ML20090E905

Safety Evaluation W/Increased Enthalpy Rise Factor

ML20090E905
Person / Time
Site:	Prairie Island
Issue date:	05/31/1984
From:	Richard Anderson, Kapitz J, Rautmann D NORTHERN STATES POWER CO.
To:
Shared Package
ML20090E885	List: ML20090E882 ML20090E891 ML20090E901 ML20090E905 ML20090E917
References
NSPNAD-8406, TAC-55445, TAC-55446, TAC-55816, TAC-55817, NUDOCS 8407200094
Download: ML20090E905 (29)

Text

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EXHIBIT C 4

PRAIRIE ISLAND UNITS 1 AND 2 SAFETY EVALUATION WITH INCREASED ENTHALPY RISE FACTOR NSPNAD-8406 May 1984 Prepared by 3 tw , A W *d Data I///8/

Reviewed by ,

AM8 Date C [ // V Approved by //

,,, A -

/ ara Date 1/e//F hh[$00bK0 P

l Page 1 of 29

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1 LEGAL NOTICE This report was prepared by o. on behalf of Northern States Power Company (NSP). Neither NSP, nor any person acting on behalf of NSP:

a. Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, usefulness, or use of any information, apparatus, method or process disclosed or contained in this report, or that the use of any such information, apparatus, method, or process may not infringe privately owned rights; or

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b. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in the report.

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DISTRIBUTION

, C. E. Larson, Director Nuclear Generation MS3 G. H. Neils, Gen Mgr Hdqtrs Nuc Group MS4 L. R. Eliason, Gen Mgr Nuclear Plants MS3 E. L. Watz1, PI Plant Manager PI M. A. Klee, Supt Nuclear Engineering (2) PI D. M. Musolf, Mgr Nuclear Support Services (25) MS4 C. A. Bonneau, Supt Core Analysis MS4 T. L. Breene, Engineer II MS4 D. H. Peterson, Mgr Fuel Supply MS4 J. L. Karalus, Adm Fuel Supply Contracts MS4 D. A. Rautmann, Supt Safety Analysis MS4 L. McCarten, Nuclear Engineer MS4 V. J. Lax, Engineer II MS4 D. W. Dean, Engineer I MS4 P. J. Riedel, Engineer I MS4 J. K. Kapitz, Engineer II MS4 T. J. Tasto, Engineer Associate MS4 J. M. Garrick, Engineer Associate MS4 C. S. Gantner, Engineer Associate MS4 P. D. Pankratz, Engineer Associate MS4 Exxon Nuclear Inc. (5) Richland, WA R. L. Grow ' '

UAI R. C. Kern UAI Nuclear Analysis Department File System (4) MS4 2

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TABLE OF CONTENTS Page

1.0 INTRODUCTION

7 2.0 CALCULATIONAL MODELS AND METHODOLOGY 8 2.1 Calculational Models 8 2.2 Methodology 8 2.2.1 Transient Analysis 8 2.2.2 Rod Bow Penalty 8

, 2.2.3 Safety Limit Curves 9 2.2.4 Overtemperature AT Setpoint Verficiation 12 3.0 RESULTS 15 3.1 Input Parameters 15 3.2 Transient Analysis 15 3.3 Rod Bow Analysis 16

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-3.4 Safety Limits Curves 17 3.5 Overtamperature AT Setpoint Verification 18 -

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4.0 CONCLUSION

S 28

5.0 REFERENCES

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LIST OF TABLES Page 3.1-1 Parameter Values Used in Transient Analysis 19 3.2-1 Summary of BOC, HFP Base Case MDNBR 20 3.3-1 Slow Rod Withdrawal Transient and Thermal Margin Results 21 3.4-1 Prairie Island Units 1 and 2 22 Overtemperature AT Trip Setpoints l

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LIST OF FIGURES Page 3.2-1 Slow Rod Withdrawal PI 2 Cycle 8 - W-3 MONBR 23 3.2-2 Slow Rod Withdrawal PI 1 Cycle 9 - W-3 MDNBR. 23 3.4-1 Core Thermal Overtemperature Limits 24 3.5-1 Thermal Overtemperature Limits - 1700 psia 25 4

3.5-2 Thermal Overtemperature Limits 2400 psia 26 3.5-3 F(AI) vs AO - PI 2 Cycle 8 27 3.5-4 F(AI) vs AD - PI 1 Cycle 9 27 f

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_ . - _ . - _ - . - _ . - . - - - . . . - - - . . - - . _ _ . . .. . . , - - _ - - . - - . . . _- ~ ._.

1.0 INTRODUCTION

This report summarizes the calculations completed by NSPNAD in support of the Margin Improvement Program - Phaso II for Prairie Island Units 1 and 2. ,

In Phase II of the Margin Improvement Program, the Nuclear Enthalpy Rise Hot Channel Factor, FAH, is increased from 1.55 to 1.65.

The LOCA analysis was performed by the current fuel vendor (Exxon) and the results are reported in reference 1.

The non-LOCA analyses were performed by the NSP Nuclear Analysis Department for Prairie Island Unit 1 Cycle 9 and Unit 2 Cycle 8, the currently operating cycles.

These analyses have been divided into four distinct parts consisting of:

calculations for the limiting transients, evaluation of the rod bowing penalty, generation of the safety limit curves, and evaluation of the adequacy of the overtemperture AT t, rip function. The basic methodology and the analytical results which were obtained are presented in this report.

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2.0 CALCULATIONAL MODELS AND METHODOLOGY 2.1 Calculational Models The increased enthalpy rise factor calculations were performed using the NSPNAD Reload Safety Evaluation Methods for PWRs.(3) There have been no changes made to the codes since the analysis of Prairie Island Unit 2 Cycle 8.(5) 2.2 Methodology 2.2.1 Transient Analysis The non-LOCA transient analyses were performed by the NSP Nuclear Analysis Department (NSPNAD) using methods (2)(3)which have been approved by the U.S. Nuclear Regulatory Commission (NRC). These methods,have been demonstrated to be conservative for a complete spectrum of Reload Safety Evaluation transients, except LOCA, so

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that they can bb used to perform analysis relating to licensing actions.

2.2.2 Rod Bow Penalty ,

The methodology for calculating the reduction in MDNBR due to red bow for PI 1 Cycle 9 and PI 2 Cycle 8 with FAH = 1.55 is described in reference 4. After the RSE reports for these cycles (5)(6)were published, Exxon received NRC approval of a new computational method for evaluating the effect of fuel rod bowing on MDNBR.(7) The new method still uses the equation:

MDNBRB = MDNBRNB (18- 6 )

where:

MDNBRNB = MDNBR for nonbowed fuel MDNBRg = MDNBR for bowed fuel 6B = fractional reduction in MDNBR due to fuel rod bowing v Page 8 of 29

, The new method has changed the calculation of 6 . 6 IS " "

, B B defined as:

68 = 0.0 for 0 s AC/C, s 0.5 and m AC/C,- 0.5 AC 6B= *6 80W fr > 0.5 0.5 C, This change in methodology was made because rod bow DNB test' results have shown that the penalty on MDNBR is negligible until the fractional gap closure is well in excess of 0.5. This new methodology substantially reduces the rod bow penalty.

2.2.3 Safety Limit Curves The safety limit curves define the regions of acceptable operation

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with respect to average temperature, power, and pressurizer pressure.

The boundaries of acceptable operation are defined in terms of thermal overpower limit (fuel melting), thermal overtemperature limit (cladding damage based on DNB considerations), and locus of points where the steam generator safety valves open. These limits are used to set the overpower and overtemperature AT trip setpoints.

The thermal overpower limit to prevent fuel melting is protected by the thermal overpower trip. Fuel melting calculations are predominantly dependent on the total peaking factor, Fg , and the fuel type analyzed, and relatively independent of the enthalpy rise factor, F AH.

Previous calculations were based on Westinghouse fuel with an Fg of 2.58.(8) The Exxon fuel type is sufficiently similar to the Westinghouse design so that for an gF of 2.32, the current Technical Specifications value, the previous analysis, including the overpower trip setpoints, will bound the present core. Therefore, increasing FAH fr m 1.55 to 1.65 will not effect the thermal overpower limit calculations.

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The opening of the steam generator safety valves impose a physical limit on the reactor power and temperature, dependent only on the

, reactor system characteristics and independent of peaking factors.

Therefore, an increase in F 3g from 1.55 to 1.65 will not effect this limit.

Increasing F AH fr m 1.55 to 1.65 will significantly impact the results of the thermal overtemperature limit calculations.

For the overtemperature limit, the following four limiting criteria are used:

1. Vessel exit temperature < 650 *F (design temperature limits).

2. Vessel exit temperature < saturation temperature (ensures power ~ AT).

3. Minimum DNBR > 1.3 (fuel damage limit).

4. Hot channel exit quality < 15% (limit for W-3 CHF correlation).

The first two criteria result in a single limit on vessel exit temperature.

This limit is easily determined from an energy balance on the vessel at different pressures. ,

h3 , + Q/m = hout where:

Q = core power, Btu /hr m = vessel design flow rate = 68.2 E + 6 lbm/hr hin= inlet enthalpy, Btu /lbm h

out

=h f } lowest value, Btu /lbm h (650 'F) page 10 of 29

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The third criteria on DNBR is evaluated using the thermal margin methodology described in Reference 3 Appendix C. The following key assumptions are made in generating the DNB limits.

An 1/8 core CDBRA IIIC/MIT hot channel model is used.

• The radial peaking factor, F N H, will obey the following equation at powers below 100%, provided the control rod insertion limits are observed.

N F AH(P) = 1.65 [1 + .2 (1-P)]

where P is the fraction of rated core power.

At power levels above 100%, the value of FN is 1.65.

AH The reference axial power distribution, an upskewed cosine with a peak to average value, F N Z f 1.365 is used. The effect of other axial power distributiofis will be discussed in section 2.2.4.

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c The coolant flow rate will be the Thermal Design Flow, which is 68.2 M1bm/hr. ,

This model is then iterated to find the inlet temperature (and hence average temperature) that will give a MDNBR of 1.3 for a given pressure and power level. The resulting functions of T,yg versus core power, at I

different pressures, define the limits for this criteria.

The fourth criteria limits the hot channel outlet quality to less than

! + 15%. This criteria assures that the quality in the hot node (point of MDNBR) will also be less than + 15% which is the upper range of appitcability of the W-3 DNB correlation for the ranga of pressures over i which the core DNB limits are required. This criteria is evaluated using the same model used for the MONBR criteria only now the hot channel model is iterated to give an exit quality of + 15%.

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2.2.4 Overtemperature AT Setpoint Verification

, The thermal overtemperature limit is protected against fuel damage by the overtemperature AT reactor trip which compares the difference of the measured hot and cold leg temperatures to a dynamically calculated setpoint. The setpoint in Laplace notation is given by.

(1+t y) s setpoint AT =

AT,0T (g _g2 (Tave(s) - T,0T)

OT (1+t 2) s

+K3 (P(s) - P,) - f0T(AI))

where:

AT,0T = Indicated AT at rated power T,0T = Reference average temperature P, = Reference pressurizer pressure T*V'(3) = Measured average temperature P(s) = Measured pressurizer pressure ,

K3,K2,K3 = Constants (> 0) t y, t2 = Time constants f0T(AI) = A function of the measured difference between top and bottom excore detectors (t0)

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For steady state operation with AI=0, AT s p nt 0T reduces to:

AT OT tpoint (steady state) =

ave AT,0T (K -K2 (T - T,0T) + K3 (P - P,))

Ky includes an error allowance for calibration and instrument errors. AT 0T s tpoint (steady state) is then compared to the safety limit curves for the region where the thermal overtemperature limits are bounding. If AT0Ts tpoint (steady state) s AT safety limit, then the current setpoint is acceptable. If not, then the setpoint constants must be recalculated. The procedure for calculation of the overtemperature AT trip setpoint constants is described in Reference 9.

For transient operation, the dynamic terms,i t and t2, in the overtemperature AT trip function compensate for inherent instrument delays and piping lags between the reactor core and the loop temperature sensors. These affects are system dependent and independent of peaking factors. Thus, an increase in F AH will not effect these parameters.

The final consideration is to determine the adequacy of the f0T(AI) function. This parameter is included in the setpoint, because the Westinghouse methodology relating to minimun DNB utilizes a conservative axial power profile which is applicable only when the axial offset (which is related to AI) is within a certain operating band near zero. Power distribution control at the PI units during normal operations keeps the axial offset essentially within this band. Thus, if the power ' distribution becomes adversely skewed outside this band, the setpoint is appropriately reduced by f(AI) to ensure that the same thermal margins are maintained. The Westinghouse methodology consists of running a selected number of operational transients which span a wide range of anticipated load follows, boron dilutions with the rod controller in the manual and automatic mode, cooldown accidents, and rod withdrawal transients.

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. . _ . . _ _ . . _ _ _ _ . _ . , , . . , _ _ _ . . _ _ - . _ . . _ , _ _ . . . , , _ _ _ _ . . , _ _ _ _ - _ . _ . , . . ,_ . _ , . ,_,_r, __,,

o .

The axial power distributions which result during these transients are then used to compute the setpoint reductions to maintain a fixed thermal margin at full power conditions. Tha f(AI) is then set to .

bound all of these points to provide the necessa.y protection.

The NSP methodology approached this analysis in a similar manner.

The core safety limit curves were developed using a reference hot channel axial power distribution (i.e. upskewed cosine) for AI core

= 0. The effect of normal and offnormal operation on.the hot channel power distribution and on AI is determined using the core methods described in Reference 2. These distributions are obtained during the normal design analysis phase and bound both normal and transient operation including Condition II events. These distributions are input to the COBRA hot channel model and the power level and inlet temperature varied until the thermal margin (section 2.2.3) is the same as the reference case (AI = 0). The tpoint steady state AT equation is then solved for for f0T(AI)"

OT ATg setpoint ave

, f0T (AI) = Kg -K2 (T - T,0T) + K3 (P - Pg ) -

AT,0T s tpoint AT is taken to be the calculated AT at the safety limit OT line. Since the overtemperature AT trip setpoint bounds the safety limit curves for all pressures (using the reference distribution),

f0T (AI=0) Reference s 0. Tnerefore, the calculated f0T(AI) distribution must be biased so that f0T(AI=0) Reference = 0.

This biased distribution is then used to determine whether the current f0T(AI)setpoint distribution is bounding (see section 3.5).

This calculation is only performed at one pressure. Varying the pressure will only change the bias and will not effect the results.

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3.0 RESULTS 3.1 Inout Parameters The analyses in this report are performed for. Prairie Island Unit 1 Cycle 9 and Unit 2 Cycle 8, the currently operating cycles. Thermal-hydraulic parameters for full power operation are summarized in Table 3.1-1. All other parameters including neutronics, setpoints and delays are outlined in References 5 and 6. In all cases, the parameters used in the analyses bound the actual values for the Prairie Island plants.

3.2 Transient Analysis For steady state operation at HFP, raising the F AH limit to 1.65 reduces the MDNBR by about 11% and the MDN8R margin by about 29% for both cycles.

Table 3.2 1 summarizes the results for PI 1 Cycle 9 and PI 2 Cycle 8.

For non LOCA transients, changes in F AH nly impact events for which fuel damage is a consideration. For the PI units, the limiting transients with respect to the thermal margin fuel damage criterion on DNBR are: The RCC

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assembly misalignment, slow rod withdrawal at power, locked rotor, the large steamline break, and the rod ejection transient. ,

The RCC assembly misalignment, large steamline break, and red ejection transient will not be effected by a change in the Technical Specification

F AH limit for the currently operating cycles, i.e. PI 1 Cycle 9 and PI 2 Cycle 8. These transients are initiated from offnormal initial conditions, i.e. dropped or misaligned RCCA at HFP for the RCCA misalignment transient a'nd HZP on rod stuck out for the steamline break transient and ejected rod at HZP and HFP for the ejected rod transient. The peak.ng factors for these l transients typically will excced the Technical Specifications criteria.

These transients are therefore evaluated using the calculated peaking factors, including reliability factors rather than the Technical Specification limits. The calculated transient response is therefore dependent on the core loading and is independent of any change in the T.S.

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., limit on FAH nce cycle operation has begun. Therefore,'the previous calculations (reference 5 and 6) are still valid for PI 1 Cycle 9 and PI 2

. Cycle 8, respectively.

The locked rotor transient, a class IV transient, is evaluated based on the number of failed fuel pins. The failure criteria used is a MDNBR of less than 1.3, evaluated using the W-3 correlation. A pins census (F AH E*" PI ")

is done based on the calculated power districution, includ_ing reliability factors, at HFP. From this the MDNBR for each pin is determined. The calculated pin power distribution, at HFP, is dependent only on the core loading, once cycle operation has begun, and will not be effected by a change in the T.S. limit on F AH.

Therefore, the previous calculations (ref 5 and 6) are still valid, for PI 1 Cycle 9 and PI 2 Cycle 8, respectively.

The slow rod withdrawal transient assumes initial operation at HFP with F3g at the 1.S. limit. This transient must therefore be reanalyzed using an F

AH f 1.6 5.' The slow rod withdrawal event was reanalyzed and the resulting

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transient DNBR is shown in Figures 3.2-1 and 3.2-2, for PI 2 Cycle 8 and

, PI 1 Cycle 9, respectively, assuming values of F3g = 1.55 and 1.65. The

( former case shows the transient terminating earlier relative to the latter case because the scram was calculated to occur earlier. The difference, i in scram times is due to the fact that the 1.65 case included a more l- conservative evaluation of the overtemperature delta T trip setpoint instrument and measurement uncertainties than the 1.55 case._ This was done in order to make the analysis more consistant with current Westinghouse methodology. The results show sufficient margin with respect to meeting the 1.3 limiting criterion.

3.3 Rod Bow Analysis The most limiting operational transient for PI 1 Cycle 9 and PI 2 Cycle 8 is the slow rod withdrawal transient. For this event the MONBR was cal-culated to be 1.441 and 1.440, respectively. A further reduction in DNBR will be caused by rod bowing at high exposures. The methodology for cal-

. culation of the DNBR reduction due to rod bcw is explained in section 2.2.2.

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.; The maximum anticipated fractional gap closure was calculated at an assembly average exposure of 49,500 GWD/MTU. This exposure will bound the peak pellet exposure limit of 55,000 MWD /MTV.

The results of the rod bow penalty calculations are shown on Table 3.3.-l.

Both plants will lose an additional 3.4% on MDNBR due to rod bowing. This brings the MDNBR during a slow rod withdrawal transient down to 1.392 and 1.391 for PI 1 Cycle 9 and PI 2 Cycle 8, respectively.

The results of this analys:s show that by using an FAH f 1.65, together with the new rod bow penalty methodology, both Prairie Island 1 Cycle 9 and Prairie Island 2 Cycle 8 are adequately protected for all the transients and accidents for which fuel damage is prohibited.

3.4 Safety Limit Curves The safety lim,it curves were generated using the methodology described in section 2.2.3. These curves were generated using a Prairie Island Unit 1 Cycle 9 hot channel model. The MDNBR limit curve was established on the basis of 1.365 which includes an additional 5% over the required limit to compensate for cycle to cycle variations in the hot channel model, thereby f

making this analysis generic to Prairie Island units with T0 PROD fuel. ,

The 5% margin was established based on historical data which shows 1%

to 2% variations in DNSR between cycles with similar fuel types. The applicability of this number (5%) will be checked as part of-the Reload Safety Evaluation process.

The solid curves of Figure 3.4-1 represent the loci of points of thermal power, coolant pressure, and coolant average temperature for which either the coolant enthalpy at the core exit is limiting or the DNB ratio is limiting. For the 1685 psig and 1985 psig curves, the coolant average j enthalpy at the core exit is equal to saturated water enthalpy below power l levels of 81% and 65%, respectively. For the 2235 psig and 2385 psig i-curves, the coolant average temperature at the core exit is equal to 650 F below power levels of 58% and 66%, respectively. For all four curves, the

. DNBR is equal to 1.3 at higher power levels. The area of safe operation is below these curves.

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3.5 Overtemperature AT Setpoint Verficiation

. The applicability of the current overtemperature AT trip setpoint was verified as described in section 2.2.4. The overtemperature AT setpoints used in the analysis are given in Table 3.4-1.

• The overtemperature AT trip limit (with AI=0) was compared to the safety limit curves generated in section 3.4. The results of this comparison for the two limiting pressures of 1700 and 2400 psia, the low and high pressurizer pressure trip setpoints respectively, are shown in Figures 3.5-1 and 3.5-2, respectively. In all cases, the overtemperature AT trip setpoint is seen to previde adequate thermal margins with respect to the safety limits, since the setpoint always lies below all the limits.

Figures 3.5-3 and 3.5-4 show the results of the f( AI) function adequacy study for Prairie Island Unit 2 Cycle 8 and Unit 1 Cycle 9, respectively.

These results ,show that the current setpoint adequately covers this core design. In fact, with the exception of the one point at AI = 31, for pI 1 Cycle 9, the reference ~ power distribution is more conservative so that the need for the the f(AI) function is only minimal. These results confirm the

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fact that the reference axial power distribution has been chosen very conservatively and bounds all shapes. The adequacy of the f( AI) function will be checked each cycle as part of the Reload Safety Evaluation process.

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. TABLE 3.1-1 Parameter Values Used in Transient Analysis Analysis Input Value Core Total Core Heat Output, Mw (102%) 1,683.0*

Heat Generated in Fuel, % 97.4

. System Pressure, psia 2,220.0**

Hot Channel Factors Total Peaking Factor, FqT 2.32 N

Enthalpy Rise Facto,r, FAH 1.65 Total Coolant Flow, lb/hr ,

68.20 x 10 6 Effective Core Flow,1b/hr 65.13 x 10 0

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Reactor Inlet Temperature, 'F 534.5***

Steam Generators -

Calculated Total Steam Flow,1b/hr 7.26 x 10 6 Steam Temperatue, 'F 510.8.

Feedwater Temperature, 'F 427.3 Includes +2% uncertainty .

• • Includes -30 psi uncertainty

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• Includes +4 'F uncertainty Page 19 of.29

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TABLE 3.2-1 Sumary of BOC, HFP Base Case MDNBR MONBR Margin FAH=1.55 FAH=1.65  % Change  % Change PI 1 Cycle 9 2.083 1.859 -10.8 -28.6 PI 2 Cycle 8 2.114 1.881 -11.0 -28.6 i

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TABLE 3.3-1 Slow Rod Withdrawal Transient and Thermal Margin Results Prairie Island 1 Cycle 9

FAH=1.55 FAH=1.65 2

Rod Ave. Heat Flux @ time of MDNBR (Btu /hr ft ) 341,335 366,953 MONBR NB 1.760 1.441 (AC/C,) 95/95 0.485 0.5493 6 g 0.328 0.348 6

8 0.159 0.0338 MDNBR B

1.484 1.392 s

Prairie Island 2 Cycle 8 ,

FAH=1.55 FAH=1.65 2

Rod Ave. Heat Flux @ time of MONBR (8tu/hr ft ) 340,930 366,588 MONBR 1.790 1.440 NB (AC/C,) 95/95 0.485 0.5493 o g 0.327 0.348 og 0.1586 0.0348 MONBR B

1.5 1.391 i

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, TABLE 3.4-1 Prairie Island Units 1 and 2 m

Overtemperature AT Trip Setpoints AT,0T = 64.3 *F*

T,0T = 567.3 P, = 2235 psig K3 = 1.202**

K2 = 0.009 e K3 = 0.000566 f(AI) = a) for qt qb within -12 to +9 percent F(AI) = 0 b) for each^ percent that the magnitude of qt qb exceeds +9 percent the AT trip setpoint shall be aut6matica11y reduced by an equivalent of 2.5%

rated power

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c) for each percent that the magnitude of qt gb exceeds -12 percent the AT trip setpoint shall be automatically reduced by an equivalent of -

1.5% rated power.

AT, at design flow

• • Including uncertainties

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4. .

Siow Rod Withdrawal

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Minimum DNB Ratio vs. Time Figure 3.2-1 AFan.1.55 Prairie Island 2. Cycle 8 X Fan =1.65 4

3.s- * * ' **' * <. * ' *

 :- * * ** ='= **

• t. .

- = t-

. **==t.** .

3 .........:........4.......4.......:.........:.................4.........;....... <

g  :  :

e . .

E 2.g. ........ ......... . . . . . . . . ........ . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . ... ......

o  :

3 .

2i .A 2- = * * = * *

• t l.1 ****>****+*l*+-ta=*=. '

+++-*****=t***

• v -

• * :* 4 * * * : - * * * * :: A  :  :. - -

< s . . . . .

X .. >

1,3

..........................4..

A......v:A.........X.

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.- ' #Av

....:.........4....4.....:....4.....:.....:.....

1 . .

e a to is 20 ss 30 as 40 4e Time (Seconds)

Figure 3.2-2 A F 3=1.55 '

Prairie faland 1, Cycle 9 XFu=1.65 4

I . . .

3.,_ .. .. .

. ...s....

t l  :  : .  : .  :  : .

a. .

.: . . . . . . :~ . . . . . .: . . . . . . ...

t  :  :  :  :  :  :  :  : ,

a . . . . .

a

• n 2

o 3.5-

• * * * * * * * * * * * * * ' * * * **E*******************

• • • • • • • • • • • • • • h************ E 2 .  :

y  :.

. . i......c....,....

y 5 . . . . . . : . . . . . .g.A. . . . . . . ....

, s y .

X:

A. A' i e . .

X;

. :,. . . . . . . . . . 4. . . . . . . X. . . . . . . X' :. . 2 i.s _ .. ... .. ..

- . X. . . . A,X. . . . . .

n 4... ...........:. ........

• ggp

=

1 a

, O s i0 is 20 2s 30 as 40 es ,

Time (Seconds)

• CL

. s.

Page 23 of 29

9 Core Average Temperature Limits vs. Core Average Power Figure 3.4 -1 680 .

650 . . . . . . . . . ..... ... ..... . . . . . . . . . . . . . ........ . . . . . . . . . ..

640 ........... .. ..... ....

630- -. .

g2Q. .............. .. .. ....*.... ... . . . . ...

......f..

,6., .

g j Q .- .... ........... .............. .!.... . . . 4.. .. . .... ...... . . . ... .. ... .. .

. .e . .

800

. ....,.....,.4... .... . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .

as . .

.mme . . . .

e  : j*  :  :  :

V w 590- -: .

,~. '

,a -

4 .

580-  :

,.*- o. .

+ .

• 2385 psig 67Q . . . . . . . . .. .

Locus of Points et which Steam * -

Generator Safety velves Open  : 2235 P e.lU 560- .  :.  :.

: 19 85 p s.ig m

560 . . . . . . .. ......... .

e

. . . 9

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. . n 540 . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . . . . .

: ta ss pe.ig .

530 . . . . . . . . . .. . . . . . . . . . .. . . . . . . . .. . . . . . . . .. .

- ~

i n

E 520 ' ' ' '

O 2O 4o 6o 8O 150 120 15 0 150 E Rated Core Power (%) .

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FIGURE 3.5-1

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.- Core Average Temperature Limits vs. Core Avera9e Power Fah = 1.65, Fq = 2.32, Upskewed 1700 pala 630 . . . . . .

. . . . o.. . .

o o .-- . .

• o Locus of Points at which Steam ,.......*:.....,f.....!...........i..........,:........

820- ,

Generator Safety Velves Open  : , -

o .

o. s .

. *o. .

o . . .

610- + --:- - ' - - - - -

<. o , , +:. - - -

.o r . .

A

,s: .

800- ,.~  %- -

,r  :

. N: . ,#  :.  :.  :

. :o

e.< N  :  :

o . .

690 .......... . .........p..f . . . . .. .%. ,. . , . . . . . . . . .. . ...... .. .. .. . . .

. + . . . .

o, m J.. . . . .

N }..........

. g ggQ- f

........ 4 .......... ...

4........... .. . .. . .... ..... ........ ...

e a  :,  :  :  :  :  :.

w * . .

e  :  :  :  : N .  :

( . . . . .

N: .

F 570-  :

>r+-

Tcxrr - T.sAr

.  :  : d. Hot Chan. nel xcxir = 15%

560- .;

i i.-

- ~

+-.

I . . . .

f . . .

\ . . . . . .

( -

i 550- -

i. --

+- - -

i .

[ . . .

540- + . . t.- a.

.- +- ,

t..

. MDNB A = (1.3)(1.05) i . . . . .

l -A 530- r- - -

><..v- -

i

, r, .

. OTAT Limit .

n - -

l .  :

E 520 E

E 20 IO 6O 80 150 120 15 0 id0 180

. 0 Power (%)

l

.A l 7 J.

eh o

a Page 25 of 29 1

r ,

, FIGURE 3.5-2 Core Average Temperature Limits

., vs. Core Average Power FAh = 1.65, Fq = 2.32, Upskewed 2400 psia 660 . . . . . .

g 650 .............;...........4...........:......................................4...........:............

340- ............ ,. %..........j. ..........j.............j...... L cus of Points at which Steam

.  :  : Generator safety valves coen

. N :  :

630- -

. v. -v. -

s : .

.  :. - p:.  :.  :

N.- . .

Ne  : -

g20 ... .. . .. . . . . . . . . . . . . . . . . g .. . . . ... . . . . ... .

. . . o

.N

. .o*

f

d. N .-  :.

m g, g30-i s 4 NT = 650*F

....:...........a.............;.....a.x..rr...............

w i

. i.o*s#

Hot Channel xmy = 15% .

o

: e .  : .  :

3, . 4

. . g .

4 .

, r' +:

h 600-  :

~

. v-

. . *o

- ,o .

. o. . . .

590- - - -

a . . .

o,o o.! .

680- F- -- -

MONBR = (1.3)(1.05) 5'/0 - .> - . i.

?, -

. i .

OTai Limit .

= . . .

.: 560- -

. < r- - - - - - -

n .

.E 550 . . . . . .

N 4 20 40 60 80 100 120 140 160 180

.3 Power (%)

d 2

t.

c Page 26 of 29

b PRIARE ISLAND UNIT 2 CYCLE 3 F(AI) calc. vs. Axial Offset Figure 3.5-3 ,,,,

1-o .4- *

,,**, c

.~

• m*e. %q 4 0

.g. .s. .i. '.;. g . .  % g a ,. ^ i, i. h i. i.**

a ....-

-* 8- Legand T.C N. S P.C.1.lu 87

• • • • ' n LINEGn.Rit'Bu

. i-d 00.RA Ilic/MIT PRIARE ISLAND UNIT 1 CYCLE 9 FIAl) calc. vs. Axial Offset F10ure 3.5-4 ,u,,

1-t I _

e.s- ,

a'

,~, a 3 '- ^

.i g .;. . .. .. a...

g .g .i. .. wgi i ..

4 ....-

2=

A

..4-t<>

I *.8- Legend 2 reew. . pre. uurt f, '-

n t'S39We'!U o 6 co.Rasiiciuir g 1-l = .

!T.g i -

?>

u 5

g as Page 27 of 29 L

3 4.0 Summary and Conclusions r

- Analyses have been performed to demonstrate an increase in the Technical

~

Specifications limit for FAH from 1.55 to 1.65 for the two Prairie Island units will not result in any reduction in safety margins. This analysis .

1 covered the entire spectrum from anticipated transients through design basis LOCA's. The LOCA analysis was performed by Exxon (I) which resulted in a slight reduction in FQ to meet all licensing acceptance criteria. The non-LOCA analysis was performed by NSpNAD utilizing our in-house analytical methods. This analysis demonstrated that the limiting transients with respect to minimum DNBR fuel damage are not adversely effected with respect to meeting the limiting criterion. In addition, the overtemperature AT trip setpoint has been shown to provide adequate protection to the safety limits and the existing f0T(AI) function protects against all possible skewed axial power distributions.

h Page 28 of 29

a S.0 References t

1. " Prairie Island Units 1 and 2 Limiting Break LOCA ECCS Analysis with Increased Enthalpy Rise Factor" XN-NF-84-03, February 1983.

2. Qualification of Reactor Physics Methods for Application to PI Units" NSPNAD-8101P, December 1982.

3. " Reload Safety Evaluation Methods for Application to PI Units" NSPNA9-8102P, December 1982.

4. XN-75-32(P)(A), " Computational Procedure for Evaluating Fuel Rod Bowing" Supplement 1, June 1979.

5. " Prairie Island Units 1 Cycle 9 Final Reload Design Report (RSE)"

NSPNAD-8313P, October 1983.

6. " Prairie Island Unit 2 Cycle 8 Final Reload Design Report (RSE)"

NSPNAD-8305P Rev.2, November 1983.

7. XN-75-32(P)(A), "Compu.tational Procedure for Evaluating Fuel Rod Bowing" Supplement 4, October 1983.

8. WCAP 8091, " Fuel Densification Prairie Island Nuclear Generating Plant Unit No. 1", March 1973. /

, 9. WCAP 8746, " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions", March 1977.

e l

l l

l l

i i

i :

Page 29 of 29

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