Text

Forwards Application for Proposed Amend 132 to License NPF-14,revising Tech Spec to Support Cycle 6 Reload
	ML18026A235
Person / Time
Site:	Susquehanna
Issue date:	07/02/1990
From:	Keiser H; PENNSYLVANIA POWER & LIGHT CO.
To:	Butler W; Office of Nuclear Reactor Regulation
Shared Package
ML17157A243	List: ML17157A242; ML17157A244; ML17157A245; ML18026A235;
References
	IEB-90-002, IEB-90-2, PLA-3407, NUDOCS 9007120229
	Download: ML18026A235 (127)
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ACCESSION NBR:9007120229 DOC.DATE: 90/07/02 NOTARIZED: YES DOCKET FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva, 05000387 AUTH. NAME AUTHOR AFFILIATION KEISER,H.W. Pennsylvania Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION BULTER,W.R. Project Directorate I-2 Forwards application for Proposed Amend 132 to License j

SUBJECT:

NPF-14,revising TS to support Cycle 6 reload.

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Pennsylvania Power & Light Company Two North Ninth Street ~ Allentown, PA 18101-1179 ~ 215i770-5151 Harold W. Keiser Senior Vice president.Nuclear 215I770 4194 July 2, 1990 Director of Nuclear Reactor Regulation Attention: Dr. M. R. Butler, Project Director Project Directorate I-2 Division of Reactor Projects U.S. Nuclear Regulatory Commission

'Washington, D.C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT 132 TO LICENSE NO. NPF-14: UNIT 1 CYCLE 6 RELOAD PLA-3407 FILES A7-BC A17-2 R41-2 Docket No. 50-387

Dear Dr. Butler:

The purpose of this letter is to propose changes to the Susquehanna SES Unit 1 Technical Specifications in support of the ensuing Cycle 6 reload. Changes to the following Technical Specifications and bases are requested:

Index 3/4.2.1 Average Planar Linear Heat Generation Rate 3/4.2.3 Minimum Critical Power Ratio 3/4.2.4 Linear Heat Generation Rate 3/4.4.1 Recirculation System B 2.1 Safety Limits B 3/4.2. 1 Average Planar Linear Heat Generation Rate B 3/4.2.3 Minimum Critical Power Ratio B 3/4.4.1 Recirculation System 5.3.1 Fuel Assemblies 5.3.2 Control Rod Assemblies The following attachments to this letter are provided to illustrate and technically support each of the changes:

July 2, 1990 SSES PLA-3407 F ILES A7-8C/A17-2/R41-2 Marked-up Technical Specification Changes No Significant Hazards Considerations PL-NF-90-003, "Susquehanna SES Unit 1 Cycle 6 Reload Summary Report," June 1990 Susquehanna SES Unit 1 Cycle 6 Proposed Startup Physics Tests Summary Description, June 1990 ANF-90-049, " Susquehanna Unit 1 Cycle 6 Plant Transient Analysis," Advanced Nuclear Fuels Corporation, May 1990 ANF-90-050, "Susquehanna Unit 1 Cycle 6 Reload Analysis Design and Safety Analyses, "Advanced Nuclear Fuels Corporation," May 1990 ANF-90-018(P), Revision 1, "Susquehanna Unit 1 8 x 8 Extended Burnup Design Report," Advanced Nuclear Fuels Corporation, June 1990*

This report is proprietary and is being submitted under separate cover pursuant to 10 CFR 2.790.

In addition to the normal analyses and considerations for Unit 1 Cycle 6, the following specific analyses and issues deserve special mentioning:

Approximately 50 original equipment control blades will be replaced in Unit 1 Cycle 6 with equivalent worth Duralife 160C control blades. A detailed evaluation of the control blade replacements is provided in the attached Unit 1 Cycle 6 Reload Summary Report.

To support single loop operation for Unit 1 Cycle 6 and future operating cycles, ANF performed the recirculation pump seizure accident from single loop operating conditions on a generic basis for the Susquehanna Units. This approach should eliminate the need to have the single loop pump seizure accident analyzed for each reload. Further discussion is provided in the attached Susquehanna specific reload reports and the Unit 1 Cycle 6 Reload Summary Report prepared by NFE.

As a result of the issuance of NRC Bulletin No. 90-02, entitled "Loss of Thermal Margin Caused by Channel Box Bow," licensees who use channels, including those who use them for only a single bundle lifetime, have been required to take into account the effects of channel bow in analyses supporting reload applications.

ANF has addressed this issue for Unit 1 Cycle 6 through a generic analysis and in a separate letter to the NRC. Additional detail is provided in the attached Unit 1 Cycle 6 Reload Summary Report.

For Unit 1 Cycle 6, ANF performed analyses to extend the burnup limit of the remaining Bx8 fuel up to an exposure of 37,000 MWO/MTU. Details are provided in the attached Susquehanna SES specific reload reports, the Susquehanna Unit 1 8x8 Extended Burnup Design Report, and the attached Unit 1 Cycle 6 Reload Summary Report prepared by NFE.

July 2, 1990 SSES PLA-3407 FILES A7-8C/A17-2/R41-2 Susquehanna SES Unit 1 is currently scheduled to be shutdown for refueling and inspection on September 8, 1990 and to restart as early as November 9, 1990 request that your approval be conditioned to become effective upon startup 'e after this outage, and we will keep you informed of any schedule changes.

Any questions regarding this proposed amendment should be directed to Hr. R. Sgarro at (215) 770-7916.

Very truly yours, n-H. W. eiser Attachments cc: NRC Oocument Control Oesk (original)

NRC Region I Hr. H. C. Thadani, NRC Project Manager-Rockville Mr. G. S. Barber, NRC Sr. Resident Inspector-SSES Hr. T. H. Gerusky, Pennsylvania OER

ANF-90-050

~

ADVANCEDNUCLEAR FUELS ."CRPCF.-i; i'-;

SUSQUEHANNA UNIT 1 CYCLE 6 RELOAD ANALYSIS DESIGN AND SAFETY ANALYSES MAY 1990 I

..9007120229 il.ltll y.gQ <

Lli

Errata to ANF-90<50 Susquehanna Unit 1 Cycle 6 Reload Analysis Design and Safety Analyses Report Include PAFF line reference for 9x9 fuel design in Sections 2.0 and 7.2.3.

Update References 9.11 and 9.13 to reflect current revisions.

ADVANCEDNUCLEARFUELS CORPORATION ANF-90-050 Issue Date: 05/30/90 SUSQUEHANNA UNIT 1 CYCLE 6 RELOAD ANALYSIS Design and Safety Analyses Prepared by:

BWR P. N. 'y Fuel Engineering Fuel Engineering and Licensing T. L. otz BWR Fuel Engineering Fuel Engineering and Licensing Hay 1990

CUSTOMER OISCLAIMER IMPORTANT NOT1CE REGARDING CONTENTS AND USE OF TMIS OOCU MENT PLEASE REAO CAREFULLY Advanced Nuclear Fuels Corporabon's warranties and representations con.

cerning the subject matter of this document are those set forth in me Agreement between Advanced Nuclear Fuels Corporation and the Customer 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 representauon. exoressed 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 wal not infnnge philately owned nghts: or assumes any liabilities wnh respect to the use of any informauon. ap.

paratus, method or process distsosed in this document.

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

In order to avoid impairment of rights of Advanced Nuclear Fuels Corporation in patents or inventions wtiich may be included in the infonnahon contained in this doctiment, the recipient. by its acceptance of this document. agrees not to publish or make public use (in the patent use of the term) of such inlormauon until so authonzed in wnting by Advanced Nuclear Fueis Corporahon or untd atter six (8) months following terminauon or expiration of the aforesaid Agreement and any extension thereof. unless otherwise expressly provided in the Agreement. No nghts or licenses in or to any palents are implied by!ne furnishing of this docu-menL ANF-3145,472A (12/87)

ANF-90-050 Page i TABLE OF CONTENTS Section ~Pa e

1. 0 INTRODUCTION

2. 0 FUEL MECHANICAL DESIGN ANALYSIS .

3.0 THERMAL HYDRAULIC DESIGN ANALYSIS . ~ ~ ~

3.2 Hydraulic Characterization . . . . ~ ~ ~

3.2. 1 Hydraulic Compatibility . . ~ ~ ~

3.2.2 Thermal Margin Performance, Comparison ~ ~ ~

3.2.3 Fuel Centerline Temperature ~ ~ ~

3.2.5 Bypass Flow . ~ ~ ~

3.3 MCPR Fuel Cladding Integrity Safety Limit ~ ~ ~ ~ ~ ~

3.3. 1 Coolant Thermodynamic Condition . . ~ ~ ~

3.3.2 Design Basis Radial Power Distribution ~ 4 ~

3.3.3 Design Basis Local Power Distribution ~ ~ ~

3.3.4 Channel Box Bow . . . . . ~ ~ ~

4.0 NUCLEAR DESIGN ANALYSIS . ~ ~ ~

4. 1 Fuel Bundle Nuclear Design Analysis ~ ~ ~

4.2 Core Nuclear Design Analysis . . . . . ~ ~ ~

4.2. 1 Core Configuration ~ ~ ~

4.2.2 Core Reactivity Characteristics ~ ~ ~

4.2.4 Core Reactivity Stability . ~ ~ ~

5.0 ANTICIPATED OPERATIONAL OCCURRENCES . ~ ~ ~ ~ ~ ~ 8

~

5. 1 Analysis Of Plant Transients At Rated Condi tions ~ ~ ~ 8 5.2 Analyses For Reduced Flow Operation ~ ~ ~ 8 5.3 Analyses For Reduced Power Operation . . . ~ ~ ~ 9 5.4 ASME Overpr essurization Analysis . ~ ~ ~ 9 5.5 Control Rod Withdrawal Error (CRWE) ~ ~ ~ 9 5.6 Fuel Loading Error . ~ ~ ~ 9 5.7 Determination Of Thermal Margins . ~ ~ ~ 10 6.0 POSTULATED ACCIDENTS ~ ~ ~ 11 6.1 Loss-Of-Coolant Accident . . ~ ~ ~ 11

6. 1. 1 Break Location Spectrum . ~ ~ ~ ~ ~ ~ ~ 11

6. 1.2 Break Size Spectrum . ~ ~ ~ 11

6. 1.3 MAPLHGR Analyses ~ ~ ~ 11 6.2 Control Rod Drop Accident ~ ~ ~ ~ ~ 12

7. 0 TECHNICAL SPECIFICATIONS ~ ~ ~ 13 7 ~ 1 Limiting Safety System Settings ~ ~ ~ ~ ~ ~ 13 7.1.1 MCPR Fuel Cladding Integrity Safety Limit ~ ~ ~ 13 7.1.2 Steam Dome Pressure Safety Limit ~ ~ ~ 13

ANF-90-050 Page ii TABLE OF CONTENTS (Continued)

Section Parcae 7.2 Limiting Conditions For Operation ~ ~ ~ ~ ~ ~ 13 7.2. 1 Average Planar Linear Heat Generation Rate Limits 13 7.2.2 Minimum Critical Power Ratio 13 7.2.3 LHGR Limits 14 7.3 Surveillance Requirements 14 7.3. 1 Scram Insertion Time Surveillance . 14 7.3.2 Stability Surveillance 15 8.0 METHODOLOGY REFERENCES 16 9.0 ADDITIONAL REFERENCES . 17 APPENDIX A SINGLE-LOOP OPERATION A-1 APPENDIX B SEISMIC-LOCA EVALUATION 8-1

ANF-90-050 Page iii LIST OF TABLES Table Pacae

4. NEUTRONIC DESIGN VALUES . . . . . . . . . . , . . . . 27 1 . . . . . . .

B.1 COMPARISON OF PHYSICAL AND STRUCTURAL CHARACTERISTICS FOR 8X8 AND 9X9 FUEL ASSEMBLIES B-2 LIST OF FIGURES

~Fi ere Pacae 3.1 SUSQUEHANNA UNIT 1 CYCLE 6 HYDRAULIC DEMAND CURVE POWER VERSUS FLOW 19 3.2 SUSQUEHANNA UNIT 1 CYCLE 6 DESIGN BAS'IS RADIAL POWER . 20 3.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-331L-10G5 ANF-5 FUEL 21 3.4 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-364L-9G4 ANF-5 FUEL . 22 3.5 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF-4 9X9 FUEL 23 3.6 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 PERIPHERAL FUEL 24 3.7 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 FUEL . 25 3.8 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-2 8X8 PERIPHERAL FUEL 26 4.1 SUSQUEHANNA UNIT 1 CYCLE 6 ENRICHMENT DISTRIBUTION FOR THE ANF92-364L-9G4 ANF-5 FUEL LATTICE o o ~ 28 4.2 SUSQUEHANNA UNIT 1 CYCLE 6 ENRICHMENT DISTRIBUTION FOR THE ANF92-331L-10G5 ANF-5 FUEL LATTICE . 29 4.3 SUSQUEHANNA UNIT 1 CYCLE 6 REFERENCE CORE LOADING PLAN . 30

5. 1 SUSQUEHANNA UNIT 1 CYCLE 6 CONTROL ROD WITHDRAWAL ERROR ANALYSIS CONTROL ROD PATTERN . 31 5.2 SUSQUEHANNA UNIT 1 CYCLE 6 FLOW-DEPENDENT MCPR OPERATING LIMIT 32

ANF-90-050 Page 1

1.0 INTRODUCTION

"~

This report provides the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF)* in support of the Cycle 6 reload for Susquehanna Unit 1, which is scheduled to commence operation in the Fall of 1990. This report is intended to be used in conjunction with ANF topical P 41 4, R 11 I, "RVTI I f h 1 Nuclear Company Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list (Section 8.0).

However, LHGR mechanical design limits (Reference 9. 1 for the 9x9 fuel and Reference 9. 13 for the Bx8 fuel) and plant transient simulation model developments pp I fh~,f14,R (Reference 9.2) have been revised by ANF subsequent to NRC 11 .hRf 9.2 have been approved by the NRC for referencing in license applications.

.Td Reference 9. 13 provides an evaluation which justifies extending the ANF Bx8 fuel LHGR limits out to an assembly exposure of 37,000 MWd/NTU. This report application.

Th P

XT BIIR I ~,VI is being submitted along with these documents in support of the U1C6 reload Section numbers in this report are the same as corresponding I I d d I,R Ih d Volume 1 and Supplements 1 and 2, as modified by Reference 9. 12, was used for neutronic calculations. The modification described in Reference 9. 12 improved the numerical fitting of cross-section data associated with void history accounting in the simulator code.

The Susquehanna Unit 1 Cycle 6 core will contain a total of 764 fuel assemblies, including 220 unirradiated ANF-5 9x9 assemblies, 228 irradiated ANF-4 9x9 assemblies, 240 irradiated ANF XN-3 9x9 assemblies, and 76 irradiated ANF XN-2 Bx8 fuel assemblies. The reference core configuration is described in Section 4.2.

Formerly Exxon Nuclear Company (ENC).

ANF-90-050 Page 2 The design and safety analyses reported in this document were based on the design and operational assumptions in effect for Susquehanna Unit 1 during the previous operating cycle with the following exceptions:

(1) Approximately 50 original equipment control blades will be replaced in Unit 1 Cycle 6 with control blades of equivalent worth.* Where appropriate, analyses presented herein have been reevaluated (Reference 9. 14) in a conservative manner using bounding input for the new control blade design, and the limiting results are provided.

(2) In previous cycles, for the feedwater controller failure event, the turbine stop valve was modelled as the limiting valve closure following the turbine trip. For this event, Susquehanna-specific plant data indicates that the turbine control valves reach their fully closed position prior to full closure of the stop valves for most initial power levels. Therefore, the turbine control valve was modelled as the limiting valve closure, where appropriate, in the Unit 1 Cycle 6 feedwater controller failure analysis.

(3) For Cycle 6, the bypass control model was revised to more accurately represent the quick opening feature of the Susquehanna plant design.

Additional information and the results of design studies detailing the development of 9x9 fuel assemblies for BWR reloads are contained in Reference 9.3.

Equivalent worth is defined as within a5X of original equipment worth.

ANF-90-050 Page 3 Revised 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable ANF Fuel Design Report: References 9. 1 and 9.13 To ensure that the expected power history for the fuel to be irradiated during Cycle 6 of Susquehanna Unit 1 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating limits (Figure 5. 1 of Reference 9. 13 for the 8x8 fuel and Figure 3.3 of Reference 9. 1 for the 9x9 fuel) have been specified. In addition, an LHGR transient operating limit for anticipated operating occurrences has been specified for ANF 8x8 fuel (Figure 5,2 of Reference 9. 13) and ANF 9x9 fuel (Figure 3.4 of Reference 9. 1).

The NRC has approved the ANF 9x9 fuel design for assembly exposures up to 40,000 MWd/MTU (Reference 9.4).

ANF-90-050 Page 4 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS

3. 2 H draul i c Characteri zati on 3.2.1 H draul i c Com atibil i t Component hydraulic resistances for the constituent fuel .types in the Susquehanna Unit 1 Cycle 6 core have been determined in single phase flow tests of full scale assemblies. Figure 3. 1 shows the hydraulic demand curves for ANF 9x9 fuel and ANF 8x8 fuel in the Susquehanna Unit 1 core. The similar hydraulic performance indicates compatibility for co-residence in the Susquehanna Unit 1 core.

3.2.2 Thermal Har in Performance Com arison Report XN-NF-87-23, "Susquehanna Unit 1 Cycle 4 Reload Analysis Design and Safety Analysis," Advanced Nuclear Fuels Corporation, Richland, Washington, April 1987, gives thermal margin comparisons.

3.2.3 Fuel Centerline Tem erature Applicable Generic Report Reference 9. 1 8x8 Extended Burnup Design Report Reference 9. 13

.5.5 ~F1 Calculated Bypass Flow Fraction at 10. I/o 104% Power/100% Flow 3.3 MCPR Fuel Claddin nte rit Safet Limit MCPR Safety Limit 1.06 3.3. 1 Coola t Thermod namic Condition Rated Thermal Power 3293 HWt Feedwater Flow Rate (at SLHCPR) 16. 1 Hlbm/hr Core Pressure (at SLHCPR) 1043 psia Feedwater Temperature 383'F 3.3.2 Desi n Basis Radial Power Distribution Figure 3.2

ANF-90-050 Page 5 3.3.3 Desi n Basis Local Power Distribution Figures 3.3-3.8 References 9. 15 and 9.16 On March 20, 1990, the USNRC issued bulletin no. 90-02 requiring assessment of loss of thermal margin caused by channel box bow (Reference 9. 15). 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 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 (Reference 9. 16).

0 ANF-90-050 Page 6 4.0 NUCLEAR DESIGN ANALYSIS

4. 1 Fuel Bundle Nuclear Desi n Anal sis Assembly Average Enrichments 3.52% and 3.21%

Radial Enrichment Distributions Figures 4.1 8 4.2 Axial Enrichment Distributions Uniform 3.64% and 3.31% with a 6" natural uranium top blanket Burnable Poisons* Figures 4. 1 8 4.2 Non-Fueled Rods Figures 4.1 & 4.2 Neutronic Design Parameters Table 4. 1 4.2 Core Nuclear Desi n Anal sis 4.2.1 Core Confi uration Figure 4.3 Core Exposure at EOC5, HWd/HTU 22,954 Core Exposure at BOC6, HWd/HTU 13)828 Core Exposure at EOC6, HWd/HTU 24,453 Maximum Cycle 6 Licensing Exposure Limit, HWd/HTU 25,069 4.2.2 Core Reactivit Characteristics BOC Cold k-eff, All Rods Out** 1.11537 BOC Cold k-eff, Strongest Rod Out** 0.98382 Reactivity Defect (R-Value) 0.07% rho Standby Liquid Control System Core k-eff, Cold Conditions, 660 ppm.Boron** 0.97931

Burnable poisons are distributed uniformly over the enriched length of the designated rods. The natural urania axial blanket sections do not contain burnable absorber material.

- Values biased to critical eigenvalue of 1.0.

ANF-90-050 Page 7 4.2.4 Core Reactivit Stabilit Power Flow State Points Deca Ratio COTRAN 60%/40% 0.77 63%/40% 0.87 66%/45% 0.73 72%/45% 0.87 70%/50% 0.60 75%/50% 0.72 Linear extrapolation along a constant decay ratio line to a higher power/flow condition is conservative compared to COTRAN calculations,

ANF-90-050 Page 8 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient References 9.5 Analysis Hethodology Report and 9.7

5. 1 'Anal sis Of Plant Transients At Rated Conditions Reference 9.6 Limiting Transients: Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LFWH)

% Rated % Rated Maximum Maximum Haximum Pressure Delta Event Power* Flow Heat Flux Power ~sia CPR** Model LRWB 100% 100% 123.3% 394% 1199 0.28 COTRANSA/

XCOBRA-T FWCF 100% 100% 117. 9% 212/e 1179 0.23 COTRANSA/

XCOBRA-T LFWH 100% 100% 121 5% 123% 1081 0.16 PTSBWR3/

XCOBRA Single-Loop Operation Appendix A 5 ' Anal ses For Reduced Flow 0 eration Reference 9.6 Limiting Transient: Recirculation Flow Increase Transient (RFIT)

Total Core Reduced Flow Recirculation Flow HCPR

/ Rated 0 eratin L'mit 8x8 Fuel 9x9 Fuel 100 1.11 1.11 90 1.15 1.15 80 1.19 1.19 70 1.24 1.24 60 1.31 1.31 50 1.43 1.43 40 1.61 1.61 30 1.92 1.92

104% power used in analysis as design basis.

- Delta-CPR results for most limiting fuel type.

ANF-90-050 Page 9 5.3 Anal ses For Reduced Power 0 eration Reference 9.6 Limiting Transient: Feedwater Controller Failure (FWCF)

Delta-CPR 1 Power Transient ANF 9x9 ANF 8x8 104 FWCF 0.23 0.23 80 FWCF 0.25 0.25 65 FWCF 0.27 0.27 40 FWCF 0.27 0.27 5.4 ASME Over ressurization Anal sis Reference 9.6 Limiting Event Full HSIV Isolation Worst Single Failure Direct Scram Maximum Pressure 1,312 psig Maximum Steam Dome Pressure 1,295 psig 5.5 Control Rod Withdrawal Error CRW Starting Control Rod Pattern for Analysis Figure 5. 1 1001 Flow Distance Rod Block Setting Withdrawn Delta 14 ~f't CPR 105 3.5 0.21 106 4'. 0 0.24 107 4.0 0.24 108* 4.5 0.26

5. 6 5555i EE Maximum Delta-CPR 0.21

Rod Block Monitor setting recommended for Cycle 6 operation.

ANF-90-050 Page 10 5.7 Determination Of Thermal Mar ins Summary of Thermal Margin Requirements:

Event Power Flow Delta-CPR* MCPR Limit LRWB 100%%u** 100% 0.28 1.34 LRW8 100%+* 100% 0.35 1.41 (RPT Inoperable)

FWCF 100%~ 100% 0.23 1.29 FWCF 100%" 100% 0.34 . 1.40 (Bypass Inoperable)

LFWH 100%* 100% 0.16 1,22 CRWE 100% 100% 0.26 at 108% RBM 1.32 MCPR Operating Limits at Rate'd Conditions:

MCPR 0 eratin imit 1.34 Reduced Flow MCPR Limits Figure 5.2 and Section 7.2.2 Power Dependent MCPR Operating Limit Results for Cycle 6:

Limiting Transient ANF 9x9 ANF 8x8 100*~/100 LRWB 1.34 1.34 80/100 LRWB*** 1.34 1.34 65/100 LRWB*** 1.34 1.34 40/100 LRWB*** 1.34 1.34

Delta-CPR results for most limiting fuel type.

- 104% power used in analysis as design basis.

- - Limited by 100/100 LRWB transient.

ANF-90-050 Page 11

6. 0 POSTULATED ACCIDENTS

6. 1 Loss-Of-Coolant Accident Seismic-LOCA Appendix B

6. 1. 1 Break Location S ectrum Reference 9.8

6. 1.2 Break Si e S ectrum Reference 9.8

6. 1.3 MAPI.HGR Anal ses ANF 9x9 Fuel Reference 9.9 Limiting Break: Double-ended guillotine pipe break Recirculation pump discharge line 0.4 discharge coefficient Peak Clad Peak Local Bundle Average Temperature MWR*

Ex osure HAPLHGR De ree F Percent

~GWd WTU kW ft XN-3** ANF-4 ANF-5*** XN-3 ANF-4 ANF-5 0 10.2 2060 2008 3.9 2.8 5 10.2 2069 1941 3.7 1.4 10 10.2 2121 2095 3.7 3.2 15 10.2 2140 2139 4.8 4.5 20 10.2 2147 2173 2141 5.2 , 5.1 4.50 25 9.6 2016 2000 2.7 2.5 30 8.9 1839 1833 1.0 1.0 35 8.2 1752 1744 0.7 0.7 40 7.5 1675 1668 0.5 0.5

Metal water reaction.

- Peak clad temperatures shown are for ANF 9x9 fuel in Susquehanna Unit 2.

Changes in PCT, due to minor fuel design changes, were investigated for ANF 9x9 fuel in Susquehanna Unit 1, and found to be insignificant. Calcula-tions were performed at exposure intervals of 5 GWd/HTU from BOL to 20 GWd/MTU,'nd at 30 and 40 GWd/MTU. All PCTs were lower; the largest change was a 21 'F decrease in PCT at 10 GWd/HTU and the smallest was a 1 'F decrease in PCT at 20 GWd/HTU.

- - The ANF-5 ANF92-331L-10G5 fuel type is similar to the XN-3 fuel type loaded in Cycle 4 and is bounded by those results. The ANF-5 ANF92-364L-9G4 fuel type is similar to the ANF-4 fuel type loaded in Cycle 5 except that the ANF-5 fuel type is slightly more edge peaked at the limiting exposure point for PCT and MWR. This exposure point was analyzed for the ANF92-364L-9G4 fuel type to confirm the ANF-4 fuel peak cladding temperatures are bounding.

ANF-90-050 Page 12 ANF SxS Fuel Reference 9. 10 Limiting Break: Double-ended guillotine pipe break Recirculating pump discharge line 0.4 discharge coefficient Peak Clad Peak Local Bundle Average Temperature* MWR**

Ex osure MAPLHG cDOFL ~Percent GWd MTU kw ft 0 13.0 2074 1.9 5 13.0 2093 2.0 10 13.0 2116 2.1 15 13.0 2136 2.2 19 13.0 2147 2.3 25 11.5 1977 1.6 30 10.4 1846 1.0 35 10.4 1852 1.2 40*** 10.'4 1832 1.7 6.2 Control Rod Dro Accident Dropped Control Rod Worth, mk 15.0 Doppler Coefficient, 1/k dk/dT -11.5 x (10)-6 Effective Delayed Neutron Fraction 0.0054 Four-Bundle Local Peaking Factor 1.26 Maximum Deposited Fuel Rod Enthalpy, cal/gm**** 205 Number of Rods Exceeding 170 cal/gm (600

Peak clad temperatures shown are for ANF XN-1 Sx8 fuel. The possible change in PCTs, due to the fuel design changes in ANF XN-2 Sx8 fuel, was investigated at three exposures. A 22 'F increase resulted at BOL, no change at 10 GWd/MTU, and a 15 'F increase resulted at 19 GWd/MTU.

- Metal water reaction.

- - Extended exposure point computed for Susquehanna Unit 1 Cycle 6.

- - - Deposited enthalpy was determined from parametric analysis where all parameters, except Dropped Control Rod Worth, were enveloped. The well-behaved relationship between rod worth and deposited enthalpy indicates that an extrapolation based on rod worth results in a conservative value of deposited enthalpy.

ANF-90-050 Page 13

7. 0 TECHNICAL SPECIFICATIONS

7. 1 imitin Safet S stem Settin s

7. 1. 1 MCPR Fuel Claddin Inte rit Safet imit MCPR Safety Limit 1.06 7,1.2 Steam Oome Pressure Safet Limit Pressure Safety Limit (as measured in steam dome) 1325 psig Analysis shows that a steam dome pressure safety limit of 1358 psig is allowed but the 1325 psig value used in previous cycles is to be conservatively retained.

7.2 Limitin Conditions For 0 eration 7.2. 1 Avera e Planar Linear Heat Generation Rate Limits Bundle Bundle Average MAPLHGR Limits Average MAPLHGR Limits Exposure (j(W/ft) Exposure (kW/ft)

~GWd WTU ANF 9x9 Fuel ~GWd WT ANF 8x8 Fuel 0 10.2 0 13.0 5 10.2 5 13.0 10 10.2 10 13.0 15 10.2 15 13.0 20 10.2 19 13.0 25 9.6 25 11.5 30 8.9 30 10.4 35 8.2 35 10.4 40 7.5 40 10.4 7.2.2 Minimum Critical Power Ratio MCPR Operating Limits at Rated Conditions:

Bypass Operable Bypass Inoperable Bypass Operable RPT 0 erable RPT 0 erable RPT Ino erable 1.34 1.40 1.41

ANF-90-050 Page 14 Revised MCPR Operating Limits at Reduced Flow:

~d Total Core Recirc. Flow 100 Bypass Operable RPT 0 erable 1.34 Bypass RPT 0 Inoperable erable 1.40 Bypass Operable RPT Ino erable 1.41 90 1.34 1.40 1.41 80 1.34 1.40 1.41 70 1.34 1.40 1.41 60 1.34 1.40 1.41 57.5 1.34 52.5 1.40 51.7 1.41 50 1.43 1.43 1.43 40 1.61 1.61 1.61 30 1.92 1.92 1.92 MCPR Operating Limits at Reduced Power:

Power Level Bypass Operable Bypass Inoperable Bypass Operable

% Rated RPT 0 erable RPT 0 erable RPT Ino erable 100* 1.34 1.40 1.41 80 1.34 1.42 1.41 65 1.34 1.44 1.41 40 1.34 1.44 1.41 25 1.34 1.44 1.41 7.2.3 LHGR Limits LHGR Limits: 9x9 Fuel Figures 3.3 and 3.4 of Reference 9. 1 8x8 Fuel Figures 5. 1 and 5.2 of Reference 9. 13 7.3 Surveillance Re uirements 7.3. 1 Scram Insertion Time Surveillance Thermal limits established in Section 5.0 are based on minimum acceptable scram insertion performance as defined in the Technical Specifications. No additional surveillance for scram insertion is required for validation of thermal limits.

104% power used in analysis as design basis.

ANF-90-050 Page 15 7.3.2 Stabilit Surveillance PPI%L 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 upon the calculation results presented in Section 4.2.4.

ANF-90-050 Page 16 8.0 HETHODOLOGY REFERENCES A complete bibliography of applicable methodology references is provided in the following document: "Exxon Nuclear Hethodology for Boiling Mater Reactors: Application of the ENC. Hethodology to BWR Reloads,"

~F-->>,F44,4111,dddd1F141 Richland, Washington 99352, Harch 1985.

Formerly Exxon Nuclear Company (ENC).

ANF-90-050 Page 17 Revised 9.0 ADDITIONAL REFERENCES 9.1 "Generic Mechanical Design for Exxon Nuclear Jet Reload Fuel>"

R "I I" I A d N I Pump F

BWR I R p Richland, Washington, September 4, 1986.

9.2 ii~ it lltidl 1 R P I I,"

"Exxon Nuclear Methodology for Boiling Water Reactors, THERHEX: Thermal

~tld -

Revision 2, Advanced Nuclear Fuels Corporati.on, Richland, Washington, Fi 1, January 1987.

9.3 "Demonstration of 9x9 Assemblies for BWRs," EPRI NP-3468, Electric Power Research Institute, Palo Alto, California, May 1, 1984.

9.4 "gualification of Exxon Nuclear Fuel for Extended Burnup - Supplement 1 R ddt tl lipi I FRN>>RIIRF I,"~t Supp 1 ement 1, Revi sion 2, Advanced Nucl ear Fuel s Corporation, 9'

9.6 Richland, Washington, Hay 1988.

->>, R I, "Exxon Nuclear Plant Transient Methodology

~R- - I Richland, Washington, November 16, 1981.

"Susquehanna d d N for Boiling I

Unit 1 Cycle 6 Plant Transient Analysis," ANF-90-049, 1 I Water Reactors,"

p Advanced Nuclear Fuels Corporation, Richland, Washington, March 1990.

9.7 "XCOBRA-T:

A I and 2, I," A Computer R~p-Advanced February 1987.

Nuclear Code for Fuels 11 BWR Transient Thermal-Hydraulic Core I

Corporation, d 11 I, ddl Richland, Washington, 9.8 "Generic LOCA Break Spectrum Analysis BWR 3 & 4 with Modified Low Pressure Coolant Injection Logic Using the EXEH Evaluation Model,"

XN-NF-84-117 P , Advanced Nuclear F'uels Corporation, Richland, Washington, December 1984.

9,9 "Susquehanna LOCA-ECCS Analysis HAPLHGR Results for ENC 9x9 Fuel,"

XN-NF-86-65, Advanced Nuclear Fuels Corporation, Richland, Washington, Hay 1986.

9.10 "Susquehanna Unit 1 LOCA-ECCS Analysis, MAPLHGR Results," XN-NF-84-119, Advanced Nuclear Fuels Corporation, Richland, Washington, December 1984.

Rl dAIIFR,R I R,'"~,RIFI,AddN 9.11 "Principal Reload Fuel Design Parameters, Corporation, Richland, Washington, Fuel Design, Susquehanna December 1989.

I Unit 1

1 Formerly Exxon Nuclear Company (ENC).

ANF-90-050 Page 18 Revised 9.12 Letter, R. A. Copeland (ANF) to M. W. Hodges (NRC), "Void History Correlation,"

Rdddddddgg Revision 1, RAC:058:88, September 13, 1988.

Advanced Nuclear Fuels Corporation, Richland, Washington, June 1990.

9.14 Letter, H. G. Shaw (ANF) to H. R. Vernick (PPBL), "DuraLife 120C Control Blade Evaluation for Susquehanna Unit 1 Cycle 6 Licensing," HGS:091:90, March 23, 1990.

9. 15 NRC Bulletin No. 90-02, "Loss of Thermal Margin Caused by Channel Box Bow," March 20, 1990.

9. 16 Letter, R. A. Copeland (ANF) to R. C. Jones (NRC), "Loss of Thermal Margin Caused by Channel Box Bow," RAC:030:90, April 9, 1990.

9, 17 Letter, H. R. Vernick (PPImL) to H. G. Shaw (ANF), "Susquehanna Unit 1 Cycle 6 Control Rod Design Information," PLE-12047, December 7, 1989, Attachment C.

ANF-90-050 Page 19 I

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120 100 80 n

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0 0.2 0.4 0.6 0.8 1.2 RRDIRL PONER PERKING FIGURE 3.2 SUSQUEHANNA UNIT I CYCLE 6 DESIGN BASIS RADIAL POWER

0 ANF-90-050 Page 21

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

FIGURE 3.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-331L-10G5 ANF-5 FUEL

ANF-90-050 Page 22

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FIGURE 3.4 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-364L-9G4 ANF-5 FUEL

e ANF-90-050 Page 23

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FIGURE 3.5 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF-4 9X9 FUEL

0 I

ANF-90-050 Page 24

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FIGURE 3.6 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 PERIPHERAL FUEL

ANF-90-050 Page 25

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FIGURE 3. 7 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-3 9X9 FUEL

ANF-90-050 Page 26

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FIGURE 3.8 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-2 SXS PERIPHERAL FUEL

ANF-90-050 Page 27 TABLE 4.1 NEUTRONIC DESIGN VALUES Fuel Pellet Reference 9, 11 Fuel Rod Reference 9. 11 Fuel Assembl Reference 9. 11 Core Data Number of fuel assemblies 764 Rated thermal power, HW 3293 Rated core flow, Hlbm/hr 100 Core inlet subcooling, Btu/ibm 24.0 Hoderator temperature, 'F 548.8 Channel thickness, inch 0.080 Fuel assembly pitch, inch 6.00 Wide water gap thickness, inch 0.562 Narrow water gap thickness, inch 0.562 Control Rod Data Absorber material B4C Total blade span, inch 9.75 Total blade support span, inch 1.58 Blade thickness, inch 0.260 Blade face-to-face internal dimension, inch 0.200 Absorber rods per blade 76 Absorber rod outside diameter, inch 0.188 Absorber rod inside diameter, inch 0.138 Absorber density, 5 of theoretical 70 Note: The above values are representative of a Susquehanna plant original equipment control blade. UIC6 will include a new control blade design with slightly different values than those shown above.

However, the new blade design is neutronically compatible with the original equipment (Reference 9. 17).

ANF-90-050 Page 28

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0 ANF-90-050 Page 29 L:

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HL : H* MH MH : MH M: ML ML : H : H : M* HH : M* : M HL : ML LL : HL : ML : H H : H : HL : HL LL LLL RODS ( 1) 1.80 W/0 U235 LL RODS ( 3) 2.00 W/0 U235 L RODS ( 2) 2.20 W/0 U235 ML RODS 16) 2.52 W/0 U235 H RODS (19) 3.25 W/0 U235 HH RODS (13) 3.66 W/0 U235 H RODS (15) 4.50 W/0 U235 H* RODS (10) 3.25 W/0 U235 + 5.00 W/0 GD203 W RODS ( 2 INERT WATER ROD FIGURE 4.2 SUS(UEHANNA UNIT 1 CYCLE 6 ENRICHMENT DISTRIBUTION FOR THE ANF92-331L-10G5 ANF-5 FUEL LATTICE

ANF-90-050 Page 30 1 2 3 4 5 6 7 8, 9 10 11 12 13 14 15 1 82 82 C1 82 C1 82 DO 82 DO 82 DO 82 DO C1 82 2 82 EO 82 EO 82 EO C1 EO 82 DO 82 EO 82 C1 A3 3 C1 82 82 C1 82 C1 82 C1 82 C1 EO C1 DO C1 A3 4 82 EO C1 EO C1 EO 82 EO 82 EO C1 EO 82 C1 A3 5 C1 82 82 C1 DO 82 C1 C1 82 C1 82 C1 DO C1 82 6 82 EO C1 EO 82 EO 82 DO 82 EO 82 EO 82 CI A3 7 DO C1 82 82 C1 82 C1 82 DO C1 DO C1 DO C1 A3 8 82 EO C1 EO C1 DO 82 EO 82 EO C1 DO C1 A3 9 DO 82 82 82 82 82 DO 82 C1 82 DO C1 A3 10 82 DO C1 EO C1 EO C1 EO 82 C1 C1 A3 A3 11 DO 82 EO C1 82 82 DO C1'O C1 A3 12 82 EO C1 EO C1 EO C1 DO C1 A3 XY Fuel Type X Burned Y Cycles 13 DO 82 DO 82 DO 82 DO C1 A3 A3 14 C1 C1 C1 C1 C1 C1 C1 A3 15 82 A3 A3 A3 82 A3 A3 Fuel Type No. of Bundles Description 76 XN-2 ANF SXB 2.89 w/o U-235 240 XN-3 ANF 9X9 3.31 w/o U-235 228 ANF-4 ANF 9X9 3.33 w/o U-235 108 ANF-5 ANF 9X9 3.52 w/o U-235 112 ANF-5 ANF 9X9 3.21 w/o U-235 FIGURE 4.3 SUSQUEHANNA UNIT 1 CYCLE 6 REFERENCE CORE LOADING PLAN

e ANF-90-050 Page 31 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 59 55 04 06 04 55 51 30 -- 32 -- 32 30 51 47 06 08 10 -- 08 -- 06 47 43 -- -- 30 36 -- 32 -- 32 36 30 -- -- 43 39 -- 04 -- 08 12 00 -- 12 -- 08 -- 04 -- 39 35 -- -- 32 32 32 32 -- -- 35 31 -- 06 -- 10 00 00 -- 10 -- 06 -- 31 27 -- -- 32 32 32 32 -- -- 27 23 -- 04 -- 08 12 ** 12 -- 08 -- 04 -- 23 19 -- --. 30 -- 36 -- 32 -- 32 36 30 -- -- 19 15 06 -- 08 -- 10 -- 08 -- 06 15 30 -- 32 -- 32 -- 30 04 -- 06 -- 04 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58

- Control Rod Being Withdrawn Rod Position in Notches Withdrawn Full in 00 Full out FIGURE 5.1 SUSQUEHANNA UNIT 1 CYCLE 6 CONTROL ROD WITHDRAWAL ERROR ANALYSIS CONTROL ROD PATTERN

NOTE: The HCPR operating limit shall be ANF 9X9 the maximum of this curve, the full flow HCPR, or the power dependent HCPR operating limit.

as C9 X

F 5

as O

LLJ 1A O

j5 h

1.2 1

30 40 50 60 70 80 90 100 110 TOTAL CORE RECIRCULATlON FLOW Fo RATED RGURE 5.2 SUSQUEHANNA UNIT 1 CYCLE 6 FLOW-DEPENDENT MCPR OPERATING UMIT

0 0

ANF-90-050 Page A-1 APPENDIX A SINGLE- LOOP OPERATION This appendix provides limits and justification of those limits for single-loop operation.

A.l ANTICIPATED OPERATIONAL OCCURRENCES AND HCPR SAFETY LIMIT HCPR 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 (Reference A-1). 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 HCPR limit at maximum power (i.e., 76%) in single-loop operation is equal to the two-pump operating HCPR limit at 100% power. The flow dependence of the HCPR limit is based on a flow increase transient from run-up of two pumps.

For single-loop operation, the 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'valuated 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.

The two-loop HCPR operating limits plus 0.01 conservatively protect the fuel from any transient in single-loop operation;

ANF-90-050 Page'A-2 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-loop operation is also applicable to single-loop operation with ANF fuel (Reference A-2).

A.2.2 Pum Seizure Accident ANF has analyzed the pump seizure accident from single-loop operating conditions on a generic basis for the Susquehanna Units (Reference A-1). 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 operation at the single-loop operating HCPR limit, the radiological consequences of a pump seizure accident from single-loop operating conditions are but a small fraction of the 10CFR100 guidelines.

A single-loop MCPR operating limit of 1.30 is 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 STABILITY 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 MCPR 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-90-050 Page A-3 A.5 REFERENCES A-1 "Susquehanna Unit 1 Cycle 6 Plant Transient Analysis," ANF-90-049, Advanced Nuclear Fuels Corporation, Richland, WA 99352, March 1990.

A-2 0. R. Swope, "Susquehanna LOCA Analysis for Single Loop Operation,"

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

ANF-90-050 Page B-I APPENDIX B SEISMIC- LOCA EVALUATION The structural response of Advanced Nuclear Fuels Corporation's (ANF's) 9x9 fuel is similar to the structural response of the GE 8x8R fuel it has replaced in the Susquehanna Unit I core. Therefore, the seismic-LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertion will not be inhibited following the occurrence of the design basis seismic-LOCA event.

The physical and structural properties of the 9x9 and the Bx8 fuel types which are important to the dynamic response of the fuel are summarized in Table B. l. The close agreement between the important parameters for the ANF 9x9 and GE 8x8R fuel types indicates that the structural response would be very similar for both fuel types.

Similarity of the natural frequencies of the two fuel types mentioned above is further assured by the stiffness of the fuel assembly channel box.

Both fuel types use the same fuel assembly channel box, and the channel box dominates the overall dynamic response of the incore fuel. ANF calculations show that approximately 97% of the stiffness of a fuel assembly is attributable to the stiffness of the channel box. For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original seismic-LOCA analysis remains applicable.

Deformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.

ANF-90-050 Page B-2 TABLE B.1 COMPARISON OF PHYSICAL AND STRUCTURAL CHARACTERISTICS FOR 8X8 AND 9X9 FUEL ASSEMBLIES fuel T es

~Pro ert ANF 9x9 ANF 8x8 GE Bx8R Assembly Weight, lbs 580 596 600 Number of Spacers 7 7 Overall Assembly Length, in 171.29 171.29 171.40 Assembly Frequencies, cps Mode 1 1.9 1.7 2 3.7 3.5 3 6.5. 6.5 4 10.4 10.8 5 15.5 16.6 6 21.9 24.2 7 29.1 33.9

GE propri etary.

ANF-90-050 Issue Date: 05/30/90 SUSQUEHANNA UNIT I CYCLE 6 RELOAD ANALYSIS Design and Safety Analyses Oi<< lI G. J. Busselman R. A. Copeland L. J. Federico A. L. B. Ho N. L. Hymas S. E. Jensen T. L. Lotz J. L. Haryott S. C. Hellinger

-T. E. Hillsaps L. A. Nielsen P. H. O'eary A. Reparaz J. A. White H. E. Williamson H. G. Shaw/PPFL (25)

Document Control (5)

0 PL-NF-9P-0Q3 Susquehanna SES Unit 1 Cycle 6 RELOAD

SUMMARY

REPORT Nuclear Fuels Engineering June 1990

~ ~ ~

Pennsylvania Power 5. Light Company 7/2/90...9007120229 I

DC

SUSQUEHANNA SES UNIT I CYCLE 6 RELOAD

SUMMARY

REPORT Prepared by: 8 'N R 6-~2-~0 K. P. Renninge Engineer Level II-Nuclear Fuels Engineering R. M. Rose/SSES Unit I Group Leader-Nuclear Fuels Engineering A. J. Roscioli/Senior Project Engineer-Nuclear Fuels Engineering Concur with:

. H. Emmett/Supervisor-Nuclear Fuels Engineering Approved by:

St ko Manager-Nu ear F ls 8 Systems Engineering PENNSYLVANIA POMER I LIGHT COMPANY

NOTICE This technical .report was derived in part from information developed during PPKL's nuclear design activities and from safety and licensing information'rovided to PP8L by Advanced Nuclear Fuels Corporation. This report is being submitted by PPSL to the U.S. Nuclear Regulatory Commission specifically in support of the Susquehanna Steam Electric Station Unit I Cycle 6 reload license amendment. In demonstrating compliance with the U.S. Nuclear Regulatory Commission's regulations, the information contained herein is true and correct to the best of PP8L's knowledge, information, and belief.

CONTENTS Pa<ac

1.0 INTRODUCTION

2.0 GENERAL DESCRIPTION OF RELOAD SUBMITTAL SCOPE 3.0 SSES UNIT 1 CYCLE 5 CORE OPERATING HISTORY ~

4.0 RELOAD CORE DESCRIPTION 5.0 'CONTROL BLADES .

6.0 FUEL MECHANICAL DESIGN .

7.0 THERMAL HYDRAULIC DESIGN . 10 7.1 Hydraul i c Compatibility . 10 7.2 Safety Limit MCPR . 10 7.3 Core Bypass Flow. 11 7.4 Core Stability. 11 8.0 NUCLEAR DESIGN . 12

'3

8. 1 Fuel Bundle Nuclear Design.

8' Core Reactivity . 14 8,3 Contrast of Cycle 6 Core with Cycle 5 15 8,4 New Fuel Storage Vault/Spent Fuel Pool Criticality. 16 8.4. 1 New Fuel Storage Vault . 16 8.4.2 Spent Fuel Pool. ~ ~ ~ 17 9.0 CORE MONITORING SYSTEM . 18 10.0 ANTICIPATED OPERATIONAL OCCURRENCES. 18

'0. 1 Core-Wide Transients. 20 10.2 Local Transients. 20 10.3 Reduced Flow/Power Operation. . 21 10.4 ASME Overpressurization Analysi s ~ 21 11.0 POSTULATED ACCIDENTS . 22

11. 1 Loss-of-Coolant Accident. . 23 11.2 Control Rod Drop Accident . 24 12.0 SINGLE LOOP OPERATION . 24 REFERENCES. ~ ~ 26

1.0 INTROOUCTION Susquehanna Steam Electric Station (SSES) Unit 1 Cycle 6 will be the third reload of Advanced Nuclear Fuels Corporation 9x9 fuel in SSES Unit 1 and the second nuclear design developed by PPKL. This report provides a general discussion and summary of the results of the reload analyses performed by PPEL and Advanced Nuclear Fuels Corporation (ANF) in support of SSES Unit 1 Cycle 6 (UlC6) operation. PPKL developed the nuclear design and performed related analyses (e.g., Shutdown Margin, Hot Excess Reactivity, and cycle length determination). ANF provided the necessary Safety and Licensing Analyses. Also included are a description of the UIC6 reload fuel and core design, a description and discussion of control blade replacements for U1C6, and a brief discussion of the license amendment (i.e., proposed Technical Specification changes). The analyses, evaluations, and results presented in this report and the reports referenced herein are similar to those submitted in support of SSES Unit 2 Cycle 4 operation (Reference 1) which were approved by the NRC (Reference 2).

The ANF U1C6 Reload Analysis Report ANF-90-050 (Reference 3), Plant Transient Analysis Report ANF-90-049 (Reference 4), and Bx8 Extended Burnup Design Report ANF-90-018(P), Revision 1 (Reference 5) along with the proposed changes to the SSES Technical Specifications serve as the basic framework for the reload core licensing submittal. Where appropriate, reference is made-to these and other supporting documents for more detailed information and/or specifics of the applicable analysis. The ANF Reload Analysis Report is intended to be used in conjunction with ANF topical report XN-NF-80-19(P)(A), Vol. 4 Rev. 1, "Application of ENC Methodology to BWR Reloads" (Reference 6) which describes in more detail the analyses performed in support of the reload and identifies the methodology used for those analyses. The list of references provided at the end of this document contains the SSES specific reload documents prepared by ANF and the applicable ANF generic reload documents (generic methodology previously approved or currently

under review) which are being used in support of the U1C6 reload core submittal.

The issue of core stability has been addressed through several calculations (Section 4.2.4 of Reference 3), previous startup tests (Section 7.4 of this report), and implementation of the interim operating guidelines presented in NRC Bulletin 88-07 Supplement 1 via Technical Specifications'his approach is consistent with the current Unit 1 Cycle 5 method for addressing core stability.

2.0 GENERAL DESCRIPTION OF RELOAD SUBMITTAL SCOPE During the fifth refueling and inspection outage at SSES Unit 1, PP&L will be replacing 220 irradiated fuel assemblies (approximately 29 percent of the previous Cycle 5 core) with 220 fresh ANF-5 9x9 fuel assemblies. The fuel being replaced includes 4 XN-1 8x8 assemblies and 216 XN-2 Bx8 assemblies. The ANF-5 9x9 fuel has similar operating characteristics (thermal-hydraulic, and nuclear) to the ANF-4 9x9 design which has previously been approved (Reference 7) for coresidence with the ANF 8x8 fuel that will remain in the core. The Cycle 6 reload core required the performance of a wide range of analyses to support UIC6 core operation. These included analyses for anticipated operational occurrences and postulated accidents. In addition, analyses were performed to support Single Loop Operation (SLO) for Unit 1 Cycle 6 and future operating cycles. Analyses for normal operation of the reactor consisted of fuel evaluations in the areas of mechanical, thermal-hydraulic, and nuclear design.

Based on PPKL's design and ANF's safety analyses of the Cycle 6 reload core, a number of proposed changes to the SSES Unit 1 Technical Specifications have resulted. The rationale used to arrive at these

proposed changes is contained in the discussions and documentation that follow.

A list of those Technical Specifications, applicable Bases, and Design Features PP&L proposes to change is given below:

Pro osed Chan es to Technical S ecifications 3/4.2. 1 - Average Planar Linear Heat Generation Rate 3/4.2.3 - MCPR Operating Limits 3/4.2,4 - Linear Heat Generation Rate 3/4.4, 1 - Recirculation System Pro osed Chan es to Technical S ecification Bases 2, 1 - Safety Limits 3/4.2 - Power Distribution Limits 3/4.4 - Reactor Coolant System Pro osed Chan es to Desi n Features 5.3 - Reactor Core 3.0 SS S UNIT 1 CYCLE 5 COR OPERATING H STORY To date, the Cycle 5 core has operated with power distributions that will yield end-of-cycle power and exposure shapes consistent with the planned operating strategy. Actual core follow operating data at the time of the reload core design analysis was used, together with projected plant operation, as a basis for the Cycle 6 core design and as input to the safety analyses. The Cycle 5 core will be operated within the assumptions of the Cycle 6 core analysis; therefore, the remainder of Cycle 5 core operation will not affect the, licensing basis of the Cycle 6 reload'ore.

0 4.0 RELOAD CORE DESCRIPTION The UlC6 core designed by PP&L will consist of 764 fuel assemblies, which include 220 fresh ANF 9x9 assemblies (ANF-5), 228 once burned ANF 9x9 assemblies (ANF-4), 240 twice burned ANF 9x9 assemblies (XN-3), and 76 thrice burned ANF Bx8 assemblies (XN-2). The ANF-5 reload fuel consists of 108 bundles which contain nine burnable poison rods with 4.0 wt% Gdz0>

(9Gd4) at a bundle average enrichment of 3.52 wt% U-235 and 112 bundles which contain 10 burnable poison rods with 5.0 wt% Gdz0s (10Gd5) at a bundle average enrichment of 3.21 wt% U-235. A breakdown by bundle type/bundle average enrichment is provided in the following table:

Number of Bundles Bundle T e 108 ANF 9x9/3.52 wt% U235 fresh ANF-5 (9Gd4) 112 ANF 9x9/3.21 wt% U235 fresh ANF-5 (10Gd5) 228 ANF 9x9/3.33 wt% U235 once burned ANF-4 (9Gd4) 240 ANF 9x9/3.31 wt% U235 twice burned XN-3 (9Gd4) 76 ANF 8x8/2.89 wt% U235 thrice burned XN-2 (6Gd4)

The anticipated Cycle 6 core loading configuration along with additional core design details is presented in Section 4.0 of the ANF U1C6 Reload Analysis Report (Reference 3). The core is a conventional scatter loading with the lowest reactivity bundles placed in the peripheral region of the core. The loading pattern was designed to obtain the

'equired energy while meeting the constraints on shutdown margin, hot excess reactivity, and power peaking. In order to successfully obtain the required energy and meet these constraints, a split batch (different enrichments and gadolinia designs) was employed.

In response to IE Bulletin 79-26, Rev. 1, PP&L committed to replacing control blades prior to exceeding a limit of 34 percent 8'epletion averaged over the upper one-fourth of the control blade (Reference 8).

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To ensure that this limit is not exceeded during Susquehanna SES Unit 1 Cycle 6 operation, PPINL plans to replace up to 50 of the original equipment control blades with GE Duralife 160C control blades. The Duralife 160C control blade is desi'gned to eliminate the B~C tube cracking and increase the control blade assembly life. The main differences between the Duralife 160C control blades and the original equipment control blades are:

a) the Duralife 160C control blades utilize improved B~C tube material (i.e. high purity stainless steel vs. commercial purity stainless steel) to eliminate cracking during the lifetime of the control blade; b) the Duralife 160C control blades utilize three solid hafnium rods at each edge of the cruciform which replace the three B~C rods that are most susceptible to cracking to increase control blade life; c) the Duralife 160C control blades contain additional B~C tubes in place of the stiffeners, have an increased sheath thickness, utilize a full length weld to attach the handle and velocity limiter, and contain additional coolant holes at the top and bottom of the sheath which result in a crevice-free structure; d) the Duralife 160C control blades utilize low cobalt-bearing pin and roller materials in place of stellite which was previously utilized; e) the Ouralife 160C control blades are longer by approximately 3. 1 inches in order to facilitate fuel moves within the reactor vessel during refueling outages at Susquehanna SES; and f) the Ouralife 160C control blades are approximately 16 pounds heavier as a result of the design changes described above..

The Ouralife 160C control blade has been evaluated to assure it has adequate structural margin under loading due to handling, and normal,

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emergency, and faulted operating modes. The loads evaluated include those due to normal operating transients (scram and jogging), pressure differentials, thermal gradients, seismic deflection, irradiation growth, and any other lateral and vertical loads expected for each condition.

The Duralife 160C control blade stresses, strains, and cumulative fatigue have been evaluated and result in an acceptable margin to safety. The control blade insertion capability has been evaluated and it has been determined to be capable of insertion into the core during all modes of plant operation within the limits of plant analyses. The Ouralife 160C control blade coupling mechanism is equivalent to the original equipment coupling mechanism and is fully compatible with the existing control rod drives in the plant. In addition, material selected is compatible with the reactor environment. The impact of the increased weight of the control blades on the seismic and hydrodynamic load evaluation of the reactor vessel and internals has been reviewed and found to have a negligible effect on existing analyses.

With the exception of the crevice-free structure and the extended handle, the Ouralife 160C control blades are equivalent to the NRC approved Hybrid I Control Blade Assembly (Reference 9). The mechanical aspects of the crevice-free structure were approved by the NRC for all control blade designs in Reference 10. A neutronics evaluation oF the crevice-free structure for the Ouralife 160C design was performed by GE using the same methodology as was used for. the Hybrid I control blades in Reference 9.

These calculations were performed for the original equipment control blades and the Ouralife 160C control blades described above assuming an array of ANF 9x9 fuel. The Duralife 160C control blade has a slightly higher worth than the original equipment design, but the increase in worth is within the criterion for nuclear interchangeability. The increase in blade worth has been taken into account in the appropriate UIC6 analyses. However, as stated in Reference 9, the current practice in the lattice physics methods is to model the original equipment all B,C control blade as non-depleted. The effects of control blade depletion on core neutronics during a cycle are small and are inherently taken into account by the generation of a target k-effective for each cycle. As discussed above, the neutronics calculations of the crevice-free structure show that the non-depleted Duralife 160C control blade has direct nuclear interchangeability with the non-depleted original equipment all B~C design. The Duralife 160C also has the same end-of-life reactivity worth reduction limit as the all B~C design. Therefore, the Ouralife 160C can be used without changing the current lattice physics models as previously approved for the Hybrid I control blades (Reference 9).

The extended handle and the crevice-free structure features of the Ouralife 160C control blades result in a one pound increase in the control blade weight over that of the Hybrid I blades, and a sixteen pound increase over the Susquehanna SES original equipment control blades. In Reference 9, the NRC approved the Hybrid I control blade which weighs less (by more than one pound) than the 0 lattice control blade. The basis of the Control Rod Drop Accident analysis continues to be conservative with respect to control rod drop speed since the Duralife 160C control blade weighs less than the 0 lattice control blade, and the heavier D lattice control blade speed is used in the analysis. In addition, GE performed scram time analyses and determined that the Ouralife 160C control blade scram times are not significantly different than the original equipment control blade scram times. The current Susquehanna SES measured scram times also have considerable margin to the Technical Specification limits. Since the increase in weight of the Duralife 160C control blades does not significantly increase the measured scram speeds and the safety analyses which involve reactor scrams utilize the Technical Specification limit scram times, the safety analyses are not affected.

Since the Ouralife 160C control blades contain solid hafnium rods in locations where the B4C tubes have failed, and the remaining B~C rods are manufactured with an improved tubing material (high purity stainless steel vs. commercial purity stainless steel), boron loss due to cracking is not expected. PPLL plans to track the depletion of each control blade and discharge any control blade prior to a ten percent loss in reactivity

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worth. Therefore, the requirements of IE Bulletin 79-26, Revision 1 do not apply to the Duralife 160C control blades.

6.0 FUEL MECHANICAL DESIGN The mechanical design and supporting analyses of the ANF-5 fuel are the same as those for the SSES Unit 1 Cycle 5 ANF-4 fuel and are described in XN-NF-85-67(P)(A), Revision 1 (Reference 11), XN-NF-84-97 (Reference 12),

PLA-2728 (Reference 13), XN-NF-82-06(P)(A), Supplement 1, Revision 2 (Reference 14). Each ANF-5 reload fuel assembly contains 79 fueled rods and two water rods in a 9x9 rod array. One of the water rods functions as a spacer capture rod. Seven spacers maintain fuel rod spacing.

Generic mechanical design analyses were performed to evaluate the steady state strain, transient strain, cladding fatigue, creep collapse, cladding corrosion, hydrogen absorption, differential fuel rod growth, and grid spacer spring design for both the ANF 8x8 and 9x9 fuel designs.

The RODEX2, RODEX2A, RAMPEX and COLAPX codes were used in the generic mechanical design analyses. All parameters meet their respective design limits as shown in References 3, 5 and 11. The Susquehanna SES specific analyses for the 8x8 design (Reference 5), which are included in this UIC6 submittal, are applicable to the XN-2 fuel and support a maximum 8x8 assembly discharge exposure of 37,000 HWD/HTU. This is an extension of 2,000 MWD/HTU beyond the 35,000 MWD/HTU 8x8 assembly exposure limit approved generically in Reference 14. The generic analyses for the 9x9 design (Reference 11) are applicable to the XN-3, ANF-4, and ANF-5 fuel designs and support a maximum 9x9 assembly discharge exposure of 40,000 HWD/HTU. Based on calculati'ons, UIC6 operation is projected to result in a peak 8x8 assembly exposure greater than 35,000 HWD/HTU but less than 37,000 MWD/MTU. In addition, the peak 9x9 assembly exposure is projected to be less than 40,000 HWD/HTU.

For the ANF Bx8 and 9x9 fuel, the design's such that margin to fuel mechanical design limits (e.g., centerline melting temperature, transient strain, etc.) is assured for all anticipated operational occurrences throughout the life of the fuel as demonstrated by the fuel design analyses (References 5 and 11), provided that the fuel rod power history remains within the power histories assumed in the analyses. The steady state design power profile for the ANF Bx8 fuel is shown in Figure 5.1 of Reference 5 and the steady state power profile for the ANF 9x9 fuel is shown in Figure 3.3 of Reference 11. The Reference 5 power history is the same as the Reference 11 Bx8 power history to a planar exposure of 42,000 MWD/MTU. These power profiles are incorporated into the Technical Specifications as operating limits. In addition, a Technical Specification provision for reducing the APRM scram and rod block settings by Fraction of Rated Power divided by Maximum Fraction of Limiting Power Density (FRP/MFLPD) was incorporated. This ensures that ANF fuel does not exceed design limits during an overpower condition for transients initiated from partial power. The LHGR curve used for calculating MFLPD for ANF fuel is based on ANF's Protect'ion Against Fuel Failure (PAFF) line shown in Figure 3.4 of Reference 11 and is incorporated into the Technical Specifications. The,.Technical Specification curve represents the LHGR corresponding to the ratio of PAFF/1.2, under which cladding and fuel integrity (i.e., 1/ clad strain and fuel centerline melting) 'is protected during AOOs. This curve is applicable to both 8x8 and 9x9 ANF fuel.

The structural response of the ANF 8x8 and 9x9 assembly designs during Seismic-LOCA events is essentially the same as the response of the GE Bx8R assembly design that comprised the initial Susquehanna SES Unit 1 core. The similar physical properties and bundle natural frequencies result in nearly identical structural responses as discussed in ANF-90-050 Appendix B (Reference 3). In addition, Reference 11 presents the Seismic-LOCA analysis for ANF 9x9 fuel which showed large design margins to fuel design limits for all ANF 9x9 fuel assembly components.

Additional justification (Reference 13) was also provided to the NRC by PPEL during the Unit 2 Cycle 2 reload licensing process. The ANF Bx8 fuel Seismic-LOCA evaluation was presented in the U1C3 reload submittal (Reference 15).

0 7.0 THERMAL HYDRAULIC DESIGN XN-NF-80-19(P)(A), Volume 4 Revisjon 1 (Reference 6) presents the primary thermal hydraulic design criteria which require analyses to determine:

(1) hydraulic compatibility of the ANF 8x8 and 9x9 fuel assemblies in the core, (2) the HCPR fuel cladding integrity safety limit, (3) bypass flow characteristics, and (4) thermal-hydraulic stability. The analyses performed to determine each of these parameters are discussed in this section.

7.1 H draulic Com atibilit Component hydraulic resistances for the ANF Bx8 and 9x9 fuel have been determined in single phase flow tests of full scale assemblies.

Reference 3 summarizes the resistances and evaluates the effects on thermal margin due to the coresidence of the ANF Bx8 and 9x9 fuel assemblies for Unit 1 Cycle 6. The NRC has previously approved coresidence of ANF 8x8 and 9x9 fuel for Unit 1 (Reference 7),

7.2 Safet Limit HCPR The MCPR fuel cladding integrity safety limi,t for UIC6 is 1.06 which is equal to the Unit 1 Cycle 5 HCPR fuel cladding integrity safety limit. The methodology and generic uncertainties used in the MCPR safety limit calculation are provided in XN-NF-80-19(P)(A), Volume 4 Revision 1 (Reference 6). The SSES U1C6 specific inputs and HCPR Safety Limit calculation are provided in ANF-90-049, Appendix B (Reference 4).

During U1C6, as in the previous cycle, the ANF 8x8 and 9x9 fuel will be monitored using the XN-3 critical power correlation. ANF has determined that this correlation provides sufficient conservatism to preclude the need for any penalty due to channel bow during UlC6.

Susquehanna SES is a C-lattice plant and uses channels for only one fuel bundle lifetime. The conservatism has been evaluated by ANF to be greater than the maximum expected bCPR (0.02) due to channel bow in C-lattice plants using channels for only one fuel bundle lifetime. Therefore, the monitoring of the HCPR limit is conservative with respect to channel bow and addresses the concerns of NRC Bulletin No. 90-02 (Reference 16). The details of the evaluation performed by ANF have been reported generically to the NRC (Reference 17).

7.3 Core 8 ass Flow Core bypass flow is calculated using the methodology described in XN-NF-524(A), Revision 1 (Reference 18). The core bypass flow fraction (including water rod flow) for U1C6 is 10. 1% of total core flow which is similar to the Cycle 5 bypass flow value of 9.9%. The bypass flow fraction is used in the HCPR safety limit calculation and as input to the cycle specific transient analyses.

~Cbi 1 i COTRAN core stability calculations were performed for Unit 1 Cycle 6 to determine the decay ratios at predetermined power/flow conditions. The resulting decay ratios (Reference 3) were used to define operating regions which comply with th'e interim requirements of NRC Bulletin No. 88-07, Supplement 1 "Power Oscillations in Boiling Water Reactors," (Reference 19). As in the previous cycle, Regions B and C of the NRC Bulletin have been combined into a single region (i.e., Region II), and Region A of the NRC Bulletin corresponds to Region I.

Region I has been defined such that the decay ratio for all allowable power/flow conditions. outside of the region is less than 0.90. To mitigate or prevent the consequences of instability, entry into this region requires a manual reactor scram. Region I for Unit 1 Cycle 6 has been calculated to be slightly larger than Region I for the previous cycle. '

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0 Region II has been defined such that the decay ratio for all allowable power/flow conditions outside of the region (excluding Region I) is less than 0.75. For Unit 1 Cycle 6, Region II must be immediately exited if it is inadvertently entered. Similar to Region I, Region II is slightly larger than in the previous cycle.

In addition to the region definitions, PP5L has performed stability tests in SSES Unit 2 during initial startup of Cycles 2 and 3 to demonstrate stable reactor operation with ANF 9x9 fuel. The test results for U2C2 (Reference 20) show very low decay ratios with a core containing 324 ANF 9x9 fuel assemblies, Analysis of data taken during U2C2 Two Loop Operation at 60% power and 47/ flow resulted in a "measured" decay ratio of 0.33 and a COTRAN calculated decay ratio of 0.33. In Single Loop Operation at 55% power and 44% flow the "measured" decay ratio was 0.30 and the COTRAN calculated value was 0.29. In addition, the use of the ANF "ANNA" software to analyze APRH signals from the U2C3 startup produced a "measured" decay ratio

'of -0.37 at 60/ power and 46% flow. The U2C3 core contained 556 ANF 9x9 assemblies. The U2C4 core contains 764 (full core) ANF 9x9 assemblies. Two loop stability tests similar to those described above were performed at BOC 4 and the test data has been sent to the NRC (Reference 21). Stability tests are not planned for UIC6.

PPKL believes that the use of Technical Specifications that comply with NRC Bulletin 88-07 Supplement 1,. and the tests and analyses described above, will provide assurance that SSES Unit 1 Cycle 6 will comply with General Design Criteria 12, Suppression of Reactor Power Oscillations. This approach is consistent with the SSES Unit 1 Cycle 5 method for addressing core stability (Reference 22 and 23).

8.0 NUCLEAR DESIGN The neutronic methods for the design and analysis of the UlC6 reload are described in PP8L topical report PL-NF-87-001-A (Reference 24), ANF

topical reports XN-NF-80-19(A), Vol. 1, and Vol. 1 Supplements 1 and 2 (Reference 25), and ANF letter RAC:058:88 (Reference 26). These methods have been reviewed and approved by the Nuclear Regulatory Commission for application to the Susquehanna SES reloads.

8.1 Fuel Bundle Nuclear Desi n The ANF-5 fuel bundle designs are a 9x9 lattice with two (2) inert (water) rods and 79 fuel rods containing 150 inches of active fuel.

The top six (6) inches of each fuel rod contain natural uranium and the lower 144 inches (enriched zone) of each rod contain enriched uranium at one of ten different enrichments. The ANF-5 reload batch consists of 108 bundles which contain nine burnable poison rods with 4.0 wt% GdzOs (9Gd4) blended with UOz enriched to 3.45 wt% U-235 and 112 bundles which contain ten burnable poison rods with 5.0 wt/

Gdz0s (10Gd5) blended with UOz enriched to 3.25 wt/ U-235. These Gdz0s-UOz rods are utilized to reduce the initial reactivity of the bundle.

The average enrichment of the enriched zone is 3.64 wt% U235 for the lattice containing 9Gd4 and 3.31 wt/ U235 for the lattice containing 10Gd5. The corresponding bundle average enrichments (including the top natural uranium blanket) are 3.52 wt% U235 and 3.21 wt% U235, respectively. The number of fuel rods at each enrichment is given below:

3.64 wt% U235 3.31 wt% U235 Lattice with 9Gd4 Lattice with 10Gd5 Rod Enrichment Rod Enrichment wt% U 35 ¹ of Rods wt% U235 ¹ o Rods 2.00 6 1.80 1 2.70 16 2.00 3 3.45 29 (9 with 2.20 2 4 wt% Gdz0s) 2.52 16

'.25 29 (10 with 5 wt% Gdz0s) 4.50 13 3.66 13 4.91 15 4.50 15

The neutronic design parameters and pin enrichment distributions are described in Section 4.0 of the U1C6 Reload Analysis Report (Reference 3).

8.2 Core Reactivit Shutdown Hargin for U1C6 was analyzed using PPIEL's core physics methods (Reference 24) and assumed a low Cycle 5 exposure of 9266 MWD/HTU, which results in a conservative cold core reactivity condition at Beginning of Cycle 6 (BOC6). Shutdown Margin is defined as the core reactivity with all control rods fully inserted, except for the strongest worth control rod, at 68'F and xenon-free conditions. The minimum value of Shutdown Margin occurs at BOC 6 and is 1.07% dk/k. The cold all-rods-in core k-effective at BOC 6 is 0.96826. The value of R, which is the difference between the BOC Shutdown Margin and the minimum Shutdown Margin during the cycle, is 0.00% hk/k. The calculated Shutdown Margin at any point in the cycle is well in excess of the minimum 0.38% hk/k Technical Specification requirement, and sufficient Shutdown Margin shall be verified by test at BOC 6.

The Standby Liquid Control System, which is designed to inject a quantity of boron that produces a concentration of no less than 660 ppm in the reactor core within approximately 90 to 120 minutes after initiation, was calculated by ANF to provide a margin of shutdown of

2. 11% dk/k with the reactor in a cold, xenon free state, and all control rods at their critical full power positions (Reference 3).

This assures that the reactor can be brought from full power to a cold, xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted. Thus for the Cycle 6 reload core the basis of the Technical Specification requirement is met.

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8.3 Contrast of C cle 6 Core with C cle 5 The core loading strategies For Cycle 5 and 6 are very similar in nature. Cycle 5 utilized a conventional scatter loading with the lowest reactivity bundles placed in the peripheral region of the core. Cycle 6 will also be based on this scatter loading principle.

Fresh reload bundles will be scatter loaded in control cells throughout the-core except on the core periphery. Thrice burned XN-2 bundles and twice burned XN-3 bundles will be utilized on the core periphery. Twice burned XN-3 bundles will also be used to constrain reactivity in interior four bundle cells. The once burned ANF-4 bundles will be distributed throughout the core in a manner which yields acceptable radial peaking and provides adequate cold shutdown margin throughout the cycle.

Briefly reviewing the previous fuel bundle designs, the Cycle 3 XN-2 fuel is the only remaining 8x8 fuel which will be loaded in Cycle 6.

This fuel initially contained 4 wt/ Gdz0s distributed uniformly over the enriched zones of designated rods. The Cycle 4 XN-3 and Cycle 5 ANF-4 fuel initially contained 4 wt/ Gdz0s distributed uniformly over the enriched zones of designated rods. The Cycle 6 ANF-5 fuel bundle designs contain two different enrichment and gadolinia types.

The higher enrichment assembly (3.52 wt/o U235 bundle average enrichment) contains 9 gadolinia bearing .rods at 4 wt/ Gdz0s, while the lower enrichment assembly (3.21 wt1. U235 bundle average enrichment) contains 10 gadolinia bearing rods at 5 wt% Gdz0s. The lower enrichment 10Gd5 assembly was primarily utilized to maintain adequate core shutdown margin, while the higher enrichment 9Gd4 assembly was utilized to obtain the desired Hot Excess Reactivity and cycle energy.

For reload cycles, the axial exposure profile of the exposed bundles provides an axial shaping effect and eliminates the need for axial varying gadolinia in U1C6. Thus, like the XN-2, XN-3 and ANF-4 fuel designs, it is not necessary to include axial varying gadolinia in

the ANF-5 fuel for the purposes of hot operating power shape control. The ANF-5 fuel utilizes an enrichment distribution to yield internal power peaking which results in a balanced and acceptable design relative to MCPR, MAPLHGR, and LHGR Limits. In addition, the XN-2, XN-3, ANF-4,and ANF-5 fuel designs contain a six (6) inch natural uranium section at the top of the fuel bundles in order to increase neutron economy by decreasing leakage at the top of the active core.

8.4 New Fuel Stora e Vault S ent Fuel Pool Criticalit 8.4. 1 New Fuel Storage Vault The original neutronics analysis of the currently installed SSES new fuel storage vault was performed by General Electric Company (GE). GE did not limit the stored fuel to a specific enrichment distribution or burnable poison content, but instead limited the k of the fuel lattice (i.e. the maximum enriched zone of the bundle) to s 1.30. This insures that, under dry or flooded conditions, the new fuel vault k-effective remains below 0.95 as specified in the SSES FSAR.

. Since the GE analysis was for an Bx8 lattice, ANF performed calculations for the new fuel vault assuming a 9x9 lattice.

The results show that 9x9 fuel with a lattice average enrichment s 3.95 wtX U235 and an,ANF calculated k s 1.388 will yield a new fuel vault k-effective s .95 under dry or flooded conditions (Reference 27).

The above mentioned k is calculated for a cold (68'F),

moderated, uncontrolled fuel assembly lattice in reactor geometry at beginning-of-life (BOL). The maximum cold, uncontrolled, BOL k of the two ANF-5 fuel assembly enriched zones, as calculated by ANF is 1. 133. This value is well below the ANF analysis criterion of 1.388. Thus for the

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ANF-5 fuel it is concluded that adequate margin to prevent new fuel vault criticality under dry or flooded conditions exists.

Although the new fuel vault has not been designed to preclude criticality at optimum moderation conditions (between dry and flooded), watertight covers are used, administrative procedures are in place to prevent this condition, and criticality monitors have been installed as an, added precaution.

8.4.2 Spent Fuel Pool The original neutronics analysis for the spent fuel pool as presented in the FSAR was performed by Utility Associates International (UAI). The basis of the analysis assumed the spent fuel pool was loaded with an infinite array of fresh Bx8 fuel assemblies at a uniform average enrichment of 3,25 wt/ U235 containing no burnable poison. The absence of burnable poisons insures that peak assembly reactivity occurs at BOL.

ANF performed an analysis to determine criteria for ANF 9x9 fuel that will ensure that the SSES Spent Fuel Pool k-effective will be s .95 (Reference 28). The resulting criterion is that the enrichment of the maximum enriched zone of a 9x9 assembly be s 3.95 wt'1. U235. The enrichment of the maximum enriched zones of each of the ANF-5 9x9 fuel designs are 3.64 wtX U235 and 3.31 wt'1. U235. These enrichments are less than the 3.95 wtL U235 requirement, and thus it is concluded that adequate margin exists to prevent spent fuel pool criticality throughout the ANF-5 fuel assembly lifetime.

9,0 CORE MONITORING SYSTEH The POWERPLEX core monitoring system will be utilized to monitor reactor parameters during Cycle 6 and for future reload cycles at SSES.

POWERPLEX incorporates ANF's core simulation methodology and is used for both on-line core monitoring and as an off-line predictive and backup tool. POWERPLEX input will be based on either the XFYRE methodology (Reference 25) or the CPH-2/PPL methodology (Reference 24). Both methodologies have been approved by the NRC.

The POWERPLEX system has been operational at SSES and utilized to monitor reactor parameters during Unit 1 Cycles 2, 3, 4, and 5 and Unit 2 Cycles 2, 3 and 4. The POWERPLEX routines are fully consistent with ANF's nuclear analysis methodology (with the exception of CPH-2/PPL input, if used) as described in XN-NF-80-19(A) Volume 1 and Volume 1 Supplement 2 (Reference 25) and use the VHIST13 void history correlation (Reference 26). In addition, the measured power distribution and monitoring related uncertainties are incorporated into the calculation of the HCPR Safety Limit as described in ANF's Nuclear Critical Power Methodology Report XN-NF-524(A) (Reference 18). If CPH-2/PPL is used to generate input to the POWERPLEX routines, the monitoring related uncertainties will be reevaluated based on the use of CPH-2/PPL and must be less than or equal to the current monitoring related uncertainties in order to maintain a HCPR Safety Limit of 1.06.

10.0 ANTICIPATED OPERATIONAL OCCURRENCES In order to determine HCPR operating limits for U1C6 fuel for two loop operation, eight categories of core-wide transients were considered as described in ANF's Plant Transient Methodology Report XN-NF-79-71(P)

(Reference 29). ANF has provided analysis results for the following four core wide transients to determine the thermal margin for UIC6:

1) Generator Load Rejection Without Bypass (LRWOB)

2) Feedwater Controller Failure (FWCF)

3) Loss of Feedwater Heater (LFWH)

4) Recirculation Flow Controller Failure - Increasing Flow As shown in XN-NF-79-71(P) (Reference 29), the other core-wide transients are non-limiting (ice., bounded by one of the above). In addition, two local events, Rod Withdrawal Error and Fuel Loading Error, were analyzed in accordance with the methodology described in XN-NF-80-19(A) Vol. 1 (Reference 25). The results of the core-wide and local transient analyses are provided in the UIC6 Reload Analysis Report ANF-90-050 (Reference 3) and in the UIC6 Plant Transient Analysis Report ANF-90-049 (Reference 4). These documents describe the correspondence between the generic documents listed above and the UIC6 specific cases. The core wide transient hCPRs are calculated with the XCOBRA-T methodology as

'escribed in Reference 30.

At rated power and flow conditions the LRWOB was determined to be the limiting event for UIC6 resulting in a hCPR of 0.28, and when combined with the 1.06 Safety Limit, results in a MCPR operating limit of 1.34.

for U1C6, this limit does not increase at less than rated power.

However, at less than rated flow, the Recirculation Flow Controller Failure is limiting. Therefore, the MCPR operating limit increases at reduced flow conditions.

Analyses at 104% power and 1007. core flow were also performed to determine the MCPR operating limits with either the turbine bypass system inoperable or with the End-of-Cycle Recirculation Pump Trip logic (EOC-RPT) inoperable. The resulting hCPRs are 0.34 for the bypass system inoperable and 0.35 for the EOC-RPT inoperable. Therefore, full power operation with the bypass system inoperable requires a MCPR operating limit of 1.40, and full power operation with the EOC-RPT inoperable requires a MCPR operating limit of 1.41. The results of the FWCF analysis are conservatively interpolated and combined with the above hCPRs to yield the Technical Specification limits for MCPR as a function of core power.

10. 1 Core-Wide Transients The plant transient model used to evaluate the Load Reject Without Bypass (LRWOB) and Feedwater Controller Failure (FWCF) event is ANF's COTRANSA code (Reference 29) which incorporates a one-dimensional neutronics model to account for shifts in axial power shape resulting from rapid pressurization and void collapse, and a multi-node steam line model to accommodate pressure waves in the steam line. The 4CPRs for the LRWOB and FWCF transients are determined by the XCOBRA-T methodology described in Reference 30.

All core-wide transients are analyzed deterministically (i.e., using bounding values as input parameters). These methods were also used to analyze the turbine bypass system inoperable and End-of-Cycle Recirculation Pump Trip function (EOC-RPT) inoperable conditions.

The Loss of Feedwater Heater event (LFWH) was analyzed deterministically with ANF's PTSBWR code (Reference 29) which uses a point-kinetics neutronics model since rapid pressurization and void collapse do not occur for this event. The LFWH event yields a smaller 4CPR than the LRWOB or FWCF.

Technical Specification scram times were used in the pressurization analyses. Therefore, the calculated operating limit HCPR is conservative for scram times less than the Technical Specification scram times, and no scram speed adjustment to the HCPR operating limit is required during operation of SSES Unit 1 Cycle 6.

10.2 Local Transients As shown in ANF-90-050 (Reference 3), the results of the Fuel Loading Error are bounded by the Rod Withdrawal Error (RWE) event and are therefore non-limiting. RWE analyses were performed to support an RBN setpoint of lOP/. The 4CPR for the RWE event with a lOP/ full flow RBH setpoint is 0.26. The RWE event is bounded by the LRWOB event for U1C6 operation.

\

- 20-'

10.3 Reduced Flow Power 0 eration ANF has provided HCPR operating limits for manual flow control reduced flow operation for Cycle 6 in ANF-90-049 (Reference 4).

These values are based on ANF's analysis of the Recirculation Pump Flow Increase event from reduced flow operation during UIC6. The operating limit consists of a plot of MCPR versus core flow.

The FWCF event is analyzed at reduced power conditions using the XCOBRA-T methodology (Reference 30). For Unit 1 Cycle 6, the results of this event at reduced power are bounded by the result of the Generator Load Reject Without Bypass (LRWOB) at rated power/flow conditions and with turbine bypass and RPT systems operable.

Therefore, the HCPR operating limit included in the Technical Specifications is constant as a function of power.

For all power/flow conditions with operable bypass and RPT systems, the U1C6 MCPR operating limit will be the maximum of the flow dependent operating limit taken from the MCPR versus flow curve and the constant operating limit taken from the MCPR versus power curve.

Since the automatic load following capability has been removed from SSES Unit 1, analyses for the automatic flow control mode of operation have not been performed.

10.4 ASM Over ressurization Anal sis In order to demonstrate compliance with the ASHE Code overpressurization criterion of ll& of vessel design pressure (1375 psig), the HSIV closure event with failure of the HSIV position switch scram was analyzed with ANF's COTRANSA code. The U1C6 analysis assumes six safety relief valves are out of service and an HSIV closure time of 2.0 seconds. The current Technical Specification HSIV minimum closure time is 3.0 seconds. PP8L elected to use a 2.0 second closure time to allow for a revision of

the SSES Unit 1 MSIV closure time Technical Specification at a future date. A previous ANF analysis of HSIV closure with direct scram shows that this faster closure time does not make the HSIV closure a limiting MCPR event. The maximum calculated pressure from the analysis (at the vessel bottom) is 1312 psig which is within the 110% design criterion (1375 psig).

The calculated steam dome pressure corresponding to the 1312 psig peak vessel pressure is 1295 psig, for a vessel differential pressure of 17 psi. This includes the effects of the ATWS RPT which is assumed to initiate at a pressure setpoint of 1170 psig. The current Technical Specification Safety Limit of 1325 psig is based on dome pressure and therefore conservatively assumes a 50 psi vessel differential pressure (1375-1325). Because the calculated vessel differential pressure is 17 psi, the choice of 1325 psig assures compliance with the ASHE criterion of 1375 psig peak vessel pressure while also maintaining consistency with the U1C6 pressure safety limit.

11.0 POSTULATED ACCIDENTS ANF analyzed two types of accidents for two loop operation of Unit 1 Cycle 6: the Loss of Coolant Accident (LOCA) and the Control Rod Drop Accident (CRDA). ANF has previously analyzed the Loss-of-Coolant Accident (LOCA) to determine the HAPLHGR limits for the Unit 1 Cycle 4 XN-3 9x9 fuel. This MAPLHGR limit is also applicable to the U1C5 ANF-4 fuel and the UIC6 ANF-5 fuel as demonstrated by heatup analyses (References 41 and 3). ANF performed the Control Rod Drop Accident (CRDA) analysis to demonstrate compliance with the 280 cal/gm Design Limit. The results of these analyses are presented in Section 6.0 of ANF-90-050 (Reference 3). ANF's methodology for the CRDA analysis is described in XN-NF-80-19(A) Vol. 1 (Reference 25) and for the LOCA analysis is provided in References 31 thru 33.

11.1 Loss-of-Coolant Accident XN-NF-84-117(P) (Reference 34) describes ANF's generic jet pump BWR-4 LOCA break spectrum analysis. This determined the limiting break for BWR-4's with modified Low Pressure Coolant Injection logic to be a double-ended guillotine break in the recirculation piping on the discharge side of the pumps. The discharge coefficient assumed was 0.4, which is equivalent to a total break area of 2.8 ft . The analysis of this event for SSES 9x9 fuel is provided in XN-NF-86-65 (Reference 35). ANF-90-050 (Reference 3) confirms that the MAPLHGR limits in XN-NF-86-65 ensure that the peak cladding temperature (PCT) for the UlC6 ANF-5 fuel remains below 2200'F, local Zr-HzO reaction remains below 17%, and core-wide hydrogen production remains below 1% for the limiting LOCA event as required by 10CFR50, The limiting operating condition was identified in XN-NF-86-65 as the highest power and highest flow permitted by the operating map.

The results reported in ANF-90-050 (Reference 3) are bounding for reactor operating conditions up to 100% rated power and 100% rated flow and assure acceptable Peak Cladding Temperatures for the ANF-5 fuel during a postulated LOCA event. The LOCA analysis of XN-NF 65 (Reference 35) was performed for an entire core of 9x9 fuel and therefore provides MAPLHGR limits for ANF 9x9 fuel only.

As discussed in Sections 7.0 and 8.0, the coresident ANF 8x8 and 9x9 fuel are hydraulically and neutronically compatible. Therefore, ANF LOCA analysis and MAPLHGR limits for the ANF 8x8 fuel remain applicable during UlC6 operation. In addition, in order to support an 8x8 fuel assembly discharge exposure of 37,000 MWD/MTU for U1C6, ANF performed an additional LOCA heatup calculation (MAPLHGR evaluation) at 40,000 MWD/MTU for the 8x8 fuel (Reference 3). Since the 8x8 fuel assembly discharge exposure limit is 37,000 MWD/MTU, the Technical Specification MAPLHGR limits for 8x8 fuel have been extended out to this exposure for UlC6.

11.2 Control Rod Dro Accident ANF's methodology for analyzing the Control Rod Drop Accident (CRDA) is described in XN-NF-80-19(A) Vol. 1 (Reference 31) and utilizes a generic parametric analysis which calculates the fuel enthalpy rise during postulated CRDAs over a wide range of reactor operating variables. The UIC6 analysis was performed using bounding assumptions similar to those used in the U1C5 analysis presented in Reference 41. For U1C6, Section 6.2 of ANF-90-050 (Reference 3) shows a value of 205 cal/gm for the maximum fuel rod enthalpy and less than 600 fuel rods exceeding 170 cal/gm during the worst case postulated CRDA. The 205 cal/gm value is well below the design limit of 280 cal/gm and less than 600 fuel rods exceeding 170 cal/gm is bounded by the 770 rods assumed in Section 15.4.9 of the SSES FSAR (Reference 36). To ensure compliance with the CRDA analysis assumptions, control rod sequencing below 20/ core thermal power must comply with GE's Banked Position Withdrawal Sequence.

constraints (Reference 37).

12.0 SINGLE LOOP OPERATION To support single loop operation for U1C6, ANF analyzed the MCPR Safety Limit considering single loop operation power/flow conditions and associated single loop operation uncertainties. The results presented in ANF-90-049, Appendix A (Reference 4) show that the HCPR Safety Limit must be increased by 0.01 when in single loop operation. The 0.01 increase in the Safety Limit is a result of the increased measurement uncertainties associated with single loop operation.

ANF performed a review of the limiting anticipated operational occurrences for single loop operation (Reference 4). Former analyses (Reference 38 and 39) have indicated that other events which could be affected by single loop operation were non-limiting when analyzed under single loop operating conditions. Under single loop operating conditions, steady state operation can not exceed approximately 76/ power 0

and 6N'ore flow because of the capability of the operating recirculation pump. Thus, it was determined that when operating at low power/flow conditions, the two loop events remain limiting. The two loop HCPR operating limits plus 0.01 conservatively protect the fuel from any transient in single loop operation.

It was determined that the single loop operation LOCA analysis presented in XN-NF-86-125 (Reference 40) is bounded by the two loop LOCA event. In addition, ANF analyzed the pump seizure accident from single loop operating conditions on a generic basis for the Susquehanna Units (Reference 4). The results of the generic analysis 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 operation at the single loop operating HCPR limit, the radiological consequences of a pump seizure accident from single loop operating conditions are but a small fraction of the 10CFR100 guidelines. Previous analyses (Reference 38) have shown that other accidents which could be affected by single loop operation were non-limiting when analyzed under single loop operating conditions.

Based on the vessel internal vibration analysis performed by GE, the 80%

recirculation pump speed restriction, previously discussed in Reference 38, is maintained for U1C6 single loop operation.

The results discussed previously in Section 7.4 on core stability also apply under single loop operating conditions. One of the stability tests performed during the startup of Susquehanna SES Unit 2 Cycle 2 was performed under single loop operating conditions. The measured decay ratio was 0.30 (a 0.064) at 55% power/44'X flow. ANF performed an analysis of these tests with their COTRAN computer code and calculated a decay ratio of 0.29. This data, the stability calculation results presented in ANF-90-050 (Reference 3), and the U1C6 Technical Specifications which comply with NRC Bulletin 88-07, Supplement 1 support single loop operation during UIC6.

0 REFERENCES

1. PLA-3209, "Proposed Amendment 24 to License No. NPF-22: Unit 2 Cycle 4 Reload," Letter from H. W. Keiser (PP&L) to W. R. Butler (NRC), June 16, 1989.

2. Letter from Mohan C. Thadani (NRC) to H. W. Keiser (PP&L), "Technical Specification Changes to Support Cycle 4 Operation (TAC No. 73588)

Susquehanna Steam Electric Station, Unit 2", November 3, 1989.

3. ANF-90-050, "Susquehanna Unit 1 Cycle 6 Reload Analysis Design and Safety Analyses," Advanced Nuclear Fuels Corporation, May 1990.

4. ANF-90-049, "Susquehanna Unit 1 Cycle 6 Plant Transient Analysis,"

Advanced Nuclear Fuels Corporation, May 1990.

5. ANF-90-018(P), Revision 1, "Susquehanna Unit 1 Bx8 Extended Burnup Design Report," Advanced Nuclear Fuels Corporation, June 1990.

6. XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, Inc., June 1986.

7. Letter from W. R. Butler (NRC) to H. W. Keiser (PP&L), "Technical Specification Changes Supporting Unit 1 Cycle 5 Operation (TAC No.

72094)," May 15, 1989.

8. PLA-623, "IE Bulletin 79-26 Revision 1," Letter from N. W. Curtis (PP&L) to B. H. Grier (NRC), February 11, 1981.

9. NEDE-22290-A, Supplement 1, "Safety Evaluation of the General Electric Hybrid I Control Rod Assembly for The BWR 4/5 C Lattice," July 1985.

10. NEDE-22290-A, Supplement 3, "Safety Evaluation of the General Electric Duralife 230 Control Rod Assembly," May 1988.

11. XN-NF-85-67(P)(A), Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, Inc., September 1986.

12. XN-NF-84-97, Revision 0, "LOCA-Seismic Structural Response of an ENC 9x9 Jet Pump Fuel Assembly," Exxon Nuclear Company, Inc., December 1984.

13. PLA-2728, "Response to NRC guestion: Seismic/LOCA Analysis of U2C2 Reload," Letter from H. W. Keiser (PP8L) to E. Adensam (NRC), September 25, 1986.

14. XN-NF-82-06(P)(A), Supplement 1, Revision 2, "gualification of Exxon Nuclear Fuel for Extended Burnup Supplement 1 Extended Burnup gualification of ENC 9x9 Fuel," Hay 1988.

15. PLA-2585, "Proposed Amendement 78 to License No. NPF-14," Letter from H.

W. Keiser (PP8L) to E. Adensam (NRC), January 16, 1986.

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

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

18. XN-NF-524(A), Revision 1, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc., November 1983.

19. NRCB 88-07, Supplement 1, "Power Oscillations in Boiling Water Reactors (BWRs)" USNRC Bulletin, December 30, 1988.

20. XN-NF-86-90, Supplement 1, "Susquehanna Unit 2 Cycle 2 Stability Test Results," Exxon Nuclear Company, Inc., January 1987.

21. PLA-3344, "Unit 2/Cycle 4 Stability Data," Letter from H. W. Keiser (PPKL) to W. R. Butler (NRC), February 28, 1990.

22. PLA-3154, "Proposed Amendment 122 to License No. NPF-14: Unit 1 Stability," Letter from H. W. Keiser (PP&L) to W. R. Butler (NRC), April 12, 1989.

23. Letter from M. C. Thadani (NRC) to H. W. Keiser (PPL), "Technical Specification Changes Reflecting Thermal Hydraulic Stability Considerations (TAC No. 73056)," November 27, 1989.

24, PL-NF-87-001-A, "gualification of Steady State Core Physics Methods for BWR Design and Analysis," April 28, 1988.

25. XN-NF-80-19(A), Volume 1, and Volume 1 Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors: Neutronic Methods for Design and Analysis," Exxon Nuclear Company, Inc., March 1983.

26. RAC:058:88, "Void History Correlation," Letter from R. A. Copeland (ANF) to M. W. Hodges (NRC), September 13, 1988.

27. XN-NF-86-44, Revision 1, "Criticality Safety Analysis Susquehanna New Fuel Storage Vault with Exxon Nuclear Company, Inc. 9x9 Reload Fuel," Exxon Nuclear Company, Inc., May 1986.

28. XN-NF-86-45, Revision 1, "Criticality Safety Analysis Susquehanna Spent Fuel Storage Pool with Exxon Nuclear Company, Inc. 9x9 Reload Fuel," Exxon Nuclear Company, Inc., May 1986.

29. XN-NF-79-71(P), Revision 2, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc., November 1981.

30. XN-NF-84-105(A), Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Advanced Nuclear Fuels Corporation, February 1987.

31. XN-NF-80-19(A), Volumes. 2, 2A, 2B, and 2C, "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, Inc., September 1982.

32. XN-NF-CC-33(A), Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Company, Inc.,

November 1975.

33. XN-NF-82-07(A), Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., November 1982.

34. XN-NF-84-117(P), "Generic LOCA Break Spectrum Analysis: BWR 3 and 4 with Modified Low Pressure Coolant Injection Logic," Exxon Nuclear Company,

'nc., December 1984.

35. XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel," Exxon Nuclear Company, Inc., May 1986.

36, Susquehanna Steam Electric Station, Units 1 and 2, Final Safety Analysis Report.

37. NED0-21231, "Banked Position Withdrawal Sequence," General Electric Company, January 1977.

38, PLA-2885, "Proposed Amendment 52 to License No. NPF-22," Letter from H. W.

Keiser (PP&L) to W. R. Butler (NRC), June 30, 1987.

39. PLA-2935, "Additional Information on Proposed Amendment 52 to License No.

NPF-22," October 30, 1987.

40. XN-NF-86-125, "Susquehanna LOCA Analysis for Single Loop Operation Analysis," Exxon Nuclear Company, Inc., November 1986.

1 41, PLA-3141, "Proposed Amendment 119 to License No. NPF-14: Unit 1 Cycle 5 Reload," Letter from H. W. Keiser (PPIEL) to M. R. Butler (NRC),

February 2, 1989.

rp0192i.kpr:el 3

ML18026A235

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