ML20207K201
| ML20207K201 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/22/1986 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20207K207 | List: |
| References | |
| NUDOCS 8607290358 | |
| Download: ML20207K201 (18) | |
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NUCLEAR REGULATORY COMMISSION p,.
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ARKANSAS POWER & LIGHT COMPANY DOCKET N0. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 License No. NPF-6 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Arkansas Power & Light Company (the licensee) dated September 16, 1985, 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; 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 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 Specifica-tions 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. 77, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
0607290358 860722 PDR ADOCK 05000368 P
- 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Geor'ge W.
nighton, Dire r
PWR Project Directorate
- o. 7 Division of PWR Licensing-B
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 22, 1986 I
]
.9
ATTACHMENT TO LICENSE AMENDMENT NO. 77 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET N0. 50-368 Replace the following pages of the Appendix "A" Technical. Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pages Insert Pages III III 2-4 2-4 2-7 2-7 2-8 2-8 2-9 2-9 B 2-7 B 2-7 3/4 3-7 3/4 3-7 3/4 3-9 3/4 3-9 6-13 6-13
INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Reactor Core.................................................. 2-1 Rea ctor Cool ant System Pressu re............................... 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints........................................ 2-3 Deleted....................................................... 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 l
ARKANSAS - UNIT 2 III Amendment No. B/V, %$, 77 l
1
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 ccnservative 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 OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
ARKANSAS - UNIT 2 2-3 c-
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THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 2-7 Amendment No. ZA, 66, 77
THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 2-9 Amendment No. 24, EE, 77
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES a.
RCS Cold Leg Temperature-Low 3.465 F b.
RCS Cold Leg Temperature-High 5.605 F c.
Axial Shape Index-Positive Not 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_
E2400 psia g.
Integrated Radial Peaking Factor-Low 3.1.28 h.
Integrated Radial Peaking Factor-High 5.4.28 1.
Quality Margin-Low 3.0 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 setting is required to enhance the overall reliability of the Reactor Protection System.
ARKANSAS - UNIT 2 B 2-7 Amendment No. //, guq, 77
TABLE 4.3-1 REACTOR PROTECTION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL liODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.
Manual Reactor Trip N.A.
N.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 1
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.
Steam 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 i
ARKANSAS - UNIT 2 3/4 3-7 Amendment No. 29, 77, 77
5 THIS PAGE LEFT BLANK INTENTIONALLY I
ARKANSAS - UNIT 2 3/4 3-9 Amendment No. y, 77
TABLE 4.3-1 REACTOR PROTECTION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL liODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.
Manual Reactor Trip N.A.
N.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 1
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
R M
1, 2 l
9.
Local Power Density - High S
D(2,4),
M,R(6) 1, 2 R(4,5)
I 10.
DflBR - Low S
S(7),
M,R(6),
1, 2 D(2,4),
l M(8),
R(4,5)
- 11. Steam 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. 25, 77, 77 i
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 shape 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 measurement 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%.
(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 measurenent uncertainties).
(9) - The correct values of addressable constants (See Table 2.2-2) shall be verified to be installed in each OPERABLE CPC.
ARKANSAS - UNIT 2 3/4 3-8 Amendment No. 79,39 l
THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 3/4 3-9 Amendment No. Jg, 77
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 applicable 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 cod 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 demanstrated 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
ADMINISTRATIVE CONTROLS 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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 (PCP) 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.
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, and changes thereto, shall be reviewed by the PSC and approved by the ANO General Manager prior to implementation and reviewed periodically as set forth in administrative procedures.
ARKANSAS - UNIT 2 6-13 Amendment No. Z$,2E,$3,E2,$0,$3, 77
ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
a.
The intent of the original procedure is not altered.
b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
c.
The change is documented, reviewed by the PSC and approved by the AN0 General Manager within 14 days of implementation.
l 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS,AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Administrator of the Regional Office unless otherwise noted.
STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments sball be included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commence-ment of commercial power operation, or (3) 9 months following initial critical-ity, whichever is earliest.
If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup) test program, and resumption or commencement of commercial power operation, supplementary reports shall be submitted at least every three months until all three events have been completed.
ARKANSAS - UNIT 2 6-14 Amendment No. E, g 9
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