ML20214F992
| ML20214F992 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/14/1986 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Arkansas Power & Light Co |
| Shared Package | |
| ML20214F994 | List: |
| References | |
| DPR-51-A-104 NUDOCS 8611250516 | |
| Download: ML20214F992 (10) | |
Text
o UNITED STATES 8
g NUCLEAR REGULATORY COMMISSION o
y WASHINGTON, D. C. 20555
%...../
ARKANSAS POWER AND LIGHT COMPANY DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No.
1 04 License No. DPR-51 1.
The huclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendments by Arkansas Power and Light Company (thelicensee)datedJuly 18, 1986 and July 31, 1986, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the applications, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by b
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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
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. DPR-51 is hereby amended to read as follows:
8611250516 861114 PDR ADOCK 05000313 P
PDR i
\\
. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.104, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications..
3.
This license amendment is effective as of its date of issuance.
F THE NUCLEAR REG ATO COMMISSION oh 7. Stolz, Direr,
Project Direct' e #6 Division of PWR Licer ;ing-B
Attachment:
Changes to the Technical Specifications Date of Issuance: November 14, 1986
ATTACHMENT TO LICENSE AMENDMENT NO.104 FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 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.
Remove Insert 10 10 13 13 14a 14a 15 15 42a 4?a 43a 43a 45f 45f s
2.2 SAFETY LIMITS - REACTOR SYSTEM PRESSURE Applicability Applies to the l'imit on reactor coolant system pressure.
Objective
+
To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.
Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.
2.2.2 The setpoint.of the pressurizer code safety valves shall be in accordance with ASME, Boiler and Pressurizer Vessel Code, Section 3
III, Article 9, Sammer 1968.
Bases The reactor coulant system (1) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere.
In the event of a fuel cladding failure, the reactor. coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME code,Section III, is 110 percent of design pressure.(2) The maximum
'5 transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110 percent of design pressure.
Thus, the safety limit of 2750
, pressure) has been established.ggjg (110 percent of the 2500 psig design The settings for the reactor high pressure trip (2355 psig) and the pressurizer code safety valves (2500 psig 11%) (3) have been established to assure that the reactor coolant system l
pressure safety limit is not exceeded. The initial hydrostatic test is conducted at 3125 psig (125 percent of design pressure) to verify the l
integrity of the reactor coolant system. Additional assurance that the reactor coolant system pressure does not exceed the safety limit is provided l
by setting the pressurizer electromatic relief valve at 2450 psig.(4)
REFERENCES (1)
FSAR, Section 4 (2)
FSAP., Section 4.3.10.1 l
(3).
FSAR, Section 4.2.4 (4)
FSAR, Table 4-1 Amendment No. 44/,104 10
punp(s). The pump monitors also restrict the power level for the nember of pumps in operation.
C.
RCS Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip is reached before the nuclear overpower trip setroint.
The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient.(2)
The low pressure (1800.psig). and variable low pressure (11.75T
-5103) trip setpointsshown.in Figure 2.3-1 have been establishE8g to maintain the DN8 ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.(a,s)
Due to the calibration and instrumentation errors, the safety analysis used a variable low reactor coolant system pressure trip i,
value of (11.75t
-5143).'
ot O.
Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (618F) shown in Figure 2.3-1 has been established to pre. vent excessive core coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620F.
E.
Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
F.
Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there i_s provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1.
Two conditions are imposed when the bypass is used:
1.
A nuclear overpower trip setpoint of 55.0 percent of rated power is automatically imposed during reactor shutdown.
2.
A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.
Amendment No. 2,21,49 J7;,104 13
'- ;:.,4.,
2500 P:2355 PSIG T=618 'F 2365 2300 o
ACCEPTABLE OPERATION ED 2100 Ca.
P:(11.75 T
-5103) PSIG og a
8
[
o 1900 UNACCEPTAGl.E g-OPERATION r
E P=1800 PSIG 1700 i
1500 560 580 600 620 640 660 REACTOR OUTLET TEMPERATURE.*F PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINT FIGURE 2.3-1 Am2ndment No. 27,#,g7,104 14a e
.---r--o,,
--w--,..,, -
.m
-,-m
---r-ne,,-,m.
---,---,w---
w-
Tabla 2.3-1 Reactor Protection System Trip Settino Limits i
J One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop I
Operating (Nominal Operating (Nominal (Nominal Operating Shutdo,wn Operating Power - 100%)
_ Operat'ing Power - 75%)
Power - 49%)
Bypass i
Nuclear power, % of 104.9 104.9 104.9 5.0(a) rated, max Nucigar Power based on 1.07 times flow minus 1.07 times flow minus 1.07 times flow minus Bypassed flow and imbalance, reduction due to reduction due to reduction due to
% of rated, max imbalance (s) imbalance (s) imbalance (s)
Nuclear Power based on NA NA 55 pump monitogs, % of Bypassed rated, max
)
High RC system 2355 2355 2355 17208 pressure, psig, max Low RC system 1800 1800 pressure, psig. min 1800 Bypassed Variable low RC 11.75 T
-5103 11.75 T
-5103 d
d d
11.75 T"" -5103 Bypassed system pressure, psig, min 1
RC temp, F, max 618 618 618 618 i
High reactor building 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 prassure, psig, max psia)
" Automatically set when other segments of the RPS (as specified) are bypassed, bR actor coolant system flow.
C l
Tha pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two pump operation.
dT is given in degrees Fahrenheit (F) cut Amehdment No. 2,2I,43,49,52,57,$e.104 15
- r,.
T e
i
- 3. 5.1. 7 The Decay Heat Removal System isolation valve closure setpoints shall be equal to or less than 340 psig for one valve and equal to or less than 400 psig for the second valve in the suction line.
The relief valve setting for the DHR system shall be equal to or less than 450 psig.
3.5.1.8 The degraded voltage monitoring relay settings shall be as follows:
The 4.16 KV emergency bus undervoltage relay setpoints 1. hall a.
be >3115 VAC but (3177 VAC.
b.
The 460 V emergency bus undervoltage relay setpoints shall be
> 423 VAC but (431'VAC with a time delay setpoint of 8 seconds il second.
- 3. 5.1. 9 The following Reactor Trip circuitry shall be operable as indicated:
1.
Reactor trip upon loss of Main Feedwater shall be operable (as determined by Specification 4.1.a and item 35 of Table 4.1-1) at greater than 5% reactor power.
(May be bypassed up to 10% reactor power.)
2.
Reactor trip upon Turbine Trip shall be operable (as i
det' ermined by Specification 4.1.a and item 41 of Table 4.1-1)
I at greater than 5% reactor power.
(May be bypassed up to 45%
,l reactor power.)
3.
If the requirements of Specifications 3.5.1.9.1 or 3.5.1.9.2 cannot be met, restore the inoperable trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or bring the plant to a hot shutdown condition.
3.5.1.10 The control room ventilation chlorine detection system instrumentation shall be operable and capable of actuating control room isolation and filtration systems, with alarm / trip setpoints adjusted to actuate at a chlorine concentration of $5 ppm.
1 3.5.1.11 For on-line testing of the Emergency Feedwater l
Initiation and Control (EFIC) system channels during power operation only one channel shall be locked into " maintenance bypass" at any one time.
If one channel of the NI/RPS is in j
malatenance bypass, only the corresponding channel of EFIC may be
{
bypassed.
3.5.1.12 The Containment High Range Radiation Monitoring instrumentation shall be operable with a minimum measurement range from 1 to 107 i
R/hr.
Amendment No. 60 6I, 69, 91, 94, 42a 104
for protective action from a digital ESAS subsystem will not cause that subsystem to trip.
The fact that a module has been removed will be continuously annunciated to the operator.
The redundant digital subsystem is still sufficient to indicate complete ESAS action.
The testing schemes of the RPS, the ESAS, and the EFIC enables complete system testing while the reactor is operating.
Each channel is capable of
'being tested independently so that operation of individual channels may be evaluated.
The EFIC is designed to allow testing during power operation. One channel may be placed in key locked " maintenance bypass" prior to testing. This will bypass only one channel of EF.W initiate logic. An interlock feature prevents bypassing more than one channel at a time.
In addition; since the EFIC receives signals from the NI/RPS, the maintenance bypass from the NI/R,PS is interlocked with the EFIC.
If one channel of the NI/RPS is in maintenance bypass.ionly the corresponding channel of EFIC may be bypassed.
The EFIC can be tested from its input terminals to the actuated device controllers. A test of the EFIC trip logic will actuate one of two relays in the controllers.
Activation of both relays is required in order to actuate the controllers. The two relays are tested individually to prevent automatic actuation of the component.
The EFIC trip logic is two (one out-of-two).
Reactor trips on loss of all main feedwater and on turbine trips will sense the start of a loss of OTSG heat sink and actuate earlier than other trip signals.
This early actuation will provide a lower peak RC pressure during the initial over pressurization following a loss of feedwater or turbine s.
trip event. The LOFW trip may be bypassed up to 10% to allow sufficient margin for brir.ging the MFW pumps into use at approximately 7%.
The Turbine Trip trip may be bypassed up to 45% based on BAW-1893, " Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip," October 1985 and the NRC Safety Evaluation Report for BAW-1893 issued from Mr. D. M.
Crutchfield to Mr. J. H. Taylor via letter dated April 25, 1986.
The Automatic Closure and Isolation System.(ACI) is designed to close the Decay Heat Removal System (OHRS) return line isolation valves when the Reactor Coolant System (RCS) pressure exceeds a selected fraction of the DHRS design pressure or when core flooding system isolation valves are opened. The ACI is designed to permit manual operation of the DHRS return line isolation valves when permissive conditions exist.
In addition, the ACI is designed to disallow manual operation of the valves when permissive conditions do not exist.
Power is normally supplied to the control rod drive mechanisms from two separate parallel 480 volt sources.
Redundant trip devices are employed in each of these sources.
If any one of these trip devices fails in the untripped state, on-line repairs to the failed device, when practical, will be made and the remaining trip devices will be tested.
Four hours is ample time to test the remaining trip devices and, in many cases, make on-line repairs.
Amendment No. 50, 60, 6I, N1,104 43a i
[
.--.s..
E i
E N
e.
g TABLE 3.5.1-1 (Cont'd) l ?"
With the number of operable channels less than required, either return the indicator to operable status rn 12.
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or verify the block valve closed and power removed within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If the block valve cannot be verified closed within the additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, de-energize the electromatic wn
- 2' relief val,ve power supply within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
, n.
?}
13.
Channels may be bypassed for not greater than 30 seconds during reactor coolant pump starts.
If the automatic bypass circuit or its alarm circuit is inoperable, the undervoltage protection shall be wa C"
restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, otherwise, Note 14 applies.
x l'
14.
With the number of channels less than required, restore the inoperable channels to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 e
hours.
15.
This trip function may be bypassed at up to 10% reactor power.
16.
This trip function may be bypassed at up to 45% reactor power.
17.
With no channel operable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the inoperable channels to operable status, or initiate and maintain operation of the control room emergency ventilation system in the recirculation rode of operation.
18.
With one channel inoperable, restore the inoperable channel to operable status within 7 days or within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.
19.
This function may be bypassed below 750 psig OTSG pressure.
Bypass is autumatically removed when pressure exceeds 750 psig.
20.
With one channel inoperable, (1) either restore the inoperable channal to operable status within 7 days, or (2) prepare and submit a Special Report to the Commission. pursuant to Specification 6.12.4 within 30 days following the event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status. With both channels inoperable, initiate alternate methods of monitoring the containment radiation level within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in addition to the actions described above.
21.
With one channel inoperable, restore the inoperable channel to ope.rable staus within 30 days or be in hot shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless containment entry is required.
Tf containment entry is required, the inoperable channel must be restored by the next refueling outage.
If both channels are inoperable, restore the inoperable channels within 30 days or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- C'