ML20140A781
| ML20140A781 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/06/1986 |
| From: | Kapitz J, Rautmann D NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20140A716 | List: |
| References | |
| NSPNAD-8600, NUDOCS 8601230371 | |
| Download: ML20140A781 (56) | |
Text
- a EXHIBIT H PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request Dated January 13, 1986 e
PRAIRIE ISLAND UNITS 1 AND 2 SAFETY EVALUATION OF INCREASED FQ, FAH, AND ISOTHERMAL TEMPERATURE COEFFICIENT NSPNAD-8600 January 1986
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Prepared by #h e &&/-, Date I/ /80
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Reviewed by }
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Date / //.e </6 7 ,
i- Approved by / 1[ C W;s Date / - _-
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8601230371 960113
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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.
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TABLE OF CONTENTS P_agg
1.0 CONCLUSION
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 3.3 Thermal Margin 12 3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance 12 3.4.1 Rod Bow as Appl 1ed to DNBR Analysis - ENC Fuel 12 3.4.2 Rod Bow as Applied to DNBR Analysis - Westinghouse Fuel 13 3.5 Fuel Temperature Analysis 14 3.6 Safety Limit Curves 14 4.0 ACCIDENT AND TRANSIENT ANALYSIS 18 4.1 Plant Transient Analysis 18 4.1.1 Input Parameters 20 4.1.2 Transient Analysis Results 20 4.1.2.1 Fast Control Rod Withdrawal 20 4.1.2.2 Slow Control Rod Withdrawal 21
, 4.1.2.3 Los of External Electric Load 21 4.1.2.4 Dropped Rod - Auto Control 22 4.1.2.5 Loss of Reactor Coolant Flow 23 4.1.2.6 Locked Pump Rotor 24 4.1.2.7 Large Steam Line Break 25 4.1.2.8 Small Steam Line Break 27 4.2 LOCA-ECCS Analysis 27 4.3 Rod Ejection Analysis 27 Page 3 of 56
LIST OF TABLES
.P.,agg
, 3.1 Prairie Island Thermal Hydraulic Reference Conditions 16 3.2 Dropped Rod - Auto Control, Transient and Thermal Margin Results 17 4.1 Sununary of Prairie Island Transient Margins 29 4.2 Parameter Values Used in Full Power Transient Analysis 30 4.3 Pra'irie Island Units 1 and 2 Trip Setpoints 31 4.4 Prairie Island Ejected Rod Analysis 32 i
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LIST OF FIGURES Pa21 4.1 Fast Rod Withdrawal - K-effective 33 4.2 Fast Rod Withdrawal - Absolute Power 33 4.3 Fast Rod Withdrawal - Core Average Heat Flux 34 4.4 Fast Rod Withdrawal - Pressurizer Pressure 34 4.5 Fast Rod Withdrawal - Vessel Average Temperature 35 4.6 Fast Rod Withdrawal - Minimum DNB Ratio 35 4.7 Slow Rod Withdrawal - K-effective 36 4.8 Slow Rod Withdrawal - Absolute Power 36 4.9 Slow Rod Withdrawal - Core Averaga Heat Flux 37 4.10 Slow Rod Withdrawal - Pressurizer Pressure 37 4.11 Slow Rod Withdrawal - Core Inlet Temperature 38 4.12 Slow Rod Withdrawal - Minimus DNB Ratio 38 4.13 Turbine Trip - K-effective 39 4.14 Turbine Trip - Absolute Power 39 4.15 Turbine Trip - Core Average Heat Flux 40 4.16 Turbine Trip - Pressurizer Pressure 40 4.17 Turbine Trip - Core Inlet Temperature 41 Page 5 of 56
LIST OF FIGURES Paig 4.18 Turbine Trip - Minimum DNB Ratio 41 4.19 Dropped Rod - K-effective 42 4.20 Dropped Rod - Absolute Power 42 4.21 Dropped Rod - Core Average Heat Flux 43 4.22 Dropped Rod - Pressurizer Pressure 43 4.23 Dropped Rod - Vessel Average Temperature 44 4.24 Dropped Rod - Minimum DNB Ratio 44 4.25 2/2 Pump Trip - K-effective 45 4.26 2/2 Pump Trip - Absolute Power 45 4.27 2/2 Pump Trip - Core Average Heat Flux 46 4.28 2/2 Pump Trip -Pressurizer Pressure 46 4.29 2/2 Pump Trip - Core Flow 47 4.30 2/2 Pump Trip - Minimum DNB Ratio 47 4.31 Locked Rotor - K-effective 48 4.32 Locked Rotor - Absolute Power 48 4.33 Locked Rotor - Core Average Heat Flux 49 4.34 Locked Rotor - Pressurizer Pressure 49 Page 6 of 56
LIST OF FIGURES P.agg 4.35 Locked Rotor - Core Flow 50 4.36 Locked Rotor - Minimum DNB Ratio 50 4.37 Large Steam Line Break - K-effective 51 4.38 Large Steam Line Break - Absolute Power 51 4.39 Large Staam Line Break - Core Average Heat Flux 52 4.40 Large Steam Line Break - Pressurizer Pressure 52 4.41 Large Steam Line Break - Pressurizer Level 53 4.42 Large Steam Line Break - Vessel Average Temperature 53 4.43 Small Steam Line Break - K-effective 54 4.44 Small Steam Line Break - Pressurizer Pressure 54 4.45 Small Steam Line Break - Pressurizer Level 55 4.46 Small Steam Line Break - Vessel Average Temperature 55 9
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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, FAH, and isothermal temperature coefficient. The new FQ and FAH limits are 2.23 and 1.60 The FAH equation includes a new multiplier on the limit at reduced power, increasing froc 0.2 to 0.3. The Technical Specification on isothermal temperature coefficient allows a +5 pcm/*F ITC from 0% to 70% power, and a 0.0 pcm/*F ITC above 70% power.
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.
Section 2 of this report describes the calculational models and me'hodology t used for this analysis.
Section 3 contains the thermal-hydraulic design analysis.
Section 4 containt the accident analysis results.
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2.0 CALCULATIONAL MODELS AND METHODOLOGY 2.1 Calculational Models These calculations have been performed using the NSPNA0 Reload Safety Evaluation Methods for PWRs (Reference 1). These methods have been submitted to the NRC for approval and are currently being reviewed.
2.2 Methodology For this analysis NSPNA0 has evaluated the limiting transients fse Prairie Island. These limiting transients have been identified previoosly by
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Westinghouse (Reference 2), Exxon (Reference 3 and 4) 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 Stea.s Line Break
- 8. Rupture of Control Rod Drive Mechanism Housing (RCCA Ejection)
The transients that are not ru nalyzed are:
- 1. Uncontrolled RCC Assembly Withdrawal From a Subcritical Position
- 2. Startup of Inactive Loop
- 3. Feedwater System Malfunction
- 4. Excessive Load Increase
- 5. Loss of AC Power.
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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 following:
A. While the new FQ and FAH will affect the initial steady state MONBR, 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 accients.
B. Through the process of creating reload safety evaluations for several cycles at Prairie Island, NSPNAD has performed transient analysis over a wide range of ITC values. Some analyses have included a positive ITC and no change in the relative severity of the accidents was observed.
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 ITC and FQ, FAH.
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3.0 THERMAL HYORAULIC DESIGN ANALYSIS This section provides results of the thermal hydraulic design analyses for Prairie Island 1.
3.1 Gesign Criteria The thermal and hydraulic design performance requirements for Prairie Island fuel are as 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 MONBR due to fuel rod 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 and shall not exceed 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 Compatibility The hydraulic compatibility of the Prairie Island fuel is discussed in Reference 5.
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] 3.3 Thermal Margin The most limiting transient for Prairie Island is the dropped rod - auto control event. The minimum DNBR was calculated to be 1.662 in an ENC assembly, using I
the W-3 correlation. Table 3.1 provides reference conditions for the analysis.
Details of the plant transient analysis for Prairie Island are given in Section 4.1.2.4 3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance 3.4.1 Rod Bow as Applied to DNBR Analysis - ENC Fuel The calculation of the DNBR reduction as a result of rod bow considers both DNB tests with rod bow and the degree of bowing.
MDNBRB = MDNBRNB (1 8- 6 )
and, 68 = 0.0 for 0 s AC/C, < 0.5 q 68 = 2.0 X (AC/C, - 0.5) x 680W for 0.5 s AC/C, s 1.0 where MDNBRNB = MDNBR for nonbowed fuel MDNBRB = MDNBR for bowed fuel
, 6B = fractional reduction in MDNBR due to bowing
, A_C 95/95 = anticipated fractional gap closure C, as a function of exposure
, 6 bow
= MDNBR reduction associated with bow to contact.
The calculation of DNB rod bow to contact penalty is based on DNB tests with rod bow as referred to in the NRC's Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors. Exxon Nuclear Company's detailed methodology for calculating fuel rod bowing and its MDNBR effect is given in Reference 4.
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The maximum anticipated fractional gap closure is based on rod bow measurements of ENC fuel similar to the Prairie Island design and currently being used in operating reactors. Reference 6 presents the results of rod bow measurements taken on ENC PWR reload fuel.
The data base obtained from the above measurements include approximately 11,000 independent measurements of rod-to-rod spacings for interior as well as peripheral rod bows. After two cycles of operation, the results indicate rod bow which is a small fraction of that required for bow-to-bow contact. Application of this data to the ENC TOPR00 design is in accordance with the aforementioned SER and includes a 1.2 multiplier to account for cold-to-hot variations in measured rod spacings and a 1.5 multiplier to account for batch-to-batch variation. The maximum anticipated fractional gap closure through the fuel lifetime is:
AC 95/95 = 0.5493 .
C, This value is valid for assembly exposures up to 49.5 GWO/MTU, which bound operation up to 55 GWO/MTU peak pellet exposure. Table 3.2 provides a comparison of key results and rod bow penalties from the analysis of the l " dropped rod" event in section 4.0. The heat flux and pressure parameters in Table 3.2 correspond to the values calculated at the time of MONBR. For conservatism, 60 psia has been added to the pressure prior to calculating the rod bow penalties shwon in Table 3.2. The bowed and unbowed MONBR results are well above the allowable 1.3 value. Thus, no reduction in allowable reactor peaking is required as a result of a change in MONBR due to rod bow for ENC fuel.
3.4.2 Rod Bow as Applied to DNBR Analysis - Westinghouse Fuel The calculation of the ONBR reduction due to rod bowing for Westinghouse fuel is similar to that for ENC fuel (Reference 7).
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MONBRg = MONBRNB (1-58 )
where 6g is given as a function of assembly average burnup. For 0.400 in 0.0. OFA fuel the value of 6 g is:
0.050 (full flow) 0.055 (Iow flow)
This represents the rod bow penalty at an average assembly burnup of 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 ca Westing-house fuel; 33,000 MWD /MTU represents the maximum burnup of concern.
The bowed and unbowed MDNBR results are well above the allowable 1.17 value. Thus, no reduction in allowable reactor peaking is required as a result of a change in MONBR due to rod bow for Westinghouse fuel.
3.5 Fuel Temoerature Analysis The fuel temperature analyses for Prairie Island are given in References 8 (ENC) and 9 (Westinghouse).
3.6 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 con-siderations. 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.
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The thermal overtemperature limit is imposed on the reactor by the overtemperature 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 AT OT s tpoint and f( AI). These two terms are combined to give a final trip setting AT0T = AT OT ~ #(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 i
'ollows the equation.
f i FAH(P) = 1.60 [1 + 0.3 (1-P)] P s 1.0
= 1.60 p > 1.0 where P = fraction of rated power (1650 MWth)
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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 Cooiant Inlet Temperature ( F) 530.5 Core Pressure (esia) 2250.0 Power Distribu:1cr Overall Peaking (Fg ) 2.32 Radial x Loca' 1.60 Engineering Facte- 1.03 i
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TABLE 3.2 Dropped Rod - Auto Control i
Transient and Thermal Margin Results i
Rod Heat Flux @ time of MONBR (Btu /hr ft8 ) 326,406 MONBR 1.662 NB (Ac/c,) 95/95 0.5493 6 0.2193 BOW 6g 0.0216 l
MONBR 1.626
] B Margin to DNBR Limit (%) 20 i
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=,.-1 -,--.T -=r - = * --
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- 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 10).
- 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?
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- d. Is the margin of safety defined in the bases for any Technical Specification reduced?
As part of all safety evaluations, NAD evaluates all accidents in the USAR to determine if the current anlayses bound the design. All accidents that are not bounded by previous analysis are reanalyzed and the results are presented in the RSE.
4.1 plant Transient Analysis This documents the NSP Nuclear Analysis Department's analysis of plant operational transients for Prairie Island.
This analysis includes the following:
ITC The current Tech Specs require that the Isothermal Temperature Coefficient (ITC) be negative (except during low power physics ' tests) at all times. The new Tech Specs are designed to allow a +5.0 pcm/*F ITC up to 70% power, and a zero or negative ITC above 70%. All transients have been analyzed using a +5.0 pcm/*F ITC at full power in order to bound operation with the new limits.
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Thimble Plug Removal NSP has recently received approval to rernve the thimble plug assemblies from the assemblies that have neither control rods or source assemblies (Reference 11). Our analysis included the effect of thimble plug removal and the associated increase in core bypass flow.
Decarture from Nucleate Boiling Limits Previous analysis of thermal margin to MONBR limits was performed using the W-3 CHF correlation in the COBRA-IIIC/MIT computer program. NSP has since replaced the COBRA-IIIC/MIT program with the VIPRE-01 program. The ENC fuel will continue to be evaluated using the W-3 correlation. The OFA fuel uses the new WRB-1 CHF correlation. This carrelation has a lower design limit of 1.17 (versus 1.3 for the W-3 correlation)'. Both types of fuel are analyzed separately for DNB margin and the fuel type showing the least margin is included in the graphs for each transient.
FQ, FAH Small and large break LOCA analysis are being performed by Westinghouse for PI 1 Cycle 11. This analysis will include both Westinghouse OFA and ENC TOPROD fuel. The NSP transient analysis used a FQ of 2.32 and an FAH of 1.60. This will bound the LOCA analysis values of 2.30 and 1.60. In addition, the FAH limit equation at reduced power was changed as shown in section 3.6.
The analysis shows that the calculated transient, the dropped rod-auto control, is well above the acceptable MONBR of 1.3 for ENC fuel and 1.17 for Westinghouse fuel.
In addition to the dropped rod described above, the pump seizure, a Class IV event, showed a calculted MDNBR of < 1.3. The number of fuel rods which would potentially experience DNB in the transient is calculated to be less than 8%.
For all transients, the maximum pressurizer pressure was less than 2750 psia.
The latter pressure corresponds to 110% of the design pressure of 2500 psia.
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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 1.379 located at X/L = 0.6, and a total peaking, Fg, of 2.32 is assumed. Other thermal hydraulic parameters for full power operation are summarized in Table 4.2.
Reactor trip setpoints for Prairie Islano 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.
4.1.2 Transient Analysis Results 4.1.2.1 Fast Control Rod Withdrawal This trar.sient assesses plant response to a control rod withdrawal, with a reactivity insertion rate of 8.2 E-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 .94 seconds. The pressurizer pressure rises to 2308 psia at 4.9 seconds. The vessel average temperature rises by less than .1.5 *F at 4 seconds and then drops off. The DNB ratio drops from its initial-value of 2.157 to a minimum of 2.046 at 2.0 seconds after the start of the transient.
The acceptance criteria for this transient are that the minimum
- DNBR be not less than 1.3 and that the maximum reactor coolant and main steam system pressure not exceed 110% of tneir design values. This transient meets all acceptance criteria.
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l 4.1.2.2 Slow Control Rod Withdrawal This transient assesses plant response to a control rod withdrawal with a reactivity insertion rate of 2.52 E-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.
The transient response of the NSSS for this case is shown in Figures 4.7 through 4.12. The reactor trip is generated on 4 overpower AT at 57.1 seconds. The pressurizer pressure rises to 2481 psia at 60.2 seconds. The vessel average temperature l rises by less than 6 *F and then drops off. The DNS ratio drops from its initial value of 2.157 to a minimum of 1.794 at 57.2 seconds after the start of the transient.
The accept 4hca criteria arc that the minimum ONBR be not less than 1.3 and th.t 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.
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! The transient response of the NSSS for this case is shown in t Figures 4.13 through 4.18. At the start of the transient, the i
turbine stop valves close and the secondary side pressure rises rapidly to the safety valve setpoint at 14 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 5.4 seconds on a high pressurizer pressure j trip signal. Pressurizer safety valve opening occurs and a peak j pressure of 2501 psia is calculated at 6.6 seconds. Core power l remains relatively constant up until the time of reactor trip.
Because of the primary system pressurization, the DN8 ratio j increases and remains above its initial 2.157 value.
The acceptance criteria for this transient are that the minimum DNBR be not less than 1.3 and that the maximum reactor coolant' and main steam system pressure not exceed 110% of their design
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- values. This transient meets all acceptance criteria.
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4.1.2.4 Dropped Rod - Auto Control i 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, 1s.not adversely effected during the short I 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 a dropped rod could conceivably challenge the 1.3 MONBR limit.
The transient response of the NSSS for this case is shown in Figures 4.19 - 4.24.
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The MONBR drops from its initial value of 1.689 to 1.675 at 35.0 seconds. After a slight rise to 1.686 at 70.0 seconds, the MONBR begins tu drop and reaches a value of 1.663 at 10 minutes. The analysis is terminated at 10 minutes because of:
- a. The slow decrease in MONBR from 70 seconds on is due to the analysis using a HFP ITC of +5.0 pcm/*F. This causes the power to continue to rise after the initial rod pull.
Thi,s would not occur if the ITC was at 0.0 pcm/*F (maxinum value allowed at HFP).
- b. It is assumed in the analysis that the operators would notice the dropped rod and correct the situation within 10 minutes.
' A maximum of pressurizer pressure of 2232 psia occurs at 101 seconds. The acceptance criteria for this transient are thatthemthimumDNBRbenotlessthan1.3andthatthemaximum reactor coolant and main steam pressure not exceed 110% of their design values. _This transient meets all acceptance criteria.
s 4.1.2.5 Loss of Reactor Coolant Flow 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 open 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 assessmenof t the delay. All automatic reactor' control systems are assumed inoperable.
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The transient response of the NSSS for this case is shown in Figures 4.25 through 4.30. The MON 8R drops from its initial value of 2.157 to a minimum of 1.865 at 3.5 seconds into the transient due to an increase in the power to flow ratio. A . ,
3 maximum pressurizer pressure of 2372 psia is calculated to a
occur at 6.5 seconds into the transient.
The acceptance criteria for this transient are that the minimum i ONBR be not less than 1.3 and that the maximum reactor coolant' I and main steam system pressure not exceed 110% of their design values. This transient meets all acceptance criteria.
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! 4.1.2.6 Locked Pump Rotor ,
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The locked pump rotor transient is a Class IV event that considers '
plant response from full power operation when one of the two I primary coolant pumps 1,s 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 4
seizure. All automatic reactor control systems are assumed inoperable.
1 The plant response for this transient is shown in Figures-4.31 to 4.36. The calculated MONBR drops below 1.3 at approxim'ately 1.1 seconds after pump seizure. The duration of time for which the DNBR is less than 1.3 is less than 4.5 seconds. The number f of fuel rods statistically calculated to experience DN8 for this Class IV transient is less than 85. A maximum pressurizer pressure of 2502 psia is calculated to occur at 3.25 seconds
- into the transient.
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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 2700 'F.
This transient meets all acceptance criteria.
4.1.2.7 Large Steam Line Break This transient considers plant response, with a withdrawn (stuck) control rod cluster from hot shutdown (full flow), due to depressurization of the secondary system such as might occur for a large steam line break. Hot shutdown is considered since the steam generator secondary side water inventory is at a maximum which maximizes the duration and magnitude of the primary loop cooldown. This cooldown in conjunction with a negative moderator temperature coefficient at end-of-cycle conditions maximizes the severity of the transient.
The transient response of the NSSS for this case is shown in Figures 4.37 through 4.41. The secondary depressurization for the large steamline break transient is calculated using the Moody curve for fL/0 = 0 for a break at the S.G. exit. Cooldown and depressurization of the primary system is fairly rapid.
The signal for high pressure safety injection is conservatively assumed to occur at 6.0 seconds as a result of high containment pressure. . Borated HPSI flow does not actually commence however until after a 10 second delay or until 16 seconds into the transient. In the interim, the cooldown, in conjunction with the large negative moderator temperature coefficient, is sufficient to overcome the assumed minimum shutdown margin and begin a return to power after about 23 seconds into the transient.
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Minimum capability for injection of high concentration boric acid solution corresponding to the most restrictive single failure in the Safety Injection System is assumed. This corresponds to the flow delivered by one high head safety injection pump. Low concentration boric acid (1950 ppm) must be swept from the safety injection lines cownstream of the boric acid tank isolation valves prior to the delivery of high
! concentration boric acid (20,000 ppm) to the main coolant loops.
This effect has been allowed for in the analysis.
Core he'at flux reaches a maximum value of 15.5% at about 39 seconds after the start of the transient after which point boron reaching the core begins to reduce power. The MON 8R at
- approximately the time of peak heat flux was calculated to be 3.7 using the W-3 correlation. In calculating this MONBR a conservative peaking factor of 9.4 was used in the calculations.
The containment pressure response has not been reevaluated for Prairie Island 1 Cycle 11. The containment pressure response is a system dependent parameter and is relatively independent of fuel type. The containment pressure response is therefore bounded by the analysis in the FSAR.
l The acceptance criteria for the large steam line break 4
are as follows:
- 1. The maximum reactor coolant and main steam-system pressures must not exceed 110% of the design values.
- 2. The maximum containment pressure must not exceed the Technical Specification limit of 46 psig.
This transient meets all acceptance criteria.
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4.1.2.8 Small Steam Line Break The small steam line break transient is an acciaent similar to the large steam line break except that the initial break flow is only 25% of normal rated steam-flow versus the 620% in the large steam line break transient. The small steam line break transient is intended to envelope a steam generator relief valve failure.
Figures 4.42 through 4.45 show plant response during this transient. Primary system cooldown is less rapid in this transient then in the large steam line break transient.
Depressurization of the pressurizer is correspondingly less rapid so that in this transient the signal for HPSI due to low pressurizer pressure is not reached until 112 seconds after the start of the transient. In this small steam line break transient primary system cooldown is not sufficient to overcome the assumed minimum shutdown margin prior to boration from the HPSI and hence the reactor does not return to power.
4 The acceptance criteria for this transient are identical to those of the large steam line break. This transient meets
, all acceptance criteria.
4.2 LOCA-ECCS Analysis Large and small LOCA-ECCS analysis for Prairie Island Unit 1 Cycle 11 have been performed by Westinghouse and transmitted under separate cover.
4.3 Rod Ejection Analysis 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 I together with an adverse core power distribution, possibly leading to
- localized fuel rod damage.
- i l
Page 27 of 56
- I
a 1
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 coefficieats 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 136 cal /gm. The HZP pellet energy deposition was calculated to be 147 cal /gm. These values bound both ENC and Westinghouse fuel. 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 28 of 56 t
. ...-- . . ~ - . - - , ,- _- -.- . - - . . . - . . - . - - . . = . - - . .
. . - . ~ . .
y.
i e
TABLE 4.1 Summa ry o r Pra i rie Island Transient Margins (Calculated Value/ Acceptance Criteria)
. Pressure (psia)
. T rtns ient MDW8R RCS MSL g Falled Pins Clad Temp. Ipel Enthalpy (1) (F) (cal /ge)
Rod Wi thd rawa l ct Powe r Fast 2.046/1.3 2308/2750 885/1210 - - - 1 i
Slow 1.794/1.3 2481/2750 1048/1210 - - --
l Turbine Trip 2.157/1.3 2501/2750 1109/1210
~ 2/2 Pump Trip 1.865/1.3 2372/2750 1069/1210 - - -
Locked Rotor -
2502/2750 1099/1210 8 NC/2700 -
Crppped Rod 1.662/1.3 2232/2750 757/1210 - - -
MSL C ree k * -
2250/2750 .1046/1710 0 NC/2700 -
Ejected Ro4-HZP -
2500/3000 - NC 761/2700 147/280 Ejected Rod-HFP -
2380/3000 -
NC ,
739/2700 136/280 NC - Not calculated for. each cycle
- - Also a Toco.. Spec.- limit of peak containment pressure 46 psig,.which is not calculated.
Page 29 of 56
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 System Pressure, psia 2,220 Hot Channel Factors Total Peaking Factor, Fg T 2.32 N
Enthalpy Rise Factor, F 1.60 AH Total Coolant Flow, Ib/hr 68.62 x'10 6 Effective Core Flow, Ib/hr 6 64.50 x 10 Reactor Inlet Temperature, 'F 534.5 j Steam 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**
- Locked Rotor is initiated from 2280 psia MSL break conservatively assumes no plugging i
i Page 30 of 56
____- .. . - . _ . - --- __- . . ._ - - .- . .. ._. . - - _ ~ . .. _ _ - __ _ -.
TAllLE 4.3 Prairie Island units 1 and 2 Trip setpoints SR1PRID1 M13dJR M4.lY111 LMay Tiog High heutron Flux 108% 118% 0.5 sec Low Reactor Coolant flow 93% 871 0.6 sec High Pressurizer Pressure 2388 psia 2425 psia 1.0 sec Low Pressurizer Pressure 1915 psia 1700 psia 1.0 sec High Pressurizer Water Level 85% of Span 100% of Span 1.5 sec Low-Low Steam Genera tor 13% or Span UK of Span 1.0 see Water Level ove rtempe ra tu re A T
- 1AVEo = 567.3F TAV[n = 567.3F 6.0 sec Po = 2250 psia Po = 2250 psia ove rpowe r A T *
- 1AVEo = 567.3 TAVfo a 567.3 6.0 sec High Pressure Safety injection 1842 psia 1800 psia 10 sec
- T he ove rtempe ra tu re A T t rip is a function of pressurizar pressure, coolant average temperattare, and axial offset. The TAVEo and Po setpoints are contained within the functional re l a t i onsh i p.
- T he ove rpowe r A T t r i p is a function of coolant average temperature and axial offset. The TAVEo setpoint is contalped within the functional relationship.
I J
Page 31 of 56
! TABLE 4.4
[ Prairie Island Ejected Rod Analysis j HZP HFP
. Maximum Control Rod Worth (pcm) 607 167 i
Doppler Defect (pcm) 1243 1243 Shutdown Margin (pcm) 1631 1631
! Delayed Neutron Fraction 0.004975 0.004884 Power Peaking' Factor, Fg 9.252 3.576 Energy Deposition (cal /gm) 147 136 i
i i
i i
- i 1-I.
i i
l Page 32 of 56
6
<l Prairie Island DYNODE-P Fast Rod Withdrawal Figure 4.1 K Effective i.ees 3 ..
..<.. ..j.
g,9,3- . .....- . . ... . . . . . . .
8.990- * * * *
- h*-
e.,es- .. . .. . - - .
i
! s.ess e i s : 4 s t.
he (Sec.nds) i i
1
]
Figure 4.2 -
Absolute Power 14 e Isa- ' +
i.... . ..
- ISO-
++ +* + +. ** * * ** * -l + + ++ -
b e .
e . .
!' c. .
e' 60- **d*
h- ** * * * - * * - * * **
i e
y eg- . . . . . . . . . . . . . . . . . . . . . ...... ... . .. ..
1 a
. se e
= e s a e s 2
l , N . cs...nds) js page 33 of 56 I #
l 2
- Prairie Island .
DYNODE-P Fast Rod Withdrawal !
1 i
Figure 4.3 Core Average Heat Flux
\
\
\ .
OR ri.. cs... a.>
4 1
Figure 4.4 -
Pressurtzer Pres.ure
. 22.
l 1
2,... . . . .. . .;.
1
, s . .
2 .
/ .
i L
/
/
....y...............
x 3 GEP
, in.
2
~ Ti.. (s...ad )
a g page 34 of 56 a.
5 m - , , _ _ _ . . . , . _ . . _ - - , _ . _ _
6A Prairie Island DYNODE-P Fast Rod Withdrawal Rgur. 4.5
- v. . . .i A. . r.g. T. . p . r. o r.
i
\
- ~
,n
\
- - . . ./. c 5
/
/ .., ...
p ri.. cs....d >
1 2
1 i
sgur. 4.6 .
MDNBR (W-3) 3 J
4h
.... . .s.. . . ..., . . .. . .. ..........y....A...
M i
i t . a L . -
a
. ...A ..-A. . . .A . A. . . . . . . ....
s z
- a E
.s .
.... . . . . .. . . . .4.. . . . . . . . . . . . . . . .
"o
?.
3
, , m 5
, Ti . cs....d.)
l' =
1
- 8 _ page 35 of 56 5
1 i
i Prairie Island i DYNODE-P Slow Rod Withdrawal l rigur. 4.7
, K Effective i....
. . . .) .
l 1 .
4 ...,,. .' . . . . . . .. . :. . . .
- . i. ,. i. .. .. .. ..
ri.. <s....,4.>
4 Figure 4.8 .
Ab.olute Power
,... . .. .. . .4 . . , .. . . . . . . . . , . . . .
i l
M s
9 ( .
(
! 3 .x N
. . .. 3. .. .. .. .. ,.
= ri.. cs... ..>
~
a g
p' age 36 cf 56 c
. . - - . . - . . - . _ ~ , - . - - - . . . . - -
l Prairie Island DYNODE-P Slow Rod Withdrowol Rgure 4.9 Core Average Heat Flux 13e l
.... .4... ., . < - . . . . . ..
m es- - -
.,_ .... , . ..i.. .
( e s - u.
e io se se .o se ao te Time (Seconds)
Rgure 4.10 .
Pressuriser Pressure 2$e8
\
34gg. .. . . .q... . . . . . .. . . . . . . ....;.... ..
,e I m
. /
e.
o.
R** .
13ee- *1" ' ' * * * ' * <*
3 I
(
I.
3 [EIP ssee e is se se .e se se to i
i Ti e (s..end.)
a
. page 37 of 56 e
i 1
j l I l Prairie Island DYNODE-P Slow Rod Withdrawol 1
1 Figure 4.11
' Core inlet Temperature ses Ste- * - E- '*(
d' ' ' ' * ;* * *
. .. . . . ...L.. . ...k... . . .
g4g.
J
- .= .
o ..e_ . . . . . . . . .
I i , .
333-j
.ie W
e to to se de Se et to Ti.. (s...nds) 4 I t
i Rgure 4.12 .
l MDNBR (W-3) i
),g a . . .4.. ...fa. . .e. ....e . . . n.'.
p i b i r.,
e ,
& ,"_ 6 4
. m. . . . e. . .
a . 6 -
s
=
z o
E o da w i
- g ,3 . . . . . . . . .0 . . . . . , . . . ..
,i I
3 .
E
- e io se se .e se 4. re u , , , , g s ,, ,, ,3 a
js page 38 of 56 6
- _ ,~ . _ _ _ - . - . . ___ . . _ _ _ , . , .
e g j i
l Prairie Island DYNODE-P Turbine Trip Figure 4.13 K Effective
, ....;.. . .; . . ..i. . ..
...,,. .. . . . . . .+. . . .
4 1 4
/
. .,e . . .. .. .
Y
. i. .. 3. .. ..
ri. . <s....a.)
l l
l l
Figure 4.14 .
, Absolute Power ii .
I i ... . . . . .
. e.
90 5
E 1 ... . . . . . .
.~
1 G2p
. i. 3. i. .. ..
= ri.. cs.... .)
a
[ ~ page 39 of 56 e
Prairie Island DYNODE-P Turbine Trip Figure 4.15 Core Average Heat Flux ire i.e. .4.. .... ... . . . . .
i l as- .. . . .. .. . . .
- . . - . * * >A- + -
DE es- . ..
.e- . . . . . . . . . . . . . .. . . . . . . . .
. .e. .
se. . , . .
e e is se se .e se Time (Seconds) i a Figure 4.16 .
Pressuriser Pressure lese a
Stes- k= *. *-k.* ** . 1. . ** ** - - * * * *
..es. .
. N N N
. ... . . . . 4.. . .. .. . .
< ,,se. . . . .. ..
N
. . . ..s . . . . . . . . . . . . .. . . . . _ . .
~
2 g h i \
am re !
- Stee- + + ' * * ...........8< * * * *a *a . +
i
- i
- sies. . . .. . ... . . . . . .. .. .
e g toes e e is as se 4e se u Ti.. (s. send )
3 page 40 of 56 a
9 5
Prairie Island :
DYNODE-P Turbine Trip Figure 4.17 Core inlet Temperature SFe o / .-
9.s~ - 4**
/
I 6
e ese- -
e.
! O 4
, . , .~ -
, .. ...,. ..s. .
/
sse W
e is to 8e *e se Tim . (5.c.nds)
- L Raure 4.18 .
- MDNBR (W-3) 3 ,,
o a
.. _ .......<....... . . .. .....4..... . . . . .
, . . . 3. . . . .
odobodaba
?
y u againaaa 3
,- s. .. . .. . . . . . . . . . . . . . . . . .a.. . . . .
% Z
- O I
E. '
-.: . .. . .<.... .. ...,. .. . .i... . . . . . . . .
i...
r i
E i
e : . . e is z ti.. (s.o.adi)
~
i
- page 41 of 56
Prairie Island DYNODE-P Dropped Rod. EOC ,
l Figure 4.19 K Effective i.see 3 .. . . . . . . _ .
i j
I g,,,,4- , .. . . . .. . . . . .. ..
I s.,,,3- * * * * * * * * * **
- 1 1 e.,,,e e ios too see dos
] Tim. (s. .ndo 1
Figure 4.20 Absolute Power i0f
\_
is,. . . . . .. .....i.. . . .
..i... .
4 M
e , .
as ,
e ,0- ** *** *** * * * * * * * * * * * * * ****+++a** * * * * *
- 1 J
s
-=
...i..... . . .
i <
j i .
?.
! 3 se E
e ios too ses see
- 2 Ti.. (s....do ;
~
5
! ? page 42 of 56 e
i
. . - - . .. , - - . _ . , - - - - - - - - , __ - . . . . - - . - . . , . . --. .-n-. ,. . . - , - - , . - -
Prairie Island DYNODE-P Dropped Rod. EOC Figure 4.21 Core Averog. Heat Flus see ion- >- ->- s- +
ise. .. . . . . . .
m es- - - -
96- - - *.
- lJ .
e iso eso see see Tim. (Sec.nds)
Figure 4.22 .
Pressuriser Pressure
- e 2 3.. . . . Y. .. . .
..f.. . .. . . . . .
t i
a
- 3
.h i :ee- .. . . . , . ...i.. . . . ..i.. . ..
e e
m i 5 l
~
.f.
, si,,
W e e ice see see ese ti
.e Ti. . (s...nd.) ;
1 1
i I
8 y page 43 of 56 L
Prairie Island DYNODE-P Dropped Rod. EOC Figure 4.23 v....i A..r... T..p.r.,ur.
/...
m 1
/
Q l
l p
ri.. cs... 4.)
Fiaur. 4.24 .
MDNBR (W-3) t i
! ..._ .. . . . . . . . . . . . . . .....4..... .. . .
I a .
l
[~ A a o a a26 A AiA A A!A A alA A t
. . .. . . . . . .. . . . . . .. . . . . ..+... ..
z
- O i s (
l t i .._ . .
0 .
W 3
- u. ... ... ...
ti.. cs....d.)
i 1
!. page 44 of 56
I l
Prairie Island DYNODE-P 2/2 Pump Trip Figure 4.25 K Effective l
4
- . . .I. . .
. i. .. .. .. ...
ri.. < s... .a .)
Figure 4.26 .
Absolute Power it s
+
300- +' +a*$+++ + + + + + +8+++ + + + + + + + + + + + + +
i
... ., . , . ... ..... .. .c... . . . .
M 60- + + + * + + + + + ++ + + + ++ + + + + + + ++
f9 r
i
.. . .( . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
?.~
s .
W e se .e .. es ies
- Ti. . (s....d.)
' J L
page 45 of 56
.g. ----- - . - _ , -- - -_
Prairie Island DYNODE-P 2/2 Pump Trip I
Hgure 4.27 Core Average Heat Flun 12 0 i.._ .
. , . . . . . ... . .i .
9
. , . _ . . .. . .. . 4. .
j .
i .
I
,,_ . . . w. . . . .. .. ..
l j e. i. .. .. .. ...
- Ti. . (s...ad )
l l
1
'l i.
4 Figure 4.28 .
Pressuriser Pressure
- 2400 2390- * *"* * ** j* *+ *** + = + +
- 2300- - * * * * * - * * ' * ** * * ** *a****
- v I
% b g 333g. . .,. . .. .. . . . . ....j.... .. . , .
s -
e.
1 3:3g- . . . . . . . . . . . .. . . . . . . . ...4..... . . . . , . . . . . . . .
-/
~. :
3
,, site E
- e. e se se se se see u Ti.. (s...adi) b E
page 46 of 56 l a".
)
0 a
.-. - . . - , , . - - . - 4- . - . - - , . . _ - - . . - - - . . . - --
Prairie Island DYNODE-P 2/2 Pump Trip Figure 4.29 Core Flow ee og. ..<.. . . , . ...i.. . .
b z
N e
".e- * * * *
- l* ** **
e.
c.
- t * .
' te-
. N _.
4 g e
- e se 4 es se see Tim e (s...nd.)
i i
Figure 4.30 .
MDN8R (W-3) 1
- i.e - ** *v * * - * *
- h ** * * * ** ** .* *
- *g >
e
- A M a 0
I
.h , . ' '.OAA . .. . .A...A % ,. g.3......... .....
............3.g............
wdA-Ad4AAAgw 4
m .
$ g t -
J; i
L 3, g < .. . .. . . .. . .>.. . . . . . . . . . . . . . . , . .
1 .
4 3 i G2!P
- e i : . . .
'i n . (s....d )
a, t.
page 47 of 56
f Prairie Island DYNODE-P Locked Rotor 4
Figure 4.31 1
K Effective 4 .....
j ...... . . . . . . . . . . ... ..
l . .
i y 4
J
, ...n. . . . , i, j n.. cs....s,>
L 4
'I
] Figure 4.32 .
Absolute Power <
1 ii.
i
\ .
i
..............,............7..........
A i ,, ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
y ' ,
1 e M
i 3 ... . . . . . . . . . . . , . . . . . . . . . - . . . . ....
) J 3 t
~
1 4 .
. . . . . . . . . . . . . .... w . . . . . . . .
t.
3 W
- E b
'n.. (s....dd l
, a I
Pace 48 of 56
, rI. -
6 E
f
i Prairie Island e
DYNODE-P Locked Rotor i
Figure 4.33 Core Averog. Heat Flum
- IIe s .
I 4
) .g.
s ..s.. . . 4., .. ..
J j
j 40 f 9 3 e e 4 10 i Tim. (S s.nds) i i
f 4
i j
'h
' Eigure 4.34 ,
Pressurlier Pressure tilt l
l test- .+...<... . .
.............t.... . . . . . . . . . . . . . . . . . .
i
- ..... ....4 . . . . . . . . . . . . . . . .. . . . . > . . . . . . .
< 34 g 3..
~
es e ** .
) %. .
6 n *
- 333g .. . ,,,...j. . .....g... . . . . . . . . . .
1
.et s.
"I l 2..
J ' '
b Bisa O
e : . . e is u w. (s...ade)
O I page 49 of 56 p
l 4
Prairie Island DYNODE-P Locked Rotor Figure 4.35 Cor. Flow re 9..
s -
a
- se- ~ - - - - - -- -
s -
4 s a e .
4 d
e.
- y ae- - * -
4 .
i 30- .
- i g
4 .
0 1 4 6 4 te Time (Seconds)
Floure 4.36 '
MDNBR (W-3) 3 3,s. ..
. 9 . . . ..
' i O
7, o -
O
=
g g_ ..A,. .. . . . . . . .. . .. w e e n=4 .
2 o .
8 k '
r 6 ,
.:: w i .s -- - ,+ - - - s - - - -
c->- -
]
< b .
- e. g a *
! 3 G9fP 3
u
'e i : : . e a r
- Time (Seconds)
A 9
e h page 50 of 56
I e
Prairie Island DYNODE-P Large MSL Break Agure 4.37 K Effective i...
jA ...,..
x
..,,. s ..;. .
\ . .K . . . . .
..,.. . . . . . . ...a..
s =
ri. . cs....a.)
Rgure 4.38 ,
i Absolute Power t.
I is. ... . . . . . .
. . . .. ;. . 4. . ... . ..
, ... .......4..... . .. .. . . .
1 ,
c; .
i 2
1 .. ..<... .. . . . .<.. . ..
2 4 \
m
- N
. ~_ M I . 2. 4, 6. .. I.. 32 .
i Tim. (S.co.de) a 1 page 51 of 56 2
Prairie Island DYNODE-P Large MSL Break Figure 4.39 !
Core Average Heat Flux se
. t is. . . . . , . . . . . ..i.. , .. . ..
. . . [
. i x.. . . . . . . . . . : . . ....;..
j
- g. .<. .. h . . .
e to 40 60 se soo iso Time (Seconds)
Rgure 4.40 Pressurizer Pressure 2$00 N
2000- . . --
. .- . . . .. ....i... . ..4.... . . . . . , . . . . .
< i300-m f- . . . . L . - .. . . .. . ..k...
- a '
=
s 2
i
- .000- .. ... ...-r..
.i. . ..s ... . .
~i
=
t 5
- e to 40 so so soo tro i Ti.. (s.condi)
.3 w
e.
page 52 of 56
Prairie Island DYNODE-P Large MSL Break 1
Rgure 4.41 Pressurl er Liquid Level it to- . , <
o- ,. .. . . .a . . ... . . . ... . . . .
4 . . . 1. . .
. . .) . .
C e- e . . .s . $. . .
3 ...e.. , . . . . . .; . . .. .
I s
a no ao ao no see iso Ti.. (s.c nds)
Rgure 4.42 .
Vessel Average Temperature see soo- .. . . . . ..
. ........i..... . i. . .. . . ..
, ..o- .....q...........;..... . .. . . . .. . ... .
i . Ns 2
s c .co. . . . . . . . , . .. . . .. .
2 1 -
2
- so. .. . . . . . . . . . . , . . . .. .. . . . . , . . . . . .
,m
- E W
l l 5
, 2o0
- a to 40 so so eso 12 0 i
a Tim. (Seconds) w i page 53 of 56 e
a
Prairie Island DYNODE-P Small MSL Break Floure 4.43 K lif fective
..,,. /. . ..;... . . . . .. .
\
. .. i.. ... ...
Tim. cs....s.)
Figure 4.44 .
Pressurizer Pressure
= ,.co. . . . . . . . . . . . . . . . . . . .
2 e.... . . .... . . .. .. . . ..h b
m isso 5
o so too iso 200
{
a Time (Seconds) w t page 54 of 56 z
l Prairie Island DYNODE-P Small MSL Break Figure 4.45 Pressurlier Liquid Level 12
- i. ..,. ..i.. ..;.
P i
3 . . .
~
o N N r _
r,._ i 0 30 t80 ISO 200 Tim. (s.c nd )
Figure 4.46 .
Vessel Average Temperature Sea b
.... . . . . . ...i.. . . . ..i.. . . . ...i. .. . . .
n m 2 o.... . . . . . . . . . . . . .. .. .. .
A h
5 o so 10e eso too 2
, Tim. (S.conds)
.a i
5
- page 55 of 56 2
h I) NSPNAg
- 8Pch n Methods for Application to PI Units".
i
) Prag7,,
5 XN~Nf.7g.
Units j ashe Prairie Island Nuclear Plant, XN'NF-go.gz TOPR00 Fue;7rairie Island Units 1 are 2 with ENC S)
- Stin9hou,,
and ENC 7(tempatability of Westinpouse 14x14 i,1985.
6)
XN~Np.,g_32(P)(A Suppyement J Mfer valuating Fuel Re Bowing",
2, ,
l)
WCAP.ggg, Roy 3 ,
,,,on , 31979, 8)
XN-Np.80-61' "#rdiet March 2881. ,Tanti Safety Analysisepert", Rev 1, 9)
Westinghouse Letter 85 po\tD00*"
- 10)
Northern States Powe NIAWI 5.1.9, " Safety va Er Cgse O(e V0fk 11)
NSPNAD-8412. " PrairieM I sig g61swation RCC Guide Thimble g Reag O' Plu I
i i
A 56 of 56 l
- , m - . r