Sincerely, /s/ Victor Nerses, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects - 1/11. Office of Nuclear Reactor Regulation

Transcription

1 Mr. Robert G. Byram Senior Vice President-Generation and Chief Nuclear Officer Pennsylvania Power and Light Company 2 North Ninth Street Allentown, PA Ma" 4, 1998 SUBJECT: SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 Dear Mr. Byram: (TAC NOS. M97365 AND M97366) The Commission has issued the enclosed Amendment No. 176 to Facility Operating License No. NPF-14 and Amendment No. 149 to Facility Operating License No. NPF-22 for the Susquehanna Steam Electric Station (SSES), Units I and 2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 27, 1996, as supplemented by letter dated February 12, These amendments modify the SSES, Units I and 2, TSs by modifying the Rod Block Monitor (RBM) flow biased trip setpoints and also the RBM channel calibration frequency and allowed outage times. A copy of our safety evaluation is also enclosed. Notice of Issuance will be published in the Federal Register. Sincerely, /s/ Docket Nos / Victor Nerses, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Enclosures: 1. Amendment No. 176 to License No. NPF Amendment No. 149 to License No. NPF Safety Evaluation 4. Notice of Issuance ' ' I cc w/encls: See next page DISTRIBUTION Docket File PUBLIC DlI-_3 DI andinn JZwolinski JStolz MO'Brien JMarea~c OGC GHiII(4) GGolub WR rpcn.r ACRS CAnderson, RGN-I THarris ( SE) OFFICE P -2/PR I PDI OGC SRXB J PD1-2/D NAME VNerss:rb M, W' lf l14? SE input dtd RCapra w DATE 4/6/98 g//98 "Y///"98 11/26/97 5/ 4/98 OFFICIAL RECORD COPY - DOCUMENT NAME: SU97365.AMD PDR ADOCK P PDR.1 I i t I K

2 N1.- UNITED STATES 0 NUCLEAR REGULATORY COMMISSION It WASHINGTON, D.C May 4, 1998 Mr. Robert G. Byram Senior Vice President-Generation and Chief Nuclear Officer Pennsylvania Power and Light Company 2 North Ninth Street Allentown, PA SUBJECT: SUSQUEHANNA STEAM ELECTRIC STATION, UNITS I AND 2 (TAC NOS. M97365 AND M97366) Dear Mr. Byram: The Commission has issued the enclosed Amendment No. 176 to Facility Operating License No. NPF-14 and Amendment No. 149 to Facility Operating License No. NPF-22 for the Susquehanna Steam Electric Station (SSES), Units I and 2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 27, 1996, as supplemented by letter dated February 12, These amendments modify the SSES, Units I and 2, TSs by modifying the Rod Block Monitor (RBM) flow biased trip setpoints and also the RBM channel calibration frequency and allowed outage times. A copy of our safety evaluation is also enclosed. Notice of Issuance will be published in the Federal Register. Sincerely, Docket Nos / Enclosures: 1. Amendment No.176 to License No. NPF Amendment No.149 to License No. NPF Safety Evaluation 4. Notice of Issuance cc w/encls: See next page Victor Nerses, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects - IIl Office of Nuclear Reactor Regulation

3 Mr. Robert G. Byram Pennsylvania Power & Light Company Susquehanna Steam Electric Station, Units 1 & 2 cc: Jay Silberg, Esq. Shaw, Pittman, Potts & Trowbridge 2300 N Street N.W. Washington, D.C Bryan A. Snapp, Esq. Assistant Corporate Counsel Pennsylvania Power & Light Company 2 North Ninth Street Allentown, Pennsylvania Licensing Group Supervisor Pennsylvania Power & Light Company 2 North Ninth Street Allentown, Pennsylvania Senior Resident Inspector U. S. Nuclear Regulatory Commission P.O. Box 35 Berwick, Pennsylvania Director-Bureau of Radiation Protection Pennsylvania Department of Environmental Resources P. 0. Box 8469 Harrisburg, Pennsylvania Mr. Jesse C. Tilton, III Allegheny Elec. Cooperative, Inc. 212 Locust Street P.O. Box 1266 Harrisburg, Pennsylvania Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania General Manager Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania Mr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company Rural Route 1, Box 1797 Berwick, Pennsylvania George T. Jones Vice President-Nuclear Operations Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania Dr. Judith Johnsrud National Energy Committee Sierra Club 433 Orlando Avenue State College, PA Chairman Board of Supervisors 738 East Third Street Berwick, PA 18603

4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C PENNSYLVANIA POWER & LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC. DOCKET NO SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 176 License No. NPF The Nuclear Regulatory Commission (the Commission or the NRC) having found that: A. The application for the amendment filed by the Pennsylvania Power & Light Company, dated November 27, 1996, as supplemented by letter dated February 12, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied PDR ADOCK P PDR

5 -2-2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows: (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 176 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and is to be implemented within 30 days after its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Attachment: Changes to the Technical Specifications Date of Issuance: May 4, 1998 Robert A. Capra, Director Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

6 ATTACHMENT TO LICENSE AMENDMENT NO. 176 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO Replace the following pages of the Appendix A Technical Specifications with enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. REMOVE 3/ / / /4 4-1c B 3/4 1-4 INSERT 3/ / / /4 4-1c B 3/4 1-4

7 REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION Both rod block monitor (RBM) channels shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER. ACTION: a. With one RBM channel inoperable, restore the inoperable RBM channel to OPERABLE status within 7 days and verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN; otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour. b. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within 48 hours. SURVEILLANCE REQUIREMENTS Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a: a. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and for the OPERATIONAL CONDITIONS specified in Table b. CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when the reactor is operating on a LIMITING CONTROL ROD PATTERN. SUSQUEHANNA - UNIT 1 3/ Amendment No

8 C,) C C,) 0 C m z z C z CA) CA) CL 3 (I-I z P *-. TABLE CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1. ROD BLOCK MONITOR I a. Upscale # # _! 0.58W + 52% 0.58W + 55% b. Inoperative NA NA c. Downscale -e 5/125 divisions of full scale >_ 3/125 of divisions full scale 2. APRM a. Flow Biased Neutron Flux Upscale # # 1) Flow Biased 0.58 W + 50% < 0.58 W + 53% 2) High Flow Clamped 108% of RATED THERMAL POWER _< 111% of RATED THERMAL POWER b. Inoperative NA NA c. Downscale 2:5% of RATED THERMAL POWER > 3% -of RATED THERMAL POWER d. Neutron Flux-Upscale Startup 12% of RATED THERMAL POWER < 14% of RATED THERMAL POWER 3. SOURCE RANGE MONITORS a. Detector not full in NA NA b. Upscale < 2 x 105 cps < 4 x 10 5 cps c. Inoperative NA NA, d. Downscale > 0.7 cps** >_ 0.5 cps** 4. INTERMEDIATE RANGE MONITORS a. Detector not full in NA NA b. Upscale _< 108/125 divisions of full scale < 110/125 divisions of full scale c. Inoperative NA NA d. Downscale >5/125 divisions of full scale >3/125 divisions of full scale 5. SCRAM DISCHARGE VOLUME a. Water Level-High 44 gallons < 44 gallons 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW a. Upscale < 114/125 divisions of full scale < 117/125 divisions of full scale b. Inoperative NA NA c. Comparator < 10% flow deviation < 11% flow deviation The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification ** Provided signal-to-noise ratio is > 2. Otherwise, 3 cps as trip setpoint and 2.8 cps for allowable value. et See Specification a for single loop operation requirements. ( (

9 TABLE CO C C', 0 C m z C => z I CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL OPERATIONAL TRIP FUNCTION CHANNEL CHECK FUNCTIONAL TEST CALIBRATION(a) CONDITIONS FOR WHICH SURVEILLANCE REQUIRED 1. ROD BLOCK MONITOR a. Upscale NA Q R 1" b. Inoperative NA Q NA 1" c. Downscale NA Q R 1* 2. APRM a. Flow Biased Neutron Flux - Upscale b. Inoperative S Q SA 1 c. Downscale NA Q NA 1,2,5** d. Neutron Flux - Upscale, Startup S Q SA 1 S Q SA 2,5*** ( ca C. 3. SOURCE RANGE MONITORS a. Detector not full in NA S/UtJ,W NA 2,5 b. Upscale NA SIU(b),W Q 2,5 c. Inoperative NA S/U(b),W NA 2,5 d. Downscale NA S/U(b),W Q 2,5 CD) 4. INTERMEDIATE RANGE MONITORS a. Detector not full in NA S/U >,W NA 2,5 3 (D a' Z :3 2 "tq o. 31 k ON b. Upscale S S/U(b),W Q 2,5 c. Inoperative NA S/U(b),W NA 2,5 d. Downscale S S/U( 0 ),W Q 2,5 5. SCRAM DISCHARGE VOLUME a. Water Level-High NA Q R 1,2,5** 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW a. Upscale NA Q 0 1 b. Inoperative NA Q NA I c. Comparator NA Q Q 1

10 REACTOR COOLANT SYSTEM RECIRCULATION LOOPS-SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION One reactor coolant recirculation loop shall be in operation with the pump speed 80% of the rated pump speed and the reactor at a THERMAL POWER/core flow condition outside of Regions I and II of Figure , and a. the following revised specification limits shall be followed: 1. Specification 2.1.2: the MCPR Safety Limit shall be increased to the value.shown in Figure ' Table : the APRM Flow-Biased Scram Trip Setpoints shall be as follows: ITrip Setpoint Allowable Value 50.58W + 54% < 0.58W + 57% 3. Specification 3.2.2: the APRM Setpoints shall be as follows: Trip Setpoint Allowable Value S < (0.58W + 54%) T. S. (0.58W + 57%) T SRB < (0.58W + 45%) T SRB < (0.58W + 48%) T 4. Specification 3.2.3: The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the applicable Single Loop Operation MCPR limit as specified in the CORE OPERATING LIMITS REPORT. 5. Specification 3.2.4: The LINEAR HEAT GENERATION RATE (LHGR) shall be less than or equal to the applicable Single Loop Operation LHGR limit as specified in the CORE OPERATING LIMITS REPORT. 6. Table : the RBM/APRM Control Rod Block Setpoints shall be as follows: a. RBM - Upscale b. APRM-Flow Biased *Trip Setpofntf Allowable Value _ 0.58W + 47% 0.58W + 50% Trip Setpoiint _:..Allowable Value < 0.58W + 45% < 0.58W + 48% I APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*+, except during two loop operation.# ++ Only applicable for Unit 1 Cycle 11 operation. Controls to preclude single loop operation shall be maintained as stated in PP&L letter PLA-4872, dated March 19, SUSQ UEHANNA - UNIT c Amendment No. 176

11 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.4 CONTROL ROD PROGRAM CONTROLS (Continued) 280 cal/gm design limit to demonstrate compliance for each operating cycle. If cycle-specific values of the above parameters are outside the range assumed in the parametric analyses, an extension of the analysis or a cycle-specific analysis may be required. Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are referenced in Specification The RBM is designed to automatically block erroneous rod withdrawal at power. However, its operation is not required to prevent fuel damage as a result of such an event. Two channels are provided. This system backs up the written sequence used by the operator for withdrawal of control rods. 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted. To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes. A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown requirement. There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing. The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41.2 gpm. The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel. The temperature requirement for the sodium pentaborate solution is necessary to ensure that the sodium pentaborate remains in solution. With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable. Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours assures that the solution is available for use. Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges. SUSQUEHANNA - UNIT I B 3/4 1-4 Amendment No. 110, Mii, 176

12 . UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C PENNSYLVANIA POWER & LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC. DOCKET NO SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 149 License No. NPF The Nuclear Regulatory Commission (the Commission or the NRC) having found that: A. The application for the amendment filed by the Pennsylvania Power & Light Company, dated November 27, 1996, as supplemented by letter dated February 12, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

13 -2-2. Accordingly, the license is amended by changes to thetechnical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-22 is hereby amended to read as follows: (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.149 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and is to be implemented within 30 days after its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Attachment: Changes to the Technical Specifications Date of Issuance: May 4, 1998 Robert A. Capra, Director Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

14 ATTACHMENT TO LICENSE AMENDMENT NO. i4a FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO Replace the following pages of the Appendix A Technical Specifications with enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. REMOVE / / /4 4-1c B INSERT 3/ / / /4 4-1c B 3/4 1-4

15 REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION Both rod block monitor (RBM) channels shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER. ACTION: a. With one RBM channel inoperable, restore the inoperable RBM channel to OPERABLE status within 7 days and verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN; otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour. b. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within 48 hours. SURVEILLANCE REQUIREMENTS Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a: a. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and for the OPERATIONAL CONDITIONS specified in Table b. CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when the reactor is operating on a LIMITING CONTROL ROD PATTERN. SUSQUEHANNA - UNIT 2 3/ Amendment No.1 49

16 0) C M m z z C -4 z a. CD Z a I TABLE CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1. ROD BLOCK MONITOR a. Upscale # # _ 0.58W + 52%!< 0.58W + 55% b. Inoperative NA NA c. Downscale _5/125 divisions of full scale > 3/125 of divisions full scale 2. APRM a. Flow Biased Neutron Flux Upscale # # 1) Flow Biased W + 50% _< 0.58 W + 53% 2) High Flow Clamped _ 108% of RATED THERMAL POWER 111% of RATED THERMAL POWER b. Inoperative NA NA. c. Downscale 2! 5% of RATED THERMAL POWER 2:3% of RATED THERMAL POWER d. Neutron Flux-Upscale Startup 12% of RATED THERMAL POWER _< 14% of RATED THERMAL POWER 3. SOURCE RANGE MONITORS a. Detector not full in NA NA b. Upscale _52x 10 5 cps :54x 105 cps c. Inoperative NA NA d. Downscale >_ 0.7 cps** _ 0.5 cps** 4. INTERMEDIATE RANGE MONITORS a. Detector not full in NA NA b. Upscale < 108/125 division of full scale < 110/125 division of full scale c. Inoperative NA NA. d. Downscale _5/125 division of full scale Ž3/125 divisions of full scale 5. SCRAM DISCHARGE VOLUME a. Water Level-High 44 gallons _< 44 gallons 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW a. Upscale < 114/125 divisions of full scale < 117/125 divisions of full scale b. Inoperative NA NA c. Comparator _< 10% flow deviation _< 11% flow deviation * The Average Power Range Monitor rod block function is varied as a function of recirculation loop'flow (W). The trip setting of this function must be maintained in accordance with Specification ** Provided signal-to-noise ratio is;> 2. Otherwise, 3 cps as trip setpoint and 2.8 cps for allowable value. ## See Specification a for single loop operation requirements. ( (

17 TABLE Cn C Cn 0 C: m IT z z I~% ca --I M CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP FUNCTION CHANNEL CHECK CHANNEL FUNCTIONAL CHANNEL OPERATIONAL TEST CALIBRATION(a) CONDITIONS FOR WHICH SURVEILLANCE jrequired 1. ROD BLOCK MONITOR a. Upscale NA Q R 1" b. Inoperative NA Q NA 1" c. Downscale NA Q R 1" 2. APRM a. Flow Biased Neutron Flux Upscale S Q SA 1 b. Inoperative NA Q NA 1,2,5"* c. Downscale S Q SA 1 d. Neutron Flux - Upscale, S Q SA Startup I ( Cl' C,' 3. SOURCE RANGE MONITORS a. Detector not full in NA S/U(b),W NA 2,5 b. Upscale NA SU(b),W SA 2,5 c. Inoperative NA S/U(b),W NA 2,5 d. Downscale NA S/U(b),W SA 2,5 4. INTERMEDIATE RANGE MONITORS a. Detector not full in NA S/U(b),W NA 2,5 b. Upscale S SIu(b),w SA 2,5 C> (D) 0. :3 (I, z k o0 c. Inoperative NA S/U(b),W NA 2,5 5. d. Downscale S S/U(b),W SA 2,5 SCRAM DISCHARGE VOLUME a. Water Level-High NA Q R 1,2,5** 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW a. Upscale NA Q Q 1 b. Inoperative NA Q NA 1 c. Comparator NA Q 0 1

18 -REACTOR COOLANT SYSTEM RECIRCULATION LOOPS-SINGLE LOOP OPERATION LIMITING CONDITION FOR.OPERATION One reactor coolant recirculation loop shall be in operation with the pump speed :9 80% of the rated pump speed and the reactor at a THERMAL POWER/core flow condition outside of Regions l and 11 of Figure , and a. the following revised specification limits shall be followed: 1. Specification 2.1.2: the MCPR Safety Limit shall be increased to the value shown in Figure *. 2. Table : the APRM Flow-Biased Scram Trip Setpoints shall be as follows: ITrip Setpoirit Allowable Value 0.58W + 54% :50.58W + 57% 3. Specification 3.2.2: the APRM Setpoints shall be as follows: Trip Setpoint. Allowable Value S:5 (0.58W + 54%) T S - (0.58W + 57%) T SRB:5 (0.58W + 45%) T SRB (0.58W + 48%) T 4. Specification 3.2.3: The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the applicable Single Loop Operation MCPR limit as specified in the CORE OPERATING LIMITS REPORT. 5. Specification 3.2.4: The LINEAR HEAT GENERATION RATE (LHGR) shall be less than or equal to the applicable Single Loop Operation LHGR limit as specified in the CORE OPERATING LIMITS REPORT. 8. Table : the RBMIAPRM Control Rod Block Setpoints shall be as follows: a. RBM - Upscale b. APRM-Flow Biased Trip Setpoint Allowable Value 0.58W + 47% 0.58W +50% Trip Setpoint Allowable Value 0,58W + 45% 0.58W + 48% APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*+, except during two loop operation.# ACTION: a. In OPERATIONAL CONDITION 1: 1. With a) no reactor coolant system recirculation loops in operation, or b) Region I of Figure entered, or c) 'Region 11 of Figure entered and core thermal hydraulic instability occurring as evidenced by: ++Only applicable for Unit 2 Cycle 9 operation. SUSQUEHANNA - UNIT 2 3/4 4-I c Amendment No. 01, ~ ~ 149

19 REACTIVITY CONTROL SYSTEMS BASES CONTROL ROD PROGRAM CONTROLS (Continued) The RSCS and RWM logic automatically initiates at the low power setpoint (20% of RATED THERMAL POWER) to provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted. Parametric Control Rod Drop Accident analyses have shown that for a wide range of key reactor parameters (which envelope the operating ranges of these variables), the fuel enthalpy rise during a postulated control rod drop accident remains considerably lower than the 280 cal/gm limit. For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effective delayed neutron fraction, and maximum four-bundle local peaking factor are compared with the inputs to the parametric analyses to determine the peak fuel rod enthalpy rise. This value is then compared against the 280 cal/gm design limit to demonstrate compliance for each operating cycle. If cycle-specific values of the above parameters are outside the range assumed in the parametric analyses, an extension of the analysis or a cycle-specific analysis may be required. Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are referenced in Specification The RBM is designed to automatically block erroneous rod withdrawal at power. However, its operation is not required to prevent fuel damage as a result of such an event. Two channels are provided. This system backs up the written sequence used by the operator for withdrawal of control rods. 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted. To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes. A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown requirement. There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing. The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41.2 gpm. The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel. The temperature requirement for the sodium pentaborate solution is necessary to ensure that the sodium pentaborate remains in solution. With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable. SUSQUEHANNA- UNIT 2 B 3/4 1-4 Amendment No. 01, 0%, 149

20 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 176 TO FACILITY OPERATING LICENSE NO. NPF-14 AMENDMENT NO. 149 TO FACILITY OPERATING LICENSE NO. NPF-22 PENNSYLVANIA POWER & LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC. SUSQUEHANNA STEAM ELECTRIC STATION, UNITS I AND 2 DOCKET NOS AND INTRODUCTION By letter dated November 27, 1996, as supplemented by letter dated February 12, 1997, the Pennsylvania Power and Light Company (PP&L, the licensee) submitted a request for changes to the Susquehanna Steam Electric Station (SSES), Units I and 2, Technical Specifications (TSs). The requested changes modify the Rod Block Monitor (RBM) flow biased trip setpoint and also modify RBM channel calibration frequency and allowed outage times (AOT). Currently, PP&L does not take credit for the RBM system in the Rod Withdrawal Error (RWE) transient event. The supplemental letter dated February 12, 1997 providing clarifying information that did not change the scope of the November 27, 1996 application. The RBM is designed to prevent violation of the safety limit minimum critical power ratio (SLMCPR) and the cladding 1% plastic strain fuel design limit that may result from a single control RWE event. The RWE analysis described in topical report PL-NF A (reference 1) assumes that the RBM automatically actuates to stop control rod motion during the RWE event before any fuel design limits are exceeded. RBM operability is also assumed in the RWE analysis evaluated in Section of the SSES Final Safety Analysis Report (FSAR). In 1993, SSES experienced unplanned control rod drift events where control rods failed to settle properly in the 00 position following a scram. These events only involved limited rod movement. The staff notes that SSES has not experienced a full rod drift withdrawal. As part of PP&L's compensatory actions, PP&L used the approved licensing RWE methodology to evaluate the effect of control rod drift events on the SLMCPR (reference 2). PP&L now bases the RWE event on the control rod drift events in each reload analysis, and has committed to the use of the rod drift event in future cycles in Supplement 2 of PL-NF A (reference 3), stating that "PP&L does not take credit for the Rod Block Monitor, but assumes a complete withdrawal of the control rod for the (rod drift) event" PDR ADOCK P PDR

21 EVALUATION The licensee proposed to remove the RBM TS requirements and bases from the SSES TSs. PP&L analyzed the rod drift event using approved licensing RWE methodology (references I and 3), which requires the following input assumptions: 1. The RWE event is initiated at rated power and flow. 2. The control rod pattem: a) forces the limiting core MCPR location to be within one control cell of the error rod; b) forces the fuel bundles near their MCPR and LHGR operating limits; and c) results in the core being near critical. 3. The error rod is the maximum worth rod and is within a centrally-defined region of the core. 4. The error rod is initially fully inserted. 5. Zero xenon is assumed as the xenon concentration for 100% power rod line cases. 6. Deep control rods are inserted quarter-core symmetrically. During a conference call with the licensee on July 3, 1996, PP&L stated that the RWE event is now based on a rod drift event in which a control rod drifts from position 00 to 48, e.g. fully inserted to fully withdrawn (references 2 and 4). In this event, the RBM cannot stop control rod motion and, therefore, no credit is taken for the RBM system. The licensee also confirmed that even though the RWE is now based on the rod drift event, the above assumptions remain the same. As part of the approach for the RWE analysis, PP&L calculates the effect of the control rod withdrawal from position 00 to 48. The methodology approved in reference 1 for the RWE takes credit for the RBM high alarm setpoint. Due to the control rod drift events of 1993, the licensee now conservatively evaluates the RWE based on rod drift events without taking credit for the RBM high alarm setpoint. The results of this analysis are operating limits that prevent fuel damage from an RWE in which control rod motion is not stopped by the RBM. The results of the RWE based on the rod drift event are documented in the Reload Summary Report which is submitted for each fuel cycle. The modification of the RBM setpoint flow-biased setpoint curves in the TSs is only acceptable as long as the RWE analysis is based on the rod drift event. PP&L implemented a system to perform the RWE as a rod drift event in reference 3, and must notify the staff if the methodology changes. The licensee would also be required to submit appropriate TS amendments for adequate RBM setpoints if the RWE is again analyzed using the conventional NRC-approved methodology. Based on the information above, the modification of the RBM system flow-biased setpoints in the SSES TSs and bases is acceptable since the RBM is not required for the protection of the MCPR safety limit and the cladding 1% plastic strain fuel design limit.

22 SUMMARY The staff has reviewed PP&Us request to modify the RBM system flow-biased setpoints contained in SSES TSs. Since each reload analysis currently evaluates the RWE based on the rod drift event and does not take credit for the RBM system, modification of the RBM system flow-biased setpoints is acceptable. However, this is only acceptable as long as the RWE analysis is based on the rod drift event. The licensee has committed to analyzing the rod drift event in future cycles and is required to submit appropriate TS amendments to include adequate RBM setpoints if the RWE is analyzed using the conventional NRC-approved methodology. Based on the above, the staff concluded that operation in the proposed manner-will not endanger the health and safety of the public and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 4.0 STATE CONSULTATION In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendments. The State official had no comments. 5.0 ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on May 1, 1998 (63 FR 24197). Accordingly, based upon the environmental assessment, the staff has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment. 6.0 CONCLUSION The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor. G. Golub Date: May 4, 1998

23 -4 REFERENCES 1. PL-NF A, "Application of Reactor Analysis Methods for BWR Design and Analysis," dated July PL-NF , ususqehanna SES Unit I Cycle 9 - Reload Summary Report," Pennsylvania Power & Light Company, March PL-NF A, Supplement 2-A, "Application of Reactor Analysis Methods for BWR Design and Analysis," dated July PL-NF , Revision 1, "Susquehanna SES Unit 2 Cycle 8 - Reload Summary Report," Pennsylvania Power & Light Company, March 1996.

24 P UNITED STATES NUCLEAR REGULATORY COMMISSION PENNSYLVANIA POWER AND LIGHT COMPANY DOCKET NOS AND NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The U.S. Nuclear Regulatory Commission (Commission) has issued Amendment No. 176 to Facility Operating License No. NPF-14 and Amendment No. 149 to Facility Operating License No. NPF-22 issued to Pennsylvania Power and Light Company (PP&L, the licensee), which revised the Technical Specifications (TSs) for operation of the Susquehanna Steam Electric Station, Units 1 and 2, located in Luzeme County, Pennsylvania. The amendment is effective as of the date of issuance. The amendment modified the TSs by changing the Rod Block Monitor (RBM) flow biased trip setpoints and also the RBM channel calibration frequency and allowed outage times. The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License and Opportunity for a Hearing in connection with this action was published in the FEDERAL REGISTER on April 11, 1997 (62 FR 17885). No request for a hearing or petition for leave to intervene was filed following this notice.

25 -2- The Commission has prepared an Environmental Assessment related to the action and has determined not to prepare an environmental Impact statement. Based upon the environmental assessment, the Commission has concluded that the issuance of the amendment will not have a significant effect on the quality of the human environment ( 63 FR ). For further details with respect to the action see (1) the application for amendment-dated November 27, 1996, and supplemented by letter dated February 12, 1997, (2) Amendment No.176 to License No. NPF-14, (3) Amendment No.149 to License No. NPF-22, (4) the Commission's related Safety Evaluation, and (5) the Commission's Environmental Assessment. All of these items are available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street NW., Washington, DC, and at the local public document room located at the Osterhout Free Library, Reference Department, 71 South Franklin Street, Wilkes Barre, PA Dated at Rockville, Maryland, this 4th day of May FOR THE NUCLEAR REGULATORY COMMISSION Victor Nerses, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects - I/I1 Office of Nuclear Reactor Regulation