ML20095D324

Rev 1 to Radiation Chemistry Procedure 78.000.15, Determination of Extent of Core Damage

ML20095D324
Person / Time
Site:	Fermi
Issue date:	06/12/1984
From:	Heins J DETROIT EDISON CO.
To:
Shared Package
ML20095D308	List: ML20095D306 ML20095D324
References
78.000.15, NUDOCS 8408230301
Download: ML20095D324 (28)

Text

. . - . . . - -,- .. . . _ . . - _ . .. - .

4 1 b' Attachment 2 Radiation Chemistry Procedure 78.000.15, Rev. 1

" Determination of Extent of Core Damage" 1

4 1

P l

.4 i

(

8408230301 840816

~

PDR ADOCK 05000341 PDR E

.

• 78.000.15 3 Rsv. 1 mm Safety-Deeignation a

( .

. INFORMfl10f0Ni.Y .

ENRICO FERMI ATOMIC POWER PLANT UNIT 2 PON PROCEDURE - Radchem

~

TIT 1h: Determination of Extent of Core Damage PROCEDURE NUMBER: 78.000.15 REVISION NUMBER: I PREFARED BY: J. Heins /s/ MTg : 4_3 7.g 4 APPROVED BY J. D. Leman /s/ MTE 4-18-84

- _(Responsible Section Isad/ Delegate) ,

RECOMMENDED BY: W. E. Miller /s/ MTE 6-12-84 (Supervisor-Operational Assurance / Delegate)

• RECOMMENDED BT: R. S. Lenart /s/ MTE 6-12-84 g (0510 Chairman / Alternate) eAPPROYED BY: R. S. Lenart /s/ MTE 6-12-84 (Superintendent-Nuclear Production / Delegate)

• Signatures required for Safety-Related or Superintendent-Designated procedures._

MS - f.':~;r. '...', . . . . S.,.g,,,~,

t FjCQ 4 M- Oe>% L*Y lw"*.

ll

..L JU:!231:.34 l I

. s o EDISON cq

m. n .x

~

~

r .,

q f(,

gg W g

INFORMATION~ ONLY

• • jlg'$s**S

-M w

78.000.15 a Rev. 1 TABLE OF CONTENTS 1.0 Purpose........................................ 1

- 2.0 Discussion..................................... 1 3.0 References..................................... 3 4.0 Equipme nt Required. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l 4.1 Apparatus.................................. 4 4.2 Reagents................................... 4 5.0 Pre cautions and Limit ations . . . . . . . . . . . . . . . . . . . . 4 6.0 Prerequisities.... 1........................... 4 7.0 Procedure...................................... 4 8.0 Acceptance Criteria............................'.9 Enclosures Core Inventory of Major Pission

( Products in a Reference Plant Operated

, at 3651 MWt for Three Years . . . . . . . . . . . . . . . . Enclosure 1 i

Pission Product Concentrations in Reactor Water and Drywell Gas Space During Reactor Shutdown Under Normal Conditions... Enclosure 2 Ratios of Isotopes in Core Inventory a nd Puel Gap . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .Enclos ur e 3 Plant Parameters........................... Enclosure 4 Relationship Between I-131 Concentration in the Primary Coolant (Reactor Water +

Pool Water) and the Extent of Core Damage in Reference Plant.................. Enclosure 5 Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water +

Pool Water) and the Extent of Core Damage -

in Reference Plant . . . . . . . . . . . . . . . . . . . . . . . . . Enclosure 6 Relationship Between Xe-133 Concentration in the Containment Cas (Drywell + Torus Gas) and the Extent of Core Damage in -

( Reference F1 ant.......,..................... Enclosure 7 100/KS23/6.0 ,

061584

78.000.15

. Rsv. 1

. TABLE OF CONTENTS Enciasures (continued)

- Relationship Between Kr-85 Concentration in the Containment Gas (Drywell + Torus Gas) and the Extent of Core Damage in Refe rence Plant. . . . . . . . . . . . . . . . . . . . . . . . . Enclosure 8 Best-Estimate Fission Product Release Fractions.......................... Enclosure 9 Samples Most Representative of Core Conditions During An Accident For Estimation of Core Damage................................ Enclosure 10 X Metal-Water Reaction as Related to Containment Rydrogen Concentration..................... Enclosure 11 Containment Radiation Level as Related to I Fission Froducts Released.......................... Enclosure 12

(

Sequence of Analysis for Estimation of Core Damage.................. Enclosure 13

!(

4 y-..- - . ,,,-- - - ~ ~ -- -+- -n--e wi----*---*-+--

78.000.15 hv. 1 Peg 2 1

. 1.0 Purpose c

'( The purpose of this procedure is to provide a guide for estimating the haount and type of reactor core damsge under accident conditions.

2.0 Discussion The estiastion of core damage is based on the isotopic analysis of water and/or gas' samples from the primary system and the integration of these results with other known plant parameters. The following matrix any be used in the assessment of core damage:

Degree of Minor Intermediate Major Degradation (<10%) (10%-50%) (>50%)

No fuel damage 4 1 7 Cladding Failure 2 3 4 Fue1 0verheat 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 50 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 anximum extent possible, those categories which apply to the actual in plant situation.

l

The initial core damage assessment based on radionuclide measurement will provide one or several candidate categories which most likely

, represent the actual in plant condition. The other parameters should l then be evaluated to corroborate and further refine the initial estimate.

I For example, fission product osasurement using FASS may indicate i

Category 4 core damage and, additionally, the potential for fuel overheat and fuel amit (i.e., categories 5 through 10). Measurement of hydrogen in containment and use of the hydrogen correlation, Enclosure 11 could be used to verify that extensive clad damage had

! occurred. Use of the containment radiation monitor reading along with

the correlation provided in Enclosure 12 of this procedure would j verify that a significant fission product release to the containment had occurred, further verifying the intitel assessment.

!( .

9

--.w-r-rw-,o-e---,,>,e a m, e -,w,v mv w mmwu w w~ w w ' w w = w w-www w-m - v em' m > ww ~m e --~ w ==v mwr-wv=*'~ms, w-=r-~

78.000.15 Rav. 1 Pega 2 The isotopic estimation of core damage will be calculated by comparing the measured concentrations of major fission products in either gas or liquid samples, after appropriate normalization, with reference plant data from a BWR-6/238 with a Mark III containment. Fission product inventories in the primary system were calculated based on postulated loss of coolant accident (LOCA) conditions after three years (1095 days) of continuous operation at 3651 MWt. or 102% of rated power by using a computer code developed at Los Alamos and adapted to the GE computer system. The inventories of major fission products in the core at the time of reactor shutdown are given in Enclosure 1.

The pertinent reference and EF2 plant parameters are given below:

Reference Plant EF2 Rated Reactor Thermal Power 3579 MWt 3292 MWt Number of Fuel Bundles 748 Bundles 764 Bundles Total Primary Coolant Mass (Reactor Water plus Suppression Pool water) 3.92 x 10 9 g 3.51 x 10 9 g J

Total Containment and Drywell Gas Space Volume 4.0 x 1010cc 8.35 x 109cc 4

Gas / water samples'taken from the Post Accident Sampling system are analyzed for major fission product concentrations by gamma ray spectrometry. If the concentration of a fission product in reactor water or drywell, decay corrected to the time of reactor shutdown', is measured to be higher than the baseline concentration shown in IN ' Enclosure 1, the extent of fuel or cladding damage can be determined directly 2 rom Enclosures 5-8 based on isotopes I-131, Co-137, Xe-133, and Kr-85. Measurements of Cs-137 and Kr-85 are not very likely until the reactor has been shut down for longer than a few weeks and most of the shorter-lived isotopes have decayed.

If the concentration falls into a range where the release of the l fission product from the fuel gap or molten fuel cannot be definitely determined, additional data may be needed to determine the source of fission product release.

l l NOTE: The fuel gap fission products are assumed to be released

! upon the rupture of fuel cladding; the majority of fission products in the core will be released when the fuel is i melted at higher temperatures.

For example, if less volatile fission products such as isotopes of Sr, Ba, La, and Ru are found to have unusually high concentrations in the water sample as compared to baseline reactor water concentrations, a fuel meltdown may be assumed. The presence of 2.7hr Sr-92 (1.385MeV) and 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> La-140 (1.597MeV) will be relatively easy to identify and measure from a gamma ray spectrum.

l

-r, , , ,. ----n-,, a--,,+,------r- .,w, .,,--,,----n,,,..v,,,,,, , ,,._ -__w,,,m,emn___,-,-. -,,,,,,,,,,_n_,_ww.~ ,-,~.v,n -+ - - , - - ,-- - - - --

78.000.15

, R:v. 1 Pago 3

- In addition to the longer-lived isotopes, some shorter-lived isotope

[ concentrations may be measured in the sample. The ratios of isotopes

.( released from either the fuel gap or the molten fuel are significantly different as shown in Enclosure 3, thus the source (fuel or gap) of release may be identified.

. Samples acquired for the estination of core damage shall be taken from locations that are consistent with break case and system conditions (Enclosure 10). This will ansure the viability of results reported and provide the best isotopic estimation of core damage.

Correlations similar to those which are provided for the radionuclide eersurements can be developed which provide confirmation of the initial core damage estimate. Such correlations can be developed for the parameters of containment radiation Ievel 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 mach smaller fraction of the particulates) released from the fuel to the containmen::. Containment hydrogen levels, which are measurable by the PASS or the containment gas analyzers, provide a measure of the extent of metal watar reaction which, in turn, can be used to estimate the degree of clad damage.

(

. Another significatit parameter indicating the possibility of core damage is rasetor vessel water level. This parameter is used to establish if Enere has been an interruption of adequate core cooling.

Significcat 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 leval measurement would be particularly useful in distinguishing between bulk core damage situations caused by loss of adequate cooling to tae entire core and localized core damage situations caused by a flow blockage in some portion of the core.

Enclosure 13 indicates how the analysis of these other significant l parameters relates to the estimation of core damage based on radionuclide measurensats.

3.0 References

• 3 .1 Plant Operations Manual (PON) Procedure 78.000.14 (Post Accident Sampling and Analysis) ,

e3.2 PON Procedure 76.000.05 (,0peration of the Chemistry ND6685) -

(later)

• Denotes "Use" Refareace e g e e e

78.000.15 Rev. 1 Peg 2 4

• 3.3 POM Procedure 76.000.06 (Operation of the Chemistry ND680) 3.4 Lin, Chien C, Procedures for the Determination of Core Damage

. Under Accident Conditions, General Electric Co., 1982 4.0 Equipment Required 4.1 Apparatus - Gamma Spectroscopy system.

4.2 Reagents - None 5.0 Precautions and Limitations None 6.0 Prerequisites 6.1 Accident conditions exist and a decision has been made to take a sample by the General Supervisor of Chemistry or designee.

6.2 Specific location and additional instructions for the acquisition of samples have been given to Operations and Chemistry consistent with the break case and system conditions as described in Enclosure 10.

7.0 Procedure ,

7.1 Estimation Procedure 7.1.1 Obtain samples, consistent with Enclosure 10, from the Post Accident Sampling System per Reference 3.1.

7.1.2 Perform gamma spectroscopy (per References 3.2 and 3.3) and determine the concentration of fission products t I-131, Co-137, Xe-133, and Kr-85. (Cwi in water or Cgi  ;

in gas.)

NOTE: Measurements of Cs-137 and Kr-85 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. i l

7.1.3 If the temperature and pressure of the gas sample vial i are different from that in the containment, correct the measured gaseous activity concentration for temperature and pressure per paragraph 7.3. l I

1

• Denotes "Use" Reference

78.000.15

.. R:v. 1 F:32 5 7.1.4 Calculate the fission product inventory correction ,

{ factor fit per paragraph 7.4. )

7.1.5 Calculate the plant parameter correction factors (Fg or Fw) per paragraph 7.5.

~

7.1.6 Calculate the normalised concentration, C$t f or or Cl{f for I-131, co-137, Xe-133, and Kr-85 per paragraph 7.6 .

7.1.7 Interpretation of Clff or C5ff

1. If the normalised concentrations, C5t f or Citf .

obtained in paragraph 7.1.7 are higher than the i baseline concentrations shown in Enclosure 2, the extent of fuel or eladding damage can be determined directly from Enclosures 5-8.

2. If the normalized concentrations fall into a range where release of the fission product from the fuel O p or the molten fuel cannot be definitely dotermined, the presence of Sr, Ba, La and Ru  !

s'r.ould be established., Fission products 2.7hr. l Sr-92 (1.385, NeV) and 40hr La-140 (1.597MeV) are  ;

relatively easy to identify and measure from a gamma rey spectrue and are indicative of fuel

/ meltdown. These results should be compared to

!\ baseline reactor water concentrations.

7.2 Identification of Release Source ,

7.2.1 Determine the concentrations of the following short-lived isotopes by gamma spectroscopy:

Kr-87 I-134 4

Kr-88 I-132 Kr-85m I-135

Ke-133 1-133  ;

I-131 7.2.2 Correct the asasured fission products to the time of reactor shutdown. i 7.2.3 Calculate isotopic ratio's per paragraph 7.7.

i

_ _ _ - - . ~ . _ _ _ . . . _ _ - _ -

78.000.15 R2v. 1 Pcg2 6 7.2.4 Determine release source by comparing results obtained in paragraph 7.2.3 to ratio's supplied in Enclosure 3.

7.3 Temperature and pressure correction for gas sample vial.

=Cgg gyg,y) ,P2 Tg

- gi

'l '2 where Cgi (vial) = Sample vial isotopic concentration Cgi = Containment isotopic concentration f '

(P , T g ) = Sample vial pressure and temperature 3

(P , T ) = Containment pressure and temparature 2 2 7.4 Fission Product Inventory Correction Factor F = Inventory in reference plant 11 Anvent.ory an rr4

= 3651 (1-e )

(jIf}1-e -ATj) -11Tjj where Fj = steady' reactor power operated in period j (MWt)*

Tj = duration of operating period j (days)*

Tj=timebetweentheendofoperatingperiodjand time of reactor shutdown (days)*

11 = decay constant for a particular isotope (days -1).

For a particular short-lived isotope, i, a calculation for only a period of 6 half-lives of reactor operation time before reactor shutdown should be accurate enough. It should be pointed out that the computer calculation of core inventory takes into account the fuel burnup, plutonium fission and neutron capture reactions. The correction factor calculated from this equation may not be entirely accurate, but the error is insignificant in comparison to the uncertainties in the fission product release frections (Enclosure 9) and other assumptions.

• In each period, the variation of stesdy power should be limited to +_ 20%.

i

- . . - . - , , _ _ _ - . . .-....__-,_-m. _ , _ . , . _ _ . _ _ , _,_

78.000.15 R v. 1

,. Peg 2 7 9

7.5 Plant Parameter Correction Factors

. F ,= EF2 coolant mass (g) reference plant cooJant mass (3.92x107 g)

F5 = EF2 gas volume (ce) reference plant containment gas vol. (4x10tu ec) l In case the fission product concentrations are measured separately'for the reactor water and suppression pool water or the drywell gas and the torus gas, the measured concentrations C ,1 or C51 would be averaged from the separate measurements:

C,1 = (Cone. in Rx water)(Rz. water mass) + (Conc. in pool)(Pool water mass)

React 6r water mass + pool water mass Cgi = (Conc. in drywell)(Drywell gas vol) + (Conc. in Torus)(Torus gas vol)

Drywell gas volume + Torus ps volume 7.6 Calculation of Normalized Concentration C ,1 and Cgi C${f=C,te Y x Fgt x F, C)[f=Cgie YxFIt x F5 li( whereCyf=~concentrationofisotopeiinthereferenct plant reference plant coolant (Ci/g)

CR ef = concentration of isotops i in the reference 31 plant containment gas (Ci/cc)

C ,1 = nessured concentration of isotope i in EF2 coolant at time, t (Ci/g)

C 5i = asasured concentration of isotope i in EF2 containment gas at time, t (C1/cc)

I

! ehi" = decay correction to the time of reactor l shutdown i

l 11=decayconstantofisotope1(day) l t = time between the reactor shutdown and the sample time (day)

Fgi = inventory correction factor for isotope i Fg = containment gas volume correction factor F, = primary coolant mass correction factor l

1 78.000.15 R:v. 1 Pcg2 8 7.7 Calculation of Isotopic ratios Noble gas ratio = Noble gas isotopic concentration  ;

Xe-133 Concentration  !

Iodine ratio = Iodine isotopic concentration j

_ I-131 Concentration 7.8 Metal-Water Reaction 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.

A correlation has been develoepd which relates containment hydrogen concentration to the percent metal-water reaction for Mark I type containments. That correlation is shown in Enclosure 11*. Note A to that Enclosure indicates the major assumptions used in developing the correlation. 7.8.1 through 7.8.3 details the method to determine the extent of clad damage.

7.8.1 Obtain containment hydrogen reading, [H], in %.

7.8.2 Using the curve on page 1 of Enclosure 11. determine the metal-water reaction for the reference plant, NWref at [H].

7.8.3 The metal-water reaction for EF2 (EMW) is determined s, .

from the following equation:

V XMW = (NWref)(500N- (350,000 N = number of fuel bundles at EF2 = 764 V = total containment free volume at EF2 in cubic feet

= 292,000 7.9 Containment Radiation Level Another indicator 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. Enclosure 12 provides the

• Correlation is based on the following formula:

N gg (1641) 741r (MWR) 2 y 7yg (1641) (748) (MWR) + (3176) 1.36 x 100

78.000.15 Rsv. 1 Pag 2 9

l. 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. To apply this correlation to EF2 perform the following:

7.9.1 Obtain containment radiation monitor reading, [R] in Rem /hr.

NOTE: Reading should be taken from CHRRM D11-N443A.

. Readings taken from D11-N4435 may be in error (high or low) due to line 331-06-E-40-B13 blocking the detectors line-of-sight to the shield wall.

7.9.2 Determine elapsed time from plant shutdown to the containment radiation monitor reading [t] in hours.

7.9.3 Using Enclosure 12 determine the fuel inventory release for the reference plant [I]ref in X.

7.9.4 Determine the inventory release to the containment [1]

usiv the following formula:

[I] = [I]ref 1670 T/ V (6/D)

/ (237 ,450

'( where P = reactor power level, MWth V = total containment free volume, f t3=

292,000 D = distance of detector from reactor biological shield wall, ft. = 6.5' for D11-N443A and 6' l for D11-N4435.

7.10 Refer to Enclosure 13 for estimating core damage.

l 8.0 Acceptance criteria NA 1

)

t

. e w 9

,---~.,..ap.-- - - . ,n., n-,--,-m-w.,- ,,,,,,_r- ,--m---- s aw ,-- ,enm,-- - - - - - - ws w rw w -e-ee,,w m - m * -w

78.000.15 R2v. 1 CORE INVENTORY OF MAJOR FISSION PRODUCTS IN A REFERENCE PLANT OPERATED AT 3651 MWt FOR THREE YEARS MAJOR GAMMA RAY ENERGY INVENTORY (INTENSITY)

CHEMICAL GROUP ISOTOPE

• HALF-LIFE 106 Ci EeV (T /d)

Noble gases Er-85m 4.48h 24.6 151(0.755)

Er-85 10.72y 1.1 514(0.0043)

Er-87 76. m 47.1 403(0.494)

Er-88 2.84h 66.8 196(0.203),1530(0.109) xe-133 5.25d 202. 81(0.371) xe-135 9.09h 26.1 250(0.906)

Halogens 1-131 8.04d 96. 364(0.824) 1-132 2.29h 140 668(0.99),773(0.762) 1-133 20.8 h 201 530(0.87)

I-134 52.6 a 221 847(0.954),884(0.653) 1-135 6.59h 189 1132(0.231),1260(0.293)

Alkali Metals Cs-134 ~

2.06y 19.6 605(0.98),796(0.88)

Co-137 30.17y 12.1 662(0.85)

Cs-138 32.2 a 2990.** 463(0.267),1436(0.75)

Tellurium Group Te-132 78. h 138 228(0.88)

Noble Metals Mo-99 66.02h 183 740(0.138)

Eu-103 39.4 d 155 497(0.9)

Alkaline Earths Sr-91 9.52h 115 750(0.24)

S r-92 2.71h 123 1385(0.9)

Ba-140 12.8 d 173 537(0.238)

Rare Earths Y-92 58.6 d 118 934(0.137)

La-140 40.2 h 184 487(0.453),1597(0.953)

Ce-141 32.5 d 161 145(0.49)

Ce-144 284.4 d 129 134(0.108)

^

Refractories Zr-95 46. d 161 724(0.435),757(0.543)

Zr-97 16.8 h 166 743(0.933)

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

[ **1 hr af ter shutdown I, ,

Enclosure 1 Page 1 of I 9

- - . - , s.-, ..--, - - - - - _ . - _ _ _ , , __-.-_,.- ...,_. ,.-- ,--,, _---_ y-,.-my. -,--,.w----,-,.,,,---,-...--_,,,,-.m ,--,___..r.-.-,--_

78.000.15 R;v. 2 FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER

( AND DRYWELL GAS SPACE DURING REACTOR SHUTDOWN UNDER NOFM CONDITIONS DRYWELL GAS ISOTOPE _ REACTOR WATER, uCi/g uC1/cc

- UPPER LIMIT NOMINAL UPPER LIMIT NOMINAL I-131 29 0.7 ---

Co-137 0.3* 0.03** ---

Ze-133 -- --

10-4* 10-5

• Kr-85 --

4x10-5. 4x10-6**

[

• 0bserved experimentally, in an operating BWR-3 with MK I containment, data obtained from GE unpublished document, DRF 268-DEV-0009.

i

• • Assuming 10% of the upper limit values.

.(

Enclosure 2 Page 1 of 1 1

- - . _ _ . _ _ _ . . _ . . . - - - - - . - _ . _ _ . _ . . . _ . - _ . - - _ . , , . . - , - - . - _ _ . _ , _ _ ~ . . - - . . - . _ . - _ _ . - ,

78.000.15 R;v. 1 RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP ACTIVITY RATIO

• IN ACTIVITY RATIO

• IN ISOTOPE HALF-LIFE CORE INVENTORY FUEL CAP Er-87 76 m 0.233 0.0234 Kr-88 2.84h 0.33 0.0495 Kr-85m 4.48h 0.122 0.023 Ze-133 5.25d 1.0* 1.0*

I-134 52.6 a 2.3 0.155 I-132 2.28h 1.46 0.127 I-135 6.59h 1.97 0.364 I-133 20.8 h 2.09 0.685 I-131 8.04d 1.0* 1.0*

• Ratio . noble gas isotope concentration for noble gases Xe-133 concentration

. Iodine isotope concentration for iodines 1-131 concentration Enclosure 3 Page 1 of I l

l

78.000.15 R;v. 1 1

1

• l PLANT PARAMETERS (NEDO-2215)

(

PRIMARY COOLANT

• CONTAINMENT GAS

• TORUS /

RATED REACTOR SUPPRESSION DRYWELL CONTAINMENT REACTOR TYPE / CON- POWER . WATER MASS POOL WATER GAS VOL. GAS VOLUME PLANT TAINMENT DESIGN (MWt) (108 g) (109 g) (to9ee) (to9ee)

EP2 BUR 4 3292 2.77 3.23 4.64 3.71 MKI

(

• Total Primary Coolant Mass = Reactor Water + Suppression Pool Water Total Containment Gas Volume - Drywell Gas + Torus (or Primary Containment in MKIII) gas Rnclosure 4

( Page 1 of 1

78.000.15 R:v. 1 10 FUEL MELTDOWN UPPER RELEASE LIMI

, BEST ESTIMATE /

/

3 ,4 LOWER RELEASE LIMIT ,/  ;

/

/

To / /

h *

,' / /

S

/ / /

/ /

1. /

10 r / /

~

/ /

/ p l

/ / /

/

l E 10 2

/

,/

/

5 *

/

~

/

8 w

/

/

/ /

~ # #

j CLADDING FAILURE

/

10 t /

1. UPPER RELEASE LIMIT u */

BEST ESTIMATE s g T

(

~

/

,# LOWER RELEASE LIMIT

- /

• NORMAL SHUTDOWN

/ CONCENTRATION

/ IN REACTOR WATER 1.0 [ /

/ UPPER LIMIT

29.0 pCi/g

, ,' HOMINAL: 0.7 pCi/g

- /

/

/

/

0.1 i . e *>i - a > >>>>' a >'==" ' ^ ' ' ' ' ' ' ' ' ' ' ' ' ' "

0.1 1.0 to 100

' N 1 CLADDING FAILURE I

' 1.0 11) 100 p I FUEL MELTDOWN y Relationship Between 1-131 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant Enclosure 5 Page 1 of 1

_=___, . _ -_.

78.000.15

, R*V* 1

. lo' , -

l FUBL NRLTDOWN k' WPPER RELEASE LIMIT l

/

, BEST ISTIMATE /

r- LOWER RELEASE L }T j

. / l

/v

/ /

/ / /

- / /

h- 10 / j / '

p /

}  !

/

/ ,' /

,' / s' / /

l 10 e p

/

/

/

/

/

E  ;

/ /  :

3  : / /

m *

/ /

N  ! / CLADDINC FAILURE

,, j

.E e

( -

1.0 / , UFFER RELEASE L MIT I /

f B st ESTIMATE .

[

j,! LOWER RELEASE LIMIT

% / NORMAL SNUTDOWN , ,

C / CONCENTRATION O.1 r- y IN REACTOR WATER

/ UFFER LIMIT: 0.3 pCL/s y  :

" . NOMINAL: 0.03 pCi/s '

7

/

,/

10

-2 / , , , , ,,,,l , , , , , , , ,1 , j , , , , , ,1 . , ,,,,,a , , , . . . . .

c.1 1.0 to 100 p I CLAD 0 DIG FAILURE N o 10 100 h I FUIL M LTDOWN y Relationship Between Co-137 Concentration in the Primary coolant (Reactor Water + Pool Water) and the Extent of Core Damage in

. Reference Plant k Enclosure 6 Page 1 of 1 i

78.000.15

. Rev. 1 10' l FUEL MELTDOWN

• /

UPPER MLEASE LIMIT /

, ,. i BEST ESTINATE / /

3 /

I LOWER M LEASE LIMIT /

f

~

/,

{/ / '

f/ l

, // ,

y /s /

g 10 ,_ /

~

/ges'/

// / ,

b ~

~

/ ,' /

/ / /

h -

/ /

/

10 j /

/ /

p /

- / ~

CLADDING FAILURE

, /

P / / UPPER M LEASE LIMIT 1.0 / /

U / BEST ESTIMATE

> I /

/ LOWER MLEASE LIMIT j

I "/ / /

NORMAL OPERATING CONCENTRATION m 0.1 / 1N DRYWELL D  ?'/

UPP:R LIMIT:

j r 10'-5 pCi/cc pCi/cc

, NOMINAL: 10 A 0 0 AAa A II' A A A A AAAA A A A A AAAA I I

$.AJ AAA A I 0 I A 0.1 1.0 40 100 g I CLADDING TAILURE 1.0 13 100 W X TUEL MELTDOWN --->l Relationship Between Xe-133 Concentration in the containment Gas Drywell + Torus Gas) and the Extent of Core Damage in Reference Plant Enclosure 7 Page 1 of 1

78.000.15

, Rcv. I 10 .

.. ,' FUEL MELTDOWN .

( _

,' UPPER RELEASE LIMIT BEST ESTIMATE /

l 10 LOWER RELEASE LIMIT ,/ ,

E

. /

5 / '

~

~

/ /

// /

/ /

s/

/,/ s 1.n _ / //

/ / /

n  ; // /

0 /

5

~

,4 9

//

/

/

m f

-l / /

5 10 ~ /

/

5 /

l E

~

. /

/

,/

/

/

/

~

/ / CLADDING FAILURE

~

~

E 10 - / / UPPER RELEASE LIMIT i /

g /

/ BEST ESTIMATE

,/ /

1. LOWER RELEASE LIMIT p

l e

10

~

f f

! NORMAL OPERATION CONCENTRATION IN DRYWELL i .' / -

E  ; / UPPER LIMIT

4x104 pCi/cc

,/ NOMINAL: 4x10 pCi/cc

-4 10 ,,

, ,,m..I . , , ,,,,,1 ,,,,,,,,1 , , , ,,,,,i , , , , , ,

0.1 1.0 10 100 l

i b I CL4DDING PAILURE O P 9 1.0 10 100 I FUEL MELTDOWN %

l L

Relationship Between Kr-85 Concentration in the Containment Gas

! (Drywell + Torus Gaa) and the Extent of Core Damage in Reference l(

~

Plant l\ Enclosure 8 Page 1 of 1

BEST-ESTIMATE FISSION PRODUCT RELEASE FRACTIONS Cap Release Meltdown Release Oxidation Release Vaporization Release Lower Upper Lower Upper Lower Uppe r Lower Upper Nominal Limit Limit Nominal Limit Limit Nominal Limit Limit Nominal Limit -Limit Noble Cases 0.030 0.010 0.12 0.873 0.485 0.970 0.087 0.078 0.097 0.010 0.010 0.010 (Xe,Kr)

Halogens 0.017 0.001 0.20 0.885 0.492 0.983 0.088 0.078 0.098 0.010 0.010 0.010 (I,HR)

Alkali Metals 0.050 0.004 0.30 0.760 0.380 0.855 -- --- --

0.150 0.190 0.190 (Cs,Rb)

Tellurium 0.0001 3x10-7 0.04 0.150 0.05 0.250 0.510 0.340 0.680 0.340 0.340 0.340 Croup (Te,Se,Sb)

Noble Metals - - --

0.030 0.01 0.10 0.873 0.776 0.970 0.005 0.001 0.024 (Ru,Rh,Pd,Mo,Tc)

Alkaline lx10-6 3x10-9 0.0004 0.100 0.02 0.20 ---

0.009 0.002 0.045 Earths (Sr,Ba)

! Rare Earths -- --

0.003 0.001 0.01 --

0.010 0.002 0.050

, (Y,La Ce,Nd, Pr,Eu,Pa,Se, l Np,Pu)

! Refractories -- --

0.003 0.001 0.01 --- -- -- -- ---

gg (Zr,Nb)

! *E i

~E EM R2 *

.- o

-. . ~8 h

__ __ _ ___ m --__ _ _ _ - - - __

l SMIPLES MDST REPRESENTATIVE OF CORE CDlIDITIONS DURING All ACCIDENT i

POR THE ESTIMAT10ll 0F CDRE DMIAGE i

t

• l l Break Category / System Conditions Sample location Other Instructions l Supp. Supp.

l Jet Pool Fool l , l Pump Liquid Atmos. RHR Drywe11l I -

Sus 11 Liquid Line Break, Reactor Power >1% Yes YesI Yes2 Sas11 Liquid Line Break, Reactor Power <1% YesI Yes Yes2 A. RRR aust be in shutdown cooling j mode.

j B. Reactor water level must be raised and flow from moisture

! separators.

I i San 11 Steam Line Break, Reactor i Power >1% Yes YesI Yes2

Saml1 Steam Line Break. Reactor i Power <1% YesI Yes Yes2 A. RHR aust be in shutdown cooling mode.

B. Reactor water level mast be raised and flow from moisture separators.

i Large Liquid Line Break, Reactor

! Power >1% Yes3 Yes4 Yes! Yes2 A. Suppression pool asst be in suppression cooling mode.

Large Liquid Line Break, Reactor

,, Power (1% Yes' Yesl Yes3 Yes2 A. RHR aust be in shutdown cooling j gg ww mode.

B. Stappression pool aust be in l ~h suppression cooling mode. 7%

i gg C. Reactor water level must be 4*

o

, ,, raised and flow from moisture 8 o

s epa rato rs. g

/

I SAMPLES MDST REPRESENTATIVE OF CORE CONDITIONS DURING AN ACCIDENT POR THE ESTIMATION OF CORE DAMAGE Break Category /Systea Conditions Sample Location Other Instructions Supp. Supp.

Jet Pool Pool l Pump Liquid Atmos. RHR Drywelll Large Steam Line Break, Reactor Power >l% Yes3 Yes4 Yes

Large Steam Line Break, Reactor Power <1% --

YesI Yes Yes2 A.

' RRR aust be in shutdown cooling mode.

l B. Reactor water level must be raised and flow from motsture s eparators.

1. Use if SRV's are vented to the suppression pool.

2. Use if SRV's are not vented to suppression pool.

j 3. Use if makeup water flow is <50% of core flow present.

4. Use if makeup water flow is >50% of core flow present.

i i

EU l 2SO g"

e R  ? 'o yw O o wO L

l w l

l i

m s

• a 68 se a

5# -

8 60 .

8 56 a .

s .

n

• 52 .

m.

46 .

'o

) N 44 -

l M

~

j mu 40 .

5 i

m 36 -

a J

j 32 -

mD 26 -

. v E 24 .

8 l g 20 .

n i 16 .

12 .

, 8 -

mm 4 '

i =~

. A. -R . *.

i 2 0 i a = = = = = = =

_f ." o

, oS m.

0 10 20 30 40 50 60 70 PO 90 100 .8

! w .-- 1 METAL-WATER REACTION vi

~

i Hydrogen Conccentration for a Mark. I Containment as a Function of Metal-Water Reaction

78.000.15 Rcv. I  !

Note A-Analytical Assumptions

1. Containment Volume = 350,000 ft3 (gg I_II)

2. Number of bundles = 500 (MK I-II)

3. Fuel type = 8 x 8 R

k. 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 Enclosure 11 Page 2 of 2

78.000.15

.. Ecv. 1 Fercent of Fuel Inventory Airborne in the Containment t 1001 Fuel Invent.ry - 1001 Nome Cases g .y + SSE ledlae

. . II ..o.1 ..

a' N

f M. .

g .

I a. ,,

] a'. I

..n 10 =

.. 13 j a' ...cu 1.

_g, p, , , , , , , ,

Time After Shutdown (Hrs) 2 Fuel Inventory Approximate Source and Damage Estimate Released

-[. 100. 100% TID-14844, 100% fuel damage, potential core melt.

50. 50% TID noble gases, 'IMI source.

10. - 10% TID, 100% NRC gap activity, total clad failure, partial core uncovered,

3. 3% TID, 100% WASH-1400 gap activity, major clad failure.

1. 1% TID, 10% NRC gap, Max. 10% clad failure.

t

.1 .1% TID, 1% NRC gap, 1% clad failure, local heating of 5-10 fuel assemblies.

.01 .01% TID, .1% NRC gap, clad failure of 3/4 fuel element (36 rods).

10-3 .01% NRC gap, clad failure of a few rods.

1d 100% coolant release with spiking.

5x10-6 100% coolant inventory release.

i 10-6 Upper range of normal airborne noble gas activity in containment. .

I '

Enclosure 12 Page 1 of 1 l

SEQUENCE OF ANALYSIS FOR FSTIMATION OF CORE DAMACE t

ir gr

, Hydrogen Containment Water NORMAL OPERATION s

Analysis Yes s Radiation Yes s Level MINOR CLAD DAMACE (Confirm) (ConfIra) '

(Confirm)

R a

E E E Determine Cote Damage Optimum s Estimate I if Sample From PASS -

Point

^ ^ ^

o o o -

z z z S

'y Hydrogen Containment Water Analysis For Analysis Yes9 Radiation ._Yes s Level Yes 's s

Ba, Sr, La, Ru (Confirm) (Confirm) (Confira) 2

/\ pg

\r i MAJOR CLAD DAMAGE Determination i FUEL OVERHEAT g Of Fisston ,

i FUEL MELT r uc attos

! m s'

= n V g o.

%7 CLAD DAMAGE Qg

! E POSSIBLE FUEL *

,o oy OVERHEAT *g

"' y NO CORE MELT ww l

i 5%-