ML20082Q643

Affidavit of LI Loflin Supporting Motion for Summary Disposition of Chapel Hill Anti-Nuclear Group Effort Contention 44 & Eddleman Contention 132.Certificate of Svc Encl

ML20082Q643
Person / Time
Site:	Harris
Issue date:	12/05/1983
From:	Loflin L CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20082Q614	List: ML20082Q611 ML20082Q620 ML20082Q637 ML20082Q643
References
ISSUANCES-OL, NUDOCS 8312120290
Download: ML20082Q643 (26)

Text

_ - _ _ _ _ _ .

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING E0APD In the Matter of )

)

CAROLINA POWER & LIGHT COMPANY ) Docket Nos. 50-400 OL ano NORTH CAROLINA EASTERN 50-401 OL MUNICIPAL POWER AGENCY )

(Shearon Harris Nuclear Power )

Plant, Units 1 and 2) )

AFFIDAVIT OF LEONARD I. LOFLIN County of Wake )

) ss.

State of North Carolina )

LEONARD I. LOFLIN, being duly sworn according to law, d_e-P poses and says as follows:

1. I am Manager-Engineering, Shearon Harris Nuclear P_over Plant, Carolina Power & Light Company. A sunmacy of g professional qualifications and experience is attached hereto ,

as Exhibit "A". I have personal knowledge of the natters stat-  ;

ed herein and believe them to be true and correct. I make this Affidavit in support of Applicants' Motion for Sur: mary Disposi-tion of CHANGE Contention 44 and E_ddicce.n Contention 132.

_2_ . In 1983, CP&L decided that the Westinghouse-designed Peactor Vessel Level Instrumentation System (RVLIS) would b_e installed at the Harris plant as a part of the enhanced inade-quate core ecoling instrumentation system responsive to Item II.F.2 of NUREG-0737. The RVLIS being installed et the Harris plant represents the ccat recent Westinghouse design. It is a fully qualified and redundsne c: stem for monitoring water in-ventory in the reactor vessel.

3_ . E::hibit "B" to this Affidavit is the documentation 8312120290 831207 PDR ADOCK 05000400 0 PDR

C_P&L provided to the NRC Staff in response to Item II.F.2 of NUREG-0737. It describes the Harris plant's instrumentation system, including R,LVIS, for inadequate core cooling. Exhibit "C" clarifies and supersedes the information provided in Exhibit B.

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Lecnard I. Loflin Subscribed and sworn to before me this & day of December, 1983

.. 44Notary

$ Public $A w My Commission expires: FJ a - N e

EXHIBIT A LEONARD IRA LOFLIN Manager - Harris Plant Engineering Section BIRTH DATE: October 16, 1941 I. EDUCATION A. B.S. Degree in Electrical Engir.eering from Clemson University - February 1964 B. Professional Degree in Nuclear Engineering from North Carolina State University - June 1969 C. Reactor Operator Training Programs

1. Westin;; house Corporation, Saxton Plant:

AEC Reactor Operator License - February 1970

2. Virginia Electric & Power Company, Surry Plant:

AEC Senior Reactor Operator License - April 1972 II. EXPERIENCE A. 1960 to 1963

1. Duke Power Ccmpany

a. Three summer work periods at Buck Steam Plant, Spencer, N.C.

b. One summer work period at Greenville, F.C., Distribution Engineering Office B. February 1964 to June 1973

1. Virginia Electric & Power Company

a. Assistant Engineer, Yorktown Power Plant (two 165 MWe fossil firea uffits): February 1964 to November 1964 Participated in maintenance and modifications of plant control systems. Responsible for plant performance testing and monitoring,

b. Assistant Engineer: November 1964 to May 1965 Associate Engineer: May 1965 to January 1967 Engineer: January 1967 to May 1967

t Mt. Storm Power Plant (two 565 MWe fossil fired units)

Assignraent to Mt. Storm was made prior to initial phases of first unit startup. As the only nonsupervisory utility engineer assigned to the plant during startup of both units, was integrally involved in all engineering, cperations, and maintenance facets of startup on both units.

c. _ Engineering Supervisor, Mt. Storm Power Plant: May 1967 to September 1968 Supervisory and technical responsibility for all station engineers, chemists, instrument technicians, laboratory tecnnicians, coal handling foremen, and coal handling union personnel. Handled contract interface and execution on station level between fuel vendors and VEPCO.

d. Staff Engineer, Richmond, VA September 1968 to June 1969 Assigned to N. C. State University

e. Assistant Operating Supervisor, Surry Nuclear Power Plant (two 2441 MWt Pressurized Water Reactors):

June 1969 to September 1972

. Responsible involvement: Core loading; initial criticality; escalation to power: preoperation&1 startup of all plant systems, both secondary and primary; scheduling and organization of operations department; interface relations with Stone & Webster (A/E), Westinghouse, and Atomic Energy Commission; organization and coordination of nuclear training. ,

f. Operating Supervisor, Surry Nuclear Power Plant (two 2441 MWt Pressurized Water Reactors): September 1972 to June 1973 Responsible for all plant operational functions. Conducted escalation to rated power of Unit 1. Directly supervised core loading, initial criticality, and escalation to power of Unit 2. Personnel responsibility fe forty-five (45) operators and eleven (11) first-line supervisors.

C. June 1973 to Present

1. Carolina Power & Light Company

a. Employed as Principal Engineer in the Power Plant Engineering Department. Assigned to Nuclear Plant Engineering Section II as Principal Engineer on the South River Project. Supervised first year of project aativity including nuclear steam supply (NSS) negotiation; i

creating functioning relationships between CP&L, NSS vendors, and the architect / engineer (A/E); generating plant conceptual design.

b. July 1974 - Assigned as Principal Engineer to the Brunswick Steam Electric Plant as Engineering Start-Up Coordinator. Responsible for g identifying engineering areas which were not properly supporting start-up effort and rectifying the situation. Emphasis was given to interfacing with CP&L construction and operation organizations to optimize feedback and execution.

c. August 1975 - Promoted to Manager - Corporate Nuclear Safety Section of the Special Services Deoartment. Transferred from Power Plant Engineering Department.

d. June IG76 - Assigned as Manager - Corporate Nuclear Safety Scction in the TechnicalServices Department.

e. Decemb~er 1976 - Manager - Corporate Nuclear Safety Section, System Plannir:g & Cocedination Department.

f. December 14, 1976 - Transferred to the Power Plant Engineering Department as Manager - Nuclear Plant Engineering Section.

g. January 13,1977 - Assigned as Manager of Engineering Pool Section of the Power Plant Engineering Department, accountable for cesigning and engineering safe, economical, constructible, and operable power plants that have as little impact on the natural environment as possible by interacting with necessary project entities and by managing, supervising, developing, training, and motivating and organization of Company employees to carry out such activities.

h. December 1,1979 - Assigned as Manager - Harris Plant Engineering Section of the Nuclear Power Plant Engineering Department, Shearon Harris Nuclear Power Plant, New Hill, N. C.

l III. PROFESSIONAL SOCIETIES A. ANS B. P.E. - California - 1976 l

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f EX9IBIT B

> 3 31 y3- ,1 ] SERIAL: LAP-83-354

. : .. a H m a .. ;. : . m f!)G 111983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNII NOS. 1 AND 2 DOCKET NOS. 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT P2SPONSES

Dear Mr. Denton:

Carolina Power & Light Company (CP&L) hereby trans. tits one original and forty copies of responses to Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report Open Items. The response numbers are listed on the cover page of the attachment along with the corresponding review branch and reviewer for each response.

We will be providing responses to other Open Items in the Draft

. Safety Evaluation Report shortly.

Yours very truly, M. A. McDuffie Senior Vice President Engineering & Construction J1E/ cc.c (7117JDK)

Attachment cc: Mr. E. A. Lici ra (NRC) Mr. Wells Eddleman Mr. G. F. Maxwull (NRC-SBNPP) Dr. Phyllis Lotchin Mr. J. P. O'Reilly (NRC-RII) Mr. John D. Runkle Mr. Travis Payne (KUDZU) Dr. Richard D. Wilson Mr. Danf el F. Read (CHANGE /ELP) Mr. G. O. Bright (ASLB)

Chapel Hill Public Library Dr. J. H. Carpenter (ASLB)

Wake County Public Library Mr. J. L. Kelley (ASLB) 7 _- -_ _ --_ ,---- - , 4 - _- -- j, .,3

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Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER)

Core Performance Branch open Item 31 (DSER Section 4.4.8, pages 4-47, 4-49, & 4-50)

Provide the itemized docuc;entation required by Item II.F.2 of NUREG-0737.

Response

The following information describes the instrumentation utilized for monitcring ICC and is organized per NUREG-0737, Ices II.F.2 " Documentation Raquired."

(1) Infot1aation utilized to give the operator an advance warning of the approach to ICC and to monitor the recovery from ICC, if it occurs, is obtained via a qualified instrumentation _ package. The information is ootained by the use of the Reactor Vessel Level Indicating System .(RVLIS) and core exit thermocouples.

The RYLIS is a fully qualified and redundant sysces for monitoring water inventory in the reactor vessel. Each of the two channels provide differential pressure calls and transmitters for narrow and teide range monitoring over the full length of the vessel, with the reactor coolant pumps off (natural circulation) and on respectively.

Additionally, narrow range monitoring is provided for each channel of the upper planum during natural circulation. Each channel's microprocessor utilizes these D/P signals in conjunction with other inputs such as RCS pressure, RCS temperature (loop RTD's or incore thermocouples), and RVLIS reference leg temperature sensors, to compensate for density changes in the system reference legs so as to provide direct water level readings available for operator use.

These water level readings will be displayed by redundant, qualified, alpha /atsaaric displays in the control room.

Incore thermocouples are utilized to determine core exit temperature. These 51 thermocouples outputs will be data linked to the ERFIS computer system for primary display on the SPDS CRT located on the MCB. Additionally, the thermocouple outputs are transmitted to a back up display in the control room. The display has a number of switches for selection based upon thermocouple identification and location cdte map.

The input to 'the ERFIS computer will'also be used to determine the margin of saturation which can be displayed on demand on the SPDS CRT or continuously on a s trip chart recorder.

The operator can also use Main Control Room display information in conjunction with s team tables to determine the margin to saturation.

m _ a

Shearon Harris Nuclear Power Plant Draft SER Open Item 31 (Cont'd)

Response (Cont'd)

(2) Design analysis and an evaluation of instruments to monitor water level, and available cast data to support the design described in itse 1 above may be found in NURIG CR-2628 regarding the Westinghouse design (RVLIS).

(3) A description of test programs conducted for evaluation and qualifiestica of the RVLIS was provided in NUREG CR-2626. For qualification of the thermocouples, see item (4) below.

Although the system sansors and microprocessors are not directly testable at power for calibration, the calculated parameter of margin to saturation can be readily verified at power through use of the stems tables and observation of tie independent indications of pressure and temperature. These cSeervations should show higher mergin to saturation since the system uses conservatively auctioneered values.

(4) An evaluation on the conformance of ICC instrumentation to item II.F.2 Attachment I and NUREG 0737 Appendix B is provided in NUREG CR-2628 for the RVLIS. RVLIS meets the intent of Regulatory Guide 1.97.

Technical Specifications will be prepared for the instrumentation specifically installed for the detection of inadequate core cooling. The Technical Specifications will be prepared considering the recommendation d NRC's Standard Tec,hnical Specifications (STS's) for Westinghouse P4R's (Rev.4) . CP&L is currently reviewing the Technical Specifications in Chapter 16.0 of the FSAR in view of the recommendations of Revision 4 of the Westinghouse STS; a revision of the Technical Specifications will be submitted to the NRC in the second quarter of 1984.

The thermoqouples meet the intent of design criteria outlined in II.F.2 Attachment 1 as indicated below:

i A.1 Thermocouples utilizad for the core exit for each core quadrant (in conjunction with core inlet temperature data) are sufficient to provide indication of radial distribution of the coolsac 3 enthalpy (temperature) rise across representative regions of the Core.

A.2 The primary display has the following capabilities:

(a) A spatially oriented core map indicating the temperature or temperature difference across the core (at each theraccouple location) is displayed on the CRT.

Shearon Harris Nuclear Power Plant Drafr SER Open Item 31 (Cont'd)

Response (Cont'd)

(b) A selective reading of core exit temperature which is consistent with parameters pertinent to operator actions in i connection with plant-specific inadequate core cooling procedures, will be displayed on demand.

(c) Direct readout and hard copy capability is available for all thermocouple temperatures. The range extends from '

200*F to 2300*F.

3ard copy will be provided by computer princout.

(d) Trend capability showing the temperature-time history of representative core exit temperature values is available on demand.

(e) Alarms are provided in the control room consistent with operator procedure requirements.

(f) The operator display device (CRT) interface will be located

' in accordance with humat-factor design in order to provida rapid access to requested displays.

A.3 A backup display is provided with the capability for selective reading of each of the operable thermocouples.

Backup display is provided in the control room.

A.4 The types and locations of displays and alarms will take into account:

(a) the use of this information by an operator during both normal and abnormal plant conditions; (b) integration into emergancy procedures; (c) integration into operator training; and (d) other alarms during emergencies and the need for prioritization of alarma Normal operating and emergency operating procedures are currently being developed and will be submitted at a later date. They will be available for onsite review six months prior to fuel load.

. _ . . . . . . .. . . . . . _J

Shearon Harris Nuclear Power Plant Draf t SER Open Item 31 (Cont'd)

Response (Cont'd)

A.5 The RVLIS instrumentation meets the requirements of Appendix B,

" Design and Qualification Criteria for Accident Monitoring L2strumentation," as modified by the provisions of items 6 through 9 below.

A.6 The primary and backup display channels are electrically independent, energized from independent station Class 1E power sources, and physically separated in accordance with Regulatory Guide 1.75 up to and including the isolation devices. The primary display and associated hardware beyond the isolation device are energized from a high reliability power source. The backup display and associated hardware is Class 1E.

A.7 Prisary and Backup displays are located in the control room envelope. Backup display will be completely qualified in accordance with IEEE 323 (1974) and 344 (1975) as defined in WCAP 8587/8687. The isolation device is located in an area which is accessible for maintenance following an accident.

A.8 The primary and backup display channels are designed to provide 99% availability for each channel with respect to functional capability to display a ainumum of four thermocouples per core quadrant.

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A.9 Quality assurance meets the requirements of 10CFR50 as applicable.

(5) For a description of the computer functions associated with ICC monitoring, refer to item (1) above.

(6) ICC Instrumentation, including associate.d testing, and calibration, will be installed before plant operation.

(7) Guidelines for the use of instrumentatica for monitoring ICC and the basis for these procedures will be provided af ter their completion.

(8) A description of key operator action instructions in the emergency procedures for ICC will be provided when available.

(9) Additional information to support the acceptability of the ICC' monitoring system will be provided by August 30, 1983 in the applicants response to Supplement 1 of NUREG 0737 (i tem R.G. I.97 Rev. 2)

(7292PS Anda)

" ^ ' ' " -- _- __.______-h _mO _ _ _ _ _ _ _ , _ _ _ , _ _ _ _ _ _ _ _ _ _ _ __ ___ _ _ _ _ _ _ _ _

EXHIBIT C SERIAL: LAP-83-521 Carolina Power & Light Company NOV 041983

)

Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation United States Nuclear Regulatory Conunission Washington, DC 20555 SHEARON RARRIS NUCLEAR POWER PLANT UNIT NO3. 1 AND 2 DOCKET NOS. 50-400 AND 50-401 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION

Dear Mr. Denton:

Carolina Power & Light Company hereby transmits one original and forty copies of additional information requested by the NRC as part of the safety review of the Shearon Harris Nuclear Power Plant. The cover sheet of the attachment sunnnarizes the related Open Items addressed in the attachment along with the corresponding review branch and reviewer for each response.

We will be providing responses te other requests for additional information shortly.

Yours very truly,

[ '

M. A. McDuffie Senior Vice President Nuclear Generation

! JHE/mf (8419COM)

Enclosures cc: Mr. B.C. Buckley (NRC) Mr. Wells Eddleman Mr. G.F. Maxwell (NRC-SHNPP) Dr. Ehyllis Lotchin Mr. J. P. O' Reilly (NRC-RII) Mr. John D. Runkle Mr. Travis Payne (KUDZU) Dr. Richard D. Wilson Mr. Daniel F. Read (CHANGE /ELP) Mr. G. O. Bright ( ASLB)

Mr. R. P. Gruber (NCtJC) Dr. J. H. Carpenter (ASLB)

Chapel Hill Public Library Mr. J. L. Kelley ( ASLB)

Wake County Public Library 411 Fayetteville Street

• P. O. Box 1551 e Raleegn. N C. 27602

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Shearon Harris Nuclear Power Plant Draf t SER Open Item No. 31 Provide the itemixed documentation required by Ite:s II.F.2 of NUREG-0737.

Response

The following information describes the instrumentetion utilized for monitoring ICC and is organized per NUREG-0737, Item II.F.2, " Documentation Required."

(1) Information utilized to give the operator an advance warning of the approach to ICC and to monitor the recovery from ICC, if it occurs, is obtained via a qualified instrumentation package. The information is

'obtained by the use of the Reactor Vessel Level Indicating System (RVLIS) and incore exit thermocouples.

(a) The Westinghouse RVLIS being installed at SHNPP represents the most recent Westinghouse design. It is a fully qualified and redundant system for monitoring water inventory in the reactor vessel. Each of the fewo channels provide differential pressure cells and trans-mitters for narrow and wide range monitoring over the full length of the vessel, with the reactor coolant pumps off (natural circulation) and on, respectively. Additionally, narrow range monitoring is provided for each channel of the upper plenum during natural circulation. Each channel's microprocessor utilizes these D/P signals in conjunction with other inputs such as RCS pressure, RCS temperature, (loop RTDs or incore thermocouples), RVLIS reference leg temperature sensors, to compensate for density changes in the system reference legs so as to provide direct water level readings available for operator use.

Qualified incore thermocouples are utilized to determine core exit temperature. These 51 thermocouples (26 channel A, 25 channel B) are inputs to and processed by the RVLIS microprocessors. Both RVLIS water level readings and incore exit thermocouple data will be data-linked to the ERFIS computer for primary display on the SPDS CRT which is located on the MCB. The data link is supplied from an isolated non-Class IE output from the qualified RVLIS micro-processors. Although ERFIS is non-class IE, it is powered from a high reliability power source. The isolation device cabinets and ERFIS are readily accessible and adjacent to the Main Control Room.

Additionally, qualified microprocessor outputs (RVLIS water level and thermocouple data) will be transmitted to dedicated redundant backup displays. These backup displays are alpha-numeric and qualified (class IE), and are located in the control room. The primary and backup displays have a selective capability for provid-ing RVLIS water level, thermocouple data, and temperature mapping functions.

l The input to the ERFIS computer will also be used to determine the margin of saturation which can be displayed on demand (at operator request) on the SPDS CRT or continuously on a strip chart recorder. The plant computer (ERFIS) processes and calculates subcooling data using temperature and pressure signals from the reactor coolant system. Displayed information includes margin of subcooling data both graphically and in engineering units.

7.n accordance with the provision of Regulatory Guide 1.97 Rev. 3 operator confirmation of subcooling data is provided through the use of qualified pressure and temperature signals and AS!fE steam tables.

(b) Existing instrumentation which provides operating information.

pertinent to ICC considerations consist of the non-safety incore thermocouple system and a digital list of thermocouple temperatures readout. This is being replaced by the system as described in Item (1)(a) above.

(c) Modifications to the instrumentation systems described in Item 1(b) above include upgrading the incore thermocouples, connectors, reference junction boxes RTDs and cables in order to be qualified in accordance with the IEEP 344 (1975) and IEEE 323 (1974), the procurement of a qualified RVLIS, and the procurement of redundant -

integrated plant process /;.ergency response computers.

(2) Design analysis and an evaluation of instruments to monitor water level, and available test data to support the design described in Item (1),

above may be found in NUREG CR-2628 regarding the Westinghouse RVLIS design and will be available later for the incore exit thermocouple instrumentation.

(3) A description of test programs conducted for evaluation and qualification of the RVLIS was provided in NUREG CR.-2628. For qualification of the thermocouples, see Item (4) below.

Although the system sensors and microprocessors are not directly testable at power for calibration, the calculated parameter of margin to satura-tion can be readily verified at power through use of the steam tables and observation of the independent indications of pressure and temperature.

These observations should show higher margin to saturation since the system uses conservatively auctioneered values.

(4) An evaluation on the conformance of ICC instrumentation to item II.F.2, Attachment 1, and NUREG-0737, Appendix B , is provided in NUREG CR-2628 for the RVLIS. RVLIS meets the intent of Regulatory Guide 1.97.

Technical specifications will be prepared for the instrumentation specifically installed for the detection of inadequate core cooling. The technical specifications will be prepared considering the recommendations of NRC's Standard Technical Specifications (ETS) for Westinghouse PWRs (Rev. 4). CP&L is currently reviewing the technical specifications in Chapter 16.0 of the FSAR in view of the recommendations of Revision 4 to the Westinghouse STS; a revision to the technical specifications will be submitted to the NRC in the second quarter of 1984.

l The thermocouples meet the intent of design and qualification criteria outlined in II.F.2, Attachment Ic as indicated below: _

A.1 Thermocouples utilized for the core exit for each core quadrant (in conjunction with core inlet temperature data) are sufficient to provide indication of radial distribution of the coolant enthalpy (temperature) rice across representative regions of the core.

A.2 The primary display has the following capabilities:

(a) A spatially oriented core map indicting the temperature or temperature difference across the core (at each thermocouple location) is displayed on the CRT.

(b) A selective reading of core exit temperature, which is consistent with parameters pertinent to operator actions in connection with plant-specific inadequate core cooling procedures , will be continuous on demand.

(c) Direct readout and hard copy capability is available for all thermocouple temperatures. The range extends from 200 F to 2300 F.

Hard copy will be provided by computer printout.

(d) Trend capability showing the temperature-time history of representative core exit temperature values le available on demand.

(e) Alarms are provided in the control room. Thest alarms will be set to be consistent with the decision points i t the emergency operating procedures (refer to Items A.4 and (2 6 below) .

(' f) The operator display device (CRT) interface will be located in accordance with human-f actor design in order to provide rapid access to requested displays. CP&L's human factors methodology for the main control board has been provided to the NRC in a submittal dated June 1, 1983. This document identified the methodologies and human engineering requirement specifications which apply to items such as ICC instrumentation, which were not-defined when the Detailed Contrcl Room Design Review was performed.

A.3 A backup display is provided with the capability for selective reading of each of the operable thermocouples. The range extends 6

from 200 F to 2300 F.

The backup display provided, which is in the control room, is described in Item 1 above.

A.4 The types and locations of displays and alarms will take into account the following:

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4 (a) The use of this information by an operator during both. normal and abnormal plant conditions (b) Integration into emergency procedures (c) Integration into operator training (d) Other alarms during an emergency and need for prioritization of alarms.

Normal operating and emergency operating- procedures are currently being developed and will be available for onsite review six months prior to fuel load (January 1985).

A.5 The instrumentation meets the requirements of Appendix B, " Design and Qualification Criteria for Accident Monitoring Instrumentation,"

as modified by the provisions of Items (6) through (9) below.

A.6 The primary and backup display channels are electrically indepen-dent, energized from independent str.ti.on Class 1E power sources, and physically separated in accordance with Regulatory Guide 1.75 up to i and including the isolation devices. The primary display and i associated hardware beyond the isolation device are energized from a high reliability power source. The backup display and associated hardware is Class 1E. Refer to Item 1 above.

A.7 Primary and backup display are located in the control room envelope.

Backup display will be completely qualified in accordance with IEEE 323 (1974) and 344 (1975) as defined in WCAP 8587, " Methodology for Qualifying Westinghouse WRD Supplied Safety Related Electrical Equipment" and WCAP 8687, " Equipment Qualification Test Reports."

i The isolation device is located in an area.which is accessible for maintenance following an accident.-

A.8 The primary and backup display channels are designed to provide 99%

availability for each channel with respect to functional capability to display a minimum of four thermocouples per core quadrant. This can be accomplished since each quadrant will contain a minimum of four thermocouples for each of Train A and Train B. ICC systems will be~ addressed in the technical specifications.

A.9 Quality assurance meets the requirements of 10 CFR 50 as applicable.

This is further addressed in the applicants response to Supplement 1 to NUREG-0737 (R' e g. Guide 1.97) dated September 6,1983.

(5)- For a description of.the computer functions associated with ICC monitoring, refer to Item (1) above.

1 l (6) ICC instrumentation will be installed and preoperational tests will be completed before fuel load. Startup tests and calibrations which require

.the core to be in place will be completed prior to operation above 10

' percent of full power.

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l l-Procedures (FRPs) will incorporate the Westinghouse Owners' Group Emergency Response Guidelines and Functional Restoration Guidelines.

These procedures employ inadequate core cooling (ICC) instrumentation -

(RVLIS, the core exit thermocouples, and the subcooling data) along with other post-accident monitoring capabilities (i.e. , reactor coolant system

, pressure, reactor coolant pump status, and safety injection flow).

Therefore, SHNPP instrumentation for monitoring ICC will be used in accordance with the emergency response guidelines developed by the Westinghouse Owners' Group. The emergency response guidelines were accompanied by extensive analysis of the setpoints used in the critical safety function status tree and the functional restoration guidelines.

These analyses are refetenced in WOG Revision 1 (High Pressure Plant)

Emergency Response Guidelines. SHNPP E0Ps and FRPs will be completed by January 1985. The development of these procedures will include details

-such as the specification of SHNPP-specific setpoints; these setpoints will account for instrumentation uncertainties which are specific to

'SHNPP equipment. A draft copy of the EOP for ICC will be provided for NRC's information by January 1985.

I

' (8) The SENPP E0P for ICC will refer the operator to functional restoration procedures based on the readings on the ICC instrumentation. The SHNPP functional restoration procedures will incorporate the Westinghouse Owners' Group Functional Restoration Guidelines C.1, C.2, and C.3. The actions specified for the operator are fully addressed in WOG submittals and are briefly described below.

F RG C.1 This guideline will be used when indicated by the core cooling critical safety function status tree ( refer to Attachment 1). The operator actions specified include:

(a) Verify safety injection actuation and flowpath alignment.

(b) Align and actuate systems required to support reactor coolant pump operation, l

(c) Monitor containment hydrogen concentration.

(d) Operate pressurizer PORV, if necessary.

(e) Operate steam system PORVs.

(f) Actuate reactor coolant oumps.

The sequence, priority, and action levels for the listed actions'are based on ICC instramentation and other post-accident monitoring capability.

FRG C.2 This guideline will be used when indicated by the core cooling critical safety function status tree (see Attachment 1).

The operator actions specified include:

(a) Verify safety injection actuation and flowpath alignment.

l l

(b) Align and actuate systems required to support reactor coolant pump <

operation. l (c) Observe trend in inadequate core cooling instrumentation. j (d) Operate steam system PORVs.

The sequence, priority, and action levels for the listed actions are based on ICC instrumentation and other post-accident monitoring capability.

FRG C.3 This guideline will be used when indicated by the core cooling critical. safety function status tree (refer to Attachment 1).

The operator actions for this guideline include:

I (a) Verify safety injection actuation and flowpath alignment.

(b) Verify that pressurizer PORVs and reactor vessel head vents are closed.

(9) Additional information to support the acceptability of the ICC monitoring system was provided in the applicant's response to Supplement 1 of NUREG-0737 (Reg. Guide 1.97), dated September 6,1983. Further information regarding test data for the incore exit thermocouples will be available

.by March 31, 1984. A draft emergency operating procedure for ICC will be submitted by Jan'uary 4, 1985. Changes subsequent to the design and operation of the ICC instrumentation as described in the FSAR will be reported to the NRC in accordance with 10 CFR50.59.

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ATTACHMENT 1 PAGE 1 Number:

Title:

Rew Issue /Date:

HP/LP, REV.1 p.0.2 CORE COOLING 1 Sept.,1983 0070 FR4.1 R C.

RIIXlT 3

+ ksLISg L3ANCE THAN12000F GREAllR YES THAN t2) l Y,3 s LESS A*

TMAM 7004 YES c0 TO l

FR C.2 i ,

No AT LEAST ONERCP -

Rytts NO RUNNING FUU. RANGE I YES GREATER THAN (2) ygg RCS NO Sil5C00 LING l GREATEA THAN W YEs G0jj 3,

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'RVLIS DYN AMIC HEAD RANGE NO QREATER THAN (3)-4 RCP (4)-3 RCP YES (5)-2 RCP (6)- t RCP R.$

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ATTACHMENT 1 PAGE 2 CRtTICAL HP/LP, REV.1 F-0 SAFETY FUNCTION STATUS TREES 1 Sept.,1983 FCOTNOTES F-0.2 CORE COOLING (1) Enter sum of temperature and pressure measurement system errors, including allowances for normal channel accuracles and post accident transmitter

errors, translated into temperature using saturation tables,

! (2) Enter plant specific value which is 3-1/2 feet above the bottom of active fuelin core with zero vold fraction, plus uncertainties.

(3) Enter plant specific value corresponding to en average system vold fraction of 50 percent with 4 RCPs running, plus uncertainties.

(4) Enter plant specific value corresponding to an average system vold fraction of

. 50 percent with 3 RCPs running, plus uncerialnties.

(5) Enter pf ant speciffo value corresponding to an average system vold fraction of 50 percent with 2 RCpa running, plus uncertaintles.

(6) Enter pf ant specifte value corresponding to an average system void fraction of 50 percent with 1 RCP running, plus uncertainties.

l l F-0.3 HEAT SINK l (1) Enter plant speciflo value showing SG loveljust ln the narrow range, including allowancesfornormatchannelaccuracy, post accidenttransmittererrors,and reference leg process errors, not to exceed 80%.

(2) Enter the minimum safeguards APW flow requirement for heat removal, plus allowancesfornormatchannef accuracy (typleallyone MD AFWpumpeapaa-Ity at 80 desigrt pressure).

(3) Enter plant speciffe pressure for highest steamilne safety valve setpoint.

(4) Enter plant specific value for SG high hlgh level faWWwater Isolation setpoint.

l

'(5) Enter plant speciffo pressure for lowest steamline safety valve setpol,nt.

[ .

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E' y.,

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

CAROLINA POWER & LIGHT COMPANY ) Docket Nos. 50-400 OL and NORTH CAROLINA EASTERN ) 50-401 OL MUNICIPAL POWER AGENCY )

)

(Shearon Harris Nuclear Power )

Plant, Units 1 and 2) )

CERTIFICATE OF SERVICE I hereby certify that copies of " Applicants' Motion for Summary Disposition of CHANGE Contention 44 and Eddleman Contention 132," " Applicants' Statement of Material Facts -

as to Which There is No Genuine Issue to be Heard (CHANGE Contention 44 and Eddleman Contention 132)," " Affidavit of Clarence G. Draughon" and " Affidavit of Leonard I. Loflin" were served this 7th day of December, 1983, by deposit in

+

the U.S. mail, first class, postage prepaid, to the parties on the attached Service List.

1 "h:- em ,

Thomas A. Baxter, P.C.

4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

CAROLINA POWER & LIGHT COMPANY ) Docket Nos. 50-400 OL and NORTH CAROLINA EASTERN ) 50-401 OL MUNICIPAL POWER AGENCY )

)

(Shearon Harris Nuclear Power )

Plant, Units 1 'and 2) )

SERVIC LIST James L. Kelley, Esquire John D. Runkle, Esquire Atmlic Safety and Licensing Board Conservation Council of North Carolina U.S. Nuclear Regulatory Cmmission 307 Granville Road Washington,-D.C. 20555 Chapel Hill, North Carolina 27514 Mr. Glenn O. Bright M. Travis Payne, Esquire Atanic Safety and Licensing Board Edelstein and Payne U.S. Nucle _c Regulatory Cm mission P.O. Box 12607 Washington, D.C. 20555 Raleigh, North Carolina 27605 Dr. James H. Carpenter Dr. Richard D. Wilson Atanic Safety and Licensing Board 729 Hunter Street U.S. Nuclear Regulatory Ctmnission Apex, North Carolina 27502 Washington, D.C. 20555 Mr. Wells Eddlanan

[

Charles A. Barth, Esquire 718-A Iredell Street Janice E. Moore, Esquire Durhan, North Carolina 27705 >

Office of Executive Iegal Director U.S. Nuclear Regulatory Camission Richard E. Jones, Esquire Washington, D.C. 20555 Vice President and Senior Counsel Carolina Power & Light Canpany

, Docketing and Service Section P.O. Box 1551 l Office of the Secretary Raleigh, North Carolina 27602 l U.S. Nuclear Regulatory Camission j Washington, D.C. 20555 Dr. Phyllis Ictchin 108 Bridle Run Mr. Daniel F. Read, President Chapel Hill, North Carolina 27514 i CHANGE /ELP l 5707 Waycross Street Dr. Linda W. Little l Raleigh, North Carolina 27606 Governor's Waste Management Board l 513 Albenarle Building l

325 North Salisbury Street l Raleigh, North Carolina 27611 l

l

Bradley W. Jones, Esquire U.S. Nuclear Regulatory Ctmuission Region II 101 Marrietta Street Atlanta, Georgia 30303 Ruthanne G. Miller, Esquire Atcmic Safety and Licensing Board Panel U.S. Nuclear Regulatory Cmmission Washington, D.C. 20555 Mr. Robert P. Gruber Executive Director Public Staff - NCUC P.O. Box 991 Raleigh, North Carolina 27602 O

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