ML20064N165

Amends 61 & 55 to Licenses DPR-42 & DPR-60,respectively, Revising Tech Specs Re Steam Generator Tube Surveillance, Steam Exclusion Sys & Auxiliary Feedwater Sys Reporting Requirements & Administrative Controls

ML20064N165
Person / Time
Site:	Prairie Island
Issue date:	02/02/1983
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20064N171	List: ML20064N165 ML20064N174 ML20064N176
References
NUDOCS 8302160052
Download: ML20064N165 (27)

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• UNITED STATES g

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' NORTHERN' STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. I s

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. DPR-42 1.

The Nuclear Regulatory Commission (the Commission) has found that:

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A.

The application for amendment by Northern States Power Company i

(the licensee) dated September 14, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; i

B.

The facility will operate in conformity with the application, i

the provisions of the Act, and the rules and regulations of i

the Commission; i

C.

There is reasonable assurance (i) that the activities authorized by this amendnent can be conducted without endangering'th,e health f

I and safety of the public, and (ii) that such activities will be-l conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and sect. ity or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part i

51 of the Commission's regulations and all applicable requirements have been satisfied.

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8302160052 830202 DR ADOCK 05000282 PDR

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2-2.

Accordingly, the license is amended by changes to'th'e Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Faci.lity Operating License e

No. DPR-42 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 61, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Speci ficat' ions.

3.

Within 90 days after the effective date of thi.s amendment, or such later time as the Commission may specify, the Licensee shall satisfy any applicable requirement of P.L.97-425 related to pursuing an agree-ment with the Secretary of Energy for the disposal of high-level radio-active waste and spent nuclear fuel.

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4.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION qb b

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.t Robert A. Clark, Chief

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Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

February 2, 1983 e

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION j -Cp./, E WASHINGTOfJ, D. C. 20555 f..f..l NORTHERNSTATESPOWERC0htPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 License No. DPR-60 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Northern States Power Company (the licensee) dated September 14, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

-r B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Conmission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements.

l have been satisfied.

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 55, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Within 90 days after the effective date of this amendment, or such

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later time as the Commission may specify, the Licensee shall satisfy any applicable requirement of P.L.97-425 related to pursuing an agree-ment with the Secretary of Energy for the disposal of high-level radio-active waste and spent nuclear fuel.

4.

This license amendment is effective as of the date of its issuance.

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FOR THE NUCLEAR REGULATORY COMMISSION M. - _ _

t-Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date'of Issuance: February 2, 1983 T

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ATTACHMENT T0 LICENSE AMEND!iENTS AMENDMENT N0. 61 TO FACILITY OPERATING LICENSE NO. DPR-42 AMENDMENT NO. 5 5 TO FACILITY OPERATING LICENSE N0. DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Appendix A Technical Specifications

. with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Insert TS 3.1-3 TS 3.1-3 TS 3.3-1 TS 3.3-1 TS 3.3-2 TS 3.3-2 TS 3.3-5A TS 3.3-5A TS 3.4-2 TS 3:4-2_

TS 3.4-3 TS 3.4-3 Table TS.4.1-1 (Pg 2 of 5)

Table TS.4.1-1 (Pg 2 of 5)

Table TS.4.1-1 (Pg 3 of 5)

Table TS.4.1-1 (Pg 3 of 5)

Table TS.4.1-1 (Pg 4 of 5)

Table TS.4.1-1 (Pg 4 of 5)

Table TS.4.1-1 (Pg 5 of 5)

Table TS.4.1-1 (Pg 5' of 5)

Table TS.4.2-1 Table TS.4.2-1

- I' TS 4.5-2

.TS 4.5-2 Table TS.4.12-1 Table TS.4.12-1 TS 5.6-2 TS 5.6-2 TS 6.1-1 TS 6.1-1 Figure TS 6.1-1 & TS 6.1-2 Figure TS 6.1-1 & TS 6.1-2 i

TS 6.2-1 TS 6.2-1 TS 6.2-3 TS 6.2-3 TS 6.2-4 TS 6.2-4 TS 6.2-5 TS 6.2-5 TS 6.5-1 TS 6.5-1 l

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Basis e

When the boron concentration of the reactor coolant sy. stem is to be reduced, the proc,ess must be uniform to prevent sudden reactivity changes in the.

reactor. Mixing-of the reactor coolant will-be sufficient to maintain a uniform boron concentration if at least ona recctor coolant pump,or one residual heat removal pump is running while the change is taking place.

The residual heat removal pump will circulate the equivalent of the primary. system volume in approximately one-half hour.

" Steam Generator Tube Surveillance", Technical Specification 4.12, identifies steam generator tube imperfections having a depth >50%'of the 0.050-inch tube wall thickness as being unacceptable for power operation.

The results of steam generator burst and tube collapse tests submitted to the staff have demonstrated that tubes having a wall thickness greater than 0.025-inch have adequate margins of safety against failure due to loads imposed by normal plant operation and design basis accidents'.1 Part A of the specification requires that both reactor coolant pumps be operat-irg when the reactor is critical to provide core cooling in the event that a loss of flow occurs.

In the event of the worst credible coolant flow loss (loss of both pumps from 100% power) the minimum calculated DNBR remains.well above 1.30.

Therefore, cladding damage and release of fission products to the reactor coolant will not occur. Critical operation, except for low power physics tests, with less than two pumps is not planned. Above 10% power, an-automatic reactor trip will occur if flow from either pump is lost. Below 10% power, a shutdown under administrative control will be made if flow from either pump is lost.

.t The pressurizer _is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients.

Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point.

Below 350*F and 450 psig in the reactor coolant system, the wsidual heat removal system l

can remove decay heat and thereby control system 6emperature and pressure.

If no residual heat were removed by any of the means available,"the amount of steam which could be generated at safety valve relief pressure would be less than half the. valves capacity.

One valve therefore provides adequate defense against over-pressurizatio'n of the reactor coolant system for reactor coolant temperatures less than 350*F.

The combined capacity of both safety valves is greater;than the maximum surge rate resulting from complete loss of load.2 i

Prairie Island Unit 1 Amendment No. 47, 4V. 61 Prairie Island Unit 2 Amendment No. 41, #1 5 5 m, yyy.gm. m gm

TS.3.3-1 4

3.3 EXCT."EERED SAFETY FEATURES Anplicability Applies to the operating status of the engineered. safety features.

Objective To define those limiting conditions that are necessary for operation of' engineered safety features:

(1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.

Specifications A.

Safety Iniection and Residual Heat Renoval Systems

1. A reactor shall not be made or maintained critical nor shell it be heated or maintained above 2000F unless the following conditions are satisfied except as permitted in Specification 3.3.A.2.

a.

The refueling water tank contains not less than 200,000 gallons of water with a boron concentration of at least 1950 ppm.

b.

Each reactor coolant system accumulator shall be operable when reactor coolant system pressure is greater than 1000 psig.

Operability requires:

e-(1) The isolation valve is open (2) Volume is between 1250 and 1282.9 cubic feet of borated i

water (3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of at least 700 psig Two safety injection pumps are operable except that pump c.

control switches in the control room shall meet the require-ments of Section 3.1.G whenever the reactor coolant system temperature is less than )TT.

d.

Two residual heat removal pumps are operable.

l Two residual heat exchangers are operable.

e.

1 f.

Automatic valves, interlocks and piping associated with the' above components and required to function during accident conditions, are operable.

g.

Manual valvec in the above systems that could (if one is improperly positioned) reduce injection flow below that i

assumed for accident analyses, shall be blocked and tagged s

5 in the proper position for injection. RHR system valves, however, may be positioned as necessary to regulate plant x,.

heatup or cooldown rates when the reactor is suberitical.

All changes in valve position shall be under direct ad=ini-strative control.

Prairie Island Unit 1

' Amendment No. 35, 61 Prairie Island Unit 2 Amendment No'. 32, 5 5 g

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a TS.3.3-2 d.

Any. redundant valve in the system required for safety injection, may be inoperable provided' repairs are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Prior to initiating repairs, all valves in the system that provide redundancy shall be tested to demonstrate operability.

One accumulator may be inoperable for up to.one hour when-e.

i ever pressurizer pressure is greater than 1000 psig, f.

One safety injection system and one residual heat system may be inoperable for a time interval not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided the redundant safety injection system and heat removal system required for functioning during accident co'nditions is operable.

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B.

Containment Cooling Systems

l. A reactor shall not be made or maintained critical nor shall it be heated above 200 F unless the following condi'tions are 0

satisfied c:: cept as permitted by Specification 3.3.B.2.

Two containment spray pu=ps are operable.

a.

b.

Four fan cooler units are operable.

t Prairie Island Unit 1 Amendment No.. 21, 25, 6 1 Prairie Island Unit 2 Amendment No. 15, If, 5 5 e

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TS.3.3-5A One diesel-driven cooling water pump may be inoperable for a a.

period not to exceed seven days (total for both diesel-driven cooling water pumps during any consecutive 30 day period) provided:

(1) the op,erability of the other diesel-driven pump and its associated diesel generator are demonstrated immediately and at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

(2) the engineered safety features associated with the operable diesel-driven cooling water pump are operable; and (3) both off-site power supply paths frem the grid to the 4Kv emergency buses are operable.

(4) two motor-driven cooling water pumps shall be operable.

b.

One of the two required motor-driven cooling water pumps may be inoperable for a period not to exceed seven days provided:-

(1) the operability of both diesel-driven cooling water pumps is demonstrated i==ediately and at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

I One of the two required cooling water headers =ay be inoperable c.

for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided:

i (1) the operability of the diesel-driven pump and the diesel e

generator associated with safety features on the operable header is demonstrated ic=ediately.

(2) the horizontal motor-driven pump associated with the operable.

header and the vertical motor-driven pu=p are operable.

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f Prairie Island Unit 1 - Amendment No. 17, 27, 6 1

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Prairie Island Unit 2 - Amendment No. II,15, 5 5 4

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TS.3.4-2 For Unit 1 operation notor operated valves MV32242 and MV32243 e.

shall have valve position monitor lights operable and shall be locked in the open position by having the motor control center supply breakers manually locked open..For. Unit 2, corresponding valve conditions shall exist.

f.

Essential features including system piping, valves, and inter-locks directly associated with the above components are operable.

g.

Manual valves in the above systems that could (if one is im-properly positioned) reduce flow below that assumed for accident analysis shall be locked in the proper position for emergency use.

During power operation, changes in valve position will be under direct administrative control.

h.

The condensate supply cross connect valves C-41-1 and C-41-2

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to the auxiliary feedwater pumps shall be blocked and tagged open. Any changes in position of these valves shall be under direct administrative control.

3. Steam Exclusion System a.

Both isolation dampers in each ventilation duct that penetrates rooms containing equipment required for a high energy line rupture outside of containment shall be operable.

b.

If one of the two redundant dampers i.s removed from service for testing and maintenance purposes or-found inoperable, the operable redundant damper may remain open for a period not to exceed 24

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hours.

If after 24 hcurs, the inoperable damper is not returned-to service, one of the two dampers shall be closed.

c.

The actuation logic for one train of steam exclusion may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided the other train is tested and found operable prior to intiating repair of the inoperable channel.

4.

Radiochemistry The specific activity of the secondary coolant system for that reactor shall be < 0.10 uC1/gm DOSE EQUIVALENT I-131.

B.

If, during startup operation or power operation, any of the conditions of Specification 3.4.A.,

except as noted below for 2.a, 2.b or 4 cannot be '

met, startup operations shall be discontinued and if operability cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the affected reactor shall be placed in the cold shutdown condition using normal operating procedures.

With regard to Specifications 2.a or 2.b, if a turbine driven AFW pump is not operable, that AFW pump shall be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the affected reactor shall be cooled to less than 350*F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If a motor driven AFW pump is not operable, that AFW pump shall be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or one, unit shall be cooled to less than 350*F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Prairie Island Unit 1 - Amendment No. 17, 46, 51, 53,0 1 l

Prairie Island Unit 2 - Amendment No. 11, 46, 46, 47,3 5

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9 TS.3.4-3 If 4 is not met, the affected reactor shall be placed in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Basis ~

1 A reactor shutdown from power requires removal of decay heat. Decay heit.

removal requirements are normally satisfied by the steam bypass to_ the condenser and by continued feedwater flow to the steam generators.

Normal feedwater flow

- to the steam generators is' provided by operation of the turbine-cycle feedwater system.

The ten main steam safety valves have a-total combined rated capability of 7,745,000 lbs/hr.

The total full power steam flow is 7,094,000 lbs/hr; therefore, the ten main steam total steam flow if necessary.gg{ety valves will be able to relieve the In the unlikely event of complete loss of offsite electrical power to either or both reactors, continued removal of decay heat would'be assured by avail-4 ability of either the steam-driven auxiliary feedwater pump or the moto,r-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the main steam safety valves. -One auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from one reactor.

The motor-driven auxiliary feedwater pump for each. '

reactor can be made available to the 'other reactor.

The minimum amount,of water specified for the condensate storage tanks,is sufficient to remove the decay heat generated by one reactor in the first e.

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

Essentially unitnited replenishment of the condensate storage. supply is available from the intake structures through the cooling water system.

The two power-operated relief valves located upstream of the main steam isola-i tion valves are required to remove decay heat and cool th followingahighenergylineruptureoutsidecontainment.{2}eactordown Isolation dampers are required in ventilation ducts that penetrate those rooms contain-ing equipment needed for the accident.

i The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of.10 CFR'Part 100 limits in the event of a. steam line rupture.

This dose also' includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam.

generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.

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References I

(1) FSAR, Section 10.4 (2) FSAR, Appendix I Prairie Island Unit 1 - Amendment No. 46, 52, 6 1 Prairie Island Unit 2 - Amendment No. 40, 46, 5 5 i

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Mit1IFUM FREQUENCIES FOR CllECKS, CALIBRATIONS AND TEST OF INSTRUMENT CllAMMELS Channel Functional

Response

mm

$2 Description Check Calibrate Tent Test Remarks G n-

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4KV Voltage &

NA R-M NA Reactor protection circuits only Frequency

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$ $'8a.

RCP Breakers NA R

T NA

n. o.

@ F 9.

Analog Rod S(l)

R T(2)

NA

1) With ste'p counters Q['l 5* s' Position M(2)

2) Rod Position Deviation Monitor Tested by updating computer i

u-j bank count and comparing with i,

analog rod position test fj gg cignal 2

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E & 10.

Rod Position S(1,2)

NA

.T(3)

NA

1) With analog rod position

.J MS Bank Co6nters M(3)

2) Following rod motion in excess ij SS of six inches when the computer

.Y is out of service

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3) Control rod banks insertion limit monitor and control rod position 11, uu h.j

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deviation monitors un m

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Steam Generator S

R M

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Level ti j

12.

Steam Generator S

R H

NA H

h Flow Mismatch in m

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13.

Charging Flow S

R NA NA g

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14.

Residual lleat S(l)

R NA.

NA

1) When in operation g

Removal Pump Flow 4

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i 15 '. Boric Acid Tank D

R(1)

M(1)

NA

1) Transfer logic to Refueling Water y

Level Storage Tank g

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I S Channel Functional

Response

i Description Check Calibrate Test Test Remarks i.

i$ e 16.

Refueling Water W

R M(1)

NA

1) Functional test can beiperformed 1 S.

Storage Tank Level by biceding transmitter n

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Volume Control Tank S

R NA NA 7

8 d$

18a. Containment Pressure S

R M(1)

NA Wide Range Containment Pressure it SI Signal

1) Isolation Valve. Signal Y

O 18b. Containment Pressiire S

R M

NA Narrow Range Containment Pretsurc Steam Line Isolation

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j 18c. Containment Pressure S

R M

NA jM Containment Spray 7,;

$4 19.

Deleted l} [-

$j) 20.

Boric Acid Make-up Flow NA R

NA NA Channel Ji C q}

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21.

Containment Sump Level NA R

R NA Includes Sumps A, B, and C fi N

22.

Accumulator Level S

R R

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8 and Pressure

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23.

Steam Cenerator Pressure S

R H

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8

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Turbine First Stage Pressure S

R H

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25.

Emergency Plan Radiation

• M R,

M NA Includes those named in the emergency b'

) 0 Instruments procedure (referenced in Spec. 6.5 A.6)

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Protection Systems NA NA M

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j Logic Channel Testing Includes auto load sequencers i

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Response

Channel j

Description Check Calibrate-Test Test Remarks NA tj j

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27. Turbine Overspread NA R'

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Protection Trip Channel s

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Deleted 1

y 29.. Deleted

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Deleted g

31.

Seismic Monitors R

R NA NA Includes those reported'in Item 4 s

a of Table TS.6.7-1

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11 z

S R

11 NA O

d 32.

Coolant; Flow - RTD

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Byp,1ss Flowmeter t

33.

CRDM Cooling Shroud S

NA R

NA FSAR page 3.2-56 n

j Exhaust Air Temperature 34.

Reactor Gap Exhaust S

NA R

NA FSAR page 5.4-2 b[

jI Air Temperature i

35.

Post-Accident Monitoring M

E NA NA Includes all those in FSAR Table e

7.7-2 and Table TS.3.15-1 not in-3 S.

cluded elsewhere in 'this Table.

j n

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Steam Fxclusion W

R H

NA See FSAR Appendix I, Section

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g Actuation System 0

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Overpressure NA R

R NA Instrument Channels for PORV g

Control Including Overpredsure g;

g Mitigation System

}!itigation System p

N' a

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38.

Degraded Voltage NA R

H

'NA d

g 4KV Safeguard Busses w

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'39. Uoss of Voltage NA R

M NA

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f' 4KV Safeguard Busses g

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TABLE TS.4.1-1 (Page 5 of 5)

A Functional

Response

Channel

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Description Check Calibrate Test Test Remarks

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Auxiliary Feedwater NA R

R NA

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77 Pump Suction Pressure Ej s.-.

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Auxiliary Feedwater NA R

R NA 7;

@y Pump Discharge Pressure D-0-

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Each Shift

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Daily

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Weekly t8 o

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Monthly

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Quarterly un

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Each refueling shutdown g

1 ww P

Prior to each startup if not donc previous week P

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Prior to each startup following shutdown.in excess of 2 days if not done in the previous 30 days oa NA Not Applicable O

xh See Specification 4.1.D m

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Tablo TS.4.2-1 n

SPECIAL INSERVICE INSPECTION REQUIREMENTS Component Method of Extent and Ixamination Frequency REACTOR COOLANT PUITS 1.

Punp Flywhiel U.T.

An in-place ultrasonic volumetric

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examination of the areas of higher stress concentration at the bore and key way at approx. 3 year intervalr., during '.he refueling or maintenance shutdown coinciding with the in-service inspection schedule as required by the ASME B & PV Code Sectien XI.

M.T. or P.T.

A surface examination of all exposed U.T.

surfaces and complete ultrasonic volumetric examination at approx.

10 year intervals, during the pihnt shutdown coinciding with the in-service inspection schedule as required by the ASME B & PB Code Section XI.. Removal of the fly-wheel is not required to perform these examinations.

Notes:

1.

The following definitions shall app 3, to the inspection methods employed in Table TS.4.2-1.

a.

U.T. - Ultrasonic examination per IWA-2230.

b.

P.T. - Liquid Penetrant examination per IWA-2220.

c.

M.T. - Magnetic Particle examination per IWA-2220.

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TS.4.5-2 3.

Containment Fan Coolers Each fan cooler unit shall be tested during each reactor refueling shutdown to verify proper operation of all es.ential features including low motor speed, cooling water valves, and normal ventilation system dampers.

Ihdividual unit performance will be monitored by observing the terminal temperatures of the fan coil unit and by verifying a cooling water flo,w rate of greater than or equal to 900 gpm to each fan coil unit.

4.

Comoonent Cooling Water Syste m System tests shall be performed during each reactor refueling a.

shutdown.

Operation of the system will be initiated by tripping the actuation instrumentation, b.

The test will be considered satisfactory if control board indica-tion and visual observations indicate that all components have operated satisfactorily.

~

5.

Cooling Water Svstem a.

System tests shall be performed at each refueling shutdown. Tests shall consist of an automatic start of each diesel engine and automatic operation of valves required to mitigate accidents including those valves that isolate non-essential equipment from the system.

Operation of the system will be initiated by a simulated accident signal to the actuation instrumentation.

The tests will be considered satisf actory if control board indication and visual observations indicate that all components have operated satisfacterily and if cooling water flow paths required for accident mitigation have been established.

b.

Each diesel engine shall be inspected at each refueling shutdown.

B.

Cocoonent Tests 1.

Pumps a.

The safety injection pumps, residual heat removal pumps and contain-ment spray pumps shall be started and operated at intervals of one month. Acceptable levels of performance shall be that the pumps start and reach their required developed head on mini =um recircula-l tion flow and the control board indications and visual observations indicate that the pumps are operating properly for at least 15 minutes.

b.

A test consisting of a manually-initiated start of each diesel engine, and assumption of load within one minute, shall be conducted monthly.

Prairie Island Unit 1 Amendment No. 4f,01 Prairie Island Unit 2 Amendment No. 43,5 5 i

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1 STEAM CENERATOR TUBE INSPECTION Q{

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,9 IST SAMPLE INSPECTION 2ND SA}JfLE_ INSPECTION 3RD SAMPLE INSPECTION bj Sample Size Result Action Required Result Action Required Result Action Required i','

i A minimum of C-1 None N/A N/A N/A-N/A (3[k S Tubes per j

S.G.

C-2 Plug defective tubes l]

and inspect additional C-1 None N/A N/A ij 2S tubes in this S.G.

f C-2 Plug defective tubes C-1 None O

and inspect additional C-2 Plug defective tubes

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C-3 Perform action for 4

c-3 result of first-j sample h

C-3 Perform action for N/A N/A

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sample

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j C-3 Inspect all tubes in All other - None N/A

'A this S.G., plug de-S.G.s are II) fective tubes and C-1 3

inspect 2S tubes in Some S.G.s Perform action for N7A N/l each other S. G.

C-2 but no C-2 result of second H.

D additic,nal sample N

'4 Prompt notification S.G; are to NRC.

C-3 Hus.

Additional.

Inspect all tubes in N/A N/A p.G. os C-3, each S.G. and plug

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Prompt notification to NRC.

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, j S=3%; When two' steam generators are inspected during that outage.

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.S+6%; When one steam generator is inspected'during that outage.

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Unit 1 - Aihendment No. 31, G 1 i

Unit 2 - Amendment No. 25, 5 5 A

TS.5.6-2 The spent fuel pool has a reinforced concrete bottom'elab nearly 6 feet thick and has b.een designe.d to minimize loss of water due to a dropped cask accident'. Such water loss, if it did occur, would be from the smaller of the wo compartments, leaving the stored fuel in the larger.

compartment still covered with water. Piping to the pool is arranged so that failure of any pipe cannot drain more than 3 feet of. water from the pool. This leaves a margin of 22 feet of water above the tops of the stored fuel assemblies.

C.

Fuel Handling The fuel handling system provides the means of transporting and handling' fuel from the time it reaches the plant in an unirradiated condition until

. it leaves after post-irradiation cooling. The system consists of the refueling cavity, the fuel transfer system, the spent fuel storage pit, and the spent fuel cask transfer system.

Major components of the fuel handling system are the manipulation

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crane, the spent fuel pool bridge, the auxiliary building. crane, the fuel transfer system, the spent fuel storage racks, the spent fuel cask, and the rod cluster control changing fixture. The reactor vessel stud tensioner, the reactor vessel head lifting device, and the reactor inte 221s lifting device are used for preparing the re. tor for refueling and for asse=bling the reactor aftcr refueling.

Upon arrival in the storage pit, spent fuel will be redoved from the transfer system and placed, one assembly at'a time, in storage racks using a long-handled manual tool suspended frem the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded. into shipping casks for re= oval from the site.

The casks 'will be handled by the auxiliary building crane.

.The spe,nt fuel cask will be lowered 66 feet from th~e auxiliary building to the railroad car for offsite transportation. Specification 3.8 will limit this loading operation so that if the cask drops 66 feet, there will not be a significant release of fission products from the fuel in l

the cask.

Unit 1 - Amendment No.6 1 s.'

Unit 2 - Amcadment No. 5 5 i

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TS.6.1-1 6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION A.

The Plant Manager has the overall full-time onsite respons.ibility for safe operation of the facility.

During periods when the Plant Manager is unavailable, he may delegate this responsibiliity to other qualified supervisory personnel.

B.

The Nortnern States Power corporate organizational structure relating to the operation of this plant is shown on Figure TS.6.1-1.

C.

The functional organization for operation of the plant shall be as shown in' Figure TS.6.1-2 and:

1.

Each on duty shift shall be composed ofoat least the minimum shif t crew composition shown on Table TS.6.1-1.

2.

For each reactor that contains fuel:

a licensed operator in the control room.

3.

At least two licensed operators shall be present in the control room during a reactor startup, a scheduled reactor shutdown, and during recovery from a reactor trip.

These operators' are in addition to those required for the other reactor.

-t 4.

An individual qualified in radiation protection procedures F

shall be on site when fuel is in a reactor.

5.

All refueling operations shall be directly supervised by a licensed Senior Reactor Operator or a Senior Reactor Operator Limited to Fuel Handling who has no other concurrent respons-ibilities during this operation.

6.

A fire brigade of at least five members shall be maintained on site at all times." The fire brigade shall not include the six members of the mirimum shif t crew for safe shutdown of the reactors.

• Fire Brigade composition may be less than the minimum requirements for a perio'd 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 i=zediate action is taken to restore the Fire Brigade to within the minimum requirements.

Prairie Island Unit 1 Amendment No. 39, gg, 61 Prairie Island Unit 2 Amendment No. 33, 47, 5 5

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Unit 2 - Amendment !!o. 33, 43, 5 5

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s TS.6.2-1 6.2 Review and Audit Ccganizational units for the review and audit of facility operations shall-be constituted and have the respo'nsibilities and authorities outlined below:

A.

Safety Audit' Committee (SAC)

The Safety Audit Committee provides the independent review of plant operations from a nuclear safety standpoint. Audits of plant operation are conducted under the cognizance of the SAC.

1.

Membership The SAC shall consist of at least five (5) persons.-

a.

b.

The SAC chairman shall be an NSP representative, not havfhg line responsibility for plant operation, appointed by the Director of Euclear Generation. Other SAC members shall be appointed by the Director of Nuclear Generation or by such other person as he may designate. The Chairman shall appoint a Vice Chairman from the SAC membership to act in his absence.

c.

No more than two members of the SAC shall be from groups holding line responsibility for operation of the plant.

t d.

A SAC member may appoint an alternate to serve in his absence, with concurrence of the Chairman. No more than one alternate shall serve on the SAC at any'one time. The alternate member shall have voting rights.

~

2.

Qualifications a.

The SAC members should collectively have the capability required to review activities in the following areas:

nuclear power plant operations, nuclear engineering, chemistry and radiochemistry, =etallurgy, instrumentat'.on and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and other appro-priate fields associated with the unique characteristics of the nuclear power plant.

5 Prairie Island Unit 1 Amendment No. 13, 41,6 1 Prairie Island Unit 2 Amendment No. 7, 41, 5 5 -

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'TS.6.2-3 e

f.

Investigation of all events which are required by regulation or technical specifications (Appendix A) to be reported to NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

S.

Revisions to the Facility Emergency Plan, Facility Security Plan, and.the Fire Protection Program.

h.

Operations Committee minutes to determine if matters considered by that Committee involve unreviewed or unresolved safety questions.

i. Other nuclear safety matters referred to the SAC by the Operations Committee, plant management or company management.

j. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures systems, or components.

k.

Reports of special inspections and audits conducted in accordance with specification 6.3.

1.

Changes to the Offsite Dose Calculation Manual (ODCM).

m.

Review of investigative reports of unplanned releases of radioactive

=aterial to the environs.

~

D 6.

Audit - The operation of the nuclear power plant shall be audited formally f-under the cognizance of the SAC to, assure safe facility operation.

a.

Audits of selected aspects of plant operation, as delineated in Paragraph 4.4 of ANSI N18.7-1972, shall be performed with a frequency commensurate with their nuclear safety significance and in a manner to assure tnat an audit of all nuclear safety-related activities is completed within a period of two years. The apdits shall be performed in accordance with appropriate written instructiens and proce,dures.

b.

Audits of aspects of plant radioactive effluent treatment and radio-logical environmental monitoring shall be performed as follows:

1. Implementation of the Offsite Dose Calculation Manual at least once every two years.

2. Implementation of the Process Control Program for solidificatio'n of radioactive wastes at least once every two years.

3. The Radiological Environmental Monitoring Program and the results thereof, including quality controls, at least once every year.

c.

Periodic review of the audit program should be performed by the SAC at least twice a year to assure its adequacy, i

d.

Written re, ports of the audits shall be reviewed by the Director of Nuclear Generation, by the SAC at a scheduled meeting, and by cembers of management having responsibility in the areas audited.

f Unit 1-AmendmentNo.fg,f$sC1 Unit 2 a Amendment No. 43, $ 1 5 5

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Authority The SAC shall be advisory to the Director of Nuclear Generation.

l 8.

Records ttinutes shall be prepared and retained for all scheduled meetings of The minutes shall be distributed within one l the 3sfety Audit Committtee.

month of the meeting to the Director of Nuclear Generation, the General Manager Nuclear Plants, each member of the SAC and others desigt.ated by.

the Chairman. There shall be a formal approval of the minutes.

9.

Procedures A written charter for the SAC shall be prepared that contains:

a.

Subjects within the purview of the group.

b.

Responsibility and authority of the group.

c.

Mechanisms for convening meetings.

d.

Provisions for use of specialists or subgroups.

Authority to obtain access to the nuclear pode,r plant operating e.

record files and operating personnel when assigned audit func tions.

f.

Requirements for distribution of reports and minutes prepared by the group to others in the 'ISP organization.

1 Unit 1 - Amendment No. Z3, 6 J,61 Unit 2 - Amendment No. 7, 11.55

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Operations Conmittee (OC) 1.

Members' hip The Operations Committee shall consist of at least six. (6) members drawn from the key supervisors of the ensite staff.

The Plant Manager shall serve. as Chairman of the OC and.shall appoint a Vice Chairman from the OC membership to act in his absence.

~2.

Meeting Frequency The Oper'ations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

3.

Quorum A majority of the permanent members, including the Chairman or Vice Chairman 4.

Responsibilities - The following subjects shall be reviewed by.

the Operations Comnittee:

a.

Proposed tests and experiments and their results, b.

Modifications to plant systems or equipment as described t-in the Updated Safety Analysis Report and having nuclear l

safety significance or which involve an unreviewed safety question as defined in Paragraph 50.59 (c), Part 50, Title 10, Code of Federal Regulations.

c.

Preposals which would effect permanent cha'nges to normal and emergency operating procedures and any other proposed i

changes or procedures that will af fect nuclear safety as

, determined by the Plant Manager.

I d.

Proposed changes to the Technical Specifications or operating licenses.

e.

All reported or suspected violations of Technical Specifica-tions, operating license requirements, administrative procedures, operating procedures.

Results of investigations, l'

including evaluation and recommendations to prevent recurrence will be reported in writing to the Director of Nuclear Generation l,

and to the Chairman of the safety Audit Committee.

{

Prairie Island Unit 1 Amendment No. Ag, 61 l

Prairie Island Unit 2 Amendment No. 43, 5 5 l

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TS.6.5-1

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6.5 P'LA:iT OPERATING PROCEDUFES Detailed writt.en procedures, including the applicable checkoff lists and instructions, covering areas listed below shall be prepared and

-followed. These procedures and changes thereto, except as specified in TS 6.5.D., shall be reviewed by the Operations Committee and approved' by a mecher of plant manage =ent designated by the Plant Manager.

A.

Plant Operations 1.

Integrated and. system procedures for normal startup, operation and shutdown of the reactor and all systems and components involving nuclear safety of the facility.

2, Fuel handling operations 3.

Actions to be taken to correc0 specific and foreseen potential or actual malfunction of systems or components including responses to alarms, primary system leaks and abnormal reactivity changes and including follow-up actions required after plant protective system actions have initiated.

4.

Surveillance and testing requirements that could have an effect en nuclear safety.

5.

Implecenting procedures of the security plan.

e 6.

I:plementing procedures of the Facility Emergency. Plan, including procedures for coping with e=ergency conditions involving potential or actual releases of radioactivity.

7.

Inplementing procedures of emergency plans for coping with earthquakes and floods. The flood e=argency plan shall require plant shutdown for water levels at the site higher than 692 feet above 15L.

8.

Implementing procedures of the fire protection prograc.

9.

Implementing procedures for the Process Control Program and Offsite Dose Calculation Manual including quality control ceasures.

Drills on the procedures specified in A.3. above, shall be conducted as a part of the retraining program.

B.

Radiological Radiation control procedures shall be maintained and made available

?,

to all plant personnel. These procedurer shall show permissible radiation exposure and shall be consistent with the requirements of 10CFR20. This radiation protection program shall be organized to meet the requirements of 10CFR20.

Unit l'- Amendment No. 26,f 7, Unit 2 - Amendment No. 20, f y,5 b

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