ML20101R157
| ML20101R157 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/10/1996 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20101R139 | List: |
| References | |
| NUDOCS 9604160076 | |
| Download: ML20101R157 (7) | |
Text
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TABLE 4.1-1
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INSTRUMENT SURVEILLANCE REQUIREMENTS l
CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS nF 1.
Pmtection Channel NA M
NA j3 Coincidence logic j! 2.
Control Rod Drive Trip NA M
NA (1)
Includes independent testing of shunt Bmaker and Regulating trip and
- S undervoltage trip features. Rod Power i
SCRs 3.
Power Range Amplifier D(1)
N (2)
(1)
When reactor power is greater than 15%.
1 e
(2)
When above 15% mactor power run a heat balance check once per shift. Heat balance calibration shall be performed whenever heat balance exceeds indicated neutron power by more than two percent.
4.
Power Range Channel S
M M(1)(2)
(1)
When reactor power is greater than 60%
verify imbalance using incore instrumentation.
(2)
When above 15% reactor power calculate i
axial offset upper and lower chambers after each startup if not done within the previous seven days.
5.
Intermediate Range Channel S(l)
PS/U NA (1)
When in service.
6.
Source Range Channel S(l)
PS/U NA (1)
When in service.
7.
Reactor Coolant Temperature S
M R
Channel 9604160076 960410 PDR ADOCK 05000289 P
3:
h TABLE 4.1-1 (Continued) 3:
CHANNEL DESCRIlrTION CHECK TEST CALIBRATE REMARKS y 19.
Reactor Building Emergency Cooling and Isolation o
C System Analog Channels J
j a.
Reactor Building S(l)
M(1)
R (1) When CONTAINMENT INTEGRITY is required.
L 4 psig Channels j3 b.
RCS Pressure 1600 psig S(l)
M(1)
NA (1) When RCS Pressure > 1800 psig.
- 3 c.
Deleted la d.
Reactor Bldg. 30 psig S(l)
M(1)
R (1) When CONTAINMENT INTEGRITY is required.
M pressure switches
{
e.
Reactor Bldg. Purge W(1)
M(1)(2)
F (1) When CONTAINMENT INTEGRITY is required.
lg Line High Radiation (AH-V-1A/D) f.
Line Break Isolation W(1)
M(1)
R (1) When CONTAINMENT INTEGRITY is required.
Signal (ICCW & NSCCW) 20.
Reactor Building Spray NA Q
NA y
System Logic Channel 21.
Reactor Building Spray NA M
R 30 psig pressure switches 22.
Pressurizer Temperature S
NA R
Channels 23.
Control Rod Absolute Position S(l)
NA R
(1) Check with Relative Position Indicator.
24.
Control Rod Relative Position S(l)
NA R
(1) Check with Absolute Position Indicator.
25.
Core Flooding Tarls a.
Pressure Channels S(l)
NA R
(1) When Reactor Coolant system pressure is greater j than 700 psig.
- b. Level Channels S(l)
NA R
26.
Pressurizer Level Channels S
NA R
>K Eo K
TABLE 4.1-1 (Continued) k o
CHANNEL DESCRIIYrION CHECK TEST CALIBRATE REMAR_KS b
30.
Borated Water Storage W
NA R
Tank I.evel Indicator 31.
Boric Acid Mix Tank:
DELETED 32.
Reclaimed Boric Acid Storage Tank:
DELETED 33.
Containment Temperature NA NA F
34.
Incore Neutron Detectors M(1)
NA NA (1)
Check functioning; including functioning
[
of computer readout or recorder readout when reactor power is grater than 15%.
6 35.
Emergency Plant Radiation M(1)
NA F
(1)
Battery check.
Instmments 36.
Strong Motion Accelerometer Q(1)
NA Q
(1)
Battery check.
37.
Reactor Building Sump NA NA R
Ixvel 6
n
m TABLE 4.1-1 (Continued)
Kl CHANNEL DESCRIPTION CHECK TEST CAllBRATE REMARKS 8 38.
OTSG Full Range Ievel W
NA R
is
]g 39.
Turbine Overspeed Trip NA R
NA 3 40.
BWST/NaOH Differential NA NA F
Pressure Indicator g
-i 41.
Sodium Hydroxide Tank NA NA F
j Ievel Indicator a
42.
Diesel Generator Protective NA NA R
Relaying 43.
4 KV ES Bus Undervoltage Relays (Diesel Start)
- a. Degraded Grid NA M(1)
R (1) Relay operation will be checked by local test pushbuttons.
- b. Loss of Voltage NA M(1)
R (1) Relay operation will be checked by local test pushbuttons.
44.
Reactor Coolant Pressure S(l)
M R
(1) When reactor coolant system is DH Valve Interlock Bistable pressurized above 300 psig or T,y, is greater than 200 F.
45.
Loss of Feedwater Reactor Trip S(l)
M(1)
R (1) When reactor power exceeds 7% power.
46.
Turbine Trip / Reactor Trip S(l)
M(1)
F (1) When reactor power exceeds 45%
power.
47.
- a. Pressurizer Code Safety Valve S(l)
NA R
(1) When T,y, is greater than 525 F.
l and PORV Tailpipe Flow Monitors
- b. PORV - Acoustic / Flow NA M(1)
R (1) When T,y, is greater than 525 F.
48.
PORV Setpoints NA M(1)
R (1) Per Specification 3.1.12 excluding valve operation.
1
3:
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g
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TABLE 4.1-3 Cont'd z
o ja Item Check Fmquency l8 jg 4.
Spent Fuel Pool Water Sample Boron concentration greater than Monthly and after each makeup.
jg or equal to 600 ppmb Jo t
ij 5.
Secondary Coolant System Isotopic analysis for DOSE At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when Activity EQUIVALENT I-131 concentration reactor coolant system pmssure j.
is gmater than 300 psig or Tav is Ja ll gmater than 200*F.
- 6. Deleted 7.
Deleted t
8.
Deleted 9.
Deleted i
- 10. Sodium Hydroxide Tank Concentration Semi-Annually and after each makeup.
- 11. Deleted i
l
- 12. Deleted t'
Until the specific activity of the primary coolant system is mstored within its limits.
i Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the reactor was last subcritical fcr 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
f 1.--.
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TABLE 4.1-4 H
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o POST ACCIDENT MONITORING INSTRUMENTATION
~
FUNCTION INSTRUMENTS CHECK _TE_ST CALIBRATE REMARKS
- t h
- 1. Noble Gas Effluent m
3 a.
Condenser Vacuum Pump Exhaust W M
F (1) Using the installed check source when background is (RM-A5-Hi) less than twice the expected increase in cpm which would msult from the check source alone. Background readings I
greater than this value are sufficient in themselves to show that this monitor is functioning.
b.
Condenser Vacuum Pump Exhaust W(1)
M F
(RM-G25) 9 c.
Auxiliary and Fuel Handling W
M F
Building Exhaust (RM-A8-Hi) d.
Reactor Building Purge Exhaust W
M F
(RM-A9-Hi) e.
Reactor Building Purge Exhaust W(1)
M F
(RM-G24) f.
Main Steam Lines Radiation W(1)
M F
(RM-G26/RM-G27) 2.
Containment High Range Radiation W
M R
(RM-G22/G23) 3.
Containment Pmssure W
N/A R
4.
Containment Water Level W
N/A R
5..
Containment Hydrogen W
M F
6.
Wide Range Neutron Flux W
N/A F
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION CERTIFICATE OF SERVICE IN THE MA' ITER OF DOCKET NO, 50-289 GPU NUCLEAR CORPORATION LICENSE NO. DPR 50 This is to certify that a copy of Technical Specification Change Request No. 243 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addmssed as follows:
I l
Mr. Darryl LeHew, Chairman Mr. Russell L. Sheaffer, Chairman Board of Supervisors of Board of County Commissioners j
Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 l
l Director, Bureau of Radiation Protection PA. Dept of Environmental Resources Rachael Carson State Office Building,13th Floor P.O. Box 8469 Harrisburg, PA 17105-8469 l
Att: Mr. Stanley P. Maingi i
GPU NUCLEAR CORPORATION j
BY:
ice President and Director, TMI DATE:
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