ML20094B458
| ML20094B458 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/02/1984 |
| From: | Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| OL, NUDOCS 8408070033 | |
| Download: ML20094B458 (39) | |
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PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101
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C UGENE J. BR ADLEY assoc Ave eassumat copuss6 CONALD BLANKEN
~ s.upoLPH A. CHILLEM6 E. C. KIR K H ALL -
- T. H. M AMER CORNELL PAUL AUERBACH August 2,1984 assasvany essenmak counsak EDW ARD J. CULLEN. J R.
]
THOM AS H. MILLER J R.
8REME A. McKENN A assestaNT CO*)NsEb.
Mr. A. Schwencer, Chief Docket Nos.: 50-352
~ icensing Branch No. 2 50-353 L
Division of Licensing U. S. Nuclear Regulatory Cm mission Washington, D.C.
20555
Subject:
Limerick Generating Station, Units 1 and 2 Information for ContaintNnt Systems Branch (CSB) and Accident Evalcation Branch (AEB) Regarding SGTS Drawdown
Reference:
.(1) Telecon Between L. Bell (NRC/AEB) and D. R. Helwig (PECO),-7/11/84 File:
GOVT 1-1 (NRC)
Dear Mr. Schwencer:
In the reference-(1) telecon, Mr. Bell of your staff requested that we provide revised FSAR pages which reflect an increase to the reactor enclosure design inleakage rate-from 50 to 100 percent..ee volume per day and a doubling of the SGTS fan flow during drawdown of the Unit 1 reactor
-enclosure. In response to that request, I am enclosing draft revised FSAR
- pages.
The information contained on these draft FSAR changes will be incorporated into the FSAR, exactly as it appears on the attachments, in the revision scheduled for August, 1984.
Very truly yours, 1
o i
E ene J. B a 1
' FaIB:pkc Attachment /cc: See Attached Servica List g
8408070033 840002 PDR ADOCK 05000 g
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Lect Judge Lawrence Brenner (w/o enclosure)
~
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' Judge Richard F. Cole-(w/o enclosure)
(I
' Troy B._-Conner,JJr., Esq.
(w/o enclosure)
-Ann;P..Hodgdon, Esq.
(w/o enclosure) 77 LMr.; Frank R. Romano (w/o enclosure)
~
Mr.uRobert'L. Anthony-(w/o enclosure)
~
-Charles'W. Elliot,'Esq.
(w/o. enclosure)
Zori G. Ferkin, Esq.
(w/o' enclosure)
Mr.L Thomas Gerusky
.(w/o enclosure) 547d-
~
- ' Director,iPenna.-Emergency-(w/o enclosure)
' Management Agency
' Angus R. Love, Esq..
(w/o enclosure)
-David Wersan,:Esq.-
(w/o enclosure)
Robert J. Sugarman, Esq..
!(w/o enclosure) 4
- Spence <W. Perry,.Esq.
(w/o enclosure)
. 1
' Jay M. Gutierrez, Esq.
(w/o enclosure)
Atomic Safety & Licensing (w/o enclosure)
' Appeal' Board
~
(w/o enclosure) 1 <-.
A'tomic-Safety & Licensing-
- Board Panel.
-Docket & Service Section (w/o enclosure)
Martha W.-Bush, Esq.-
(w/o enclosure)
'H-Mr.' James:Wiggins' (w/o enclosure)
.Mr.. Timothy.R. S. Campbell (w/o enclosure)
.Es. Phyllis Zitzer (w/o enclosure)
Judge Peter A. Morris ~
(w/o enclosure) r
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I.G3 FSAR TABLE 1.J-4 tront'd)
( Page 2 o f 28 l
ItII!bElU E!.!EQl![USNNA EI944_R 1 P{ACN 99_Tgf!
E T
Design temperature of suppression 220 220 275 2A1
- cramber,
'F Downconer vent pressure loss f actor 2.1R
- 2. 5
- 2. 17 6.21 l
0.0159 0.016
- 0. 00 0
- 0. 0 19 Break area / tota l vent area Calculated maximum pressure aft er blowdown to 44.0 44 40.4 to drywell, s:sig Calculated maximum suppression chamb=r.
- 10. f-29 35.6 25 pressure af ter LOCA b1cwlown, psi n 43 40 35 32 Initial suppression pool temperature rise during IDCA blowdown, *F Leakage rate, 5 free volume / day 0.5 0.5
- 0. 615 at 0.5 45 psig a nd 340*F SECONDAM.CQEAlt#_qM (See FSAR Sect i on 1. 83) controllei controlled cont rolled cont rol led teamage, roof le.s ka ge,
leak. age,
- leakage, Type level release elevated elevated ele va ted release release retrace Construction ke in f orceil lieinforced Reinforced Pel tif orced lower levels conr rete concrete concrete con cr et e Hein f or ce=1 Steel super-St eel super-Steel superA-Upper levels cotw rete supr r-st rtet ure and structure and structure and at ruct ure i n<1 siting si ding siding sidinq g
W Re in f orced Steel deckina Steel decking Stect decking poor concr et e Internal design pressure, psig telow at minoleric 0.25 0.25 0.25 0.25 N
Desian inleakage rate,1 free volume / day at M
10O 100 100 yp
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refueling area provide primary containment for the unit being refueled.
6.2.3.1 Desion Bases I
a.
The conditions that could exist following a LOCA or fuel handling accident require the establishment of a method of controlling any fission products that may leak into 4
b.
The functional capability of the ventilation system to maintain negative pressure in the secondary containment with respect to the outdoors is discussed in Sections 6.5.1.1 and 9.4.2.
c.
The seismic design, leaktightness, and internal and external design pressures of the secondary containment structure are discussed in Section 6.2.3.2 and Chapter 3.
d.
The capability for periodic inspection and functional testing of the secondary containment structure is discussed in Chapter 16.
-4. 2. 3. 2 System Desian 6.2.3.2.1 Secondary. Containment Design l-Thesecondarycontainmentisdesignedandconstructedin accordance with the design criteria outlined in Chapter 3.
All of the major structural walls are constructed of reinforced concrete.
All of the major structural floor slabs and roof slabs are constructed of. reinforced concrete supported by structural Msee{framin toduw o
Iud M i
L She condary containment onesAare designed to limit the in) age to percent o their zone free volume per day a
t-negative interior pressure of 0.25 in, wg, while operating the standby gas treatment system (SGTS).
Following a LOCA or a fuel
' handling accident, the affected zone is maintained at this negative pressure by operation of the SGTS.
The secondary containment zones are identified in Figures 6.2-27 through 6.2-35.
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An analysi of the post-pressure transient in the# reactor enclosureg as performed The length of time following isolation l
signal initiation of the SGTS that the pressure in the secondary l
containment would exceed minus 0.25 in. wg. is +0v6 minutes,6esede 3
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O LGS FSAR h iation aT three minutes.
At two minutes, the SGTS ha
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redu,-Nrea ctor e to a ve pres e, but the additional he of the RJRS an motors ree minu causes t eactor enclosure pressure to' rise again ri t m *
.. before re A ng minus D.25 in. wg. at. - mi utes igure 6.LW' The RERS is specificaLly p'r,ovided to reduce swettor osure ai borne activity post-LOCA, and it is on red desi e to init te the system as soort as Me
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.nt.
this extends the-period in which assume r
or en The guidelines stated in the Standard Review Plan 6.2.3 have been followed in calculating the drawdown time as noted:
1.a.
(1)
The heat transfer coefficients found in Branch Technical Position CSB 6-1 apply to an atmosphere with high-energy blowdown where condensation on the_ _.
primary containment surface is expected.
B e r a n s e _.__ z..z.c3 the drawdown analysis was only eib44iinutes long, the primary containment heat load was calculated as the steady state load during normal operation when there are no condensation effects.
This is accurate because LOCA conditions inside the primary containment will not affect the exterior surface temperature of the 6-foot containment wall significantly in minutes.
For steady state heat load a conserva 've value was assumed.
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(2)
Steady state conduction and convection was calculated.
(3)
Radiant heat transfer was considered.
l b.
Adiabatic boundary conditions were used.
l l
c.
There will be negligible expansion of the 6-foot thick primary containment concrete walls in M nut s.
d.
Inleakage was considered.
l e.
No credit was taken for outleakage.
l l
f.
The analysis was based on the assumptions and delays l
indicated in the acceptance criteria.
g.
Heat loads generated within the secondary containment were considered.
h.
Fan performance characteristics were considered.
l 6.2-41 Rev. 15, 12/82
> ~
LGS FSAR TABLE 6.2-14 (Page 1 of 2)
SECONDARY CONTAINMENT DESIGN DATA
- REACTOR ENCLOSURE VENTILATION ZONES I AND II, AND REFUELING AREA VENTILATION ZONE III Free-volume, fta: Zones I and II 1,900,000 each Zone III 2,200,000 Prermure Normal operation: Negative 0.25 in:wg Post-accident: Negative 0.25 in:wg es/ day (Ean IU':
0.5 air chang /dw r
Leak rate at post-accident pressure:
l.0 & Me
( 7.oruZa JL)
SGTS Exhaust fans - common l
Number: 2 Types centrifugal, single inlet single width (SISW)
Secondary containment atmosphere filtration prior to release to outdoors via-SGTS fans Number: 2 Type:
Zone I and II prefilter, high efficiency particulate air (HEPA),
charcoal, HEPA in RERS followed by'HEPA, charcoal,~HEPA in SGTS l
Zone III prefilter(8), high efficiency particulate air (HEPA),
charcoal, HEPA in SGTS TRANSIENT ANALYSIS Initial Conditions Pressure: negative 0.25 in, wg 1
Temperature: 104*F Outside air temperature 950F j
Thickness of secondary containment wall: 36 in.
Thickness of primary containment wall: 72 in.
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LIMERICK GENER ATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT REACTOR ENCLOSURE DRAWDOWN FIGURE 6.242 REV.15,12/82
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-As described in Section 9.4.2.1, the secondary containment consists of three ventilation zones.
Iones I and II surround the
-primary containment of Units 1 and 2, respectively, below the floor at El. 352 ft.
Zone III consists of the common refueling area above the floor at El. 352 ft.
The SGTS is designed to accomplish the following objectives:
Exhaust sufficient filtered fir from the reactor a.
enclosure ;;....
..1.. ::: or refueling area '
- r.
- "-
- 'to maintain a negative pressure of about 0.25 inches w.g. in the affected volumes during secondary containment isolation (see Section 9.4.2 for discussion of the secondary containment isolation signals) b.
Filter the air exhausted to remove radioactive particulates and both radioactive and nonradioactive forms of iodine from the following areas:
1.
Reactor enclosure (Zone I and Zone II) l 2.
Refueling area (Zone III) 3.
Main steam isolation valve leakage control system discharge area, following filtration by the reactor enclosure recirculation system 4.
Primary containment during purging and ventilating 5.
Discharge from the high pressure coolant injection (HPCI) barometric condenser c.
Ensure that the failure of any component of the filtration train, assuming loss of offsite power, cannot impair the ability of the system to perform its safety function d.
Remain intact and functional in the event of a safe l
shutdown earthquake (SSE) m L
Rev. 15 12/82 6.5-2 t
F5-w7 DRAET e.
Automatically start in response to any one of the following signals:
l 1.
LOCA signal as described in Section 9.4.2, LOCA signal takes precedence and blocks all signals from the refueling area.
If no LOCA signal has been received, the SGTS exhausts either the refueling area or the reactor enclosure based on the ordered signal priority falling in 2 through 5.
2.
High radiation level in refueling area exhaust air l
3.
High radiation level in reactor enclosure (Zone I or Zone II) exhaust air 4.
Low differential pressure in the reactor enclosure (Zone I or Zone II) 5.
Low differential pressure in refueling area (Zone III)
(The SGTS fans can also be started manually in the control room by tripping the refueling area isolation system or the reactor enclosure isolation system or by setting an SGTS filter in cooldown mode and starting an SGTS fan.)
f.
The design bases employed for sizing the filters, fans, and associated ductwork are as follows:
1.
Each filter train is sized and specified for
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treating the incoming air-steam mixture at F
11,000 cfm maximum and 1350F for drywell purge (drywell purge is discussed in Section 9.4.5).
The SGTS fans are sized for 3000 cfm maximum flow at 7 inches w.g. static pressure)Wsfn -rg, t.sg.5 tc owcATie0.
TM. %T5 Fled Wiw EE Afyp_OKtHNTE.Ly (2. Boo C.FM AT 7 INS W.4.
Sferic, ra455 n t V0nTMT IEEr_5 ed oPEcAfton R.EES is 14tTLMEP 3 L. j The system capacity is maintained with all filters fully loaded (dirty).
MiOOTE.S AFTEP 'THE STAET o V:. p o d N.r-t o E C T ).
3.
For high efficiency particulate air (NEPA) filters, maximum free velocity does not exceed 300 fpm with 6.5-3 Rev. 15 12/82
g LGS FSAR R),f n Y d i IC,I) / Q 9 A b0 J
provided for use in conjunction with the SGTS filter trains.
Each fan has a controllable capacity of 500 to 3000 cfm, which is sufficient to restore and maintain the Unit " 2;f Crit f reactor enclosure # 12: ;; I ;;d lit or the common refueling area T
... iiifat the required negative pressure in relation to atmospheric pressure during secondary containment isolation.
The air flow varies in response to secondary containment differential pressure controls, which modulate a control damper in the run-around bypass duct provided for each fan.
The SGTS is actuated automatically in its safety-related mode of operation.
Both SGTS filter trains are maintained in the open position.
Upon receipt of a secondary containment isolation signal (Section 6.5.1.1.1.e), both of the SGTS fans are started and the associated controls are activated to open or modulate appropriate dampers and valves so that the system function is accomplished.
Following the initial fan start, the operators may elect to place one of the SGTS f ans in -the standby position.-
~
For its non-safety-related mode of operation (described in Section 6.5.1.a), two redundant 100% capacity drywell purge fans are provided for use in conjun'etion with the SGTS filter trains.
Each fan has a capacity of 11,000 cfm which is sufficient for the drywell purge operation.
The SGTS is manually actuated for its non-safety related mode of operation.
If one of the SGTS filter train isolation valves fails closed,'
the redundant filter train is automatically placed into operation.
If one of the SGTS fans fails to establish flow, because of either fan or fan damper failure, the standby SGTS fan automatically starts.
9.4-2 The SGTS is shown schematically on Figure C.4.
Specific SGTS component design parameters are shown in Table 6.5-1.
1 The equipment and materials conform to the applicable requirements and recommendations of the guides, codes, and standards listed in Section 3.2.
Conformance with Regulatory Guide 1.52 is discussed in Table 6.5-2.
i I
(
Compnents for each SGTS train are designed as discussed in the following paragraphs.
)
l 6.5-5 Rev. 15 12/82 i
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IAS FSAR ThaLE 6.5-9 Ciace 1 c.( Ja Ile5TitOMENTATION FOR ESF ATMOSPHERE CLEAIEIP SYSTB08 f
EERIDBL11tRS PER ERP W
- L 5.1-t lusfallstaEA11tal PAQVIBRD IM idRE BRA 2 ant a 8 COIST901 20G4 ENERGRIICY SEMI M jacATICE IACAL RERDOR/AIARN CQWTBQL ACOM PAREL gg23 333 FR.Esat AIR Onit inlet or outlet Flow rate (Indication) teot providedtes Flow rate 31ou pate indication latica tion at outlet at outlet Unit inlet or outlet Flow rate (recorded Flow indication Low flow low ilow alare indication, high and at outlet alarataato3 taeaea low alarms Iow flow alarm testaste8 Electrical heater Status indication Status indication IE/A Not provided in the control strouble alare room in the control I
roomt ( F A soace between heater moe rature (indication Indication onlytes N/A Indica tion and profilter high and low alarm eienalet ontwte8 Space between heater Temperature (indication Provided N/A Not providedt*8 and profilter high and low alarms, trip alara siqualel Profilter Pressure droo (indication, -
Indication oniv Indication Indication 0
high alarm signall (satse8 onlyteatses onlytestaes 9
l First IEEPA Pressure drop (indication.
Indication only same as same as high alarm signali testsea sagsg e st s ea sagsgestaea J
First BEPA Pressure drop (recorded Not providedtsel Not 3:rovided Not providedtses j
indication)
(sel Soece between adsorber Temperature (two stage Not providedte8 Not provided Not providedtes j
and second REPA high alarm signal) tea i
Space between adeorber Temperature (two atace Three stage hich Same as Same as and second HEPA high alarm elonal) alarm and SGIS SGIS indacation Second REPA Pressure drop (indication. -
Indication oniv Same as Same as hiqh alara s1gna11 8888aoa scqseaaaeea solst e a t a e 3
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125 FSAP T API.E 6. 5-9 fcont'dl
'(Page 2 of 39 -
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loot provided Not provid)
Fen (ootional hand switch and toot provided status indication)
's Fen Hand switch, status aProvided Provided Provided indication N
Valve / damper operator (optional status Not provided Not provided Not provid indicat lan t velve/damoer 03erator Status indication Provided Provided Provided Deluge valves Hand switch, status Manaal valves Same as same as indication Indication oniv SCTS SGTS Deluge valves Hand switch, status Alarm (883 Same as Same as indication SGTStas sa;3tssa sveten inlet to outle*t Summation of pressure Provided Provided Provided droo across total system, high alarm signal.
(83 Regulatory Guide 1.52, ANFI-r509, and standard Review Plan Table 6. 5.1 were arisins11v issued af ter the Limerick gvstem design and therefore were not specifically consi1eral ici the Jaalon.
(P)
The S3TS flaw rate is variable to draw down and maintain the reactor enclosure /sofuelina area at a negattve 0.25 in.wa.
Local flow indication does not provide meaningful information in
' M 4
terms of system operability. Flow indication is provided in the contrat ro w where this L
information, an addition to the reactor enclosure /ref ueling area pressure dif ferential indjcators, as available for operator evaluation of system perfctmar ce.
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Iow flow switen operates on loss of flow oniv.
(*5 The SGTS does not operate during normal olant operation. Maximum MT3 flow cccurs only fcr the W
first M ainutes of drawdown. As thermodynamic equilibrium in approached withis: the reactor g-C enclosure during isclation, the SGTS flow decreases to the desion inicak age rate, which is IcS9
'p L
D 2,,7 p than half the rated caoacity of 3GTS fans.
y (58 The REFS does not operat e durina norm! pl. ant ogwrit iors.
The kHS tiow as t ecircial..ted wit his N
the reactor eviclosure during isolation and as not directiv release I tc the? enviranment.
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I1 0 24
ff-l-Y7 LGS FSAR The assumptions and calculation methodology used are as follows:
a.
Spiking factor The activity released from the fuel to the coolant as a consequence of reactor scram and vessel depression was based on measurements during plant shutdowns (Ref 15.6-2).
It was shown that for a 95 percentile probability, a total of 7 Ci of I-131 is released to the coolant for every 1,Ci/see of prespike I-131 release.
This conservative ratio was applied for all the iodine isotopes for the dose analysis.
The prespike iodine releases were those that correspond to a 0.35 Ci/sec noble gases release, a design basis accident assumption.
b.
Iodine concentration in coolant The total iodine released from the fuel to the coolant was assumed to take place in a span of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, resulting in continued buildup of coolant activity during that period.
The coolant activity during 0-2 hours was assumed to be constant and equal to that at the end of the first hour.
The coolant activity during 2 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> was assumed to be equal to that at the end of 3-1/2 hours.
This is a conservative assumption, since the rate of increase in coolant activity decreases with time.
c.
Partition factor It was assumed that 100% of the activity in the coolant that flashed into steam remains airborne and that 105 of the activity carried by the coolant water into the secondary containment becomes airborne (corresponding to a conservative partition factor of 0.1).
d.
Activity in secondary containment and released to the environment ac The secondary containmen volume was assumed to consist of hee 6e reactor enclosuref ;..d th f
-^- rr
- f :?irg cr:2 7 as oiscussed in Section 15.6.5.5.1.2.
The activity airborne in the secondary containment was assumed to be uniformly mixed in th:.i..;P;;; pu, uy"-
4j thr
"with auc--Li..ud" airflow of 120,000 T6T5- ]s'g.o and a 954 efficient filter. Secondary containment air i released to the environment via the SGTS at the rate of one M secondary containment volume change per day.
The SGTS filter has an efficiency of 994. The SGTS draws air from the RERS exhaust.
The activity airborne in the secondary containnent and the activity released to the environment /are presented in Tables 15.6-2 and 15.6-4, respectively.
15.6-5
n-.
^
concentration throughout the volume resulted relatively quickly.(minutes) indicating almost complete mixing in a room sized volume.
Convective air mixing due to thermal gradients is also discussed in NUREG-1575 (Ref. 15.6-8) for hydrogen.
A concentration gradient of less than 0.25% was found for
- a. temperature differential of 50F.
It is expected that convective air mixing effects would be similar for iodine and hydrogen.
However, it is also expected that the temperature differentials within the reactor enclosure would be larger than 5'F, and also of many and varied geometrical configurations.
Considering that any containment leakage sources are also likely to be heat sources,' mixing by convection should be a major effect.
Forced air mixing will result from two systems, local ESF fan coolers and the RERS.
Local coolers will remove heat and mix the air in selected rooms of the reactor enclosure as indicated in Section 9.4.5 and the RERS will mix the air as described in Section 6.5.1.3.
The fan coolers.will be operational for heat removal within minutes after the accident.
The RERS, which is initiated 3 minutes after the accident, will mix the air d
between the various compartments of the reactor enclosure and the HEPA, and charcoal filters will remove particulates and iodines from the air.
SRP 6.5.3 Section III.2.C indicates that mixing credit for small annulus type secondary containments typical of some PWRs and Mark III BWRs will not be given, and that for large BWR reactor enclosures a positive period (implying unfiltered exfiltration) is not assumed.
In accordance with this guidance, and as explained in Section 6.5.3, the Limerick evaluation assumes that the mechanisms discussed above will ensure the assumed 50%
mixing within the large reactor enclosure at all times during the period when the reactor enclosure pressure is above minus 1/4 inch, as well as when it is below.
However, it will also be conservatively assumed that there is unfiltered enfiltration at air changedI/ day i
l
" ;;;?, in addition to the SGTS aust, during L
periods when the pressure is aboveiminus 1/4 inch w.g.
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c.
Plateout within the reactor enclosure I
l Iodine and particulates in the reactor enclosure
[
atmosphere will also be removed from the air by many Rev. 15 12/82 15.6-18 I
.... - - _. -. _ _ _ _ _ _ _ _ -...... _ ~,,...,
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LGS FSAR reduce the amount of airborne activity available for release to the atmosphere.
f.
As-discussed in Section 6.2.3, no bypass leskage around the SGTS filters is assumed for this analysis.
J g.
Because each unit is equipped with redundant containment hydrogen recombiners, as discussed in Section 6.2.5, calculation of hydrogen purge doses is not necessary.
. h.-
Leakage from the MSIV-LCS:
The MSIV-LCS routes any leakage through the MSIVs to the steam tunnel where it mixes with the steam tunnel air before being filtered cen$edtat and mixed by the RERS and then exhausted by the SGTS.
paith
-The maximum permissible leakage is specified in the Limerick Technical Specifications and is.:ksee-een the 11,5 spetsefh per valve conservatively assumed for this safety analysis.
Leakage past the inboard MSIVs is assumed to begin immediately after the accident to simplify this analysis.
In reality a delay would be associated with this release.
A total of 444rscfh has been assumed for the four main steam lines.d This MSIV leakage is in addition to the 0.5 percent per day containment leakage.
Wa The fission product activity in the reactor enclosure at any time (t) is a function of the leakage rate from the primary
-containment, cleanup in the reactor enclosure, and the volumetric
' discharge rate from the reactor enclosure.
Upon receipt of appropriate signals, the reactor enclosure ventilation isolation valves. isolate the reactor enclosure atmosphere in 3 to 5 seconds.
This rapid closure time prevents possible uncontrolled escape of radioactivity.
Upon reactor enclosure isolation, the RERS is. designed to circulate the reactor enclosure air to provide iodine removal by the charcoal filters and a delay mechanism ~whereby radioisotopes are retained in the reactor
. enclosure and undergo radioactive decay rather than direct escape through.the SGTS.
'A further function of the RERS is to provide thorough mixing of the recirculated flow to ensure that the SGTS L
cannot extract an unmixed quantity of radioactivity.
Any fission product removal effects in the reactor enclosure, such as plateout, are neglected.
However, the effects of decay are considered.
A mixing efficiency of 50% has conservatively been assumed in the analyses, although a higher efficiency is espected.
The system removal efficiency is designed to be in excess of 99%
removallof all forms of iodine and 0.3 micron or larger Rev. 23, 08/83 15.6-20 i
F va--
-...--n--,.-.n,__n--n__v.,_,_
.nnn-
j v-
?YifS1[;h h O'
OMC particulates.
The SGTS has a design flow of one-he+f air change per day of the reactor enclosure.
The SGTS draws air from the exhaust of the reactor enclosure recirculation system.
The following equations describe the activity buildup in the reactor anclosure due to primary containment leakage.
dAE,=-A A
- lAD+ Ac t ) Act (15.6-3) c where Ac, Activity in reactor enclosure at tim-t, Ci
=
SGTS vent rate from reactor enclosure, hr-1 A,,
=
Recirculation removal rate, hr-1 A,,
=
Unfiltered releases from reactor enclosure during A,3
=
drawdown phase, hr-8 Plateout removal rate, hr-1 A
=
P Aca A
+ Ars + Ass + A rt P
The solution of Equation 15.6-3 is A
+
~*
c2 *
(A +A -Ac2)DF 3
g c
where DF Iodine decontamination factor for leakage
=
c through cracks Activity contributed from recirculation leakage of the ECCS is modeled as follows.
Following a LOCA, 50% of the core iodine 15.6-21 Rev. 16, 01/83
tm m Q [ifY
!NN hJ h'ha3 _
)
J e
Considering that approximately 40% of the released liquid flashes to steam, it is conservatively assumed that 40% of the released iodine activity is airborne initially.
However, as a result of lateout and condensation effects, only 50% of the activity
(
s p'nitially airborne remains available for release to the i
environment.
As a consequence of reactor scram and depressurization, additional iodine activity is released from those rods that experienced cladding perforation during normal operation.
\\
Measurements performed (Ref 15.6-2) at operating BWRs during 4
reactor shutdown have been used to develop an analytical model for the prediction of iodine and noble gas spiking as a consequence of reactor scram and vessel depressurization.
Because no measurements have been obtained during a pressure transient as rapid as the LOCA, it is difficult to predict the actual release rate from the fuel as a consequence of iodine spiking.
It is, therefore, arbitrarily assumed that 100% of the spiking source term is released during the time period that 40%
of the discharged coolant is flarhing into steam.
It is also assumed that plateout and condensation removes 50% of the airborne iodine activity in the primary containment.
The total activity airb,orne in the containment is presented in Table 15.6-17.
15.6.5.5.2.2 Fission Product Transport to the Environment The leak rate from the containment to the reactor enclosure is 0.5%/ day, where 50% mixing efficiency is assumed to occur.
Release rate from the reactor enclosure to the environment via a 95% iodine efficient recirculation filter and a 99% iodine efficient SGTS. filter is of the reactor enclosure volume per day.
The ac,tivity buil in the reactor enclosure is presented j
in Table 15.6-18.
The integrated isotopic activity released to the environment is less than that presented in Table 15.6-19.
AApfe i
15.'6.5.5.3 Results 1
.?
/
15.6.5.5.3,14 Cffsite Doses 1
The radiological exposures resulting from the activity released to the environment as a consequence of the LOCA have been determined for the design basis and realistic cases.
The design 15.6-25 Rev. 15 12/82
O l
y ff-$ v 7
((% @ jk $
. b ]O h.i."n [#du -
TABLE 15.6-2 (Page 1 of 2)
INSTRUMENT LINE BREAK ACCIDENT: PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSES DESIGN REALISTIC BASIS BASIS ASSUMPTIONS ASSUMPTIONS
'I.
Data and Assumptions used to Estimate Radioactive Source from Postulated Accidents A. Power Level NA NA B. Burnup NA NA C. Fission Product Release NA None from Fuel. fuel damagedi q
g D. Release of Activity by Table 15.6-F Table 15.6-4 Nuclide to k Envlw.me.nt E.
Iodine Fractions li Organic NA 0
- 2, Elemental NA 1
'3, Particulate NA 0
F. Reactor Coolant Activity 15.6.4.5.1 15.6.4.5.2 Before the Accident II.
Data and Assumptions Used to Estimate Activity Released A.
Primary Containment Leak NA NA Rate Su day )
100 100 B. Secondary Containment Release M
SR Rate (%/ day)
C. Valve Movement Times NA NA D. Adsorption and Filtration Efficiencies (SGTS SYSTEM) 15 Organic lodine 99 99 (2) Elemental iodine 99 99 (3) Particulate iodine 99 99 (4) Particulate fission 99 99 products E. Recirculation System Parameters ill Flow rate (cfm) 60,000/ unit 60,000/ unit (2) Mixing efficiency 50 50 (3) Filter efficiency 95 95 F. Containment Spray Parameters NA NA (flow rate, drop size, etc)
G. Secondary Containment Volume (ft3)
Reactor enclosure, Unit 1 1.8x106 1.8x106 m
^
e r-
=
a....
_c
)
t, 6
.,g) 4 LGS FSAR TABLE 15.6-2 (Cont'd)
(Page 2 of 2)
DESIGN REALISTIC BASIS BASIS ASSUMPTIONS ASSUMPTIONS M_,m
.- e
__- o _
...r H. All Other Fertinent Data NA NA and Assumptions III. Dispersion Data A. EAB 'LP2 Distance s mi 731/2043 731/2043 B. X/Os for Time Intervals of
- 1) 0-2 hrs - EAB 2.9x10-*
1.2x10-*
s2) 0,-8 hrs - LP2 4.0x10-5 2.0x10-5 31 8-24 hrs - LP2 2.9x10-5 1.6x10-t 3
4.I-4 days - LPZ 104x10-5 9.0x10-*
s5) 4-30 days - LP2 5.4x10-*
4.2x10-*
IV.
Dose Data A. Method of Dose Calculation Section 15.10 Ref 15.6-3 B. Dose Conversion Assumptions Section 15.10 Ref 15.6-3 C. Peak Activity C:----t :t::.;I Table 15.6-)3: Table 15.6-)W:5 D. Dosel ___,
- y Table 15.6-7 Table 15.6-7 EAc7aR Eucto suas
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hy my i
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.a Ua IABLE 15.6-6 INSTRUMENT LINE EREAK:
ACTIVITY RELEASED TO THE ENVIRONMENT (1)
REALISIIC Ab'AHjlj M919E S-2 3M 2.5_jg 5-8 HM I-131 1.w h:58 x 10-5 6.ir a:::sts x 10-s 5.n2266 x 10-*G I-132 2,4tt:26 x 10-*
r.e4 %:49.x 10-*
6.093-96 x 10-8 I-133 f.o? k$4 x 10-*
9.r5=5r5 x 10-4 y.c4%-69 x 10- 85 I-134 6.7th48 x 10-*
/.(,7 M x 10-*J i.122r60 x 10-* 4 I-135 J.ti m x 10-*
7,qi N-te x 10-*
i.rlA44 x 10- F.f' TOTAL 1.1 M x 10-43 3.13 2+49 x 10-a 3 00 h-14 x 10-e1 U919H 8-24 HBS 1-4 DAYS 30 DAYS I-131 4.9) 592 x 10-8'1 4.10 a 40- :
- 0. o 0.0 1-132 2.34 *vM x 10-F ic
.i. -i G :
to
- 0. 0 I-133 3.s9 h40 x 10-, se g ; ;;- e 3.o 0.0 s
I-134 9,9r.k44 x 10-8 88
-0.32 - M ::
- 0. 0 0.0 I-135
- 4. if 3:46 x 10-* '#
.10 : 2 ; ;, 15 c.c 0.0 TOIAL
- l. l 'l. y.yH x 10-
- S
-1.0 0 ;; ^4 : : C.C 0.0 i
=
(1) Units for activities are in curies.
-,----.-.-n.-,
me=~m 3fg2{
y TABLE 15.6-7 INSTRUMENT LINE FAILURE: RADIOLOGICAL EFFECTS DESIGN BASIS ANALYSIS WHOLE-BODY INHALATION DOSE (rem)
DOSE (remi Exclusion Area Boundary
.r.e z.2 3 (731 meters hr dose)
Wx 10-7 m x 10-5 Low Population Zone 3.37 5
(2043 meters day dose)
-M*9 x 10- ?
x 10-s REALISTIC ANALYSIS WHOLE-BODY INHALATION DOSE (rem)
DOSE (rem) 6.96
- 2. 7r Exclusion Area Boundary
&mML x 10-s W x 10-*
(731 meters hr dose)
M x 10-s N x to-o Low Population Zone (2043 meters day dose) e 9
~-'
-,.s-.
FS -& *'I h[
~
h / Li 3
TABLE 15.6-13 (Page 1 of 2)
LOSS-OF-COOLANT ACCIDENT: PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSES DESIGN REALISTIC BASIS BASIS ASSUMPTIONS ASSUMPTIONS I.
Data and Assumptions Used to Estimate Radioactive Source from Postulated Accidents A. Power Level 3458 3458 B. Burnup NA NA j
C. Fission Products Released 1005 O
__ ~-.
from Fuel (fuel damaged) 16 iq ' ~
~
Nu cl i'de +cW hirdnWh l J t ) * * * **
- l ' ' ' ' i ' T'.Ta bl e _15.
- p. Release of Activity,_oy Table 15.6-J
~ " " ~ - ~
E.
Iodine Fractions Ili Organic 0.04 C.01
'2s Elemental 0.91 0.99
- 3) Particulate 0.05 0.0C F. Reactor Coolant Activity Section Section i Before the Accident 15.6.5.5.1 15.6.5.5.$
II.
Data and Assumptions Used to Estime.te Activity Released A. Primary Containment Leak 0.5 0.5 Rate excluding MSIVs (t/ day)
B. Secondary Containment Release RateU)
- f..t.5
- 1) DuringgDraw?down IPoo thee ci:
1.
1 (0-tcvS min)
(2) After Drawd wn 2 5898$ cfm NA Z..Z.5 (98=5 min - 30 days) io0'SGt/ day )
(3) Unfiltered Release r7.504485 cfm NA During Drawdown too(30%/d sy )
t.2-5 N min )
(4) Release Rate for NA two7.921/ day Realistic Analysis C. Valve Hovement Times NA NA D.!SGTS Adsorption and Filtration Efficiency (%)
99 99 E. Recirculation System Parameters (1) Flow rate (cfm) 6.0 x 10*
6.0 x 10*
l (2) Mixing efficiency 50%
50%
(3) Filter efficiency 95%
95%
F. Containment Spray Parameters NA NA Rev. 15, 12/82
FS-&YI LGS FSAR p S,r-7._7 WUO$
f TABLE 15.6-13 (Cont'd)
(Page 2 of 2)
DESIGN REALISTIC BASIS BASIS ASSUMPTIONS ASSUMPTIONS (flow rate, drop size, etc)
G. Containment Volumes (ft8) 3.976 d3. '? 75 Primary.
4-tC3 x 10s m x 10s Secondary (Total)
Unit I reactor enclosure 1.8 x 10*
1.8 x 108 H. All Other Pertinent Data and Assumptions tion +
- NeC PIStV lis
/t. 5 $CF//
11 5 Sc f n III. Dispersion Data %< feA. Va//e, A. EAB/LP2 Distance (m) 731/2043 731/2043 B. X/Os for Time Intervals of (1) 0-2 hrs - EAB e.9 x 10-*
1.2 x 10-*
(2) 0-8 hrs - LP2 4.0 x IO-*
2.0 x 10-s (3) 8-24 hrs - LPZ 2.9 x 10-9 1.6 x 10-s (4) 1-4 days - LPZ 1.4 x 10-s 9,0 x 10-e (5) 4-30 days - LPZ 5.4 x 10-*
4.2 x IO-*
IV.
Dose Data A. Method of Dose Calculation Section 15.10 Ref 15.6-3 B. Dose Conversion Assumptions Section 15.10 Ref 15.6-3 C. Peak Activity C - - '
_ _h Table 15. 6-M' Table 15. 6-HL '7 Pr2 * "
i Containment
/4 es Table 15.6-20 Table 15.6-20 C
(
L.D. TL k. Aeu,$,h Rec c s n L,,,,,
y,,,,, _ _ g 7;. s g i s s., e Neto '.
(_tY Aerwoosa Tee R.nwa causn_e. pwoou
))
hE ts a ;;MS mooses Crsse_ secn@
6.2. 3. >. t T'm % poses Q~t-# E. MS ER dMidEW DArsE0
~
oA A PWtpouh9 Tt Hre_ e s A f'g-ext a w r E t.y F
Minotes.
_..m.__
c
/
IAS FSAR TABLE 15.6-14 IASS-OF-COOLANF ACCIDENT:
ACTIVITY AIRBORNE IN PRIMARY CONFAINMENT(a 3 DESIGN BASIS ANALYSIS, ISOTOPE 2 BRS 8 HRS 1 DAY 4 DAYS 36 M J
I--131 3.95?4006
- 3. HAY 4006
- 3. 6 W i 00Aj' 2.76?i00/*
2.533+005 E -132 3.2?Ve006
- 5. 3.5Y e 00*i 4.15'.'
- 00?
1.337-006
.000 I--133 R.3655006 6.0464006 4. 01 ? ' 00/C.
3.6?3400'i 3.224-004 1--134 2.145+006 1.H4Ye004 5./00-00:li 9.756-027
.000 1--135 6.5794006 3.50?4006 6.520800%
3.379+002 1.135-026 KR-83M 6.7294006 6.925500*5 1.6114003;-
2.2/8-OOY
.000 KR--85 1.418+006 1.4164006 1.410*006[
1.385+006 1.189+004 KR-85M 3.290*007 1.2??>007
- 1. OHV t OO.t 1.5576001
.000 MR--fT.'
2.713+007 1.OP94006
- 1. 474 i OOC 1.4YO-015
.000 KR--88 6.812'00?
1.5894007 3.?30'00**'
O.530-003
.000 KR--- 8?
4.045-004
.000
.000 C
,000
.000 XE131M 8.925600*5 8./82f00*5 H. 41 ? 8 00*r 6.Y31*005 1.293+005 XE133M 4.659+006 4.???e006 3.1604004 1.3194006 3.036+002 XE-133 1.9224003 1.8576000
- 1. 694 6 00f E 1.1214 OOH 3.121+006 XE135M 2.6504005 3.160-00?
1.088-02Q
.000
.000 XE-135 1.592+00H 1.007:00H 2.?664007-1.213&005 2.430-016 XE-137 6.175-002
?.446-O50
.000
.000
.000 XE-138 4. 66*i F 00*5 1.052-002 4.26'7-02 1
.000
.000 TOIAL 5.183+000 3.38f,t00H 2.1459003,'
1.187e008 D.693+006 4
d I;'Mj s-su (93 Units for activities are in curies.
O
%w i.
fi y
l N ld i
A e
N D
t Na ill la
M S FSAR TABLE 15.6-15 LOSS-OP-COCLANT ACCIDElff:
ACTIVITY. AIRBORNE IN REAC'ICR ENCLOSURE (s 3 DESIGN BASIS ANALYSIS IsororE 2_BER A_EEg 1 mr 4 MYS
-h 1--131 3 9364002 3.7484002 4.229+502 4.14h4bO2 3,703+001 I--132.
3.1964002 5.172&OO1 4 H25-001 2.0j0-010
.000 I--133' O.104+002 6.6324002 4.6626002 5.43H+001 4.714-008 I--134
?.078+002 1.7914000 6.717-006 1.464-030
.000 I--135 6.374+002 3.393+002 7.577+001 5.072-00?
I.659-030 KR-83M 2.945&OO3 9.616*002 3.981+000 6.511-01?
.000 KR--85 6.206+002 1.966+003 3.4R4+003 3.9614003 3.400+003 KR-85M 1.440FOO4 1.803+004 2.6906003 4.452-002
.000 KR--87 1.1884004 1.429+003 4.134-001 4.260-01H
.000 KR--88 2.982+004 2.207*004 8.102+002 2.441-005
.0(X)
KR--99 2.646-007
.000
.000
.000
.000 XE131M 3.907+002 1.219+003 2.0784003 1.981+003 3.697+002 XE133M 2.03?+003 5.969+003 H 566+003 3.7714003 8.681-001 XE-133 8.412+004 2.578'005 4.185*005 3.203+005 8.926+003 XE135M 1.160+002 4.387-005 2.688-023
.0(X)
.000 XE-135 6.967+004 1.3986005 7.327e004 3.468+002 6.948-019 XE-137 2.703-005 3.674-033
.000
.000
.000 XE-138 2.0424002 1.461-005 1.055-025
.000
.000 TOTg 2.186+005 4.5074005 5.1034005 3.309+005 1.273+004 q:rg (t
asa Units for activities are in curies.
es p!M Rev. 15, 12/82 g-,
a[5 h
9 6xN
fS - 6 V7 g
Ty 747 Y
41o w
4 o
22:
a M
M M
MMM M
MMM M
M b OH "V d 4NQQh h4 N b
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TAELE 15.6-18 PEAc 7&R E A/C LOS uRE LOSS-OF-COOLANT ACCIDENT:
ACTIVITY AISBCRNE IN SECONDstgMpet a 3 p,.y JtIALISTIC MMYSIS J3GTORE 2 E32 W
_1 DNY 4 DAYS 30 DAYS I--131 2.288-00?
?.237-002 2.104-00?
1.597-00?
1.465-003 I--132 1.905-00' 3.085-003 2.399-005 7.7.50-015
.000 1--133 5.090-00?
4.167-002
?.442-00?
2.205-003 1.963-01?
I--134 1.218-002 1.051-004 3.284-010 5.488-035
.000 I--135 4.214-002 2.245-002 4.179-003 2.166-006 7.201-035 KR-83M 1.418-001 4.628-002 1.916 004 3.134-016
.000 KR--85 1.637-001 5.183-001 9.I85-001 1.0414000 8.965-001 KR-8'W 5.483-001 6.863-001 1.024-001 1.694-006
.000 KR--87 4.866-001 5.855-00?
1.693-005 1.74'i-02?
.000 KR--88 1.260+000 9.321-001 3.421-002 1.031-009
.000 KR--8?
1.170-011
.000
.000
.000
.000 XE131M 2.195-002 6.850-002 1.167.-001 1.113-001 2.0,'6-002
~XE133M 1.061-001 3.106-001 4.457-001 1.962-001 4.516-005 XE-133 3.820+000 1.170+001 1.899:001 1.4544001 4.051-001 XE135M 2.959-003 1.119-009
- 6. Wi4-028
.000
.000 XE-135 3.148+000 6.313+000 3.3094000 1.566-002 3.138-023 XE-137
'1.227-009
.000
.000
.000
.000 XE-138 1.005-002 7,190--010 5.191-030
.000
.000 TalAL 9.855+000 2.0734001 2.3974001 1.593+001 1.324+000 cm3 Units for activities are in curies.
h t
s
./
LG3 FSAR DRAF~T',
~
TABLE 15.6-g LOSS-OF-COOLANT ACCIDENT:
ACTIVITY RELEASED TO THE ENVIROIStENT4 a 3 li'E A L /S 72c AA/,4L Y3/ S BDSMBe>44966 ANALYSIS ISOTOP3 2 MRS 8 HRS 1 DAY 4 DAYS M
I--131 1.669-004 5.6h3-006
- 1. e. 4.'- 005 5.51H-005 1.579-004 I--13?
1.037-00.".
2.1Y3-006 4.?01-007 3.2?3-00?
1.062-018 I--133 3.809-006 1.154-005
'?.152-005 2.77?-005 2.752-006 I--1'34 1.YS5 006 6.353-00'/
5.S26 00Y
- 1. ~72/- 014
.000 I--135 3.367-004 7.817-006 7.244 006 1.656-006 8.588-010 KR-83M 1.5Y5-002 5.000-00?
1.207-00?
4.332 00N 6.8Y1-017 KR--85 1.402-003 1.779-001 1.0144000 6.1776000 5.037+001 KR-85M 5.228-002 3.586-001 4.476 001 5.80?-002 9.115-007 I
KR--87 6.237-002 1.135-001 1.020-002 2.642-006 2.667-023 KR--88 1.27? 001 6.367-001 3.Y46-001 1.223-00?
3.542-010 KR- -89 6. 9'?Y- 004 7./16-014
.000
.000
.000 XE131M 1.883 003 2.345-00?
1.315-001 7.1Y1-001 2.804+000 l
XE13.3M
?.173-003 1.10?-001 5.440-001 1.9244000 1.218+000 XE-133 3.284-00; 4.07?t000
?,1884001 1.053+002 2.053+002 XE135M 3.626-003 1.0H4-004 3.622--011 2.160-O?Y
.000 XE-135 2.040-001 2.642: 000 6.783+000 3,8894000 1.709-002 XE-137 1.328-003
?.786-01?
.000
.000
.000 XE-138 1.785-003 3.??O-004
?.101-011 1.4R0-031
.000 10fAL 9.195-001 H.1916000 3.1224001 1.180400?
2.597+002 (a3 Units for activities are in curies.
hw a
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FS-6 9'I DRAFT LGS FSAR TABLE 15.6-20 LOSS-OF-COOLANT ACCIDENT:
RADIOLOGICAL EFFECTS DESIGN BASIS ANALYSIS WHOLE-BODY INHALATION DOSE (rem)
DOSE (rem)
~i 0 7*80~I
- 1. 6 x 80 Exclusion Area Boundary (731 meters hr dose) 5.1. 10 ;
- 1. ',0 10 --
l
~
Low Population Zone J.~1 4.0 d 80 (2043 meters day dose) 4,44
-. 01 x 10 :-
l REALISTIC ANALYSIS WHOLE-BODY INHALATION DOSE (rem)
DOSE (rem)
Exclusion Area Boundary
/, 2 x fo
/. 6 < /o '
-6 (731 meters hr dose) 3.27 x 10-e 1.15 : 10 '-
-S
-7 Low Population Zone 4, # e so
- 6. o n fo (2043 meters day dose) 2.00 r 10-*
3.01 : 10 '
Rev. 15, 12/82 m
f; 73-GY?
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LGS FSAR L' ' 1 TABLE 15.6-22 LOSS-OF-COOLANT ACCIDENT:
CONTROL ROOM DOSES DESIGN BASIS ANALYSIS THYROID SKIN BETA WHOLE-BODY (rem)
(rem)
GAMMA (rem)
OPERATOR DOSES 2.07 ; 10-5 3-r64 T 72 x iM 44.3 y10-3
~;i. G 2.2 x;6$
d Rev. 18, 03/33
' An
. 7,.
a 6 J TABLE 15.6-27 (Page 1 of 2)
LOSS OF COOLANT ACCIDENT:
SEQUENCE OF RTS FOR RADIOLOGICAL CONSEQUENCE ANALYSIS il Time Events and Assumptions 0
- Instantaneous RG 1.3 source term assumed.
- Instanteneous suppression pool DF assumed.
- No leakage to atmosphere assumed due to transport time for activity to travel.from core to outside atmosphere.
18 Sec.
SGTS is initiated.
- Activity leaks from containment with leakage pathway DF.
- Transport of activity from leak points (penetration rooms) to exhaust and exfiltration points (outside walls) diffusion and convective mixing result in an assumed 50% mixing in the reactor enclosure.
- Elemental iodine pla_teout = 2.75/hr on reactor building surfaces O P TO D F = 5 0.
- Reactor eng_losure) air e giltrates unfiltered at l
pf0MFM (oNti alt changep7 day).
- SGTS exhausts M cfm through 95S. RERS filter train and 99% SGTS filte train.
lMSEE% NEXT FAAE +
2.8003 l
3 min. w RERS is initiated l
Etap O F-Acetogey
- 17. ;,1;;:_: : ::
- _T; rr. ',_;_ ; ; _ Tr' rt 'r-l I'Z E.m__ iii;' 2_ ?!ZT_'Z ::: : '.f EE.m 121..T "_'" ;',7 7 5 E Ekb50EU A IEMd'o((35Asch UdAs.A 5-[~~Y4 in. W.3
~
- 50% mixing in reactor enclosure guaranteed by RERS operation and natural transport, diffusion and mixing mechanisms.
RERS filtration (95%) of reactor enclosure air.
Iodine plateout on reactor enclosure surfaces up to L
l DF = 50.
SGTS exhausts efm through 95% RERS and 99% SGTS.
\\t6C) l Rev. 15, 12/82 L
l
W FS-6 S
SAR g of n
)@[se TABLE 15.6-27 (cont'd)
(Page 2 of 2) 2.2 5 I
M ain.
Reactor enclosure reaches -1/4 in, w.g.
M
- Unfiltered exfiltration ceases.
W
- SGTS exhausts g efm through 95% RERS and 99% SGTS filters.
//D F
- LOP (ME 9( ArEo0T ors Re.Acttg.EAc-toEJLE SOILF+cE.5 op To PF = 60.
- 50% wi as s ecs eaccosm Dx-
"{~o p><tQRAL T(?_AQSf 9CTj 9l F F 0 5 1o 0 ; M 0---
MigleoA MecaMisss.
NcrTE.S ".
(O 'TtW SEadEh of EVENTS ftErcEI.-r A9 Ac row fEAcroit E0C L OS.OltLE OW00@@ TiHE_ OF-d.M Hs OJCES CSEC* top 6.'2 3.~2.)), gggy THE. WCA DOSE.5 me v__ MSAR.dA-ro/Ety Bass _o 04 N SE E0MLE_ OF EMS @Sigy Qg A Re# cts ecwsa,_i_ o%aooea ms,_ op l
N % IM A'ffAY 3 M it0 0 T E.5..
Rev. 15, 12/82 t
00ESTION 480.22 (Section 6.2.3) h[
FSAR Section 6.2.3.2.1, page 6.2-40, states "An analysis of the post-LOCA pressure transient in the secondary containment will be performed to determine the length of time following isolation signal initiation of the SGTS that the pressure in the secondary containment would exceed minus 0.25 in, wg."
Provide the results
- of this analysis of the pressure and temperature response of the secondary containment to a loss-of-coolant accident (LOCA) occurring inside the primary containment, and describe specifically how each of the guidelines of SRP Section 6.2.3 Item 11.1 has been followed.
RESPONSE
w 12E/cCoA EdLC5cy Section 2.3.2.1 has been changed to provide the results of the post-LOCA secondary containment pressure transient analysis.
The LOCA radiological analyses in Chapter 15 have been changed to account for the radiation released from the secondary containment during the time that the precsure exceeds minus 0.25 in, wg.
In addition to the pressure transient analysis, a detailed review has been performed to identify potential leakage paths from either the primary containment or the secondary containment to the common refueling area.
This review resulted in the following changes which ensure that no leakage paths exist:
a.
A vent path from the reactor well to the reactor enclosure was added b.
Normally closed valves on the reactor well skimmer drain lines were added.
g Periodic tests are performed in accordance ith the plant Technical Specifications to verify that the Jecondary containment inleakage is less than percent of its free volume per day at a negative interior pressur of 0.25 in. wg.
tOD g, ()Rodl4MS IMP _, 6 E 6d N Ar>E. To ASSU M DM
-ren a tseco ru a PATg Ttwe%s Tac b9%Q SYSTE%
BsT@ e Ew Twe-ho E@or hi rcIN (bah o inG-. &G'hcxOYL. E OC1.c%Q c4..
8W C_.c.O tp Astr/ Cc.dTAtGM MTg 480.22-1 Rev. 21, 06/83
g g y,
=....
OUESTION 480.24 (Section 6.2.3)
The FSAR states in Section 6.2.3.2.1 that the reactor enclosure is' designed to limit the inleakage to 50% of the reactor enclosure free volume.per day at a negative interior pressure of 0.25 in, wg. while the SGTS is operating.
Verify that this stated inleakage limit applies also to the refueling area, or provide a separate inleakage limit for the refueling area in terms of a percentage of the refueling area free volume per day (reference SRP Section 6.2.3 Item II.3.b and BTP CSB 6-3 Position B.4).
RESPONSE.
Theinleakagellimitof50percentofthezonefreevolumeperday at a negative interior pressure of 0.25 in.wg. while the SGTS is operating aboo applies to the refueling area.
Section 6.2.3.2.1 has been clarified to reflect this fact.
w 6
/Gr k kalaye 4;,,;6/ po put of Me zo>>e m&me pv a(sy 4., 4 de xce/x escAswes.
^
J 0
480.24-1 Rev. 11, 10/82 7
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