ML17301A141

Rev 0 to Emergency Plan Implementing Procedure 1302, Offsite Emergency Organization Psl Core Damage Assessment. W/841206 Ltr

ML17301A141
Person / Time
Site:	Saint Lucie
Issue date:	12/06/1984
From:	Williams J FLORIDA POWER & LIGHT CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
1302-01, 1302-1, L-84-372, NUDOCS 8412110386
Download: ML17301A141 (141)

Text

REGUlATORY FORMATION DISTRIBUTION S EM (RIDS)

ACCESSION NBR; 8412110386 DOC ~ DATE t 8$ /12/06 NOTARIZED: NO DOCKET FACIL't50-335 'St, Lucie Plant~ Un) t 1, Fl or ida Power & L)ght Co. 05000335 50>>389 'St. Lucie Planti Uni,t 2> Florida Power & Light Co. 05000389 AUTH',NAME."- AUTHOR,. AFFILIATION WILLIAMS' ~ W ~ Flor ida Power & light Co,

'RECIP ~ NAME'ECIPIKNT. AFFILIATION O'REILLY<J P, Region 2'i Of f i ce of Dir ector Assessments tr'a.-i Plan Implementing Procedure SUBJECT)= Rev O~to Emer gency 130?i "Offsite Emergency Organization'SL Core Damage "

TITLK".-": OR W/841206 l DI'STRIBUTION CODE: A045D:COPIES RECEIVED!LTR Submittal Emergency Prep Correspondence (ENCL ~ SIZE; R~

05000335 NOTES:'L:02/01/76 05000389 OL;04/06/83 RECIPIENT'D

~

COPIES RECIPIENT COPIES CODE/NAME> L'TTR ENCL< ID CODE/NAME LTTR ENCL NRR ORB3 BC 1 1 NRR ORB3 BC 1 1 KREUTZERgP 1 0 SELLSpD 01 1 INTERNALS IEi DEPER/EPB 06" 2  ? IE/DEPER/IRB 12 1 1-1 NRR/DSI/RAB 1 RE G. 04 1 1 RGN2 1 1 R 2/DRSS/EPRPB, 1 1 EXTERNAL: FEMA TECH HAZ, 1 1- LPDR 03 NRC PDR 02! 1 1 NSIC NT'I S 05. 1 1 TOTAL NUMBER 'F COP IES REQUIRED ~ 'TTR 1 7 '"

ENCL 1 6

Mh P. IOX l4000, JUNO BEACH, FL 33408 FLORIOA POWER 8I LIGHT COMPANY December 6, 1984 L-84-372 Nr. James P. O'Reilly Regional Administrator, Region,II U. S. Nuclear Regulatory Commission Suite 2900, 101 Harietta Street NW Atlanta, Georgia 30323

Dear Nr. O'Reil.ly:

Re: St. Luci e Units 1 and 2 Docket Nos. 50-335, 50-389 Emergency Plan Implementin Procedures In accordance with 10 CFR 50, Appendix E, enclosed is one copy of Emergency Pl an Impl ementing Procedure:

Number 'itle 1302 PSL Core Damage Assessment - Of f-Si te Emergency Organi zati on Two copies of the enclosed proced'ure have been forwarded to the Document Control Center by copy of this correspondence.

Very 'truly your s, J. W. IIIlliams, Jr.

Group Vice President Nuclear Energy JWM/RDH/js Attachment cc: > Document Control Desk (2 copies of attachment)

Harold F. Rei s, Esquire P NS-L I-84-431-1 PEOPLE... SERVING PEOPLE

t y I

To: Distribution

Subject:

L 84 372 December 6, 1984 From: J.W. Williams, Jr.

NRC CORRESPONDENCE ST. LUCIE UNITS I &2 GO: J. W. Dickey PSL I: R.R. Jennings J.H. Francis Vault Custodian J.E. Moaba C. Fierabend (NRC)

D.C. Po tera lski JB: S.G. Brain PSL 2: C.T. Hamilton R.F. Englmeier E. Preast F.G. Flugger N.T. Weems F.P. Green G.E. Crowell (ISEG)

C.S. Kent W.B. Lee PTP 3/4: Document Control (K. York)

H.D. Mantz F.A. Panzani J.L. Parker J.E. Vessely OTHER: T.P. Gates (Combustion) SL I W.J. Harris (Combustion) SL 2 Harold Reis, Esq. (Newman)

M. Horrel I (Ebasco)

,nC i

e EMERGENCY PLAN IMPLEMENTING PROCEDURES PROCEDURE 1302 R ev.

OFF-SITE EMERGENCY ORGANIZATION PSL CORE DAMAGE ASSESSMENT Oa<e 10/24/84 FLORIDA POWER & LIGHT COMPANY Page of 1 80

1.0 Tit1e

PSL CORE DAMAGE, ASSESSMENT - OFF-SITE EMERGENCY ORGANIZATION 2.0 A royal and List of Effective Pa es:

2.1 ~Arovai s Revi wed by Emergency Planning Supervisor o 1984.

Approved by: Chief Engineer Power Plants s 1984+

2.2 List of Effective Pa es

~Pa e Date i

1 through 80 inclusive 10/24/84

3.0 Scope

3.1 ~por ose:

This procedure identifies the responsibility and methodology to perform core damage assessment for St. Lucie Plant (both Units 1 and 2).

Methods for estimating core damage assessment are based upon post accident raiionuclide concentrations within the reactor coolant system and containment, and auxiliary indicators, 'ncluding core exit thermocouple, hydrogen, and containment high range radiation monitors.

An estimate of core damage can then be used to assist in evaluating protective action recommendations, severity of plant conditions, and/or plan operations. 'ecovery 3.2 Discussion The Of f-Si te Emergency Organi zati on provides an .

expanded emergency response'apabi1 i ty to assi st the plant in admini stration, cenmuni cati ons, engineering, techni cal support, securi ty, and publ i c.

relations. This organization, which consists of the Emergency Technical Manager 'nd hi s staff provides engineering and technical support at the r equest of the Emergency Control Officer (ECO) and/or Recovery Manager (RM). Specifically, this support includes estimating core damage, using the methodology provided in the appendix, to, differentiate among four major fuel conditions. These are:

No Damage Cladding Failures 8412110386 841206 Fuel Overheating PDR ADOCK 05000335 F PDR Core Melt

EMERGENCY PLAN IMPLEMENTING PROCEDURES PROCEDURE 1302 Rev.

OFF-SITE EMERGENCY ORGANIZATION PSL CORE DAMAGE ASSESSMENT Date 10/24/84 FLORIDA POWER a LIGHT COMPANY Page The methodology attached is site specific and is based upon Combustion Engineering Owners Group (generic) core damage assessment guidelines.

3.3 Authority At the request of the ECO or RM, the Emergency Technical Manager will direct his staff to perform core damage assessment using the applicable guidelines in the attached appendix. The ECO or RM will request that appropriate input parameters be provided by the plant in order for the ETM's staff to perform the assessment.

\

4.0 Pr ecauti ons 4.1 The assessment of core damage obtained by using the attached methodology is only an estimate. The techniques employed are only-accurate to locate the core condition wi thin one or more of the 10 categories of core damage described in the methodology.

4.2 Core damage assessment using indicators that are readily available (e.g., containment high range radiation monitor) represent, only preliminary estimates. Other plant indicators (e.g., radionuclide concentrations) should be obtained to improve upon estimation of core damage.

4 ' Measurements obtained duririg rapidly changing plant conditions should not be weighted heavily into the. assessment of core damage. If deemed necessary, these pertinent indicators should be measured within a minimum time period particularly during rapidly changing conditions.

It is recommended that measurements be made, if possible, when plant conditions stabilize.

5.0 Responsibil i ties 5.1 The Emergency Control Officer or Recovery Manager will request the Emergency Technical Manager to perform core damage assessment using the methodology attached.

5.2 The Emergency Control 'f ficer or Recovery Manager wi l l request the plant to provide appropriate data in order to perform the assessment.

5.3 The Emergency Technical Manager will direct his staff to perform core I damage assessment (when staffed accordingly) using the attached I methodology.

- I I

will f

5.4 The RM use this information as deemed appropriate in evaluating severity of plant conditions, protective action recommendations, and/or recovery operations.

EMERGENCY PLAN IMPLEMENTING PROCEDURES PROCEDURE 1302 Rev.

OFF-SITE EMERGENCY ORGANIZATION

-PSL CORE DAMAGE ASSESSMENT Date 10/24/84 FLORIDA POWER a LIGHT COMPANY Page 3 80 6.0 References 6.1 Turkey Point Plant Radiological Emergency Plan 6.2 St. Luci e Plant Radiological Emergency Plan 6.3 Procedure 1101, Duties of the Emergency 'ontrol Officer, Off-Site Emergency Organization.

6.4 Procedure 1102, Duties of the Recovery Manager, Off-Site Emergency Organization.

6.5 Procedure 1105, Outi es of the Emergency Technical Manager, Off-Si te .

Emergency Organization.

6.6 Appendix I, St. Lucie Units 1 and 2 Core Damage Assessment Guidelines, Rev. 0, October 1984.

7.0 Recor ds Al 1 informati on used to estimate core damage including appropri ate worksheets wi 1 1 be documented by the Emergency Techni cal Manager or hi s staff designee.

8.0 Instructions 8.1 The ECO or RM can request that an estimate of core damage be performed by the Emergency Technical Manager's staff, when deemed appropriate.

The ECO or RM will request pertinent data frott the plant to perform the assessment.

8.2 The Emergency Techni cal Manager wi l 1 di rect hi s sta ff desi gnee to perform the estimate using the methodology provided in the appendix.

8~3 The staff designee will perform the estimate using this methodology.

Available pertinent plant data needed to perform the assessment will be provided to the staff designee through the ECO or RM.

8.4 All pertinent data available should be used in estimating 'core damage. This includes radionuclide data and auxiliary indicators including core exit thermocouple, hydrogen, and containment high range radiation monitor.

8.5 Results in terms of fuel condition should be provided to the Emergency Technical Manager and Recovery Manager (and/or Emergency Control Officer) as timely as possible.

8.6 Updated estimates to,core damage may be requested periodically by the ECO or RM as plant conditions change and/or stabilize. These updates should be performed using the most recent available plant data and the methodology. Results should continue to be reported to the ETM and RM (or ECO).

I

'l 4

'I

(4 at: so),

APPENDIX CORE DAMAGE ASSESSMENT METHODOLOGY - ST. LUCIE PLANT REVISION 0 OCTOBER 1984

(5 of 80)

The core damage assessment methodology is divided into four main sections. Each section contains its own table of contents and list of enclosures, A fifth section, summary of results, has" been added as a guideline for a comprehensive evaluation of results.

INOEX SECTION PAGE Radiological Analysis of Samples Core Damage Assessment using Hydrogen 28 C. Core Damage Assessment, using Core Exit 55 Thermocouple Temperatures D. Core Damage Assessment using Radiation ~o3 Dose Rates E. Summary of Results

(6 of 80)

SEC'PION, h RADIOLOGICAL ANALYSIS OF SAMPLES

~

~ (

I

(7 of 80)

TASLE OF CONTENTS PAGE 1 ~ 0 PURPOSE 4 2.'3 REFERENCES

3. 0 DEFINITION 4 4.0 PRECAUTIONS AND LIMITATIONS 5 5.0 INSTRUCTIONS F 1 Record of Plant Condition 5 ' Selection of Sample Location 5.3 Sample Analysis 6 5 ' Decay Correction 7 5.5 Identification of the Fission Product 7 Release Source 5 ' Quantitative Release Assessment 7 5 ' Plant, Power Correction 13 5 AS Assessment of Core Damage l.

2'IST OF ENCLOSURES Enclosure Al Radiological Characteristics of HRC Categories of Fuel Damage Enclosure A2 - Input Parameters Enclosure A3 Sample Locations Recommended for 16 Core Damage Assessment Fnclosure A4, Record of Measured'Specific Acta.vasty 17 (Decay. Corrected)

Enclosure A5 Record of Fission Product Release Source Identification Enclosure A6 Quantitative Release Assessment Norksheet Enclosure g7g ord R ecor of Core Release Inventory 21 Enclosure AB Density Correction Factor for Reactor 22 Coolant, Temperature Enclosure A9-PSL1 Record of Steady State Power Correction for St. Lucie Unit 1 Enclosure A9-PSL2 Record of Steady State Power Correction for St. Lucie Unit, 2

(8 of 80)

TABLE OF CONTEHTS {CONTINUED)

PAGE Enclosure AlQ-PSLl Record of Transient Power 25 Correciton for St. Lucie Unit L Enclosure ALQ-PSL2 Record of Transient Paver 26 Correction for St. Lucie Unit 2 Enclosure All Record of Percent Release 27

I l~

I

(>0 of 80) 3 ' Source Inventory: The source inventory is the total quantity of fission products expressed in curies of each isotope present in either source; the fuel pellets or the fuel rod gas gap.

4.0 PRE AUTIOHS AND LIMITATIONS 4~L The assessment of core damage obtained by using the methodology in this section is only an estimate. The techniques employed in this section are only accurate to locate the core condition within one or more of the 10 categories of core damage described in Enclosure AL. The methodology is based on radiological. data.

Other plant indications may be available which, can, improve upon estimation of core damage. These include incore temperature indicators, the total quantity of hydrogen released from zirconium degradation and containment radiation monitors. Whenever possible these additional indicators should be factored into the assessment.

4~2 The methodology in this. section relies upon samples taken from multiple locations inside the containment building to determine the total quantity of fission product's available for release to the environment.

The amount of fission products present at each sample

.location may be changing rapidly due to transient plant conditions. Therefore, it is recommended that the samples should be obtained within a minimum time period and if possible under stabilized plant conditions. Samples obtained during rapidly changing plant conditions should not be weighed heavily into the assessment of core damage.

4' A number of factors influence the reliability of the chemistry samples upon which this section is based.

Reliability is influenced by the"ability to obtain representative samples due to incomplete mixing of the fluids and equipment limitations. The accuracy achieved in the radiological analyses are also influenced by a number of factors. The equipment employed in the analysis may be subjected to high levels of radiation exposure over extended periods of time. Chemists are recommended to exercise considerable caution to minimize the spread of radioactive materials. 'Samples have the potential of being contaminated by numerous sources and they may not result from a uniform distribution of the sample fluid. Cooling or reactions may take place in the long sample Lines. Therefore, the results obtained may not'e representative of plant conditions. To minimize these effects multiple samples should be obtained over an extended time period from each location, conditions permitting.

(9 of 80)

I p.g PURPOSE This section provides a method under post-accident plant conditions to determine the type and degree of reactor core damage which may have occurred by using fission product isotopes measured in samples obtained from the Post Accident Sampling System {PASS). There are three factors considered in this section which are related to the specific activity of the samples. These are the identity of those isotopes which are released from'the core, the respective ratios of the specific activity of those isotopes, and the percent of the source inventory at the time of the accident which is observed to be present in the samples. The resulting observation of core damage is described by one or more of the ten categories of core.

damage in Enclosure Al.

2 ' REFERENCES 2.1 Development of the comprehensive procedure guideline for core damage assessment, CE Owners Group Task 467, July 1983.

2 ' Post Accident Sampling System Operating PSL-1 and 2-C-113, PSL-2 Procedures'-C-112,

~

3.0 DEFINITIONS

3. 1 Fuel Damage:. For the purpose of this methodology, fuel damage is defined as a progressive failure of th material boundary to prevent the release of radioactive fission products into the reactor coolant starting with a penetration in the zircaloy cladding.

The type of fuel damage as determined by this methodology is reported in terms of four (4) major categories which ares. no damage; cladding failure; fuel overheat; and fuel melt. Each of these categories is characterized by the identity of the fission products released, the mechanism by which they are released, and the source inventory within the fuel-rod from which they are released. The degree of fuel damage is .measured by the percent of the fission product source inventory which has been released into fluid media and therefore available for immediate release to the environment. The degree of fuel damag as determined by this methodology is reported in terms of three levels which ares. initial; intermediate; and major. This results in a total of. ten possible categories as characterized in Enclosure A1.

(ll of 80) 5, g INSTRUCTIONS 5 ' Obtain the following plant indications and source of indication. Record on Enclosure A2. Because of transient conditions the values should be recorded as close as possible to the time at which the radiological samples are obtained.

SOURCE:

Pressure PSIG Temperature (Tavg) oF Reactor Vessel Level Shows:. (Full, Void or Below Recorder)

Pressurizer Level

5. l.'2 Containment Bnildincn:

Atmosphere Pressure PSXG Atmosphere Temperature OF 5.l.3 Prior 3'3~da s Power Resistor POWER, PERCENT DURATIONs DAYS Estimated average power level during last 30 days:.

Estimated average power level during last 4 days:

S.l.4 Time of Reactor Trio Date: Times.

5.1.5 Change in Refueling Water Tank (RWT) Volume gal. Time:

5-1.6 Change in Boric Acid Makeup Tank (BANT) volume gal. Time:

5.1 7 Safety Injection Tanks injected (yes/no):

~

5.2 Select the most appropriate sample locations required for core damage assessment using the guidelines provided in Enclosure A3.

5 ' Obtain and analyze the selected samples for fission product specific activity using the procedures for Post Accident Sample System operati.on described in Reference 2.2. Record the required sample data,

I (12 of 80) corrected to Standard Temperature and Pressure (ST?,),

and time of sample collection on Enclosure A'2. All of the isotopes listed in the enclosure may not be observed in the sample.

5.4 Correct the sample specific activity at STP for decay back to the time of reactor trip which is recorded in step 5.1.4 using the following equation., Enclosure A4 is provided as a worksheet.

A A0 e

Where: Ao ~ the specific activity of the sample corrected back to the time of reactor trip, uci/cc.

the measured specific activity, uci/cc the radioactive decay constant, 1/sec.

the time period from reactor trip to sample analysis, sec.

5.5 Identification of the Fission Product Release Source.

S.5.1 Calculate the following ratios for each noble gas and iodine isotope usi;ng the specific activities obtained in step 5.4. Enclosure A5 is provided as a worksheet.

Noble Gas Ratio ~ Noble Gas Isotone Specific Actxvit Xe 1.33 Specific Activity I-131. Specific Actxvity 5.5.2 Determine the source of release (gas gapor fuel pellet) by comparing the results obtained in step 5.5.1 to the predicted ratios provided in Enclosure AS..An accurate comparison is not anticipated. Within the accuracy o f this methodology it is appropriate to select as the source of release that ratio which is closest to the value obtained in step 5.5.1.

5.6 Quantitative 'Release Assessment 5.6.1 Calculate the total quantity of fission products available for release to the environment. Enclosure A6 is provided as a worksheet.

r (13 of 80) 5~6 ~ L.L lf the vessel water level in the reactor recorded in step 5.1.1 indicates that the vessel is fulL, the quantity of fission products found in the reactor coolant is calculated by the following equation:

Total Activity. AT RCS ( i) 0 (uci/cc) X RCS volume X 1.0 (-6)

Where: ~ the specific activity of the reactor coolant sample corrected to time of Reactor trip obtained in step 5.4, uci/cc.

RCS Volume ~ in units of cc, the ful1 reactor coolant system water volume corrected to standard temperature and pressure using'Enclosur A9.

RCS Volume ~ Water Volume X density ratio (Enclosure AS).

is 2.945

'LL Water Volume (8) cc SL2 Water Volume is 2.993 (8) c 1 ' (-6) ~ i/uci 5.6.L.2 Zf the water levels in the reactor vessel and pressurizer recorded in step 5.L.L indicates that a steam void is present in the reactor vessel, then the quantity of fission products found in the reactor coolant is also calculated by step 5.6.1.1.

However, it must be recognized that the value obtained will overestimate the actual quantity released.

Therefore, this sample should be repeated at such time when the plant operators have removed the void from the reactor vessel.

5.6.1.3 Xf the water level i.n the reactor vessel recorded in st p 5.1.1 is below the low end capability of the indi-cator. i.t is not possible to determine the quantity of fission products 'from thi.s sample because the volume of

~ 4o'ha rassWrave sn4o

(14 of 80) unknown. Under this condition, assess-ment of core damage is obtained using the containment sump sample.

4 5.6.2 The quantity of fission products found in the containment building sump is determined as follows:

5 6 2 1

~ ~ ~ The water volume in the containment building sump is determined from the sum of the following sources as applicable:

(gal) RCS Volume (gal) the injected S.I.T.

tanks volume (step 5. 1. 7)

(gal) the Delta volume change in B.A.M.T. (step 5.1.6)

(gal) the Delta volume change of the RWT (step 5.1.5)

Sump Volume ~ (gal) Total gal X 3785 cc/gal ~ cc Maximum Values in gallons for each unit from ap'plicable FSAR Ch. 6.

SL1 SL2 RCS Volume (cold)(gal) 58,300 .57, 400 Safety Injection Tanks 34, 049 46,564 (SIT) volume (gal)

Caution: Values reported indicate, maximum volumes in applicable FSAR.

Water volume in containment building sump is only applicable in recirculation mode.

5~6~2~2 The quantity of fission products in the sump is calculated by t'e following equation:,

Total Activity, A> sum (Ci) ~ Ao (uci/cc) X Sump Volume 7cc) X 1.0 (-6)

Where:. Ao ~ the specific activity of the containment sump sample corrected time of reactor trip obtained in to'he step 5.4, uci/cc.

0 (15 of 80) 5~6~3 The quantity of fission products found in the containment building atmosphere is determined as follows:

A 5~6~3 ~ 1 The volume of, gas in the containment at the time of the accident 'uilding, is corrected to standard temperature and pressure using the foLlowing equation:

Gas Volume (STP) ~ Gas Volume X {F

• P2 (TL + 460)

Where' Gas Volume ~ 7.096 X LO cc TL, Pl ~ Containment Atmosphere temperature and pressure

'ecorded in step 5.1.2 T2 P2 ~ Standard temperture, 32 P, and Standard pressur 14.7 psia.

5.6.3.2 The quantity of fission products {Ci) in the containment atmosphere by: is'alculated Total Activity AT co (Ci) Ao (tlci(cc) K Gas Volume (CTP,cc) X 1~0

( i/uci)

Where: A ~ The specific activity of .the containment atmosphere sample corrected to Standard Temperture and Pressure and for decay to the time of reactor trip.

(Enclosure A4) 5 '.4 The total quantity of fission products available for release to the environment is equal to the sum of the values obtained from each sample location (liquid and gas).

Enclosure A7 is provided as a worksheet.

5.7 Plant Power Correction A

The quantitative release of the fission products is expressed as the percent of the source inventory at the time of the accident. The equilibrium source inventories are to be corrected for plant power history.

I'l 0

(16 of 80) 5.7.1. To correct the source inventory for the case in which plant power level has remained constant for a period greater than four radioactive half lives the following procedure is employeds

--Enclosure K9 is provided as a worksheet.

5~7 ~ l. 1. The fission products are divided into two groups based upon the radioactive half lives. Group 1 isotopes are to be employed in the case where core power had not changed greater than

+ 10'4 within the last 30 days prior to.the reactor trip. Group 2 isotopes are to be employed in the case where core power had not changed greater than + 10'4 within the last 4 days prior to the reactor trip 5 ~ 7~1~2 The following equation may be applied:.

to the fission product Group, which meets the criteria stated in 5.7.1. l.-

Group 1 Power Corr'ection Factor .~

Avera e Power Level For Prior 3~3Da s 100 Group 2 Power Correction Factor ~

100

5. 7. 2 To correct the source inventory for the case in which plant power level has not remained constant prior to reactor trip, the following equation is employed. The entire 30 days power history should be employed.'nclosure A10 is provided as a worksheet.

Power Correction Factor ~-

gj Pj (1- e-ht 3

~

) e" ~to 7'00 C

Where' 3

~ Steady reactor power in period j t 3

~ ~ duration of period j (sec) t- ~ time from end. of period j to.

reactor trip (sec) decay constant (Enclosure A4)

(17 of 80) 5.8 Comparison of Measured Data with Source Inventory The. total quantity of fission products available for release".to the environment obtained in step 5.5.4 (Enclos'ure A7) is compared to the source inventory corrected for plant power history obtained in step

. 5.7 (Enclosure A9 or Al'3) This comparison is made

~

by dividing the total quantity available for release by the power corrected source inventory. Enclosure All is provided as a worksheet.

5.9 Core Damage Assessment The conclusion on core damage is made using the three parameters developed above. These are:

l. Identification of the fission product isotopes which most characterize 'a given sample, step 5.3..

(Enclosure A2).

2. Identification of the source of the release, step 5.5 (Enclosure A5).

3. Quantity of fission product available for release to the environment expressed as a percent of source inventory, step 5.8 (Enclosure All).

Knowledgeable judgement is used to compare the .above three parameters to the definitions of the 10 NRC categories of fuel damage found in Enclosure Al. Core damage is not anticipated to take place uniformly.

Therefore when evaluating the three parameters listed above the methodology in this section is anticipated to yield a combination o f one or more o f the LO categories defined in Enclosure Al. These categories will exist simultaneously.

The type of core damage is described in terms of the 10 SRC categories defined. in Enclosure Al. The degree of core damage is described as the percent of the fission products in the source inventory at the time of the accident which is'ow in the sampled fluid and therefore available for release to the environment.

I

ENCLOSURE Al RADIOLOGICAL CHARACTERISTICS OF NRC CATEGORIES OF FUEl. DAMAGE I 'ELEASE OF CIIARACTERISTIC NRC CATEGORY OF HECllhNISH OF SOURCE OF CllhRACTERISTIC ) ISOTOPE EXPRESS)D AS A FUEL DAHAGF. RELHASE RELEASE ISOTOPE PERCENT OF SOURCE INVENTORY

1. No Fuel Damage I Halogen Spiking, I Gas Gap ) I-131, Cs-131, Less thaq 1 Tram Uranium Rb 88

2. Initial Cladding Failure( Gas Gap I I Less than 10 I I

3. Intermediate Cladding Clad Hurst and Gas Ga's Gap Xe-131m, Xe-133, 10 to 50 Fa ilure I Gap Diffusion Release) I I-1311 I-133 I 4~ Ha)nr Cladding Failure Gas Gap I I Greater than 50

5. Initial Fuel Pellet Fuel Pellet I Less than 10 Ovcrhcatin ' 'Grain'oundary Cs-134, Rb-88

6. Intermediate Fuel Diffusion Fuel Pellet ) Te-129, Te-.132 10 to 50 Pellet Overhentin; 7~ Ha]or F<<cl Pellet ) Dtffus lonal Release Fuel Pellet Greater than 50 Ovcrhcnting From UO Grains

8. Fuel Pellet Melt Fuel Pellet Less than 10 J

I I I

9. 1ntcrmc<l l at e Fuel Fscape from Fuel Pellet aa-140, 1~-140, (

10 to 50 Pcllrt Melt Molten Fuel I La-142 Pr" 144 I

I ln. M.i ) ~ ir Furl Pcl let Melt luel Pellet I Gtcat.cr than 50 I

.{19 ot'0)

ENCLOSURE A2 {sheet I of 2)

INPUT PARAMETERS (ref. step 5.1)

Unit:

Reactor Coolant System:

SOURCE:

Pressure PSIG Temperature (Tavg) op Reactor Vessel Level Shows Full (Circle One) Void Below Recorder Pressurizer Level Containment Building:

Atmosphere Pressure PSIG Atmosphere Temperature Prior 30 Days Power History Power, Percent Duration, Days Estimat d Average Power Level During Last 30 Days 3 Estimated Average Power Level During Last 4 Days '

Time of Reactor Trip: Date: T ime:.

Change in volume of RNT gal. Time:

, Change in volume of BAHT 'al. T ime SIT in)ected (yes/no):

'I t

(20 of 80)

. ENCLOSURE A2 (sheet 2 of 2)

INPUT PARAMETERS - RADIONUCLIDE DATA (ref. step 5.3) 0 UNIT:

SAMPLE NUMBER-SAMPLE LOCATION RCS, SUMP i CONTAINMENT ):

TIME OF SAMPLE COLLECTION:

MEASURED SPECIFIC ACTIVITY 8 STP ISOTOPE A(uCi/cc)

Xr 87 Xe-131m Xe-.133 I-131 I-132 I-133 I-135 Cs-134 Rb-88 Te-129 Te-132 Sr-89 Ba-140 La-140 La-142 Pr-144 NOTE: NI means not identified Performed by:

Date:

A ENCLOSURE A3 SAMPLB LOChTIONS RECAQKNDED POR CORE DANlGE ASSESSNHHT {ref. ate 5.2)

I I I SIIIITDOWN I STEAM I ACCIDENT SCENARIO I RCS RCS CONTAINMENT I CONTAINMENT I COOI.INC I: CENFRATOR I KNOMN HOT ISC PRESSIIRIZER SI)MP ATMOSPIIERH SYSTEM SECONDhRY Small Break LOCh, I YES I YES YES I YEs I

-"- I Reactor Power )IX I Smnl i Break LOCh, I. YES YES YF.S I I

Reactor Power (IX I I Smail Steam I YFS I YES I

-- I I I.lne Break Large Break LOCh, I'ES YES YES I

'YES I I

I Reactor Power >IX I I.erg<<Brcak LOCA, I YES YES YES I I

) Reactor Power (lX I I large Steam I YEs I YES I I

I I.lne Break I I

'team Cenerator YES YES YES I Tulle Rupture I

• available only on recirculation

(22 of 80)

~ ~

ENCLOSURE A4 (REF STEPS 5.3 AND 5.4)

RECORD OF MEASURED SPECIFIC ACTIVITY (DECAY CORRECTED)

UNIT:

TIME OF REACTOR TRIP, ENCLOSURE A2-SAMPLE NUMBER: CONTAZ'i%MENT):

SAMPLE LOCATION RCS, SUMP, TIME OF SAMPLE COLLECTION:

DECAY MEASURED SPECIFIC DECAY CORR CTEO CONSTANT. ACTIVITY 9 STP- SPECIFIC ACTIVITY, ISOTOPE g(1/sec) A (uci/cc) Ao (uci/cc)

Kr 87 1.5 Xe 131m 6.7 ( -7)

Xe 133 1 ' -6)

I 131 9' I 132 8 4~ (-5)

I 133 9' -6 I 1.'3 5 Rb 88 5.5 Te 129 1 ~ 7 -4)

Te 132 2 '

Sr 89 1.6 Ba 140 6.3 4.

La 142 1 ~ 2 Pr 144 6.7 (

'I Ao ~

e-gt A, where A and k are as above, and t ~ time period seconds from reactor trip to sample collected.

in Performed by:

Date:.

NOTEs NI means not identified

ENCLOSURE A5 (ref. step 5.5)

RECORD OF FISSION PRODUCT RFLEASE SOURCE IDENTIFICATION Unit:

Sample Nuwher:

Inca t ion:

DECAY CORRECTED Activity Rati.o ACTIVITY RATIO IDENTIFIED I in I SPECIFIC ACTIVITY ) ChLCULATED ) FUEl'ELLET SOURCE (GAS GAP)

) I ISOTOPE (ENCLOSURErlf) nci/cc ISOTOPE RATIO* INVENTORY** INVENIORY ** OR FUEL PELLET)I Kr 87 0.2 < 0.001

) Xe 131>a 0.003 0.001 0.003 Xe 133 1.0 1.0 1.0 NA I I 131 1.0 1.0 1.0 NA I I 132 0.01 0.05 I 133 2.0 0.5 1.0 I 135 ~'.8 0.1 0.5 NA ~ NOT APPLICABLE

• Noble Gas Ratio Deca Corrected Nohle Gas S ecif ic hctivit Decay Corrected Xe 133 Specific Activity Iodine Ratio Decay Corrected Iodine Isoto e Specific Activity De@ay Corrected I-131 Specific Activity

~~ Table 3.3 of Reference 2.1 Performed By:

Date:

(24 of 80)

ENCLOSURE A6 (Sheet I of 2)

QUANTITATIVE RELEASE ASSESSMENT WORKSHEET (ref. step 5.6) a RCS ACTZVITY (AT RC )

RCS Tavg F (ref. step 5.1.1, Enclosure A2)

Vessel, Level Indication (Pull, Void, Below Recorder)

{ref. step 5.F 1, Enclosure A2)

ZF FULL OR VOID, perform the following calculation, for each isotope measured:'T RCS (Ci) Ao (uci/cc) X RCS Volume X 1 ' (-6) (Ci/uci)

Where:. Ao ~ decay corrected specific activity of RCS sample (Enclosure A4)

RCS volume = Water Volume X Density Ratio at RCS Tavg (Enclosure A8). PSLl water volume is 2.945 (8) cc and PSL2 water volume is 2. 888 { 8) cc.

Enter results in Enclosure A7 (AT, RCS)

IF BELOW RECORDER, Use Containment Sump Calculation Below.

SUMP ACTIVITY (AT )

Determine sump water volume by adding the following (ref.

step 5.6.2).

SL1. SL2 RCS SIT Volume Injected Volume BAMT Injected Volume

. gal gal gal

58. 300 34, '349 (Enclosure

'6,57,4'30 A2) 564 RWT Volume Change gal (Enclosure A2)

Vs ~ Total Sump Volume ~ gal x 3755 cc/gal cc sump Ao uci/cc ) X Vs X 1 ~ 0 ( -6 ) ( Ci/uci )

Where Ao .= decay corrected specific activity 'of SUMP sample (Enclosure A4)

Enter resul"s in Enclosure A7 (AT I sump)

r

~g I

(25 of 80)

ENCr,pSURE A6 (Sheet 2 of 2)

COMTAINMEST ACTIVITY {AT ~~~t.)

Calculate Containment Volume in cc, including pressure and temperature corrections (Ref. step 5.6.3)

Vc ~ Containment Volume (cc) 7 ~ 096 (10) X 14 ~ 7 + PL X 32 + 460 14.7 TL +, 460 Where: PL ~ Containment pressure in psig (ref. step 5.1.2, Enclosure A2)

Tl ~ C'ontainment temperature in P (ref.

step 5.1.2, Enclosure A2) cont ~ Ao (uci/cc) X Vc X 1.0 (-6)

Where:. Ao ~ decay corrected specific activity for containment sample (Enclosure A4)

Enter results in Enclosure K7 (AT I cont)

Performed .by:

Date:

'I (26 of 80)

WCXDSURE A'7 RECORD OP CORE RELEASE INVENTORY (ref. step 5.6.4)

EGT:

KRCK)R COOLANT CG87AXHMEST SUMP nmrAIX~Wr SA~ iiEERg ISOTOPE (

{Ci)Q~

SAMPLE Ã3MBER, RE a6)

+

(

(Ci)

RE A.e)

AXKSPHEZE SAMPLE NIJNBER~~

(m )

(Ci)

IOEAL QUANPI1Y

('i) "

Kr S7 Xe 131m Xe 133 I 131 I 132 I 133

-I 135 s 134 Rb 83 Te 129 Te 132 Sr 89 Ba 14'3 La 149 La 142 Pr 144 AT, RCS

+ r, SUM.

+ "r, CXm Perfoaned by:

ate:.

II III

(27 of 80)

EHCLOSURE AS-RATIO OP,H20 DEHSITY TO H20 DEHSITY, AT STP vs TENPERATURE 700

~ 'II ~

600 .I '

500 400 0

300 I '::

I . (

'": I

~ ~

'00

~ I..:

100 0 .25 .50 .75 1 '

PACT f7 STP g DENSITY RATIO

(28 of 80) 28CXDSURE A9-PSLl (ref. step 5.7)

RECORD OP STEADY'TATE POWER CORRECI'ION 0

UNIT: Performed By:

AVERAGE 30 DAYS POWER LEVEL: Dates AVERAGE 4 DAYS POWER LE>VEL SLL HJEL POWER EQUILIBRIUM CORRIC7ED HISIOKC CO~erXON X SOUSE RUTE ISOIOPE GRXJPING .PAt IOR INVIWIORY INVETIORY Xe 131m 4.6 4 Xe 133 L.3 7 I 131 7.0 6 I 132 7.7 3 133 6.7 6 135 1.1 (6)

HJEL PEIL2 P INVEN'IORf Kr 87 3.2 7 Xe 131m 4.9 5 Xe 133 L.5 8 I 13L 7. 7 I 132 1.1 8)

I 133 I 135 1.4 8 CS 134 1'2 7 Rb 88 Te 1 9 Te 132 Sr 89 6.6 (7) 1.4 8 1.4 (8)

La L42 1 5 8 Pr 144 9.6 7)

Corrected source Inventory ~ Pcwer Correction Pactor X Equilibrium Source ventory alues fran Table 3.4 and 3.5 of Reference 2.1 Group 1 Pmer Correction, Pactor ~ Avera e Level for Prior 30 da s 100 Group 2 Pcwer Correction, Pactor ~ Avera e Level for Prior 4 Da s

~'

0 4

(29 of 80)

ZmLOSURE A9-PSL2. (ref. step 5.7)

RECQ@) OF SEEM% SPATE POWER CORRECTION PiGT: Performed By:

AVERAGE 30 DAYS POWER LEAK: Dat'e:

AVERAGE 4 DAYS POWER LEVEL:

SL2 PC%DR SQUILIBRIUH CDRR C1ED CORRECTION X SOUFFLE SOUKZ FACTOR IHVEN'IVORY" INVZRIQRY GAS GAP INRRVIORY Kr 97 6.3 Xe 131m Xe 133 1.3 7 I 131 I 132 7' 133 6.7 6 135 L.L (6)

HJEL PEELZH'HVEKLQRY Xe 131m Xe 133 ~ L.5 8 I 131 7~

I L32 L.O 8 I L35 M 134 1'9 7 Te 129 Te 132 L.O 8 Sr 89 Ba 140 L.

La 143 L.3 9 La 142 Pr L44 9.1 (7)

Corrected source Inventory ~ Pcwer Correction Factor X Ecpilibrium Source nventory

~ values fran Table 3.4 and 3.5 of Reference 2.1 Group 1 Pmer Carrecticn, Factor ~ Avera Level for Prior 30 Da s 100 Group 2 Pcwer Correction, Factor ~ Avera Level for Prior 4 D s

(30 or d0)

ENCLOSURE A13-PSLL (ref. step 5.7.2)

RECORD OF TRANSIENT POWER CORRECTION UNIT: Performed By:

'ates Prior 30 Days Power History: POWER DURATIONS TIME TO TRIP, Days (t~) Days (t~)

SLl POWER EQUILIBRIUM CORRECTED CORRECTION X SOURCE SOURCE'"

ISOTOPE FACTOR INVENTORY INVENTORY GAS GAP INVENTORY Kr 87 6.5 0 Xe 131m 4.6 4)

Xe 133 1' 7 I L31 7.0 (6)

I 132 7.7 3)

I 133 6 ' 6 I 135 1 ~ 1 6)

FUEL PELLET INVENTORY Kr 87 3 ' 7 Xe 131m 4~9 5)

Xe 133 1 ~ 5 L31 7.6 7)

I L32 8)

I 133 1 ~ 5 I 135 1~4 8)

CS 134 L~ 2 Rb 88 4.8 Te 129 2 '

Te 132 1 "2 (8)

Sr 9 6 ' 7 Ba 140 1 ' 8)

La L 0 L ~ 4 S La 142 1.6 (8)

Pr 144 9~6 7 Corrected Source Inventory ~ Power Correction Pactor X Equilibrium Source Inventory

• Values from Table 3.4 and 3.5 of Reference 2.1

(31 of 80)

ENCLOSURE A10-PSL2 (ref. step 5.7. 2)

RECORD OF TRANSIENT POWER ORRECTION e A UNIT: Performed By:

Date:

Prior 30 DaYs Power History: POWER, '8 DURATION TIME TO TRIP, Days (t~) Days (t )

SL2 POWER EQUILIBRIUM CORRECTED CORRECTION X SOURCE SOURCE ISOTOPE FACTOR INVENTORY INVENTORY AS AP INVENTORY Kr 87 Xe 131m 4' 4 Xe L33 1 ~ 3 7 I L31 6.7 6 I 132 7.

I 133 6 ' 6 I 135 1.1 (6)

FUEL PELLET INVENTORY Kr 87 3 ~ 1 7 Xe 131m 4 6 ~ 5 Xe 133 I L31 7 ' 7 I L32 1 ~ 0 8 I l33 1 ~ 5 8 I 135 1 ~ 3 9 CS 134 4' 7)

Rb 8 6' 7 TG 129 2~4 7 Te 132 1 ' 9)

Sr 89 1 ~ 9 Ba 140 1 ~ 3 8)

La 140 1 ~ 3 La 142 1~6 8 Pr 144 9.1 (7)

Corrected Source Inventory ~ Power Correction Factor X Equilibrium Source Inventory

• Values from Table 3.4 and 3.5 of Reference 2.1

J(~

.0

k (32 of 80)

ENCLOSURE All (ref. step 5.8)

RECORD OF PERCENT RELEASE UNIT: Performed By:

Date:

TOTAL QUANTITY POWER CORRECTED AVAILABLE FOR RELEASE SOURCE INVENTORY Ci ISOTOPE ( ENCLOSURE A7) Ci (ENCLOSURE A9 OR A10) PERCENT*

GAS GAP INVENTORY Kr 87-Xe 131m Xe 133 I 131 I 132 I 133 FUEL PELLET INVENTORY Kr 87 Xe 131m Xe 133 I 131 I 132 I 133 I 135 CS 134 Rb 88 Te 129 Te 132-.

Sr 89 Ba 140 La 140 La 142 Pr 144 Percent Total Quantity Available for Release Power Correcte Source nventory

(33 of 80)

S" CTION 8 CORE DAMAGE ASSESSMENT USING HYDROGEN

I I

{34 of 80)

III TABLP OF CONTENTS l.0 PURPOSE "31

2.0 REFERENCES

3l 3.0 DEFINITIONS 3L 4' PRECAUTIONS AND LIMITATIONS 32 5 ' INSTRUCTIONS 33 5.1 Record of Plant Condition 33 5.2 Hydrogen Sampling 34 5.3 ,Containment Hydrogen Generation 34 5.4 Radiolysis Hydrogen Generation 34 5.5 Core Clad 35 Percent of Ruptured Fuel Rods Oxidation'.6 35 5.7 Percent of Embrittled Fuel Rods 36 5.8 Procedure Bias Adjustment 36 5.9 Core Damage Assessment 36 6.0 ADDENDUM TO SE TION B ESTIMATION OF AMOUNT OF HYDROGEN IN REACTOR VESSEL HEAD VOID LIST OF ENCLOSURES Enclosure Bl Clad Damage Characteristics of NRC . 37 Categories of Fuel Damage Enclosure B2 "Record of Core Uncovery Conditions Enclosure B3 Record of Sampling Conditions and Measured Hydrogen Enclosure B4 Ratio of Water Density at Sample Temperature to Density at STP, as a Function of Temperature Enclosure B5 Record and Calculation Worksheet for 4l Hydrogen Generated in Containment Enclosure B6-SLl Hydrogen Production Rate in 42 Containment as a Function of.

Temperature for St. Lucie Unit'l Enclosure B6-SL2 Hydrogen Production Rate in Containment as a Function of

~~maerature for St. Lucie Unit

(35 ot'0)

TABLE OF CONTENTS (CONTINUED)

PAGE Enclosure B7 Record and Calculation Worksheet for 44 Hydrogen Generated by Radiolysis Enclosure BS Hydrogen Production by Radiolysis as a Function Time After Reactor'rip Enclosure B9 Core Damage Assessment from Hydrogen 46 Measurement Enclosure B10 Percent of Ruptured Rods as a Function 47 of the Percent of Core Clad Oxidi,zed Enclosure Bll Percent of Embrittled Rods as a-Function of the Percent of Core Clad Oxidized Enclosure BBL Saturated Water Pressure and 53 Temperature Enclosure BB2 Record of Hydrogen In Void 54

0 (36 of 80)

I ~ 0 PURPOSE:

This section provides the methodology for use under post accident plant conditions to determine the extent of fuel clad damage which may have occurred. It utilizes hydrogen measured in samples obtained with the Post Accident System (PASS) and containment hydrogen analyzers.

'ampling The measured hydrogen is related to the amount of fuel clad oxidation. Clad oxidation is in turn related to clad damage which is expressed in terms of the percent of fuel rods which are ruptured and the percent which are embrittled. The resulting observation of damage is described by one or more of the seven categories of core damage in Enclosure Bl.

2.0 REFERENCES

2.1 Development of the comprehensive procedure guideline for core damage assessment, CE Owners Group Task 467, July 1983.

2.2 Operation of the CE Post Accident Sampling System (PASS). Chemistry Procedure No. 1-C112 for PSLl and No. 2-C113 for PSL-2.,

2.3 Clarification of TMI action plan requirements'UREG 0737, Item II.B.3.

2.4 Determination of Hydrogen gas in containment.

Chemistry Procedure No. 1-C-80 for PSL-1 and 2-C-80 for PSL-2.

3.0 DEFINITIONS

F 1 internal gas pressure exceeds the external coolant pressure and the clad yield strength is reduced because of elevated temperatures. Clad rupture results in release of gaseous fission products from the gas gap and possibly some fragments of fuel pellets but does not otherwise destroy the structure of the fuel assembly.

3.2 Clad Embrittlement:. At temperature above the rupture temperature signi:ficant oxidation of the clad occurs. I f the oxidation exceeds the embrittlement threshold. fragmentation of embrittled clad may subsequently occur from thermal shock, hydraulic pressure forces or handling such that the structure of the fuel assembly is destroyed and substantial fuel pellet fragments are released to the coolant.

I (37 of 80)

"4 0

~ PRECAUTIONS AND LIMITATIONS:

4'I The assessment of core damage obtained by using this methodology is only an estimate. The techniques employed in this section are only accurate to locate the core condition within one or more of the 7 categories of core damage in Enclosure Bl. The methodology is based on hydrogen data. Other plant indications may be available which can improve upon estimation of core damage. These include radiological sample. characteristics, incore temperature indicators, and containment radiation monitors. Whenever possible these additional indicators should be factored into the assessment.

4' This methodology relies upon hydrogen samples taken from the containment atmosphere and the reactor coolant system hot leg. Those samples may contain mixture of hydrogen generated within the core by clad oxidation and also hydrogen from radiolytic dissociation of water and oxidation of aluminum and zinc in the containment. ,'The estimate of clad damage is influenced by the amount of hydrogen generated by ex-core'sources and, by the ability to identify plant conditions conducive .to such hydrogen generation.

Therefore, a hydrogen measurement is not a unique indicator of the amount of core clad oxidation.

4 ' There are areas of aluminum components in the containment building. This aluminum would oxidize rapidly at temperatures about 200 F and would be consumed within about two hours. The remainder of the aluminum and other oxidizing material react at a rate determined by temperature and over a longer time. The methodology in this section assumes all of the sport term transient hydrogen is generated within the first two hours and is added'o the slower accumulation as function of time. Hence, the methodology is valid for hydrogen samples taken after about two. hours with temperatures about 200 F, or'after the short term oxidation r

is complete.

4.4 The methodology in this section yields estimates of the percentages of fuel rods with ruptured clad and embrittled clad. Simultaneous with embrittling of the clad, there may'e clad melting and pellet overheating occurring. This section provides an estimate of only the percentage of rods which have progressed to at least clad rupture or clad embrittlement, and does not attempt to predict and physical configuration of .those rods which have progressed beyond local clad fragmentation.

Depending on the accident scenario, a given total amount of hydrogen produced by oxidation of fuel clad

(38 o.f 80) 4 can represent varying local amounts and distributions of clad damage.

4.6 The methodology in this section is applicable under conditi'ons for which there are no voids measurable by the Reactor Vessel Level Monitoring System. It is assumed that if such voids had been found, their

• removal would be accomplished by using the Reactor Vessel Vent System as prescribed elsewhere in the actions to mitigate the consequences of if the hydrogen samples are taken under accidents'owever, conditions in which measurable void does exist, a guideline for analysis is provided in the addendum attached to this section to estimate the contribution of that source to be added to the total hydrogen

'measured.

5.0 INSTRUCTIONS

5.1 Obtain the Following Plant Indicators.

5 ~ I ~ 1 Core damage can occur following reactor trip only when the coolant level within the reactor vessel drops below the top of the active fuel.

Several instrument records are available from which an estimate of the core uncovery and recovery times might be made. The instruments are:.

Reactor Vessel Level Monitoring System Core Exit Thermocouple Temperature Core Exit Thermocouple Saturation Margin Obtain data from these instruments according to the instructions on .the worksheet of Enclosure B2.

5 ' ~ 2 The magnitude of Reactor Coolant System (RCS) pressure during the core uncovery period can influence the number of early clad ruptures.

Interpret -the data from Step 5.1.1 to determine the best estimate for the time period of core uncovery and determine the range of RCS pressure during this time period. Record on the Enclosure M worksheet.

5 ' ' The presence of some subcooled inlet flow while the core is uncovering can slow the uncovery and cause greater local clad oxidation for a given total amount of core oxidation, thereby leading to a greater underestimate of the number of damaged rods predicted by this procedure. Observe available instrument records to determine if there was some reactor vessel inlet flow during the rising temperature

C gh

(39 of 80) portion of the core uncovery period. Include net flow from charging and letdown systems, HPSI, LPSI, spray, etc. Record-the data on the

..Enclosure B2 worksheet.

5.1.4 Record the conditions in the containment and the reactor coolant system at the time the

'ydrogen samples are obtained in Step 5.2 fol-lowing., Enter on the worksheet of Enclosure B3.

5 ' Obtain a liquid sample from the RCS hot leg and a gas sample from the containment atmosphere and analyze them for hydrogen concentration using the procedures for Post Accident Sample System operation described in Reference 2.2. Record the results on the worksheet of Enclosure B3. Follow the instructions on Enclosure B3 to obtain the total amount of hydrogen measured in units of cubic feet of hydrogen at standard temperature and pressure.

5 ' The total measured hydrogen in, Step 5.2 includes the hydrogen generated by three processes: 1) core clad oxidation, 2) radiolysis of water and 3) oxidation of containment materials such as aluminum and zinc. The

-amount of hydrogen generated by the last two processes is calculated and then subtracted from the total measured to yield the amount generated by core clad oxidation.

Enclosure B5 is a worksheet for calculating the amount of hydrogen generated by oxidation of materials within the containment. It utilizes measured data for, the containment temperature as a function of time up to the sampling time and a plant 'specific curve of the rate of production as,a function of containment.

temperature in Enclosure B6. Record the. data required on Enclosure B5 and complete the indicated calculations to obtain the cubic feet of hydrogen at STP generated by containment materials oxidation.

5.4 The hydrogen generated by radiolysis is a function of operating power and decay time. Record the data required on the worksheet of Enclosure B7, and utilize the -curve of Enclosure BS to obtain, the cubic feet of hydrogen at STP generated by radiolysis. The appropriate power is determined as follows:.

5.4.1 For the case in which the operating power is constant or has not changed by more than"+ 10 percent for a period greater than 30 days, that power is used.

' t ~

(40 of 80) 5.4 ~ 2 For the case in which the power has not remained constant during the 30 days prior to the reactor, trip engineering judgement is used

--to determine the most representative power level. The following guidelines should be considered in the determination.

5.4.2.1 The average power during the 30 day time period is not necessarily the most representative value for determining radiolysis by fission products.

5.4.2.2 The last power levels at which the reactor operated should weigh'more heavily in the judgement than the earlier levels.

5.4.2-3 Continued operation for an extended period should weigh more heavily in the judgement than brief transient levels.

5' ' For'he'ase in which the reactor has produced power for less than 30 days, this methodology may be employed. However, the estimate of hydrogen from radiolysis will be too high'and therefore the cal'~ulated hydrogen by core oxidation will be too low. Hence an underprediction of core damage may result.

5.5 Enter the amounts of hydrogen from Steps 5. 2, 5. 3 and 5.4 on the worksheet of Enclosure B9. Subtract the amounts in'5.3 and 5.4 from 5.2 as indicated on the worksheet to yield the cubic feet of hydrogen generated by core clad oxidation. Adjust with the plant specific constant as shown on the worksheet to obtain the estimated percent of the core clad which is oxidized. This value represents the quantity of hydrogen produced per percent of zirconium oxidized.

5.6 Enter the abscissa of the curve on Enclosure BLQ'ith the percent of core clad oxidized from Step 5.5. Use the curve labeled with the pressure closest to but greater than the RCS pressure during the core uncovery period as obtained in Step'5.1.2 and recorded on Enclosure B2,- e.g. if pressure during core uncovers is 1300 psia, the curve labeled with temperature 1SOO F is used. Read on the ordinate of Enclosure B10, the percent of fuel- rods with ruptured clad. Record on-the worksheet of Enclosure B9. Mote that the sensitivity of measurement of hydrogen is comparable to the range of oxidation on Enclosure B10. Hence, small amounts of clad rupture are not reliably predicted by the methodology in this section.

E

~

0

41 of 80) 5.7 Enter the abscissa of the curve on Enclosure B11 with the percent of core clad oxidized from Step 5. 5. Read on the ordinate the lower and upper values of the range indicated by the curve for the percent of fuel rods which have embrittled clad. Record on the worksheet of Enclosure B9.

5.8 For a given percent oxidation of the core clad, the lower limit estimated of embrittled clad in Step 5.7 is, for most accident scenarios, the least amount of potential fuel structural failure. Actual values are probably greater. The upper limit of the range i*n Step 5.5 may be interpreted as follows:

5.8 ~ 1 When the pressure during uncovery, from, Step 5.1.2 and recorded on Enclosure B2, is less than about 100 psia, .a rapid core uncovery by blowdown is concluded. Heatup with minimum clad oxidation occurs. The extent of potential clad structural failure by melting may be greater than the upper limit of embrittlement from Step 5.7 as determined by oxidation.

Hence, use the upper limit from Step 5.7.

5.8.2 When there is inlet flow while the core is uncovering, the rate of uncovery is slower than assumed in the derivation of the curves on Enclosures B10 and B11. For a measured total .

amount of oxidation, the local percentage oxidation is probably greater along a shorter length of the upper portions of the fuel.

Hence, favor the upper limit from Step 5.7.

5.9 Core Dama e Assessment The conclusion on core damage is made using t'e two results from above. These are:

1.

5 '.

Percentage of fuel rods with ruptured clad, St'ep 2 ~ Percentage of fuel rods with embrittled or structurally damaged clad, Step 5.7.

Knowledgeable judgement is used to compare the above two results to the definitions of the 7 ~VRC categories of fuel damage found in Enclosure Bl. Core damage does not take place uniformly. Therefore when evaluating damage using these results, Enclosure Bl may yield a combination of categories of damage which exist simultaneously.

ENCLOSURE iz CLhD OWNS CHARhCTERlSTlCS Of NC CATEGORlES Of RlEL QhNGE NRC Category Tcaperaturo Nechanisa Characteristic Ncasurint Percent of of Fuel Oaaa ka 'F of Oaaa Neasurea.nt Ra Oalla Rods

1. No Fuel Oamge I Less Than 1 I

2. Initial Cladding Rupture Oue to  :

Naxima Core <1550 F* Less Tea Failure Gas Gap 'xit 10 .

3. lntenaed}atc 1200-1800 Ovcrpressurization Thelcouple <<1700F>>  ; 10 to 50 Cladding Failure . Teapcrature I. Na)or Cladding <2300'F Failure I <2% Greater Than 50 I

I Oxidatioa I

5. 'nitial Fuel Pcllct Loss of Structural ~

Aaount of Equivalent Core Less Than 10

6. lntc I Overheating diate Fuel Pellet Overheating 1800-3350 lntcgrity Ouc to Fuel Clad Oxidation I Hydrogen Gas Produced (Equivalent to

$ Oxidation of Core) .

Oxidation

<35 Na)or Fuel Pellet <65% Greater Than 50 Overheating

'epends on Reactor Pressure and Fuel Burnup.Va)ves Gtven for pressure <]200 ps)a and Burnup ~0.

0 (43 of 80)

ENCXQSURE B2 CORE UNCOVERS QNBITIONS Step 5.1.1'ime period of core uncovery. Ccaplete the following table using recorded instrument 4

data.

Estimated Estimated Instrument Core Uncove Time Core Recove Time Reactor Vessel Level Lower Limit Elevation Lmmr Limit Elevation Hanitoring System Uncovers (core Recovers.

uncovery) TICK TiIl&

Core Exit ThermxxIuple Start of Ccntinuous Rapid Tetlperature Temperature Rise or Exceed 660 F Drop to Saturation TlJT& Time Temperature Telrperature Core Exit Theamcouple Start of Superheat Return to Saturation Saturation Margin TiJlle or Subcooling Till&

Step 5.1.2 Interpret above data to obtain best estimate for time period of core uncoyery and obtain pressurizer pressure range during that period. The superheat derived fran the thenracouple 'tengerature and corresponding system pressure is considered as the best indicator for core uncovery during boiloff and should be used, but should be compared with the other indicators to help identify possible ancmalies. The pressure during uncovery is used later in Enclosure Bl'3, Step 5.6, to determine the appropriate curve for assessment of the nun¹r of clad ruptures.

Pressure Step 5.1.3 Estimate vessel inlet flee rates during core uncovery heatup period, up to approxiInately the time of peak core exit thermcouple teaperature. Net inlet f1~ ~icates that the methodology may have additional bias which underpredicts clad damage.

Charging Flmr Rate Letdown Fleer Rate HPSI Flmr Rate LPSI Fleur Rate Other Inlet Flcws Net inlet flow ~ Ch-~ing flottr + HPSI and LPSI flour + other inlet flmr - letdcam flew Performed By: Dates

(44 of 80)

EHCXDSURE B3 SMPLLD CXN3TTIOMS 5%3 MEASURED HYDRQGE8 0A Step 5.1.4 Obtain the RCS and containment conditions at, the time'of sanpling for hydrogen.

Reactor Coolant a stem Ccntainmnt Sanpling Time Pressure Taqpterature, Tavg

~ig F Sanpling Time htrlaspllere Pressure Atnasphere Tenperature

~S Lg Reactor Vessel Has Hydrogen Reconbiner Yes/No Coolant Level Operated Pressurizer Level Does Pressure or Tenpera- Yes/Mo ture History Indicate a Hydrogen Burn Step 5.2 Hydrogen Sample Data Reduction. ~

Cont. Sanple x Cont. Vol. x (32 + 460) (Hormal Temp.) Ft Hydrogen (Vol. a/100) (Ft3) + 460 at STP x 2.5 (6) x 492 Ft3 Hot Leg Sanple x KS Vyl.* x Density Ratio 1000 Ft Hydrogen (cc/kg '3 SZP) (Ft ) (Enclosure B4) at STP x act/ SPP 130'3 ft3 Total = Cont. Sanple (ft ) + Hot Leg Sample = +

Also record total an Enclosure B9.

• RCS licpxid value is: SLl ~ 10,401 ft SL2 ~ 10,198 ft Mote:. If the Reactor Vessel Coolant Level indication shcws a mea'surable void, refer to the addendum to this section. This addendum contains instructions to calculate Hydrogen in the void. The void volume should be subtracted from the

'CS volund above. The estimated hydrogen in the void is to be added to the total hydrogen measured above.

Perfarmed By: Date

~ s 4 c}

0

~ ~

(45 of 80)

ENCLOSURE $4 RATIO OP 820 DEHSITY TO H20 DEESITY J

AT STP vs. TBNPERATURB 700 600 500 400 0

300 200 100 I~

0 0 .25 .50 .75 :1.0 PACT PSTP, Density Ratio W

~ ~ ~y 0

~ I ~ ~

(46 of 80)

PSCfDRRE BS HYDKGEl %REREAD Qi CÃZKQRWP 0 ~

Step 5.3 Record the ccntainment terrperature at selected time intervals and calculate the hydrogen generated by oxidation of ccntainnent materials utilizing the plant-specific production rates fran Enclosure B6.

~(~+i', "

4 5 6 Avg. Ccntainnant Time at Start Ccntainrrrent Interval Tarp o During Q Produced ~

of Intervals T ture ( P) Duration (hr) Interval ( F) Enclosure B6) (CoT 3) X (Col 5) hccident Starts anple Time Long Term Hydrogen Producticn in Contaiarant, Total (Surmaticn of Colurrn 6)

Short tep rapid hydrogen p~ction by ccntainrrent aluminum, 2,277 ft for SLl and 5235 ft for SL2 (Reference 2.1 +

Table 4-3, )

Total Hydrogen Producticn in Ccntainrnent Record total cn Enclosure B9 also.

1 and 2 Items in Columns 1 and 2 are input plant data 3 Interval Duration is the line difference between consecutive tenperture readings.

Performed By:

Date:

E ~ 4

~ I (47 of 80)

EHCLOSURE B6 - SLl HXDBOGEH PRODUCTION RATE FROM, ALUMINUM AHD ZINC VS TEMPERATURE FOR ST LUCIE UNIT 1 7200

~ ~

6400

~ I tt'I 5600 O

N

~ ~

4800

~

I 4000 I

I I

3200 I

I I

2400 1600 i

I I

800

~ ~

I 120 140 160 180 200 220 240 260 280 300 CONTAINMENT TEMPERATURE, F

g ~

(48 of 80) e ENCLOSURE B6 SL2 HYDROGEN PRODUCTION RATE PRON ALUMINUM AHD ZINC VS TEMPERATURE POR ST LUCIE UHIT 2 I l I

7200 6400 I

I ~ ~

I U I, 5600 R

0 H ."e I

4800 5

8 4000 I I l4 CI 3200 2400

~ I 1600 8OO I ~

~

' t 100 120 140 160 180 200 220 240 260 2 0 3 0 CONTAINMENT TEMPERATURE P

~ f (49 ot 8O)

BlCXDSURE B7 HYDKQEN GWEfWZED BY RADIOLYSIS Step 5.4 Record the following data and utilize the curves of Enclosure BB to determine the hydrogen generated by radiolysis.

Prior 30 days power history Note:. No calculation is recpxired to determine power level, guidance on judgement is provided in Step 5.4.

Estimated Pester Level based on a power history:

ratin Pcver (Mwt):

Pmer to use in evaluating long term hydrogen production by radiolysis ~ (Full P~er, Mwt) X Pcarer Level.

100 (Full Paver:. SL1 = 2690 %t; SL2 ~ 2560 Mwt]

T = Tire of Reactor Trip Time Ti = Time Sample "Taken (see Enclosure B3)

Decay Tame T~ Int rval'i To Hours Enter abscissa on Enclosure BB with above decay tom and read two values of hydrogen produced by radiolysis, one frcxn each curve, in cubic feet of hydrogen at STP per Rat ooerating pcwer. Multiply by above poorer and record as fOll~s:.

Hydrogen Produced x Operating Total Hydrogen Limit Curve (SCF/~, Enclosure BG) P~er (Mwt) Produced (~)

Upper Lower Results fran Radiological Analysis of Samples are used to estimate whether the upper limit for major fuel overheat or lower limit for intermediate fuel overheat is appropriate. Circle corresponding value of hydrogen above and also record on Enclosure B9..

Performed By- Date:

ENCLm3QRE BS SPECIFIC RADIOMTIC EYDROGEQ PRODUCTION VS TINB 13 u

11 10 8

I I

7 Ij ~

INT!3RHEDIATE FUEL OVER33EAT j'

~

~

~ ~ ~

~~

INITIlQs tUEL OVB563E3IT

~

l 0 100 200 300 400 500 600 700 800 O

h DECAY TIHEi HOURS 00 CD

s%

(51 of 80)

ENCXDSURE B9

'ORE DR%BE ASSESStKNT FR%

M3RQGBl ME~REMEÃP Step 5.5 Hydrogen Measured, Step 5.2, Enclosure B3(Total)

Hydrogen Produced in Containment, Step 5.3, Enclosuie B5 Hydrogen Produced by Radiolysis, Step 5.4, Enclosu're B7 Subtract Step 5.3 and 5.4 fran 5.2 to Get Hydrogen Produced by Core Clad Oxidation Divide by t.4.21 E3 for PSL13 or I.4.64 E3 for PSL2j. These values represent the quantity of hydrogen produced per percent 0 Core Clad Oxidized of Zirconium oxidized for St. Lucie Unit 1 and Unit 2, respectively. Reference 2.1, able 4.2.

Step 5.6

~ Enter abscissa on Enclosure B10 with "4 Core Clad Oxidized" and read

~

ordinate frnn curve labeled with pressure during core uncovery as

~

given on Enclosure B2, Step 5.1.2. Record here Percent of Fuel FLds with Ruptured Clad Step 5.7 Enter abscissa on Enclosure B11 with above "0 Core Oxidized" and read range of values on ordinate. Record here Percent of fuel rods errbrittled Range - Upper

- urer Step 5.8 Review Step 5.1 of these instructions to determine which of these limits is narc likely to be representative of the core damage.

Step 5.9 Frcxn Enclosure Bl select the core clad danage categories based on the above percentages of rods ruptured and rods eabrittled. e Aggroved By: Date:

~ ~

(52 of 80)

EHCLOSURB H10 PERCEHT OF FUEL RODS WITH RUPTURED CLAD VS CORB CLllD OXIDATIOH 1200 F 100 t '::I.

RUPTURE TEMPERATURB 80 1500 F 60 1800 F D

Q 40 a

0 Oo WHEN THB PRESSURE USE CURVB LABELED IH STEP 5.1 2 IS TEMPERATURB 'ITH 0 +100 PSIA. 1200'F 20 C1200 PSIA 1500'F 41650 PSIA 1800 F 0

0 0.5 1.0 1.5 2.0

\ OXIDATIOH OF CORB CLAD VOLUME

~ ~

~

(53 of 80)

EHCLOSURB all

'I OP THB PUEL RODS WITH OXIDATIOH EKBRITTLBHEHT VS TOTAL CORB OXIDATIOH POR 1% TO 3% DECAY HEAT AHD 300 PSIA TO 2500 PSIA WHBH.COOLAHT LBVEL DROPS BY SOILOFP WITH HO IHLBT PLOW UHTIL CORB IS RAPIDLY QUEUED 100 80 0H A 60-M 0

40 g

20 0

0 100 0 20 40 60 80 4 OXIDATIOH OF CORE CLAD VOLUME

~~

(54 of 80) 6.'3 ADDENDUM TO SECTION B ESTIMATION OF AMOUNT OF HYDROGEN IN REACTOR VESSELS HEAD VOID

~ ~,

(55 of 80)

The purpose of this addendum is to provide the methodology to calculate t'e amount of hydrogen gas contained in a void in the top of 'the reactor vessel. This hydrogen is added to" the measured amount in Step 5.2 of Section 8 to determine the total hydrogen-generated by all sources.

2 ' LIMITATIONS:

2.1 The preferred method of determining the amount of hydrogen in the primary system is to sample liquid from the hot leg when the system is full. However, the system cannot be filled, a method based on this if addendum could be used to estimate the hydrogen which is in the vessel void and which would not be evident from the hot leg liquid sample.

2 ' This method applies when the coolant level is above the hot leg and the remainder of the primary system is filled. Verification that the steam generator tubes are filled can be provided by the existence of natural convection flow in the primary system. If the coolant level is below the hot leg, this method does not apply.

2' A reactor vessel level monitoring system is required which can provide the coolant level. The volume of the void is obtained by relating the volume in the vessel above the coolant level to the value of level.

\

2.4 The methodology in this addendum provides the analytical means for only an estimate of the hydrogen contained in the void. The presence of other gases

=including-helium, nitrogen'and fission product gases will add uncertainty to the result.

3 ' INSTRUCTIONS:

3 1 Determine the conditions of the void as follows:.

V ~ Void volume (Pt 3 ) derived from measurement of coolant level TL ~ Temperature of liquid at coolant surface ( P) as measu'red by CET Water saturation pressure at temperature TL (Enclosure 881)

Pt t ~ Reactor coolant system pressure (psia)

(56 of 80) 3.2 A first approximation is made ass6ming the following:

3 '.1 The partial pressure of vapor in the assumed equal to saturation pressure void is at the liquid temperature, TL. This implies no

.heating of the void gas by the reactor vessel walls and head. They are normally at reactor outlet temperature and could remain above the temperature of the void causing the vapor to be

'uperheated.

3.2 ' All the non-condensible gas in the void is hydrogen. This implies no helium or fission product gas from ruptured fuel rods and no nitrogen from Safety Injection Tanks. A second approximation which eliminates this assumption is given in 3.4.

3.3 Calculate the amount of hydrogen as follows:

H tot sat' H

Ft H2 9 STP ~ (V)( )( )

14.7 TL + 46O Add this amount to the total hydrogen in Step 5.2 of Section B.

3 ' A second approximation can be made in plants with a CE designed PASS.(i.e. PSL2) which measures both total gas and hydrogen which are dissolved in the hot leg liquid sample. This approximation includes the following assumptions regarding the relative solubilities of the non-condensible gases in the-liquid.

3.4.1 The gases are assumed to have the same values of Henry's law constant which relates the partial pressure of gas to the amount of gas dissolved in the liquid sample at equilibrium.

3.4.2 When, the dissolved gas is not in equilibrium with the gas in the void, the dissolved

, concentrations are in the same relative proportion as if equilibrium did exist.

3.5 The partial pressure of hydrogen is calculated from:

H tot sat g

=X ~ g H (cc/kg) t t 1 and the amount of hydrogen in the vessel head void is

~q ~ ~

(57 of 80) 3.6 This procedure can be extended to include specific values of Henry's law constants but the assumption of equilibrium at the gas liquid interface would still be questionable. Also, to utilize detailed values of the gas cogptants, the individual gases in the sample would have to be identified and measured. This would require additional measurement capability.

ENCLOSURE BBl SATURATED MATER PRESSURE MlD TEMPERATURE 3000 2500 I

~

lg 2000 I:

jl I~ ,ei I~

I i ifl I I

I ~

h el I:

~ e I 1500 I e e! hl, ~

~

I~

li j:I: e

~ e I&l ~

I ~ ~

I \ I I i,

~~

I' le ~  ;, I I

!i jl il I I~ g]

~ ~ ~ I ~

I!

I I e e

e I!

e I

y, 1000

'-I'.I II "i

l4:

~

III iL' 500 II li I

j i' I I~

300 350 400 450 500 550 600 650 700

'50 TEMPERATURE, P QO C)

~ t (59 of 80) mCIDSURE BB2 RECORD QF HYDKGEN IM VOID Reactor Vess 1 Ccolant Level Indication ( 1 tn 8 ): VL Percent Void Height VL x 100 8 8

Note:. 18 correspcads to approximately 0. 135 Pt.

Void height above fuel aaligrrrrent plate ~ 8 x 0.188 Ft lt V ~ Void Volute ~ 105 ft X Void Height (Ft) ~

ft TI Temperature of Liquid at coolant surface ( P) as measured by CET.

Pressure at TI, ran Enclosure BBl P~ ~ KS pressure

=

(psia)

Anuunt of Hydrogen Calculation: PH ~ P~ - P p

Ft H 8 m y ( > ) ( 492 )

14e7 TI, + 460 Md this armunt to the total hydrogen in Step 5.> of Section B (Enclosure B3)

~U&18d Bye Dater

r, (60 of 80)

SECTION C ASSESSMENT OP CORE DAMAGE USING CORE EXIT THERMOS OUPLES TEMPERATURES

~ g ~ ~

0

(61 of 80)

TABLE OF CONTENTS PA" E 1.0 PURPOSE 57 2 ' REFEREH ES 57 3.0 DEFINITIONS 57 4 9

~ PRECAUTIONS AND LIMITATIONS 57 5 0 INSTRUCTIONS LIST OF ENCLOSURES Enclosure Cl Clad Damage Characteristics of NRC Categories of Fuel, Damage Enclosure C2 Record of Temperature, Pressure and Damage Estimate Enclosure C3 Percent of Fuel Rods vith Ruptured Clad 62 as a Function of Maximum Core Exit Thermocouple Temperature

I ~ ~ ~

4

(62 of 80)

This section provides the methodology for use under post accident plant conditions to determine the number of fuel rods with ruptured clad.. It provides an estimate of damage up to about the time when the peak core temperature reaches about 2300 F. At that time most of the" rods probably have ruptured clad but little other structural degradation has occurred. Therefore this procedure applies to the relatively less severe accidents although it may be used for other accidents to confirm that damage exceeds this minimum amount. The resulting observation of core damage is described by. categories l through 4 of the seven NRC categories in Enclosure Cl.

2.0 REFERENCES

2 ~ l Development of the Comprehensive Procedure Guidelines for Core Damage Assessment, CE Owners Group Task 467, May, 19S3.

F 2 Inadequate Core Cooling, St. Lucie Unit 2, Final Safety Analysis Report, Appendix l.9.B.

2 ' Generic Thermal-Hydraulic Functional Design Objectives for Inadequate Core Cooling Instrumentation, CE-NPSD-l99, prepared for the CE Owners Group.

3.D DEFINITIONS:

the fuel rod clad at least sufficient to release the internal gas pressure. Rupture may be preceded by ballooning of the clad if the internal gas pressure exceeds the external coolant pressure during an accident, and the temperature is higher than normal.

4.0 PRECAUTIONS AVD LIMITATIONS:

4.l The assessment of core damage obtained by using this

'n method is only an estimate. The techniques employed this section are only accurate to locate the core condition within the first four of the 7 categories of core damage described in Enclosure Cl.: 7he methodology is based on core exit temperature data.

Other plant indications may be available which can improve upon estimation of core damage. These include radiological sample characteristics, the total quantity of hydrogen released from zirconium degradation and containment radiation monitors.

Whenever possible these additional indicators should be factored into the assessment.

(63 of 80) 4~2 The assessment of damage provided by this procedure extends up to the time of clad rupture on most of the fuel rods. This time occurs early in very severe core uncovery accidents. +fore severe coze damage cannot be

'uantified by this procedure.

4' The relationship between the core exit thermocouple-temperature and the clad temperature varies- with the core uncovery senario. This procedure applies to slow core uncovery by boiloff of the coolant. Por other more rapid uncovery scenarios this procedure could yield a very,low estimate of the number of ruptured rods. In general, for core uncovery at pressures below about 1200 psia there is high confidence that at least the predicted estimate of rods are actually ruptured.

5.0 IHSTUCTZOMS

-5 ~ 1 Obtain the following from the instrument recordings:

?

5.1.1 From the recording of maximum cor exit thermocouple temperature as a function of time, obtain and record on Enclosure C2 the maximum temperature and the time it, occurs. As many thermocouples as possible should be used, in, this way equipment malfunction may be detected if a thermocouple reads 'greater than 1650 P or varies considerably from its neighboring thermocouples.

5 ~ 1 ~ 2 Prom the recording of reac tor coolant system pressure as a function of time, obtain and record on Enclosure C2 the pressure during the period of maximum thermocouple temperature.

5.2 Sele"t'he curve on Enclosure C3 which is labeled with a pressure'pproximately equal to or-greater than the pressure in Step 5.1.2. Enter -the abscissa't the maximum temperature from Step 5. 1. 1 and read on the ordinate the percent of the fuel rods which have ruptured clad. Record on Enclosure C2.

5.3 The result in 5.2 is probably a lower limit stimate of damage. Some judgement on the bias is available as follows.

,5.3.1 This procedure applies most directly for relatively slow core uncovery with a maximum temperature below the rapid oxidation temperatures at about 1800 P and above. A smooth core exit thermocouple recording and an uncovery duration of 20 minutes or longer are indicators for a good prediction of clad ruptures.

~ 4 (64 of 80) 5-3 ~ 2 If the about pressure in 5.1.2 drops to less than l,QO psia within less than about two minutes of accident initiation, a large break is. indicated. This causes undetected core heatup followed by flashing during refill.

Depending on the rate of refill, the thermo-couple temperature may rise rapidly then quench when the core is'recovered. This procedure could yield a very low estimate for the percent of rods ruptured.

5 ' ' If the pressure in Step 5.1.2 is above about 1650 psia, it could exceed the rod internal gas pressure dependirig on rod burnup, causing clad collapse onto the fuel pellet instead of outward clad ballooning. The clad rupture criteria ar'e less well defined for such conditions, but at temperatures above 1800 F where the highest pressure curve applies on Enclosure C3, clad fai.lure sufficient to release fission gas is likely and this procedure may be used to obtain estimates of damage.

5.4 Core Dama e Assessment Use the percent of rods ruptured from Step 5.2 and the clad damage characteristics of Enclosure Cl to determine the NRC category of cladding failure. Thi,s procedure yields damage estimates in categories 2, 3 or 4 (Enclosure Cl).

I ~ ~~

ElKiOSNE CX CUS DhNhGE ONRhCTERlSTlCS OF NC CATEGORlES Of fllEL OHMAGE Characteristic of of NC Category Fuel Daai ka 'f Temperature NcctQhi $ M of Daai a t ~a Ncasuremet Percent

~lh

1. No Fuel DNIage ~150 Less TINh 1

2. initial Cladding Rupture Due to Naxiaa Core <<1550'F* Less Thah ]O failure Gas Gap . Exit S

3. late ldiatc l200-1800 Ovcrpressuriration. '%eraxouple <<1700'F>> 10 to 50 Cladding failure Temperature I. Na)or Cladding <<2300'F Failure <<25 Greater Thah 50 Oxidatioh

5. initial Fuel Pellet Loss of Structural haeat of Equ)valent Core Less Tbah l0 Overheatiag integrity Duc to Hydrogen Gas Oxidation Fuel Clad Produced <<35 C lllteIdiate 1800-3350 Oxidation '(Equivalent to <<185 Fuel Pcl lct %Oxidation Overheating of Core)

7. Na)or Fuel Pellet <<65% Greater TlNn 50 I

• Depends on Reactor Pressure and Fuel burnup Valves G)ven,for Pressure <<l200,psla and burnup >0.

T~

(66 of 80)

CmP GE ESraeZZ Step 5.1 Record the follcwing data Mmcimm Core Exit Therrmcouple Temperature*

(see Instruction 5.l, in the text for guidelines)

Time of Maxim'arperature Reactor Coolant System Pressure at Above Tina psia Step 5.2 Frcxn Enclosure C3, at maxirmm thezaacouple teaperature and at appropriate pressure Read percent of ruptured rods Step 5.3 Ccament on (see p~le bias of results paragraph 5.3 in text).

in 5.2 Step 5.4 NRC category of cladding failure fran Enclosure Cl P

• As many theamcouple readings as possible should be recorded. In this way equiprrent malfunction may'be detected its ifneighkaring a therrmcouple reads greater than 1550 P or varies considerably fran therrmcouples.

MaxirrLrm Core Exit Thexnacmple Temperature Performed By: Date:

~ ~ re (67 of 80)

ENCLOSURE C3 PERCEST OP FUEL RODS WITH RUPTURED CLAD VS MAZIMUM CORE EZIT TEERMOCOUPLE TEMPERATURE WHEN THE PRESSURE USE CURVE LABELED IN STEP 5 1 2 IS WITH TEMPERATURE 100 PSIA 1200 F

<1200 PSIA 1500 F

<1650 PSIA 1800'F 100 80 1200 F 60 CLAD RUPTURE T&fPERATURE 4

0 1500 F

.1800 F 20 l5 1200 1400 1600 1800 2000 2200 MAZIMUM CORE EXIT ~98lOCOUPLE TEMPERATURE

~ ~

(68 of 80)

SECTION D ASSESSiIENT OF CORE DAMAGE USING RADIATION DOSE RATES

~ ~

A ~

(69 of 80)

TABLE OF CONTENTS.

PAGE 1 '0 PURPOSE 65 2' REFERENCES 65 3 ' DEFINITIONS 65 4.0 PRECAUTIONS AND LIMITATIONS 5' INSTRUCTIONS 5.1 Record of Plant Condition 67 5' Plant Power Correction 6S 5.3 Decay Correction 69 5.4 ,Assessment of Core Damage 69 LIST OF ENCLOSURES Enclosure Dl Radiological Characteristics of NRC 71 Categories of Fuel Damage Enclosure '2 Conta inment H igh Range Radiation 72 Monitor (Core Damage Assessment)

Worksheet Enclosure D3 Post Accident Dose Rate Inside the 73 Containment Building (Containment High Range Radiation Monitor)

I' 1

~ ~

(70 .3f 80) pURPOSE:

This section provides the methodology for use under post accident plaat conditions to determine the type and degree core damage which may have occurred by using radiation

'f dose rates measured inside the containment building using the containment high range radiation monitor. The radiation dose rate is related to the quantitative release

- of fission products from the core expressed as the percent of the source inventory at the time of the accident. The resulting observation of core damage is described by one or more of the seven categories of core damage in Enclosure DL.

2.0 REFEREMCESi 2.1 Development of the comprehensive procedure guideline for core damage assessment, CE Owners Group Task 467, July 1983.

3 ' DEFZ3ITXONS 3.1 damage is defined as a progressive failure of the material boundary to prevent the release of radioactive fission products into the reactor coolant starting with a penetration in the zircaloy cladding.

The type of fuel damage as determined by this methodology is reported in terms of three major categories which are:. no damage, cladding failure, and fuel overheat. The categories are characterized by the resuLting radiation dose rate inside the containment building. The degree of core damage is determined by making a comparison between dose rates measured following an accident and analytically determined values of the realistic or best estimate dose rates that would correspond to the specific categories of core damage. The degree of core damage as determined by this section is report d in terms of three levels which are:. initial; intermediate;. and major. This results in a, total of seven possible categories as characterized in Enclosure Dl.

3 ' souroe Invent~or:. The source inventory's the total quantity of fission products expressed in curies of each isotope present in either source; the fuel pellets or the fuel rod gas gap.

~ ~

(71 of 80) 4.9 PRECAUTIONS AND LIi4ITATIOQS-

4. l The. assessment of core damage obtained by using the methodology in this section is only an estimate. The techniques employed in this section are only accurate to locate the core condition within one or more'f the 7 categories of core damage described in Enclosure Dl. The procedure is based on radiation dose rate.

Other plant indications may be available which can improve upon the estimation of core damage. These include sample radiological analysis,'ncore temperature indicators, and the total quantity of hydrogen released from zirconium degradation.

Whenever possible these additional indicators should be factored into the assessment.

4.2 This section relies upon radiation dose rate measurements taken from the higher of two high range radiation monitors located inside the containment building to determine the total quantity of fission products released from the core and therefore available for release to the environment. The amount of fission products present at the location of the monitors may be changing rapidly due to transient plant conditions. Therefore, multiple measurements should .be obtained within a minimum time period and when possible under stabilized plant conditions.

Samples obtained during rapidly changing plant

.conditions should not be weighed heavily into the assessment of core damage.

4' A number of factors influence the reliability of the measured radiation dose rates upon which this procedure is based. Reliability is influenced by the ability to obtain representative measurements due to incomplete mixing of the measured media, and equipment limitations. Additionally the procedure relies, upon analytically determined values of the best'stimate dose. rates that, are anticipated to correspond to the specific categories'of core damage. These analytical values are based upon assumptions made about the identity and relative proportions of the fission products released from the core and their transport within the containment building Therefore, the

~

method is only accurate to within the validity of the assumptions.

4.4 The methodology in this section is limited to the upper'ound condition of- fission product release from the core due'o fuel overheat. Simultaneous with fuel overheat, there may be localized fuel pellet melting within the core'. The transport of the non-volatile fission products released due to melting is not known. The dose rates measured under conditions of

(72 of 80) fuel pellet melting are anticipated to exceed those, shown in Enclosure D3, for major fuel overheat.

However, -this procedure does not attempt to identify the extant of any potential fuel melting.

4.5 This section is limited to the interpretation of the dose rate measurement resulting from a mix of fission products. The methodology cannot accurately distinguish between the conditions of fuel cladding failure and fuel'overheat when the resulting dose rates are the same. The methodology does provide an upper limit estimate of the progressive core damage.

Concurrent conditions of cladding failure and overheat should be anticipated due to the radial distribution of heat generation within the core. Distinction between the type of core damage requires the identification of the characteristic fission products. The procedure for core damage assessment using radiological analysis of fluid samples is required to explicitly distinguish between the categories, This methodology is limited in applicability to those conditions in which the fission product inventory in the core has had sufficient time to reach equilibrium. Equilibrium fission product inventory is a function of reactor power and burnup. Based upon the fission products of concern equilibrium conditions are achieved after thirty days of operation at constant power. Constant power is considered to include changes of no greater than + LO percent. The methodology may be used following non-constant periods of operation by using engineering judgement to s '.ect the most representative power level during the p='=iod. This method may also be used if the reactor has produced power for less than thirty days, how'ever, the resulting assessment of core damage would be an underprediction of the actual conditions.

5.0 INSTRUCTIONS 5 ' Record the following plant indications.

5. l. 1 Containment Building:

Radiation Dose Rate Rads/hr'.

Time of Measurement Date Time

(73 of 80) 5..1 2

~ Prior 30 days power history:

Power, Percent Duration, Da s 5.1,3 Time of reactor trip Date T ime Record these values in Enclosure D2.

5.2 Plant Power Correction The measured radiation dose rate inside the containment building is to be corrected for the plant power history. A correction factor is used to adjust the measured dose rate to the corresponding value had the plant been operating at 100 percent power.

• 5.2 ~ 1 To correct the radiation dose rate for the case in which plant power level has remained constant for a period-greater than 30 days a simple ratio of the power may be employed. The reactor power is considered to be constant if it has not changed by + 10 percent within the last thirty days prior to the reactor trip.

5~2~2 To =correct the radiation dose rate for the case in which reactor power level has not remained constant during the 30 days prior to the reactor shutdown engineering, judgement is used to determine the most representative power level. The following guidelines should be considered in the determination.

5~ 2~ 2~1 The average power during the 30 day time period is not necessarily the most .representative value for correction to equilibrium conditions.

5. 2 ~ 2-2 The last power levels at which the reactor operated should weigh more heavily in the judgement than the earlier levels.

~ ~

i

Ck (74 of 80) 5 ~ 2.2 ~ 3 Continued operation for, an extended period should weigh more heavily in the judgement than brief transient 4A levels.

5 ~ 2'3 In the case in which reactor has produce'd power for less than 30 days the procedure may be employed. However, the es timat e of. core damage obtained under this condition may be an under prediction of the actual condition.

5.2.4 The following equation is applied to determine the radiation dose rate corresponding to equilibrium full power source inventory conditions'quilibrium Measured 100 Dose'ate (Rad/Hr) (Rad/Hr) m The reactor 'power level and the resulting dose rate correction factor used above will be the same for all subsequent measurement of the dose rate. Record these values to reduce the work required to .evaluate the subsequent meas'urements.

5.3 The decay correction for the radiation dose rate the determination of the time duration 'requires between the reactor trip and the measurement of the dose rate. This is done simply using the time of reactor shutdown recorded in Section 5.lan 5.4 The conclusion on the extent of core damage is made using the equilibrium dose rate, the duration of '.

reactor shutdown, and the analytically determined dose rates provided in Enclosure D3. The equilibrium dose'rate is plotted on Enclosure D3 as a function of time following reactor shutdown. Engineering judgement is used to determine which'ategory of, core damage shown on Enclosure D3 is most representative of the particular value that has been plotted. The following criteria should be considered in the determination.

5.4.1 Dose rate measurements may have been recorded during peri.ods of transient conditions within the plant. Measurements made during stable plant conditions should weigh more heavily in the assessment of core damage.

QS 4 ,l

(75 .of 80)

Dose rates significantly above the lower bound for the category of majorfuel overheat may indicate concurrent fuel pellet melting. The

-methodology in this section may not, be emmployed to estimate the degree of fuel pel.let melting.

Dose rates within any category of fuel overheating may be anticipated to include concurrent fuel cladding failure. The methodology in this section may not be used to distinguish the relative contributions of the two categories to the total dose rate. The methodology does give the estimate of the highest category of damage.

Dose rates corresponding to the two categories of major cladding failure and initial fuel overheat axe observed to overlap on Enclosure D3. The evaluation of other plant paramet rs may be required to distinguish between them.

However, concurrent conditions may be ant'icipated.

4r 4 l

Knclosure P1 ~

~

Radiol ic Characteristics of NRC Cate ries of Fuel Oaaa NRC Category of Hcchanisa of Source of Percent of Source Inventory Distribution of Fission Fuc Oaaa Relcasc Froa Core Release Released to Containment Products ia Contaiint No Fuel Damage . Halogen Spiking Gas Gap Less than 1 Airborne Traap Uraniua initial Cladding Gas Gap Less than 10 Airborne Failure 30 lntermdiatc Clad burst and Gas Gap 10 to 50 Airborne Cladding Failure Gas Gap Diffusion Release

4. Na)or Cladding Gas Gap Greater than 50 Airborne Failure

5. initial Fuel Pellet Fuel Pellet less than 10 Overheating Airborne:

Grain Boundary 100K Noble Gas Diffusion 25% Halogen

6. lntermdi ate Fuel Fuel Pellet 10 to 50 Pellet Overheating Plated Out 25% Halogen 1$ Solids

7. Na)or Fuel Pellet Diffusjonal Release Fuel Pellet Greater than 50 Pellet Overheating . Froa ij02 Grains

s r 0

(77 af 80)

ENCLOSURE D2 CONTAINMENT HIGH RANGE RADIATION MONITOR (CORE DAK%GE ASSESSMENT) WORKSHEET Highest Radiation Dose Rate (CHRRM) Rads/hr Time of Measurement Date:. Time:

Prior 30 Days Power History:

Power, Percent Duration, Da s Time of Reactor Trip Date: Time:

Equilibrium Dose Rate (Rad/hr)

M easured Dose Rate (Rad/hr) x 100 Reactor Power Level Refer to Enclosure D3 to obtain category of core damage. See step 5.4.

Performed by:

Date:

0' (78 of 80)

ENCLOSURE D3 ANALYSIS POR POST ACCIDENT DOSB RATS INSIDE CONTAI%69iT (CONThINMENT HIGH RANGE RADIATION MONITOR)

~ ~

I i I:

I

( ~

I I

~ I l

~ ~

105 ~ ~ 1.

E~ Og, Ey

+

I

! I Eg

l ~ Og, (

'l

}

104

~ ~ I C' i I

I

~

I i '

l l I

I g }, l

~ ~

~ ~

I 103 I ~ I C' '

~ I I I I ~ ~ ~ I

'I I ~

}

10 100 1000 TINE POST REACTOR TRIPg HOURS

SECTIONAL E SUMMA,RY OP RESULTS

(80 of 80)

SUMMARY

OP RESULTS:

Section A, or'adiological analysis of samples, is the most complete and possibly the most accurate of 'he methods used to assess the degree of core damage. This section of the methodology provides the instructions required to determine core damage up to the major fuel melt category identified in the NRC guidelines for core damage assessment.

Other indicators which are described in Sections B, C and D of these methodology are limited to the fuel overheat category o' core damage. Section B,'hich uses the hydrogen content of both the reactor coolant and containment-bu'ilding atmosphere for an indication of fuel cladding oxidation is most applicable within the fuel overheat. category. Section C, which uses the

.information from reactor coolant core exit temperatures is most applicable within the cladding failure'ategory. While Section D, use information from area dose rates within containment and..;.

it is most applicable within both cladding failure and fuel overheat categories of damage.

It is important to note that core damage is not anticipated to take place uniformly. Therefore a combination of one or more of the categories of fuel damage will most likely exist simultaneously. The results obtained from. the should be compared with the results of the evaluation radionuclide'nalysis of other available indicators for a comprehensive assessment.

If the results are in agreement, the core damage assessment is complete. If the results are not in agreement, a recheck 'of both analyses may. be performed or certain indications may be discounted or weighted more heavily based on engineering judgement.

.jO V

+0 Q rQ-

.!+ 71983 Page 1 of 1 PC7 ( 0( QQ~P~

( r November 198

'MERGENCY<<PLAN. IHPLEMEN'ZXNG PROCEDURE" INDEX '53,,;~';-

.Applicable to'oth'Units,k, and'2 4 ~

REV. TZILE.

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P 4 LI 4 1 24 te Em rgericj Oqpmization Rostei;-"- ~

41 4

Faaijjjfcies

'100024E

\ J Wiooozsz '8  ;..'~ Ere es' 100026E Cifteriya For,: .of Evacuations >;;

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