ML072880418
| ML072880418 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/15/2007 |
| From: | NRC/NRR/ADRO/DORL/LPLI-2 |
| To: | |
| Bamford, Peter J., NRR/DORL 415-2833 | |
| Shared Package | |
| ML072680011 | List: |
| References | |
| TAC MD4910 | |
| Download: ML072880418 (7) | |
Text
(2)
AmerGen Energy Company. LLC, pursuant to the Act and 10 CFR Parts 30.40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as reactor fuel, sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required for reactor operation; (3)
AmerGen Energy Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess at either TMI-I or TMI-2, and use in amounts as required for TMI-I any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis, testing, instrument calibration, or associated with radioactive apparatus or components. Other than radioactive apparatus and components to be used at TMI Unit 2 in accordance with the TMI-2 License, the radioactive apparatus and components that may be moved from TMI Unit I to TMI Unit 2 under this provision shall be limited to: (l) outage-related items (such as contaminated scaffolding, tools, protective clothing, portable shielding and decontamination equipment); and (2) other equipment belonging to TMI Unit I when storage of such equipment at TMI-2 is deemed necessary for load handling or contamination control considerations; (4)
AmerGen Energy Company, LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess at the TMI Unit I or Unit 2 site, but not separate, such byproduct and special nuclear materials as may be produced by the operation of either unit. Radioactive waste may be moved from TMI Unit 2 to TMI Unit I under this provision for collection, processing (including decontamination), packaging, and temporary storage prior to disposal. Radioactive waste that may be moved from TMI Unit 1 to TMI Unit 2 under this provision shall be limited to: (1) dry active waste (DAW) temporarily moved to TMI Unit 2 during waste collection activities, and (2) contaminated liquid contained in shared system piping and tanks. Radioactive waste that may be moved from TMI Unit 1 to TMI Unit 2 under this provision shall not include spent fuel, spent resins, filter sludge, evaporator bottoms, contaminated oil, or contaminated liquid filters.
The storage of radioactive materials or radwaste generated at TMI Unit 2 and stored at TMI Unit I shall not result in a source term that, if released, would exceed that previously analyzed in the UFSAR in terms of offsite dose consequences.
The storage of radioactive materials or radwaste generated at TMI Unit I and stored at TMI Unit 2 shall not result in a source term that, if released, would exceed that previously analyzed in the PDMS SAR for TMI Unit 2 in terms of off-site dose consequences.
- c.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(I)
Maximum Power Level AmerGen Energy Company, LLC is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A. as revised through Amendment No2 6 2 are hereby incorporated in the license. The AmerGen Energy Company, LLC shall operate the facility in accordance with the Technical Specifications.
Amendment No. 262
LIST OF TABLES TABLE TITLE PAGE 1.2 Frequency Notation 1-8 2.3-1 Reactor Protection System Trip Setting Limits 2-10 3.1.6.1 Pressure Isolation Check Valves Between the 3-15a Primary Coolant System and LPIS 3.5-1 Instruments Operating Conditions 3-29 3.5-1A DELETED 3.5-2 Accident Monitoring Instruments 3-40c 3.5-3 Post Accident Monitoring Instrumentation 3-40d 3.5-4 Remote Shutdown System Instrumentation and Control 3-40i 3.21-1 DELETED 3.21-2 DELETED 3.23-1 DELETED 3.23-2 DELETED 4.1-1 Instrument Surveillance Requirements 4-3 4.1-2 Minimum Equipment Test Frequency 4-8 4.1-3 Minimum Sampling Frequency 4-9 4.1-4 Post Accident Monitoring Instrumentation 4-10a 4.19-1 DELETED 4.19-2 DELETED 4.21-1 DELETED 4.21-2 DELETED 4.22-1 DELETED 4.22-2 DELETED 4.23-1 DELETED vi Amendment No.
2 00, 6nn 137,142, 447,
.n 473, 97-, 2!T-- 262
2500 23004
& 2100 1900 6
U, UNCCPTBL OPRAIO I--
7 - - - - - - - - - - -
I --------
1700 +---
1500 590 600 610 620 630 640 650 Reactor Outlet Temperature (0F)
CORE PROTECTION SAFETY LIMIT TMI-1 FIGURE 2.1-1 2-4a Amendment No. 50, 142, 167; 211, 238, 2-7, 262
TABLE 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS (5)
Four Reactor Coolant Pumps Operating (Nominal Operating)
Power - 100%
Three Reactor Coolant Pumps Operating (Nominal Operating)
Power - 75%
One Reactor Coolant Pump Operating in Each Loop (Nominal OperatinQ Power - 49%)
Shutdown Bypass
- 1. Nuclear power, max. %
of rated power
- 2. Nuclear power based on flow (1) and imbalance max. of rated power
- 3. Nuclear power based (4) on pump monitors max.
% of rated power
- 4. High reactor coolant system pressure, psig max.
- 5. Low reactor coolant system pressure, psig min.
- 6. Reactor coolant temp. F.,
max.
- 7. High Reactor Building pressure, psig max.
- 8. Variable low reactor coolant system pressure, psig min.
105.1 105.1 105.1 5.0(2)
Power/Flow Setpoint in COLR times flow minus reduction due to imbalance Power/Flow Setpoint in COLR times flow minus reduction due to imbalance Power/Flow Setpoint in COLR times flow minus reduction due to imbalance Bypassed NA NA 55%
2355 Bypassed 2355 1900 618.8 2355 1900 1720(3) 1900 Bypassed 618.8 618.8 618.8 4
4 4
4 (16.21 Tout - 7973)(6)
(16.21 Tout - 7973)(6)
(16.21 To t-7973)(6)
Bypassed (1)
(2)
(3)
(4)
(5)
(6)
Reactor coolant system flow, %
Administratively controlled reduction set during reactor shutdown.
Automatically set when other segments of the RPS (as specified) are bypassed.
The pump monitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation.
Trip settings limits are limits on the setpoint side of the protection system bistable connectors.
Tout is in degrees Fahrenheit (F).
2-10 Amendment No. 45, 78, 90, 126, 135, 112, 184, 247, 262
2500 2300 I
T=618.8 OF SACCEPTABLE
- 1 (OPERATION
'U I
I 0.
I-2100 I
up LP=2Tw - 7973 psig
- a.
I IIL 0
U P
1900psig 1900 o
p a,
0 E1l
- UNACCEPTABLE OPERATION 1700 I
1500 1
1 540 560 580 B00 620 640 Reactor Outlet Temperature, OF PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS TMI-1 FIGURE 2.3-1 2-11 Amendment No 13.7. 2, 31.1
, v7-1 78, 4 26,365
- 142, 1 2 17,2 17, 262
TABLE 4.1-1 (Continued) 3 CD CD z3 0
ro CHANNEL DESCRIPTION
- 8.
High Reactor Coolant Pressure Channel
- 9.
Low Reactor Coolant Pressure Channel
- 10. Flux-Reactor Coolant Flow Comparator
- 11. Reactor Coolant Pressure-Temperature Comparator
- 12. Pump Flux Comparator
- 13. High Reactor Building Pressure Channel
- 14. High Pressure Injection Logic Channels
- 15. High Pressure Injection Analog Channels
- a. Reactor Coolant Pressure Channel
- 16. Low Pressure Injection Logic Channel
- 17. Low Pressure Injection Analog Channels
- a. Reactor Coolant Pressure Channel
- 15. Reactor Building Emergency Cooling and Isolation System Logic Channel CHECK S
TEST S/A CALIBRATE R
REMARKS S
S/A S
S/A S
S/A R
F R
See Notes (a) and (b).
S S
NA S(1)
S/A S/A Q
M R
F NA R
(1) When reactor coolant system is pressurized above 300 psig or Tare is greater than 200°F NA NA 0
0 S(1)
NA M
Q R
(1) When reactor coolant system is pressurized above 300 psig or T.ve is greater than 200OF NA
TABLE 4.1-1 (Continued)
CD 3
CD z
0 r
A D
A 4
I, 13 CHANNEL DESCRIPTION
- 49. Saturation Margin Monitor
- 50. Emergency Feedwater Flow Instrumentation
- 51. Heat Sink Protection System
- a. EFW Auto Initiation Instrument Channels
- 1. Loss of Both Feedwater Pumps
- 2. Loss of All RC Pumps
- 3. Reactor Building Pressure
- 4. OTSG Low Level
- c. EFW Control Valve Control System
- 1. OTSG Level Loops
- 2. Controllers
- d. HSPS Train Actuation Logic
- 52. Backup Incore Thermocouple Display CHECK TEST CALIBRATE S(1)
M(1)
R NA M(1)
F (1) When T,,, is greater than 525 0F.
(1) When Tave is greater than 2500F.
REMARKS (1) Includes logic test only.
NA Q(1)
NA Q(1)
NA Q
W Q
NA Q
F R
F R
R R
R R
W W
Q NA NA Q(1)
M(1)
NA R
(1) When Tave is greater than 2500F.
- 53. Deleted
- 54. Reactor Vessel Water Level NA NA R
Notes (a) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. Enter condition into Corrective Action Program.
(b) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conversative than the NSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures to confirm channel performance. The NSP and the methodologies used to determine the as-found and the as-left tolerances are specified in a document incorporated by reference into the UFSAR.