ML19305C328

Forwards Supplemental Response Commiting to Category a Requirements of NUREG-0578,short-term Lessons Learned Requirements

ML19305C328
Person / Time
Site:	Farley
Issue date:	03/14/1980
From:	Clayton F ALABAMA POWER CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578, TASK-2.E.4.2 NUDOCS 8003260556
Download: ML19305C328 (36)

Text

.,. .e Aatama Power Comery 6CO Norm f B:n S:reat P st C%Ce OCs 2641 Omrgeam. AaDama 35291 Te'eprone 205 323-5341 th M" "dR, AlabamaPower A

tre scumem e!ecinc sniem March 14, 1980 Docket No. 50-348 Dirr : tor, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Darrell G. Eisenhut

Dear Mr. Eisenhut:

As a result of discussions held with the NRC Lessons Learned Review Team on February 29, 1980, Alabama Power Company submits Enclosure (1) documenting the Category A items requir-ing additional information for resolution. This response sup-piements responses of October 24, 1979, November 21, 1979, and December 31, 1979, on this subject. It is Alabama Power Company's position that with this submittal, the Category A requirements of NUREG-0578 are satisfactorily resolved.

Yours very truly, k U2dNN F. L. Clayton, Jr.

BDM:n Enclosure cc: Mr. R. A. Thomas Mr. G. F. Trowbridge 037 S

800326066@

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a o Mr. Darrell G. Eisenhut March 14, 1980

References:

(1) NUREG -0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations",

July, 1979.

(2) " Follow-Up Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident", September 13, 1979.

(3) Handouts at Atlanta Regional Meeting, " Regional Meeting's TMI Short-Term Implementation Action",

September 28, 1979.

(4) " Discussion of Lessons Learned Short-Term Require-ments", October 30, 1979.

(5) Discussions on Implementation of Lessons Learned (Category A Items) with NRC Lessons Learned Westinghouse Review Team, February 29, 1980.

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• ENCLOSURE (1)

SHORT TERM LESSONS LEARNED COMMITMENTS Section 2.1.1 - Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer

' Level Indicators in PWR's Required' Action:

Document Class IE Interface Equipment for Pressurizer Headers (A&B Groups), PORV's, and PORV Block Valves.

Response

Pressurizer Heater Power Supply Pressurizer Heater Back-Up Groups A and B motive and control power interfaces with the emergency buses are accomplished through devices that have.been qualified in accordance with safety grade requirements.

Power Supply for Power-Operated Relief Valves and Block Valves The PORV's and associated block valves' motive and control power connections to the emergency buses for the PORV's and their associated block valves are through devices that have been qualified in accordance with safety grade requirements.

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., e 2.1.3. A - Direct Position Indication of Relief and Safety Valves Required Action:

Document that installed equipment meets requirements in NRC letter of October 30, 1979.

Response

The Pressurizer Power Operated Relief Valves (PORV) and Safety Valves ave stem mounted limit switches which operate red and green indicating lights on the Main Control Board. These indicating lights provide the operator with unambiguous indication of valve position (open or closed) so that appropriate operator actions can be taken. Also a main control board annunciator is provided in conjunction with this indication.

The PORV and safety valve position indication is powered from the plant Class IE D.C. distribution system, and backup methods of determining valve position are available and are diccussed in the plant emergency procedures as an aid to operator diagnosis and action. The limit switches are seismically qualified consistent with the PORV and safety valve requirements. The PORV and safety valve limit switches are environmentally qualified for any containment environment.

Section 2.1.3.B - Subcooling Monitor Required Action:

Document that installed equipment meets requirements in NRC letter of October 30, 1979.

Response

The subcooling meter provides continuous main control board indication 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 (2) meters that provide a continuous pressure indication of margin to saturation and degrees superheat. Multiple core exit thermocouples wide range T ADT HOT COLD' 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 qualified 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 ensures that the addition of the subcooling meter does not adversely impact the reactor protection or engineering safety features systems.

Table 1 provides a summary of information required for the subcooling monitor. Emergency procedures provide for backup methods to determine subcooling. Appropriate training has been conducted for these procedures.

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o TABLE 1 PAGE 1 of 2 SUBC00 LING MONITOR Display

1. Information displayed P-Psat subcooled T-Tsat superheat

2. Display type . Analog and ligital

3. Continuous or on demand Analog - Continuous Digital - On demand

4. Single or redundant display Redundant

5. Location of display Meter - Main Control Board Microprocessor - Main Control Room Instrument Racks

6. Alarms (include setpoints) Caution: 25 F subcooled for RTD 150 F subcooled for T/C 0

Ala rm: 0 F subcooled for RTD and T/C

7. Overall uncertainty Digital - 4 0 F for T/C; 3 0 F for RTD Analog - 5 0 F for T/C; S O F'for RTD

8. Range of display Calibrated region - 1000 psi subcooled to 20000F superheat overall - never offscale

9. Qualifications None at present Calculator

1. Type Dedicated Digital

2. If process computer is used, N/A specify availability

3. Single or redundant calculators Redundant

4. Selected logic Highest temp for RTD or T/C and lowest pressure

5. Qualifications None at present.

6. Calculational technique Functional fit - ambient to critical point

TABLE I PAGE 2 of 2 SUBC00 LING MONITOR Input

1. Temperature (RTD's or T/C's) RTD, T/C, and Tref

2. Temperature (number & location RTD - 2 hot and 2 cold legs of sensors) per channel T/C - 8 per channel

3. Range of temperature sensors RTD 7000F T/C 16500 F (calibration 0 unit range 0-2300 F)

4. Uncertainty of temperature 10.7% RTD sensors

5. Qualifications IEEE 323 1971

6. Pressure (specify instrument RCS Wide Range Pressurizer used)

7. Pressure (number and location 2 wide ran'eg - Loops 1 & 3 of sensors) l' narrow range - Pressurizer (per channel)

8. Range of pressure sensors Wide range 3000 psi Narrow range - 1700-2500 psi

9. Uncertainty of pressure Wide range - 1%

sensors Narrow range - 1.5%

Pressurizer - 11.0%

10. Qualifications IEEE 323 1971 Backup Capability

1. Availability of temperature Temp - Swap between T/C and RTD and pressure Press - Can defeat any of the three inputs. System uses auctioneered low pressure.

2. Availability of steam tables Saturated steam tables and tables to verify required subcooled conditions are included in Emergency Procedures.

3. Training of operators Operators have been trained on the use of the subcooling monitor to determine required subcooling conditions.

4. Procedures Emergency procedures have been revised to describe the utilization of the subcore cooling monitor readout and appended portion of the steam tables to determine subcooling cenditions. A system operating procedure has been written to guide operators in operation of the subcooling monitor. Appropriate personnel have been trained on these procedures.

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Section 2.1.4 - Containment Isolation Provisions for PWR's and BWR's Required Action:

i Document that-FNP has individual containment isolation valve control.

4 switches.

$ Response:

All containment isolation valves at Farley Nuclear Plant have.

individual control switches that do not operate more than one containment i isolation valve.

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Section 2.1.5.C - Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant Required Action:

Document that installed hydrogen recombiners meet requirements in NRC letter of October 30, 1979.

Response: ,

Farley Nuclear Plant has redundant electric hydrogen recombiners located inside containment for use in removing hydrogen gas from the containment atmosphere during post-accident conditions. These recombiners meet all engineered safety feature requirements and the controls and instrumentation for each are located on separate panels in in the Main Control Room. Hydrogen concentration can be adequately monitored in the Main Control Room. The emergency procedure for loss-of-coolant accident contains detailed instructions for operating the recombiners. Since this system is located inside containment and does not require mechanical hookup after an accident, personnel exposure during use is not a consider-ation at Farley.

Alabama Power Company has documented in its December 31, 1979 submittal, the review, upgrading, and training on emergency procedures for operation of electric hydrogen recombiners located inside containment.

9 2.1.6.A - Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWR's'and BWR's

-Required Action:

Submit a Detailed Description of Farley Nuclear Plant Preventive Maintenance

'and Leak Reduction Programs.

Response

Alabama Power Company has instituted a' leak reduction program which consists of the following:

A. Systems Included in Program (Scoped Systems)

We have reviewed plant systems and identified the following systems outside containment that could potentially contain highly radioactive fluids following a serious accident:

1. High Head Safety Injection System (Recirculation Portion Only)

2. Low Head Safety Injection System (Recirculation Portion Only) s

3. Residual Heat Removal System t

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Reactor Coolant System Letdown and Makeup System

5. Reactor Coolant Sampling System

6. Containment Spray System (Recirculation Portion Only)

7. Radioactive Waste Gas System B. Systems Excluded From System The.following systems have been excluded from the program. They will not preclude any option of cooling the reactor core nor will they prevent the use of needed safety systems.

Section 2.1.6.A - Page 2

1. Radioactive Liquid Waste System - Excluded by NRC in Regional Meeting

2. Radioactive Waste Gas System - Portions of system not required to process Volume Control Tank off-gas have been excluded.

Off gas processing (A.7) would be the means of handling highly radioactive gases resulting from various accidents.

C. Immediate Leak Reduction Measures

1. All vent and drain lines have been capped to prevent release due to seat leakage.

2. The packing of all valves (except Kerotest which is a packless, stainless steel diaphragm valve) in the scoped liquid systems have been inspected for leakage or evidence of leakage such as boric acid accumulation. Maintenance has been performed on packing of liquid system valves identified as requiring work.

3. The seals and packing on all pumps in the scoped liquid systems have been inspected for leakage or signs of leakage. Maintenance has been performed on a containment spray pump seal identified during the inspection.

.4. Valves, fittings and compressor seals in the scoped gaseous systems have been " snooped" for leakage. Maintenance has been performed on gas system valves and instrument fittings identified during leak tests as requiring work.

D. Procedures for Determining (Measuring) Leakage During the inspections described under C, leakage rates were recorded

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in drops per minute or hour. After completion of the inspection, the following additional leak rate tests were performed:

Section 2.1.6.A - Page 3

1. High Head Safety Injection System - Integrated Leak Rate Test

2. Low Head Safety Injection System - Integrated Leak Rste Test

3. Residual Heat Removal System - Integrated Leak Rate Test

4. Reactor Coolant System Letdown and Makeup System - Integrated Leak Rate Test

5. Reactor Coolant Sampling System - Integrated Leak Rate Test

6. Contain Spray System - Integrated Leak Rate Test

7. Radioactive Waste Gas System - Bubble (" Snoop") test of individual valves, fittings and seals. Quantitative value obtained from bubble rate.

E. Leakage Measurement Results:

Leakage rates for the systems following the maintenance described under C were as follows:

1. High Head Safety Injection (Recirculation Portion Only) - 10 drops / minute ~ 0.15 gallons / hour (attributable to charging pump leakage)

2. Low Head Safety Injection (Recirculation Portion Only)

Train A: no measurable leakage-(Feb. 1980 retest results)

Train B: no measurable leakage (Feb. 1980 retest results)

3. Residual Heat Removal System - system tested as part of low head safety injection.

4. Reactor Coolant System Letdown and Makeup System-only identified leakage is that of charging pumps reported in item 1.

5. Reactor Coolant System Letdown and Makeup System - no measurable leakage l

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6. Containment Spray System (Recirculation Portion Only) - no l measurable leakage l

7. Radioactive Waste Gas System - Portions of system which could '

receive high level gaseous waste except Recombiner A (out of service) were " snooped" for leakage. Detected leakage has been corrected. Current measurable leakage is zero SCFM.

Recombiner A will be leak tested after returned to service.

F. Continuing Leak Reduction A General Maintenance Procedure has been written and approved incorporating the inspections discussed under item C and the leak tests discussed under item D into the FNP preventative maintenance program. This program is under the supervision of the Maintenance Supervisor and includes removal of Boric Acid residue on components in order to facilitate leak detection during subsequent inspections.

Leak rate measurements will be performed periodically at intervals not to. exceed each refueling outage. Records of leakage rates will be retained in the plant maintenance files and will be retrievable using computer indexing.

G. IE Circular 79-21 All relief lines coming off the scoped systems have been walked down'against P & ID's as required by IE Circular 79-21. No piping discrepancies exist.

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- Section 2.1.6b - Design Review of Plant Shielding of Spaces for Post-Accident Operations Required Action:

Submit a Summary Description of all aspects of the Shielding Design Review including environmental effects.

Response

1. A design review for the Farley Plant Unit #1 has been conducted by Bechtel Power Corporation, which used the TID source terms and the 10CFR-20 and GDC19, 60-64 of Appendix A to 10CFR50, dose criteria.

This review established zone boundaries based on the following:

A-I The first zone is consistent with the personnel radiation exposure guidelines for vital areas requiring continuous occupancy to 10.015 rem /hr.

A-II The second zone is consistent with the personnel radiation exposure guidelines for vital areas requiring infrequent access or corridors to these areas. 0.015 to 5 0.100 rem /hr.

A-III The third zone is consistent with the personnel radiation exposure guidelines as given in General Design Criteria 19.

This included a time motion study to insure that the integrated exposure was not greater than 5 rem. 0.100 to i 5 rem /hr.

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Section 2.1.6b -Page 2 NOTE: Subsequent zones were selected by grouping them by powers of 10 so that rapid assessment of additional shielding measures could be used via " tenth value layers" of common shielding materials.

A-IV 5 to 50 rem /hr A-V 50-to 500 rem /hr A-VI 500 to 5000 rem /hr i

A-VII 5,000 to 50,000 rem /hr A-VIII 50,000 to 500,000 rem /hr (CAUTION): These zone designations should not be confused with those used for normal plant operation zone maps found in chapter 12 of the FSAR.

2. Scope of design review

1. Selection of systems for shielding review.

The criteria applied in selection of plant systems used in the shielding review resulted in several clar,sifications of systems as outlined in the following:

Category A (Recirculation Systems)

The first group of systems are those required by plant design  ;

l to mitigate a design basis loss of coolant accident and which l might contain highly radioactive sources in excess of the current design basis. A first priority safety concern was to ensure that operation of these systems containing a significant 1

Section 2.1.6b - Page 3 source will not adversely impact operator or equipment functions required outside the containment. Therefore, the following systems have been reviewed to ensure that this first priority safety concern was adequately addressed by the existing plant shielding design.

Those portions of the containment spray system used to recirculate water from the containment sump back into containment.

Those portions of the residual heat. removal system used to recirculate water from the containment sump back into containment.

Those portions of the high head safety injection system used to recirculate water from the containment sump via the RHR system, back into the containment.

Catagory B (Extensions of containment atmosphere)

In addition to systems listed above, there are other systems or portions of other system = which could contain radioactivity by virtue of their connection to the containment following an accident. Proper operation of the emergency core cooling systems would prevent extensive core damage and mean that these systems would not be expected to contain the significant radioactive sources required by this special analysis. Never-theless, such sources have been postulated in the following systems.

Those portions of the post accident containment combustible gas control system external to the containment which would contain the atmosphere from the containment.

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Section 2.1.6B - Page 4 Those portions of the containment ventillation systems external to the containment up to the first closed isol-ation valve which could contain the atmosphere from the containment.

Those portions of the sampling system used to obtain a containment atmosphere sample.

Catagory C (Liquid Samples)

Lessons learned task 2.1.8 requires that certain post accident liquid samples be obtained from the reactor coolant system or containment systems. Those portions of the sampling system which must be used to meet the intent of task 2.1.8 were reviewed.

Catagory D (Letdown)

The following portion of the letdown system was analyzed.

That portion of the letdown system from the reactor coolant system past the Letdown Heat Exchanger into the VCT and into the suction of the charging pump.

2. Quantification of Potential Radioactive Source Release Fractions The following release fractions were used as a basis for for determining the concentrations for the shielding review:

Source A: Containment atmosphere - 100% noble gases, 25%

halogens Source B: Reactor coolant - 100% noble gases, 50% halogens, 1% solids.

Source C: . Containment sump liquid - 50% halogens, 1% solids.

Section 2.1.6b - Page 5 The above release fracrians were applied to the total curies available for the particular chemical species (i.e. noble gas, halogens, or solid) for an equilibrium fission product inventory for an LWR Core.

The release fractions for Cs and Rb were assumed to be 1% for the.

purpose of this shielding review. Further evaluations of the TMI radioactivity releases may conclude that higher release fractions are appropriate. However, the overall effects of higher release fractions on radiation levels or integrated exposures are not expected to be significant. Therefore, the Regulatory Guide 1.7 solids release fraction, 1%, was used in this review. Similarly, no noble gases were included in the containment sump liquid (source C) because Regulatory Guide 1.7 has also set this precedent in modeling liquids in the containment sua, Furthermore cursory analyses have indicated 'that the halogens dorminate all shielding requirr Sats and that contributions to the total dose rates from noble gases are negligible for the purposes of a shielding design review.

3. SOURCE TERM MODELS Section 2 above outlines the assumptions used for release fractions for the Shielding design review. These release fractions are, however, only the first step in modeling the source terms for the activity concentrations in the systems under review. The important.

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Section 2.1.6b - Page 6 modeling parameters, decay time and dilution volume, obviously also affect any shielding analysis. The following sections outline the rationale for the selection of values for these key parameters.

A. Decay Time For the first stage of the shielding review process, no decay time credit was used with the above releases. The primary reason for this was to develop a set of accident radiation zone maps normalized to no decay that could be used as.a tool by plant staff along with a set of decay curves to quantitatively assess the plant status quickly following any abnormal occurrence.

Zone maps and decay curves have been developed in accordance with boundaries and radiation level criteria as discussed in 2.1.6b.1 for identifying problem areas, however, the following decay times were used in assessing anticipated potential personnel radiation exposure due to those operator actions required post LOCA.

For analyses of personnel exposures in vital areas outside the control room, radioactive decay equivalent to 10 minutes that is allowed for operator action was used as the minimum decay time.

Additional decay time was also allowed for the review of all those ECCS systems previously outlined which are used to recirculate water from the containment sump back into the

Section 2.1.6b - Page 7 containment. That decay time was 24 minutes, which is-consis-tent with the time- for initiation of recirculation as per FSAR Chapter 6.

b. Dilution Volume The volume used for dilution is important, affecting the calculations of dose rate in a linear fashion. The following

, , dilution volumes were used with the release fractions'and decay times. Listed above to arrive at the final source terms for the shielding reviews:

Source A: Containment free volume. The volume occupied by the ECCS water was neglected.

Source B: Reactor coolant system volume based on reactor coolant density at the operating temperature and pressure.

Source C: The volume of' water present at the time of recircula-tien (reactor coolant system and refueling water storage tank and safety injection tanks).

c. Sources used in piping and equipment for each system under review

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Section 2.1.6b - Page 8 In defining the limits of the connected piping subject to contamination listed below, normally shut valves were assumed to remain shut.

Containment Spray System: At the initiation of recirculation, source C was used.

High Head Safety Injection System: At the initiation of recirculation, source C was used.

Residual Heat Removal System: Source C was used for sump recirculation mode.

Sampling Systems: -The sources ~used in the shielding design review for sampling systems were as follows:

Containment air samples - Source A.

Reactor coolant sample - Source B.

Letdown System: The liquid source was B.

l B. The Shielding Design Review Methodology

1. Analylical Shielding Techniques i

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The previous sections outlined the rationale and assumptions '

j- for the selection of the systems that would undergo a

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Section 2.1.6b - Page 9 shielding design review as well as the formulation of the sources for those systems. The next step in the review process was to use those sources along with the standard point kernel shielding analytical techniques to estimate dose rates from those selected systems. For compartments containing the systems under review, estimates were made g for general area dose rate rather than to superimpose the maximum dose rate at contact with the surfaces of all individual components of that system in the compartment.

For corridors outside compartments, reviews were done to check the dose rate transmitted into the corridor through the walls of adjacent compartments. Checks were also made for any piping or equipment that could directly contribute to corridor dose rates; i.e. piping that may be running directly in the corridor or equipment / piping

, in a compartment that could shine directly into corridors with no attenuation through the cmpartment walls.

2. Accident Radiation Zone Maps One of the two principal products of this review is the series of accident radiation zone maps. These zone maps represent the correlation of the dose rates as estimated above with the required operator actions and resultant necessary accessibility to vital areas. By using these zone maps along with the decay curves, potential problem

I Section 2.1.6b - Page 10 areas were identified and reviewed. The results of these reviews are being used to formulate plant shielding modification recommendations.

C. Environmental Qualification Methodology Most organic materials used for insulation, seals, etc. are generally unaffected by integrated gamma doses below 100,000 rads (carbon). For the purpose of estimating integrated doses to safety related equipment near the system components analyzed in the Shielding design review, the following criteriun was used:

Those areas shown in the zone maps with a zone designation of VI, VII or VIII (dose rates between 500R/hr and 500,000 R/hr) were identified as potential problem areas. The rationale for this is that the total integrated exposure for sources listed in the selection of systems for shielding reviews under catagory A could exceed 100,000 rads integrated dose in about one or two days. This was ascertained by developing a corres-ponding set of integral energy release curves as a function of time for the same sources for which decay time curves were developed.

Those ateas identified as zones VI, VII or VIII are the subject of a co :inuing review to quilify the potential severity of -

any rad ation damage _to safety related equipment in that area.

Critics pieces of equipment identified within those areas are also the subject of individual reviews to establish their respective qualifications.

e 5ection 2.1.6b -~Page 11

, Details concerning the foregoing shielding review including specific systems.in the source. term, zone maps, decay curves',

and specific component locations, will be available at the.

, ~Farley plant for review and inspection.

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Section 2.1.7B - Auxiliary Feedwater Flow Indication to Steam Generators fer PWR's Required Action:

Provide a detailed description of Steam Generator Level Instrumentation Power Supply. Provide the Frequency _ Intervals of Calibration for Steam Generator Level and Auxiliary Feedwater Flow Instrumentation.

Response

Auxiliary feedwater injection lines to each steam generator are provided with flow indication. This flow indication is read out on the Main Control Board and is powered from the Class IE plant emergency power system. These flow instrument loops are testable. Redundancy requirements are met by qualified steam generator level instrumentation (Safety Grade).

Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements are met:

Position 1: Auxiliary feedwater flow indication to each steam generator shall satisfy the single failure criterion.

' Response: Local and control room indication of auxiliary feedwater flow to eacit of the steam generators is provided by flow orifices Q1N23FE-3229A, B and C in the auxiliary feedwater supply line, located just upstream of the auxiliary feedwater stop check valves. Instrumentation and controls

Section 2.1.7B - Page 2 for the system are shown on P & ID drawing D-175007. The auxiliary feedwater flow indication is backed up by three redundant safety grade narrow range steam generator level channels and one safety grade wide range steam generator level channel per steam generator which have control room readouts.

Position 2: Testability of the auxiliary feedwater flow indication channels shall be a feature of the design.

Response: Testing of the auxiliary feedwater flow indication is performed in accordance with the FNP preventative mainte-nance program on 18-month intervals by injection of a test signal at the primary sensor. The instruments are calibrated if the output signals do not meet the required accuracy for the instrument. The steam generator narrow range level channels are functionally tested every 31 days and calibrated every 18 months in accordance with FNP Technical Specification requirements. The steam generator wide range level channels are calibrated every 18 months in accordance with FNP Technical Specifications.

Position 3: Auxiliary feedwater flow instrument channels shall be powered from the vital instrument buses.

Response: The auxiliary feedwater flow instrumentation channels receive their power from the class IE vital instrument bus. The Bechtel Corporation drawing A-177076 gives a block diagram of the steam generator feedwater flow indication loop and the power supply-to the channels (B0P m

Section 2.1.7B - Page 3 Instrument Panel IJ) is indicated on drawing D-177024.

The steam generator narrow range channels receive their power from the class IE vital instrument buses. The power supply to the channels (process I&C cabinets numbers 1, 2 and 3) is indicated on drawing D-177024. Each cabinet contains one channel for each steam generator.

The steam generator wide range channels also receive their power from vital instrument buses as indicated on drawing D-177024 (process I&C cabinets numbers 5, 7 and 8).

Position 4: Each auxiliary feedwater channel should provide an indi-cation of feed flow with an accuracy on the order of i 10%.

Response: The present control grade system has transmitters with a 1 5% full scale accuracy. The Westinghouse power supply has a gain accuracy of ! .1% of full scale and the control board indicator has an accuracy of i 1.5% of full scale.

The additive accuracy of the flow loop is t 2.1% of full scale which is well within the 10% range. The steam generator level channels have an indication accuracy of i 4%.

The steam generator level and auxiliary feedwater flow transmitters are seismically and environmentally qualified. The auxiliary feedwater line flow transmitter cables will be rerouted as a result of jet impingement studies by January 1, 1981. This will meet all safety grade requirements.

It is the opinion of Alabama Power Company that the requirements of NUREG-0578 are satisfied of this section.

2.1.8.b Increased Range of Radiation Monitors 4 Required Action:

. Develop a procedure to estimate the release rate for the Steam Jet Air Ejectors and Main Steam Power Operated Relief Valves.

Response

A. Interim Methods for Determining Releases All major potential release points are piped'to the plant vent stack, which was addressed in the December 31, 1979, response.

The main condenser air ejector and the Atmospheric Relief Valves are addressed as follows:

1. Main Condenser Air Ejector

!. It is highly improbable that this monitor would go offscale 1

for any anticipated accident. Nevertheless, a procedure will be written by April 18, 1980 whereby, in the event that the monitor goes offscale, the main condenser air ejector will be monitored with a portable Beta / gamma survey instrament, and by means of a mR/hr vs.,uci/ml conversion table, prompt determination of release rates will be obtained. The proce-dure will provide a designated location on the discharge line of each main condenser air ejector where a technician will measure the dose rate. By taking a discharge flow rate measurement on an installed manometer, the current and/or total release to the environment may be quickly ascertained.

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Section 2.1.8.b - Page 2 These release rates will be calculated every 15 minutes or as of ten as required by the emergency director.

2. Atmospheric Steam Relief Valves

-Procedures will be written to provide locations for taking contact dose rate measurements with a portable survey

! instrument on each of the three (Unit #1) steam lines down stream of the Atmospheric Steam Relief Valves. The dose rates may be converted using tables provided to obtain s uci/ml. A flow rate will be obtained using a correlation between steam pressure and relief valve flowrate. With suci/ml and the flow-rate known, a release rate to the

! environment can be quickly determined. These release rates will be calculated every 15 minutes or as often as required by the emergency director. Procedures and capability to determine release rates will be available by April 18, 1980.

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.- a Section 2.1.9 - Reactor Coolant System Venting Required Action:

Document that Reactor Vessel Head Vent will meet the requirements in NRC letter of October 30,'1979.

Response

The Reactor Vessel Head Vent System (RVHVS) is designed to remove noncondensable gases from the reactor vessel and head area. The RVHVS is designed to vent in excess of one-half the reactor coolant system volume of hydrogen at 6500 F. in one hour from one of two available flow paths. The RVHVS is orificed to limit the blowdown from a break or inadvertently opened flow control valve downstream of the orifice to the capacity of one charging pump.

The system is operated from the control room and has control room indication of valve position. The system is completely safety grade and meets the single failure criteria for venting initiation and termination.

The reactor vessel head vent valves are powered by train A and B of the plant class IE D.C. distribution system and fail closed. The head vent valves in series are powered from the same train. This method of powering valves in conjunction with the fail closed design allows the RVHVS to meet the single failure criteria. Each RVHVS flow path consists of two (2) normally closed, normally de-energized valves. The two series valve arrangements eliminate the possibility of a spuriously opened flow path due to the spurious movement of one valve. The RVHVS will discharge into a well-ventilated area of the containment in order to ensure optimum dilution of combustible gases. Inside the containment, hydrogen can be recombined by means of the post-accident hydrogen recombiners.

The discharge point will be designed for adequate drainage of reictor

coolant in the cases of inadvertent discharges, i

Section 2.1.9 - Page 2 The vent piping will be supported to result in acceptable stresses on the RVHVS piping and the vent / vessel' connection resulting from seismic, thermal, dynamic, pressure and deadweight loads. The piping downstream of the fit'st anchors downstream of the second isolation valves may be non-nuclear safety; however, piping downstream of the second isolation valves must not transmit loads back to the Safety Class 2 piping. Each vent line will be seismically qualified.

As the potential exists for water flow through the vent line after the vessel has been refilled after venting, the vent system piping shall be supported for water flow up to the Safety Class 2 - non-nuclear safety boundary.

The PORVs will be used to vent the pressurizer.

The Reactor Coolant Leakage Detection System provides the capability of detecting the presence of significant radioactive or non-radioactive leakage from the reactor coolant loops to the containment atmosphere during no rmal operation. Variations in the particulate activity, gaseous activity, and specific humidity of the containment atmosphere above a preset level give pssitive indications in the control room to the reactor operators. These leakage detection provisions are sufficiently sensitive so that small increases in leakage rates can be detected while the total leakage rate is still below a value consistent with safe operation of the plant. The part.culate and gaseous activity are monitored by the containment air. particulate (R-11) and radiogas (R-12) monitors. Humidity is monitored b r the containment ' sir coolers' condensate level measuring system. lut increase in R-11, R-12, or containment' air cooler condensate level is an indication of reactor coolant leakage into containment.

Section 2.2.1.B - Shift Technical-Advisor Required Action:

Develop a formal communicacions system from the Operating Assessment Group to the Shift Technical Advisor (STA).

Response

Farley Nuclear Plant LER's and significant LER's of other plants are presently being routed for STA review and signature in accordance with existing plant procedures. In addition, when personnel perform-

ing operational assessment conclude that information exists which may be relative to the function of the STA, such information will be issued to the STAS. This requirement will be incorporated in plant procedures by March 31, 1980.

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Section 2.2.2.B - On-Site Technical Support Center Required Action:

Provide a description of the following:

1. Communications Systems

2. Radiation Monitoring Equipment.

3. Parameters Available to TSC Personnel

Response

1. A Temporary Technical Supporc Center (TSC) has been established which meets all interim TSC criteria as follows:

A. Description - The temporary TSC has been established in an 11' x 10' office on the west side of the Control Room (see attached Figure 1). This office is separate from the controls area and provides sufficient space for the Emergency Director and his staff to coordinate emergency activities. Though not required prior to January 1,1981, additional emergency response personnel could work from the Unit 2 controls area in the event of an emergency prior to Unit 2 licensing. The Unit 2 controls area is separated from the Unit 1 controls area by a security barrier.

B. Procedures for TSC Staffing and Support - FNP-0-EIP-0, " Emergency Organization and Control Room Access" has been issued. It specifies which plant personnel shall report to the TSC during an emergency and delineates their duties and responsibilities.

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. io Section 2.2.2.B - Page 2 C. TSC Communit.tions - The TSC contains the following communication equipment:

(1) NRC hot line (red phone)

(2) Two plant public addrecs system units for intra-plant communication (3) One intra plant phone capable of reaching other phones within the plant, other phones in the Southern Company operating system and the Alabama Power general office switchboard. From the gencral office switchboard, it can be connected to commercial phone facilities.

(4) Two pushbutton phones, each with the following features:

(a) Two commercial lines on the Graceba Co. Ashford, Alabama exchange (b) One commercial line on the GTE Dothan, Alabama exchange (c) One Alabama Power general office extension with the intra plant system capabilities discussed under C(3).

(d) One Daniel Construction Company of Alabama (DCCA) extension capable of reaching other DCCA extension on site and outside commercial lines on the Graceba Co. Ashford, Alabama exchange.

Control room communication can be made by intra plant line, public address line or by voice due to the close proximity of the TSC to the controls area. Communication with the operation support centers can be made by intra plant phone. ,

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• Section 2.2.2.B - Page 3 D. Radiation Monitoring and Action Levels - The TSC is habitable to the same degree as the main control room controls area for

postulated accident conditions. The TSC EVAC is common with the control room and includes a process monitor which automati-cally initiates recirculation mode on detection of high radiation.

An area monitor on the northwest corner of the Unit one controls area provides radiation monitoring and alarm. Action levels for retreat to the controls area are unwarranted since conditions will be identical to those in the TSC.

E. Access to Technical Data and Display of Plant Parameters - A complete set of P & ID's is maintained in the TSC for emergency use. Any other prints or technical data are readily available from the document control facility in the plant Service Building which may be reached by intra-plant phone from the TSC.

Plant data can be displayed in the TSC by relocating one of two Unit 1 controls area CRT computer readouts to the TSC or Unit 2 Controls Area. Cable has been fabricated and is in place to permit this. The CRT is mounted on a dolly and could be placed in service within 5 - 10 minutes giving access to any parameter monitored by the plant computer. Display format is as follows:

(1) One group of points is displayed and immediately below the group the most recent 16 alarms (2) Any one of 16 groups may be selected for display (3) A group may consist of up to 26 parameters selected by the computer operator

s ..o Section 2.2.2.B - Page 4 F. Procedures for Accident Assessment from the Control Room -

Since the temporary TSC is, for all practical purposes, part of the conurol room, this requirement is not applicable.

When the permanent TSC is established, procedures will be modified to reflect accident assessment from the temporary TSC facility should the permanent TSC become uninhabitable.

G. Long range plans for establishing an upgraded TSC are contained in APC's 12/31/79 NUREG-0578 response.

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