ML20245D982

Cycle 4 Plant Transient Analysis

ML20245D982
Person / Time
Site:	Susquehanna
Issue date:	05/31/1989
From:	Roberts C SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17156B218	List: ML20245D982 ML20245E007
References
ANF-89-057, ANF-89-57, NUDOCS 8906270291
Download: ML20245D982 (46)

Text

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SUSQUEHANNA UNIT 2 CYCLE 4 PLANT TRANSIENT ANALYSIS 1

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i ADVANCEDNUCLEARFUELSCORPORATION L. ANF-89-057-Issue Date: 5/11/89 i I

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i p SUSQUEHANNA UNIT 2 CYCLE 4 PLANT TRANSIENT ANALYSIS j i

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i Prepared by

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A&cM M /E~

C' C. Roberts, JrN/

BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services q

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May 1989 l

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CUSTOMER DISCLAIMER I

IMPOR* ANT NOTICE REGARDfMG CONTENTS AND USE OF THIS DOCUMENT I

PLEASE READ CAREFULLY i

Advanced Nuclear Fuees Corporetson's warrances and representatens cori-  !

coming the subject matter of the document are those set forth in the Agreement between Advanced Nuclear Fuste Corporation and the Customer pursuant to whsch that document is issued. Accordingly, except as otherwise expressly pro-vided in such Agreement, neither Advanced Nuclear Fuels Corporaten nor any person acting on its behalf makes any warranty or representation. expressed or implied, with respect to the accuracy, cornpieteness, or usefulness of the infor. .

maten contasned in the document or that the use of any information, apparatus, l method or process disclosed in this document will not intnnge pnvately owned rights; or assumes arty liebelitsee with respect to me use of any information, ap.

paretus, memod or procese diecioned in mis document.'

The informetton caritained herein is for tne soie use of Customer.

In order to avoid impaarment of nghts of Advanced Nuclear Fuets Corporation in patents or inventens which may De included in the information contained in this document me recipient by its acceptance of mis document. agrees not to publish or make public use (in me patent use of tne term) of such information until so authonzed in wnting by Advanced Nuclear Fuels Corporation or untsi at:er six (6) months following terminadon or expsraton of the aforesaid Agreement and any .

extensen moroof, unless omerwoe expressly provided in the Agreement. No nghts or licensee in or to any patents are implied by me fumishing of this docu-mont.

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ANF-89-057 Page i TABLE OF CONTENTS Section Pace '.

1.0 INTRODUCTION

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2.0

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 ",

3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN . . . . . . . . . . . . . . . . 5 3.1 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.2 Anticipated Transients . . . . . . . . . . . . . . . . . . . . . 5 3.2.1 Load Rejection Without Bypass . . . . . . . . . . . . . . 6 3.2.2 Feedwater Controller Failure .............. 7 3.2.3 Loss of Feedwater Heating . . . . . . . . . . . . . . . . 8 3.3 Calculational Model ...................... 8 l 3.4 Safety Limit . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.0 MAXIMUM OVERPRESSURIZATION ..................... 21 4.1 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . 21 4.2 Pressurization Transients ................... 21 5.0 RECIRCULATION PUMP RUN-UP . . . . . . . . . . . . . . . . . . . . . . 23

6.0 REFERENCES

............................. 25 APPENDIX A SINGLE-LOOP OPERATION ................... A-1 APPENDIX B MCPR SAFETY LIMIT ..................... B-1 1

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ANF-89-057 Page ii

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LIST OF TABLES Table EiLqt 2.1 TRANSIENT ANALYSIS RESULTS AT DESIGN BASIS CONDITIONS ....... 4

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3.1 REACTOR DESIGN AND PLANT CONDITIONS FOR SUSQUEHANNA UNIT 2 . . . . .

3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR 9 l

'SUSQUEHANNA UNIT 2 . . . . . . . . . .-. . .-. . . . . . . . . . . . 10 3.3 RESULTS'0F SYSTEM PLANT TRANSIENT ANALYSES . . . . . . . . . . . . . 13 3.4 FEEDWATER CONTROLLER FAILURE ANALYSIS RESULTS AT 100% FLOW . . . . . 14

'A.1 SINGLE-LOOP OPERATION RE. ACTOR AND PLANT CONDITIONS . . . . . . . . A-5 f

I LIST OF FIGURES

[ fiaure Eagg 3.1 LOAD REJECTION WITHOUT BYPASS ................... 15 3.2 LOAD REJECTION WITHOUT BYPASS ................... 16 3.3 FEEDWATER CONTROLLER FAILURE . . . . . . . . . . . . . . . . . . . . 17 3.4 FEEDWATER CONTROLLER FAILURE , . . . . . . . . . . . . . . . . . . . 18 q 3.5 LOSS OF FEEDWATER HEATING ..................... 19  !

I. 3.6 LOSS OF:FEEDWATER HEATING ..................... 20 5.1 SUSQUEHANNA UNIT 2 CYCLE 4 REDUCED FLOW MCPR OPERATING LIMIT . . . . 24 A.1 SINGLE-LOOP OPERATION - PUMP SEIZURE . . . . . . . . . . . . . . . A-6 i

'A.2 SINGLE-LOOP OPERATION - PUMP SEIZURE . . . . . . . . . . . . . . . A-7 B.1 SUSQUEHANNA UNIT 2 CYCLE 4 DESIGN BASIS RADIAL POWER HIST 0 GRAM . . B-3 B.2 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS f

ANF-3 3.33/9Gd5 FUEL . . . . . . . . . . . . . . . . . . . . . . . B-4 )

B.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF-3 3.17/9Gd4 FUEL . . . . . . . . . . . . . . . . . . . . . . . B-5 q B.4 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-2 FUEL . . ......... ................ B-6 {

B.5 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-1 CENTRAL FUEL ........................ B-7 B.6 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS )

XN-1 PERIPHERAL FUEL . . . . . . . . . . . . . . . . . . . . . . B-8 )

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1.0 INTRODUCTION

This report presents the results of Advanced Nuclear Fuels Corporation's

) (ANF's)* evaluation of system transient events for Susquehanna Unit 2 Cycle 4 operation. The evaluation, together with core transient events, determines f

the necessary thermal nargin (MCPR limits) to protect against the occurrence of boiling transition during the most limiting anticipated transient. Thermal margins are calculated for operation within the allowed regions of the power / flow operating map up to the full power / full flow operating condition.

The evaluation also demonstrates the vessel integrity for the most limiting pressurization event using a two-second main steam isolation valve closure time. The bases for these analyses were provided in Reference 1. -

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'2.0 StMIARY-

The Susquehanna Unit 2 Cycle 4 core is composed of all 9x9 fuel as follows:

Number of Bundle Average b.

Fuel Tvoe- Assemblies Enrichment ANF-3 100 3.17/nGd4*

p 104 3.33/9Gd5 ANF XN-2 236 3.33 ANF XN-1 324 3.31

'Using ANF's methodology ' and considering the Cycle 4 core, the most limiting anticipated plant system transient with regard to thermal margin at

rated power' and flow conditions was confirmed- to be the generator load rejection without bypass- (LRWB) transient with recirculation pump trip (RPT) operable. 'The . minimum critical power ratio (MCPR) limits for potentially

( limiting anticipated plant'_ system transient events are shown in Table 2.1,for comparison ' purposes. These transients 'were evaluated with all co-resident fuel types modeled- and the most limiting condition _was used to determine the reported MCPRs. t Results with RPT out of service are reported in Section 3.2.1. The c'ontrol rod withdrawal error (CRWE) analysis and delta I

CPR results are reported in Reference 2.

Maximum system pressure has been calculated for the containment isolation event, which is a rapid closure of all main steam isolation valves (MSIVs),

j using the -scenario _ as specified by the ASME Pressure Vessel Code. This analysis verified that the safety valves of Susquehanna Unit 2 have sufficient

. capacity and performance to prevent the pressure from reaching the established transient pressure safety limit of 110% of design pressure (1.1 x 1250 =

-1375 psig). The maximum system pressures predicted during the event are l

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• The first number following the slash (/) states the number of gadolinia rods L per bundle and the second number states the weight percent gadolinia per rod.

The gadolinia concentrations and number of rods per bundle are stated for fresh fuel only. The gadolinia concentrations in the others are not significant since most of the gadolinia has been burned.

I ANF-89-057 Page 3 provided in Table 2.1. The analysis conservatively assumed a valve closing time of two seconds and six safety relief valves out of service.

Results of the single-loop operation (SLO) analysis are shown in Appendix A of this report. The safety limit analysis justifies single-loop operation with an increase in the MCPR safety limit of 0.01.

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TABLE 2.1 TRANSIENT ANALYSIS RESULTS AT DESIGN BASIS CONDITIONS

• i- l L Transient A CPR/MCPR**

Load Rejection Without Bypass 0.27/1 33 With Recirculation Pump Trip f;

Feedwater Controller' Failure 0.22/1.28 With Bypass-Loss of Feedwater Heating 0.16/1.22-t i

Maximum Pressure (osia)

' Vessel i Transient Dome Vessel Lower Plenun.] Steam line MSIV Closure 1,295 1,311 1,299 1 .y

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• 104% power /100% flow.

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y ** Based ~on the MCPR safety limit of 1.06 confirmed herein.

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ANF-89-057 Page 5 b

I 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN

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3.1 Desian Basis Consistent with the FSAR plant transient analysis, thermal margin operating MCPR limits are determined based on the 104% power /100% flow

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operating point. This thermal margin operating MCPR limit is then modified as a function of power and flow as required to protect against boiling transition resulting from anticipated transients occurring from allowed conditions on the i power / flow operating map. The plant conditions for the 104% power /100% flow point are as shown in Table 3.1. The most limiting point in Cycle 4 is at the maximum Cycle 4 licensing exposure limit when control rods are fully withdrawn f-om the core. The thermal margin limit established f or this exposure cordition is conservative for cases where control rods are partially inserted.

Ob ,ervance of a MCPR operating limit for ANF 9x9 fuel of 1.33 or greater conservatively protects against boiling transition during anticipated plant systens transients for Susquehanna Unit 2 Cycle 4.

The calculational models used to determine thermal margin include ANF's plant transient and core thermal-hydraulic codes described in previous documentation.(1,3-6) Fuel pellet-to-clad gap conductances used in the analyses were based on calculations with RODEX2.(7) Table 3.2 summarizes the values used for important parameters that provided a bounding analysis.

Recirculation pump trip (RPT) flow coastdown input was based on measured Susquehanna Unit 2 startup test data. To confirm the neutronics as requested by the SER issued for Reference 8, the Susquehanna system transient model was l

benchmarked to appropriate Unit 2 startup test data. XCOBRA-T(9) was used to calculate the change in critical power ratio (delta CPR) for pressurization I event analyses, 3.2 Anticipated Transients f

ANF considers eight categories of potential system transient occurrences for Jet Pump BWRs in XN-NF-79-71.(1,8) The three most limiting transients are described here in detail to show the thermal margin for Cycle 4 of Susquehanna Unit 2. These transients are:

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ANF-89-057 Page 6

- Load rejection without bypass (LRWB)

Feedwater controller failure (FWCF)

Loss of feedwater heating (LFWH).

A summary of the transient analyses is shown in Table 3.3. Section 5.0 contains a discussion of the recirculation pump run-up event which is limiting at less than rated flow conditions. Other plant transient events are inherently nonlimiting or clearly bounded by one of the above events.

3.2.1 Load Re.iection Without Bvoass The load rejection without bypass event is the most limiting cf the class I of transients characterized by rapid vessel pressurization. The generator load rejection causes a turbine control valve trip, which initiates a reactor scram and a recirculation pump trip (RPT). The compression wave produced by the fast control valve closure travels through the steam lines into the vessel and creates the vessel pressurization. Turbine bypass fl ow, which could mitigate the pressurization effect, is not allowed. The excursion of core power due to void collapse is primarily terminated by reacter scram and void g 5I

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growth caused by RPT. Figures 3.1 and 3.2 depict the time var'iance of critical reactor and . plant parameters during the load rejection transient calculation with bounding assumptions. The bounding assumptions tre consistent with ANF's COTRANSA code uncertainties analysis methodology reported in Reference 8 and approved by the NRC. The bounding assumptions '

include:

Technical Specification minimum control rod speed Technical Specification maximum scram delay time g Integral power increased by 10%. 3 E'

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At design basis conditions (104% power /100% flow), a delta CPR of 0.27 was calculated for the load rejection without bypass event when RPT is operable.

> The load rejection without bypass event was also analyzed at the design basis conditions when RPT is not operable. The resulting delta CPR is 0.35.

3.2.2 Feedwater Controller Failure Failure of the feedwater control system is postulated to lead to a maximum increase in feedwater flow into the vessel. As the excessive t

feedwater flow subcools the recirculating water returning to the reactor core, the core power will rise and attain a new equilibrium if no other action is taken. Eventually, the inventory of water in the downcomer will rise until the high level vessel trip setting is exceeded. To protect against spillover of subcooled water to the turbine, the turbine trips; closing the turbine stop valves and initiating a reactor scram. The compression wave that is created, though mitigated by bypass flow, pressurizes the core .and causes a power

• excursion. The power increase is terminated by reactor scram, RPT, and pressure relief from the bypass valves opening.

s The evaluation of the flow event at design basis conditions was performed with bounding values and resulted in a delta CPR of 0.22. Figures 3.3 and 3.4 l

present key variables for this feedwater controller failure event. This event was also examined for reduced power conditions at full flow. The results for the FWCF transients from reduced power conditions are shown in Table 3.4. The calculated results show that FWCF delta CPRs vary with decreasing power at full flow conditions. The highest delta CPR was calculated at 40% power and b

100% flow conditions.

This transient event at full power and full flow conditions was also analyzed assuming bounding conditions and failure of the bypass valves to open. With these assumptions, a delta CPR of 0.32 was calculated.

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I ANF-89-057 Page 8 3.2.3 Loss of Feedwater Heatina I: l The loss of feedwater heating leads to a gradual increase in the subcooling of the water in the reactor lower plenum. Reactor power slowly  !

rises to the thermal power monitor system trip setpoint. The gradual power change allows fuel thermal response to maintain pace with the increase in .)

neutron flux. Using the methodology in Reference 1, the delta CPR for this j event in Cycle 4 is 0.16. Figures 3.5 and 3.6 depict key variables for the loss of feedwater heating event. {

The bypass valves do not significantly affect the loss of feedwater heating results. Therefore, the delta CPR limit is applicable whether the  ;

bypass valves are operable or not. '

3.3 Calculational Model The plant transient code used to evaluate the generator load rejection and feedwater flow increase was ANF's code COTRANSA.(l) The axial one-dimensional neutronics model predicted reactor power shifts toward the core middle and top as pressurization occurred. This was accounted for explicitly g in determining thermal margin changes in the transient. The loss of feedwater 5 heating event was evaluated with the PTSBWR3 and XCOBRA codes (Reference 1).

Appendix A of the Susquehanna Unit 1 Cycle 2 plant transient analysis (Reference 10) delineates the changes made to the COTRANSA code for the Susquehanna analyses. Reference 9 describes the XCOBRA-T code used to l

calculate the delta CPRs for the pressurization transients.

3.4 Safety limit The safety limit is the minimum value of the critical power ratio at g which the fuel could be operated where the expected number of rods in boiling 5 transition would not exceed 0.1% of the fuel rods in the core. The safety limit is the MCPR permitted to occur during the limiting anticipated h operational occurrence. A MCPR safety limit of 1.06 for all fuel types in Susquehanna Unit 2 Cycle 4 was supported by the methodology presented in Reference 3. The input parameters used to support the MCPR safety limit are presented in Appendix B of this report.

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TABLE 3.1 REACTOR DESIGN AND PLANT CONDITIONS FOR SUSQUEHANNA UNIT 2 L-Reactor Thermal Power (104%) 3439 MWt Total Core Flow (100%) 100.0 Mlb/hr Core In-Channel Flow 90.0 Mlb/hr Core Bypass Flow 10.0 Mlb/hr i Core Inlet Enthalpy 518.0 Btu /lbm Vessel Pressures Steam Dome 1035 psia Upper Plenum 1046 psia Core 1053 psia

>. Lower Plenum 1067 psia Turbine Pressure 975 psia Feedwater/ Steam Flow 14.15 M1b/hr Feedwater Enthalpy 362.6 Btu /lbm Recirculation Pump Flow (per pump) 15.85 M1b/hr O

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I ANF-89-057 Page 10 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 2 High Neutron Flux Trip 125.3%

Control Rod Insertion Time 3.49 sec/90% inserted Control Rod Worth nominal Void Reactivity Feedback nominal Time to De-energized Pilot Scram Solenoid Valves 200 msec (maximum)

Time to Sense Fast Turbine g Control Valve Closure 30 msec 5 Time from High Neutron Flux Time to Control Rod Motion 290 msec Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open Turbine Control Valve Stroke Time (Rated Power) 70 msec Fuel / Cladding Gap Conductance Core Average-(Constant) 770.2 Btu /hr-ft2-F g Safety / Relief Valve Performance Technical Specifications 3 Settings Relief Valve Capacity 225.4 lbm/sec (1110 psig) l Pilot Operated Valve Delay / Stroke 400/150 msec 1

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ANF-89-057 Page 11 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 2 (Continued)

MSIV Stroke Time 2.0 sec MSIV Position Trip Setpoint 85% open Turbine Bypass Valve Performance Total Capacity 936.11 lbm/sec Delay to Opening (80% open) 300 msec

! Fraction of Energy Generated in Fuel 0.965 Vessel Water Level (above instrument zero)

High Level Trip 58.7 in Normal 35 in* ~

Low Level Trip 8 in Maximum Feedwater Runout Flow Three Pumps 5049 lbm/sec ,

Recirculation Pump Trip Setpoint Vessel Pressure 1170 psig I

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• COTRANSA plots are giving water level above separator skirt and the value here is above instrument zero.

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I ANF-89-057 Page 12 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 2 (Continued)

I Control Characteristics Sensor Time Constants Pressure 500 msec Others 250 msec Feedwater Control Mode Three-Element Feedwater Master Controller Proportional Gain 50.0 (%/%) (%/ft)

Reset Rate 1.70 (%/sec/ft)

Feedwater 100% Mismatch Water Level Error '4.0 ft Steam Flow Equivalent 4.03 ft/100%

Recirculation Flow Control Mode Manual Pressure Regulator Settings Lead 3.0 sec Lag 7.0 sec Gain 3.33%/psid I

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I TABLE 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES Maximum Maximum Maximum Core Average System Neutron Flux Heat Flux Pressure l frent (% Rated) (% Rated) (osia) A_QPR Load Rejection 367 121.7 1198 0.27 Without Bypass g .

Feedwater Controller 210 116.7 1178 0.22 Failure I Loss of Feedwater Heating 123 121.4 1080 0.lo MSIV Closura With 507 138.5 1326 NA Flux Scram I

l NOTE: All events are bounding case at 104% power /100% flow.

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I TABLE 3.4 FEEDWATER CONTROLLER FAILURE ANALYSIS RESULTS AT 100% FLOW

% Powqt A_G28 104 0.22 80 0.31 1 65 C.33 40 0.36 I

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ANF-89-057 Page 21 _

l 4.0 MAXIMUM OVERPRESSURIZATION Maximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenario as specified by the ASME Pressure Vessel Code. This analysis showed that the safety valves of Susquehanna Unit 2 have sufficient capacity and performance to prevent pressure from reaching the established transient pressure safety limit of 110% of the design pressure (1375 psig). The maximum

system pressures predicted during the event are shown in Table 2.1. This analysis assumed six safety relief valves were out of service and a fast valve closing time of two seconds.

I 4.1 Desian Basis The reactor conditions used in the valuation of the maximum pressurization event are those shown in Table 3.1. The most critical active component (scram on MSIV closure) was assumed to faii during the transient. ~

The calculation was performed with ANF's advanced plant simulator code COTRANSA,(1) which includes an axial one-dimensional neutronics model. .

I 4.2 Pressurization Transients ANF has evaluated several pressurization events and has determined that I closure of all main steam isolation valves without direct scram is the most limiting. Although the closure rate of the MSIVs is substantially slower than the turbine stop valves or turbine control valves, the compressibility of the additional fluid in the steam lines results in a less severe transient for the faster turbine stop/ control valve closure transients. Essentially, the rate of steam velocity reduction L concentrated toward the end of the valve stroke, generating a substantial coyression wave. Once the containment is I isolated, the subsequent core power production must be absorbed in a smaller volume than if a turbine trip had occurred. Calculations have determined that the overall result is to cause isolation (MSIV) closures to be more limiting for system pressure than turbine trips.

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E ANF-89-057 Page 72 l Closure of All Main Steam Isolation Valves I

This more limiting calculation assumed that six relief valves were out of service and that all four steam isolation valves were isolated at the containment boundary within two seconds. For overpressurization, the g two-second valve closing time is conservative compared to longer closing W times. At about 2.1 seconds, reactor scram initiated due to a high flux trip.

Since scram performance was degraded to its Technical Specification limit.

effective power shutdown is delayed until after 3.4 seconds. Substantial thermal power production enhances pressurization. Pressures reach the recirculation pump trip setpoint (1170 psig) before the pressurization is reversed. Loss of coolant flow leads to enhanced steam production as less subcooled water is available to absorb core thermal power. The maximum pressure calculated in the steam lines was 1,299 psig occurring near the g vessel at about 5.7 seconds. The maximum vessel pressure was 1,311 psig W occurring in the lower plenum at about 5.4 seconds. The maximum vessel pressure is significantly less than the established vessel pressure limit of 1,375 psig.

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ANF-89-057 Page 23 5.0 RECIRCULATION PUMP RUN-UP Analysis of pump run-up events for operation at less than rated recirculation pump capacity demonstrates the need for an augmentation of the full flow MCPR operating limit for lower flow conditions. This is due to the potential for large reactor power increases should an uncontrolled pump flow increase occur.

This section discusses pump excursions when the plant is in manual flow control operation mode. Results obtained from previous analyses (10) showed the two-pump run-up bounds the single-pump run-up. Only the two-pump run-up is evaluated for Susquehanna Unit 2 Cycle 4. These results indicate that MCPR would decrease below the safety limit if the full flow reference MCPR is observed at initial conditions. Thus, an augmented MCPR is needed for partial flow operation to prevent violation of the MCPR safety limit for the two-pump excursion event. The analysis of the two-pump flow excursion indicates that the limiting event is a gradual power increase in which the heat flux tracks power.

The Susquehanna Unit 2 Cycle 4 analysis conservatively assumed the run-up event initiated at 57% power /40% flow and reached 114% rated power at 100%

rated flow. The event terminated at 105% of rated flow with a minimum CPR of 1.06. The results.of the two-pump run-up analyses for manual flow control are presented in Figure 5.1. The cycle specific MCPR limit for Susquehanna Unit 2 Cycle 4 shall be the maximum of the reduced flow MCPR operating limit, the full flow MCPR operating limit, or the power dependent MCPR operating limit.

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6.0 REFERENCES

1. R. H. Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71(P), Revision 2, Advanced Nuclear Fuels Corporation *, Richland, WA 99352, November 1981.

2. S. E. Jansen, T. L. Lotz, C. C. Roberts, Jr., "Susquehanna Unit 2 Cycle 4 Reload Analysis, Design and Safety Analyses," ANF-89-058, Advanced Nuclear Fuels Corporation, Richland, WA 99352, April 1989.

3. J. A. White, " Exxon Nuclear Methodology for Boiling Water Reactors,

._ THERMEX: Thermal Limits Methodology, Summary Description,"

XN-NF-80-19(P)(A), Volume 3, Revision 2, Advanced Nuclear Fuels

-- Corporation, Richland, WA 99352, January 1987.

~

4. T. W. Patten, " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors," XN-NF-524(A), Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1983.

5. T. H. Keheley, "L.isquehanna Unit 2 Cycle 2 Plant Transient Analysis,"

XN-NF-86-55, Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, May 1986.

6. T. H. Keheley, "Susquehanna Unit 1 Cycle 4 Plant Transient Analysis,"

XN-NF-87-22, Advanced Nuclear Fuels Corporation, Richland, WA 99352, April 1987.

7. K. R. Merckx, "RODEX2 Fuel Rod Mechanical Response Evaluation Model,"

XN-NF-81-58(A), Revision 2, Advanced Nuclear Fuels Corporation, Richland, WA 99352, April 1984.

8. S. E. Jensen, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71(A), Revision 2, Sucolements 1. 2 & 3, Advanced Nuclear Fuels Corporation, Richland, WA 99352, March 1986.

9. M. J. Ades, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," XN-NF-84-105(P)( A), Volume 1 & Volume 1, Supplencnt I anu Supplement 2, Advanced Nuclear Fuels Corporation, Richland, WA 99352, February 1987.

10. T. H. Keheley, "Susquehanna Unit 1 Cycle 2 Plant Transient Analyses,"

XN-NF-84-118, including Supplement 1, Advanced Nuclear Fuels

- Corporation, Richland, WA 99352, December 1984.

m

• Formerly Exxon Nuclear Company (ENC).

m W

i ANF-89-057 Page A-1 APPENDIX A SINGLE-LOOP OP_ERATION The NSSS supplier has provided analyses which demonstrate the safety of I

plant operation with a single recirculation loop out of service for an extended period of time. These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are allowed when both recirculation systems are in operation. The physical interdependence between core power and recirculation flow rate inherently limits the core to less than rated power.

l ANF fuel was designed to be compatible with the co-resident fuel in thermal hydraulic, nuclear, and mechanical design performance. The ANF methodology has given results which are consistent with those of the previous analyses for normal two-loop operation. Many analyses performed by the NSSS l supplier for single-loop operation are also applicable to single-loop operation with fuel provided by ANF. A discussion of the relevant events and limits for single-loop operation and the results of ANF analyses follow.

)

A.1 ABNORMAL OPERATING TRANSIENTS MCPR limits established for full-flow two-loop operation are conservative for single-loop transients because of the physical phenomena related to part-power part-flow operation; not because of features in reactor analysi: model s f

or compatible fuel designs. A review of the most limiting delta CPR

! transients for single-loop operation was conducted. Under single-loop l

conditions, steady state operation cannot exceed approximately 76% power and

61% core flow because of the capability of the recirculation loop pump. Thus, f the MCPR limit at maximum power is higher than the two-pump operating MCPR limit due to the power and flow dependent MCPR functions. The flow dependence is based on a flow increase transient from run-up of two pumps.

Flow run-ups from a single recirculation pump would be much less severe, though the conservative two-pump limit is retained.

I Ahf-89-057 Page A-2 A.1.1 Load Reiection Without Bvoass The limiting anticipated system transient for the Susquehanna units is ,

the load rejection without bypass (LRWB) pressurization transient. In this transient, the primary phenomena is the pressurization caused by abruptly gI l

stopping the steam flow through rapid closure of the turbine control valve. 5 When the rapid pressurization reaches the core it cat , is a power excursion due j to void collapse.

At reduced power and flow conditions, there is a corresponding reduction in steam flow. With the lower steam flow, the maximum pressurization of the core is reduced in comparison to rated conditions when the control valve is 1 closed. The resulting power excursion and associated margin change are reduced below those of the full power case. Analysis has shown that the g difference in the two-loop operation and single-loop operation core flow 5 characteristics do not adversely effect the single-loop operation case. Thus, for the Susquehanna units, the MCPR limits based on LRWB analyses at full power are conservatively applicable to the lower powers / lower flows associated with single-loop conditions. Furthermore, LRWB analyses by ANF at l

reduced power and flow conditions in other BWRs with single-loop operation confirm'this trend.

A.I.2 Feedwater Controller Failure The second worst limiting transient at full power and flow is the feedwater controller failure (FWCF) to maximum demand. This transient is also )

less severe at the power and flow conditions associated with single-loop I operation.

This transient essumes the feedwater controller fails to maximum demand and results in the maximum amount of subcooled feedwater in the downcomer.

When this cooler water reaches the core the power rises. The core power rise l is terminated by a reactor scram initiated by a turbine trip. The turbine trip is the result of the high water level trip caused by the additional 1

amount of feedwater being injected.  !

l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - . _ _ _ J

2

)

ANF-89-057-Page A-3

" At the reduced recirculation flows, the subcooling' in the downcomer due

to the high feedwater flow takes longer to transverse the core so that a high f water. level' trip' occurs before core power can' rise as high as it does in the  !

full-flow case.- As with.the LRW8, the pressurization event resulting from the n , turbine trip is less severe for the reduced power in single-loop operation.

Thus,. because of the slower enthalpy transport phenomena caused by the ilower recirculation flow and because of the lower steam line flow in the pressurization portion of ' the transient, the FWCF has larger margin to the operating limit in single-loop operation than in two-loop operation.

A.1.3 MCPR Safety Limit l

- For single-loop operation, the NSSS vendor found that an increase of 0.01 I in the MCPR safety- limit was needed to account for the increased flow measurement uncertainties and increased TIP uncertainties associated with {

single pump operation. ANF has evaluated the effects of the increased flow  ;

measurement uncertainties on the MCPR safety limit and found that the NSSS l vendor determined increase in the allowed MCPR safety limit is also applicable to ANF fuel- during single-loop operation. Thus, increasing the MCPR ' safety l limit by 0.01 for~ single-loop operation (1.07) with ANF fuel is sufficiently l l conservative to also bound the increased flow measurement uncertainties for l single-loop operation.

i A.l.4 Summary The full power full flow two-loop MCPR operating limit plus .01, together with ~the MCPRr curve for two-loop operation plus 0.01 and the MCPRp curve for two-loop ' operation plus 0.01, conservatively bound all transients.

.- A.2 POSTULATED ACCIDENTS k

A.2.1 Loss Of Coolant Accident ANF performed ' LOCA analyses for single-loop conditions and determined that the MAPLHGR limit curve for two-loop ope. ration is also applicable to single-loop operation with ANF fuel (Reference A-1).

J

ANF-89-057 Page A-4 A.2.2 Pumo Seizure Accident The seizure of a recirculation pump is considered as a design basis accident e'v ent. It is a very mild accident relative to other design . basis g accidents such as the loss of coolant accident (LOCA). The pump seizure event 3 is a postulated accident in which the recirculation pump impeller speed is rapidly reduced to zero. This causes a rapid decrease in core flow and a decrease in the heat removal rate from the fuel rods.

A pump seizurc accident event was analyzed for Susquehanna Unit 2 Cycle 4 to confirm the significance of this event relative to the design basis LOCA. g The initial plant conditions are shown in Table A.1, and the plant response to a the pump seizure accident is shown in Figures A.1 and A.2. The core remains ,

covered during the accident and any fuel rods which experience boiling transition would be expected to be in the film boiling mode for a short '

period. In addition, the film boiling would be limited to small localized areas in the affected fuel ass'emblies. Because of this short duration, fuel failures due to overheating or clad strain would not be expected as a result of this accident. Thus, the consequences of this event are bounded by the LOCA where fuel failures are assumed to be extensive.

l A.3 STABILITY PP&L will establish stability surveillance requirements for Susquehanna Unit 2 Cycle 4 in conformance with the interim operating guidelines presented in NRC Bulletin 88-07 Supplement I based on the calculation results prepared by ANF.

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'ANF-89-057 Page A-5 TABLE A.1 SINGLE-LOOP OPERATION REACTOR AND PLANT CONDITIONS' 1

Reactor Thermal: Power 2489 MWt Total Recirculation Flow 60.35 Mlb/hr

/~

. Core Bypass Flow 5.61 Mlb/hr Core Inlet Enthalpy 507.3 Btu /lb p Vessel Fressurer.

Steam Dome 994.5 psia Lower Plenum 1010.9 psia 965.5 psia '

Turbine Pressure l Steam Flow 9.85 Mlb/hr. .

Feedwater Enthalpy 330.7 Btu /lb l l-I l

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Page A-8 i A.4 REFERENCES I

A-1. D. R. Swope, "Susquehanna LOCA Analysis- for Single Loop Operation,"

XN-NF-86-125, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1986.

A-2. S. E. Jensen, T. L. Lotz, C. C. Roberts, Jr., "Susquehanna Unit 2  !

Cycle 4 Reload Analysis, Design and Safety Analyses," ANF-89-058, l

' Advanced Nuclear Fuels Corporation, Richland, WA 99352, April 1989. l

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l ANF-89-057 Page B-1 f

APPENDIX B i

l MCPR SAFETY LIMIT l-i B.1 INTRODUCTION The MCPR fuel cladding integrity safety limit was calculated using the methodology and uncertainties described in Reference B-1. In this j methodology, a Monte Carlo procedure is used to evaluate plant measurement and power predictions uncertainties such that during sustained operatior, at the

)

MCPR cladding integrity safety limit, at least 99.9% of the fuel rods in the -

I core would be expected to avoid boiling transition. This appendix describes the calculation and presents the analytical results.

1 B.2 CONCLUSIONS During sustained operation at a MCPR of 1.06 with the design basis power ,

distribution described below, at least 99.9% of the fuel rods in the core are l expected to avoid boiling transition at a confidence level of 95%. The 1.06 l

MCPR safety limit is justified based upon a full power full flow MCPR operating limit of 1.34 or below, r i

B.3 DESIGN BASIS POWER DISTRIBUTION '

Predicted power distributions were extracted from the fuel management analysis for Susquehanna Unit 2 Cycle 4. Th.e radial power distributions were ..

evaluated for performance as the design basis radial power map, and the distribution at 11,086 mwd /MTU exposure was selected as the most severe j expected distribution for the c,ycle. The distribution was skewed toward I

higher power factors by the addition of bundles with a radial peaking factor

, approximating an operating MCPR level of 1.34 at full power. The resulting design basis radial power distril"Jtion is shown in Figure B.1.

f The fuel management analysis indicated that the maximum power ANF bundle (ANF-3 3.33/9Gd5) in the core at the end-of-cycle exposure (11,086 mwd /MTV) was predicted to be operating at an exposure level of 14,200 mwd /MTV, so a

E ANF-89-057 Page B-2 local power distribution typical of a nodal exposure of 15,000 mwd /MTU was selected as the design basis local power distribution. This distribution is shown in Figure B.2. For ANF-3 3.17/9Gd4, the limiting locals were also found to occur at 15,000 mwd /MTU, and this distribution is shown in Figure B.3.  !

Uncontrolled local power peaking distributions for XN-2 fuel were reviewed. I The limiting locals were found to occur at 25,000 mwd /MTU for XN-2 fuel, and this distribution is shown in Figure B.4. Local power distributions for centrally located XN-1 fuel were reviewed. The limiting locals were found to occur at 35,000 mwd /MTU, and this distribution is shown in Figure B.5. A '

bounding flat local power distribution was selected for the XN-1 fuel in the '

peripheral low power region. This distribution is shown in Figure B.6.

The limiting axial power profile used in the analysis had an axial peak j greater that the expected axial peak at E0C; and an axial offset more than

+2.99% greater than the expected axial offset at EOC.

I!

B.4 CALCULATION OF THE NUMBER OF RODS IN BOILING TRANSITION The methodology of Reference B-1 was used to analyze the number of fuel rods in boiling transition. The XN-3 correlation (B-2) was used to predict critical heat flux phenomena. Five hundred Monte Carlo trials were performed to support the MCPR safety limit. Non-parametric tolerance limits (B-3) were ,

used in lieu of Pearson curve fitting. The uncertainties used in the analysis for normal conditions were those identified in Reference B-1. At least 99.9% l of the fuel rods in the core are expected to avoid boiling transition .with a confidence level of 95%.

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• : 0.86 : 0.89 : 0.95 : 1.04 : 1.03 : 1.04 : 0.95 : 1.00 : 0.95 :

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1.03 : 0.92 : 1.03 : 1.00 : 0.00 : 0.99 : 1.06 : 1.08 : 1.04 :

1.04 : 1.08 : 1.04 : 1.01 : 0.99 : 0.00 : 1.04 : 0.95 : 1.05 :

0.95 : 0.98 : 1.04 : 1.05 : 1.06 : 1.04 : 1.07 : 0.99 : 0.96 :

1.00 : 1.04 : 0.98 : 0.94 : 1.08 : 0.95 : 0.99 : 0.93 : 1.00 :

0.95 : 1.00 : 0.95 : 1.05 : 1.04 : 1.05 : 0.96 : 1.00 : 0.95 :

FIGURE B.2 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF-3 3.33/9Gd5 FUEL I

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ANF-89-057 Page B-5

, *: 0.87 : 0.91 : 0.95 : 1.04 : 1.03 : 1.04 : 0.95 : 0.99 : 0.97 :

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• 0.95 : 0.97 : 0.90 : 1.04 : 1.03 : 1.04 : 1.04 : 0.98 : 0.95 :

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FIGURE B.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF-3 3.17/9Gd4 FUEL

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• : 0.95 : 0.98 : 0.93 : 1.05 : 1.05 : 1.06 : 1.05 : 0.99 : 0.96 :

• : 1.01 : 1.05 : 1.05 : 1.03 : 1.03 : 1.03 : 1.06 : 0.96 : 1.01 :

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1.01 : 1.06 : 1.06 : 1.03 : 1.01 : 0.00 : 1.04 : 0.98 : 1.02 :

0.95 : 0.97 : 1.05 : 1.06 : 1.06 : 1.04 : 1.06 : 0.98 : 0.96 :

0.97 : 1.01 : 0.99 : 0.96 : 1.06 : 0.98 : 0.98 : 0.95 : 0.98 :

0.95 : 0.97 : 0.96 : 1.01 : 1.01 : 1.02 : 0.96 : 0.98 : 0.96 :

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, I l FIGURE B.4 DESIGN BASIS LOCAL POWER DISTRIBUTION l ADVANCED NUCLEAR FUELS XN-2 FUEL 1

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• 0.91 : 0.92 : 0.98 : 0.97 : 1.06 : 0.96 : 0.99 : 0.93 : 0.95 :

1 i *: 0.93 : 0.98 : 0.95 : 1.08 : 1.08 : 1.09 : 1.06 : 0.98 : 0.94 :

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• 0.99 : 1.06 : 1.08 : 1.06 : 0.00 : 1.03 : 1.09 : 1.07 : 1.00 :

• 1.00 : 0.96 : 1.09 : 1.07 : 1.03 : 0.00 : 1.06 : 0.97 : 1.00 :

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0.94 : 0.99 : 1.06 : 1.09 : 1.09 : 1.06 : 1.06 : 0.99 : 0.95 :

0.95 : 0.93 : 0.98 : 1.07 : 1.07 : 0.97 : 0.99 : 0.94 : 0.96 :

0.94 : 0.95 : 0.94 : 1.00 : 1.00 : 1.00 : 0.95 : 0.96 : 0.94 : 1 FIGURE B.5 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-1 CENTRAL FUEL

I ANF-89-057 Page B-8 3

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=********************** ,

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• :...........___..___..........................................-: E

• :  :  :  :  :  : :  :  : E

• 1.00 : 1.00 : 1.00 : 1.00 : 0.00 : 1.00 : 1.00 : 1.00 .: 1.00 :

• 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 0.00 : 1.00 : 1.00 : 1.00 :

1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 3

:  :  :  :  :  : :  :  : 3

1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 :

1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 :

FIGURE B.6 DESIGN BASIS LOCAL POWER DISTRIBUTION .

ADVANCED NUCLEAR FUELS XN-1 PERIPHERAL FUEL I

I I

ANF-89-057 Page B-9 B.5 REFERENCES B-1. T. W. Patten, " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors," XN-NF-524(A), Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1983.

B-2. R. B. Macduff and T. W. Patten, "The XN-3 Critical Power Correlation,"

XN-NF-512(A), Revision 1, and Supplement 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, October 1982. p B-3. Paul N. Somerville, " Tables for Obtaining Non-Parametric Tolerance Limits," Annals of Mathematical Statistics, Volume 29, Number 2 P.

(June 1958), pages 599-601.

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ANF-89-057 .1 Issue Date: 5/11/89 1 i

SUSQUEHANNA UNIT 2 CYCLE 4 PLANT TRANSIENT ANALYSIS 1 Distribution:

D. J. Braun 0.-C. Brown.

R. E. Collingham L. J. Federico R. G. Grummer K.' D. Hartley M.'J.'Hibbard A. L.'B. Ho M..L. Hymas

+. S..E.-Jensen T. H. Keheley T.'L. Lotz T. E. Millsaps.-

- L. A. Nielsen G. L.'Ritter-C. C. Roberts, Jr.

C.'J..Volmer R. B. Stout H. E. Williamson H. G. Shaw/PP&L (20)

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