ML17179A998
| ML17179A998 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/12/1993 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | Wallace M COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 9307190159 | |
| Download: ML17179A998 (12) | |
Text
- )
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 Docket Nos. 50-237 and 50-249 July 12, 1993 Mr. Michael J. Wallace, Vic~ President Chief Nuclear Officer Commonwealth Edison Company Executive Towers West Ill, Suite 500 1400 OPUS Place Downers Grove, Illinois 60515 Dear Mr.
Wallace~
SUBJECT:
DRESDEN NUCLEAR POWER STATION, CONTAINMENT COOLING WATER SYSTEM DESIGN BASIS CONFIGURATION This letter responds to the Commonwealth Edison Company (CECo) submittal of March 5, 1993, which provided information supporting the assertion that the design basis configuration of the Dresden Containment Heat Removal System (CHRS) is one Low Pressure Coolant Injection (LPCI) pump and one Containment Cooling Service Water (CCSW) pump for long-term containment heat removal.
During an Enforcement Conference held on February 22, 1993, CECo made this assertion and the NRC requested that the safety review package supporting this conclusion be provided for staff review.
The CECo submittal included a listing of discrepancies in licensing documents indicating both the 1 LPCI/
1 CCSW and 1 LPCl/2 CCSW pump design basis configurations and a reanalysis of the design basis loss of coolant accident (LOCA) using modern codes demonstrating that the 1 LPCl/l CCSW pump configuration for CHRS was adequate.
Commonwealth Edison Company requested NRR staff concurrence of the analysis and stated its intention to modify the Updated Final Safety Analysis Report (UFSAR) to remove the inconsistencies.
The NRC staff concluded that the inconsistencies identified by CECo did not adequately support the assertion that the design basis configuration of the CHRS for the Dresden Nuclear Power Station has always been a 1 LPCl/l CCSW pump configuration. A summary of the staff's review is provided as.
Overall, the staff finds that the CECo examples of inconsistencies supporting 1 LPCl/l CCSW pump configuration appeared to be sentences taken out of context, misinterpretations or only implied that the design basis configuration was a one CCSW pump configuration.
On the other hand, the examples supporting the 1 LPCl/2 CCSW design basis configuration appeared to be very expli~it and were consistent with existing Technical.
Specification requirements.
Additionally, the listing of inconsistencies omitted examples supporting the 1 LPCl/2 CCSW design basis configuration such as inputs and safety evaluations for the Systematic Evaluation Program (SEP) and the recent Individual Plant Examination (IPE) submittal.
The NRC staff does agree with CECo that the existing UFSAR provides inconsistent information that, if not read as a whole, could be confusing and should be made clearer.
93071:90.1_59. 9,3071"2
\\.~.
)
>ADOC~:
1.'05000237.1 I P
PDR,
OFC NAME DATE Mr. Michael J. Walla~ auly 12, 1993 The NRC staff review of the reanalysis of the containment performance and calculations of net positive suction head (NPSH) for the 1 LPCl/l CCSW pump configuration identified concerns about the adequacy of your review.
The specific concerns involved questionable methods and inputs for the reanalysis and are identified in Enclosure 2 to this letter. Overall, the staff found that the reanalysis was not sufficiently conservative to support the conclusion that the 1 LPCl/l CCSW pump configuration would provide the necessary NPSH margin to ensure adequate long-term cooling for the core.
During the meeting at NRC on April 7, 1993, these concerns were discussed with the CECo staff.
At that meeting, CECo presented information that the Dresden Nuclear Power Station emergency diesel generator design could support the 1 LPCl/2 CCSW pump configuration. This presentation confirmed an earlier NRC review of this issue in Inspection Report 50-237/50-249/92034.
Based on this information, the NRC staff concluded that the Dresden Nuclear Station CHRS cnuld adequately remove decay heat from the containment during a design basis accident using two CCSW pumps.
This conclusion is consistent with the Dresden Nuclear Power Station Technical Specifications.
Based on the*review of the information provided in the CECo letter of March 5, 1993, and subsequent discussions held on April 7, 1993, the NRC staff concludes that the design basis configuration for the CHRS has always been 1 LPCI/2 CCSW pumps.
The staff does not concur with the analysis or proposed UFSAR changes provided in the CECo letter. Commonwealth Edison Company should reanalyze their design basis accident conditions for this configuration using appropriate analytical methods and assumptions and revise the Dresden Nuclear Station UFSAR to reflect an accurate analysis for the 1 LPCI/2 CCSW pump configuration.
The staff plans no further review of this submittal.
NRC staff approval of a 1 LPCI/l CCSW design basis configuration for the Dresden CHRS will require a license amendment.
The staff review of this proposed amendment will be based on a comprehensive submittal which clearly addresses all safety concerns and follows 10 CFR 50.90.
DISTRIBUTION:
NRC & Local PDRs J. Roe J. Oyer C. Moore Docket File PDill-2 r/f J. Zwolinski J. Stang OGC Sincerely, Original signed by John A. Zwolinski, Assistant Director for Region III Reactor ACRS (10)
R. Jones B. Clayton Rill R. Barrett Division of Reactor Projects - IIl/IV/V Office of Nuclear Reactor Regulation
Enclosures:
- 1.
Design Basis Configuration Comments
- 2.
CCSW Reanalysis Comments cc w/enclosures:
See next page
- See previous concurrence
- Bc:SRXB*
- RJONES
- 06/18/93
- Bc:SCSB*
- RBARRETT
- 06/14/93
- D: PDII I-2
- JOVER ::J#t'
- ? /tv/93
- AD: RI I*
- JZWOLINSKI
- 07 /07 /9~ *~
'\\ \\1--\\.
l Mr. Michael J. Wallace Commonwealth Edison Company cc:
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 Bo~rd of Supervisors of Grundy County Grundy County Courthouse Morris, Illinois 60450 Regional Administrator 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, 1llinois 60601 L. 0. DelGeorge, Vice President Nuclear Oversight and Regulatory Services Commonwealth Edison Company Executive Towers West III, Suite 500 1400 OPUS Place Downers Grove, Illinois 60515 Dresden Nuclear Power Station Unit Nos. 2 and 3
COMMENTS ON CECo'S
SUMMARY
OF DESIGN BASIS INFORMATION FOR CCSW AT DRESDEN STATION ENCLOSURE 1 to Conunonwealth Edison Company's (CECo) letter of March 5, 1993, provided a listing of inconsistencies within Dresden licensing documentation showing that both one and two Containment Cooling Service Water (CCSW) pumps are required for design basis accident mitigation. A CECo review of this information concluded that the design basis configuration for Dresden long~
term containment heat removal was one low pressure coolant injection (LPCI) pump and one CCSW pump.
The NRC staff has performed a review of the information provided and disagrees with the CECo conclusion.
The following paragraphs outline the NRC staff positions on the excerpts from Dresden Nuclear Power Station licensing documentation provided by CECo to support the assertion that the design basis configuration of the CCSW system is one pump.
- 1.
SAR Section 6.2.7.3: "After a period not greater than two hburs. two of the LPCI pumps can be shutdown and one or two containment cooling service water pumps put into service to cool the suppression pool."
The staff concludes that this excerpt could imply that one CCSW pump is the design bases configuration for the CCSW system, but firm support for this conclusion is not apparent.* Section 6.2.7.3 of the Updated Final Safety Analysis Report (UFSAR), discusses ECCS load sequencing during a large break LOCA.
The excerpted sentence does not appear relevant to the ongoing discussion. Although the title for Section 6.2.7.3.1 is "Design Basis Accident", it appears that the discussion involves less restrictive electrical lineups where both emergency diesels are available. It is not clear whether other nonconservatisms in areas such as decay heat load, torus water temperature, or service water temperature were considered in this discussion.
These factors could influence whether one or two CCSW pumps would be required.
- 2.
SAR Section 8, EOG Loading Table:
"Containment Cooling Water Pump #2 (Manual) 600 BHP Cif within the capabilitv of the diesel generator)."
The staff concludes that this excerpt does not support one CCSW pump as the design basis configuration of the CCSW system.
A further review of the emergency diesel generator (EOG) loading table indicates that both CCSW pumps #1 and #2 are listed as starting at the same time after the accident (within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) and the running loads include two LPCI pumps.
The note on the table is below the entire list of diesel generator loads and could be applied to more than just the CCSW Pump #2.
Accident analyses assume securing one of the two LPCI ~umps before starting a CCSW pump.
The adequacy of EOG loading was reviewed during the April 7, 1993, meeting with CECo and in Inspection Report 50-237/50-237/92034.
- 3.
SAR Figure 5.2.11-Curve 'd' is labeled:
"1/2 cont. cooling loop-I core sprav."
The shape of curved corresponds to resultant affects with 1/1 pump operation.
The NRC staff concludes that this assertion does not support one CCSW pump as the design basis configuration of the CCSW system.
Although the shape of the curves may be similar, the peak temperatures achieved with the l LPCl/l CCSW pump configuration appear to be different. A detailed discussion of this issue was provided in NRC Special Inspection Report 50-237 /50-249/92034.
The UFSAR, Section 5.2.3.3, defines 1/2 containment cooling loop as one LPCI pump and two service water pumps.
- 4.
SAR Table 8.2.l: "After a period not exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> the operator can manually stop one LPCI pump and start a containment cooling water pump (460 bhp).
This would achieve the containment cooling capability as specified in Section 5.2.3.3."
The NRC staff concludes th~t this excerpt supports one CCSW pump as the design basis configuration of the CCSW system.
However, it conflicts with the UFSAR Section 5.2.3.3 which states that the analyses considers two CCSW pumps in each loop.
- 5.
Original SR 4.5:8.2:
"When it is determined that one CCSW pump is inoperable, the remaining components of that subsystem and the other containment cooling subsystem shall be demonstrated to be operable immediately and daily thereafter."
(Note:
Implies two of three remaining pumps as stated in original TSB 3.5.B. is for the affected containment cooling subsystem.)
The NRC staff concludes that this excerpt does not support one CCSW pump as the design basis.configuration of the CCSW system.
This surveillance requirement pertains to alternate train testing of the CCSW system in addition to the other components of the CCSW system.
During a limiting condition for operation, the staff does not consider a single failure to the safety system so all three pumps are presumed to be available for accident mitigation. This configuration provides one additional pump beyond the required two CCSW pumps for design basis heat removal.
- 6.
Current TSB for 3.5.B.:
"Loss of one containment cooling service water pump or one LPCI pump does not seriously jeopardize the containment cooling capability as any 2 of the remaining three pumps can satisfy the cooling requirements. Since there is some redundancy left, a 30-day repair period is adequate."
The NRC staff concludes that this excerpt does not support one CCSW pump as the design basis configuration of the CCSW system.
The two sentences before the quoted excerpt state "The containment cooling subsystem consists of two sets of cooling equipment.
Each set contains 2 service water pumps, I heat exchanger and two LPCI pumps.
Either set of equipment is capable of performing the containment cooling function."
These
.. sentences indicate that the two CCSW pumps are required for adequate containment cooling.
- 7.
Special Report No. 33:
There are three other possible sources of flood water in the condensate pump room.
They are:
I. The three contaminated condensate storage tanks. 2. The condenser hotwell and condensate piping to the condensate pumps. and 3. The CCSW system piping to the CCSW pumps.
The NRC staff concludes that this excerpt does not support one CCSW pump as the design basis configuration of the CCSW system.
Section 2.2.2.1 of Special Report No. 33 states, "Each unit at the Dresden Station has four containment cooling service water pumps, any two of which will provide the required containment cooling."
As you noted in the enclosu~e to your letter, this sentence appears to provide an explicit statement of the design basis for the CCSW system as two CCSW pumps.
- 8.
SR 4.5.B.1:
"Each containment cooling water pump shall deliver at least 3500 gpm against a pressure of 180 psig."
CNo corresponding SR for 7000 gmLl.
The NRC staff concludes that this excerpt does not support one CCSW pump as the design basis configuration of the CCSW system.
The surveillance test requirement was intended to verify that there would be no degradation in the performance of either pump.
Combined pump testing could mask a single degraded pump.
The test acceptance value for each pump should be such that the pumps would collectively meet their intended safety function flows and pressures.
- 9.
SAR Page 5.2-22 states:
"The containment pressure and temperature are shown as curve 'd'. Figures 5.2.11 and 5.2.12 respectively. It is shown that. following the initiation of the single containment cooling pump and its associated heat exchanger. the containment pressure decreases initially then slowly increases to the maximum shown in Table 5.2.5 due to decay-energy addition to the containment. Thereafter. energy removal by the single containment spray cooling pump and heat exchanger exceeds the addition rate from all sources. resulting in decreasing containment pressure."
The NRC staff concludes that this excerpt does not support one CCSW pump as the design basis configuration of the CCSW system.
This excerpt refers to the LPCI pump in the containment spray mode and is consistent with the 1 LPCl/2 CCSW pump design basis configuration specified in Table 5.2.3:1.
Table 5.2.3:1 specified two service water pumps per heat exchanger in a footnote referring to all the LPCI and Core Spray pump combinations.
- 10. QCS 9/1/89 SER:
"... one RHR pump and one RHR Service Water pump will provide adequate containment cooling following a loss of coolant accident."
The NRC staff concludes that this excerpt does support one RHR Service Water pump as the design basis configuration for the Quad Cities Nuclear Power Station.
It does not provide a persuasive argument for the design basis configuration of the Dresden CCSW system.
ENCLOSURE 2 COMMENTS ON CECo'S CALCULATIONAL INFORMATION FOR VALIDATING THE DESIGN BASIS FOR CONTAINMENT HEAT REMOVAL SYSTEM TO BE ONE LPCI PUMP AND ONE CCSW PUMP to the CECo letter of March 5, 1993, provided the bases for the determination that the differences between the Dresden Nuclear Power Station Containment Cooling Service Water (CCSW) system and the Updated Final Safety Analysis Report (UFSAR) identified in April 1992 did not constitute an unreviewed safety question. This determination was based on the premise that the design basis configuration of the Containment Heat Removal System (CHRS) was one Low Pressure Coolant Injection (LPCI) pump and one CCSW pump.
Included in the attachment were the summary results of a General Electric (GE) analysis of the Dresden Nuclear Station CHRS for post-loss of coolant accident (LOCA) containment heatup and pressurization using more realistic computer codes for decay heat and containment performance modeling, and calculations demonstrating adequate net positive suction head (NPSH) for the LPCI under design basis conditions.
Also provided were the proposed changes for the Dresden UFSAR and supporting calculations and correspondence for inputs used in the calculations and analysis.
Based on the information provided, the NRC could not conclude that the one CCSW pump could provide adequate containment heat removal for the Dresden Nuclear Power Station.
The following concerns were identified during the NRC staff review:
- 1.
The GE analysis, GENE-770-26-1092, utilized the ANS 5.1-1979 Decay Heat Model and SHEX containment response computer code and adequate justification was not provided to show proper application to Dresden Nuclear Power Station.
No benchmarking of the code for Dresden application was provided except for a reference that a comparison of the peak temperature and pressure determined by the new analysis for the 2 LPCl/2 CCSW pump case with the 2 LPCl/2 CCSW pump case in the original FSAR yielded similar results.
The NRC staff concluded that the similarity of this limited comparison, which utilized different inputs for the fluid system flows and heat exchanger duties, did not provide adequate assurance of proper application.
- 2.
The NPSH calculations did not ensure that the most limiting conditions were considered.
Instead the NPSH calculations and containment analysis were coupled in that the pressure output at the time of the maximum temperature were both considered as inputs to the NPSH calculations.' The NRC staff was concerned that use of this coupled input for only four accident configurations may not reflect the most limiting NPSH conditions that could occur in a design basis accident.
- 3.
The calculations assume that the LPCI pumps are the most limiting case for the NPSH determination which appears inconsistent with earlier licensing analyses for the Dresden Nuclear Power Station. Calculation No.
NED-M-MSD-43, Revision 1, "Dresden LPCI/Core Spray Pumps NPSHA Evaluation-Post DBA-LOCA," states the core spray case is bounded by LPCI because similar suction losses, similar NPSHR curves, identical pump centerline elevations and core spray pumps run at lower flow than the LPCI pumps.
This assumption appears to be inconsistent with UFSAR Figure 6.2.7:29, "MINIMUM CONTAINMENT PRESSURE AVAILABLE AND CONTAINMENT PRESSURE REQUIRED FOR PUMP NPSH," which shows that the core spray pumps have a more limiting required NPSH than the LPCI pumps during a design basis accident.
- 4.
The application of flow uncertainties due to instrument inaccuracies for the LPCI and CCSW pump flows does not appear to represent the most limiting conditions for determining whether adequate NPSH exists.
In all cases instrument inaccuracies were subtracted from nominal or observed LPCI and CCSW pump flows.
For the determination of adequat~ NPSH, increased LPCI flows with reduced CCSW flows would appear to be more appropriate bounding conditions.
- 5.
The acceptance criteria for the NPSH calculations does not appear to provide assurance of a conservative conclusion.
In Calculation No.
NED-M-MSD 49, Revision 0, "Dresden LPCI/Core Spray NSPH Evaluation w/o Overpressure-Post LOCA," the acceptance criteria for adequate NPSH is defined as a success if the available NPSH is one percent below the vendor specified required NPSH.
It does not appear that the specified acceptance criteria provides sufficient assurance for continued pump operation in a design basis accident situation.
..tt** -
e Mr. Michael e July 12, 1993 The NRC
~taff review of.the reanalysis of the containment ~erformance and cal~ulations of net positive suction head (NPSH) for the 1 LPCI/1 CCSW pump configuration identified concerns about the adequacy ~f your review.
The*.
specific concerns involved.. questionable methods and inputs for the reanalysis and are identified in Enclosure 2 to this letter. Overall, the staff found that the reanalysis was not sufficiently conservative to support the conclusion that the 1 LPCI/1 CCSW pump configuratioh would provide the
- .. necessary NPSH margin to ensure adequate long-term cooling for the core.
- During. the meeting at NRC on April 7, 1993, these concerns were discussed with
-*. the CECo staff.
At th~:t:m.e~ting, CECo presented information that the Dresden Nuclear Power*
.S}at**io_n :emergency d_ieset generator aesign could sup_port the 1 LPCl/2 ccsw pump
. configuration;:. Jhis.,presentation confirmed an ear.lier NRC review of this
- ~*issue in Inspection -Report -50.,.237/50-249/92034.
Based on this information,
- . the* NRC *staff.*concludedth.at the 'Dresden Nuclear Station CHRS coul_d adequately r~rnoY.e decay heat.*from.:the. containment during a design basis accident using.
. -.,two_ C~SW pumps.* }his* conclusion is.consistent with_* the Dresden Nuclear Power
- _Statio.n Technical *specif,ications.
,, * *~, Based on the review, of the i nformat i cin provided in* the CE Co letter of March 5, 1993, and subsequ~pt discyssions h~ld on April 7, 19~3, the NRC staff conclu~e~_.that the de.sign basis configuration for the CHRS has always*-been 1 LPCl/2 CCSW pumps;,, The* staff does* not concur with the ana lysj s *or. proposed UFSAR changes provided in the CECo letter.* Commonwealth 1Edi*s.on Company should reanalyze their desi~n basis a~cident tonditirins for this configuration using appropriate analyti.cal.'methods *and assumptions and revise the Dresden Nuclear*
Station UFSAR to reflect an ac~urate analysis for the 1 LPCl/2 CCSW pump OFC NAME DATE.
. corifiguratiOI)... The.staff plans.no further review of this submittal.
NRC staff approval *of. a 1 LPC!fl*CCSW:design basis configuration for.the Dresden CHRS ~ill. require. a license amendment.
The staff review of this proposed amendment will be based on*a comprehensive submittal which clearly addresse~ all safety con~erns and follo~s 10 CFR 50,90.
DISTRIBUTION:
. Docket Fi*le NRC & Local PDRs **PDJ.II-2 r/f*
J. Rqe J. Zwolinski J. Dyer
. J. Stang. * "
- C. *Moore
- oGC Sjncerely, Original SiQ!led by..
John A.* Zwolinski, Assistant* Director
- for Region III-Reactor ACRS (10)
B. Claytrin Rill.
Division of Re~ctor Projecis - 111/IV/V Office of Nuclear Reactor Regulation R. Jones R. Barrett
Enclosures:
- 1.
Design Ba~is Configuration Comments *.
- l.
CCSW Reanalys~s Comments cc w/enclosures:
See nex*t. page C\\.ln*JI_/
~9Y
- f\\/\\1193
- See. previous concurrence iBC:SRXB*
iRJONES 106/18/93
- Bc:SCSB*
iRBARRETT
- 06/14/93 ID:PDIII-2
/JDYER ~
- I /tv/93
- AD: RI I*
- / JZWOLINSKI
- 07 /07 /9)'
/'\\ \\,1,\\
~-,
OFC NAME DATE
~
e Mr. L. 0. DelGeorge
- The NRC staff review of the reanalysis of the containment peiformance and calculations of net positive suction head (NPSH) for the 1 LPCl/l CCSW pump*
confi~uration identified concerns about the adequacy of your review.
- The specific concerns involved questionable methods and inputs for the reanalysis and are identified *in Enclosure 2 to this letter. Overall, the staff found that the reanalysis was not sufficiently conservative to support th~
conclusion that the 1 LPCl/l CCSW pump configuration would provide the necessary NPSH margin to ensure adequate long-term cooling for the core.
During the meeting at NRC on April 7, 1993, these concerns were discussed with the CECo staff.
At that meeting, CECo presented information that the D~esden Nuclear Power Station emergency.diesel generator design could support the 1 LPCl/2 CCSW pump configuration. This presentation confirmed an earlier NRC review of this*
issue in Inspection Report 50-237/50~249/92034. Based on this information, the NRC staff concluded that the. Dresdeh NucJear Station CHRS_ could adequately remove decay heat ftom the containment during a design basi*s accident using two CCSW pumps.
This conclusion is consistent with the Dresden Nuclear Power Station Technical Specifications.
Bas~d o~ the review of the information provided in the CECo letter of March 5, 1993, *and subsequent discussions: held on April 7, 1993 1 the NRC staff concludes that the design basis configuration for the CHRS has always been 1 LPCl/2 CCSW pumps.
The staff does not concur with the analysis or propos.ed UFSAR changes provided in the CECo letter. Commonwealth Edison Company should reanalyze their design b~sis accident*conditions for this configur~tion using appropriate analytical methods and assumptions and revise the Dresden Nuclear Station UFSAR to reflect an accurale analysis for the 1 LPCl/2 CCSW pump configuration.
The staff plans ~o further review of this submi.ttal.
NRC staff approval of a 1 LPCl/l CCSW design basis* configuration for the Dresden CHRS will require a license amendment.
The staff review of this proposed amendment will be based on a comprehensiv~ submittal which clearly addresses all safety concerns and follows 10 CFR 50.90.
- DISTRIBUTION:
NRC & Local PDRs J. Roe J. Dyer C. Moore Docket File PDill-2 r/f J. Zwolinski J. Stang OGC Sincerely,'
John A. Zwolinski, Assistant Director for Region III Reactor ACRS (10)
R. Jones B. Clayton Riii R. Barrett*
Division of.Reactor Projects - 111/IV/V Office of Nuclear Reactor Regulation
Enclosures:
- 1.
Design Basis Configuration Comments
- 2.
CCSW Reanalysis Comments cc w/enclosures:
See next page JLA:PDIII-2 JCMOORE l I
/93 JPM:PDIII-2 JJSTANG l I
/93
- See previous concurrence JBC:SRXB*
JRJONES J06/18/93 JBC:SCSB*
JRBARRETT 106/14/93 JD:PDIII-2 l JDYER
. : I
/93
- AD:R~
J"JZwoC ~Yi-,
- 111./93
OFC NAME DATE Mr. L. 0. DelGeorge
- The NRC staff review of the reanalysis of the contaiDment performance and calculations of net positive suction head (NPSH) for the 1 LPCl/l CCSW pump*
configuration identified concerns about the adequacy of your review.
Th~
~
specific concerns involved questionable methods and inputs for the reanalyiis and are identified in Enclosure 2 to this letter. Overall, the staff found that the reanalysis was not sufficiently conservative to support the conclusion that the 1 LPCl/l CCSW pump co~figuration would provide. the.
necessary NPSH margin to ensure adequate long-term cooling for the core*.
During the meeting at NRC on April 7, 1993, these concerns were' dis'cussed with.
the CECo staff.
At*that meeting, CECo presented information that the Dresden Nuclear Power Station emergency diesel generator design could support the 1. L'PCl/2 CCSW--p;ump * *.
configuration. This presentation confirmed an earlier NRC*re~ie~ of this.. v issue in Inspection Report 50-237/50-249/92034.
Based ori this infor111,ati.on,
.the NRC staff concluded that the Dresden Nuclear Station CHRS.could adequately remove decay heat from the containment during a design basis accident using.*
two CCSW pumps.
This conclusion is 'con~istent with the Dresden Nuclear Pqwer Station Technical Specifications.
Based on the review of the information provided in the CECo 1-etter of March-5, 1993, and subsequent discussions held on April 7, 1993, the NRG staff concludes that the design.basis configuration for the CHRS has al~ays b~en 1 LPCI/2 CCSW pumps.
The staff does not concur with the* analysis or proposed UFSAR changes provided in the CECo letter~ Commonwealth Edison CompahY ~~ould ~,
reanalyze their design basfs accident conditions for this configuration usi~g Y appropriate analytical -methbds and assumptions and revise the Dresden Nucle~r Station UFSAR to reflect an accurate analysis f6r the l LPCl/2 ccsw*pump configuration. Approval of a 1 LPCI/l CCSW design basis configuration for the presden CHRS will require a license amendment.
The staff.plans no further review of this submittal.
DISTRIBUTION:
- . NRC & Local PD Rs
- J. Roe J. Dyer C; Moore ACRS (10)
R. Jones
Enclosures:
Docket File.
PDIII~2 r/f J. Zwolinski.
J. Stang*
OGC B. Clayton Riii R. Barrett
- 1.
Design Basis Confi~uration Comments
- 2.
~CSW Reanalysis Comments cc w/enclosures:
See next page Sincerely, John A.
Zwoli~ski, Assistant Di~ector for Region III Reactor Division of Reactor Projects - lll/IV/V Office of Nuclear Reactor Regulation IAD:RIII I JZWOLINSKI I I
/93