ML17179A770

Forwards GENE-770-26-1092, ...Lpci/Containment Cooling Sys Evaluation, Rev 0 to Calculation NED-M-MSD-43 & Rev 0 to Calculation NED-M-MSD-49,in Response to 930222 Enforcement Conference Re Insp Repts 50-237/92-34 & 50-249/92-34

ML17179A770
Person / Time
Site:	Dresden
Issue date:	03/05/1993
From:	Piet P COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML17179A771	List: ML17179A770 ML17179A772 ML17252A753 ML20012G662
References
NUDOCS 9303110227
Download: ML17179A770 (10)

Text

_).

CommonvAth Edison 1400 Opus !We Downers Grove, Illinois 60515 March 5, 1993 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk

Subject:

Dresden Nuclear Power Station Units 2 and:*3 Usage of ANS 5. 1-1979 and Other Inputs to Validate

• the Design Basis of the Containment Heat Removal System for Long-Term Cooling Using One Low Pressure Coolant Injection <LPCI> Pump and One Containment Cooling Service Water CCCSW> Pump NRC Docket Nos. 50-237 and 50-249

References:

(a) Enforcement Conference held between members of Commonwealth Edison CCECo> and the NRC Staff to discuss matters related to the CCSW System at Dresden Station, dated February 22, 1993.

(b)

• NRC Inspection Report, 50-237/92034, and 50-249/92034{

dated February 11, 1993. . ....

Dear Dr. Murley:

As discussed during the Reference* <a> meeting; CECo is providing design input assumptions, supporting information, and analysis results to validate the design basis of the Containment Heat Removal System <CHRS> for long-term*

cooling as one LPCI pump and one CCSW pump. Major input parameters and output results from our re-analysis are summarized as follows:

LPCI/l CCSW 2 LPCI/2 CCSW LPCI Flow (gpm) 3881 8916 CCSW Flow Cgpm) 3071 4795 Torus Temp <°F> 186 171 Torus Pressure <psig max> 9.4 7.6 Torus Pressure <psig min) 5.9 4.4 Torus Design Temp (°F) 281 281 Torus Design Pressure <psig) 62 62 NPSH Required (ft) 25.7 26.9 NPSH Actual (ft). 39.72 40.3 090006  !.

( . 930~i 10221 930305 i I\

~DR

• ADOCK 05000237 . 11 1 r

\&. * .. PDR 1

\ L

Dr. Murley 2 - March 5, 1993

• The decay heat model was. chosen to- be more realistic than the original May-Witt curves. The ANS 5.1-1979 decay heat input was used and applied as described in the Attachments to this letters.

• The following additional conservatisms were then utilized to ensure overall conservative results of the analysis. <The conservatism is demonstrated by a comparison of the original FSAR peak temperature and pressure for the 2 LPCI/2 CCSW pump case with the revised analytical results in the 2 LPCI/2 CCSW pump case with nominal flows.

All of the water in the Feedwater system which could contribute to a higher calculated pool temperature was added to the reactor pressure vessel <RPV) and containment system. In addition, a conservative calculation of the energy in the feedwater system piping, heaters and other metal structures is added-.to the RPV/containment system.

The original FSAR analysis used an initial suppression pool temperature of 90°F. The current analysis used the Technical Specification maximum pool temperature of 95 °F.

The heat removal capability was conservatively calculated using General Electric methodology. The heat removal capability calculations used LPCI/CCSW flows which were reduced to actount for loop flow measurement inaccuracies. These conservatisms are documented in the Attachments to this letter.

The FSAR Figure indicates a peak temperature of 165 °F and a pressure of 6.5 psig; the revised analysis results in a peak temperature of 168 °F and a pressure of 7.2 psig.

The inputs to the recent analysis were chosen to obtain analytical results which have sufficient conservatism to ensure adequate margin to safety. Where variable choices for input values were available, the value which would maximize the temperature and pressure response was chosen. The critical parameter which determine the pressure response were then adjusted to provide a minimum pressure value_ to be input in the

• NPSH calculations.

The three critical areas for ensuring acceptability of the LPCI/CCSW containment cooling system are temperature and pressure response which are bounded by the structural design limits for the Drywell and Suppression Chamber (including Mark I condensation stability transients), providing/ /'.:

adequate NPSH for the LPCI and Core Spray pumps, and preventing radiological releases in excess of 10 CFR Part 100 limits. The pressure and temperature response aie addressed above <below 62 psig and 281°F). NPSH is also addressed above (25.7 ft required,*39.72 ft actual). Atmospheric radiological releases are bounded by the original .accident analysis results at the peak blowdown pressure of 47 psig. Effluent ra_diological releases are controlled

• by providing a leak tight heat exchanger and by ensuring a positive CCSW to LPCI differential discharge pressure at the heat exchanger <See Reference 16 of Attachment 2 for description)~

Dr. Thomas E. Murley - 3 - March 5, 1993 This calculational information is provided for your Staff's review *

~nd concurrence for Dresden Nuclear Power Station. A complete summary of all information associated with the re-analysis is provided in Attachment 2 to this letter. The attached calculations have been reviewed by Dresden'~

On~Site Review in accordance with Station procedures.

• The CHRS calculations were necessitated due to the inconsistencies discovered within the Updated Final Safety Analysis Report <FSAR> at Dresden Station related to the CHRS <CCSW). Several examples exist in Chapter 8 of the SAR that d~scribe the electrical configuration of the Station

<specifically, emergency diesel generator loading requirements> where a single CCSW pump is required for containment heat removal purposes. Chapter 5 of the SAR implies in some places that two CCSW pumps are required for long-term containment cooling purposes. There are other examples in Chapter 6 of the SAR that support either a single CCSW pump or a dual CCSK pump configuration.

A**summary of these. i neons i stenci es is presented in Attachment l to this letter.

As part of CECo's attempts to resolve the noted inconsistencies associated with CHRS as it relates to CCSW, a design basis reconstitution effort was undertaken to fully define all open issues. A summary of more notable discrepancies is noted in Attachment 1 to this letter. CECo concluded that confirmatory calculations using state-of-the art methodology were appropriate and proper in validating the design basis of CHRS Cl LPCI/l CCSW pump). This is further evidenced by the calculational results provided in .

CECo requests NRR concurrence of the analysis. Following NRR concurrence, it is Dresden's intention to modify the FSAR to remove inconsistencies. The revisions include clarifications of discrepancies pertaining to references made in various descriptions on the definition of the most limiting design basis LPCI/CCSW configuration.

To the best of my knowledge and belief, the statements contained above are true and correct. In some respect these statements are not based on my personal knowledge, but obtained.lnformation furnished by other Commonwealth Edison employees, contractor employees, and consultants. Such information has been reviewed in accordanc~ with company practice, and I believe it to be reliable.

Dr. Thomas E. Murley - 4 - March 5, 1993 If there are any questions related to this issue, please contact this off1 ce.

Sincerely,

@g~

Peter L. Pi et Nuclear Licensing Administrator Attachments: l. Summary of Design Basis Information Related to the CCSW System at Dresden Station

2. CECo Calculational Information for Validating the Design Basis for the Containment Heat Removal System to be One LPCI Pump and One CCSW Pump cc: A. Bert Davis, Regional Administrator-RI!!

J. Stang, NRR Project Manager-Dresden P. Hiland, RIII M. Leach, Senior Resident Inspector-Dresden Office of Nuclear Facility Safety-IONS

ATIACHMENT l Summary of Design Basis Information For CCSW at Dresden Station

1 LPCl/2 CCSW Pumps 1 LPCl/1 CCSW Pump SAR Tabla 6.2.4: "2 [pumps, emergency service SAR Section 6.2. 7 .3: "After a period not greater than water] required to provide required cooling capacity." two hours two of the LPCI pumps can be shut down Secondary side flow (river water) = 7000 gpm and one or two containment cooling service water pumps put into service to cool the suooression pool."

SAR Page 6.2-17: "Two LPCl/containment cooling SAR. Section 8, EOG Loading Table : "Containment service water pumps will deliver cooling water to each Cooling Water Pump #2 (Manual) 600 BHP (if within heat exchanger." the capability of the diesel aenerator)."

SAR Page 5.2-20: "d. Operation of only one of the SAR Figure 5.2. 1 1 - Curve 'd' is labeled: "1 /2 cont:

two core spray cooling system loops and one-half of cooling loop - 1 core spray." The shape of curve d one containment cooling loop. Namely one LPCI pump corresponds to resultant affects with 1 /1 pump and 2 service water pumps." operation.

NRC SER for Dresden Amendments Nos. 8/6 to DPR- SAR Table 8.2. 1: "After*a period not exceeding 2 19/25 (Change Nos. 34/23), dated 5/16/75: "Two hours the operator can manually stop one LPCI pump CCSW pumps provide adequate cooling capacity." and start a containment cooling water pump (460 bhp). This would achieve the containment cooling capability as specified in Section 5.2.3.3."

Original TSB for 3.5.B: "Loss of one containment Original SR 4.5.B.2: "When it is determined that one cooling service water pump does not seriously CCSW pump is inoperable, the remaining components jeopardize the containment cooling capability as any 2 of that subsystem and the other containment cooling of the remaining three pumps can satisfy the cooling subsystem shall be demonstrated to be operable requirements. Since there is some redundancy left a immediately and daily thereafter". [Note: Implies 2 of 30-day repair period is adequate. 3 remaining pumps as stated in original TSB 3.5.B, is for the affected containment coolina subsystem].

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 adeauate."

Special Report No. 33 - "Each unit at the Dresden Special Report No. 33 - "There are three other Station has four CCSW pumps, any two of which will possible sources of flood water in the condensate provide the required containment cooling. Each unit pump room. They are: 1. The three contaminated will have two of the above pumps in an isolated condensate storage tanks. 2. The condenser hotwell vault." and condensate piping to the condensate pumps. 3.

The CCSW system pipina to the CCSW pumps.

SR 4.5.B. 1: "Each containment cooling water pump shall deliver at least 3500 gpm against a pressure of 180 Psia." [no correspondina SR for 7000 aom)

SAR Page 5.2-22 states: "The containment pressure and temperature and 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."

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."

c:\ccsw3.doc

ATTACHMENT 2 CECo Calculational Information for Validating the Design Basis for the Containment Heat Removal System to be One LPCI Pump and One CCSW Pump

INDEX AND DESCRIPTION FOR REFERENCES ON LPCI/CCSW CONTAINMENT COOLING SYSTEM POST LOCA LONG TERM COOLING ANALYSIS

.1. GE Report GENE-770-26-1092. dated November 1992. Provides the summary results of the General Electric Analysis for Post LOCA Long Tenn Containment heat up with detailed descriptions of inputs and assumptions.

2. UFSAR Update and Supporting 10 CFR 50.59. Consists of mark-ups of the UFSAR pages for revision with 10 CFR 50.59 Evaluation.

3. Calculation NED-M-MSD-43. Revisions 0 and 1 on NPSH for LPCI and Core Spray Pumps. This calculation provided input for the UFSAR update in Reference 2.

• 4. *calculation NED.:M-MSD-49. Revision 0 on NPSH for LPCI and Core Spray Pumps.

This calculation provides the documentation for stating that the use of overpressure is not required to provide adequate NPSH for the nominal pump configuration of2 LPCI and 2 CCSWpumps. *

5. Letter from T. Rieck to C.W. Schroeder dated November 24. 1992 with Recommendations for Tube Replacement versus Plugging on LPCI Heat Exchangers.

The directions provided in this transmittal were utilized to revise the maintenance procedure used for performing heat exchanger retubing.

6. Letter. C.R. Parker to S. Eldridge. "LOCA Long Tenn Containment Response Analysis K-Values for LPCI/Containment Cooling System Heat Exchangers Dresden Nuclear.

Units 2 & 3. "October 6. 1992. (Reference 2 ofGENE-770-26-1092). Transmittal o(

proposed K values to be used as input for the GE analysis for CECo concurrence.

Concurrence was provided verbally on October 13, 1992 with reference.to Reference 7.

7. Memo to file dated October 13. 1992. K. Ramsden NFS-CECo. Provides verification of the GE calculated K values to be used in the Long Tenn Containment Post LOCA analysis.

8. Letter. S. L. Eldridge/B. M. Viehl to T. Allen. "Inputs for Heat Exchanger Parameters for CCSW Flow Issue. Dresden Units 2 & 3." August 31. 1992. (Reference 3 of GENE-770-26-1092) Provides the flows to be used in the calculation of heat exchanger duty for input into the GE analysis.

9. Letter. C.R. Parker to S. Eldridge. "LOCA Long Tenn Containment Response Analysis Input Parameters Dresden Nuclear Power Station. Units 2 & 3 (Final Values, September 21, 1992 (With CECo Approval Letter dated September 9, 1992) (Referen~e 7 of GENE-770-26-1092). Provides CECo approval of the input parameters to be used in the GE Containment analysis.

10. LPCI Containment Cooling System Process Diagram. GE Drawing 729E583. Rev 1.

February 24. 1969. (Reference 10 ofGENE-770-26-1092). Provides system design input infonnation which resulted from the original containment analysis.

11. Letter from S. Eldridge to C. Schroeder on "Submergence ofLPCI Discharge Line Post LOCA Dresden Units 2 & 3 11 dated September 29. 1992. (Reference 3 of NED-M-MSD-43 Calculation on NPSH). Provided suppression pool water level infonnation for use in calculation of the NPSH available for the LPCI and Core Spray pumps.

• 12. "Dresden LPCI/Containment Cooling System.". GE Nuclear Energy letter From S. Mintz

.to T. L. Chapman dated January 25. 1993. (Reference 2 ofNED-M-MSD-49 Calculation on NPSH) .. Provide additional infonnation on water level to be assumed in the supp~ession pool during containment cooling mode.

13. Sumnuu:y oflnput Parameters used in comparison with those used in the original FSAR analysis. This summary provides a discussion of the justification and/or source for the input parameters used in the Reference 1 analysis.

14. Letter from B. M. Viehl to C. W. Schroeder on CCSW Reduce Flow Issues. dated May

15. 1992 includes the results of the General Electric reconstituted heat exchanger duty

_ calculations. This letter ~!so provides the justification used to authorize use of ANS 5 .1 for the containment re-analysis.

15. Letter from S. Mintz (GE) to S. Eldridge (CECo) dated February 17. 1993. describing the use of ANS 5.1 in the Containment analysis. A technical description of the application of the decay heat from ANS 5.1-1979 for the Dresden Post LOCA Long Term Containment Cooling analysis provided.

16. Control of Effluent Releases. provides a discussion of the protection against effluent releases which would violate 10 CFR 100 release limits.