ML20044H090
| ML20044H090 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 05/25/1993 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044H091 | List: |
| References | |
| NUDOCS 9306070376 | |
| Download: ML20044H090 (11) | |
Text
- pa Mc o
UNITED STATES E'
NUCLEAR REGULATORY COMMISSION 3.r.
h c
WASHING TON. D. C. 20555 o
9....,&
ILLINDIS POWER COMPANY. ET At.
DOCKET NO. 50-461 CLINTON POWER STATION. UNIT NO. 1 AMENDMENT TO FAClllTY OPERATING LICENSE Amendment No. 75 License No. HPF-62 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Illinois Power Company * (IP),
and Soyland Power Cooperative, Inc. (the licensees) dated February 11, 1993, 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 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
NPF-62 is hereby amended to read as follows:
- Illinois Power Company is authorized to act as agent for Soyland Power Cooperative, Inc. and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
9306070376 930525 PDR ADDCK 05000461 P
PDR (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 75, are hereby incorporated into this license.
Illinois Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
V MM Am James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 25, 1993
h i
ATTACHMENT TO LICENSE AMENDMENT NO. 75 FACILITY OPERATING LICENSE NO. NPF-62 P
I DOCKET NO. 50-461 1
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.
The corresponding overleaf pages, as indicated by an asterisk, are.provided to maintain document completeness.
Remove Pages Insert Paaes B 2-7 B 2-7 i
- B 2-8
- B 2-8 l
3/4 3-7 3/4 3-7
.l
- 3/4 3-8
- 3/4 3-8
- 3/4 3-9
- 3/4 3-9 3/4 3-10 3/4 3-10 5-5 5-5
- 5-6
- 5-6 l
t i
I f
b
?
k r
N P
Y
LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Averaae Power Ranae Monitor (Continued)
The APRM trip system is calibrated using heat balance data taken during steady-state conditions.
Fission chambers provide the basic input to the system and therefore, the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High setpoint; i.e; for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow-Biased Simulated Thermal Power-High setpoint, a time constant specified in the COLR is introduced into the flow-biased APRM in order to simulate the fuel thermal transient characteristics.
A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.
In these flow biased equations, the variable W is the loop recirculation flow as a percentage of the total loop recirculation flow which produces a rated core flow of 84.5 million lbs/hr.
The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.
3.
Reactor Vessel Steam Dome pressure-Hich High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase during operation will also tend to increase the power of the reactor by compressing voids, thus adding reactivity.
The trip will quickly reduce the neutron flux, counteracting the pressure increase.
The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips.
The setting provides for a wide margin to the maximum allowable design pressure and takes into account the lo'c'ation of the pressure measurement compared to the highest pressure that occurs in the system during a transient.
This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure and turbine control valve fast closure trips are bypassed.
For a turbine trip or. load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
4.
Reactor Vessel Water level-Low The reactor vessel water level trip setpoint has been used in transient analy-ses dealing with coolant inventory decrease. The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
CLINTON - UNIT 1 B 2-7 AMENDMENT NO. 75
LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 5.
Reactor Vessel Water Level-High A reactor scram from high reactor water level, approximately two feet above normal operating level, is intended to offset the addition of reactivity effect associated with the introduction of a significant amount of relatively cold feedwater.
An excess of feedwater entering the vessel would be detected by the level increase in a timely manner.
This scram feature is only effective when the reactor mode switch is in the Run position, because at THERMAL POWER levels below 10% to 15% of RATED THERMAL POWER, the approximate range of power level for changing to the Run position, the safety margins are more than adequate without a reactor scram.
6.
Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events.
The MSIV's are closed automatically from measured parameters such as high steam flow, high steam line radiation, low reactor water level, high steam tunnel temperature and low steam line pressure.
The MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.
7.
Main Steam Line Radiation-High The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding.
When the high radiation is detected, a trip is initiated to reduce the continued failure of fuel cladding.
At the same time, the main steam line isolation valves are closed to limit the release of fission products.
The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures iy.the fuel cladding.
8.
Drywell Pressure-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or loss of drywell cooling.
The reactor is tripped in order to minimize the possit'ility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment.
The trip setting was selected as low as pussible to minimize heat loads of equipment located within the primary containment and to avoid spurious trips.
CLINTON - UNIT 1 B 2-8
L TABLE 3.3.1-2 8
REACTOR PROTECTION SYSTEM RESPONSE ilMES 5o 7
RESPONSE TIME FUNCTIONAL UNIT (Seconds) c 5'
1.
~
a.
Neutron Flux - High NA b.
Inoperative NA 2.
Average Power Range Monitor *:
a.
Neutron Flux - High, Setdown NA b.
Flow Biased Simulated Thermal Power - High 5 0.09**
c.
Neutron Flux - High 5 0.09 d.
Inoperative NA 3.
Reactor Vessel Steam Dome Pressure - High 5 0.33 4.
Reactor Vessel Water Level - Low, level 3 5 1.03 w1 5.
Reactor Vessel Water Level - High, level 8 s 1.03 6.
Main Steam Line Isolation Valve - Closure 5 0.04 w
i 4,
7.
Main Steam Line Radiation - High-NA
-8.
Drywell Pressure - High NA 9.
-Scram Discharge Volume Water Level - High a.
Level Transmitter NA b.
Float Switches NA-10.
Turbine Stop Valve - Closure 5 0.04 11.
Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low s 0.05,
- 12. Reactor Mode Switch Shutdown Position NA j
- 13. Manual Scram-NA g
ii5
- Neutron detectors are exempt from response time testing.
Response time shall be measured from the.
g detector output or from the input of the first electronic component in the' channel.
{
- Not including a simulated thermal power time constant specified in'the COLR.
- Measured from start of turbine-control valve fast closure.
IABLE 4. 3.1.1-1 0y REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVElllANCE REQUIREMENTS E
CilANNEL OPERATIONAL CllANNLL FUNCTIONAL CHANNEL CONDITIONS IN WHICH Q
FUNCTIONAL UNIT CilECK TEST CAllBRATION SURVEILLANCE REQUIRED 1.
IC) a.
Neutron Flux - High S/U,5,(b)
S/U
,W R
2 S
W R
3,4,5 b.
Inoperative NA W
NA 2,3,4,5 2.
Average Power Range Monitor: III IC) a.
Neutron Flux - High, S/U.S,(b)
S/U
,W SA 2
Setdown S
W SA 3, 4, 5 b.
Flow-Blased Simulated
{
Thermal Power - High S
S/U
,Q W(d)(e) 34, g(1) y IC)
IC)
Id)I'}
4 c.
Neutron Flux - High S
S/U
,Q W
SA.
I d.
Inoperative NA
'Q NA 1,2,3,4,5 3.
Reactor Vessel Steam Dome I9)
Id}
Pressure - High S
Q R
1, 2 4.
II)
Low, Level 3 S
Q R
1, 2 k
5.
- n.$
II)
High, level 8 S
Q R
1
- ?.
6.
Main Steam Line Isolation Valve - Closure NA Q
R 1
,8wg 7.
Main Steam Line Radiation -
High 5
Q R
1 '2(y)
II) 1,2(I) 8.
firywell Pressure - High
-S Q
R
. -, _ ~,.
,_,-s-w-
.,,,,,..,,s,-.
._,,.,....,y-
,,,,._,.%.M.
,,,...,.,.m
,,s..
E nLE 4.3.1.1 'l (Continued) p REACTOR PROTECil0N SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
i$
2 CilANNfL OPERATIONAL CilANNE L FUNCTIONAL CilANNEL CONDITIONS IN WHICH g
FUNCTIONAL UNIT CllECK IEST CALIBRATION (*)
SURVEILLANCE REQUIRED 4
9.
Scram Discharge Volume Water g
Level - High I9)
Ik) a.
Level Transmitter S
Q R
1, 2, S Ll b.
Float Switches NA Q
R 1, 2, SIk)
I Q ")
10.
Turbine Stop Valve - Closure NA R ")
I 1
11.
Turbine Control Valve Fast Closure Valve Trip System 011 Q ")
I R ")
I Pressure - Low NA 1
l-2 Y
12.
Reactor Mode Switch
[
Shutdown Position NA R
NA 1,2,3,4,5 13.
Manual Scram NA Q
NA 1,2,3,4,5 b
a
$E
-.c c.,._
-.m..-...
-,---e
..n,,..
.-.--..~.-.ww..
-r,,.
~ -._.,. _,
,,,.,.-,s....~-----__.----_._LL-'-,a
TABLE 4.3.1.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE RE0VIREMLNTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
i (b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decade during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least I decade during each controlled shutdown, if not performed within the.
previous 7 days.
(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2 25% of RATED THERMAL POWER.
Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.
(e) This calibration shall consist of a setpoint verification of the Neutron Flux-High and the Flow Biased Simulated Thermal Power-High trip functions.
The Flow Biased Simulated Thermal-High trip function is verified using a calibrated flow signal.
(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.
(g) Calibrate the analog trip module at least once per 92 days.
f (h) Deleted.
l (i) This calibration shall consist of verifying that the simulated thermal power time constant is within the limits specified in the COLR.
(j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(k) With any control rod withdrawn. Not applicable to control rods removed i
per Specification 3.9.10.1 or 3.9.10.2.
j (1) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required to be OPERABLE per Special Test Exception 3.10.1.
(m) The CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION shall include the turbine first stage pressure instruments.
i i
CLINTON - UNIT 1 3/4 3-10 Amendment No. 75
{
~
1 DESIGN FEATURES SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the Fuel Building, the ECCS pump rooms and the containment gas control boundary, including extension, and has a h
minimum free volume of 1,710,000 cubic feet.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 624 fuel assemblies.
Each assembly shall consist of a matrix of zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel. material, and water rod (s).
Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases.
A limited number of lead-use assemblies that have not completed representative testing may be placed in non-limiting core regions.
CONTROL R0D ASSEMBLIES 5.3.2 The reactor core shall contain 145 cruciform shaped control rod assemblies as approved by the NRC. The control material shall be boron carbide powder (B,C) and/or hafnium metal.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURF, 5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the
~:
FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of:
1.
1250 psig on the suction side of the recircuiation pump.
2.
1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.
3.
1550 psig from the discharge shutoff valve to the jet pumps.
c.
For a temperature of 575'F.
VOLUME i
5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 16,000 cubic feet at a nominal steam dome saturation temperature of 549'F.
CLINTON - UNIT 1 5-5 AMENDMENT NO. 75
DESIGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with:
A k,7f equivalent to less than or equal to 0.95 when flooded with unborated a.
water, including all calculational uncertainties and biases as described in Section 9.1.2 of the FSAR.
b.
A nominal 6.4375 inch center-to-center distance between fuel assemblies placed in the storage racks in the Fuel Building storage pool.
A nominal center-to-center spacing between rows of 12.25 inches and within the rows of 7.00 inches for fuel assemblies placed in the storage rack in the Upper Containment Fuel Pool.
- 5. 6.1.1 The k,ff for new fuel for the first core loading Stored dry in the spent fuel storage' racks shall not exceed 0.98 when aqueous foam moderation is assumed.
DRAINAGE r
5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 754'0".
t 2
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2522 fuel assembifes.
1 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT
- 5. 7.1 The components identified in Table 5.7.1-1 are designed end shall be 1
maintained within the cyclic or transient limits of Table 5.7.1-1.
6 f
I CLINTON - UNIT 1 5-6 i
y-
,