ML20071J998

Changes to Sys & Procedures Described in FSAR Per 10CFR50.59

ML20071J998
Person / Time
Site:	Three Mile Island
Issue date:	05/10/1983
From:	Hukill H GENERAL PUBLIC UTILITIES CORP.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
5211-83-099, 5211-83-99, NUDOCS 8305270203
Download: ML20071J998 (17)

Text

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GPU Nuclear Corporation G U quclear artnze"A#8 s Middletown, Pennsylvania 17057

/17 944-7621 TELEX 84 2386 Writer"s Direct Dial Number:

Fhy 10,1983 5211-83-099 Mr. R. C. Haynes Region 1, Regional Administrator U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406

Dear Sir:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 10 CFR 50.59 Report In accordance with the requirements of 10 CFR 50.59, enclosed please find two copies of changes to TMI-l systems and procedures as described in the FSAR. There were no tests or experiments performed not described in the FSAR.

Sincerely, s

H. D. Hu ill Director, TMI-l HDil: RAS:vj f Enclosures cc: Director, Office of Inspection and Enforcement (40 copies)

U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Director, Office of Management Information and Program Control U. S. Nuclear Regulatory Commission Washington, D.C. 20555 John F. Stolz, Office of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Washington, D.C. 20555 l

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" w GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation i

Modification: B/A 412074 - Limit Reactor Building Purge and Vent Valve Angular Opening to 30 (PM-3)

Description of Project:

The Reactor Building Purge and Vent Valves AH-V1A/B/C/D have been restricted to 30 open to ensure that the valve operators can close against the fluid dynamic forces and pressure differential that are created during a LOCA condition. Air-operated valves AH-VlA&D have been restricted by installing mechnical travel stops (bolts) in the cylinder head of each air-operator. The stops prevent the piston from traveling full stroke, i.e., only to the 30 maximum open and have been locked in position. Motor-operated valves AH-VlB6C have been restricted by resetting the Limitorque operator rotor switches to a trip position that corresponds to the 30 maximum open.

The new setting will limit the valve opening via action of the limit switch contacts within the valve operator control circuit.

}Bt wiring changes have been added and/or deleted. Limit switches which signal the valve "open" position in the control room have been reset to indicate the valve "open" position at the 30 opening.

Furthermore, a new air flow sensor and recorder will be installed in the ductwork and control room, and will measure the throttled flow rate during purging / venting operations.

NOTE: This modification is complete. We have subsequently committed to close the valves, however, we have not received a response from the NRC.

Safety Evaluation Summary:

This modification has been implemented so that the torque output to stroke the valves closed from the 30 angle under high differential pressure and flow conditions is within the capability of both the Bettis and Limitorque valve operators. The signals to open/close the valves have not been modified and the ES closure signal remains the same, i.e., as modified via the Containment Isolation modifica-tion. Purging / venting of the Containment Building for reducing air-borne activity may still be accomplished, however, at a reduced flow rate. During shutdown, the valves may be opened to the full open 90 position. No safety margins will be reduced as a result of this modification.

Modification: B/A 412045 - BWST Level Indication (LM-4A)

Description of Project:

This modification added Barton differential pressure switch to the BWST to provide an additional low level alarm.

Following an accident, the initial injection of water by the decay heat removal system involved pumping water from the borated water storage tank (BWST) into the reactor vessel. With all engineered safeguards pumps operating, and assuming the maximum break size, this mode of operation lasts for a minimum of about 25 minutes. When the BWST reaches a low level, an alarm will be annunciated in the control room. At this time the operator will take action to open the suction valves (DH-V6A/B) from the reactor building emergency sump, permitting recircu-lation of the spilled reactor coolant and injection water from the reactor build-ing sump. The BWST low level alarms, one from each BWST level transmitters, will be supplemented by a low level alarm from the new Barton differencial pressure switch. This switch will provide additional notification that the succion valves to the R.B. sump must be opened, thus removing a potential conflict if one of the existing transmitters indicates an erroneous high level.

Safety Evaluation:

This change provides for an additional alarm for BWST low level to the operator.

This alarm will serve to confirm either of the existing low level alarms for the contingency of mis-operation of one of the alarms. Since this change does not adversely affect Nuclear Safety, it will serve to enhance Nuclear Safety; this change does not constitute an Unreviewed Safety Question.

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Modification: B/A 412053'- Audio Monitoring System for the Main Steam

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System Code Safety Valves (Task RM-6)

Description of Project:

The, audio monitoring system is an acoustic system that monitors noise resulting from high velocity steam flow when the main steam system code safety valve (s) lif t when-relieving steam flow. The noise information is transmitted by microphone to the control room to indicate the lif t of the valve (s) associated with the specific steam generator (A or.B) and the main steam line to the emergency feed pump turbine.

The safety valves are in four compartment areas located within the Intermediate Building. Each compartment comprises.1 to 6 safety valves. In each compartment a microphone is provided in close proximity to the safety valves for monitoring the noise of the relieving steam flow from the valves (s). '

An audio processor cabinet is provided when a signal from a microphone or channel pick-up is converted in order to provide; input to data logger, and signal- to the audio monitoring cabinet in the control room. The audio monitoring cabinet includes a speaker, channel selector switch and indicating lights which provide the' operator with information concerning valve position indication for 'open' and 'close' conditions.

An uninterruptibl'e power supply is provided as the power source to the monitoring system.

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, Safety Evaluation Summary This modification has been implemented so that the control operator can monitor the lifting of the main steam safety code valve (s) in the event the main steam pressure is exceeded in both the A and B steam generator system. The affected steam generator system which the operator can identify from the data logger printout or audio channel indicating light. The operator acts per plant procedures to mitigate tne event to preclude cooling of the RCS.

This mod'ification is considered to be non-safety related. Therefore, single failure criteria or seismic conditions are not part of the design modification.

However, to ensure the electrical functioning of the monitoring system the power source is taken from a vital AC source (inverter).

The probability or consequences of an accident have not increased.

The modification will not change the operation of the steam generator (A or B) systems. Therefore no new accident conditions will occur as a result of this change modification.

l No safety margins have been reduced.

i I Accordingly, the implementation of this modification does not involve unreviewed safety considerations with regard to the criteria of 10CFR Part 50 Section 50.59 l (a) (2).

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Modification: B/A 412021 - Reactor Coolant System High Point Venting System Pressurizer Vent Only (LM-21A)

Description of Project:

The pressurizer vent line has been installed in order to improve the plant's ability to vent a mixture of reactor coolant liquid / steam and/or non-condensible gases from the Reactor Coolant System, with-out having an adverse impact on core cooling. This safety grade modification satisfies Seismic Class I criteria and is supplied with Class 1E electrical and instrumentation power.

The pressurizer vent line is controlled by motor-operated isolation valve RC-V28 and solenoid-operated isolation valve RC-V44, which are mounted in series, and vents initially to the Reactor Coolant Drain Tank and ultimately to the Reactor Building atmosphere. Flow ele-ment FE-1079 and differential pressure transmitter DPT-1079 provide an input signal to a " flow /no flow" lamp installed on panel "PC".

The vent valves are activated from the Main Control Room "PC" panel which includes open/ closed key lock switches and indicating lights.

Annunciation of inadvertent valve opening has been provided on light i box "G" in the Main Control Room. The Pressurizer vent line will be maintained and operated via strict administrative controls.

l Safety Evaluation Summary:

l The purpose of this modification is to provide remote power opera-tion of the vent line from the Pressurizer to the Reactor Coolant Drain Tank. This modification makes no changes in the procedure as described in the Safety Analysis Report. Administrative procedures l

will be implemented for controlling the operation of the pressurizer vent line from remote controls in the Main Control Room. This modi-fication does not create the possibility of an accident or malfunc-tion different from any previously evaluated in the SAR, i.e., fail-l ure of the PORV which is already an evaluated accident. No safety l margins have been reduced as a result of this modification.

l Strict administrative controls and key lock switches will govern I actuation of the pressurizer vent line.

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CHANGE MODIFICATION: B/A 412028 - ICS/NNI Loss of Power Critical Plant Parameters Independent of ICS/NNI

, (LM-43C) ,

e DESCRIPTION OF CHANGE: ,

The function of the Safety Critical Plant Variables in the control room is to provide the operator with indication of vital information in the event that all power internal to the NNI/ICS is lost. This design is designated as Nuclear Safety Related, Class lE. (Digital indicators are not yet qualified to Class lE and they will be replaced with qualified units at a later date, if available).

The indicators are located on the "PCL" back panel (except Pressurizer and Makeup Tank Level which are on main console "CC".) The signal source is from the Safety Grade Remote Shutdown Signal Conditioning cabinets (Task LM-38) via isolated outputs.

The existing NNI/ICS indicators and controls remain. The new digital indicators do not share power or signals with the existing equipment.

The critical plant variables are:

OTSG Loop A/B Pressurizer Level RC Pressure (wide) loop B RC Hot Leg Temperature Loop A/B Makeup Tank Level RC Cold Leg Temperature Loop A/B SAFETY EVALUATION:

This modification provides the control room operators with an alternate source with vital information so that corrective action can be taken to effect a controlled and orderly shutdown of TMI-1, if necessary.

Therefore, this modification does not increase the probability or occurrence or consequence of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis.

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CHANGE MODIFICATION: B/A 412013 - Iodine and Particulate Sampling (LM-25B)

DESCRIPTION OF CHANGE:

Three.(3) new sampling stations have been installed. Each station provides iodine and particulate samples of high and low radiation levels, and also, provide samples of noble gas. Two of the stations are located in the special building located on top of the concrete exhaust duct. Of these 2 stations, one samples the auxiliary building ventilation exhaust and the other the reactor building ventilation exhaust. The third station is associated with radiation monitor RM-A5 and provides samples of the condenser vacuum pump discharge.

The post accident iodine and particulate monitoring is accomplished passing a flowing stream representative of the effluent discharge stream through silver Zeolite cartridges with particulate filters. +Jhe range of measurement for the sample system as defined in NUREG 0737 is 10 Ci/cc. After sampling the cartridges will be removed and counted in a laboratory. Each of the three previously mentioned effluent discharge paths have a post accident iodine and particulate sample system. These sample systems are normally idle and are energized by particulate, iodine and gas monitors RM-A5, A8 or A9 when these monitors detect high radiation.

_ SAFETY EVALUATION:

The addition of the post accident iodine and particulate monitors in no

, way adversely affects plant safety. The probability or occurrence or the

( consequence of an accident or malfunction of equipment important to safety l previously evaluated in the Safety Analysis Report has not been increased.

l The ability of the effluent radiation monitoring system to perform its function l under accident conditions has been enhanced.

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MODIFICATION: B/A 412013 - High' Range _ Containment Monitoring (LM-23)

DESCRIPTION OF PROJECT:

Section 2.1.8b of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," July,1979, contains NRC recommendations on installations of in-containment radiation level monitors. The guidance on in-containment radiation level monitors is as follows:

In-containment radiation level monitors with.a maximum range of 10 8 rad / hour total radiation shall be installed. Al radiation measurement with a maximum range of 10pernately, rad /hr may be a photon only utilized.

In response to the above recommendations, two Victoreen Model 877 radiation monitor detectors'have been installed to monitor containment radiation levels during and following a postulated accident.

7 The detectors have a range of 10 rad /hr. to 10 rad /hr. photon measurement only.

The high range radiation detectors and high range area monitors meet:the design and quality assurance requirements for a Nuclear Safety Related System. They-meet the single failure and testability criteria of IEEE 279-1971.. Separation of redundant circuits,is maintained in accordance with Section 8 of the FSAR.

All components except readout devices and recorders are ~ qualified to IEEE Standard 344-1975 " Recommended Practices for Seismic Qualification of Class lE Equipment for Nuclear Power Generating Stations." The readout devices and recorders are mounted on the PRF Panel. The electrical installation is designated " Nuclear Safety Related" to include the detectors and their associated electrical cabling. The detectors are qualified in accordance with the requirements of IEEE-323 1974 for operation in the_ containment environment for normal and post accident environment.

SAFETY EVALUATION:

The system design does not increcse the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.

! The use of the high range radiation monitor will assist plant operators to l assess plant conditions during and following an accident.

i No accidents or malfunctions of a different type than that which was previously evaluated in the safety analysis report, will be introduced. The design and.

installation features of the radiation monitor are designed so as to preclude

-the compromising of containment integrity or other safety features.

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(LM-23) B/A 412013 - High Range Containment Monitoring (cont'd.)

t No safety margins as defined in the basis for any Technical Specifications have been reduced. The radiation monitor provides continuing in-containment post-LOCA radiation' assessment and is not involved in consideration

'of any safety margin.

Based upon the above, we conclude that plant modification needed for installation of the high range radiation monitors do not involve any unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59 (a)(2).

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i CHANGE MODIFICATION: B/A No. 412044 - Saturation Margin Monitor (LM-I)

DESCRIPTION OF CHANGE:

This modification provides all necessary instrumentation for detection of inadequate core cooling by . operators in Control Room.. The Reactor Coolant System Saturation margin monitor provides direct on-line reactor coolaot system margin to saturation indication. This is accomplished in each reactor coolant loop by means of continuous, computation, display

> and comparison to the limiting saturation condition for the reactor coolant system. These computations are based on e xis ting reactor coolant system Non-Nuclear Instrumentation (NNI) pressure and temperature signals. Continuous digital display of Reactor Coolznt Sys tem Saturation Margin (RCSSM) is provided to the plant operator in terms of 0F margin in.

the Control Room. An audible and visual alarm actuates when the computed saturs_. tion margin is less than the limiting value. The RCSSM provides, saturation margin and alarm status output suitable for-trending and computer input.

SAFETY EVALUATION:

This modification is to provide the controk room operator with information to assist in identifying inadequate core cooling conditions. The saturation margin monitor will display, in the control room,the margin between the actual r e a'c t o r coolant' system tempera;ure (Th) and the saturation temperature (Tsat) for the exis, ting reactor coolant system pressure. An alarm will b e initiated if the margin should decrease below a preset value. The signal conditioning equi # ment and cabinets are seismically qualified and powered from4 IE buses. The digital panel meters though not seismically qualified are the best quality industrial grade available. Adequate redundancy and diversity exist to ensure that Ts,at margin will be available to operator in the ControltRoom of TMI-I. This modification does not degrade the integrity of any safety-related system or any existing instruments which are required for safe and reliable

. operation and shutdown of the plant.

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Modification: B/A B00225 - Second Level Undervoltage Protection (RM-22)

Completion Date:

Descrip' tion of Project:

NRC letters dated June 3, 1977, August 8, 197,7 and May 2, 1979 on failure oT Class 1E equipment to perform their required function due to offsite electric grid system voltage degradation, require a second level of undervoltage protec-tion to protect against such an incident. This modification has been designed on the basis of NRC's ' Safety Evaluation and Statement of Staff Positions, Re-lative to the Emergency Power Systems' as it pertains to second level under-voltage protection and which was enclosure 1 of June 3, 1977 letter NRC to J. G. Herbein.

s On safety related 4160V buses ID and 1E, there is an addition of six solidstate undervoltage relays, which includes replacement of existing three electromechanical undervoltage relays, one overvoltage relay and associated test switches and time delay relays. Also, it modifies panel wiring to provide re-required logic and alarm.

Safety Evaluation Summary:

This modification provides protection of safety loads from sustained degraded voltage from the offsite electric grid system without causing voltages in excess of maximum voltage ratings of safety loads and without causing spurious separation of the safety buses from offsite power.

When a degraded voltage condition is detected on a 4160V safety bus, that bus will be tripped and the associated diesel generator started. The diesel generator then provides the capability to adequately power safety loads on that bus. The changes that have been implemented by this modification do not create the possibility for an accident or malfunction, and do not have any adverse effects on any system / components.

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httftention: B/A 412012 -

OTSC Level indication in the Control Room Independent of ICS (101-111)

Royf],ctionliato: February 4, 1982

, Description of Project,: c -

The OTSG level indLeation independent of ICS has been inntalled on the main control console in the control room and assures that the plant operators have proper indicat ton to control OT!;C level . Th l .6 modification is redundant, provides full range (0-640" of water) i level indicatien, and satisfies single failure criteria.

Two independent and rodundant equipment trains (' Red and Green')

provide assurance that at least one (1) display unit per GT8t: 18 availabic to the plant operators. The Remote Shutdown Panel naiety grade level transmitters and signal proccasing equipment provide an input ::ource for tha , level indicat ion sir,nals. The ' Red' level ,

transmittern, LT-775 and LT-7hS, on 0T80'8 (RC-II-1A/ll) and corre-sponding isolation n:odules provide signals to control connole indi-cators LI-7758 and L1-7883. Similarly, the 'Creen' level transmit-ters, LT-776 and LT-789, and corresponding isolation modules provide ,

signals to control consolc indicators LI- 776B and LI-789B.

Safety Evaluation Summary:

This safety grade. Scismic Class I modification, provides OTSC water icvel indication (full rance) in the control room independent of ICS. Redundant indicators have been installed on the control con-sole for each OTSG. This modification provides only a display func-tion and does not interface with any existing control and/or protec-tion systems. Therefore, this modification does not degrade any existing safety margins. The load requtrements for the additLonal instrumentation have been reviewed and the Irapact on the diesel generator loading is acceptable.

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v Modifiention: B/A 412008 - Decay Heat and Core Flood Check Valves Leak Monitoring System (LM-42)

Description of Change:

This test system has been installed in order to determine the leak rate across isolation check valves DH-V22, CF-V4 and CF-VS, and to preclude over pressurization and rupture of the low pressure decay heat removal system piping outside of containment. The test system consists of commer-cial grade tubing up to the points where it ties into the Dl! and CF system piping. At the tie-in points, the pipe and fittings are nuclear class to maintain the existing system integrity and pipe classification. All por-tions of'this modification are designed to Seismic Class I requirements.

- Safety Evaluation Summary:

This modification covers the installation of a test system for the purpose of performing Icak rate tests of the DH and CF check valves. The test system is a commercial grade system which has ryg safety related function during plant operation and does not alter any safety functions nor reduce the margin of safety as defined in the FSAR station technical specifica-tions. The system is disconnected from the CF and DH system piping during normal plant operation, and is seismically supported to avoid potential interaction with safety systems, RCS pressure boundary isolation is mai'ntained via existing double isolation valves or by an isolation valve and blank flan ge arrangemen t. The entire system is located within the Reactor Building.

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9 Change Modification: RB Spray System D/P Instrumentation Modification a

B/A 412073; LM-7A Description of Change: ,

This modification adds a differential pressure instrument between the

BWST and SHST to provide a simple and direct means of measuring the .

fluid level differential between the two tanks, which will aid in main-taining the differential within prescribed limits.

Safety Evaluation:

This is an Important to Safety instrument only from a pressure boundary integrity standpoint. It is not required to function in order to achieve the RBSS safety functions. The DHRS and RBSS pressure boundary integrity is assured by the "Important to Safety" classification and quality assurance pertaining to it, and the seismic class S-I design criterion.

The addition of a d/p instrument does not change any of the safety analysis described in the FSAR, nor in any way increase the potential for any of the hypothetical accidents described therein.

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Modif'ication: B/A 412039 - Tripping of loads on engineered safeguards

{ buses (LM-32)

Completion Date: ,

Description of Project:

THI-1 Licensee Event Report (LER) 80-01/0lT reported a potential overload condition on the 1P480V bus during loss of IS480V bus. Subsequently, Techni-cal Data Report 185 Rev. O confirmed the overloading of IP and 1S buses when 1S and IP were not available respectively.

Following six loads will trip automaticclly, due to this modification, on ES signal.

I) Spent Fuel Cooling Pump Air Unit AH-E-8A II) Penetration Cooling Fan AH-E-9A III) & IV) Spent Fuel Pit Cooling Pumps, SF-P-1A & 1B V) Boric Acid Mix Tank Heater VI) Boric Acid Tank Mixer The automatic tripping of selected loads along with administratively controlled manually applied loads e.g. 480V power receptacles vill bring the 480V buses IP and 1S within their. design limits.

Struthers-Dunn auxiliary relays and indicating lights have been added to the IAES and IBES motor control centers. Cabling runs between ES actuation A cabinet 4 and IAES MCC and ES actuation B cabinet 5 and 1B ES MCC. Necessary coordination curves have been developed and the setting of the incoming breaker for the 1B ES MCC has been increased corresponding to its load.

Safety Evaluation Summary:

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This modification restores the integrity of the safety related buses under all modes of plant operation as is required per general design criteria 17 of 10CFR 50 Appendix A.

Loads that trip automatically can safely be tripped without any design concerns.

In the event both spent fuel cooling chains were out of service simultaneously, the heat capacity of the water contained in both pools is such that approximately 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> would elapse before the water in them would heat up to an excessive temperature. Abnormal spent fuel pool water temperature is presently alarmed in the control room (per FSAR article 9.4.7). GAI letter #GAI/TMI-1/ICS/3354

,, - W. A. Steidle and R.M. Rogers to D.G. Slear on " Loss of Penetration System" dated 6/17/1980,, indicates that restart of this system can be delayed six hours without design concerns.

The changes that have been implemented by this modification do not create the

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possibility for an accident or malfunction, and do not have any adverse effects on any system / components.

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1. RCS Overcooling Guidence PCR No. Procedure Procedure Title 1-OS-82-0007 1202-2 Station Blackout 1-OS-82-0008 1202-2A Station Blackout With Loss of Both Diesel r Generators 1-OS-82-0010 1202-4 Reactor Trip 1-0S-82-0011 1202-5 OTSG Tube Leak-Rupture 1-OS-82-0004 1202-6A Loss of RC/RCP Within Capability of Makeup Sys (RC Pressure Above ESAS Set Point) 1-0S-82-0005 1202-6B Loss of RC/RC Pressure (Small Breakloca.)

Causing Automatic High Pressure Injection 1-OS-82-0006 1202-6C Loss of RC/RCP Causing Auto HP Inj. Core Flood

& LP Inj.

1-0S-82-0009 1202-14 Loss of Reactor Coolant Flow /RCP Trip 1-0S-82-0012 1202-26A Loss of Steam Generator Feed to Both OTSG's 1-OS-82-0013 1202-26B Loss of Feed to One Steam Generator 1 1-0S-82-0014 1202-36A Loss of Instr. Air-Backup Air Available 1-OS-82-0015 1202-36B Loss of Instr. Air-No Backup Air Available 1-OS-82-0016 1202-39 Inadequate core Cooling (No Local) 1-05-82-0017 1203-24 Steam Line Rupture Detection System (SLRD)

Actuation These changes improve the margin of safety since they improve the operator guidance on plant response and plant operation within the expected operating limits.

Potential for challenges to the plant safety limits and limiting safety system settings are thus reduced.

The RCS overcooling guidance helps to ensure that the heat balance between Reactor Heat Generation is equal to the RCS Heat Rejection though the OTSG's, thus assuring adequate decay heat removal without overcooling the RCS.

The HPI throttling criteria assures that sufficient RCS inventory control is established under abnormal plant conditions to maintain maximum decay heat removal without damage to either the HPI pumps or the Reactor Vessel.

2. Loss of OTSG's as a Heat Sink l 1-0S-82-0493 1202-39 Inadequate Core Cooling (No Local) ,

This change provides specific operator guidance for conducting a safe plant ecoldown without using the OTSG's as a heat sink. Since a total and extended loss of the OTSG's as a heat sink was not previously addressed in the emergency procedure, this change improves the margin of safety by providing guidance for this contingency.

3. Failura of Fredwstar and PORV 1-OS-82-0495 1202-29 Pressurizer Systems Failure This change improves the operator guidance for plant emergencies in which a coincident failure of all OTSG feedwater and failure of RC-RV-2 (PORV) were to occur.

By alerting the operator to proceed to the Inadequate Core Cooling Procedure (EP 1202-39 under the above circumstances, the operator vill maximize the available core cooling methods. The safety margin is improved by this change, since it provides guidance for a contingency which previously was not provided in this Emergency Procedure.

4. Reducing Challenge to MSRV's .

t 1-0S-82-0694 1202-5 OTSG Tube Leak-Rupture These changes provide more guidance for taking the turbine off line and shutting down the reactor while reducing the potential challenge to the main steam safeties.

The margin of safety is not reduced since the change minimizes the main steam .

safety valve discharges assumed in the F.S.A.R.

5. Condensate Inventory 1-0S-82-0791 1202-26A Loss of Steam Generator Feed to Both OTSG's This change improves the margin of safety since it more accurately reflects the available condensate inventory within the condensate storage tanks when the low-low level alarm is actuated.

6. Pressurizer Heater Power Source

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1-05-82-047.0 1202-29 Pressurizer Systems Failure i

This change improves the margin of safety since it provides additional power l

sources to the pressurizer heaters, thus increasing the ability of the operator to maintain the plant subcooled.

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