ML17254A906

Forwards Response to 840604 Request for Addl Info Re NUREG-0737,Item II.F.2 Concerning Inadequate Core Cooling Instrumentation

ML17254A906
Person / Time
Site:	Ginna
Issue date:	08/07/1984
From:	Kober R ROCHESTER GAS & ELECTRIC CORP.
To:	Paulson W Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8408130357
Download: ML17254A906 (15)

Text

j REGULATORY AFORMATION DISTRIBUTION SY EM (RIDS)

ACCESSION NBR;8008130357 DOC,DATE: 84/08/07 NOTARIZED: NO, DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Planti Unit li Rochester G

05000244 AUTH'AME AUTHOR AFF ILIATION,

'OBERiR

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WE Rochester Gas L Electric Corp.

RECIP ~ NAME RECIPIENT AFFILIATION PAULSONiW,AD Operating Reactors Branch 5

SUBJECT:

Forwards response to 840604 request for addi info re NUREG-0737i Item IFFY 2 concerning inadequate coro cooling inst'mentation

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DISTRIBUTION CODE:

A002S COPIES RECEIVED:LTR ENCL SIZE; TITLE:, OR Submittal: Inadequate Core Cooling (Item II.F.2)

GL 82 28 NOTES:NRR/DL/SEP icy.

OL:09/19/69 05000240 RECIPIENT ID CODE/NAME NRR ORB5 BC DICKiG INTERNAL+

ADM LFMB NRR/DHFS/HFEB15 NRR/DL/ORAB 08 NRR/DSI DIR 09 N

SB 10 REG 00 COPIES LTTR ENCL 1

1 1

1 1

0 1

1 1

1 1

1 1

1 1

1 RECIPIENT ID CODE/NAME NRR ORBS LA NRR SHEAg J 01 NRR/DHFS/PSRB16 NRR/DL/ORB5 NRR/DSI/CPB 10 NRR/DSI/RSB 13 RGN1 '7 COPIES LTTR ENCL 1

1 1

1 1

1 1

=

1 3

3 1

1 1

1 EXTERNAL: ACRS NRC PDR NTIS NOTES:

17 10 10 02 1

1 05 1

1 LPDR NSIC 03 06 TOTAL NUMBER OF COPIES REQUIRED:

LTTR

>32 ENCL 31

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ROCHESTER GAS AND ELECTRIC ROGER W. KOBER VKE PRESIDENT ELECTRIC 6 STEAM PRODUCTION NIT L ~ t lp le%

CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649.0001

'3 TEI.EPHONE ARE* CODE 7ld 546-2700 August 7

1984 Director of Nuclear Reactor Regulation Attention:

Mr. Walter A. Paulson, Acting Chief Operating Reactors Branch No.

5 U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

Inadequate Core Cooling Instrumentation, NUREG-0737, Item II.F.2 R. E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Paulson:

An NRC letter from Dennis M. Crutchfield dated June 4,

1984 requested that RGGE provide additional information regarding Inadequate Core Cooling Instrumentation.

Attachment A to this letter responds to the information requests of enclosures 1 and 2

of that letter.

er truly yours, I

@N/

Roger W. Kober Attachment 8408130357 840807'DR, ADOCK.05000244 P

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ATTACHMENT A ENCLOSURE 1

REQUEST FOR ADDITIONAL INFORMATION ROCHESTER GAS AND ELECTRIC CORPORATION PROPOSED INADEQUATE CORE COOLING INSTRUMENTATION FOR THE R. E.

GINNA NUCLEAR POWER PLANT 1.

Describe the scope of the CET upgrading during the 1984-85 refueling outages and note any deviations from the requirements of II.F.2.

The ex>.sting core exit thermocouple (CET) system at Ginna Station utilizes commercial grade connectors, a heated reference junction inside containment and a single display in the control room.

None of the system components have qualification documentation and no isolation is provided

'between the display, and plant: computer (P250).

'i During the 1984-8S refueling outages, the CET system will be upgraded from the reactor head, connectors to the control room !'displays.~

Qualified, thermocouple extension wire will

",, be, run from the.CET's, through a new penetration to the control room eliminating the need for the heated reference junction boxes inside containment.

The 39 CET's will be split into 2 trains. outside of containment and run to separate digital scanning displays in the control room.

The displays will provide isolated CET outputs to the plant computer for normal plant operations and safety assessment.

The CET

displays, cable, containment penetrations, and connectors at the reactor head will be seismically and environmentally qualified.

Operating range of the new CET system, including the displays, will be 0-2300'F.

2.

Document the redundant (installed)

SMM and note any deviations from the requirements of II.F.2.

A redundant subcooling margin monitor (SMM) system was installed during the 1980 and 1981 refueling outages at Ginna Station.

RCS wide range pressure (0-3000 psig) is input to a function generator which computes the saturation temperature (Tsat) for that. pressure.

RCS hot. leg temperature (Thot, 300-700'F) is subtracted from Tsat to obtain the margin to saturation (Tsat-Thot).

The saturation margin is displayed in the control room on a vertical scale indicator (0-100'F subcooling) and input to a bistable which alarms when Tsat-Thot is less than 40'F.

The SMM system hot leg RTD's, wide range pressure transmitters, signal processing

cards, and instrument cable are seismically and environmentally qualified.

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3.

Describe the display system for the ICC related parameters.

Class IE displays for the inadequate core cooling (ICC) related parameters will consist of vertical scale indicators for saturation margin monitoring and reactor vessel inventory

trending, and digital scanning displays for the core exit thermocouples.

The SMM and CET outputs will be directly input to the plant computer for normal operations and safety assessment.

All inputs used to compute reactor vessel inventory will also be input to the plant computer for an independent level calculation.

4.

Provide a detailed description of the proposed dp system and clarify the schedule planned for the system installation, calibration, and testing.

Include the system accuracy for the level measurement and the component uncertainty which contribute to the overall system accuracy.

The proposed dp system will trend coolant inventory within the reactor vessel during all phases of plant operation including post accident conditions with quasi-steady-state conditions, and during relatively slow developing transients.

The system will consist of two redundant Class IE differential pressure trending channels, each consisting of one dp trans-mitter per channel.

Each channel will drive a separate indicator in the main control room showing reactor vessel level (0-100%)

and refueling level (0-24 ft.)

Signals will be processed to compensate for reference leg temperature differences, primary coolant flow and temperature, and safety injection and residual heat removal system operation.

Instrument tubing required will be redundant in the area of the transmitter manifolds only.

There will be only one upper and lower reactor vessel tap with a single line to the transmitter manifolds.

The areas available for reactor vessel taps are small and preclude separation of redundant taps.

Single taps at the reactor top and bottom are considered acceptable because redundant taps would be susceptible to the same events and would introduce additional potential leakage locations.

Likewise, the CETs are in close proximity to each other at the reactor head and are not split into separate trains until after they have exited containment.

Routing of the CETs is such that they are not susceptible to other single events, such as high energy line breaks inside containment.

Although the dp inventory trend instrument will be capable of generating process signals over a greater

span, trend information will be used in accident decision making during quasi-steady state conditions only between the upper tap and the vessel piping penetrations.

Emergency Operating Proce-dures will be revised, based upon Westinghouse Owner's Group Emergency

Response

Guidelines, to include instrument responses

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in this range even though one consultant who studied inadequate core cooling situations (see RG&E letter dated November 29, 1983) concluded that "existing plant instrumentation and procedures are adequate to advise operators of how to respond to voids in the reactor vessel head or distributed-through the reactor coolant system." Installation and use of the system in this manner will meet the requirement for a reactor coolant inventory tracking system to monitor coolant inventory over the range from the vessel upper head to the bottom of the hot leg as the requirement was imposed by Mr. Eisenhut's December 10, 1982 letter for operation of pressurized water reactors.

Rochester Gas and Electric has previously stated its position in letters dated July 2,

1980, December 15,

1980, December 30, 1980 and January 19, 1982, that, an instrument to accurately measure reactor vessel water level could serve a useful

purpose, but that such a device is not necessary for proper response to emergency situations.

RG&E also is not convinced that reactor vessel water level (inventory trend) instruments provide a clear, unambiguous indication of inadequate core cooling, although they may indicate coolant void formation in the limited span above the vessel piping penetrations.

For these

reasons, indications from the differential pressure system'ay be used during normal plant operations to measure coolant inventory over the fullrange of the instrument, but during accident conditions only the limited span required by the NRC will be used.

In this span, coolant level or inventory trend will be calculated assuming a fluid specific gravity determined using the average of three core exit thermocouples.

For use during non-accident conditions, equations and a

functional block diagram have been developed for at least two modes of operation; forced or natural circulation with and without safety injection (SI) or residual heat removal (RHR) system operation.

The vessel is assumed to be par-titioned into three temperature zones; below the core, within the core and above the core.

The fluid temperature (and hence the fluid density) within each of these zones is derived from a combination of cold leg RTDs and core exit thermocouple readings and is dependent, upon the operation of the SI and RHR systems.

However, because of limitations in determining representative temperatures in the lower regions during SI and RHR system operation, inventory indications will not be relied upon for operator action.

The dp system accuracy in the span above the vessel piping is dependent upon environmental conditions.

All signal processing modules are assumed to have the manufacturer's speci, fied accuracy of 20.5% under all conditions.

The dp and wide range pressure transmitters under normal conditions have uncertainties of X0.5%.

For accident conditions, the transmitter uncertainty is specified as i7.0% worst, case for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a LOCA, and i2.5% worst case 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3

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after a LOCA.

The overall system accuracies for the above described operating and environmental conditions have been calculated in accordance with guidelines provided in NRC letter from D.M. Crutchfield to J.E. Maier dated 2/23/83 and are tabulated below.

LOCA LOCA Normal

< 24 hrs.

> 24 hrs.

RCP's Running or Natural Circulation

%2.8%

25.13%

t3.18%

The above system uncertainties were determined using manu-facturer's component specification data.

An acceptable uncertainty for the inventory trending system is considered to be at least 28.0%,

which is based on the distance between the top of the core and "the centerline of the hot leg.

An 8% unc'ertainty will assure

that, a coolant trend indication at the vessel outlet will not represent.

an actual coolant inventory trend which has-dropped to the top of the'ore.

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ENCLOSURE 2

MILESTONES FOR IMPLEMENTATION OF INADEQUATE CORE COOLING INSTRUMENTATION 1.

Submit final design description (by licensee)

(complete the documentation requirements of NUREG-0737, Item II.F.2, including all plant-specific information items identified in applicable NRC evaluation reports for generic approved systems).

RGSE Res onse This description is provided in response to Enclosure 1,

. item 4 included in this letter.

2.

Approval of emergency operating procedure (EOP) technical guidelines-(by NRC).

Note:

This EOP technical guideline which incorporates the selected system must be based on the intended uses of that, system as described in approved generic EOP technical guidelines relevant to the selected system.

System procedures will be based upon Westinghouse Owners Group Emergency

Response

Guidelines.

NRC has the above guidelines available for review.

3.

Inventory Tracking Systems (ITS) installation complete (by licensee).

Installation completion is scheduled for the end of the 1986 refueling outage.

4.

ITS functional testing and calibration complete (by licensee).

Completion of functional testing and calibration is scheduled for three months after installation of ITS.

5.

Prepare revisions to plant operating procedures and emergency procedures based on approved EOP guidelines (by licensee).

Revise.ons to plant procedures are scheduled for 3 months after completion of sy'tem testing and calibration.

6.

Implementation letter report to NRC (by licensee).

qRGGE.Res onse An implementation report will be sent; 6 months after completion of functional testing and calibration.

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7.

Perform procedure walk-through to complete task analysis portion of ICC system design (by licensee).

Procedure walk-through will be performed 3 months after revision of plant operating and emergency procedures.

8.

Turn on system for operator training and familiarization.

RG&E Res onse The system will be available for operator training and familiarization following completion of testing and calibration.

9.

Approval of plant-specific installation (by NRC).

NRC to supply date.

10.

Implement modified operating procedures and emergency procedures (by licensee).

- System Fully Operational RG&E Res onse Implement modified procedures 3 months after procedure walk-through.