ML19350C919

Tech Spec Changes,Administrative in Nature,Including Revised Plant Staff Organization

ML19350C919
Person / Time
Site:	Point Beach
Issue date:	04/06/1981
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML19350C918	List: ML19350C917 ML19350C919
References
TAC-43742, TAC-43743, NUDOCS 8104100293
Download: ML19350C919 (13)

Text

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power distribution', the reacto trip limit, with allowance for errors,(2) is always below the core safety limit as shown on Figure 15.2.1-1.

If axial peaks are greater than design, as indicated by difference between top and bottom power range nuclear detectors, the reactor trip limit 1

is automatically reduced. (6) (7)

The overpower, overtemperature and pressurizer pressure system setpoints have been revised to include effect of reduced system pressure operation (including the effects of fuel densification). The revised setpoints as given daove will not exceed the revised core safety lbmits as shown in Figure 15.2.1-1.

The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur. The reactor is prevented from reaching the overpower limit condition by action of the nuclear overpower and overpower AT trips.

The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves ( 2485 psig) such that the reactor is tripped before the safety valves actuate.

The low pressurizer pressure reactor trip tripe the reactor in the unlikely event of a loss-of-coolant accident.I4)

The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or both reactor coolant pumps.

The set point specified is consistent with the value used in the accident analysis. (8) The low loop flow signal is caused by a condition of less than 90% flow as measured by the loop flow instrumentation. The loss of power signal is caused by 15.2.3-6 810.4100 M

i D.

During power operation the requirements of 15.3.2-B and C may be modified to allow the following components to be inoperable for a specified time.

If the system is not restored to meet the requirements of 15.3.2-B or C within the time period specified, the appropriate reactor (s) except as 1

otherwise noted, shall be placed in the hot shutdown condition.

If the requirements of 15.3.2-B or C are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the appropriate reactor (s) shall be pleced in the cold shutdown 4

condition.

1.

One of the two operable charging pumps associated with an operating reactor may be removed from service provided a charging pump associated with that same reactor is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

One of the boric acid transfer pumps designated in B.2 or C.2 may be out of service provided a pump is restored to operable status within 1

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

For the system piping and valve operability requirements (B.4 and C.4):

The flow path from the boric acid tank to a reactor coolant l

a.

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system may be out of service provided the flow path is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l b.

The flow path from the refueling water storage tank to the reactor coolant system may be out of service provided the flow path is restored to operable status within one hour.

If the flow path cannot be restored to operable status within one hour, the I

reactor shall be placed in cold shutdown with the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

r i

1 15.3.2-3 i

15.4.9 REACTIVITY ANOMALIES l

Applicability Applies to potential reactivity anomalies.

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objective To require evaluation of reactivity anomalies within the reactor.

Specification Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be periodically compared with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one percent in reactivity, an evaluation as to the cause of discrepancy shall be made and reported to the Nuclear Regulatory Commission.

Basis To eliminate possible eerros in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation I

between fuel burn-up and the boron concentration necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core condition. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured baron concentration is compared with the predicted concentration and the slope of the curve relating burn-up and reactivity is compared with that predicted. This process of normalization should be completed after about 10% of the total core burn-up.

Thereafter, 15.4.9-1 t

t

actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anomaly greater than 1% would be unexpected, and its occurrence would be thoroughly investigated and evaluated.

The value of 1% is considered a safe limit since a shutdown margin of at least 1% with the most reactive rod in the fully withdrawn position is always maintained.

Reference FSAR - Section 3.2.1 15.4.9-2

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15.5 DESIGN FEATURES 15.5.1 SITE g licability Applies to the location and extent of the reactor site.

Objective 1

To define those aspects of the site which affect the overall safety of the installation.

Specification The Point Beach Nuclear Plant is located on property owned by Wisconsin Electric Power Company at a site on the shore of Lake Michigan, approximately 30 miles southeast of the city of Green Bay.

The minimum distance from the reactor containment center line to the site exclusion boundary as defined in 10 CFR 100.3 is 1200 meters.

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15.5.1-1 i

BOARD OF DIRECTORS PRESIDENT WISCONSIN ELECTRIC POWER COMPANY EXECUTIVE VICE PRESIDENT OFF-SITE OTilER WISCONSIN ELECTRIC POWER COMPANY REVIEW DEPARTMENTS COMMITTEE DIRECTOR NUCLEAR POWER DEPARTMENT SUPERINTENDENT MANAGER MANAGER QUALITY AFSURANCE NUCLEAR ENGINEERING NUCLEAR OPERATIONS DIVISION SECTION SFCTION MANAGER'S SUPERVISORY MANAGEF1ENT ORGANIZATION CllART STAFF Figure 15.6.2-1

President I

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i Treasurer Executive Vice President I

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Superintendent System Security & Fire Protection Superintendent of Insurance and Claims (NEL-PIA and NML)

System Fire Protection Officer Director - Nuclear Power Department i

J

(----

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Fire Insurance Manager l

Nuclear Engineering Section Consultants, Designers &

Inspectors Manager -

Nuclear Operations Point Beach Nuclear Plant l

Fire Protection Design Consultant WISCONSIN ELECTRIC POWER COMPANY OFF-SITE MANAGEMENT FIRE PROTECTION ORGANIZATION

- - - - Administrative Crganization Figure 15.6.2-3

POINT BEACII NUCLEAR PIANT Manager -

FIRE PROTECTION ORGAN 4ZATION Nuclear Operations Figure 15.6.2-4 F-------------

l General Superintendent i

1

______________q l

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Security Supervisor SuWntendent - Operadons l

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Fire Protection Supervisor Training Engineer e

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1 Fire Brigade Chiefs (5)

(Operations Supervisors - 1 per shift) i I

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l Asst. Fire Brigade Chiefs (5)

Two creeks Volunteer g

(Operations Supervisors - 1 Fire Department per shift) 1 I

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t C'ntract Security Guards Operations Brigades (1 per shift)

(25 members total)

---

Administrative Organization Fire Protection Organization

i 15.6 ADMINISTRATIVE CONTROLS l

15.6.1 RESPONSIBILITY 15.6.1.1 The Manager - Nuclear Operations shall be responsible for overall facility operation and shall delegate in writing the succession to this i

f responsibility during absences from the Point Beach Nuclear Plant area of greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and where ready contact by telephone or other means is not assured.

15.6.2 ORGANIZATION j

OFFSITE 15.6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 15.6.2-1.

FACILITY STAFF l

15.6.2.2 The facility staff organization is shown on Figure 15.6.2-2.

Key I

)

personnel essential to the safe operation and maintenance of the facility are indicated on Figure 15.6.2-2 by an asterisk. Changes to the facility I

j staff organization which affect these key personnel shall be approved in advance by the Nuclear Regulatory Commission. The Licensee may otherwise alter this organizational structure for administrative purposes without prior Nuclear Regulatory Commission approval.

This structure for conduct of plant operation is subject to the following additional provisions.

a.

Each on-duty shift shall normally be composed of at least the minimum shif t crew composition as noted in Figure 15.6.2-2.

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b.

At least one licensed Operator shall be in the control room when fuel is in either reactor.

c.

At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.

l d.

An individual qualified in radiation protection procedures shall 1

i be on-site when fuel is in either reactor.

l 15.6.1/2-1 i

s e.

All core alterations af ter the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Oper-ator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

15.6.2.2.f A Fire Brigade of at least 5 members shall be maintained on-site at all times *. This excludes 3 members of the minimum shif t crew necessary for safe shutdown of the plant and any personnel required for other essential functions during a fire emergency.

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l Fire Brigade composition may be less than the minimum requirements for a l

period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements.

15.6.2-2 m

CONDUCT OF PLAlfr OPERATICWS FIGURE 15.6.2-2 Manager -

(1)

Nuclear Operations g------==-~--------------

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Manager's l

Supervisory g

Staff l

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l 8

1 OPERATIONS & SUPPORT I

l ADMINISTRATIVE & EtlGINEERING SERVICES l

l General (1) 8 Superintendent I

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TECPNICAL SERVICES OPERATIONS THAINING MAIKTENANCE 6 CCESTRUC'1103 l

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Superintendent -

(1)

I Engineering, Quality Superintendent - (1)

Superintendent -(1)

Training Iuperintendent-Technical Services Operations (SRO)

Engineer (1)

Maintenance &

I

& Regulatory-Constructson (1) 1 l

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s Assistant to Maintenanca I

Superintendent-l Supervisors l

CHEMISTRY & HEALTH PHYSICS REA R DIGINEERING INSTRUME

& CONTROL g

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• Superintendent -

Reactor

• Instrument &

l Chemistry & licalth Engineer (1)

Control Engineer (1) g Shift Supervisor Duty Technical Physics (1) g One per shift (SRO)

• Advisors I

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i PUTES:

l Nuclear Plant Technicians 1.

The Operations Group shift makeup is Engineers visors one por the minimum size for operation in all l

s..

Health (1) shift. (SRO or W) sodvs except cold shutdowr. of Point Physicist g

gg i

Beach Unit Nos. 1& 2.

The Oger-2.

In administrativo matters & routine HP control Operator

• tions group shif t makeup suy teu less Radioches ist concerns the HP reports directly to the One per shift-one that the minimum requirements for a ggg Supt.-Chemistry & Ilealth Pt'ysics.

In unit period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> matters of radiological health & safety Two per shift-two in order to accocenodate ur. expected policy determination, interpretation, or units (PN absence of on-duty shif t crew sembers, Nuclear Plant implementation, the HP (based on HP judg-l provided isnmediate action is taken to restore the shift makeup to within the Engineers ment) reports directly to the Manager -

inuxiltary Operator

• Nuclear Operations Two per shift-one i

4.

An unexpected absence of a Duty 3.

SRO = NRC Senior Reactor Operator License unit Technical Advisor shall be treated RO = NRC Reactc,r Operator License Three per shift-two

• See Specification 15.6.2.2 units similar to Note 1.

The Duty Technical Advisor is located on-site on ten minuta call to the control room.

COMPOSITION 15.6.5.2.2 The Manager's Supervisory Staff shall be selected from the following:

Chairman:

Manager - Nuclear Operations Member:

General Superintendent

{

Member:

Superintendent - Operations

}

Member:

Superintendent - Maintenance & Construction i

Member:

Superintendent - Engineering, Quality and Regulatory l

Member:

Superintendent - Chemistry and Health Physics Member:

Superintendent - Technical Services Member:

Instrument and Control Engineer Member:

Health Physicist Member:

Reactor Engineer Member:

Training Engineer ALTERNATES 15.6.5.2.3 Alternate members shall be appointed in writing by the MSS Chairman to serve on a temporary basis, however, no more than two alternates shall participate in MSS activities at any one time.

MEETING FREQUENCY 15.6.5.2.4 The MSS shall meet at least once per calendar month and as convened by the MSS Chairman.

QUORUM 15.6.5.2.5 A quorum of the MSS shall consist of the Chairman and four members including alternates.

RESPONSIBILITIES l

15.6.5.2.6 The Manager's Supervisory Staff shall:

a)

Review existing and proposed normal, abnormal and emergency l

operating procedures. Review maintenance procedures and l

proposed changes to these procedures and other procedures or changes thereto as determined by the Manager to affect i

l plant operational safety.

(Re: Section 15.6.7 for area of review.)

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15.6.5-2

15.6.9.2 Reportable Occurrences A.

prompt Notification with Written Followup The types of events listed in items 1 through 9 below shall be reported as expeditiously as possible within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director, Regulatory Operations, Region III, or his designate, no later than the first working day following the event. A written followup report must be submitted within two weeks. This written followup report shall include a completed copy of the licensee event report form, and may include additional narrative material to provide complete explanation of the circumstances surrounding the event.

1.

Failure of the reactor protection system or other systems subject to limiting safety-system settings to initiate the required protective function by the time a monitored para-meter reaches the setpoint specified as the limiting safety-system setting in the technical specifications or failure to complete the required protective function.

2.

Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the Technical Specifications.

3.

Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

15.6.9-4

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