ML17146B094

Rev 1 to Susquehanna Unit 2 Cycle 3 Plant Transient Analysis.

ML17146B094
Person / Time
Site:	Susquehanna
Issue date:	11/30/1987
From:	Jason White ADVANCED MEDICAL SYSTEMS, INC.
To:
Shared Package
ML17146B090	List: ML17146B089 ML17146B092 ML17146B094 ML17146B095 ML17146B096 ML17146B102 ML18040B194
References
ANF-87-125, ANF-87-125-R01, ANF-87-125-R1, NUDOCS 8712310156
Download: ML17146B094 (60)

Text

ANF-87-125 R E V IS ION AQMHCSDo HAIL,EARPUDDLES CORPORATION SUSQUEHANNA UNIT 2 CYCLE 3 PLANT TRANSIENT ANALYSIS NOVEMBER 1987 AN AFFILIATEOF KRAFTIVERK UPIION Q~ KMfU 87i23iOi56 87i'223 05000389',

ADOCK PDR

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ADVANCEDNUCLEARFUELS CORPORATION ANF-87-125 Revision 1 Issue Oate: 11iIOi87 SUSQUEHANNA UNIT 2 CYCLE 3 PLANT TRANSIENT ANALYSIS Prepared By:

J. A. White BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services AH AFFIUATE OF KRAFTWFRK VHIOH Qxwu

CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con-ceming the subject matter of this document are those set forth In the Agreement between Advanced Nuclear Fuels Corporation and the Customer pursuant to which this document ls 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 implied. with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rights; or assumes any liabilities with respect to the use of any information, ap-paratus, method or process disclosed in this document.

The information contained herein Is for the sole use of Customer.

In order to avoid impairment of r/ghts of Advanced Nuclear Fuels Corporation in patents or inventions which may be included in the information contained In this document. the recipient, by its acceptance of this document, agrees not to publish or make public use gn the patent use of the term) of such information until so authorized in writing by Advanced Nuclear Fuels Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this docu-ment.

XN.NF.F00-765 (1/Bi

ANF-87-125 Revision 1 TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

~ t~~~~~~~~~~~~~~~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t~~~~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 2.0 S UMMARYo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN.............................. 5 3.1 D esign Basss....................... 1

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ 5 3.2 Anticipated Transients............. ~ ~ ~ ~ ~ ~ ~ ~ ~ t~\~~~~~~~~~~~~ ~ ~ ~ ~ ~ ~ ~ 6 3.2.1 Load Rejection Without Bypass...... ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 3.2.2 Feedwater Controller Failure....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ 7 3.2 ' Loss Of Feedwater Heating.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3.3 Calculational, Model................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3.4 Safety Limit......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I~ ~ ~ 4 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9 MAXIMUM OVERPRESSURIZAT ION................................... 22 D

0 esign Bases................................................... 22 4.2 Pressurization Transients. ~ ............... ~ ~ ~ ~ ~ 22 4.2. 1 Closure Of All Hain Steam Isolation Valves.......,. 23

5. 0 RECIRCULATION PUMP RUN-UP......................................... 24

6.0 REFERENCES

............. ~ ~ ~ ~ ~ ~ ~ ~ t~~~~~~~~t~~~~~~~~~ ~ ~ ~ ~ ~ ~ ~ ~ 26 APPENDICES A. SINGLE LOOP OPERATION..........,.......'........................ .. A-1 B. HCPR SAFETY LIMIT.............................. ........ .. B-l

g t

~ ~

11 ANF-87-125 Revision 1 LIST OF TABLES Table ~Pa e 2.1 Transient Analysis Results At Design Basis Conditions... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 3.1 Reactor Design And Plant Condi.tions Susquehanna Unit 2.. ~ ~ o ~ ~ ~ ~ ~ ~ ~ 10 o t 3.2 Significant Parameter Values Used In The Analysis For Susquehanna Unit 2...................................... o ~ ~ ~ ~ ~ ~ ~ ~ ~ ll 3.3 Results Of System Plant Transient Analyses.............. o ~ ~ ~ oo ~ ~ ~ ~ 14 3.4 Feedwater Controller Failure Analysis Results At 100% Fl owo ~ ~ ~ ~ ~ ~ o 15 A.l SLO Reactor And Plant Conditions...:.................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

LIST OF FIGURES Ficiur'e ~Pa e 3.1 Load Rejection Without Bypass............................. 16 3.2 Load Rejection Without Bypass.......................... 17 3.3 Feedwater Controller Failure...... ... 18 3.4 Feedwater Controller Failure...................................... 19 3.5 Loss Of Feedwater Heating.. ~ ........................... 20 3.6 Loss Of Feedwater Heating.................. 21 5.1 Susquehanna Unit 2 Cycle 3 Reduced Flow MCPR Operating 25 A.l Single Loop Operation - Pump Seizure....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o A.2 Single Loop Operation - Pump Seizure.. .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ A 9 A.3 Core Power Versus Core Flow........... .. ............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ A 10 B.3-1 Susquehanna Unit 2 Cycle 3 Design Basis Radial Power Hi stogram.... B-4 B.3-2 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-2 9x9 Fuel...............,. B-5 B.3-3 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-1 9x9 Fuel.. ~ ~ ~ ~ ~ ~ ~ ~ ~

ill 0 ~ ~

ANF-87-1P Revision LIST OF FIGURES (Continued)

~Fi ere Parcae B.3-4 Design Basis Local Power Distribution General Electric (Central) 8xSR Fuel.............................. B-7 B.3-5 Design Basis Local Power Distribution General Electric (Peripheral) SxSR Fuel........................... B-8

ANF-87-125 Revision 1

1.0 INTRODUCTION

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

evaluation of system transient events for Susquehanna Unit 2 Cycle '

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. The bases for these analyses have been provided in Reference l.

'Formerly Exxon Nuclear Company (ENC).

f~

ANF-87-125 Revision 1 2.0

SUMMARY

The Susquehanna Unit 2 Cycle 3 core can be .described as follows:

No. of Bundle Average

~Feel T e ~Ass Enrichment ANF XN-2 140 3.33/9Gd4*

XN-2 96 3.33/10Gd5 ANF XN-1 324 3.31 GE 8x8R 196 2.19 e~

8x8R 8 Using ANF's methodology and considering the Cycle 3 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. The values in Table 2. 1 were determined assuming bounding conditions in the analyses. Results with RPT out of service are reported in Section 3.2. 1. These transients were evaluated with all co-resident fuel types modeled and the most limiting condition was used to determine the reported MCPRs. The Control Rod Withdrawal Error (CRWE) analysis and resulting 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, using the scenario as specified by the ASME Pressure Vessel Code. This analysis shows that the safety valves of Susquehanna Unit 2 have sufficient capacity and

• The first number 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.

The others are not significant.

ANF-87-1 Revision performance to prevent the pressure from reaching the established transient pressure safety limit of 110% of design pressure (l. 1 x 1250 = 1375 psig).

The analysis also assumed six safety relief valves out of service. The maximum system pressures predicted during the event are shown in Table 2. 1.

Results of the single loop operation (SLO) analysis are shown in Appendix A.

The safety limit analysis for single loop operation supports an increase in the HCPR Safety Limit of .01.

ANF-87-125 Revision 1 TABLE 2.1 TRANSIENT ANALYSIS RESULTS AT DESIGN BASIS CONDITIONS*

h CPR MCPR**

Transient ANF 9x9 GE 8x8R Load Rejection Without Bypass 0.24/1.30 0.21/1.27 with Recirculation Pump Trip Feedwater Controller Failure 0.23/1.29 0.20/1.26 with Bypass Loss of Feedwater Heating 0.16/1.22 0.15/1.21 Maximum Pressure si Transient Vessel Dome Vessel Lower Plenum Steam Line MSIV Closure 1281 1297 1284

• 104% power/100% flow.

• Based on the MCPR Safety Limit of 1.06 confirmed herein.

pW ANF-87-125 Revision I 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1 Desi n Basis Consistent with the FSAR plant transient analysis, thermal margin operating MCPR limits are determined based on the 104% power/100% flow 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 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 3 has been determined to be at the maximum Cycle 3 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 artially inserted. Following requirements established in the Plant Operating

~

License and associated, Techn'ical Specifications, observance of a MCPR operating limit of 1.30 for ANF 9x9 fuel and 1.27 for GE 8x8R fuel or greater

~

conservatively protects against boiling transition during anticipated plant systems transients from design basis conditions for Susquehanna Unit 2 Cycle 3.

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

Recirculation Pump Trip (RPT) flow coastdown was input based on measured Susquehanna Unit 2 startup test data. To confirm the neutronics as requested by the SER issued for the supplements of Reference l(8), the Susquehanna system transient model was benchmarked to appropriate Susquehanna Unit 2 startup test data. XCOBRA-T( ) was used to calculate the change in critical ower ratio (delta CPR) for pressurization'vent analyses.

i

ANF-87-1 Revision 3.2 Antici ated Transients ANF considers eight categories of potential system transient occurrences for Jet Pump BWRs in XN-NF-79-71(1~ ). The three most limiting transients are described here in detail to show the thermal margin for Cycle 3 of Susquehanna Unit 2. These transients are:

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. 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 This 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 recirculating 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 due to 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 as reported in Reference 8 and approved by the NRC. The bounding assumptions include:

ANF-87-125 Revision 1 Technical Specification minimum control rod speed Technical Specification maximum scram delay time Integral power increased by 10%

At design basis conditions (104% power/100% flow) this results in a delta CPR of 0.24 for the load rejection without bypass when RPT is'perable for ANF 9x9 fuels. The corresponding delta CPR for GE 8xSR fuel is 0.21.

The load rejection without bypass event was also analyzed at the design basis conditions when RPT is not operable. The resulting delta CPR's are 0.37 and 0.32 for ANF 9x9 and GE SxSR fuels, respectively.

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 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.

The evaluation of the flow event at design basis conditions was performed with bounding values and resulted in a delta CPR of 0.23 for ANF 9x9 fuels and 0.20 for GE Sx8R fuel. 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 for all 9x9 and Sx8 fuels.

he calculated results show that FWCF delta CPR's vary with decreasing power

8 ANF-87-1 Revision 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. This results in a delta CPR of 0.28 for ANF 9x9 fuels and 0.25 for GE 8x8R fuel.

3.2.3 oss 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 of Reference 1 the delta CPR for the event in Cycle 3 is 0.16 for ANF 9x9 fuel and 0.15 for GE 8x8R fuel. Figures 3.5 and 3.6 depi key variables for the loss of feedwater heating event.

The bypass valves do not significantly affect the loss of feedwater heating results. Thus, 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( ). 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 PTSBWR3 and XCOBRA (Reference 1). Appendix A(1 ) of the Susquehanna Unit 1 Cycle 2 analysi's delineates the changes made to COTRANSA(1) to merge the PTSBWR3 code with the COTRANSA code, to refine numerical techniques and to improve input. Reference 9 describes the XCOBRA-T code us to calculate the delta CPR's for the pressurization transients. Appendix B o

9 ANF-87-125 Revision 1 Reference 10 delineates the plant related changes made to these codes for the Susquehanna Units 1 and 2 analyses.

3.4 Safet Limit The safety limit is the minimum value of the critical power ratio (CPR) 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 which would be permitted to occur during the limiting anticipated operational occurrence. A HCPR safety limit of 1.06 for all fuel types in Susquehanna Unit 2 Cycle 3 was supported by the methodology presented in Reference 3. The input parameters and uncertainties used to support the safety limit are presented in Appendix B of this report.

10 ANF-87-l~

Revision~

TABLE 3.1 REACTOR DESIGN AND PLANT CONDITIONS SUSQUEHANNA UNIT 2 Reactor Thermal Power 3439 Mwt (104%)'otal Core Flow (1005) 100.0 Mlb/hr Core In-Channel Flow 89.9 Mlb/hr Core Bypass Flow 10.1 Mlb/hr Core Inlet Enthalpy 518.0 Btu/ibm Vessel Pressures Steam Dome 1035 psia Upper Plenum 1045 psia Core 1052 psia Lower Plenum 1066 psia Turbine Pressure 975 psia Feedwater/Steam Flow 14.15 Mlb/hr Feedwater Enthalpy 360.8 Btu/ibm Recirculation Pump Flow (per pump) 15.75 Mlb/hr

ANF-87-125 Revision 1 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 2 High Neutron Flux Trip 125.3%

Control Rod Inserti'on Time 3.49 sec/90% inserted Control Rod Worth nominal Void Reactivity Feedback nominal Time to Deenergized Pilot Scram Solenoid Valves 200 msec (maximum)

Time to Sense Fast Turbine Control Valve Closure 30 msec 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 (Total) 70 msec Fuel/Cladding Gap Conductance Core Average (Constant) 758.0 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

12 ANF-87-1 Revision TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 2 (CONTINUED)

MSIV Stroke Time 3.0 sec MSIV Position Trip Setpoint 90% open Turbine Bypass Valve Performance Total Capacity 936.11 ibm/sec Delay to Opening (80% open) 300 msec Fraction of Energy Generated in Fuel 0.965 Vessel Mater Level (above i'nstrument zero)

High Level Trip 58.7 in Normal 35 in*

Low Level Trip 8 in Maximum Feedwater Runout Flow Three Pumps 5049 ibm/sec Recirculation Pump Trip Setpoint Vessel Pressure 1170 psig

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

13 ANF-87-125 .

Revision 1 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN THE ANALYSIS FOR SUSQUEHANNA UNIT 2 (CONTINUED)

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 4 034 ft/100%

Flow Control Mode Manual Pressure Regulator Settings Lead 3.0 sec Lag 7.0 sec Gain 3 '3%/psid

ANF-87-1 Revision TABLE 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES Maximum Maximum h CPR Maximum Core Average System For Neutron Flux Heat Flux Pressure 9x9 Event  % Rated  % Rated ~si a Fuels Load Rejection 267 116.2 1194 0.24 Without Bypass Feedwater Controller 233 116.8 1179 0.23 Failure Loss of Feedwater 123 121.3 1078 0.16 Heating MSIV Closure with 342 133.2 1312 Flux Scram NOTE: All events are bounding case at 104% power/100% flow.

15 ANF-87-125 Revision 1

'ABLE 3.4 FEEDWATER CONTROLLER FAILURE ANALYSIS RESULTS AT 100% FLOW

-% Power Delt'a CPR ANF 9x9 GE 8x8R 104 0.23 0.20 80 0.25 0.23 0.28 0.26 40 0.31 0.28

l,. NEUTRON FLUX LEVEL

2. HEAT FLUX

3. RECIRCULATION FLOW

4. VESSEL STEAM FLOW

5. FEEDWATER FLOW CI Ol Cl

~O

~n LK O 45 123 23 2 Z O 5 LU 3 4 CJ CE LU CL CI I

CI 0 0 0.2 0.5 0.7 1.0 1.2 1.5 1.7 2.0 2.2 2.5 TIME, SEC Figure 3.1 Load ion Without Bypass

i. VESSEL PRESSURE CHANGE (PSI)

2. VESSEL HATER LEVEL (IN)

Cl LO 1

W.O 0.2 0.5 0.7 1.0 1.2 1.5 1.7 2.0 2.2 2.5 TIME, SEC Figure 3.2 Load Rejection Without 8ypass

i. NEUTRON FLUX LEVEL

2. HEAT FLUX

3. RECIRCULATION FLOW

4. VESSEL STEAM FLOW

5. FEEDWATER FLOW 5

5

<2 4 i 4 12 i6 20 24 28 32 36 40 TIME, SEC figure 3.3 Feedw .ontrol 1er Failure

ID

i. VESSEL PRESSURE CHANGE (PSI)

2. VESSEL WATER LEVEL (IN)

Ct CU CI OJ oP 12 16 20 24 28 32 36 40 TIME, SEC Figure 3.4 Feedwater Controller Failure

i. NEUTRON .FLUX LEVEL

2. HEAT FLUX

3. RECIRCULATION FLOW

4. VESSEL STEAM FLOW

5. FEEDWATER FLOW 3 3 12 45 45 12 Cl 3 Cl I- O tO C3 IX UJ 0

2 1

C)

I t0 10 -20 30 40 50 60 70 80 90 100 TIME, SEC Figure 3.5 Loss eedwater Heating

i. VESSEL PRESSURE CHANGE (PSI)

2. VESSEL WATER LEVEL (IN) 2 2 LA ILL I

CI LA I

LA I

IQ 10 20 30 40 50 60 70 80 90 100 TIME, SEC Figure 3.6 Loss Of Feedwater Heating

l KV

-. $1 r ~

0

22 ANF-87-125" Revision 1 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 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 out of service.

4.1 Desi n'asis The reactor conditions used in the evaluation of the maximum pressurization vent 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 advanced plant simulator code COTRANSA( ), 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 Hain Steam Isolation Valves (HSIVs) 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 or system pressure than turbine trips.

23 ANF-87-1P Revision

'.2.

1 Closure Of All Hain Steam Isolation Valves This calculation assumed that six relief valves were out of service and that all four steam isolation valves were isolated at the containment boundary within 3 seconds. At about 3.0 seconds, the reactor scram is initiated by reaching the high flux trip setpoints. Since scram performance was degraded to its Technical Specification limit, effective power shutdown is delayed until after 4.4 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 1284 psig occurring near the vessel at about 6.5 seconds. The maximum vessel pressure was 1297 psig occurring in the lower plenum at about 6.3 seconds.

ANF-87-125 Revision 1 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 2 Cycle 3. These results indicate that HCPR would decrease below the safety limit if the full flow reference MCPR is observed at initial conditions. Thus, an augmented HCPR is needed for partial flow operation to prevent violation of the

~

HCPR Safety Limit for the two pump xcursion 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 3 analysis conservatively assumed the run-up event initiated at 57% power/40% flow and reached 111% 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 HCPR limit for Susquehanna Unit 2 Cycle 3 shall be the maximum of the reduced flow MCPR operating limit, the full flow HCPR operating limit, or the power dependent HCPR operating limit.

9X9 FUELS ANF GE 8X8R FUEL 1.50 1.40 1.30

~L 1.20 uo 40 60 BO 70 80 90 100 TOTAL CORE RECIRCULATION FLOW (% RATED)

Figure 5.1 Susquehanna Unit 2 . 3 Reduced Flow HCPR Operating Limit

26 ANF-87-125 Revision 1 6.0 . REFERENCES R. H.

Ilt R t,"

Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling

~XN-NF.I-II, R I I Corporation*, Richland, WA 99352, November 1981.

2, Ad d N I I

2. J. A. White, "Susquehanna Unit 2 Cycle 3 Reload Analysis, Design and Safety Analyses, " 8NF-87-126, Advanced . Nuclear Fuels Corporation, Richland, WA 99352, October 1987.

3. J. A. White, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Hethodology, Summary Description," XN-NF ~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 R,"~52-525A, R 11 I,Ad dN I F I C 5 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'orporation, Richland, WA 99352, Hay 1986.

, 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. Herckx, "RODEX2 Fuel Rod Mechanical

~NF.I-N,R 99352, April II 1984.

2,Ad dll I

Response

F I C 5 tl,Ill Evaluation Model," XN-ll d.,IIA

8. S.

Il t E.

R t,"

Jensen, "Exxon Nuclear

~XN-II -I Advanced Nuclear Fuels Corporation,

- I, Plant Transient Methodology for Boiling R I I Richland, I,

WA 99352, March 1986.

9. H. J. Ades, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," XN-NF-84-105 P A , Volume 1 & Volume 1, Supp.

1 8 Supp. 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).

I A-I ANF-87-125 Revision I APPENDIX A SINGLE LOOP OPERATION The NSSS supplier has 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 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. 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 hose 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 and analyses provided by ANF.

A discussion of the relevant events and limits .for single loop operation follow. Also included are results of ANF analyses which confirm the NSSS vendor conclusions.

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 61% core flow because of the capability of the recirculation loop pump. Thus, the MCPR limit at maximum power is higher than the two pump operating MCPR limit due to

~

the flow dependent MCPR function. This flow dependence is based on a flow

~

A-2 ANF-87-1~

Revision~

increase transient from runup'f two pumps. Flow runups from a single recirculation pump would be much less severe, though the conservative two pump limit is retained.

Load Re 'ection Without B ass 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 the steam flow through rapid closure qf the turbine control valve. 'topping When the rapid pressurization reaches the core it causes a power excursion due to void collapse.

The reduced power and flow analyses for the Susquehanna Units described in Reference A-1 under two loop operation shows that the resulting pow excursion and associated delta CPR are reduced below those of the ful power/full flow case. Thus for the Susquehanna Units the HCPR limits based on LRWB analyses at full power are conservatively applicable to the lower powers/flows associated with single loop conditions. Furthermore, LRWB analyses by ANF at reduced power and flow conditions in other BWR's with single loop operation confirm this trend.

A.1.2 Feedwater Controller Failure The second worst limiting transient at full power and flow is the Feedwater Failure (FWCF) to maximum demand. This transient is also less 'ontroller 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 is terminated by a reactor scram initiated by a turbine trip. The turbine tr'

A-3 ANF-87-125 Revision I is the result of the high water level trip caused by the additional amount of feedwater being injected.

I 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 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 SLO.

. Thus, because of the slower enthalpy transport phenomena caused by the lower 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.

.1.3 Pum Seizure Accident Pump seizure is a postulated accident where the operating recirculation pump suddenly stops rotating. This causes a rapid decrease in core flow, a decrease in the rate at which heat can be transferred from the fuel rods and a decrease in the critical power ratio. Analyses with COTRANSA and XCOBRA-T show that for Cycle 3 the CPR for ANF fuel would. decrease by 0.30 during a pump seizure for single loop operation.

The COTRANSA code was used to simulate system response to a pump seizure in single loop operation from the conditions specified in Table A. l. The operating recirculation pump rotor was stopped in 0. 1 seconds causing a sudden decrease in active jet pump drive flow. At about 6.7 seconds the inactive jet pump diffuser flow went from negative flow to positive flow. In 7.3 seconds the dome pressure decreased to a minimum value of 970.3 psia and then started to increase again. Figures A. I and A.2 present a graphical representation of important system parameters during the transient.

A-4 ANF-87-1, Revision The delta CPR for this event was calculated using XCOBRA-T. The ANF 9x9 fuel reached a maximum delta CPR of 0.30 at 2.2 seconds into the transient. The GE 8x8R fuel reached a maximum delta CPR of 0.29 at 2. 15 seconds into the transient.

A.1.4 MCPR Safet Limit For single loop operation, the NSSS vendor found that an increase of 0.01 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 safety limit MCPR and found that the NSSS vendor determined increase in the allowed safety limit MCPR is also applicable to ANF fuel during single loop operation. Thus, increasing the safety limit MCPR by 0.01 for single loop operation (1.07) with ANF fuel is sufficient conservative to also bound the increased flow measurement uncertainties fo single loop operation.

A.1.5 ~Summar The limiting MCPR operating limit For single loop operation is conservatively set using the limiting pump seizure accident delta CPR plus the single loop operation MCPR safety limit. This limit together with the MCPRf curve for two loop operation plus .01 and the MCPRp curve for two loop operation plus .Ol conservatively bound all transients.

A-5 ANF-87-125 Revision 1 A.2 MAPLHGR LIMITS ANF performed LOCA analyses for single loop conditions and determined that the MAPLHGR limit curve for two-loop operation is also applicable to single loop operation for ANF fuels (Reference A-2).

A-6 ANF-87-1 Revision A.3 STABILITY The Technical Specifications require APRM/LPRM surveillance to the left of the 45% Constant Flow line and above the 80% Rod Block line. Based on core hydrodynamic stability analyses for Cycle 3, operation at power/flow combinations above and to the left of the line connecting the 66% Power/45%

Flow and 69% Power/47% Flow points needs to be added to the APRM/LPRM surveillance requirements. Figure A.3 shows the core power versus core flow established for Cycle 3.

A-7 ANF-87-]25 Revision 1 TABLE A. 1 SLO REACTOR AND PLANT CONDITIONS Reactor Thermal Power 2489 HMt Total Recirculation Flow 60.35 Hlb/hr Core Bypass Flow 5.70 Hlb/hr Core Inlet Enthalpy 507.3 Btu/lb Vessel Pressures Steam Dome 994.5 psia ,

Lower Plenum 1011.3 psia Turbine Pressure 965.4 psia Steam Flow 9.8 Hlb/hr Feedwater Enthalpy 330.7 Btu/lb

l. NEUTRON FLUX LEVEL

2. HEAT FLUX

3. RECIRCULATION FLOW

4. VESSEL STEAM FLOW

5. FEEOWATER FLOW 5 i io 14 i6 iB 20 TIME, SEC Figure A.l Single L eration - Pump Seizure

CI

1. VESSEL PRESSURE CHANGE (PSI)

2. VESSEL HATER LEVEL (IN)

CU O

IQ 10 12 14 16 18 20 TIME, SEC Figure A.2 Single Loop Operation - Pump Seizure

A-10 ANF-87-125 Revi si'o'n' 0 ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ' g ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ 0 ~ ~ ~ ~Q ~ ~ ~ ~ ' ~

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ega 10)X Xe APRM.'OD BLOCK r ROD LINE 80 Ie ~ op 0 ~ 0 0 ~ ~ ~ $ ~ ~ ~ 0 ~ 0 \~00 ~ 0A $1 0 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ 0)0 ~ ~ ~ ~ ~ ~ ~ $ ~ 0 ~ ~ ~ ~ ~ ~

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80K ROD LINE 40 0 ~ 0 ~ ~ ~ Joe 0 ~ 0 0 J ~ 0 0 ~ 0 0 el 0 0 0 ~ ~ 0 0 el ~ 'L ~ ~ ~ ~ ~ ~ ~ ~ J ~

30 ~ ~ ~ ~~ J 20 ~ >0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ L .~ ~ ~ ~ ~ J~ ~>~ ~ ~ ~ ~ ~ ~ ~ ~

N$ T CER C 2'-,PUMP MIN FLOW:

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0 0 10 20 30 40 60 60 70 80 90 100 CORE FLOW, % RATED Figure A03 Core Power Versus Core Flow

A-ll ANF-87-125 Revision 1 A.4 REFERENCES A-1. J. C. Chandler, "Susquehanna Unit 1 Cycle 3 Reload Analysis," XN-NF-85-132, Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, December 1985.

A-2. R. Swope, "Susquehanna LOCA Analysis for Single Loop Operation,"

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

A-3. K. D. Hartley, et al., "Susquehanna Unit 2 Cycle 2 Stability Test Results," XN-NF-86-90, Supplement 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, January 1987.

0 B-1 ANF-87-125 Revision 1 APPENDIX B MCPR SAFETY LIMIT B. 1 INTRODUCTION The HCPR 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 HCPR 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.

i

B-2 ANF-87-1 Revision 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%.

B-3 ANF-87-125 Revision 1 B.3 DESIGN BASIS POWER DISTRIBUTION Predicted power distributions were extracted from the fuel management analysis for Susquehanna Unit 2 Cycle 3. The radial power distributions were evaluated for performance as the design basis radial power map, and the distribution at 8,000 HWd/HTU exposure was selected as the most severe expected distribution for the cycle. The distribution was skewed toward higher power factors by the addition of bundles with a radial peaking factor approximating an operating HCPR level of 1.32 at full power. The resulting design basis radial power distribution is shown in Figure B.3-1.

The fuel management analysis indicated that the maximum power ANF bundle (XN-

2) in the core at the end-of-cycle exposure (10;829.6 HWd/HTU) was predicted to be operating at an exposure level of 14,319 HWd/HTU, so a local power istribution typical of a nodal exposure of 15,000 HWd/HTU was selected as the

~

design basis local power distribution.

~

Uncontrolled local power peaking distributions for both ANF 9x9 XN-2 9GD4% fuel and 10GD5% fuel were'eviewed.

The limiting locals were found to occur at 15,000 HWd/HTU for 9GD4% fuel.

This distribution is shown in Figure B.3-2. Local power distributions for two three cycle irradiated fuel were chosen conservatively for ANF XN-1 and GE

'nd 8x8R fuel, and are shown in Figures B.3-3 and B.3-4. A boundingly flat local power distribution was selected for the General Electric fuel in the peripheral low power region. This distribution is shown in Figure B-3.5.

The limiting axial power profile selected for the 8,000 HWd/HTU statepoint of Cycle 3 was conservatively selected based on established criteria.

80 70 60 50 00 C)

So 20 10 S

C Ao tA Q) 0 O 0 0.2 0.0 0.6 0.8' 1.2 6 RRD IRL POWER PERKING Figure B.3-1 Susquehanna Unit le 3 Design Basis Radial Power Histogram

8-5 ANF-87-125 Revision 1

• ~ 0

0.88  : 0.91  : 0.96  : 1.04 : 1.02 : 1.04  : 0.96  : 1.00  : 0.96

• 0.91:

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0.93: 0.98: 1.07  : 0.91 1.07  : 0.97  : 1.04  : 1.01

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0.96  : 0.98  : 0.90  : 1.04  : 1.03  : 1.04  : 1.04  : 0.99  : 0.96  :

• ~

• : 1.04  : 1.07  : 1.04  : 1.00  : 0.99  : 1.00 1.05  : 0.94  : 1.04  :

• ~

• : 1.02

~  : 0.91

~ 1.03 : 0.99  : 0.00 : 0.98  : 1.05  : 1.07 : 1.04

• ~

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• .04 : 1.07 1.04 : 1.00  ; 0.98 : 0.00  : 1.03  : 0.94 : 1.05

• ~

0.96  : 0.97  : 1.04 : 1.05  : 1.05 : 1.03  : 1.06  : 1.00 : 0.97 1.00  : 1.04  : 0.99 : 0.94  : 1.07 : 0.94-: 1.00  : 0.94 : 1.01 0.96  : 1.01 0.96 : 1.04  : 1.04 : 1.05  : 0 '7  : 1.01 : 0.97 Figure B.3-2 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-2 9X9 Fuel

B-6 ANF-87-17 Revision '

~

• : 0.91  : 0.92  : 0.95  : 1.01 : 1.01 : 1.01 : 0.96 : 0.98 : 0.95

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• ". 0.92 : 0.94  : 0.98  : 0.97 : 1.05 : 0.95 : 0.99 : 0.95  : 0.98

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• : 0.95  : 0.98  : 0.93  : 1.06 : 1.05 : 1.06 : 1.05 : 0.97 : 0.96

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1.01 : 0.97  : 1.06  : 1.03 : 1.03 : 1.04 : 1.07 : 1.06 : 1.02

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1.01 1.01 1.05 0.95 1.05 1.06 1.03 1,04 0.00 1.01 1.01 0.00 1.07 1.04 1.06 0.96 1.01 1.02 0

• ~

0.96 : 0.99  : 1.05  : 1.07 : 1.07 : 1.04 : 1.06 : 1.00 : 0.96

~ ~

0.98 : 0.95  : 0.97.: 1.06  : 1.06 : 0.96 : 1.00 : 0.95 : 0.98 0.95 : 0.98  : 0.96  : 1.02 : 1.01 : 1.02 : 0.96 : 0.98 : 0.96 Figure B.3-3 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-1 9X9 Fuel

B-'7 ANF-87-125 Revision 1

• ~

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1.03 : 1.00  : 1.00  : 1.00  : 1.00  : 1,00  : 1.01 : 1.03

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1.00 : 0.98  : 1.00  : 1.02  : 1.02 : 1.03  : 1.00 : 1.01

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1.00 : 1.00 1.01  :; 1.01 1.01 : 0.90  : 1.03 : 1.00  :

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1.00 : 1.02  : 1.01  : 0.89  : 0.00 : 1.01  : 1.02 : 1.00  :

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1.00 : 1.02 1.01  : 0.00 0.89 : 1.01  : 0.99 : 1.00  :

• ~

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~

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1.00 : 1 '3  : 0,90  : 1.01 1.01 : 0.98 1.00 : 1.00  :

1.01 : 1.00 1,03  : 1.02 0.99 : 1.00  : 0.98 : 1.00  :

1.03 : 1.01 1.00  : 1.00  : 1.00 : 1.00,: 1.00 1.03 Figure B.3-4 Design Basis Local Pow'er Distribution General Electric (Central) 8XSR Fuel

B-S, ANF.-'87-1 Revision ,

• ~

• : 1.00 : 1.00  : 1.00  : 1.00  : 1.00  : 1.00  : 1.00 : 1.00  :

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• : 1.00 : 1.00  : 1.00  : 1.00  : 1.00  : 1.00  : 1.00 : 1.00  :

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

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1.00 : 1.00  : 1.00  : 0.00  : 1.00 ; 1.00  : 1.00 : 1.00

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

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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.3-5 Design Basis Local Power Distribution General Electric (Peripheral) SXSR Fuel

B-9 ANF-87-125 Revision 1 B.4 CALCULATION OF THE NUMBER OF RODS IN BOILING TRANSITION The methodology of Reference 8-1 was used to analyze the number of fuel rods in boiling transition. The XN-3 correlation(B ) 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 ) were used in lieu of Pear son curve fitting. The uncertainties used in the analysis for normal conditions were those identified in Reference B-l. At least 99.9% of the fuel rods in the core are expected to avoid boiling transition with a confidence level of 95%.

B-10 ANF-87-'1 Revision B.5 REFERENCES B-1. T.

Ilt W.

R t,"

Patten, "Exxon Nuclear Critical Power Methodology

~XN-Npt A, R 11 I, Ad Corporation, Richland, WA 99352, November 1983.

d N for Boiling I F Critical B-2. R.

5 B.

Nuclear I ti,"

Macduff Fuels and X~N-Np-51 T.

A, W:

R Patten, 11 Corporation, Richland, I,

WA "The d

XN-3 Rppl t I, Ad 99352, October 1982.

Power d

B-3. Paul N. Somerville, "Tables for Obtaining Non-Parametric Tolerance Limits," Annals of Mathematical Statistics, Vol. 29, No. 2 (June 1958), pp. 599-601.

ANF-87-125 Revision 1 Issue Date:

SUSQUEHANNA UNIT 2 CYCLE 3 PLANT TRANSIENT ANALYSIS Distribution:

D. A. Adkisson D. J. Braun R. E. Collingham L. J. Federico S. F. Gaines R. G. Grummer K. 0. Hartley H. J. Hibbard S. E. Jensen T. H. Keheley J. N. Morgan L. A. Nielsen D. F. Richey G. L. Ritter C. J. Volmer J. A. White H. E. Williamson H. G. Shaw/PP8L (20)

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