ML20117C942

Forwards Proposed Radiological Source Term Input for Facility.Encl Amended to Reflect NRC Suggested Changes

ML20117C942
Person / Time
Site:	Millstone, 05000000
Issue date:	04/03/1984
From:	Pratt W BROOKHAVEN NATIONAL LABORATORY
To:	Barrett R Office of Nuclear Reactor Regulation
Shared Package
ML20117C836	List: ML20117C942
References
CON-FIN-A-3748, FOIA-84-243 NUDOCS 8505090600
Download: ML20117C942 (30)

Text

.. . __ l:j:03yyL 2 BROOKHAVEN NATIONAL LABORATORY vr"1? , ASSOCIATED UNIVERSITIES, INC. Upton. Long Island. New York 11973 Department of NuctectEnergy ) 2630 April 3,1984 Dr. R. Barrett

  ,            Reactor Systems Branch Division of Systems Integration U.S. Nuclear Regulatory Commission Mail Stop P-1132 7920 Norfolk Ave.

Bethesda, Maryland 20814

Dear Rich,

Please find enclosed (Enclosure 1) proposed radiological source-tern; in-put for the Millstone-3 DES. The enclosure has been amended to reflect NRC- - suggested changes. If you have any questions on the report, please don't hesitate to call me or Monsen.

-l                                                      Very truly yours, N       .

W. Trevor Pratt, Group Leader Accident Analysis Group WTP:tr-Encl. cc: W. Y. Kato (w/ enclosure) R. A. Bari ,"

                   % E. Hall                  ."

M ,

                                        ,     a R. Pall                 ,

P. Easley ,,

                                                   \

y hV 8505090600 PDR FOIA 841119 N' SHOLLYB4-243 PDR Q[h~f4-2T3 I p f M e+1 4# AW .

s Enclosure 1 Containment Failure Matrix and Radiological Source Term for tne Millstone-3 DES Outline i. t I. Introduction II. Description of Plant Damage State III. Containment Failure Probabilities (C-Matrix) IV. Source Term Probabilities V. Radiological Source Term V.1 MPSS Release Fractions V.2 Discrete Probability Distributions Used in tne MPSS V.3 Suggested Source Terms for Input to Millstone-3 DES VI. Further Work

 .           VII. References i.

I 4/3/84 l l 1

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c --- . . , -- , - - , --, ---- , - . . ., T - ,, ,, -,,- - - , - - -

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I. Introduction- , l The Draft' Environmental Statement (DES) for the Millstone Unit 3 will include a severe accident risk estimate based on site consequence analyses  ! performed by the Accident Evaluation Brancn ( AEB). As input to these calcula-- tions, the Reactor Systems Branch (RSB) and the Containment . Systems Branch (CSB) are providing AEB with an estimate of the conditional probabilities of various potential containment building failure modes (C-matrix). The radio-logical source term is being specified jointly by RSB and AEB. The Reliabil-ity and Risk Assessment Branch (RRAB) will eventually provide the plant damage

    ;            state probabilities based on a review of Millstone-3 Probabilistic Safety Study. (MPSS)[13 by RRAB staff and contractors at Lawrence Livermore National Laboratory (LLNL).[2] The data presented herein is preliminary and may be changed prior to final input to the DES and refers only to internally ini-tiated events.

The data presented in this enclosure are based largely on the MPSS, which

has been reviewed by RSB and CSB staff and contractors at.Brookhaven National i . Laboratory (Reference 3). Several adjustments to the MPSS results have been made, for reasons which will be described in the following sections.

4 II. Description of Plant Damage States I In the Millstone Probabilistic ' Safety Study (MPSS), each core melt acci-1 dent sequence is assigned to one of the twenty-seven plant damage states de-scribed in Table 1. Summation over all of the frequencies of core melt acci-dents associated with a given plant damage' state yields the annual frequency of the damage state. These frequencies, which are listed in Table 2, are preliminary and based on the LLNL[2] review. Tney are currently under re-view by RRAB staff and may be adjusted prior to input to the DES. The 4/3/84

original MPSS plant damage state frequencies are also included in Table 2 for reference. Note that in the MPSS, twenty-seven plant damage state frequencies were identified, whereas in the LLNL review, only seventeen plant damage state frequencies were given. The LLNL review eliminated twelve damage states (namely, AEC' , AE, ALC", AL, SE, S'E, SLC", SL, V2E, V2LC' , V2LC" and V2L) from further consideration because of low probability (<10-7) but also added two additional damage states (namely, S'EC and TLC). The plant damage states classify events according to three parameters; i (1) Initiating Event, namely: A, large break Loss-Of-Coolant Accidents (LOCAs) S, small break LOCAs S', incore instrument tube LOCA T, transients V2, Steam Generator Tube Rupture (SGTR) V, Interfacing Systems LOCA (2) Timing of Core Melt, namely: E, failure of Energency Core Cooling Injection (ECCI) L, failure of ECC recirculation (3) Status of Containment Heat Removal (CHR) l l _

                                 , complete loss of Containment Sprays (CS)

C', loss of recirculation CS C", loss of quench CS C, all spray systems available In the following sections the process of relating the plant states to potential containment building failure modes and fission product release characteristics is described. 4/3/84

III. Containment Failure Probabilities (C-Matrix) In the MPSS, the twenty-seven plant states identified in Taole 1 were related to potential containment building failure modes by using containment event trees. It was considered unnecessary to analyze each individual plant state because of common characteristics relative to primary system response, containment response, and source term. The primary system response cnaracter-istics were grouped using accident sequence classes (A-G in the MPSS). Acci-dent sequences were classified in the MPSS according to:

   ,                  (1) the initiating event, (2) time of onset of fuel melt, and
   ,                  (3) RCS conditions at time of vessel failure, particularly RCS pressure.

Five of the sequence classes (A-E) required furtner analysis to cnarac-4 terize the containment response. Accident classes F (interfacing system LOCA) and G (ruptured steam generator tube) bypass the containment and hence were allocated directly to an appropriate release path and fission product source te rm. Characterization of containment response for the five accident classes ( A-E) required four possible combinations of quench spray system and recircu-

   .           lation spray system operation. These quench and recirculation spray system i

l combinations are:

   !                  (1) both quench sprays and recirculation sprays on (2) both sprays off (3) quench sprays on, recirculation sprays off (4) recirculation sprays on, quench sprays off This characterization by accident sequence and containment response for five of the accident classes defines twenty distinct accident groups or 4/3/84
                                                                                 -               J

categories. Again, because of common characteristics, it was no~t considered necessary to assess all of the possibilities and hence only ten containment re-sponse classes were quantified using containment event trees in the MPSS. These containment response classes are defined in Table 3. Table 4 summarizes the containment response classes with the correspond-ing plant damage states and their associated mean frequencies as provided in the LLNL review (see Table 2). Therefore, these containment response classes can be related to the ra-diological release categories to form the containment matrix. The quantification of the MPSS containment event trees was a significant task, and it was necessary to use a computer code, ARBRE, to group the various path probabilities into the thirteen release categories [1] However, the containment matrix 'C' is a concise summary of the quantification process. Table 5 is a reproduction of the 'C' matrix for the MPSS.[13 It lists the conditional probabilities of the release categories (defined in Table 6) given the plant damage state, with the plant damage states defined earlier in Table 2. - A simplification to the C-matrix is obtained in Table 7 by disregarding all of the very low probability values (CP<10-2). This simplification is I not expected to influence the risk calculations. Table 7 indicates that tne containment classes 1 through 3 lead to inter-mediate and late overpressure failures or basement melt-through in the absence of CHR operation, with an early failure being more likely as a result of hy-drogen burn for classes 1 and 3. Furthermore, the containment response classes 4, 5, 7, 8, and 10 are dominated by intermediate or late overpressure failures without full CHR operation, with basemat penetration being less 4/3/84

likely. However, successful operation of containment recirculation spray sys-tem leads to basemat failure for class 9 states. It should also be noted that the most probable sequence (class 6) leads to the lowest failure probability. IV. Source Term Probabilities In Table 7, conditional probabilities for the various release categories given a containment response class were assigned. In order to determine the frequency of occurrence of the source terms summarized in Table 8, the con-tainment class frequencies listed in Table 4 must be multiplied by the condi. tional probability of the containment failure modes given in Table 7. These i frequencies are included in the source term characterization in Section V. V. Radiological Source Term V.1 MPSS Release Fractions For most of the release categories, the applicant's evaluation of radio-nuclide release fractions was based on CORRAL-II calculations. - For a few re-lease categories, the release fractions were taken directly from WASH-1400. These two approaches are consistent insofar as both account for the same mecn-j anisms of fission product release, transport, and deposition. Three. compo-nents of release from the core were included: gap release, core melt release, and vaporization release. Radionuclide attenuation due to deposition on con-4 tainment surfaces, gravitational settling and washout by containment sprays

  !       was calculated.      ,

4 Because of uncertainties in the chemical form of iodine, two sets of re-lease fractions were calculated; one characteristic of gaseous elemental iodine and one representative of Csl aerosol. The latter source term was used for all calculations in the MPSS. The principal difference between the two options is that tne aerosol model yields significantly higher iodine releases 4/3/84

l i for release categories M-5, M-6, and M-7; the intermediate and late overpres-surization failure modes without sprays. Because M-7 is the most likely mode of failure, these differences could be important. Since the publication of WASH-1400, it has become apparent that iodine will have a strong tendency to form Cesium Iodide and subsequently adhere to aerosols. However, the Accident Source Term Program Office (ASTP0) is cur-rently in the process of assessing this question, as well as numerous otner issues related to the source term. Until these results can be quantified and i submitted to peer review, the agency will continue to base licensing decisions on WASH-1400 methodology. Consequently, we have used the release fractions characteristic of elemental iodine (Table 9) as the starting point for our review. Comparisons of Table 9 with other studies performed with WASH-1400 metn-odology have led us to conclude that the iodine releases in M-5, M-6, and- M-7 are too low. In references [4] and [5], the iodine releases for late over-pressure failure at Indian Point were an order of magnitude higher than.the i MPSS results (Table 10). The releases of all other radionuclides were of com-parable magnitude. We have used the higher iodine release fractions for this DES input (refer to Section V.3). V.2 Discrete Probability Distributions (DPD) Used in the MPSS The release fractions in Table 9 do not reflect all mechanisms of source term attenuation. Retention of fission products in the primary system was not credited. Furthermore, the enhancement of gravitational settling in contain-ment due to aerosol agglomeration was not included. To account for these fac-tors and their associated uncertainties, the applicant employed the method of discrete probability distrioutions. In this method, the actual release 4/3/84 1 l

fractions for a given release category can assume values which are a fraction

                '(F) of the values given in Table 9. The allowed fractions are 1,1/2.1/4, 1/10 and 1/100. A prob' ability (P) is associated with each F, and the proba-bilities are different for each release category (Table 11).- For example, in-a failure to isolate containment (M-4), there is an assumed 40% probability

, that F is equal to unity, and a 60% probability that F is 1/2. This small reduction in fission product release reflects an assumed retention of fission

                . products in the primary system, but very little effect of agglomeration. For' late failure without sprays (M-7), agglomeration is assumed to play a signifi-cant role, and the source term is reduced by a factor of 1/10 to 1/100. The values of F and P are based largely on engineering judgment. In all cases, the discrete probability distributions lead to a reduction in the radiological source term.

We have examined the DPD metnodology and concluded that it should not be factored into the release fractions used for the DES. Fission product reten-tion in the primary system and aerosol agglomeration in containment are credi-ble mechanisms for fission product attenuation, and are currently under study

      ;          by the Accident Source Term Program Office (ASTP0). Until the ASTP0 evalua-I          tion of tne existence and magnitude of these mechanisms is complete, we will not have a sound basis for quantifying the reduction in the source term. We recognize that the decision not to factor in the DPD's represents a conserva-tive approach to the source term.                    ,

V.3 Suggested Source Terms for Input-to Millstone-3 DES In this section, the approach utilized to determine the fraction of fis-sion products originally in the core and leaked to the outside environment will be outlined. The fission product source to the environment, as calcula- a ted by this approach, will be compared with those for similar plants. The 4/3/84 l

t-calculations to be included in this comparison are those done for the Zion and Indian Point Probabilistic Risk Assessments, (ZPSS[6] and IPPSS,[5] re-spectively), and the Indian Point Study (IPS) carried out for the NRC and pre-sented as testimony [43 at the Indian Point hearings. These calculations are

    ,"'     based on the methods used in the Reactor Safety Study (RSS), which was pub-lished as WASH-1400.[73 In the RSS, the CORRAL-II code was the mathematical model used to deter-mine fission product leakage to the environment. This code takes input from l     the thermal-hydraulic analysis carried out for the containment atmosphere.        In addition, it needs the time dependent emission of fission products. The fis-sion product release is divided up into three phases, namely, Gap, Melt, and Vaporization releases. The time dependence of these phases is determined by the core heatup, primary system failure and core / concrete interactions. In all, thirteen releases were determined in the MPSS using these methods ranging from the containment bypass sequence (V-sequence) to the no fail sequence.

The results are shown on Table 9. l Some of the thirteen MPSS releases outlined in Table 9, namely M-1A (PWR-2), M-10 (PWR-6), and M-11 (PWR-7) are identical in both fractional re-lease and timing to equivalent PWR rele'ases in the RSS. The release M-1B, which corresponds to a steam generator tube rupture, is determined by dividing

     ]      PWR-2 or M-1A by ten. Noble gases and organic iodine are not subject to this
      ;     reduction in release.

There are two areas of significant disagreement between the MPSS and the staff review. These are the iodine release for the overpressurization failure sequences (M-5, M-7) and the energy of release for these sequences. It is felt tnat the fraction of iodine released to the enviror.nent snould be 4/3/84

4 increased from .015 to .1 for these sequences. This recommendation is based t on a comparison between the MPSS results and those determined in the IPPSS and IPS. Shown in Table -10 are tne fractions of fission products released for the l I M-5 and M-7 sequences comparec with similar sequences in the IPPSS and IPS. I From an inspection of Table 10, it can be seen that the release fractions for all the species agree well, except for iodine. l 1 The energy of release for the overpressurization failures are high com-pared to those used in the RSS, IPPSS, and IPS. In fact, the values are more j characteristic of the values used for a steam explosion failure mode in the j RSS. The effect of a high energy of release on the plume is to lift it nigher into the atmosphere and tnus spread it over a larger area. By comparing the NPSS values with those used in the above studies, it is felt that the energy of release should be reduced to 150x108 Stu/hr. This value is higher than the values used in~ the IPPSS and IPS, however, it is felt to be a reasonable ! value for the overpressurization failure mode. I In Tables 12-15, release characterizations for the dominant sequences are l shown.

    .l L                   Table 12 shows release fractions and timing for two containment bypass

; sequences; the first' being an interfacing loss of coolant accident (Event V) and the second representing a steam generator tube failure (Event V2).

g Shown on Table 13 are release fractions for overpressure failures of the containment during various time frames ranging from 4.3 hrs to 20.1 hrs. No I . spray operation is assumed during these sequences. Tables 14 and 15 show release fractions and timing for casemat penetra-tion and no containment failure, respectively. Table 16 shows the release fraction to be used for a steam explosion initiated failure mode. This i

                                                                                                -g-4/3/84

s ~. release fraction and timing are based on the WASH-1400 PWR-1 release. The frequency of a steam explosion release was assumed to be 10-4 of the total core melt frequency, whicn is consistent with previous DES analyses (e.g., Limerick) . VI. Further Work The source terms given in Tables 12 througn 15 represent our recommended , input to the DES for Millstone-3 at this time. The source terms are based, in large part, on the MPSS and on a rather limited review of the MPSS by the NRC l staff and contractors. However, the Millstone-3 DES has been postponed and - thus provides additional time to refine our source term estimates. In this section, we indicate those areas in which our Millstone-3 source term esti-mates will receive further investigation. The results of these investigations will be factored into our final report. External Events - The present assessment is limited to the internal initiating events; however, the containment response to accidents l l initiated by external initiating events (fires, floods, and seismic j events) must also be reviewed.

     ;            Hydrogen - In the MPSS for accidental sequences without CHR, the condi-tional probability of an intermediate failure (M6) from a H 2 burn relative to a late failure (M7) due to overpressurization, varies sig-l i           nificantly depending on the initiator (LOCA vs. Transient).                                          If the accident sequence is initiated by a large break LOCA, then the condi-tional probability of a H2 burn failure mode is 0.62 compared with 0.06 for a small break LOCA, and negligible probability for sequences initiated by transients.                  In tne IPPSS, ZPSS and IPS, no such distinction was made for these accident sequences. We therefore will determine if we 4/3/84

can support the conditional probabilities of a H 2-burn failure for con-tainment classes 1-4. Containment Failure Distribution - In MPSS-3, the containment failure probability distribution has been calculated. This failure distribution will be carefully evaluated. Debris Quenching - The quantitative significance of dabris quenching in the reactor cavity will be examined. Elemental ~ Iodine - The acceptability of the relatively low release fraction of elemental iodine for sequences M-5, M-7, and M-9 compared to releases for similar sequences determined by other investigators (IPPSS) will be determined.

   ;               Energy of Release - The higher energy of release for the overpressuri-zation failures, compared to energy releases for similar failure modes determined by other analysts (IPPSS) will be examined.

Warning Time - For sequence M-6, the release time is 4.3 nrs and the 1 warning time is 4.1 hrs. This timing implies that the operating staff j responds quite rapidly to the accident. The feasibility of sucb a rapid response and its acceptability for use in the MPSS-3 will be investi-gated. The LLNL review also introduced two new plant states (namely, S'EC and TLC), which we binned into containment class 6. We will confirm that this is an appropriate containment class for these sequences. In addition, the difference in response for TEC' and SEC' will be resolved. 4/3/84

VII. References (1) " Millstone Unit 3 Probabilistic Safety Study," Northeast Utilities, August 1983. (2) "A Review of the Millstone-3 Probabilistic Safety Study," Incomplete Preliminary Draft, January 25, 1984. (3) M. Khatib-Rahbar, H. Ludewig, and W. T. Pratt, " Preliminary Review and Evaluation of tne Millstone-3 Probabilistic Safety Study," Brookhaven National Laboratory, Informal Report, December 1983. (4) Direct Testimony of J. F. Meyer and W. T. Pratt concerning Commission Question 1, Indian Point Hearings, Docket Numbers 50-247 and 50-286, 1983. (5) " Indian Point Probabilistic Safety Study," Power Authority of the State of New York and Consolidated Edison Co., March 1982. (6) " Zion Probabilistic Safety Study," Commonwealth Edison Company, September 1981. (7) Reactor Safety Study, "An Assessment of Accident R!sks in U.S. Commercial

 !             Nuclear Power Plants", WASH-1400, NUREG/75-014, October 1975. -

i 4/3/84

Table 1 Notation and definitions for plant states (internal) Symbol Description AEC Large LOCA, Early Melt AEC' Large LOCA, Early Melt, Failure of Recirculation Spray AE Large LOCA, Early Melt, No Containment Cooling ALC Large LOCA, Late Melt ALC' Large LOCA, Late Melt, Failure cf Recirculation Spray ALC" Large LOCA, Late Melt, Failure of quencn Spray Al Large LOCA, Late Melt, No Containment Cooling s SEC Small LOCA, Early Melt SEC' Small LOCA, Early Melt, Failure of Recirculation Spray SE Small LOCA Early Melt, No Containment Cooling S'E Incore Instrument Tube LOCA, Early Melt, No Containment Cooling SLC Small LOCA, Late Melt SLC' Small' LOCA, Late Melt, Failure of Recirculation Spray SLC" Small LOCA, 'te Melt, Failure of Quench Spray SL Small LOCA, Late Melt, No Containment Cooling

    -f         S'l             Incore Instrument Tube LOCA, Late Melt, No Containment
     !                        Cooling
     !         TEC            Transient, Early Melt TEC'           Transient, Early Melt, Failure of Recirculation Spray TE             Transient, Early Melt, No Containment-Cooling V2EC           Steam Generator Tube Rupture, Steam Leak, Early Melt V2EC'          SGTR, Steam Leak, Early Melt, Failure of Recirculation Spray V2E            SGTR, Steam Leak, Early Melt, No Containment Cooling V2LC           SGTR, Steam Leak, Late Melt V2LC'          SGTR, Steam Leak, Late Melt, Failure of Recirculation Spray V2LC"          SGTR, Steam Leak, Late Melt, Failure of Quench Spray

., V2L SGTR, Steam Leak, Late Melt No Containment Cooling I y Interfacing Systems LOCA 4/3/84

               .~  ..

Table 2 Plant damage state frequencies for internal events (perreactor-year) MPSS Provided Synbol (Mean) by RRAB AEC 1.92E-06 8E-7 AEC' 4.17E-09 AE 2.68E-09 ALC 5.44E-06 2E-6 i ALC' 4.88E-7 IE-7 3.42E-09

• ALC"

    .        AL                           3.36E-10 i        SEC                          1.12E-06                           2E-5 i        SEC'                         2.76E-09                           6E-7 SE                           1.17E-07 i        S'EC                             -                              4E-7 S'E                          1.83E-09 SLC                          9.81E-06                           IE-4 SLC'                         4.79E-07                           1E-5 SLC"                         5.77E-08 SL                           2.73E-09 S' L                         3.35E-10                           1E-7 TEC                          1.81E-05                           4E-5 TEC'                         3.46E-07                           2E-7 i        TE                           5.31E-06                           7E-6
    }        TLC                              -                              4E-5 V2EC                         1.11E-07                           4E-6 V2EC'                        1.03E-09                           3E-7 V2E                          1.29E-08 V2LC                         2.76E-09                           2E-7 V2LC'                        1.49E-10
   .j

• V2LC" 1.77E-11 i V2L 8.40E-13

    !        V                            1.90E-06                           4E-7 i

i

    '                              TOTAL  4.53E-05                           2.3E-4

• Indicates frequency values <10-7, 4/3/84

Table 3 Containment response classes Dominant Class Sequence Reference Definitions 1 AE Initiating event is typically a large break LOCA without safety injection and without minimum con-tainment safeguards operating throughout the transient. 2 SE Same as the AE sequence except that the initiating event is typically a small break LOCA or transient 3 event. Note that the containment sprays do not

   -l                                       operate.
     !              3         AL            Same as the AE sequence except that safety injec-
    ]i                                      tion is initiated but operate only until switch-over to recirculation is attempted, at wnich time it becomes inoperative for the remainder of the transient.

4 TE The initiating event is typically a transient in

which all power is lost. There would therefore be i

no safety injection and no containment safeguards initiation at any time during the transient. 5 SL Same as the Al sequence except that the initiating event is typically a small break LOCA or transient i event. Note that the containment sprays are ac.

     !                                      tuated but do not deliver water to the spray headers.

6 TEC Same as the TE sequence except that all contain-ment heat removal systems are availaDie. 7 TEC' Same as TE sequence (Class 4) except that AC power is available and containment quencn spray system is functioning. 8 SEC' Same as SE sequence (Class 2) except that contain-

     ,                                      ment quench spray system is functioning.

9 TEC" Same as TE sequence (Class 4) except that AC power is available and recirculation spray system is functional. 10 S'l Same as SL sequence (Class 5) except that rupture is as incore instrumentation tube rupture. 4/3/84

Table 4 Containment class mean frequencies for internal events (per reactor year) Containment Class Plant Damage States Mean Frequency (yr-1) AE 1 SE 2 3 AL 4 TE 7.0E-6 SL

• 5 6 AEC, ALC, SEC, SLC, 2.03E-4 SEC, TEC, TLC, S'EC 7 TEC', SLC' 1.02E-5 8 AEC', ALC', SEC' 7.0E-7 AEC", ALC", SEC",
• 9 SLC", TEC" -

i 10 S'E, S'L 1.0E-7 V2EC, V2EC', V2E, 4.5E-6 V2LC, V2LC', V2LC",

  !                                V2L j                                           V                      4.0E-7

• Indicates frequency value less than 10-7, 4/3/84

Table 5 Reproduced from MPSS Table 4.7.2-2 4 1 4 i 4/3/84

Table 6 Notation and definitions for release categories i Release Category Description M1A Containment Bypass, V-Sequence M1B Containment Bypass, SGTR

     ;.                  M2                      Early Failure /Early Melt, No Sprays l                   M3                      Early Failure / Late Melt, No Sprays M4                      Containment Isolation Failure MS                      Intermediate Failure / Late Melt, No Sprays M6                      Intermediate Failure /Early Melt, No Sprays l                  M7                      Late Failure, No Sprays M8                      Intermediate Failure With Sprays M9                      Late Failure With Sprays
       '                 M10                     Basemat Failure, No Sprays M11                     Basemat Failure With Sprays M12                     No Containment Failure 2

4/3/84

                    ..   ..- -.            . . . . . . - - -      . - - . - .           a.      . - . - . . . . . . . .   . . w .. . . L : -..    . - .. . - , . . -

j-Table 7 Simplified containment matrix for MPSS i . Containment Response Class M1A M1B MS M6 M7 M10 Mll M12 i 1 0.62 0.29 0.09' [ 2 0.06 0.89 0.05 3 0.54 0.35 0.11

        .      4                                                                           0.90               0.10

) 5 0.01 0.79 0.20 . h 6 0.05 0.95 7 1.0 4 8 1.0 9 0.99 0.01 10 0.99 0.01 l V 1.0 V2 1.0 R

! 2 l

Table 8 Source term frequencies Containment Frequency Response Class M1A MlB MS M6 M7 M10 Mll M12 (yr-1) 1 * * *

• 2 * * * *

             -3                                                      *      *            *

• 4 6.3E-6 7.0E-7 7.0E-6 5 * * *
• 6 4

O 1.01E-5 1.93E-4 2.04E-4 7 1.02E-5 1.02E-5 8 7.0E-7 7.0E-7 9 * * *

• 10 * *
• V 4.0E-7 4.0E-7 V2 4.5E-6 4.5E-6 Release (yr-1)

Frequency 4.0E-7 4.5E-6 * ' 1.72E-5 7.0E-7 1.01E-5 1.93E-4 K M

 $       .

• Indicates frequency value less-than 10-7

1 Table 9 - Reproduced from MPSS i f t 1

 'I I

t I i I I i 1 1 4/3/84

e Table 10 Intermediate and late overpressurization j (no sprays) Sequence MPSS MPSS IPS[4] IPPSS[5] M-5 M-7 TMLB'- 6 2RW* f Xe-Kr .9 .9 .96 1.0 I 0I+1 .016 .015 1.05(-1) 9.3(-2) i Cs-Rb .5 .3 .34 .26 i Te-Sb .5 .3 .38 .44 Ba-Sr 5(-2) 3(-2) 3.7(-2) 2.5(-2) Ru 4(-2) 2(-2) 2.9(-2) 2.9(-2) La 6(-3) 4(-3) 4.9(-3)

   ,                                                                         1.0(-2) i I

i 4/3/84

Table 11 MPSS release category DPDs l 4 Discrete Probability Distributions Release Category F* 1 1/2 1/4 1/10 1/100 M-1A 0.17** 0.55 0.28 0 0 M-2 0.25 0 0.25 0.50 0 l M-3 0.0 0 0.06 0.63 0.31 4 M-4 0.40 0.60 0 0 0 M-5 0.0 0.0 0.05 0.64 0.31 M-6 0.11 0.14 0.27 0.48 0 M-7 0 0 0 0.11 0.89

    ]!

• Release Fraction (F) -

                    ** Probability Values (P) i

^ t 4 4 I e 4/3/84

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

Table 12 containment bypass sequences

 ;           Failure Mode and Release Paths                    M-1A               M-1B Xe-Kr                       9(-1)              9(-1)
  ;                 I+0!                       7.07(-1)           7.07(-1)

Cs-Rb 5(-1) 5(-2)

Te-Sb 3(-1) 3(-2) i

  ,                 Ba-Sr                      6(-2)              6(-3)

Ru 2(-2) 2(-3) La 4(-3) 4(-4) Release Time (br) 2.5 2.5

    ,          Warning Time (hr)               1.0                1.0 j           Duration (br)                   1.0                1.0 Energy (106 Btu /hr)            20.0               20.0 I           Probability                    4E-7               4.5E-6 0

4/3/84

Table 13 Intermediate and late overpressurization failure M-5 M-6 M-7 Xe-Kr 9(-1) 9(-1) 9(-1) i OI+I 0.1 0.1 0.1 Cs-Rb 5(-1) 5(-1) 3(-1) Te-Sb 5(-1) 5(-1) 3(-1)

   . .i Ba-Sr                                  5(-2)            5(-2)                                   3(-2)
      ?
    ~ !,                         Ru                                     4(-2)            4(-2)                                   2(-2)

La 6(-3) 7(-3) 4(-3) . Release Time 8.3 4.3 20.1 (br) Warning Time 4.1 4.1 16.0 (br) Duration (hr) 0.5 0.5 0.5 Energy 150 150 150 l (106 Btu /hr) i Probability <10-7 <10-7 1.72x10-5 4/3/84

1 Table 14 Basemat penetration Failure Mode and Release Paths M-10 M-11 Xe-Kr 3(-1) 6(-3) 1+0I 2.8(-3) 4(-5) Cs-Rb 8(-4) 1(-5)

 !.                               Te-Sb                                1(-3)                       2(-5)

Ba-Sr 9(-5) 1(-6) {' Ru 7(-5) 1(-6) La 1(-5) 2(-7) Release Time (br) 95 95 i Warning Time (nr) 80 80 Duration (br) 10 10 Energy (106 Btu /hr) - - Probability 7x10-7 1.01x10-5 4/3/84

         . _ .   . _ _ . _ ..         r- _   ____.___...._ _ ____.            __. _ . . _ _ _ _.                 _

                    ' ' ~

i~ Table 15 No containment failure L Failure Mode and Release Path M-12 j. .  ; Xe-Kr 1(-3) OI+I 1.5(-5) Cs-Rb 1(-6) Te-Sb 9(-7) l Ba-Sr 2(-7) Ru 8(-8) La 1(-8) i t Release Time (hr) .5 Warning Time (hr) - Duration (hr) 5.0 Energy (106 Btu /hr) - i i l Probability 1.93x10-4 i i i 4/3/84

  '             o                                _.

_..___..o i Table 16 Steam explosion failure mode Failure Mode and Release Path PWR 1 i Xe-Kr 9(-1)

    ~ '

OI+I 7(-1) l Cs-Rb 4(-1) { Te-Sb 4(-1) Ba-Sr 5(-2)

        !                                Ru                                                    4(-1)

La 3(-3) j Release Time (br) 2.5 i Warning Time (br) 1.0 Duration (nr) 0.5 Energy (106 Btu /hr) 520 Probability 2x10-8 l t 4/3/84

                                                                      . -.- ._ . - - - - - _ _               . _ .   ._ - -.}}