Text

Advises Commission of Intention to Grant Request by Toledo Edison Co & Sacramento Municipal Util Dist for Extension of Time for Response to Order for Mod of Licenses Re Instrumentation for Detection of Inadequate Core Cooling
	ML20209C028
Person / Time
Site:	Davis Besse, Rancho Seco, 05000000
Issue date:	02/28/1983
From:	Dircks W; NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	;
Shared Package
ML20134H339	List: ML20209C028;
References
	TASK-PINC, TASK-SE SECY-83-084, SECY-83-84, NUDOCS 8303110439
	Download: ML20209C028 (4)
	v • d • e

. . , -

ENCLOSURE 2

~

[3-3 '2 eS4 9 SRG February 28, 1983 *****

SECY-83-84

_ POLICY ISSUE (NEGATIVE CONSENT)

For: The Comissioners From: William J. Dircks, Executive Director for Operations

Subject:

REQUEST FOR EXTENSION OF TIME FOR RESPONSE TO ORDER FOR ,

MODIFICATION OF LICENSE REGARDING INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING - TOLEDO EDISON COMPANY AND SACRAMENTO MUNICIPAL UTILITY DISTRICT

Purpose:

To advise the Comissioners of the Staff's intention to grant a request by Toledo Edison and Sacramento Municipal Utility District (SMUD) for an extension of time for response to the subject Order.

Issue: By letters dated January 17,1983, Counsel for Toledo Edison Company and Sacramento Municipal Utility District responded to our December 10, 1982 Orders requiring the installation of an inadequate core cooling instrumentation system at the Davis-Besse. Unit 1 facility and Rancho Seco.

Counsel for the licensees stated that both plants already include subcooling margin monitors and core-exit thermocouples.

The licensees intend to install a reactor coolant inventory tracking system to provide inventory tracking with reactor coolant pumps on and off. However, their proposed system will not include a reactor vessel head monitor.

Both utilities requested an extension of time to respond to the NRC Order in order to permit them to complete some analyses designed to show that when certain plant modifi-cations are made a vessel head monitor is not needed. The licensees also made contingent requests for hearings if the extension of time is not granted.

Contact:

L. E. Phillips, DSI:CPB X-29472 36k(DD /

VM M &fM M M M M M M M M M M M MA . . .

.. s, '. '

f Comissioners

Background:

The staff has concluded that voiding in a B&W PWR system could appear initially in either the candy cane of the raised hot leg loop or the head of the reactor vessel, depending on the location of the leak and the nature of the transient. The B&W reactors are unique in that the raised

" candy-cane" contains a significant inventory of primary coolant which must drain back into the reactor before the core becomes uncovered. Voids forming within the circu-lating system would tend to accumulate in the top of the candy cane where detection would be an early indication of inventory loss and possible approach to inadequate core

! cooling (ICC). However, overcooling transients such as the l Ginna SGTR event would result in cooling of the circulating i system while the hottest coolant remains in the upper head, i leading to initial voiding in the upper head. An upper i head steam bubble under these conditions (or with an upper head leak) would grow until it reaches the hot leg elevation i

before it could migrate to the " candy-cane." Early detection

of the problem under these conditions would require upper i head instrumentation in addition to the hot leg monitor.

It was on this basis that the staff concluded that both a l " candy-cane" monitor and an upper head monitor are needed for inventory tracking in a B&W system. This recomendation

, was provided in SECY-82-407, which has been approved by the Comission.

j Discussion: In response to a request by the NRC staff following receipt of the extension request, representatives of Toledo Edison Company and SMUD met with the NRC staff in Bethesda to l

discuss the issue. Major points of clarification are:

(1) The utilities intend to submit their conceptual

designs for a hot leg level monitor and a reactor 2

coolant pump monitor to provide inventory tracking i with pumps off and on, respectively, on the schedule i

required by the Order.

(2) The utilities described a cost / benefit analysis in progress to evaluate the merits of including reactor vessel head instrumentation. They described problems involving accessibility, routing, and refueling considerations relating to installation of a reactor vessel head monitor. Preliminary estimates on instal-

. lation costs and personnel exposure were presented, but were well within the bounds of estimates used by the staff when the requirement was justified.

1 i

. . ~ . . . . . - . . . . . . - - . - . - .

Comissioners .

(3) The response of the required instrument system during an approach to ICC was discussed. The licensees pointed out that the large coolant inventory in the hot leg reduced the importance of a vessel head

monitor to indicate the approach to ICC. While this

is an important design feature, the information i presented had been considered by the staff when the requirement was established.

(4) An alternate design concept consisting of a vent line from the reactor vessel head to the top of the hot leg was discussed. An analysis is being performed to .

ediuate the design and perfomance feasibility of this concept in lieu of a reactor vessel head measure-ment dp system.

The licensees requested a short delay (until April 15) to complete their analysis and finalize their conceptual design.*

Th'e Order for Modification of License requires that the licensees complete their conceptual design review for a

- reactor coolant inventory tracking system, identify the

. design selected, and submit detailed schedules for engi-neering, procurement, and installation within 90 days. The licensees' evaluation of an alternate design concept is needed.to complete that portion of their conceptual design review involving monitoring of the coolant in the vessel

~

upper head. The licensee propeses to address that aspect of its design and to provide other specified information relating to design evaluation and operator interaction in the deferred submittal (by April 15). The staff has determined that only that portion of the deferred submittal which describes the vessel head monitoring concept fails to meet the intent of the Order provided that other portions of the submittal are identified and included in the de-talled schedule for engineering and documentation of the 3

inventory tracking system, which is to be provided within the 90 day period.

Although the Order requires certain actions "within 90 days of the date of this Order", one reading of the Order is that submittals are not due until April 16 because the Order by its tems is not effective until January 16, 1983. Under such a reading, Commission approval of the licensee's requests would not be necessary. However, in view of the underlying nature of the requests, consultation with the Comission was deemed appropriate.

\.:u . .. ,. . . . . ~ . . . . . - . . . . - - . . _ ~ . . . . . . . . . . . . . . . . . . . _ , . .

Comissioners The staff believes that the alternate design concept under consideration by the licensees, consisting of a vent line from the reactor vessel to the top of the hot leg, has potential merit and may satisfy the Comission approved design requirement to monitor the coolant inventory over the range from the vessel upper head to the bottom of the hot leg as a minimum.

Recomendation: That the Comission approve transmittal of letters to the licensees (Enclosures) which approve the delay in their partial response to the Order. The expected.

content of the "within 90 day" and April 15 submittals 4

is described in attachments to the enclosed letters.

\

Scheduling: In view of the nature of the request, and the April 15 submission date proposed by the licensees, it is

desirable to respond as soon as possible. Unless we

~

are advised.to the contrary, the staff intends to send

- the letters five days from the date of this paper.

Will m J. Dircks, Executive Director for Operations

Enclosures:

Ltrs to Toledo Edison Coe & SMUD SECY NOTE: In the absence of instructions to the contrary, SECY will notify the staff on. Tuesday, March 8, 1983 that the Commission, by negative consent, assents to the action proposed in this paper. .

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[ l owiss or~e w FINAL SAFETY MRRTURE - ~ =nc=l= .,oo ,,, o, rowumrcoao orc,acarin nan ANALYSIS REPORT

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! MM n so83u'of. .so ,3r outer or RADWASTE EVAPORATOR BUILDING g ;

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I BFNP-3 12.3 SHIELDING AND RADIATION PROTECTION 12.3.1 Desian Basis Plant shielding is designed to provide for personnel access to the plant to. perform maintenance and carry out operational duties i with personnel exposures limited to the dose criteria set forth

' in the Code of Federal Regulations Title 10 Part 20 (10CFR20) and g the Radiation Protection Plan, as appropriate. The shielding design criteria for the Reactor, Turbine, and Radwaste Buildings and for the offgas stack for both normal and shutdown conditions are established considering the total activity in the core, coolant, liquid waste, and offgas systems. The shielding and radiation protection design criteria al so consider the radiation conditions in the control room following a design basis accident.

Within these criteria the plant is shielded to provide work areas for operation and maintenance of the plant and for the control of the plant during the design basis accident.

4 Full power operation design conditions a s s um e that the core is operated at the design power of 3440 MW t . At this power level the **N coolant a c t iv i ty leaving the reactor vessel is 1,47 x 108 microcuries per second per unit (see subsection 9.4,

'" Gaseous Radwaste Sy s t em") . The offgas system shielding is l

designed for a. stack release rate of 370,000 microcuries per s second per unit after a 30-minut e holdup (see Appendix E).

Reactor water fission product concentrations and activated corrosion product concentrations are assumed to have maximum values of 7.0 pCi/cc and 0.07 pCi/cc, respectively. The latter condition sets the maximum shielding requirements in the j condensate sy s t em f il t e r domineralizers sad reactor water cleanup system.

12.3.2 Descrintion The design basis for shielding work areas is based on the l-expected radia tion level s and estimated occupancy times. The i plant i s divided into zones dependent upon the intensity of radiation within the given area. Zone classifications are listed I in Table 12.3-1. The entrances into all zones are marked in accordance with the regulations of 10CFR20. Regulated zones are a

also appropriately identified at the points of access. Barriers restrict entry to controlled zones. In addition, radiation work I 7

l permits and health physics surveillance may be ne ce s sary be f ore entry.

i Entrances to prohibited zones are barred by locked or guarded 1 3 doors. Radia tion work permit s and health physics surveillance I may be required t o enter prohibited zones. The entrances to the various zones in the buildings are shown on Figures 1.6-1 through 1.6-7 and Figures 1.6-11 through 1.6-14.

j O 12.3-1 i

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BFNP-1 O

Exposure of personnel to concentrations of radioactive materials in air or w a ter is limited to the v alue s in Appendix B of 10 CFR

20. All ow a n c e of protection factors for respirator equipment as presented in 10 CFR 20.103 (c) will be used for limiting esposure of individuals to concentrations of airborne radioactive material in restricted areas.

The radiation monitoring sy st em s are described in subsectiona 7.12, ' Process Radiation Monitoring'; 7.13, ' Area Radiation Monitoring System *; 7.14, ' Site Environs Radiation Monitors's and 7.15, ' ll c a l t h Physics and Labora t ory Analysis Radiation Monitors.'

The design basis accident, which is the loss of coolant, de f ine s the protection required for the plant Main Control Room and is descrfbed in subsection 14.6, ' Analysis of Design Basis Accidents.' The Main Control Room is located on the top floor of the control 1ay. The entire control bay is shielded from se conda ry containment by concre t e w all s, roof, and floor. The roof is 27 inches thick, the wall separating the control bay and se con d a ry containment is 30 inches thick, the wall between the Turbine Building and the control bay is 18 inches thick, the floor over the steam tunnel is 54 inches thick, and the remainder of the floor is 30 inches thick. Penetrations from the secondary containment enter the control bay on the lower tw o floor s which are separated from the main control by 8-inch-thick concrete floors.

The control bay shielding was a nalyred using the maximum normal sources and wa s calculated to be less than 0.5 mr em/ h r.

The control bay shielding was analyred for the design basis accident using the source s treng th s listed in Appendix F of the Units 1 and 2 Design and Analysis Report. The primary containment contribution to the control bay accident dose was found to be less than 2 mrem / hr. The dose from the secondary containment activity was found to be less than 500 mrem in any continuous 8-hour period.

During accident conditions the exit route from the control room and pl a n t is thet of lesst possible exposure and is determined at the time of the accident, j

The Reactor Building contains five maj or shielding structures--

the reactor sacrificial shield, the d ryw e ll biological shield, the main steampipe tunnel, the fuel pool, and the cleanup domineralizer system equipment rooms. The concrete d ryw ell biological shield with the sacrificial shielding provides the main protection for the ares surrounding the reactor vessel, the primary coolant piping, and the recirculation system. More than ,

8 feet of concrete is used to reduce the radiation levels in 12.3-2

BFNP CONDUCT OF OPERATIONS LIST OF TABLES I

Title l Ighle l 13.5-1 Control Rod Drive System Tests (Unit 1) 13.5-2 Maj or Plant Transients 13.5-3 Stability Tests 13.5-4 Startup Test Program (Unit 1) 13.5-5 Startup Test Program (Unit 2) 13.5-6 Startup Test Program (Unit 3)

CONDUCT OF OPERATIONS LIST OF ILLUSTRATIONS Flaure Title 13.2-1 Organization of the Tennessee Valley Authority 13.2-2 Office of Nuclear Power Organization Chart 13.2-3 Division of Nuclear Services Organization Chart 3 13.2-4 Site Director Organization for Browns Ferry Nuclear Plant 13.2-5 Plant Manager Organization for Browns Ferry Nuclear Plant 13.4-1 Typical Preoperational Test Sequence 13.5-1 Probable Startup Test Sequence - Unit 1 13.5-2 Probable Startup Test Sequence - Units 2 and 3 13.6-1 Plant Procedures O

O 13.0-111

BFNP-3 13.0 CONDUCT OF OPERATIONS O 13.1 Summary Descristion

(

' 13.1.1 General The Browns Ferry Nuclear Plant is designed, constructed, and operated to produce electric power reliably and economically, and ;

with safety to the public and plant personnel.

13.1.2 Ormanization Relationshins r

General Electric (GE) supplied the Te nne s see Valley Authority ;

(TVA) with, fully licensable and operable nuclear steam supply J

systems. - The TVA Office of Engineering served as the plant architect engineer and f ul filled the technical requirements of '

the nuclear steam supply system contracts. The TVA Office of 3 Construction was responsible for constructing the plant in accordance with design specifications supplied by the Office of g !

Engineering.

General Electric is also conmitted to supply fully licensable and operable nuclear fuel for the initial core loads and several reload batches.

The TVA Office of Nuclear Power is responsible for the safe i operation and maintenance of the plant in compliance with the O operating licenses, technical specifications, and other requirements, i Plant organization and responsibilities f or normal plant ;

operation are described in paragraph 13.2.1. Plant organization t and responsibilities during preoperational testing, startup, and initial operation are described in paragraph 13.2.2. All operations are performed by assigned plant personnel of the 3 Office of Nuclear Power or other TVA divisions. General Electric t i

provided technical direction and guidance during preoperational and startup testing, and during initial operation until all plant equipment was fully accepted by TVA. GE continues to supply <

technical direction and guidance as necessary to support continued operation of Browns Ferry. Technical direction and

guidance are defined in paragraph 13.2.2. ,

l General Electric supplied TVA with preoperational and startup tett procedures for the nuclear steam supply system. The Office I l

! of Engineering provide d t e s t-s copin g documents and acceptance .

criteria for the balance of plant sy st em s . The Office of g l-Construction was responsible for preoperational testing including

preparation of a detailed set of test procedures. The Office of l

Nuclear Powe r assume s re sponsibility for plant startup testing and operation commencing with fuel loading and prepares detailed

' startup test procedures and operation manuals.

i l

' 13.1-1

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BFNP-3 O

The Browns Ferry Nuclear Plant is staffed consistent with TVA's policy for tristing steam plants, but taking into consideration the additional nuclear talent which is necessary for a nuclear station. Support in the areas of operation, maintenance, and engineering is provided by the Office of Nuclear Power's central 3 office staff. Consultation in other areas such as design impr ov em e nt s , radiological safety, and reactor physics is av a il a bl e from other TVA divisions.

13.1.3 operation Normal and abnormal operations are carried out according to standard practices and written operating procedures which conform to the opera tional quality assurance plan. Normal operating procedures for planned operations are prepared for sy s t em s and the integrated plant. Emergency operating procedures are written for the general categories of accidents or abnormal occurrences which could lead to the i nj ury of plant personnel or the public.

TV A ' s Emergency Plans contain the precautionary planning, authority, and responsibility, and delegation plans of action to protect the public, plant employees, and equipment in case of ususual incidents.

Pl ant maintenance also adheres to standard practices and detailed procedures which conform to the operational quality assurance plan, where necessary.

Refueling operations are carried out according to detailed procedures to ensure a safe and orderly refueling. Each fuel loading or unloading step is done utilizing strict administrative and procedural control to avoid inadvertent criticality and hazardous conditions. Fuel inventory and control measures are used in order that all special nuclear material is accounted for at all times during refueling a nd d ay- t o-d ay o p e ra t i ou s .

Records reflecting pl ant operation, maintenance, tests, and inspections are maintained to support plant operations and to show compliance with the plant licenses.

O 13.1-2

. __ -~ . .. - ._ _ . - _ _ . _ . . _ _ . _ . . _ _ . . . _ _ _ . . . _ . _ _ . _ _

k BFNP-3

(' 13.2 Ormanization and Resoonsibility 4

13.2.1 Plant Oooration, Ormanization. and Responsibility l 13.2.1.1 General TVA'is's corporate agency of the Federal Government whose m aj or policies, programs, and organizations are determined by a full-I time, th ree-member Boa rd of Directors. Members of the Board are appointed by th e President and confirmed by th e Senate for 9-year i- terms. The general organiza tion of the Tennessee Valley Authority is shown in Figure 13.2-1.

The Office of Nuclear Power within the Office of Power is l3 responsible for operating and maintaining TVA's nuclear power The organiza tions of the Office of Power and the plants.

Office of Nuclear Power is shown in Figures 13.2-2. l0

13.2.1.2 Plant Ormanization The Browns Ferry Nuclear Plant organization charts is shown in Figure 13.2-3. The site director functionally directs the Site Service s G roup, the Design Services Group, the Quality Assurance Group, the Modifica tions Group, the Plant Manager, Personnel, and j Financial Planning. The principal groups that normally function

directly under the s up e rv i si o n of the plant manager are the j Opera tions a nd Engine ering Group and Maintenance Groups. Staff

' services are provided by the Industrial Safety Staff, the Plant

+ \ Compliance Staff, and the Planning and Scheduling Section. The Browns Ferry Plant organization follows the pattern developed through experience and is in use at all TV A nu cl e a r pl ant s. The

! r e s po n sibil i t i e s and personnel qualifica tion r equirement s of each 3 of the se groups are generally described in the following paragraphs.

2 13.2.1.3 Site Director i

! The site director is responsible for the overall management of l

the Browns Ferry nuclear plant planning and directing the operations, maintenance, engineering, repair, modification, refueling, and administration at the plant. He plans, coordina te s and directs overall activities through the plant f

manager and supporting staffs, i

The site director shall have had six years of experience in the l

operation of steam generator stations, three years of which shall

. be nuclear experience. He should have a ba chel or's degree or higher or the eq uiv al ent in an engine ering or scientific field generally associated with power production.

i i

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( 13.2-1 i

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BFNP-3 13.2.1.4 Plant Manager The plant manager is responsible for the mana gement of the Browns Ferry Nuclear Plant. He is responsible for safeguarding the general public and station personnel from radiation exposure, and for adherence to all requirements of the operating license s g and technical spe cifica tions. The plant manager shall h av e 10 years of responsible power plant experience, of which a minimum of 3 years shall be nu cl ea r power plant experience. A maximum of 4 years of the remaining 7 years of experience m ay be f ul filled by acadenic t r ai ni ng on a one- f or- o ne time basis. This a cadem ic training shall be in an engineering or scientific field generally associated with power production. The plant manager shall have acquired the experience and training normally required for e xamina tion by the NRC for a Senior Reactor Operat or's License whether or not the examination is taken.

If one of the plant superintendents meets the nuclear power plant experience and NRC examina tion requirements e stablished for the plant manager, the requirements of the plant manager may be g reduced, so that only 1 of his 10 years of experience need be nuclear power plant experience and eligibility f or NRC e xam ina ti on is not needed.

The plant manager, or a plant superintendent should have a 3 recognized baccalaureate or higher degree or the eq uiv al ent in an engineering or scientific field generally as socia ted with power production.

Plant Superintendents The plant superintendents assist the plant manager in planning, 3 coordinating, and directing the plant activities.

A plant superintendent shall have a minimum of 8 years responsible power plant experience of which a minimum of 3 years shall be nuclear plant experience. A maximum of 4 years of the remaining 5 years of the power pl ant experience m ay be f ul fill ed by satisfactorily completing academic or related t e ch ni c al training on a o ne- f or- o ne time basis. A degree in engine ering or the equivalent is de si rabl e. A plant superintendent, the plant b manager, or the assistant plant m ana ge r shall be capable of f ul filling the requirements of a Senior Reactor Operator License whether or not the e xamina tion is taken. If the plant manager has the required 3 years nucl ea r plaat experience, the req uirement s of the plant superintent m ay be reduced so that only 1 of the 8 years of experience needs to be nuclear plant experience.

13.2-2 O

r-BFNP-3 13.2.1.4.1.a Plant Superintendent (Ensincerina and Oper ations)

() This pl ant superintendent assists the plant manager in planning, coordinating, and directing the opera tional and engineering activities of the plant.

S The plant superintendent must have a good knowledge of the g.

nuclear processes involve d in the ge nera tion of steam reactor safety, and control systems. He is responsible for maintaining strict adherence to all spplicable re gula tions re garding safe operation of the plant.

13.2.1.4.1.b Plant Superintendent (Maintenance) I The plant superintendent (maintenance) assists the plant manager g in c a r ry in g out the r e sponsibil itie s in pl anning , directing, and coordina ting mainte nance and oth er a c tivi tie s of the plant. He is re sponsibl e for maintaining strict adherence to all appl icabl e regulations regarding safe opera tion of the plant.

13.2.1.4.2 Doerations Section I The operations section is responsible for all plant operations.

It provides operating personnel for the preoperational testing, fuel loading, startup, and opera tional testing performed under the technical direction of the General Electric Company. It is r e s po n s ibl e for coordina ting and scheduling the training program for all operations personnel. It provides the nucleus of emergency teams such as the plant rescue and fire fighting n\_/ organiza tions.

Within the operations section are f iv e shift crews. The minimum shift crew requirements for 1, 2, and 3 unit plant operation are g ive n in table 6.8.A of the Browns Ferry Technical Specifications.

A licensed senior opera tor will be on duty a the sta tion a t all times. Th er e w ill also be one licensed operator in the control room for each reactor that is not in the shut dow n mode , or when any po s s ibil i ty of r e a c t iv i ty addition to the core exists. Plant management and technical support w ill be pre se nt or on call at all times.

The operations section is under the direction of the operations section supervisor. He is assisted in all ph a se s of operation by in-line opera tions supervisors.

13.2.1.4.2.a Onorations Section Sunervisor The operations section s upe rv i so r is r esponsible for the safe and efficient opera tion of the station in accordance with the operating licenses, technical specifications, and approved 13.2-3

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procedures. He is responsible for the prepara tion and maintenance of up-to-date operating procedures and the Ol l prepara tion of operating records. He is also responsible for operator training programs and operating p er so nnel schedules and is charged with the r e spo n sibil i ty of keeping the plant manager fully informed in all matters of opera ting significance. He I shall have a minimum of 8 years of responsible power plant experience of which a minimum of 3 years shall be nuclear power plant experience. A maximum of 2 years of the remaining 5 years l of power plant experience may be f ulf illed by sa ti sf a ct ory conpletion of academic or related technical training on a one-f o r-o ne time basis. At the time of initial core loading or appointment to the active position, the operations section supervisor shall hold a Senior React or Opera t or's License. The required nuclear experience for this po sition may be reduced to 1 year, and the NRC licensing requirement may be w a iv e d if an operations supervisor has the nuclear plant experience and holds a Senior Reactor Opera t or's Li ce n se.

13.2.1.4.2.b Assistant Operations Supervisors g The assistant opera tions supervisors assist the opera tions section supervisor in reviewing, coordinating, and pl anning the l2 activities of the operations se c tion. In the absence of the operations section supervisor, a designated operations supervisor assumes the responsibilities of that po si tion. At the time of initial core loading or appointment to the active position, an opera tions supervisor shall have a minimum of 6 years of responsible power plant experience, of which a minimum of 1 year shall be nuclear plant experience. A maximum of 3 years of the remaining 5 years of power plant experience m ay be f ul filled by sa ti sf act ory compl e ti on of academic or related technical training on a o n e- f o r- o n e time basis.

13.2.1.4.2.c Shift Engineer The shift engineer on duty is in direct charge of the plant, including the startup, operation, and shutdown of the reactors, turbogenerators and their auxiliaries.

In addition to the normal shift operating personnel under his supervision, the shift engineer has control over the actions of other personnel while they are inv olve d wi th plant sy s t em s or components. During offshifts, he is in functional control of all o n- s i t e personnel. He has the pr e ro ga t ive of instituting immediate action in a ny given situation to elimina te difficulties or remove equipment from service to preclude vi ola tion of the operating l icense s , technical spe cif ica tions, or to avert possible i nj u ry to personnel or equipment.

The shift engineers have worked up through the ranks and have be come qualified for th eir supervisory po si tious by f ulf illing O

13.2-4

BFNP-3 the requirements of TV A' s formal operator tr aining pl an. This is

, ('N a comprehensiv e work-st udy training and advancement program with

. ( ,) rigorous qualifying examina tions administered by a central accrediting committee. .The program builds on a foundation of 2 years technical education in the opera tion of s t e am- e l e c t ri c generating stations. Satisfactory completion of technical a s s i g nme nt s , in grade service period requirements, and formal examinations at each level of advancement assure competence in both technical and supervisory abil i ty . Specializ ed training is used to supplement work experience, as required, to ensure that nuclear knowledge is adequate for the responsibilities of this position. At the time of initial core loading or appointment to the active position, the shift engineer shall have 5 years of poner plant experience of which a minimum of 1 year shall be nuclear plant experience. A Senior Reactor Operator's License is required.

13.2.1.4.2.D Assistant Shift Enmineer The assistant shift engineer is under the immediate supervision Ho f the shift engineer and the general supervision of the operations section supervisor. He supervises the work of operators or others assigned to him and performs mainipulative operation of equipment as required. He has completed the same format training activities and qualifying e xamina tions a s indicated for the' shift engineer in paragraph 13.2.1.5.3, except possibly for the shift engineer accrediting examination and the ingrade service period requirement to be eligible for shift engineer.

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At the time of appointment to the active position in the licensed plant, the assistant shift engineer shall h av e a minimum of a high school diploma or eq uiv al ent and four years of responsible pow er plant experience, or which a minimum of one year shall be nuclear power plant experience. Al though he will no rm ally hold a se ni or operator license, he shall as a minimum be licensed at the reactor opera t or l evel.

13.2.1.4.2.e Unit Onerators The unit operator is under the immediate supervision of the assistant shift engineer and the general supervision of the shift engineer. He s upe rvi se s the work of one or more assistant unit operators or others a ssigned to him. He is responsible for the safe and efficient opera tion of one unit and appurtenant equipment which he normally operates from the Main Control Room.

He may pe rf orm work out side the main control room as assigned.

The unit operator has completed the requirement s of TVA's c o nv e n t i o na l operator training plan to the unit operator level of competence. This is the same c ompr ehe n s iv e w ork- study program described for the shift engineers in paragraph 13.2.1.5.3.

13.2-5

BFNP-3 Spe ci al iz e d training, as 4.n d i c a t e d in Section 13.3, is used to supplement work experiences, as required, to ensure that nuclear knowledge is adequate for the responsibilities of the position.

A the time of appointment to the a c t iv e position, the unit operator shall have a high school diploma or eq uiv al ent and two years of pow e r plant experience, of which a minimum of one year shall be nuclear power plant experience. The latter w il l consist of a basic nuclear course, a plant technology course, the simulator course, and an extensive prestartup, onsite plant f amiliariz ation training phase. Subseq7ently, the unit operator w ill complete training and be licensed before assuming the re spo n s i bil i ti e s of the position on a licensed unit.

13.2.1.4.2.f Assistant Unit Operators The assistant unit operator is under the immediate supervision of the unit operator and the ge ner al supervision of the assistant shift engineer. He normally has very little supervision of others, but m ay s up e rv i se the work of lahorers or others assigned to him. He pe rf orm s work requir eme nt s and assists in the operation of equipment w ithin w ell-ds f ine d ar ea s throughout the plant. He must have completed the 2 year Student Genera ting Plant Operator Training Program as stipulated in the f orm al TVA training plan or h av e had three years of steam plant operating experience, six months of which w as in a position similar to that of assistant unit operator. Upon assuming the full r e s po n s ibil i t i e s of the position in the licensed plant, the assistant unit operator working within the plant (water treatment plant excepted), will have a minimum of a high school diploma or eq u iv al e n t and h ave completed a basic nuclear course, plant technology co ur se , and have had several weeks onsite plant f amilia riza tion. This position does not require a reactor operator license.

13.2.1.4.2.g Technicians and Repairmen Each TVA technician and repairman is a skilled journeyman. These experienced j ourneymen will initially be predominately transferees from other TVA generating plants and installations.

The primary source of new journeymen is the TV A a ppre n ti c e ship l program. This program, jointly administered by a TV A l a b o r-l management council, normally requires in excess of three years for completion. The program requires the apprentice to progress through a series of on- th e-j o b w ork and w ri t t e n assignments j designed so that he w il l develop skills equal to the recognized l journeyman standard. Related classroom and/or correspondence lesson assignments provide the technical information needed in the actual work being done on the j ob. Only employees who l succe s sf ully complete the apprenticeship program are promoted to l journeymen.

! In addition, the TVA Service Shops Section, located some 40 mil e s l

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BFNP-3 distant at Mus cl e Shoals, Alabama, is a secondary offsite source of manpower for emergency plant outages. This section provides (3) shop and field services for maj or repairs at all TVA generating plants. The work force of this se c tion varie s as the workload demands, but it usually consists of approximately 50 electricians, 5 4 machini st s, 13 machinist welders, 2 blacksmiths, 9 bo11ermakers, 9 iron workers, and 7 steamfitters.

13.2.1.4.3 Enmineerina Section The Engineering Se c tion is responsible for plant and equipment performance tests, in- pl a nt fuel operations i nv ol v in g fuel receipt and storage, core loading, core calculations, and power distribution control; other responsibilities include waste management and chemical control. It is responsible for the preparation and maintenance of up- to-d a t e procedures related to these responsibilities.

The Engineering Section provides technical support for plant operations. It carries out a comprehensive program of plant tests, st udie s, and i nv e s t i g a t i o n for the purpose of monitoring the reactor, engineered safeguards, and plant operating conditions to assure compliance with the operating licen se s and technical specifications and to improve the efficiency of the pl ant . The Engineering Section is under the direction of the

. Engineering Section Supervisor.

13.2.1.4.3.a Enaineerina Section Supervisor

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( ,) The Engineering Section Supervisor serves as supervisor of the Engineering Section and as a staff e n g i ne e r in providing engineering advice and assistance to the power plant mananger.- ,g He is responsible for initiating, planning, and coordinating the I technical support function of the plant. His experience and training must provide him with a good understanding of nuclear reactor technology, hazards, safeguards, and licensing requirements, and a knowledge of the co nt r ol sy st em s used in a nuclear plant. He will be responsible for analysis of the performance of the reactor and turbine cycle and associated equipment during the test, startup, and operation of the plant.

The Engineering Section Supervisor should have a minimum of 8 years of responsible power plant experience of which a minimum of 2 years shall be nucl e a r power plant experience. He should be a graduate with a degree in science or engineering. A maximum of 4 years of the remaining 6 years of power plant experience m ay be fulfilled by sa ti sf a ct orily completing academic training on a one- f or-one basis.

13.2.1.4.4 Power Plant Maintenance Sections The power plant maintenance sections are responsible for plant 13.2-7 N.

BFNP-3 maintenance work and in s pe c tion in the plant. This includes the coordination of scheduling and conduct of the periodic tests on the sy st em s a ssigne d t o these sections which are a ssociated with the reactor and engineered saf e guards, as required by the technical specifications and license s. The se ct i ons develop and carry out a pr eve nta tiv e m ainte nance program tha t ensures all repair work and replacement parts are consistent with the intent of applicable codes and basic requirements of the original equipment. These se ctions maintain a record file on electrical and me chanical equipment, inservice tests, ins pe c tion s, and maintenance reports.

The maintenance supervisors shall have a minimum of 7 years of responsible power plant e xpe rie nce or applicable industrial experience, including at least 1 year of nu cl e a r power plant experience. A ma ximum of 2 years of the remaining 6 years of power pl ant or industrial experience may be f ul filled by sa ti sf a ct ory completion of academic or related training on a o ne- f or- one basis.

13.2.1.4.4.s Electrical Maintenance Section Supervisor The Electrical Maintenance Section Supervisor supervise the work of the Electrical Maintenance Se c tion. He is responsible for electrical mai nt e na nce and inspection work and determines h ow th e maintenance program is carrled out. He initiates methods and procedures r el a ting to the dismantling, a s sembl ing , and repair of equipment. He is responsible for the stringent and de t a il e d inspections and equipment tests required by the NRC on the el e c t ri c al systems associated with the reactor. If maj or equipment f all s, he determines the extent of damage, degree of di sa s sembly ne ce s sa ry, manpower required, and the procedures to be f ollowed and keeps f amiliar with the progress of each j ob and the te chni c al difficulties encountered.

l 13.2.1.4.4.b Mechanical Meintenance Section Supervisor i

The Mechanical Maintenance Section Supervisor supervises the work l of the Mechanical Maintenance S e c tion. He is responsible for mechanical maintenance and inspection work and de termine s how th e maintenance program is carried out. He initiates methods and procedures r el a ting to the dismantling, a s sembl ing , and repairing of equipment tests required by the NRC on me chanical systems l a s so cia ted with the reactor. If maj or equipment f ail s, he l determinea the extent of the damage, degree of di sa s s emb ly

, necessary, manpower required, and the procedures to be followed i and keeps f amiliar with the progress of each job and the technical difficulties encountered.

l 13.2.1.4.4.c Instrument Maintenance Section Supervisor The Instrument Maintenance Section Supervisor supervises the work 13.2-8 l

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BFNP-3 of the Instrument Maintenance Section. He is responsible f or all inst rum ent m aint enance and inspection work and determinea how the O maintenance program is carried out. He initiates methods and

. procedures relating to the dismantling, assembling, and repair of equipment. He is responsible for the stringent and de tailed inspections and equipment te st s .r equired by the NRC on the

'in s trume nta tion systems associated with the reactor. If maj or equipment f ail s, he determines the extent of the damage, degree of disassembly ne ce s s a ry, manpower required, and the procedures to be followed and koops f amiliar with the progress of each job and the technical dif ficul tie s encountered. 3 13.2.1.4.4.d General Foreman Janitor. Laborer and Suncort Craft The General For eman J anitor, Laborer and Support Craft Section, under the direction of a section supervisor, is responsible for 2 planning, organizing, and supervision of the work of all plant janitors, laborers, painters and truck drivers.

13.2.1.4.5 Plant Conoliance Section The Plant Compliance Section is responsible for dealing with TVA and non-TVA group s in compliance and re gula t ory a c t ivi tie s and a s si st s other plant sections in the interpretation and implementation of regulatory commitments, licensing a c t iv iti e s, division requir eme nt s and special processes.

13.2.1.4.5a Plant Compliance Sunervisor O' 'The Plant Compliance Supervisor provides e xper ti se on various regulatory and other requirements, their implementation, and int erpre ta tion. He s upe rvi se s and directs the a c t ivi tie s of a staff of indiv idual s experienced in the following areas: Final safety analysis report and technical specifications, environmental activities a nd requirement s, special license requirements, licensing a ct iv i ti e s , emergency planning, div isi on requirements, and special processes. He provides direct assistance to the various plant sections through the expertise of the staff.

13.2.1.4.6 Industrial Safety 3 The Industrial Safety se ction, under the direction of a section supervisor, it responsible for providing t e chni c al and administrative direction for the safety and fire protection programs at the plant.

13.2.1.4.7 Health Physics The Health Physics section supervisor is responsible for direction of an adequate program of radiological hygiene 13.2-9

BFNP-3 surveillance for all plant operations i nv ol v in g potential radiation haz ards. He keeps the plant manager informed at all times of radiological haz ards and conditions related to potential personnel exposure, cont amina tio n of plant and equipment, or cont amina tion of site a nd e nv irons. His duties include training and supe rvi sing he al th physics t e ch ni ci a n s; planning and scheduling monitoring and surveillance se rvice s; scheduling technicians to assure around- the-cloc k shif t coverage as required; maintaining current data f il e s on radiation and contamina tion level s, personnel, exposures, and work restrictions; and ensuring that operations are co rried out within the provisions of developed radiological hygiene standards and procedures. He provides monitoring assistance and technical advice to plant, opera tions and medical staff in emergencies where radiation a nd contamina tion h az ards are involved.

The m inimum qualifica tions of the plant hesith physics supervisor 2 comply wi th requir eme nt s set forth in Regulatory Guide 1.8,

' Personnel Sel ection a nd Training, ' Revision 1, September 1975.

13.2.1.4.8 Plannina and Schedulina Section The Planning and Scheduling Section, under the direction of a' section s up e rv i so r , is responsible for the planning and b scheduling of all maintenance and modifica tion work a s signed to the nuclear plant.

13.2.1.5 Modification Group 3 The modifica tion group is responsible for refueling outages and the coordina tion and ins talla tion of modifications at the plant. This incl ude s directing the work of central office support forces, outside support forces, and assigned plant forces as may be needed to a c c ompl ish the schedul ed work. The sections a s s um e responsibility f or modifica tions and additions prior to fuel loading of the applicable unit. Th e se sections develop all records required by regulatory and division c ommi tme nt s , and assemble all records and prepares the final outage report.

13.2.1.5.a Modification Mananer The Modifications Supervisor supervises the work of the 3 Modifi ca tion s Group. He is responsible as the proj e c t s up e rv i so r for a ssigned proj ect s t h a t m ay include modifications of significant scope or schedule. He is responsible for identifying the total scope of a s signe d proj ect s, determining resources required (manpower, materials, tools, equipment, services, etc.) and executing the proj ect to completion. He prepares and provides input into plans, schedules, and budgets for a s si gne d w or k, coordinating the preparation of j ob orders and proj e c t a uthoriza tion s required. He is responsible for the technical adequacy, quality, and productivity of all work O

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assigned to the se c tion.

() 13.2.1.6 Site Services Site Services functions by two principal groups: the Technical Service s Group and the Support Services Group.

Technical Services includes regulatory engine e r ing support to the plant, impl eme nta tion of applicable codes and standards, and proj ect engineer support in the following. areas: r a dw a st e; h e al th physics; security; emergency planning; ch emi s try ;

inst rum ent a tion a nd cont rol s; c iv il , mechanical, and electrical pr oj e c t s ; and fire pro t e c tion. Technical-Services administers per sonal . se rvice s contracts in these areas or obtains NUC PR j services.

Support Services includes medical, administrative, document l control, drawing control, ma teri al s, and power stores support to t

! the plant, l

13.2.1.6.a Site Services Mananer The Site Service a manager report s to the site director and is responsible for establishing and carrying out the duties and functions of the Technical Service s Group and Support S e rvi c e s Group.

13.2.1.7 Desian Services Group 3 i .The Design Service s Group is responsible for providing as many as i

practical of the engineering and design functions that are the responsiblity of-the Office of Engineering.

13.2.1.7.a Desian Services Mananer The Design Services Manager reports to the Site director and provides the e xe c u t iv e and admini s tra tiv e re sponsibility for proj ect staff and professional functional supervision for the a c t ivi tie s of the De sign Se rvi ce s Group.

13.2.1.8 Quality Assurance Qua l i ty Assurance is responsible for r ev i ew of procedures and work instructions t o ensure prope r QA requirement s are met, welding inspections, electrical inspections, receipt inspections, t o rq uin g , and other se cond par ty inspections, and routine surveillances of all plant activities affecting quality.

13.2.1.8.a Quality Assure ace Mananer The Quality Assurance msnager functionally reports to the site director, but to provide independence, al so has a reporting pa th 13.2-11 i

BFNP-3 through TV A's Quality As s urance Organiza tion to the Manager of Nuclear Power. He supervises the Quality Assurance Staff.

13.2.1.9 Personnel Staff The Personnel Staff has the responsibility for accomplishing the day-to-day personnel f unc tions which are more of f e c tiv ely performed at the site.

13.2.1.9.a Personnel Staff Supervisor The Personnel Staff supervisor reports to the site director and supervisea the Site Personnel Staff.

13.2.1.10 Financial Plannina Financial Planning is responsible for establishing and carrying out site guidelines for the budgeting, financial planning,

, accounting co st control, and financial reporting of the site resources.

13.2.1.10.a Financial Plannina Supervisor The Financial Pl a n nin g supervisor reports to the site director and is r esponsible f or the Fi na nc i al Planning group.

13.2.1.11 Services Available from Other TVA Divisinns The basic responsibility for operating the Browns Ferry Nuclear Plant rests w i th the site director. Other TV A groups provide se rvice s a s r equir ed t o suppl em ent e xi st ing facilities or staff capabilities at the site. The organizations principally involved are as follows:

1. Dcveloping and maintaining the nuclear quality assurance program which e stablishe s and defines policy, interpretations, and requirements.

3

2. Maintaining this quality assurance topical report.

3. Providing the quality engineering, quality control, and s u rv e ill a n c e functions for the Division of Nuclear Services.

4. Derf orming ongoing monitoring of quality performance in NU C PR.

5. Providing quality analysts to assist line m ana ger s in resolution of problems associated with implementation of quality assurance program requirements.

13.2.1.12 Division of Nuclear Services O

13.2-12

BFNP-3 13.2.1.12.1 Director of Nuclear Services The Director of Nuclear Servicea providea policy direction and a s si st a nce necessary for essential st a ndardiz ation in practices tad procedures among operating plants a nd e stablishe s maintenance

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and te chnical standards which incorporate re gula t ory requirements. He provides capability for monitoring and

' ovalua ting perf ormance of equipment and personnel for the Manager of Nuclear Power. He is a ssisted by the Managers of Maintenance and Engineering, Licensing Branch, Emergency Preparedness and Prote ction B ranch, Nuclear Training Branch, and Radiological He al th Staf f.

13'.2.1.12.2 Licensina Branch The Licensing Branch has the following functions and r e spo nsi bil i ti e s :

1. Obtains const ruction permits, nucl ear material and fuel licenses, reload lice nse s, special proj e c t licenses, and operating lice nse s for nuclear f acilitie s.

2. Coordinates the preparation and handles the distribution of th o se licensing documents supporting the license applications described in item 1, including safety analysis reports, 3 technical specifications, license applications and amendments, topical reports, etc.

O '3. Provides engineering support in the ev alua tion of industry experience review items and new or r evi se d re gula tion s for plant operations and e nv ironmental issues.

4. Provides the Office of Nuclear Powe r with a r e gul a t o ry commitment tracking and closure sy st em to ensure TVA has fully met its re s pon sib ili tie s to r e g ula t o ry agencies.

13.2.1.12.3 Radioloalcal Services Staff The Ra di ol o gi c al Services Staff plans and develops policy for radiation protection and control programs for TV A a ctiv itie s where radiation and radioa ctiv e ma te rial s are involved. They define th e program for personnel dosimetry, define and conduct environmental radiological dose assessments in the event of a nuclear accide nt , and prov ide ce ntraliz ed labora tory se rvice s in I support of e nv i r o nm e n t al radioactivity monitoring.

13.2.1.12.4 Risk Protection Branch The Risk Protection Branch is responsible for developing and maintaining cuclear power protection programs a nd policie s, including em er ge ncy planning, nuclear security, no nr a di ol o gi c al e nv i r o nme n t al protection, and loss preve ntion and nuclear 13.2-13

B FN P- 3 insurance. The branch develops and coordinates the implementation of the TVA Radiol ogi cal Emergency Pl a n. In carrying out these re s po nsibil iti e s, these organizations verify that NUC PR comply with regulatory and TV A r equir eme nt s and policy.

13.2.1.12.5 Nuclear Trainina Branch The Nuclear Training Branch is responsible for the development, impl em ent a tion, and administration of NUC PR training activities.

The branch administers the nuclear student ge nera ting plant operator ( NSG PO ) training program, nucl e a r operator licansing training, QA/QC training, NDE training and c er t ifica tion, and other technical training program s. The brarch is responsible for the opera tion end mainte nance of the Power Operations Training Center.

The Nuclear Power Training Center organization is not a part of Browns Ferry plant or ganiza tion.

The following training is pl anned a t the Nuclear Power Training Ce n t er :

1. Senior reactor operator licensing training

2. Reactor operator licensing training

3. Senior react or operator and reactor operator r equal if ica ti on training

4. Nucl e a r student generating plant operator training

5. Simulator sh or t course for engineer and ma na g eme nt training

6. Instrument and senior instrument mechanic training

7. Shift t e ch ni c al advisor training

8. Managers and engineering certifica tion training The B rowns Ferry training simulator w ill be availabio for use as required in the training programs.

13.2.1.12.6 Maintenance and Ennineerina Mananer The Manager, Maintenance and Engineering, is responsible for developing and maintaining program s and standards and providing di rect technical support to the nucl ea r plants in the areas of maintenance and engi-neering, including mechanical engineering, electrical, instrumentation, and cont r ol s engineering, fuel management, and core de sign. He super-vises the Mechanical, Nuclear Fuel, and Electical and Instrument and 13.2-14 O

BFNP-3 L

Controls Branches, and proj e ct managers as designated.

13.2.1.17.7 Mechanical Branch The Mechanical Branch is responsible for providing programs, f standards, and direct engineering assistance to the nuclear plants for maintenance and engineering activities on safety-rel a t ed compo ne nt s and for structures, sy st ems, fire protection, in- s e rv ic e inspection, chemical and chemical support, radioactive waste, metallurgy, and containment testing activities. It develops maintenance and technical standards and procedures for this equipment as equired.

13.2.1.12.8 Electrical and Instrument and Controls Branch The Electrical and Instrument and Controls Branch is responsible for providing programs, standards, and direct engineering assistance to the nuclear plants for m aint e na nc e and engineering a c t ivi tie s on s af e ty-r el a t e d compone nt s and for electrical and instrument and controls equipment, computers, communications, and vibrations and dia gno s tic activities. It develor s mainte nance and technical standards and procedures for this equipment as required.

13.2.1.12.9 Nuclear Fuel Branch 3 The Nuclear Fuel Brecch is responsible for providing programs, standards, and direct engineering support to the nuclear plants O f or reactor analysis, core operations, and fuel ec.gineering support. It plans and implements nucl e a r fuel cycle design ano supply activities, determines de t aile d rea ctor f uel supply requitcments, proj ect s fuel cycle costs, and develops nuclear fuel e conomi c ma na g eme n t and supply programs. It performs the audit and surveillance of nuclear fuel suppliers under the cognizance of the Quality Audit s Branch.

13.2.1.13 Office of Power Operations The Office of Power Operations provides support services (e.g.,

fabrication, repair, assembly, rework) to the nuclear plants, is responsible for maintaining laboratory standards and calibrating portable measuring and test equipment and provides metallurgical and chemical a n a ly si s of various types for the Offices of Canst ruction a nd Nucl ear Power.

The Office of Power Opera tions is responsible for the calibration, functional checking and other tests required to ensure sa tisf a ct ory pe rf ormance of electrical cont rol s, instrumentation and relaying of TV A ' s transaission sy st em s and sw i t chy ards , power transformers, and main generators, including the auxiliary power sy st em within the nuclear plan:, as assigned by the plant manager.

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BFNP-3 13.2.1.14 Office of Engineerina 0

The Office of Engineering is responsible for designing all TVA power plants and other maj or f acilitie s. The services of this organization are available for a dv i c e and consultation on all engine ering problems and for carrying out the design of the maj or plant changes or additions which may be required. ,

13.2.1.15 Office of Construction The se rv ice s of the Office of Construction are available for any maj or plant construction that may be required at the site.

13.2.2 Nuclear System Startup Ornanization and Resoonsibility The Tenne s se e Valley Authority had overall r e s po n s ib ili ty for Planning, scheduling, carrying out, and documenting the plant startup programs. All aspects of plant startup conformed to the requirements of the operational quality assurance program in effect at the time of plant startup. General Electric assisted by providing technical direction a nd guidance in support of the following operations:

1. The storage, protection, installation, cleaning, initial calibration, testing, and opera tion of the nuclear system equipment, instrumentation, and material supplied by GE.

2. Tht preope ra tional testing of the nuclear plant systems in which GB-supp!ieu equipment is installed. This includes the right of review and comment on the preoperaticnal tosting of all plant systems that are related to the safety and performance of the nuclear system.

3. All operational check-outs of the nuclear sy st em, from the initial fuel loading and startup to the completion of the warranty demonstration test.

4. The e n- s i t e training of TVA personnel during the nuoleer sy st em preoperational testing, initial fuel loading, and startup activities.

Technical direction is de f ined a s that engineering and technical guidance related to the work to be performed in conne ction with the installation, initial testing, fuel loading, starting up, and initial operation of the equipment up to the time it is accepted by TV A a s an integral part of the total plant, but technical direction excludes any supe rv i si on, management, regulation, arbitra tion and/ or measur eme nt of TVA operating personnel, agents, or contractors and work rel ated there to, and excludes any responsibility for planning, scheduling, management of the work, or operation of the plant. In all instances, TV A w a s responsible O

13.2-16

BFNP-3 for performing all actual work and opera tions related to the functions f or which technical direction was furnished by GE.

The TVA Division of Engineering Design (now Office of g Engineering) was primarily responsible for scoping the preoperational test program. The Div i si on of Construction (now Office of Construction) and the General Electric Company assisted by providing details as required. Test procedure specifications for TV A-de s i gne d sy st ems were prepared by the Office of Engineering Design and reviewed by General Electric. The General Electric Company provided the preoperational test procedures, which w ere reviewed and approved by the Division of Engineering Design, for all items within the nucioar steam supply system.

The Division of Construction f inaliz ed and approved all detailed pre ope ra tional test procedures, and was responsible for scheduling and executing each preoperational test. Operators and technical assistance were supplied by the Division of Power Production (which has since been reorganized into the Office of Nuclear Power and the Division of Fo s s il and Hydro Power) and l3 technical guidance and direction was received from General Electric repr e se nt a tiv e s. The Div ision of Construction analyzed, ev al ua t e d, and approved the test data, with assistance from General Electric and the Div ision of Engineering Design. Final acceptance of teat re s ult s wa s made by the Division of Engineering Design.

The TV A Div ision of Power Production was responsible for the O overall conduct of the startup tests (fuel loading and power tests). Technical direction of the tests was provided by the General Electric Company, but the plant manager was responsible g

f or mana geme nt and safety of operation at all times. Startup test procedures provided by the General Electric Company were selectively reviewed by the Division of Engineering Design. The g plant manager had final approval authority on all startup test procedures before use. With the assistance of GE, the results of each startup test were evaluated and approved by the Division of Power Production and, on a selective basis, by the Division of Engineering Design.

13.2-17

TVA Board of Directors General Manager l

. l l Office of Corporate Manager of Power And Services Engineering g

Division of Purchasing O

i l

l Office of Project Office of Office of Office of Nuclear Power Management Power Engineering Construction Organization Operations I

Division of Quality Assurance GENERAL ORGANIZATION ,

Figure 13.2-1 Revised by Amendment 3

l O

3 t

n F e ORS m AE d REC n OLI e 2 TCV m -

CUR A 2 ENE R S y 3 I b 1 D

d e E s R i U v

e GI R F R

O T T R C O E P P R N P I L U D B S E L T A I C S I

- N l

l C

E E T C

N D F A N O R R A U O S T T O

N S C N O A E P E -

I R N M S Y I B E R

I T V I D W G E I L E A W D A T N O U I A R R P Q S M E E G Y G WN R T A O I A i I

N P R E L A E L I M F E I C C O N U A L I N F A RG R E N F R R E G E O O O N A T F F N D E C G A N C Y E P N MA I T R N O F E I Q I F F D S T 0 A A S E - Z T I F I N

RF S A AA G ET R LS O C

U E N T I

S R F O F T O C

E P R N I F

DIVISION OF QUALITY SITE ASSURANCE DIRECTOR QUALITY ASSURANCE I l l

MODIFICATIONS SITE MANAGER SERVICES DESIGN MANAGER SERVICES MANAGER !

7____________

l I PLANT e 1

MANAGER I

i e

ISEC /

s- _ _

p COMPLIANCE l

OPERATIONS &

MAINTENA!!CE ENGINEERING SUPERINTENDENT SUPERINTENDENT I

I I OPERATIONS ENGINEERING I

ASSISTANT S

SQR

Plant Compliance and ISEG are the same group

, with the ISEC function reporting to the SHIFT Site Director CR EtJS (6)

- Shift Engineer (6) i ASE (24) FACILITY ORGANIZATION FIGURE 13.2-3 UO (18) t AUO (66)

I 1

BFNP"3

1. 3.5 Nuclear Systems Operator Trainina - Hot License The purpose of this program is to prepare operators to take the NRC hot license examination. Norm ally, the students entering }

this program are assistant unit operators who are attempting to become certified to take the NRC RO examina tion.

13.3.5.1 Reactor Operator Hot License Program ibis is a t least a 30-week program conducted j ointly at BFNP and 3 a t the TVA's IVTC.

The plant portion of the reactor operator hot license program is accanplished by using the plant control room, associated plant areas, and classroom.

The POTC portion of the reactor operator hot license program is corducted utilizing both classroom and simulator training. 1 Related technical training f or trainees in the hot license pro gr am a t th e R0 l ev el c ov er s th e f oll owing :

1. Principles of reactor operation

2. De sign f eature s of nuclear plant involved

3. General characteristics of nuclear plant involved 8 3

4. Instrumentation and control systems

5. Saf ety and emergency sy stems

6. Standard and emergency opera ting procedures O 7.

8.

Radiation control and saf ety provisions Principles of heat transfer and fluid mechanics At the conclusion of the simulator training, a certification examination is administered by an NRC examiner and/or by TVA personnel who have not been involved with the applicant's training.

A general outline of the hot license program is as follows:

l 1 Plant control room observa tion training

2. Plant classroom training

3. POTC classroom training

4. POTC simulator training

5. Plant walkthrough NRC license examination

}

I 6.

l l

l O

13.3-5

BFNP-3 13.3.5.2 Senior Reactor Operator Hot License Fronram 1 This is at least a 27-week program conducted jointly at BFNP and IVA's POTC. The plant portion of the senior reactor operators hot license program is conducted in the control room, associated plant areas, and plant classrooms. no POTC portion of the program is conducted in the classroom and on the BFNP simulator. l' Related technical training for trainees in the hot license program at the senior reactor operator level is the same as for the cold license program.

Included in the 27-week senior reactor operator training program 3 is one week of supervisory training consisting of the following topic:

j' O

0 13.3-5A

B FN P-3

7. Radiation control and safety .

8. Technical spe cif ica tions O 9.

10.

Applicable portions of NRC rules and regulations Principles of heat transfer and fluid mechanics

11. The o ry of fl uid s and thermodynamics

12. Mitigation of accidents i nv ol v ing a de gr ade d core g

13. Licensee event reports (LERs), significant operating event reports (SOER), and s i gnif ici ant event reports (SER)

14. Radiological Emergency Plan A licensee is required to attend a minimum of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> during 3 each of his or her s che dule d r equalifica tion wee ks.

The f ir st two weeks of retraining each year are conducted at the The f ina l two weeks of this annual program, consisting of b POTC.

classroom sessions, including the final annual writte n e xamina tion, may be conducted at the plant or at the POTC during the last six months of the year.

Training sh all be scheduled for groups. These groups shall be relieved from regular dutiea for the training se s sions. Each license h ol de r sh all be assigned to a group and scheduled for training.

13.3.6.2 Experience Reanirements Ea ch licensed reactor operator (RO) shall manipulate the pl a nt / s imul a t or controls and each licensed senior reactor operator ( S RO) ei ther manipul ate s the pl ant / s imul a t or controls or O directs the a c t iv i ty of individuals during plant control manipulation through at least 10 re a c t iv ity co nt r ol m ani pul a t i o n s that demonstrate skill and/or f amiliarity with the reactivity control sy st em. These m ani pul a t ion s shall consist of any combina tion of reactor startups, reactor shutdowns, or other control manipulations.

A list of the NRC acceptable pl ant / s imul a t or co nt r ol manipulation- ,

and plant evolution is contained in TVA's Program Manual PM-0200, 3 2 :

Procedure 0202.05.

This procedure identifies the items required to be perf ormed on an annual basis and the items pe rf orme d on a two-year basis.

1 l

13.3.6.3 Evaluation An annual wri tten examina tion shall be administered to all license h ol d e r s after completion of retraining.

The annual written examination w il l consist of the following categoriea for reactor operators:

O 13 .3 -7

BFN P-3 3

1. Principios of nuclear power plant opera tion, thermodynamics, heat transfer, and fluid flow

2. Plant design, including safety a n'd emergency systems

3. Instruments and controls 2

4. Procedures- normal, abnormal, emergency, and radiological control There shall be a time limit of six hours for'this annual e xamina tion.

The senior reactor operator annual examina tion will include the following categories:

1. Theory of nucl e a r power plant operation fluids and th e rmodyn ami c s

2. Pl an t sy st em s, design, control, and instrumentation

3. Pro ce dur e s- normal, abnormal, emergency, and radiological control 1

4. Administrative procedures, conditions, and limitations There is a time limit of six hours for this e xamina tion.

A score of 70 percent on each category of the annual w ri t t e n e xam ina ti on is considered passing; however, an average score of 80 percent or above for all categories of the examination must be maintained by the licensee. If a licensee scores below 70 percent on any category or averages below 80 percent overall, this i ndiv idu a l is removed from license activitics and placed in accelerated training.

The plant Training Review Board shall r ev i ew any unsatisfactory evalua tion of licensee pe rf ormance. The need for additional 1 training w il l be e stablished by this board or the Chief, Nuclear Training Branch.

Any licensee assigned to accelerated retraining by the Training Review Board shall achieve a score of 80 percent or more overall on a writ te n e xamina tion in the area or areas of indicated weakne s se s before resuming the duties of the licensed position.

Approval by the Training Review Board, after review of the training and e xamina tion, is required before the licensee m ay reassume the duties of the licensed position.

The Training Review Board consists of the pl ant manager, plant superintendent (operations and engineering), power plant operations section supervisor, and operations training section supervisor. A minimum of three of the four members shall be pr e se nt when a ny action is taken by the Training R ev i ew Board.

Review shall be made within 10 days after the date of an evaluation.

O 13.3-8

... _- . .. - . .- - - .- . ~_ -. ..- _ _ _ ___-_ _ - -. - - -

t BFNP-3

1. Neutron monitoring system Vital instrumentation

()

2.

3. Primary chemistry

4. Radiation monitoring ,

S. Gas ge nera tion This training program was developed as a r. s al t of the Three Mile Island Incident and consist s of cl a s sr o om training in the above ;

listed areas with a f i nal e xamina tion. t i

13.3.9 Records ,

t 13.3.9.1 TVA Records Official records of employee qual if ica tions, experience, and ,

retraining are maintained in the official TVA Per sonal Hi st ory ,

Record (PER) by the Division of Personnel. The PHR prov ide s a ;

standardized arrangement for inf ormation of ficially recognized in !

recording and supporting employee status. The PER is maintained l In current and accurate status and is controlled as to av ail abil ity. The ma terial admitted to this record is restricted to items f or which authenticity has been confirmed through e stablished procedures, e.g., official TV A f orm s , signed ;

statements from the employee, ma na g eme n t re pre se nta t ive s , etc.

13.3.9.2. Trainina Records Records of employee participation in training activ itie s leading ;'

to promotion to a higher level of competence are maintained f or O all personnel succe s sf ully completing training.

Records supporting requests for NRC se nior operator and operator licenses are maintained in the Browns Ferry training f ile s.

l3 These records include training courses a t te nde d, retraining classes, number of reactor startups, and other inf orma tion necessary to ensure that training requirement s have been met.

Some of these records are duplicated in the PE R.

A training f il e for each member of the plant organization is 3 maintained in the plant training f il e. Inf orma tion re garding l

participation in training and retraining activities will be maintained in this training f il e, f 13.3.9.3 Trainina Proarse Evaluation i l

The training program is approved by the Chief, Nuclear Training Branch. This ensures that the content and the i nt e nt of the training program provides the necessary training for personnel !

a ssociated with reactor operations. ;

The ef f e c t ive ne s s of training programs is evalua ted by the performance of em pl oy e e s in carrying out their assigned duties and by pe rf ormance on TV A and NRC e xamina tions.

I lO' 13.3-11 I i

.. - - - . ~ _ . . _ . . , . _ _-.. _ . _ _. m_ _ _ __._ ., _ _ _ , _ . _ _ . _ . . . _ . - , - - . _ _ _ _ _ ,y_,_,.-

BFMP-3

<* 13.6 Normal Ooerations

\# 13.6.1 General Day-to-day operations are carried out by the various plant '

sections. Each section, within its assigned area of responsibility, operates with some degree of independence and freedom from clo se supe rvi sion, yet their actions are clo s e ly coordinated to beat achieve the common pu r po s e,.

The Site Director issues instructions governing employee actions and establishing standards for plant activities. Managers of principal organizations w il l issue lastructions governing a ct iv itie s unde r their cognizance. The plant manager will issue instructions which contain administrative restrictions and station requirements established to ensure safe operation of the y plant within the limits set by the facility licensea and technical specifications. They provide that plant activities will be conducted in a manner to protect the general public, i plant personnel, and equipment, j

A formalized system of written procedures conforming to the requirements of the operational quality assurance plan (Appendia D-4) is employed in support of the standard practices. Figure 1

t 13.6-1 shows the organizational structure of these procedures. g r

Procedures covering pl ant activities which might adv er se ly affect safety are put into effect only after being reviewed by l

appropriate members of the plant staf f and written authorisation of the plant manager. The se activities include operation, l

maintenance, tests, equipment changes, etc.; but not necessarily service activities such as design, procurement, modification, or other services not directly related to nuclear safety. All site procedures judged to be safety relat ed will be submitted to the plant manager through PORC. The plant manager has the responsiblity to ensure that safety related procedures prepared by his staff or submitted to him by other organizations receive required reviews and approvals before authorizations are issued.

}

The Plant Operations Review Committee (PORC), as outilmed in the technical specifications i s re sponsible for reviewing all propo sed change s to plant procedures. On the basis of the recommendations received from this group, the superintendent is responsible for determining if f urther review is required before approving a change. The plant superintendent is not authorised to approve any change which could r e s ul t in exceeding the

limitations of the operating licenses or technical specifications.

There i s, in addition to planned changes in the plant and j p r o c e it u r e s , the area of accidental or gradual changes in' plant j equipment characteristics or conditions. Each supervisor and j employee has the responsibility to be continually alert for such <

!O i

13.6-1 l

s

DFNP-3 changes and for reporting them upon detection. The periodic inspection of pl ant equipment and the continuing review and analysis of operating data from plant logs, instruments, and tests provide regular sources of information on plant conditions.

13.6.2 Normal Operatint Instructions Instructions are prepared for integrated plant operation and indiv idual system operation.

The instructions for integrated pl ant operation outline the principal st e p s required for startup and shutdown of the reactor, turbine-generator, and supporting a uxil iar i e s as an integral unit.

The system operating instructions contain preoperational requirements and instructions for startup, operation, shutdown, and anticipa ted abnormali tie s of the system concerned.

13.6.3 Emeraency Operatina Instructions Emergency operating instructions are written for conditions which may lead to inj ury to plant personnel or the public or to the rol e s se of radi oa ctiv ity in excess of established operating limits. Included are core reactivity problems, reactor wa ter level, primary and secondary containment control during a primary sy st em rupture, main steam sy s t em rupture, main steam system or y

feedwater sy s t em m a,1 f u n c t i o n , ECCS component f ailure and auxiliary sy s t em component failure.

The primary re sponsibility for initiating the c orr e c t iv e action rests upon the ope ra tor who f irst becomes aware of the situa tion. He will then notify his supervisor of the existing condition and the action he has taken. All operating personnel through training and experience h av e learned t o re co gniz e and evaluate impending f ailures or m al f un ct i on s and to initiate proper cor re c t iv e actions.

I The emergency instructions are used to train the operating personnel and make them aware of the accidents or situations that l could occur, and the propsr course of action.

13.6.4 Maintenance Instructions l

The plant maintenance program is designed to safely and officiently provide maintenance and repair to keep the pl ant in l good operating order. Maintenance is initiated through trouble reports and a preventive maintenance program. Safe working I conditions are assured by th e use of TVA's Hold Order, Clearance, and Special Work Permit procedures. Compiex maintenance operations require s t e p-b y- s t o p performance and therefore are

! de t a il e d in written in s truc tions. These instructions, covering l mechanical, electrical, and instrumentation maintenance w il l l

13.6-2 l

l

BFNP-3 -

provide inf orna tion to assure proper coordination of operating ,

and maintenance personnel as well as s t e p-by- s t e p procedures for l Items such as removal and installation of control rod drives. !

13.6.5 Radiolonical Emeraener Plan f

k i

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1 1

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13 .6-2 A i

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. M.

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OUTSIDI LT RIQU191 MENT 1

($fANDARDS, CODES, CUIDES, ETC.)

_____l_____________________

TTA CODg LETEL IT

_ TECW TOPICAL AND SPEC REPORy PSAR OTHERS

! l l l LITEL Nf}AM KP1De F*tr OTHERS g

TIER g RIQUI DIENTS l g l

___________________l__________________

G I StIE DIRECTOR STANDARD PRACTICES I

SITE DIRECTOR F%ICT N II Site Tier Precedures I i l l I SITE SERVICES MDDIFIC ATIcMS DESICM SERVICES Site Servlees Mod Services Desten Services Isot reet tone last reet tons Inst reet tone

($31) (MSI) (DSI)

FtANT MANACER LEVEL 1 Putneteel* Tier Plant Maneser (FMI) Inst mt tns inst ructione Maintenance Inst ructione Sur ret tlance SHOWNS FE RRY NALE AR PLAN T knot tuc t ione ,y,g,g g ,, g ,,,,, g g gg g,,, ,g g g gg FIN AL S AFETY operat tna defined as Flant Manaser, Site ANALYSf 3 RFPORT gnaggvet tone ,,,, g ,,, q,4 g g g,,g g ,,, ,,4 Radioloalcal Deeten Servicee Orsantastione.

PtJut? FROCEDt'RES It st ruc t tons Figure 13.6-1 Sec urit y Mher ette staff trouse maY Revised by Amendment ) Inst ruct tons toeve inservet tons at the ette g,,, g ,

Techn tes t

_ inat twet tone Sec t inn Inst ruct tone Others

if orns a i 13.7 Records t

l The TVA records program observes all acts of Congress, Executive l 1 orders. and regulations of Federal 'agencie s having j urisdiction

! in re cords adminis tra tion. TVA complie a with Federal Power l Commission regulations concerning the pre servation and dispo sal of records of public utilitie a and licensee s, insofar as thes, regulations a pply to TVA records r el a ting to the generation, lg transmission, and sale of electric energy.

l The Site' Support Services Supervisor has responsibility for l general- supe rvi sion and coordina tion of all plant re cords including those required by the Nuclear Regul atory Commission i pertaining to the opera tion of a nuclear plant.

l.

Records reflecting plant or equipment performance and records of l te st s and inspe c tions which support compliance with the plant licenses, including records of radioactivity rel ea se to the environs, are evalua ted by a cognizant reviewer in the I appropriate plant se ctions and f orw arded to the Site Support l' Service s Staf f for storage and re tention. Such records may be origina ted by any plant section.

The duty shif t engineer maintains an operating los book which is a chronological record of significant plant eve nt s and conditions. The unit operators maintain similar j ournal s containing de t a il s pertaining to the opera tion of their i ndiv idual units. The plant operators al so maintain operating data sheets which ensure their f requent observations of equipment condition and operating value s. Th e se records are examined daily by the opera tions se c tion supervisor and are support do c ume nt s for performance analysis.

The st a tion computer printouts and the operators' data sheets serve as the normal source of operating data and statistics. To l ensure continuity of inf orma tion, provision i s made for

- s uppl em ent a ry data sheets to be maintained if the computer becomes inoperative. In addition, this information is supported by las ta11ed recording and data logging ins tr ume nta tion. Th e s e records are se nt to the engineering section on a regular basis for' review and then forwarded to the Site Support Services $

Staff for storage and retention.

Each of the maintenance se ctions will maintain equipment history records describing repairs effected, derangement s occurring, alterations made, te st s conducted, and such items as are

considered noceseary to provide a comprehensive material history of the item concerned.

Radiological Health personnel maintain records of indiv idual 3 re. dis tion e xpo sur e a nd e nv i ronm e nt a l radioactivity surveillance.

I 13.7-1 t

O '

t

13.8 OPERATIONAL REVIEW AND AUDITS 13.8.1 Inolant Reviews A continuing review of operations will be performed by the station operating staff. The Plant Operations Review Committee (PORC), composed of plant employees, will al so review operations and s e rv e in an advisory capacity to the plant superintendent.

This f unction is described in Appendix D and TVA Topical Report g TR7 5-1 A, Rev. 8.

13.8.2 Onorational Quality Assurance Opera tions will be carried out according to procedure s which conform to the Operational Quality Assurance Manual and the quality assurance policies and requirements as set forth in 3 Appendix D and in Topical Report TR 75-1A Rev. 8. The Opera tional Quality Assurance Pro gram will be audited by the Operations Quality Assurance Branch.

13.8.3 Indeoendent Review The Nuclear Safety Staf f (NSS) reviews nuclear saf e ty-rel a ted a c t iv i ti e s, programs, and events, including tho s e required by pl ant technical specifications, to independently ev al ua te the 2 safety of licensed TVA nuclear plants. Audits of a c t iv i tie s identified in the plant technical specifica tions are perf ormed under the cognizance of the NSS. The NSS reports to and advises 3 (f%,'s) the Manager of Nuclear Power.

The Independent Nuclear Safety Review Program Procedure defines I the responsibility, authority, and method of operation of the NSS. This program compl ie s with the requirements of Am e r i c a n National Standard ANSI N18.7-1976/ANS-3.2, " Admi ni s t ra t ive Controls and Quality Assurance for the Operational Phase of Nuclear Power Pl ant s," as endor sed by NRC Regulatory Guide 1.33, Rev. 2, and with American National Standard ANSI /ANS-3.1,

" Selection, Qualification, and Training of Personnel for Nuclear Power Plants." 8 13.8.4 Audit Pronram The Opera tions Quality Assurance Branch of the Office of Quality Assurance is responsible for a system of planned and periodic audits of the nuclear plant and the supporting organizations. A detailed description of this audit program including areas to be audited, frequency of audits, and disposition of the audit findings is contained in Appendiz D and TVA Topical Report TR75- l 1 A, Rev. 8. The audit program complies with the requirements of l3

, ()

V 13.8-1 1

B FN P-3 ANSI N18.7-1976," Admini s t r a t iv e Controls and Quality Assurance 9

for the Operational Phase of Nuclear Power Pl ant s."

3 l

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13.8-2

I Test Number 72: Drewell Atmosshere Coolina System NOTE: RTI-72 r e pl a c e d by TI-82. 3 f

Purnose i

This test verifles the ability of the drywell atmosphere cooling i system to maintain design conditions in the d ryw ell during i operating conditions.

Descrintion l

The drywell astmosphere cooling sy st em will be placed in [

t operation and its ability to maintain normal operating temperatures laside the drywell is verified. For this test, 8 of 10 fans (and associated coils) are on, thereby demonstrating 1 25 ,

percent standby heat removal capability.

Accentance Criteria f

4 Level 1 Not a ppl icabl e.

Level 2 4

i 8 i

3. The drywell cooling system shall maintain the bulk volumetric

! average temperature in the d ryw ell below design values a during normal operation.

1

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l 13.10-9 i

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PROPOSED RULE

. . L ..o vi( "t C % t D- . ' ' - "

[ \ ATOMIC ENERGY COMMISSION

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[10CFRPART50[

LICENSING OF PRODUCTION AND UTILIZATION FACILITIES General Design Criteria 3 for Nuclear Power Plant Construction Permits 7 The. Atomic. Energy Commission has unde'r consideration an amendment to its regulation, 10 CFR Part 50, " Licensing of Production and Utilization Facili-ties," which would add an Appendix A. " General Design Criteria for Nuclear Power Plant Construction Permits." The purpose of the proposed amendment would be to provide guidance to applicants in developing the principal design criteria to be included in applications for Commission construction permits.

s These General Design Criteria would not add any new requirements, but are

i

_- intended to describe more clearly present Commission requirements to assist applicants in preparing applications.

The proposed amendment would compicment other proposed amendments to Part 50 which were published for public comment in the FEDERAL REGISTER on August 16, 1966 (31 F.R. 10891).

1/ Inasmuch as the Commission has under consideration other amendments to 10 CFR Part 50 (31 F.R.10891), the amendment proposed herein would be a further revision to Part 50 previously published for comment in the FEDERAL REGISTER.

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I The proposed amendments to Part 50 reflect a recommendation made by a seven-member Regulatory Review Panel, appointed by the Comission to study: (1) the programs and procedures for the licensing and regulation of reactors and (2) the decision-making process in the Ccmmission's regula-tory program. The Panel's report recommended the development, particularly at the construction permit stage of a licensing proceeding, of design j criteria for nuclear power plants. Work on 'the development of such criteria had been in process at the time of the Panel's study.

As a result, preliminary proposed criteria for the design of nuclear power plants were discussed with the Commission's Advisory Comittee on Reactor Safeguards and were informally distributed for public comment in

- Commission Press Release H-252 dated November 22, 1965. In developing the proposed criteria set forth in the proposed amendments to Part 50, the Comission has taken into consideration coments and suggestions from the Advisory,Comittee on Reactor Saf eguards, f rom members of industry, and from the public.

Section 50.34, paragraph (b), as published for comment in the FEDERAL REGISTER on August 16, 1966, would require that each application for a constru:- l l

tion permit include a preliminary saf ety analysi s report. The minimum informa- l tion to be included in ,this preliminary safety analysis report is (1) a descrip-tion and safety assessment of the site, (2) a sumary description of the f acility, 1 (3) a preliminary design of the f acility, (4) a preliminary saf ety analysis and evaluation of the f acility, (5) an identification of subjects expected O

-v-1 l

l

e ,.,i. ,

. to be technical specifications, and (6) a preliminary plan for the organiza-tion, training, and operation. The following information is specified for 4 inclusion as part of the preliminary design of the f acility:

" (i) The principal design criteria for the f acility; (ii) The design bases and the relation of the design bases to the principal design criteria; (iii) Information relative to materials of construction, general arrangement and approximate dimensions, suffi-

_, cient,to provide reasonable assurance that the final design will conform to the design bases with adequate margin for safety;"

The " General Design Criteria for Nuclear Power Plant Construction Permits' proposed to be included as Appendix A to this part are intended to aid the All applicant in development item (1) above, the principal design criteria.

criteria established by an applicant and accepted by the Commission would be In considering the incorporated by reference in the construction permit.

t

/

issuance of an operating license under the regulations, the Commission would assure that the criteria had been met in the detailed design and construction I of the f acility or that changes in such criteria have been justified.

Section 50.34 as published in the FEDERAL REGISTER on August 16, 1966, would be further amended by adding to Part 50 a new Appendix A containing the General Design Criteria applicable to the construction of nuclear power plants and by a specific reference to this Appendix in $50.34, paragraph (b).

i The Commission expects that the provisions of the proposed amendmer.ts relating to General Design Criteria for Nuclear Power Plant Construction

-A3-G V

l m L.:.M i

Pemits will be useful as interim guidance until such time as the Comission takes further action on them.

Pursuant to the Atomic Energy Act of 1954, as amended, and the Administrative Procedure Act of 1946, as amended, notice is hereby given that adoption of the following amendments to 10 CFR Part 50 is contemplated.

All interested persons who desire to submit written coments or suggestions in connection with-the proposed amendments sh'ould send them to the Secretary, United States Atomic Energy Commission, Washington, D.C. 20545, wi thin 60 days af ter publication of this notice in the FEDERAL REGISTER. Coments received af ter that period will be considered if it is practicable to do so, but assurance of consideration cannot be given except as to coments filed within the period specified. Copies of comments may be examined in the Comission's Public Document Room at 1717 H Street, N.W., Washington, D.C.

1. 550.34(b)(3)(1) of 10 CFR Part 50 is amended to read as follows:

{50.34 Contentgjof applications; technical information saf ety analysis report (b) Each application for a construction permit shall include a preliminary safety analysis report. The report shall cover all pertinent 2/ Inasmuch as the Commission has under consideracios. other amendments to

$50.34 (31 F.R. 10891), the amendment proposed herein would be a f urther revision of $ 50.34(b)(3)(1) previously published for comment in the FEDERAL REGISTER.

O

-A4-

' . . ~ . . .m .

'-~

s.. -

% f subjects specified in paragraph (a) of this section as fully as available !

I information permits. The minimum information to be included shall consist

-d ,

of the following:

(3) The preliminary design of the f acility, including:

(i) The principal design criteria for the facility.

Appendix A, " General De' sign Criteria for Nuclear

Power Plant Construction Permits," provides guidance for establishing the principal design criteria for nuclear power plants.. .

2. A new Appendix A is added to read as follows:

I

-A5-e

T

.. % . s s APPENDIX A GENERAL DESIGN CRITERIA FOR NUCLEAR POWEP PLANT CONSTPUCTION PERMITS1!

Table of Contents INTRODUCTION .

Group Title Cri terion No.

I. OVER ALL PLANT REQUIREMENTS '

Quality Standards 1 Performance Standards 2 Fire Protection 3 Sharing of Systems 4 Records Requirements 5 j II. PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS Reactor Core Design 6 Suppression of Power Oscillations 7 Overall Power Coef ficient 8 Reactor Coolant Pressure Boundary 9 Co nt ainment 10 III. NUCLEAR AND R ADI ATION CONTROLS Control Room 11 Instrumentation and Control Systems 12 Fission Process Monitors and Controls 13 Core Protection Systems 14 Engineered Saf ety Features Protection Systems 15 Monitoring Reactor Coolant Pressure Boundary 16 Monitoring Radioactivity Releases 17 Monitoring Fuel and Waste Storage 18 1/ Inasmuch as the Commission has under consideration other amendments to 10 CFR Part 50 (31 F.R. 10891), the amendment proposed herein would be a further revision to Part 50 previously oublished for comment in the FLDERAL REGISTER.

-A6- Appendix A

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Criterion No.

( Group Title IV. . RELI ABILITY AM) TESTABILITY OF FROTECTION SYSTEMS Protection Systems Reliability 19 Protection Systems Redundancy and Independence 20 Single Failure Definition 21 Separation of Protection and Control Instru- 22 mentation Systems -

Protection Against Multiple Disability for 23 Protection Systems Emergency Power. for Protection Systems . 24

.. Demonstration of Functional Operability of. 25 Protection Systems l Protection Systems Fail-Safe Design -26 l l

V. REACTIVITY CONTROL Redundancy of Reactivity Control 27 Raactivity Hot Shutdown Capability 28 Reactivity Shutdown Capability 29

. Reactivity Holddown Capability 30 Reactivity Control Systems Malfunction 31

Maximum Reactivity Worth of Control Rods 32 VI. REACIOR C001 ANT PRESSURE BOUNDARY Reactor Coolant Pressure Boundary Capability 33 Reactor Coolant Pressure Boundary Rapid 34 Propagation Failure Prevention l' Reactor Coolant Pressure Boundary Brittle 35 Fracture Prevention l Reactor Coolant Pressure Boundary Surveillance 36 l.

i VII. ENGINEERED SAFETY FEATURES l

A. General Requirements for Engineered Saf ety Features Engineered Safety Features Basis for Design 37 Reliability and Testability of Engineered 38 Safety Features

l. Emergency Power for Engineered Safety Features 39' Missile Protection 40 Engineered Safety Features Performance Capability 41 t Engineered Safety Features Components Capability 42 Accident Aggravation Prevention 43

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Group Title Cri te rio n No .

VII. ENGINFERED SAFETY FEATUFES B. , Emergency Core Coolina Systems Emergency Core Cooling Systems Capability 44 Inspection of Emergency Core Cooling Systems 45 Testing of Emergency Core Cooling Systems 46 Components Testing of Emergency Core Cooling Systems 47 Testing of Operational Sequence of Emergency 48

, Core Cooling Systems C. Conta i nment Containment Design Basis 49 NDT Requirement for Containment Material 53 Reactor Coolant Pressure Boundary outside 51 Containment -

Containment Heat Removal Systems 52 Containment Isolation Valves 53 Containment Leakage Rate Testing 34 Containment Periodic Leakage Rate Testing 55 Provisions for Testing of Penetrations 56 Provisions for Testin,g of Isolation Valves 57 D. Containment Pressure-Reducina Systems Inspection of Containment Pressure-Reduci ng Sa Systems

, Testing of Containment Pressure-Reducing Systems 59 Testing of Containment Spray Systems 60 Testing of Operational Sequence of Containment 61 Pressure-Reducing Systems E. Air Cleanup Systems Inspection of Air Cleanup Systems 62 Testing of Air Cleanup Systems Components 63 Testing of Air Cleanup Systeas 64 Testing of Operational Sequence of Air Cleanup 65 Sys tems

-g8- Aprendix A

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Croup Title Criterion No.

VIII. FUEL AC WASTE SICRAGE SYSTD4S Prevention of Fuel Storage Criticality 6 6 Fuel and Waste Storage Decay Heat 67 Fuel and Waste Storage Radiation Shielding 68 Protection Against Radioactivity Release from 69 Spent Fuel and Waste Storage IX. PLANT EFFLUENTS ,

-. Control of Releases of Radioactivity to the 70 Environment i

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-$9- Appendix A O

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INTRODUCTION Every applicant for a construction permit is required by the provisions of 50.34 to include the principal design criteria for the proposed f acility in the application. These General Design Criteria are intended to be used as guidance in establishing the principal design criteria for a nuclear power plant. The General Design Criteria reflect the predominating experience with water power reactors as designed and located to date, but their applicability is not limited to these reactors. They are considered generally applicable to all power reactors.

Under the Commission's regulations, an applicant must provide assurance that its principal design criteria encompass all those f acility design features required in the interest of public health and safety. There may be some power reactor cases for which fulfillment of some of the General Design Criteria may not be necessary or appropriat'e.~ There will be other cases in which these criteria are insufficient, and additional criteria must be identified and satisfied by the design in the interest of public safety. It is expected that additional criteria will be needed particularly for unusual sites and environ-mental conditions, and for new and advanced types of reactors. Within this context, the General Design Criteria should be used as a ref erence allowing additions or deletions as an individual case may warrant. Departures from the General Design Criteria should be justified.

The criteria are designated as " General Design Criteria for Nuclear Power Plant Construction Permi ts" to emphasize the key role they assume at thi s stage of the licensing process. The criteria have been categorized as Category A or Category B. Experience has shown that more definitive informa-tion is needed at the construction permit stage for the items listed in Category A than fo r th. m2 in Category B.

- A lo - Appendix ^

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I. OVERALL PLANT REQUIREMENTS CRITERION 1 - QUALITY STAbDARDS (Category A)

Those systems and components of reactor f acilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be identified and then designed, f abricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, f abrication, and inspection are used, they shall be identifi ed. Where adherence to such codes or standards does not suffice to

~

assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified.

A showing of sufficiency and applicability of codes, standards, quality

[%.'/)

assurance programs, test procedures, and inspection acceptance levels used is required.

CRITERION 2 - PERFORMANCE STANDARDS (Category A)

Those systems and components of reactor f acilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, f abricated, and erected to performance standards that will enable the f acility to withstand, without loss of the capability to protect' the public, the additional forces

'that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design

-Att- AP pendix A

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bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.

CRITERION 3 - FIRE PROTECTION (Category A)

The reactor f acility shall be designed (1) to minimize the probability of events such as fires and explosions and (2) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical throughout the facility, particularly in areas con-taining crit! cal portions of the f acility such as containment, control room, and components of C.ugineered cafety features.

CRITERION 4 - SHARING OF SYSTNiS' (Category A)

Reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing.

CRITERION 5 - RECORDS REQUIREMENTS (Category A)

Records of the design, fabrication, and construction of essential com-ponents of the plant shall be maintained by the reactor operator or under its control throughout the life of the reactor.

II. PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS CRITERION 6 - REACIDR CORE DESIGN (Category A)

The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been f12- Appendix A

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~/ stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-

~

site power.

CRITERION 7 - SUPPRESSION OF POWER OSCILLATIONS (Category B)

The core design, together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.

CRITERION 8 - OVERALL POWER COEFFICIENT (Category B)

)~~s .

The reactor shall be designed so that the overall power coefficient in the

( )

power operating range shall not be positive.

CRITERION 9 - REACTOR COOLANT PRESSURE BOUNDARY (Category A)

The reactor coolant pressure boundary shall be designed and ccnstructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime.

CRITERION 10 - CONTAINMENT (Cctegory A)

Containment shall be provided. The conta inment structure shall be designed to sustain the initial ef f ects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as l the situation requires the functional capability to protect the public.

x . A 13 - ^Poend>x ^

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III. NUCLEAR AND RADIATION CONTROLS CRITERION 11 - CONTROL ROOM (Category B)

The f acility shall be provided with a control room from which actions to maintain saf e operational status of. the plant can be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the f acility without radiation exposurn-of personnel in excess of 10 CFR 20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause.

CRITERI0b 12 - INSTRUMENTATION AND CONTROL SYSTEMS (Category B)

Instrumentation and controls shall be provided as required to monitor and i .

maintain variables within prescribed operating ranges.

CRITERION 13 - FISSION PROCESS MONITORS AND CONTROLS (Catestory B)

Means shall be provided for monitoring and maintaining control over the fission process throughout core lif e and for all conditions that can reasonably be anticipated to cauce variations in reactivity of the core, such a's indica-tion of position of control rods and concentration of soluble reactivity control poisons.

CRITERION 14 - CORE PROTECTION SYSTD4S (Category B)

Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

- A t' - ^PPendix ^

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'l CRITERION 15 - EEINEERED SAFETY FEATURES PROTECTION SYSTEMS (Category B)

Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features.

CRITERION 16 - MONITORING REACTOR COOLANT PRESSURE BOUNDARY (Category B)

Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage. .

CRITERION 17 - MONITORIE RADIOACTIVITY RELEASES (Category B)

Means shall be provided for monitoring the containment atmosphere, the facility effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions.

- t CRITERION 18 - MONITORIE FUEL AND WASTE STORAGE (Category B)

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Monitoring and alarm instrumentation shall be provided for fuel and waste storag'e and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures.

IV. RELIABILITY AND TESTABILITY OF PROTECTION SYSTENS ,

CRITERION 19 - PROTECTION SYSTEMS RELIABILITY (Category B)

Protection systems shall be designed for high functional reliability and in-service testability commensurate with the saf ety functions to be performed.

CRITERION 20 - PROTECTION SYSTEMS REDUNDANCY AND INDEPENDENCE (Category B)

Redundancy and independence designed into protection systems shall be sufficient to assure that no single failure or removal from service of any A 15 - Appendix ^

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- e r w e e . - .-yr-- ,-c y----- - - - - _ _ . - - - - - - . - - - - - - - _ . - -,

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component or channel of a system will result in loss of the protection function.

The redundancy provided shall include, as a minimum, two channels of protection for each protection function to be served. Different principles shall be used where necessary to achieve true independence of redundant instrumentation Components.

CRITERION 21 - SINGLE FAILURE DEFINITION (Category B)

Multiple f ailures resulting f rom a single event shall be treated as a single failure.

CRITERION 22 - SEPARATION OF PROTECTION AND CONTROL INSTRUMENTATION SYSTEMS (Category B)

Protection systems shall be separated from control instrumentation systems to the extent that f ailure or removal f rom service of any control instrumenta-tion system component or channel, or of those common to control instrumentation and protection circuitry, leaves intact a system satisfying all requirements for the protection channels.

CRITERION 23 - PROTECTION AGAINST MULTIPLE DISABILITY FOR PROTECTION SYSTEMS (Category B)

The ef fects of adverse conditions to which redundant channels or protec-tion systems might be exposed in common, either under normal conditions or those of an accident, shall not . result in loss of the protection function.

CRITERION 24 - EMERGENCY POWER POR PROTECTION SYSTEMS (Category B)

In the event of loss of all off site power, sufficient alternate sources of power shall be provided to permit the required functioning of the protec-tion systems.

- A 16 - Appendix A

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CRITERION 25 - DENONSTRATION OF FUNCTIONAL OPERABILITY OF PROTECTION SYSTEMS (Category B)

Means shall be included for testing protection systems while the reactor is in operatien to demonstrate that no failure or loss of redundancy has occurred. .

CRITERION 26 - PROTECTION SYSTEMS FAIL-SAFE DESIGN (Category B)

The protection systems shall be designed' to f ail into a safe state or into a state established as tolernble on a defined basis if conditions such as dis-connection of the system, loss of energy (e.g. , electric power, instrument air),

or adverse environments (e.g. , extreme heat or cold, fire, steam, or water) are expe ri enced .

V. REACTIVITY CONTROL

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fL CRITERION 27 - REDUNDANCY OF REACTIVITY CONTROL (Category A)

(mj)

At least two independent reactivity control systems, preferably of dif ferent principles, shall be provided.

CRITERION 28 - REACTIVITY HOT SHUTDOWN CAPABILITY (Categorv_ Al At least two of the reactivity control systems provided shall independently be capable of making and holding the core subcritical f rom any hot standby or hot operating condition, including those resulting f rom power changes, suf fi-ciently f ast to prevent exceeding acceptable fuel damage limits.

CRITERION 29 - REACTIVITY SHUTDOWN C APABILITY (Category A)

At least one of the reactivity control systems provided shall be capable of making the core suberitical under any condition (including anticipated operational transients) suf ficiently f ast to prevent exceeding acceptable fuel

- A 17 - ^PPendix ^

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e

e damage limits. Shutdown margins greater than the maximum worth of the most effective control rod when fully withdrawn shall be provided.

CRITERION 30 - REACTIVITY HOLDDOWN CAPABILITY (Category B)

At least one of the reactivity control systems provided shall be capable of making and holding the core suberitical under any conditions with appropriate margins for contingencies.

CRITERION 31 - REACTIVITY CONTFOL SYSTEMS MALFUNCTION (Category B)

The reactivity control systems shall be capable of sustaining any single malfunction, such as, unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient which could result in exceeding acceptable fuel damage limits.

CRITERION 32 - MAXIMUM REACTIVITY WORTH OF CONTROL RODS (Category A)

Limits, which include considerable margin, shall be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the ef fectiveness of emergency core cooling.

VI. REACTOR COCIANT PRESSURE BOUNDARY CRITERION 33 - REACIOR COOLANT PRESSURE BOUNDARY CAPABILITY (Catestory A)

The reactor coolant pressure boundary shall be capable of accommodating without rupture, and with only limited allowance for energy absorption through plastic def ormation, the static and dynamic loads imposed on any boundary

- A 18 - ^Ppendix A

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(x component as a result of any inadvertent and sudden release of energy to the coolant. As a design reference, this sudden release shall be taken as that which would result from a sudden reactivity insertion such as rod ejection

.(unless prevented by positive mechanical means), rod dropout, or cold water addition.

~ CRITERION 34 - REACTOR COOLANT PRESSURE BOUNDARY RAPID PROPAGATION FAILURE PREVENTION (Category A)

The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagating type f ailures. Consideration shall be given (a) to the notch-toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loadings, (c) to the quality control n

specified for materials and component fabrication to limit flaw sizes, and

't s

\s,/ (d) to the provisions for control over service temperature and irradiation effects which may require operational restrictions.

CRITERION 35 - REACTOR COOLANT PRESSURE BOUNDARY BRITTLE FRACTURE PREVENTION (Category A)

Under conditions where reactor coolant pressure boundary system components !

constructed of ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperatures shall be at least 120 F above the nil ductility transition (NDT) temperature of the component material if the resulting energy release is expected to be absorbed by plastic deforma-tion or 60 F above the NDT temperature of the component material if the resultkng energy release is expected to be absorbed within the elastic strain j energy' range.

- j[ 19 - Appendix A 6

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CRITEFION 36 - REACTOR COOLANT PRESSURE BOUNDARY SURVEILLANCE (Category A)

Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance program conforming with ASTM-E-185-66 shall be provided.

VII. ENGINEERED SAFETY FEATURES CRITERION 37 - ENGINEERED SAFETY FEATURES BASIS FOR DESIGN (Category A)

Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall l be designed to cope with any size reactor coolant pressure boundary break up to and including the circamferential rupture of any pipe in that boundary assuming unobstructed discharge f rom both ends.

CRITERION 38 - RELIABILITY AND TESTABILITY OF ENGINEERED SAFETY FEATURES (Category A)

All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a f acility for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered saf ety af forded by the systems, including engineered safety features, will be inf luenced by the known and the demonstrated perfomance capability and reliability of the systems, and by the extent to which the operability of such systems can be . tested and inspected where appropriate during the life of the plant.

Appendix A A 20 -

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CRITERION 39 - EMERGENCY POWER FOR ENGINEERED SAFETY FEATURES (Catenorv A)

Alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the f unctioning required of the engineered safety f eatures. As a minimum, the onsite power system and the offsite power system shall each, independently, provide this capacity assuming a f ailure of a single active component in each power system.

CRITERION 40 - MISSILE PROTECTION (Category A)

Protection for engineered safety features shall be provided against dynamic ef fects and missiles that might re sult from plant equipment failures.

CRITERION 41 - ENGINEERED SAFETY FEATURES PERFORMANCE CAPABILITY (Catenory A)

Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accom-r3 . ,

'v

) modate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.

CRITERION 42 - ENGINEERED SAFETY FEATURES COMPONENTS CAPABILITY (Category A)

Engineered saf ety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident.

CRITERION 43 - ACCIDENT AGGRAVATION PREVENTION (Category A)

Engineered safety features shall be designed so that any action of the engineered safety f eatures which might accentuate the adverse af ter-ef fects of the loss of normal cooling is avoided.

, ~s - A 21 - Appendix A t i

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CRITERION 44 - EMERGENCY CORE COOLItG SYSTEMS CAPABILITY (Category A)

At least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be previded. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere wi th the emergency core cooling function and to limit the clad metai-water reaction to negligib1h amount's f or all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe. The perf orra-ance of each emergency, core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shall not share active components and shall not share other f eatures or components unless it can be demonstrated that (a) the capability of the shared f eature or component to perfom its required function can be readily ascer,tained during reactor operation, (b) f ailure of the shared f eature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared feature or component to perfom its required function is not impaired by the effects of a loss-of-coolant accident and is not lost during the entire period this function is required following the accident.

CRITERION 45 - INSPECTION OF EMERGENCY CORE COOLING SYSTEMS (Category A)

Design provisions shall be made to f acilitate physical inspection of all critical parts of the emergency core cooling systems, including reactor vessel internals and water injection nozzles.

- A 22 - Appendix A

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i CRITERION 46 - TESTING OF EMERGENCY CORE COOLING SYSTEMS COMPONENIS (Category A)

Design provisions shall be made so that active components of the emergency core cooling systems, such as pumps and valves, can be tested periodically for operability and ' required functional performance.

CRITERION 47 - TESTING OF EMERGENCY CORE COOLING SYSTEMS (Category A)

.- A capability shall be provided to test periodically the delivery capability of the emer$ency core cooling systems at a location as close to the core as is p rac ti cal .

CRITERION 48 - TESTING OF OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS (Category A) .

A capability shall be provided to test under conditions as close to design as practical the full operational sequence that would bring the emergency core

~s cooling systems into action, in,cluding the transfer to alternate power sources.

CRITERION 49 - CONTAINMENT DESIGN BASIS (Category A).

The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that t he containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for ef fects f rom metal-water or other chemical reactions that could occur as a consequence of f ailure of emergency core cooling systems.

CRITERION 50 - NDT REQUIREMENT FOR CONTAIN4ENT MATERIAL (Category A)

Principal load carrying components of ferritic materials exposed to the external environment shall be selected so that their temperatures under normal

/ - A 23 - Appendix A

Q operating and testing conditions are not less than 30 F above nil ductility transition (NDT) temperature.

CRITERION 51 - REACTOR COOLANT PRESSURE BOUNDARY OUTSIDE CONTAIhMENT (Category A)

If part of the reactor coolant pressure boundary is outside the containment, appropriate features as necessary shall be provided to protect the health and safety of the public in case of an accidental rupture in that part. Determina-tion of the appropriateness of features such as isolation valves and additional containment shall include consideration of the environmental and population conditions surrounding the site. -

CRITERION 52 - CONTAINMEffr HEAT REMOVAL SYSTEMS (Catenorv A)

Where active heat removal systems are needed under accident conditions to prevent exceeding containment, d.esign pressure, at least two systems, preferably of dif ferent principles, each with full capacity, shall be provided.

CRITERION 53 - CONTAIhNENT ISOLATION VALVES (Category A)

Penetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus.

CRITERION 54 - CONTAINMENT LEAKAGE RATE TESTING (Category A)

Containment shall be designed so that an integrated leakage rate testing can be conducted at design pressure af ter completion and installation of all penetrations and the leakage rate measured over a suf ficient period of time to verify its conformance with required performance.

O

- A 24 - Appendix A

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CRITERION 55 - CONTAINMENT PERIODIC LEAKAGE RATE TESTING (Category A)

L The containment shall be designed so that integrated leakage rate testing '

l can be done periodically at design pressure during plant lifetime.

l-CRITERION 56 - PROVISIONS FDR TESTING OF PENETRATIONS (Category A)

Provisions shall be made for testing penetrations which have resilient seals or expansion bellows to permit leaktightness to be demonstrated at design pressure at any time.

- CRITERION 57 - PROVISIONS FOR TESTING OF ISOLATION VALVES (Category A) .

Capabjlity shall be provided for testing functional. operability of valves i s

and arisociated apparatus essential to the containment function for establishinc l

no f ailure has occurred and for determining that valve leakage does not

/

exceed acceptable limits. ,

CRITERION 58 - INSPECTION OF CONTAINMENT IRESSURE-REDUCING SYSTEMS (Category A)

- Design provisions shall be made to f acilitate the periodic physical inspection of all important components of the containment pressure-reducing

- systems, such as, pumps, valves, spray nozzles, torus, and sumps.

CRITERION 59 - TESTING OF CONTAINMENT PRESSURE-REDUCING SY3TEMS COMPONENTS '

(Category A)

The contairment pressure-reducing systems shall be design'ed. so that active components, such as pumps and valves, can be tested periodically for operability and required functional performance.

CRITERION 60 - TESTING OF CONTAINMENT SPRAY SYSTE!!S (Category A)

A capability shall be provided to test periodically the delivery capa-l bility of the containment spray system at a position as close to the spray

- A25- Appendix A

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W nozzles as is oractical.

CRITERION 61 - TESTING OF OPERATIONAL SEQUENCE OF CONTAINMENT PRESSURE-REDUCING SYSTEMS (Catenory A)

A cepability shall be provided to test under conditions as cicse to the design as practical the full operational sequence that would bring the contain-ment pressure-reducing systems into action, including the transfer to alternate power sources. -

CRITERION 62 - INSPECTION OF AIR CLE ANUP SY STOKS (Catego ry A)

Design provisions shall be made to f acilitate physical inspection of all critical parts of containment air cleanup systems, such as, ducts, filters, fans, and dampers.

CRITERION 63 - TESTING OF AIR CLEANUP SYSTEMS COMPONENTS (Category A) l Design provisions shall ba made so that active components of the air cleanup systems, such as fans and dampers, can be tested periodically for ope rability and required functional performance.

CRITERION 64 - TESTING OF AIR CLEANUP SYSTEMS (Catego ry A)

A capability shall be provided for in situ periodic testing and surveil-lance of the ai r cleanup systems to ensure (a) filter bypass paths have not developed and (b) filter and trapping materials have not deteriorated beyond acceptable limits.

CRITERION 65 - TESTING OF OPERATIONAL SEQUENCE OF AIR CLEANUP SYSTEMS (Category A)

A capability shall be provided to test under conditions as close to design as practical the full operational sequence that would bring the air cleanup

. A 26 - Appendix A

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systems into action, including the transf er to alternate power sources and the design air flow delivery capability.

VIII. FUEL AND WASTE STORAGE SYSTEMS CRITERION 66 - PREVENTION OF FUEL STORAGE CRITICALITY (Catenory B)

Criticality in new and spent fuel storage shall be prevented by physical systems or orocesses. Such means as geometrically saf e configurations shall be emphasized over procedural controls.

CRITERION 67 - FUEL AND WASTE STORAGE DECAY HEAT (Category B)

Reliable decay heat removal systems shall be designed to prevent d amage to the fuel in storage facilities that could result in radioactivity release i

to plant operating areas or the public environs.

l,,') CRFIERION 68 - FUEL AND WASTE

STORAGE RADI ATION SHIELDING (Category B)

\ >'

Shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities as required to meet the requirements of 10 CFR 20.

CRITERD3N 69 - PROTECTION AGAINST RADIOACTIVITY RELEASE FROM SPENT FUEL AND WASTE STORAGE (Category B)

Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs.

- A 27 - Appendix A

','* .s s IX. PLANT EFFLUENTS CRITERION 70 - CONTROL OF RELEASES OF RADIOACTIVITY TO THE EINIR0tNENT (Category B)

'Ihe facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaceous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous,. liquid, or solid effluents, particularly where unfavorable environ-mental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In.all cases, the design for radioactivity control shall be, justified (a) on the basis of 10 CFR 20 requirements for normal operations and for any transient situation that might reasonably be anticipated to occur and (b) on the basis of 10 CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recomended dosage levels may be required where high population densities or very large cities can be af fected by the radioactive effluents.

(Sec. 161, 68 Stat. 948; 42 U.S.C. 2201)

Dated at Washington, D. C. this twenty-eighth day of June, 1967.

For the Atomic Energy Comission.

W. B. McCool Secretary

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/ i AEC PUBLISHES GENERAL DESIGN CRITERIA

'd FOR NUCLEAR POWER PLAST CONSTRUCTION PERMITS _

The AEC is publishing for public comment a revised set of proposed General Design Criteria which have been developed to assist in the prepara-tion of applications for nuclear power plant construction permits.

In November 1965, the AEC issued an announcement requesting cot = tents These on General Design Criteria developed by its regulatory staff.

criteria were statements of design principles and objectives which have evolved over the years in licensing nuclear power plants by the AEC.

It was recognized at the time the criteria were first issued for -

, The comment that further efforts were needed to develop them more fully.

i e revision being published today reflects extensive public comments received

/g t from twenty groups or individuals, suggestions clade at meetings with the Atomic Industrial Forum, and review within the AEC.

The regulatory staf f has worked closely with the Commission's Advisory Committee on Reactor Safeguards on the development of the criteria and the revision of the proposed criteria reficcts ACRS review and concent.

The General Design Criteria reflect thc predominating experience to date with water reactors, but they are considered to be generally appli-cable'to all power reactors. The proposed criteria are intended to be used as guidance to an applicant in establishing the principal design criteria for a nuclear power plant. The framework within which the l

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criteria are presented provides sufficient flexibility to permic applicants to establish design requirements using alternate and/or additional criteria.

In particular, additional criteria will be needed for unusual sites and envirorn. ental conditions and for new or advanced types of reactors. In each case an applicant will be required t'o identify its principal design criteria and provide assurance that they encompass all those f acility design features required in the interest of public health and safety.

The criteria are designated as " General Design Criteria for Nuclear Power Plant Construction Permits" to emphasize the key role they assucc at this stage of the licensing process. The criteria have been categorized as Category A or Category B. Experience has shown that more definitive information has been needed at the construction permit stage for certain of the criteria; these have been ' designated as Category A.

Development of these criteria is part of a longer-range Co=:::ission program to develop criteria, standards, and codes for nuclear reactor plants. This includes codes and standards tha* industry is developing with AEC participation. The ultimate goal is the evolution of industry codes and standards based en accumulated knowledge and experience as has

' occurred in various fields of engineering and construction, i

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The provisions of the proposed amendment relating to General Design Criteria are expected to be useful as interim guidance until such time as the Comrission takes further action on them.

The proposed criteria, which would become Appendix A to Part 50 of the AEC's regulations, will be published in the rederal Register on Interested persons may submit written coments or sugges-

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tions to the Secretary, U. S. Atomic Energy Comission, Washington, D. C.

20545, within 60 days. A copy of the proposed " General Design Criteria for Nuclear Power Plant Construction Permits" is attached.

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. </73-33M or (Monday, November 22,1965)

73 -3 4.'.6 L " , 'l F.IUG PimLIC COXXENT ON PROPOSED DESIGN CRITERIA M HifCLEAR POWER PLANT CONSTRtJCTION PERMITS T,.e Atomic Energy Commission is seeking comment from the

. lene industry and other intorested persons on proposed i.

waral design criteria which havo been developed to assist ' l

.n the evaluation of applications for nuclear power plant mnstruction permits.

The proposed criteria have bee.n developed by the AEC l i +mlatory staff and discussed with the Comnission's Advisory i i n:anittee on Reactor Safeguards ( ACRS). They represent an .

-' Tort to set forth design and performance criteria for i r" actor systems, components and structures which have evolved meer the years in licensing of nuclear power plants by the t rr . As such, they reflect the predominating experience to onto with water reactors but most of them are generally appli-

r. Le to other reactors as well.

It is recognized that further efforts by the AEC regu-tatory staff and the ACRS will be necessary to fully develop eoese criteria. However, the criteria as now proposed are #

mfficiently advanced to submit for public comment. Also, Lney are intended to give interim guidance to applicants and reactor equipment manufacturers.

The development and publication of criteria for nuclear i cower plants was one of the key recommendations of the special -

liegulatory Review Panel which studied ways of streamlining l tne Commission's reactor licensing procedures.

In the further development of these criteria, the AEC intends to hold discussicns with organizations in the nuclear l industry and to issue from time to time explanatory informa- g tion on each criterion. Following such discussions with industry and receipt of other public comment, the AEC expects to develop and publish criteria that will serve as a basis for evaluation of applications for nuclear power plant con- t struction permits. [

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O It is recognizea that additional criteria may also be i neerted, particularly for reactors other than water reactors.

and that there may bo instances whore one or more of the presently proposed criteria may not bo applicable. Applica-tion of the criteria to a specific deci n continues 5 to involvo a considerable amount of engineering judgment.

These proposed criteria are part of a longer-range Com-nicsion program to develop criteria, standards and codes for nuclear reactors, includin6 identification of codes and standards that industry will be encouraged to undertake.

The ultimate goal is the evolution of industry codes based on necumulated knowledge and experience, as has occurred in various fields of engineering and construction.

A copy of the proposed " General Design Criteria for

(~N Nuclear Power Plant Construction Permits" is attached. Com-Q ments should be sent to the Director of Regulation, U. S.

Atomic Energy Commission, Washington, D. C. 20545, by O

February 15,1%6.

11/22/65 i .

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MLiql.,,.il. SIGN CHITF.RIA F,0R NUCLEAR P0h'liR PLANT CONSTRtlCTION PLIB!lTS, At t.iebed hereto are general design critoria used by the AEC in judging wbce i-r .. rroposed nucicar power facility can be built and operated without undue risk to the henith ano safety of the public, n cy represent design and performance criteria for reactor systems, ccaponents and structuros which have evolved over the years in licensing of ' nuclear power plants by the AdC. As aeh they reficct the predominating experience to date with wa:s r re.c:o.e. sut uost of tacm aro generally applicablo to other reactors as well.

It shou 4d bc recognized that additional critoria will be nooded for evaluation of a detailed design, particularly for unusual sites and environmental conditions, and for new and advanced types of reactors.

I Moreover, there n.sy be instances in which it can be demonstrated that one or more of the criteria need not be fulfilled. It should also be recognized that the application of these criteria to a specific design involves a considerable amount of engineering judgment.

An applicant for a construction permit should present a design approach together with data aod analy*is sufficient to give assurance that the design can reasonably be expected to fulfill the criteria.

FACl a.ITY CRITERIONJ Those features of reactor facilities which are essential to the prevention of accidents or to the mitigation of their consequwces i must be designed, fabricated, and crected to:

(a) Quality standards that reflect the importance of the safety functicn to be performed. It should be vc.y ". *cd, in this respect, that design codes commonly used for nonnuclear applications may not be adequate. -

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(b) Performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces imposed by the most severe earthquakes, flooding conditions, winds, ice, and other natural phenomena anticipated ct tha proposed site.

CRITEH10N 2 Provisions must be included to limit the extent and the consequences -

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of credible chemical reactions that could cause or materially augment the "

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release of 6ignificant amounts of fission products f rom the f acility.-

CRITERION 3 Protection must be provided against possibilities for damage of the safeguarding features of the facility by missiles generated through equipment failures inside the containment.

O RI: ACTOR CRITERION 4 I The reactor taust be designcd to accocoodate, without fuel f ailure or I

primary system dama6e, deviations from steady state norm that might be i occasioned by abnormal yet anticipated transtant events such as tripping.

4 of the turbine-generator and loss of power to the reactor, recirculation r

system pumps.

CRITERION 5 The reactor must be designed so that power or process variable 'M'

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oscillations or transients that could cause fuel failure or pramary system. .

desage are not possible or can be readily suppressed. p I

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{tilEHHsN 6 Clad fuel must be designed to accom::1odata throughout its design inietime all normal and abnormal modes of anticipated reactor operation, includin,; m design overpower condition, without experiencing significant cladoing f at tures. Unclad or vented fuela must bs dasigned with the

I similar oblective of providing control over fission products. For unclad and vented solid fuels., normal and cbncn.1al modea cf anticipated reactor operation must be achieved without excccdtn;; design release ratas of fission products from the fuct over core lifettma.

CRITERION 7 The maximum reactivity worth of control rods or elements and the rates

, with which reactivity can be inserted must be held to values such that no

, single credible mechanical or electrical control system malfunction could cause a reactivity transient capable of damaging the primary system or .

t causing significant fuel f ailure.

CRITERION 8 Heactivity shutdown capability must be provided to make and hold the core subcritical from any credible operating condition with any one control element at its position of highest reactivity.

CRITEr<f0N 9 Backup reactivity shutdown capability must be provided that is independent of normal reactivity control provisions. This system must heve the capabi11ty to shut down the reactor from any operating condition.

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.4 rRITENION 10 Heat removal systems must bo provided which are c'apable of accom.

swidating core decay heat under all anticipated abnormal and credible accident conditions, such as isolation from the main condenser and complete or partial loss of primary coolant from the reactor.

rRITERTON 11 Components of the primary coolant and containment systems must ha designard and operated so that no substantial pressure or thermal stress util be imposed on the structural materials unless the temperatures are i

well above the nil. ductility temperaturcs. For ferritic materials of the coolant envelope and the containment, minimum temperatures are tiDT + 60 F and NDT + 30 F, respectively.

CRITERION 12 O a Capability for control rod insertion under abnormal conditions must be provided.

CRITERION 13 The reactor facility must be provided with a control room fror.

which all actions can be controlled or monitored as necessary to maintain i r' safe operational status of the plant at all times. Ilx control room inunt be provided with adequate protection to permit occupancy under the condi.

tions described in Criterion 17 below, and with the means to shut down the plant and maintain it in a safe condition if such accident were to be e

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,CAITr.YtDN 84 Heans must be included in the contr.1 room to show the relative reactivity status of the reactor such as position indication of aschanical rexis or concentrations of chemical poisons.

CHI M IUN 15 A reliable reactor protection system must be provided to automatically initiate appropriate action to provent safety limits from being exceeded.

Capability must be provided for testing functional opesability of the system and for determining that no component or circuit f ailure has occurred. For instruments and control systems in vital areas where the potential conse.

quences of f ailure require redundancy, the redundant channels must be independent and must be capable of .being tested to determine that they rossin iridependent. Sufficient nedurdancy must be provided that failure or removal from service of a single component or channel will not inhibit

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necessary safety action when required. These critoria should, where applicable, be satisfied by the instru:nentation associated with containment closure and isolation systems, af terheat removal and core cooling systess, systems to prevent cold. slug accidents, and other vital systems, as well sa the reactor nuclest and process safety system.

CRITERION 16 The vital instrumentation systems of Criterion 15 must be designed so that no credible combination of circumstances can intafere with the performance of a safety function when it is needed. In particular, the effect of influences cocoon to redundant channels which are intended to

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be independent must not ncRate the operahility of a safety system.

1hc effects of grons disconnection of the system, loss of energy Q (electric power, instrument air), and adverse environment (heat from ions of instrument cooling, catreme enld, fire. steam, water, i etc.) must cause the system to go into its safest state (fail-safe) or be d.amonstrably tolcrahle on some other basis.

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CIGT!:RION 17 The containment structure, including access openings and ponstra-i tions, mt.st be designed and fabricated to accomrcdato cr dissipate without failuro the pressures and temperatures associated with the largest credible energy release including the effcets of credible metal-water or other chemical reactions uninhibited by activo quenching systens. If part of the primary coolant system is outside the primary reactor containment, appropriate safeguards must be provided for that part if necessary, to protect the health and safety of tho public, in case of ar accidental rupturo in that part of the system.

l The appropriateness of safeguards such as isolation valves, additional O containment, etc., will depend on enviror. mental and population conditions surrounding the site.

CRIT,ERION 18 Provisions must be made for the removal of host from within the containment structure as necessary to maintain the integrity of the structure under the conditions described in Criterion 17 abuve. If engineered safeguards are needed to prevent contau m nt vessel failuro due to heat released under such conditions, at least two

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independent systems must be provided, preferably of different principles, llackup equipment (e. g., water and power systems) to such engineered safeguards must also be rodundant. f ,.

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C,HITERION 19 The vnaxterum integrattd Icake.go f rom the containmont structura doder

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the conditions described in Criterion 17 chova cust meet the site exposure criterie set forth in 10 CFR 100. The containmant structure suust be designed so that the containment can be leak tested at least to design pressure conditions af ter cotepletter. and instal'1ation of all penetrattuns, anel the leakage rate measured over sacitable period to verify its con =

fo m co with required performance. The plant must be designed for later .

I,dtsatsuitablepressures.

s' CRITERION 20 All contalement structure penetrations subject to failurs such as resilient sesis and expansion bellows must be designed and constructed so that lesk-tightness can be demonstrated at design pressure at any tiine throughout operating life of the reactor.

9 CRITEEION 21 Sufficient normal and emergency sources of electfical power aunt be providsd to assure a capsbility for prompt shutdown and corttinued maintenance of the reactor facility in a safe condition under all credible circumstances.

CRITERTON 22 Valves and their associated apparatus that.are essential to the containment function edust be redundant and so arranged that no credible combination of circumstances can interfere with their necessary function.

ing. Such redundant. v.alves and associated apparatus must be ind'ependent

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anality of thenc valves and associated equipecnt to determine that no O f alluso has occurreJ and that leakage is within acceptable limits.

h dundant valves and auxiliaries must be independent. Contaimnent closure valves must be actuated by instrumentation, control circuits nnd voorgy sources which satisfy Critorion 15 and 16 above.

CRITERION 23

In determining the suitability of a facility for a proposed site the acceptance of the inherent and engineered safety afforded by the systems, materials and components, and the associated angineered safeguards built into the f acility, sill depend on their demonstrated performance capability i

and reliability and the extent to which the operability of such systems, a

materists, components, and engineered safeguards can be tested and inspecteo during the life of the plant.

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"'t(WTIVITY CONTROL QtITERION 24 All fuel storage and waste handling systems must be contained if pr:essary to prevent the accidental release of radioactivity in amounts

e,hich could affect the health and safety of the pub 1tc. l CRITERION 25 The fuel handling and storage facilities must be designed to prevent triticality and to maintain adequate shielding and cooling for spent fuel $

u. der all anticipated normal and abnormal conditions, and credible accident t.[ , ;

i conditions. Variables upon which health and safety of the public depend ,, .. . N F

vast oe monitored.

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CRITPJtION 27 The plant must be provided with systems capable of monitoring the release of radiocetivity under accident conditions.

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APPENDIX D QUALITY ASSURANCE PLAN

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s_/ FOR THE BROWNS FERRY NUCLEAR PLANT 4

D.1 Quality Assurance Durina Desian and Construction The original QA program for design and construction was described g in Appendix D of the Preliminary Safety Analysis Report. The program has been revised and upgraded several times to reflect both orga niza tional changes and changes in requirements. The program was included as Appendix D of the Final Safety Analysis l g Report and subsequently revised and included in Revision 8 to l TV A-TR7 5-1 A, " Qual i ty As sur ance Program Description for Design, g Construction and Operation of TVA Nuclear Power Plants."

Design activities by Office of Engineering after licensing shall l

be in acc*ordance with the latest approved revision of TVA-TR75-1A.

D.2 General Electric Quality System for BWR Nuclear Steam Sunoly Proiects The original Quality Assurance Program implemented by General Electric for the Browns Ferry Nuclear Plant was described in Appendix D of the Preliminary Safety Analysis Report. Over the course of performing the design and initial procurement activities for the Browns Ferry Nuclear Plant, the General Electric Quality Assurance Program was upgraded to reflect

( changes in regulatory requirements and industry standards. These changes first culminated in the G. E. Boiling Water Reactor QA Manual, NEDE-20586, Revision 0 (Reference 1) which was applicable 8, to activities starting September 2'i, 1974.

The present G. E. Boiling Water Reactor Quality Assurance Program is described in NEDE-20586, Revision 14 and in the G. E. Topical Report, NEDO-11209 (Reference 2)

' The General Electric Nuclear Fuel and Special Proj ect s Division Quality Assurance Program is described in Section 6.0 of the G.

E. BWR QA Manual. Revision 14 (Reference 1) .

Reference

1. General Electric Boiling Water Reactor Quality Assurance Manual, NEDE-2 0 5 86, Revision 14, December 15, 1982,

2. General Electric Topical Report NEDO-11209-04A, Revision 4, ,

December 31, 1982, i l

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D.3 Quality Assurance Pronram for Station Operation D.3.1 Como11ance Chapters 17.1 and 17.2 of TVA Topical Report - TV A-TR 7 5-1 A,

" Quality Assurance Program Description for Design, Construction, and Operation of TVA Nuclear Power Pl a n t s" presents an accurate and complete description of the quality assurance program for operation of Browns Ferry:

(1) In Section 17.2.12 the following applies:

The plant instrument engineer and/or plant electrical engineer shall supervise the calibration of plant instrumentation.

(2) Browns Ferry Nuclear Plant takes the following exception to item F. of Table 17.D-3 (Appendix A of Regulatory Guide 1.33):

2 TVA's alternative at Browns Ferry to separate procedures for each alarm annunciation is to provide instructions for safety-related alarms in the appropriate sy s t em operating instruction. The alarm procedures are event-l l

oriented, not specifically oriented to a given alarm annunciator; and plant operating personnel are appropriately trained in the system operating instructions.

(3) Browns Ferry Nuclear Pl ant takes the following exception to item Q of Table 17.D-3 ( Re gula t ory Guide 1.88) :

Permanent record storage facilities at Browns Ferry will l provide 2-hour minimum fire protection.

Small nonpressure floor drains (not associated with the fire protection system) are routed through the permanent record storage facility. These lines contain water only when the floor above is washed down for housekeeping.

The possibility of water being released onto the records facility is extremely remote.

l D.3.2 Critical Structures. Systems and Components TVA prepares a Critical Structure, Systems, and Components (CSSC) list for each operating plant. Items designated on this list are l treated under the operational QA program as set forth in the 3 Nuclear Quality Assurance Manual (NQAM), Reference 1.

The only requirement for inclusion of items on the CSSC list is that they be safety related. For this purpose, " s a f e ty- rel a t ed 2 items" have been defined as those that meet the following l criteria.

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/~'\ s. Those items that are necessary to ensure:

b, (1) The integrity of the reactor coolant pressure boundary, i (2) The capability to shut down the reactor and maintain it in a safe condition.

(3) The capability to prevent or mitigate the consequences 8 of an incident which could result in potential offsite exposures comparable to those specified in 10 CFR Part 100.

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b. Those items which TVA considers should receive the same level of quality assurance coverage as those listed i n " a" above.

A detailed list for Browns Ferry Nuclear Plant is contained in the Nuclear Quality Assurance Manual.

Reference 3

1. Nuclear Quality Assurance Manual.

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, Board of Directors General Manager I

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Office of Project Office of Office of Office of Nuclear Power Management Power Engineering Construction Organization Operations I

Division of l Quality Assurance j i ,

l GENERAL ORGANIZATION I Figure D.0-1 Revised by Amendment 3

TENNESSEE VALLEY AUTHORITY CHATTANOOGA, TENNESSEE 374ot 400 Chestnut Street Tower II August 21, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

In the Matter of the ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 In accordance with the requirements of 10 CFR 50.71, we are submitting under separate cover 13 copies of Amendment 3 to the Browns Ferry Nuclear Plant Final Safety Analysis Report. These copies have been assigned controlled Nos.

59 through 71. The revisions on each page are identified by a vertical bar and the number 3. On July 19, 1985, Mike Hellums of my staff and Bill Long of your staff discussed an extension for submitting this amendment.

Very truly yours, TENNESSEE ALLEY AUTHORITY VJ. A. Domer, Chief Nuclear Licensing Branch Subscribe pnq sworn to f a m,e this , f 4Tday of 4 / 1985.

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An Equal Opportunity Employer

Mr. Harold R. Denton August 21, 1985 1 cc: U.S. Nuclear Regulatory Commission Region II ATTN: Dr. J. Nelson Crace, Regional Administrator 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 Mr. R. J. Clark Browns Ferry Project Manager U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 g '8D

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