ML20081F518
| ML20081F518 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/27/1983 |
| From: | Schroeder C COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 7521N, NUDOCS 8311030114 | |
| Download: ML20081F518 (8) | |
Text
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _. _ _ _ _ _ _ _ _ _ _ _ _
CN Commonwealth Edison
[
) one First National Plaza, Chic *go, tilinois 7
(
l Address Riply to: Post Offica Box 767 3
'y/ Chicago, litinois 60690 October 27, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission l
Washington, DC 20555
Subject:
LaSalle County Station Units 1 & 2 i
i Response to Special Scenerio'on Turbine l
Control System Failure of Non-Safety )
Equipment (Multiple Failures Concerns l
NRC Docket Nos. 50-373 and 50-374 i
Dear Mr. Denton:
Following the June 21, 1983 discussion of multiple failures of non-safety related equipment as a result of HELB or failures of common sensors / power supplies (Reference FSAR responses to QO31.288 through 296 es revised), Dr. Bournia forwarded a special scenerio of HELS at feedwater heaters where the turbine building zone H7 environmental conditions become equivalent to reactor building zone H4.
The hypothesis then assumes failure of all non-safety related turbine control equipment (recirculation system control equipment and feedwater system control equipment are not co-located in this zone).
The concern is that with the turbine bypass valves unavailable coincident with loss of feedwater preheating (hence increased reactivity hence a power increase) that an ovent beyond the previously analysed safety envelope could result.
The NSSS contractor has evaluated this scenerlo specifically for LaSalle and has provided the analytical results for two cases:
Auto flux control on the feedwater system and manual flow control on the feedwater system.
The ODYN results indicate, respectively, that the peak vessel non-bounding.ak heat fluxes and peak neutron fluxes are acceptable andThe auto-flux control case is b pressures pe clad temperature and MCPR.
The manual flow coqtrol case produces a peak clad temperature of less than 10000F with approximately 1% of the fuel rods subject to transition boiling.
The MCPR is approximately 1.00+0.02.
This means that there is no challenge to fuel integrity in either case.
The auto-flux control maintains the neutron flux nearly constant by reducing core flow to compensate for subcooling effect.
That case is a transient.
The manual flow control case, on the other hand, because of the decreased MCPR is categorized as an accident even though there is significant margin to the fuel PCT limit of 22090F.
Therefore, it is concluded that there is no safety concern to LaSalle from this HELB event in the Turbine Building.
8311030114 831027 hk PDR ADOCK 05000373 l
IJ I PDR k
u, H. R.'Oenton October 27, 1983 a
L The' sequences of events for these two cases, along with plots of significant" plant parameters, are also provided in the attachment.
The
/[>
-worst timing for the turbineGtrip without bypass was arbitrarily imposed
- 'in,these sequences.
The 1000F loss of feedwater heating was shown to be a worse situation than a<l500F loss of feedwater heating (see FSAR response to Questions 212.56 and.142).
Additional information on these same. events is documented in FSAR ' responses to Q212.15,.58,.61,.105 and
.144.cwhere special~ failure hypotheses were applied but the bounding case of Chapter 15 was never compromised (see Q212.61).
A similar series of hypotheses were explained in FSAR respnnses to QO31.124,.145,.215, and g
fr
.256 regarding multiple failures of non-safety equipment.
T' The reactor protection system devices for tho safety related
, functions from the turbine building. include the following:
4 a.
C71-N005A-D Turbine Control VGive movement by EHC pressure.
b.
C71-N006A-D Turbine Stop Valve closure position switch.
C71-N006E-H Turbine Stop Valve closure position switch, c.
C71-N003A D Trip bypass' for,a and b (above) when at low 7
reactor power below turbine pressure 106 psig.
ThesedevkcesareseismicallymountedandareincludedintheLaSalleEQ Program for environmental qualification.
Se'e response to Q.031.97 for oroll statement on seismic consideratons.
Although these devices are gocategr{nzoneH-7, they are qualified, or are being qualified to the H4A environment or,ifor the pressure switch, being upgraded with a qualified replacement.
The general conclusion from this specific scenerio, along with those events referehced above, is that the. Chapter 15 accidents representing the LaSalle design' basis envelope'these hypothesized transients.(and accidents).evenlltiple failure assumptions.-
- f. hough some are unquestionably beyond the design basis w~ith res'pe'ct to mu i
As stated in the June 21, 1983' discussions with the' staff, the feedwater control t
system and the~ turbine cont.tol, system are not colocated, nor is the l
turbine control system colocated with/the recirculation control system, hence Tallure of one of these non-safety related control systems from T.
localized events _does not directly threaten the other non-safety related control system.
Specifically for the post 01ated scenerio, a HELB in the turbine building does not directly affect tt)e feedwater control system nor the recirculation' control system.
)
To the best of my knowledge and belief the statements contained hereinland in the attachment are_ true and ' correct.
In some respects these statements are not based on my personal knowledge but upon information furnished by other' Commonwealth Edison and contractor employees.
Such information has been reviewea in accordance with Company practice and I believe,it to_be reliable.
E s
- 4.
- - 9.
ar H ; -- R '. t De n t on -
- October 27, 1983
.. Enclosed for.your use are one (1); signed original and forty (40) copies;.of this. letter and..the enclosures.
1Weitrust that this.information provides the basis for your l
Lf c osure o.all concerns regarding multiple failures at LaSalle County
-Station.
It is therefore: requested that you-provided, in writing, confirmation of closure of.the : Unit-1 and potential: Unit 2 -license condition;on this topic..'If'you have.any further questions, please-
' contact this-office.-~
~'
very truly yours, ml2-7/o3 C.-W. Schroeder Nuclear Licensing Administrator
- 1m
- cc:: NRC Resident Inspector'- LSCS.
4
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TABLE 1 LASALLE RELE ANALYSIS RESELT3 Case 1 Case 2 (Maman1 Flaw centrell (Anka-Flum Caakral)
Peak Neutron Flux 346.0 330 7
($ of Initial)
Peak Heat Flux 128.5 113.0 (5 of Initial)
Peak Dome Pressure 1165 1162 (psig)
Peak Yessel Pressure 1192 1187 (psig)
Initial CPR 1.24 1.24 4LCPR 0.25 0.16 Minimum CPR Option A Adjustment 0 98 1.08 Factor Option B Adjustment 1.02 1.11 Factor
% of Fuel Rods Subject to Transition Boiling l
Option A Adjustment 1.15
( 0.1%
Factor Option B Adjustment 0.4%
( 0.1%
i Factor l
Peak Cladding (1000 l
Temperatures j
(OF) i l
l
j TABLE 1 Sequence of Events for LaSalle Loss of 100*F FW Heating at Manual Flow Control Mode With Turbine Trip a~t the Worst Timing and With Bypass Failure j
Tine (sec)
Event 1
0 100*F temperature reduction in the feedwater system caused by the extract steamline break in the feedwatee heaters.
>5 Initial effect of unheated feedwater starts to raise core power level and steam flow.
~37 APRM initiates reactor scram on high thermal power.
i 37 Turbine trip is conservatively assumed to occur at the same time as that of thermal power monitor scram due to the extract steamline break ef fect.
37 Turbine bypass valves fall to operate due to extract steamline break ef fect.
37.01
- Turbine stop valves reach 90% open position and initiate a recirculation pump trip (RPT).
37.1 Turbine stop valves closed.
37.19 Recirculation pump motor circuit breakers open causing decrease in core flow to natural circulation.
38.3 Group 1 relief valves actuated.
38.4 Group 2 relief valves actuated.
38.5 Group 3 relief valves actuated.
38.6 Group 4 relief valves actuated.
38.8 Group 5 relief valves actuated.
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1 LEVELl!NCH-REF-SEP-SKIRT I VO!O MFK TIVIT EACTI]ITT 2 H R SENSED LEVELiINCM SI 2 DOPPLER f 3 SERAM REFCTIVI i
3 N R SENSF0 LEVELIIE E S) g*
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TIME 15EC1 HR1 DV06 HMR CEC 102 ORF 672C-9 LOSS OF 100 F FW HEATING W/ TT W/0 BP. IFC
' LOSS OF 100*F FW llEATING AT MANUAL FLOW CONTROL MODE WITil TURBINE TRIP AT Tile WORST
$ise.Yo.'.
FTGURE 1
r TABLE 2 Sequence of Events for LaSalle Loss of 100*F FT Heating at Auto-Flux Control Mode With Turbine Trip at'the Worst Timang and With Bypass Failure
_ Time (sec)
Event 0
100*F temperature reduction in the feedwater system caused by the extract steamline break in the feedwater heaters.
~5 Initial ef fect of unheated feedwater starts to raise core power level but auto-flux control system automatically reduces core flow to maintain initial neutron flux.
~20 Reactor variables (mainly neutron flux and core flow) settle into new steady state.
20 Turbine trip is assumed to occur due to the extract steamline break ef fect.
20 Turbiae bypass valves f ail to operate due to extract steamline break ef tect.
20.01 Turbine stop valves reach 90% open position and initiate a reactor scram.
l 20.01 Turbine stor valves reach 90% open position and initiate a recirculation pump trip (RFI).
20.1 Turbine stop valves closed.
~
20.19 Recirculation pump motor circuit breakers open causing decrease is core flow to natural circulation.
21.2 Group 1 relief valves actuated.
21.3 Group 2 relief valves actuated.
21.5 Group 3 relief valves actuated.
21.6 Group 4 relief valves actuated.
21.7 Group 5 relief valves actuated.
l l
1 VESSEL PFES SE'(PSil I NEUTRON FL JC NTER TE W 2 STM LINE Pf1E SE IPSI) 2 PERM FUEL j : NEAT FLUX LDTPASS yFQE
_(,I 3 Ry_E_SURffC 125*
4 C0ftE INLt I t' U/LB) 150*
4 FEEDWATEf
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TIME l'ECl TIME ISECl I LEVEL (INCH-REF-SEP-SMIRT I VOID RER(ji!VITT 2 W R SENSt0 LEVELl!NCHESI 2 DOPPLER NFFCTIVlif 3 SCAAM Affc".VITT 3 N R SENSFO LEVEll!NCHESI I*
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TIE (SEC) 1511 DV06 iMB CEC 202 DF 672C-9 LOSS OF 100 F FW ERTING W/TT W/0 BP. FC IIsse!E.
FIGUkE 2 LOSS OF 100*F FW !! EATING AT AUTO-FLUX CONTROL MODE WITl! TURBINE TRIP AT Ti!E WORST TTkfiNG AND WITII BYPASS FAILURE