ML19321A041

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Forwards Updated Pages to Util 800523 Response to TMI Concerns.Matl Addresses NUREG-0694 Requirements.Another Revision Scheduled for 800801 & Fuel Loading for mid-Sept 1980
ML19321A041
Person / Time
Site: McGuire, Mcguire  
Issue date: 07/18/1980
From: Parker W
DUKE POWER CO.
To: Harold Denton, Youngblood B
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0694, RTR-NUREG-694 MNS-TMI-80-02, MNS-TMI-80-2, NUDOCS 8007220365
Download: ML19321A041 (44)
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{{#Wiki_filter:. DUKE POWER COMPANY Powra Dun.nswo (- t 422 Socra Cnuncu Srazzi. CnAntoriz. N. C. 2e24a N.)! wi we. o. e.a n ca. s a. July 18, 1980 VsCE Pets 1QENT 7tL EP=O8s t:A8CA 704 Strana 8-ooowevio*e 373-4063 Mr. Harold R. Denton, Director MNS-TMI/80-02 Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 i Attention: Mr. B. J. Youngblood Licensing Projects Branch No. 1

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370

Dear Mr. Denton:

Enclosed with this letter are forty copies of updated responses to the document A " Duke Power Company, McGuire Nuclear Station, Response to TMI Concerns." This document was transmitted to the NRC by my letter of May 23, 1980. In June of 1980 the NRC consolidated the TMI-related licensing requirements for NTOL plants into NUREG-0694: TMI-Related Requirements for New Operating Licenses. " Duke Power Company, McGuire Nuclear Station, Response to TMI Concerns" as updated by the attached material addresses each of the NUREG-0694 requirements as completely as is possible at this time. Another revision to this document is currently scheduled for August 1, 1980. Fuel loading for McGuire Unit 1 is currently scheduled for mid-September 1980. Please schedule your review of this document accordingly. Very truly yours, s/ William O. Parker, Jr. William O. Parker, Jr. THH:scs Attachments (40) O) THIS DOCUMENT CONTAINS ( POOR QUAllTY PAGES ,8007220'3.f5 t

Mr. Harold R. Denton, Director' July 18, 1980 Page Two WILLIAM 0. PARKER, JR., being duly sworn, states that he is Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this document, Duke Power Company McGuire Nuclear Station Response to TMI Concerns, and that all state-ments and matters set forth therein are true and correct to the best of his knowledge, i s/ William O. Parker, Jr. William O. Parker, Jr., Vice President i s Sebscribed and sworn to before me this 18th day of July, 1980. s/ Sue C. Sherrill 4 Notary Public s t( My Commission Expires: September 20, 1984 4 i .I l j i l f f I I s

~ l i DUKE PO'nER COMPANY I MCGUIRE NUCLEAR STATION i RESPONSE TO TMI CONCERNS 4 Changes and Corrections: Remove and insert pages in accordance with the following table. Re:nove These Pages Insert These Pages Table of Contents, Page 2-Table of Contents, Page 2 07/18/80 I-4 I-4 07/18/80 I-10 I-10 thru I-10F 07/18/80 I-17 I-17 07/18/80 i II-2 II-2 07/18/80 II-3 II-3 07/18/80 II-8A, I -8B 07/18/80 II-9 II-9, II-10, II-11, 11A 07/18/80 II-12, II-13 II-12, 11-13, II-13A, II-13B, II-13C 07/18/80 II-15, II-16 II-15, II-16, II-16A 07/18/80 11-17 II-17, II-17A, II.7B 07/18/80 11-18 II-18, II-18A 07/18/80 III-4 III-4 07/18/80 Appendix A Appendix A Station Directive 3.1.4 Station Directive 3.1.4, Rev. 1 Appendix C March 28 letter of H. R. Denton Appendix C Appendix C SECY80-230 Enclosure i NUREG-0694 9e s s y ,,7

.=. ) Appendix A McGuire Nuclear Station Procedures Station Directive 3.1.4, Conduct of Operations Station Directive 3.8.2, Station Emergency Organization Station Directive 3.1.9, Relief at Duties of Plant Operation Periodic Test PT/1/A/4700/10, Shis: Turnover Verification 4 b Appendix B Control Room Design Preliminary Report \\ i i. I I i N.s i u Appendix C NRC Requirements for McGuire Nuclear Station September 27, 1979 letter from D. B. Vassallo to al) Pending Operating License Applicants November 9, 1979 letter from D. B. Vassallo to all Pending Operating License Applicants March 28, 1980 letter from H. R. Denton to all Power Reactor Applicants and Licensees NUREG-0694: IMI-Related Requirements for New Operating Licenses i i 2 07/18/80

=- SHIFT MANNING ' ( 'N )

Reference:

Action Plan - I.A.1.3 The shift' crew composition for operation of McGuire Unit I will be in accor-dance with Section 6.0 of the McGuire Technical Specifications. The minimum shift manning for Unit I will be one shift supervisor (SRO), one senior reac-tor operator in the Control Room (SRO), two reactor operators (R0 License - i one in Control Room at all times), two nuclear equipment operators (non-licensed personnel). In addition, a shif t technical advisor will be assigned to each shift in a strictly advisory capacity. A licensed SRO will be stationed in the Control Room whenever the unit is operating in Modes 1, 2, 3 or 4. If the shift technical advisor is the only SRO in the Control Room, the shift supervisor or assistant shift supervisor will be available to the Control Room within ten minutes. An administrative procedure will be written before fuel loading to govern the amount of overtime worked by licensed operators. This procedure will limit work hours to no more than twelve hours of continuous duty with at least twelve hours between work periods, no more than seventy-two hours in any seven-day period, and no more than fourteen consecutive days of work without at least two consecutive days off. However, this procedure will include provisions which allow for circumstances which may require deviation from the above limitations. This deviation will be authorized and documented by either the 'h Superintendent of Operations or the Station Manager of McGuire Nuclear Station. wJ 4 1 1 J t A t \\ / V I-4 07/18/80

TRAINING DURING LOW POWER TESTING O ?

Reference:

Action Plan - I.G.1 Duke Power Company will conduct a series of special low power tests at McGuire Nuclear Station for the purposes of providing technical information beyond that obtained in the normal startup test program and to provide supplemental operator training. These tests will be performed at reactor power levelp no greater than five percent of full power. Each Reactor Operator and Senior Reactor Operator at McGuire will either observe or participate in each test. 'The low power tests to be performed at McGuire are listed below: Natural Circulation Verification Effect of Steam Generator Isolation on Natural Circulation Natural Circulation with Simulated Loss of Offsite Power Natural Circulation at Reduced Pressure Natural Circulation with Loss of Pressurizer Heaters Simulated Loss of All Onsite and Offsite AC Power One page abstracts of each of these tests are provided in the pages which follow. A safety analysis of this program is being prepared and will be submitted for NRC review by mid August, 1980. f-~s\\ V l fQj) I-10 07/18/80

T. IP/1/A/2150/20 s ) Natural Circulation Verification Purpose Verify that natural circulation is established in the primary system following the loss of forced reactor coolant flow. 4 Prerequisites Reactor Power <3% Normal primary system temperature and pressure. Reactor Coolant Pumps operating. Pressurizer Heaters controlling pressure. Steam Generators being fed by normal feedwater. l Test Method The test will be initiated by tripping all reactor coolant pumps. The establishment of natural circulation will be verified by observing the response of the hot and cold leg temperature instrumentation in each loop. Acceptance Criteria ,a Natural circulation is successfully established in all four reactor coolant loops. l i 1 l l 1 s.,) I-loa 07/18/80 ~

TP/1/A/2150/21 [ } Effect of Steae Generator Isolation on Natural Circulation tV Purpose To observe the effects of steam generator isolation on natural circulation and to verify that these ef: facts do not adversely affect the performance of the operable primary loops. Prerequisites Reactor Power <1% Primary system temperature -515 F and pressure ~2000 psig. Reactor Coolant Pumps shutdown. Pressurizer Heaters controlling pressure. Steam Generators being fed by normal feedwater. Natural Circulation established in all 4 loops. Test Method With natural circulation successfully established in all four primary loops, cool the reactor coolant system down to provide sufficient margin to the steam generator safeties. Isolate steam generators one at a time while p adjusting feed flow to the unisolated steam generators to maintain a coolant ( { primary system average temperature. Continue this until two steam generators N._ / remain unisolated or until the average primary system temperature can no s ~l longer be prevented from increasing. The hot and cold leg temperatures will be observed to ensure the core is being sufficiently cooled by the natural l circulation process. The steam generators will then be returned to service one at a time and natural irculation will be verified to reestablish in each loop. l Acceptance Criteria 1. Natural circulation in the unisolated loops is not interrupted or sig-nificantly impaired by the isolation of steam generators on other loops. 2. Natural circulation can successfully be reestablished on the previously isolated loops. j 1 l l 1 1 ( 3 LJ I-10B 07/18/80

TP/1/A/2150/23 O) Natural Circulation with Simulated Loss of Offsite Power ( Purpose Verify that natural circulation cooling can be maintained following a loss of offsite power. i Prerequisites Reactor Power <1% Normal primary system temperature and pressure, j Reactor Coolant Pumps operating. Pressurizer Heaters controlling pressure. Steam Generators being fed by auxiliary feedwater. Test Method The test will be initiated by simulating a loss of offsite power. The reactor coolant pumps-will be tripped and the auxiliary feed pump and pressurizer heater loads will be transferred to diesel power. Natural circulation will be verified by observing the response of the hot and cold leg temperature instrumentation on each loop. Acceptance Criteria Natural circulation is successfully reinstated in all four loops following the interruption of feedwater flow to the steam generators. e. l I. I-loc 07/18/80 ,,_,r.

f i TP/1/A/2150/24 [{)-. Natural Circulation at Reduced Pressure g ~ %.f Purpose i

1. - 'To provide Operations personnel with online experience in using the

',0perator Aid Computer Saturation Monitor Program to monitor and control margin to saturation. 2. To _ verify that changes in saturation margin will not affect natural circulation provided that adequate margin to saturation exists. Prerequisites 1 Reactor Power <3% Normal primary system temperature and pressure. Reactor Coolant Pumps = operating. i Pressurizer Heaters controlling pressure. Steam Generators being fed ty normal feedwater. Test' Method-The test will be initiated by tripping all reactor coolant pumps an (N verifying the establishement of natural circulation. As the primary 1 ( system temperature is reduced, the primary system pressure will be reduced. The Saturation Monitor Program results will be monitored and compared against hand calculations. The effect of pressure reduction on natural circulation will be observed. Acceptance Criteria 1. Natural Circulation is successfully established ir all four reactor coolant loops. 2. The Operator Aid Compucer Saturation Monitor Program provides reliable information concerning the margin to saturation. 3. Reducing _the margin to saturation does not impair natural circulation as long as sufficient margin exists. 4 1 e u I-100 07/18/80

TP/1/A/2150/25 p-Natural Circulation with Loss of Pressurizer Heaters Purpose Verify establishment of natural circulation and determine the rate of decrease of margin to saturation while in this mode and the ability to reestablish margin through cooldown and makeup. Prerequisites l Reactor Power <3% Normal primary system temperature and pressure. Reactor Coolant Pumps operating. Pressurizer Heaters controlling pressure. Steam Generators being fed by normal feedwater. Test Method 4 The test will be initiated by tripping all reactor coolant pumps and the pressurizer heaters. Establishment of nctural circulation will be verified by observing the response of the hot and cold leg temperature instrumentation in each loop. Operations personnel will observe the Saturation Monitor to assure margin. Prior to reaching saturation, secondary side steam flow will be increased to affect a cooldown and a reestablishment of saturation g' margin will be verified. Any shrinkage associated with cooldown will be i compensated for by reactor water makeup. Acceptance Criteria 1. Natural circulation is successfully established in all four reactor ] coolant loops. 2. Natural circulation is successfully maintained during the depressuriza-tion associated with the loss of pressurizer heaters. 3. Margin to saturation is successfully reestablished following the depressurization associated with the loss of pressurizer heaters. - v) 1 1-10E 07/18/80

TP/1/A/2150/26 .) SIMULATED LOSS OF ALL ONSITE AND OFFSITE AC POWER (,,j Purpose 1. To demonstrate that auxiliary feedwater can be controlled by manual means to remove heat energy from the primary system in order to maintain hot standby conditions. 2. To demonstrate that critical plant operations can be performed using emergency lighting. '3. To demonstrate the ability of 125 volt battery to supply the 125 volt vital AC. 4. To demonstrate that selected equipment areas do net exceed maximum design temperatures. Prerequisites Reactor subcritical by > 1.6% Ak/k. Normal primary system tempera ure and pressure. -Reactor Coolant Pumps operating. Pressurizar Heaters controlling pressure. ss ) Steam Generators being fed by auxiliary feedwater. %J Test Method Test will be initiated by simultaneously performing the following: 1. Tripping the pressurizer heaters. 2. Removing AC power from the auxiliary feedwater components. 3. Tripping selected space and equipment coolers. 4. Tripping vital battery chargers and AC power to inverter. 5. Isolating main feedwater and main steam lines. Operations personnel will then establish manual control of the auxiliary feedwater system and provide sufficient flow to the steam generators to maintain a constant primary system temperature. These conditions shall be ma'intained for one hour after which AC power will be restored and equipment returned to normal service. i Acceptance Criteria 1. Emergency lighting in the station is sufficient to operate critical equipment with the loss of all normal lighting. 2. Hot standby conditions can be maintained for a 1-hour period with critical equipment operating off of vital battery power. 3. Manual operation of auxiliary feedwater valves and main steam power [ h reliefs can be coordinated by the operators. '\\_,) 4. Cirtical equipment areas do not exceed design. temperature limits. I-10F 07/18/80

CONTROL ROOM DESIGN \\j

Reference:

Action Plan - I.D.1 A detailed design review has been instituted for the McGuire control room to: 7 (A) Evaluate the amount, type, and form of information available to the unit operator. (B) Review the requirements for control by the operator, and assess the degree to which the design of the control boards enables the operator to perform (or constrains him from performing) his duties safely and efficiently, and minimizing the potential for operator error. 4 (C) Identify and implement those changes and modifications to equip-ment, its arrangement and identification, on a schedule addressing the concerns for safety and operability. This review has revealed two classifications of corrective actions, those to be implemented before fuel loading, and those to be implemented before startup following the first refueling. A human factors engineering design review of the McGuire control room was conducted during the week of June 2-6, 1980 by the NRC/ Human Factors Engineering Branch. Duke has received and is currently evaluat-ing the review team's initial report. i i 8 V 1 j I-17 07/18/80 ( 1 I l. L i-

RELIEF AND SAFETY VALVE TESTING 1./ Referr s: NUREG-0578 - 2.1.2 Action Plan - II.D.1 Test Program Duke Power Company is participating in and monitoring an EPRI testing program to qualify the McGuire relief and safety valves under expected operating con-dicions for design basis transients and accidents. By letter dated December 17, 1979, Mr. W. J. Cahill,'Jr., Chairman of the EPRI Safety and Analysis Task Force submitted " Program Plan for the Performance 7erification of PWR Safety / Relief Valves and Systems," December 13, 1979. EPRI formally presented an updated safety / relief valve program plan to the NRC on February 28, 1980. Duke will assure that this program is applicable to the McGuire design and that it will provide sufficient information to accomplish its purpose. The EPRI program calls for testing both the McGuire relief and safety valves. Upon completion of the test program and evaluation of the test results Duke will submit test data to the NRC. This data will provide evidence that the McGuire relief and safety valves will open and reclose under the expected McGuire flow conditions for the expected McGuire operating and accident (non-ATWS) conditions. Test data submitted would include criteria-for success and failure of valves tested and would permit evaluation of discharge piping and } supports which are not tested directly. l The EPRI program also calls for the development of correlation codes for analyzing the effect of relief and safety valve discharge piping on valve operability. Duke will use these codes to correlate the test loop piping with the McGuire discharge piping. Duke Valve Testing l Duke Power Company has established a full scale valve testing facility at Marshall Steam Station IInit No. 2. This facility will be used to perform the EPRI steam flow testing of power operated relief valves (PORV). Valves identical to both the McGuire PORV and PORV block valve were tested at Marshall for~ steam flow at McGuire full temperature and pressure conditions. Several iterations of valve modifications were performed and tested until both valves met'all functional and design requirements. These modifications were then performed on the PORV's and PORV block valves at McGuire. t j'"%. k q_,/ II-2 07/18/80 ) l l ~ -n-

AUXILIARY FEEDWATER INITIATION AND INDICAT.~.ON ,~. \\' } l

References:

NUREG-0578 - 2.1.7a and 2.1.7) Action Plan - II.E.1.2 Automatic Initiation Safety-grade automatic initiation and safety-grade emergency power for the Auxiliary Feedwater System are features of the McGuire Nuclear Station design (reference FSAR Ch. 10). The automatic initiation circuitry for the Auxiliary Feedwater System meets the single failure criteria. Additionally, for most failures which could prevent the automatic start of an individual auxiliary feedwater pump, manual initiation of the affected pump is available from the Control Room. However, should the auxiliary feedwater pump in one safety train not be available due to any single failure, the redundant safety train is available with no loss of system function. In the final stages of plant shutdown, the motor-driven auxiliary feedwater pumps must be tripped. For this reason the automatic auxiliary feedwater pump start upon trip of both main feedwater pumps or steam generator low-low level must be bypassed. This bypass is accomplished manually by means of a bypass switch located in the Control Room. This bypass is administratively controlled ( g by use of operating procr ures. When the bypass switch is in the bypass posi-i (s, / tion, an annunciator is actuated and a status light is illuminated on the by-pass status' panel as required by Regulatory Guide 1.47. The turbine-driven auxiliary feedwater pump does not have a bypass feature. Indication For the short term, existing control grade flow instrumentation in the lines to each steam generator will be relied upon. Backup for these instruments, in case of single failure, is available from existing control grade flow instrumen-tation in the suction piping to each auxiliary feedwater pump. Further backup is available in the form of safety-grade steam generator level indication. The present control grade flow instrumentation is powered from the highly reliable battery-backed 120VAC Auxiliary Control Power System (Ref. FSAR Section 8.3.2.1.3). Provisions for tests and calibration are included in the design of the control grade flow instrumentation. Safety grade indication of auxiliary feedwater flow to each steam generator will be provided in the Control Room by Janua v 1,1981. Provisions for calibration and testing will be incorporated into the design of this instrumentation. t v II-3 07/18/80

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INADEQUATE CORE COOLING INSTRUMENTS \\ x'

References:

NUREG-0578 - 2.1.3b Action Plan - II.F.2 u Subcooling Monitor The margin to saturation will be calculated from reactor coolant system pressure and temperature measurements (wide-range and low-range pressure and wide-range hot and cold leg temperature and temperature f rom in-core thermocouples). The thermocouple readings (approximately 60) are averaged and compared with the wide range RTD values. The highest of these temperatures and the lowest pressure are then used to calculate margin to saturation. Averaging of the thermocouple read-ings and calculation of margin to saturation are performed by the plant computer. The computer output consists of a CRT graphic display of margin to saturation conditions, i.e., a plot of plant pressure and temperature in relation to a computer generated satura. on curve. Additionally, this display also indicates and in numerical terms RCS temperature, pressure, power level, margin to Tsat, Alarm status is indicated by flashing the alarming parameter on margin to Psat. the CRT display and by printout on the typewriter. Two alarm setpoints are pro-vided for both T and P The alarm setpoint is dependent on reactor power. sat sat. Further details on this subcooling meter are provided in the table which follows. /} Normal control board instrumentation for RCS temperatare and pressure will be ( j used in conjunction with a control room copy of the steam tables and a written procedure to determine margin to saturation as a backup to the computer calculation. This system for determining the degree of subcooling will be fully operational by fuel loading. Rcactor Vessel Level Measarement Duke will install the Westinghouse designed reactor vt.el level measurement system in McGuire Unit 1. This system is designed to monitor directly the water level in the reactor vessel, or the approximate void content under forced circula-tion conditions, during certain postulated accident conditions. Included is equipment to monitor both the upper plenum (head) level, as well as the entire height of the reactor vessel. The system instrumentation permits vessel level measurement from the bottom to the top of the reactor vessel, utilizing taps off of an existing spare head penetration and a tap off of a thimble tube at the seal table. Two sets of differential pressure transmitters are provided which have differing measurement ranges to cover different flow behavior with and without pump operations. The narrow range cells indicate water level when zero or one reactor coolant pump is operating. The wide range cells indicate the combined core and inte nals pressure drop for any combination of operating reactor coolant pumps. The upper plenum measurement is taken by two differential pressure transmitters between ~' fy) the same spare head penetration, and taps off two hot legs. \\.s' To minimize containment post-accident environment effects in measurement, accuracy, the' system design is based upon locating the transmitters outside the II-9 07/18/80

1 containment. Hydraulic isolators in the impulse lines provide the required double barrier protection between the RCS and outside containment. Reference leg temperature measurements, together with the existing RCS temperature and pressure, are utilized to automatically compensate for difference in coolant and reference leg temperature effects. The best schedule for delivery of the sensors and transmitters utilized in this system is April, 1981. The design support for installation of this system will be as complete as possible prior to receipt of this equipment. Installation and functional testing will be performed during the first subsectent outage of sufficient duration. I ' O J. II-10 07/16/80 1 I

INFORMATION PROVIDED ON THE SUBC00 LING METER V Display T-Tsat, P-Psat Temp., Press (Wide Range) Information Displayed (T-Tsat, Tsat, Press, etc.) % Power, Alarms Display Type (Analog, Digital, CRT) CRT Continuous or on Demand DEMAND Single or Redundant Display SINGLE i Location of Display CONTROL ROOM >0X FP <0% FP F 5 Alarms (include setpoints) Setpoints: .200 to.750 PSIA Overall uncertainty (OF, PSI) .175 to.575oF Range of Display PROGRAMMABLE Qualifications (seismic, environmental, IEEE279) N/A O Calculator HONEYWELL Type (process computer, dedicated digital or analog calc.) HS4400 PROCESS COMPUTER If process computer is used, specify availability. (% of time) 99.26% (1979 Average) Single or redundant calculators SINGLE HIGHEST VALID TEMPERATURE Selection Logic (highest T., lowest press) LOWEST VALID PRESSURE Qualifications (seismic, environmental, IEEE279) N/A Calculational Technique (Steam Tables, Functional STEAM TALLES (1967 ASME) Fit, ranges) j Input Temperature (RTD's or T/C's) T/C & RTD Approx. 60 IN-CORE T/C; Temperature (number of sensors and locations) Two wide range RTD's per loop T/C: 0-2300 F 0 Range of temperature sensors RTD: 0-700 F .Ol ) V II-11 07/18/80 1

<2.0*V T/C Uncertainty

  • of temperature sensors (OF at lo)

<1.5 F RTD j RTDs (seismie, environmental) (y Qualifications (seismic, environmental, IEEE279) T/cs (nora) RCS wide range press. Pressure (specify instrument used) RCS low range press. Pressure-(number of sensors and locations) 2-R_eactor Coolant System t.ow Range 0-800 PSIG Range of Pressure sensors Wide Range 0-3000 PSIG Uncertainty

  • of pressure sensors (PSI at Ic)

Il span ide range (seismic, environmental) Qualifications (seismic, environmental, IEEE279) Low range (none) i Backup Capability INCORE T/C-CONTROL ROOM METER WITH SELECTOR SW. HOT AND COLD LEG TEMP. (RTDs)-CONTROL ROOM RECORDER Availability of Temp & Press PRESSURE-CONTROL ROOM METER AND 1 CHANNEL RECORDED Copy available in Availability of Steam Tables etc. Control Room j Training of operators Yes Procedures Yes

  • Uncertainties must address conditions of forced flow and natural circulation, j

1 i 1 l V II-11A 07/18-80

,m ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION i N \\v]

References:

NUREG-0578 - 2.1.8b Action Plan - II.F.1 Noble Gas Monitors Vent monitors for noble gases will be provided with a range adequate to cover both normal and postulated accident conditions. The presently installed noble gas monitors at McGuire cover the range of 10-7 pCi/cc to id*J pCi/cc. A gross gamma detector will be added to these monitors to extend the range up to 10*5 pCi/cc. This detector will be attached to the outside of the unit vent and shielded to minimize count rate contribution from other possible sources. The detector will be sensitive to the 80 Kev energy range of noble gases and will have a minimum of one decade overlap with the existing noble gas monitor. If an event were to occur to cause the activity being released to be in the range of this additional detector, the noble gas monitor sample will be isolated. This action will prevent the noble gas monitor from becoming contaminated and rendering erroneous indications when activity starts decreasing. The additional detectors will be installed by January 1,1981. Procedures for estimating noble gas release rates if the existing instrumentation goes off scale will be written to cover the interim period between fuel loading and ['~'g installation of the new detectors. ( In addition, procedures for quantifying radioactive releases through all of the atmospheric steam release valves are being developed. Coupled with the modifications to the unit vent monitors these procedures will provide Duke with the capability to quantify the noble gas releases from McGuire Nuclear Station. Contairment High Range Radiation Monitors Two physically and electrically separated radiation monitors will be installed inside the McGuire containment. These monitors will be supplied by General Atomics and will feature GA detector model number RD23. Each monitor will utilize an ionizatiog chamber to measure gamma radiation and will cover the 0 to IT R/hr. No overlapping of ranges is required. Monitor range from 10 . sensitivity to 62 Kev is 9.8X10-12 Amps / Rad /hr and the sensivity to 52 Kev is 9.0X10-12 Amps / Rad /hr. Seismic qualification of the monitor is in accord-ance with IEEE344-1975 and environmental qualification is per IEEE323-1971. One monitor will be powered from the Train A vital instrument bus, and the other mor.itor will be powered from the Train B vital instrument bus. Analog meters (one per train) will continuously indicate monitor output in the control room. A continuous strip chart recorder (one train) will also be located in the control room. Calibration of these monitors will be performed at least every other refueling (s) outage according to procedures currently being finalized by General Atomics. N- / In no case will the calibration frequency exceed 36 months. The current schedule II-12 07/18/80

I f i / ' for implementing this requirement calls for equipment shipment on September 1, 1980 and installation, calibration, and functional testing to be complete by January 1,1981. The detectors'will be mounted on the primary shield wall at an elevation of at at 00 least7g0+2'(10feetor-moreabovethemaximumpost-LOCAwaterlevel and 180 in the lower containment. The following McGuire General Arrangement drawings show the plan and sectional views with the monitor locations drawn in. Containment Pressure Continuous indication of containment pressure will be provided in the control j room. Measurement and indication range will extend from -5 psig to 60 psig. Each of the redundant differential pressure transmitters will be located in the annulus where a filled capillary system will connect its associated trans-mitter with a bellows sensor located inside containment. Continuous indica-tion from each transmitter will be provided in the control room. In addition, one channel of containment pressure will be recorded. These instruments will be completely independent of the existing containment pressure transmitters. They.will be installed by January 1, 1981 contingent upon timely equipment delivery. ) Containment Water Level Two containment floor and equipment sumps are provided on the floor of the lower containment (El 725') to collect floor drains and equipment drains. However, these sumps and their associated pumps and instrumentation serve O. no safety function. The containment emergency recirculation sump at McGuire encompases the entire floor of the lower containment. The two ECCS recirculation lines take ' suction just inaide the Containment wall at elevation 725' and are oriented horizontally. They are not located in the bottom of a recess or sump in the floor. Redundant safety grade level instrumentation is provided to measure emergency recirculation sump level. The range of this instrumenta-tion is 0-20 feet (E1.725' to El 745') which is equivalent to a lower contain- . ment volume of approximately 1,000,000 gallons. The accuracy of this instrumen-tation is 10% over the full range. The redundant differential pressure transmitters utilized in this instrumentation have been relocated to the annulus where a filled capillary system will connect its associated transmitter with bellous sensors located inside containment. Continuous indication from each transmitter will be provided in the control room. In addition, one channel of containment water level will be recorded. Containment Hydrogen Monitoring . Continuous indication of hydrogen concentration in the containment atmosphere will be provided lLn the control room. This hydrogen monitoring system will consist of two redundant Comsip, Inc./Delphi Systems Division K-ill analyzer systems with a range of 0 to 30% hydrogen by volume. These analyzers operate independent of the recombiner system and will be powered from redundant Class lE' power supplys. Each analyzer will have its.cmm containment sample and return lines, and will be able to monitor either of two identical containment . sampling headers or,the calibration gases. Each analyzer will have a local control panel-indicator and alarm and a separate control room indicator and alarm. In addition, one channel of containment hydrogen concentration will be ' recorded. II-13 07/18/80

1 j Each containment sample header will have five inlet samples available for monitoring. 1. Top of containment 2. Operating level 3. Basement 4. Radiation Monitor /Recombiner Inlet header 5. Radiation Monitor /Recombiner Discharge header All sample selection and switching is accomplished manually by the operator from the local analyzer control panel. This instrumentation will be installed by January 1,1981 contingent upon timely equipment delivery. i I 1 r 'i t i 4 i 1 O II-13A 07/18/80 1 1 1

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PLANT SHIELDING p t. 'N

References:

NUREG-0578 - 2.1.6b Action Plan - II.B.2 General Design Criteria (GDC) 19 of Appendix A to 10CFR 50 and 10CFR 20 require control of radiation exposure to personnel associated with nuclear station opera-tions. In addition, GDC 4 of Appendix A to 10CFR 50 requires safety equipment and systems to function in the environmental conditions to which they either will or may be subjected during the station lifetime, A review of the McGuire Nuclear Station has been conducted to determine if any areas of the station fail to meet the above criteria. Personnel access criteria is as recommended in Harold Denton's October 30, 1979 letter te all operating nuclear power plants. These criteria are:

1) Less than 15 mR/hr for areas requiring continuous occupancy and
2) GDC 19 (5 rem whole bcdy or equivalent to any organ) for areas requiring infrequent access.

Equipment suitability criteria is by comparison of calculated environmental conditions with the equiptent design and/or qualification. The accident scenario selected to yield the greatest release of radioactivity from the Reactor Coolant System (RCS) is the Loss of Coolant Accident (LOCA) ['~')S with subsequent fuel damage. The basis for selecting this particular scenario as the Design Basis Accident (DBA) is discussed in TID-14844. The resulting ( _ airborne activity assumed to be released to the containment is 25% core inven-tory of iodines and 100% core inventory of noble gases. These values are con-sistent with Regulatory Guide 1.4 and TID-14844. Typically, the liquid activity has been assumed to be 50% core inventory of iodines and 1% core inventory of the j remaining fission products. These values are consistent with Regulatory Guide 1.7 and TID-14844. However, Harold Denton's October 30, 1979 letter recommended the inclusion of 100% core inventory of noble gases with the previous liquid activity. Our calculations show that less than 2% of the noble gas inventory will remain in solution post-LOCA. Although we consider the inclusion an unnec-essary conservatism, we have accepted the NRC Staff recommendation for our initial station review. As a result, the fission product distribution assumed for the initial McGuire Nuclear Station review is: Airbo rne: 100% core inventory of noble gases 15% core inventory of iodines (These activities are assumed to be homogeneously distributed throughout the containment free volume. Liquid: 100% core inventory of noble gases 50% core inventory of iodines 1% core inventory of remaining fission products (These activities are assumed to be homogeneously distributed throughout a water volume consisting of: RCS, Core Flood Tanks, rN water injected by the Safety Injection System, and water from the ( ) Ice Condenser melt.) %./ II-15 07/18/80

Plant systems or portions of systems which might contain significant levels of /x radioactivity as a result of a Design Basis Accident were selected for the ( ) station accessibility review. lacluded in the review were:

1) those portions of the Residual Heat Remova), Reactor Building Spray, Safety Injection, and Chemical and Volume Control Systems which could be aligned for recirculationof water from the containment sump to the Reactor Coolant
System, r
2) those portions of the Liquid Waste Recycele System which would collect and store leakage from the systems mentio ed in item no. 1, 3) those portions of the Nuclear Sampling System which would be used in determining radiation levels inside containment or those systems mentioned above, and 4) those portions of the Chemical and Volume Control System which supply seal water to the Reactor Coolant Pump seals.

The Waste Gas System and the entire Chemical and Volume Control System were not included in the accessibility review. The purpose served by these systems at TMI-2 was to remove and store noncondensible gases from the Reactor Coolant System. Duke Power Company believes that the design of McGuire Nuclear Station as augmented by post-TMI modifications does not require the use of these two systems in a post-DBA situation. Tn aid in identifying potential personnel access problems, the station was divided ('~'} in*o post-LOCA radiation zones. Included in the radiation zones were all areas %',j necessary for personnel access in controlling and mitigating a possible accident. Two types of area access have been designated:

1) continual occupancy and
2) infrequent occupancy.

Areas designated as requiring continual access are:

1) Control Room, 2) On-site Technical Support Center (TSC), 3) Onsite Operational Support Center (OSC),

and 4) Personnel Access Portal (PAP). The Control Room, TSC, and OSC are located in the Control Complex. The PAP is located in the Administration Building and serves cd the main personnel access point to the station proper. The Control Room serves as the initial onsite center of emergency control and is designed to evaluate, control, and respond to various accident conditions. A detailed descroption of Control Room design and functions is presented in Section 7.8 of j the McGuire FSAR. Areas requiring infrequent access are generally located in Auxiliary Building corridors. Two er.ceptions to this are:

1) Contcinment Hydrogen Recombiner controls and 2) Emergency Diesle Generators. Redundant hydrogen recombiners are located in the Upper Containment. Power control panels for these recombiners are located in the Electrical Penetration Rooms. Redundant diesel generators are located in t. section of'the Auxiliary Building designated as the Diesel Generator Area. Various control functions associated with the diesel generators and supporting systems are located in the Diesel Generator Area. Typical func-tions centered in Auxiliary Building corridors are:
1) station radioactive waste control panels, 2) motor control centers, and 3) instrumentation panels for various station systens.

I II-16 07/18/80

The major emphasis of the McGuire Nuclear Station plant shielding review was / to assure that station personnel would be able to carry cut their emergency (') procedures. The review featured the consistent use of 6ne defined source terms in conjunction with the KAP-6 computer code. Listed below sce all areas where radiation problems may exist and their present status. 1. Sample Room - Unit No. 1 New sampling panels have been designed to allow analysis of reactor coolant and containment atmosphere samples under accident conditions. However, samples from shared process systems are taken from the Unit I sample room. A large cable tray penetration is located in the wall separating the sample room from the Unit 1 penetration area. The locations of the cable tray opening and recirculation piping within the . penetration area allow significant radiation steaming into the sample room. Using the radiation sources discussed earlier, the sample room will be inaccessible for slightly more than one week. However, the following samples have been identified as needing to be analyzed before one week: 1. Recycle Holdup Tank 2. Waste Evaporator Holdup Tank 3. Waste Drain Tank 4. Baron Recycle Evaporator Condensate Demineralizer Outlet 5. Waste Evaporator Conuensate Demineralizer Outlet Due to the complexity of the geometries involved, shielding of the Unit 1 s s } sample room is not a viable solution. Modifications which would allow s,_/

nalysis of these samples in the Unit 2 sampling root would not be completed until mid 1982.

An alternate method for collecting these samples through tell-tale drain lines has been established. A review of tell-tale valve locations has shown that all valves are located in the Auxiliary Building corridors and are therefore accessible. Procedures for collecting these samples will be written, and no design modifications are necessary. 2. Floor Drain Tank Room The RHR sump pumps discharge to the Floor Drain Tank. Manual valves associated with isolating the Floor Drain Tank, and directing its contents elsewhere for storage or processing, are located in this room. As a result of the location and manual operation of these valves, per-sonnel exposures coule exceed GDC 19. Reach rods vill be added to the valves by fuel loading. Duke Power Company is currently conducting a through review of the environ-mental qualifications of Class 1E equipment at McGuire. This review includes possible radiation environments resulting from the source terms assumed in the plant shielding review. A I \\ II-16A 07/18/80

I f% CONTROL ROOM HABITABILITY

Reference:

Action Plan - III.D.3.4 Duke has conducted an evaluation of the McGuire Control Room to determine the level of protection provided against the toxic gas hazards described in the following documents. -1. Standard Review Plan - Sections 2.2.1, 2.2.2, 2.2.3, and 6.4. 2. Regulatory Guides 1.78 and 1.95. This evaluation revealed that the design of the McGuire Control Room both meets the criteria specified in the above documents and provides adequate protection of the Control Room from_ toxic gas hazards. A detailed description of this protection is provided in the applicable sections of the McGuire FSAR. A summary description is provided below. The only path of entry of a toxic gas to the Control Room is the outside air intakes. These intakes are.'.ocated adjacent to the Reactor Building opposite each unit vent. They are at El. 772+0 and.are separated by 360 ft. The only toxic material;normally' used at the station which might be drawn into an air intake of the Control-Room is chlorine. It is stored in standard 150 lb. cylinders at two locations on site. The shortest distance from a Control Room air intake to a chlorine storage area is 500 feet. Each air intake has a redundant chlorine detection and isolation system. A single component failure will_not impair the function of the system. The chlorine detectors are seismically qualified and have a five second response time to 5 PPM chlorine. The isolation valves conform to ASME code, section III, class 3. They have electric motor operators and a closure time of 8.4 seconds. During normal ~ operation 500 CFM is taken from'each intake to maintain the Control Room at 1/8 in, w.g. positive pressure. The outside air plus 1000 CFM of return air is circulated through carbon filters before delivery to the Control Room. .Upon chlorine detection the affected intake is automatically isolated and the full 1000 CFM of pressurization air is brought in throuft the second intake. In the unlikely event-that both intakes become contaminated tha Control Room will be isolated with 2000 CFM circulated through carbon filters. Additional protection includes manual isolation at the discretion of the operator and full face respirators with a six hour air supply for five men which can be replenished from outside the Control Room if necessary. This habitability. review of.the McGuire Control Room has revealed that no design modifications are necessary. II-17 07/18/80

COMPARISON OF CONTROL ROOM PROTECTION I AGAINST T0XIC GAS HAZARDS WITH REGULATORY GUIDE 1.78 i l i j Paragraph Compliance Status C-1 in Compliance l C-2 In Compilance i C-3 Chlorine Detectors Are Located in the Control Room Outside Air intakes C-4 in Compliance C-5-a in Compliance C-5-b in Compl'ance C-6 in Compliance j C-7 in Compliance l C-8-a .In Compliance C-8-b in Compliance C-9 in Compliance C-10 in Compliance C-11 In Compliance C-12 in Compliance C-13 in Compilance C in Compliance l C-15 in Compliance 1 i f I II-17A 07/18/80-l l l l l l

- -. _ =. COMPARISON OF CONTROL ROOM PROTECTION AGAINST T0XIC GAS HAZARDS WITH REGULATORY GUIDE 1.95 mO Paragraph Compliance Secjus_ C In Compliance C-2 In Compliance C-3-a-1 In Compliance j C-3-a-2 The McGuire Fresh Air Inlets Are 12 Feet Above Grade. They are redundant and separated by 360 ft. C-3-a-3 The equivalent and exchange rate at McGuire is 935 CFM or 0.47/hr. See section 6.4.3 of the FSAR. ll C-3-a-4 In Compliance C-3-b Not Applicable C-3-c Not Applicable C-3-d Not Applicable \\ C-3-e Not Applicable C-3-f Not Applicable C-4-a In Compliance C-4-b In Compliance C-4-c In Compliance C-4-d-1 In Compliance C-4-d-2 The McGuire System response time is less than 15 seconds. C-4-d-3 In Compliance C-4-d In Compliance i C-4-d-5 The detection system is designed to operate up to 120 F &nd 100% R. H. and is installed in an environment which does not exceed these conditions under normal plant operation. 3 C-4-d-6 In Compliance C-5 In Compliance C-6 In Compliance II-17B 07/18/80

IE BULLETINS ON !EASURES TO MITIGATE SMALL-BREAK eN LOCA'S AND LOSS OF FEEDWATER ACCIDENTS ( \\

Reference:

Action Plan - II.K.1 C.1.5 During the planning and procedure development stage of the integrated Engineered Safety Features (ESF) test a complete review of all valves receiving a safety injection actuation signal and containment isolation signal was conducted. This review primarily evaluated the response time requirements for each of these valves. However, in order to verify the response times, valve positioning requirements were also reviewed. As a result of this review and the successful performance of the integrated ESF test, disect verification of correct valve positioning requirements and valve positions under ESF conditions was obtained. Correct valve positioding requirements, valve positiono, and valve response times are verified during the ESF test via the Operator Aid Computer (OAC). The correctness of the OAC indication is verified through the use of operating procedures which require visual verification that the valve position indication in the control room and on the OAC is identical to the actual valve position. These operating procedures are required to be performed on all ESF valves after any maintenance activities'which could affect proper operationof the ( valve. U} C.l.10 Procedures for repositioning valves following maintenance or test activities have been reviewed and revised as necessary to provide assurance that these valves are returned to their correct position. These procedures require verification of the operability of a redundant system prior to the removal of any safety-related system from service, verification of the operability of all safety-related systems when they are returned to service, and noti-fication of the reactor operators whenever a safety-related system is removed from and returned to service. The operabil'ty of redundant systems and safety-related systems is verified by performing an initial functional test and subsequent periodic tests. A Removal and Restoration procedure governs the repositioning of valves follow-ing these tests and follouing any maintenance activities performed on these valves. This procedure utilizes a formal checklist to provide assurance that these systems are properly aligned. Notification of operators when safety-related systems are removed from, or returned to, service is accomplished by the use of Removal and Restoration checksheets, red tags and red tag logbook, white tags and white tag logbook, out of service stickers, and the 1.47 bypass panel. Log entries denoting the removal and restoration are made in the Reactor Operator's Log. All of f$ the above documents are reviewed during shift turnovers. ( i Q) II-18 07/18/80

t The McGuire Work Request Program governs all maintenance activities performed at McGuire. These work requests describe the maintenance to be performed and the procedures for performing it. Upon completion of the maintenance all work requests are entered into the plant computer. This program provides for a 4 historical record of all maintenance performed on safety-related systems. I u C.1.17 ] 1 The design of McGuire Nuclear Station does not feature safety injection initiation on coincident pressurizer level and pressure signals. Therefore, no modifications are necessary to assure that safety injection is initiated whenever the low pressurizer pressure trip setpoint is reached. i i f i l i i i i l 1 l I l I j 4 1 4 I \\ II-18A 07/18/80 i . - - -. ~

IN-PLANT RADIATION MONITORING O

References:

NUREG-0578 - 2.1.8c Action Plan - III.D.3.3 (Partial) Portable air samplers with silver zeolite radioiodine sampling cartridges are used at McGuire for sampling air when the presence of noble gases is suspected. McGuire Health Physics perscnnel are knowledgeable in the appropriate station procedures and are trained in the equipment required to determine airborne iodine concentrations in the plant under all conditions. Two independent counting rooms are available for performing detailed sample analysis. These counting rooms have been designed with shielding to reduce radiation levels to 0.02 mR/hr from plant sources during normal operation. Also included in the counting rooms are shielded GeLi detectors, and shielded sample storage areas. A procedure to determine airborne radioiodine concentrations will be established which does not rely on the availability of a counting room. This procedure will utilize portable " survey-type" instrumentation with energy discrimination for iodine to determine a "go" or "no go" iodine concentration for respiratory equipment use. The results of this analysis will be available within ten minutes. This instrumentation in conjunction with the portable air samplers is a fully adequate method to monitor iodine in-plant. \\ To reduce counting system saturation, sample sizes will be varied to minimite counting system problems. In addition, nitrogen purging of the counting room GeLi detector shields can be used to reduce airborne activity interferences. III-4 07/18/80 + y.

STATION DIRECTIVE 3.1.4 APPROVAL DATE or424n 1 <==n.A t t /in /?o REVISION 1 DATE DUKI POWER CCMPANY McGUIRE NUCIJAR STATION CONDUCT OF OPERATIONS OBJECTIVE The objective of this directive is to outline the responsibilities, authority, handling of special orders and rules of practice for licensed operators at Mc-Guire Nuclear Station. This directive also outlines control room authority, succession control, access procedure and on-call professional and supervisory personnel requirements. IMPLEMENTATION ) Duties of licensed operators and licensed senior operators described in this '- /' directive are derived from requirements in 10CFR55, ANSI N18.1-1971, McGuire FSAR Section 13, McGuire Technical Specifications Section 6. Regulatory Guide 1.114, NUREG-0578 and ANSI NIS.7-1976. RESPONSIBILITY AND AL'THORITY SUCESSION A. Station Manager j The Station Manager reports to the Manager, Nuclear Production and has direct responsibility for operating the station in a safe, reliable and efficient nanner. He is responsible for protection of the station staff and the gen-eral public from radiation exposure and/or any other consequences of an accident at the station. He bears the responsibility for compliance with the facility operating license, l B. Operating Superintendent i 1 The Operating Superintendent has the responsibility for directing the actual day-to-day operation of the station. In the event of the absence of the Station Manager, the Operating Superintendent, if so designated, assumes the responsibilities and authority of the Station Manager. C. Operating Engineer The Operating Engineer assists the Operating Superintendent in directing j

_2 p I station operation and may assume complete responsibility for the actual day-(V co-day operation of the station in the absence of the Operating Superintendent. He also serves as "On Call" Duty Engineer for Operations. D. Assistant Operating Engineer The Assistant Operating Engineer assists the Operating Engineer in directing station operation and may assume complete responsibility for the Operating Engineer in his absence. He also serves as "On Call" Duty Engineer. "On Call" Duty Engineer An Operations "On Call" Duty Engineer is responsible for directing actual shift and station activities. He will act as a liasion between Operations and other groups at the station and is available for coordinating and assisting the Shift Supervisor during accident or e=ergency conditions. He shall hold a current Senior Reactor Operators License. E. Shife Supervisor A Shift Supervisor is responsible for the actual operation of the station on his assigned shift. He has the responsibility to be cognizant of all opera-tional conditions affecting the safety of the plant as a matter of highest /'~'N priority when on duty. x_,/ A Shift Supervisor directs the activities of the operators on his assigned shif t and must be cognizant of all =aintenance activities affecting plant operation being perfor=ed while he is on duty. TheShiftSupervisorondutyhasboththeauthorityandtheob11gac5cnto shutdown a unit if, in his opinion, conditions warrant this action. During accident or emergency conditions, the Shif t Supervisor is responsible for directing activities of control room personnel. He shall remain in the t?n-trol room at all c1=es during these situations and should not become total!.y involved in any single operation when =ultiple operations are required during emergencies. The Shift Supervisor can only '2e relieved by another Shift Supervisor or by a member of sanagement that holds a current Senior Reactov Operators license. During routine operations, the s.'.ft Supervisor should remain in the Control l Room at all times. However, he say delegate the " Operator at the Controls" the authority of control roon operations during te=porary absences. F. Assistant Shif t Supervisor An Assistant Shift Supervisor assists the Shift Supervisor in operation of ~ \\v/ the station on his assigned shift. The Assistant Shift Supervisor on duty has both the authority and the obligation to shut down a unit if, in his 9 9

0 opinion. conditions warrant this action. The Assistant Shift Supervisor assumes the responsibilities r authority of the Shif t Supervisor, as de-fined for the Shift Supervis-, if so designated. G. Nuclear Control Operator A licensed Nuclear Concrr-Operator is responsible for the actual operation of a unit on his assigued shift. The licensed Nuclear Control Operaccr has authority and obligation to shut down a unit if, in his opinion, conditions warrant this action. He may piso serve as " Operator at the Controls" per Station Directive 3.1.17, if so designated. H. Assistant Nuclear Control Operator A licensed Assistant Nuclear Control Operator is responsible for the actual operation of a unit on his assigned shift as directed by the licensed Nuclear Control Operator. He has both the authority and obligt. tion to shut down a unit if, in his opinion, conditions warrant this action. He may also serve as " Operator at the Controls", if so designated, per Station Directive 3.1.17. LICENSED OPERATOR DUTIES

1. A licensed Nuclear Control Operator or Assistant Nuclear Control Operator shall

.act as " operator at the controls" in accordance with Station Directive 3.1.17 3 except in situations where a licensed senior operator elects to assume these duties. 2. The " operator at the controls" shall continue in his duties until relieved 4 by a qualified operator in accoreoEce with Station Directive 3.1.19. 3. " Operator at the controls" duties shall not be delegated to non-licensed persons. However, a non-licensed person may manipulate the controls of the facility under the direction and in the presence of a licensed operator or licensed senior operator as a part of his training to qualify for an operator license under 10CFR55. 1 The " operator at the controls" shall monitor instrumentation displays and alarms to assure safe operating conditions for his assigned unit. NOTE: All station personnel performing functions which may affect unit operation or control room indications are reauired to notify the " operator at the controls" prior to initiating such action in accordance with McGuire Station Directive 3.1.5. 5. The " operator at the controls" shall know and comply with limits and set- [O \\ points associated with safety-related equipment and systems as specified in Technical Specifications and designated in the Operating Procedures. m t- -e e---- i~-g

r

.m--.------ - - - -

. m' / He shall review routine operating data in order to assure safe operation tN of his assigned unit. All licensed reactor operators shall participate in the requalification program outlined J McGuire FSAR Section 13.2.2. 6. The " operator at the controls" shall be re, onsible for the manipulation of controls which directly or indirectly a2 rect core activity. He shall operate or direct others to operate other equipment associated with his assigned unit. He shall direct che activities of persons assigned under him in the operation of this assigned unit. These operations pertain to startup opercion, power operation, shutdown operation, or during a shut-down condition. 7. The " operator at the controls" shall acknowledge all alar =s. He shall notify licensed senior operators on duty of unexpected alar =s or alarms of unknown cause. He shall initiate prompt corrective action on receipt of any indication (instrument movement or alarm) of an irregular operating condition in accordance with applicable Operating, Alarm, or Emergency Procedures. This corrective action may include tripping the reactor should, in his best judgement, a situation exist requiring prompt action and the licensed senior operators on duty are not available for consultation. The " operator at the controls" shall verify that appropriate automatic l ~ action has taken place in case of an alarm. When automatic action is deter =ined to have failed or to have been ineffective, he shall manually initiata this action. 8. The " operator at the controls" shall notify the licensed senior operators on duty of all abnormal operating conditions. The '.perator at the controls" j shall notify the area dispatcher of conditions which could significantly affect station load and other informacien which the area dispatcher may request. 1 9. The " operator at the controls" may stop fuel handling operations if, in l his best judgement, control room indication or communications show warrant-ing conditions.

10. When two or more licensed operators are assigned to the same unit, one shall act as " operator at the controls" while the others assist in all the above duties.
11. The " operator at the controls" shall maintain the Reactor Operator's Log-book in accordance with Station Direceive 3.1.1.

[ \\

12. The " operator at the controls" can authori:e the removal of any instru=ent or

( ) components from service in accordance with Station Direceive 3.1.5 or 3.1.19.

~ l 1 5-l. 4 LICENSED SENIOR OPERATOR DUTIES 1. Shift Supervisors and Assistant Shif t Supervisors responsible for the units shall be licensed senior operators and shall only be relieved by management personnel who hold Senior Reactor Operators license. l 2. The Shift Supervisor on duty shall oversee station operations. The Assist-anc Shift Supervisor on duty shall oversee operations associated with his assigned unit. Persons in these positions while on duty fullfill the on-site requirement s for licensed senior operators. 3. All licensed senior operators within their area of authority and whila en duty shall keep themselves informed of operating status and shall direct the licensed activities of licensed operators. All licensed senior oper-acors shall within their areas of authority and while on duty organi:e, direct, and control activities to insure safe, efficient operation in compliance with administrative and technical requirements for operation. When required a licensed senior operator may perform any duty of a licensed operator. 4. The Shif t or Assistant Shif t Supervisor can authorite the removal of any instrument or conponents from service in accordance with Station Direct-ive 3.1.5 or 3.1.19. 5. The Shift or Assistant Shift Supervisor shall naintain the Unit Supervisor's Logbook in accordance with Station Directive 3.1.13. 6. The Shift Supervisor will have administrative assistance on shift to assist in cimo tickets, scheduling days off and other associated duties. RULES OF PRACTICE Work schedules, hours at " Duty Stations" and rules of practice are dicected by station management. A normal tour of duty is (3) eight hours at the assigned " Duty Station". The Shift Supervisor may extand these hours as necessary to maintain proper manning requirements and station output. The discretion of t.- Shift Supervisor shall ensure that no tour of duty shall axtend to a period such that duty station personnel's judgement and availability to properly perform their function are impaired. U ON CALL OF PROFESSIONAL AND SUPERVISCRY PERSCNNEL. Eachgroupsuperintendentshallensurech.at the Operations Shift Supervisor is i -e e.-~n e ,---.~e-- -- ~ - - - - - - w -w,.gy

~ . ( supplied with an updated "on call" list for professional and supervisory per-s sonnel in their respective group. This "on call" list shall include name, tele-phone numbers and beeper numbers of these "on call" individuals. CONTROL ROOM ACCESS Access to the area deemed " Surveillance Area", per Station Directive 3.1.17 (Definition of "At the Controls") shall not be permitted by non-licensed per-sonnel, while fuel is in either reactor, without paraission having been given to enter such area by the on-duty licensed " operator at the controls" or the senior licensed unit supervisor. SPECIAL ORDERS Management written instructions, other than procedures, Station Directives, etc., which encompass special operations, housekeeping, personnel actions or other similar instructions shall be kept in an accessable document for Operaticus per- [' sonnel. This document shall be reviewed on a semi-annual basis for applicability of in-structions. Updates to the document shall be placed in this document, after review by appli-cable personnel, for future reference. Any cancellation of a written order shall be removed from this document and com-sunciated to applicable personnel. USAGE AND TESTING OF EMERGENCY NOTIFICATION SYSTEM (NRC PHONE) Any event listed in Attachment 1 shall be reported as soon as time permits and in all cases within one hour to the NRC Operation Center via the Emergency No-tification System (ENS) - NRC Phone. In the event that the (ENS) is not oper-able, the report should be made via commercial telephone lines. The NRC Oper-acion Center telephone number is (301)-492-8111. The event shall bc reported to the NRC stating that the report is being =ade pursuant to 10CFR50.72(a). This notification should be sade by the Operations m - Shif t-Supervisor or the Operations Unit Supervisor. Documentation of the no-tification should be made in the Unit Supervisur's log. O n..

. Any event reported pursuant to this directive should also be reported to cua Licensing Engineer, the Projects and Licensing Group during normal work hours or to the Projects and Licensing " duty person" if the report is made outside of regular working hours. The Projects and Licensing Group is responsible for notifying the NRC resident inspector as soon as practical. NRC ENS PHCNE TESTING Daily, between the hours of 0400 and 0800 hours the NRC will call the station to obtain the status of each unit. This call will also be used as a check of the dedicated phone line. This call should be answered by either the Operations Shif t Supervisor or an Operations Unit Supervisor. A brief operating status of each unit should be provided. In addition, the NRC =ay request that we hang up and return the call to verify operability with the station as the initiator. ~'N Occasionally, a check of each extension at the facility will also be made. This check will be scheduled in advance by the NRC and should consist of verifying s_s communication from each cxtension at the station. If the ENS is found to be inoperable, notify the Duty Officier at the NRC Oper-acions Center at the telephone number given in Section " Usage and Testing of Emergency Notification System (ENS Phone)". U O

Page 1 of 2 \\ Q STATION DIRECTIVE 3.1.4 ATTACIDfENT 1 Events Requiring Immediate NRC Notification Via ENS Phone

  • l.

Any event requiring initiation of the station emergency plan or any section of that plan.

  • 2.

The exceeding of any Technical Specifications safety limit.

  • 3.

Any situation whereby a reactor is not in a controlled or expected condition of operation. A situation such as this is defined as any unscheduled event involving a reactor which cannot be stabilized by use of normal operating procedures or the followup actions of existing Emergency Procedures. I

  • 4.

Any act that threatena che safety of the station, or station personnel, or the security of special nuclear material, including instances of sabotage, and attempted sabotage. 5. Any event requiring initiation of shutdown of a Unit in accordance with Technical Specification limiting conditions for operations. 6. Any event involving personnel error or procedural inadequancy which, during normal operations, anticipated operational occurrences, or accident conditions p) prevents or could prevent, by itself, the' fulfillment of the safety function lO of those structures, systems, and components important to safety that are needed to (a) shutdown the reactor safely and maintain it in a safe shutdown condition, (b) re=ove residual heat following reactor shutdown, and (c) 11mic 1 the release of radioactive material to acceptable levels or reduce the potential for such release. 7. Any event resulting in manual or automatic actuation of engineered safety features, including the Reactor Protective System. 8. Any accidental, unplanned, or uncontrolled radioactive release. (Normal i or expected releases from =aintenance or other operation activities are j not included). 9. Any fatality or serious injury occurring at the site and requiring transport to an off'. site medical facility for trcatment.

10. Any serious personnel radioactive contaminacien requiring extensive onsite decontamination or outside assistance.
11. Any event involving licensed material which may have caused or threatens to cause:

4 (1) Exposure of the whole body of an individual to 5 re=s or = ore of radi-i acion; exposure of the skin of the whole body of an individual to 30 rems or more of radiation, or exposure of the feet, ankles,' hands, or forearms to 75 rams or more of radiation; or e ' e n e- ~

Pass 2 of 2 STATION DIRECTIVE 3.1.4 ATTAC11 MENT 1 (2) The release of radioactive material in concentrations which if averaged over a period of 24 hours, would exceed 500 ti:nes the limits specified for such :nacerials in 10CTR20, Appendix 3, Table II; or (3) The loss of one day or more of the operation of any facilities affected; or (4) Damage to property in excess of $2,000. 12. Strikes of operating employees or security guards, or honoring of picket lines by these employees.

  • NOTE: When an event is reported as a result of items (1), (2), (3), or (4),

establish and ::uintain an open, continuous communication channel with the NRC Operation Center and close the channel only when notified to do so by tha NRC. 4 U i i l i l M -m ,,m.

/ +nats\\ UNITED STATES C. c_ - y /, O / f ',, ~,., g NUCLEAR REGULATORY COMiv11SSION

  • , j j

WASHING TON. D. C. 20555 a , 4.'.u y 1 +! >Q,) WR 2 ~ ?;30 'ch, y m b i%Q LI :CWER C0 E L.! m.c.m : n ved r,w ALL ElR PIACIDR APPLICAMS AND LICEiSEES Centleren: SUBJECI': QJALIFICATICNS OF REACIDR CFERA'IEPS In a letter dated Septe er 13, 1979, we infor:ed you of NRR requirennnts established as of that date based on cur review of the 'IMI-2 accident. to the letter outlined the staff reccmuendations cencerning irprovements in the area of operator training for your information. Since that eine, the Ccmnission has acted on the staff recocz:endations. It is the purpose of this letter to set forth the revised criteria to be used by the staff in evaluating reactor operator training and licensing that can be ig lemenced under the current regulations and to establish an effective date for their incle::entation. Other criteria that will be established require addition.il staff work are also addressed.

Fcaever, oIV) irolenuntatica dates cannot be provided at this tire.

Cecmissica review in' the area of cperator training and qualificaticn is continuing and can be expected to result in additional criteria. Finally, requirerents will be established thrcugh rule making proceedings. 'J details the revised criteria and the effective date fer their implementatien. Your attention is specifically directed to Sections A, B and C of Enclosure 1 since these call out new criteria that will be inplenented in the near future; therefore, your plans regarding trairdng and licensing activities should be prceptly revised to conform to these criteria. Enclosures 2 and 3 provide guidance for establishing training progra::s in heat transfer, flutd ficw and then:odynamics; and mitigating ccre da age. enclosure 4 deta:.ls control manipulaticns for requalification programs. Based on our understanding of the industry's reasons for establishing the Institute of Nuclear Pcwur Operations and cur review of the latest revisions to applicable ANSI standards,.;e believe ycu share cur desire to significantly upgrade the requirements for operations perscrael. I .mm. :W**irG%. iWT.W-lS. ;5?5 n,. 7 /^'N DUPLICATE DOCUMENT ( ) 4 Entire document previously entered into system under: l ANO N o. of pages: ..}}

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