ML18040B150

Forwards Application for Proposed Amend 39 to License NPF-22,revising Tech Specs to Support Cycle 2 Reload.Fee Paid

ML18040B150
Person / Time
Site:	Susquehanna
Issue date:	06/19/1986
From:	Kenyon B PENNSYLVANIA POWER & LIGHT CO.
To:	Adensam E Office of Nuclear Reactor Regulation
Shared Package
ML17146A415	List: ML17146A414 ML17146A416 ML17146A417 ML17146A419 ML17146A421 ML17146A422 ML17146A423 ML18040B150
References
PLA-2661, NUDOCS 8606240301
Download: ML18040B150 (56)

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ACCESS I QN NBR: 8606240301 DOC. DATE: Sb/06/19 NOTAR I ZED: YES DOCKET ¹ FACIL: 50-388 Susquehanna Steam Electric Stations Unit 2. Pennsglva 05000388 AUTH. NAt'tE AUTHOR AFFILIATION KENYON'. D. Pennsylvania Potoer 5 Light Co.

RECIP. NANE RECIPIENT AFFILIATION ADENSAI'1p E. BWR Prospect Directorate 3

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SUBJECT:

Forwards application for proposed Amend 39 to License NPF-22'evising Tech Specs to support Cycle 2 reload. Fee paid.

DISTRIBUTION CODE: ACOID TITLE:

COPIES RECEIVED: LTR Submittal: l'eneral Di stribution 8 ENCL R SIZE:

OR, NOTES: icy NNSS/FCAF/PN. LPDR 2cgs Transcripts. 05000388 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NANE LTTR ENCL ID CODE/NANE LTTR E 'CL BWR EB BWR EICSB 2 BWR FOB BWR PD3 LA BWR PD3 PD 01 S CANPAQNONE BWR PSB BWR RSB INTERNAL: ACRS ELD/HDS4 09 Sly ADN/LFNB NRR/ T/TSCB NRR/ORAS 04 ROND EXTERNAL: EQ8(Q BRUSKE> S LPDR 03 NRC PDR 02 NSIC 05 NOTES: 3 3 0/> 1 TOTAL NUNBER OF COPIES REQUIRED: LTTR 33 ENCL

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Pennsylvania Power 8 Light Company Two North Ninth Street ~ Allentown, PA 18101 ~ 215 i 770.5151 Bruce D. Kenyon Senior Vice President-Nuclear 215/770-41S4 JUN $ 9 1986 Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Project Director BWR Pro)ect Directorate No. 3 Division of BWR Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT 39 TO LICENSE NO. NPF-22 PLA-2661 FILES R41-2, A7-8C Docket No. 50-388

Dear Ms. Adensam:

The purpose of this letter is to propose changes to the Susquehanna SES Unit 2 Technical Specifications in support of the ensuing Cycle 2 reload. Changes to the following Technical Specifications are requested:

Index 1.0 Definitions 3/4.1.2 Reactivity Anomalies 3/4.2.1 Average Planar Linear Heat Generation Rate 3/4.2.2 APRM Setpoints 3/4.2.3 Minimum Critical Power Ratio 3/4.2.4 Linear Heat Generation Rate 3/4.3.4.2 End-of-Cycle Recirculation Pump Trip System Instrumentation 3/4.4.1.1.2 Recirculation Loops Single Loop Operation 3/4.7.8 Main Turbine Bypass System 5.3.1 Fuel Assemblies B 2.1 Safety Limits B 3/4. 1. 1 Shutdown Margin B 3/4. 1. 2 Reactivity Anomalies B 3/4.1.3 Control Rods B 3/4.1.4 Control Rod Program Controls B 3/4.2.1 Average Planar Linear Heat Generation Rate B 3/4.2.2 APRM Setpoints B 3/4.2.3 Minimum Critical Power Ratio ~e0 B 3/4.2.4 Linear Heat Generation Rate B 3/4.4.1 Recirculat:ion System B 3/4.7..8 Main Turbine Bypass System

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Page 2 SSES PLA-2661 Files R41-2, A7-8C Ms. E. Adensam As discussed in a 'telecon hei'd with your staff on June 16, 1986, and in the attached'reload summary report, this submittal does not contain Minimum Critical Power Ratio (MCPR) Technical Specification Limits. The methodology which will be used to derive these limits is being provided at this time for your review; the actual values will be supplied in mid-July.

The following attachments to this letter are provided to illustrate and technically support each of the changes:

Marked-up Technical Specification Changes No Significant Hazards Considerations Susquehanna SES Unit 2 Cycle 2 Reload Summary Report XN-NF-86-60, "Susquehanna Unit 2 Cycle 2 Reload Analysis," May, 1986 XN-NF-86-55, "Susquehanna Unit 2 Cycle 2 Plant Transient Analysis,"

May, 1986 XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for 9X9 fuel," May, 1986 Susquehanna SES Unit 2 Cycle 2 Proposed Startup Physics Tests Summary Description, May, 1986 Please note that with respect to thermal hydraulic stability of the Exxon Nuclear Company 9X9 fuel being inserted during this reload, PP&L has already submitted (PLA-2637, dated April 30, 1986) a stability test program which will supplement the results presented in the pertinent analyses attached. Certain other supplementary data will also be submitted to the NRC at their request in accordance with our discussions on May 30, 1986. It is not our intent to treat any of this supplementary information as revisions to this proposal.

Also, sufficient analysis has not been completed to support Single Loop Operation (SLO) with the 9X9 fuel design. The Technical Specifications have been altered accordingly, and we will provide a separate submittal on this issue based on appropriate analysis when it is available. Again, this submittal is not to be considered a revision to this proposed amendment.

Susquehanna SES Unit 2 is currently scheduled to be shutdown for refueling and inspection on August 2, 1986 and to restart as early as October 3, 1986.

We request that your approval be conditioned to become effective upon startup after this outage, and will keep you informed of any schedule changes.

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Page 3 SSES PLA-2661 Files R41-2, A7-SC Ms. E. Adensam Any questions with respect to this proposed amendment should be directed to Mr. R. Sgarro at (215) 770-7855. Pursuant to 10CFR170, the appropriate fee is enclosed.

Very truly yours, B. D. Keny Senior Vice President-Nuclear Attachments cc: M. J. Campagnone USNRC R. H. Jacobs USNRC T. M. Gerusky Bureau of Radiation Protection Pennsylvania Department of Environmental Resources P.O. Box 2063 Harrisburg, PA 17120

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XN-NF-86-55 Issue Date: 5/]5/86 SUSQUEHANNA UNIT 2 CYCLE 2 PLANT TRANSIENT ANALYSIS Prepared by:

T.H. Keheley, Team Leader BWR Safety Analysis Concur:

R.ED ollingh , Manager BWR Safety A lysis Concur:

J.N. Horgan, H ager Customer Services Engineering Concur:

G.N. Ward, Manager Reload Licensing Approve:

H.E. Williamson, Manager Licensing & Safety Engineering Approve:

G.L. Ritter, Manager fuel Engineering & Technical Services thk/ml n EQON NUCLEAR COMPANY, INC.

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NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub.

mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nudear.fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, informaaon, and begef. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstradon of compliance with the USNRC's reguladons.

Without derogating from the foregoing, neither Exxon Nuclear nor any person acting nn its behalf:

A. Makes any warranty, express 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 B. Assumes any liabilities with respect to the use of, or for dan'ages resulting from the use of, any information, ap.

paratus, method, or process disclosed in this document.

XN. NF- FOO, 7BB

XN-NF-86-55 TABLE OF CONTENTS Section Pacae

1.0 INTRODUCTION

.......................................... 1 2.0 S UHMARY' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN.................. 4 3.1 Design Basis t ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 3.2 Anticipated Transients ................................ 5 3.2.1 Load Rejection Without Bypass ........ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 3.2.2 Feedwater Controller Failure .......................... 6 3.2.3 Loss of Feedwater Heating ............................. 7 3.3 Calculational Model.................................... 8 3.4 S afety Limit .......................................... 8 4.0 ANALYSES FOR INCREASED CORE FLOW (ICF) AND FINAL FEEDWATER TEMPERATURE REDUCTION (FFTR)........... 19 5.0 MAXIMUM OVERPRESSURIZATION ............................ 22 5.1 esign Basis .....................................

0 D 22 5.2 Pressurization Transients ............................. 22 5.3 Closure of All Hain Steam Isolation Valves............. 23 6.0 RECIRCULATION PUMP RUN-UP.."............................ 24

7.0 REFERENCES

............................................ 26 APPENDIX A ...:............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

XN-NF-86-55 List of Tables Table Pacae 2.1 Transient Analysis Results at Design B asis Conditions........................................ 3 3.1 Reactor Design and Plant Conditions Susquehanna Unit 2 ................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ 0 9 3.2 Significant Parameter Values Used in Susquehanna Unit 2 Analysis .......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 10 3.3 Results of Plant Transient Analysis .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 13 3.4 FWCF Results at 100% Flow ............ 14 4.1 Results of System Plant Transient Analysis at ICF and FFTR................................ 20 4.2 Feedwater Controller Failure Delta CPR Results of ICF and FFTR Analyses.............. 21

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XN-NF-86-55 List of Fi ures Ficiure Pa(ac 3.1 Load Rejection Without Bypass .......................... 15 3.2 Load Rejection Without Bypass . . . . ....... 16 3.3 Feedwater Controller Failure .. ... ....... ... ... .... 17 3.4 Feedwater Controller Failure ........................... 18 6.1 Reduced Flow HCPR Operating Limit.............. ....... 25 A-3.1 Design Basis Radial Power Histogram..................... A-4 A-3.2 Design Basis Local Power Distribution (ENC XN-1 9x9 Fuel) ............................... ... A-5 A-3.3 Design Basis Local Power Distribution (GE Sx8 Fuel)

XN-NF-86-55

1.0 INTRODUCTION

This report presents the results of Exxon Nuclear Company's (ENC's) evaluation of system transient events for Susquehanna Unit,2 during Cycle 2 operation with a reload of ENC 9x9 BWR fuel. This evaluation together with an evaluation of 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. Resulting Thermal margin analyses are also presented for operation in the Increased Core Flow (ICF) region of the power/flow operating map and for operation with a Final Feedwater Temperature Reduction (FFTR). Analyses are also reported for operation with the Recirculation Pump Trip (RPT) out of service and with the turbine bypass capability inoperable. An evaluation is also made to demonstrate the vessel integrity for the most limiting pressurization event. The bases for these analyses have been provided in Reference l.

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XN-NF-86-55 2.0

SUMMARY

Using ENC methodology and considering Cycle 2 fuels, the most limiting plant system transient with regard to thermal margin at rated power and flow conditions was determined to be the generator Load Rejection Without Bypass (LRWB). The Minimum Critical Power Ratio (HCPR) limits for potentially limiting 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. 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 Cycle 2 HCPR operating limit 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 during Cycle 2 the safety valves of Susquehanna Unit 2 have sufficient capacity and performance to prevent the pressure from reaching the established transient pressure safety limit of 110% of design pressure (1. 1 x 1250 = 1375 psig). The 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 for RPT out of'ervice are reported in Section 3.2. 1, and results for operation at ICF and FFTR are reported in Section 4.

XN-NF-86-55 Table 2. 1 Transient Analysis Results at Design Basis Conditions*

Transient CPR MCPR ENC 9x9 GE Bx8 Load Rejection Without Bypass 0.17/1.23 0.16/1.22 Feedwater Controller Failure 0.15/1.21 0.14/1.20 Loss of Feedwater Heating NA /1.14 NA /1.14 Maximum Pressure si Transient Vessel Dome Vessel Lower Plenum Steam Line MSIV Closure 1301 1315 1305 104% power/100% flow.

• • Based on a safety limit MCPR of 1.06.

XN-NF-86-55 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 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 2 has been determined to be at end of full power capability when control rods are fully withdrawn from the core. The thermal margin limit established for end of full power conditions is conservative for cases where control rods are partially inserted. Follow-ing requirements established in the Plant .Operating License and associated Technical Specifications, observance of a MCPR limit of 1.23 for 9x9 fuel and 1.22 for 8x8 fuel or greater conservatively protects against boiling transi-tion during anticipated plant systems transients from design basis conditions for Susquehanna Unit 2 Cycle 2.

The calculational models used to determine thermal margin include ENC's plant transient and core thermal-hydraulic codes as described in previous documentation (1,4-7)

' Fuel pellet-to-clad gap conductances used in the analyses are based on calculations with RODEX2 ). Table 3.2 summarizes the values used for important parameters to provide a bounding analysis.

Recirculation Pump Trip (RPT) coastdown was input based on measured Susquehanna Unit 2 startup test data. To confirm the neutronics as 'required by the SER issued for the supplements of Reference I the Susquehanna system transient model was benchmarked to appropriate Susquehanna Unit 2 startup test data. All transients were analyzed on a bounding basis using the COTRANSA hot channel delta CPR model as described in Reference 9.

XN-NF-86-55 3.2 ntici ated Transients ENC considers eight categories of potential system transient occurrences for Jet Pump BWRs in XN-NF-79-71((g)'. The loss of feedwater heating transient has been analyzed on a generic basis as reported in Reference 10.

Results shown for this transient are from the ENC generic analysis.

The two most limiting transients are described here in .detail to show the thermal margin for Cycle 2 of Susquehanna Unit 2. These transients are:

Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

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 Rejection Without Bypass 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 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 ENC's COTRANSA code uncertainties analysis methodology as reported in XN-NF-79-71(P) Rev. 2, Supplements 1-3 and approved by NRC. The bounding assumptions include:

XN-NF-86-55 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. 17 for the load rejection without bypass when RPT is operable for ENC 9x9 fuel. The corresponding delta CPR for GE 8x8 fuel is 0. 16.

The load rejection was then analyzed assuming the same bounding conditions but

.with both RPT and. bypass inoperable. This resulted in delta CPRs of 0.31 for both ENC 9x9 and GE 8x8 fuel.

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. 15 for ENC 9x9 fuel and

0. 14 for GE 8x8 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

XN-NF-86-55 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 CPRs were calculated at the 40% power/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 resulted 'in a delta CPR of 0.18 for ENC 9x9 fuel and 0.17 for GE 8x8 fuel.

3.2.3 Loss of Feedwater Heating 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.

ENC has analyzed the loss of feedwater heating event on a generic basis as described in Reference 10. Based on the generic analysis and the Cycle 2 safety limit of 1.06, the MCPR limit. for Susquehanna Unit 2 Cycle 2 will be

1. 14 for both the ENC 9x9 fuel and the GE 8x8 fuel for the loss of feedwater heating event.

The bypass valves do'ot significantly affect the loss of feedwater .heating results. Thus, this MCPR 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 ENC'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

XN-NF-86-55 was evaluated generically because rapid pressurization and void collapse do not occur in this event. Appendix A of the Susquehanna Unit 1 Cycle 2 analysis delineates the changes made to COTRANSA to merge the PTSBWR3 code with the COTRANSA code, to refine numerical techniques and to improve input.

Appendix A of Referen'ce 9 describes the refinement made to the hot channel model to calculate the delta CPR's during the transient. Appendix B of Reference 3 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 occurrenc'e. The safety limit for all fuel types in 'usquehanna Unit 2 Cycle 2 was determined by the methodology presented in Reference 4 to have a value of 1.06. The input parameters and uncertainties used to establish the safety limit are presented in Appendix A of this report.

XN-NF-86-55 Table 3.1 Reactor Design and Plant Co'nditions Susquehanna Unit 2 Reactor Thermal Power (104%) 3439 HWt Total Core Flow (100%) 100.0 Hlb/hr Core In-Channel Flow 89.7 Mlb/hr Core Bypass Flow 10.3 Hlb/hr Core Inlet Enthalpy 518.0 Btu/ibm Vessel Pressures Steam Dome 1031 psia Upper Plenum 1049 psia Core 1058 psia Lower Plenum 1067 psia Turbine Pressure 974.7 psia Feedwater/Steam Flow 14.15 Hlb/hr Feedwater Enthalpy 360.8 Btu/ibm Recirculation Pump Flow (per pump) 15.7 Hlb/hr

10 XN-NF-86-55 Table 3.2 Significant Parameter Values Used in Analysis Susquehanna Unit 2 High Neutron Flux Trip 125.3%

Control Rod Insertion Time 3.5 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 Notion 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) 443.8 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

XN-NF-86-55 Table 3.2 Significant Parameter Values Used in Analysis (Cont.)

Susquehanna Unit 2 HSIV Stroke Time 3.0 sec HSIV 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 Water Level (above Separator Skirt)

High Level Trip 58.7 in Normal 36.5 in Low Level .Trip 8 in Haximum Feedwater Runout Flow Three Pumps 4118 ibm/sec Recirculation Pump Trip Setpoint 1170 psig Vessel Pressure

12 XN-NF-86-55 Table 3.2 Significant Parameter Values Used in Analysis (Cont.)

Susquehanna Unit 2 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 48 in Steam Flow Equiv. 100%

Flow Control Mode Manual Pressure Regulator Settings 3.0 sec

'ead Lag 7.0 sec Gain 3.33%/psid

13 XN-NF-86-55 Table 3.3 Results of System Plant Transient Analyses Maximum Maximum Maximum Core Average System Neutron Flux Heat Flux Pressure Event  % Rated  % Rated ~sia 6 CPR Load Rejection 274 114.3 1213 .17 Without 8ypass Feedwater Controller 245 114.7 1180 .15 Failure MSIV Closure with 368 130.7 1330 Flux Scram Note: All events are bounding case at 104% power/100% flow.

14 XN-NF-86-55 Table 3.4 Feedwater Controller Failure Analysis Results at 100% Flow

% Power Delta CPR CE ENC Sx8 9x9 100 .14 .15.

80 .22 .24 65 .23 .25 40 .26 .29

30 2 HEA FLUX

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'8.0 0.2 0.5 0' 1 ' 1 ' 1 ' 1 ' 2.0 2.2 2' TINE. SEC Figure 3. 1 Load Rejection Without Bypass

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l' 19 XN-NF-86-55 4.0 ANALYSES WITH INCREASED CORE FLOW ICF AND FINAL FEEDWATER TEMPERATURE REDUCTION FFTR As part of the Susquehanna Unit 2 licensing analysis, ENC evaluated transients for operation in the Increased Core Flow (ICF) operating region up to 108% of rated flow. Transient analyses were also performed for a feedwater temperature reduction of up to 65 degrees F at both nominal flow and increased core flow conditions at the end of the operating cycle. This 65 degree F temperature reduction was conservatively held constant at all power levels evaluated. A summary of the transient analyses is shown in Table

4. 1. Comparison of the results in Table 2. 1 and 4. 1 indicate that ICF had no significant effect on the LRWB delta CPR results and FFTR condition slightly reduced the impact of this document. The corresponding maximum overpressurization event is discussed in Section 5.0 and the pump run-up analysis is reported in Section 6.0.

The effects of the final feedwater temperature reduction were evaluated by analyzing the FWCF transient over the allowed power range for both nominal feedwater temperature and a 65 degree F final feedwater temperature reduction.

Calculations were performed for both the 100% core flow and for the 108% core flow conditions. The results of these calculations are shown in Table 4.2.

The calculated FWCF transient delta CPR generally increases with decreasing power at both flow conditions, and an increased MCPR limit is indicated for low power operating conditions. Thus, for increased core flow operation, increased MCPR limits are indicated. A further, but small, delta CPR increase is generally indicated to operate with reduced feedwater temperature for both rated core flow and increased core flow.

20 XN-NF-86-55 Table 4. 1 Results of System Plant Transient Analysis at ICF and at FFTR Load Re'ection Without B ass Maximum Minimum Maximum Neutronic Core System Flux Average Pressure Delta

(% rated) (% rated) (psia) CPR 104/100 (FFTR) 253 112.9 1191 0.15 104/108 241 112.1 1210 0.17 104/108 (FFTR) 222 110.8 1187 0.15 ASME Over ressure MSIV Closure si Vessel Dome Vessel Lower Plenum Steam Line 104/100 (FFTR) 1264 1279 1265 104/108 '1290 1307 1296 104/108 (FFTR) 1257 1274 1259

21 XN-NF-86-55 Table 4.2 Feedwater Controller Failure Delta CPR Results of ICF and FFTR Analyses Nominal Feedwater Tem . FFTR

~/' E , ENC 9x9 GE 8XS ENC 9X9 100 / 100 0.14 0.15 0.16 0.17 80 / 100 0.22 0.24 0.20 0.22 65 / 100 0.23 0.25 0.24 0.26 40 / 100 0.26 0.29 0.26 0.29 100 / 108 0.15 0.16 0.16 0.17 80 / 108 0.20 0.22 0.20 0.22 65 / 108 0.23 0.25 0.24 0.26 40 / 108 0.27 0.30 0.26 0.30

• 65'F reduction in Feedwater Temperature.

l I

22 XN-NF-86-55 5.0 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. The maximum system pressures predicted during the event are shown in Table 2. 1.

This analysis also assumed six safety relief valves out of service.

5.1 Desi n Basis 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 ENC's advanced plant simulator code COTRANSA , which includes an axial one-dimensional neutronics model.

5.2 Pressurization Transients ENC has evaluated several pressurization events and has determined that closure of all Hain Steam Isolation Valves (MSIVs) without direct scram is the most limiting. Though 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 causes the severity of these faster closures to be less. 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 isolation had occurred. Calculations have determined that the overall result is to cause isolation (MSIV closures) to be more limiting for system pressure than turbine isolations.

23 XN-NF-86-55 5.3 Closure of All Main 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 5.5 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 7. 1 seconds. Substantial thermal power production enhances pressurization. Pressures reach the recirculation pump trip setpoint (1170 psig) before the pressurization has been 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 1305 psig occurring near the vessel at about 10. 1 seconds. The maximum vessel pressure was 1315 psig occurring in the lower plenum at about 10.0 seconds.

The analysis was repeated for ICF and FFTR conditions and the results are summarized in Table 4. 1. Compaison of the results in Table 2.1 and Table 4. 1 show that the design basis conditions are more limiting than ICF or FFTR conditions. At about 5.5 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 6.5 seconds. Substantial thermal power production enhances pressuriza-tion. Pressures reach the recirculation pump trip setpoint (1170 psig) before the pressurization has been 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 1296 psig occurring near the vessel at about 10.2 seconds. The maximum vessel pressure was 1307 psig occurring in the lower plenum at about 9.8 seconds.

XN-NF-86-55 6.0 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 HCPR 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. Based on the results obtained from previous analyses which showed two pump excursions were the limiting pump run-up event, only two pump excursions are evaluated for Susquehanna Unit 2 Cycle 2. These results indicate that MCPR would decrease below the safety limit if the full flow reference MCPR was observed at initial conditions. Thus, an augmented HCPR is needed for partial flow operation to protect the two pump excursion event.

The evaluation of the two recirculation pump flow excursion for Susquehanna Unit 2 showed that establishment of HCPR limits for this event which prevents boiling transition will also bound single pump runups. The analysis of the two pump flow excursion indicates that the limiting event scenario is a gradual quasi-steady run-up due to the inlet enthalpy lag associated with a more rapid run-up.

The Susquehanna Unit 2 Cycle 2 analysis conservatively assumed the run-up event initiated at 57% power/40% flow and reached 111% rated power at 110%

rated flow. 110% flow is consistent with increased core flow analysis; This power to flow relationship bounds that calculated by XTGBWR for the constant Xenon assumption.

The results of the two pump run-up analyses for manual, flow control are presented in Figure 6. 1. The cycle specific HCPR limit for Susquehanna Unit 2 Cycle 2 shall be the maximum of the reduced flow MCPR operating limit and the full flow HCPR operating limit.

1.4 R

1.3 K

Cl 1.2 1.1 1.0 4

OC Total Core Recirculating Flow (I Rated) I Tl I

Figure 6.1 Reduced Flow MdPR Operating Limit CX) l, Ch I

CJl CJl

26 XN-NF-86-55

7.0 REFERENCES

W R.H.

R t," X~,

Kelley, "Exxon Nuclear Plant Transient Methodology Nuclear Co., Inc., Richland, WA R 1*1 2 (

99352, November 1981.

ppi for Boiling d), E

2. T.H. Keheley, "Susquehanna Unit 2 Cycle 2 Reload Analysis, Design and Safety Analyses," XN-NF-86-60, Exxon Nuclear Co., Inc., Richland, WA 99352, April 1986.

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

XN-NF-84-118 including Supplement 1, Exxon Nuclear Company, Richland, WA 99352, December 1984.

T.L.

P I i,"~,EE,R Krysinski and J.C.

Boiling Water Reactors; Inc., Richland, WA 99352, Chandler, THERMEX April 1981.

"Exxon Nuclear Methodology for 11 (,EN Thermal Limits Methodology; I

Summary C.,

5. T.W.

W Richland, R," X~, 11, Patten, WA "Exxon Nuclear 99352, November 1979.

Critical Power Methodology E N I for Boiling 2

6. R.H. Kelley, "Dresden Unit 3 Cycle 8 Plant Transient Analysis 2 ," (' - - , I I I, E N I ., I ., Ri hl d, IIA 99352, December 1981.

7. R.H.

2,1,"~X---,(21.,1.,(tihi Kelley and N.F. Fausz, "Plant Transient Analysis for Dresden d,llA

-, 1, 99352, October 1982.

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

~X- R I E II I ., I ., IW hl d, IIA 99352, March 1984.

9.

10. R.G.

WA Grummer, "A Generic 99352, February 1986.

Loss of Feedwater d I,,

T.H. Keheley, "Susquehanna Unit 1 Cycle 3 Plant Transient Analysis,"

XN-NF-85-130, Exxon Nuclear Company, Richland, WA 99352, November 1985.

W<<,"~X- ->>, Heating Transient I hi For d,

E A-I XN-NF-86-55 APPENDIX A HCPR SAFETY LIHIT A.l INTRODUCTION The HCPR fuel cladding integrity safety limit was calculated using the methodology and uncertainties described in Reference A. l. 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

A-2 XN-NF-86-55 A.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%.

A-3 XN-NF-86-55

'.3 DESIGN BASIS POWER DISTRIBUTION Predicted power distributions were extracted from the fuel management analysis for Susquehanna Unit 2 Cycle 2. These radial power distributions were evaluated for performance as the design basis radial power map, and the distribution at 10,500 MWD/HT cycle 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 1.26 at full power.

The resulting design basis radial power distribution is shown in Figure A.3-1.

The fuel management analysis indicated that the maximum power ENC bundle in the core at this statepoint was predicted to be operating at an exposure level of 12,600 HWD/HT, so a local power distribution typical of a nodal exposure of l5,000 MWD/MT'as selected as the design basis local power distribution. This distribution is shown in Figure A.3-2.

A boundingly flat local power distribution was selected for the co-resident G.E. Fuel. This distribution is shown in Figure A.3-3.

Because the predicted power distributions during the cycle were not all characterized by bottom peaked axial distributions, representative safety limit evaluations were performed at several representative cycle burnup statepoints throughout the cycle, including all points at which the power was skewed toward the upper half of the core. These analyses confirmed the that most severe power distribution conditions were those which are predicted to exist at the end of Cycle 2. The 1.06 safety limit was confirmed at all the points evaluated.

90 80 70 V) 60 SO C) 40 30 20 10 0.2 0.4 0.6 0.8 1.2 RRDIRL PERKING F'RCTOR Figure A.3-l Design Basis Radial Power Histogram

A-5 XN-NF-86-55

0.97  : 1.01  : 0.97 : 1.04 : 1.04 : 1.05 : 0.97 : 1.02 : 0.97 :

1.01  : 0.94  : 0.97 : 1.07 : 1.06 : 0.95 : 1.00 : 0.95 : 1.02 0.97  : 0.99  : 1.04 : 1.05 : 1.05 : 1.02 : 1.06 : 1.00 : 0.97 1,04  : 0.93  : 1.05 : 1.01 : 0.97 : 0.00 : 1.02 : 0 '5 : 1.05 :

1.03  : 1.05  : 1.03 : 1.00 : 0:00 : 0.97 : 1.05 : 1.06 : 1.04 :

1.04  : 0.94  : 1.04 : 1.00 : 1.00  : 1.01 : 1.05 : 1.07 : 1.04  :

0 '7  : 0.98  : 0.90  : 1.04  : 1.03  : 1.05  : 1.04  : 0.97  : 0.97  :

0.91  : 0.94  : 0.98  : 0.94  : 1.05  : 0.93  : 0.99  : 0.94  : 1.01

0.88  : 0.91  : 0.97  : 1.04  : 1.03  : 1.04 : 0 '7  : 1.01  : 0.97  :

Fiaure A.3-2 DESIGN BASIS LOCAL POWER DISTRIBUTION ENC XN-1 9X9 FUEL

• Rod adjacent to control blade corner location

A-6 XN-Nf-86-55 1.03  : 1.00  : 0.99  : 0,99  : 0.99  : 0.99 : 1.00  : 1.03 1.00  : 0.99  : 1.03  : 1.02  : 0.99  : 0.99 : 0,97  : 1.00 0.99  : 1.03  : 0.91  : 1 '2  : 1,01  : 0.98 : 0.99  : 0.99 0.99 1.03 1.02 0.00  : 1.02  : 1.01 : 0.99  : 0.99 0.99  : 1.02  : 1.01  : 0.91  : 0.00  : 1.02  : 1.02  : 0.99 0

0.99  : 0.99  : 1.02  : 1.01  : 1.02  : 0.91  : 1.03  : 0.99 0 ~ ~ ~

4 1.00  : 0.97 0.99  : 1.02  : 1.03  : 1.03  : 0.99  : 1.00 0

1.03  : 1.00  : 0.99  : 0.99  : 0.99  : 0.99  : 1.00  : 1.03

~

• Figure A.3-3 DESIGN BASIS LOCAL POWER DISTRIBUTION G.E. 8X8 FUEL

• Rod adjacent to control blade corner location

A-7 XN-NF-86-55 A.4 CALCULATION OF THE NUHBER OF RODS IN BOILING TRANSITION The SAFTLIH computer code was used to analyze the number of fuel rods in boiling transition. The XN-3 correlation was used to predict critical heat flux phenomena. Five hundred Honte Carlo trials were performed to support the HCPR safety limit. Non-parametric tolerance limits were used in lieu of Pearson curve fitting. The uncertainties used in the analysis for normal conditions were those identified in Reference A-1. At least 99.9% of the fuel rods in the core were expected to avoid boiling transition with a confidence level of 95%.'

A-8 XN-NF-86-55 A.5 REFERENCES A-l. "Exxon R

Richland, Nuclear WA R 11 Critical 1,

Power

~X-(November 1983).

, Methodology X

for Boiling N 1 C Water P

A-2. "TR XN- 11 1 1 C Exxon Nuclear Company, Richland, WA (March 1981).

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

XN-NF-86-55 Issue Date: 5/15/86 SUS(UEHANNA UNIT 2 CYCLE 2 PLANT TRANSIENT ANALYSIS Distribution D.J. Braun J.C. Chandler R.E. Collingham S.F. Gaines R.G. Grummer K.D. Hartley S.E. Jensen T.H. Keheley J.E. Krajicek T.L. Krysinski J.N. Morgan L.A. Nielson T.W. Patten G.L. Ritter H.G. Shaw/PP&L (40)

D.R. Swope G.N. Ward H.E. Williamson Document Control (5)

I 1