ML20215H226

Rev 2 to Post-Accident Core Damage Assessment Methodology

ML20215H226
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Site:	Beaver Valley
Issue date:	11/30/1984
From:	Lipke E WESTINGHOUSE OPERATING PLANTS OWNERS GROUP
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r WESTINGHOUSE WNER'S GROUP POST ACCIDENT CORE DAMAGE ASSESSMENT METHODOLOGY (CDAM) l MODIFIED FOR BVPS l

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WESTINGHOUSE OWNERS GROUP (WOG)

POST ACCIDENT CORE DAMAGE ASSESSMENT METHODOLOGY

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l Developed by the Core Damage Assessment Working Group of the WOG, Dr. E. J. Lipke (NEP) Chiarman (S.0. MUHN-2049) l 1

Fcbruary 1984, Rev. O March 1984, Rev. 1 November 1984, Rev. 2 l

4255g:12

TABLE OF CONTENTS Page

1.0 INTRODUCTION

AND PURPOSE 1

1.1 Methodology 1

2.0 TECHNICAL BASIS FOR CORE DAMAGE ASSESSMENT METHODOLOGY 3

2.1 Characteristic Fission Products 3

2.2 Core Inventories 4

2.3 Power Correction for Core Inventories 4

2.3.1 Power Correction Factor 7

2.4 Relationship of Clad Damage With Activity 10 2.4.1 Gap Inventory 10 2.4.2 Spiking Phenomena 10 2.4.3 Activity Associated With Clad Damage 13 2.4.4 Gap Activity Ratios 26 2.4.5 Adjustments to Determine Activity Released 26 2.5 Relationship of Fission Product Release With Overtemperature Conditions 37 2.6 Relationship of Nuclide Release With Core Melt Conditions 40 2.7 Sampling locations 43 3.0 AUXILIARY INDICATORS 49 3.1 Containment Hydrogen Concentration 49 3.2 Core Exit Temperatures and Reactor vessel Water Levels 53 3.3 Containment Radiation Monitors and Core Damage 56 4.0 GENERALIZED CORE DAMAGE ASSESSMENT APPROACH 62 5.0 LIMITATIONS 64 6.0 EXAMPLE OF CORE 0AMAGE ASSESSMENT METHODOLOGY 66

7.0 REFERENCES

90 m

LIST OF TABLES Table Title h

2-1 Selected Nuclides fer Core Damage Assessment 5

2-2 Fuel Pellet Inventory for Westinghoisse Plants 6

2-3 Gap Inventory 11 2-3-1 Gap Inventory Minimum and Maximum 12 2-4 Expected Iodine Spike 14 2-5 Normal Operating Activity 24 2-6 Isotopic Activity Ratios of Fuel Pellet and Gap 2) 2-7 Parent-0aughter Relationships 33 2-8 Source Inventory of Related Parent Nuclides 36 2-9 Expected Fuel Damage Correlation with Fuel Rod Temperature 38 2-10 Percent Activity Release for 10d Percent Overtemperature 39 Conditions 2-11 Percent Activity Release for 100 Percent Core Melt 44 Conditions l

2-12 Suggested Sampling Locations 48 3-1 Average Containment Volume and Zirconium Mass 52 3-2 Instantaneous Gamma Ray Source Strengths Due to a 100 57 Percent Release of Noble Gases at Various Times Following an Accident

LIST OF TABLES (continued)

Table Title Page 3-2A Instantaneous Gamma Ray Fluxes Due to 100% Release of 58 Noble Gases at Various Times Following an Accident 4-1 Characteristics of categories of Fuel Damage 63 6-1 Retults of Sampling Analysis Taken 6 Hours Af ter Reactor 67 Shutdown 6-2A Source Inventory 69 6-28 Decay Corrected Specific Activities of Sampling Analysis 72 6-3 Adjusted Specific Activity Due to Pressure and Temperature 74 Differences (Containment Atmosphere)

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6-4 Adjusted Specific Activity Due to Pressure and Temperature

- 75 Differences (Sump) 6-5 Adjusted Specific Activity Due to Pressure and Temperature 76 Differences (RCS) 6-6 Containment Atmosphere Activity 77 6-7 Containment Sump Activity 78 6-8 RCS Activity 79 6-9 Total Activity Released 81 6-10 Activity Ratios of Released Fission Products 83 9

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LIST OF TABLES (continued)

Table Title

'Page 6-11 Fission Product Inventory at Reactor Shutdown 84

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6-12 Release Percentage 87 e

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LIST OF FIGURES Fiqure Title Pace 2-1

_ Power Correction Factor for Cs-134 Based on Average Power 8

During Operation 2-2 9elationship of % Clad Damage with % Core Inventory 15 Released of Xe-133 2-3 Relationship of % Clad Damage with % Core Inventory 16 Released of I-131 2-4 Relationship of % Clad Damage with % Core Inventory 17 Released of I-131 with Spiking 2-5 Relationship of % Clad Damage with % Core Inventory 18 Released of Kr-87 2-6 Relationship of % Clad Damage with % Core Inventory 19 Released of Xe-131m 1

2-7 Relationship of % Clad Damage with % Core Inventory 20 Released of I-1~52 2-8 Relationship of X Clad Damage with % Core. Inventory 21 Released-of I-133 2-9 Relationship of % Clad Damage with % Core Inventory 22

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LIST OF FIGURES (continued)

Fiqure Title Page 2-12 Relationship of % Fuel Overtemperature with % Core 42 Inventory Released of Ba or Sr 2-13 Relationship of % Fuel Melt with % Core Inventory Released 45 of Xe, Kr I, Cs, or Te 2-14 Relationship of % Fuel Melt with % Core Inventory Released 46 of Ba or Sr 2-15 Relationship of % Fuel Melt with % Core Inventory Released 47 of Pr 3-1 Containment Hydrogen Concentration Based on Zirconium 50 Water Reaction 3-2 Distribution of Thermocouples and Flux Thimbles 4-Loop 54

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Plant 3-3 Percent Noble Gases in Containment for Containment Volume 60 6

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1.0 INTRODUCTION

AND PURPOSE In March 1982 the NRC issued a " Post Accident Sampling Guide for Preparation of a Procedure to Estimate Core Damage" as a supplement to the post accident sampling criteria, of NUREG-0737(II. The stated purpose of this guide was to aid utilities in preparation of a methodology for relating post accident core damage with measurements of radionuclide concentrations. The primary interest of the NRC was, in the event of an accident, to have some means of realistically differentiating between four major fuel conditions: no damage, cladding failure, fuel overheating, and core melt. The methodology developed is intended to enable qualified personnel to provide an estimate of this damage.

In order to comply with the NRC request for such a methodology, Westinghouse, under contract to the Westinghouse Owners Group (WOG), prepared the following generic technical report.

This report is cognizant of NRC's initial intention. Additionally, the report reflects input by NRC and various representatives of the WOG provided during several meetings held on this subject during the past year.

This report has been arranged to present the technical basis for the methodology (Section 1 through 5), and to provide a step-by-step example, which can be made applicable to various sizes and types of Westinghouse pressurized water reactors (Section 6).

1.1 METHODOLOGY The approach utilized in this methodology of core damage assessment is measurement of fission product concentrations in the primary coolant system, and containment when applicable, obtained with the post accident sampling system. Greater release of fission products into the primary coolant can occur if insufficient cooling is supplied to the fuel elements. Those fission products contained in the fuel pellet - fuel cladding interstices are presumed to be completely released upon failure of cladding. Additional fission products from the fuel pellet are assumed to be released during overtemperature and fuel melt conditions.

These radionuclide measurements, l

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together with auxiliary readings of core exit thermocouple temperatures, water level within the pressure vessel, containment radiation monitors, and hydrogen production are used to develop an estimate of the kind and extent of fuel damage.

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2.0 TECHNICAL BASIS FOR CORE DAMAGE ASSESSMENT METHODOLOGY 2.1 CHARACTERISTIC FISSION PRODUCTS Depending on the extent of core damage, characteristic fission products are expected to be released from the core. An evaluation was conducted to select the fission product isotopes which characterize a mechanism of release relative to the extent of core damage. Nuclides were selected to be associated with the core damage states of clad damage, fuel overheat, and fuel melt. The selection of nuclides for this methodology was based on half-life, energy, yield, release characteristics, quantity present in the core, and practicality of measurement using standard gamma spectrometry techniques.

The nuclides selected for this methodology have suf ficient core inventories and radioactive half-lives to ensure that there will be sufficient activity for detection and analysis of the nuclides for some time following an accident. Most of the nuclides selected have half-lives which enable them to -

reach equilibrium quickly within the fuel cycle. The list of selected nuclides contains nuclides with half-lives of I day or less which are assumed to reach equilibrium in approximately 4 days. These nuclides are used to assess core damage for cores that have been operational in a given cycle for less than a month. For cores that have been operating for more than a month, the list contains nuclides with half-lives greater than 1 day which reach equilibrium at some time during the first month of operation depending on the half life of the nuclide. Both groups of nuclides are used to assess core i

damage for cores that have been operational in a given cycle for more than a month. Other factors considered during the selection process were the energy and yield of the nuclides along with the practicality of detecting and analyzing the nuclides.

Nuclides were chosen based on their release characteristics to be representative of the specific states of core damage. The Rogovin Report noted that as the core progressed through the damage states certain nuclides associated with each damage state would be released. The volatility of the nuclides is the basis for the relationship between certain nuclides and a particular core damage state.

3

A list of the selected nuclides for this core damage assessment methodology is shown in Table 2-1.

2.2 CORE INVENTORIES Implementation of the core damage assessment methodology requires an estimation of the fission product source inventory available for release. The fission product source inventory of the fuel pellet was calculated using the ORIGEN computer code, based on a three-region equilibrium cycle core at end-of-life. The three regions were assumed to have operated for 300, 600, and 900 effective full power days, respectively. For use in this methodology the fission product inventory is assumed to be evenly distributed throughout the core. As such, the fission product inventory can be applicable to other equilibrium cores with different regional characteristics. The fuel pellet inventory of the selected fission products and some additional fission products of interest is shown in Table 2-2.

2.3 POWER CORRECTION FOR CORE INVENTORIES i

The source inventory shown in Table 2-2 presents inventories for an I

equilibrium, end-of-life core that has been operated at 100 percent power.

For this methodology a source inventory at the time of an accident that accounts for the power history is needed. For those cases where the core has reached equilibrium, a ratio of the steady state power level to the rated power level is applied. Within the accuracy of this methodology, a period of four half-lives of a nuclide is sufficient to assume equilibrium for that nuclide. For nuclides with half-lives less than one day the power ratio based on the steady-state power level of the prior four days to reactor shutdown can be used to determine the inventory. To use a simple power ratio to determine the inventories of the isotopes with half-lives greater than i day, the core I

should have operated at a constant power for at least 30 days prior to reactor shutdown. The assumption is made that constant power exists when the power level does not vary more than 110 percent of the rated power level f rom the time averaged value. For transient power histories where a steady state power condition has not been obtained, a power correction f actor has been developed to calculate the source inventory at the time of the accident.

TABLE 2-1 SELECTED NUCLIOES FOR CORE OAMAGE ASSESSMENT Core Damage State Nuclide Half-Life

• Predominant Gammas (Kev) Yield (%1*

Clad Failure Kr-85m**

4.4 h 150(74), 305(13)

Kr-87 76 m 403(84),2570(35)

Kr-88**

2.8 h 191(35), 850(23), 2400(35)

Xe-131m 11.8 d 164(2)

Xe-133 5.27 d 8)(37)

Xe-133m**

2.26 d 233 (14)

Xe-135**

9.14 h 250(91) 1-131 8.05 d 364(82) 1-132 2.26 h 773(89),955(22),1400(14)

I-133 20.3 h 530(90)

I-135 6.68 h 1140(37),1280(34),1460(12),1720(19)

Rb-88 17.8 m 898(13),1863(21)

Fuel Overheat Cs-134 2 yr 605(98), 796(99)

Cs-137 30 yr 662(85)

Te-129 68.7 m 455(15)

Te-132 77.7 h 230(90)

Fuel Melt Sr-89 52.7 d (beta emitter)

Sr-90**

28 yr (beta emitter)

Ba-140 12.8 d 537(34)

La-140 40.22 h 487(40), 815(19), 1596(96)

La-142 92.5 m 650(48), 1910(9), 2410(15), 2550(11)

Pr-144 17.27 m 695(1.5)

Values obtained f rom rable of Isotooes Lederer, Hollander, and Perlman, Sixth Edition.

• • These nuclides are marginal with respect to selection criteria for candidate nucildes; they have been included on the possibility that they may be detected and thus utilized in a manner analogous to the candidate nuclides.

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TABLE 2-2 FUEL PELLET INVENTORY

• Inventory, Curies 3-Loop Nuclide 2652 Mwt)

Kr 85m 1.6(7)

Kr 87 3.0(7)

Kr 88 4.2(7)

Xe 131m 4.7(7)

Xe 133 1.4(8)

Xe 133m 2.l(7)

Xe 135 2.7(7)

I 131 7.3(7)

I 132 1.1(8)

I 133 1.4(8)

I 135 1.3(8)

Rb 88 4.2(7)

Cs 134 1.7(7)

Cs 137 7.9(6)

Te 129 2.4(7)

Te 132 1.1(7)

Sr 89 5.9(7)

Sr 90 5.4(6)

Ba 140 1.3(8)

La 140 1.3(8)

La 142 1.l(8)

Pr 144 9.l(7)

Ce 144 7.9(7)

• Inventory based on ORIGEN run for equilibrium, end-of-life core. _

There are a few selected nuclides with half-lives around one year or longer which in most instances do not reach equilibrium during the life of the core.

For these few nuclides and within the accuracy of the methodology, a power correction factor which compares the ef fective full power days of the core to the total number of calendar days of cycle operation of the core is applied.

Due to the production characteristics of cesium-134, special consideration must be used to determine the power correction f actor for Cs-134. This power correction factor can be obtained from Figure 2-1.

2.3.1 POWER CORRECTION FACTOR A) Steady state power prior to shutdown.

1) Half-life of nuclide < 1 day nace Pown W M fu oHu 4 davs Power Correction Factor =

Rated Power Level (Mwt)

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Steady state power condition is assumed where the power does not vary by more than 110 percent of rated power level f rom time averaged value.

8) Transient power history in which the power has not remained constant prior to reactor shutdown.

For the majority of the selected nuclides, the 30-day power history prior to shutdown is sufficient to calculate a power correction f actor.

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2.4 RELATIONSHIP OF CLA0 DAMAGE WITH ACTIVITY 2.4.1 GAP INVENTORY Ouring operation, volatile fission products collect in the gap. These fission products are isotopes of the noble gases, iodine, and cesium.

To determine the fission product inventory of the gap, the ANS 5.4 *)

I Standard formulae were used with the average temperature and burnup of the fuel rod. The average gap inventory for the entire core for this methodology was estimated by assuming the core is divided into three regions - a low burnup region, a middle burnup region, and a high burnup region. Using the i

ANS 5.4 Standard, the gap fraction and subsequent gap inventory were calculated for each region. Each region is assumed to represent one-third of the core. The total gap inventory was then calculated by summing the gap inventory of each region. For the purposes of this core damage assessment methodology, this gap inventory is assumed to be evenly distributed throughout -

the core. Table 2-3 shows the calculated gap inventories of the noble gases and lodines. Table 2-3-1 shows the minimum and maximum gap inventories. The minimum and maximum gap inventory were determined by assuming the entire core l

was operating at the low burnup condition and the high burnup conditions, respectively.

4 2.4.2 SPIKING PHENOMENA Reactor coolant system pressure, temperature, and power transients may result in todine spiking. (Cesium spiking may also occur but is not considered in this methodology.) Spiking is noted by an increase in reactor coolant iodine concentrations during some time period after the transient.

In most cases, the iodine concentration would return to normal operating activity at a rate f

based on the system purification half-life. Spiking is a characteristic of the condition where an increase in the normal primary coolant activity is noted but no damage to the cladding has occurred.

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• I Gap Inventory, Curies 3-Loop Nuclide (2652 Mwt)

Kr 85m**

2.82(3) i Kr 87 2.68(3) l Kr 88**

5.93(3)

Xe 131m 6.58(2)

Xe 133 1.31(5) l Xe 133m**

1.25(4)

Xe 135**

6.68(3) l I-131 2.11(5)

I-132 3.39(4) 1-133 1.43(S)

I-135 7.29(4)

• Total core inventory based on 3 region equilibrium core at end-of-life.

Gap inventory based on ANS 5.4 Standard

• • Additional nuclides -

TABLE 2-3-1 GAP INVENTORY MINIMUM AND MAXIMUM Gap Inventory, Curies (Minimum - Maximum)**

3-Loop Nuclide (2652 Mwt)

Kr 85m*

5.14(2)-7.12(3)

Kr 87 5.07(2)-6.86(3)

Kr 88*

1.06(3)-1.48(4)

Xe 131m 1.18(2)-1.65(3)

Xe 133 2.48(4)-3.36(5)

Xe 133m*

9.51(2)-1.32(4)

Xe 135*

3.06(3)-4.07(4)

I 131 4.01(4)-5.47(5)

I 132 6,36(3)-8.70(4)

I 133 2.62(4)-3.65(5)

I 135 1.33(4)-1.86(5)

• Additional nuclides

• • Minimum values are based on the low burnup region (5,000 MWD /MTU).

Maximum values are based on the high burnup region (25,000 MWD /MTV).

1 For this methodology consideration of the spiking phenomena into the radionuclide analysis is limited to the I-131 information found in l

WCAP-9964 WCAP 9964 presents releases in Curles of I-131 due to a j

translent which results in spiking based on the normal primary coolant activity of the nuclides. The WCAP gives an average release and 90 percent confidence interval. These values are presented in Table 2-4.

The use of this data is demonstrated in Section 2.4.3.2.

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l 2.4.3 ACTIVITY ASSOCIATED WITH CLA0 DAMAGE i

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Clad damage is characterized by the release of the fission products which have accumulated in the gap during the operation of the plant. The cladding may j

l rupture during an accident when heat transfer from the cladding to the primary i

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coolant has been hindered and the cladding temperature increases. Cladding l

failure is anticipated in the temperature range of 1300 to 2000'F depending j

upon the conditions of the fission product gas and the primary system I

pressure. Clad damage can begin to occur in regions of high fuel rod peak clad temperature based on the radial and axial power distribution. As the 4

accident progresses and is not mitigated, other regions of the core are expected to experience high temperatures and possibly clad failure. When the l

cladding ruptures, it is assumed that the fission product gap inventory of the l

damaged fuel rods is instantaneously released to the primary system. For this

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methodology it is assumed that the noble gases will escape through the break I

of the primary system boundary to the containment atmosphere and the lodines will stay in solution and travel with the primary system water during the

accident, i

j To determine an approximation of the extent of clad damage, the total activity l

of a fission product released is compared to the total source inventory of the fission product at reactor shutdown.

Included in the measured quantity of the total activity released is a contribution from the normal operating activity i

of the nuclide. An adjustment should be made to the measured quantity of release to account for the normal operating activity. Direct correlations can then be developed which describe the relationship between the percentage of I

total source inventory released and the extent of clad damage for each nuclide. Figures 2-2 through 2-9 present the direct correlations for each I

nuclide in graphical form. The contribution of the normal operating activity 13 i

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i TA8LE 2-4 EXPECTED I0 DINE SPIKE Averaae. uti/am I-131 Total Release. Curies 0.5 < SA* < 1.0 3400 0.1 < SA < 0.5 300 0.05 < SA < 0.1 200 0.01 < SA < 0.05 200 0.005 < SA < 0.01 100 0.001 < SA < 0.005 100 SA < 0.001 2

90/90 Unner Confidence level. uti/am 0.5 < SA < 1.0 6500 0.1 < SA < 0.5 950 0.05 < SA < 0.1 650 0.01 < SA < 0.5 650 0.005 < SA < 0.001 300 0.001 < SA < 0.005 300 SA < 0.001 10 SA is the normel operating 1-131 specific activity (uct/gm) in the primary coolant.

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t

/

/

e

/

/

w

.01 f

o

/

v f

007 f

N,M

/

.005 c

.003

.002

.001 n.

e.

n.

u n m n a o

o o o o o

o o

o o a n a ~ o CladDamage(%)

FIGURE 2-6 RELATIONSHIP OF % CLAD DAMAGE WITH % CORE INVENTORY RELEASED OF XE-131M Rev. 2 19

~.

01

.07 4:

/..

.05

/

.33

/

/

.02

/

/

/

/

.01

.007

.005

/

d:

/

/

.003 l

/

/

i l

.002 5

/

/

1 001

+

6

/

b

/

9

/

Rev. 2 E

7.0 :

/

/

~

O 5 0-4

/

s y 3.0-4

+, '

t 2 0-4 i

~

5

/

/

/

es 1 0-4 '..'

/

s.

0 7.0-5

/

5 0-5 s

,/

s s'

3 0-5

/

s' 2 0-5 p

1.0-5 m,,

e s

=

=

• ..==

N P1 art A o O

O O O O

o e

o o o

~

n n ~ O Clad Damage (%)

FIGURE 2-7 RELATIONSHIP OF % CLAD DAMAGE WITH % CORE -INVENTORY RELEASED OF I-132 Rev. 2 20

3 0.7 E

lb 05 l

1 03 02

/

l 01 s'

l

.07

/

/

.35

/

/

.33

/

/

.02

/

m

/

w

/

w

/

.01

/

j 7

m

.Go?

/

e s

s G

.005

/

/

4s

/

/

Y

/

a:

.cos

/

• 00*

Rev. 2

/

/

l

+*

/

/

h

/

/

.001

/

/

,7 0-4 y'

q 8

5.0-4

/

/

2

.. /

/

3 3 0-4

/

2.0-4

/

/

s' 1 0-4 7 0-5 s

5 0-5 3.0-5 2 0-5 1 0-5 n.

n.

e.

u.

m n

e A o

o o

o o o a

n n

A o

e o

e o o Clad Damage (%)

FIGURE 2-8 RELATIONSHIP OF % CLAD DAMAGE WITH % CORE INVENTORY RELEASED OF I-133 Rev. 2'

'l t

l 1

1 1

1.

0.7

5 t

-~

1 l

05 03 02

/

01

/

.07

[E s

.05 s

/

.03 s

.02 s

s s

M

.01

/

w 007 y

/

/

.005

/

s

/

/

e 6

/

~

e 003 s

O s

~

s g

.002 p

Rey, 2 s

h/

/

u

/

/

O

.001 s

/

e 7 0-4 s'

b s' m

E 5 0-4 s'

s' s

/

e 3.

-4 s

s 5-

/

/

0 2 0-4 ~

s

.s

/

s I.0-4 2

7.0-5 5 0-5

.s

.e 3 0-5 2 0-5 1.0-5 m.

n.

m. n.

m n m n o

o o

o o o

o o

o o o m

n e n a_

Clad Damage (%)

FIGURE 2-9 RELATIONSHIP OF % CLAD DAMAGE WITH % CORE INVENTORY RELEASED OF I-135 Rev. 2 22 e

has been factored into the correlations shown in Figures 2-2 through 2-9.

Examples of how to construct the correlations shown in Figures 2-2 through 2-4 are presented in the next two sections. Figures 2-5 through 2-9 were

]

determined in the same fashion as described in the examples.

It should be noted that not all of the fission products listed in Table 2-3 need to be analyzed but as many as possible should be analyzed to determine a reasonable approximation of clad damage.

2.4.3.1 Xe-133 A graphical representation can be developed which describes the linear relationship of the measured release percentage of Xe-133 to the extent of clad damage. Since the linear relationship is based on percentage of inventory released, the linear relationship applies to all Westinghouse standard plants. The Westinghouse standard 3-Loop plant is used as the base plant for developing the relation. The total source inventory of Xe-133 for a 8

Westinghouse standard 3-Loop plant is 1.6 x 10 Curies (Table 2-2).

For 100 ~

percent clad damage, all of the gap inventory, which corresponds to 0

1.43 x 10 Curies (Table 2-3) would be released. For 0.1 percent clad 2

damage, 1.43 x 10 Curies would be relecsed. These two values can be used to represent two points of the linear relationship between percentage of total inventory released and the extent of clad damage. However, the normal operating activity needs to be accounted into the relation. From Table 2-5 the normal operating activity of Xe-133 is 18 pCi/gm(6)

The average 8

primary coolant mass of a 3-Loop plant is 1.78 x 10 grams. The total normal operating contribution to the total release of Xe-133 is 3200 Curies.

5 Thus the adjusted releases are 3340 Curies and 1.46 x 10 Curies for 0.1 percent clad damage and 100 percent clad damage, respectively. This

-3 corresponds to 2.2 x 10 percent for 0.1 percent clad damage and

-2 9.1 x 10 for 100 percent clad damage. This relation is shown in Figure 2-2.

Figure 2-2 also shows a minimum and a maximum relation which bound the best estimate line. The minimum and maximum lines were determined by bounding the fission product gap inventory. The minimum gap inventory was determined by assuining the entire core was operating at the low burnup condition used to calculate the average gap inventory as described in Section 2.4.1.

The 23

~O

TABLE 2-5 NORMAL OPERATING ACTIVITY

• Specific Activity in Reactor Coolant Nuclide (uCi/am) s Kr 85m 1.1 (-1) i Kr 87 6.0 (-2)

Kr 88 2.0 (-1)

Xe 131m 1.1 (-1)

Xe 133 1.8 (+1)

Xe 133m 2.2 (-1)

Xe 135 3.5 (-1)

I 131 2.7 (-1)

I 132 1.0 (-1)

I 133 3.8 (-1)

I 135 1.9 (-1) 1 l

l

• Values obtained from ANS 18.1 24

maximum gap inventory was determined by assuming the entire core was operating at the high burnup condition of Section 2.4.1.

For the 3-Loop plant, the 4

minimum gap inventory for Xe-133 is 2.71 x 10 C1, and the maximum value is 3.67 x 10' C1. Table 2-3-1 shows the maximum and minimum values for the gap inventories. The normal operating activity is bounded by assuming a w ter 8

mass of 1.23 x 10 grams (2-Loop plant) for the minimum value and 2.6 x 8

10 grams (4-Loop plant) for the maximum value. The points of the minimum and maximum linear relations are calculated in the same manner as discussed above.

2.4.3.2 I-131 The gap inventory for a Westinghouse standard 3-Loop plant f rom Table 2-3 for 5

I-131 is 2.31x10 Curies. The minimum and maximum gap inventory for a 4

3-Loop plant for I-131 is 4.38x10 Ci and 5.98x10 Ci, respectively. The I

source inventory of I-131 for a 3-Loop plant is 8.0 x 10 Curies (Table 2-2).

The normal operating specific activity for I-131 from Table 2-5 is 0.27' 8

uC1/gm. With a primary coolant mass of 1.78 x 10 gm for a standard 3-Loop plant, the normal operating actisity of I-131 is 48 Curies. The points of the average, minimum, and maximum relations are calculated in the same manner as described in Section 2.4.3.1.

Figure 2-3 shows the percentage of I-131 activity as a function of clad damage. The percentage release of I-131 calculated from the radionuclide analysis would be compared to Figure 2-3 to estimate the extent of clad damage.

For I-131, the possibility of iodine spiking should be considered when distinguishing between no clad damage and minor clad damage. The contribution I

of iodine spiking is discussed in Section 2.4.2 and is estimated to be as much as 950 Curies of I-131 released to primary system with an average release of 350 Curies based on a normal operating I-131 activity of 0.27 uCi per gram ( ). The linear relationships of Figure 2-3 are adjusted to account for the release due to iodine spiking by adding 950 Curies of I-131 to the maximum release and by adding 350 Curies of I-131 to the minimum and average release.

Figure 2-4 shows the percentage of I-131 released with lodine spiking versus clad damage.

Iodine spiking was not considered during the calculations of the correlations for the remaining iodines, I-132,1-133, and 1-135, Figures 2-7 through 2-9, respectively.

25

2.4.4 GAP ACTIVITY RATIOS Once equilibrium conditions are reached for the nuclides during operation, a fixed inventory of the nuclides exists within the fuel rod. For these nuclides which reach equilibrium, their relative ratios within the fuel pellet can be considered a constant.

Equilibrium conditions can also be considered to exist in the fuel rod gap.

Under this condition the gap inventory of the nuclides is fixed. The distribution of the nuclides in the gap are not in the same proportion as the fuel pellet inventory since the migration of each nuclide into the gap is dependent on its particular dif fusion rate. Since the relative diffusion rates of these nuclides under various operating conditions are approximately constant, the relative ratios of the nuclides in the gap are known.

In the presence of other indicators of a major release, the relative ratios of the nuclides can be compared with the relative ratios of the nuclides analyzed' (corrected to shutdown) during an accident to determine the source of the fission product release. Table 2-6 presents the relative activity ratios for both the fuel pellet and the gap. The relative ratios for gap activities are significantly lower than the fuel pellet activity ratios. Measured relative ratios greater than gap activity ratios are indicative of more severe failures, e.g., fuel overheat.

2.4.5 A0JUSTMENTS TO DETERMINE ACTIVITY RELEASED When analyzing a sample for the presence of nuclides, the isotopic concentration of the sample medium is expressed as the specific activity of the sample in either Curies per gram of liquid or Curies per cubic centimeter of atmosphere. The specific activity of the sample should then be adjusted to determine the total activity of that medium. The measured activity of the sample needs to be adjusted to account for the decay f rom the time the sample was analyzed to the time of reactor shutdown and adjusted to account for pressure and temperature difference of the sample relative to temperature and i

26 I

~

I

TABLE 2-6 ISOTOPIC ACTIVITY RATIOS OF FUEL PELLET AND GAP Nuclide Fuel Pellet Activity Ratio Gao Activity Ratio Kr-85m 0.11 0.022 Kr-87 0.22 0.022 Kr-88 0.29 0.045 Xe-131m 0.004 0.004 Xe-133 1.0 1.0 Xe-133m 0.14 0.096 Xe-135 0.19 0.051 I-131 1.0 1.0 I-132 1.5 0.17 I-133 2.1 0.71 I-135 1.9 0.39 Noble Gas Ratio = "

5 "Y

Xe-133 Inventory U'

Y Iodine Ratio =

I-131 Inventory i

• The measured ratios of various nuclides found in reactor coolant during i

normal operation is a function of the amount of " tramp" uranium on fuel rod cladding, the number and size of " defects" (i.e. " pin holes"), and the i

location of the fuel rods containing the defects in the core. The ratios derived in this report are based on calculated values of relative concentrations in the fuel or in the gap. The use of these present ratios for post accident damage assessment is restricted to an attempt to differentiate between fuel overtemperature conditions and fuel cladding failure conditions. Thus the ratios derived here are not related to fuel defect levels incurred during normal operation.

27

pressure conditions of the medium. Also the mass (liquid) or volume (gas) of the sample medium is required to calculate the isotopic activity of that medium. The following sections discuss the required adjustments.

2.4.5.1 DILUTION OF SAMPLE MEDIUM The distribution of the total water inventory should be known to determine the water amount that is associated with each sample medium.

If a sample is taken from the primary system, an approximation of the amount of water in the primary system is needed and a similar approximation is required for a sump sample. For the purposes'of this methodology the water is assumed to be distributed within the primary system and the sump. However, consideration should be taken if a significant primary system to secondary system leak rate is noted as in the case of a steam generator tube rupture. The amount of water that is available for distribution is the initial amount of primary system water and the amount of water that has been discharged from the Refueling Water Storage Tank (RWST). Also, an adjustment must be made for water added via the containment spray systems, accumulators, chemical addition tanks, and ice condensers. To approximate the distribution of water, the monitoring systems of the reactor vessel, pressurizer, sunip, and RWST can be employed.

If not all of the monitoring systems are available, the monitoring systems which are working can be used by assuming that the total water inventory is distributed in the sump and the primary system with consideration given if a significant primary system to secondary system leak rate is noted.

The approximate total activity of the liquid samples can then be calculated.

RCS activity (Curies) = Specific Activity (Ci/cc or Ci/gm) x RCS water volume or mass (cc or gm).

Sump activity (Curies) = Specific Activity (Ci/cc or Ci/gm) x Sump water volume or mass (cc or gm).

Total water activity = RCS activity + Sump activity +

Activity leaked to Secondary System + Activities from other sources (accumulators, ice condensers, spray additive tanks, etc.).

23 l

n

\\

Note:

The sp cific activities sh uld b2 decay corrected to reactor shutdown, and the RCS amount should be corrected to account for temperature and pressure differences between sample and RCS.

The containment atmosphere activity can then be added to approximat.e the total activity released at time of accident.

Total Activity Released = Total Water Activity +

Contninment Atmosphere Activity 2.4.5.2 PRESSURE AND TEMPERATURE ADJUSTMENT The measurements for the containment atmosphere samples need to be adjusted if the pressure and temperature of the samples at the time of analysis are different than the conditions of containment atmosphere. The adjustments to

~

the specific activity and the containment volume are as follows.

P T + 460 Specific Activity (Atmosphere) = Specific Activity (Sample) x E p

x (T2 + 460 I

1 where:

T, P) measured sample temperature (*F) and pressure (psia)

=

j c ntair. ment atmosphere temperature (*F) and pressure (psta).

T.P

=

2 2

Rev. 2 For those plants with ice condensers, consideration should be given to account for a decrease in free volume due to the ice melting occupying a portion of th? containment volume.

29

The total activity released to the containment atmosphere is Rev. 2 Total Containment Activity = Specific Activity (Atmosphere) x Free containment Volume where the specific activity (atmosphere) has been decay corrected to time of reactor shutdown.

The specific activity of the liqu.id samples requires no adjustment if the specific activity is reported on a per-gram basis (wC1/ge).

If the specific activity is reported on a per-volume basis (vC1/cc), an adjustment is performed to convert the per-volume specific activity to a per-gram specific activity.

The conversion is performed for consistency with later calculations.

If the temperature of the sample is above 200*F, an adjustment is required to the conversion.

In most cases the sample temperature will be below 200*F and no adjustment is necessary. Figure 2-10 shows a relation of, water density at some temperature relative to the water density at standard temperature and pressure.

The mass of the liquid medium (RCS or sump) can be calculated from the volume of the medium.

If the medium (RCS or sump) temperature at time of sample is above 200*F, an adjustment is required to the conversion.

I A.

RCS or Sump temperature > 200*F 3

RCS or sump mass (ge) = RCS or Sump Volume (f t )

3 28.3 x 10 cc x A (2) x p TP x 3

1. STP ft where:

1. ---( 2 ) = water density ratto at medium (RCS or sump) temperature,

1. STP Figure 2-10 pSTP

= water density at STP = 1.00 ge/cc.

i 1

1

i o

l I

I I

i I

a00 700.

600 500 i

i w

0 400.

L 3

w, 6*

l

{

300 200 l

100.

I i

0 6

7 3

g x

o 9

o e

3 6

l

~

l i

/:

STP l

FIGURE 2-10 WATER CENSITY 2A? 0 (TEMPERATURE VS. STP) l l

l l

31 1

1 i

i

8.

RCS or sump temperature < 200*F 3

RCS or Sump Mass (ge) = RCS of Sump Volume (f t ),,

p, 3

28.3 x 10 cc 3

ft where:

p

= water density at STP = 1.00 gm/cc.

STP The total activity of the RCS or sump is as follows.

RCS or Sump Activity = RCS or Sump Specific Activity (pC1/ge) x RCS or Sump Mass (ge) where the specific activity has been decay corrected to time of shutdown.

2.4.5.3 DECAY CORRECTION The specific activity of a sample is decay adjusted to time of reactor shutdown using the following equation.

Specific activity at shutdown = Soecific activity (measured)

-A t g

l l

where:

A g radioactive decay constant, 1/sec

=

time period from reactor shutdown to time of sample analysis, t

=

sec.

Since this correction may also be performed by some analytical equipe nt, care must be taken to avoid duplicate correction. Also, consideration must be given to account for precursor effect during the decay of the nuclide. For this methodology, only the parent-daughter relationships are considered.

Table 2-7 lists the significant parent-daughter relationships associated with the methodology. The decay scheme of the parent-daughter relationship is described by the following equation.

32

TABLE 2-7 PARENT-0AUGHTER RELATIONSHIPS Parent Daughter M

Half Life

• Dauehter Half Life

• C Kr-84 2.8 h Rb-88 17.8 m 1.00 i

1-131 8.05 d Xe-131m 11.8 d

.008 1-133 20.3 h Xe-133m 2.26 d

.024 1-133 20.3 h Xe-133 5.27 d

.976 Xe-133m 2.26 d Xe-133 5.27 d 1.00 1-135 6.64 h Xe-135 9.14 h

.70 Xe-135m 15.6 m Xe-135 9.14 h *.

1.00 I-135 6.64 h Xe-135m 15.6 m

.30 Te-132 77.7 h I-132 2.26 h 1.00 Sb-129 4.3 h Te-129 68.7 m

.827 l

Te-129m 34.1 d Te-129 64.7 m

.680 Sb-129 4.3 h Te-129m 34.1 d

.173 1

Ba-140 12.6 d La-140 40.22 h 1.00 8a-142 11 m La-142 92.5 m 1.00 Co-144 284 d Pr-144 17.27 m 1.00

• Table of isotones. Lederer. Hollander, and Perlman Sixth Edition

• • Branching decay factor 33 n

A,

-A t

-1,t) + 0

-1,t 3

OA ('

Q s " A,- A i'

3 where:

I activity (C1) or specific activity (vCi/gm or pC1/cc)

Q

=

of the parent at shutdown Q'

activity (C1) or specific activity (vCi/gm or 9C1/cc)

=

of the daughter at shutdown l

activity (Cl) or specific activity (VC1/gm or WC1/cc)

Q

=

g of the daughter at time of sample

~

k decay constant of the parent, sec

=

g

~

decay constant of the daughter, sec 1

=

8 time period f rom reactor shutdown to time of sample t

=

analysis, sec.

Since the activity of the daughter at sample time is due to the decay of the parent and the decay of the daughter initially released at shutdown, an estimation of the fraction of the measured activity at sample time due to only the decay of daughter is required, to use the above equation to determine the fraction, an assumption is made that the fraction of source inventory released of the parent and the daughter at time of shutdown are equal (for the nuclides used here within a factor of 2). The following steps should be followed to calculate the fraction of the measured activity due to the decay of the i

daughter that was released and then to calculate the activity of the daughter released at shutdown.

1.

Calculate the hypothetical daughter concentration (Q ) at the time of g

the sample analysis assuming 100 percent release of the parent and daughter source inventory.

34

.O

k

-A t

-k t

-k t o <t) = x g o; g

g)+oge a

(e

-e g

I where:

1005 source inventory (Cl) of parent, Table 2-2 or 2-8 0

=

1005 source inventory (C1) of daughter, Table 2-2 or 2-8 O

=

g O (t) hypothetical daughter activity (C1) at sample time

=

g l

K if parent has 2 daughters, K is the branching factor,

=

l Table 2-7

~

k parent decay constant, sec a

g daughter decay constant, sec

k

=

g I

time period from shutdown to time of sample, sec.

t

=

2.

Determine the contribution of only the decay of the initial inventory of the daughter to the hypothetical daughter activity at sample time l

1 t

0 e 8

Og (t) l l

3.

Calculate the amount of the measured sample specific activity associated with the decay of the daughter that was released.

1 M6 = Fr x measure specific activity (uci/gm or pCi/cc) i 4.

Oecay correct the specific activity (,8) to reactor shutdown.

(

p m;=.a x.,

e I

35 l

0

i I

t TABLE 2-8 i

SOURCE INVENTORY OF PARENT NUCLIDES NOT LISTED IN TA8LE 2-2 3-Loop Nuclide (2652 MWt)

Xe-135m 3.1(7)

Sb-129 2.4(7)

Te-129m 5.9(6)

Ba-142 1.2(a)

Co-144 7.9(7)

I I

L l

6

2.5 RELATIONSHIP OF FISSION PRODUCT RELEASE WITH OVERTEMPERATURE CONDITIONS The current concept of the mechanisms for fission product release from UO 2

fuel under accident ecnditions has been sunmarized in 2 documents, draf t NUREG-0956 and 10COR Task II.l N.

These documents describe five principal release mechanisms; burst release, diffusional release of the pellet-to-cladding gap inventory, grain boundary release, diffusion from the UO2 grains, and release from molten material. The release which occurs when the cladding fails, i.e., gap release, is utilized to quantify the extent of clad failure as discussed in Section 2.4.

Table 2-9 presents the expected, fuel damage state associated with fuel rod temperatures.

7 Fission product release associated with overtemperature fuel conditions arises initially from that portion of the noble gas, ceslum and iodine inventories that was previously accumulated in grain boundaries. Forhighburnuprods,(t is estimated that approximately 20 percent of the initial fuel rod inventory of noble gases, cesium, and halogens would be released. Release from icwer burnup fuel would no doubt be less. Following the grain boundary release, additional diffusional release from U0 grains ccurs. Estimates of the 2

total release, including U0 diffust nal release, vary from 20 to 40 percent 2

of the noble gas, lodine and cesium inventories.

Additional information on the release of fission products during overtemperature conditions was obtained from the TMI accident In this instance current opinion is that although the core had been overheated, fuel melt had not occurred. Values of core inventory fraction of various fission products released during the accident are given in Table 2-10.

These values, derived f rom radiochemical analysis of primary coolant, sump, and containment gas samples, provide much greater releases of the noble gases, halides, and ceslums, than is expected to be released solely from cladding failures.

In addition, small amounts of the more ref ractory elements, barium-lanthanum, and strontium were released.

In the particular case of TMI, the release mechanism, in addition to diffusional release from grain boundaries and UO2 grains, is believed to arise from U0 grain gr wth in steam, j

2 l

37 l

_n

TABLE 2-9 i

EXPECTED FUEL DAMAGE CORRELATION WITH FUEL ROD TEMPERATURE (8)

Fuel Damaae Temperature 'F*

No Damage

< 1300 Clad Damage 1300 - 2000 Ballooning of zircaloy cladding

> 1300 Burst of zircaloy cladding 1300 - 2000 0xidation of cladding and hydrogen generation

> 1600 Fuel Overtemperature 2000 - 3450 Fission product fuel lattice mobility 2000 - 2550 Grain boundary diffusion release of fission 2450 - 3450 products i

Fuel Melt

> 3450 Dissolution and liquefaction of UO in

> 3450 2

i the Zircaloy - Zr0 eutectic 2

Melting of remaining U0 5100 2

i i

These temperatures are material property characteristics and are non-specific with respect to locations within the fuel and/or fuel cladding.

i 38

--, ~,.

,,_,.___,____..,7,

,,m____

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

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

TA8LE 2-10 PERCENT ACTIVITY RELEASE FOR 100 PERCENT OVERTEMPERATURE CONDITIONS Nuclide Min.*

Mg Nominal **

Min.***

Max.***

Ar-85 40 70 Xe-133 42 66 52.

40 70 1-131 41 55 Cs-137 45 60 Sr-90 0.08****

1 1

sa-140 0.1 0.2 0.15 0.08 0.2

• Release values based on TMI-2 measurements.

Nominal value is simple average of all Kr Xe, I, and Cs measurements.

Minimum and maximum values of all Kr Xe, I and Cs measurements.

• • • • Only value available.

l 39

^

The relationship between extent of fuel damage and fission product release for several radioisotopes during overtemperature condition is depicted graphically in Figures 2-11 and 2-12.

To construct the figures, the extent of fuel damage, expressed as a percentage of the core, is plotted as a linear function of the percentage of the source inventory released for various nuclidas. The values used in constructing the graphs were obtained f rom Table 2-10.

For example, if 100 percent of the core experienced overtemperatures, 52 percent of Xe-133 core inventory would be released.

If 1 percent of the core experienced overtemperature 0.52 percent of Xe-133 core inventors would be released. The assumption is also made that nuclides of any element, e.g.,

1-131 and 1-133, have the same magnitude of release.

In order to apply these figures to a particular plant, power, decay, and dilution corrections described earlier in this report must be applied to the concentrations of nuclides determined from analysis of radionuclide samples. The maximum and minimum estimates of release percentages are those given in Table 2-10 as the range of values: nominal values of release are simple averages of the miminum and maximum values.

2.6 RELATIONSHIP OF NUCLIOE RELEASE WITH CORE MELT CON 0!TIONS Fuel pellet melting leads to rapid release of many noble gases, halides, and cesiums remaining in the fuel after overheat conditions. Significant release of the strontium, barium-lanthanum chemical groups is perhaps the most distinguishing feature of melt release conditions.

Values of the release of fission products during fuel melt conditions are derived f rom ex-pile experiments perfomed by various investigators.

These release measurements have been expressed as release rate coefficients for various temperature regimes. These release rate coefficients have been represented by a simple exponential equation in draft NUREG-0956. This equation has the form:

BT K(T)

Ae where

=

release rate coefficient K(T)

=

A&B constants

=

T temperature.

=

m 4

i I

1 I

l l

100 L

70...

/.

i 50.

/

t

/

/

/

30

/

/

t

/

/

l

/

/

20

/

/

/

/

/

/

/

/

/

/

10

/

/

l

/g'd

/

/

3 7.~:

/

5 f

5!

g 9,' +/p,* /

/

/

/

l g

3..

/

/

/

/

/

/

2 *-

/

/

sa

/

/

/

/

/

/

C

/

/

~

s.

/

/

e

. /

/

o

0. f-

u

/

/

0.s./

./

01 02 0.1 N

b b

b b

b b

b n

n o

~

g i

Fuel Overtemperature (*.)

l FIGURE 2-11 RELATIONSHIP OF % FUEL OVERTEMPERATURE WITH %

CORE INVENTORY RELEASED OF XE, KR, I, OR CS 11 n

i i

1.

0.E:

05 04 l

02

/

/

/

/

O. l.

s 0Z e::

0i s'

,/

,s

.01

/

//s' s'

T

.02 t

6

/

a go* D 9

/

/

l

/

"T

/

/

• o g~

a s'

/

i s

r w

.00f,.

/

s s

o

.005

/

~

e s

g g

.003 e'

i l

/

.002.'

/

e

,s u

i 0

l 001.

/

..s 7.0-4:

5 0-C l

3.0-4.

2 0-4 I

1.0-A N

E N

b k

h Y

k Y

l i

FuelOvertemperature(%)

FIGURE 2-12 RELATIONSHIP OF 4 FUEL OVERTEMPERATURE WITH %

CORE INVENTORY RELEASED OF BA OR SR l

1 n

I

[

These release rate coef fleients were utilized wtth core temperature proflies to develop fission product release estimates for various accident sequences i

for which core melt is postulated in draf t NUREG-Og56.

I Fission product release percentages for three accident sequences which lead to

(

100 percent core melt are given in Table 2-11.

The menon, krypton, cesium, todine, and tellurium elements have been arranged into a single group because of similarity in the expected magnitude of overtemperiture release. The assumption is also made that nuclides of any element e.g., lodine 131 and lodine 133, have the same magnitude of release. The differences in the calculated releases of the various elements for the dif ferent accident sequences were used to determine minimum and maximum values of expected release; nominal values of release are simple averages of all release values within a group.

l The percentage release of various nuclides has been correlated to percentage of core melt with the linear extrapolations shown in Figures 2-13 through 2-15.*

l l

l 2.7 SAMPLING LOCATIONS A survey of a number of Westinghouse plants has indicated that the post l

accident samp1tng system locations for liquid and gaseous samples varies for each plant. To obtain the most accurate assessment of core damage. It is reconsnended to sample and analyze radionuclides from the reactor coolant system, the containment atmosphere, and the containment sump (if available).

Other samples can be taken dependent on tne plant's capaht11 ties. The specific sample locations to be used during the initial phases of an accident should be selected based on the type of accident in progress.

If the type of accident scenario is unknown, known plant parameters (pressure, temperature, level indications, etc.) can be used as a basis to determine the prime sample locations. Consideration should be given to sampilng secondary system if a significant leak from the primary system to secondary system is noted. Table 2-12 presents a list of the suggested sample locations for dif ferent accident scenarios based on the usefulness of the information derivable from the sample.

U A

TA8LE 2-11 PERCENT ACTIVITY RELEASE FOR 100 PERCENT CORE MELT CON 0!TIONS Large*

Small*

Nominal **

Min.***

Max.***

Snecies M

Transient

• M Release Reltale Release I

Xe 88.35 99.45 78.38 Kr 88.35 99.45 78.38 87 70 99 1

88.23 99.44 78.09 l

Cs 88.55 99.46 78.84 1

Te 78.52 94.88 71.04 i

Sr 10.44 28.17 14.80 24 10 44 l

l l

8a 19.66 43.87 24.08 j t

l Pr 0.82 2.36 1.02 1.4 0.8 2.4 l

l

[

Calculated releases for severe accident scenarios without emergency

(

safeguard features, taken f rom draf t NUREG-0956 Nominal release are averages of Xe, Kr I, CS, and To groups, or $r l

and 84 groups.

l Maximum and minimum releases represent extremes of the groups.

=0

100,

s

/

7 0..

s 50.-

/,'..

s s

/

e 30.<

f

/

ll

/

20.-

/

'/

/

/ NO 10

/

'/

' #/

g 7.-

/

f s

s/

E s.-

/,i

/

g

,'ls g

3.-

/

l D

g.

,/,s 3a

/

/

3

i..'

e

/

o. r::'

o.s o.a 01 01 5

5 5

E N

N N

N N

Fuel Melt (%)

FIGURE 2-13 RELATIONSHIP OF % FUEL MELT WITH 4 CORE INVENTORY RELEASED OF XE, KR, I, CS, OR TE 45 P

---

a

I 100.0 t

B

/..

j

/

/

/

10.0

,s s

s

,,0'*C ' ',0 a

so s'

s 2

s' s'

h bp ',

p 1.0 o

y

,s s

y

=

/

o

/

0.1 0.01 1.0 10 0 100.0 Fuel Melt (%)

FIGURE 2-14 RELATIONSHIP OF % FUEL MELT WITH % CORE INVENTORY RELEASED OF BA OR SR 46 O

._. =-

100.0 10.0

(

g

~

1.0 Y.

,c.

e T

x b

$t*g,#'

o

'go \\B*s T

0.1 e

i E

b g& '

i e

o' t

u 0.01 l

0.001 l

1.0 10.0 100.0 l

Fuel Melt (%)

i FIGURE 2-15 RELATIONSHIP OF *. FUEL MELT WITH % CORE INVENTORY RELEASED OF PR t

47

TA8LE 2-12 Succested Samolina locations Principal Other Scenario Samplina locations Samolina locations Small Break LOCA Reactor Power > 1%*

RCS Hot Leg, Containment RCS Pressurizer Atmosphere Reactor Power < 1%*

RCS Hot Leg **

RCS Pressurizer large Break LOCA Reactor Power > 1%*

Containment Sump, Containment Atmosphere, RCS Hot Leg Reactor Power < 1%*

Containment Sump, Containment Atmosphere Steam Line Break RCS Hot Leg, RCS Pressurizer Containment Atmosphere Steam Generator Tube RCS Hot Leg, Secondary Containment Rupture System Atmosphere s

Indication of Signifi-Containment Sump, Containment cant Containment Sump Atmosphere I

Inventory Containment Building Containment Atmosphere, Radiation Monitor Alarm Containment Sump Safety Injection RCS Hot leg RCS Pressurizer Actuated Indication of High RCS Hot Leg RCS Pressurizer Radiation Level in RCS Assume operating at that level for some appreciable time.

• • If a RCS hot leg sample is unavailable and a RCS cold leg sample is available, obtain a RCS cold leg sample. However, for a cold leg sample to be a good representation of the RCS, the primary water should be circulating through the system.

)

28 n

3.0 AUXILIARY INDICATORS There are plant indicators monitored during an accident which by themselves cannot provide a useful estimate but can provide verification of the initial estimate of core damage based on the radionuclide analysis. These plant indicators include containment hydrogen concentration, core ;xit thermocouple temperatures, reactor vessel water level, and containment radiation level.

l When providing an estimate for core damage, these plant indicators, if available, should confirm the results of the radionuclide analysis. For example, if the core exit thermocouple readings and reactor vessel water level indicate a possibility of clad damage and the radionuclide concentrations indicate no clad damage, then a recheck of both indications may be perfomed or certain indications may be discounted based on engineering judgment.

3.1 CONTAINMENT HYOR0 GEN CONCENTRATION An accident, in which the core is uncovered and the fuel rods are exposed to steam, may result in the reaction of the zirconium of the cladding with the steam which produces hydrogen. The hydrogen production characteristic of the zirconium vater reaction is that for every mole of zirconium that reacts with water, two moles of hydrogen are produced. For this methodology it is assumed that all of the hydrogen that is produced is released to the containment atmosphere. The hydrogen dissolved in the primary system during normal operation is considered to contribute an insignificant amount of the total hydrogen released to the containment.

In the absence of hydrogen control measures, monitoring this containment hydrogen concentration during the accident can provide an indication of the extent of zirconium water reaction.

The percentage of zirconium water reaction does not equal the percentage of clad damaged but it does provide a qualitative verification of the extent of clad damage estimated from the radionuclide analysis.

Figure 3-1 shows the relationship between the hydrogen concentration and the percentage of zirconium water reaction.

The relaticnship shown in Figure 3-1 does not account for any hydrogen depletion due to hydrogen recombiners and hydrogen ignitions. The recombiners that new exist are capable of dealing 49

26.0 I

24.0 22.0 20.0 18.0 l

l 16.0 HYDROGEN CONCENTRATION (v/o) 14.0 ICE CONDENSER 2-LOOP 12.0 10.0 8.0 4-LOOP 6.0 1

4.0 2.0 10 20 30 40 50 60 70 80 90 100 ZIRCONIUM WATER REACTION (".)

FIGURE 3-1 CONTAINMENT HYDROGEN CONCENTRATION BASED ON ZIRCONIUM WATER REACTION 50

effectively with the relatively small amounts of hydrogen that result from radiolysis and corrosion followiag a design basis LOCA. However, they are incapable of handling the hydrogen produced in an extensive zirconium-steam reaction such as would result from severe core degradation. Current recombiners can process gas that is approximately 4 to 5 percent hydrogen or less(10)

Each recombiner unit can process an input flow in the range of 100 SCFM to 200 SCFM. Within the accuracy of this methodology, it is assumed that recombiners will have an insignificant effect on the hydrogen concentration when it is indicated that extensive zirconium-steam reaction could have occurred. Uncontrolled igniticn of hydrogen and deliberate ignition will hinder any quantitative use of hydrogen concentration as an auxiliary indicator. However, the oxygen amount depleted during the burn, if known, can be used to estimate the amount of hydrogen burned. If the oxygen amount depleted is not known, it can be assumed that for ignition of hydrogen to occur a minimal concentration of 4 percent hydrogen is needed. This assumption can be used qualitatively to indicate that some percentage of zirconium has reacted, but it is difficult to determine the extent of the reaction.

Containment hydrogen concentrations can be obtained from the Post Accident Sampling System or the containment gas analyzers. Figure 3-1 shows the relationsnip between the hydrogen concentration (percent volume) and the percentage of zirconium water reaction for the Westinghouse Standard two loop, three loop, and four loup plants and the ice condenser containment plant. The l

hydrogen concentration shown is the result of the analysis o'f a dry containment sample. Th; curves were based on cverage containment volumes and the average initial zirconium mass of the fuel rods for each type of plant, which are shown in Table 3-1.

Table 3-1 also presents the correlation between hydrogen concentration and percentage of zirconium water reaction. To use the auxiliary indicator of hydrogen concentration, the assumptions were that all hydrogen from zirconium water reaction is released to containment, a well-mixed atmosphere, and ideal gas behavior in containment.

51 m

O

TABLE 3-1 AVERAGE CONTAINMENT VULUME AND ZIRCONIUM MASS Plant Tvoe Averace Zirconium Mass (1bm)

Average Containment Volume (SCF) 6 2-Loop 23,900 1.2 x 10 6

3-Loop 37,500 1.7 x 10 6

4-Loop 47,300 2.8 x 10 6

Ice 47,300 1.2 x 10 Condenser 1

i l

Relationship between hydrogen concentration of a dry sample and fraction of I

zirconium water reaction is based on the following formula.

(FZWR)(IM)(H)

%H x 00 2 " (FZWR)(ZM)(H) + V where: FZWR = f raction of zirconium water reaction ZM

= total zirconium mass, 1be H

= conversion factor, 7.92 SCF of H per pound of zirconium 2

reacted V

= containment volume, SCF 52

1 3.2 CORE EXIT TEMPERATURES AND REACTOR VESSEL WATER LEVELS Core exit thermocouples measure the temperature of the fluid at the core exit at various radial core locations (Figure 3-2).

The typical thermocouple system is qualified to read temperatures as high as 1650*F. This is the ability of the system to measure the fluid temperatures at the incore thermocouples locations and not core temperatures.

Most reactor vessel level indication systems (RVLIS) use differential pressure (d/p) measuring devices to measure vessel level or relative void content of the circulating primary coolant system fluid. The system is redundant and includes automatic compensation for potential temperature variations of the j

impulse lines. Essential information is displayed in the main control room in a form directly usable by the operator.

RVLIS and CETC readings can be used for verification of core damage estimates in the following ways o

Due to the heat transfer mechanisms between the fuel rods, steam, and therscouples, tne highest clad temperature will be higher than the CETC readings. Therefore, if thermocouples read greater than 1300*F, clad failure may have occurred. 1300*F is the lower limit for cladding failures.

o If any RCPs are running, the CETCs will be good indicators of clad temperatures and no core damage should occur since the forced flow of the steam-water mixture will adequately cool the core.

If RCPs are not running, the follcwing apply.

o No generalized core damage can occur if the core has not uncovered. So if RVLIS full range indicates that the collapsed liquid level has never been below the top of the core and no CETC has indicated temperatures corresponding to superheated steam at the corresponding RCS pressure, then no generalized core damage has occurred.

53

l i

2 3

4 5

6 7

8 9

10 11 12 13 14 15 A

A T

E A

D C

g T

T U

C T

T 8

E 0

T T

D

^

E T

T T

T T

T B

C 5

0 5

C 5

p T

T T

T T

A E

D 0

T T

T T

T E

A D

E A-C T

T T

T T

T T

D i

' G-0 t

c g

T T

T-K l

e

' i i t

i T T

C E

A 5

E 8

A g

T T

T T

M 8

T T

T N

D E

D A

E T

T T

T T

T C

R T

A - FLUX THINBLE DETECTOR A E - FLUX THIMBLE DETECTOR E I

8 - FLUX THIM8LE DETECTOR S CP - CAllSRATION FLUX THlHSLE C - FLUX THIH8LE DETECTOR C (CONMON PATH) 0 - FLUX THIMBLE DETECTOR 0 T - THERMOCCUPLE J

Figure :To. 3-2 Core Distribution of Flu Sir.bles and Ther= occupies For BVPS 41 & #2 l

l l

o If RVLIS indicates less than 3.5 ft, collapsed liquid level in the core or CETCs indicate superheated steam temperatures, then the core has uncovered and core damage may have occurred depending on the time after reactor trip, length and depth of uncovery. Best estimate small break (1 to 4 inches) analyses and the Three Mile Island (TMI)(I2) accident data indicate that about 20 minutes after the core uncovers clad temperatures start to reach 1200*F and 10 minutes later they can be as high as 2200*F.

These times will shorten as the break size increases due to the core uncovering faster and to a greater depth.

o If the RVLIS indication is between 3.5 ft collapsed liquid level in the core and the top of the core, then the CETCs should be monitored for superheated steam temperatures to determine if the core has uncovered.

As many thermocouples as possible should be used for evaluation of the core temperature conditions. The Emergency Response GuidelinesO3) recomend that a minimum of one thermocouple near the center of the core and one in each-quadrant be monitored at identified high power assemblies. Caution should be taken if a thermocouple reads greater than 1650*F or is reading considerably different than neighboring CETCs. This may indicate that 'he thermocouple has t

failed. Caution should also be used when looking at CETCs near the vessel walls because reflux cooling from the hot legs may cool the fluid in this area. CETCs can also be used as an indicator of hot areas in the core and may be used to determine radial location of possible local core damage.

Therefore, core exit thermocouples and RVLIS are generally regarded as reliable indicators of RCS conditions that may cause core damage. They can predict the time of core uncovery to within a few minutes by monitoring the core exit thermocouples for superheat after RVLIS indicates collapsed liquid level at the top of the core. The onset and extent of fuel damage after core uncovery' depend on the heat generation in the fuel and the rapidity and duration of uncovery. However, if the core has not uncovered, no generalized fuel damage has occurred. Core exit thermocouples reading 1300*F or larger indicate the likelihood of clad damage.

55 m.

3.3 CONTAINMENT RADIATION MONITORS AND CORE DAMAGE Post accident radiation monitors in nuclear plants can be used to estimate the xenon aad krypton concentrations in the containment.

An analysis has been made to correlate these monitor readings in R/hr with gaseous concentrations. For this analysis the following assumptions were made:

1.

Radiogases released from the fuel are all released to containment.

2.

Accidents were considered in which 100% of the noble gases, 52% of noble gases, and 0.3% of the noble gases were released to the containment.

3.

Halogens and other fission products are considered not to be significant contributors to the containment monitor readings.

A relation can be developed which describes the gamma ray exposure rate of a detector with time, based on the amount of noble gases released. The exposure rate reading of a detector is dependent on plant specific parameters: the nperating power of the core, the ef ficiency of the monitor, and the volume seen by the monitor. The plant specific response of the detector as a function of time following the accident can be calculated from the instantaneous gamma ray source strengths due to noble gas release, Table 3-2, and the plant characteristics of the detector. The gamma ray source strengths presented in Table 3-2 are based on 100 percent release of the noble gases.

To determine the exposure rate of the detector based on 52 percent and 0.3 percent noble gas release, 52 percent and 0.3 percent, respectively, of the gamma ray source strength are used.

Alternately, the energy rates in Mev/ watt-sec given in Table 3-2 can be expressed in terms of an instaneous flux by assuming the energy is absorbed in 3

a cm of air. These energy rate values, in Mev/ watt-sec-cm, when divided by discrete values of Mev/ photon and the gamma absorption coefficient for air, u, considered as a constant (3.5 x 10 c,-1),

rovide values of the

-5 2

photon flux, photons / watt-cm.-sec, as shown in Table 3-2A.

The discrete values of Mev/ photon were obtained by using the average values of the energy groups Mev/ gamma, from Table 3-2.

56

TABLE 3-2 INSTANTANEOUS GAMMA RAY SOURCE STRENGTHS DUE TO A 100 PERCENT RELEASE OF NOBLE GASES AT VARIOUS TIMES FOLLOWING AN ACCIDENT Enerav Group Source Strength at Time After Release (Mev/ watt-sec)

Mev/cansna 0 Hours 0.5 Hours 1 Hour 2 Hours 8 Hours 9

9 8

8 8

0.20 - 0.40 1.2 x 10 3.0 x 10 2.6 x 10 2.4 x 10 2.0 x 10 9

8 8

8 7

0.40 - 0.90 1.5 x 10 3.4 x 10 2.6 x 10 1.9 x 10 5.9 x 10 9

7 I

7 6

0.90 - 1.35 1.3 x 10 9.4 x 10 6.7 x 10 4.7 x 10 9.8 x 10 9

8 8

7 7

1.35 - 1.80 1.8 x 10 3.4 x 10 2.1 x 10 1.4 x 10 2.9 x 10 9

8 8

8 7

1.80 - 2.20 1.4 x 10 5.4 x 10 3.6 x 10 2.4 x 10 5.2 x 10 9

8 8

8 8

2.20 - 2.60 1.3 x 10 8.5 x 10 7.1 x 10 5.3 x 10 1.1 x 10 8

6 6

6 5

2.60 - 3.00 4.0 x 10 6.6 x 10 5.1 x 10 3.5 x 10 5.0 x 10 8

5 6

6 4

3.00 - 4.00 3.5 x 10 6.3 x 10 4.5 x 10 2.6 x 10 9.7 x 10 7

4 2

4.00 - 5.00 3.1 x 10 4.4 x 10 3.6 x 10 0

0 5.00 - 6.00 0

0 0

0 0

Mev/camma 1_,Qgy, 1 Week 1 Month 6 Months 1 Year 8

7 6

0.20 - 0.40 1.3 x 10 3.0 x 10 1.5 x 10 0

0 I

4 4

4 4

0.40 - 0.90 1.1 x 10 1.5 x 10 1.5 x 10 1.5 x 10 1.4 x 10 5

0.90 - 1.35 1.8 x 10 0

0 0

0 5

1.35 - 1.80 5.5 x 10 0

0 0

0 5

1.80 - 2.20 9.9 x 10 0

0 0

0 6

2.20 - 2.60 2.0 x 10 0

0 0

0 3

2.60 - 3.00 8.5 x 10 0

0 0

0 3.00 - 4.00 0

0 0

0 0

4.00 - 5.00 0

0 0

0 0

5.00 - 6.00 0

0 0

0 0

57

.a'

TABLE 3-2A INSTANTANEOUS GAMMA RAY FLUXES DUE TO 100% RELEASE OF NOBLE GASES AT VARIOUS TIMES FOLLOWING AN ACCIDENT Enerav Group Photon Flux at Time After Release (Dhotons/cm -watt-sec)

Mev/aamma 0 Hours 0.5 Hours 1 Hour 2 Hours 8 Hours I4 13 13 13 13 0.3 1.1 x 10 2.7 x 10 2.4 x 10 2.2 x 10 1.8 x 10 I4 13 I3 13 12 0.65 1.0 x 10 2.3 x 10 1.7 x 10 1.3 x 10 3.9 x 10 13 12 12 12 II 1.13 3.3 x 10 2.4 x 10 1.7 x 10 1.2 x 10 2.5 x 10 13 12 12 II Il 1.58 3.3 x 10 6.2 x 10 3.8 x 10 2.5 x 10 5.3 x 10 13 12 12 12 Il 2.0 2.0 x 10 7.7 x 10 5.1 x 10 3.4 x 10 7.4 x 10 13 13 12 12 12 2.4 1.5 x 10 1.0 x 10 8.4 x 10 6.3 x 10 1.3 x 10 12 10 10 10 9

2.8 4.1 x 10 6.7 x 10 5.2 x 10 3.6 x 10 5.1 x 10 12 9

10 10 8

3.5 2.9 x 10 5.3 x 10 3.8 x 10 2.2 x 10 8.1 x 10 II 9

6 4.5 1.9 x 10 2.8 x 10 2.3 x 10 0

0 Mev/camma M

1 Week 1 Month 6 Months 1 Year 13 12 II 0.3 1.2 x 10 2.7 x 10 1.4 x 10 O

O II 3

9 9

9 0.65 7.3 x 10 1.0 x 10 1.0 x 10 1.0 x 10 1.0 x 10 9

1.13 4.5 x 10 0

0 0

0 10 1.58 1.0 x 10 0

0 0

0 10 2.0 1.4 x 10 0

0 0

0 10 2.4 2.4 x 10 0

0 0

0 2.8 8.7 x 10 0

0 0

0 3.5 0

0 0

0 0

4.5 0

0 0

0 0

58

l In general, values below 0.3% releases are indicative of clad failures, values between 0.3% and 52% release are in the fuel pellet overtemperature regions, while values between 52% release and 100% release are in the core melt regime. To represent the release of the normal operating noble gas activity in the primary coolant as obtained from ANS 18.l( }, 1.0 x 10 % of the

-3 gansna ray source strength is used.

In actual practice it must be recognized that there is overlap between the regimes because of the nature in which core heating occurs. The hottest portion of the core is in the center due to flux distribution and hence greater fission product inventory. Additionally heat transfer is greater at the core periphery due to proximity of pressure vessel walls. Thus conditions could exist where there is some molten fuel in the center of the core and overtemperature conditions elsewhere. Similar conditions can occur which lead to overtemperature in the central portions of the core, and clad damage elsewhere. Thus, estimation of extent of core damage with containment radiation readings must be used in a confirmatory sense -- as backt.p to other measurements of fission product release and other indicators such as pressure vessel water levels and core exit thermocouples.

An example of the relationship of the exposure rate of a detector as a function of time following reactor shutdown is presented in Figure 3-3.

The exposure rates, which are expressed in units of R/hr-MWt, are representative of a point located 57.5 feet below the apex of the containment dome of a 6

3 containment volume of 2 x 10 ft. No objects or components shield the detector from the noble gas sources which are assumed to be uniformly distributed throughout the containment building.

The methodology of using the relationship of containment radiogas monitors readings shown in Figure 3-3 is:

1.

Determine time lapse between core shutdown and radiation reading.

2.

Record containment monitor reading in R/hr at this time.

3.

Correct the monitor reading for specific plant power via the relationship:

R N M = Radiation Monitor Reading Plant Power (MWt) 59

h 1000.0u 100% Noble Gas Release 100.0; 52% Noble 10.0 Gas Release C

i E

4 1.0 W

2 f

1.0-1:

0.3% Noble Gas 1

g 8

Release g

)

1.0-2:!

r

{

NS 18.1 Normal Operating 1.0-3:5

s Noble Gas Release s*

'E 1.0-4!!

T

?

1.0-5 1.0 10.0 100.0 1000.0 TIME AFTER ACCIDENT (HOURS)

FIGURE 3-3 PERCENT NOBt.E GASES IN CONTAINMENT 1

60

1 l

i 4.

Determine core damage regime from Figure 3-3 at the time interval j

ascertained in step 1.

For plants which have the same monitor characteristics as the monitor described above, except for the containment volume Figure 3-3 can be used provided a correction is made to the exposure rate (R/hr) as follows.

R/

t = Radiation Monitor Readina (R/hr) x Containment Vol. (ft )

6 3

Plant Power (MWt) x 2 x 10 ft

~

• 4.0 GENERALIZED CORE DAMAGE ASSESSMENT APPROACH Selected results of various analyses of fission product release, core exit thermocouple readings, pressure vessel water level, containment radiogas monitor readings and hydrogen monitor readings have been summarized in Table 4-1.

The intent of the summary is to provide a quick look at various criteria intended to define core damage over the broad ranges of:

No Core Damage l

0-50%

clad failure l

50-100%

clad failure 0-50%

fuel pellet overtemperature 50-100%

fuel pellet overtemperature 0-50%

fuel melt 50-100%

fuel melt Although this table is intended for generic applicability to most Westinghouse pressurized water reactors, except where noted, various prior calculations are required to ascertain percentage release fractions, power, and containment volume corrections. These corrections are given within the prior text of this technical basis report.

j The user should use as many indicators as possible to differentiate between the various core damage states. Because of overlapping values of release and potential simultaneous conditions of clad damage, overtemperature, and/or core melt, considerable judgement needs to be applied.

62

T Ast t 4-1 CNASACTIRISIICS of CAIEGeelts Of f u(L DAflAGE*

Core Damage Contalanent Indicator Percent Radlegas and Type Ilonttor Core tatt leydrogen Cua of fIsston itsslen (t/hr - shot)

Therustouples Core leonitor D. mage Products Product to brs after Seadings Uncovery (vol 5 sey)***

Category Deleased Batto shutdoun**

(Beg f)

Indication

& Plant lype eso clad damage Er-el < 1 10-3 mot Appilcable

< ISO so uncovery megilglble se-133 < 3 10-3 I-131 < 1a10-3 I-133 < 1x10-3 0-50s clad damage Er-87 10 0.08 Er-87 = 0.022 0

.00 150 - 1300 Core uncovery 0-7 se-133 10 0.1 I-838 10 0.3 I-133 = 0.Il 1-133 10 0.1 50 1005 clad damage Er-SP 0.01 - 0.02 Er-07 = 0.022 0.00 to 0.16 1300 - 1650 Core uncovery 7-14 se-133 0.1 - 0.2 I-838 0.3 - 0.5 1-833 = 0.18 I-133 0.1

- 0.2 0 50s f uel pellet se Er.Cs.I Er-el = 0.22 0.16 to 21

> 1650 Core uncovery 7-14 overtemperature B - 20 Sr-Ba 0 - 0.5 I-133 = 2.1 m

w 50 100% fuel pellet se Er.Cs.I Er-al = 0.22 25 to 42

> 1650 Core uncovery 7-14 overtemperature 20 - 40 Sr-Ba 0.1 - 0.2 I-833 = 2.1 0-50s fuel melt me.Kr.Cs.I 40 - 70 Er-83 = 0.22 42 to 10

> 1650 Core uncovery 7-]4 Sr-Ba 0.2 - 0.8 Pr 0.1 - 0.8 I-833 = 2.1 50-100% fuel melt se Er.Cs.I,le Kr-87 = 0.22

> 10

> 1650 Core encovery 7-14

> 70 Sr ta > 24 I-133 = 2.1 Pr > 0.8

• Ihls table is intended to supplement the methodology outlined la this report and should not be used without referring to this report and without considerable engineering judgement.

• • values should be revised per plant specif ic parameters and times other than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. These values are from flgure 3-3 and represent a specific de t ec t or geome t r y,

Ignitors may obulate these values.

.... er er 1:11) ar Ill' i 131 I

5.0 LIMITATIONS The emphasis of this methodology is on radiochemical analysis of appropriate liquid and gaseous samples. The assumption has been made that appropriate post-accident systems are in place and functional and that representative samples are obtained. Of particular concern, in the area of representative sampling, is the potential for plateout in the sample lines.

In order to preclude such plateout, it is assumed that proper attention to heat tracing of the sample lines and maintenance of sufficient purge velocities is inherent in the sampling system design.

Having obtained a representive sample, radiochemical analysis via gama spectrometry are used to calculate the specific activity of various fission products released from the fuel.

Radiochemical analyses of fission products under normal plant operating conditions are accurate to 110 percent. Radiochemical analyses of post accident samples which may be much more concentrated, and contain unfamiliar nuclides, and which must be performed expeditiously may have an error band of 20 to 50 percent.

Having obtained specific activity analysis, the calculation of total release requires knowledge of the total water volume from which the samples were taken. Care must thus be exercised in accounting for volumes of any water added via ECCS and spray systems, accumulators, chemical addition tanks, and melting ice of ice condenser plants. Additionally estimates of total sump water volumes have to be determined with data from sump level indicators.

Such estimates of water volume are probably accurate to 110 percent.

The specific activity also requires a correction to adjust for the decay of the nuclide in which the measured specific activity is decay corrected to time of reactor shutdown. For some nuclides, precursor ef fects must be considered in the decay correction calculations.

The precursor ef fect is limited to parent-daughter relationships for this methodology. A major assumption is made that the release percentages of the parent and daughter are equal. For overtemperature and melt releases, this assumption is consistent with the technical basis presented in Sections 2.5 and 2.6, but the gap releases could be different by as much as a factor of 2.

e4

~.

The models used for estimation of fission product release from the gap activity are based on the ANS 5.4 standard. Background material for this report indicate the model, though empirical, is believed to have an accuracy of 20-25 percent.

In our application of these models to core wide conditions.

the core has arbitrarily been divided into three regions of low, intermediate, and high burnup. This representation predicted nominal values of release with maximum and minimum values that approach 1100 percent of the nominal value.

Therefore these estimates of core damage should only be considered accurate to a factor of 2.

The models employed for estimates of release at higher temperature have not been completely verifled by experir.ent. Additionally, calculations of expected core temperatures for severe accident conditions are still being refined. These uncertainties are exacerbated by the manner in which various accident scenarios leading to core melt have been combined to produce fission product release predictions for the core melt condition. Consideration of the melt release estimates shown in Table 2-11 for the ref ractory nuclides indicate a range of approximately i 70 percent.

From these considerations it is clear that the combined uncertainties are such that core damage estimates using this methodology are sufficient only to establish major categories of fuel damage. This categorization, and confirmation of subcategorization will require extensive additional analysis for some several days past the accident date.

l l

i l

l l

65

~

6.0 EXAMPLE OF CORE DAMAGE ASSESSMENT METHODOLOGY The following example is presented to illustrate the use of this methodology in assessing the extent of core damage.

6.1 SAMPLING RESULTS For this example, a Westinghouse 3-Loop plant has experienced an accident where the plant's monitoring systems indicate that safety injection has initiated and a significant amount of water has accumulated in the sump.

Samples are available from the primary coolant (RCS hot leg), the containment sump, and the containment atmosphere 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after reactor shutdown. The results of the sampling are presented in Table 6-1.

6.2 DECAY CORRECTION The specific activities determined by the sampling analysis are decay corrected to the time of reactor shutdown. A sample calculation is presented here.

^o"

-A t g

where:

measured specific activity, uti/gm or uti/cc A

=

decay constant of isotope i, sec-K

=

g time elapsed from reactor shutdown to time of sampling, sec.

t

=

decay corrected specific activity pCi/gm or pC1/cc.

A

=

i 66 l

TABLE 6-1 l

RESULTS OF SANPLING ANALYSIS TAKEN 6 HOURS AFTER REACTOR SHUTOOWN Specific Activity Isotope Atmosphere uC1/cc Sumo. uC1/am RCS. uCi/am Kr 87 1.8(1)

Xe 133 1.9(3)

I 131 2.6(4) 6.9(4)

I 132 4.4(4) 1.2(4)

Cs 137 2.6(3) 6.5(3)

Ba 140 4.4(4) 1.3(S) 1 l

67

...,-.---,.,..--w.O.

~-.--n-

For I-131 primary coolant specific activity. Table 6-1 6.9E4 A

=

-I k

1.0E-6 sec

=

g 21600 sec.

t

=

6.9E4 3

,-(1.0E-6)x(21600) o 7.0E4 A

=

For I-132, parent-daughter relationship must be considered in calculation of

- decay adjustment. Following the methodology outlined in Section 2.4.5.3, the decay correction calculation is as follows.

Parent-Daughter: Te-132 - I-132 Hypcthetical activity of I-132 (daughter) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> af ter shutdown, assuming 100 percent release of Te-132 and I-132.

A

-A^t

-X t

-A t 0)+QB' 0

Q (t) = K 0

(e

-e g

where:

O 100% source inventory of Te-132, Table 6.2A = 1.2E8 Ci Q

=

100% source inventory of I-132, Table 6.2A = 1.2E8 Ci Q

=

decay branching factor, Table 2-7 = 1.00 K

=

-I Te-132 decay constant = 2.48E-6 sec K

=

A 68 a.

l TABLE 6-2A SOURCE INVENTORY

• Nuclide Inventory. Ci Kr-85m 1.8(7)

Kr-87 3.3(7)

Kr-88 4.6(7)

Xe-131m 5.1(5)

Xe-133 1.6(8)

Xe 133m 2.3(7)

Xe-135 3.0(7)

I-131 8.0(7)

I-132 1.2(8)

I-133 1.6(8)

I-135 1.4(8)

Rb-88 4.7(7)

Cs-134 1.9(7)

Cs-137 8.7(6)

Te-129 2.7(7)

Te-132 1.2(8)

Sr-89 6.4(7)

Sr-90 5.9(6)

Ba-140 1.4(8)

La-140 1.4(8)

La-142 1.2(8)

Pr-144 1.0(8)

• The source inventory of a 3-Loop (2900 Mwt) plant is used in this example.

69

I-132 decay constant = 8.52E-5 sec'I 1

=

8 time from shutdown to sample time = 21600 sec t

=

9.75E7 C1 + 1.91E7 C1 O

=

g 1.17E8 C1 of I-132 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after shutdown O

=

g 2.

Contribution of decay of only I-132 to hypothetical activity of I-132.

~

Q e 8

08 (t) 1,91E7

" 1.17E8 Fr = 0.16 3.

Amount of measured sample specific activity due to decay of just I-132.

M = Fr x measured specific activity (i.e., RCS), Table 6-1 g

= 0.16 x 1.2ES M = 1.9E4 B

4.

Decay correct specific activity (A) of I-132 in RCS to shutdown.

"8 M

=

,-A t

g

= Lili

.16 M = 1.2E5 B

70

Table 6-28 lists the decay corrected specific activities of the sampling analysis.

6.3 PRESSURE AND TEMPERATURE CORRECTION As discussed in Sectionj2.4.5.2, a correction is needed to the sample specific activity only if the temperature and pressure of the actual sample are different than the temperature and pressure of the medium from which the sample was taken. Since the measured specific activity of the RCS and sump samples are based on gram of water, no adjustment to the specific activities is required. The conditions of the medium and the sample are listed below.

Containment Atmosphere Atmosphere Samole Correction Factor Pressure = 20 psia Pressure = 15 psia 1.1 Temperature = 200*F Temperature = 100*F Containment Sumo Sumo Sample Correcti6n Factor Pressure = 20 psia Pressure = 20 psia 1.0 Temperature = 125'F Temperature = 125'F Primary Coolant RCS Sample Correction Factor Pressure = 1500 psia Pressure = 500 psia 1.0 Temperature = 350*F Temperature = 150'F Correction factor calculations are shown below.

For containment atmosphere sample, P2 (T) + 460)

Correction Factor = P)

(T2 + 460) l 71 o

TABLE 6-28 DECAY CORRECTED SPECIFIC ACTVITIES OF SAMPLING ANALYSIS Measured Decay Corrected Specific Parent-0aughter Decay Specific Nuclide location Activitva X Factor **

X Factor *** =

Activitv*

i Kr-87 Atmosphere 1.8(1)

N/A 26.6 4.8(2)

Xe-133 Atmosphere 1.9(3) 0.97 1.03 1.9(3)

I-131 Sump 2.6(4)

N/A 1.02 2.7(4)

I-131 RCS 6.9(4)

N/A 1.02 7.0(4) 1 I-132 Sump 4.4(4) 0.16 6.25 4.4(4) 1-132 RCS 1.2(5) 0.16 6.25 1.2(5) l l

Cs-137 Sump 2.6(3)

N/A 1.00 2.6(3) l Cs-137 RCS 6.5(3)

N/A 1.00 6.5(3)

Ba -140 Sump 4.4(4)

N/A 1.01 4.4(4)

Ba-140 RCS 1.3(5)

N/A 1.01 1.3(5) pC1/cc for atmosphere sample or uC1/gm for sump and RCS sample.

Fraction of measured specific activity due to decay of only the daughter.

Decay factor = 1/e' 1 where t = 21600 sec.

72

e where:

sample pressure = 15 psia P

=

j sample temperature = 100*F T

=

3 c ntainment pressure = 20 psia P

=

2 c ntainment temperature = 200*F.

T

=

2 CorrectionFactor=h

(

l

) = 1.1 Tables 6-3, 6-4, and 6-5 lists the corrected specific activities due to pressure and temperature differences.

6.4 ACTIVITY OF EACH MEDIUM The volume of the containment atmosphere and the mass of the sump and the primary coolant need to be known to determine the amount of Curies in each medium. Tables 6-6, 6-7, and 6-8 lists the activity of each medium.

1.

Containment Volume 3

46 0

28.3 x 10 10 V = 1.7 x 10 SCF x cc,

,(2 q6 ) = 4.7 x 10 cc 5p 2

3 where:

P2 = c ntainment pressure = 20 psia T = containment temperature = 200*F 2

P3 = standard pressure = 14.7 psia T = s an ard temperature = 32 %

3 2.

Sump Mass The sump water level monitor indicates the sump is 50% full.

For the purposes of this example, this corresponds to a water volume of 50,000 ft. The sump temperature is below 200*F and no adjustment is necessary in converting the sump volume to sump mass.

73

~

TABLE 6-3 ADJUSTED SPECIFIC ACTIVITY OUE TO PRESSURE AND TEMPERATURE DIFFERENCES Containment Atmosphere, pCi/cc Specific Activity Specific Activity Isotone From Table 6-28 Correction Factor Adiusted Kr 87 4.8(2) 1.1 5.3(2)

Xe 133 1.9(3) 1.1 2.1(3)

I 131 1 132 Cs 137 Ba 140 La 140 4

TABLE 6-4 ADJUSTED SPECIFIC ACTIVITY DUE TO PRESSURE AND TEMPERATURE DIFFERENCES Containment Sump, pC1/gm Specific Activity Specific Activity ligigp_g From Table 6-2B Correction Factor

• Adiusted Kr 97 Xe 133 I 131 2.7(4) 1.0 2.7(4)

I 132 4.4(4) 1.0 4.4(4)

Cs 137 2.6(3) 1.0 2.6(3)

Sa 140 4.4(4) 1.0 4.4(4)

• No correction is necessary since the nuclide analysis was performed on a per gram basis.

l l

l I

s

TABLE 6-5 ADJUSTED SPECIFIC ACTIVITY DUE TO PRESSURE AND TEMPERATURE DIFFERENCES RCS, pCi/gm Specific Activity Specific Activity Isotone From Table 6-28 Correction Factor

• Adiusted Kr 87 Xe 133 I 131 7.0(4) 1.0 7.0(4)

I 132 1.2(5) 1.0 1.2(5)

Cs 137 6.5(3) 1.0 6.5(3)

Ba 140 1.3(5) 1.0 1.3(5)

• No correction is necessary since the nuclide analysis was performed on a per gram basis.

76

TABLE 6-6 CONTAINMENT ATMOSPHERE ACTIVITY Adjusted Isotone Specific Activity. uCi/cc Atmosphere Volume, cc Activity. Ci Kr 87 5.3(2) 4.7(10) 2.5(7)

Xe 133 2.1(3) 4.7(10) 1.0(8)

I 131 I 132 Cs 137 Ba 140 l

77

~

TABLE 6-7 CONTAINMENT SUMP ACTIVITY Adjusted Isotope Specific Activity. uCi/am Sumo Water Mass, am Activity. Ci Kr 87 Xe 133 l

I 131 2.7(4) 1.4(9) 3.8(7)

I 132 4.4(4) 1.4(9) 6.2(7)

Cs 137 2.6(3) 1.4(9) 3.7(6) l Ba 140 4.4(4) 1.4(9) 6.2(7)

I e

73

TABLE 6-8 RCS ACTIVITY Adjusted Isotope Specific Activity. uCi/am RCS Water Mass. am Activity. Ci Kr 87 l

Xe 133 I 131 7.0(4) 2.3(8) 1.6(7)

I 132 1.2(5) 2.3(8) 2.8(7)

Cs 137 6.5(3) 2.3(8) 1.5(6) 8a 140 1.3(5) 2.3(8) 2.9(7)

8.3 x 10 cc Sump mass = 50,000 ft3,pSTP x ft

= 1.4 x 10' gm where:

pSTP = 1.00 y I

3) Primary Coolant Mass f

The primary system monitors indicate the system is full. The volume of the 3

primary system of a 312 plant is 8910 ft,

At the temperature of the RCS at time of sample (350*F)

RCS mass = 8910 ft x (2) x p TP

• S STP ft 8

= 2.3 x 10 g,

where:

(2) = water density ratio at RCS temperature (350*F), Figure 2-10

1. STP

= 0.9 pSTP

= water density at STP,1.00 gm/cc.

6.5 TOTAL ACTIVITY RELEASED The total activity released is determined by adding the activity of the atmosphere, sump, and the reactor coolant system. Table 6-9 presents the total activity released.

i 30

~.

TABLE 6-9 TOTAL ACTIVITY RELEASED Isotone Atmosphere. Ci Sumo. Ci RCS. Ci Total. Ci Kr 87 2.5(7) 2.5(7)

Xe 133 1.0(8) 1.0(8)

I 131 3.8(7) 1.6(7) 5.4(7)

I 132 6.2(7) 2.8(7) 9.0(7)

Cs 137 3.7(6) 1.5(6) 5.2(6)

Ba 140 6.2(7) 2.9(7) 9.1(7)

O Si

6.6 ACTIVITY RATIOS OF THE RELEASED FISSION PRODUCTS The activity ratios of the released fission products are shown in Table 6-10.

The use of the ratios is demonstrated in Section 6.9.

6.7 INVENTORY AVAILABLE FOR RELEASE To determine the total inventory of fission products available for release at reactor shutdown, the power history prior to shutdown needs to be known. For this example, the reactor has been operating continuously for 400 days with the following power history prior to shutdown.

20 days at 75%

power = 2175 h t 10 days at 100%

power = 2900 ht 10 days at 50%

power = 1450 ht

_5 days at 75%

power = 2175 h t 45 days The new inventories are calculated by applying the power correction factors discussed to the equilibrium, end-of-life core inventories. The following sections present examples in determining the power correction factor for this scenario. The corrected core inventories are listed in Table 6-11, 1)

Isotopes with half-lives <1 day For isotopes with half-lives less than 1 day, it is assumed that they reach equilibrium in approximately 4 days. For this scenario the reactor is operating at 2175 h t for 5 days prior to shutdown. Thus, the power correction is as follows:

Power Correction Factor =

= 0.75 0

t For I-133 (t

= 2 h),

U2 8

Corrected Inventory = 1.2 x 10 Curies x 0.75 I

= 9.0 x 10 Curies 32

TABLE 6-10 ACTIVITY RATIOS OF RELEASED FISSION PRODUCTS Isotone Total Activity. C1 Activity Ratio

• Kr 87 2.5(7) 2.5(-1)

Xe 133 1.0(8) 1.0 I 131 5.4(7) 1.0 I 132 9.0(7) 1.7 e a ctivity

• Noble Gas Ratio =

Xe-133 Activity Iodine Activity Iodine Ratio

= I-131 Activity i

l 1

83 1

~

• TA8LE 6-11 FISSION PRODUCT INVENTORY AT REACTOR SHUT 00hN Equilibrium Inventory Power Corrected Isotope at End-of-Life. Ci*

Correction Factor Inventory. Ci Kr 87 3.3(7) 0.75 2.5(7)

Xe 133 1.5(8) 0.68 1.0(8)

I 131 8.0(7) 0.68 5.4(7)

I 132 1.2(8) 0.75 9.0(7)

Cs 137 8.7(6) 0.60 5.2(6)

Ba 140 1.4(8) 0.65 9.1(7)

Inventories for a 3-Loop plant

2) Isotopes with half-lives >1 day i

Since the power is not constant during the 30-day period prior to shutdown, the transient power correction equation is applied.

I)P3 (1-e d) e-A'td

-A t Power Correction Factor =

-A'Itd)

RP (1-e For I-131 (t1/2 = 8d, Ag = 8.7 x 10 day ~I)

-2 0 93 since I t) = 45 days > 4 x

= 32 days,

-A t)) e -A t) g g

IPg 3 (1-e Power Correction Factor =

RP 2n 5 (1-e (8.7E-2)x(20)),-(8.7E-2)x(25).

2900 2900 (1-e (8.7E-2)x(10)),-(8.7E-2)x(15) 2900 1450 (1-e (8.7E-2)x(10)),-(8.7E-2)x(5) +2175(1-e (8.7E-2)x(5)),-(8.7E-2) (0)

+

2900

= 0.68

=

3)

Isotopes with half-lives around 1 year For this scenario, the core has operated for 240 effective full power days during the 400 days of cycle operation.

For Cs-137 (t1/2 "

Y"'}

Power Correction Factor =

= 0.6 0 0 55 e

6.8 PERCENTAGE OF INVENTORY RELEASED The corrected inventories are used to determine the percentage of inventory released for each isotope. The inventory released percentages are compared to Figures 2-2, 2-3, 2-5, 2-7, and 2-11 through 2-14 to estimate the extent of core damage. Table 6-12 presents the release percentages for the isotopes of this example.

6.9 CORE DAMAGE ASSESSMENT BASED ON RADIONUCLIDE ANALYSIS The results of the radionuclide analysis are used to determine an estimate of the extent of core damage. Table 6-12 shows the inventory released percentages of this accident scenario.

These percentages are compared to Figures 2-2, 2-3, 2-5, 2-7, and 2-11 through 2-14 to estimate the extent of core damage.

The fission products analyzed after the accident are Kr-87, Xe-133. I-

.1, I-133, Cs-137, and Ba-140. The noble gases, lodines, and cesium are released during all stages of core damage with Ba-140 being a characteristic fission product of fuel overtemperature and fuel melt. The calculated release of Ba-140 is used to estimate the extent of fuel temperature and fuel melt. From Figures 2-12 and 2-14 the 0.025 percent release of Ba-140 corresponds to approximately 20 percent fuel overtemperature and less than 1 percent fuel melt. Based on the Ba-140 release percentage, the fission. product release is primarily due to clad damage and fuel overtemperature.

The release percentages of the noble gases, iodines and cesium indicate from Figure 2-11 that approximately 15-25 percentage of the core has experienced overtemperature conditions. The activity ratios shown in Table 6-10 indicate that the release has progressed beyond gap release to fuel pellet release.

Comparing the release percentages of the noble gases and lodines to Figures 2-2, 2-3, 2-5, and 2-7 clad damage greater than 100 percent is indicated.

However, as stated previously, it is recognized that in actuality there is an overlap between the regimes of core damage states. Unfortunately, it cannot be estimated from the radionuclide analysis the extent of clad damage. The release due to overtemperature dominates the release due to clad damage.

36 m

• TABLE 6-12 RELEASE PERCENTAGE Corrected Release Isotone Total Activity Released. Ci Inventory. Ci Percentaae. 1 Kr 87 2.0(6) 2.5(7) 8.0 Xe 133 8.3(6) 1.0(8) 8.3 I 131 4.6(6) 5.4(7) 8.5 I 132 7.4(6) 9.0(7) 8.2 Cs 137 4.1(5) 5.2(6) 7.8 84 140 2.3(4) 9.1(7) 2.5(-2)

The conclusion drawn from the radionuclide analysis is that the core has experienced some clad damage (but the extent is not known from solely the radionuclide analysis), less than 50 percent fuel overtemperature, and the possibility of very minor fuel melt (less than 1 percent).

6.10 AUXILIARY INDICATORS To verify the conclusion of the radionuclide analysis, the auxiliary indicators (containment hydrogen concentration, core exit thermocouple temperature, reactor vessel water level and containment radiation monitor readings) are used.

6.10.1 CONTAINMENT HYOROGEN CONCENTRATIONS The containment hydrogen monitor indicated a hydrogen concentration in the containment of 10 v/o. From Figure 3-1,10 v/o hydrogen concentration corresponds to approximately 75 percent zirconium water reaction. Thus, the hydrogen concentration indicates that there is a high probability that greater than 50 percent of the clad is damaged. Table 4-1, 6.10.2 CORE EXIT THERMOCOUPLE REA0!NGS AND REACTOR VESSEL WATER LEVEL The core exit thermocouple readings during this accident reached 1650'F for the center half of the core and ranged between 900*F to 1100*F for the outer regions of the core. The reactor vessel water level monitor indicated that the core uncovered during the accident for an extended period of time. From Table 4-1, these readings indicate a possibility of the core experiencing fuel overtemperature in the center regions and clad damage in the outer regions.

Also, the high hydrogen concentration measured in the containment confirms that the core had uncovered during the accident.

6.10.3 CONTAINMENT RA0!ATION MONITOR The containment radiation monitor indicated a gross gamma dose rate of 1.02 x 10" R/hr at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> af ter reactor shutdown. To use Figure 3-3, the dose rate is normalized based on power and containment volume.

It is assumed 33

that the only difference between this plant and the plant of Section 3.3 is 6

3 the containment volume. This plant has a volume of 1.7 x 10 ft. The normalized dose rate is calculated as follows:

4 6

3 Dose Rate (Normalized) = 1.02 x 10 R/hr x 1.7 x 10 ft 6

3 2900 W t x 2 x 10 ft

= 3 R/hr - Wt From Figure 3-3, 3 R/hr-W t corresponds to an overtemperature release and a significant gap release which confirms the radionuclide analysis.

6.11 SUMARY The combination of the radionuclide analysis and the auxiliary measurements indicated greater than 50 percent clad damage, less than 50 percent fuel overtemperature, and a possibility of very minor fuel melt.

This example was provided to illustrate the use of this core damage assessment methodology in determining the extent of core damage. Although this example was for a Westinghouse 3-1.oop plant, the methodology can be applied to the other Westinghouse pressurized water reactors.

39 A

7.0 REFERENCES

1.

" Clarification of TMI Action Plan Requirements," NUREG-0737, USNRC, November 1980.

2.

"A Report to the Connission and to Public NRC Special Inquiry Group,"

M. Rogovin, 1980.

3.

"0RIGEN !sotope Generation and Depletion Code " 04k Ridge National Laboratory, CCC-217.

4.

Method of calculating the f ractional release of fission products from oxide fuel, ANS!/ANS 5.4 - 1982.

5.

WCAP-9964 Westinghouse Electric Corporation.

6.

" Source Term Specification " ANS 18.1 Standard 1976.

7.

"Radionuclide Release Under Specific LWR Accident Conditions," Draft NUREG-0956, USNRC, January 1983.

8.

" Release of Flssion Products From Fuel in Postulated Degraded Accidents,"

10COR ORAFT Report, July 1982.

9.

"TMI-2 Accident: Core Heat-up Analysis," NSAC/24, January 1981.

10. " Light Water Reactor Hydrogen Manual," NUREG/CR-2726 August 1983.

11. Westinghouse Emergency Response Guidelines.

12. Analysis of the Three Nile Island Accident and Alternative Sequences, Prepared for NRC by Battelle, Columbus Laboratories, NUREG/CR-1219.

10

,.