ML20150F277
| ML20150F277 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/30/1988 |
| From: | Shandeth Montgomery NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20150F258 | List: |
| References | |
| NSPNAD-8705, NSPNAD-8705-R01, NSPNAD-8705-R1, NUDOCS 8807180211 | |
| Download: ML20150F277 (32) | |
Text
- _. . ,_.
PRAIRIE ISLAND UNITS 1 AND 2 SAFETY EVALVATION OF ,
INCREASED FQ AND FAH t
NSPNAO-8705 ,
Revision 1 June 1988 Prepared by , Date /$0/f/(
Reviewed by , Y e Date [9/9 97 Approved by / ,
di Mf87 Date d d [
2, -
8807180211 880705 -j_
LEGAL NOTICE This report was prepared by or on behalf of Northern States Power Company (NSP). It is intended for use by NSP personnel only. Use of any information, apparatus, method, or process disclosed or contained in this report by non-authorized personnel shall be considered unauthorized use, unless said personnel have received prior written permission from NSP to use the contents of this report. With respect to unauthorized use, 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
- 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.
EXECUTIVE
SUMMARY
This report summarizes the calculations performed by the Northern States Power Nuclear Analysis Department (NSPNAD) in support of Technical Specification changes to FQ and FaH. The~new FQ and FAH limits are 2.50 and 1.70. The new FQ limit of 2.5 no longer contains the K(z) multiplier. The FAH limit equation at reduced power includes a multiplier of 0.3.
NSPNA0 has analyzed the transient response of Prairie Island with the new Technical Specifications. The results show that Prairie Island still meets all transient acceptance criteria with the new Technical Specifications and therefore they involve no unreviewed safety questions.
The results presented in this report are based on Prairie Island Unit 1 Cycle 13, and the acceptability of the new limits for Prairie Island Unit 2 Cycle-12 has been verified. This analysis covers both the Westinghouse OFA fuel and ENC TODR00 fuel.
The original report has been revised to update the results to reflect the changes in the analysis methodology since February 1987. These changes include: correcting an error (error 87002) in the surface heat transfer coefficient used in the analysis of the rod ejection event, an increase in the reliability factor (RF) applied to the Doppler coefficient, an increase in the low setpoint for the high neutron flux trip (power range), an updated rod bow penalty fac*.or for the MDNBR and a revised SCRAM reactivity insertion curve j to insure conservatism in all cases.
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TABLE OF CONTENTS Paae
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1.0 !NTRODUCTION 1.1 1 2.0 CALCULATIONAL MODELS AND METHODOLOGY 2.1 2.1 Calculational Models 2.1 2.2 Methodology 2.1 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.1 3.1 Design Criteria . 3.1 3.2 Core Hydraulic Compatability 3.1 3.3 Thermal Margin 3.2 3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance 3.2 3.5 Safety Limit Curves 3.3 4.0 ACCIDENT AND TRANSIENT ANALYSIS 4.1 4.1 Plant Transient Analysis 4.1 4.1.1 Input Parameters 4.2 4.1.2 Transient Analysis Results 4.3 4.1.2.1 Fast Control Rod Withdrawal 4.3 4.1.2.2 Slow Control Rod Withdrawal 4.3 4.1.2.3 Loss of External Electric Load 4.4 4.1.2.4 Dropped Rod - Auto Control 4.5 4.1.2.5 Loss of Reactor Coolant Flow 4.5 4.1.2.6 Locked Pump Rotor 4.6 4.1.2.7 Main Steam Line Break 4.7 4.2 Rod Ejection Analysis 4.7
5.0 REFERENCES
5.1
-iv-L- ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. - . . . . . - . . - . . - . . . . = . . - . . . .. . .- .- . _ .-.
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LIST OF TABLES Page 3.1 l Prairie Island Thermal Hydraulic Reference Conditions 3 .~ 4
'4 d Summary of Prairie Island Transient Margins ~ 4. 8
~ 4 '. 2 Parameter _ Values Used in Full. Power. Transient Analysis 4.9 4' . 3 Prairie. Island Units 1 and 2 Trip Setpoints . 4.10 4.4 Prairie Island Ejected Rod Analysis, HFP 4.11 4.5 Prairie Island Ejected Rod' Analysis, HZP 4.11
.y.
LIST OF FIGURES Page 41 Fast Rod Withdrawal 4.12 4.2 Slow Rod Withdrawal 4.13 4.3 Turbine Trip 4.14 4.4 Dropped Rod - E0C 4.15 4.5 2/2 Pump Trip 4.16 4.6 Locked Rotor 4.17 l
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-1,0. INTRODUCTION This report summarizes the calculations performed by the Northern States' Power Nuclear Analysis Department (NSPNAD) in support of Technical Specification changes to FQ and FAH. The new FQ and FAH limits are 2.50 and 1.70. The new FQ limit of 2.5 no longer contains the K(z) multiplier. The FAH equation includes a multiplier on the limit at reduced power of 0.3.
NSFNAD has analyzed the transient response of Prairie Island with the new Technical Specifications. The results show that Prairie Island still meets all transient acceptence criteria with the new Technical Specifications and therefore they involve no unreviewed safety questions. The results presented in this report are based on Prairie Island Unit 1 Cycle 13 and the acceptability of the new limits for Prairie Island Unit 2 Cycle 12 has been verified. This analysis covers both the Westinghouse OFA fuel and ENC TOPR00 fuel.
Section 2 of this report describes the calculational models and methodology used for this analysis. Section 3 contains the thermal-hydraulic design analysis. Section 4 contains the accident analysis results.
The original report has been revised to update the results to reflect the changes in the analysis methodology since February 1987. These changes include: correcting an error (error 87002) in the surface heat transfer coefficients used in analysis of the rod ejection events, an increase in the reliability factor (RF) applied to the Doppler coefficient from 10% to 25%, an increase in the low setpoint for the high neutron flux trip from 25% of full reactor power to 40%, a reduction in the rod bow penalty factor to 2.6%, and a revised SCRAM reactivity insertion curve to insure conservatism in all cases. Error 87002 had the effect of not forcing the fuel rod into the film boiling heat transfer regime during a control rod ejection event. Of the transients analyzed in this report, the change in the low setpoint for the high neutron flux trip (power range) only affected the rod ejection event from hot zero power (HZP) conditions. The change in the Doppler reliability factor and the SCRAM reactivity insertion curve affects almost every transient in this report. The SCRAM reactivity curve used in the original analyses assumed constant control rod velocity as they were dropping into the core. The revised curve models the effect dra), buoyancy and gravity forces and tends to delay the SCRAM reactivity insertion.
1.1
2.0 CALCULATIONAL MODELS. AND METHODOLOGY 2.1 Calculational Models Thsse' calculations have been performed using-the NSPNAD Reload Safety Evaluation Methods for PWRs (Reference 1). These methods have been reviewed'and approved by the~NRC.
2.2 Methodoiogy For this analysis NSPNAD has evaluated the limiting transients.for' Prairie Island. These limiting transients have been identified previously by Westinghouse (Reference 2), Exxon (Reference 3, 4 and 7) and NSPNAD (Reference 1). The limiting transients are:
- 1. Fast Control Rod Withdrawal
- 2. ' Slow Control Rod Withdrawal
- 3. Loss of Power to Both Reactor Coolant Pumps
- 4. Locked Rotor in One Reactor Coolant Pump
- 5. Loss of Electric Load
- 6. Large Steam Line Break
- 7. Small Steam Line Break
- 8. Rupture of Control Rod Drive Mechanism Housing (RCCA Ejection)
The transients that are not reanalyzed are:
- 1. Uncontrolled RCC Assembly Withdrawal From a Subcritical Condition
- 2. Startup of Inactive Loop
- 3. Feedwater System Malfunction
- 4. Excessive load Increase
- 5. Loss of AC Power.
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. 2.1
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These transients have.not been limiting in the past. The new Technical
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Specifications will not change the' relative worth of various transients, so these transients will continue to be non-limiting. This conclusion is supported by the large amount of transient analysis NSP, has performed to date for the Prairie Island units. Thesanalysis has shown that while the'new FQ and FAH wil'1 affect-the initial steady state MDNBR, it will not change the relative change in MDNBR. NSPNAD has performed Prairie Island transient analysis at severa'l different values ~ of FQ and FAH without seeing any change in the relative severit'y-of the accidents.
These results form the basis for concluding that the limiting transients identified in the past analysis will continue to be limiting under the aew Technical Specifications on FQ and FaH.
2.2
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3.0- THERMAL HYORAULIC DESIGN ANALYSIS - '
- This section provides results of the. thermal hydrau'ic design analyses for.
Prairie Island.
n 3.1 Design Criteria The-thermal a'nd hydraulic design performance requirements for; Prairie
( Island' fuel are is follows.
'1. The minimum departure from nucleate boiling ratio'(MONBR) will be:
1.3 at overpower for ENC fuel using the W-3 correlation with corrections for non-uniform axial heating, cold wall effects, and a reduction in MDNBR due to fuel rod bowing.
1.17 at overpovor for Westinghouse fuel using the WRB-1 correlation with corrections for non-uniform axial heating and a reduction in MDNBR due to fuel rod bowing.
- 2. -The fuel must be thermally and hydraulically compatible with the existing fuel and the reactor core throughout the life cycle of the fuel.
- 3. The maximum fuel temperature at design overpower shall not exceed the fuel melting temperature for ENC fuel or 4700 F for Westinghouse fuel.
- 4. For ENC fuel the cladding upper temperature limits shall not exceed.
Inner surface temperature 850 F Outer surface temperature 675 F Average volumetric temperature 750 *F 3.2 Core Hydraulic Comoatibility The hydraulic compatibility of the Prairie Island reload fuel is discussed in Reference 5.
3.1
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' 3.3 Thermal Margin The.most limiting transient for Prairie Island is the . dropped rod event at E0C conditions. The minimum DNBR was calculated to be 1.445, using the WRB-1 correlation.
Table 3.1 provides reference' conditions for the analysis. Details of the plant transient analysis.for Prairie Island are~given.in Section 4.0.
3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance The calculation of the ONBR reduction due to rod bowing for Westinghouse fuel is described in detail in Reference 6. The form of the rod bow penalty is as follows:
MDNBRB = MONBRNB O-68)
MONBRB = MDNBR for bowed fuel MONBRNB = MDNBR for nonbowed fuel 68 = fraction reduction in MDNBR due to rod bowing where 68 is given as a function of assembly average burnup.
The value of 6 used represents the rod bow penalty at an average assembly burnup of 8
24,000 MWD /MTU. While the amount of rod bowing increases beyond this exposure, the fuel is not capable of achieving limiting peaking factors due to the decrease in fissionable isotopes and the buildup of fission product inventory. The physical burndown effect is greater than the rod bowing effects which would be calculated based on the amount of bow predicted at those burnups. Therefore, for the purpose of evaluating effects of rod bow on Westinghouse fuel, 24,000 MWD /MTU represents the maximum burnup of concern.
For all transients, the bowed and unbowed MONBR results are well above the allowable 1.17 limit. Thus, no reduction in allowable reactor peaking is required as a result of a change in MDNBR due to rod bow.
3.2
en o st 3.5 Safety Limit Curves Safety limit curves for Prairie Island are given in Technical Specifications section5.1. These curves define the region of acceptable eperation in terms of core average temperature, power, and pressure. One of these limits is the thermal overtemperature limit, which prevents cladding damage based on DNB considerations. Due to.the fact that DNB' ratios are a function of core loading, the applicability of these' limits will be verified by NSP on a cycle specific basis. ,
The ' thermal over-temperature limit is imposed on the reactor by the over-temperature AT reactor trip. This trip function is designed to trip the reactor before it exceeds the limits defined in the Technic'al Specifications in order to prevent DNB induced cladding damage. This trip function consists of two parts,-AT trip: AT 0T 0T setpoint-f(AI).
.NSP has evaluated both components of the overtemperature AT trip function and found that they remain valid for operation of Prairie Island with an FAH . hat follows the equation.
FAH(P) = 1.70 [1 + 0.3 (1-P)] P s 1.0
= 1.70 P > 1.0 where P = fraction of rated power (1650 MWth) l l
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3.3
TABLE 3.1'
- Prairie Island Thermal Hydraulic Reference Conditions Reactor-Conditions -Nominal Rated Core Power (MWt) 1650(100%)
Total Reactor Flow Rate (Mlb/hr) 68.62 Active Core Flow Rate (M1b/hr) 64.50 Core Coolant Inlet Temperature ( F) 530.5 Core Pressure-(psia) 2250.0 Power Distribution Overall Peaking (Fg ) 2.50 Radial x Local (FAH) 1.70 Engineering Factor 1.03 l
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d 4.0: ACCIDENT AND TRANSIENT ANALYSIS This safety evaluation was performed to help answer the following questions that must be -addressed as part of all safety evaluations (Reference 8).
- a. Does it create a possibility fer an accident or malfunction of-a different type than evaluated previously in.the USAR or subsequent commitments?
b, Does it increase the probability of occurence of an accident or malfunction.of equipment important'to safety previously analyzed in the USAR or subsequent commitments?
- c. Does it increase the consequences of any accident or malfunction of equipment important to safety previously analyzed in the USAR or subsequent commitments?
- d. Is the margin of safety defined -in the bases for any Technical Specification reduced?
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As part of all safety evaluations, NSPNAD evaluates all accidents in the USAR to determine if the current analyses bound the reload design. All accidents that are not bounded by previous analysis are reanalyzed and the results are presented in the safety evaluation.
The Tech Specs are also reviewed to determine if any changes are necessary.
If a change -is ~made, the supporting ' analysis is also presented in the RSE.
~4.1 Plant Transient Analysis This d9:uments the NSP Nuclear Analysis Department's analysis of plant operationai transients for Prairie Island.
4.1
Westinghouse is performing new LOCA/ECCS analyses for Prairie Island. The new FQ and FAH limits will be 2.5 and 1.7. In order to support the Technical Specifications changes required for the new limits, NSPNAD has analyzed the spectrum of bounding transients for Prairie Island. This analysis will be submitted along with the proposed Technical Specification changes as part of the safety evaluations.
The analysis.shows that the limiting coadition II and III transient, the dropped rod at EOC has a calculated MDNBR of 1.445. With a rod bow penalty, the calculated MDNBR is still well above the design limit of 1.17.
In addition to the limiting condition II and III event described abeve, the Locked Rotor accident, a condition IV event, shows a calculated MDNBR < 1.17. The number of fuel rods which would potentially experience ?NB in the accident is calculated to be 1.54%. This is well below the acceptance criteria of less than 20% failures.
For all transients, the maximum pressurizer pressure is less than 2750 psia.
The latter pressure corresponds to 110% of the design pressure of 2500 psia.
A summary of transient margins is shown in Table 4.1.
4.1.1 Input Parameters The steam line breaks are initiated from hot shutdown conditions. All other transients are initiated from 102% of full power conditions. For full power operation, an axial peaking factor, FZ, f 1.428 located at X/L = 0.56, and a total peaking, Fg, of 2.50 is assumed. Other thermal hydraulic parameters for full power operation are summarized in Table 4.2.
Neutronics parameters calculated for this cycle are given in References 12 and 13. Reactor trip setpoints for prairie Island Units 1 and 2, along with setpoints and delay times used in the analysis, are given in Table 4.3.
The setpoints used in the analysis are essentially the same as those in the FSAR analysis. In all cases, the setpoints used in the analysis bound the actual setpoints for the Prairie Island plants to account for instrumentation errors and uncertainties.
4.2
4.1.2 Transient Analysis Results 4.1.2.1 Fast Control Rod Withdrawal This transient assesses plant response to a control rod withdra.wal, with a reactivity insertion rate of 8.2E-4 AK/sec, from full power.
All automatic reactor control systems are assumed inoperable.
The transient response of the NSSS for this case is shown in Figure 4.1. The reactor trip is generated on high neutron power (setpoint at 118%) at .96 seconds. The pressurizer pressure rises to 2284 psia at 4.16 seconds. The vessel average temperature rises by about 1.0 F at 3 seconds and then drops off. The DNB ratio drops from its initial value of 1.90 to a minimum of 1.783 at 2.0 seconds after the start of the transient.
l l The acceptance criteria for this transient are that the minimum DNBR be not less than 1.17 and that the maximum reactor coolant and main steam system pressure not exceed 110% of their design values. This
! transient meets all acceptance criteria.
4.1.2.2 Slow Control Rod Withdrawal This transient assesses plant response to a control rod withdrawal with a reactivity insertion rate of 3.31E-5 AK/sec during full power l operation. The reactivity insertion rate was selected to minimize DNBR during the transient. All automatic reactor control systems are assumed inoperable.
The transient response of the NSSS for this case is shown in Figure 4.2.
The reactor trip is generated on overpower AT at 45 seconds. The pressurizer pressure rises to 2432 psia at 47.5 seconds. The vessel average temperature rises by less than 7.5 F and then drops off. The DNB ratio drops from its initial value of 1.90 to a minimum of 1.566 at 45.1 seconds after the start of the transient.
4.3
3 ;,1 The acceptance criteria are that the minimum DNBR be not less than or 1.17 and that the maximum reactor coolant and main steam system pressure not exceed 110% of their design values. This transient meets all acceptance criteria.
4.1.2.3 Loss of External Electric load This transient considers plant response from full power when a loss of load results in.a turbine trip. Simultaneous reactor trip initiated by.the turbine stop valves is conservatively neglected. Rather the reactor is scrammed later in the transient by the pressurizer overpressure trip signal. All automatic reactor control systems, as well as the steam renerator relief valves, are assumed inoperable.
Steam dump and bypass e.e also neglected.
The transient response of the NSSS for this case is shown in Figure 4.3. Atl the start of the transient, the turbine stop valves-close and the secondary side pressure rises rapidly to the safety valve setpoint at 15 seconds and is limited to that pressure by relief through the safety valves. The primary system pressurizes rapidly due to the loss of heat sink and the reactor 11s scrammed at 5.5 seconds on a high pressurizer pressure trip signal. pressurizer safety valve opening occurs and a peak pressure of 2500 psia is calculated at 6.75 seconds. Core power remains relatively constant up until the time of rehctor trip.- Because of the p mary system pressuriration, the DNB ratio increases and remains above its initial 1.90 value.
The acceptance criteria for this transient ara that the cinimum DNBR be not less than 1.17 and that the maximum reactor coolant and main steam system pressure not exceed 110% of their design values. This transient meets all acceptance criteria.
4.4
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,e "4.1.2.4 Oropoed Rod - Auto Control In this transient a full length RCCA is. assumed to be released by the stationary gripper coils and to fall into a fully inserted position in the core, A dropped RCCA typically results.in a reactor trip signal due to the power range negative neutron flux rate circuitry. The core power distribution, from an absolute value point.of view, is not adversely effected during the short interval prior to reactor trip.
The drop of a single ROCA may or may not result in a negative flux rate reactor trip. If a trip does not occur, a single failure of the controller circuitry can cause a transient power overshoot. The power overshoot combined with the higher peakir.g factors associated with a dropped rod could conceivably challenge the MDNBR limit. The transient response of the NSSS for this case is shown in Figure 4.4.
The MONL'R drops t rom its initial value of 1.497 to 1.445 at 10 minutes.
A maximum of pre surizer pressure of 2244 psia occurs at 217 seconds.
The acceptaIce c*iteria for this transient are that the minimum DNBR be n L i.:s tnan 1.17 and that the maximum reactor coolant and main steam pressure not exceed 110% of their design values. This transient meets all acceptance criteria.
4.1.2.5 Loss of Reacter Coolant Flow - 2/2 pump Trip This transient considers the loss of reactor coolant flow associated with the simultaneous coastdown of both primary system coolant pumps.
Following the loss of two pumps at power, a reactor trip is actuated by either low voltage or cp n pump circuit breakers since the incident
- is due to the simultaneous loss of power for all pump buses. Both the low voltage and pump breaker reactor trip circuitry meet the single failure criteria and therefore cannot be negated by a single failure.
The time from the loss of power to all pumps to the initiation of control rod assembly motion to shutdown reactor is taken as 2.1 seconds. This is a conservative assessment of the delay. All automatic reactor control systems are assumed inoperable, 4.5
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The transient response of the NSSS for this case is shown in Figure 4.5.
The MONBR drops from its initial valoc of 1.90 to a minimum of 1.583 at 3.5 seconds into the transient due to an increase in the power to flow ratio. A maximum pressurizer pressure of 2333 psia is calculated to occur at 5.8 seconds into the transient.
The acceptance criteria for this transient are that the minimum DNER be
, not less than 1.17 and that the maximum reactor coolant and main steam system pressure not exceed 110% of their design values. This transient meets all acceptance criteria.
4.1.2.6 Locked Pump Rotor The locked pump rotor transient is a Class IV event that considers plant response from full power operation when one of the two primary coolant pumps is postulated to abruptly seize. Reactor scram and trip of the feedwater pumps due to low prfmary coolant flow is conservatively assumed to occur at 0.9 second after pump seizure.
All automatic reactor control systems are assumed inoperable.
The plant response for this transient is shown in Figure 4.6. The calculated MDNBR drops below 1.17 at approximately 1.4 seconds after pum; seizure. The duration of time for which the ONBR is less than 1.17 is roughly 2.00 seconds. The number of fuel rods statistically calculated to experience ONB for this Class IV transient is 1.54%. A maximum pressurizer pressure of 2501 psia is calculated to occur at 3.35 seconds into the '
transient. A maximum cladding temperature of 1940 F is calculated to occur at 3.53 seconds into the transient.
The acceptance criteria fer the locked rotor analysis are as follows:
- 1. The maximum reactor coolant and main steam system pressures must not exceed 110% of the design values.
- 2. The maximum 'ad tampe ature calculated to occur at the core hot spot must not exceed 2750 F.
This transient meets al' acceptance cetteria.
4.6
4.1.2'.7 Main Steamline Break
'This transient is bounded by the analysis for PI 2. Cycle 10. .The
- results of this analysis are shown in Reference 11.
_This transient meets all acceptance criteria.
4.2 Rod Ejection Analysis l
A Control Rod Ejection Accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a Rod Cluster Control Assembly (RCCA) and drive shaft. The consequence of this mechanical failure is a rapid reactivity insertion together with an adverse l core power distribution, possibly leading to localized fuel rod damage.
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The rod ejection accident has been evr.luated with the procedures developed in Reference 1. The ejected rod worths, hot pellet peaking factors, delayed I
neutron fractions and Doppler coefficients were taken as conservative values which bound the Prairie Island analysis.
The pellet energy deposition resulting from an ejected rod was evaluated explicitly at HFP and HZP initial conditions. The HFP pellet energy deposition was calculated to be 165 cal /gm. The HZP pellet energy deposition was calculated to be 53 cal /gm at 800 and 97 cal /gm'at EOC. The rod ejection accident was found l to result in energy deposition of less than the 280 cal /gm limit as stated in Regulatory Guide 1.77. The significant parameters for the analysis, along with the results are summarized in Tables 4.4 and 4.5.
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TABLE 4.1 ,
Summa ry of Pra i rie island Transient Ma rains (Calculated Va lue/ Acceptance Criteria )
Pressure (psia )
T rans ient MONBR PCS MSL # Faited Pins Clad Temp. Fuel Enthalpy
(%) (F) (cal /ga)
Rod Wi thd rawa l et Power Fast 1.78S/1.17 2?3'4/2750 883/1210 - - -
Slow 1.566/1.17 L432/2750 1087/1210 - - -
Turbine Trip 1.90/1.17 2500/2750 1108/1210 2/2 Pump Trip 1.606/1.17 2320/2750 1067/1210 - - -
Locked Rotor -
2501/2750 1089/1210 1.54 1940/2750 -
Dror ped Rod 1.445/1.17 2244/2750 759/1210 - - -
MSL Brea k*+ -
2250/2750 1046/1210 0 NC/2750 -
Ejscted Rod-HZP,BOC -
2351/3000 -
NC 1091/2750 53/280
-HZ P, E OC -
2385/3000 -
NC 1660/2750 97/280 Ejected Rod-HIP -
2371/3000 -
NC 2334/2750 165/280 NC - Not calculated for each cycle
- - Iso a Toch. Spec. l imi t or peak conta inment pressure 46 psig, which is not calculated.
+ - Pt 2 Cycle 10 Results 4.8
TABLE 4.2 Parameter Values Used in Full Power Transient Analysis Analysis i Input Value Core Total Core Heat Output, Mw (102%) 1,683.0 Heat Generated in Fuel, % 97.4 System Pressure, psia 2,220 Hot Channel Factors Total Peaking Factor, Fg T 2.50 Enthalpy Rise Factor, FAH .O Total Coolant Flow, Ib/hr 68.62 x 10 6 6
Effective Core Flow, lb/hr 64.50 x 10 Reactor Inlet Temperature, F 534.5 5 team Generators Calculated Total Steam Flow, Ib/hr 7.23 x 10 6 Steam Temperature, F 510.8 Feedwater Temperature, F 427.3 Tubes Plugged, % 5.0**
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- Locked Rotor is initiated from 2280 p3ia
- MSL break conservatively assumes no plegging 4.9
TABLE 4.3 ,
Pra i ri e Island Units 1 and 2 Trip Setpoints Setpoint Ustd in Analysig Delay Timo High Pleutron tlux I
(High Setpoint) 108% 118% 0.5 sec (Low Setpoint) 40% 50% 0.5 sec l t ow Reactor Coolant flow $3% 87% 0.6 sec liigh Pressurizer Pressure 2388 psia 2425 psia 1.0 sec t ow Pressurizer Pressure 0815 .sia 1700 psia 1.0 sec High Pressurizer Water Level 85% or Soan 100% or Span 1.5 see Low-Low Steam Generator 13% of Span 0% of Span 1.0 sec Wa te r 8 ave l Ove rc empe ra ttero I* TAVEo = $67.3F TAVEo = $67.31' 6.0 see Po = 2250 psia Po = 2250 psia ovo rpowe r T** TAVEo = 567.3 TAVEo = $67.3 6.0 sec liigh Pre'sure Sa fety injection 188:2 psia 1800 psia 10 sec Negativo Neutror. IIu. Rato 5% / 2 sec 7% / 2 sec O.1 sec The overtempera ture T trip is a reinction or pressurizer pressure, coolant average temperatesre, and axial orrset. The TAVEo and Po setpoints are contained within the functional re la t ionsh ip.
- The overpower T trip is a function of coolant averagt. temperature and axial orrset. The TAVEo setpoint is contained within the functional relationship.
4.10
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TABLE 4.4' :s, Prairie Island Ejected-Rod Analysis, HFP Maximum Control Rod Worth (pcm) 193 Doppler Defect-(0-100% Power,:pcm) 1138 Shutdown Morgin (pem) 1030 Delayed Neutron Fraction 0.004775
~ Power Peaking Factor, Fg 3.839
. Energy Depositior, (cal /gm) 165
< TABLE 4.5 Prairie Island Ejec+ed Rod Analysis, HZP BOC EOC Maximum Control Rod Worth (pcm) 569 655 Doppler Defect (0-100% Power, pcm) 1303 1871 ShutdownMargin(pcm) 2668 2111 Delayed Neutron Fraction 0.005721 0.004828 Power Peaking Factor, _Fg 5.830 9.299 Energy Deposition (cal /gm) 53 97 4.11
Figure 4.1 Fast Rod Withdrawal K Effective Absolute Power i.0 i - ii0 i , : .. ..
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0 e 2 s 4 s 0 t 2 3 4 nm. (s.condi) nm. (s.condi)
Core Average Heat Flux Pressurizer Pressure 110 2300 . .
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1 i Yessel Average Temperatere MDNBR CWRB-D 468 3
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l l
-- o-- m- -
Figure 4.2 Slow Rod Withdrawal K Effective Absolute Power i.0i ise 3
. . . . . i40-C- h-
- : M : :
0,,. .. . . . . . .p . . . . .. .... .. .. 50 . . . ...
j l- l l :
0.00 0 0 20 40 40 0 20 40 40 n.. (s.condi) Tim. (3.conds)
Core Average Heat Flux Pressurizer Pressure
= '
iOO_ . .. ....:... ... . .. . . . . . . .. 3 00_ . . . ..
w : -< . .
. m .
- a. . .
00_ .. . . . . . . .. . . . . . . . . . . .. 2:00 .. ... . . . . . . . . . . . . . . . . . . .
j . y : :
0 ::00 0 30 40 40 0 20 40 40 nm. (secondi) Tim. (seconds)
Core Inlet Temperature MONBR (WRB-D 570 4 000_
...........7 O. 3 .. ... . . .. ..
s e .
e .
- ! * . i e 0,0 . . . .....>.. . . . . . ..<.. . . . . . .. - .
= : : = .
e e .
G : .
Z . .
O l g Oc0 . . . .
5 E3 030 1 0 30 40 40 0 20 40 40 l nm. (s.condi> tm. (s.condi) !
4.13
l a
i l
1 Figure 4.3 Turbine Trip :
K Effective- Absolute Po u -a 1.01 150 --
- : . l '
g . . . . . . .... .. :........ .
l00 .
3: :
- l : :
. . . M .
0.99- * ' + ' * * * * * * * * * * * * * * * * * $0- ' ** * * * * ** - **
- O.99 0 0 10 20 30 40 50 0 10 to 30 40 50 n.. (s.condi) na. (sn ndi)
Core Average Heat Flux Pressurizer Pressure ISO 2600 iOO_ .... . . . . . . . . .. . . . . . ... . . . . . . . . . . 3400 ... . .
x - -
A . . . .
.0 . . . . . . . . . . . . . . . . . . . . . . 3:00 . . . . . . . . . .. ... .. . . . . . .
I i ! ! : i %
0 2000 l 30 40 50 0 10 TO 30 40 50 0 10 to n.:.. (s.condi) nm. (s.Gondi)
Core Inlet Temperature MDNBR (WRB-D sr0 :
.00_ ......,.......,. .... ... .. . . . . . . :
ll
= ,
E : : . .
I e ! : : : * ,_
! : i e . e_ . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . .
- I II ....... EBEE
.. .. .. . . . E E M m E i3 ra 3 5. . ...
d :
2 : . .
. . . o,, .
ist. . i. . ..:..
1 .
l .
530 1 50 s 10 i e le to 30 40 0 2 4 4 ns. (s.conds) nm. (secondi) 4
Figure 4.4 Dropped Rod - EOC X Ef fective Absolute Powe.-
l -
i j i_ ;_
iOS. f.....
~ . . .:. .
x -
,0, . . .
...4.. .......:.. . . ..
0,,,, . . . . . ... . . . + -. . . . . . . . . . . . . . . . . . . . . .
0.9990 to 400 400 0 100 400 600 9 300 n= e (s.conds) nm. (s.conds)
Core Average Heat Flux Pressurizer Pressure 105 2260 at. .. .........;..... . . .. .. . . . . . ::40 . . . . . .. . . . . . . . . . . . .
! :- i Ni x . .
. . m . .
(g. ,, .... ..........y.... . .- .. . . . . 3130- .... . . . . . . . . . . . . . . . . . . . . .
90 2200 400 600 0 200 400 600 0 300 nm.(s.cende) nm. cs.cond )
Vessel Averogo Temperature MDNBR (WRB-D
$60 f.4
. l 1
- 15E E E E E E5501
. e i,4 .
o .
3 -
- fee- * ... ++ + *** * + *i- .. * *
=
ei : : .
e a O -
Z .
. O t.3 - = ~ < =
$ I . .
r 964 i 400 400 0 200 400 600 0 100 noe (second ) nm. (seconds) 4.15
Figure 4.5 2/2 Pump Trip X Effective Absolute Power i.o 1so 3....
- : : x : : . :
l .
0.99- * * * * * . So- * *
- i i :
0.94 0. .
0 20 40 60 so too o to 40 40 80 100 Time (Seconds) Tim e ($* con J )
Core Average Heat Flux Pressurl er Pressure 180- .
23s0 j
- l I . . !
l 2300- J- + . :- - * * . .- *t .i-I ;
io . . . . , . ... .
y *
< ssge- , . . ...i. ...i.. nyn. .ul.n. . .
- G '
- o. .
l l 50- " "+ :"*"." " -
2200- .- . . . . " * * .
, . -. : . +
e tiso 48 60 80 80 0 0 20 40 40 80 tot O 20 Time (Seconds) Time (Seconds)
Core Flos MDNBR (WRB-D se : El ll
- E E
.o. . . . ..}. .
= 5. m m
. i g, . _ . . as ... . . . . . ..m.
=
m E.... .
$ l .
. 3 E -
N . . .
.. . ..f-a E 40- * *
- 6
.i... - . i. . m e
+
E
- a
( . . Z .
. . . .E3 B E.E ..
. o i... .
t .
- }+
10- * + + . . ,
N. m I'd O
o so "e4 so so too e 4 : : 4 s use (Seconds) Time (Seconds) 4.16
Figure 4.6 Locked Rotor K Effective Absolute Power 1.01 15 0
- . l 3 l !
g ...
.. .. get. A. .
l l l .
. M :
- - +- + +- - +-
g,0g_ . . . . . . . .... .. .. . .. 30- .
0.00 . .
0 .
4 6 8 10 0 2 4 4 4 10 O A Due ($econdt) hme (Seconds)
Core Average Heat Flux Pressurizer Pressure l2 0 1600 i00 .. . . . . . . . .
2000- t- - t- -
.t .
t-40- . . . . . . .. . . . . . , , ,
y ; -< 2400- - F ** l-
.) - -
.1 + -
. w . .
O- . .
Q4- ... . . . . . .. .. .. . . .
. : . 2300- *- . - .
.; ...1 40- *** * * - * *
- - / j ,
- I
- i : ! ! !
10 2200 0 2 4 6 4 10 0 2 4 e l 10 Dee ($econds) Tim e (Seconds)
Core Flow MDNBR CWRB-D at 3.s
= ,
+-
40- + + - - ++
- - - e 3- - - - - - - -
. . m : .
= : :
ll .
s, . .
a a
as m E ,
U
(
z : E g 3-40 . . . . . . .. . .. . . . . .
g i... 5 .- - - - -
. g g "E .
N5 E -
5 E= E
_E 1
20 4 4 10 0 1 2 3 4 5 0 2 4 De e (Seconds) Dee (Seconds) 4.17
,, w 1,e 15.0- REFERENCES 1)~ NSPNAD-8102P Rev 5, "Reload Safety Evaluation Methods for Application to PI Units",
1 March 1987.
- 12) ' Prairie Island-Final Safety Analysis Report.
1
- 3) XN-NF-78-35, "Plant Transient Analysis for the Prairie Island Nuclear Plant, Units 1 and 2", November 1978.
4)_ XN-NF-80-60, "Plant Transient Analysis for Prairie Island Units 1 and 2 with ENC TOPROD Fuel", February 1981.
.5) Westinghouse' Letter 85NS*-G-015, "Hydraulic Compatability of Westinghouse 14x14 0FA' and ENC TOPROD Fuel Assemblies " June 25, 1985.
6). WCAP-8691 Rev 1, "Fuel Rod Bow Evaluation," July 1979.
- 7) XN-NF-80-61, "Prairie Island Nuclear Plants TOPROD Safety Analysis Report", Rev 1, March 1981. ;
- 8) Northern States Power Company Corporate Administrative Work Instructions, N1AWI 5.1.9, "Safety Evaluations."
- 9) - Deleted -
- 10) - Deleted -
- 11) NSPNAD-8504P, Rev.1, "Prairie Island Unit 2-Cycle 10 Final Reload Design Report,"
October 1985.
i
- 12) Internal Correspondence, T.L.Breene to C.A.Bonneau, "Transfer of CAS Input to SAS:
Core Physics Parameters Applicable to the Prairie Island Unit 1 Cycle 13 Reload Safety Evaluation," April 8, 1988.
- 13) Internal Correspondence, T.L.Breene to C.A.Bonneau, "Transfer of Additional CAS Core Physics Parameters Applicable to the Prairie Island Unit 1 Cycle 13 Reload Safety Evaluation," April 26, 1988, 5.1
Exhibit E
.s Prairie Island Nuclear Generating Plant License Amendment Request Dated July 5,1988 Demonstration of the Conformance of l
Prairie Island Units 1 and 2 to Appendix K and 10CFR50.46 for Large break LOCAs
_ _ _ _ _ _ _ __ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ .-