ML20127B689
Text
{{#Wiki_filter:' Distribution 'E {
- Docket File
$4 RSB R/F RSB P/F: Millstone 3 RBarrett R/F RBarrett MAY 1 4 1994 JRosenthal BSheron FIEMORANDUM FOR: Lewis G. Hulman, Chief Accident Evaluation Branch, DSI ( M FROM: Brian W. Sheron, Chief Reactor Systems Branch, DSI
SUBJECT:
MILLSTONE 3 DES; CONTAINMENT FAILURE MATRIX AND SOURCE TEPJ!S On April 5,.1984, we sent you a preliminary version of the containment failure matrix (C-Matrix) and source terms for use in calculating severe accident risk estimates for the Millstone 3 Draft Environmental Statement. The enclosed final input differs from the preliminary results in the following ways: A separate C-Matrix for external events is included, The release times and warning times for radiological releases have been redefined and are listed according to accident sequence. ~ The release factors for the V sequence have been recalculated. All elements of the C-Matrix have been included, not just those with values greater than 0.01 The Iodine release fractions for M-5, M-6 and M-7 release categories have been reduced to the values quoted in the Millstone 3 probabilistic safety study. For internally initiated events, we have also included the annual frequencies for each release category, based on our C-Matrix and the plant dacage state frequencies provided in the April 25, 1984 me::orandum from RRAB. When the plant damage state frequencies for external events become available in mid-May, we will assist you in calculating the corresponding release category frequencies. g,,,, w;.,, g,, w ac.
- . ;." c. ::. : -
Brian W. Sheron, Chief Reactor Systems Branch, DSI gef 8 e 8
- PREVIOUS CON RRE C SHEET ON FILE W/RSB.
ex Pa Barrett, RSK [$L10 531 o LW xD CONTACT:
- MT:TS B-
- OSI
..Q.8 "T** RBarrett,,jf ldRosenthal ,,,l BSh,\\,, ,,[,, ..t. . 3,, ,, j,, j,,,, n ,,) 05//,f/,84 ,,j,,, ,,{ 7"* 05/p,9/84 05/10/84 eee.* n t ase: acc eno OF71CIAL RECORD COPY.
P i ( MAY 14 1lB4 "~ L. Hulman, Chtf 5 -- cc: R. Mattson R. Houston e D. Muller P. Easley J. Read B. Hardin J. Mitchell A. Thadani G. Kelly E. Doolittle n,: t p [............I i' a*.*th,............*............. PATE)I,......................... t
- a oc'.t :n oo toi N mcr/. c240 mppar Al p g rn p rg rn py.
e
o.; -- 4, c E Containment Failure Matrix and Radiological Source Term for tne 14111 stone-3 DES i l Outline 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 the MPSS V.3 Suggested Source Terms for Input to riillstone-3 DES 'VI. References l b e w i e 5/10/ 84 n L
- i. -..
g C-I. Introduction The Draf t Environmental.itatement (DES) for the Hillstone U include a severe accident risk estimate based on site consequence analy performed by the Accident Evaluation Branch (AEB). As inpu* to these calcula-tions, the Reactor Systems Branch (RSB) and the Containment Systems Bran (CSB) are providing AEB with an estimate of the.conditionaT 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) has provided the plant damage state frequencies for internal events based on a review of -the Millstone-3 Proba-bilistic Safety Study (MPSS)E13 by RRAB staff and contractors at Lawrence Livermore National Laboratory (LLNL).[2] The corresponding frequencies for external events will be provided by RRAB in the middle of May. The data presented in this enclosure are based largely on the NPSS, which has been reviewed by RSB and CSB staff and contractors at Brookhaven National Laboratory (Reference 3). Several adjustments to the MPSS results have been made, for reasons which will be described in the following sections. II. Description of Plant Damage States In the Millstone Probabilistic Safety Study (MPSS), each core melt ac- -cident sequence is assigned to one of the plant damage states described in Sumation over all of the frequencies of core melt accidents Tables 1 and 2. associated with a given plant damage state yields the annual frequency of the 2 damage state listed in Tables 3 and 4. For internal
- events the original MPSS
. plant damage state frequencies are also included in Table 3 for reference Note that in the MPSS, twenty-seven plant damage state fregmcies were ida. ~ tified, whereas in the LLNL review, only seventeen plant damage state fre- ,i quencies were given. The LLNL review eliminated twelve damage states (namely,
- kefers to all initiating events except seismic events.
-1 + ' ' ~,. 5'nb'M o
} :,, '... - (, AEC', AE, ALC".. AL, SEC', S'E, SLC", SL, V2E, V2LC', V2LC" and V2L) from further consideration because of low probability ((10-7) but also added additional damage states, namely, S'EC and TE with and without the AMSAC. The plant damage states classify events according to three parameters; (1) Initiating Event, namely: A, lar'ge break Loss-Of-Coolant Accidents (LOCAs) 5, small break LOCAs S*, incore instrument tube LOCA T, transients V2, Steam Generator Tube Rupture (SGTR) V3, Seismic induced AE combined with containment bypass V, Interfacing systems LOCA (2) Timing of Core Melt, namely: E, failure of Emergency Core Cooling Injection (ECCI) L, failure of ECC recirculation '(3) Status of Containment Heat Removal (CHR) , 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 damage states to potential containment building failure. modes and fission product release g characteristics is described. III. Containment Failure Probabilities (C-Matrix) In the MPSS, the plant states identified in Tables 1 and 2 were related to potential containment building failure modes by tsing containment ev.ent c op e 2. 5/1 0/ 84 .~. :. -
- r. r e
(. r o t. o It was considered unnecessary to analyze eacn individual plant state trees. because of comm'on. characteristics relative to primary' system response, con-tainment response, and source term. The primary system response characteris-tics were grouped using accident sequence classes ( A-G in the MPSS). Acci-cent 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 further analysis to charac-terize the containment response. Accident classes F (interfacing system LOCA) and G (ruptured steam generator tube) bypass the containment and nence were allocated directly to an appropriate release path and fission product source term. Characterization of containment response for the five accident classes (A-E) required four possible combinations of quench spray system and recircu-lation spray systein operation. These quench and recirculation spray system 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 (for internal. events) by accident sequence and con-tainment response for f.ive uf the accident classes defines twenty distinct ac-cident groups or categories. Again, because of common characteristics, it was not considered necessary to assess all of the possibilities and hence only ten -3 5/10/84
s. e q containment response classes were quantified using containment event trees in the MPSS. These containment, response classes are (efined in Table 5. Tables 6 and 7 summarize the containment response classes with the cor-responding plant damage states and their associated mean frequencies for in- _ternal and external events, respectively. Therefore, these containment re-sponse classes can be related to the radiological release categories to form the containment matrices for both internal and external events. 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 C-matrix is a concise summary of the quantification process. Tables 8 and 9 are the C-matrices for internal and external events, respectively.3, They list the conditional probabilities of the release categories (definec in Table 10) given the plant damage state, with the plant damage states definea earlier in Tables 3 and 4. A steam explosion release category M2B has been added for both internal and external events. Simplifications to the C-matrices are ob-tained in Tables 11 and 12 by disregarding all of the very low probability values (CP<10-2). This simplification aids 'n identification of the domi-nant sequences and their corresponding release categories. Tables 11 and 12 indicate that the containment classes 1 through 3 lead to intermediate and late overpressure failures or basement melt-througn in the absence of CHR op-eration, with an intermediate failure being more likely as a result of hydro-gen burn for classes 1 and 3. Furtnermore, the containment response classes 4, 5, 7, 8, and 10 (for internal initiators) are dominated by intermediate or late overpressure failures without full CHR operation, with basemat penetra-tion being less likely. However, successful operation of the containment y, J 5/10/84 ? -~ ~ .: : ^. .: r: v -
r.- {,. recirculation spray system (class 9) leads to basemat failure, and full spray operation (class 6) leads to the lowest failure pr6bability. Furthermore, the difference in small. break LOCA's (SE) and transient (TE) events results-from the significant delay in core melt (51/2 hours) for the TE plant damage state due to operation of the turbine-driven auxiliary feed-water pump. The probabilitie's of various release categories depend to a great extent on the probability distribution of containment failure as a function of pres-The distribution assumed in the MPSS-3 is currently being reevaluated sure. by NRC contractors, but the results are not yet available for inclusion in this report. It is not anticipated, however, that the reevaluation will have a major impact on risk estimates. IV. Source Term Prcbabilities In order to determine the frequency of occurrence of the radiological source terms for both internal and external initiators, the conditional proba-bilities for the various release categories given a containment response class summarized in Tables 8 and 9 must be multiplied by the mean frequencies of the plant damage states (see Tables 3 and 4). At present, only the source term frequencies associated witn the internal initiating events are calculated as listed in Table 13. The source term frequencies for the externally initiated events can be calculated in a similar fashion when the revised damage state frequencies become available from RRAB. 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-Il calculations. For a few i 2 5/10/84
f. c c-release categories, the release fractions were taken directly from WASH-1400. These two approaches are consistent insofar as bot 6' account for the same mech-anisms of fission product release, tra,nsport, 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-tainment surfaces, gravitational settling and washout by containment sprays we. calculated. 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 the aerosol model yields significantly higher iodine releases 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-(except for basemat melt-thrcugh), these differences could be im-portant. 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 other issues related to the source term. Until these results can be quantified and submitted to peer review, the agency will continue to base licensing decisions i on WASH-1400 methodology.[63 Consequently, we have used the release frac-tions characteristic of elemental iodine (Table 14) as the starting point for ~ our review. n 5/1 6/ 84
.4 g c. Comparisons of Table 14 with other studies performed with WASH-1400 meth-cdology[6] show discrepancies in the iodine releases for the M-5, M-6, and M-7 release categories. In references [4] and [5], the iodine releases for late overpressure failure at Indian Point were an order of magnitude higher than the MPSS results (Table 15). The releases of all other radionuclides were of comparable magnitude. We have no technical basis for adjusting the MPSS release fractions at this time. However, plant specific MARCH / CORRAL calculations will be performed at BNL to resolve these discrepancies. V.2 Discrete Probability Distributions (DPD) Used in the MPSS The release fractions in Table 14 do not reflect.all mechanisms of source term attenuation. Retention of fission products in the primary system was not credited. Furtnermore, 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 metnod of discrete probability distributions. In this method, the actual release frac-tions for a given release category can assuine values which are a fraction (F) of the values given in Table 14 Tne allowed fractions are 1, 1/2, 1/4, 1/10 and 1/100. A probability (P) is associated with each F, and the probabili-ties are different for each release category (Table 16). 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 reduc-tion in fission product release reflects an assumed retention of fission pro-ducts in the primary sys. tem, but very little ef fect 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, 4 - 5/10/84 4 m m
-+ t c-the discrete probability distributions lead to a reduction in th g e radiological source term. We have examined the UPD methodology and concluded that it sho not be factored into the release fractions used for the DES. Fission procuct reten- ~ tion in the primary system and aerosol agglomeration in contai nment are credi-ble mechanisms for fission product attenuation, and are curre tl n y under study by the Accident Source Term Progran Office (ASTP0). Until 'the ASTP0 evalua-tion of the existence and magnitude of these mechanisms is complet e, we will not have a sound basis for quantifying the reduction in the source term Me recognize that the decision not to factor in the DPD'.s represents a con tive approach to the source term. V.3 Suggested' Source Terms for Incut 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 environm will be outlined. In the RSS, the CORRAL-Il code was the mathematical model used mine fission product leakage to the environment. This code takes input from the thernal-hydraulic analysis carried out for the containment atmosph ere. In addition, it needs tne 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 m from the containment bypass sequence (V-sequence) to the no fail seque The results are shown on Table 14 e 5/10/84 Im;; .m- ... ~ ' ~
Q* Some of the irteen MPSS releases outlined in Table 14, namely M 1A '(PWR-2), M-10 (PWR-6), and M-11 (PWR-7) are ident,ical in both frac onal re-lease and timing to equivaTent PWR releases in the RSS The release M-1B, wnich corresponds to a steam generator tube rupture, is determined b y dividing PWR-2 or M-1 A by ten. Noble gases and organic iodine are not subject to this reduction in. release. There are three areas of significant disagreement between the M the staff review. These are; the release fractions for the even V, the iodine release for the overpressurization failure sequences (M 5 , M-7) and the energy and tne release timings 'for these sequences. The staff notes that the WASH-1400 PWR-2 releases us report for the release catagory M 1A were defined for sequences other th . interfacing system LOCA (event V). The Oconee-RSS-MAP [73 study produced a set of releases specifically defined for the characteristics of a V sequence Consequently, the staff has decided to substitute the Oconee release fr for the MPSS-3 values in release catagory M-1A (see Table 17) Tne energy of release for sequences M-5 through M-7 is high when com pared to similar sequences (overpressurization failure) in other PRA's The release energy is more characteristic of a steam explosion release (ot) evaluated in the RSS. This value is a function of the failure pressure and the assumed rupture size. Furthermore, it has been found that a high energy of release lofts the plume to a higher altitude than a lower release en i The overall effect of this is to disperse the plume, and thus reduce the con L centration of the dose received by the surrounding population. This reduction in the dose affects consequences which are a function of a threshold dose i.e., early fatalities. In view of the uncertainty and the possible affect on P 0 0 9 0 5/10/84 L=,=rma ~.--. - - - r i.
1- ^ (.' L early fatalitiss,. we have reduced the release -energy from a 450x106 to 150x106 BTU /hr as shown in Table 17. Due to the-significance of the release time of the early release cate-gories for acute fatalities, and also the importance of the warning time for the same categories on the effectiveness of evacuation, the release and warn-ing times quoted in the MPSS were reevaluated for each response class. Tne results of the reevaluations are summarized as follows: M1A and M1B - The MPSS release time and warning times are acceptable. M2A and M3 - The release time is calculated as the time wnen containment at-mosphere becomes flammable; otherwise, it is set to the vessel failure time for a-given containment response class. The warning time is defined as the-time after core melt starts to the time of release. M2B and M4 - The release is assumed to occur following vessel failure, and the -warning time is the time following core melt to the radiological release time. M5 M6, and M8 - The same as M2A and M3. M7, M9 through M12 - The MPSS release and warning times are acceptable since their influence on acute fatalities is negligible. Table 18 summarizes the proposed release and warning times as determined using the MPSS best-estimate accident chronology in Table 4.4.2-1 of the PRA. Table 18 clearly indicates the dependence of the radiological release characteristics on the containment response class. These ' values are equally applicable to the external initiating events. ~ ~ k e 5/1 0/ 84 ~ t
V: c C' VI. R7ferances (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 the' 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) Reactor Safety Study, "An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants", WASH-1400, NUREG/75-014, October 1975. (7) G. S. Kolb, et al " Reactor Safety Study Methodology Application Program: Oconee !3 PWR Plant", NUREG/CR-1659/2 of 4. tt p I 1: 3 5/10/ 84
- E -,
,;,g-:, i
c t'-.,', . s (; (-L Table 1 Notation and definitions for plant states (internal) Symbol Descriotion AEC Large LOCA, Early Melt AEC' Large LOCA, Early Melt, Failure of Recirculation Spray AE Large LOCA, Early Melt, No Containmerit Coolino ALC ' Large LOCA, Late Melt ALC' Large LOCA, Late Melt, Failure of Recirculation Spray ALC" Large LOCA, Late Melt, Failure of Quenen Spray Al Large LOCA, Late Melt, No Containment Coolinc, 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 S'EC Incore Instrument Tube LOCA, Early nelt SLC Small LOCA, Late Melt SLC' Small LOCA, Late Melt, Failure of Recirculation S; ray SLC" Small. LOCA, Late Melt, Failure of Quench Spray SL Small LOCA, Late Melt, No Containment Cooling S'l Incore Instrument Tube LOCA, Late Melt, No Containment Cooling TEC Transient, Early Melt TEC' Transient, Early Melt, Failure of Recirculation Spray 1 : 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 s: V2E SGTR, Steam Leak. Early Melt, No Centainment 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 Y2L SGTR, Steam Leak, Late Melt, No Containment Cooling V Interfacing Systems LOCA 9 ' ~ ?.a , 5/10/84,7 O
C C' s Table 2 Notation and definitions for plant damage states with seismic initiator (external) Symool ' Descripti on AE Large LOCA, Early Melt, No Containment Cooling SE Small LOCA, Early Melt, No Conteinment Cooling TE T"ansient, Early Melt, No Containnent Cooling A. Large LOCA, Late Melt, No Containment Cooling SL1 Consequential LOCA Due to Opening PORV's to Perforn Feec and Bleed Subsequent to a Seismic Inducee LOSP SL2 Late Core Melt Following an S Initiator AEC Large LOCA, Early Melt SEC Small LOCA, Early Melt TEC Transient, Early Melt AEC' targe LOCA, Early Melt, Failure of Recirculation Spray SEC' Small LOCA, Early Melt,. Failure of Recirculation Soray TEC' Transient, Early Melt, Failure of Recirculation Spray ALC Large LOCA, Late Melt SLIC Same as SL1'Except for Fall CHRS Operation SL2C Same as SL2 Except for Fall CHRS Operation ALC' Large LOCA, Late Melt, Failure of Recirculation Spray SLIC' Same as SLIC Except Failure of Recirculation Spray SL2C' g Same as SL2C Except Failure of Recirculation Spray V3 AE Combined with Containment Bypass g.: O e 4 l 13 5/10/84 4 . c..a n!.y. 7 :, ; O
t (- Table 3 Plant _ damage state frequencies _ for internal even s (per reactor-year) r MPSS Provice: Symbol (Mean) Dy RRA3 AEC 1.92E-05_ IE-6 AEC' 4.17E-09 = AE 2.68E-09 ALC 5.44E-06 E-6 ALC' 4.88E-7 3E-7 ALC" 3.4 2E-09 AL 3.35E-10 SEC 1.12E-05 SE-E SEC' 2.76E-09 SE 1.17E-07 2E-5 S'EC 9E-7 S'E 1.83E-09 SLC 9.81E-05 2E-5 SLC' 4.79E-07 AE-7 SLC" 5.77E-08 SL 2.73E-09 S'l 3.35E-10 2E-7 TEC 1.81E-05 -E + J E-f TEC' 3.46E-07 2E-7 TE with AMSACt 5.31E-05 4 -5 / e '7 TE without AMSAC 3E-5 V2EC 1.11E-07 4 E-6 V2EC' 1.03E-09 3E-7 V2E 1.29E-08 V2LC 2.76E-09 2E-7 V2LC' 1.49E-10 V2LC" 1.77E-11 V2L 8.40E-13 V 1.90E-05 4E-7 TOTAL 4.53E-05' 1.38E-4 With AMSAC [ ' Indicates frequency values (10-7 b i Auxiliary ATWS jjitigating System Actuation Circuirty t e V e 5/10/84 2 ., y.....,.
s r.. ( (. ' Table 4 Plant damage state frcquenci,es for external events (per reactor year) r Plant Damage State MPSS-3 h Provided by RRAS Oz,lru L-t w L. L'-- gp AEC 1.06E-9 ~ AEC' 3.24E-10 AE 1.22E-6 ALC is f 2.53E c ALC' 1.57E 7 AL 1.62E 9 SEC 2.97E-7 f, c.1 SEC' 1.0-E-7 Ae 7 SE 7.33E-6 .IE-4 SLC 1.25E-8 F d '? SLC' 3.03E-7
- 2. 6
~1 SL 5.16E-9 TEC 4.69E-10 2d-3 TEC' 6.90E-11 TE 7.80E-6 10 -' 76 -f k. V3 7.14E-8 3E-( 4 Total 1.73E-5 L e i s f Gy- = 4 I E 9 e G. 5/10/84 a
y x.-., .y. c E. n:. *
- Table 5 Containment response classes m
r 1 Don.inant Class Seouence-Re.ference 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 event. Note that the containment sprays do not operate. 3 .AL Same as the AE sequence except that safety injet-tion is initiated but operate only until swit:..- over to recirculation is attempted, at wnicn ti.:e 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 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 f event. Note that the containment sprays are ac-tuated but do not deliver water to the spray headers. 6 TEC Same es the TE sequence except that all contain-ment heat removal systems are available. 7 TEC' Same as TE sequence (Class 4) except that AC power is available and containment quench spray system is functioning. 8 SEC' Same as SE sequence ment quench spray sys(Class 2) except that contain- [: tem is functioning. 9 TEC" Same as TE sequence Class 4) except that AC power is available an(d recircu,lation spray system is functional. r [ t 10 S'l Same as SL sequence (Class 5) except that rupture is as incore instrumentati,on tube rupture. e i 1 9 5/10/84...,_..,.....
r- ' ~~ ^ ( Q- -o Table 6 Containment class mean frequencies for internal events (per reactor year) Containment Class Plant Damage States flean Frequency (yr-1) .S & O ,p 1 AE O 2.6SE-9* gg.y 2 SE 2E-5 3 AL
- 3. 3 5E-10
- J G-'
25-E 4 (6 b with AMSAC TE 3E-5 withcut AMSAC 5 SL 3 2.73E-9* 2. ~16 ' ' 6 AEC,.ALC, SEC, SLC, W TEC, S 'EC 1,1cd-4 Ed-7 7 TEC', SLC' 6.0E-7 ge-7 8 AEC', ALC', SEC', 3.07E-7* 9 ALC", SLC" 6.11E-B+ 10 S'E, S'l 2.02E-7* V2EC, V2EC', V2E, 4.51E-6* V2LC, V2LC', V2LC", V2L V 3E-6 4.0E-7 f
- Based upon 14PSS values (see Table 3).
6 4 f 6 0 di p.- -17 D .. 5/10/84,'...'
c c- ^ Table 7 Containment class nean frequencies for the external events r e Plant Containment' Class Damage States Mean Frecuency (yr-I) 1 AE 2 SE 3 AL 4 TE 5. SL 6 AEC,ALC.SEC,SLC,TEC 7 TEC',SLC' 8 AEC ', ALC '. S EC ' V3 ~ i r j. t t 4 18 5/10/84 i 2 mam .,......,.;,,.. 'u.., s - ,s.. n. . _ _, - - _ _ ~ _ _ _. - - ~-
. + e g. .e r -i e. e. -8 e e e = e e e '. 4 4 4. I. ~ - - ~ ~ m o m ~ o O..i.i e e. . ~ e. z e e o ~. s. -. m E o. e A E - o - e m sii6 i a i i c. z ~ ~ ed em .i i
- E 4
4 4 ~ s E R E e - ee .e e e e ~. m aii6 6 6 i a a h* m. a R I e E .i a6 J
- g m
m i c E e e =* 3 't E .i i.i .i i i i.i i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 5 m .i e. .i w z ~ m e ~e. 3 2 .i.i.i.i.i.i.i.i i.i a .i." i .i i i.i a e. ~. =. e. u m n u m ~ 2 T T i e 8 M M 4 e g 9 E 4 e O
- 8 g.
G em u* : . = m. e e ~ =, e. .L., V S 1 w G e B 4 g m 5/10/84 f p-f . e m.*. e' 6. ... J a. ,, ; ; *pp p, '_.s*,~
w. e Q s-m. ~_s g e e e. E s gp e e e e opt sn. S,$ N N N d $ $ NT., ~ E e. 3 w e e m~ e. e t. g - _ _E O. e. e e o e i I %a .T .T E D .es OR W D .=
- Rbe, E
e.W.8 =W' e. P ~. W.'#' = e i e.
- m. h. =9 e, g sg w
a. g O i O.cO O eO. n e w s e
- 2..%**
f.% x e et.es y dwwwd o c. t g m o 1 ..c. n t ~ ~ ~ ~ ~ ~ ~ ~ I .I., s ., x. ~ 2 n. ~ t x== w aaA a a = = r x Ei i ~ -.,,, .~.M e. e 1:.' 9 C. 20 5/10/84 p .
- e.. I's... .*.
i.,. /,...c)..' j 'k .a
( (* .~- p Table 10 Notation and definitions for release categories i Release Category Description M1A Containment Bypass, V-Sequence M1B Containment Bypass, SGTR i M2A Early f ailure/Early Melt, tio Sprays M2B Steam Explosion Failure M3 Early Failure / Late Melt, tio Sprays M4 Containment Isolation failure MS Intemediate Failure / Late Melt, tio Sprays M6 Intemediate Failure /Early Melt, tio Sprays M7 Late Failure, flo Sprays M8 Intemediate Failure With Sprays M9 Late Failure With Sprays 'e M10 Basemat Failure, No Sprays i. M11 L Basemat Failure With Sprays e, M12 No Containment Failure 21- '? 5/10/R4 f- .' ?,i I
-o--.; ay P l Table 11 Simplified containment matrix for Hillstone 3 (Internal Events) l l .i Containment Response Class HlA HIB H5 H6 H7 HIO Hil Hl2 A 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 e 5 0.01 0.79 0.20 ~ U. 0.05 0.95 7 1.0 8 l.0 'p 9 0.99 0.01 10 l U.99 0.01 V 1.0 1 -2 V2 1.0 Y$~ \\ l A R l T [
- r.
1 [ l g ye. _ =<*e* ~ a
w. +' r -{ Table 12 Simplified contairvient matrix for Hillstone 3 (External Events), 1 1 Containment Response Class MIA MIB M4 MS M6 M7 M10 Mll M12 A 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 4, 5 0.01 0.79 0.20 6 0.05 0.95 7 1.0 8 1.0 i p, V3 1.0 e i
- i
'R s ? O t + ss JP i
- A s-ce'.
j ~~ -me.-rs.f ~ +
- J
Q e .e f e (.' P p (*. 6 an. amm anun ' mann men om wan aman i g e eq e. .em 9ee a== =m en e N a = min - - t
==m-er ame
==
un I e e o e *
- e e
e e e ey my g
==,, en. I. am. ens.
==.==e==w ene
==> ee emme n es.e e. e. a=. e. e eim.ime e m W e== W W. em h e es 9 O. ee een=== h m
=
o*e .u. .= m -== m w e e e o ame a= emme ese eue
=
e M esse F. g e. e. O. e sy en er' H w en 1 g man en ammemm e a.= = em e=== maman e e e e e w e. g
- ens, one w==========
amm e S= ett N. e.
- e. S*
enseene en e e. g o e ar.t e. ev es M emie en h am m amm amen ew e amm en e. (- e.eno eue o M P'b une e M e e o se sie em asum aun mm O em een e
=
e e e e W o ene em esp E es. e.
==. S. e. W
- W gr W
.un, mm
== g amb g ammmunamn suam en asum s, en - e en e= so sin ammen i e e e o e o e e e e est en, y em eue esse eue emeenpane ene me e ta E e. e. N. ap sm. N O. se. ee n
- emen, amn Eu*
en W p W M h. ept. %, 9
- e. e i
8 am. 4r '"n* 8" e O 8 46 an e w== a e e e w
== e 9 me ene eus one E en b o Se e ese Pe. se e e e me am .o e-N e aun spun g sm amm am g man eum e s= ee en am Y e em E emm e.e e e e ease ene e meine enee e. em m e S.oen N M e e N am "s en sum sun O a em -amm e aum ey e me C aun so me ee se ene g= ao e aso m - e e me g, sy e e =e fs am e ame an se e e ame .e N. e. o e.e. .e to. eme ems one N e og .e e as. e .m w I c. a e a ~ .O..m. E e. m. a.o .e a e 444 4 e". x i e m ~ g p .t. l m. a-n 2, .u,m m .u. e ~ .m. N E o.e e .e. e. - - - e.e. e e e. e. e.e e e e o p -e .=, - em, e o. e emm - e e e o. a aun a.un a.m.ma.s.m man m amm m. e. a. - e a. e 2 .e e e.. W We ee. e = e.e n - m ~ e e .e .e e. e a.n. .m. e e. 4 en e E ene 96 er em.y se. tw e se N en. e ene o e e -O N M W W ek sun sun asE pun en e ' eue e e em. W W m - f j.I .tre$.
- 4..
I -.g - = m e en. e. es p. g m e. bi !i .m. p 9 =% e* W =tr 8 33 4,
== 9-lm.= h e e t w m ee i r a e e e i 24 5/10/84 t s oe <..o.e se e em.e. . e.e4 e e .*4n' s*.e 1-e e e f ee w ' e **, w
s1. ' g -4 a so \\. 7 _i a m. m. m. m. n. e. m. o W'- W M M W W W js' M M M M N 7 7 7 7 = . it r: M M M M x = w w f s
- m. i
~. m. ~. ~. ~. ~. ~. ~. e. ~. m. E 3 3 W W W N M M M M M W g m, .e T 7 m ~ e. m m e m T 2' d M d d d d d 6 2 2 = g W i ~ m. k m m e 'I w" r.- 3 o o e. m. e. m. 9 e. w w C; 5 i i e g o o o 2 M W W W'. Eo B 6 ~. ~. ~. m. m. m. L 4: c' x 4 4 4 N W M W M 3 e m = e u s. u 2 o m m m m m m m m m m m + o u e; = = M' 3 3 3' 3 x' 3' M M y e. u t b 6.' S' Y. 6 e. e. o. e. e. m. 6 I .o o o o o o o o e o o 3 W =. I e .N.v. 2. .J. .J. .e .d .e, h e W 5-3 L Si .s.a.u. e o R o r. ~ S ,o o, ~ ~ ~ .a .d g g .a e e. ~ esv e a, o m + 2, 1. 9 e o o e ~ ~ ~ o e o o o g g =o e ac t e=e
- =
c 2TO 3 _E 3, 5 N o a. o s, e o a v.- o s R S g k. I. a .s. s s. 9, U. k N b e o o e T, * - ~ c o e o o e o W W f C a e ~ N e c ( I, ' p ee3 1 f ev m e e ~ o es U 3 i i i i i i a i i i i i i
- 25 k'
5/10/84 I t ... p .... s td'.1 # t.- ,-.t, .ww.*. .*r t
y c-ble 15 Intermediate and late overpressuri:ation (nosprays) Secuence IPS[4f. IPPSS[5] .MPSS MPSS M-5 M-7 T;1s '- 3 2q g Xe-Kr .9 .9 .96 1.0 10+I .016 .015 1.05(-1) 9.2(-2) Cs-Rb .5 .3 ,39 ,3s Te-Sb .5 .3 .;a ,.: 4 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)
- 9( 3) 1,0 (.2 )
6 l. t. i. 6 -26 c 5/10/8'4 ... [ ) ' {.,:. 9,1?. .~.
r
- s..
{. h-Table 16 MPSS release category OP0s v Discrete Probability Distributions Release Catecory F* 1 1/2 1/4 1/10 i/100 M-1A 0.17 " 0.55 0.28 0 0 M-2 0.25 0 0.25 0.50 0 M-3
- 0. 0 0
0.06 0.63 0.31 M4 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)
"Probacility Values (P) e 't. i. i i t q F ,. 5/10/84,. 4 'l ~ ,n.
- e
~ i e %. w D g s e 4 V' .'9 w M M. M. 88.h M. M. M. ~. .6 ~. g t e s d W d a., e, .e, d .s. d e M M 8". er N si.d. m ~.= N M. ~., N. N. ~. N. N s .M., 3 3 .y S ned .sd W d .ed d g, A d W ~ ~ M .A 9 T ~.* W. ~ m 3 I' e ~. ~,- ~ ~. ~ ~. ~. ~ e. .= ~., p. s E g e . ~ o. e ~ e e m - ~ 5 ~ ..a t A. ~. i ~. e. e. 3 e e. e o... ( ~. ~. e. .e., .
- e e o.
e. e. m. e e.. ~ g w ~. t t ~. ~. ~... ~. e ~ c ~ j 4 % d d 4 4
- K E E
- e a.,
a. a., a. a. a. m. e. w - - ~ e M M 2 * .... ~ h. b M. M. e. . ~.. =., e. e. s. 3 - e e o e e o e e e 3 g o v e.s .i, .W2 s 3 3 S 2 R S S 2 2 * =
- gt.
.t. ,8 e. o. e. e. o. e. e. e. e. e. e. . e. e. e. i g}i - - ~ e =******2 'r' 1 p t i 2 a..- ~. .i a d :
- e.. ~... a a a g
r I wi t 20 J 5/10/84 ~ i
r 4 .o** t
- * * * * *
- e e e l-o.d'g*g'g'g*
a e e o e f l s c a.e o. e S # o . ee 3. E ESESSEEEEE $3 p I -2 - se ~
== t I
- e...
- e..
e. I .. o. i .~ ~ee as assi !i e . o.. o.. o.. o.. o. . o.. o. . e. .a.x we.o.a.o. 2 2. R. -~-~-~-~ _w .we.=.=.=.. =. d --ooca e .i b .s, e ew+ ,.6 .. o... w co=~ e. e e e I
- e. n. e.
e e s I . m. e. m. o.... e.. c. o. e. R. o R. +. o = e s e e m o e o. o ~. #. e. e. ~. e e oe. o. me .s. t ~ az .. o.. ~. ~. co-~ c. .o c a.e s op. e. 88 . =. =. =. e.... e.. J. d. 4..==. - e e e e m o e e. o e. m. m. e. e
- e. -.
. ~. e . se ee. e o.
- e. a. ee
. w e. ~ s -ooem . =. . m. e. e. e os osee s e ce esce aa e ce g ad adAJ Ad ad 4 i' a ce la g 4d r.1 n e 3, s ,.T n. i . e. a. .? .smses 3=sesenersewasases .e ? a# .v yi g e e u. w - ew n. e. a. e e -. 3. r l i r 2r). s
- o.
.. 5. 1. 0/. 84..,.*. ;. ..y.* ,.:.. s. s
- b..
m}}