ML20197H623

Rev 1 to Palisades Large Break Loca/Eccs Analysis W/ Increased Radial Peaking

ML20197H623
Person / Time
Site:	Palisades
Issue date:	02/28/1990
From:	Fausz N, Schmitt B, Slater C INTERMOUNTAIN TECHNOLOGIES, INC., SIEMENS CORP.
To:
Shared Package
ML18057A580	List: ML20197H616 ML20197H623
References
ANF-88-107, ANF-88-107-R01, ANF-88-107-R1, NUDOCS 9011200007
Download: ML20197H623 (57)

Text

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ANF-88-107 J REVISION 1 3 NIMp L'.

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3 j PALISADES LARGE BREAK LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAKING I

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ADVANCEDNUCLEARFUELS CORPORATION ANF 88-107 Revision 1 Issue Date: 02/12/90

$We PALISADES LARGE BREAX LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAKING

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. R. C. Gottula, Team Leader

PWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Service-Contributors

N. F. Fausz B. E. Schmitt (Intermoutain Technologies Inc.)

M Calvin Slater Ross Jensen M

January, 1990

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l ANF 88 107 Revision l' Table of Contents

1.0 INTRODUCTION

' . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 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' Sununary of Analysis Parameters ........ 8 3.4 Break Spectrum Results ... ................... 9

3. 5_ ' Axi al Shape S tudy Results . . . . . . . . . . . . . . . . . . . . . 10 3.6 Exposure Limits . . . . . . . . . . . ... . . . . . . . ..... 10

4.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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5.0 REFERENCES

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-l -ii- ANF 88 107 Revision 1 List of Tables IAhlt E.dS.t

. 2.1 Summary of Results for'0.6 DECLG Limiting Break Size ....... 4 3.3.1 onlisades System Analysis Parameters ..............12

-3.4.l Palisades Break Spectrum Analysis Results . . . . . . . . . . . . 14 3.4.2 Calculated Event Times for 0.4 DECLG Break ...........15 7-DN-3.4.3 Calculated Event Times for 0.6 DECLG Break ...........16 Y

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lii- ANF 88-107 Revision 1 list of Fiaures Floures g

/ 2.1 Allowable LHR as a Function of Peak Power Location . . . . . . . . 5 3.5.1 Normalized Power (EOC), 0.6 DECLG Break . . . . . . . . . . . . . - . 17 3 . 5 . 2 '. Double ' Intact Loop Accumulator Flow Rate, 0.6 DECLG Break .... 1B

'q 3.5.3 Single intact Loop Accumulator Flow Rate, 0.6 DECLG Break- .,.. 19 N: 3.5.4 Broken loop Accumulator Flow Rate, 0.6 DECLG Break . . . . . . . . 20 t 3. 5. 5. Total Intact Loop HPSI Flow Rate, 0.6 OECLG Break ........ 21 3'5.6

. Total-Intact Loop LPSI Flow Rate, 0.6 DECLG Breck ........ 22 3 . 5'. 7 Broken Loop SIS Flow Rate, 0.6 DECLG Break . . . . . . -. . . . . . 23

m. '3.5.B Upper Plenum Pressure during Blowdown, 0.6 DECLG Break . . . . . . 24 3.5.9 Total Break Flow Rate during Blowdown, 0.6 Break . . . . . . . . . 25

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3.5.10 Pressurizer Surge Line Flow Rate during Blowdown, 0.6 DECLG Break- . . . .-. .- . . . . . . . . . . . . . . . . . . . . . . . 26 3.5.11 Downcomer Flow Rate during Blowdown, 0.6 DECLG Break ... . . . . .

27 3.5.12 Average Core inlet Flow Rate during Blowdown, 0.6 DECLG Break,

'X/L = 0.8 ............................ .

2B 3.5.13 . Hot' Channel inlet Flow Rate during Blowdown, 0.6 DECLG Break,

'X/L = 0.8 ............................ 29 k

3.5.14 Hot Volume Inlet Flow Rate during Blowdown, 0.6 DECLG Break, X/L = 0.8 .-........................... 30

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. j iv- ANF 88 107 Revision 1

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List of Fiaures _,

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j. 3.5.15. Hot Node Fluid Quality during Blowdown, 0.6 DECLG Break, F 'X/L = 0.8 - . . . . . . .-.....................

. 31

~i 3.5.16 PCT Node Fluid Temperature during Blowdown, 0.6 DECLG Break, X/L = 0.8 . . . . . . ...................... 32 ~

l 3.5.17 PCT- Node- Fuel Average Temperature during Blowdown, 0.6 DECLG  !

Breat, X/L = 0.8 . . . . . . . . . . . . . . . . . . . . . . . . . 33 g 3.5.18 _ PCT Node Cladding Temperature during Blowdown, 0.6 DECLG Break, .t X/L = 0.8 . . . . . . - . . . . . . . . . . . . . . . . . . . . . . . . 34 3.5.19 . PCT Node Heat Transfer Coefficient during Blowdown, 0.6 DECLG Break, X/L = 0.8_. . . . . . ... . . . . . . . . . . . . . . . . . 35 s

3'5.20 PCT. Node Heat Flux during Blowdown, 0.6 DECLG Break, X/L = 0.8 . .

. 36 l 3.5.21' Containment Pressure, 0.6 DECLG Break, X/L = 0.8 . . . . . . . . . 37 1 3.5.22 ~ Upper' Plenum Pressure after E0BY,- 0.6 DECLG Break, X/L = 0.d . . . 38.

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-3.5.23 Downcomer Mixture Level after E0BY, 0.6 DECLG. Break, X/L = 0.8 . . 39

-3.5.24- _-Core Flooding Rate after E0BY, 0.6 DECLG Break, X/L = 0.8

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3.5.25 Core Mixture'. Level after E08Y, 0.6 DECLG Break, X/L = 0.8 . ... 41 3.5.26 PCT Node Cladding Temperature after E0BY, 0.6 DECLG Break, ii

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X/L = 0.8 . . . . . . . ....................... 42 I

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Revis1or. 1 r:t .3 Sumary of Revisions

1. Page 7, changed sentence from:

Single fr.ilure criteria is met by assuming that one HPSI pump and one '

LPSI ramp are not available for operation.

to:

Single failure criteria is met by assuming that one LPSI pump is not available for operation.

Reason for change: >

The single failure assumption of the loss of one LPSI pump was used in the ANF licensing analysis and -is. consistent with- the original CE licensing analysis. The single failure assumption -used in the ANF analysis was incorrectly stated in the original ANF 88-107. report. .

L 2. Page 9, changed sentence from:

The pump performance curves are characteristic of CE pumps.

to:

The primary pump performance curves are characteristic of Ct. pumr,s.

Reason.for change:

To clearly distinguish the pump performance curves as being for the primary pumps as opposed to the HPSI or LPSI pumps.

M. 3. Page 13 Table 3.3.1:

Added initiation times for containment fan' coolers and sprays.

Reason for change:

ANF received a question from Consumers Power Co. .regarding the-initiation time of containment sprays in the- LOCA analysis. The l

initiation times for the fan coolers and containment sprays were added-for additional information.

H 4. Changed distribution list.

L Reason for change:

The internal ANF distribution list was changed due to changes .in personnel since the original report was issued.

ANF 88 107 Revision 1 R

PALISADES LARGE BREAK LOCA/ECCS ANAL.YSIS WITH INCREASED RADIAL PEAKING y '

l.0 INTRODUCTION This document _ presents the results of a large break loss-of coolant accident: (LOCA) analysis for the Palisades plant operating with Advanced Nuclear Fuels. Corporation (ANF) fuel. The primary purpose of the analysis was to support an increase in the total radial peaking factor from 1.77 to 1.83.

The analysis was performed at a total radial peaking factor of 1.92 to -bound potential' future increases in the total radial ' peaking f actor. The analysis

. supports a maximum LHR of 15.28 kW/ft and a modification in the _ axial LHR limit curve shown in Figure 3.23-1 of the technical -specifications. The analysis also provides justification for removal of Figure 3.23-2 (allowable I.HR as aLfunction of burnup) and Figure 3.23-3 (allowable LHR as a function of peak power location, for interior and narrow water. gap fuel rods) from the technical specifications. The analysis was performed.for the Palisades plant operating at 2581 MWt (2530 MWt plus 2% uncertainty) and a maximum average steam generator tube plugging level of 29.3% with'up to 4.5% asymmetry, n Numerous changes have occurred in the ANF LOCA methodology since the previous' licensing calculations were performed for the ' Palisades plant.

Therefore, the scope of this analysis includes a mini-break spectrum analysis.

Calculations were performed for a 0.4, 0.6, and 0.8 double-ended cold leg guillotine break (DECLG) at the pump discharge to verify the previously determ'ined 0.6 DECl.G limiting break size (I) .

The analysis also includes calculations at the limiting- break size for 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 conn.vatively used the

.2 ANF 88 107 Revision 1 maximum fuel stored. energy near BOC where maximum densification occurs.

Justification is provided to support operation with ANF fuel up to a bundle

'. average exposure of 52,500 mwd /MTU with regard to the large break LOCA.

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ANF 88 107 1 Revision-1 1 .

2.0- SIM ERY OF RESULTS The- results of the analysis verified the 0.6 OECLG break as the limiting break size- . The analysis demonstrates that the 10 CFR 50.46(b) criteria 1 are satisfied for the Palisades! plant with the axially dependent

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power peaking limit curve shown in Figure 2.1. The analysis supports a maximum LHR cf 15.28 kW/ft up to a relative core height of 0.6 and a LHR of 14.75 kW/f t^ at a . relative core height of 0.8. The analysis supports a total radial peaking factor of 1.92 and 'a maximum average steam generator tube plugging level of 29.3% with up to 4.5% asymmetry. Results of the analysis i for both the' BOC and EOC axial profiles at the limiting 0.6 OECLG break size

. are shown _inLTable 2.1. The peak cladding temperature was calculated to be:

1914*F for the BOC profile and 2114*F for the EOC profile. The analysis supports Cycle 8 operation and is intended to support operation - for_ future L cycles.

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t ANF-88-loy Revision 1 TABLE 2.1

SUMMARY

OF RESULTS FOR 0.6 OECLG LIMITING BREAK SIZE

". BOC Stored Energy B0C Stored Energy B0C Axial shape EOC Axial Shape X/L = 0.6 X/L = 0.8 Peak LHR (kW/f t) 15.28 14.75 Hot Rod Burst Time (Sec) 41.77 41.37 Elevation (ft) 7.4 8.9

- Channel Blockage Fraction 0.31 0.34 Peak Cladding Temperature

- Temperature ('F) 1913.7 2114.2 Time (Sec) 52.47 57.57 Elevation (ft) 7.4 8.9

- Metal Water Reaction Local Maximum (%) 2.23 4.14 Elevation of Local Max. (ft) 7.4 8.9 Hot Pin Total (%) 0.46, 0.47,,

Core Maximum (%) <1.0 <l.0

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- 6- ANF 88 107 Revision 1  ;

I 3.0 AMLYSIS Section 3.1 of this report provides a cosciiption o' 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 sumary of the system, parameters used in the analysis. Section 3.4 provides a summary af the results of the mini break

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spectrum calculations. Section 3.5 summarizes the results of the limiting EOC p

axial power shape and section 3.6 provides justification for the burnup i

i independence of the LHR limit for ANF fuel, i

!r" i 3.1 Descriotion of LBLOCA Transient ,

c Q 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 hreak occurs on the pump discharge side of a cold leg pipe. The LOCA is assumed to result f*om an earthquake and is co-incident with loss-of-offsite power. Primary coolant pump coastdown occurs

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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 conservatively 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). Following CHF, the fuel rods dissipate heat through the transition and film boiling modes of heat transfer. Rewet is precluded during

[~~l .bicwdown by Appendix K of 10 CFR 50.

Q p- A Safety Injection System (SIS) signal is actuated when the appropriate d setpoint (high-containment pressure) is reached. Due to loss of offsite m power, a time delay for startup of diesel generators and SIS pumps is &ssumed.

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,Q Once the time ofiay criteria is met and the system pressure falls below the shutoff head of the High. Pressure Injection 3yste (HPSI) and Low Pressure injection System (LPSI) pumps, SIS flow is injected into the cold legs.

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I 7- ANF 88 107 #

Revision 1 Single f ailure criteria is met by assuming that one LPSI pump is not

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available for operation. When the system pressure falls below the accumulator Flow 3 #4 , pressure, flow from the accumulators is injected into the cold legs.

from the Emergency Core Cooling System (ECCS)is assumed to b>.. ass the core and

<p , flow . to the break until the end of bypass (E0BY) is gredicted to occur g' P Following E0BY, ECCS flow fills the (sustained downflow in the downcomer).

downcomer and lower plenum until the liquid level reaches the bottom of the During the refill period, core (beginning-of-core-recovery or 80CREC time).

heat is transferred from the fuel rods by radiation heat transfer.

The reflood period begins at BOCREC time. ECCS fluid fills the downcomer and As the mixture

' prov'.ies the driving head to move coolant through the core.

Steam binding occurs as the level moves up the core, steam is generated.

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 10 CFR 50)

The fuel rods are eventually cooled virvh tends to reduce the reflood rate.

ara quenched by radiation and convective heat transfer as the quench front moves up the core. The reflood heat transfer rate is predicted through carry over rate fraction experimentally determined heat transfer and

. correl ations .

The purpose of the LBLOCA analysis is to demonstrate that the eriteria stated in 10 CFR 50.46(b) :.re met. The criteria are:

1)

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

The amount of fuel element cladding which reacts chemically with water or 2) steam does not exceed 1% of the total amount of zircaloy in the core.

3)

The cladding temperature transient is terminatedTheathota time fuel when the core rod cladding geometry is still amenable to cooling.

oxidation limit of 17% is not exceeded during or after quenching.

The core temperature is reduced and decay heat is removed for an extended 4) period of time, as required by the long-lived radioactivity remaining in the core.

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3.2 Descriotion of Analytical Models l The ANF EXEM/PWR evaluation model(2) was used to perform the analysis. This evaluation model consists of the following computer codes:

1) R00EX2(3) for computation of initial fuel stored energy, fission gas release, and gap conductance;

2) RELAP4 EM for the system and hot channel blowdown calculations:

3) CONTEMPT /t.T-22 as modified in accordance with NRC Branch Technical  ;

Position CSB 61 for computation of containment back pressuret

4) REFLEX for computation of system reflood; and F 5) T000EE2 for the calculation of fuel od heatup during the refill and k reflood portions of the LOCA transient. ,

The quench time, quench velocity, and carayover rate fraction - (CRF) correlations in REFLEX, and the heat transfer correlations in T000EE2 are

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based on ANF'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

.g K of 10 CFR 50. The reactor core in RELAP4 is modeled with heat generation W rates determined from reactor kinetics equations with reactivity feedback, and g with actinide and decay heating as required by Appendix K. Appropriate

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conservatisms specified by Appendix K of 10 CFR 50 are incorporated in all of the EXEM/PWR models.

I3 LJ 3.3 Plant Descriotion and Summary of Analysis Parameters lf~ The Palisades plant is a Combustion Engineering (CE) designed pressurized water reactor which has two hot leg pipes, two U tube steam generators, and l four cold leg pipes with one _ recirculation pump in each cold lag. The plant utilizes a large dry containment. The reactor coolant system was nodalized 7

into control volumes representing reasonably homogeneous regions,

{ interconnected by flow paths or " junctions". The two cold legs connected to

9- ANF 88 107 Revision 1 i

the intact loop steam generator were assumed to be symmetrical and were l

i modeled as one intact cold leg with appropriately scaled input. The model considers four accumulators, a pressurizer, and two steam generators with both

,g primary and secondary sides of the steam generators modeled. The high pressure safety injection (HPSI) and residual heat removal (LPSI) pumps were s modeled as fill junctions at the accumulator lines, with conservative flow "5

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 y channels, each with three axial rodes. A steam generator tube plugging level

! of 29.3% was assumed with an asymetric steam generator tube plugging of 4.5%.

The break was conservatively assumed to have occurred in the most highly 7

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 f;a"f '3.3.1. I i

3.4 Break Soectrum Results A mini-break spectrum study was performed to confirm the previously

.,3 determined 0.6 DECLG break as the limiting break size since numerous changes ,

have occurred in- the ANF LOCA methodology since the previous licensing calculations were performed for the Palisades Plant. Calculations were  ;

performed for 0.4, 0.6, and 0.8 DECLG break sizes with an axial power shape  ;

peaked at a relative core height of 0.6. Also, ANF methodology previously and l currently shows that split breaks are less limiting the guillotine breaks. J Therefore, split break calculations were not included in this analysis.  ;

f System bloWown calculations were first performed to the end of bypass (E0BY) to confirm the 0.6 DECLG as the limiting break size. Fuel and cladding temper:tures between the 0.4 and 0.6 DECLG break sizes were fairly close at the end of-bypass such that it was not conclusive that the 0.6 DECLG break was the limiting break. Tharefore, calculations were performed through the refill

" and reflood periods fo* these two break sizes. The results of the b-eak

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! 10 ANF 88 107 Revision 1 spectrum study are shown in Table 3.4.1. The 0.6 DECLG break size is confirmed as the limiting break. The peak cladding temperature (PCT) for the 0.6 DECLG break with an axial power shape peaked at a relative core height of 0.6 was calculated to be 1914 'F. Thus, a maximum LHR of 15.28 kW/f t is supported up to a relative core height of 0.6. Calculated event times for the M 0.4 DECLG break are shown in Table 3.4.2. Calculated event times for the 0.6 DEClG break are shown in Table 3.4.3.

3.5 Axial Shane 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 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 2114'F. Plots of parameters depicting calculations for the limiting 0.6 DECLG break and the EOC shape are shown in Figures 3.5.1 through 3.5.26.

3.6 Exoosure Limits The r6sults of previous exposure analyses for the Palisades plant (I) required a reduction in the LHR limit at high exposures. This was a result of the use of the previous ANF fuel rod code, GAPEX. Er90sure calculations have been performed with the current EXEM/PWR methodology using R00EX2 for two plants with a maximum bundle average exposure of 52,500 mwd /MTV. The current ANF methodology predicts maximum fuel storage energy to occur near BOC where maximum densification occurs. Closure of the fuel cladding gap at higher exposures significantly reduces the fuel stored energy. At high exposures, gap closure significantly outweighs the effect of higher concentrations of fission gases which tend to reduce the gap conductance and increase fuel stored energy. Also, the reduced stored energy at high exposures outweighs any adverse effects of increased rod internal pressure at high exposures.

Thus, the peak cladding temperature will be lower at high exposures than for sta

=

11 ANF 88 107 Revision 1 l

. I the limiting case reported in Section 3.5 which assumes a BOC fuel stored 4 energy. Since this phenomena is fuc1 related rather than system related, the exposure study results for other plants are applicable to the Palisades plant.

1 Thus, the LHR limit is independent of exposure up to a maximum bundle average l exposure of 52,500 mwd /MTU.

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12- ANF 88 107 Revision 1 TABLE 3.3.1 PALISADES SYSTEM ANALYSIS PARAMETERS Primary Heat Output, MWt 2530*

Primary Coolant Flow Rate, 1bm/hr 1.203 x 108 (318,770 gpm)

Primary Coolant System Volume, ft 3 8808**

Operating Pressure, psia 2060 Inlet Coolant Temperature (hottest loop), 'F 544 Reactor Vessel Volume, ft3 4782 W Pressurizer Total Volume, ft3 Pressurizer Liquid Total, ft3 1504 803 Accumul< tor Total Volume, f t3 (one of four) 2011

. Accumulator Liquid Volume, ft3 1116 Accumulator Pressure, psia 215 Accumulator Fluid Temperature, 'F 90 Total Number of Tubes per Steam Generator 8519 Steam Generator Tube Plugging 33.8 24.8 % split Number of Tubes Plugged (33.8 % SGTP) 2878 M Number of Tubes Plugged (24.8 % SGTP) 2114 SteamGeneratorSecondarySidegeat 48,661 Transfer Area, 33.8% SGTP, ft Steam Generator Secondary Side eat.

Transfer Area, 24.8% SGTP, ft 55,245 Steam Generator Secendary Flow Rate, 1bm/hr 5.241 x 106 (33.8% SGTP) *

(47 53% power split) 5.949 x 106 (24.8% SGTP)

Steam Generator Secondary Pressure, psia 730 Steam Generator Feedwater Enthalpy, 8tu/lbe 414 Primary Heat Output used in RELAP4-EM Model - 1.02 x 2530 2580.6 MWt.

Includes pressurizer total volume and 29.3% average SGTP.

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13- ANF.88 107 Revision 1 TABLE 3.3.1 PALISADES SYSTEM AV.ALYSIS PARAMETERS (CONTINUED)

Reactor Coolant Pump Rated Head, ft 260 Reactor Coolant Pump Rated Torque, ft-lbf 32,530 Reactor Coolant Pump Rated Speed, rpm 880 ReactorgeolantPumpMomentofInertia, 98,000 1bm-ft Containment Volume, ft3 1.64 x 106 ]

Containment Temperature, 'F 90 l SIS Fluid Temperature, 'F 70 l

HPSI Delay Time, Sec. 27.0 LPS! Delay Time, Sec. 28.0 Containment fan coolers initiation time, sec. 0.0 Containment sprays initiation time, sec. 0.0 l

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14- ANF 88 107 Reviston 1 m TABLE 3.4.1 PALISADES BREAK SPECTRUM ANALYSIS RESULTS DECLG 0.4 DECLG 0.6 DECLG 0.8 X/L = 0.6 X/L = 0.6 X/L = 0.6 Peak LHR (kW/ft) 15.28 15.28 15.28 E08Y Time (Sec) 24.77 19.17 16.74 m Fuel Average Temperature at E08Y (*F)

Cladding Temperaturi at 1424.3 1429.1 1386.8 M- E08Y ('F) 1250.9 1245.1 1176.8 Hot Rod Burst Time (Sec) 48.77 41.77 Elevation (ft) 7.4 7.a

- Channel Blockage Fraction 0.32 0.32 Peak Cladding Temperature

- Temperature (*F) 1851.0 1913.7 Time (Sec) 57.57 52.47 Elevation (ft) 7.4 7.4 Metal Water. Reaction

- Local Maximum, (%) 1.93 2.23

- Elevation of Local Max. (ft) /.4 7.4 Hot Pin Total (%) 0.43, 0.46 Core Maximum (%) <l.0 <l.0*

sta At 350 Sec W

15 ANF 88 107 Revision 1 TABLE 3.4.2 CALCULATED EVENT TIMES FOR 0.4 DECLG BREAK

=. .g g Time (Sec.)

Start 0.0

' ' 'b Break is Fully Open 0.05 Safety injection Signal 0.81 Pressurizer Empties 12.6 Accumulator injection Begins, Broken Loop 18.5 Accumulator injection Begins, Single intact Loop 20.3 Accumulator Injection Begins, Double Intact loop 20.3 End-of-Bypass (E0BY) 24.77 Start of Reflood 44 41 Peak Cladding mperature is Reached (X/L = 0.6) 58.47 Accumulators Empty, Broken loop 74 74-

- Accumulators Empty, Single intact loop 77 41' Accumulators Empty, Double intact Loop -

78.35 2

_^

w i isiami --------- --

.a 16- ANF 88 107 Revision 1 TABLE 3.4.3 CALCULATED EVENT TIMES FOR 0.6 DECLG BREAK Lyggt Time (see.)

Start 0.0 Break is Fully Open 0.05 Safety injection Signal 0.62 Accumulator injection Begins, Broken loop 11.70 Pressurizer Empties 12.26 Accumulator Injection Begins, Single Intact Loop 15.65 Accumulator injection Begins, Double Intact Loor 15.65 End of-Bypass (E0BY). 19.17 c3 Start of Reflood 27.70 Peak Cladding Temperature is Reached (X/L = 0.6) 52.47 Peak Cladding Temperature is Reached-(X/L = 0.8) 57.57 Accumulators Empty, Broken loop 68.9 Accumulatorr Empty, Single Intact Loop 72.85 Accumulators Empty, Double intact Loop 73.75 8$8 sh ses she see NR .

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4.0 CONCLUSION

S 4 The analysis with the current EXEM/PWR models for the Palisades plant

. confirms the 0.6 DECLG break size as the limiting break size. The analysis supports operation of the Palisades plant at a power level of 2530 MWt, an increase in the total radial peaking factor from 1.77 to 1.83, 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 kk/ft with the- axially dependent power peakirg limit shown in Figure 2.1. The analysis supports Cycle 8 operation and is intended to support operation for future cycles.

Operation-of the Palisades plant with ANF 15x15 fuel at or below the LHR limits shown in Figure 2.1_ assures that the NRC acceptance criteria (10 CFR 50.46(b)] for coss of-Coolant Accident pipe breaks up to and including the double ende'd severance of a reactor coolant pipe will be met with the emergency core cooling system for the Palisades plant.

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5.0 REFERENCES

1 LOCA Analysis for- Palisades at 2530 MWt usina the ENC WREM-Il PWR ECCS Evaluation Model, XN NF-77-24, Exxon Nuclear Company, Richland, WA 99352,

' July 1977, 2 Dennis M. Crutchfield (USNRC Asst. Director division of PWR Licensing B)

'N " Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation

% Model EXEM/PWR and Acceptance for Referencing of Related Licensing

' Topical Reports", dated July 8,1986.

v

3. R00EX2: Fuel Rod Thermal Mechanical Resoonse Evaluation Model, XN-NF 81-

• 58(P)(A), Revision 1, and Supplements 1-4, Exxon Nuclear Company, y Richland, WA 99352, February 1983.

j D

K n

a s

w n

M M

45- ANF 88 107 Revision 1 Issue Date: 02/12/90 PALISADES LARGE BREAK LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAKING Distribution TH Chen RA Copeland NF Fausz M LJ Federico RC Gottula c JW Hulsman JD Kahn TR Lindquist LD O' Dell GL Ritter HG Shaw (1)/ Customer (15)

El Tolman HE Williamson W

Docuement Control (5).

BE us sus

h ATTACHMENT 5 Consumers Power Company Palisades Plant Docket 50-255 t

SAFETY INJECTION TANK LEVEL (TAC NO. 76806) REVISION 1 A COMPARISON OF CALCULATED EVENT TIMES USING THE EXISTING AND PROPOSED MINIMUM TANK LEVELS November 9, 1990 k

2 Pages

gaggjguu ne m oono n onwut u nm orrr November 8,1990 HOS:372:90 I

k Mr. Guy Packard

.Pallsades Nuclear Plant

=

Consumers Power Company 27780 Blue Star Memorial Hwy.

Covert, MI 40043 9830 -

Subject:

Changes in Timing of Safety injection Tank (SIT) Flow Due to Reduced SIT Uquid Level for the Palisades Nuclear Plan:

References:

'1) ANF Letter (H. G. Shaw) to CPCo (R. J. Ger!!ng) dated April 11, 1990, HGS;117:80

2) Palisades Larga Break LOCA/ECCS Analyels with incrossed - Redial Peaking, ANF-88107, Revlelon 1, February 1990 Reference 1 transmitted the results of a disposition of Chaeter 15 evente for a reduced Safety injection Tank (SIT) level from 188 In. to 174 in. The disposition concluded that the reduced liquid level would have no impact on the large break LOCA analysis. The timing of SIT flows to the intact and broken loops for the 186 In. liquid levol are documented in Table 3.4.3 of Reference 2.- SIT tiows to the broken loop, cingle latacrloop, and doubje immet loop were reported to end at 68.9,72.85, and 73.75 seconds, respectively. With a reduced liquid level of

174 in., the Safety injection Tanks will empty about 4 seconds earlier. Therefore, the times at which the intact loop SITS empty are beyond the time at which the peak cladding tentperature (PCT) occurs, in addit!on, the time at which flow from the SIT lines and intact cold legs to the -

downcomer ende is well beyond the time at which the SITS empty, if you have any additional questions, please do not hesitate to contact me.

Very truly ours,-

H. G. Shaw -

Contract Adminletrator tlm c: T. A' Buczwinski E. R. VanHoof .

A Stomens company 7)

N

?

• CALCULATED EVENT TIMES-FOR 0.6 DECLG BREAK-Eggal Time (Sec.1

• OLD NEW Start 0.0 0.0

+ Break is Fully Open 0.05 0.05

- Safety Injection Signal 0.62 0.62 Accumulator Injection Begins, Broken Loop '11.70 11.70

- Pressurizer Empties 12.26 12.26 Accumulator Injection Begins, Single Intact Loop 15.65 15.65 Accumulator Injection Begins, Double Intact Loop 15.65 15.65

, End-of-Bypass (E0BY) 19.17 19.17 Start of Reflood 27.70 27.70

' Peak Cladding Temperature is Reached (X/L = 0.6) 52.47 52.47 Peak Cladding Temperature is Reached (X/L = 0.8) 57.57 57.57 Accumulators Empty, Broken loop 68.9 64.9

< Accumulators Empty, Single Intact Loop 72.85 68.85

' Accumulators Empty, Double Intact Loop 73.75 69.75

• 0ld time is from ANF-88-107, Rev. I and the New time is representative of SITS.

at the proposed lower-limit of 174".

! LICENSING CORRESPONDENCE - RECORD

SUMMARY

DAIL: November 9, 1990 DOCKET 50-255 LICENSE-DPR PALISADES PLANT TECHNICAL SPECIFICATION CHANGE REQUEST - SAFETY INJECTION TANK (SIT) LEVEL (TAC N0.

76806) -REVISION 1

SUMMARY

Requests change to Technical Specification 3.3.1.b to reduce the minimum safety injection tank level from 186 inches (55.5%) to 174 inches (52%)

COMMITMENTS MADE: (Identify Close-out Document) RESIDENT COMMITMENT Ves/No None COMMITMENTS CLOSED:

None Previous NRC/CP Co Correspondence Resident Commitment Information NRC letters dated None Air No. -

CPC letters dated 06/i3/90 Resident Document -

Responsible Individual -

Due Date -

AIR No UFI No 950-70*0l*07*03,02216/00000 Individuals Providina Information: Individuals Assianed Resoonsibility for imolementina Commitment:

GCPackard, Pal RJGerling, Pal Concurrences: Individual Responsible for Obtainina RJGerling, . Pal DJVandeWalle, Pal M GCPackard, Pal JLKuemin, Pal RAVincent, Pal Budget - FSAR/FHSR Chanae (Identifv):

NSB -

PS&L Log 90-0391 Yes/No Category #

Special Distribution RRFrisch, Pal Oriainator: Individual Responsible for Initiatina Chance Reauest:

RWSmdey, Pal N/A y

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