ML20137M481

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Rev 1 to Procedure 5.7.5, Estimating Core Damage
ML20137M481
Person / Time
Site: Pilgrim
Issue date: 08/21/1985
From: Seeny J
BOSTON EDISON CO.
To:
Shared Package
ML20137M477 List:
References
5.7.5, NUDOCS 8509130248
Download: ML20137M481 (82)
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{{#Wiki_filter:'e BOSION EDISON NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure No. 5.7.5 ESTIMATING CORE DAMAGE List of Effective Pages 5. 7. 5-1 5.7.5-2 5.7.5-3 Attachments 5.7.5A-1

5. 7. 5 B-1
5. 7. 5 C-12 5.7.5C-29 5.7.50-6 5.7.5A-2 5.7.58-2 5.7.5C-13 5.7.5C-30 5.7.5D-7 5.7.5A-3 5.7.5B-3 5.7.5C-14 5.7.5C-31 5.7.50-8
5. 7. 5 A-4 5.7.58-4 5.7.5C-15 5.7.5C-32 5.7.50-9 5.7.5A-5 5.7.5B-5 5.7.50-16 5.7.5C-33
5. 7. 5 D-10 5.7.5A-6
5. 7. 5C-1 5.7.5C-17 5.7.5C-34 5.7.50-11 5.7.5A-7 5.7.5C-1 5.7.5C-18 5.7.5C-35 5.7.50-12 5.7.5A-8 5.7.5C-2 5.7.5C-19 5.7.5C-36 5.7.50-13 5.7.5A-9 5.7.5C-3 5.7.5C-20 5.7.5C-37 5.7.50-14
5. 7. 5 A-10 5.7.5C-4 5.7.5C-21 5.7.5C-38 5.7.50-15
5. 7. 5 A-11 5.7.5C-5 5.7.5C-22 5.7.5C-39 5.7.50-16
5. 7. 5 A-12 5.7.5C-6

'5.7.5C-23 5.7.5C-40 5.7.50-17

5. 7. 5 A-13 5.7.5C-7 5.7.5C-24 5.7.50-1

[/)

5. 7. 5 A-14 5.7.5C-8 5.7.5C-25 5.7.50-2 5.7.5A-15 5.7.5C-9 5.7.5C-26 5.7.50-3 Approved L/0RC Chairhan
5. 7. 5 A-16
5. 7. 5 C-10 5.7.5C-27 5.7.50-4
5. 7. 5 A-17
5. 7. EC-11 5.7.5C-28 5.7.50-5

/ 75~ l 8509130248 850909 yDR ADOCK O 293

5. 7. 5-1 Rev. 1 l

1 's ' 1. PURPOSE The purpose of this procedure is to provide a mechanism to determine the degree of Reactor Core Damage from samples collected using the Post Accident Sampling System (PASS) and other available plant parameters. II. DISCUSSION Two procedures have been developed by G.E. which provide the user with a rapid mechanism to preliminarily identify what (if any) core damage has occurred during and after a suspected degraded core event. This procedure (Attachment C) has been developed to predict the degree of core damage based on fission product concentrations in either the water or gas samples. These concentrations are determined from samples collected by the PASS System (Reference Procedures NO. 5.7.4.1.0 through 5.7.4.14) and analyzed in the Radio Chem. Lab. as described in Procedure No 's 7.11.6 through 7.11.9. The second procedure (Attachment D) supplements the above by providing the user with a mechanism to verify and refine the initial estimates of core damage. This verification is based on correlations established between the radioisotopic information and other significant plant parameters available during and after the event. III. REFERENCES A. Emergency Procedures 5.7.4.1, 5.7.4.1.1 through 5.7.4.1.14 B. Chemical Analytical Procedures 7.11.6 through 7.11.9 C. Operating Procedures 2.2.133 0. BWR Owners Group Letter BWROG-8324 dated June 17, 1983 IV. PREREQUISITES A. Samples collected and analyzed as directed by the Watch Engineer. V. APPARATUS None VI. PRECAUTIONS None l VII. LIMITATIONS None l t 5.7.5-2 Rev. 1 me

~ VII

I. PROCEDURE

A. Use data.from the Drywell and Torus Hydrogen Monitors (panels C-174 or C-175) and/or the Containment High Radiation Monitors (panels C-170 and C-171 to determine the extent of the core damage. This data is to be used with the instructions from Attachment B.. Perform this step prior to collecting PASS samples in determining ALARA considerations. B. Obtain PASS system samples (Proc. 5.7.4.1.x series) - It is recommended that both the water and gas phase samples be taken and analyzed in order to reduce the uncertainty in core damage estimations. C. Using the information from the PASS samples assess the degree of core damage as described in Attachment A. s i r i % ~ 5.7.5-3 Rev. 1

ATTACHMENT A 1. Procedure to determine Core Damage based on samples collected from the Jet Pump, RHR and Torus Sample Points. A. Set CwI-131 and CwCS-137 as t'he measured concentrations (in uCi/g) of I-131 and CS-137 from the Jet Pump Sample Points. Cw1 131 = CwCS-137 = B. Correct the measured concentrations of the above radionuclides to the time of reactor shutdown as follows: JP JP 1. Let; Cwil-131 and CwiCS-137 Concentration of the radio- = nuclide of interest at the time of reactor shutdown, t= Time period (hrs.) between reactor shutdown and the deter-i mination of the-radioactivity of the radionuclide of interest JP 2. Find twil-131 where JP Cwi -131 = CW -131 x e3.5914 E-03 x t l I Cwi -131 " l JP 3. Find CwiCS-137 where JP CwiCS-137 = CwCS-137 x e2.6221 E-06 x t ,v JP CwiCS-137 = 5.7.5A-1 Rev. 1 f .-m, .---r.,

l C. Set Cwl-131 and CwCS-137 as the measured. concentrations (in uCi/g) of I-131 and CS-137 f rom the RHR Sample Points. T Cwl-131 = CwCS-137 = I D. Correct the measured concentrations of the above radionuclides to the time of reactor shutdown as follows: RHR RHR Concentration of the radio-1. Let; Cwil-131 and CwiCS-137 = nuclide of interest at the time of reactor shutdown. t= Time period (hrs.) between reactor shutdown and the deter-mination of the radioactivity of the radionuclide of interest RHR 2. Fird Cwil-131 where RHR Cwi -131 = CW -131 x e3.5914 E-03 x t l I w RHR Cwil-131 = RHR 3. Find CwiCS-137 where - J t RHR -CwiCS-137 - CwCS-137 x e2.6221 E-06 x t RHR CwiCS-137 = s 5.7.5A-2 Rev. 1 M O e4

.E. Set CwI-131 and CwCS-137 as the measured concentrations (in uCi/g) of I-131 and 'CS-137 f rom the Torus Sample Points. Cwl-131 = CwCS-137 = F. Correct the measured concentrations of the above radionuclides to the time of reactor shutdown as follows: T T 1. Let; Cwil-131 and CwiCS-137 Concentration of the radio- = nuclide of interest at the time of reactor shutdown. t= Time period (hrs.) between reactor shutdown and the deter-mination of the radioactivity of the radionuclide of interest T-2. Find Cwil-131 where i T Cwi -131 = CW -131 x e3.5914 E-03 x t l I T Cwil-131 = T 3. Find CwiCS-137 where T CwiCS-137 = CwCS-137_ x e2.6221 E-06 x t T CwiCS-137 = 5.7.5A-3 Rev. 1 l

G. Determine average fission product concentration .of the nuclide of interest in the total coolant mass as follows: 1. Let Cwia: Average fission product concentration in the plant coolant mass for the isotope of interest. a. Find CwiaI-131 as follows: Cwia = Cwi -131(Step B. or D.) x 2.05 E u8 + Cwi -131(Step F) x 2.38 E 09 l-131 2.565 E 09 4 CwiaI-131 " I b. Find CwiaCS-137 as follows: Cwia = CwiCS-137(Step B. or D.) x 2.05 E 08 + CwiCS-137(Step F) x 2.38 E 09 CS-137 2.585 E 09 CwiaCS-137 " 5.7.5A-4 Rev. I t

l j H. Calculate the fission product inventory factor n follows: 1. Let FIi -131 and FliCS-137 Fission Product inventory = l correction factor for the radionuclide of interest at the time of reactor shutdown. Pi Steady state reactor power level = operated during period number j in MWt - Note the variation in power level should be limited to 120% to meet steady state conditions. j NUMBER of operating periods at a = steady state power level since startup from the last refueling outage to the last shutdown prior to sampling. Tj Duration of operation 9 steady = state power for the operating period j in HOURS. Tj Time between the end of the = "jth".operatir:g period and the last reactor shutdown prior to sampling in HOURS. Find FIi -131 where FIi -131 2. l l 3.651 E 03 j p) g _,-3.5914 E-03 x Tj -3.5914 E-03 x TJ) FIi -131 " l 3. Find FliCS-137 where FliCS-137

  • 2.431 E 02 j) p) g _,-2.6221 E-06 x Tj

-2.6221 E-06 x TJ) FliCS-137 " 5.7.5A-5 Rev. 1

1. Calculate the normalized concentration of the radioisotope of interest as follows: } REF REF t 1. Let Cwi -131 & CwiCS-137 Normalized concentration of the l = radioisotope of interest at the time of reactor shutdown. 'REF REF Cwi -131 2. Find Cwi -131 where l l = Cwial-131 x Fli -131 x 6.594 E-01 l 1 I REF Cwil-131 - REF REF 3. Find CwiCS-137 where CwiCS-137 = CwiaCS-137 x FliCS-137 x 6.594 E-01 i REF-CwiCS-137 = J. Refer to Figures 1 and 2 to estimate the extent of ' fuel or cladding damages. t J 1 1 5.7.5A-6~ Rev. 1 -w-- a w - - - ~, - - ,,.w--, --w w -~ -o--. e

II. Procedure to determine Core Damage based on samples collected from the Torus and Drywell Atmosphere Sample Points. A. Set CgXe-133 and CgKr-85 as the measured concentrations (in uti/g) of Xe-133 and Kr-85 f rom the Drywell Sample Points. CgXe-133 CgKr-85 = 8. Correct the measured concentrations of the above radionuclides to the time of reactor shutdown as follows: D D 1. Let; CgiXe-133 and CgiKr-85 Concentration of the radio- = nuclide of interest at the time of reactor shutdown. t= Time period (hrs.) between reactor shutdown and the deter-mination of the radioactivity of the radionuclide of interest D i 2. Find CgiXe-133 where D CgiXe-133 = CgXe-133 x e5.5 E-03 x t D CgiXe-133 = t D 3. Find CgiKr-85 where CgiKr CgKr-85 x e7.3796 E-06 x t D l CgiKr-85 = 5.7.5A-7 Rev. 1 i

i C. Correct the measured atmospheric samples for temperature and pressure difference between the sample vial and Drywell. D D 1. Let CgicKr-85 and CgicXe-133 Concentration of the radionuclides of interest 9 the Drywell atmospheric temperature and pressure 9 the time of reactor shutdown. Tj Temperature of the gas in the = sample vial.

  • F P1 Pressure of the gas in the

= sample vial. PSIA T2 Temperature of the gas in the = containment. 'F P2 Pressure of the gas in the = containment. PSIA D 2. Find CgicKr-85 where: D D CgicKr-85 = CgiKr-35 P T) 2 P)T2 D CgicKr-85 = D-3. Find CgicXe-133 where: D D CgicXe-133 = CgiXe-133 P T) p 12 0 CgicXe-133 = L 5.7.5A-8 Rev. 1 i

D. Set CgXe-133 and C9Kr-85 as the measured concentrations (in uti/g) of Xe-133 -and Kr-85 f rom the Torus Sample Point. CgXe-133 -CgKr-85 = E. Correct the measured concentrations of the above radionuclides to the time of reactor shutdown as follows: T T Concentration of the radio-1. Let;~CgXe-133 and CgKr-85 nuclide of interest at the time of reactor shutdown. t= Time period (hrs.) between reactor shutdown and the deter. mination of the radioactivity of the radionuclide of interest T 2. Find CgiXe-133 where T T CgiXe-133 = CgXe-133 x e5.5 E-03 x t T CgiXe-133 " T 3. Find CgiKr-85 where T T CgiKr-85 = CgKr-85 x e7.3796 E-06 x t T CgiKr-85 = '5.7.5A-9 Rev. 1 a h

F. Correct the measured atmospheric samples for temperature and pressure difference between the sample vial and Torus. T .T 1. Let CgicKr-85 and CgicXe-133 Concentration of the = radionuclides of interest e the Torus atmospheric temperature - and pressure 9 the time of reactor shutdown. Ti Temperature of the gas = in the sample vial.

  • F P1 Pressure of the gas in

= the sample vial. PSIA T2 Temperature of the gas = in the containment. 'F l P2 Pressure of the gas in = the containment. PSIA T 2. Find CgicKr-85 where: l T T CgicKr-85 = CgiKr-85 P T) 2 '12 T T CgicKr-85 = T 3. Find CgicXe-133 where: T T .CgicXe-133 = CgiXe-133 P T) p '12 T T CgicXe-133 =

5. 7. 5 A-10 Rev.1 L

G. Determine average fission product concentration of the nuclide of interest in the total contain-ment volume as follows: 1. Let Cgia: Average fission product concentration in the containment volume for the isotope of interest. a. Find CgiaXe-133 as follows-CgiaXe-133" I ' e-133* I 'Xe-133 * + 7.34 E 09 CgiaXe-133 " b. Find CgiaKr-85 as follows: CgiaKr-85 " 9 ' r-85

  • 8 'Kr-85
  • 7.34 E 09 CgiaKr-85 "

5.7.5A-11 Rev.1 i' r L.

c H. Calculate the fission product inventory factor as follows: 1. Let FliXe-133 and FliKr-85 Fission Product inventory = correction factor for the radionuclide of interest at the time of reactor shutdown. Pi Steady state reactor power level = operated during period number j in MWt - Note the variation in power level should be limited to i 120% to meet steady state conditions. j NUMBER of operating periods at a = steady state power level since startup from the last refueling outage to the last shutdown prior to sampling. Tj Duration of operation 9 steady = state power for the operating period j in HOURS. Tj Time between the end of the = "jth" operating period and the last reactor shutdown prior to sampling in HOURS. Find Fli r-85 where Fli r-85 " 2. K K 6.436 E 02 j) p) $ _,-7.3796 E-06 x Tj),-7.3796 E-06 x'T] Fli r-85 = K Find Fli e-133 where Fli e-133 3. X X 1 3.651 E-03 j) - p g _,- 5.5 E-03 x Tj ),- 5.5 - E-03 x Tj) FIixe-133

  • I I

5.7.5A-12 Rev. 1

I. Calculate the normalized concentration of the radioisotope of interest as follows: REF REF 1. Let CgiKr-85 & CgiXe-133 Normalized concentration of the = radioisotope of interest at the time of reactor shutdown. REF REF 2. Find CgiKr-85 where CgiKr-85 = CgiaKr-85 x Fli r-85 x 1.835 E-01 K REF CgiKr-85 = ^ REF REF 3. Find CgiXe-133 where CgiXe-133 = CgiaXe-133 x FIi e-133 x 1.835 E-01 X f REF CgiXe-133 = J. Refer to Figures 3 and 4 to estimate the extent of fuel or cladding damages. 5.7. 5 A-13 Rev. 1 1 )

e, 5 1: FUE L WE LT*Aws i UPPEm RELE ASE LtuiT SEST ESTIMATE / / \\ lower RELE ASE LIMIT s o*- / / 1 / / I } 1~ / / / / \\ / / / l / / i / / / i b / / / p',/ 'C' / k / / / / / 5 / / / c / / / E / / / / / / y r sc / f h / ~ E / 5 / / / / / CL ADO'NG E AILus t w / to \\ p p UE R R E LE ASE LiWIT [ SEST ESTouaTE / L0nta mE LE ASE 6 Mit / / / / I / N ostW A L lauTDo*N / CONCENTRATi0% / ME ACToa navge / UP'ER LIMIT 20 0.C.,g / NOWIN A L o 7.C.<s / i /

,,1

,,,,I ,..,,1 ,,,,,,,11 l o, oi 1.o to too r e i.u: ( )= s cL.*coi%c 5 A LumE l l l-i.o to too W s s uE L ut LToows 1l Figure 1. Relationship Between 1-131 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant. 5.7. 5 A-14 Rev. 1 '-

1:8, FUE L Ff f LTDow% WetEm RELE ASE Liw T sEST ESTtWATE / / ~ LCwvEm mE LE ASE LIMIT / ic? P / / ~ / / / / / / / p / / / / / l / / 1 / / / I 8 / / / 2 / / / 'C' ~ y / / / ~ 4 / / / E ~ / / / ~ C l / / / ~ l / ~ c / / / / c / l / z / l to I / / / / / ~ / / ~ / CL AD0sNO 8 Aituut z O f t C / WP*E m m E LE A5E Ltw.* y f f C 'C f / BEST ESTiwATE g f / L0nta *E LE A5E Livi* C. ,/ ~ / / / / N0mW AL SNTDOWN o, / CONCENTmATioN sN REAC'Os waTE* / / U** E m Ltu t' C3 .C. s / NQwsNA. 0 03.C. g / '/ / ,,,,,1 ic-2 / ei io ic too - % CL ADDING 8 At' um[ f 10 to 10C 'g %fufLwELTDoeN r 'l Figure 2. Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant. l 5.7. 5 A-15 Rev. 1 1 1 i 4

1C*, F ut t tf E LTDow % ~ UPPEm mELE ASE LIMLT j Jest ESTIM ATE / p / ~ / LonEm mELE A3E LIMIT / 3 10 / ~ / // ~ ~ // / ~ // / // / ~ \\ // / 1 // / / 2 1C /,// O 2 / ff / / / / / / 5 / l / / / ~ E / l h / / te / E / } [ / ~ C / / ~ p / CLAODING F AILust ~ f / / UPPEm mE LE ASE LtWIT l y y c / 1O' / BEST ESTlwATE j E / LOwEn RELE ASE Ltu:T f -( / ~ ~ / / / / / monw Ag opEm ATsNG ~ C1 / CONCE%TmATio% f th om vwg LL uppEm LiuiT 10 '.C.tcs ~ l howi=AL t o-5,c., c ,c-2 ,,,,,,1 ,,,,,,1 ,,,,,,,,1 Ot tc 10 100 L-s CLADDING F AttumE

II i.o to 100 rl

'g s t uf L ut LTocas Figure 3. Relationship Between Xe-133 Concentration in the Containment Gas (Drywell + Torus Gas) and the Extent of Core Damage in Reference Plant. 5.7.5A-16 Rev. 1 l l 1 -

1C' p ut L WE s700=% UNEm mE LE ASE LluiT o, SE5T ESTlW ATE LOWER RE LE ASE LlwiT j f 10 / ~ 5 / / / / / / f / 5 / / ,c - / / / / / / l / 5 / / h / / / f ~ f / / g ic ' / / v E i / / i / / / / / / CLAD 0iNO 8 Attunt f / c b / / UP'Em mE LE A5E Liuti 1C'I / / k [ / / SE5T EST8W ATE ~/ / / LOWEa mELE ASE LiutT / / ~ / 1C*3 / NO*W AL 0*E a ATION f CONCE NTR ATION 7 th DAYWE LL .j 4. t o-S.C.,cc / . WPPEm LiWiT NOWi%AL 4 a 10'. C. < st it,erl , e e el tt,t*! d ,ei,,1 e i e t t t t t 1 t i 1 ee... ge 01 1C to too fc % CLADDING 8 AikumE -l t.o to too 'g fc % F UE L WE LTDOWN Relationship Between Kr-85 Concentration in the Containment Gas Figure 4. (Drywell + Torus Gas) and the Extent of Core Damage in Refe,rence Plant. 5.7.5 A-17, Rev.1 l l I l l l l ( 1

Beco #26 (PR02-1) ATTACHMENT B I. Procedure to determine the degree of claddirig damage from analysis of the torus atmosphere H2 monitor (in panel C-174 or C-175 indicators AIT-2A-5082A or AIT-2A-50828 respectively) readings. A. Set H0 as the measured concentration (%) of hydrogen in the torus as indicated ont he torus atmosphere monitors (panel C-174 or C-175) HD" B. From Figure 1 determine the metal water reaction (%) for the reference

plant, j

MWREF = C. Find the amount of cladding that has reacted (therefore is damaged) to j form hydrogen as follows: 1. Let %MW = The percentage of the cladding that has reacted to form hydrogen. 2. Find % fed where %MW = MWREF x 6.3845 E-01 5MW = l f i f i l

5. 7. 5 B-1 Rev. I l

4 l l

B:co #26 (PR02-1) II. Procedure to determine the degree of cledding damage from analysis of the drywell atmosphere H2 monitor (in panel C-174 or C-175 indicators AIT-2A-5082A or AIT-2A-50828 respectively) readings. A. Set HD as the measured concentration (%) of hydrogen in the torus as indicated ont he torus atmosphere monitors (panel C-174 or C-175) HD" B. From Figure i determine the metal water reaction (%) for the reference plant. MWREF " C. Find the amount of cladding that has reacted (therefore is damaged) to form hydrogen as follows: 1. Let %MW = The percentage of the cladding that has reacted to form hydrogen. 2. Find SMW where %MW = MWREF x 6.3845 E-01 %MW = 4 5.7.58-2 Rev. 1 i 1

III. Procedure to determine the degree of Fuel Inventory released from analysis of the Containment High Range Monitoring System (CHRMS). NOTE A Drywell CHRM and a Torus CHRM should be used together to determine the degree of Fuel Inventory released. A. Obtain the associated CHRM reading. Ch. A / D

Drywell, RD=

/ R/hr.

Torus, RT =

/ R/hr. B. Record the time, in hours, after reactor scram that readings were taken. t= hrs C. From Figure 2 determine the following: 1 DRYWELL TORUS CH. A / 8 CH. A / B 1. Fuel Failure Type: / / 2. 5 of Noble Gas: / -/ 3. % of Halogens: / / 4. % of Solids: / / 4 no us 5.7.58-3 Rev. 1 i ) l

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  • a x s

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h p e-c L I n I, .n s. 5 -.t. + W ATTACHMENT 1 v' E: y ? 4 w.. I e PROCEDURES JOR THE DETERMINATION OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS t i e 1 3

f. '

s a q -.b .t -g ' NM -'u. .. Lr --..--..j Ma* ' .*,.== -j. e ..e.. ,* j. A'A % I. ,t *' (3)iksel Fr3 e. u., "... *.

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= . y~ r ~ N< ,y .n = z u.- b

NEDO8 83NEl CU AUGUST J ~ ~ i l iL;s z.x m..= # m w m e <= wmme.xmmmere i PROCEDURES FOR THE DETERMINATION OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS 1 C. C. LIN l_2._. A:,~2L 54% h* l~;Y'% $%3ML :i 4 ^"G.TUIMES.Li':iUY L T*'E TIE l i l 't. ~ GENERAlh ELECT 5.7.5c-2 Rev. l h

NEDo-22215 82NED090 Class I August 1982 k PROCEDURES FOR TE DETERMINATION OF TE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS i ~ Chien C. Lin ( Approved: D R. L. Cowan, Manager Plant Chemical and Radiation Technology-Approved: E. Kiss, Manager Plant Technology i NUCLEAR P0kTR SYSTEM ENGINEERING DEPARTMENT, GENERAL ELECTRIC COMPANY VALLECITOS NUCLEAR CENTER, PLEASANTON, CA 94566 GENERAL h ELECTRIC 7 .s 5.7.5C-3 Rev. 1

t NEDO-22215 i i DISCLAIMER OF RESPONSIBILITY Neither the This doc:r:ent was prepared by or for the General Electric Company. General Electric Company nor any of the contributors to this doc:ount: M:kesanyvarrantyorrepresentation,expressorimplied,withrespectto) A. the accuracy, corpinteness, or usefulness of the infom:: tion contained in his doccunt, or that the use of any infomation disc 1osed in this docu-ment my not infringe privately owned rights; or Ass:c:es any responsibility for liability or da~: age of any kind which my E. result from the use cf any infomation disclosed in this doc: cunt. i I l l. i L-2 ( 11 5.7.5c-4 Rev. 1 l

NEDO-22215 i COSTENTS i P, a ge 1. OBJECTIVE AND SCOPE 1-1 j, 2. PROCEDURES FOR DETERMINATIONS OF CORE DAMAGE 2-1 2.1 Reference Plant (BWR-6/238. Mark III) 2-1 i 2.1.1 Reference Plant Parameters 2-1 2.1.2 Procedura 2-1 2-1 2.1.3 Supplementary Data 2-8 2.2 Specific Plant Application 2.2.1 Plant Parameters 2-8 2-11 2.2.2 Procedure 3-1 3. TECHNICAL BASIS 3.1 Tission Product Concentrations in the Primary System During Reactor Shutdown Under Normal Operation Conditions 3-1 ) i 3.1.1 Tission Product Concentrations in Reactor Water 3-1 3.1.2" Noble Gas Concentrations in Drywell and Torus 3-1 Cas Phase I 3.2 Tission Product Release Source Terms Under Accident 3-2 Conditions 3-2 1 3.3 Isotopic Distribution in Tuel Cap 3.4 Anticipated Chemical Behavior of Iodine and Coolant 3-3 Chemistry Under Accident Conditions 4-1 4. DISCUSSION AND SUGGESTIONS FOR TUTURE WORK 5-1 I 5. RITERENCES I APPENDICES i A. SAMPLE CALCULATION OF TISSION PRODUCT INVENTORY CORRICTION A-1 TACTOR B-1 i 3. TOTAL I-131 RILEASE DURING A SPIKING SEQUENCE C. ESTIMATION OT MAXIMUM 10 DINE-131 SPIKE CONCINTRATION IN C-1 i REACTOR WATER I i 1 i iit/iv 5.7.5C-5 Rev. 1

NEDo-22215 TABLIS Title Pjuge Table 1 Core Inventory of Major Tission Productssin a Reference 2-2 Plant operated at 3651 MWt for Three Years. Tission Product Concentrations in Reactor Water and Drywell 2 2-3 Gas Space During Reactor Shutdown Under Normal Conditions 2-9 3 Ratios of Isotopes in Core Inventory and Tuel Cap-2-10 J 4 Plant Parameters 2-15 i 5 Best-Estimate Tission Product Release Tractions ?. r' k ILLUSTRATIONS Title P_gee Ti __ sure i 1 Relatienship Between 1-131 Concentration in the Primary 1 Coolant (Reactor Water + Pool Water) and the Extent of 2-4 Core Damage in Reference Plant 2 Relationship Between Cs-137 Concentration in the Primary i Coolant (Reactor Water + Pool Water) and the Extent of Core 2-5 Damage in Reference Plant 3 Relationship Between Xe-133 Concentration in the Containment Gas (Drywell + Torus Gas) and the Extent of Core Damage in 2-6 1 Reference Plant 4 Relationship Between Kr-85 Concentration in the Containment Gas (Drywell + Torus Gas) and the Extent of Core Damage in 2-7 i Reference Plant i l 4 i l .v/vi 5.7.5C-6 Rev. 1 T a e e v c----,-~---- -m , ~, - ~, - - -,,----n--- -e

NEDD-22215 s 1. OBJECTIVE AND SCOPE The purpose of this procedure is to determine the degree of reactor core da= age from the measured fission product concentrations in either the water or gas sa=ples taken from the primary system under accident conditions. The pro-cedure involves calculations of fission product inventories in the core and the release of inventories into the primary system under postulated loss-of-coolant accident (LOCA) conditions. The fuel gap fission products are assumed to be released upon the rupture of fuel cladding. The majority of fission inventories in the fuel rods would be released when the fuel is melted product at higher temperatures. ABWR-6/238withaMarkIIIcontainmentisused~as}a reference plant in the demonstration of this procedure. Application of the procedure for any other type or size of boiling water reactor (BWR) is described. a 4 1-1/1-2 5.7.5C-7 Rev. 1 i .. ~..

i NEDO-22215 2. PROCEDL*RES FOR DETERMINATIONS OF CORI DAF. AGE .2.1 RETERINCE P1 NT (BkT-6/238 MARK III) t 2.1.1 Reference Plant Parameters i I The pertinent plant parameters for the re'erence plant are given below: l Rated reactor thermal power 3579 MWt Wumber of fue1 bundles 748 bundles ~ Total primary coolant mass (reactor 9 water plus suppression pool water) 3.92x10g} w s. Total containment and drywell gas space 10 volume 4.0 x 10 cc The fission product inventories in the core are calculated based on three ' years (1095 days) of continuous operation at 3651 MWt, or 102% of rated f l power, by using a computer code developed at Los Alamos and adapted to the GE computer system. The inventories of some major fission products in the core at the time of reactor shutdown ard' given in Table 1. 2.1.2. Procedure Either3the gas or water samples taken from the post accident sampling systa= are analyzed for major fission product concentrations by gamma ray i spectrometry. If the concentration of a fission product in reactor water i or drywell, corrected the decay to the time of reactor shutdown, is measured to be higher than the baseline concentration shown in Table 2 (see See-j tion 3.1 for details), the extent of fuel or cladding damage can be deter-mined directly from Tigures 1 through 4 based on isotopes 1-131. Cs-137, Xe-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. w 2-1

5. 7. 5 C-8.

Rev. 1 l

1 NEDO-22215 Table 1 CORE INVENTORY OT MAJOR FISSION PRODUCTS IN A RETERENCE Pl. ANT OPERATED AT 3651 MWt FOR THREE YEARS Major Gamma Ray Energy Inventory ** (Intensity) Chemical Group Isotope

  • Half-Life 106 Ci kev (y/d)

Noble gases Kr-85m 4.48h 24.6 151(0.753) Kr-85 10.72y 1.1 514(0.0044) Kr-87 76.3m 47.1 403(0.495) Kr-88 2.84h 66.8 196(0.26).1530(0.109) Xe-133 5.25d 202.0 81(0.365) Xe-135 9.11h 26.1 250(0.899) Halogens I-131 8.04d 96.0 364(0.812) I-132 2.3h 140 668(0.99.773(0.762) I I.133 20.8h. 201 530(0.86) I-134 52.6m 221 847(0.354) 884(0.653) 1-135 6.61h 189 1132(0.225).1260(0.286) J Alkali Metals Cs-134 2.06y 19.6 605(0.98),795(0.85) ~ Cs-137 30.17y 12.1 662(0.85) Cs-138 32.2m 178.0 463(0.307).1436(0.76) Telluriu= Group Te-132 78.2h 138 228(0.88) Noble Metals Mo-99 66.02h 183 740(0.128) Ru-103 39.4d 155 497(0.89) Alkaline Earths Sr-91 9.Sh 115 750(0.23),1024(0.325) St-92 2.71h 123 1388(0.9) Ba-140 12.8d 173 537(0.254) Rare Earths Y-92 3.54h 124 934(0.139) La-140 40.2h 184 487(0.455).1597(0.955) 1 Ce-141 32.5d 161 145(0.48) Ce-144 284.3d 129 134(0.108 Refractories 2r-95 64.0d 161 724(0.437),757(0.553) 2r-97 16.9h 166 743(0.928)

  • 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.

J 1 2-2 5.7.5C-9 Rev. 1

NIDo-22215 Table 2 FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER AND DRYWEL1, GAS SPACE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS Reactor Water, vCi/g Drywell Gas (uci/cc) Isotope Upper Limit Nominal Upper Limit Nominal I-131 29 0.7 b Cs-137" 0.3* 0.03 -5b 10 ' 10 ~ Xe-133 -5a -6b 4x10 4x10 Kr-85 \\ " Observed experimentally.'in an operating BWR-3 with HK I containment, data obtained from GE unpublished document. DRT 268-DEV-0009. bAssu=ing 10% of the upper Itait values. " Release of Cs-137 activity would strongly depend on the core inventory which is a function of fuel burnup. ] l i l 2-3 5.I.5c-10 Rev. 'l f

NEDO-22215 5 1c FUE L MELTDOWN UPPER RELE ASE LIMIT BEST ESTIMATE / / LOWER RELE ASE LIMIT l 10", l / / / ~ / / / ~ i / / / p' '/ ~ / / / p 2 / / / S' to / 3 / / / / / / } E / / / "g / / / 5* / / / E / / / / 4 1 / / / f 'O 1 E l / ~ l / ~ i y / ~ G / / / l / CLADDI AG F AILumE / u / UPPER RELE ASE LlutT T.; io :- / / ~ k / SEST ESTIMATE / LOWER RELEASE LIMIT / / / / / NonMALSMUTDowN 1.o 7 CONCENTRATION 7 IN mE ACTO # W ATgm p UPPER LIMIT 20 0 mCits / NOMINAL o 7 eCars i 7 / \\ / ......i , s,,,i ,,,,,,,i ,,,,,,,i l ot 1.0 to 100 % CLADDING F ALLURE - 9, g s.o to 100

  • g

% F UE L ME LTD0nN l i Relationship Between 1-131 Concentration in the Primary Coolant Tigure 1. (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant 5.7.5C-11 Rev. 1 2-4

EEDO-22215 1 l l so". F UE L ME LTDOWN [ UPPER RELE ASE LIMIT BEST ESTIMATE / / ~ LOWER RELE ASE LIMIT / to' / / / / / / / / / p / / / / / / / / y / / / / / /

  • t o,

/ / / / / / k ~ / / / s / / / } l / / / ~ / / s / / i / / / / / I to p / p / g 9 / / 3 / / / x / CLADoiNG S AttumE j j UPPER RELE ASE LIMIT u 7 f a t.o ( / BEST ESTIMATE f l / LOWER RELE ASE LIMIT / / / / / / NORM AL SMUTDowN l c1 / CONCENTRATION = IN RE AcTOm mATEm / / UPPER LIMIT o 3 eCeig / NOMihAL c o3.C.,e / / l . / iiinnt i ...it l io-2 ,,,,ii ,.....I i i i i>>ini i f i, i e o1 1.o to too l I I % CLADDING S ALLURE r 'g i.o to too 'g % FutL MELTDOWN Tigure 2. Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in l Reference Plant 2-5 5.7.5c-12 Rev. 1

1

._ -. ~, ,.. _.. _ _ _., _ _ _ _ _ _.., ~.. _,. - -, ___,, __

NEDO-22215 10 FUE L MELTDOWN ~ UPPER RELE AsE LlulT j BEST ESTIM ATE / / / / LOWER RELE ASE LIMIT / 12 / / / ~ // / // / // / / / 5 / / / / ,c' / / / / 0 / / / / / k e / x / / l / / p / 3 / 8 / l / 10 / E / l ~ l / / / / / CL ADDING F AILumE / ~ / w / UPPER RELEASE Llutt / j/ u / SEST ESTIMATE 10 2 / ( / LOWER RELE ASE LIMIT I / ~ / / ~ / / / NORM AL OPE R ATING oJ,- / CONCENTRATION f ~I/ IN DavwtLL l UPPER LIMIT. to # C.tec NOMIN AL. 10 S eC./cc I io-2 ,,,,,,,1 ,,,,,,,1 ',,,,,,,1 ,,,,,,i ,0.1 1.0 to too % CLADDING F AILumE r 1.o to 100 -l fc % FUEL ME LTDOWN Tigure 3. Relationship,Between Xe-133 Concentration in the Containment Gas (Dryveil',+ ToYus~. Gas) 'and the Extent' of, Core Da= age in Reference i Plant,"'"i' 2-6 5.7.5C-13 Rev. 1 i -.~.-

NEDo-22215 7 10 - FUE L ME LTDOWN UPPER RELE ASE LIMIT BESY ESTIM ATE LOWER RELE ASE Leutt f / 10 : - / / / / / / / 3 / O / j 3 10 ~~ E 5 / / 1 / / ? / / ~ 5 l / h j / / 2 / / / l 10-' / -/ u E i / / / / l / / E / CLAD 0tNG F AILumE / / UPPER RELE ASE LIMtf u go-2 / / f [ / / BEST ESTIM ATE u =l f LOWER RELE ASE LIMIT / / / / / ) in-3 / NORM AL OPER ATION d. / CONCE NTR ATION 5 l 7 IN DRYWELL

  • /

UPPER LlulT 4 a 10-5 Cacc i ( NOMINAL 4 a 10'. Case j l ~ i A 'I 10 O.1 1.0 10 100 % CLAODING F ALLURE ~l lc 1.0 10 100 -l 'c % FUEL MELTDOWN g Tigure 4. Relationship Between Kr-85 Concentratien in the Containment cas (Dryvell + Terus Gas) and the Extent of Core Da= age in Reference Plant 2-7 5.7.5C-14 Rev.1

NEDo-22215 If the concentration f alls into the range where release of the fission product from the fuel gap or the molten fuel cannot be definitively deter-mined, additional data may be needed to determine the source of fission product release (see below). It is recommended that both the water and gas phase samples be measured in order to reduce the uncertainty in core damage estimations. 2.1.3 Supplementarv Data in addition to the longer-lived isotopes, some shorter-lived isotope concantrations may be measured in the sample. Theratiosofisotopesreleased) fro = either the fuel gap or the molten fuel are significantly different as J shown in Table 3 (see Section 3.3 for details), thus the source (fuel or gap) of release may be identified. Furthermore, some less volatile elements in If the less the core may also start to release as the fuel starts to melt. volatile fission products, such as isotopes of Sr. Ba, La, and Ru (either soluble or insoluble), are found to have unusually high concentrations in the In a mixture of water sample, some degree of fuel melting may be inferred. fissien products 2.7h Sr-92 (1.384 Mev) and 40h La-140 (1.597 MeV) should be More work, relatively easy to identify and measure from a gamma ray spectrum. however, is needed to establish the baseline concentrations for those isotopes. i I 2.2 SPECITIC PLANT APPLICATION i 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. Si=ilar information is available for all 3'JRs. 5.7.5C-15 Rev.1 ~ 2-8 I l

NEro-22215 Table 3 RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP Activity Ratio

  • in Activity Ratio
  • in Tus1 Cap Isotope Half-Life Core Inventory 0.0234 0.233 Kr-87 76.3 m 0.0495 Kr-88 2.84h 0.33 0.023 Kr-85m 4.48h 0.122 1.0*

Xe-133 5.25d 1.0* 0.155 l 1-134 52.6 a 2.3 I 0.127 I-132 2.3 h 1.46 0.364 I-135 6.61h 1.97 0.685 I-133 20.8 h 2.09 1.0* 1-131 b.04d 1.0* 1 ( i noble gas isotope cencentration g,,,,31, g,,,, Xe-133 concentration Iodine isotope concentration for iodines 1-131 concentration y 2-9 5.7.5C-16 Rev. I

NEDo-22215 Table 4 2 Pl. ANT PARAMITERS Primary Coolant

  • Containment Gas
  • Reactor Reactor Drfwell Torus /

Type / Rated Water Suppression Gas Containment Containment Power Mass Pool Water volume cas Volume Plant Design (MWt) (108 g) (109 g) (109 cc) (209 ee) 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 I 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 Dresden-2/3 BWR 3/1 2527 2.61 3.18 4.48 3.30 Duane Arnold BWR 4/I 1593 1.45 1.67 2.67 2.67 Termi-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 k 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 3/1 2011 2.05 2.78 4.16 3.06 McE.ticello 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 2/1 1933 2.05 2.32 5.10 3.85 Peach Bottom-2/3 BWR 4/I 3293

2. 6,7 3.48 4.98 3.62 Pilgrim BWR 3/I 1998 2.05 2.38 4.16 3.18 Susquehanna-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 Total Containment Cas Volume = Drywell Gas + Torus (or Primary Containment in Mark III gas -10 5.7.5c-17 Rev. 1

i NEDO-22215 2.2.2 Procedure The extent of core damage in an operating B'n'R can be determined by comparing the measured concentrations of major fission products in either the gas or water samples, after appropriate normalization, with the reference plant data. The following procedure is reconnended. 1. Obtain the samples from the post accident sampling system, and the concentration of a fission product i (C in water or C in gas is yg g determined). the measured concentration for decay to the time of reactor I 2. Correct shutdown. 3. Correct the measured gaseous activity concentration for temperature and pressure dif ference in the sample vial and the containment (torus) gas phase (see f ootnote on Tage 2-12). e 1, 4. Calculate the fission product inventory correction factor (Section 2.2.2.2). 5. Calculate the plant parameter correction factor (Section 2.2.2.3). 6. By using the correction factors, calculate the normalized concen-tration, C,*g or C 'g (Section 2.2.2.1). 7. Use Tigures 1 through 4 to estimate the extent of fuel or cladding damage. 2.2.2.1 Comparison with Reference Plant Data 3 The extent of core damage can be estimated from the measured fission product concentrations in either the gas or water samples, as described for the reference plant. 'However, the measured concentration must be corrected 5.7.5c-18 Rev. I 2-11

NDo-22215 for the differences in operation power level, time of operation, primary coolant mass and containment gas volume. Ref i C =C e xT xT yg yg 7g y or Cf = C A e xT xT 3g where k C = concentration of isotope i in the reference plant coolant (VCi/g) Re y C 'g = concentration of isotope i in the reference plant containment gas (uCi/cc) C = =essured concentration of isotope i in the operating coolant at yg time, t (uCi/g) C = measured concentration of isotope i in the operating containment gas at time, t (uCi/cc)* I = decay correction to the time of reactor shutdown e A = decay constant of isotope 1 (day ~) g l

  • The following correction for the measured concentration is needed if the tem-perature and pressure in the sample vial (T.P ) are different from that in 3

3 the containment (T,P )* 2 2 i '2 g T gi gi(vial)

  • P T gy 2-12 5.7.5C-19 Rev.1 i

. ~.,, _

NEDO-22215 t = time between the reactor shutdown and the sample time (day) T = inventory correction factor for isotope 1 (see Section 2.2.2.2) g T = containment gas volume correction factor (see Section 2.2.2.3) T, = primary coolant mass correction f actor (see Section 2.2.2.3) 2.2.2.2 Inventory Correction Factor k , Inventory in reference plant 7li Inventory in operating plant e 3651 (1-e I) -).g )) e->g ) T T I j P) (1-e where P) = steady reactor power operated in period j (Mk't)* T) = duration of operating period j (day)* O T) = time between the end of operating period j and time of the last reactor shutdown (day) For a particular short-lived isotope, i, a calculation for only a period of % half-lives of reactor operation time before reactor shutdown should be accurate enough. An example of calculation is illustrated in Appendix A. It should be pointed out that the computer calculation of core inventory takes into account the fuel burnup, plutoniu= fission and neutron capture reactors.

  • In each period, the variation of steady power should be limited to :200.s, 2-13 5.7.5C-20 Rev. 1

NEDo-22215 The correction f actor calculated from this equation may not be entirely accurate. but the error is insignificant in comparison to the uncertainties in the fission product release fractions (Table 5) and other assumptions (Section 3.2). 2.2.2.3 71 ant Pacameter Cerrection Factors G operating plant coolant mass (g) . 1;5[F '9 7", reference plant coolant mass (3.92 x 10' g) J#'d' perating plant containment gas volume (ce) .))[# # ^' ,gp3r se 8 reference plant containment gas volume (4 x 1010,c) y,,o ) F = In case the fission product concentrations are measured separately for the reactor vster and suppression pool water or the drywell gas and the torus or C would be averaged from the gas, the measured concentrations Cyg g separate measurements: i C Reactor water mass + ts1 water = wi (e ne. in drywell)x(dryvell gas vol)+(cone. in torus)x(torus gas vol) gi dryvell gas volume + torus gas volume C = l 2-14 5.7.5C-21 Rev. 1 l

NEDO-22215

8. e.4 C

C C C 4 e e e e N e a I I W t== e C. E C. C. m. O. C. C. I e

c. =

G D 3, O C C C C C C C C C

== N N g b.M ' .C e e C I a. I E C. C. o. C. t = eJ 1 C C C C C C C w = e C C e C e e W = 8 k C

== = m 4 4 O. O.

==. m. C. I 5 O C C C C C C e >l z ww e O C e e i e N I I I b E C C 1 e I I I

== 8 I i 1 E

c. =.

E C.i2 C C C C 46 e C 5! e m e m I l 8 a b u E. M. I u 8

== m C o. 4 1 8 i m E C. I J ta E C C C C C g

g 3 :l I
  • l b

8'l ~ = e m N I I E C C I =j C C l e m I I = I I I I w l a A C i C C C C la: z,. W b k Mi C m e

== & =; N E C C C L E, e e E N N C C Ca' E @ ' D.'.':! O C C C C C C C e e n = u, e ~ e

==* e e et f

== N C C .c C u E. 4 C. C. C. C. C. 6 f'* 5 = "g j g g E 4 c' M tr* C C m f'* m C O u M C W C M E. E. N. C. C. C. C C C C C C C e 4 ,j A f la: l t== ; N O C 4 8 I I z 6 u, i 1 8 C. E.' N. m. C. 1 I e

== sa: D=1 C C C C C g,. I N e f e' bw C = c I I C 1 C I i 1 I = E 6 = =. 1 w I w I e E C. s a J,. C C C r* m e i L. g t I f I C l 1 nl C h I I c, m -I C. C. C. m I i O C C C al' e C-u n en i b q E 'O L W E C U

== u e c

  • b

.C z z = e gc w *= = *

  • b g

F. M b 4 h & C b' U f* um t m c e =

  • Cm U.C b

@m

==.c 5. L .c = fl f CE b gg et I.

== 5 3.% WW -E 4 *-{ b

  • CE M
  • M *

.as F. .c = .ar b t-t= b -U LF C ar -M f: >= = bN .c b CM j =- m-m - l.: 5-z- =. 2-15/2-16 5.7.5C-22 Rev.1

t NEDO-22215 s. 3. TECHNICA!. BASIS 4 FISSION PRODUCT CONCENTRATIONS IN THE PRIMARY SYSTEM DURING REACTOR 3.1 SHUTDOWN UNDER NORMA 1. OPERATION CONDITIONS 3.1.1 Tission Product Concentrations in Reactnr_ Water is well known that some volatile and water soluble fission products, It mainly iodine and cesium isotopes, will be released (called spiking) from Based on defect fuel rods when the reactor is shutdown and depressurized. Pasedag of NRC,3 the maximum I-131 release would be 10 Ci per each pCi/see release rate during normal power operation (see Appendix B). According to the I GE design basis of 1-131 release rate at 700 uC1/sec' a maximum of 7000 Ci of I-131 may be released during reactor shutdown, and the concentration in reac-ter water would be 29 uCi/g. An analytical model to predict the magnitude of I-131 spiking f ollowing The reactor shutdown in operating BWRs has been reported by Brutschy at al. "best estimate" concentration for I-131 has to be calculated based on the analytical mode for the individual reactor according to its fuel condition 5 (see Appendix C). However, if one adopts a standard 1-131 concentration of 6 5x10 uCi/s or s18 uti/sec) as proposed by ANS, the nominal I-131 spiking is -3 estimated to be %0.7 uCi/g in the reference plant water. This concentration is consistent with an average spiking concentration observed experimentally (see Appendix C Tigure C-1). The results of these estimations, including the Cs-137 concentration, have been sumnarized in Table 2. Potential future research in this area vill be discussed in Section 4. Noble Gas Concentrations in Drvvelk and Torus Gas Phase 3.1.2 Similar to the spiking magnitude, the noble gas activities in the drywell and the torus gas may vary significantly from reactor to reactor, mainly depending on the fuel condition and the steam leakage rate. In an operating BWR when the Xe-133 release rate measured at the stems jet air ejector (SJAE) l i i 5.7.5C-23 Rev. 1 3-1 1

? NEDO-22215 e. was 1.5 x 10' pC1/sec* (compared to design basis release rate at 8200 uci/sec), the neble gas concentrations in the dryvell were determined to be N10 ' uCi/cc ~ for Ke-133 and 44 x 10 uC1/cc for Kr-85. These data may be considered as ~ the upper limit values. 3.2 FISSION PRODUCT RELEASE SOURCE TERMS UNDER ACCIDENT CONDITIONS The source terms for the damaged core under accident conditions have been proposed by several investigators. ' The "best estimate" release source terms for dif ferent chemical groups of fission products are su.marized in Table 5. g The release of fission products from the damage core has been estimated to be a function of temperature,' and time af ter the loss-of-coolant accident. In the present procedure, the fraction of fission product release from the core is assumed to be proportional to the f raction of core damage as suggested by Malinauskas, et. al.' It is further assumed that the core is homogeneous so that each fuel rod has an identical exposure history. The fuel cladding rupture has been assumed to occur over the temperature range from abeut 780* to 1100*C,9 and the entire fission product noble gas inventory in the fuel gap would be released. All other fission products in the fuel gap, which may 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. Accr.,rding to a model calculation, portions of the fuel may start to melt before the cladding is totally destroyed. 3.3 150TCPIC DISTRIBUTION IN TUEL CAP Diffusion equations predict that the fractional release of radioactive isotopes from the fuel to the plenum and void spaces should be inversely pro-portional to the square root of the decay constant f or isotope reaching production-decay equilibrium.10,11 This prediction has been substantiated by experimental data reported by several investigators.12-17 A comparison of

  • Data ottained from CE unpublished document DRT 268-DEV00009. The fission product release pattern was found to be mostly " recoil."

5.7.5c-24 Rev.1 3-2

l a t 1 NEDO-22215 1 e. j isotepic distributions in the total fuel inventory and the predicted j 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 j

water samples, the source of fission product release may be semi-qqantita-tively' determined (see more discussion in Section 4). i 3.4 ANTICIPATED CHEMICAL BEHAVIOR OF IODINE AND COOLANT CHEMISTRY UNDER ACCIDENT CONDITIONS f. The results of measurements of Three Mile Island-2 (THI-2) ' " indicate that the airborne radiciodine release was much lower when compared to the 0 I noble gas activity release (by a factor of %20 ). Extensive investigations at the Oak Ridge National Laboratory (ORNL) on the nature and quantity of j fission product release from the over-heated fuel have concluded that cesium iodide Cs1 (B.P. = 1280*C) is the primary volatile species released from the O j fuel at elevated temperatures. The behavior of iodine under loss-of-coolant accident (LOCA) conditions has been evaluated by Lin and Campbell et. 'a1.22 1 21 j i ( Tor iodine at a concentration of a few ppm in aqueous solutions, the i redox reactions should be more predictable and formations of anomalous or 1 organic species should be much smaller than that at very low concentrations as generally assumed for radiciodine release. If iodine is released as CsI. it would stay in water as the I ion in a slightly basic solution (mainly due to ~ I Cs ions which may be released as elemental Cs or Cs oxides in addition to f CsI). Air oxidation or radiation-induced oxidation of 1 to 1 is not j ~ 7 l very likely to occur in a basic solution. In addition to the reducing nature i of zirconium and iron metals in the core. the production of hydrogen from Zr l steam reactions should make the chemistry, environment in the primary system f avorable to reducing reactions f or fodine. t There are at least three known velatile forms of iodine, 1. HIO and 2 f organic iodine. The formatie1 of 1 from I is not very likely in basic solu-2 l tions. The"ex'isthn~c'sofHIOhasneverbeenchemicallyidentifiedduetoits lowstableioSc's"n1 ration. TheairbornespeciescalledM10'ison2whichbehaves i } differently from 1 and organic iodine determined by using the iodine species 7 i l l 5.7.5C-25 Rev. l' 3-3 L

t NEDo-22215 sampling method developed by Keller, et. al. $ However, some convincing 0 evidence has been given by Lin that HIO, a product of 1 hydrolysis, is the 2 was initially added second volatile inorganic species in the gas phase when 12 to water in equilibrium partitioning studies. The partition coefficient increases with decreasing iodine concentration; at very low todine concentra-tions, the total iodine paJtition coefficiones have been dettrained to be %4000 at 21*C and %1600 at 72*C. It must be pointed out that since both 17 and M10 are very reactive species, any reducing impurities in water or on construction material surfaces would reduce 1 or HIO to I~ and significantly 2 reduce the airborne iodine concentration. The mechanisms of converting inorganic iodine to organic iodine, which is } generally observed in gas phase at very low concentrations, are largely unknown. It is certain, however, that at least more than a stoichiometric amount of organic species (or carbon-containing compounds) sheuld be readily available for reaction with iodine. As such organic species are limited, the results of several experiments indicate that the yield of organic iodine decreases with increasing iodine concentration in the gas phase. Less than 0.1% con-version is expected when the airborne fodine concentration is 1 g/m or larger. The total iodine concentration could be 43 g/m in the centainment free air space if all iodine is assumed to become airborne. It is also important to realize that the organic iodine, e.g., CH I is readily hydrolyzed 3 28 29 in water and basic solutions at higher temperatures. The half-time of hydrolysis is %20 min in water at 100'c and 43 see at 200*C, based on Heppolette and Robertson's data. O It is obvious that the very low release of iodine activities to the atmosphere -in the TM1-2 accident can be easily explained in tenns of the nature of iodine released from the fuel and the subsequent stabilization in I water. Water plays an important role in preventing iodine from release to the atmosphere. In the present procedure, all the iodine activities are assumed to stay in water, and the airborne activities are dominated by the neble gas fission products. 4 3-4 5.7.5C-26 Rev.'l

I O NEDO-22215 n e chemistry in the primary coolant may be significantly changed under accident conditions. Mainly due to the release of cesium, the water pH may 21 increase to %10.5 and the water conductivity may increase from %10 us/cm (terus water quality specification) to as high as %170 uS/ca. ll 4 1 i i I i I l l l f f i i i 5.7.5c-27 Rev.1 3-5 L

N NEDO-22215 i i 4. DISCUSSION AND SUGGESTIONS FOR FUTURE WORK It is obvious that the uneettainty of gas release fractions for iodine l and cesium are too large for an accurate calculation of the extent of core demage. While additional experimental work in fuel gap measurements is .apparently needed. *he lower li: nit release fraction for iodine may be j ' reevaluated by, examining the iodine spiking release data from defective fuel rods.following normal operation shutdowns. 1. Although the =1-131 spiking' data has been well documented previously, the analytical model may be refined to reflect more recent experimental data. The maximum spiking release of 1-13L estimated by Paseda;;3. based on pre-1973 g data,isobvicusiftoohigh.particularlywhntheimprovedfuelswhichare currently used in most of the operating BWRs are considered. i l ~ The accuracy of core damage estimation may be significantly improved by I measuring' more than iodine, cesium, and noble gas activities. ' Some less i volatile but easy to measure isotopes of Sr. Ba, 3.a. and Ru may be determined 2 in the water sample. More work, however, is needed to establish the release ( fractions as well as the baseline (shutdown spiking) concentrations for those 4 + isotopes. Additionally, some chemical analysis data, such as hydrogen pro-duction frem water-zirconium reaction may be used to improve the estimate. The degree of core damage between fuel cladding failures and core melt. namely fuel overheating, may be determined.,' i As sentioned in Section 3.3. it is possible to determine the source of 4 fission product release by measuring the activity ratios of noble gases or 1 iodine isotopes. It must be cautioned, however, each isotope should be accurately measured. Particular care must be exercised when the Xe-133 activity is determined in a sixture of other fission' products with high con-centration because of its low gamma ray energy (81 kev). Additional' work is required to perfect this procedure. (. i os' i l t. I l 5.7.5C-28 "Rev. 1 4-1/4-2 I j

1.__.. t NEDO-22215 I +. 5. REFERENCES 1. H. A. Carevay, " Calculation of Tission Product Inventory and Spectra -- RADC101 Program." NEDO-25176 (October 1980). 2 E. L. Strickland and N. R. Cash, " Operating Plants Parameters," NEDE-21398 Rev. 2 (September 1978) S. A. Hucik, private communication. 3. V. G. Pasedag " Iodine Spiking in BWR and PWR Coolant Systems," Proc. Topical Meeting on Thermal Reactor Safety, July 31-August 4,1977 Sun Valley Idaho. Conf-770708. k 4. J. M. Skarpelos and R. S. Gilbert, " Technical Deriviation of BWR 1971 Design Basis Radioactive Material Source Terms." NEDO-10871. (March 1973). 5. T. J. Brutschy, C. R. Hill, N. R. Horton, and A. J. Levine, " Behavior of Iodine in Reactor Vater During Plant Shutdown and Startup," NEDO-10585 (August 1972). 6. A=erican Nuclear Society. "American National Standard Source Term Specification " ANSI N237-1976. i 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. Commercial Nuclear Power Plants, WASH-1400. Appendix VII, p. VII-13. U.S. Nuclear - Regulatory Commission (October 1975)'. 9.- A. P. Malinauskas, R. A..Lorenz, H. Albrecht and H. Wild " LWR Source Terms for Loss-of-Coolant and Core Melt Accident," Proc. CSNI Specialists Meetings on Nuclear Aerosols in Reactor Safety, held at Gatlinburg, Tenn#., April 15-17, 1980, NUREG/CR-1724. 't 4 / 5.7.5C-29 Rev. 1 5-1 ~ ~.-

] NEDO-22215 J 10. A=erican Nuclear Society. " Methods for Calculating the Release of Tission Products from Oxide Fuels " proposed ANS-5.4 standard (November 1979). 11. R. A. Lorenz, J. L. Collins and A. P. Malinauskas Nucl. Tech., 46, 404 (1979). 12. Table VII 1-1, Reference 8. 13. C. M. Allison and H. K. Rae, "The Release of Tission Cases and Iodine frem Defected UO Tuel Elements of Dif ferent lengths " AECL-2206 (1965). y k 14 W. A. Yuill, et. al., " Release of Noble Cases from UO Tuel Rods," y 15-1346 (1969). 15. C. Jackson, D. Davico, M. J. Waterman, "The Ef fect of Neutron Flux on the Tission Product Cas Emission from UO at High Temperatures," y AERE-M-1607 (July 1965). ( 16. N. V. Krasnoyarov, V. V. Kenyashov, V.1. Polyakov, and u V. Chechetkin, Sov. At. Eng. (English), 38, 89 (1975). 17. General Electric Company unpublished data. 18. A. P. Hull, Trans. Am Kucl. Soc., 34, 91 (1980). 19. A. D. Miller, Ibid, 34, 633 (1980). 20. R. A. Lorenz, et. al., "Tission Pro' duct Release from Highly Irradiated LWR Tuel," NUREG/CR-0722 (ORN1./NUREG/TM-287/R2) February 1980. 21. C. C. Lin, " Anticipated Chemical Behhvior~ of Iodine Under LOCA Conditions," NEDO-25370 (January 1981). 22. D. O. Campb(11, A. P. Malinauskas, and W. R. Stratton, Nucl. Tech., 53, 111 (1981). 5.7.5C-30 Rev, t 5-2

NEDo-22215 23. M. Kahn and J. Kelinberg, " Radiochemistry of Iodine," NAS-NS-3062, ERDA (1977), p. 9. 24 C. C. Lin, J. Inorg. Nucl. Chem. 42,, 1101 (1980). 25. J. H. Keller, et. al., "A Selective Absorbent.ampling System for c Differentiating Airborne Iodine Species " Proc. lith Air Cleaning Conference CONF-700816 (1970), p. 621. 26. C. C. Lin, " Volatility of Iodine in Aqueous Solutions," J. Inorg. Nucl. Chem., 4_3_, 3229 (1981). 27. A. K. Postma and R. W. Zavadoski, " Review of Organic Iodide Formation Under t.ccident Condition in Water-Cooled Reactors," WASH-1233 (UC-80), 0:tober 1972. 28. R. L. Heppelette and R. E. Robertson, Proc. Roy Soc. (1959), A 252, 273. ( 29. M. Adachi, et. al., J. Chem. Eng. Japan, 7. (5), 364 (1974). i 5-3/5-4 5.7.5C-31 1 Rev.1

NEDO-22215-Appendix A SAMPLE CALCULATION OF TISSION PRODUCT INVENTORY CORRECTION TAC 70R Invent ry of nu lide i in reference plant T11 Inventory of nuclide i in operating plant = -1095 A 3651 (1-e I) 1 -A T))e-A T) g g 7 P) (1-e } where P) = steady reactor power operated in period j (Wt) ~ of nuclide i (day ) A, = decay constant T) = duration of operating period j (day) TO = time between the end of operating period j and time of last j reactor shutdown (day) 3651 = ave. operation power (in Wt) for the reference plant. 1095 = continuous operation time (in day) for the reference plant. Assuming a reactor has the following power operation history: Average Power Operation Time 0 y (gg) Operation T) (day) Feriod Days Since Startup 1A 1 - 60 60 254 1000 0 IB 61 - 70 2A 71 - 270 270 44 2000 0 2B 271 - 300 3 301 - 314 14 0 3000 A-1 5.7.5C-32 Rev. 1 m 9 e.- a-a

NEDO-22215 e For I-131 (A = 0.0862 day-I) 3651(1-e-0.0862x1095) F (1-131) = 1000(1-e-0.0862x60),-0.0862x m + 2000(1-e-0.0862x200) I ,-0.0862x44 + 3000(1-e-0.0862x14),-0.0862x0 3651 " % +.4 5 + 2103 " I " \\ -5 day'I) e For Cs-137 (1 = 6.29 x 10 -5 -6.29x10 x1095) 3651(1-e F ( '~ -O = 1000(1-e x60),-6.29x10 x254 -6.29x10 4 j ( + 2000(1-e x200),-6.29x10 x44 -6.29x10 + 3000(1-e x14),-6.29x10 x0 -6.29x10 i 243.16 " 3.74 + 24.93 + 2.64 l l l l i' l 5.7.5C-33'* Rev. 1 A-2

NEDO-23215 Appendix C ESTIMATION OF MAXIML'M 10DLNE-131 SPIKE CONCENTRATION IN REACTOR WATER The magnitude of iodine spiking in BWRs can be predicted by an empirical relationship proposed by 3rutschy, et. al.5 The data basis for the empirical relaticr. ship ara shown in Figure C-1 In order to predict the maximum I-131 concentration in reactor water (assuming no reactor water clean-up system in operation) during a shutdown spiking, the following information should be known for the reactor during steady state operation: 1-131 concentration in reactor water, 1-131 release source ter=, and the knowledge of the fission gas release characteristics. } Tht' fission product release from the d'efective fuel rods in a BWR during steady cperation is empirically characterized by ~ A (Ci/sec) = KY A g gg or R (fission /sec) = 7 = KA g where R = fission product release source term g K = a dimensional constant establishing the level of release b = a dimensionless constant establishing the relative amount of each nuclide in the mixture of similar chemical groups, i.e., noble gases or iodine isotopes. '~ t.- C-1. 5.7.5C-35 Rev. 1 l

NEDo-22215 ' O, Appendix B TOTAL I-131 RELEASE DURING A SPIKING SEQUENCE 'S 4 4 II 04 0 4 S \\ 5 0[ 0 0 4 4 8 4 4 n s 8 E ~ ( 4 o O o o o 5 O O O i O O 2 8 O S i 40 i l o i 1 ? "S i "E toiasvi7awista101 J B-1/2-2 5.7.5C-34 Rev. 1 l

a n NEDO-22215 io* O o \\ so' E o 0 o 5 I o w O I ( E s. 5 S 0 <u 1 = 1 I i 10 2 s 0 1o' 10 ,a 10 FISSION GASSODINE 131 RATIO i 1 1 Tigure C-1. Tission Cas to Iodine-131 Ratio versus Calculated Spike Magnitude (Assuming one peak, and cleanup system out of snrvice) I C-2 5.7.5C-36 Rev. 1

NEDo-22215 By plotting A /Y l against l for noble gases or iodine isotopes on a g gg g log-log paper, a straight line can be obtained with a negative slope of b as de:enstrated in Figure C-2. Experimentally, the noble gas activity release rate (concentration in the offgas times offgas flow rate) is measur.d at she steam jet air ejector (SJAE) sa:ple point. The measurement of iodine activity release rate is more co= plicated. By assumtng no iodine activity is returned from the feedvater into the reactor vessel.* The release rate can be calculated by A = C W(1 +S + S,) g g g \\ where A = release rate to coolant from the core, UCi/sec. g W = reactor water mass, Kg -1 decay constant of species i, sec 1 = S, = reactor water cleanup (RWCU) system removal time constant, -I, which is defined as 8, = f /W, assuming 4100% efficient. see f = RWCU flow rate, Kg/sec -I, which is defined as B, = cF/W 8, = steam removal time constant, sec c = fodine carryover defined as the ratio of the concentration of species i in condensate the concentration of species i in reactor water l F = steam flow rate, Kg/sec l l l This is true only f or non-forward pumping plants. l s. l C-3 5.7.5C-37 Rev. 1 i

NEDO-22215 The source term ratio (R /R,333) in Tigure C-1 is defined as the ratio y of the noble gas source term estimated at the A of I-131 to the I-131 source term (see Figure C-2). By knowing R /R _g3g and the magnitude of I-131 spik-7 ing (maximum spiking concentration / steady state concentration) from Figure C-1, the maximum I-131 concentration in reactor water during a shutdown, spiking can be estimated. d k i \\ 1 l l l l l i l I l l l l 5.7.5C-3D. Rev. I 1 C-4

NEDO-22215 I l l \\ i k n, E e I l E I _; n,.ui E ( I FOR NOSLE GASES I I I FORIODINE l l 1 I I Al131 LOGtv.sEc-' l Figure C-2. Plot of Log (R ) versus Log (A ) g g C-5/C-6 5.7.5C-39 Rev. 1

I NEDo-22215 DISTRIBUTION M/C W. E. Chamberlin 532 R. L. Cowan 117 R. M. Fairfield 855 T. A. Creen 277 E. Kiss 145 D. Knecht 738 K. E. Kolb 772 C. C. Lin (10) V04 W. J. Marble 165 L. B. Nesbitt 195 887 } H. D. Ongman R. N. Osborne V15 W. A. Pitt 195 K. Salahi 276 117 M. Siegler J. M. Skarpelas 117 E. T. Stell 858 G. E. Taylor 858 D. R. Wilkins 158 VNC Library (2) V01 NEBG Library (3) 528 k. TIE (5) SCH t I 5.7.5C-40 Rev. 1 7

ATTACIDENT 2 INTEGRATION OF OTHER Pl. ANT PARAMETERS INTO CORE DAMAGE ESTIMATE l i ) 5.7.5D-l' Rev. 1

.=

1. 0 INTEGRATION OF "0THER PLANT PARAMETERS" INTO CORE DAMAGE ESTIMATE

1.1 Purpose and Scope

1 The purpose of this section is to address NRC Clarificlition 2(a)2 regarding the integration of other plant parameters into the determination of an estimate of core damage for a suspected degraded core event. This additional information would provide verification t of the initial estirate of core damage based on radionuclide measure-ments using the post-accident sampling system (PASS). Theprocedureforestimationofcoredamagebasedonradionuclide) f measurements from the PASS has been previously described and provided to those utilities which use that system. That procedure involves 1 { calculations of fission product inventories in the core and the { release of those fission' prod'ucts into the primary system under postulated loss of-coolant (LOCA) conditions. For that procedure, a BWR-6/238 with a Mark III containment is used as the reference plant. For that plant, plots of core damage estimate versus radio- ~ 1 nuclide concentrations in reactor water and containment atmosphere are provided. The procedure describes the method for determining the extent of core damage for each unique' plant by comparing the measured concentrations of major fission products in either gas or l water samples, after appropriate normalization, with the reference 4 plant data. Consequently, core damage estimates can be made for each unique plant based on radionuclide measurements using the PASS. 1.2 Identification of Other Significant Parameters { There are several other plant parameters which are measured in the BWR which can provide sufficient information to confirm his initial core damage estimate based on radionuclide measurements. 5.7.5D-2'- Rev.1 1-1

1 s. For some of these parameters, correlations similar to that which is provided for the radionuclide measurements can be developed which provide confirmation of the initial core damage estimate. Such correlations can be developed for the parameters of containment ra'diation level and containment hydrogen level. Containment radiation level provides a measure of core damage, because it is an indication of the inventory of airborne fission products (i.e., noble gases, a fraction of the halogens and a much smaller fraction of the particulates) released from the fuel to the containment. Containment hydrogen levels, which are measurable by 4 the PASS or the containment gas analyzers, provide a measure of theI extent of metal water reaction which, in turn, can be used to estimate the degree of clad damage. Another significant parameter for the estimation of core damage is reactor v'essel water level. This parameter is used to establish if there has been an interruption of adequate core cooling. Significant periods of core uncovery, as evidenced by reactor vessel water level readings, would be an indicator of a situation where core damage is likely. Water level measurement would be particularly useful in distinguishing between bulk core damage situations caused by loss of adequate cooling to the entire' core and localized core damage situations caused by a flow blockage in some portion of the core. i There are other parameters which may provide an indication that a core damage event has occurred. These are main steam line radiation level and reactor vessel pressure. The usefulness of main steam line radiation measurement is limited because the main steam line radiation monitors are downstream of the main steam isolation valves (MSIVs) and would be unavailable following vessel isolation. Reactor vessel pressure measurement would provide an ambiguous indication of core damage, because, although a high reactor vessel pressure may be indicative of a core damage event, there are many non-degraded core events which could also result in high reactor vessel pressure. l 1-2 5.7.5D-3 Rev. 1

t There are other measurements besides radionuclide measurements which are obtainable using the PASS which would further aid in estimating core damage. As noted in the procedure already supplied, detection of such elements in the reactor coolant as Sr, Ba, La and Ru is evidence of fuel melting. These indications could be factored into the final core damage estimate. 4 1.3 Application of Other Sionificant Parameters to Core Damage Estimate As noted in Section 1.1, procedures have already been developed which provide an estimate of core damage based on radionuclide measurements. Based on these procedures, an initial assessment of ) core damage is made. Based on a clarification provided by the NRC, that assessment would appear in a matrix as follows: 4 Degree of Minor Intermediate Major Degradation (<10%) (10%-50%) (>50%) No fuel damage 2 1 Cladding Failure 2 3 4 Fuel Overheat 5 6 7 Fuel Melt 8 9 10 As recommended by the NRC, there are four general classes of damage and three degrees of damage within each of the classes except for the "no fuel damage" class. Consequently, there are a total of 10 possible damage assessment categories. For example, Category 3 would be descriptive of the condition where between 10 and 5C percent of the fuel cladding has failed. Note that the conditions of-more than one category could exist simultaneously. The objective of the final core damage assessment procedure'is to narrow down, to the maximum j extent possible, those categories which apply to the actual'in plant j situation. i l 5.7.5D-4 Ren.1 - 1-3 l en e -y 4 -s-p 4m- ,-4 ye-w,--yw y 9 -m ew-,- +4 y , - -m-

8 The initial core damage assessment based on radion'uclide measurement will provide'one or several candidate categories which most likely represent the actual in plant condition. The other parameters should then be evaluated, as identified in Section 1.2, to corroborate and further refine the initial estimate. For' example, fission product measurement using P. ASS may indicate Category 4 core damage and, additionally, the potential for fuel overheat and fuel melt (i.e., Categories 5 thraugh 10). Measurement of hydrogen in containment and use of the hydrogen correlation provided in Appendix A of'this report could be used to verify that extensive clad damage had occurred. Useofthecontainmentradiatioh monitor reading along with the correlation provided in Appendix B of this report would verify that a significant fission product release to the containment had occurred, further verifying the initial assessment. Further analysis of the PASS samples for concentrations of.Ba, Sr, La and Ru and consideration of the relative amounts of fission t products released would indicate if any fuel melt had occurred. i i The flow diagram in Figure 1 indicates how the analysis of the other. significant parameters relates to the estimation of. core damage I based on radionuclide measurements. 'As noted earlier,-Appendices A l and B provide correlations for the determination of the degree of core damage based on containment hydrogen and radiation' levels. 1.4 Determination of Sample Location In. order to assure a representative sample which reflects the actual in-core condition, care must be taken in selecting a suitable sample i location. The selection of a sample location should account for the type of event which will determine where the fission products will i concentrate. I ?- 1-4 , 5.7.5D-5 Rev. 1 T

1 0 l For gas sampling, the recommended sampling locations are as follows: Event Type Sample Location Non-Breaks (e.g., MSIV Closure) Suppression Pool Atmosphere Small Breaks Drywell (before depress.) Suppression Pool Atmosphere (after depress.) Large Breaks (liquid or steam) Drywell in Containment Large Breaks outside containment Suppression Pool Atmosphere Forliquidsampling,theoptimumsamplepointforalleventsisthg jet pumps as long as there is sufficient reactor pressure to provide a sample from that location. If there is not sufficient reactor pressure to allow a sample to be taken from the jet pumps, then the sample should be taken from the sample point in the RHR system. In order to ensure a representative liquid sample from the jet pumps at low (<1%) power conditions for small break or non-break events, i the reactor water level should be raised to the level of the moisture separators. This will fully flood the moisture separators and will provide a thermally induced recirculation flow path for mixing. Several requisition plant licensees have already committed to the NRC to perform this procedure. i 1-5 5.7.5D-6 Rev. 1

I ~ 1. 5 References 1. C. C. Lin, " Procedures for the Determination of the Extent of Core Damage Under Accident Conditions". NEDO-22215, August 1982. 1 \\- ( + r 'l-6 5.7.5D-7 dev. 1'

.o FIGURE 1-1 SEQUENCE OF ANALYSIS FOR E_STIMATION OF CORE D_AMAGE Hydrogen Containment Water NORMAL OPERATION Y'S Radiation M Level Y'S MINOR CLAD DAMAGE Analysis (Confirm) (Confirm) (Con firm) E E E E a Detemine Core Damage Optimum Estimate lI Sanple From PASS e o n Point E E E \\ Hydrogen Containment Wa ter Analysis For Analysis y,, : Radiation y,, : Level y,, : Ba, Sr, La, Ru 4 5 (Confirm) (Confirm) (Confirm) in o NAJOR CLAD DAMAGE Detemina tion FUEL OVERHEAT ,Yes Of Fission FUEL E LT Product Ratios E v. o F CLAD DAMAGE i y POSSIBLE FUEL OVERHEAT = NO CORE MELT W

i i-.e. - p;.,- 1. i< io [ I t i 4, 1 i 1. 1-l, t j c I ) i -APPENDIX A i 3 INTEGRATION OF' CONTAINMENT ATMOSPHERE HYDROGEN MEASUREMENT INTO t CORE DAMAGE ESTIMATE t t I t i i I i. i V t i 5.7.5D-9s Rev.l ~ i E I

1 4

SUMMARY

The extent of fuel clad damage as evidenced by the extent of metal water reaction can be estimated by determination of the hydrogen concentration in the containment. That concentration is measurable by either the containment hydrogen monitor or by the post accident sampling system. i A correlation has been developed which relates containment hydrogen concentration to the percent metal water reaction for Mark I, II and III-type containments. That correlation is shown in Figure A-la. Note A to that figure indicates the major assumptions used in developing the correlation. Note B indicates the method by which individual utilities k can use the correlation to determine the extent of clad damage. (

  • Correlation is based on the following formula:

(1641) %H2, g g) N V (1641) (748) (MWR) + (3176) I.36 x 106) I Where: N = number of fuel bundles V = containment volume, ft3 MWR = fraction of cladding in active fuel zone reacting %H = concentration of hydrogen in containment atmosphere, mole % z A-1 5.7.5D-10 Rev. 1

04 8 M n y M g W = a a O O N g M e M S2 n I M g g g j n = a 5 4 M ~ j w 88 M 2 te 33 j 14 p N q 5 = 12 g N 3 5 = m h 2 e 12 4 o 4 4 2 = i a a a a a a a a a f 0 to 20 30 40 50 80 70 30 90 100 ~ i 4 NTAt.- WATER Rf ACTMM Munte star 1 -~ Figure A-l. Hydrogen Concentration for Mark 1/II and III Containments an a Function of Metal-Water Reaction i

o fD i

= ? W l.

t 0 i i.'. Note A-Analytical Assumptions 1. Containment Volume = 350,000 ft2 (MK I-II) 1,360,000 ft3 (MK III) 2. Number of bundles 500 (MK I-II) = 748 (MK III) 3. Fuel type = 8x8 R 4. All hydrogen from metal-water reaction released to containment \\ 5. Perfect mixing in containment 6. No depletion of hydrogen (e.g., containment leakage) 7. Ideal gas behavior in containment. ( 4 A-3 5.7.5D-12 Rev. 1

f g_

e. O w.

j Note B-Determination of Clad Damage from Hydrogen Monitor Reading i Step 1. Obtain containment hydrogen monitor reading, [H],'in %. Step 2. Using the appropriate curve in Figure A-1, determine the metal water reaction for the reference plant, MW at [H]. ref Step 3. The metal water reaction for the actual in plant conditions (MW) is detarmined from the following equation: \\ %MW = (MWref) ( ) ( 350 000 ) I'II e 4 %MW = (MWref) ( N ) I 1,36.000 ) MK III where N = number of bundles V = total containment free volume, ft2 A-4 5.7.5D-13 Rev. 1

-i g .g: 1 (4 l s I APPENDIX 8 i .. INTEGRATION OF CONTAINMENT ATMOSPHERE. . RADIATION MEASUREMENT INTO-C0RE DAMAGE ESTIMATE ~ t I 3 1 i i l i i i i i g i 4 g it .5.7.5D-14 Rev. 1 [ r- ?

J V. %d I

SUMMARY

l As noted in the Task 1 response, an indication of the extent of core damage is the containment radiation level which is a measure of the inventory of fission products released to the containment. Several plant procedures already contain a correlation of the containment radiation monitor dose rate to the percent of fuel inventory airborne in the containment. The purpose of this appendix is to present that correlation and provide a method whereby individual plants can use that correlation to determine the degree of core damage. Figure B-1 provides the results of a correlation performed for the ) Monticello plant. The key parameters which impact the containment dose rate are reactor power, containment volume and monitor location within the containment. The method whereby individual plants can' apply this correlation is provided in Note A in Figure B-1. ~ ( 4 B-1 5.7.5D-15 Rev. 1

g FIG 1)#E B-1 \\1 I tsrcent of Puel 2nvu.ntory Airborne la, the Comtainment b' 1906 Puel 2nventory = 3006 Noble Gases b']

  • 2St I'd "'

e 3t particulates );s 20, s is' q e= 24 $ Wl se \\ L o i 10' [ j o. ! lo' q E e.o, 5 i j IC', g 8.o024 t

b..

10 I ) 'i i hilf f I 3 A h h)W10' ) ) k 6i'rePitt ) ) A66sei10' lC nme After Shutdoun (Hrs) 4 Pse! Invenscry Ap;roximate Sotree and Da:tage Estimate Released l 200. 2004 T2D-24944, 2006 fuel damage, posaatial core melt. 50. 50% T3D amble gases, gM2 sourse. 10. 104 T2D, 2004 3rRO gap activity, total elad failure, partial more ancovered. 3. 34 T2D, 2004 MA33-1400 gay activity, major elad fallare. 1. In TD,194 ERO gay, aina.104 clad f;silare.

  • 1

.3% TID, 3t une gep, 24 elad fallart, Sesal heating rf 5-30 fuel assemblies. .51 .824 fuel T2D,.34* IrAC gap, 32nd fallare of 3/4 element (36 reds). 10-3 .91t une gay, slad fai2ure of a few reds. St*4 2006 anotaat release with spiking. 5:33-6 2004 esolaat is,vatory release. 10*8 Opper range of normal airborne amble gas activity la acetainammt. B-2 5.7.5D-16 Rev.1

i 1 r O' N Note A-Determination of Clad Damage From Containment Radiation Monitor Reading The procedure for determination of fraction of fuel inventory released to the containment is as follows: Step 1: Dbtain containment radiation monitor reading,.[R] in Rem /hr. Step 2: Determine elapsed time from plant shutdown to the containment radiationmonitorreading[t]inhodrs. \\ Step 3: Using Figure B-1, determine the fuel inventory release for the reference plant [I]ref I" I-Step 4: Determine the inventory release to the containment [I] using the following formula:* [I] = [I]ref I ) (237 450) (6/D) i where P = reactor power level, MWth i V = total containment free volume, ft2 D = distance of detector from reactor biological shield wall, ft.

  • Assumes a wall-mounted detector and no major equipment in detector line-of sight except for reactor vessel shield wall.

1 B-3/B-4 5.7.5D-17 Rev.I -}}

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