ML20045J299

Proposed Tech Specs Pages iii,3-128,3-129 & 4-7a,w/regard to Requirements for Operability of Rv Water Level Indication

ML20045J299
Person / Time
Site:	Three Mile Island
Issue date:	07/15/1993
From:	GENERAL PUBLIC UTILITIES CORP.
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ML20045J296	List: ML20045J295 ML20045J297 ML20045J299
References
NUDOCS 9307230311
Download: ML20045J299 (4)

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l TABLE OF CONTENTS Section Paae 3.16 SHOCK SUPPRESSORS (SNUBBERS) 3-63 3.17 BEACTOR BUILDING AIR TEMPERATURE 3-80 3.18 FIRE PROTECTION (DELETED) 3-86 3.19 CONTAINMENT SYSTEMS 3-95 3.20 SPECIAL TEST EXCEPTIONS (DELETED) 3-95a 3.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATION 3-96 3.21.1 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION 3-96 3.21.2 RADI0 ACTIVE GASE0US PROCESS AND EFFLUENT 100 MONITORING INSTRUMENTATION 3.22 RADI0 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 RADIOLOGICAL ENVIRONMENTAL MONITORING 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 REC 0MBINER SYSTEM 4-38 4.5 EMERGENCY LOADING SE00ENCE 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 R0D SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS 4-48 4.7.2 CONTROL R00 PROGRAM VERIFICATION 4-50

-iii-Amendment /'f, Jff, M , @ , Jg/, ///, fM , 158  ;

1 i

9307230311 930715 E PDR ADOCK 05000289 5 P PDR Li

. .3.24 Reactor Vessel Water level Indication Applicability 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. j Specification Two channels of the Reactor Vessel Water Level Instrumentation System shall be OPERABLE.

If one channel becomes INOPERABLE that channel shall be returned to l l OPERABLE within 30 days. If the channel is not restored within 30 days, I details shall be provided in the Monthly Operating Report. These details shall include cause, action being taken and projected dato 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 l days, details shall be provided in the Monthly Operating Report. These l details shall include cause, action being taken and projected date for return to OPERABLE status.

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, 157

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 LCO action statement provides the level of emphasis required for an information system.

The Reactor Vessel Water Level is a Regulatory Guide 1.97 Category I variable.

Reference (1) UFSAR, Update Section 7.3.2.2(c)10(d) " Reactor Coolant Inventory Tracking System".

(2) USNRC Regulatory Guide 1.97.

k 3-129 Amendment No. JJ7, 157

Table 4.1-1 (Continued) .

CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS

49. Saturation Margin Monitor S(1) M(1) R (1) When T,,, is greater than 525 F. ,

50. Emergency Feedwater Flow NA M(1) R (1) When T,,, is greater than 250 F.

Instrumentation

51. Heat Sink Protection System

a. EFW Auto Initiation (1) Includes logic test only.

Instrumentation

1. Loss of Both Feedwater NA Q(1) R Pumps

2. Loss of All RC Pumps NA Q(1) R

3. Reactor Building Pressure NA Q R

4. OTSG Low Level W Q R ,

b. MFW Isolation OTSG Low Pressure NA Q R

c. EFW Control Valve Control System

1. OTSG Level Loops W Q R

2. Controllers W NA R

d. NSPS Train Actuation Logic NA Q(1) R

52. Backup Incore Thermocouple Display M(1) NA R (1) When T,,,is greater than 250 F.

53. Chlorine Detection System .W M R(1) (1) Calibration is a one concentration Instrumentation point check (need not be traceable to NBS standards)

54. Reactor Vessel Water Level NA NA R Amendment No.: 75,/61l,Jff,IJf,14/,147 4-7a

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