ML19257A503

Responds to NRC 790913 Ltr Re TMI Lessons Learned Task Force short-term Recommendations.Addresses & Commits to Implementation of NUREG-0578 800101 Requirements

ML19257A503
Person / Time
Site:	Farley
Issue date:	12/31/1979
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 8001040489
Download: ML19257A503 (39)

Text

4 Alabama Power Company 600 North 18th Street

• Post Omco Box 2641 Birmingham Alabama 35291 Telephone 205 323-5341 F. L Ct.AYTON, JR. M3h3m3 %gygf Senor Vice President the southern electrc system December 31, 1979 Docket No. 50-348 Mr. Darrell G. Eisenhut, Acting Director Division of Operating Reactors U. S. Nuclear Regulatory Commission .

Washington, DC 20555

Dear Mr. Eisenhut:

As required by your letter of September 13, 1979, Alabama Power Company submits Enclosure (1) documenting implementation of NUREG-0578 January 1,1980 requirements . In addition NUREG-0578 January 1, 1981 requirements are addressed with appropriate conceptual design. This response supplements responses of

, October 24, 1979 and November 21, 1979 on this subject.

As additional information is supplied by the Division of Operating Reactors regarding requirements in the areas of Lessons Learned or as further study by Alabama Power Company requires the commitments contained in Enclosure (1) will be anended.

Yours very truly,

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Mr. Darrel G. Eisenhut Page Two

References:

(1) NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommer.dations", 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 Requirements", ,

October 30, 1979.

cc: Mr. R. A. Thomas Mr. G. F. Trowbridge 1687 138

4 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 PUR's Pressurizer Heaters Power Supply Pressurizer Heater Back-up Groups A & B are capable of being powered from emergency buses. Procedures have been revised to make the operator aware of when and how to connect the required p nssurizer heaters to emergency buses.

These procedures address what loads can be shed from the emergency buses to provide -

power for the heaters. Training on the new procedures has been conducted.

Power Supply for Power-Operated Relief Valves and Block Valves No additional response required for this item.

I- Pressurizer Level Indication Power Suoply No additional response required for this item.

6 1687 139

Section 2.1.2 - Performance Testing for BWR and PUR Relief and Safety Valves

, By letter dated December 17, 1979, Mr. William J. Cahill, Jr., Chairman of the EPRI Safety and Analysis Task Force submitted " Program Plan for the Performance Verification of PWR Safety / Relief Valves and Systems", December 13, 1979.

Alabama Power Company' considers the program to be responsive to the requirements Presented in IMREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short Tern Recommendstions" dated July,1979, Item 2.1.2 which recommended in part, commit to provide performance verification by full scale prototypical testing for all relief and anfety valves. Test conditions shall include two phase slug flow and subcooled +

liquid flow calculated to occur for design basic transients and accidents."

The EPRI Program Plan provides for a completion of the essential portions of the test program by July,1981. Alabama Power Company vill be participating in the EPRI Program to provide program review and to supply plant specific data as required.

1687 140

6 Section 2.1.3A - Direct Position Indication of Relief and Safety Valves The Pressurizer Power Operated Relief Valves have stem mounted limit switches which operate red and green indicating lights in the valve control switch on the Main Control Board (MCB). These switches provide positive open and shut indication for these valves. The present switches are not environmentally qualified; there-fore, they will be replaced with environ =entally qualified limit switches during a plant shutdown scheduled to begin March 14, 1980. At this shutdown, required alarm equipment will be installed.

The Pressurizer Safety Valves do not have positive position indication at the present time. The installation of positive position indication equipment for these safety velves is scheduled for a plant shutdown to commence March 14, 1980.

This installation uses stem mounted limit swi.tches to provide open and shut r- indication and provide the associated alarm function. The formal design proposal will be provided to Alabama Power in January,1980 with delivery of equipment (critical path) expected by the end of February 1980 or early March 1980. The installation is expected to require two (2) weeks with the plant in cold shutdown.

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Section 2.1.3B - Instrumentation for Detection of Inadequate Core Cooling In PWR's

a. Procedures and Description of Existing Instrumentation The Westinghouse Owners' Group, of which Alabama Power Company is a member, has performed analyses as required by Item 2.1.9 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 sub-mittal 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 and operator training required to recognize the existence of inadequate core cooling and restore core cooling based on existing instru-mentation. Alabama Power Company has incorporated the key considerations of this instruction into our LOCA procedures, and has provided training to the operators in this area.

, . - . b. Subcooling Meter Alabama Power Co=pany will install a primary coolant saturation meter that meets the requirements of NUREG 0578. This saturation meter will be L2 stalled during a plant outage scheduled to begin March 14, 1980. Alabama Power Company ;

I will monitor the existing hot leg RTD's, core exit thermocouples, and reactor coolant pressure instrumentation to compute temperature margin to saturation i

until installation of the saturation meter is completed.

c. Additional Instrumentation to Indicate Inadequate Core Cooling The submittal references in section (a) above described the capabilities of the core exit thermocouples in determining the existence of inadequate core cooling conditions and their superiority in some instances to the loop RTD's for measuring true core conditions. Other means of determining the approach to or existence of inadequate core cooling could be:

1687 I U-

Section 2.1.3.3 Page 2

1. Reactor vessel water level

2. Incore detectors L

3. Excore detectors

4. Reactor coolant pump motor currents

5. Steam generator pressure ,

A discussion of the possible use of these measurements are addressed below.

-- The use of incore movable detectors to determine the existence of inadequate core cooling conditions appears doubtful. The detectors could be driven into the tcps of the incore thimbles, which are located at the top of the core, following an accident in which concern for inadequate core cooling exists.

The problem comes in the lack of sensitivity of the detectors to very low neutron levels and changes that would occur due to core uncovery. Camma

--- detectors could perhaps be employed, but they suffer from similar sensitivity problems, and the fact that ga=ma levels in the fuel region change insignifi-cantly between the covered and uncovered condition. As a result, it does not appear worthwhile to pursue incore movable detectors as a means of determining inadequate core cooling conditions.

The use of excore detectors has been mentioned as a possibility in responding to core uncovery. The only detectors which would have the required sensitivity are the source range monitors. Since the intermediate and power range monitors are not sensitive enough to the low level changes resulting from vessel voiding.

The use of the source range monitors will be investigated further as part of the more indepth study of inadequate core cooling being performed by the Westing-house Owners' Group. However, their use is probably limited to those instances when significant voiding exists in the downcomer region, since normally water in the downcomer would effectively shield the detectors from the core region whether voids existed or not.

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Section 2.1.3B Page 3 Reactor coolant pump motor current, which could be indicative of core voiding, is inappropriate for a reliable means of determining inadequate core cooling, since a loss of off-site power or pump trip due to a LOCA blowdown shut the pumps down.

I Steam generator pressure, which already exists, is useful in the case where heat transfer from primary to secondary is interrupted due to loss of natural circulation. This, however, does not satisfy requirements to indicate the approach to inadequate core coolins, nor does it indicate the true condition of the core.

Reactor vessel water level determination is the most promising of the items discussed to provide additional capability of determining the approach to and the existence of inadequate core cooling. Several systems t -~ for determining water level are under review by the Westinghouse Owners' Group. A conceptual design of one system is given below:

Vessel Level System Description Af ter examining many different methods and principles for determining the water level in the reactor vessel, a basic delta pressure measure-ment from the bottom of the vessel to the top of the vessel appears to provide the most meaningful and reliable information to the operator.

One of the reasons for choosing this system is that the sources of potential errors are better known for this system than for any o'ther new or untested system.

The attached figure shows a simplified sketch of the proposed vessel level instrumentation system. The bottom tap of the teatrument would use a thimble of the incore movable detector system either at the seal table or in the thimble below the vessel. Use of the thimble as part of the incore flux

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Section 2.1.3B Page 4 monitoring would not be lost. The flux thimble guide tube would be tapped below the vessel and an instrument line connection made. The instrument line would have an isolation valve and slope down to a hydraulic coupler connected to a sealed reference leg. For connection at the seal table, a special fitting would be utilized which would be connected to an isolation valve and sealed reference leg. The sealed reference leg would go to the differential pressure transmitter located at a higher elevation above the expected level of containment flooding. A similar sealed leg would go to the top of the vessel and penetrate the head using the vent line or a special connection on a spare RCC cechanism penetration. Two trains of vessel level instrucentation would be provided.

The behavior of the signal generated by this 1cvel instrument under normal r --, and accident conditions is being evaluated. The usefulness of this instru-cent to provide an unambiguous indication of inadequate core cooling is being evaluated as part of Item 2.1.9. The potential errors and accurancy of a final system configuration are being evaluated to assess its usefulness to provide information to the operator for proper operation of a vessel venting system and for nor=al water level control during periods when the primary system is open and a water level may exist in the vessel. The connection of the level system to the vessel head should be designed to be compatible with the head vent system. Operation of the vent system should not upset all indications of vessel level. This can easily be avoided by using a separate instrument cap or by using more than one location.

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Section 2.1.4 Containment Isolation Provisions for PUR's and BliR's Alabama Power Company has previously documented containment isolation design provisions meeting NUREG-0578 requirements. In addition Alabama Power has reviewed conreif e nt isolation design against a generic study generated by the Westinghouse DfI Owners Group.. Farley containment isolation design meets all requirements of this study. A list of the systems identified as essential and non-essential is attached.

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Section 2.1.4 - Essential Systems

1. Normal Letdown

2. Excess Letdown / Seal Water Return

3. Pressurizer Sa=ple

4. Hot Leg Sample

5. Contain=ent Air Sample

6. Normal Charging

7. Residual Heat Removal (normal suction)

8. Residual Heat Removal (ctmt sump recirculation)

9. High Head Safety Injection

10. Low Head Safety Injection

11. Containment Spray

12. Contain=ent Spray (ctut sump recirculation) 13'. Service Water to Ctmt Coolers

'~~ 14. Reactor Coolant Pu=p (RCP) Seal Water Supply

15. Contain=ent Pressure

16. Instru=ent Air Supply

17. Component Cooling To RCP Thermal Barrier

18. Post-Accident Air Sample

19. Post-Accident Ctut Vent 1687 148

Section 2.1.4 - Non-Essential Systems

1. Accumulator Test Lines

2. Accumulator Make-up

3. Accu =ulator Sa=ple

4. Nitrogen to Accumulators

5. Nitrogen to Pressurizer Relief Tank (PRT)

6. PRT Make-up

7. Reactor Coolant Drain Tank (RCDT)

8. Contain=ent Differential Pressure

9. Component Cooling to Excess Letdown and RCDT Heat Exchanger

10. Component Cooling from Excess Letdown and RCDT Heat Exchanger

11. RCDT Vent

12. Demineralized Water to Containment

13. Service Water to RCP Motor Air Coolers e -

14. Containment Purge and Mini-Purge

15. Containment Su=p Pu=p Discharge

16. Contain=ent Su=p Pump Sample Recurculation

17. Charging Pu=p/ Residual Heat Removal Relief

18. Valve Discharge to PRT

19. Containment Leak Rate Test

20. Service Air 1687 149

Section 2.1.5.A - Dedicated Penetrations for External Recombiner or Post-Accident External Purge Systen No additional response required for this iten.

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1687 150

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Section 2.1.5 C - Capability to Install Hydrogen Reconbiner at Each Light Water Nuclear Power Plant Alabama Power has reviewed and upgraded the energency procedure for operation of electric hydrogen reconbiners located inside containnent. Training on this procedure revision has been conducted.

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1687 151

2.1.6. A - Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PUR's and BWR's Alabama Power Company has instituted a leak reduction program for the systems identified in our November 21, 1979 submittcl. This leak reduction program consists of the following actions:

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

2. Maintenance has been performed on packing in liquid system valves identified as requiring work during leak tests.

3. Maintenance has been perfor=ed on a pump seal identified during leak tests.

4. Maintenance has been performed on gas system valves identified as requiring work during leak tests.

Leakage rates for the systems identified in our November 21, 1979 submittal is as follows:

1. High Head Safety Injection (Recirculation Portion Only) - 13 drops / minute

• - 0.20 gallons / hour (attributable to charging pump leakage)

2. Low Head Safety Injection (Recirculation Portion Only)

Train A: 1.59 GPM (highly conservative)*

Train B: 0.15 GPM

• Train A leakage measurement is suspect due to the possibility of a tempera-ture decrease during testing.

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 tha of charging pumps reported in item 1.

5. Reactor Coolant Sampling System - no measurable leakage

6. Containment Spray System (Recirculation Portion Only).

120 drops / minute

• 1.8 gallons / hour

. 1687 152

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Section 2.1.6.A Page 2

7. Radioactive Waste Gas System - Portions of syst.em which could receive high level gaseous weste except Recombiner A (out of service) were

" snooped" for leakage. Detected Icakage has been corrected. Current r.easurable leakage is zero SCDf. O. combiner A will be leak tested after returned to service.

As part of the preventive maintenance program on the scoped systems, leak rate measure = ants will be performed periodically at intervals not to exceed each refueling outage.

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

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Section 2.1.6B - Design Review of Plant Chielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Post-Accident Operations Alabana Power Company has completed a design review of plant shielding using the source tems postulated by the NRC. The needed design changes resulting from this review have been identified and will be implemented by January 1,1981.

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Section 2.1.7A - Automatic Initiation of the Auxiliary Feedwater System The requirements of NUREG-0578 are met by the Farley Nuclear Plant as docu:::ented by our suhnit'd of Deceber 14, 1979.

1687 155

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Section 2.1.7B - A xiliary Feedwater Flow Indication to Steam Generators for PWR's ,

i Auxiliary feedwater injection lines to each steas generator are provided '

with flow indication. 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 contained in our submittal of December 14, 1979.

The auxiliary feedwater line flow indicators will be seismically and environ-4 mentally qualified by January 1,1981. This will meet all safety grade requirements.

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Section 2.1.8A - Post-Accident Sampling Capability Methods of obtaining a reactor coolant (i.e. pressurized and unpressurized) and contain=ent atmosphere highly radioactive samples have been established. The reactor coolant sample is obtained by using the existing sample system with modi-fications to allow for remote operation of the sample valves and transportation of the shielded sample to a shielded lab area for chemical analysis and sample dilution. The containment at=osphere sample is obtained by modifying the existing monitoring system to allow particulate and radiciodine samples to be taken using small gas volumes while minimizing personnel exposure. The particulate filter and silver zeolite radiciodine sample will be transported in a shielded container to a remote temporary area for Ge(L1) Camma Ray Spectroscopy analysis. Procedure modifications for handling and analysis of samples, plcnt modifications, and a design review of the sample system have been completed. Additional infor=ation on r-these modifications is contained in the attached sketches and system operation de-scription.

The modified, operational sampling systems meet January 1, 1981 requirements.

i 1687 157

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Attachment (Section 2.1.8. A) - Description of Post-Accident Sanpling Systems i

A. Reactor Coolant (Pressurized and Unpressurized)

The Unit I reactor coolant sampling system has been modified by addition of a post-accident sampling panel and shielded sample pass-through (Figure 1) .

A schematic of the post-accident sampling panel is shown in Figure 2.

The above system has the capability for remotely taking a pressurized sample (RCS) and an unpressurized sample (RHR/ Containment Sump). Based upon the

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results of the shielding design study with subsequent modifications (2.1.6.B) and time studies for drawing the samples, the estimated whole body or extremities radiation doses to any individual will not exceed 3 and 18 3/4 rems, respectively.

B. Vent Stack Effluent The Unit I vent stack air particulate and noble gas monitor (RE 21 & 22) has been modified by installation of a post-accident particulate and iodine sampler, f .-.

as well as addition of a septum for drawing a gas sample. (Figure 3). Subsequent to an accident the normally used radiation monitor (RE 21 & 22) will be valved out and the effluent flow directed through the post-accident sampler for a specified time period by operation of the remote control panel. Based upon the results of the shielding design study with subsequent modifications (2.1.6.B) and sampling time studies, the estimated whole body or extremities radiation doses to any individual will not exceed 3 and 18 3/4 rems, respectively.

C. Containment Air The Unit I containment air particulate and noble gas monitor (RE 11 & 12) has been modified by installation of a post-accident particulate and iodine sampler, as well as addition of a septum for drawing a gas sample, (Figure 4). Subsequent to an accident the normally used radiation monitor (RE 11 & 12) will be valved '

out and the containment air flow directed through the post-accident sampler for i687 158

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4 Section 2.1.8B - Internim Procedures for Quantifying High Level Accidental Radioactivity Releases To measure. the noble gas radioactive effluent release, Alabama Power Company has mounted a Jordan Rad-Gun on the vent stack at the 175 foot level. This monitor uses a shielded isokinetic sampler to obtain a representative sample and uses lead shielding to reduce background interference, and provides a continuous readout at the monitor. The monitor is DC powered with a battery life in excess of 30 days.

Monitor readings are provided to the control room via verbal communications using any one of the three existing communications systems: (1) plant phone system; (2) sound powered phone system; or (3) plant public address. The monitor has a range of 0.01 MR/HR to 10,000 R/HR over an energy range of 80 KEV to 1.2 MEV.

Calibration is done at installation and annually using a Cs 137 calibration source.

Predetermined calculational methods are used to convert the radiz. tion level reading f - to radioactive effluent release rate.

The measurement of radiciodine and particulate effluents is accomplished by a modification to the normal vent stack monitor RE 21/22 which allows the f

collection of small gas samples on particulate filters and silver zeolite cartridges, which are analyzed using a Ce(L1) Gamma Ray Spectroscopy system. Procedures for operation of the system have been developed to provide calculational methods to  !

determine release rates.

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l Section 2.1.8B1 - High Range Effluent Monitor To meet the January 1, 1981 requirements for a high range noble gas effluent monitor, Alabama Power Company has ordered an Eberline SPING-4 sampler. This sampler will monitor the vent stack effluent and has a range of 10-7 pCi/cc to 105 nCi/cc by using multiple ranges. The monitor readout will be located in the control room and will be powered from a vital instrument bus. Calibration is by

_ use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage. Procedures will be developed for use, calibration of the system, and dissemination of release rate information.

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i Section 2.1.8B2 - High Range Effluent Radiciodine and Particulate Sampling and Analysis To meet the January 1,1981 requirements for a high range effluent radioiodine and particulate sampling system, Alabama Power Company has ordered an Eberline SPING-4 sampler. This sampler provides the capability to monitor effluent radio-activity in the form of noble gases, radioiodines, and particulates by use of individual channels for each type of radioactivity. The sampler will monitor the

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vent stack with the monitor readout located in the control room area. The sampler will be powered from a vital instrument bus. The particulate channel uses a filter paper in the air streamwhich is counted by a beta scintillation detector with an alpha detector fersubtraction of the radon-thoron daughter activity contribution.

The range of the channel is 2.6x105 counts per minute per microcurie on the paper for Csl37 The radiciodine channel monitors a silver xcolite cartridge in the

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air stream with a sensitivity of approximately 80K CPM per uCi of 1131 Both the particulate channel and the radiciodine channel use external sources for calibration and can be compensated for background radiation. Calibration is performed upon installation and at intervals not exceeding each refueling outage. Procedures will be developed for use o. the sytem, calibration of the monitor, and dissemination of release rate information.

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4 Section 2.1.8B3 - High Containment Radiatiot A thorough evaluation has been made of all the equipment available at this time and projected to be available during the next year. This evaluation was made by a point-by-point comparison of all published material and af ter considerable dise;ssions uith each vendor. None of the equipment designs are finalized yet.

None of the designs have completed the environmental qualification tests or per-formance tests. It is quite likely that the design will have to be modified, perhaps considerably, in order to meet both the NRC performance requirements and environmental qualifications simultaneously. Some of the equipment has been 4 shown not to meet the performance requirements, while other fail the environ-mental qualifications. Some information in vendor brochures has been shown by cests to datemto be incorrect. Vendors have had to retract certain published reports on perfor=ance and revise their figures drastically in the non-conservattne direction.

i __,

Of particular note, none of the present monitor designs can meet the temperature requirements for LOCA conditions and also temperature ratings in brochures have been reduced. As a result, there'is nct equipment that is known to be able to satisfy both sets of requirements (performance and environmental conditions) simul-

._ taneously. Theretore, a commitment at this time to buy specific equipment from a particular vendor with the presently proposed designs with known and possible flaws or shortcomings is definitely premature. It is not likely that simple design changes will make any of the designs acceptable. Therefore delivery dates and perhaps even design concepts may have to be changed drastically.

The present technical specifications require calibration of monitors during each refueling outage. Calibration of these high range containment radiation level monitors will require removcl of the detectors from the containment to a shielded calibrator in a calibration room. At this time there are no calibrators available to calibrate the upper decades of these monitors. For safety purposes, it would be better to have the monitors calibrated at some centrally located facility offsite.

1687 L67

, Section 2.1.8B3 Page 2 Alabama Power commits to installing a containment radiation monitor with the range specified in NUREG-0578 upon availability of a production type monitor.

1687 168

.,n. _ - . .- -

Section 2.1.8C - In-Plant Iodine Instrumentation Alabama Power Company has a portable monitoring system available which uses an iodine silver xeolite sampler and single channel analyzer. Emergency procedures have been revised to address the use of this portable monitor. Appropriate shift personnel have been trained on the use of this analyzer.

By January 1,1981, Alabama Power Company will have the capability of purging these samples of entrapped noble gases by the use of nitrogen gas and analysis by Ge(L1) gana-ray sepctroscopy in a low background counting facility.

4 f ===

1687 169

Section 2.1.9 - Transient and Accident Analysis Analyses of small break loss-of-coolant accidents, symptoms of inadequate core cooling and required actions to restore core cooling, and analysis of transfent and acci!ent scenarios including operator actions not previously analyzed are being performed on a generic basis by the Westinghouse Owners' Group, of which Alabama Power Company is a member. T L small break analyses have been completed and were reported in WCAP-9600, which was submitted to the Bulletins and Orders Task Force by the Owners' Group on June 29, 1979. Incor-

~~

porated in that report were guidelines that were developed as a result of small break analyses. These guidelines have been reviewed and approved by the B&O .

Task Force and have been presented to the Owners' Group utility representatives in a seminar held on October 16-19, 1979. Following this seminar, Alabama Power has revised emergency procedures and trained personnel on these procedures. Revised procedures and training are in place in accordance with

' ~~

rhe requirement in Enclosure 6 to Mr. Eisenhut's letter of September 13, 1979, and Enclosure 2 to Mr. Denton's letter of October 30, 1979.

The work required to address the other two areas--inadequate core cooling and other transient and accident scenarios--has been perforned in conjunction with schedules and requirements established by the Bulletins and Orders Task Force. Antlysis related to the definition of inadequate core cooling and guide-lines for recognizing the symptoms of inadequate core cooling based on existing plant instrumentation and for restoring core cooling following a small break LOCA were submitted on October 31, 1979. This analysis is a less detailed analysis than was originally proposed, and will be followed up with a more extensive and detailed analysis which will be available during the first quarter of 1980. The guidelines and training are in place as re s a ired by the B&O Task Force.

With respect to other transient accidents contained in Chapter (15) of the J. M. Farley FSAR, the Westinghouse Owners' Group has performed an evaluation

_ _ . . . 1687 170

Section 2.1.9 Page Tuo of the actions.which occur during an event by constructing sequence of event trees for each of the non-LOCA and LOCA transients. From these event trees a list of decision points for operator action has been prepared, along with a list of information available -to the operator at each decision point. Following this, criteria have been set for credible misoperation, and time available for operator decisions have been qualitatively assessed. The information developed was then used to test Abnormal and Emergency Operating Procedures against the event sequences and determine if inadequacies exist in the A0P's and E0P's. The

. results of this study will be. provided to the Bu!1etins and Orders Task Force March 31, 1980 as required. Alabama Power Company will revise procedures and complete retraining within 3 months of receipt of new guidelines.

The Owners' Group has also provided test predictions analysis of the LOFT L3-1 nuclear small break experiment. This analysis was provided on December 15, 1979, in accordance with the schedule established mutually with the Bulletins i _- and Orders Task Force.

. i687 171

Containment Pressure Indication (ACRS)

The present containment pressure indication provides continuous redundant indication in the Main Control Room and has an indication range of -5 psig to 60 psig. Additional monitoring capability with Control Room indication having a range of 0 to 210 psig will be installed by January 1, 1981. This additional monitoring equipment will be safety grade and meet the design provisions of Regulatory Guide 1.97, Revision 1.

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1687 172

L-- l . . . . - .. .. -. - . - - - . . . - . . . . . _

Containment Water Level Indication (ACRS)

No additional response required for this item.

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I 1687 173

Containment Hydrogen Indication (ACRS)

No additional response required for this item.

4 f

1687 174

Reactor Coolant System Venting Alabama Power Company will install a Reactor Vessel Head Vent System by January 1, 1981. The following is a description of the system proposed:

System Description

The Reactor Vessel Head Vent System is designed to remove gases from the reactor coolant system via remote manual operations from the Main Control Room.

The Reactor Vessel Head Vent System 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 reactor coolant in the case of inadvertent discharges.

The Reactor Vessel Head Vent System flow diagram is shown in Figure 1.

The system arrangement provides for venting the reactor vessel head by using only safety grade equipment. The system mainly consists of 1-inch piping with

' ~~

four safety Class 2 " fail closed" isolation valves. The system utilizes all normally closed valves. The isolation valves No. 1 and No. 2 are powered by Train A and the isolation valves No. 3 and No. 4 are powered by Train B. The system is designed such that any single active failure will not prevent vessel gas venting nor prevent venting isolation. The system also provides the necessary manual venting functions during vessel filling operations.

The system connects to the reactor vessel head at the existing vent pipe with redundant flow paths through orifices. The orifice restricts the flow rate from a pipe break downstream of the orifice to within the makeup capacity of one centrifugal charging pump. All piping and equipment between the orifice and the discharge point is of Safety Class 2 as defined by ANSI N18.2A.

The reactor vessel head vent system isolation valves will be supported from the seismic support platform. The system can be disconnected downstream of the second isolation valve to accommodate refueling. In this manner the necessary flanged connections will be cutside of the reactor coolant pressure boundary.

1687 175

System Description (Con'd)

~

All piping and valves upstream of the flanges will remain integral with the reactor vessel head at all times.

0 "

1687 176

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TO CONTAINMENT ,

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REACTOR

. VESSEL l HEAD NOTES:

1. EXISTING VENT LINE Figure : 1 Flow Diagram of the R'eactor Vessel Head Vent System 1687 177

, . . . - - . . _ _ . _ . . _ _ . ~ . _ --

Section 2.2.1A - Shif t Supervisor Responsibilities

1. Corporate management has issued and will periodically reissue a management directive that emphasizes the primary management responsibility of the Shift Supervisor for safe operation of the plant under all conditions on his shif t and that clearly establishes his comand duties.

2. Plant procedures have been revised to assure that the duties, responsi-bilities, and authority of the Shif t Supervisor and control room operators

- are properly defined to effect the establishment of a definite line of command and clear delineation of the comand decision authority of the Shif t Supervisor in the control room relative to other plant management personnel.

3. Training programs for Shift Supervisors emphasize and reinforce the responsi-bility for safe operation and the management function the Shif t Supervisor is to provide for assuring safety.

r- 4. Corporate management has reviewed the administrative duties of the Shift Supervisor. Administrative functions that detract from safe operation have been delegated to other personnel.

1687 178

. . ~ ~ t . . . . .. . - ...... ... _ _ --........ ..-- ... _ . . -

Section 2.2.lB - Shift Technical Advisor i

An individual on-shif t has been designated to serve in a dedicated capacity as Shift Technical Advisor during emergency conditions. This individual reports to the Shift Supervisor during emergency conditions and serves in a technical advisory capacity. Personnel designated for this function are not part of the minimum shift complement as specified in the Technical Specifications. Designated individuals have other duties during normal plant conditions.

Alabama Power has placed Shife Technical Advisors on shif t.

By January 1,1981, the Shif t Technical Advisor will have received additional training ,

and periodic retraining commensurate with NUREG-0578 requirements.

The operating experience assessment function is being performed by the plant's g li system performance group which is composed of supervisory, engineering and technical i personnel. This group is not functionally a part of the plant operations group.

8 Procedures have been revised to incorporate this function.

179 1687

Section 2.2.lC - Shift and Relief Turnover Procedures Plant procedures for shif t and relief turnover have been revised as neces-sary to assure that comitments in our October 24, 1979 submittal are in effect.

A system has been established to evaluate the effectiveness of the turnover procedure by requiring operations quality assurance audits on a periodic basis of this function.

~

f W 1687 180

Section 2.2.2A - Control Room A" cess Plant procedures have been developed to assure that the commitments in our October 24, 1979 submittal are in effect.

f -

1687 181

Section 2.2.2B - On-Site Technical Support Center An interim Technical Support Center has been established to meet commitments made in our October 24, 1979, sabmittal. The plant Emergency Plan and appropriate implementing procedures have been revised to incorporate the role and location of the interim Technical Support Center as part of the emergency preparedness improvements.

The following is a description of the proposed long term onsite technical ,

support center. It covers the anticipated manning requirements, document storage, monitoring systems, communication, habitability and structural arrangement. The center has been designed to be used far post accident and normal plant conditions.

Conceptual Design

' -~

1. Location - The TSC is located in the Auxiliary Building, Elevation 155'-0" immediately north of the Unit 2 control room area. Personnel will normally enter and exit via the Primary Security Access Point and Unit 1/ Unit 2 control room area. Passage through the control room will not interfere with the operators who are manning the control boards. An emergency or alternate exit is from the north side of the TSC and through an exit door in the northwest wall of the Unit 2 Auxiliary Building.

2. Size and Room Layout - The TSC is designed to accommodate 25 people for the performance of post-accident monitoring, evaluation of plant status, coordination of damage assessment and emergency actions, interface with the NRC and emergency operations center and onsite/offsite communications.

Figure 1 shows a proposed room' layout.

An overall space 22 feet x 65 feet, with a 9 foot ceiling height, has been provided. Roco layout is as follows:

1687 IB2

System 2.2.2B Page 2 A. Monitoring Area - Includes CRT, trend line printer and alarm type-writer for each unit, and provision for communication with control room, emergency operations center, NRC and the State of Alabama.

B. Planning and Coordination Area - Includes desks, reference tables, files for plant procedures and manuals and monitors for TV cameras in the control room. Tables and additional folding chairs and movable partitions will be provided so that the area can be quickly set up f-r usage. Phones will be provided for full communication 4

capability. A chalk board and projection screen will also be located here. Figures 2 and 3 describes the TV monitoring system.

C. Document Room - Includes files for microfilm, drawings, data sheets, indexes, microfilm reader and printer.

P. Storage Area - Includes cabinets for paper, office supplies, etc.

E. Stairway Enclosure - The TSC area is presently used for access to the control room mechanical equipment room. A structural steel stairway and open equipment hatch at Elevation 175'-0" are provided for this purpose. The enclosure will isolate the stairway and hatch

~~

from the TSC while still alloaing access during normal plant conditions.

3. Construction - Structural and architectural design criteria of the TSC will be similar to that used for the office areas in the control room, except that it will not meet seismic category I requirements.

4. Activation - Activation is in accordance with the Emergency Plan. Equipment will be capable of displaying vlt-l ilant parameters frod the time an accident begins TV camers and monitors will be turned on as needed.

} ()h7

-- ~. - ,_.

- System 2.2.2B Page 3

5. Monitoring and Display Equipment - The philosophy of data monitoring is that a person would be located at the main control room computer console and communicate with the TSC via sound powered phones. TSC personnel would of the two request display of the desired data in the TSC. One/CRTs normally located in the. main control room (one per unit) would be disconnected, rolled into the TSC, and hooked up upon activation of the TSC. A two pen recorder located in the TSC would have the capability to trend two parameters. The line printers (one per unit) now located in the computer room (auxiliary bldg.121') would be moved to the TSC. Parameters could be trended on the line printers providing a hard copy of data for review.

4

6. Communications - Co==unications will be provided to contact the NRC, State of Alabama, and APCO General Office. Scund powered phones will be provided between the TSC and Main Control Room. An intercom system will be available between the TSC, the three Operations Support Centers, and the Emergency Operations Center.

, __ Commercial and APCO Telecommunications lines will be available for redundant offsite communications. An APCO security radio will also be located in the TSC.

7. Habitability - The TSC will not be exposed to any areas in the Auxiliary Building that could contain highly radioactive sources post-accident.

~~

The ventilation system will include a deep-bed charcoal filter to remove airborne contamination, and it will have the capability of pressurzing the TSC area and recirculating the room air through the charcoal filter.

A permanent radiation monitor will be provided to continuously indicate radiation dose rates and airborne activity. A radiation alarm in the makeup air supply duct will automatically initiate room pressurization and recirculation. The HVAC system is described in detail under system description.

1687 I84

System 2.2.2B Page 4

8. Fire Protection - The TSC is enclosed by 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated walls with Class A airtight firedoors. The present area, Room 2452, is provided with fire and smoke detectors and a wet pipe sprinkler system. .A water hose cabinet and CO 2 hose reel are presently located within the TSC area.

These will be moved to immediately outside of the TSC. One CO2 and one dry chemical fire extinguisher are provided inside of the TSC. Emer-gency breathing apparatus and spare air bottles are now provided for the control room personnel.

9. Electrical Power - Normal and emergency provisions are as follows:

A. Closed Circuit TV and Computer Hardware - Regulated Voltage Distribution Panels lA and 2A, Channel 1, or Panels 1B and 2B, Channel 2.

B. HVAC - 600 Volt MCC's IF and 1G. These are shared MCC's and can

, _ _ be powered from the diesel generators upon LOSP.

C. Wall Receptacles - 120/208 Volt Distribution Panels 1R, 1S, IEE or 1FF. These are shared panels and can be powered from the diesel generators upon LOSP.

D. Lighting - Fluorescent - 26 fixtures, 4 - 40W lamps each. 277 V-AC

~-

from MCC - IF or 1G (Unit 1) and MCC-2CC or 2DD (Unit 2).

Emergency - 25 watt.

8 hr. battery packs; 2 - 2 head, 2 - 1 head. Tied to MCC-2DD (Unit 2) ,

"B" train.

Wall Receptacles - Fed from MCC-2E (Unit 2), " normal" train. 125V AC, 15A.

HVAC System Description This system is composed of one (6.4 ton) A/C unit with air cooled condenser, one emergency filtration unit equipped with a separate fan, two six inch butter-fly valves for positive isolation from contaminated outside air, associated ductwork and controls. Refer to Figure 4

. 1687 185

System 2.2.2B Page 5 A positive pressure is maintained in this area by using outside air which is vented directly to the outside from the occupied area. Upon receipt of a high radiation signal from a sensor in the outside air duct, butterfly valves and dampers change position and the outside air plus an equal quantity of re-circulated air is directed through the emergency filter, thus maintaining the pressure within the conditioned space and establishing a clean-up mode of operation simultaneously.

The te=perature is maintained by a room thermostat which cycles either the compressor or the electric heating coil within the A/C unit upon demand.

The filtration unit contains a 50% orefilter, a 99.97% HEPA. filter plus a carbon filter respectively.

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