ML110050430

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HC-EP-EP.ZZ-0205(Q), Rev. 04, TSC-Post Accident Core Damage Assessment
ML110050430
Person / Time
Site: Salem, Hope Creek  
Issue date: 07/27/2004
From:
Public Service Enterprise Group
To:
Office of Nuclear Reactor Regulation
References
LR-N10-0355 HC-EP-EP.ZZ-0205(Q), Rev 04
Download: ML110050430 (42)
v • d • e


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I I PSEG Internal Use O(nly Page.1-of.1-HOPE CREEK GENERATING STA-TION HC.Ep*EP.ZZ-0205 (Q)I Rev. 04 TSC - POST ACCIDENT CORE DAMA!.9E ASSESSMENT USE CATEGORY: II REVISION

SUMMARY

Biennial Review Performed: Yes.-!-

No' Made changes to reflect the elimination of the Post:.J\\ccident Sampling System (PASS) as the primary means of assessing post accident corel.IJlarnage as per LeR 149.

Made enhancements to remaining core damage assess ment methods to reflect guidance provided in NEDC-33045P, "Methods of Estimating Core Damage in SWRs", July, 2001.

Restructured the procedure to retain gas and water based isotopic sampling as a contingency core damage assessment method.

This revision is considered a major re-write and tharrsfore, revision bars are not utilized.

IMPLEMENTATION REQUIREMENTS Effective Date: o 71 a.., J 8 (J 0'1

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TSC - POST ACCIDENT CORE DAMAGE ASSESSMENT TABLE OF CONTENTS Section ID!!

1.0 PURPOSE......... II *******************,..............................., ********************** " ****************** 11.3 2.0 PREREQUISITES.'1 *********************, *********************************, **., **** 1'111 ************************ 3 3.0 PRECAUTIONS AND LIMITATIONS................................................................ 3 4.0 EQUIPMENT REQUIRED................................................................................ 4 5.0 PROCEDURE.,.......................,...............,... '..*. _,.. _............................................* 5 6.0 7.0 5.1 Core Thermal-Hydraulics Engineer Should Perform the Following to Initiate Core Carnage Assessment (CDA) and CDA Sample Results............................................ ~................................ 5 5.2 Estimating the Type and Extent of Core Damage Based on the Drywell Atmosphere Post Accident (DAPA)

Equivalent Calculation **..**.**....**.********.***.*.**..**.****** ~............................. 6 5.3 Determining the Percent of Zirconium Oxidation from 'the Hydrogen Concentration In the Primary Containment Free 5.4 5.5 5.6 5.7 5.8 5.9 Volume..................................................................... ".............................. 7 Estimating If an Interruption of Adequate Core Cooling Has Occurred............. I............................ ~ ************* I111....................

111............ 7 Estimating the Type arid Percent of Core Damage From Fission Product Concentrations.....................,.................................... 8 Utiiizing the Normalized Concentrations......................................... 11 Estimating Release Source (Gap or Fuel Pellet) From the Isotopic Ratios...................................................................................

11... 11 Determine If Less Volatile Fission Products are Present in the Reactor Coolant............................................................................. 11 Performing an Assessment of the Type and Extent of Core Damage Based Upon All Available Indicators........ "..... 11........ 12 5.10 Reporting the Results of the Assessment and Recommending Further Actions................................ iII....................... 12 RECORDs...,..,.... '.... I" ** ' **** ' ** " ***************** i.' **. t **************** 1., *********, **, ************* 4.'... 12 REFERENCES....................,..........................

111 ******* "...................

,......................... 13 Hope Creek Page 1 of 41 Rev. 04

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ATTACHMENTS ATTACHMENT 1.. POST ACCIDENT RADIONUCLIDE SAMPLE REQUEST.........*......... 14 ATTACHMENT 2.. DAPA MONITOR DOSE RATE TO FUEL INVENTORY AIRBORNE... 15 ATTACHMENT 3.. PRIMARY CONTAINMENT HYDROGEN CONCENTRATION TO 0/0 ZIRCONIUM OXiDATION..................................................................... 17 ATTACHMENT 4 - FISSION PRODUCT CONCENTRATIONS........................................... 20 ATTACHMENT 5.. FISSION PRODUCT INVENTORY CORRECTION FACTORS............ 22 ATTACHMENT 6 - NORMALIZED CONCENTRATION OF FISSION PRODUCTS............ 25 ATTACHMENT 7 -1.. 131 CONCENTRATION VS. INDICATION OF CORE DAMAGE........ 28 ATTACHMENT 8.. 1.. 133 CONCENTRATION VS. INDICATION OF CORE DAMAGE........ 29 ATTACHMENT 9.. 1-135 CONCENTRATION VS. INDICATION OF CORE DAMAGE........ 30 ATTACHMENT 10.. C8-134 CONCENTRATION VS. INDICATION OF CORE DAMAGE... 31 ATTACHMENT 11-CS-137 CONCENTRATION VS. INDICATION OF CORE DAMAGE... 32 ATTACHMENT 12-KR.. 8Sm CONCENTRATION VS. INDICATION OF CORE DAMAGE.. 33 ATTACHMENT 13-KR-85 CONCENTRATION VS. INDICATION OF CORE DAMAGE..... 34 ATTACHMENT 14-XE-133 CONCENTRATION VS. INDICATION OF CORE DAMAGE... 35 ATTACHMENT 15-XE.. 135 CONCENTRATION VS. INDICATION OF CORE DAMAGE..* 36 ATTACHMENT 16-ISOTOPIC RATIO INDICATION OF RELEASE SOURCE................... 37 ATTACHMENT 17-CORE DAMAGE ASSESSMENT

SUMMARY

, DETERMINATION &

RECOMMENDATIONS................................ ~...................................... 39 Hope Creek Page 2 of 41 Rev. 04

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o z HC.EP-EP.ZZ-0205(Q) 1.0 PURPOSE This procedure provides guidance for core damage assessment after an ALERT or higher level of emergency has been declared with the reactor shut down.

2.0 PREREQUISITES Implement this procedure:

At the discretion of Core Thermal.. Hydraulics Engineer (CTE)

Upon staffing of your Emergency Response Facility.

3.0 PRECAUTIONS AND LIMITATIONS 3.1 Precautions 3.1.1 It is recommended that initials be used in the place keeping sign-offs, instead of checkmarks, if more than one person may implement this procedure.

3.1.2 Personnel who implement this procedure shall be trained and qualHied lAW the Emergency Plan.

3.1.3 If additional support is needed for performing Fuel Damage Assessment, contact the Nuclear Fuels Manager.

3.2 Limitations

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3.2.2 The core damage assessment methodology assumes reactor coolant cleanup systems are isolated.

3.2.3 Measurement of Cs-137 and Kr-85 activities may not be possible until shorter.. Uved isotopes have decayed.

Page 3 of 41 Rev. 04

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HC.EP.. EP ~ZZ-0205(Q) 3.2.4 Clad damage of less than 1 % is not considered to be a loss of the fuel cladding boundary.

3.2.5 Radiation level measurements may underestimate core damage if:

A.

The primary containment or RPV has been vented.

B.

Primary system isolations have been defeated to permit continued use of the main condenser under failure.. to.. scram conditions.

c.

Primary containment integrity has been lost.

3.2.6 Radiation level meaSurements may overestimate core damage if:

A.

The suppression pool has been bypassed.

B.

Suppression pool water level is low.

3.2.7 Hydrogen concentration measurements may underestimate core damage if:

A.

The primary containment has been vented.

B.

Primary containment integrity has baen lost.

C.

Significant amounts of hydrogen remain trapped in the RPV.

3.2.8 Hydrogen concentration measurements may overestimate core damage if:

A.

Significant amounts of hydrogen have been generated by radiolysis.

B.

C.

The hydrogen injection system is leaking.

Steam is present in the drywell but the drywall atmosphere is not at saturation conditions

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Hope Creek Page 4 of 41 Rev. 04

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He. EP-EP.ZZ-0205(Q) 5.0 PROCEDURE NOTE:

Due to the multiple and, at times, unpredictable failure n1echanism associated with core damage this procedure has been developed to provide GUIDANCE for Core Damage Assessment. The sequenoe and extent of procedure performance should be based on the knowledge and experience of the Core Thermal-Hydraulics Engineer.

5.1 Core Thermal-Hydraulics Engineer Should Perform the Following to Initiate Core Damage Assessment (COA) and CDA Sample Results 5.1.1 PERFORM HCGS plant-specific calculations and estimations of the types and extent of reactor fuel damage utilizing the guidance of this procedure. [CD-385Y]

[CD..

548X]

5.1.2 IF the Drywell Atmosphere Post Accident Monitor has been declared inoperable by Operations, THEN GO TO step 5.2.

5.1.3 ESTIMATE the type and extent of core damage based on the Drywell Atmosphere Post Accident (OAPA) Radiation Monitor Reading.

5.1.4 OBTAIN and record on Attachment 2, DRYWELL ATMOSPHERE POST ACCIDENT (OAPA) MONITOR A AND BREADING (RlHR), the time of the reading and the time of reactor shutdown.

NOTE DAPA monitor A and B provide indication for two different locations in the Drywell.

If adverse conditions exist in the Drywall (average Drywell air temperature greater than or equal to 245°F) validate with the Radiological Assessment Coordinator that EPIP 302H J Attachment 5, DAPA CORRECTION CALCULATIONS has been utilized.

Hope Creek Page 5 of 41 Rev.. 04

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HC.EP*EP.ZZ*0205(Q) 5.1.5 DETERMINE the percent of fuel inventory airborne in containment using Attachment 2. Record the result on 7.

5.2 Estimating the Type and Extent of Core Damage Based on the Hope Creek Drywell Atmosphere Post Accident (DAPAl Equivalent Calculation 5.2.1 IF the DAPA monitors were operable, GO TO step 5.3, OTHERWISE CONTINUE with step 5.2.2.

5.2.2 INFORM the TSTL of the need to determine drywall atmosphere radiation levels without the DAPA monitor.

5.2.3 REQUEST from the Radiological Assessment Coordinator a "Contact Dose Rate" at the Drywell Personnel Airlock and a Particulate, Iodine, and Noble Gas air Sample of the 120' EI.

Rx. Bldg. for the purposes of determining a DAPA EQUIVALENT READING. Calculate a "DAPA EQUIVALENT" value and document the value on as a DAPA EQUIVALENT in the following manner:

EQUIV = 100 x (CDR - {20 R1HRI~i/cc x (NGC))

WHERE: EQUIV = DAPA Equivalent (RlHR) for use in Attachment 2 (if CDR== normal bkg then CDR = 0) 5.2.4 CDR = Contact Dose Rata (R/HR)

NGC = Nobel Gas Concentration (fLCi/cc)

(if NGC is < 1 E-04 JA,Ci/cc then NGC = 0)

DETERMINE the percent of fuel inventory airborne by using. Record the result on Attachment 17.

Page 6 of 41 Rev. 04

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HC.EP-EPRZZ-0205(Q) 5.3 Determining the Percent of Zirconium Oxidation from the Hydrogen Concentration In the Primary Containment Free Volume.

5.3.1 OBTAIN the hydrogen concentration in the primary containment from the Hydrogen-Oxygen Analyzer System and record it on Attachment 3.

5.3.2 RECORD the time of the reading or sample and the sample point on Attachment 3.

5.3.3 RECORD on Attachment 3 any drywall venting or hydrogen recombiner operation.

5.3.4 DETERMINE the percent Zirconium oxidation by using. Record the result on Attachment 17.

5.4 Estimating If an Interruption of Adequate Core Cooling Has Occurred.

5.4.1 OBTAIN a history of the reactor vessel water level from the initiation of the accident from SPDS or the VAX LA 120.

5.4.2 DETERMINE if the top of active fuel (TAF) has been uncovered.

5.4.3 RECORD the level history, duration of level below the TAF and an estimate of cooling adequacy on Attachment 17.

NOTE Significant or core-wide damage is not expected unless the TAF has been uncovered. Core.. wide clad damage can occur within 30 minutes of uncovering the fuel. However, unless level is below the bottom of the active fuel, boiling heat transfer will provide cooling and significantly extend the duration that a partial uncovering can be withstood without significant core damage.

Hope Creek Page 7 of 41 Rev. 04

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Core 'U:ncovery Time vs. Core Da'mage Time 20o/t) of Active core lUlcovered 0.5 to 0.75 hrs 0.5 to 1.5 hl'S 1 to 3+hrs Core Temperature 1800.. 2400°F NOTE Core,Dalllage Condition Rapid oxidation Cladding datllage (gap reJease)

Overheating damage Eutectic formation Core geometry changes Core lllelting RPV breach (ex... vessel release)

The primary methods for assessing core damage are provided in Steps 5.1 thru 5.4 above. Core damage assessment utilizing analysis of fission product concentrations obtained through sampling reactor coolant or drywall atmosphere provided in Steps 5.5 thru 5.8 below, can be pe~ormed as a supplementary method if so desired.

5.5 Hope Creek Estimating the Type and Percent of Core Damage From Fission Product Concentrations.

5.5.1 I F no radionuclide sarnpling/analysis is to be perlormed at this time, THEN GO TO Step 5.9.

5.5.2 PROVIDE recommendations to the Radiological Assessment Coordinator (RAC) to initiate post accident radlonuclide samples and review all requests for radionuclide samples for the purpose of contingency core damage assessment. [CO-4430]

5.5.3 DETERMINE need and frequency for post accident radionuclide samples with consideration that the application of the results will be for contingenoy Core Damage Assessment Only.

Page 8 of 41 Rev. 04

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NOTE Depending on the severity of the accident, radionuclide sample results may not be available for several da s followin the accident.

5.5.4 OBTAIN the post accident radionuclide sample results with consideration as to how representative the sample will be of the core condition.

5.5.5 Recommend to the RAe sample points based upon reactor condition or event type.

A.

SELECTION of Liquid Sample Point NOTE Residual Heat Removal (RHR) samples: If RHR is in the Low Pressure Coolant Inlection (LPC1) or Suppression Pool Cooling modes, it should be operating an estimated 30 minutes minimum prior to sampling to ensure a representative sample.

[CD.. 384Y]

LIQUID SAMPLE LOCATION SAMPLE PANEL Reactor Water Recirc.

10.. C-2S1 Reactor (RHR.. LPCI/Shutdown Cooling) 10.. C.. 250 Torus Water (RHR)

  • aO.. C-350
  • [CD-384Yl GAS SAMPLE EVENT SAMPLE LOCATION SmalVLarge Break Drywall Atmosphere B.

RECORD on Attachment 1 the current time, selected sample point, the desired frequency of sampling and the basis for the selection and frequency.

Hope Creek Page 9 of 41 Rev. 04

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PROVIDE a copy of Attachment 1 to the RAC and the Technical Support Team Leader (TSTL).

5.5.6 IF a liquid sample has been selected as identified on, obtain from the Chemistry Supervisor in the TSC the concentratJon of 1.. 131,1-132, 1-133,1.. 134,1-135, C8-134, Cs.. 137, sample point, sampling time, sample analysis time, type of decay correction performed and the time of f[nal reactor shutdown. Record the information on.

5.5.7 IF a gas sample has been selected as identified on, obtain from the Chemistry Supervisor in the TSC the concentration of Kr.. 85m, Kr.. 85, Kr-87, Kr.. 88, Xe-133, Xe-135, sample point, sampling time, sample analysis time, type of decay correction petiormed, and the time of final reactor shutdown. Record the information on.

5.5.8 CALCULATE the pressure/temperature corrected fission product concentrations for gas sam pie radioisotopes as per.

NOTE Pressure/temperature corrections will not be necessary if the corrections have been performed by the Chemistry Department.

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5.5.9 CALCULATE the decay corrected fission product

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Hope Creek 5.5.10 CALCULATE the fission product inventory correction factors (FI) as per Attachment 5.

5.5.11 CALCULATE the normalized concentrations of the fission products (Cwn) as per Attachment 6.

Page 10 of 41 Rev. 04

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HC.Ep*EP.ZZ-0205(Q) 5.6 Utilizing the Normalized Concentrations 5.6.1 Following the instructions on Attachment 6 and Attachments 7 through 15 estimate the percent cladding failure and percent fuel melting.

5.6.2 Record the results on Attachment 17.

NOTE The lines on the graphs are set up in the following manner:

Upper Dashed Line - maximum fission product release for a given fuel condition.

Lower Dashed Line - minimum fission product release for a given fuel condition.

Center Solid Line - nominal fission product release for a given fuel condition.

5.7 Estimating Release Source (Gae or Fuel Pellet} From the Isotoglc Ratios.

5.7.1 CALCULATE the isotopic ratios as per Attachment 16.

5.7.2 COMPARE the calculated isotopic ratios to the values listed in the table on Attachment 16 to estimate the release source.

Record the results on Attachment 17.

5.8 Determine If Less Volatne Fission Products are Present In the Reactor Coolant.

5.8.1 IF the less volatile fission products, such as Sr, Ba, La, or Ru (either soluble or insoluble), are found to have unusually high concentrations in the reactor coolant some degree of fuel melting may be inferred.

5.8.2 RECORD observations of less volatile fission products on 7.

Hope Creek Page 11 of 41 Rev. 04

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HC.Ep.. EP.ZZ-0205(Q) 5.9 Performing an Assessment of the Tvpe and Extent of Core Damage Based Upon All Available Indicators 5.9.1 CLASSIFY the type and extent of core damage relative to the following matrix.

Degree of Core Damage Minor <<10%) Intermediate (10°A,.. 50%)

Major (>500k)

None <<1 % clad) 1 1

Clad Fa'ilure 2

3 Fuel Overheat 5

6 Fuel Melt 8

9 5.10 5.9.2 EVALUATE the other indicators or parameters to corroborate and fu rther refine the assessment as determined in section 5.9.1.

5.9.3 REQUEST that the TSTL INITtATE appropriate confirmation of accuracy if conflicting indications are identified.

5.9.4 RECORD the assessment and bases on Attachment 17.

Reporting the Results of the Assessment and Recommending Further Actions.

5.10.1 REPORT the results to the TSTl for dissemination to the TSS and the RAe.

5.10.2 REVI EW the current accident status in order to make recommendations for further actions to refine or continue the assessment.

5.10.3 RE-ENTER the procedure as appropriate.

6.0 RECORDS Return completed procedure, original copies to the EP Manager.

Hope Creek Page 12 of 41 1

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7.0 REFERENCES

7.1 Refe renees 7.1.1 General Electric Document, NEDO.. 22215 82NED090, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions, August 1982.

7.1.2 General Electric Document, C&RE Transmittal, RPE 81 CL01, November 1981 7.1.3 PSEG Nuclear Radiation Protection/Chemistry Services File NRP-88-0048, Preplanned Alternate Monitoring Methods for the DAPA Monitoring System, March 3, 1988.

7.1.4 Hope Creek UFSAR 1.14.1.49.2 7.1.5 Genera) Electric Document, N EDC-33045P, Methods of Estimating Core Damage in SWRs, July 2001 7.2 Cross References 7.3 Hope Creek 7.2.1 EPIP NC.EP-EP,ZZ"0302(Q), Radiation Assessment Coordinator Response.

7.2.2 PSEG Nuclear Enlergency Plan Closing Documents 7.3.1 Hope Creek CD.. 443D 7.3.2 Hope Creek CD-384Y 7.3.3 Hope Creek CD.. 385Y 7.3.4 Hope Creek CD-548X.

Page 13 of 41 Rev. 04

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Time of Request Sample Point Frequency Bases Comments Sample Request No.

Time of Request Sample Point Frequency Bases Comments Hope Creek ATTACHMENT 1 Page 1 of 1 RADIONUCLIDE SAMPLE REQUEST Page 14 of 41 HC.EP*EP.ZZ.. 0205(Q)

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ATIACHMENT2 Page 1 of 2 HC.EP-EP.ZZ.. 0205(Q)

DAPA MONITOR DOSE RATE TO FUEL INVENTORY AIRBORNE Time of Reactor Shutdown _________ _

Dat9 _____ _

Time of DAPA reading: A _________ _

Date ------

B _______________ __

Dat9 _____ _

4.

Complete the following Table for each reading.

1 2

3 4

5 6

5.
6.
7.
8.

DAPA EaUIV Monitor Dose Rate Time after Cladding/Over (Y/N)

(AlB)

(RlHr)

Shutdown heating damage (Hrs)

(0/0)

For cladding damage estimation, determine the drywall radiation levels corresponding to 10Qok cladding damage from Figure 1 below.

Estimate the amount of cladding damage as follows and record in table above:

01' ctl dd'

'0 Indicated Radiation Level

'I (')0 10

  • a Lng 31nage =

X '

100% Clackhng Dmllage RadJubon Level For overheating damage estimation, determine the drywell radiation levels corresponding to 100% overheating damage from Figura 1 below.

Estimate the amount of overheating damage as follows and record in table above:

Indica.ted Radiation Level

% Overllcating 'Damage =

x 100 100% Ove:rheatillg Radiation Level Hope Creek Page 15 of 41 Rev. 04

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l Hope Creek ATTACHMENT 2 Page 2 of 2 Figure 1: Drywell Radiation Levels Page 16 of 41 HC.EP.. EP.ZZ-0205(Q)

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ATTACHMENT 3 Page 1 of3 PRIMARY CONTAINMENT HYDROGEN CONCENTRATION TO % ZIRCONIUM OXIDATION Time System and Sample H2 (Ok)

Point Zirconium Oxidation (0/0)

Comments and Drywell Ventlng/RecomblnerOperatlon Note Page 17 of 41 Rev. 04

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HC.Ep.. EP.ZZ-0205(Q) 10 20 30 40 50 60 70 80 90 100 Average Pnmaty Contaln~~ ~droQ~~ Concentration (%)

Hope Creek Page 18 of 41 Rev. 04

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J Hope Creek ATTACHMENT 3 Page 3 of 3 Zirconium Oxidation Fractions He. EP*EP.ZZ.. 0205(Q)

Damage Phase Total Zr Oxidation (% )

No damage

<1 %

Clad. damage l.. S%

Overheating damage 5-10%

Core melt 10-20%

Page 19 of 41 Rev. 04

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ATTACHMENT 4 Page 1 of 2 FISSION PRODUCT CONCENTRATIONS Time of Reactor Trip or Shutdown _________ _

Sample No. ______ Sample Time ______ _

Sample Analysis Time ____________ _

Sample Type _____ Sample Point _____ _

Sample Vial P1 _____ psi T1 _____ 0K Sample Point P2 _____ psi T2 OK Environment PTMULT = P2 T1/ P1 T2:::: _""""--____ (PTMULT = 1.0 for liquid)

Decay Correction _________________ - ____ _

DMULT = eAt

'A = decay constant of the isotope of interest (1/days) t:::: time of decay (days)

NOTE The time of decay must represent the elapsed time from reactor trip or shutdown to the sample analysis time.

NOTE The deoay correction must account for the activity decrease during the time period from reactor trip or shutdown to the sample analysis time.

Hope Creek Page 20 of 41 Rev. 04

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Decay Corrected

("Cl/g)

Corrected (J.tCi/g)

(IlCi/g)

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FISSION PRODUCT INVENTORY CORRECTION FACTORS

1.

Calculate the inventory correction factor (Fi) for each 'fission product listed in steps 3 and 4 of Attachment 5 using the following:

1.1 Bases

FI = reference inventory of isotope i in HCGS actual inventory of isotope i in HCGS 1.2 If the total operating time for all batches is greater than or equal to the power correction time:

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  • For each time period, Tt the variation of steady reactor power, Pt, should be limited to +/- 20%.

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  • For each time period, T]", the variation of steady reactor power, Pj*,

should be limited to +/- 20%.

2.

Each fission product must be corrected for either 6 half-lives or 3 fuel cycles whichever is shorter. The times are delineated in steps 3 and 4 as the "Power Correction Time".

3.

Liquid Sample Fission Product Power A

3293

  • Fj Correction Time (1/day)

(1_e-1095J.i) 1.. 131 49 days 8.621E.. 2 3.293E3 1-133 6 days 7.998E-1 3.293E3 1-135 2 days 2.517E+O 3.293E3 Cs.. 134 3 fuel cycles 9.219E-4 2.089E3 Cs.. 137 3 fuel cycles 6.294E.. 5 2.192E2 Hope Creek Page 23 of 41 Rev. 04

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  • F, (1_e-1095Ai) 3.293E3 5.800E2 3.293E3 3.293E3 Rev. 04

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ATTACHMENT 6 Page 1 of3 NORMALIZED CONCENTRATION OF FISSION PRODUCTS

1.

For each fission product in steps 2 and 3 of Attachment 6 perform the following calculation using tha applicable data from Attachment 4 and Attachment 5.

Where:

Cw = the normalized concentration of the fission product (uCi/g for liquids and uCi/cc for gases)

Ct = the decay and pressure/temperature corrected fission product concentration from Attachment 4.

Fl = the inventory correction factor from Attachment 5.

2.

Liquid sample - Activity concentrations dispersed equaHy through reactor water and torus water Fission Product Ct PI Cw 1-131 1-133 1-135 Cs.. 134 Cs.. 137

3.

Gas Sample - Activity concentrations dispersed equally through drywell and torus free volumes Fission Product Ct F,

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4.

Additional normalizations may be required if plant parameters indioate that the specific activity from a liquid sample represent a sample environment different than the reference environment. Concentration or dilution corrections should be petiormed and documented in step 7. Reference and typical constants required for the correotions are delineated in step 6.

5.
6.

Reference mass = the total mass of the reactor water and torus water If the actual mass of liquid water does not equal the reference mass a correction factor should be applied.

Fdlo = actual mass (9) I reference mass (g)

Additional normalizations may be required if plant parameters indicate that the specific activity from a gas sample represents a sample environment different than the reference environment. Concentrations of dilution corrections should be petiormed and documented in step 7. Reference and some typical constants required for the corrections are delineated in step 6.

Reference volume = drywall plus torus free volume If the actual volume of gas does not equal the reference volume a correction factor should be applied.

FdiC = actual mass (co) / reference mass (cc)

Dilution/Concentration Data Reference liquid mass Reactor liquid n1ass At Power Hot Standby Cold Shutdown Torus liquid mass Reference gas volume Torus free volume Drywall free volume 3.B33E9 9 (8.01 E6 Ibs) 2.93E8 9 !6.46E5 IbS}

3.0SES 9 6.68E5 Ibs 4.09E8 g 9.02E5 Ibs 3.34E9 9 (7.36E61bs) 8.57E9 co (3.03E5 ft3) 3.78E9 cc (1.33E5 fts) 4.79E9 cc (1.69E5 fts)

Hope Creek Page 26 of 41 Rev. 04

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ATTACHMENT 11 Page 1 of 1 Cs-137 CONCENTRATION VS. INDICATION OF CORE DAMAGE CI.137 Concentration,G. IndicatiDn af Core Damage I

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Xe--133 CONCENTRATION VS.INDICATION OF CORE DAMAGE Xe-133 Cancentration VI. Indication of Core Damage

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ATTACHMENT 15 Page 1 of 1 Xe-135 CONCENTRATION VS. INDICATION OF CORE DAMAGE Xe-135 Concentration vs. indication of Cora Damage

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ISOTOPIC RATIO INDICATION OF RELEASE SOURCE

1.

Obtain the decay corrected fission products from Attachment 4 and calculate the ratios as described in step 2.

2.

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'-131 1-134 1-131 1.. 135 IM131 Hope Creek ATTACHMENT 16 Page 2 of 2 Ratio In Pellet (indicates fuel melt) 0.122 0.233 0.33 1.46 2.09 2.3 1.97 Page 38 of 41 HC.EP-EP.ZZ-0205(Q)

Ratio in Pellet/Clad Gap (indicates clad damage) 0.023 0.0234 0.0495 0.127 0.685 0.165 0.364 I

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CORE DAMAGE ASSESSMENT

SUMMARY

, DETERMINATION AND RECOMMENDATIONS Date: -------

Time: _____ _

Summary No.: ___ _

1.

Assessment of amount and type of core damage based on DAPA readings.

2.

Assessment of the % Zirconium oxidation and corresponding clad failure (determine in conjunction with assessment of adequacy of core cooling if possible).

3.

Assessment of the adequacy of core cooling.

4.

Assessment of release source based on isotopic ratios.

Hope Creek Page 39 of 41 Rev. 04

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5.

Observations of less volatile fission products.

6.

Core damage estimates based on fission product concentrations from samples as determined utilizing Attachments 7-15.

1~133 1-135 Cs-134 Cs-137 Kr-85 Xe-133 Xe-135 NOTES Hope Creek Page 40 of 41 Rev, 04

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7.

Summary, Determinations and Recommendations

8.

Final Core Damage Estimate Core Thermal-Hydraulics Engineer Hope Creek Page 41 of 41 HC.EP*EP.ZZ-Q205(Q)

Date Time Revw04

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