ML19330C694

Forwards Util Info Re TMI-2 Action Plan Per NRC

ML19330C694
Person / Time
Site:	Farley
Issue date:	08/01/1980
From:	Clayton F ALABAMA POWER CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8008110351
Download: ML19330C694 (41)

Text

Alabama Power ComDany 600 Nortn t e:n street Post Cece Box 2641 Ehrmngham. Alaoama 35291 Telephone 205 323-5341 A

AlabamaPower

e scu:rwn ew:x system August 1, 1980 Docket No. 50-364 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C.

20555 Attention: Mr. A. Schwencer JOSEPH M. FARLEY NUCLEAR PLANT - UNIT 2 TMI - 2 ACTION PLAN Gentlemen:

In accordance with Mr. Tedesco's letter dated July 16, 1980, Alabama Power Company submits the attached information concerning the TMI - 2 Action Plan for Farley Nuclear Plant - Unit 2.

If you have any questions, please advise.

Yours very truly, n

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a-t K / s.:,zunc-r e.

F. L. ' C1 ayton,.Jr.

RWS:sc Attachment cc: fir. R. A. Thomas Mr. G. F. Trowbridge Mr. L. L. Kintner fir. W. H. Bradford SON s

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POSITION

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1.b By January 1,1981, complete the installation of the high range noble gas monitors ll.F.1(f) and provide all the information required in Item 2.1.8b, Sections 1.B and 2.B. gisen in our November 9, 1979 let-ter. Your response should contain calculational methods for converting instrument readings to release rates based on exhaust air flow and radionuclide spectrum distribution, correction for background radiation, j

and procedures for dissemination of'information.

Clarify your intent to meet the January 1,1981 requirements for the atmospheric steam re-1.ief/ safety valves and main condenser air ejector release monitoring.

See NUREG-0694 "TMI Related Requirements for New Operating Licenses",

June 1980.

RESPONSE

i To meet the January 1,1981 requirements for a high range noble gas effluent monitor, Alabama Power Company has ordered an Eberline SPING-a range of 10-}s sampler will monitor the vent stack effluent and has 4 sampler. Th uCi/cc to 103 uCi/cc using multiple ranges for noble 4

gases.

The monitor readout will be located in the control room area and will be powered from a vital instrument bus.

Calibration is by use of an external calibration source and is perfomed.upon installation i

and at intervals not exceeding each refueling outage.

Procedures will l

be developed for use, calibration of the system, and dissemination of release rate information. The SPING-4 will also meet the January 1, 1981 requirements for a high range effluent radiciodine and particu-late sampling system. The particulate channel uses a filter paper in the l

air stream which is counted by a beta scintillator detector with an alphe detector for subtraction of the radon-thoron daughtgr activity contri-bution. The sensitivity of the channel is 2.6 x 10 counts per minute i

per microcurie of Csl37 on the paper. The radiciodine channel monitors mately 80K counts per minute per mci of Ig with a sensitivity of approxi-a silver zeolite cartridge in the air str Both the particulate channel and the radioiodine channel use external sources-for calibra-tion and can be compensated for background radiation. The calculational method will be to directly covert #Ci/cc indicated by the monitor to curies / min. discharge rate by multiplying by the flow rate and the ap-t propriate conversion factor.

It is highly improbable that the main condenser air ejector monitor will go off scale for any anticipated accident. A procedure has been implemented l

for monitoring the air ejector exhaust line with a portable gamma survey instrument using a mR/hr vs. pCi/ml conversion table. With JJCi/ml and the flow rate known, a release rate to the environment can be quickly determined.

This procedure will be utilized if the existing monitor goes off scale.

It is Alabama Power's position that such action will result l

in prompt determination of release rates as required by NUREG 0694.

It.is highly improbable that an accidental release would occur through the main steam atmospheric relief valves. A procedure has been implemented for monitoring the atmospheric relief lines with portable ganna

survey instruments using a mR/hr vs. pCi/ml conversion tabla. With pCi/ml and the flow rate known, a release rate to the environment can be quickly determined. This procedure will be. utilized if the necessity arises to bleed steam to the atmosphere due to condenser being unavailable following an accident.

It is Alabama Power's position that such action would result in prompt determination of release rates as required by NUREG-0694.

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POSITION 3

Before full power operation prior to January 1,1981, provide a descriptive summary of the interim procedures for obtaining, handling and analyzing the reactor coolant and the containment atmosphere. By January 1, 1981, provide a description and final system design of the new post accident sampling panel, and modifications to the sample handling and counting facilities to achieve analysis within the time specified in Item 2.1.8a'given in the Nov-ember 9, 1979 letter.

See NUREG-0694 "TMI Related Requirements for New Operating Licenses", June 1980.

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RESPONSE

Reactor Coolant Sampling System Before full power operation of Farley Nuclear Plant Unit #2, Alabama Power Company will have a post-accident RCS sampling system installed and operating which complies with the short term requirements of NUREG 0578.

It is the opinion of Alabama Pcwer Company that this system also meets long term recommendations of NUREG 0578. The Attachment "FARLEY NUCLEAR PLANT POST ACCIDENT SAMPLING SYSTEM UNIT #2" fully describes the system, its theory of operation and its capabilities. Based upon the results of the shielding design study with subsequent modifications (2.1.6.8) and sampling time estimates, the estimated whole body or extremities radiation dose to any individual will not exceed 3 and 18-3/4 rems, respectively.

Containment Atmosphere Samoling System The Unit #2 containment air particulate and noble gas monitor (RE-11 & 12) has been modified by installation of a post-accident particulate and iodine sampler which also provides a septum for drawing a gas sample (See figure 5 attached). Subsequent to an accident the normally used radiation monitor RE-ll 312 will be valved out and tFc containment air flow directed through the post-accident sampler for a specified period of time 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 estimates, the estimated whole body or extremities. radiation doses to any individual will not exceed 3 and 18-3/4 rems, respe?'ively.

Sample Handling ana Counting Facili ties The post accident sampling system and the containment atmosphere sampling system both 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 samples which include the use of lead glass windows and manipulating apparati that will insure that no analyst will receive exposures exceeding 3 and 18-3/4 rem to the whole body and extremities respectively.

Farley Nuclear Plant will have two (2) counting rooms (one per unit).

In the event of an accident, it is anticipated that the background radicactivity level in the non-affected unit's counting room will be low enough to perform the measurements.

If the background radioactivity level is too high in the non-affected unit, the analysis will be performed at the water treatment plant or the emergency operations facility upon its completion.

Both of the above systems will enable sampling within one (1) hour and sample analysis within an additional hour.

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FORWARD During normal operations, the primary sample system is typically aligned to the RCS flow path.

This alignment is also the desired path in the event of a radiological accident. The design of the Post Accident Sample System, however, is convenient for normal routine gas sampling as well as post accident sampling; therefore additional flow paths have been provided to enhance the sampling capability as well as to reduce contamination and exposure during normal plant operations. The Post Accident Sampling System has been designed to align on loss of power to the RCS flow path to accommodate the post accident requirements. That is, with all power off the valves will automatically align to the RCS flow path.

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THEORY OF OPERATI0tt 1.0 A pressurized liquid sample is circulated through a 40cc. stainless steel sample bomb until it is representative of the system being sampled (see Figure 1). The sample is then isolated and expanded into an evacuated 250ml. glass bomb where degassing occurs. When the pressure on the liquid in the 40cc. sample bomb suddenly drops below the boiling point, the sam-pie rises in the glass bomb due to expanding steam bubbles and begins to boil vigorously. The vigorous boiling agitates the sample. Coupled with the reduced pressure, agitation facilitates the degassing. Cooling of the sample is accomplished by enthalpy and by conduction from the liquid to the glass sampling bomb. When the sample is cooled below the boiling point, the boiling stops. As the last steam bubble colapses with cessa-tion of boiling, the liquid is drawn back into the 40cc. sample bomb. The recession of the liquid from the glass sample bomb provides a clearly de-fined end point to the degassing operation. When degassing is complete argon is injected into the sample to normalize the pressure to one atmos-phere.

(When the temperature of the circulating liquid is low, the liquid may not boil when depressurized.

If boiling does not occur, argon may still be injected very slowly until the pressure reaches one atmosphere.

The argon bubbling through the depressurized liquid will agitate and sweep the gases out). An aliquot of the gas sample is then drawn off by micro-syringe and injected into a Gas Chromatography for analysis. The liquid portion of the sample may then be transported to a laboratory in a lead shield for chemical analysis. The VCT gas space may also be sampled.

By-passing the liquid process, one can normalize the VCT gasses to one at-mosphere pressure and draw an aliquot in a micro-syringe for analysis in a Gas Chromatograph.

OPERATI0ft 2.0 Referring to Figure 1, valves CV4,CV6 should be energized. All other valves should be deenergized. Open vlave TV1 or TV2 very slowly until the pressure gage reads about 10 psig. Then open or energize DV1. This will drain any residual liquid from the system.

DVl should then be de-energized, and CV7 and CV8 and CV10 should be energized.

The vacuum pump should then be energized and the system should be evacuated to about 30 inches of mercury.

Valves CV7, CV8, and CV10 should then be deenergized and the vacuum pump should be deenergized. The system at this point is purged and ready to take a sample.

2.1 Selecting one of the several available sample lines by proper alignment of valves CV1, CV2, and CV3 a small amount of liquid is circulated through the previously evaculated 40cc. sample bomb by deenergizing CV4-6.

After a sufficient time has elapsed to insure a representative sample, valves CV4-6 are energized and the samole is expanded into the glass bomb. When the boiling stops, valve TV1 or TV2 may be opened very slowly to allow the pressure to rise to zero on the gage or to about one atmosphere (14.7 psia).

If boiling does not occur, argon should be added even more slowly by open-ing either TV1 or TV2 until the pressure on the gage is zero.

2.2 The gases may not be drawn off by micro-syringe for analysis in a Gas Chromatograph.

2.3 The liquid sample may be collected by rolling the lead shield containing a sample vial (see Figure 2) into the core drilled hole in the sample room wall until it reaches the stop where it aligns with the sample spout. Lower the sample spout into the mouth of the vial by pulling the Bowden wire con-trol marked SPOUT all the way to the stop.

Then open TV1 or TV2 very slowly until the pressure gage reads about 4 psig (this will provide the necessary pressure to push the sample into the vial).

Then valve CV5 should be energized and the sample will be dumped into the sample vial.

Valve CV5 should then be deenergized, the spout retracted by pushing the the spout control back into its nomal position and the lead shield rolled out of the hole or.to the transport cart which can then transport the sample to the lab for analysis.

2,4 In reference to Figure 1, several specific points concerning the assembly of the Post Accident Sample System are clarified as follows:

1.

All valves are shown either in the deenergized position or manually closed.

2.

The nomal or deenergized condition lines up the RHR/RCS Hot leg for sample recire.

3.

CV-1 and CV-2 are mechanically interlo:ked to energize and deenergize simultaneously.

4.

CV-4 and CV-6 are mechanica~lly interlocked to energize and deenergize simultaneously.

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POSITION 4

Your " Response to TMI - 2 Action Plan" does not explicitly respond to Part 4, " Dated Requirements" of NUREG-0694 "Ti1I-Related Requirements for New Operating Licenses" and does not respond at all to Item I.A.2.1, "Immediate Upgrading of Operator and Senior Operator Training and Qualification," Item I.A.2.3 " Administration of Training Programs for Licensed Operators, " and Item I. A.3.1, " Revised Scope and Criteria for Licensing Exams."

Provide an explicit response to Part 4 of NUREG-0694, giving a description of the program, status of the program, and statement of intent to meet the required completion date for the program.

References to other portions of your response will be acceptable.

1 RESPONSE TO POSITION 4 I.A.1.1 SHIFT TECHNICAL ADVISOR Position The Shift Technical Advisor shall have a technical education, which is taught at the college level and is equivalent to about 60 semester hours in basic subjects of engineering and science, and specific training in the design, function, arrangement and operation of plant systems and in the expected response of the plant and instruments to normal operation, transients and accidents including multiple failures of equipment and operator errors.

This requirement shall be met by January 1,1981.

(See NUREG-0578, Section 2.2.lb, and letters of September 27 and November 9, 1979.)

Response

Shift Technical Advisor requirements were addressed on page 1 of A'.aoama Power Company's TMI-2 Action Plan response dated June 20, 1980.

In this submittal, APC0 committed to implement these requirements by January 1,1981.

This program will include mitigation of accidents involving a degraded core, as outlined in our response dated July 29, 1980, item II.B.4.

I.A.2.1 IMMEDIATE UPGRADING 0F OPERATOR AND SENIOR OPERATOR TRAINING AND QUALIFICATION 4

Position Applicants for SR0 license shall have 4 years of responsible power plant experience, of which at least 2 years shall be nuclear power plant experience (including 6 months at the specific plant) and no more than 2 years shall be academic or related technical training.

Certifications that operator license applicants have learned to operate the controls shall be signed by the highest level of corporate management for plant operation.

These requirements will be met on or after May 1,1980.

(See March 28, 1980 letter.)

Revise training programs to include training in heat transfer, fluid flow, thermodynamics, and plant transients.

This requirement shall be met by August 1,1980.

(See March 28, 1980, letter.)

Response

Item 1:

Applicants for an SR0 License will have 4 years of responsible power plant experience, of which at least 2 years will be nuclear power plant experience, including 6 months of. site specific experience, and no more than 2 years shall be academic or related technical training.

Item 2: Certifications that operator license applicants have learned to operate the controls are signed by the highest level of corporate management for plant operation which is the Vice President-Nuclear Generation.

Item 3: Training Programs have been revised to include training in heat transfer, fluid flow, thermodynamics, and plant transients.

Alabama Power Company meets all these requirements.

I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS FOR LICENSED OPERATORS Position 1.

Training instructors who teach systems, integrated responses, transients and simulator courses shall successfully complete a SR0 examination.

Applications shall be submitted by August 1, 1980.

(See March 28, 1980 letter. )

2.

Instructors shall attend appropriate retraining programs that address, as a minimum, current operating history, problems and Changes to procedures and administrative limitations.

In the event an instructor is a licensed SRO, his retraining shall be the SR0 requalification program.

Programs shall be initiated by May 1,1980.

(See March 28, 1980, letter.)

Response

Item 1:

Permanent plant instructors involved in training programs for licensed operators will be SRO licensed or will make application for licenses by August 1,1980.

Instructors obtained from other sources will be SR0 licensed, will make application for licenses, will be SR0 cold license certified by August 1, 1980, or will make application for NRC Instructor certification prior to August 1, 1980; certification is expected to include satis-4 factory completion of a NRC senior operator examination and adherence to INPO standards.

Item 2:

Instructors will attend license retraining program.

All licensed SR0 instructors will attend the SR0 requalification program.

Instructor License Numbers:

Sam Gates SOP-2920-1 Tom Horne 50P-3270-1 Mike Eidson SOP-3380 Randy Wiggins SOP-2927-1 Lee Williams 50P-2930-1 1

1.A.3.1 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMS i

Position i

Applicants for operator licenses will be required to grant permission to the NRC to inform their facility management regarding the results of examinations.

Centents of the licensed operator requalification program shall be modified to include instruction in heat transfer, fluid flow, thermo-dynamics, and mitigation of accidents involving a degraded core.

These requirements shall be met by May 1,1980.

(See March 28, 1980, letter.)

The criteria for requiring a licensed individual to participate in accelerated requalification shall be modified to be consistent with the new passing grade for issuance for a license.

This requirement shall apply to all annual requalification examinations conducted after March 28, 1980.

(See March 28, 1980, letter.)

Requalification programs shall be modified to require specific reactivity control manipulations.

Normal control manipulations, such as plant or reactor startups, must be performed. Control manipulations during abnormal or emergency operations shall be walked through and evaluated by a member of the training staff.

An appropriate simulator may be used to satisfy the requirements for control manipulations.

This requirement shall be met by August 1,1980.

(See March 28, 1980, letter.)

Response

Item 1: Applicants for operator licenses will grant permission to the NRC to inform plant management regarding the results of exami-nation.

Item 2:

Content of the licensed operator requalification program has been modified to include instruction in heat transfer, fluid flow, thermodynamics, and mitigation of accidents involving a degraded core, as outlined in our response dated July 29, 1980, item II.B.4.

Item 3:

The criteria for requiring a licensed individual to participate in accelerated requalification have been modified to be consis-tent with the new grading criteria.

Item 4:

Requalification programs have been mcdified to require specific reactivity control manipulations.

Normal control manipulations such as plant or reactor startups will be performed.

Control manipulations during abnormal or emergency operations are r

- walked thecugh and evaluated by.a member of the training.

J staff. An appropriate simulator is used -to satisfy the require-ments for control manipulaticn to the extent that simulation is

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I.C.1 SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION Position Analyze the design basis transients and accidents including single active failures and considering additional equipment failures and operator errors to identify appropriate and inappropriate operator actions.

Based on these analyses, revise, as necessary, emergency procedures and training.

This requirement was intended to be completed in early 1980, however, some difficulty in completing this requirement has been experienced.

Clarification of the scope and revision of the schedule are being developed and will be issued by July 1980.

It is expected that this requirement will be coupled with the I.C 9, Long-Tenn Upgrading of Procedures.

(See NUREG-0578, Sections 2.1.3b and 2.1.9, and letters of September 27 and November 9,1979.)

Response

Short-Term Accident Analysis and Procedure Revision requirements were addressed on page 15 of Alabama Power Company's TMI-2 Action Plan response dated June 20, 1980.

A completion schedule of this item can only be obtained after review of the NRC clarification scheduled for issuance during July,1980.

II.B.1 REACTOR COOLANT SYSTEM VENTS Position Install reactor coolant system and reactor vessel head high-point vents that are remotely operable from the control room.

This requirement shall be met before January 1,1981.

(See Enclosure 4 to letters of September 27 and November 9, 1979.)

Response

Reactor Coolant System Vent requirements were addressed on page 65 of Alabama Power Company's TMI-2 Action Plan response dated June 20, 1980.

The reactor vessel head vent system design is incomplete but is in progress.

Alabama Power Company expects design completion and material delivery in order to support installation by January 1,1981.

Specifically, the design is scheduled for completion by August 8,1980 and all material is currently scheduled for shipment by September 1,1980.

Material Status Description Scheduled Shioment Date Valves 8-15-80 Pipe & Fittings 9-1-80 Terminal Blocks Onsite Hoffman Enclosures Onsi te Handswitch Onsite Indicator 9-1-80 J

i 11.8.2 PLANT SHIELDING Position Complete modifications to assure adequate access to vital areas and protection of safety equipment following an accident resulting in a degraded core.

This requirement shall be met by January 1,1981.

(See NUREG-0578, Section 2.1.6b and letters of September 27 and November 9,1979.)

Response

A review, which identified areas needing improved shielding, has been completed.

This shielding review identified areas requiring access for operation of essential safe shutdown equipment.

Shielding modifications developed in the shielding review will be implemented by January 1,1981.

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II.S.3 POSTACCIDENT SAMPLING Position Complete corrective actions needed to provide the capability to promptly obtain and perform radioisotopic and chemical analysis of reactor coolant and containment atmosphere samples under degraded-core conditions without excessive exposure.

This requirement shall be met by January 1,1981.

(See NUREG-0578, Section 2.1.8a, and letters of September 27 and November 9, 1979.)

Response

Methods of obtaining highly radioactive samples of the reactor coolant (i.e., pressurized and unpressurized) and containment atmosphere have been established for use when Unit 2 becomes operational. The reactor coolant sample will be obtained by using the existing sample system with modifications 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 atmosphere sample will be obtained by modifying the existing monitoring system to allow particulate and radioiodine samples to be taken using small gas volumes while minimizing personnel exposure.

The particulate filter and silver zeolite radiciodine sample will be trans-ported in a shielded container to a remote temporary area for Ge(Li) gamma ray spectroscopic analysis.

Procedure modifications for handling and analysis of samples, plant modifications, and a design review of the sample system are being completed.

The procedures and plant systems modifications described above will be completed by January 1,1981.

Design and material for the postaccident sampling system is available to support installation by January 1,1981.

II.D.1 RELIEF AND SAFETY VALVE TEST REQUIREMENTS Position Complete tests to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.

This requirement shall be met by July 1,1981.

(See NUREG-0578, Section 2.1.1, and letters of September 27 and November 9, 1979.)

Response

Relief and Safety Valve Test requirements were addressed on page 26 of Alabama Power Company's TMI-2 Action Plant Responses dated June 20, 1980, and July 17, 1980.

Subsequent to these submitals, Alabama Power Company's letter of July 23, 1980 from fir. F. L. Clayton, Jr. to Mr. A. Schwencer, committed to provide appropriate test data for safety and relief valves in-stalled at Farley Nuclear Plant by July 1,1981.

II.E.1.2 AUXILIARY FEEDWATER INITIATION AND INDICATION Position Upgrade, as necessary, automatic initiation of the auxiliary feedwater system and indication of auxiliary feedwater flow to each steam generator to safety-grade quality.

This requirement shall be met by January 1,1981.

(See NUREG-0578, Sections 2.1.7a and b, and letters of September 27 and November 9,1979.)

Response

Auxiliary Feedwater Initiation and Indication requirements were addressed on page 30 of Alabama Power Company's TMI-2 Action Plant response dated June 20, 1980. Design and material are available for installation by January 1, 1981. Prior to exceeding 0% power, Alabama Power Company will provide power supply train separation for the Auxiliary Feedwater flow control valve solenoid valves.

Desian and Material Status Conceptual design for power supply train separation is currently being re-viewed by Alabama Power Company. After design is finalized, material requirements will be determined.

II.E.4.1 CONTAINMENT DEDICATED PENETRATION Position Install a containment isolation system for external recombiners or purge systems for postaccident combustible gas control, if used, that is dedicated to that service only and meets the single-failure criterion.

This requirement shall be met before January 1,1981.

(See NUREG-0578, Section 2.1.5a and 2.1.5c, and letters of September 27 and November 9,1979.)

Response

Containment dedicated penetration requirements were addressed on page 32 of Alabama Power Company's TMI-2 Action Plant response dated June 20, 1980.

All applicable requirements have been met by Alabama Power Compiny as indi-cated in this description.

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II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUtiENTATION Position Install continuous indication in the control room of the following parameters:

a.

Containment pressure from minus 5 psig to three times the design pressure of concrete containments and four times the design pressure of steel containments; b.

Containment water level in PWRs from (1) the bottom to the top of the coritainment sump, and (2) the bottom of the containment to a level equivalent to 600,000 gallons of water; Containment water level in BWRs from the bottom to 5 feet above the normal water level of the suppression pool; c.

Containment atmosphere hydrogen concentration from 0 to 10 volume percent; 8

d.

Containment radiation up to 10 Rad /hm e.

Noble gas effluent from each potential release point from normal concen-5 trations to 10 pCi/cc (Xe-133).

Provide capability to continuously sample and perform onsite analysis of the radionuclide and particulate effluent samples.

This instrumentation shall meet the qualification, redundancy, testability and other design requirements of the proposed revision to Regulatory Guide 1.97.

This requirement shall be met before January 1,1981.

(See NUREG-0578, Section 2.1.8b, and letters of September 27 and November 9, 1979.)

Response

a.

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 -5 to 225 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.

I The containment pressure indication design is available and all equip-ment is expected to be delivered in order to support installation by January 1, 1981.

b.

The present design has two level transmitters that will be installed to monitor postaccident narrow range containment sump level. The present design provides for both indication and recording capability in the main control room.

The design for the narrow range ;ontainment level is complete.

All material is currently scheduled to be shipped by November 7,1980 in order to support installation by January 1,1981.

Wide Range containment water level (bottom of containment to 600,000 gallons) have been installed.

c.

Two independent, redundant systems for containment hydrogen monitoring are provided in the present design with a range of 0 to 10 percent hydrogen concentration. The design of these systems follows, as applicable, the requirements for safety-related protective systems and meets the requirements of IEEE 279-1971.

The output signal of the analyzers are indicated at the analyzer panel location and are recorded and alarmed in the main control room.

Each system is supplied electrical power from independent and redund;nt Class lE power supply.

The system meets the single failure criteria and remains operable under postulated accidents. Any single failure in one hydrogen monitoring system does not affect its reoundant and independent counterpart.

The system described above currently meets the January 1,1981, require-ments.

d.

Alabama Power Company has ordered redundant Victoreen 875 Detector System 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. These locations ensure the monitors are not protected by massive shielding and that they will provide a reasonable assessment of area radiation conditions inside containment.

1.

Each detector is designed to measure gamma radiation.

7 2.

The range of each detector is 1 R/hr to 10 R/hr for photon radiation.

3.

The energy response is -15% at 80 key and 8% from 100 key to 3 Mev.

4.

The calibration frequency will be at a maximum interval of 18 months. At this time, we intend to return the monitors to the vendor for calibration.

S.

The containment high radiation monitor is scheduled for installation l

by January 1,1981.

The design for the high range containment radiation monitor is incomplete but is scheduled to be complete by September 1,1980.

Alabama Power Company expects that all equipment will be shipped by Sep-i tember 15, 1980.

Design and material will be available for installation by January 1, 1981.

e.

To meet the January 1,1981, requirements for a high range noble gas effluent monitor, Alabama Power Company has ordered an Eberline SPING-4 range of 10-}s sampler wilg monitor the vent stack effluent and nas a sampler. Th pCi/cc to 10 xCi/cc by using multiple ranges for noble gases.

The monitor readout will be located in the control room area 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 by January 1,1981,.for use, calibration of the system, and dis-semination of release rate infomation.

To meet the January 1,1981, requirements for a high range effluent radiciodina and particulate sampling system, Alabama Power Company has ordered an Eberline SPING-4 sampler.

This sampler provides the capability to minitor effluent radioactivity in the form of noble gases, radioiodines, and particulates by use of individual channels for each type of radio-activi ty.

The sampler will monitor the 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 stream which is counted by a beta scintillation detector with an alpha detector for subtraction of the radonthorog daughter activity con-tribution.

The range of the channel is 2.6 x 10 counts per minute per microcurie on the paper for Cs137 The radiciodine channel monitors a silver zeolite cartridge in the air stream with a sensitivity of approximately 80k CPM per DCi of I 131 Both the particulate channel and the radiciodine channel use external sources for calibration and can be compensated for background radiation.

Calibration will be performed upon installation and at intervals not exceeding each refueling outage.

Procedures will be developed for use of the system, calibration of the monitor, and dissemination of release rate information.

The design for the high range noble gas effluent monitor is incomplete but is scheduled for completion by August 8,1980. All equipment is scheduled to be shipped by October 15, 1980.

To measure the noble gas radioactive effluent release, Alabama Power Company will mount a Jordan rad-gun on the vent stack at the 175-foot level prior to receipt of an operating license.

This monitor will use a shielded isokinetic sampler to obtain a representative sample and lead shielding to reduce background interference; this will provide a continuous readout at the monitor.

The monitor will be DC powered with a battery life in excess of 30 days.

Monitor readings will be 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

? MeV.

Calibration will be done at installation and annuallyusingaCs{N calibration source.

Predetermined calculational methods will be used to convert the radiation level reading to radioactive effluent release rate.

The measurement of radiciodine and particulate effluents will be accomolished by a modification to the nomal vent stack monitor RE 21 & 22 which allows the collection of small gas samples on particulate filters and silver zeolite cartridges, which are analyzed using a Ge(Li) gamma ra spectroscopy system.

Procedures for operation of the system have been developed to provide calculational methods to determine release rates.

These modifications will be completed prior to receipt of an operating license.

To measure the release rates from the main condenser air ejector, procedures have been developed to utilize a portable beta / gamma survey instrument and a

mR/hr vs. pCi/cl conversion table to promptly determine the release rates in the highly improbable event the installed monitor goes offscale.

The procedures provide a designated location on the discharge line of each main cond?nser air ejector where a technician will measure the dose rate.

By using this dose rate and the discharge flow rate measurement, the current and/or total release to the environment may be quickly ascertained. These release rates will be calculated every 15 minutes or as often as required by the emergency director.

Procedures have been written to obtain the release rates from the atmospheric steam relief valves.

These procedures provide for taking contact dose rata measurements with a portable survey instrument on each of the three (Unit 2) steam lines down stream of the atmospheric steam relief valves. The dose rates may be converted using tables provided to obtain Ci/ml. A flow rate will be obtained using a correlation between steam pressure and relief valve flow rate.

With pCi/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.

Material Status Description Scheduled Shioment Date (a) Containment Pressure Transmitter / Sensor 12-15-80 Indicator 8-8-80 (b) Containment Level Level Switch / Convertors 11-1-80 (narrow range)

Recorder 8-8-80 (c) Hydrogen Monitoring N/A (d) Containment Radiation Radiation Monitor 9-1-80 (e) High Range Noble Sampler / Analyzer 10-15-80 Gases, Radioiodine and Particulate Effluents l

l

II.F.2 INADEQUATE CORE COOLING INSTRUMENTS Position Install, if required, additional instruments or controls needed to supplement installed equipment in order to provide unambiguous, easy-to-interpret indication of inadequate core cooling.

This requirement shall be met by January 1,1981.

(See NUREG-0578, Section 2.1.3b, and letters of September 27 and November 9, 1979.)

Response

In Alabama Power Company's letter from F. L. Clayton, Jr., to A. Schwencer dated July 17, 1980, the commitment to install a level system and the development of a schedule for installation, testing and operation was addressed. The connitments in this letter will be formally transmitted during the week of. August 4, 1980.

1

---r-

J III.A 1.2 UPGRADE EMERGENCY SUPPORT FACILITIES Position Prt, vide radiation monitoring and ventilation systems, including particulate and charcoal filters, and otherwise increase the radiation protection to the onsite technical support center to assure that personnel in the center will not receive doses in excess of 5 rem to the whole body or 30 rem to the thyroid for the duration of the accident.

Provide direct display of plant safety system parameters and call up display of radiological parameters.

For the near-site emergency operations facility, provide shielding against direct radiation, ventilation isolation capability, dedicated communications with the onsite technical support center and direct display of radiological and meteorological parameters.

This requirement shall be met by January 1, 1981 (See NUREG-0578, Section 2.2.2b and 2.2.2c and letters of September 27 and November 9,1979, and April 25, 1980.)

Response

Emergency Support Facility requirements were addressed on page 51 of Alabama Power Company's TMI-2 Action Plan response dated June 20, 1980, where Alabama Power Company committed to meet the January 1,1981 'date.

Material Status Description Scheduled Shioment Date 1.

Hvac Filtration Unit 8-15-80 2.

Closed Circuit Television System 8-22-80 NOTE: Equipment delivery dates do not take into account major construction effort required for facility completion.

Schedule completion of 1-1-81 is projected.

III.D.3.3 IN-PLANT RADIATION MONITORING Position Provide the ' equipment, training, and procedures to accurately measure the radiofodine concentrations areas within the plant where plant personnel may be present during an accident.

This requirement shall be met before January 1, 1981.

(See NUREG-0578, Section 2.1.8c, and letters of September 27 and November 9, 1979.)

Response

In-Plant Radiation Monitoring requirements were addressed on page 61 of Alabama Power Company's TMI-2 Action Plan response dated June 20, 1980.

Alabama Power Company currently meets these requirements as stated in this response.

position 6.

Your response to Item II.F.1 " Additional Accident Monitoring Instrumentation" provices a description of high-range radiation monitors for containment atmosphere but does not give their location.

Provide the location of the monitors on plant layout drawings.

The monitors should be located in a manner so as to provide a reasonable assessment of area radiation conditions inside containment.

Monitors should not be placed

.1 areas which are protected by massive shielding.

Response

The plant layout drawing, Figure 6, provides the approximate locations of the high-range radiation monitors for containment atmosphere.

These monitors are located approximately six (6) feet above the operating deck such that they are not protected by massive shielding and will provide a reasonable assessment of area radiation conditions inside containment t

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Position 7.

Your response to Item II.B.2, " Plant Shielding" describes the source terms used in evaluating shielding but does not adequately address all the requirements.

Provide a summary of the shielding design review required by our letter dated November 9,1979, implementing the Lessons Learned Item 2.1.6.b of NUREG-0578, and provide a description of the results of this review.

Include in your description:

a.

source terms used in the evaluation (NUREG-0578 specified that source tems in Regulatory Guide 1.3,1.4 and 1.7 be used).

b.

systems assumed to contain high levels of radioactivity in a post-accident situation including, but not limited to, residual heat removal, safety injection, CVCS, demineralizers, charging systems, reactor coolant filter, seal water filters sample lines, liquid radwaste systems, gaseous radwaste systems, and standby gas treat-ment systems.

If any of these systems or others that could contain high radioactivity were excluded, explain why such systems were excluded.

You should verify that field run piping and indirect radiation (such as shine over shield walls) were included in the

analysis, c.

specify areas where access was considered necessary for vital system operation after an accident. Your evaluation of areas to determine the necessary vital areas should include but not be limited to, con-sideration of the control room, Technical Support Center, Operational Support Center, recombiner hookup and control stations, hydrogen purge control stations, containment isolation reset control area, sampling and sample analysis areas, manual ECCS alignment area, motor control centers, instrument panels, emergency power supplies, security center and radwaste control panels.

If any of these areas were not considered areas where access was necessary after an accident, explain why such areas were excluded.

d.

designation of the codes used for analysis, such as ORIGEN, IS0 SHIELD, QUAD or others.

I e.

the projected doses to individuals for necessary occupancy time in vital areas.

i f.

a brief description of the proposed plant modifications resulting from the design review and confirmation that these modifications will be complete by January 1,1981.

Response

7a.

Source Tems Used for Plant Shieldino Evaluation l

The following release fractions, based on Regulatiory Guides 1.4 and 1.7, were us d as a basis for determining the concentrations for the shielding j

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 These release fractions 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 sheilding 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 inte-grated 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 sump. Furthermore, cursory analyses have indicated that the halogens dominate all shielding requirements and that contributions to the total dose rates from ncble gases are negligible for the purposes of a shield-ing design review.

The release fractions outlined above are, however, only the first step in modeling the source terms for the activity concentrations in the systems under review.

The important modeling parameters, decay time and dilution volume, obviously also affect any shielding analysis. The following sections outline the rationale for the selection of val.ues for these key parameters, a.

Decay Time For the first stage of the shielding design 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 the plant staff along with a set of decay curves to quantitatively assess the plant status quickly following any abnormal occurrence.

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 cnalyses 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 that are used to recirculate water from the containment sump back into the containment. That decay time was 24 minutes, which is consistent with the time for initiation of recirculation as per FSAR Chapter 6.

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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 tenns 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 recirculation (reactor coolant system + refueling water storage tank +

safety injection tanks).

c.

Sources Used in Piping and Equipment for Each System Under Review 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 inituation of recirculation, Source C was used.

- High Head Safety Injection System: At the initiation of recirc-ulation, Source C was used.

- Residual Heat Removal System:

Source C was used for sump recir-culation mode.

- Sampling 5ystems:

The sources used in the shielding design review for sampling systems were as follows:

Containment Air Sample - Source A Reacto. coolant Sample - Source B

- Letdown System: The liquid source was Source B.

7b. As discussed in the TMI Action Plan Response, the shielding study con-sidered systems based on the following criteria:

Selection of Systems for Shieldino Review The criteria applied in selection of plant systems used in the shielding review resulted in several classifications of systems as discussed below:

Cateoory A (Recirculation Systems}

The first group of systems are those required by plant design to mitigate a design basis loss of coolant accident and which might l

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contain highly radioactive sources in excess of the current design basis. A first priority safety concern is to ensure that operation of these systems containing a significant source will not adversely impact operator or equipment functions required outside the con-tainment. Therefore, the following systems have been adequately addressed by tha existing plant shielding design:

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

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

Category B (Extensions of Containment Atmosohere)

In addition to systems listed above, there are other systems or portions of systems which would contain radioactivity by virtue of their connec-tion 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 signifi-cant radioactive sources required by this special analysis.

Nevertheless, 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.

- Those portions of the containment ventilation systems external to the containment up to the first closed isolation valve which could contain the atmosphere from the containment.

- Those portions of the sampling system used to obtain a contain-ment atmosphere sample.

Category C (Liouid Samoles)

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 selected for this shielding review, Category D (Letdown)

The following portion of the letdown system has been selected for analysis:

- That portion of the letdown system from the reactor coolant system past the Letdown Heat Exchanger up to the inlet val.ves to the

Letdown Demineralizers. The flow path from the letdown heat exchanger, into the VCT and into the suction o.f the charging pump, was also considered.

The liquid and gaseous radwaste systems were not included in the analysis.

The gaseous system was eliminated since the reactor vessel head vent would be used for de-gassing 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 rather than the letdown system and VCT, eliminate the need to consider the liquid waste processing system.

The effects of field run piping were included in the shielding analysis.

The determination of radiation zones throughout the plant were analyzed for both direct radiation and sources of indirect radiation.

The review found that there were no partial height shield walls bordering corridors or stairways which require access.

In addition, all radiation zones were reviewed for the effects of streaming around labyrinth structures.

7c. Access Areas The following areas in the auxiliary building have 'oeen designated by the plant operators as being required for post-accident access:

Area Occupancy Period Control Room 24 hr./ day Technical Support Center 24 hr./ day Health Physics Area 24 hr./ day Passageway to Unit 1 (2402) 1 hr./ day i

Hallway 2409 1 hr./ day Electrical Penetration Rooms I hr. (approximately 1 hr.

after accident)

Hallway 2322 (Outside Sample Room)

I hr./ day High Activity Radioactive Lab 4 hr./ day Counting Room 24 hr./ day Spectro Photometer I hr./ day Cable Spreading Room h hr.*

Filter Rooms 2 hr./ day

• Switchgear Rooms (Elev. 121')

h hr.*

Hot Shutdown Panel 24 hr./ day

• CCW Pump Room h hr.*

Corridor 2161 hr.*

RHR Heat Exchanger Room hr.*

Primary Access Point 24 hr.*

Service Building (Operational Support 24 hr.*

Center)

Stairway No.1 Transit to elevation at west side of aux. b1dg.

Stairway No. 2 Transit to elevations 77'/83' Stairway No. 8 Transit to elevations at north and east sides of aux. bldg.

• Access to these areas may be required post-accident, but no other specific activity is anticipated.

Each of these areas have been analyzed to determine the dose rates following an accident.

7d.

Discussion of Comouter Programs Used Source term concentrations in uCi/cc for each isotope along with the volumetric source strengths in Hev/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 of 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.

Dose rates were calculated with the CYLSO conputer program. This program uses the Rockwell Point Kernel theory for one dimensional cylindrical volumetric source ;.

Self attenuation in the source as well as the shielding effects of various construction materials such as steel, lead, concrete and water are considered in the code. The code output is in terms of dose rate vs. distance for various piping diameters and shielding configurations.

This comouter program is similar to the SDC code.

7.e A time / dose rate study will be performed by January 1,1981 to determine the projected doses to individuals for necessary occupancy time in vital areas.

7.f The following recommendations will be implemented.

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Recontendations to be Implemented 1.

Add shielding at hydrogen analyzers, around reactor coolant and containment air sample lines and around lines to hydrogen analyzers.

This shielding will reduce dose rates in the Electrical Penetration Rooms and the areas near these rooms.

2.

Add 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.

3.

Place temporary shielding at the containment radiation monitor to reduce dose rate in the corridor (RE-Oll,.012).

4 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.

In addition, a design modification was initiated to add additional shielding outside the auxiliary personnel hatch to minimize potential effects at the Elevation 155' TSC area.

Ecuicment Oualification The effects of radiation on equipment is being considered as part of the NUREG 0588 response. The results of this study will be submitted along with NUREG 0588 September 15, 1980.

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