ML20004G121
| ML20004G121 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/17/1981 |
| From: | Swart F PUBLIC SERVICE CO. OF COLORADO |
| To: | Kuzmycz G Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0588, RTR-NUREG-588 IEB-79-01B, IEB-79-1B, P-81168, P-81186, NUDOCS 8106290228 | |
| Download: ML20004G121 (23) | |
Text
_ _ - - - - - _ _ _ _
puhuc sen* e company ce cdwtdo 12015 East 46th Avenue,' Suite 440; Denver, CO 80239 June 17, 1981 Fort St. Vrain Unit No.1 P-31158 Mr. George Kuzmycz, Project Manager Special Projects 3 ranch Division of Project Management Office of nuclear Reactor Regulation
-U.S. Nuclear Regulatory Cmanission Washington, D.C.
20555 Docket No. 50-257
Subject:
Environmental Qualification of Class 1E Equipment IE Bulletin 79-01B
Reference:
P-80350, 10/3/80
Dear Mr. 'Wzaycz:
/
The following information is being subnitted to assist you in the review of PSC's responses to IE Bulletin 79-013:
1.
Summary of FSAR and Other Documentation Pertaining to High Energy Line Breaks and Design Basis Accidents 1 and 2.
A.
Design Basis Accident Number 1
" Permanent loss of Forced Circulation (LOFC)"
The analysis of this accident is contained in section 14.10 and Appendix 0 of the FSAR.
B.
Design Basis Accident Number 2
" Rapid Depressurization/ Blowdown" The analysis of this accident is in section 14.11 of the FSAR.
8106290 2.M
P-81168 Page 2 June 17, 1981 C.
Response to DRL Question 6.1 (Supplied as Attachment 1 to this letter)
D.
Qualification of Fort St. Vrain Safe Shutdown Equipment for Steam Environment Resulting from Pipe Ruptures Gulf General Atomic Report - GA-A12405 published fiay 30, 1972.
E.
Evaluation of the Consequences of Postulated Pipe Failures Outside of the Reactor Building for Fort St. Vrain Unit No.
1 This was provided as Attachment C to kiendment 25 of the FSAR.
2.
Summary of DBA1 and DBA2 and Environments The opening discussion pertaining to DBA1 and DBA2 are quoted from the FSAR and supplied as Attachment 2 to this letter.
Temperature Environments Comparison of the DBA and steam line rupture temperature tran-sients are provided by Figure 6.1-1 of Attachment 1 to this letter.
It can be seen from this figure that the Steam Line Rup-ture Temperature environment envelopes the DBA Temperature Envi ronment.
Steam line rupture temperature transient curves for distances of 20 feet and greater from the leak source are provided by Figures 6.1-2 and 6.1-3 of Attachment 1 to the. letter.
Steam Line Rupture Temperature Transient curves for distances of less than 20 feet were provided by figures 3 and.4 of Attachment D to our letter P-80350 dated October 3,1980. These curves are provided as Attachment 3 to this letter for ease of reference.
Radiation Environments The following is excerpted from pages 7 and 8 of our letter P-80250.
h_.
P-81168 Page 3 June-17, 1981
" Radiation:
There are no radiological concerns directly associated with a high energy line break at FSV. That is, the process fluid (steam or feed-water is not contaminated. _ To postulate a radiological incident DBA
- 1 " Permanent Loss of Forced Circulation" and DBA #2 " Rapid Depressurization/ Blowdown" were considered. DBA #1 provides the worst case radiological conditions, but the overall radiological concerns are minimal.
Complete details of this accident may be found in Section 2.1.6.b (Attachment C) of P-79312 (Swart to Varga) dated December 1979.
In summary, the peak doses in the Reactor Building following DBA #1 are as follows:
Peak Gamma 180 Day Accumulated Location Cose Rate Time of Peak Dose (REM)
Reactor Building 1.4 R/ hr 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 400 In conclusion, the reactor building will be accessible for short-term operations following such an accident. The accumulated doses indi-cated above would have no operational effect on the Reactor Building equipment."
As indicated during our phone call of June 9,1981, the above doses represent the worst case in the reactor building. The 400 REM dose includes the contribution from the gas borne activity in the reactor building along witn the contribution from the activity of the gas in the refueling penetrations.
Equipment at other locations in the reactor building would be subject to radiation fields from gas borne activity only. Therefore, the resultant accumulated dose would be less than 400 R84.
3.
How to Determine Temperature Transient for Each Equipment Item.
The following fields from Enclosure 3 (P-80350, October 3,1980) define the time temperature qualifications for components.
Field Description LOC (35)
Defines general environmental location.
That is, either turbine building (TB2) or reactor building (RX2) steam line break environments TEST-DIST(10)
Defines the time-temperature curve utilized for qualification of equipment
m P-81168 Page 4 June 17, 1981 NOTE:
Components at a distance of 20 feet or greater from a steam line were qualified to the 20 foot curve.
Items between 15 and 20 feet were qualified to the 15 foot curve. The same logic holds for the remainina curves.
4.
Basis For Not Qualifying Equipment to Category 1 of ilVREG-0588 As indicated during our phone call of June 9,1981, NUREG-0588 was written for LWR technology and postulated accidents.
It is PSC's position that the only items in NUREG-0588 that apply to the Fort St. Vrain HTGR are the environmental condi tions out-side of containment.
It does not seem prudent to commit to a document in whicn the vast majority of the requirements do not apply to Fort St. Vrain.
)
l.
PSC is not taking exception to qualifying equipment environmen-tally, we are stating that it is our intent to continue qualifying equipment to the accident environmental conditions ap-plicable to Fort St. Vrain. These accident conditions and our environmental qualification programs have been presented to, and reviewed by, the flRC Staff many times.
It is PSC's position that l
l thepresent program is adequate for providing environmentally qualified equipment.
Very truly yours, a
brtW u
Frederic E. Swart Nuclear Project Manager FES/l1EN:pa Attachments 1
i l
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4 A=endment 25 P-81168 Attachment A Page 6.1-1 Page 1 of 13 6.0 ENGINEERED SAFETY FEATURES 6.1 Oues tion:
Identify all Recctor Protection and Engineered Safety Feature equipment and components (e.g. =ctors, swi:chgear, cables,
filters, pump, seal) located in the reactor or turbine buildings which are required to be operable during md subsequent to a less-of-coolant (depressurization), feedwater line break, or a steam line b reak accident. Describe the qualification tests which have been or will be performed on each of these ite=s to ensure their availability in the resulting environments (i.e. helit=, te=cerature, pressure, humidity).
Answer; Environ = ental steam tes ts have 1 een perfor=ed to verify that all Safe Shu:down Cooling Equipment located in the reactor and turbine buildings will operate during and folicwing a design basis acciden: (DBA),
feedwater line break or s:eam pipe rupture.
A listing of the Safe Shutdown Ccoling Equip =ent is presented in Table 6.li..
The test requirements and descrip:ica of the test facili:y, as well as :he complete resul:s of the environ = ental steam :ests are presented in Gulf General Atomic Report GULF-GA-A120a3, " Qualification of Fort St. Vrain Safe Shutdown Equipment for Steam Environment Resulting from Pipe Ruptures," dated P.ay 1972.
~he resul:s of':he tests show that all components tested have been qualified either unprotected or protected. The components which required scee procaction or =inor design change in order to survive the tempera:ure transien: following a steam pipe rupture accident are listed in Table 6.1-2.
A more de: ailed g
description of the required changes is given in GULF-GA-A12045.
With the planned incorporation of these changes, it is concluded that the environmental steam test program has successfully demonstrated that the Safe Shutdown Cooling Equipment has suf ficient high-ce:perature endurance to survive any anticipated steam oipe rupture accident in the reactor or turbine buildings.
The temperature transients used for the environ = ental steam tests were analytically determined. The results of that analysis led to the conclusion that a ccid reheat steam pipe rupture in the reactor building and a hot reheat steam pipe rupture in the turbine bailding are :he worst postulated accident conditions. Although the reactor building atmosphere temperature is considerac.ly higher for a DBA (rapid depressurization of the pri=ary cooling system) than for a steam leak accident, che reverse is true for the surface te=peratures of the components. The reason is that l
the heat transfer coefficient for staam is much larger than for helium.
Because the surface temperature of a component and not the building a:=osphere l
temperature deter =ines the temperature endurance limit of the cc=conent, l
- he cold reheat steam pipe rupture is repre.:. ctative of the worst case and was selected for the environmental test program. The results of the comparative analysis between the temperatures of a DBA r.nd a cold reheat steam pipe rupture are shown in Figure 6.1-1.
P00RDHINAL
L 5
I P-81168 Amendment 25
' Attachment 1 Attachment A Page 2 of 13 Page 6.1-2 s
CONTEMPT-G computer code was used to analy:e the reactor and turbine building atmosphere response following a steam pipe rupture accident.
This is a modified version of the original CONTEMPT code which has been adapted for high-temperature gas-cooled reactors. For the analysis the building pressure was assumed constant because the building pressure relieves through the louvers which open at three inches water gage g
pressure. The results of the analysis are shewn by a f amily of curves in Figure 6.1-2 and 6.1-3 for the reactor and turbine buildings, respectively.
The complete analysis and the rationale for selecting these two sets of curves for the environmental steam testa are given in the referenced report. Each curve represents a te=perature transient for a specific distance from the postulated steam leak. In both the reactor and turbine buildings the minimum distance from the steam leak of any Safe Shutdewn ecmponent is approximately 20 ft.
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P00R ORGINAL 1.
1 P-81168 Amendment 25 Accach=ent A Page 3 of 13 Page 6.1-3 r
Table 6.1-1 SAFE CE7!DOW 0:0L2NG EQUIP ENT L13T 25 Syeces Ina trumen ta tion ins tr'mentaeion or or or Equipment Pmetion Notes Equiosent 4'anb e r to e st ien i
l 'aactor Sids Buffer
?so loope-equip-Suller He recir:ula*i PD10-2143 8e System nant for only one cor bypass valve PDV-2163 iRaactor Sidg loop noted Racirculator inlet IV-21?3 Reactor 8144 block valve HS-2133 control Roon.
Racirculator inlet NY-21213
,Kaactor 8143 block valve HS-21213 control Roos
?eo loops-equip-l 24aring. ace r pmps.l l Reset r 51d3 Searing P-2101 for only sne,-
P-2101S
{ Rasetor 313 Water.
ment React-r 31Ja Sys tem loop noted P-2106 3
q"'
Surge tank ezergencV LO-21245
- Raaccor 31ds drain concrol LV-11
- 45 Raactor Sids Norsal surge tank LC-2135 Reactor Sids level controla LV-21J5-152 Reactor 3114 Surge tank level LSL-2137 Reactor 3143 I
switch 5 isolatien LV-2137 Reactor 31d3 valve Bearing water pump RS-2131-1.213 Control Roca switenen g
Bearing vater make-P-2105 Raaccor sids up pumps & switches P-2108 Reactor Sids 15-21331 Control Room ES-21394 Reactor Sids low flow bypane bear + T3-21333 Reactor Sida ing water make-up TV-21333 Raactor Bids pW Low flow bypass F5L-21297 Rasetor Sids (bearing water pumpa) TV-2129 7 Reactor Sids l
Ol Turbine Turbine water re-LC-21119 Rasctor Sida soval low-pressure,,
Water LV-21119 Raetter 31d3 Samov 11 separator 'ackup s
drain controle (to l turbine water drainf tank)
I Drain controle (to ' LSH-21113 Rasceor 3144 reactor building l LV-21113
!2aa:torBldg Nle j
munp) l LT-21118 g
../'
P00R BR K
P-81168 A=endment 25 Attachment A Page 4 of 13 Page 6.1-4 m
i Table e.1-1 (continued) 25 System Ins tr*mancation Ins trzentation or or or Equipment Fection Notes feutoment Vu-s e r
_ location Turbine Pressure relief PS-21123 Reactor 3143 Vater turbine water drain PV-21123 Reactor 3144 Ramavel header (continued)
Turbine water drain MV-21277-1,253 Sanctor 31-4 tank equalisatien Standby pep s cart 13H.1122 Reactor Sids (turbine water removal pu F Normal pu:no start 17-21129 Reactor 31ds (turtine water
- 3'-11129
?aaccor 3144
.d removal pus:o)
Normal turbine water 10-21114 Reactor 3134 renaval tank level 17-11114 Reactor 3144 controls Backup turbine wate: 13-111 %
Reactor 314; removal tank level
- 7-21123 Raactor 313g controls.
Tuthine wate r re=o-
?-1133,33 Reactor 3144
/
vel pua:ps & switrhes ES-21111 C.:ncroi Rocs ES-21112 Centrol Roos "g
Equalization line
,Ev-21277-4,5, Kaactor 31ds o
valve to euroine 667 w& tar drain tank Circula-Tour circulators Bearing water supply F3-2133 Reactor Sids tor bear-
-equipment for isolation ing water enly one circu-control lacor noted-typ.
Main drain control FDT-2175-1 Reactor 3143 ical of four FDC-2175 circulators PDV-2175 Steme/ water drain FOT-2179-1 Raactor 31d4 control FOC-2179 FDV-2179 FM-1179 P00R OR'GM
P-81168 hend=en: 25 Attachment A Page 5 of 13 Page 6.1-5 Tabla 6.1-1(continued) 25 System Instrueantation
[ Instr mentation et or i or Equipment Ftmeties Votes h uitse9t I
'fu:s e r
?.oes tion Circulator Four circulators Circulator C-2101 Pasctor 31ds Veter
-equipmen t for Turbina only one circu-Circulator water 13-2109 Rasctor 3143 lator noted-tuihine block E7-2109-1.2 Reactor Sidg typical of four valves circulators Water cuttine on-SC-2109 Control han trol valve EC-2109 Rasetor sid; 57-2109
?4 actor 3143 Eigh-pressure separ-
- C-21:0 3
?.eactor Sid; ator level control
'.V-2120 3 Reactor 814g Circulator serrica 15-2137-1 system isolati:n thru -n-4 Control Room valves, circulator 23-21192. E g
braka and static
-21203 Control Roon seal a5-2137-1 thru -7 P4 actor aids 37-21192-192
-2120 %11,-2
?maccor 31dg.
Emergency :andensate 33-21257 paactoc 31d3 ad=tanton valve to 37-21237 Raactor Sids circulator water turbine Plant Procactive
?:IS-21149 Fasetor 31dg Spetas (P?5) cir:u-
?::5-211.51 74 actor 3133 lacor trip inputs
? DIS-21153 Reactor 31d3 PDIS-211/3 Rmactor Sidg PDIS-21175 Reactor 31dg FDIS-21177 Raaccor 3143 PDIS-21319 Esactor 31d3 FDIS-21321 Reactor Sidg PD15-21323 Reactor 31d3 Emergency Two loops-equip-Emergency condensate 35-2237 Control T on Condensate sent for only one header admission 57-2237 Turoine 314g Beede r loop noted velve to staan generator Pee &acer Two loope-equip-Feedveter sensure:: mon T*-2205 utbine 31dg Control sent for only one control m -2005-1 outrol Room loop noted FC-2005 outral Roce m-2005-2 outrol Room FM-2205-7 ontrol Room F7-2005 purdine31d3 l
P00RBR M
P-81168 Amend ent 25 Attach =ent A Page 6 of 13 Page 6.1-6 s
Tabla 6.1-1 (centinued) 23 System Instrematation b erunentation or or or F.quipment Fa ction Notes Yo.11-men t Tud o r toestion Two loops-equip-l Superhascar tempera-TI-2225-1 :n-4 Raaetor 31d3 Steen Cenerator sent for only one cure indicator T!-2223-I to-4 Control h, control loop noted 3f-2225-1, -2 control Room T2-2225 Control Room Stoma generator PT-22129 Turbine Sids g
pressure control FC-2229 Control Room 7V-2229
!arbine Sids P2-22129-1 Turbine 3143 PI-22129-1 Control Room FM-2229 Control Room PI-22129 Control Roon 3ypese block valve 37-2293 Turbine Sids 33-2293 control Room Superheater outlet 37-2223 Turbine 31ds seen deck valve 35-2223 Control ? con Emergency Teo loops-equip-E.sergency condensate iT-2239 Rasctor Sids Condensate sent for enl/ oes flow control Dt-2239 Ccuerol Room loop noted FC-2239 control Room TV-2229 Reactor Sids Emergency condensate 33-2291 Control loca block valve EV-2291 Reactor Sids j
Steam Cold raheat isola-35-2249 Ccatrol Roca
%nerator tion (inist)
EV-2241 Reactor 31d4 Cantrol (Raheater)
Cold reheat isola-15-2251 Control Room tion (islec)
SV-2251 Reactor 31dg Cold reheat isola-
?Y-2243 Reactor 31ds tion (1 ale t)
PM-2243-1 Control Room FC-224 3 Control Roon 5 teen Two loops-equip-Raheat bypass pres-FT-2267 Turbica 31dg Cenerecor amat for only one sure control FC-226 7 Control Room Control loop noted PV-2267 Turbine Sids (Rahe ste r)
Stesa
- <o loops-equip-Raheater bypass EV-22131 Turbine Sidt i
Generator zent for only one pressure control r.5-22131 Control Room i
Control loop noted (Raheat er; Raheater (outlet)
E7-1253 Turbine 31ds stop check valve 35-2253 jcontrol Room l
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l P00R ORGM
P-81168 Amendment 25 Attachment 't Attac.hent A Page 7 of 13 Page 6.1-7 Table 6.1-1 (continued) 25 System Ins trummatation Ins crmen tation of or or Equipment Pa ctica Notes Equipeant Number Location h ector FC27 Cooling F 4e F-4601,13,2,23 Raactor 31d3 Plant Cooling Cooling water out-
"I-4637-1,38-1 Reactor Sids Ol Water let temperature TI-4637-3,38-3 Reactor 31d3 TT-4437-1,3 8-1 Control Room TI-4637-1,33-1 Control Room Hydraulic Hydraulic pumps P-91012, 3%
. Reactor 31ds Systee 8 drau11c pump
,H5-9101-1 Control kom 7
switches 135-9103-1 Control Roca i
Instrument Air H3-3211-1,-2, f
l Control Room Ins trument:
Compressors and j
-3 2l Air aftercoolers C3201 i Turbine Eld 4 C32013
' Turbine Bids C3203 Turbine Bid; l
Tire Vacart Fire water pump P-4501 fire Water Pu:sp Ecuse RS-4504-2 Control Roos PS-4504 Turbine 31d4 1.5-4504-1 Turbine 31d;
!.5-4504-2 Turbine 31dg
$l Enstne driven fire F-4501-5 Fire Water P mp pg and controls House RS-4504-1 Control Room Control panel for.
engine driven pump l
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?00R OR M L
Amendment 25 P-81168 Attachment A Page 6.1-8 Page 8 of 13 Table 6.1-1(c::atinued) 25 Systen Inecrmantation Inetramaeation-or or or Equipeast hection Notes fouissent Sumer tocation Diesel Fire water backup FIC-4236 Turbine 3144 Cenerator to seretce water for FCV-4256 Tuttina 3144 Coolers diesel generator coole rs F5-4266 Turbine 31d4
%l FIC,4266 Tursice 3144 FCV-4266 Turbine 5143 Diesel Fire water backup None Generator to ir.acrusent air Air Cooler coole rs circula-Fierpe A,3.C Circulating Vater ting Water Makeup Fizp Souse Makeup Circulating water 2*-4102-1 Control Room sakeup pu=ps S 15-4102-2 Control Roca switches l
Service Service Wate r pumps F -4201,2,25 Service a..er Water and :entrols Fump Souse 85-211-1,2,3 Control Room Service Isolation to decay HV-4225 Turbine 3144
)
Water heat removal ex.
MS-4225 Ccuerol Room chanser Isolation noe.ats..w EV-4257 Turbine 31ds tial turbine water H3-4257 Control Roca dervice gl F5V-8211-1 Turbine 31.sg Isolation to instru-75V-3211-2 Turdine Stig amat air coolers F5V-4211 3 Turbine Bid; Temperature control TCV, */,!E-Turbine Sidt valve for instru-4234, 4235, Turbine Bids gl ment air compressors 4274 Diesel steerator TIC-4266 Otesel Cen. Rooms coolers ta=perature
!!C-42'2 7 Diesel Cen. Roome instrunancation and TIC-4269 Diesel Cen. Rocca control TIC-42 70 Diesel Cen. Room Diesel generator g TCV-4256 Diesel Cen. Rooms coolers tagerature TCV-4257
'01esel Cen. Roou control valve TCV-4261 Diesel Cen. Room TCV-4270 Diesel Can. Roorts l
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c P-81168 Amendment 25 Attachment A Page'9 of 13 Page 6.1-9 Tela 6.1-1(continued) 23 System Iastrummatation be trimenta tion or or or Equipment F metica Vo*es Yo 21? tan t
%-s e r taceefen Service Diesel generator TE-6266 01esel Sn. Roots
' deter coolers temperature II -257 Oiesel Gen. Roor.s (continues outlet sensors TE-4269 Otesel Cen.1toce.s t
TE-4270 Diesel Oen. Roome Isolation from emer-15-4221 Coners1 Roos gency circulating 37-4221-2 Tartice aids water supply S tandby Mesel oil taansfer F-9201%.5X Olesel Gen. Roc +
Cenerato r pumpe and cosersis d3-9299 Otesel Cen. Roc ia 011 6 Air E5-92100 Diesel Oen. Roci:n g
Diesel goveratar at:
C-9201.15.2 Otesel Cen. Rooes cog resso s and 25 controls ES-92101-1 -2,0iesel Cen. Rooms 02-1.-2 Auxiliary boiler P ail.04X Turbine Sldg fuel oil feed pumps 7-l!.C13X Turbine 31d3 and switches SS-342 7 Turtise 31dg ES-8428 Turbine lids P00RBR M
P-81168 Amend =ent 25 Attachment A Page 10 of 13 Page 6.1-10
-Table 6.1-2
SUMMARY
OF COP 2ECTIVE ACTION PLXiNED FOR SAFE SHLTDOWN CCOLING EQUIP:ENT TO WITHSTMID STEAM PIPE RUPTURE ACCIOE::T Planned Corrective Af fected Test Tat No.
Component
. Action Cemronent PIC-4256 Pneumatic Pressure Con-apply insul-PIC-4256 croller ation PIC-4266 TCV-4234 Te=perature Controlled none, because TCV-4234 Valve failure is in TCV-4235 safe direction TCV-4274 None Class A31 A-C Corbination design codif-HV-2253 Motor Starter ication (providoHV-2254 for remote by-lHV-22131 pass of overloadHV-22132 HV-4225 circuit if HV-4257 needed)
N EV-2312 Motor Driven valve design =odif-HV-2290 Operator ication HV-2291 (short out
!EV-2253 overload
' HV-2254 element in EV-22131 valves - over-HV-22132 load protection provided else-where)
LV-21114 Control Valve Actuator design modif-
?V-2257 i
TV-2227 ication (re-
?V-2268 move pressure gage) i I
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700
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oo 600 TEMPERAluRE OF ATHOSPilERE me
SURfAEE TEMPERArullE 500 HELIUM LEAK
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Fig. 6.1-1--Temperatures in t he ort St. Vrain c
reactor building following a flesign Basis Accident or a cold reheat steam pipe ruptarc inside PCRV support ring at 30 ft from leak s-803 Amenilment 25
600 a
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Fig. 6.1-2--Temperature of reactor building atmosphere during postulated cold reheat steam pipe rupture accident as a fuuction of distance M
"R from possible source of steam leak S-804 Amendment 25
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P-81168
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P-81158 Page 1 of 4 14.10.
CESICN 3 ASIS ACCIOE'IT ':0. 1
"?ERMA'IENT LOS3 0F FORCED CIRC"LATICN (LOFC)" (?:ver Level 379 '*e(:))
14.10.1' Intredue: ion A hypo:hetical per=anen: loss of f:::ed circulati:n of primary c: clan:
helic = vould require :he extended f ailure of all feu: heliu= cir:ulators, their s:aas and water drives or their nul:1ple sources of =ctive ;cver. ::
f ailure of be:h the main steam and reheat s:eam see: ions of be:h s:ea:
generators. This c:ndi:1on is an extensl:n f :he 30 min :e pera:7 *:ss of nornal shu:d:vn :coling ac:iden: described in See: ion 14.'.
The decalled dese:17:1:n, consequeness and supplementar r inforesti:n
- ertaining
- o :his a::ident and supplemen:arv ac:iden:s are given in Appendix 0, de: ailed and supplemen:a y informa:i:n per:sining :o :na pe rmanen: 1:ss of f:::ed cir:ula:1:n (LOFC) f:: :he ?or: 3:. Vrain ETC?..
The ::n:en:s of A;;endin 0 include:
0.1.
Ce: ailed Oescripci:n of Design 3 asis Acciden: No. 1
0.2. Supplemen
1 LOFC Acciden: Resul:s 0.3.
LQ?C Acciden: Enperi.ran:a1 Oata and Analv:: 11 Nethods 14.10.2.
Cendt:1ons o f the Acciden A: cae time of chia hypo:netical loss of for:ed circula:1:n, the rea::::
vould have s::a:med, nasc pr:baoly on ":vo-L:op-t :uble" as defined in Tamle
- 7. L-3.
Loss of forced ci:cula: ion in :ne loop causes isolati:n of :ha: 1:cp while subsequent less of circula:1on in :he second loep cons:1:uces :v:-1:cp trouble.
'a* hen 1: becomes apparen: :s the plant operator that the loss of forced circulation is permanen:,
e.g., af:e about 5 h: when resumpti:n of c:oling would cause steam ge:eracor damage (see Appendix 3, See:1on 0.2.5). the pri=ary ecolan: system would be depressuri:ed to scorage (in a few hou:s) in the normal sanner chrough the helium purificatica sys:em as described in Section 9.4.3.5.
The reserve shu:down system would be operated after :nis initial period to assure an adequace shu:devn =argin.
The PCRV cooling va:e system '.ould continue in operation and would be closely nonitored since its operation is vi:a1 :o :he PCRV in:egri:y during the accident. This system is dese:1 bed in Section 9.7 as a Class I sys en connected to tne essential elec::ic. ' bus.
Two separa:e 1:antical closed
-loops supply cooling va:e: :: hree separa:e enes of :he PCRV: :he ::p head penetrations; :he core supoort floor; PCRV liner on :he side vali and
- vp head; and :ne PC27 bot::s head and bo:::m pene::sti:ns.
Ha'.f capa i:y cooling (one of ree identical loops opera:ing) is assu ed as :ha :ense:"2-tive lisi:ing case.
The reactor plant ren:11a: ice sretem would con:inus :s cperace normally during the accident in order :s provide fil::acien tnd elevated release f::
any fission product activity escaping ft:= dr.e ?C2V during :he c urse o f the accident. This system is dese:1 bed in See: ion 6.1.3.2.
P00R ORLGINAL
P-81168 Attacnment 2 A=end=en: la Page 2 of 4 14.10-2 No other r_ se:or plan: equi;=en is required to func:ica during this accident. Continued opera:1on of non-essen:ial equipment, instru=ents and controls nor= ally operating during reactor shutdown is assu=ad for purposes of monitoring plan: conditicas, but :his equipmen: has no effect on che course or consequences of :he ac:1 den:.
The following opera:or ac:1ons have been deter =ined to be either ne:essary or desirable :o =1:igate the consequences of this acciden:t 1.
Normal pos: sera = operatiens.
2.
Actions required to re-establish helium circula: ion (assumed to be unsuccessful for chis hypothetical acciden:).
3.
? i=ary coolan: system depressurization.
4 Operation of the reserve shutdown sys:em.
5.
Adj us t=en: of :he PCRV cooling systa= water flow ra:es and cover pressure :o increase cooling abili:y in areas affected.
None of the above 1:ees require rapid operator response and :hus :hese ac:1ons could be carried ou: in a logical and thoughtful =anner.
14.10.3.
Accident pesults and Consecuences Su==4rf.
This acciden: involves boch core damage and fission credue:
release causing off-site doses. The core hot regions sl:wly haa: up to abou: 5400*F, =axi=u=, occurring af:e 83 hr.
Approxi=a:elv 95P. of the fuel particles in the core will suffer f ailed coatings, resul:ing in ralease cf abou: 23* of the core fission produe: inven:ory. Of :his 23% of 'he inven-tory, less :han 5% re=ains gas borne in the PCRV. This 5 is essen:1 ally all noble gas wi:h a small amoun: of iodine.
Other chan =el:ing of the steel components of :he control rod asse=bly and some local f ailure of the PCLV liner insulation, ne c:her f ailures in the core or PCRV internals will occur. The core will re=ain suberi:ical during all periods of the acciden: due to control rod and reserve shutdown syste= poisons.
The doses resulting from this acciden: are orders of =agnitude lower than the guidelines of 10 CTR 100. The total duration (6 eench) doses at
- he low popula: ion =ene boundary (16,000 =eters) based on initial o;aration of the plant at 379 Mv(:), are listed below.
Whole body ga==a
- 0. 37 =re is Thy roid 36.
=re:
l 3cne
- 1. 0
=re:
f 1
P00R OR M
(=
P-81168 Page 3'of 4 14.11.
OESIG't 3 ASIS ACCIDENT NO. 2 "?aPID OEPRESSLT ".ATICM/3tCWCTi"
(?:ver Level 379 Mv(:))
In :his see:1on the Oasign Basis Acciden: No. 2 (Rapid Depressu:1:2:1:n) to taaly:ed and evaluated. This acciden: consists of a hypothetical sudden f ailure of be:h cl:sures in a ?CIV penetratica so :hac :he pri ary :colan:
sys:ec is spidly depressurizad, and any potential for air ingress is developed.
This acciden: vas orig 1: ally ;;esented in :he PSA:' as a " Max 1='.:2 Hype:hetical Accident" to illustrace : hat the essentt,* 1y inscantaneous relasse of the : actor :1::ula:1mg fission produ:: inventory would ne:
cx: cod 10 071100 limits. The doses are at lease an order of sagni:ude bol:v :he 10 Cy1 100 11:1:s. := assumed : hat the coolan: vi:h 1:a ac:1vi:7 is 211 ved :: escape directly f::s :he building in:o the at=csphers 2:
g und level <i:heu: any credi: for holdup or fil::a:ica by :he ventila:1cn systen. The of f-si:e deses resul:ing from :his acciden:, based on " design" ac:ivi:ies 2:e given in Table 11.11-1.
These doses veuld be even 1:ver if based on "expe :ed" pri=ary coolan: ac:ivities.
Me:e:::1:gi:a1 ::ndi:icas assu=ed for :he MRA are assumed :s be 1 niset vind speed and s:abili:y c ndi: ton C vi:h -he release ini:ia:ing as an area source.
Table 14.11-1 CCSIS TRCM MAX 2ft'M HYPCTHET! CAL ACCIOCIT
(? cue Level 379 Mv(c))
(Release of " Design" prinarv -colant gas-borne ac:ivity only)
Total Duracien Dose (ren)
At Exclusion At Lev Population Tvee of Dose Area 3ounderv Zone Boundarv Whole body ga==a (WBG)
- 2. 5 0. 0 73 Thyroid
- 5. 0 0.30 3cas 0.0 75 0.006 A nachanical basis was not censidered to exist for the Max 1=u= Hype-thetical Accident, and therefore only the hypothetical radiclogicni censcquences were censidered. However, ac :he :sques: ef the DRL the hypocnetical sudden failure of both closures in any pene::acien was sunty:ed. The resultant pri=ary system depressuri:ati:n ra:es vers annly:sd nachanistically vi:h respect to the free flow area presu=ed to d:volop in each pene:racien as lisi:ed by :he "71cv Rescric:or" engineered
.stfaguards. The insul:35g analyses showed :ha: even vi:h failure of bc:h cl:sures in a penetra:ica:
t The in:egrity of the fuel and core configura:1cn are not disturbed i _
by any th sc forces developed by the prima:y ecolan:, and imposed upon reac:or internals, during the depressu:1:ation.
l P00R~0 E MI.
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l P-81168 Page 4 of 4 14.11-2 2.
Adequa:e pri=ary circuit cooling would be =aintained following the depressuri:a: ion by use of the circulators on ei:her stea=
or water-turoine drive with :he reduced coolan: densi:y. No da= age to':he circula: ors would occur during the depressurization.
3.
- he continuation of :he nor=al return-flow of clean heliu= :o :he PCRV would al=os: totally exclude ingress of at: in:o :he PCRV.
However, even if all return flow to the PCRV were eli=ina:ed, the air ing:ess would be insignificant fro both a hea: generation and graphite conbustion viewpoint.
4 A1: hough :here are a number of se: ions which :he operator could perfor= to fur:her ensure or i= prove the saf e shu:down of :he plan: :here are no i==ediate or necessary ac:1ons which are
- equired of hi::.
5.
~he pressurs differen:ials and je: forces due :o :he depressu:1:ing pr1=ary coolan: vill no da= age or overprtssuri:e any required co=ponents of the PCR7 or the reactor building 6.
- he radiological :ensequences of :he rapid depressurization, al: hough increased in severi:y vi:h respe:: to the Maxi =u:
Hypothe:ical Accident, would be well within the li=1ts prese:1 bed by 10 CTR 100.
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