ML20009G245
| ML20009G245 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/21/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20009G246 | List: |
| References | |
| NUDOCS 8108040022 | |
| Download: ML20009G245 (11) | |
Text
-
I5 gP fig UNITED s1 Af t s
- C NUCLEAR REGULATORY COMMISSION
+[ *, c,<
',' g WASHINGTON. D. C. 205%
C a'
SA%,*..../
k I
ARKANSAS POWER AND LIGHT COMPANY DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 26 License No. NPF-6 The Nuclear Regulatory Commission (the Commission) has found that:
1.
The application for amendment by Arkansas Power and Light Company A.
(the licensee) dated February 20 and March 5,1981, as supplemented, 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; The facility will operate in conformity with the application, D.
the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized C.
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; The issuance of this amendmert will not be inimical to the D.
common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR E.
f Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8106040022 010721 PDR ADOCK 05000368 P
Accordingly. Facility Operating License No. NPF-6 is amended by changes to the Technical Specifications as indicated in the 2.
attachment to this license amendment, and by revising Paragraphs 2.L.(1) and 2.C.(2) to read as follows:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of Prior to attaining this power 2815 megawatts themal.
level the licensee shall comply with the conditions in Paragraph 2.C.(3).
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 26, are hereby The licensee shall operate incorporated in the license.
the facility in accordance with the Technical Specifications.
This license amendment is effective as of the date of its issuance.
3.
FOR THE NUCLEAR REGULATORY C0ftMISSION e
r j( o bM d t$ k c 7
G
..s N l
Rocert A. Clark, Chief Operating Reactors Branch #3 l
Division of Licensing
Attachment:
1 Changes to the Technical Specifications Date of Issuance:
July 21, 1981 l
l
[
ATTACHMENT TO LICENSE AMENDMENT NO. 26 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET N0. 50-368 Replace the following pages of the Appendix "A" Technical Specifications The revised pages are identified by ame7dment with the enclosed pages.
The number and contain vertical lines indicating the area of change.
corresponding overleaf pages are also provided to maintain document completeness.
Pages_
2-6 3/4 2-8 B3/4 2-3 6-13
5 TABLE 2.2-1 95 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS i
E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES _
1.
Manual Reactor Trip Not Applicable Not Applicable 2.
Linear Power Level - High a.
Four Reactor Coolant Pumps
< 110% of ;1ATED THERMAL POWER
< 110.712% of RATED THERMAL POWER Operating b.
Three Reaci.or Ccolant Pumps Operating c.
Two Reactor Coolant Pumps
.i Operating - Same Loop 1
m d.
Two Reactor Coolant Pumps M
Operating - Opposite Loops I
3.
Logarithmic Power Level -
1 75% of RATED THERMAL POWER 1 0.819% of RATED THERMAL POWER 0
High (1) 4.
Pressurizer Pressure - High 1 2362 psia
< 2370.887 psia
[
5.
Pressurizer Pressure - Low
> 1766 psia (2)
> 1712.757 psia (2) o 1 8.4 psia 1 19.024 psia 1
[
6.
Containment Pressure - High
[
7.
Steam Generator Pressure - Low
> 751 psia (3) 1 729.613 psia (3) o b
8.
Steam Generator Level - Low
> 46.7% (4)
> 45.811 % (4 )
f O
1 4
- These values lef t blank pending NRC approval of safety analyses for operation with less than four reactor coolant pumps operating.
4 e
1
TABLE 2.2-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E
ALLOWABLE VALUES 9
FUNCTIONAL UNIT
-TRIP SETPOINT 9.
Loca. Power Density - High 1 20.3 kw/ft (5) 1 20.3 kw/ft (5)
> 1.24 (5)(6)(7)
> 1.24 (5)(6)(7) h 10.
DNBR - Low 11.
Steam Generator Level - High 1 93.7% (4) 1 94.589% (4)
TABLE NOTATION (1) Trip may be manually bypassed above 10~4% of RATED THERMAL POWER; bypass shall be automatically
-4 removed when THERMAL POWER is 1 10 of RATED THERMAL POWER.
Value may be decreased manaally; to a minimum value of 100 psia, during a planned reduction in l
l (2) pressurizer pressure, provided the margin between the or essurizer pressure and this value is maintained at 1200 psi; the setpoint shaP be increased automatically as pressurizer pressure is increased antil i
7 the trip setpoint is reacl;ed. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is > 500 psia.
'3) Value may be decreased manually during a planned reduction in steam generator pressure provided the 200 psi; the setpoint margin between the steam generator pressure and this value is maintained at 1 l
l shall be increased awomatically as steam generator pressure is increased until the trip setpoint is l
- reached, k
(4) % of the distance between steam generator upper and lower level instrument nozzles.
i As stored within the Core Protection Calct.lator (CPC).
Calculation of the trip setpoint includes measure-t R
(5)
Tripmaybemanuallybypassgd ment, calgulational and processor uncertainties, and dynamic allowances.below 10' % of RATED
?+
i 2
F of RATED THERMAL POWER.
The minimum allowable value of the addressable constant BERR1 in each OPERABLE channel is 1.086.
l (6)
The opproved SCU eat:ivalent DNBR limit is 1.26 which includes a two percent rod bow compensation.
(7)
A DNBR trip setpoint of 1.24 is allowed provided that the difference is compensated by an increase of the addressable constant BERR1 to a minimum allowable value of 1.065.
O
__m
POWER DISTRIBUTION LIMITS DNBR MARGIN LIMITING CONDITION FOR OPERATION The DNBR margin shall be maintained by operating within the 3.2.4 region of acceptable operation of Figure 3.2-3 or 3.2-4, as applicable.
l APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With operation outside of the region of acceptable operation, as indicated by e'ither (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 DNBR' limit, within 15 minutes initiate corrective action to reduce the DNBR to within the limits and either:
Restore the DNBR to within its limits within one hour, or a.
b.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on all OPERABLE DNBR channels, is within the limit shown on Figure 3.2-3.
At least once per 31 days, the COLSS Margin Alarm shall be 4.2.4.3 verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR.
i ARKANSAS - UNIT 2 3/4 2-7 Amendment No. 24 l
l
~~ "'
~ --
m_. -
POWER DISTRIBUTION LItilTS 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 (%)
0-3.1 0
3.1-5 2.0 5-10 5.9 10-15 8.8 15-20 11.4 20-25 13.6 25-30 15.6 30-35 17.4 The penalty for each batch will be determined from the batch's maximum burnup assembly and applied to the batch's maximum radial power peak A single net penalty for COLSS and CPC will be determined assembly.
froia the penalties associated with each batch, accounting for the offsetting l
margins due to the lower radial power peaks in the higher burnup batches.
An alternate method is to detennine the penalty for each individual assenbly in the core based on that assembly's burnup and apply that penaly to that assembly's radial power peak.
1 i
l ARKANSAS - UNIT 2 3/4 2-8 Amendment No. 74, 26
O POWER DISTRIBUTION LIMITS BASES P
/P is the ratio of the power at a core location in the til t untilt presence of a tilt to the power at that location with no tilt.
3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences, of which the loss of flow transient is the most limiting.
Oper-ation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR.
Reactor operation at or below this calculated The power level assures that the limits of Figure 3.2-3 are not violated.
COLSS calculation of core ' power operating limit based on DNBR includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the core power at which a DNBR of less than 1.24 could occur, as calculated by COLSS, is less than or euqal to that which would actually be required in the core.
To ensure that the design margin to safety is maintained, the COLSS computer program includes an F measurement uncertainty factor of 1.053, an engineering uncertainty *Yactor of 1.03, a THERMAL POWER measurement uncertainty factor of 1.02 and appropriate uncertainty and penalty factors for flux peaking augmentation and rod bow.
Parameters required to maintain the margin to DNB and total core power are also monitored by the CPCs.
Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-4 can be maintained by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels.
The above listed uncertainty and penalty factors are also included in the CPC.
The DNBR penalty factors listed in section 4.2.4.4 are penalties used to accommodate the effects of rod bow.
The amount of rod bow in each individual fuel assembly is dependent upon the burnup experienced by that assembly.
Higher burnup assemblies will experience a higher degree of rod bow and should be assigned a higher penalty factor.
Conversely, low burnup assemblies will experience a lesser degree of rod bow and should be assigned a lower penalty factor.
ARKANSAS - UNIT 2 B 3/4 2-3 Amendment No. 27, 26
(
-s s-v x
-,w
,-=,.-r
, -+
POWER DISTRIBUTION LIMITS BASES 3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the LOCA safety analyses.
3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg tenperature is maintained within the range of values used in the safety analyses.
3/4.2.7 AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.
3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the range of value's used in the s'afety analyses.
l i
I l
i ARKANSAS - UNIT 2 B 3/4 2-4 Anendment No. 24
ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
The unit shail be placed in at least HOT STANDBY within a.
one hour.
b.
The Safety Limit violation shall be reported to the Commission, the Assistant Vice-President, Nuclear Operations, and to the SRC within 24 hourr.
A Safety Limit Violation Report shall be prepared.
The report c.
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 Assistant Vice-President, Nuclear Operations, within 14 days of the viol 6 tion.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
The applicable procedures recommended in Appendix "A" of a.
of Regulatory Guide 1.33, Revision 2, February 1978.
b.
Refueling operations.
c.
Surveillance and test activities of safety related equipnent.
d.
Security Plan implenentation.
e.
Emergency Plan impicaentation.
f.
Fire Protection Program implementation, Modification of Core Protection Calculator (CPC) Addressable g.
Constants NOTE:
Modification to the CP C addressable constants based on information obtained through the Plant Computer -
CPC data link shall not be made without prior approval of the Plant Safety Committee.
6.8.2 Each proccdure of 6.8.1 above, and changes thereto, shall be revieweo by the PSC and approved by the General Manager,nrior to inplenentation and reviewed periodically as set forth in administrative procedures.
ARKANSAS - UNIT 2 6-13 AmendmentNo.E,//,@9,25
_ ADMINISTRATIVE CONTROLS 6.8.3 Temporary chancas to procedur?s of 6.8.1 above may be made pro-vided:
.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.
The change is documented, reviewed by the PSC and approved by c.
the 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 Director of the Regional Office of Inspection and Enforcement unless othezuise 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 manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-mance 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 l
program and a comparison of these values with design predictions and speci fications.
Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details re.luired in license conditions based on other commitments shall be included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or comencement of commercial power operaticn, or (3) 9 months following initial criticality, 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 comencement of comercial 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. 5 L