ML20154S210

Proposed Tech Specs Clarifying Conditions That Constitute Operable Irpi Sys Channels

ML20154S210
Person / Time
Site:	Prairie Island
Issue date:	10/23/1998
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20154S204	List: ML20154S199 ML20154S206 ML20154S210
References
NUDOCS 9810270347
Download: ML20154S210 (16)

Text

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l EXHIBIT B PRAIRIE ISLAND NUCLEAR GENERATING STATION l

License Amendment Request dated October 23,1998 i

l Appendix A, Technical Specification Pages I Marked Up Pages

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(shaded material to be added, strikethrough material to be removed)

TS-iv TS-x TS.3.10-6 TS.3.10-6A B.3.10-9 B.3.10-9A i B.3.10-9B 9810270347 981023 PDR ADOCK 05000282 P PDR

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TS-iv IABLE OF CONTENTS (Continugdl TS SECTION TITLE EAGE i

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 l 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 P. Inepereble Rod Position IndicaterM 3 Ghennele fy M TS.3.10-68 G. Control Rod Operability Limitations T3.3.10-7 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 l

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TS x TABLE OF CONTENTS (continued)

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Td' BASES SECTION TITLE EAGE l

l 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYFTEM 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 l i 2.2 Safety Limit Violations B.2.2-1 2.3 Limiting Safety System Settings. Protective 3.2.3-1 Instrumentation l

3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 l 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 i 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 i and Fluoride Concentration I F. Isothermal Tempetature 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 j 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. RodMisalignmentLimitationl B.3.10-9 F. I*epecable Rod Position Indicate @ Ghennel-e M B.3.10-9 G. Control Rod Operability Limitations B.3.10-95 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 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.15-1

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r TS.3.10-6 3.10.E. Rod Misalignment Limitations

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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. and Specification 3.10.B applied. If peaking factors are not determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the high neutron flux trip setpoint 4

shall be reduced to 85 percent of rating.

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2. c. If the bank-demand pccition i geest-ee--t-hen-er-equal :: 215 eteps, cr lees-then er : quel tc 30 tcp: and the red pee 4-t4en 4*dieetee-channel-44-f-fers by scre than 2'. stepc. that rcd eentA4*etee-aee:=bly (RCCA) ch:14 be concidered

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b. If the-bank d

:nd pecitien is between 30 ced 215 stepe-end j the red pccition in44 eat-er channel differs by :::: then -44 eteper-t-het RCCA ch:11 be censidered =icalignedy

[3. If the misaligned RCCA is not realigned within a total of 8

hours, the RCCA shall be declared inoperable, b

F-r hepercble Red Pccition Indicater Channel:

1. If a red pcsit4en indicater--(RFI4--ehann 1 is-cut cf servie: then

c. Fcr cperatica between 50% :nd-100% cf RATED-THERMAIr-POWERT the p;;ition of--the-RCGA-ehall be ch::ked-directly by ecce j 4 net +umentet4en-feueer: dete:tc: and/c: theemeeeuples-and/ee i movable-ineeee-deteetere)-every-ehi-f t er cubeequent tc red 1 met 4en-exeeeding--a-tetel-ef--24--etepc , whiehever cecur: Meet,- I b . During-operet4en - beleu 50%-ef-RATED-THERMAIr-POWER , nc pec ie4-monitee4ng-le-requir+dy

2. The plent--ehell be brought--to-the-HOT-GHUTDOUM Ccndit4en-eheu14 mere-then-ene-RFI-ehennel p e-group cr =cr: then-tuc RPI channele per bank be found tc be inepeceble-ducing-POWER OPERATIONr

3. If a full length red having a red pce4t4en-indieetee-ehannel inepeceble ic fcund te be mic:Mgned f rc= 1re. abcve, then apply Spee4ffeation 3.10 R.

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B.3.10-9 3.10 C.ONTROL ROD AND POWER DISTRIBUTION LIMITS Eases continued D. Rod Insertion I.imits (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. RodMisalignmentLimitationl Rod misalignment requirements are specified to ensure that power distributions n. ore severe than those assumed in the safety analyses do not occur.

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F. -Ineperable-Rod Position IndicatIole Ghannele M:"ti The rod position indicator channel is sufficiently accurate to detect a rod i inches away from its demand position

. A misalignment less than 15 inches oes not lead to over-limit power peaking factors.

If the rod position 1ndicator 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 excore 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 end of the core, because a 15-inch misalignment would have no effect on power distributions. Therefore, it is necessary to apply the indirect checks following significant rod motion.

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G. Control Rod Operability Limitations 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 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 time 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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EXHIBIT C PRAIRIE ISLAND NUCLEAR GENERATING STATION License Amendment Request dated October 23,1998 Appendix A, Technical Specification Pages Revised Pages TS-iv TS-x TS.3.10-6 TS.3.10-6A B.3.10-9 B.3.10-9A B.3.10-9B i

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TS-iv TABLE OF CONTENTS (Continued)

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. 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 Monitoting Instrumentation TS.3.15-1 l

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TS-x IABLE OF CONTENTS (continued)

IS_,,3 ASIS SECTIQH TITLE PAGE 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 B.2.1-5 ,

2.2 Safety. Limit Violations B.2.2-1 '

2.3 Limiting Safety System Settings. Protective B.2.?-1 Instrumentation i

i 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION l 3.0 Applicability B.3.0-1 j 3.1 Reactor Coolant System B.3.1-1 i A. Operational Components B.3.1-1 l 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 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 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.15-1

TS.3.10-6  ;

3.10.E. Rod Misalignment Limitations

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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and l

Specification 3.10.B applied. If peaking factors are not

' determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the RCCA shall be declared-inoperable.

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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 with 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or

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

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

b. Verify the position of rods with inoperable position indicators indirectly using the moveable incore detectors within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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.

B.3.10-9 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Haken 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 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 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 regien 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 excore 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 end of the core, because a 15-inch misalignment would have no effect on power distributions. Therefore, it is necessary to apply the indirect checks following significant rod motion.

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B.3.10-9A Sp9e,ifications 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 l 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 l 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The verification of rod position every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 ellowance 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.

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j '. o B.3.10-9B f

When more than one rod position indication channel per group is inoperable, the po'sition of the rod (s) can still be determined by use of g the moveable incore detectors. Based on experience, normal power j 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 j position within every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per Specification 3.10.F.3.a is adequate 1 for allowing continued full power operation for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, since

the probability of simultaneously having a rod significantly out of i

position and an event sensitive to that rod position is small. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed out of service time also provides sufficient time to l troubleshoot and restore the IRPI system to operation following a a

component failure in the system, while avoiding the challenges associated with a plant shutdown.

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j G. Control Rod Operability Limitations One inoperable control rod is acceptable provided that the power j distribution limits are met. trip shutdown capability is available. and j provided the potential hypothetical ejection of the inoperable rod is not worse than the cases analyzed in the safety analysis report. The i 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

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

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