ML20077H244

Rev 1 to Procedure NE-5-301, Technical Procedure - Prediction of Core Damage

ML20077H244
Person / Time
Site:	Waterford
Issue date:	07/20/1983
From:	LOUISIANA POWER & LIGHT CO.
To:
Shared Package
ML20077H235	List: ML20077H244 W3P83-2583, Forwards Rev 1 to Procedure NE-5-301, Technical Procedure - Prediction of Core Damage, Per TMI Action Item II.B.3
References
TASK-2.B.3, TASK-TM NE-5-301, NUDOCS 8308090695
Download: ML20077H244 (63)

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. , WATERFORD 3 SES PLANT OPERATING MANUAL LOUISIANA POWER & LIGHT NSY NE-5-301 POM VOLUME 14 REVISION 1 M g POM SECTION 5 APPROVAL DATE

EFFECTIVE DATE:

LPal. W-3 RECOROS UNCONTROLLED CC'Y DO NOT IN W WY S AcrT'/ cui .a TP1 Tvr- .3, MA!NTEN(iww i yri vW. <d. ..c , e. v n Y -

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NUCLEAR RECORDS l JUL %6 1983

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WATERFORD 3 SES PLANT OPERATING MANUAL CHANGE / REVISION / DELETION REQUEST '

Procedure No. Ib b0 Title bENc77dN M bfE AA E Effective Date . (if different from approval date)

Camnieta 1. B. or C A. Change No. AllA g

B. Revision No. I I 1 -10 C. Deletion (A RFAMON FOR CH ANCE 'REYTMTON . OR DELKTION .

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• REQUIRED SIGNATURES Originator- M tte Y ocb Date b ~ I b 8'3 Technical Review N dWA Date (e- 6-?3 SAFETT EVALUATION  !

Does this change, revision, or deletion: YES NO  ;

m 1. Change the facility as described in the FSAR7 /

J 2. Change the procedures as described in the FSAR7- / l

3. Conduct tests / experiments not described in the FSAR7 b l

4. Create a condition or conduct an operation which ex-  !

coeds, or could result in exceeding, the limits in Technical Specifications?

If the answer to any of the above is yes, complete and at-tach a 10 CFR 50.59 Safety Evaluation checklist.

Safety Evaluation b OY1- Date (e- T- V 3 Group /Dep't. Head Review Mb ##Nb Date 4/39M'T Temporary Approva18 Date (NCS)

Temporary Approvale Date QC Review AM A W - / Date '/- v - 3 7 PORC Review MN Date Y/ Y% Meeting No. W - M' PlantManager-NucIsarAp oval N/ F Date AC

' Temporary approval must be followed by Plant Manager-Nuclear approval

.' within 14 days.

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UMT-1-003 Revision 6' Attachment 6.9 (1 of 1) 3G

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O Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 I

TABLE OF CONTENTS 1.0 PURPOSE l 2.0 REFEREN CES 3.0 DEFINITIONS 4.0 RES PONSIBILITIES 5.0 PREREQUISITES 6.0 PRECAUTIONS AND LIMITATIONS 7.0 INITIAL CONDITIONS 8.0 MATERIAL AND TEST EQUIPMENT 9 .0 ACCEPTANCE CRITERIA 10.0 PROCEDURE 10.1 Plant Conditions 10.2 Core Damage Based on Radioisotopic Analysis 10.3 Core Damage Based on Hydrogen Analysis 10.4 Core Damage Based on Radiation Dose Rate in Containment 10.5 Core Damage Based on the Core Exit Thermocouples 11.0 SETPOINTS 12.0 ATTACHMENTS 12.1 Radiological Characteristics of NRC Categories of Fuel Damage (Radioisotopic) (1 page) 12.2 Sample Locations Appropriate for Core Damage Assessment (1 page) 12.3 Record of Sample Specific Activity (1 page) 12.4 Density Correction Factor for Reactor Coolant Temperature (1 page) 12.5 Record of Sample Temperature and Pressure Correction (1 page) 12.6 Record of Sample Decay Correction (1 page) 1

'. Technical Procedure NE-5-301 Prediction of Core Damage Revision 1  ;

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12.7 -Record of Fission Product Release Source Identification (1  ;

page) 12.8 Record of Release Quantity (1 page) 12.9 Containment Building Water Level vs. Volume (LATER) l 12.10 Record of Steady-State Power Correction (1 page) l 12.11 Record of Transient Power Correction (2 pages) '

12.12 Record of Percent Release (1 page) 12.13 Clad Damage Characteristics of NRC Categories of Fuel Damage (1 page) 12.14 Core Uncovery Conditions (1 page) 12.15 Sampling Conditions and Measured Hydrogen (1 page) 12.16 Calculation Worksheet for Hydrogen Generated in Containment (1 page) 12.17 Hydrogen Production Rate in Containment as a Function of Temperature (1 page) 12.18 Calculation Worksheet for Hydrogen Generated by Radiolysis (1 page) 12.19 Specific Radiolytic Hydrogen Production vs. Time (1 page) 12.20 Core Damage Assessment from Hydrogen Measurement (1 page) 12.21 Percent of Fuel Rods with Ruptured Clad vs. Core Clad Ox-idation (1 page) 12.22 Percent of the Fuel Rods with Oxidation Embrittlement vs.

Total Core Oxidation (1 page) 12.23 Estimation of the Amount of Hydrogen in a Reactor Vessel Head Void (4 pages) 12.24 Radiological Characteristics of NRC Categories of Fuel Damage (Dose Rate) (1 page) 12.25 Post Accident Dose Rate Inside the Containment Building (1 page) 12.26 Record of Temperature, Pressure and Damage Estimate (1 page) 12.27 Percent of Fuel Rods with Ruptured Clad vs. Maximum Core Exit Thermocouple Temperature (1 page) 13.0 COMMITMENTS AND REFERENCES 2

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. Technical Procedure. NE-5-301 Prediction of Core Damage . Revision 1-LIST OF EFFECTIVE PAGES Title Revision 1 1-60 Revision.1 J

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1 Technical-Proceduro NE-5-301 Prediction of-Core Damage Revision 1 1.0 PU R POS E 1.1 This procedure is to be followed under postaccident plant conditions to determine the type and degree of reactor core damage which may have occurred, using:

1.1.1 Fission product isotopes measured in samples obtained from the Post Accident Sampling System (PASS) 1.1.2 Hydrogen measured in samples obtained with the PASS 1.1.3 Radiation dose rates measured inside the containment building using the wide range radiation monitor 1.1.4 The core exit thermocouples, to determine the number of fuel rods with ruptured clad 1.2 The resulting estimate of core damage is described by one or more of the 10 categories of core damase in Attachment 12.1.

2.0 REFERENSES .

2.1 Development of the Comprehensive Procedure Guidelines for Core Damage Assessment, C.E. Owners Group Task 467, May 1983, transmitted by C.E. letter SE-83-094 dated May 13, 1983 2.2 NUREG-0737, Item II.B.3 2.3 Regulatory Guide 1.97 2.4 Post Accident Sampling System Procedures 2.4.1 CE-3-900, Operation of the Post Accident Sampling System 2.4.2 CE-3-903, Post Accident Gamma Spectroscopic Analysis 2.4.3 CE-3-904, Post Accident Analysis of Dissolved Hydrogen in the Reactor Coolant 2.4.4 CE-3-901, Post Accident Sampling of Containment Atmosphere 4

'a Tcchnical Procedure NE-5-301 Prediction of Core Damage Revision 1 2.5 (LATER), Wide Range Containment Radiation Dose Rate Monitor Operating Procedure 2.6 (LATER), Inadequate Core Cooling Instrumentation, Core Exit Thermocouple Operating Procedure 2.7 EP-1-001, Recognition and Classification of Emergency Conditions 3.0 DEFINITIQHE 3.1 Fuel Damage: For the purpose of this procedure, fuel 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 zircalloy cladding. The type of fuel damage, as determined by 'this procedure, is reported in terms of four major categories, which are: 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 is available for immediate release to the environment. The degree of fuel damage as determined by this procedure is reported in terms of three levels, which are: initial, intermediate, and major. This results in a total of ten possible categories, as characterized in Attachment 12.1.

3.2 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.

3.3 Clad Rupture: The fuel clad ruptures when the 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 5

'. Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 some fragments of fuel pellets but does not otherwise destroy the structure of the fuel assembly.

3.4 Clad Embrittlement: At temperatures above the rupture temperature, significant oxidation of the clad occurs. If 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 Essembly is destroyed and substantial fuel pellet fragments are released to the coolant.

3.5 Uncovery

A condition where water level in the reactor vessel is below the top of the active fuel.

4.0 RES PONSIBILITIES 4 .1 EMERGENCY COORDINATOR The Emergency Coordinator is responsible for directing the performance of this procedure during accident (or other) conditions which indicate the possibility of core damage.

4 .2 CHEMISTRY Chemistry personnel are responsible for operating the Post Accident Sample System to obtain the results necessary for performance of this procedure.

4.3 NUCLEAR ENGINEERING i-Nuclear Engineering is responsible for maintaining this procedure and for aiding in evaluating the results of this procedure. l 4.4 SHIFT TECHNICAL ADVISOR (STA)  !

i The STA is responsible for aiding the Emergency Coordinator in the  ;

performance of this procedure.

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Technien1 Procedure NE-5-301 Prediction of Core Damage Revision 1 5.0 EEIBIQ.UISITES 5 .1 The Post Accident Sampling System shall be operable with the capability to obtain and analyze the identity and concentration of fission product isotopes in fluid samples which have the potential to be highly radioactive. The system should meet the requirements of NUREG-0737, Item II. B.3, Ref erence 2.2.

5.2 The Post Accident Sampling System shall be operable with the capability to obtain and analyze the concentration of hydrogen in fluid samples which have the potential to be highly radioactive.

The system should meet the requirements of NUREG-0737 Item II.B.3, Reference 2.2.

5.3 The Wide Range Radiation Dose Rate Monitor System shall be operable with the capability to measure the area dose rates inside the Containment Building resulting from fission products dispersed in the building atmosphere and plated out on building surf aces. The system should meet the requirements of Regulatory Guide 1 97, Reference 2.3.

5.4 An Inadequate Core Cooling Instrumentation System shall be operable which includes core exit thermoconples and which can select and permanently record the highest thermocouple temperature for convenient later inspection.

5.5 The Reactor Vessel Vent System shall-be operable.

5.6 The Reactor Vessel Level Indication System shall be operable.

l 6.0 PRECAUTIONS AND LIMITATlQN3 +

l 6 .1 GENERAL

6.1.1 This procedure should not be performed until such time as the plant has been returned to a stable condition. ,

6.1.2 The assessment of core damage obtained by using this procedure is only an estimate. The techniques employed in this procedure are 7

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Tcchnical Procedure NE-5-301 Prediction of. Core Damage Revision 1 1

accurate only to locate the core condition within one or more of the 10 categories of core damage described in Attachment 12.1.

6.2 The following precautions apply to section 10.2:

6.2.1 Section 10.2 relies upon samples taken from multiple locations inside the Containment Building to determine the total quantity of fission products 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 required 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 asse'ssment of core damage.

6.2.2 A number of factors influence the reliability of the chemistry samples upon which section 10.2 is based. Reliability is influenced by the ability to obtain representative samples due to incomplete mixing of the fluids, equipment limitations, and lack of operator familiarity with rarely used analytical procedures.

The accuracy achieved in the radiological analyses is 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 required 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 distribu-tion of the sample fluid. Cooling or reactions may take place in ,

the long sample lines. Therefore, the results obtained may not be representative of plant conditions. To minimize these effects, l multiple samples should be obtained over an extended time period j from each location.

6.3 The following precautions apply to section 10.3: i 1

6.3.1 Section 10.3 relies upon hydrogen samples taken from the containment atmosphere and the Reactor Coolant System Hot Leg.

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1 Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 Those samples may contain a mixture of hydrogen generated within i the core by clad oxidation and also hydrogen from radiolytic  !

dissociation of water and oxidation of aluminum and zine 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.

6.3.2 Section 10.3 yields estimates of the percentages of fuel rods with ruptured clad and embrittled clad. Simultaneous with embrittling

of the clad, clad melting and pellet overheating may occur. Sec-tion 10.3 provides an estimate of only the percentage of rods which have progressed to at least clad rupture or clad embrit-tlement and does not attempt to predict the physical configuration of those rods which have progressed beyond local clad fragmenta-tion.

6.3.3 Depending on the accident scenario, a given total amount of hydrogen produced by oxidation of fuel clad can represent varying local amounts and distributions of clad damage. Section 10.3 at-tempts to bias the damage estimates such that the results represent lower limit estimates of clad damage. Actual damage could be greater, depending on the accident scenario.

4 6.3 4 Section 10.3 is applicable under conditions 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 accidents. However, if the hydrogen samples are taken under conditions in which measurable void does exits, a guideline for analysis is provided in Attachment 12.23 to estimate the contribution of that source to be added to the total hydrogen measured.

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Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 6.4 The following precautions apply to section 10.4:

6.4.1 Section 10.4 relies upon radiation dose rate measurements taken from one or more monitors located inside the Containment Building l-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.

6.4.2 A number of factors influence the reliability of the measured radiation dose rates upon which section 10.4 is based.

Reliability is influenced by the ability to obtain representative measurements due to incomplete mixing of the measured media, equipment limitations, and lack of operator familiarity with rarely used procedures. Additionally, section 10.4 relies upon analytically determined values of the best estimate 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, section 10.4 is accurate only to within the validity of the assumptions.

6.4.3 Section 10.4 is limited to the upper bound condition of fission product release from the core due to fuel overheat. Simultaneous with fuel overheat, there may be localized fuel pellet melting within the core. The transport of the nonvolatile fission products released due to melting is not known. The dose rates measured under conditions of fuel pellet melting are anticipated to exceed those shown in Attachment 12.2 for major fuel overheat.

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Tochnical Procedure NE-5-301 Prediction of Core Damage Revision 1 However, this procedure does not attempt to identify the extent of any potential fuel melting.  ;

6.4.4 Section 10.4 -is limited to the interpretation of the dose rate measurement resulting from a mix of fission products. Section 10.4 cannot accurately distinguish between the conditions of fuel cladding failure and fuel overheat when the resulting dose rates are the same. Section-10.4 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 types 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.

6.5 The following precautions apply to section 10.5:

6.5.1 The assessment of damage provided by section 10.5 extends up to 1

the time of clad rupture on most of the fuel rods. This time oc-curs early in very severe core uncovery accidents. More severe core damage cannot be quantified by this procedure.

6.5.2 The relationship between the core exit thermocouple temperature l  !

and the clad temperature varies with the core uncovery scenario. l Section 10.5 applies to slow core uncovery by boiloff of the {  !

coolant. For other more rapid uncovery scenarios, section 10.5 ll could yield.a very low estimate of the number of ruptured rods.

In general, for core uncovery at pressures below about 1200 psia, l there is high confidence that at least the predicted estimate of rods is actually ruptured. l )

i 6.6 This procedure 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 30 11 I

', Technical Procedure NE-5-301 Prediction.of Core Damage Revision 1 i

days of operation at constant power. Constant power is considered  !

to include changes of no greater than 2 10 percent. The procedure may be used following nonconstant periods of operation by using engineering judgement to select the most representative power level during the period. The procedure may also be used if' the reactor has produced power for less than 30 days; however, the resulting assessment of core damage would be an underprediction of the actual conditions.

70 INITTAL .GQEITIOM This procedure is to be employed for analysis of radiochemistry, hydrogen, dose rate in containment, and core exit thermocouple data when it is determined that a plant accident with the potential for core damage has occurred. The following is a list of plant symptoms to assist in this determination. One or more of these symptoms may exist at or before the time the sample is obtained. Under these conditions, sampling should be performed using the Post Accident Sampling System.

7 .1 HIGH alarm on the Containment Radiation Monitor 7.2 HIGH alarm on the CVCS Letdown Radiation Monitor 7.3 HIGH alarm on the Main Condenser Air Ejector Exhaust Radiation Monitor 74 Pressurizer level low 7.5 Safety Injection System may have automatically actuated 7.6 Possible high quench tank level, temperature, or pressure 7.7 Possible noice ~ indicative of a high energy line break 7.8 Decrease in Volume Control Tank level 79 Standby -charging pumps energized 7.10 Unbalanced charging and letdown flow {

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. Toohnical Procedure NE-5-301 Prediction of Core Damage Revision 1 7.11 Reactor Coolant System subcooling low or zero 8.0 MATERIAL AND TEST EQUIPMERJ NONE 9 .0 ACCE PTAHfE CRITERIA NONE 10.0 PROCEDURE 10.1 PLANT CONDITIONS Record the following plant conditions. Because of transient conditions, the values should be recorded as closely as possible to the time at which the radiological samples are obtained from the Post Accident Sampling System.

10.1.1 Reactor Coolant System:

Pressure psia Temperature *F Reactor Vessel Level  %

Pressurizer Level  %

Core Exit Thermocouple Temperature _ F Core Exit Thermocouple Saturation Margin F subcooled 10.1.2 Containment Building:

I l Atmosphere Pressure psia l l

.~ 1 Atmosphere Temperature F l Sump Level  %

Radiation Dose Rate ___ rads /hr l 1

Time of Measurement Date __ Time l l

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Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 10.1.3 Prior 30-day power history: Power. Persgnt Duration _D3ys 10.1.4 Time of reactor shutdown Date Time 10.2 CORE DAMAGE BASED ON RADI0 ISOTOPIC AN ALYSIS 10.2.1 Select the most appropriate sample locations required for core damage assessment, using the guidelines provided in Attachment 12.2.

10.2.2 Obtain and analyze the selected samples for fission product specific activity using the procedures for Post Accident Sample System operation described in Reference 2.4.1. Record the required sample data for each selected sample. Attachment 12.3 is provided as a worksheet. All of the isotopes listed in the enclosure may not be observed in the sample.

NOTE This step is required only if it is not included in the procedures for Post Accident Sample System operation in Reference 2.4.1.

10.2.3 Correct the measured sample specific activity to standard temperature and pressure.

10.2.3.1 Reactor coolant liquid samples are corrected for RCS temperature and pressure using the factor for water density .

provided in Attachment 12.4. The correction factor obtained  !

I from the attachment is multiplied by the measured value to i obtain the density corrected value.

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i Prediction of Core Damage Revision 1 i i

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1 10.2.3.2 Containment Building sump samples do not require correction for  ;

i temperature and pressure within the accuracy of this procedure.

10.2.3.3 Containment Building atmosphere gas samples are corrected using the following equation.

Specific Activity (STP) =

Specific Activity x P2 x Ti + 460 Pl + Pg, Tg + 460 where:

Ty , P, = Measured sample temperature and pressure recorded in step 10.2.2 Tz, Pg = Standard temperature, 32 F and standard pressure 14.7 psia 10.2.3.4 Attachment 12.5 is provided as a worksheet.

-10.2.4 Correct the sample specific activity at STP for decay back to the time of reactor shutdown which is recorded in step 10.1.4 using-the following equation. Attachment 12.6 is provided as a worksheet.

o" ,- t where:

A = the specific activity of the sample corrected back to the time of reactor shutdown, pC1/cc.

A = the measured sperific activity, UCi/cc.

Ag = the radioactive decay constant, 1/hr.

t = the time period from reactor shutdown to sample analysis, hr.

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.' Technicel Procedure NE-5-301 Prediction of Core Damage Revision 1 10.2.5 Identification of the Fission Product Release Source l 10.2.5.1 Calculate the following ratios for each noble gas and iodine isotope only, using the specific activities obtained in step 10.2.4. Attachment 12.7 is provided as a worksheet.

Noble Gas Ratio = Nobl G pg silylly Iodine Ratio = Iodine Iso tsag_3Jpasifis_As111112 I-131 Specific Activity 10.2.5.2 Determine the source of release by comparing the results obtained to the predicted ratios provided in Attachment 12 7 An accurate comparison is not anticipated. Within the accuracy of this procedure, it is appropriate to select as the source that ratio which is closest to,the value obtained in step 10.2.5.1.

10.2.6 Calculate the total quantity of fission products available for release to the environment. Attachment 12.8 is provided as a worksheet.

10.2.6.1 If the water level in the reactor vessel recorded in step 10.1.1 indicates that the vessel is full, the quantity of fis-sion products found in the reactor. coolant is calculated by the following equation:

Total Activity (Ci) = Ao (pCi/cc) x RCS Volume 1

where:

1 Ao = the specific activity of the reactor coolant sam ple j corrected to time of reactor shutdown obtained in step l j 10.2.4, pCi/cc l 1

RCS Volume = the full Reactor Coolant System water volume t i

= 3.1899 x 18 cm3

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'. Tcchnical Procedure NE-5-301 Prediction of Core Damage Revision 1 NOTE

1. If the water levels in the reactor vessel and pressurizer recorded in step 10.1.1 indicate that a steam void is present in the reactor vessel, then the quantity of fission products found in the reactor coolant is again calculated by step 10.2.6.1. 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 re'a ctor vessel.

2. If the water level in the reactor vessel recorded in step 10.1.1 is below the low end capability of the indicator, it is not possible to determine the quantity of fission

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products from this sample because the volume of water in the reactor coolant system is unknown.

Under this condition, assessment of core damage is obtained using the containment sump sample.

l 10.2.6.2 The quantity of fission products found in the Containment )

Building sump is determined as follows: )

l A. The water volume in the Containment Building sump is determined from the sump level recorded in step 10.1.2 and the curve provided in Attachment 12.9.

B. The quantity of fission products in the sump is calculated by the following equation:

. Total Activity, Ci = Ag (pCi/cc) x Sump Volume where:

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Technical Proceduro NE-5-301 Prediction of Core Damage Revision 1 A = the specific activity of .the containment sump sample corrected l to the time of reactor shutdown obtained in step 10.2.4,

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pCi/cc j 10.2.6.3 The quantity of fission products found in the Containment Building atmosphere is determined as follows:

Gas Volume (STP) = Gas Volume x (P< - Pg+ Pg) x ((It + 46D)

T, + 450) where:

Gas Volume = Containment net free volume = 2,6 80,000 f t 3 Ti , Pg = Containment atmosphere temperature and pressure recorded in, step 10.1.2 j Tg , Pt = Standard temperature, 32 F, and standard pressure 14.7 paia 10.2.6.4 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.

10.2 7 Plant Power Correction 4

The quantitive 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 i power history.

i 10.2.7.1 To correct the source inventory for the case in wri .n plant j power level has remained constant for a period greater than four radioactive half-lives of the longest lived isotope present in the sample, the following procedure is employed.

Attachment 12.10 is provided as a worksheet.  !

l A. 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 i 10 percent l within the last 30 days prior to the reactor shutdown. Group 2 ,

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'. TGchnical Procedure NE-5-301 Prediction of Core Damage Revision 1 isotopes are to be employed in the case where core power had not changed greater than 2 10 percent within the last 4 days prior to the reactor shutdown.

B. The following equation may be applied to the fission product group which meets the criteria stated in 10.2.7.1 A only.

Group 1 Power Correction Factor = Standv-Stalg_hger Level _ Prior ~40_Daya 100 Group 2 Power Correction Factor : Steady-Stg1g_ b ygr_L3 risr_.q_ Days 10.2 7.2 To correct the source inventory for the case in which plant power level has not remained constant prior to reactor shutdown, the following equation is employed. The entire 30-day power history should be employed. Attachment 12.1 is provided as a worksheet. This calculation must be made for each isotope of interest.

gP 3(1-e i*j )e" 1*j Power Correction Factor =

100 where:

P = steady reactor power in period j , %.

t = duration of period j, hr.

t)' = time from end of period j to reactor shutdown, hr.

Ag = the radioactive decay constant, 1/hr.

(from Attachment 12.6) .

10.2.8 Comparison of Measured Data with Source Inventory The total quantity of fission products available for release to the environment obtained in step 10.2.6.4 is compared to the 19

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l source inventory corrected for plant power history obtained in I step 10.2.7.1 or 10.2.7.2. This comparison is made by dividing the two values for each isotope and calculating the percent of the corrected source inventory that is now in the sampled fluid and therefore available for release to the environment. At-tachment 12.12 is provided as a worksheet.

10.2.9 Core Damage Assessment NOTE The type of core damage is described in terms of the 10 NRC categories defined in Attachment 12.1.

The degree of core damage is d'e scribed as the percent of the fission products in the source inventory at the time of the accident which are ,

now in the sampled fluid and therefore available for release to the environment.

10.2.9.1 The conclusion on core damage is made using the three parameters developed above. These are:

A. Identification of the fission product isotopes which most characterize a given sample, step 10.2.2 B. Identification of the source of the release, step 10.2.5 C. Quantity of the fission product available for release to the environment, expressed as a percent of source inventory, step 10.2.8 10.2 9.2 Knowledgeable judgement is used to compare the above three parameters to the definitions of the 10 NRC categories of fuel damage found in Attachment 12.1. Core damage is not anticipated to take place uniformly. Therefore, when evaluating the three parameters listed above, anticipate the l

! procedure to yield a combination of one or more of the 10 l

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• t

/ Toohnicel Procadure NE-5-301 Prediction of Core Damage Revision 1 categories defined in Attachment 12.1. These categories will exist simultaneously.

10.3 CORE DAMAGE BASED ON HYDROGEN ANALYSIS 10.3.1 Complete Attachment 12.14.

10.3.1.1 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 10.1.1 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 Attachment 12.14.

10.3.1.2 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 portion of the core uncovery period. Include net flow from charging and letdown systems, HPSI, LPSI, spray, etc. Record the data on Attachment 12.14.

10.3.1.3 Record the conditions in the containment and the Reactor Coolant System at the time the hydrogen samples are obtained in step 10.3.2 following. Enter on the worksheet of Attachment 12.15. .

NOTE If the Reactor Vessel Level Indication System indicates that a void exists in the RCS and the void cannot be removed, refer to Attachment 12.23.  ;

i 21

. Technical Proceduro NE-5-301 Prediction of Core Damage Revision 1 4

10.3.2 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 2.4.1. Record the results on the worksheet of Attachment 12.15. Follow the instructions on Attachment 12.15 to obtain the total amount of hydrogen measured in units of cubic feet of hydrogen at standard temperature and pressure.

10.3.3 The total measured hydrogen in step 10.3 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 oxida-tion.

Attachment 12.16 is a worksheet for calculating the amount of hydrogen generated by oxidation of materials within the t 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 Attachment 12.17. Record the data required on Attachment 12.16 and complete the indicated calculations to obtain the cubic feet of hydrogen at STP generated by containment materials oxidation.

10.3.4 The hydrogen generated by radiolysis is a function of operating i power and decay time. Record the data required on Attachment

' 12.18, and utilize the curve of Attachment 12.19 to obtain the cubic feet of hydrogen at STP generated by radiolysis. The appropriate power is determined as follows:

10.3.4.1 For the case in which the operating power is constant or has not changed by more than i 10 percent for a period greater than

30 days, that power is used.

22

Toghnical Procedure NE-5-301 Prediction of Core Damage Revision 1 1

10.3.4.2 For the case in which the power 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.

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

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

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

D. For the case in which the reactor has produced power for less than 30 days, the procedure may be employed. However, the estimate of hydrogen from radiolysis will be too high and therefere the calculated hydrogen by core oxidation will be too low. Hence an i underprediction of core damage may result.

10.3.5 Enter the amounts of hydrogen from steps 10.3.2, 10.3.3 and 10.3.4 on the worksheet of Attachment 12.20.

Subtract the l amounts in 10.3.3 and 10.3.4 from 10.3.2, as indicated on the

! worksheet, to yield the cubic feet of hydrogen generated by core clad oxidation. Adjust with the constant shown on the worksheet to obtain the estimated percent of the core clad which is ox-idized.

l 10.3.6 Enter the abscissa of the curve on Attachment 12.21 with the percent of core clad oxidation from step 10.3.5. Use the curve labeled with the pressure closest to but greater than the RCS pressure during the core uncovery period as obtained in step 10.1.2 and recorded on Attachment 12.14. Read on the ordinate of Attachment 12.21 the percent of fuel rods with ruptured clad.

Record on the worksheet of Attachment 12.20. Note that the f 23 f

,e - . - , , - - - - , - - - - - . , , . . , - . . . . - + - . . , . . , ~ . . . . . . , , . . - , . . - . . , . -,.,...,.--,-.-,n.,.-, .,----,.,-n n, - - - - - - . - -.-- -

. Technical Proceduro NE-5-301 Prediction of Core Damage Revision 1 sensitivity of measurement of hydrogen is comparable to the range

, of oxidation on Attachment 12.21. Hence, small amounts of clad rupture are not reliably predicted by this procedure.

10.3.7 Enter the abscissa of the curve on Attachment 12.22 with the percent of core clad oxidized from step 10.3.5. 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 Attachment 12.20.

10.3.8 For a given percent oxidation of the core . lad, the lower liv.it of embrittled clad estimated in step 10 3.7 is, for most accident scenarios, the least amount of potential fuel structural failure. Actual values are probab1'y greater. The upper limit of the range in step 10.3.7 may be interpreted as follows:

10.3.8.1 When the pressure during uncovery, from step 10.1.1 and recorded on Attachment 12.14, 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 10.3.7 as determined by oxidation. Hence, use the upper limit from step 10.3.7 10.3.8.2 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 Attachments 12.21 and 12.22. 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 10.3 7 10.3.9 Core Damage Assessment The conclusion on core damage is made using the two results from j above. These are: l 10.3.9.1 Percentage of fuel rods with ruptured clad, step 10.3.6 24 .

.- _ -.- _- _ -. _ ___ _ ._ _ .__ __, _ - - . . ~ , _ _ - _ . . _

.~ Tcchnical Proceduro NE-5-301 Prediction of Core Damage Revision 1 10.3.9.2 Percentage of fuel rods with embrittled or structurally damaged clad, step 10.3.7 10.3.9.3 Knowledgeable judgement is used to compare the above two results to the definitions of the seven NRC categories of fuel damage found in Attachment 12.13. Core damaga does not take place uniformly. Therefore, in evaluating damage using these results, Attachment 12.13 may yield a combination of categories of damage which exist simultaneously.

10.4 CORE DAMAGE BASED ON RADIATION DOSE RATE IN CONTAINMENT 10.4.1 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.

10.4.1.1 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 2 10 percent within the last 30 days prior to the reactor shutdown.

10.4.1.2 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.

A. The average power during the 30-day time period is not necessarily the most representative value for correction to equilibrium conditions. ,

25 i _ - _ - _ _ _ . _ _ _ . . _ . _ . - _ _ _ _ . _ - . . - . _ _ . - - ~

.' Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 B. The last power levels at which the reactor operated should weigh more heavily in the judgement than the earlier levels.

,C. Continued operation for an extended period should weigh more heavily in the judgement than brief transient levels.

10.4.1.3 In the case in which reactor has produced power for less than 30 days, the procedure may be employed. However, the estimate oof core damage obtained under this condition may be an underprediction of the actual condition.

10.4.1.4 The following equation is applied to determine the radiation 1

dose rate corresponding to equilibrium full power source inventory conditions.

Equilirium = Measured x J00 Dose Rate Dose Rate Reactor Power Level The reactor power level and the resulting dose rate correction factor used above will be the same for all subsequent measurements of the dose rate. Record these values to reduce the work required to evaluate the subsequent measurements.

Dose rate correction factor = 100 Reactor Power Level

=

10.4.2 The decay correction for the radiation dose rate requires the determination of the time duration between the reactor shutdown and the measurement of the dose rate. This is done simply using

.the time of reactor shutdown recorded in step 10.1.4 10.4.3 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 Attachment 12.25.

The equilibrium dose rate is plotted on Attachment 12.25 as a function of time following reactor shutdown. Engineering judgement is used to determine which category of core damage  ;

shown on Attachment 12.25 is most representative of the 26 l

l Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 particular value that has been plotted. The following criteria should be considered in the determination.

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

10.4.3.2 Dose rates significantly above the lower bound for the category of major fuel overheat may indicate concurrent fuel pellet melting. This section may not be employed to estimate the degree of fuel pellet melting.

10.4.3.3 Dose rates within any category of. fuel overheating may be anticipated to include concurrent fuel cladding failure. This section may not be used to distinguish the relative contributions of the two categories to the total dose rate.

This section does give the estimate of the highest category of damage.

10.4.3.4 Dose rates corresponding to the two categories of major cladding failure and initial fuel overheat are observed to overlap on Attachment 12.25. The evaluation of other plant parameters may be required to distinguish between them.

However, concurrent conditions may be anticipated.

10.4.4 The type of core damage is described in terms of the seven NRC categories defined in Attachment 12.24. The degree of core damage is determined by the dose rate present in the containment and is read from Attachment 12.25.

l 27

.' Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 10.5 CORE DAMAGE BASED ON THE CORE EXIT THERMOCOUPLES 10.5.1 Obtain the following from the instrument recordings:

10.5.1.1 From the recording of maximum core exit thermocouple temperature as a furction of time, obtain and record on At-tachment 12.26 the maximum temperature and the time it occurs.

I 10.5.1.2 From the recording of Reactor Coolant System pressure as a  !

function of. time, obtain and record on Attachment 12.26 th'e i pressure during the period or maximum thermocouple temperature.

10.5.2 Select the curve on Attachment 12.27 which is labeled with a pressure approximately equal to or greater than the pressure in step 10.5.1.2. Enter the abscissa 'at the maximum temperature from step 10.5.1.1 and read on the ordinate the percent of the fuel rods which have ruptured clad. Record on Attachment 12.26.

10.5.3 The result in step 10.5.2 is probably a lower limit estimate of damage. Some judgement on the bias is available as follows. .

10.5.3.1 This procedure applies most directly for relatively slow core uncovery with a maximum temperature below the rapid oxidation temperatures at about 1800 F 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.

10.5.3.2 If the pressure in step 10.5.1.2 drops to less than about 100 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 thermocouple 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.

10.5.3.3 If the pressure in step 10.5.1.2 is above about 1650 psia, it could exceed the rod internal gas pressure, depending on rod 28 i

,

• Technical Procadure NE-5-301 1 Prediction of Core Damage Revision 1 i

l i

t I burnup, causing clad collapse onto the fuel pellet instead of )

outward clad ballooning. The clad rupture criteria are less well defined for such conditions, but at temperatures above 1800 F, where the highest pressure curve applies on Attachment 12.27, clad failure sufficient to release fission gas is likely and this procedure may be used to obtain estimates of damage.

10.5.4 Core Damage Assessment Use the percent of rods ruptured from step 10.5.2 and the clad damage characteristics of Attachment 12.1 to determine the NRC category of cladding failure. This procedure yields damage estimates in categcries 2, 3 or 4.

10.6 If the performance of any portion of this procedure indicates that core damage exists, see Reference 2 7, EP-1-001, Table C.

11.0 SETPOINIS NONE 12.0 ATTACH MENTS 12.1 Radiological Characteristics of NRC Categories of Fuel Damage j (Radioisotopic)

[ 12.2 Sample Locations Appropriate for Core Damage Assessment l

12.3 Record of Sample Specific Activity 12.4 Density Correction Factor for Reactor Coolant Temperature l

12.5 Record of Sample Temperature and Pressure Correction 12.6 Record of Sample Decay Correction l 12.7 Record of Fission Product Release Source Identification 12.8 Record of Release Quantity 12.9 Containment Building Water Level vs. Volume l 29 1

l

,' Technical Procedure NE-5-301 Prediction of Core Damage Revision 1 9

12.10 Record of Steady-State Power Correction 12.11 Record of Transient Power Correction 12.12 Record of Percent Release 12.13 Clad Damage Characteristics of NRC Categories of Fuel Damage 12.14 Core Uncovery Conditions 12.15 Sampling Conditions and Measured Hydrogen 12.16 Calculation Worksheet for Hydrogen Generated in Containment 12.17 Hydrogen Production Rate in Containment as a Function of Tempera ture 12.18 Calculation Worksheet for Hydrogen Generated by Radiolysis 12.19 Specific Radiolytic Hydrogen Production vs. Time 12.20 Core Damage Assessment from Hydrogen Measurement 12.21 Percent of Fuel Rods with Ruptured. Clad vs. Core Clad Oxidation 12.22 Percent of Fuel Rods with Oxidation Embrittlement vs. Total Core Oxidation 12.23 Estimation of the Amount of Hydrogen in a Reactor Vessel Head Void 12.24 Radiological Characteristics of NRC Categories of Fuel Damage 12.25 Post Accident Dose Rate Inside the Containment Building 12.26 Record of Temperature, Pressure and Damage Estimate 12.27 -Percent of Fuel Rods with Ruptured Clad vs. Maximum Core Exit Thermocouple Temperature 13.0 COMMITEESIS_JLHJ) REFIEIH.CES 1

30 l L

RADIOLOGICAL CH ARACTERISTICS OF NRC CATEGORIES OF FUEL DAMAGE (RAD 10 ISOTOPIC)

Release of Characteristic NRC Category of Mechanism of Source of Characteristic Isotope Expressed as a Ensi Dnenns Esisans Esksnas inn %nns terssak sE Doorss insators

1. No Fuel Damage Halogen Spiking Gas Gap I-131, Cs-137 Less than 1 Tramp Uranium Rb-88

2. Initial Cladding Gas Gap Less than 10 Failure Clad Burst and Gas Gas Gap Xe-131m, Xe-133 10 to 50

3. Intermediate Cladding Failure > Gap Diffusion 1-131, 1-133 Release

4. Major Cladding Gas cap Greater than 50 Failure Fuel Pellet Cs-134, Rb-88, Less than 10

5. Initial Fuel Pellet Overheating Te-129, Te-132

'5 Grain Boundary Diffusion Fuel Pellet to to 50

6. Intermediate Fuel Pellet Overheating Major Fuel Pellet Diffusional Release Fuel Pellet Greater than 50 7

Overheating From UO Grains Fuel Pellet Less than 10

8. Fuel Pellet Melt Intermediate Fuel Escape from Molten Fuel Pellet Ba-140, La-140 10 to 50 9

Pellet Melt ' Fuel La-142, Pr-144 Fuel Pellet Greater than 50

10. Major Fuel Pellet Melt t

NE-5-301 Revision 1 31 Attachment 12.1 ( 1 of 1 )

,, 4p ,

RADIOLOGICAL CH AR ACTERISTICS OF NRC CATEGORIES OF FUEL DAMAGE (RADI0 ISOTOPIC)

Release of Characteristic NRC Category of Mechanism of Source of Characteristic Isotope Expressed as a Eusk Dannns Bsisnan Rsissan inetnos Entsent 9E Boorss intsD19t1

1. No Fuel Damage Halogen Spiking Gas Gap I-131, Cs-137 Less than 1 Tramp Uranium Rb-88 Initial Cladding Gas Gap less than 10 2.

Failure Clad Burst and Gas Gas Gap Ie-131m, Ie-133 to to 50

3. Intermediate Cladding Failure Gap Diffusion 1-131, I-133 Release 4 Major Cladding Gas Gap Greater than 50 Failure Initial Fuel Pellet Fuel Pellet Cs-134, Rb-88, Less than 10 5.

Overheating Te-129, Te-132

'5 Grain Boundary Diffusion Fuel Pellet to to 50

6. Intermediate Fuel Pellet Overheating Diffusional Kelease Fuel Pellet Greater than 50

7. Major Fuel Pellet Overheating From UO Grains Fuel Pellet Less than 10

8. Fuel Pellet Melt Intermediate Fuel Escape from Molten Fuel Pellet Ba-140, La-140 to to 50 9

Pellet Melt ' Fuel La-142, Pr-144 Fuel Pellet Greater than 50

10. Major Fuel Pellet Melt l ,

,, gy , NE-5-301 Revision 1 31 Attachment 12.1 ( 1 of 1 )

[

• SAMPLE LOCATIONS APPROPRIATE FOR CORE DAMAGE ASSESSMENT SHUTDOWN ACCIDENT SCENARIO RCS CONTAINMENT CONTAINMENT COOLING XHDWj HOT LEG SHER @ SEBIEE SJSTEM Small Break LOCA, Yes --- Yes Yes Reactor Power >1%

Small Break LOCA, Yes --- --- Yes Reactor Power <1%

Small Steam Line Break Yes --- --- ---

Large Break LOCA, Yes Yes Yes Yes Reactor Power >1%

Large Break LOCA, --- Yes Yes Yes Reactor Power <15 Large Steam Line Break Yes --- Yes ---

Steam Generator Tube Yes --- Yes ---

Rupture NE-5-301 Revision 1 32 Attachment 12.2 (1 of 1) l

. RECORD OF SAMPLE SPECIFIC ACTIVITY Sample Number:

Location:

Time of Analysis:

Temperature,

• F:

Pressure, psig:

Sample Activity, pCi/cc:

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 NE-5-301 Revision 1 33 Attachment 12.3 (1 of 1) i l

k

DENSITY CORRECTION FACTOR FOR REACTOR COOLANT TDiPERATUPE I

t 700 600 - l 1

500 -

400 .

u.

o I-300 -

200 -

. l 100 -

1 I e i t )

00 0.25 0.50 0.75 1.0

~

)

1. ACT/#STP l DENSITY CORRECTION FACTOR 1

I l

l l

l 34 i4E-3-301 Revision 1 1 Attachment 12.4 (1 of 1) ,

RECORD OF SAMPLE TEMPERATURE AND PRESSURE CORRECTION

• Sample Number:

Location:

Time of Analysis:

Tem pera ture, F:

Pressure, psig:

-Measured S'pecific Activity Correction __

Specific Activity Isotops (Attachment 12-3LE1Zss -

Eas.tst 4 STP. uCi/cc 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 NE-5-301 Revision 1 35 Attachment 12.5 (1 of 1)

L

, . RECORD OF SAMPLE DECAY CORRECTION Time of Reactor Shutdown, Step 10.1.4:

Sample Number:

Location:

Time of Analysis:

Decay Specific Activity Decay Corrected Constant, 8 STP (Attachment 12.5),i Specific Activity, L1g.gnan lLhr , 11Ct/cc MCt/cc l Kr-87 5.4 (-1)

Xe-131m 2.4 (-3)

Xe-133 5 .4 ( -3 )

I-131 3.6 (-3)

I-132 3.0 (-1)

I-133 3.3 (-2) 1.0 (-1 )

I-135 Cs-134 4.0 (-5)

Rb-88 2.3 (0)

Te-129 6 .1 ( -1 )

Te-132 9 0 ( -3 )

Sr-89 5.8 (-4)

Ba-140 2.3 (-3)

La-140 1.7 (-2)

La-142 4.3 (-1 )

Pr-144 2.4 (0)

NE-5-301 Revision 1 36 Attachment 12.6 (1 of 1)

L

RECORD OF FISSION PRODUCT RELEASE SOURCE IDENTIFICATION Sample Number:

Loca tion Decay Corrected Specific Activity Calculated Fuel Pellet Activity Ratio Identified lan1Das ih11asbesnk-.12 bh.pC1Lss inninos B21192 insniets in Gan Gap sontss Kr-87 0.2 0.001 Ke-131m 0.003 0.001-0.003 4

Xe-133 1.0 1.0 I-131 1.0 1.0 I-132 1.4 0.01-0.05

, I-133 2.0 0.5-1.0 I-135 1.8 0.1-0.5 i ' Noble Gas Ratio = Dgsa2 IgtrasigdgMgbl,9)gsg Sogg((1g;((1((112 Iodine Ratio a Dasag Egtrgstad {cgiggyjyngges ((esg((gg((111112 l

L ,g gy NE-5-301 Revision 1 37 Attachment 12.7 (I of 1) i M

,' . RECORD OF RELEASE QUANTITY Reactor Coolant Containment Sump Containment Total Sample Number, t Sample Number, + Atmosphere Sample : Quantity Isotsms .G.i .Q1 Number . Ci .G.i

~ ~ ~ -

~ ~ ~

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 NE-5-301 Revision 1 38 Attachment 12.8 (1 of 1)

L -- _ _ _ - . . ~ _ _ _ _ , - . - -_ _,

. RECORD OF STEADY-STATE POWER CORRECTION Sample Number:

Location:

Steady-State 30-Day Power Level:

Steady-State 4-Day Power Level:

Fuel Power Equilibrium Corrected History Correction x Source = Source Jess -

Grouning East,sr Inventsu InventDrJ Das DAD Invent 9a Kr-87 2 9.5 ( 0)

Xe-131m 1 6.6 (4)

Xe-133 1 1.8 (7)

I-131 1 9.0 (6)

I-132 2 9.9 (3)

I-133 2 8.9 (6)

I-135 2 1.6 (6)

E.u s l 1 s l l a.%

Inventsu Kr-87 2 4 7 (7)

Xe-131m 1 7 0 (5)

Xe-133 1 2.0 (8)

I-131 1 9.9 (7)

I-132 2 1.4 (8)

I-133 2 2.0 (8)

I-135 2 1 9 (8)

Cs-134 1 6.8 (7)

Rb-88 2 9.4 (7)

Te-129 2 3 .1 (7)

Te-132 1 1.4 (8)

Sr-89 1 1.8 (7)

Ba-140 1 1.7 (8)

La-140 1 1.8 (8)

La-142 2 2.2 (8)

Pr-144 2 1.2 (8)

NE-5-301 Revision 1 39 Attachment 12.10 (1 of 1)

I

- - - . ~ _ - . - .- -

RECORD OF TRANSIENT POWER CORRECTION Sample Number:

Location:

Prior 30-Day Power History: Eevsc_1 Ducarlon. Hanra Power Correction x Equilibrium Source a Corrected Source lacknen _

EasLoc _ _ innsnLocs _ _ innsatocs _

Gna_ Gas.inssnLacs Kr-87 9.5 (0)

Xe-131 6.6 (4)

Xe-133 1.8 (7)

I-131 90(6)

I-132 9.9 (3) 1-133 8.9 (6)

I-135 1.6 (6)

Euni.2211sk innsnLott Kr-87 4.7 (7)

Xe-131m 7.0 (5)

Xe-133 2.0 (8)

I-131 9.9 (7)

I-132 1.4 (8) 1-133 2.0 (8)

I-135 1.9 (8) cs-134 6.8 (7)

Rb-88 9.4 (7)

Te-129 3 1 (7)

Te-132 1.4 (8)

Sr-89 1.8 (7)

Ba-140 1.7 (8)

La-140 1.8 (8)

La-142 2.2 (8)

Pr-144 1.2 (8) 6--s AD , NE-5-301 Revision 1 40 Attachment 12.11 ( l of 2)

RECORD OF TRANSIENT POWER CORRECTION POWER CORRECTION FACTOR CALCULATION l PCF g

= j I I3(I"* \ 3)*~ 13 100 (Step 10.2.7.2)

(

Isotoper ,

(from Attachmettt 12.6)

I

• j 2 (from pase 1 or Step 10.1) t -

l j hr (from page 1 cr step 10.1)

E'*

j hr (from page I or Step 10.1) i I

=  %

PCF, E-5-301 Revision 1 41 Attachment 12.11 (& of 2)

. _ - _ _. . _ __i. _ _ _ _ _ _ _ _ _. E' M

RECORD OF PERCENT RELEASE ,

Total Quantity Power Corrected Available For Release Source Inventory, a 100 =

1sotons LALLanbesaL 1LB1 CA GA LALLasbesaL 1L3D Dr 1211) tatsanL J ,_ _

Dan DDD lD2tDLDrat Kr-87 Xe-131 Xe-133 I-131 1-132 I-133 I-135 Eucl EsLint InsnDLor.2 Kr-87 Xe-131m Xe-133 I-131 1-132 I-133 I-135 Cs-134 Eb-88 Te-129 Te-132 Sr-89 Ba-140 La-140 La-142 Pr-144 4

4 4

ww NE-5-301 Revision 1 42 Al' -hment 12.12 (1 of 1) 1

CLAD DAMAGE CH AR ACTERISTICS OF NRC CATEGORIES OF FUEL DAMAGE

• Tem perature Mechanism Characteristic Heasurement Percent of NRC Category Rann L*El nE.Daansa nasaursasnL Ransa Danaan Rods 9LEnsLDamass v750 None -- -- Less than 1

1. No Fuel Damage Rupture due to Gas Maximum core <1550*Fe Less than 10

2. Initial Cladding Failure gap exit Overpressurization Thermocouple <1700*F8 10 to 50

3. Intermediate )> 1200-1800 Cladding Failure tem perature

4. Major Cladding "<2300 *F

--~ "<21 Greater than 50 oxidation

5. Initial Fuel Pellei Overheating Amount of Equivalent Less than 10 1800-3500 Loss of structural integrity due to hydrogen gas core

)>

fuel clad oxids- produced oxidation tion (equivalent to (35 5 oxidation of (185 core)

6. Intermediate Fuel Pellet Overheating Major Fuel Pellet 'E655 Greater than 50 7

Overheating

• Depends on reactor pressure and fuel burnup. Valves given for pressure f1200 psia and burnup 20.

wp NE-5-301 Revision 1 43 Attachment 12.13 (1 of 1)

1

. CORE UNCOVERY CONDITIONS l

1 12.14.1 Time period of core uncovery. Complete the following table using recorded instrument data.

Estimated Estimated Ingj;rygtesj; Core Unco 23rv Time Core R3 ss23ry Time Reactor Vessel Level Lower Limit Elevation Lower Limit Eleva-Monitoring System Uncovers. tion Recovers.

Time Time Core Exit Thermocouple Start of Continuous Rapid Temperature

' Temperature Rise or Exceed 6600 F. Drop to Saturation.

Time Time Temperature Temperature Core Exit Thermocouple Start of Superheat. Return to Saturation Saturation Margin or Subcooling.

Time Time 12.14.2 Interpret above data to obtain best estimate for time period of core uncovery and obtain pressurizer pressure range during that period. The superheat derived from tne thermocouple temperature and corresponcing 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 anomalies. The pressure during uncovery is us6d later on Attachment 12.19 step 10.3.6 to determine the appropriate curve for assessmen,t of the num6er of clad ruptures.

Core Uncovery Core R3sD23r3

\ Time, __

Pressure

\

12.14.3 Estimate vessel inlet flow rates during core uncovery heatup period up to approximately the time of peak core exit

. thermoc,nuple temperature. Net inlet flow indicates that procedure may have additional bias which underpredicts clad damage.

Charging Flow Rate a.

. Letdown Flow Rate l HPSI Flow Rate LPSI Flow Rate Other Inlet' F).ows NE-5-301 Revistor. 1 44 Attachment 12.14 (1 of 1) s 9

l SAMPLING CONDITIONS AND MEASURED HYDROGEN I l

12.15.1 Obtain the RCS and containment conditions at the time of sampling for l I

hydrogen.

E,gastsr Coolant Sy s133 Contaimment Sanpling Time Atmosphere pressure psig Pressure psig Atmosphere temperature F Temperature F Has hydrogen recombiner Yes/No operated?

Reactor Vessel Coolant Level  % Does pressure or tempera-ture history indicate a hydrogen burn? Yes/No Pressurizer Level 5 12.15.2 Hydrogen Sample Data Reduction.

Cont. Sample x Cont. Vol. x (32 + 460) - (Normal Temp. ) = ft Hy'rogen d

IVol. Q 1QDJ (ft3) +459 alEP x 2.6d0.000 x 492 y =

3 Hot Leg Sample x RCS Vol. x Density Ratio 1000 = ft Hydrogen (cc/kg..$ STP) (ft3) (Attashment 12.45 at STP x JJM x -

.DDD =

Total =

Also record total on Attachment 12.20.

-NE-5-301 Revision 1 45 Attachment 12.15 (1 of 1)

CALCULATION WORKSHEET FOR HYDROGEN GENERATED IN CONTAINMENT 12.16.1 Record the containment temperature at selected time intervals (i.e.,

5,10 or 15 minutes as required) and calculate the hydrogen generated by oxidation of containment materials utilizing the production rates

., from Attachments 12.17.

~ ' - -- -

2 3 4 5 Avg. Containment H2 Prod. Rate Time at Start , Interval Temp. During (ft /hr, j Hz Produced =

of.. Int erv al s Duration (hrl  ; higryal ( F) , A11EhE12D1_112 ) ( 2) r (41

' I

_L _

Accident Starts l

l l Sanpling Time l

l l

1 NE-5-301 Revision 1 46 Attachment 12.16 (1 of 1)

HYDROGE!i PRTUCTIN RATE IN CONTAINMENT AS A FUNCTION 4600 - T TEMPERATURE 4400 -

4200 -

4000 -

3800 -

3600 -

3400 -

3200 -

e

=

2 3000 -

ISs z' 2800 - -

o i-

@ 2600 c

o e

n.

2400 -

2 e 2200 o .

a:

g 2000

=

a 1800 m

, k= 1w0 -

l 1400 -

l 1200,-

1000 - .

800 -

l wo - 400 -

200 -

f I l f I I e f 1 140 160 180 200 220 240 260 280 300 l 100 120

! TEMPERATUR E,

• F

.E3-5-301 Revision.1 _ _ _ _ _ , . _

47 ,_ _ _

Attachment 12.17 (1 of 1)

CALCULATION WORKSHEET FOR HYDROGEN GENERATED BY RADIOLYSIS 12.18.1 Record the following data and utilize the curves of Attachment 12.19 to determine the hydrogen generated by radiolysis.

12.18.1.1 Prior 30-day power history Power. Pgrsgnt Duration. Days 12.18.1.2 Power to use in evaluating long term hydrogen production by radiolysis = 3390 MWt x 5 (from above) 12.18.1.3 Reactor Trip Time hr 12.18.1.4 Ssmpling Time (see Attachment 12.15) hr 12.18.1.5 Decay Time (Sampling Time - Trip. Time) _hr 12.18.2 Enter abscissa on Attachment 12.19 with above decay time and read two values of hydrogen produced by radiolysis, one from each curve, in cubic feet of hydrogen at STP per MWt operating power. Multiply by above power and record as follows:

Hydrogen Produced x Operating = Total Hydrogen limit C n (SCF/MWt. Attachmgat 12.19) Power (MWt) Produced (320, Upper ._

Lower 12.18.3 Using results from section 10.2, estimate which results should be used, upper limit for major fuel overheat or lower limit for intermediate fuel overheat. Circle corresponding value of hydrogen above and also record on Attachment 12.20. )

NE-5-301 Revision 1 48 Attachment 12.18 (1 of 1)

r'

, N - ,

+

z 14 -

i -

T SPECIFIC RADIOLYTIC HYDROGEN PRODUCTION vs TIME -

8 13 -

~

E

$ 12 -

a MAJOR FUEL OVERHEAT 8

~

g1 g 1 -

I u.

M 10 -

z' 9

h 9 3

o O

e 8 -

a.

z su c.

87 m

• O INTERMEDIATE FUEL OVERHEAT

- I 6 -

u

_a 6 -

9 g 4 -

52 is.

o 3 -

us 05 k

g n

2 - g INITIAL FUEL OVERHEAT

[

$ 1 n

7 r

~

~ ' ' ' ' ' ' ' '

• 0 O 100 200 300 400 500 800 700 800 -

S TIME, HOURS l0

CORE DAMAGE ASSESSMENT FROM HYDROGEN MEASUREMENT 12.20.1 Hydrogen Measured, step 12.15.2, Attachment 12.15 SCF 12.20.2 Hydrogen Produced in Containment, step 12.16.1, Attachment 12.16 SCF 12.20.3 Hydrogen Produced by Radiolysis, step 12.18.2, Atta chment 12.18 SCF 12.20.4 Subtract steps 12.20.2 and 12.20.3 from 12.20.1 to get hydrogen produced by core clad oxidation SCF 3

12.20.5 Divide by 5.04 x 10 SCF/1% clad oxidized = 5

= 5 Clad Oxidized 12.20.6 Enter abscissa on Attachment 12.21 with "% Clad Oxidized" and read ordinate from curve labeled with pressure during core uncovery as given on Attachment 12.14, step 12.14.2. Record here Percent of Fuel Rots with Ruptured Clad 5.

12.20.7 Enter abscissa on Attachment 12.22 with above "% Clad Oxidized" and read range of values on ordinate. Record here:

Percent of fuel rods embrittled Range - Upper 5

- Lower  %

12.20.8 Review step 10.3.1 to determine which of these limits is more likely to be representative of.the core damage.

12.20.9 From Attachment 12.13, 3 elect the core clad damage categories based on the above percentages of rods ruptured and rods embrit-tied.

NE-5-301 Revision 1 50 Attachment 12.20 (1 of 1)

PERCENT OF FUEL RODS WITH RUPTURED CLAD vs CORE CLAD OXIDATION Psico Psi A 7 RUPTURE PRESSURE 80 .

O a -

$ P512OO PSA w /

5#- Ps; lG50 Psi A t

' o -

E k

l B C

O E

O 4t 20 -

l 0 8 O 0.5 1.0 1.5 2.0

% OXIDATION OF CORE CLAD VOLUME I

IiE-5-301 Revision 1 51 Attachment 12.21 (1 ot' 1)

^

_ ,i. '~ ~ T T~ ' l . _ ~ L _ ~ ~i-- E ZZTIlT--- - --- - -- - l lT - - - - ~ -- -

PERCENT OF THE. FUEL RODS WITH OXIDATION DiBRITTLEMENT VS, TOTAL CORE OXIDATION FOR 1% TO 3% DECAY HEAT AND 300 PSIA TO 2500 PSIA WHEN COOLANT LEVEL DROPS BY BOILOFF WITH NO INLET FLOW UNTll CORE IS RAPIDLY QUENCHED 100, W

5 a

G .0 _

=

Z 9

7 80 _

e s

8

" 40 _

h E

un 8

m d 20 . ,

2 Ei at 0 100 O 20 40 60 80

% OXIDATION OF CORE CLAD VOLUME i

l NE-5-301 Revision 1 52 Attacament 12.22 (i of .1)

. .-. .- . ~, , . - . . _ , . - _ _ , . . ~ . . . - . , . . . . . - . . . - - . _ _ - . . - . - . _ . . - _ _ _ _ . .

ESTIMATION OF THE AMOUNT OF HYDROGEN IN A REACTOR VESSEL HEAD VOID 12.23.1 PURPOSE The purpose of this attachment is to provide an analytical procedure to calculate the 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 10.3.5 of the procedure to determine the total hydrogen generated by all sources.

12.23.2 LIMITATIONS 12.23.2.1 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, if the system cannot be filled, this attachment can be used to estimate the hydrogen which is in the vessel void and which would not be evident from the hot leg liquid sample.

, 12.23.2.2 This guideline 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, the guidelines of this attachment do not apply.

12.23.2.3 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 the level for each specific reactor vessel design.

12.23 2.4 This guideline 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.

NE-5-301 Revision 1 53 Attachment 12.23 (1 of 4)

l ESTIMATION OF THE AMOUNT OF HYDROGEN IN A REACTOR VESSEL HEAD VOID 1

12.23.3 PROCEDURE The following is a guideline for an analytical procedure to be followed to estimate the amount of hydrogen contained in a void in the top of the reactor vessel. Calculational details and plant specific information must be included to implement this guideline.

12.23.3.1 Determine the conditions of the void as follows:

V = Void volume (ft') derived from measurement of coolant level Tt = Temperature of liquid at coolant surfaces (*F)

Psat = Water saturation pressure at temperature TL Pw, = Reactor coolant system pressure (psia) 12.23.3.2 A first approximation is made assuming the following:

A. The partial pressure of vapor in the void is assumed equal to saturation pressure at the liquid temperature, T. g 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 superheated.

B. All the noncondensible gas in the void is hydrogen. This implies no helium or fission product gas from ruptured fuel rods and no l

nitrogen from Safety Injection Tanks. A second approximation which eliminates this assumption is given in step 12.23.3.4 below.

12.23.3.3 Calculate the amount of hydrogen as follows:

P,g = Prop - Pgat ;

fP 3I ft# Hg @ STP = (V)! lH L 442

( 14 7/ ( Tg + 430 Add this amount to the total hydrogen in step 10.3.5.

12.23.3.4 A second approximation can be made in plants with a C-E designed PASS which measures both total gas and hydrogen which are dissolved in the hot leg liquid sample. This approxima-NE-5-301 Revision 1 54 Attachment 12.23 ( 2 of 4)

ESTIMATION OF THE AMOUNT OF HYDROGEN IN l A REACTOR VESSEL HEAD VOID tion includes the following assumptions regarding the relative solubilities of the noncondensible gases in the liquid.

A. 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.

B. 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.

12.23.3.5 The partial pressure of hydrogen is calculated from H

Pg3 = ( Py -Pg) (cc/kg) 2 (cc/kg) total and the snount of hydrogen in the vessel head void is calculated using the equation above in step 12.23.3.3.

NE-5-301 Revision 1 55 Attachment 12.23 ( 3 of 4)

REACTOR VESSEL LEVEL VS. RCS VOID VOLUME (LATER)

NE-5-301 Revision 1 56 Attachment 12.23 (4 of 4) 3

suww. --.-,

RADIOLOGICAL CHAHACTEk1STICS OF NHC CATEGONIEs.Ot' ruck utrmut Parcant cf Sourc2 Invsntory Distribution of Figaica *'

Hechinisa of Source of Ersdusts_in_fectaises:1 NRC Category of Esisaac Bslsased_is_C9atainssal ',

fus1_Pasans Esisaac_fros fors Airborne Gas Gap Less than 1 No Fuel Damage Italogen Spiking 1.

Tramp Uranium Less than 10 Airborne Gas Gap

2. Initial Cladding Failure 10 tp 50 Airborne Gas Gap Intermediate Clad Burst and Gas

3. 3 cladding Failure Gap Diffusion Release Airborne Gas Gap Greater than 50

4. Hajor Cladding Failure Airborne Fuel Pellet Less than 10 1005 Noble Gas

5. Initial Fuel PellII 255 Halogen Overheating Grain Boundary Diffusion 10 to 50 Plated out Fuel Pellet 255 Halogen

6. Intermediate Fuel Pellet Overheating 15 Solids fuel Pellet Greater than 50 Major Fuel Pellet Diffusional Release 4

7.

overheating From U0) Grains i

1 J

4 k

i 57 Attachment 12.24 (1 of 1)

NE-5-301 kevision 1 l

4 5 _ _ _ _ _ _ _ _ _ _

1 POST ACCIDENT DOSE RATE INSIDE THE CONTAINMENT BUILDING 1x108 l

1 l

E l

l i 1e -

'%+  %,.

'4/p ~ #

4g #

47 4r  !< #+g

'%. c<4 , *%,

1x104 - #,

4 09 p

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7

1. 4 4 4

4/g Y+g I

1x103  !

1 10 100 1000 1 TIME POST ACCIDENT, HOURS j CYLINDRICAL CONTAINMENT l

___ NE+ntamspaI - - - -

5 Attad"""e 2.23 c1 a u /

E

, RECORD OF TEMPERATURE, PRESSURE AND DAMAGE ESTIMATE 12.26.1 Record the following data:

Maximum Core Exit Thermocouple Temperature 0F Time of Maximum Temperature Reactor Coolant System Pressure at Above Time psia 12.26.2 From Attachment 12.27, at maximum thermocouple temperature and at appropriate pressure, read percent of ruptured rods. 5 12.26.3 Comment on probable bias of result in 12.26.2 (see section 10.5.3 in text) .

12.26.4 NRC category of cladding failure from Attachment 12.1 NE-5-301 Revision 1 59 Attachment 12.26 (1 of 1)

b PERCENT OF FUEL RODS WITH RUPTURED CLAD vs MAXIMUM CORE EXIT THERMOCOUPLE TEMPERATURE t

100 _ _

I u- "

P6100 PSIA S

E .

k B

= 80 -

mzoo es*

1 8

a l

a 40 - ~

i E p41650 PSA s

o l

)

3 20 -

5

a. )

)

\

0 1200 1400 1903 1800 ll000 2200 MAXIMUM CORE EXIT THERMOCOUPLE TEMPERATURE j

l

.iE-5-301 Revision 1 60 Attachment 12.27 (1 of 1)

I ..

e LOUISI AN A POWER & LIGNT C O.

IE WATERFORD 3 M b* MISSING INFORM ATION LIST PROCEDURE NC. / PROCEDURE TITLE V IC /O M O CbII #9 Rev. I D

ITE M NO. ITEM PARAGRAPH DATE INITIAL I Attachvp,ty 12,9 l 2. 9 2 A tk ch r,g.r/ /2. z.3 (3 o,<3 ) '

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ALL MIS $1NG INFORMATICN ON THl3 LIST HAS SEEN CLEARED )

STARTUP ENGINEER / ASSIGNED AUTHOR l

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