ML19309B773
| ML19309B773 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/31/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309B770 | List: |
| References | |
| NUDOCS 8004070165 | |
| Download: ML19309B773 (14) | |
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l EVALUATION OF LICENSEE'S COMPLIANCE WITH I
CATEGORY "A" ITEMS OF NRC RECOMMENDATIONS RESULTING-FROM TMI-2 LESSONS IEARNED ARKANSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR.ONE - UNIT 1 DOCKET No. 50-313
--Date: March 27,1980 8004070 1 6 5-
Introduction I.
INTRODUCTION December 5(3 4
By letters dated Oc{gger 1{ ), Ngyymber(jg( a}n,d February 19(99nd 17and 29 1979 and January 17.
, 18
, 29
, 31 1980, Arkansas Power and Light Company (licensee) submitted commitments and documentation of actions taken at Arkansas Nuclear One, Unit:1 (ANO-1) to 2
implement our requirements resulting from TMI-2 Lessons Learned (L ).
To expedite our review of the licensee's actions, members of the staff visited
.the licensee's facility on January 21, 22 and 23, 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.
II.
EVALUATION Each of the Category "A" requirements: applicable to PWRs is identified below. The staff's requirements are set forth in Reference 10; the accept-ance criteria is documented in Reference'12. The numbered designation of each item is consistent with the identifications used in NUREG-0578.
2.1.1 EMERGENCY POWER SUPPLY REQUIREMENTS (Pressurizer Heaters)
The licensee has determined, based on B&W calculations and startup test experience, that at least 126 kilowatts of pressurizer heaters be available from an assured power source within two hours after loss of off-site power to establish and maintain natural circulation at hot standby conditions. We have reviewed this information and note that this calculated heat toss is similar to and has been accepted for other pressurized water reactors. Based on our review, we believe that sufficient heater capacity has been provided to main-tain pressure control in the pressurizer during normal hot standby conditions. This is in accordance with our position.
The design has available 84 kilowatts of proportional heaters on each of the class IE buses. In addition, 42 kilowatts of (on-off) heaters were added to the swing bus (connected to either one of the diesel generators). The combination will thus produce the necessary 126 kilowatts. The added 42 kilowatts will not exceed the seven day rating of the diesel generators.
In addition, the licensee has available additional 42 kilowatts of heaters which can be manually connected to either of the safety buses to achieve 126 kilowatts of independent and redundant pressurizer heaters on i
each class IE bus.
4 Based on our review of the licensee's submittal, we conclude that i
the licensee is in conformance with the emergency power supply requirements for the pressurizer heaters as outlined in NUREG-0578.
Our Office of Inspection and Enforcement will ver(fy at a later date the installation of this design and that.the procedures to connect j
these extra heaters are included in the operating procedures. This will be documented in an appropriate inspection report.
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2-EMERCENCY _ POWER SU_PPLY REQUIRDiENTS (Pressurizer Level and Power Operated
' Relief Valves and Black Valves)
The PORV and the pressurizer level instrumentation receive emergency power upon loss of offsite power as a feature of the approved existing design.
The PORV motive and control power is provided from the Channel 1 (red)
DC system. The pressurizer instrument channels are powered from vital buses. The PORV. block valve power supply has been modified to be powered from the channel 2 (green) AC system. Thus the power supplies for PORV, PORV block valve and pressurizer level indicating instrumentation are capable of being powered from both offsite and onsite emergency sources.
We have reviewed'the above design and conclude that it is in accordance with our requirement for providing emergency power to the PORV, PORV block valve, and pressurizer level instrumentation. We find the ANO-1 emergency power supply design for these components to be in conformance with our requirements outlined in NUREG-0578. Our Office of Inspection and Enforcement will verify that the power supply modification to PORV block valve have been correctly installed and document this in an appro-priate inspection report.
2.1.2 PERFORMANCE TESTING FOR RELIEF AND SAFETY VALVES The licensee has stated in his response to this item that it will participate in the EPRI program to conduct performance testing of PWR relief and safety valves.
A description of the test program was provided by EPRI in December 1979.
At present, this program is under NRC review to ensure that the NUREG-0578 requirements are met.
We believe that -this commitment provides assurance that the requirements for performance testing of relief and safety valves will be satisfied.
l The basis for this acceptance is that we will review the test program to confirm the applicability to ANO-1.
Completion of the test program is on a schedule different from Category "A" items. Therefore, we conclude that the licensee has satisfied the Category "A" requirements of this item.
2.1.3.a DIRECT INDICATION OF POWER-OPERATED RELIEF VALVES AND SAFETY VALVES POSITION To meet thi-TUREG 0578 requirement the licensee has installed an acoustical titoring system to monitor the position of PORV and safety valv.,. The acoustical system consists of a hermetrically sealed piezoelectric sensor mounted immediately down stream of the PORV and safety valves piping. Redundant sensors are provided for each valve. The sensors are connected to the preamplifiers (one for each sensor) which are located inside containment and transmit redun-dant signals to a signal conditioner located in the control room. The acoustical valve position indication equipment is powered from a class IE bus. ~ This system will ' provide the operaEor with positive indication of valve position and an annunciation-of an open valve in the control room. All equipment will also be mounted as seismic class IE
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.: -installations. In addition, the equipment is environmentally suited for its application. 'The licensee has stated that sufficient qualification i
documentation is not available at present, however, generic qualifica-tion should begin in February 1980, and will require six months to complete.
Backup valve position indication is provided from temperature elements located downstream of the PORY and safety valves. These are monitored in the control room and alarm on high temperature.
Based on our review of the licensee's design, we conclude that the licensee is in compliance with the direct indication of power operated relief valve and. safety valve position and schedule requirements as outlined in NUREG-0578, and is, therefore, acceptable. Our Office of Inspection and Enforcement will verify (1) installation of the above -design, (2) that the procedures to use backup valve posi-tion indication are included in the operating procedures, and (3) the adequacy of the qualification documentation of this equipment when made available. This will be documented'in appropriate inspection reports.
2.1.3.b INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (Existing Instrumentation)
The licensee has stated that its emergency procedures and associated operator training with regard to existing instrumentation to be used by the operator to detect inadequate core cooling conditions, have been updated based on guidelines provided by B&W. Our evaluation for this item will be reported separately.
ADDITIONAL INSTRUMENTATION The licensee has looked at several conceptual designs for reactor vessel water level indication. We conclude that the licensee has satisfied our short term requirement. However, the need to supplement existing instrumentation and to provide unambiguous indications of inadequate core cooling are still under review. We will complete this item during the review of Category "B" items.
SUBC00 LING METER The licensee has installed two primary coolar.t saturation meters. These saturation meters continuously display the margin between actual primary coolant temperature and the saturation temperature (the tempera-ture at which boiling occurs). Two channels of saturation margin measurement and indication'are provided. Each channel consists of l
, a calculator and a digital display.. The pressure input for each calcula-tor is derived from safety grade,. wide-range (0-2500 psig) primary coolant -
pressure input from existing buffered output present in the Engineered Safeguard System. The temperature inputs for-each calculator are two wide-range (120' to 920*F) RTD's from each reactor coolant hot leg loop.
These temperature inputs, however, use non-safety grade temperature The signals available in the non-nuclear instrumentation system.
temperature inputs will be upgraded to safety grade requirements not later than next refueling outage.
The temperature margin to saturation conditions from each calculator is continuously recorded on a two channel strip chart recorder located in
,the control room. This condition can also be individually read from digital indicators mounted in cabinets located in the control room.
Annunciation of low margin to saturation is provided in the control room.
In addition to the above redundant saturation meters backup capability already exists to detect inadequate core cooling conditions from the plant computer. In addition, primary coolant temperature and pressure are directly available to the operators by means of existing console indica-Steam tables are also provided in the control room for use by the tors.
operators to determine saturation margin manually.
Based on our review of licensee's design, we conclude that the design of subcooling meter is in compliance with our requirements as outlined in NUREG-0578 with the exception of safety grade temperature inputs to the subcooling meter which are not provided. A commitment has been made to upgrade the temperature inputs to safety grade requirements no later than the next refueling stage. We find this commitment acceptable for the short term. Our Of fice of-Inspection and Enforcement will verify at a later date (1) installation of the subcooling meters, (2) the adequacy of the qualification documents of the meters, calculators and cabinets, (3) future upgrading of the temperature signals to safety grade, and (4) procedures to use the subcooling meters are included in the operating procedures. This will be documented in an appropriate inspection report:.
i 2.1.4 CONTAINMENT ISOLATION The licensee has modified the containment isolation provisions so that diverse parameters will be sensed to assure automatic isolation of non-essential systems under all postulated accident conditions. The para-meters to be used are containment high pressure (>4 psig) signal or low
. reactor coolant system pressure (<l500 psig) signal. Resetting of the
. isolation system will not cause any valve to automatically reopen. A separate manual action is required to open the individual valves.
The licensee has identified 16 systems that have automatically actuated valves which provide penetration isolation. The reactor coolant pump
s-controlled bleed-off and reactor coolant letdown were originally designed to isolate on diverse containment isolation signals.
~Another group of six systems are closed loop cooling systems and are
. isolated only on high reactor building pressure. These systems are normally open during power operation and are necessary for " normal,"
orderly cooldown of the reactor. Justification for this is that none of these systems have direct contact with the reactor coolant such that non-isolation would allow contaminated fluid to escape unless the integrity of the closed loop system was also-violated. We find this justification acceptable.
The remaining eight systems are normally closed during operation and require specific manual operation for valve opening. They currently receive an isolation signal on high reactor building pressure and the licensee has modified the system to provide a low' reactor coolant pressure signal to isolate in case these system valves were inadvertently open.
It should be noted that seal injection water to the reactor coolant pumps is not isolated. This flowpath is considered part of the high pressure injection and backflow is prevented by check valves. We find this item acceptable.
We conclude that the licensee has satisfied our requirements with regard to containment isolation, as set forth in NUREG-0578. Our Office of Inspection and Enforcement will verify the installation of this design and document this in an appropriate inspection report.
2.1.5.a DEDICATED H, CONTROL PENETRATIONS ANO-1 was licensed to use a hydrogen purge system for post-accident combus-tible gas control of the contaiament atmosphere.- Therefore, the plant was required to have a redundant, safety-grade and dedicated hydrogen ti:7a system. The containment penetrations to the hydrogen purge system eer; the single-failure criteria for containment isolation and operation o" the purge system. The system is sized to meet the flow requirements for the system during an accident. This has been verified by the staff.
$45t1 on the above considerations, we conclude for ANO-1 that the 2 M asee has met the Category "A" and Category "B" requirements of chis Lessons Learned item. No further action needs to be taken.
2.1.5.b -INERTING BWR CONTAINMENTS
.This item does not apply to ANO-1.
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- .. 2.1.~5.c MhPURCEPROCEDURES 1
The 11censee has reviewed the procedures and shielding for operating the hydrogen purge system which provides post-accident combustible gas control
- of the containment atmosphere during' an accident'. The licensee states that no changes to shielding for this' system are needed.
The licensee has identified one change to the procedures to operate this system.. Solenoid valves must be added to the seal water supply lines to the system inlet and outlet fans so that manual valve operation will no longer be required. The solenoid valves will be operated from the control room. The licensee will have the solenoid valves installed at the first outage of sufficient duration.
Based on the above considerations, we conclude that the licensee has met the Category "A" requirements for this. item. The changes to the seal water supply lines.of the hydrogen purge system must be completed by January 1, 1981 to meet the Category "B" requirements for this item.
Our Office of Inspection and Enforcement will inspect the plant to determine that the above change is completed by 1/1/81. Verification of the licensee's procedures will also be performed by the'0ffice of Inspection and Enforcement. Thase will be documented in appropriate inspection reports.
2.1.6.a SYSTEM INTEGRITY The licensee has listed the plant systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident. These systems are the dirty liquid radwaste, gas radwaste, makeup and purification, decay heat system (low pressure safety injection) and reactor building spray system. High pressure safety injection is included in the makeup system. The licensee has Laplemented an immediate leak reduction program for these systems -to reduce their present leakage.
The licensee has reported the measured "as found" leakage and the "as reduced" leakage for these systems to NRC.
The licensee has established and implemented a permanent leak reduction program to ' keep future leakage from these systems to as-low-as-practical levels. This program includes integrated leak rate tests once per refueling cycle; identification of leakage from visual surveillance by plant personnel, area radiation monitors and effluent monitors and correc-tive-actions taken; and the existing plant preventive maintenance program.
The licensee has also reviewed the plant design for potential leakage release paths from the above systems due to design and operator deficiencies as discussed in an NRR letter to'the licensee regarding North Anna and Related Incidents dated October 17, 1979. No corrective actions by the licensee were needed.
Based on the above considerations, we conclude that the licensee has met the Category "A" requirements for this item. There are no Category "B" requirements. Verification of the procedures which implement the licensee's
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2.1.6.b PLANT SHIELDING REVIEW
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The licensee has performed a radiation and shielding design review of the spaces around the plant systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident.
These systems are discussed in the evaluation of item 2.1.6.a. <This design review has been provided to NRC which includes maps of the estimated dose 1
rates in different areas of the plants. The radioactive source terms j
assumed for the design review are consistent with the source terms given uba the NRC clarification letter dated October 30, 1979. The licensee will propose long-term changes, if any are needed, to NRC by March 31, 1980. For the short-term, the licensee will pro' vide portable shielding onsite for use'in areas.co reduce personnel' exposure.
The licensee did identify the location of vital areas and equipment and instrument areas in which personnel occupancy may be lLaited. These areas are the sample room, decay heat pump room (i.e., valves must be manually operated), and'certain areas of the control room. Corrective actions for the long term, beyond 1980, to provide adequate access to the control room and sampling area, will be provided in the shielding report. We discussed with the licensee during our plant trip corrective actions for the long term, beyond 1980, for the decay heat pump room to not require access to this room during an accident. The current thinking to correct this problem is to replace certain manual valves by motor operated valves 4.n this room. The licensee will identify any areas in which safcty equipment may be unduly degraded by radiation fields during post-accident operations in the shielding report.
The major contributor to the radiation fields is the makeup and purifica-tion system. The licensee has concluded for this year that the levels of radioactivity in the makeup and purification system can be limited with existing plant equipment and procedures.
Based on the above considerations, we conclude that the licensee has met the Category "A" requirements. An evaluation of the above design review and the licensee's corrective actions will be performed as part of the
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review of the Category "B" Lessons Learned requirements.
2.1.7.a AUTO INITIATION OF AUXILIARY FEEDWATER SYSTEMS
-This item has been reviewed separately by the NRC Bulletins and Orders Task Force..The ANO-l' design was accepted to meet the control grade requirement for auto initiation of the auxiliary feedwater systems.
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2.1.7.b AUXILIARY FEEDWATER FLOW INDICATION TO STEAM GENERATORS The licensee has modified the auxiliary feedwater system flow indication
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system to confore to safety-grcde requirements with the exception of i
the four panel mounted indicators. This system will be testable and is located outside containment. Our review of the system design indicates that the system does meet safety grade requirements.
Based on our review of the above design, we conclude that the present modified design meets our short term requirements of NUREG-0578 and is acceptable. For the long tenn the upgrading of the panel mounted indicators will be required for fully meeting safety grade requirements.
Our Office of Inspection and Enforcement will verify the installation of the design and document this in an appropriate inspection report.
2.1.8.a POST-ACCIDENT SAMPLING The licensee has performed a design and operational review of the reactor coolant and containment atmosphere sampling. The licensee has provided procedures to provide the capability in 1980 to promptly take a reactor coolant sample during a serious transient or accident and to minimize personnel radiation exposure. The licensee has provided interim procedures to take a containment air sample in 1980 in an area to minimize personnel exposure.
In addition, portable shielding will be available onsite for use in the plant reactor coolant sampling room and analysis facilities to provide a capability to take a sample and to analyze it during an accident. A modified sampling system to minimize personnel exposure will be proposed in the March 31, 1980 shielding report. New reactor coolant system sampling points will be included in this proposal to allow collection of a representative sample without operating the letdown into the makeup and purification system. Any long term changes which are needed for the plant analysis facilities will be proposed in the shielding report.
The licensee has performed a design and operational review of the plant radiological analysis facility. This facility has been moved to an area just off the turbine deck adjacent to the control room to reduce the possible radiation background in this facility during an accident. Proce-dures are available to provide the capability to promptly quantify radio-nuclides in a highly radioactive sample during a serious transient or accident.
The licensee has performed a design and operational review of the plant chemical analysis facility. Procedures are available to provide the capability to promptly quantify certain chemical analyses in a highly radioactive sample during a serious transient or accident.
Based on the above considerations, we conclude that the licensee has met
'the Category "A" requirements. Verification of the procedures for taking
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and analyzing a reactor coolant and containment sample will be performed
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by the Office of Inspection and Enforcement and will be documented in an appropriate inspection report.
2.1.8.b HIGH RANGE EFFLUENT MONITORS The licensee has provided an interim method for quantifying high level noble gas effluent from the Auxiliary Building ventilation line and the hydrogen purge line. These lines and the secondary side steam dump line are the only ones used during a serious transient or accident involving the reactor.- The hydrogen purge is used to control the hydrogen concentration inside containment after an accident. The licensee is studying how to monitor gross radioactivity releases from the plant steam dump valves. The licensee will provide a design and install a radiation monitoring system on this line and provide interim procedures to use,this system to quantify noble gas releases from this,line by March 31, 1980.
The licensee has installed a new iodine / particulate sampling system for the above lines to provide sampling at a more accessible location. The iodine /particulates are collected on cartridges and taken to the plant radiological counting facility for analysis. Procedures have been developed for collecting and analyzing these samples.
Based on the above considerations, we conclude that the licensee has met the Category "A" requirements. Verification of the procedures for quantifying high-level noble gas and iodine / particulate affluents from the plant will be performed by the Office of Inspection and Enforcement and will be documented in an appropriate inspection report.
2.1.8.c IMPROVED IODINE INSTRUMENTATION The licensee has equipment which can be dedicated to analyzing air samples for radioisdine concentrations during an accident. The licensee states that sample collection and counting times can be minimized, in an acci-dent situation, so that a rapid analysis can be made to determine if significant inplant airborne radiciodine concentrations are present. The licensee has also ordered two single channel' analyzers which can be used to promptly and accurately analyze air' samples for airborne radiciodine during an accident. These analyzers will be located onsite: one in the ANO-1/ANO-2 control room area and one in the Technical Support Center where plant personnel will be stationed during an accident. The analyzers will be available by May 1,1980.
Based on the above considerations, we conclude that the licensee has' met the Category "A" requirements. There are no Category "B" requirements.
Verification of the procedures that. state the licensee has equipment dedicated.to analyzing air samples during an accident, that the above equipment is in place in the ANO-1/ANO-2 control room area and in the Technical Support Center, and is periodically checked and calibrated will be performed ' by the Office of Inspection and Enforcement and will be documented in an appropriate inspection report.
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. 2.2.1.a SHIFT SUPER'lISOR RESPONSIBILITIES The licensee has revised plant procedures, as necessary, to set forth the responsibilities of the Shift Supervisor-such that he can provide direct.. command oversight of operations and perform management review of ongoing operations that are important to safety and not be distracted from these important responsibilities by administrative details.
We conclude that the licensee has satisfied the Category "A" requirements to provide revised responsibilities and authority for the Shift Super-visor. Verification of the licensee's procedures will be performed by the Office of Inspection and Enforcement and will be documented by-an appropriate inspection report.
2.2.1.b SHIFT TECHNICAL ADVISOR
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l The licensee has provided an on-shif t technical advisor (STA) to the shift supervisor to fulfill the two required functions of accident assessment and operating experience assessment. The STA should be able to report to the control room within 10 minutes to assist in diagnosing an off-normal event.
For the interim period of 1980, the licensee has provided additional site engineering staff personnel (i.e., degreed engineers) to be the STA. They will serve a 24-hour duty day on a rotating basis and will be onsite at all times during their duty. The same person will be the STA for both ANO-1 and ANO-2.
Sleeping quarters are available onsite and the STA will be available to the control room within 10 minutes of being called. In the l
event of an accident, the designated STA is removed from the chain of command. He then acts only as an advisor to the Shif t Supervisor who is in charge of all actions on the affected unit. The operational experience analysis is the primary responsibility by the on-site Plant Performance group. The designated STAS are responsible for the review and evaluation of plant operational experience and of off-normal events.
For the permanent situation, degreed engineers will be provided with addi-tional training and assigned on shift as STAS. This group of engineers will also perform the operational analysis function.
We conclude that the licensee has satisfied the Category "A" requirements for the shift technical advisor. Verification of the licensee's proce-dures for implementation of this item will be performed by the Office of Inspection and Enforcement and will be documented in an appropriate inspection report.-
2.2.1.c SHIFT AND RELIEF TURNOVER PROCEDURES The licensee has revised plant procedures 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.
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. Based on the above considerations, we conclude that the licensee has satisfied the requirements of Item 2.2.1.c ta provide new procedures.
Verification of the implemented procedures will be performed by the 4
Office of Inspection and Enforcement and will be documented by an
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appropriate. inspection report.
~ 2.2.2.a ' CONTROL ROOM ACCESS -
The licensee has implemented procedures which will limit control room access during an emergency. These procedures establish the authority and responsibility of the person in charge of the control room to control access and establish the line of succession for the person in charge of the control room. We therefore conclude that the licensee has satisfied j
our requirements of this item. Verification of licensee's procedures will J
be performed by the Office of Inspection and Enforcement and will be documented in an apprppriate inspection report.
l 2.2.2.b ON-SITE TECHNICAL SUPPORT CENTER (TSC)
The licensee has established an onsite technical support center on the 4th floor of the Plant Administrative Building which is located south of the turbine building. Direct telephone communication between TSC and the control room, near site Emergency Operations Center and the NRC has been established. Portable airborne and radiation monitors have been provided for the TSC to provide warning and monitoring capability. Access to technical data (plant drawings and records) is available from the plant's records management system. This is located within one floor of the TSC. Plant parameters can be monitored by CRT display in the TSC from the plant computer.
For the long term Category "B" requirements, the licensee has proposed two Technical Support Centers. A Primary Technical Support Center (PTSC) and a Secondary Technical Support Center (STSC). The PTSC will be located in the administrative building as discussed above. The STSC will be located approximately.65 miles-from the plant. It should be noted here, however, that PTSC is not habitable because of shielding problems in the administrative building, while the STSC will be habitable when completed. This proposal is under NRC consideration and will be reviewed as a Category "B" item.
Based on our review of licensee's submittal and our site visit, we have concluded that TSC at the ANO-1 satisfies our short term requirements set forth in NUREG-0578. Verification of the licensee's procedures for the short term TSC will be performed by the Office of Inspection and 4
Enforcement and will be documented in an appropriate inspection i
report.
2.2.2.c OPERATIONAL SUPPORT CENTER The licensee has established an on-site Operations Support Center (OSC) located in the vicinity of the on-site technical support center. Opera-
' tions Support Personnel will be located in the OSC for response to the
. control room and/or TSC needs. Telephone communications and the station page system presently exist with the control room and the TSC.
Based on our review of licensee's submittal and our site visit, we con-clude that the liceasee has satisfied our requirements set forth in NUREG-0578. Verification of licensee's procedures to cover this center will be performed by the Office of Inspection and Enforcement and will be documented in an appropriate inspection report.
NRR REACTOR COOLANT SYSTEM HIGH POINT VENTS The licensee has proposed installation of remotely operated vents for the top of the pressurizer and for the high point of each hot leg. Each vent point will have parallel redundant paths with two valves in series. A single safety grade vent path will be added to the pressurizer with the PORV and the block valve providing a redundant vent path. The hot leg vents will be safety grade.
The proposed design does not include a reactor vessel head vent as required by the current Lessons Learned position.
Unless it can be shown that the hot leg vents provide a sufficient substitute for a head vent, it is the staff's position that a reactor vessel head vent be installed.
This proposal is currently under review by the NRC for determination of the adequacy of the bases for not installing the reactor head vent.
The B&W Owners Group will assist in resolving this item. Our findings will be presented in the report on Category "B" requirements.
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REFERENCES 1.
Letter, APLC (Cavanaugh) to NRC/NRR dated October 17, 1979.
2.
Letter, APLC (Cavanaugh) to NRC/NRR, dated November 20, 1979.
3.
Letter,'APLC (Cavanaugh) to NRC/NRR dated December 5, 1979.
4.
Letter, APLC (Cavanaugh).to NRC/NRR, dated December 17, 1979.
5.
Letter, APLC (Cavanaugh) to NRC/NRR dated January 17, 1980.
6.
Letter, APLC (Trimble) to NRC/NRR dated January 18, 1980 7.
Letter, APLC (Trimble) to NRC/NRR dated January 29, 1980.
8.
Letter, APLC (Trimble) to NRC/NRR dated January 31, 1980.
9.
Letter, APLC (Trimble) to NRC (E/NRR) dated February 19, 1980.
i 10.
Letter, APLC (Trimble) to NRR (E/NRR) dated February 29, 1980.
- 11. Letter, NRC (Eisenhut) to ALL OPERATING NUCLEAR POWER PLANTS, dated September 13, 1979.
12.
Letter, NRC (Denton) to ALL OPERATING NUCLEAR POWER PLANTS, dated October 30, 1979.
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