ML20090H363

Responds to SER Open Item 6 Re NUREG-0737,Item II.F.2, Instrumentation for Detection of Inadequate Core Cooling, Providing Updated FSAR Pages Concerning Core Exit Thermocouples & Reactor Vessel Level Measurement

ML20090H363
Person / Time
Site:	Catawba
Issue date:	10/19/1983
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8310280096
Download: ML20090H363 (13)

Text

DUKE Pow 1m GOMPANY P.O. HOX 33180 CHARLOTrE. N.C. 28242 IIAL II. TUCKER TELEPHONE vn:n parmaarre (yo4) 373-4531 AMLEAR PRODtWOM October 19, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414

Dear Mr. Denton:

Section 4.4.3.4 of the Catawba Safety Evaluation Report discusses Open Item 6, Instrumentation for Inadequate Core Cooling Detection. This item was also identified as License Condition 5. As noted therein, the Staff required additional information on core exit thermocouples and reactor vessel level measurement. This information is provided in the attached FSAR pages.

Very truly yours, Ak e-c ,

Hal B. Tucker ROS/php Attachment cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.

Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 Palmetto Alliance 213515 Devine Street Columbia, South Carolina 29205 8310280096 831019

{DRADOCK 05000413 PDR 8i

. Mr. Harold'R.'Denton, Director

' October 19,'1983-Page.2 cc: _Mr. Jesse L.-Riley

< Carolina Environmental Study Group

'854 Henley Place Charlotte, North Carolina 28207 4

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Table 1.9-1 (Page 11)

, Response to TMI' Concerns I

margin to Psat, and the minimum allowable margins to P sat and T sat. Alarm

' - status is indicated by flashing the alarming parameter on the CRT graphic dis- ,

play, the Alarm CRT, and by printout on the Alarm Typer. Two alarm setpoints j are provided for both Tsat.and Psat. The alarm setpoints are dependent on re- l

' actor power. Further details on this subcooling monitor are provided in Table ,

! 1.9-2. '

. Normal' control board instrumentation for RCS temperature and pressure will be

used in conjunction with a control room copy of the steam tables and a written i-procedure to determine margin to saturation as a backup to the computer cal-

- culation.

~

- This system for determining.the degree of subcooling will be fully operational l by fuel loading.

i Reactor Vessel Level Measurement

The reactor vessel level instrumentation system (RVLIS) is of Westinghouse design. The RVLIS is of standard Westinghouse design for upper head injec-tion (UHI) reactor systems and utilizes a microprocessor for data processing.

. The RVLIS uses differential pressure (DP) transmitters to measure the pres-sure drops from the bottom of the reactor vessel to the hot legs for UHI plants and from the hot legs to the top of the reactor vessel.' Under natural-

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i circulation or no-circulation conditions, these pressure drops will provide indication of the collapsed liquid level or relative void content in the i reactor vessel above and below the hot legs. Under forced-flow conditions, the pressure drops will provide indication of the vessel void content above the hot. legs and the relative void content of the circulating primary coolant system fluid. Automatic compensation for changes in the temperature of the impulse lines leading from the reactor vessel and not legs to the DP trans-mitters-is incorporated in the system. Strap-on RTD's are mounted on the vertical runs of the impluse lines for measuring impulse-line temperatures.

Automatic compensation for changes in the reactor coolant system fluid den-sities is also incorporated in the system. Following a hypothetical accident which causes a loss of primary coolant, the RVLIS will be used by the plant operators to assist in detecting a gas bubble or void in the reactor vessel and assist in detecting the approach to a condition of inadequate core cool-ing. If forced-flow conditions are maintained after the accident, the RVLIS will also be used-to assist in detecting the format'on of void in the cir-culating-primary coolant system fluid. The equipment which comprises the RVLIS includes the DP transmitters, impulse lines, impluse-line RTD's, in-containment sensor bellows units, out-of-containment hydraulic isolators, and all the necessary electronic signal conditioning, processing and display equipment. A technical description of the system appears in Westinghouse's manual entitled, "RVLIS -' Summary Report, December,1980."

An item-by-item discussion of NUREG-0737, II.F.2, is provided in Table 1.9-3.

t Rev. 8 l

Table 1.9-1 (Page lla)

Response tn TMI Concerns Incore Thermocouple System I. Present Design The present incore thermocouple system has 65 T/C's (thermocouples) positioned to sense exit flow temperature of selected fuel assemblies.

The T/C's penetrate the reactor vessel head in 5 locations known as instrument ports. Each instrument port has 13 T/C's. Electrical connection to the T/C's is made at the instrument ports by qualified connectors. The class 1E thermocouples are cabled to qualified thermo-couple penetrations. In accordance with NUREG-0737, the system will utilize 16 thermocouples per train and will be able to handle normal attrition. The non-safety thermocouples are cabled to reference junction boxes to allow transition to copper for the remainder of the cabling including the run to' an' instrument penetration. Outside containmbnt, the class 1E T/C's are cabled to reference junction boxes to allow the transition to copper wire. These cables are cabled to the backup display along with the non-safety T/C's. The backup display will display a T/C by use of push-to read switches. All T/C's are cabled from the backup display to the primary display, the plant computer.

A. Present System Capabilities (NUREG 0737 II.F.2 Attachment 1 format)

1. Core inlet temperature data is used with core exit temperature to give radial distribution of coolant enthalpy rise across the core. This is available to the operator via CRT or hard copy.

2. The plant computer via CRT is the operator's primary display having the following capabilities:

a. A spatially oriented core map is available on demand indicating temperature and enthalpy rise at each core exit thermocouple location.

b. The incore thermocouples are an input into the saturation monitor program to assist operator actions for inadequate core cooling procedures.

c. Direct readout via CRT and hard copy print out capability is provided for all thermocouple temperatures. This read-out range extends from 200 degrees F. to 2300 degrees F.

d. Trending of selected thermocouples to show temperature -

time history is available on demand.

e. Alarm capabilities are provided thru the saturation monitor program.

f. Addressed in the Control Room Design Review.

Rev. 8

Table 1.9-1 (Page 11b)

Response to TMI Concerns

3. A backup display is provided to read any of the thermocouples.

With push-to-read switches, readings can be taken well within the six minute time guidance. The range of this backup display extends from 200 degrees F. to 2300 degrees F.

4. A human factor analysis will be part of the ongoing Control Room Review (See Response to Supplement 1 to NUREG-0737 which was transmitted by letter dated June 1, 1983 from H. B. Tucker to H. R. Denton).

5. See Duke Power Company, Topical Report, Quality Assurance Program, Duke - 1A.

6. The primary and backup displays are electrically independent and are on separate highly reliable power systems. Isolation devices are not necessary in the present system as the backup display is not in the circuit until a push-to-read switch is used.

7. Documentation is unavailable regarding the environmental cap-abilities of the present system, but all readouts are located .

in a mild controlled environment.

8. Availability of the plant computer has been greater than 99%.

This documentation has previously been submitted via NUREG 0578/TMI Action Plan response.

9. QA provisions cited in Appendix B do not apply to the present system.

II. Incore Thermocouple System Upgrade The present incore T/C system will be upgraded in the following manner. From outside of containment the non-safety thermocouple cabling will not be altered. However, the class 1E T/C cables will be cabled to a class 1E backup display directly from the T/C penet-rations. These T/C's will be cabled to the primary display using qualified isolation devices. The backup display will be selected as part of the ongoing control board review.

The upgrade of the incore T/C system will be completed on Unit 1 by or during the first refueling and on Unit 2 prior to fuel load.

A. Upgraded System Capabilities (NUREG 0737 II.F.2 Attachment 1 format) ,

Display capabilities will be the same as the present design description with exceptions listed below. Exception numbers correspond to original design numbers.

Rev. 8

Table 1.9-1 (Page 11c)

Response to TMI Concerns 2(f) & 4 The control room review will be completed and all operator display devices will be human factor designed.

6. The prirr.ary and backup display will be energized from in-dependent battery backed power sources. The backup display and associated hardware will be on IE power. Separation as defined in Chapter 8 will be met from the refueling canal to the isolation device.

7. The device 1E T/C instrumentation (T/C qualification assumed) will be seismically and environmentally qualified up to and including the isolation device. Seismic quali-fication will be consistent with the methodologies descr~ib'ed in Section 3.10. Instrumentation subject to a harsh environment will be environmentally qualified consistent with the Duke Power Company position on the Category II Guidelines of NUREG 0588 as detailed in the Duke submittal of June 30, 1982. The isolation device will be in an accessible area following an accident.

8. Quality Assurance provisions cited in Duke Power's QA program for nuclear safety related equipment will be met for all equipment installed except the primary display and hardware beyond the isolation device.

Procedures See Section 13.5.

II.G EMERGENCY POWER FOR PRESSURIZER EQUIPMENT Pressurizer PORV The pressurizer power operated relief valves are air-operated with DC control solenoids. Power for the solenoid valves is supplied from the 125VDC Vital Instrumentation and Control Power System (See Section 8.3.2). The solenoid operators and their controls are safety-related.

Pressurizer PORV Block Valves The pressurizer PORV block valves are motor-operated valves with both motive and control power supplied from the 600VAC Ecsential Auxiliary Power System (See Section 8.3.1). The block valves including their power and control cir-cuits are safety-related.

Rev. 8

Table 1.9-1 (Page lid)

Response to THI Concerns Pressurizer Level Indication Three redundant channels of pressurizer level instrumentation are provided.

These channels are part of the safety-related portion of the Process Control System which receives its power from the Vital Instrumentation and Control Power System (See Section 8.3.2).

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Rev. 8 l

Table 1.9-3 (Page 1)

Reactor Vessel level Instrumentation System Reference Deviations

1. Description of the proposed final system including:

a. a final design description of Items A, B Functionally, additional instrumentation and None displays; b, detailed description of existing Items H, I Functionally, instrumentation systems. None

c. Description of completed or Same as 1(a) --------------

planned modifications.

2. A design analysis and evaluation of Items A,B,C, Functionally, inventory trend instrumentation, and D,E,F None test data to support design in item 1.

3. Description of tests planned and Items C,D,E,F None results of tests completed for evaluation, qualification,'and calibration of additional instru-mentation.

4. Provide a table or description Items G, K None covering the evaluation of con-formance with NUREG-0737: II.F.2, Attachment 1, and Appendix B (to be reviewed on a plant specific basis)

5. Describe computer, software and Items A, B None display functions associated with ICC monitoring in the plant.

6. Provide a proposed schedule for Item L None installation, testing, and cali-bration and implementation of any proposed new instrumentation or information displays.

7. Describe guidelines for use of Item J None reactor coolant inventory tracking system, and analyses used to develop procedures.

Rev. 8 New Page

Table 1.9-3 (Page 2)

Reactor Vessel Level Instrumentation System Reference Deviations

8. Operator instructions in emergency Item J, M None operating procedures for ICC and how these procedures will be modified when final monitoring system is implemented.

9. Provide a t'nedule for additional N/A submittals required.

Explicit confirmation of conformance to the Appendix B (NUREG-0737) items listed below for the Reactor Vessel Level Instrumentation System.

Confirmation Deviations

1. Environmental qualification Item K None

2. Single failure analysis Item G None

3. Class IE power source Yes None

4. Availability prior to an accident One channel None (Per Tech. Spec.)

5. Quality Assurance Yes, as described None in Duke Power Co.

Topical Report, Quality Assurance Program, Duke-1A

6. Continous indications Yes None

7. Recording of instrument outputs Yes (Single train) None

8. Identification of instruments Iter A None

9. Isolation Yes None For the Westinghouse Differential Pressure (dp) system a detailed response to the plant specific items stated below is provided.

Response

A. Westinghouse dp System

1. Describe the effect of instru- Item B ment uncertainties on the

measurement of level.

Rev. 8 New Page

Table 1.9-3 (Page 3) -

Reactor Vessel Level Instrumentation System j Response

2. Are the differential pressure Yes transducers located outside containment?

3. Are hydraulic isolators and Yes sensors included in the impulse lines?

9 Rev. 8 New Page

i Table 1.9-3 (Page 4)

Reactor Vessel Level Instrumentation System REFERENCE ITEM A-

SUMMARY

REPORT Westinghouse Reactor Vessel Level. Instrumentation System -

for Monitoring Inadequate Core Cooling December, 1980.

Submittal Letter T. M. Anderson (W) to Darrell G.

Eisenhut (NRC)

NS-TMA-2358 of December 23, 1980.

REFERENCE ITEM B 25 RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION On the Westinghouse R.V.L.I.S. Summary Report REFERENCE ITEM C SUPPLEMENTARY INFORMATION Submittal Letter E. P. Rahe (W) to L. E. Phillips (NRC) NS-EPR-2579 of March 1971982.

REFERENCE ITEM D W EVALUATION OF TESTS S-UT-3, S-UT-6, S-UT-7, S-NC-2, S-NC-3, S-NC-8 Submittal Letter E. P. Rahe (W) to L. E. Phillips

-(NRC) NS-EPR-2526 of December 9, 1981.

REFERENCE ITEM E W EVALUATION OF TEST S-UT-8 Submittal Letter E. P. Rahe (W) to L. E. Phillips (NRC) NS-EPR-2542 of January 13, 1982.

REFERENCE ITEM F W EVALUATION OF TEST S-1B-1 AND FUNCTIONAL TEST Submittal Letter E. P. Rahe (W) to L. E. Phillips (NRC) SED-SA-0081 of June 28, 1982.

REFERENCE ITEM G RESPONSE TO REQUEST TO WOG FOR ADDITIONAL INFORMATION Submittal Letter E. P. Rahe (W) to Dennis M.

Crutchfield (NRC) NS-EPR-2597 of May 14, 1982.

REFERENCE ITEM H RCS WIDE RANGE PRESSURE INSTRUMENTATION

. Submittal Letter of E. P. Rahe (W) to R. C. Deyoung (NRC) NS-DPR-2586 of April 21, 1982.

REFERENCE ITEM I RELEVANT DRAWINGS OF EXISTING INSTRUMENTATION SYSTEMS 8757D55 Revision 4 W Process Block Diagrams. (See i Sheet for W. R. RCS7 HOT)

Rev. 8 New Page

Table 1.9-3 (Page 5)

Reactor Vessel Level Instrumentation System 1098E74 Sh 1 Rev 8 W RCS Flow Diagram i FSAR Fig. 5.1-3 RCS Flow Diagram.

Note that W Process Block Diagram which shows W. R.

RCS Pressure is not recommended as suitable inter-facing instrumentation. Refer to Item H.

REFERENCE ITEM J Critical safety function status trees for core cooling and system inventory volumes 1, 2, and 3 Emergency Response Guidelines (ERG) Developed by Westinghouse Owner's Grcup (WOG) Sections FR-C.1 and C.2, FR-I.3, FR-P.I.

Refer also E20.26 (which is a new section to be added, for natural circulation cooldown regarding depressurization in the upper head. Although this shows the use of the RVLIS instrumentation, it is not specifically for inadequate core cooling (ICC) monitoring.

Volume 1 and 2 was under cover of OG-64, 11/30/81

0. Kingsley (WOG) to D. Eisehut (NRC)

Volume'3 was under cover of OG-83, 1/4/83, O. Kingsley I (WOG) to D. Eisenhut (NRC)

All member utilities received copies under cover of letters:

WOG-81-235, 12/2/81 WOG-83-100, 1/4/83 REFERENCE ITEM K ENVIRONMENTAL QUALIFICATION (Note Various Equipment Data Qualification Packages)

- i.e., EQDP - for RVLIS instrumentation are sub-mittals as supplements to the Environmental Quali-fication topicals WCAP-8587, non proprietary (which provides summary EQDP-s) and WCAP-8687, proprietary (which provides detailed EQDP's).

EDQP References are as follows:

1. ESE-2 D/P Transmitters Outside Containment

2. ESE-42 Strap-on RTDs

3. ESE-48 High Volume Pressure Sensor

4. ESE-49 Differential Pressure Indicator Switches Rev. 8 New Page

Table 1.9-3 (Page 6)

Reactor Vessel Level Instrumentation System

5. ESE-50C RVLIS-86 Microprocessor

6. ESE-53 Plant Safety Monitoring System Electronics REFERENCE ITEM L IMPLEMENTATION SCHEDULE The RVLIS will be fully operational by fuel load.

REFERENCE ITEM M ICC OPERATING PROCEDURES Duke Power plans to upgrade the Catawba emergency operating procedures based upon the Emergency Response Guidelines (ERG) developed by the Westinghouse Owners Group. The RVLIS will be incorporated into procedures according to these guidelines.

9 Rev. 8 New Page

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