ML20087A156
| ML20087A156 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 07/27/1995 |
| From: | Dick G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20087A158 | List: |
| References | |
| NPF-37-A-073, NPF-66-A-073, NPF-72-A-065, NPF-77-A-065 NUDOCS 9508040205 | |
| Download: ML20087A156 (28) | |
Text
{{#Wiki_filter:, paarc a. -j*k UNITED STATES , h..I Tf- )7 +,[- NUCLEAR REGULATORY COMMISSION in f: WASHINGTON, D.C. 20555 6 1 y 'sr.....J COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No. NPF-37 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Commonwealth Edison Company (the licensee) dated February 15, 1995, as supplemented February 28, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's. rules and 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 rules and regulations of the Commission; t C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health t and safety of the public,- and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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. 2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.14PF-37 is hereby amended to read as follows. -r 9500040205 950727 PDR ADOCK 05000454 P PDR
pp i j i (2) Technical Soecifications The Technical Specifications contained in Appendix A as revised l through. Amendment No. 73? and the Environmental-Protection Plan i contained in Appendix B, both of which are attached hereto, are i hereby' incorporated-into this license. The licensee shall-operate the facility in accordance with the Technical Specifications and ~ he Environmental Protection Plan. t 3. This license amendment is effective as of the date of its ' issuance and shall be implemented within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION l George . Dick Jr., Senior Project Manager-Project Directorate III-2 Division of Reactor Projects - Ill/IV Office-of Nuclear Reactor Regulation j
Attachment:
Changes to the Technical Specifications Date of Issuance: July 27, 1995' h l 4 k
.e .I 'I*' p3 Kf c [*.. UNITED STATES
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[ NUCLEAR REGULATORY COMMISSION ) ^' "[t' WASHINGTON, D.C. 205S4001 9.../ T: ( COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-455 SYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE h Amendment No.73 License No. NPF-66 1. The Nuclear Regulatory-Commission (the Commission) has found that: A. The application for. amendment by Commonwealth Edison Company (the licensee) dated February 15, 1995, as supplemented on February 28,- 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B. The ' facility will operate in conformity with the application, the provisions of the Act and the rules and regulations of the Commission; C. There is reasonable assurance (1) 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; 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;. t. and i E. The issuance of this amendment is in accordance with 10 CFR l Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment:to this license' amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows: l l
'4l (2) Technical Specifications The Technical. Specifications contained in Appendix A (NUREG-Ill3), as revised through Amendment No.73 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37,- dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appeadix A whici' is-hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION /U L- . ):1 e George F. Dick Jr., Senior Project Manager Project Directorate-III-2 Division of Reactor. Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: Jcly 27, 1995 r
ATTACHMENT TO LICENSE AMENPMENT NOS. 73 AND 73 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. The page indicated by an asterisk is provided for convenience only. Remove Paaes Insert Paaes IV IV 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-Sa
- 3/4 1-6
- 3/4 1-6 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 6-22 6-22 6-22a
=
i -LIMITING' CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS o SECTION: lAE l 3/4.0 APPLICABILITY................................................ 3/4 0-1: a i 3/4.1 REACTIVITY CONTROL SYSTEMS 1 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T,,,> 200*F............................ 3/4 1-l' Shutdown Margin - T,,,$_200*F............................ 3/4 1-3 l Moderator Temperature Coef ficient........................ 3/4 1-4 FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT VERSUS POWER LEVEL......................................... 3/4 1-Sa-y Minimum Temperature for Criticality...................... '3/4.1-6 l 3/4.1.2 B0 RATION SYSTEMS l Flow Path.- Shutdown..................................... '3/4 1 ! Flow Paths - Operating................................... '3/4 1-8 l Charging Pump - Shutdown................................. 4/4 1-9 Charging Pumps - Operating............................... 3/4 1-10 f [ Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources.- 0perating........................ -3/4 1-12 Boron Dilution Protection System......................... 3/41-13a 3/4.1.3 MOVABLE CONTROL' ASSEMBLIES Group Height..........=................................... 3/4'l-14 -TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE l ~ EVENT OF AN IN0PERABLE FULL-LENGTH R00.............. 3/4 1-16 Position Indication Systems - Operating.................. 3/4 1-17 Position Indication System - Shutdown..................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits............................. 3/4 1-21' l FIGURE 3.1-1 R0D BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION.......................... 3/4'l-22 u BYRON - UNITS I & 2 IV AMENDMENT N0. 73. 1 ..p
~_ - REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT UMITINGCONDITIONFOROpERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the Operating Limits Report (0LR). The maximum upper limit shall be less than or equal to that shown in Figure 3.1-0. APPLICABILITY: Beginning of Life (BOL) limit - MODES 1 and 2* only". End of Life (E0L) limit - MODES 1, 2, and 3 only". ACTION: a. With the MTC more positive than the BOL limit specified in the OLR, l operatica in MODES I and 2 may proceed provided: 1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL l limit specified in the OLR within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in-addition to the insertion limits of Specification 3.1.3.6; 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition. 4. The provisicns of Specification 3.0.4 are not applicable. b. With the MTC more negative than the E0L limit specified in the OLR, l be in HOT SHUTDOWN within 12 hours. i
- With K,,,ial Test Exceptions Specification 3.10.3.
greater than or equal to 1.
- See Spec 3
BYRON - UNITS 1 & 2 3/4 1-4 AMENDMENT NO. 73
REACTIVITY CONTROL SYSTEMS l SURVEJLLANCE REQUIREMENTS L 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: a. The MTC shall be measured and compared to the predicted MTC to establish administrative rod withdrawal limits, as necessary, to assure that the BOL limit specified in the OLR, is met throughout I core life, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, and l l b. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the OLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium baron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit specified in the OLR, the MTC shall be remeasured, and compared to the E0L MTC limit specified in the OLR, at least once per 14 EfPD during the remainder of the fuel cycle. l BYRON - UNITS 1 & 2 3/4 l'-5 AMENDMENT NO. 73
'.) 4 10, jgL f Unacceptable Operation S ~ C 17 Ct } 6 t-U l e 5 E d7 Acceptable Operation 3
- o i
2h 1 h 4 0 O 10 20 30 40 50 60 70 E 90 100 Percent RTP 1 1 i 1 FIGt!RE 3.1-0 MODERATOR TEMPERTURE COEFFICIENT vs. POWER LEVEL BYROH - UNIT 5 1 & 2 3/4 1-Sa AMENDMENT NO.73 1 I
4 L +* m+ -GLc. m A-REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY i LIMITING CONDITION FOR OPERATION s 3.1.1.4' The Reactor Coolant System lowest operating loop temperature (Tavg) shall be greater than or equal to 550'F. i APPLICABILITY: MODES I and 2#*. ACTION: With a Reactor Coolant System operating loop temperature (T,yg) less than 550 F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS i 4.1.1.4 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 550 F: Within 15 minutes prior to achieving reactor criticality, and a. b. At least once per 30 minutes when the reactor is critical and the i Reactor Coolant System T,yg is less than 557*F with the T,yg-Tref l Deviation Alarm not reset. i i i i I f
- With K,ff greater than or equal to 1.
l
- See Special Test Exceptions Specification 3.10.3.
BYRON - UNITS 1 & 2 3/4 1-6 4 2
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN HARGIN A sufficient SHUTDOWN MARCIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,. The most restrictive condition occurs at EOL, with T,,, at no load operaEing temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% ok/k is required to control the reactivity transient. l Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting t condition and is consistent with FSAR safety analysis assumptions. With T,y less than 200*F, the reactivity transients resulting from a postulated steam, line break cooldown are minimal and a 1% Ak/k SHUTDOWN MARGIN provides adequate i protection provided that boration dilution paths are isolated. A 1.3% ok/k SHUIDOWN MARGIN is required to ensure the OPERABILITY of the automatic Boron Dilution Protection System. 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the UFSAR accident and transient analyses. The limitations on MTC i also ensure that the Anticipated Transient Without Scram (ATWS) risk is acceptable. A cycle specific Unfavorable Exposure Time (UET) value will be calculated to ensure < 5% of the cycle operations occur when the reactivity l feedback is not sufficient to prevent exceeding an ATWS overpressurization i condition of 2 3200 psig in the RCS. This UET value will be updated for each core reload and appropriately considers the effects of changes in MTC, including any variations that are more adverse than those originally modeled in the analyses supporting the basis for the final ATWS rule. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. 1 1 BYRON - UNITS 1 & 2 B 3/4 1-1 AMENDMENT NO. 73
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the UFSAR analyses to nominal operating conditions. These corrections l involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting E0L MTC value specified in the OLR. The 300 ppm surveillance limit represents a conservative value (with corrections for burnup and soluble boron) with an equilibrium boron concentration and is obtained by making these corrections to the limiting E0L MTC value. The Surveillance Requirements for measurement of the MTC at the beginning i and near the end of the fuel cycle are adequate to confirm that the MTC can be maintained within its limits. The BOL MTC measurement, combined with the predicted MTC throughout core life, will be used to impose administrative limits on rod withdrawal, as required during core life to ensure that MTC will always remain within the limits specified in the OLR. This coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 550*F. This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, (4) the reactor vessel is above its minimum RT,P-12. temperature, and (5) the plant is above the cooldown steam dump permissive, 3/4.1.2 B0 RATION SLSHMS The Boron Injectiw System ensures that negative reactivity control is available during each MODE of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 350*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement is 13,487 (15,780*) gallons of 7000-ppm borated water from the boric acid storage tanks or 54,014 (70,450*) gallons of 2300-ppm (2000-ppm,) borated water from the refueling water storage tank. A Boric Acid Storage System level of 40% ensures that there is a volume of greater than or equal to 13,487 (15,780*) gallons available. A RWST level of 89% ensures that there is a volume of greater than or equal to 395,000 gallons available.
- Not applicable to Unit 1.
Applicable to Unit 2 until completion of cycle 5. BYRON - UNITS 1 & 2 B 3/4 1-2 AMENDMENT NO.73
ADMINISTRATIVE CONTROLS REPORTING RE0VIREMENTS (Continued) ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted prior to May 1 of each year.
The report shall include summaries, interpreta-i tions, and analysis of trends of the results of the Radiological Environmental l Monitoring P ogram for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT ** f .6.9.1.7 A Radioactive Effluent Release Report covering the operation of the facility during the previous year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous. effluents and solid waste released from the facility. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP' i and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50. MONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, i including documentation of all challengas to the PORVs or RCS safety valves, i shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report. OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle. The analytical methods used to determine the operating limits shall be those l previously reviewed and approved by the NRC in Topical Reports: l 1. WCAP-9272-P-A, " Westinghouse Reload Safety Evaluations Methodology" dated July 1985. 2. WCAP-8385, " Power Distribution Control and Load Following Procedures-Topical Report" dated September 1974. i
- A single subm.ittal may be made for a multi-unit station.
- A single submittal may be made for a multi-unit station.
The submittal should combine those sections that are common to all units at the station; however, for units wit _h separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. BYRON - UNITS 1 & 2 6-22 AMENDMENT N0.73 t
.~ ~. 9 ADMINISTRATIVE' CONTROLS OPERATING tlMITS REPORT (Continued) i 3. NFSR-0016, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods" dated July 1983. 4. NFSR-0081, " Commonwealth Edison Company Topical Report on Benchmark of PWR i ~ Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," dated July 1990. 1 5. Comed letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and Comed application of the UET methodology addressed in " Additional Information Regarding Application for Amendment to Facility Operating ~ Licenses-Reactivity Controls Systems." r The operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, i nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall. be provided upon issuance,- for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. i 1 i i l 4 l f 1 l i BYRON - UNITS 1 & 2 6-22 a AMENDMENT NO.73 1
r. )k.f?*'%s[t-UNITED STATES ..f If). [ NUCLEAR REGULATORY COMMISSION e w AsninoTow, o.c. 2o555-o005
- g%g...../
g COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 - I License No. NPF-72 1. The Nuclear Regulatory Commission (the Commission) has found that: l A. The application for amendment by Commonwealth Edison Company.(the - l licensee) dated February 15, 1995, as supplemented on February 28, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's .l rules and regulations set forth in 10 CFR Chapter I;. B. The facility will operate in conformity with the application, the l provisions of the Act and the rules and 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; d 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 3 E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable l requirements have been satisfied. 2. Accordingly, the-license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and-paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows: l i l l t
? w. s 4 .;.g. (2) Technical Soecifications i The: Technical Specifications contained in Appendix A as revised' through Amendment No. 65 and the Environmental Protection Plan. contained in Appendix B,-both of which are attached hereto, are-hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental-Protection ~ Plan. 7 .3. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION j i l /7 /s a - ~, c Ramin R. Assa, Project-Manager Project Directorate III-2 Division of Reactor Projects - III/IV l Office of Nuclear Reactor Regulation: l .)
Attachment:
f, Changes to the Technical Specifications Date of Issuance: July 27, 1995 i i I i 5 -l 'i
~ '*Wp.g 8-UNITED STATES 1 [.. i * 'j - NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055!WW1 f %g*... + gi' COMMONWEALTH EDIS0N COMPANY-DOCKET-N0.' STN 50-457 BRAIDWOOD~ STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE' Amendment No. 65 License No. NPF-77 '1. The Nuclear Regulatory Commission'(the Commission) has found that: [ A. The application for amendment by Commonwealth Edison Company (the licensee) dated February 15, 1995, as supplemented on' February 28, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B. The facility will operate in conformity with the application,-the provisions of the Act and the rules and 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; 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 l Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2.
- Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated-in the attachment to this license amendment and paragraph 2.C.(2) of Facility.0perating License No. NPF-77 is hereby 1
amended to read as follows: s 1 m _o
(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 65 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2,1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of the date if its issuance and shall be implemented within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION Ramin R. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 27, 1995 4 E
~ l F ATTACHMENT TO LICENSE AMENDMENT NOS. 65 AND 65 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Appendix "A" Technical Specifications with [ the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The page indicated by an asterisk is provided for convenience only. Remove Paaes Insert Paaes IV IV 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-Sa
- 3/4 1-6
- 3/4 1-6 B 3/4 1-1 B 3/4 1-1
] B 3/4 1-2 8 3/4 1-2 -{ 6-22 6-22 6-22a l 'I 1
^ LIMITING C9NDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS l l SECTION-PME 3/4.0 APPLICABILITY................................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS-i 3/4.1.1 B0 RATION CONTROL y Shutdown Margin - T.,, > 200*F............................ 3/4 1-1 Shutdown Margin - T,y, s 200*F............................ 3/4 1-3 Moderator Temperature Coefficient........................ 3/4 1-4 FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT VERSUS POWER LEVEL......................................... 3/4 1-Sa o Minimum Temperature foi Cri ticality...................... 3/4 1-6~ 3/4.1.2 B0 RATION SYSTEMS Flow Path - Shutdown...................................... 3/4 1-7 l Flow Paths - Operating................................... 3/4 1-8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - Operating............................... 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources - 0perating........................ 3/4 1-12 . Boron Dilution Protection System......................... 3/4 1-13a 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN IN0PERABLE FULL-LENGTH R0D.............. 3/4 1-16 Position Indication Systems - Operating.................. 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 l Shutdown Rod Insertion Limi t............................. 3/4 1-20 q Control Rod Insertion Limits............................. 3/4 1-21 FIGURE 3.1-1 R0D BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION.......................... 3/4 1-22 I BRAIDWOOD - UNITS 1 & 2 IV Amendment No.65 l )
l 14- ' REACTIVITY CONTROL SYSTEMS l MODERATOR TEMPERATURE COEFFICIENT l LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the Operating Limits Report (0LR). The maximum i upper limit shall be less than or equal to that shown in Figure 3.1-0. l t APPLICABILITY: Beginning of Life (BOL) limit - MODES I and 2* only". End of Life (E0L) limit - MODES 1, 2, and 3 only". l ACTION-I t a. With the MTC more positive than the BOL limit specified in the'OLR, I operation in MODES I and 2 may proceed provided t 1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the OLR within 24 hours or be in HOT STANDBY-within the next 6 hours. These withdrawal limits shall be in i addition to the insertion limits of Specification 3.1.3.6; 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3. A Special Report is prepa. red and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for + restoring the positive MTC to within its limit for the all rods withdrawn condition, i 4. The provisions of Specification 3.0.4 are not applicable. b. With the M1C more negative than the E0L limit specified in the OLR, l be in HOT SHUTDOWN within 12-hours.
- With K,,,ial Test Exceptions Specification 3.10.3.
greater than or equal to 1.
- See Spec BRAIDWOOD - UNITS 1 & 2 3/4 1-4 Amendment No.65 e
w y er -,
F REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS i 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: a. The MTC shall be measured and compared to the BOL predicted MTC to establish administrative rod withdrawal limits, as necessary, to i assure that the BOL limit specified in the OLR, is met throughout core life, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, and b. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the OLR (all rods withdrawn, i RATED THERMAL POWER condition) within 7 EFPD after reaching an i equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit specified in the OLR, the MTC shail be remeasured, and compared to the E0L MTC limit specified in the OLR, at least once per 14 EFPD during the remainder of the. fuel cycle. l f i s l b l-i i ( f BRAIDWOOD - UNITS 1 & 2 3/4 1-5 Amendment No. 65 i
10 -9 Unacceptable Operation 8 2 i 7,' 6'h o 5 a o 4 Acceptable Operation 3' B-2 h 1 l l i ,,.,i .I t 0 10 20 30 40 50 60 70 80-90 100 Percent RTP 1 FIGURE 3.1-0 MODERATOR TEMPERTURE COEFFICIENT vs. POWER LEVEL j BRAIDWOOD - UNITS 3 & 2 3/4 1-Sa AMENDMENT NO.65
O REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be greater than or equal to 550 F. APPLICABILITY: MODES 1 and 2#*. ACTION: With a Reactor Coolant System operating loop temperature (T,yg) less than 550*F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.1. 4 The Reactor Coolant System te: erature (Tavg) shall be determined to be greater than or equal to 550 F: Within 15 minutes prior to achieving reactor criticality, and a. b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 557 F with the T,yg-Tref Deviation Alarm not reset.
- With Keff greater than or equal to 1.
- See Special Test Exceptions Specification 3.10.3.
BRAIDWOOD - UNITS 1 & 2 3/4 1-6
I 4-3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN .. A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable I limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the. shutdown condition. b SHUTDOWN MAF. GIN requirements vary throughout core life as a function of fuel depletion, WCS boron concentration, and RCS T,. The most restrictive condition occurs at E0L, with 1,y, atnoloadoperaYingtemperature,andis associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN ' MARGIN of 1.3% Ak/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,y less than 200*F, the reactivity transients resulting from a postulated steam, line break cooldown are minimal and a 1% Ak/k SHUTDOWN MARGIN provides adequate protection provided that boration dilution paths are isolated. A 1.3% Ak/k SHUIDOWN MARGIN is required to ensure the OPERABILITY of the automatic Boron Dilution Protection System. 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the UFSAR accident and transient analyses. The limitations on MTC also ensure that the Anticipated Transient Without Scram (ATWS) risk is acceptable. A cycle specific Unfavorable Exposure Time (UET) value will be calculated to ensure < 5% of the cycle operations occur when the reactivity feedback is not sufficient to prevent exceeding an ATWS overpressurization condition of 2 3200 psig in the RCS. This UET value will be updated for each core reload and appropriately considers the effects of changes in MTC, including any variations that are more adverse than those originally modeled in the analyses supporting the basis for the final ATWS rule. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. BRAIDWOOD - UNITS 1 & 2 B 3/4 1-1 Amendment No.65 l
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC i used in the UFSAR analyses to nominal operating conditions. These corrections l involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting E0L MTC value specified in the OLR. The 300 ppm surveillance limit represents a conservative value (with corrections for burnup and soluble boron) with an equilibrium boron concentration and is obtained by making these corrections to the limiting E0L MTC value. The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC can be maintained within its limits. The BOL MTC measurement combined with the predicted MTC with core burnup, can be used to impose administrative limits on rod withdrawal to ensure that MTC will always remain within the limits specified in the OLR. This coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 550*F. This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, (4) the reactor vessel is above its minimum RT,P-12. temperature, and (5) the plant is above the cooldown steam dump permissive, 314.1.2 BORAT10N SYSTEMS The Boron injection System ensures that negative reactivity control is available during each MODE of facility operation. The components required to perform this functica include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 350*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% 6k/k after xenon decay and cooldown to 200,*F. The maximum expected boration capability requirement is 15,730 (13,487) 70,450 (54,014),of 7000-ppm borated water from the boric acid storage gallons tanks or gallons of 2000-ppm (2300-ppm) borated water from the refueling water storage tank. " Applicable to Unit I and Unit 2 starting with Cycle 6. BRAIDWOOD - UNITS 1 & 2 B 3/4 1-2 Amendment No. 65
n ATHINISTRATIVE CONTROLS REPORTING RE0UIREMENTS (Continued) ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
SEMI ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT ** 6.9.1.7 A Radioactive Effluent Release Report covering the operation of the facility during the previous year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50. MONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or RCS safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report. OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle. The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in Topical Reports: 1. WCAP-9272-P-A, " Westinghouse Reload Safety Evaluations Methodology" dated July 1985.
- A single submittal may be made for a multi-unit station.
" A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. BRAIDWOOD - UNITS 1 & 2 6-22 AMENDMENT NO. 65
~ -. ~ .A o ADMINI_STRATIVE CONTROLS j OPERATING LIMITS REPORT (Continued) i 2. WCAP-8385, " Power Distribution Control and Load following Procedures-l Topical Report" dated September 1974. 3. NFSR-0016, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods" dated July 1983. t 4. NFSR-0081, " Commonwealth Edison Company. Topical Report on Benchmark' of PWR i Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," dated July 1990. l S. Comed letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-Il992/11993 and Comed application of the UET methodology addressed in " Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Controls Systems." i e The operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The OPERATING. LIMITS REPORT, l including any mid-cycle revisions or supplements thereto, shall be provided -( upon issuance, for each reload cycle, to the NRC Document Control Desk with' l copies to the Regional Administrator and Resident Inspector. l P i t a i BRAIDWOOD - UNITS 1 & 2 6-22a AMENDMENT N0. 65 .}}