ML19246A358
| ML19246A358 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/02/1979 |
| From: | Quittschreiber Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-SM-0105, ACRS-SM-105, NUDOCS 7906180617 | |
| Download: ML19246A358 (15) | |
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UNITED STATES E h w d' NUCLEAR REGULATORY COMMISSION h
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WASHINGTON, D. C. 20555 May 2, 1979 P
ACRS Merbers ACP.S Technical Staff RESPONSE OF CGGJSTION ENGINEERING PIANTd 10 THREE MILE ISIR;D TYPE TRANSIENTS NRC Staff and Combustion Engineering met an ?'.av 1,1979 in Bethesda,10 to discuss generically the effect that the transients which occurred at Three Mile Island would have on Ccmbustion Engineering plants.
IESS OF FEEDGTER Historically, the Combustion Engineering operating plants response to loss of fee 6ecter has been that within 15 to 30 seconds they get reactor trip ca low steca generator water level. The turbine will trip on reactor trip and steam bypass will occur. The power operated relief valve (PCRV) will not open under this condition.
At least 10 to 15 minutes are available t, t ually initiate auxiliary feedeater before steam generator heat remval cgability is degraded.
SIUCK OPEN PRESSURIZER RELIEF VALVE Rapid depressurization would occur with a reactor trip an low pressure in ab ut 5 seconds. ECCS would be initiated by low pressure in about 15 to 30 seconds. Flashing in the RCS would occur with an insurge into the pressuri;er. Ccmbustion Engincering anticipates that the operator would close the block valve to stop the transient.
CE PIW4T GEERIC DESIGN FEATURES MiICH MITIGATE TEREE MILE ISL\\ND TYPE TPMSIENTS 1.
The anticipated response to rederate frequency events, such as loss of feedeater, does not open the PChV.
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A large secondary water inventory allows sufficient ti:nc to manually 2.
initiate auxiliary fecaeater (pbout 13 minutes for those plants that don't have autenatic initiation and a little less for those with automatic initiatica).
3.
The pressuri:cr level is not used to automatically initiate safety systems. Iow pressurizer pressure nonnally initiates ECCS.
4.
The low HPSI shutoff head will not lift the pressurizer relief or safety valves.
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He NSSS elevation layout requires only 20-25% of RCS inventory to cover the core. The top of the active core is about 14 feet lower than the bottom of the tube sheet in System 80 plants.
S RLL PIPE ERE ES Combustior Engineeging has analyzed 14 diffemnt pipe break sizes and has y
found tF.c 0.05 ft cold leg break is the worst (much worse than break at top of the pmssurizer) 0.1 ft2 breau and larger do not require steam generator heat remval since sufficient heat will be mmved through the break.
The advantages of "U" tube steam generators, where the hottest e.er in the
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primary interfaces with the secondary side of the steam generator at the lowest level first, over a straight-through type design were noted as follows:
a.
On depmssuri:ation, a steam bubble will form in the reactor vessel and not in the top of the steam generator, since the ;op of the steam generator is cooler.
4 b.
hbat transfer to the secondary occurs with any secondary side water level.
c.
A moderate voline of non-condensibles does not cause a catastrophic loss of heat sink.
NNIURAL CIRGIATION Ioss of the all AC power will allow something greater than 40 minutes before the steam generators boil dry, assuming no auxiliary feedwater.
Natural circulatic. following loss of AC power will be about 2% of full flow. Flow can be verified by AT masuremnts across the m ictor and steam generators. No direct flow measurements are available.
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Combustion Engineering indicated that Michelson's concem, that during
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a small break, water slugs and non-condensibles when going from pool boiling to natural circulation might prevent establishing natural circulation, would not be a pmblem due to the high differential pres-sure driving force from the hot stean in the reactor to the mlatively cold steam generator abes t* 1ch would be rapidly condensing steam and would push gases from the tubes.
1 CONTRCL SYSTDiS Cperational history of CE plants control systems failures indicates that there have been cnly 6 out of 180 total (3%) scrams in 28 years of cperation attributabic to NSSS control system sensor and logic induced protective system challenges. nese 6 scrams kem due to main feedwater control challenges.
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t 3-Canbustica Engneering has been invited to discuss some of these same itens at the 229th ACRS Weting on May 10, 1979..,
g h.~uQ G. R. Quittschreiber Senior Staff Engineer Attachments:
Handout Material 9
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c C-E OPERATING PLAfiT RESP 0ilSE TO A STUCK PRESSURIZER RELIEF VALVE RAPID DEPRESSURIZATIO!1 0F RCS e
TM/LP TRIP ON LOW THERMAL MARGIN ECCS INITIATI0ft AT APPROXIMATELY 1600 PSIA FLASHIliG IN RCS RESULTIflG IN PRESSURIZER INSURGE FILLING OF PRESSURIZER RESULTIfiG Ift TU0-PHASE RELIEF THROUGH PORV FAILURE OF DRAlft TANK RUPTURE DISK AND FLUID RELEASE TO THE CONTAlfiMENT OPERATOR CLOSES PRESSURIZER BLOCK VALVE TERMINATIt!G
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DEPRESSURIZATI0ft 223 198
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SIZE, FT PSIA
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M SIMULTANEOUS HOT LEGICOLD 2
20 LEG INJECTION COOLS CORE 1
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20 0.5 FROM VESSEL O.2 75 0.1 125 O.05 150 0.02 225 c
REFILL OF RCS DISPERSES 0.01 410 BORIC ACID 9ROUGHOUT SYSTEM AND SG'S ARE 4s 0.005 8@
ABLE TO COOL RCS TO 0.002 1425 SDC TEMPERATURE 0.001 1620 0.0005 1680
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c 228 l99 OVERLAP OF ACCEPTABLE LTC MODES IN TERMS OF COLD LEG BREAK SIZE
SOURCES OF lloil C0r'DEllSIBLES 3
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STP SAT a 1300 PSI DISSOLVED H2
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S.B. ASSUMPTIONS REACTOR TRIP DN L.P.P.
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SIAS ON L.P.P.
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PUP.P TRIP ON REACTOR TRIP LOSS OF FAIN F.W. ON REACTOR TRIP FAIN STEAM VALVES CLOSED ON REACTOR TRIP
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N0 OPERATOR ACTION FOR FIRST PHASE SEC. SIDE AT SAT. C0i1DIT'ONS OF LOWEST SAFETY VALVE SETPolliT 228 201 N
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C-E CONTROL SYSTEMS OVERVIEW O AUTOMATIC TURBIE!E/ REACTOR CONTROL BETWEEN 15% AND 100% POWER EXCEPT FOR MANUAL COM-s TROL OF BORON C0i!CENTRATIO!! A!1D AXIAL POWER DISTIRBUTION 0 CONTROL SYSTEMS LARGELY SEPARATE FROM AND INDEPENDENT OF PROTECTIVE SYSTEMS O
o REDUNDANT 2/2 LOGIC EMPLOYED TO ENHANCE SAFETY A!1D AVAILABILITY BY PREVENTING SI!1GLE CONTROL SYSTEM FAILURES FROM CAUSING STRONG, RAPID C0!iTROL SYSTEM ACTIO:iS STEAM BYPASS CONTROL SYSTEM o CONTROL SYSTEMS DESIGt!ED TO ENHANCE SAFETY Ai!D AVAILABILITY BY RESTORIi!G ACCEPTABLE COMDITIONS
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WITHOUT CHALLENGING THE PROTECTIVE SYSTEM DURING A i!UilBER OF DESIGN BASIS EVENTS o EXTEB!SIVE DIAGNOSTIC / TEST FEATURES PROVIDED TO ENHANCE RELIABILITY 228 203 9
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LOSS OF 0FFSITE POWER 9
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MAIN CONDENSER IS LOST
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hAlta FEE 5 WATER IS LOST AUXILIARY FEEDWATER IS diT_ INITIATED PRESSURIZER ASSuneTIONS m
NO CHARGING / LETDOWN
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NO HEATERS / SPRAY NO WALL HEAT IRANSFER STEAM UENERATOR PRESSURE IS MAINTAINED BY 1.
SECONDARY CODE SAFETY VALVES, OR 2.
SECONDARY DUMP VALVES a
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