ML20085A736

Rev 2 to Vol 2 to CRBRP-3, Assessment of Thermal Margin Beyond Design Base (Tmbdb)

ML20085A736
Person / Time
Site:	Clinch River
Issue date:	07/06/1983
From:	ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:
Shared Package
ML20085A733	List: ML20085A729 ML20085A736
References
CRBRP-3, CRBRP-3-V02-R06, CRBRP-3-V2-R6, NUDOCS 8307070105
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{{#Wiki_filter:~ CRBRP-3 O i I i l HYPOTHETICAL CORE DISRUPTIVE I ACCIDENT CONSIDERATIONS IN CRBRP VOLUME 2:

O i

l ASSESSMENT OF THERMAL MARGIN BEYOND l THE DESIGN BASE I l CLINCH RIVER BREEDER REACTOR PLANT SBS7258$R*oS8858$7 A PDR

l I h TITLE Hypothetical Core DOCUMENT NO.' s/ Disruptive Accident Considerations in CRBRP CRBRP-3 CHANGE CONTROL RECORD volume E volume 2 Assessment of Thermal Margin Beyond the Design Base CHANGE REV N0/DATE RELEASE PAGES AFFECTED REMARKS DOCUMENT 0/ March 80 All First formal issue of document Rev 1/5-81 i, 2-4, 2-6, 2-7, 2-10 Pages added 2-16A thru 2-16, 2-24, 2-25, thru 2-16G, 2-34A 2-26, 2-32, 2-34, 2-43, 2-44, 2-76 Rev 2/10-81 2-16B, 2-16C, 2-16F, 2-16G Rev 3/2-82 i, 2-4, 2-18 Page added 2-18A /) Rev 4/6-82 ii, iii, vii, xi, xiii, Pages added 2-10A, 2-10B (__ / xxiii, 2-7, 2-8, 2-10, Revised Appendix A 2-76, 2-78, 3-24, 4-10, Added Section H.3 4-19, 4-24, A-1 to A-25, C.1-3, C.1-12, C.1-18, C.3-19, G.2-1, G.2-3, G.4-4 to G.4-6, H.0-1, H.1-4, H.1-5, H.1-9, H.3-1 to H.3-7 Rev 5/3-83 ii, iv, ix, 4-1, 4-4 to Pages added 4-10A, E-4A, 4-7, 4-9, 4-10, 4-13 to Revised Section G.4 4-20, 4-22, 4-24, E-2 to E-9, E-11, G.4-1 to G.4-6 Rev 6/6-83 vii, 2-7, 2-34 Added Appendix K bo 01142

CRBRP-3 Vol. 2, Rev. 6 TABLE OF CONTENTS (Continued) Page H.2 FLAME LENGTHS FOR TMBDB SCENARIO H.2-1 H.2.1 Introduction H.2-1 H.2.2 Flame Length Calculation H.2-2 H.2.3 Effects on Containment H.2-6 H.2.4 Conclusions H.2-8 H.2.5 References H.2-9 H.2.A Flame Length Correlation H.2-16 H.3 POTENTIAL FOR HYDR 0 GEN STRATIFICATION H.3-1 H.3.1 Introduction H.3-1 H.3.2 Pre-Hydrogen Ignition Time Period H.3-1 H.3.3 Hydrogen Ignition to Sodium Boildry Time H.3-4 H.3.4 Post Sodium Boildry H.3-4 H.3.5 Conclusions H.3-5 4 H.3.6 References H.3-5 I. ASSESSMENT OF CONSEQUENCES OF FUEL IN PHTS PIPING FOLLOWING I-l g REACTOR VESSEL DRAINING I.1 INTRODUCTION I-l I.2 EFFECT ON PHTS PIPING I-3 I.3 EFFECT ON CELL LINERS I-5 I.4 SENSITIVITY ASSESSMENT OF FUEL IN PHTS PIPING I-7 J. NRC'S REQUESTS FOR ADDITIONAL INFORMATION J-l (RAI's) ON CORE MELTDOWN CONSIDERATIONS K. INFORMATION SUPPLIED TO THE NUCLEAR REGULATORY COMMISSION IN K.0-1 DEPARTMENT OF ENERGY LETTERS K.1 TMBDB INSTRUMENTATION DEVELOPMENT K.1-1 K.2 TMBDB MARGINS ASSESSMENT DOCUMENT K.2-1 K.3 TMBDB MELTING SCENARIO K.3-1 K.4 INFORMATION ON TMBDB K.4-1 6 vii l

p a 9 CRBRP - Vol.2, Rev.0 LIST OF TABLES Table Number Page 2-1 CRBRP FEATURES PROVIDING THERMAL MARGIN BEYOND THE 2-42 DESIGN BASE 3-1 PUMP C0ASTDOWN DATA 3-76 3-2 PARTICULE DISTRIBUTION FOLLOWING IN-VESSEL 3-77 SETTLING 3-3 DISTRIBUTION OF UPWARD EJECTED DEBRIS 3-78 3-4 PARTICLE DRAG COEFFICIENTS 3-79 3-5 MAXIMUM CRBR FULL EQUILIBRIUM CYCLE DECAY POWER BY 3-80 REGION 3-6 STABLE DEBRIS BED DFPTHS 3-81 3-7 COMPARISON OF FUEL RETENTION CAPABILITY WITH 3-82 PREDICTED FUEL DISTRIBUTION 3-8 SENSITIVITY OF CONTAINMENT PARAMETERS AT 24 HOURS 3-83 AFTER AN HCDA TO VESSEL PCNETRATION TIME 3-9 ANNULUS COOLING SYSTEM ANALYSIS DATA 3-84 3-10 CONTAINMENT STRUCTURAL CAPABILITY 3-85 3-11 Keffective OF VARIOUS GE0METRIES 3-86 3-12 REACTOR CAVITY AND PIPEWAY CELL LINER FAILURE TIMES 3-87 3-13 HEAT TRANSFER COEFFICIENT USED IN THERMAL ANALYSIS 3-88 0F THE REACTOR CAVITY LEDGE 3-14

SUMMARY

OF RESULTS FOR SUBMERGED LINER (WITHOUT 3-89 CREEP) 3-1"

SUMMARY

OF RESULTS FOR LINER ABOVE SODIUM POOL 3-90 (WITHOUT CREEP) 4-1 CORE SOURCE TERMS RELEASED TO THE REACTOR 4-17 CONTAINMENT BUILDING FOR HYP0THETICAL ACCIDENT SCENARIOS CONSIDERED 4-? ATMOSPHERIC DILUTION FACTORS 4-18 l l 4-3 DOSE

SUMMARY

FOR HYPOTHETICAL ACCIDENT SCENARIOS 4-19 CONSIDERED O viii

CRBRP-3 Vol. 2, Rev. 6 b\\ %d 2. To insure containment atmosphere mixing during purging and venting, the purge air penetration shall be located so as to minimize the potential for direct flow from the purge lines to the vent lines. 6 3. The purge system shall prevent backflow from containment to the outside atmosphere. 4. The purge system, in combination with the containment vent and cleanup systems, shall maintain containment at a negative pressure after the containment pressure is reduced by the initial venting after 24 hours. 5. The purge system operations shall be by remote manual actuation from the control room. 1 2.1.2.8 Containment Vent System 1. To prevent containment failure by excessive pressure, the vent system shall have a capacity between 24,000 and 26,400 acfm with a containment 3 pressure of 30 psia, a containment atmosphere density of 0.07 lb/f t and a viscosity of 0.06 lb/ft-br. It shall remain functional if up to 1 300,000 pounds of aerosol enter the system at a maximum rate of 5,600 lb/hr. 2. The vent system shall exhaust the containment atmosphere from below the polar crane into the containment cleanup system. 6 3. The containment vent system shall be compatible with the following gases, vapors and aerosols: Ar, N, H, H 0, CO, CO, 0, Na 0, Na 0 ' 2 2 2 2 2 2 22 NaOH, Na 00, fissi n products, and compounds resulting from fission 2 3 product reactions. The system must remain functional for inlet g6s temperatures and pressures given on Figures 2-5 and 2-6, and, beyond 150 hours for temperatures up to 250 F. 4 '4. The vent system operations shall be by remote manual actuation from the /G control room. (") 1 2-7

CRBRP-3 Vol. 2, Rev. 4 9 2.1.2.9 TMBDB Containment Cleanup System 1. The containment cleanup system efficiency shall be a minimum of 99% for vented materials in the solid or liquid state, 97% fo' vapors (Nal, r Se0, and Sb 0 ) subject to condensation in the cleanup system, 2 23 and 0% for noble gases. These efficiencies shall apply when subjected to the vent rates on Figure 2-7 and containment atmosphere temperatures 3 on Figure 2-5 with a containment atmosphere density of 0.07 lb/ft ; 0 beyond 150 hours, containment atmosphere temperatures up to 250 F 4 shall apply. It shall be capable of performing all of its intended functions in the presence of Ar, N, H, H 0 CO, CO, 0, Na 0, Na 0 ' 2 2 2 2 2 2 22 NaOH, Na Co3, fission products, and compounds resulting from fission 2 product reactions. 2. The containment cleanup system shall remain functional at an aerosol mass flow rate of up to 5,600 lb/hr and a total mass of 300,000 pounds of aerosol entering the cleanup system. The principal constitutents of the aerosol are Na0H and Na 0, the proportions of which can. vary from 2 0 to 100% of the aerosol, and Na CO which can vary from 0 to 8% of 2 3 the aerosol. The aerosol particle properties are: Mass Mean Radius (microns): 5 < r50 < 10 Aerodynamic Equivalent Radius (microns): 2.3 < AER < 4.7 l Density (g/cc): 2.1 < p < 2.5 Mass Geometric Standard Deviation: 3.0 < a < 3.5 l l Aerodynamic equivalent radius is based on AER = r50 (** where p = 2.21 and a = 0.1 0 2-8

CRBRP-3 Vol.2, Riv.0 O The redundant filter train monitors provide for determination of the U radioactivity being released from the filter train. Monitoring will be acenmplished using isokinetir sampling norries and assor.iated three channel continuous air monitors (CAMS) which provide one channel each for particulates, radiciodines, and radiogases. The detectors and associated electronics are shielded to reduce the accident induced radiation background to levels suitable for system operation. The three channel CAMS will provide gross count rates for each channel. The predicted radioisotopic inventories within the RCB coupled with gross count rate data will allow estimates of off-site doses to be made and will provide early identification of rapid and/or significant changes in release concentrations. In addition, a suitably shielded plutonium air particulate monitor (PAPM) specifically designed to measure very low concentration of the long half-life alpha emitters, such as Pu-239, will be provided and will 'lso a continuously isokinetically sample the common exhaust. The PAPM provides capability for identifying the plutonium releases at the point where such m.) releases would be the most concentrated and in this way maximizes the sensitivity of the measurement. Redunda'ncy is provided forsthe CAMS by the common exhaust monitor and is required due to the inaccessibility of the channels under accident conditions. Redundant PAPMS are not required due to the inherent redu'ndancy of a typical PAPM,;hich is provided as a means of accounting for the natural radon' thoron background (switching collection between dual channels allows the radon-thoron on the " idle" channel to decay (leaving behind the longer lived isotopes)). The power requirements for the plant radiation monitoring system are supplied by the IE power distribution system. 1 N 2-33 9 h l -.--.------.---r,

s CRBRP-3 Vol. 2, Rev. 6 O Provisions for off-site monitoring are described in the TVA Radiological Emergency Plan, as discussed in Section 13.3.11 of the PSAR. 2.2.12.2.2 High-Range Containment Area Radiation Three High-Range Containment Area Radiation Monitors are provided to indicate the radiation levels within containment to essist in determining actions to protect the public. These monitors have a seven decade range to 7 10 R/Hr gamma. The detectors are located approximately 120 degrees apart around the Reactor Containrrent Building periphery in the annulus space, to take advantage of the relatively benign environment. The monitors are classified as Safety Class lE and powered from three independent divisions of power. All three monitors have continuous display in the Control Room and one channel is recorded. I 2.2.13 Electrical Power System The electrical power requirements for motors, controls, and instruments will be distributed as part of the Class lE electric power system using the appropriate standards of quality assurance, structural support, and physical separation. TMBDB Instrumentation is connected to C1' ass lE electrical power and is energized during both normal and emergency plant operation. Other electrical loads for TMBDB features are connected to IE electrical power and under normal plant operation remain de-energized. When the equipment is required to operate during the TMBDB event, it will be remote manually energized from the Control Room. 6 2.2.14 Containment Structures As a result of the structural analysis of the containment building, a few changes in the design have been made to provide increased thermal neargins. These include: O 2-34

CRBRP-3 Vol. 2, Rev. 6 O APPENDIX K INFORMATION SUPPLIED TO THE NRC IN DOE LETTERS Appendices K.1 through K.4 contain copies of the four major DOE letters supplying information to the NRC in the course of their review and preparation of the CRBRP Safety Evaluation Report (SER). Appendix K.4 contains only Enclosures 2 and 5 of the DOE letter because Enclosures 4 and 6 have already been incorporated into the appropriate licensing documents, and Enclosures 1 and 3 have been published as separate reports. O j (::') K.0-1 m w g.,--------,.r., w#y y7_ _-y,, %.w-- w---g yre..r wy-,g e g, e, y r-y-v g'-y -e e--wrw.

7 -. 1 i CRBRP-3 l . Vol.*2, Rev. 6 i!O i j K.1 TMBDB INSTRUMENTATION DEVELOPMENT I i { This appendix contains the DOE letter supplied to the NRC, dated f j September 29, 1982. f i i j s b r I l i I i 1 Ile 1 1 i i I l 4 l l };- t i 1 l K.1-1 i f i i l 4 5-I . ~..

_= CRBRP-3 Vol. 2, Rev. 6 .il W e Department of Energy Washington, D.C. 20545 SEP 9 o 1932 Docket No. 50-537 HQ:S:82:096 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Mr. Check:

TMSDB INSTRUMENTATION DEVELOPMENT Letter HQ:S:82:028 J. R. Longenecker to P. S. Check.

Reference:

" Future Information for Review of CRBRP-3, Volume 2 " O dated May 14, 1982 in the above reference, a list of topics and reports to be submitted to the NRC in support of the CRBRP Thermal Margins Beyond the Design Base (TMBDB) was supplied. This letter transmits item 1 of the list in the above reference. Sincerely, 1. A l J hn R. Longen ker Acting Director, Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy i G ~ K.1-2

) ' CRBRP-3 VOI 2, Rev. 6 O LETTER REPORT TF3DB INSTRUME,NTATION DEVELOPMENT SEPTEPSER 20, 1982 O e O K.1-3

? CRBRP-3 Vol. 2, Rev. 6 TMBDB INSTRUMENTATION DEV$LOPMENT I. INTRODUCTION TMBDB instrumentation requirements for monitoring containment vessel shell temperature, containment atmosphere temperature, pressure, and hydrogen content are defined in CRBRP-3, Vol. 2. The purpose of thir report is to describe the development tests for these instruments and to demonstrate that it is feasible to develop instrumentation that meets the requirements in CR'BRP-3, Vol. 2. II. SIGNAL CONDITIONING EQUIPMENT All signal conditioning equipment is located in the Steam Generator Building. The Steam Generator Building is not subjected to the products of sodium combustion or to 'the extreme temperatures and pressures whichcharacterize the containment. The most severe TMBDB environmental concern in the Steam Generator Building arises from the radiation levels emanating from the containment. While these levels are not suf ficient to affect the operation of equipment, they will limit operator access.* Unattended operation is a requirement and the systems have been designed for remote manual actuation from the Control Room. III. HYDROGEN MONITORING INSTRUMENTATION Two instruments are provided at different locations in the Steam Generator Building. The principal operating requirements of the Hydrogen Monitoring System under TMBDB conditions are listed in Table 1.The cabinets may be operated locally or remotely from the Centrol Room. Remote control features include actuation, calibration ard filter blovback. The principal concern with this system is the containment sampling arrangement. To prevent plugging of the sampling tube, a filter assembly will be mounted at the entry end. The development of the filtering system has been in three phases: The first, or scoping phase, completed in 1980, aimed at identifying basic problems of filtering sodium aerosols; cat the Hydrogen Analyzer Sampling Station, the maximum whole body dose has been calculated to be =280 mrem assuming 2 minutes for ingress and egress with a 2 minute stay. These doses fall off rapidly following venting to 60 mrem for the same time interval at 50 hours. K.1-4

. CRBRP-3 Vol. 2, Rev 6 4500g 2 AC6 40 = 190 1.2 NaOH/1.0 H O > 5000g 2 Each test lasted for about 50 hours and different filters were used. As the tests progressed, less ef fective tilters were deleted and new types w'ere tested, based on developir.g expericnce. Whan increases ir, pressure were observed, various blowback techniques were used to clean the filters. The principal conclusions from the tests were: Aerosol composition is the dominating factor controlling filter loading. 3 and Na202 aerosols are filterable and filters Joaded with Na2 CO these aerosols are readily blown back. Wet aerosols tend to plug filters and the plugged filters are dif ficult to blow back. The sample gas flowrate probably does not af fect filter loading at velocities of 18 m/hr and below. The filter configuration and orientation are important only with respect to, blowback and aerosol settling and plating. ~ (t,,/ Use of settling chambers improves filtering ef ficiency. I K.1-5

CRBRP-3 Vol. 2 Rev. 6 0 \\ 1 From these conclusions, a further series of tests were planned based on the TMBDB limiting containment atmospheric conditions. The presence of sodium hydroxide in liquid state in these tests provided more severe plugging and corrosion problems than were experienced in the scoping tests. MEDIA TESTS To determine the performance of specific filter. media under TMBDB conditions, seven tests, designated HFT3 and HFT9, were conducted at the Large Sod ium Fire Facility (LSFF). The te s t me thod wa s to burn sodium vapors in a heated chamber, which held the test filters. Air flow through the chamber was controlled and steam and CO were added 2 to control test atmospheres and aerosol composition. Tablas 2 and 3 summarize the conditions of the HPT series of tests. In a typical test, test conditions were first established. The test filter was blown back with nitrogen to clean the surf ace. Flow through the filter was started and the filter pressure drop monitored. When the pressure drop reached 100 in. H 0, a brief 2 blowback with nitrogen gas was used to clear the filters, then sample flow ~was resumed. The filters were considered to be plugged when the pressure drop could not be reduced below 100 in. H O by 2 blowback. The duration of each test was approximately 10 hours. Total loading on each filter was calculated from the known filter flow and the measured aerosol concentration. The results of the tests are summarized in Table 4. In this table, the filter loading is for the first blowcack cycle only. The following conclusions were reached: Filter loadings of 1200 g/m2' can be expected for the worst case when moist NaOH aerosol filter deposits are heated 'o temperatures above the NaOH melting point. Filter loading for " dry" aerosols or sodium carbonate aerosolr 2 will be 24 00 g/m or greater. Filter efficiencies above 99% can be obtained. Fibrous types of filter media generally provide the highest loadings before plugging, followed by sintered powder media. Corrosion will occur for stainless steel media exposed to molten NaOH aerosols. Corrosion will be less for nickel powder media. Cleaning the filters by blowback generally will extend the filter life. Cleaning was most successful for fibrous media and for the dry type of aerosols. K.1-6 d c

CRBRP-3 Vol. 2,'ev. 6 R l O Several of the test filters ruptured on blowback due to the combined ef fects of temperature and corrosion. Filters should be designed for the application. I 1 PROTOTYPE TEST flow The capacity requirement of the filter system is based on the requirements of the hydrogen instrument, the length of time for which is required, and the aerosol concentration. In the case measurement of the CRBRP requirements listed in Table 1, the total aerosol capacity required by the sampling filter should be less than 1.4 kg. filter media performance determined in the previous test Based on the and a the prototype design will utilize nickel powder filters,

phase, settling chamber which significantly reduces the filter loading.

The prototype units will be tested under conditions of the TMBDB environment and the load and time exposure associated with TMBDB scenario. Present plans are for the prototype tests to start before the end of FY82 and full system tests to be completed in FY83. f With the availability of blowback and the capability of increasing ) sampling filter unit capacity and/or utilizing a backup filter, successful development of this component poses minimal risk. I V.' PRESSURE AND TEMPERATURE SENSOR' PERFORMANCE The successful performance of conventional thermocouples and pressure sensors over the extended test program period provides a basis for the survivability of equivalent containment insttunients under TMBD3 cen6itions, as discussed below. Temperature Measurement Temperature measurements in CSTF and LSFF are made using seamless stainless steel sheathed, M O insulated, Type K thermocouples, generally with ungrounded junctions. Where possible, continuous lengths are used inside the test vessel. In the CSTF, 1/16" OD thermocouples were used for 14 test runs. The temperature varied from 100 to 400*F in sodium oxide and hydroxide aerosols. The total exposure time was about 300 hours and the few failures observed have been from mechanically breaking the thermocouples while revising the test arrangements. The 1/16" OD stainless sheathed thermocouples were satisf actory for this service. O K.1-7 I

CRBRP-3 Vol. 2, Rev. 6 O In the case of the LSFF tests, 1/8" OD thermocouples are being used. These thermocouples have now survived 100 hours of testing without failure in average temperatures between 800 and 900* F and with peaks of 1200* F. The performance of the test atmosphere thermocouples provides assurance that the existing containment atmosphere and ve'ssel shell temperature thermocouples will operate adequately. Pressure Measurement Pressure measurements within the various experimental vessels have used Bourdon gauges and diaphram gauges. These are connected to the vessel.shell with 1/4 to 1/2" stainless steel tubing. Lengths vary from 10 to 50 feet. During hundreds of hours of testing with various aerosols, plugging of the sensor tube has not occurred probably due to the small pressure sensor system volume and the very small gas displacement required to reflect pressure changes. The design and performance of these sensors again provides assurance that existing pressure measurement technology is adequate for TMBDB containment measurement. K.1-8

CRBRP-3 . Volt 2, Rev. 6 O TABLE 1 Y HYDRDGEN' SAMPLING SiSTEM REQUIREMENT ~ Sample System Dela Time *I 10 minutes (max) I I Sample Line size " 61 m (200 ft),long 6.3 mm (0.25-in) ID Filter Pressure Drop, Max *I 34 kPa (5 psi) I Containment Temperature 16

• C to 59 3
• C ( 60
• F to 1100
• F)

Containment Atmosphere 0-10 v/o H 0, 0-6 v/o CO2 2 Operation Duration 500 hrs (with aerosol) 8000 hr total 3 Aerosol Concentration 46 g/m (at CV conditions) Aerosol Composition Na 0 NaOH, and Na CO 2 2, 2 3 Aerosol Size

• AMMD = 5 pm, og = 3.0 Required Instrument Flow *I 150 cc/ min (minimum) at 150*C (300*F)

I () Filter Efficiency ("I Sufficient to protect system i (a) Adapted f rc:a CRBRP-3, Vol. 2 Aerodynamic Mass Median Diameter O K.1-9 -s.

? --- q TADLE 2 IIYDROCEN FILTER TEST CONDITIONS Condition when Peak Conditions Filter Started Aerosol Characteristics Major Steam Aerogol Aerogol AMMD( b) Chemical g/m g/m Temp g (STP)(a) Temp Test or CO2

• C (STP)
• C pm g

Form Notes No. Addition .FT 3 S 107 2,00 44 100 2.9 2.3 Na0llaxif 0 Dase case 2 b HPT 4 S 175 250 63 75 2.2 2.0 Na0lf xil 0 Higher concentration 2

• 9 n

IIFT 5 S 17G 650 58 350 5.6 2.1 Na0ii (c) Effect of molten NaOH .N g O

o e o

HPT 6 S 188 650 66 245 6.3 1.9 Na0ti Effect of molten Na0ll W t 0 /Na0H Lw temp start to molten NaOH o IIFT 7 146 440 20 76 3.0 2.1 Ma2 2 1 ) addition Na CO /NaOH, CO IIFT 8 S, CO2 166 660 40 388 5.1 2.1 2 3 2 IIPr 9 S 166 620 154 330 4.8. 2.1 Na0ti fligh concentration start g (a) STP; 0*C, 101 kPa (b) Aerodynamic mass median diameter. Molten Na0ll, MP - 318'c (c) Na0li = g O O O l t

( i 1 TADLE 3 IIFT AVERACE ATHOSPlfERE COMPOSITION Test Chamber Air oxygen Steam CO, 3 No. Flow, m / min (STP) Vol t g/ min Vol t II,0 T/ min vol t HFT 3 1.0 19 85 11 0 0 ~ p sFT 4 0.6 17 85 18 0 0 g w 4 HPT 5 0.7 17 85 15 0 0 Q e. N to HPT 6 1.0 17 85 11 0 0 4 x, I *I 0 0 I HPT 7 0.6 17 0 HFT 8 1.1 19 85' 10 110 5 m HrT 9 0.8 20 85 13 0 0 } (a) Est. inlet air dew point of 10*C I i 1 4 4 i 4 4i

A i TABLE 4, IIFT RESULTS SUMMARI2ED DY FILTER MEDIA 2 AVERAGE FILTER AEROSOL FIRST CYCLE LOADINGS, g/m Media Type itFT3 IlFT4 ffFT5 ffFT6 IlFT7 IIFT8 !!PT9 4o Powder 2590 3140 600 740 1110 1300 N

x2 Po' der, Pleated 1600 1970 240

.N E w , ~o 1280 7030 $b b Screen 460 1740 2190 N ~ Fiber, Filterite 2610 3170 1280 9160 880 1090 m 1790 320 1020 Fiber, HR '450 1400 9950 Nickel Powder Nickel Screen 1810 O O O

r l CRBRP-3 Vol: 2,*Rev. 6 l K.2 TMBDB MARGINS ASSESSMENT DOCUMENT J This appendix contains the DOE letter supplied to the NRC, dated L October 20, 1982. L The radiological consequences section has been updated to present results i of dose calculations that are consistent with the latest methodology for gas sparging and plutonium release as presented in Section 4, Appendix G.4, l and Appendix E. This appendix is annotated to identify changes from the October 20, 1982 DOE letter. 4 6 l i j l 4 I K.2-1 4

CRBRP-3 Vol. 2,, Rey. 6 CQl n O Department of Energy Washington, D.C. 20545 Docket No. 50-537 2 013B'A HQ:S:82:112 OCT Mr. Paul S. Check Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Mr. Check:

THERMAL MARGIN BEYOND DESIGN BASE (TMBDB) MARGINS ASSESSMENT DOCUMENT Enclosed is a document entitled "TMBDB Sodium - Concrete Penetration Margins Assessment for the Clinch River Breeder Reactor Plant (CRBRP)." The information in this document was presented to the NRC during the September 15, 1982, meeti ng. This document is being sent to you for your early review and will be incorporated into CRBRP-3, Volume 2 "TMBDB" in the future. Sincerely, n John R. Longenecker b Acting Director Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy O 1 K.2-2

CRBRP-3 Vol. 2 Rev. 6 'INBDB SODIUM-ONCRETTE PENEIRATICN MMGINS ASSEESMENT FOR ' DIE CRBRP by T. W. Ball J. F. Gross D. P. Koshurba R. G. Vasey G. Freskakis (B&R) August 1982 i K.2-3

CRBRP-3 Vol. 2, Rev. 6 'INBDB SODIUM-CONCRET. E PENETIRATIm MARGINS ASSESSMENT EDR 'IHE CRBRP Table of Contents Pace Introduction............................. K.2-5 K.2-7 Model and Assumptions Resulta K.2-8 Radiological and Aerosol Consequences K.2-10 K.2-11 Structural Consequences Conclusions K.2-12 References.............................. K.2-13 Accendix A Annulus Cooling Analysis....................... K.2-31 Results K.2-31 Accendix B Structural Assessments........................ K.2-45 Reactor Cavity and Liner....................... K.2-46 Pipeway Cell Walls.......................... K.2-47 Pipeway Cell Floor and Liner..................... K.2-48 Confinement Structure K.2-48 Containment Steel Shell K.2-49 References.............................. K.2-49 O K.2-4

CRBRP-3 Vol. 2, Rev. 6 s 'INBDB SODIUM-CXECRETE PENETRATICN MARGINS ASSESSMENT INIRCDUCTICN Experiments (References 1 and 2) have indicated that sodium-concrete reactions tend to be self-limiting with limestone concrete under conditions prevailing during a postulated 'INBDB scenario in CRBRP. 'Ihe sodium-concrete penetration was represented in the base case CACE00 analysis as a constant reaction penetration of 0.5 inch per hour for a period of 4 hours (Reference 3). Sensitivity studies (Reference 3) show that a reaction penetration rate of 1 inch per hour for 12 hours (12 inches total penetration versus 2 inches in the base case) can be arw== lated by the 'INBDB containment features based on their design requirements. S ee sodium-concrete experiments have exhibited considerably higher penetration rates (up to approximately 0.2 inches per minute) for short b periods of time (References 4 and 5). For all tests that had an excess of V sodium, the rapid reaction penetration was not sustained for more than a few minutes. A@arently the reaction penetration was self-limiting after a short period of rapid penetration. In other tests, where only a limited amount of sodim was used, the available sodium was all consumed in a short time if rapid penetration cccurred (in many tests virtually no penetration at all occurred, and very little sodim was consumed). In all tests to date, the total sodim reaction penetration has not exceeded the energy and hydrogen releases associated with the 1 inch per hour for 12 hours 'used in the sensitivity calculations. Frm the large body of available data, Reference 6 states that sodim-concrete reaction penetration into horizontal surfaces appears to progress in three stages, each characterized by successively decreasing rates of penetration. These three stages and the recomended upper bound rates and durations are as follows: A. An initial rapid penetration probably due to spallation and breakup of the surface layer of concrete. The upper bound rate and duration of this stage is 7 inches per hour for 20 minutes. (m\\ K.2-5

) CRBRP-3 Vol. 2 Rev. 6 0 B. An intermediate stage where spallation is no longer occurring but the ) concrete surface is not fully protected by the developing layer of reaction products. The upper bound for this stage is 1 inch per hour for 3 hours. C. In the longer term (the third stage) the rate of penetration is much slower and decreases with time because of a continually growing layer of reaction products which inhibits transport of unreacted sodium to the unreacted concrete surface. Upper bound-0.1 incVhour indefinitely. Based on the above mentioned analysis of the data and the analyses of 'IMBDB sequences in Reference 3, it is apparent that the sensitivity analyses adquately cover the range of data observed to date. It is also apparent that considerable margin exists in 'INBDB analyses for accommodating higher sodium-concrete reactions at the expense of vent time (less than the base case 36 hours). With a view to assessing the margins available in the current CRBRP design relative to assumed variations in the sodium-concrete penetration rate, a margin study was conducted. In assessing the margins available in the design, two distinct considerations energe: A. If there were no rquirement to maintain antainment integrity for a fixed time without venting, what range of reaction rates could be accanmodated by the design? B. There must be some period of time allowed for venting decisions, activation of emergency plans, etc. How early could venting reasonably be assumed, and would such an assumption give sufficient margin to cover any remaining uncertainties regarding sodium-concrete reactions? l To assess these considerations a study was initiated with following objectives: A. Artificially contrive a sodium-concrete reaction scenario which would so far exceed any observed in tests to date so as to result in a need O K.2-6

CRBRP-3 Vol. 2, Rev. 6 O V to vent contaiment in 10 hours. (This is compatible with evacuation times quoted in Chapter 13 of the PSAR.) B. Assess the capability of the design to accomodate such a scenario. This report documents the analysis of this margins assessment which is a scenario with an initial sodium concrete penetration rate of 7 inches / hour for 3 hours, followed by a rate of 1 incyhour. For this case, no credit is taken for the reaction inhibiting effects of the reaction product layer, such that the 1 incVhour reaction was assumed to continue until all sodium was consumed. It is emphasized that this artificially contrived case does not represent test data, but is simply a margin study to assess the design capability. MODEL AND ASSUMPTIONS The CACECO code model defined in Reference 3 (Appendix C.1) was modified to perform this sensitivity study. The model used in the base case includes a sodium-concrete reaction rate of 0.5 inches per hour for 4 hours, starting at the time of reactor cavity liner failure (assumed to be at the time of penetration of the reactor vessel and guard vessel). Tb calculate the containment conditions for the penetration margins assessment, the CACECO model had to be modified to include the more severe reactions associated with the margins assessment case. 'Ihe criteria for venting were similar to those for the base case scenario and, as in the base case, the RCB hydrogen concentration was the limiting factor on vent time. Other design requirements (heat load to the confinement building, venting rates, aerosol discharge to the cleanup system, etc.) were not imposed on the assumption that the 'INBDB features external to the RCB could be enhanced, if necessary, to accomodate greater operational loads. Other variations from the base case scenario included were: 1) the soditn-concrete penetration rate for the pipeway floor concrete is the same as for the RC floor (with the exception that the reaction occurs only when a sodita pool exists; 2) initiation of the RCB annulus cooling system at 10 hours; 3) increasing the thermal conductivity of the concrete nodes as the assumed penetration front moved past the node to simulate movment of the sodium boundary (the node thickness K.2-7

CRBRP-3 Vol. 2 Rev. 6 0 used in the region of concrete penetration was two inches); 4) consumption of 3 3 sodium by reaction with the concrete residue (0.38 ft sodium per ft concrete) in addition to reactions with the water and carbon dioxide driven off frm the concrete. (In the base case 'IMBDB, the reaction energy 331 Btu /lb concrete, was accounted for, but sodium consumption from this source was ignored due to the small amount of concrete involved); and 5) RCB purge initiated at 13.5 hours (immediately after blowdown to atmospheric pressure). RESUL'IS The results confirm that with an initial sodium concrete penetration rate into the reactor cavity floor of seven inches per hour for the first three hours and 1 inch per hour thereafter, R G venting would be required at 10 hours. Considerations which are important during the 'IMBDB scenario include the consequences from the initial hydrogen ignition, conditions which cause the initiation of RCB venting, and long term or maximum RCB conditions during vcating. These results are summarized in Table 1. Due to the additional energy fr m the more severe sodium-concrete reactions causing the sodium pool to heat up faster and allowing sodium vapor to enter containment sooner, the criteria for initial hydrogen ignition are met at 1.4 hours as empared to 10 hours in the base case. The corresponding hydrogen build-up in containment prior to ignition is less than in the base case (2.5% versus 4.5%). Upon ignition, the containment responses would be less severe. The resulting containment temperature and pressure would be 5700F and 13.9 psig as empared l to 8450F and 22.4 psig for the base case. l The initiation of RCB venting is the next important consideration. The criteria that can dictate venting are excessive RCB pressure and steel shell temperature, or excessive hydrogen buildup. In this case the hydrogen buildup was the limiting condition causing venting to occur at 10 hours. With the sodium and hydrogen rates into containment averaging about 4000 lb/hr and 350 lb/hr, respectively, over the first ten hours, the containment oxygen was depleted to below 8% due to the chmical reactions. After the oxygen is depleted to below 8% in containment (8.5 hours) the hydrogen is O K.2-8

CRBRP-3 Vol. 2, Rev. 6 predicted to accumulate due to incomplete burning. As in the base case 'INBDB sequence, venting would be initiated well in advance of reaching the hydropn limit because the depressurization of containment results in hydrogen flowing up from the reactor cavity, which in turn causes the hydrogen concentration to increase during the blowdown period before purging can be initiated. (For purging, the RCB must be at a negative pressure.) In the base case the 3 resulting peak hydrogen concentration is 4.5%, well below the objective maxim a of 6.0. In this case the calculated hydrogen concentration has reached 2.6% at 10 hours when venting is initiated. %e blowdown excursion exceeded 6 for approximately 2 hours. (%e peak was 8.7% at 13.5 hours.) During the time hydrogen was above 6%, the oxygen was below 5% so that the i mixture in contaiment would not be flmmable (calculationally, the reason the hydrogen exceeded 6% was because there is not enough oxygen to meet burning criteria). Wis short excursion beyond 6% hydrogen is considered acceptable for this margins assessment case since it occurs at a time when j RCB oxygen has been depleted. Shortly after purging was initiated, the incoming air raised the oxygen concentration to 5% at which time the hydre/jen in excess of 4% burned with acceptable consequences to containment conditicns. Figure 1 shows the hydrogen and oxygen concentrations as a ] function of time. The peak containment pressure was higher at venting than i the base case (18.7 versus 13.1 psig), but well within the ultimate pressure capability of the steel shell (Reference 3). % e reactor cavity and contalment atmosphere temperatures are shown in Figure 2. We containment temperature is several hundred degrees higher than the base case due to the increased rate of sodium burning. We higher contalment atmosphere temperature also results in a higher steel shell temperature than in the base case (4900F versus 4000F baad on the one-dimensional CACE00 calculations). The steel temperature is shown on Figure 2. Figure 3 shows the reactor contaiment pressure which is higher than found in the base case because of the increased sodim burning and more energetic concrete reactions. Additionally, the heat loads to the RCB steel shell for the base case and for the margins assessment case are shown on Figure 4. While the peak flux to the steel shell is not significantly different from that observed for the l base case, it occurs earlier in time and is sustained for a longer period. Sodim boildry occurs at 51 hours for this case as compared to 133 hours for the base case. %e amounts of sodim aerosols ingested into the cleanup i' K.2-9

o CRBRP-3 Vol. 2, Rev. 6 system are discussed in the Radiological and Aerosol Consequences section below. Figures 5 to 15 show structural tenperatures for the various cavity and pipeway floors and, walls..The total penetration of sodium into the reactor cavity floor is 5.7 feet while the pipeway floor is penetrated approximately 2 feet. These temperatures are used in Appendix B to assess the integrity of the structures. The effects of the heat loads on the steel containment shell and concrete confinement structure were assessed and the results are shown in Appendix A. RADIOLOGICAL AND AEROSOL CCNSIOUENCES 2 Hour Exclusion Boundary Doses The doses resulting from the margins assessment case (10 hour vent) scenario are compared with the '1NBDB base case (36 hour vent) doses in Table 2.* The penetration margins assessment scenario produces a higher RCB pressure for the first 2 hours resulting in a greater release rate of the initially suspended 1000 lbs. of sodium and noble gases. The larger leakage during this time interval produces the higher 2 lx)ur Exclusion Boundary doses for the margins assessment case compared to the base case. 30 Day Iow Poculation Zone Doses ** Two different effects are seen in the 30 day LPZ doses. The first effect results from the increase in early leakage. The whole body dose is largely dep3ndent on external gamma exposure from the released noble gases. The higher RCB pressure and early vent time release more of these gases early in the event, when the calculated radiological impact is greater. The result is an LPZ dose about 4 times greater than the base case. The liver dose, which is dependent on inhalation of noble gases, also increases. The second effect influences the remaining organ doses. For example, the thyroid dose decreases from 85.4 rem in the base case to 81.8.

• This comparison used Case 2 from Table 4-3 of Reference 3 as the base case.
  • This section and the calculated dose values have been updated (see page K.2-1).

K.2-10

I' CRBRP-3 Vol. 2, Rev. 6 O rem for the penetration margins assessment case. W is decrease is attributed to the fact that this case has higher aerosol concrentrations, resulting in l greater overall agglomeration and fallout than the base case. %e lung dose is heavily dependent on the solid fission products which are not released early in the event during more radiologically significant time intervals. However, the greater suspended aerosol concentrations produce a higher RCB agglomeration and fallout rate with less release to the environment. Consequently, the net effect on the lung dose is very small and it is essentially the same as the base case dose. % e higher aerosol concentrations in the penetration margins assessment case result in greater aerosol depletion and somewhat less aerosol discharge to the RCB clean-up system. Table 1 provides a comparison of the total aerosol transported into the clean-up system and the rate of transport. SDUCIURAL ONSEQUENCES % e consequences of the penetration margins assessment case on containment and confinement structures have been assessed and the details are presented l in Appendix B. The results indicate that design changes would be required in two areas of the structures to ace -hte the margins assessment case. We required modifications are: A. Reactor Cavity Floor (1) Extend the wall liner to 6.5 feet into the floor structural concrete., (2) Eliminate the construction joint at Elevation 733 and rearrange the rebar in the floor. (3) Provide a design feature on the RC floor liner near the wall to inhibit the spreading of the duel debris to the region of the wall-floor junction.

O K.2-11 l

CRDRP-3 Vol. 2, Rev. 6 B. Pipeway Cell Floors (1) Provide a second layer of insulating concrete below the second liner which separates the two layers of structural concrete. (2) Increase the thickness of the floor by the thickness of the second layer of insulating concrete (lower bottom). These modifications are not considered major changes and would, in conjunction with other existing TMBDB features, result in acceptable 'INBDB margins to accomodate the margins assessment case. CCNCLUSIONS Base on this margins assessment, the following conclusions have been reached: A. The sodium-concrete penetration rate for the margins assessment case, which is 7 inches /hr for 3 hours followed by 1 incyhr thereafter until sodium boildry, could be accommodated by venting the containment at about 10 hours. B. Hydrogen ignition would occur earlier and would result in less severe containment conditions following rapid burning. C. The clean-up system peak flow rate is somewhat higher than the base case, but below the design basis value for the system. The total aerosol loading would be less than the base case. l D. The sodium boildry time is re3uced to 51 hours empared to 133 hours in the base case. 1 E. With modest design modifications in two areas (the Reactor Cavity floor and the pipeway cell floors) the margins assessment scenario would result in acceptable containment and confinement structural margins. O K.2-12

CRBRP-3 Vol. 2, Rev. 6 F. Radiological doses are comparable to the base case except the 30 day LPZ whole body dose which is greater than the base case due to the earlier venting, but still within the guideline values for this beyond design bam event. REFERENCES 1. J. A. Hassberger, R. K. Hilliard and L. D. Muhlestein, " Sodium Concrete Reactions Tests," HHXe'INE-74-36, June 1974 2. J. A. Hassberger, " Intermediate Scale Sodium-Concrete Reaction Tests," HEIL ' DIE-77-99, March 1978 3. CRBRP-3, Volume 2, " Hypothetical Core Disruptive Accident Considerations in CRBRP, Voltane 2 Assessment of Thermal Margin Beyond the Design Base," March 1980 [ 4. B. M. Butcher, et al., " Sodium Containment and Structural Integrity," NJREG-0181-3, Advanced Safety Research Program Quarterly Report, April-June,1977, SAND 77-1134 Sandia National Laboratories, Albuquerque, IM, pp. 145-156, November 1977 5. D. A. Dahlgren, et al., "Soditan Contairunent and Structural Integrity," Advanced Reactor Safety Research Program Quarterly Report, July-September,1977, SAND 77-l!T15 Sandia National Laboratories, Albuquerque, N4, pp.111-118, May 1978 6. HEIL 'IME 82-15, L. D. Muhlstein and A. K. Postma, "SoditmH%mcrete Reaction Executive Stanary Report: Application to Limestone Concrete," June 1982 llO K.2-13 1

CRBRP-3 Vol. 2, Rev. 6 TABLE 1

SUMMARY

OF RESULTS WITH INITIAL SODIUM-CONCRETE PENETRATION OF SEVEN INCHES PER HOUR Penetration Base Margins Case Assessment initial Hydrogen Ignition Time (hrs.) 10.0 1.4 RCB Atmosphere Temperature (*F) (before/after) 120/845 145/570 RCB Pressure (psig) (before/after) 2.2/22.4 2.4/13.9 Hydrogen Concentration (Vol.%) (before/after) 4.5/0.0 2.5/0.0 Initiation of RCB Venting Time (hrs.) 36 10 RCB Atmosphere Temperature ('F) 617 710 RCB Steel Shell Temperature ('F) 400 390 RCB Pressure (psig) 13.1 18.7 RCB Hydrogen Concentration (%) 0.0 2.6 RCB 0xygen Concentration (%) 8.4 7.4 Maximum Conditions During Venting Maximum Venting Rate (ACFM) 24,000 27,500 Purge Rate Assumed (SCFM) 8000 8000 Peak Hydrogen Concentration (Vol.%)/ Time (hr.) 4.0/40 8.7/13.5 RCB Atmosphere Temperature (*F)/ Time (hr.) 915/40 1020/14.6 Aerosol Comparisons j Maximum Rate to the RCB Cleanup System (Ib/hr) 4400 5100 Total Aerosols to the RCB Cleanup System to 260,000 167,000 Boildry (lb) O K.2-14

l i i CRBRP-3 Vol. 2, Rev. 6 i O l i TABLE 2 COMPARISON OF RADIOLOGICAL CONSEQUENCES 2_ Hour EB Doses (rem)* 36 Hour Vent 10 Hour Vent Organ Base Case Penetration Margin Assessment i Bone Surface 0.19 0.36 ) Red Bone Marrow 0.040 0.074 Lung 0.032 0.055 Liver 0.060 0.11 Thyroid 0.020 0.025 L Whole Body 0.82 1.38 O 30 Day LPZ Doses (rem)* 36 Hour Vent 10 Hour Vent Organ Base Case Penetration Margin Assessment Bone Surface 0.95 0.86 Red Bone Marrow 0.19 0.19 Lung 1.55 1.50 Liver 0.36 0.42 Thyroid 85.4 81.8 Whole Body 2.09 9.63

• Calculated dose values have been updated (see page K.2-1).

U O K.2-15

l CRBRP-3 l l Vol. 2, Rev. 6 i O i l 20.0 l i I 16.0 l 3 12.0 E 'e E 0xygen 8.0 19ll h 5 Is ~ I I I I 4.0 - / / Hydrogen ~ /g I l / / 1 !I I I I l 0.0 O 10 20 30 40 50 Time (Hrs.) l l Figure 1 1 RCB Hydrogen and Oxygen Concentration 1 K.2-16

CRBRP-3 Vol. 2, Rev. 6 O 1800 Reactor Cavity Atmosphere s - s. / 1600 - [ I / / 1200 ' g Containment Atmosphere '~~ ~~~~~*w u l f 800 l N ) U / Contairamnt Steel Dome Il - f, i/ 400 la I I -l gl l l I l I 1 i 0.0 0 10 20 30 40 50 Time (Hrs.) Figure 2 Reactor Cavity and Containtnent Atmosphere and l ( Containment Steel Dome Temperature K.2-17 l l

l i i CRBRP-3 Vol. 2, Rev. 6 l O 24 20 4/ I / I 16 / g / I i / l O -l, f / I 2 i g 'l I l e l I I/ \\ c. 8 l 11 / I lIl l II l 4 -l t i \\ / \\ \\ / 0 L----------------- i I I i 1 I I 0 10 20-30 40 50 Time (Hrs.) Figure 3 Containment Pressure O K.2-18

I CRBRP-3 Vol. 2, Rev. 6 O 1 12 i f 11 - ( \\ l\\ 10 \\ \\ 9 \\ Penetration Margins l 8 - \\ Assessment Case q xy7 - 4 \\ Base f.,ase O C 6 c a 5 -) y 4 ) l 3 I 2 1 j i I I I I I I I i 0 l 0 20 40 60 80 100 120 140 160 Time (Hours) O Figure 4 Heat Load Through the Containment Structure Steel Shell -K.2-19 . _.. ~...,..,. _........ _,.,.., _ _, _ _. ~....,

CRBRP-3 Vol. 2, Rev. 6 O Figure 5 i REACTOR CAVITV FLOOR (from liner to 3.6 ft.) LEGEND l>00. 0 Hours Y*:-y, I600- ^~'"# ~' 10 _ _ - - - ~.....:-__ l i .,,, ~ 20 1900. N... j I---- 30 1 I stos. -~ 40 l u. I i 50 8 1000. j

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CRBRP-3 Vol. 2, Rev. 6 OG Figure 6 I l 1 1 1 REACT 0R CAVITY FL00R (from 3.6 to 7.6 ft.) LEGEND h_, 0 Hours ,_i

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g l 10 g i s j l l -------- 20 l 30 l i iroo. l 40 i C i ) L i i 50 I ) e i s) 5 aoo. u i ,\\ i l 2 w i \\ it i .\\ ,l =* 600. L \\ i i i \\ j i \\ 900. g i \\ i \\ N i s .;, - -...'.s,..'_'? _ _ ' __ ~ - N toc. 6 s i i I G s.A 9.o 9.9 9.8 9.2 4.6 6.0 6.9 6.8 F.2 F.6 DISTANCE (FT) ) K.2-21

CRBRP-3 Vol. 2, Rev. 6 O Figure 7 REACTOR CAVITY FLOOR (from 7.6 to 11.6 ft.) LEGEND O Hours sim s 10 lrt % 20 30 i,. % 40 4:

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CRBRP-3 Vol. 2, Rev. 6 Figure 8 i l l l REACTOR CAVITY NONSUBMERGED WALL LEGEND l a r,0. - 0 Hours I L i co 10 gg l l i oc. 30 40 i200. 50 O L 8000. E B 2 80G. cL 8 600. =00. ^ '.'s\\ '.'N '.\\ - *.s% /00. '.. = = _ _ _ _ _ _ _ _ = 's 1 t, e, 1.0 2.0 3.0 9.0 5.0 6.0 F0 l l DISTANCE FT) l } m K.2-23 i

CRBRP-3 Vol. 2, Rev. 6 l h l Figure 9 l REACTOR CAVITY SUBMERGED WALL (upper) LEGEND IMOG. 0 Hours L i>00. 10 ........ 20 l*00.

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CRBRP-3 Vol. 2, Rev. 6 (' %J Figure 10 REACTOR CAVITY LOWER SUBMERGED WALL LEGEND tr 00 0 Hours L 10 i 00. 20 1400. 30 0 00. -! 40 C 50 8 1000. Oi E t00. --- N i 600. t + l s\\

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CRBRP-3 Vol. 2, Rev. 6 Figure 11 REACTOR CAV1TY PIPEWAY CELL FLOOR (from liner to 4 ft.) LEGEND O Hours sasc. \\ 10 ~_ :_ 4 _ _4___ 1600. 20 .\\ i \\ 30 anoo. \\ 40 \\ 20c. 50 o' s \\ s t e s. toco \\ .o. m \\ \\ l u E. \\ e aco. g s o \\ 6-g \\- i \\ I '\\ -s ?- eco. s i \\; '\\ \\ s s. co. g v, s N '). 's N m ..--p'~~..:*~ t p,0. .i.. ts .o e o, o.n i.i i.e 2.0 2.* 2.s L2 3.6

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O K.2-26

CRBRP-3 i Vol. 2, Rev. 6 mi Figure 12 REACTOR CAVITY PIPEWAY CELL ROOF LEGEND 0 Hours 1600. L 10 20 i,00. 30 1200. 40 m E 50 g 1000. a L b .L-800. it i i* 600. -l

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CRBRP-3 Vol. 2 Rev. 6 0 Figure 13 REACTOR CAVITY PIPEWAY CELL OUTSIDE WALLS-2.5 FT. THICK LEGEND 0 Hours 1600. 1 4 10 L 20 i.00. T 30 g u00. 40 t

\\

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L f, 0.2 0.9 0.6 0.A 1.0 1.2 1.9 1.6 1.8 2.0 2.2 2.9 2.6 OISTANCE (FT) l O K.2-28

CRBRP-3 Vol. 2, Rev. 6 A (s Figure 14 REACTOR CAVITY PIPEWAY CELL OUTSIDE WALLS-4 FT. THICK LEGEND a st,. 1 10 20 30 lico \\ 40 \\ 4W" m i 50 o ice o. m u 3 k l 80 4 L.o AGC a. i *.i e ss ..\\ e-i.i i eco $h 1

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CRBRP-3 Vol. 2, Rev. 6 Figure 15 REACTOR CAVITY PIPEWAY CELL DOUBLE HEATED WALL LEGEND l=00. 0 Hours f I 10 i>00. q 20 4 t=00. 33 t ~ ~ 40 1200. -u. i o 1 co as 5 8000. l u o m L. 1 C# aey co0. 4 600. i i.: / er i .- %m ",, '; p '00-A.,s,- ,,.i* ,.. ' ~?W -...l '..:s.- ~,'" : -{.,{... i.~. ** * * ~ i es0. -~......... i o ') G 0.9

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9 K.2-30

CRBRP-3 Vol. 2, Rev. 6 O APPENDIX A TO APPENDIX K.2 S0DIUM-CONCRETE PENETRATION MARGINS ASSESSMENT ANNULUS COOLING ANALYSIS The extreme sodium-concrete penetration case results, i.e., heat loads to the containment steel shell as a function of time, were used as input to a detailed thermal calculation using the TRUMP computer code. The thermal model was identical to that of Reference 3 except an updated model was used in the l vicinity of the annulus cooling outlet structures. The nodal arrangement of 1he thermal model is shown in Figure A1. An annulus cooling system flow of 400,000 SCFM was used as was in the base case. RESULTS The temperature versus tima plote, for key tones are presented in Figures A2 through A13. The structural corsequences of these temperatures are described in Aprendix 8. O O K.2-31

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081000f 81980s.10002 AND 10B83 ARE SOUSOART $0MS I l l Figure Al Annulus Cooling System Thennal Model 36591 O K.2-32

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[ f _.. iii. 1 / ) . - = = i.. i f t y/

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== ais / y, ~s i.. ,% ~ ~ N.., I a' O 'l l. ',/c-Q{.%,,. f 'l s su. x 4 i N% '"' OMV ~ n. n. in. n ass.o.as Figure A4 Nodes 1313-2013 Extreme Penetration Case O K.2-35

CRBRP-3 Vol. 2, Rev. 6 0 Li6ie0 .6.. M.. ili,

== i.ie u.. ........n t.-

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wu 5

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~ / ., \\ f un m, p .x u.... h... l j s,.- ~~ / , ~ h. h x-j /! ,T M /-. ~y s in. i s, x , / f F m !* Y m. e. m. u.. r e ns e...n. : l Figure A5 Nodes 1317-2017 Extreme Penetration Case K.2-36 l l l l l

CRBRP-3 Vol. 2, Rev. 6 anestes cesuas an.. LEGEe8 n... u..

==... = ........n 8 l m. f ,/ ) l =

== ui / ( es a*.. -u t.. L \\ g / ,-, g\\ on... ~

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N. m. in. in. r a ns e ssess e Figure A6 i Nodes 301-1001 Extreme Penetration Case O_ K.2-37

CRBRP-3 Vol. 2, Rev. 6 0 Ltstas ow. =n n. F\\ mu... f ~ ) ........n

u..

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n... / un... / m. -- a o, i 'g j t an.... t ..%.N.. i ._ -l / ~' ~'.5 o

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A. go, Q j l / m. en. T Iati st..$ 3 Figure A7 Nodes 304-1004 Extreme Penetration Case l i 8 l

CRBRP-3 Vol. 2, Rev. 6 OV ....t.....u.. Listes am..

==... [\\

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assme.as Figure A8 l

Nodes 309-1009 Extreme Penetration Case t J I K.2-39 l l

CRBRP-3 Vol. 2, Rev. 6 0 enestes costies tsssee me. =u

==... l

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3 "'T; / -- un... mu.. i i fi . }%_ f s .~. J J ./ -s K / ~ % A, / A i... ~N 'g.,,k '/ / ' ONN w NN. i /' 3. ,lf 2 in. m ~ y %f/ N. a. m. in. t anee nsens e i I i Figure A9 Nodes 311-1011 Extreme Penetration Case K.2-40

CRBRP-3 Vol. 2, Rev. 6 ) / l OESULWs COOLIt$ LE6tes I IM. N E

• • • • • • • • W 86 938 e

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t / /" i / r / f / lll *,' i f / > F w. .i N. a-a. m. in. 5:asin$ses Figure A10 Nodes 312-1012 Extreme Penetration Case K.2-41

CRBRP-3 Vol. 2, Rev. 6 0 LEGEme en.. un.. wu...

u..

........ w u.. n = - - - - = =.. -\\ / / 2 --== n. i p, .n l / .I.s'. N., = = f.- / s -n -.K ,,,, a f i j ,e =u i.n / X. .~ / ~ _. S ! l; /D ~.. 9 p m m >h

) li,

/0,. ~ ~ -~ j,:,' i f l*.!'Y u. u. an. en. ssass mouas s Figure All l l l Nodes 313-1013 Extreme Penetration Case t 1 l l I l 9 K.2-42

CRBRP-3 Vol. 2, Rev. 6 s enantes testine tseses m

== si. f l_-,'T. - - - - = =... m. /

== ei. 1 -/ =e ei. l / ./ 1 '.h*- ~ [,/ = 439 - ~ = = u. j j wu.... A //.' ~. ~ ~.x ~ / /.' awn A c ,'f y% .l!,',' s . /f l f / e. m. m. m. m. m.

asseeves Figure A12 Nodes 314-1014 Extreme Penetration Case m) l K.2-43

CRBRP-3 Vol. 2. Rev. 6 0 Lasses MM N ........ e..g it. lI I l' .... wu... ll; = -- un n. ,.-s - = = n. &a r' e e.

a. =

u. i ,1 g.:.,

== i... .e l l'.t ,l i ~;. s a / 'k l / j t. . l s'f ll O' ~ -_.[.,1./. . /I._.'"7'_'?k in. - s / N. i m. w. n. in. l ran.ne.asa Figure A13 Nodes 315-1015 Extreme Penetration Case O K.2-44

3 CRBRP-3 Vol. 2, Rev. 6 APPENDIX B TO APPENDIX K.2 'mBDB SODIUM-CWCRETE PENE1 RATION MARGINS STRUCIURAL ASSESSMENTS '1he soditan-concrete penetration margins assessment case considers penetration rates of sodium into concrete of 7 inches per hour for 3 hours and 1 inch per hour thereafter. 'Ihe temperature transients for this case are shown in Figures 5 through 15. The structural requirements are as follows: A. Wall liners in the Reactor Cavity to maintain integrity tmtil boildry time (50 hours). While this is not a base case TMBDB requirement, it is inposed here to preclude extreme penetration into the RC walls concurrently with extreme penetration into the floor. B. Pipeway cell wall integrity to be maintained as follows: Wall between RC and pipeway cell (double heated wall) - 50 hours All others - 30 hours C. 'Ihe Reactor Cavity floor and Pipeway Cell floor to prevent sodium leakage to Cell 105 until boildry time. D. Containment and Confinement integrity to be maintained for long term. The structural assessments for the soditat-concrete penetration margins assessment case include the Reactor Cavity, Pipeway Cells, and the Confinement structure and containment shell above the operating floor. 'Ihe containment and confinement below the operating floor are not subjected to any significant tenperatures in the short term, and long term effects are not expected to be different than the base case. 'Ihe purpose of the structural assessments was to evaluate whether the l structures, as designed, can withstand the inposed conditions, and if not, to determine what modifications would be necessary to accomplish that. Due to the scoping nature of the work the assessments are based on sinplified conputer models ancVor comparisons with the base case structural evaluations, the' material properties and criteria are as described in Appendix C.3 of K.2-45 l . _ _, _ _ _ _... _ _ _, -.. _.. __-_-_, _, - _ _...,. _.._.--,_ _...__ _. _.. __ _....-...-_.-_,~.

CRBRP-3 Vol. 2, Rev. 6 CRBRP-3 (Reference B1). A brief sumary of the evaluations is given below. REACIOR CAVITY AND LINER The Reactor Cavity wall above the floor is subjected to tmperature transients (Figures 8 through 10) which are about the same as in the base case, since the surface tmperature is governed by the sodium boiling temperature which is independent of sodium-concrete penetration rate. Integrity for the Reactor Cavity wall and liner in the base case has been demonstrated for 50 hours and longer, so the 50 hours (boildry time) integrity requirment in the margins assessment case is met. The Reactor Cavity floor liner is assumed to fall at the onset of the accident as in the base case. The Reactor Cavity floor thermal transients of Figure B1 indicate that 5.5 to 6.0 feet of concrete would be totally degraded by the penetration of sodium and heat and this would leave only about 2 feet of concrete to the floor fill construction joint (Figure B2) which is not adequate to prevent sodium leakage through the construction joint. Further, if the sodium penetration extends radially into the wall, the base of the wall would be undermined and leakage to the adjacent cell 105 might occur. In order to meet the scenario requirments it would be necessary to introduce modifications in the floor of the Reactor Cavity and the junction with the wall. The basic modifications, in the conceptual stage, consist of the following (Figure E3): l A. TLe wall liner would be extended to 6.5 feet into the structural concrete. B. The constructicn joint at Elevation 733 would be eliminated and the floor rebar rearranged. C. Provide a design feature on the RC floor near the wall to inhibit the spreading of the fuel debris to the region of the floor wall junction. K.2-46

CRBRP-3 Vol. 2, Rev. 6

O Evaluations were performed to determine the adequacy of the concrete below l

the reactor cavity floor under the temperature transient at boildry time (50 hours). In these evaluations the undergraded portion of the floor was ] represented by the axisynnetric restrained section in Figure B4 and the stress analysis was performed using elastic procedures and the computer program ANSYS (Reference B2). The behavior was bracketed by two extreme conditions of no radial restraint and full radial restraint. Capacity calculations were performed using the computer program MPHI (Reference B3). ) 'Ihe results of the analyses demonstrate that the floor can withstand the imposed temperatures with sufficient margin. PIPEWAY CELL WMLS In the base case, integrity for the wall and liner between the Reactor Cavity and the pipeway cells has been demonstrated for 70 hours. For the penetration margins assessment case, the wall liner is subjected to the same temperatures as in the base case and the concrete wall at boildry (50 hours) is subjected to temperatures which are lower than the base case 70 hour f temperature (Figure B5). It may be concluded that this concrete wall and wall liner meet the margin assessment case requirenents. I In the base case, integrity of the pipeway cell walls and wall liners (other than the liner between the RC and the pipeway cell) has been demonstrated for 40 hours, at which time the liners may fail, but collapse of the walls is not expected until after the 133 hour sodium boildry time. In the penetration margins assessment case, the wall liners are subjected to the same temperatures as in the base case. A comparison of the penetration margins assessment case transient with the base case transient (Figure B6) indicates that the pipeway cell wall temperatures at 30 hours are somewhat lower than l the 40 hour base case transients. It may be concluded from this that the l pipeway cell liners meet the penetration margin assessment requirements. Because of the much shorter boildry time, the wall temperatures are less severe at boildry (50 hours) in the margins assessment case, so that the concrete walls also meet the requirements for the penetration margins assessment case. K.2-47 ~ -, _. _ -.. _. _..

CRBRP-3 Vol. 2 Rev. 6 0 PIPEWAY CELL FII)OR AND LINER The penetration margins assessment case considers that the pipeway cell floor experiences the same penetration rate in the reaction as the reactor cavity so there is substantial sodium penetration. In order to acca modate the t m perature transients it is necessary to introduce the following modifications to the present design (Figure B7): A. Provide a second layer of insulating concrete below the second liner. B. Increase the thickness of the floor by the thickness of the second layer of insulating concrete (lower bottom). With the above modifications the t aperature transient at 50 hours is as shown in Figure B8. Scoping computer analysis was performed using the programs ANSYS and MPHI, with the floor represented by a restrained section similar to that of the Reactor Cavity floor model ch cribed earlier. The results of this analysis indicate that structural int.grity of the modified floor would be maintained and leakage to Cell 105 would be prevented for at least 50 hours as required. CINFINENENT STRUCIURE An evaluation was performed to determine whether the confinement structure could sustain the thermal transients of Appendix A with the same annulus cooling as in the base case. The evaluation consisted of computer analysis using simplified models and comparisons with the base case evaluation described in Reference Bl. Specifically, analysis was performed using the cmputer program ANSYS and models of restrained sections similar to that shown in Figure B4 for the RC floor to calculate the thermal m ments and forces at various levels. These values were then adjusted based on the results from the base case which considered both restrained sections and the full structure and were compared to allowables frm the program MPHI. The 1 l results indicate that for the 50 hour transients integrity will be maintained. The results also indicate that the worst conditions, frm the O K.2-48

CRBRP-3 Vol. 2, Rev. 6 p V structural stanchoint, occur at 50 hours or earlier. As shown in Appendix A, the temperatures decrease after the 50 hour sodium boildry time. CNTAINMENT STEEL SHELL 'Ihe temperature transients in the steel shell are shown in Appendix A, with 4 i the annulus cooling activated at 10 hours. %e peak temperatures prior to j venting (when the peak pressure of 18.7 psig occurred) were about 6000, F 1 Reference B1 gives a pressure capability of the steel shell of over 34 psig 1 at 6000F so there is no significant threat to the steel shell as a result of the penetration margins assessment case. Also, the peak temperature at the steel shell-grade level intersection was about 1700, well below the 2400F F critical buckling tem prature of Reference Bl. It is concluded that the margins assessment case does not present a significant challenge to the steel shell. i REFERENCES B1 CRBRP-3, Volume 2, Assessment of %ermal Margin Beyond the Design Base B2 Computer Progran ANSYS, Revision 3, Swanson Analysis Systems Inc., Houston, Pennsylvania H3 Computer Program MPHI, Burns and Roe, Inc., Oradell, NJ O K.2-49

CRBRP-3 Vol. 2, Rev. 6 O l 1800 1600 1400 1200 ed E P 1000 I E E 800 k30 HOURS 50 HOURS (Boil Dry) 600 ~ j \\ 400 ~ \\ \\ 200 0 O 20 40 60 80 100 120 i DISTANCE FROM TOP OF FLOOR, IN. FIGURE B1 REACTOR CAVITY FLOOR - TEMPERATURE TRAFSIENTS O K.2-50 l

CRBRP-3 Vol. 2, Rev. 6 O l l l l l J NSULATING CONCRETE l l 9 i l EL. 7408" l 740 =. - w- ~7 - / + g sii s +1 ,e 6 / o EXTENT OF Na PENETRATION 4 ~ 2 CONST. JT 730 di 730 'I ~ CONTAINMENT LINER f l FIGURE B2 l SECTION - REACTOR CAVITY AND WALL O CURRENT DESIGN AND EXTENT OF Ha PENETRATION EXTREME PENETRATION CASE K.2-51

CRBRP-3 Vol. 2, Rev. 6 O ~ Top of Floor EL. 740'-8" 3/8" Steel Liner 3/8" Steel Liner Insulating [ Concrete X fl rw I x x x tw n X io f . 733' EL et,_ 11 a l EL. 730' Containment Liner Figure 83 Section-Reactor Cavity Floor and Wall Modified Design K.2-52

CRBRP-3 Vol. 2, Rev. 6 R Ved s - / 40/ fare ~ ss,, f. a ,1 = C 54 4

• . d V

t 9, s% Ni e M I .p ~ rw coec. .? b i -,i Yl c.uae i s M 1 _ja < w <<<,,,,, (SA. 7Jo o /2 " I FIGURE B4 AXISYMMETRICAL MODEL FOR RC FLOOR I K.2-53 r ,-,n-, -.,------.e---,-,-,n,,-._ -,,,, - -~ r----,, --.,---,,,, -, -na,- ,.o--.--

.) CRBRP-3 Vol. 2, Rev. 6 O MEACTOR CAv!TV PIPEWAY CELL DOUBLE HEATED WALL LEGEND taoo. O Hours f I 10 i co. 20 inoo. 30 I 40 izoo. t u. l ma 50 1 l $ icoo, i i a ,v i CL j .i ecc. .70 Hour Base Case .co. r. l..

1. +

noo' / a' '0l ~ { ',? k %~- s ..T'. ... ) '...' ' ~. ~.. ~, ~ ~ ' " :......:...... ioa. c. <>. o o.. o.e s.: s.s a.o a.. i.e 1.t s.s e.o DISTANCE (FT) NOTE: Penetration Margins Assessment Transients Unless Noted Otherwise Figure B5 K.2-54

CRBRP-3 Vol. 2, Rev. 6 O REACTOR CAVITY PIPEWAY CELL OUTSIDE WALLS-9 FT. THICK LEGEND 1600. O Hours 1 10 1*CO. \\ 20 q 30 t/CG. I i 40 C i

icco, 50 2

i o N%' l 3 ace h W a *., .i 40 Hoor Base, Case 600 i -i ij'\\ + y s 400. i ?. %.......:Nw_x._.. ~. ..... - -.= ;. a.- I b 0.0 0.* o.a 1.7 1.a 7.o 7.* 7.a 1.2 3.6 9.0 DISTANCE (FT) t NOTE: Penetration Margins Assessment Transient Unless Noted Otherwise Figure B6 Pipeway Cell Wall Temperature K.2-55 m---,

+ 3/8" FLOATING LINER I 3/8" LINER ] _,.se { 'd~ e -f -Y p INSULATING l w 2 = CONCRETE INSULATING (PERLITE) CONCRETE (PERLITE) N .. 1/4" FIXED LINER g 1/4" LINER] I g x- -l 'u gg 7: N INSULATING E g g 3 CONCRETE f 4 g (PERLITE) cn %h PRESENT DESIGN 1 PROPOSED MODIFICATION FOR EXTREME PENETRATION CASE (CONCEPTUAL) FICURE B7 PIPEWAY CELL FLOOR CROSS SECTION

CRBRP-3 Vol. 2, Rev. 6 O 4" INSULATING CONCRETE SECOND LINER 2000 1 1 1600 c' w 1200 O=5 C"i e 800 400 0 i e i i i i i ) 20 30 40 50 60 70 DISTANCE FROM TOP OF FLOOR, IN. FIGURE B8 PIPEWAY CELL FLOOR - THERMAL TRANSIENTS K.2-57 l

1 i 4 3 CRBRP-3 { Vol. 2, Rev. 6 y-

9 i

s K.3 TMBDB MELTING SCENARIO

)-

1 ' This appendix contains the DOE letter supplied to the NRC on i i j November 23, 1982. i o ) k 4 l 1 i I i I !e i h i i l 1 I i l 1 i-l- I e i l_ K.3-1 n v M w f te-9 *m mweew-y-ww,g ge mey m weer_ _. _, .., -wv,wwerm -ve eewaew-w w me.3-e, -_.. _. _ --wamur

l CRBRP-3 Vol. 2, Rev. 6 hl Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:127 ) NOV 2 31982 1 l Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Mr. Check:

CLINCH RIVER BREEDER REACTOR PLANT TRANSMITTAL OF THERMAL MARGIN BEYOND THE DESIGN BASE (TMBDB) MELTING SCENARIO Enclosed is a document entitled "TMBDB Melting Scenario." This document is in response to a commitment made at the last TMBDB meeting with the Nuclear Regulatory Commission on September 15, 1982. Sincerely, Jo n R. Longen ker Acting Directo. Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy O K.3-2

CRBRP-3 I Vol. 2, Rev 6 O TMBDB MELTING ECENARIO I. Introduction The CRBRP Thermal Margin Beyond the Design Base (TMBDB) scenario assumes that a coolable core debris bed forms on the reactor cavity floor following a melt-through of the reactor vessel and the reactor vessel guard vessel. To ensure that a coolable bed does form, the reactor vessel guard vessel support skirt is provided with flow ports to assure that () sufficient core debris is swept under the support skirt to produce coolable bed depths on the RC floor both inside and outside the support skirt (this is discussed at length in Appendix G.1 of Reference 1). Following the establishment of a coolable core debris bed, the RC floor liner is assumed to fail. In order to maximize the sodium-concrete contact in the Base Case scenario (Reference 1), the RC floor liner is conservatively assumed to " vanish" at the time of The core debris is then assumed to remain at failure. the interface of the sodium-concrete reaction product' layer and the sodium, such that the debris bed remains cooled by the sodium pool. This report presents the results of an analysis to examine the ef fects of a i K.3-3

CRBRP-3 Vol. 2, Rev. 6 9 hypothetical scenario where the core debris is as'sumed to sink beneath the sodium-concrete reaction products, whereupon the reaction product layer would insulate the core debris from the cooling effects of the sodium pool. The purpose of Obis " melting scenario" analysis is to evaluate the consequences of a non-coolable debris bed early in a TMBDB scenario in terms of TMBDB margins (Reference 1). Thus, the melting scenario is compared with the TMBDB base case scenario in terms of (1) containment vent time (36 hours in the base case), (2) containment atmospheric and steel shell tempe ratur es, (3) containment pressures, and (4) RCB atmosphere hydrogen concentration. This scenario is presented as a bounding case on the ef fects of core debris on the TMBDB scenario and is not assumed to be a realistic scenario aince: t I i 1. The RC floor liner is not expected to " vanish" at time zero and this scenario ignores the capability l of the unfailed portions of the floor liner to support the core debris. 2. The sodium-concrete reaction products are believed to be a viscous liquid which would O K.3-4 m.

CRBRP-3 Vol. 2, Rev. 6 O \\ provide a measure of support for the core debiis. II. Conclusions f In the event that core debris on the reactor cavity floor assumed a non-coolable geometry (melting scenario), the following conclusions are drawn from the melting scenario study discussed in detail below: l containment conditions would be acceptable o throughout the scenario, containment venting could be delayed for at least o 24 hours, penetration into the basemat prior to sodium boildry o would be about 5 feet, existing TM'JCB systems would be capable of handling o the TMBDB loads. III. Scenario ) The melting scenario was developed from the TMBDB base K.3-5

CRBRP-3 Vol. 2, Rev. 6 O case spenario of Reference 1, Section 3.2.1, so as to permit direct comparison with the base case, thereby evaluating the impact of early melting on the beyond design base margins. Thus, the only changes were to replace the sodium-concrete reactions in the reactor cavity floor (af ter a 20 minute initial rapid sodium-concrete reaction as described below) with an extreme heat input to represent progression of a melt front (molten core debris and melting concrete) into the floor. All other features of the scenario, such as sequential liner failures, the resulting l sodium-concrete reactions in other structures, the 50-hour vent bypass through the reactor vessel head, etc. were retained. Therefore, only the scenario details involving the molting into the RC floor will be described here. 1. Upon failure of the reactor and guard vessels, it is assumed that sodium and core debris pour onto the reactor cavity floor and spread uniformally over it as in the base case. l l 2. For twenty minutes, sodium-concrete reactions occur at a rapid rate of 18 cm/ hour for a total K.3-6

CRBRP-3 Vol. 2, Rev. 6 O penetration of 6 cm. This rate and duration are the recommended initial upper bound from Reference 2, page 50. 3. The reaction products formed from the initial 20 minute sodium-concrete reaction form a layer 5 inches thick (approximately twice the volume of the original concrete reaction tests), and the core debris sinks to the bottom of the reaction product layer. At this point the core debris [} 1ayer (3.4 inches thick per Reference 1, page G.1-7 ) is separated by the 5 inch layer of reaction products, which has a thermal conductivity much less than sodium (assumed value 1.0 BTU /hr-f t-oF vs. M 30 for liquid sodium). 4. Also at this point the sodium pool is mechanically isolated from the concrete so that sodium-solid concrete reactions cease. The steam and C02 driven out of the concrete, by the heat of the core debris, pass through the debris and reaction product layers and react with the sodium pool, to generate additional sodium reaction products, hydrogen and heat (i.e., the sodium does not ] K.3-7 I~

CRBRP-3 Vol. 2, Rev. 6 O migrate down to the concrete since, if it did; the debris would be cooled by the sodium and there would be no melting of concrete as described below). 5. The insulating effect of the reaction product layer causes the core debris layer to heat up (due to core decay heat). When temperatures reach the I assumed melting point of concrete (2200 F)*, the core debris melts into the concrete. 6. As the concrete melts, the core debric layer sinks lh through the molten concrete, and remains in contact with the unmelted concrete. 7. The core debris melt front continues down into the concrete, driving water and CO up into the sodium 2 pool. Sideways melting is not considered since the extension of the wall liner anchor for the Margin Assessment Case (see Reference 3) would preclude sideways melting. l

• This is the lowest eutectic temperature of the concrete-core debris mixture (see section 2.2 of Reference 1).

K.3-8

i CRBRP-3 i Vol. 2, Rev. 6

O i

l ) i After sodium boildry, the scenario is essentia'lly f 8. the same as the base case post-boildry, except the i i i post-boildry melt front starts at a lower elevation by the amount of melt front penetration l prior to boildry. } I IV. Analysis Methods 1. The analysis was conducted in two steps, a TRUMP t' t i () analysis of the thermal ef fects of fuel debris melting into concrete to determine fraction of decay heat i driven into the concrete followed by a CACECO analysis f to determine containment conditions resulting from the RC floor thermal ef fects. 1. TRUMP Analvnis The TRUMP code was utilized to study the situation where fuel debris on the reactor cavity floor is separated from an overlying pool of sodium by an In this insulating layer of reaction products. situation, heat transfer from the fuel to the sodium is by conduction through the insulatin'g layers of molten concrete and reaction products. K.3-9

CRBRP-3 Vol. 2, Rev. 6 O Under these conditions the fuel would be expected and to melt into the concrete, releasing more CO2 water than that calculated for the base case, and the resulting effecta on containment using the CACECO code are to be calculated. The analysis using TRUMP, however, is a prerequisite in order } to develop the input for CACECO. A one-dimensional TRUMT model was developed as shown in Figure 1. It consists of one node representing the reaction product layer which is connected to a constant temperature boundary (the sodium pool); a second her.t generating node representing the core debris layer, and 52 nodes representing 26 feet of concrete. The thermal properties for the fuel and concrete l are the same as has been used in previous analyses of Reference 1. The thermal properties for the reaction product (assumed to be sodium carbonate) are: l l l 3 p = 158 lb/ft 0 C =.27 Btu /lb F p K.3-10

CRBRP-3 Vol. 2, Rev. 6 O k = 1.0 Btu /hr f t oF The density and heat capacity of sodium carbonate were obtained from page 3.121 of Perry's Chemical The thermal Engineer's Handbook, Fourth Edition. 1 conductivity value was selected on the basis that it would be similar to that of concrete (0.6 to 1.5 Btu /hr ft oF) as no data were found specifically for sodium carbonate. O The results of the calculation are shown in Table 1 as the fraction of the generated heat which is transferred to the overlying sodium, and the fraction that is transferred to the concrete beneath the fuel. It is these two quantities that are used in developing the CACECO input. Except i for the early phases (several hours) of the transient, approximately 60% of the heat generated is transferred into the underlying concrete. 2. CACECO Analvais In the TMBDB scenario, fuel in the reactor- ~civity would be expected to be in particulate form in a K.3-11 - - - -. -. -. ~. -... - -.

l CRBRP-3 Vol. 2, Rev. 6 O l coolable debris bed until all of the sodium wouId' l be boiled away (Reference 1). However, as described earlier, this analysis will consider the core debris bed spread over the entire cavity l floor to be uncoolable after sinking beneath a five inch layer of sodium-concrete reaction l Based on the scenario described in products. i the CACECO code was used to model Section III, this effect. l i The CACECO model used in the analysis is the same i as used in the Reference 1 base case except for the following: (a) the cavity floor was divided into 5 heat structures made up of a series of 2 inch nodes; (b) the sodium-concrete reaction was redefined (described later); (c) the boundary conditions applied to the RC floor were changed to l venting simulate a melt front progression; and (d) was initiated at 24 hours. The RC floor thermal model involved 5 connected heat structures to represent the basemat. Each of the heat structures was made up of" nodes 2 inches thick so as the melt front progressed, the volumetric release of water and carton dioxide reacting with 9 K.3-12

I CRBRP-3 Vol. 2, Rev. 6 l the sodium would occur in realistic increments. l l l For this scenario, sodium-concrete reaction rates of 18 cm/hr were utilized for a 20 minute duration. (See Reference 2, page 50).* The reaction energy for this interaction was the base case value of 331 Btu /lb. of concrete. With a layer of reaction product formed from the initial rapid sodium-concrete reaction, the core debris was assumed to sink below this layer and become isolated from the sodium pool. At this time the core debris cooling would be greatly reduced, I causing the core debris temperatures to increase and melt the concrete. Based on the TRUMP results shown in Table 1, the fission product decay power was proportioned between the sodium pool and the concrete in the cavity floor to simulate a melting process in the CACECO code. The melting process into the RC floor is simulated by forcing a heat flux into the first node of concrete. As the melting point of the concrete (2200F) was reached

• For the. Base Case scenario, the sodium-concrete reaction rate is assumed to be 1/2" per hour for 4 hours.

K.3-13

CRBRP-3 l Vol. 2, Rev. 6 e i i for a particular node, the thermal conductivity of that node was subsequently changed from approximately 1 Btu /hr ft *F to 100, thus allowing the melt front to progress. As the concrete heated released passed through the up, the water and CO2 core debris, the reaction product layer and the melted concrete to react exothermally with the sodium pool until sodium boildry. Vent and annulus cooling were initiated at 24 hours with purging starting at 27 hours. All other aspects of the scenario were identical to the base case. V. Results The containment conditions from this CACECO analysis are compared with the TMBDB base case in Table 2. As expected, containment pressures, temperatures, and hydrogen concentrations are more severe for the I m'elting scenario than for the base case, but the conditions are still acceptable in terms of vent time i (see Reference 3). The peak hydrogen during the. l l K.3-14 O m--

. =. CRBRP-3 Vol. 2, Rev. 6 O venting blowdown exceeded 6 v/o (maximum 6.4 v/o) 'for a short period, but this occurred at a time when l containment oxygen was below 5 v/o. The initial (autocatalytic) hydrogen burn occurred earlier in time with a greater accumulation of hydrogen and resulted in higher, but still acceptable, pressure and temperature spikes (conservatively assuming instant burning). Beyond 24 hours the upper containment conditions are acceptable. The peak vent rate to the cleanup system and the peak heat load to the steel shell were slightly higher than the base case, but within other extreme sensitivity studies for which TMBDB systems These results indicate that were found adequate. existing TMBDB systems could accommodate the melting i scenario. Temperatures are, of course, much higher deeper into'the concrete basemat in the melting scenario, but the shortened boildry time would make the structural temperatures in the RC walls and pipeway cells less severe since the structures would be exposed to sodium boiling temperatures for a The steel'shell temperatures are higher shorter time. but could be reduced by starting than the base cast, i K.3-15 ~ \\ i y ~ -3 ..,, _.,. - ~ - _ _

CRBRP-3 Vol. 2, Rev. 6 O the annulus cooling sooner without venting contaisiment (in this calculation, both were started at 24 hours). It is concluded that the TMBDB systems would mitigate the threats to containment in the unlikely event that the core debris on the reactor cavity floor resulted in a non-coolable geometry (melting scenario). References CRBRP-3, Vol. 2, " Assessment of Thiermal Margin 1. Beyond the Design Base," dated March 1980. 2. BEDL-THE 82-15, " Sodium Concrete Reaction Executive Summary Report: Limestone Concrete," dated June 1982. 3. Letter BQ:S:82:ll2, J. Longenecker to P. Check, "TMBDB Margin Assessment Document," dated October 20, 1982. K.3-16 O

CRBRP-3 Vol. 2, Rev. 6 O Table 1 Beat Split Between Concrete and Sodium As Determined by the TRUMP Model Heat Split Time, hr Fraction Fraction Into to Concrete sodium 1-10 .79 .21 10-25 .63 .37 25-35 .62 .38 35-50 .63 .37 l l 50-+ .60 .40 l l l K.3-17 I l l

Table 2 CWarison of Containment Conditions for Base Case and Helting Scenario All Parameters: Base Case / Melting Scenario Initial Hydrogen Bu;n_ 10.0/8.4 Time (Hrs.) 22.4/27.9 Pressure After Burn (psig) 845/1031 RCB Atmosphere Temp. After Burn ("F) $n 4.5/5.7 RCB Hydrogen Conc. Before Barn (v/o) "g k"9 xL Initiation of Venting cn 36/24 Time (Hrs.) 4.5/6.4 RCB Hydrogen Conc. Peak During Vent (v/o) Remainder of Scenario RCB Atmos. RCB Pressure (psig)_ Steel Temp. (*F) Hydrogen (.v/o)_ I Timb (Hrs.)_ Temp. (*F)_ 0/2.4. 270/460 11.1/18.3 0/4.0 24 450/830 400/445 l 13.1/0 36 620/875 4.0/4.0 315/410 0/0 50 515/790 360/495 3.7/4.0 0/0 71 670/1000

O o o f Table 2(continued) Base Case Melting Scenario 133 71 Bo11 dry Time (Hours) 4 RCB Maximum Vent Rate (ACFM) 24000 29000 0 8 Peak Heat Load on RCB Steel Shell (8tu/Hr.) 1.1 x 10 1.2 x 10 i 5 5 8.6 x 10 6.71 x 10 sodium Vapor to RC8 at Boildry (Lbs.) 19300 29000 Sodium Vapor to RC8 Max. (Lb/ Hour) 2.3 5.5 RC Floor Temperature Above 500'F at Bolldry g l (Ft. from Liner) 1.2 0.6 h,9 I' RC Submerged Wall Temperature Above 500*F at Bolidry -g (Ft. from Liner) x, m 0.33 5.0 h" RC Floor Penetration at Bo11 dry (Ft.) cn I l l l f

T' l CRBRP-3 Vol. 2, Rev. 6 TMBDB MELTING SCENARIO Sodium Overlay-simulated by constant temperature boundary b Reaction Product Layer } h Molten Concrete Increa tsing Fuel Debris } 3.4 " O Solid Concrete 26 ft. (and decreasing) 1 Figure 1 Schematic of TRUMP Model 0 K.3-20 ~

CRBRP-3 Vol. 2, Rev. 6 'd K.4 INFORMATION ON TMBDB This appendix contains the DOE letter supplied to the NRC on December 7, 1982 and Enclosures 2 and 5 of that letter. Enclosure 4 is not included here since it was included in Revision 5 of CRBRP-3, Volume 2. is not included here because it was incorporated in the revised response to Question CS760.144 in Amendment 74 to the PSAR. Enclosures 1 and 3 are not included since they are now available separately as published reports: : L. E. Muhlestein and R. Colburn, " Aerosol Release from Sodium-Concrete Reactions," HEDL SA-7351, dated May 1983. : G. R. Bloom, " Hydrogen Distribution in Breeder Reactor Containments," HEDL TME-83-5, dated April 1983. G K.4-1

CRBRP-3 Vol. 2, Rev. 6 Departrnent of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:140 DEC 0 71992 Mr. Paul S. Check Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Check:

SUBMITTAL OF INFORMATION ON THERMAL MARGINS BEYOND THE DE On September 15, 1982, the project and the Nuclear Regulatory Comission (NRC) met to discuss the open issues on TMBDB. The enclosur,es to this letter docket all of the action items the project owes NRC on this topic. The following enclosures are: : Aerosol Release from Sodium-Concrete Reactions, by Muhlestein and Colburn, dated October 1982 : Prototype Hydrogen Filter Tests - preliminary results Hydrogen Distribution in Breeder Reactor Containments,) : by G. R. Bloom, dated October 1982 (applied technology : Miscellaneous updates to CRBRP-3, Vol. 2 (includes updates on sparged plutonium and new requirements on the vent and purge lines) : Information on the Containment Cleanup System : Revised response to Question CS760.144 on axisymmetric distribution of temperatures Sincerely, h C John R. Longenedker Acting Director, Office of ~ Breeder Demonstration Pr,ojects Office of Nuclear Energy O K.4-2

CRBRP-3 Vol. 2, Rev. 6 O 4 ENCLOSURE 2 PROTOTYPE HYDROGEN FILTER TESTS - PRELIMINARY RESULTS O i i 'l i .l l O K.4-3

CRBRP-3 Vol. 2, Rev. 6 Prototvoe Hydrocen Filter' Tests Introduction The Project has provided a letter report on the 'IMBDB instrumentation developnent program.* The purpose of this report is to provide preliminary information on the results of the prototype hydrogen filter test performed in September 1982 as a part of the 'IMBDB instrumentation develognent program. The objective of the prototype test was to demonstrate the performance of the sodium aerosol filters for the hydrogen monitoring system during prototypic 'IMBDB conditions. This test was designed to simulate the total filter exposure anticipated for 'IMBDB conditions by 100 hours of exposure to sodium aerosol and 600 hours of exposure to high tm perature. The test involved 3 filter configurations which are described in Table 1. Test Results For the first 40 hours of the test, the three filter assemblies performed adequately. The filter assemblies were repeatedly blown back successfully whenever the pressure drop reached 100 inches of water, thus demonstrating the adequacy of the filter blow back technique. At 40 hours into the test, water vapor injection into the test chamber was stopped and carbon dioxide injection began. At this time, the filter blow back became increasingly difficult and, at 45 hours, the blow back technique could no longer maintain the pressure drops less than 100 inches of water. The test was terminated at this time. Table 2 provides the primary parameters of the test and Table 3 surtrnarizes the results of the test. Figures 1 and 2 provide a cmparison of the prototype test chamber temperature and aerosol concentration versus the 'IMBDB Base Case conditions. Following cool-down, filter assemblies 1 and 2 were disassembled, examined, photographed, and flow tested. (Filter assembly 3 was preserved for future examination and testing.) The filter disassembly showed that: o Filters were mechanically intact. o Areas of corrosion and small holes were present in the filter media. o Filter surfaces appear darkened and glossy. o Settling chambers and filter surfaces appear relatively free of caked aerosol build-up, o Filters lost approximately 2% of their weight.

• Letter HQ:S:821096, J. R. Longenecker to P. S. Checs, "'IMBDB Instrumentation Development," dated September 29, 1982.

K.4-4 i

CRBRP-3 Vol. 2, Rev. 6 l Test conclusions 'Ihe prototype test deonstrated the feasibility of designing filters to accm nodate the high taperature aerosol loading anticipated for 'INBDB i conditions for significant periods of time (See Table 2). The Project is currently beginning a series of laboratory-scale bench tests to obtain a more cm plete understanding of the phenomena involved in limiting the time of effectiveness of the individual units in the test. In addition, the Project is evaluating whether it is realistic to expect free carbon dioxide to exist in the vicinity of the hydrogen filter assemblies during 'INBDB conditions. Following cmpletion of these tests and studies in January 1983, the Project will re-evaluate the design of the filter assemblies and perform additional testing of candidate filter assemblies. In the unlikely event that a single filter assembly cannot be designed to accMte all the 'INBDB environmental conditions for the time required, alternate design configurations, such'as a series of parallel filters, could be utilized to rteet the 'INBDB conditiors for the required length of time. O

CRBRP-3 Vol. 2 Rev. 6 0 O TABLE 1 FILTER TYPES AND PARAMETERS SETTLING CHAMBER DFSIGN HPE OPEN BOTTOM FILTER 101 SINTERED WIRE MESH CLOSED BOTTOM FILTER 102 SAME AS 101 WITH A NICKEL POWDER LAYER BONDED TO UPSTREAM SURFACE OF SINTERED WIRE MESH O OPEN BOTTOM FILTER 103 SINTERED NICKEL POWDER l K.4-6

l CRBRP-3 Vol. 2, Rev. 6 i O TABLE 2 PROTOTYPE TEST CONDITION $LM ERY Maximum Temperature 1140*F 3 Maximum Aerosol Concentration (1) 55 g/m 3 Average Aerosol Concentration (1)

• 40 g/m Steam Flow
• 15 vol 5 start at 11.2 hours Stop at 41 hours CO Flow 1.9 vol %

2 Start at 39.8 hours Stop at 47.6 hours Aerosol Composition NaOH, Na 0 at 4.8 hours 22 Na CO at 40.8 hours 2 3 Na 00 at 45 kurs 2 3 (1) At chamber temperature O K.4-7

CRBRP-3 Vol. 2, Rev. 6 O TABLE 3 SU W RY OF PROTOTYPE FILTER RESULTS 2 F'11ter Area, m 0.053 0.054 0.058 Flow Rate, cc/ min (1) 3000(2) 3000 700 Flow Duration, hrs 42 42 44.7 Calculated Mass Filtered, g 528 904 220 Time at 1st Blow Back, hrs 10.0 2.1 11.4 Total Number of Blow Back (3) 33 35 14 2 Loading at 1st Blow Back kg/m 3.0 0.24 0.07 2 Final Loading, kg/m 10.0 16.8 3.8 Filter Efficiency 5 95.0 92.1 98.8 (1) At 70*F; the TMBDB filter flow is estimated to be 700 cc/ min at 70*F (2) Flow reduced to 1500 cc/ min at 21 hrs (3) When fi)ter AP increased to 100 inches H O 2 l l K.4-8 l l

CRBRP-3 Vol. 2, Rev. 6

O Temperature ('F) 1800 1400

= i Test Chamber Temperature 1200 =, -Test FWter Row Stopped 1M -b 000 E" TM808 Containment Ar.: f. Temperature (With Margins includedt gog l. 200 L I 0 O 20 40 60 80 100 120 140 100 Time (Hourst i Figure 1. C-. /;: Of TMBOB Containment Atmosphere Temperature To Test Chamber Temperature lO K.4-9 l i

CRBRP-3 Vol. 2, Rev. 6 f O ( l Aerosol Mese Concentration 3 (gm/m At Chamber Temperature) 100 l 90 - TM808 Bees Case p i 1 80 - Filter Test / ,/,

• s l\\

j s s n f x; \\v; <0 i i e i 1 20 / / Start Steam Start CO2 Stop Steam I ik p njection injectionq p:-;;;I I II I I Y I I 0 O 10 20 30 40 50 80 70 Time (Hours) Figure 2. Aerosol Mese Concentration At Chamber Temperature K.4-10

j CRBRP-3 Vol. 2, Rev. 6 ENCLOSURE 5 INFORMATION ON THE CONTAINMENT CLEANUP SYSTEM i

4 4 i 4 K.4-11 --*,-,,,--.---w-,-,e.-------- -,.m,,-, -.=.e .w-,, ,w*----, - - - --w-r, ,. - --~-m-e==w.w-wew-

CRBRP-3 Vol. 2, Rev. 6 O and show how system performance changes when

== Description:== containment cleanup system. Theoretical correlations currently exist in the technical literature which j estimate the filtration perfomance of components similar to those foun

Response

These correlations are in the CRBRP Containment Cleanup System. One of the ob.jectives of the referenced in Chapter 6.3 of HEDL TME 81-1. CSTF test at HEDL was to verify the applicability of these correla to sodium aerosol environments. shows that the theoretical modeling of the HEDL C system performance substantially exceeded the specifications. On the basis of the foregoing, the applicability of the theoretical correlations has been established and the actual size (scaled-up) CRB Containment cleanup System was evaluated to determine its estimated performance during the expected operating range of the system us The evaluation indicated thjt the predicted performance of all CRBRP Containment Cleanup System compon theoretical correlations. the operating range of the system is expected to be even higher than t actual performance of the HEDL CSTF cleanup system. On the basis of the above, it could be concluded that the specified performance of the CRBRP Containment Cleanu ance requirements. The comparison of the CRBRP vs. HEDL system parameters affecting the performance of the system is provided in the attached Table 1. l l K.4-12

l I t CRBRP-3 Vol. 2, Rev. 6 i 1 i i f TABLE 1 l i pOMPONENT/ PARAMETER BEDL CRTF CRBRP i I Duench Tank i Gas Flow Rate Max. /s 0.663 25 1 Gas Flow Rate Min. m /s 0.073 16.6 i Liquid Flow 1/s 0.125 31.5 Spray Fall Height a 3.35 5.80 l Spray Drop Diameter a 0.00114 0.00114 Relative Particle Velocity W s 12 12 4 Particle Diameter Max. microns 6.40 9.4 i Particle Diameter Min. microns 1.80 4.6 l Residence Time sec. 6.8 6.3 Venturi Ecrubber Gas Flow Rate Max. /s 0.612 11.3 Gas Flow Rate Min. m /s 0.056 M.3 3 l Liquid Flow a /s 0.225 63 Venturi Beight a 1.71 3.05 4 1 Spray Drop Diameter a 0.0014 0.0014 j Relative Particle Velocity m/s 1 1 Particle Dismeter Max. microns 5.0 9.4 Particle Diameter Min, microns 1.85 4.6 j Fiber Bed Scrubber i Fiber Fill Coeff. 0.15 0.15 i Fiber Bed Thickness em ~ 7.62 7.62 i Fiber Diameter em 0.002 0.002 Face velocity Max. cm /s 21.2 22.2 2 Face Velocity Min. cm /s 2.1 16.7 i Particle Diameter Max. microns 3.65 9.4 j Particle Diameter Min. microns 0.64 4.6 i f l i O K.4-13 t

1 CRBRP-3 Vol. 2, Rev. 6 O'

== Description:== During the presentation of the ' dynamic analysis" model of the cleanup system by l Peter Fazekas (B&R), equations from the EEDL air cleaning reports were referenced. In HEDL-TME-81-1, there are several equations to I correlate pressure differential, gas cooling, and aerosol removal efficiency (e.g., 97, pg. l 87, 48, pg. 90, 99, pg. 92, til, pg. 94, 017, pg. 100, #18, pg.105). Many of these l equations are based on specific conditions used in the BEDL system. What equations have been used in the CRBR, TMBDB dynamic analysis?

Response

The purpose of the CRBRP Containment Cleanup System dynamic analysis was to determine the thermal / hydraulic performance of the scaled-up system. It was not the objective of the analysis to scale-up the filtration ' efficiency of the system or to determine the pressure I drop thru the scaled-up components. The correlations for the containment cleanup system dynamic analysis utilized the test results from the BEDL CSTF AC-1 thru AC-4 tests and they were the following: QGO = 0.333 L/G + TQSO III "Ud l T VGO = TVSO i Where: TQGO = Gas tegp. out of Quench tank ( F) L/G = Liquid gas ratio (SCFM/GPM) Solution temp o QSO = Quench tank (6 )ut of T 7 TVGO " G8s temp. out of Venturi, ( F) Solution VSO = Venturi (gesp. out of T F) The dynamic analysis indicated that correlation (1) is valid for cases when the residence time in the quench tank is the same; therefore the CRBRP containment cleanup system quench tank was sized to yield the same residence time at maximum gas O K.4-14

CRBRP-3 Vol. 2, Rev. 6 flow as of the BEDL CSTF system quench tank at maximum gas flow. All other equations used in the CRBRP dynamic No equation analysis are fundamental equations. was used f rom HEDL TME-81-1 for the CRBRP dynamic analysis. 1 O I O K.4-15 4 ---.----e- -p -,,n--,,-, ,.,,-,,,,,m+,em,,,,-,,w ,,,,nsem.,,e,-,ma m.r_,n, w,,,,-,m w,,e,,_m-- -}}