitrogen supply Batteries  ?  ?

) CTG  ?  ? -diverse design

(90-016) -one safe-shutdown division Concrete  ?  ? -protect liner and pedestal (93-087) -production of non-condensible gases

! COPS  ? 6.2.5.2.6 -dual rupture disk (90-016) 6.2.5.3 -burst pressure j 6.2.5.4 -material

-re-closure isolation valves DW Area  ?  ? -final area, with no major (93-087) obstructions ,

DWPressCap ?  ?

(19.2) ,

DW/WW Vents ?  ?

FireProt  ?  ? -alternate shutdown capability l

.: (90-016) -features to prevent migration of  !

smoke, hot gasses, or fire suppressant  ;

InertCont  ?  ? -atmospheric control system for (90-016) inerting ISLOCA 7  ? -upgraded piping design '

(90-016) -capability for leak testing

-valve position indication

-high pressure alarms 4

, - - -., . __ , , , , _ _ , , _,. . _ _ . _ ,m.. ,.

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l LeakRate  ?  ? -most severe accident leakage in this j range -

t LDFlooders ? 9.5.1.2 -cavity flooding (93-087) -flow rates

-design of fusible plug

-testing / qualification of plug l

l Penetrat.  ?  ?

RecircTrip ?  ? -auto trip and runback of pumps l (90-016) i Rx.Vesskirt ?  ? -solid to prevent debris ejection to I (93-087) upper drywell and water migration from l UD to LD l l

i' ScramSystem ?  ? -hydraulic (90-016) -electric run-in l SLCS  ?  ? -automatic initiation i i

(90-016)

Sprays  ? 5.4.7.1 -WW and DW .

l (93-087)  ? 5.4.7.1.1.6 -supply from ACIWA SumpDesign ?  ?

(93-037) .

SupPool  ? ?

VacBreak  ? ? -limit swithes for .9 cm clearance (93-087) l l

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.-