ML20205C687
| ML20205C687 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/05/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20205C693 | List: |
| References | |
| NPF-06-A-077 NUDOCS 8608120440 | |
| Download: ML20205C687 (13) | |
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{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1 i THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 3/4 3-9 Amendment No. M, 77 k$e12044086o805 ADOCK 05000368 p PDR
f 3 O INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instru-mentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3. ACTION: a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the appitcable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip set-point adjusted consistent with the Trip Setpoint value. b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3. SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES Fed at the frequencies shown in Table 4.3-2. 4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. ARKANSAS-UNIT 2 3/4 3-10 1
O TABLE 4.3-1 (Continued) TABLE NOTATIONS - With reactor trip breakers in the closed position and the CEA drive system capable of CEA withdrawal. (1 ) - If not performed in previous 7 days. (2) - Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 15% of RATED THERMAL POWER; adjust the Linear Power Level signals and the CPC addressable constant multipliers to make the CPC AT power and CPC nuclear power calculations agree with the calorimetric calculation if absolute difference is > 2%. During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau. (3) - Above 15% of RATED THERMAL POWER, verify that the linear power sub-channel gains of the excore detectors are consistent with the values used to establish the shape annealing matrix elements in the Core Protection Calculators. (4 ) - Neutron detectors may be excluded from CHANNEL CALIBRATION. (5) - After each fuel loading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be used to determine the snape annealing matrix elements and the Core Protection Calculators shall use these elements. (6) - This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify OPERABILITY including alarm and/or trip functions. (7) - Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differen-tial pressure instrumentation (conservatively compensated for measure-ment uncertainties) or by calorimetric calculations (conservatively compensated for measuremei.t uncertainties) and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate. The flow measurement uncertainty may be included in the BERR1 term in the CPC and is equal to or greater than 4%. l (8) - Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by calorimetric calculations (conservatively compensated for measurement uncertainties). (9) - The correct values of addressable constants shall be verified to be installed in each OPERABLE CPC. ARKANSAS - UNIT 2 3/4 3-8 Amendment No. 79,fg, 77 l
INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS R e a c t o r Co r e.................................................. 2-1 Reactor Coolant System Pressure............................... 2-2 2.2 LIMITING SAFITY SYSTEM SETTINGS Reactor Trip Setpoints........................................ 2-3 l Deietea....................................................... 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS Reactor Core................................................. B 2-1 Reactor Coolant System Pressure.............................. B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints....................................... B 2-2 Deleted...................................................... B 2-7 i i ARKANSAS - UNIT 2 III Amendment No. f//, $$, 77 l
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY........................................... 3/40-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL 3/41-1 S h u td ow n Ma rg i n - T,yg > 20 0* F....................... 3/4 1-3 Sh u tdown Ma rg i n - T,,g <_ 2 00* F....................... Boron Dilution...................................... 3/4 1-4 Moderator Temperature Coef ficient................... 3/4 1-5 Minimum Temperature for Criticality.................. 3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Paths - Shutdown................................ 3/41-7 Flow Paths - Operating............................... 3/4 1-8 C ha rging Pump - Shu tdown............................. 3/4 1-9 Ch argi no Pump s - Operati ng........................... 3/4 1-10 Boric Acid Makeup Pumps - Shutdown................... 3/4 1-11 Boric Acid Makeup Pumps - Operating.................. 3/4 1-12 Borated Water Sources - Shutdown..................... 3/4 1-13 Borated Water Sources - Operating.................... 3/4 1-15 l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES C EA P o s i ti o n......................................... 3/4 1-17 Posi tion Indicator Channels - Operating.............. 3/4 1-20 Position Indicator Channels - Shutdown............... 3/4 1-22 [ l CEA Drop Time........................................ 3/4 1-23 Shu tdown C EA In se rtion Limi t......................... 3/4 1-24 Regul ating CEA Insertion Limi ts...................... 3/4 1-25 Part Length CEA Insertion Limits..................... 3/4 1-28 ARKANSAS - UNIT 2 IV Amendment No. N,60 --{-
) SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERT.BLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. ARKANSAS - UNIT 2 2-3
THIS PAGE LEET BLANK INTENTIONALLY ARKANSAS _ UNIT 2 2-4 Amendment No. 25, 77
e THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 2-7 Amendment No. N, $$, 77
THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 2-8 Amendment No. 24, 77
THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 2-9 Amendment No. H. H. 77
I SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES a. 'RCS Cold Leg Temperature-Low > 465*F b. RCS Cold Leg Temperature-High 7 605*F c. Asial Shape Index-Positive liot more positive than +0.6 d. Axial Shape Index-Negative Not more negative than -0.6 e. Pressurizer Pressure-Low > 1750 psia f. Pressurizer Pressure-High 7 2400 psia g. Integrated Radial Peaking ~~ h. Integrated Radial Peaking -> 1.28 Factor-Low Factor-High < 4.28 s_0 1. Quality Margin-Low 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 reliable means for providing protection to the turbine from excessive moisture carry over. This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for oper-ation of this trip. Its functional capability at the specified trip 4 setting is required to enhance the overall reliability of the Reactor j Protection System. l Anendment No. g, g, [7 ARKANSAS - UNIT 2 B 2-7 l
TABLE 4.3-1 REACTOR PROTECTION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECX CALIBRATION TEST REQUIRED 1. Manual Reactor Trip H.A. H.A. S/U(1) N.A. 2. Linear Power Level - High S D(2,4), M 1, 2 l M(3,4), Q(4) 3. Logarithmic Power Level - High S R(4) M and S/U 1,2,3,4,5 (1) and
- l 4.
Pressurizer Pressure - High S R M 1, 2 l 5. Pressurizer Pressure - Low S R M 1, 2 and
- l 6.
Containment Pressure - High S R M 1, 2 l 7. Steam Generator Pressure - Low S R M 1, 2 and
- l 8.
Steam Generator Level - Low S P. M 1, 2 l 9. Local Power Density - High S D(2,4), M. R(6) 1, 2 R(4,5) I 10. DNBR - Low S S(7), M. R(6), 1, 2 D(2,4), l ~ M(8), R(4,5) 11. Stean Generator Level - High S R M 1, 2 l 12. Reactor Protection System Logic N.A. N.A. M 1, 2 and
- l 13.
Reactor Trip Breakers N.A. N.A. M 1, 2 and
- l 14.
Core Protection Calculators S W(9) D(2,4) M. R(6), 1, 2 R(4,5) l 15. CEA Calculators S R M,R(6), 1, 2 l ARKANSAS - UNIT 2 3/4 3-7 Amendment No. 24, 77, 77
ADMINISTRATIVE CONTROLS l l 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated: a. The unit shall be placed in at least HOT STANDBY within one hour. b. The Safety Limit violation shall be reported to the Commission, the Vice President, Nuclear Operations and to the SRC within 24 hours. c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PSC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence. d. The Safety Limit Violation Report shall be submitted to the Commission, the SRC and the Vice-President, Nuclear Operations within 14 days of the violation. 6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below: a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978. b. Refueling operations. c. Surveillance and test activities of safety related equipment. d. Security Plan implementation. e. Emergency Plan implementation. f. Fire Protection Program implementation, g. Modification of Core Protection Calculator (CPC) Addressable Constants. These procedures should include provisions to assure that sufficient margin is maintained in CPC Type I addressable constants to avoid excessive operator interaction with the CPCs during reactor operation. NOTE: Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P that has been determined to be applicable to the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval. h. New and spent fuel storage. 1. ODCM and PCP implementation. J. Postaccident sampling (includes sampling of reactor coolant, radio-active iodines and particulates in plant gaseous effluent, and the containment atmosphere). 6.8.2 Each procedure of 6.8.1 above, the changes thereto, shall be reviewed by the PSC and approved by the Director, Site Nuclear Operations or the responsible General Manager prior to implementation and reviewed periodically as set forth in administrative procedures. ARKANSAS - UNIT 2 6-13 Amendment No. 2d, 28, #3, 52, 69, $3, TI. 77 .}}