Text

Summary of 900716 Meeting W/Epri Re Hydrogen Generation & Containment Performance.List of Attendees,Nrc Slides & EPRI Presentation Encl
	ML20059C672
Person / Time
Issue date:	08/23/1990
From:	Kenyon T; Office of Nuclear Reactor Regulation
To:	; Office of Nuclear Reactor Regulation
References
	PROJECT-669A NUDOCS 9009050256
	Download: ML20059C672 (68)
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.f August 23,19' 0 9

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Project No. 669 APPLICANT:

Electric Power Research Institute PROJECT:

ALWR Requirements Document

SUBJECT:

SUMMARY

OF MEETING TO DISCUSS HYDROGEN GENERATION AND j

CONTAINMENT NRFORMANCE WITH EPRI On July 16, 1990, representatives of the imC and EPRI met to discuss the staff's conclusions on hyPogen generation and containment performance as it pertained to the EPRI ALWR Nequirements Document and its reviews of the evolutionary light water reactors (LWRs). is the list of attendees. is a copy of the slides presented by the staff in their discussion of hydrogen generation and deflagration. is a copy of EPRI's presentation to the staff on containment performance criteria.

The staff stated that although it had already given many of the reasons for its position on hydrogen generation and control during the May 30, 1990 meeting with EPRI, it would provide additional insights to its positions by discussing some of the studies that have been performed to support NUREG-1150 efforts.

During the meeting, the staff _noted that there was no one piece of information that led it to conclude that evolutionary plants should comply

.with the requirements of 10 CFR 50.34(f). Rt.ther, this conclusion came from evaluation of an accumulation of information from experiments and analysis and the uncertainty in our understanding of the physical phenomena associated with severe accidents. The staff stated that 10 CFR 50.34(f) does not represent an I

upper bound, but is a good representation of whr.t the staff has concluded J

might be produced during a wide range of severe accidents.

1-During the meeting, the participants discussed hydrogen generation, combustion and control, containment performance criteria, and the ABWR vent. Enclosure 4 provides a suneary of the technical discussians. The staff agreed to provide EPRI with some feedback regarding input it has received in its evaluations of the hydrogen generation and detonation issues.

i.

I Odo!nal59ned By:

Thomas J. Kenyon, Project Manager Standardization Project Directorate Division of Reactor Projects - III, IV, V, and Special Projects

Enclosures:

As stated cc w/ enclosures:

d See next page

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DOCUMENT NAM ' b UMMARY HYDR 0 GEN G&C pM82 Milik NRC Hl.E CENTER COPY

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UNITED STATES 5

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g NUCLEAR REGULATORY COMMISSION y5 j-WASHINGTON, D. C. 20555 g

August 23, 1990 j

' Project No. 669

[

e APPLICANT:

Electric Power Research Institute PROJECT:

ALWR Requirements Document

SUBJECT:

SUMMARY

OF MEETING TO DISCUSS HYDROGEN GENERATION AND CONTAINMENT PERFORMANCE WITH EPRI On July 16, 1990, representatives of the NRC and EPRI met to discuss the i

staff's conclusions on hydrogen generation and containment performance as it pertained to the EPRI Al.WR Requirements Document and its revien of the evolutionary light water reactors (LWRs). is the list of attendees. is a copy of the slides-presented by the staff in l

their discussion of hydrogen generation and deflagration. is a copy of EPRI's presentation to the staff on containment performance criteria.

The staff stated that although it had already given many of the reasons for its position on hydrogen generation and control during the May 30, 1990 meet' g with EPRI, it would provide additional insights to its positions by-discussing some of the studies that have been performed to support HUREG-1150 efforts. During the meeting, the staff noted that there was no one piece of information that led it to conclude that evolutionary plants should comply withtherequirementsof10CFR50.34(f).

Rather, this conclusion came from i

evaluation of an accumulation of information from experiments and analysis and the uncertainty in our undtrstanding of the physical p(henomena associated with severe accidents. The statf stated that 10 CFR 50.34 f) does not represent an upper bound, but is a good representation of what the staff has concluded might be produced during a wide range of severe accidents.

During the meeting, the participants discussed hydrogen generation, combustion and control, containnent performance criteria, and the ABWR vent.

provides a summary of the technical die assions. The staff agreed to provide EPRI with some feedback regarding input it has received in its evaluations of the hydrogen generation and detonation issues.

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^

Th as J. K y o

t Manager L

Standardization Project Directorate i

Division of Reactor Projects - III, ll IV, V, and Special Projects

Enclosures:

As stated cc w/ enclosures:

See next page l

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k ALWR Utility Steering Committee EPRI Project No.-669 cc:

Mr. E. E. Kintner ALWR Utility Steering Committee GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054 Nr. William Sugnet Nuclear Power Division Electric Power Research Institute P.O. Box 10412 Palo Alto, CA 94303 L

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August 23,.1990

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5 NRC PDR 4

FMiraglia, 12G18 JPartlow, 12G18 f

DCrutchfield, 13A2 WTravers, 11B19-PDST Reading TKenyon-OGC, 15B18 EJordan,'MNBB3701 JKudrick, 801 BHardin', NLS169 JLyons 8

'CNichols,D1 8D1 DScaletti IYoshida THsia FEltawila, NLN346 KParczewski, 704 PWorthington, NLN344 PWright, NLN344 PShea.

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ENCLOSURE 1 Meeting Atiendees July 16, 1990 11, Generation &

Control Name Organizatior, Thomas K nyon NRC/NRR/PDST Dave Leaver Tenera Steve Additon Tenera/ARSAP Bob Hammersley FAI/ARSAP Martin Plys FAI/ARSAP Stan Ritterbusch Combustion Eng., Ir.c.

John Trotter EPRI Jack Kudrick NRR/SPLB Brad Hardin RES/ARGIB Lyle Gerdes ABB-Combustion Eng.

Jim Lyons NRR/SPLB Charles Nichnis NRR/SPLB Dino Scaletti NRR/PDST Isao Yoshida NRR/PDST Tony Hsia NRR/PDST Fhrouk Eltawila RES/AEB Kris Parczewski NRR/DET/EMCB Douglas Stamps Sandia NL Pat Worthington RES/EAB R. T. Curtis AECL-Tech R. W. Wright RES/EAB Dave Blanchard Tenera Trevor Pratt BNL

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ETICLOSURE 2 i

l BROOKHAVEN NATIONAL LABORATORY i~

i DEPARTMENT OF NUCLEAR ENERGY SAFETY AND RISK EVALUATION DIVISION l

XYDM@GHE GREMM&TUOM BUiMDEG &

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SHVHMH ACC:DHETS DE LWMs wet.PMATT l

PRESENTED TO:

i U. S. Nuclear Regulatory Commission I

July 16,1990 g

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OUTLINE

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Background

Analytical Models Experimental Basis

% CP Applications f

MELCOR Applications Uucertainty Estimates Summary l

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BACKGROUND i

Aim is to reduce likelihood of early containment failure from hydrogen combustion l

Options:

l Provide sufficient containment design margin (volume, design configuration, and ultimate pressure j

capability) to withstand effects of hydrogen burns and maintain global and local concentrations belo'w detonable concentrations, or Provide fo~ controlled igniting which maintains globa!

r and local hydrogen concentrations below detonable concentrations and controf,s hydrogen burning such that containment intepi::y is maintained, or l

l Provide for containment inerting that prevents l

oxygen concentrations from exceeding 4%

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7 BACKGROUND JCont.)

Keyconsid&Faians:

Amount and rate of metal-water reaction (MWR)

Detonable concentrations (global and local)

Powersupplyif'gnitorsare used Equipmentsurvivability EPRI ALWR docannsdss-;-;6s:

MWR 75%

l H2concedu4i00 <13%

10 CFR 50.34(f) calls for:

MWR100%

H2 concentration <10%

Should 10 CFR 50.34(f) H2 requirements be applied to evolutionary LWRs?

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ANALYTICAL MODELS Integrated Risks Codes:

STCP Sponsored by NRC MELCOR Sponsored by NRC MAAP Sponsored by EPRI l

Mechanistic Separate Effect Codes:

SCDAP Sponsored by NRC MELPROG Sponsored by NRC A

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,3 ANALYTICAL MODELS (Cont.)

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STCP:

l Extensively used by NRC in PRA and issues Related to H2 Rule

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Now being replaced by MELCOR j

i MELCOR:

i, MELCOR 1.8.0 Released in February 1990 Large User Group i

Peer Review, Technical Assessment and I

Application Underway

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.l EXAMPLE OF BENCHMARKING ANALYSIS WITH 1

EXPERIMENTAL DATA AND MORE MECHANISTIC CODES i

Compare with results of PBF SFD 1-1 TEST i

l Previously analyzed by SCDAP CODE (mechanistic) i Also analyzed by STCP code (risk code) l l

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l NRC sponsored, under SFD prograin j

Conducted in power burst facility at INEL a

Large scale severe fuel damage experiment i

Transient initiated by coolant boiloff and bundle uncovery i

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Fuel was trace-irradiated 1

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l BRIEF DATA 32 zircaloy-clad UO2 fuel rods Active fuel length 0.914m Initial two-phase level 0.117m Nominal inlet flow 0.64-0.67 g/s Pressure 67.8 MPa m

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TEST DESCRIPTION

.High density ZrO cytInders Low density ZrO fiber 2

2 Pressure regulator Bundle coolant line 6/ ' 'Q Inlet lines rO DWe bypass tube 2

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CROSS-SECTIONAL VIEW OF THE SHROUD AND BUNDLE

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COMPARISON OF CALCULATED AND MEASURED.TWO-PHASE WATER LEVEL DURING TESI P

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RESULTS HYDROGEN GENERATION 0.09 i

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EXPERIMENTAL BASIS l

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All Codes (STCP, MELCOR, and MAAP) can predict H2 l

Generation during Early Phase of Core Heatup and i

Degradation i

Lack of Prototypical Experimental Dat$ for late Phase In-Vessel Meltdown l

Largest Uncertainty in Predicting H2 Generation During Core Relocation into Lower Vessel Head and Vessel Penetration a

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i STCP APPLICATIONS

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l 1-5) in Support of NUREG-1150 i

Comparison with MAAP Analysis in SP/90 PRA 4

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Table 4.7 Sunrnary of MARCH Results and Comparison with HCOG Results

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Case No Reflood

' 300 gpm at 42 min 5000 gen at 56.7 min Code MARCH HCOG MARCH HCOG MARCH-IlCOG Transient Time, min 70 70 72 72 61 61 4

Total H Production, 1b 524 478 1310 585 412 416 2

4

-Peak H Generation 4

2 Rate,Ib/ min 28 23.6 100 76 151 308 Time, min 50 52 52 53 58 57 4

0xidation, t Claddins *1 19 13.5 24 16.5 8.5 11.8 Chanr.el Box 5

21 2.9 Control Blade 1

12 3.1 I

Melt %

Core *2 29 54.7 0

29 0

29 Channel Box 86 24 0

j Control Blade

- 93

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1 - HCOG: Active zircaloy including both cladding and channel box.

2 - MARCH: Fuel rod with eutectic melting temperature given as 4130*F, 75rcaloy melting temperature 3365*.

IICOG:

Active zircaloy with melting temperature given as 3451*F.

- Reproduced f rom BNL Technical Report A-3808 and A-3809, dated September 1985.

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Table 2.6 MARCII in-Vessel N Predictions 2

BWR Mark I, NUREG/CR-4624, Volume 1 Metal Water Accident Sequence Seauence Description Reaction (MWR)

TC)

ATWS, failure to reduce power level, containment failure

.26 TC2 ATWS, failure to reduce power level, injection systems fail

.58 TC3 Same as TC2 but with wetwell venting

.58 TB1 Blackout, injection fails when batteries depicte core melts

.25 with containment initially intact, containment fails later TB2 Same as TBl. but containment fails when vessel fails

.26 V

Interfacing system LOCA through low-pressure ECCS into

.33 reactor building

f Table 2.6 (Continued)

MARCH In-Vessel H2 Predictions FWR Ice Cortd., NUREG/CR-4624, Volume 2 i

Metal Water Accident Reaction (MWR)

Scauence Sequence Description S311F RCP seal LOCA, failure of injection and containment spray

.74 recirculation S3B Station blackout. RCP seal LOCA. battery depletion

.75 TMLU Transient, loss of secondary side cooling, failure of injection

.43 to primary TBA Station blackout, no RCP seal'LOCA, battery depletion

.69 4

9

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Table 2.6 (Continued)

MARCH In-Vessel H2 Predictions PWR Subatmospheric, NUREG/CR-4624, Volume 3 Accident Metal Water Sequence Scauence Description Reaction (MWR)

AG Large break LOCA, failure of containment heat removal

.71 BWR Mark III, NUREG/CR-4624, Volume 4 I

i Accident Metal Water Scouence Scouence Description Reaction (MWR)

'IC ATWS

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TB Station blackout, injection fails when batteries depicte

.32

'mS Blackout with immediate injection failure, operator

.39 depressurizes when core uncovers e

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o Table 2.6 (Continued) l MARCH In-Vessel H2 Predictions PWR Large Volume, NUREG/CR-4624, Volume 5 Accident Metal Water Scauence Scauence Description Reaction (MWR)

S DCa RCP seal LOCA, failure of injection and containment spray

.47 2

recirculation, fan. coolers fail at vessel rupture S2 0 D

Same as S DCR, except fan coolers fail initially

.47 2

TMLU Transient, loss of secondary side cooling, failure of injection

.52 i

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4 18 SP/90 Table 4.1 Comparison of Major Events for TE Base Case **

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MAAP STCP 1

Steam generator dryout, min 141 132 Core uncovery, min 194 222 Start melt, min 284 Support plate fails, min 355 327 Reactor vessel fails, min 356 343 1

Containment fails, min no failure no failure Cavity dryout, min 81 0 1243 i

Hydrogen burn none 5 in the icwer compartment 1 in the upper compartment Peak containment pressure, psia 56 59

(

Pressure spike at vessel failure, psia 30' 27 l

l Cavity axial penetration, f t 0.831 8.2 Cavity radial penetration, f t 8.2

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% Zr reaction in-vessel 5.6 13.3 iotal H2 in containment at 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,1b 401 2758 l

l Total gas release from corium/ concrete interaction at 48 hr, 1b l

l C02 17,480 C0 85 H2 t

3400 11 0 2

68,270 1

Not Available

- Reproduced from BNL Technical Report A-3805, dated 1985.

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Table 4.7 Coiparison of Major Events for AE Base Case **

MAAP STCP Steam generator dryout, min 1286

8 J

Core uncovery, min 141 96 1

1 Start melt, min 151 Support plate fails, min 293 199 Reactor vessel fails, min 294 221

{

i e2 Containment fails, min No No I

1 Cavity dryout, min 294 282-

% Zr reaction in vessel 8.8 15 j

c Hydrogen burn in containment No No i

Peak containment pressure, psia 64 110 3

Pressure spike at vessel failure, psia 57 48 Cavity axial penetration, f t 4.75 8.1 Cavity radial penetration, ft 7.4 Total H in containment at 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , Ib 2727 (3450)*1 2

4097

-Note:

1 2)) No containment failure within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,2727 lb in Table 5.5.2-AE-3, 3450 lb in T potential failure at 51 hrs (Based upon extrapolation of pressure rise).

3) Not available.

' ** Reproduced Irom BNL Technical Report A-3805, dated 1985.

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l Provide benchmarking analyses with experimental data and more mechanistic codes Apply MELCOR to simulation of accident sequences l

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~

TIME (hr)

.l i.

I 4

4 L

o.

i l t.

4

=: -

?-

7

.;g UNCERTAINTY

~

NUREG-1150 Performed Extensive Uncertainty Assessments for Various Areas of Core Meltdown Phenomena Extensive Use of Expert Opinion Experts Drawn from National Labs, Industry, Universities, Etc.

Expert Panel specifically for in-Vessel H2 Generation bni aui

y NUREG-1150 Uncertainty Ranges for 17 PWR Hydrogen Generation

z, (in % Zr Oxidized)

Cum. Prob.

1P4 SP4 25P4 50P4 75 %

95P4 99P4 a

Case 1a 6

10 28 39 50 96 125 Case 1b 8

20 34 45 62 93 125 Case 1c 6

10 28 29 52 96 125 Case 2a 7

14 23 30 39 57 73 Case 2b 8

17 27 36 47 64 78 Case 2c 7

15 24 30 40 57 73-Case 5 10 16 24 29 41 63 81 Case 3a 12 18 27 40 64 97 114 Case 3b 12 20 34 48 65 97 114 Case 4 11 15 24 33 63 97 114

Reproduced from NUREG/CR-4551, Volume 2 to be published.

bnl aui' 1

.c f.1 i

SUMMARY

c-

~

Given Uncertainty in:

I Accident Progression j

1 Core Meltdown Phenomena j

1 Ex-Vessel Coolability l

100% of Active Clad is a Reasonable Estimate for H2 o

Generation which Envelops Wide Range of Potential i

Severe Accidents a

e f

i b

i

!l'i l

bnl BUI t

,o - Ac:

EXPERIMENTAL DATA.0N HYDROGEN GENERATION (ZIRCALOY OXIDATION FRACTION)

TEST HIGH POWER HOLD (MIN.)

Z OXIDIZED (%)

UNCOVERED

Zgy OXIDIZED RY Z

0xidized (%) after Z Melt RY ry PBF SFD l-1 4.

28.

.85 SFD l-3 4.

26.

SFD l-4 4.

38.

>.95 NRV FLHT-2 4.

10.

15.

.90 FLHT-4 30.

68.

90.

>.95 FLHT-5 50.

86.

~100.

>.95 LOFT FP-2 58.

TMI-2 45.

195 Table 5-6 PWR Hydrogen Generation (In 4 Zirconium Oxidized) c Aggregate f

Cumulative Probability

i 1%

5%

25%

50%

75%

95%

99%

Case la 6

10 28 39 50 96 125 Case Ib 8

20 34 45 62 93 125 Case 1c 6

10 28 39 52 96 125 Case 2a 7

14 23 30 39

,57 73 Case 2b 8

17 27 36 47 64 78 Case 2c 7

15 24 30 40 57 73 Case 5 10 16 24 30 41 63 81 Case 3a 12 18 27 40 64 97 114 Case 3b 12 20 34 48 65 97 114 Case 4 11 15 24 33 63 97 114 i

Cumulative Probability refers to the first row in the table.

The numbers I

below that row are the values of the parameter defined in the-table heading.

i i

i L

w er

-.7=

.r,

.e

~

50%

90%

10%

Source of H Prob.

Prob.

7 Early Phase (SCDAP) 0.38 0.50 0.30 Late Phase 0.10 0.20 0.05 Melt-Water 0.08 0.15 0.05 Stainless Steel 0.09 0.25 0.00 i.

l Total Hydrogen 0.65 1.10 0.40

- l-i l

l O

l i-__..___,__

c:v s

Hydrogen Concentration Limits May Be Established by Three General Criteria D_elonation Data and Analysis Heated Detonation Tube (HDT) Detonability Limits Data Detonation Cell Size Data / Analysis Flame Acceleration Data Potential for Deflagration-to-Detonation Transition

'(DDT)

Potential to Generate Detonation-Like Impulsive Loads Potential to Exceed AICC Pressures Flammability Limits Data n

6 I

Hydrogen-Air-Steam L

HDT Detonability Limits Data (P = 1 atm, T= 100 C) l Equivalence Hydrogen Steam Detonation Ratio Concentration Concentration 0.231 8.8%

0%

No 0.248 9.4%

0%

Yes 1.02 17.8 %

40.5 %

No 1.01 18.2 %

38.8%

Yes An estimate of the detonability limits for other steam q

concentrations can be obtained from a linear interpolation of the HDT data i

For example, the detonability (NO GO) limit for 20%

steam is estimated to be 13.2% H in the HDT 2

Detonability limits widen with increasing scale. How do you extrapolate HDT detonability limits to reactor scale?

E

r, Detonation Cell Size i

Data / Analysis l

-NAS Method (1987)

Extrapolate detonation cell size for lean hydrogen-L air mixtures to reactor scales using the ZND model Calculate a threshold tube diameter (d ) and a e

minimum cloud diameter (d ) using empirical s

correlations 1

Compare de and d to an assumed reactor 3

containment diameter of 50m i

NAS Recommendation Ensure "that any mitigation strategy should avoid the attainment of hydrogen levels greater than 9 to 11 percent."

Shenherd Method (1990)

Assume that a detonation is possible if the cell size is less than two channel widths Assume a representative channel width of 5 m Extrapolate the detonation cell size using the ZND model and compare to the channel width Shenherd Recommendation Maximum hydrogen concuatration should be limited to less than 10% H by volume 2

Mechanisms / Conditions

-Known to:Cause Detonations High explosive of sufficient strength 6

Turbulence generated by fans, etc. -

e Obstructed geometries (obstacles)

Hot jets Shock focussing in convergent geometries (corners, edges, etc.)

Rapid venting Long run in smooth tubes M

r

N Hydrogen-Air-Steam Flame Acceleration Data Potential to o,enerate detonation-like impulsive loads Accelerated flames in obstacle-filled ducts can produce impulses that are comparable to those produced by CJ detonations.in smooth ducts Impulse values increase fairly linearly with increasing mixture sensitivity even though the peak velocity can be discontinuous Potential to exceed AICC pressure

2000 <

~

o

'~-..,.,,-~r-Q detonation velocity 1

- Isobaric sound speed

-=

1500 -

,,E~

~~~ ~

-1 T

e 1

n en N

..,k -

4. = 1.0 E

a

-a../

m 0

4 = 0.5 I

t-

.1000 -

O

=

T O

1 HJ i

500 -

i

~

i j

=,

l 1-------g------......___

l 0

3 1

0 10 20 30 40 50 60 i

%HO 2

l Maximum Flame Velocities in H -Air-Steam Mixtures

~

2 P = 760 mm Hg, T = 110*C i

t 432 mm Dia. Tube, B.R. = 0.30,- (L/D)u = 1.0

(--o-Data of N. Brehm, 66.6 mm Dia. Tube, B.R. = 0.29, (L/D)

= 1.5) i

,. l

-m.

m

.m a--Oa c-.

.A h.

.,,w

.miu-g=5

~

,I

,,I

,,,1

,,,f 70U

.n s

x 4

N N

N.

D N.

Nx N

N..

x'N.

N..

1 4 = 0.5 N

N N.,

N "s.

^*

500 -

+ - to E,

(

.O_.

u) 4\\.,_~~~.._, -

O, w

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

l g

N (f)

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

y 250 -

~ ~ ~~

\\

'~

a.

x 2

Ne

.N N

Data Fit

~

C-J Detonation, S!de-On Impulse C-J Detonation, Reflected impulse o

,...,.v.

0 10 20 30 40 50 60

%'H O 2

Impulse Values From H -Air-Steam Mixtures 2

6 ms of integration. 7.81 m from igNtion Point, T = 110*C i

P = 760 mm Hg, 43.2 cm Dia. Tube. B.R. e 0.30 (L/D)a., = 1.0 i

i

,s

-~;

u llm W

is i

e aull l

p i-M u

ii ei i

n iiq u

i lli iu e q, ii e

i i i

l i.s si, ua e

iq lmI i

ll sie i

g, l

ls Equivalence Ratio,4 0

.1 2

.3

.4 5

6 2000 ib Linear Regression to 25'C Data

- CJ. Velocity, Ti = 25'C

- CJ. Velocity, T = 110*C o

Isobaric Sound Speed, T = 25'C 1500 - ---Isobaric sound speed,T - tio'c

~

'y i

ig#g'

^

a Data, Ti = 25'C 0

Data, T = 110*C v

o T

1000 -

I O

I 0

, s. =

g

~'

500 -

9 o.

.=

0 0

5

- 10 15 20

% H2 Maximum Flame Velocities in Lean Dry H -Air Mixtures 2

P, = 760 mm Hg, 43.2 cm Dia. Tube, B.R. = 0.30, (UD)u = 1.0 (o Data of N. Brehm, 66.6 mm Dia. Tube, B.R.=0.29, (UD)m=1.5)

Equivalence Ratio,4

[.;

800

- C-J Det Side-On impulse, T = 25'O i

C-J Det Side-On impulse, T = 110*C i

600 -

^

Experiment, T = 25'O

=

i g

_a

---+--- Experiment, Ti = 110*C m

a

... 4 '

400 -

til a)

,. -P,. /,.-

=

j f

a.

/

J 200 -

e O

O 0

5 10 15 20

% He impulse Values for Lean Dry.H -Air Mixtures 2

Pi =.760 mm HQ,.43.2 cm Dia. Tube, B.R. = 0.30, (11D)

= 1.0

.,e e

f

,t i i t

I i

1 4 ~ t f

f

't O.8 c

IMPOSSIBLE P

MIXTURES 0.7 Il o

~

's,

~

g

+

~

O 4

0.5 -

.,S

<C T

o 4

0.4 -

LLJ

R J

'e.,

o g

g 0.3 -

g,

~

.O 0.2 -

/*

"4 i.o 9

9 g

!"....,,,,....,,a

.' s Q

p

3.,

'~.. -

3 0.1

-s..

FLAggABLE "D0> 8~...

~

NONFLAMMABLE 0.0

...i....,...

i i..>

i i>

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 STEAM MOLE FRACTION FIGURF 16. 30% Blockage Ratio Ring / Disk Obstacle Test Results Deton tlon cell widte. \\. obtaine d from ZND theoretical calculations, e

Observod DDT limit E 00 served dynamic pressure limit a Propagation down entite length of tube, V propagation part way Hydrogen rich data at 3 atm.; lean and stoichiometric at 1 atm.

Ref.1: B. Marshall Jr.. Ref.10: R. Kumar 4

-s

1 0.8 u

lMPOSSIBLE MIXTURES 0.7 -

H 0.6 -

'o '

)

g Z

?

o- 's -

4 3 1

O

~

%*,? y 4

\\

] 0.5 ;

,'g

(

'o 4

g a

[1] 0.4 -

i

,,#p J

O y

l 0.3 -

7

- " ~ ~.

tyC Rg

'o.,

~

L.U Der tyg

~

om 0

%.o x

.2 -.

Q

r.

3

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

y 0.1-

_ _ ' ~... "

., ~.. '

ptAMMABLE

'" N S...,,

I I

NONFLAMMABLE 0.0 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 STEAM MOLE FRACTION FIGURE 17. 60% Blockage Ratio Ring / Disk Obstacle Test Results Detonation cell width, A, obtained from ZND theoretical calculations.

Observed DDT limit 8 Observed flame acceleration limit a Propagation down entire length of tube, V propagation part way initial pressure = 1 atm. except as noted.

Ref.1: B. Marshall. Jr.. Ref.10: R. Kumar

Summary of Hydrogen-Air-Steam HDT Flame Acceleration Data Potential to generate detonation-like impulsive loads Hydrogen Steam Comments 11%

0%

Limited HDT data at elevated temperatures 13.5 %

20%

Estimate from linear interpolation 13.7 %

21%

HDT data Potential to exceed AICC pressure Hydrogen Steam Comments 9%

0%

HDT data 10.9 %

20%

Estimate from linear interpolation 11.9 %

31%

HDT data Caveats on the use of HDT data for reactor conditions Limited data are available at elevated temperatures Experiments indicate flames will accelerate over a wider range of mixtures as the geometric scale increases Results are only for two obstacle configuations

EFFECT OF TEMPERATURE ON DETONABILITY LIMITS OF HYDROGEN-AIR MIXTURES 9 DETON AtttlTY LIWit (3)

T.,

AUTO.10 NITION LlWly a DETON A.lLITY LIMIT (16) 4..*

6.. -j LOWER FLAWW AtlLITY LIMIT i

(V'W ARD PROP AG ATION)

^

4..

(DOWNW ARD PROPAG ATION)

UPPER FLAWW AtiLITY Lluff 3

,i s.

i g...

1 r

a 1.C l

le a

a j

1..

s i

3...

i.

HYDROGEN CONCENTRATION (VOLUME PERCENT)

Tlie range of detonable concentrations widens with increasing scale and temperature (11.7%-75% at 20 C to 9.5%-77% at 100 C)

.=

'l HYDROGEN-AIR-STEAM FLAMMABILITY LIMITS DATA 100 i

H2w REF (15), T=110 C, P=83 kPA O

L a~

REF (9hT=100 C, P=20-51 kPA-w l

1 80 REF (9), T=200 C, P=20-51 kPA R E F,,(1,6) g T,=,14,9,,g,,, P =1,p,,k,P,A,,,

1 3

R_ _E_F_ (1,6), T_=_1_4_9_ C_ _, P_=7_ 8 8_k_P_A_ _

s g

- s

')

00 z

qss O

N'4N l

h s

N'sN z

40

'%N m

9 Oz N

l oU NN h-20'

\\sN\\

\\

O N

L

/

x

?

g=_. -. J uh-= &+.= E=== -= - =O O

0 20 40 60 80 STEAM CONCENTRATION (VOLUME PERCENT)

Hydrogen concentration at the downward propagation lower flammability limit is 8.8% H for both 0% and 2

20% steam E

e+ @h B

O

////f

$g IMAGE EVALUATION 4

9+

9f77 g; p/

'k

"# y TEST TARGET (MT-3)

NNNY gys qf

////q

10 g rem for cumulative accident for cumlative accident fre ary dose <2 Address best-estimate hydrogen generation from both in-vessel and ex-vessel sources and demonstrate that ALWR severe accident mitiga-tion requirements are met Provide features such that early containment failure sequences-(e.g., containment bypass, direct containment heating) are prevented through engineered means and thus are. very remote in probability w c _i..E. m Ad-. -..--J- - ~ -,, _, / . _ ~ . m. ,a e ._g; .~ Evolutionary Plant Mitigating features Directed at Limiting Severe Accident Chat tenges to the Centainment Severe Accident . Limit Meanitude ~

Chattence Timing

. Limit Possibility Interfacing System Sort tene PWR/8Wt ~ Interlocks Pressure Relief LOCA Piping capable of withstanding esposure to peactor pressure ATWS Short ters 8WR Diverse RPS, Stan&y Ligsid Control-Diverse CRDMS ' Diversity within RPS. Steam Generator /EFW PWR Borated SI Nigh Pressure Short tenn SWR Depressurizati g System- . Suppression Poot Melt Ejection Automatic . p. Inerted Contairunent (No. Chemical Neat Nigh Capacity ' ~ Addition) .steudoun from relatively tou pressure PWR PORY

. Cavity Configuration to entrain debrfs Sized for depressurizing-Manually initiated Nydrogen Short tenn BWR Inerted Containment Detonation Containment Size PWR Core Concrete Long tern BWR

Passive flooder Interaction

Water addition to dryuelt from sources Noncondensible Gas enternal to contairunent -

Generation Baserent Penetration PWR Reactor Cavity Flooding from 1RWST Water addition to cent from sources enternet to containment Liner Melt through Long term SWR. Liner intiedded in concrete Passive flooder PWR Precluded by concrete reactor cavity floor Swp Line Short teen BWR/PWR Stmp Coniiguration Reactor Cavity /DryvetI ftooding - ~. a ~ - 7 cc-, -e [-f.. : ~':u-x .s_ ,~ ~-;. ^ . - 8 ~' -; ; i _; , _ 3;. Evolutionary Plant Design features Directed at Limiting Severe Accident Chattenges 2' Severe Accident timit staanitude thattence Timine Limit Potential - Water Addition to Contaitument' ' Decay Heat Long term BWR Main Condenser ' RNR Injection capable of pumping saturated fluid, ' ~ ~

' Reactor Cleantp System AAC 3

Water Addition to Contairunent PWR Steam Generations /EFW/MfW RNR

~

. Injection capable of ptsuping saturated fluid

' AAC i itbe Ripture from Not ' Short tene PWR Steam Generator /ETW Gases Depressurization uith PORVs Design requirement idiich exceeds capability of current generation of LWRS + m.e.% _-.. _ _.%s..* ..mm.m,_.mm u--.m-r wd-a m. Utility /EPRI Containment Performance Criterion Status: S The 'ALWR Program is reviewing options for a containment i performance criterion as part of evolutionary plant Requirements Document rollup The options for the containment performance criterion include: - Conditional Containment Failure Probability (CCFP)- - Time-based deterministic criterion - Demonstrate that individual containment challenges ara j accommodated - Existing requirements (which specify containment features and l 10E-6 / 25 rem) NRC staff has said CCFP of 0.1, or an equivalent deterministic criterion would be acceptable Our understanding of expected Commission response to SECY 90-016 is that it will generally accept the use.of CCFP for overall regulatory. guidance, not individual plant evaluations i Advanced LWR Program 1 l 4L Utility /EPRI Containment: Performance Criterion, continued: Current ALWR Position: I The CCFP.is.not being considered for.the Requirements Document since it can be a disincentive to the Plant Designer to reduce core damage frequency The need for, and effectiveness of a time-based criterion is being i evaluated: e._., for selected sequences, containment remains intact for n x flours A containment challenge approach is also being e'.

uated, e.g.,

i require Plant Designer to. systematically define potential l containment challenges and demonstrate that design features exist to render each of these challenges negligible. The existing ALWR requirements (with potential changes based on discussions with NRC)7rovide the technical basis for adequate containment performance; specifying a containment performance criterion could change the format but is not i ~ expected to change the substance of this technical basis ACRS 7/90-2 w. ~ .w.., -a.,.- ~ --si - - e., s,-- e r &~ w +,,;c I ENCLOSURE 4 i SUMARY OF TECHNICAL DISCUSSIONS OF JULY 16, 1990 MEETING t Hydrogen Generation t The-staff first discussed some of the tests and analysis regarding hydrogen generation. It opene( the discussion by stating that it is difficult to. i predict what occurs during the late phase of meltdown during a severe' accident due to the effects of crusting, uncertainty in amount of surface area that would be exposed to react, and the uncertainty in predicting how the damaged core would relocate in the reactor vessel. Modelling of late phase meltdown is more parametric than mechanistic. Therefore, there is a wide variation of opinion among the analysts in predicting hydrogen generation phenomena. i Because of these uncertainties, the staff believed it appropriate to use a design criteria that gives the staff reasonable assurance t1at, es more information is obtained through experimentation and analysis, tFa designer will not need to use up the margin incorporated into this crit < ria. H Uncertainties in the models include such items as in-vessel sources, calculations on ex-vessel quenching of the corium, the coolability of the debris bed, and the core-concrete interaction mechanisms if water is provided after vessel. failure. There is no data base of information concerning the s effect of depressurization on accident sequences. I Since the staff approval of designs is being sought today by the vendors the staff and the designers must agree on a solution to the issue that will let t the vendors complete the design. EPRI felt that the plant features _in the l design of the-plant should have allowed a deviation from the Hydrogen Rule. EPRI indicated that they believed ex-vessel hydrogen generation was limited by L the cavity flooding features which would assure cooling of the debris over a reasonable time. The staff indicated that the phenomena of debris cooling was l difficult to predict since, depending on the scenario, water could be added l before and after vessel failure, and at different time intervals. L Furthermore, EPRI has not determined whether cavity flooding would be provided prior to or subsequent to failure of the reactor pressure vessel. I EPRI-requested feedback on their February 5,1990 report on hydrogen control. The staff indicated that they did not plan on providing specific feedback on the-report since they treated it a part of EPRI's input in support of their position on hydrogen generation and control. The staff agreed to determine the best method for providing the requested feedback. Hydrogen Detonability The staff first discussed some of the tests and analysis regarding hydrogen deflagration and detonability. During the discussion, the staff indicated that'it has determined that the requirements to ensure the plant design ensures-that hydrogen concentrations do not exceed 10 percent in the containment again comes from its evaluation of an accumulation of information from experiments and analysis and the uncertainty in our understanding of the physical phenomena associated with severe accidents. The National Academy of Science has determined that a hydrogen concentration around 9 - 11% should not detonate and J. Shepherd provided the staff with his opinion that a 10% ---N re-a m m w .o. 4,s Wf ' Hb concentration of hydrogen in the containment should provide a margin of safety '

l against detonation. The staff acknowledged that it was difficult to draw a-direct correlation of lab experiments to full scale containment sizes, but.so little data is-available that it makes it necessary to use all available data in its considerations.

't At the close of the discussion, the staf f agreed to look into providing information it has obtained f rom Battelle-Frankfurt and J. Shepherd regarding i detonability of hydrogen. In addition, the staff agreed to provide feedback 1 on the level of analysis necessary to be performed on a design with ignitors and without them. Containment Performance /ABWR Vent The staff stated that it: regarded its proposed conditional containment failure i probability (CCFP)of0.1asanobjective,andnotare designisclosetomeetingthisobjective(buthasnot)quirement. If a plant i , and the designer has taken reasonable steps to improve or,-at least, resolve weaknesses in the ' I design, the staff would likely a pprove the design. The staff's position was defined in SECY-90-016 (and furtler detailed in the DSER on Chapter 5 of the EPRI ALF" Requirements Document) and endorsed by the Commission. A Regulatory Guide 3 aing developed on containment performance, but is not planned for C . a couple of years. >There are difficulties in defining a large release. Over the years, several i differentproposalshavebeenproffered,includingonedgfinedintermsofone prompt fatality, the EPRI goal ~that the design has a 10~ arobability of releasing 25 Rem at 0.5 mile from the site boundary, and tie staff reconnended criteria that the containment stresses do not exceed ASME service level C 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the onset of core damage and provide controlled venting capability after that. initial 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . EPRI indicated that the ALWR Requirements Document is fairly complete and will not need to be changed no matter what the definition is used for large 1 release. EPRI further stated that it would propose to use a CCFP = 0.1 if the i definition of large release were based on one prompt fatality. The staff explained to EPRI that it did not require-GE to incorporate the vent into the design of the ABWR. The vent was necessary to ensure the containment survived a TW avent. The staff is concerned that the ABWR containment may not be able to withstand such an event without the vent. However, the staff has not completed its review of the vent for the ABWR. The technical basis for inclusion of the vent in the ABWR design includes GE's position that it improves safety, it meets the Commission's regulations, and that technical "downsides" to the use of the vent may be minimized through design features and specific emergency operating procedures. EPRI asked if there are any accident scenerios for which the vent is the only solution. In response, the staff stated that the TW scenerio is the worst accident, and GE relied on a slow leak in the containment due to containment failure to provide a solution to the accident. The staff feels it prudent to provide for a method that would allow for a controlled release, rather than an uncontrolled one. However, there is no regulatory requirement for the vent. GE has agreed to put it % on their own. F.,, ?O, f,t. ~, 4 4 o y s,,' j,_ itf- [ 4 s i i [4 / - 3- '_k.. U The staff-then discussed-the results of. their review of 5 the RESAR SP/90 regarding.the.need.for a vent.-_ It: takes a long time to build up to its s maximum pressure-inside the containment (greater than 2 days),~ which. even p then, would'not be expected to exceed its ultimateL pressure capacity.; The; = staff noted it-has:not completed its revi_ew of this matter.- y fy ' .The. staff closed the discussion. stating.that it was. developing a response to- .the Commission to provide a recomended definition on-large rel. ease. ?? .i I,^. hl ') l ', I, l., t l$ i-5 _0 (;, t f(I q !L p y l - 1 t A

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