ML20039G521

Forwards Procedures for Determination of Extent of Core Damage Under Accident Conditions. Encl Procedure Was Inadvertently Omitted in Util 820111 Submittal & Should Accompany Util Response to NUREG-0737,Item II.B.3

ML20039G521
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Site:	Susquehanna, Shoreham  File:Long Island Lighting Company icon.png
Issue date:	01/13/1982
From:	James Smith LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML17139A510	List: ML17139A509 ML20039G521
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM SNRC-660, NUDOCS 8201180360
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{{#Wiki_filter:_ -. nc,/w.-a u x x :s,r, a ssrm I LONG ISLAND LIGHTING COM PANY [ ,, (O SHOREHAM NUCLEAR POWER STATION w s P.O. BOX 618, NORTH COUNTRY ROAD + WAD!NG RIVER, N.Y.11792 y,.. .g .,ww. January 13, 1982 SNRC-660 P Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation 0g g ED = U.S. Nuclear Regulatory Commission g", gi $1MP. ( Washington, D.C. 20555 EgOasse8 n Shoreham Nuclear Power Station - Unit 1 W / Docket No. 50-322 ~ Wl

Dear Mr. Denton:

Enclosed herewith are sixty (60) copies of a document entitled " Procedures for the Determination of the Extent of Core Damage under Accident Conditions". This report was inadvertently omitted in our January 11, 1982 submittal (SNRC-657) and should accompany our response to Item II.B.3 regarding Post Accident Sampling. Very truly yours, A J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station CC:mp Enclosure cc: J. Higgins I l an98e# J880h A FC-8 935

..8 C6RE TRANSMITTAL RPE 81CCL01 ~ November 1981 e PROCEDURES FOR THE DETERMINATION OF THE EXTEST OF CORE DAMAGE UNDER ACCIDENT CONDITIONS PREPARED BY: ~CHIEN C. LIN Principal Engineer REVIEWED BY: 'Dtd T. A. GREEN, MANAGER Chemical Retrofit /) [

• REVIEWED AND C h Ik APPROVED BY:

C. P. RUIZ,9.ANAGER Radiological Proc *:ss Engineering h. W CC APPROVED BY: R. L. COWAN, MANAGER Chanical and Radiological Engineering l i APPROVED BY: L. F. FIDRYCH, ACTING MANAGER FLANT DESIGN AND ANALYSIS NUCLEAR POWER SYSTEMS ENGINEERING DEPARTMENT GENERAL ELECIRIC COMPAh7 VALLECITOS NUCLEAR CENTER, PLEASANTON, CALIFORNIA 94566 988AB82R'ol586$g N E R A L $6' E LE CTfilC A

<4 L, DISCLAIMER This document was prepared by the General Electric Company. Except as may be otherwise provided, neither the General Electric Company nor any of the con.tributors to this document nor any of the sponsors of the work makes any warranty or representation (express or implied) with respect to the accuracy, completeness, or usefulness of the information contained in this document or that the use of such infor-mation may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from the use of any of the information contained in this document. /

t. TABLE OF CONTENTS 1.0 OBJECTIVE AND SCOPE 1 2.0 PROCEDURES FOR DETERMINATIONS OF CORE DAMAGE 1 2.1 REFERENCE PLANT (BWR-6/238, MARK III) 1 2.I.1 REFERENCE PLANT PARAMETERS 1 2.1.2 PROCEDURE 1 2.1.3 SUPPLEMENTARY DATA 2 2.2 SPECIFIC PLANT APPLICATION 2 2.2.1 PLANT PARAMETERS, 2 2.2.2 PROCEDURE 2 2.2.2.1 COMPARISON WITH REFERENCE PLANT DATA 3 2.2.2.2 INVENTORY CORRECTION FACTOR 4 2.2.2.3 PLANT PARAMETER CORRECTION FACTORS 5 3.0 TECHNICAL BASIS 5 3.1 FISSION PRODUCT CONCENTRATIONS IN THE PRIMARY SYSTEM DURING REACTOR SHUTDOWN UNDER NORV.AL OPERATION CONDITIONS 5 3.1.1 FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER 5 3.1.2 NOBLE CAS CONCENTRATIONS IN DRYWELL AND TORUS GAS PHASE 6 3.2 FISSION PRODUCT RELEASE SOURCE TERMS UNDER ACCIDENT CONDITIONS 6 3.3 ISOTOPIC DISTRIBUTION IN FUEL GAP 6 3.4 ANTICIPATED CHEMICAL BEHAVIOR OF IODINE AND COOLANT CHEMISTRY UNDER ACCIDENT CONDITIONS 7 4.0 DISCUSSION AND SUGCESTIONS FOR FUTURE WORK 8

5.0 REFERENCES

9

1.0 OBJECTIVE AND SCOPE The purpose of this procedure is to determine the degree of reactor core damage from the measured fission product concentrations in either the water or gas samples taken from the prim ary system under accident conditions. The procedure involves calenistions of fission product inventories in the core and the release of inventories into the prim ary system under po stulate d loss-of-coolant accident (LOCA) conditions. The fuel gap fission products are assumed to be released upon the rupture of fuel cladding, and the majority of fission product inventorie s in a -{uel rod wod be released T fu<l *<l btw"*4ooce " M A d vod. A BWR-6/238 with a Mark III containment is used as a reference plant in the demonstration of this procedure. Application of the procedure for any other type or size of boiling water reactor (BTR) is described. 2.0 PROCEDURES FOR DETERXINATIONS OF COPE DAMAGE 2.1 Reference Plant (UTR-6/238. Hark III) 2.1.1 Reference Plant Parameters The pertinent plant parameters for the reference plant are given below: Rated reactor thermal power 3579 MTt Number of fuel bundles 748 bundles Total primary coolant mass (reactor water pins suppression pool water) 3.92x10's Total containment and drywell gas space volume 4.0x10 "c c The fission product inventories in the core are calculated based on three years (1095 days) of continnons operation at 3651 Mit, or 102% of rated power, by using a com uter code developed at Los Alamos and adapted to the GE computer system.(1 The inventories of some major fission products in the core at the time of reactor shutdown are given in Table 1. 2.1.2 Procedure Either the gas or water samples taken from the post accident sampling system are analyzed for major fission product concentrations by gamma ray spectrometry. If the concentration of a fission product in reactor water or drywell, corrected p decay.(w mthe time of reactor shutdown, is measured to be higher than the baseline concentration shown in Table 2 (see Section 3.1 for details), the extent of fuel or cladding damage can be determined directly from Figures 1-4 based on isotopes I-131, Cs-137 Ie-133, and Kr-85. Measurements of Cs-137 and Kr-85 activities are not very likely until the reactor has been shut down for longer than a few weeks and most of the shorter-lived isotopes have decayed.

If the concentration falls into the range where release of the fission product from the fuel gap or the molten fuel cannot be definitively determined, additional data may be needed to de t erm ine the source of fission product release (see below). It is recommended that both the water and gas phase sample s be measared in order to reduce the nacertainty in core damage estimations. 2.1.3 Supplementary Data In addition to the longer-lived isotopes, some shorter-lived isotope con-contrations may be measured in the sample. The ratios of isotope s released from either the fuel gap or the molten fuel are significantly different as shown in Table 3 (see Section 3.3 for detail), thus the source (fuel or gap) of release may be identified. Furthermore, some less volatile elements in the core may also start to release as the fuel starts to melt. If the less volatile fission products, such as isotopes of Sr, Da, La, and Ru (either soluble or insoluble), are found to have unusually high concentrs-tions in the water sample, Soec cleyEC-o-{ fu-el v^<tl4mq ety b4_ ' v/r4WEcl. i In a mixture of fission products 2.7h Sr-92 (1.385 MeV) and 40 h La-140 (1.597 MeV) shonid be relatively easy to identify and measure from a gar.na ray spectrun. More work, however, is needed to establish the baseline concentrations for those isotopes. 2.2 Specific Plant Aeolication 2.2.1 Plant Parameters The pertinent reactor parameters for selected plants currently being retrofitted with the post accident sampling system are tabulated in Table 4,(2) 2.2.2 Procedure The extent of core damage in an operating BTR can be de termined by comparing the measured concentrations of major fission prodnets in either the gas or water samples, after appropriate normalization, with the reference plant data. The following procedure is recommended. (1) Obtain the samples from the post accident sampling system, and the concentration of a fission product i (Cai in water or C g in gas) g is determined. I l (2) Correct the measured concentration for decay (mthe time of reactor shutdown. (3) Correct the measured gaseous activity concentration for temperature and pressure difference in the sample vial and the containment (torus) gas phase (see footnote on p. 3). ,.n- ,-,,,.,,-,,.,-e.- m.,.,-,--,,,,,.-a,.,,-.,,-w,,.,v ,n - - > ~,,, -,,,

(4) Calculate the fission product inventory correction factor (Section 2.2.2.2). (5) Calenlate the plant partmeter correction factor (section 2.2.2.3). (6) By using the corgegtfon fgg} ors, calculate the normalized concentration, or (section 2.2.2.1). (7) Use Figures 1 through 4 to estimate the extent of fuel or cladding d am a g e. 2.2.2.1 Comparison with Reference Plant Data The extent of core damage can be e stima te d from the measured fission product conentrations in either the gas or water samples, as described for the reference plant. However, the measured concentration must be corrected for the differences in operation power level, tine of operation, pr imary coolant mass and containment gas volume. C *g C,ge' xF x F, = or yg C 'g' = C e zF xF g yg g I Ref where C concentration of isotope i in the reference plant = coolant (Ci/g) b concentration of isotope i in the reference plant = l 8 containment gas (Ci/cc) l C,g = measured concentration of isotope i in the operating coolant at time, t (C1/g) I gg e h-g e 3 y--- ,,,.,,9+ --,,,m-. t, iye,. --wy -.aa .%--.-----.--,----m -.--_y_ --w.

3..

4_ C = measured concentration of isotope i in the operating Ig containment gas at time, t (Ci/cc)** = decay correction to the time of reactor shutdown e A = decay constant of isotope i (day) g t = time between the reactor shutdown and the sample tic;e (day) F = inventory correction factor for isotope 1 yg (see Section 2.2.2.2) F = containment ga s voinne correction f actor I (see Section 2.2.2.3) F, = primary coolant mass correction factor (see Section 2.2.2.3) 2.2.2.2 Inventory Correction Factor Inventory in reference plant p 11 Inventory in operating plant 3651 (1-e ) f[P(1-e i j) c' ib] j where P = steady reactor power operated in period j (MWt)* T = duration of operating period j (day)* j T* = time between the end of cperating period j and d time of reactor shutdown (day) For a particular short-lived isotope, i, a calculation for only a period of ~6 half-lives of reactor operation time before reactor shutdown should be accurate enough. It should be pointed out that the computer calculation of core inventory takes into account the fuel burnup, pintonium fission and mentron capture reactions. The correction factor calculated from this egnation may not be entirely accurate, but the error is insignificant in comparison to the nacertainties in the fission product release fractions (Table 5) and other assumptions (Section 3.2).

• In each period, the variation of steady power shon1d be limited to 120%.
  • The following correction for the measured concentration is needed if the (T,P ) are different from temperature and pressure in the sample vial 1 1 (T, P ):

that in the containment 2 2 si gi(vial) 12 w -, --,, --+ .w. ,, -,, - - -wm- .w. -~,,_ +w.

~5-l l 2.2.2.3 Plar.t Parameter Correction Factors eneratina clant coolant mass (a) w reference plant coolant mass (3.92x10' g) l l operatina plant containment ans volume (cc) g" reference plant containment gas vol.(4x102' cc) l In case the fission product concentrations are measured separately for the reactor water and suppression pool water or the drywell gas and the torus gas, the measured concentrations C,g or C would be averaged from the separate gg j measurements: (Conc, in Rx veter)x(Rx water mass)+(cone. in pool)x(nool water mass) C wi Reactor water mass + pool water { (Conc. In drywell)r(drYvell Ras vol)+(Cone.in torus)r(torus nas vol) ( gi drywell gas volume + torus gas volume i 3.0 TECHNICAL BASIS 3.1 Fission Product Concentrations in the Primary System Durine Reactor Shutdown Under Normal Overstion Conditions 3.1.1 Fission Product Concentrations in Reactor Water l l It is well kncwn that some volatile and water soluble fission products, mainly iodine and cesium isotopes, will be released (called spiking) from defect fuel rods when the reactor is shut down and depressurized. Based on Pasedag of NRC,(4) the maximum I-131 release would be 10 Ci per each pCi/sec telease rate during normal power operation. Acc ordi'c g to the GE design basis of I-131 releaso rate at 700 pCi/sec(5) a maaimum of 7000 Ci of I-131 may be released during reactor shutdown, and the concentration in reactor water wonid be 29 pC1/3 An analytical model to predict the magnitude of I-131 spiking following l reactor shutdown in operating IMRs has ' been reported by Brutschy et al l (3). The "best e st imate" concentration for I-131 has to be calculated based on the analytical mode 1(3) for the individual reactor according to its inel condition. However, if one adopts a standard 1-131 concentration I of 5x10-3 pCi/g or ~18 pCi/sec) as proposed by ANS(6), the nominal I-131 spiking is e stimated to be ~0.7 pC1/3 in the reference plant water. ' Ibis concentration is consistent with an average spiking concentration observed expe rimenta11y. (3 ) The resnits of these e stimations, including the Cs-137 soncentration, have been summarized in Table 2. Potentsal future research in this area will be discussed in Beetion 4. i a e

' 3.1.2 Noble Gas Concentrations in Drywell and Torus Gas Phase similar to the spiking magnitude, the noble gas activities in the drywell and the torns gas may vary significantly from reactor to reactor, mainly depending on the fuel condition and the steam leakage rate. In ont operating BTR when the Ie-133 release rate measured at the steaa jet air ejector (SJAE) was 1.5x10* pCi/sec* (compared to design basis release rate of 8200 pCi/sec), the noble gas concentrations in the dryvell were determined to be ~10-4 pCi/cc. f or Ze-133 and ~4x10-5 pCi/cc for Kr-85. These data may be considered zr upper limit values. at the time of measurement was the worst ever observed at an operating BTR. 3.2 Fission Product P.elease Source Terms Under Accident Conditions The source terms for the damaged core under accident conditions have been proposed by several inve stigators. (7.8) The "best estimate" release source terms for different chemical groups of fission products are summarized in Table 5. The release of fission products from the damaged core has been estinated to be a function of temperature,(9) and time after the loss-of-coolant accide nt. (7 ) In the present procedure, the fraction of fission product release from the core is assumed to be proportional to the fraction of core damage as suggested by Maliaanskas, et.al.(9) It is further assumed that the core is homo 5eneous so that each fuel rod has an identical exposure history. It hatemed Nadding rupture would occur over the temperature range from about 780' to 1100*C, and the entire fission product noble gas inventory in the fuel gap would be released. All other fission products in the fuel gap, which nay be present in a condensed phase, or as vapor in equilibrium with a condensed phase, will not be released as quickly as noble gases until the temperature is further increased. Acc ording to a model calculation,(7) pcrtions of the fuel may start to melt before the cladding is totally destroyed. 3.3 Isotopic Distribution in Fuel Gao Diffusion e gnations predict that the fractional release of radioactive isotopes from the fuel to the plents and void spaces should be inversely proportional to the square root of the decay constant for isotope reaching production-decay e quilibrism. (10,11) This prediction has been substanti-isted by e xperimental data reported by several investigators.(12-17) A soeparison of isotopic distributions in the total fuel inventory and the predicted distribution for some major fission products has been shown in Table 3. Thus, by measuring the ratios of fission product activities in either the gas or water samples, the scurce of fission product release may be semi quantitatively determined (see more discussion in Section 4).

• The fission product release patterm was found to be mostly

" recoil."

l . 3.4 Antseipated Chemical Behavior of Iodine and Coolant Chemistry Under Actident Conditions The results of measurements of Three Mile Island-2 (TMI-2)(18,19) indicate that the airborne radioiodine release was much lower when compared to the noble gas activity release (by a f actor of ~108 ). Extensive investigations at the Oak Ridge National Laboratory (ORNL) on the nature and quantity of fission product release from the over-heated fuel have concluded that cesium iodide CsI (B.P. = 1280'C) is the primary volatile species released from the fuel at elevated temperatures.(19) The behavior of iodine under loss-of-coolant accident (LOCA) conditions has been evaluated by Lin(21) and Campbell et al.(22) For iodine at a concentration of a few ppm in aqueous solutions, the redor reactions should be more predictable and formations of anomalous or organic specie s should be much smaller than that at very low concentrations as generally assumed for radioiodine release. If iodine is released as CsI, it should stay in water as the I-ion in a slightly basic sointion (mainly due to Cs ions which may be released as elemental Cs or Cs oxides in addition to CsI). Air oxidation (23) or radiation-induced oxidation of I,(24) is not very likely to occur in a basic solution. In I-to addition to the reducing nature of zirconium and iron metals in the core, the production of hydrogen from Zr-steam reactions should make the chemistry environment in the primary system favorable to reducing reactions for iodine. There are three known volatile f orm s of iodine, I,, HIO, and organic iodine. The f orm a tio n of I, from I-is not very likely in basic solutions. The existence of HIO has never been chemically identified due to its low stable concentration. The airborne species called HIO is one which behaves differently from I and organic iodine determined by using 3 the iodine species sampling method developed by Keller, et al.(25) l However, some convincing evidence has been given by Lin (26) that HIO, a l product of I, hydrolysis, is the second volatile. inorganic species in the gas phase when I, was initially added to water in equilibrium partitioning studies. 'Ib e partition coefficient increases with decreasing iodine concentration; at very low lodine concentrations, the total iodine partition coefficients have been determined to be ~8000 at 21*C and ~1600 at 72'C.(26) It must be pointed out that since both I, and HIO are very reactive species,

• a ny reducing impuritie s in water or on construction material surfacs would reduce I, or H10 to I and significantly reduce th e airborne iodine concentration.

-,,,,,r

~- -g-The mechanisms of converting inorganic iodine to organic iodine, which is generally observed in gasUhase at very low concentrations, are largely saknown. H oweve r, at least more than a stoichiometric amount of organic species (or carbon-contsicing compounds). species are should be readily available for reaction with iodine. As such organic limited, the resnits of several e xpe rim ent s (27) indicate that the yield of organic iodine decreases with increasing iodine concentration in the gas phase. Less than 0.1% conversion is expected when the airborne iodine larger.(27) The total iodine concentration concentration is 1 g/mi or conid be ~3 g/m* in the containment free air space if all iodine is assumed to become airborne. It is also important te realize that the organic

iodine, e.g.,

CH I is readily hydrolyzed in water (28) and basic solutions (29i at higher temperatures. The half-time of hydrolysis is ~20 i l min in water at 100'C and ~3 sec at 200*C, based on Heppolette and i Robertson's data.(28) %us, the very low release of iodine activitie s to the atmosphere in the Tf!I-2 accident can be explained in terms of the nature of iodine released from the fuel and the subsequent stabilization in water. Yater plays an important role in preventing iodine from release to the atmosphere. In the present procedure, alphe iodine activities are j assumed to stay in water, and the airborne activities are dominated by the noble gas fission products. The chemistry in the primary coolant any be significantly changed under accident conditions. Mainly due to the release of cesina, the water pH may i increase to ~10.5(21) and the water conductivity may increase from ~10 PS/cm (torns water quality specification) to as high as ~170 pS/cm. l 4.0 DISCUSSION AND SUGGESTIONS FOR FUTURE VORI It is evidcet that the uncertainty of gas release fractions for iodine and cesium are too large for an accur ate 'Talenla tion of the extent of core asurements is damage. Whil e additional experimental work in feel gap me,9t(( be appsrently r.asdad, the lower limit release fraction for iodi re-evsinsted by exanining the iodine spiking release data from defective fuel rodc tollowing normal operation shutdowns. Although the I-131 spiking data has been well documented previously(3), the analytical model may be refined to reflect more recent e xpe riment a l data. The maximum spiking release of I-131 estimated by Pasadag(4), based on pre-1973 data, is , too high, particularly when the improved fuels which are entrently used in most of the operating BTRs are considered. l - - - -

_9 The accuracy of core damage e stimation may be significantly improved by measuring more than iodine, cesium, and noble gas activities. Some less volatile but easy to measure isotopes of Sr, Ba, La, and Ru may be deter-mined in the water sample. More work, however, is needed to establish the release fractions as well as the baseline (shutdown spiking) concentrations for those isotopes. As mentioned in Section 3.3, it is possible to determine the source of fission product release by measuring the activity ratios of noble gases or iodine isotopes. It must be cautioned, however, each isotope should be accurately measured. Particular care must be excerised when the Ic-133 ac t iv ity is d e t e rmined is a mixture of other fission products with high coterntration because of its low gamma ray energy (81 kev). Additional work is required to perfect this procedure. 5.0 References 1. H. A. Careway, "' Calc ula ti o n of Fission Product Inventory and Spec tra-RADC101 Program", lEDO-25176 (October 1980). 2. E. L. Strickland and N. R. Cash, " Operating Plants Parameters ", NEDE-21398, Rev. 2 (September 1978), S. A. Hucik, private communi-cation. 3. F. J. Bruts chy, C. R. Hill, N. R. Horton, and A. J. Levine, " Behavior of Iodine in Reactor Water During Plant Shutdown sad Startup," NEDO-10585 (August 1972). 4. W. G. Pasedag, " Iodine Spiking in BWR and PWR Coolar.t Systems," Proc. Topical !!eeting on Thermal Reactor Safety, July 31-August 4,1977, Sun Valley Idaho. Conf-770708. 5. J. N. Skarpelos and R. S. Gilbert, " Technical Derivistion of ENR 1971 Design Basis Radioactive Material Source Terms," NEDO-10871. (!! arch 1973). 6. American Nuclear Society, "American Nation Standard Source Term Speci-fication," ANSI N237-1976. 7. L. L. Bonzon and N. A. Lurie, "Best-Estimate LOCA Radiation Signature" NUREG/CR-1237 (January 1980). 8. Reactor Safety Study An Assessment of Accident Risks in U. S. Commer-cial Nuclear Power Plant s, WASH-1400 Appendix VII, p. VII-13, U.S. Nuclear Regulatory Commission (Oct. 1975).

~ 9. A. P. Malinanskas, R. A. Lorenz, H. Albrecht and H. Wild, "LTR Source Terms for Loss-of-Coolant and Core Melt Accident, Proc. CSNI Special-ists Meetings on Nuclear Aerosols in Reactor Safety, held at Gatlinburg, Tenn., April 15-17, 1980, NUREG/CR-1724. 10. American Nuclear Society, '" Methods for Calenlating the Release of Fission Product s from Oxide Fuels," proposed ANS-5.4 standard (Nov. 1979). 11. R. A. Lorenz, J. L. Collins and A. P. Malinanska s, Nucl. Tech., f 6, 404 (1979). 1 12. Table VII 1-1, Reference 8. 13. G. M. Allison and H. K. Rae, "The Release of Fission Gases and Iodine from Defected U0: Fuel Elements of Dif ferent Lengtha," AECL-2206 (1965). 14. T. A. Yuill, et al, " Release of Noble Gases From U0 Fuel Rods," s IN-1346 (1969). 15. G. Jackson, D. Davico, M. J. Waterman, "The Effect of Neutron Finx on the Fission Product Gas Emission From UO at High Temperature s," s AERE-N-1607 (July 1965). f 16. N. V. Krasnoyarov, V' V. Konyashov, V. I. Polyakov, and Yu V. Chechetkin, Sov. At. Eng. (English),18., 89 (1975). 17. General Electric Company unpublished data. 18. A. P. En11, Trans. Am Nu-1. Soc.,M 91 (1980). 19. A. D. Miller, Ibid,M 633 (1980). 20. R. A. Lorenz, et al, " Fission Product Release from Highly Irradiated l LTR Fuel," NUEEG/CR-0722 (02NL/NUREG/TM-287/R2), February 1980. 1 i 21. C. C. Lin, " Anticipated Chemical Behavior of Iodine Under LOCA Conditions, " NEDO-25370 (Jan 1981). l 22. D. O. Campbell, A. P. Malinanskas, and W. R. Stratton, Nucl. Tech., }}. 111 (1981). l 23. M. Kahn and J. Kelinberg, " Radiochemistry of Iodine," NAS-NS-3062, ERDA (1977), p. 9. l L

.. 24. C. C. Lin, J. Inors. Nucl. Chem.42, 1101 (1980). 25. J. H. Keller, et al., " A Selective Absorbent Sampling System for Differentiating Airborne Iodine Specie s!' Proc. lith Air Cleaning Conferenc, CONF-700816 (1970), p. 621. ~ 26. C. C. Lin, " Volatility of Iodine in Aqueous Solutions," J. Inorg. Nucl. Chem., in press (1981). 27. A. K. Postma and R. W. Zavadoski, " Review of Organic Iodide Formation Under Accident Condition in Yater-Cooled Reactors," TASH-1233 (UC-80), October 1972. 28. R. L. Heppolette and R. E. Robertson, Proc. Roy Soc. (1959), A 252, 273. 29. M. Adachi, et al., J. Chem. Eng. Japan,1(5), 364 (1974). 30. A. P. Malinanskas, R. A. Lorenz, and J. L. Collins, Trans. Am. Nucl. Soc. }l, G51 (1979). ~ 6 e

TABLE 1 CORE INVENTORY OF MAJOR FISSION PRODUCTS IN A REFERENCE PLAhT OPERATED AT 3651 MWt FOR THREE YEARS MAJOR GAMMA RAY ENERGY INVEhTORY* * (INTENSITY) CHEMICAL GROUP ISOTOPE

• RALF-LIFE 108 Ci kev (v/d)

Noble gases Kr-85m 4.48h 24.6 151(0.755) Kr-85 10.727 1.1 514(0.0043) Kr-87

76. m 47.1 403(0.494)

Kr-88 2.84h 66.8 196(0.203),1530(0.109) Ie-133 5.25d 202. 81(0.371) Ie-135 9.09h 26.1 250(0.906) Halogens I-131 8.04d 96. 364(0.824) I-132 2.29h 140 668(0.99).773(0.762) I-133 20.8 h 201 530(0.87) I-134 52.6 m 221 847(0.954),884(0.653) I-135 6.59h 189 1132(0.231),1260(0.293) Alkali Metals Cs-134 2.06y 19.6 605(0.98),796(0.88) Cs-137 30.17y 12.1 662(0.85) Cs-138 32.2 m 178. 463(0.267),1436(0.75) Tellurium Group Te-132 78. h 138 228(0.88) Noble Metals Mo-99 66.02h 183 740(0.138) Ra-103 39.4 d 155 497(0.9) Alkaline Earths S r-91 9.52h 115 750(0.24) Sr-92 2.71h 123 1385(0.9) Ba-140 12.8 d 173 537(0.238) i Rare Earths Y-92 58.6 d 118 934(0.137) La-140 40.2 h 184 487(0.453),1597(0.953) Co-141 32.5 d 161 145(0.49) Ce-144 284.4 d 129 134(0.108) l Refractories Zr-95

46. d 161 724(0.435),757(0.543) l Zr-97 16.8 h 166 743(0.933)

• 0nly the representative isotopes which have relatively large inventory and considered to be easy to measure are listed here.
  • At the time of reactor shutdown.

-v- ,--w,.

I TABII 2 FISSION PRODUCI CONCENTRATIONS IN REACIDR VATER AND DRYWELL GAS SPACE DURING REAcr0R SHITIDOWN UNDER NORMAL CONDITIONS ISOTOPE REACTOR VATER. uCi/ r _DRYTELL GAS (uci/ce)__ UPPER LIMIT NOMINAL UPPER LI!!IT NOMIRAL I-131 29 0.7 Cs-137 0.3* 0.03** 10-4* 10-5** Ie-133 Kr-85 4x10-5* 4x10-6** Observed experiments 11y, in an operating BTR-3 with MK I containment, data obtained from GE unpublished docusent.

• • Assuming 10% of the upper limit values

, TABLE 3 RATIOS OF ISOTOPES IN CORE INVDEORY AND FUEL GAP ACTIVITY

• RATIO IN ACTIVITY RATIO
• IN ISUTOPE HALF-LIFE COPE INVESTORY FUEL GAP Kr-87 76 m

0.233 0.0234 Kr-88 2.84h 0.33 0.0495 Kr-85m 4.48h 0.122 0.023 Ze-133 5.25d 1.0* 1.0* I-134 52.6 m 2.3 0.155 I-132 2.28h 1.46 0.127 I-135 6.59h 1.97 0.364 I-133 20.8 h 2.09 0.685 I-131 8.04d 1.0* 1.0* n ble gas isotope concentration g,, l

• Ratio =

Ie-133 concentration Iodine isotope concentration for iodines I-131 concentration l l l

.'E TABLE 4 PLANT PARAMETERS (REF. 2) PRIMARY COOLANT

• CONTAINMENT CAS*

TORUS / RATED REACTOR SUPPRESSION DRYWELL CONTAISMENT REACTOR 'IYPE/ CON-POJER WATER MASS POOL WATER GAS VOL. GAS VOLUME B 9 (10 g) (l@cc) (103 cc) PLANT TAINTMENT DESIGN (MWt) (10 g) Standard BWR 6/III 3579 2.46 3.67 7.77 32.5 Brunswick-1/2 BWR 4/I 2436 2.14 2.48 4.65 3.46 Chinshan-1/2 BWR 4/I 1775 1.76 1.93 3.68 2.69 Cofrentes BWR 6/III 2894 2.04 3.14 6.91 32.43 Cooper BWR 4/I 2380 2.00 2.48 3.75 3.03 Drasden-2/3 BWR 3/I 2527 2.61 3.18 4.48 3.30 Duane Arnold BWR 4/I 1593 1.45 1.67 2.67 2.67 Fermi-2 BWR 4/I 3293 2.77 3.23 4.64 3.71 Fitzpatrick BWR 4/I 2436 2.14 -3.00 4.37 3.20 Hrnford-2 BWR 5/II 3323 2.74 3.17 5.75 4.08 Hatch-1 BWR 4/I 2436 2.00 2.47 4.07 3.20 Hatch-2 BWR 4/I 2436 2.00 2.47 4.12 3.11 Hope Creek-1/2 BWR 4/I 3293 2.93 3.34 4.79 3.78 Kuo sheng-1/2 BWR 6/III 2894 2.04 3.74 6.74 40.50 Limerick-1/2 BWR 4/II 3293 2.93 3.63 6.66 4.23 Millstone-1 BWR 3/I 2011 2.05 2.78 4.16 3.06 Monticello BWR 3/I 1670 1.75 1.93 3.80 2.76 NMP-1 BWR 2/I 1850 2.17 2.34 5.10 3.33 Oyater Creek BWR 2/I 1933 2.05 2.32 5.10 3.58 Peach Bottcm-2/3 BWR 4/I 3293 2.67 3.48 4.98 3.62 Pilgrhn BWR 3/I 1998 2.05 2.38 4.16 3.18 Susquehana-1/2 BWR 4/II 3293 2.92 3.60 6.79 4.36 Vermont Yankee BWR 4/I 1593 1.77 1.93 3.79 3.18 CTotal Primary Coolant Mass = Reactor Water + Suppression Pool Water Total Containment Gas Volume = Drywell Gas + Torus (or Primary Containment in MKIII) gas .o

e o R o, R EE ~ l o o 3 o q o

• . D =J o

o o o o o o I e s. a e o g o ~ ~ .!.gi s. s. .x. 8 8 8 o1 -g a a o o o o o o o yg E B. ~ e o m. o o R o, wE I e o 7 8 o te o i ~ g o o o o o e J, !? e ~ e o ~. 51 4 S I 9 S. I i l IE mx ch = t ra e o o o o -r T L2

• ~ * *

.0 e0 O g ~ ~ o, -o.E ~ a 8 8.!! 9 o 9 l l l e w = a s o o o o 3 ,3 4 R w m ~. ~o o 2 ~.o =.c 9 l I i 58 o -~ 5 9 l 9 y o a e o o o so w E g. . r. s o e o c I e. e. 9 7 9 9 .5== g e. ~ .o em e o a=*

== y a a c o o o o o o o o e. o

• :a a-e i.

2 a C N o t,: =. g g a. 8 8 31 z a e 0 p o -- o o o o o o o o Zg a a m g g~~ m s 1 o. a g 8 = s. 8 8 ,o ~. m ~ o y G s s s s s s s s oL z

c s

t. 0 [i 8! R 8 I 8 l l al ~ "~ 4 p."j o o o o o EL w

• 8 wm 4

2 t o I.a E a. 8 8 2 l l i T3 o m s u

• g

,.s J o o o w n J M$ ( Q m 2 g R o I. e I I Il 6 4 e ~ -} o o a o o g s- .g p. .j ylI f .o .a is a riz I e 2 ogl - z* v g * +E g X w E gn L La = c f i "$.e =r m u. [ gg .c,w o ~- e o _A eE =g 3 og Q c pE I-t ; i.- 9 ~8 y 25 $d 28 $D 25 Ed EDW EU i

4 FIGURE 1. Relationship between 1-131 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant .v 5 10 a. i Fuel Meltin ... l' t 4 ~ Upper Release Limit .h.._.._-__. . s._. 'I l s i_=- 1 13 e ' n. :. .-4 Best Estinate 9' ~. =. _-.. - j -4...2. 2.2:..- Lower Release Limit N.. a-10" 3,,y_ _ 1 // \\/ ~ 1 e i y_..q,,r.--y...' // s e 1 a 3 y'....g___,__._.._....____._____.. e __ u i i ./ e s .y.. ~ ..,/ I o,

1

- 3

-

1

,r -

y 3 - -J' - 10 2 -1 ,1 u s' // l // ./ 2 J y i ,m __.__.--._,:. f, c _ _. __./. ,Y c ^ '.:..y'f y - f._ _ } ,/.____...__.__...__.... -... '._ - - _ _... -. j x / /- u g 7 , jo ; .__s _ v c c /... A N /., .,/ '.. E 9 -- o / \\ e y/ Cladding Failure /. ~) i ' r., .. \\./'.' ' c . i..,,!. i li ~ l [... i. 8 ~ s . - Upper Release Limit g 10 / -m v s. x x Best Estimate f + , x m .i/ i .N __ Lower Release Limit ___.. _ _ _ _. i .r 4 .7 / m: +. r m1 .;__.____...2___. 1.. i.;/. 4 m.+ _J. -a. . t ?>: i i i r .. =:1 : 1:Yi n - 1.) 1 - -

l-i i

5.5 5 f l % I N i$ --l W :-i d / i i i ~

1. 0.;-

z-1 /

/

Normal Shutdown Conc. in Reactor Water

. - y, i s i

,..i,;.t s . i ;; i t : N,r. l. Upper Limit

29.0 uti/g i E

. r . 1 ] i 1. ? i:!.s; Nominal

0.7 uti/g

._1 =2 _ 9_5.. ;-?. -' ,3 3 _ . b1 n 2- .. ~ - / .s. ...a a ; dii-f4Msfj e..n: +_sW.s-152 1 i Hn"iWWI~I~ll!U . T~,i -T3h 44 .. j!...,.. _ l. 1 ....y. _m m...._...

r. : _3.. _.,.

..... _. -A. :.=:.r__ q_....2. : _:..:..:..n.. _=._ _ g=_=- = :,....:.. . 1 g*) Q.1 1.0 10 100 d h % Cladding Failure ' 1.0 10 100 4 % Fuel Melting e i

FIGURE 2. Relationship between Cs-137 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damace in Reference Plant. 10,g Fuel Meltin i 5 Upper Release Limit k--- ~ Best Estimate b4-

• x x

,7 -:-1 . - = Lower Release Limit 251 M a 5Oi : \\ ~I' ,1.. .i 10.3 ,/ ,/ \\. I t // T ./ / a s / _f ,.'/-/._.- .,/ _.. /[f. 1 ........ ~ _ _... -... i . / // 1 1 ,f a

// _,/

.), / . / I ~ ' j C* 10 1 FI / u = '/ / ' 1 Y/ / r ) ./ /.'. / .,.f _....Jl l4-. 1 L. i'l ./ I -+ 2 N-i t a; g i.. .,/ 3 .. ~ -.,., a _..~ ; 10- ,1 ~ l- /s s / \\ \\ / N N C \\ -..._N Cladding Failure

y' i,'

C x, ^. y ,7 / 3y .b . 3 -.4 n:-...i/ i . :: :, 'N N '_ Upper Release Limit _h.N Best Estimate N: c .j ;4 :.t 2.r#, :: . 7 e 1.Q [, Lwer Release Limit 0 e i J. i e

/

- a e NC l / - 4 !/ t .~,-/ ! :. i.. + 8 u . /. + - / i j ! + 8 id! : 8-M.2-!_kj l 4 ? ! ! 4 i i I i !i)' 4 !.i i t t 3*-- /. ** 3. e :p ::: :.... :. :: :-.

--- - : ?:

--1 3.. : : :. < :.:: : -:. :..:.. O. ..::3.x.1-: Norval Shutdown Conc. in Roactne Water i i# 4 + i' 4 q iii. Upper Limit

0.3 uCi/g

. aq.,'.m i y- .-i--- 2 2, ai + 4.. 4... .v

<;;4 ;;

4,i ; @ Nominal

0.03 uCi/g A /,

ai,

m

......m p-3 ;;. ip e.cw? : a m -N. .I3I. m. 8 1 EW--+_:_-:.L: _$3-1 li.'dhMTI ::. 4 ~i-t -if - F. :.1.,#E~ h4* 25

• E

^ g.r :

- ~ ~ =r -* --- -* L, 44

: - ~ ~

F-i 4:n; r.- 4 y - :n.-in-: 3 '::; 9 1 0' r : 2 2~-r ; ':-'-~ ~ ~- ~' 97 31; 2' i G 10-20.1 1.0 10 109 ._ t. F % Cladding Failure 1.0 10 100 % Fuel Melting t i l.

FIGURE 3. Relationship between Xe-133 Concentration in the Containnant Gas (Dryuell + Torus Gas) and the Extent of Core Damace in Rafarence Plant ~ 4 Fuel Melti.g .6 ~~ i -_ Upper Release Limit x mV Best Estimate d-[ 7 4- \\'7 - i --4 ' i-d:i. j..= l ~= i : : - Lower Release Limit & -/;- - -. ] ' - l - - ~ _ 1 ,m ,,u .}/ 1

1. //

/ ,,'r' q - _a.,e. s _,. _. _,.. _ _ _ _ _. _.... _. _ _ _. .I f // / .i;. / . __.s. _ u 2 > r,- as s .s sIO y u .s/ / e / s' / i s / / / l p q._.....,.. i... -a g w 'i . _ J. _ __ s' r .+ a .f .s- / i C e l i <- o i ';;;10 c .s S / 'y 'x Claddino Failure - - _... - --i / i N N 1 e ,.x x,-Upper Release Limit. _ 4 . _.. a .g /..___ / _ _ _ \\ ' Best Estimate / N a E /

i.

' Lower Release Limit j ~., ' ' j f: i, /. ..] .; q i :: v 1. 0. c Q I O r / ./ p ,,/ C ~f 5 ra . s 4 8 j 4 t 7, t. ' 4 j jli' ~ ,..,. / { d j/e a sy 4 -- ' _.; i.? I j j j. n ~4

ji l

f. 6 V '- i:.21:

• a ':

t-i 1 =1 ~ 3, .: :t._ _........ ..w_ . -. -~ . : ' ' ' ~ ~ ~ ~ ~ ~~ ~~ ~'"' 0*1 ~ 5,f !!or ul Operation Conc. in Drywell a.33 j 10.g uCi/cc . i c Upper Limit 9.:.i 4 10-5 uCi/cc - wn s. ,.. i.,' ::. Nominal n v 2 i. u. 4.i.a , 42. i ;- ..$ i

ii4.

L +i j jjiji.p-ifef7

' j 1 8.]j

.-J j ?t-pj 4... . -9 . 6 3 ,a tt ' C =- + 4.Ml := 3 rt:=-i -:!_i ti i4: :#.~ m.__ @. - H H il ~ T1-M-T741n .i +ti* 2 m._. m. ) g-2. :: =.. 2.:_._:. :.. 2.;.. : =...q._: _: :r..: = _-. _. __ _.- : _:.,..e..m..

-- + -. 2 :.,.2.u..i. n_..

2, 7 e., m _

u;.

0.1 1.0 10 100 'I i- % Cladding Failure l.0 10 100 h % Fuel Neitig

I

~

Containrent Gas (Drywell ion in th FIGURE 4. Relationshio between Kr-85 Concentra+danage in beference Plant. + Torus Gas) and the Extent of Core f 10 . 1 l Fuel Melt __ a s i ~ Upper Release Limit 1 i ' Best Estimate ...i j _j \\ % /, 10.- ~ T2 '" #-j.. Lower Release Limit g~ ' '.. s* ~8, i _ :.2..n ;. >1 x 1 1 ,I s, - . '/ s i t r/ / // / _} i ?- e - - _ - -.. _ _ _ _ _. - - _. 2_ -.__. . p .,y.., ! - g } n-s1 -) 3 1... ,f 's r 1

• f\\

-i . '.4' -/

• -i;

>J / ) u 'f , e/ // r V / ./ l .. --. _ -..,,y. '...., :../ r / r/ e i. y --_ l,u

7. y_.-

. : = --- / __l .: f. e d 4------ s 10~I o ,y -1 / "I c } E r c 's _s /* C1 adding Failure ^ -/ --- __.___+_m.._.-.-< J._ _----i.. O / d. . ' Upper Release Limit c /

/.

-\\ \\: Best Estimate ~ i e . j 2., Lower Release Limit 8 10 ; / i. 2 ./ / l I I f / / M 5 I I 4 / f. /, i. 1 c. ,f .P 4 +- .

• i:'

g. + i/ - s' r -.- i - ' i 2 1 i =t z-! - LU t e. ~ i cj - ) -/. iM. o f.= ... - -.d .gii'

1.. : - ?

i i. f_ :.,:s'.: .1 l ..d....... w ,,., j..;, g x' 10-1-/. ! :. M. ::

=:

2x i 5 ?!orral Operation Conc. in Dryviell .j r; I 4 x 10-5 uCi/cc a r e.. . f,? "'y' iiT i Upper Limit l 4 x 10-6 uCi/cc h-- m.: 3 a ll. Nominal NJ. J.' l.i

1-

- a !- i Zi -...4. i mr= =-_2 W.tpi!4 =s:- 2-18 d 4LHV G.Ma~~l Mbi!; ~%4= m i 4 i!!g ' u. I i =::w 2.:.:.2 2 4 : u : = ---:=: .:; u: : - -- w - :..::.p.::... y../.2..2 .,- w M :p -- - .:-1.

, 1 g,=: 1 0.

1.0 10 100 in 100 ! Cladding Failure ), P % Fuel Melt h g t 1}}