ML17179B008
| ML17179B008 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/22/1993 |
| From: | Stang J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9307290187 | |
| Download: ML17179B008 (34) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION Docket Nos. 50-237 and 50-249 WASHINGTON, D.C. 20555--0001 July 22, 1993 LICENSEE:
Commonwealth Edison Company
SUBJECT:
SUMMARY
OF MEETING WITH COMMONWEALTH EDISON COMPANY TO DISCUSS THE CONTAINMENT COOLING SERVICE WATER SYSTEM AT DRESDEN, UNITS 2 AND 3 On April 7, 1993, Commonwealth Edison Company (CECo, the licensee) met with NRR and Region III personnel to discuss the design of the containment cooling service water (CCSW) system at Dresden, Units 2 and 3.
In addition, the licensee made a presentation on information contained in its March 5, 1993, submittal to the NRC.
The March 5, 1993, letter provided information concerning the reconstituted design basis of the CCSW system.
The package also contained supporting documentation concerning the normal and post-acc i dent o.perat ion of the system.
The purpose of the meeting was for the NRC Lu uGLdir1 a better understanding of the CCSW system operating and design basis. is the handout provided by the licensee. is a list of attendees.
The licensee described the configuration of the system for normal and post-accident operation.
The licensee described the analyses performed in the March 5, 1993, submittal for the CCSW system heat exchanger duty, long-term containment cooling following a loss of coolant accident (LOCA), and net positive suction head (NPSH) calculations for the Emergency Core Cooling System (ECCS) pumps.
The NRC staff expressed its concerns with input parameters used in the March 5, 1993, submittal with respect to the NPSH evaluations for the ECCS pumps.
The staff's concerns dealt with the licensee's basis for choosing certain assumptions and input parameters, including the licensee's evaluation techniques.
The NRC staff requested that the licensee provide additional information regarding the CCSW reconstituted design basis. Specifically, the staff requested that the licensee document the pertinent parameters for evaluating the low pressure core injection (LPCI) pump NPSH margin derived from a consideration of the as-built plant design.
This was to be accompanied by a listing of the input parameters and assumptions actually used in the licensee's analysis of the available NPSH margin.
Finally, the effect on the available LPCI NPSH margin of varying each of these input parameters was to be submitted.
This request arose out of the staff's concern that it was unable to determine with the information available in the March 5, 1993, submittal, the overall conservatism in the licensee's reconstituted design basis of the CCSW system.
Specifically, in light of the licensee taking credit for containment over-pressure in evaluating the LPCI NPSH margin, it is necessary for the staff to evaluate the overall degree of conservatism and impact of each of the licensee's assumptions in the March 5, 1993, submittal.
The staff indicated 9307290187 PDR ADOCK p
- ,, 6*' 0 n {..-,
,(,
l I!~
'.)
~
~* __:_
\\.,J 930722 05000237 PDR
Commonwealth Edison Company in the meeting that adequate information was not available in the March 5, 1993, submittal to allow the staff to reach the conclusion that 1 CCSW would provide adequate long-term containment cooling.
The licensee indicated they would provide the additional information by May 21, 1993.
The information has not yet been docketed.
During the meeting, questions were posed to the licensee concerning loading of two CCSW pumps onto 1 diesel generator.
The licensee verbally responded that one diesel generator had marginal capacity to operate two CCSW pumps during a design basis accident.
However, the licensee indicated that operators were concerned about manually loading two CCSW pumps onto 1 diesel generator because the resultant loading would be close to the diesel generator load rating.
The staff indicated that 2 CCSW pump operation, even with degraded flow, would provide adequate heat removal capability.
Enclosures:
As stated cc w/enclosures:
See next page
~\\\\ c-?~
~J.$
Bohn F. StaQ, Project Manager Project Directorate 111-2 Division of Reactor Projects 111/IV/V Office of Nuclear Reactor Regulation
OFC NAME DATE Commonwealth Edison Company in the meeting that adequate information was not available in the March 5, 1993, submittal to allow the staff to reach the conclusion that 1 CCSW would provide adequate long-term containment cooling.
The licensee indicated they would provide the additional information by May 21, 1993.
The information has not yet been docketed.
During the meeting, questions were posed to the licensee concerning loading of two CCSW pumps onto 1 diesel generator.
The licensee verbally responded that one diesel generator had marginal capacity to operate two CCSW pumps during a design basis accident.
However, the licensee indicated that operators were concerned about manually loading two CCSW pumps onto 1 diesel generator because the resultant loading would be close to the diesel generator load rating.
The staff indicated that 2 CCSW pump operation, even with degraded flow, would provide adequate heat removal capability.
Enclosures:
As stated cc w/enclosures:
See next page DISTRIBUTION Docket File PDllI-2 r/f JPartlow JZwolinski JStang OGC ACRS(lO)
BClayton, Rill RJones SBurgess, Rill TCollins MPeck
/tJ-/93 Original Signed By~
John F. Stang, Project Manager Project Directorate III-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation NRC & Local PDRs TMurley/FMiraglia JRoe JDyer CMoore EJordan GGrant, EDO JKudrick MD Lynch CPatel MRazzaque
/-,-~See previous concurrence 2
IBC:SRXB*
IBC:SCSB*
ID:PDllI-2 IRJONES IRBARRETT IJDYER j7t-106/04/93 106/04/93 17~93
OFC NAME Commonwealth Edison Company The licensee indicated they would provide the additional information by May 21, 1993.
John F. Stang, Project Manager Project Directorate III-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/enclosures:
See next page DISTRIBUTION Docket File PDIII-2 r/f JPartlow JZwolinski JStang OGC ACRS(lO)
BCl ayton, RII I RJones SBurgess, RIII TCo 11 ins MPeck NRC & Local PDRs TMurley/FMiraglia JRoe JDyer CMoore EJordan GGrant, EDO JKudrick MD Lynch CPatel MRazzaque ID: POI II-2 IJDYER DATE I lx{
/93 16/r/93 I /93
Commonwealth Edison Company cc:
Mr. D. L. Farrar Manager, Nuclear Regulatory Services Commonwealth Edison Company Executive Towers West III, Suite 500 1400.0PUS Place Downers Grove, Illinois 60515 Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60690 Mr. C. Schroeder Plant Manager Dresden Nuclear Power Station 6500 North Dresden Road Morris, Illinois 60450-9765 U. S. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station 6500 North Dresden Road Morris, Illinois 60450-9766 Chairman Board of Supervisors of Grundy County Grundy County Courthouse Morris, Illinois 60450 Regional Adm1nistrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Illinois 60137
.Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Robert Neumann Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300
. Chicago, Illinois 60601 Dresden Nuclear Power Station Unit Nos. 2 and 3
I. Brief system description :
AGENDA APRIL 7, 1993 A. Physical system layout and operation of CCSW ENCLOSURE 1 B. Description of the containment cooling subsystem configuration nonnal and post LOCA.
C. Discussion of Diesel Generator loading capability
- 2. Description of Analyses perfom1ed:
A. Heat Exchanger Duty Calcualtion
- 1) Key Inputs
- 2) Results B. Long term containment cooling I). inputs used, ie key parameters such as pump flows, torus water temperatures and levels, decay heat inputs and HX duty.
- 2) Results of analysis.
a) peak long term pool temperature b) peak long term pool pressure C. N"PSH Calcualtion
- 1) Key Inputs
- 2) Results
- 3. Current actions to improve system
- 4. Conclusion Commonwealth Edison Pa11icipants:
Paul Dietz Mechanical & Strnctural Group Sharon Eldridge Pete Piet Rick Ralph Kevin Ramsden Joann Shields Brian Viehi Site Engineering Nuclear Licensing Technical Staff Nuclear Fuel Services Regulatory Assurance Site Engineering
_J
LPCI SYSTEM FIGURE I DRY\\JELL SPRAY LPCI INJECTION
--T-E-ST_/_T_D_R_u_s-ill_::oflof-o-L-IN-!11Gfl-M-------L_--- ------- J----
TORUS SPRAY
/
fO OTHER LPCI SUB:O:YSTEM I'
M LPC! HX
.----------------------'-----=--...[~
I I
LPCI Pl IMPS
~
\\
/ nl FLOV ;RIFICE Li-------ij cornROL
~
r rDO", HVAC L,,:;":;'.;~;" ~
7
'~
~
L e
cc-~\\./ P_UMP.,__:f p:c; \\J LDDLER
~UMF'S
~
SUCl !UN FROM CRIB HOUSE
LPCI SYSTEM FIGURE 1 L.PCI PUMPS TD OTHER LPCJ SUBSYSTEM SUCTION
LPCI SYSTEM FIGURE I l!CACTIJI LPCI PUMPS
--- DISCHARGE TD OTHER LPCI SUBSYSTEM SUCTION
LPCI SYSTEM FIGURE 1 LPCI PUMPS TO OTHER LPCl SUBSYSTEM SUCTJON
l i
I I I I !..
\\
I SYSTEM DESCRIPTION e
Containment Cooling Service Water Containment Cooling Subsystem
'1 Operation
~
CONTAINMENT COOLING SERVICE WATER e
Flow Patl1 e
Component Description o
Power Supplies L.------
4 ~
o I
CONTAINMENT COOLING SUBSYSTEM Flow Paths
'9 Component Description 9
Power Supplies I
' i I
I
OPERATION
~ Loss of Coolant Accident e
Loss of Off Site Power
e e
NORMAL CONFIGURATION 2 LPCI 2 CCSW PUMPS
- THE NORMAL SUB-SYSTEM CONSISTS OF 2 LPCI PUMPS, 2 CCSW PUMPS AND 1 HEAT EXCHANGER.
~ THE SUB-SYSTEM IS 1 OF 2 REDUNDANT, ELECTRICALLY SEP ARA TED AND DIVISIONALIZED SUB-SYSTEMS
- ACHIEVE A HEAT EXCHANGER DUTY OF 82.48 MBTu/Hr WITH FULL COMPLEMENT OF 2 LPCI, 2 CCSW PUMPS INCLUDING FLOW MEASUREMENT INACCURACIES.
- ONLY AVAILABLE WITH OFF SITE POWER
- ~
I LOCA/LOOP CONFIGURATION
- AUTOMATIC INTIATION STARTS DIESEL GENERATORS AND LOADS:
- 1 CORE SPRAY PUMP AND 2 LPCI PUMPS PER DIESEL
~ THIS LINE UP PRODUCES THE MAXIMUM DESIGN LOADING ON THE DIESEL o ADDITION OF CCSW PUMP TO THIS LINE UP WOULD EXCEED THE DIESEL GENERA TORS CONTINUOUS RA TING.
e e
LOCA/LOOP CONFIGURATION (CONT'D)
~ CONTAINMENT COOLING MODE IS ENTERED BY MANUAL ACTIONo
- 1 LPCI PUMP IS SECURED
- SECOND LPCI PUMP IS LINED UP FOR DRYWELL AND l,ORUS SPRAYS
- 1 CCSW PUMP IS STARTED
- ACHIEVE A HEAT EXCHANGER DUTY OF 55.2 MBTu/Hr WITH 1 LPCI PUMP AND 1 CCSW PUMP INCLUDING FLOW MEASUREMENT INACCURP~CIES.
n!sEL GENERATtfa. LOADING
- ORIGINAL SPECIFICATION (K-2183, 3/31/67)
- 2500 KW CONTINUOUS
- 2750 KW FOR 2 HOURS
- NAMEPLATE (AS DELIVERED)
- 2600 KW CONTINUOUS
- 2860 KW FOR 2000 HOURS
- LOCAJLOOP IS MOST LIMITING DESIGN CONFIGURATION FOR DIESEL GENERATORS.
LOA!NG FROM LA TES~ALCULA TION OF RECORD FOR LOCA/LOOP Diesel Peak Load (KW)
Long Term Load (KW)
Generatcr Autoloaded Pumps Manually Set Line Up (Core Spray & 2 LPCI) (Core Spray, 1 LPCI &
1 CCSW)
Unit 2 2247 2119 EDG Swing EDG 2340 2157 (U2)
Swing EDG 2343 2144 (UJ)
Unit 3 2260 2060 EDG
- DIESEL GENERA TORS HA VE ADEQUATE CAPACITY FOR LPCI LONG TERM COOLING ( 1 LPCI PUMP AND 1 CCSW PUMP )
SUMMARY
OF ANALYSES
- Initial Concern:
992 pump CCSW flow Less Than FSAR LPCI HX Design Table"
.. Noted by operating personnel during 1naintenance activities.
$ Revie\\v of.6.!ialyses Performec!
- l\\1ark I Torus Temperature Transients Affected
- Tube replacement authorization withdrawn
- pending further revie'\\V
- J\\1ark I limiting case was immediately reanalyzed by CECo
- NTU method used to extrapolate HX performance from data sheet
i--'OCA heatup calculations to confir~n data in process diagram
- Results of Document Search
- GE unable to locate original calculations
- Results of HX Performance Calculation
- The reconstituted va!ue of HX performance was approximately 9°/o less
- Detailed review of GE cal cs performed by CE Co, conclusion was that conservative methods were correctly employed.
- Decision was made to have GE reperform the long term post-LOCA calculations based on the revised heat exchanger data.
NOTE: At this point, there was nothing to indicate that the original FSAR calculations were in error.
There also was no way to prove that they were alJsolutely correct.
- Reconstituted Long_Term post-LOCA Calculation
- Worst case in calculation based on 1 LPCl/l CCSW pump combination
- ANS 5.1-1979 Decay heat models
CASE NO 3
3a 4
4a
- Results:
- Peak Suppression pool temperature of 186F.
Prior maximurn was 180F.
- "Second Pressure Peak" increased by 2 psi from 8 to 10 psig.
- CCSW to LPCI delta-P across I-IX tubes confirmed to be positive, precluding leakage path.
SUMMARY
OF ANALYTICAi_; RESULTS
- LPCI FLOW
- CCSW FLO'V PEAK PUMPS (gpm)
PUMPS (gpm)
TEMP
(°F) 2 10,000 2
5,600 168 2
8,916 2
4,795 171 l
5,000 l
3,500 180 l
3,88 l 1
3,071 186 PEAK PRESSLRE (psig) 7.2 7.6 8.6 9.4
- Revised Tube Replacemen~ Criteria
- Tube replacement criteria recalculated to keep the effects of replacement within the design 6°/o plugging allowance
Table 1 Key Analysis Assumptions, CECo/GE Recirc Line Break Initial Conditions Reactor Power Reactor Dome Pressure Suppression Pool Initial Temperature Pool Atmosphere Initial Conditions Drywell Free Volume Suppression Pool Airspace Volume Suppression Pool Volume Time to initiate pool cooling Drywell initial conditions Downcomer submergence/flow area Feedwater mass/energy Decay Heat ECCS pump heat 102%
1020 psia 95 F 95 F, 100% RH, 0. 15 psig 158236 ft3 120097 ft3 112000 ft3 600 seconds 135 F, 20% RH, 1.25 psig 3.67 ft, 301.6 ft2 feedwater mass and energy added to containment ANS 1979 I 00% of motor HP added as heat input Revised Heat Exchanger Capability with maximum flow uncertainties (1 LPCUl CCSW)
(2 LPCU2 CCSW)
Heat Removal (at reference 55.238 1\\1BTU!HR 82.48 MBTU!HR temperatures below)
Primary Flow (LPCI) 3881 GPM 8916 GPM Secondary Flow (CCSW) 3071 GPM 4795 GPM CCSW Temperature 95 F 95 F LPCI Temperature 165 F 165 F
Table 2 Key Analysis Assumptions, CECo 0.01FT2 Steamline Break Transient Analysis Initial Conditions Reactor Power Reactor Pressure Pool Initial Temperature Suppression Pool Temperature Monitoring System Error Stearn flow feed flow CRD flow Offsite power available for all cases Normal Auto operation of ECCS systems MSIV Closure Time Feedwater temperature Decay Heat LFCI pumps heat 102%
1050 psia 101 F
( 1 F Steady State Error) 6F due to all causes (Transient Error) 102% of licensed conditions 1 l.11 lb/sec 3 seconds with.5 sec delay operating condition until feed system volume is swept, then 170 F used for conservatism Approximately 1.1 ANS 1979 for infinite operating history 100% of motor I-IP added to pool as heat input Revised Heat Exchanger Capability (1 LPCI/1 CCSW)
(2 LPCl/2 CCSW)
Heat Rernoval (at reference 63.63 ~*IBTU/HR 89 06 MBTU/HR temperatures below)
Primary Flow (LPCI) 4500 GPM 9000 GPM Secondary Flow (CCSW) 3450 GPM 5000 GPM CCSW Temperature 95 F 95 F LPCI Temperature 165 F 165 F
e NPSH ANALYSIS ASSUMPTIONS & INPUTS
- RESULTS OF CONTAINMENT RESPONSE ANALYSIS
- MAXIMUM SUPPRESSION POOL TEMPERATURE
- CORRESPONDING MINIMUM PRESSURE
- NPSH REQUIRED NOT ADJUSTED (REDUCED) FOR TEMPERATURE.
- SUCTION PIPING LOSSES
. DETERMINED AT 90° F.
e e
NPSH ANALYSIS RESULTS NPSH (FEET)
AVAILABLE REQUIRED MARGIN ONE LPCI PUMP NOMINAL FLOW 38.62 30.0 8.62 (5000 gpm)
ONE LPCI PUMP MINIIVIUl\\1 FLOW 39.48 25.7 13.78 (3881 gpm)
TWOLPCI PU1\\1PS 38.14 30.0 8.14 NOI\\1INAL FLOW (10000 gpm)
TWOLPCI PUMPS 39.35 26.9 12.45 l\\1INil\\1Ul\\'1 FLOW (8916 gpm)
CURRENT ACTIONS TO INCREASE MARGIN e
Flow Balance Heat Transfer et Pump Fouling
'(
e
SUMMARY
- THE NORMAL AND POST LOCA/LOOP CONFIGURATIONS ARE ORIGINAL DESIGN s DRESDEN HAS A SIMILAR CONFIGURATION ( 1 LPCI AND 1 CCSW PUMP COMBINATION) FOR CONTAINMENT COOLING AS:
- QUAD CITIES STATION
- OYSTER CREEK
4
(
1
~
THE DESIGN BASIS OF THE PLANT HAS BEEN RECONSTITUTED IN A CONSERVATIVE MANNER AND PROVIDES ASSURANCE THAT EVEN IN THE MOST LIMITING CASES, KEY CONTAINMENT PARAMETERS REMAIN WITHIN ACCEPTABLE LIMITS.
John Stang Jim Dyer Dave Lynch Chandu Patel Jack Kudrick Robert Jones Tim Collins M. M. Razzaque Sonia Burgess Michael Peck K. B. Ramsden M. S. Tucker R. Ralph Pete Piet JoAnn Shields Sharon Eldridge Paul Dietz Brian Viehl S. Mintz D. J. Robare LIST OF ATTENDEES MEETING WITH COMMONWEALTH EDISON COMPANY TO DISCUSS THE CONTAINMENT COOLING SERVICE WATER SYSTEM AT DRESDEN, UNITS 2 AND 3 APRIL 8, 1993 Company NRC/NRR/PDIII-2 NRC/NRR/PDI II-2 NRC/NRR/PDI II-2 NRC/NRR/PDI I I-2 NRC/NRR/SCSB NRC/NRR/SRXB NRC/NRR/SRXB NRC/NRR/SRXB NRC - Region II I NRC - Region II I CECO/Nuclear Fuels CECO CECO/Dresden CECo/Nuclear Licensing CECo/Dresden CECo/Dresden CECo/Downers Grove CECo/Dresden GE/Technical Services GE/Licensing ENCLOSURE 2