NLS2008067, Response to Nuclear Regulatory Commission Request for Additional Information Reactor Equipment Cooling System

Response to Nuclear Regulatory Commission Request for Additional Information Reactor Equipment Cooling System

ML082730526
Person / Time
Site:	Cooper
Issue date:	09/11/2008
From:	Minahan S Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2008067, TAC MD8374
Download: ML082730526 (56)

Text

N Nebraska Public Power District "Always there when you need us" 50.90 NLS2008067 September 11, 2008 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Response to Nuclear Regulatory Commission Request for Additional Information Re: Reactor Equipment Cooling System (TAC No. MD8374)

Cooper Nuclear Station, Docket No. 50-298, DPR-46

References:

1.

Letter from Carl F. Lyon, U. S. Nuclear Regulatory Commission, to Stewart B. Minahan, Nebraska Public Power District, dated July 7, 2008, "Cooper Nuclear Station - Request for Additional Information Re: Reactor Equipment Cooling System (TAC No. MD8374)"

2.

Letter from Stewart B. Minahan, Nebraska Public Power District, to the U.S. Nuclear Regulatory Commission, dated March 24, 2008, "License Amendment Request to Revise Technical Specification 3.7.3, Reactor Equipment Cooling System"

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District to submit a response to a request for additional information (RAI) (Reference 1) from the Nuclear Regulatory Commission (NRC). The RAI requested information in support of the NRC review of a license amendment request (Reference 2) to revise Cooper Nuclear Station Technical Specification 3.7.3, Reactor Equipment Cooling System.

The response to the specific questions in the RAI is provided in Attachment 1. Excerpts from applicable calculations are provided in Enclosures 1 and 2. No regulatory commitments are made in this submittal.

The information submitted by this response to the RAI does not change the conclusions or the basis of the no significant hazards consideration evaluation provided with Reference 2.

COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5271 K

ý-K www.nppd corn

NLS2008067 Page 2 of 2 If you have any questions concerning this matter, please contact David W. Van Der Kamp, Licensing Manager, at (402) 825-2904.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on

/Y' 7')o- ý-'

(date)

Sincerely, Stewart B. Minahan

,WILMA M. WERNER Stewart B.MinahanMY COMMISSION EXPIRES Vice President - Nuclear and E...R:

%'/'.

k'*{*," October 26, 2010 Chief Nuclear Officer MY

/rr Attachment Enclosures cc:

Regional Administrator w/ attachment, w/o enclosures USNRC - Region IV Cooper Project Manager w/ attachment and enclosures USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment, w/o enclosures USNRC - CNS Nebraska Health and Human Services w/ attachment, w/o enclosures Department of Regulation and Licensure NPG Distribution w attachment, w/o enclosures CNS Records w/ attachment and enclosures

NLS2008067 Page 1 of 5 Response to Nuclear Regulatory Commission Request for Additional Information Re: Reactor Equipment Cooling System (TAC No. MD8374)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 NRC Question #1 Please provide numerical values of any changes in the normal environment, such as temperature, humidity, radiation due to this LAR.

Response

This license amendment request (LAR) proposes a change in how a beyond-design-basis loss-of-coolant accident (LOCA) would be mitigated. The LAR proposes no changes to normal operation of Cooper Nuclear Station (CNS). Therefore, this LAR will result in no changes to normal environment parameters, including temperature, humidity, or radiation.

NRC Question #2 Please provide numerical values of any changes in environmental conditions, such as temperature, pressure, humidity, radiation etc. during LOCA including profile comparison for temperature and pressure. Please confirm your conclusion that these profiles are bounding in meeting the requirements of Title 10 of the Code of Federal Regulations Section 50.49.

Response

The current Equipment Qualification (EQ) pressure and temperature profiles for LOCA remain bounding for this LAR. No radiation calculations have been performed. However, no changes in radiation would be expected since the LAR involves a change in the Emergency Core Cooling System (ECCS) pump and room cooling which has no effect on radiation. Pressure, temperature, and humidity are inherent elements of the Reactor Building (RB) thermal environments.

Comparisons have focused on the 180-day Arrhenius equivalent aging temperature. The changes due to this LAR are negligibly small, as reflected in Table 2 presented later. Based on this, pressure and humidity changes would also be negligibly small, and have not been plotted or integrated.

The zones that will be most affected by the LAR are the four rooms in the RB where the Residual Heat Removal (RHR) and Core Spray (CS) pumps are located (referred to as the Quad Rooms), plus the High Pressure Coolant Injection (HPCI) pump room.

The Nebraska Public Power District (NPPD) confirms that the current EQ profiles for CNS are bounding in meeting the 1 OCFR5 0.49 requirements.

NLS2008067 Page 2 of 5

Background

The post-LOCA essential function of the Reactor Equipment Cooling (REC) System is to cool the ECCS pump rooms and equipment. (The REC System was called the Reactor Building Closed Cooling Water System in the original CNS design.) The critical (i.e., essential) heat loads are as follows:

The RHR pumps; The RHR pump bearing oil coolers;

• The RHR pump room fan coil units (FCUs);

" The CS room FCUs; and

• The HPCI room FCUs.

The above cooling functions are provided by critical, closed loops of Class IS, seismically designed piping that supply clean, demineralized water to all of the heat exchangers in the RB Quad and HPCI Rooms. This heat is ultimately rejected to the Missouri River through one or both of the REC heat exchangers. These heat exchangers are essential equipment that is cooled by the Service Water (SW) System.

The original post-LOCA mission time of the REC System was 30 days. This mission time was changed to seven days in Amendment No. 185. Calculations prepared in support of that amendment assumed that the REC system was lost after seven days, and at that time, SW backup cooling was assumed to be aligned to the REC critical piping loops after a 20-minute delay to make the alignment. (SW backup cooling was installed in July of 1973, in response to FSAR Question 10.5b from the U.S. Atomic Energy Commission during their review of the CNS operating license application.)

The current LAR changes the post-accident initiation time of SW from seven days following a LOCA to as early as time zero following a LOCA, with a one-hour time delay now allotted to align SW backup cooling instead of a 20-minute delay. Thus, the REC System is effectively relieved of its essential (i.e., safety-related) post-accident functions, these being replaced by the SW System.

With SW backup cooling, Missouri River water would be piped directly into the REC critical piping loops for cooling of the ECCS pumps and the Quad Rooms and HPCI Room heat exchangers.

The need to use SW backup cooling presumes that both divisions of the REC System have been lost. When evaluating essential cooling to the ECCS Quad Rooms, additional equipment failures have also been assumed, namely, loss of individual room coolers. (These postulated events that could lead to the use of backup cooling are beyond-design-basis.)

Quad Room Thermal Environments - Calculation NEDC 00-095E Calculation NEDC 00-095E, Revision 2, "CNS Reactor Building Post-LOCA Heating Analysis,"

was performed to determine RB heating for various loss-of-REC conditions. This calculation develops the RB thermal environments for EQ analysis. This analysis has been performed using the GOTHIC Version 7.0 computer code and models the entire RB, consisting of 50 different thermal volumes. Nine simulations have been performed in support of the LAR, with focus on the four

NLS2008067 Page 3 of 5 ECCS Quad Rooms and the HPCI Room. Each simulation was done for a six-month, post-accident time period, with plots generated of the transient thermal response. These simulations are summarized in the following Table 1.

Table 1 Simulation Cases for NEDC 00-095E, Revision 2 Case Case Definition Assumptions Switchover to Service Water (hours)

P1 Base Case, Switchover to SW Backup at 20 No FCU Failures 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> P2A Loss-of-FCU in South West Quad Fail SW FCU at 20 (RHR pumps B and D) t=0 P2B Loss-of-FCU in North West Quad Fail NW FCU at 20 (RHR pumps A and C) t=0 P3 Loss-of-FCU in HPCI Room Fail HPCI FCU at 20 t=O P4A Loss-of-FCU in South East Quad Fail SE FCU at 20 (Core Spray pump B) with flow from CRD t=0 P4A-1 Same as P4A, but without flow from CRD Fail SE FCU at 20 t=0 P4B Loss-of-FCU in North East Quad Fail NE FCU at 20 (Core Spray pump A; RCIC) t=0 P5 Failure of Drywell FCUs to Isolate; 1 15°F No FCU Failures No SW Backup REC Inlet Temperature to ECCS P6 Same as Case P1, with REC Cooling for No FCU Failures No SW Backup I Event Duration Case P1 of Table 1, the base case, does not assume any FCU failures. Cases P2A through P4B of Table 1 address the five thermal volumes in which the ECCS pumps and room coolers are located, with the assumed FCU failures identified. For these cases the REC System is assumed to fail 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after accident occurrence. Scoping analysis was performed in this regard, with various failure times investigated, from time zero up to seven days. Twenty hours was selected as an appropriate, limiting case. After 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> the rooms and equipment would have had substantial soak time, and the suppression pool temperature, which affects the process piping temperatures in the quads and therefore drives this large heat load, is still about 200'F and near maximum. After seven days, for example, the suppression pool temperatures and the piping heat loads are appreciably less. Case P5 is a sensitivity calculation with no FCU failures assumed, and the REC cooling water assumed to be at 11 5°F instead of the normal licensing basis maximum of 1 00°F, due to assumed inability to isolate the non-safety, drywell coolers. (Technical Specification Surveillance Requirement 3.7.3.2 limits the temperature of REC to 100°F.) Case P6, the more realistic calculation and the expected case, assumes successful REC cooling for the event duration and no need for SW backup cooling.

While the ECCS rooms are the focus of this analysis, the GOTHIC analysis determines the temperature of all RB volumes. However, since REC post-accident cooling is only active in the

NLS2008067 Page 4 of 5 quads and in the HPCI Room, the thermal effects in the other volumes of losing REC are negligible.

The EQ analysis does account for the other volumes, however.

Similar calculations were performed by CNS in support of Amendment No. 185, with REC assumed to fail at seven days, and with 20 minutes assumed to make the SW backup cooling alignment. The change from 20 minutes to one hour for SW system alignment is small, and comparing 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> to seven days, the main items of change are the suppression pool temperature and process piping heat loads. Over a simulation time period of 180 days, these are all small effects, as verified by the EQ analysis.

EQ Analysis - Calculation NEDC 03-027 Calculation NEDC 03-027, Revision 2, "Assessment of Post-LOCA Heatup Temperature Profiles for Reactor Building," examines all of the RB thermal environments developed in calculation NEDC 00-095E and calculates a six-month, Arrhenius equivalent aging temperature for each volume or group of volumes. The calculation also assesses the impact of the new environments on the existing EQ profiles.

The following Table 2 provides a comparison of the Arrhenius Temperature in the Quad Rooms from the previous Revision 1 of calculation NEDC 03-027 to the results from the current Revision

2. In Revision 1 REC failure was assumed to occur after seven days. Table 2 reflects that the differences in the 180-day Arrhenius temperatures are negligible.

Table 2 EQ Temperature Comparisons in ECCS Rooms Room Previous Current EQ PBD Arrhenius Arrhenius Temp.

Arrhenius Temp.

Temp. ('F)

(°F)/Case (Rev. 1)

(°F)/Case (Rev. 2)

North East Quad 133.82 / (4B) 133.48 / (P4B) 156 South East Quad 147.62 / (4A) 147.76 / (P4A) 156 South West Quad 157.06 / (2A) 157.20 / (P2A) 158 North West Quad 172.39 / (2B) 172.42 / (P2B) 173 HPCIRoom 121.20 / (2A) 121.28 / (P2A) 122 180-day The maximum calculated (worst-case) Arrhenius temperature for the nine cases reflected in the above Table 1 is listed on pages 9 and 10 of calculation NEDC 03-027, Revision 2 (provided as ). This worst-case temperature is compared to the Arrhenius temperature from the current EQ Program Basis Document (PBD). NEDC 03-027 includes plots of temperature versus time with a comparison of the Arrhenius temperature to the temperature for each volume.

The calculation concludes that the Arrhenius equivalent aging temperatures currently used in the EQ PBD remain unchanged, and that they are bounding for all equipment qualification documentation used in the EQ Program to demonstrate 1 OCFR50.49 compliance. The calculation.

NLS2008067 Page 5 of 5 also notes that some peak temperatures have increased in conjunction with the current LAR, but it is concluded that these are insignificant and that no qualification changes are needed.

NRC Question #3 Please identify all EQ equipment that will be affected by this LAR. Confirm that no EQ equipment needs replacement and no new equipment is added to the EQ program related to this LAR.

Response

The essential equipment located in the RB is potentially affected by this LAR. The equipment located in the four ECCS Quad Rooms and the HPCI Room, at elevations 859' and 881'in the RB, has the greatest potential for being impacted by this LAR.

Because the current EQ Profiles for the RB, including the four ECCS Quad Rooms and the HPCI Room, remain bounding for the cases analyzed in support of the subject LAR, no equipment is impacted. As a result no equipment is being replaced, and no new equipment is being added to the EQ program.

NRC Request for Support Documentation Please provide detailed EQ evaluations, justifications, calculations, and comparisons for the above items 1, 2, and 3

Response

Excerpts from calculations NEDC 00-095E, Revision 2, "CNS Reactor Building Post-LOCA Heating Analysis" and NEDC 03-027, Revision 2, "Assessment of Post-LOCA Heatup Temperature Profiles for Reactor Building" are provided as Enclosures 1 and 2, respectively. These excerpts include the purpose, assumptions, methodology, results, and conclusions of these calculations.

ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@

0.ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@

Correspondence Number: NLS2008067 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None

.1-i 4

4-4 4-I 4

I 4

I 4

L

£ PROCEDURE 0.42 REVISION 22 PAGE 18 OF 25

NLS2008067 Enclosure I J

NEDC 00-095E, Revision 2, "CNS Reactor Building Post-LOCA Heating Analysis" (Excerpted pages 1 through 34)

Cooper Nuclear Station, Docket No. 50-298, DPR-46

Page:

1 of 34

Title:

CNS Reactor Building Post-LOCA Calculation Number: NEDC 00-095E Heating Analysis CED/EE Number:

N/A System/Structure: SC/Reactor Building Setpoint Change/Part Eval Number:

N/A Component: EO Equipment in Reactor Bldg.

Discipline: Mechanical Classification: [ X ] Essential; [ ] Non-Essential SQAP Requirements Met? [ X ] Yes; [

N/A Proprietary Information Included? [ ] Yes; [ X ] No

==

Description:==

Revision 2 is performed in support of the REC Margin Licensing Amendment [Draft NLS2007004]. The allowable time for switchover from REC Cooling to Service Water Backup Cooling is extended from 20-minutes to 1-hour. Whereas previous versions have generally assumed that Loss-of-REC Cooling occurs after 7-days or else at time zero, in this revision it is assumed that Loss-of-REC cooling can occur at any time. Worst-case is approximately 20-hours after accidentoccurrence, based on supporting analysis in NEDC 97-085, Revision 8.

==

Conclusions:==

Post-LOCA reactor building temperatures are determined using the GOTHIC Version 7.0 computer code.

GOTHIC results for post-LOCA Cases P1 through P6 are attached (nine cases total). The graphs provide a time history of the post-LOCA reactor building temperatures. The GOTHIC results are the primary input to calculation NEDC 03-027, which evaluates the resulting post-LOCA Arrhenius Equivalent temperatures for Equipment Qualification purposes.

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Edward Holcomb

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2 3

Ken Don07 K. Hilgenfeld K. Hilgenfeld S. Domikaitis Ken Done' K. Hilgenfeld K. Hilgenfeld T. Stevens 1C2 3-13-06 5-2-06 5-2-06 5-8-06 Tim McClure I

I MPR 6/23/03 G. Levy 3/3/04 N/A 4/15/04 G. Levy 6/26/03 Tim McClure 0

3 MPR 6/23/03 J. Wright 6/26/03 N/A 6/26/03 Rev.

Status Prepared By/Date Reviewed By/Date IDVed By/Date Approved By/Date Number Status Codes:

1. Active

2.

Information Only

3.

Pending

4.

Superseded or Deleted

5. OD/OE Support Only

6.

Maintenance Activity Support Only

7.

PRA/PSA 1 ý t7lý 111i X

IE

Page:

2 of '34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX DESIGN INPUT ITEM REV.

PENDING CHANGES NO.

DESIGN INPUT NO.

TO DESIGN INPUT

/1 NEDC 00-095A 4

,2 NEDC 00-095C 1

3 NEDC 00-095D 1

,4 NEDC 02-006 1

,5 Bums & Roe Drawing 2020 N58 6

NEDC 94-034B 1

,7 NEDC 94-034C 3

J8 NEDC 94-034D 2

/9 NEDC 02-008A (MPR Calculation 1

0315-0302-001) 40 NEDC 02-008B (MPR Calculation 1

0315-0302-002)

/ 11 NEDC 02-008C (MPR Calculation I

0315-0302-003)

MPR Calculation 0315-0302-004 (see Attachment A) 13 Terry Steam Turbine Co. C-888-X 1

,14 GE Drawing 729E719BC,

N02 115 GE Drawing 729E720BB /

N04 16 NEDC 92-093 9

17 CNS USAR Section V-2 loep.xxii2 18 CNS USAR Section V-2.4.4 loep.xxii2 19 CNS USAR Section V-3 loep.xxii2 20 CNS USAR Section V-3.3.4 loep.xxii2 21 CNS USAR Section X-5 loep.xxii2 Meteorological Program for the Cooper 22 Nuclear Station, January 1, 1994-N/A December 31, 1994 Meteorological Program for the Cooper 23 Nuclear Station, January 1, 1995-N/A December 31, 1995 Meteorological Program for the Cooper 24 Nuclear Station, January 1, 1996-N/A December 31, 1996

-W t rio PILAL

Page:

3 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District ITEM PENDING CHANGES NO.M DESIGN INPUT REV. NO.

TONDIG C NPUT NO.

TO DESIGN INPUT Meteorological Program for the Cooper 25 Nuclear Station, January 1, 1997-N/A December 31, 1997 Meteorological Program for the Cooper 26 Nuclear Station, January 1, 1998-N/A December 31, 1998

'27 MEL 33

/28 Procedure 2.4HVAC 13

'29 Procedure 2.3 R-2 12

,30 Procedure 6.HV.602 3

'31 USAR XIV-6-8 loep.xxii2

'32 USAR X-6.5.3 loep.xxii2

-33 USAR X-10-1 loep.xxii2

,34 GE Specification 22A2858 1

'35 GE Drawing 729E590BA 1

-36 Bums & Roe Drawing 2037 N59

-37 NEDC 94-021 4

4C1, 4C2

,38 NEDC 92-050AC 2

,39 Procedure 6.REC.201 16

/40 Procedure 6.SW.202 14 41 NEDC 97-085 8

W-4 7U,

Page:

4 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska, Public Power District AFFECTED DOCUMENTS ITEM REV.

NO.

AFFECTED DOCUMENTS NUMBER I

NEDC 03-027 1

2 EQ Program Basis Document PBD-EQ, Volume 1 4

3 EQ Program Basis Document PBD-EQ, Volume 2 3

4 NEDC 93-050 5

5 Procedure 2.4TEC 16 6

Procedure 2.2.73 45 I

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1

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1

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Page:

5 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District AFFECTED DOCUMENT SCREENING The purpose of this form is to assist the Preparer in screening new and revised design calculations to determine potential impacts to procedures and plant operations.

SCREENING OUESTIONS YES NO UNCERTAIN

1. Does it involve the addition, deletion, or manipulation of a component or components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?

2. Could it impact system operating parameters (e.g., temperatures, flow rates, pressures, voltage, or fluid chemistry)?

3. Does it impact equipment operation or response such as valve closure time?

4. Does it involve assumptions or necessitate changes to the sequencing of operational steps?

5. Does it transfer an electrical load to a different circuit, or impact when electrical loads are added to or removed from the system during an event?

6. Does it influence fuse, breaker, or relay coordination?

7. Does it have the potential to affect the analyzed conditions of the environment for any part of the Reactor Building, Containment, or Control Room?

8. Does it affect TS/TS Bases, USAR, or other Licensing Basis documents?

[]

[X]

[ ] [x]

[] [Xl

[] [x]

[ ] [x]

[ ] [x]

[x] []

[x] [ ]

ix] []

[ ] [x]

[1

[1

[1

[]

[]

[

[ ]

[]

9.

10.

Does it affect DCDs?

Does it have the potential to affect procedures in any way not already mentioned (refer to review checklists in Procedure EDP-06)? If so, identify:

[

[I

]

]

If all answers are NO, then additional review or assistance is not required.

If any answers are YES or UNCERTAIN, then the Preparer shall obtain assistance from the System Engineer and other departments, as appropriate, to determine impacts to procedures and plant operations. Affected documents shall be listed on the Design Calculation Cross Reference Index.

UT a

Page:

6 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Table of Contents D E S IG N IN P U T.........................................................................................................................

2 AFFECTED DOCUMENTS..................................................................................................

4 AFFECTED DOCUMENT SCREENING..............................................................................

5 R E V IS IO N S U M M A R Y..............................................................................................................

6 P U R P O S E.................................................................................................................................

7 A S S U M P T IO N S........................................................................................................................

7 M E T H O D O LO G Y....................................................................................................................

10 C O M P U T A T IO N S...................................................................................................................

14 C O N C L U S IO N S......................................................................................................................

16 R E F E R E N C E S.........................

16 F IG U R E S......................................

16 A T T A C H M E N T S.....................................................................................................................

16 Revision Summary Revision 0: - Original issue.

Revision 1: - Changed calculation from Status 3 to Status 1.

Calc Change Notice - ICI: Clarified MPR Assumption 4.9 to indicate that the Group 6 trip, which isolates the RR MG set to trip the inlet and outlet ventilation dampers to close is received either due to high drywell pressure, low reactor level (Level 2) or high Reactor Building radiation. The Group 6 low water level signal was previously at Level 3 but was changed as a result of CED6014280 to address SIL 131.

Calc Change Notice - 1C2: This change notice added a sensitivity study to determine the Reactor Building temperatures if using Service Water Backup cooling to the Reactor Building Quads and the HPCI Room, with the fan coil units (FCUs) starting 1-hour after a LOCA.

Revision 2: Revision 2 is performed in support of the REC Margin Licensing Amendment [Draft NLS2007004].

The allowable time for switchover from REC cooling to Service Water Backup cooling is extended from 20-minutes to 1 -hour. Whereas previous versions have generally assumed that loss-of-REC cooling occurs after 7-days or else at time zero, in this revision it is assumed that Loss-of-REC can occur anytime after accident occurrence, with 20-hours used as the failure timing for this event. (Refer to Assumptions #2 and #3).

Revision 2 supersedes all previous revisions of this calculation. Due to insufficiently fine print edit frequencies in earlier revisions, the peak temperatures were not captured, thus, it is not meaningful to attempt comparisons of peak temperatures to the earlier revisions. Changes to the print edits will also impact temperature file integration (and thus the Arrhenius temperatures) by some amount, since the GOTHIC print edit files (i.e. output data files) have been used for this purpose.

Page:

7 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Purpose This calculation will determine the post-LOCA Reactor Building temperatures if using Service Water (SW) for essential cooling in the Quad Rooms and the HPCI Room. Area temperatures are computed for 50 volumes. Nine cases are analyzed with different equipment alignment and different failure assumptions. The resultant temperature profiles are used in the environmental qualification evaluation for the EQ classified essential components in the CNS Reactor Building Secondary Containment.

This analysis supersedes all previous editions of this calculation and provides new design basis information for the CNS Equipment Qualification (EQ) Program.

During the setup runs for this analysis, it was discovered that previous work had not captured the peak temperatures. This is because the print edit frequency (described in the GOTHIC 'Run Control Parameters') was too coarse. For example, refer to page 107 of Attachment C of NEDC 00-095E, Revision 0, for Case 1 R. Revision 0 took print edits every 6e+005 seconds after 7-days, during the time of switchover to SW Backup, whereas the Graph Interval was 60-seconds. Thus, while the switchover event is captured on the graphs, the maximum temperature is not captured by the printed data files.

Since the printed data are used for EQ input in NEDC 03-027, some impact to the older EQ results is expected in this regard.

Event Description A Loss-of-Coolant Accident (LOCA) is assumed to occur. Consistent with the CNS Licensing Basis, a Safe Shutdown Earthquake (SSE) is also assumed to occur, although it has no particular consequences here since all essential piping in the Reactor Building is Class 1 Seismic or Class 1 Restrained. To maximize building heat loads, preferred power is assumed to be available, i.e., there is no Loss-of-Offsite Power (LOOP).

Since this analysis is developed mainly for EQ purposes, the input parameters are generally structured so that they are bounding, in that a compendium of events is considered. For example, from DI #1, the suppression pool temperature used in the GOTHIC model represents the most conservative profile of several cases, including the CNS Small-Break, the Large-Break and the NPSH Analysis.

With no LOOP, a postulated single-failure could include Loss-of-REC (one division) or else a single fan coil unit (FCU) failure, but not both. Some cases evaluated in this calculation do consider failure of both divisions or else equipment failures on both. However, this approach is well outside 10 CFR 50 Appendix A criteria, hence, most cases simulated are beyond-design-basis.

Assumptions

1.

In support of the REC Margin Amendment [Draft NLS2007004], this calculation assumes that post-LOCA Loss-of-REC cooling can occur at any time after accident occurrence. (Previously analyzed cases have generally assumed Loss-of-REC after 7-days or else at time zero.)

Page:

8 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET

2.

Scoping analysis, performed in NEDC 97-085, Rev. 8 and NEDC 92-093, Rev. 9, indicates that the worst time for Loss-of-REC is approximately 20 to 22-hours after accident occurrence. In this time window, the suppression pool and the torus area temperatures are both near maximum, and the equipment soak time is also maximized. (Because the suppression pool temperature would remain above 200'F for most of the first day following a large LOCA, the peak temperatures in the ECCS Quad Rooms are not especially sensitive to the failure timing.) Further scoping analysis, performed with GOTHIC preparatory to the production runs in this calculation, has confirmed that Loss-of-REC after 20-hours (compared to time zero or after 7-days) yields higher peak temperatures in the Quad Rooms.

3.

Service Water Backup cooling is assumed to be initiated within 1-hour after Loss-of-REC cooling.

This also supports the REC Margin Amendment.

Pertinent assumptions and clarifications from Revision 0 of NEDC 00-095E and Sections 4 & 6 of MPR Calculation 0315-0302-004 (DI #12 - Attachment A) are carried forward:

4.

The 1507F fuel pool temperature assumed for post-LOCA conditions corresponds to the maximum allowable fuel pool temperature during fuel handling operations. (Refer to GE Spec. 22A2858 [DI

1. 34] and GE Dwg. 729E590BA [DI #35]). The viability of post-accident fuel pool cooling and makeup is discussed in References 2 through 5. This question is also an active issue at CNS (10/2007), but its resolution is beyond the scope of this calculation.

5.

RWCU Pump A is assumed to operate in all cases until the room heats up to 200'F or is tripped.

RWCU Pump B is assumed to be off.

6.

When MPR Calculation 0315-0302-004 (DI #4), "CNS Reactor Building Post-LOCA Heatup Analysis Results", refers to demineralized water for the REC system, it is referring to the normal REC cooling source as opposed to use of SW Backup Cooling.

7.

The outside ambient conditions are based on the maximum CNS summer design temperature of 97°F. (Reference USAR X-10-1). The Service Water supply is assumed to be at the maximum allowable temperature of 95'F and the REC at 1 00'F. These are conservative and are assumed to persist for the 180-day accident duration, ignoring seasonal variations.

8.

RCIC is assumed to operate for up to 24-hours, except for Case P4A-1, which assumes 2-hours.

9.

For RHR pump operation, it is assumed the one RHR pump in a quad would be kept running and one secured if the quad temperature reaches the temperature alarm setpoint.

10.

Each quad has its own FCU that only cools that quad.

11.

The trips in the model are based on Sections 6.2 and 6.13 of MPR calculation 0315-0302-004.

Certain equipment such as HPCI, RCIC and the RWCU pumps would automatically trip off at an area temperature of 200'F. It is assumed that the CS and RHR pumps would be manually tripped 97111:

-7,

=T

-I_

Page:

9 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET after receiving high area temperature alarms at 180 0F.

Software Ouality Assurance To complete this calculation, GOTHIC verification computations have been performed on nine different computers at CNS. Correlation of the machine to the Case Number is provided In Table 1.

Table 1. CNS Computers Used for NEDC 00-095E, Rev. 2 GOTHIC Computations Case Computer I.D.

P1 ctrain24 - #10372424 P2A ctrain27 - #10372360 P2B ctrain29 - #10372405 P3 ctrain25 - #10372425 P4A G. Levy - #10357087 P4A-1 J. Wright - #10313415 P4B ctrain29 - #10372405 P5 ctrain29 - #10372405 P6 ctrain2l - #10372356

Page:

10 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Methodology I.

This analysis uses the GOTHIC Version 7.0 computer code and the analytical model of the CNS Reactor Building that was developed by MPR Associates in NEDC 02-008A (DI #9). Most of the model remains as originally documented in DI #9, as modified through Revision 1 of NEDC 00-095E and as documented in MPR Calculation 0315-0302-004 (DI #12 - Attachment A), except with the changes noted for this calculation. Table 2 lists the cases covered by this calculation. All have been modeled for 180-days' simulation time. The primary changes are updated print edits, updated HPCI forcing function (approved in CCN-1 C2), trips changed as noted in Table 2 to account for a one-hour delay in switching from REC Cooling to SW Backup Cooling, plus an REC failure timing of 20-hours.

Table 2. Simulation Cases for NEDC 00-095E, Revision 2 FCU Failure Switchover to Case Case Definition Assumptions Volume Service Water (hours)

Base Case, Switchover to SW P1 Backup at 20-hours No FCU Failures N/A 20 Loss-of-FCU in SW RHR Fail SW FCU at P2A Room t=0 TV3 20 Loss-of-FCU in NW RHR Fail NW FCU at P2B Room t=0 TV4 20 Fail HPCI FCU P3 Loss-of-FCU in HPCI Room at t=0 TV5 20 Loss-of-FCU in SE Core Spray Fail SE FCU at P4A Room, CRD On t=0 TV2 20 Fail SE FCU at P4A-1 Same as P4A, but CRD is OFF t=0 TV2 20 Loss-of-FCU in NE Core Spray Fail NE FCU at P4B Room t=0 TV1 20 Failure of Drywell FCUs to Isolate; 1 157 REC Inlet P5 Temp. to ECCS No FCU Failures N/A None Same as Case P 1, with REC Cooling for Event Duration; P6 No SW Backup No FCU Failures N/A None W'T' 7--C "a vz

.W

- ER:ff-f

Page:

11 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION, SHEET

2.

Various segments of each transient were examined during production run preparation. The print edit frequency is listed in Table 7 and was set as fine as practical to capture various events of interest such as timing of peak temperatures, early transient behavior, etc. For example, edits were taken every 10-seconds during a two-hour time period enveloping the switchover to SW Backup cooling, and other events of interest were plotted every 30 to 60-seconds. Compromises were necessary here, because too many edits yielded files so large (over 6 GB) that Windows could not open or close them.

3.

Trips #3 and #4 are adjusted so that REC trips off after 20-hours (72,000-seconds), with a one-hour delay (3600-seconds) before initiating SW Backup cooling.

4.

Cooler/Heater (C/H) numbers 50H thru 57H, the fan cool units in the ECCS Quad Rooms, are controlled by Trips #3 and #4. These parameters control the quad cooling by REC or Service Water. Trip timing is summarized in Table 5. The trips themselves are detailed in the attached input decks under 'Component Trips' and 'Cooler/Heater'.

5.

The RCIC Turbine (C/H #58H, 61H, 63H, and 64H) is controlled by Trip #10 and is tripped off after 24-hours. This is overly conservative since RCIC would not run for this long following a large break. [Conversely, during a small break, other heat loads (principally Core Spray Pump 'A' in the NE Quad) are unnecessary for RPV level control and would be secured earlier. Area temperatures would be reduced in this case.] RCIC is assumed not to operate above 200°F ambient room temperature. Case P4A-1 is more realistic and models RCIC off after 2-hours in the NE Quad.

6.

The REC Pump (#70H) is controlled by Trip #3, this heat load being set to zero coincident with the assumed timing of REC failure.

7.

A CRD pump is assumed to run continuously in the SE Quad. This could happen if power remains available, but it is unrealistic if power is lost because the pump would trip. However, if REC failure is assumed, the CRD pump would seize within a short time due to loss of bearing cooling.

Hence, more realistically, Case P4A-1 assumes that the CRD pump is unavailable at time zero due to the assumed Loss-of-REC cooling in the SE Quad at the start of the event.

8.

HPCI room cooling performance in the GOTHIC model is adjusted to account for reduced cooling, conservatively assuming that SW is used instead of REC. The HPCI turbine heat load to the HPCI room is modeled using Forcing Function #36, which is an input to Heater/Cooler 60H.

Heater/Cooler 60H trips after 8-hours of operation or when the HPCI room temperature exceeds 200'F. Forcing Function #36 includes the combined effects of room heat up sources as well as the room cooling sources as shown in calculation NEDC 02-008A (DI #9). To account for less heat removal by SW compared to REC, Forcing Function #36 has been modified with the lower cooler performance.

For the period between time-zero and 8-hours (mission time of HPCI) the forcing function is adjusted to provide cooling only when the HPCI room is at elevated temperatures. This is

~:ii j

Page:

12 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET accomplished by adjusting the forcing function as indicated below, so that no cooling is provided until a room temperature is reached that would occur if no cooling were provided.

The HPCI room heat loads and FCU cooling performance are adjusted based on Attachment A of calculation NEDC 02-008A (DI #9) for Volume 5 (HPCI Room). A quadratic equation of the form Q=A + B*T + C*T 2 is used to define the heat loads as well as the FCU cooling performance.

The cumulative heat loads in Forcing Function #36 for Volume 5 are defined for temperature ranges as follows, based on Table A-I (p. 106) and Table F-2 (p. 197) of NEDC 02-008A:

0 0

0 T < 135 0F; A = 197,627 and B = -734.4 (C = 0) 135°F < T < 1700F; A = 132,536.7 and B = -255.4 (C = 0)

T > 170'F; A = 126,447.6 and B = -219.6 (C = 0)

Cooling performance for REC water at 100°F is determined to be A = 371,500; B = -3715 and C = 0 for room temperatures equal to or greater than 100°F as shown in Table A-I of NEDC 02-008A. For room temperatures less than I 00°F no cooling is assumed. Similar treatment is used for Case P5 with REC water at 11 5°F. The coefficients are summarized in Table 3, with all three sets shown for completeness, although the service water coefficients are the ones used for HPCI.

Table 3. Coefficients for HPCI Room FCU Heat Removal Coefficient SW at 95F REC at 100F REC at 115F A

318,535 371,500 427,225 B

-3353

-3715

-3715 C

0 0

0 Summing the heat loads and the cooling rates, the revised Forcing Function #36 values are provided in Table 4 for the net heat transfer rate into and out of the HPCI room in Btu/sec:

Table 4. Forcing Function #36 Values for Volume 5 Heat Load Room Temperature REC @ 100F REC @ 115F No Cooling degF Heat Rate (Btu/sec)

Heat Rate (Btu/sec)

Heat Rate (Btu/sec) 32 48.368 48.368 48.368 95 35.516 35.516 35.516 100 34.496 34.496 34.496 104 29.553 33.680 33.680 115 15.957 31.436 31.436 125 3.598 19.077 29.396

Page:

13 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET 135

-8.762 6.718 27.356 145

-19.909

-4.430 26.529 150

-25.423

-9.944 26.174 165

-41.966

-26.487 25.110 180

-58.411

-42.932 24.144

>200

-80.270

-64.791 22.924 Cases:

All Others P5 P3 Table 5. Trip Tim ing Summary for REC, Service Water and FCUs Time to Align REC Service Time SW Backup Case (Motor Trips)

Cooling FCU Failure Figures hours hours P1 20 1

None 1 & 2 P2A 20 1

SW Quad Cooler at t=0 3 & 4 P2B 20 1

NW Quad Cooler att=0 5 & 6 P3 20 1

HPCI Room Cooler at t=0 7

P4A 20 1

SE Quad Cooler at t=0 8 & 9 SE Quad Cooler at t=0 CRD Off at t=0 P4A-l 20 1

RCIC Off after 2-hours 10 & 11 P4B 20 1

NE Quad Cooler at t=0 12 & 13 180-days (never DW FCUs Fail to Isolate P5 off)

N/A REC at l150 F 14 & 15 180-days (never P6 off)

N/A None 16 & 17

• lf<

~

Page:

14 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Computations The simulations have been performed with GOTHIC Version 7.0 for a 180-day time period in each case.

GOTHIC results for post-LOCA Cases P 1 through P6 (nine cases total) are attached. GOTHIC graphical output is provided in the attachments for the 50 volumes modeled in the Reactor Building.

The graphs provide a time history of the post-LOCA. Reactor Building temperatures. The print-edit files from GOTHIC have been transposed to spreadsheets for input to calculation NEDC 03-027 Revision 2.

NEDC 03-027 will determine the corresponding, resultant post-LOCA Arrhenius-Equivalent temperatures over 6-months for these cases. The GOTHIC output files and the attendant spreadsheets are maintained electronically and have been written to a CD that accompanies the file copy of this calculation. The electronic files contain the output information for all 50 volumes.

Because of the logarithmic time scale and the 180-day simulation time period, the GOTHIC plots are small and difficult to read. Therefore, more detailed figures have been provided (on appropriate time scales) that illustrate the main features of each computational run. Figures 1 through 17 illustrate the thermal response of the ECCS Quads and the HPCI Room, i.e. GOTHIC volumes TV1 through TV5.

These figures also illustrate the timing of the assumed REC failure (20-hours) and the volume in which it occurs, e.g. GOTHIC TVl in Case P4B for an assumed FCU failure in the Northeast Core Spray Quad, etc. These plots provide an easy way to check for correct implementation of the REC trip at 20-hours, providing Service Water Backup Cooling one-hour later, and other events of interest. Some plots illustrate behavior early in the event, and others are provided to capture peak temperatures in the volume where an FCU failure is assumed. The peak calculated temperatures generally occur at 21 -hours, i.e.,

one-hour after REC failure and just prior to commencing Backup Cooling, the exceptions being the rooms in which REC is assumed to fail at time zero. Peak calculated temperatures are listed in Table 6.

The time occurrence of these temperatures is accurate within about 10-seconds to 60-seconds, depending on the print edit frequency. Inspection of Figures 1-17 facilitates judgment of adequacy of the print edits, which are listed in Table 7.

Case P4A-1 is interesting. In Figure 11, it can be seen that the reduced heat load in the SE Quad (TV2 -

due to securing the CRD Pump) causes the Core Spray pump to run for a longer duration before reaching the 180'F trip compared to Case P4A (Figure 9). As a result, the thermal load is increased in TV2 beyond the usual expectation. Although the difference is expected to be small, this effect can be evaluated from the results of Calculation NEDC 03-027 by comparing the Arrhenius temperatures for the two cases.

Page:

15 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Table 6. GOTHIC Maximum Calculated Temr eratures in the ECCS Quads and HPCI Room Case TV1 (NE)

TV2 (SE)

TV3 (SW)

TV4 (NW)

TV5 (HPCI) degF degF degF degF degF P1 180.6 165.1 177.2 193.6 148.2 P2A 179.9 169.5 184.1 194.1 151.9 P2B 180.4 165.7 188.9 209.3 148.2 P3 180.5 164.9 188.7 194.4 165.1 P4A 179.9 179.0 188.4 194.0 147.1 P4A-1 176.8 180.0 188.4 193.8 147.3 P4B 179.4 174.5 171.4 193.6 146.8 P5 156.5 150.3 162.2 164.3 153.6 P6 146.2 138.7 150.9 152.7 148.2 Table 7. GOTHIC Print Edit Freq ency Time Interval (seconds)

Print Edit Interval Print Edit Interval seconds 0.0 - 0.01 0.01 0.01 sec.

0.01 - 1.0 1.0 1.0 sec.

1.0- 10 1.0 1.0 sec.

10 - 1,000 30 30 sec.

1,000 - 7,200 300 5-min.

7,200 - 21,600 600 10-min.

21,600 - 71,800 1800 30-min.

71,800 - 79,200 10 10-sec.

79,200 - 142,000 3600 1-hour 142,000 - 148,000 60 1-min.

148,000 - 864,000 3600 1-hour 864,000 - 15,552,000 86400 1-day Note: Small exceptions to the above print edits will be noted in some of the input decks.

Refer to the 'Run Control Parameters' section of the input in this regard.

Page:

16 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Conclusions Figures 1 thru 17 provide detailed plotted output for the nine cases examined. Attachment A provides parent Calculation 0315-0302-004 by MPR Associates. Attachments B thru J provide the GOTHIC input decks and the plots produced by the GOTHIC output processor for the respective cases. The tabulated output is provided in text files and Excel spreadsheets. These are both on a CD that accompanies the file original of this document. These results are considered conservative, they are consistent with the proposed REC Margin Licensing Amendment, and they provide input for the CNS EQ analysis.

References

1. CNS EQ Program Basis Document, Volumes 1 and 2

2. CNS OER, "Potential Loss of Spent Fuel Pool Cooling Following a Loss of Coolant Accident (LOCA)", dated March 16, 1993

3. General Electric Company, "Information Regarding the Licensing Basis of GE Boiling Water Reactors Regarding Loss of Coolant Accidents and Losses of Offsite Power", GE-NE-A0005683-01, dated June 30, 1993

4. Letter from L. A. England, BWROG, to A. C. Thadani, USNRC, "BWR Owners' Group Assessment of the Spent Fuel Pool Cooling System", BWROG-94024, dated March 4, 1994

5. USNRC Memorandum from James M Taylor to Chairman Jackson, "Resolution of Spent Fuel Storage Pool Action Plan Issues", dated July 26, 1996 Figures Figures 1 through 17 illustrate plotted results for GOTHIC Volumes TV1 thru TV5, as delineated in Table 5.

Attachments A.

MPR Calculation 0315-0302-004 (Body Only)

(25 pages)

NEDC 00-095E, Revision 2 Computations (GOTHIC Input Decks and Plotted Results)

B.

Case P1 Base Case C.

Case P2A Fail FCU in SW Quad (TV3)

D.

Case P2B Fail FCU in NW RHR Room (TV4)

E.

Case P3 Fail FCU in HPCI Room F.

Case P4A FCU Failure in SE Core Spray Room (TV2)

L CZ

Page:

17 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET G.

Case P4A-1 Fail FCU in SE Quad (TV2) IRCIC & CRD OffJ H.

Case P4B FCU Failure in NE Quad (TVI)

I.

Case P5 Drywell FCU Failure J.

Case P6 Base Case with REC Only K.

CD-ROM Data Disk with Electronic Files and Results

Page:

18 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 1 Case PI: Base Case - All FCUs On Loss-of-REC after 20-hours 220 200 180 h..

= 160 E

140 120 TV1-NE TV2-SE TV3-SW TV4-NW TV5-HPCI 100 4-0.01 0.1 1

10 100 1000 Time (seconds) 10000 100000 1000000 10000000 100000000

Page:

19 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 2 Case PI: Base Case - All FCUs On Loss-of-REC after 20-hours 220 200 180 I-

= 160 140 120 100 TV1 -NE TV2-SE TV3-SW TV4-NW TV5-H PCI 0

10000 20000 30000 40000 50000 Time (seconds) 60000 70000 80000 90000 100000

Page:

20 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 3 Case P2A: Fail SW FCU at Time Zero Loss-of-REC after 20-hours 220 200 180 I.

= 160 E

140 120 100 TV1-NE TV2-SE TV3-SW TV4-NW TV5-HPCI 0.01 0.1 1

10 100 1000 Time (seconds) 10000 100000 1000000 10000000 100000000

Page:

21 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 4 Case P2A: Fail SW FCU at Time Zero Loss-of-REC after 20-hours 220 200 180 IL

" 160 E

140 120 100 TV1-NE TV2-SE TV3-SW TV4-NW TV5-HPCI 0

10000 20000 30000 40000 50000 60000 70000 80000 90000 100000 Time (seconds)

Page:

22 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 5 Case P2B: Fail NW FCU at Time Zero Loss-of-REC after 20-hours 220 200 E

I.-

180 160 140 TV1-NE TV2-SE TV3-SW TV4-NW

---

TV5-HPCI 120 100 -!-

0.01 0.1 1

10 100 1000 Time (seconds) 10000 100000 1000000 10000000 100000000

Page:

23 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 6 Case P2B: Fail NW FCU at Time Zero Loss-of-REC after 20-hours 220 200 U-S "0

S 4-.

a E

S I-180,

160 TV1-NE TV2-SE TV3-SW TV4-NW TV5-HPCI I -#'U 120 100 0

20000 40000 60000 80000 100000 Time (seconds) 120000 140000 160000 180000 200000

Page:

24 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 7 Case P3: Fail HPCI FCU at Time Zero Loss-of-Rec after 20-hours 220 200 180 CU-

"0 V

E 0I-160 140 TV1-NE TV2-SE TV3-SW TV4-NW

--.-- TV5-HPCI 120 100 0.01 0.1 1

10 100 1000 10000 100000 1000000 Time (seconds) 10000000 100000000

Page:

25 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 8 Case P4A: Fail SE FCU at Time Zero Loss-of-REC after 20-hours 220 200 180 160 140

IL0, S

I-a a

E SI-.

TV1-NE TV2-SE TV3-SW TV4-NW TV5-H PCI 120 100 1 1.OE-02 1.OE-01 1.OE+00 1.0E+01 1.OE+02 1.OE+03 1.OE+04 1.OE+05 1.0E+06 1.OE+07 1.OE+08 Time (seconds)

Page:

26 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 9 Case P4A: Fail SE FCU at Time Zero Loss-of-REC after 20-hours 220 200 180 0,

S 10

• = 160 E

S I-140 120 100 - TV1-NE TV2-SE TV3-SW TV4-NW

-- E-TV5-HPCI 0

10000 20000 30000 40000 50000 60000 70000 80000 90000 100000 Time (seconds)

Page:

27 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 10 Case P4A-1: Fail SE FCU at Time Zero Loss-of-REC at 20-hours; CRD & RCIC Off 220 200 180 U-S 0

1.

4-I!&E S

jI if 7:

160 140 TV1-NE TV2-SE TV3-SW TV4-NW TV5-H PCI

,~f.

moor-, X" r

N111111111WNV*ý 120

• 100 0.01

-4q*ý

-y~

10000 100000 1000000 10000000 100000000 0.1 1

10 100 1000 Time (seconds)

Page:

28 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 11 Case P4A-1: Fail SE FCU at Time Zero Loss-of-REC at 20-hours; CRD & RCIC Off 220 200 IL 0*

S b.

• 1 E

S I-180 160 140 TV1-NE

-U-TV2-SE TV3-SW TV4-NW TV5-HPCI 120 100 0

10000 20000 30000 40000 50000 60000 70000 80000 90000 100000 Time (seconds)

Page:

29 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 12 Case P4B: Fail NE FCU at Time Zero Loss-of-REC after 20-hours 220 200 SL

• 0 E

180 160 140

- TV1-NE TV2-SE TV3-SW TV4-NW TV5-HPCI 120 100 0.01 0.1 1

10 100 1000 Time (seconds) 10000 100000 1000000 10000000 100000000

Page:

30 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 13 Case P4B: Fail NE FCU at Time Zero Loss-of-REC after 20-hours 220 200 180 La-S0 0a 6S I-160 140 TV1-NE TV2-SE TV3-SW TV4-NW

--E--- TV5-H PCI 120 100 0

10000 20000 30000 40000 50000 Time (seconds) 60000 70000 80000 90000 100000

Page:

31 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 14 Case P5: REC Inlet Temp. at 115F All FCUs Operable in ECCS Rooms 180 170 160

" 150 140 I-S140 E 130 120 110 100 TV1-NE

-a-TV2-SE TV3-SW TV4-NW TV5-HPCI 0.01 0.1 1

10 100 1000 Time (seconds) 10000 100000 1000000 10000000 100000000

Page:

32 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 15 Case P5: REC Inlet Temp. at 115F All FCUs Operable in ECCS Rooms 180 170 160

. 150

= 140 0.

E0 130 120 110 100 TV1 -NE TV2-SE TV3-SW TV4-NW TV5-HPCI 0

2000000 4000000 6000000 8000000 10000000 12000000 14000000 16000000 Time (seconds)

Page:

33 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 16 Case P6 - REC Cooling for Event Duration No Switchover to SW Backup 160 150 140 IL E,

130 120 TVI-NE TV2-SE TV3-SW TV4-NW TV5-HPCI 110 100 4-0.01 0.1 1

10 100 1000 Time (seconds) 10000 100000 1000000 10000000 100000000

Page:

34 of 34 NEDC: 00-095E Rev. Number:

2 Nebraska Public Power District DESIGN CALCULATION SHEET Figure 17 Case P6 - REC Cooling for Event Duration No Switchover to SW Backup 160 150 140 L-0.

E 0*

I-130 120 TV1-NE

-U-TV2-SE TV3-SW TV4-NW TV5-HPCI 110 100 0

2000000 4000000 6000000 8000000 Time (seconds) 10000000 12000000 14000000 16000000

NLS2008067 NEDC 03-027, Revision 2, "Assessment of Post-LOCA Heat Up Temperature Profiles for Reactor Building" (Excerpted pages 1 through 12)

Cooper Nuclear Station, Docket No. 50-298, DPR-46

ATTACHMENT 1 DESIGN CALCULATION COVER SHEET Page I of 12

Title:

Assessment of Post-LOCA Heat Up Temperature Profiles Calculation Number:

NEDC 03-027 for Reactor Building CED/EE Number: N/A System/Structure: Reactor Building / Secondary Containment Setpoint Change/Part Eval Number: N/A Component:

Equipment in Secondary Containment Discipline:

EQ Classification: [ X ] Essential; [ ] Non-Essential SQAP Requirements Met? [ ] Yes; [ X ] N/A Proprietary Information Included? [ ] Yes; [X] No

==

Description:==

The purpose of this calculation is to perform an assessment of Post-LOCA Heat Up (PLHU) temperatures due to REC switchover to Service Water (SW) at t=20 hours with switchover taking 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Case scenarios include FCU failures as well as no failure baseline without switchover. The assessment will determine if these scenarios will change the current PBD-EQ qualification values.

The purpose of Revision 1 was to:

(1) include the temperature effect to the critical switchgear rooms. The results show that the Arrhenius equivalent aging temperature for the PLHU 180 days for these rooms is about 1200F, which is still within the definition of a temperature mild environment; (2) include an Arrhenius equivalent aging temperature for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the HPCI room; and (3) eliminate Attachment B evaluations as they are now included in the Status 1 EQDP's implemented by EE04-012.

The purpose of Revision ICI was to:

perform a sensitivity study of PLHU temperatures resulting from the use of SW instead of REC for essential quad cooling. SW was assumed available 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after event initiation. The results showed that there was no appreciably difference from the existing design basis cases using REC for essential quad cooling.

Conclusions and Recommendations:

The assessment performed has determined no impact to existing EQ-PBD Post-LOCA Heat Up values. No charges are required for the EQ-PBD as the original Conclusions and Recommendations remain bounding.

Attachment A - Post-LOCA Heat Up Profiles

~

~

//'

~

11111,c~ 7 4/11/0/6' 1 Cl 3

Ken Done 5/4/06 Tim Pospisil 5/5/06 Tim Pospisil 5/5/06 T. Stevens 8/1/06 1

1 Calc Body & Attachment A -

Mark E Unruh Mark E Unruh J.E. Lechner Gregory Chinn 3-25-04 3-25-04 7-29-04 Attachment B -

Mike Baldwin Mike Baldwin Tim Pospisil 6/26/03 6/26/03 6/26/03 Tim McClure 0

Calc Body & Attachment A -

Steve Nelson Steve Nelson 6/27/03 Gregory Chinn 6/26/03 6/26/03 6/26/03 Rev.

Status Prepared By/Date Reviewed By/Date IDVed By/ Date Approved By/Date Number Status Codes

1. Active

2. Information Only

3. Pending

4. Superseded or Deleted

5. OD/OE Support Only

6. Maintenance Activity Support Only

7. PRAIPSA

$3~

ATTACHMENT 2 DESIGN CALCULATION CROSS-REFERENCE INDEX Page:

2 of 12 NEDC:

03-027 Rev.

2 Number:

Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM REV.

PENDING CHANGES NO.

DESIGN INPUTS NO.

TO DESIGN INPUTS 1

NEDC 00-095E 2

Concurrent approval.

2 EQ Program Basis Document 4

3 NEDC 02-069 3

__ I:

I

4.

4

.4

.4 4

1-

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ATTACHMENT 2 DESIGN CALCULATION CROSS-REFERENCE INDEX Page:

3 of 12 NEDC:

03-027 Rev.

Number:

2 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM AFFECTED DOCUMENTS REV. NUMBER NO.

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I ATTACHMENT 3 AFFECTED DOCUMENT SCREENING Page:

4 of 12 NEDC:

03-027 Rev.

Number:

2 The purpose of this form is to assist the Preparer in screening new and revised determine potential impacts to procedures and plant operations.

design calculations to SCREENING QUESTIONS

1)

Does it involve the addition, deletion, or manipulation of a component or components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?

2)

Could it impact system operating parameters (e.g., temperatures, flow rates, pressures, voltage, or fluid chemistry)?

3)

Does it impact equipment operation or response such as valve closure time?

4)

Does it involve assumptions or necessitate changes to the sequencing of operational steps?

5)

Does it transfer an electrical load to a different circuit, or impact when electrical loads are added to or removed from the system during an event?

6)

Does it influence fuse, breaker, or relay coordination?

7)

Does it have the potential to affect the analyzed conditions of the environment for any part of the Reactor Building, Containment, or Control Room?

8)

Does it affect TS/TS Bases, USAR, or other Licensing Basis documents?

9)

Does it affect DCDs?

10)

Does it have the potential to affect procedures in any way not already mentioned (refer to review checklists in Procedure EDP-06)? If so, identify:

YES NO UNCERTAIN

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[x]

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[x]

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[x]

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[x]

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[1 I:]

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[I I]

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If all answers are NO, then additional review or assistance is not required.

If any answers are YES or UNCERTAIN, then the Preparer shall obtain assistance from the System Engineer and other departments, as appropriate, to determine impacts to procedures and plant operations. Affected documents shall be listed on Attachment 2.

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Page: 5 of 12 NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET PURPOSE The purpose of this calculation is to perform an assessment of a new set of Reactor Building Post-LOCA Heatup (PLHU) temperature profiles developed in NEDC 00-95E (Ref. 1). NEDC 00-95E has identified revised scenarios involving Post-LOCA Fan Coil Unit (FCU) failure and REC switchover to Service Water (SW) different from those evaluated in the previous revision of this calculation. These new cases are identified in Table 1. The previously evaluated results are identified in the EQ Program Basis Document, Volume 2 (Ref. 4) and are also identified in Tables 2 and 3 of this calculation. Attachment A to this calculation provides the new PLHU temperature profiles and develops the Arrhenius equivalent aging temperatures for the profiles.

This calculation compares the new Arrhenius equivalent aging temperatures to those previously developed (Table 2). This calculation performs an assessment of the impact of the new temperature profiles on EQ.

ASSUMPTIONS

1. The post-LOCA heat up profiles shows a gradual heat up within a volume. Maximum temperatures are not reached until about 1 E5 to 1 E6 seconds (approximately 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> to 11.6 days). This type of gradual heat up is not indicative of "thermal shock" conditions.

Temperature decays occur from about 11 days to 180 days with the temperature difference typically less than 40°F. This slow decay in temperature is also assumed not indicative of "thermal shock".

2. The Arrhenius equivalent aging time/temperature methodology is applicable to steady state temperature conditions. Based on Assumption #1 above, continuous discreet segments of the post-LOCA heat up profiles can be assumed to be relatively steady state and therefore the Arrhenius methodology is applicable.

3. An activation energy of 2.0 eV is used for the Arrhenius equation. This value is assumed conservative for the calculation of the Arrhenius equivalent aging temperature. Based on the Reference 3 Appendix B Histogram of Activation Energies, this assumption is valid.

Use of a comparable high activation energy results in higher equivalent temperature predictions.

4. Peak temperatures for the PLHU conditions are not significant in the evaluation of EQ equipment. This assumption is valid because High Energy Line Break peak temperatures envelop the PLHU peak temperatures with substantial margin (see Table 3). EQ equipment is qualified to the HELB peak temperatures, which provides evidence of qualification to the PLHU peak temperature.

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Page: 6 of 12 NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET METHODOLOGY Reference 1 provides the post-LOCA heat up profiles evaluated in this calculation. The Case descriptions are provided in the following table:

TABLE 1 - PLHU CASE DESCRIPTIONS Case ID Case Description Notes Baseline case, REC demineralized water cooling No single failure P1 switchover to Service Water (SW) at t=20 hours postulated (Note: switchover takes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)

Loss of Fan Cooling Unit (FCU) in SW RHR Room Single failure of a P2A (HELB Vol. 3) at t=O; switchover to SW for non failed FCU or relay failure FCU at t=20 hours. (Note: switchover takes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)

Loss of FCU in NW RHR Room (HELB Vol. 4) at t=0; Single failure of a P21 switchover to SW for non failed FCU at t=20 hours.

FCU or relay failure (Note: switchover takes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)

Loss of FCU in HPCI Room (HELB Vol. 5) at t=O; Single failure of a P3 switchover to SW for non failed FCU at t=20 hours.

FCU or relay failure (Note: switchover takes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)

Loss of FCU in SE CS Room (HELB Vol. 2) at t=0; Single failure of a P4A switchover to SW for non failed FCU at t=20 hours.

FCU or relay failure (Note: switchover takes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)

Same as P4A but with CRD OFF from t=0 (no heat Single failure of a P4A-1 load in SE CS Room from CRD pump or motor.

FCU or relay failure Note: switchover takes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)

Loss of FCU in NE CS Room (HELB Vol. 1) at t=O; Single failure of a P4B switchover to SW for non failed FCU at t=20 hours.

FCU or relay failure (Note: switchover takes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)

Failure of Drywell FCU's to isolate, 115 0F REC inlet No FCU failures.

temperature to ECCS. (no switchover to SW)

Same as Case P1, with REC cooling for event No FCU failures.

duration. (no switchover to SW)

Reference 2 section A5.2 provides an equation that provides the single Arrhenius equivalent aging temperature for the same duration as for a variety of temperatures. This equation is used to evaluate the post-LOCA heat up profiles to derive a single temperature representative of the entire profile. Development of the Arrhenius Equivalent temperature formula used in the Excel Spreadsheets was performed per NEDC 02-069 Attachment A.

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Page: 7 of 12 NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET Since the profiles depict relatively static temperature conditions, adjoining volumes with large flow paths can be grouped together in one Excel worksheet for Arrhenius evaluation. An example of this are the 903' elevation general area volumes (HELB volume designation);

volumes 14, 16, 18 and 20; another example are the four Torus room volumes, 6, 7, 8, and 9.

A maximum temperature profile is derived from the combined profiles and from the maximum profile an Arrhenius equivalent aging temperature is calculated.

Results Table 2 - Summarizes the Arrhenius equivalent temperatures for new cases P1 through P6.

Table 3 - Summarizes the peak temperature for new cases P1 through P6.

Attachment A - Provides time history graphs of the post-LOCA temperatures for cases P1 through P6.

Discussion Table 2 compares the previously developed Arrhenius equivalent aging temperatures which are used for EQ and documented in the EQ Program Basis Document, Volume 2 (Ref. 4) to the Arrhenius equivalent aging temperatures calculated for the Table 1 Case scenarios. The Table-2 column "Worst Case P" identifies the worst case temperature of all of the scenarios. In every applicable case, the worst case value is less than (or if rounded up, is at most equal to the) EQ Program Basis Document Arrhenius equivalent aging temperature value. In this regard, the Table 1 Case scenarios do not have an impact on the qualification of EQ equipment.

Table 3 compares the previously identified PLHU peak temperatures to the Table 1 Case scenario derived peak temperatures. In some instances, the Table 1 Case peak temperatures are greater by a few degrees with the worst case in the CS Quads (Volume 1 & 10 and Volume 2 & 11) where the temperatures are about 150F greater. As identified in Assumption 1, these peak temperatures are reached gradually and are not symptomatic of "thermal shock". In this regard, the heat effect of the peak temperature is captured in its impact on the Arrhenius equivalent aging temperature. Since Table 2 shows that all of the Table 1 Case scenario Arrhenius equivalent thermal aging temperatures are bounded by the existing values the impact of the higher peak temperature is not significant. With regard to peak temperature qualification profile enveloping, as identified in Assumption 4 and shown in Table 3, the PLHU peak temperatures are enveloped with substantial margin by the HELB peak temperatures. In this regard, that some of the Table 1 Case scenario peak temperatures are greater is not significant and has no impact on EQ.

Page: 8 of 12 NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET CONCLUSION Attachment A to this calculation provides the time history graphs of the post-LOCA heat-up temperatures and the Arrhenius equivalent temperatures for the Table 1 Case scenarios.

Tables 2 and 3 summarize the results and provide a comparison to PBD-EQ used values.

Based on the comparisons, the Arrhenius equivalent aging temperature current values used in the PBD-EQ will remain unchanged and bounding for all equipment qualification documentation used by the EQ Program to demonstrate 1 OCFR50.49 compliance. Some of the new post-LOCA heat-up scenario peak temperatures are higher than the previous values, however, they were evaluated as not being significant and no qualification changes are needed.

REFERENCES

1. NEDC 00-095E, CNS Reactor Building Post-LOCA Heatup Analysis Results, Rev. 2

2. EPRI TR-100516, EQ Reference Manual (section A5.2)

3. EPRI NP-1558, A Review of Equipment Aging Theory and Technology, Sept. 1980.

4. CNS EQ Program Basis Document, PBD-EQ, Volume 2 of 2, Rev. 3

5. NEDCOO-95A, EQ Normal Temp., Relative Humidity, Pressure and Radiation, Rev. 4 ATTACHMENTS Attachment A - PLHU Profiles ly.-w

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Page: 9 of 12.

NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET TABLE 2 - COMPARISON OF ARRHENIUS VALUES Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Worst EQ PBD Volumes HELB Case P1 Case P2A Case P2B Case P3 Case P4A Case Case P4B Case P5 Case P6 Case P (Rev. 3)

Volumes P4A-1 Arrhenius

( OF)

( OF)

( OF)

( OF)

( OF)

(OF)

( OF)

( OF)

( OF)

( OF)

(OF)

Torus 6, 7, 8, 9 134.38 134.96 135.20 134.40 134.40 133.83 134.30 136.14 133.76 136.14 137.00 Rm 1,2,3,4, QUADS 10,11, 150.76 N/A N/A 150.76 N/A N/A N/A 155.37 145.37 155.37 see below 12,13 QUADS 2, 4, N/A 151.31 N/A N/A N/A N/A N/A N/A N/A 151.31 see below 10, 11,13 QUADS 1 0,1,12 N/A N/A 147.17 N/A N/A N/A N/A N/A N/A 147.17 see below 10, 11, 12_____

1D 2,3,4, N/A N/A N/A N/A N/A N/A 150.35 N/A N/A 150.35 see below UAS 11, 12,,13 QUADS 1, 3,4, N/A N/A N/A N/A 150.72 150.10 N/A N/A N/A 150.72 see below 10,12,13 QUADS -

failed 3,12 N/A 157.20 N/A N/A N/A N/A N/A N/A N/A 157.20 158.00 FCU QUADS -

failed 4,13 N/A N/A 172.42 N/A N/A N/A N/A N/A N/A 172.42 173.00 FCU QUADS-failed 1,10 N/A N/A N/A N/A N/A N/A 133.48 N/A N/A 133.48 156.00 FCU QUADS-failed 2,11 N/A N/A N/A N/A 147.76 132.21 N/A N/A N/A 147.76 156.00 FCU HPCIRm 5

117.01 121.28 117.42 119.25 116.90 116.60 116.62 119.13 116.10 121.28 122.00 180 day HPCI Rm 5

145.67 148.68 145.65 161.09 145.15 145.33 144.77 150.59 145.67 161.09 N/A Rm 8hrI 903' Gen 14,16, 129.26 131.61 133.61 129.27 129.52 129.74 129.77 134.45 129.80 134.45 135.00 Area 18,20 1

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NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET TABLE 2 - COMPARISON OF ARRHENIUS VALUES Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius Arrhenius worst PBD Volumes HELB Case P1 Case P2A Case P2B Case P3 Case P4A Case Case P4B Case P5 Case P6 Case P ev. 3)

Volumes (F)

(OF)

(OF)

(OF)

(OF)

P4A-1 (OF)

(°F)

(OF)

(OF)

Arrhenius 9 e 22 1. 23 2

16

24.

27 35 2

130.51_

131.0 931'Gen 253, 28, 123.60 125.33 126.10 123.60 123.74 122.73 123.71 130.51 127.65 130.51 131.00 Area 31,32 958' and 33, 35, up Gen 36: 37, 120.40 122.36 121.48 120.40 120.41 119.62 120.15 123.74 121.78 123.74 124.00 Areas 38 39, 40, 41, 48 RWCU 26,27, Rms 29, 30,34, 129.51 130.47 130.81 129.52 129.65 129.35 129.60 132.74 131.00 132.74 133.00 Rms 45,46 MISC 47,19, 128.81 130.64 131.73 128.81 129.06 128.52 128.96 132.55 129.46 132.55 133.00 Rms 22, 17 Stun 15 175.77 175.78 175.78 175.77 175.77 175.74 175.77 175.80 175.76 175.80 176.00 Tunnel Valve Injection 21 132.72 133.69 135.44 132.73 132.80 132.95 132.98 136.21 133.07 136.21 137.00 Rm I

I SGTS 44 132.38 133.04 132.51 132.38 132.34 131.94 132.13 133.83 132.77 133.83 134.00 Rm Crit SWGR 23,24 117.11 117.97 119.14 117.11 117.21 117.17 117.34 120.44 118.14 120.44 N/A Rooms Note: To reconcile HELB Volumes to EQ Zones identified in the EQ PBD, refer to Attachment A of NEDC 00-095A (Ref. 5).

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NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET TABLE 3 - COMPARISON OF PEAK TEMPERATURE VALUES Case Case Case Case Case Case Case Case Case Worst EQ PBD EQ PBD (Rev. 3)

Volumes HELB P1 P2A P2B P3 P4A P4A-1 P4B P5 P6 Case P (Rev. 3)

Lowest HELB Volumes (OF)

(°F)

(°F)

(°F)

(°F)

(°F)

(OF)

(OF)

(OF)

(OF)

Peak (F)

Temp (°F)

Torus 6,7,8, 9 166.56 166.34 166.47 166.71 166.53 166.23 166.47 166.75 164.188 166.75 167.0 281 Rm (Vol. 6) 1,2,3,4, QUADS 10, 11, 12, 193.58 N/A N/A 194.44 N/A N/A N/A 164.34 152.658 194.44 see below see below 13 QUADS 1, 2, 4,10, N/A 194.07 N/A N/A N/A N/A N/A N/A N/A 194.07 see below see below 11, 13 1

QUADS 1,2, 3, 10, N/A N/A 188.86 N/A N/A N/A N/A N/A N/A 188.86 see below see below 11, 12.

1 QUADS 2,3,4,11, N/A N/A N/A N/A N/A N/A 193.57 N/A N/A 193.57 see below see below 123 10 QUADS 1,3,4,10, N/A N/A N/A N/A 193.97 193.77 N/A N/A N/A 193.97 see below see below 12, 13 QUADS -

failed 3,12 N/A 184.07 N/A N/A N/A N/A N/A N/A N/A 184.07 183.0 298 (Vol. 3)

FCU 290 (Vol. 12)

FCU QUADS -

245 (Vol. 4) failed 4,13 N/A N/A 209.34 N/A N/A N/A N/A N/A N/A 209.34 206.0 263 (Vol. 13)

FCU QUADS-285 (Vol. 1) failed 1,10 N/A N/A N/A N/A N/A N/A 179.38 N/A N/A 179.38 164.0 285 (Vol. 1)

FCU 288 (Vol. 10)

FCUL QUADS-246 (Vol. 2) failed 2,11 N/A N/A N/A N/A 179.03 179.99 N/A N/A N/A 179.99 164.0 260 (Vol. 11)

FCU 2

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1 HPCIRm 5

148.19 151.91 148.19 165.11 147.14 147.29 146.82 153.562 148.191 165.11 166.0 298 180 day I

HPCI RmCh 5

148.19 151.91 148.19 165.11 147.14 147.29 146.82 153.562 148.191 165.11 N/A N/A R~m 8hr 903' Gen 14,16,18, 135.02 140.17 140.98 135.05 135.77 135.77 135.12 138.674 135.032 140.98 138.0 243 (Vol. 18)

Area 20 1

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NEDC 03 - 027 Rev. 2 Nebraska Public Power District DESIGN CALCULATIONS SHEET TABLE 3 - COMPARISON OF PEAK TEMPERATURE VALUES Volumes HELB Case Case Case Case Case Case Case Case Case Worst EQ PBD EQ PBD (Rev. 3)

P1 P2A P2B P3 P4A P4A-1 P4B P5 P6 Case P (Rev. 3)

Lowest HELB Volumes

(°F)

(OF)

(OF)

(°F)

(OF)

(OF)

(OF)

(°F)

(OF)

(OF)

Peak (°F)

Temp (°F) 931'Gen 25,28,31, 128.17 131.09 131.92 128.18 128.94 128.39 128.24 133.739 131.548 133.74 134.0 211 (Vol. 25 and Area 32 Vol. 32) 958' and 33, 35, 36, up Gen 37, 38, 39, 123.91 126.67 125.90 123.99 124.21 123.78 124.12 126.738 125.266 126.74 127.0 198 (Vol. 39)

Areas 40,41,48 RWCU 26, 27, 29, Rms 30,34, 45, 171.61 171.79 171.63 171.62 171.63 171.62 171.63 171.832 171.846 171.85 172.0 222 (Vol. 34)

R 46 MIsc 47,19,22, 137.60 141.13 141.57 137.61 138.26 137.78 137.66 140.445 137.972 141.57 141.0 210 (Vol. 47)

Rms 17 Stm Tu 15 266.88 266.88 266.88 266.88 266.88 266.88 266.88 266.878 266.878 266.88 263.0 312 Tunnel Valve Injection 21 154.50 153.56 154.59 154.59 154.85 154.64 154.28 154.386 153.459 154.85 154.0 300 Rm SGTS 44 137.63 136.91 137.15 137.64 137.77 137.62 137.67 138.769 138.074 138.77 139.0 205 Rm Crit SWGR 23, 24 120.45 122.36 123.19 120.45 120.84 120.62 120.51 122.849 121.016 123.19 124.0 N/A Rooms Note: To reconcile HELB Volumes to EQ Zones identified in the EQ PBD, refer to Attachment A of NEDC 00-095A (Ref. 5).