ML20212R689
| ML20212R689 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/26/1987 |
| From: | Kapitz J NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20212R565 | List: |
| References | |
| NSPNAD-8705, TAC-65078, TAC-65079, NUDOCS 8704270325 | |
| Download: ML20212R689 (45) | |
Text
{{#Wiki_filter:. + Exhibit D Prairie Island Nuclear Generating Plant. License Amendment Request Dated April 13, 1987 P Safety. Evaluation of Increased FQ and FdH I NSPNAD-8705 l Prepared by: Nuclear Analysis Department Northern States Power Company J i I I ( 8704270325 870413 PDR ADOCK 05000282 - p PDR 4 l
. _ _ , ~ . , .
PRAIRIE ISLAND UNITS 1 AND 2 SAFETY EVALUATION OF INCREASED FQ AND FAH NSPNAD-8705 i Prepared by 6h d.J4 Date ) .2jV 8 7 Reviewed by (I / ws ca cA Date / f ~7 Approved by
, i. . .
Date J ,f /, $ Y 4 Page 1 of 44 1 _ _ _ _ - . . -- - , , . _ _ _ , , -. _ _,_ _ - ~ . _ . , _ _ . , . , _ _ - . . _ . _ _ _ _ _ _ ,
r LEGAL NOTICE This report was prepared by, or 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 us 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.
Page 2 of 44
. o. .
TABLE OF CONTENTS Pagg
1.0 INTRODUCTION
8 2.0 CALCULATIONAL MODELS AND METHODOLOGY 9 2.1 Calculational Models 9 2.2 Methodology 9 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS -11 3.1 Design Criteria 11 , 3.2 Core Hydraulic Compatability 11 I 3.3 Thermal Margin 12 3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance 12 3.5 Safety Limit Curves 13 4.0 ACCIDENT AND TRANSIENT ANALYSIS 15 4.1 Plant Transient Analysis 15
- 4.1.1 Input Parameters 16 4.1.2 Transient Analysis Results 17 4.1.2.1 Fast Control Rod Withdrawal 17 4.1.2.2 Slow Control Rod Withdrawal 17 l 4.1.2.3 Loss of External Electric Load 18 l 4.1.2.4 Dropped Rod - Auto Control 19 2 4.1.2.5 Loss of Reactor Coolant Flow 19 4.1.2.6 Locked Pump Rotor 20 4.1.2.7 Main Steam Line Break 21
] 4.2 Rod Ejection Analysis 21 4 i Page 3 of 44
. e. .
LIST OF TABLES - P_a21 3.1 Prairie Island Thermal Hydraulic Reference Conditions .14 4.1 Summary of Prairie Island Transient Margins 22 4.2 Parameter Values Used in Full Power Transient Analysis 23 4.3 Prairie Island Units 1 and 2 Trip Setpoints 24 4.4 Prairie Island Ejected Rod Analysis .25 i } l i i I i 1 1 ) i Page 4 of 44 i 1
w a:. . LIST OF FIGURES P_alL* 4.1 Fast' Rod Withdrawal - K-effective 26 - 4.2 Fast Rod Withdrawal - Absolute Power .26 4.3 Fast Rod Withdrawal - Core Average. Heat Flux 27 4.4 Fast Rod Withdrawal - Pressurizer Pressure' 27 4.5 Fast' Rod Withdrawal - Vessel 4verage Temperature 28 4.6 Fast Rod Withdrawal - Minimum DNB Ratio 28 4.7 Slow Rod Withdrawal - K-effective 29 4.8 Slow Rod Withdrawal - Absolute Power -
-29 4.9 Slow Rod Withdrawal - Core Average Heat Flux 30 4.10 Slow Rod Withdrawal - Pressurizer Pressure 30
? 4.11 Slow Rod Withdrawal - Core Inlet Temperature 31 ,. i
- i. 4.12 Slow Rod Withdrawal - Minimum DNB Ratio 31
.g j 4.13 Turbine Trip - K-effective 32 1
l 4 4.14 Turbine Trip - Absolute Power 32 4.15 Turbine Trip - Core Average Heat Flux 33 , i l l ' \ l 1 l t 1 j 'Page 5 of 44
-. r - , . .
1 LIST OF FIGURES Pagg 4.16 Turbine Trip - Pressurizer Pressure 33 4.17 Turbine Trip - Core Inlet Temperature 34 4.18 Turbine Trip - Minimum DNB Ratio 34 4.19 Dropped Rod - K-effective 35 4.20 Dropped Rod - Absolute Power 35 4.21 Dropped Rod - Core Average Heat Flux 36 4.22 Dropped Rod - Pressurizer Pressure 36 4.23 Dropped Rod - Vessel Average Temperature 37 4.24 Dropped Rod - Minimum DNB Ratio 37 4.25 2/2 Pump Trip - K-effective 38 4.26 2/2 Pump Trip - Absolute Power 38 4.27 2/2 Pump Trip - Core Average Heat Flux 39 4.28 2/2 Pump Trip - Space Pressurizer Pressure 39 'l 4.29 2/2 Pump Trip - Core Flow 40 4.30 2/2 Pump Trip - Minimum DNB Ratio 40 I Page 6 of 44
LIST OF FIGURES Pagg 4.31 Locked Rotor - K-effective 41 4.32 Locked Rotor - Absolute Power 41 4.33 Locked Rotor - Core Average Heat Flux 42 4.34 Locked Rotor - Pressurizer Pressure 42 4.35 Locked Rotor - Core Flow 43 4.36 Locked Rotor - MDNBR (WRB-1) 43 i I l , 1
- l l
l Page 7 of 44
- - - - - i- - - , *w y 3 , *, ry
? . e. l
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. NSPNAD 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. This analysis covers both the Westinghouse OFA fuel and ENC TOPROD fuel. Also, the acceptability of the new limits to Prairie Island Unit 2 Cycle 11 has been verified. 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. I 1 Page 8 of 44 l
-2.0 CALCULATIONAL MODELS AND METHODOLOGY 2.1 Calculational Models These 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 Methodology 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.
Page 9 of 44
These transients have not been limiting in the past. The new Technical 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. The analysis has shown that while the new FQ and FAH will affect the initial steady state MDNBR, it will not change the relative change in MDNBR. NSPNAD has performed Prairie Island transient analysis at several different values of FQ and FAH without seeing any change in the relative severity 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 new Technical Specifications on FQ and FAH. l Page 10 of 44 l
3.0 THERMAL HYDRAULIC DESIGN ANALYSIS This section provides results of the thermal hydraulic design analyses for-Prairie Island. 3.1 Design Criteria The thermal and hydraulic design performance requirements for Prairie Island fuel are as follows.
- 1. The minimum departure from nucleate boiling ratio (MDNBR) 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 red bowing. 1.17 at overpower 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 : l 3.2 Core Hydraulic Compatibility The hydraulic compatibility of the Prairie Island reload fuel is discussed in Reference 5. Page 11 of 44
3.3 Th:rmal Margin The most limiting transient for Prairie Island is the slow rod withdrawal event. The minimum DNBR was calculated to be 1.576, using the WRB-1 correlation. Table 3.1 provides reference conditions for the analysis. Details of the plant transient ana13 sis for Prairie Island are given in Section 4.0. 3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance The calculation of the DNBR 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 = MDNBRNB 8(1-6 ) MDNBRB = MDNBR for bowed fuel MDNBRNB = MDNBR for nonbowed fuel 6B = fraction reduction in MDNBR due to rod bowing where 6B 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 B 33,000 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, 33,000 MWD /MTU represents the maximum burnup of concern. For all transients, the bowed and unbowed MDNBR 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. . Page 12 of 44
3.5 Safety Limit Curves Safety limit curves for Prairie Island are given in Technical Specifications section 2.1. These curves define the region of acceptable operation 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 Technical Specifications in order to prevent DNB induced cladding damage. This trip function consists of two parts, AT0Ttrip = AT OT 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 that 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)
Page 13 of 44
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 (Mlb/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 i > Page 14 of 44 if
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 for 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?
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 4 This documents the NSP Nuclear Analysis Department's analysis of plant operational transients for Prairie Island. , I l 5 1 Page 15 of 44 4
- . , - - _ . - . .m_ _ _ , _ _ _ . _ _ . _ __ . _ _ _ _ _ . . _ .
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 condition II and III transient, the slow rod withdrawal, has a calculated MDNBR of 1.576. 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 above, the Locked Rotor accident, a condition IV event, shows a calculated MDNBR < 1.17. The number of fuel rods which would potentially experience DNB in the accident is calculated to be less than 1%. 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 shovn 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, F ,Z f 1.428 located at X/L = 0.56, and a total peaking, F g, of 2.50 is assumed. Other thermal hydraulic parameters for full puwer operation are summarized in Table 4.2. Neutronics parameters calculated for this cycle are given in References 9 and 10. 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 j the FSAR analysis. In all cases, the setpoints used in the analysis l bound the actual setpoints for the Prairie Island plants to account for instrumentation errors and uncertainties. l l Page 16 of 44 ;
4.1.2 Transient Analysis Results 4.1.2.1 Fast Control Rod Withdrawal This transient assesses plant response to a control rod withdrawal, 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 Figures 4.1 through 4.6. The reactor trip is generated on high neutron power (setpoint at 118%) at .99 seconds. The pressurizer pressure rises to i 2280 psia at 4.1 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 1.914 to a minimum of 1.802 at 2.0 seconds after the start of the transient. 3 The acceptance criteria for this transient are that the minimum DNBR 4 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 4.51E-5 AK/sec during full power operation. The reactivity insertion rate was selected to minimize DNBR during the transient. All automatic reactor control systems are assumed inoperable. f The transient response of the NSSS for this case is shown in Figures 4.7 through 4.12. The reactor trip is generated on overpower AT at 36.1 seconds. The pressurizer pressure rises to 2427 psia at 44.4~ 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 1.914 to a minimum of 1.576 at 42.2 seconds after the start of the transient. , Page 17 of 44
The acceptance criteria 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. 4.1.2.3 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 generator relief valves, are assumed inoperable. Steam dump and bypass are also neglected. The transient response of the NSSS for this case is shown in Figures 4.13 through 4.18. At 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 is scrammed at 5.5
- seconds on a high pressurizer pressure trip signal. Pressurizer safety valve o'pening occurs and a peak pressure of 2500 psia is calculated at 6.8 seconds. Core power remains relatively constant up until the time of reactor trip. Because of the primary system pressurization, the.DNB ratio increases and remains above its initial 1.914 value.
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. 1 i l Page 18 of 44
4.'1.2.4 Dropped 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 > 0 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 RCCA 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 peaking factors associated with i a dropped rod could conceivably challenge the MDNBR limit. The transient response of the NSSS for this case is shown in Figures i 4.19 - 4.24. The MDNBR drops from its initial value of 1.761-to 1.742 at 10 minutes. i A maximum of pressurizer pressure of 2229 psia occurs at 80 seconds. [ 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. pressure not exceed 110% of their design values. This transient meets all acceptance criteria. 4.1.2.5 Loss of Reactor Coolant Flow - 2/2 Pump Trip , This transient considers the loss of reactor coolant ~ flow associated l 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 open purep 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. { l The time from the loss of power to all pumps to the initiation of j control rod assembly motion to shutdown reactor is taken as 2.1 j seconds. This is a conservative assessment of the delay. All I automatic reactor control systems are assumed inoperable. l l Page 19 of 44
The transient response of the NSSS for this case is shown in Figures 4.25 through 4.30. The MDNBR drops from its initial value of 1.914 to a minimum of 1.630 at 3.0 seconds into the transient due to an increase in the power to flow ratio. A maximum pressurizer pressure of 2322 psia is calculated to occur at 5.5 seconds into the transient. 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. 4.1.2.6 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 primary coolant flow is conservatively assumed to occur at 0.9 second after pump seizure. I All automatic reactor control systems are assumed inoperable. The plant response for this transient is shown in Figures 4.31 to 4.36. The calculated MDNBR drops below 1.17 at approximately 1.1 seconds after pump seizure. The duration of time for which the DNBR is less than 1.17 is less than 2.5 seconds. The number of fuel rods statistically calculated to experience DNB for this Class IV transient is less than 1%. A maximum pressurizer pressure of 2500 psia is calculated to occur at 3.8 seconds into the transient. The acceptance criteria for 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 clad temperature calculated to occur at the core hot spot must not exceed 2750 *F.
This transient meets all acceptance criteria. Page 20 of 44 4
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 ejsetien 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 core power distribution, possibly leading to localized fuel rod damage. The rod ejection accident has been evaluated with the procedures developed in Reference 1. The ejected rod worths, hot pellet peaking factors, delayed 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 148 cal /gm. The HZP pellet energy deposition was calculated to be 187 cal /gm. The rod ejection accident was found 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 Table 4.4. ! Page 21 of 44
TABLE 4.1 '. Summa ry o r P ra i r i e I s l a nd T ra n s i en t Ma rg i n s (Calculated Value/ Acceptance Criteria) Pressure (psia ) T rans ient MONBR RCS MSL # Failed Pins Clad Temp. Fuel Enthalpy (%) (F) (cal /gm) Rod Withdrawal . at Power Fast 1.802/1.17 2280/2750 882/1210 - - - Slow 1.576/1.17 2427/2750 1092/1210 - - - l Turbine Trip 1.914/1.17 2500/2750 1108/1210 2/2 Pump Trip 1.630/1.17 2323/2750 1067/1210 - - - Locked Rotor - 2501/2750 1085/1210 1 NC/2750 - Dropped Rod 1.742/1.17 2229/2750 757/1210 - - - MSL Break * - 2250/2750 1046/1210 0 NC/2750 - Ejected Rod-HZP - 2501/3000 - NC 863/2750 187/280 Ejected Rod-flFP - 2375/3000 - NC 756/2750 148/280 NC - Not calculated for each cycle
* - Also a Tech. Spec, limit or peak containment pressure 46 psig, which is not calculated.
Page 22 of 44 a
TABLE 4.2 Parameter Values Used in Full Power. Transient Analysis Analysis Input Value Core Total Core Heat Output, Mw (102%) 1,683.0 Heat Generated in Fuel, % 97.4 e System Pressure, psia 2,220 Hot Channel Factors T Total Peaking Factor, F 2.50 n i Enthalpy Rise Factor, F AH .O 6 ! Total Coolant Flow, lb/hr 68.62 x 10 i 6 Effective Core Flow,1b/hr 64.50 x.10 Reactor Inlet Temperature, F 534.5 4 Steam Generators Calculated Total Steam Flow, lb/hr 7.23 x 10 6
- Steam Temperature, F 510.8 Feedwater Temperature, *F 427.3
! Tubes Plugged, % 5.0** i
- Locked Rotor is initiated from 2280 psia
** MSL break conservatively assumes no plugging i
i i Page 23 of 44 1 l , i
i TA2LE 4.3 Prairie Island Units 1 and 2 Trip detpoints i Sejpoint Used in Analysis Delay Time' f , High Neutron Flux 108% 118% 0.5 sec t Low Reactor Coolant Flow 93% 87%. 0.6 sec ( l
~
High Pressurizer Pressure 2388 psia 2425 psia 1.0 sec Low Pressurizer Pressure 1915 psia 1700 psia 1.0 see High Pressurizer Water Level 85% of Span 100% of Span 1.5 sec Low-Low Steam Generator 13% of Span 0% of Span 1.0 sec Water Level
- Ove rtempe ra tu re T* TAVEo = 567.3F TAVEo = 567.3r 6.0 see Po = 2250 psia Po = 2250 psia Ove rpowe r T** TAVEo = 567.3 TAVEo = 567.3 6.0 see High Pressure Safety injection 1842 psia 1800 psia 10 sec-
{ Negative Neutron flux Rate 5% / 2 sec 7% / 2 sec 0.1 sec 1
- The overtemperature T trip is a function of pressurizer pressure, coolant average temperature, j and axial orrset. The TAVEo and Po setpoints are contained within the functional re la t ionsh i p.
i ** The overpower T trip is a function or coolant average temperature and axial ortset. The TAVEo setpoint is contained within the functional re l a t ionsh i p. I J Page 24 or 44 a e
^
" - - - + - __ o, _ _._ ___ _ _ _ _
~ . - - . . . ..
TABLE 4.4 I Prairie Island Ejected Rod Analysis HZP HFP Maximum Control Rod Worth (pcm) 781 234 Doppler Defect ('pcm) 1294 1294 Shutdown Margin (pcm) 1763 1763 I Delayed Neutron Fraction 0.004837 0.004782 Power Peaking Factor, F g 10.58 4.109 l l Energy Deposition (cal /gm) 187 148 l l i i Page 25 of 44 i
DYNODE-P Prairie Island 1 Cycle 12 Fast Rod Withdrawal Figure 4.1 K Effective
- i. _. . . .. -
...... 4.. ..>. .. ..
1,. . (s... a .> Figure 4.2 Absolute Power
. .j . . . . . . ..
9 l ... ,
. .i . .i.., . . .:. . .
4
. g;gp Tr.. (s.e.w )
Page 26 of 44
DYNODE-P Prairie Island 1 Cycle 12 Fast Rod Withdrawal Figure 4.3 Core Average Heat Flux Ile i.,_ .
. .g . . . i.. . . . ..
m ... . . . .. I so. .<.. ..>. . i.. . . to , , . e i . h . (s.e.nds) Figure 4.4 Pressuriser Pressure 3300 h ssee- 5.. . .
.t. . .
$ ee. . ! .. .
ssen- .<.. . we - g 8 I 2 3 4 9 h . (s. .ad ) Page 27 of 44
.. . - - ... ,_ ___ - - _-__ - _ _ , . . _ . - . - _ _ _ - _ _ . . - . - - = . - _ -
DYNODE-P Prairie Island 1 Cycle 12 Fast Rod Withdrawal ; Figure 4.5 Vessel Average Temperature see esp. . .<.. , .i. . 4 t.ess. o
. i es,. .<.. , ..s. . . ,..
ss4 , , , , e I 3 3 4 S Time (seconds) Figure 4.6 MDNBR (WRB-D s i ll l 1 5 ..,. . A E 3 - 3 . .g. . . l E E E lI ( E E E
,,._ . i. . . . ,.
. : p l
's i i 4 e Ti.. (s.condi)
- Page 28 of 44 l l
l l l
DYNODE-P Prairie Island 1 Cycle 12 Slow Rod Withdrawal Figure 4.7 K Effective
...o.
.j. . . . .i . . . . . . . . . .. ..
...... . . . . . . .c . ..
...... 4 . . . , .. .<.. . .
e.,rs , . e to so so 4o so ao Time (Seconds) Figure 4.8 Absolute Power 1 iso i.e. . . . . l
. : . I i
... . , . . . :. . . .p . . ..
1
. t
. l I
i n .e- . . . . . . . ... . i 1 4o. ..i. .
. i. . . . . .!. . ..
i- . . 4. . . . . .. e g e io to ao as so se Ti. . (s...nd.) Page 29 of 44 i
a DYNODE-P Prairie Island 1 Cycle 12 Slow Rod Withdrawal Figure 4.9 Core Average Heat Flux ISO io . .
..>.. .,. .<.. ..s. .
ee. . . . .. . . . . . . .u . . n so- - ~- -- EQ- ~ i ! 9 . , , , , e is so so do so ao Time (Seconds) Figure 4.10 Pressuriner Pressure 2450 , t400- k- >- -4.- - . . 3330- - - - + . -
-[ - +.
.(
- E
- 9300- ++ < * +
- i. . - -
i . 3350- * * + - * * * < +
-{
. . : l l
- .0 i e to to so 40 so as Ti.. (s.c..d.)
1 1 Page 30 of 44
DYNODE-P Prairie Island 1 Cycle 12 Slow Rod Withdrawal Figure 4.11 Core inlet Temperature sto e.g. . .: . . .
. . j. . . .
E sse- * ~ -
- e e .
o - e4e- j- .- t. sao , e to so so 4e so so Time (Sec.nds) Figure /.12 MDNBE (WRB-D 3
- i i
. . i 3.g. s , .i.. .
. .p . . . .
= .
e, : : m , 1
) : : )
- - . . . . /. .
m ll - z o a - 3 - g i.e - s .i- .(.. ea .
- [jgp
. i. i. 3. 4. ..
n.. ts.e.n4.) Page 31 of 44
DYNODE-P Prairie Island 1 Cycle 12 Turbine Trip
; Figure 4.13 K Effective
, ....4.. . . . , . . . . . . . . . . .
. i
- . : 1
' i ! !
. i. ... . . ..
1,.. <s... 4 3 ( I Figure 4.14 ' Absolute Power
. 1 i i
.+.... ,
)
r . . 1 ... . . : .. . .. ..:...... ..
- . : ggy
- i. ... ... .. ..
1,. . < s....s o l \ l Page 32 of 44 l 1 l l I \ l l
l
.4 DYNODE-P Prairie Island 1 !
Cycle 12 Turbine Trip Figure 4.15 Core Average Heat Flux , ne '
~ -
i.e. . ...>. . ... . . . .s. . . .
+ .
.7, . .. .
. . 9. . .
r -
... .. . : . . . c. . .. .
4o. .<.. ..>. . i so- , , . . . .:. . . e . . . . 8 10 30 30 40 90 g Tim. (sec.nds) Rgure 4.16 Pressuriner Pressure ts00 j sieo- .<.. ..>.. . t .
)
l 2400- . . . . .
. Q ..
\
l 4 9300- -
' ' . j.- . .h . .
- l
- 1 ases- : :
..i.. . : . . .... . . . .
l' tie s . i. i. 3. .. .. ii.. (s.c.ndi) Page 33 of 44
DYNODE-P Prairie Island 1 Cycle 12 Turbine Trip Figure 4.17 Core inlet Temperature are
!, see- - . ...........
o : -
. i. . .. . . > . .
530 , . e se so ao 4e so Tim. (s.t.nds) i Figure 4.18 l MDNBR (WRB-D !
. 1 1
11 n... .
. . 5. . .3 ,
=. -
E
= -
= :
y - a .E. .E...
" ..m.
Z l IEE a - E : i :
.... 4.. ...>. . . .;... . . .
gl gip i , , , e 4 e e is Ts.. (seconds) Page 34 of 44
DYNODE-P Prairie Island 1 Cycle 12 Dropped Rod EOC Rgure 4.19 K Effective s.osot i. [.... i o.,,,,. . .. .. . . . . . e.,,,o. . s . . , . . .:. . . ... o.,,, . j .. ..>. . , ., . ..:...... e.,,,: , , , . e iso soo soo .co soo een Tim. (s.c.nds) i Rgure 4.20 l Absolute Power j ses l
. . . i iso. . g. . ..;. . . ;. . . ; ..
I
- , ,s. . . . '. . . . . .
- . . .
i : l .
,o_ .<.. . .:. . ..;.. . . . . . .
r
.o W j e too soo soo 4eo soo ano Tim. (S.conds)
Page 35 of 44 l
DYNODE-P Prairie Island 1 Cycle 12 Dropped Rod EOC Rgure 4.21 Core Average Heat Flux 104 i,3 . i. . .. i . . . 4. . . . . .:. . m io,. . . . .c . . . . j-
,s- s .1- - -
. i 96 ,
o 100 too soo 4o0 soo ese Time (Seconds) Rgure 4.22 Pressuriser Pressure 2230 1325- ,
$- k-
,2 9 ,
..7 .. . .. . .
< :i. . . . . .:. .
e.n i: !. . ;. . s ..i . . ;. . . l
.. v. : . . . . . . .... ..
$ G5 iip O ISO 200 300 400 500 400
- m. <s.cond )
Page 36 of 44
l i l DYNODE-P Prairie Island 1 < Cycle 12 Dropped Rod EOC Figure 4.23 Vessel Average Temperature ser ess.s- < n- -<-- w sse- --- - -
-+ -
e, e - o ..s.o_ . . , . . . . . . i ess- ,- -?- . , ..>. .
~.
ses.s , , , , e soo soo,. soo 4c0 soo ese Time (Seconds) Figure 4.24 MDNBR (WRB-D s
,, s. ..: . . . :. . . ..;... . .. . . .
=
1 1E E E E E E E E E E l l D m - Z . O ' ( i_ . i. . ..;. . i. . . . . . . g o . . . . . e iso soo soo ano eso aeo Time (seconde) l l l Page 37 of 44
1 DYNODE-P Prairie Island 1 Cycle 12 2/2 Pump Trip Rgure 4.25 K Effective i..: 4...... O ..F . so do so so a.o Time (S.cends) i Rgure 8.26 Absolute Power iso . i .. .<.. .3 . . ... .. ..
. ... . . . . e. . .
... ..<.. .3 . . . . . .
. [jlgy
. no 4. a. a. iee Ti.. (s.c.nd )
Page 38 of 44 l l
----- . -. ~ . - , . . . - , - . . , _ _ _ , - . . , . . . . . , _ _ _ , , , , . . _ , , . . . . . , . - , . - . . _ _ . - . . - - , . . - .
DYNODE-P Prairie Island 1 l l Cycle 12 2/2 Pump Trip Figure 4.27 Core Average Heat Flux no i.e. . . l ee. . . .. .. . . .. . :. . ;
. . i i
. I y e,_ . . . .. . . . .; . . . .
! 1 4
- i
. . . 1 1
e - o so 4e so so see i Tim (Seconds)
'l, Figure 4.28 l Pressuriner Pressure i 2:50 I 1
4 l
\
t 33... . c. . . . j
- i ! l 1
. . . . l I
$ ::se- . - . - . . .; . . .
e
)
- e- .>. .i. . ... .
. [ imp su.
e no .o ao no i.e Tim. (s.c.nd ) I i i Page 39 of 44 1 1 l
,______,_,._.__,___..._........,-.._.J
DYNODE-P Prairie Island 1 Cycle 12 2/2 Pump Trip i Rgure 4.29 Core Flow 60 l
.0 < . . ..>. ..i.. . . ..
I N j .0 . . .. 4 a . I
- l 30 . !. .
. .i .
.i. .
\
- 1 i
0 0 30 40 40 80 00 0 Tim. (s.c.nds) Rgure 4.30 MDNBR (WRB-D 3.5
- I
- 1 l
. ll
= i- , ,
, .E- i
$1 1
BEgE E i E l E E BEE# E o . EEE.EE ( i.s - . i. . . .i . ..i.. . .i. . . 1
. . 1
- g I , , ,
4 9 2 3 e 3 l Ti.. (s.c..d.) 4 Page 40 of 44 _. _ - . _ - . . . . . _ _ . . , __ ~ _ . _ _ _ - . , . _ . . . _ _ . _ . . _ _ _ _ . . _ _ _ . . _ . . _ . . _ _ _ . _ . - - _ _ , . . _ . _ _
I i 1 1 DYNODE-P Prairie Island 1 Cycle 12 Locked Rotor Figure 4.31 K Effective i.oi
. l
.... 4.. . ..,.
9.97 , , , ,
. . . . i.
Tim. (seconds) Figure 4.32 Absolute Power ite l00- -<. '? t* , i i 33- , , ..l.. . .j .. .. .
.+. . ..
("
.0- .t- t- c ** -
l
... . . . . e. . . . .
g e : is 11.. (s.c.ad )
- Page 41 of 44 4
e 1 DYNODE-P Prairie Island 1 Cycle 12 Locked Rotor Figure 4.33 Core Average Heat Flux iso i.. .. t . m . so- - - '-
-- - ~
i : . 4,_ , . . .; . . . !. . . . . . , . . l se , , , . o a 4 . a is j Tim e (Seconds) t Figure 4.34 3 Pressuriner Pressure esse
. i : :
... .e.
34,s. , .;.... . .; . . . . . . g ..... . . : . . .:. . . A. . . .
. : e. .e. ..>..
ison. . , . . . .:. . .
. : : i .
- . : [jgp
>>i. . , ,
e . . . is is.. (s.c.adin Page 42 of 44 f
-.---v ---- - - - - , - , , , - ,,,,n-,-, . , . , , , . - - . . - , , . . , , , - . , - , , - , . _ . - - - . , , - , . - - , , . , , , , . , . , _ , . - , - , , -
---,--,---_.y , - , -.n,-
DYNODE-P Prairie Island 1 Cycle 12 Locked Rotor Figure 4.35 Core Flow 70
..e. .
. . .y . ... . . .
SS- - * ** *-
- s, .- .
m . . .
*4.. . . .- .e.
so- ,
?- .
so , , , , o 4 . s se Tim e (Seconds) Figure 4.36 MDNBR (WRB-D s.s r . er *
. ll
) -
3 . m - O + 4 i,s _ .'.. . .! . .
.i. .
.....E., '
E. - - a a m - e n i g
'o a i i e 1 w e(secondo i
Page 43 of 44
- 1) NSPNAD-8102P Rev 4, " Reload Safety Evaluation Methods for Application to PI Units",
March 1985.
- 2) Prairie Island Final Safety Analysis Report.
- 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) Internal Corespondence, C A Bonneau to D A Rautmann, " Transfer of CAS Input to SAS for PI 1 Cycle 12 Reload Evaluation: December 12, 1986.
- 10) Internal Correspondence, C A Bonneau to D A Rautmann, " Error in FRDR CAS-SAS Transfer Letter" December 19, 1986.
- 11) NSPNAD-8504P, Rev.1, " Prairie Island Unit 2 Cycle 10 Final Reload Design Report,"
October 1985. Page 44 of 44}}