Text

Summary of 900531 Meeting W/Epri Re Source Term to Be Used on Future Lwr.List of Attendees & Viewgraphs Encl
	ML20058N102
Person / Time
Issue date:	08/08/1990
From:	Kenyon T; Office of Nuclear Reactor Regulation
To:	; Office of Nuclear Reactor Regulation
References
	PROJECT-669A NUDOCS 9008130223
	Download: ML20058N102 (60)
	v • d • e

-

.1 Project No. 669 August 8,1990 APPLICANT:

Electric Power Research Institute PROJECT:

ALWR Requirements Document i

SUBJECT:

SUMMARY

OF MEETING TO DISCUSS THE SOURCE TERM TO BE USED ON FUTURE LICHT WATER REACTORS On May 31, 1990, representatives of the NRC and EPRI met to discuss potential modifications to the current source term to be used on future light water i

reactors (LWRs). is the list of attendees. are copies of the staff's presentation. are copies of EPRI's slide presentation.

The staff made a presentation discussing the current status of its work concerning u? dating source term guidance and practice for future LWRs. The staff sees t1e updating effort as a two step process. The first-step consisting of a technical update of information related to the source term, i.e., updating the TID, regulatory guides, and the SRP. The second step would involve decoupling siting from the plant design thrcugh proposed rule changes.

EPRI opened their presentation stating that they were seeking a more rational basis for the source term to se used on future LWRs.

In order to do that, EPRI proposed to establish a source term based on the best knowledge available.

EPRI indicated that they focussed their efforts on those items in TID 14844 that the applicant felt were physically incorrect.

EPRI then discussed in more detail these areas, including fission product release fractions from the fuel into the RCS, timing of the release, RCS retention of the fission products, fission product speciation, and fission product transport into the containment.

At the end of the mceting, it was agreed to continue discussions of the work of both the NRC and EPRI at future meetings.

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

Enclosures:

As stated cc w/ enclosures:

See ner:t page co O DISTRIBUTION:

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SOURCE TERM MODif! CATIONS MAY 31, 1990 Narre Organization Charles Willis NRC/DREP Jay Lee NRC/DREP Frank Congel FRC/NRR Willieni Peckner NRC/RES/ SAIL John Trotter EPP1 Len Soffer NRC/RES/SAIB Tom King NRC/RES Ray Ng NUliARC Jerry kilson NRC/RES/ARG1P ve'n-Lic Caron NRC/RES/ARG1B Araian Heyraer NUMARC Tom Essig flRR/PRPB Charles Nichols NRR/SPLB Jim A. Martin NEP/PRPB Norm Edwards Censultant to EPRI Ralph Architzel NRR/SPLB David Shum NPR/SPLB David Leaver Tenera, San Jose, CA Bruce llonty Westinghoust Daniel J. Osetek LATA o

Richard Hobbins INEl Robert Ritman EPRI Jim Metcalf Stone A Webster llossein Nourbakhsh BNL Brad Haroin RES/ARGIC Walt Pasedag 00E Stephen Additon Tenera/ARSAP l

Xavier Pouget-Ababie EPRI Lee' Fiorych SLI L

Charles Miller NRR/PDS-Jim Martin NRR/DREP/PRPB Gary Yine EPRI Bill Sugnet EPRI Tom Kenyon NRR/FDS I

Ralph Meyer NRC/RES Stephen Koscielny NRR/EllCB Frank Witt NRR/ENCB 1

l I

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J STATUS AND APPROACH OF NRC SOURCE TERM j

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i UPDATE EFFORTS i

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i T. L. KING, USNRC-MAY 31, 1990 PRESENTATION TO i

i EPRI l

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STATUS l

1 i

o THE NRC STAFF IS EVALUATING UPDATING l

GUIDANCE AND PRACTICE FOR FUTURE-LWRs l

l l

RELATED TO THE USE OF THE SOURCE f

TERM:

STEP 1 - TECHNICAL UPDATE l

STEP 2 - DECOUPLING SITING FROM PLANT I

DESIGN o

WORK UNDERWAY FOR PAST THREE MONTHS l

1 l

4

TECHNICAL UPDATE l

i o

APPROACH:

l MAINTAIN CURRENT 10 CFR 100 REQUIRE-MENTS (INCLUDING USE OF R.G.4.7)

AND DESIGN BASIS.

CONTINUE TO EVALUATE CONTAINMENT PERFORMANCE UNDER LIKELY SEVERE ACCIDENT CONDITIONS ON A CASE BY CASE BASIS.

i UPDATE TECHNICAL INFORMATION RELATED TO SOURCE TERM:

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TID 0

RGs j

o SRP j

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TECHNICAL UPDATE (CON'T)

I EFFORT BASED PRIMARILY ON ANALYSES OF 1

i EXISTING LWRs i

4 1

PRIMARY EMPHASIS IS TO SUPPORT REVIEW 0F EVOLUTIONARY UWRs i

l APPLICATION TO. EXISTING PLANTS NEEDS TO BE CONSIDERED i

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3 i

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i TECHNICAL UPDATE (CON'T) o EVALUATION IS EXAMINING:.

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i TIMING OF RELEASE 1

BASED UPON LOCA CALCULATIONS (FOR ONSET OF PIN FAILURE)

AND NUREG

-1150 RESULTS FOR MELT PROGRESSION 1

(UPDATE OF NUREG/CR-4881)

FORM AND CONTENT OF RELEASE BASED UPON-UPDATE OF NUREG/CR-4881 AND ADDITIONAL CALCULATIONS ON CHEMICAL FORM..OF IODINE.

1 OTHER ASSUMPTIONS USED IN

~

CALCULATING OFFSITE RELEASE 4

3 i,

)

i DECOUPLING l

o LONGER TERM POSSIBILITY TO SUPPORT REVIEW OF PASSIVE PLANTS 1

t o

NO DECISION NOW - FURTHER STUDY REQUIRED l

IN CGNJUNCTION WITH REVIEW OF PASSIVE ALWR i

REQUIREMENTS DOCUMENT t

o DECOUPLING WOULD INVOLVE:

l CHANGE TO 10 CFR 50 TO SPECIFY DIRECTLY' DESIRED PLANT FEATURES CHANGE.TO 10 CFR 100 TO SPECIFY l

DIRECTLY DESIRED SITE CHARACTERISTICS 5

O,'

o FUTURE ACTIVITIES PLAN TO RECOMMEND TWO STEP PROCESS AND SCHEDULE TO COMMISSION (JUNE 1990).

SUPPORTING RESEARCH:

l ADDITIONAL SEVERE ACCIDENT SEQUENCES BEING ANALYZED BY BNL FOR COMPLETENESS WITH

(

REGARD TO FISSION PRODUCT TIMING AND l

COMPOSITION (UPDATE OF NUREG/CR-4881 l

SEPTEMBER 1990).

FISSION PRODUCT TIMING - ONSET OF FISSION PRODUCT RELEASE FROM FUEL-RE-ANALYZED BASED ON LOCA CALCULATIONS.

SOME l

RELAXATION OF TIMING EXPECTED (STAFF PAPER i

l JULY 1990, INEL ANALYSES DECEMBER 1990).

i INVESTIGATION OF IODINE CHEMICAL FORM BEING l

ACCELERATED AT COMMISSION REQUEST (ORNL l

DECEMBER.1990).

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R EX-VESSEL RELEASES E

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IN-VESSEL RELEASES S

E GAP ACTIVITY I

COOLANT ACTIVITY SECONDS MINUTES HOURS MANY HOURS TIME i

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Estimated Duration of Gap Release (from NUREG/CR-4881)

Plant Sequence Time (min)a Zion S 0C 6

p J

TMLU 6

s PWRs Surry TMLB 7

V 6

b-AG 27 Sequoyah TMLB 6

Peach Bottom TC1 10 TC2/TC3 13 TB1/TB2 24 BWRs V

8 Grand Gulf TC 19 TB 30 TBS 16 a

Time predicted between peak and core-wide average temperature reaching clad failure temperature, b

Assumed most representative of DBA for PWRs since ECC operates, e

+

0 Simplified PWR Fission Product Releases to Containment for Severe Accident Conditions (frmn PUREG/CR-4BBI) a-Groups FRCS FCCI Basaltic Limestone H,1 L

Concrete Concrete b

NG 1.0 1.0 0~

0 Cs, I 0.35 0.9 0

0 Te 0.3 0.65 0.15 0.35 Sr, Ba 2x10-3 0.01 0.15 0.4 Ru, Ce, La 3x10-5 3x10-5 6.x10-3 0.05 Release Duration 40 mins.

2 hrs.C a

H, I and L refer to high, intermediate or low RCS pressure, espectively.

b All entries are fractions of the initial core inventory.

^

c Except for Te where the duration of ex-vessel release ie

.anded to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for PWRs and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for BWRs.

}

~

6 i

Simolified BWR Fission Product Releases to Containment for Severe Accident Conditions (fromNUREG/CR-4881) a GROUPS FRCS FCCI Basaltic Limestone H,1 1.

Concrete Concrete b

NG 1.0 1.0 0

0 Cs,1 0.7 0.8 0.15 0.15

'Te 0.1 0.15 0.12 0.5 Sr,Ba 6x10-3 6x10-3 0.2 0.7 Ru,Ce,La 3x10-5 3x 10- 5 6x10-3 0.06 Release Duration 1.5 hrs.

3 hrs.C l

H, I and L refer to high, intermediate or low RCS pressure, respectively, a

b All entries are fractions of the initial core inventory.

c Except for Te where the duration of ex-vessel release is extended to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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AREAS WHICH COULD BE AFFECTED BY

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ST REEVALUATION; TIMING OF ST FORM QT[Q BILfAS1 COMPOSITION ASSUMP.

BWRs Ilg. tis.

- SGTS START AND-PERFORMANCE X

X (R.G.1.52).

- MSIV CLOSURE TIME X

- - NEED FOR'MSIV LEAKAGE

-X cu CONTROL SYSTEM (R.6. l.M)

- FP PLATE 0VT IN CONTAINMENT X

X

.(R.G.I.3)

-CONTAINMENTLEAKRATE(R.G.I.3)

X t

- SHIELDING DESIGN (10 CFR X

50.34(F)(2)(VII))

SAMPLING (10 CFR-X

=

m 50 34(F)(2)(Vill))

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- -10'(R.t1.d9.'

X

- CR HABITABILITY'(10 CFR X

S0.34(F)(2)(XXVill))

- LK CONTROL AND DETECTION X

_{

(10CFRCO.34(F)(2)(XXVI))

PWRs r

' : CONT PURGE VALVE CLOSURE TIME X

e P

N FP:PLETE0VT IN CONTAINMENT (R.G.l.4)

X X

- CONTAINMENT LEAK RATE (R.G.I.4)

X

- FILTEri SYS, DESIGN (R.G.I.52)l X

- SHIELDING DESIGN (10CFR50.34(F)(2)(VII))

X

- SAMPLING (10CFR53.34(F)(2)(VIII))

X

- EQ'(R.G.1.~89)

X e CR HABITABILITY-(10CFR50.34(F)(2)(XXVIII))

X

- LK CONTROL AND DETECTION 810CFR50.34(F)(2)(XXV))

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CURRENT AND FUTURE " ACTIVITIES

? ~-

-(CONTINUED)-

~

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CHANGES TOLREGULATIONS/ GUIDANCE:

a STEP 1 -

INITIATE CHANGES - 9/90 TECHNICAL BASIS FO'R TIMING CHANGE i

AVAILABLE 12/90

' DRAFT REGULATORY GUIDES, SRPs, TID REVISION AND SUPPORTING DOCUMENTS 1

AVAILABLE 7/91 PROPOSED CHANGES TO' COMMISSION 12/91 FINAL CHANGES TO COMMISSION 12/.R2*

~

STEP 2 -

-RECOMMENDATION TO COMMISSION 9/92 1

PRIOR-TO DESIGN CERTIFICATION OF CE SYS 80+

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.l MODIFIED TID 14844 1

SOU.RCE TERM:FOR q

ADVANCED ll LIGHT WATER REACTORS

+

Presented by the Advanced Reactor Severe Ace! dent Program o

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on behalf of the Utility / EPRI ALWR Program s

D. E. Leaver / Tenera, L.P.

u Presented to NRC Staff May 31~,1990

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Utility /EPRI --

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'I ALWR Program Presentation Licensing tesign Basis Source Term I

Advanced LWR Program Utility /EPRI 1

1 ALWR Program Objectives in Raising Source Term lasue Incorporate source term knowledge developed over the e

i last 10 years

A general desire for ALWR safety to be based on best available technical information as opposed to regulatory requirements which may be out of date

. Desire for more rational design basis for accident

- mitigaton systems AdvancedLWR Program 1

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Utility /EPRI t

i Range of Possibilitlee for Source Term l

Compete 4y enemy eemed wleressel Tin leselflod TIO Term

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if

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leadifled TlO charmetarietlan gvaleanw aamad LDa Rauram Tarse lia Jastatmam

Use 70 as etering point end pstfy a t es essete escuences as eieren0 poet; oenesono eentnese sequence typee Desse on uent aseen j

eneer nomeemos argumente Quien rosese and 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months eseoeure

. Permessy bened seevenne amme end re*ese le O., many hours)

Lime or no couanne of contamment a Bene aerosol removal and leake0e on -

I tnorme nyereuhce and source term Niemment conosons for the escuences Peon seeuence specin: source w.

. Em weis source term for imponent seewonce somiceoes to el piente typen tot a gwen eienderstod plant oceagn Sepse reope n Aeo. Guides

. Roope in Reg.Guedes, but not ouise es em i

AdvancedLWR Program

?

Utility /EPRI i

Use of Mudified TID vs.

a Physically-Based Source Term i-

A modified TID approach vs. a physically based approach is being suggested for the Evolutionary ALWR Requirements Document at this time because:

it is a first step which can be agreed upon in the near term 10CFR100 -2 hour siting dose tends to be compatible with a

[

quien release, non mechanistic source term t

For specific Evolutionary Plant designs which are essentially a

completo, TlO 14844 is still acceptable from a siting Mandpoint and could be used for certification with the modified TID used to I

demonstrate margin For future ALWRs, a more physically based source term should be considered Advanced LWR Program i, i t f.

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Relationship of. Release Magnitude and Timing

{

Relative Salggt Release idannitude Release Timing Coolant -

Very Low.

Very early Gap Low Early

_[

Fuel High Late l

Objective: Define release timing which reflects this order Advanced LWR Program 1

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Utility /EPRI Proposed Release Timing Based on a review of a variety of existing plant PRA anahses and excludin0 sequences which are very remote in probability (e.g., core damage from large LOCA and ECCS failure) or are elimmated for the ALWR (recirculation pipe break):

Fission product release from fuelliquefaction or molting begins at the earliest at about 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

One sequence had gap release at - 40 minutes 1

Belggt Release Timina Coolant On order of 1 minute (to be defined) -

i Gap Burst at 30 minutes Fuel-Uniform release from 60-90 minutes AdvancedLWR Program J

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(

i Summary of Evolutionary ALWR LDB Source Term 1

(continued);

j Ealoting l

LD rce Term ce arm Aerosol Removal

.' I

' in PWR Containment

from Sprays 0 5 minutes after x = 40 per hour A = 7.5 por hour beginnittg of release 5 30 minutes after A = 20 per hour A = 7.5 per hour beginning of release 4

30120 minutt. after :

A = 3 per hour x = 0.75 per hour after begintung of release

' 2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after -

A = 1 per hour A = 0,75 per hour

' beginning of release

-l Advanced LWR Program Utility /EPRI Summary of Evolutionary ALWR LDB Source T64m (continued)

ALWR Exleting Regulatory LDa source Term Soures Term Mixing Rate x = 10 unsprayed A = 2 unsprayed between sprayed volumes per hr volumes per hr and unsprayed volumes in PWR -

Suspended Aerosol Concentration in BWR

~ 0 30 minutes 40% of aerosol 10% of aerosol release to release to containment containment 30 120 minutes 10% os yprose:

10%of aerosol i

release to release to contaira n containment

- ~ ~ - - A&anced LWR Program 1--- - -

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EVOLUTIONARY PLANT LSOURCE' TERM L

RCS RETENTION ll[

EXPERIMENTAL RESULTS.

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ANALYTICAL RESULTS RECOMMENDED 'ALWR RETENTION' FACTORS k

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DOMINANT FISSION PRODUCTS (Csi, A,1 AND CsOH) WILL FORM AEROSOLS IN THE RCS L

l-e ALTHOUGH EXPERIMENT RETENTION FACTORS ARE NOT DIRECTLY-7 l

APPLICABLE TO ACCIDENT ANALYSIS...

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AEROSOL BEHAVIOR IN EXPERIMENTS IS INDICATIVE OF EXPECTED ACCIDENT BEHAVIOR 1-i I

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SUMMARY

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RETENTION FRACTIONS DEPOSITION

(% OF SOURCE)

CLOSE TO FUEL TOTAL TEST SPECIES (SOURCE)

PIPING

= LACE LA3A CsOH/MnO =.21

~ 26 77-d f

LA3B CsOH/MnO =.13 15 51 LA3C CsOH/MnO =.61 46 83 LA1 CsOH/MnD =.43 99 MARVIKEN ALL AEROSOL

~74 SFD 1-4 IODINE 10 95 CESIUM 30 95

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SUMMARY

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RCS RETENTION CAN VARY -WIDELY l

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-HIGH RETENTION IS NOTED FOR-LOW TO MODERATE

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VELOCITIES AND LONG TORTUOUS PATHWAYS l

e LOW RETENTION IS NOTED-FOR VERY HIGH VELOCITIES AND SHORT PATHWAYS l

e IODINE, CESIUM AND BULK AEROSOL GENERALLY BEHAVE.

1 (DEPOSIT) SIMILARLY ~

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CODES THAT INCLUDE AEROSOL BEHAVIOR MODELS.

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DEPOSITION AND OVERPREDICT RELEASES

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MEDIAN RETENTIOM FACTORS.

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CASE CONDITIONS IODINE CESIUM AEROSOLS l

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BWR1 HIGH PRESSURE, 91 97 97 EARLY MELT BWR2 LOW PRESSURE, 59 70 74 l

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ANALYTICAL RESULTS

SUMMARY

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~RCS RETENTION IS HARDWARE AND SEQUENCE DEPENDENT l

B-E CALCULATIONS UNDERPREDICT EXPERIMENTAL RESULTS

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e B-E CALCULATIONS PREDICT 15 TO 96% RETENTION FOR EXISTING LWR SEQUENCES-e LOW RETENTION RESULTS FROM:

i CLOSE-TO-VESSEL PIPE BREAKS, AND REVAPORIZATION AT LONG TIME e

NUREG-1150 EXPERTS SPECIFY MEDIAN RCS-RETENTION IN LWRS

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i OF 48-96%

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t-t EVOLUTIONARY PLANT SOURCE TERM. ASSUMPTIONS 1

l LARGE BREAK-SEQUENCES ARE LOW PROBABluTY, <10-7 e

e LEAK BEFORE BREAK TECHNOLOGY REDUCES PROBABILITY

FURTHER, l

e REVAPORIZATION OCCURS LATE IN SEQUENCE, j

e TERMINATED-SEQUENCES REFLOOD BEFORE REVAPORIZATION, o

EVOLUTIONARY PLANT DESIGN FURTHER REDUCES. PROBABILITY OF LARGE RELEASES FROM THE RCS.

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3- - -

J RECOMMENDED RCS RETENTION FOR EVOl_UTIONARY -PLANT SOURCE TERM e

70% ' RETENTION-FOR IODINE, CESIUM AND BULK AEROSOL BASED ON THE FOLLOWING:

EXPERIMENTAL EVIDENCE INDICATING a 70% RETENT(ON 6N VAPOR PATHWAYS, NEARLY COMPLETE AEROSOL RETENTION IN LIQUID PA1HWAYS, e.g., TMI-2, EXTREMELY. REMOTE PROBABILITY OF CORE DAMAGE FROM LARGE, CLOSE-TO-VESSEL PIPE BREAKS, EXTRAPOLATION OF NUREG-1150 ESTIMATES REDUCED REVAPORIZATION IN TERMINATED SEQUENCES, LOW PROBABILITY OF LARGE BREAKS, ALWR RCS DEPRESSURIZATION TO 1RW'ST.

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3

SPRAY REMOVAL-(JAN 88 EPRI STUDY) l r

MIXING BETWEEN SPRAYED AND UNSPRAYED

' l VOLUMES EVOLUTIONARY BWR l

)

. -SEQUENCE THERMAL-HYDRAULICS j!

POOL DECONTAMINATION i

SUMMARY

OF CONTAINMENT FISSION PRODUCT l

REMOVAL 1

^

1 u_

~,

~

lT i

SUMMARY

LOFLKEY ASSUMPTIONS

~

j ACCIDENT-ARRESTED 1,ARGE BREAK LOCA ( fif

~

RELEASE TIMING

.y

~ START '- 60 MIN 1

DURATION - 30 MIN (UNIFORM RELEASE) l RELEASE FRACTIONS TO CONTAINMENT 1, Cs - 0.25' O. 2-k l

/

l Te 0.12' Ba, Sr, Sb, Ru

-0.003 0

Ce, La, Actinides - 0.0003 i

F i

IODINE FORM

~

PARTICULATE - 97 PERCENT L

ELEMENTAL -2.85 PERCENT.

~

i ORGANIC

- 0.15 PERCENT 4

~

.+,4

,_y,

...~..m_.,

y

EVOLUTIONARY PWR - SPRAY REMOVAL 0

1:

LOC &-bb/$

x BASED ON JAN 88 EPRI STUDY (CASE D)

CORE POWER = 3425 Mw(t) l

CONTAINMENT VOL = 3.5E6 FT3,70 PERCENT SPRAYED j f v#1+

)

]

3

SPRAY FLOW = 3130 GPM p

p' ', pg

VOL SPRAY FLOW / SPRAYED VOLUME = 0.01/HR c '

SPRAY DROPLET RADIUS = 150 MICRON j

i

RELEASE START AT 20 MIN i

RELEASE DURATION = 20 MIN j

TOTAL RELEASE = 420 KG I

- NUMBER MEDIAN RADIUS = 0.21 MICRON l

- SIGMA = 1.7 l

75 PERCENT CLAD REACTED, H2 BURN AT40 MIN

HYGROSCOPICITY NOT INCLUDED 1

1 EMPLOYED NAUA WITH NUREG-0772 SPRAY MODELING i

-l

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s = $V)weCszw2 ZOO r.

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EVOLUTIONARY PWR - SPRAY REMOVAL (CONT)

. SPRAY LAMBDAS.FOR AEROSOL.

8 40/HR FOR FIRST 5 MIN OF RELEASE e 20/HR FOR NEXT25 MIN (UNTILT=30 min;i e

3/HR FOR T = 30 MIN UNTIL'T ='2 HR e

1/HR FOR T = 2 HR TO T = 24 HR ELEMENTAL IODINE - SAME AS CURRENT.

PRACTICE ORGANIC IODINE - NO REMOVAL-

p.

~;

EVOLUTIONARY: PWR MIXING BETWEEN SPRAYED AND UNSPRAYED VOLUMES I

i l

l CURRENT REGULATORY PRACTICE ALLOWS 2 UNSPRAYED l

VOL/HR j

MIXING IS THE DOMINANT REMOVAL MECHANISM FOR UNSPRAYED VOLUME (FOR A PURIFICATION STREAM, MIXING RATE = LAMBDA)

IF CONT COOLDOWN RATE AND BUOYANCY DRIVEN FLOW IS i

CONSIDERED, MIXING RATES GREATER THAN 10/HR CAN BE JUSTIFIED FOR THE PERIOD BETWEEN START OF RELEASE j

AND TWO HR AFTER START OF RELEASE

]

MOMENTUM EXCHANGE FROM SPRAYS AND FORCED FLOW.

l PROVIDED IN MANY DESIGNS WILL SERVE-TO ENHANCE MIXIN3 l

l

1 l

EVOLUTIONARY BWR - SEQUENCE THERMAL-HYD LARGE LOCA SMALLER THAN CURRENT PLANTS LONGER TIME TO CORE MELT (ASSUME ONE HOUR)

DURING CORE MELT (ASS..WE ONE-HALF HOUR)

APPROXIMATELY 2E4 LBS BOILED OFF - LEADS TO AVERAGE CONCENTRATION IN DRYWELL OF 40 PERCENT OF RELEASE DUR!NG QUENCH, MINIMUM STEAMING IS 1.7E5 LBS/HR - LEADS TO DRYWELL TURNOVER OF 10 VOLUMES /HR DEPOSITION RATES IN DRYWELL ARE AN ORDER OF MAGNITUDE LESS THAN SWEEPOUT SWEEPOUT AND REDISTRIBUTION ASSUMED TO BE INSTANTANEOUS AFTER QUENCH AFTER QUENCH, CONCENTRATION IN DRYWELL = (100/ POOL DF) PERCENT OF RELEASE

--um

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EVOLUTIONARY BWR - POOL DECONTAMINATibN N

r L

j a

a CURRENT REGULATORY PRACTICE -

DF=10 BELIEVE THAT HIGHER DFs CAN BE i

JUSTIFIED j

NUREG/CR-4624, VOLUME 1 SENSITIVITY STUDY, CASE.5A a

i e PERIOD ~OF INTEREST BETWEEN 250 MIN -

1 4

350 MIN e DF ABOUT100 l

HIGHER CLAIMED DF WILL REQUIRE j

CLOSER LOOK AT BYPASS I

1

-g..,~a

~

. u.-

f CASE 5A DECONTAMINATION FACTORS

~

10000_

4 1

1000.

g O

3 1

0 I

I

[t 8

4

~

10 0:

\\

(P

\\

I 6

fJ 1

e t

=

8

~

(C

\\

8

(

i O

lb0 2b0 3b0 Ab0 Sb0 6h0 7b0 800 i

TIME (min)

FIGURE 6.10.

DECollTAMillATI0lt FACTORS FOR litTEREDIATE ELLIPTICAL BUBBLES IIICLUDIIIG'

~

CIRCULAT1011 AllD SOLUBLE PARTICLES

y i

l

l; u-

EVOLUTIONARY ALWR LDB: SOURCE-TERM j

J

. Proposed ALWR Existing -

p LQR Source Itan Regulatory 1929.t ItEn L

E

I: Aerosol' Removal'in PWR

l. Containment from' Sprays y,.

0-5 minutes after A = 40 per hr.

A = 7.5 per hr.(2)

+

I,'

beginning of release.

' 5-30 minutes after A = 20 per hr.

A = 7.5 per hr.(2) beginning of release 30-120 minutes A = 3 per hr.

A = 0.75 per hr.(2) j,

'after beginning of l

release..

V I

' 2-24 hours after A = 1 per hr.

A = 0.75 per hr.(2)

T beginning of release o[

Mixing rate between A = 10 unsprayed A =-2 unsprayed volumes

'l

sprayed and unsprayed volumes per hour.

per hr volumes in PWR E

> ? Suspended Aerosol LC6ncentration-in-BWR

. 0-30 minutes 40% of aerosol release to 10% of aerosol release to Containment Containment (3) j m

s. E 30'-120 minutes 10% of aerosol release to 10% of aerosol release to containment (3) -

containment (3)

'Fotes: '(2) Based on SRP 6.5.2 for aerosol removal

(3) Based on suppression pool scrubbing decontamination factor of 10-4

~

1 l

I Fission Product Release; s

from Fuel into the RCS 1

i Presented by-R.R.Hobbins May 31,1990 j

Idaho Rockville, MD.

j National

'\\

Engineering l

MOO 00 O Laboratory Advanced Reactor Severe ccident Program j

A Source Term Expert Group with USNRC

i

ll' ll)

1 l t

, l k

c c

7 m;:3,

f r

i f.

e i

o t

e

7 a

w n

u s

eo l

s e

m^

~-

s ae.

l'-

a l

'r, t

e a

r w

d t

o c

la y

u w

h t

gd

-+

r c

n o

o e

a el r

-r o

p n

o md r

i i

p r

'o n

l p

eh o

i t

a pt s

/

x e

u s

d e mf i

O n

f u

od d

o e

e

~

ws s

rg e

o o

k p

p i

c v

o o

a e

r r

B RP P

w p

r,

.s

~.

,lll iilLl1llll

\\lll lj I

c

. e i

j I

I

.i l

Models of fission product rele.ase 1

L are becoming increasingly complex.

3 i

l CORSOR.

Release as a function of te.aperature

^l n

l MAAP

Release as a function of temperature l

and oxidation-enhanced grain growth-3 GRASS

Release as a function of temperature, j

fuel liquefaction, and oxidation-enhanced diffusion-VICTORIA : Release as a function of temperature, i

fuel li.quefaction, oxidation-enhanced diffusion, fuel chemistry, coolant, thermal-hydraulics

j

~

,h y

Approach Values for fission product release in the. case ofi t

" substantial core melt" for an accident terminated in-vessel deveiened on the basis of results from t

integral in-pile experiments and the TMI-2: accident' TMI-2, PBF-SFD, and LOFT-LP-FP-2 indicate

~10% Cs and.I remain in previously molten fuel.

l No residual fission:; gases remain in the m'elted

?

l material.

l

)

>:, y l

.+

Release Fractions from the Core in the TMI-2 Accident Element Fraction es Kr-0.54 I

0.55 i

'S

'7 Cs 0.55 i

132Te 0.06 8 Sr 0.001*

12sSb 0.016 1

i

'8 Ru 0.005 l

'"Ce 0.0001 i

Leaching from damaged core after reflood increased Sr release a

to 0.032 two months after accident 4

.l

[

Release Fractions from the Fuel in l

1 the PBF Severe Fuel Damage Tests Element SFD-ST S FD-1 -1 SFD 1-3 SFD 1 i Kr,Xe 0.50 0.026-0.093 0.08-0.19 0.23-0.44 I

0.51 0.12 0.18 0.24-Cs 0.32 0.09 0.13-0.18 0.39-0.51 j

Te 0.40 0.01 0.01-0.09 0.03 i

Ba 0.11 E-1 0.06E-1 0.04E-1 0.07E-1 Sr 0.02E-3 0.24E-3 0.88E-2 1

Sb 0.19E-3 0.13E-2 1

Ru 0.03E-2

. 0.02E-2 0.03E-3 0.07E t Ce 0.02E-4 0.09E-3.

0.08E-3 0.13E-3 Actinides

<0.01 E-2

<0.01 E-3 Experiment Conditions-Zr oxidized-(%')

.75 26 22 32 i

Fuel melted (%)

15 16 18 18 i

i

~-

m

~

y. #

.; x n

q Used simple empirical correlation j

to calculate releases of Cs, I, NG 1

1

%I, Cs released = % fuel melting x 0.9 +

too I

(F- % fuel. melting) x 0.25 i

% NG released = % fuel melting +

/00

(;P- % fuel melting) x 0.25 i

Compared. calculated results with e

experimental results j

i Assumed 75% core melt

-J

=,

__ ^'

Calculated results are in fair l

agreement with measurements.

% Release

% Release Measured

. Calculated 1

Test / Accident

% Oxid. % Melt. _L Cs_

NG I.Cs NG i

TMI-2 50 45 55 55-54 54 59 SFD ST 7fpr 15 51 32 50-35 36' l

SFD 1-1 26 16 12 9

9 35 37 SFD 1-3 22 18 18 18 19 37

39-SFD 1-4 32 18 24 51 42 37-

.39 LOFT LP-FP-2 46 20.

16 16

-12 38-

4. 0

--c

-~ --

I Release from Fuel (%)

\\

n i

75 %

60%-

~

l Elemental Group Core Melt Core Melt

~ :

Noble gases (Xe, Kr) 0.81 0.70 i

Volatiles (I,Cs) 0.75 0.64-Tellurium 0.40 0.35 Semi-volatiles (Sr,Ba,Sb,Ru) 0.01 0.01 ~

I

(

Low volatiles (Ce,La, Actinides) 0.0001 0.0001

~

y i

M083 rrMMi90-10 l

a

+

n 1

1 1,

Fission Product Speciation JJ!

L in Core Damage? Accidents

=

i i

a R. L. Ritzman a

I Electric Power Research Institute Palo Alto, California l

May 31,1990 i

i i

~

RLRS31/90-1

-m._,s..

t-wi.

m

.mi...

-%.s.

..us.

.m ii,,.

..a..n__.m,.

m.

._m,mm%,_

~.

.m

.%,+,a

~

.uw-s y,,,

y

~

Fission Product Speciation in Core Damage Accidents Important factors-Accident terminated in-vessel e

i Containnient spray available j

RCS processes affect speciation l

i i

l Containment processes affect speciation l

l Qualitative vs quantitative assessment i

-j FM.R5/31/90-7

^

-w 1

Volatile Fission Products-1 RCS Speciation - Experience i

4 Fission Dominant Data Normal Water

]

Product Species Sources (a)

State Solubility l

1 Csl ORNL, STEP, SFD, TMl Solid High l

)

Agl LOFT Solid Low j

Cs CsOH LOFT, SFD, STEP, ORNL Solid High l

Te SnTe BCL, ORNL, SFD Solid Low i

~

CsTe, Csj e ORNL, STEP Solid High T

i (a) _ The systems which provide the data sources contained no boric acid i

except for theTMI-2 accident and the LOFT-FP2 test.

1 FER5/31/90-3 m

.N r -

.--M..

,g.,_..

~

y.;=~

~

'{

~ :5.

=

RCS?ReactionGroup

~

m.

i i)

Csi + HaBO3 CsBO

+ H 04 Hi

- 4

=

2 2

CsBO '~+ 2H 0.

2)

CsOH +-H BO3=

2 2

3 4

3)

Csl + H O CsOH + HI

=

2 4)

CsOH + oxidized S.S. =

cesium silicates + H O 2

N 2HI = H

+1 2

2 6)

HI + metals = metaliodides + H 2

N) 1 '+ metals = metaliodides 2

HLR/5/31/90-4

_~'

a i

.j j

RCS lodine Assessment 1

i Excess CsOH vapor will minimize HI formation via reactions (3) and (1) i H1 while gaseous is quite reactive and water soluble. Also, iodine still in its most reduced state i

Presence of hydrogen in gas mixture will minimize 1 formation via i

2 reaction (5) and maintain a chemically reducing environment Technicaljudgment: Almost all.the iodine leaving the RCS will be particulate.(Csl or possibly some Agl). A few percent may exit as HI and an insignificant fraction (<.1%) as'l.

j 2

l t

nuvsawns l

~

. n s ' -

g.

..G_

~

l ~~ l

';- - j]

. -i

, _ ;f?

l: [

~

4-Containment Reaction Group; 5+5 i

IN Csl -+ -water = ' neutral' solution

~t 9)

CsOH + water =

alkaline solution 10)

Hi + water =

acidic solution t

1 t

i 1i) 312 + 3H O = SI + 103 + 6H+

2 12) 12 organic =

organic iodides I

I 2;

a b

i 1

[f

~

--W-g p

.w.

,g 3

a-~4.g.a ogq A

_ _ _,=. -- __

yq w

_my

.ar 9

g-.mys.6 cy y__..

g,.yo g

aes u-#

_^_ys

r-.b'.

x-L Containment Reaction Group (continued) 13)

- CO2 + water =

weakiy acidic soiution 14)

H BO3+ water =

weakly acidic solution 3

l I

l 15)

Alkali metal borates + water = alkaline buffer solutien l

?

16) 2CsOH + CO2=

Cs2CO3+HO 2

s l

17)

- Cs2CO3 + water =

alkaline buffer solution 1

18)

Agi + water =

water + precipitate a

t t

i FER/5/31/90-7

-.~n.,.

p -s 3--%

-~g-c, ow,

-x m

7

=

3 1

Containment Reaction Group

~

. (centinued) i H 0, H, H, OH, e;q, HO, etc.

j H O + radiation =

19) 2 2

2 2

2 l

i

.i l

20) lodide solutions + radiation =

l,103, HOI, etc.

j 2

(Yields very sensitive to solution pH) l 1,105, etc.

l 21)

. Airborne Csl + H burn =

2 2

(Steam appears to be protective.)

22)

Water puddle. evaporation to dryness =1, etc.-

2 (Sensitive to pH, radiation. and presence of Ag.)

RLR/5/31/90-8 i

~~

e' J

w ion O

l

~ [

INITIAL t CONCENTRATION-~

g O

io-' m so

{

~ 10" m O

O' 50" m

~

an z

!?

eo 1

i 3

5'

_~ ao

~

r:

2e O

l l

O o

y 1

1 o

2-4 s

e to 17 pH Effects of pH on the ir yield in deserated iodide solutions iriadiated at 4.5 x if R/h for I hr.

l i

i From C. C. Lin, J. Inorg. Nucl, Chem. 42,1101-7 (1980) fit RS/31/90-10

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

...--..n...-.

,,r 4

e Containment Reaction Group i

(continued)

l 23)

Radiolysis of aerosol droplets =

1,105, HOI, etc.

l 2

(Sensitive to solution pH) l 24)

Radiolysis of moist air = nitric acid + other products

- j i

(Long term pH decreaser) l 1

25) 12 + organics +: radiation organic iodides

=

l (Yields depend on contact mode.)

[

l

~

I ntFvsates

-.-l D

-m,e.

.s

-m-N-p a

n, g_.

__..___..y

.+

d 1

Containment lodine Assessment l

o I

i Assume containment water pools are maintained alkaline for the accident duration such that 1 2 release to the gas space is very low (<1%).

j Technical judgment: Hydrogen combustion is not likely when Csl is airborne and steam. appears protective even if a burn occurs. The yield of 12 i

vapor from this process should be.very low (<1%).

l RLRS/31/90-11

+

_.4---*

.,.,-y.

w 4.,..-',i g

e q -g 4y a

m l

Containment lodine Assessment (continued) l Technical judgment: The fraction of total contain-l ment water containing iodide that might evaporate to dryness will be small. Hence the yield of 1 vapor 2

from this process should be very low (<1%).

j Technical judgment: Any HI released to containment i

l will be readily neutralized by interaction with alkaline l

l aerosol. The yield of 1 vapor from this process i

2 should be insignificant (<.1%)

2 ntascimi:t 1

g7

~

=

, 1

.a

.m

^'d'

+

Organic lodide Formation C

a l

Yields depend on a variety of parameters including 1 concentration, temperature, radiation

]

i 2

dose, type of organic, geometry effects, etc; 1

l Strategy: Use the empirical procedure;of-i WASH-1233 to estimate the yield from the com'oined assessed amounts of 1 2 vapor

~

l a

k 4

FEFW/31/90-13

~.-.....,

.,,..,,_.._.,4--.,,

~...

e

^ -

~

RCB PROCESS-

- Organic lodide Formation BNWL-B-213 (Oct.1972) Postma & Zavadowski T

33

.e

.==i>wi.

i%

e r---

s en N

g N

-s W enesensmou reactvv wetasm ro one.secmers s

o

~

m !!D !*m* cW****

r n us n'ene sis f *n ***"rs h "' N s_.

~_

g'

~

.3

~

' wijh s

e

-s.s w w srs y Si,,4 5

unis.w..c.a.wr.nsov.ow r

~~,

4 ',

%,s.,

r,,

evera mesvanareenou

+

, 5.-

<cwranverur earvnruenrs

~

/

s

~

~*

f.

j-comaarw ouw mamt m e -

't 41

..L-c q

..u.==< couctuvaunw #'

i W

h M

- ~ -

~e

,.w.

w.

m m.-a e

m,.

7-,

. w r,s r

+,".

.+

^

i i

~

r j

L

Chemical Form Summary

]

1 y

a Fission Containment 1

Product Suspended Form 1

Xe,Kr Gases 4

I(a)

Aerosol 97.0 %

I 2.85%.

2

- CHal 0.15% -

- q Cs Aerosol Te Aerosol Others Aerosol 3

(a) Assumes alkaline containment water pools-FER5/31/90-14 i

--*w--

-aw-ir-6%, e e

-e c.e.

w u.am w,:-

=ws.

4.*w

_a we,,--

.g ti.

,p_

+~-%-

a wg --

g-mc-w am,w W-

_

ML20058N102

Contents

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SUBJECT:

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ML20058N102

Text

SUBJECT:

SUMMARY

Enclosures:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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