Text

Amends 139 & 130 to Licenses DPR-42 & DPR-60,respectively Clarifying Conditions That Constitute Operable Irpi Sys Channels & Provide Compensatory Measures to Be Taken When Any Channel Determined to Be Inoperable
	ML20195D367
Person / Time
Site:	Prairie Island
Issue date:	10/30/1998
From:	Carpenter C; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20195D371	List: ML20195D367; ML20195D376;
References
	NUDOCS 9811180018
	Download: ML20195D367 (20)
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1 UNITED STATES

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NUCLEAR REGULATORY COMMISSION l

"g g,*****/g WASHINGTON, D.C. 20066 4 001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT 1 i

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.139 License No. DPR-42 1.

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

A.

The application for amendment by Northem States Power Company (the licensee) dated October 23,1998, as supplemented October 26,1998, complies -

with the standards and requirements of the Atomic Energy Act of 1954, as l

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

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.

l 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-42 is hereby amended to read as follows:

l 9811100018 981030 PDR ADOCK 05000282 4

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2 Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 139

, 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.

FOR THE NUCLEAR REGULATORY COMMISSION Cynthia A. Carpenter, Director Project Directorate ill-1 Division of Reactor Projects -lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: October 30, 1998

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ATTACHMENT TO LICENSE AMENDMENT NO. 139 FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-282 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 verticallines indicating the area of change.

REMOVE INSERT TS-iv TS-iv TS-x TS-x TS.3.10-6 TS.3.10-6 TS.3.10-6A B.3.10-9 B.3.10-9 l

B.3.10-9A B.3.10-98 i

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l TABLE OF CONTENTS (Continued) 1 TS SECTION TITLE PAGE 3.10 Control Rod and Power Distribution Limits TS.3.10-1

'A. Shutdown Margin TS.3.10 1 B. Power Distribution Limits TS.3.10-1 C. Quadrant Power Tilt Ratio TS.3.10-4 D. Rod Insertion Limits TS.3.10-5 E. Rod Misalignment Limitations TS.3.10-6 F. Rod Position Indication System TS.3.10-6A l

G. Control Rod Operability Limitations TS.3.10-7 H. Rod Drop Time TS.3.10-7 I. Monitor Inoperability Requirements TS.3.10-8 J. DNE Parameters TS.3.10-8 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 A Control Room Special Ventilation System TS.3.13-1 3.14 Deleted 3.15 Event Monitoring Instrumentation TS.3.15-1 j

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c Prairie Island Unit 1 Amendment No. 120, 121, 139 l

Prairie Island Unit 2 Amendment No. 113, 114, 130 t

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TS-x TABLE OF CONTENTS feontinued)

TS BAEES SECTION TITLE EAGE

'2.0L BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits B.2.1-1 A.

Reactor. core Safety Limits B.2.1-1 B.

Reactor Coolant System Pressure Safety Limits B.2.1-5 2.2 Safety Limit Violations 3.2.2-1 2.3' Limiting Safety System Settings. Protective B.2.3-1 Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 t

A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C.. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant oxygen. Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC)

B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 3~3 Engineered Safety Features B.3.3-1 3.4

' Steam and Power Conversion Systems B.3.4-1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7

~ Auxiliary Electrical System B.3.7-1 3.8 Refueling.and Fuel Handling B.3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitations B.3.10-9

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F. Rod Position Indication System B.3.10 9 G. Control Rod Operability Limitations B.3.10 9B

'H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10-10 1

. J. DNB Parameters B.3.10-10 j

3.11 Core Surveillance Instrumentation B.3.11-1 j

3.12 Snubbers B.3.12-1 i

3.13 Control Room Air Treatment System B.3.13-1 i

3.14 Deleted

-3.15 Event Monitoring Instrumentation B.3.15-1 i

i Prairie 1 Island Unit 1 Amendment No. 123, 139 JPrairieLIsland Unit 2 Amendment No. 116, 130

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TS.3.10 6 3.10.E. Red Mimaliennent L4mitations

1. If a rod cluster control assembly (RCCA) is misaligned from its bank by more than 24 steps, the rod will be realigned or the core power peaking factors shall be determined within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Specification 3.10.B applied.

If peaking factors are not determined within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the high neutron flux trip setpoint shall be reduced to 85 percent of rating.

2. If the misaligned RCCA is not realigned within a total of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the RCCA shall be declared inoperable.

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Prairie Island bnit 1 Amendment No. 49, 91, 139

. Prairie Island Unit 2 Amendment No. 43, 84, 130

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TS.3.10-6A 3.10.F. Rod Position Indication System

1. In MODE 1 each channel of the Rod Position Indication System shall be OPERABLE, capable of determining the control rod positions within the following (except as specified in 3.10.F.2 or 3.10.F.3 below):

a. With bank demand position greater than or equal to 215 steps, or less than or equal to 30 steps, the difference between the individual rod position indication and the demand position for the corresponding group step counter shall be no greater than i 24 steps, or

b. With bank demand position between 30 and 215 steps, the difference between the individual rod position indication and the demand position for the corresponding group step counter shall be no greater than i 12 steps.

2. In MODE 1 vith one rod position indicator per group inoperable for one or more groups either:

a. Verify the position of rod (s) with inoperable position indicator (s) indirectly using the moveable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

b. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3. In MODE 1 with more than one rod position indicator per group inoperable for one or more groups:

a. Verify the position of rods with inoperable position indicators i

indirectly using the moveable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

b. Verify the position of rods with inoperable position indicators indirectly using the moveable incore detectors within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of their position, and

c. Monitor and reecrd the demand position for the corresponding group s'tep counters for rods with inoperable position indicators at least once per hour, and

d. Monitor and record reactor coolant system average temperature at least once per hour, and

e. Restore inoperable position indicators to OPERABLE status witTin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months such that a maximum of one rod position indicator per group is inoperable.

4. If the requirements of Specification 3.10.F.3 cannot be met, then place the affected unit in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

5. If a control rod with an inoperable rod position indicator is found to be misaligned during the verification of rod position required by Specifications 3.10.F.2.a. 3.10.F.3.a or 3.10.F.3.b above, then apply the requirements of Specification 3.10.E.

i Prairie Island Unit 1 Amendment No. 139 Prairie Island Unit 2 Amendment No. 130 l

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B.3.10-9 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases continued D.

Rod Insertion Limits (continued) as stated above. Therefore.'this specification has been written to further minimize the likelihood of any hypothesized event during the j

i performance of these tasts later in life. This is accomplished by limiting'to two. hours per year the time the reactor can be in this type of configuration. and requiring that a rod drop test is performed on -

the rod to be measured prior to performance of test.

Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special precautions are taken during the test.

E.

Rod Misalignment Limitations l

Rod misalignment requirements are specified to ensure that power distributions more severe than those assumed in the safety analyses do not occur.

The rod misalignment limitations are linked closely with the Rod Position Indication System operability requirements of 3.10.F.

The relaxed rod position indicator channel operability requirements at less than or equal to 30 steps or greater than or equal to 215 steps of up to

,i 24' steps indicated position is allowed since the reactivity worths of control rods in these ranges are sufficiently small that this misalignment will have-no appreciable effect on core power distributions.

F. Rod Position Indication System l

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The rod position indicator channel is sufficiently accurate to detect a rod 17.5 inches away from its demand position in the center region of the core. A misalignment less than 15 inches in the center region of the core does not lead to over-limit power peaking factors. In the peripheral core regions (less than or equal to 30 steps or greater than or equal to 215 steps) a misalignment less than 22.5 inches will not 1ead to over-limit power peaking factors due to small control rod reactivity worth in this region of the core. If the rod position

' indicator channel is not operable, the operator will be fully aware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on

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escore nucleat detectors, and/or core thermocouples, and/or movable incore detectors, will be used to verify power distribution symmetry.

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These indirect measurements do not have the same resolution if the bank is near either and of the core, because a 15-inch misalignment would E

.have no effect on power distributions. Therefore, it is necessary to

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apply the indirect checks following significant rod motion.

I Prairie Island Unit 1 Amendment No. 91, 94, 139 Prairie Island Unit 2 Amendment No. 84, 87, 130

i-B.3.10-9A Specifications 3.10.F.2 and 3.10.F.3 provide actions to be taken when rod position indicators are determined to be inoperable. The actions to be taken are dependent on how many rod position indicators are inoperable per group. When dealing with rod position indicators associated with a control rod bank that does not contain multiple groups, the bank should be considered a single group for the purposes of entry into Specifications 3.10.F.2 or 3.10.F.3.

Specification 3.10.F.3.c requires that the demand position for the corresponding group step counters for rods with inoperable position indicators be monitored and recorded on an hourly basis.

This requirement is intended to provide a periodic assessment of rod position such that it can be determined if rod movement in excess of 24 steps has occurred since the last determination of rod position.

If rod movement in excess of 24 steps has occurred since the last determination of rod position._the requirements of Specification 3.10.F.3.b are to be implemented.

Specification 3.10.F.3.d requires that reactor coolant system average temperature be monitored and recorded on an hourly basis.

Monitoring and recording of the reactor coolant system average temperature may provide early detection of mispositioned or dropped rods.

Specifications 3.10.F.2.a and 3.10.F.3.a require that the position of irods with inoperable position indicators be verified indirectly using the moveable incore detectors every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The verification of rod position every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for continued plant operation since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

The reduction of THERMAL POWER to less than 50% of RATED THERMAL POWER required by Specification 3.10.F.2.b puts the core into a condition where rod position is not significantly affecting core peaking factors.

The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable. based on operating experience, for reducing power to less than 50% RATED THERMAL POWER from full power conditions without challenging plant systems.

Specification 3.10.F.3.b ensures that verification of rod position is initiated promptly following the movement of rods with inoperable position indication in excess of 24 steps in one direction, since the rod position was last determined or was last available from an OPERABLE rod position indication channel.

The four hour allowance for completion of this action allows adequate time for personnel to be called in and for them to complete the rod position verification using the moveable incore detectors.

Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 139 Amendment No. 130 i

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When more than one rod position indication channel per group is inoperable, the position of the rod (s) can still be determined by use of the moveable incore detectors.

Based on experience, normal power operation does not require excessive movement of control rods.

If one or more banks has been significantly moved, the action specified by Specification 3.10.F.3.b is required.

Therefore, verification of rod position within every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per Specification 3.10.F.3.a is adequate for allowing continued full power operation for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed out of service time also provides sufficient time to troubleshoot and restore the IRPI system to operation following a component failure in the system, while avoiding the challenges associated with a plant shutdown.

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Control Rod Operability Limitations j

one inoperable control rod is acceptable provided_that the power distribution limits are met, trip shutdown capability is available, and provided the potential hypothetical ejection of the inoperable rod is I

not worse than the cases analyzed in the safety analysis report. The rod ejection accident for an isolated fully-inserted rod will be worse if the residence t:Lae of the rod is long enough to cause significant

,non-uniform fuel depletion. The four-week period is short compared with the time interval required to achieve a significant non-uniform fuel depletion.

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Prairie Island Unit 1 Amendment No. 91, 94, 139 Prairie Island Unit 2 Amendment No. 84, 87, 130

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UNITED STATES g

l NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D.C. 20006 4001 s,...../

NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAlRIE ISLAND NUCLEAR GENERATING PLANT. UNIT 2 -

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 License No. DPR-60 I

1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated October 23,1998, as supplemented October 26,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's ru!es and regulations set forth in 10 CFR Chapter I; B.

The facil4y will operate in confonnity 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 Commission'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 requirementa have been satisfied.

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:

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Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 130, 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.

FOR THE NUCLEAR REGULATORY COMMISSION 4

,g Gk Cyn la A. Carpenter, Director Project Directorate ill-1 Division of Reactor Projects -lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: October 30, 1998 4

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l-ATTACHMENT TO LICENSE AMENDMENT NO.130 FACILITY OPERATING LICENSE NO. DPR-6Q QQCKET NO. 50-306 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 verticallines indicating the area of change.

REMOVE INSERT TS-iv TS-iv TS-x TS-x TS.3.10-6 TS.3.10-6 TS.3.10-6A B.3.10-9 B.3.10-9 B.3.10-9A B.3.10-9B l

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TS-iv l

TAELY OF CONTENTS (Continued)

TS SECTION T7TLE M

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3.10 Control Rod and Power Distribution Limits TS.3.10-1 A. Shutdown Margin TS.3.10-1 B. Power Distribution Limits TS.3.10-1 C. Quadrant Power Tilt Ratio TS.3.10-4 D. Rod Insertion Limits TS.3.10-5 l

E. Rod Misalignment Limitations TS.3.10-6 F. Rod Position Indication System TS.3.10-6A l

G. Control Rod Operability Limitations TS.3.10-7 H. Rod Drop Time TS.3.10-7 I. Monitor Inoperability Requirements TS.3.10-8 J. DNB Parameters TS.3.10-8 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 A Control Room Special Ventilation System TS.3.13-1 3,14 Deleted 3.15 Event Monitoring Instrumentation TS.3.15-1 i

Prairie Island Unit 1 Amendment No. 120, 121, 139

-Prairie Island Unit 2 Amendment No. 113. Ild, 130

l-TS-x TABLR OF CONTENTS (continued)

TS BASES SECTION M

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2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits B.2.1-1 A.

Reactor Core Safety Limits B.2.1-1 B.

Reactor Coolant System Pressure Safety Limits 3.2.1-5

.2.2 Safety Limit Violations 3.2.2-1 2.3 Limiting Safety System Settings. Protective B.2.3-1 Instrumentation 3.0 BASES FOP LIMITING CGNDITIONS FOR OPERATION 3.0 Applicability E.3.0-1 l

3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 L

3. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 i

E. Maximum Reactor Coolant Oxygen. Chloride B.3.1-8 j

and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC)

B.3.1 9 3.2 Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 l

3.4 Steam and Power Conversion Systems B.3.4-1 l

3.5 Instrumentation System B.3.3-1 3.6 Containment System B.3.6-1 j.

3.7 Auxiliary Electrical System B.3.7-1 i

3.8 Refueling and Fuel Handling B.3.8 1 3.9 Deleted 3.10 control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 l

B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 l

D. Rod Insertion Limits B.3.10 8 E. Rod Misalignment Limitations B.3.10 9 F. Rod Position Indication Syst'em B.3.10 9 G. Control Rod Operability Limitations B.3.10 9B l-H. Rod Drop Time B.3.10 10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 l_

3.12 Snubbers B.3.12-1 l-3.13 Control. Room Air Treatment System B.3.13-1 l

3.14 Deleted I

3.15 Event Monitoring Instrumentation B.3.13-1 l

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l Praf rie Island l' nit 1 Amendment No. 123, 139 Prairie Island Unit 2 Amendment No. 116, 130

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l-TS.3.10-6 3.10.E. Rod Minaliennent 14m4tations

1. If a rod cluster control assembly (RCCA) is misaligned from its bank by more than 24 steps, the rod will be realigned or the core l

power peaking factors shall be determined vithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Specification 3.10.3 applied. If peaking factors are not i

determined within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the high neutron fluz trip setpoint shall be reduced to 85 percent of rating.

2. If the misaligned RCCA is not realigned within a total of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the RCCA shall be declared inoperable.

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l Prairie Island Unit 1 Amendment No. 49, 91, 139 Prairie Island Unit 2 Amendment No. 43, 84, 130

_______________q TS.3.10 6A 3.10.F. Red Position Indication System

1. In PCDE 1 each channel of the Rod Position Indication System shall be O!ERABLE. capable of determining the control rod positions within the following (except as specified in 3.10.F.2 or 3.10.F.3 below):

a. With bank demand position greater than or equal to 215 steps, or less than or equal to 30 steps, the difference between the individual rod position indication and the demand pocition for the corresponding group step counter shall be no greater than i 24 steps, or

b. With bank demand position between 30 and 215 steps, the difference between the individual rod position indication and the demand position for the corresponding group step counter shall be no greater than i 12 steps.

2. In MODE 1 with one rod position indicator per group inoperable for one et more groups either:

a. Verify the position of rod (s) with inoperable position indicator (s) indirectly using the moveable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

b. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3. In MODE 1 with more than one rod position indicator per group inoperable for one or more groups:

s. Verify the position of rods with inoperable position indicators e

indirectly using the moveable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

b. Verify the position of rods with inoperable position indicators indirectly using the moveable incore detectors within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of their position, and

c. Monitor and record the demand position for the corresponding group s~tep counters for rods with inoperable position indicators at least once per hour, and

d. Monitor and record reactor coolant system average temperature at least once per hour, and

e. Restore inoperable position indicators to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months such that a marimum of one rod position indicator per group is inoperable.

4. If the requirements of Specification 3.10.F.3 cannot be met, then place the affected unit in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

5. If a control rod with an inoperable rod position indicator is found to be misaligned during the verification of rod position required by

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Specifications 3.10.F.2.a. 3.10.F.3.a or 3.10.F.3.b above, then apply the requirements of Specification 3.10.E.

Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 139 Amendment No. 130

3.3.10-9 3.10 CONTROL ROD AND POVER DISTRIBUTION LIMITS Basas continued D.

Rod Insertion Limits (continued) as stated above. Therefore, this specification has been written to further minimize the likelihood of any hypothesized event during the performance of these tests later in life. This is accomplished by limiting to two hours per year the time the reactor can be in this type of configuration. and requiring that a rod drop test is performed on the rod to be measured prior to performance of test.

Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special precautions are taken during the test.

E.

Rod Misalignment Limitations l

Rod misalignment requirements are specified to ensure that power distributions more severe than those assumed in the safety analyses do not occur.

The rod misalignment limitations are linked closely with the Rod Position Indication System operability requirements of 3.10.F.

The relaxed red position indicator channel operability requirements at less

,than or equal to 30 steps or greater than or equal to 215 steps of up to i 24 steps indicated position is allowed since the reactivity worths of control rods in these ranges are sufficiently small that this misalignment will have no appreciable effect on core power distributions.

F. Rod Position Indication System l

The rod position indicator channel is sufficiently accurate to detect a rod 17.5 inches away from its demand position in the center region of the core. A misalignment less than 15 inches in the center region of the core does not lead to over limit power peaking factors. In the peripheral core regions (less than or equal to 30 steps or greater than or equal to 215 steps) a misalignment less than 22.5 inches will not lead to over-limit power peaking factors due to small control rod reactivity worth in this region of the core. If the rod position i

indicator channel is not operable. the operator will be fully aware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on encore nuclear detectors. and/or core thermocouples, and/or movable incore detectors, will be used to verify power distribution symmetry.

These indirect measurements do not have the same resolution if the bank is near either and of the core, because a 15-inch misalignment would have no effect ou power distributions. Therefore, it is necessary to apply the indirect checks following significant rod motion.

Prairie Island Unit 1 Amendment No. 91, 94, 139 Prairie Island Unit 2 Amendment No. 84, 87, 130

l' 9

B.3.10 9A Specifications 3.10.F.2 and 3.10.F.3 provide actions to be taken when rod position indicators are determined to be inoperable. The actions to be taken are dependent on how many rod position indicators are inoperable per group. When dealing with rod position indicators associated with a control rod bank that does not contain multiple groups, the bank should be considered a single group for the purposes of entry into Specifications 3.10.F.2 or 3.10.F.3.

Specification 3.10.7.3.c requires that the demand position for the corresponding group step counters for rods with inoperable position indicators be monitored and recorded on an hourly basis.

This requirement is intended to provide a periodic assessment of rod position such that it can be determined if rod movement in excess of 24 steps has occurred since the last determination of rod position.

If rod movement in excess of 24 steps has occurred since the last determination of rod position, the requirements of Specification 3.10.F.3.b are to be implemented.

Specification 3.10.F.3.d requires that reactor coolant system average temperature be monitored and recorded on an hourly basis.

Monitoring and recording of the reactor coolant system average temperature may provide early detection of mispositioned or dropped rods.

Specifications 3.10.F.2.a and 3.10.F.3.a require that the position of

rods with inoperable position indicators be verified indirectly using the moveable incore detectors every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The verification of rod position every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for continted plant operation since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

The reduction of THERMAL POWER to less than 50% of RATED THERMAL POWER required by Specification 3.10.F.2.b puts the core into a condition where rod position is not significantly affecting core peaking factors.

The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experience, for-reducing power to less than 50% RATED THERMAL POWER from full power conditions without challenging plant systems.

Specification 3.10.F.3.b ensures that verification of rod position is initiated promptly following the movement of rods with inoperable position indication in excess of 24 steps in one direction, since the rod position was last determined or was last available from an OPERABLE rod' position indication channel.

The four hour allowance for completion of this action allows adequate time for personnel to be called in and for them to complete the rod position verification using the moveable incore detectors.

d b

Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 139 Amendment No. 130 S

a B.3.10-9B 1

When more than one rod position indication channel per group is inoperable, the position of the rod (s) can still be determined by use of the moveable incore detectors.

Based on experience, normal power i

operation does not require excessive movement of control rods.

If one or more banks has been significantly moved, the action specified by l

Specification 3.10.F.3.b is required.

Therefore, verification of rod l

position within every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per Specification 3.10.F.3.a is adequate for allowing continued full power operation for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed out of service time aluo-provides sufficient time to troubleshoot and restore the IRPI system to operation following a 1

compt. Tent failure in the system, while avoiding the challenges l

associated with a plant shutdown.

G.

Control Rod operability Limitations l

One inoperable control rod is acceptable provided that the power l

distribution limits are met, trip shutdown capability is available, and l

provided the potential hypot etical ejection of the inoperable rod is not worse than the cases analysed in the safety analysis report. The rod ejection accident for an isolated fully inserted rod will be worse if the residence time of the rod is long enough to cause significant l

pon uniform fuel depletion. The four week period is short compared with the time interval required to achieve a significant non-uniform fuel depletion.

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Prairie Island Unit 1 Amendment No. 91, 94, 139 Prairie Island Unit 2 Amendment No. 84, 87, 130

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