ML20072S893

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PRA Evaluation of Population Dose Risk from Severe Accidents at San Onofre Nuclear Generating Station Units 2 & 3
ML20072S893
Person / Time
Site: San Onofre  Southern California Edison icon.png
Issue date: 08/18/1994
From: Chien S, Fenstermacher, Hook T
SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML20072S884 List:
References
PRA-2-3-94-012, PRA-2-3-94-12, NUDOCS 9409140282
Download: ML20072S893 (30)
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Text

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9 A NUCLEAR SAFETY GROUP PROBABluSTIC RISK ASSESSMENT REPORT PRA EVALUATION OF POPULATION DOSE RISK FROM SEVERE ACCIDENTS AT SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 August 18, 1994 NSG/PRA REPORT PRA-2/3-94-012 Prepared by:

(T. G. Hook)

Prepared by: 6v r

. Chaen)

Reviewed by: ' '

s by .

(T. 'R. Fenst'ermacher, PLG)

Reviewed by: d/-

[.J.

/

Lee) .

Approved by: W (C./Chiu) l 9409140282 940902 PDR P

ADOCK 05000361 Pnp

,. 4 NSG/PRA REPORT PRA-2/3-94 012 l i

PURPOSE i

The purpose of this evaluation is to determine the population dose risk from severe accidents initiated by internal events at San Onofre Nuclear Generating Station (San Onofre) Units 2 and 3.

This type of evaluation is characterized in probabilistic risk assessment (PRA) terminology as a Level 3. The result of the evaluation is the estimation of the likelihood that the general population will be exposed to radioactivity from releases following a severe accident initiated by internal events at San Onofre Units 2 and 3.

BACKGROUND A Level 1 and Level 2 PRA of internal events, identified as an Individual Plant Examination (IPE), was conducted for San Onofre Units 2 and 3 in response to NRC Generic Letter 88-20 (Ref. 1).

The IPE was submitted to the NRC in May 1993. The Level 1 portion of the PRA determined the likelihood of core damage from internal initiating events. The Level 2 portion of the PRA determined the likelihood, magnitude, and timing of radioactive releases from the plant following the core damage events evaluated in the Level 1 PRA.

A Level 3 PRA determines the impact of radioactive releases ,

evaluated in the Level 2 portion of the PRA on the' population and land use. The results from the~ Level 3 PRA include estimates of the following parameters associated with each Level 2 release type: population doses, early fatalities,' latent fatalities, land interdiction costs, and other impacts associated with the dispersion of radioactive material beyond the site boundary.

The inputs to the Level 3 PRA include: the likelihood, timing, and magnitude of radioactive releases from the units; the typical weather pattern'for the site; the population distribution surrounding the site (up to 100 miles) ; the land use surrounding the site (up to 100 miles); the emergency declaration procedures; and the. emergency evacuation plans. With the exception of the likelihood of each type of radioactive release, all the inputs are fed into'a consequence analysis code which determines the average radiological impact of each release type from the Level 2 PRA. The MACCS (MELCOR Accident Consequence Code System) code was chosen for the calculation'since it is was developed for and utilized by the NRC for consequence evaluation (Ref. 2).

The quantity of information required to generate the input to the MACCS code is enormous. A significant amount of effort is required to evaluate historical weather _ data from the site, determine population distributions in radial sectors around the ,

site up to 100 miles, and determine land use up to 100 miles from i

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l MSG /PRA REPORT PRA-2/3-94 012 the site. Also a large number of parameters associated with each release type in the Level 2 PRA must be evaluated and reformatted I for input in the MACCS code. I 1

The development of a sample MACCS code input applicable to the SONGS site was performed in 1991 by NUS Corporation (Ref. 3).

The sample code input utilized actual site weather data, 1990 land use information, and the 1990 population census to accurately reflect the characteristics of the SONGS site and the surrounding area. The sample code input also evaluated the SONGS emergency evacuation plan to determine the timing and directional factors affecting population evacuation and/or sheltering during an emergency. The only inputs in the sample code which were not determined were the characteristics of the radionuclide releases (magnitude, timing, energy of plumes), since the Level 2 portion of the SONGS PRA was not complete at that time.

Since the Level 2 portion of the San Onofre Units 2 and 3 PRA is now complete, all the information required to generate the MACCS input is now available. This report documents the development of the MACCS input files, performance of the MACCS runs, and review of the results.

METf0DOLOGY The Level 2 portion of the San Onofre Units 2 and 3 PRA resulted in the determination of the likelihood, timing, and magnitude of radioactive releases from severe accidents. Fifteen distinct source term release categories were specified in the Level 2 PRA.

Each source term release category was evaluated using the MAAP (Modular Accident Analysis Program) code. The MAAP code simulates the accident conditions in the core, reactor coolant system, other plant systems, and the containment to determine how, when, and how much radioactive material would be released from the plant in a severe accident. The MAAP code was specifically developed by the nuclear industry for this use and was utilized by almost all utilities in the performance of their Level 2 PRAs in response to NRC Generic Letter 88-20.

The parameters from the MAAP runs which are used in the Level 3 MACCS code input include:

o cumulative radionuclide release fractions for each major fission product group over time (up to 48 hrs) o sensible heat in the released radionuclides over time o timing associated with the start and duration of major radioactive release plumes o timing associated with conditions where a General Emergency would be declared NSG/PRA REPORT PRA-2/3 94-012 The definition of the radionuclide release groups in MAAP and j MACCS differ slightly. MAAP uses 12 radionuclide release groups, whereas MACCS uses 9 radionuclide release groups. The differences between the release groups are summaried in Table 1.

Table 1 '

Radionuclide Release Groups Radionuclide Type MAAP Release Group MACCS Release Group Nobles Gases Nobles Xe/Kr I CsI I Cs CsI, CsOH Cs Te TeO 2, Te2 , Sb Te Sr Sr0 Sr Ru, MO2 MO 2 Ru Ba Ba0 Ba La La20 3 La ,

Ce CeO 2 Ce UO 2 UO 2 None There is a direct correspondence between all the radionuclide release groups except Cs and Te. For these two radionuclide release groups, mass balance equations in Appendix A correspondence from PLG were utilized. The release fractions and masses for each radionuclide type were taken from the MAAP analyses. Table 2 provides the masses used in the calculation:

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NSG/PRA REPORT PRA-2/3-94-012 Table 2 Radionuclide Type Masses From MAAP Radionuclide Type Mass (kg) Applicable Source Term Sequence (s)

CsI 39.1 All CsOH 331.35 All TeO 2 0.0 All except:

1.9 MLO-4 40.38 PCS-35 1E-5 SBO-17 3.12 LLO-4 0.61 ATWS-26 Te2 0.0 All except:

33.83 MLO-4 3.03 PCS-35 34.45 SBO-17 32.88 LLO-4 34.9 ATWS-26 ASSUMPTIONS

1. The start time and duration of radioactive material release plumes were determined by reviewing the cumulative radioactive release plots from each source term sequence MAAP run. Dr. Edward Fenstermacher of PLG, a consultant expert in consequence analysis, reviewed the cumulative release plots and recommended plume start times and durations (see Appendix A for correspondence).
2. The sensible heat associated with each plume segment for each source term sequence was calculated from the MAAP code.  !

Mr. Cris Henry of FAI provided a minor modification to the MAAP code which would calculate the sensible heat of the radionuclide release (see Appendix A for correspondence).

The average sensible heat during a release period (i.e.,

plume) was used in the MACCS input for each plume.  !

3. The time associated with the declaration of a General l Emergency was determined based on the failure or imminent failure of all three fission product barriers (fuel cladding, reactor coolant system, and containment) per SONGS Emergency Plan Implementing Procedure SO123-VIII-1 (Ref. 4).

Since the imminent failure of a fission barrier is I subjective, the predicted failure time was used for l conservatism. The failure times of the fission product

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NSG/PRA REPORT PRA-2/3-94-012 barriers were determined from Table 4.7-6 (Source-Tenn Analysis Results MAAP Run Summary Table) of the San Onofre Units 2 and 3 IPE submittal (Ref. 1). The assumed time when a General Emergency is declared for each source term sequence is provided in Table 3.

Table 3 Time Assumed When General Emergency Declared Source Time of Time of RCS Time of Time Term Cladding or Vessel Containment General Sequence Failure Failure Failure Emergency (hr) (hr) (hr) Declared (hr)

PCS-4 1.7 2.5 * **

MLO-4 1.3 0.0 * **

LLO-4 1.2 0.0 * **

ATWS-6 0.6 1.6 * **

LOP-48 1.7 2.6 39.8 39.8 PCS-35 10.3 11.5 * **

SLO-20 1.3 3.6 26.5 26.5 LLO-13 0.6 0.0 22.3 22.3 SBO-17 1.7 2.7 0.0 2.7 LLO-32 0.6 0.0 0.0 0.6 VSEQ-2 5.6 0.0 0.0 5.6 SGTR-20 10.4 0.0 0.0 10.4 SGTR-33 1.7 0.0 0.0 1.7 SGTR2-48 10.4 0.0 0.0 10.4 SGTR2-66 10.4 0.0 0.0 10.4 Notes: l

  • Denotes no containment failure predicted by MAAP within 48 l hours. )
    • Denotes conditions not sufficient for declaration of General l Emergency per SO123-VIII-1, Rev. O. Time General Emergency declared is set to 7 days in MACCS input.  !

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NSG/PRA REPORT PRA-2/3-94-012 l

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4. The radionuclide release height for each plume was assumed to be 10 meters, except for steam generator tube rupture i events where is was assumed to be 30 meters. San Onofre j Units 2 and 3 are situated below a bluff next to the ocean '

I on the Southern California coast. Winds blowing in the direction of population centers (inland or along the coast) must in general pass up over the bluff surface unless released from the upper areas of the containment such as the main steam relief valve discharges. The most likely.

locations for containment failure during post-accident pressurization are associated with the penetrations located on the low portion of the containment. Due to the large ,

number of bluffs, hills, and mountains between the site and i the major population centers, the impact of the release height on the population dose is assumed small. l ANALYSIS The sample MACCS input for SONGS was used as the basis for the analysis. The release information from the MAAP runs for each of the fifteen source term sequence analyses were incorporated into the MACCS sample input.

The fifteen source tenn sequence MACCS input files were analyzed using MACCS to determine the population dose consequences. The risk measure of interest in the MACCS outputs is the mean population dose (in units of sieverts) from 0 to 1609 kilometers (i.e., 100 miles) from the site using the combined emergency ,

response cohorts (i.e., the "OVERALL RESULTS" section of output).

The total population dose risk was calculated by multiplying the frequency of each source term by its population dose consequence, and summing the risks for all fifteen source term sequences.

RESULTS The results of the fif teen source term sequence MACCS runs are provided in Appendix B.

The frequency, mean population dose, and mean population dose risk (i.e., frequency time dose) for each of the fifteen source term sequence MACCS runs are provided in Table 4.

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NSG/PRA REPORT PRA 2/3-94-012 Table 4 Summary of Results from MACCS Runs Source Term Frequency Mean Mean Sequence (/yr) Population Population Dose to 100 Dose Risk Miles (man- (man-rem /yr) rem)

PCS-4 9.0E-6 3.27E+2 2.94E-3 MLO-4 6.8E-6 7.14E+2 4.86E-3 LLO-4 4.3E-6 5.37E+3 2.31E-2 ATWS-26 2.8E-6 5.12E+2 1.43E-3 LOP-48 2.1E-6 2.02E+4 4.24E-2 PCS-35 2.7E-6 1.40E+4 3.78E-2 SLO-20 6.9E-7 2.83E+6 1.95E+0 LLO-13 6.2E-8 1.60E+6 9.92E-2 SBO-17 1.9E-8 3.55E+6 6.75E-2 LLO-32 1.4E-9 1.78E+6 2.49E-3 VSEQ-2 6.5E-7 1.74E+7 1.13E+1 SGTR-20 7.9E-7 3.90E+6 3.08E+0 SGTR-33 2.2E-7 1.40E+4 3.08E-3 SGTR2-48 1.5E-7 4.78E+6 7.17E-1 SGTR2-66 2.BE-7 4.53E+6 1.27E+0 TOTAL 3.0E-5 NA 1.86E+1 CONCLUSION The estimated mean population dose risk from severe accidents due to internal initiating events at San Onofre Units 2 and 3 is 19 man-rem /yr.

The accident initiators contributing over 98% to the total population dose risk at San Onofre Units 2 and 3 are:

o Interfacing System LOCA (61%)

o Steam Generator Tube Rupture (27%)

o Small LOCA (11%)

NSG/PRA REPORT PRA 2/3-94-012 A comparison of population dose risk for San Onofre Units 2 and 3 with other nuclear plants is provided in Table 5. The population dose risk from San Onofre Units 2 and 3 is lower than all other plants in Table 5 with the exception of Grand Gulf.

Table 5 Comparison of Population Dose Risk Plant Population Dose Risk (man-rem /yr)

Grand Gulf 6 San Onofre 2 &3 19 Peach Bottom 28 Surry 31 Sequoyah 80 Zion 136 Note: Population Dose Risk for other plants was taken from Draft NUREG-1493 (Ref. 5).

REFERENCES

1. " Individual Plant Examination Report for San Onofre Nuclear Generating Station, Units 2 and 3, in Response to Generic Letter 88-20", Southern California Edison, April 1993.
2. NUREG/CR-4691, "MELCOR Accident Analysis Consequence Code System," Sandia National Laboratories, Prepared for the Nuclear Regulatory Commission, February 1990.
3. " SONGS Consequence Analysis Using MACCS", Halliburton NUS Environmental Corporation, August 1991.
4. SONGS Emergency Plant Implementing Procedure SO123-VIII-1,

" Recognition and Classification of Emergencies", Revision 0.

5. Draft NUREG-1493, " Performance-Based Containment Leak Test Program," Nuclear Regulatory Commission, Draft for Comment.

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NSG/PRA REPORT PRA-2/3-94 012 APPENDIX A Technical Correspondence I

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VANAGEMEttr CONSUJANTS August 8,1994 l

i Mr. Sam Chien San Onofre Nuclear Generating Station l Southern California Edison Company l P.O. Box 126 San Clememe, C A 92674 )

Dear Sam-I discovered after reading your fax of Augiast 2,1994 that there was indeed an error in the calculation of the Cs and Te release fractic as from the release fractions in MAAP j categories 2 and 6 ( for Cs) and 3 and 11 ifor Te), but it was in equation 3. The amended )

equations are presented below.

We first determine the fractions of Cs in the mass of Csl and CsOH. These are

(!)

f[' = _w, c

+'_w, W, e W' c (2) j,c, , _,

W,+Y+Y e o n Wc , + 17.0073 g I gmot k

where eW , , W,,o W and W,, are the average atomic weights of cesium, iodine, oxygen j and hydrogen in the material rt: leased. Then derme the fraction of Cs in each of the MAAP release categories:

ff'm, (3)

  1. 2 = ll'm, +fl'm, (4)
  1. = 1-42 in MAAP categories 2 and 6 respeedvely, and where mi and m, are the total mass of Csl and CsOH available for release. T% Cs release fraction is then (5)

S c, " #sf2 + #68, 6 A- 2. I 1

C0d C00 GW VOS3Hl.33 ONI old- 0500 206 IOC1 CC:Ot :E e :

{

)

where (3 and 4 are the release fractions for MAAP categories 2 and 6, respectively. i f

As we discussed, this operation needs to be performed for each fission product group:

each phase of each release. The MAAP release fraction for the first phase i '

taking the release fraction from the graph (or printout) at a time equal to thel the becond phase. For each subsequent phase, the release fraction is taken at the l beginning of the next phase (or the end of the relcase), and subtracting all prior ph I hope this error has not delayed your analysis. Also, Keith told me that you m the values for m, and m, needed. Please call me and we'll see if we can come strategy for getting the values you need.

Keith indicated that you had gone back to Fauske for ar explanation of the anomal resu;ts of the sensitivity runs. Do we need to do anyth;ng additional about this, o bee resolved?

I have expanded the tables to include the last two release scenarios. These tab .

Start time for Phase, Variable RDPDELAY001 -

Values in seconds Phase 3 Phase 4 Phase 1 Phase 2 Scauence 47700 83700 8100 11700 l PCS-4 72000 151200_

7200 36000 LOP 48 21600 57600 7200 14400 MLO-4 45900 81900 8100 9900 SGTR 33 10800 43200 5400 7200 '

VSEQ2 61200 97200 38700 49500 PCS-35 148500 165600 39600 116100  !

SGTR2-48 72000 93600 '

39600 41400 SGTR2-66 18000 36000 2700 6300 LLO-34 39600 NONE 8100 21600 SBO-17 95400 136800 5400 81000

__MLO-20 97200_

3600 18000 82800 LLO-13 57600 93600 39600 43200 SGTR 20 55800 91800_

4500 8100 LLO-4 8100 18000 2700 6300 ,

ATWS26  ;

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l 20d COO 'CW VCS3H138 ONI CGd 0500 LOG 10C1 EC:01 t-5,80 T M l

l Duration of Phase, Variable RDPLUDUR001 i

Values in seconds Phase 2 ( Phase 3 Phase 4 Sequence Phase 1 36000 36000 89100 PCS-4 3600 36000 79200 21600 LOP-48 28800 7200 36000 115200 i MLO-4 7200 36000 36000 92700 SGTR-33 1800 3600 7200 10800 VSEQ2 1800 11700 36000 75600 ,

PCS-35 10800 32400 17100 7200 SGTR2-48 7200 30600 21600 79200 SGTR2-66 1800 11700 18000 136800 3600

_LLO-34 133200 NONE 13500 18000 S8 0-17 36000 27000 14400 41400 MLO-20 14400 64800 14400 10800__

LLO-13 +

3600 14400 18000 79200__

SGTR-20 47700 36000 81000 LLO-4 3600 1800 2700 9000 ATWS26 1800 RDNUMRELOOl, the number of plume segments, wit! be 4, except as noted. The risk dominant segment, RDMAXRSK001, is indicated by having the start time in bold face type. The representative time point, RDREFTIM001, may be conservatively set in zero for all segments of all plumes.

Very truly yours, z{

T. Edward Fenstermacher l

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DATE: July 25,1994 TO: Sam Chien, San Onofre Nuclear Generating Station FROM: Christopher E. Henry, FAI, MAAP Maintenance Group (JI,8

SUBJECT:

MAAP variables representing mass and sensible energy flow *, the environment from a containment leak or breach I believe that I have acquired the requested information regarding MA AP variables which represent mass and energy flow from a containment leak or breach to the environment.

While there may not be a direct correspondence between these variables and the required input parameters for the MACCS code, the following information should be sufDcient to generate the appropriate input parameters.

The first part will focus on energy transport to the environment via sensible heat transrer of gaseous plume emanating from either a containment leak or breach. Note that a containment breach can reside in either the upper compartment (A compartment) or the annular compartment (D compartment). Containment leakage is confined to the D compartment.

The second part will then focus on energy transfer to the environment via transport of 6ssion product gases and aerosols, and their associated fraction of decay power, within the plume.

Plume Sensible Heat Transfer: NormalIrakace in the D Comnartment 1

As noted above, normal containment leakage to the environment is assumed to occur in the D compartment, presumably due to the fact that most containment penetrations are located at  ;

this elevation. As shown in Attachment 1, the gas mass flow rate through the leakage, WGLK, is calculated via a call to GPLOW within subroutine EQUIL. Notice that the input parameter of normal leakage area, ALKNOM, which is an argument within the call to GFLOW, is used to compute WGLK.

Attachment 1 then shows a section of coding from regional subroutine DCOMPT. The I sensible heat transfer associated with WGLK is included as part of the calculation of FUGD, which is the rate of change of gas internal energy within the D compartment control volume.

Specifically, the gas sensible heat transfer is the first circled term within the rate equation.

A breakdown of the term shows that it is simply the product of WGLK and the summation of each constituent's mass fraction multiplied by its specific enthalpy. (Note, for l noncondensable gas constituents, the specific enthalpy is the product of the gas temperature, TGLK, and the constituent's specific heat (CPH2 for hydrogen, for instance). For steam, the specific enthalpy, HSTLK, is known directly.)

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Page 2 In addition to the gas sensible heat transfer, a second component to plume sensible heat transfer is suspended water, which constitutes a fog resulting from steam condensation in the D compartment atmosphere that is also transported within the plume. As shown in the EQUIL coding within Attachment 1, the suspended water mass flow rate, WSWLK, is calculated from WGLK in the line immediately following the call to GFLOW. The sensible heat term for the suspended water is shown as the second circled term in the DCOMPT coding within Attachment 1, immediately after the noted gas sensible term.

Plume Sensible Heat Transfer: Containment Failure in the D Comonrfment In the case of a containment failure in addition to the normal leakage in the D compartment, calculation of sensible heat transfer via gas and suspended water transport through the containment breach is performed in essentially the same manner as that for normal leakage.

The gas and suspended water flow rates, WGCFD and WSWCF respectively, are calculated in subroutine EQUIL, as shown by the section of coding in Attachment 2. The respective terms for sensible heat transfer are shown as the two circled terms in the section of coding from subroutine DCOMPT in Attachment 2. Like the corresponding leakage terms, both gas and suspended water flow through the containment breach contribute to the calculation of the rate of gas internal energy change in the D compartment, symbolized by FUGD.

Plume Sensible Heat Transfer: Containment Failure in the A Comnartment The containment failure can also be specified as occurring in compartment A rather than D.

As shown in Attachment 3, gas and suspended water flow rates, WGCFA and WSWCF respectively, are also calculated in subroutine EQUIL. The respective sensible heat transfer terms are circled in the included section of subroutine ACOMPT coding. Like the sensible heat terms in the D compartment, these terms contribute to the rate of change in the gas internal energy for the A compartment, symbolized by FUGA.

Plot File Outout of De_stred MA AP Parameters The noted sections of coding reveal that the desired sensible heat terms have not cast in terms of discrete parameters that can be casily added to user-defined plot files, such as Plot File 77. It is possible to add all contributing parameters, such as flow rates, constituent mass fractions and specific heats, temperature, enthalpy, etc., to the plot file and then perform the sensible heat transfer calculations external to MAAP.

However, a much more efficient method entails making assignments of the sensible heat transfer to elements of the PLT array. For instance, suppose that leakage and containment failure occur in the D compartment. To utilize the PLT array for plotting, the named j common /XPLTX/, which contains array declaration PLT (500), must be added to subroutine i DCOMPT, as shown by the first "CREV SONGS" code modification in Attachment 4 l (Note, the sections of DCOMP'T coding in Attachment 4 are essentially the same as those shown in Attachment 2.) As shown, this line is added adjacent to the other common declarations at the top of DCOMPT.FOR.

In the latter "CREV SONGS" code modification, PLT (1) and PLT (2) are assigned the plume sensible heat transfer through the normal leakage and the containment breach, respectively.

PLT (1) and PLT (2) can then be added to a user-defined plot file, yielding the desired data in A- G

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Page 3 terms of two discrete parameters. Of course, since the DCOMPT.FOR has been modified, a new MAAP executable must be generated by recompiling DCOMPT and linking it to the MAAP object library. This limited amount of code modification will render a code that is -

tailored to your specific needs.

If you need further assistance on any of the points discussed above, please contact me.

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Page 4 Attachment 1 MAAP coding corresponding to plume sensible heat transfer

'from normal leakage in the D compartment A-9 1

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EQUIL. TOR CODING CALL GTLOW ( 0,1.D0, PTINL, TGD, VOLGT, CTD, PE, T3E,1. D 6,1. D0, TD,

"." 126 2VGLK,MTSTLK MTH2 LK, MTO2 LK, MTCOLK, MTC2 LK, MrN2 LK, ALLO 40M, 0. D 0, WG LK)

WSWLK-WSWLK*WGLK WAWLX=WAWLK*WGLK i

cccNPT.FOR CODING r e

FUGD=QTLD-QDIFD+QBRND-QGOW-QGIW+

CREV P16.04 RAA/MAM 11/30/90 END e 1WGBD*(MFSTBD'HSTBD+TGBD*(MTH2BD*CPH2+MTO25D.CPO2+MTCOSD*CPCO+

2MFC2BD* CPC2 +MTN2 BD* CPN2) ) +

1WGAD*(MFSTAD*HSTAD+TGAD*(MTH2AD*CPH2+MT02AD*CP02+MTCOAD-CPCO+

2MFC2AD*CPC2+MTN2AD*CPN2))-

3WGCTD* (MTSTCT*HSTCr+TGCT* (MTH2Cr*CPH2+MTO2CT *CPO2+MFCOCT*CPCC+

CREV P16.04 TJD 11/01/90 START '

C 6MTC2CT*CPC2+MTN2CT*CPN2))+QGFN*ITCD+GGTPD+WSTFP=HSTW-C 7 LAMS D* MSWD

  • KWS TD +WSW B D
  • HW S T B D +W SWAD* HWS TAD +

C GWGntA/MGA MSWA*ITCD*HWSTA-WSTAW'HWSTD-CREV P16.05 TJD 1/7/91 START C 6MFC2CT

  • CPC2 +MTN2 C r* C PN2 ) ) +QGTPD+WSTTP
  • HSTW+QTNODE* QGRD-6MTC2CT* CPC2 +MTN2 C T* CPN2 ) ) +QGE PD+MSTEP.HSTW +NQT* QGRD-CREV P16.05 FJD 1/7/91 END 7 LAMS D*MSWD* KWSTD+WSWBD* HW STBD+WSWAD* HWSTAD-WSTAW* HWS TD-CREV P16.04 ran 11/n1/4n FNn 3WGLK* ( MFSTLK' HST LK+T GLK* (MTJi? r,x. c pH 2 +MT02 LK' CP02 +MFCO LK
  • CPrO+

6MTC2LK* CPC2 +MTN2 LK' CPN2 ) )'* WSWLR*ReisTLK -

GAS SENSIBLE HEAT

  • SUSPENDED WATER SENSIBLE HEAT
  • a l

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  • FAI GUIDANCE TO MODIFY MAAP TO ENSURE COMPATIBILITY WITH MACCS. , )

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Page 6 Attachment 2 MAAP coding corresponding to plume sensible heat transfer from a containment breach in the D compartment 1

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. c.w. u Page 7 EQUIL.FOR CODING

- CALL GFLOW ( 0,1. DO, PFINL, TGD, VOLGT, CFD, PC , TGE ,1. D6,1. D0, TD,

. 116 2VGCF, MFSTCF, MFH2CF, MF02CF, MFCCCF, MFC2 CF , MFN2C F, CD* ACT, O . DO, WGC FD) f.

WSWCF=WSWCF*WGCFD WAWCF=WAWCF *WGCFD l

DCcseT.rcR CODING FUGD=0FLD-QDITD+QBRND-QGDW-QGIW+

CREV P16.04 RAA/MAM 11/30/90 END l 1WGBD*(MFSTBD*HSTBD+TGBD*(MFH2BD*CPH2+MF02BD*CPO2+MFCCBD*CPCO+ j 2MFC2BD*CPC2+MFN2BD*CPN2))+ j

~

1WGAD* (MFSTAD*HSTAD+TGAD* (MFH2AD*CPH2+MF02AD*CPC2+MFCOAD*CPCO+ i l

2Mrr2AD

  • cPc2 +wrN1n 0* c ?!4' )

(~f4GCFD* (MrSTt*F*liSyrr+TGCT* f MTH2CF* CPH2+MF02CE* C702+MFCOCr*CPCO+

CREY 'c'lo . u s FJD 11/01/90 START 4 C 6MFC2 C F

  • C PN2 ) ) +QG FN' I FCD+QG FPD+WST F F
  • HST4- l C 7 LAMSD
  • MSW D
  • HVS TD+WSWBD* HWSTBD+WSWAD
  • MWST AD+ ,

C BW G FNA/ MGA* MSWA* I FC D

  • HWSTA-W S T AW ' W4S TD-CREV P16.05 FJD 1/7/91 START

,, Aurc7Cr* cPC2 '"rui"** "*"' 1 ) +QGTPD+WSTEP'HSTW+QTNODF.*QGRD-6MFC2 C E

  • C PC 2 +MFN2 CF
  • C PN2 ) ) FQGF FD+WST T P
  • HSTW +NQT
  • QGRD-c.xEV P16.05 PJD 1/7/91 ENDr 7 LAMS D
  • MSUD
  • HWSTD+WSWBD
  • HWST BD+W SWAD* HWSTAD-WSTAW
  • HWST D-CREV P16.04 TJD 11/01/90 END 3WGLM* (MFSTLK+ HSTLK+TGLK* (MFH2 LK* CPH2*MF02LK* CPO2 +MFCOLK* CPC0+

6MFC2LK*CPC2+ MEN 2LK*CPN2))-WSWLK'HWSTLK IF (WGCFD.EQ.0.DO) GO TD 1111 FMSWD=FMSWD-W5WCF FMAWD=FMAWD-WAWCF o rMwc=rW G-Wsure.w">--

FvGD-FUGD wSWCF'HWSTCr SUSPENDED WATER 1111 I F (ISORT . c2-+! - men SENSIBLE HEAT.*

t

} *FAI GUIDANCE TO MODIFY MAAP TO ENSURE COMPATIBILITY WITH MACCS.

. A- i t

~

sus a:1cau,u ' ub. 5s . . a o, w r- "L na i ,n -.1.-' .i< av .ic,,= u.t.

4 s Page 8 Attachment 3 MAAP coding corresponding to plume sensible heat transfer from a containment breach in the A compartment 2

l l

l 4

Pago 9

- EQUIL.roK CCDING ,

.114- CALL GFLOW (0,1.D0, PFINL, TGA, VOLGT, CFA, Pr., TGE,1. DS,1.D0, TD, 2VGC F, MFSTCF, MrH2 C F, Mro2 C r, MrCoC r, MrC2 C T , MFN2 C T , CD* ACF , 0. 00, WGCFA)

WSWCT=W5WCF*WGCrA WAWCr=WAWCF*WGCFA ACCNPT.FCR CCDTNG q

. . l

! FUGA=QFLA-QDITA-QGSP1A-QGSP2A+QBRNA+

l CREV P16.04 RAA/MAM 11/30/90 END I 1WGBA* (MTSTBA* HSTRA+TGEA* (MFH2BA* C FH2+MF025A* CPO2 +MFCOBA* CPC0+

l 2MIC2BA*CPC2+MFN2BA*CPN2))- I IWGAD* (MFSTAD*HSTAD+TGAD* (MTH2AD*CPH2+MiO2AO* CP02+MFCOAD CPCO+

) 2MFC2AD* CPC2 +MTN2AD* CPN2) ) +

l l 3WGUA* (MFSTUA* HSTUA+TGUA* (MrH2UA* CPH2+MF02UA* CPO2 +MTCOUA* CPCO+

4HFC2VA*CPC2+MFN2UA*CPN2)l-CREV P16.04 TJD 10/23/90 START ,

C SVGrNA* (MFSTA*HSTA+TGA* (MrH2A* CPH2+MTO2A*CP02+MFCCA* CPCO+

C 6MFC2A*CPC2+MTN2A*CPN2))-

CREV P t A . 0 4 " *" 1 ' '" # co **

'sWGCrA* (MrSTCrahSTCr+TGCF* (Mruxt*ren+Mro2Cr Cpo2+MrCoCr*CPCo+ GAS SENSIBLE ~

HEAT

  • 6MFC2CT*CPC2+ MEN 2CF*CPN2)' QGTPA+WSTrP=HSTW+0STA+QdCNA-

't QGUW= w2 wr-gu A wv-Qww,w .- QGEQ- LAMS A* MSWA* HWSTA +WSWBA* HWST B +

) CREV P16.04 RAA/MAM 11/30/90 START l CCREV P16.04 TJD.10/23/90 START CC SWSWU A* HWS TUA-WSWAD* HWSTAD-WGFNA/MGA*MSWA* h",iSTA-WSTAW" HW STA+QGDCH C BWSWUA*HWSTUA-WSWAD*HWSTAD-WSTAW'HWSTA+0GDCH CCREV P16.04 FJD 10/23/90 END' GWSWUA*HWSTUA-WSWAD*HWSTAD-WSTAWA*HWSTA+CGDCHA

)

IF (WGCFA.EQ. 0.DO) RETURN

  • 1HSWA= mSWA-WSWCr FMAWA=FMAWA-WAWCF FMWG-mWG-vsWCr-Wm-ruga =rUnr wSWCF'HMSTCF SUSPENDED WATER RETURN SENSIBLE HEAT
  • 1 1

I

l 4!

i, EI 1

.'4 i

. *FAI GUIDANCE TO MODIFY MAAP TO ENSURE COMPATIBILITY WITH MACCS. 1

  • A- I L d R'

n~

Page 10 l

l l

Attachment 4 Sample DCOMPT code modification for utilizing PLT ()

array assignments as a means of retrieving sensible heat transfer data from D compartment j leakage and containment failure plumes l

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l, l,

z I

1

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%. w. -

Page 11 DCCNPT.FOR CODING COMM5N/D3/TroWDo,TFoWBo,7FIWDo CREV S014GS *

  • NEW CODING FOR SENSIBLE HEAT OUTPUT *
  • START C ADD /XPLTX/ COMMON To SUBROUTINE NEW CODING
  • C CCMHON/XPLTX/ PLT (500)

C END CREV SONGS ** NEW CCDING FOR SENSIBLE HEAT CUTPUT **

FUGD=QFLD-QDITD+QBRND-QGCW-QGIW+

CREV P16.04 RAA/MAM 11/30/90 END 1WGBD* (MTSTBD* HSTBD+TGBD* (MFH2BD* CPH2+MF02BD*CPo2+MFCOBD*CPCo+

2MFC2 BD*CPC2+MFN2 BD

  • CPN2 ) ) +

1WGAD*(MFSTAD*HSTAD+TGAD*(MFH2AD*CPH2+Mro2AD*CPc2+MFC0AD*CPCC+

2MFC2AD*CPC2+MFN2AD*CPN2))-

3WGCFD* (MFSTCF'HSTCF+TOCF* (MFH2CF *CPH2+MF02CF* CPo2+MFCOCF* CPC0+

CREV P16.04 FJD 11/01/90 START C 6MFC2CF*CPC2+MFN2CT*CPN2))+QGFN*IFCD+QGTPD+WSTFP*HSTW-C 7 LAMSD*MSWD

  • HWSTD+WSWBD* HWSTBD +W SW AD
  • HWSTAD*-

C BWGFNA/ MGA* MSWA* I TCD

  • HW S TA-WST AW ' HWS TD-CREV P16.05 PJD 1/7/91 START C 6MFC2CF*CPC2+MFN2CF*CPN2))+QGFPD+WSTTP'HSTW+QTNODE*QGRD-6MFC2CF*CPC2+MTN2CT*CPN2))+CGTPDtWSTYP*HSTW+NQT*QGRD-CREV P16.05 FJD 1/7/91 END

? LAMSD* MSWD* HWSTD+WSW BD

  • HWSTBD+WSWAD
  • HWSTAD-WSTAW ' HW STD-CREV P16.04 FJD 11/01/90 END 3WGLK*(MFSTLK*HSTLK+TGLK*(MFH2LK*CPH2+MF02LK'CP02+MFCOLK*CPC0+

EMFC2 LK

  • C P C2 +MFN2LK* CPN2 ) ) -WSW LK* HWSTLK START CREV SONGS ** NEW CODING FOR SENSIBLE HEAT OUTPUT **

C PLT (1) - SENSIBLE HEAT TRANSFER THRCUGH CONT. LEAK IN D CCMPT NEW CODING

  • C PLT (2) - SENSIBLE HEAT TRANSFER THROUGH CONT. BREACH IN D CCMPT C

PLT (1) =WGLK+ (MFSTLK* HSTLK+TGLK* (MFH2 LK*CPH2 +Mro2LK'C P02 +MFCOLK* C

. MFC2 LK* CPC2+MFN2LK'CPti2) ) +WSWLK* HWSTLK PLT ( 2 ) =WGCFD * (MFSTCr* HSTCF+TGCF* (MFH2 CF

  • CPH2 +MF02CF* CPO2 +MFCO

. MTC2CF*CPC2+HFN2CF*CPN2))+WSWCT*HWSTCF CREV SONGS ** NEW CODING FOR SENSIBLE HEAT OUTPUT ** END IF (NGCFD.EQ.0.00) Go To 1111 FMSWD=FleWD-WSWCF INAWD==D90fD-WAWCr FNWG=FlSO-WSWCF-WANCr FUGD-rudD-WsWCr*HWsTCr 1111 IF (ISoRT. EQ.1) THEN .

  • FAI GUIDANCE TO M0DIFY MAAP TO ENSURE COMPATIBILITY WITH MACCS.

, A-15 1

e'LU. ;nc. 7J13 VW50ers n Avrive, Suite C20 Eas*. Detheace. Mcy a,'d not M200

, Tel 331407 9100

  • Fax 301 rJ7 0050 DLG, M0 , Newport Beach. CA. Office Tel 714 833 2020. Fax 714 033 2085 ENGINEE AS. APPUED SCLEFmSTS
  • VANAoEMENT CONSULTANTS August 8,1994 Mr. Sam Chien San Onofre Nuclear Generating Station Southern California Edison Company P.O. Box 126 San Clemente, C A 92674 Dear Sam-I discovered after reading your fax of August 2,1994 that there was indeed an error in the calculation of the Cs and Te release fractions from the release fractions in categories 2 and 6 ( for Cs) and 3 and 11 (for Te), but it was in equation 3. The a equations are presented below.

We first determine the fractions of Cs in the mass of Csl and CsOH. These are-(1)

/['=h'c,+W j

a Cs ,

5fs __ ,

cs W,+W+W g Wcs+ 17.0073 g / gmol e o where eW , , W,, c W, and W,, are the average atomic weights of cesium, iodine, oxy hydrogen in the material released.. Then define the fraction of Cs in each of th release categories:

ff'm, (3) y2 = fl'm, +fl'm, (4) 9 = 1 - 4, i

in MAAP categories 2 and 6 respectively, and where m and m, are the total mas 2

and CsOH available for release. The Cs release fraction is then (5) t c, = p3t , + p,t ,,

A- up

~

co# coo cw vcsas.Las oni e,d osco 20s.toct cc:nt - .

m e4 3 ,

PLo,Inc.,731$ Wisconsin Avsnua. Sv'te 020 Ea:t, Ostheads. Miryt:no 20a14 3200 Tel,301-907 9100 e Fax 301907-0000 PLG. Inc., Newport Beacn CA, Off.co Tot.114 833 2020 e Fda 714 633 2063 EgW APP., LED SCIENTISTS.

MENT CONSULTANTS August 25,1994

' Mr. Sam Chien San Onofre Nuclear Generating Station Southern California Edison Company P.O. Box 126 San Clemente, CA 92674

Dear Sam:

I received your fax this morning, and thought I should extend the equations I gave you earlier to include three components for2 the TeO -Te2-Sb group. The equations for cesi I sent you on August 8 still hold.

We first determine the fraction of Te in the mass 2 of TeO . This is:

Te.

-- g^ =

y" (1) ff = W u+ 2 Wo Wr , + 319988 where W and W o are the average atomic weights of tellurium and oxygen in the material 3

released. Then define the fraction of Tc/Sb in each of the MAAP release categories:

f{'m, ._ _

f,"m, ( ,

  • ~ f3"m,+ffm +fym io ~ nf3"m, + m,, + m n mio (3)

N'* ,~_ f(*m, + m,, +um _

m" (4) -

n = fs"m, +m,, + m a {

l in MAAP categories 3,10 and 11 respectively, and where ma , m, and m are the total 3 '

masses of TeO 2

, Sb, and Te2 available for release. The Te/Sb release fFaction is then

($) j l ,7 = p s t, + p,ols, + pu tu , i A-t 3

CO'd C00- aw VGS3H138 ONI Did 0500 206 1001 2C:01 #6.5f .Y

A 4J A a a, y where t 3, t andio t aren the release fractions for MAAP categories 3,10 and i1, respectively.

This gives the release fractions in terms of the mana released. If we want it in terms of the I activity released, we need to modify equations 2 5 as fo!!ows:

f[*m,All f{'m,All ,

A f

{ll*m, + mn ) All + m,n

' ~ (f[*m, + fymn ) All + flmaAZ o mmA2

  • ~ (fl'm, +m,,)All +mn Ay m,, All p', = (ff*m + m n) A7 3 7

+m,, Ay (5')

27, = p;l ,+ plntw + p[,4n ,

where Afl and Ag are the specific activities (in Ci/g) of the tellurium and antimony available for release. Of course, this could be extended to include not just activity, but some measure of the relative dose-effectiveness of antimony and te!!urium.

The discrepancies I found extended to the other groups as well. I don't know the cause, but the symptom I see is that some of the total relcase fractions used in the MACCS runs are not the same as the corresponding relcase fractions at the end of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the graphs you sent me last month. For instance, in the case of LLO-34, we have:

Graph at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> MACCS/ Graph Ratio Group MACCS Total 6.3E-3 1.30 I 8.2E-3 6.4E-3 -1.27 Cs 8.1E-3 1.6E-3 1.44 Te 2.3E-3 1.6E-4 1.44 Sr 2.3E-4

<1E 7 Ru 8.5E-6 1.7E-5 1.76 La 3.0E-5 8.0E-5 1.62 Ce 1.3E-4 8.6E-5 1.40 .

1.2E-4 Ba The root of this discrepancy may not be in your computer 3 rogram, It might just be that I was sent graphs from a MAAP run that was later changed._ lionetheless, you need t a consistent audit path for the calculations before you sont th.'s report out.

A-18 I' _ ___ [*"" " ma s o zos toci sciot es , s e mm

I am proceeding with the review of the remaining 6 cases I haven't checked yet, and will .

let you know any inconsistencies that appear. Please call if you have any other questions.

Very tmly yours, I

T. Edward Fenstermacher - l

~I i

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q__ osoo-zos toct ge ot ,g,.3, 3 3

4 #

G PLG. Inc. 7315 Wisconsin Av$nua. Suita 620 East. Bzthmsd2. Miryttnd 20814-3209 Tel. 301907-9100 + Fax 301907-0050 PLG. Inc.. Newport Beach. CA. office EgggfygfpfEcg psW e 14 s -2020. Fax 714-833-2085 August 25,1994 Mr. Tom Hook San Onofre Nuclear Generating Station Southern California Edison Company P.O. Box 126 i San Clemente, CA 92674 l

Dear Mr. Hook:

I have completed my review of the reports "PRA Evaluation of Population Dose Risk from Severe Accidents at San Onofre Nuclear Generating Station Units 2 and 3" (referred to herein as the Base Case Report) and "PRA Evaluation of Risk Impact of Proposed One-Time Exemption form the Requirement of 10CFR50, Appendix J for ILRT Testing at San Onofre Nuclear Generating Station Units 2 and 3" (referred to herein as the ILRT Report).

I found the methodology employed in both reports to be sound. I have sent you minor corrections to several of the source terms employed in the Base Case Report, based on a comparison of the integral release fraction of each isotope group as a function of time implied by the release fractions, the start time and the duration of each phase with the graphical data sent to me by Sam Chien from July 22,1994 to August 2,1994. A corrected set of data is enclosed with this report. Subject to the revision of the Base Case Report to include these results, the resulting mean population dose should be correct within the type of uncertainty limits characteristic of consequence analysis methodology.

Since no corresponding curves were available to me for the MAAP results for the ILRT report, I was unable to check the input for this report in detail. However, a careful internal consistency check, along with a comparison of the ILRT Report source terms to the Base Case Report source terms, should reveal any problems which may be present.

IfI can be of any further assistance to you, please feel free to call.

Very truly yours, T. Edward Fenstermacher Enclosure (AVMLA6tE uPuU PEQJM l

A - 2. o

, u .

NSG/PRA REPORT PRA 2/3-94-012 APPENDIX B MACCS Output File Listings (AVAILABLE UPON REQUEST)

B-1

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