ML19318A801

Forwards Response to TMI-2 Action Plan.Info Addresses Fuel Loading,Low Power Testing License Requirements & fuel-power License Requirements

ML19318A801
Person / Time
Site:	Farley
Issue date:	06/20/1980
From:	Clayton F ALABAMA POWER CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8006240176
Download: ML19318A801 (117)

Text

{{#Wiki_filter:Alabama Power Company 600 North 18th Street Post Office Box 2641 8 fmingham, Alabama 35291 Telephone 205 323-5341 k h y,cy"g;fj, Alaba.aa Power g o \\d June 20,1980 f*'e sa#hern ekttcytem Docket No. 50-364 Director Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 RE: TMI-2 Action Plan Items Joseph M. Farley Nuclear Plant - Unit 2 Centlemen: Alabama Power Company submits the enclosed information addressing the IMI-2 Action Flan Items for the Joseph M. Farley Nuclear Plant - Unit 2 as resquested by the NRJ staff. The enclosed inforu.ation addresses " Fuel-Ioading and Low-Power Testing ps License Requirements" and " Full-Power Licen e Requirements". d If you have questions, please advise. Very truly

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g F. L. Clayton, Jr. FLCJr/TNB:aw cc: Mr. R. A. Thomas Mr. C. F. Trowbridge THIS DOCUMENT CONTAINS POOR QUALITY PAGES g a l)8006c40 a

O 4 l JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2 00CKET N0. 50-364 I 4 v RESPONSE TO TMI-2 ACTION PLAN lO l

'y TABLE OF CONTENTS %J Page No. FUEL-LOADING AND LOW-POWER TESTING LICENSEE REQUIROtENTS:

1. A. l.1

' Shif t Technical Advisor................. 1

1. A.l.2 Shift Supervisor Administrative Duties.........

3

1. A.l.3 S h i f t Ma n n i ng......................

4 I. A. 3.1 Revise Scope and Criteria for Licensing Examinations.. 7 I.B.l.2 Evaluation of Organization and Management Improvements of Near-Term Operating Licensee Applicants...... 8 I.C.1 Short-Term Accident Analysis and Procedure Revision... 15 I.C.2-Shift Relief and Turnover Procedurcs.......... 17 4' I.C.3 Shift Supervisor Responsibilities............ 18 I.C.4 Con trol Room Acces s. -.................. 19 k I.C.5 Procedures for Feedback of Operating Experience to Pl a n t S ta f f..................... 20 'I.C.7 NSSS Vendor Review of Procedures............ 22 I.D.1 Control Room Ocsign................... 23 I.G.1 Training During Low-Power Testing............ 24 11.8.4 Training for Mitigating Core Damage........... 25 I I.D.'l Relief and Safety Valve Test Requirements........ 26 II.D.3 Relief-and Safety Valve Position Indication....... 27 i II.E.1.2 Auxiliary Feedwater Initiation and Indication...... 28 II.E.4.1 Containment Dedicated Penetrations........... 32 II.F.1 Additional Accident Monitoring Instrumentation..... 33 II.F.2 Inadequate Core Cooling Instrumentation......... 37 i II.G ' Emergency Power for Pressurizer Equipment.... 43 f'1\\- \\_- i - y

TABLE OF CONTENTS (Cont'd) {m} Page No. II. K.1 IE Bulletins on Measures to Mitigate Small-Break -LOCAs and Loss of Feedwater Accidents........ 44 II. K.3 Final Recommendations of B&O Task Force........ 47 -III.A.1.1 Upgrade Emergency Preparedness............. 49 III.A.l.2 Upgrade Emergency Support Facilities.......... 50 III.D.3.3 In-Plant Radirtion Monitoring............. 6: FULL-POWER LICENSE REQUIREMENTS: I.C.7 NSSS Vendor Review of Procedures 62 I.C.8 Pilot Monitoring of Selected Emergency Procedures for Near-Tenn Operating License Applicants....... 63 I.G.1 Training During Low-Power Testing........... 64 4 II.B.1 Reactor Coolant System Vents.............. 65 II.B.2 Pl a n t Sh i el d i ng.................... 68 II.B.3 Pos t-Accident Sampl ing................. 75 II.B.4 Training for Mitigating Core Damage.......... 81 s II. E.1.1 Auxiliary Feedwater System Reliability Evaluation... 82 II.E.3.1 Emergency Power for Pressurizer Heaters........ 83 II.E.4.2 Containment Isolation Dependability.......... 84 II. K.3 Final Recommendations of B&O Task Force........ 86 III.AA.1 Upgrade Emergency Pre'paredness............. 87 4 l III.D.l.1 Primary Coolant Sources Outside Containment... 88 III.D.3.4 Co ntrol Room Ha bi tab il i ty............... 90 b) v

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Faricy Nuclear Plant Unit 2 j Docket No. 50-364 \\ Vl I. A.l.1 SHIFT TECHNICAL ADVISOR Position A technical advisor to the shift supervisor shall be present on all shifts and available to the Control Room within 10 minutes. Although minimum training require-ments have not been specified, shif t technical advisors should enhance the accident assessment function at the plant. This requirement shall be met before fuel loading. (See NUREG-0578, Section 2.2.1.b, and-letters of September 27 and November 9,1979.)

Response

An individual on shif t has been designated to serve in a dedicated capacity as shif t 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 shif t complement, in accordance with technical specifications. Designated individuals have other duties during normal plant conditions. Alabama Power has placed shift technical advisors on shift. By January 1,1981 the shif t technical advisor (STA) will have received additional training and periodic retraining, commensurate with NUREG-0578 requirements, in the basic physical concepts of mathematics, chemistry, metallurgy, atomic physics, reactor p) physics, heat transfer, fluid mechanics, thermodynamics and thermohydraulics. In t addition, training will be provided in the response and analysis of the plant for various transients and accidents and in basic plant design and layout including the capabilities of instrumentation and controls in the control room. The operating experience assessment function will be performed by the plant's system performance group which is composed of supervisory, engineering, and technical personnel. This group is not functionally a part of the plant operaticas group. The system's perfomance group is a multi-disciplined group which has overview of all plant systems including mechanical, electrical, and instrumentation and co ntrol. This group is dedicated to the operating experience assessment function which includes, but is not limited to, the fcilowing: 1. Engineering evaluation of the operating history of the plant (equipment failures, design problems, operations errors, etc.) and Licensee Event Reports from other plants of similar design, with suitable dissemination of the resul ts of such evaluations to other members of the plant staff. Engineering evaluation of the adequacy of the policy for maintenance, testitig, 2. equipment procurement, etc. 3. Engineering evaluation of continuing adequacy of plant operations quality assurance. 4. Engineering evaluation of adequacy of plant emergency and operating procedures. k

S FarleyNu-1 car ' Plant 2 Unit 2 Decket No.-50-364 I. A. l.1 Shif t Technical Advisor (Continued) The' operating experience assessment function will be implemented upon receipt of an operating license. Farley Nuclear Plant Licensee Event Reports (LERs) and significant LERs of other phnts are presently being routed for STA review and signature in accordance with existing plant procedures. In addition, when personnel performing operational assessment conclude that information exists which may be relative to the function of the STA, such information will be issued to the STAS. 2 O 4 l A N 4 1 - L.) 9 4 s u ,,mv -awr,, --n-- e-w-,-- -g+.,, g--. ~y v -4w,

f Farley Nuclear Plant 3 Unit 2-Docket No. 50-364 . / I.A.l.2 ' SHIFT SUPERVISOR ADMINISTRATIVE DUTIES Position Review the administrative duties of the shift supervisor and delegate functiens that detract from or are subordinate to the management responsibility for assuring safe i operation of the plant to other personnel not on duty in the control room. .This; requirement shall be met before fuel loading. (See NUREG-0578, Section 2.2.1.a, Item (4), and letters of S. 4mber 27 and November 9,1979.)

Response

Corporate management has reviewed the administrative duties of the shif t supervisor. Administrative functions that detract from safe operation have been delegated to other personnel not on duty in the control room. i O E 1 f 1 a 0 v-

Farl<y Nuclear Plant Unit 2 4 Docket No. 50-364 j I.A.1.3 SHIFT MANNING Posi tion The minimum shif t crew for a unit shall include three operators, plus an additional three operators when the unit is operating. Shift staffing may be adjusted at multi-unit stations to allow credit for operators holding licenses on more than one unit. In each control room, including common control rooms for multiple units, there shall be at all times a licensed reactor operator for each reactor loaded with fuel and a senior reactor operator licensed for each reactor that is operating. There shall also be onsite at all times, an additional relief operator licensed for each reactor, a licensed senior reactor operator who is designated as shift supervisor, and any other licensed senior reactor operators required so that their total number is at least one more than the number of control rooms from which a reactor is being operated. Administrative procedures shall be established to limit maximum work hours of all personnel perfonning a safety-related function to no more than 12 hours of continuous duty with at least 12 hours between work periods, no more than 72 hours in arty 7 day period, and no more then 14 consecutive days of work without at least 2 consecutive days off. These requirements shall be met before fuel loading. (Detailed guidance to licensees is expected to be issued in May 1980.) d( b

Response

The control room will be staffed as required by Technical Specification Table 6.2-1 (a ttached). These staffing levels meet the intent of this staffing requirement. Alabama Power Company is in total agreement with and fully supports the Commission's requirement to limit maximum work hours of personnel performing safety-related functions. At this time the Company is able to commit to working personnel under normal conditions tn no more than 72 hours in any seven-day period and to working these personnel no more than 14 consecutive days without at least two consecutive days off. At the present time the Company is not able to fully commit to limit the work of personnel to more than 12 hours under continuous duty with at least 12 hours off be-tween work periods. Our inability to make this commitment is due primarily to a binding contract with the International Brotherhocd of Electrical Workers (IBEW) which does not provide for such scheduling of personnel. The company is initiating negotiations with the IBEW to meet the 12 hour provisions of your requirement. In addition, newly constituted training and licensing requirements raise uncertainties about total manpower needs which make your 12 hour provisions difficult to meet. n k

~ 5 TABLE'6.2-1 p.1

1. FOR SHARED CONTROL ROOM

((j - TOTAL MINIMUM CREW COMPOSITION Condition of Unit 2 - No Fuel in Unit 1 ~. APPLICABLE liODES LICENSE CATEGORY 1, 2, 3 & 4 5&6 SOL 2 1* OL 2 1 Non-Licensed 2 1 Condition of Unit 2 - Unit 1-in MODES 1, 2, 3 or h APPLICABLE MODES LICENSE CATEGORY 1, 2, 3 & h 5&6 v SQL** 2*** 2* OL** 3*** 2*** 4 Non-Licensed 5 h . Condition of Unit 2 - Unit 1 in MODES 5 or 6 . APPLICABLE MODES LICENSE CATEGORY 1, 2, 3 & 4 5&6 SQL** 2 1***** OL** 2*** 2 'Non-Licensed h 3 /"'h .(j;

6 TABLE 6.2-1 (Continued)

• Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS after the initial fuel loading.
  • Assumes each individual is licensed on the unit to which he is assigned.

1. Shift crew composition for shared control room (including an individual

~ qualified in radiation protection procedures) may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accomodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. """ One individual licensed on both units. The licensed individuals, one licensed on Unit 1 and one licensed on Unit 2, may be substituted in lieu of that individual. l< j ~

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c- ~ - - - -. - ~ M 2 Farley Nuclear Plant ' 7 . Unit 2. 4 Docket Noi 364 f t d. I.A.3.1

REVISE SCOPE AND CRITERIA FOR LICENSING EXAMINATIONS

. Pos ition. = All reactor ornatorilicense applicants shall take a written examination with a new category dealing with the principles of heat transfer and fluid mechonics, a time limit of. nine ho'urs, a_nd ~a passing. grade of.80 percent overall and 70 percent 1 .in each category. All -senior reactor operator' license applicants shall take the reactor operator examination, an operating test, and a senior reactor operator written examination with a' new category dealing with the theory of fluids and thermodynamics, a time- . limit of seven hours, and a passing grade of 80 percent overall and 70 percent in 4 each category. These requirements shall.be met before fuel loading. (See letter of March 28,1980.) l

Response

i The training scope and criteria for_ reactor operator and senior reactor operator licenses examinations are ~ being changed to comply with requirements stipulated in the March 28 1980 letter on this subject. All requirements will be implemented prior to fuel load. 1

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Farlcy Nucicar.P1 ant Unit 2 8 Docket No. 50-364 h I.B.l.2 EVALUATION OF ORGANIZATION AND MANAGEMENT IMPROVEMENTS OF NEAR-TERM OPERATING LICENSEE APPLICANTS Position The licensee organization shall comply with the findings and requirements generated in an interoffice NRC review of. licensee organization and management. The review will be based on an NRC document entitled " Draft Criteria for Utility Management and Technical Competence." The first draft of this document was dated February 25, 1980, but the document is changing with use and experience in ongoing reviews. These draft criteria address the organization, resources, training, and qualification of the plant staff, and management (both onsite and offsite) for routine operations and the resources and activities (both onsite and offsite) for accident conditions. Establish a group that is independent of the plant staff but is assigned on site to perform independent reviews of plant operationalactivities and a capability for evaluation of operating experiences at nuclear power plants. Organizational changes are to be implementeo' on a schedule to be determined prior to fuel loading.

Response

Alabama Power Company meets or exceeds the general technical and organizational con-cepts outlined in the " Draft Criteria for Utility Management and Technical Competence." hd A description of the company's organization to operate and support the Farley Plant is contained in the following sections and in other sections of the Action Plan. A. Criteria for Routine Operations Figure 1 ~ describes the Farley Nuclear Plant (FNP) Organization established for routine day-ta-day operational activities. Such organization reponsibilities, authorities and expertise levels described in ANSI N 18.7-1972 are fully met by this plant organi-zation. The qualification, selection and training requirements of Regulatory Guide 1.8 and ANSI N 18.1-1971 are met within the plant organization staff. - An organizational chart of the offsite support staff is presented in Figure 2. This staff is sufficiently involved in the day-to-day operational activities, including plant modifications to assure a continual understanding of plant conditions and safety considerations. This figure indicates clear lines of authority and communication between organizational units involved in management, technical support and operation of the plant.. Personnel depicted on this chart are of sufficient numbers and have appropriate educational and experience associated with power plant operating and design necessary to ensure the proper handling of each work assignment by competent personnel This support staff, as indicated in the group titles -in Figure 2, provides support in the areas of operating abnormalities, maintenance problems, design changes and modi-fications, licensing support, security and emergency planning. This organization is supplemented-by the staff of Southern Company Services (Figure 3) which serves in an architect engineering and licensing advisory role within the Southern Company. (7 This figure indicates separate staffs dedicated to design and licensing support. In V add] tion the staffs of Bechtel Power Corporat' ion, and Westinghouse prov'ide technical supportin the areas of general architect enginee-ing and Nuclear Steam Supply System Support. .In order to provide thorough onsite safety review, Alabama Power Company has established

r Farl ^y Nuclcar Plant Unit 2 9 Docket No. 50-364 I.B.1.2 Evaluation of Organization and Management Improvements of Near-Term Operating Licensee Applicants (Continued) a comprehensive system of reviews and evaluations as follows: a. The responsibilities of the System Performance and Planning Group, which rerorts to plant management, include: 1. Evaluation of the technical accuracy and clarity of procedures important to the safe operations of the FNP 2. Evaluation of operating experience of the FNP and plants of similar design 3. Review of changes to the FNP as described in the FSAR, procedures as described in the FSAR, and tests and experiments not described in the FSAR which are completed without prior NRC approval under provisions of 10CFR50.59(a)(1). b. The Plant Operation Review Committee (PORC) composed of key plant management personnel, advises the Plant Manager on matters affecting nuclear safety which include: 3 1. Evaluations for technical accuracy and clarity of procedures impor- !j tant to the safe operations of the facility. 2. Review of changes to the FNP as described in the FSAR, procedures as described in the FSAR,and tests and experiments not described in the FSAR which are completed without prior NRC approval under provisions of 10CFR50.59(a)(1). i 3. Review of changes in the FNP and proposed tests and experiments of which involve a change in the Technical Specifications or an unreviewed safety questior as defined in 10CFR50.59(c). 4. Review of changes in the Technical Specifications and license amendments relating to nuclear safety prior to implementation, except in those cases where the change is identical to a previously reviewed pmposed change. 5. Review of violations, deviations, and reportable events, which require reporting to the NRC in writing such as:

1) Violations of applicable codes, regulations, orders, technical specifications, license requirements or I

internal procedures or instructions having safety significance. ii) Signifcant operating abnormalities or deviations from nonnal or expected performance of plant structures, systems, iii) Reportable events which require notification to the NRC in writing within 24 hours, as defined in the plant technical specifications. Operations Evaluation Teams, composed of highly trained and specialized c. m

Farl:y Nuclear Plant U nit 2 10 Docket No. 50-364 . f3 V I.B.l.2 Evaluation of Organization and Management Improvements of Near-Term Operating Licensee Applicants (Continued) personnel, are constituted by and report to the Vice President-Nuclear Generation for the purpose of evaluating matters affecting the operation of the FNP which include: 1. Evaluation of plant operations from a nuclear safety perspective. 2. Overall assessme.e of the FNP staff perfonnance regarding conformance to requ*rements related to nuclear safety. d. The Safety Audit and Engi>.eering Review Group (SAERG), which is independent of the FNP staff, is constituted to peJform review functions and verification that other functions are prformed correctly by other applicable groups. These review and verification responsibilities include: 1. Evaluation of plant operation from a nuclear safety perspective. 2. Evaluation of the QA program is performed by Meager-0QA, to whom the SAERG reports, by periodic audits performed by his offsite organi-zation, Joint Utility Management Audits (JUMA), and providing an annual formal summary of the 0QA program which is reviewed by the Senior Vice President. 3. Assessment of the FNP staff performance regarding conformance to rec;airements relating to nuclear safety. 4. Evaluation of personnel changes in key management positions and evaluation of changes in plant organizational structure via annual audit of FNP training and qualification. 5. Verification of adequate evaluation of the technical accuracy and clarity of proceduresimportant to safety operations of the FNP. l 6. Verification of adequate evaluationof tte experience of the FNP l and plants of similar design. f

7. ' Verify by audit the handling of changes to the FNP as described in the l

FSAR, changes in procedures as described in the FSAR, and tests and experiments not described in the FSAR which are completed without prior NRC approval under provisions of 10CFR50.59(a)(1) and changes in the facility and proposed tests and experiments, any of which involve a change in the Technical Specifications or an unreviewed safety question as defined in 10CFR50.59(c). 8. Verify by audit the handling of changes in Technical Specification and license amendments relating to nuclear safety prior to implementa tion. q U 9. Verify by audit the handling o'T violations, deviations, and reportable events which require reporting to the NRC in writing. These organizations are shown in Figures 1 and 2.

Faricy Nuclear Plant Unit 2 Docket No. 50-364 Il '(] I.8.1.2 Evaluation of Organization and Management Improvements of Near-Term Operating Licensee Applicants (Continued) Persons perfonning these independent reviews and evaluations meet the qualification requirements described in Section 4.7 of ANSI 18.1 of 1971 and Reg. Guide 1.8 as verified by an annual audit. The SAERG prepares written summaries of reviews and evaluations performed as noted above. These summaries include the results of, and recommendatons resulting from such reviews and evaluations. Monthly reports containing a summary of work completed and reconnendations made are prepared and distributed to corporate manage-ment with infonnation copies to plant management. The Ntclear Operations Review Board (NORB) is the senior level oversight group that provides a means for corporate management to be involved in nuclear power plant safety considerations, and to assure that safety considerations are effectively applied to plant operational activities, n The NORB is comprised of knowledgeable senior management personnel. The group meets Q quarterly to perform, as a minimum, the following functions: a. Provide oversight to activities performed by the PORC. b. Review corrective actions and recommendations of the PORC. c. Assure, as appropriate, that corrective actions and recommendatio1s of the PORC are effectively implemented. B. Criteria for Accident Conditions Alabama Power Company currently meets many of the concepts provided in the draft as described in Farley Nuclear Plant Emergency Plan. To more fully document compliance with these criteria, the Emergency Plan is planned for revision by July 15,1980. This revision will address Alabama Power Company's dedicated offsite and onsite resources for accident mitigation and recovery activities. O

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Farley Nuclear Plant 1,5 Unit 2 ' Docket No. 50-364 ch I.C.l' SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION Positio n Analyze small-break LOCAs over a range of break sizes, locations and conditions (including some specified multiple equipment failures) and inadequate core cooling due to both low reactor coolant system inventory and the loss of natural circulation to determine the important phenomena involved and expected instrument indications. Based on these analyses, revise as necessary emergency procedures and training. These requirements shall be met before fuel loading. (See NUREG-0578, Sections 2.1.3b and 2.1.9, and letters of September 27 and November 9,1979.)

Response

Analyses of small break loss of coolant accidents, symptoms of inadequate core cooling and required actions to restore core cooling, and analysis of transient and accident 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. The small break analyses have been completed and were reported in WCAP-9600, which was submitted to the Bulletins and Orders (B & 0) Task Force by the Owners' Group on June 29, 1979. Incorporated 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 & 0 Task Force and have been presented to the Owners' Group utility representatives in a seminar held on October 16-19, 1979. O Following this seminar, Alabama Power has revised emergency procedures and has trained personnel or, these procedures. Revised procedures and training are in place in accordance with the requirement in Enclosure 6 to Mr. Vassallo's letter of September 27, 1979 and included in the Unit 2 operator training program. The work required to address'the other two areas--inadequate core cooling and other transient and accident scenarios--has been performed in conjunction with schedules and requirements established by the Bulletins and Orders Task Force. Analyses related to the definition of inadequate core cooling and guidelines 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 second quarter of 1980. The guidelines and training are included in the Unit 2 operator training program required by the B & 0 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 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 informa-tion-available to the operator at each decision point. Following this, criteria have _been set for credible misoperation, and time available for operation decisions have.been qualitatively assessed. The information developed has been used to test CX Abnormal and Emergency Operating Procedures against the event sequences and determine NJ C

. Faricy Nuclear Plant " Uni t 2 16 Docket No. 50-364- .I.C.1 . Short-Term Accident' Analysis and Procedure Revision (Continued) ~if inadequacies. exist. There were no inadequacies determined in the study. The results of this study were provided to the Bulletins and Orders Task Force on March 31, 1980.as required. The Owners'. Group.has also provided test predictions analysis of the LOFT L3-1 4 nuclear small break experiment. This analysis was provided on December 15, 1979 in accordance with the schedule established mutually with the Bulletins and 1 Orders Task Force. i l 1 l I r. i'O e i e LO ,.,_....-,--,,,-.._.,.,.....-c.

!Faricy Nuclear Plant 17 -Unit 2 Docket No. 50-364 .(~). k- > I.C.2 - SHIFT RELIEF AND TURNOVER PROCEDURES Position Revise plant procedures for shift relief and turnover to requirc signed check-lists and logs to assure that the operating staff (including auxiliary operators and maintenance personnel) possess adequate knowledge of critical plant parameter status; system status, availability and alignment. This requirement shall be met before fuel loading. (See NUREG-0578, Section 2.2.lc, and letters of September 27 and November 9,1979.)

Response

Plant procedur's for shift and relief turnover have been reviewed and revised to assure that. 1. A checklist has been provided for the shift relief of control room operators and the shift supervisor. The check list includes the following items: a. Assurance that critical plant parameters are within allowable

limits.

x b. Assurance of the availability and proper alignment of all Os systems essential to the prevention and mitigation of operational transients and accidents. c. Review of systems and components that are in a degraded mode of operation permitted by the technical specifications. 2. Checklists or logs have been provided for shift relief of assistant plant operators and equipment operators. Equipment that could de-grade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient has been listed on the appropriate operators' checklist. 3. A system has been established to periodically evaluate the effectiveness of the turnover procedures. l l (~%).

Faricy iiuclear Plant Unit 2 18 Docket No. 50-364 1 V I.C.3 SHIFT SUPERVISOR RESPONSIBILITIES Posi tion Issue a corporate management directive that clearly establishes the command duties of the shif t supervisor and emphasizes the primary management responsibility for safe operation of the plant. Revise plant procedures to clearly define the duties, responsibilities and a'Jthority of the shift supervisor and the control room operators.

Response

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 pl.snt under all conditions on his shift and that clearly establishes his command duties. Plant procedures have been revised to assure that the duties, responsibilities, 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 shift supervisor in the control room relative to other plant management personnel. Particular emphasis has been placed on the following: 1. The responsibility and authority of the Shift Supervisor shall be to maintain p) a broad perspective of operational conditions affecting the safety of the ( plant as a matter of highest priority at all times when on duty in the control room. The idea shall be reinforced that the Shif t Supervisor snould not become totally involved in any single operation in times of emergency when multiple operations are required in the control room. 2. The Shif t Supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to relieve the Shift Supervisor shall be speci fied. 3. If the Shift Supervisor is temporarily absent from the control room during routine operations, a licensed person shall be designated to assume the control room command function. These temporary duties, responsibilities, and authority shall be clearly specified. Training programs for shift supervisors emphasize and reinforce the responsibility for safe operation and the management function the shift supervisor is to provide for assuring safety. 7

i Farley: Nuclear 'Planti - Unit 2 19

Docket?No. 50-364 I.C.4 ' CONTROL ROOM ACCESS Position Revise plant procedures.to-limit access to the control room to those individuals responsible for the direct operation of the plant, technical advisors, speci-

. fied NRC. personnel, and. to. establish a clear line of authority,' responsibility, and succession in the~ control room. This-requirement shall be met before fuel loading. (See NUREG-0578, Section 2.2.2a, and letters of. September 27, and November 9,1979)

Response

- Plant procedures have been reviewed and revised to assure the following: 1. The authority and responsibility of the persons in charge of . limiting access to the"at the controls area" have been established. 2. A clear line of authority and responsibility in the control room in normal and emergency conditions has been established. The line of succession for persons in charge of the control room has been established. Lines of communication and authority for plant management personnel not in direct command of operation are defined. i 4 4 L 4 LO: -s. m. m-n v e - w , -,- p

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Farley Nuclear Plant Unit 2 20 . Docket No. 50-364 y) I.C.5 PROCEDURES FOR FEEDBACK 0F OPERATING EXPERIENCE TO PLANT STAFF Position In accordance with Task Action Plan I.C.5, Procedures for Feedback of Operating Experience to Plant Staff, each licensee shall review its procedures and revise them as necessary to assure that operating information pertinent to plant safety originat-ing both within and outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs. These procedures shall: Requirement: (1) Clearly identify organizational responsibilities for review of operating experience, the feedback of pertinent information to operators and other personnel and the incorporation of such information into training and retrain-ing programs. (2). Identify the administrative and technical review steps necessary in translat-ing recomendations by the operating experience assessment group into plant actions (e.g., changes to procedures, operating orders). (3) Identify the recipients of various categories of information from operating experience (e.g., supervisory personnel, STA's, operators, maintenance personnel, f) H.P. technicians) or otherwise provide means through which such information can v be readily related to the job functions of the recipients. (4) Provide means to assure that affected personnel become aware of and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs. (5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency. (6) Provide suitable checks to assure that conflicting or contradictory information is not conveyed to operators and other personnel until resol Jtion is reached. -(7) Provide periodic internal audit to assure that the feedback program functions effectively at all levels.

Response

Requirement: (1) - The operating experience assessment function will be performed by the plant's system performance group which is composed of supervisory, engineering and technical personnel. This group is not functionally a part of the plant p operations group. The system performance group is a multi-disciplined group

Faricy Nuclear Plant Unit 2 21 Docket No. 50-364 Pd I.C.S. Procedures for Feedback of Operating Experience to Plant Staff (Continued) which has overview of all plant systems including mechanical, electrical, and instrumentation and control. This group is dedicated to the operating experience assessment function which includes, but is not limited to, the following: A. Engineering evaluation of the operating history of the plant (equipment failures, design problems, operations errors, etc.) and Licensee Event Reports from other plants of similar design, with suitable dissemination of the results of such evaluations to other members of the plant staff; B. Engineering evaluation of the adequacy of the policy for maintenance, testing, equipment procurement, etc. C. Engineering evaluation of continuing adequacy of plart operations quality as:,urance; and D. Engineering evaluation of adequacy of plant emergency and operating procedures. (2) Steps for implementing recomended procedural changes and changes in plant design are outlined in plant procedures. Reports of significant events from other bq plants, vendor notifications, and regulatory notifications are reviewed and evaluated in accordance with appropriate plant procedures. (3) Conduct of Operations procedures identify plant superintendents / supervisors as the recipients of pertinent information. The functional group super-intendents/ supervisors will distribute pertinent infonnation as required. (4) The Conduct of Operations procedures for appropriate functional groups will be revised as necessary to assure timely distribution of information. (5) Plant procedures provide for screening of extraneous and unimportant informa-tion. (6) Checks are performed by both the systems / performance group and by the individual group superintendent / supervisor to ensure that conflicting or contradictory information is not conveyed to plant personnel. (7) Procedures will be revised to provide for periodic audits by Safety Audit and Engineering Review Group to ensure that the feedback program is functioning effectively. h v

e-4 ~ Faricy Nuclear. Pl ant x ' Unit:2L 22 Docket No. 50-364. D :v 'I.C;7-NSSS VENDOR RF. VIEW OF PROCEDURES . Position: i Obtain nuclear steam supply. system'(NSSS) vendor review of low-power testing ~ procedures to further' verify their adequacy. 'This f requirement nust be met before fuel loading. ? l Response < i. Low-power physics testing procedures will be reviewed by the NSSS vendor i (Westinghouse) before fuel loading. i f b I i-i i

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Faricy Nucicar Plant Unit 2 23 Docket No. 50-364 rmy I.D.1 CONTROL ROOM DESIGi Position Perform a preliminary assessment of the control room to identify significant human factors deficiencies and instrumentation problems and establish a schedule approved by the NRC for correcting deficiencies. This requirement must be met before fuel loading.

Response

The preliminary human factors and operations review of the Unit 2 control room has been conducted by Alabama Power Company. The report was transmitted to the NRC via letter from Mr. F.L. Clayton, Jr. to Mr. A. Schwencer dated June 10,1980. A schedule for implementing enhancement items identified during the control room review is provided in the report. O A

Farity Nuclear Plant Unit 2 24 Docket No. 50-364 - ~s I.G.1 TRAINING DURING LOW-POWER TESTING Posi tion Define and commit to a special low-power testing program approved by NRC to be conducted at power levels no greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training. This requirement shall be met before fuel loading. Position Alabama Power Company commits to perform the special low-power testing program which constitutes supplemental operator training, as delineated in the letter from Mr. F. L. Clayton, Jr. to Mr. A. Schwencer dated June 17, 1980. t t d

- Farley Nuclear Plant Unit'2 25 Docket No. 50-364 p; Q II.B.4 TRAINING FOR MITIGATING CORE DAMAGE Position Develop a training program to instruct all operating personnel in the use of installed systems, including systems that are not engineered safety features, and instrumentation to monitor and control accidents in which the core may be severely damaged. This requirement shall be met before fuel loading.

Response

A training program to instruct all licensed operating personnel in methods for mitigating core damage as stated will be implemented before fuel loading. i l D i 1

Farl'ey Nuc1 car Plant - 26 Unit 2 Docket No. 50-364 U,7 I I. D.'l

RELIEF AND SAFETY VALVE TEST REQUIREMENTS Position Describe a test program and schedule for testing to qualify reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.

This requirements shall be met before fuel loading. (See NUREG-0578, Section 2.1.2, and letters of September 27, and November 9,1979).

Response

By letter dated December 17, 1979, Mr. William J. Cahill, Jr., Chairman of the EPRI Safety and Analysis' Task Force submitted the " 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 NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations" dated July 1979, Item 2.1.2 which recommended, in part, to " commit to provide performance verifications by full scale proto-typical testing for all relief and safety valves. Test conditions shall include two phase slug flow and subcooled liquid flow calculated to occur for design basic transients and accients." The EPRI Program Plan provides for a completion of the essential portions of the test program by July 1981. Alabama Power Company is participating in the EPRI program to provide program review and to supply plant specific data as required. r

s Farley-Nuclear Plant l Lunit 2-27-Docket No. 50-364, p ~ L'(, ) ; II.D.3 RELIEF:AND SAFETY VALVE POSITION INDICATION 1 . Posi tion Instal 1~ positive in'dication in the control room of relief and safety valve posi-tion derived from a reliable valve position detection device or a reliable indication of flow in the valve discharge pipe. 4 LThis requirement shall 'be met before' fuel loading. (See NUREG-0578, Section _ 2.1.3a, and letters of September 27, and November 9,1979).

Response

The ' pressurizer power-operated ' relief valves have sten-mounted limit switches which control red and green. indicating lights in the valve control switch on the main' control board (MCB). This ' indicating system provides positive open i ' 1and closed indication for these valves. The presently installed limit switches-will be replaced with qualified limit switches and an alarm function associated with the valve position will be added prior to receipt of an operating license. j This alarm function will be provided on the-main control board. I' _ The pressurizer safety _ valves _will be modified to install qualified stem-mounted j . limit switches which will' provide. positive open and shut indication and the associated alarm function.on the main control board. This modification will be cor _leted prior to fuel load.- As backup indication, there are temperature detectors in all relief and safety lines to the pressurizer relief. tank. - In addition, temperature, pressure, and level indication for the pressurizer relief tank are provided on the MCB and alarmed in the plant cunputer. These indications are available to the operator to confinn the position'of the~ pressurizer power-operated relief valves and safety valves. The pressurizer power-operated relief valves and' safety valve indicating lights -provide the operator.with unambiguous indication of valve position (open or closed) so that appropriate operator actions can be taken. These indications are powered from the plant class 11E de distribution system. They are seismically qualified consistent with the power operated relief valve and safety valve requirements, and evnironmentally qualified for. any containment environment. ) I i a ) N ,a 4,-- ~

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Faricy Nuclear Plant Unit 2 28 Docket No. 50-364 _f II.E.1.2 AUXILIARY FEEDWATER INITIATION AND INDICATION position 1 The design shall provide for the automatic initiation of the auxiliary feedwater system.

Response

With the control switches in the auto position, the motor-driven auxiliary feedwater pumps will automatically start on any of the following signals: A. Low-low water level signals from two-out-of-three level transmitters on any one steam generator. 8. Any of the conditions as defined in FSAR Section 7.3 that cause a safety injection signal. C. Loss of offsite power. Operation of the turbine-driven auxiliary feedwater pump is initiated by the opening of the steam supply valves to the turbine drive. Steam from the main steam header is automatically admitted to the turbine drive on either of the following signals: 7-~3 A. Loss of power signal (2/3 reactor coolant pump bus undervoltage), or \\ l B. Low-low water level signals from two-out-of-three of the level transmitters of any two-out-of-three steam generators. Details of the emergency operation of the auxiliary feedwater systems are contained in FSAR Sections 6.5.2.2.5 and 6.5.2.3.3. Instrumentation and controls for the system are shown on drawings D-205007, 0-205033, D-207186, D-207188, D-207189, D-207590, D-207591, and D-207857. Position 2 The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.

Response

The auxiliary feedwater system is designed to meet the single failure criteria so that no single failure will prevent the supply of sufficient feedwater to at least two of the three steam generators. A detailed design evaluation is contained in FSAR Section 6.5.3. A failure analysis of the auxiliary feedwater system is pro-vided in Table 6.5-2 of the FSAR. r I I w"

Faricy Nuclear Plant 29 Uni t 2 - Docket No. 50-364 .(, V II.E.1.2 Auxiliary Feedwater Initiation and Indication (Continued) ?osition 3 Testability of the initiating signcis and circuits shall be a feature of the design.

Response

In order to ensure the operability of the auxiliary feedwater system, periodic testing of the system will be performed in accordance with the Farley Nuclear Plant Standard Technical Specifications, Section 3/4.7.1.2. These technical specifications list the limiting conditions for operation and the surveillance requirements for the aux 111ary feedwater system. These surveillance requirements ensure that both the motor-driven pumps and turbine-driven pump are operable and develop at least 93 percent of design discharge pressure; that each valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and that the motor-driven and turbine-driven pumps start automatically upon receipt of a test signal which simulates emergency operation of the systen. Position 4 The initiating signals and circuits shall be powered from the emergency buses. p

Response

v The initiating signals and circuits for the auxiliary feedwater system are powered from the emergency buses as shown in drawings D-207001, D-207005, and D-207006. Position 5 Manual capability to initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manuai circuits will not result in the loss of system functions.

Response

The auxiliary feedwater system can be operated locally from the hot shutdown panel or' remotely from the control room. The operation of the system is described in FSAR Section 6.5.2.3. The system is designed to meet the single failure criteria and a failure analysis of the system is contained in FSAR Table 6.5-2. Position 6 The ac motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads on the emergency buses. ( Respons e The ac motor-driven pumps and valves in the auxiliary feedwater system are included in the automatic _ actuation of the loads onto the emergency buses as indicated on the emergency load sequence drawings D-207001, D-207005, D-207006, D-207645, D-207646, D-207649, and D-207650.

.-. - -.. =- Farley Nuclear Plant 30-Unit'2<.. - Docket No.:50-364-p II.E.1.2 Auxiliary Feedwater Initiation and Indication (Continued) Position 7 The automatic initiating signals and circuits shall be designed so that.their ' failure will not result in the -loss.of manual capability to initiate the AFWS fron' the control room.

Response

The auxiliary feedwater system can be operated locally from the hot shutdown panel-or remotely from the control room as detailed in FSAR Section 6.5.2.2.5. Instru-mentation and controls for the systen are shown on P & ID drawing D-205007 and the . initiating logic on Figure 7.2-14 of the.FSAR. i 7 Auailiary feedwater. injection lines to each steam generator are _ provided with j - flow indication. This flow indication is on the main control board and is powered from the plant energency power. These flow instrument loops are testable. Redundancy requirements are met by qualified steam generator level instrumentation (safety i grade). LA description of the presently installed equipment is contained below. 1 The auxiliary feedwater-line flow indicators will be seismica;1y and environ-mentally qualified by January 1,1981. This will meet all safety grade requirements. Position i 1 2 Auxiliary feedwater flow indication to each steam generator shall satisfy the single failure criterion.- i

Response

Local and control room indication of auxiliary feedwater flow to each of the steam generators. is provided by flow orifices Q2N23FE-3229A, B, and C in the auxiliary feedwater supply line,. located,just upstream of the auxiliary feedwater stop check valves.. Instrumentation and controls ~for the system are shown on P & ID drawing D-205007. The auxiliary feedwater flow indication is backed up by three redundant safety grade narrow range steam generator. level channels and one safety grade wide range steam generator level channel per steam generator vnich'have control room readouts. e Pos'i tio n' 2 ' Testability of the-auxiliary fe'edwater flow indication channels shall be a feature of-the design.. ~ Response' Testing of!the, auxiliary.feedwater. flow indication,will be performed in accordance ~~ lwith the. Farley Nuclear Plant preventative maintenance program on 18-month intervals by injection of a test signal.at the primary sensor.. The instruments are calibrated it?the. output. signals do not meet the' required accuracy for the instrument. The s-- T g i 9 y g .w- /-y-e y y-, - w, r 3 - -y g-g -. = 3 -, r.p rw g wei. w y- ,y-y,u--w--i -m 9 9 r-w-,- vgw,g

Farley-Nuclear Plant Unit 2 31 Docket No. 50-364 7 _

-( ) II.E.1.2 Auxiliary Feedwater Initiation and Indication (Continued) steam generator narrow range level channels are functionally tested every 31 days and calibrated every 18 months in accordance with Farley Naclear Plant Technical Specification requirements. The steam generator wide range level channels are calibrated every 18 months in accordance with Farley Nuclear Plant Technical Specifications. Position 3 Auxiliary feedwater flow instrument channels shall be powered from the vital instru-ment buses.

Response

The auxiliary feedwater flow instrumentation channels receive their power from the class IE vital instrument bus..The Bechtel Corporation drawing A-207076 gives a block diagram of the steam generator feedwater flow indication loop and the power supply.to the channels (B0P Instrument Panel lJ) is indicated on drawing D-207024. The steam generator narrow range channels receive their power from the class IE vital instrument buses. The power supply to channels (process I & C cabinet numbers 1, 2, and 3) is indicated on drawing D-207024. Each cabinet contains one channel for-each steam generator. The steam generator wide range channels also receive 9 ~3 their power from vital instrument buses as indicated on drawing D-207024 (process (,) I & C cabinet number 5, 7, and 8). Position 4 Each auxiliary feedwater channel should provide an indication of feed flow with an accuracy on the order of t 10 percent. +

Response

The present control grade system has transmitters with a 15 percent full scale accuracy. The Westinghouse power supply has a gain accuracy of i 0.1 percent of full scale and the control. board indicator has an accuracy of i 1.5 percent of full scale. The additive accuracy of the flow loop is i 10 percent range. The steam generator level channels have an indication accuracy of i 4 percent. The steam generator level and auxiliary feedwater flow transmitters are seismically 'and envirorynentally qualified. (D V 4 w

+ A Farley Nuclear Plant Unit 2 32 Docket No. 50-364 7V II. E.4.1 CONTAItNENT DEDICATED PENETRATIONS Posi tion - Install a containment isolation system for external recombiners or purge systems for post-accident combustible gas contml, 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

Farley Nuclear Plant has redundant electric hydro' gen.recombiners located inside contairinent for use in removing hydrogen gas from the containment atmosphere during post-accident conditions. These recombiners meet all engineered safety feature requirements and the controls and instrumentation for each are located on separate panels in the main control room. Hydrogen concentration can be adequately monitored in the main control room. The emergency procedure for loss of coolant accident contains detailed instructions for operating the recombiners. Since this system is located inside containnent and does not require mechanical hookup after an accident, personnel exposure during use is not a consideration at Farley. Alabama Power has reviewed and upgraded the emergency procedure for operation of electric hydrogen recombiners located inside containment. Training on this procedure revision has (m been incorporated into the Unit 2 operator training program. b In addition, the post-accident venting system is provided as a backup system for the redundant hydrogen recombiners. It consists of a supply line through which pressurized air may be admitted to the containment and an exhaust line through which hydrogen-bearing gases may be vented from the containment. The gases are filtered to limit radioactive discharges, /

U Farlcy Nuclear Plant ~ 33 Unit 2-.

Docket-No. 50-364 II.F.1 ADDITIONAL. ACCIDENT MONITORING INSTRUMENTATION Position

- Provide procedures for estimating noble gas, radiciodine, and particulate release rates if the existing effluent instrumentation goes off the scale. This requirement shall.be met-before fuel loading. Install continuous indication in the control room of the following parameters: 3 a. Containment pressure from minus 5 psig to three times the design pressure of concrete containments and four times the design pressure of sted contain-ments;- b. Contairinent water level in PWRs from (1) the bottom to tb top of the contain-ment sump, and (2) the bottom of the containment to a level equivalent to 600,000 gallons of water; Contairunent 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 /hr; e. Noble gas effluent from each potential release point from normal -'ncentrations to 105 pCi/cc(Xe-133). ~ Provide capability to continuously sample and perform onsite analysis of i.he 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

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. 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 moni tor. 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 1.2 MeV. Calibration will-be done at installation and annually using a Csl37 calibration source.- Predetermined calculational methods will be used to convert the radiation { level reading to' radioactive effluent release rate.

Firley Nuclear: Plant ~ Uni t12 - Docket No. 50-364 34 g Q II.F.1 Additional Accident Monitoring Instrumentation (Continued) The measurement of radioiodine and particulate effluents will be accomplished by a modification to the normal 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 ray 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. (Note: If the requirements of NUREG 0578 -- Section 2.1.8.B1 are met prior to re-ceipt of the Unit 2 Operating License, the.above interim procedu're will not be used. To measure the release rates from the main condenser air ejector, procedures will be developed prior to the receipt of an operating license to use a portable beta /gama survey instrument and a mR/hr vs. pCi/ml conversion table to promptly determine the release rates in the highly improbable event the installed monitor goes offscale. The procedures will provide a designated location on the discharge line of each main condenser air ejector where a technician will measure the dose ra te. 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 of ten as required by the emergency director. Procedures have been written to obtain the release rates from the atmospheric O steam relief valves. These procedures provide for taking contact dose rate V 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 of ten as required by the emergency director. The.present containment pressure indication provides continuous redundant indica-tion 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. The Farley Nuclear Plant present design has two wide range containment water level detectors. These' detectors provide indications in the main control room that meet the wide range requirements as specified in Mr. Vassallo's September 27, 1979 letter. These level transmitters and associated readout are safety grade and measure volumes in excess of 600,000 gallons. The Farley Nuclear Plant will provide a narrow range containment level indication . meeting the ranges specified in Mr. Vassallo's September 27, 1979 letter by January 1,1981. ,/

~Farley Nuclear Plant Unit 2 35 Docket 'No. ' 50-364 ,,O I I '. F.1 Additional Accident Monitoring Instrumentation (Continued) 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 an independent and redundant Class IE 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 redundant and independent counterpart. The system described above currently meets the January 1,1981 requirements. Alabama Power Company has ordered redundant Vitoreen 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. a. Each detector is designed to measure gamma radiation, 7 b. The range of each detector is 1 R/hr to 10 R/hr for photon radiation. c. The energy response is -15% at 80 key and i8% from 100 key to 3 Mev. d. 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. e. The contairinent high radiation monitor is scheduled for installation by January 1, 1981. 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 pCi/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 for use, calibration of the system, and dissemination of release rate information. To meet the January 1,1981 requirements for a high range effluent radiciodine and particulate sampling system, Alabama Power Company has ordered an Eberline SPING-4 sampl er. This sampler provides the capability to monitor effluent radioactivity in [- ) the form of noble gases, radioiodines, and particulates by use of individual channels for each type of. radioactivity. The sampler will monitor the vent stack with the

Farley Nucitar Plant- ' Unit 2 36 Docket No. 50-364 A() II.F.1 Additional Accident Monitoring Instrumentation (Continued) 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 radonthoron daughter activity contribution. The range of the channel is 2.6 x 105 counts per minute per microcurie on the paper for Csl37, The radioiodine channel monitors a silver xeolite cgridge in the air stream with a sensitivity of approximately 80k CPM per pCi of I Both the particulate channel and the radioiodine channel use external sources for calibration and can be compensat-4 ed for background radiation. Calibration will be performed upon installation and at internis not exceeding each refueling outage. Procedures will be developed for use of the system, calibration of the monitor, and dissemination of release rate information. '4 i i i a 1 f I !O 2 --rm e. 3.. ,,-4,_~_,__,_,,.-- --me._ ,,.,w. . c-., w m m, - - n -m e

Farley Nuc1 car Plcnt Uni t 2 - Docket No. 50-364 37 q U II.F.2 INADEQUATE CORE COOLING INSTRUMENTATION Posi tion Develop' procedures to be used by operators to recognize inadequate core cooling with currently installed instrumentation in PWRS. Install a primary coolant saturation meter. Provide a description of any additional instruments or controls needed to supplement installed equipment to provide unambiguous, easy-to-interpret' indication of inadequate core cooling, procedures for use of this equipment, analyses used to develop these procedures, and a schedule for installing this equipment. This requirement shall be met before fuel loading. (See NUREG-0578, Section 2.1.3b, and letters of September 27 and November 9,1979).

Response

Procedures and Description of Existing Instrumentation The Westinghouse Owner's Group, of which Alabama Power Company is a member, has performed analyses as required by Item I.C.1 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 Owner's Group, an " Instruction to Restore Core Cooling during a Small LOCA" was included. This instruction pmvides the basis for procedure nQ changes and operator training required to recognize the existence of inadequate core cooling and restore core cooling based on existing instrumentation. Alabama Power Company has incorporated the key considerations of this instruction into the Unit 2 operator training program. Subcoolina Meter . Alabama Power Company will install a primary coolant saturation meter that meets the requirements of NUREG-0660. This saturation meter will be installed prior to fuci load. The subcooling meter provides continuous main control board indication of margin-to-saturation conditions. Two identical channels operate independently except that the same sensor may be input into each channel. Main control . board indication consists of two meters that provide a continuous pressure indica-tion of margin-to-saturation and degrees superheat. Multiple core exit thermo-couples wide range T and T , and redundant safety grade system pressures are used for inputs $Tthe subbling monitor. Two thermocouples per quadrant nearest the center of the core were selected to monitor exit temperatures. The subcooling monitor is a-highly reliable and testable system powered from a vital instrument bus and environmentally suited to the service condition for the main control room.. Vital power to the subcooling meter is separated from the IE electrical distribution system with fuses. Signals for the core subcooling monitor are picked up c.. the isolated (control) side of the protection channel which ensure that the addition of the subcooling meter does not adversely impact - Q the reactor protection or engineering safety features systems. Table 1 provides C a summary of information required for the subcooling monitor. Emergency procedures will provide for backup methods to determine subcooling. Appropriate training will be conducted for_ these procedures prior to fuel load.

k Farley Nudlear.' Plant. -Unit 2-38 -Docket No.-50-364' p

TABLE 1

.INFORMATION REQUIRED FOR THE 4 SUBC00 LING MONITOR (1 of 3) DISPLAY l.; .Infomation displayed P - Psat subcooled T'- Tsat superheat l 2. Display type. - Analog and Digital -3. Continuous or on demand Analog - continuous Digital - on demand

4. -Single or redundant display Redundant 5..

Location of display Meter - main control board Microprocessor - main control room instrument racks ~ 0 6. Alarms (include setpoints) Caution: 25 F subcooled for RTD i 150F subcooled for T/C i Alarm: 00F subcooled for RTD and T/C ~ 7. 0verall uncertainty Digital - 40F for T/C; 30F for RTD Analog - 50F for T/C; 50F for RTD - 8. - Range of ~ display - Calibrated region - 1000 psi sub-0 cooled to 2000 F superheat .i overall; never offscale. -9. Qualifications None at present CALCULATOR i 1. Type Dedicated digital =2. -If process computer is used, N/A . specify availability 4 3. Single or redundant calculators - Redundant 4. Selected logic Highest Temperature for RTD or T/C and lowest pressure

15. Qualifications -

None at present M-6. Calculational technique Functional fit - ambient to critical ~ . point n .i ..,s ....,-_,,,2_..,...,,_._ _._,_.,,,_-._,.____..,,..,___,m..

' Faricy Nuclear Plant. 39 ~. Unit 2, . Docket No. 50-364 [] .V TABLE 1 INFORMATION REQUIRED FOR THE SUBC00 LING MONITOR (2of3) INPUT 1. Temperature (RTDs or T/Cs) RTD, T/C, and T ref 2. Temperature (number and location-RTD - 2 hot and 2 cold legs ofsensors) per channel 0 3. Range of temperature sensors RTD 700 F T/C 16500F (calibration 0 .(unit range 0-2300 F) 4. Uncertainty of' temperature 10.7% RTD sensors 5. Qualifications IEEE 323 1971 6. Pressure (specify instrument used) RCS Wide Range Pressurizer (# - 7. Pressure (number and location 2 wide range - Loops 1 and 3 of sensors) 1 narrow range - Pressurizer (per channel) 8. Range of pressure sensors Wide range 3000 psi Narrow range - 1700-2500 psi 9. Uncertainty of pressure Wide range - 11% Narrow range - il.5% Pressurizer - 11.0%

10. Qualifications IEEE 323 1971 BACKUP CAPABILITY 1.

Availability of temperature Temp - Swap between T/C and RTD. and pressure' Press - Can defeat any of the three inputs. System uses auctioneered low pressure. 2. Availability of steam tables Saturated steam tables and tables to verify required sub-cooled conditions are included in Emergency Procedures. U J

_. = _ _ ~ !Il ^ IFarlfy; Nucl' ear 'P1 art 10 nit 2.. 40 ' Docket No. 50-364 TABLE 1 INFORMATION REQUIRED FOR THE SUBC00 LING MONITOR (3 of 3) BACKUP CAPABILITY (CONTINUED): 3. . Training and operators Operators have been trained on the use of the subcooling monitor to determine required subcooling co nditions. ?

4. ~ Procedures-

. Lmergency procedures have been revised to describe the utilization. of the subcore cooling monitor readout and appended portions of the steam tables to determine subcooling conditions. A system operating procedure has been written to guide operators in the operation . of the subcooling monitor. Appro-priate personnel have been trained in these procedures. l i V f 6 l ~ _ _ _., - - -. ~.. - _

j -Farity Nucl:ar Plant Unit 2 41 ' Docket No. 50-364 g V .II.F.2 ' Inadequate Core Cooling Instrumentation (Continued) Additional Instrumentation to Indicate Inadequate Core Cooling The submittal references in " Procedures and Description of Existing Instrumenta-tion" ~ bove. described the capabilities of the core exit thermocouples in a determining the existence of inadequate core cooling conditions and their super-iority in some instances to the loop TRDs for measuring true core conditions. Other means of determining the approach to or existence of inadequate core cooling could be: 1. Reactor vessel water level. 2. Incore detectors. 3. Excore detectors. 4. Reactor coolant pump motor currents. 5. Steam generator pressure. A discussion of the possible use of these measurements is addressed below. The use of the incore movable detectors to determine the existence of inadequate core cooling conditions appears doubtful. The detectors could be driven into the tops of the incore thimbles, which are located at the top of the core, following an eccident 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 cora uncovery. Gama p) y detectors could perhaps be employed, but they suffer from similar sensitivity problems and the fact that gamma levels in the fuel region change insignifi-cantly between the covered and uncoverd 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 powerrange 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 Westinghouse Owner's 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. 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 los_s of offsite power or pump trip because of a LOCA blowdown shuts ' the pumps down. 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 cooling, 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

=- u Farley Nuclear Plant -Unit 2 42 Docket No.E 50-365 : II.F.2 Inadequate Core; Cooling Instrumentation (Continued) . the existence of inadequate core. cooling. Several systems for determining water level are under review by the Westinghouse Owner's Group. These . systems are still under research and development and for this reason Alabama i Power Company has not comitted to install a particular system at this time. 1 LO J 4 i i I " MANAGEMENT DISCUSSION"

i ' u; % Mi..W s M !Farltylfiuclear Pll ant 7 (Unit}21,. fW tDocket No.'(50-364i 43 ..rg g .(m); JII;G -EMERGENCYiPOWERIFOR-PRESSURIZER EQUIPMENT Po'sition Y: - LMothe and controlicomponents of the power-operated relief valves and associated block valves and the pressurizer level indication shall be. capable of being Esupplied from the. offsite power ' source. or from the emergency power buses when ioffsite ' power' is notYavailable. Thisirequirementisha11;lbe met before fuel loading. (See NUREG-0578, -Section-2.1.1, and letters of SeptemberL27 and November 9,1979).

Response

.e Power Supply for'PoweN0perated Relief Valves (PORV) and Block Valves

La Farley Nuclear

Plant ha's two PORVs installed. Their associated circuits meet the separation req'uirements for redundant systems with one' PORV powered

'from train "A" and tiheiother powered - from. train "B". Each PORV is air operated and equipped with two-solenoid valves in their respective air lines. The solenoids are-powe're_d from Train A and B._125-Ydc -buses, respectively. These

buses are powered through associated battery chargers which can also be supplied

.from.the train A'or.B l diesel generator upon. loss, of offsite power or from l25-Vdc batteries. A system' will be. installed as a backup system to the instrument air l system for operating the PORVs by fuel load. ~ Each PORV has a-motor ~ operated block valve locateil in its respective piping.

These valves' associated circuits. meet the -separation requirements for redun-dant' systems. Thetoperator 'aiid its associated control circuits for each block -

valve are powered fromfits associated train A or. B emergency ~ bus. The motive and control power interfaces (for the PORVs and associated block

valves) with the, emergency buses are accomplished through safety grade devices.

Pressurizer Level Indication Power Supply There are three channeIlized pressurizer' level indicators that receive signals from their-correspondi.ng channelized level transmitter. The power supply for each channelized indicator is-~provided by the correspondent 120-Vac vital and. ~ regulated channel '(e.g.,: emergency bus). A e + p' .(- .y i'.. g !O: J ~ n- -+ - 4;T: y c ,- a x . ; Q. !l, e .= !(Q (

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1 Farley-Nuclear P1aht: ' Uni t E2 44' Docket No. 50-364-(,') I' I. K.1 ~IE BULLETINS ON MEASURES TO MITIGATE SMALL-BREAK LOCAs AND LOSS OF ~',

FEEDWATER ACCIDENTS -

Position -C.I.5' Re' view all-valve position. positioning requirements, positive controls and related test and maintenance procedures to assure proper ESF functioning. (See Bulletin-79-06A ' Item 8, 79-068 Item 7, 79-08 Item 6).

Response

Valve positions for-safety related systems are specified in the FNP Operating Procedures. System Operathg Procedures have valve check lists that establish by sign-off the initial valve positions for the start of system operation. The body of these procedures is written such that startup of the system and/or subsequent shutdown will maintain the required. valve position' for the existing mode and/or return the system to the ini tial, conditions. All' engineered safeguards systems have Surveillance Test Procedures which require verification by sign-off correct valve alignment for required flow paths at an interval established by Plant Administrative Procedures and Technical Specifications. . Plant Surveillance Test Procedures are written in such a manner as to return system valves to the required. position for the existing mode. This is accomplished by sign-off steps in the procedure. -O Requirements for oositive controls-(locks on valves, locked out power to M0V's, d etc.).are specified in the Technical Specifications and in plant procedures. The implementations of positive controls is accomplished by sign-off steps in the appropriate plant procedures, e.g., Unit Operating Procedures, System Operating Procedures and Surveillance Test Procedures. It has been the Plant's policy to perform valve check lists or flow path verifica-tion checklist following major. outages, significant maintenance in areas containing safety related equipment, or prior to returning a. safeguards system to service from maintenance or off-normal operation. This policy is incorporated into Plant Administrative Procedures. I'n -addition, shif t relief' requirements include a walk-down of main control room control boards to check proper alignment of remotely operated equipment. For additional controls relative to system status during maintenance activities, . refer to the response provided for item C.l.10. Position C l.10 Review and modify, as required, procedures for removing safety-related systems from service (and restoring to service) to assure operability status is known. (See Bulletin 79-05A Item 10,' 79-06A Item 10, 79-06B Item 9, 79-08 Item 8).

Response

iI Administrative. Procedure 52, Equipment Status Control and Maintenance Authorization, .and Administrative Procedure 16, Conduct of Operations-0perations Group, has been

• reviewed with respect to the requirements for the following:

. Verificatio'n, by test or' inspection, of the operability of redundant safety-a. g

^~ Farley Nuclesr Plast " Uni t' 2 ' ~ Docket No.'50-364 45

1).

'IE Bulletins on Measures to Mitigate Small-Break LOCAs and Loss of' II. K. l!

Feedwater Accidents T(Continued)-

~ .related ' systems' prior to removal.of any safety-related system from service. . b. Verification of the operability of all safety-related systems when they are - returned to service following maintenance or testing.

c. Explicit' notification of-involved reactor operational personnel whenever a safety-related system' is removed from and returned to service.

The following actions with respect to _the above review requirements have been taken: ~ - a. AP-52,LEquipment Status Control and Maintenance Authorization, was revised to ~ require that prior to removing a safety-related system from service for any-reason other than scheduled surveillance testing that the operability of the redundant safety-related system has been verified by one of the following methods: 1. Placing the redundant-safety-related system in service for equipment required to function during normal operation. -2. Testing the redundant safety-related system for fluid systems required to. remove decay heat from the reactor under accident conditions. These - systems consist of auxiliary feedwater and residual heat removal. 3. Inspecting the redundant safety-related system. .These comitments do not apply when a safety-related system is declared inoperable due to.a ' malfunction which requires going under a limiting condition for operation to repair or to instrumentation which is placed in'the tripped condition when removed from service. . b. J AP-52, Equipment Status Control and Maintenance Authorization presently requires that the _ operability of safety-related systems be verified prior to their being returned to service following maintenance or testing (excluding surveillance testing)~ by performance of. applicable surveillance test procedures. This . requirement has been reviewed and incorporated in the training program. c. AP-16, ' Conduct of Operation-- Operations Group, has been changed to make it clear that the Plant Operator at-the-Controls is required to log removal and return to service of all safety-related systems.and equipment. AP-52,

Equipment Status Control and Maintenance Authorization, has been revised to require explicit notification of the operator at the controls by the Shift L Foreman whenever a' safety-related system is removed from service or found to be inoperable and returned 'to service.

Shift turnover and relief require- - ments ensure that awareness of, out-of _ service safety-related systems is main-tained until the. system is declared operable. r . Position u - C.l.17 JFor: Westinghouse-designed reactors, trip the pressurizer low-level

conincident' signal bistables, so that. safety _ injection would be initiated when.the: pressurizer low-pressure setpoint is reached regard-elessLof)the: pressurizer level. (See-Bulletin 79-06A and Revision 1,

. Item 3.

Far'lcylNucicar Plant -Unit 2. 46 ? Docket No. 50-364 p: - () I I. K.1 IE Bulletins on Measures to Mitigate Small-Break'LOCAs and Loss of

Feedwater Accidents-(Continued)

. Response- ~ Placing the low pressurizer level bistables in the tripped condition when these bistables areLrequired to be operable by technical specifications is not required. A design change-has been implemented and technical specification change approved - for the removal of automatic-initiation of. Safety Injection on coincident Low w Pressurizer Pressure and Level and to add the automatic initiation of Safety Injection on 2/3 low pressurizer pressure signals. Pasition C.1.20 For B&W-designed reactors, provide procedures and training to operators .for: prompt manud reactor trip for loss of feedwater, turbine trip, main steamline isolation valve closure, loss of offsite power, loss of steam generator level, and low pressurizer level. (See Bulletin 79-05B, Item 4) C.I.21 For B&W-designed reactors, provide automatic safety-grade anticipatory reactor trip for loss of feedwater, turbine trip or significant decrease in steam generator level. (See' Bulletin 79_-05B, Item 5). (d~3 C.l.22 For. boiling water, reactors, describe the automatic and manual actions necessary for proper functioning of the auxiliary heat removal.

Response

Due to the fact that Alabama Power Company's Joseph M. Farley Nuclear Plant is - a Westinghouse designed Pressurized Water Reactor (PWR), these requirements are not applicable. . j^\\ ) N)

Farity Nucl ar Plant Unit 2 47 Docket No. 50-364 p_ V II.K.3 FINAL RECOP91ENDATIONS OF B&O TASK FORCE Position C.3.9 For Westinghouse-designed reactors, modify the pressure integral derivative controller (PID), if installed on the PORV, to eliminate spurious openings of the PORV.

Response

Upon recommendation of Westinghouse, Alabama Power Company has incorporated into the Farley Unit 2 design, " rate-time constant" in the PID Controller of zero seconds. This, in effect, removes the derivitive action from the controller which decreases the likelihood of opening the PORV since the actuation (opening) signal will not be sensitive to the rate of change of the pressurizer pressure. Position C.3.10 For Westinghouse-designed reactors, if the anticipatory reactor trip upon turbine' trip is to be modified to be bypassed at power levels less than 50 percent, rather than below 10 percent as in current design, demonstrate that the probability of a small break LOCA resulting from a stuck open PORV is not significantly changed by this modification.

Response

The licensing basis for Farley Nuclear Plant includes an anticipatory reactor trip upon turbine trip which has been modified to be bypassed at power levels of 50 per-cent or less. The Westinghouse design criterion is that load rejections up to 50 per-cent should not require a reactor trip if all other functions operate properly. The power mismatch is taken up by steam dump (40 percent) and automatic rod insertion (10 percent). The PORV's are provided to reduce the likelihood of tripping the reactor on the high pressurizer pressure signal and opening the pressurizer safety valves. A series of 50 percent load rejections from full power performed by Westinghouse (WCAP-8142) indicate an increase in reactor coolant pressure of less than 100 psi. A turbine trip at half-power without immediate reactor trip results in a peak pressurizer pressure of 2324 psia, or 26 psi below the PORV opening setpoint. Therefore, the PORVs would not be expected to open and the probability of a small break LOCA due to a stuck open PORV is not significant-ly changed by th e modification. Position C.3.11 Demonstra te that the PORV installed in the plant has a failure rate that is not significantly more than the valves for which there is an operating his tory.

Response

-The two (2) PORV's installed at Unit 2 of The Farley Nuclear Plant were supplied (')'s. by Westinghouse. Vendor information is as follows: L Hanufacturer - Copes-Vulcan Model No. - D-100-160 Size 3" 1500 lb. Type Air operated globe valve

~ Farley Nuclcar.P1 ant' Unit 2 48 Docket No. 50-364 pG.. .II.K.3 FINAL RECOMMENDATIONS OF B&O TASK FORCE (Continued) ' Response (Continued) These PORV's are the same type as those used in the majority of Westinghouse designed plants (including Farley Unit 1) and for which an operating history exists and.a-failure rate has been determined. Westinghouse has provided data to - the NRC which shows that Westinghouse-provided PORV's have opened approximately 60 times during normal operation for various reasons. For each of these openings, the valve reseated correctly. (See NUREG-0611, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants). Position C.3.12 For Westinghouse-designed reactors confirm that themis an anticipa-s tory reactor trip on turbine trip. These requirements shall be met before fuel loading. Respcnse The Farley Nuclear Plant has an anticipatory reactor trip on turbine trip. v 1

.Farley Nuclear. Plant Unit 2 49 Docket 50-364 (I .III.A.1.1 ' UPGRADE EMERGENCY PREPAREDNESS Position Comply with Appendix. E, " Emergency Facilities," to 10CFP Part 50, Regulatory Guide 1.101, " Emergency Planning for Nuclear Power Plants," and for the offsite plans, meet essential elements of NUREG-75/111 or have a favorable finding from FEMA. - This requirement shall be met before fuel loading: -Response The Joseph M. Farley Nuclear Plant Emergency Plan meets the requirements of Appendix E to 10 CFR Part 50 as documented in the Farley SER (page 13.2), Supplement 3 (page 13.1) and has been accepted by the NRC. Revision 3 of the Emergency Plan, submitted in December,1979, includes the requirements of Regulatory Guide 1.101, March 1977. A revision to the State of Georgia Radiological Dnergency Operations Plan is being submitted to FEMA for review. This plan was written to meet the essential requirements of NUREG-75/lll. .The State of Alabama Radiation Emergency Plan which is approved by the NRC is ' (~cT in compliance with the essential requirements of NUREG-75/111. xJ i i {v- %y + y,--- .e a -y y wyv-1 v-'w +-

Farley Nuclear Plant Uni t 2-50 Docket No. 50-364 bh .III.A.1.2: UPGRADE EMERGENCY SUPPORT FACILITIES Position 1 Provide 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 pareters and call up display of radiological parameters. Response. An interim technical support center (TSC) has been established to meet requirements for receipt of an operating license (Figure 4). 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 prepared-ness improvements. The following is a description of the interim TSC. l. Description - The temporary TSC has been established in an 11-x 10-foot office on the west side of the control room (see attached Figure 4). This office is separated from the controls area and provides sufficient space for the emergency director and his staff to ccordinate emergency acti ities. v ~2. Procedures for TSC Staffing and Support - FNP-0-EIP-0, " Emergency Organization f) and Control Room Access," has been issued. It specifies which plant personnel shall report to the TSC during an emergency and delineates their duties and v responsibilities. 3. TSC Communications - The TSC contains the following comunication equipment: a. NRC hot line (red phone). t j b. Two plant public address system units for intraplant communication. c. One intraplant phone capable of reaching other phones within the plant, other phones in the Southern Company operating system and the Alabama Power general office switchboard. From the general office switchboard, it can be connected to comercial phone facilities. -d. Two pushbutton phones, each with the following features: .(1) Two comercial lines on the Graceba Co. Ashford, Alabama exchange. (2) One comercial line on the GTE Dothan, Alabama exchange. (3) One. Alabama Power general office extension with the intraplant system capabilities discussed under 3C. (4): One Daniel Construction Company of Alabama (DCCA) extension capable O of reaching other DCCA extensions on site and outside comercial lines V on the Graceba Co. Ashford, Alabama exchange.

- -.. = - - v~ Farleylfbclear Plant - . Uni t~' 2 ? 51 ' Docket No. 50-364 _ \\ ' .III.A.1.2 Upgrade Emergency Support _ Facilities (Continued) Control room communication can be made by intraplant line, public address line, or by voice due to.the close' proximity of the TSC to the controls v area. Communication with the operation support centers can be made by 4 intraplant phone. 4.:l degree as the main control room controis area for postulated accident conditi Radiation Monitoring ~ and Action Levels The TSC:is habitable to the same The.TSC HVAC is common with the control room and includes a process monitor which Jautomatically initiates recirculation mode on detection of high radiation. ~ An 7-area monitor on the. northwest corner of the Unit 1 contmls area provides radiation monitoring and alarm. Action-levels' for retreat to the controls area are unwarranted since conditions will be identical to those in the TSC. .5. Access to Technical Data and Display of Plant Parameters - A complete set of. P & ids is maintained in. the TSC for emergency use. Any other prints or ? technical-data are readily available from the document control facility in the . Plant Service Building ~which may be reached by intraplant phone from the TSC. ~ Plant data can be displayed in the TSC by relocating one of two Unit 2 i. controls area-CRT computer readouts to the TSC. Cable has been fabricated and is in place to permit this. The CRT.is mounted on a dolly and could be .placed in service with 5-10 minutes :giving access to any parameter monitored by the plant computer. Display format is as follows: a._ One group of points is displayed and immediately below the group the most recent 16 alarms. L -b. Any one'of -16 groups may be selected for display. E. c. A group may consist of up-to 26 parameters selected by the computer operator. J. 6. Procedures fon Accident Assessment from the Control Room - Since the temporary 4 TSC is,; for all-practical. purposes, part of the control room, this requirement .is ' not applicable. lhen the' permanent TSC is established, procedures will be ' modified to reflect accident assessment from ~the temporary TSC facility should the permanent TSC become uninhabitable'. - The following is a description of-the proposed long-term onsite technical support center to meet the January 1,1981 requirements. This design covers the anticipated manning' requirements, document stcrage, monitoring systems, comunication,_ habitability, L - and structural arrangements'. Le center has been designed to be used for post-j~ accident ~and normal-plant :.onditions. - Conccotual Design !1'. Location _.The TSC Lis;1ocated in the. Auxiliary Building, elevation 155',

imediately

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 i h~:

room' area.; Passage through the control room will not interfere with the E operators 'who are_ manning the control boards. An emergency or alternate. exit Lis from the north side of the TSC and through 'an exit door in the northwest - g wall of:the Unit 2 Auxiliary Building.. ~ 1 b ^ s -rv e wvei V,+ww=y <r=g-v ,r--.ipre,pv v y---~1eT* -'rty5-* vt 4 =-' q N-S F

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Faricy Nuclear Plant Unit 2 52 Docket No. 50-364 f3 V III.A.l.2 Upgrade Emergency Support Facilitics (Continued) 2. Size aid 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 5 shows a proposed room layout. An overall space of 22 feet x 65 feet, with a 9-foot ceiling height, has been provided. Room layout is es follows: a. Monitoring Area - Includes CRT, trend line printer for each unit, and provision for communication with the 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 for usage. Phones will be provided for full consnunication capability. A chalk board and pro-jection screen will also be located here. Figures 6 and 7 describe the TV monitoring system. c. Document Room - Includes files for microfilm, drawings, data sheets, 3 indexes, ar/ microfilm reader and printer. (~) d. Conference Area - Includes area where phones will be provided for full ^ communication capability. 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 vital plant parameters from the time an accident begins; TV cameras and monitors will be turned on as needed. 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 request display of the desired data in the TSC. One of the two CRTs normally located in the main control room (one per unit) would be disconnected, rolled into the TSC, and hooked up apon 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 Building, elevation 121') would be moved to the TSC. Parameters could be trended on the line printers providing a hard copy of data for review. 6. Communications - Communications will be prov:ded to contact the NRC, State of Alabama and Alabama Power General Office. Sound-powered phones will be (9 provided between the TSC and main control room. An intercom system will be v available between the TSC, the three operations support centers, and the emergency operations center. Comercial and APC telecommunications lines will be available for redundant offsite communications. An APC security radio will also be located in the TSC.

Faricy Nuclear Plant Unit 2 53 Docket No. 50-364 h (O III.A.l.2 Upgrade Emergency Support Facilities (Continued) 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 pressurizing the TSC area and recirculating the room air through the charcoal filter. A pennanent 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. 8. Fire Protection - The TSC is enclosed b -hour fire-rated walls with Class A airtight fire doors. The present area,.oom 2452, is provided with fire and smoke detectors and a wet pipe sprinkler system. A water hose cabinet and C02 hose reel are presently located within the TSC area. One CO2 and one dry chemical fire extinguisher are provided inside of the TSC. Emergency 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 1B and 2B, channel 2. b. HVAC - 600-V MCCs-lF and 1G. These are shared MCCs and can be powered from ( the diesel generators upon LOSp. c. Wall Receptacles - 120/208-V distribution panel LL. This shared panel can be powered from the diesel generator upon LOSP. d. Lighting - 26 fluorescent fixtures, four 40-W 1 amps each; 277-Vac from MCC-1F or 1G (Unit 2) and MCC-2CC or 2DD (Unit 2); 25-W emergency lighting; 8-hour battcry packs, two 2-head, two 1-head; tied to MCC-2DD (Unit 2), train B. HVAC System Description This system is composed of one (6.4-ton) air conditioning unit with air-cooled condenser, one emergency filtration unit equipped with a separate fan, two 6-inch butterfly valves for positive isolation from contaminated outside air, and associated ductwork and controls. Refer to Figure 8. 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 recirculated air is directed through the emergency filter, thus maintaining the pressure within the conditioned space and establishing a cleanup mode of operation simultaneously. The temperature is maintained by a room thermostat which cycles either the compressor or the electric heating coil within the air conditioning unit upon demand. The filtration unit contains a 50 percent prefilter, a 99.97 percent HEPA filter, plus a carbon filter.

r Faricy Nuclear Plant 54 Unit 2 Docket No. 50-364 ,m 1 >V III.A.l.2 Upgrade Emergency Support Facilities (Continued) Position 2 Establish an onsite operational support center, separate from,but with comunications to, the control room for use by operations support during an accident.

Response

The southeast corner of the control room has been designated as the operations support center for operations and chemistry and health physics personnel (Figure 9). This area is separate from the controls area and will provide rapid response to requests for operations or chemistry and health physics support by the shift supervisor while precluding congestion of the controls area. All other plant support personnel will report to operations support center areas designated in the plant Service Building. Comunications exist between these areas and the controls area. The plant Emergency Plan will be revised to reflect the operational support centers and to establish methods of management and lines of communication as part of the improved emergency preparedness effort prior to receipt of an operating license. Position 3 Establish a near-site emergency operations facility with communications with the O plant to provide evaluation of radiation releases and coordination of all onsite C/ and offsite activities during an accident.

Response

Alabama Power Company Construction Offices have been designated as the interim Emergency Operations Facility for Plant Farley. These offices are located on the plant site outside the primary security area. Offices are available to accommodate federal, state and local officials and communications to the plant and offsite are provided. A permanent Emergency Operations Facility is presently under construction. This facility will accommodate a minimum of sixty (60) officials and will provide: (1) Separate offices for each group of officials, (2) A large assembly room for the dissemination of information, (3) Reliable onsite and offsite communications capabilities including a dedicated intercom to the plant Technical Support Center and Operation Support Centers, dedicated lines to the APCo General Office, and a State civil defense " hot-line" phone, (4) Conference rooms, and (5) An isolable ventilation system with HEPA filters. . O, This facility will be located on the plant site outside the primary security area. j e 1 i

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61 Farley Nuclear Plant Unit 2 Docket No. 50-364 ,V ~ III.D.3.3 IN-PLANT RADIATION MONITORING Position Provide the equipment, training, and procedures to accurately measure the radioiodine 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

A.abama 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. Alabama Power Company presently has the capability of purging these samples of entrapped noble gases by the use of nitrogen gas and analysis by Ge(Li) gama ray spectroscopy in a low background counting facility. oG

4 h 2 .A.ah w h,&wa-a- ~-4 +-4<J,.,._A hhm m._y,m .-a,-. ._a l 1 i i i I n i a s ) 1 I l 1 l d i i I l FULL-POWER LICENSE REQUIREMENTS 9 4 I i e i ( i 4 i t 1 l t F ~,.- ,__4... ._,.._.__.m._.,-,. vv., ..-y-

Faricy Nucitar Plant Unit 2 62 Docket No. 50-364 ,-~ N I.C.7 NSSS VENDOR REVIEW 0F PROCEDURES l Posi tion l Obtain nuclear steam supply system (NSSS) vendor review of power ascension test i and emergency proceduresto further verify their adequacy. l This requirement must be' met before issuance of a full-power license. . Response Power ascension tests associated with NSSS vendor, and emergency procedures will be reviewed by the NSSS vendor (Westinghouse) prior to issuance of a full power i license. l l 1 l i l i t

Farley Nucicar Plant Unit 2 63 Docket No. 50-364 I. C.8. PILOT MONITORING 0F SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPLICANTS Position Correct emergency procedures, as necessary, based on the NRC aucit of selected

-plant emergency operating procedures (e.g., small-break LOCA, loss of feedwater, restart of engineered safety features following a loss of ac power, steam-line' break, or steam-generator tube rupture). This action will be completed prior. to issuance of a full-power license.

Response

i Selected plant emergency operating procedures will be revised, as necessary, based on NRC audits prior to issuance of a full-power license. 4

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.y/ Farity Nuc1 car Plant 4 Unit 2 ~ Docket No. 50-364 64 (3 i I.G.1 TRAINING DURING l.0W-POWER TESTING Posi tion. Supplement operator training by completing the special low-power test program. Tests may be observed by other. shifts or repeated on other shifts to provide training to the operators. This requirement shall be met before issuance of a full-power license.

Response

Operator training will b'e supplemented by completing the low-power test program as outlined in the response to I.G.1 (Fuel Loading and Low Power Testing Requirements). l O

FarlGy NuclGar Plant Unit 2 65 Docket No. 50-364 II.B.1 REACTOR COOLANT SYSTEM VENTS Posi tion Provide (1) a description of the design of reactor coolant system and reactor vessel head high point vents that are remotely operable from the control room; (2) anlayses of loss-of-coolant accidents initiated by a break in the vent pipe; and (3) analyses demonstrating that direct venting of noncondensable gases wie, perhaps high byfrogen concentration limits does not result in violation of combustible gas concentration limits in containment. This requirement shall be met before issuance of a full-power license. (See to letters of September 27 and November 9,1979).

Response

Alabama Power Company will install a reactor vessel head vent system by January 1, 1981 or prior to full power operation, whichever is later. The following is a description of the sytem proposed:

System Description

The reactor vessel head vent system (RViiVS) is designed to remove non-condensable gases from the reactor vessel and head area. The RVHVS is designed to vent in excess of one-half the reactor coolant system volume of hydrogen at 6500F in one hour from one of two available flow paths. The RViiVS is orificed to limit the blow-down from a break or inadvertently opened flow control valve downstream of the orifice to the capacity of one charging pump. The system is operated from the control room and has control room indication of valve position. The system is completely safety grade and meets the single failure criteria for venting initiation and termination. The reactor vessel head vent valves are powered by train A and B of the plant class IE dc distribution system and fail closed. The head vent valves in series are powered from the same train. This method of powering valves in conjunction with the fail-closed design allows the RViiVS to meet the single failure criteria. Each RVHVS flow path consists of two nonnally closed, normally deenergized valves as shown in Figure 10. The two-series valve arrangement eliminates the possibility of a spuriously opened flow path due to the spurious movement of one valve. The RVHVS will discharge into a well-ventilated area of the containment in order to ensure optimum dilution of combustible gases. Inside the containment, hydrogen can be recombined by means of the post-accident hydrogen recombiners. The discharge point will be designed for adequate drainage of reactor coolant in the cases of inadvertent di scha rges. The vent piping will.a supported to result in acceptable stresses on the RVHVS piping and the vent / vessel connection resulting from seismic, thermal, dynamic, pressure, and deadweight loads. The piping downstream of the first anchors down-stream of the second isolation valves may be nonnuclear safety; however, piping downstream of the second isolation valves must not transmit loads back to the Safety 3 ass 2 piping. Each vent line will be seismically qualified. As the potential exists for water flow through the vent line af ter the vessel has been refilled after venting, the vent system piping shall be supported for water flow up to the Safety Class 2 nonnuclear safety boundary. L

. Faricy Nuclear Plant' Unit 2 66 Docket No. 50-364 10 11.B.1 Reactor _ Coolant System Vents (Continued) i .The PORVs will be 'used-to vent the pressurizer. The reactor coolant leakage detection system provides.the capability of detecting the presence of significant radioactive or nonradioactive leakage from the i reactor. coolant loops to the containment atmosphere during normal operation. i

Variations in the particulate activity, gaseous activity, and specific humidity of. the containment atmosphere above a pre-set level give positive indications

- in the control room to the reactor operators. These leakage detection provisions are sufficiently sensitive so that 'small increases in leakage rates can be detected while the total leakage rate is still below a value consistent with safe operation of the plant. The particulate and gaseous activity are monitored by the containment air particulate (R-ll) and radiogas (R-12) monitors. Humidity is monitored by the containment air coolers' condensate level measuring system. _ An increase -in R-ll, R-12, or containment air cooler condensate level is an indication of reactor coolant leakage -into containment. T i 2 4 i k i - -c +.,, ,,-v ,+ ,-.a .,.7 .,,.n,...,. w. - -- ..m..-

67 TO CONTAINMENT / o o 2 4 S S 1 3 ,g; , d' S S Jk EXISTING VENT LINE N-ORIFICES O .s t,. E' REACTOR h, VESSEL ~' HEAD 1 !O t Figure 10 Flow Diagram of the Reactor Vessel Head Vent System f

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Faricy NucicarPlant Unit 2 Docket No. 50-364 68 II.B.2 PLANT SHIELDING Posi tion Provide (1) a radiation and shielding design review that identifies the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by radiation during operations following an accident resulting in a degraded core, and (2) a description of the types of corrective actions needed to assure adequate access to vital areas and protection of safety equipment. This requirement must be met before issuance of a full-power license. (see NUREG-0578, Section 2. i.6b, and letters of September 27 and November 9,1979.) R_esponse Previous shielding design studies did not take into account the source terms now required by the NRC. A shielding review has been initiated to identify areas which need improved shielding. This shielding review will use NRC source terms. The review will identify areas requiring access for operation of essential safe s%tdown equipment. The radiation sources in these areas will be identified by 6 field walkdown or design review. The design review will be completed prior to receipt of an operating license. Aspects of all major acci-dents will be covered in the design review, and shielding additions developed in e vJ the design review will be implemented by.lanuary 1,1981 or prior to full power operation, which ever is later. A summary of the design review is given below. A. Design Review Zone Boundaries A design review for the Farley Plant Unit 2 will be conducted by Bechtel Power Corporation, using the TID source terms and the 10CFR20 and GDC19, 60-64 of Appendix A to 10CFR50, dose criteria. This review will establish zone boundaries based on the following: A-I - The first zone will be consistent with the personnel radiation exposure guidelines for vital areas requiring continuous occupancy (<0.015 rem /hr. ) A-II The second zone will be consisteat with the personnel radiation exposure guidelines for vital areas requiring infrequent access or corridors to these areas (0.015 to <0.100 rem /hr). A-III The third zone will be consistent with the personnel radiation exposure guidelines as given in General Design Criterion 19. Thisincluded a time motion study to ensure that the integrated exposure was not greater than 5 rem (0.100 to <5 rem /hr). Subsequent zones were selected by grouping them by powers of 10, so that rapid assessment of additional shielding measures could be used via " tenth value h layers" of conmon shielding materials. They include: A-IV 5 to 50 rem /hr A-V 50 to 500 ram /hr A-VI 500 to 5,000 rem /hr A-VII 5,000 to 50,000 rem /hr A-VIII 50,000 to 500.000 remIbr----

Faricy Nuclear Plant Unit 2 69 Docket No. 50-364 I i V II.B.2 Plant Shielding (Continued) CAUTION: These zone designations should not be confused with those used for normal plant operation zone maps found in the Farley FSAR, Chapter 12. B. Scope of the Design Review 1. Selection of Systems for Shielding Review The criteria applied in selection of plant systems used in the shieldirg review has resulted in several classifications of systems as cuttined in the following: a. Category 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 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 /3 not adversely impact operator or equipment functions required q,/ outside the containment. Therefore, the following systems will be reviewed to ensure that this first priority safety concern is adequately addressed by the existing plant shielding design. (1.) Those portions of the containment spray system used to recirculate water from the containment sump back into containmen t. (2.) Those portions of the residual heat removal system used to recirculate water from the containment sump back into containment. (3.) 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. b. Category B (Extensions of Containment Atmosphere) In addition to systems listed above, there are other systems or portions of other system which could contain radioactivity by virtue of their connection to the containment following an accident. Proper operation of the emergency core cooling systems would prevent extensive core damage and mean that these systems would not be expected to contain the significant radioactive sources required by l ? this special analysis. Nevertheless, such sources have been postulated in the following systems. (1.) Those portions of the post-accident containment combustible gas control sy*, tem external to the containment which would contain the atmosphere from the containment.

Farlcy Nuclcar Plant Unit 2 70 Docket No. 50-364 ) II.B.2 Plant Shielding (Continued) (2.) 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. (3.) Those portions of the sampling system used to obtain a containment atmosphere sample. c. Category C (Liquid Samples) Lessons Learned Task 2.1.8 requires that certain post-accident liquid samples be obtained from the reactor coolant system or containment systems. Those portions of the sampling system which must be used to meet the intent of Task 2.1.8 will be reviewed. d. Category D (Letdown) That portion of the letdown system from the reactor coolant system past the letdown heat exchanger into the VCT and into the suction of the charging pump will be analyzed. ) 2. Quantification of Potential Radioactive Source Release Fractions The following release fractions will be used as a basis for determining the concentrations for the shieldina review: a. Source A: Containment atmosphem - 100 percent noble gases, 25 percent halogens. b. Source B: Reactor coolant - 100 percent noble gases, 50 percent halogens,1 percent solids, c. Source C: Containment sump liquid - 50 percent halogens,1 percent sol ids. The above release fractions will be 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 are assumed to be 1 precent for the purpose of this i,hielding review. Further evaluations of the TMI radio-activity releases may conclude that higher release fractions are appropriate. However, the overall effects of higher release fractions on radiation levels or integrated exposures are not expected to be significant. There-fore, the Regulatory Guide 1.7 solids release fraction, I percerit, will be used in the review. Similarly, no noble gases are includalin the containment sump liquid (Source C) because Regulatory Guide 1.7 has also set this precedent J in modeling liquids in the containment sump. Furthermore, cursory analyses have indicated that the halogens dominate all shielding requirements ar.d that contributions to the total dose rates from noble gases ar7 negligible for the purposes of a shielding design review. l

Farlcy Nucl:arPlant Unit 2 71 Docket No. 50-364 ) II.B.2 Plant Shielding (Continued) 3. Source Term Models Part 2 above outlines the assumptions used for release fractions for the shielding design review. These release fractions are, however, only the first step in modeling the source terms for the activity concentrations in the systems under review. The important modeling parameters, decay time and dilution volume, obviously also affect any shielding analysis. The following sections outline the rationale for the selection of values for these key parameters. a. Decay Time For the first stage of the shielding review process, no decay time credit will be used with the above releases. The primary reason for this is to develop a set of accident radiation zone maps normalized to no decay that could be used as a tool by plant staff along with a set of decay curves to quantitatively assess the plant status quickly following any abnormal occurence. Zone maps and decay curves will be developed in accordance with boundaries and radiation level criteria as discussed in part A of this section for identifying problem area. However, the following decay times will be used in assessing ,,T (V anticipated potential personnel radiation exposure due to those operator actions required post-LOCA. (1.) for analysis of personnel exposures in vital areas outside the control room, radioactive decay equivalent to 10 minutes allowed for operator action will be used as the minimum decay time. (2.) Additional decay time will also be allowed for the review of all those ECCS systems previously outlined which are used to recir-culate water from the containment sump back into the containment. That decay time is 24 minutes, which is consistent with the time for initiation of recirculation in accordance with the Farley FSAR, Chapter 6. b. Dilution Volume The volume to be used for dilution is important, affecting the calculations of dose rate in a linear fashion. The following dilution volumes will be used with the release fractions and decay times listed above to arrive at the final source terms for the shielding reviews: (1.) Source A: Containment free volume. The volume occupied by the ECCS water is neglected. pC) (2.) Source B. Reactor coolant system volume based on reactor coolant density at the operating temperature and pressure. (3.) ' Source C: The volume of water present at the time of recirculation (reactor coolant system and refueling water storage tank and safety injection tatiks).

Farley Nuclcar Plant 72 Unit 2 Docket No. 50-364 w-w s II.B.2 Plant Shielding (Continued) c. Sources Used in Piping and Equipment for Each System Under Review In defining the limits of the connected piping subject to containment listed below, nonnally shut valves are assumed to remain shut. (l.) Containment Spray System: At the initiation of recirculation, Source C is to be used. (2.) High Head Safety Injection System: At the initiation of recircu-lation Source C is to be used. (3.) Residual Heat Removal System: Source C is to be used for sump recirculation mode. (4.) Sampling Systems: The sources to be used in the shielding design review for sampling systems e ae as follows: -Containment air samples - Source A. -Reactor coolant "mple - Source B. (5.) Letdown System: .1 siquid source to be used is Source B. ,3b 4. The Shielding Design Review Methodology a. Analytical Shielding Techniques The previous sections outlined the rationale and assumptions for the selection of the systems + hat would undergo a shielding design review as well as the formulation of the sources for those systems. The next step in the review process will be to use those sources along with the standard point kernel shielding analytical techniques to estimate dose rates from those selected systems. For compartments containing the systems unaer review, estimates will be made for general area dose rate rather than to superimpose the maximum dose rate at contact with the surfaces of. all individual components of that system in the conpartment. For corridors outside compartments, reviews will be done to check the dose rate transmitted into the corridor through the walls of adjacent compartments. Checks will be made for any piping or equip-ment that could directly contribute to corridor dose rates, i.e., piping that may be running directly in the corridor or equipment / piping in a compartment that could shine directly into corridors with no attenuation through the compartments walls, b. Accident Radiation Zone Maps ^ T One of the two principal products of this review will be the series of (V accident radiation zone maps. These zone maps will represent the corre-lation of the average dose rates as estimated above with the required operator actions and resultant necessary accessibility to vital areas. By using these zone maps along with the decay curves, potential problem areas will be identified and reviewed. The results of these reviews will be used to formulate plant shielding modification recommendations.

Faricy Nuclear 'P1 ant Unit 2 Docket No.~50-364-73 m II.B.2 Plant Sh'ielding (Continued) c. Environmental Qualification Methodology Most organic materials used'for insulation, seals, etc., are generally unaffected by integrated gama doses below 100,000 rads (carbon). For the purpose of estimated integrated les to safety related equip-ment near the system components analyzed . the shielding design review, the following criterion will be used: Those areas shown in the zone' maps with a zone designation of VI, VII, or VIII (dose rates between 500R/hr and 5000,000 R/hr will be identified and reviewed).- The results of this are that the total integrated ex-posure fo r sources -listed in the selection of systems for shielding reviews under. category A could exceed 100,000 rads integrated dose-in about one or two days. This was ascertained by developing a . corresponding set of integral energy release curves as a function of time for the same sources for which decay time curves were developed. Details concerning the foregoing shielding review including specific systems in the source term, zone maps, decay curves, and specific component locations,.will be available at the Farley Nuclear Plant for reviewand inspection. bs) 4 A study will be conducted to determine the effect of pest accident dose rates on v components located in areas outside the containment from systes processing primary coolant and containment sump fluids. The following is a discussion and sumarization of the study. Based on the calculated dose rates, the total integrated doses over periods of one day, ten days, and thirty days will be calculated for the various Class lE component locations, and compared to the qualification or radiation tolerance levels for those componeots. i Five major areas have been detennined to be affected by post-acci. lent high radia-tion fluids outside the containment. These are as follows: (1) Containment Spray Pump Rooms _. ( 2 ) RHR/LHSI Pump Roams (3) RHR Heat Exchanger Room (4 ) Charging /HHSI Tump Rooms (5) Piping Penetration Rooms- . The sourca term used in producing the dose rates calculated in our. analysis includes 50% Mlog,..s,1% fuel, and the RCS-inventory diluted by the RWST and Safety Injection - tanks. The does rate for each component will be calculated as follows: (1) The piping systems surrounding the component for a distance of approximately 25 feet will be bmken down into a number of ' straight-line elements. (2) Each element will be categorized according to the pipe diameter and distance from the V component. lit was originally intended to consider each elemental pipe length in' the. calculation; however, in_ order to simplify the calculation, and the numberof dose-rate graphs required, and at the same time provide additional j conservatism,: all pipe elemer.ts will be considered a be infinitely long. (3) The sourcerate contribution for each element will be determined from 'an appropriate .N

' Farley Nuclear Plant Unit 2 - 74 ' Docket No. 50-364 II.B.2 Plant Shielding (Continued) dose rate oraph based on Bechtel' Power Corporation computer program "CLYS0". The total dose rate for the component is the arithmetical sum of all contri-butors. (4) The integrated dose for periods of one day, ten days, thirty days, and 180 days will be determined by integrating the dose rates for the appropriate times,taking into consideration the decay rates of the various fission products. The Unit 1 study verified the documented radiation qualification levels to be less than the calculated thirty day dose. The Unit 1 study also identified considerable margin between the calculated dose and the documented qualification level. Based on the similar designs between Units 1 and 2, we do not anticipate any Unit 2 equipment will be identified with documented radiation qualification levels less than the calculated thirty day dose. O cd

Farley Nuclcar Plant Unit 2 Docket No. 50-364 75 II.B.3 POSTACCIDENT SAMPLING Position Provide (1) a design and operational review of the capability to promptly obtain and perfom radioisotopic and chemical analyses of reactor coolant and containment atmosphere samples under degraded core accident conditions without excessive exposure, (2) a description of the types of corrective actions needed to provide this capability, and (3) procedures for obtaining and analyzing these samples with the existing equipment. This requirements must be met before issuance of a full-power license. (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 unpres urized) 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 q atmosphere sample will be obtained by modifying the existing monitoring system Q 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 transported in a shielded container to a remote temporary area for Ge(Li) gamma ray spectroscopic analysis. Procedure modifi-cations for handling and analysis of samples, plant modifications, and a design review of the sample system are being completed. A description of the post acci-dent sampling system is as follows: R_eactor Coolant (Pressurized and Unpressurized) The Unit 2 reactor coolant sampling system will be modif:ad by addition of a post-accident sampling panel and shielded sample pass througt (Figure 11). A schematic of the post accident sampling panel is shown in Figure 12. The above system has the capability for remotely taking a pressurized sample (RCS) and an unpressurized sample (RHR/ containment sump). Based opon the results of the shielding design study with subsequent modifications (2.1.68) and time studies for drawing the samples, the estimated whole body or extremities radiati doses to any individual will not exceed 3 and 18 3/4 rems, respectively. Vent Stack Effluent The Unit 2 vent stack air particulate and noble gas monitor (RE 21 & 22) will be modified by installation r,f a post-accident particulate and iodine sampler, as well as addition of a septum for drawing a gas sample (Figure 13). Subsequent to an accident the normally used radiation monitor (RE 21 & 22) will be valved n I i out and the effluent flow directed through the post-accident sampler for a V specified time period by operation of the remote control panel. Based upon the r,esults of the shielding design study with subsequent modifications (2.1.68) 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.

Farley Nuclcar Plant Unit 2 76 Docket No. 50-364 II.B.3 ~ Postaccident Sampling (Continued) Contr.inment Air The Unit 2 containment air particulate and noble gas monitor (RE 11712) will be modified by installation of a post-accident particulate and iodine sampler, as well as addition of a septum for drawing a gas sample (Figure 14). Subsequent i 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 a specified time period by operation of the remote control panel. Based upon the results of the shielding design study with subsequent modifications (2.1.68)'and sanpling time studies, tne estimated whole body or extremities radiation doses to any individual will not exceed 3 and 18 3/4 rems, respectively. i The procedures and plant system modifications described above will be completed prior-to receipt of an operating license or January-1,1981 which ever is later, and are intended to meet the requirements for a full power operating license. /~s I i 4 k G 9 g_ ~..

.. ~. _ _ .. -.. _.. ~,. ~.. . ~ ~ - -. ~.. 1 SIDE VIEW '. 1 REMOTE CONTROL PANEL-POST ACCIDENT SAMPl.ING ] PANEL () : REMOVABLE SHIELDED PLUG 7lF t LEAD SAMPLE PIG (S MOVED FORWARD ON TRACK oe TRACK ON CART j MATES UP TO TRACK PERMANENTLY MOtJNTED f IN CORE DRILLED y PASSAGE 1 PIG COMES TO REST i HERE, WHERE LIQUID SAMPLE I IS DISPENSED TO SAMPLE VIAL ( [ i o o f EAST WALL OF SAMPLE ROOM i a u 1 N I T l I 1 I 4 Figure 11 Reactor Coolant Sampling System i i 4 .I 1840-0 1

"s t)f% ()w .J 7-------- TV 2 GAGE g g TO VACUUM g-g~ Rvt L...------J REMOTE PANEL (.-. --- m SVs TV 1 GAGE l I ARGCN L.-_ = -- I suPPLv NORMALLY. LOCALPANEL CLOSED SURGE 80M8 SV7 SEPTUM LINE FOR GAS SAMPLE O G S EXPANSION BOMB SAMPLE RETURN LINE ARROWS INDICATE THE DEENERGl2ED FLOW PATH 6 RHR/RCS HOT LEG l SV1 VCT GAS SV4 WACE gg DV1 l SV3 j [ STEAM LIOUlO SAMPLE LINE f TO LEAD SHIELD SAMPLE SINK y c2 SV2 I d ( P2R LIQUID Figure 12 Reactor Coolant Post-Accident Sampling Panel f 1 1840-0' [

~- _.~..,_ J 79 4-I es f[ d 4 l 1 i + l 1 i l 4 i CONTROL PANEL MOUNTING BOARD FOR SAMPLE SYSTEM SAMPLE CONTAINER 4 f FILTER n 1 _.J SV1 1 5, WALL l p------ y SAMPLE FROM l VENT STACK /j# l g I RE 21 & 22 i f [ RETURN TO a l VENT STACK [Sv 2 l l I l f n i i i l I I I I l l I ,I L ) ) L____--_-_J I s / PUMP i i N6 Figure 13 Vent Stack Effluent Sampling System 4 384@-@

80 X I i SAMPLE CONTAINE R SHIELD WALL FILTER ASS'Y (TO BE INSTALLED) WITH QUICK gy, DISCONNECT /s SV1 WALL EXISTING RE 11 & 12 EXISTING INLET PUMP OUTt.ET /% i OUTLET SAMPLE LINE CONTAINMENT I '3 INLET SAMPLE LINE O G Figure 14 Containment Air Sampling System

Farlcy Nuclear Plant j Unit 2 Docket No. 50-364 81 ,\\ ( II.B.4 TRAINING FOR MITIGATING CORE DAMAGE Position Complete the training of all operating personnel in the use of installed systems to monitor and control accidents in which the core may be severly damaged. This requirement shall be met before issuance of a full-power license. Respcose The training of all licensed operating personnel in the use of installed systems to monitor and control accidents in which th e core may be severly damaged will be completed prior to issuance of a full-power license. O g

Farley Nuclear Plant Unit 2 82 Docket No. 50-364 0 II.E.1.1 AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION Position Provide a simplified auxiliary feedwater system reliability analysis that uses (1) event-tree and fault-tree logic technique to determine the potential for AFWS failure following a main feedwater transient, with particular emphasis on pote.atial failures resulting from human errors, common causes, single point vulnerability, and test and maintenance outage. (2) Provide an evaluation of the AFWS using the acceptance criteria of Standard Review Plan Section 10.4.9. (3) Describe the design basis accident and transients and corresponding acceptance criteria for the AFWS. (4). Based on the analyses performed modify the AFWS, as necessary. These requirements shall be met'before issuance of a full-power license.

Response

The NRC letter of March 10, 1980 addressed the objective of the staff to evaluate Auxiliary Feedwater (AFW) system designs on a plant specific basis. The staff reactors including initiated their efforts by directing them towards operating (Mr. F. L. Clayton's Fa rl ey Uni t No.. l. Information already provided on Unit 1 lettersof December 14,1980, April 1,1980, and May 27, 1980 is directiv aoolicable to Unit 2 since the AFW systems for both Un its 1 and 2 are essentially identical. The NRC study (NUREG-0611 ) concluded that the Farley Unit 1 AFW system is highly reliable. This will meet the requirements for a AFW system reliability evaluation for Unit 2.

Farlcy Nuclear Plant Unit 2 83 Docket No. 50-364 II.E.3.1 EMERGENCY POWER FOR PRESSURIZER HEATERS Posi tion Install the capability to supply from emergency power buses a sufficient number of pressurizer heaters and associated controls to establish and maintain natural circulation in hot standby conditions. The requirement shall be met before issuance of a full-power license. ( ce NUREG-0578, Section 2.1.1, and letters of September 27 and November 9,1.9.)

Response

3 Based on the analysis performed by Westinghouse for a 3-loop PWR with a 1400-ft pressurizer, a minimum of 125 kW pressurizer heater capacity is required to establish and maintain natural circulation following a loss of offsite power. This capacity is also sufficient to maintain natural circulation with small break LOCA and subsequent isolation. Westinghouse recommends that these heaters have the capa-bility to be energized within 1 hour following the loss of power. The Farley Unit 2 pressurizer is equipped with 78 individual heater elements assembled in 5 groups. Heater groups A and B each contain 15 elements with a heater capacity of 269.25 kW per group. These heater groups have the capability of being po'.tered from the emergency section of the 600-V load centers 2A and 2C, respectively. These heater groups can be powered from either the normal or emergency bus by operation of breakers fm ( ) controlled from the control room. The pressurizer heater backup groups A and B are redundant, and their associated circuits meet the separation requirements for redundant systems. The motive and control power interfaces with the emergency buses are accomplished through safety grade devices. Procedures and training re included in the Unit 2 operator training program to make the operator aware of when and how the required pressurizer heaters are to be connected to the emergency buses. The procedures identify under what conditions other emergency loads may be shed from the buses in order to provide sufficient capacity for connection of the required heaters. l 'V

Faricy Nuc1 car Plant Unit 2 84 Docket No. 50-364 bq II.E.4. 2 CONTAlfNENT ISOLATION DEPENDABILITY Position Provide (1) containment isolation on diverse signals, such as containment pressure or ECCS actuation, (2) automatic isolation of nonessential systems (including the bases for specifying the nonessential systems), (3) no automatic reopening c f containment isolation valves when the isolation signal is reset. This requirement shall be met before issuance of a full-power license. (See NUREG-0578, Section 2.1.4, and letters of September 27 and November 9,1979.)

Response

There are two phases of containment isolation at Farley. Phase A isolates all penetrations except component cooling water, containment spray, and systems essential for safe shutdown. Phase B isolates all remaining process lines except safety injection, containment spray, service water lines to containment coolers, and auxiliary feedwater. Phase A isolation is initiated by all safety injection signals or manual initiation. Phase B isolation is initiated by containment pressure or manual initiation. A high radiation signal is also used for purge isolation. Resetting of isolation signals will not reopen isolation valves Manual action is needed to open each valve. In addition Alabama Power has reviewed containment isolation design against a generic study generated by the Westinghouse TMI Owners' Group. Farley containment isolation design meets all requirements of this study. A list of the systems identified as essential and nonessential appears below. All contaiment isolation valves have individual control switches that do not operate more than one containment isolation valve. Essential Systems 1. Nomal letdown 2. Excess letdown / seal water return 3. Pressurizer sample 4. Hot leg sample 5. Containment air sample 6. Nomal charging 7. Residual heat removal (normal suction) 8. Residual heat removal (containment sump recirculation) 9. High head safety injection

10. Low head safety injection

11. Containment spray

12. Contaiment spray (containment sump recirculation) 13.

Service water associated with containment coolers

14. Reactor coolant pump (RCP) seal water supply

15. Contaiment pressure 16.

Instrument air supply O

17. Component cooling associated.with RCP thermal barrier V

18. Post-accident air sample 19. Post-accident containment vent

Farley Nuclear Plant Unit 2 85 Docket No. 50-364 (..) II.E.4.2 Contaiment Isolation Dependability (Continued) Nonessential Systems 1. Accumulator test lines 2. Accumulator makeup 3. Accumulator sample 4. Nitrogen to accumulators 5. Nitrogen to pressurizer relief tank (PRT) 6. PRT makeup 7. Reactor coolant drain tank (RCDT) drain 8. Cohtainment 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 t

12. Demineralized water to containment

13. Service water to RCP motor air coolers

14. Containment purge and mini-purgc

15. Containment sump pump discharge

16. Containment sump pump sample recirculation

17. Charging pump / residual heat removal relief valve discharge to PRT

18. Containment leak rate test

19. Service air Containment emergency air breathing 20.

,(- '\\ ) ,.___,__.__..yy

Faricy Nuclear Plant Unit 2 86 Docket No. 50-364 II.K.3 FINAL RECOMMENDATIONS OF B&O TASK FORCE Position Any failure of a PORV or safety valve to close should be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report. This requirement shall be met before issuance of a full-power license.

Response

Alabama Power Company commits to notifying the NRC promptly upon any failure of a PORV or safety valve to close on the Steam Generators or Pressurizer. Challenges to these valves will be documented in the annual report. f ( <m ks

). Fcriey Nuclear' Plant - runit 2 - DocketLNo. 50-364 87 73O III. A'.111 UPGRADE' EMERGENCY' PREPAREDNESS . Position Provide an emergency response plan in substantial compliance with NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" (which may be modified af ter .May 13, 1980 based on _ public.canments) cxcept that only a description of and completion _ schedule for the means for providing prompt notification to the population-(App. 3), the staffing for energercies in addition to that c21 ready . required (Table B.1), and an upgraded meteorological program (App. 2) need be provided.- NRC will give substantial weight to FEMA findings on offsite plans in judging the adequacy against NUREG-0654. Perform an emergency response exercise -to test the integrated capability and a major portion of the basic elements ex sting i within energency preparedness-plans and organizations. This requirement shall. be met before issuance of a full-power license.

Response

Alabama' Power Company is currently developing a revision to the existing approved Emergency. Plan for the-Joseph M. Farley Nuclear Plant to address NUREG-0654 require-ments. However, as noted in the TMI-2 Action Plan, dated May 2,1980, this NUREG may be modified af ter May 13, 1980 based on public comment. 'q() Alabama Power Company plans to submit the revision to' the Joseph M. Farley Nuclear Plant Bnergency Plan as soon as possible after receipt of the final criteria as specified in the approved' NUREG-0654. The current schedule for submittal of this revision is July 15, 1980. \\_/ y gy,- y u-= y it <+7,y Wsw-

Farley[ Nuclear Plant

Unit 21 88

-Docket.No. 50-364- -(M )i 'III.D.l.l. _ PRIMARY COOLANT SOURCES OUTSIDE CONTAINMENT Position-Reduce [ leakage from systems outside containment that would or could contain highly, radioactive' fluids during a serious transient or accident to as-low-as-practical levels, measure actual' leak-rates and establish _a program to maintain 11eakage.at as-low-as-practical levels and monitor leak rates. 1 This requirenent shall be~ met before issuance of a full-power license. (See NUREG-0578, Section 2.1.6a, and letters of September 27 and November 9,1979.) Response-- Alabama power Company has-instituted a program to maintain leakage rates of systems outside contairinent to.as low as practical which consists of the following: A. Systems included in ~ the program: (We have reviewed plant systems and identified the following systems outside containment that could potentially contain highly radioactive fluids following at serious accident.)

1..High head safety injection system (recirculation portion only).

2. Low head safety-injection system (recirculation portion only). - 3. - Residual heat' removal system. h. 4. Reactor coolant. system letdown and makeup system. V 5. Reactor coolant sampling system. ~ ~ 6. Containnent spray system (recirculation portion only). 7. Radioactive waste gas ~ system. B. Systems excluded from the program: -(They will not preclude any option of cooling .the reactor core-nor will-they prevent the use of needed safety systems.) 11. Radioactive liquid.. waste system (excluded by NRC in regional meeting). 2.~ Radioactive waste gas system. (Portions of' system not contaminated by volume control tank off-gas processing will-be excluded. Off-gas processing would be'the means of handling highly radioactive gases resulting from various accidents.) C.- immediate 1eak reduction measures: 1. All vent and drain lines will be capped to prevent release due to seat leakage. ' 2.'. TheLpacking of all-valves (except Kerotest which is' a packless, stainless steel diaphragm valve) in the scoped liquid systems will be inspected for leakage or evidenceTof leakage such as boric acid accumulation. Maintenance will be performed on-the packing of liquid system valves identified as requiring. work. 3. The ' seals and packing on all pumps in the scoped -liquid systems will be inspected: for leakage or signs _of leakage. L4.. Valves fittings, and compressor seals in'the scoped gaseous systems will /~Y be ;" snooped'.' for leakage.. Maintenance will be performed on gas system valves A._/ - andLinstrument fittings identified during leak tests as requiring work. 1

Farley Nuclear Plant Unit 2 89 Docket No. 50-364 n t \\ / III.D.l.1 Primary Coolant Sources Outside Containment (Continued) D. Procedures for determining (measuring) leakage: During the inspections described in item C of this section, leakage rates will be recorded in drops per minute or hour. After completion of the inspection, the following additional leak rate tests will be performed: 1. High' head safety injection system - integrated leak rate test. 2. Low head safety injection system - integrated leak rate test. 3. Residual heat removal system - Integrated leak rate test. 4. Reactor coolant system letdown and makeup system - integrated leak rate test. 5. Reactor coolant sampling system - integrated leak rate test. 6. Containment spray system - integrated leak rate test. 7. Radioactive waste gas system - bubble (" snoop") test of individual valves, fittings, and seals. Quantitative value obtained from bubble rate. E. Leakage measurement results: Actual leakage rates will be determined during plant startup testing prior to initial criticality and reported to the NRC. F. Continuing leak reduction: g l l U A general maintenance procedure has been written and approved incorporating the inspections discussed under item C and the leak tests discussed under item D into the Farley Nuc7 ear plant preventative maintenance program. This program is under the supervision of the maintenance supervisor and includes removal of boric acid residue on components in orderrto facilitate leak detection during subsequent inspections. Leak rate measurements will be perfomed periodically at intervals not to exceed each refueling outage. Records of leakage rates will be retained in the plant maintenance files and will be retrievable using computer indexing. G. IE Circular 79-21: All relief lines coming off the scoped systems will be walked down against P & ID's as required by IE Circular 79-21. ,a

Farley Nuclear Plant Unit 2 90 Docket tb. 50-364 III.D.3.4 CONTROL ROOM HABITABILITY Position Identify and evaluate potential hazards in the vicinity of the site as described in SRP Sections 2.2.1, 2.2.2, and 2.2.3, confirm that operators in the control room are adequately protected from these hazards and the release of radioactive gases as described in SRP Section 6.4, and, if necessary, provide the schedule for modifications to achieve compliance with SRP Section 6.4. This requirement shall be met by issuance of a full-power license.

Response

The following is a description of the existing control room habitability systems. The control room habitability systems are designed to provide maximum safety and comfort for operating personnel during nonnal operations and during postulated accident conditions. These habitability systems G < the control room include shielding, charcoal filter systems, heating, ventiiation anc' air conditioning, storage capacity of food and water, kitchen, sanitary facilities and fire protection. The control room habitability systems are designed to meet AEC General Design n Criterion 19. Sufficient shielding and ventilation are provided to permit 1 occupancy of the control room for a period of 30 days following a design basis accident (DBA) without receiving more than 5 rem whole body dose or its equivalent to any part of the body. Design Bases - The following design bases were used to determine the functional design of the habitability system: 1. The postulated accident that determines the habitability design requirements is the Design Basis LOCA. 2. The assumptions regarding the sources and amounts of radioactivity that could pose a hazard to the control room are discussed in Section 12.1.3 of the FSAR. 3. In the event of an accident, the ventilation system in tit. control room will be triggered by the "T" signal which will automatically isohte the normal air systems and start both trains of the control room A/C system, pressurization system and filtration system. The control room air will thus be filtered through a HEPA charcoal filter system which is capable of removing 99.9 percent of both the inorganic and organic iodine. On a long-term basis (eight-day continuous flow) the filter efficiency for both organic and inorganic iodine is 99.0 percent. The filter system and the control room shielding are capable of keeping the [,j dose to the operators less than 5 rem whole body dose or its equivalent to any part of the body for the duration of the accident.

Farley Nuclear Plant Unit 2 Docket No. 50-364 91 ,~) !N-III.D.3.4 Control Room Habitability (Continued) 4. Following postulated accidents, the most severe being a LOCA, the limitations on control room pressure, tanperature, radioactivity concentrations, doses, and concentrations of carbon dioxide are as follows: Parameter Allowable Control Room pressure >l/3" w.g. Control Room Temperature 51200F Radioactivity Concentration As stated in 10CFR20, Appendix B Doses 5 rem Concentration of carbon dioxide 1 wt percent 5. The' fire protection system in the control room consists of an early warning ionization-type detection system with hand portable water extinguishers located in the control room itself. The ionization detectors used will detect products of combustion in approximately four seconds. Detectors are located on the false ceiling to detect smoke in the control room itself. Each detector is equipped with a light to indicate which detector has operated. There is a subpanel that indicates which detector has operated in spaces 'not readily visible. All detectors will operate the visible and audible alarm on the main fire protection annunciator panel located in the control room. In addition, a fixed carbon dioxide hose reel with 100 feet ("' of hose is located in the innediate vicinity outside the control room and can be ( j} used to back up the hand extinguishers. Non-combustible materials are used in construction and equipment as much as . pos sibl e. The quantity of combustible material such as paper and other flammable supplies are kept to a minimum. A person trained in fire fighting will be on duty in the control room at all times. Control Room HVAC - Design Bases t The control room air conditioning and filtration system is designed with sufficient redundancy and separation of components to provide reliable operation under normal conditions and to ensure operation under emergency conditions. The system is designed to provide an environment with controlled temperature and humidity to ensure both the comfort and safety of the operators and the integrity of the control room. components. Design ambient conditions are approximately 74 F and 50 percent relative humidity, taken from ASHRAE Standard 55-66, " Thermal Comfort Conditions." Provisions are made in the system to detect and limit the introduction of airborne radioactive material into the control room. Provisions are also made in the system for the removal of radioactive and foreign material from the control room environment. () The system is designed to permit periodic inspection of the principal system components.

Farley: Nuclear Plant . Unit 2 . Docket No.~50-364 92 \\ I II.D.3.4'- Control Room Habitability (Continued) Control Room' HVAC System Description - During normal plant operation,-one (1) of the two (2) 100-percent capacity, Category I air conditioning packaged units recirculates 21,000 cfm of cooled filtered air through the control room. Each packaged unit consists of a recirculation-fan, prefilter, refrigeration unit with a water-cooled condenser, cooling coil, and the associated instrumentation and controls. Should the operating unit faillto function in some way, the second 100-percent capacity unit can be manually started by the operator. Two full-capacity, redundant, seismic Category I air pressurization systems are provided to maintain the control room at a positive pressure post-LOCA. Each train is capable of supplying 300 cfm through electric heating coils, prefilter, HEPA filter, and 6-inch deep bed charcoal filters, designed in accordance with the requirements of Regulatory Guide 1.52. Installed in parallel to the suction side of each control room main air conditioning unit are 1000 cfm filtration units consisting of prefilters, HEPA filters, and 2-inch charcoal filters, and a 2000 cfm filtration unit incorporating the same composite filter elements as the 1000 cfm units. There-fore, the overall recirculation filtration capacity is 3000 cfm. During'. normal operation, an air supply system delivers fresh outside air to the control room and to the computer room. The air supplied to the control room by this ~ - ("] system maintains the control room at a slight positive pressure,. thereby preventing G the introduction of air into the control room from sources other than the design fresh air makeup system. A smoke detector near the return-air duct to each recirculation fan will sound an alarm in the control room on high smoke level. If necessary, the operator can ex-E huast air from the control room by manually opening the motor-operated exhaust damper and starting one (1) of the two (2) 100-percent-capacity exhaust fans. An area l radiation monitoring system and: redundant control room charcoal filter recirculation systems are provided to detect and reduce radiation levels in the control room. The filters are composite units and have been furnished to the same specifications as the filters for the penetration room fil. stion system. An area radiation monitor located in the control room alarms on nigh radiation level and alerts the operator to the need for filtration of recirculated air. Redundant Category I chlorine detectors and process radiation monitors are provided at the control room normal fresh air intake. Redundant Category I smoke detectors are provided in the fresh air intake and return air duct of the computer room air l handling unit. The air-operated isolation damper on the computer room recirculation line is interlocked Wh the computer room fire detectors, and will automatically t_ close in _ the event of smoke detection. Closing of this damper will also initiate closure'of the. computer _ room outside air supply line isolation valve and shutdown of the computer-room air conditioning system. The computer room fire detectors are also the means for actuating the halon release into this area. The halon storage tank is located outside the computer room. Inadvertent release of halon is discussed o. below. Tripping any one of the detectors will cause an alarm to sound in the control ( room and the closing of all air-operated and motor-operated isolation valves in the non-engineered safety: features HVAC penetrating the-control rocm boundary, thereby

Farley Nuclear Plant Unit 2 93-Docket No. 50-364 73-3 III.D.3.4 Control Room Habitability (Continued) isolating the control room. If required, and if contaminants are within safe levels, the operator can draw outside air by manually starting the emergency pressurization system.-- in which the air is filtered through deep bed filters. During post-LOCA conditions, the normal makeup air is cut off and all control room isolation valves are closed. In this event, the positive pressure in the control room is _ maintained by the automatic startup of one of the emergency pressurization systems, each consisting of an air inlet, an isolation valve, and a 300-cfm, deep-bed, charcoal filtration unit. In like manner,- the standby filtration units will automatically: recirculate 3000 cfm out of the 21,000 cfm total room recirculation flow rate through charcoal filters. All penetrations into the control room are sealed to eliminate inleakage of outside air. Mechanical penetrations are sealed by the use of silicone-rubber foam or by fiberglass impregnat<d boots. Electrical penetrations are sealed with silicone-rubber foam. HVAC sytten duct penetrations are provided with air-tight automatic butterfly valves. Control Room HVAC --Safety Evaluation The redundant system has been designed to provide minimum filtering and ventilation, g) and ensures that no single failure will prevent the safe occupancy of the control 3V room under any mode of plant operation. Power for the fans and refrigerant compressors of each air conditioning packaged unit is supplied from redundant emergency power supplies. A separate, independent system of distribution ducts is installed for each unit. Cooling water for the condensers of the air conditioning packaged units is supplied from redundant headers in the service water system. The duct from the computer room air conditioning unit that normally supplies fresh makeup air to the control room contains two pneumatic-operated valves in series. A single pneumatic-operated isolation valve is provided for each duct connected to the smoke purge exhaust system. Two pneumatic-operated valves in series are provided for the utility exhaust subsystem. These valves are powered from redundant power supplies and close automatically on a containment isolation signal to isolate the control room from outside air. The control room habitability is maintained by continually monitoring radiation levels and smoke concentration inside the room, plus contirPlly monitoring radia-tion levels, including monitoring of smoke concentration in ~hc control room air intake duct and computer room return duct, and chlorine cont.entration at the i control room air intake. To prevent inleakage, the ca trol room is provided with normal and emergency pressurization systems designed to maintain greater than 1/8-inch w.g. positive pressure. /', Upon toxic gas -(chlorine or smoke) detection, an alarm is annunciated in the control ! ) room and all_ failsafe, airtight isolation valves are closed automatically and remain closed until they are reopened manually. The normal makeup air is cut off. In the case of smoke detection, the operator can manually star +he exhaust fan to purge smoke; in the case of chlorine detection, the operator can make use of self- . contained breathing.appara tus. Af ter a safe level of toxic gas concentration is

Faricy Nuclear Plant. 94 Unit 2 Docket No. 50-364 'V III.D.3.4 Control' Room Habitability (Continued) reached, the operator can draw outside air either from the normal makeup air sub-system or from the. emergency (pressurization system) makeup air subsystem. Upon a high radiation signal.from the makeup' air inlet an alarm is sounded in the control room and all. isolation valves are closed automatically and remain closed until they are reopened manually. This will result in a loss of positive pressure in the control room. In this event, the operator will manually start one of the redundant pressurization systems and one of the redundant recirculation filtration systems (one 2000 cfm and one 1000 cfm system). The HEPA and 6-inch deep bed charcoal filter unit in the pressurization system and the HEPA and 2-inch charcoal filters in the recirculation system provide additional assurance that the dose received by control room personnel will -not exceed the guidelines of Criterion 19. Upon receipt of a containment isolation signal, the control room is automatically isolated as described above. The air pressurization system and recirculation system are automatically actuated to maintain the positive pressure and to provide control. room cleanup, respectively. A flow control, automatic damper mounted on the bleed-off leg of the pressurization system will respond automatically to the pressure controller.ir the control room. If the positive pressure drops below 0.25 in, w.g., the bleed-of f damper will automatically restrict the bleed-off flow to divert maximum air supply into the room, in order to achieve a rapid pressure .Q-buildup. On a rising room pressure, the bleed-off damper will function in .Q the opposite manner. Therefore, the control room pressure will be sensed and maintain-ed, as well as exhibited for operator information. Upon receipt of a smoke detection signal from the computer room smoke detector, the computer room HVAC is automatically isolated. In addition, re-dundant seismic Category I smoke detectors downstream of the return air subsystem from the computer room will automatically isolate redundant seismic Category I isolation valves in the computer room recirculation line, in the event of smoke recirculation following a ' computer room fire. Radiation monitors are provided within the control room boundary. Radiation monitors are also provided within each of the various ventilation systems serving all radiation release points in the plant. These monitors provide indication in the control room, and alarm whenever predetermined radiation levels are exceeded. These HVAC systems discharge through the plant vent stack. Additional radiation. monitors are provided at the vent stack discharge which will provide a back-up means of detecting abnormal plant releases. These monitors are designed to detect releases in excess of the MPC Guidelines established under 10 CFR 20, -Appendix B. Based on the availability.and sensitivity of the monitoring system provided, the operator will have adequate indication and information'to evaluate the magnitude of any abnonnal plant releases, and will manually isolate the control room if required. An analysis of dose levels in the control room under accident conditions is pre-sented in the applicable _ sections of FSAR Chapter 15.0. 4 Redundant chlorine detectors will be provided in the circulating water chlorination house designed to transmit an alarm in the control room if the chlorine spill should These alarms will. provide ' sufficient time for operators to put on self-occur. contajned breathing apparatus. a

.Farley Nuclear Plant Unit 2 95 Docket No. 50-364 / Control Room Habitability (Continued) III.D.3.4 An analysis of a chlorine release accident has been performed. Because of the proximity of the closest circulating water chlorination house the analysis was performed for the release of two tcns of chlorine (the maximum amount of chlorine headered together at one time). Twenty-five percent of the chlorine was assumed to flash to gas. This is analyzed as a puff release. The remainder is assumed to form a 200 square foot pool where it evaporates due to the heat load from the sun and from the ambient air and ground temperature. No credit is taken for the channeling of the dense chlorine gas around buildings and along ditches. No credit is taken for an elevated air intake, even though the intake is at an elevation of approximately 177 ft. To evaluate the control room habitability, considering the closest circul-ating chlorination house, the following cases have been analy;ed using the techniques outlined in References 1, 2, and 3 to Section 9.1.4.5: A. Two-ton chlorine spDl, 450 feet from control room air intake, 0.5 m/sec wind, Pasquill Class F meteorology shared Unit 1 and Unit 2 control room, 70 cfm unfiltered inleakage and no credit for building wake effect. B. The same as Case A except a wind speed of 1.0 m/sec it used. 7 C. The same as Case A except building wake effect is considered. D. The same as Case B except building wake effect is considered. The time (af ter chlorine accident) required to reach 15, 30, 45, 60 and maximum chlorine concentration in ppm by volume inside the control room is given in FSAR Table 9.4-0. This table also sh',ws the peak chlorine concentration. The analysis assumes a 5-second detcctor response time from 5 ppm setpoint (chlorine detector at air intake), a 3-second transport time of chlorinated air from air intake to the isolation valve, and a 6-second closing time of the isolation valve. In Case B, the operators will have over 2 minutes to put on self-contained breathing apparatus before a concentration of 15 ppm is reached in the control room, after allowing 5 seconds detection time at 5 ppm for the chlorine detectors at the chlorination house. This is in accordance with Regulatory Guide 1.78. An inleakage rate of 70 cfm is used based on 0.06 air exchange per hour for Type A control room defined in Regulatory Guide 1.78. Self-contained breathing apparatus with a minimum 30 minutes air supply are stored in the control room. Sufficient apparatus is provided to allow one spare per three units required. An additional six hour air supply is provided in the auxiliary building as close as practical to the control room. An offsite source of air will also be available. An analysis of the Halon 1301 concentration in the control room following an accidental spill in the computer room, without fire, has been performed. It is assumed that the total amount of Halon 1301 flooding the computer room mixes with e the air in the computer room. This yields a maximum initial concentration of 6 percent by volume in the computer room. Using the maximum control room makeup air quantity of 1650 cfm, the maximum concentration of Halon 1301 in the control room is.47 percent by volume after 15 minutes. This is well below the 7 percent maximum recommended concentration for normally occupied areas of NFPA Standard 12A.

Faricy Nuclear Plant Unit 2 96 Docket No. 50-364 III.D.3.4 Control Room Habitability (Continued) Control Room HVAC - Instrumentation One train of the contrcl room air conditioning system is operating during normal conditions and both trains are automatically started during post-LOCA conditions. Both trains of the control room recirculation filtration and emergency pressurization systems are automatically started upon receipt of a CIAS signal. The following are displayed and/or located in the control room: a. Control room air temperature b. Air intake chlorine concentration indication c. Air intake smoke concentration indication d. Air intake radiation level e. Room smoke detectors f. Alarm for high differential pressure across each filter train q g. Differential pressure between the control room and atmosphere i Position indication of all isolation valves and dampers is locally displayed. Differential pressure across each filter train and fan is locally displayed in the control room mechanical equipment room. Control Room Shielding - Control room shielding design is based on the requirements set forth in 10 CFR 50, Appendix A, Criterion 19, which requires occupancy of and access to the control room under accident conditions. The dose to personnel will be limited to 5 rem whole j body or its equivalent to any part of the body for the duration of the accident. The accident ar,alysis in FSAR Chapter 15 indicates that this dose to personnel will be less than 5 rem. Protection of control room personnel from the fission-product release in the containment is provided by the concrete walls between them. Storage Capacity of Food and Water - There will be sufficient storage capacity (including a food freezer) for two shifts of operators for 30 days. Water is drawn from the Pctable and Sanitary Water System. 7

Farley Nuclear Plant Unit =2 97 Docket No. 50-364 O y III.D.3.4 Control Room Habitability (Continued) Kitchen - Kitchen facilities are available for refrigeration, cooking, and the cleaning of cooking and eating utensils.

Sanitary Facilities -

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