ML19319C270

Chapter 15 of Davis-Besse PSAR, Tech Specs. Includes Revisions 1-8

ML19319C270
Person / Time
Site:	Davis Besse
Issue date:	08/01/1969
From:	TOLEDO EDISON CO.
To:
References
NUDOCS 8002110737
Download: ML19319C270 (62)

Text

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

g - N

,E ,

,M ,;p

t4' a, s

.: y.e .-

-f

,=QQl, 4,

• Q gg&%+

$j,+

a r ,y , -

- se >

be $, $ h , '

• l$ . ,

g'., c,f *. .

.y e

9% .; , .

- ,y[v8?

- .s.

O.' y,

'h'-

.rrw :-

n.

y .; og - ,'I

,; . w' ..

+s -

b '. a i.

< ~26.

4

,~

'..?-

I s

• ,3

'. , d .I t

~ .*

8 5 i 0140 15 m

a w e -e + , - g- we,evi. # a w ,w., ,, y r y Sw m - a 4 e-= # ee e - .,.e,,,w,=, - A e. ,- r e me e , we ---,e *e . --- - w e- - - -- - - -

WB TABLE OF CONTENTS n.

I Section Page 15 TECHNICAL SPECIFICATIONS 15-1 15.1 DEFINITIONS 15-1 15 2 SAFETY LIMITS AND LIMITING SAFETY SYSTD4 SETTINGS 15-5 15.2.1 SAFETY LIMITS, REACTOR CORE THERMAL MARGIN 15-5 15.2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE 15-6 15.2.3 LIMITUTG SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 15-7 15.3 LIMITING CONDITIONS FOR OPERATION 15-8 15.3.1 R*: ACTOR COOLANT SYSTD4 15-8 15.3.2 CHEMICAL ADDITION SYSTEM AND MAKEUP SYSTEM 15-13 15.3.3 EMERGENCY CORE COOLING SYSTEM 15-14 15.3.h TURBINE CYCLE EQUIPMENT 15-16 3

, 15.3.5 CONTAINMENT COOLING AND SPRAY SYSTEG 15-17 t

15.3.5 INSTRUMENTATION AND CONTROL 15-19 15 3 7 CONTAINMENT 15-22 15.3.8 AUXILIARY ELECTRICAL SUPPLY 15-23 15.3.9 REFUELING 15-26 15.3.10 WASTE MATERIALS RELEASE 15-28 15.h SURVEILLANCE STANDARDS 15-29 15.h.1 OPERATIONAL SAFETY ITEMS 15-29 15.4.2 EMERGENCY CORE COOLING SYSTD4 PERIODIC TESTING 15-30 15..h.3 CONTAINMENT COOLING SYSTEMS 15-32 15.h.h CONTAINMENT TESTING 15-3h 15.k.5 EMERGENCY POWER SYSTEM PERIODIC TESTING 15-37 15.h.6 ;AUXI'IAR1 FEED PUMP PERIODIC TESTING 15- 39 l 15.h.7 REAC,IO" COOLANT SYSTEM IN-SERVICE INSPECTION 15-40 l 15-1 0141 Amendment No. 3 l

D-B TABLE OF CONTENTS (Continued)

Oj Section M 15.h.8 REACTOR COOLANT SYSTEM INTEGRITY TESTS 15-h1 15.4.9 REACTOR CONTROL ROD SYSTEM 15-h2 15.h.10 REACTIVITY ANOMALIES . 15-h3 15.k.11 ENVIRONMENTAL RADIATION SURVEY 15-hh 15.5 DESIGN F mr 15-h5 15 5.1 SITE 15-45 15 5.2 CONTAINMENT 15-h6 15 5 3 REACTOR 15-50 15 5.h NEW AND SPENT FUEL STORAGE 15-51 15.6 ADMINISTRATIVE STANDARDS 15-52 15.6,1 ORGANIZATION, REVIEW AND AUDIT 15-52 3

15.6.2 STATION OPERATING PROCEDURES 15-55 3 15.6.3 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL 15-56 OCCURRENCE IN STATION OPERATION 15.6.h ACTION TO BE TAKEN IN THE EVENT A SAFEIY LIMIT IS EXCEEDED 15-57 15.6.5 STATION OPERATING RECORDS 15-58 i

0142 Amendment No. 3 15-11

D-3 15 TECETICAL SPECIFICATIONS 15 1 U m NITIONS The following ter=s are defined for unifom interpretation of these Technical Specifications:

15 1.1 SAFETY LIMITS Safety limits are the necessary quantitative restrictions placed upon those process variables that must be controlled in order to reasonably protect the integrity of certain of the physical barriers which guard against the uncon-trolled release of radioactivity. If any safety limit is exceeded, the reactor shall be shut down until the AEC authorizes resumption of operation.

15 1.2 LIMITING SAFETY SYSTEM SETTINGS Limiting safety cystem settings are maximum or minimum set points for auto =atic protective devices responsive to those variables having significant nuclear safety functions. These set points are so chosen that auto =atic protective actions will correct the most severe abnor=al situations antici-pated so that a safety limit is not exceeded.

15 1 3 LIMITING CONDITIONS FOR OPERATION Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. 3 15 1.4 SURVEILIANCE REQUIREMENTS Surveillance requirements are tests, calibrations, or inspections necessary to verify perfor=ance, availability, or status of equipment and systems required for safe operation of the facility.

15 1 5 DESIGN FEATURES Design features are those features of the facility important to nuclear safety, such as materials of construction and geometric arrangements, which are not otherwise covered in the technical specifications.

15 1.6 ADMINISTRATIVE STAITDARDS Administrative standards are the provisions relating to organization and management, procedures, record keeping, review and audit, and reporting that are necessary to assure responsible, safe operation of the facility.

15 1 7 OPERABIE A system or component is operable when it is capable of performing its intended function.

, 15 1.8 CONTAINMEITT INTEGRITY I

Containment integrity is defined to exist when all of the following r 0143 Amendment No. 3

D-B conditions are satisfied: .m

1) The required non-automatic containment isolation valves and blind j flanges are closed.

2) The required automatic containment isolation valves are operable or are secured closed.

3) The Equipment Hatch is properly closed.

4) At least one door of both the Personnel Lock and Emergency Lock is properly closed.

5) The uncontrolled contain=ent leakage satisfies Technical specification 15 4.4.

15 1 9 saurDoWN

1) Cold Shutdown The reactor is in the cold shutdown ecndition when it is suberitical by atleast(later)%ok/kandTavg is less than (later)F.

2) Hot Shutdown 3 The reactor is in the hot shutdown condition when it is suberitical by atleast(later)%ok/kandTavg is greater than (later)CF. 3

)

3) Refueling Shutdown The reactor is in the refuelig shutdown condition when the primary coolant is at refueling boron concentration and T ay is below (later)OF.

A refueling shutdown refers to a scheduled reactor $hutdown to replace and/or shuffle fuel.

15 1.10 RATED POWER Rated power is defined as a steady state reactor system output of 2,650 MWt (including 2,633 MWt core power and 17 MWt reactor coolant pump power).

15 1.11 THERMAL POWER Thezzal power, the total core heat transferred from the fuel to the reactor coolant, is 2,633 MWt at rated power.

15 1.12 REAC'10R CRITICAL The reactor is considered critical for purposes of administrative control when the neutron chain reaction is self-sustaining; that is, keff 1.0.

15 1.13 POWER OPERATION The reactor is at power operation when it is critical and the neutron flux  ;

/

Amendment No. 3 15-2 0144

D-3 power range instrumentation indicates greater than 2% of rated power.

15 1.14 REFUELING OPERATION Refueling operation consists of any operation involving movement of core components when the reactor vessel head is unbolted or removed.

15 1.15 INSTRUMENT CHANEL An instru=ent channel is the combination of sensor and its power supply, vires, a=plifiers, and output devices which are connected for the purpose of measuring the value of a process variable. An instrument channel may be either analog or digital in nature.

15 1.16 PEOTECTIVE CHANNEL A protective channel is a combination of instrument channels feming a single digital output to the protective system's coincidence logic. .

15 1.17 MEASURED VALUE The measured value of a process variable is the value of the variable as it appears on the output of an instrument channel.

15 1.18 REACTOR PEOTECTIVE SYSTEM 3 i The reactor protective system is that combination of protective channels

\ and associated circuitry which foms the auto =atic system that protects the reactor by control rod trip.

15 1.19 DEGREE OF REDUNDANCY The difference between the number of operable channels and the minimum number of channels monitoring a specific parameter which when tripped will cause an automatic system trip.

15 1.20 REACIOR TRIP The de-energizing of the control rod drive mechanism which releases the con-trol rods and allows them to drop into the core.

15 1.21 INSTRUMENTATION SURVEILIANCE TEFMS

1) Channel Check Channel check is a qualitative determination of accept. ,e operability by observation of channel behavior and/or state -

during operation. This determination shall include, where feasible, comparison of the channel with other independent channels measuring the same variable. .

2) Channel Functional Test

- A channel functional test consists of injecting an internal or 0145 15-3 Amendment No. 3

D-B external test signal into the channel to verify that it is operable, incluiingalarmand/ortripinitiatingaction.

3) Channel Calibration Channel calibration consists of a functional test and the adjust-ment, if necessary, of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel asures or an accurate simulation of these values. Calibratu shall encompass the entire channel, including alarm or trip, but may exclude the sensor provided calibration data for the sensor is available.

15 1.22 EEAT BALANCE CECK A heat balance check is a comparison of the indicated neutron power and thermal power. The neutron power as indicated by the power range channels will be compared and adjusted, if necessary, to agree with the thermal power calculated from a heat balance. .

15 1.23 QUADRANT POWER TILT The difference between nuclear power in any core quadrant and the average in all quadrants.

3 15 1.24 ABNOINAL OCCURRENCE An abnormal occurrence is defined as any of the following:

1) An unplanned reactor trip.

2) Failure of the reactor protective system to trip the reactor when a " limiting safety system setting" is exceeded.

3) A shutdown or other permitted remedial measure taken whenever the requirements for " limiting conditions for operations" cannot be met.

4) Uncontrolled or unanticipated release of radioactivity from the site; overexposure of personnel; station damage or loss, as definef. in 10 CFR 20, Section 20.403 0146 i

l l 15_h I

Amendment No. 3 l

9

D-B 15 2 SAFETY LIMITS AND WETING SAFETY SYSTEM SEfrINGS 15 2.1 SAFETY GETS, REACTOR CORE TEER4AL MABGIN Applicability Applies to reactor ther=al power, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the station.

Objective To =aintain the integrity of the fuel cladding.

Specification The combination of reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figures (later). If the actual pressure-temperature point is below or to the right of the line for the specified power and flow, the safety limit is -

exceeded.

Bases The departure from nucleate boiling ratio (DNBR), although fundamental, is 3 not an observable plant variable. For this reason, limits have been placed on reactor coolant outlet temperature, reactor coolant flow, reactor system

( pressure, and reactor themal power, these being the =easurable process variables related to detemination of the DNBR. A DNBR of (later) has been chosen as a reasonable lower limit for nor=al operation and operational transients. The curves presented in Figures (later) represent the condi-tions at which a minimum DNBR of (later) is predicted by i? W-3 DNB correlation at indicated ther=al power and nomal reactor coolant flow.

In order to minimize the probability of any significant amount of fission products being released from the fuel to the reactor coolant, it is desirable to prevent clad overheating both during nor=al operation and while undergoing system transients. Clad overheating and potential failure could occur if the heat transfer mechanism at the clad surface departs from nucleate boiling. System parameters which affect departure from nucleate boiling (DNB) have been correlated with experimental data to provide a means of determining the probability of DNB occurrence. The ratio of heat flux at which DNB is expected to occur for a given set of conditions to the actual calculated heat flux experienced at a point in the core is the DNB mtio and reflects the probability that DNB will actually occur.

m

, 0147 L

15-5 Amendment No. 3

15 2.2 SAFETY IJMIT, REACTOR C00IANT SYSTEM PRESSURE Applicability ,Q)

Applies to the maximum allovable pressure in the reactor coolant g stem.

Objective To maintain the integrity of the reactor coolant system boundaries.

Specification 3

i The reactor coolant system pressure shall not exceed (later) psig with fuel assemblies in the reacter vessel.

Bases The reactor coolant system serves as a barrier which prevents release of .

reactor coolant and radionuclides contained in the reactor ecolant to the '

containment atmosphere. A limit of (later) psig((later) percent of design pressure) has been established as the maximum transient pressure allowable in the reactor coolant system under the ASME Code,Section III.

03/18 T I Amendment No. 3 15-6

D-B 15 2 3 LIMITHIG SAFETY SYSTEM SETTEIGS, PROTECTIVE INSTRUENTATION ,

(- Applicability Applies to instru=ents monitoring reactor power; reactor coolant system pressure, outlet temperature, and flow; and number of reactor coolant pumps in operation.

Ob.jective To provide automatic protective action to prevent any combination of process variables from exceeding a safety limit.

Specification The limiting safety system settings of protective instrumentation are as follows:

Reactor Trip Set Points -

1. ReactorOverpower*,%, Max. (later)

2. Reactor Overpower

• Based on Pumps in Service,%, Max.

(a} Three Pumps, Max. (later)

(b) Two Pumps in One Loop, Max. (later) 3 (c) One Pu=p in Each Loop, Max. (later)

(

3 Reactor Power *-to-Flow Ratio, Max. (later)

4. Reactor Coolant Outlet Temperature, F, Max. (later) 5 Reactor Coolant Pressua a) High Pressure, psig, Max. (later)

(b)

( Pressure-Temperature, psig, Min. (later)

(c) Low Pressure, psig, Min. (later)

6. Feactor Coolant Pumps, Simultaneous Loss, (later)

Number of

• As measured by neutron flux.

Bases (Later) h OJA9 15-7 Amendment No. 3

D-B 15 3 LIMITING CONDITIONS FOR OPERATION 15 3 1 REACTOR C00IANT SYSTEM O) 15 3 1.1 Reactor Coolant System Activity Applicability Applies to the measured maximum activity in the reactor coolant system.

Ob.jective 3

To assure that the reactor coohnt fission product activity does not exceed a level commensurate with the health and safety of the public.

Specification The total specific activity of the reactor coolant due to nuclides with .

half-livesofmorethan30minutesshallnotexceed(hter)/EpCiec-MeV whenever the average reactor coolant temperature is greater than later)0F (where E is the measured average of the beta and gamma enezgies per dis-integration in MeV).

Bases (Later)

D

1. ^

l 0g0 l'~

Amendment No. 3

D-B

( 15 3 1.2 Combined Heatup, Ccaldown and Pr*ssurization Limitations Applicability Applies to combine. **eatup, cooldown, and pressurization of the reactor coolant system.

Objective To limit reactor coolant system temperature and pressure cond.itions during heatup and cooldown in oztier to =aintain structural integrity of the reactor coolant system.

Specification

.c (Later)

(

H l

. 1 k 0151 l

15-9 Amendment No. 3

D-B 15 3 1 3 Leakage Applicability Applies to leakage from the reactor coolant system.

Ob.jective i

To specify allowable leakage limits for the reactor coolant system.

l Specification A. The reactor coolant system shall be monitored for evidence of leakage.

B. Detectable leakage from the reactor coolant system shall be 3 investigated and evaluated. If the leakage exceeds (later) gym and the source of leakage is not identified within (later) hours, ,

the reactor shall be brou6ht to the hot shutdown condition. If -

the sources of leakage have been identified and the results of the evaluations by the Shift Supervisor are that continued operation is safe, operation of the reactor with a total leakage rate not l

exceeding (later) gpn shall be pemitted.

I C. If the activity of the secondary system coolant exceeds (later) pCi l

of I-3 31/cc of coolant, the reactor shall be brought to the hot l sW' lovn condition within (later) hours. s.

)

/

Bases (Iater) l 1

i l

l l

l 1

'T.,i 1 .. -

Amendment No. 3 15-10 OSouy l

D-B 15 3 1.4 Operational Components Applicability Applies to the operating status of the reactor coolant system equipment.

Objective To specify the minimum functional capability of the reactor coolant system components which will assure the structural integrity of the system.

Specification A. Reactor Coolant Pumps

1. The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one low pressure injection pump is _a operation.

2. Reactor power shall be limited in accordance with the number of operating reactor coolant pumps as stated in Technical Specification 15 2 3 3 During reactor startup and power operation, a minimum of two 3 reactor coolant pumps shall be in operation before more than a single control rod can be removed from the core.

lv B. Steam Generator

1. One steam generator shall be capable of performing its heat transfer function whenever the reactor is critical and the reactor coolant average temperature is above (later).

C. Pressurizer and Relief Valves

1. At least one pressurizer code relief valve shall be operable when all reactor coolant system openings are closed, except for hydrostatic tests.

2. The reactor shall not be critical unless both pressurizer code relief valves are operable.

Bases A reactor coolant pump or low pressure injection pump is required to be in operation before the boron concentration is reduced by dilution with makeup water since either pump will provide mixing which will prevent sudden posi- 1 tive reactivity changes caused by dilute coolant reaching the reactor. One low pressure injection pump will circulate the equivalent of the reactor coolant system volume in one half hour or less.

Since no single control red sufficient worth to render the reactor 0153 15-11 Amend =ent No. 3

i l

D-B critical in a startup condition, a minimum of two reactor coolant pumps will be in operation should a startup accident occur. Under such conditions, the 'N flux-flow monitor and the fit'x-pump monitor will act to teminate the '$zil accident well before system damage can occur.

One pressurizer code relief valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than 3

that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat. Both pressurizer code relief valv9s are required to be in service prior to criticality to confom to the sys em design relief capabilities. The code relief valves are sized for a rod withdravs1 accident at less than (later) percent full power.

l

y ,

l l i l

l i

l 1 l

0154 Amendment No. 3 .

15-12

D-B

._ 15 3 2 CHEMICAL ADDITION SYSTEM AND MAKEUP SYSTEM

( Applicability Applies to the operational status of the chemice' addition and the makeup systems.

Objective To specify those limiting conditions for operation of the chemical addition system and the makeup system that must be met in order co assure safe reactor operation.

Specification A. When fuel is in the reactor, there shall be at least one flow path to the core for boric acid injection.

B. The reactor shall not be critical unless the following conditions are met:

1. One of the two boric acid addition tanks contains at least the equivalent of (later) ft3 of (later) ppm boron as boric acid solution, and the associated boric acid pump is operable.

2. One of the two makeup pumps is operable.

k Bases 3 The makeup system and the chemical addition system provide control for the reactor coolant system boron concentration. This is accomplished by using either one of the two boric acid pumps in the chemical addition system. The solution is added to the reactor coolant by the makeup system pumps.

A. The boric acid pu=ps can deliver boric acid from the boric acid addition tanks in the chemical addition system to the reactor coolant makeup tank.

B. The two makeup pumps take their suction directly from the makeup tank.

The quantity of boric acid in storage in either of the two sources given in the Specification is sufficient to borate the reactor coolant in order to l reach cold shutdown at any time during the core life.

The ~ hum boron addition required to attain a one percent subcritical margin in cold condition at end of the. core life is the equivalent of (later) -

ft3 of (later) ppm boron as borie acid solution (available in the chemical

~

addition system). A minimum of (later) ?t3 of (later) ppm bcron as boric acid solution in each of the boric acid addition tanks will satisfy the requirements.

Boric acid is maintained at temperatures to e.ssure solubility, and sufficient l(' mixing is provided in the makeup system to assure dilution of the concentrate, f * ,

, 15-13 0155 Amendment f No. 3

D-B 15 3 3 DERGENCY CORE COOLING SYSTEM p 44 Applicability "

Applies to the operating status of the emergency core cooling system.

Objective To define those conditions necessary to assure immediate availability of l emergency core cooling.

Specification A. The reactor shall not be :::ade critical except for low temperature physics tests, unless the following conditions are met:

1. Injection Systems (a) Borated water storage tank shall contain a minimum of >

(later) gallons. The boron concentration in the tank shall be a minimum of (later) ppm boron. l (b) Two high pressure inacation pu=ps are operable. ,

l 3

(c) Two low pressure injection pumps (decay heat pumps) are operable.

1 (d) Both decay heat removal coolers and their cooling water '

f supplies are operable (see item 3 below). i l 2. Core Flooding System (a) Two core flooding tanks each containing (later) ft3 or j

borated water at (later) psig shall be available.

I l

(b) Core flooding tank boron concentration shall not be less than (later) ppm boron.

! 3 Component Cooling Water System 1

(a) Two of the three component cooling pumps must be operable.

4. Service Water System (a) Two of the three service water pumps must be operable.

5 Valves and interlocks associated with each of the above systems <

are operable.

B. Maintenance shall be allowed during power operation on any com-ponent(s) in the high pressure and low pressure injection systems which will not remove more than one train of each system from service. At such times, the service water pump and component cool-J Amendment No. 3 15-lh OM

D-B ing water pump associated with the bus serving the unaffected train

(

shall be in service. Components shall not be re=oved from service such that the affected system train is inoperable for longer than (later) consecutive hours.

C. Prior to initiating maintenance on any of the components, the duplicate (redundant) components shall be tested to assure operability.

Bases The requirements cf Specification A assure that, before the reactor can be made critical, adequate engineered safety features are operable. Two high pressure injection pumps and two low pressure injection pumps are specified.

However, only ;ne of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident. Both core flooding tanks are required as a single core flooding tank has insufficient inventory to recover the core.

The borated water storage tank is used for two purposes:

A. As a supply of borated water for the emergency core cooling system.

B. As a supply of borated water for flooding the refueling canal dur-ing refueling operation.

3 (Later) gallons of borated water are required to supply emergency core cooling and containment vessel spray in the event of a loss-of-coolant

( accident. The borated water storage tank capacity is (later) gallons. The borated water supply is =aintained at a temperature to prevent freecing.

The boron concentration is based on the amount of boron required to maintain the core 1 percent suberitical at 700 F vithout any control rods in the core.

This concentration is (later) ppm boron while the minimum value specified in the tanks is (later) ppm boron.

When the reactor is critical, maintenance is allowed per Specification B provided requirements in Specification C are met which assure operability of the duplicate components. Operability of the specified components shall be based on the results of testing as required by Technical Specification 15 A.2.

In the event that the need for emergency core cooling should occur, function-ing of one train (one high pressure injection pump, one low pressure injection pump, and both core flooding tanks) vill protect the core, and in the event of a main coolant loop severence, limit the peak clad temperature to less than (later)0F.

Three service water pumps are provided; one is required during normal oper-ation. A single pump will also provide the essential cooling requirements following a loss-of-coolant accident.

Three component cooling pumps are provided; one is required during normal operation. A single pump will also provide the essential cooling require-(- ments following a loss-of-coolant accident.

g 0157 15-15 Amendment No. 3

D-B 15.3.h TURBINE CYCLE EQUIPMENT Applicability Applies to the operating status of turbine cycle components for removal of reactor decay heat.

Objective To specify minimum conditions of the turbine cycle equipment necessary to assure capability to remove decay heat from the reactor core.

Specification The reactor coolant shall not be heated above (later)OF unless the following conditions are =et:

A. Capability to supply feedvater to one of the two steam generators at a rate corresponding to a decay heat load of (later) percent .

full reactor power. The feedvater may be supplied by:

1. One of the two main feed pu=ps/ booster pu=ps taking suction from one of the two deaerator storage tanks.

2. One of the two auxiliary feed pu=ps with suction available from 3 ene of the deaerator storage tanks and from the service water system.

B. Two of the three service water pu=ps are operable.

C. Rated relief capacity in the steam system safety ;alves is available.

D. The main steam at=ospherie du=p and turbine bypass system shall have a capacity equivalent to (later) percent rated reactor power.

Bases The feedvater system and the turbine bypass system are nored.ly used for decay heat re= oval and cooldown above (later)0F.

If neither main feed pu=p is available, feedvater can be supplied to the steam generators by the auxiliary feed pu=ps.

In the event of complete loss of electrical power, feedvater is supplied by two turbine driven auxiliary feed pu=ps which take suction from the deaerator 8l storage tanks, condensate storage tanks, fire protection system and service water system. Either auxiliary feed pump is capable of supplying the necessary feedvater for decay heat removal. Main steam relief would be through the atmospheric dump valves.

The major backup water supply to the auxiliary feed pu=ps is from the service water system having three service water pumps provided with suction from Lake ,j Erie, and having a backup power source from the diesel e=ergency generators.

Each service water pump has more than adequate decay heat removal capacity.

Amendment No. 8 01.58 15-16

D-B f

15 3 5 c0ttrADDENT CCOLING AND SPRAY SYSTEMS

\

Applicability Applies to the operating status of the contain=cnt cooling aad spray systems.

Ob.jective To define those conditions necessary to assure i= mediate availabilicy of the containment cooling and spray systems.

Specification A. The reactor shall not be made critical except for lov temperature physics tests, unless the following conditions are met:

1. One of the following combinations of subsystem components must be operable.  :

(a) Two containment spray pumps and their associated spray no::le headers, and two of the three containment cooling fans and associated cooling units.

(b) One containment spray pump aligned to an operable spray nozzle header and three containment cooling fans and 3 associated cooling units.

2. Two of the three service water pumps must be operable.

3 Borated water storage tank shall contain a minimum of (later) gallons. The boron concentration in the tank shall be a minimum of (later) ppm boron.

4. Two shield building and penetration room fan-filter systems must be operable.

5 All valves and piping associated with the above components are operable .

B. During power operation, maintenance shall be allowed on containment spray and cooling components provided that:

1. No less than* one of the following combinations of equipment remain operable:

(a) Two containment spray pumps. )

1 (b) Two containment air cooling units. )

< l (c) One containment spray pump and one containment air cooling l unit. )

.3 1 i,

~

\

2. The above condition does not persist for longer than (later)  ;

consecutive hours. j 0159 g i 15-17 Amendment No. 3

D-3 C. During power operation, maintenance shall be allowed on one shield building and penetration room fan-filter system, provided it is ,N inoperable for no longer than (later) hours. /

D. Prior to initiating maintenance on any of the components, the duplicate (redundant) components shall be tested to assure oper-ability.

Bases The post accident containment cool 1D6 may be accomplished by two cooling units, by two spray units, or by a combination of one cooling unit and one spray unit. The specified requirements assure that the required post acci-dent components are available.

Three service water pumps are provided; one is required during nomal operation. A single pump will also provide the essential cooling require-ments following a loss-of-coolant accident.

(Later) gallons of borated water are required to supply emergency core cooling and containment spray in the event of a loss-of-coolant accident.

The borated water storage tank capacity is (later) gallona based on refueling 3 volume and IOCA requirements. The borated water supply is maintained at a temperature to prevent freezing. The boron concentration is based on the amount of boron required to maintain the core 1 percent subcritical at 700F vithout any control rods in the core. This concentration is (later) ppm boron while the minimum value specified in the tanks is (later) ppm boron. .

The shield building and penetratinn room fan-filter system consists of two J redundant full capacity fan-filter assemblies each capable of performing the system function following a IOCA.

When the reactor is critical, maintenance is allowed per Specification B and C provided requirements in Specification D are met which assure operability of the duplicate components. Operability of the specified components shall be based on the results of testing as required by Technical Specification 15 A 3 Whenever containment cooling and spray systems components are inoperable, the service water pump and component cooling water pump associated elec-trically with the unaffected components shall be in service.

0160 j Amendment No. 3 ,

15-18

D-B f

~

15 3 6 INSTRUENTATION AND CONTROL

'~

15 3.6.1 Operational Safety Instrumentation Applicability Applies to station instrumentation and control systems related to nuclear safety.

Ob.jective To delineate the conditions of the station instrumentation and safety cir-cuits necessary to assure reactor safety.

Specification This specification relates to the items listed in Table (later).

A. Startup is not pemitted unless the requirements of Table (later) are met.

B. For on-line testing or in the event of a protective instrument or channel failure, stat'on operation at rated power shall be permitted to continue in accordance with Table (later). 3 C. In the event the number of protective channels in service falls

( below the limit given under Table (later), operation shall be limited according to the requirement shown in Table (later).

Bases Startup is not pemitted unless at least three each of the following reactor protection system channels are functioning correctly: four power range neutron channels, four reactor coolant temperature channels, four reactor coolant flow channels, four reactor coolant pressure channels, four reactor coolant pump monitor channels, and four reactor protection channels. The engineered safety features system must have at least three each of the following channels functioning correctly prior to startup: four reactor coolant pressure channels, four high containment pressure channels, four high high containment pressure channels, four borated water storage tank level channels, and four high radiation channels.

Opeistion at rated power is pe=1tted as long as the systems have at least the redundancy requirements of Table (later). This is in agreement with redundancy and single failure criteria of .u:.a. 279 as described in PSAR Section 7, and does not ecmpromise safety since such operation has been pro- "

vided for in the design.

(

.(, 0161 15-19 Amndment No. 3

D-B 15 3 6.2 Control Rod croup Limits i

Applicability ,

Applies to the insertion and withdrawal of the control rod groups during power operation.

Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core suberiticality after a reactor trip.

Specification A. Whenever the reactor is critical, except for physics tests and control rod exercises, the shutdown control rods shall be fully withdrawn. ,

B. Except for below 15% of rated power and/or determination of "just critical" rod positions, operation of the control rod groups shall be in accordance with specifications C, D, and E below.

C. The position of the Axial Power Shaping Rod Assemblies (APSPA) vill 3 be coordinated with the position of the regulating group to main-tain the power level in the top and bottom of the core in a speci-fled ratio.

D. Control rod group withdrawal overlap shall not exceed 25% between two sequential groups.

E. If a dropped rod condition occurs (individual rod deviation from its group reference position by more than (later) inches), the unit load demand will be run back to (later)% of rated load.

F. Eot shutdown margin shall be not less than 1% ok/k with the highest worth control rod fully withdrawn from the core.

G. No more than one control rod may be inoperable as defined in Technical Specification 15 4.9 H. If a control rod cannot be moved by the drive mechanism, the shut-down margin shall be adjudted to maintain the necessary 1% Ak/k shutdown margin as required by Item F of this specification.

Bases A. The reactivity control concept is that short-tem reactivity changes caused by changes in reactor power, xenon poisoning, and 0 to 15 percent power moderator temperature changes are compensated by control rod motion. Long-tem reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (cold to hot critical) are compensated by Amendment No. 3 15-20  !

L

D-B .

changes in the soluble boron concentration. During power opera-( tion, the shutdown groups are fully withdrawn and control of reactor power is by the control groups. A reactor trip occurring during power operation vill put the reactor into the hot shutdown condition.

B. The axial position of the APSRA vill depend on the power condition of the core as well as the maneuvering sequence and the resulting xenon condition. APSEA vill be moved to obtain a power balance between the top and bottom of the core as determined by in-core instru=entation.

C. The 25% overlap between successive control rod groups is allowed since the worth of the rod is lower at the upper and lower part of the stroke.

D. If a control rod drop occurs during power operation, a rapid de- 3 crease in reactor ther=al power would occur acccmpanied by a ,

decrease in core average coolant temperature. The movement of the control rod groups could result in a distorted power distribution and subsequent return to full power could lead to localized power densities and heat fluxes in excess of design limitations. The action provided by the system which detects a dropped rod results in the reactor ther=al power assuming a lover value to match the load demand and assures that the DNB ratio will not be less than (later) and the reactor coolant system pressure vill not exceed code limits.

(

E. The minimum available rod worth provides for achieving hot shut-down by reactor trip at any time assuming the highest worth control rod remains in the fully withdrawn position.

1 OM3 k

21 Amendment No. 3

D-B 1537 CottrAINMENT Applicability Applies to the integrity of reactor contairment.

Ob,jective To define the operating status of the reactor containment for station opera-tion.

Specification A. Contairment Integrity (As defined in Section 15 1)

1. Containment integrity shall not be violated unless the reactor is in the cold shutdown condition.

2. Containment integrity shall be maintained when the reactor  :

coolant system is open to the containment atmosphere while a 3

• • " *8 "" ' # " " '" "* " ** 8' 3 Positive reactivity changes shall not be made by rod drive motion or boron dilution whenever the containment integrity does not exist.

B. Internal Pressure The reactor shall not be critical if the containment internal "

')

pressure exceeds (later) psig or an internal vacuum exceeds (later) in. Hg.

Bases The containment is designed so that should the =aximum hypothetical accident occur when containment integrity is =aintained, the containment design, along with the engineered safety features provided, vill prevent exceeding the guidelines of 10 CFR 100.

H.S y 15-22 Amendment No. 3

D-B

,- 15 3.8 AUXILIARY EIECTRICAL SUPPLY t

('

Applicability Applies to the availability of electrical power for operation of the station components.

Objective To define those conditions of electrical power availability necessar/ to assure:

(a) Safe station operation.

(b) Continuous availability of engineered safety features.

Specification -

A. The reactor shall not be made critical without: .

1. One 345 KV - 13 8 Ky startup transromer in sertice.

2. One 13 8 KV bus energized.

3. 4160-volt engineered safety features buses cl and D1 energized. 3

4. 480-volt engineered safety features El and F1 energized.

('

5 (IAter) 120 volt ac Vital Instrumentation Buses energized.

6. Two emergency diesel generators operable with on site supply of (later) gallons of fuel available.

7 Both250/125 volt station batteries and the de systems operable, and at least two battery chargers on each de bus operable .

8. Both switchyard batteries and de systems with one battery charger on each de bus operable.

B. During reactor operation the above requirements may be modified as follows:

1. Power operation may continue with both startup transfomers out of service provided the following conditions are met:

(a) the failure shall be reported to the AEC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with an outline of the plans for prompt restoration of -

offsite power and the additional precautions to be taken while the transfomers are out of service.

(b) the operability of both diesel generators is demonstrated i= mediately.

03.65 15-23 Amendment No. 3

D-B Under conditions of fulfillment of (b) and non-fulfillment of (a),

continued power operation shall not extend beyond (later) hours. Q Non-fulfill =ent of (b) shall be deemed sufficient cause for i=med- )

inte reactor shutdown.

2. 4160 V engineered safety features bus Cl or D1 may be inoperable for up to (later) hours provided the operability of the diesel genemtor associated with the operable bus is demonstrated i= mediately, and there are no inoperable engineered safety features components associated with the operable bus.

3 480 V engineered safety features El or F1 may be inoperable for up to (later) hours provided there are no inoperable safety features components associated with the operable bus.

4. One of the four Vital Instrumentation Buses may be inoperable ,

for (later) hours provided the reactor protection and engineered safety features syste=s supplied by the remaining ,

three buses are all operabic. -

5 Pov a operation =ay continue if one diesel generator is out of service provided the following conditions are met:

(a) the re=aining diesel generator and its associated sequence 3 loader is tested dail not less than (later)$y by loading full the diesel generator to load rating (b) either startup transformer is operable ')

J (c) such operation is not in excess of (later) days (+otal for both diesels) during any month l

l 6. One of the 250/125 volt station batteries may be inoperable for (later) hours, providing all three battery chargers on the l affected bus are in operation.

7 One of the switchyard batteries may be inoperable for (later) l hours provided the battery charger for the inoperable battery l and the backup charger are both operable.

i Bases The electrical system is arranged so that no single contingency can de-activate sufficient engineered safety features components to jeopardi::e station safety. The 480 V equipment is arranged on six buses. Two separate double-ended unit substations, each with a single bus, provide for the supply of power to 480 volt engineered safety features and emergency equip- J ment. Power for each of these two buses is normally supplied from the correspondin6 substation transfomer fed from one of the 4160 volt engineered safety features buses. The second transformer of each substation is cross-connected to the other kl60 volt engineered safety features bus through a nomally open interlocked breaker to prevent undesired paralleling of sources. The 4160 V equipment is arranged on five buses. Three 4160 volt T Amendment No. 3 15-24

D-B buses are emergency (engineered safety features) buses, which supply equip-ment essential for the shutdown of the station-The nomal source of auxiliary power, with the station at power, is from the station auxiliary transfomer being ad from the =ain generator, with standby power from two startup trans-formera and emergency power from either one of two emergency diesel generators. The design of the electrical system has been carried out with reliability as a prime consideration. Station power is provided from five independent sources. The outage of any three sources vill not cause loss of service to the station power supply of 4160 V or below, down to and including 125 V de supply.

There are two emergency power sources on site which do not require outside power or use of a startup transformer. Upon loss of normal and standby power sources, the 4160 V engineered safety featums buses are energized from the emergency diesel generators. The two engineered safety features distribution systems are connected to different 4160 volt emergency buses and they are redundant in that completion of the starting and loading sequence of one of the diesels is adequate to satisfy minimum engineered safety features .

requirements.

The total of all loads required for opemtion in the LOCA condition is below 3 the capacity of one diesel generator. Spare capacity is available, with either diesel generator, for supplying vital loads not required for IOCA operation. The minimum diesel fuel oil inventory 3.t all times is maintained to assure the operation of both diesels etrrying design load of all engi-neered safety features equipment for at least (later) hours. Co-arcial oil

, supplies and trucking facilities are available to assure deliveries within t (later) hours.

The station can be safely shut down without the use of offsite power since all vital loads (safety systems, instruments, etc.) can be supplied from the emergency diesel generator.s:

Four station battery chargers and two switchyard battery chargers shall be in service so that the batteries vill always be at full charge. This en-sures that adequate de power will be available for emergency uses.

Y 0167 1

i 15-25 Amendment No. 3

r D-B 15 99 EEFUELING Applicability

(

Applies to operating limitations during refueling operations.

Ob,jective To ensure that no incident could occur during refueling operations that would affect public health and safety.

Specification A. Radiation levels in the containment refueling area and spent fuel ctorage area shall be monitored.

B. Three of the four high radiation channels of the containment iso-lation system must be functioning correctly, and the shield build- ,

ing and penetration room ventilation and filtration system must be operable.

C. Core suberitical neutron flux level shall be continuously monitored by at least two neutron flux monitors, each with continuous indication available, whenever core geometry or boren concentration 3

are being changed. At other times when fuel is in the reactor, at least one neutron flux level monitor shall be in service.

m D. At least one decay heat pump and decay heat removal cooler shall be operable.

j E. Euring reactor vessel head renoval and while loading and unloading fuel from the reactor, a minimum boron concentration of (later) ppm shall be =aintained in the pri=ary coolant system.

F. Direct communication between the control room and the refueling personnel in the containment shall be available whenever changes in core geometry are taking place.

G. If any of the specified limiting conditions for refueling are not met, movement of fuel into the reactor core shall cease, action shall be initiated to correct the vio Mted conditions so that t h specified limits are met, and no operations which =ay increase the reactivity of the core shall be made.

Bases e

Detailed written procedures will be available for use by refueling per-sonnel. These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9 6 of the PSAR, incorporat-ing built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to pub.lic health and safety.

O M /3 Amendment No. 3 15-26

D-B If no change is being made in core geometry one neutron flux monitor is h sufficient. This permits maintenance on the instrumentation. Continuous monitoring of radiation levels and neutron flux provides i= mediate indica-tion of an unsafe condition.

The decay heat pump is used to maintain a uniform boron concentratica. The boron concentration indicated in Specification E will keep the core sub-critical, even with all control rods withdrawn. Although '.his concentration is sufficient to maintain the core heff = 0 99 if all the c rtrol rods were 3 removed from the, core, only a few control rods vill be removed at any one time during fuel shuffling and replacement. The keff with all rods in the core and refueling boron concentration is approximately (later).

Specification F allows the control room operator to inform the containment personnel of any impending unsafe condition detected from the main control board indicators during fuel movement. .

o O

(

l i

l 0.1.69 1

W l

i -

15-27 Amendment No. 3

D-B 15.3.10 WASTE MATERIALS Pl: LEASE Aeplicability, Applies to the controlled release of radioactive liquids and gases, and con-tainerized wastes from the station.

Objective To ensure that radioactive wastes are discharged and/or shipped from the station in a controlled manner in conformance with 10 CFR 20.

Specification A. Liquid Wastes (Later)

B. Gaseous Wastes (Later)

C. Containerized Wastes (Later)

Bases

.~

A. Liquid Wastes (Later)

3. Gaseous Wastes (Later)

C. Containerized Wastes (Later) e a

Amndment No. 3 15-28 0.1.7()

D-B 15 4 SURVEILIANCE STANDARDS I Unless otherwise specified, surveillance intervals may be adjusted plus or minus 25% to accomodate normal test schedules.

15 4.1 OPERATIONAL S/LW ITEMS Applicability Applies to items directly related to safety limits and limiting conditions for operation.

l Objective 3

To specify the minimum frequency and type of surveillance to be applied to station equipnent and conditions.

Specification A. Calibration, testing, and checking of analog channels and testing of logic channels shall be performed as specified in Table (later).

B. Equipment and sampling tests shall be conducted as specified in Table (later).

Basis

( (Iater) l l

d 1.. 03.71 l

!j,- .

15-29 Amendment No. 3

D-B 15.h.2 EMERGENCY CORE COOLING SYSTEM PERIODIC TESTING Applicability

)

Applies to periodic testing requirement for emergency core cooling systems.

Objective To verify that the emergency core cooling systems are operable.

Specifi . tion A. System Tests

1. High Pressure Injection System (a) At intervals not to exceed the normal interval between re-fueling, a system test shall be conducted to demonstrate that the system is operable. The test vill be performed -

in accordance with the procedure summarized below:

(1) With the high pressure injection pu=p motor breakers in the test position, a test signal vill be applied to demonstrate actuation of the high pressure injec-tion system for emergency core cooling operation.

3 (b) The test will be considered satisfactory if control board _s indication and/or visual observations verify that all com- )

ponents have responded properly to the actuation signal. _s'

2. Low Pressure Injection System (a) At intervals not to exceed the normal interval between refueling, a system test shall be conducted to demonstrate that the system is operable. The test shall be performed in accordance with the procedure summarized below:

(1) A test signal vill be applied to demonstrate actuation of the low pressure injection system for emergency core cooling operation.

(2) Verification of the engineered safety features function of the following syste=s shall be made to demonstrate their operability:

a. Component cooling vater system which supplies cooling water to the decay heat coolers.

3

b. Service water system which supplies cooling water to the component cooling water system.

(b) The test vill be considered satisfactory if control board indication and/or visual observations verify that all com- x

, ponents have responded properly to the actuation signal. ,,

Amendment No. 3 15- 30

D-B 3 Core Flooding System (a) At intervals not to exceed the normal interval between re-fueling, a system test shall be conducted to demonstrate proper operation of the system. The test shall be performea in accordance with the procedure summarized below:

(1) Verification shall be made that the check valves in the core flooding tank discharge lines function properly.

(b) The test will be considered satisfactory if control board indications and/or visual observations verify that~all com-ponents have operated properly.

B. Co=ponent Tests

1. Pu=us ,

(a) The high pressure injection, decay heat, co=ponent cooling, and service water pumps shall be started at intervals not to exceed (later) months to verify that they are operable.

(b) Acceptable performance of each pu=p will be indicated if 3

the pu=p starts, reaches rated shutoff head or recircula-tion flow head, and operates for at least five minutes.

/

i 2. Valves - Motor Operated (a) At intervals not to exceed (later) conths, each engineered safety features valve associated with emergency core cooling in the component cooling and service water syste=s shall be tested to verify operability.

(b) The acceptable performance of each motor operated valve vill be that motion is indicated or observed upon actuation by appropriate signals.

Bases The emergency core cooling systems are the principal station safety features.

They provide the =eans to limit core damage in the event of a loss of coolant accident. Pre-operational performance tests of the c rponents are performed in the manufacturer's shop. Initial systems tests demonstrate proper dynamic functioning of the systems. Thereafter, periodic tests demonstrate that all components are functioning properly. The tests specified above vill demon-

strate that the core cooling systems , which are .not normally and routinely .

! operated, vill operate properly.

I

~

1l 0J.73 15-31 Amend =ent No. 3

D-B 15.h.3 CONTAINMENT COOLING SYSTEMS -

Applicability ,

Applies to testing of the containment cooling syste=s.

Objective To verify that the containment cooling systems are operable.

Specification A. System Tests

1. Containment Spray System (a) At intervals not to exceed the normal interval between re-fueling, a test signal vill be applied to demonstrate actuation of the containment spray system (with the spray header inlet valves closed to prevent water entering spray nozzles).

(b) At least every (later) years, station compressed air vill be introduced into the spray headers to verify the avail-3 ability of the headers and spray nozzles.

(c) The tests vill be considered satisfactory if control board -

indications and/or visual observations verify that all com-ponents have responded properly to the actuation signal or ')

test condition.

2. Contain=ent Air Cooling System (a) At intervals not to exceed the normal interval between re-fueling, a system test shall be conducted to demonstrate proper operation of the system. The test shall be performed in accordance with the procedure su=marized below:

(1) A test signal vill be applied to actuate the contain-ment air cooling system for containment air cooling operation.

(2) Verification of the engineered safety features function of the service water system to supply coolant to the containment air coolers shall be made to demonstrate operability of the coolers.

(b) The test vill be considered satisfactory if control board indication and/or visual observations verify that all com-ponents have responded properly to the actuation signal.

T v'

M O.J.74 Amendment No. 3 15-32

D-B B. Co=ponent Tests .

k 1. Pu=es (a) The contain=ent spray pumps and service water pumps shall be started at intervals not to exceed (later) =onths to verify that they are operable.

(b ,' Acceptable perfor=ance of each pump vill be indicated if the pu=p starts , reaches rated shutoff head or recirculation flow head, and operates for at least five minutes.

2. Fans (a) The contain=ent air cooler fans shall be started at intervals not to exceed (later) conths to verify that they are oper-able .

(b) Acceptable performance of each fan vill be indicated if the fan starts, develops the specified head, and operates at half speed for at least five minutes. 3 3 valves - Motor Operated (a) At intervals not to exceed (later) months each engineered safety features valve in the containment spray and con-tain=ent air cooling system and each engineered safety

( features valve associated with containment cooling in the service water system shall be tested to verify that it is operable .

(b) The acceptable perfor=ance of each motor operated valve vill be that motion is indicated or observed upon actuation by appropriate signals.

Bases The containment cooling systems are the principal contain=ent safety features.

They provide the means of assuring containment integrity in the event of a loss of coolant accident. Preoperational perfor=ance tests of the components are performed in the manufacturer's shop. Initial systems tests and periodic tests demonstrate that all components are functioning properly. The tests specified above de=onstrate that the contain=ent cooling systems, which do not nor-ally and routinely operate in the LOCA mode, vill operate when needed.

03.75 i 4 15-33 A=endment No. 3

D-B 15.h.h CONTAINMENT TESTING Applicability Applies to containment leakage.

Ob.iective To verify that potential leakage from the containment is maintained within specified limits.

Specification A. Containment Vessel Integrated Leakage Rate Test,

1. Tests' (Later)

2. Acceptance Criteria (Later) 3 Corrective Action (Later) 3 4. Test Frecuency (Later)

B. Shield Building Leak Testing ,

1. Tests (Later)

2. Acceptance Criteria (Later)

3. Corrective Action (Later)

h. Test Frecuency (Later)

C. Local Leak Detection Tests

1. Tests (a) Local leak rate tests shall be performed at a pressure of not less than (later) psig.

(b) Acceptable methods of testing are halogen gas detection, soap bubble, pressure decay, ultrasonics, or equivalent.

(c) The local leak rate shall be measured for each of the following components:

g

D-B (1) Containment vessel penetrations that employ resilient seal gaskets, sealant compounds, or bellows.

(2) Personnel and emergency locks, and equipment hatch seals.

(3) Fuel transfer tubes.

(h) Isolation valves on the testable lines of fluid systems penetrating the containment.

(5) Other containment components, which require leak repair in order to meet the acceptance criterion for any integrated leak rate test.

2. Acceptance Criterion The total leakage from all penetrations and isolation valves shall not exceed (later). -

3. Corrective Action (a) If at any time it is determined that (later) is exceeded, 3 repairs shall be initiated immediately.

(b) If repairs are not completed and confor=ance to the above Acceptance Criterion is not demonstrated within (later)

( hours, the reactor shall be shut down and depressurized until repairs are effected and the local leakage meets this Acceptance Criterion.

h. Test Frecuency (Later)

D. Containment System Ventilation and Filtration Test

1. Tests (Later)

2. Acceutance Criteria (Later)

3. Corrective Action (Later) *

h. Test Frequency (Later) 0177

\

15-35 Amendment No. 3

D-B E. Report on Test Results Each integrated leak rate test will be the subject of a summary '&d technical report, including method used, test procedure, test 3 results, and analysis used to verify that specified leakage rates were not exceeded.

Bases (Later) 4 1

0178 .

v :

. l Amendment No. 3 15-36 l

D-B 15.4.5 EMERGENCY POWER SYSTEM PERIODIC TESTING

- Applicability Applies to periodic testing and surveillance requirements of the emergency power system.

Objective To verify that the emergency power diesel generator system and station batteries will respond promptly and properly when required.

Specification A. Diesel Generators

1. Every (later) months each diesel generator shall be manually started and synchronized with normal power sources and loaded to (later) rating. The signal initiating the start shall be -

varied from one test to another to verify that all starting circuits are operable.

2. During each refueling shutdown, each diesel generator shall be started automatically by a simulated loss of all normal AC power supplies together with a simulated safety injection signal, and loaded sequentially with vital loads. Proper oper-ations shall be verified by bus load shedding and automatic

( starting of selected motors and equipment to establish that 3 rustoration with emergency power has been accomplished within (later) seconds.

3 Each diesel generator shall be given a thorough inspection at least annually following the manufacturer's recommendations for t,his class of standby service.

h. The diese] oil transfer pu=ps shall be verified to be operable monthly.

5 The above tests will be considered satisfactory if all applicable equipment operates as designed.

6. Diesel generator electric loads shall not be increased beyond the continuous rating of (later) KW.

B. Diesel Fuel Tanks A minimum oil storage shall be as stated in Technical Specification 15.3.8. -

C. Station Batteries

1. Every (later) months the voltage of each cell (to the nearest 0.01 volt), the specific gravity and temperature of a pilot cell in each battery shall be measured and recorded.

[^ g 15-37 ,.

A=end=ent No. 3

D-B

2. Every (later) months the specific gravity of each cell, the temperature reading of every fifth cell, the height of electrolyte, and the amount of water added shall be measured and recorded. n,J 3 Every (later) months the battery shall be subjected to a load test. The battery voltage as a function of time shall be monitored to establish that the battery perfor=s as expected.

during heavy discharge, and that all electrical connectior:, are tight.

h. At each time data is recorded, new data shall be compared with old to detect signs of abuse or deterioration.

Bases The tests specified are designed to demonstrate that the diesel generators will provide power for operation of equipment. They also assure that the emergency generator system controls and the control systems for the engineered safety features equipment will function automatically in the event of a loss of nor=al power.

The testing frequency specified is often enough to identif. and correct 3

any mechanical or electrical deficiency before it can result in a system failure. The fuel supply, and starting circuits and controls, are continuously monitored and alarmed for abnormal conditions.

The specified fuel supply will ensure power requirements for at least (later). s Station batteries will deteriorate with time, but precipitous failure is _)

extremely unlikely. The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell beccming unserviceable long before it fails.

As a check upon the integrity of all 'oattery cells, the battery should be loaded rather heavily and the voltsge monitored as a functior. of time.

If a cell has deteriorated or if a connection is loose, the voltage under load will drop excessively indicating need for replacement or maintenance.

i j

l 1

i i

l 1

l I

l J

01.80  ;

Amendment No. 3 15-38 l

D-B

. 15.h.6 AUXILIARY FEED PCMP PERIODIC TESTING C

( Applicability Applies to the periodic testing of the turbine-driven auxiliary feedvater Pu=ps.

Objective To verify that the auxiliary feed pu=ps are operable.

Specification A. At least every (later) =enths the urbine-driven auxiliary feed pu=ps vill be operated on recirculation to the condensate storage tanks.

B. At least every (later) =onths, the auxiliary feed pu=ps discharge valves shall be exercised. ~

C. The test vill be considered satisfactory if control board indication 3 and/or visual observations verify that all components have operated properly.

Bases The specified testing of the auxiliary feed pumps by recirculation to the

(- condensate storage tanks vill verify their operability. The capacity of any one of the two auxiliary feed pu=ps is sufficient to meet decay heat removal requirements.

Proper functioning of the turbine ad=ission valve and the operation of the pu=ps vill de=onstrate the integrity of the system. Verification of correct operation vill be made from instrumentation within the control room and direct visual observation of the pu=ps.

0f.h$

15- 39 A=endment No. 3

, D-B 15.h.7 REACTOR COOLANT SYSTEM IN-SERVICE INSPECTION m

Applicability 1 v'}

Applies to in-service inspection of the Reactor Coolant System.

Objective To define the ins 7ections required to assure the continuing integrity of the Reactor Coolant S stem.

Specification (Later)

Bases I

(Later)

I

]

i i

O i

03.82

~

Amendment No.-3 15-40

D-B 15.h.8 REACTOR COOLANT SYSTEM INTEGRITY TESTS Aeplicability Applies to test requirements for Reactor Coolant System integrity.

Objective To specify tests for Reactor Coolant System integrity after the system is closed following normal opening, modification, or repair.

Specification A. When the Reactor Coolant System is closed after having been opened, the system vill be leak tested at not less than (later) psig at NDTT requirements for te=perature prior to the reactor being =ade critical,

~

B. When Reactor Coolant System modifications or repairs have been made vhich involve new strength velds on components with diameters greater than (later) in. , the new welds vill receive both a surface 3

and 100 percent volumetric examination.

C. When Reactor Coolant System modifications or repairs have been made which involve new strength velds on components with diameters of (later) in, or less , the new welds vill receive a surface examina-tion.

N Bases For normal opening, the integrity of the reactor coolant system, in ter=s of strength, is unchanged. If the system does not leak at (later) psig (operating pressure + (later) psi; t (later) psi is normal cystem pressure fluctuation),

it will be leak tight during normal operation.

For repairs on components with dia=eters greater than (later) in., the thorough non-destructive testing indicated gives a very high degree of con-fidence in the integrity of the reactor coolant system, and will detect any significant defects in and near the new welds.

Repairs on components with diameters of (later) in. or less are relatively minor in comparison and the surface examination assures a similar standard of integrity. In all cases, the leak test will ensure leak tightness during nor=al operation.

M 01.83 15-h1 Amendment No. 3

l D-B 15.h.9 REACTOR CONTROL ROD SYSTEM Applicability f Applies to the surveillance of the reactor control rod system.

Objective To assure operability of the reactor control rod system.

Specification l

A. The control rod trip insertion time shall be checked following each l refueling outage. The maximum trip insertion time for an operable control rod drive mechanism from the fully withdrawn position to (later) insertion ((later) inches travel) shall not exceed (later) l seconds at reactor coolant full flow conditions. For no flow con-ditions thi's trip insertion time shall not exceed (latcr) seconds.

B. Each control rod drive mechanism shall be exer'ised by a movement of approxi=ately two (2) inches of travel each week. This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions.

C. If a control rod is misaligned with its group refe rence position by

= ore than (later) inches , the red shall be considered inopet able.

3 D. If a control rod cannot be exercised or the trip insertion time in A above is not met, the rod shall be considered to be inoperable. /

Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has completed (later) inches of travel from the fully withdrawn position. The specified trip time is based upon the safety analysis in PSAR Section 14.

Exercising the control rod drive mechanis=s in the prescribed manner provides assurance of reliability of the mechanisms.

The drive mechanism is provided with two position indication syste=s; one, an absolute position indicator and the other, a relative position indicator.

The absolute system provides a readout of the direct position of the control rod in its travel. .The relative system consists of a small pulse-stepping motor driving a potentiometer. Additional switches also provide infor=ation as to the rod extre=e positions. The relative position indicator, absolute position indicator, or limit switches can bs used to determine rod position.

A rod is considered inoperable if it canno' be exercised, the trip insertion time is greater than the specified allowable time, or the rod deviates from its group reference position by more than (later) inches. Conditions for operation with an inoperable rod are specified in Technical Specification 15.3.6.2. .

J 01.84 Amendment No. 3 15 h2

l. -

r D-B 15.h.10 REACTIVITY ANOMALIES Applicability Applies to potential reactivity anomalies.

Objective To require the evaluation of reactivity anomalies of a specified magnitude occurring during reactor operation.

Specification Following the normalization of the calculated boron concentration versus burn-up curve to actual core conditions, the actual concentration of boron in the primary coolant will be periodically compared with the predicted value.

If the difference between the observed and predicted steady-state concentra-tions reaches the equivalent of (later)% in reactivity, an evaluation as to the cause of the discrepancy shall be made and reported to the Atomic Energy Com=ission within (later) days.

Bases The predicted relation between fuel burnup and the primary coolant boron concentration, necessary to maintain adequate control characteristics, vill (s be nor=alized to actual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted carve is adjusted to this point.

As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burnup and reactivity is co= pared with that predicted. This process of nor=alization should be completed after about 10 percent of the total core burnup, at which point actual boron concentration can be co= pared with prediction, and the reactivity status of the core can be continuously evaluated.

Any reactivity anomaly greater than (later)% vould be unexpected, and its occurrence would be thoroughly investigated and evaluated. The value of (later)% is considered a safe limit since a shutdown margin of at least (later)% vith the most reactive rod in the fully withdrawn position is always maintained.

Y 0185

(

15-h3 A=end=ent No. 3

a D-B

15.k.11 ENVIRONMENTAL RADIATION SURVEY Applicability

-A' i

i Applies to routine testing of station environs.

Objective To establish a sampling schedule which will assure cognizance of changes in 3 radioactivity in the environs.

Specification i

3 Environmental samples shall be collected and analyzed according to Table (later).

i Bases l (Later) ,

3, s'

3 03.86 Amendment No. 3~ 15-kk l..

l L.

D-B 15 5 LESIGN FEATUES 15 5 1 SITE Applicability Applies to the location and extent of the reactor site.

Objective To define the location and the size of site area as pertains to nuclear safety.

Specification The Davis-Besse station site is located along the south shcre of Lake Erie 3 in Ottawa County, Ohio, approximately six miles northeast of Oak Harbor, Ohio, and 21 miles east of Toledo, Ohio. The site provides a minimum exclusion radius of approximately 2,400 feet.

Reference PSAR Section 2.2 b

v.

j l

C osm t

15-h5 Amendment No. 3

D-B 15 5 2 CONTARDENT Applicability Applies to those design features of the reactor containment relating to operational and public safety.

Ob.jective To define the significant design features of the reactor containment.

Specification A. Structures The reactor containment consists of a steel containment vessel within a reinforced concrete shield building; both independent structures are supported by a common concrete foundation.

1. Shield Building The shield building surrounds the containment vessel and is 3 designed to provide biological shielding from hypothetical accident conditions, biological shielding during nor=al operation, environ-mental protection for the containment vessel for adverse atmospheric conditions and external missiles, and a means for collection and filtration of fission product leakage from the contaic=ent vessel following a hypothetical accident.

The shield building is a reinforced concrete structure of right cylinder configuration with a hemispherical dome roof. There are no structural ties between the containment vessel and the shield building above the foundation slab. An annular space is provided between the steel containment vessel and the interior face of the concrete shield building of approximately 4 5 feet to permit periodic visual inspection of the containment vessel. The volume contained within this annulus is approximately 465,000 cu. ft.

The shield building is designed so hat following a loss of coolant accident, a slight negative pressure, between -1/2 and 1/2 inches vg, can be maintained in the shield building.

2. Containment Vessel The containment vessel is a lov leakage cylindrical steel pressure vessel with hemispherical dome and ellipsoidal bottcm which com- l pletely encloses the reactor and its associated primary coolant 5 system, and is designed to withstand a postulated loss-of-coolant accidentsand to confine a postulated release of radioactive material. j The containment vessel and shield building is supported on a concrete foundation on competent rock structure.

0188 l Amendment No. 3 15 h6

D-B The containment vessel will be designed in accordance with the ASbE Boiler and Pressure Vessel Code,Section III, Class B. The "=aximum internal pressure" as defined in Article N1311 of that code is 40 psig.

The coincident temperature is 264 F. The " design internal pressure" as defined in that code is 36 psig. The coincident design tempera-ture is 264 F. The vessel is capable of withstanding an external The vessel pressureof050psigreaterthantheinternag.

net free volume is approximately 2,800,000 ft pressure.

In addition to the pressure and temperature conditions specified, the contain=ent vessel will be designed to safely withstand seismic loads based on the following:

a. Maximum probable (smaller) seismic horizontal ground accelera-tion is 0.08 g.

b. The maximum possible (larger) seismic horizontal ground acceleration is 0.15 g.

c. A vertical component of 2/3 of the horizontal ground accelera-tion is applied simultaneously with the horizontal acceleration. 3 B. Penetrations

1. The shield building and penetration room penetrations for piping, ducts, and electrical cable are designed to withstand the nomal environmental conditions which may prevail during station operation and also to retain their integrity during postulated accidents.

The openings in the shield building and penetration rooms, including personnel access openings, equipment access openings and penetrations for piping, ducts, and electrical cable are designed to provide con-tainment which is as effective as the shield building and consistent with the leakage rate specified.

2. All containment vessel penetrations have the following design characteristics:

a. Capable of withstanding the maximum internal pressure and temper-ature which would occur F e to the postulated rupture of any pipe inside the containment vessel.

b. Capable of withstanding the applicable jet forces associated with the flow from a postulated rupture of the pipe in the penetration or adjacent to it, while still maintaining the integrity of containment. q

c. Capable of safely accomodating all themal and mechanical stresses which may be encountered during all modes of operation ,

and test.

~~

L U3.SS 1

\

15-47 Amendment No. 3 Y-

D-B 3- The containment isolation system, actuated by high containment pressure and high radiation, closes all penetrations, not required for operation of the engineered safety features systems, to minimize the leakage of radioactive materials to the environment. All isola-c) tion values, except those required for engineered safety features, fail closed on loss of actuating power.

C. Containment Systems

1. The containment vessel cooling and ventilation system is composed of the air recirculation cooling system and the purge system, which accomplishes three functions:

a. To re=ove heat released by equipment and piping in the con-tainment vessel during normal operation.

b. To purge the containment vessel with clean fresh air whenever access is desired. -

c. To cool and reduce the pressure of the containment vessel atmosphere after a loss-of-coolant accident.

The air recirculation cooling system consists of three fcn-cooler 3 units located throughout the containment vessel, outside the secondary shield. Any two units will satisfy the requirements of normal station operation. In the event of a loss-of-coolant accident, any two units will be capable of recirculating and cooling the containment vessel atmosphere to reduce the containment pressure.

The purge system is designed to provide clean fresh air to the con-tainment vessel or to the shield building and penetration rooms.

2. The containment vessel cooling system is designed to limit and subsequently reduce the containment vessel internal pressure, and consists of the air recirculation cooling system and the containment spray system. The air recirculation cooling system has the same coolingcapacityasthespraysystem,whichis(later) Btu /h. -

Two spray pumps and three air recirculation coolers are provided.

All of the following combinations of equipment will provide equal heat removal capability:

a. Two containment spray pumps.

b. Two air cooling units.

c. One air cooling unit and one containment spray pump.

Removal of heat by the spray system is accomplished by directing borated water spray into the containment atmosphere. The system consists of two half-capacity pumps with suction from the borated water storage tank and the containment sump, two half-capacity spray s

D-B headers, isolation valves, and the necessary piping, instrumentation y and controls.

3 The shield building and penetration room ventilation and filtration system is designed to provide a negative pressure within the shield building and penetration rooms following a loss-of-coolant accident and to reduce airborne fission product leakage to the environment by filtration prior to release of air through the station vent. Air 3 from the shield building and. penetration rocms will be drawn through the filter assembly consisting of roughing filters, high efficiency particulate filters and charcoal filters in series, and then dis-charged through the station vent.

Reference PSAR Section 5 9 l

l l

i 0191

{ .

S

. 15-h9 Amendment No. 3 L_ .

D-B 15 5 3 REACTOR Applicability Applies to the reactor vessel, reactor core and internals, and the reactor coolant system and components, including auxiliary systems'which ccnnect to and are exposed to the temperature and pressure conditions of the reactor coolant system.

Objective To describe those design features of the reactor system which are important to nuclear safety and which are not otherwise defined in other Technical Specifications.

Specification A. React r Core

1. The reactor core contains approximately 94.1 =etric tons of slightly enriched uranium dioxide pellets. Tne pellets are encapsulated in Zirealoy-4 tubing to form fuel rods. The reactor core is made up of 177 fuel assemblies. Each fuel assembly contains 208 fuel rods.

3

2. The fuel assemblies fem an essentially cylindrical lattice with an active height of 144 in and an equivalent diameter of 128 9 in.

3 Reload fuel will be similar in design to the initial core.

4. Burnable poison rod assemblies are incorporated in the initial core. The control rod positions in 72 of the fuel assemblies not equipped with control or axial power shaping rod assemblies will be utilized for location of burnable poison rod assemblies.

5 There are 49 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSRA) distributed in the reactor core a shown in PSAR Figure 3-46. The full-length CRA contain a 134-inch length of silver-indium-cadmium alloy clad with stainless steel. The APSRA contain a 36-inch length of silver-indium-cahim alloy clad in stainless steel.

B. Reactor Coolant System

1. The design of the pressure components in the reactor coolant system shall be in accord with the code requirements. "

2. The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pressure, are des 16 ced for a pressure of 2,500 psig and a

+

temperature of 650T. The pressurizer and pressurizer surge line are designed for a temperature of 670 F. j 3 The maximum reactor coolant system volume is (later) ft3, Amendment No. 3 1550

)

'm f,9G D-B 15 5 4 NEW AND SPENT FUEL STORAGE Applicability Applies to the capacity and design of the new and spent fuel storage areas.

Ob.iective To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas.

Specification A. New fuel assemblies are stored in the new fuel storage area, which is a separate and protected area for.the dry storage of new fuel assemblies. The new fuel assemblies are stored in racks in parallel rows having spacing sufficient to maintain a k eff of less than (later) even if flooded with non-borated water. The racks are designed to prevent insertion of r fuel assembly in other than the

• prescribed locations, thereby insuring a safe geometric array.

New fuel assemblies may also be stored in the spent fuel pool racks in a geometry which insures a k of less than (later) with the eff fuel flooded with non-borated water. The racks are designed to 3 prevent insertion of a fuel assembly in other than the prescribed locations, thereby insuring a safe geometric array.

3. After removal from the reactor, spent fuel is stored in the spent fuel storage pool.

The spent fuel storage pool is a reinforced concrete pool, lined with stainlecs steel, located in the fuel handling area of the auxiliary building. The spent fuel r.ssemblies are stored in racks having spacing and/or poison sufficient to maintain a k eff of less than (later) if i=mersed in non-borated water.

Although not required for safe storage of spent fuel assemblies, the spent fuel storage pool water will also be borated so that the refueling canal water will not be diluted during fuel transfer operation. The racks are designed to prevent insertion of a fuel assembly in other than the prescribed locations, thereby insuring a safe geometric array.

~'

0193 15-51 Amendment No. 3

D-B 15 6 ADOTISTRATIVE STANDARDS _,

15 6.1 ORGANIZATION, REVIEW AND AUDIT -

Applicability Applies to the organization of the station for safe operation and maintenance and establishes the means to providing a continuing review and audit of the nuclear safety aspects of station operation.

Ob.jective To define the administrative controls necessary to assure adequate manning of the station at all ti=es and to provide continuing review and audit of nuclear safety related conditions and operations.

Specification A. In all matters pertaining to actual operation and maintenance and to these technical specifications, the Station Superintendent shall report to and be directly responsible to the General Super-intendent, Power Production. The relationship between this supervisor and other levels of company management is shown in Figure (later).

3 B. The station organization shall include personnel with adequate training and experience to provide for the safe operation and maintenance of the station. These personnel vill be assigned to four basic groups:

Operation Maintenance Technical Chemical and Health The lines of responsibility within the station organization are shcvn in Figure (later).

C. AEC licensed personnel shall be provided as follows:

1. At least two licensed reactor operators shall be at the sta-tion, one of whom shall be in the control room at all times when there is fuel in the reactor vessel.

2. Two licensed reactor operators shall be in the control room during startup and scheduled shutdown of a reactor, and during recovery from reactor trips caused by transients or emer- -

gencies. A licensed senior reactor operator, who may be one of the above two licensed operators, shall be at the station at these times.

3 The Shift Supervisor shall be a licensed senior reactor oper-ator. Any person assigned to act in this capacity on a \

15-52 0194 ,

Amendment No. 3

D -B temporary basis shall be at least a licensed reactor operator and during this period, a senior reactor operator shall be

>' readily available on call.

4. A licensed senior reactor operator shall be in direct charge of any operation involving manipulation of fuel or control rods in the reactor vessel during refueling.

D. Review and audit of station operation, maintenance, and technical matters shall be provided by two review boards as follows:

1. A Station Review Board shall have the following membership:

Station Superintendent Operations Engineer Technical Engineer Chemical and Health Engineer Maintenance Engineer This board shall meet as required, but not, less than once each month. The current chairman or acting chairman plus two mem-bers shall form a quorum. The Station Superintendent shall appoint a chaiman and an acting chaiman to serve in the absence of the chaiman.

The board shall have the following responsibilities:

(a) Review all proposed new procedures or proposed changes to existing procedures which may affect nuclear safety.

(b) Review station operation and nuclear safety considerations.

(c) In accordance .ith Technical Specification 15.6 3 and 15.6.4, review abnomal occurrences and reported instances of violations of Technical Specifications, make reco= men-dations to prevent recurrence, and submit appropriate re-ports.

(d) Review all proposed emergency maintenance procedures that relate to nuclear safety.

(e) Review all proposed tests that relate to nuclear safety.

(f) Review proposed changes to Technical Specifications and changes or modifications to the station design, and sub-mit proposed changes or modifications to the Company Nuclear Review Board.

(g) Recommend to the Station Superintendent approval or dis-approval of the above proposals.

In the event of disagreement between the recommendations of the Station Review Board and the actions contemplated r[, by the Station Superintendent, the course detemined by 0195 g l 15-53 , A=endment Iro. 3 l

,.fri l

D-B the Station Superintendent to be the more conservative will be followed with i= mediate notification to the Chair- /

an of the Company Nuclear Review Board. -

2. A Ccmpany Nuclear Review Board shall have the following mem-bership which shall be appointed by the Vice President, Power Group:

Vice President, Power Group Chief Mechanical Engineer General Superintendent, Power Production Davis-Besse, Station Superintendent Station Electrical Engineer Nuclear Engineer, Mechanical Engineering Division Others deemed advisable (may be outside consultants)

(Later) members shall constitute a quorum.

The board shall have the following responsibilities:

(a) Review proposed new procedures or proposed changes to existing procedures which may affect nuclear safety.

3 (b) Review proposed station design changes or modifications.

(c) Review abnomal occurrences and violations of Technical Specifications and make reco=mendations to prevent \

recurrence. ,/

(d) Approve or disapprove reco=mendations in items (a), (b),

and (c).

(e) Conduct periodic station operations audits.

(f) Review all matters pertaining to radiation safety of station personnel and the general public.

l (g) Prepare requests to the AEC for changes in Technical i Specifications that involve unzeviewed safety questions I (10 CFR 50 59).

(h) Make reco=mendations concerning the above items to the Vice President, Power Group.

l a

01f6  !

3!

s

' Amendment No. 3 . 15-54

r-D-B 3

15 6.2 STATION OPERATING PBOCEDURES C~ Applicability Applies to approved written procedures for normal and emergency station oper-ations, maintenance, and testing operations that could affect the nuclear safety of the station.

Objective To assure that station personnel are provided with approved detailed written procedures governing the operation, maintenance and testing of components and systems that could have an effect on nuclear safety.

Specification A. The following detailed written procedures with appropriate check- '

off lists and instructions shall be provided:

l. Operating procedures for startup, normal operation, and shut-down of the reactor and of station systems and components involving nuclear safety.

2. Operating procedures for abnormal and emergency conditions 3 including specific malfunctions of systems or components.

3 Preventive or corrective maintenance procedures which have an effect on station or personnel safety.

4. Refueling procedures.

5 Radiation control procedures organized to meet the requirements of 10 CFR 20.

6. Emergency procedures involving possible or actual release of radioactive materials.

B. All procedures described in "A" above, and changes thereto, shall be reviewed by the Station Review Board and approved by the Station Superintendent prior to implementation, except ac provided in "C" below.

C. Written procedures shall be strictly adhered to in all =atters relating to nuclear safety. Temporary minor changes which do not 4 change the intent of the original procedure and which are not safety related are permitted only on approval of the Operations Engineer or Station Superintendent. Such changes shall be documented and subsequently reviewed by the Station Review Board.

C '

15-55 Amendment No. .

D-B 15 6.3 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL CCCURRENCE IN STATION OPERATION f)

Applicability Applies to administrative acticn to be followed in the event of an abnormal occurrence in station operation.

Ob.jective To provide for a review of all abnomal occurrences, to detemine their cause, to determine appropriate action to minimize the probability of recurrence, and to provide for reporting of such occurrences.

Specification A. Any abnomal occurrence (% defined in Section 151) shall be reported im=ediately by the parties involved to the Station  :

Superintendent who will notify through appropriate channels the General Superintendent, Power Production; Vice President, Power 3 Group; and the Chaiman of the Company Nuclear Review Boani.

B. The Station Review Board shall review each abnomal occurrence and prepare a report which shall include an evaluation of the cause of the occurrence and make recommendations for appropriate action to minimize the probability of a recurrence.

C. Copies of the Station Review Board report shall be submitted to the Station Superintendent; the General Superintendent, Power Production; the Vice President, Power Group; and the Chaiman of the Company Nuclear Review Board.

D. The Vice President, Power Group shall report the circumstances of any abnor=al occurrence to the AEC in acconi with regulations.

I

/ .

0198 x s

Amendment No. 3 15-56

D-B 15 6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY IIMIT IS EX ED Applicability Applies to administrative action to be taken in the event a safety limit is exceeded.

Objective D dw ine the appropriate action, review, and evaluation in the event a safety limit is exceeded, and to provide for reporting of such occurrence.

Specification

}

If a safety limit is exceeded:

A. The reactor shall be shut down and operation shall not be resumed until authorized by the AEC.

B. An i==ediate report shall be made by the parties involved to the Station Superintendent who will notify through appropriate channels the General Superintendent, Power Production; Vice President, Power Group; and the Chair =an of the Company Nuclear Review Board.

C. The Vice President, Power Group shall promptly report the occurrence to the AEC.

3 D. A complete analysis of the circumstances leading up to the occur-rence and the conseq;.ences shall be perfomed by the Station Review Board, and a report prepared. This report shall include an analysis of the effects of the occurrence and recommendations con-cerning operation of the unit and prevention of a recurrence.

Tnis report shall be submitted to the Station Superintendent; General Superintendent, Power Production; the Vice President, Power Groq; and the Chairman of the Company Nuclear Review Board.

Appropriate analyses or reports vill be submitted to the AEC by the Vice President, Power Group.

I l

y 03.99 C_

~-

15-57 Amend =ent No. 3 L

lwB 15 6 5 STATION OPERATING RECORDS O

.j Applicability Applies to all station records pertaining'to nuclear safety.

Objective To define those station records that will be provided and maintained for evaluation of nuclear safety.

Specification Station records and logs relating to the following items shall be retained for a period of time specified by Company policy or AEC regulations, which-ever is longer.

A. Records of nor=al station operation, including power levels and .

periods of operation at each power level.

B. Records of principal =aintenance activities, including preventive

=aintenance.

C. Records and reports of abnor=al occurrences and safety limits exceeded.

D. Records of periodic checks, tests, and calibrations performed to 3 verify that surveillance requirements are being met. n.

E. Reports of station design changes.

7. Records of nuclear fuel assemblies from receipt to shipment from site.

G. Records of routine station radiation monitoring and surveying.

H. Records of environ = ental monitoring surveys.

I. Records of radiation exposure of all station personnel, contractors, and visitors who enter radiation control areas.

J. Records of radioactive liquid and gaseous vastes released to the environ =ent, and containerized wastes shipped from the site.

K. Records of reactor physics tests and other special tests pertain-ing to nuclear safety.

L. Records of changes made in station operating procedures relating to Technical Specification 15.6.2.

\

^

-02007 Amendment No. 3 15-58 ,

1