ML20030A354

From kanterella
Jump to navigation Jump to search
Chapter 11 to Final Hazards Summary Rept for Big Rock Point, Operating Procedures
ML20030A354
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 11/14/1961
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090376
Download: ML20030A354 (30)
v • d • e


Text

g

\\-

V SECTION 11 OPERATING PROCEDURES This section describes the Big Rock Point plant operr. ting p:ocedurea, which include procedural operating safeguards to be established, the plans for pre-operational testing and initial startup of the reactor system, the procedures for normal operation, and the ple.ns for handling emergency situations which may arise in the operation of the plant.

11. 1 BASIC OPERATING PRINCIPLES

'1he basic operating principles for the plant are as follows:

11. 1. 1 Before being placed into regular service, the plant will be subjected to appropriate individual component and systems tests, and an initial operation program.

11. 1. 2 Operation and control of the reactor and most of the process equipment is to be centralized in the control room, which is located in the turbine building. The control room is equipped with a main control console,with corresponding instrument panel,for control of the generator, turbine, nuclear steam supply system, reactor control systems, reactor plant auxiliary systems, and t he electrical system.

The control room is shielded so that it is tenable in event of a maximum credible accident.

11. 1. 3 There will be at least two operators (one of whom will be an AEC licensed operator) in the control room for startup and normal shutdown. There will be at least one AEC licensed operator in the control room at other times during power operation and also during refueling. No licensed operator is require d in the control room when reactor is in the cold, xenon free shutdown condition with the control rods fully inserted and rod withdrawal circuits locked.
11. 1. 4 Operators may perform certain operating functions at control panels and valve racks outside of the control room but only at the direction of or with prior knowledge of the operator in in the contro? rocm.
11. 1. 5 Sta rtap, normal shutdown, and all other repetitive operations will be performed in accordance with specific procedures.
11. 1. 6 Maintenance of most facilities outside the shielding may be undertaken by contact methods and without over-all plant shutdown.when such work can be accomplished without ex-ceeding permissible radiation exposure limits.

WOl090 b s

Soction 11 Pege 2

11. 1. 7 All tests and routine mainteunce of protective devices and critical operating equipment will be done in accordance with established schedules.
11. 1. 8 With some exceptions, the reactor enclosure and turbine building areas are habitab! during normal operation.

Radiation monitoring by f 4xed or portable instrumentation will be provided for iritM entry to all r.<diation zones.

(

11. 1. 9 All personnel leaving radiation zones, and all equipment being removed from such zones,will be surveyed to an extent f

adequate for control of contamination.

11. 1. 1 0 Irradiated fuel is to be changed by semi-remote methods, viz.,

by operators using long grappling poles through water and a lead shielded transfer cask. The water serves as both shield-ing and coolant for the irradiated fuel. To assure the pro-tection of the environs against the effects of an accident, enclosure integrity provisions will be in effect during all times that it is physically possible for the reactor to be l

made critical, or whenever the reactor system is pres-l surized.

1 f

11. 1. 11 Operation of the radioactive waste handling system will be such as to assure that the disposal of radioactive materials will not result in the exposure of any persons on or off the plant to radiation in excess of permissible limits. These operations will be performed in accordance with AEC regulations (10CFR Part 20). "

Most liquid wastes are handled in discrete batches to facili-tate control. Most gaseous and air-borne wastes are contin-uously monitored and discharged from a high stack to facili-tate atmospheric diffusion. Solid wastes are stored in underground tanks or vaults,

11. 1, 1 2 All incidents, unsafe ;

and excessive exposures to radiation will be investigated to effect procedures to prevent recurrence,

11. 1. 1 3 In the event of any situation which may compromise the safety of continued operation, it will be required procedure to shut the plant down and to take other planned emergency action to protect persons and property.

c

Section 11 Page 3

11. 2 PROCEDURAL SAFEGUARDS '

The following procedural safeguards have been or will be established to assure the operating safety of the Big Rock Point plant.

11. 2. 1 Detailed Operating and Emergency Procedures
11. 2. 1. 1

~ Detailed written procedures for all. normal and emergency.

' operations which may involve nuclear safety will be prepared'

-and issued prior to startup of 'the plant.

11.2.1.2-Instructions for normal operations will consist of detailed procedures required for the operation.of all equipment associated with the plant. Radiation control procedures will be compiled to cover all aspects wf the plant's radiation

. protection program. The bases for these procedures are the various AEC and State of Michigan regulations pertaining to j

radiation protection, implemented by the procedures and practices.used in the atomic energy industry and adapted to Consumers Power Company's operations.

11. 2. 1. 3 The emergency procedures veill be st. grated into two parts.

The first part will describe the immediate action which is to be take- 'o shut the plant down and to place it in a safe condition, the second part will describe the follow-up action which is to be taken to maintain the plant in a safe condition and to prepare to return it to operation. It is recognized that action after placing the plant in a safe con-dition will be dictated largely by the circumstances existing at the time and that to this extent, prepared procedures can-not an'd will not be substituted for the responsible judgment of plant management personnel. In addition' to the emergency procedures related to plant operations, procedures and pre-cautions related to emergencies confronting any industrial plant, such as fire, explosion and flood, will be developed.

These procedures will include specific instructions as to special precautions and procedures which must be followed because of the potential presence of radioactivity.

11. 2. 2 Measures to Prevent Operating Error Thorough training of the operating staff and systematically planned operating and maintenance procedures will combine to keep to a minimum the possibility of operator errors.

i w

I

S;cti n 11 Page 4 Each operator will be well acquainted with his specific duties and responsibilities and the action to be taken in the event of off-standa rd conditions. The following paragraphs discuss the design measures and administrative controls which will promote the safety of plant operation.

11. 2. 2. 1 Design Measures (a) The reactor safety system is designed to take appropriate safety action automatically in the event a hazardous situation is approached through operator error or equip-ment malfunction.

l (b) The control rod system is designed so that operator l

error cannot cause the addition of positive reactivity at a rate which.would be hazardous.

J (c) Equipment has been designed so that automatic controls function to eliminate the requirement of operator action j

in situations which require fast action.

i

)

(d) Controls which require constant attention during routine l

l operations have been kept to a minimum to reduce operator fatigue.

(e) In systems where a misoperation could possibly lead to serious consequences, sequential operations have l

been interlocked and/or annunciated, the rate of change l

of the controlled variables has been limited and the con -

trols have been physically separated from those of other systems.

11. 2. 2. 2

_ Administrative Control l

(a) The duties and responsibilities for each operating posi-l tion will be clearly set down in writing.

(b) Decision-making authority will be defined for the various position levels and reservations of decision-making authority specified.

(c) Minimum standards will be established for the performance of various operational and maintenance programs.

(d) All repetitive operations such as start up, shutdown, and routine maintenance will be carried out according to normal operating procedures. Specific procedures will be prepared

(

as necessar/ for non-routine cperations.

f

S3ctien 11 P:gs 5

11. 2. 3 Measttres to be Taken Following Incidents
11. 2. 3. 1 To prevent or limit adverse consequences following incidents, it will be standard procedure to:

(a) Investigate all such incidents.

(b) Suspend any unsafe operation pending an investigation.

(c) Establish any additional procedures necessary to pre-vent recurrences.

11. 2. 3. 2 Notification will be made to the AEC as required by the Facility Operating License or by 10 CFR Part 20.
11. 2. 4 Anti-Sabotage Measures
11. 2. 4. 1 Protection against acts of sabotage which might result in serious reactor accidents is afforded both by design features of the plant and by plant operating procedures and policies.
11. 2. 4. 2 The design of the reactor safety system is such that it is not considered possible to prevent a safety shutdown of the reactor by a single act of sabotage of'the system. The reactor safety system is an inherently fail safe system so disabling or destroying a single component will not place the plant in an unsafe condition. In addition, the design of the plant is such that critical equipment is located in areas, access to which is available only through controlled routes.,
11. 2. 4. 3 The operating policies and procedures established for the plant will provide the following safeguards against sabotage:

(a) Entrance into the plant will bs under the surveillance and I

control of plant personnel.

(b) Frequent patrols of the plant property will be made by plant personnel.

b.

3 :.

S2cticn 11 Page 6 (c) All personnel other than those cleared for unrestricted access to the plant will be escorted at all times while visiting the plant.

a

-(d} Plant personnel will be instructed to challenge unescorted personnel to ascertain the reason for their presence.

(c)

Provision will be made for the routine checking for the operability and response of control instrumentation and control systems to assure that failures of such equip-ment are discovered as early as possible.

11.3 DISASTER PLANS

11. 3. 1 Gene ral
11. 3. 1. 1 A disaster plan will be prepared for the Big Rock Point plant so that necessary equipment, procedures and training will have been provided should they ever be needed.

N1. 3.1. 2 The objective of actions taken under the disaster plan will be

'x to limit insofar as possible the magnitude of the accident, to prevent or limit the release of radioactive materials from the containment vessel or other plant structures, and to minimize personnel exposure to radiation.

11. 3. 1. 3 "Ihe disaster plan will be prepared prior to plant startup. It will be coordinated with state authorities prior to adoption.

Training in the disaster plan will be a part of the on-site phase of the plant personnel training program.and will include practice drills in establishing the communication channels required to make the plan effective.

11. 3. 2 Emergency Action Plans for Plant Personnel
11. 3. 2.1 Control Room Personnel Control room personnel will be responsible for the following actions:

(a) Assure that the reactor is subcritical.

(b) Assure that the containment sphere is isolated and all penetration isolation valves are closed.

m.

Ssction 11 Page 7 (c) Notify plant personnel.

(d) Notify senior member of plant management on the site.

(e) Assure that cooling of the reactor has been initiated.

(f) ' Assure that cooling of the containment vessel is maintained.

(g) Collect data from radiation monitoring equipment to assure

-that such data are ava'ilable' for determining subsequent action.

11. 3. 2. 2 Action by Plant Management The senior member of plant management present will be respon-sible for the following acti ons:

(a) Determine extent and severity of the radiological hazard.

(b) Order partial or complete evacuation of the site as required by (a).

(c) Formulate and initiate appropriate course of action.

(d) Notify Plant Supe rintendent.

(e) Notify off-site Consumers m (nagement.

(f) Notify the AEC as required by the operating license or by 10 CFR, Part 20..

11. 3. 2. 3 Control of Radiation Exposure e

Radiation exposure of plant personnel will be minimized by assembling personnel as quickly as possible after the incident in a well shielded area or areas. Normally this would be expected to be in the vicinity of the control room but might be elsewhere, depending on results from action taken in accordance with above paragraph 11. 3,2. 2 (a). If required, all or part of the station personnel may be evacuated. Provision will be made for personnel monitoring and decontamination at some check 1

point as close as possible to the site to assure that contamination is not allowed to remain on the personnel and is not spread furthe r.

11, 3. 3 Action Plans Involving the Public

11. 3. 3. 1 Depending on the nature of the incident, the magnitude of any release of radioactive material and its direction of travel, it may be necessary to take action involving the public. Such 1

3:ction 11 Page 8 i

action could include general warnings, warnings to remain indoors for a specified length of tirne, warnings to avoid certain roads or areas, warnings against using food or water from certain areas, or the evacuation of areas. If evacuation is required, control points will be established for monitoring and decontamination of personnel leaving the areas.

11. 3. 3. 2 Decisions as to courses of action can only be made based on information about the release and spread of radioactivity.

To obtain such information, p: ovision will be made for formation and equipping of survey teams to gather and report the radio-activity data.

11. 3. 3. 3 The evacuation of the public will be arranged through civil authorities with Consumers Power Company personnel acting in an advisory capacity.
11. 3. 3. 4 Recovery from such an accident will include:

(a) Continued radiological surveys of the environment including food and water supplies.

NN (b) Provision for medical examination of any personnel who received radiation dosages equal to or greater than allow-3 N able or who are known to have been in a significant radia-

\\

tion field without being equipped with dosimeters.

(c} Decontamination of areas bf habitation to permit re-entry.

(d) Controlled re-entry into the evacuated areas.

11. 4 PREOPERATIONAL TESTING
11. 4. 1 General
11. 4. 1. 1 A program of preoperational testing of equipment and systems will take place prior to initial operation of the plant. It is the purpose of this program to demonstrate that the plant has been built according to specifications and that-it is ready for initial fuel loading and startup.
11. 4. 1. 2 The test program includes those checks, adjustments, cali-brations and operations necessary to assure that initial fuel load-ing and subsequent operation can be safely undertaken. Each system will be. functionally tested with special ernphasis placed on systems important to the safety and operability of the plant.

S2cticn 11 Page 9

11. 4. 1. 3 The following is a summary cf the preoperational test pro-g ram. This information is presented to indicate the nature and scope of the program. Details may be modified as final planning and execution progresses.
11. 4. 2 Nuclear Steam Supply System
11. 4. 2. 1 Inst rumentation
11. 4. 2. 1. 1 Prior to initial loading of the reactor, all reactor instru-mentation and controls will be checked to assure their satisfactory operation. Each primary sensing element, transmitter, rec eive r, indicator, and nontrol and instrumentation device will be thoroughly inspected.. Checks will also be made of all interconnecting piping and wiring.

Esch device

.will then be thoroughly tested and calibrated in accordance with the manufacturer's recommendations. Operating con-ditions for each system will be simulated to the degree possi-ble under cold conditions. As an example, temperature devices using thermocouples for the primary element will be checked by applying a known d-c voltage to the input of the instruments to simulate the thermocouples. Limit switches used for initiating indicating lights and alarms will

^be checked by actual or simulated operating conditions.

All control devices will be exercised to assure proper operation with the accuracy and response characteristics required by the system. Operation of the feedwater control system, for example, will be simulated by varying simulated feedwater flow, steam flow, and drum water level' signals and checking the feedwater control valve response to these changes.

Set points for all devices will be checked and adjusted to their ope rating.value s.

11. 4. 2. 1. 2 Each individual circuit of the reactor safety system will be tested to see that a scram signal will initiate a reactor scram.

For example, the reactor vessel pressure sensors in the safety system will be individually tested by varying a simulated pres-j sure si; uai io the sensors.

Ar, a cir,nal level corresponaing to a scram condition-is reached, scram functions and annun-I ciation should appear for the particular channel under test.

In this manner, all scram sensors except the short reactor period and high neutron flux sensors will be tested.

j

11. 4. 2. 1. 3 A simulated scram signal will be supplied to the input of each j

of the short reactor period and high neutron flux amplifiers l

to check the scram function of these devices. All reactor safety sys-l tem coinc'idence features and the fail-safe feature of each component of this safety system will also be checked for proper action.

l l

l

-i Secti n 11 P go 10

11. 4. 2, 2 Control Rods The complet. control rod hydraulic system (accumulators, scram valvet. dump tank, p,iping, pressure regulating valves and interlocks) will be checked for proper operation. With channels in the core, each control rod will then be moved over its full travel and its speed of withdrawal and insertion will be adjusted and set. After the control rods have been checked in this mode of operation, scram testing of the control rods will be carried out. Necessary test instrumentation will be provided to check the scram time of each control rod in order that any malfunctions of the rods may be detected and repaired.

11, 4. 2. 3 Recirculation System The recirculation system piping will be filled with demineralized water. The drum and piping support action will be observed to check spring hang er settings. At normal operating water level in the drum, the recifrculation pumps will be tested.

11. 4. 2. 4 Liquid Poison System The liquid poison system will be filled with water and checks will be made for proper operation of the control valves, relief valves, chec valves, level switches, heating system and alarms. When these are shown to operate satisfactorily, the water will be removed from the system and it will be filled with the liquid poison solution. The liquid poison system will be operable prior to fuel loading.
11. 4. 2. 5 Core Spray and Post.-Incident Cooling System Preoperational testing of the core spray and post -incident cool-ing system will be done before final cleaning of the reactor water system.

The electrical interlocks and controls on the pumps and valves I

for initiating the operation of the system will be checked prior to the actual functional test of the system.

11. 4. 2. 6 Reactor Cooling Water System

~

A check will be made for proper flow through the reactor cooling water system.

I s

l Section 11 Page 11 11, 4, 2, 7 Fuel Handling and Reactor Servicing Equipment l

The extension tank will be filled with water and checked for l

leakage. Dummy fuel assemblies will be run through a com-plete cycle from the new fuel storage area into the reactor I

core, shuffled within the core and finally removed to the spent fuel storage pool. Reactor head removal tools and equipment, and control rod blade,and drive,ochannel and in-core ion chamber handling tools will be checked. Such testing will assure the proper operation of the equipment and provide training for operating personnel in all phases of tk w

11. 4. 2. 8 Shutdown Cooling System The shutdown coolirig system will bti: filled:with. water. The shutdown phmps will be started and water ciictilated through'the system. Cooling water will be admitted to the shutdown heat exchanger. The pump and heat exchanger will be checked for proper operation in accordance with manufacturer's instructions.
11. 4. 2. 9 Emergency Cooling System l

A check will be made of the steam, condensate and vent valves on the emergency condenser and of the make-up water systems for pr' per operation.

o

11. 4. 2. 1 0 Reactor Cleanup System and Initial System Cleanup The cleanup demineralizer pump will be checked in accordance with manufacturer's instructions for total differential head, l

leakage, and proper cooling. All va-1ves in the system will be operated and a check will be made of piping and valves for leakage. The demineralizer will be checked for flow, pressure drop and effluent water quality. The resin sluice system will be checked for satisfactory action in emptying and recharging of the demineralizer.

Initial cleanup of the system equipment will have been performed using only chemicals which are compatible with the chemical e

properties of the materials in the system.

11. 4. 2. 11 Steam Drum Relief Valves i

Each valve will be individually inspected prior to installation.

Valves will have been set and tested previously.

(

(

i.

Section 11 Page 12

- 11. 4. 2. 1 2

. System Heating and Pressurization At the conclusion of t. sis work with all systems in the cold con-dition, the nuclear steam supply system will be heated and t

pressurized. Expansion of systems will be checked during this period, heat exchange equipment performance will be verified,

f' and control rods will be checked for proper operation.

11. 4.'3 Turbine Plant Equipment.

+

11. 4. 3. 1 Instrumentation and Control Instrumentation and control systems: will be checked for proper

-installation of primary elements,- transmitters, indicators, f

. switches, recorders and interconnecting piping. Controls will be calibrated and contro1~ systems exercised where feasible, thus reducing instrument work to a minimum during. initial s ta r tup.

11. 4. 3. 2 Condensate Demineralizers The demineralizers will be charged with resins and the re-generation equipment operated to demonstrate proper function-ing and' accessibility. of equipment, valves and instrumentation.

4

11. 4. 3. 3 Condensate Pumps Proper connections (such as vents, equalizing lines, and bearing cooling) will be verified and operation of the. pumps will be checked. A check of condensate pump recirculation control operation will be made.
11. 4. 3. 4 Feed Pumps Proper connections ( such as vents, equalizing lines, and bearing cooling) will be verifie. - The feed pumps will be operated with d

discharge valves closed and with recirculation to the condenser hotwell.

11. 4. 3. 5 Turbine - Gene rator Preoperation checks of the turbine-generator auxiliaries and accessories will be conducted in accordance with procedures -

established by the tu2 bine-generator manufacturer.

11, 4. 3. 6 Electrical System Switchgear, relays, interlocks and other devices will be checked by using simulated signals. All emergency power sources will be thoroughly checked for proper functioning during simulated emergency conditions, i

~.

~.

S2ction 11 Page 13

11. 4. 3. 7 Circulating Water System The circulating water pumps, traveling water screens, screen wash pumps, and valves will be operated to check for proper installation and performance.
11. 4. 4 Service Equipment
11. 4. 4. 1 All auxiliary systems will be operated and checked to assure satisfactory condition and compliance with procurement and erection specifications and drawings with resp ect to performance, cleanliness, accessibility of valves, cont rols and instrumentation.
11. 4. 4. 2 These auxiliary systems are listed below:

(a) Waste Disposal (b) Service Water (c) Make-up Water (d) Instrument Air (e) Service Air e

11. 5 INITIAL CORE LOADING AND CRITICAL TESTS
11. 5. 1 Basic Test Conditions
11. 5. 1. 1 Accurate knowledge of reactor parameters and characteristics is required for, safe operation of the plant. Consequently, an extensive program of tests and measurements is planned for execution during and immediately following the initial fuel loading. The loading and critical testing program will begin when the special initial loading instrumentation and the neces-sary reactor equipment have been thoroughly checked and found to be in a safe and operable condition. Conditions applicable to this program are indicated in the following paragraphs.
11. 5. 1. 2 At the start of loading, all necessary equipment will be in a safe and operable condition and the reactor vessel will be filled with water to a minimum of three feet above the active portion of-the core.

The normal equipment and techniques will be, in general, used

. i Section 11 Page 14 for the installation of fuel. Mino-modifications will be made as required; for example:

(a) The extension tank above the reactor vessel will not be filled since the fuel will not have been irradiated to any appreciable extent.

- t (b) The fuel may be transported directly to the react or area, since the fuel will be new and can be more efficiently handled in this manner.

11. 5. 1. 3 Neutron sources of sufficient strength to provide appreciable readings at all times on the neutron sensitive chambers will be installed before loading starts. Chamber-source geometric relationships will be maintained such that the chambers measure neutron multiplication by the fuel.
11. 5. 1. 4 During this period, the control rod scram circuit will be operated by at least four neutron sensitive channels whose chambers see neutron multiplication by the fuel. Additional neutron channels will be provided for information. This instrumentation,

as well as the scram circuits and control rods, will be checked at frequent intervals. In addition, each control rod, before and after it is encompassed by fuel, will be checked for proper function; ing in all modes and for the presence of poison material in the rod.

11.5.2 Core Loading and Test Program The initial loading and critical testing program will be divided into three major parts as follows:

11. 5. 2. 1 Loading to minimum critical size cores Criticality calculations will be used to guide the experimental work. Fuel will be installed in an area provided with zirconium channels, selected to be adjacent to or surrounding a neutron source. All, or all but one or two, control rods will be inserted during fuel additions. Loading will take place in a stepwise fashion; neutron multiplication measurements will be made with all control rods withdrawn between steps. Fuel will be added in steps of one element between multiplication measurements until a minimum critical size core is obtained. When the minimum critical size core for this fuel and channel arrange-ment has been established, the keff will be estimated from period data and/or control rod calibrations. This core will be then i

disas s embled.

b l

Section 11 Page 15 The zirconium channels will be replaced with steel channels and a second (larger) minimum critical size core will be assembled in the same manner and genera 11ocation. The keff will be estimated from period and/or control rod calibration data and may be increased by the addition of one fuel elen ent to provide reac ivity for measurement of void coefficients, and temperature coefficients. Temperature coefficient measurements will probably involve employing an external heat source.

11. 5. 2. 2. _

Loading h intermediate core size Fuel will be added in increments of one-to-four elements with all control rods inserted. The loading procedure is as follows:

11.5.2.2..l. A control rod adjacent to the area to be loaded will be completely withdrawn to prove suberiticality with one rod out; the rod will be then inserted.

{

l

)

11. 5. 2. 2. 2 The control rod in the area to be loaded,will be functionally tested and inserted.

11.5.2,2.3 Fuel will be loaded; the control rod in the area loaded will be then withdrawn to prove subcriticality with the rod out and to provide I

functional testing after fuel surrounds the rod. The rod will be I

then inserted, and these three steps will be repeated.

l 11.5.2.2.4 Loading will be continued until a pre-calculated core configuration is reached. Tests will be made in this core using any of the l

following techniques.

(a) Multiplication measurements (b) Positive or negative periods (c) Low power level irradiations of detectors During this testing, uniform and non-uniform control rod patterns may be used; comparisons may be made of fuel and controls with standard and/or reference fuel and controls.

11. 5. 2. 3 Loading to " Full Core" Size The initial " full core" will contain 56 elements. Fuel will be added in increments of one-to-four elements with all control rods inserted. The procedure is identical with that described in paragraph 11. 5. 2. 2.

Loading will be continued until the core contains 56 elements. If the criterion in 11. 5, 2. 2. 1 cannot be met, the loading will be discontinued and additional steel channels will be installed. When the core is fully loaded the following tests, as a minimum, will be performed.

i

Section 11 Page 16

11. 5, 2. 3. 1 Void and temperature coefficients of reactivity are measured.
11. 5. 2. 3. 2 Critical configurations with uniform and non-uniform control-rod patterns are determined, a normal startup rod-withdrawal pattern is selected, and reactivity addition rates near critical are measured.
11. 5. 2. 3. 3 The fully shutdown and one-rod-out reactivity margins are estimated.
11. 5. 2. 3. 4 When full core critical testing is completed the operating neutron sources are installed. The neutron-sensitive chambers used in the reacter vessel for testir.g are transfe t red, ir_ a stepwi< e fashion, to their design locations concurrent with measure-ments of neutron attenuation between core and design location.

Aa initial power calibration of the nuclear instrumentation is provided.

11.6 POWER OPERATION TEST PROGRAM

11. 6. 1,

Gene ral The power test program, consisting of five phases, will commence only after the initial loading and critical test program has been completed and the results of this program evaluated and found to be satisfactory.

11. 6. 2 Phase A - Heating Ppwer 0 to 5,Mwt, O to 1035 psig The tests in this phase are run while the reactor is brought up to rated pressure and temperature for the first time. The nor-mal cold startup procedures described in the following para-graph 11. 7. 2 will be used. Reactor power will be maintained in the range of approximately Oto 5 Mwt.

The test program during this phase will include the following:

11. 6. 2. 1 Control rod performance (the ability to scram rods in the hot pressurized condition will be demonstrated).
11. 6. 2. 2 Radiation surveys.

11, 6. 2. 3 Temperature coefficient det erminations.

11, 6. 2. 4 System expansion checks.

i

11. 6. 2. 5 Reactor vessel temperature differential measurements.
11. 6. 2. 6 Emergency condenser performance test.

S:icticn 11 '

Page 17

11. 6. 2. 7 Flux wire irradiations to determine power distribution and to make preliminary calibration of the in-core system.
11. 6. 2. 8 Critical control rod configurations in the hot condition.
11. 6. 2. 9 Performance testing of reactor auxiliary equipment and cont rols.

i

11. 6. 2. 1 0 Padiochemical checks.

Extrapolation of data obtained from this phase of the test pro-gram to higher p ower will be made where applicable.

11. 6. 3 Phase B, Power 0 to 120 Mwt (0 to 75% rated},1,035 psig The reactor power will be increased in increments of approxi-mately 40 Mwt during this phase of testing. The data taken at each step will be evaluated and the results found satisfactory with respect to plant safety before proceeding to the next step.

The turbine will be placed in service during this phase. The test program at each step will include the followingt

11. 6. 3. 1 Determination of reactivity changes resulting from variations of steam flow and transient poison in the core. The effect of minor variations of reactor pressure will also be determined.
11. 6. 3. 2 Radiation level surveys throughout the entire plant.
11. 6. 3. 3 Flux wire irradiations to make preliminary calibration of the in-core system and determine power distribution.
11. 6. 3. 4 Radiochemical tests.
11. 6. 3. 5 System transient tests.

11.6.3.6 Observation of reactor stability.

11. 6. 3. 7 Standard turbine stntup tests.
11. 6. 3. 8 Ga.mma p ohe er gamma scan to de. term;ne pcn er distributies.
11. 6. 4 Phase C, Final In-core Calibration at 120 Mwt (75% rated),

1035 psig

(

11. 6. 4. 1 During this phase the reactor will be run at a steady power level such that the steam void and reactor power distribution approaches that expected at rated. power. Extensive flux wire irradiations will be performed and the data obtained will be used for a final calibration of the in-core system before the

{

approach to rated power.

'(

- Section 11 Page 18

11. 6. 4. 2 Since this will be the first time the plant will have been at high power for an appreciable length of time, extensive chemi-cal and radiochemical tests will be run. These test results will be used as reference values for later cor.parison.
11. S. 5 Phase D, Power 120 Mwt to157 Mwt (7(.,% rated to rated),1035 psig i

This phase covers the approach to and attainment of full power.

This will be done in two equal increments with the tests described for Phase B carried out at each step. Full power will be approached cautiously, the data at each step being carefully analyzed and the results found satisfactory before proceeding to the next step.

11. 6. 6 Phase E, Full Power Demonstration Test A demonstration test consisting of a rated power run to prove that the plant meets design and contractual requirements will take place during Phase E.
11. 7 NORMAL OPERATION
11. 7. 1,

Gene ral Detailed written operating procedures for all modes of plant operation will be prepared prior to the initial start-up and critical testing period. By the time the power operation test program has been completed, tlye normal operating character-istics of the plant will have been established. Appro:priate changes in these procedures will be made based on the experience of the power operation test program.

i The following is an outline of the principal normal operating procedures having a potential effect on the safe operation of the plant. This information is presented to indicate the general method of operation.

11. 7. 2 Cold Start-up A cold start-up will occur each time the reactor is returned to service following t r.

ex erded e hu t dow-Ar outline of the procedure for a normal cold start-up is as follows:

11' 7'. 2. 1 A

de ta i:ed start-up check list will be followed prior to beginning the actual start-up to assure that all appli-cable equipment and systems are in the proper condition for start-up; all a-c and d-c power systems are operating normally; l

Page.19

- Sect 2W 11 instruments, annunciators, safety circuitry and interlocks are operable and in service; and..all valve and control settings are prope r for start-up. Reactor protection systems must be available and operable before start-up. Containment vessel integrity provisions must be in effect.

11, 7. 2. 2 Each control rod will be exercised and Gnally one or more control rods will be scrammed as a fir.al check of the control rod hydraulic system and the reactor safety system.

11. 7. 2. 3 The start-up check list, when completed, will be reviewed and approved by the Shift Supervisor prior to start-up.

The reactor will be brought critical by control rod withdrawal 11, 7. 2. 4 following a prescribed _ withdrawal pattern. A coupling inte-grity check will then be made on all control rods which have An alternate been scrammed since the last coupling check.

procedure involving control rod withdrawal verification by nuclear instrumentation response during start-up may be substituted for this check. The withdrawal of rods will be based on the estimated critical position and on the response of the start-up nuclear instrumentation.

11.-7. 2. 5 The power will be adjusted once criticality is reached to maintain a coolant temperature rise rate not to exceed 100*F pe r hour. The purpose of this limit is to minimize thermal stresses in the reactor vessel walls.

11. 7. 2. 6 The turbine shaft sealing system will be placed in service as soon as sufficient steam pressure is available. ( Approxi- =

mately 150 psig- )

11. 7. 2. 7 The condenser will be evacuated with the mechanical vacuum pump and the air ejector will be placed in service.

I 11.7.2.8 Turbine heating can be started at any convenient time during this operational sequence. Af ter turbine heating is completed, and the reactor reaches rated pressure, the turbine is grad-ually brought up to speed. The load limit is on its no load 1

setting and upon reaching rated speed the generator is synchronized and connected to the line. During this time the turbine is under its speed governor control.

t

(

i

S:cticn 11 P2go 20

11. 7. 2. 9 The mode of turbine control is next transferred to the initial pressure regulator, and the solenoid transfer and reset devices are latched in their standby positions.
11. 7. 2. 1 0 The speed governor is then run up to l*s high-speed stop. The bypass valve pressure controller is then backed off 10 psi above desired reactor operating pressure.
11. 7. 2. 11 As reactor power is raised by control rod withdrawal, and more power is available, the turbine load limit is backed off so that it remains at all times slightly above the reactor power setting. Electrical load will not be permitted to exceed 5 Mwe and reactor power will be limited to 20 Mwt until the reactor reaches operating pressure.
11. 7. 2. 1 2

_ The control rods are adjusted to provide the demed powe; distribution within the core.

11. 7. 3 Hot Start up Whenever the plant has been shutdown for a short period of time and the reactor vessel and auxiliaries remain at or near operating temperature, a hot start-up procedure may be followed to return the plant to service. This procedure will be c uen-tially independent of the cause of shutdown assuming that the cause is recognized and any non-standard conditions have been co r rect ed.

The reactor instrumentation will be reset and downscaled as required and a hot start-up check list will be completed prior to the withdrawal of control rods. A coupling integrity check will then be made on all control rods wnich have been scrammed since the last coupling check. An alternate procedure involving control rod withdrav at verifi-cation by nuclear instrumentation response during startup may be substituted for this check. The critical rod position will be estimated based on the critical position for the previous run and on an evaluation of the xenon build-up. The start-up will then proceed in accordaace with paracrEpho IL 7. 2. 4 through

11. 7. 2. 1 2 of the normal cold start-up procedure outlined above.
11. 7. 4 Normal Power Operation During normal power operation, the initial pressure regulator will maintain the reactor pressure at its rated value by operat-ing the turbine admission valves. The turbine-generator load is established by the control rod positions. The principal function of the operating personnel during this period will be as follows:

i

Section 11 Page 21

11. 7. 4. 1 The maintenance of a continuous shift watch in the control room to assure prompt attention to any annunciated alarms which occur.
11. 7. 4. 2 The adjustment of the control rod pattern to accommodate changes in reactivity or load demand and to maintain the desired. power distribution.

11'. T. 4'. 3 The surveillance of all equipment for proper functioning. This will include routine tours of the operating area and the recording of data provided by the plant instrumentation.

11. 7. 4. 4 The evaluation of any abnormal conditions and the initiation of corrective action as required.
11. 7. 4. 5 Operation of the waste disposal system.
11. 7. 4. 6 Regeneration of demineralizer resins as required.
11. 7. 4. 7 Maintenance of assigned stations in a clean and orderly fashion.
11. 7. 5 Extended ~ Shutdown A normal shutdown of long dtiration will be accomplished as follows:
11. 7. 5. 1 Reactor power will be reduced as required by manipulation of the control rods, and the main generator load will be de-cre'ased simultaneously. The turbine-generator will 'oe separated from the system.
11. 7. 5. 2 All control rods will be inserted and the control rod with-drawal circuit will be locked.
11. 7. 5. 3 The removal of reactor decay heat and the reduction of reactor pressure will be accomplished by controlling reactor steam flow to the main condenser through th6 turbine by-pass line. '(This steam will be condensed, and i

returned to the reactor vessel by the reactor feed pumps. )

The rate of cooling of the reactor will not be allowed to exceed 100 *F per hour. When the condenser ceases to be an effective heat sink, the steam bypass valve will 'oe closed.

11. 7. 5. 4 The reactor shutdown cooling sy' stem may be placed in operat-ion when'the reactor pressure drops to 300 psig. This system will complete the cooling of the reactor water to 125'F.

I

(

Sectirn 11 Pago 22

11. 7. 6 Short Duration Shutdown A normal shutdown of short duration may be accomplished while maintaining system pressure. The turbine-generator will be unloaded and then separated from the system.

Reactor decay heat will be accommodated by bypassing stearn to the main condenser.

(

11.8 PROCEDURES FOR RESEARCEI & DEVELOPMENT PROGRAM TESTING

11. 8. 1 Gene ral 11.' 8. 1. 1 This testing program is for a period of 41/2 years after initial startup of the Big Rock Point plant. Descri'ption of the scope of testing to be performed in the Big Rock Point plant is given in Section 10 of this report; the Research &

Development Program (R&D) is described in detail in Contract AT(04-3)-361 between the AEC and General Electric Company.

11. 8. 1. 2 The R&D testing to be performed at the Big Rock Point plant site is scheduled in two major parts, Phase I and Phase II.

Phase I will begin after Phase E of the Power Operation Test Program (Section 11. 6), thus, there will be no interference with the initial engineering tests as a result of R&D activities.

11. 8. 2 Procedures For Phase I Testing Detailed written procedures will be prepared and issued prior to the initiation of each test involving the o- ' ention of the Big Rock Point plant. These procedures w_

vide the information for performing each test, detaile.

.o the extent appro-priate, according to the cutline as followc:

11. 8. 2. 1 The objective of each particular test.
11. 8. 2. 2 The step by step operational method for the test including specific terminal aspects of the test.
11. 8. 2. 3 The expected response of the object undergoing test and, as appropriate, the expected response of the system within which the test is being conductec' i
11. 8. 2. 4 The possible responses of the test object or system which would be considered abnormal, and the course of action tobe taken in the event that such a response becomes evident.

.~.

Secticn 11 Page 23

11. 8. 2. 5 The radiological aspects that may be associated with the personnel activities involved in the test will be included in the written procedure and the proper protective measures to be taken will be specified. These protective rneasures will be consistent with plant radiation control procedures.
11. 8. 2. 6 The responsible individual having authority to direct the continuance or discontinuance of a particular test will be specified.
11. 8. 2. 7 In addition to the details covered above in this outline, there will be sufficient instructive information provided with any special equipment, that may be employed in the course of testing.
11. 8. 2. 8 In order that each particular test will proceed according to the prearranged procedure, appropriate meetings amohg the participants will be held to minimize the likelihood of an unusual incident resulting f rom human errors during either the operation of the reactor system or the handling of radioactive materials.
11. 8. 3 Procedures For Phase II Testing Detailed written procedures will be prepared and issued prior to the initiation of these tests. The form and content of the Phase 11 procedures will reflect operational experience and knowledge gained during Phase I testing. In general, the safe operation of the Big Rock Point plant during Phase II testing will be assured by cdnsistent application of the same safety principles which will guide the entire operation of the plant.

11, 9 REFUELING 11.9.1 General Principles The refueling operation will be supervised by a Shift Supervisor and conducted in accordance with the following basic principles:

11. 9. 1. 1 Detailed written procedures will be prepared by the Nuclear Engineer prior to each refueling outage.

11.9.1.2 The insertion and removal of fuel bundles and channels will be done through the top of the reactor vessel after the removal of reactor vessel head. Water shielding is provided by flooding the reactor vessel and the refueling extension tank. Fuel l

t

Section 11 Page 24 bundles and channels will be handled by means of a grapple, transfer cask, and crane controlled by operators.

1191.3 The following reactor instrumentation will be in service and connected to the reactor safety system during all refueling operations.

(a) Reactor low water level.

(b) Scram dump tank high level.

(c) Loss of potential on the 115 volt a-c reactor protectior system buses.

(d) High flux (three channels).

(e) Short period (two ch annels).

(f)

Manual scram.

In addition, at lea st ore pf the start-up nuclea-instrument channels hl! be in service during all refueling operations.

11.9.1.4 The procedure which will be used for core alterations which increase reactivity will involve withdrawal of a central con-trol rod of maximum worth in the vicinity of the alteration before and after the alteration to verify that the core is sube ritical. All rods will be interted during the actual re-activity addition. The core will be signifidantly' subc ritica l when the control rod in the vicinity of the alteration is withdrawn if the cold shutdown requirement is being met. This requirement will be checked at frequent intervals during core alterations. Under no circumstances will the reactor be permitted to rs.ach criticality during.the withdrawal of a single control rod. Communications between the control room and the loading area will exist during all core alterations.

1191.5 The liquid poison system will be available and ready for use.

11. 9 1. 6 The radiation exposure of the refueling operators will be con-trolled and minimized by these considerations:

(a) The grappling operation within the reactor will be done through about 20 feet of water.

~

Section 11 Page 25 (b) During transfer, theirradiated fuel will normally be under a minimum of about 10 feet of water or contained in a water

_ filled lead-shielded transfer cask. The cask is equipped with a low water level alarm and a hose connection for the addition of water if needed.

t (c) Fuel removal and transfer operations will be monitored by fixed and portable instruments.

11. 9. 1. 7 To assure the protection of the environs against any possible refueling accident, containment isolation provisions will be in effect during refueling operations.
11. 9. 1. 8 Unirradiated fuel will be stored in air in a new-fuel storage area within the reactor containment vessel.
11. 9. 1. 9 Irradiated fuel and irradiated channels will be stored in the spent fuel storage pool. Storage will be at a sufficient depth to shield personnel from the irradiated material during operations over the pool.
11. 9. 2 Fuel Handling A summary of the procedural sequence for normal fuel handling is given below.
11. 9. 2. 1 New fuel will be transported to the spent-fuel storage pool from the new-fuel storage area by the reactor crane prior to the actual reactor refueling operation. As an alternate to transferring new fuel to the spent-fuel storage pool, new fuel may be loaded into the cor,e directly from the new-fuel storage ama using the monorati crane as described in paragraph 11, 9. 2.10, below.
11. 9. 2. 2 The reactor head w.ll be removed and stored near the reactor or in the spent-fuel pool depending on the level of radioactive contamination involved, 11.o.2.3 The refueling platform will be placed on the extension tank.

This platform has an opening which allows the transfer cask to pass through into the reactor.

11. 9. 2. 4 Prior to the initial insertion of the transfer cask into the reactor, the baffle will be opened using the appropriate tools.
11. 9. 2. 5 For refueling, the transfer cask will be lowered into the ex-tension tank to a depth sufficient to provide water submergence when irradiated fuel is hoisted inside. The cask lower door will be opened and the fuel grapple will be lowered into the reactor.

4 I

i

=

w

.*w e-

,y

-,,.-y

Section 11 Page 26

11. 9. 2. 6 Using a long actuator pole, an ope rator will attach the grapple to a fuel bundle to be removed,
11. 9. 2. 7 The fuel bundle will then be withdrawn vertically from the core. When clear of the core, the actuator pole will be removed and the fuel hoisted into the transfer cask.
11. 9. 2. 8 The lower door of the cask will be closed and sealed; the cask will then be moved to the fuel storage pool.
11. 9. 2. 9 At the fuel storage pool, the cask will be positioned over a storage rack and lowe red into the water. The cask door will be opened and the fuel bundle lowered into the rack.
11. 9. 2. 1 0 The fuel grapple will be removed from the fuel bundle and engaged with a new fuel bundle, which will be transferred to the reactor with the return trip of the transfer cask. How-ever, during refueling with new fuel, the monorail crane may be used to move the unirradiated fuel directly from the new-fuel storage area to the reactor core, thus eliminating the temporary storing of unirradiated fuel in the spent-fuel storage pool.

During such refueling the transfer cask will be empty during the return trip of the overhead crane from the spent-fuel pool to the reactor. With the overhead crane positioned above the reactor, the monorail crane would then be moved to the new-fuel storabe area, gica.ura new fuel bundle, transfer the bundle to the reactor, and insert the bundle into the prescribed location in the reactor core.

11. 9. 2. 11 If reshuffling of the fuel in the core is required, the operator will remove the designated fuel bundle. He will reinsert the fuel bun dle into a previously vacated position by using the fuel shuffling winch located on the refueling platform.
11. 9. 2. 1 2 Channel removal or replacement as necessary will be accom-plished with the transfer cask in a manner similar to that with fuel bundles.

11, 9. 2. 13 This procedure will be repeated until the fuel bundle movement s are complete and channel movements are complet.e. 7The top guide beams and baffle will then be re-installed to complete the refueling operation.

11. 9. 3 Relocation and Orificing of Irradiated Fuel Elements Removable orifices are installed in the core fuel channels in the inlet to each fuel bundle. These orifi'ces can be changed individually when the corresponding fuel bundle has been removed. This permits redistributing core flow if necessary during core lifetime.

^

1 t

~

~l

.L l ' ?

(

S2ction 11 Page 27 Rev 1 (3/19/62)

I1.10 WASTE DISPOSAL SYSTEMS OPERATION 11.10.I Gaseous Radioactive Wastes

('

11.10.1.1 The normal sources of gaseous radioactive wastes from the operation of the plant are:

(a) Air ejector off-gas.

(b) Gland seal condenser. and condenser mechanical vacuum pump exhaust.

I (c) Plant ventilation system exhaust.

I1.10.1.2 The air ejector off-gas monitor (dual monitoring is provided by the fuel rupture detection and air ejector off-gas monitoring systems) alarms if the off-gas release rate is such that stack release rate would reach 2.4 curies per second and initiates automatic closure of the off-gas isolation valve if the r elease rate reaches 10 curies per second.

11.10.1.3 Gaseous radioactive wastes from the plant ventilation system are discharged to atmosphere through a 240-foot stack. A stack gas monitoring system is provided. In the event that the release rate approaches 10 curles per second at the air ejector off-gas monitor, the operator will attempt to reduce the release rate or isolate the source of radioactivity. Operator action might include reduction of r eactor power, manual reactor scram and isolation.

11.10.2 Liquid Wastes

(

11.10.2.1 The plant is designed so that all radioactive contaminated or potentially contaminated liquid wastes are collected in tanks.

Each batch of liquid waste collected will be sampled to deter-I mine the concentration of radioactivity pt esent. Analysis of samples will be performed in the plant's radiochemical

{

laboratory and counting room, 11.10.2.2 Disposal of each waste batch will be dependent on the physi-(

cal and chemical properties of the solution and concentration of radioactivity. Methods available for treatment of these wastes to permit their safe disposal are:

(a) Holdup prior to disposal.

t (b) Removal of radioactivity by filtration and ion exchange.

(c) Concentration of wastes to a small volume by evaporation.

I

Secticn 11 Page 28 (d) Dilution to permissible concentrations in the circulating water discharge.

(e) Suitable combinations of the above methods.

11. 1 0. 2. 3 After treatment, each batch will be pumped to one of two waste hold-up tanks. Liquid waste collected in a waste hold-up tank will be sampled and analyzed to determine its subsequent disposal. Complete records documenting the radiochemical analysis and the volume of each waste batch which is discharged will be maintained. Detailed procedures for the release of liquid wastes will be prepared.

The Shift SuPervis,or will have the authority to release liquid wastes in conformanc.e with these procedures.

11. 10. 3 Solid Wastes Spent resins will be the principal sources of high-level soliu ruio-active waste that must be disposed of routinely. Such wastes will be slu.ced to an unde rground resin storage tank 2

of 10,000 gallon capacity for long term storage. The excess liquid from these tanks will be periodically drawn off and appropriately treated in the liquid waste treatment facilitie s.

Other sources of solid radioactive waste include such items as spent waste filter units, defective equipment, laboratory waste, and trash from the controlled area of the plant. It is planned that such material will be placed in long-term storage in an underground vault. In all cases, such mate rial will be monitored and contamination control measures and radiation dose rates established before handling or removal from the radiation zone. Trash may be compacted to the extent feasible to minimize volume.

11. 11 MAINTENANCE PRINCIPLES
11. 11. 1 The policy which applies to the maintenance of the conventional power plants of Consumers Power Company will be extended to the Big Rock plant. The Company's policy emphasizes a com-plete restoration of defective equipment together with an effective progran-of systematic, thorough preventive main-tenance as a means of assuring economical p' ant performance with maximum availability and maximum protection of per-sonnel and equipment. The following principles will be followed in the maintenance program at the Big Rock Point plant,
11. 11. 2 Damaged or defective equipment will be restored to tae the orignial condition as nearly as possiale.
11. 11. 3 Maintenance check lists will be used wherever practicabit to assure that all equipment is included in the systematic pre-ventive maintenance program and to guard against error or damage in carrying out the maintenance effort.

Section 11 Page 29

11. 11. 4 A system of equipment history records will be kept in which Will be recorded the extent of and type of repair, the regular preventive maintenance actions, as well as any nonroutine maintenance which is required. Such a record system is a valuable tool in the scheduling of work and as a means of checking for completion of work, i
11. 11. 5 The preventive maintenance program will include a schedule for exercising of normally idle components. Installed spares can be exercised with the plant in operation; other components will be exercised during periods of plant shutdown.
11. 11. 6 '

It is not contemplated that major mechanical maintenance on the principal items of plant equipment will be done with the plant in ope ration.

11. 11. 7 Instrumentation and control systems, especially the neutron power level instrumontation and the reactor safety system.

can be tested periodically with the plant..t operation, and certain portions of the systems can be replaced with spare units while the plant is in operation should malfunctions occur.

11. 11. 8 Accepted radiological protection practices will be observed in maintenance activities. The first approach to equipment for maintenance after shutdown will be made only after a radiation survey of the area has been completed and necessary radiation zone markings have been placed.
11. 11. 9 Protective clothing including coveralls, shoe' covers, gloves, and caps will be available as required for maintenance activities, as will protective equipment such as respirators and air masks.
11. 11. 1 0 Decontamination facilities will be provided in the plant and will be used whenever practicable to ' reduce the contamination level of components requiring maintenance to values such that the components do not present a significant radiation hazard and can be worked on a regular schedule, 11.12 STORAGE
11. 1 2. 1 Fuel
11. 1 2. 1. 1 New fuel is brought to the site in shipping containers by truck or rail.

It rs taken into the reactor enclosure through the equipinent lock, afte'r having been removed directly fiom the truck or car posi-tioned below the external equipment lock crane. The contai.c1 = r. re rolled on a dolly through the lock into the enclosure. Inside the enclosure, the elements are removed from the containers and inspected. The fuel bundles are then hoisted and stored in the new-fuel storage area in a vertical position by the over-head crane.

s k

^

Section 11 Page 30

11. 12. 1. 2 The new fuel will arrive in containers designed to preclude criticality while in the container. Similarly, the racks for storage of fuel and for transfer of fuel to the reactor are designed to preclude the development of a critical assembly.

l

11. 1 2. 1. 3 Irradiated fuel is removed from the reactor and transferred to the fuel storage pool. It remains in this storage pool for a suitable decay period, probably 90 or more days. When the fuel activity has decayed enough to allow transport to a reprocessing plant, it is. transferred to a shipping cask positioned alongside the pool storage racks. The cask is closed, then lifted from the pool over the pool wall. The cask is dried before removal from over the pool. It is then lowered to the operating floor level, laid on its side on a dolly, and then rolled out through the equipment lock.

The external crane is used to load the cask onto the rail car.

11, 1 2. 1. 4 The safety of the above operation is assured by:

(a) Handling and storing the irradiated fuel under water.

(b) Continuously monitoring operator action for any unusual radiation levels.

(c) The handling and storage facilities which have been desi;fncd to preclude any critical arrangement.

11. 1 2. 2 Contaminated Material
11. 12. 2. 1 Material which becomes radioactively contaminated is sur-veyed at its point of contamination, and handled under standard operating procedures. In some circumstanc's, temporary storage may be arranged in the reactor e tosure or turbine building (depending on the point of remc

'the equipment). The contaminated material will be sea.

'fnecessary for such storage.

11. 1 2. 2. 2 It is assumed above that the material discussed is so large that the dry burial vaults cannot be used. Small items (such as rags, paper and scrap) may be packaged and stored in the dry dis-posal area.
11. 1 2. 3 Hydrogen and Other Explosive Gases Hydrogen is stored in cylinders on the site. Conventional safety procedures are followed in the use and handling of this gas.

Other explosive gases such as acetylene and propane which may be used for soldering, welding and other work are stored in accord-an ce wi'th conventional safety procedures.

s

Secti:n 11 Paga 31

11. 12. 4 Flammable Material Fuel oil storage tanks are buried outside of the building limits.

The lubrication oil storage tanks and the transformer are protected by automatic sprinklers. In addition, all buildings are protected by the plant fire loop.

Small quantities of paints, solvents, and flammable cleaning solutions are stored in metal cabinets in well ventilated areas.

11. 1 2. 5 Acid and Caustics Pumps, tanks, valves, and piping have been especially selected for the handling of these corrosive materials. All routine safety practices, applicable to handling of such materials, will be utilized (rubber aprons, gloves, shoes, protective masks, emergency showers, etc).

4 s,.

,

Retrieved from "https://kanterella.com/w/index.php?title=ML20030A354&oldid=4606328"