Text

Large Break Loca/Eccs Analysis W/Increased Radial Peaking & Reduced ECCS Flow
	ML18057B352
Person / Time
Site:	Palisades
Issue date:	10/31/1991
From:	Chen T, Lindquist T, Stitt B; SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	;
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ML18057B349	List: ML18057B348; ML18057B350; ML18057B351; ML18057B352;
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	EMF-91-177, NUDOCS 9111080152
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SIEMENS Palisades Large Break LOCA/ECCS Analysis With Increased Radial Peaking And Reduced ECCS Flow October 1991 Siemens Nuclear Power Corporation EMF-91-177

SIEMENS PALISADES LARGE BREAK LOCNECCS ANALYSIS WITH INCREASED RADIAL PEAKING

/dlw AND REDUCED ECCS FLOW T. R. Lindquist, earn Leader PWR Fuel Engineering Fuel Engineering and Licensing Contributors:

T. H. Chen

8. D. Stitt R. T. Welzbacker October 1991 Siemens Nuclear Power Corporation Engineering and Manufacturing Facility EMF-91-1n Issue Date: 1O/Ol 191 2101 Horn Rapids Road, PO Box 130 Richland, WA 99352-0130 Tel: (509) 375-8100 Fax: (509) 375-8402

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I CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Siemens Nuclear Power Corporation's warranties and representations concerning the subject matter of this document are those set forth in the Agreement between Siemens Nuclear Power Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such Agreement, neither Siemens Nuclear Power Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rights; or assumes any liabilities with respect to the use of any information, apparatus, method or process disclosed in this document.

The information contained herein is for the sole use of the Customer.

In order to avoid impairment of rights of Siemens Nuclear Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by Siemens Nuclear Power Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document

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Table of Contents EMF-91-177 Page i

1.0 INTRODUCTION

................................................... 1 2.0

SUMMARY

OF RESULTS............................................ 3 3.0 ANALYSIS........................................................ 6 3.1 Description of LBLOCA Transient................................. 6 3.2 Description of Analytical Models................................. 8 3.3 Plant Description and Summary of Analysis Parameters................ 9 3.4 Axial Shape Study Results...................................... 9

4.0 CONCLUSION

S................................................ **.... 41

5.0 REFERENCES

42

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l EMF-91-177 Pageii List of Tables Table Page 2.1 Summary of Results for 0.6 DECLG Limiting Break Size..................... 4 3.1 Revised HPSI and LPSI Pump Flow Curves..............................

11 3.2 Palisades System Analysis Parameters.................................

12 3.3 Calculated Event Times for 0.6 DECLG Break............................ 14

List of Figures EMF-91-177 page iii.

Figures Page 2.1 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 3.10 3.11 3.12 3.13 3.14 3.15 3.16 3.17 3.18 3.19 3.20 3.21 3.22 3.23 Allowable LHR as a Function of Peak Power Location.......................

5 Double Intact Loop SIT Flow Rate, 0.6 DECLG Break......................

15 Single Intact Loop SIT Flow Rate, 0.6 DECLG Break.......................

16 Broken Loop SIT Flow Rate, 0.6 DECLG Break...........................

17 Single Intact Loop HPSI Flow Rate, 0.6 DECLG Break......................

18 Double Intact Loop SIS Flow Rate, 0.6 DECLG Break......................

19 Broken Loop LPSI Flow Rate, 0.6 DECLG Break..........................

20 Broken Loop HPSI Flow Rate, 0.6 DECLG Break..........................

21 Upper Plenum Pressure During Slowdown, 0.6 DECLG Break................

22 Total Break Flow Rate During Slowdown, 0.6 DECLG Break.................

23 Pressurizer Surge Line Flow Rate During Slowdown, 0.6 DECLG Break.........

24 Downcomer Flow Rate During Slowdown*, 0.6 DECLG Break.................

25 Average Core Inlet Flow Rate During Slowdown, 0.6 DECLG Break, X/L = 0.8...

26 Hot Channel Inlet Flow Rate During Slowdown, 0.6 DECLG Break, X/L = 0.8... ~

27 Hot Volume Inlet Flow Rate During Slowdown, 0.6 DECLG Break, X/L = 0.8. ~...

28 Hot Node Fluid Quality During Slowdown, 0.6 DECLG Break, X/L = 0.8....... ;. 29 PCT Node Fluid Temperature During Slowdown, 0.6 DECLG Break, X/L = 0.8...

30 PCT Node Fuel Average Temperature During Slowdown, 0.6 DECLG Break, X/L

= 0.8...........................................................

31 PCT Node Cladding Temperature During Slowdown, 0.6 DECLG Break, X/L =

0.8............................................................ 32 PCT Node Heat Transfer Coefficient During Slowdown, 0.6 DECLG Break, X/L =

0.8............................................................ 33 PCT Node Heat Flux During Slowdown, 0.6 DECLG Break, X/L = 0.8..........

34 Containment Pressure, X/L = Q.8.....................................

35 Upper Plenum Pressure after EOBY, X/L = 0.8............................

36 Downcomer Mixture Level after EOBY, X/L = 0.8..........................

37

EMF-91-177 Pageiv List of Figures (Cont.)

Figures Page 3.24 Core Flooding Rate after EOBY, X/L = 0.8..............................

38 3.25 Core Mixture Level after EOBY, X/L = 0.8...............................

39 3.26 PCT Node Cladding Temperature after EOBY, X/L = 0.8....................

40

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1.0 INTRODUCTION

EMF-91-177 Page 1 This document presents the results of a large break loss-of-coolant accident (LOCA) analysis for the Palisades plant operating with Siemens Nuclear Power Corporation (SNP) fuel beginning with Reload M. The primary changes supported by this analysis are:

1)

A reduction in emergency core cooling system (ECCS) flow occurring from a change in the Low Pressure Safety Injection (LPSI) flow curve and the loss of a High Pressure Safety Injection (HPSI) pump.

2)

To bound future cycles, assembly (F /')and peak rod (FrT) radial peaking limits of 1.76 and 2.04, respectively, were used.

3)

A 5 mil increase in the pellet diameter (i.e., a 5 mil decrease in pellet-to-clad gap).

4)

An increase in pellet density to 94.5% of theoretical density.

5)

Minimum Technical Specification SIT level.

The analysis was performed for the Palisades plant operating at 2581 MWt (2530 MWt plus 2% uncertainty) and incorporates a maximum average steam generator tube plugging level of 29.3% with up to 4.5% asymmetry in the system blowdown, hot channel and reflood calculations. The blowdown data reflects operation with the original steam generators and since

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the PCS inventory is reduced with extensive tube plugging, the data is conservative relative to operation with the replacement steam generators. Because tube plugging increases steam binding during reflood, use of 29.3% tube plugging in the original steam generators is conservative relative to operation with the replacement steam generators.

Single failure criteria is met by assuming that one LPSI pump is not available for operation(6). The configuration of a loss of a LPSI and a HPSI was, however, analyzed in order to provide an analysis consistent with the conditions considered by the transient analyses for other events. The loss of a LPSI and a HPSI is reflective of a loss of diesel generator. Losing a diesel generator results in the availability of one LPSI.pump (2 of 4 control valves) and one HPSI pump (4 control valves). However, the containment spray pumps and air coolers, which

EMF-91-1n Page 2 are also lost with the loss of a diesel generator, were conservatively assumed to be operational since lower containment pressures and higher peak clad temperatures result.

The changes supported by this analysis do not affect the limiting ~reak size identified by SNP's LOCA methodology since the changes listed above will not effect the system blowdown.

Therefore, the analysis used a 0.6 double-ended cold leg guillotine break (DECLG) at the pump discharge, which was previously identified to be the limiting break size(1). The analysis includes calculations at both a BOC axial power shape peaked at a relative core height of 0.6 and an EOC axial power shape peaked-at a relative core height of 0.8. The calculations conservatively used the maximum fuel stored energy near BOC where maximum densification occurs. Justification is provic:fed to support operation with SNP fuel up to an assembly average exposure of 52,500 MWd/MTU with regard to the large break LOCA.

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SUMMARY

OF RESULTS EMF-91-177 Page 3 The analysis demonstrates that the 1 o CFR 50.46(b) criteria are satisfied for the Palisades plant with the axially dependent power peaking limit curve shown in Figure 2.1. The analysis supports a maximum LHR of 15.28 kW/ft up to a relative core height of 0.6 and an LHR of 14.75 kW/ft at a relative core height of 0.8. A total radial peaking factor of 2.04 and a maximum average steam generator tube plugging level of 29.3% with up to 4.5% asymmetry are supported.

Results of the analysis for both the BOC and EOC axial profiles at the limiting 0.6 DECLG break size are shown in Table 2.1. The peak cladding temperature was calculated_ to be 1926.5 °F for the BOC profile and 2110.6 °F for the EOC profile. The analysis supports Cycle 1 o operation and is intended to support operation for future cycles.

EMF-91-177 Page4 Table 2.1 Summary of Results for 0.6 DECLG Limiting Break Size BOC Stored Energy BOC Stored Energy BOC Axial Shape EOC Axial Shape

~L = 0.6}

{X/L = 0.8}

Peak LHR (kW/ft) 15.28 14.75 j

Hot Rod Burst I

Time (sec}

46.00 47.20 Elevation (ft) 7.45 8.95 I

Channel Blockage 0.30 0.32 Fraction Peak Cladding Temperature J

Temperature (°F) 1926.5 2110.6

J Time (sec) 58.20 64.20 Elevation (ft) 7.45 8.95

,l Metal-Water Reaction Local Maximum (%)

2.30 4.25 Elevation of Local 7.45.

8.95 Maximum (ft)

Hot Pin Total (%)

0.36 0.43 Core Maximum (%)

< 1.0

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0.8 EMF-91-177 Page 5 1.0 FIGURE 2.1 ALLOWABLE LHR AS A FUNCTION OF.PEAK POWER LOCATION

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3.0 ANALYSIS Section 3.1 of this report provides a description of the postulated large break loss-of-coolant transient. Section 3.2 describes the methodology and major assumptions used in the analysis. Section 3.3 provides a description of the Palisades plant and a summary of the system parameters used in the analysis. Section 3.4 summarizes the results of the limiting EOC axial power shape.

3.1 Description of LBLOCA Transient A loss-of-coolant accident (LOCA) is defined as the rupture of the reactor coolant system primary piping up to and including a double-ended guillotine break. The limiting break occurs on the pump discharge side of a cold leg pipe. The LOCA is assumed to result from an earthquake and is co-incident with loss-of-offsite power. Primary coolant pump coastdown occurs co-incident with the loss-of-offsite power. Following the break, depressurization. of the reactor coolant system, including the pressurizer, occurs. A reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached. Reactor trip and scram are cc;mservatively neglected in the LOCA analysis. Early in the blowdown, the reactor core experiences flow reversal and stagnation which causes the fuel rods to pass through critical heat flux (CHF).

Followjng CHF, the fuel rods dissipate heat through the transition and film boiling modes of heat transfer. Rewet is precluded during blowdown by Appendix K of 1 o CFR so.

A Safety Injection System (SIS) signal is actuated when the appropriate setpoint (high containment pressure) is reached" Due to loss-of-offsite power, a time delay for startup of diesel generators and SIS pumps is assumed. Once the time delay criteria is met and the system pressure falls below the shutoff heads of the HPSI and LPSI pumps, SIS flow is injected into the cold legs. Single failure criteria is met by assuming that one LPSI pump is not available for operation<6>. The LPSI flow is assumed to be evenly split between one intact loop and the broken loop. Also, only one HPSI pump was assumed operable for this analysis, the flow of which was split evenly between the three intact loops and one broken loop. The SIS flow curves are depicted in Table 3.1 <4>. When the system pressure f~lls below the Safety lnj~ction Tank (SIT) l

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EMF-91-177 Page 7 pressure, flow from the SITs is injected into the cold legs. Flow from the Emergency Core Cooling System (ECCS) is assumed to bypass the core and flow to the break until the end-of-bypass (EOBY) is predicted to occur (sustained downflow in the downcomer). Following EOBY, ECCS flow fills the lower downcomer and lower plenum until the liquid level reaches the bottom of the core (beginning-of-core-recovery or BOCREC time).

During the refill period, heat is transferred from the fuel rods by radiation heat transfer.

The reflood period begins at BOCREC time. ECCS fluid fills the downcomer and provides the driving head to move coolant through the core. As the mixture level moves up the core, steam is generated. Steam binding occurs as the steam flows through the intact and broken loop steam generators and pumps. The pumps are assumed to have a locked rotor (per Appendix K of 1 o CFR 50) which tends to reduce the reflood rate. The fuel rods are eventually cooled and quenched by radiation and convective heat transfer as the quench front moves up the core. The reflood heat transfer rate is predicted through experimentally determined heat transfer and carry-over rate fraction correlations.

The purpose of the LBLOCA analysis is to demonstrate that the criteria stated in 1 o CFR 50.46{b) are met. The criteria are:

1)

The calculated peak fuel element cladding temperature does not exceed the 2200 °F limit.

2)

The amount of fuel element cladding which reacts chemically with water or steam does not exceed 1 % of the total amount of zircaloy in the core.

3)

The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching.

4).

The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

3.2 Description of Analytical Models EMF-91-177 Page 8 The SNP EXEM/PWR evaluation mode1<2> was used to perform the analysis.

This evaluation model consists of the following computer codes:

1)

RODEX2(3) for computation of initial fuel stored energy, fission gas release and gap conductance;

2)

3)

4)

RELAP4-EM for the system and hot channel blowdown calculations; CONTEMPT/LT-22 as modified in accordance with NRC Branch Technical Position CSB 6-1 for computation of containment back pressure; REFLEX for computation of system reflood; and

5)

TOODEE2 for the calculation of fuel rod heatup during the refill and reflood portions of the LOCA transient.

The quench time, quench velocity, and carryover rate fraction (CRF) correlations in REFLEX, and the heat transfer correlations in TOODEE2 are based on SNP's Fuel Cooling Test Facility (FCTF) data.

The governing conservation equations for mass, energy, and momentum transfer are used along with appropriate correlations consistent with Appendix K of 1 O CFR 50. The reactor core in RELAP4 is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback, and with actinide and decay heating as required by Appendix K. Appropriate conservatism specified by Appendix K of 1 O CFR 50 is incorporated in all of the EXEM/PWR models.

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3.3 Plant Description and Summary of Analysis Parameters EMF-91-177 Page 9 The Palisades plant is a Combustion Engineering (CE) designed pressurized water reactOr which has two hot leg pipes, two U-tube steam generators, and four cold leg pipes with one recirculation pump in each cold leg. The plant utilizes a large dry containment. The reactor coolant system was nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow paths or "junctions". The two cold legs connected to the intact loop steam generator were assumed to be symmetric and were modeled as one intact cold leg with appropriately scaled input. The model considers four SITs, a pressurizer, and two steam generators with both primary and secondary sides of the steam generators modeled. The high pressure safety injection (HPSI) and low pressure safety injection (LPSI) pumps were modeled as fill junctions at the SIT lines, with conservative flow rates given as a function of system back-pressure. The primary pump performance curves are characteristic of CE pumps. The reactor core was modeled radially with an average core and a hot assembly as parallel flow channels, each with three axial nodes. A steam generator tube plugging level of 29.3% was assumed with an asymmetric steam generator tube plugging of 4.5%. The break was conservatively assumed to have occurred in the most highly plugged loop since this results in more steam binding during reflood and a higher peak cladding temperature. Values for system parameters used in the analysis are given in Table 3.2. The values in Table 3.2, together with Reference 5, represent a comprehensive summary of analysis inputs.

3.4 Axial Shape Study Results An EOC (top-skewed) axial power shape was analyzed to define the axially dependent LHR limit curve shown in Figure 2.1. The axial power shape was peaked at a relative core height of 0.8 with an LHR of 14.75 kW/ft. The axial shape was selected from those axial shapes allowed by the Tinlet LCO barn. A BOC fuel stored energy was conservatively used in conjunction with this axial shape. The results for the EOC shape are shown in Table 2.1. The PCT was calculated to be 2110.6 °F. The effect of reduced pellet-to-clad gap (reduced stored energy) more than offsets the effects of radial peaking, reduced ECCS flow and minimum SIT level. The calculated event times for the 0.6 DECLG break size are given in Table 3.3. Plots of parameters depicting

EMF-91-1n

Page 10 calculations for the limiting 0.6 DECLG break and the EOC shape are shown in Figures 3.1 through 3.26.

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EMF-91-177 Page 11 Table 3.1 Revised HPSI and LPSI Pump Flow Curves HPSI Flow LPSI Flow Primary Coolant System (One Pump)

(One Pump-2 Valve)

Injection Point Pressure (psia}

(gpm}

1237.7 0.0 1173.7 100.0 1068.7 200.0 922.7 300.0 561.7 400.0 200.0 0.0 197.7 1008.0 188.7 1257.0 177.7 1509.0 164.7 1765.0 159.0 500.0 146.7 2069.0 127.7 2344.0 94.7 2732.0 58.7 2960.0 14.7 535.0 3202.0

EMF-91-177 Page 12 Table 3.2 Palisades System Analysis Parameters Primary Heat Output, MWt 253o(a)

Primary Coolant Flow Rate, lbm/hr 1.203 x 108 (318, 770 gpm)

Primary Coolant System Volume, ft3 Operating Pressure, psia Inlet Coolant Temperature (hottest loop), °F Reactor Vessel Volume, ft3 Pressurizer Total Volume, ft3 Pressurizer Liquid Total, ft3 SIT Total Volume, ft3 (one of four)

SIT Liquid Volume, ft3 SIT Pressure, psia SIT Fluid Temperature, °F Total Number of Tubes per Steam Generator Steam Generator Tube Plugging Number of Tubes Plugged 33.8 % SGTP.

24.8 % SGTP Steam Generator Secondary Side Heat Transfer Area, tt2 33.8% SGTP 24.8% SGTP 8800Cb,c) 2060 544 4782 1504 803 2011 1040 215 90 9519Cc) 33.8 %(c) 24.8 %(c) 2878(c) 2114(c) 45 942(c) 51 :931 (c)

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Steam Generator Secondary Flow Rate, lbm/hr (47-53% power split) 24.8 % SGTP 33.8 % SGTP Steam Generator Secondary Pressure, psia Steam Generator Feedwater Enthalpy, Btu/lbm Reactor Coolant Pump Rated Head, ft Reactor Coolant Pump Rated Torque, ft-lbf Reactor Coolant Pump Rated Speed, rpm Reactor Coolant Pump Moment of Inertia, lbm-tt2

. Cont.ainment Volume, tt3 Containment Temperature, °F SIS Fluid Temperature, °F HPSI Delay Time, seconds LPSI Delay Time, seconds Containment fan coolers initiation time, seconds Containment sprays initiation time, seconds 5.949 x 1 as (24.8 % SGTP)(c) 5.241 x 1 as (33.8 % SGTP)(c) 73a(c) 414 26a 32,53a 88a 98,aaa 1.64 x 1as 9a 7a 27.a 28.a a.a a.a

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Primary Heat Output used in RELAP4-EM Model - 1.a2 x 253a = 258a.6 MWt.

b.

Includes pressurizer total volume and 29.3% average SGTP.

c.

Data for original steam generators.

Table 3.3 Calculated Event Times for 0.6 DECLG Break Event Transient.Initiation Break is Fully Open Safety Injection Signal SIT Injection Begins, Broken Loop Pressurizer Empties SIT Injection Begins, Single Intact Loop SIT Injection Begins, Double Intact Loop End-of-Bypass (EOBY)

Start of Reflood (BOCREC)

Peak Cladding Temperature (X/L = 0.6)

SIT Empties, Single Intact Loop SIT Empties, Double Intact Loop Peak Cladding Temperature (X/L = 0.8)

SIT Empties, Broken Loop EMF-91-1n

Page 14 Time (seconds) 0.00 0.05 0.65 11.65 12.26 15.65 15.65 22.60 41.65

. 58.20 61.60 63.70 64.20 64.70

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4.0 CONCLUSION

S EMF-91-177 Page 41 The analysis supports operation of the Palisades plant at a power level of 2530 MWt, a modification of the LPSI flow curves, the loss of a HPSI pump, and an average steam generator tube plugging level of 29.3% with a maximum asymmetry of 4.5%. The analysis supports a peak LHR of 15.28 kW/ft with the axially dependent power peaking limit shown in Figure 2.1 and radial peaking limits of 2.04 (FrT) and 1.76 (FrA). The analysis supports Cycle 10 operation and is intended to support operation for future cycles.

Operation of the Palisades plant with SNP 15 x 15 fuel at or below the LHR limits shown in Figure 2.1 assures that the NRC 1 O CFR 50.46(b) acceptance criteria for LOCA pipe breaks up to and including the double-ended severance of a reactor coolant pipe will be met with the emergency core cooling system for the Palisades plant.

I I,.

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I 5.0

1.

2.

REFERENCES EMF-91-1n Page 42 Palisades Large Break LOCNECCS Analysis with Increased Radial Peaking, ANF-88-107, Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, February 1990.

Dennis M. Crutchfield (USNRC Asst. Director division of PWR Licensing-B) "Safety Evaluation of Exxon Nuclear Com'pany's Large Break ECCS Evaluation Model EXEM/PWR and Acceptance for Referencing of Related Licensing Topical Reports", dated July 8, 1986.

3.

RODEX2: Fuel Rod Thermal Mechanical Response Evaluation Model, XN-NF-81-58(P)(A),

Revision 2, Supplements 1 and 2, March 1984, Supplements 3 and 4, June 1990, Advanced Nuclear Fuels Corporation, Richland, WA 99352.

4.

CPCo Letter, G. C. Packard to H. G. Shaw, "Large Break LOCA Analysis with Reduced HPSI and LPSI Flow", GCP90*020, dated October 11, 1990 (see CPCo EA-A-PAL-90-049).

5.

6.

Palisades Principal Parameters, EMF-90-084(P), Revision 2, Siemens Nuclear Power Corporation, Richland, WA 99352.

Calculative Methods for the CE Large Break LOCA Evaluation Model, Volume 1, CENPD-132P, August 1974.

I I

r*

I PALISADES LARGE BREAK LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAKING AND REDUCED ECCS FLOW Distribution TH Chen SE Cole RA Copeland RC Gottula JS Holm JW Hulsman TR Lindquist JN Morgan KC Segard HG Shaw/Customer (1 /20)

BD Stitt CJ Volmer RT Welzbacker Document Control (5)

EMF-91-177 Issue Date:

10/01 /91

ATTACHMENT 5 Consumers Power Company Palisades Plant Docket 50-255 CYCLE 10 TECHNICAL SPECIFICATIONS CHANGE REQUEST CONSUMERS POWER COMPANY ENGINEERING ANALYSIS EFFECT OF INCREASED SIRW TANK BORON CONCENTRATION ON LONG TERM COOLING (EA-PAH-91-04)

November 1, 1991

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l'lllMJaSI PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Sheet

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Title EFFECT OF INCREASED SIRW TANK BORON CONCENTRATION LIMIT ON LONG TERM COOLING INITIATION AND REVIEW Initiated Initiator Review Method Check CJ>

Technically Reviewed Reviewer Rev Appd Alt Det Qual Appd By Date Description By Date By Cale Rvw Test By O

Original Issue

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u PURPOSE:

To determine the effect of raising the Safety Injection Tanks (SITs) and, SIRW Tank boron concentration limit, from 2000 ppm to 2500 ppm, on the post-LOCA long term cooling analysis.

PROCEDURE UTILIZED:

Administrative procedure 9.11

SUMMARY

OF RESULTS:

The current long term cooling analysis was evaluated to determine the available margin in PCS boric acid concentration before precipitation would occur for the large and small break LOCAs.

The increased SIT and SIRW tank boron concentration and any changed plant parameters were evaluated against the available margin in the long term cooling analysis, considering the conservatism incorporated, to determine the effect on the margin.

It was determined that the increase in SIT and SIRW Tank boron concentration is insignificant compared to the available margin and consevatism employed.

It is therefore concluded that increasing the SIT and SIRW tank boron concentration limit to 2500 ppm will not change the conclusions of the current long term cooling analysis.

SPECIAL MEDIA ATTACHED (DRAWINGS, MICROFICHE, ETC) x NO YES - List of Attachments included

1.0 2.0 3.0 4.0 5.0 5.1 5.2 6.0 7.0 8.0 PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET Table of Contents EA-PAH-91-04 Rev. 1 Sheet _2_ of 11 Objective..................................................................................................................................................... 3 References.................................................................................................................................................. 3 Background................................................................................................................................................ 3 Analysis Input............................................................................................................................................ 4 Analysis....................................................................................................................................................... 4 I..arge Break LOCA................................................................................................................................ 6 Small Break LOCA................................................................................................................................. 8 Summary......................................................... :........................................................................................... 10 Conclusion............................................. ~.................................................................................................... 11 List of Attachments.................................................................................................................................. 12

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet ~3~ of 11 EFFECT OF INCREASED SIRW TANK BORON CONCENTRATION LIMIT ON.LONG TERM COOLING 1.0 OBJECTIVE The objective of this Engineering Analysis is to determine the effect of raising the Safety Injection Tanks (SITs) and Safety Injection and Refueling Water (SIRW) Tank boron concentration limit, from 2000 ppm to 2500 ppm, on the post-LOCA long term cooling analysis. This analysis will aid in the justification for Technical Specification changes to increase these limits.

2.0 REFERENCES

2.1 Palisades Plant Technical Specifications, T.S. 3.3.

.I 2.2 P-CE-5627, "Palisades Long Term Cooling Performance Evaluation,". Combustion Engineering Report dated June 5, 1981.

2.3 DBD-2.02, Palisades Design Basis Document - "High Pressure Safety Injection System," section 3.3.2. Rev. 0 - April 1990.

2.4 P-PEC-170 Rev. 0, "Head Losses and Flow Requirements for the Hot Leg Injection Line for Palisades," Combustion Engineering Calculation dated January 30, 1979.

2.5 EOP 4.0 Rev. 2, "Loss of Coolant Accident Recovery," Palisades Plant Emergency Operating Procedure. July 20, 1990.

2.6 "Thermal Hydraulic Analysis of Pressurizer PORV Relief System," Final Report, Rev. 0. NUS Corporation, November 2, 1990.

3.0 BACKGROUND

Before the boron concentration in the SITs and SIRW Tank can be increased, the effect of the added boron on post-LOCA long term cooling (LTC) must be addressed. The current long term cooling analysis for Palisades [Ref. 2.2] was performed by Combustion Engineering to demonstrate acceptable core cooling and that boric acid precipitation would be avoided. Reference 2.2 concluded that between

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet ~4~ of 11 51h and 6V2 hours post-LOCA the operators should detenirine whether to take mitigating measures for a large break (by initiating hot leg injection) or small break (by initiating shutdown cooling), based on PCS pressure. This time requirement, however, was not based on boric acid precipitation. According to the Design Basis Document on high pressure safety injection (HPSI) [Ref. 2.3], the 61h hour time requirement is based on the time at which the steam generators (SGs) could exhaust feedwater supply from the condensate storage tank and can no longer function as a source of heat removal. For small breaks, shutdown cooling (SDC) should be initiated when the SGs are no longer available since the break would not be large enough for once through cooling. To avoid operator confusion this was also used to establish the time by which simultaneous hot leg and cold leg injection should have been started for a large break. Also according to Reference 2.3, the 51h hour time requirement was conservatively chosen because high steam flows in the PCS could cause entrainment of the normal flow of coolant if LTC were initiated before four hours.

Since the time requirement for appropriate operator actions to mitigate either a large or small break LOCA is not based on the time at which boric acid begins precipitate in the core, only the margin for the peak predicted PCS boric acid concentrations needs to be evaluated. In the evaluation, all other plap.t parameters or conditions that have changed which could affect the LTC analysis will be addressed.

4.0 ANALYSIS INPUT 4.1 The maximum boric acid storage tank concentration of 12 wt % and the maximum safety injection tank level of 200" are from the plant Technical Specifications [Ref.. 2.1].

4.2 Parameters used in the current long term cooling analysis are from Reference 2.2.

4.3 The time at which charging suction is realigned from the boric acid storage tank to the safety injection and refueling water tank, 30 minutes, is from Reference 2.5 [step 23].

4.4 The effective throat areas of the old PORVs, 1.5 in.2, and the new PORVs, 5.774 in.2, are from Reference 2.6.

S.O ANALYSIS To evaluate the effect of raising the boron concentration limit of the SITs and SIR W tank on the current LTC analysis, the first step is to identify the parameters used in that analysis and compare them to the parameters as proposed with the increased SIT and SIR W tank boron concentration. Table 1 below lists the sources and concentrations of boric acid used in Reference 2.2 [Ref. 2.2., Table 1].

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet ~5~ of 11 Table 1 Boric Acid Source Parameters From P-CE-5627 Source PCS SIRW Tank SITs BAST Boric Acid wt%

1.02 1.13 1.13 12.0 Liquid Mass(lb.J 445,588 2,489,684 269,077 109,101 The *values listed in Table 1 must be compared to the current and proposed Technical Specifications to *determine which values need to be changed. From Technical Specification 3.2, the boric acid storage tank (BAST) maintains 10 wt% boric acid. From the basis for Tech Spec 3.2, the system is capable of handling 12 wt % boric acid, so 12 wt % was used in Reference 2.2, which is very conservative. To convert boron concentration to weight percent boron, the following equation can be used [Ref. 2.4]:

boric acid wt%= ppm boron 1748 (1)

The proposed increased SIT and SIR W tank boron concentration is 2500 ppm, which corresponds to 1.43 wt % boric acid using Equation (1).

The PCS boric acid concentration listed in Table 1 corresponds to approximately 1783 ppm boron using Equation ( 1 ). This value is sufficiently high enough to bound refueling boron concentrations, and therefore considered conservative. Therefore the only change in boric acid concentrations from the value used in the current LTC analysis is that for the SITs and the SIR W tank to account for the proposed increase.

It should also be noted that the maximum liquid inventory of the SITs used in current LTC analysis is no longer correct. The LTC analysis used 269,077 lbm as the maximum inventory from all four SITs, which was based on the SIT maximum level at the time, 198" or 1166 ft3. Since then, the SIT level limit has been raised to 200" corresponding to 1176 ft3 [Ref. 2.1]. Using the same density as was used to obtain 269,077 lbm, results in a current maximum liquid inventory of the four SITs of 271,385 lbm. This increased liquid inventory of the SITs will affect the total amount of boron available. However, since the proposed boric acid concentration of this increased available liquid, 1.43 wt %, is still low relative to the BAST, adding the additional volume to the PCS would actually slightly decrease the boric acid concentration of the PCS.

The sources of boric acid, with the maximum concentrations and liquid inventories as currently allowed or proposed (exceptfor PCS, which is taken as a minimum inventory), are shown in Table 2 below.

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET Table 2 EA-PAH-91-04 Rev. 1 Sheet _6_ of 11 Current and Proposed Parameters for Boric Acid Sources Source PCS SIRWTank SITs BAST Boric Acid wt%

1.02 1.43 1.43 10.0 Liquid Mass(lb.J 445,588

. 2,489,684 271,385 109,101 To get a comparison of the total available boric acid concentration of the PCS in Reference 2.2 and as proposed, without considering boiloff, the sum of the weight percentages multiplied by the respective liquid inventory of each source must be divided by the total liquid inventory from all of the sources.

Performing this for the values in Table 1 and Table 2 results in 1.473 wt % and 1.657 wt % for that used in the current LTC analysis and that proposed, respectively. The difference is less than 0.2 %,

which is only a slight total increase. However, this is the total available boric acid concentration if all sources were mixed together at once and does not represent an appropriate means of comparison. The times at which the different sources of boric acid are being added to the PCS and the amount of core boiloff that is occurring have the most significant effects.

In the current LTC analysis [Ref. 2.2], it is assumed that charging suction is not realigned from the BAST to the SIRW Tank until 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months post-LOCA Therefore, concentrated boric acid, at 12 wt%

as used in the current LTC analysis but 10 wt % in reality, is being delivered to the PCS via two charging pumps during that time. At the same time, all safety injection water that is not exiting through the cold leg break is being delivered from the SIR W Tank and has been delivered from the SITs to the PCS. However, the SITs and the SIRW Tank are at a much lower boric acid concentration than the BAST, even with the proposed Technical Specification change to increase it by 0.3 wt %, to 2500 ppm boron. The boric acid concentration of the water being delivered by charging is therefore going to be dominating the PCS boric acid concentration since boiloff in the core is also occurring which increases the boric acid concentration. The extent of the effects depend upon the break size, PCS pressure, and amount of safety injection circulating through the core. However, the assumption of concentrated boric acid injection for 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is conservative since the emergency operating procedures [Ref. 2.5, pg 7]

direct the operators to realign charging pump suction to the SIR W Tank after 30 minutes into the event.

If injection of concentrated boric acid from the BAST is disco:ritinued after 30 minutes, a much lower PCS boric acid concentration would result.

5.1 LARGE BREAK LOCA For the case of a large break LOCAs, Figure 3 of Reference 2.2 must be examined. This figure shows the PCS boric acid concentration versus time for three cases following the largest possible cold leg break: no core flushing, a hot leg flushing flow of 5 gpm, and flushing flow provided from 50/50 hot

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet 7

of 11 and cold leg injection. From Figure 3 of Reference 2.2 can be seen that the precipitation limit of boric acid will be 32 wt %, for the saturation temperature at 20 psia. The PCS boric acid concentration increases fairly sharply during the first 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ,*while concentrated boric acid from the BAST is being delivered via charging. After the first 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , charging suction is taken from the SIR W Tanlc. After this, the PCS boric acid concentration increases as core boiloff continues, but the increase is at a slower rate since the PCS boric acid concentration is well above 10 wt % and water reaching the core from the SIRW Tanlc or the SITs would dilute this. For the case of no flushing flow, the PCS boric acid

. concentration continues to increase due to boiloff, with Reference 2.2 predicting boric acid precipitation to begin after 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months post-LOCA For the large break case with a hot leg flushing flow of 5 gpm started at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months post-LOCA, the PCS boric acid concentration levels off after the flushing flow begins, and Reference 2.2 predicted a maximum boric acid concentration of approximately 22 wt %. Increasing the boric acid concentration of the SITs and the SIRW tanlc by 0.3 wt % could increase the maximum boric acid concentration of the PCS, but the increase would be small since the boric acid concentration of the water being injected

-would still be much lower than that of the PCS. With the precipitation limit of boric acid being 32 wt

%, increasing the SITs and the SIRW Tank boric acid concentration by 0.. 3 wt % would not significantly decrease the margin for this case.

If 50/50 hot and cold leg injection flow is used, as is directed by plant procedures, Reference 2.2 predicted the maximum PCS boric acid concentration to be less than 20 wt % prior to initiation of the hot leg injection.

After the start of the 50/50 hot and cold leg injection, the Pcs* boric acid concentration decreases sharply, with the concentration leveling off close the value of the water being injected (sump water if during recirculation) as core boiloff decreases and flushing flow consequently increases. Increasing the SITs and the SIRW Tank boric acid concentration by 0.3 wt% would not significantly increase the peak PCS concentration for this case, leaving ample margin before the precipitation limit would be reached.

To cover the range of large break LOCAs, Figure 5 of Reference 2.2 shows the PCS boric acid concentration versus time for the smallest cold leg break for which PCS pressure remains low enough such that the HPSI pump can cool and flush the core. Reference 2.2 predicted the PCS boric acid concentration to peak at approximately 17 wt % for this smallest range of large breaks when using 50/50 hot and cold leg injection. However, it can be seen that the PCS boric acid concentration increases at a much faster rate during the first 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months than for the largest break. The faster rate of increase of the PCS boric acid concentration for the smaller break occurs since the PCS pressure is higher causing lower HPSI flow, with a smaller amount of safety injection water reaching the core.

Thus, concentrated boric acid from the BAST injected via charging will have a smaller amount of dilution in the core since a smaller amount of SIRW Tank water with a relatively low boric acid concentration will reach the core. This is also complicated by more boiloff and voiding in the core which increases the boric acid concentration. However, once hot leg injection is initiated, the PCS boric

. acid concentration decreases at a fairly steady rate. Having greater margin before the boric acid

precipitation limit is reached than the two previously mentioned large break LOCA cases, increasing I the SIT and SIR W Tank boric acid concentration would not significantly decrease the margin for this case either.

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet ~8~ of 11 For a large break, the minimum margin for post-LOCA boric acid concentration in the PCS, 10 wt%,

occurs with minimum flushing flow of 5 gpm initiated at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for the largest cold leg break in the current LTC analysis. However, 50/50 hot and cold leg injection is used by plant procedures, which has a margin of 12 wt % boric acid in the current LTC analysis before precipitation would occur. An increase of 0.3 wt % in the SITs and the SIR W Tank boric acid concentration would increase the maximum PCS boric acid concentration by a small amount as boiloff occurs, but there would not be a significant increase since the major boric acid contributor is the conservative assumption of charging from the BAST for 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The SIT and the SIR W Tank boric acid concentrations will still be relatively low compared to that from the BAST and in the PCS so as to dilute the PCS boric acid concentration once flushing flow begins. Thus, with the amount of available margin, increasing the SIT and SIRW Tank boric acid concentrations by 0.3 wt % will not affect the conclusions of the LTC analysis for the large break LOCA, and initiation of simultaneous hot and cold leg injection between 51h and 6~ hours post-LOCA is still more than adequate for prevention of boric acid precipitation after a large break LOCA S.2 SMALL BREAK LOCA The plant emergency operating procedures direct the operators to initiate hot leg injection regardless of the break size of the LOCA However, if auxiliary feedwater is lost, shutdown cooling must be initiated for a small break since once-through-cooling through the break location as is done for a large break LOCA would not be feasible for this size of break. Core cooling and boric acid dispersion can be accomplished by doing so. A third alternative for small breaks is considered, though, since the shutdown cooling system is not single-failure proof. By opening the PORVs and continuing HPSI flow to the cold legs, a once-through-cooling path can be created and the core can be cooled and boric acid

flushing accomplished. Reference 2.2 evaluated these alternatives for a range of small break LOCA sizes. For the temperature predicted for small break cases, the solubility limit of boric acid is 47.5 wt

% [Ref. 2.2, Figures 9 & 10].

For the case of shutdown cooling initiation at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , Reference 2.2 predicted a maximum boric acid concentration of 17.5 wt %.

As can be seen on Figure 9 of Reference 2.2, the PCS boric acid concentration increases rapidly during the first 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , as was the case for the smallest large break LOCA The rapid increase in PCS boric acid concentration occurs since charging is delivering concentrated boric acid to the PCS, while the higher PCS pressures limits HPSI flow, which would dilute the PCS boric acid concentration. After 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , the PCS boric acid concentration starts to level off as charging suction is realigned with the SIR W Tank and safety injection reaching the core starts to match the boiloff. From Figure 8 of Reference 2.2, the PCS pressure prohibits release of the SITs until around 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , at which time the PCS boric acid concentration is over 15 wt% from Figure 9. Thus the water from the SITs will dilute the boric acid concentration even with the proposed increase. The minimum margin before boric acid precipitation occurs is 30 wt % for this case with the worst small break. An increase of 0.3 wt % boric acid in the SITs and the SIR W Tank would have no significant effect on this margin.

For the case of a small break with auxiliary feedwater no longer available and shutdown cooling rendered inoperable, two PORVs are opened. Figure 10 of Reference 2.2 shows the PCS boric acid concentration versus time for a range of small break sizes in which the PORVs are opened to cool and

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet 9

of 11 flush the core. It can be seen from that figure, that the largest break for this case results in the highest PCS boric acid concentration, of 19 wt %. This leaves a margin of more than 28 wt % before boric acid precipitation will begin, and increasing the SIT and SIRW Tank boric acid concentration by 0.3 wt%

would not significantly decrease this margin either. However, Reference 2.2 analyzed this case with the PORVs having an effective throat area of 1.5 irf [Ref. 2.6, pg. 1-1], but since then, new PORVs having an effective throat area of 5.774 irf [Ref. 2.6, pg. 1-2] have been installed.

To evaluate the effect of the larger effective throat area of the PORVs, the characteristics of plot of the worst case small break in Figure 10 of Reference 2.2 must be explained, with the aid of the PCS pressure profile in Figure 8 of that reference. *As with the LOCA cases discussed previously, during the first 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months the PCS boric acid concentration increases rapidly while concentrated boric acid from the BAST is being injected via charging. After 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , the concentration increases at a much slower rate as HPSI and charging flow provide dilution and boiloff continues to occur. Around 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , the PCS boric acid concentration starts to slowly decrease as the reactor refills for the smaller breaks. It can be seen from Figure 8 of Reference 2.2 that safety injection flow is greater than the break flow rate for breaks smaller than 0.018 ft2 since the PCS refills after 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . As the PCS refills it repressurizes causing the HPSI flow rate to be reduced, which limits core cooling and flushing. At about 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , the PORVs are opened and a large portion of the water in the core flashes and exits through the PORVs~ and the core water level drops. This allows a large slug of cooler safety injection water at a low boric acid concentration to enter the core which causes a sharp decrease in the PCS boric acid concentration. For the larger range of small breaks, the PCS boric acid concentration starts to increase again. For the 0.018 ft2 break, the rate of increase in the PCS boric acid concentration is very similar to that of a large break LOCA without hot leg injection (Figure 3 of Reference 2.2), since opening the PORVs has in a sense created a large break. For the larger breaks, the amount of safety injection flow reaching the core and not exiting through the break is determined by the available hydraulic head in the reactor vessel downcomer. As water in the core boils off and core exit pressure decreases due to the decreasing decay heat level, it allows water in the downcomer to circulate to the bottom of the core.

The boric acid concentration continues to increase however, as this water boils off. Once decay heat decreases enough to significantly lower the boiloff, the boric acid concentration stops increasing.

With larger PORVs, the initial decrease in PCS boric acid concentration could be expected to be more rapid, since more mass would be released during a shorter period of time allowing a faster surge of safety injection water into the core. For the larger range of small breaks, the increase in PCS boric acid concentration, after 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months on Figure 10 of Reference 2.2, would be expected to be at the same or a lower rate with the larger PORVs. Opening the larger PORVs would decrease the pressure in the reactor vessel upper plenum allowing the same amount or more safety injection water in the downcomer to reach the core. This would result in the at least the same amount of dilution in the core. The margin in Reference 2.2 before boric acid would begin to precipitate is more than 28 wt %. With the larger PORVs not affecting or decreasing the PCS boric acid concentration, an increase of 0.3 wt% in the SIT and the SIR W Tank boric acid concentration would not significantly affect the 28 wt % margin.

This case is also a last resort that requires multiple failures since shutdown cooling is assumed to be inoperable and auxiliary feedwater supply is assumed to be exhausted with backup supplies unavailable.

For the worst small break LOCA in which shutdown cooling is initiated, the minimum predicted margin in PCS boric acid concentration before precipitation begins was 30 wt % in the current LTC analysis.

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet 10 of 11 Therefore, more margin in boric acid concentration is available for a small break LOCA than for a large break, and increasing the SITs and the SIR W tank boron concentration will not significantly affect the margin. For the worst small break in which the PORVs are opened for core cooling and flushing, the margin is more than 28 wt %. Larger PORVs have been installed since Reference 2.2 was completed, which were determined to have no significant affect on the margin. Also, for use of the PORVs, multiple failures would have to be assumed as it is a last resort. With as large of a margin as is available, the increased boron concentration in the SITs and the SIRW tank will not significantly affect the margin in the LTC analysis for a small break LOCA 6.0

SUMMARY

The current post-LOCA long term cooling analysis on record was examined to determine which parameters used in that analysis have changed or will change with the proposed increase in SIT and SIR W tank boron concentrations. Several plant parameters were identified to be changed. The SIT liquid inventory used in the LTC analysis was based on a maximum SIT level of 198" which has since been increased to 200", corresponding to a total liquid inventory increase of approximately 2308 lbm for the four SITs. The boron concentration limit of the SITs and the SIR W tank has been proposed for an increase from 2000 ppm to 2500 ppm, corresponding to approximately 1.43 wt % boric acid. The boric acid concentration of the BAST is 10 wt % by Technical Specifications, whereas 12 wt % was used in the current LTC analysis. Another parameter is the installation of new PORVs that occurred since the current LTC analysis was performed. The new PORVs have a much larger effective throat area.

The available margin in boric acid concentration from the LTC analysis for both large break and small break LOCAs was evaluated for the effect from increasing the boric acid concentration of the SITs and the SIRW tank by 0.3 wt %. The minimum margin in the current LTC analysis before boric acid precipitation would occur, 10 wt %, was for the largest break using a minimum hot leg flushing flow of 5 gpm. However, plant procedures direct the use of 50/50 hot and cold leg injection which resulted in a margin of more than 12 wt % boric acid in the current LTC analysis. The available margin for both of those cases is ample for increasing the boric acid concentration of the SITs and the SIR W tank by 0.3 wt%.

For the worst small break in which shutdown cooling is initiated by 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , the available margin before boric acid precipitation occurs is 30 wt % in the current LTC analysis. This margin is much greater than that for the large break case. For the worst small break in which the PORVs must be opened, the available margin is 28 wt% in the current LTC analysis. The installation of larger PORVs since Reference 2.2 was completed would either not affect or decrease the PCS boric acid concentration for that case resulting in at least the same amount of margin. The case of using the PORVs is conservative and requires the assumption of multiple failures since shutdown cooling and auxiliary feedwater would not be available. Also, additional margin could be acquired if the conservative assumption of concentrated boric acid injection for 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months were changed to 30 minutes as is directed by plant procedures, and the boric acid concentration of the BAST is decreased to the Tech Spec value of 10 wt%.

7.0 CONCLUSION

S PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS WORK SHEET EA-PAH-91-04 Rev. 1 Sheet 11 of 11 For the worst case large break, a minimum margin of 10 wt % boric acid exits in the current post-LOCA long term cooling analysis before boric acid precipitation would begin. For the worst case small break, the margin is almost three times that for a large break. Larger PORVs that have been installed result in at least the same amount of margin for the case of a small break in which shutdown cooling is inoperable and auxiliary feedwater is unavailable. Also, use of the PORVs requires the assumption of multiple failures which is conservative. Regardless of the break size, operators are directed by procedures to realign charging suction from the BAST to the SIR W Tank after 30 minutes, whereas the LTC analysis assumed that realignment did not occur for 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Additional margin would also be

I added to the analysis if that conservative assumption were removed and the boric acid concentration I of the BAST were decreased from 12 wt % to the Tech Spec value of 10 wt %.

The effect of increasing the boron concentration limit of the SITs and the SIR W tank from 2000 ppm to 2500 ppm, an increase of 0.2 wt % boric acid in the containment sump, on the available margin in the long term cooling analysis is insignificant, when considering the available margin and the conservatism incorporated. Therefore, initiation of simultaneous hot leg and cold leg injection or initiation of shutdown cooling between 5Vz and 6¥2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is still adequate to maintain core cooling and prevent boric acid precipitation with SIT and SIR W tank boron concentrations of 2500 ppm.

J

ATTACHMENT 6 Consumers Power Company Pa 1 i sades Pl ant Docket 50-255 CYCLE 10 TECHNICAL SPECIFICATIONS CHANGE REQUEST BORIC ACID SOLUBILITY VS TEMPERATURE November 1, 1991

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34 32 30 28 26 24 22 20 18 16 14 12 10 8

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2 BORIC ACID SOLUBILITY (1% BORIC ACID = 1748.3 ppm)

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~v 32 50 68 86

. 104 122 140 158 176 194 212 TEMPERATURE (F)

Weight of boric acid soluble per saturated solution weight vs saturation temperature.

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ML18057B352

Text

1.0 INTRODUCTION

SUMMARY

4.0 CONCLUSION

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

4.0 CONCLUSION

SUMMARY

2.0 REFERENCES

3.0 BACKGROUND

SUMMARY

7.0 CONCLUSION

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