ML20128N934
| ML20128N934 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/06/1985 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20128N917 | List: |
| References | |
| 5.7.5, NUDOCS 8506030423 | |
| Download: ML20128N934 (25) | |
Text
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~
,. l NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION l Procedure No. 5.7.5 ESTIMATING CORE DAMAGE List of Effective Paces
- 5. 7. 5-1 5.7.5-2 Attachments 5.7. 5 A-1 5.7.5B-1 5.7.5C-12 5.7.5C-29 5.7.5D-6 5.7.5A-2 5.7.5B-2 5. 7. 5 C-13 5.7.5C-30 5.7.50-7 5.7.5A-3 5.7.5B-3 5. 7 . 5C-14 5.7.5C-31 5.7.50-8 5.7.5A-4 5.7.5B-4 5.7,5C-15 5.7.5C-32 5.7.50-9 5.7.5A-5 5.7.5B-5 5.7.5C-16 5.7.5C-33 5.7.50-10 5.7.5A-6 5. 7. 5 C-1 5.7.5C-17 5.7.5C-34 5. 7. 5 D-11 5.7. 5 A-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. 5 C-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. 5 A-15 5.7.5C-9 5.7.5C-26 5.7.50-3 , j, 5.7.5 A-16 5.7. 5 C-10 5.7.5C-27 5.7.50-4 ,
.f/ ,
- 5. 7. 5 A-17 5.7. 5C-11 5.7.5C 28 5.7.50-5 Approved '*
ORCCyairman Date / // blU3 7 8506030423 850530 5.7.5-1 Rev. O PDR ADOCK 05000293 P PDR f i
.. ,.-.. . ~ - -
- 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 othe,r. available plant parameters.
~
1 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 FASS 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 infomation 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 , F C. Operating Procedures 2.2.133
- ~ D. 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 VII. LIMITATIONS None VI!!. PROCEDURE A. Assess the degree of core damage as described in Attachment A. 4 B. Verify the assessment by integrating other plant parameters as described in Attachment 8. 5.7.5-2 Rev. 0 1
ATTACHMENT A 1,- Procedure to determine Core Damage based on samples collected from the Jet Pump, RHR and Torus Sample Points. A. Set Cwg_131 and CwCS-137 as the measured concentrations (in uC1/g) of I-131 and CS-137 f rom the Jet Pump Sample Points. , Cwl-131 CwCS-137
- 8. Correct the measured concentrations of the above radionuclides to the time of reactor shutdown as follows:
JP JP
- 1. Let; Cwi l-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 JP
- 2. Find Cwi l-131 where JP Cwil -131 = CW l -131 x e3.5914 E-03 x t Cwil -131 "
JP
- 3. Find CwiCS-137 where Cwi -137 = CwCS-137 x e2.6221 E-06 x t JP CwiCS-137 "
5.7.5A-1 Rev. 0 n: 'n -- n. m l *i f ulI f y i
.1 "lV .
, u
. . - - . . C. Set Cwg_131 and CwCS-137 as the measured ccncentrations (in UC1/g) cf 1-131 and CS-137 from the RHR Sample Points.
. Cwg_131 - =z CwCS-137 D. Correct the measured concentrations of the above ra:lionuclides to the '."
time of reactor shutdown as follows: RHR RHR
.1. Let; Cwi l-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-RHR
- 2. Find Cwi l-131 where 3
Cwi _ 31 = CW I _131 x e .5914 E-03 x t RHR Cwit_131 =
. RHR 3.- Find CwiCS-137 where 2
Cwi -137 = CwCS-137 x e .6221 E-06 x t RHR CwiCS-137 " 5.7.5A-2 Rev. 0
,Q1 " !!"9 T / .c gj n
'"l iUJhild-" Uh J\ _Y
n-l E. Set Cwg_j33 and CwCS-137 as the measured concentrations (in uCi/g) of I-131 and CS-137 f rom the Torus Sample Points.
=
Cwg_131 ,
" 1 CwCS-137 ,
F. Correct the measured concentrations of the above radionuclides to the time of reactor shutdown as follows: T T
- 1. Let; Cwig_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 twig _131,where Cwi _131 = CW I -131 x e3.5914 E-03 x t T .
Cwil -131
- I
- 3. Find CwiCS-137 where Cwi CS-137 " CWCS-137 x e2.6221 E-06 x t l
T Cw1CS-137
- 5.7.5A Rev. 0
'*l- n
((^ 7 iut iiCL]Ri,nM'--v<\g.niv _y
* ,,-ww-.w,--, , , . , , , , , , - , , - . r-,,. -- e , ,
- - - . - . - v - - - - - . , - - - -y- r-----.p... -
-,,-------,fwr, m-3
m- . . . . .. 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 l isotope of interest. I
- a. Find Cwial-131 as follows:
Cwia l-131= Cwig ,33)(Step B. or 0.) x 2.05 E 08 + Cwl;_)3)(Step F) x 2.38 E 09 2.585 E 09 Cwial-131, "
- b. Find twiaCS-137 as follows:
CwiaCS-137 * " CS-137( ep B. or D.) x 2.05 E 08 + CwiCS-137(Step F) x 2. 2.585 E 09 CwiacS-13f " l l I _01 \0MK'0\oty
/
5.7.5A-4 Rev. 0
H. Calculate the fission product inventory factor as follows:
,= Fission Product inventory
- 1. Let Fli -131 l and FliCS-137 correction factor for the radionuclide of interest at the
- time of reactor sh'utdown. , , ,
Pi = Steady state. reactor power le' vel operated during period number j in MWt - Note the variation in power level should he limited to 120% to meet steady state conditions, f- = NUMBER of operating periods at a
~F --
n n. steady state power level since
,I ~
l I startup f rom the last refueling u V- - 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.
- 2. Find FIi l-131 where Fli l-131 l
i 3.651 E 03 g p) g , ,-3.5914 E-03 x Tj -3.5914 E-03 x Tj) i Fli l-131 = l
- 3. Find FliCS-137 where FliCS-137 =
2.431 E 02
<j
< [Pj (1 - e-2.6221 E-06 x Tj)('-2.6221 E-06 x Tj)
FliCS-137 = 5.7.5A-5 Rev. O
- 1. Calculate the normalized concentration of the radioisotope of interest as follows:
e REF REF
- 1. Let Cwil-131 & CwiCS-137 = Normalized concentration of th'e' radioisotope of interest at the time of reactor shutdown.
REF REF
- 2. Find Cwil -131 where Cwil-131 ,
Cwial-131 x Fli l-131 x 6.594 E-01 REF Cwil -131 " 4 REF REF
- 3. Find CwiCS-137 where CwiCS-137 "
CwiaCS-137 x FliCS-137. x 6.594 E-01 REF CwiCS-137 "
.1. Refer to Figures 1 and 2 to estimate the extent of fuel or cladding damages.
- .m ,o t: i
-n- : n1 1
-lu 1
j,;
-! yl ; ,
L L L I V }lt
\ .
.- 5.7.5A-6 Rev. 0
II. Procedure to determine Core Damage based on samples collected from the Torus i 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. .
>p CgXe-133
=
i CgKr-85 { 8. Correct the measured concentrations of the above radionuclides to the I 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.
l ' t= Time period (hrs.) between reactor shutdown and the deter-l mination of the radioactivity of l
- the radionuclide of interest l
D
- 2. Find CgiXe-133 where D
CgiXe-133 = CgXe-133 x e5.5 E-03 x t D CgiXe-133 = 0
- 3. Find CgiKr-85 where D
CgiKr-85 = CgKr-85 x e7.3796 E-06 x t 1 D C91Kr-85 = l l F "
- n. a , 5.7.5A-7 Rev. O Any
.' L; 7 \ vi* lV A - An c
C. Correct the measured atmospheric samples for temperature and pressure dif ference between the sample vial and Drywell. D D f
- 1. Let CgicKr-85 and CgicXe-133 = Concentration of the radionuclides of interest 8 the Drywell atmospheric temperature, and pressure 6 the time of reactor shutdown.
T1
= 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-Cgic r-85 = Cgifr-85 PpT)
PT j2 D CgicKr-85 " D
- 3. Find CgicXe-133 where:
0 D CgicXe-133
- E9 Xe-133 PpT)
PT j2 D CgicXe-133 *
' ' '^~' "-
FnJ Folv, A~ o_n nY._Y m
D. Set CgXe-133 and CgKr-85 as the measured concentrations (in uti/g) of Xe-133 and Kr-85 from 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
- 1. f.et; CgXe-133 and CgKr-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 T'
- 2. Find CgiXe-133 where
( T T CgiXe-133 = CgXe-133 x e5.5 E-03 x t i T CgiXe-133 " l T ! 3. Find CgiKr-85 where T T CgiK r-85
- CgKr-85 x e7.3796 E-06 x t l
T CgiKr-85 = FDR \20lVK70N ONLY l l l
F. Correct .the measured atmospheric samples fer temperatura and pressure difference between the sample vial and Torus. T T '
- 1. l Let CgicKr-85 and CgicXe-133
= Concentration of the radionuclides of interest 9 the Torus,,,
atmospheric temperature and pressure 9 the time of reactor shutdown. T1
= Temperature of the gas in the sample' vial. *F
= Pressure of the gas in' Pi the sample vial. PSIA-I T2
= Temperature of the gas in the containment. *F
= Pressure of the gas in P2 the containment. PSIA T
- 2. -Find CgicKr-85 where:
T , CgicKr-85 = Cgigr-85 PpT)
'12 T
T
.CgicKr-85 " ^
t' T
- 3. Find CgicXe-133 where:
CgicXe-133 = CgiXe-103 P2T) P)T2
- - T CgicXe-133
- Ij N ! 3 I\ [O3"i/A";*\l \/ 5. 7. 5 A-10 Rev. O i'"
u\ lT uOMi u ,4 ]'v :\l' L.I
---- - r -- - - - . - - - - - - - -
7.
- 6. 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= (Cgicle-133x 4.16 E 09) + (Cgic Xe-133 x 3.18 E 09).
7.34 E 09 CgiaXe-133 "
- b. Find CgiaKr-85 as follows-x 3.18 E 09)
CgiaKr-85 = (Cgiclr-85 x 4.16 E 09) + (CgicKr-85 7.34 E 09 CgiaKr-85 " l l i .-
, t f
5.7 . 5 A-11 Rev. O i
H. Calculate the fission product inventory factor as follows:
- 1. Let Fli e-133X and FliKr-85 ,= Fission Product inventory correction factor for the radionuclide of interest at the time of reactor shutdown. ,
4 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 = NUM8ER of operating periods at a steady state power level since startup from the last refueling q 11 outage to the last shutdown
't ;
-~ "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.
- 2. Find FliKr-85 where FliK r-85
- 6.436-E 02
-7.3796 E-06 x Tj y{,-7.3796 E-06 x'TJ NJ [PJ (1 - e FliKr-85 "
1 1
=
-3. Find Fli e-133X where Fli Xe-133 3.651 E 03 E-03 x Tj - 5.5 Ij[Pj(1-e - 5.5 E-03 x TJ)
FliXe-133 "
- 5. 7. 5 A-12 Rev. O i
l t-
I. Calculate the normalized concentration of the radioisotope of interest as follows: r 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 FI1Kr-85 x 1.835 E-01 REF CgiKr-85 " REF . REF
- 3. Find CgiXe-133 where CgiX e-133
- CgiaXe-133 x Fli Xe-133 x 1.835 E-01 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. 0
r -- - S C' , putLt'ELTDCH=% WPPER RELE ASE Liutt SEET ESTIMATE
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(Reactor Water + Pool Water) and the Extent of Core Damage in ! Reference Plant.
- 5. 7. 5 A-15 Rev. 0
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% % S WE L WE LTDOWN Cl Figure 1. Relationship Between 1-131 Concentration in the Primary Coolant l (Reactor Water + Pool Water) and the Extent of Core Damage in l
! Reference Plant. l FOR N:0RMg m OK y
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Figure 4. (Drywell + Torus Gas) and the Extent of Core Damage in Reference Plant. 1 5.7. 5 A-li Rev. 0 20R N 0 NATL 0K ON Y
~ _ _ _ . .
Beco #26 (PR02-1) ATTACHMENT 8 r I. Procedure to determine the degree of cladding damage from analysis of the ' 5' torus atmosphere H2 monitor (in panel C-174 or C-175 indicators AIT-2A-5082A or AIT-2A-50828 respectively) readings. A. Set He as the measured concentration (%) of hydrogen in the torus as indicated ont he torus atmosphere monitors (panel C-174 or C-175) HD" .
- 8. From Figure 1 determine the metal water reaction (%) for the reference plant.
WREF " . 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 %MW where 5MW = WREF x 6.3845 E-01
.%MW =
FOR TORI!KE\! OEY . - - . . . - - _ 5.7.58-1 Rev. 0
=
- Seco #26 (PR02-1)
II. Procedure to determine the degree of cladfing 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 H0 as the measured concentration.(5) of hydrogen in.the torus as
, indicated ont he torus atmosphere monitors (panel C-174' or C-175)
HD"
- 8. From Figure 1 determine the metal water reaction (%) for the reference plant.
WREF " 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 %MW where %MW = MWREp x 6.3845 E-01 f
l.
=
l l I i-FOR \20RIVATEN ONLY i 5.7.58-2 Rev. O
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 / B 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: ORYWELL TORUS _
- CH. A / 8 CH. A / B
/ /
- 1. Fuel Failure Type:
/ /
- 2. 5 of Noble Gas:
/
- 3. 5 of Halogens: /
/
- 4. 5 of Solids: /
l l 1 FDR FCWAT DN ONLY ' 5.7.5B-3 Rev. 0 1 v_ _ l
l l l . iiMYDmoCEM ley mean) aAARE ull. 30 Age PT3 Am suseDLES
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REV. 0
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FIGURE 2 CONTAINMENT RADIATI0tt YS. TDE 701.LOUTNG ACODENT -- . 107- - --
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~
ATTACHMENT 1 k PROC'EDURES FOR THE DETERMINATION OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS 5.7.5C-1 Rev. O
J. NEDO3 82NE: eu AUGUST l t; m .:a...s we m.w cw w sum.ra mm mes \m w PROCEDURES FOR THE DETERMINATIO OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS .
/
1 C. C. LIN k T* ~~ E , l 1Ef: . L':_. !ii'i Ni?_i?<'iEiTZ'aNltU /Ed '~C >PG-efX.io-iE' GENERALh ELEC 5.7.5c-2 Rev. O t.wu a v i-i w ii ,p pp.
- t NEDO-22215 82NED090 f
Class I August 1982 l I k PROCEDURES FOR TE DETERMINATION OT TE EXTENT OF CORE DA%\GE UNDER ACCIDENT CONDITIONS l I
~
1 i I Chien C. Lin Approved:
*h R. L. Cowan, Manager Plan't Chemical and Radiation Technology-I l
Approved: L. E. Kiss, Manager Plant Technology l l NUCLEAR P0kTR SYSTEM ENGINEERING DEPARTMENT. CENERAL ELEC VALLECITOS NUCLEAR CENTER, PLEASANTON. CA 91566 GENERAL h ELECTRIC - 5.7.5C-3 Rev. O l WW
.. .. NEDO-22215 r
DISC 1J.D!ER OF RESPONSIBILITY Neither the T*nis dor.ennt was prepared by or for the General Electric Company. General Electric Company nor any of the contributors to this document: A.. Makesanywarrantyorrepresentation,expressorimp2ied,withrespect the accuracy, cor"pteteness, or usefulness of the infomation contained in this dor. ment, or that the use of any infomation disclosed in this docu-ment may not infringe. privately omed rights; or
. B.
Asszenes any responsibility for liability or danage of any kind which m:y reeuit from the use of any infomation dis Zosed in this dor.cnent. 1-2 ii 5.7.5C-4 Rev. 0
NEDO-22215 CONTENTS : Zag r 1-1
- 1. OBJECTIVE AND SCOPE '
. 2-1 '
- 2. PROCEDURES FOR DETERMINATIONS OF CORE DAMAGE 2-1 2.1 Reference Plant (BWR-6/238, Mark III) 2-1 2.1.1 Reference Plant Parameters. 2-1 2.1.2 Frocedure 2-1 2.1.3 Supplementary Data 2-8 2.2 Specific Flant Application 2-8 2.2.1 Plant Parameters 2-11 2.2.2 Procedure 3-1
- 3. TECHNICAL BAS 15 3.1 Tission Product Concentrations in the Primary System During 3-1 )
Reactor Shutdown Under Normal operation Conditions 3-1 3.1.1 Tission Product Concentrations in Reactor Water 3.1.2 " Noble Gas _ Concentrations in Drywell and Torus
- 3-1 Cas Phase 3.2 Tission Product Release Source Terms Under Accident
' 3-2 Conditions 3-2 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
- 5. RETERENCES APPENDICES SAMPLE CALCULATION OF TISSION PRODUCT INVENTORY A-1 CO A.
TACTOR
. B-1
- 3. TOTAL I-131 RELEASE DURING A SPIKING SEQUENCE C-1 C. - ESTIMATION OT MAXIMW110 DINE-131 SPIKE CONCENTRATI REACTOR WATER iiiliv 5.7.5C-5 Rev. 0
NEDO-22215 TABLES
/ Page Title 73}le Core _ Inventory of Major Fission Products in a Reference 1 . 2-2 >>
Plant Operated at 3651 MWt for Three Yeara 2 Fission Product Concentrations in Reactor Water and Drywell 2-3 Gas Space During Reactor Shutdown Under Normal Conditions l' 2-9
-3 Ratios of Isotopes in Core Inventory and Tuel Cap 2-10 I 4 Plant Parameters 2-15 l 5- Best-Estimate Tission Product Release Fractions
\
l ILLUSTRATIONS . l-Paje Title I i Tirure i 1 Relationship Between I-131 Concentration in the Primary
- 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 [" 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 Reference Plant 4 Relationship Between Kr-65 Concentration in the containt.ent Gas (Drywell + Torus Gas) and the Extent of Core Damage in 2-7 Reference Plant l i l v/vi 5.7.5C-6 Rev. 0 l
l NEDO-22215
- 1. OBJECTIVE AND SCOPE )
/
The purpose of this procedure is to determine the degree of reactor core damage from the measured fission product concentrations inTheeither pro-
'the wa gas samples taken from the primar ' systas under accident conditions.
cedure involves calculations of fission product inventories in the core and the release of inventories into the primary system under postulated loss-of-The fuel gap fission products are assumed coolant accident (LOCA) conditions. The majority of fission to be released upon the rupture of fuel cladding. l product inventories in the fuel rods would be released when the fuel is m l at higher temperatures. A Bk'R-6/238 with a Mark III containment is used as}a Application of the reference plant in the demonstration of this procedure. procedure for any other type or size of boiling water reactor (Bk'R) is described. , l l l
- 1. .
l I l l l- . 1-1/1-2 5.7.5C-7 Rev. 0 i
NEDO-22215
- 2. PROCEDURES FOR DETERMINATIONS OF CORE DAMAGE RETERENCE PLANT (BWR-6/238, MARK III) <
2.1
.y 2.1.1 ' Reference Plant Parameters The pertinent plant parameters for the reference plant are given below:
3579 MWt Rated reactor thermal power 748 bundles Number of fuel bundles ,
' Total primary coolant mass (reactor 9 vater plus supprnsion pool water) 3.92x10g\
Total containment and drywell gas space 4.0 x 1010cc volume The fission product inventories in the core are calculated based on three f rated
- years. (1095 days) of continuous operation at 3651 MWe, or 102% o power, by using a computer code developed at Los Alamos and adapted to the GE The inventories of some major fission products in the core computer system.
at the time of reactor shutdown are given in Table 1.- 2.1.2 Procedure Either the gas .or water samples taken from the post accident sampling i by gamma ray l
. system are analyzed for major fission product concentrat ons l
spectrometry. If the concentration of a fission product in reactor water or drywell, corrected the decay to the time of reactor shutdown, is measured f to be higher than the baseline concentration shown in Table 2 (see See-tion 3.1 for details), tl., extent of fuel or cladding damage can be deter-mined directly from Tigures 1 through 4 based on isotopes I-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. 2-1 Rev. O 5.7.5C-R. i~
*- ** NEDo-22215 Table 1 CORE IhTENTORY 07 MAJOR TISSION?RODUCTS IN A
' REFERENCE PLANI OPERATED AT 3651 MWt TOR THREE YEARS Major Gansna Ray Energy ,,
Inventory ** (Intensity) kev (y/d)
'. Isotope
- Balf-Life 106 Ci Chemical Group 4.48h 24.6 151(0.753)
Noble gases .Kr-35m 1.1 514(0,0044) i Kr-85 10.72y l 76.3m 47.1 403(0.495) Kr-87 66.8 196(0.26) 1530(0.109) i Kr-88 2.84h 5 25d 202.0 81(0.365) Xe-133 250(0.899) i 9.11h 26.1 Xe-135 8.04d 96.0 364(0.812) I Nalo8 ens 1-131 668(0.99,773(0.762) 2.3h 140 I-132 201 530(0.86) i-1.133 20.8h 52.6m 221 847(0.954) 884(0.653) I-134 189 1132(0.225).1260(0.286) I-135 6.61h
~
2.06y 19.6 605(0.98).796(0.85) Alkali Metals Cs-134 662(0.85) 30.17y 12.1 l Cs-137 Cs-138 32.2m 178.0 463(0.307).1436(0.76) .' 78.2h 138 228(0.88) Tellurium Group. Te-132 183 740(0.128) Mo-99 66.02h
' Noble Metals 155 497(0.89)
Ru-103 39.4d 115 750(0.23).1024(0.325) Sr-91 9.5h 1388(0.9) Alkaline Earths 2.71h 123 Sr-92 173 537(0.254) Ba-140 12.8d 124 934 (0.1~39) Y-92 3.54h 487(0.455).1597(0.955)
' Rare Earths 40.2h 184 La-140 161 145(0.48)
Ce-141 32.5d 134(0.108 284.3d 129 Ce-144 64.04 161 724(0.437).757(0.553) Refractories Zr-95 166 743(0.928) 2r-97 16.9h d
*0nly the representatise isotopes which have relatively large inventory an d here.
considersd.co be easy to measure are liste
**At the time of reactor shutdown.
5.7.5C-9 Rev. O 2-2 _. _. __ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - ~ _. --
NEDo-22215 Table 2 FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER AND DRYWELL GAS SPACE DURING REACTOR SMUID6WN UNDER Reactor Water, UCi/s Drywell Gas (uci/ce) i Upper Limit Nominal Upper Limit- Nominal Isotop,e 29 0.7 1-131 0.3* 0.03 - Cs-137* -5b 10 j
--- 10* -6b Xe-133- -5a 4x10 l 4x10 Kr-85 \ l
*0bserved experimentally, in an operating BWR-3 with MK I containment, data l obtained from GE unpublished document. DRT 268-DEV-0009.
Assuming 10% of the upper limit values. I
" Release of Cs-137 activity would strongly depend on the core inventory which is a. function of fuel burnup.
i I l . i i I l l t-I 2-3 5.I.5C-10 Rev. O
h-h.-e
- i. . .. NEDO-22215 l
I 10 { FUE L ME LTD0ww P : WPPER RELEASE LIMIT sEst ESTIMATE / , . .. p . LOWER RELE ASE LIMIT ' p
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7 CONCEN tm ATION ) c ~ f IN RE ACTom Watta 1 UPPER LIMIT 29 0 sci /g
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-- % CLADDING F AILURE -
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'g - % FUEL MELTDOWN [
t-Tigure 1. Relationship Between I-131 Concentration in the of CorePrimary Da/ Age Coolant in : l (Reactor Water + Pool Water) and the Extent l L Reference Plant 1 l 5.7.5C-11 Rev. 0 l 2-4 , l l l k "-
' - -~
"i NEDO-22215' e
+e .
so* , ~ FUEL MELTGG h-p UPPER RELECJ LIMIT BEST ESTIMATE
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LOWER RELE ASE LIMIT
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Tigure 2.. Relationship Between Cs-137 Concentration in the Pri6ary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant 5.7.5c-12 Rev. 0 2-5 l
, , _ _ - . . . - _ - _ . - ~ . . _ _ _ . , , _ . . _ - _ , __
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NEDo-22215 ee w*
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- FUE L MELTDOWN UFFER RELE ASE LIMIT
. SEST ESTIMATE
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'100 g
i.o ' 'g - s FUEL MELTDOWN-l Tigure 3. Relationship Between Xe-133 Concentration in the Containment Gas (Dryvell + Torus Gas) and the Extent of Core Dar, age in Reference l-Plant 2-6 j: 5.7.5c-13 Rev. O r l i
l .. -.* NED0-22215 ~ I' so? .
' FUE L ME LTOOWN l
UPPER MELEME LIMIT 4 oEST ESTIM ATE . LOWER RELEASE LIMIT f
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/- CONCENTRATION
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r NOMINAL 4 a to* aC./sc l ....,1 i e i
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% CLADDING F ALLURE t.o to too
'f - % FUEL MELTDOWN Tigure 4. Relationship Between Kr-85 Concentration in the Containment Gas (Dryvell + Torus Gas) and the Extent of Core Damage in Reference Plant 2-7 Rev. 0 5.7.5C-14
. _ - ~ _ ,
NEDO-22215 e 3f the concentration falls 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 deterudne the source of fission product release (see below). . It is recommended that both the water and gis phase samples be measured in order to reduce the uncertainty in core damage estimations.
- 2.1.3 Supplementary Data f:
In addition to the longer-lived isotopes, some shorter-lived isotope Theratiosofisotopesreleased} i concentrations may be measured in the sample. from either the fuel gap or the solten fuel are significantly different as f shown in Table 3 (see Section 3.3 for details), thus the source (fuel or gap) of release may be identif'ied. 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
- j 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.
fission products 2.7h Sr-92 (1.384 MeV) and 40h La-140 (1.597 MoreMeV) work, should b relatively easy to identify and measure from a gamma ray spectrum. however, is needed to establish the baseline concentrations for those isotopes.
~
2.2 SFICIPIC PLANT APPLICATION 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. Similar information is available for all BWRs. 5.7.5C-15 Rev. 0 ! l 2-8
~ . . _ _ . _ ,
- ... .. NEDo-22215 Table 3 RATIOS OF ISOTOPES IN CORE INVEjiiTORY AND TUEL GAP Activity Ratio
- in
. Activity Ritio* in Fu&l Cap _
Core Inventory Balf-Life Isotope __ 0.0234 0.233 Kr-87 76.3 a 0.0495 - 0.33. Kr-88 . 2. 84h 0.023 4.48h 0.122 Kr-85m 1.0* 5.25d-1.0* Xe-133 0.155 I-134 52.6 s 2.3 0.127
\
1.46 1-132 2.3 h 0.364 1.97 1-135 6.61h 0.685
- 2.09 1-133 20.8 h 1.0*
8.04d 1.0* I-131
"
- E** I ' # * * *" * *" * # ' * ' "- for noble gases
=
- Ratio Xe-133 concentration Iodine isotope _ concentration for iodines 1-131 concentration 2-9 Rev. 0 5.7.5C-16 .
- 1. .o NEDO-22215 Table &
I Pl. ANT PARAMETERS Primary Coolant
- Containment Gas
- Drfwell Torus /
Reactor Reactor Containment Water Tuppression Gas Type / Rated Containment Power Mass Fool Water Volume Cas Volume " Design _ (MWt). ~(10 8 g) (109 g) __ (109 cc) (309 ee) Plant 3.67 7.77 32.5 BWR 6/III 3579 2.46 Standard 4.65 3.46 2436 2.14 2.48 Brunswick-1/2- BWR 4 /I 1.93 3.68 2.69 BWR 4 /I 1775 1.76 Chiashan-1/2 6.91 32.43 I 2894 2.04 3.14 Cofrentes BWR 6*III 2.48 3.75 3.03 BWR 4/I 2380 2.00 Cooper -4.48 3.30 2527 2.61 3.18 Dresden-2/3 BWR 3/I - 1.67 2.67 2.67 BWR 4/I 1593 1.45 Duane Arnold 4.64 3.71 3293 2.77 3.23 Termi-2 Inm 4 /I 3.20 2.14 3.00 4.37 , 2436 Fitzpatrick BWR 4 /I 5.75 4.08 3323 2.74 3.17 Hanford-2 BWR 5/II 2.47 4.07 3.20 BWR 4/I 2436 2.00 Eatch-1 4.12 3.11 2436 2.00 2.47 Match-2 BWR 4 /I 3.34 4.79 3.78 BWR I/I 3293 2.93
. Hope Creek-1/2 6.74 40.50 2894 2.04 3.74 Kuo aheng-1/2 BWR 6/III 4.23 !
2.93 3.63 6.66 BWR 4/II 3293 Limerick-1/2 4.16 3.06 2011 2.05 2.73 Millstone-1 3WR'3/I 2.76 1.75 1.93 3.80 BWR 3/2 1670 Monticello 5.10 3.33 1850 2.17 2.34 NMP-1 BWR 2/I 2.32 5.10 3.85 BWR 2/I 1933 2.05 Oyster Creek 4.98 3.62 3293 2. 6,7 3.48 Peach Bottom-2/3 BWR 4/I 4.16 3.18 1998 2.05 2.38 Pilgrim BWR 3/I 3.60 6.79 4.36 3293 2.92
'Susquehanna-1/2 BWR 4 /11 3.15 1.77 1.93 3.79 BWR 4/I 1593 Vermont Yankee
. = Reactor Water + Suppression Pool Water Total Primary Coolant Mass Total Containment Cas Volume = Drywell Gas + Torus (or Primary Containment in Mark III gas 2-10 5.7 5c-17 Rev. 0 w- *-s-- w -----w- , ,,__u _____ - , _ , - _ , _ - - ..,.--m,,,, , , , , , , , - , - . _ _ , _ . _ .
MEDo-22215 2.2.2 Procedure The extent of core damage in an operating BWR can be determined by comparing the measured concentrations of major fission products in either the ,, gas or water samples, after appropriate normalisation, with the reference plant data. The following procedure is reconnended.
~
.1. Obtain the samples from the post accident sampling system, and the in gas is cencentration of a fission product 1 (Cyg in water or C g determined).
'2.
Correct t.he measured concentration'for decay to the time of reacl , shutdown. f l
- 3. Correct the measured gaseous activity concentration for temperature
- l and pressure difference in the sample vial and the containment !
(torus) gas phase (see footnote on Page 2-12). *l l e t ! 4. Calculate the fission product inventory correction factor i (Section 2.2.2.2). l
- 5. Calculate the plant parameter correction factor (section 2.2.2.3).
I
- 6. By using the correction factors, calculate the normalized concen-tration.Cf,*g or C '3' (Section 2.2.2.1).
I of fuel or cladding I a 7. L'se Tigures 3 through 4 to estimate the extent damage. , 2.2.2.1 Comparison with Reference Plant Data The extent of core damage can be estimated from the measured fission l s, as described for
. product concentrations in either the gas or water samp e the reference plant. However, the measured concentration must be corrected i .i 5.7.5C-18 Rev. 0 2-11 a_ _ ._- _--_. __ _____ __.
c-E EC-22215 for the dif ferences in operation power level, time of operation, primary coolant mass and containment gas volume.- r A
. 0-I C,*g = C,ge
- xF 3 xT y C* =C gg e xT yg x F, where \
(uCi/g) I C ,' = concentration of isotope i in the reference plant coolant gas C
= concentration of isotope i in the reference plant containment (uCi/cc)
Cyg
= measured concentratien of isotope i in the y erating coolant at tin.s. t (uci/s)~
C
= peasured concentration of isotope i in the operating containment gg gas at time, t (uCi/cc)*
1t l e = decay correction to the time of reactor shutdown A = decay constant of isotope i (day" ) g
*The following correction for the measured concentration is needed fromif the thattem- in perature-and pressure in the sample vial (T 3 .P 3 ) are different the containment (T y Py ):
I 2T, 31 31(vial)
- P 3y T 2-12 5.7.5c-19 Rev. 0
- : 1. .. . _ _
NDO-22215 t=timebetweenthereactorshusdownandthesampletime(day) r T yg
= inventory correction factor for isotope 1 (see Section 2.2.2.2) ,*.
T, = containment gas volume correction f actor (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 Tactor k y , Inventory in reference plant 11 Inventory in operating plant
- a 3651 (1-e A) 0" ,
Z g 3
'j T -A T)) e-1 7)
. $'(1-e .
where P g = steady reactor power operated in period j (Wt)* T) = duration of operating period j (day)* 0
- 7) = time between the end of operating period j and time of the lastl reactor shutdown (day) l Tor a particular short-lived isotope, 1, a calculation for only a peried of % half-lives of reactor operation time before reactor shutdown should be It j accurate enough. An example of calculation is illustrated in Appendix A.
l should be pointed out that the computer calculation of core inventory takes into account the fuel burnup, plutonium fission and neutron capture reactors.
*In each period, the variation of steady power should be lir.ited to :200.
2-13 5.7.5C-20 Rev. O
NEDO-22215 The correction factor calculated from this equation may not be entirely i accurate, but the error is insignificant in comparison to the uncertaint es fractions (Tabla 5) and other assumptions
.in the fission product release
~
(Section 3.2). . 2.2.2.3 71 ant Pacameter Cerrection Factors o
- 1;56 < J r ,- 9 P'#**'"8 '1*"*
- 1*"* **** ('} _,
I' 6 1 4
'F" = reference plant coolant mass (3.92 x 10' g)
.))I I#
, ,fpps operating plant containment gas volume (cc) 7 , 10 ,c) q , ,, o 8 reference plant containment gas volume (4 x 10 In case the fission product concentrations are measured separately for the reactor water and suppression pool water or the drywell gas and the torus or C gg veuld be averaged from the gas, the measured concentrations C,g ,
separate measurements:
'" * "'*'#'* * "'*'# **** ^ # "#* '" ##
C = Reactor water mass + pool water wi (e ne. in drywell)x(drywell gas vol)+(cone. in torus)x(torus gas vol) I C = drywell gas volume + torus gas volume gi-c i ( 5.7.5C-21 Rev. O 2-14
f" NEDo-22215 O O e en O f
! O C
== e N q em g ' b .e. ' -
- O. O. .O.
a E O. C. ' . . O O g = O O O- O
. Da O O y O - 'N N a 'u' .O 8
liw .5 5 7., 8 8 1
.l
== 3 O O O O O O O . , , -
w g , O in O t O - k O
- = g l
* . 5 .
O O O k' s O C O O Pl
> e O O
\ su O N I 9 4
- - @ l
@ 1 8 8 8 e. I e I 6 E .
O. 8 e 3 =. . O- O O st . 3 dj O 4* 8
- W @ e i
- e. 'm3! 6. u. =
E N N .O e N 4 I I g N E O. f".t N. g 3 O. 8 e - O O. O i ad. O e l O m
- b. '
Y
~c ;
N e 8
.- N l 8
I 8
=L , .-
8' tM. E. I e 3 i C. C C O
- e .
In: i 44 m e C - -
- h. b u] R .
e C N. C. C. E! EE; - e. . =. N. C C C O - g . O C C
.at O in el- - -
N C s, I,' a 6 s'. no s e- a e - N 5-'. *. g *. F. C. C. C. . .
.c C C C C g C C C E i Jl C D e r*
E c) C c m to e
- .m= .
u e N E e N O. . . C en . i
- c. E. .
C O C O m zj O e C e
- i u ul'. N O C 4 s e
# 1
- I e
> g. E , ==
N. e. O. a M p,.! . C C C sa: C O e l'
- N P
! c I i an: g* b .=
C 6
.C 8 4
- .C=
8 m .w= - I = w 8 t I
.t ' E O. . . m I C O e ' 8"i l
r s : .a ' C i g,l *
- e b 'C s
\ gf k' I C. t 4 t z I t 8
.. , C. C. .
i O C C C
* .n
- c. x
>* b =
s .E. O m b S.* C. E
.m= 6. EN c. W e C - 9
- E e n c .
em In. .c e E
- E. s.
t,, e g.c s.. s, 6 c, e un< m n - m. an eu J -. v .c t,,. a e., .c -e cz 6 en . .m i
= v. wm -m E ou g *6
-z ~
3; EE t ~q '. 5E 3 u* u
,g 5E
-u s'
- ,, y cm - v. nw:
z - .:
- a. e<
m- )
- . <- >- z- .c -
2-15/2-16 5.7.5C-22 Rev. O
* *
- NEDO-22215
- 3. TECHNICAL BASIS 3.1 FISSION PRODUCT CONCENTRATIONS IN THE PRIFARY SYST SHUTDOWN UNDER NORMAL OPERATION CONDITIONS
- 3.1.1 Fission Product Concentrations in Reactar_ Water It is well known that some volatile and water soluble fission products, mainly iodine and cesium isotopes, will be released (called Based spiking) on from defect fuel rods when the reactor is shutdown and depressurized.
Fasedag of NRC,3 the maximum I-131 release would be 10 Ci per According to theeach uCi/see t release rate during normal power operation (see Appendix B). 4 a maximum of 7000 Ci of I CE design basis of I-131 release rate at 700 uCi/sec I-131 may be released during reactor shutdown, and the concentration in reac-tor water would be 29 uCi/g. An analytical model to predict the magnitude of I-131 spiking following The , reactor shutdown in operating BWRs has been reported by Brutschy et al . d on the best estimate" concentration for 1-131 has to be calculated base analytical mode 5 for the individual reactor according to its fuel condition However, if one adopts a standard I-131 concentration of (see Appendix C). 5x10'3 uCi/g or s18 uti/sec) as proposed by ANS , the nominal 1-131 spiking This concentration f estimated to be %0.7 uti/g in the reference plant water. is consistent with an average spiking concentration observed experimentally i The results of these estimations, including (see Appendix C, Pigure C-1). the Cs-137 concentration, have been summarized in Table 2. Potential future research in this area vill be discussed in Sect 3.1.2 Noble cas Concentrations in Drvvel1 and Torus Cas Phase Similar to the spiking magnitude, the noble gas activities in the dryvell and the torus gas say vary significantly from reactor to reactor, mainly In an operating depending on the fuel condition and the steam leakage rate. BWR when the Xe-133 release rate measured at the steam jet air ejector ($ 5.7.5C-23 Rev. 0 3-1
NEDo-22215 was 1.5 x 10' WC1/sec* (compared to design basis release rate at 8200 uCi/sec), ~ uCi/cc the noble gas concentrations in the dryvell wen determined to be %10 for Xe-133 and %4 x 10'3 uCi/cc for Kr-85. These data may be considered as ' the upper limit values. 3.2 TIS $10N PRODUCT RII. EASE SOURCE TERMS UNDER ACCIDENT CON The source terms for the damaged core under accident conditions have been proposed by several investigators. '0The "best estimate" release source terms for dif ferent chemical groups of fission products are summarized in g Table 5. The release of fission products from the damage core has been estimated ' to be a function of temperature. I and time after 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 fraction of cora damage as suggested - by Malinauskas, et. al.I It is further assumed that the core is homogeneous The fuel cladding so that each~ fuel rod has an identical exposure history. rupture has been assumed to occur over the temperature range from about 780 to 1100*C.I and the entire fission product noble gas inventory in the fuel gap All other fission products in the fuel gap, which may be would be released. 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. Accc,rding to a model calculation, portions of the fuel say start to melt before the cladding is totally destroyed.
3.3 150TCPIC DISTRIBUTION IN TUIL 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 12-17 A comparison of ' experimental data reported by several investigators.
. The fission Data obtained from GE unpublished document. DRT 268-DEV00009.
product release pattern was found to be mostly " recoil." Rev. 0 J 5.7.5C-24 3-2
^ ~"^- ~
EEDO-22215 isotopic distributions in the total fuel inventory and the predicted 3. Thus,
. distribution for some major fission products ha,s been shown in Table by measuring the ratios of fission product activities in either the gas or ,
water samples, the source of fission product release may be semi-qqantits-tively determined (see more discussion in Section 4). l i. 3.4 ANTIFIPATID CHEMICAL BEHAVIOR OT IODINE AND COOLANT CHEM UNDER ACCIDENT CONDITIONS indicate The results of measurements of Three Mile Island-2 h (TMI-that the airborne radiciodine release was much6 lower when compared to t e Extensive investigations I noble gas activity release (by a factor of %20 ). at the Oak Ridge National Laboratory (ORNL) on the nature and quantity of fission product release from the over-heated fuel have concluded that cesium l todide Cs1 (5.P. = 1280*C) is the primary volatile species released from the O fuel at elevated temperatures. The behavior of iodine under loss-of-coolant and Campbell et. 'a1. , accident (LOCA) conditions has been evaluated by Lin For iodine at a concentration of a few ppm in aqueous solutions, the redox reactions should be more predictable and formations of anomalous or organic species should be much smaller than that at very low concentrations as If iodine is released as CsI, it generally assumed for radiciodine release. would stay in water as the l' ion in a slightly basic solution (mainly due to Cs ions which may be released as elemental Cs or Cs oxides in addition to
~
to I y is not CsI). Air oxidation or radistion-induced oxidation of I In addition to the reducing nature very likely to occur in a basic solution. of airconium and iron metals in the core, the production of hydrogen from 2r in the primary system steam reactions should make the chemistry environment f avorable to reducing reactions for iodine. There are at least three known volatile forms of iodine, y l . H10, and from I' is n t very likely in basic solu-organic fodina. The formation of 12 tions. The existence of HIO has never been chemically identified due to its The airborne species called HIO is one which behaves lov stable concentration. differently from Iy and organze iodine determined by using the iodine species 5.7.5C-25 Rev. 0 3-3
* * ** NEDo-22215 However, some convincing sampling method developed by Keller, et. al.25 0 hydrolysis, is the evidence has been given by Lin that HIO a product of 17 was initially added second volatile inorganic species in the gas pffase when 22 to water in equilibrium partitioning studies. The partition coefficier t
-increases with decreasing iodine concentration; at very low iodine <:en:entra-tisns, the total iodine partition coef ficiones have been determined to be 4000 at 21'C and %1600 at 72*C. 6 It must be pointed out that since both 17 and MIO are very reactive species, any reducing impurities in water or on or HIO to I~ and significantly construction material surfaces would reduce 12 .
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 As such organic species are limited, the results for reaction with iodine. of several experiments indicate that the yield of organic iodine decreases
- l 1.ess than 0.1% con-with increasing iodine concentration in the gas phase. 3 or version is expected when the airborne iodine concentration is I g/m l
larger. The total iodine concentration could be %3 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 todine, e.g., CH3I is readily hydrolyzed 29 in water 28 and basic solutions at higher temperatures. The half-time of L hydrolysis is %20 min in water at 100*C and %3 see at 200'C based on 8
*
- Neppolette and Robertson's data.
It is obvious that the very low release of iodine activities to the atmosphere in the THI-2 accident can be easily explained in terms of the nature of iodine released from the fuel and the subsequent stabilization in 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. f l l . 3-4 5.7.5c-26 Rev. 0 l l
J' '
NED0-22215 The chemistry in the primary coolant may be significantly changed under Mainly due to the releasp of cesium, the water pH may accident conditions.
' 21 and the water conductivity may increase from %10 US/cm increase to %20.5 - (torus water quality specification) to as high as %170 US/ca. ,
\
e
~
i i i i I L S.7.5C-27 Rev. O 3-5
.. .. NEDO-22215
- 4. DISCUSSION AND SVGGESTIONS FOR TUTURE WORK It is obvious that the uncertainty of gas' release fractions for todine and cesium are too large for an accurate calculation of the extent of core
damage. While additional experimental work in fuel gap measurements is apparently needed, the lower limit release fraction for iodine may be reevaluated by examining the iodine spiking release data from defective fuel rods following normal operation shutdowns. Although the 1-131 spiking data has been well documented previous the analytical model may be. refined to reflect more recent experimental data. The maximum spiking release of I-131 estimated by Pasedag , based on pre-1973 } data, is obviously too high, particularly when the improved fuels which are currently used in s.est of the operating BWRs are considered. The accuracy of core damage estimation may be significantly improved by Some less measuring more than iodine, cesium, and noble gas activities.
- volatile but easy to measure isotopes of Sr. Ba, I.a. and Ru may be determined in the water sample. More work, however, is needed to establish the release fractions as well as the baseline (shutdown spiking) concentrations for those isotopes.
Additionally, some chemical analysis data, such as hydrogen pro-duction from 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. As mentioned in Section 3.3, it is possible to determine the source of i fission product release by ceasuring the activity ratios of noble gases or 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 mixture of other fission products with high con-Additional work is centration because of its low gamma ray energy (81 kev). required to perfect this procedure. 5.7.5C-28 Rev. 0 4-1/4-2
NEDO-22215
- 5. RETERENCES M. A,.
Careway. " Calculation of Fission Prd" duct Inventory and Spectra --
- 1. .
RADC101 Progras," NED0-25176 (October 1980). .
" NEDE-2
- 2. L. Strickland and N. R. Cash, " Operating Plants Parameters,
- 21398, Rev. 2 (September 1978), S. A. Hucik, private communication.
. W. G. Fasedag, " Iodine Spiking in BWR and PWR Coolant Systems.
-3. '
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 Derivistion of BWR 1971 Design Basis Radioact1ve Material Source Terms," NEDO-10871. (March 1973). 5.- T. J. Brutschy, C. R. Hill, N. R. Morton, and A. J. Levine, *
" Behavior of Iodine in Reactor Water During Plant Shutdown and Startup,"
NEDD-10585 (August 1972).
- 6. An'rican Nuclear Society, "American National Standard Source Term Specification," ANSI N237-1976.
7. L. L. Bonzon and N. A. Lurie "Best-Estimate LOCA Radiation Signature" KUREG/CR-1237 (January 1980).
- 8. Reactor Safety Study. An Assessment of Accident Risks in U.S. Commerc Nuclear Power Plants, WASH-1400 Appendix V11, p. VII-13. U.S. Nuclear Regulatory Commission (October 1975)'.
- 9. A. P. Malinauskas, R. A. Lorenz, H. Albrecht and H. Wild "LVR Source Terms for Loss-of-Coolant and Core Melt Accident " Proc. CSNI Specialists Meetings on Nuclear Aerosols in Reactor Safety, held at Gatlinburg, Tenn.,
f April 15-17,1980, NUREG/CR-1724. 4
- 5.7.5C-29 Rev. 0 5-1
4 NEDO-22215
- 5. REFERINCES
- 1. H. A,. Carevay, " Calculation of Tission Prd' duct Inventory and Spectra -
i
~
' RADC101 Program " NED0-251 6 (October 1980). .
2 Z. L. Strickland and N. R. Cash, " Operating Plants Parameters " NEDE-S. A. Hucik, private communication. 21398, Rev. 2 (September 1978) 3. W. C. Pasedag " lodine 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 Radioact,1ve Material Source Terms," NED0-10871. (March 1973)..
- 5. F. J. Brutschy, C. R. Hill, N. R. Norton, and A. J. Levine, *
" Behavior of Iodine in Reactor Water During Plant Shutdown and Startup,"
NEDO-10585 (August 1972). 1
- 6. American Nuclear Society, "American National Standard Source Tarn ,
Specification," ANSI N237-1976.
- 7. L. L. Bonzon and N. A. Lurie "Best-Estimate LOCA Radiation Signature" NURIG/CR-1237 (January 1980).
4
- 8. Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants WASH-1400. Appendix Yll, 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 3pecialists Meetings on Nuclear Aerosols in Reactor Safety, held at Catlinburg, Tenn.,
April 15-17, 1980, NUREC/CR-1724 >
' 5.7.5C-29 Rev. 0 5-1 i
u ._ --.. l
** ** NEDO-22215 10.
American Nuclear Society, " Methods for Calculating the Release of Tission Products from oxide Fuels," Proppsed ANS-5.4 standard (November ?979). - \ 11. R. A. Lorenz, J. L. Collins and A. P. Malinauskas Nucl. Tech., 46,, ! 404 (1979).
- 12. Table VII 1-1, Reference 8.
- 13. G. M. Allison and H. K. Rae, "The Release of Tission Gases and Iodine f rom Defected Vo y Tuel- Elements of Dif ferent Lengths," AECL-2206 (1965).
k Tuel Rods "
- 14. W. A. Yuill, et. al., " Release of Noble Gases from UO y IN-1346 (1969). -
- 15. G. Jackson, D. Davico, M. J. Waterman, "The Effect of Neutron T1ux on at High Temperatures,"
the Tission Product Gas Imission from UOy AERE-M-1607 (July 1965).
- 16. N. V. Krasnoyarov, V. V. Kenyashov, V. 1.'Polyakov, and Yu V. Chechetkin Sov. At. Eng. (English), 38, 89 (1975).
- 17. General Electric Company unpublished data.
- 18. A. P. Hull Trans. Am Nucl. Soc., 34, 91 (1980).
- 19. A.~ D. Miller Ibid, jbi, 633 (1980).
-20. R. A. Lorenz, et. al., "Tission Pro' duct Release from Highly Irradiated LWR Tuel," NUREG/CR-0722 (ORNL/NUREG/TM-287/R2), Tebruary 1980.
- 21. C. C. Lin, " Anticipated Chemical Behavior of lodine Under LOCA conditions," NID0-25370 (January 1981).
- 22. D. O. Campbell, A. P. Malinauskas, and W. R. Stratton, Nucl. Tech. ,
53, 111 (1981). 5.7.5C-30 Rev. 0 5-2
'. e' .*
NEDo-22215 l 23. M. Kahn and J. KeIinberg, " Radiochemistry of Iodine," NAS-NS-3062, t ERDA (1977), p. 9. , 1 24 C. C. Lin, J. Inors. Nucl. Chem. 42, 1101 (1980). ,
~
- 25. J. H. Keller, et. al., "A Selective Absorbent Sampling System for Dif ferentiating Airborne Iodine Species " Proc. lith Air Cleaning Conference, CONT-700816 (1970), p. 621.
- 26. C. C. Lin " Volatility of Iodine in Aqueous Solutions," J. Inorg.
Nucl. Chem., 43, 3229 (1981).
\
27.- A. 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), A 252, .
273.
- 29. h. Adschi, et. al., J. Chem. Eng. Japan, 7, (5), 364 (1974).
t 5-3/5-4 5.7.5C-31 ' Rev. O
NED0-22215 ..- .+ Appendix A SAMPLE CALCULATION OF FISSION PRODUCT INVENTORY CORRECTI r Inventory of nuclide 1 in reference plant - 7 3g = Inventory of nuclide i in operating plant , ,,
-1095 A 3651 (1-e A)
AT 0' T E i 3)e g3
) P) (1-e where
}.
7 = steady reactor power operated in period j (Et) 3 A = decay constant of nuclide 1 (day ~) g T) = duration of operating period j (day) T O~ = time setveen the end of operating period j and time of last reactor shutdown (day) 3651 = ave. operation power (in W t) for the reference plant.
- 1095 = continuous operation time (in day) for the reference plant.
. Assuming a reactor has the following power operation history:
Averar,e Power Operation Time O Operation Ij P (Wt) Days Since Startup . (day) Feriod 60 254 1000 1A 1 - 60
--- 0 IB 61 - 70 ---
270 44 2000 2A 71 - 270
--- --- 0 28 271 - 300 14 0 3000 3 301 - 314 A-1 S.7.5c-32 Rev. O
r . _ _ _ _ _
..: ...- ' NEDO-22215 e For 1-131 (A = 0.0862 day ~I) 3651(1-e -0.0862x1095) y1(1-131) , 2000(1-e-0.0862x60),-0.0862x254 + 2000(1* 0.0862x200)
,-0.0862x44 + 3000(1-e
-0.0862x14 -0.0862x0 3651
" % + .45 + 2103 " I'7
-5 day ~I) e For Cs-137 (A = 6.29 x 10
-5
-6.29x10 x1095)
=
3651(1-e ,
-5 T1(Cs-137) '
1000(1-e
-6.29x10 x60),-6.29xio-5x254
-O x44
+ 2000(1-e
-6.29x10 x200),-6.29x10 x0
+ 3000(1-e -6.29x10 x14),-6.29x10 243.16 ,7,77
= 3.74 + 24.93 + 2.64 .
9 5.7.5C-33 Rev. 0 l A.
. . = - - . = = . . - - = = .=. . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - - - - -
. .. .* ' NEDo-22215
- Appendix B is TOTAL I-131 RE1. EASE DURING A SJIKING SEQUGCE
< lI O<
4 0 4 - 1 )
- 4
~
4 . 4 4 OO g 4 i
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~
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O o O 'g 0 OO O ! O O 4
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O 4 4 4 40 1 i .
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tem asvauw 15ttAU# 5.7.5C-34 Rev. 0 B-1/B-2
NEDo-22215 Appendix C' ESTIMATION OF MAXIMLH 10 DINE-131 r SPIKE CONCENTRATION The magnitude of iodine spiking in BWRs can be predicted by an empirical
,o relationship proposed by Brutschy, et. al.5 The data basis for the espirical relatienship 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 1-131 concentration in reactor water, 1-131 release steady state operation: source term, and the knowledge of the fission gas release characteristics. } The' fission product release from the defective fuel rods in a BWR during steady operation is empirically characterized by Ag(Ci/sec) = KY ggA i
- or
\ b Rg (fission /sec) = 7 i1 = K1g
~
vhere Rg = fission product release source term K = a dimensional constant establishing the level of release b = a dimensionless constant establishing the relative amount of l i.e., each nuclide in the mixture of similar chemical groups, i noble gases or iodine isotopes. l 5.7.5C-33 Rev. O C-1 l l l L
NED0-22215 r to' O O
\
O 3 10 - g O C o i s O w O g e l C i ! b O ? 5 u 2 - 10 l l I I 1 1o' 2 ,g3 tc ic 0 to' Fis5 ION G ASIl0 DINE.131 R ATIO Tigure C-1. Tission Gas to Iodine-131 Ratio versus Calculated Spike Magnitude (Assuming one peak, and cleanup system out of service) I C-2 5.7.5C-36 Rev. 0
' *" NIDO-22215 By plotting A g/Y igg against l gfor noble gases or iodine isotopes on a log-log paper, a straight line can be obtained r with a negative slope of b as demonstrated in Figure C-2.
Experimentally, the noble gas activity release rate (concentration in - the offgas times offgas flow rate) is measur=d at he steam jet air ejector (SJAI) sample point. The measurement of iodine activity release rate is more complicated. By assuming no iodine activity is returned from the feedwater into the reactor vessel.* The release rate can be calculated by Ag = C gW(13 + S, + 6,)
* \
where Ag = release rate to coolant from the core, UCi/sec. W = reactor water mass. Kg
~
>g = decay constant of species i, sec removal time constant, B, = reactor water cleanup (RWCU) syste
-I = /W, assu=ing 1002 efficient.
sec . which is defined as S c
' f = RWCU flow rate, Kg/see sec ~ , which is defined as 2,= cT/W B, = steam removal time constant, c = iodine carryover, defined as the ratio of the concentration of species i in condensate the concentration of species i in reactor water F = steam flow rate, Kg/sec This is true only for non-forward pumping plants.
"C-3 5.7.5C-37 Rev. 0 l.
** ** NEDo-22215 The source term ratio (Rg /R ,33g) in Figure C-1 is defined as the ratio of the noble gas source term estimated at the A of I-131 to the I-131 source and the magnitude of I-131 spik-term (see Figure C-2). By knowing R,/;ig,333 C-1, ing (maximum spiking concentration /ste.(dy state concentration) d iking can . from
.. Figur the maximum I-131 concentration in rerctor water during a shut own ,sp f be estimated.
4 k l Rev. O S.7.SC-38 C-4
.. .* -NEDO-2221*
r R, . - U l - 1 2 al.23, . . _ E
! FOa N00LE GASES E
a - 1 I l l l FOR IODINE I
~
l I N.isi
- LosM.sEc-'
l l i i I L I Tigure C-2. Plot of Log (R g) versus Log (13 ) A t C-5/C-6 5.7.5C-39 Rev. O i l
NEDo-22215 DISTRIBUTION , M/c . 832 W. E. Chamberlin 117 R. L. Cowan 855 R. M. Fairfield 277 T. A. Green 145 E. Kisa 736 D. Knecht 772 K. E. Kolb V04 C. C. Lin (10) 165 W. J. Marble 195 L. B. Nesbitt 887 ) H. D. Ongman V15 R. N. Osborne 195 W. A. Pitt 276 K. Salahi 117 M. Siegler' - 117 J. M. Skarpelas 858 E. F. Stall 858 C. E. Taylor 158 D. R. Wilkins Vol VNC Library (2) 528 NEBG Library (3) SCH TIE (5) l ? l l i l l l 1/2 5.7.5c-40 Rev. 0
o r ATTACHMENT 2 INTEGRATION OF OTHER PLANT ' PARAMETERS INTO CORE DAMAGE ESTIrdTE
/
~ ~j ::-
I l ff[/ /.IT!0n ogj y ( 5.7.5D-1 ~ Rev. O
- . ~ .-.-_ _ - . -
. . 1.0 INTEGRATION OF "0THER PLANT _ PARAMETER 5" INTO CORE DAMAGE ESTIMATE
1.1 Purpose and Scope
The purpose of this section is to address NRC Clarificiitkon 2(a)2
- regarding the ir.tegration 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 of the initial estimate of core damage based on radionuclide measure-ments using the post-accident sampling system (PASS).
Theprocedureforestimationofcoredamagebasedonradionuclide) measurements from the PASS'has been previously described and provided to those utilities which use that system 2. That procedure involves calculations of fission product inventories in the core and the release of those fission prod'ucts into the primary system under postulatedloss-of-coolant (LOCA) conditions. For that procedure, a ' r- BWR-6/238 with a Mark III containment is used as the reference plant. For that plant, plots of core damage estimate versus radio-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 measuredconcentrationsofmajorfissionproductsineithergasor water samples, after appropriate normalization, with the reference plant data. Consequently, core damage estimates can be made for each unique plant based on radionuclide measurements using the PASS. [- L f ! 1.2 Identification of Other Sionificant 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. i m 5 7.5D-2 Rev. 0 l _ _ , . . 11
~ ,-,
- For some of these parameters, correlations similar to that which is provided for the radionuclide measgrements can be developed which
-provide confirmation of the initial core damage estimate. Such correlations can be developed for the parameters of contafnment 9 radiation 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
'>- lj containment. Conteinment hydrogen levels, which are measurable by the PASS or the containment gas analyzers, provide a measure of the Ul extent of metal water reaction which, in turn, can be used to estimate the datgree of clad damage.
Z, . CD l Another significant parameter for the estimation of core damage is g
<C.' I reactor v'essel water level. This parameter is used to establish if .
there has been an interruption of adequate core cooling. 'Significant
.E l I periods of core uncovery, as evidenced by reactor vessel water level La_, readings, would be an indicator of a situation where core damage is 2
likely. Water level measurement would be particularly useful in distinguishing between bulk core damage situations caused by loss of M adequate cooling,to the entire' core and localized core damage g LL situations caused by a flow blockage in some portion of the core. 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 4 radiation monitors are downstream of the main steam isolation valves (M51Vs) and would be unavailable following vessel isolation. i 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. 5.7.5D-3 Rev. O
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 procedyre 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. 1.3 Application of Other Significant Parameters to Core Damage Estimate As noted in Section 1.1, procedures have already been developed j which provide an estimate of core damage based on radionuclide I measurements. Based on these procedures, an initial assessment of ) Mj core damage is made. Based on a clarification prcvided by the NRC, O that assessment would appear in a matrix as follows: Z> ' O I Degree of Minor Intermediate Major [Ql
%l Degradation (<10%) (10%-50%) (>50%)
ec - LW No fuel damage 1 r o 2 Cladding Failure 2 3 4 6 7 L Fuel Overheat 5 Fuel Melt 8 9 10 g ! O Lt As recommended by the NRC, there are four general classes of damage l and three degrees of damage within each of the classes except for the "no fuel damage" class. Consequently, there are a total of f For example, Category 3 l 10 possible damage assessment categories. would be descriptive of the condition where between 10 and 50 percent of the fuel cladding has failed. Note that the conditions of more than one category could exist simultaneously. Theobjectiveofthefinal core damage assessment procedure is to narrow down, to the maximum extent possible, those categories which apply to the actual in-plant situation. i 5.7.5D-4 Rev. 0 1-3
- -~..- .
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 corroborat.e ' and further refine the initial estimate. ,- For example, fission product measurement using PASS may indicate Category 4 core damage and, additionally, the potential for fuel overheat and fuel melt (i.e., Categories 5 through 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 O
monitor reading along with the correlation provided in Appendix B of Z this report would verify that a significant fission product release i g to the containment nad occurred, further verifying the initial g_
<~f, ,
assessment.
*=::
Further analysis of the PASS samples for concentrations of Ba, Sr, LA La and Ru and consideration of the relative amounts of fission 7 products released would indicate if any fuel melt had occurred. The flow diagram in Figure 1 indicates how the analysis of the other
- LL. significant parameters relates to the estimation of core damage based on radionuclide measurements. As noted earlier, Appendices A
- and B provide correlations for the determination of the degree of
- core damage based on containment hydrogen and radiation levels.
1.4 Determination of Samole Location In order to assure a representative sample which reflects the actual in-core condition, care must be taken in selecting a suitable sample location. The selection of a sample location should account for the type of event which will determine where the fission products will , concentrate. I I"4 5.7.5D-5 Rev. 0
For gas sampling, the recommended sampling locations are as follows: Event Type e Sample Location Non-Breaks (e.g.,MSIVCIosure) Suppression Pool Atmosphere g 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 jetpumpsaslongasthereissufficientreactorpressuretoprovide a sample from th e location. If there is not sufficient reactor pressuretoallowasampletobetakenfromthejetpumps,thenthe 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, 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. . t l __, FOR POW AT JN CLY t I I l 1-5 ' 5.7.5D-6 Rev. 0 t
-- . -.- - , - . _ _ . , _ , . , , - , ~ . _ _ _ , - _ _ _ _ . _ _ _ _ _ _ - _ - -
1.5 References
- 1. C. C. Lin, " Procedures for the4etermination of the Extent of Core Damage Under Accident Conditions" NED0-22215, August 1982.
\-
]
l [ l L - 4 l l l 1-6 5.7.5D-7 dev. 0 l
FIGURE 1-1 SEQUENCE OF ANALYSIS FOR QTIMATION OF COR_E DAMAGE Hydrogen Containeent Water I'S NORMAL OPERATIOR ! I'S 'S e : MINOR CLAD DN%GE ! r Analysis : Radiation Level (Confim) (Confim) (Confim) i 5 E E E l I O ' Detemine Core Damage , Optimum Estimate Sample From PASS n n lI Point
- E E E Containment Water halysis For Hydrogen Analysis y,3 Radiation y,, : Level y,,: Ba, Sr La, Ru g a : :
(Confim) (Confim) 3 (Confim) b o MAJOR CLAD DAMRGE Detemination FUEL OVERHEAT ,Yes Of Fission FUEL MELT Product Ratios
/ ,//g-
\ ' ',p ,
Eo
? CLAD DAMAGE POSSIBLE FUEL OVERIEAT y L, h(/// s NO CORE E LT s
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ify, , , l
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g e 'l xv hj]/), N s
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APPENDIX A INTEGRATION OF CONTAINMENT ATMOSPHERE . HYDROGEN MEASUREMENT INTO CORE DAMAGE EST] MATE
\%lgsie?@L9P: -
5.7.5D Rev. O
~
$UMMARY Theextentoffuel'claddamageasevide$cedbytheextentofmetal-water reaction can be estimated by determination of the hydrogen con' centration.
in the containment. That concentration is measurable by either the containment hydrogen monitor or by the post accident sampling system. 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 can use the correlation to determine the extent of clad damage.
" Correlation is based'on the following formula:
% N, = (1641) 748 (MWR)
N V , (1641) (7T5) (MWR) + (3176) I .36 x 106) I Where: N = number of fuel bundles V = containment volume, ft3 S'R = fraction of cladding in active fuel zone reacting XH 3 = concentration of hydrogen in containment atmosphere, mole 1 A-1 5.7.5D-10 pey, o
O ' t t T
- N I m lv =
i Z 1
* - l
~ "
4 2 $" a o v ,g g as - - m 2
,v i .
; Z '
-c.
[ g s2 - - m [ {
~
= a - - m ,
. y-i I
= - - n
! g 9 cs, ..
. e k @
- x j- -
x - - se
- x. .
I m - 14 J j y m- (, I a - q, 5 a g n -
- n g E, m -
- = E, s
) - - e x
,e 12 - - e s - - 4 4 -
= 3 a a a a a a a a a g Y e to se m se se se se so so see . .
u - , 16 nattat.- spann seeactions esuseoesse
. Figure A-1. Mytirogen Concentration for Mark 1/II and III Containements as a Function of Metal-Water Reac tion s .
23 O O
Note A-Analytical Assumptions e
- 1. Containment Volume = 350,000 ft* (MK I-II) 1,360,000 ft* (MK III) ,- ,
- 2. Number of bundles = 500 (MK I-!!)
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. .
-- FOR TONAil0N OF.'..Y -
L __ _ . _ . . . . . I i A3 5.7.50-12 Rev. O
e.s '- o, .. Note B-Determination of Clad Damage from Hydrogen Monitor Reading r Step 1. Obtaincontainmenthydrogenmonitorreading,[H),'in%. Step 2. Using the appropriate curve in Figure A-1, determine the metal-water reaction for the reference plant, Wref at(H). Step 3. The metal-water reaction for the actual in plant conditions (W) is determined from the following equation:
\
NE I~II
%W " (Wref) ( )I 350000) ,
748 V MK III
" " ( *ref) ( T ) ( 1,360,000 )
where N = number of bundles . V = total containment free volume, ft3 FOR IFORB:"ATION DE.Y . l [ I I i A4 S.7.SD-13 Rev. 0 l t
s
\
APPENDIX B INTEGRATION OF CONTAINMENT ATMOSPHERE RADIATION MEASUREMENT INTO CORE DAMAGE ESTIMATE n
;* ]
u ,\
\' Q*tD T yl\A.H
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.)
S.7.SD-14 Rev. O
. . . SIM4ARY AsnotedintheTaskIresponse,enkndicationoftheextentofcore damage is the centainment 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 con'ainment 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. 2 I l l { s1 Rev. O 5.7.5D-15 L
, i FI@% B-1 parvent of Pue1 Imentory 361rtprae la,the custalament D',
; 2004 Puel Inventory = lett sable 5ases s. -
b' ,'
- 384 sealm.
- 34 partiestates
, to8 - Seet w
2 *, r' < i 3.. I b' , 34 ) 6 j 6' . . 9.3 . . 0.02 e 1' e' ,
.. u .
....it EE." - h* i lo" ) ,a ih Deriv ) ) i h Hepto' ) ) i hheht ) ) i 6 6) i te' Q hmt After Shutdown (Hrs)
- .o1 g .
O!
*~~**
t Pse! Inventory Appremiante Searse and Danage satiante I . 1- ' . ' as ses..a e--
;' 300. 2006 T2D-34044, 2006 fuel damage, poteasial v ;* sore molt.
! T. 80. 806 T23 amble gases, m2 new ee.
- 38. Set 729, 3904 Unc gap attivity. tota 3 elet g
6-- 3. sessw., parsses sore sev.res. St TD, 2004 MAat.3400 gap activity, major stad fallare. i Z 1. 34 T3D, let me gap, maa. 394 else fattare. l C 43 I
- u
.36 T33, a u.ts., .,t am.e J4 3 83 I' 1 ee.el.ed ea 1f e..
allare, Sesal
.93 .914 fuel,TED, .3t* (34 eteness wneredal.
gap, stad fa13ere of 3/4 30-3 .914 Esc gay, stad failure of a few reds. 10-4 Stet esotaat release with opthing. f Salead geog ans3 nag Anw etery ret.. . I 19*8 ppper rense of aermet alsborse amble Tse l sesivity la sentatammet. 8-2 S.7.50-16 Rev. 0
l
)
i Note A-Determination of Clad Damage From Containment Radiation tionitor Reading 1,*,* The procedure for determination of fraction of fuel inventory released to the containment is as follows: Step 1: Obtaincontainmentradiationmonitorreading,[R)inRem/hr. Step 2: Determine elapsed time from plant shutdown to the containment radiationmonitorreading[t]inhours.
\
Step 3: Using Figure B-1, determine the fuel inventory release for the reference pl. ant (1]ref I" *
- Step 4: Determine the inventory release to the containment [1] using the following formula:" ,
1670 V EI3
- EI3ref IT) (237,450) Ib#0) where P = reactor power level, Wth V = total containment free volume, ft3 0 m distance of detector from reactor biological shield wali,ft.
' Assumes a wall mounted detector and no major equipment in detector line-of sight except for reactor , vessel shield wall.
j , , , . . L . B-3/B 4 5.7.$D-17 Rev. O
0 't~ ATTACHMENT D
ATTADDINT D I ' TOTAL IMDIVIDLIAL nnCrt TO OBTAIN. TnABCDORT. AMD AMA1Y2E A MIWIttJM MUMarD CF SAMPLES AT t = 1 HR ' *' ME RODY (GAletA) EXTREMETIES (gal #4A1 RACKGROUND SAleLE TOTAL BACKGROUND SAMLE TOTAL TOTALS FOR MINIMUM NUMeER TIME DOSE DOSE DOSE DOSE DOSE DOSE OF SAs#LE PROCED. (MIN.) (REM) (REMI fREM) (REM) fREM) iREM) TOTALS REF G. E. MANUAL
- 1. PROCEDURE 4.3 60. 0.4484 0.6610 1.1095 0.4484 32.152 32.60 10 ML UNDILUTEF' l D
,RYWELL ATM. ,
- 2. PROCEDURE 4.4 52. 0.4007 2.6592 3.060 0.4007 10.2403 10.642
' IODINE CART" l DRYWELL ATM
- 3. PROCEDLRE 4.5 79. 0.6032 0.2047 0.8077 'O.6032' 36.87 37.47 15 ML UNDILUTEti DEGASSED RCS
, LIQUID SAMPLE ,
- 4. PROCEDURE 4.5 103. 0.6196 0.05199 0.6715 0.6196 2.2508 15 ML DILUTED" 2.8704 DEGASSED RCS LIQUID SAMPLE, 5.. PROCEDURE 4.6 &,4.7 139. 0.8630 0.1045 0.9675 0.6630 0.1715 1.0346 10 ML DISSOLVED GAS SAMPLE FROM DILUTED RCS LIQUID SAMPLE lot NG _
*** TOTALS FOR 433. 2.935 3.681 6.616 2.935 81.685 84.617 PROCEDURES (7.22 HR) TOTAL INDIVIDUAL TOTAL INDIVIDUAL AT t = 1 HR 6 MOLE BODY DOSE J EXTREMETIES lAT t = 1 HR DOSE AT t = 1 HR TOTAL MAN-REM FOR SAMPLES TAKEN FOR 7 DAYS, POST-LOCA, SHOULD NOT EXCEED 11. MAN REM W.B.T and 141.' MAN-REM EXTREMETIEST
'** Total of all individual samples are not required to meet the dose and time criteria specified in NUREG-0737. Section II.B.3, Post Accident Sampling Capability. The dose criteria of 5. Rem whole body and 75. Rem for Extremettes, and the 3 hour time criteria, to obtain and analyze a single sample can easily be met for any sample being obtained, transported, and analyzed, above.
- - . . .}}