NRC-95-4519, Forwards Preliminary Marked Up Radiological Sections of Chapter 15 of AP600 Ssar
| ML20086T336 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 07/28/1995 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NTD-NRC-95-4519, NUDOCS 9508020318 | |
| Download: ML20086T336 (51) | |
Text
1 Energy Systems Ba 355 Westinghouse ""SDufEh Pennsylvania 15230-0355 Electric Corporation NTD-NRC-95-4519 DCP/NRC0371 Docket No.: STN-52-(X)3 July 28,1995 Document Coctrol Desk U.S. Nuclear R(gulatory Commission Washing:on, D.C. 20555 A*ITENTION: T. R. QUAY SUIUECT: PRELIMINARY MARKED UP SECTIONS OF SSAR CHAIYTER 15 (ACCIDENT ANALYSES)
Dear Mr. Quay:
Enclosed are preliminary marked up radiological sections of Chapter 15 of the AP600 Standard Safety Analysis Report (SSAR). These sections complement those transmitted on June 2,1995 (References NTD-NRC-95-4480).
The subsections provided in this submittal and their topics are:
15.1.5.4 Steamline break radiological consequences 15.3.3.3 Locked rotor radiological consequences 15.4.8.3 Rod ejection radiological consequences 15.6.2 Small break outside containment 15.6.3.3 SGTR radiological consequences 15.7 Radioactive release from a subsystem or component i
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4 NTD-NRC-95-4519 DCP/NRC0371 July 27,1995 l l
I Please contact Brian A. McIntyre on (412) 374-4334 if you have any questions concerning this {
transmittal. ;
bi '
Nicholas J. Liparuto, Manager Nuclear Safety Regulatory and Licensing Activities i
/nja Enclosure ,
t cc: T.' Kenyon, NRC (w/o enclosures) ;
K. Coyne, NRC (IEl). ,
R. C. Jones, NRC (w/o enclosures) l G. D. McPherson, NRC (w/o enclosures) .[
r L. Shotkin, NRC (3El)
A. Levin, NRC (IEl) i J. Burtt, INEL (3EI)
P. Boehnert, ACRS (5EI)
B. A. McIntyre, Westinghouse (w/o enclosures) l i
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- c. 15. Accident Analyses 1
l h u..r:2,.ynder the low flow (natural circulation) condi ions actually present in the AP600 transient, the retum to power is severely limited by the ge negative feedback due to flow and power. The results of the bounding, E!! "~" ::= emonstrate that the departure from nucleate boiling design basis, as described in Section 4.4, is met for the steam rystem piping failure event.
15.1.5.3 Conclusions The analysis shows that the departure from nucleate boiling design basis is met for the steam system piping failure event. Departure from nucleate boiling and possible clad perforation following a steam pipe rupture are not precluded by the criteria. The preceding analysis shows that no departure from nucleate boiling occurs for the rupture assuming the most reac-tive rod cluster control assembly stuck in its fully withdrawn position. Th- P!ci;H
-cc g_..w J dm iimiting evem are widiin Um Juwaitci. cNO C" !m ,
15.1.5.4 Radiological Consequences The evaluation of the radiological consequences of a postulated main steam line break outside containment assumes that the reactor has been operating with the design basis fuel defect level (0.25 percent of power produced by fuel rods containing cladding defects) and that leaking steam generator tubes have resulted m a buildup of activity in the secondary coolant.
Following the rupture, startup feedwater to the faulted loop is isolated and the steam generator is allowed to steam dry. Any radioiodines carried from the primary coolant into the faulted steam generator via leaking tubes are assumed to be released directly to the environment. It is conservatively assumed that the reactor is cooled by steaming from the intact loop.
15.1.5.4.1 Source Term l 1
l The only significant radionuclide releases due to the main steam line break are the iodines that ;
become airborne and are released to the environment as a result of the accident. Noble gases are also released to the environment but their impact is secondary since any noble gases i entering the secondary side during normal operation are rapidly released to the environment. i The accident conditions do not result in an increase in the releases of noble gases beyond the !
levels associated with normal operation. j I
The analysis considers two different reactor coolant iodine source terms, both of which consider the iodine spiking phenomenon. In one case the initial iodine concentrations are assumed to be those associated with the design fuel defect level. The iodine spike is assumed to be initiated by the accident with the spike, causing an increasing level of iodine in the reac- !
tor coolant.
The second case assumes that the iodine spike occurs prior to the accident and that the maximum reactor coolant iodine concentration exists at the time the accident occurs.
Revision: 4 osi nev4usoito nosos3 95 August 31,1995 15.1-I8 T W85tlagt100Se
a
. 15. Accident Analyses The reactor coolant noble gas concentrations are assumed to be those associated with the design fuel defect level.
The secondary coolant is assumed to have an iodine source term of 0.04 pCi/g dose equivalent I-131. This is 10 percent of the design basis primary coolant activity.
15.1.5.4.2 Release Pathways t%r<.K-There are omponents to the accident releases:
- The secondary coolant in the steam generator of the faulted loop is assumed to be released out the break as steam. Any iodine contained in the coolant is assumed to be released.
- The reactor coolant leaking into the steam generator of the faulted loop is assumed to be released to the environment without any credit for iodine partition.ng or iodine plateout onto the interior of the steam generator.
\nWe 6 Credit is taken for decay of radionuclides until release to the environment. After release to the environment no consideration is given to radioactive decay or to cloud depletion by ground deposition during transpon offsite.
15.1.5.4.3 Dose Calculation Models The models used to calculateeyemedosesed qu. in. e Lc!: bady deser. :.:.u!:L.; f.=
^^i.:mo uf ;udma and J.c ;nadc: ad :a e u!;:: --to:c4mdy Ja:. duc :: m::r" . . i
'he re!: .~d ncL c 3os ocauf are provided in Appendix 15A. ;
15.1.5.4.4 Analytical Assumptions and Parameters The assumptions and parameters used in the analysis are listed in Table 15.1.5-1.
15.1.5.4.5 Identification of Conservatisms The assumptions and parameters used in the analysis contain a number of significant conservatisms:
- The reactor coolant activities are based on a fuel defect level of 0.25 percent whereas the expected fuel defect level is far less than this (See Section 11.1). )
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- The assumed leakage of 500 gallons of reactor coolant per day into each steam generator j is conservative. The leakage is expected to be a small fraction of this during normal operation. ;
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- The conservatively selected meteorological conditions are present only rarely.
c Warrev4u 50lin ROM 5M95 Revision: 4 3 W8Stingh0US6 15.1-19 August 31,1995
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Insert into subsection 15.1.5.4.2 i
- The reactor coolant leaking into the steam generator of the intact k>op is assumed to mix with the secondary coolant, raising the radioactive iodine concentration in the secondary water. The steam release from the intact loop includes dthe release of r radioactive iodine but credit is taken for iodine partitioning in the steam generator. j i
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. 15. Accident Andyses S
15.1.5,4.6 Dos!Cek.le;;en MaJci; e/Y Using the assumptions from Table 15.1.5-1, l the aka:a:cd rclcasc of ac;;.!!y d"e 'n h imaulaicd mam acern !!nc b.mok ,, 3..cn ... Tsbic li! 5 2. '":: :hyr:!d hes duc to-
- " ' ' " n of m.d crr adina und the ,,huic.budy du>ca due ' " et nL;uadou oi mnuer.icn-an "r "eNe gm c!eud : 9 'hc equ.valen: chale Eady dese retu!!;ng fr :-halat:c e n edir: -
. m e "n'ly m i for 'h e ;;c buundmy (U iu 2 hum asaum inuc) enc for iLc lcw popuk;ian ae"- nnter hundarj (0 ;0 " Laur rc: Loa inny. , uv uuss .., p,c ided i- Talk 15.1.5-3:
A descs dus to e nisin siconi ;nc brcok 4'b ' ,nra avMng iodme spib ere 4t Mn de 4trideliuc ,aiuo uf l0 CI R l00. Thc desc:; due '^ : .naia aivam linc Livok uib on acciden:
mitioiva iodme yuc am a 3msll f.ucdon (lcs, then l0 g. uni) uf ilm 10 CP !004tmtty TLcx duws nic !!- "W" 'he dese cr!!eria deF rd ;u LRI' Scc;;c- !515, ApponAv A__
15.1.6 Inadvertent Operation of the Passive Residual lleat Removal 6ystem M6@
15.1.6.1 Identification of Causes and Accident Description r
The inadvertent actuation of the passive residual heat removal system causes an increase in the core reactivity by decreasing reactor coolant temperature. The overpower /overtemperature protection (neutron overpower, overtemperature, and overpower AT trips) prevents any power increase that could lead to a departure from nucleate boiling ratio less than the safety analysis 4[pf limit. In addition, since the reactor coolant system depressurizes during this event, the low h pressurizer pressure function could generate a reactor trip.
These protection functions do not terminate operation of the passive residual heat removal system, however. So the RCS continues to cool down and depressurize. The safety injection low pressurizer pressure setpoint actuates the core makeup tank and brings the plant to a stable condition.
h(Sioo*4t* hMP The inadvertent actuation of the passive residual heat removal system could be caused by rato r a false actuating signal. Actuation of the passive residual heat removal gow" gerrgormvoTl e's opening the isolation valves, which establishes a flow path from coolant system hot leg, through the passive residual heat removal *y*m heat exchangerp, and back into the associated steam generator cold leg plenum. L.c 15 The passive residual heat removal gheat exchange av located above the core to promote natural circulation flow when the reactor coolant pumps are not operating gg reactor coolant pumps in operation, flow through the passive residua @l heb' t re oval i
% s enhanced. The heat sink for the passive residual heat Ayuem vr, provi d .$y eu-4 containment re ' al water storage tank, in which the passive residual heat re oval meem-w gvrhanyt submerggSince the passive residual removal 7 ey+m is conn ted to Yf only one reactor coolant system loop, the cooldown resulting from its actuation is asymmetric with respect to the core.
Revision: 4 o wrre,4sisoivn novossi9<
August 31,1995 15.1-20 W W85tingt100S8
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insert into subsection 15.1.5.4.6 !
r calculated total effective dose equivalent (sEDE) doses for both the case with pre-existing iodine spike and the case with accident-initiated lodine spike are determined to be less than 0.2 rem at the site boundary and less than 0.1 rem at the low population zone outer boundary.' ,
Dese doses are a small fraction of the proposed dose guidelipe of 25 remJEDE identified in the draft revision to 10 CFR Part 50 (see SECY-94-194). A small fractp, is taken gbeing ten percent or less7CM SlW Y @5 W M $ W6t\M iM i
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. 15. Accide:t Analyses Table 15.1.5-1 PARAMETERS USED IN EVALUATING TIIE RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE BREAK Reactor coolant iodine activity Accident-initiated spike: Initial activity equal to the design basis reactor coolant activity of 0.4 pCi/g dose equivalent 1-131 (see Table 11.1-2) with an assumed iodine spike that increases the rate of iodine release from fuel into the coolant by a factor of 500 (see Appendix 15A)
Preaccident spike: An assumed iodine spike that has resulted in an increase in the reactor coolant activity to 24 pCi/g of dose equivalent 1-131 (see Appendix 15A)
Reactor coolant noble gas activity Design basis activity (See Table 11.1-2)
Swondary coolant initial iodine activity 0.(M pCi/g dose equivalent 1-131 (10% of design basis reactor coolant concentrations listed in Table 11.1-2)
Secondary coolant mass (Ib) 3,[fl.M6E+05 Duration of accident (br) 8 Atmospheric dispersion factors See Table 15A-5 Steam generator in faulted loop Initial water mass (Ib) 1.T ';L 79 E+05 Primary to secondary leak rate (Ib/hr) 130(a) lodine panition coefficient 1.0 Steam generator in intact kop Primary to secondary leak rate (Ib/hr) 130(a)
Steam released (Ib) 0 - 2 hr 3. Col MY E+05 2 - 8 hr g,6 64& E+05 lodine partition coefficient 0.01 1
(a) Equivalent to 500 gpd at $61.5*F and 2250 psia l
ownnv4sisoirn nows 3i95 Revision: 4 )
[ W85tingh00S8 15.1-27 August 31,1995 l l
- 15. Accident Analyses
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Table 15.1.5-2 -),~'
ACTIVITY RELEASED TO TIIE ENVIRONMENT DUE TO A STEAM LINE BREAK lodines (accident-initiated iodine spike)
Isotope 0 - 2 hr (Cjl 2 - 8 hr (Ci) l-131 6.4 8.2 l-132 16.5 14.4 I-133 11.8 15.7 I-134 4.9 1.7 1-135 8.9 I lodine (preaccident iodine spike)
Isotope 0 - 2 hr (Ci) 2 - 8 hr (Ci) 1-131 7.8 12.5 1-132 12.0 6.0 1-133 1 18.6 1-134 .4 0.4 I-135 8.6 8.9 Noble gases (both cas )
Isotope 0 - 2 hr (Ci) 2 - 8 hr (Ci)
Xe-13 m 0.1 0.4 Xc- 3m 1.2 3.4 Xe 33 19 54
-135m 0.(M 0.0 e-135 0.05 1.1 Xe-138 0.007 0.0 Kr-85m 0.1 0.2 Kr-85 0.005 1.4 Kr-87 0.5 0.02 Kr-88 0.(X)7 0.2 l
Revision: 4 a =4usoiraows3i93 August 31,1995 15.1-28 y Westinghouse
. 15. Accide:t Analyses Table 15.1.5-3 RADIOLOGICAL CONSEQUENCES OF A STEAM LINE B
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Thyroid doses (rem)
Case 1 - Accident-initiated iodine spike Site boundary (0 - 2 hr) 3.2 low population zone (0 - 8 hr 1.0 Case 2 - Preaccident spike Site boundary (0 - 2tr) 3.9 Low population zone (0 - 8/ hr) 1.3 Whole body doses rem)
Case 1 - ccident-initiated iodine spike Site boundary (0 - 2 hr) 0.10 Low population zone (0 - 8 hr) 0.03 Case 2 - Preaccident spike Site boundary (0 - 2 hr) 0.12 Low population zone (0 - 8 hr) 0.(M c hmv4\l50lfn R03 053195 Revision: 4 Westirigh0US8 15.1-29 August 31,1995
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- 15. AccidInt An Jyses l
15.3.3.2.7 Results 1 yf} l F gures 15.3.3-1 through 15.3.3- show the trans'ent results for one locked rotor with four loops in operation without offsite power available, case bounds the results forgtase with offsite power. The results of these calculations are also summarized in Table 15.3E The peak reactor coolant system pressure reached during the transient is less than that which causes stresses to exceed the faulted condition stress limits of the ASME Code.Section III.
Also, the peak clad surface temperature is considerably less than 2700 F. The clad temperature is conservatively calculated assuming that departure from nucleate boiling occurs at the initiation of the transient. These results represent the most limiting conditions with respect to the locked rotor event or the pump shaft break.
The calculated sequence of events for the case analyzed is shown in Table 15.3-1. With the reactor tripped. a stable plant condition eventually is attained. Normal plant shutdown may then proceed.
15.3.3.3 Radiological Consequences The evaluation of the radiological consequences of a postulated locked reactor coolant pump rotor accident assumes that the reactor has been operating with the design basis fuel defect level (0.25 percent of power produced by fuel rods containing cladding defects) and that leaking steam generator tubes have resulted in a buildup of activity in the secondary coolant.
As a result of the accident,18 percent of the fuel rods are conservatively assumed to be damaged in such a way that the activity contained in the fuel-clad gap is released to the reactor coolant. Activity carried over to the secondary side because of primary-to-secondary leakage is available for release to the environment via the steam line safety valves or the power-operated relief valves.
15.3.3.3.1 Source Term The significant radionuclide releases due to the locked rotor accident are the iodines and the
. noble gases. The reactor coolant iodine source term assumes a pre-existing iodine spike. The initial reactor coolant noble gas concentrations are assumed to be those associated with the Sfuel defect level. These initial reactor coolant activities are of secondary importance compared to the release of the gap inventory of fission products from the portion of the core assumed to fail because of the accident.
MSCW@
The initial secondary coolant activity is assumed to be 0.04 pCi/g dose equivalent I-131. This is 10 percent of the design basis primary coolant activity.
15.3.3.3.2 Release Pathways There are two components to the accident releases:
Revision: 4 owm3siso3rnno3.o33i95 August 31,1995 15.3-8 Yjv Westingh0Use
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Insert into subsection 15.3.3.3.1 Based on NUREG-1465, the fission product gap fraction is three percent of fuel inventory. For ihis analysis, the gap fraction is increased to 3.6 percent of the inventory to address concerns identified in NUREG-1465 regarding the applicability of the three percent gap fraction to high burnup fuel (i.e., fuel with burnup in excess of 40 gigawatt days per metric ton of uranium).
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. 15. Accidelt Analyses
- The activity initially in the secondary coolant is available for release as long as steam releases continue.
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- The reactor coolant leaking into the steam generators is assumed to mix with the secondary coolant. The iodine from the primary coolant mixes with the secondary l coolant. As steam is released, a portion of the iodine in the coolant is released. The i fraction of iodine released is defined by the partition coefficient assumed for the steam l generator. The noble gas activity entering the secondary side is released to the l environment. These releases are terminated when the steam releases stop.
Credit is taken for the decay of radionuclides until release to the environment. After release to the environment, no consideration is given to radioactive decay or to cloud depletion of iodines by ground deposition during transport offsite.
15.3.3.3.3 Dose Calculation Models The models used to calculate oses ' A ~"n a!cr' h!c t~d dv>w y .auhtng fem-irha! N nf i~% oud ;he . 8' ed : ; Jcub:c hak body dc= d"e * ...uiicra;c- S hreia-d ~At- onc nceivitpare provided in Appeadix 15A.
15.3.3.3.4 Analytical Assumptions and Parameters The assumptions and parameters used in the analysis are listed in Table 15.3-3.
15.3.3.3.5 Identification of Conservatisms The assumptions used in the analysis contain a number of significant conservatisms.
= Although fuel damage is assumed to occur as a result of the accident, no fuel damage is anticipated.
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- The reactor coolant activities are based on a fuel defect level of 0.25 percent, whereas '
the expected fuel defect level is far less than this. (See Section 11.1.)
= The leakage of reactor coolant into the secondary system, at 1000 gallons per day, is conservative. The leakage is normally a small fraction of this.
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- It is unlikely that the conservatively selected meteorological conditions are present at the time of the accident.
15.3.3.3.fi ik..~ ...d Doses
\O Using the assumptions from!=! Table :cd r!~15.3-3,bc er sc'irity d= =SN :c the nnernIn'ad !cchcd .viv ccMe=' ic ig ven in T+!c 15.3-4. TLv d,yiv;d dv3c3 Juc :c !""""
cf unh= ;eJmu3 .uid dew 'e-b~4y doses one 'n 'ha enmhinatinn nr im.-er ;-,a :n :' _
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4 Insert into subsection 15.3.3.3.6 the calculated total effective dose equivalent (TEDE) doses are determined to be less than 1.0 rem at the site boundary and less than 0.5 rem at the low population zone outer boundary.
These doses are a small fraction of the proposed dose guidelipe of 25 rem,JEDE identified in the draft revision to 10 CFR Part 50 (see SECY-94-194). A small fraction is taken as being ten percent or less/'ccessWW w (kh h StWciQVt'W 9M, L
- 15. Accident Analyses en !yzed fu. ilm iic 'vuuuda.j and foi ic 'ow populatica : sac cu:= 5ao.Jary. Ine ousc3 .
re !!::ed in Teblc 15.M.-
The dos, duc av o lvacd suivi auddem um o = !! f w.ien # 6 guMelima:sca ci m.
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15.3.4 Reactor Coolant Pump Shaft Break 15.3.4.1 Identification of Causes and Accident Descri: tion The accident is postulated as an instantaneous failure of a reactor coolant pump shaft, as discussed in Section 5.4. Flow through the affected reactor coolant loop is rapidly reduced, though the initial rate of reduction of coolant flow is greater for the reactor coolant pump rotor seizure event. Reactor trip occurs on a Low-flow signal in the affected loop.
Following the reactor trip, heat stored in the fuel rods continues to be transferred to the coolant, causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generators is reduced: first, because the reduced flow results in a decreased tube-side film coefficient; and second, because the reactor coolant in the tubes cools down while the shell-side temperature increases. (Turbine steam flow is reduced to zero upon plant trip.)
The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the steam generators, causes an insurge into the pressurizer and a pressure increase throughout the reactor coolant system. The insurge into the pressurizer compresses the steam volume, actuates the automatic spray system and opens the pressurizer safety valves, in that sequence. For conservatism, the pressure-reducing effect of the spray is not included in the analysis.
This event is classified as a Condition IV incident (limiting fault), as defined in Subsectien 15.0.1.
15.3.4.2 Conclusion With a failed shaft, the impeller could be free to spin in a reverse direction as opposed to being fixed in position as assumed in the locked rotor analysis. However, the net effect on core flow is negligible, resulting in only a slight decrease in the end point (steady-state) core flow. For both the shaft break and locked rotor incidents, reactor trip occurs very early in the transient. In addition, the locked rotor analysis conservatively assumes that departure from nucleate boiling (DNB) occurs at the beginning of the transient. The calculated results esented for the locked rotor analysis bound the reactor coolant pump shaft break event.
15.3.y References o
- 1. Burnett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Nonproprietary), April 1984.
Revision: 4 ow.riev3siso3rnno3.os3i95 August 31,1995 W Westinghouse g g 15.3-10
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. 15. Accidet A alyses Table 15.3-3 1, i
PARAMETERS USED IN EVALUATING TIIE RADIOLOGICAL l l
CONSEQUENCES OF A LOCKED ROTOR ACCIDENT l l
Initial reactor coolant iodine activity An assumed iodine spike that has resulted in an increase in the reactor coolant activity to 24 pCi/g of Dose Equivalent L 131 (See Appendix 15A)
Reactor coolant noble gas activity Design basis activity (See Table 11.1-2)
Secondary coolant initial iodine 0.N pCi/g Dose equivalent activity 1-131 (10% of design basis reactor coolant concentrations listed in Table 11.1-2)
Fraction of fuel rods assumed to fail 0.18 Core activity See Table 15A-3 Fission product gap fractions 3. (,
-S:~ ; " . ;J nu:!!& "~ ) ?-
! ^n;; !".n! nuc!!&' "* ) T-Iodine chemical form Elemental (%) $ 4.T5 Organic (%) 1r C O )
Particulate (%) 97- 9 5 i Reactor coolant mass (Ib) 3.39 E+05 )
i Secondary coolant mass (Ib) .7.15 GM E+05 j Condenser Not available l
Duration of accident (hr) 8
{
Atmospheric dispersion factors See Table 15A-5 !
Primary to secondary leak rate (Ib/hr) 260(a)
I Steam released (lb) 0-2 hours 4.0 LOT 5.M+ 05 l 2-8 hours '?.3 C+05 \ .oq f., kb y i lodine partition coefficient in steam 0.01 generators (a) Equivalent to 1000 gpd at 561.5*F and 2250 psia Revision: 4 ei arrev3si303rnno3453i95 August 31,1995 15.3-14 3 Westinghouse
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- 15. Accident Acalyses -
Table 15.3-4
[
ACTIVITY RELEASED TO TIIE ENVIRONMENT DUE TO A LOCKED REACTOR COOLANT PUMP ROTO /
Iodines Release (
Isotope 0-2 hrs 2-8 hrs I-131 6.0 E00 5.7 E+01 I-132 6.0 E00 1.8 E+01 1-133 1.2 1 9.6 E+01 I-134 .3 E+01 3.6 E00 l-135 9.5 E+01 5.9 E+01 Noble gases Release (Cl)
Isotope 0-2 hrs 2-8 hrs Xe-131 2.3 E00 1.0 E+01 Xe-1 m 9.1 E+01 3.8 E+02 j 133 6.3 E+02 2.7 E+03 Xe-135m 1.1 E+01 1.1 E-01 Xe-135 1.9 E+02 6.3 E+02 ,
l Xe-138 4.1 E+01 2.8 E-01 l Kr-85m 6.6 E+01 1.7 E+02 Kr-85 1.1 E+01 4.7 E+01 l
Kr-87 8.2 E+01 6.1 E+01 Kr-88 1.6 E+02 2.9 E+02 owmv3uso3rn.no3e3 95 Revision: 4 W8Silligt100S8 15.3-15 August 31,1995
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' 15. Accident Analyses j
Table 15.3-5 RADIOLOGICAL CONSEQUENCES O ?
A LOCKED ROTOR ACCIDEN Thyroid Doses Site boundary 3.1 rem low population zone 4.2 rem Whole Body Dose S' oundary 0.2 rem w population zone 0.2 rem I
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- 15. Accident Anilyses 15.4.8.2.1.9 Pressure Surge A detailed calculation of the pressure surge for an ejection worth of about one dollar at beginning of cycle, hot full power, will demonstrate that the peak pressure does not exceed that which would cause the stress to exceed the Service Limit C as described in the ASME Code, Section Ill. Since the severity of the present analysis does not exceed the worst-case analysis, the accident for this plant will not result in an excessive pressure rise or further damage to the reactor coolant system.
15.4.8.2.1.10 Lattice Deformations A large temperature gradient exists in the region of the hot spot. Since the fuel rods are free to move in the vertical direction, differential expansion between separate rods cannot produce distortion. However, the temperature gradients across individual rods may produce a differential expansion tending to bow the midpoint of the rods toward the hotter side of the rod.
Calculations indicate that this bowing results in a negative reactivity effect at the hot spot since Westinghouse cores are undermoderated, and bowing tends to increase the undermoderation at the hot spot. In practice, no significant bowing is anticipated, since the structural rigidity of the core is sufficient to withstand the forces produced.
1 Boiling in the hot spot region would produce a net flow away from that region. However, l the heat from the fuel is released to the water relatively slowly, and it is considered l inconceivable that crossflow will be sufficient to produce sufficient lattice forces. Even if i massive and rapid boiling, sufficient to distort the lattices, is hypothetically postulated, the large void fraction in the hot spot region produces a reduction in the total core moderator to i fuel ratio and a large reduction in this ratio at the hot spot. The net effect would therefore be a negative feedback.
l It is concluded that no mechanism exists for a net positive feedback resulting from lattice deformation. In fact, a small negative feedback may result. The effect is conservatively ignored in the analysis.
1 15.4.8.3 Radiological Consequences ;
1 The evaluation of the radiological consequences of a postulated rod ejection accident assumes that the reactor is operating with the design basis fuel defect level (0.25 percent of power produced by fuel rods containing cladding defects) and that leaking steam generator tubes result in a buildup of activity in the secondary coolant.
As a result of the accident,15 percent of the fuel rods are assumed to be damaged such that the activity contained in the fuel-clad gap is released to the reactor coolant. In addition, a i small fraction of fuel is assumed to melt and release core inventory to the reactor coolant. l l
o% arm 3\15Nfn R03 053195 Revision: 3 3 Westinghouse 15.4-37 August 31,1995
- 15. Accident Analyses I i l
l Activity released to the containment via the spill from the reactor vessel head is assumed to I be available for release to the environment because of containment leakage. Activity carried l over to the secondary side due to primary-to-secondary leakage is available for release to the l environment through the steam line safety or power-operated relief valves. l 15.4.8.3.1 Source Term The significant radionuclide releases due to the rod ejection accident are the iodines and the noble gases. The reactor coolant iodine source term assumes a pre-existing iodine spike. The initial reactor coolant noble gas concentrations are assumed to be those associated with the design fuel defect level. These initial reactor coolant activities are of secondary importance compared to the release fission products from the portion of the core assumed to fail.
The initial secondary coolant activity is assumed to be 0.04 pCi/g dose equivalent I-131. This is 10 percent of the design basis primary coolant activity.
15.4.8.3.2 Release Pathways There are three components to the accident releases:
- The activity initially in the secondary coolant is available for release as long as steam releases continue.
- The reactor coolant leaking into the steam generators is assumed to mix with the secondary coolant. The iodine from the primary coolant mixes with the secondary coolant and, as steam is released, a portion of the iodine in the coolant is released. The fraction of iodine released is defined by the partition coefficient assumed for the steam generator. The noble gas activity entering the secondary side is released to the environment. These releases are terminated when the steam releases stop.
- The activity from the reactor coolant system and the core is released to the containment atmosphere and is available for leakage to the environment through the assumed design basis containment leakage.
Credit is taken for decay of radionuclides until release to the environment. After release to the environment, no consideration is given to radioactive decay or to cloud depletion of iodines by ground deposition during transport offsite.
15.4.8.3.3 Dose Calculation Models The models used to calculate thy:md doses and cquivakn' L6 Lvdy Jc;ca scoultia; fron; W'"!ca of iodiac and Sc mc&! =ed :c c d:c ech& Scdy deso duc w u..mcr5 :n s the ickecd nubic ;= 2c!H:y are provided in Appendix 15A.
Revision: 3 owmm3mnou)33i,3 August 31,1995 15.4-38 T Westinghouse
1 l
. I t i
- 15. Accident Analyses 15.4.8.3.4 Analytical Assumptions and Parameters l The assumptions and parameters used in the analysis are listed in Table 15.4-4.
15.4.8.3.5 Identification of Conservatisms The assumptions used in the analysis contain a number of conservatisms:
- Although fuel damage is assumed to occur as a result of the accident, no fuel damage is anticipated.
- The reactor coolant activities are based on an assumed fuel defect level of 0.25 percent, whereas the expected fuel defect level is far less than this. (See Section 11.1.)
- The leakage of reactor coolant into the secondary system, at 1000 gallons per day, is conservative. The leakage is normally a small fraction of this.
- It is unlikely that the conservatively selected meteorological conditions are present at the time of the accident.
- The leakage from containment is assumed to continue for a full 30 days. It is expected that containment pressure is reduced to the point that leakage is negligible before this time.
15.4.8.3.6 Lk .,c., ..d Doses b
\CS Using the assumptions from Table 15.4-4, theb!cula;cd acicac cf mi.;;j doc a.1776 p~*u!:rdr^d ejr::ica accids.n u e,s.. i- T25!e !5.A 5 Re thyrcid dcx: duc :: :-h:!::!c -
cf .rLv.uc mau.cs and :he -hc!c body desca duc :c 1: :cdin; ivu vi inu..cr;ien :- 6:
e nc5k rs cleud end ihc cv.,e:cn; who!c body de::e recu!::n; frc.T inh;!a:ica of iodincs c.rc -
2 'y=d for :hc s.i Luuudaiy and fu. ;he !ci" pcpu! :ica cn cc:= $~" d j h A=;
e na a in w: :5ps---
Sc dc= de ;e e rad s.,cci.en Accidcni =: ">:!! "*- *he gr! dan ~ ;;!u= cf 10 C"1 100.
(".,J ;;hiu" m Jcf;uwd ou SRP 5ccuun i3.4.5.3 Ls.ng 25 p==n*, er k=.; -
References 15.4 7 tD
- 1. Risher, D.11., Jr., and Barry, R. F., " TWINKLE--A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Nonproprietary),
January 1975.
- 2. Hargrove, H. G., "FACTRAN--A FORTRAN-IV Code for Thermal Transients in a UO2 Fuel Rod, WCAP-7908 (Proprietary) and WCAP-7337 (Nonproprietary), June 1972.
ohmarrev3\l504fn R03-053195 Revision: 3 W W85tiligtl00Se 15.4-39 August 31,1995
- { (, o toraddiO M bCNS M D N
. . .- _ ., - ~.
=
k 1 i
insert into subsection 15.4.8.3.6 l calculated total effective dose equivalent (TEDE) doses are determined to be less than 2 rem at ;
the site boundary and less than I rem at the low population zone outer boundary. 'Ihese doses
, are well within the proposed dose guideline of 25 rem TEDE identified in the draft revision to 10 CFR Part 50 (see SECY-94-194). The phrase "well within" is taken as being 25 percent or less [ C.o nb M M A P Wsh A hhck b'sw Plu,
- i. .
4 t
I
- 15. Accident Analyses Table 15.44 (Sheet 1 of 2)
PARAMETERS USED IN EVALUATING TIIE RADIOLOGICAL CONSEQUENCES OF A ROD EJECTION ACCIDENT Initial reactor coolant iodine activity An assumed iodine spike that has resulted in an increase in the reactor coolant activity to 24 pCi/g of dose equivalent I-131 (See Appendix 15A)
Reactor coolant noble gas activity Design basis activity (See Table 11.1-2)
Secondary coolant initial iodine activity 0.04 pCi/g dose equivalent I-131 (10% of design basis reactor coolant concentrations listed in Table 11.1-2)
Fuel clad failure Fraction of fuel rods assumed to fail 0.15 Fission product gap fractions <3.t ,
g . i:.g - 3.g3 -
L=g!%d 0.10-Core melting Fraction of core melting 0.00375 Fraction of activity released lodines 0.5 Noble gases 1.0 lodine chemical form Elemental (%) 4-tt5'- 4 I b Organic (%) 0.15 Particulate (%) 97 4 S Core activity See Table 15A-3 Reactor coolant mass (Ib) 3.39 E+05 Revision: 3 owarrev3tiso4rnno34153i95 August 31,1995 15.4-48 3 Westingh0Use
+ 15. Accident Analyses Table 15.4-4 (Sheet 2 of 2)
PARAMETERS USED IN EVALUATING TIIE RADIOLOGICAL CONSEQUENCES OF A ROD EJECTION ACCIDENT Condenser Not available Duration of accident (days) 30 Atmospheric dispersion factors See Table 15A 5 Secondary system release path Primary to secondary leak rate (Ib/hr) 260(a)
Secondary coolant mass (Ib) p.4 9-42 E+05 Steam release time period (min) 10 Steam released (Ib) 1.2 E P -- D 9 fb lodine partition coefficient in 0.01 steam generators Containment leakage release path Containment leak rate (% per day) 0 - 24 hrs 0.12 1 - 30 days 0.06 lodine removal coef6cients ( ht~
Elemental tir ~2.0 Organic 0 Particulate 0-35 O I C\ W DF limit forhodine removal go gi; ;. _ . u m phcr Pen!:r - .0 E . '" -
(a) Equivalent to 1000 gpd at 561.5'F and 2250 psia o barrev3\15041n.R03453195 Revision: 3
[ W65tingt100Se 15.4-49 August 31,1995
- 15. Accident Analyses l
Table 15.4-5 ,-
ACTIVITY RELEASED TO TIIE ENVIRONMENT FOR A ROD EJECTION ACCIDENT -
/
Releases (Cl) /
Isotope 0 - 2 hr ,
,2 720 hr I-131 3.0 E+01 3.0 E+01 1-132 3.5 E+01 1.0 E+01 1-133 5.8 E+01 4.3 E+01 1-134 3.8 E . 3.2 E00 l-135 4. E+01 2.8 E+01 Xe-131m 3.7 E-01 2.9 E+01 Xe-133m 1.6 E+01 3.1 E+02 Xe-133 1.0 E+02 4.3 E+03 Xe-135m 6.2 E+00 1.4 E-02 Xe-135 3.4 E+01 1.5 E+02 Xe-I38 2.5 E+01 4.0 E-02 Kr-85m 1.2 E+01 2.7 E+01 Kr-85 5.8 E+0) 8.7 E+01 Kr-87 1.8 E+01 6.4 E00 Kr-88 2.9 E+01 3.8 E+01 Revision: 3 os nn3uso4rnnow)33i,3 August 31,1995 15.4-50 3 Westinghouse
. +-+ + wr++
- 15. Accident Analyses Table 15.4-6 RADIOLOGICAL CONSEQUENCES OF A ROD EJECTION ACCIDENT Thymid doses (rem) /
Site boundary 15.1 Low population zone 3.7 Whole body doses (rem)
' e boundary 0.5 Low population zone 0.12 I
i l
l l
l otsarrev3\l504fn.R034)33195 Revision: 3
[ W85tingh00S8 15.4-51 August 31,1995
e
-- p
- 15. Accident Analyses
, l }
}F.6.2 Failure of $ mall Lines Carrying Primary Coolant Outside Containment Th sma!! lines carrying primary coolant outside containment are the reactor coolant system sample line and the discharge line from the chemical and volume control system to the liquid radwaste system. These lines are used only periodically. No instrument lines carry primary coolant outside the containment.
When excess primary coolant is generated because of boron dilution operations, the chemical and volume control system purification flow is diverted out of containment to the liquid radwaste system. Before passing outside containment, the flow stream passes through the chemical and volume control system heat exchangers and mixed bed demineralizer. The flow leaving the containment is at a temperature of less than 140'F and has been cleaned by the demineralizer. The flow out a postulated break in this line is limited to the chemical and volume control system purification flow rate of 100 gpm. Considering the low temperature of the flow and the reduced iodine activity because of demineralization, this event is not analyzed. The postulated sample line break is more limiting.
The sample line isolation valves inside and outside containment are open only when sampling.
The failure of the sample line is postulated to occur between the isolation valve outside the containment and the sample panel. Since the isolation valves are open only when sampling, the loss of sample flow provides indication of the break to plant personnel. In addition, a break in a sample line results in activity release and a resulting actuation of area and air radia-tion monitors. The loss of coolant reduces the pressurizer level and creates a demand for makeup to the reactor coolant system. Upon indication of a sample line break, the operator would take action to isolate the break.
The sample line includes a flow restrictor at the point of sample to limit the break flow to less l than 130 gpm. Offsite doses are based on break flow isolation after 30 minutes.
15.6.2.1 Source Term The only significant radionuclide releases are the iodines and the noble gases. The analysis assumes that the reactor coolant iodine is at the maximum technical specification level for continuous operation. In addition,it is assumed that an iodine spike occurs at the time of the accident. The reactor coolant noble gas activities are assumed to be those associated with the design basis fuel defect level.
15.6.2.2 Release Pathway The reactor coolant that is spilled from the break is assumed to be at high temperature and pressure. A large ponion of the flow flashes to steam and the iodine in the flashed liquid is assumed to become airborne.
l Revision: 3 ow .uso6no3-osso95 May 31,1995 15.6-4 W Westinghouse
ii ! !"i"'!"i i 5 Accident Analyses De iodine and noble gases are assumed to be released directly to the environment with no credit for depletion, although a large fraction of the airborne lodine is expected to deposit on building surfaces. No credit is assumed for radioactive decay after release.
15.6.2.3 Dose Calculation Models #
De models used to calculate 4hyreid doses d "n r;eadn* "*c!: N;dj dc:.x ic3uldn; f- -
.hiakte.; af iudiac.: and 'ha nudd ;:;d :c c;kub:: mbb N:dy dc_.; duc iv .a.uici. den b.
k sclca:;;d Mkee ac'iet y are t provided in Appendix 15A.
15.6.2.4 Analytical Assumptions and Parameters ne assumptions and parameters used in the analysis are listed in Table 15.6.2-1.
15.6.2.5 Identification of Conservatisms
%e assumptions used contain the following significant conservatisms:
- Re reactor coolant activities are based on a fuel defect level of 0.25 percent, whereas the expected fuel defect level is far less than this. (See Section 11.1.)
. It is unlikely that the conservatively selected meteorological conditions would be present at the time of the accident. ,
i 15.6.2.6 Releases and Doses Q s Using the assumptions from Table 15.6.2-1, the calculated --! cre d 2c9 ;;y duc : 'he -
r ;= :r'e' _ug;v iiuc i,4v.L oudk c: ""^a' ic given in Tabk 'S 6 ' 1 R: Syrs ;
^"^ " Iv ildlaiaisvas of airhGsnc audauca and Asv ivE.GIc UGdy d6ac5 duc iv use vun;I k.; MI i of_ im merei- '- Sc asbic r a cloud mid ic equ;v;ka "*^b Sady &::: rc:;;:dng from inhein, inn nrintiines nre m!yc.cd for m :;i:e Faund.uj .d for die iow popuiadou cvuc ou:e' hnnaa a ry m A- ~ ': acd ,,, T;bk 1512 ' -
erneancre cine en n emen i: : brca; uui3ide tuniou,u,cui n;di un accidem-imitatediodinu apav=
are a a;nrJ: fiscueeface 'hna 1(i,aa c- .:) af ic 1^ C.T 100 !!silis. Theau du3cs aiu nidda Me ese cra.sa dur.ns-;f in the UD&^cuen ;5.5.2.
15.6.3 Steam Generator Tube Rupture 15.6.3.1 Identification of Cause and Accident Description 15.6.3.1.1 Introduction De accident examined is the complete severance of a single steam generator tube. De accident is assumed to take place at power with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited number of defective fuel rods owm.mmaoso33o95 Revision: 3 W W85tingh0USe 15.6-5 May 31,1995
Insert into subsection 15.6.2.6 total effective dose equivalent (TEDE) doses are determined to be less than 1.0 rem at the site boundary and less than 0.2 rem at the low population zone outer boundary. These doses are a small fraction of the proposed dose guideline of 25 rem TEDE identified in the draft revision .
to 10 CFR Part 50 (see SECY-94-194).
l
i 4
- 15. Accident Aryvses I
Table 15.6.2-1 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SMALL LINE IIREAK OUTSIDE CONTAINMENT
] Reactor coolant iodine activity Initial activity equal to the design basis reactor coolant activny of 0.4 pCi/g dose equivalent 1-131 (see Table 11.1-2) with an assumed iodine spike that increases the rate of iodine release from fuel into the coolant by a factor of 500 (see Appendix 15A).
Reactor coolant noble gas activity Design basis activity (See Table 11.1-2)
Fraction of reactor coolant flashing 0.4 Duration of accident (br) 0.5 Atmospheric dispersion factors See Table 15A-5 ow-v3uso6 Rous3095 Revision: 3
[ W85tingh0US8 15.6-59 May 31,1995
- ,f-- 9,
- v: ; 15. Accidert Analyses A p *, < i Table 15.6.2-2 ACTIVITY LEASED TO TIIE ENVIRONMENT DUE T A SMALL E IIREAK OUTSIDE CONTAINMENT Release (C' lodines 1-131 2.2El I-132 1.2E2 1-133 4.7El 1-134 5.4El 1-135 4.5El Noble gases Xe-131m 5.8E0 Xe-133m 5.3El Xe-13 8.2E2 Xc- Sm 4.9E-1
-135 2.4El e 138 3E-1 Kr-85m 5.9 Kr-85 2.lEl Kr-87 2.9E0 Kr-88 9.7E0 U.eVISion: 3 ow .uso6.no3-os3095 May 31,1995 15.6-60 W Westinghouse
- 15. Accident Analyses n-Table 15.6.2-3 RADIOL ICAL CONSEQUENCES F A SMALL LINE EAK OUTSIDE CO AINMENT Thyroid doses (re Site boundary 11.9 Low population n 1.6 Whole body dos (rem)
Sit undary 0.4 w population zone 0.05
/
) ownsusos. nous 3095 Revision: 3 W W85tingh00S8 15.6-61 May 31,1995
s w
- 15. Accid:tt Analyses 15.6.3.3 Radiological Consequences The evaluation of the radiological consequences of the postulated SGTR assumes that the reactor is operating with the design basis fuel defect level (0.25 percent of power produced by fuel rods containing cladding defects) and that leaking steam generator tubes result in a buildup of activity in the secondary coolant.
Following the rupture, any noble gases carried from the primary coolant into the faulted steam generator via the break flow are released directly to the environment. The iodine entering the secondary side is also available for release, with the amount of iodine released dependent on the flashing fraction of the reactor coolant and on the partition coefficient for iodine in the steam generator. In addition to the activity released through the faulted loop, there is also a small amount of activity released through the intact loop.
15.6.3.3.1 Source Term The significant radionuclide releases from the SGTR are the noble gases and the iodines that become airborne and are released to the environment as a result of the accident.
The analysis considers two different reactor coolant iodine source terms, both of which consider the iodine spiking phenomenon. In one case the initial iodine concentrations are assumed to be those associated with the design fuel defect level. The iodine spike is assumed to be initiated by the accident, the spike causing an increasing level of iodine in the reactor ;
I coolant.
The second case assumes that the iodine spike occurs before the accident and that the maximum reactor coolant iodine concentration exists at the time the accident occurs.
The reactor coolant noble gas concentrations are assumeu to be those associated with the design fuel defect level.
5 %< Tul. Spr.c. vel-c.
The secondary coolant is assumed to have an iodine source termpf 0.04 pCi/g dose equivalent 1 131. This is ten percent of the design basis primary coolanf activity.
15.6.3.3.2 Release Pathways The noble gases contained in the reactor coolant that leaks into the intact steam generator and that enters the faulted steam generator through the break are assumed to be released immediately as long as a pathway to the environment exists. There are three components to the modeling of iodine releases:
- Intact loop steaming, with credit for iodine partitioning (includes continued primary to secondary leakage at the maximum rate allowable by the technical specifications)
- Faulted loop steaming, with credit for iodine partitioning (includes modeling of increasing activity in the secondary coolant due to the break flow)
Revision: 3 mmmsooowsms May 31,1995 15.6 14 W-Westingh0Use
- 15. Accident Analyses
. Release of flashed reactor coolan through the faulted 1 , with no credit for s rubbing ,.
(>Mr &
ww_: <.
'"CuAr*. h =- n . --
.(rwk s.4 d *t%s Credit is taken for decay of radionuelides until release to the environment. After release to the environment, no consideration is given to radioactive decay or to cloud depletion of iodines by ground deposition during transpon offsite.
15.6.3.3.3 Dose Calculation Models he models used to calculate thytmd doses =d q= ;6r* -he!e 'ady her rc_dnus busi-inhotatinn nf bjjn;; ggj 1, ,,,gj;1 neaA +n ont. eit nen whnle.hndu doc ~ jo; ig nng;;; .nn in _
.h, r,1,..m.a .We y e hy..are provided in Appendix 15 A.
15.6.3.3.4 Analytical Assumptions and Parameters he assumptions and parameters used in the analysis are listed in Table 15.6.3-3.
15.6.3.3.5 Identification of Conservatisms ne assumptions used in the analysis contain a number of significant conservatisms, such as:
. The reactor coolant activities are based on a fuel defect level of 0.25 percent, whereas the expected fuel defect level is far less than this (see Section i1.1).
= lt is unlikely that the conservatively selected meteorological conditions are present at the time of the accident.
15.6.3.3.6 Releases and Doses Using the assumptions from Table 15.6.3-3, the calculated =' esc acLi.iy duc-;o
" r=: !2H SU ' .; glicn in Tabic ;5.6.3-4. L byrcid dc= h duc ^f-Jialaucr. cf =$
odines an e whole-body doses due to combination ofi e non in the noble g loud and the e ent whole body dose res ting from inhalation o ~ i ines are analy for the site bo dary to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> residene me) and for the low pulati zone o r boundary (0 to hour res ence time). The oses are listed in Tab 15.6.3-5.
e doses due to an GTR ith a pre-existing iodin spike are wi the\uideline values if 10 CFR 100. The do s . ue to an SGTR with a ccident initia iodine s ' e are a smal l action (less than ten . nt) of 10 CfF G iimits. Inese doses are tim - _ __ n t he do o : arin danned la-SiG 3ccLea-;5.65 15.6.3.4 Conclusions The results of the SGTR analysh show that the overfill protection logic and the passive r system design features provide d-~, protection to prevent steam generator overfill for the j AP600. Following an SGTR secident, the operators can identify and isolate the faulted steam o wartev)\l506 Robo $1los Revision: 3 W W85tingh00S8 15.6-15 May 31,1995
o l .
l Insert into subsection 15.6.3.3.6 total effective dose equivalent (TEDE) doses for the case in which the iodine spike is assumed to be initiated by the accident are determined to be less than 1.0 rem at the site boundary and less than 0.1 rem at the low population zone outer boundary. 'Ihese doses are a small fraction of the proposed dose guideline of 25 rem TEDE identified in the draft revision to 10 CFR Part 50 (see SECY-94-194).
For the case in which the steam generator tube rupture is assumed to occur coincident with a pre-existing lodine spike, the TEDE doses are determined to be less than 2.0 rem at the site boundary and less than 0.2 rem at the low population zone outer boundary. These doses are within the proposed dose guideline of 25 rem TEDE identified in the draft revision to 10 CFR Part 50 (see SECY-94-194).
s
z=ii
- 15. Accide~t Analyses generator and complete the required actions to terminate the primary to secondary break flow before steam generator overfill or ADS actuation occurs.
ftfven*S when no operator actions are assumed, the AP600 protection system and passive design eatures initiate autornatic actions that terminate a steam generator tube leak and stabilize the
. RCS in a safe condition while preventing steam generator overfill and ADS actuation. The h resulting offsite radiological doses are within the 10 CFR 100 limits.
15.6.4 Spectrum of IlWR Steam System Piping Failures Outside of Containment This section is not applicable to AP600.
15.6.5 Loss of Coolant Accidena Resulting from a Spectrum of Postulated Piping lireaks Within the Reactor Co',iant Pressure Boundary 15.6.5.1 Identification of Causes and Frequency Classification A loss of coolant accident (LOCA)is the result of a pipe rupture of the reactor coolant system (RCS) pressure boundary. For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 square foot.
This event is considered a Condition IV evem (a limiting fault) because it is not expected to occur during the lifetime of the plant t'ut is tvsstulated as a conservative design basis. See Subsection 15.0.1.
A minor pipe break (small break), as considered in this subsection,is defined as a rupture of the reactor coolant pressure boundary (Section 5.2) with a total cross-sectional area less than 1.0 square foot in which the normally operating charging system flow is not sufficient to sustain pressurizer level and pressure. This is considered a Condition III event because it is an infrequent fault that may occur during the life of the plant.
The acceptance criteria for the LOCA are described in 10 CFR 50.46 (Reference 1) as follows:
- 'lhe calculated maximum fuel element cladding temperature shall not exceed 2200*F.
- Localized cladding oxidation shall not exceed 17 percent of the total cladding thickness before oxidation.
- The amount of hydrogen generated from fuel element cladding reacting chemically with water or steam shall not exceed one percent of the total amount if all metal cladding were to react.
- The core remains amenable to cooling for any calculated change in core geometry.
- The core temperature is maintained at a low value and decay heat is removed for the extended period of time required by the long-lived radioactivity remaining in the core.
Revision: 3 owarrevuso6.nos.os3095 May 31,1995 15.6-16 3 Westingh0Use
i l
15.6.3 Steam Generator Tube Rupture Replacement 25 Even when no operator actions are assumed, the AP600 protection system and passive design features initiate automatic actions that can terminate a steam generator tube leak and stabilize the RCS in a safe condition while preventing steam generator overfill and ADS actuation.
s & m--Tes_
The resulting offsite radiological doses for the limiting case analyzed are within 10 C"' idlimits.
4 i
4 4
4 4
. I h,, 15. Accid:nt Analyses hk Table 15.6.3-3 (Sheet I of 2)
PARAMETERS USED IN EVALUATING TIIE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE Reactor coolant iodine activity Accident initiated spike: Initial activity equal to the design basis reactor coolant activity of 0.4 pCi/g Dose Equivalent 1-131 (See Table 11.1-2) with an assumed iodine spike that increases the rate of iodine release from fuel into the coolant by a factor of 500 (See Appendix 15A)
Preaccident spike: An assumed iodire spike that results in an increase in the reactor coolant activity to 24 pCi/g of Dose Equivalent 1-131 (See Appendix 15A)
Reactor coolant noble gas activity Design basis activity (See Table 11.1-2)
Secondary coolant initial iodine 0.04 pCi/g Dose Equivalent activity 1-131 (10% of design basis reactor coolant concentrations listed in Table 11.1-2)
Reactor coolant mass (Ib) 3.39 E+05 Offsite power Available Condenser Lost on reactor trip Condenser partition coefficient for iodine 1.0 E-02 Time of reactor trip See Table 15.6.3-1 j Duration of activity releases (hr) @8 l l
Atmospheric dispersion factors See Table 15A-5 l l
i 1
i l
l Revision: 3 ow .vuso6.aos-os,;o95 l May 31,1995 15.6-64 W Westinghouse i
l
- 15. Accident Analyses Table 15.6.3-3 (Sheet 2 of 2)
PARAMETERS USED IN EVALUATING TIIE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR tulle RUPTURE Steam generator in faulted loop Initial secondary coolant mass (Ib) 94444- 51[ ' M 7, h b N Primasy to secondary break flow See Figure 15.6.3-5 Flashing fraction for break flow See Figure 15.6.3-11 Steam released (Ib) See Figure 15.6.3-7 lodine partition coefficient 1.0 E-02 Steam generator in intact loop Initial secondary coolant mass (Ib) W ').6 0 0* 7,8 b Y#
Primary to secondary leak rate (lb/hr) 130(a)
Steam released (Ib) See Figure 15.6.3-8 lodine partition coefficient 1.0 E-02 (a) Equivalent to 500 gpd at $61.5*F and 2250 psia (fr) .T ,./ ,*.. ...
~
h ow-rev3u sano3-oswis Revision: 3
[ W85tlligh0USe 15.6-65 May 31,1995
- 15. Accident Analyses Table 15.6.3-4 ACTIVITY REl EASED TO TIIE ENVIRONMENT DUE TO A STEAM GENERATOR TUBE RUPT E lodines (accident initi- 'odine spike)
Isotope Rele- (Ci)
E401 1-131
-132
/ 7.5 E+01 33 3.3 E+01 1- 4 3.0 E+01 1-13 3.1 E401 lodines (preaccident iodine spike)
Isotope Release (Ci) 1-131 2.1 E+01 1-132 2.8 E+01 1-133 3.5 E+01 1-134 4.4 E00 1-13 2.2 E+01 i
Noble gases (both cases)
Isotope Release (Ci) l Xe-131m 3.2 E+01 Xe-133m 2.9 E+02 Xe-133 4.7 E+03 Xc-135m 1.7 E00 Xe-135 13 E+02 Xc-138 .8 E00 Kr-85m 3. E+01 j Kr-85 1.2 +02 i Kr-87 1.4 - 1 '
Kr-88 5.1 E4 1 Revision: 3 owevuso6.novos30%
May 31,1995 15.6-66 Westinghouse
- $$$M
- :
- 15. Accident Analyses Table 15.63-5 RADIOL GICAL CONSEQUENCES OF A STEAh! GEN TOR TUBE RUPTURE Thyroid doses (rem)
Case 1 - Accident itiated iodine spike /
Site boundary I 8.5 Low population zone 1.2 Case 2 - Preaccident spike Site boundary 10.2 Low population zone 1.4 Whole body doses (rem)
Case 1 - Accident ' itiated iodine spike i
i Site bou 33 E-01 Low pulation zone 4.4 E-02 Case 2 Preaccident spike Site boundary 3. -01 Low population zone 5.1 E l
l i
i l
1 ow-v3u506.no3453095 Revision: 3 l htay 31,1995
@ W8stinghouse 15.6-67
)
=v
, 10. Accidext Analyses m-l 4
i 15.7 Radioactive Release from a Subsystem or Component Ris group of events includes the following:
- Gas waste management system leak or failure
- Liquid waste management system leak or failure (atmospheric release)
. . Release of radioactivity to the environment via liquid pathways
- Fuel handling accident
- Spent fuel cask drop accident, 15.7.1 Gas Waste Management System Leak or Failure De Standard Review Plan no longer includes this event as part of the reviewx [The AP600 gaseous radwaste system is a low pressure, low flow charcoal delay process which is passive in nature. Failure of the gaseous radwaste system results in a minor release of activity which
'is not significant. M> analysis is provided.
% W 15.7.2 Liquid Waste Management System Leak or Failure (Atmospheric Release Re Standard Review Plan no longer includes this event ashof the reviewafThe AP600
& ^W iquid radwaste system tanks do not contain significant levels of gaseous activit since liquids I
expected to contain gaseous radioactivity are processed by a gas stripper before being directed to storage. The tanks are open to the atmosphere so that any evolution of gaseous activity is continually released through the monitored plant vent. #tranalysis is provided.
15.7.3 Release of Radioactivity to the Environment Due to a Liquid Tank Failure Tanks containing radioactive fluids are located inside plant stmetures. "ic = ks h = 0
. J m.,; eimme ,n cm.. : m m,.
en ,hnt it is cm ~
wev- " L - mod that the snillq redi bh[pccif ' podm. - Tid ent is .ebbe an uncontrolled r e contents of an ent tank in e liquid radw system. suming that the tank is 80 ent full at time of the failure, ere is t 16,000 gallo released. The spectrum of nu ' des is assumed to be presen 'n th tank is conservative assum to be the same as th for the design basis reactor coo t activity presented in S 11.1.
This assump ' rItak s no credit for purification mixed bed demine z in the chemical volume ntrol system nor for the fects o dioactive dec i
l s vf J.c la; pat of t. s cvcat i: t renonsibiliDtef'he _mt="ad I ica-e vpFca 15.7.4 Fuel Handling Accident A fuel handling accident (FHA) can be postulated to occur either inside the containment or I
in the fuel handling area inside the auxiliary buiMing. He fuel handling accident is defined as the dropping of a spent fuel assembly such that every rod in the dropped assembly has its cladding breached so that the activity in the fuel / clad gap een+e7eleased.
A
\
1 a w.rrev3s1507tn no3-o7:395 Revision: 4 '
15,7-1 August 31,1995 3 West lIngh0US8
e r
Insert into subsection 15.7.3 In the event of a tank failure the liquid would be contained by the tank room until drained by the floor drains to the auxiliary building sump. From the sump, the water would be directed to the waste holdup tank. He basemat of the auxiliary building is six feet thick, the exterior walls are three feet thick, and the building is sesimic category I. The ex*.erior walls are sealed to prevent leakage. Thus, it is assumed that there is no release of the spilled liquid waste to the environment. His approach is a departure from Section 15.7.3 of the SRP which states that credit cannot be taken for liquid retention by unlined building foundations. His SRP position is considered to not accurately reflect what would actually happen following a tank rupture.
1 6
h
44
'. 15. Accidsnt Analyses The possibility of a fuel handling accident is remote because of the many adrainistrative controls and equipment operating limits that are incorporated in the fuel handling operations (See Subsection 9.1.4). Only one spent fuel assembly is lifted at a time and the fuel is moved at low speeds, exercising caution that the fuel assembly not strike anythin . ring movement.
'Ihe containment, auxiliary building, refueling pool, and spent fuet pi,r e ( esigned to seismic Category I requirements which assure their integrity in the event of a safe shutdown earthquake. The spent fuel storage racks are located to prevent a credible external missile from reaching the stored fuel assemblies. The fuel handling equipment is designed to prevent the handlino e ipment from falling onto the fuel in the reactor vessel or that stored in the spent fuel prt- ie facility is designed so that heavy objects, such as the spent fuel stdpping cask, can$ot be carried over or tipped .into the spent fuel 141--- .
15.7.4.1 Source Terms
-15.7.4.1. !ai::al Airbara "clccc Searcc Twn; The inventory of fission products available for release at the time of the accident is dependent on a number of factors such as the power history of the fuel assembly, the time delay between reactor shutdown and the beginning of fuel handling operations, and the volatility of the nuclides.
l l The fuel handling accident source term is derived from the core source term given in Table 15A-3 by taking into account the factors below. The assumptions used to define the fuel handling accident initial airborne release source term are provided in Table 15.7-1 along with the derived source term.
15.7.4.1.1 Fission Product Gap Fraction During power operation a portion of the fission products generated in the fuel pellet matrix diffuses into the fuel / clad gap. The fraction of the assembly fission products found in the gap depends on the rate of diffusion for the nuclide in question as well as the rate of radioactive decay. In the event of a fuel handling accident, the gaseous and volatile radionuclides contained in the fuel / clad gap are free to escape from the fuel assembly. The radionuclides of concern are
? !cs 'han 9goble gases ree=' af ee(kryptons fue! rsemMy and xenons) i nven'ery + 'Nived and iodines. e ;sOen radirnuc!!&s assumed
'ha:r Wat @cr 7
nng yn,r n,ir_tirgc3ca in g;c gap,(parerence ) n, ,mm3d:gc3 . jim 2 h,ir_yfe g :a:c7 Ah-r "ne yeon terc i; ; css um ;u riun: of d:c assembly inven:c y :" me gnp_.
15.7.4.1 lodine Chemical Form ,
5 Re .adinc in 0.c gap . . ==med :c exist prhrerily in cuiubinauon witn ccsnm = cesium- l Jodii. ^_ Fmull fr?c' inn M We . odin 0 !': rEumed P bc in dic ciemenmi-fts- I4 unuous that nono nf the irviino ic in ic nrgonicr nre etace icic aiu nu vapnic ""'eriolc in the fue_l-wi@ ahich40 inriine mayw Revision: 4 w=ousnrenms August 31,1995 15.7-2 W Westingh0USS
84 i
Insert into subsection 15.7.4.1.1 Based on NUREG-1465 (Reference 1), the fission product gap fraction is three percent of fuel inventory. For this analysis, the gap fraction is increased to 3.6 percent of the inventory to
. address concerns identified in NUREG-1465 regarding the applicability of the three percent gap fraction to high burnup fuel (i.e., fuel with burnup in excess of 40 gigawatt days per metric ton of uranium).
Insert into subsection 15.7.4.1.2 Consistent with NUREG-1465 guidance, the lodine is assumed to be 95% cesium iodide, 4.85% elemental iodine, and 0.15% organic iodine. There is no organic iodine present in the fuel itself. The organic iodine is assumed to be formed from the reaction of elemental iodine with organic contaminants in the spent fuel pool.
l i
t
yi!!Fl!E
- 15. Accident Analyses b
ni, ,,
15.7.4.1 lodine Behavior in the Gap A Cesium iodide is nonvolatile and the iodine in thi form dissolves in water but does not readily become airborne. It is assumed that the cesium iodide is not entrained in the air bubbles released from the damaged fuel assembly.j)demental iodine is expected to p'. te eut on the relatively cold surfaces of the fuel cladding instead of remaining airbornej A :scugh 4he &=nta! ;adeac k expected "" ;-la:: cc: n ic c:edding and no; bc avaihi:; for -
- nnerhaic y&;g jg is Conservatively assumed that the elemental iodine in the gap remains as an airborne component.
15.7.4.1 4 Assembly Power Level All fuel assemblies are assumed to be handled inside the containment during the core shuffle so a peak power assemblygis considered for the accident. Any fuel assembly can be transferred to th spent fueljpr(; during a core off-load all fuel assemblies are discharged to the spent . To obtain a bounding condition for the fuel handling accident analysis,it is assumed at the fuel handling accident involves a fuel assembly that operated at the maximum rated fuel rod peaking factor. This is conservative since the entire fuel assembly does not operate at this level.
15.7.4.1 .5 Radiological Decay The fission product decay time experienced prior to the fuel handling accident is at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
N 6 Q,u;ce Tarm rne gg== fr- o gnm,ng spr.q nd ; a '
I : is conservatis ly assume- at subsequent to the el handlin nt the is '
s pent fuel pit coo ' ability for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> res 'ng,i sling of the ter in the spent fuel pit. 1 n of the iodine activity released spent fuel pit water . m the damaged f assembly i eleased to the environm t is assu that all of the affi 'ne
,ctivi , oth elemental and - 'culate, is av le for release due to e boili of the poo
' xiiuc somcc icmefer-;~ qFing ._ aws is nmviaM in Tahic is .1 15.7.4.2 Release Pathways The spent fuel handling operations take place underwater. (Because of are first sybbed,b 1 thg cogwater 23 feet in deg This has no effect on the releases of noble gasc$blh tlier,e is a sigmlicant removal ofgodine. Bgspd ongsge scrubbing provided by a 23 foot column of water reduces the concentration of iodirie in the gas bubbles f
by a factor of greater than 500. For the fuel handling acci ent analysis the pool sembbing decontamination factor (DF) is assumed to be.266 which i. ess thanit:rtf-the value supported by tests (see Reference 2). O A A
TA. w y' / 3 3 A Rr $s /f
- fd /27 Y*
mmv3uso7to no3-o7:395 Revision: 4 3 W85fingh0Use 15.7-3 August 31,1995
.. i
- 15. Accident Analyses r
After the gases released from the fuel assemb escape from the water pool, it is assumed that they are released directly to the environmen without credit for any additional iodine removal process. A If the fuel handling accident occurs in the contaimnent, the release of activity can be j terminated by closure of the containment purge lines on detection of high radioactivity. No credit is taken for this in the analysis. Additionally, no credit is taken for removal of airborne iodine by the filters in the containment purge lines.
For the fuel handling accident postulated to occur in the spent fuelgthere is assumed to be no filtration in the release pathway. Activity released from the pool is assumed to pass directly to the environment with no credit for holdup or delay of release in the building.
-Mcm wig oc f J has ting accid (nt, it is assumco a.cic w - lcas of33m m ja $
- ooling ability. water in the poolI su to contain the to uel as bly gar nventory o e which is available for r to the environment as esult of poo
) oiling. pool Qis based on eat input the com let sch d reacto.'
ge_ , - me that becomes airborne dern en nnni hnmn g ic relevad dire ly en the envir 15.7.4 3 Dose Calculation Models he models used to calculate egm2 doses dec ;o ec iriiden af i~'N: =! &
.whole-6,dj d= du. hvi u ir =:rin in 1: rinsed subic gn: --d 'e iahai"sa cf
,cdh are provided in Appendix 15A.
l Table 15.7-1 lists the assumptions used in the analysis.
15.7.43.1 Differences from the Guidance of Regulatory Guide 1.25 The significant differences from Regulatory Guide 1.25 are discussed below.
15.7.43.1.1 Gap Fraction The gap fraction specified in Regulatory Guide 1.25 is 10 percent for short-lived nuclides and 30 peregnt for long-lived nuclides. He assumption used in the fuel handling accident analysis of/ rcent f~ Se ^r" "=! nudii: P= P.:f;;;.ac D =d 1^ ia.w.; far ic '"ng Hve+
-eneis based on review of fuel performance and is conservatively selected.
'f." U.L2 Pc -: Sm. .bbing ac;;;=':29- Facte. w jre ' rubbing oe aT nauon h ui riflea oy negulat Guiuc i.25-f flodine is 3 selection of a deco ation factor of 250, w ' e be most a factor l of two gr e an that provi for in aulatory Guide 1. , 1. ti( conservative onsi ng that Ac .cM3 re GiEd sui'iN' 2 C^""""i"E f?C'O' h 0< S "' 5*
.- -rn ~~- g l
Revision: 4 ow.w3uso7r..mo3.ovi395 August 31,1995 15.7-4 W W85tingh00S8 l
. 15. Accident Analyses 2-15.7.43.lf lodine Chemical Form In Regulatory Guide 1.25 it is assumed that 99.75 percent of the iodine is in the elen: ental form and that the remaining 0.25 percent is in the organic form. A'" gr!: .adinc :n concid-ed !- 'he fud h=d!!n; accide-' 2n1ycic imii- h mc fuc; oasc..$!y 6 an' in 5 1
nrenic form harme $::: cc no 0 3aiuv inaivuo s prc=n' :-'he fue! re.-5!y for ic icdiaa h: r0CCl 'db ,^..'lef-rCIC&ic iv d.c pug} vi;l0" C"'} : $0 hu!! ding, 'he fl amental indine i; [re;-
en renct with orf.an!" mC:crids to fouu vig.uuc audiuc cun pounda. !!av,cics,10 preccee of Snrming nronnig iodinne le time C:;naunung and .o .C. .U"'.0d LGi "G oCCC: '" any significant _
citen! - I 'h0-Mgr ;=rimi nr orentect enneara II; ,.i51 iWO is6uia afiu d.C JCC5dCni).--
'4% e m'% M,k %B57[ M46 A 8'/T Re assumption of "7 feimui uf J.c vd::: Sdn;; !: 2hc form of cecinm 'edW is consistent with the :.cVer =ciden' surcc icia. uscd in ic l055 of ceclan: ;=iden* i:= =dysis in- ~
3 95=ction 15.5 51 eM 9 WWREG ~/V67 15.7.4 3.1. Iodine Dose Conversion Factors M N( h l ,
Re g do.e conversion factors specified in Regulatory Guide 1.25 are taken from TID-14844
- f -n(Reference c 3) andmrthese re recon, aorin;,;g- valuesf=:urs gye ggmygrc:^- areprovidcd derived from in ICRP material PJaciuvu 36 in ICRP JPeference 5) =perscdc,ic Pub: cci:- 212!ue:. n; du,c men,ua.cn f=:e~ "- !CPP 3"
_orn need in ther unt h33dt:ng ig,e 333 %, 1 45.7.43.1." Reica3e vi iocine Actiiity frosu a I;oi in;; Spe-! Fue' "it -
c34atacuvuv reim: s.cchani>>u is not cousiduud in-Regulatory-Guide-fr25-sincc l' sun s ete precence nr en oc,;er enretv-related soent fuel nit cooling function.
15.7.4 3.2 Identification of Conservatisms Ec fuel handling accident dose analysis assumptions contain a number of conservatisms.
Some of these conservatisms are described in the following subsections.
15.7.4 3.2.1 Fuel Assembly Power Level The source term is based on the assumption that all of the fuel rods in the damaged assembly have been operating at the maximum fuel rod radial peaking factor. In actuality, this is true for only a small fraction of the fuel rods in any assembly. He overall assembly power level is less than the maximum radial peaking factor.
15.7.4 3.2.2 Fission Product Gap Fraction 3, l.
The assumption of 4 tree percent gap fraction for the short-lived nuclides is conservative by a factor of-l+or more, depending on the nuclide,(Refereace '
2.
o 'usarrevh1507fn R03-071395 Revision: 4 W W85tiligh0USe 15.7-5 August 31,1995
' 5 - 15. Accident Analyses 15.7.43.23 Amount of Fuel Damage It is assumed that all fuel rods in a fuel assembly are damaged so as to release the fission product inventory in the fuel / clad gap. In an actual fuel handling accident it is expected that there would be few rods damaged to this extent. 15.7.4.3.2A Iodine Plateout on Fuel Cladding Although it is expected that virtually all elemental iodine plates out on the fuel cladding and is unavailable for atmospheric release, no credit is taken for platcout. 15.7.43.2.5 Pool Scrubbing Decontamination Factor for Iodine , 1.?3 The selection of a sembbing decontamination factor of.eStt provides & a factor offwo- %
, conservatism in determining iodine releases based on the pool scrubbing test data (Reference 2).
15.7.43.2.6 Meteorology It is unlikely that the conservatively selected meteorological conditions are present at the time of the accident. 15.7.43.2.7 Time Available for Radioactive Decay he dose analysis assumes that the fuel handling accident involves one of the first fuel assemblies handled. If it were one of the later fuel handling operations, there is additional decay and a reduction in the source term. 15.7.4.4 Offsite Doses
< Using the assumptions from Table 15.7-1, *he c u!=cd r&= -f ac ..;;y d= '^ 'he
' pm'"I*d f=1 hs;d!!ng neelden ' gieca i_a T2M e ' 5 ' 2. nc w!io:c-bcdy dosc duc :a combinntion of immg~:^3 *g 3;c ,;ghie ggg &gd gnd @c jnhq17tiga mr ;;373g,gg g,3;;;g, ,,g __
_ Wg iyrn!d rinen sine in inhalatina ^r ;;;73g77; jggjng;; gg 3333y73g y . ;;j3;;g gggggg77 ;;ng__ ror me inw nonnation 7nne n"'r r he"ndry. He rac"Irino rincec nra licted !a Tah:c 15.2-1 E1^ d"cac Ne t : 0 f=l hand!!ng 70cident nre well uithi;; 3,g gu;jg!!ng jig;; er }n gy;; ;g pace tan 25 ;v rcenn Thren dmce ye miehjn ehn anco criegrig dgnngj * ; pp [g,7,4 15.7.5 Spent Fuel Cask Drop Accident The spent fuel cask handling crane is a single failure proof device. No radiological consequences analysis is necessary for the dropped cask event. Revision: 4 ow.,rev3uso7r.Rc3-m395 August 31,1995 15.7-6 W Westirigt10tise
l ( Insert into subsection 15.7.4.4 the calculated total effective dose equivalent (TEDE) doses are detennined to be less than 1.0 rem at the site boundary and less than 0.2 rem at the low population zone outer boundary.
'Ihese doses are well below the proposed dose guideline of 25 rem TEDE identified in the draft revision to 10 CFR Part 50 (see SECY-94-194).
l [ l f l l l
U !![z::
- 15. Accident Analyses EM --
15.7.6 References
-! "D-iw Al urp_ gg,gg; 7 7;7;,* {c .; 7, g {,, . ;j,, g gg[jg jgg]},r p p,,m,, Iggj D 2. Malinowski, D. D., et al., " Radiological Consequences of a Fuel Handling Accident,"
WCAP-7828, December 1971.
- 3. DiNunno, J. J., et al., " Calculation of Distance Factors for Power and Test Reactor Sites "
TID-14844, March 23,1962.
- 4. " Report of ICRP Committee 11 on Permiscible Dose for Internal Radiation," International Commission on Radiological Protection, ICRP Publication 2,1959.
- 5. " Limits for Intakes of Radionuclides by Workers," International Commission on Radiological Protection, ICRP Publication 30,1978-1981.
/, .$ 3 b 3 *)
< -r- r - DA A
( /%~ 42, " NN/fE G -/W5~ 9Wc DSC, s I l l oksarrev311507fn.R03-071395 Revision: 4 l 3 Westhighouse 15.7-7 August 31,1995 i i
1 1
'e 15. Accidsnt Analyses Table 15.7-1 (Sheet I of 2)
ASSUMPTIONS USED TO DETERMINE FUEL IIANDLING ACCIDENT RADIOLOGICAL CONSEQUENCES Source term assumptions I Core power (MWt) 1972 Decay time (hr) . 100 4-we- te r 100 hours decay Ci) 4 I-131 -4.0-Er07- E,7 6 to f 1-132 .3.4-E+47 2. 2 E + o 6' l-133 4.1 E ; 00- 2,7 6 +o 4 Xe-131m -e.7 C+05 3 ,*p 6 -+- o 3 Xe-133m 4." E ' 'M- C), f 6 -t- o 3 Xe-133 2 " E : 0J f, ( 6 -t- o 7 Xe-135 4.5 C:05- 53,5 E -f-o? Kr-85 2.
- E : 05- f,,!)E-f-o 3
=d= e fmc: aw...L:.m., ... cu.m :45 Maximum rod radial peaking factor 1.65 Percentage of fission products in gap 3,5
!; q " =d (!L- G - -
"~ "ad ed!;icr.0 , .4---
lodine chemical form (a) Elemental (%) G Y86 Organic (%) -e o / [ Particulate (%) A7- Y Source term for gaseous release from the failed fuel assembly (Ci) 1-131 B,o -iHM E+02 I-132 6.ff-hN E+02 1-133 g..o 2-04 E+01 Xe-131ta 2,2 +-26 E+02 Xc-133m 5' 41E+oZ - 2.7 Ex04 Xe-133 g,oyC;42.34 E;03 Xc-135 fg,g+of 5.: Ce Kr-85 3,9g . M+ E+02 Revision: 4 owm,3uso7to.Ro3 07:395 August 31,1995 15.7-8 W W65tingh0US8
1 0=; 9 !
+
- 15. Accidert Analyses n_
l Table 15.7-1 (Sheet 2 of 2) ASSUMPTIONS USED TO DETERMINE FUEL IIANDLING ACCIDENT RADIOLOGICAL CONSEQUENCES Pool decontamination factor 45tt- I33 Initial activity release period (hr) Aff.d 7
-Tin._ : ::=h teilbg h & ;x;,;: (hr) - -
%rce ter F :::g pr' re!:= c' h di=; (Cil-
' u i.37 Ev M -
- 132 i.io t.+ w
'133 iE:^? -
Per' ta;;cff .mu (iiviu; 4.2s', E : ^2 Ti'" O O! pik5 liviiinig as ws.. "'"'ad (hr) 72 - D A ' -- l Imyce p1reieinn enefficient l
' M i n i m i_"-- .. ;;;c; olisme in uv nt finet pi* j,G ErO' c=for tiffuer" (ga') -
Atmospheric dispersion factors See Table 15A-5 (a) Only the elemental formhv& olatile and available for airborne release. 4 4 o%arrev3\1507fn.R03 071395 Revision: 4 W Westinghouse 15.7 9 August 31,1995
- 15. Accident Analyses Table 15.7-2 IVITY RELEASED TO TIIE ENVIRONMEN !
D TO A FUEL HANDLING ACCIDE Nuclide Initial Relea. 'i) Pool Boiling Release (Cl) 1-131 1.6 E00 23 E+02 1-132 1.4 E . 2.3 E00 I-133 ' 01 9.4 E00 Xe-131m 1.3 E+02 N/A Xe-133m 2.3 E+03 N/A Xc-133 2.7 E+G8 N/A Xe-135 5.1 E+01 Kr-85 6.1 E+02 N/A Revision: 4 ow.mv3uson .no3-o7:395 August 31,1995 15.7-10 3 W65tingh00S8
. l
!!-- i
- 15. Accident Analyses
,is -
I Table 15.7-3 - RADIOLOGICAL CONSEQUENCES OF A FUEL IIANDLING ACCIDENT Thyroid doses (rem) Site boundary 0.7 Low populatio ne 4.0 Whole bod ses Dem) Site boundary 0.3 Low population zone 0 o%arrev3\l507fn.R03-071395 Revision: 4 W85%)gh00S8 15.7-11 August 31,1995
1 t insert into subsection 15.7.4.1.1 Based on NUREG 1465 (Reference 1), the fission product gap fraction is three percent of fuel inventory. For this analysis, the gap fraction is increased to 3.6 percent of the inventory to address concerns identified in NUREG 1465 regarding the applicability of the three percent gap fraction to high burnup fuel (i.e., fuel with burnup in excess of 40 gigawatt days per metric ton of uranium). Insert into subsection 15.7.4.1.2 Consistent with NUREG-1465 guidance, the iodine is assumed to be 95% cesium iodide, 4.85% elemental iodine, and 0.15% organic iodine. 'Ihere is no organic iodine present in the fuel itself; the organic iodine is assumed to be formed from reaction of elemental iodine with organic contaminants in the spent fuel pool. 1 i
Insert into subsection 15.7.4.3.1.3 With the change in accident dose methodology to calculate Total Effective Dose Equivalent (EDE) doses instead of thyroid and gamma whole body doses, the use of thyroid dose conversion factors is no longer used. 'Ihe TEDE doses are a combination of the Committed Effective Dose Equivalent (CEDE) doses and the gamma whole body doses. 'Ihe CEDE dose are calculated using dose conversion factors based on ICRP Publication 30 (Reference 5) which include contributions from all significant organ pathways, including the thyroid. ICRP Publication 30 supersedes ICRP Publication 2. t I l
}
@M u 'X Q ,
insert into subsection 15.7.4.4 the calculated total effective dose equivalent (TEDE) doses are determined to be less than 1.0 rem at the sie boundary and less than 0.2 rem at the low population zone outer boundary. These doses are well below the proposed dose guideline of 25 rem TEDE identified in the draft revision to 10 CFR Part 50 (see SECY-94-194). From the Standard Review Plan, "well below" is taken as being 25 percent or less. 1
e l TABLE OF CONTENTS Section Title Page 15.7 Radioactive Release from a Subsystem or Component . . . . . . . . . . . . . . . . . . 15.7-1 15.7.1 Gas Waste Management System Leak or Failure . . . . . . . . . . . . . 15.7- 1 15.7.2 Liquid Waste Management System Leak or Failure (Atmospheric Release) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.7-1 15.7.3 Release of Radioactivity to the Environment Due to a Liquid Tank Failure . . . . . . . . . . . . . . . . . . . . . . . . . . 15.7- 1 15.7.4 Fuel Handling Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.7- 1 15.7.4.1 Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.7-2 15.7.4.2 Release Pathways . . . . . ........ . . . . . . . . . . . . . 15.7-3 15.7.4.3 Dose Calculation Models . . . . . . . . . . . . . . . . . . . . . . 15.7-4 15.7.4.4 Offsite Doses . . . . . . .... ... . . . . . . . . . . . . . . 15.7- 6 15.7.5 Spent Fuel Cask Drop Accident ... ........ . . . . . . . . . . . 15.7-6 15.7.6 References . . . . . . . . . . . . . . . . . . . . ............... . . . 15.7-7 l l : l l i I l Revision: 4 ow.m,3u507t.Ros-o7:395 1 August 31,1995 W Westingh0USe
*e h r.*._._m..~
._7 re-LIST OF TABLES Table No. Title Page 15.7-1 Assumptions Used to Detennine Fuel Handling Accident Radiological Consequences . . . . . . . . . . . . . . . ..................... 15.7-8
!f.7 2 Ac9/!:y !!cicssco to tne Environment aue to a ruet
_ u.....e...3
. . .. __ m uuca..... ... . . . . . ... .w. ,, m. m
'S ' 3 Pmiiningcd Ceregences M ? F' >c' "=E4 Acciuent . . . . . . . . . . . . . . A r l
1 l 1 1 1 l l l l l o%arrev3\1507fn R03-071395 Revision: 4 W Westinghouse August 31,1995
==
o
.' -__=
o LIST OF FIGURES Firure No. Title Page i I l l 4 l 1 Revision: 4 ow=,suso7r..no3-o7:395 August 31,1995 3 Westinghouse}}