ML20082H053
| ML20082H053 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/10/1983 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082H055 | List: |
| References | |
| NUDOCS 8312010014 | |
| Download: ML20082H053 (10) | |
Text
{{#Wiki_filter:e ma, u s jf o UNITED STATES 'g y'* g NUCLEAR REGULATORY COMMISS!ON n. cj W ASHINGTON, D. C. 20555 \\..**/ l ARKANSAS POWER & LIGHT COMPANY DOCKET N0. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No. NPF-6 j 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Arkansas Power & Light Company (the licensee) dated February 23, 1983 as supplemented April 18, 1983 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, j the provisions of the Act, and the rules and regulations of the Commission; l 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 tha 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. 8312010014 831110 PDR ADOCK 05000368 P PDR
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 Facility Operating License No. NPF-6 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 49, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions, except where otherwise stated in specific license conditions. 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULA Y COMMISSION v James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: ' November 10, 1983
- l ATTACHMENT TO LICENSE AMENDMENT NO. 49 FACILTIY OPERATING LICENSE N0. NPF-6 DOCKET NO. 50-368 Replace the following pages of the Appendix "A" Technical Specifications The' revised pages are identified by Amendment with the enclosed pages.
The number and contain vertical lines indicating the areas of change. co.rresponding overleaf pages are provided to maintain document completeness. Pages 3/4 2-7 3/4 2-13 3/4 2-14 3/4 3-Sa 3/4 4-25 B 2-3 B 2-4 B 2-7 8 3/4 4-10 1 .,--,---a g-r,---, .g- .w-- r
SAFETY LIMITS AND LIMITING SAFETY SIS' TEN SETTI'NGS BASES Pressurizer Pressure-High The Pressurizer Pressure-High trip, in conjunction with the pres-surizer safety valves and main steam safety valves, provides reactor coolant-system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at 1 370.887 psia g 2 which is below the nominal lift setting (2500 psia) of the pressurizer safety valves and its operation avoids the undesirable operation of the pressurizer safety valves. Pressurizer Pressure-Low The Pressurizer Pressure-Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a Loss of Coolant Accident. During norma 1' operation, this trip's setpoint is set at il712.757 psia. This trip's setpoint may be manually i decreased, to a minimum value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pres-surizer pressure and this trip's setpoint is maintained at i 200 psi; this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached. Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint for l this trip is identical to the safety injection setpoint. Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an. excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setpoint is sufficiently be hw the full load operating point of approximately 900 psia so as not to inter-fere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually. decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at 1 200 psi; this set-point increases automatically as steam generator pressure increases until the trip setpoint is reached. ARKANSAS - UNIT 2 B 2-4 Amendment No. 49 i L
SAFETYLIMITSANDLIMITINGSAFETYS'YSTEkSETTINGS BASES To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density - High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CEN-44(A)P, " Core Protection Calculator Functional Description," January 7,1977, Supplement 1P, May 13,1977, Supplement 2P, May 19,1977, Supplement 3P, September 2,1977; CEN-45(A)P, " Control Element Assembly Calculator Functional Description," January 7,1977; CEN-53(A)P, "AN0-2 Cycle 1 CPC and CEAC Data Base Document," May 20, 1977, Amendment 1P, June 28, 1977, Supplement.2P, September 2,1977. Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. Linear Power Level-High The Linear Power Level-High trip provides reactor core protection against rapid reactivity excursions which might occur as the result of an ejected CEA. This trip initiates a reactor trip at a linear power level l of < 110.712% of RATED THERMAL PO!4ER. Logarithmic Power Level-High The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown con-dition. A reactor trip is initiated by the Logarithmic Power Level - High trip at a THERMAL POWER level of < 0.819% of RATED THERMAL POWER unless this trip is manually bypassed liy the operator. The operato4 may manually bypass this trip when the THERMAL POWER level is above 10 % of RATED THERMAL POWER; this bypass is g% of RATED THERMAL POWER.utomatically THERMAL POWER level decreases to 10-ARKANSAS - UNIT 2 B 2-3 Amendment No.4 9
u SAFETYLIMITSANDLIMITINGSAFETYSYSTENSETTI5GS BASES a. RCS Cold Leg Temperatu're-Low > 465*F 'b. RCS Cold Leg Temperature-High 7 605*F c. Axial Shape Index-Positive N'ot more positive than +0.6 d. Axial Shape Index-Negative Not more negative than -0.6 e. Pressurizer Pressure-Low > 1750 psia g.- Integrated Radial Peaking -7 2400 psia f.. Pressurizer Pressure-High h. Integrated Radial Peaking -> 1.28 Factor-Low 4 Factor-High. < 4.28 1. Quality Margin-Low-E0 Steam Generator Level-High The Steam Generator Level-High trip is provided to protect the turbine from excessive moisture carry over. Since the turbine is auto-matically tripped when the reactor is tripped, this trip provides a i reliable means for providing protection to the turbine from excessive ~ moisture carry over. This trip's setpoint does not correspond to a Safety Lim'it and no credit was taken in the accident analyses for oper-i i ation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. 2.2.2 CPC Addressable Constants The Core Protection Calculator. (CPC) addressable constants are-provided to allow calibration of the CPC system to more accurate indications such as-calorimetric measurements for power level and RCS flowrate and incore - detector signals for axial flux shape, radial peaking factors and CEA deviation penalties. Other CPC addressable' constants allow penalization of-the calculated DNBR and LPD values based on measurenent uncertainties or inoperable equipment. Administrative controls on changes and periodic l checking of addressable constant values (see also Technical Specifications 3.3.l.1 and 6.8.1) ensures that inadvertent misloading is unlikely. The methodology for determination of CPC addressable constant values is described in MSS-NA2-P, " Arkansas Nuclear One-Unit 2 Core Protection Calculator Addressable Constant Determination Methodology" dated August 1981. 2 L n ARKANSAS --UNIT 2 B 2-7 Anendment No gA,4 g i i a
POWER DISTRIBUTION LIMITS DNBR MARGIN, LIMITING CONDITION FOR OPERATION 3.2.4 The DNBR margin shall be maintained by operating within the region of acceptable operation of Figure 3.2-3 or 3.2-4, as applicable. APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER. l ACTION: With operation outside of the region of acceptable operation, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on DNBR; or (2) when the COLSS is not being used, any OPERABLE Low DNBR channel exceeding the DNBRilimit, within 15 minutes initiate corrective action to reduce the DNBR to within the limits and either: a. Restore the DNBR to within its limits within one hour, or l b. Be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable. 4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verify-ing at least once per 2 hours that the DNBR, as indicated on all 0PERABLE DNBR channels, is within the limit shown on Figure 3.2-4 l 4.2.4.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR. ARKANSAS - UNIT 2 3/4 2-7 Amendment No. 24
=. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.4.4 The following DNBR penalty factors shall be verified to be included in the COLSS and CPC DNBR calculations at least once per 31 days: GWD Burnup(MTU} DNBR Penalty (%) 0-30 2.0 30-40 3.5 40-50 5.5 The penalty for each batch will be determined from the batch's maximum burnup assembly and applied to the batch's maximum radihl power peak j assembly. A single net penalty for COLSS and CPC will be determined from the penalties associated with each batch, accounting for the offsetting i margins due to the lower radial power peaks in the higher burnup batches. I f ARKANSAS - UNIT 2 3/4 2-8 Amendment No. 2's, #, 32
POWER DISTRIBUTION LIMITS AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The core sverage AXIAL SHAPE INDEX (ASI) shall be maintained within the following l!mits: a. COLSS OPERABLE -0.28 < ASI < + 0.28 l. COLSS OUT OF SERVICE (CPC) -0.20 < ASI < +0.20 APP'LICABILITY: MODE I above 20% of RATED THERMAL POWER
- ACTION:
With the core average AXIAL SHAPE INDEX (ASI) exceeding its limit, restore the ASI to within its limit within 2 hours or reduce THERMAL POWER to less than 20% of RATED THERMAL POWER within the next 4 hours. SURVEILLAN'CE REQUIREMENTS 4.2.7 The core average AXIAL SHAPE INDEX shall be determined to be within ] its limits at least once per 12 hours using the COLSS or any OPERABLE Core Protection Calculator channel.
- See Special' Test Exception 3.10.2.
ARKANSAS - UNIT 2 3/4 2-13 Amendment No. pp, 4 g 1-w -em wm y y m- -i-?-
I _ POWER DISTRIBUTION LIMITS PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION 3.2.8 The average pressurizer pressure shall be maintained between 2225 psia and 2275 psia. APPLICABILITY: MODE 1 ACTION: With the average pressurizer pressure exceeding its limits, restore the pressure to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.8 The average pressurizer pressure shall be determined to be within its l limit at least once per 12 hours. ARKANSAS - UNIT 2 3/4 2-14 Anendment No. 24, 18' 49'
1 TABLE 3.3-1 (Continued) ACTION STATEMENTS b. With both CEACs inoperable, operation may continue provided that: 1. Within 1 hour the margins required by Specifi-cations 3.2.1 and 3.2.4 are increased and main-tained at a value equivalent to > 11% of RATED THERMAL POWER. 2. Within 4 hours: a) All full length and part length CEA groups ~ are withdrawn to and subsequently main-tained at the " Full Out" position, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2 or for control when CEA group 6 may be inserted no further than 127.5 inches withdrawn. b) The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to the inoper-able status, c) The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the "Off" mode except during CEA group 6 motion permitted by a) above, when the CEDMCS may be operated in either the " Manual Group" or " Manual Individual" mode. 3. At least once per 4 hours, all full length and part length CEAs are verified fully withdrawn except during surveillance testing pursuant to Specification 4.1.3.1.2 or during insertion of CEA group 6 as permitted by 2. a) above, then l verify at least once per 4 hours that the inserted CEAs are aligned within 7 inches (indicated position) of all other CEAs in its group. ACTION 6 With three or more auto restarts of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL TEST within the next 24 hours. ARKANSAS - UNIT 2 3/4 3-Sa Amendment No. 71*4 0
1 g= TABLE 3.3-2 5ll gg REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES y; lE FUNCTIONAL UNIT RESPONSE TIME na 1. Manual Reactor Trip Not Applicable 2. Linear Power Level - High 5,0.40 seconds
- 3.
Logarithmic Power Lctai - High 5,0.40 seconds
- 4.
Pressurizer Pressure - High 3,0.90 seconds 5. Pressurizer Pressure - Low 5,0.90 seconds 6. Containment Pressure - High 5,1.59 seconds ,,1 7. Steam Generator Pressure - Low 5.0.90 seconds 8. Steam. Generator Level - Low-5.0.90 seconds 9. Local Power Density - High a. Heutron Flux Power from Excore Neutron Detectors < 2.58 seconds
- I b.
CEA Positions 51.58 seconds ** e
=
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REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to: a. A maximum heatup of 200*F in any one hour period, and b. A maximum cooldown of 200'F in any one hour period. ' APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to detennine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatur6s shall be determined to be within the limits at least once per 30 minutes during system heatup or cool-down. ARKANSAS - UNIT 2 3/4 4-25 Amendment No. 4 g
REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 AND 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 com-ponents shall be maintained in accordance with Specification 4.4.10.1. APPLICABILITY: ALL MODES ACTION: a. With the structural integrity of any ASME Code Class 1 com-ponent(s) not conforming to the above requirements, restore thestructuralintegrityoftheaffectedcomponent(s)to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50"F above the minimum temperature required by NDT considerations. b. With the structural integrity of any ASME Code Class 2 com-ponent(s) not conforming to the :bove requirements, restore thestructuralintegrityoftheaffectedcomponent(s)to ' within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F. c. With the structural integrity of any ASME Code Class 3 com-ponent(s) not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component from service. d. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. ARKANSAS - UNIT 2 3/4 4-26
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REACTOR COOLANT SYSTEM BASES The actual shif t in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to tha adjacent section of the reactor vessel. The heatup and cooldown curves must be recalcu-lated when the ARTNDT determined from the surveillance capsule is dif-ferent from the caTculated ARTNDT for the equivalent capsule radiation exposure. The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50. The maximum RTNDT for all reactor coolant system pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 50*F. The Lowest Service Temperature limit line shown on Figure 3.4-2 is based upon this RTNDT since Article NB-2332 (Sumer Addenda of 1972) of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 100 F for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. The limitations imposed on the pressurizer heatup and cooldown rates are provided to assure that the pressurizer is operated within the design l criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. ARKANSAS - UNIT 2 B 3/4 4-10 Amendment No. 1g . _. - _ - _ -}}