ML17157A245

Cycle 6 Plant Transient Analysis.

ML17157A245
Person / Time
Site:	Susquehanna
Issue date:	05/31/1990
From:	Oleary P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17157A243	List: ML17157A242 ML17157A244 ML17157A245 ML18026A235
References
ANF-90-049, ANF-90-49, NUDOCS 9007120262
Download: ML17157A245 (63)

Text

ANF-90-049 PU ADVANCEDNUCLEAR FUELS CORPORATION SUSQUEHANNA UNIT 1 CYCLE 6 PLANT TRANSIENT ANALYSIS MAY 1990

~ 007120262 900702 F'DR ADOCK 050003 7 F( PDi".

A Siemens Company

ADVANCEDNUCLEARFUELS CORPORATION ANF-90-049 Issue Date: 05/14/90 SUSQUEHANNA UNIT 1 CYCLE 6 PLANT TRANSIENT ANALYSIS Prepared by

/rc"7 /I/c X P. H. O'Lea BMR Fuel Engineering Fuel Engineering and Licensing Hay 1990

CUSTOMER OISCLAIMER IMPORTANT NOTICE REGAROINQ CONTENTS ANO USE OF T)fIS OOCU MENT PLEASE REAO CAREFULLy Advanced Nuclear Fuels Corporation's warranties and representations con-ceming the subject matter of this document are those set fonh in!he Agreement between Advanced Nuclear Fuels Corporation and the Custoiner pursuant to which this document is issued. Accordingly, except as otherwise expressly pro.

vided in such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation. expressed or iinplied, with respect to the accuracy, completeness. or usefulness of the infor-madon contained in this document, or that the use of any information. apparatus.

method or process disclosed in this document vnll not infnnge pnvately owned rights; or assumes any liabilities with respect to the use of any informauon, ap.

paratus, method or process disclosed in this document.

The informatktn contained herein is for the sole use of Customer.

In order to avoid impairment of rights of Advanced Nuctear Fuels Corporation in patents or invenbons which may be included in the information conttuned in this document, the recipient, by its acceptance of this document. agrees not to publish or make public use (in the patern use of the term) of such informauon until so authorised in wnting by Advanced Nuclear Fuels Corporabon or unul after six (8) months foliovring termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreeinent. No rights or licenses in or to any patents are implied by!he fumisning of this docu-ment.

ANF-3145.472A (12/87)

ANF-90-049 Page i TABLE OF CONTENTS Section ~Pa e 1;0 INTRODUCTION 1 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 (FWCF) 7 3.2.3 Loss of Feedwater Heating . . 8 3.3 Calculational Model ~ ~ ~ 8 3.4 Safety Limit . 8

4. 0 MAXIMUM OVERPRESSURIZAT ION 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

ANF-90-049 Page ii LIST OF TABLES Table Pacae 2.1 TRANSIENT ANALYSIS RESULTS AT DESIGN BASIS CONDITIONS ~ ~ 4 3.1 REACTOR DESIGN AND PLANT CONDITIONS FOR SUSQUEHANNA UNIT 1 9 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 1 10 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES . 13 3.4 FEEDWATER CONTROLLER FAILURE ANALYSIS RESULTS AT 100% FLOW 14 A.1 SINGLE-LOOP OPERATION REACTOR AND PLANT CONDITIONS . A-6 IST OF FIGUR S

~Fi ere Pacae 3.1 LOAD REJECTION MITHOUT BYPASS 15 3.2 LOAD REJECTION WITHOUT BYPASS 16 3.3 FEEDWATER CONTROLLER FAILURE . 17 3.4 FEEDWATER CONTROLLER FAILURE . 18 3.5 LOSS OF FEEDWATER HEATING 19 3.6 LOSS OF FEEDWATER HEATING 20

5. 1 SUSQUEHANNA UNIT 1 CYCLE 6 REDUCED FLOW MCPR OPERATIN G LIMIT . 24 8.1 SUSQUEHANNA UNIT 1 CYCLE 6 DESIGN BASIS RADIAL POWER HISTOGRAM 8-4 8.2 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-364L-9G4 ANF-5 FUEL . 8-5 8.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-331L-10G5 ANF-5 FUEL 8-6 8.4 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF-4 9X9 FUEL 8-7 8.5 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 FUEL . 8-8 8.6 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 PERIPHERAL FUEL 8-9 8.7 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-2 PERIPHERAL 8X8 FUEL 8-10

ANF-90-049 Page 1

1.0 INTRODUCTION

This report presents the results of Advanced Nuclear Fuels Corporation's (ANF's)* evaluation of system transient events for Susquehanna Unit 1 Cycle 6 operation. The evaluation, together with core transient events, determines the necessary thermal margin (HCPR 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.

• Formerly Exxon Nuclear Company (ENC).

ANF-90-049 Page 2 2.0

SUMMARY

The Susquehanna Unit 1 Cycle 6 core is composed of the following ANF fuel types:

Number of Bundle Average

~Fuel T e Assemblies Enrichment ANF-5 108 3.52/9Gd4*

112 3.21/10Gd5 ANF-4 228 3.33 ANF XN-3 240 3.31 ANF XN-2 76 2.89'pplication of ANF's methodology to the Cycle 6 core confirmed that the most limiting anticipated plant system transient with regard to thermal margin at rated power and flow conditions is the generator load rejection without bypass (LRWB) transient with recirculation pump trip (RPT) operable. The minimum critical power'atio (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 at design basis conditions, at end of cycle with all rods out. The results were used to determine the reported MCPRs. Results with RPT out of service are reported in Section 3.2. 1. The control rod withdrawal error (CRWE) analysis and delta-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),

using the scenario as specified by the ASME Pressure Vessel Code. This analysis verified that the safety valves of Susquehanna Unit 1 have sufficient capacity and performance to prevent the pressure from reaching the established transient pressure safety limit of 110% of design pressure

• The first number following the slash (/) states the number of gadolinia rods 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.'he gadolinia concentrations in the others are not significant since most of the gadolinia has been burned.

ANF-90-049 Page 3 (1. 1 x 1250 1375 psig). The maximum system pressures predicted during the event are 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 HCPR safety limit of 0.01.

0 ANF-90-049 Page 4 TABLE 2.1 TRANSIENT ANALYSIS RESULTS AT DESIGN BASIS CONDITIONS*

Transient A CPR MCPR**

Load Rejection Without Bypass 0.28/1.34 With Recirculation Pump Trip Feedwater Controller failure 0.23/1.29 With Bypass Loss of Feedwater Heating 0.16/1.22 Maximum Pressure si Vessel Transient Dome Vessel Lower Plenum Steam Line MSIV Closure 1,295 1,312 1,298

• 104% power/100% flow.

• • Based on the MCPR safety limit of '1.06 confirmed herein.

ANF-90-049 Page 5 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN

3. 1 ~iB Consistent with the FSAR plant .transient analysis, thermal margin operating HCPR limits are determined based on the 104/ power/1005 flow operating point. This thermal margin operating HCPR 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 power/flow operating map. The plant conditions for the 104% power/1005 flow point are as shown in Table 3. 1. The most limiting point in Cycle 6 is at the maximum Cycle 6 licensing exposure limit when control rods are fully withdrawn from the core. The thermal margin limit established for this exposure condition is conservative for cases where control rods are partially inserted.

Observance of a HCPR operating limit for ANF .9x9 and Bx8 fuel of 1.34 or greater conservatively protects against boiling transition during anticipated plant systems transients for Susquehanna Unit 1 Cycle 6.

The calculational models used to determine thermal margin include ANF's plant transient and core thermal-hydraulic codes described in previous documentation.(1~3 ) 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 1 startup test data. The Susquehanna system transient model was benchmarked to appropriate Unit 1 startup test data, as required by the SER issued for Reference 8. XCOBRA-T(9) was used to calculate the change in critical power ratio (delta-CPR) for pressurization event analyses.

3.2 Antici ated ransients 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 6 of Susquehanna Unit 1. These transients are:

ANF-90-049 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'ection Without B ass The load rejection without bypass event is the most limiting of the class 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 flow, which could mitigate the pressurization effect, is not allowed. The excursion of core power due to void collapse is primarily terminated by reactor scram and void growth caused by RPT. Figures 3. 1 and 3.2 depict the time variance of critical reactor and plant parameters during the load rejection transient calculation with bounding assumptions. The bounding assumptions are 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 Integral power increased by IOXo.

At design basis conditions (104% power/100% flow), a delta-CPR of 0.28 was calculated for the load rejection without bypass 'vent when RPT is operable.

ANF-90-049 Page 7 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 FWCF Failure of the feedwater control system is postulated to lead to a maximum increase in feedwater flow into the vessel. As the excessive 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 the turbine control valves, thus initiating a reactor scram. The analyses for Cycle 6 modelled the closure of the turbine control valves rather than the turbine stop valves because of the faster valve stroke times as a function of steam flow, as provided by the utility. 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. For Cycle 6, the bypass control model was revised to more accurately represent the quick opening feature of the Susquehanna design.

The evaluation of the FWCF event at design basis conditions was performed with bounding values and resulted in a delta-CPR of 0.23. Figures 3.3 and 3.4 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 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 0.34 was calculated.

ANF-90-049 Page 8 3.2.3 Loss of Feedwater Heatin 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 event in Cycle 6 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 Hodel The plant transient code used to evaluate the generator load rejection and feedwater flow increase was ANF's code COTRANSA.(1) The axial one-dimensional neutronics model predicted reactor power shifts toward the core middle and top as pressurization occurred. This was accounted for explicitly in determining thermal margin changes in the transient. The loss of feedwater 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 calculate the delta-CPRs for the pressurization transients.

3.4 F Li The safety limit is the minimum value of the critical power ratio at which the fuel could be operated where the expected number of rods in boiling transition would not exceed 0.1/ of the fuel rods in the core. The safety limit is the HCPR permitted to occur during the limiting anticipated operational occurrence. A HCPR safety limit of 1.06 for all fuel types in Susquehanna Unit 1 Cycle 6 was supported by the methodology presented in Reference 3. The input parameters used to support the HCPR safety limit are presented in Appendix B of this report.

e ANF-90-049 Page 9 TABLE 3.1 REACTOR DESIGN AND PLANT CONDITIONS FOR SUSQUEHANNA UNIT 1 Reactor Thermal Power (104K') 3439 HWt Total Core Flow (100/) 100.0 Hlb/hr Core In-Channel Flow . 89.9 Hlb/hr Core Bypass Flow 10. 1 Hlb/hr Core Inlet Enthalpy 518.0 Btu/1 bm Vessel Pressures Steam Dome 1035 psia Upper Plenum 1046 psia Core 1053 psia Lower Plenum 1068 psia Turbine Pressure 975 psia Feedwater/Steam Flow 14.15 Mlb/hr Feedwater Enthalpy 362.6 Btu/ibm Recirculation Pump Flow (per pump) 15.75 Hlb/hr

ANF-90-049 Page 10 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUS(UEHANNA UNIT 1 High Neutron Flux Trip 125.3/o 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 Control Valve Closure 30 msec Time from High Neutron Flux Trip to Control Rod Notion 290 msec Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 905 open Turbine Control Valve Stroke Time (Rated Power) 71 msec Fuel/Cladding Gap Conductance Core Average (Constant) 776.8 Btu/hr -ft2-F Safety/Relief Valve Performance Technical Specifications Settings Relief Valve Capacity 225.4 ibm/sec (1110 psig)

Pilot Operated Valve Delay/Stroke 400/150 msec 1

,e ANF-90-049 Page 11 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 1 (Continued)

HSIV Stroke Time 2.0 sec HSIV Position Trip Setpoint 85% open Turbine Bypass Valve Performance Total Capacity 936.05 ibm/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 Naximum Feedwater Runout Flow Three Pumps 5049 ibm/sec Recirculation Pump Trip Setpoint Vessel Pressure 1170 psig I

• COTRANSA plots represent water level relative to the steam separator skirt and the value here is relative to instrument zero.

ANF-90-049 Page 12 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 1 (Continued)

Control Characteristics Sensor Time Constants Pressure 500 msec Other.s 250 msec Feedwater Control Hode Three-Element Feedwater Master Controller Proportional Gain 50 0 (/o//o) Po/ ft )

Reset Rate 1.70 P/sec/ft)

Feedwater ION'ismatch Water Level Error 4.0 ft Steam Flow Equivalent 4.03 ft/1001o Recirculation Flow Control Hode Manual Pressure Regulator Settings Lead 3.0 sec Lag 7.0 sec Gain 3.33//psid

0 ANF-90-049 Page 13 TABLE 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES Maximum Maximum Maximum Core Average System Neutron Flux Heat Flux Pressure Event / Rated / Rated ~si a A CPR Load Rejection 394 123.3 1199 0.28 Without Bypass Feedwater Controller 212 117.9 1179 0.23 Failure Loss of Feedwater 123 121.5 1081 0.16 Heating MSIV Closure With 550 138.9 1312 NA Flux Scram NOTE: All events are bounding case at 1045 power/100/ .flow.

ANF-90-049 Page 14 TABLE 3.4 FEEDWATER CONTROLLER FAILURE ANALYSIS RESULTS AT 100% FLOW I Power h CPR 104 0.23 80 0.25 65 0.27 40 0.27

1. NEUTRON FLUX LEVEL

2. HEAT FLUX

3. RECIRCULATION FLOW

4. VESSEL STEAN FLOH S. FEEOWATER FLOW C) 8Ol 12 45 123 3 5 2 8

CD C7 i 0.00 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2-25 2.50 TINEi SEC FIGURE 3.1 LOAO REJECTION WITHOUT BYPASS

l. VESSEL PRESSURE CHRN PEPSI)

2. VESSEL HRTER LEVEL (IN)

~0.00 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2-M 2-25 2.50 TINE, SEC FIGURE 3.2 LOAD REJECTION WITHOUT BYPASS

1. NEUTRON FLUX LEVEL

2. HEAT FLUX

3. RECIRCULATION FLOH

4. VESSEL STERN FLOH

5. FEEOHATER FLOH 12 0 12 q 1 2

3 12 16 20 24 28 32 TIHEp SEC FIGURE 3.3 FEEDWATER CONTROLLER FAILURE

1. VESSEL PRESSURE CHANGE (PSI)

2. VESSEL HATER LEVEL ( IN) ll I

Dl LD 12 16 20 28 32 40 cQ DI ED TIMEp SEC 00 %D FIGURE 3.4 FEEOWATER CONTROLLER FAILURE

1. NEUTRON FLUX LEVEL

2. HEAT FLUX

3. RECIRCULATION FLOH

4. VESSEL STERN FLOH

5. FE'fDHATER FLOH 5 3 rl I

40 50 70 SV lO lQ CP D

TINE, SEC ED M tD fIGURE 3.5 LOSS Of fEEDWATER HEATING

1. VESSEL PRESSURE CHANGE (PSI)

2. VESSEL WATER LEVEL (IN) 40 50 60 Zp TINE, SEC FIGURE 3.6 LOSS OF FEEOMATER HEATING

ANF-90-049 Page 21

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 ASHE Pressure Vessel Code. This analysis showed that the safety valves of Susquehanna Unit 1 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.

The reactor conditions used in the evaluation of the maximum pressurization event are those shown in Table 3. 1. The most critical active component (scram on HSIV closure) was assumed to fail during the transient.

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

4.2 Pressurization Transients ANF has evaluated several pressurization events and has determined that closure of all main steam isolation valves without direct scram is the most limiting. Although the closure rate of the HSIVs 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 is concentrated toward the end of the valve stroke, generating a substantial compression wave. Once the containment is 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 (HSIV) closures to be more limiting for system pressure than turbine stop or control valve closures.

ANF-90-049 Page 22 Closure of'll Hain Steam Isolation Valves 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 two-second valve closing time is conservative compared to longer closing times. At about 2. 1 seconds, reactor scram initiated due to a high flux trip.

Since control rod scram insertion times were assumed at their Technical Specification limit, effective power shutdown is delayed until after 3.6 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 1,298 psig occurring near the vessel at about 5.5 seconds. The maximum vessel pressure was 1,312 psig 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.

ANF-90-049 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( ) showed the two-pump run-up bounds the single-pump run-up. Only the two-pump run-up is evaluated for Susquehanna Unit 1 Cycle 6. 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 HCPR 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 1 Cycle 6 analysis conservatively assumed the run-up event initiated at 57! power/4Ã 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 curve was conservatively extrapolated down to a core flow of 30% rated flow to calculate the flow dependent HCPR. The cycle specific HCPR limit for Susquehanna Unit 1 Cycle 6 shall be the maximum of the reduced flow MCPR operating limit, the full flow HCPR operating limit, or the power dependent HCPR operating limit.

ANF 9X9 1.S j 16 O

O 1A 9

8 1Z 1

30 50 60 70 80 90 100 110 TOTAL CORE RECIRCULATION FLOW,%-RATED FIGURE 5.1 SUSQUEHANNA UNIT 1 CYCLE 6 REDUCED FLOW MCPR OPERATING UMIT

ANF-90-049 Page 25

6.0 REFERENCES

R.

II "2,"~X-->>-1,2112,Add H. Kelley, Corporation*,

"Exxon Nuclear Plant Transient Methodology Richland, WA 99352, November 1981.

for Boiling I F

2. P. M. O'eary and T. L. Lotz, "Susquehanna Unit 1 Cycle 6 Reload Analysis, Design and Safety Analyses," ANF-90-050, Advanced Nuclear Fuels Corporation, Richland, WA 99352, March 1990.

3.

THERMEX:

~tN-NF- ->>,

J. A. White, "Exxon Nuclear Methodology for Boiling Water Reactors, Thermal Corporation, Richland, 2 I Limits Methodology, Summary Description,"

WA 2, R I I 2, Ad 99352, January 1987.

2 N I F for Boiling

"~XII-F-5,Critical T. W. Patten, "Exxon Nuclear Power Methodology lit R R 11 I, Ad 2 II I I Corporation, Richland, WA 99352, November 1983.

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

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

~t- -

Richland, K. R. Herckx, "RODEX2 Fuel Rod Mechanical Response WA R

99352, I I April 2, Ad 1984.

d II I I Evaluation Model,"

I 5 for Boiling

8. S.

R<<,"~X---7,2112,pp1<<1,2d E. Jensen, "Exxon Nuclear Plant Transient Methodology

~5, II Advanced Nuclear Fuels Corporation, Richland, WA 99352, March 1986,

9. H. J. Ades, "XCOBRA-T: A Computer Code for BWR Transient Thermal-ll d 11 I i*,'" 21 I d V I Supplement 1 and 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.

• Formerly Exxon Nuclear Company (ENC):

ANF-90-049 Page A-I APPENDIX A SINGLE-LOOP OPERATION Both ANF and the NSSS supplier have provided analyses which demonstrate the safety of 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 nodal power levels than are allowed when both recirculation systems are in operation. The physical interdependence between core power and recirculation flow rate, coupled with the technical specification constraints for the Susquehanna Units, inherently limit the core to less than rated power when operating with one recirculation loop out of service.

ANF fuel is designed to be compatible with the initial core and current 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 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.l 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 analysis models or compatible fuel designs. A review of the most limiting delta-CPR transients for single-loop operation was conducted. Under single-loop conditions, steady state operation cannot exceed approximately 76% power and 60% core flow because of the capability of the recirculation loop pump. Thus, the MCPR limit at maximum power (i.e., 76%) in single-loop operation is equal to the two-pump operating MCPR limit at 100% power. The flow dependence of

ANF-90-049 Page A-2 the MCPR limit 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; therefore, the conservative two-pump limit is retained.

A. l. 1 Load Re 'ection Without B ass The limiting anticipated system transient for the Susquehanna units at full power and full flow is the load rejection without bypass (LRWB) pressurization transient. In this transient, the primary phenomena is the pressurization caused by abruptly stopping the steam flow through rapid closure of the turbine control valve. When the rapid pressurization reaches the core it causes a power excursion due 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 closed. The resulting power excursion and associated margin change are reduced below those of the full power case. Analysis has shown that the difference in the two-loop operation and single-loop operation core flow characteristics do not adversely affect 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 reduced power and flow conditions in other BWRs with single-loop operation confirm this trend.

A. 1.2 Feedwater Controller Failure The second most limiting system 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 operation.

This transient assumes 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

ANF-90-049 Page A-3 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 amount of feedwater being injected.

At the reduced recirculation flows, the subcooling in the downcomer due to the high feedwater flow takes longer to traverse the core so that a high water level trip occurs before core power can rise as high as it does in the full flow case. As with the LRWB, the pressurization event resulting from the turbine trip is less severe for the reduced power in single-loop operation, Thus, because of the slower enthalpy transport phenomeha caused by the lower recirculation flow and because of the lower steam line flow in the pressurization portion of the transient, the consequences of a FWCF are less severe in single-loop operation than in two-loop operation.

A.1.3 MCPR Safet Limit For single-loop operation,'he NSSS vendor found that an increase of 0.01 in the HCPR safety limit was needed to account for the increased flow measurement uncertainties and increased power distribution uncertainties associated with single pump operation. ANF has evaluated the effects of the increased uncertainties on the HCPR safety limit and found that the NSSS vendor determined increase in the allowed HCPR safety limit is also applicable to ANF fuel during single-loop operation. Thus, increasing the HCPR safety limit by 0.01 for single-loop 'operation (1.07) with ANF fuel is sufficiently conservative to also bound the increased uncertainties for single-loop operation.

A.1.4 ~Summar The two-loop HCPR operating limit plus .Ol conservatively protects the fuel from any transients in single-loop operation.

ANF-90-049 Page A-4 A. 2 POSTULATED ACCIDENTS A.2. 1 Loss Of Coolant Accident ANF performed LOCA analyses for single-loop conditions and determined that the HAPLHGR limit curve for two-1'oop operation is also applicable to single-loop operation with ANF fuel (Reference A-l).

A.2.2 Pum Seizure Accident The seizure of a recirculation pump is considered as a design basis accident. It is a very mild accident relative to other design basis accidents such as the loss of coolant accident (LOCA). The pump seizure event 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.

ANF has analyzed the pump seizure accident from single-loop operating conditions on a generic basis for the Susquehanna Units. The results of the generic analyses show that single-loop operation of the Susquehanna Units with single-loop HCPR operating limits protects against the effects of the pump seizure accident. That is, for oper ation at the single-loop operating HCPR limit, the radiological consequences of a pump seizure accident from single-loop operating conditions're but a small fraction of the 10CFR100 guidelines.

ANF used safety limit HCPR methodology to determine the extent of rods which might experience boiling transition should the minimum critical power ratio reach a value of 0.90. Using input from the Unit 1 Cycle 6 analysis (including single-loop uncertainties), less than 2% of the fuel rods in the core were calculated to experience boiling transition at the 95% confidence level.

The core remains covered during the single-loop pump seizure accident, and any rods which experience boiling transition would be expected to be in the film boiling mode for only a short period of time. Because of the short duration of the accident, the number of fuel rods that would fail is

ANF-90-049 Page A-5 significantly less than the number of rods predicted to experience boiling transition. However, assuming all rods calculated to experience boiling transition fail, the number of rods failing is less than 2% of the rods in the core. The radiological consequences of 2% fuel failure are only a small fraction of the 10CFR100 guidelines, especially since no uncontrolled release path to the environment is created during the single-loop pump seizure accident.

The single-loop pump seizure accident has been analyzed using ANF's transient methodology for several cycles of operation for the Susquehanna Units. The calculated delta-CPRs have varied from 0.29 to 0.35. A delta-CPR of 0.40 is expected to bound this and future cycles. Therefore, a single-loop MCPR operating limit of 1.30 (i.e., 0.90 + 0.40) would be required to protect against the single-loop pump seizure accident. This generic MCPR operating limit for the single-loop pump seizure accident provides sufficient margin to offset any cycle specific variability of delta-CPR results.

A. 3 STAB I LITY PPKL will establish stability surveillance requirements for Susquehanna Unit 1 Cycle 6 in conformance with the interim operating guidelines presented in NRC Bulletin 88-07 Supplement 1 based on the calculation results prepared by ANF.

A. 4 TECHNICAL SPECIFICATIONS The single-loop HCPR operating limit for Unit 1 Cycle 6 is the greater of the two-loop HCPR operating limit plus 0.01 and the generic single-loop pump seizure HCPR limit of 1;30. The single-loop HCPR operating limit provides assurance that the consequences of a single-loop pump seizure accident would be a small fraction of 10CFR100 guidelines and in addition protects against all anticipated operational occurrences during single-loop operation.

ANF-90-049 Page A-6 TABLE A.1 SINGLE-LOOP OPERATION REACTOR AND PLANT CONDITIONS Reactor Thermal Power 2489 HWt Total Recirculation Flow 60.35 Hlb/hr Core Bypass Flow 5.66 Hlb/hr Core Inlet Enthalpy 507.3 Btu/lb Vessel Pressures Steam Dome 994.5 psia Lower Plenum 1010.9 psia Turbine Pressure 965.4 psia Steam Flow 9.85 Hlb/hr Feedwater Enthalpy 330.7 Btu/lb

ANF-90-049 Page A-7 A.4 REFERENCES 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. P. M. O'eary and T. L. Lotz, "Susquehanna Unit 1 Cycle 6 Reload Analysis, Design and Safety Analyses," ANF-90-050, Advanced Nuclear Fuels Corporation, Richland, WA 99352, April 1990.

ANF-90-049 Page B-1 APPENDIX B MCPR SAFETY LIMIT B. 1 INTRODUCTION The MCPR fuel cladding integrity safety limit was calculated using the methodology and uncertainties described in Reference B-1. In this methodology, a Honte Carlo procedure is used to evaluate plant measurement and power predictions uncertainties such that during sustained operation at the MCPR cladding integrity safety limit, 'at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. This appendix describes the calculation and presents the analytical results.

B. 2 CONCLUSIONS During sustained operation at a HCPR of 1.06 with the design basis power distribution described below, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition at a confidence level of 95%. The 1.06 MCPR safety limit is justified based upon a full power full flow MCPR operating limit of 1.34 or below.

8.3 DESIGN BASIS POMER DISTRIBUTION Predicted power distributions were extracted from the fuel management analysis for Susquehanna Unit 1 Cycle 6. The radial power distributions were evaluated for performance as the design basis radial power map, and the distribution at 9,625 HWd/HTU exposure was selected as the most severe expected distribution for the cycle. The distribution was skewed toward higher power factors by increasing the power of the limiting bundle to a radial peaking factor approximating an operating MCPR level of 1.34 at full power. The resulting design basis radial power distribution is shown in Figure 8.1.

The fuel management analysis indicated that the maximum power ANF bundle (ANF-5 3.52/9Gd4) in the core at the worst case exposure (9,625 HWd/HTU) was

ANF-90-049 Page B-2 predicted to be operating at an exposure level of 14,200 HWd/HTU, so a local power distribution typical of a nodal exposure of 15,000 HWd/HTU was selected as the design basis local power distribution. This distribution is shown in Figure B.2. For ANF-5 3.21/10Gd5, the limiting locals were also found to occur at 15,000 HWd/HTU, and this distribution is shown in Figure B.3.

Uncontrolled local power peaking distributions for ANF-4 fuel were reviewed.

The limiting locals were found to occur at 25,000 HWd/HTU for ANF-4 fuel, and this distribution is shown in Figure B.4. Local power distributions for centrally and peripherally located XN-3 fuel were reviewed. The limiting locals were found to occur at 35,000 HWd/HTU, and these distributions are shown in Figures B.5 and B.6, respectively. A bounding flat local power distribution was selected 'for the XN-2 fuel in the peripheral low power region. This distribution is shown in Figure B.7.

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

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

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 Honte'arlo trials were performed to support the HCPR safety limit. Non-parametric tolerance limits(B ) 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%

of the fuel rods in the core are expected to avoid boiling transition with a confidence level of 95%.

B.5 ASSESSHENT OF THE EFFECTS OF CHANNEL BOW On March 20, 1990, the USNRC issued bulletin no. 90-02 requiring assessment of loss of thermal margin caused by channel box bow.(B ) In response to this bulletin, ANF generically evaluated the effects of channel box bow. ANF determined that for reactors monitoring ANF 8x8-2 and 9x9-2 fuel assemblies using the XN-3 critical power correlation, sufficient conservatism

ANF-90-049 Page 8-3 exists in the correlation that more than offsets the 0.02 maximum delta-CPR effect due to channel box bow when channels are used only one lifetime in C-lattice plants. The details of this evaluation have been reported generically to the NRC.(8 5)

120 Cl Q3 60 C3 lL 40 0 l1l ll I

a 0.2 0.4 0.6 0.8 1 1.6 ELl 0 D I RADIAL PONER PERKING U7 I

O M LD FIGURE B.I SUSQUEHANNA UNIT I CYCLE 6 DESIGN BASIS RADIAL POWER HISTOGRAM

ANF-90-049 Page B-5

• ~

• : 0.95 : 0.90  : 0.96  : 1.04 : 1.02 : 1.04 : 0.96 : 1.01  : 0.95  :

• ~

• ~

• ~

• : 0.90 : 0.93  : 0.98  : 1.07  : 0.91 : 1.08 : 0.96 : 1'.04 : 1.01

• ~

• ~

• ~

• : 0.96 : 0.98  : 0.89  : 1.03  : 1.02 :- 1.04 : 1.04 : 0.98  : 0.96  :

• ~

• ~

• ~

• : 1.04 : 1.07 1.03  : 0.99  : 0.99 : 1.00 : 1.05 : 0.93 1.04  :

• ~

• ~

• ~

• : 1.02 : 0.91 1.02  : 0.99  : 0.00 : 0.97 : 1.04 : 1.08  : 1.04  :

• ~

• ~

• ~

• : 1.04 : 1.08 1.04  : 1.00  : 0.97 : 0.00 : 1.03 : 0;94  : 1.05  :

• ~

• ~

0.96 : 0.96  : 1.04  : 1.05  : 1.04 : 1.03 : 1.06 : 0.99  : 0.97  :

1.01 1.04  : 0.98  : 0.93  : 1.08 : 0.94 : 0.99 : 0.94  : 1.02  :

0.95 : 1.01  : 0.96  : 1.04  : 1.04 : 1.05 : 0.97 : 1.02  : 0.96  :

FIGURE B.2 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-364L-9G4 ANF-5 FUEL

ANF-90-049 Page B-6

• ~

0.95  : 0.96 : 0.97  : 1.05 : 1.04 : 1.05  : 0.96 : 1.01 : 0.98

• ~

• ~

• : 0.96  : 0.94  : 0.98  : 1.02 : 0.93 : 1.01  : 0.98 : 1.05 : 1.01

• ~

• ~

• ~

• : 0.97  : 0.98  : 0.94  : 1.04 : 1.03 : 1.04  : 0.98 : 1.00 : 0.97  :

• ~

• ~

• ~

• : 1.05 : 1.02 1.04  : 1.00  : 0.99 : 1.01  : 1.05 : 0.95 : 1.05  :

• ~

• ~

• ~

• : 1.04  : 0.93 1.03  : 0.99  : 0.00 : 0.93  : 1.05 : 1.01 : 1.04

• ~

• ~

• ~

• ' 05 ' 01 1.04  : 1.01  : 0.93 : 0.00  : 0.98 : 0,96 : 1.05  :

• ~

• ~

0.96  : 0.98  : 0.98  : 1.05  : 1.05 : 0.98  : 1.00 : 1.01 : 0.97  :

1.01  : 1.05  : 1.00  : 0.95  : 1.01 : 0.96  : 1.01 : 0.95 : 1.01

0.98  : 1.01  : 0.97 : 1.05  : 1.04 : 1.05  : 0.97 : 1.01 : 0.98  :

FIGURE B.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-331L-10GS ANF-5 FUEL

ANF-90-049 Page 8-7

• ~

• : 0.91  : 0.92  : 0.95 : 1.01 : 1.00 : 1.01 : 0.95 : 0.97 : 0.95  :

• ~

• ~

• ~

• : 0.92  : 0.94  : 0.98 : 1.06 : 0.94 : 1.06 : 0.97 : 1.01 : 0.97  :

• ~

• ~

• ~

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

• ~

• ~

• ~

• ~

1 01 1.06  : 1.06 : 1.03 : 1.03 : 1.04 : 1.06 : 0.96 : 1.01

• ~

• ~

• ~

• : 1.00  : 0.94  : 1.05 : 1.03 : 0.00 : 1.01 : 1.06 : 1.07 : 1.01

• ~

0

• ~

• ~

• ~

1 01 1.06 1.06 : 1.04 : 1.01 : 0.00 : 1.04 : 0.96 : 1.02  :

• ~

• ~

0.95  : 0.97  : 1.05 : 1.06 : 1.06 : 1.04 : 1.06 : 0.99 : 0.96  :

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

0.95  : 0.97  : 0.96 : 1.01 : 1.01 : 1.02 : 0.96 : 0,98 : 0.95  :

FIGURE B.4 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF-4 9X9 FUEL

ANF-90-049 Page B-8

• ~

• : 0.96  : 0.94 : 0.96  : 1.00 : 1.00 : 1.00  : 0.96 : 0.98 : 0.98  :

• ~

• ~

• ~

• 0.94 : 0.95  : 0.98  : 1.04 : 0.95 : 1.04  : 0.97 : 1.00 : 0.98  :

• ~

• ~

• ~

• : 0.96 : 0.98  : 0.94  : 1.05 : 1.05 : 1.06  : 1.04 : 0.99 : 0.96  :

• ~

• ~

• ~

~

1.00 1.04 1.05  : 1.04 : 1 '4 : 1.04  : 1.06 : 0.96 : 1.00

• ~

• ~

• : 1.00 : 0.95  : 1.05  : 1.04 : 0.00 : 1.02  : 1.06 : 1.05 : 1.00  :

• ~

• ~

• ~

• : 1.00 1.04  : 1.06  : 1.04 : 1.02 : 0.00  : 1.04 : 0.99 : 1,00  :

• ~

• ~

0.96 : 0.97.: 1.04  : 1.06 : 1.06 : 1.04  : 1.05 : 0.97 : 0.97  :

0.98 : 1.00  : 0.99  : 0.96 : 1.05 : 0.99 *: 0.97 : 0.96 : 0.98  :

0.98 : 0.98  : 0.96  : 1.00 : 1.00 : 1.00  : 0.97 : 0.98 : 0.98  :

FIGURE B. 5 DESIGN BASIS LOCAL POMER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 FUEL

ANF-90-049 Page B-9

• ~

~

0.93  : 0 '3  : 0.95  : 1.00 : 1.00 : 1.00 : 0.96 : 0.97 : 0.96  :

• ~

• ~

• : 0.93  : 0.94  : 0.98  : 1.05 : 0.95 : 1.05 : 0.97 : 1.01 : 0.97  :

• ~

• ~

• ~

• : 0.95  : 0.98  : 0.94  : 1.05 : 1.05 : 1.06 : 1.04 : 0.99 : 0.96  :

• ~

• ~

• ~

• : 1.00  : 1.05 1.05  : 1.03 : 1.03 : 1.04 : 1.06 : 0.96 : 1.01

• ~

• ~

• ~

• : 1.00  : 0.95 1.05  : 1.03 : 0.00 : 1.02 : 1.06 : 1.05 : 1.01

• ~

• ~

• ~

• ' 00 ' 05 1.06  : 1.02 : 1.04 : 0.00 : 1.04 : 0.98 : 1.01

• ~

• ~

0.96  : 0.97  : 1.04  : 1.06 : 1.06 : 1.04 : 1.06 : 0.98 : 0.96  :

~ ~

0.97  : 1.01  : 0.99  : 0.96 : 1.05 : 0.98 : 0.98 : 0.95 : 0.98  :

0.96  : 0.97  : 0.96  : 1.01 : 1.01 : 1.01 : 0.96 : 0.98 : 0.97  :

FIGURE B. 6 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 PERIPHERAL FUEL

0 ANF-90-049 Page B-10

• ~

• ~

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

• ~ 4

• ~

• ~

• ~

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 : 1.00 : 0.00 : 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  : 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.7 DESIGN BASIS LOCAL POMER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-2 PERIPHERAL 8X8 FUEL

ANF-90-049 Page B-11 B.6 REFERENCES B-1. T. W.

R,"

Patten, Corporation, Richland, B-2. R. B. Macduff and T.

-512, I

~,

"Exxon Nuclear WA W. Patten, Critical R 11 99352, November 1983.

"The XN-3 Power I,

t I, Methodology Ad Critical d

Power II for Boiling I I Correlation,"

~tN-N R I I, 5 51 Ad d II I F Corporation, Richland, WA 99352, October 1982.

B-3. Paul N. Somerville, "Tables for Obtaining Non-Parametric Tolerance Limits," Annals of Mathematical Statistics, Volume 29, Number 2 (June 1958), pages 599-601.

B-4. NRC Bulletin No. 90-02, "Loss of Thermal Margin Caused by Channel Box Bow," March 20, 1990.

B-5. Memo, R. A. Copeland (ANF) to R. C. Jones (NRC), "Loss of Thermal Margin Caused by Channel Box Bow," RAC:030:90, April 9, 1990.

ANF-90-049 Issue Date: 05/14/90 SUS(UEHANNA UNIT 1 CYCLE 6 PLANT TRANSIENT ANALYSIS Distribution:

L. J. Federi co A. L. B. Ho H. L. Hymas S. E. Jensen T. L. Lotz S. C. Mellinger T. E. Hillsaps L. A. Nielsen P. H. O'eary C. C. Roberts, Jr.

R. B. Stout H. E. Williamson H. G. Shaw/PP&L (25)

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