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Rev 1 to Procedure NFG-0026, Post-Accident Core Assessment Procedure. W/One Oversize Drawing.Aperture Card Is Available in PDR
	ML18089A334
Person / Time
Site:	Salem
Issue date:	08/19/1983
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18089A332	List: ML18089A331; ML18089A333; ML18089A334; ML18089A335;
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NFG-0026, NFG-26, NUDOCS 8309080096
	Download: ML18089A334 (32)
	v • d • e

{{#Wiki_filter:-*, NUMBER: NFG-00 26 NUCLEAR FUEL GROUP FUEL PERFORMANCE REPORT NFG-0026 Revision 1 Page 1 of 10 TITLE: POST ACCIDENT CORE ASSESSMENT PROCEDURE COPY NO i 0.0 Approval Revision 1 Effective Date Prepared by Date <fo~Szt~J Reviewed by Uc~ Date Reviewed by Approved by 8309080096 830831 PDR ADOCK 05000272 P PDR ,..Cwl£ ) C_ I,-{;C. t. (_,1

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Table of Contents NFG-0026 Revision 1 Page 2 of 10 1

0 Pu rp o s e............................................ 3 2.0 Discussion......................................... 3 3. 0 Prerequisites............ ~......................... 5 4.0 Estimation Methodology...........**......*..*...... 5 5.0 Final Estimation of Core Damage...*.*.**.... ~****** 9

6. O References..*.*.......... Ill **************************

10 Enclosures Half-lives and reference core inventories for isotopes required in the methodology.*.......**.*..*..*..*.****.. Enclosure 1 Relationship between I-131, I-133, Kr-87, Kr-88, Xe-133, Cs-134 and Cs-137 released and the type and extent of core damage in the reference data..*....*..**..** ~.. Enclosure 2 Relationship between Mo-99, Ce-144 St-92, La-140 and Ba-140 released and the type and extent of core damage in the reference data ***..*****.*....*.* Enclosure 3 Ratios of isotopes in the fuel pellet and fuel gap.......**.*..****........*.**.*.*.* Enclosure 4 Relationship between H2 produced and clad reacted................................. Enclosure 5 Radiation Monitor data required for the methodology....**...........***..*..... Enclosure 6

1. 0 Purpose NFG-0026 Revis ion 1 Page 3 of 10 The purpose of this report is to provide an interim methodology to assess post accident core conditions.

The report shall provide guidelines for the estimation of core damage which will consider fission product inv.entory release, hydrogen produced and thermal-hydraulic parameters (local core temperatures and coolant levels). The methodology of the report shall be referenced in the Salem Nuclear Generating Station 11 Emergency Plan. 2.0 Discussion In making a knowledgeable judgement of the extent and type of core damage, the indications from all sources outlined in the report must be evaluated as a whole. The redundancy of indicators provides a mechanism to preclude a judgement based on one source. Each indivi-dual source is realized to have either interpretation or measurement limitations. However, the sources included in the report, when analyzed as a whole, were chosen to ensure the most viable results for the best estimation of core damage. The first source of data to be analyzed will be the fission product inventory released to the coolant. The concentrations of the fission products listed in will be determined by gamma *spectroscopy. The Post Accident Sampling System will be used to obtain samples of the primary coolant and containment air. The containment sump will be sampled at the discharge of the RHR pump as specified in Salem Nuclear Generating Station Procedure PD-3.5.071. The sum of the concentrations from each sampling location will be. used to calculate the total release of each fission product in uCi/g(cc). The concentrations will be adjusted to account for time since reactor shutdown, primary coolant volume changes and differences in pressure-temperature between sample and sample location environments.

2.0 Discussion (Continued) NFG-0026 Revis ion 1 Page 4 of 10 The estimation of core damage can then be calculated by normalizing and comparing the adjusted measured concen-trations to reference data developed from the Salem FSAR and WASH-1400. The reference data supplies fission product inventory information to ascertain extent and type of core damage. If the measured concentrations fall into a range where release from the fuel gap or molten fuel cannot definitely be determined from Enclosure 2 (areas of "Questionable Origin"), contains reference data for comparison which will be present in significant proportions only during a fuel melt. In addition to comparing the fission product concentra-tions to the reference concentrations for determining type of fuel damage, the ratios of certain isotopes are significantly different depending on whether released from the fuel gap or fuel pellet. Thus, another indication of type of fuel failure is available as shown in Enclosure 4. Hydrogen concentrations are also provided with the fission product data. The hydrogen concentrations in the reactor primary coolant, reactor sump, and contain-ment air will be summed to provide the total amount of hydrogen produced. The zirconium-water reaction is assumed to be the dominant hydrogen production mechanism during the first post accident 24 hours. Therefore, all hydrogen produced is assumed to be from the zircaloy cladding. The effects of the recombiner and scavenging by the containment sprays will not be considered for the interim methodology. Plant physical data will also be used to complement the chemistry indicators. Local core temperatures and coolant conditions will be evaluated to determine cladd"ing and fuel pellet temperatures. Furthermore, plant area radiation monitor data will be evaluated for indications of high coolant activity.

3. 0 Prerequisites NFG-0026 Revision 1 Page 5 of 10 3.1 Accident conditions exist and the Salem Nuclear Generating Station Emergency Operating Facility (EOF) has been activated.

3.2 The Fuel Support Team Leader has requested a chemistry sample (executed upon arrival at EOF) consistent with accident case and chemistry procedures for post accident sampling. 4.0 Estimation Methodology 4.1.l Obtain fission product inventory data.from chemistry. Data will be supplied corrected back to the time of reactor shutdown and for any pressure-temperature differences in the sample vial. 4.1.2 Convert each inventory from uci;g or uci/cc to total-curies released. This step is repeated for each fission product isotope reported. Vprimary Total curies = ? curies i l i = sample location: primary, containment, or containment sump = (Vnominal primary Vcontainment = Vnorninal - Vsump containment

4.1.2 (Continued) NFG-0026 Revision 1 Page 6 of 10 Vsump ~ Vactual (Volume of coolant in sump will be reported by plant operations)

for'sample calculations lee. = lg.
- ;6.VRWST and

~VBAST is the volume of coolant added to the RCS from the Reactor Water Storage* Tanks and Boric Acid Storage Tanks, respectively, as reported by operations. 4.1.3 Normalize each fission product inventory to correlate to the reference inventories. Repeat this step for each fission product isotope reported. Normalized total curies = total curies x Icorr* Icorr = 3558 (1 _ e-Ai65o) - e- /iiTj )e - fhTjs] i = isotope of interest Pj = steady state power in period j (MWt) Tj = time of period j (days) Tjs= time from end of period j to shutdown Pj should be chosen such that the variation in steady state power does not exceed ~20%. For short -lived isotopes Icorr should be calculated for approximately 6 half-lives. For longer-lived isotopes Icorr should be calculated back to the beginning of cycle (BOC). This criterion in conjunction with choice of Pj defines j periods for the calculation.

4.1. 4 4.2 NFG-0026 Revision 1 Page 7 of 10 Estimate extent and type of core damage from fission product inventory released. Using, locate the total curies of I-131, I-133, Kr-87, Kr-88, Xe-133, Cs-134 and Cs-137 on their respective graphs and note the type and extent of fuel damage indicated. Use Enclosure 3 to verify indications of questionable origin (fuel melt or gap release) or to further confirm extent of damage. Locate the total number of curies of St-92, Ce-144, Mo-99, Ba-140 and La-140 on their respective graphs and note the fuel damage indicated. Calculate the isotopic ratios of the following: I-133* I-131, Kr-88

Xe-133 and Kr-87*

Xe-133 Compare "the ratios to the data in Enclosure 4, and. note the type of failure indicated.

All ratios are based on total normalized curies.

4.3 Estimate clad damage from hydrogen production 4.3.l Obtain hydrogen concentrations from chemistry for the primary coolant, containment air and reactor sump. Concentrations will be supplied cor-rected for any pressure-temperature differences in sample vial.

4. 3. 2 NFG-0 0 26 Revision 1 Page 8 of 10 Calculate total volume of hydrogen produced NOTE:

Volumes are calculated as desGribed in Section 4.1.2. Hpt = Hp x Vprimary x P2T1* P1T2 Hct = %H x Vcontainment x P3T1* P1T3 Hst = Hs x Vsump x P4T1* P1T4 Ht = Hpt + Hct + Hst (T1,P1) *-standard temperature and pressure* ( STP) (T2,P2) - temperature and pressure of primary system (T3,P3) - temperature and pressure of containment (T4,P4) - temperature and pressure of sump

hydrogen concentrations are corrected to STP to correlate with reference data.

Hp,Hs - concentration of hydrogen in the primary and sump respectively (CC/g) %H - percent of containment volume that is hydrogen Hpt1 Hct1 Hst1 Ht - hydrogen in primary, containment, sump, and total, respectively.

4.3.3 NFG-0026 Revision 1 Page 9 of 10 Using Enclosure 5, estimate the amount of clad damage by locating total grams of hydrogen and noting the percent of fuel failure (clad reacted). 4.4 Estimate Coolant Conditions in Core 4.4.1

4. 4. 2 Obtain thermal-hydaulic parameters from plant operations.

Evaluate local temperatures (thermocouple data), core pressure, reactor coolant level and core flow to ascertain if the heat transfer characteristics indicated confirm the type and extent of damage noted in 4.1 through 4.3 4.5 Estimate core condition - from area radiation monitors 4.5.1 4.5.2 Obtain and record area radiation monitor data from plant operations as indicated on Enclosure 6. Evaluate departures from nominal (or lack of) to confirm other indications of core conditions. 5.0 Final Estimation of Core Damage Collectively evaluate all indications* of the type and extent of fuel damage and make a knowledgeable assessmen~ of the condition of the core.

6.0 References 6.1 Salem Final Safety Analysis Report 6.2 WASH-1400, USNRC, App.7-10 NFG-0026 Revision 1 Page 10 of 10 6.3 Salem Chemistry Procedures: 3.1.001, 3.1.002, 3.1.003, 3.1.005, 3.1.006, 3.1.008, 3.1.009 6.4 Salem Nuclear Generating Station Procedure: PD-3.5.071 6.5 Salem Emergency Plan Procedure EP V-6 \\ 1

NFG-0026 Revision 1 Page 1 of 1 HALF LIVES AND REFERENCE CORE INVENTORIES ISOTOPE HALF LIFE REFERENCE INVENTORY ( lo7ci) I-131 8.04d 8.8 I-133 20.8h 19.7 Kr-87 76.m 7.59 Kr-88 2.84h 10.8 Xe-133 5.25d 20.3 Cs-134 2.06y 20.9 c*s-137 30.17y 19.5 Mo-99 66.02h . 19. 2 Ce-144 284.4d 19.0 Ba-140 12.8d 20.1 La-140 40.2h 20.1 St-92 58.6d 16.8

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ISOTOPE Kr-87 Kr-88 Xe-133 I-133 I-131 Ratio = NFG-0026 Revision 1 Page 1 of 1 RATIOS OF ISOTOPES IN THE FUEL PELLET AND FUEL GAP RATIO IN FUEL PELLET* (INDICATES FUEL MELT) 0.233 0.33 1.0 2.09

1. 0 isotope concentration Xe-133 concentration isotope concentration RATIO IN FUEL GAP*

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NFG-0026 Revision 1 Page 1 of 1 RADIATION MONITOR DATA MONITOR. Rl2A Containment Noble Rl2B Containment Iodine R31B Letdown Iodine R41B Plant Vent Iodine R41C Plant Vent Noble R45A Plant Vent Noble (Background) R45B Plant Vent Noble (medium) R45C Plant Vent Noble (high) NOMINAL READING* ACTUAL READING* Supplied from plant operations during post accident recovery.

APPENDIX B Pass Procedures Listing Procedure No. Title

1.

CH-3.1.001 - PASS Diluted Liquid Sampling During Accident Conditions.

2.

CH-3.1.002 - PASS Undiluted Liquid Sampling During Accident Conditions.

3.

CH-3.1.003 - PASS Containment Air Sampling Procedure

4.

CH-3.1.004 - PASS pH/Conductivity/YSI Dissolved Oxygen Analysis During Accident Conditions.

5.

CH-3.1.005 - PASS Reactor Coolant Stripped Gas Sampling During Accident Conditions.

6.

CH-3.1.006 - PAss* Gas Chromatographic Hydrogen Analysis During Accident Conditions.

7.

CH-3.1.007 - PASS Ion Chromatographic Chloride Analysis During Accident Conditions.

8.

CH-3.1.008 - PASS Interim Transfer and Dilution of Contain-ment Air Samples.

9.

CH-3.1.009 - Transfer and Dilution of Diluted Reactor Coolant Samples for Boron and Isotopic Activity Analyses

10. CH-3.1.010 -

PASS Reactor Coolant Grab Sampling During Normal Conditions.

11. CH-3.1.011 -

PASS, pH, Conductivity/Rexnord Dissolved Oxygen Analysis During Normal Conditions.

12. CH-3.1.012 -

Assembly and Calibration of the Sentry Shielded Syringe.

13. CH-3.1.013 -

Operation of the RCT Gas Partitioner for Obtain-ing Containment Air Samples.

14. CH-3.1.015 -

PASS Disposal of Containment Air Samples B-1

Pass Procedure (continued) Procedure No. Title

15. CH-3.1.016 -

PASS Disposal of Reactor Coolant Stripped Gas Samples.

16. CH-3.1.017 -

Calibration of the Baseline Gas Chromatograph

17. CH-3.1.018 -

Calibration of the PASS Dionex Ion Chroma_to-gr:aph.

18. CH-3.1.019 -

PASS Equipment Storage.

19. CH-3.1.020 -

System Checkout of the RCT Containment Air samples.

20. CH-3.1.021 -

Handling and Preparation of RCT Gas Samples.

21. CH-3.1.071 -

Interim Post Accident Sampling

22. PD-3.2.080 -

Boron Analysis Fluoroborate Selective Ion Electrode. (Copies of the procedures listed available upon request from PSE&G. ) B-2

APPENDIX C Drawings

1.

Flow Diagrams:

2.

A. Liquid Sampling System B. Containment Air Sampling System

c.

Chemical Analysis System Arrangement Drawings: A. PASS Facilities - El 100'-0", Partial Plan C-1

[ REFERENCE [)RAWINGS :, 1l REACTOR COOLANi PIPING DIAGRAMS 205301-A-8762 & 205201-A-8760.* 2l COMPONENT~ COC~J~G _ PIPING D!AvAA~M -_ 20533!-A-876"

3) PASS INSTRUMENT

SCriEMATICS 227510-8-9491 !.

6'10081-8-94 78 4l PSB?. 172063 - LIDUID 5AMP~!NG PANEL813

PSBP l7206A0NALVS!S CHEMICAL PANEL 815 61 PSBP 172047 COOLER RACK PANEL 811

(-*- L._, NOTE 3

-,-~*

v1;1 -~ _l'JTE.~ Vl.2 ATMOSPHERIC VENT NOTE 3 '/7 ~ ~. I _, I u

.i QT C j---PANEL 811

--*~---...J SPfJ,RE ~ V!.5

_ VB.2 CP-L-lS

) V4 Fs-L-3 CYl.2 6 ' R~ACTOR COOLA~..

TO-CHEt.i. AN~!.. YS!:i PANEL 615 643!K-A 2CC375 2CC372 CP-L-21 CP-L-25 PRv-3 CP-L-24

~ LO\\o/, ~ ES;;J 'GSl,4 G5!.3 t FJ:.2 J ~ -n LO EOE 1155908 DELAY COIL "TC LO TC 2CC373 CP-L-13 ~1 ~, ~I ~ 3QT1LE uf ~ llRC6 !3RC6 G-L-1 cv-s UNIT N0. l SUMP 11 HOT LEG 21!15201-A-8760 UNIT N0. 1 SUMP 13 HOT LEG 2'!15201-A-8761l UNIT N0, 2 SUMP 21 HOT LEG 205301-A-8762 UNIT Nil. 2 SUMP COMPONENT COOU~G \\/PTER OUT ~1!5331-A-6763 TO O!LUTEO _GAS B~TTLE Sw.PL~ N0.2 ,,,_.I.I~ FLVS~ FOG~ r;"; c~ni--o..,Cti>, FAKE:.l. SIS Hi GAS CtftOMAT, ICAPi PANEL 615 \\ ~* ' \\ ~.. *, ~- ,~; *" ".i A

OFF-GAS SAMPLE ARGON <CHROMO. GRAOEl FROM LS~.::: G-L-lfil IX) I -' I c.:> Vl F4 GAS CHROMATOGRAPH AE-1 I AT-1 REACTOR COOLANT FROM uaum SAMPLE PANEL 813 OG-1 I ~-I-----T-------__, ---J I I EXHAUST TO 1 P1E ---$ I I PLENUM ~----© REl=ERENCE DRAWllJGS: l)PS~P \\7'ZO<od C:HEMIC:AL ANAi.. VS iS PA.i.JEL Bio.

2) PSE>P 17'2063 L.IQUIO S.O.MPLll-lC. PANE:I. 61&.

Ki=Y: AIR LINE: ,V 1S' #,f' AP-L-17 WASTE )> I CL <I: CV15 N ~ I CL <I CV7 co I -' I Cf) u DRAIN INSTRUMENT AIR V10 AP-L-6 CS-L-21 -<t I -'

I Ill LL N

I cva _J N 69 I (/) I LL. I (!) I I AP-L-17 CHEMICAL ANALYSIS SYSTEM (PANEL 815) l--+'-"'-'-1-- - -© AP-L-13 CVl CV2 CV11 CV12 AP-L-l4 --CV $ I AP-L-B A 2 .1 VB V7 CS-L-4 AP-L-11 V2 I -' I CL <t !*I .\\,.

~J;:,.

[ NOTES:~ I> AIR OPERATED SAMPLE LINE VALVES ARE CONTROLLED FAOM PANEL 81-4, 2l AIR CFERATOO VALVES IN PANEL 812 ICASP> ARE CONTROLLED FROM PANEL 616 IAS ltl:>ICATEDI, 31 ELECTRIC HEAT TRACE IS ENERGIZED FROM P~LS '11"4 A-1121 l %4 S-1121 ANO CONTROi.LEO/MONITORED FROM '113 A-l!Z> l '113 B-llZ> REF. DWG:

I. CONTROL AIR PJPIND DIAGRAM 2£)53'43-A-8763

2. WASTE OISPOSAL. GAS PIPING DIAGRAM 2063'46-A-6763 3, PASS IHSTRUMENf S(Ha.1ATIC 207!510*B-'1"4'll !UNIT ll

-4. PASS INSTRUMENT SC+IEW.TIC SOOl!JSl-B-'i-478 !UNIT 2> ~. P S B P 1720!54 CONTAINMENT AIR SAMPLING PANEL, 8l21CASPl. 6, P S B P 1720"4'!, CASP CONTROL PANEL BIS 7, P S B P 172050, CASP GRAPHIC V4 .KEY:. HEAT TRACE C\\J\\f'(;;, -. l::LECTRlCAL - ~. -~.. AIR f' f' ~',; I' TC=TEST. CDN~c~nc~ // '

1.

UNIT l UNIT 2 I-UNIT 2 UNIT 1 ~-----'"'-----~ A a'i ~ :E z TC ~ >-:z TC TC TC TC 0 u ~ ~ u; z I- ¢ I-TC 0 z 0 u .-----.i1<----,F--J 1,.__,f'---!><ll---rl--< CONTROL AIR 205343-A-0763 I VENT : CASP CONTROL PANEL 8J6 I ~ VENT ~*-*-*-*-*-*---1-1-1.. - 1*-1-*1 -*--*-*-*-*-** ~---*-' r--+--.--::12t¥J0i3t---#---.I'----#"----#'-. _. --,,I'-- :-#-:.ff<+-n<--l-n<-l-n<-l---------,----#--JCJ)---,f'---#---lif---+--,~lol:¥:~-#---I I I I I L ________ _ -------.-----:-- sv-1~ I 1, I I I l I L--.,.---- 1 I I

L-IB L-10

r--1 L-11 I

I

--....,.---.-1 _I*

L-16 I FM-I I SV5 L-40 L-32 I V2 L-12 L-42 L-13 I I L-41 SV-10 L __ _ EV-1. L-38 01.8 2CA845 L-115 I V3 I ~--_,.PANEL 816 I L-14 .---+--::+:-+--r:-.3;-. - * - * -.* - * - * - * -: * - * - * - * -

L ;:::!:-a~_,

V7 SF! SF2 Vl3 SF4 Vl5 PI U!llod wlth PE-I - PE-4 CONTAINMENT AIR SAMPLING SYSTEM

' ~.

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ML18089A334

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