ML18037A248
| ML18037A248 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/21/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Dise D NIAGARA MOHAWK POWER CORP. |
| References | |
| NUDOCS 8004100032 | |
| Download: ML18037A248 (32) | |
Text
4 Docket No. 50-220 Hr. Donald P. Disc Vice President - Engineering Niagara Hohawk Pbwer Corporation 300 Erie Boulevard ilest
- Syracuse, New York, 13202
Dear Hr. Disc:
~Docket ORB g3'NRRReading Local PDR'RC PDR DEisenhut RTedesco l<Gammill RVollmer
'JHiller
.BGrimes LShao TIppol,ito SNorris PPolk DVerrel,li
- Atty, OELD'I8E.(3)
Ctlel son CLong NAnderson LBarrett TTelford DVerrelli Enclosed for your information is the Staff's evaluation for the Nine Hile Point Nuclear Station, Unit No.
1 of the. actions you have taken to satisfy the Category "A" items of the NRC recomnendations resulting from THI-2 Lessons Learned.
This evaluation is based on your submitted documentation'nd the discussions'etween our staffs at a site visit on Harch 12, 1980.
A l.ist of meeting attendees is also enclosed.
e Based on our review, we conclude that you have satisfactorily met all Category "A" requirements.
The adequacy of certain implemented procedures and modifications will be verified by our Office of Inspection and Enforce-ment.
Each of these is discussed in our evaluation.
Should you. have any questions regarding our evaluation, please contact us.
e Sincerely, Thomas A. Ippolito, Chief Operating Reactors Branch 83 Division of Operating Reactors
Enclosures:
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Eugene B. Thomas, Jr., Esquire
- LeBoeuf, Lamb, Leiby 8 MacRae 1757 N Street, N.W.
Washington, D. C.
20036 State University College at Oswego Penfield Library - Documents
- Oswego, New York 13126
4
EVALUATION Of LICENSEE'S COMPLIANCE WITH CATEGORY "A" I'TENS OF NRC RECOMMENDATIONS RESULTING FROM TMI-'2 LESSONS LEARNED Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Unit No.
1 Docket No. 50.-220 P
Date:
March </, 1980
0
INTRODUCTION By 1qtters dated October 18(1),
November 26( ), December 19( ), 20, 31(5),
- 1979, and January 31(6), 1980, Niagara Mohawk Power Corporation (licensee) submitted commitments and documentation of actions taken at Nine Nile Point Power Station, Unit No. 1, to implement our requirements resulting from TMI'-2 Lessons Learned.
To expedite our review of the licensee's
- actions, members of the staff visited the licensee's facility on Parch. 12, 1980.
This report is an evaluation of the licensee's efforts to implement each Category "A" item which was to have been completed by January 1980.
EVALUATION Each of the Category "A" requirements applicable to BWRs is identified below.
The staff's requirements are set forth in Reference 7; the acceptance criteria is documented in Reference 8.
The numbered designation of each item ts consistent with the identifications used in NUREG-0578.
EMERGENCY POWER SUPPLY The NRC requirement, as it is applicable to BWR's, is that provisions must be made such that the power-operated relief valves can be supplied emergency power when off-site power is not available.
Further, for air-operated
- valves, emergency power must be available to the air compressors in order to provide a long term supply of air.
The reactor water level instrumentation must also be capable of operating from emergency power.
The licensee has stated(~)
that the relief valves at Nine Mile Point are electromatic operated relief valves-, supplied by emergency electrical power.
This type relief valve does qqf. require instrument air for operation.
The licensee also stated<'~
that vessel level indication instrument channels for safety system activation and control are also powered by emergency power.
Based on our review we have determined that no modifications are necessary to satisfy the requirements for this item.
PERFORMANCE TESTING FOR BWR RELIEF AND SAFETY VALVES The staff's position is that Boiling Water Reactor licensees shall functionally test the reactor coolant system relief and safety valves to demonstrate operability under expected operating and flow conditions.
The Category "A" requirement fs for the licensee to commit to perform an appropriate test program.
The licensee is a member of a GE BWR Owners grqup and has committed(2) to a test program adopted by this Owners Group.<9J
0 I
We conclude that the licensee has satisfied the Category "A" requirements for this item.
2.1.3.a Direct indication of Power-0 crated Relief Valves and Safet Valve ositson or WR s The staff's position is that BWR licensees shall provide a positive indi-cation for reactor coolant system relief and safety valves.
The valve position should be indicated and alarmed in the control room and derived from a reliable valve posi:tion detection device or a reliable indication of flow in the di'scharge pipe so that the operator is provided with an unambiguous indi.cation of valve position. If the valve position indication is not safety grade, a reliable single channel dtrect indication powered from the emergency bus may be provided if backup methods of determining valve position are available.
Further, the valve position indication should be setsmtcally qualified consistent with the components or system to which it is attached.
If seismic qualifications are not feasible by January 1, 1980, then justiftcation should be provided and a schedule submitted for upgrading the system to meet the seismic requirements.
To meet the above position, the licensee has provided an acoustical system to monitor the position of each of the relief and spring safety valves.
The acoustical system consists of a hermetically sealed piezoelectric sensor mounted on the downcomer piping of the relief valve and on the flange of the spring safety valves.
The sensor is held in place by a special stainless steel band clamp.
The sensor is connected to the preamplifier through the use of high temperature, low noise coaxial cable.
The signal output from the preamplifter goes to individual Flow Detector Modules which provide indication of the position of each valve.
The open or closed position for the relief valves is also indicated on the main control panel.
For the twenty two relief and safety valves, the preamplifiers are located outside containment, where the temperature during an accident is lower than inside containment.
Each Flow Detector module is located on an auxiliary rack below the main control room and pr ovides indicator lights for "closed,"
"open" positions, and a
"memory circuit."
The "memory circuit" for each valve when ectivated, stays on until manually reset; thus it provides an indication of valve actuation even though the valve may have since closed.
If any of the flow detectors indicate a valve in the open position, a common dedicated single window of the plant annunciator is activated and when modifications are completed its signal will be inputed to the plant process computer and event recorder.
An on-line system test circuit for alarm has been provided.
The Nine Mile Plant has 6 electromatic relief valves and 16 spring safety valves.
The relief valves have the capability to be-operated manually; however, three of these valves have been dedicated to the ADS function while the remaining three provide redundancy.
The acoustic monitor ing system installed on each relief or safety valve is not fully safety grade.
- However,
the licensee has stated and we agree that the system is a reliable single channel system that provides direct indication of valve position.
The system is powered from the emergency AC bus that has automatic transfer capabili'ty to a DC power supply in the event of a failure of AC power system.
Back-up valve position indication information is provided and discussed in the emergency procedures so that the operator can make a diagnosis and'ake appropriate action.
The back-up valve position indication is Pro-vided by temPerature indicators.
Each individual valve has gn pmbeddgd type thermocouple attached to the tailpipe downstream of the valve discharge point.
Signals derived from the embedded thermocouple are readout and alarmed on the process computer.
The power for the back-up temperature monitor position indicators is provided from a non-class lE instrument bus.
Therefore, tn the event of a single failure of a power supply, at least one indicating system is available to provide the reactor operator with valve status.
The temperature indication instrumentation is already seismically and environmentally qualified and is available for backup verification of valve position.
The acoustic monitoring system valve position indicators have not yet been seismically or environmentally qualified.
The licensee stated that the position indicati'on system and components both inside and outside the containment ts presently being environmentally and seismically qualified by partictpating in the BSW qualification program.
The position indication system components mT1 be seismically qualified in accordance with IEEE 344, 1975 and qualified for their appropriate environment in accordance with I'EEE 323, 1974 by January 1981.
This schedule meets our requirements.
Based on our review of the licensee's submittal,, we conclude that the licensee is in compliance with the direct indication of power-operated relief valves and safety valve position and schedule requirements for upgrading the system to meet the seismic requirements as outlined in NUREG-0578, and is, therefore, acceptable.
2.1.3.b Instrumentation for Inade uate Core Coolin The NRC requirements, licensee actions and our evaluation thereof for this item will be evaluated separately by the NRC Bulletins and Orders Task Forcy agd reported in NUREG-0645 which is incorporated herein by referenceOO).
2.1.4 CONTAINMENT ISOLATION The NRC requirements are that the licensee is to:
(a) carefully reconsider the determination of which systems should be considered essential or non-essential for safety, (b) modify systems as may be necessary, to isolate all non-essential systems by automatic, diverse, safety-grade isolation signals, and (c) modify systems, as may be necessary, to assure that the resetting of the containment isolation signals does not cause. the inadvertent re-opening of containment isolation valves.
The licensee's classification of essential systems was based on determination of'hich are required or could be of direct aid in mitigating the conse-quences of an accident.
All other systems which penetrate the primary contginment are non-essential.
The licensee's letter dated November 28, 1979'Ull includes an i'dentification of each penetration, classification as essential (engineered safety function} or non-essential, and identification of the i'solation signals for each.
As stated in NUREG-0578 our goal is to use information provtded Gy licensees to develop a consistent set of guide-lines for the selection of essential and non-essential systems.
Accordingly, the licensee has sattsfied this aspect of the requirements for this item.
The ltcensee stated(5} that certain systems (reactor building closed loop cooling to the recirculation pump coolers and dry well coolers) do not require containment isolation signals since they are closed systems that do not comnunicate with the reactor coolant pressure boundary or free space of the containment.
Two systems (the recirculation sample and suppression chamber to waste system lines) are normally closed during operation.
The licensee has comaitted to install automatic isolation valves in these systems.
In the interim the licensee coranitted at the site visit, that when such systems are tn use, an operator will be dedicated to the function of assuring that tsolation valves are closed in the event of a containment isolation signal.
Based on our review of References 5 and 11 as well as discussions at the site visit we conclude that non-essential penetrations are either (lE automatically isolated by diverse signals, (2) adequate compensatory measures have been instituted or (3} a rationale for deviation from the general requirements has been provided.
We find that the licensee has satisfied the basic intent of this aspect of the requirements for this itern.
The licensee's design of control switches for containment isolation valves was discussed at the site visit.
The design involves the use of sprtng-return-to-neutral control switches and holding relays.
For this type design resetting of a containment isolation signal does not result in the automatic reopening of containment isolation valves.
Me find that no modification to the design is required.
Based on the above, we conclude that the licensee has adequately conformed to the requirements of this litem The Office of Inspection and Enforcement will verify the adequacy of procedures and the completion of design changes as discussed above.
2.1.5.a Dedicated Penetr attons for External Recombiner of Post-Accident External Pur e
S stem The staff's position is that licensees whose plant uses external recombiners or purge systems for post-accident control of combustible gas in the contain-
ment atmosphere should provide a containment isolation system that is dedicated to that function only.
The system's design should be redundant and meet out single fai'lure requirements to that criterion 54 and 55 of the General Design Criter ta are met and that the system is sized to satisfy the flow requirements of the recombiner or purge system.
This requirement is applicable to those plants whose lfcensing basis includes requirements for external or purge systems for post-accident control of combustible gas. tn the primary containment.
The Nine Mile Point Unit is designed to use a Containment Atmosphere Control (CAC) system prior to each startup and during routine operations to maintain the oxygen concentration in the primary containment atmos-phere to less than 5 percent to ensur e that combustion of the hydrogen and oxygen cannot occur.
Me have determined that the CAC system consists of the following major subsystems:
The normal containment purge and exhaust subsystem, the containment inerting subsystem and the containment atmospheric make-up subsystem.
These subsystems do not perform any safety function.
Only those components associated with maintaining the containment isolation integrity (up to and including second containment isolation valve) are safety related and have been.designed to seismic Category I requirements.
The Containment Atmospheric Dilution (CAD) system performs the safety function of limiting initial oxygen concentration to less than 5 percent in order-to preclude a flammable mixture in the containment immediately following a LOCA and to matntai'n this inerted primary containment mixture on a long term basis following a LOCA.
The CAD system is used during emergencies and as such has been designed to seismic Category I require-ments; electrical components meet applicable portions of I'EEE-279, and have suitable redundancy and interconnections so that a single failure of an active component will not render the system inoperable.
The CAD system is functionally independent from the normal inerting system and its components include a storage vessel, electric vaporizers, redundant lines and valves and associated instrumentation.
The nitrogen from the CAD system is injected into the drywell or torus using the purge air system lines.
The CAD system branch lines are connected to the purge lines downstream of their redundant containment isolation valves.
Two solenoid actuated isolation valves for each of the reduiidant torus and drywell CAD lines have remote control switches located in the main control room.
In addition, two analyzers for hydrogen and oxygen have been provided.for the. contai'nment-drywell/torus that are redundant to each other and are designed to meet seismic and I'EEE 279 requirements.
Initially, the licensee stated that the CAD system at Nine Mile Point was reviewed to verify that isolation provisions for piping and inter-connected lines are single failure proof during CAD operations.
The licensee reported tlie results of the review in the December 31, 1979 submittal.
They have verified that tlie containment isolation provisions for all lines (including vent and purge) are single failure proof during
CAD system operations except for a single pathway.
During this operation a single failure of a Blocking valve could result in an uncontrolled pathway from containment for radioactivity to escape through the ventilation system.
Tlie licensee proposed a modification to interlock normally open valves to assure that they automatically close during nitrogen addi'tion to the containment, so that two isolation valves in series exists for all pathways.
The modification will be completed as soon as practicable But pri'or to January 1, 1981.
In the interim, the licensee has committed to modifying procedures to assure that the pumpback system is not in operation during CAD operations.
With the pumpback system not in operation, two additional isolation valves are automatically closed.
This pr events a single failure pathway from existing during the nitrogen addition operations.
Based on the above, we conclude that the licensee has satisfied the Category "A" requirements related to this item.
2.1.5.c Recombiner Procedures Tfie NRC requirements for thi;a item apply only to those plants that include hydrogen recombiners as a design basis for licensing.
Me have determined that this item is not applicable to the Nine Nile Potnt Plant.
2.1.6.a S stems Inte rit The NRC objective is to eliminate or prevent the release of significant amounts-of radioactivity to the environment via leakage from engineered safety systems and auxiliary systems, which are located outside reactor containment.
The requirements are to implement practical measures to reduce
- leakage, report leakage measurements to the NRC and establish a preventive maintenance program to maintain leakage at as-low-as practicable levels.
Based on our review of licensee submittal and discussion with licensee during the NRR/OIE site visit, we find that the licensee has tested and measured leak tightness of systems, developed leak reduction, and initiated a preventive maintenance program.
We conclude that the licensee has satisfied the requirements of this Category "A" item.
The Office of Inspection and Enforcement (OIE) will review the licensee's procedures to verify adequacy.
OIE will also verify the implementation of a preventative maintenance program and the completion of personnel training.
Results will be reported in appropriate inspection reports.
2.1.6.b Plant Shieldin
'Review.
The Category "A" requirements for this item are to perform a design review of current plant shielding to identify where corrective actions are needed to permit personnel access to vital areas, and to protect safety equipment.
The licensee has completed a general plant shielding review of the vital areas requiring continuous and infrequent access.
Problem areas such as coolant sampling systems wtll Be relocated from the containment, or a new, shi'elded sampling stnk will be provided in the turbine building.
The present shielding of the control room is sufficient to maintain dose to personnel within requi'red ltmit.
The TSC and OSC may require additional shielding; detai'led evaluation i's in progress.
Shielding review of safety system components for degradation from TI'D sources is presently in the final stages of completion.
The results will Be reported as part of the acti'ons under OlE Eullettn 79-018.
Based on the above, we conclude that the licensee has satisfied the Category "A" requirements for this item.
2.1.7.a Auto Initiation of AFM 2.1.7.b AR< Flow These items (2.1.7.a and 2.1.7.b) are unique to PMRs and are not applicable to the Nine Nile Point Plant.
2.1.8.A Post-Accident Sam 1 in The NRC objective is to quantify the degree of core damage in the course of an accident by radiological and chemical analysts of samples of reactor coolant and containment atmosphere.
The Category "A" requirements are:
(a) to review the desi'gn of reactor coolant and containment sampling system to determi'ne the capability of personnel to oBtai'n a sample (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) under accident conditions without exposing an individual in excess of 3 Rem and 18 3/4 Rems to the whole body or extremities; (b) to review operational procedures of the radiological spectrum and chemical analysis facilities to determine the capability to quantify radioisotopes that are indicators of the degree of core damage; and (c) to describe proposed plant modifications.
The licensee'has completed a design review of ment atmosphere sampling systems.
Additional
- however, because personnel radiation exposure limit further modifications may be necessary.
obtaining containment atmosphere samples from monitoring system.
reactor coolant and contain-shielding has been provided, may exceed the required The capability exists for the containment H2 and 02 The licensee stated that all samples can be obtained and analyzed for required isotopes within 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Interim procedures have been developed for collecting, transporting and analyzing samples.
Based on our review. we conclude that the licensee has satisfied the requirements of this. Category "A" item.
2.1.8.B. 'Hi'.'Ran e.'Radiation'Moni'tors
'he NRC objective is to have available adequate instrumentation to follow the course of the accident.
The Category "A" requirements are to have procedures quanti'fying effluent releases tn case existing instrumentati'on would go off scale ("provisional fi'x").
This includes a description of System/Method
- employed, and description of procedures for conducting all aspects of the measurement/analysis for noble gases, radioiodtnes, and particulate effluents.
The existing in-line monitors are capable of detecting 50 Ci/sec, and have read out and alarm capability in the main control room.
guantification of higher effluent releases is provided by portable gamma survey instru-
- ments, positioned at a predetermined location to monitor, a portion of the effluent sample line.
This "provisional fix" has been installed and cali,brated.
Conversion factors have been calculated, for detecting up to 10,000 Ct/sec NG effluent releases-at the stack.
Since all station effluents are dis-charged via the stack, only one high range effluent monitor is required.
Interim methods and appropri;ate procedures have been written, approved and implemented.
Personnel tr aintng has been conducted for sampling, quantifying and analyzing effluent releases.
Based on our review, we conclude that the licensee has satisfied the Category "A" requirements of this item.
2.1.8.C Im roved Iodine Instrumentation The NRC Category "A" requirements are that each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where personnel may be present following an accident.
The licensee has available approximately 12 portab)e air samplers, using charcoal cartridges.
These samplers are not equipped with single channel analyzers.
The charcoal cartridges will be analyzed in the counting room which. i's located approximately 1 minute walking distance from the control
The licensee stated that the air samples can be collected, transported, flushed and analyzed for iodine concen-tration wi'tFiin 10 mtnutes.
>tertm procedures for obtaining, transporting, preparing and analyztng samples have been developed and implemented.
Based on our review. we conclude -that the licensee has satisfied the requirements of this item.
2.2.1.A Shift Su ervisor Res onsibilit The NRC requirement for this item is to revise, as necessary, the responsibilities of the Shift Supervisor such that he can provide
- direct, command oversight of operations and perform management review of ongotng operations that are important to safety and not be dis-tracted from these important responsibilities by administrative details.
Tfie ltcensee fiat revised Plant Procedure APN 2A to satisfy this requirement.
We conclude that the licensee has satisfied the requirements of Item 2.2.1.A to provide revised responsibilities and authority for the Shift Supervisor.
Verification of the adequacy of the licensee's procedures will be performed by the Office of Inspection and Enforcement and will be documented by appropriate Inspection Reports.
2.2.1.b Shift Technical Advisor The NRC requirement is for the licensee to provide an on-shift technical advisor (STA) to the shift supervisor to ser ve the two functions of accident assessment and operating experience assessment.
As a supplement to the operating staff, the STA must be able to report to the control room within 10 minutes to assist in diagnosing an off-normal event.
The licensee stated~") that shift manning will be augmented by an Assistant Shift Supervisor to satisfy the accident assessment function of the Shift Technical Advisor.
The operating experience assessment function is performed by station professional personnel and corporate level engineering personnel coordinated by a person on the statton technical staff whose primary duty will be operating experience assessment.
We conclude that the licensee has satisfied the Category "A" requirements for this item.
2.2.1.c Shift and Relief Turnover Procedures The NRC requirement is for the licensee to assure that procedures are adequate to provide guidance for a complete and systematic turnover between the off-going and on-coming shift to assure that critical plant parameters are within limits and that the availability and alignment of safety systems are made known to the on-coming shift.
The licensee has revised Plant Procedure APN 2A to implement this item.
We conclude that the licensee has satisfied the requirements of Item 2.2.1 to provide new procedures.
Verification of the adequacy of the implemented checklists and logs will be performed by the Office of Inspection and Enforcement and will be documented by appropriate Inspection Reports.
2.2.2.A Control Room Access The NRC requirement includes implementing procedures to limit access to the control room and establishing clear lines of authority in the control room in the event of an emergency.
The licensee has revised Plant Procedure APN 2A to implement this item.
We conclude that the licensee has satisfied the requirements of Item 2.2.2.A.
Verification of the adequacy of the implemented procedures will be performed by the Office of Inspection and Enforcement and will be documented by appropriate Inspection Reports.
2.2;2.b Technical Su ort Center The NRC requirement is that each licensee establish and maintain an onsite technical support center (TSC) separate from and in close proximity to the control room.
The TSC should have reliable coomuni-cation systems and plant as-built technical data to provide information to those individuals knowledgeable and responsible for engineering and management support to reactor operations in the event of an accident.
Further, the licensee must describe the long range plan to upgrade the TSC to meet the Category "B" requirements The licensee has designated the Training Room in the Administration Building as the Onsite Technical Support Center (TSC).
During the NRR/
OIE site visit we toured the TSC.
The center is habitable to the same degree as the control room.
Direct telephone communications and airborne and radiation monitoring capability have been provided.
Access to permanent plant records, as-built drawings and procedures is available.
The TSC staff has access to technical data by two selector typers which print out the same information available in the control room.
In addition, display of plant parameters can be provided by means of a camera with focus, zoom and pen and tilt controls.
The existence and staffing of the center are included in the Nine Mile Point Unit 1
The licensee's submittal dated January 31, 1979 also includes a discussion of his plans to upgrade the Center to satisfy our Category "B" requirements.
We conclude that the licensee has satisfied the Category "A" requirements for this item.
2.2.2.c 0 erational Su ort Center The NRC requirement is to establish an area in which shift personnel can report for further ins tructi ons from the operations staff.
- 11 During the NRR/OIE site visit the licensee stated that the Plant lunchroom has been designated Onsite Operational Support Center.
The center has telephone comnunications.
The licensee's Emergency Plan covers this Center.
I He conclude that the licensee has satisfied the requirements of 2.2.2.c.
NRR ITEM:
REACTOR COOLANT SYSTEM VENTING As specifically related to BNRs, the Category A requirements of this item is to provide current design information to demonstrate that non-condensable gases can be vented from the primary coolant system, including isolation condensers.
The licensee's submittal dated December 31, 1979, provided design infor-mation on the capability of the Nine Mile Point design for remotely venting non-condensables from the reactor coolant system.
Reactor vessel head high points can be vented by relief valves and the head vent system.
The ltcensee's review of the capability to vent the isolation condensers indicated that modifications are necessary to assure venting capability during accident conditions.
The submittal described the modifications.
The schedule for completion is consistent with our requirements for a
'Category- "B" item.
Based on our review we have determined that the licensee has satisfied the Category "A" requirements for this item.
4
1.
- Letter, NMPC (Rhode) to NRC (Eisenhut)
October 18, 1979.
2.
- Letter, NMPC (Bartlett} to NRC (Denton),
November 26, 1979.
3.
- Letter, NMPC (Disc} to NRC (Denton),
December 19, 1979.
4.
- Letter, NMPC (Rhode} to NRC (Denton}, December 20, 1979.
5.
- Letter, NMPC (Disc) to NRC (Denton),
December 31, 1979.
6.
- Letter, NMPC (Disc) to NRC (Denton), January "31, 1980.
- 7. 'etter, NRC (Eisenhut) to ALL OPERATING NUCLEAR POWER PLANTS, September 13, 1979.
8.
- Letter, NRC (Denton) to ALL OPERATING NUCLEAR POMER PLANTS, October 30, 1979.
9.
- Letter, BHR OQIERS GROUP (Keenan} to NRC (Eisenhut},
December 14, 1979.
10.
NUREG-0645 Report of the Bulletins and Orders Task Force, January 1980.
ll.
- Letter, NMPC (Disc} to NRC (Ippolito}, November 28, 1979.
e
~
V I
Lessons Learned Site Visit Nine Mile Point 1
NRC P. J. Polk Louis B. Riani Frank C. Skopec Peter Francisco Melvin A. Silliman Walt Baumack T. J. Perkins D. M. Verrelli E. Leach B. Taylor NRC NRC NRC NMPC NMPC NRC NMPC NRC NgPC NgPC