ML19351G228
| ML19351G228 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/09/1981 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19351G229 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.3, TASK-1.A.3.1, TASK-2.B.3, TASK-2.B.4, TASK-TM NUDOCS 8102230318 | |
| Download: ML19351G228 (60) | |
Text
{{#Wiki_filter:e- . Alibim3 Power Co npany 600 Nortn 18tn street Post Offico Box 2641 Birmingnam. Alabama 35291 Teisonone 205 250-1000 F. L CLAYTcN, JR. m sen,cr vice pres.oent Alabama Power me soutrem entre sniem February 9, 1981 Docket No. 50-364 50-348 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. A. T. Schwencer JOSEPH M. FARLEY NUCLEAR PLANT - UNIT 1 & 2 CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS (NUREG-0737) Gentlemen: ( Based on discussions with NRC Staff personnel during the week of February 2, 1981, additional information was requested with regard to several items in the APC0 original submittal of January 14, 1981. Alabama Pcwer Company submits the enclosed revised responses on each of these issues. These positions have been discussed with the NRC ! Staff and are considered by Alabama Power Company to satisfy the Staff's questions. If you have any questions, please advise. 30 1 i Yours very tr' , y y / / (, - ADb! kL F L. Clayton, Jr. IR D6 I O
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I.A.l.3 SHIFT MANNING Previous Response _ In previous letters dated June 26, 1980 and January 14, 1981 for Unit 1 and June 20, 1980, August 7, 1980, August 14, 1980, September 8, 1980 and January 14, 1981 for Unit 2 Alabama Power Company previously submitted commitments and documentation of actions taken at the Farley Nuclear Plant. Clarification Response Alabama Power Company's letter dated February 5, 1981 responsed to this item for Unit 2. 1 e es* W e - y +y++--- -r v-y W - wy- -,qy tw y- y- w r ry w , ev- ---iw--w-w y y - y y- - -
t 10 1.A.3.1 REVISED SCOPE AND CRITERIA FOR LICENSING EXAMS Previous Resconse By letters to the NRC dated June 20, 1980, July 15, 1980, and August 1, 1980 related to Unit 2 and the letter of June 26, 1980 for Unit 1, Alabama Power Company addressed this item for the Farley Nuclear Plant. Clarification Response Conditional upon availability of a simulator which is adequately close to FNP to avoid a degradation of operator skills due to significantly contrasting parameters and procedures, Alabama Power will make arrangements as necessary to give examinations on a non-plant specific simulator. Specifics for this simulator examination requirement will be contained in individual operation license applications for examination after October 1981. l me s l
, , 19 I.C.7 NSSS VENDOR REVIEW OF PROCEDURES (UNIT 2 ONLY)
Previous Response Alabama Power Company by letters of June 17,1980, July 17,1980,
-- September 2,1980, September 11, 1980, October 13, 1980, and November 6, 1980 has responded to the NRC on this issue.
Clarification Response The review of the low-power physics tests, applicable emergency operating procedures, and power ascension tests has been completed by [ the NSSS vendor (Westinghouse). I 4 i l w- - . ,. , ,.,g--,...w -m. y ,.
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26 II.B.2 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES / SYSTEMS MiICH MAY BE USED IN POST-ACCIDENT { OPERATIONS Previous Resconse By letters of June 20, 1980 and August 1, ~1980, for Unit 2 and October 24, 1979, November 21, 1979, December 31, 1979, March 14, 1980, and May 5,1980 for Unit 1
- Alabama Power Company dr.:umented ccmmitments and actions taken for the Farley Nuclear Plant related to this #, tem.
Clarification Resoonse A design review for the Farley Plant - Units 1 and 2 was conducted by Bechtel Power Corporation, using the TID source terms and the 10 CFR 20 and GCC19, 60-64 of Appendix A to 10 CFR 50, dose criteria. This shielding design review considered several classifications of systems which included recirculation systems, systems which are extensions of the containment atmosphere, portions of the liquid sampling system, and portions of the letdcwn system. The liquid and gaseous radwaste systems were not included in these analyses.
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The gaseous system was eliminated since the reactor ' vessel head vent would be used for degassing operations rather than the VCT. The leak reduction program instituted at Farley Nuclear Plant, and venting of the reactor by the reactor vessel head vent and/or PORVs rather than the letdcwn system and VCT, minimized the need for the liquid waste processing system and therefore it was not considered. The high activity radioactive lab and counting roo:n for the affecte unit was not included among..those ~ areas where access is considered vital aft'er an accident since for two unit operation these areas in'th'e unaffected unit will be utilized for post-accident analyses.
~_ ' Access areas with their corresponding post-accident occupancy time for Units .
.1 and 2 are listed below: -
Onit 1 Area Occupancy Period Control Room 24 hr/ day Health Physics Area 24 hr/ day
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Primary Access Point % 24 hr/ day Passageway to Unit 2 1 hr/ day Hallway 409 1 hr/ day ( Electrical Penetratien Recms ** 1 hr (approximately 1 hr. after accident)
"* Design change to eliminate occupancy requirement is being considered, Zone maps will be updated as necessary. -- .
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U.S.2 (Continued) Unit 1 Area Occupancy Period Hallway 322 (Outside Sample Room) *** { 1 hr/ day Gas Analysis Room *** 1 hr/ day Cable Spreading Room 1/2 hr.* Filter Rooms 2 hr/ day *
- SwitchgearRooms(Eley. 121') 1/2 hr.*
Hot Shutdown Panel 24 hr/ day
- CCW Pump Room 1/2 hr.*
Corridor 161 1/2 hr.* RHR Heat Exchanger Room 1/2 hr.* Stairway No.1 Transit to elevations at west side of aux. bldg. Stairway No. 2 *** Trans-it to elevations 77'/83' Stairway No. 8 Transit to elevations at { north and east sides of aux. bldg.
- Infrequent-brief access to these areas may be required post-accident.
*** Complete evaluation of these areas is based on resolution of shielding associated with the electrical penetration rooms. .
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- Unit 2 -
l Area Occupancy Period [ . Control Room 24 hr/ day Technical Support Center 24 hr/ day Health Physics Area 24 hr/ day Primary Access Point % 24 hr/ day Passageway to Unit 1 (2402) I hr/ day Hallway 2409 1 hr/ day ( Electrical Penetration Rooms ** 1 hr (approximately 1 hr. after accident)
** Design change to eliminate cccupancy requirement is being considered.
Zone maps will be updated as necessary. -:
11.3.2 (Continued) 28 Unit 2 Area Occupancy Period Hallway 2322 (Outside Sample Room for Liquid Sample) *** 1 hr/ day Spectro Photometer *** 1 hr/ day _ Cable Spreading Room 1/2 hr
- Filter Roo3s 2 hr/ day
- Switchgear Rooms (Elev.121') 1/2 hr* .
' Hot Shutdown Fanel 24 hr/ day
- CCW Pump Room 1/2 hr
- Corridor 2161 1/2 hr
- RHR Heat Exchanger Room 1/2 hr
- Stairaay No.1 Transit to elevation at west side of aux. b1dg.
Stairway No. 2 *** ' Transit to elevations _f 77'/83'. Stairway No. 8 Transit to elevations at north and east sides of aux. b1dg.
- Infrequent-brief access to these areas may be required post-accident.
*** Complete evaluation of these areas is based on resolution of shielding
, . _ associated with the electrical penetration rooms. ,
. Each of these areas has been analyzed to detennine the dose-rates following an accident. .
l l The Primary Access Point was not initially considered but was later designated a I-A area for direct radiation. The secu- enter will be included in the ' l radiatior tone maps. The main control room the technical support center were considered as areas requiring continuous occ yancy. As a result of these studies the following shielding modifications are listed for Units 1 and 2: ( 1. /.cd shielding to the portion of line 3" GCC-12 which is exposed in the area of the Seal Injection Filter valve station to reduce the dose rate in this area.
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29 II.B.2 (Continued)
- 2. Place temporary shielding at the containment radiation monitor to reduce dose rate in the corridor (RE-011, 012).
- 3. Re-route the RCS sample discharge line so that spent samples are returned by a more direct route to the VCT without entering the letdown line.
_. 4. Add additional shielding outside the auxiliary personnel hatch to minimize potential effects at the Elevation 155' for the access control area and Technical Support Center.
- 5. The requirement to add shielding at hydrogen analyzers; around lines to hydrogen analyzers; and around reactor coolant and containment air sample lines was due to the fact personnel access was required within one hour after the accident. Further analysis has shown that design modifications which remove the necessity of operator action from this area was more practical than the shielding modifications. These modifications include relocation of eight breakers for normally locked out valves in the ECCS ficw path.
The following is a discussion of the computer progra=s used for these analyses:
- 1. Source term concentrations in uCi/cc for each isotope along
__ with the volumetric source strengths in Mev/cc/sec. were calculated using the NUCLYD computer program. This program calculates values of the specific activity and volumetric source strength at any given decay time for a given mixture or isotopes. The program also provides an integral energy release for that given mixture of isotopes from t = 0 to any specified time. This computer program is analogous to ORIGEN. , _ 2. Dose rates were calculated with the CYLSO computer program. , This program uses the Rockwell Point Kernel theery for one dimensional cylindrical volumetric sources. Self attenuation in the source as well as the shielding effects of various ccnstruction materials such as stell, lead, concrete and water are considered in the code. The code output is in terms of dose rate vs. distance, for various piping diameters l and shielding configurations.
- 3. ThiscomputerprogramigsimilartotheSDCcode.
In addition to the above study, the effects of radiation on equipment are being considered as part of the IE Bulletin 79-OlB and NUREG-0588 review. Source terms for LOCA events in which the primary system may not depressurize will be addressed during the above review. The shielding desige evaluation is a complex iterative process. All modifications listed above and modifications required to resolve the outstanding design issues will be completed for Units 1 and 2 by January 1, 1982. e - -- y. 9 ,+yy , y w - , . -
30 II.B.3 POST-ACCIDENT SAf1PLING CAPABILITY C Previous Resconse . By letters to the NRC of June 20, 1980 and August 1,1980 for Unit 2 and of October 24, 1979, November 21,1979 and December 21,1979 for Unit 1, Alabama Power Company provided information related to post-accident sampling .
.- capability. .
Clarification Resconse t As describcd in previous submittals, a post accident sample of reactor ' coolant is drawn from the same sample line used for the gross failed fuel - detector. The gross failed fuel detector is normally on recirculation at all times. Any deposition or plating in these lines will be in equilibrium. This system is isolated with the containment isolation signal but can be , reinitiated without startup of the letdown system. Recirculation is acccmplished by routing the sample return to the volume control tank. Tne sample line is 3/8" stainless steel tubing all the way from its origin in containment, therefore, loss of coolant would be limited to the flow through this 3/8" line in case of sample line rupture. The flow is normally limited to about 0.6 gallons per minute and the length of the sample line has been adjusted so that a representative sample may be obtained approximately one minute after initiating sample flow. The sample . f " return line has been rerouted to minimize exposure and line length. The sample system was designed to produce a predetermined sample volume to prevent overflow and spillage. The sample system is ccmpletely closed and pressurized to the point where the sample is extracted. The sample is degassed and depressurized at this point. The gasses taken from the sample - are used as the sample for H2 and 02 in the RCS as well as for The Noble gas analysis. The gasses excaping from the system during the sampling , evolution are drawn into the radwaste HVAC system. l -- The post accident containment atmosphere sample is collected through ' .- the sanie sample lines that are used for the normal containment atT.asphere monitors REll and RE12. These lines have been designed according to ANSI 13.1
- with no sharp bends in order to provide smooth laminar flow, to prevent impaction and to be as short as possible, according to the required location .
of sampling equipment, to minimize plateout. A plateout study, however, is-1 - in progress and a plateout factor will be included in the calculation of ' radioactivity content of the containment atmosphere if the results of that . study indicate that plateout will be significant. 'The inlet line is located at about the 134' elevation in a protiW;ed area above a ventilation duct and j belcw a steel grating with the actual inlet turned down so that no debris will
- be picked up which could block the sample line. The sanple line is nade of la l I.D. rigid stainless steel tubing with isolation valves which can be closed in case of line rupture. The sample system includes a pump which returns the l residue to containment. Any gases released during the sampling evolution will be collected in the radwaste HVAC system.
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31 II.B.3 (continued) 4 (' '
'Tse post-accident sampling system and the containment atmosphere sampling
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system provide collection of small aliquots of the sampled media which will be shielded for transportation to the laboratory for analysis. Special , procedures have been developed for analysis of highly radioactive samoles which-include the use of lead glass windcws and manipulating apparatus that T_ will insure that no analyst will receive exposures exceeding 3 and 18 3/4 rems to the whole body and extremities respectively. Both of the above systems will enable sampling and sample analysis within three hours. The - above analysis capability includes the ability to detect 1) radionuclides ... 4 in the reactor coolant system that ma 2) 4 hydrogen gases, 3) chlorides
- and boren 4) y beconcentration indicators of of core damage, liquids. The -
ability of dilute samples is also provided. The onsite liquid sample analysis program has the capability to permit sensitivity measurements of nuclide concentrations in the range frem approximately 1 vei/g to 10ci/g. . , The sample analysis will provide results within a factor of two error. Emergency implementing procedures describe hcw adequate information will be provided to the operator to describe the post accident reactor coolant system radiological and chemical status. The sampling system for reactor ecolant and containment atmosphere , for Unit 2 is installed, initial testing completed, and is anticipated to , l j[ , be fully operational by March 10,1981,cn no later than exceeding 5% power. , The sampling system for reactor coolant and centainment at=csphere is ' installed and operational in Unit 1. i i
- Due to the location of the Farley Nuclear Plant, arrangements have been ~
- c . made to perform offsite analyses within four days following sampling. _
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32 4 II.B.4 TRAINING FOR MITIGATING CORE DAMAGE Previous Resoonse By letters dated July 29, 1980, August 6, 1980, September 24, 1980, September 25, 1980, and August 1,1980 for Unit 2, Alabama Pcwer Company documented commitments and actions taken for the Farley Nuclear Plant. Clarification Resoonse This training will be completed for applicable perse nel for Unit 1 by October 1,1981 and prior to Unit 2 operation above 5 percent power. The following is a description of the subject training for Instrumentation and Control and Chemistry and Health Physics personnel. INSTRUMENTATION AND CONTROL TRAINING FOR MITIGATING CORE DAFAGE Systems
,o-- A. ESF Features B. ECCS Systems C. ECCS Systems Response to Accident Conditions Procedures A. Emergency Operating Procedures B. Emergency Plan Implementing Procedures Instrumentation
, A. Excore NIS l B. Incore NIS C. Incore Thermocouples ( D. Post Accident Hydrogen Analyzers E. Containment Environment Effects on Vital Instrumentation F. Alternate Methods for Pressure, FigTemperature & Level Determination G. I&C Calibration and Surveillance Proradures 5 t
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- 32a CHEMISTRY AND HEALTH PHYSICS TRAINING FOR MITIGATING CORE DAMAGE Post Accident Primary Chemistry A. Baseline Plant and Assumptions B. Normal Plant Chemistry C. In-Core Release and Escape Mechanisms D. Mechanisms for Extensive Core Damage E. Radiological Hazards of Sampling Mitigating Core Damage - Response of Process and Area Radiation Monitors A. Pcstulated Plant and Accident Conditions B. Concentrations of Fission Products in Containment
& Isotopic Break Down C. Effects on Area Mcnitnrs of Containment Radiogas
& Gas Generation D. Containment Centamination Levels
",__. E. Estimate of in-Containment Conditions from Exterior Measurements
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II.D.1 PERFORMANCE TESTING OF BOILING WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES \ Previous Response By previous response dated July 17, 1980, July 23, 1980 and August 1, 1980 1 - for Unit 2 and October 24, 1979 and December 31, 1979 for Unit 1, Alabama l Power Company described commitments and actions taken for the Farley Nuclear Plant. Clarification Response > As indicated in the Dacember 15, 1980 letter from R. C. Youngdalh (EPRI) to D. G. Eisenhut (NRC), the present EPRI program does not formally include the testing of block valves. However, a number of block valves have been tested i at the Marshall Steam Station Test Facility, and a preliminary scope and cost estimate study for a block valve test program has been completed by the EPRI staff. A detailed block valve test program will not be resolved until after July 1, 1981. Alabama Power Company commits to participating in such an EPRI program and will supply further details as they become available; however, in any event, Alabama Power Company will provide by July 1, 1981 a program , description for ensuring block valve qualification by July 1,1982. 1
,__. While Alabama Pcwer Company does not support additional ATWS valve testing until regulatory issues are resolved, the major test facility for the EPRI program was designed to provide the potential for addif tonal valve testir.g at higher pressures for ATWS conditions. ,
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. 35 II.E.1.2 AUXILIARYFEEDWATERSYSTEMAUTOMA[!CINTTTATIONANDFLOWINDIC f
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By letters of August 1,1980, April 1,1980, March 3,1980, April 21,1980, liay 27, 1980, June 20, 1980 and August 6,1980, for Unit 2 and July 29, 1980, . October 24, 1970, December 14, 1979, December 31, 1979, November 20,1979, and December 4,1980, for Unit 1 Alabama Power Company provided response to this item related to the Farley Nuclear Plant.
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Clarification Resconse
*' Initiation
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The autcmatic initiation signals and circuits associated with the auxiliary feedwater system meet safety grade requirements (i.e., IEEE 279-1971, seismic and environmental qualification) with the exception of the isolation circuitry . associated with motor-driven auxiliary feedwater pump auto-start on main feedwater pump trip. The non-safety grade portion of these circuits will be ' modified to include isolation frcm the safety grade.portiens thrcugh isolation devices in accordance with the requirements of IEEE 279-1971. It is intended that these ecdifications will be completed prior to startup following the current refueling outage but no later than July 1,1981, on Unit 1 and prior-to exceeding 5 percent pcwer en Unit 2. '
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Indication Auxiliary feedwater injection lines to each steam generator are provided with safety-grade flow indication for both units. This flow indication is on the main control board and is powered from the plant emergency power. These flow instrument loops are testable. Redundancy requirements are met by qualified steam generator level instrumentation (safety-grade). A description of the presently installed equipment is provided below. '
.f Local and control room indication'of auxiliary feedwater flow to each of the - , steam generators is provided by flow orifices in each auxiliary feedwater -
supply line, located just upstream of the auxiliary feedwater stop check valves. The auxiliary feedwater flow indication is backed up by three redundant safety- . grade wide range steam generator level channels..per steam generator which have. control room readouts. . Testing of this equipment is conducted in accordance with the Farley Nuclear Plant Technical Specifications. The abiliary feedwater. flow indication channels and steam generator wide range level channels are calibrated every 18 months. The steam generator narrow range level channels are functionally - tested every 31 days and calibrated every 18 months. The displays and controls asscciated with auxiliary feedwater system flowrate indication were considered as part of the human-factor analysis conducted in ( response to NUREG-0737, Item I.D.1. ,
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r I 37 II.E.1.2 (Continued) h
- The auxiliary feedwater flow instrumentation channels and the steam generator narrow range channels receive their power from the Class 1E vital instrument buses. The steam generator wide range channels also receive their power from vital instrument buses.
The additional Unit 2 condensate storage tank level transmitters have been installed. 1
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1 e e II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION ( Previous Response By letters dated August 1, 1980, August 19, 1980, June 20, 1980 and July 24,
- 1980, for Unit 2 and October 24, 1979, November 21, 1979, December 31, 1979 and March 14, 1980, Alabama Power Company described commitments and actions j- taken for the Farley Nuclear Plant.
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j Clarification Response Noble Gas Effluent Monitor - 2' A. Vent Stack Monitor 9 Alabama Power Company will install for both units an Eberime Sping 4 sampler to monitor noble gases in the plant vent stack. This sampler has a range of 10-7 to 105 uct/cc using multiple detectors. The monitor draws a sample from the vent stack to a monitor unit located in the mechanical equipment room at elevation 175 of the auxiliary building. The readout for this unit is located in the main control room. An auxiliary readout is located in the low activity counting laboratory. 4 The noble gas measurement is perfomed by several detectors viewing a sample 1 volume. The low and medium range detectors view the same sample volume located in the SA-13 sampler assembly. The high range detector views the sample volume located in the SA-9 sampler assembly. (1) LOW RANGE NOBLE GAS: The gas chamber is monitored by a BETA _. scintillation detector (Eberline Model RDA-3A). Background correction for this channel is derived from the garma background .'. detector, an energy-compensated GM detector (Eberline Model l '10450-828). Since the external (ambient) gama radiation has a - measurable effect on the BETA measurement (particulate and gas),
- . the gama background channel is used as a source of subtraction i for both the gas measurement and the particulate measurement.
l (2) MEDIUM RANGE NOBLE GAS: An energy-compensated GM detector monitors ' the gas volume for the mediugange noble gas measurement, with its output proportional to the ganta content of the sample. An additional identical detector is provided in the sampler shield as a measure of ! the external background at the sampler; this is the background l detector. Thus the effects of a fluctuating external background on the medium range gas channel are nullified by measuring and subtracting the background, i l L
II.F.1(Continued) 43 Noble Gas Effluent' Monitor (ContinJed) ( (3) HIGH RANGE NOBLE GAS: An energy-compensated GM detector monitors the gas volume of a section of 1" stainless steel tubing for the high range noble gas measurement. Its output is proportional to the gama content of the sample. ~- An area monitor radiation detector assembly (Eberline Model DAl-1-CC) is mounted on the Sping 4 and provides a measure of the gama field at the instrument. This detector is an energy-compensated GM tube and is calibrated in radiation dose rate. Calibration is by use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage. The Eberline Sping 4 monitor is
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capable of functioning both during and following an accident. Frequent filter replacement will ensure operability of the monitor's electronics after an accident. The monitor's accuracy is t 2% of span . - The following list sunmarized by channel number and type which calibration sources are provided. - CHANNEL CHECK SOURCE Number Tyce Content Isotoce 1 Beta Particulate 30 microcuries 137Cs
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2 Alpha Particulate 3 Iodine (Ga::na) 0.5 micurcurie 133Ba 4 Iodine Subtraction (Gama) 5 Beta Gas (Low Range Noble Gas) 30 microcuries 137Cs 6 . Gama Area 0.5 microcurie 90s 90y
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7 Gama Gas (Medium Range Noble Gas) , 8 Gama Background 9 Gama Gas (High Range Noble Gas) .05 microcurie 90s 90y The plant vent noble gas concentrata(on in uCi/ml is detennined by sampling and/or by obtaining a value frcm thF plant vent stack high range monitor. The plant vent flow rate is determined by the~ number of operating auxill y building exhaust fans. The release rate in curies per second is determined by the following equation: Release rate (Ci/sec) = Concentration (uCi/ml)X flew rate (cfa) X conversion factor { so w- , , .- - -,.w.-- - w-- ,
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4 e e 44 If.F.1(Continued) - Noble Gas Effluent Monitor (Continued) ( The above method to detemine noble gas release rate is described in emergency implementating procedures. During emergencies the release rate is calculated periodically as directed by the Emergency Director to determine if the accident classification should be upgraded. The monitors have been environmentally qualified by the vendor for the environment in which it is located. ! B. Main Condenser Air Removal Monitor (SJAE) The main condenser air removal exhause systems for Units 1 and 2 are , i monitored using the existing monitor (described in the FSAR) on the steam jet air ejector exhaust for the normal range. of radioactivity. The accident range of radioactivity will be monitored for Units 1 and 2 by intemediate i and high range detectors with overlapping ranges and located at the comon vent duct for the turbine building. The accident monitor consist of 2 Eberline detectors and readouts. The intemediate range detector will be model Dal-1CS with an EDl-1 readout module with a range of indication of 0.1 to 100 mR/hr. The high range detector is a model Dal-4CS with an ECl-20 readout module with a range of 10 mR/hr. to 1,000 R/hr. The relationship between mR/hr. and uCi/cc will be established for the noble gas isotopes present during an accident. The range of the accident monitors in uCi/cc is from 10-5 to 103 with the normal range' monitor measuring concentrations doun to 10-6 uCi/cc. This is the required range for the case where the SJAE exhaust is combined with turbine building ventilation exhaust.
- - C The readout modules will be located in the control room and will provide continuous indication. The accident detectors will be shielded from back-ground radiation with 6 inches of lead. Calibration is by use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage.
C. Steam Generator Atmospheric Relief and Safety Valve Monitors, The discharge from steam generator safety relief valves and atmospheric
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dump valves for Units 1 and 2 will be monitored by measuring the radiation . levels from these steam plumes. There will be four Eberline model DAl-4CS
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detectors per unit mounted on the main steam roof with a range of 10 mR/hr. to 1,000 R/hr. The relationship between mR/hr. and pCi/cc has been established for the noble gas isotopes present during an accident. The range of the monitors in uCi/cc will more than cover the required range from 10-1 to 103 for cases wi?h just the PORC open to cases with the PORY and all safties open. Each d3tector will be connected to an Eberline ECl-20 readout module in the control room - providing continuous indication. Since the safety re1% valve and atmospheric dump valve discharges are grouped together for each of the three steam generators, one detector will be used to monitor the combined effluent steam plume from each steam generator. The fourth detector is used to monitor the plume from the steam driven auxiliary feedwater pump turbine exhaust. Each detector is collimated and background shielded with 7.5 inches of lead. Calibration is by use of an external calibration source and is performed upon installation C and at intervals not exceeding each refueling outage.
45 II.F.1 (Continued) (Noble Gas Effluent Monitor (Continued) ( D. Desian and Installation Schedule for Nobles Gas Effluent Monitors The noble gas effluent monitors will be powered frca a vital instrument bus. Procedures will be developed for use, calibration o." the system, and dissemination of release rate information. The Sping-4 for both
~ units is onsite hardware to support installation of the main condenser
_ air removal monitors and the steam generator atmospheric relief and safety valve monitors are onsite or in the process of being shipped. The installation of the vent stack monitor for Unit 2 is scheduled for March 10,1981, or prior to exceeding 5 percent power. This instrument, is currently scheduled for installation in Unit 1 for prior to the end of the current refueling outage but not later than January 1,1982. The original Alabama Pcwer Company position was to monitor the main condenser air removal exhaust and the discharge frem the steam generator safety relief valves and atmospheric relief valves with a portable gamma survey instrument. Alabama Pcwer Company, however, finalized the above position based on NRC questions during the latter part of 1980. Based en the current material availability and status of the complex shieldinc design required, installation fcr both units is scheduled for completion by January 1, 1982. Alabama Pcwer Ccepany purchased the best available monitors upon finalization of this position. In order to ensure accurate reading of each of these monitors, a cceplex shielding design is required to discriminate actual readings frca backgrcund including containment shine. O m C
II.F.1 (Continued) 46 Samoling and Analysis of Plant Effluents k Alabama power Co1pany has the capability to provide continuous sampling of plant gaseous erfluent for post accident releases of radioactive iodine and particulates at the plant vent and the condenser air removal system. The sampling method involves passing the effluent gases through a filter assembly and transporting the filter to a counting roca for analysis. The sampling system has the following capabilities: (1) Effective iodine absorption of greater than 90% for all fonns of gaseous iodine. (2) Greater than 90% retention of particulates for 0.3 micron diameter particulates. (3) Design intent meets sampling requirements of ANSI N 13.1-1969. (4) Continuous collection whenever exhaust flow occurs. (5) Analytical facilities and procedures censidered the design basis sample. (6) Shielding factors were considered in the design. I On-site laboratory capability exists to analyze or measure these samples. The sampling system design is suc> that plant personnel can remove samples, replace sampling media, and transport the samples to the on-site analysis facility with radiation exposures that are not in excess of the GDC 19 criteria of 5 rem whole body and 75 rem to the extremities during the duration of the accident assuming the design basis shielding envelope of NUREG-0737. The Eberline Sping 4', which samples vent stack effluents, uses an isokinetic nozzle in the stack to draw its sample into its filter system and the flow rate can be adjusted at the pumping unit to attain a sample velocity that will match stack flow rates. There are presently two exhaust fans that determine effluent velocities. In addition, there will be a Victoreen vacuum pump with i charcoal filters that will allow the Chemistry and Health physics Group to draw 15 minute iodine and particulate samples to be analyzed in the laboratory. This pump has bypass lines that allow % awing an isokinetic sample ' by passing portions of the sample back to the stack.- } The steam jet air ejector sample point is located on the vertical section of the turbine building exhaust ventilation duct. Locating the sample point on the vertical section of the exhaust duct ensures that the absorber material is not degraded with entrapped water.
~ - .l 47 II.F.1 (Continued)
Sampling and Analysis of Plant Effluents (Continued) The primary sampling system for the vent stack (Sping-4) is scheduled to be installed by March 10, 1981, but prior to exceeding 5 percent power for Unit 2 and prior to return to power in Unit 1 following the current refueling outage but no later than January 1,1982, to provide indication (pC1/ml) in the main control room and the counting room. The sampling system for the condenser air removal system is scheduled to be installed by March 10, 1981, bat prior to exceeding 5 percent power for Unit 2 and prior to return to power in Unit 1 following the current refueling outage, but no later than January 1,1982. l 8 k 5 i. 6 i
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. TI.F.1 (Continued) 48 c
Containment Pressure Monitor h j ( The present containment pressure indication provides continucus redadant
.< indication in the main control reca and has an indication range of -5 psig
!j to 60 psig. Additional =cnitoring capability with control reca indication
, having a range of 0 to 210 psig is scheduled to be installed for Unit 1 by
, return to power after the current refueling outage but no later than January 1, i 1982, and for Unit 2 by March 10, 1981, but no later than exceeding 5 percent '
power. Continuous display and recording of the containment pressure is provided in the control rocm. The indication accuracy of both the wide and narrew range instruments is !3.5% with a response time of less than 180 milliseconds for a 10% to 90% step function change in pressure. The environmental qualfication for these items are being addressed as a part of Alabama Power Company's response to I.E. Sulletin 79-013 and N'JREG-0528. l e I l e m
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k9-II.F.1 (Continued) (- Containment Water Level Mcnitor The Farley Nuclear Plant present design has two wide range containment (ECCS sump) water level detectors. These detectors provide indication in the main control room that meets the wide range requirements as specified in the various clarification letters. These level transmitters and associated readout are safety grade and measure volumes up to and above 600,000 gallons. In addition, a narrow range containment (reactor vessel cavity sump) level system meeting the various clarification letters is scheduled to be installed for Unit 1 by return to power after completion of the current refueling outage but no later than January 1, 1982. The schedule for Unit 2 installation is > March 10, 1981, or no later than exceeding 5 percent power. The accuracy of the narrew range level instrumentation is + 1/2 inch with an instantaneous response time. Qualification will be addressed in Alabama Power. Company's response to IE Bulletin 79-01B and NUREG-0588. p W e C 6
I (< II.F.1 (Continued) Containment Hydrogen Monitor t Two independent, redundant systems for containment hydrogen monitoring are __ provided for Units 1 and 2. The design of these systems meets the requirements for safety-related protective systems as defined by IEEE 279-1971. The output signal of the analyzers are indicated at the analyzer panel location and are alarmed and recorded in the main control room. Each system is supplied electrical power from an independent and redundant Class lE Power Supply. . The system meets the single failure criteria and remains operable under the postulated accident. Any single failure in one hydrogen monitoring system does act affect its redundant and independent counterpart. The accuracy of the hydrogen monitor is +2% of span with a response time of 0.45 minutes. The range of indication Is 0-10%. Qualificatan requirements are being addressed in Alabama Power Company's response to I.E. Bulletin 79-013 and NUREG-0588. The indication and recording of hydrojen concentration will be initiated as required by energency procedures in less than one hour after a safety injection initiation. e j=. e 5 e
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51 II.F.1 (Continued) ( Containment High-Rance Radiation Monitor Alabama Power Company has ordered redundant Victoreen Model 875 Radiat1ori Detection Systems to meet the requirements for a high containment radiation monitor. Each system consists of an ion chamber detector, readout panel, and _ interconnecting cables. The monitors will be located inside containment about _ six feet above the operating deck and approximately 900 apart. These locations ensure the monitors are not protected by massive shielding and that they will provide a reasonable assessment of area radiation conditions inside the containment during and following an accident. . (a) Each detector is designed to measure ga:=a radiation. (b) The range of each detector is 1 R/hr. to 107 R/hr. for photon 4 radiation. , (c) The energy response is -15% to 80 key and 8% from 100 key to l 3 Mev. - (d) The calibration frequency will be at a maximum interval of 18 mcnths. Presently it will be necessary to return the monitors to the vendor for calibration. (e) The containment high radiation monitors are being installed and should
, :__C be operational for Unit 2 by March 10, 1981, or prior to exceeding 5 percent power. Such monitors are being installed and should be operational prior to return to power following the current refueling cu bge for Nit * '"+ no later than January 1 1982.
Victoreen has ccmpleted the prelimleiary qua' lit 1uuan review and is near completion for the fir.al qualification program. The radiation monitors satisfy the requirewits of the vendor qualification program. The only
. remaining component to be qualified is the electrical connection between the - e
' power cable and the radiation monitor. Victoreen is currently in the process ',
of qualifying this component. Alabama Power Company will update the NRC on
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the qualification program as information beccmes available. - Capability exists for on-site calibration of the radiation menitor to ICR/hr. Calibration above 10R/hr. will be completed by utilizing an electronic signal. As part of the vendor testing program, Victoreen has stated that at least one point per decade of the range between 1 R/hr. and 103 R/hr. had the calibration certified. '85 - C
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II.F.2 INADEQUATE CORE COOLING INSTRUMENTATION Previous Resoonse: By letters of December 31, 1979, for Unit 1, and of June 20, 1980, July 17, 1980, July 24, 1980, August 1, 1980, August 6, 1980, and August 19, 1980, for Unit 2, Alabama Power Company provided response to this item for the Farley Nuclear Plant. Clarification Response: The information below is the response to the clarification items listed in NUREG-0737 and NUREG-Oll7, Supplement 4, concerning Inadequate Core Cool-ing Instrumentation. 4 A. Descriotion of the Proposed Final System: l The system for monitoring inadequate core cooling (ICC) for the Farley Nuclear Plant consists of core subecoling monitor and incore thermocouples and the reactor vessel water level instrumentation to de-tect inadequate core cooling conditions. In response to the need to develop additional instrumentation, Alabama Power Company, in coordina-tion with EPRI, has undertaken a vessel level measurement program for the Farley Nuclear Plant to demonstrate the capability of a non-invasive pro-totype system on an experimental basis. EPRI will continue to fund the analytical investigations associated with the prototype demonstration pro-gram and will assist APC in establishing the necessary test programs and evaluation of the collected data to determine the feasibility of such a system. Initial testing on a temporary special test setup for Farley Unit 1 was performed during the period of November 10-14, 1980. The objective of the test was to obtain definitive measurements of the relationship between neutron count rate above the reactor vessel water level. A second objec-tive was to discover which variables such as core reactivity, shielding, and neutron background affected this relationship. General Description of the Non-Invasive Water Level Measurement System A description of the non-invasive water level measurement system was submitted to the NRC for its r$gjew and approval by Alabama Power Company letter dated August 6,1980. This is a prototype system unique to the Farley Nuclear Plant, and its use has been supported by the NRC in issuance of Supplement 4 to the Safety Evaluation Report associated with the low power operating license for Farley Unit 2.
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The non-invasive ofexternallymountedgBFter3 level measurement neutron system detectors above andconsists of reactor below the a set vessel. The principle used is the detection of photoneutrons from the reaction of high energy gannas with the deuterium impurity present in the reactor coolant system. A simplified diagram of the system is shown in Figure 1. sists of eight Each detector (8) 2-inch set above diameter, or active 24-inch below length the reactor I BF3fessel filled con-thennai neutron counters. These detectors are made from stainless steel and are filled to 70 cm. Hg. pressure. They are shielded by a 1/2-inch thick lead sleeve, and they are surrounded by a plastic moderator. Addi-tional 1/2-inch thick lead shielding is provided between the vessel and the detectors. The ratio of the count rates from these two sets of de-tectors is used to determine the water level in the vessel above the core. The detector assembly above the reactor vessel consists of two counters sheathed in 0.065-inch thick steel. Figure 2 shows the preliminary sketch of the cross-section of the top detector assembly. Four detector assemblies are then mounted in position above the vessel head as shown in Figure 3. The top detectors are expected to have a combined sensitivity of approxi-mately 500 counts per NV. The bottom detectors are mounted in the area below the reactor vessel. All eight of these detectors are consolidated into a single assembly as shown in Figure 4. Since these large detectors are too sensitive for use during operating power levels, a pair of small, less sensitive 10B lined detectors will be used to measure level during power operation, if necessary. While details of electrical wiring diagrams for the system have not yet been finalized, Figure 5 shows the features that will be incorporated in this system. The 10BF3 detector assemblies are divided into pairs with their associated amplifiers so that the redundancy requirements will be satis-fied. The amplifiers drive two counter timers. One counter operates on the top detectors, set-up for a present number of counts. Operation of this counter gates the second counter which accumulates counts from the bottom detector during the period when the first counter is counting a present number of counts. Consequently, the counts from the second scaler is proportioned to: Bottom Count Rate Top Count Rate The reduction in the amount of water above the reactor core increases the top count rate, thereby reducing the ratio. Thus, the second counter display counts increase with increased water level. Subcoolinq Meter "h A description of the primary coolant saturation meters installed at the Farley Nuclear Plant was submitted to the NRC for its review and approv-al by Alabama Power Company letters dated December 31, 1979, for Unit 1, and June 20, 1980, for Unit 2.
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- The subcooling meter provides continuous main control board in- l dication of margin-to-saturation conditions. Two identical channels operate independently except that the same sensor may be input into each channel. Main control board indication consists of two meters that pro-vide a continuous pressure indication of margin-to-saturation and degrees superheat. Multiple core exit thermocouples wide range TH0T and TCOLD' and redundant safety grade system pressures, are used for inputs to the subcooling monitor. Two thermocouples per quadrant nearest the center of the core were selected to monitor exit temperatures. The subcooling monitor is a highly reliable and testable system powered from a vital instrument bus and environmentally suited to the service condition for the main control room. Vital power to the subcooling meter is separated from the IE electrical distribution system with fuses. Signals for the core subcooling monitor are picked up on the isolated (control) side of the protection channel which ensure that the addition of the subcooling meter does not adversely impact the reactor protection or engineering safety features systems. Emergency procedures provide for backup methods to determine subcooling using steam table;.
The installed core subcooling monitors have been tested by a cali-bration and functional tert procedure for both units. These tests are a comprehensive software ana hardware performance verification which in- ' cludes initial calibration of inputs and functional testing of micro-processor self test features, calculation outputs, display capabilities and alarm outputs. The maximum thermocouple indication error found during testing of the Unit 1 subcooling monitor was 7 F at 2000 F in a conserva-tive (high) direction. The maximum error found during the Unit 2 test i- was 6*F. Incore Thermocouoles A description of the Incore Thermocouple System installed at the Farley Nuclear Plant was submitted to the NRC for its review by Alabama Power Company letters dated July 17, 1980, and July 24, 1980. The thermo-couple system utilizes 39 thermocouples in Unit 1 and 51 thermocouples in Unit 2 positioned to measure fuel assembly coolant outlet temperature at preselected core locations (FSAR Figures 4.4-22 and 4.4-22A). The thermo-couples are the chromel-alumel type and have an accuracy of 2*F. As a result of Alabama Power Company's response to I&E Circular 80-15, a modification was implemented for realignment of certain thermocouples to the upper head region as a means of indicating upper head voiding during natural circulation. There are 16 thermocouples inputs for the core sub-cooling meter (eight per channel), two of which are upper head thermocouples. The primary means of monitor % incore thermocouple temperature is the core subcooling monitor system. Each channel of the subcooling monitor receives inputs from 8 thermocouples (2 per core quadrant per channel, for a total of 16 thermocouples). A digital readout of any of the 16 single thermocouple temperatures may be obtained at the subcooling monitor panel located behind the control board. The upper limit of the readout is in excess of 2300 F. .
- 4 All of the control equipment for the Thermocouple System is located on a rack in the control room. A multipoi:t precision indicator has been provided to indicate the temperature sensad by the themocouples. Only one thermocouple at a time can be connected to the indicator. Switches have been provided on the front of, and above, the indicator to select the thermocouple desired to be read.
An additional selector switch located on the front of the panel allows either the low (100-400*F) or high (400-700 F) range measuring circuit to be used. Besides being directed to the indicator, the thermo-couple outputs are also applied to the plant computer (up to 1900*F). B. Design Analysis of Water Level Measurement System p, the National During Nuclear the Summer Corporation (NNC)ofconducted 1979, under testsEPRI usingsponsorgCf a' neutron source in a water tank mockup to determine whether neutron measurements outside a reactor vessel would provide an unambiguous measure of the water level inside the vessel. The method chosen is shown schematically on Figure 1. An array of neutron detectors were placed above and below the vessel, and the ratio of the counts from these arrays were related to the water level. Figure 6 shows count rates measured above the vessel as a function of water level. The initial rapid drop off for about four feet was due to shielding, by the water, of fissio;i neutrons from the source (or reactor). Beyond four feet, neutrons were principally produced from the action of high energy (over 2.2 MeV) ' gamma rays on the deuterium within the water. Since these gammas travel further than neutrons in water, these photo-neutrons predominate when the water in the tank was deeper than four feet over the core. When the counts from the lower detector are divided by counts from the upper detector (as in Figure 7), a relationship roughly proportional to water depth is obtained, except just before core uncovery when a much greater effect is observed. As shown by this data from tanks tests at NNC, for water levels over four feet above the core, most of the neutrons detected arise from inter-action of high energy gammas with the deuterium impurity in the water, while below five feet (where there is a danger of core uncovery) the neutron level above the reactor rises very rapidly due to neutrons produced by fission in the reactor core. Thus, this system provides a vivid warning well before core uncovery. Less dramatic indications are provided to gauge water level in the range between five feet and full. This is shown in the correlation on Figure 7. In this test, top counts were referenced against a side de-tector, instead of the bottom detectors used in the actual installation. Following these tests and ad$ional tests at Prairie Island and Rancho Seco, equipment was built and used to demonstrate successfully the operation of this system at Trojan during initial drain-down. Data from the Trojan test is shown on Figure 8. Based on this data, and on Trojan side counter data, the curve on Figure 9 has been projected to indicate the performance of the actual system one day after shutdown.
Following the Trojan test, analytical studies have been made to further investigate the system's capabilities. It has been brought out that the system reads weight of water above the core, thus giving a valid indication before core uncovery. The preliminary system discussed in A (above) is being installed in Farley Unit 2 prior to exceeding 5 percent power. In order to provide an indication of lowering reactor vessel water level on Unit 1, an abbreviated version of the system installed on Unit 2 has been installed on Unit 1 during the current refueling outage. This system consists of one detector assembly installed above the Unit i reactor vessel head and an alarm set at a predetermined count rate to give an indication of decreasing water level. The detector installation f offour(4) detector assemblies, each containing two '5 thecounters, BF3 test consisted distributed around the reactor vessel head area on top of the vessel head insulation. After the detectors were installed and calibrated, the reactor vessel water level was lowered in two (2) foot intervals until reactor vessel water level was at the centerline of the vessel nozzles. The water level was then raised in two (2) foot intervals until the vessel was full. At each level a 1,000 second count was taken. The test was repeated with the detectors shielded for a total of three test runs. The results of the tests on Farley Unit 1 demonstrate that neutron detectors mounted above the reactor vessel respond to changes in water level within the vessel. Improvements in the system may be made through better threshold adjustments of the detectors and detector shielding. ,- Arcas that require further investigation include a clearer understanding of the mechanism whereby neutrons reach the detectors when the core is covered by 10-20 feet of water. C. Description of Additional Test Programs for Evaluation of Water Level Measurement System Following the installation of the complete system on Farley Unit 2, the effects of density change with temperature will be investigated during plant startup. Also, during a full-power operation period, the detectors will be evaluated for their performance. Additional necessary water level measurement tests will be performed during unplanned outages, but no later than the next refueling outage currently planned to commence during the fall of 1982. During this entire period continued engineering evaluations of the system will be performed towards improvement of the design, installa-tion, and qualification of the system. If the system proves to be viable and reliable during the Unit 2 tests, then the same system will be installed in Unit i during the refueling outg,ge currently planned to commence in the fall of 1982. However, if the system fails to meet performance expectations, then Alabama Power Company will install, in both units, the best system available as rapidly as practicable. It is recognized that th? non-invasive water level system is still in its developmental stage and the effort undertaken by Alabama Power Company i
could be construed as the first prototype field testing program. As further infomation becomes available, Alabama Power Company will keep the NRC fully informed. If the test program shows that this reactor vessel water level system is a viable and reliable system, Alabama Power Company will make additional submittals to the NRC to provide a summary of the key operator action instructions in the emergency procedures for inade-quate core cooling and to demonstrate the ability of the system to provide unambiguous, easy-to-interpret indication of inadequate core cooling. These __ additional submittals will also describe the program to qualify the reactor vessel water level system in accordance with Appendix B of NUREG-0737 and the human factors considerations in the design of the vessel level displays. Such qualification will be performed in conjunction with EPRI. Table 1 is a schedule of development, installation, and testing of the vessel level systems for Units 1 and 2. There are tests planned at LOFT concerning this method of vessel level detection which should yield a better understanding of behavior at lower water levels, shorter times after shutdown, and under transient conditions. D. Evaluation of Conformance of ICC Instrumentation System To Regulatorv Guice 1.97, Rev. 2 The installed subcooling meter is a highly reliable and testable system powered from a vital instrument bus and environmentally suited to the ser-vice conoition for the main control room. Signals for the core subcooling '_ monitor (hot and cold leg RTDs, thermocouples, pressurizer pressure) are picked up on the isolated (control) side of the protection channel which ensures that the addition of the subcooling meter does not adversely im-pact the reactor protection or engineering safety features systems. The hot and cold leg RTDs and pressurizer pressure inputs are c.ualified to IEEE 323-1971. Westinghouse will initiate a generic program in January, 1981 to qualify the present thermocouples to the requirements of IEEE 323-1974. Alabama Power Company will notify the NRC as more information becomes available regarding a schedule for completing this program. The Reactor Vessel Water Level Measurement System is not qualified; however, a program to qualify the system in accordance with NUREG-0737, Appendix B, and Regulatory Guide 1.97 will be developed in cooperation with EPRI if the system proves to be a viable water level measurement system. E. Descriotion of the Computer Functions Associated with ICC Monitoring Core Subcooling Monitor - Initial Testing The installed core subcooling monitors have been tested by a calibration and functional . test procedure for both units. These tests are a comprehensive ee --+e
software and hardware performance verification which includes initial calibration of inputs and functional testing of microprocessor self test features, calculation outputs, display capabilities and alarm outputs. The maximum thermocouple indication error found during testing of the Unit 1 subcooling monitor was 7*F and 2000*F in a conservative (high) direction. The maximum error found curing the Unit 2 test was 6*F. Peri tdic Retesting Retesting is to be accomplished under the Preventive Maintenance Program and will be performed at refueling intervals. This test will be a full functional t est as outlined in A (above). Diagnostic Capabilities The subcooling monitor has a range of diagnostic capabilities includ-ing channel self-test, channel processor failure, power failure, sensor input failure and calculational output failure. These diagnostics along with vendor-recommended maintenance will be utilized to ensure the'subcool-ing monitor channels are operable. Incore Thermocouple Read'uto Panel - Initial Testing Testing of the Incore Thermocouple Readout Panel was performed for Unit 2 by functional test procedure. Thermocouple inputs and readout panel
,- performance was verified to be within expected tolerances. Input verifica-tion was performed on each thermocouple input. Maximum thermocouple indica-tion error over the temperature range tested was 4.5 F in the conservative (high) direction on Unit 2. Similar tests will be performed on Unit 1 at the next refueling outage.
Periodic Retesting Retesting is to be accomplished under the Preventive Maintenance pro-gram and will be performed at refueling intervals. l Diagnostic Capabilities l There are no specific diagnostics internal to the Incore Thermocouple Readout Panel. Good industry practice will be utilized to ensure operability of the readout panel. Plant process Comouter (P2500) -
- Initial Testing l The plant process computers for both units have been tested using a l combined input verification and acceptance test procedure. These tests are a comprehensive software and hardware performance verification which includes
, calibration of inputs, individual program performance testing, and calculated e N e e
value, display, and alarm outputs verification. The thermocouple program package and incore thermocouple inputs are verified and calibrated at this time. Unit 1 and Unit 2 preoperational test results indicated errors of less than 4*F over the range tested. Periodic Retesting Incore thermocouple input verification to the process computer will be performed at refueling intervals beginning at the next refueling outage. Diagnostic Cacabilities Incore thermocouple inputs are monitored and alarmed by the process computer for sensor input failure and temperature alarm limits. Thermo-couple program outputs are also monitored by the computer for alarm limits. Preventive maintenance and good industry practice is utilized to ensure pro-cess computer system operability. F. Schedule for Installation of Additional Instrumentation Table 1 provides the latest schedule for the development and installa-tien of the water level measurement system at the Farley Nuclear Plant. G. Procedure Guidelines for Use of Additional Instrumentation As part of the development of the non-invasive water level measurement system, procedure guidelines for use of the system by the operators will be developed. These procedures will be developed and implemented prior to plac-ing the water level system into operation following NRC approval of ti,e system. H. Procedures for Use of Current Instrumentation The Westinghouse Owners Group, of which Alabama Power Company is a member, has performed analyses as required by Itera I.C.1 of NUREG-0737 to study the effects of inadequate core cooling. These analyses were provided to the NRC " Bulletins and Orders Task Force" for review on October 31, 1979. As part of the submittal made by the owners group, an " Instruction to Restore Core Cooling During a Small LOCA" was included. This instruction provides the basis for procedure changes required to recognize the existence of in-adequate core cooling and restor core cooling based on existing instrumenta-tion. When the Reactor Vessel Water Level System is operational, the use of this system will be included idbthe procedures for recogniring the existence of inadequate core cooling and restoring core cooling. I. Additional Submittals to NRC As further information becomes available, Alabama Power Comoany will
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; keep the NRC fully informed. If the test program shcws that this reactor ! vessel water level system is a viable and caliable system, Alabama Power Company will make additional submittals to the NRC to provide justification ;
of the acceptability of the prototype vessel level system currently under development. Tnese additional submittals will also describe the program to qualify the reactor vessel water level system in accordance with Appen-dix B of NUREG-0737 and the human factors considerations in the design of j the ves:e1 level displays, I J I l - 4 1 i 1 J es-woce,, e,,-, ...--,,w- , ,rww ,-e,n+, yw.. ,,,r., , . .v e n -p-,- - , , - - , m- --y gm-scr.-,-,.. w. v---- . , -w.,y- ,,v, e,-,,, y e, -,.- ,, y
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II.F.2 (continued) 56 TABLE 1 REACTOR VESSEL LEVEL SYSTEM DEVELOPMENT AND PLANNED INSTALLATION SCHEDULE FARLEY NUCLEAR PLANT FARLEY NUCLEAR PLANT UNIT ACTIVITY SCHEDULE DATE Unit 1 REACTOR VESSEL DRAIN DOWN November 1980 Unit 2 INSTALL PROTOTYPE SYSTEM Prior to Exceeding S% Power Unit 1 REACTOR COOLANT SYSTEM HEATUP End of second refueling outage TESTS Unit 1 PROVIDE RESULTS OF HEATUP July 1, 1981 TESTS TO NRC Unit 1 COMPLETE INSTALLATION OF End of second refueling outage ABBREVIATED PROTOTYPE SYSTEM Units 1 and 2 COMMENCE DEVELOPMENT F h March 1981 PERMANENT SYSTEM Unit 1 ADDITIONAL TESTING AND DATA Forced Outages (between end of COLLECTION WHEN FEASIBLE' second refueling and beginning of third refueling) 4- Unit 2 PROTOTYPE TESTING WHEN Forced Outages (between initial FEASIBLE criticality and beginning of first refueling when decay heat is available) Unit 2 FINAL TESTING OF PROTOTYPE Refueling Outage 1982 SYSTEM l Units 1 COMPLETE DEVELOPMENT OF January 1, 1982 ' - and 2 PERMANENT SYSTEM Unit 1 INSTALL PERMANENT SYSTEM Refueling Outage 1982 Units 1 PROVIDE COMPLETE DESCRIPTION 90 days after completion of the and 2- 0F FINAL TEST DATA & REQUEST Final Testing of prototype during NRC APPROVAL T0 the Unit 2 1982 refueling outage UTILIZE THIS SYSTEM FOR FNP UNITS 1 & 2 Units 1 PROVIDE FOR NRC CONCU ENCE A 90 days after approval of the and 2 PROGRAM TO QUALIFY THE prototype system by the NRC for APPLICABLE COMPONENTS OF THE use in FNP Units 1 & 2 REACTOR VESSEL LEVEL SYSTEM TO R.G. 1.97 INCLUDING QUALIFI-l CATION REQUIREMENTS OF NUREG ! 0737, APPENDIX B AND HUMAN FACTORS CONSIDERATIONS ! Units 1 DEVELOP AND MODIFY, AS APPLI- 90 days after approval of the _, i and 2 CABLE, EMERGENCY OPERATIONS prototype system by the NRC
- l PROCEDURES AND PROCECURES FOR for use in FNP Units 1 & 2 USE OF THE VESSEL LEVEL SYSTEM L
66 II.F.2 (Continued) B. Core Subcooling Meter The installed core subcooling monitors have been tested by a calibration and functional test procedure for both units. These tests are a comprehensive sof tware and hardware performance verification which includes initial calibra-tion of inputs and functional testing of microprocessor self test features, calculation outputs, display capabilities and alarm outputs. The maximum thermocouple indication error found during testing of the Unit 1 Subcooling Monitor was 70F. at 20000F in a conservative (high) direction. The maximum error found during the Unit 2 test was 6 F. C. Incore Thermocouples As a result of Alabama Power Company's response to ISE Circular 80-15, a modification was implemented for reallignment of certain thermocouples to the upper head region as a means of indicating upper head voiding during natural ci - a.la tio n. There are 16 thermocouples inputs for the core subcooling meter (eight per channel) two of which are upper head thermocouples. The following is a response to the criteria of NUREG-0737, Appendix B,
" Design and Qualification Criteria for Accident Monitoring Instrumentation":
Criteria (1) Response Alabama Power Company does not fully comply with this criteria but has instrumentation in existence used to measure ICC is basically structured as follows:
- 1. Incore thermocouples aad reference junction boxes.
- 2. Core cooling monitoring system (electronics and display).
- 3. RTD's.
The incore ther=ocouples and reference junction boxes were supplied much earlier than the issue by the NRC of any requirements concerning their use as TMI related items. The core cooling monitoring system (electronics + display) were supplied to Farley Unit 2 qualified to IEEE-323-1971 standard for environmental qualification and to Reg. Guide 1.100 (IEEE-344-1975) for l seismic. The design of such system was reviewed to ensure that the l equipment meets the required qualification conditions. l l The RTD's were supplied qualified to IEEE-323-1971 and the qualification documentation appears in WCAP-9151g, To more fully comply Alabama Power Company will participate in the l Westinghouse generic program to qualify the present thermocouples which will be initiated in January,1981. Alabama Power Company will notify the NRC as more information becomes available regarding a schedule for ccmpleting this progra=. I i , a
66(a) Criteria (2) Response A single failure of a reference junction will not totally negate the value of the readings from the thermocouple system as described below: The individual thermocouple wires are separated and routed to either of two reference junctions. The thermocouple reference junction boxes are provided to permit transition from chromel-alumel thermocouple extension wiring to copper field wiring. These units provide a controlled 1600F temperature reference for the incore thermocouples. Each reference junction box contains three platinum resistence temperature detectors (RTD). Two of the RTD's from each unit are connected directly to the plant ecmputer for monitoring of reference junction temperature; the third RTD in each unit is an installed spare. The selection of the thermocouples to be connected to each reference junction was made to eliminate the possibility that a single failed reference junction would totally negate the value of the readings received from the thermocouple system. In the event a reference junction does fail, the thermocouples connected to the remaining reference junction have been chosen so as to provide a meaningful representation of the core temperatures. The core cooling monitoring system (electronics and display) has redundant channels for surveillance and indication. It has capability to detect discrepancy in the measure =ents using the calculations obtained by the pressure. (- The system is only designed to be a surveillance system and as such does not provide plant protection functions or claims to be a class IE equipment. This equipment was built and supplied to the utility previous to the existence of any NRC requirement of TM1 related equipment to be included as class IE equipment. Criteria (3) Response Alabama Power Company's systen meets the requirements that no required power source be non Class IE. The instrumentation required for displaying thermocouple readouts on the Main Control Board is powered from inverter IE, a non Class IE power source. However by manual means by the use of an RTD bridge the hot junction box temperature may be determined thereby
-allowing an accurate determination of accurate thermocouple temperatures without reliance on power.
Criteria (4) Response One instrumentation channel s available prior to an accident as described below: An incore thermocouple readout panel is located adjacent to the safeguards section of the main control board. This readout panel has a display capability for 51 thermocouples which would ensure one channel is available for readout prior to an accident. (See Criteria (6) response for further detail.) e y - - - - -n
66(b) Criteria (5) Responso Alabama Power Company's incore thernoccuple system meets the Q.A. requirements in this criteria, documented in APCO and Westinghouse Q.A. Manuals which have been reviewed by the NRC, as described below: Incore thermocouples are classified Category D items within the Operations Quality Assurance Program as documented in FSAR Section 17.2 and the Operations Quality Assurance Policy Manual. A detailed description of the procurement, installation, inspection, testing, storage, and records activities associated with D items are documented
~~
in Operations Quality Assurance and Farley Nuclear Plant procedures. The core tooling mon;toring system complies with 10CFR50 Appendix C and is incic 2ed under Westinghouse Quality pressure programs QCS-2 Rev. I and program directive 10458-001. Criteria (6) Response Alabama Power Company meets this criteria by having the continuous availability of readout as described below: The incore thermocouple readout panel consist of a display and toggle switches which are utilized to select individual ther=ocouples. A display for any one of the 51 thermocouples is provided by positioning the appropriate toggle switch. An additional selectot switch located i on the front of the panel allows either the low (100-4000 F.) or high I (400-7000F.) display to be utilized. l. h-In addition, a display of up to 26 thermocouple valves can be selected for display on either control room CRT. Thermocouple readouts using the 1RT via the plant computer have a range up to 19000F. Criteria (7) Response f
- Alabama Power Company has the capability to record and trend
. appropriate readouts as described below: Utflizing the plant ccmputer, the Farley Nuclear Plant has the capabilaty for trending up to 51 of the thermocouples with output on the trend typewriter. i Criteria (8) Response Alabama Power Company's system of detremining thermocouple readouts includes instruments which can be easily discerned under accident con-dicions as described below: k The instruments, described in response to Criteria (6), are in the Control Room, readily available and easily discernable to the operator in the event of accident conditions.
66(c) Criteria (9) Resoonse Alabama Power Company's system m: ets this criteria as described below: The transmission of signals from the thermocouples for use for da. vices other than the meter on the Control Panel are isolated by devices which would preclude the impairment of the thermocouple system under faulted conditions experienced in the other devices. Criteria (10) Response Alabama Power Company's 'vsten has the capability for checking operational availability of eaci monitoring channel as described below: The licensed reactor operator utilizing the meter on the Control Room Control Panel has the ability to cross-check thermocouple readings in locations of close proximity to verify the adequacy of such irdividual readings. Criteria (11) Response Alabama Power Company has a calibration check program which enhances the capability of monitoring instrumentation as described below: i A calibration chack is performed on the thermocouple system by providing a know simulated signal to determine proper functioning of this system at such refueling. Criteria (12) Resoonse Alabama Power Company has a systen for controlling access of removal and modification of this equipment as.decribed below: The design of the core sub-cooling monitor is such that channels cannot be taken out of service without removing the front panel of the
- monitor. Changes to the output of the process computer cannot be made i without an administratively controlled key switch.
Criteria (13) Response Alabama Power Company has a system to administratively control acces: to setpoint adjustments as described below: 1 Thermocouple readouts are sent as a voltage input to the core sub-cooling monitor. ture or to " degrees toThesevoltag saturation' greadings inputs are viaconverted a printed to a tempera-circuit card called a " Programable Read Only Memory" '(PROM). Thus, setpoint adjust-ments cannot be made without a hardware change to the PROM circuit card. The process computer also processes the voltage input from the core thermocouples. Changes to the computer programming cannot be made without an administratively controlled key switch.
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e
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, , 66(d)
Criteria (14) Response Alabama Powex Company's system is designed in a manner so as to minimize development of conditions that would cause potentially confusing indications to the operator as described below: The design of the core sub-cooling monitor is such that if the input signals are not within a preset range the operator will be alerted by an " INVALID" display on the analog meter and an " ERROR" display on the digital catput. The process computer input and the core cooling monitor also use reference junction box temperatures compensation in determining thermoc.ouple readouts. Criteria (15) Response Alabama Power Company's system is designed and constructed so as to facilitate recognition, location, replacement and repair of components with out of service time as described below: The process control instrumentation circuitry for the in-core thermocouples and core cooling monitor is located outside containment in accessable areas. Only the reference junction box and the thermo-couples (and RTD's for the core cooling monitor) are located inside containment. Criteria (16) Response Alabama Power Company's system is designed to directly measure - - - desired variables as described below: The incore thermocouples are positioned directly above the outlet of the fuel aseemblies in order to provide as " direct as possible" measurement of the fuel temperature. Criteria (17) Response Alabama Power Company's in-core thermocouple system described in the response to' criteria (6) which is utilized during applicable normal operational evaluations is also utilized during accident conditions. Criteria (18) Response Alabama Power Company has a program for periodically testing this_ system as outlined in the response to Criteria (11).
~ '
..D. Incore Thermocouple' Readout Paneng f. - -
Ir.itial Testing . , Testing of the Incore Thermocouple Readout Panel was performed for ! Unit 2 by functional test procedure. Thermocouple inputs and readout panel performance was ve.<fied to be within expected. tolerances. Input, verification was performec on each thercocouple input. Maximum thermocouple indicatica error over the temperature range tested was , 4.~50F in the conservative (high) direction on Unit 2. Similar tests ' will be performed for. Unit L at the next refueling outage. __;
II.F.2 (Continued) 67 reriodic e Retesting Periodic retesting of the subcooling monitor will be performed at refueling intervals under the Preventive Maintenance Program, utilizing an Instrument Meintenance Procedure (IMP). The IMP is essentially a repeat of the initial calibration and functional test procedure used for acceptance testing in the initial set-up of the subcooling monitor. This test is a full functional checkout of proper input conversion, calculational output accuracy, MCB meter indication accuracy, MCB annunciation and lacal alann outputs, temperature and pressure auctioneering, display function outputs, and diagnostic output verification. Testing is accomplished following input calibration verification by using analog test signals generated from calibrated plant test equipment to simulate various combinations of temperature and pressure conditions which enccmpass the nomal operating (various subcooled conditions), approach to saturation, saturation and superheated ranges. For each temperature and pressure test point combination, the subcooling monitor calculational outputs and MCB meter indications are verified to be in agreement with expected values as determined from the steam tables. Alert and alarm conditions are simulated to verify proper MCB meter indication, annunciation actuation and local subcooling monitor alarn indications. In addition, all diagnostic and self testing features and display function outputs are verified to be functional. Self-test features are used to test alarm circuitry and MCB meter operation on a routine basis. Inputs which are ccmmon to the subcooling monitor and Process Computer are routinely verified to be in agreement. Subcooling monitor display functions are also routinely used to check monitor operability. - Diagnostic Caoabilities There are no specific diagnostics internal to the Incore Thermccouple Readout Panel. Good industry practice will be utilized to ensure operability of the readout panel. E. Plant Process Comouter (P2500) Initial Testing The plant process computers for both units have been tested using a combined input verification and acceptance test procedure. These tests are a com-prehensive software and hardware performance verification which includes calibration of inputs, individual program performance testing, and calculated value, display, and alann outputs erification. The themocouple program package and incore themocouple taputs are verified and calibrated at this time. Unit I and Unit 2 preoperational test results indicated errors of less than 4*F over the range tested. Periodic Retesting Incore themocouple input verification to the process computer will be performed at refueling intervals beginning at the next refueling outage. Diagnostic Capabilities Incore thermocouple inouts are monitored and alarmed by the process c mputer for sensor in;;ut failure and temperature alarm limits. Thermocouple program outputs are also monitored by the computer for alarm limits. Preventive maintenance and good industry practice is utilized - . to ensure process computer system operability.
76 II.K.3.5 AUTOMATIC TR:P Or nC'CTOR PUMP DURING LOSS-0F-COOLANT Previous Response By letter of June 26, 1980 for Unit 1, Alabama Power Co.npany addressed this item for the Fcriey Nuclear Plant. This is a new item for Unit 2. Clarification Response The Westinghouse Owners Group, of which Alabama Power Company is a member, has performed analyses using the Westinghouse small-break evaluation model (WFLASH) to show ample time is available for the operator to trip the reactor coolant pumps following certain size small breaks (see WCAP-9584). In addition, the owners group is supporting a best-estimate study using the NOTRUMP computer code to demonstrate that tripping the reactor coolant pump at the worst trip time after a small-break will lead to acceptable results. For both of these analysis efforts, the Westinghouse Owners Grcup is performing bling post-test predictions of LOFT experiment L3-6. The input data and model to be used with WFLASH on LOFT L3-6 was submitted to the staff on December 1, 1980. The information to be used with NOTRUMP on I.0FT L3-6 will be submitted prior to performance of the L3-6 test, as stated in owners group letter 08-45, dated December 3, 1980. The LOFT prediction frcm both models will be submitted to the NRC on c~ February 15, 1981, given that the test is performed on schedule. The best estimate study is scheduled for completion by April 1,1981. Within three months after NRC approval of these models, Alabama Power Company will provide justification that these models envelope the FNP. Based upon these studies, Alabama Power Company believes that resolution of this issue will be achieved without any design modifications. In the event that automatic trip of the reactor coolant pumps is required after the NRC determination of model acceptability, a schedule will be provided for potential inodifications. l l
- 1 l
l
-. , , - . . - , . ,~ . - . .,_,. ._
II.K.3.17 REPORT OH OUTAGE OF ECC SYSTEMS - LICENSEE REPORT AND PROPOSED TECHNICAL SPECIFICATION CHANGES Previous Response Alabama Power Company's letter dated June 26, 1980 responded to this item for Farley Nuclear Plant Unit 1. Clarification Response Table 1 is a listing of ECC systems outages for Farley Nuclear Plant - Unit 1 since December 1,1977 during modes in which the systems are required to be operable in accordance with technical specifications. Included in this listing is the duration of the outage, the affected ECCS component, the cause of the outage, and corrective action taken to return the component to operable status. ECC system outages routinely required for surveillance testing are not included in the listing. It should be noted that no accurate records exist for the duration of outages of ECCS equipment during modes of operation in which this equipment is not required by technical specifications. A review of Unit 2 limiting conditions for operation status sheets will be performed and documented for ECCS outages occuring during modes of operation in which the systems are required to be operable in accordance with technical specifications. This review will cover a five year time period beginning with initial criticality. This documentation will include the duration of the outage, the 6ffected ECCS component, the cause of the outage, and corrective action taken to return the component to operating status. ECCS outages routinely required for surveillance testing will not be included in this review. It should be noted that no accurate records exist for the duration of outages of ECCS equipment during modes of opera-tion in which this equipment is not required by technical specifications. The documentation of this review will be maintained onsite for NRC review.
=
, , II.".3.17 (continued) -
82 TABLE 1 ( LISTIt:G OF ECCS OUTAGES Date Begun Date Finished Time Date Time Date System Duration Cause & Corrective Action
;_0145 12/6/77 0323 12/6/77 BIT 1.5 Boron concentration below 20,000 ppm. Boric a~cid batched to raise concentra ~
' tion..,
1145 12/28/77 0245 12/29/77 BIT 15 Inadvertent S.I. caused low
- boron concentration. Boric '
. Acid batched to raise con-centration.. ,
1000 1/21/78 2200 1/23/78 RHR 60.0
. RHR Pump 1A discharge valve -
. controller output re-adjusted
' and returned to service.
1020 4/25/78 1245 4/25/78 BIT 2.4 Baron concentration below 20,000 ppm. Boric Acid batched to raise concentra-
, . tion.
1140 C 5/29/78 2010 5/29/73 BIT 8.5 ~ Boron concentration belcw 20,000 ppm. Boric Acid batched to raise concentra-tion. , 1355 6/7/78 1650 6/7/78 RUST 2.9 RMST volume below minimum
~~ value. Level increased with blended makeup from CVCS
' .. blender.
1250 6/10/78 2120 6/10/78 RHR 8.5 RHR Pump 1A tagged out to perform maintenance on - leaking CCU relief valve. . ' ' The valve was repaired, - tested satisfactorily, and i returned to service, i 0330 7/ 21 /78 2215 BIT % 85.8 7/24/78 BIT heat tracing de-energized - to perform modification to' r' flow transmitter. t 3900 7/25/78 1455 BIT 5.9 7/25/78 BIT heat tracing de-energized to perform modification to f
. ij
{ flow transmitter. ' .j it 1 i:
.!<eI
(
, II.K.3.17 { continued) TABLE 1 83 i LISTIf4GOFECCSOUTAGES(CONTI:UED) l
( . I Date Begun Date Finished ! Time Date Time Date System Duration t Cause & Correctivs Action . . s 0745 8/4/78 1705 8/4/78 RHR 9.3 Modification to level switch required tag out of Rh2 . suction valves. - , 0840' 8/7/78 1500 8/7/78 - RHR 7.3~ !!cdification to level switch -
' required tag out of RHR .
- suction valves. , ,-
0800 RHR suction valves tagged . 8/9/78 1125 8/10/78 RHR 27.4. -
.,' . out for work on RWST level switches. . -
~,.
1040 8/14/78 1000 8/15/78 RHR 23.3 - RHR suction valves tagged cut for work on RWST level sviitches.
~
l 1415 ~ 8 / 21 /78 0312 8/22/78 ACC 13.0 Accc:ulator 1B nitr: gen . l vented to allow for main-- .-
' tenance of outlet valve. -
Returned to service. ,.
-k .
2245 12/21/78 0400 12/21/78 ' BIT 1.3 Lost BIT recirculation flow due to performance of . surveillance test. 050 1/11/79 1120 1/11/79 .RER 0.5 RHR Pump 18 breaker racked . out to perforra preventive
~
maintenance. i ~530 1/23/79 . 2340 1/23/79 ~ '. BIT 18.2 . BIT. inlet valve failed to '
' operate. Motor actuator !
repaired and BIT returned to service. 730 1/23/79 s.
'1030 1/23/79 BIT 3.0 Inl.et sample icw boron' I
* - - concentration. Borated.and l g returned to service. , , .
l 335 2/23/79 1707 2/23/79 BIT 0.8 BIT declared incperable ~. I voluntarily in order to ' 5 repair air 1ine to recirc ' valve. Air line repaired ! and returned to service. ! C - 2 [ ~ ; e
l 11.K.3.17 (' continued) 84 - TABL51 . { LISTINGOFECCSOUTAGED(CONTINUED) Date Begun Date Finished - Time Date Time Date System ' Duration Cause & Corrective Action ' 1045 2/27/79 - 1145 2/27/79 RHR 1.0 RHR pump 18 tagged out to - 7 change oil. - ., :. 0100 2/28/79 0145 2/28/79 . RHR 0.8 RHR pump 1A tagged out.to -
. change oil. .: . .
1427 S.I. 10/24/79 1432
,10/25/79 24.0. Spurious SI on low y'- ,
pressurizer pressure. Due.. to defective block card. - . 1325 11/6/79 2200 11/6/79'. RWST 8.5 RWST Level indicators out -
.. of calibration calibrated
. and returned to service. ,
3532 11/21/79 1745 11/21/79 BIT 12.2 Boron concentration below 20,000 ppm. Boric Acid
' batched to raise concentra-
- tion. - -
_(- - ,
;'105 .12/4/79 1920 12/5/79 . RHR .23.0 RHR pump 1B declared inopersble due to tag out for repair of RHR miniflow valve. . .
~
.505 12/21/79 2105 12/21/79 CHG PMP 2.0 Charging pump 1B tagged out for oil change. -
w - Il25 12/25/79 0500 12/27/79 7.5' CHG PMP ' -Only one train of operable ' charging pumps. Pump 1C - racked out while 1B was
' racked in to train A. Due l ; to administrative error by shift foreman.
035' 2/2/80 2015 2/2/80 BIT g 10.0 Boron concentration belou - 20,000 ppm. Boric Acid batched to raise concentra-tion. .
-( .
. . ,L'J ,-
. , fl.K.3.17 (continued) 85 TABLE 1
( LISTINGOFECCSCUTAGES(CONTINUED) Date Begun Date Finished ! Tirce Date Time Date System Duration Cause & Corrective Action " 2020 3/4/80 2100 3/4/80 RHR 0.7 Power supply breaker to RHR pump 1A racked out for ... performance of general " maintenance procedure. . - .'
- 3 000 3/5/80 0420 3/5/80 RHR .4.4 - RHR pump 1B tagged out for '-
. oil change. . .
1600~ 3/20/80 1620 3/20/80 RHR 0.3 RHR pump 1B tagged out for . . routine preventive main- ,
. ~tenance.
- 400 3/29/80 0242 3/30/80 BIT 12.8 No recire flow through BIT. ~
Due to clogged drain line ^
. -and seal leakage.
200 5/28/80 0950 5/29/80 RHR 12.0 RHR heat exchanger discharge _(~ valve actuator failed.
~ Following maintenance valve was returned to service.
125 10/20/80 21b0 10/20/80 ~ CHG PMP 9.5 -
~
RWST supply to charging pump -
- suction valve would not open during surveillance testing.
Repaired contact arm on valve
- limit switch. Test completed -
satisfactorily. , g . 3 - e m 4
- a g
90 III.A.1.1 EMERGENCY PREPAREDNESS -- SHORT-TEPli C ed III.A.2 IMPROVING LICENSEE EMERGENCY PREPAREDNESS -- LONG-TERM . Previous Response By letters dated October 24, 1979, June 20, 1980, October 28, 1980, November 7,. 1980, and December 16, 1983, Alabama Power Company submitted the Radiological Emergency Response Plan and implementing procedures which describe our. program regarding emergency preparedness. Clarification Response
~
Alabama Power Company submitted the Farley Nuclear Plant Radiological Emergency
; Response Plan and implementing procedures via letters dated October 28, 1980 and November 7, 1980, respectively. The compensating actions provided in the alternate to milestone (3) will be utilized by Alabama Pcwer Company. In no case, however, will the alternative be exercised after July 1,1982, without prior approval by the NRC. Milestones (4) through (8) are seneduled for
_ completion in accordance with the dates specified in NUREG-0737. In the event that the above schedule cannot be met the NRC Staff will be notified. Attachment 1 provides a description of the compensating actions taken as an ' alternatetomilestone(3).
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A:cach=en: 1 1 (i) if only ele =snt 1 or ele =ent 2 is in uss: SI O The licensee (the person who will be responsible for making - ( offsite dose projections) shall check co=munications with the cognirant National Weather Service (NUS). first order station and NWS forecasting station on a monthly basis to ensure tha* ' routine meteorological observations and forecasts can be accessed.
RESPONSE
-- The cognizant National Weather Service (NWS) first order station and forecasting station are identified in the appropriate emergency plan i=plementing procedure.
A change to the emergency plan implementing procedure which pro- - " vides instructions and describes responsibilities for testing . communications networks is being processed to specify the monthly connunications check with the appropriate NWS stations. NOTE: NRC is on controlled distribution for EIPs. .
~
0 The licensee shall calibrate the meteorological measurements pro-gran at a frequency no less than quarterly and identify a readily - available' source of meteorological data (characteristic of site conditions) to which they can gain access during calibration , periods. RESPONSE: - Past calibration frequency was established as semi-annual. This calibration frequency will be increased to. meet the quarterly calibration require =ent.
'. o.rA source of meteorological data (characteristic of site conditions) is identified in the appropriate emergency plan imple=enting ~
procedure. 0 During ccnditions of' measurements system unavailability, an . _ ~
'~
alternate source of meteorological data which is characteristic .. i - of site conditions shall be identified to which 'the licensee can* i gain access. . . .
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RESPONSE
See response immediately above. , 0 The licens~ee shall maintaluga site inspection sheedule for evaluation - of the meteorological measurements program at a frequency no less than weekly. -
RESPONSE
l The current inspection schedule specified in approved plant procedures meets this require =ent. Specific instructions are included in the ( ' appropriate environ = ental procedure which verifies operability of wind speed, wind direction and te=perature gradient instrucentation.
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92 It shall be a reportable occurrence if the meteorological data unavailability exceeds the goals outlined in Proposed Revision 1 to Regulatory Guide 1.23 on a quarterly basis.
RESPONSE
A change is being processed to include this requirement in the __ appropriate environmental monitoring procedure to ensure that the requirement is promptly recognized. (ii) The portion of the DCM relating to the transport and diffusion of gaseous effluents shall be consistent with the characteristics of the Class A model outlined in element 3 of Appendix 2 to NUREG-0654. RESPONSE: . Attached is a copy of the proposed DCM for NRC review and acceptance. An interim hand calculational method will be docketed by separate correspondence as requested by the NRC staff in meetings the week of February 2,1981. (iii) Direct telephone access to the individual responsible for making offsite dose projections (Appendix E to 10 CFR Part 50(IV)(A)(4) shall be avail-able to the NRC in the event of a radiological emergency. Procedures for establishing contact and identification of contact individuals shall
,__ be provided as part of the i=plementing procedures.
RESPONSE
The NRC ringdown telephone provides this capability. The shift supervisor and the Emergency Director are identified as primary contact individuals; however, the individual responsible for making offsite dose projections shall be available to the NRC for direct telephone communications when necessary. All other items concerning emergency preparedness will be handled via separate letter. A status of prompt reporting will be included in Alabama Power Company's l response on interim hand calculational method. _. r
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