ML20071N392
| ML20071N392 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/01/1994 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20071N397 | List: |
| References | |
| DPR-50-A-191 NUDOCS 9408050261 | |
| Download: ML20071N392 (7) | |
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UNITED STATES yo j
j NUCLEAR REGULATORY COMMISSION l
t WASHINGTON, D.C. 20555 0001 4
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9.....g METROPOLITAN EDISON COMPANY 1
JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY i
GPU NUCLEAR CORPORATION DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 191 License No. DPR-50 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.
(the licensee), dated July 15, 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; l
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 l
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; t
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 i
of the Commission's regulations and all applicable requirements have been satisfied.
I i
f 9408050261 940001 PDR ADOCK 05000289 P
l 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-50 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as t
revised through Amendment No.191, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION j?
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l J6 F. Stolz, Directo ject Directorate If4 ivision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 1,1994 i
ATTACHMENT TO LICENSE AMENDMENT NO. 191 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 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 iii iii 3-128 3-128 3-129 3-129 4-7a 4-7a i
i
TABLE OF CONTENTS Section Pace 3.16 SH0CK SUPPRESSORS (SNUBBERS) 3-63 3.17 REACTOR BUILDING AIR TEMPERATURE 3-80 3.18 FIRE PROTECTION (DELETED) 3-86 3.19 CONTAINMENT SYSTFMS 3-95 3.20 SPECIAL TEST EXCEPTIONS (DELETED) 3-95a 3.21 BADI0 ACTIVE EFFLUENT INSTRUMENTATION 3-96 3.21.1 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION 3-96 3.21.2 RADI0 ACTIVE GASE0US PROCESS AND EFFLUENT 3-100 MONITORING INSTRUMENTATION 3.22 BADI0 ACTIVE EFFLUENTS 3-106 3.22.1 LIQUID EFFLUENTS 3-106 3.22.2 GASE0US EFFLUENTS 3-111 3.22.3 SOLID RADI0 ACTIVE WASTE 3-118 3.22.4 TOTAL DOSE 3-119 3.23 RAQ10 LOGICAL ENVIRONMENTAL MONITOR' i 3-120 3.23.1 MONITORING PROGRAM 3-120 3.23.2 LAND USE CENSUS 3-125 3.23.3 INTERLABORATORY COMPARIS0N PROGRAM 3-127 3.24 REACTOR VESSEL WATER LEVEL 3-128 4
SURVEILLANCE STANDARDS 4-1 4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION 4-11 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4-28 4.4 REACTOR BUILDING 4-29 4.4.1 CONTAINMENT LEAKAGE TESTS 4-29 4.4.2 STRUCTURAL INTEGRITY 4-35 4.4.3 DELETED 4-37 4.4.4 HYDROGEN RECOMBINER SYSTEM 4-38 4.5 EMERGENCY LOADING SE0VENCE AND POWER TRANSFER.
4-39 EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 4.5.2 EMERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEAT REMOVAL SYSTEM LEAKAGE 4-45 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL R00 DRIVE SYSTEM FUNCTIONAL TESTS 4-48 4.7.2 CONTROL R00 PROGRAM VERIFICATION 4-50
-111-Amendment //, gl, /M, ///, /4/, /M, /M, Idd.191
3.24 Reactor Vessel Water Level Indication I
Acolicability Applies to the operability requirements for the Reactor Vessel Water Level Indication when the reactor is critical.
Ob.iectives To assure operability of the Reactor Vessel Water Level instrumentation which may be useful in diagnosing situations which could represent or lead to inadequate core cooling.
Soecification r
Two channels of the Reactor Vessel Water Level Instrumentation System i
shall be OPERABLE.
If one channel becomes INOPERABLE that channel shall be returned to 1
OPERABLE within 30 days.
If the channel is not restored within 30 days, details shall be provided in the Monthly Operating Report. These i
details shall include cause, action being taken and projected date for return to OPERABLE status.
With no channels OPERABLE, one channel shall be restored to OPERABLE status within 7 days.
If at least one channel is not restored within 7 days, details shall be provided in the Monthly Operating Report. These details shall include cause, action being taken and projected date for return to OPERABLE status, t
Bases The Reactor Vessel Water Level Indication (Reference 1) provides indication of the trend in water inventory in the hot legs and reactor vessel during the approach to inadequate core cooling (ICC).
In this manner additional information may be available to the operator to diagnose the approach of ICC and to assess the adequacy of responses taken to restore core cooling.
Each Reactor Vessel Water Level channel is comprised of a hot leg level indication and a reactor vessel level indication.
The system is required to be operable (as defined previously) when the plant is critical.
The system is an information system to aid the operator during the approach to inadequate core cooling. There is no regulatory limit for this system.
Inoperability of the system removes the availability of an information system. Other useful instrumentation for inadequate core cooling will be available. The Subcooling Margin Indication System is relied upon to determine subcooling margin when the reactor coolant pumps are operating or when natural circulation can be verified. When natural or forced circulation cannot be verified, the margin to saturation is determined by manual calculation, based on reactor coolant temperature (incore thermocouples) and pressure indications available in the control room and steam tables. See Tech. Spec. 3.5.5.
3-128 Amendment No. H 7, W. 191
I The system is not a required system to mitigate evaluated accidents.
It may be useful to have the system operable but there will be no adverse impact if it is not operable.
The LC0 action statement provides the level of emphasis required for an information system.
The Reactor Vessel Water Level is a Regulatory Guide 1.97 Category 1 variable.
Reference (1) UFSAR, Update Section 7.3.2.2(c)10(d)
" Reactor Coolant Inventory Tracking System".
(2) USNRC Regulatory Guide 1.97.
i s
T e
6 3-129 Amendment No. 177, f57, 191 i
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l I
TABLE 4.1-1 (Continued)
CHANNEL DESCRIPTION CHECK lESI CAllBRATE REMARKS
- 49. Saturation Margin Monitor S(l)
N(1)
R (1)When T is greater than 525*F.
50.
Emergency Feedwater Flow NA N(1)
F (1)When T is greater than 250*F.
2 Instrumentation E
2
- 51. Heat Sink Protection System 5
2 a.
FFW Auto initiation (1) Includes logic test only.
Instrument Channels i
- 1. Loss of Both Feedwater NA Q(1)
F y"
Pumps
- 2. Loss of All RC Pumps NA Q(1)
R
~3p
- 3. Reactor Building NA Q
F Pressure M*
- 4. OTSG Low Level W
Q R
l b.
R wM Pressure O
c.
EFW Control Valve Control 7
System
- 1. OTSG Level Loops W
Q R
~
O
- 2. Controllers W
NA R
O d.
HSPS Train Actuation Logic NA Q(1)
R
- 52. Backup Incore Thermocouple M(1)
NA R
(1)When T is greater than 250*F.
gs Display
$ 53. Deleted l
- 54. Reactor Vessel Water level NA N'
R i
-