ML17179A772

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Lpci/Core Spray Npsha Evaluaton W/O Overpressure - Post DBA-LOCA.
ML17179A772
Person / Time
Site: Dresden  Constellation icon.png
Issue date: 02/22/1993
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML17179A771 List:
References
NED-M-MSD-49, NED-M-MSD-49-R, NED-M-MSD-49-R00, NUDOCS 9303110237
Download: ML17179A772 (61)
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Text

COMMONWEALTH EDISON COMPANY TITLE PAGE

  • I CALCULATION NO.

~ SAFETY RELATED BED-H-MSD-49 CALCULATION TITLE l PAGE 1 NON-SAFETY RELATED OF 8 Dresden LPCI/Core Spray NPSBA Evaluation w/o overpressure - Post DBA-LOCA EQUIP NUMBER(S) STATION/UNIT SYSTEM 2(3) - 1502AlB/C/D Dresden 2 & 3 LPCI/Core Spray 2(3) - 1401A/B REV. CHRON # I PREPARER DATE I REVIEWER DATE I APPROVER DATE

  • I 0 198390 .~fik '-1;>/B f<IJ.Z rd.:'-- ~/~fiJi<S-trfl{-r;i.zfoi

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COMMONWEALTH EDISON COMPANY TABLE OF CONTENTS CALCULATION NO:

SECTIONS l

NED-M-MSD-49 TITLE PAGE

-l DESCRIPTION REV o I PAGE 2 OF 8 PAGES l

2 TABLE OF CONTENTS 2 3 REVISION

SUMMARY

3 4 CALCULATION SHEET(S) 4-7 5 REVIEW CHECKLIST 8 QE-Sl.D EXHmrrc REV. 3

COMMONWEALTH EDISON COMPANY REVISION

SUMMARY

CALCULATION NO: NED-H-HSD-49 I REV 0 I PAGE 3 OF 8 DESCRIPTION OF REVISIONS/REASON FOR CHANGE

~

AFFECTED PAGES PAGES REV. DESCRIPTION 1.,1£-51.D EXHmrro REV.3

calculation No. NED-M-MSD-49 Rev o Dresden LPCI/Core spray NPSB Evaluation w/o overpressure-Post LOCA Purpose/Obiective:

The Net Positive suction Head Available {NPSHA) for the Dresden LPCI/Core Spray pumps under post-OBA LOCA conditions was determined in Reference 1. The purpose of this calculation is to examine the effect ori NPSHA of removing the drywell/wetwell overpressure assumed in Reference l for the cases involving two pump operation, and comparing with NPSH required (NPSHR) to ensure pump protection.

Assumptions/Inputs:

The assumptions and inputs used in Reference l are also used for thi~ calculation with the following inputs changed:

1) Torus pressure is atmospheric pressure - 14.7 psia.

'* 2*) *'Maximum torus drawdown at time of analysis is 1 ft.

(versus 2.1 ft. in Ref. l) as provided and justified in Reference 2. This 1.1 ft. reduction in drawdown* results in a 1.1 ft. increase in the static head available to the pumps from 13.29 ft. to 14.39 ft *

References:

1) "Dresden*LPCI/Core Spray Pumps NPSHA Evaluation - Post DBA-LOCA", Nuclear Engineering Department Calculation No.

NED-M-MSD-43 Rev 1.

2) "Dresden LPCI/Containment Cooling System," GE Nuclear Energy letter from s. Mintz to T. L. Chapman dated January 25, 1993.
3) Hydraulic Institute Test Standards, 1988, Centrifugal Pumps - 1.6.
4) Hydraulic Institute Standards, 14th Edition, 1983.

5). General Electric Report No. GENE-770-26-1092 "Dresden Nuclear Power Station Units 2 & 3 LPCI/Containment Cooling System Evaluation," November, 1992 Equations:

Net Positive suction Head Available {NPSHA) is determined using the following equation:

{Pt - Pv)

NPSHA = 144 * + Z - hL (1)

Calculation No. NED-H-MSD-49 Rev o Dresden LPCX/Core Spray NPSB Evaluation w/o overpressure-Post LOCA where: Pt = Torus Pressure (psia)

Pv = Vapor Pressure* (psia) z = Static Head (ft) hL = suction losses (ft)

= density (lb/ft 3 )

J Acce2tance *criteria:

NPSH re~ired (NPSHR) is determined bf the pump vendor through testing as outlined in the Hydraulic Institute Standards (Reference 4*, pps. 73-77). Typically, the total head developed by a pump is monitored .at a constant flow while the supplied NPSH is reduced. At some point, the developed head begins to drop off as the pump starts to cavitate. Due to the difficulty in

~determining*exactly wpen the change begins, a value of 3% drop in head is usuallf accepted as evidence of the onset of cavitation.

The NPSH at this point is defined as the NPSH required "(NPSHR).

Decreasing the NPSH further results in a continuous (but not instantaneous) drop in head as the pump moves into full cavitation (Figure 1).

Due to this inexact method for determining the NPSHR limit, the small performance degradation beyond the 3% drop in head that does not occur instantaneouslf (Figure 1), and the conseryatisms used in Reference l and in this calculation, the acceptance limit for this calculation is defined as the following: if the NPSB available is ireater than or equal to the required NPSB (within an error of minus 1% of .the NPSBR), then pump protection is ensured and the ability of the pumps to perform their safety function is unaffected. It should also be noted that the use of zero torus.overpressure is extremely conservative for a post DBA-LOCA analysis as illustrated in Reference 5.

NPSBA Calculations; Using Equation l and the inputs provided above and in Reference 1, the NPSHA is calculated for each of the two pump cases (Table 1). The required NPSH is also provided and the difference between the two is calculated as "Margin".

summary/Conclusions:

Available NPSH for the LPCI/Core Spray pumps was determined in Reference l for post DBA-LOCA torus conditions. In the "calculation above, an additional conservative restriction of zero

  • drywell/wetwell overpressure was used to develop a set of NPSHA values for the cases involving two-pump operation. The results in Table l indicate that the available NPSH meets the acceptance criteria described above and is therefore adequate to protect the pumps under these conditions.
  • Calculation No. -M-MSD-49 Dresden LPCl/Core Spray NPSH Evaluation w/o Overpressure .. Post DBA-LOCA Total Single Torus Torus Static Specific Vapor Suction -1%

Flow Pp Flow Temp Press Head Volume Press Losses NPSHA NPSHR Margin Criteria Case (gpm) (gpm) . (f) (psia) (ft) (ft3/lb) (psla) (ft) (ft) (ft) (ft) (ft) 3 10000 5000 168 14.7 14.39 0.01644 5.722 5.87 29.8 30.0 -0.2 . -0.3 3A 8916 4458 171 14.7 14.39 0.016457 6.132 4.67 30.0 26.9 3.1 -0.27 Table 1

QJ

(

.. r Qz --

Q c .. t l&J 100% CAP Q:s

r:

...J f

3% Q4

  • )

I-0 ~r a.

z NPSH Fig. I TYPICAL NPSH TEST

REVIEW CHECKLIST CALCULATION NO: rJ E.t> - ~ - ft\ so* -if'I I REV. 0 I PAGE ~ OF ~

REVIEWED BY: J)~- f.< ~ DATE: 2/2L/73

.m  !:!Q REMARKS ID/' Cl t. IS lltE OBJECTIVE OF 'TltE ANALYSIS CLEARLY STATED7 g/ Cl Z. ARE ASSUMPTIONS AND ENGINEERING .JUDGEMENTS VAUD AND DOCUMENTE>7 Cl ic/"3. ARE THERE ASSUMPTIONS THAT NEED VERIFICATION7 a Cl 4. ARE THE REFERENCES CLE. DRAWINGS. CODES, STANDARDS!

LISTED BY REVISION eomoN. DATE. ETC.7 cY" Cl 5. IS THE DESIGN ME'nfOD CORRECT AND APPROPRIAlC FOR THIS ANALYSIS7

!SY" Cl 8. IS THE CALCULATION IN.COMPLIANCE WITH DESIGN auTEUA, CODES. STANDARDS. AND REG. GUIDES7

~Cl 7. ARE THE UNITS CLEARLY IDENTIFIED, AND EQUATIONS PROPERLY DERIVED AND APPUED7 Q/" Cl 8. ARE THE DESIGN INPUTS AND THEIR SOURCES IDENTIFIED AND IN COMPLIANCE WITH UFSAR & lCat SPECS7

~Cl 9. ARE THE RESULTS COMPATIBLE WITH THE INPUTS AND RECOMMENDA T!ONS MADE7

  • 10. INDICAlC TYPE OF CALCULATION CHAND-PREPARED AND/OR COMPUlCR-AIJEDJ AND METHOD OF REVIEW:

isniA~o PBEe!BEC CESIG~ t:!LCUbAIIQ~

THE REVIEW OF 'TltE HAND-PREPARED DESIGN CALCULA T!ON WAS ACCOMPLISHED BY ONE OR A COMBINA TlON OF THE

  • FOLLOWING IAS atECXEDI:

. ~ETAILED REVIEW OF lltE ORIGINAiCALCULATION

(

Cl A REVIEW BY AN AL'TERNA lC. SIMPLIFIED OR APPROXIMA1C ME'nfOD OF CALCULATION Cl A REVIEW OF A REPRESENTATIVE SAMPLE OF REPEiliiVE CALCULATIONS

/.REVIEW OF 'TltE CALCULA T!ON AGAINST A SIMILAR CALCULA TI~ PREVIOUSLY PERFORMED CJ COMPUTER AIDED DESIGN CALCIJbATION m  !:!Q m t!2 Cl Cl 11. IS THE PROGRAM APPLICABLE TO THIS PROBLEM7 0 Cl t&.* ARE THE RESULTS CONSISTENT WITH THE ASSUMPTIONS AND THE INPUT DATA7 Cl Cl 12. IS THE COMPUnR PROGRAM VALIDATED PER OP :J.i47 a 0 18. IS A UST OF THE PROGRAMS USED AND OA TE ..

a Cl 13. IS 1ME COMPUnR PROGRAM VAUDATED BY OTHER AFS I OF EAat COMPUnR RUN REFERENCED IN Tl4E ORGANIZATIONS AND HAS IT BE£N PREVIOUSLY APPllEO TO CALCULA T!ON7 NUCLEAR PROJECTS7 0 a 17. IS THE PAOGRAM VERSION ANO 1rs REVISION a Cl 14. IS THE INPUT DATA IN CONFORMANCE WITH IDENTIFIED ON THE COMPUTER RUN7 THE DESIGN INPUTS?

Q[-61.D EXHIBIT F REV.l

November 24, 1992

  • Mr. Brian Viehl

Subject:

Recommendations for Tube Replacement versus Plugging on LPCI Heat Exchangers

Reference:

CCSW Followup Actions Request for Assistance, Dresden Station, Commonwealth Edison, Brian Viehl;-May 1, 1992.

The reference letter requested that Nuclear Fuel Services investigate the relationship between tube replacement with AL-6XN material versus plugging. The intent was to develop an expression and/or graph of the number of tubes that can be replaced

, with new lower conductivity material while remaining within the design basis

' *. ;,.6*percent1 plugging .(of higher conductivity Monel tubes) assumption used to derive the LPCI heat exchanger thermal performance in various pump flow situations.

This work has been completed and is enclosed as an attachment to this transmittal.

As a rule of thumb, it has been determined that replacing 1 3 tubes yields the same reduction in overall heat transfer rate as plugging a single tube. With the design

  • basis of 144 plugged tubes, this allows for a considerable amount of tube replacement without affecting the design.

replacements versus plugging is *provided.

A curve of allowable numbers of It should be noted that the material thermal properties for the existing and replacement tube materials were taken from information provided by Dresden Station to NFS via fax. on 4/3/92. This transmittal consisted of excerpts of an S&L tube replacement modification review document. The thermal conductivities utilized in this analysis were:

Existing tubes (70/30 Cu-Ni) 19.0 BTU/hr-ft-F Al 6XN tubes (Stainless) 7.9 BTU/hr-ft-F These values have been reviewed for consistency with other sources and are believed to be conservative for the application presented. In particular, most handbooks present a*slightly lower conductivity value for Monel (by approximately 1 5%), which if used in this evaluation would lead to even higher allowable numbers of tube replacements per tube plugged .

November 24, 1992

  • This work has been performed and reviewed in accordance with NFS practices for the performance of safety-related work. If you have any questions regarding this matter,
  • please contact K. Ramsden or P. Kong of my staff.

~~*

\' - Terrance A. Rieck

  • T""° Nuclear Fuel Services Manager 1'~/l TAR: KBR: pc Attachment cc: NFM-CF S. Eldridg~

G. Lupia H. Massin C. Schroeder

. B. Wo11g

Dresden LPCI Heat Exchanger Tube Replacement with AL 6XN versus Plugging

~

150~~~~~~~;~~~~~~-~~~~--.~--.

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~ 120 --- -*- ___;::o~

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oL---~-1-~--'-~~~----1...~~--J__-...l..-~__J_~~~__J 0 200 400 600 800 1000 1200 . 1400 1600 1800 2000 Number of Tubes Replaced

  • I- Limit Point I based on 6% design plugging allowance
  • GE Nuciear Energy October 6,. 1992 Ms. Sharon Eldridge Commonwealth Edison Company Dresden Nuclear Power Station Rural Route #1 Morris, IL 60450

Subject:

LOCA Long-Term Containment Response Analysis K-values for LPCI/Containment Cooling System Heat Exchangers Dresden Nuclear Power Station, Units 2 &3

References:

  • 1.. Letter, G. G. Chen to S. Mintz, "K Va 1ues for Dresden Units 2 &3 Containment Heat Exchangers," September 14, 1992.
2. Letter, Sharon L. Eldridge/Brian M. Viehl to T. Allen, "Inputs for Heat Exchanger Parameters for CCSW Flow Issue Dresden Units 2 &3," August 31, 1992.
3. Letter, M.A. Wrightsman to B.M. Viehl, "CCSW Follow-Up Actions, Budgetary Est*imate 295-1C3PO-EBO," June 29, 1992.

Dear Ms. Eldridge:

1 This letter transmits the LPCl/Containment Cooling System heat exchanger

    • ,*K~values calculated by GE (Ref. 1) for the different flow conditions specified in Ref. 2. These K-values will be used to perform the Loss-of-Coolant Accident (LOCA) containment response analysis as described in the "Deliverables" section (for Item 1.) of Ref. 3.

Case #1 Case #1 Case #2 Case #2 (with loop (w/o loop (with loop (w/o loop accuracy) accuracy) accuracy) accuracy)

LPCI Flow (gpm) 3881 5000 8916 10000 CCSW Flow* (gpm) 3071 3500 4795 5600 K-value (Btu/sec-°F) 219.2 249.6 327.3 356.l

Sharon Eldridge

.. October 6, 1992 Pa e 2 These K-values ara based on a service water inlet flow temperature of 95°F and a LPCI inlet flow temperature of 165°F. The K-values are nearly constant as a function of the two inlet flow temperatures for the expected range of inlet flow temperatures. The heat removal rate, Q, of the heat exchanger is calculated using the K-value as follows:

Q = K (THX ,1n . - Tsw ,1n

. ) (Btu/sec) where THX,in =Heat exchanger inlet temperature (°F)

Tsw , 1. n =Service water inlet temperature (°F).

Please r.espond with your concurrence of the use of these heat exchanger K-values for the LOCA containment response analysis. If you have any questions on the above, please contact me.

  • My *fax number is (408)925-1674.

~1<..t~

Craig: Parker . *

... Plant *Analysis*Services (408) 925-2025, M/C 469 cc: GE C. C--~A 11 en S. Mintz J.E. Torbeck ORF T23-00685

'815s:l22s202s2~;# 11 I Co.

Dllpt. -

Fnaf/:Jf~ 2 .,.2 i._

October 13, 1992

Subject:

Verification of GE Derived LPCI Heat Exchanger Parameters To: File I have run MATHCAD flies to develop projected LPCI HX operating points for

, the flows given to GE by Dresden ENC. GE transmitted their new K*values In the attached fax received on 10/9/92. Using NTU methods with corrected heat

... , ,. exchanger tube.lengths from the drawings, values of performance were calculated for the same flow rates as GE. The results are attached, and the values are provided in the following tabl~.

Case 1 a Case lb Case 2a Case 2b LPCI flow 3881 5000 8916 10000 Ca pm)

CCSW flow 3071 3500 4795 5600 (QDm)

GE K Value 219.2 249.6 327.3 356.1 CBTU/sec*F}

GE HX at 70 degree OT 55.24 62.9 82.48 89.74 CM BTU/HR)

NFS NTU K Value 230.5 263.6 346.8 378.2 CBTU/sec*F>

NFS NTU HX at 70 58.08 66.43 87.4 95.31 degree OT

'(MBTU/HR>

The above table values demonstrates that GE Is more conservatively calculating the new performance points than would be predicted by the NFS NTU methods. This Is consistent *with the review of GE's methods performed earlier this year. Therefore we believe that the values developed by GE are appropriately conservative and will result in limiting pool temperature analysis. It appears that GE has about 5*6 percent more conservatism in their methods compared to the NFS NTU approach using public domain heat transfer coefficients.

August 31, 1992 To: T. Allen. General Electric

Subject:

Inputs For Heat Exchanger Parameters For CCSW Flow Issue Dresden Units 2 and 3 As you requested, below are the values for flow and temperatures to be used for your analysis of the Dresden Containments: Please note that two options are given for for each case, one with loop accuracies and one wothout. Calculations for both options are being requested.

  • Case l ( I LPCI - 1 CCSW )

FLOW WITH LOOP ACCURACY WITHOUT LOOP ACCURACY LPCI 3881 5000

. CCSW 3071 3500

  • Case 2 ( 2 LPCI I 2 CCSW )

FLOW WITH LOOP ACCURACY WITHOUT LOOP ACCURACY LPCI 8916 10000 ccsw 4795 5600 These values include instrument loop inaccuracies based on Tech Spec limits and values used for system surveillances. If you have any questions, please call S. Eldridge at Dresden extension 2956.

UL_o?~

,. Sharon L. Eldridge BWR Site Engineering BWR Site Engineering Design Supervisor cc: K. Ramsden S. Rhee NEDCC/Chron Sys Sup

September 9, 1992 In reply refer to CHRON # 0\155 t 7 To: Craig R. Parker, General Electric

Subject:

Approval of LOCA Long-Term Containment Response Analysis Input Parameters For Dresden Units 2 and 3.

Reference:

1. GE Letter from Craig R. Parker to S. Eldridge, dated August 7, 1992.

As per the agreement under Shipment Release YY-25 for the above scope of work, Commonwealth Edison has reviewed and approves for use the inputs provided with Reference

l. Please sign the final page and return a copy to me for our records. If there are any additional questions, please call me at (708)-942-2920 extension 2956.

/~L BWR Site Engineering

. .TJ-=-~----'*=-----=-V

. . . M. Viehl __';J.Q~~~

.Brian 1 c...

. BWR Site Engineering Design Supervisor cc: K. Ramsden - 111 NEDCC/Chron Sys Sup

September 21, 1992 Ms. Sharon Eldridge Commonwealth Edison Company Dresden Nuclear Power Station Rural Route #1 Morris, IL 60450

Subject:

LOCA Long-Term Containment Response Analysis Input Parameters Dre~den Nuclear Power Station, Units 2 &3 (Final Values)

Reference:

. Letter, M.A. Wrightsman 'to B.M. Viehl, "CCSW Follow-Up Actions, Budgetary Estimate 295-1C3PO-EBO," June 29, 1992.

Dear Ms. Eldridge:

  • This letter transmits the final signed-off version of the Containment Analysis Input Parameters Verification Form for Dresden Uni~s 2 & 3. This information is to be used to perform the Loss-of-Coolant Accident (LOCA) containment response analysis as described in the "Deliverables" section (for Item 1.) of the referenced* letter. The value(s) for "LPCl/Containment Cooling Heat Exchanger K in Containment Cooling Mode" have been determined by GE and will be transmitted to you (with a request for your concurrence) in a separate

.1 etter.

If you have any questions on the above, please contact me.

My fax number is (408)925-1674.

Sincerely,

  • ~~r~

Cr~Parker Plant Analysis Services (408) 925-2025, M/C 469

  • cc: GE C.C. Allen S. Mintz J.E. Torbeck ORF T23-00685

August 7, 1992 Ms. Sharon Eldridge Commonwealth Edison Company Dresden Nuclear Power Station Rural Route #1 Morris, IL 60450

Subject:

LOCA Long-Term Containment Response Analysis Input Parameters Dresden Nuclear Power Station, Units 2 &3

Reference:

. Letter, M.A. Wrightsman to B.M. Viehl, "CCSW Follow-Up Actions, Budgetary Estimate 295-1C3PO-EBO," June 29, 1992.

Dear Ms. Eldridge:

This letter requests your confirm~tion of the accuracy of, and agreement with the use of, the containment response analysis input parameters on the attached form. This information is to be used to perform the Loss-of-Coolant Accident (LOCA} containment response analysis as described in the "Deliverables" section (for Item l.} of the referenced letter. Please fill in the column of values entitled "CECo Proposed", sign the last page under "Verified and

'..Commented .by,". .,and return the form to me.

If you have any questions on the above, please contact me.

My fax number is (408)925-1674.

Sincerely,

~~.'/'~

Cr~Parker Plant Analysis Services (408} 925-2025, M/C 469 cc: GE C.C. Allen

s. Mintz J.E. Torbeck DRF T23-00685
  • CONTAINMENT ANALYSIS INPUT PARAMETERS VERIFICATION FORM FOR DRESDEN UNITS 2 &3 Values Resolved To Be Used, For GE ( 1) CE Co Containment Parameter Units UFSAR Proposed Proposed . Analysis Core Thermal Power MWt 2527 (2) 2578 (3)  ;).s7f?

Vessel Dome Pressure psi a 1020 (3) fO;:)..Q Drywell Free (Airspace) Volume (including vent system) ft3 158236 (4) 158236 \c:;~03G Initial Suppression Chamber Free (Airspace) Volume ft3 Low Water Level (LWL) 116300 (4) 120097 I;;loo:I7 High Water Level (HWL) 112800 (4) 116645 ll l, (c;tf-s-Initial Suppression*Pool Volume ft3.

Max. Water Level 119800 (4) 115655 ll c; "SS-Min. Water Level 116300 (4) 112203 2 ooo (2.*i) l l J. 000

  • CONTAINMENT ANALYSIS INPUT PARAMETERS VERIFICATION FORM FOR DRESDEN UNITS 2 &3 (CONTINUED)

Values

.Resolved To Be Used For GE CE Co Containment Parameter Units UFSAR Proposed Proposed Analysis Initial Drywel l Pressure psig 1.25 (5), \. z.c;;;

(6)

Initial Drywe 11 Temperature "F 135 (7) l 3S-Initial Drywe 11 Relative Humidity  % 20 (8) ;2-o Initial Suppression Chamber Pressure psig 0.15 (5) O.ls-Initial Suppression Chamber Airspace Temperature "F 95 (9) 9S-Initial Suppression Chamber Airspace Relative Humidity  % 100 (9) l ()C)

Initial Suppression Pool Temperature "F 95 (10) l1c:>

No. of Downcomers 96 (4) 96 '}~

Total Downcomer Flow Area ft 2 301.6(11) '30/.(p Initial Downcomer Submergence ft (HWL) 4.00 (4) 4.00 4.00 (LWL) 3.67 (4) 3.67 3-" 7

CONTAINMENT ANALYSIS INPUT PARAMETERS VERIEICATION FORM FOR DRESDEN UNITS 2 &3 (CONTINUED)

Values Resolved To Be Used, For GE CE Co Containment Parameter Units UFSAR Proposed Proposed Analysis Down comer 1.0. ft 2.00 (4) 2.00 ;J.oo Vent System Flow Path Loss Coef-ficient (includes exit loss) (12) 5.. 17 ( 13) S.17 Supp. Chamber (Torus) Major Radius ft 54.50 (4) 54.50 sj.so Supp. Chamber (Torus) Minor Radius ft 15.00 (4) 15.00 tS-.oo Suppression Pool Surface Area (in contact with supp.

chamber airspace) ft2 't~ 11.'t 9. *~r11.Lj.

Supp. Chamber-to-Drywell Vacuum Breaker Opening Diff. Press.

- start psid 0 .15 (14) O.L'5

- full open psid 0.5 (4) ff. 5 (0.S'"

Supp. Chamber-to-Drywell Vacuum Breaker Valve Opening Time (15) sec 1.0 (16) LO Supp. Chamber-to-Drywell Vacuum Breaker Flow Area (per valve assembly) ft 2 3 .14 (4) 3.14 3.1'-f

CONTAINMENT ANALYSIS INPUT PARAMETERS VERIFICATION FORM FOR DRESDEN UNITS 2 &3 (CONTINUED)

Values Resolved To Be Used.

For GE CE Co Containment Parameter UFSAR Proposed Proposed Analysis Supp. Chamber-to-Drywell Vacuum Breaker Flow Loss Coefficient (including exit loss) 3.47 No. of Supp. Chamber-to-Drywell Vacuum Breaker Valve Assemblies (2 valves per assembly) (17) 6 (4) 6 LPCI/Containment Cooling Heat To* be de- lo \>-L Exchanger Kin Containment Cooling _termi ned .111.\<L.-....... ~.l.

Mode (18) Btu/s-°F by GE. \c..\- e..r.

LPCl/Containment Cooling Service Water Temperature OF 95 (19)

LPCl/Containment Cooling Pump Heat (per pump) hp 700 (20) ""]oo Core Spray Pump Heat (per pump) hp 800 (21)

Time for Operator to turn on LPCI/Containment Cooling System in Containment Cooling Mode (after LOCA signal) sec 600 (22) 600 (23)

CONTAINMENT ANALYSIS INPUT PARAMETERS VERIFICATION FORM

. FOR DRESDEN UNITS 2 &3 (CONTINUED)

Values Resolved To Be Used For GE CE Co Containment Parameter UFSAR Proposed Proposed Analysis Feedwater Addition (to RPV after See o~.

start of event; mass and energy) attached table.

.~.: .,.

~ ._i_ _.._.*_r_,.._~_u_....1_*._ri_'(J_z._"_~_*.....--*-....-l!it :-=~~~~,~- .

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~,-,-,~,---i-,-t-,.-c-. ...,--,-,-t-,,--~-~.
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CONTAINMENT.ANALYSIS INPUT PARAMETERS VERIFICATION FORM FOR DRESDEN UNITS 2 & 3 (CONTINUED)

Notes:

1. The proposed values are derived from GE-NE documents 22A5743, "Containment Data" (Dresden 2),

rev. 1, 4-30-79 and 22A5744, "Containment Data" (Dresden 3), rev .. 1, 4-30-79 .

. 2. UFSAR Section 5.2.3.2, p. 5.2.3-4. Not~: This corresponds to 100% rated th~rmal power (see N6te 3).

3. GE-NE document, 459HA997, "Heat Balance, Reactor System", rev. O, 1-21-81 (102% rated thermal power).
4. UFSAR Section 5.2.2, Table 5.2.2:1.
5. GE-NE document, NED0-24566, "Mark I Containment Program Plant Unique Load Definition, Dresden Nuclear Power Station: Units 2 & 3", rev. 2, April 1982.
6. Based on operating pressure"differential of 1.1 psid (see Tech. Spec._Dresden Unit 3, Sect. 3.7, Amend. 75, p. 3/4.7-17).
7. Nominal value.
8. Minimum value.
9. Suppression chamber airspace assumed to be in thermodynamic equilibri~m with the suppression pool.
10. Maximum Tech. Spec. value (Dresden Unit 3, Sect. 3.7, Amend. 75, p. 3/4.7-2).
11. 301.6 ft 2 = n x (2.00 I.D.) 2/4 x (96 downcomers).
12. Based on downcomer flow area.

13~ GE-NE document, NED0-21888, "Mark I Containment Progra~ Load Definition Report", rev. 2,* Nov. 1981.

14. Not a critical parameter.

CONTAINMENT ANALYSIS INPUT PARAMETERS VERIFICATION FORM FOR DRESDEN UNITS 2 &3 (CONTINUED)

Notes:

15. Time required for valve *assembly to go from a fully closed position to a fully open position with the full-open differential pressure applied across the valve assembly. The full-open differential pressure is defined in the item above. (A valve assembly is two valves in series.)
16. Estimate of maximum opening time.
17. A vacuum breaker valve assembly is two valves in series.
18. K = Q I (Tpool - Tsw) where: Q = heat exchanger heat removal rate (Btu/sec)

Tpool =heat exchanger hot-side (supp. pool) inlet temperature (°F)

Tsw heat exchanger cold-side (service water) inlet temperature (°F)

19. .UFSAR Section 6.2.4, Table 6.2.4:1.
20. GE Motors document, 992C510, "Outline (Induction Motor)" (LPCI), rev. 4, 1-3-68.
21. GE Motors document, 992C510AB, "Outline (Induction Motor)" (Core Spray), rev. 6, 6-25-68.
22. UFSAR Section 5.2.3.3, p. 5.2.3-12 (One Core Spray and One Containment Spray Cooling Pump Operation case).
23. Standard assumption for operator action time for Mark I containment plants.
24. The feedwater system table is derived from data in the'Nutech letter, G.R. Edwards (Nutech) to T.J.

Mulford (GE), "Dresden and Quad Cities Containment Data" (Feedwater System - Metal and Liquid Masses, and Temperatures)*, COM-01-156, August 7, 1978. The actual feedwater addition to the RPV used in the containment analysis ts based on the data in the feedwater system table. i

CONTAINMENl ANAlYSIS INPUT PARAMETERS VERIFICATION FORM FOR DRESDEN UNITS 2 &3 (CONTINUED)

Prepared by: ~-l-'i2 Signa ure, Name, Organization, GE Nuclear Energy Date Verified and Commented by:

Final Values Resolved by:

1-2.i-<12.

ure, Name, Organization, GE Nuc ear Energy. Date

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September 29, 1992 In Reply Ref er to Chron 110115532 TO: C. Schroedel'.

SUBJECT:

Submergence of LPCI Discharge Line Post LOCA Dresden Units 2 and 3

REFERENCES:

1. Letter from S. Mintz, GE, to S. Eldridge dated September. 24 1992
2. Dresden FSAR/UFSAR, Section 5.2.3 for Minimum Submergence of Torus Downcomers of 3.67 ft.
3. Dresden Technical Specification, Section 3.7, Bases, pages 4.

B 3/4.7-33 for minimum submergence Drawings M-3302, M-3502

'\

~f

5. CB&I Drawings for Contract No. 9-3600 Numbers: 1, 200, 211

\

6. Dresden Primary Containment Pathways Evaluation, On Site Review No. 92-42,for Penetrations X-310A and X-3108 At the request of the Station Technical Staff, BWR Engineering has determined the minimum submergence of the LPCI discharge nozzles in the Suppression SI Chambers, Dresden Units 2 and 3. This evaluation is required to support the  ;;re information provided in the "Dresden Primary Containment Pathway Evaluation",

Reference 6. The conclusion in the above report states that if the nozzle -L¥.

remains covered post LOCA and is therefore not exposed to containment atmosphere, the valves on the associated line are not required to be Local leak Rate Tested (LLRT) per 10CFRSO Appendix J.

From Refer~nce 1 the maximum draw down expected post LOCA is 2.1 feet.

Apply-ing this draw down to the low water level as delineated in the Tech Spec and FSAR/UFSAR (References 2 and 3) the post LOCA minimum water level at Elevation 491 '-5" provides a minimum submergence of 6" *for the subject nozzles. The atta*ched sketch shows the geometry of the respective systems. and is a compilation from References 4 and 5.

The information in this evaluation is approved for the Stations use. If you have any other questions, please call me at extension 2956.

J:>~ v~ 't/

Br_i.._an--M-.-V-i-eh_l_....-==-=------- I '1 /

.... "C BWR Site Engineering /~ "'-

Design Supervisor BMV/SLE/maf (ZENCNE91/98) cc: M. Strait M. Andjelic H. Hassin NEDCC/Chron Sys. Supv.

September 24, 1992 Ms. Sharon Eldridge Commonwealth Edison Company Rural Route # 1 Dresden Nuclear Power Station Morris, IL 60450 *

Subject:

Dresden Units 2 &3 - Post LOCA Pool Drawdown

References:

!) Telecopy, J.C. -Elliott to S. Mintz, "Small Job Authorization," September 21, 1992. (Small Job Task Number DR127)

J

Dear'Ms. Eldridge:

In response to your request and per Reference 1, this letter provides an estimate of the maximum reduction in the Dresden suppression pool level following a loss-of-coolant accident (LOCA). - The maximum suppression pool level decrease during a LOCA for Dresden is estimated to be 2.1 feet. This pool level decrease is the same as the value estimated for Quad Cities previously. A comparison of the key parameters between Quad Cities and Dresden confirmed that these plants are essentially identical and therefore allows application of the Quad Cities pool level decrease estimate to Dresden. This

_pool level decrease can be used together with the lower Technical Specification limit on pool level to determine the minimum post-LOCA pool level.

This estimate of 2.1 feet pool level reduction post-LOCA is based on a conservative analysis of another BWR with a 251 inch ID reactor pressure vessel

  • i
  • and*'a Mark I*tontainment~ This analysis considered the water which may be transferred to_the reactor pressure vessel, drywell and drywell-to-wetwell vent system from the suppression pool during a LOCA. A more detailed calculation could be performed with the Dresden specific geometry, but it is not expected that the results would change significantly.

Documentation and evidence of verification of this letter is included in the GE design record file, DRF-T23-00692.

Sincerely,

,.-1~~~

s. Mintz ----

cc: g J. E. - Torbeck C. T. Young J. E. Nash

  • GE NUCLEAR ENERGY Sarr Jose, CA

________________PLANT PERFOR/tlANCE ANALYSIS PROJECTS January 25, 1993 To: T. L. Chapman From: S. Mintz

Subject:

Dresden LPCl/Containment Cooling System Attachment 1 provides the responses to CECO questions requested by Sharon Eldridge (CECO) in connection to the Dresden LPCl/Containment Cooling System evaluation. Please provide thes~ responses to Sharon Eldridge of CECO.

Documentation and evidence of verification is contained in DRF-123-685.

~/!~

S. Mintz

  • cc J. E. Torbeck J. E. Nash C. T. Young DRF-123-685

ATIACHHENT 1 RESPONSE TO CECO QUESTIONS DRESDEN LPCI/CONTAINHENT COOLING SYSTEM EVALUATION Question 1:

What is the impact on the peak suppression pool temperature calculated for Case 4a of GENE-770-26-1092 of assuming Containment CooHng Service Water (CCSW) initiation at 1800 seconds for containment cooling instead of the time of 600 seconds used for the GENE-770-26-1092 analysis?

Response to Question 1:

It is estimated that the peak suppression pool temperature for Case 4a of GENE -770-26-1092 w.ill increase by no more than 2*F to no more than 1sa*F. This temperature increase was estimated by deterinining the energy removed by the LPCI/Containment Cooling System heat exchanger between 600 and 1800 seconds for Case 4a. This energy was then added to the suppression pool at*

the time of the peak suppression pool temperature to determine

.the increase in the peak suppression pool temperature.

It should be noted that this temperature increase is conservative since it is more representative of the increase in the pool temperature at 1800 seconds (i.e., at the new initiation time for CCSW) than the increase in the peak pool temperature. Once CCSW is initiated this increase in the suppression pool temperature will be reduced due to the increased effectiveness of the heat exchanger with a higher suppression pool temperature.

  • 'Question 2:

What is the impact of initiation of CCSW at 1800 second (with estimated increase in the suppression pool temperature to 188.F) on the containment pressure at the time of the peak suppression pool temperature relative to the value given in Appendix B of Reference 1 for Case 4a?

Response to Q~estion 2:

At the time of the peak suppression pool temperature the suppression chamber airspace temperature is very close to the water temperature of the containment sprays. A change in the peak suppression pool temperature of 2*f will produce a change in temperature of the spray water temperature of approximately l*f. Therefore the suppression chamber airspace temperature will also increase by this amount. The change in the suppression chamber vapor pressure for this change in airspace temperature is approximately 0.1 psi. This is therefore the estimated increase in the suppression chamber airspace pressure due to an increase in the suppression pool temperature of 2*f to lss*F.

  • It should be noted that this estimate conservatively neglects (with respect to minimizing the containment pressure) the effects of the increase in the drywell temperature and drywell vapor pressure with an increase in the suppression pool
  • ; temperature.* An increase in the drywell pressure would result in an additional increase in the suppression chamber *airspace pressure due to the flow through the suppression

. chamber-to-drywell vacuum breakers.

Question 3:

What is.the maximum drawdown expected for a LOCA for Dresden?

Response to Question 3:

Reference 2 reported *a maximum drawdown of 2.1 feet for Dresden.

This was based on a conservative application of pool drawdown analyses conducted for a range of break sizes for another plant with a 251 inch ID reactor pressure vessel and a Mark I

  • * ' ***containment. However this maximum drawdown (which occurs only for breaks smaller than the DBA-LOCA) .would only occur early in the LOCA (< 1000 seconds) when suppression pool temperatures are significantly cooler than the maximum pool temperature. During the times of peak suppression pool temperatures when NPSH is of greatest concern a maximum drawdown of 1 foot is expected.

Question 4:

What is the impact of using May-Witt or ANS 5.1 + 2a on the peak suppression pool temperature determined for Case 4a of GENE-770-26-1020?

Response to Question 4:

The impact of using May-Witt on the peak suppression pool temperature is estimated based on a comparison of the decay heat near the time of the peak suppression pool temperature. The basis for using this comparison is that at the time of the peak suppression pool temperature the heat rejection rate to the suppression pool from the reactor pressure vessel (RPV) is equal to the rate of containment cooling which is itself established by the suppression pool temperature. The peak suppression pool temperature for Case 4a occurs at approximately 30000 seconds.

The value of the May-Witt decay heat is approximat_ely 15~ higher than the ANS 5.1 value at this time. This would produce an increase of approximately 14*f in the peak suppression pool temperature. The decay heat for ANS 5.1 + 2u is less than the May-Witt decay heat therefore the increase with ANS 5.1 + 2u will be bounded by the 14.F increase estimated for the May-Witt decay heat *

References:

1) "GE Report .GENE-770-26-1092, *oresden Nuclear Power Station, Units 2 and 3, LPCl/Containment Cooling System Evaluation,* November 1982.
2) Letter, S. Mintz to s. Eldridge, *oresden Units 2 & 3, - Post LOCA Drawdown,* September 24, 1992.
  • CONTAINMENT ANALYSIS INPUT PARAMETERS PARAMETERS UNITS UFSAR NEW JUSTIFICATION Section Value.

Cote Thermal Power MWt 2,527 2,578 GE-NE document, 459HA997, "Heat Balance, Reactor System", Rev. o, 1/21/81 -

102% rated thermal power Vessel Dome Pressure psia 1020 GE-NE document, 459HA997, "Heat Balance, Reactor System", Rev. o, 1/21/81 -

102% rated thermal power Drywell Free (Airspace) ft 3 158,236 158,236 N/A Volume (including vent system)

Initial Suppression Changed due to conservatism Chamber Free (Airspace)

Volume ft 3 Low Water Level (LWL) 116,300 120,097 High Water Level (HWL) 112,800 116,645 Initial Suppression Pool Changed due to conservatism Volume ft 3 Max. Water Level 119,800 115,655 Min. Water Level 116,300 112,203

CONTAINMENT ANALYSIS INPUT PARAMETERS PARAMETERS /

UNITS UFSAR NEW JUSTIFICATION Section Value Initial Drywell Pressure psig 1.25 GE-NE document, NED0-24566, "Mark I Containment Program Plant Unique Load Definition, Dresden Nuclear Power Station: Units 2.&3 11 ,

Rev. 2, April 1982 Based on operating pressure differential of 1.1 psid (Dresden Tech Spec. Unit 3, Sect 3.7, Amend 75, p.

3/4.7-17)

Initial Drywell OF 135 Nominal value Temperature Initial Drywell Relative  % 2*0 Minimum value Humidity Initial Suppression psig 0.15 GE-NE document, NED0-24566, Chamber Pressure "Mark I Containment Program Plant Unique Load Definition, Dresden Nuclear Power Station: Units 2&3",

Rev. 2, April 1982 Initial Suppression op 95 Suppression chamber airspace Chamber Airspace in thermodynamic equilibrium Temperature with the suppression pool Initial Suppr~ssion  % 100 Suppression chamber airspace Chamber Airspace in *thermodynamic equilibrium Relative Humidity with the suppression pool

CONTAINMENT ANALYSIS INPUT PARAMETERS PARAMETERS UNITS UFSAR NEW JUSTIFICATION Section Value Initial Suppression Pool op 95 Maximum Tech. Spec. value, Temperature Section 3.7, Amend. 75, p.

3/4.7-2 Number of Oowncomers 96 96 N/A Total Downcomer Flow ft 2 301. 6 This value is e~ual to Area "x (2.00 I.O.) /4 x (96 downcomers)

Initial Downcomer N/A Submergence ft (HWL) 4.00 4.00 (LWL) 3.67 3.67 Down comer I.D. ft 2.00 2.00 N/A Vent System Flow Path 5.17 GE-NE document, NED0-21888, Loss Coefficient "Mark I Containment Program (includes exit loss) Load Definition Report",

Rev. 2, November 1981 S&L Calculation:

NSLD-3C2-0978-001, Rev. o, 12/8/78 Suppression Chamber ft 54.50 54.50 N/A (Torus) Major Radius .

Suppression Chamber ft 15.00 15.00 N/A (Torus) Minor Radius

  • CONTAINMENT ANALYSIS INPUT PARAMETERS

. I PARAMETERS UNITS UFSAR NEW JUSTIFICATION Section Value Suppression Pool Surf ace ft 2 9~971.4. S&L Calculation:

Area (in contact with NSLD-3C2-0978-001, Rev. o, suppression chamber air 12/8/78 space) '

Suppression Chamber-to-Drywell Vacuum Breaker Opening Differential Pressure psid i

- start 0.15 Not a critical parameter

-' full open 0.5 0.5 N/A suppression Chamber-to- sec 1.0 Estimate  ;

Drywell Vacuum Breaker Valve Opening Time Suppression Chamber-to- ft 2 3.14 3.14 N/A Drywell Vacuum Breaker Flow Area (per valve assembly)

Suppression Chamber-to- 3.47 Drywell Vacuum Breaker Flow.Loss Coefficient (including exit loss)

Number of Suppression 6 6 N/A Chamber-to-Drywell Vacuum Breaker Valve Assemblies (2 valves per assembly)

CONTAINMENT ANALYSIS INPUT PARAMETERS PARAMETERS UNITS UFSAR NEW JUSTIFICATION

{

  • section Value LPCI/Containment Cooling Btu/s-°F To be Heat Exchanger K in determined Containment Cooling Mode by G.E.

LPCI/Containment Cooling OF 95 UFSAR Section 6.2.4, Table Service Water 6.2.4:1 Temperature LPC.I/Containment Cooling hp 700 GE Motors document, 992C510, Pump Heat (per pump) "Outline (Induction Motor)"

(LPCI), Rev. 4, 1/3/68 Corie Spray Pump Heat hp 800 GE Motors document, (per pump) 992C510AB, "Outline (Induction Motor)" (Core Spray), Rev. 6, 6/25/68 Time.for Operator to sec 600 600 N/A turn on LPCI/Containment Cooling system in Containment Cooling Mode (after LOCA signal)

Feedwater Addition (to See Attachment RPV after start of event; mass and energy)

May 15, 1992 To: c. w. Schroeder

Subject:

ccsw Reduced Flow Design Issues Dresden Units 2 and 3

References:

1) Letter dated April 7, 1992 from B. M. Vlehl to
c. w. Schroeder transmitting results of reduced CCSW flow review.
2) Letter # E12-00126 from General Electric to B. M. Viehl, dated May 13, 1992,attached.
3) Letter from T. Rieck, Nuclear Fuel Services Providing acceptance of use of ANS 5.1 Heat Decay Methodology and a Sensitivity Evaluation of the Effect of LPCI Heat Exchanger Performance during TransientsJdated May 15, 1992, attached. .
4) Letter from s. Powers to B. M. Viehl, dated May 15, 1992, with ENC-QE-81 review of GE letter (reference 2) attached.

The purpose of ~his letter is to transmit the responses to items 2: and 3 follow-up items from Reference 1, concerning the LPCI Heat exchanger duty Calculations and use of instrument accuracy in these calculations.

R~sults of Calculation I

The original LPCI heat exchanger duty calculation could not be

.retrieved. ; .A .new calculation was performed and resulted in a 9%

d~crease in heat removal capability for 1 pump/l pump operation

(~ode C). The decay heat methodology used in the original Containment Analysis is not consistent with today's accepted d~cay heat methodology. Today's decay heat methodology provides approximately a 15% margin. This results in a 15% - 9% = 6%

overall positive margin in the Containment Analysis.

Discussion The analysis (Reference 2) was performed to .obtain an appropriate heat exchanger duty for support of the parameters shown on the original LPCI system Process Flow Diagram, GE drawing 729E583, Rev 1. A current analysis was performed using current heat exchanger and decay heat methodologies after a search of both General Electric and Senior Engineering (the current heat

  • exchanger vendor company name) records failed to locate any original design calculations. The analysis shows that the heat

exchangers have sufficient heat removal capability under the reauced flow conditions of l LPCI pump and l CCSW pump (Mode C) to mitigate the required accident conditions. The calculated heat duty available under l pump/l pump operation is less, but use of the current decay heat methodology from ANS 5.1 provides a greater reduction in the heat input. This combination results in a margin of approximately 6% between heat removal capability and decay heat input. The use of ANS 5.1 for calculation of heat decay under accident conditions has been validated by Nuclear Fuel Services in Reference 3. The calculation results, assumptions and input parameters of Reference 2 have been reviewed by the NED/Mechanical & Structural Design Group (reference 4) in accordance with ENC-QE-81 and are found acceptable.

  • The General Electric Letter in reference 2, also provides justification for the use of analytical values for the flow rates without specific consideration for instrument inaccuracies. The design philosophy used for the original design basis calculations, provided sufficient conservatisms to obviate the need for specifically accounting for any postulated instrument inaccuracies by maximizing/minimizing the analysis conditions.

In accordance with our proprietary agreements with General Electric, the calculation supporting Reference 2 will be available for review at Dresden after May 18, 1992. The

  • remaining issues from reference l concerning the FSAR/UFSAR discrepancies and the maximum number of heat exchanger tubes which can be plugged will be completed by September 1, 1992. If you have any questions concerning this assessment, please call me at extension 2956.

~~*~Z--

,,. Sharon C Eidge'

  • BWR Site Engineering 73~

Brian M. Viehl VA BWR Site Engineering Design* Supervisor cc: M. Strait G. Gates L. Gerner H. Massin B. Rybak c. A. Moerke G. Smith K. Ramsden

w. Groszko J. Kotowski NEDCC/Chron system supervisor

GE Nuclea1 En11gy

.;
111a. f*r,;
1,. ~*,,,*r:*r

'-~ *:*::l"t: J. ~;*,.-c SJ** .ilJSI. l:IJ 9! .' :s

'l!ay 13, 1992 CCA9203 GE-NE-668-07-0592 ORF I ElZ-00125 Mr. Brian M. V1eh1 Dresden Site Engineering Des;gn Supervisor Nuclear Engineering Department Conmonwealth* Edison Company Rural Route 1 Morris, Illinois 60450

Subject:

Design Basis for Low Pressure Coolant Injection /

Containment Coo11ng System Heit Exth1nger Sizing

Reference:

l. S. Mintz letter to J. E. Nash dated 4/6/92: Same Subject

.2 I Br1an M. V1eh1 letter to M. w. Hansen dated 4/14/92:

CCSW Fo11ow-up Actions Request for Cost Estimate LPCI Containment Cooling System Process 011gr1m 3*

729E583, Rev. 1

4. Gordon Chen's letter to Tim Allen dated May 13,1992:

Dresden 2 Containment Heat Exch1nger Mode C Heat Transfer Capabi 1i ty . *

5. NEDC-31897P-1, "Gener1c Gu1de11nes for *General Electric Boiling Water Reactor Power Uprate",

June 1991

Dear Brian:

In Reference 1, GE sunnarized the current understand1ng of the basis for the Dresden LPCI/Conta1nment Cooling System heat exchanger sizing and associated analyses based on an expedited review of readily ava11able information.

In Reference 2, Cormnonwealth Edison Company requested a cost estimate on pursuing and closing four open items relating to a CCSW reduced flow concern not fully addressed by Reference 1. Advance approval was obtained to work on question 2 and 3 tn order to provtde a ti~e1y response by May 13, 1992. These questions are repeated below w1th GE's response.

Br~an M. V1ahl *2* May 13, 1992 Commonwealth Ed;son Company CCA9203 Qtiest1on 2:

A copy of the or1g1na1 General Electric heat exchanger duty caleulat1on for modes B and c of the or1g1na1 LPCI system Process Flow D1agr111.

Response

Design Procedures extant at the time the original calculations were accol!IJ)lished did not requ1r1 th1 f11a retent;on of ealculation11 backup data to design drawings. Nevertheless, GENE 1n1t11ted 1 search of known files and interviewed engineering personnel who may have had knowledge of the origtnal calculational methods. After several days of effort, it became apparent that efforts tn this direction would prove fruitless.

In 11ght of the abov1, GENE elected to use current methodology to establish Containment Cooling HX performance under the conditions shown

  • in "Modes. B and C" in GE Drawing 729E!83 (Reference 3). Th1s effort involved construct1on of a heat exchanger performance analysis. model from known parameters 1i sted on the orig1na1 Heat Exchanger Data Shut and GE approved, fluid heat transfer coeff1c1ents. Calculations based upon this new model yielded ter~1na1 temperatures and heat transfer rates that were ;ncons1stent with the or1g1na1 Heat Exchanger Data Sheet and with Modes B ind C of the above drawing. How1v1r, the calcu1at1on results 11 showed that the original methods did in fact reduce the value of u* (Overal 1 Heat Transfer Coeff1c1ent) from Modi B cond1t1ons to Mode C, in order to reflect the large scale reductions in flow, both on the tube and shell side of the heat exchanger. The results of these new ea1eulat1ons for Mode C flow conditions are shown in Reference 4 attached. '

To resolve the above incons;steney between new calculations and

  • ' '. I

~ . th.e .. cond1.t 1onsL .shown on the Heat: Ex~hanger D1t1 Sheet, GENE contacted the or1g1nal heat 1xeh1ng1r manufacturer (now Senior Eng1neer1ng).

These dtscussions ltd to the manufacturer's search of his records, both tn Los Angeles, C&ltforn1a and 1n Berlin, W1scons1n. Again, these data recovery efforts were not successful. The manufacturer agreed to run a set of ~alcul1ttons, ustng its current proprietary codes. in an attempt to dupl1cate* the heat exchange rates and terminal conditions shown on the Heat Exchanger Data Sheet. These efforts also showed 1nconststency with the original data sheets, however, the results .closely agreed with those produced by GENE methods. *

  • In that the original heat exchanger des1gn methodology was not recoverable and cu~rent calculations indicated an 1pprox1m1te 91 difference 1n heat transfer at the or1g1na1 design cond1t1ons. GENE and

Brian M. V1eh1 Hay 13, 1992 Commonwealth Ed1son*Company CCA9203

-CECO elected to confirm the current capab;11ty of the original Heat Exchang1r. Th1 s- effort was undertaken to concl us1 vely show that OBA maximum suppr1s11on pool temperature ~argtns (for LPCI pum~ NPSH purposes) art conservative. Th* *'fort ut111ztd the results of the more rea11st1c decay heat calculation m1thodology_(ANS S.1) to compare.with the results of tht older (May-Witt) ~*thod used tn the original design of the Dresden untts. Tht use of ANS 5.1 decay heat 1n eval~attons of the long-term containment pressure and temperature response with power uprate was presented to the NRC in a submittal (R1fer1nc1 5) which has been approved by the NRC. ihh comparhon shows that 1n th1 post-OBA t1me frame of raax1mum suppression pool temper1tur1, d1cay heat 1s at least 151 less than the value used in tht original dts1gn

{S4.5E6 BTU/hr). Hance, given the currently calculated heat exchanger heat transfer rate of 77£6 BTU/hr (at Mode C conditions) for the original heat exchanger, and a more r111tsttc decay heat rate generation, the maximum su.pprassion pool temperature, post OBA, would be less than the 180F maximum temperature spec1f11d in Mode C of GE Dr1wtng 729ES83 (Reference 3). In addtt1on, since Mode C specifies the use of one Containment Coo11ng Heat Exchanger, on1 Containment Cooling Service Water pump, and one LPCJ pump, it is concluded that this equipment comb1nat1on is adequate for post-OBA suppr1ss1on pool containment cooling purposes.

Question l:

Address how Instrument Inaccuracy was taken tnta account with the original des1gn a~alysis.

Response

Instrument inaccuracy was nil taken tnto account 1n the or1gtna1 design anl1ysts of the long term pressure/temperature response of the primary contafnment. The ana1yses made in that time ptr1od dtd not exp11c1t1y consider Instrument accuracy although in1tt&1 cond1ttons used 1n analyses ware nonn111y maximized or min1m1ztd to insure conservat1v*

results.

Typically when instrument inaccuracy 1s taken into account and the safety analysts is revised 1t has been shown that margins to design and.

safety l1m1ts sttll extst. The evaluation performed *1n response to question Z above 1s typical. Using current methodology there 1s an estimated 61 marg1n, even taking 1nto account the lower flow rates of the LPCI/Conta1nment Cooling System Heat Exchangers. Computer analysts for the Emergency Cort Cooling System also show stailar results. 1.1.

SAFER and GESTR have much more realtst1c results and lower values than the or1g1na1 safety analysis perfonned for Dr1sd1n.

Br11n M. Viehl M&y 13, 1992 Convnonwealth Edison Company CCAH03 s.~Aart~

Prepared by: Ver; f11d by:

z Plant Ana1ys1s Services C. C~len Plant Systems Das1in Ver1f1ed by:~,~&..

  • an Heat E~tr and Pump Design Reviewed ~C~t..4-j; E. Torbeck Plant Analysis Services Reviewed by: ~~e.a*/
  • owa Pl1nt Systems oes1gn

Attachment:

Gordon Chen's letter to Tim Allen dated May 13,1992:

Dresden 2 Containment Heat Exchanger Modi C Heat Transfer Capab111ty

  • cc: S. S. Dua M. G. McBride M. w. Hansen W. G. Myers G. L. Hayes R. S. V1j

GE Nuclear En*r<JY ce: R. L. Hughe*

Date& May 13, 1992 TOI Tim Allen From: Gordon Chen Heat Exchan;er ' Pwnp Deai;n s~bject: Dresden 2 eontaimnent Heat Exchanqer Mode c Heat Tran*fer

. Capability In response to Dresden site request, the heat tranater-eapability ot Dresden 2 Containment Heat Exchanger under Mode e operation has been calculated. The desiqn data, analyzed condition* and the calculated results are qiven belows *

  • DESIGN DATA Th* followinq heat exchanger de1i9l\ data obtained from PERFEX Heat Exchanger Specification Sheet were used in pertorming the calculations.

~*

1) Effective TUbe Surface Area
  • 9,897.3 tt (not including 6' excess tubes) *
2) Tube Side Plow Velocity At A Flow Rate ot 3,soo,ooo lb*/hr
  • 5.1 ft/**c
3) Shell Side Flow Velocity At A Flow Rate ot 5 350000 l~*/hr
  • s.o tt/aec
4) Foul!n9"ieeiatance Inside Tubes* .002

!5) Fouling Re*istance Outside Tubes * .0005

6) T~e*
  • 70*30 CU. Ni. 3/4 BWG 18
7) Tube OD = .7!" . .

B) TUbe ID * .6!2"

9) No. ot Tube* Per Pass
  • 1184 (not including 6' excess tubes)

MODE C CONDITIONS The tollowinc; conditions, taken from the Containment Heat Exchanger process diaqram, are analyzed tor Mode c operaticnt Mode c Conditions Tube *Side Flow 3!500 GPM Shell Side Flow 5000 GPM Tuba Side Inlet Water Temperature Shell Sida Inlet Water Temperature 180 °r

CALCUtATED RESULTS Presint*d below.are the calculated heat exchanq*r pertormance results

. tor Mode c operation:

Calculated Value*

overall coeff. ot Heat Tra~ster, U, 8tu/hr* f~-'F 189.

Heat Tranater Rate, (service), Q, Btu/hr 77.0 x 10*

T~* Sida Water Inlet Temperature, "I 95

-TW,e Side Water outlet Temperature,'F 139.29 Shell Side Water Inlet Temperature, -r 180 Shell Side water outlet Temparatura, *r 148. JS The ealeulatione tor the above heat exchanger performance are contained in ORF No. A00-0,3&9.

Prepared Bya f.[~ c?~&, j ~/tz..

Q.

  • en, Pr nc pa in9 neer Heat ~~ ' Pump Deai9n
  • Verified .Bya~ 'r ~~ ff-0.

S. F. r ce

  • Heat Exchan;er ' Pump oeaiqn

.-. v. . v May 15, 1992 Mr. Charles W. Schroeder

Subject:

Sens1t1v1ty Calculation on LPCI*Heat Exchanger Performance during Transients R1f1r1nc1s: 1. L1tt1r, T. R11ck to c. $chro1d1r dated 4/7/92 transmitting RSA-D-92-01 , "Evaluation of Reduced CCSW Flow at Dresden Station*.

2. Telephona conversation, NFS and GE on 5/12/92.
3. Microfiche of RETRAN Calculations, NFSKRB J(2575) and NFSKRC J(2&32) *

. NFS has been requested to provide an assessment of the impact of the revised LPCI haat exchanger statepoint recently calculated by General Electr1c. The statepo1nt provided by GE (Reference 2) was:

0.90 The statepo1nt calculated and used by NFS in Reference 1 for evaluation of the pool haatup transient was:

. The r11at1valy close heat exchanger performance values are pr1nc1pally a result of very conservative selection of heat exchanger flow rates used 1n the

  • or1g1na1 NFS calculat1on. The 11mit1ng transient from Reference 1 was re-performed utilizing the heat exchanger performance values provided. The data1led output 1s contained in Reference 3, appended to the NFS file version of Reference 1. As anticipated the impact was negligible, as demonstrated in the following table.

May 15, 1992 S1nca the revised calculation results 1n less than 0.1 F change in.the predicted bulk pool temperature, the conclusions of th1 or1g1na1 calculation rema;n v111d. The plant can effectively mitigate the limiting pool heatup

  • transient with a single loop of pool cooling with 1 CCSW and 1 LPCI pump. The local pool temperature limits remain sattsfted, and condensate stab111ty of the T-Qu1nchers 11 assured.
  • The site_ ang1neer1ng personnel had add1t1onal questions concerning the use of the ANS !.1 1979 decay heat curves. General Electric indicated that the May-Witt curves were originally utilized far the long term post-LOCA suppression pool heatup calculations. These curves are known to conta;n

.,signif;cant margin relative to the 1979 ANS curves. This is 1 result of th1 original uncertainty in decay heat prediction (+/- 201) applied in.the 1973 curves, versus the reduced uncertainty for the 1979 curves(+/- 21). The attached figure graphically illustrates the reduction in uncertainty associated with the later standard. The later decay heat standards have bean applied to a variety of recent analyses and ara generally acceptable to the HRC. The exception 1s *1n the area of 10CFR 50.46 LOCA analysis, whera 1 120%

decay heat mu1t1plier 1s proscribed. except for best estimate LOCA appl1cat1ons where nominal values are utilized with uncertainties addressed in statistical combination methods.

If you have any questions regarding th1s matter, please contact K. B.

Ramsden (x-3851) of my staff. . *

!
.:::-;, ~*Irv Nuclear Fuel Services Manager

'~

TAR:KBR:pc Attachment cc: NFM-CF G. P. Wagner B. M. V1ehl S. Eldridge_

K. Kovar S. Powers

)

r

)

  • .. ... ... 111 i91
  • \

1 I

Ml t

...... < ... - '91 + ~* I OIW-W+S!..-..,.19

--&LON-

,..,..18 tJW.I() l""'JIV Ult- * ...

t. Jr MAfiR

.oz ....... -.......n t

  • pn11iwt

......N-

--* m-n .-.e.1c1 *

~.-----------------------~~~~~-------------~----------i-a

...=

~*

I

May 15, 1992 In reply refer to CHRON# 185905 To. B. Vlehl ENC- Site Engineering

Subject:

Dresden Unit 2 LPCI Heat Exchanger Mode C Heat Transfer C&pablllty

Reference:

Letter dated 5/13192 prepared by C.C. Allen (General Electric) to B.M. Vlehl The Nuclear Engineering Department (NED-M&S Group) was requested to evaluate the Heat Exchanger ~uty calculation for the LPCI Heat Exchanger Mode C (1 CCSW and L LPCI Pump). General Electric did not have a calculation available for review. However, their results were summarized In Reference 4 of the above reference letter. In

.accordance with Nuclear Engineering Department procedure ENC-QE-81, a preliminary

  • ***calculation was prepared to adequately assess the LPCI heat exchanger capacity results supplied by General Electric.
  • The calculation method employed used the LMTD (Log Mean Temperature.Difference) approach because inlet shell, tube temperatures and flows ,and U (heat transfer
  • coefficient} was specified. The preliminary results yielded a value of 78.7 MBTU/HR heat capacity or 2.15%.higher as compared with the GE calculated value of 77.0 MBTU/HR. NED feels that this slight difference is negligible and that based on the design inputs their calculated value of 77 .0 MBTU/HR is acceptable.
  • This confirmatory calculation will be formally transmitted in accordance with Nuclear Engineering Department procedure ENC-OE-51.d.

If there are any further questions please contact me on extension 7666, Nuclear Engineering Department.

Prepared by:~~~~*=:=*=:::.t..~C?.~.~~~!:j~::!;l~ Date: s-J/

I I r-/9.,_

S.P. Powers Mechanical & Structural Engineer Approved ~y: £/R~. Date: ,,fb6/'-;a.

' J P.R. Donavin Mechanical & Structural Design Supervisor

  • . cc: H.L Massln ( BWR Systems Design)

S. Eldridge ( Dresden Site Engineering)

GE NUCLEAR ENERGY San Jose, CA

_ _ _ _ _ _ _ _ _ _ _ _ _ _ PLANT PERFORMANCE ANALYSIS PROJECTS February 17, 1993 Ms. Sharon Eldridge Conunonwealth Edison Company Dresden Nuclear Power Station Rural Route No. 1 Morris, IL 60450

Subject:

Reconunended Use of ANS 5.1 in Dresden 2/3 Containment Long-Term Post Accident Analyses.

References:

1) NEDC-31897P-l, *Generic Guidelines for General Electric Boiling Water Reactor Power Uprate,* June 1991.
2) Letter, W. T. Russell (NRC) to P. W. Marriott,* Staff Position Concerning General Electric Boiling Water Reactor Power Uprate Program (TAC No. 79384),* September 30, 1991.

Dear Ms. Eldridge Attachment 1 is a technical description of the use of the GE Decay Heat Model in the subject containment analysis (GE report GENE-770-26-1092). This model is founded on the industry accepted decay heat curve of ANSI/ANS 5.1- 1979 (ANS 5.1). We believe that the attached technical description of the decay heat model and conservatisms in the subject containment analysis relative to the original analysis justifies its application. These conservatisms are confirmed by a comparison of the results of Case 4 of GENE-770-26-1092 to the original

.* '*'fSAR results~ "Case *4 assumed the same pump flow rates used in the original FSAR analysis for the 1 LPCI/Containment Cooling Spray pump /1 CCSW pump configuration

  • The results of Case 4 matched the original analysis value of the peak suppression pool temperature of lSO*f. This agreement was observed even though the net effect of an approximate 15% reduction in decay heat and a 9%.reduction in the Dresden LPCl/Containment Cooling System heat exchanger performance will by themselves result in a lower peak suppression pool temperature. Known conservatisms in the current analysis relative to the FSAR analysis include a 5*f higher initial suppression pool temperature and addition of all feedwater which can contribute to heating the suppression pool.

Information forwarded to GE from Co11111onwealth Edison Company (CECO), (NRC Letter EA No.93-019) indicates that CECO is being notified of an apparent

violation due to the use of the ANS 5.1 decay heat model in the subject anilysis. The alleged violation (E.3) appears to be based upon the NRC not having approved the decay heat model nor the computer methodology in the

,analyses. GE Nuclear Energy (GENE) has used the same or similar methodology on several occasions in reports generated for other utilities. On at least one of these occas i ans the_ report was reviewed and approved by the NRC.

GENE has had a number of technical discussions with the NRC (NRR) Staff concerning GENE's use of ANS 5.1 and other computer methodology used in analyses similar to that performed for Dresden. GENE and two domestic utilities met with the NRC technical staff in Rockville Maryland in December of 1991 to discuss the GENE containment analysis approach, including computer codes, the use of ANS/ANSI 5.1-1979 (ANS 5.1) decay heat and other assumptions.

Although the GE generic power uprate program review was the motivation for the meeting non-power uprate applications were also discussed. The NRC agreed that the GE codes and input assumptions would be generally acceptable for power uprate and other applications, providing that sufficient justification was provided to assure the overall conservatism of the conclusions. It was our understanding that a detailed explanation of the current analysis was to be available for review by the NRC if requested.

The power uprate generic licensing topical report (Reference 1) specifically identified the ANS 5.1 decay heat assumption. That report was approved by the NRC (Reference 2). Several subsequent meetings and phone calls between GE and the NRC have provided GE assurance that the use of GENE containment analysis methods, including computer codes and input assumptions such as ANS 5.1 decay heat, would be acceptable to the NRC technical staff for power uprate or

  • non-power uprate analyses provided the overall conservatism of results was demonstrated.

At no time has the NRC indicated to GENE that prior NRC approval for specific

  • appl icat.ions was required. Hence, in using this methodology for the Dresden
  • 2/3 analyses we were acting in the belief that we were within known NRC mandates. To date GENE is not aware of any notification by the NRC that the

.use of ANS 5.1 or GENE computer methods used in the subject analyses are i'nvalid.

Evidence of verification for Attachment 1 is contained in the GENE design record file; ORF 123-00685.

,.//{~

S. Mintz Plant Performance Analysis Projects (408)-925-1791 M/C 469 cc J. E. Torbeck (GE)

J. E. Nash (GE)

T. L. Chapman (GE)

C. T. Young (GE)

R. T. Hill (GE)

D.. J.* Robare (GE)

DRF-123-00685 .

Attachment 1 - Use of ANS 5.1 Decay Heat Model in Containment Analysis

  • The General* Electric analysis of the Dresden long-term containment pressure response for the limiting LOCA event which is described in Reference 1 use~_a decay heat model based on the ANS/ANSI 5.1 - 1979 (Reference 2) decay heat model. This decay heat model was chosen since it is the current model approved by the American National Standards Institute and both the industry and the NRC have acknowledged that ANS 5.1 standard is a more realistic model than previous models. Although the ANS 5.1 decay heat model is more realistic than that used for the original Dresden FSAR analysis, conservative results for the Dresden containment heatup analysis were assured by use of conservatism in the input and a~alysis assumptions. The following discussion provides more detail on the use of the ANS.5.1 decay heat model.
1. Background and Description of ANS 5.1 Decay Heat Model The American National Standards Institute approved and the ANS published the new revised ANS 5.1 standard "American National Standard for Decay Heat Power in Light Water Reactors*, (Reference 2). *This standard includes significant technical advantages over the older standards in that it deals in great detail with the physics involved and has a significant data base. To use these technical advantages, the General Electric Company developed a generic decay power curve-based on the ANS 5.1 standard to provide a more accurate assessment-of the decay-heat during a LOCA. General Electric used the decay heat correlation developed from the ANS 5.1 decay heat model in ECCS analysis to determine expected ~esults based on realistic methods (Reference 3). These realistic methods were used to establish real safety margins. This methodology is described in Reference 3 which has been reviewed and approved by the NRC.

The generic decay power curve developed by GE, which is based on the ANS 5.1 decay heat model, is described in detail in Appendix B of Reference 3. This decay power curve not only includes fission product decay heat but also includes other major contributors to post-LOCA heat generation. The other cootributors incl_ude decay of actinides, decay of activation products, and fission heat due to delayed neutrons. Additional details from Appendix B of Reference 3 are provided below on the major contributors to the decay heat.

_Decay Heat from Fission Products The fission product decay heat is based on the ANS 5.1 decay heat model.

In the ANS 5.1 standard, values are provided for decay heat power from fission products from fissioning of the major fissionable nuclides present in Light W~ter Reactors {LWRs}, i.e., u235, Pu239 thermal and u238 fast.

A method is also prescribed for evaluating the total fission product decay heat power from the data ghen for these specific nuclides. There are

_fissions produced from other nuclides; however, it is assumed, as directed in ANS 5.1, that all nuclides other than Pu239 and u238 have the same

~ission product characteri~tics as u235.

Decay of Actinides Actinides are the heavy elements produced from neutron capture in uranium and plutonium isotopes. The actinide concentration following shutdown is calculated assuming that at shutdown the actinide concentration of each actinide is at its equilibrium value. This equilibrium value is determined assuming no captures in the radioactive nuclide. These assumptions conservatively result in a higher actinide concentration.

Decay of Activated Structures The principal structural material in the reactor is zirconium and is therefore the principal source of decay heat from activated structures.

Other materials in and around the reactor are the steel in the control blade, the shroud, and the bottom support plate. However, the contribution to the total decay heat from the activation of control blades or materials outside the core is negligible and is therefore not included.

Fission Heat Induced by Delayed Neutrons When a reactor shuts down, the power level does not drop to zero immediately. Instead it decays away with time. due to fissions caused by delayed neutrons. The contribution from delayed neutrons is conservatively determined for LOCAs since* it was calculated assuming a slow blowdown rate which t:_esults in a smaller void negative reactivity feedback and hence a slower decrease in the neutron flux following the break ..

Based on the description given above it was determined that the ANS 5.1 decay heat model provides a more accurate representation of the decay heat during a

. LOCA than. previous models. Additionally, it is noted that the ANS 5.1 standard was conservatively applied in developing the GE decay heat power curve.

Therefore there are conservatism retained in the resulting decay heat model.

2. AppHcation of ANS 5.1 Although a more realistic decay heat was used in the Reference 1 analysis, the results are still conservative. The limiting values of key input parameters were used for the Reference 1 analysis. Considering the unlikelihood that each

.parameter w111. be at its limiting value at the time of the LOCA this results in very conservative results. It was therefore decided to use a more realistic decay heat model than that used in the original FSAR analysis since the remaining conservatism in the analysis and input assumptions ensure that the calculatad pool temperature response is conservative. The conservatism in the Reference 1 analysis are reflected in the following input assumptions:

. 1)- The reactor . . is assumed operating at 102% of rated therma 1 power.

2) Feedwater flow into the vessel continues until all the feedwater which will increase the suppression pool temperature is injected into the vessel. In addition, a conservative calculation of the energy in the feedwater piping is added to the RPV/containment system.
3) The initial suppression pool volume is at the minimum Technical Specification (T/S) limit to maximize the calculated suppression pool temperature.
4) The initial suppression pool temperature is at the maximum T/S value to

. maximize the calculated suppression pool temperature.

5) Service water temperature is assumed to be at the maximum expected value.
6) Passive heat sinks in the containment and in the suppression pool are neglected.
7) All.ECCS and RHR system pumps have lOOS of their horsepower rating converted to a pump heat input which is added to the RPV liquid or suppression pool water.
8) Heat transfer from the primary containment to the reactor building is conservatively neglected.
10) Heat exchanger perfonnance was calculated based on design fouling factors and assumed 61 plugging of tubes.
11) As requested by CECo, uncertainties in the flow rate to the LPCl/Containment Cooling System Heat Exchanger were considered and lower bounds on the flow rates were*used in detennining heat exchanger perfonnance.
3. Other General Electric Applications of ANS 5.1 Decay Heat for Containment Analysis General Electric has used the ANS 5.1 decay heat model in several evaluations and analyses of the containment pressure and temperature response for other BWR plants. Although a more realistic decay heat was used in these analyses, in all cases conservatism were retained in other key input parameters to ensure conservative results.
4. Summary General Electric used the ANS 5.1 decay heat model in the Reference 1 analysis*

and in other similar analyses because it is the current standard and is

  • 'Considered .a superior model. Although the ANS 5.1 decay heat model h more realistic than the model used for the original Dresden FSAR analysis the calculated suppression pool temperatures are still conservative due to the conservatism retained in the inputs and analysis methods.
5. References
1) GENE-770-26-1092, *oresden Nuclear Power Station, Units 2 and 3, LPCl/Containment Cooling System Evaluation,* November 1992.
2) *oecay Heat Power in Light Water Reactors,* ANSI/ANS 5.1 - 1979, Approved by American National Standards Institute, August 29, 1979.
3) General Electric Co., *rhe GESTR-LOCA and SAFER Models for the Evaluation*

of the Loss-of-Coolant Accident,* NED0-23785-1-A, Volume III, October 1984.

CONTROL OF EFFLUENT RELEASES As .discussed during the Enforcement conference on February 22, 1993, protection against effluent releases in excess of 10 CFR 100 limits is ensured by 1) providing a leak tight heat exchanger and 2) providing a positive differential pressure across the LPCI heat exchanger (CCSW outlet pressure> LPCI outlet pressure).

The FSAR/UFSAR includes a description of the design criteria requirement of a positive

  • differential pressure across the LPCI heat exchanger. This dP is ensured by inaintaining the operating margin of 20 psid by throttling of the CCSW outlet valve or by increasing LPCI flow (as described in Section 6.2.4.2.4 #3). The original 20 psid value is based on a nominal 5 psid to ensure a positive differential with an additional 15 psid to account for instrument and control inaccuracies( see Attachment A). The "Dresden Station Unit 2 and 3 Setpoint Error Analysis",

dated December 28,

  • 1992 has determined that the dP instruments associated with monitoring this dP have a maximum error of 3.26 psid. Therefore to achieve the design criteria requirement a dP of 8.26 is ad~quate.
  • The Technical Specification Amendment transmittal dated November 4, 1974 provided an analytical basis for reducing the CCSW discharge pressure requirement to 180 psig. This calculation assumed a LPCI discharge pressure of 125 psig, a 20 psid across the heat exchanger, 10 psig for margin, and a 25 psig pressure drop in the CCSW system from pump discharge to outlet of the heat exchanger. The 125 psig LPCI pump discharge pressure represents a pump flow of 5200 gpm.
  • The value obtained from using the nominal LPCI pump flow value of 5000 gpm ( resulting in a discharge pressure at the outlet of the heat exchanger of 141.1 psig), 10 psig* for the suppression pool over pressure post LOCA and 14.0 psig calculated for CCSW line losses (at nominal flow of 3500 gpm) to the outlet of the heat exchanger results in an available dP across the heat exchanger of 14.9 psid which is greater than the 8.26 psig required and therefore demonstrates adequate
'protection against effluent discharge.
  • Kamorandua ot Conver1ation erson - Company Company Project Project No.

Dresden 9188-10 Subject Discussed 20 paid Differential Pressure Setpoint Between ccsw and LPCI Across LPCI HX Summary of Discussion, Decisions and Commitments

1. Mr. Goebbert contacted Mr. Rob Davison at GE (408-925-4587).

Mr. Davison is the DBD writer for RHR. He provided the following information.

2. The design basis for the pressure differential is that the ccsw i, always be positive .in respect to LPCI,
3. The 20 paid value was set based on consideration of instrument and control error plus 5 psi.

4, This design bases is similar for other GE plants. It will be documented in a GE internal correspondence which will be provided t9 CECo'with the LPCI/RHR DBD documents.

r cc J. Dawn B. Barth

s. Eldridge T. Behringer G. Lupia R. J. Goebbert T. R. li:isenbart L. Schwarz
o. Bianchini File si~nature.......,.. .
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