ML20083N896

Rev 1 to Procedure RTP-46, Core Estimation Damage, Post-Accident Sampling Sys

ML20083N896
Person / Time
Site:	FitzPatrick
Issue date:	03/23/1984
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20083N885	List: ML20083N882 ML20083N896 ML20083N900
References
RTP-46, NUDOCS 8404190323
Download: ML20083N896 (28)

Text

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f a l G@l l !! "' L' ' 'l ( J)ej NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT f

RADIdLOGICAL AND ENVIRONMENTAL SERVICES DEPARTMENT I

i PROCEDURE NO.: RTP-46 i

TITLE: CORE DAMAGE ESTIMATION, PASS i

PORC Review No./Date Meeting No.: N/A Date J3 8f Approved By: s Resident Manager Approved By: 4 Radiological and Environmental ~

7 Services Superintendent

[ .

t i Page No.: 1- 1 2 3 4 5 6 7 8 9 Rev. No.:

1 1 1 1 1 1 1 1 1 1 Page No.: 10 11 12 13 14 15 16 17 18 19 Rev. No.: 1 1 1 1 1 1 1 1 l' 1 Page No.: 22 21 22 23 24 25 Rev. No.: 1 1 1 1 1 1 uRev. No.: 1 Date: 03/84 8404190323 840413 -

PDR ADOCK 05000333

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RTP-46 j CORE DAMAGE ESTIMATION, PASS I

h: Table of Cont 2nts i

Page go

[ 1.0 Purpose 1 f 2.0 Requirements and Acceptance Criteria 1 E-3.0 Special Equipment 1 4.0 Procedure 1 4.1 Plant Parameters 1

'4.2 Estimation Procedure Preliminary Guides 2 f 4.3 Isotope Decay Correction 3 4.4 Gaseous Sample Temperature and Pressure Correction 3 4.5 Fission Product Inventory Correction 4 4.6 Plant Parameter Correction Factor 4 t.;

4.7 Normalized Isotopic Concentrations 5~

I 4.8 Interpretation of Normalized Isotopic Information 5 f 4.9. Identification of Release Source, Core or Fuel ~ Gap 5

. 4.10 Integration of Other Parameters into the Estimate 6 4.11 Development of_a Final Estimate 7 5.0 References '8 6.0 Attachments 8 i

.j ~

.w-Rev.-No.-. 1 LDate- 03/84~ Page "i x , .

.=

f .

RTP-46 i CORE DAMAGE ESTIMATION, PASS I

l 1.0 PURPOSE l

The purpose of this procedure is to estimate the degree of

' reactor core damage using the measured fission product con-l centrations in either the water or gas samples taken from l the primary system under accident conditions. The procedure

j. involves calculations of fission product inventories in the i

• core and the release of inventories into the primary system under postulated Loss Of Coolant Accident (LOCA) conditions.

I The fuel gap fission products are assumed to be released upon the rupture of fuel cladding. The majority of fission product inventories in the fuel rods would be released when the fuel is melted at higher temperatures. The estimation i[ of core damage will be calculated by comparing the measured concentrations of major fission products in either gas or L-liquid samples, after appropriate normalization, with refer-ence plant data from a BWR-6/238 with a Mark III contain-g ment.

2.0 REQUIREMENTS AND ACCEPTANCE CRITERIA 4 2.1 Technical Specifications

.None i

2.2 Plant-

' Compliance with NUREG-0737, Item II.B.3 3.0 SPECIAL EQUIPMENT

~None-

[

l 4.0 PROCEDURE'-

4.'l- Plant Parameters 4.1.11 -The' pertinent plant' parameters for the reference plant and the FitzPatrick Plant are given below:

Reference. FitzPatrick Plant Plant RatedLReactor Thermal Power 3579'MWt' '2436 MWt1 Number of~ Fuel-Bundles 748 Bundles 560_ Bundles

" Total. Primary Coolant Mass' 3.92E+9:g= '3.21E+9.g

.(Reactor Water'plus Suppres-

"sion: Pool Water)-

Total ~ Torus 1 Containment and ! 4.0E+10-cc- 7.57E+9 cc-Drywell Gas =SpaceLVolume: '

Rev.ENoe 1 Date

03/84 _

Page ~1

. 1, e 9 x = .- ~ a

4 f1

4.1.2- Fission product inventories in the primary system of the 1

reference plant were calculated based on postulated I.0CA

• conditions after three years (1,095 days) of continuous I}

operation at 3,651 MWt, or 102% of rated power, by using a

' computer code developed at Los Alamos and adapted to the GE computer system. The inventories of some major fin; ion products in the core at the time of reactor shutdown are

?- given in Table 1.

t 4.1.3 The inventories of some major fission products in the core 4 of the FitzPatrick Plant at the time of reactor shutdown are given in Table 2.

4.2 Estimation Procedure Preliminary Guides 4.2.1 Obtain samples from the Post Accident Sampling System (PASS), following the procedure outlined in PSP-17.

a. It is recommended that both the water and gas phase samples be taken and analyzed in order to reduce the uncertainty in core damage estimations.

f

b. Samples acquired for the estimation of core damage

" should be taken from locations that are consistent I

with break case and system conditions, as outlined in Table 3. This will ensure the viability of results i reported and provide the best estimation. of core damage.

l 4 2.2 Perform gamma ray spectrometry on -the samples and deter-g mine.the concentration,.in uCi/g, of a fission product 1, as outlined in CAP-41.

, a. In water, the concentration is represented as Cw, and i the: recommended isotopes of interest are I-131 and Cs-t- 137.

b. In'~ gas , the concentration :is represented -. as Cg, and the - recommended isotopes "of interest are Xe-133 and Kr-85.

c. In case the - fission product > concentrations ' are mea-sured' separately for the reactor water and suppression pool: water-or the.-drywell' gas and.the torus gas, the-measured concentrations - Cw or Cg '. would ~ be averaged from the separate measurements:

y .-

(Conc.-in Rx Water)-(Rx Water Mass)-

Cw = Reactor: Water Mass +. Pool Water Mass lRev.~No. L1l LDate 03/84- Page 2

~. ,

'j , (Conc. in Pool) (Pool Water Mass) q Reactor Water Mass + Pool Water Mass

'A .

(Conc. in Drywell)(Drywell Gas Vol)

Cg = Drywell Gas Volume + Torus Gas Volume

h l

4 (Conc. in Torus)(Torus Gas Vol) 1 ~ Drywell Gas Volume + Torus Gas Volume

,?

hj 4.3 Isotope Decay Correction 4.3.1 Supply the counting room technician time of shutdown for

%j automatic isotope decay correction or else decay as in next step.

4.3.2 Correct the measured concentration (Cw or Cg) for decay to the time of reactor shutdown, using the following equa-tions:

Cd = Cw e I-Ai E) or Cd = Cg e(-Ai t) 7 j Where:

Cd = the corrected concentration (C/g).

y!

Ai = the decay constant of isotope i (day 1) (given in Table 4).

G 1 t = the time between the reactor shutdown and the j sample time (day).

[ e (- Ai' D} ' = the decay correction to the time of reactor r shutdown.

1 4.4 Gaseous Sample Temperature and Pressure Correction

-4.4.1 Correct the gaseous activity concentration for temperature and pressure differences between the sample vial and the containment gas ' phase, if significant difference exists L

j between sample and containment condition,.using the fol-lowing equation:

Cdc = Cd.x- h1 Where:

Cd = containment isotopic concentration (pCi/cc).

-(P1,T1) = sarple vial pressure and temperature -(psia,*R).

(P2,T2) = containment pressure and temperature (psia,*R).

~Rev. No.

.. 1 Date 03/84 .Page 3

j f[

1 4.5 Fission Product Inventory Correction 4.5.1 Calculate the fission product inventory correction factor FIi for each isotope of interest, using the following equation:

,I

j. FIi = Inventory in Reference Plant Inventory in FitzPatrick Plant

}l , 3651 (1-e-10 9 5A1) i E Pj (1-ey - Ai Tj s)e(- Ai Tj )

I I -

Where:

l FIi = inventory correction factor for isotope i.

t Pj = steady reactor power operated in period j (MWt) i Tj = duration of operating period j (day).

Tj0 = time between the end of operating period j and r- time of the last reactor shutdown (day).

4.5.2 Pj , Tj , and the time of reactor shutdown for each oper-ating period j, excluding the operating period just concluded, are provided in Table 5.

4.5.3 Pj , Tj , and the time of reactor shutdown for the most re-cent- operating period must be obtained. Tj 0 must be calculated for each operating period.

4.5.4 i'he information given in Table 5 should- be updated to include the most recent operating period following each reactor shutdown.

~

4.5.5 For a particular short-lived isotope i, a calculation for only a period of + six half-lives of reactor operation time before reactor shutdown should be accurate enough.

4.5.6 The ' correction factor calculated' from this equation may not be entirely accurate, but the error is insignificant in comparison to the uncertainties in'the fission product release fractions (Table 6) and other assumptions.

4.6 Plant Parameter Correction Factor 4.6.1 Calculated plant parameter. correction factors were devel-oped using the equra'ons below:

p , , FitzPatrick'PI. .t Coolant Mass (3.21E+9 g)

Reference Plant Coolant Mass (3.92E+9 g)

Rev. No. 1 Date 03/84 Page 4

f

+

I j " FitzPatrick Plant Containment Gas Volume (7.57E+9 cc)

Reference Plant Containment Gas Volume (4.0E+10 cc) h k Ehere:

Fw = primary coolant mass correction factor.

h. Fg = containment gas volume correction factor.

V.

r k

These operation:

equations reduce to the following for normal plant f

f Fw = 0.819 (AND) Fg = 0.189 4.7 Normalized Isotopic Concentrations Calculate the normalized concentrations, Cnw and Cng using the equations below. These are the concentrations in the reference plant equivalent to the concentrations in the FitzPatrick Plant.

i Cnw = Cd x F x Fw (OR) Cng = Cdc x F x Fg 4 '. 8 - Interpretation of Normalized Isotopic Information i

4.8.1 If the normalized concentrations, Cn are higher than the

baseline concentrations shown in Table 7, the extent of fuel or cladding damage'(or both)'can be estimated direct-ly from Figures 1 through - 4. = These are graphs of each isotope's concentration versus % cladding failure and ~ % -

fuel meltdown.- They ' yield a best estimate for core dam-f age, as well.as-a' range of possible: values.

' 4.8.'2 If the normalized concentrations fall into a range where release of the fission - producti from the fuel gap or the -

molten fuel.cannot be definitely. determined, the. presence'

-of Sr,.Ba,.La and Ru should beLestablished'. Fission: prod-ucts'27 hr-Sr-92 (1.385 MeV) and-40 hr La-140 (1.597:MeV) are relatively easy: to identify and measure Jfrom a , gamma ray; spectrum and !are L indicative of L fuel . meltdown. 'These results.should be' compared to baseline reactor ~ water con-

~

centrations.

~ 4.9: Identification of Release Source, Core'or Fuel Gap 4;9.1 Fromi the samples obtained ~ using L the: PASS , . determine ~ the.

concentrations : 'of the following short-lived isotopas by --

,  : gamma spectroscopy: <

y P '

Rev. No. 1: L Datef 03/84' Page: 5-af

s

.i{ .

1 y

2 i Kr-87 i Kr-88 j

Kr-85m Xe-133 1-134 -

I-132 I-135

)[ I-133 E-I-131 q 4.9.2 Correct the measured fission products to the time of reac-5-

V tor shutdown, using the.following equation:

!> Ci = Ci(sample)e ~Ai U Where:

i = the isotope measured. '

Ci =

the concentration corrected for the time of reactor shutdown (uci/g).

Ci (sample) =

the measured concentration in the sample (uci/g).

Ai = _ the decay constant of isotope 1 (day 1)~

(given in Table 4),

t -=

the time between the reactor shutdown . and the sample time (day).

[ 4.9.3 Calculate the isotopic activity ratios from the following l equations:

Noble gas.rctio = Noble Gas Isotopic Concentration Xe-133 Concentration Iodine. ratio = I dine Isotopic Concentration I-131 Concentration 4.9.4 ' Compare these. ratios to the' ratios supplied in Table 8 to determine the release source-(the core or the fuel gap).

Fuel _ cladding rupture is-assumed if the source.is the fuel gap,'and some core melting.istassumed if;the' source:is the c o r e ...

4.10 Integration of'Other Parameters into the Estimate ~

i .

These methods areoonly outlined briefly-here,'and'no exact procedure for estimations is provided. Also, these methods are used for'the confirmation of estimates obtained earlier H- cin this procedure, and not'to make original estimates..

Rev. No.. I Date. :03/84 .Page 6~

1 e

1 4

l

{. -

l t 4.10,1 Containment Radiation Levels The. level of containment radiation is an indication of 4

the inventory of airborne fission products (i.e. noble gases, a fraction of the halogens, and a much smaller

[ fraction of the particulates) released from the fuel to the containment, and as such, and indication of the degree of core damage sustained.

{ 4.10.2 Reacter Vessel Water Level t

Reactor vessel water level readings indicating signifi-cant periods during which the core is uncovered would mean core damage is likely.

Bulk core damage situations could be ' caused by loss of coolant to the entire core, while localized core damage situations could be caused by a flow blockage to some part of the core.

4.10.3 Main Steam Line Radiation Level

[ -High main steam line radiation levels indicate some core

) damage may have occurred. The.usefulness of the method '

! is limited, however, because the main steam line radia-tion monitors are' downstream of the main steam isolation valves and~would be unavailable following vessel-isola-tion.

4.10.4 Reactor Vessel Pressure High reactor vessel pressure - may indicate a core damage event has . occurred. - -This indication is ambiguous, how -

ever, as there are manya non-degraded core events which could also produce'a high reactor vessel pressure.

4.10.5 . Containment. Hydrogen Concentration-Hydrogen - concentrations may be ~ obtained .~ from either the containment hydrogen monitors o'r- - from the PASS sample analysis. Curves have ~ been'. developed 1which relate this concentration, after appropriate normalization, to 'the

% metal-water. reaction undergone, and- thus to .the %

cladding failure" sustained.

'4'.11 Development ' of a Final Estilmate

.4.11.IL From.the estimates developed . -in seetion - 4.8, ' assign 'one

or morel categories:of core damage to the. core from those listed'in. Table 9.

4.11.2 ;Using the-release' sources identifiednin section:4.9, nar-

~ '

crow down:the:. range of categories assigned'above as far.as 1 -

[possible.,

4 Rev.:No.- 1 :Date 03/84 TPage 7 .

,~__ j_

a l

4.11.3 Use any or all of the methods summarized in section 4.10 j to assign the final core damage estimate. More than one j category may be assigned (For example, in order for fuel 5

5 melt to occur, some fuel overheat and cladding failure must have occurred),

f

5.0 REFERENCES

t c

5.1 NEDO-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions, August, 1982.

, 5.2 Memo RES 83-0279, NUREG-0737 Item II.B.3, BWRDG-8324 (June

17, 1983) Attachment 2,

' Integration of Other Plant Parame-ters into Core Damage Estimate.

5.3 PSP-17, PASS Operating Procedure.

5.4 CAP-41, Post Accident Sample Analysis.

6.0 ATTACHMENTS 6.1 Figure 1, Relationship Between I-131 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant.

]1 6.2 Figure'2, Relationship Between Cs-137 Concentration in the Primary Coolant-(Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant.

6.3 Figure 3, Relationship Between Xe-133 Concentration in the g Containment Gas (Drywell + Terus Gas) and the Extent cf Core i

Damage in the Reference Plant.

6 .- 4 Figure 4, Relationship Between Kr-85 Concentration in the C

Containment Gas (Drywell + Torus Gas) and the Extent of Core Damage in Reference Plant.

b

6.5 Figure-5, Core Damage Estimation Data. Sheet.

I r

6.6 Table 1, Core Inventory of -Major Fission Products in a Ref-erence Plant Operated at 3651 MWt for Three Years.

6.7 Table 2, Core Inventory of _ Maj or Fission Products in the FitzPatrick Plant Operated.at 2436 MWt for Three Years.

6.8 Table 3, Samples Most Representative of Core Conditions Dur-ing an Accident for the Estimation'of Core Damage'.

l 1 6~9-Table 4, Decay Constants of Some Radioactive Isotopes, j- 6.10 Table 5, Fission Product. Inventory Correction Data.

6.'11 Table 6, Best-Estimate Fission Product. Release Fractions.

'Rev. No. 1 Date 03/84 Page 8'

s.- -

1 ,

il

'y . .6.12 Table 7 Fission Product Concentrations in Reactor Water and m

Drywell Gas Space During Reactor Shutdown Under Normal Con-

-ditions.

?

1 f 6.13 Table 8, Ratios of Isotopes in Core Inventory and Fuel Gap.

1 6.14 Table 9, Categories of Core Damage Events. '

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L Rev. No. 1- Date 03/84 -Page 12 _ ,,

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Rev. No. 1 Date 03/84 Page 13 DM8-

hTP so 1 --

Figure 5 4

j; CORE DAMAGE ESTIMATION DATA SHEET n .

,, . 1. Name Date

[..

P t'

Date of Shutdown t-

n Time of Shutdown l, ,

2. Isotopic Information

a. _ Liquid Sample Identification:

Sample Date and Time:

Sample decayed to Shutdown YES or.NO (circle one)

I '131 pCi/g i

Cs - 137 uCi/g

[ b. Gaseous Sample Identification:

l Sample Date and Time:

l Sample decayed to Shutdown YES or NO (circle one)

Xe - 133 pCi/cc Kr - 85 pCi/cc Sample Vial Temperature R( F + 460)T1 =

Pressure psia (psig +'14.7)Pl.= , ,

Containment Temperature R( F +460) T2'=

Pressure psia (psig + 14.7)P2_ =

c. Isotopic Data-Decayed to Shutdown Cd ; I -J131. (Liquid) f uci/g a

Cd ; Cs 137 (Liquid) pCi/g Cd~;'Xe- 133 (Gas) pCi/cc Cd;; Kr- ~85 (Gas)'.pci/cc-Rev. No. 1 :Date 03/84 Page 14 e *:e r + e 's,_ *w

'. Figure 5 (con't) a .

B L y:

d. Fission Product Inventory Correction Factors Days between p Operation Days of period S/D and Average O Period Operation current S/D Power MWTh h.

1 448 670(to 9/1/83) 1985 2

3A 455 89 (to 9/1/83) 2087 ll B C

{

s D

[

F( I - 131)

F(Cs -137) i F(Xe - 133) t F(Kr - 85) t e. Plant Parameter Correction Factor Fw = 0.819 and Fg = 0.189

f. Gaseous Temperature and Pressure Correction j Cdc = Cd x.(P2 x Tl) * (P1 x T2) b Cde Xe - 133 pCi/cc =

{ Cdc Kr - 85 pCi/cc =

g. Ncrmalizing Isotopic Data-y Cnw = Cd x F-x Fw Cng = Cdc x F x Fg Cn(I - 131). = Cd x F(I - 131) x 0.819 x -x 0.819 = pCi/g Cn(Cs- 137)- = Cd x F(Cs -137) x 0.819 -

x. ~ x 0.819 = uCi/g-LCn(Xe- 133)'= Cdc x F(Xe-133) x'O.189-x x?O.189 =. pCi/g Cn(Kr 85) = Cdc'x 7(Kr- 85) 0.189-

___ X X 0.189 - 'uci/s.

I Rev. No. 'l "

Date 03/84 .Page ~15 9 ,

L__ . - _

wir-90

~i

'. Figure 5 (con't)

.L

$ 3.. Interpretation of Normalized Isotopic Data

a. I - 131

[ .

1. Within normal range 0.7 + 29pC/g YES or NO p.

4 2. Cladding Failure, % Failure k

[

Upper Best Lower

3. Fuel Meltdown, % Meltdown Upper Best Lower

b. Os - 137 I 1. Within normal range, 0.03 to 0.3 pCi/g YES or NO i

j 2. Cladding failure, % Failure

{ Upper Best Lower

3. Fuel Meltdown, % Meltdown Upper Best Lower

c. Xe - 133

1. Within normal range, lE-5 to lE-4 pCi/cc Y or N

2. . Cladding failure, % Failure t Upper Best Lower 3 3. Fuel meltdown, % meltdown Upper Best Lower

d. Kr --85

1. Within normal. range, 4E-6 to 4E-5 YES or.NO 2'. . -Cladding failure, % Failure

. Upper- Best. Lower

3. Fuel-Meltdown,'% Meltdown

Upper 3est- Lower

.Rev. No. I' ,

Late. 03/84' Page 16 . -

m 9 g >

,4_ ._ _ _ a

RTP-46 t .

o Table 1 L

!. CORE INVENIt)RY OF MAJOR FISSION PRODUCIS IN A

, REFERENCE PLANT OPERA 7ED AT 3651 MWt FOR 7IIREE YEARS I

s MAJOR GAMMA RAY ENTRGY r INVENTORY (IhTENSITY)

, CHEMICAL GROUP ISOTOPE

• HALF-LIFE 108 Ci Kev (y/d) g .

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

, Kr-85 10.72y 1.1 514(0.0043)

Kr-87 76. m 47.1 403(0.494)

Kr-88 2.84h 66.8 196(0.203).1530(0.109) l Ie-133 5.25d 202. 81(0.371)

le-135 9.09h 26.1 250(0.906)

! Halogens I-131 8.04d 96. 364(0.824) i I-132 2.29h 140 668(0.99),773(0.762) l I-133 20.8 h 201 530(0.87) f I-134 52.6 m 221 847(0.954),884(0.653) i, I-135 6.59h 189 1132(0.2;1),1250(0.293)

) Alkali Metals Cs-134 2.06y 19.6 605(0.98),796(0.88)

Cs-137 30.177 12.1 662(0.85)

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

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

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

Ru-103 39.4 d 155 497(0.9) i Alkaline Earths Sr-91 9.52h 115 750(0.24)

Sr-92 2.71h 123 1385(0.9)

Ba-140 12.8 d 173 537(0.238)

Eare Earths T-92 58.6 d 118 934(0.137)

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

Co-141 32.5 d 161 145(0.49)

Ce-144 284.4 d 129 134(0.108)

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

Zr-97 16.8 h 166 743(0.933)

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

• • 1 kr after shutdown Rev. No. 11 Date 03/84- Page 17

. . RTP-46 f Tabic 2 I

f CORE INVENTORY OF MAJOR FISSION PRODUCTS

{ IN THE FITZPATRICK PLANT OPERATED AT 2436 MWt FOR THREE YEARS L

Inventory Chemical Group Isotope T1/2 (106 Curies)

{ ,

I h Noble Gases Kr-85m 4.48h 16.4 f Kr-85 10.72y 0.73 i Kr-87 76.30m 31.4 l Kr-88 2.84h 44.6

! Xe-133 5.25d 134.8 Xe-135 9.11h 17.4 e

Halogens I-131 8.04d 64.1 I-132 2.30h 93.4 I-133 20.80h 134.1

- I-134 52.60m 147.4 g I-135 6.61h 126.1

{

Alkali Metals Cs-134 2.06y 13.1 Cs-137 -30.17y 8.1 Cs-138 32.20m 1973.

Noble Metals Mo-99 -

66.02h 122.1 Ru-103 39.40d 103.4 Alkaline Earths Sr-91 9.50h 76.7 Sr-92 2.71h 82.1 Ba-140 12.8 d 115.4

.~ . _

Rare Earths Y-92 58.6 d 82.7 La-140- 40.20h 122.8 Ce-141 32.50d 107.4 Ce-343 284.30d 86.1 Refractories Zr-95 64.00d 107.4 Zr-97 16.90h 110.8 Rev. No. I' Date. 03/84' Page 118 E '- --

. __ semM;,v,A%rannan:amem~ +mwsvatulagwMM%'escanasuaswsuwswu%w L.GLF-~

SAMPLES MOST REPRESENTATIVE OF CORE CONDITIONS DURING AN ACCIDENT FOR THE ESTIMATION OF CORE DAMAGE P? SAMPLE LOCATION

< Supp. Supp.

Break Category / System Conditions .Te t Pool Pool Other Instructions

?: Pump Liquid Atmos. RHR Drywell i l l i

l Small Liquid Line Break. Reactor 1 2 Power 21% Yes ---

Yes ---

Yes s.

Small Liquid Line Break. Reactor 1 2 Power <1% --- ---

Yes Yes Yes a. RHR must be in shutdown cool-ing mode.

b. Reactor water 1cvel must be raised and flow from moisture separators.

cy Small Steam Line Break. Reactor 1 2 W Power 21% Yes ---

Yes ---

Yes H n:

Small Steam Line Break. Reactor 1 2 Power <1% ---

Yes Yes Yes a. RHR must be in shutdown cool- g" i C} ing mode. ,

as s

oo b. Reactor water level must be raised and flow from moisture l

f separators.

Large Liquid Line Break, Reactor 3 4 1 2 Power 21% Yes Yes Yes ---

Yes a. Suppression pool must be in suppression cooling mode.

Large Liquid Line Break, Reactor 4 3 2 Power <!% ---

Yes Yes Yes Yes a. RHR must be in chutdown cool-ing mode.

??

!3 b. Suppression pool must be in suppression cooling mode.

s. c. Reactor water level must be

") raised and flow from moisture separators.

Superscripts on the Sample Location indicate system sample order of preference.

^

. y;; . -; .-@ _;.;.l;% .;i.t. ,:ykg.;ff.;@ 4.e:. gg;/.g;.gw .; L.Cs ', iqR n +: .- Q ' .: %.} '  ;+.:, -Q yQ' W -J.. x i J. .'. . ;,

. , p e.

. . c-,

. . , t.:2 b;

.. ., ,m 4

. n.e

,.. t

'~i * ,j , RTP-46 ..!r Table 4 /;-

. * :..#. -3+j

y

- .- p.

5[, DECAY CONSTANTS S":

Q .?" ; 0F SOME RADIOACTIVE ISOTOPES h

t. - bt 7 ,

.' .- f l <{

- . .y

. .:. + . .

'e 3 .T Decay Constant M

?>

..'_'[

.l...

g Isotope (Ai) (day .1) g 4 <. f e.

e- =

I-131 8.62E-2 E

.4 h ' h~ 'M

. 'c ..

I-132 . 7.23 02l S.

g m . ,io <. 4 I-133 8.00E-1 6

-.? I*

sp ?-

' t* <., f I-134 1.90E+1 af

'.f. .., ..4 ; - o-V 4 l['

, 2, I-135 2.52 D.s

.f f Cs-137 6.29E-5

,. 3

%.O p Xc-133 s)

. . ., - 1.32E-1 .b.'

.c x - . m-

. . c .. ;;

g? .

Kr-85 1.77E-4 [/h e 't:h, C .;q Kr-87 1.31E-1 y/ ~

kp. .a Kr-88 5.86 k

, i. . k 'i 4 ,

/1,

.M Kr-85m 3.71 ,,4

-t

, w.i-

. .; e'J's l,l _*; _

.:l

' . . */.: s -

. ~. lf r.

n. .h,.

..+ .

$.% N tx

.gv, j' 4'<.,

y Mr~

u. ) Q I.

j 4*

s ,c,

• s F .; ..

?e

. .i . T' "f 4 5, E~ 9

.,. Y.: . '.

..-  ?

.Y',.. _,5 n,, s t -

a-

. **  : 80-v ,

. s. l ,

Rev. No. 1 Date 03/84 Page 20 Tj ,

l +';.

'.9.#

' . 4 "<. ' [.. . k - - .1;[ - t, , e .,-/ .

.,.1,_ - -' -

,4 s, ."a

, - - 'M . '.',s

.3*..';'

,s ' ', - +'9 . '"

i l}

RTP-46 *

)

) Table 5

) .1 iA

} FISSION PRODUCT INVENTORY CORRECTION DATA Il i,}

3 Operating

), Period Pj (MWt ) Tj (day ) Time of Shutdown j'

5 1 1985 448 10/31/8] (670 days to 09/01/83) j 2 2087 455 06/04/83 ( 89 days to 09/01/83) i d

j 1

i t

i A

T J

l Rev. No. 1 Date 03/84 Page 21

.f

m- -m-==._,.

-- e 1 %

1

8 (1

4 2 BEST-ESTIMATE FISSION PRODUCT RELEASE FRACTIONS Y

Gap Release Meltdown Release Oxidation Release Vaporization Helease I.ower Upper I. owe r Upper I. owe r Upper Lower Upper Nomin.11 I.imit I.imit Nominal Limit I.!mit Nominal Limit Limit Nominal Limit Limit Noble Cases 0.030 0.010 0.12 0.873 0.485 0.970 0.087 0.078 0.097 0.010 0.010 0.010 (Xc,Kr)

. Italogens 0.017 0.001 0.20 0.885 0.088 0.010 0.492 0.983 0.078 0.098 0.010 0.010 (1.Br)

N Alkali Metals 0.050 0.004 0. ~10 0.760 0. ~180 0.855 --- --- ---

0.190 0.190 0.190

'O (Cs.Rh) pg tr m Tellurium Group 0.0001 3x10- 0.04 0.150 0.05 0.250 0.510 0.340 0.340 0.340 8 (Te.Se,Sh) 0.340 0.680 7i*

m > m ,

ao

• • Noble Metals --- --- ---

0.030 0.01 0.10 0.87'l 0.776 0.970 0.005 0.001 0.024 s (Ru,Rh,Pd Mo,Tc)

Alkallne Earties lx10' 3xt0- 0.0004 0.100 0.02 0.20 --- --- ---

0.009 0.002 0.045 (Sr,Ra)

Rare Earths --- --- ---

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

0.010 0.002 0.050 (Y,La.Cc.Nd.Pr.

Eu,Pm.Se,Np,Pu)

'}

, Refractories --- --- ---

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

Do u (7.r Nb) 2 N

N

g.. -

RTP-46 f.

p Table 7 s

k 1'

,h .

, FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER AND DRYWELL GAS SPACE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS Reactor Water, uCi/g Dryvell Gas (uC1/ce) 4 Isotope Uppe:r Li=it Nominal Upper Limit Nominal I-131 ,

29 0.7 -- --

b Cs-137* 0.3 8 0.03 ,_ ,_

~ ~

Ie-133 -- -- 10 ' 10

-5a -6b j, Kr-85 -- 4x10 4x10

'O.bserved ' experimentally, in an operating 3WR-3 with HK I containment, data

! obtained from GE unpublished document DRF 268-DEV-0009. ,

j Assu=ing 10% of the upper 11=1e values.

" Release of Cs-137 activity would strongly depend on.the core inventory which is a function of fuel burnup.

t

.Rev.-No.. 1 Date 03/84 Page 23

=

6 RTP-46 Table 8

)A n

)

,, RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL CAP k

Activity Ratio

• in Activity Ratio

• in j Isotope Half-Life Core Inventorv Fuel Cap k

Q Kr-87 76.3 m 0.233 0.0234 3 Kr-88 2.84h 0.33 0.0495 y

- Kr-85m 4.48h 0.122 0.023

.t -

Xe-133 '5.25d 1.0* 1.0*

j l'

I-134 52.6 m 2.3 0.155 E I-132 2.3 h 1.46 0.127 4

I-135 6.61h 1.97 0.364

{

-) 1-133 20.8 h 2.09 0.685

! I-131 8.04d 1.0* 1.0*

r i

-j

-

• Ratio =

n e gas isotope concentration I for noble gases

' t Xe-133 concentration i-j ,

Iodine isotope concentration for iodines 1-131 concentration

=

e

=

b i

i Rev. No. 1 Date 03/84 Page 24

l O  ;

i a

RTP-46 j Table 9 L

M y ' CATEGORIES OF CORE DAMAGE EVENTS I

i -

Degree of Minor Intermediate Major

[ Degradation (<10%) (10% - 50%) (>50%)

3 No Fuel Damage 1 +

Cladding Failure 2 3 4 Fuel Overheat 5 6 7 I Fuel Melt' 8 9 10 i

J.

/

4 s

Rev..No. 1 Date 03/84 Page 25

' Q L_ L T ___i' $ _ :__ --. -

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)

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1

'I i

l NEW YORK POWER AUTHORITY i JAMES A. FITZPATRICK NUCLEAR POWER PLANT JPN-84-23 ATTACHMENT C.

4

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