ML20054C311

Forwards Response to SER Item 1.11(17) Re post-accident Sampling & Draft Chemistry Procedure for Estimating Core Damage

ML20054C311
Person / Time
Site:	Grand Gulf
Issue date:	04/15/1982
From:	Dale L MISSISSIPPI POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0831, RTR-NUREG-831 AECM-82-153, NUDOCS 8204200343
Download: ML20054C311 (19)

Text

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MISSISSIPPI POWER & LIGHT COMPANY Helping Build Mississippi 4 EmitalladiddB P. O. B O X 16 4 0, J AC K S O N. MISSISSIPPI 3 5 yV N ,

April 15, 1982  ;' ((

NUCLEAR PRooUCTioN DEPARTMENT U. S. Nuclear Regulatory Commission f "' l Office of Nuclear Reactor Regulation {Ul . .

%,s Washington, D. C. 20555 ,

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Dear Mr. Denton:

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SUBJECT:

Grand Gulf Nuc1 car Station / %'i'"

Units 1 & 2 Docket Nos. 50-416 and 50-417 File 0260/L-334.0/L-350.0 Response to SER ltem 1.11(17)

AECM-82/153 Attached are responses or clarifications pertaining to issues discussed in either the Grand Gulf Nuclear Station Safety Evaluation Report (SER),

NUREG-0831, or the Final Safety Analysis Report (FSAR). The attachments address the final item pertaining to question 281.9 on post-accident sampling.

The attached procedure represents Mississippi Power and Light Company's final draft of the chemistry procedure for Estimation of Core Damage which was hased on the guidance provided by the Chemical Engineering Branch.

In addition, the last paragraph of NRC question 281.9 asked for documentation demonstrating the applicability of analytical procedures. At Grand Gulf, chemical parameters will be monitored by performing the grab sample analyses in the method as identified in the General Electric document NEDC-24889 entitled, " Post LOCA Sample Statistics-Compilation of Technical Information." It is our understanding, based on conversations with your Mr. F. Witt, that the analytical procedures identified in NEDC 24889 have been approved for use, therefore, no further documentation concerning the applicability of analytical procedures is deemed necessary.

If your generic review determines that the procedures and or nethods of l specific procedures are proven to be unacceptabic, please notify this office of your findings and the appropriate corrective actions will be made. As discussed with your Mr. F. Witt (CEB) and Mr. R. Phares of our staff on April 14, 1982, this concludes the information necessary to close the Grand Gulf Safety Evaluation Report, NUREG 0831, (SER) Iten 1.11(17) Post-accident sampling capability.

Should you have any questions or require additional information, please contact this office.

Yqprsftruly, 00 I 8204200343 820415 PDR ADOCK 05000416 d  %

E PDR / L. F. Dale j Manager of Nuclear Services RFP/JCC/JDR:1g [

Attachment cc: (See Next Page)

Member Middle South Utilities System

AECM-82/153 Page 2 MISSISSIPPI POWER & LIGHT COMPANY cc: Mr. N. L. Stampley (w/o)

Mr. R. B. McCchee (w/o)

Mr. T. B. Conner (w/o)

Mr. G. B. Taylor (w/o)

Mr. Richard C. DeYoung, Director Office of Inspection & Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. J. P. O'Reilly, Regional Administrator l Office of Inspection & Enforcement l U. S. Nuclear Regulatory Commission i Region II

! 101 Marietta St. , N.'A. , Suite 3100 Atlanta, Georgia 30303 i

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PLAtiT OPERATIOf!S PANUAL Attachment to A$CM-82/153 g3 .

-- h a s .. : a Volure 08 08-S-03-17 Section 03 Revision 0 Date:

CHEMISTRY PROCEDURE CORE DAMAGE ESTIMATION

~

NON-SAFETY RELATED s

Prepared: ._

Reviewed: /

Technical Review Plt. Quality S6perintendent PSRC:

Approved: .

';hemist ry/Radiat ion Cont rol Supe rintenden t Concurrence:

Assistant P 1,a r t Manager List of Ef fective Pages:

Face 1-7 Atts. 1-V List of TCN's Incorporated:

Revisien TCN No.

l

' DRAFT -

CRAND Gl!LF NUCLEAR STATION % t ' 3 r:r-- a CHEMISTRY PROCEDUPE

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Title:

Core Darage Estination No.: 08-S-03-17 : R evi s ion": O Page: 1 1.0 PURP0FE .

To estinate the pe rcent failed fuel following a major reactor accident from analyt ical result s obtained by samplinF the Containment atmosphere, Suppression Pool and/or Reactor Coolant Sy a ter.

2.0 RESPONSIBILITIES 2.1 The Plant Chemist will provide the Shif t Superintendent or Energency Director with the possible percent fuel damage as soon as the information is available and verified.

2.2 The Chemistry Supervisor will perform the initial calculations and -

provide the attached data calculation sheets and base infornation to the Plant Chemist as soon as available.

2.3 The nost Senior Laboratory Chenist on site will perform the isotopic analyses with another quali fied chemist acting as nonitor.

3.0 PEFERENCES -

3.1 NEDO - 20566 - 7, June 1978, General Elect ric Co. , Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix X, Amendrent No. 7 3.2 Duke Power Company Procedure AP/0/A/A5500, Esticate of Failed Fuel Based on 1-131 Concentration

~

3.3 General Physics Corporation, Training Fession, Pitigating Reactor Core Damage, Grand Gulf '.

3.4 NSAC/14, Novenber, 1980, Workshop on Iodine Releases in Reactor Accidents ,

3.5 1:SMRC Regulatory Guide 1.97-1980, Instrumentation for Lirbt-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident

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• CRAND

• CULF NUCLEAR STATION .m,, - up CHEMISTRY PROCEDURE q s ,

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Title:

Core Damage. Estimation No.: 08-S-03-17 Revision: 0 Page: 2 3.6 General Electric Procedures for the Determination of the Extent of Core

, Damage Under Accident Conditions, C&RE Transmittal RPE 810CL01 4.0 ATTACHMENTS 4.1 Attachment I - Isotopic Inventory Power Correction (Y). For Iodine and I

Xenon

• 4.2 Attachment II - Density Correction Factor, X, For Liquid Sample Temperature Changes 4.3 Attachment III - Data Calculation Sheet

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4.4 Attachment IV - Data Calculation Sheets 4.5 Attachment V - Core Damage Graph 5.0 DEFINITIONS ,

5.1 CGNS Base Information

• 5.1.1 3833Pht 800 Fuel Assenblies 248 Pins / Cell -

3.lE10 FISS10NS /SEC/ WATT 1.19E20 Fj$SIONS4/SEC at 100% Power l Reactor Coolant Volume (Cond 3) = 3E8m1 Suppression Pool Volume (Norm) = 4.5E9 ml ,

Containment NET FREE AIR VOL = 1.4E6 ft3 (3.96 E10 cc)

Drywell NET FREE AIR VOL = 2.7E5 ft3 ( 7.6SE5'cc)

Zironcium 178,400 lba Total -

6.0 PETAILS 6.1 The results determined from this procedure are rough estimates of core damage, and reactions based on these results should be conservative.

l l 6.2 Iodine and )(enon analyses are based upon equilibrium full power isotopic

concentrations. If fuel damage is suspected to have occurred during tires of reduced power or near the tire of significant power change , the l

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• GRAND GULF NUCLEAR STATION "

m a p = e .o CHF.MISTRY PROCEDURE

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b dd-li" 3 JTitle: Core Damage Estimation lNo.: 08-S-03-17 : R ev is ion': 0jPage: 3 f

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core iodine inventory must be compensated accordingly by using Attachment

! I to calculate power correction factor Y.

6.3 Cesium and Krypton, due to their longhalf lives, will be corrected by multiplyirg by the average capacity factor for the previ'ous three years.

6.4 l'easurements of Cs-137 and Kr-85 activitien may not be possible until the ,

i reactor has been shut down for several weeks to allow the decay of the shorter lived isotopes.

6.5 All values given are normalized to volumes of coolant at normal reactor

. coolant system pressure end temperature. To correct data, use At tachment II to determine factor X. -

6.6 The determination of percent failed fuel is biphly dependent on core

[ tenperature reached during the accident condition. Core tenperatures in excess of 1600*F indicate possible cladding damage. Temperatures in excess of 4000*F indicate possibl,e fuel nelt ing.

6.7 If the isotopic analyses show the absence of Ruthenium and Tellurium, then assume that fuel melting has not occurred . However, the presence of these nuclides does not necessarily confirm fuel melting.

6.8 Core damage below 1% is assumed to be a non-accident condition.

6.9 Post accident sampling will be performed in accordance with Cheristry j Sect ion Instruct ion 08-S-04-905, Post Accident sam.pli nr/A na l ys es.

Direction as to parameters mon,itored are included in this instruct ion.

The results obtained from the Containment H2 and the coolant system I-131 analys.es way be used in this procedure to estimate core dapage.

6.10 1(ydrogen Production ,

6.10.1 Containment hydrogen and oxygen concentrations may be used te determine the approximate extent of fuel cladding damage caused by f a metal-water reaction at elevated temperatures.

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CHEMISTRY PROCEDURE i

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Title:

Core Damage Estimation No.: 08-S-03-17 Revision: 0 Page: 4 i l

a. Hydrogen concentrations obtained from plant process instrumentation (J001 A, J002A, J001B, J002B) or the Post Accident Sample Panel Monitor may be used to complete Attachment III.

b. If the hydrogen igniters have been energized, or there is evidence of a hydrogen burn, then the containment oxygen concentration must be obt ained from the Pos t Accident Sample Panel Oxygen Monitor (to be installed) or performed by grab sample analysis and entered in the calculations.

6.11 Iodine 131 6.11.1 Due to analysis of past reactor accidents, all released iodine is assumed to remain in the reactor coolant and/cr the suppression pool; therefore, the total iodine rele as ed is the summation of the two volumes and activities i f the suppression pool was used during the accident. ,

6.11.2 Iodine concentrations day increase by a factor of 2 to 25 times above the eauilibrium level due to spiking rollowing a large power change or shutdown. This peak concentration normally occurs approximately 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> af ter the power chanFe. Do not misinterpret this temporary change for fuel f5ilure if there is no other evidence of fuel damage.

6.11.3 Iodine-131 test results perforned by on-line instrumentation or laboratory analysis will be entered on Attachment IV . Convert the measured activity to a total I-131 concentration released. Use correction factors obtained for X and Y from At tachment s I and II to normalize the data for comparison.

a. The dat,a is then compared to the 100 percent I-131 release concentration:

A = (F/S) (Yield) .

= (1.19E20) (.028)

= 3.3E18 dps/3.7E10 dps/ci

= 8.9E7Ci 1-131 e h"M hain AW $

, GRA.N.h WLF NUCLEAR STATION

' , q r . y, a 4,. CHEMISTRY PROCEDURE

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"'ritle: Core Pamage FitI'ation m No.: 08-S-03-17 lRevi s, ion i 0 Page: 5

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6.12 Cesiun 137 -

6.12.1 Due to analyses of past reactor accidents, all released Cesium is assumed to remain in the reactor coolant and/or the suppression pool; therefore, the total Cesium released is the summation of the.

two volumes and activities if the suppression pool was used during the accident.

6.12.2 Due to the long half life of Cesium and refueling schedule at 1

Grand Gulf, the 100% Cesium activity nust be adjusted based on'a non-equilibrium maximum operational concentration. This adjustment assumes three years operation at 100% power.

6.12.3 Cesium test result s performed by on-line ins t rumentation or laboratory analysis will be entered on 4dtachawnf 1V. If the suppression pool was used as an extension of the reactor system, then test results of both systems should be included. Convert the measured act ivitiy to a total Cesium concentration released. Use correction factor obtained for X from Attachment II to normalize the data for comparison.

a. The data is then compared to the 100 percent Cesium release concentration:

A0 = (F/S) (Yield)

=

(1.19E20) (.062)

= 7.38E18 dps/3.7E!O dps/ci

= 1.99E8,Ci Cs-137 G Eouilibrium Activity @ 3 Years '

A=A0 (1-e-hE)

= (1.99E8 ci)(1-e-( .693/30.24)(3)

= 1.33E7 Ci @ 3 years A

._.3- - - - - - c,.,y * * - - - - - - ~ = ' - - ^ ' * ~ - - - ' ' - '- ' '

GRA

.' 'ND CULF NUCLEAR STATION

• 3 n 7.m - CHEMISTRY PROCEDURE L, ;l.Qlf y .

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Title:

Core Damage. Estimation .

No.: 08-S-03-17 Revision: 0 Page: 6 l

i 6.13 Xenon 133 -

I 6.13.1 Based on reports from TM1, essentially all the Xe-133 released during the accident was found in the containment atmosphere.

6.13.2 Xenon-133 test results may be performed by on-line instrumentation or laboratory analyses entered on Attachment IV*and should then be compared to the 100 percent equilibrium Xenon ralease concentration:

6.67%' Fission Yield at Equlibrium A = (F/S) (Yield) ,

i =

(1.19E20) (.0667)

= 7.94E18 dps/3.7E10 dps/ci

= 2.15E8 Ci Xe-133 @ equlibrium 6.14 Krypton-85 ,

, 6.14.1 Analyses performed by CE and at past reactor accidents show that only part of the Kr is released to the containment and drywell-atmosphere. GE's core damage procedure identifies a partition 3

factor of .155.

6.14.2 Krypton-85 test results performed by on-line ins trumentation or laboratory analysis will be entered on Attachment IV . The data will then be decay and power corrected to the conditions existing at shu td own . .

4

~~

6.14.3 The data is then compared to the 100 percent non-equlibrium Krypton release concentration. .

1.1,5% Fission Yield at Equlibrium

.A0 = (F/S)(Yield)

= (1.19E20)(.0115)/(3.7E10) i = 3.7E7 Ci @ Equlibrium I

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, , ,G, rat?D GULF NUCLEAR STATION ,

-3 ,s . -; CllEMISTRY PROCEDURE t s _- i' ! -

Title:

Core Damage Estimxt ion No.: 08-S-03-17 P ev i s i on': 0 l Ps:ge : 7

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Activity @ 3 Years .

A = A0(1 - e- h E )

~ .693

= (3.7E7)(1-e 10.73 x 3)

= 6.5.E6 Ci Partition Factor to Gas Phase

= (6.5E6 ci)(.155)

= 1.01 E6 Ci Kr-85 6.15 Based on results recorded on Attcheent V., perform an 2 engineering evaluat ion of core status and notify plant management.

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, , GRAND GULF NUCLEAR STATI0ii J '

'., f r." CIIEMISTRY PROCEDURE t!a' e a ' G 1 l 08-S-03-17 Rev. O t l Attachment I Page 1 of 1 l l

ISOTOPIC INVENTORY PORER CORRECTION-(Y) FOR IODIFE AND XENON To correct for core isotopic inventory if fuel damage is suspected, calculate i

the correction factor using the folicwing equation:

Y= 100 Average thermal power over the last 30 days , -

where Y is the correction factor to be used.

Example: The plant has been at 35% full power for the last 30 days when fuel .

darrage is suspected. Therefore:

100

• Y= 35 = 2.86 e

e

• W S

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4 s

[ y , .

=

L} d'sl--W b -

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• GRAND CULF NUCLEAR STATION , s. ' -*c'd CHEMISTRY PROCEDURE t2 a <. a 1/' l l 08-S-03-17 j Rev. 0 l l Attachment Il ! Page 1 of 1 l DENSITY CORRECTION FACTOR, X, FOR LIOUID. SAMPLE TEMPERATURE CHANGES Determine the appropriate Reactor Coolant temperature at the time of sa:,pling.

Norral Peactor Coolant System sample temperature is approximately 90*F. The intersection of both nunbers is the density correction factor, X.

Reactor Coolant Sample Temperature *F REACTOR COOLANT DE NS IT Y @RFECTION -

TEMPERATURE *F FAC TO'R . -X '

, . 100 .998 150 , .985 200 -

.968

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g 250 .947 300 .923

, 350 .895 400 ,

.864 A50 .825 ,

500 .788

'550 .740 560 .729 ,

570 .718 580 .708 i

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l 590 .694 r I

, 600 i .681 i l' I

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GRAND GULF MUCLEAR STATION ;n-s -

* .. CHEMISTRY PROCEDURE h,j y*L. b '

08-S-03-17 Rev. 0 Attachment III, Page 1 of I DATA CALCULATION SHEET Hydrogen -

Monitor Readings: J001A J001B NOTE: Use the nonitor J002A J002B l with the hi hest F H 2 reading. PASM H2 Oxvren - ~~ ~ '

r.: M i' ,

Monitor Reading, pggg og Percent Core Cladding Reacted -

a. Fror F Data

%H 2 M nitor ReadinF) () .67E6) = SCF H2

( ) (1.67E6) = SCF H 2

= SCF H 2

b. From 02 Data NORFAL POST ACCID

(%0 2 M nitor Reading) - (02 M nit r Reading) = Ogdepletion

~

(0 depletion) (2) (1.67E6) = SCF H2 2 _

= SCF H2 (02) .

c. TOTAL VOLUVE of2 H liberated SCF H2 + SCF H2 (0 7 ) = TOTAL SCF H2

+ =TOTALIiCFH 2

= TOTAL SCF H 2

d. TOTAL PASS OF Zirconiun reacted Total SCF H7

=

8.0 SCF H 2

/lbn Zr reacted

=

lbn

e. Percentage of Core Cladding Reacted Ibm Zr Reacted

%= X 100 1.754E5 lbe zr in Core

%= Documed resulfs fl07o on NACb' Y DRAFT

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, . GRAND GULF NUCLEAR STATION , _ s. j - CFEMISTRY PROCEDURE f 8 08-S-03-17 l Rev. O Attachrent IV l Page 1 of 4

~ DATA CALCULATION SHEETS

\

i .

la. Iodine 131 Measured in the Suppression Pool

(_ uci/ml)(4.8E9 ml)(IE-6) = Ci (Sp & Rx) .

[.May be N/A if suppression pool not used]

I

b. Iodine 131 Measured in the Peactor

~

( uci/ml)(3E8 el)(IE-6) = Ci (Rx)

B 1

] [May be N/A if suppression pool ysed)

c. Perfore Decay Calculation to Time of Reactor Shutdown (Ci e time of count)( e At) = Ci released @ To

=

for 100% Pwr

d. Power & Density Correction (Ci 1-131 from c.)(Y)(X) = cor.rected Ci released at T,

( ) (. )( )= +

i

e. Percent of Core Damage -

100 x ( Ci released /8.9 E7 Ci cvailable) =  %

f. Docurent result s on Attachment V by indicating results from step e. above,

, with a range of + 10%.

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Q, RAND Gl'LF NUCLEAR STATION U37l ;) ' ;.jA va]y CHEMISTRY PROCEDURE

. ,+ d" j 08-8-03-17' l Ratv . O  ;

l Attachment IV l Page 2 of 4 l DATA CALCULATION SHEETS

+

2a. Cesium 137 Measured in the Suppression Pool

( uci/ml)(4.8E9 el)(IE-6) = Ci (Sp & Rx)

[May be N/A if suppression pool not used]

b. Cesium 137 Measured in the Peactor d

( uci/mi)(3E8 ml)(IE-6) = Ci (Rx)

[? fay be N/A if suppression pool ysed]

c. Power and Density Correction '

(Ci Cs-137 from a. or b.) (Z)(X) = Ci released at T0

(_ )( )( )=-

=

(Z = Average Capacity Factor f.or Previous 3 Years)

~

d. Percent of Availabe Cesiun 137 Feleased 100 x ( Ci released /l.33 E7 Ci available) =  %

e. Document result s o'n Attachment V by indicating results from step d. above with a range of + 20%.

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GR M'D GULF NUCLEAD. STATION .s n s:urm

-)"f yt q y y CHEMISTRY PROCEDURE 1 08-S-03-17' l Rev. 0 l l Attachment IV l Page 3 of 4 f DATA CALCULATION SHEETS i .

3a. Xenon 133 Measured in the Suppression Pool

(_ uci/cc)(4.73E10cc)(IE-6) = Ci (Drywell.& Cont) .

Assumes approximate equal distribution in drywell and containment

b. Xenon-133 measured in the drywell

~

( uci/cc)(7.65E9 cc) (IE-6) = Ci (Drywell)

=

c. Xenon 133 Measured in the contaigment

] (Ci Cs-137 from a. or b.) (Z)(*X) = Ci released at To 1

( uci/cc) (3.96 E10cc) (IE-6) = Ci (Cont)

NOTE: Use either a., b., or c. or the sum of b. and c. to fit the specific situatien '

d. Perform Decay Evaluation to Time of Reactor Shutdown (Ci @ time of count) e At = Ci released @ To .

e. Power Correction ,

(Ci Xe 133 from d.)(Y) = Corrected. Ci @ To

f. Percent of Availabe Cesium 137 Released 100 x ( Cireleased/2.15E8Available)=  % ,

g. Document results on Attachment V by indiccting results free. step d. above with a range of + 10%.

, . DRAFT -

, qReyD CULF NUCLEAR STATION  :,. 3 . .s ,,,,,, CHEMISTRY PROCEDURE s.4 .: ',. .A ; ' . ] .

1 08-S-03 Rev. 0 l l Attachment IV Page 4 of 4 l DATA CALCULATION SFEETS 4a. Krypton 85 Peasured in the Containment

( uci/cc)(4.73E10cc)(IE-6) = Ci (Drywell & Cont)

B Assures approximate equal distribution in drywell and containment

b. Krypton 85 Measured in the nrywell

( uci/cc)(7.65E9 cc) (IE-6) = Ci (Drywell) a

c. Krypton 85 Measured in the Containnent

'( uci/cc) (3.96 E10cc) (IE-6) = Ci (cont)

NOTE: Use either a., b., or c.' or the sum of b. and c. to fit the specific situation.

d. Power Correction ..

(Ci Kr 85 fron above) (Z) = Corrected Ci @ To ,

(Z = Average Capacity Factor 'for Previous 3 Years)

e. Percent of Availabk Kr 85 Released 100 x ( Ci released /1.01E6 Available =  %

f. Document results on Attachment V by it.dicating results fren step e. above with a range of + 20%.

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CMf 0 GULP MfCl,FAR STATION

• s ...w ,i CHE)'1STRY PROCED!'RE

. .,s s' . . ,v a . . "X , .i I 08-S-03-17 , I Rev. O  ;

l Attachment V. t Page 1 of 1 t CORE DAMAGE GRAPH

% COM DAMAGE F ' ' ' ' ' ' '

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.MP&L QUALITY ASSURANCE REVIEW SHEET GGNS PROJECT MANUALS / DOCUMENTS Part I - To be completed by the responsible Section .

8 O Manual ument Title N]k0 kP L/ k & {hi S E l O Ck et GCr W//)l/ C Review Request Letter No.: 82//3 Dated Review Due Date:

Part II - To be completed by Reviewer (s) l Approvedby:[, p at/ps ate: h -

Manager 5fof' Quality Assura6ce Reviewer's Signature [/l 4- Date:

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No coments

($( ' O Comments ,

Reviewer's Signature Date:

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' O Coments Attach Additional Sheets If Necessary.

Return to G Administration Section  ;

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8 Licensing Section J

DATE j j REV. 8- - L1/26/79' j j INTERNAL PROCEDURES MANUAL

. m, ,, a PAGE 2.8-7

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