ML19331B443
| ML19331B443 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/06/1980 |
| From: | Parker W DUKE POWER CO. |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 MNS-TMI-80-03, MNS-TMI-80-3, NUDOCS 8008120191 | |
| Download: ML19331B443 (34) | |
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DUKE POWER COMP NY Powen Du Loixo 422 Sours Cuiuncu Sinter, Cauntorre. N. C. 2824a wiwam o. emanca.sa.
August 6, 1980 Wce Pattactut TELCP=0NC; Aega 704
$ttana PacDwCtiom 373-4093 Mr. Harold R. Denton, Director MNS-TMI/80-03 Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 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 " Duke Power Company, McGuire Nuclear Station, Response to TMI Concerns." This document was transmitted to the NRC via my letter of May 23, 1980 and updated via my letter of July 18, 1980.
Fuel loading for McGuire Unit 1 is currently scheduled for September 1980.
Please schedule your review of this document accordingly.
Ve truly yours,
/O s.At - Ch. cu i
William O. Parker, Jr.
THH:scs Enclosures (40) l
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Mr. Harold R. Denton, Director August 6, 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.
.W N
jO L* 2 J (b dckt
/ William O. Parker, Jr., Vi resident Subscribed and sworn to before me this 6th day of August, 1980.
Notary Public My Commission Expires:
l September 20, 1984 l
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DUKE POWER COMPANY MCGUIRE NUCLEAR STATION RESPONSE TO TMI CONCERNS August 6, 1980 Changes and Corrections Remove These Pages:
Insert These Pages Table of Contents, Page 1 Table of Contents, pg. 1 08/06/80 Table of Contents, Page 2, 07/18/80 Table of Contents, pg. 11 08/06/80 I-1 I-1 08/06/80 I-2 I-2 08/06/80 I-3 I-3 08/06/80 I-3A 08/06/80 1-3B 08/06/80 I-3C 08/06/80 1-7 I-7 08/06/80 I-7A 08/06/80 I-9 I-9 08/06/80 I-15 I-15 08/06/80 I-18 08/06/80 II-1 11-1 08/06/80 O
II-12, 07/18/80 II-12 08/06/80 11-13, 07/18/80 II-13, 07/18/80 Carryover II-13A 07/18/80 II-13A, 07/18/80 Carryover II-19 11-19 08/06/E0 III-04, 07/18/80 III-4 08/06/80 Appendix A Station Directive 3.1.31 Station Directive 3.1.32 Appendix D (tab)
June 4, 1980 letter of B. J. Youngblood June 30, 1980 letter of B. J. Youngblood July 2, 1980 letter of B. J. Youngblood O
TABLE OF CONTENTS Page Section I.
Operational Safety Shift Technical Advisor 1-1 Shift Supervisor Duties and Responsibilities I-2 Safety Engineering Group I-3 Shift Manning I-4 Qualifications of McGuire Nuclear Station Personnel I-5 Revised Scope and Criteria for Licensing Examinations 1-6 Upgrading Operator Training and Qualifications I-7 Administration of Training Programs for Licensed Operators I-8 Training for Mitigating Core Damage I-9 Training During Low Power Testing I-10 Control Room Access I-il Shift Relief and Turnover Procedures I-12 NSSS Vendor Review of Procedures I-13 Pilot Monitoring of Selected Emergency Procedures I-14 Accident Analysis and Procedure Revision I-15 Primary Coolant Sources Outside Containment 1-16 p-~g Control Room Design I-17
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General Office Training I-18
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Section II.
Design Relief and Safety Valve Position Indication 11-1 Relief and Safety Valve Testing II-2 Auxiliary Feedwater Initiation and Indication II-3 Auxiliary Feedwater System Reliability Evaluation II-4 Containment Isolation Provisions II-5 Emergency Power for Pressurizer Equipment II-6 Reactor Coolant System Vents II-8 Inadequate Core Cooling Instruments II-9 Additional Accident Monitoring Instrumentation II-12 Post-Accident Sampling II-14 Plant Shielding II-15 Control Room Habitability II-17 IE Bulletins on Measures to Mitigate Small-Break LOCA's and Loss of Feedwater Accidents II-18 Final Recommendations of the Bulletins and Orders Task Force II-19 Section III.
Emergency Preparedness and Radiation Effects Upgraded Emergency Preparedness III-l On-Site Technical Support Center III-2 Oa-Site Operational Support Center III-3 fs In-Plant Radiation Monitoring III-4 1
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4 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, Shift Turnover Verification Station Directive 3.1.31, Duties, Responsibilities and Qualifications of the Shift Technical Advisor Station Directive 3.1.32, Station Safety Engineering Group Appendix B Control Room Design-Preliminary Report i
Appendix C NRC Requirements for McGuire Nuclear Station September 27, 1979 letter from D. B. Vassallo to all Pending Operating License Applicants November 9, 1979 letter from D. B. Vassallo to all
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Pending Operating License Applicar.ts SECY-80-230, Enclosure 1 Appendix D NRC Requests for Additional Information June 4, 1980 letter from B. J. Youngblood to W. O. Parker June 30, 1980 letter from B. J. Youngblood to W. O. Parker July 2, 1980 letter from B. J. Youngblood to W. O. Parker a
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References:
NUREG-0578 - 2.2.lb Action Plan - 1.A.1.1 A technical advisor to the shift supervisor will be present on all shifts and available to the control room within ten minutes. The shift technical advisor's primary duty will be to provide evaluation and assessment of both normal and unanticipated transients. The shift technical advisor will be detached from and independent of the normal line responsibility for plant operation.
The shift technical advisors will be selected from the group of licensed senior reactor operators at McGuire. All of the McGuire SR0 applicants have received additional simulator and academic training. The simulator training included functioning as the STA during various transients,and the academic training included instruction in heat transfer, fluid flow, thermodynamics, and plant transients.
Duke Power Company is actively pursuing the development of a program to provide supplementary technical education for both the STA's and the shift supervisors.
This program will be taught at the college level and will be equivalent to sixty semester hours in both basic engineering and plant applications of engineering principles.
1 The duties, responsibilities, and qualifications of the shift technical advisor have been defined in the McGuire administrative procedure, Station Directive
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3.1.31.
This procedure is provided in Appendix A.
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SHUT SUPERVISOR DUTIES AND RESPONSIBILITIES
References:
NUREG-0578 - 2.2.la Action Plan - I.A.l.2 and I.C.3 The Manager of Nuclear Production has issued a corporate management direc-tive that clearly establishes the command duties of the shift supervisor and emphasizes the shift supervisor's primary management responsibility for safe-operation of the plant. This directive will be reissued annually.
The shift supervisor has been provided with administrative assistance to relieve him from administrative duties which detract from or are subordinate to his management responsibility for safe operation of the plant. The duties, responsibilities, and authority of the shift supervisor and control room operators have been defined in the McGuire administrative procedure, Station Directive 3.1.4, Conduct of Operations. This procedure is provided in Appendix A.
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O OPERATING EXPERIENCE EVALUATION PROGRAM V
Reference:
Action Plan - I.B.1.2 and I.C.5 SCOPE The purpose of the Operating Experience Evaluation Program (OEEP) is to provide a formal mechanism for the systematic evaluation of off-normal events occurring at Duke Power Company nuclear units, as well as at other facilities. This evaluation serves to confirm that plant response was as expected for anticipated transients, and assures that any unexpected behavior is investi;ated thoroughly.
The results of this evaluation are then used to identify procedural and/or design changes which may mitigate or preclude the recurrence of a similar event.
The evaluation program takes on added importance in assessing any unanticipated transients to assure that they are well understood and that approp*1 ate correc-tive measures are taken. In order to assure that the program is truly effective, information gained from Duke Power Company experience will be disseminated to other organizations, as appropriate.
PROGRAM DESCRIPTION ORGANIZATION In order to achieve the objectives of the OEEP in an efficient manner, an off-
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site organization has been established, as well as an on-site organization at s m_
each of the nuclear stations.
The on-site organization includes the Station Manager, who is responsible for the operation and safety of the plant; the Shift Technical Advisor, who performs the accident assessment function; the Safety Review Engineer, who evaluates significant off-normal events; the Station Safety Engineering Group (SSEG),
which re~'ews the investigations of events and the adequacy of proposed correc-tive actions; and the supervisors of areas relevent to a significant event, who may perform the initial investigation and must implement corrective actions.
The off-site organization consists of principal engineering support groups, who interface with station personnel ard other organizations in investigating events and developing remedial actions; company management, who holds overall responsi-bility for nuclear plant safety; and the Nuclear Safety Review Board (NSRB) which performs an independent review function.
EVALUATION OF DUKE POWER SYSTEM EXPERIENCE A functional flow chart of the OEEP for events which occur at Duke Power Company plants is provided as Figure 1.
The occurrence of any event is brought to the attention of the Projects and Licensing Engineer who determines whether the event is significant enough to warrant investigation.
If investigation is warranted, he notifies the Station Manager, notifies the General Office Project Coordination and Licensing (PC&L) Section, assigns an engineer to investigate the event, and noti-(
fies the NRC if appropriate.
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The event investigator prepares a report describing the cause of the event and
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any relevant plant behavior, and outlining proposed corrective actions. The SSEG then reviews each report for accuracy and completeness, and assesses the j
adequacy of proposed corrective actions. The SSEG submits its report to the j
Station Manager and the NSRB for review and for approval of corrective actions, to the Shift Technical Advisor for review with regard to operating procedures, and to the Supervisor of Training for inclusion of relevant information in the training program. A description of the membership and duties of the SSEG is outlined in Station Directive 3.1.32 which is included in Appendix A.
Upon notification of an event, the General Office PC&L Section notifies company management, and may alert other General Office engineering and scientific support groups. The Station Manager forwards the approved event report to the General Office PC&L Section, where a Licensee Event Report (LER) is prepared and sub-mitted, if necessary.
Information is provided to other organizations, including NSAC, NRC, and the NSSS vendor. Detailed evaluations of plant transients are performed, and event occurrence data is maintained. As appropriate, other engineering support groups review the LER and station event reports for further recommendations on corrective actions, and may interface with appropriate equipment vendors. The NSRB performs an independent review of the event report, the LER, and the effectiveness of any follow-up actions.
INDUSTRY EXPERIENCE EVALUATION Figure 2 illustrates the flow path for information received concerning industry operating experience.
Significant events will be brought to the attention of (s_-}
Duke Power Company by NSAC, NSSS vendors, other utilities, or the NRC.
Informa-tion is distributed, as appropriate, to General Office engineering support groups for review and development of corrective actions and to the Training Services group for incorporation into the training program. The SSEG reviews the informa-tion for applicability to the specific station, and makes recommendations to the NSRB and the Station Manager in areas where action may be necessary. The Station Manager then developes and implements appropriate corrective actions with assist-ance from and review by the engineering support groups.
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FIGURE 1 INPLANT OPERATING EXPERIENCE EVALUATION (FUNCTIONAL FLOW CHART) d Event Omeronne l l Duae Power Conmeny Manmumment e
E it soto re hinmue Arus
_f Stamen Menegar j q hoises Caoremsten & Licareng Lianeng Enyneur o menee of cent for o lemory GO PC & L o me.ieur ownt neart spuncarum of o humore LER and mannut rt a NRC e
o Aames Irmumamer o Aseresecorreusse o Olmr*mme LER mme actier emost roomru
8"4'dd' Assummme o inertnes nem 'eEC NRC & NESE seemr ammut oWW o Deuermee unuses trio e=eit is sugese i
.l Jounens by FSAR anueves d Event inusungsmer, -
O Perfense deadme amenesen of tremanes o Osormenecasse o Evenumme se renaeus ariar : " "
o Anasvaseventmenenor and. _ _
o Oevesoo corremove estaans o nas,,emn svent ammurvunne amma sus perNurm traria anufyme o huese event report l Svanen Comutaan & Menemurum l Staten Seesty
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Ertumnmerg Gnnso assumen and..
prammmes sudsemesunt r o Reneur evest rumort for correstnam sut o Osamuuno urus spuumme sammas amamr compostunum of sunumane aseme ewaustiere o Premmes MP8105 dass o Renewthemammasy of correcthe acoors o nemovemn bhne rame duas on evanno arms __ _ _ _. _
o Deuermine rewwwum of event ror metumonin o insurfeue wie u venders courstor trening l Swame Ramens & Fuss Mengement,, -
0 Osamet any danessiums
,n,,,,,,,,,,,,,,,,,,,,,y e W emaonise omrquut and amre per9ermene
. afuey anssysis, or currergue o Osermine seus preones commerne scoore baner correcove scoors o Sueput o Prooms operatrig status resorts Nar improverunt to o imurfume wiev PASSS waruser on Station Manager med ese N oreempo Nuesomr Sefury Revuse Bempus o Interfume wie G,3 for
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1008'48D*"I88"DWWF and informsoon needs o Dourmee arie grecom comete easter correstne accons o tneurtens wie i&C mut capuntaur seneurs I
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l Ramonab6e Sucervoor,
0 Revise eveva omrmsung to l
o Imtwemert corensove resimoon proucoon accare o Osunmns arus proome poumow tweter corruerne actsore l Desen Enonmarweg Denartment [-
I Shift Tecnmcas Aawieur o Aammt we tne _ _.a, of
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o Rewww event reocrt and SSEG._._
NN l Ocuratense Rowaw Soed ! -
o Educate orMrdt pesonnes o Renew ooersoonas answrences on gw procedurus and aru$gmerets W iftformatson 3rwautore 8er moornersten e tne oemapi of
- nams unoer constructen i lucennsor et Tracer.g g
o 'icruce es.e=ent coerstwg 3 slewsw ooeranone occis**nces I
esceneace'e+ormanon arms ncorporate reievant.nformsuon l
n we *ra.n.eq orneram nto t*e tre none crogram Nucreer $arety aemew scaro ; = -- - J ONSi?! 8UNC*:CNS
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FIGURE 2 Industry Operating Experience Evaluation (Functional Flow Chart)
Industry Operating Experience Data
't NSAC e Screen data for significant events e Issue significant event report Other Utilities Vendor NRC 1r 1r M SSEG h GO
- GO Engineering N--
_PC&L
, Support Groups e Review for applicability e Distribute informa-e Review for tion to SSEG and information e Document review cognizant groups e If applicable, determine o Inform NSAC and NRC e Assist Station areas where action is of any action taken Manager to necessary and submit develop necessary recommendation to Station corrective actions Manager and Nuclear Safety Review Board
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-> Station Ebnager e Approve and implement e Incorporate corrective actions relevant informa-tion into the training program I-3C 08/06/80 r.
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UPGRADING OPERATOR TRAINING AND QUALIFICATIONS v
Reference:
Action Plan - I.A.2.1 The Vice President of Steam Production is responsible for assuring that the operators at Duke nuclear stations are qualified and capable of safe operation of these plants. This determination is made in conjunction with station manage-ment and the nuclear training director.
Subsequent to May 1, 1980 any applicant for a McGuire SRO license will have four years of responsible power plant experience of which no more than two years will be academic or related technical training. Two years will be nuclear power plant experience of which at least six months will be at McGuire Nuclear Station.
The McGuire operator training program is described in Section 13.2 of the McGuire FSAR. This program has been revised and expanded to reflect the lessons learned from the accident at TMI Unit 2.
These revisions include the following:
1.
Additional instruction in thermodynamics and heat transfer has been included in the program. This additional instruction provides greater depth to the coverage of natural circulation, heat exchange processes and the use of steam tables in problem solving.
2.
Instruction in introductory calculus has been expanded.
3.
Instruction in the methods of hydrogen generation and hydrogen flammability and explosive limits has been added to the program.
4.
A demonstration of a natural circulation experiment has been added to the research reactor training program.
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The presentation on nuclear instrumentation has been expanded to include discussion of the effects of voiding in the core on excore indication.
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Increased emphasis on thermodynamics and fluid mechanics has l
been included in the presentations on cavitation, Mollier l
diagrams and their use, analysis of throttling processes, and l
natural circulation.
7.
The emphasis on small break LOCAs (including the TMI scenario),
and abnormal and emergency plant evaluations has been increased.
8.
Demonstration of aTMI type accident on the McGuire simulator has been added to the program.
9.
One additional week of simulator training for RO applicants has been included in the program to emphasize abnormal operations, s
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control of multiple malfunctions, and transients without reactor v
trips.
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program to train SRO applicants-to recognize transients in progress, to evaluate the transient, and to implement the appropriate corrective actions.
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TRAINING FOR MITIGATING CORE DAMAGE
Reference:
Action Plan - II.B.4 Duke has modified the McGuire training program in order to place increased emphasis on the operation and significance of any McGuire systems or instru-mentatien which could be used to monitor and control accidents in which the core may be severely damaged. This additional training identifies the vital instrumentation which supplies the operator with needed information in a degraded core situation. The training also identifies alternate methods of obtaining this information as well as specific instruction in the interpreta-tion of instrument readings in degraded core situations.
The McGuire operators currently enrolled in the operator training program will receive the training for mitigating core damage at the end of the program.
These operators will be required to apply their knowledge of plant operation to hypothetical degraded core situations. Various degraded core scenarios will be presented which will require the operators to diagnose plant status and restore the plant to a safe condition utilizing both primary and alternate instrumentation as their source of information.
Currently licensed McGuire operators will receive this training in a special supplemental training class.
Each operating shift will have received this
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training prior to power escalation of Unit 1.
v The training for these operators will be similar to that described above. It will, however, include a review of their prior training with increased emphasis placed upon degraded core situations.
Training for mitigating core damage will be incorporated into the operator training program for all future McGuire license applicants.
In addition, selected Duke training instructors will participate in a degraded core training seminar spon-sored by General Physics Corporation in early October.
The existing body of knowledge regarding nuclear plant response under degraded core conditions is being enlarged. Duke is participating in this effort in conjunction with other utilities, INPO, and the NSSS vendors. The cleanup effort at TMI Unit 2 should provide significant information in this regard. As this i
information becomes available Duke will modify its training program accordingly.
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- O ACCIDENT ANALYSIS AND PROCEDURE REVISION
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References:
NUREG-0578 - 2.1.9 Action Plan - I.C.1 Duke is in the process of developing new procedures and training guidelines for controlling and mitigating small break LOCAs, incidents of inadequate core coolin3, and certain anticipated transients. Duke's effort is in con-junction with analysis and research being performed by Westinghouse.
The Westinghouse analysis of small break LOCAs in upper head injection plants, WCAP 9600 and WCAP 9639, has been submitted to tne NRC for their review. Duke has reviewed these reports and made the necessary modifications te the McGuire emergency procedures and training program.
The Westinghouse analysis of inadequate core cooling, WCAP 9753, WCAP 9754, and WCAP 9744, has been submitted to the NRC fcc their review. These reports pro-vide an analytical basis for subsequent Westinghouse development of guidelines for the detection of and recovery from inadequate core cooling. Duke will assure that the McGuire emergency procedures and training program are consistent with these forthcoming Westinghouse guidelines.
The Westinghouse analysis of selected transients and accidents is continuing
/h on a deliberate schedule. Duke is closely following the development of this t,j analysis and will modify the McGuire emergency procedures and training program as appropriate.
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GENERAL OFFICE TRAINING s
Reference:
Action Plan I.B.1.2 On June 27, 1980 the NRC concluded a management audit of Duke's readiness to operate McGuire Nuclear Station. One item discussed was general office support of the station in the event of an accident at McGuire. Duke has identified those individuals responsible for providing general office support to the sta-tion.
The NRC review team cited the need for these individuals to be cognizant of the existing design and operating policies of the station.
Duke is currently developing a training program to provide these individuals with up-to-date information regarding the design and operation of McGuire. This program will include classroom instruction and simulator demonstrations. Proper accident response will be emphasized. A periodic refresher course will be included as part of the program.
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l RELIEF AND SAFETY VALVE POSITION INDICATION
References:
NUREG-0578 - 2.1.3a Action Plan - II.D.3 PORV The position of the pressurizer power-operated relief valves is detected by seismically and environmentally qualified stem-mounted limit switches. The limit switches actuate indicator lights on the main control board. The entire circuit including power supply is safety-related. Additionally, a control room computer alarm is activated upon the opening of a PORV.
Safety Valve Flow through the safety valves is detected by an acoustic flow detection system. This system senses vibrations caused by flow through the valve which is an indication that the valve is not fully closed.
Two accelerometers have been strapped to the discharge piping of each safety valve. One of these is an installed apare and is wired to the electronics cabinet but not monitored. A charge converter processes the accelerometer output and provides the voltage input to the monitor. The RMS of this signal
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is related to the flow through the valve. This signal is filtered and ampli-
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fied and is available on a front panel BNC connector. An RMS to DC converter provides on output to drive a bar graph on the front panel. The bar graph is a set of ten vertically arranged indicator lights which are labeled to give valve position as a fraction of full open. The charge converter is located in containment and the electronics cabinet is in the electrical penetration room.
The alarm output of the monitor is used to provide indication and alarm when flow exists through any of the three safety valves. A safety grade indicator light and a non-safety annunciator are provided. The bar graphs on the monitor can be used to determine which valve is open.
The system, with the exception of the annunciator alarms, is safety-grade, meets the appropriate seismic and environmental qualification requirements, and will be installed prior to fuel loading.
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i ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION
References:
NUREG-0578 - 2.1.8b Action Plan - II.F.1 j
Noble Gas Monitors Vent monitors for noble gases will be provided with a range adequate to cover gas monitors at McGuire cover the range of 10-7 pC1/ccto10*yinstallednoble The presentl both normal and postulated accident conditions.
pCi/cc. A gross
&5 gamma detector will be added to these monitors to extend the range up to 10 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 n,,
scale will be written to cover the interim period between fuel loading and installation of the new detectors. These procedures will require the use of s_
portable high range survey instruments to measure the radiation levels on lines going to the radiation monitora for the unit vents if the radiation levels are such that personnel exposure could exceed 3 rem /qtr whole body and 18 3/4 rem /qtr to the extremities in the collection of a sample. The contact dose rate (mR/hr) on the lines will be used to estimate the concentration (p Ci/cc) of gas in the line.
If the radiation levels do not exceed the above personnel exposure limits the procedures will require collection of gas, particulate, and radioiodine samples.
Silver ziolite cartridges will be used for radioiodine sampling when noble gas interference is expected. Two independent counting rooms are available onsite at McGuire for analysis of these samples. One counting room is located in the Administration Building and the other is located in the Auxiliary Building.
In addition a third counting room is available in the Duke Technical Center located just outside the McGuire exclusion boundary.
In addition, procedures for quantifying radioactive releases through all of the atmospheric steam release valves are being developed. These procedures will involve the use of area radiation monitors. The containment hydrogen purge exhaust discharges through the unit vent and is monitored by the unit vent radiation monitors. 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.
Containment 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
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Atomics and will feature GA detector model number RD23. Each monitor will utilize an ionizatiop chamber to measure gamma radiation and will cover the 0 to 1@ 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 9.0X1g-12 Amps / Rad /hr. Seismic qualification of the monitor is in accord-ance with IEEE344-1975 and environmental qualification is per IEEE323-1971.
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.1 One monitor will be powered from the Train A vital instrument bus, and the other monitor will be powered from the Train B vital instrument bus. Analog f-~s) meters (.one per train) will continuously indicate monitor output in the control i' ' '
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 outage according to procedures. currently being finalized by General Atomics.
In no case will the calibration frequency exceed 36 months. The current schedule 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 least7j0+2(10feetormoreabovethemaximumpost-LOCAwaterlevel at 00 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 Measurement and indication range will extend from -5 psig to 60 psig.
room.
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
's be completely independent of the existing containment pressure transmitters.
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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 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 inside 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 (El 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.
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Containment Hydrogen Monitoring i
Continuous indication of hydrogen concentration in the containment atmosphere will be provided in the control room. This hydrogen monitoring system will consist of two redrndant Comsip, Inc./Delphi Systems Division K-lli 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 IE power supplys.
Each analyzer will have its own 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.
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.
a II-13A 07/10/80 Carry Over
O FINAL RECOMMENDATIONS OF THE BULLETINS AND ORDERS TASK FORCE
Reference:
Action Plan - II.K.3 C.3.3 Duke Power Company will promptly report to the NRC any failure of a McGuire PORV or safety valve to close.
In addition, all challenges to the PORV's or safety valves will be documented and reported to the NRC.
C.3.9 Westinghouse has completed its review of the pressure integral derivative (PID) controller installed on the McGuire PORVs. WCAP 8921, the NSSS Control System Setpoint Study, gives a value of "zero" for the pressurizer PID controller rate time constant. The McGuire time constant will be adjusted accordingly.
C.3.12 The design of McGuire Nuclear Station does not feature a reactor trip on turbine trip. This trip was removed from the McGuire design to prevent unnecessary reactor trips, particularly during initial startup. Unnecessary reactor trips should be avoided to minimize reactor coolant system thermal cycles and challenges
{(',g) to the reactor coolant system protective devices. The removal of this anticipatory trip was possible due to the full load rejection capability of McGuire.
The McGui s trip system keeps. surveillance on process variables which are directle celated to equipment mechanical limitations, such as pressure, pressurizer water level (to prevent water discharge through safety valves) and also on variables which directly affect the heat transfer capability of the reactor (e.g., flow, reactor coolant temperatures). Still other parameters utilized in the reactor trip system are calculated from various process variables.
In any event, whenever a direct process or calculated variable exceeds a setpoint, the reactor will be shut down in order to protect against either gross damage to fuel cladding or loss of system integrity which could lead to release of radioactive fission products into the Containment.
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a II-19 08/06/80
4 IN-PLANT RADIATION MONITORING
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 personnel 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 i
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. One counting room is located in the Administration Building and one in the Auxiliary Building.
In addition, a counting room is available in the Duke Technical Training Center located just outside the McGuire exclusion boundary.
A procedure to determine airborne radioiodine con;.ntrations 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 d
A iodine to determine a "go" or "no go" iodine concentration for respiratory
(
,1 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 minimize counting system problems.
In addition, nitrogen purging of the counting room GeLi detector shields can be used to reduce airborne activity interferences.
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O III-4 08/06/80
STATION DIRECTIVE 3.1.31 APPROVAL DATE j
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REVISION O
DATE DUKE ?0b'ER COMPANY McGUIRE NUCLEAR STATION DUTIES, RESPONSIBILITIES AND QUALIFICATIONS OF THE SHIFT TECHNICAL ADVISOR (STA)
SHIFT TECHNICAL ADVISOR FUNCTION The basic function of the 3IA is to provide additional on-shift capability for evaluation and assessment of off-normal events and normal transient.
During transient situations, the STA must be available (within ten (10) minutes of the Control Room) to advise the Shif t Supervisor of any appropriate action. The STA's primary concern is the safe operation of the plant without jeopardizing 7-~g the health and safety of the public. The STA shall be detached from and inde-(,,)
pendent of the normal line function of shif t operation.
QUALIFICATIONS The individual =ust meet ruke Power ~ Company's general requirements and in addition:
1.
Should be a high school graduate with two (2) years technical school or equivalent experience.
2.
Shall have a minimum of two (2) years nuclear power plant experience accom-panied by an overall knowledge of the plant.
3.
Shall hold a Senior Reactor Operators License.
4.
Should have a working knowledge of steam and water properties.
DUTIES AND RESPONSIBILITIES A.
During normal operation:
1.
Conduct a turnover of plant status with STA being relieved of duties.
[N 2.
Review the plant status at start of assigned shif t.
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I A.
Review Shift Supervisor and Control Room logs.
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3.
Review Out-of-Normal and Removal and Restoration logs for plant status.
C.
Review turnover sheets of Shift Supervisors, Nuclear Control Op-erators and Nuclear Equipment Operators.
D.
Make rounds in the Control Room to review Control Room status.
E.
Review unit work lists.
3.
Review with Shift Supervisors plant status and activities scheduled for shift.
4.
Inform the Shif t Supervisor of his planned activities during the shif t.
5.
Assumes no responsibilities that cannot be immediately put aside in order to advise the Shift Supervisor during off-normal events.
6.
Communicate directly with the Safety Review Engineer on operating ex-periences at the station as well as other operating plants.
B.
During Off-Normal Operation.
1.
Evaluate plant conditions from available information and determine appropriate responses.
2.
Advise Shift Supervisor of appropriate responses to the situation.
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3.
Obtain overall picture of the situation by analyzing all available information and prevent shift personnel from " locking in" on one indication.
4.
Be readily available (within ten (10) minutes) to the Control Room in the event of an off-normal or normal transient.
RELATICNSHIPS 1.
The STA will report to an Operating Engineer not responsible for the Shifts.
2.
The STA vill maintain a close working relationship with the operating shifts.
3.
The STA sill maintain a good working relationship with other station groups as well as Steam Production Department personnel.
4.
The STA will only advise the Shif t Supervisor and will not direct the actions of shif t personnel during normal and off-normal transients.
V(D
1 STATION DIRECTIVE 3.1.32 APPROVAL NN' DATE DUKE POWER COMPANY-MCGUIRE NUCLEAR STATION STATION SAFETY ENGINEERING GROUP OBJECTIVE This directive will define and establish administrative guidelines for function-ing of the on-site Station Safety Engineering Group (SSEG). The responsibilities and distribution of reports for the SSEG will be as outlined below.
APPLICABILITY The SSEG will function as a full-time group consisting of four members and a chairman. The SSEG will report directly to the Station Manager and to the Direc-tor of the Nuclear Safety Review Board. All SSEG recommendations will be sent directly to these groups and they will assure proper action is taken to address SSEG recommendations.
MEMBERSHIP All members of the SSEG shall have at least six years of technical experience with 4
a minimum of two years being nuclear station experience. A maximum of four years of the six years may be fulfilled by academic or related technical training. The four members of the SSEG will be chosen from the attached list (Attachment 1) with at least one person from each of the four areas listed. These members will serve for a period of four months with one member being replaced each month.
1 The chairman of the SSEG will be the Projects and Licensing Engineer or designee as appoinced by the Station Manager.
RESPONSIBILITIES The SSEG will function as an independent technical review group in the areas out-lined below.
A.
Licensee Event Report (LER) Summaries - Each month the General Office Licensing Group will forward a summary list of all LER's applicable to McGuire to the SSEG. The chairman will assign these LER's to members of the SSEG to determine if any specific action needs to be taken at McGuire to prevent or mitigate the consequences of a similar event. All recommendations will be forwarded directly to the responsible group, Station Manager and Director of the NSRB.
1 B.
Effectiveness of Plant Programs - The chairman will assign members of the SSEG specific station programs to review and determine if any recommendations to increase the effectiveness of the program should be considered. After review by each member of the SSEG, any recommendations will be forwarded to the respon-sible station groups, Station Manager and Director of the NSRB.
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C.
Plant Modification Review - Upon completion of the design package si and prior to implementation, all design changes involving struc-tures, systems, or components with QA conditions will be reviewed by the SSEG. All comments of the SSEG must be resolved with the modification designer prior to final approval for implementation.
D.
Station Procedures and Changes - All station procedures and/or changes to procedures that involve an unreviewed safety question must be reviewed by the SSEG.
In addition, the SSEG will review any procedure referred by a station group. All comments of the SSEG will be resolved with the group responsible prior to final approval.
E.
Plant Incident Reports - All incident reports involving reportable items as defined by Station Directive 2.8.1 or other investigations as deemed appropriate by the chairman or statica management will be assigned to a member of the SSEG for investigation and prepara-tion of the station incident report as outlined in Directive 2.8.1.
The incident report will be reviewed by the SSEG and any additional recommendations included. These reports will be sent to groups as outlined by Station Directive 2.8.1.
CONDUCT OF MEETINGS The chairman will call formal meetings of the SSEG to discharge responsibilities as outlined above. A quorum for the meetings will require three of the four
[,, '
assigned members in addition to the chairman. Any individual from a group listed in Attachment I can substitute for the assigned member from the same group. The N
Station Manager or Group Superintendent can also serve as the chairman of the SSEG.
Written minutes shall be prepared by the SSEG Chairman or his designee for each meeting of the SSEG. These minutes will document the actions taken by the SSEG and nill include, but not limited to the following:
A.
Date of SSEG meeting.
B.
Meeting number assigned according to the following format:
XXX XX Year -
Sequential Number to provide a convenient reference number for each meeting.
C.
Members of the SSEG present.
D.
Other personnel present as requested by the SSEG Chairman.
E.
Details of the documents reviewed and the action taken by the SSEG.
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F.
Signature of the presiding Chairman for the SSEG meeting.
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G. ' Copies of these minutes should be sent to the following individuals:
- Station Manager
- Superintendent of Administration
- Superintendent of Technical Services
- Superintendent of Maintenance
- Superintendent of Operations
- Chairman, Nuclear Safety Review Board
- General Office, Project Coordination and Licensing Section
('
NRC REQUEST FOR INFORMATION TRA'iSMITTED BY LETTER OF JUNE 4,1980 FROM B. J. YOUNGBLOOD, CHIEF, LICENSING 3 RANCH NO. 1 DIVISION OF LICENSING 1.
Provide additional information and evaluations for the effect of high post-accident radiation levels on access and equipment operation in vital areas as follows:
(a)
Identify those areas considered vital areas for post-accident recovery, considering the following areas: control room, on-site technical control center, operational control center, security center, emergency power supplies, radioactive waste control panels, recombiner hookup and controls, hydrogen purge controls, instrument panels, containment isolation valve reset controls, sampling stations, sample analysis stations, manual ECCS alignment, mator control center. State which areas will not require access under postulated post-accident conditions and explain why any particular area is not considered a vital area for the post-accident criteria refer-enced.
(b) Identify those syste=s which may contain high levels of radioactivity and have been evaluated for effects on access to cnd operations in vital areas. Consider (as a min 4="r) residual heat removal safety injection, CVCS, demineralizer, charging, RC filters, seal water filter, liquid Q(g radwaste, gaseous radwaste. Provide an explanation for any of the above systems not evaluated or considered.
Response
See Plant Shielding 2.
Revise and broaden your response of January 24, 195) so as to provide a desceip-tion of the two high range containment monitors required by our letter of Novem-ber 9, 1979, implementing the Lessons Learned item 2.1.8.b of NUREG-0578, and specify the location of these monitors (inside contain=ent). The description of the monitors should include:
a.
type of radiation measured b.
the range or ranges of the monitors. If two or more monitors are required to span the range in Table 2.8.1.b.3,of our November 9, 8
1979 letter (10 rad /hr total radiation or 10' R/hr photons only), the ranges of the subsystem monitors must overlap (i.e., upper value/ low value of overlap) by at least a factor of 10; c.
location of and type of readout (continuous and recording);
d.
energy response (sensitive to 60 kev);
calibration frequency and methods (refueling frequency);
e.
f.
verification that the monitors are powered by separate vital instrument buses; g.
verification that the monitors will be operational by January 1,1981; f3 h.
verification that the monitors meet :he seismic qualifications of
(^ )
Requlatory Gui:'e 1.100 (Seismic Category I) and are environmentally qualified to survive an in-containment LOCA in accordance with Regulatory Guide 1.89.
08/06/80
(O The location of the monitors should be shown on plant layout drawings. The t
monitors should be located in a manner as to provide a reasonable assessment of radiation levels inside containment. Monitors should not be placed in areas which are protected by massive shielding.
Response
See Additional Accident Monitoring Instrumentation 3.
Describe the provisions which have been made to sample and analyze airborne radioiodine in vital areas as follows:
a.
portable monitoring capability for radioiodine sampling; b.
results available within 10 minutes at a time when usage of counting analyses facilities is heavy; c.
controls to prevent counting system saturation from high sample activities; d.
clean air ventilation for counting facilities to reduce analysis inaccuracies; e.
provisions for reducing background radiation levels in the analysis facility, in the sample counting device, or from the sample; f.
procedures for keeping personnel exposures ALARA during sample taking and analysis.
Response
See In-Plant Radiation Monitoring N-i 4.
Your FSAR indicates that the Health Physics Supervisor may have direct access to the Station Manager in matters concerning any phase of radiation protection (p. 13.1-13), the station organization chart in Figure 13.1.2-1 shows Health Physics Supervisor reporting through the Technical Services Superintendent to the Plant Manager. Regulatory Guide 8.8, "Information Relevant to Ensuring Occupational Radiation Exposures at Nuclear Power Stations will be As Low As is Reasonably Achievable," states that the Radiation Protection Manager (RPM),
equivalent to your Health Physics Supervisor, should have direct recourse to the plant manager and should be independent of the station technical support division. The " Draft Criteria for Utility Management and Technical Competence" specifies that the RPM should report directly to the Plant Manager.
It is our position that your Health Physics Supervisor report directly to the Station Manager and that the Station organization be revised accordingly.
Response
The McGuire Station Health Physicist reports to the Superintendent of Technical Services. He does have direct recourse to the Station Manager should the situa-tion warrant it.
On June 27, 1980 a NRC management review team concluded an audit of Duke Power Company's readiness to operate McGuire Nuclear Station. This team utilized the " Draft Criteria for Utility Management and Technical Coti,etence" as one of the bases for their review. The team reviewed the McGuire station organiza-tion and identified no deficiencies regarding the Health Physics organization.
f 's Duke believes that the existing McGuire station organization provides for adequate
(,)
protection of the health and safety of the public and plant personnel. Therefore, no changes in the station organization are planned.
08/06/80
9 C
NRC REQUEST FOR INFORMATION TRANSMITTED BY LETTER OF JUNE 30, 1980 FROM B. J. YOUNGBLOOD s
CHIEF, LICENSING BRANCH NO. 1 DIVISION OF LICENSING i
- 1) Additional Accident Monitoring Instrumentation 1
l a) Before fuel loading, an interim method is required when the high range noble gas effluent monitors are not yet installed and operable. You should describe the interim method, addressing item 2.1.8.b enclosed in our letter dated November 9, 1979, pages 31 to 36, providing infor-mation required in 1.A.l.a and 1.A.l.b for noble gas effluents and l
2.A.1 and 2.A.2 for particu.1. ate and radioiodine effluents. Your response should contain a descriptive summary of the interim procedures i
for quantifying high level accidental radioactivity releases to meet the requirement in the Action Plan NUREG-0660, Appendix A, Table A.1, item (17) for II.F.1.(a).
d b) By January 1, 1981, complete the installation of the high range noble gas effluent monitors II.F.1.(f) and provide the information required in item 2.1.8.b sections 1.A and 2.B given in the November 9, 1979 l
letter. Clarify that the steam dump / safety and containment hydrogen purge exhaust will have high range noble gas effluent monitors.
Response
i See Additional Accident Monitoring Instrumentation
- 2) Primary Coolant Sources Outside Containment Before full power operation, provide a description of the method to be used during refueling outage leak rate tests and the weekly leak test procedure.
Discuss the test method to be used for each system or subsystem, such as hydraulic, mass spectrometer, freon, etc., and the acceptance criteria for-the test.
Compare the leak test criteria to area and effluent radiation mor.' tor levels.
Indicate the steps to be taken to minimize occupational radiation exposure, maintain test results, repair leaks and assure system completeness. Specify the staffing and training requirements.
Response
J See Primary Coolant Sources Outside Containment Both the periodic leak rate test procedure and the weekly leakage check pro-cedure will be provided prior to power escalation of Unit 1.
i
- 3) Post Accident Sampling Before full power operation prior to January 1, 1981, provide a descriptive summary of the interim provisions and procedures for sampling and analyzing
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the reactor coolant and the containment atmosphere. Consider the modifica-
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tions needed for the physical, chemical, as well as the radiological analysis steps. By January 1, 1981, provide a description and final system design of 08/06/80
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I the new accident level sampling panel, and modifications to the sample handling and counting facilities to achieve analysis within the time specified in item 2.1.8.a given in the November 9, 1979 letter.
Response
The response to this request will be provided at a later date.
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f NRC REQUEST FOR INFORMATION TRANSMITTED BY LETTER OF JULY 2, 1980 FROM B. J. YOUNGBLOOD, CHIEF, LICENSING BRANCH NO. 1 DIVISION OF LICENSING II.K.1 IE Bulletins on Measures to Mitigate Small Break LOCAs and Loss of Feedwater Accidents C.I.5 The response provided on the review of all valves involved with Engineered Safety Features (ESF) operation is qualitative in nature.
The following information is needed to assess the adequacy of the review required by the IE Bulletins.
a) Describe the review process employed by Duke Power to verify that valve positioning requirements and valve positions are met in the ESF tests. This description should include measured and visual inspections made for the verification of valve operations.
b) Describe the procedure used to verify valve response time, and the measurements and visual observations made for this verifica-tion.
c) Describe the review of positive controls and maintenance
(~N procedures performed in response to the IE Bulletins.
b)
C.1.10 a) Describe the process by which the operability of redundant systems and safety related systems is verified.
b) Describe the process for notification of operators when safety-related systems are removed from, or returned to service.
c) Describe the preparation, contents, and maintenance of historical records of safety-related systems maintenance.
Response
See IE Bulletins on Measures to Mitigate Small-Break LOCAs and Loss of Feedwater Accidents.
II.K.3 Final Recommendations of the Bulletins and Orders Task Force C.3.9 Provide a schedule for completion of your review of the pressure integral derivative controller.
Response
See Final Recommendations of the Bulletins and Orders Task Force C.3.12 Provide a discussion on analyses performed to assess plant response,
(}
and potential for PORV challenges when turbine trip, loss of feed-( )
water flow, or loss of steam generator secondary inventory occurs.
08/06/80
Response
The response to this request is not yet complete and thus will be provided at a later date.
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