ML17139A509
| ML17139A509 | |
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| Site: | Susquehanna, Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 11/30/1981 |
| From: | Fidrych L, Lin C, Ruiz C GENERAL ELECTRIC CO. |
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| RPE-81CCL01, RPE-81CCL1, NUDOCS 8201180369 | |
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Text
1 C&RE TRANSMITTAL
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RPE 81CCLOl November 1981 PROCEDURES FOR THE DETERMINATION OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS PREPARED BY:
CHIEN C. LIN Principal Engineer T. A. GREEN,. MANAGER Chemical Retrofit REVIEWED AND APPROVED BY:
C. P. RUIZ$ AGER Radiological Process Engineering APPROVED BY:
R. L. COWAN, MANAGER Chemical and Radiological Engineering APPROVED BY:
L. F. FIDRYCH, ACTING MANAGER PLANT DESIGN AND ANALYSIS Q~lc QRO((fQ44$
NUCIZAR POWER SYSTEMS ENGINEERING DEPARTMIQFZ GENERAL ELECTRIC COMPANY VALLECITOS NUCLEAR CARTER, PLEASANTON, CALIFORNIA 94566 GENERAL ~ 'LECTRIC
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y
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~0 DISCLAIHER This document,was prepared by the. General Electric Company. Except as may be otherwise provided, neither the General Electric Company nor any of the contributors 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-
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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.
TABLE OF CONTENTS 1.0 OBJECTIVE AND SCOPE 2.0 PROCEDURES FOR DETERMINATIONS OF CORE DAMAGE 2.1 REFERENCE PLANT (BWR-6/238, 1
MARK III) 2.1.1 REFERENCE PLANT PARAMETERS 2.1.2 PROCEDURE 2.1.3 SUPPLEMENTARY DATA 2.2 SPECIFIC PLANT APPLICATION 2.2.1 PLANT PARAMETERS 2.2.2 PROCEDURE 2.2.2.1 COMPARISON WITH REFERENCE PLANT DATA 2.2.2.2 INVENTORY CORRECTION FACTOR 2.2.2.3 PLANT PARAMETER CORRECTION FACTORS TECHNICAL BASIS 3.0'.1 FISSION PRODUCT CONCENTRATIONS IN THE PRIMARY SYS~
DURING REACTOR SHUTDOWN UNDER NOR?QJ. OPERATION CONDITIONS 3-.1. 1 FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER 3.1.2 NOBLE GAS 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 3.4 ANTICIPATED CHEMICAL BEHAVIOR OF IODINE AND COOLANT CHEMISTRY UNDER ACCIDENT CONDITIONS 4.0 DISCUSSION AND SUGGESTIONS FOR FUTURE WORK 8
5.0 REFERENCES
~ I 1.0 OBJECTIVE AND SCOPE The purpose of this proceduro is to dotexminc the dcgx'ee of roactor core damage from thc measured fission product concentrations in either the vates or gas samples .taken from tho primary system under accident conditions.
The procedure involves calculations of fission product inventories in the core and the release of inventoxies into the primary system undex postulated loss-of-coolant accident (LOCh) conditions. The fuel gap fission products are assumed to bo seleasod upon the rupture of fuel cladding, and the majority of fission psoduct inventories in the core vill be seleased vhen the fuel is melted at higher tempcratuses. h BWR-6/238 with a 'Mark III containment hpplication is used. as a reference of plant in the for demonstration of this procedure. tho psocedurc any othex type or sixo of boiliag vatex reactor (BWR) is described.
2.0 PROCEDURES FOR DETERMINATIONS OF CORE DhÃhGE 2.1 Reference Plant BWR-6 238 Mark III 2.1.1 Reference Plant Paxameters The pertinent plant parametess for the reference plant ax'e given below:
Rated reactor thermal pover 3579 MWt Number of fuel bundles 748 bundles Total primary coolant mass (reactor vates plus suppression pool vates) 3.92zlO>>g Total containment aad dsyvell gas space volume 4.0zlOx4cc The fission product invontoxies ia the core are calculatod based on three yoars (1095 days) of continuous opexation at 3651 MWt, or 102% of rated power, by usiag a computer codo developed at Los hlamos and adapted to the GE computes system.(1) The invontorics of some major fission psoducts in the core at tho timo of reactor shutdown aro given in Table 1.
2.1.2 Psoceduxe Eithex the gas os vates samples taken from the post accidont sampling systc'm are analyxod fos major fission pxoduct concentrations by gamma ray spoctrometry. If the conceatxation of a fission ofproduct reactor ia xeactor vatcr shutdown, is or dryvell, corsected the decay to tho time moasurcd to be higher than the baseline concontration shovn in Table 2 (see Soction 3.1 for details), the extent of fuel or cladding damage can be determined diroctly fxom Figures 1-4 based on isotopos I-131, Cs-137 Xe-133, aad Ir-85. Measuromonts of Cs-137 and Ks-85 activities are not very likely until the reactor has boon shut down for longer than a fov weeks and most of the, shoster-lived isotopes have decayod.
If the conccntxation falls into the range @here release of the fission product from the fuel gap or the molten fuel cannot bc definitively determined, additional data may bc needed to 'determine the source of fission product release (see below).
It is recommended that both the eater and gas phase samples be measured in order to reduce the uncextainty in core damage estimations.
2.1.3 Supplementary Data In addition to the longer-lived isotopes, some shorter-lived isotope con-centxations may be measured in the sample. The ratios of isotopes released from eithex the fuel gap or the molten fuel are significantly different as shorn in Table 3 (see Section 3 .3 fox 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, Ba, La, and Ru (eithex'oluble or insoluble), axe found to have unusually high concentra-tions in the water sample, a fuel meltdown to some extent may be assumed.
In a mixture of fission products 2.7h Sx-92 (1.385 MeV) and from 40 h La-140 (1.597 MeV) should be relatively easy to identify and measure a gamma ray spectrum.'ore vork, however, is needed to establish thc baseline concentrations for those isotopes.
2.2 S ecific Plant h lication 2.2.1 Plant Parametexs Thc pertinent reactor parameters for selected plants currently being retrofitted arith the post accident sampling system are tabulated in Table 4.(2) Similar information is available fox'll Bus.
2.2.2 Pxocedure The extent of core damage in an operating BVR can be determined by comparing the measuxed concentrations of major fission pxoducts in either the 'gas or eater samples, after appxopriate normalization, arith the reference plant data. The folloving proceduxe is recommended.
(1) Obtain the samples from the post accident sampling system, and the concentration of a fission product i (Qi in eater or Cgi in gas) is'etermined.
(2) Coxrect the measured concentration fox decay to the time of reactor shutdown.
(3) Correct the measured gaseous activity concentration for temperature and pressure differ'ence in the sample vt.al and the containment (torus) gas phase (see footnote on p. 3).
0~ ~~ g
~ I
~
1
~~ ~0 (4) Calculate the fission product inventory correction factox (Section 2.2.2.'2).
(5) Calculate the plant parameter corxecti.on factox (section 2.2.2.3).
(6) By using the corge~tion
~<e fgc$
0 ors, calculate the normalized concentration, C or C (section 2.2.2.1).
(7) Use Figures 1 through 4 to estimate the eztent of fuel or cladding damage.
2.2.2.1 Compari.son with Reference Plant Data The eztent of core damage can be estimated from the measured fission product concntrations in either the gas or water samples, as described for the xeference plant. However, the measured concent'ration must be corrected for the differences in operation power level, time of operation, primary coolant mass and containment gas volume.
+ f i'it C ie zFIizF Ii v or Kit
~
C C ie Zi z F g z F>i
-Re f ~
where Pi concentration of isotope i in the xeference plant coolant (Ci/g) c-Re f ~ concentxation of isotope i in the reference plant gi containment gas (Ci/cc)
C vi measured concentration of isotope i in the'pexating coolant at time, t (Ci/g)
0 C
i~ measured concentxation of isotope i in the operating containment gas at time, t (Ci/cc)~~
e ~ decay co'rrection to the time of reactor shutdown i~ decay constant of isotope i (day) t ~ time between the reactor shutdown and the sample time (day)
FIi inventory correction factor for isotope i (see Section 2.2.2.2)
F ~ containment gas volume correction factor (see Section 2.2.2.3)
I'
~ primary coolant mass correction factor (see Section 2.2.2.3) 2.2.2.2 Inventory Correction Factor Inventor in reference lant Ii Inventory in operating plant
-1095 1 3651 1-e i fP(1 ij) i j]
where P ~ steady reactor power operated in period j (Ht)e Tj ~ duration of operating period j (day)e Tj ~ time between the end of operating period j and time of reactor shutdown (day)
For a particular short-lived isotope, i, a oalculation for only a pexiod of H 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, plutonium fission and neutron capture reactions. The correction factor calculated from this equation may not be entirely accuxate, but the error is insignificant in comparison to the uncertainties in the fission product release fractions (Table 5) and other assumptions (Section 3 .2).
eIn each period, the variation of steady powex should be limited to + 20%.
+eThe following correction fox the measured concentration is needed if the temperature and pressure in the sample vial (Tl,P1) are diffexent from that in the containment (T2, P2):
C ~ C x 2 1 gi gi(vial) P1T2
~~ ~0 2.2.2.3 Plant Parameter Correction Factors F cs o eratin lant coolant mass reforoaco plant coolant mass (3.92xl0~ g)
F e o ex'atin lant containment as volume cc g reference plant containmeat gas vol.(4z10x cc)
Ia case the fission product concontrations aro measured separately for the xoactor vates and supprossion pool vator or the dryvell gas and the torus gas, the measured concentrations Cvi measurements:
i ox C i vould be avexaged from the separate Conc. in Rx vatex' Rx vater mass + conc. in ool x ool water mass C
vi Reactox vates mass + pool vater conc. in dr ell z dr ell as vol + conc. in torus x torus as vol gi dryvoll gas volume + torus gas volume 3.0 TECHNICAL BASIS 3 .1 Fission Product Concentrations in the Primar S stem Burin Reactor Shutdown Under Normal oration Conditions 3 .1.1 Fission Product Concentrations in Roactor Vates It is veil kaovn that some volatile and vatex soluble fission products, vill be roleasod (called spikiag) from mainly iodine and cesium isotopes, defect fuel xods when the reactor is shutdovn and depressurixed. Based oa Pasodag of NRC (4) the maximum I 131 xelease vould be 10 Ci per each pCi/sec x'eleaso rate during normal povor operation. According to thc GE dosign basis of I-131 xelease rate at 700 pCi/sec(5) a maximum of 7000 Ci of I-131 may bo released during reactor shutdown, and the concentration in reactor water vould be 29 pCi/g.
hn analytical modol to predict the magnitude of I-131 spiking following reactor shutdown in operating Bus has been reported by Brut schy o t al (3) . The gabe st o stimate" ooncontration f or I-131 has to be calculated based on the analytical . model(3) for tho individual reactor according to its fuel condition. Hovover, if one adopts a standax'd I-131 concentration of 5z10 3 pCi/8 or 18 pCi/sec) as proposed by ANS(~), the aominal I-131 spiking is ostimated to bo W.7 pCi/g in the xefexence plaat water. Thi>>
concentration is consisteat vith aa avexage spiking concentration observed ezporimentally. (3) The results of these e stimations, including the Cs-137 concentration, have beon summarized in Table 2.
Potential futuro xesoarch in this area will be discussed ia Section 4.
3.1.2 Noble Gas Conc'entrations in Dxyvell and Torus Gas Phase Similar to the spiking magnitude, the noble gas activities in the dr~ell and the torus gas may vary significantly from xeactor to reactor, mainly depending on the fuel condition and the steam leakage rate. In an operating NR when the Xe-133 release rate measured at the steam jet air ejector (LTAE) vas 1.5x10~ pCi/sece (compared to design basis release rate at 8200 pCi/sec), the noble gas concentrations in the dryvell vere determined to be 10 4 pCi/cc fox Xe-133 and 4x10 5 pCi/cc for Xx-85. These dati may be considered as the upper limit values since the fuel condition at this plant at the time of measuxement @as the worst ever observed at an operating BWR.
3.2 FissionProduci Release Source Terms Under hccident Conditions The source terms for the damaged core under accident conditions have been proposed by several investigators.(7i8) The ebest estimate" release source terms fox different chemical groups of fission products are summarixed in Table 5.
The release of fission products from the damaged core has been estimated to be a function of temperature,(9) and time after the loss-of-coolant accident.(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 Malinauskas, et.al.(9) It is fuxthex assumed that the core is homogeneous so that each fuel rod has an identical exposure history. The fuel cladding rupture would occur over the temperature range from about 780e to 1100oC, and the entire fission product noble gas inventory in the fuel gap would be released. All other fission products in the fuel gap, vhich may be present in a condensed phase, or as vapor in equilibrium vith a condensed phase, vill not be released as quickly as noble gases until the temperature is further increased. According to a model calculation,(7) portions of the fuel may start to melt before the cladding is totally destroyed.
3.3 Isoto ic Distribution in Fuel Ga Diffusion equations predict that the fractional release of radioactive isotopes from the fuel to the plenum and void spaces should be inversely proportional to the square root of the decay constant for isotope reaching production-decay equilibrium.(10ill) This prediction has been substanti-iated by experimental data reported by several investigators.(12 17) A comparison of isotopic distributions in the total fuel inventory and the predicted distribution fox some major fission products has been shorn in Table 3. Thus, by measuring the ratios of fission product activities in either the gas or eater samples, the source of fission product release may be semi-quantitatively determined (see more discussion in Section 4).
eData obtained from GE unpublished document, DRF 268-DEV00009. The fission product release pattern was found to be mostly "recoil."
3.4 hntici ated Chemical Behavior of Iodine and Coolant Chemistr Under hcc ident Condit ions The results of measurements of Three Mile Island-2 (TMI-2) (18~1>) indicate that the airborne radioiodine release @as much lover shen compared to the noble gas activity release (by a factor of -10s). Hztensive investigations at the Oak Ridge National Laboratory (ORNL) on the natuxe and quantity of fission product xelease from the over-heated fuel have concluded that cesium iodide CsI (B.P. ~ 1280oC) is the primary volatile species released fxom the fuel at elevated temperatuzes.(1>) The behavior of iodine under loss-of-coolant accident (LOCh) conditions has been evaluated by'in(21) and Campbell et al.(22)
For iodine at a concentrati.on of a few ppm in aqueous solutions, the xedoz xeactions should be more predictable and formations of anomalous oz organic species should be much smaller than that at very los concentrations as genexally assumed for radioiodine release. If iodine is released as CsI, it should stay in <<ater as the I ion in a slightly basi,c solution (mainly,due to Cs ions which may be released as elemental Cs or Cs ozides in addition to CsI). hir ozidation(23) or radiation-induced ozidation of I to Is(24) is not very likely to occur andin iron a basic solution. In-addition to the reducing natuxe of zirconium metals in the core, the production of hydrogen from Zr steam reactions should make the chemi.stry envi.xonment in the primary system favorable to reducing reactions for iodine.
There are three knorn volatile forms of iodinei Is, HIO, and organic iodine. The formation of Is from I is not very likely in basic identified due-solutions. The ezistence of HIO has never been chemically to its los stable concentration. The airborne species called HIO is one vhich behaves di.fferently from Is and oxganic iodine determined by usinII the iodine species sampling method developed by Iellex, et al.(>>i However, some convincing evidence has been given by Lin (26) that HIO, a product of Is hydrolysis, is the second volatile .inorganic species in the gas phase shen Is was'nitially added to rater in equilibrium partitioning studies. The parti.tion coefficient increases vith decreasing iodine concentxation; at vexy los iodine concentxations, the total iodine partition coefficients have been determined to be 8000 at 21eC and 1600 at 72oC.(26) It must be pointed out that since both Is and HIO are very reactive species, any reducing impurities in eater or on construction material surfacs would reduce Is or HIO to I and significantly reduce the airborne iodine concentration.
~ ~ l l The mechanisms of converting inorganic iodine to organic iodine, rhich is generally obsexved in gas phase at very los concentrations, are largely unknown. It is certain, however, that at least more than a stoichiometric amount of organic species (or carbon-containing compounds) should bc readily available for reaction rith iodine. hs such organic species are limited,'he results of several ezpeximents (27) indicate that the yield of organic iodine decreases arith increasing iodine concentration in the gas phase. Less than 0.1% conversion, is expected vhen the airborne iodine concentration is 1 g/ms or larger.(27) The total iodine concentration could be 3 g/ms in the containment free air apace i,f all iodine- is assumed to become aixborne. It is also important to realize that the organic iodine, e.g., CHsI is readily hydrolyxed in water(2g) and basic solutions(2>> at higher temperatures. The half-time oi hydrolysis is -20 min in eater at 1004C and 3 sec at 2004C, baaed on Heppolette and Robertson's data.(2S)
It is obvious that the very lor release of iodine activities to the atmosphere in the TMI-2 accident can be easily ezplained in terms of the nature of iodine released from the fuel and the subsequent stabilization in vates. Water plays an important role in preventing iodine from release to the atmosphere. In the present procedure, allthe iodine activities are assumed to stay in eater, and the airborne activities are dominated by the, noble gas fission products.
The chemistry in the primary coolant may be significantly changed under accident conditions. Mainly due to the release of cesium, the water pH may increase to IO.S(21) and the eater conductivity may increase from >>10 pS/cm (torus eater quality specification) to as high as 170 pS/cm.
4,0 DISCUSSION hND SUGGESTIONS FOR FUTURE VORX It is obvious that the uncertainty of gas release fractions for iodine and cesium axe too large for an accurate calculation of the eztent of core damage. Vh'lie additional ezperimental cwork in fuel gap measurements is apparently needed, the lover limi.t release fracti.on for iodine may be re-evaluated by ezamining the iodine spiking release data from defective fuel rods following normal operation shutdowns.
hlthough the I-131 spiking data has been mell documented previously(>),
the analytical model may be refined tn reflect more recent ezpeximental data. The mazimum spiking release of I-191 estimated by Pasadag(4),
baaed on pre-1979 data, is obviously too high, particularly shen the improved fuels which are currently used in most of the operating Bus are considered.
The accuracy of'ox'e damago ostimation may be significantly improved by measuring mox'e than iodino, cesium, and noble gas activt.ties. Some less vola'tile but easy to measuxe isotopes of Sr, Ba, La, and Ru may be deter mined in the wator samplo. Moro work, however, is needed 'to establish the roloase fractions as well as tho basoline (shutdown spiking) concentrations for those isotopes.
hs montionod in Section 3 3, it is possiblo to determino the source of fission product release by measuring the activity ratios of noble gases or iodine isotopes. It must be cautioned, however, oach isotope should be accurately measured. Particular care must be ezcerised whon the Xe-133 activity is dotoxmined in a mixtuie of other fi.ssion products with high concentxation because of its low gamma ray energy (81 koV).. Additional work is required to pexfoct thi.s procedure.
5.0 Rofoxences
- 1. H. h. Caroway, ">>Calculation of Fission Product Invontory and Spectra RADC101 Program", %UN-25176 (Octobez 1980) .
- 2. E. L. Strickland and N. R. Cash, " Oporating Plants Parametexs ~~,
NEDE-21398, Rev. 2 (September 1978), S, h. Hucik, private communi-cation.
- 3. F. J'. Brutschy, C.. R. Hill, N. R. Horton, and h. J'. Levine, >>Behavior of Iodiao in Reactor Water During Plant Shutdown aad Startup,>>
NEDO-10585 (August 1972),
- 4. V. G. Pasedag, >>Iodine Spiking ia SWR and PlfR Coolant Systoms," Proc.
Topical Meeting on Thormal Reactor Safoty, July 31-August 4, 1977,Sun Valley Idaho. Conf-770708.
~
- 5. X. M. Skarpolos and R. S. Gilbert, "Technical Dex'ivi,ation of BWR 1971 Design Basis Radioactivo Material Souxco Terms,>> NEDO-10871. (March 1973).
- 6. American Nuclear Society, >>hmerican Nation Standard Source Term Speci-fication,>> hNSI N237-1976.
- 7. L. L. Boaxon aad N. h. 'Lurie, "Best-Estimate LOCh. Radiation Signature" NUREG/CR-1237 (J'anuary 1980),
- 8. Reactor Safety Study, hn Assessment of hccident Risks ia U. S. Commex-cial Nuclear Powor Plants, TASH-1400, hppondiz VII, p. VII-13> U.S.
Nuclear Regulatory Commission (Oct. 1975).
- 9. A. P. Malinauskas, R. h. Lorenx, H. Albrccht and H. Wild, "LKR Source Terms for Loss-of-Coolant and Core Melt hccident," Proc. CSNI Special-ists Meetings on Nuclear herosols in Reactor Safety, held at Gatlinburg, Tenn., hpril 15-17, 1980, NUREG/CR-1724 hmcrican Nuclear Society, "'Methods for Calculating the Release of
'0.
Fission Products from Oxide Fuels,<< proposed hNS-5.4 standard (Nov.
1979).
- 11. R. h. Lorenx, J. L. Collins and h. P. Malinanskas, Nucl. Tech., ~6, 404 (1979).
- 12. Table VII 1-1; Reference 8.
13 . G. M. hllison and H. K. Rae, "The Release of Fission Gases and Iodine from Defeotcd UOs Fuel Elements of Different Lengths," hECL-2206 (1965).
- 14. V. h. Yuill, et al, " Rclcase of Noble Gases From UOs Fuel Rods, "
IN-1346 (1969).
- 15. G. Jackson, D. Davico, M. J. Vaterman, "The Effect of Neutron Fluz on the Fission Product Gas Emission From UOs at High Temperatures, "
hERE-M-1607 (July 1965).
- 16. N. V. Krasnoyarov, V. V. Konyashov, V. I. Polyakov, and Yu V.
Chcchetkin, Sov. ht. Eng. (English),38, 89 (1975).
- 17. General Electric Company unpublished data.
- 18. h. P. Hull, Trans. ha Nucl. Soo.,34 91 (1980).
- 19. h. D. Miller, Ibid,34 633 (1980).
- 20. R. h. Lorenx, et al, <<Fission Product Release from Highly Irradiated LVR Fuel, " NUREG/CR-0722 (ORNL/NUREG/TM-287/R2), February 1980.
- 21. C. C. Lin, <<hnticipated Chemical Behavior of Iodine Under LOCh Conditions, " NEDO-25370 (Jan 1981).
- 22. D. 0. Campbell,. h. P. Malinauskas, and lf. R. Stratton, Nucl. Tech.,
$ 3 111 (1981).
- 23. M. Kahn and J. Kelinberg, "Radiochemistry of Iodine, " NhS-NS-3062, ERDh (1977), p. 9.
- 24. C. C. Lin, Z. Inorg. Nucl. Chem.42, 1101 (1980).
- 25. S. H. Keller, et al., " h Selective Absorbent Sampling System for Differentiating Airborne Iodine Species," Proc. 11th hir Cleaning Conferenc, CONF-700816 '(1970), p. 621.
- 26. C. C. Lin, "Volatility of Iodine in Aqueous Solutions," X. Inorg.
Nucl. Chem., in press (1981).
- 27. h. K. Postma and R. W. Zavadoski, "Review of Organic Iodide Formation Under Accident Condition in Water-Cooled Reactors," WASH-1233 (UC-80),
October 1972.
- 28. R. L. Heppolette and R. E. Robertson, Proc. Roy Soc. (1959), h 252, 273.
- 29. M. Adachi, et al., J. Chem. Eng. Japan,g(5), 364 (1974).
- 30. A. P. Malinauskas, R. h. Lorene, and J. L. Collins, Trans. Am.
Nucl. Soc. 32, G51 (1979).
~0 TABLE 1 CORE INVENTORY OF HhJOR FISSION PRODUCTS IN A REFERENCE PLANT OPERATED AT 36S1 NWt FOR THREE YEARS MAJOR GAMMA RAY ENERGY INVENTORY' (INTENSITY)
CHEMICAL GROUP ISOTOPE+ HALF-LIFE 10'i XeV d Noble gases Xr-85m 4.48h 24.6 151(0.755)
Xr-85 10.72y 1.1 514(0.0043)
Xr-87 76. m 47.1 403(0.494)
Xr-88 2.84h 66.8 196(0.203),1530(0.109)
Xo-133 5.25d 202. 81(0.371)
Xe-13S 9.09h 26.1 250(0.906)
Halogens I-191 8.04d 96. 964(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(0.9S4),884(0.653)
I-195 6.59h 189 1132(0.231).1260(0.293)
Alkali Metals Cs-134 2.06y 19.6 605(0.98),796(0.88)
C@-137 30.17y 12.1 662(0.85)
Cj-138 32+2 XT 8. 463(0.267),1436(0.75)
Tel lur ium Group To-132 78. h 198 228 (0. 88)
Noble Metals Mo-99 66.02h 183 740(0'.138)
Ru-103 39.4 d 155 497(0.9)
Alkaline Earths Sr-91 9.S2h 115 750(0.24)
Sr-92 2.71h 123 1385(0.9)
Ba-140 12.8 d 173 537 (0.29 8)
Rare Earths Y-92 58.6 d 118 934(0.137)
La-140 40.2 h 184 487(0.453),1597(0.953)
Ce-141 32.5 d 161 145(0.49)
Ce-144 284.4 d 129 134(0.108)
Ro fractorio s Zr-95 46. d 161 724(0.435),757(0.543)
Zr-97 16.8 h 166 743(0.933) oOnly the representative isotopes which havo relativoly largo inventory and considerod to be easy to measure aro listod hero.
eeht tho time of reactor shutdown.
't ~
( ~ ~
ThBLE 2 FISSION PRODUCT CONCEKHUlTIONS IN REACTOR VhTER
-hND DRYWELL GAS SPhCE DURING REACD)R SHRUNKEN UNDER NORMAL CONDITIONS ISOTOPE REhCIOR WATER Ci DRYNELL GAS Ci cc UPPER LIMIT NOMINAL UPPER LIMIT NONINhL I-131 29 0.7 Cs-137 0.3+ 0.03++
Xe-133 10 4+ 10-5ee Xr-85 4x10-Se 4rl0 6ee Observed ezperimentally, in an operating BWR-3 with MK I containment, data obtained from GE unpublished document, DRP ?68-DEVW009.
hssuming 10% of the upper limit values
~0 ~0 IP TABLE 3 RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP ACTIVITY RATIO IN ACI'IVITYRATIONS IN ISOTOPE HALF-LIFE CORE INVENTORY FUEL GAP Kr 87 76 m 0.233 0.0234 Xr-88 2. 84h 0.33 0.0495 Xr 85m 4.48h 0.122 0.023 Xe-133 5.25d 1.0e 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 1.0e 1.0e Ratio ~ noble gas isotope concentration for noble gases Xe-133 concentration Iodine isotope concentration for iodines I-131 concentration
~e ~0 ~ e,
~
t TABLE 4 PLANT PARAMETERS (REF. 2)
PRIMARY COOLANT* CONTAINMENT GAS*
TORUS RATED REACTOR SUPPRESSION DRYWELL CONTAIEHENT REACTOR TYPE/CON- WATER MASS POOL WATER GAS VOL. GAS VOLUME PLANT TAINTNENT DESIGN POWER
~MWt 1(P ) ~P QP ) "!N*>
Standard BWR 6/III 3579 2.46 3. 67 7.77 32.5 3.46 Brunswick-1/2 BWR 4/I 2436 2.14 2.48 4.65 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 3.75 32.43
- 3. 03 Cooper BWR 4/I 2380 2.00 2.48 3.30, Dresden-2/3 BWR 3/I 2527 2.61 3.18 4.48 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 Hanford-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 Eope 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 4.23 Limerick-1/2 BWR 4/II 3293 2.93 3.63 6.66 3.06 Millstone-1 BWR 3/I 2011 2.05 2.78 4.16 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 Oyster Creek BWR 2/I 1933 2.05 2.32 5.10 3.58 Peach Bottom-2/3 BWR 4/I 3293 2.67 3.48 4.98 3. 62 Pilgrim 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
- Total Primary Coolant Mass ~ Reactor Water + Suppression Pool Water C
Total Containment Gas Volume ~ Drywell Gas + Torus (or Primary Containment in MKIII) gas
TABLE 5. BEST-ESTIMATE FISSION PRODUCT RELEASE FRACTIONS(Ref. 7,8) a Gap Release Meltdown Release Oxidation Release Va ization Release ower pper ower pper ower pper ower pper Nominal Limit Limit Nominal Limit Limit Nominal Limit Limit Nominal Limit Limit Noble Gases 0.030 0.010 O. I2 0+873 0.485 0 970 0.087 0.0180.091 O.OIO O.OIO O.OIO (Xe, Kr)
Hal s O.0I1 O.OOI 0.20 0.$ 83 0.492 0.983 0.08$ 0.078 0.09$ O.OIO O.OIO O.OIO hlkali Metals 0.030 0.004 0 30 0.160 0.380 O.gff 0. I90 O.I90 O.I90 (Cs, Rb)
Tellurium Group O.OOOI 3xlO 0.04 O.ISO 0.03 0".230 O.fl0 0.340 0.680 0.340 0.340 0.340 (Te, Se, Sb)
Noble Metals 0.030 O.OI 0. IO 0.$ 13 0.716 0.970 0.00f O.OOI 0.024 (Ru,Rh,Pd,Mo,Tc) hlkallne Earths IxlO 3xlO 0.0004 0.100 0.02 0.20 0.009 0.002 0.043 (Sr, S )
Rare Earths 0.003 0.00l O.OI I
0.010 0.002 0.050 (Y LaiCeiNdspro Eu,Pm,5m,Np,Pu)
Refractories 0.003 0.00 I O.OI (Zr, Nb)
Notet Recent values of the gap release measured at Oak Ridge National Laboratory (Ref 30) are significantly lower. For the stable and long-lived members of the chemical groups they report O.OI27 for the noble gases, 0.00033 lor the halogcns, and 0.00023 for the alkali metals.
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P ATTACEKNT D
SUMMARY
OF PASS OPERATING PROCEDURES Four procedures have been written governing the operation of the PASS. These procedures cover the requirements for obtaining representative samples from the various sample points as well as the procedural steps for operation of the sample system. The general contents of these procedures are given below:
- 1) PASS Small Volume Liquid Sample'his procedure is used for obtaining a 0.1 to 10 ml diluted liquid sample.
The procedure covers the following areas: a) Health Physics precautions before going to, at, and leaving the PASS station (protective clothing, special dosimetry, survey instrumentation, etc,) b) Preparation of the chemistry laboratory for receiving high activity samples (shielding) c) Proper valve lineups to get desired samples (reactor coolant or suppression pool) d) Conditions required to obtain representative samples from the various sample points. i) use sample from RHR system in shutdown cooling mode rather than from jet pump when reactor depressurized (<135 psig) ii) sample line flush volume equal to three times line volume iii) operation of RHR system in suppression pool cooling mode for at at least 30 minutes prior to sampling from suppression pool. iv) reactor water level at or near normal level and reactor residual power greater or equal to 1/ of rated power to assure natural circulation path prior to sampling from jet pump instrument line. v) reactor water level raised to the point where water flows from the steam separators prior to sampling from the RHR system in the shutdown cooling mode. e) Operation of sample station f) Transportation of sample to laboratory g) Sample preparation for radioisotopic counting
- 2) PASS Large Liquid/Dissolved Gas Sample This procedure covers the steps required to obtain a 10 ml undiluted liquid sample and/or a 15 ml sample of dissolved gas. This procedure covers the same general areas of radiological protection, valve lineup, flushing requirements and conditions for obtaining representative samples as the "Small Volume Sample" procedure discussed previously. In addition it contains the specific steps needed to obtain a dissolved gas sample and/or 10 ml of degassed, undiluted liquid sample.
- 3) PASS 10 ml Gas Sample This procedure describes the steps necessary to obtain a representative sample of gas from the drywell, wetwell, or secondary containment.
The procedure contains instructions in the following areas: a) Health Physics precautions b) chemistry laboratory preparation to handle high level activity gas samples c) Proper valve lineups to get desired samples: wetwell gas (2 locations), drywell gas high level, drywell gas mid level, secondary containment atmosphere d) Flush times to obtain representative samples (equivalent to 5 times volume of line) e) operation of sample station f) sample preparation for radioisotopic counting and gas chromatographic analysis.
- 4) PASS Iodine/Particulate Sample This procedure details the steps required to collect an air particulate and/or iodine sample using the PASS. This procedure generally parallels the 10 ml gas sample procedure with respect to radiological protection, chemistry laboratory preparation, valve lineups, and flush times. The procedure assures representative sampling of particulates and iodine" by requiring that the gas sample temperature as measured at the sample station is at least 10 F above the temperature of the atmosphere in the area being sampled. This will prevent condensation of moisture with attendant loss of iodine in the sample line.
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