ML20138L815

Amends 127 & 119 to Licenses DPR-42 & DPR-60,respectively, Revising TS 3.3.A & Associated Bases to Allow Safety Injection Pump Testing & Evolutions During low-temp Shutdown Conditions Provided Control for RCS Conditions in Place

ML20138L815
Person / Time
Site:	Prairie Island
Issue date:	02/20/1997
From:	Wetzel B NRC (Affiliation Not Assigned)
To:
Shared Package
ML20138L819	List: ML20138L815 ML20138L821
References
NUDOCS 9702250097
Download: ML20138L815 (10)

Text

_. _. _ _. _ _ _ _. _ _ _ _. _ _.. _ _ _ _ _ _ _... _ _ _ _ _ _

P pweg k

p UMTED STATES l

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NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. ageeH001 i

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... +

NORTHERN STATES P(EER COMPANY e

DOCKET NO. 50-282

~

PRAIRIE ISLAM) NUCLEAR GENERATING PLANT. UNIT NO.1 i

i-AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 127 l

License No. DPR-42 1.

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

i i

A.

The application for amendment by Northern States Power Company 1

(the licensee) dated February 6, 1997, as supplemented by a letter l

dated February 12, 1997, complies with the standards and requirements of the Atomic Energy-Act of 1954, as amended (the I

Act), and the Cosmiission's rules:and regulations set forth in l;

10 CFR Chapter I; i

l B.

The facility will operate in conformity with the application, the i

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

C.

There is reasonable assurance (1) 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 l

conducted in compliance with the Commission's regulations; i

i D.

The issuance of this amendment will not be inimical to the common l

defense and security or to the health and safety of the public; and I

i E.

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

/

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, j

and paragraph 2.C.(2) of Facility Operating License No. DPR-42 is hereby j

amended to read as follows:

t 4

4 T

9702250097 970220 PDR ADOCK 05000282 P

PDR

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.127, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance, with full implementation within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION SA O.

Beth A. Wetzel, roject Manager Project Directorate III-I Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 20, 1997 I

i

\\

ATTACMENT T0 LICENSE AMENDMENT NO.127 FACILITY OPERATING LICENSE NO. DPR-42 4

1 DOCKET NO. 50-282 i

1 Revise Appendix A Technical Specifications by removing the pages identified i

below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

t i

\\

REMOVE INSERT l

TS.3.3-3 TS.3.3-3 i

B.3.3-2 B.3.3-2 i

T

-m

-. 9

_ ~. _ _

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

j.

1 TS.3.3-3 l

3. 3. A. 2. g;. h valve position monitor lights or alarms for motor-operated valves specified in 3.3.A.1.g above may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided the valve position is verified once each shift.

3.

At least one safety injection pump control switch in the control room shall be in pullout vbenever RCS temperature is less than l

310*F* except that both SI pumps any be run for up to one hour while conducting the integrated SI test ** when either of the 1

following conditions i*s met:

1 l

(a) h re is a steam or gas bubble in the pressuriser and an isolation valve between the SI pump and the RCS is shut, or 4

l (b) The reactor vessel head is removed.

i 4.

Both safety injection pump control switches *** in the control Room shall be in pullout whenever RCS temperature is less than 200*F l

(except one or both pumps may be run as specified in 3.3.A.3 and j

3.1.A.1.d.(2)).

e i

i

• Valid until 20 EFPY

• • 0ther SI system tests and operations may also be conducted under these conditions.

• • • This specification does not apply whenever the reactor rossel head is removed.

Prairie Island Unit 1 Amendment No. 73, 91, 127 Prairie Island Unit 2 Amendment No. 66, 84, 119

1 i.

B.3.3-2 l

[

3.3 ENGINEERED SAFETY FEATURES i

AAAAA continued

)

(1) Assuring with high reliability that the safety system will function

{

properly if required to do so.

(2) Allowance of sufficient time to complete required repairs and testing using safe and proper procedures.

l Assuming the reactor has been operating at full RATED THERMAL POWER for j

at least 100 days, the magnitude of the decay heat decreases as follows j

after initiating HOT SHUTDOWN.

1 Time After Shutdown Decav Heat. t of RATED POWER i

1 min.

4.5 30 min.

2.0

}

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.62 l

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.96 j

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.62 i

Thus, the requirement for core cooling in case of a postulated loss-l of-coolant accident while in the shutdown condition is significantly reduced below the requirements for a postulated loss-of-coolant acci-dont during POWER OPERATION. Putting the reactor in the HOT SHUTDOWN l

condition significantly reduced the potential consequences of a loss-i-

of-coolant accident, and also allows more free access to some of the engineered safeguards components in order to effect repairs.

The accumulator and refueling water tank conditions specified are l

consistent with those assumed in the IDCA analysis (Reference 2).

l Specification 3.3.A.3 allows use of an SI pump to perform operations required at low RCS temperatures; e.g., raising accumulator levels in order to meet the level requirement of Specification 3.3.A.1.b(2) or ASME Section XI tests of the SI system check valves.

I Specification 3.3.A.3 also allows use of both SI pumps at low tempera-tures for conduct of the integrated SI test and other SI system tests j

and operations providing the pumps run for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this 4

case, pressurizer level is maintained at less than 50% and a positive j

means of isolation is provided between the SI pumps and the RCS to

{

prevent fluid injection into the RCS. This isolation is accomplished by using either a closed manual valve or a closed motor operated valve with the power removed. This combination of conditions under strict administrative control assure that overpressurization cannot occur. The j

option of having the reactor vessel head removed is allowed since in i

this case RCS overpressurization cannot occur.

(

Maintaining both safety injection pump control Room control switches in i

pullout, as specified in 3.3.A.4, will ensure that the RCS pressure /

}

temperature limitations specified in Pigures TS.3.1-1 and TS.3.1-2 will 3

not be exceeded, at low RCS temperatures, as the result of mass input 4

into the RCS from an inadvertent safety injection pump start. The j

provisions of this specification are not applicable when the reactor vessel head is removed since in that condition RCS overpressurization can not occur.

Prairie Island Unit 1 Amendment No. 94, 127 j

Prairie Island Unit 2 Amendment No. 84, 119 i

1

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[

UNITED STATES j

NUCLEAR REGULATORY COMMISSION 2

WASHINGTON. D.C. 3000H001 k...4,/

NORTHERN STATES POWER COMPANY j

DOCKET N0. 50-306 l

PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2

}

AMENDMENT TO FACILITY OPERATING LICENSE l

}

Amendment No. 119 i

License No. DPR-60 1.

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

i l

A.

The application for amendment by Northern States Power Company (the licensee) dated February 6, 1997, as supplemented by a letter dated February 12, 1997, complies with the standards and i

requirements of the Atomic Energy Act of 1954, as amended (the 4

Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 4

i B.

The facility will operate in conformity with the application, the l

provisions of the Act, and the rules and regulations of the Commission; l

C.

There is reasonable assurance (1) 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 l

conducted in compliance with the Commission's regulations;'

i 1

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 4

E.

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

requirements have been satisfied.

4-I 2.

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:

4

)

4

. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.119, are hereby incorporated in the license.

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

3.

This license amendment is effective as of the~ date of issuance, with full implementation within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION b~

0.

v Beth A. Wetzel, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical 1

Specifications Date of Issuance: February 20, 1997 l

1 J

t i

i.

1

}

4

ATTACHMENT TO LICENSE AMENDMENT NO. 119 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 l

i Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT TS.3.3-3 TS.3.3-3 B.3.3-2 Be3.3-2 l

i

i.

i TS.3.3-3 I

3.3.A.2.g.

The valve position monitor lights or alarms for motor-operated i

valves specified in 3.3.A.1.g above may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l

provided the valve position is verified once each shift.

i 3.

At least one safety injection pump control switch in the control

{

room shall be in pullout whanover RCS temperature is less than 310'F* except that both SI pumps may be run for up to one hour 2

while conducting the integrated SI test ** when either of the

{

following conditions is met:

l (a) There is a staan or gas bubble in the pressuriser and an isolation valve between the sI pump and the Ecs j

is shut, or i

(b) The reactor vessel head is removed.

4.

Both safety injection pump control switches *** in the Control Room shall be in pullout whenever ECS temperature is less than 200*F (except one or both pumps may be run as specified in 3.3.A.3 and 3.1.A.1.d.(2)).

• Valid until 20 EFPY

• • 0ther SI system tests and operations may also be conducted under these conditions.

~

• • • This specification does not apply whenever the reactor wssel head is removed.

Prairie Island Unit l' Amendment No. 73, 94, 127 Prairie Island Unit 2 Amendment No. 66, 84, 119

i c

B.3.3-2 h*[

3.3 ENGINEERED SAFETY FEATURES 1

Agang continued i

1 i

(1) Assuring with high reliability that the safety system will function j

l properly if required to do so.

(2) Allowance of sufficient time to complete required repairs and testing using safe and proper procedures.

e I

Assuming the reactor has been operating at full RATED THERMAL POWER for at least 100 days, the magnitude of the decay heat decreases as follows after initiating HOT SHUTDOWN.

i Tims After Shutdown Decay Heat. t of RATED POWER 1 min.

4.5 30 min.

2.0 l

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.62 i

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.96 l

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.62 i

Thus, the requirement for core cooling in case of a postulated loss-

+

I of-coolant accident while in the shutdown condition is significantly j

reduced below the requirements for a postulated loss-of-coolant acci-dont during POWER OPERATION. Putting the reactor in the HOT SHUTDOWN condition si nificantly reduced the potential consequences of a loss-5 l

of-coolant accident, and also allows more free access to some of the engineered safe 5uards components in order to effect repairs.

The accumulator and refueling water tank conditions specified are j

consistent with those assumed in the IDCA analysis (Reference 2).

1 j

Specification 3.3.A.3 allows use of an SI pump to perform operations l

required at low RCS temperatures; e.g., raising accumulator levels in i

order to meet the level requirement of Specification 3.3.A.1.b(2) or ASME Section XI tests of the SI system check valves.

?

Specification 3.3.A.3 also allows use of both SI pumps at low tempera-j tures for conduct of the integrated SI test and other SI system tests

)

and operations providing the pumps run for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this i

d case, pressurizer level is maintained at less than 50% and a positive means of isolation is provided between the SI peasps and the RCS to prevent fluid injection into the RCS. This iso ~iation is accomplished by i

using either a closed manual valve or a closed motor operated valve with the power removed. This combination of conditione under strict f

administrative control assure that overpressurization cannot occur. The i

option of having the reactor vessel head removed is allowed since in 1

this case RCS overpressurization cannot secur.

Maintaining both safety injection pump control Room control switches in i

pullout, as specified in 3.3.A.4, will ensure that the RCS pressure /

j temperature limitations specified in Figures TS.3.1-1 and TS.3.1-2 will i

not be exceeded, at low RCS temperatures, as the result of mass input into the RCS from an inadvertent safety injection pump start. The j

provisions of this specification are not applicable when the reactor vessel head is removed since in that condition RCS overpressurization can not occur.

Prairie Island Unit 1 Amendment No. 94, 127 l

Prairie Island Unit 2 Amendment No. 84, 119 1

-