ML20195D376

Safety Evaluation Supporting Amends 139 & 130 to Licenses DPR-42 & DPR-60

ML20195D376
Person / Time
Site:	Prairie Island
Issue date:	10/30/1998
From:	NRC (Affiliation Not Assigned)
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ML20195D371	List: ML20195D367 ML20195D376
References
NUDOCS 9811180025
Download: ML20195D376 (7)

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i UNITED STATES s

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 20086 4001 1

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 139 TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 130 TO FACILITY OPERATION LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By letter dated October 23,1998, as supplemented October 26,1998, the Northem States Power Company (NSP or the licensee) requested amendments to the Technical Specifications (TS) appended to Facility Operating License No. DPR-42 for the Prairie Island Nuclear Generating Plant, Unit 1, and Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Unit 2. The proposed amendments would clarify the conditions that constitute operable individual Rod Position Indication (IRPI) system channels, provide for an allowed out of service time for inoperable IRPl indicator channels, and provide compensatory measures to be taken when any channel is determined to be inoperable. The licensee requested that the proposed amendments be treated as emergency amendments as discussed in Section 3.0 of this Safety Evaluation.

2.0 EVALUATION 2.1 Backaround Prairie Island Units 1 and 2 have frequently experienced IRPI deviations of greater than 12 steps from the bank demand position involving more than one rod control cluster assembly (RCCA) during startups, shutdowns, and large power changes. The instrument accuracy of the IRPI at Prairie Island is affected by steady-state non-linearity in the relationship between actual rod position and analog indication and by transient thermal drift. These characteristics are generic to all Westinghouse-supplied analog IRPI systems. If the bank demand position is between 30 and 215 steps and the IRPl channel differs by more than 12 steps, the licensee considers the RCCA misaligned, in accordance with TS 3.10.E.2.b. In order to verify that the RCCA is not misaligned, the position of the RCCA is checked using core instrumentation (excore detector and/or thermocouples and/or movable incore detectors), as directed by TS 3.10.F.1.a. If the actual position of the RCCA is verified using core instrumentation, the licensee considers the RCCA to be aligned and operable. NSP also considers the IRPI to be inaccurate, but operable, as long as it tracks relative rod motion and can indicate a dropped DR P

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2-rod. Since the inaccurate IRPI is unable to provide a valid input to the Rod Deviation Monitor, the licensee considers the Rod Deviation Monitor to be inoperable for the affected IRPI channels, in accordance with TS 3.10.l.

However, on October 14,1998, the staff informed the licensee that they disagree with this interpretation and informed the licensee that it is inappropriate to consider the IRPI inoperable in order to determine actual RCCA position in accordance with TS 3.10.F.1.a, but operable to l

avoid a TS prescribed shutdown in accordance with TS 3.10.F.2.

l-The licensee agreed with the staff position and requested changes to the Prairie Island TS that would allow additional flexibility with respect to plant operation with IRPI channels. This license amendment request proposes changes based on changes previously approved for Callaway and Wolf Creek plants and on the guidance in the improved Standard Technical Specifications (NUREG-1431).

2.2 Evaluation of Proposed Chance i

in accordance with TS 3.10.F.2, should more than one RPI channel per group or more than two RPI channels per bank be found to be inoperable, the plant must be brought to the Hot Shutdown condition. As described above, this presents problems for any plant evolution requiring power changes; the current TS do not provide adequate flexibility to perform such evolutions as required plant maintenance and testing at power. Power changes needed to perform such required plant evolutions could result in an unnecessary plant shutdown. To correct this situation, the licensee has requested that its current TS 3.10.F, Inoperable Rod Position Indicator Channels, and TS 3.10.E, Rod Misalignment Limitations, be modified based on the current Callaway and Wolf Creek TS. The request would provide a completion time for conducting troubleshooting, repair, and replacement of RPI components during power operations. This safety evaluation addresses the licensee's proposed TS changes.

' The licensee justifies the TS change request based upon the fact that the safety analyses do -

not assume any dependence on the operator to monitor and properly position control rods. The safety analyses assume that the control rods are maintained within the TS Alignment and insertion Limits. Whenever an IRPI indicates that a rod is misaligned, the actual rod position must be verified utilizing the moveable incore detectors. Assurance that the control rods do not exceed their insertion limits is provided by the Rod insertion Limit Monitor in the emergency response computer system (ERCS), which receives input directly from a pulse-to-anelog converter fed from the rod control system. Proper function of the Rod insertion Limit Monitor is not dependent upon the operability of IRPI, except for the condition where power is lost to the IRPI cabinet. Maintaining control rods aligned and within insertion limits ensures that acceptable power limits are maintained and that minimum shutdown margin is maintained, and that there is not.a decrease in the minimum departure from nucleate boiling ratio (DNBR).

Continued plant operation with IRPI inoperable due to either a temperature-induced increase in the output signal nonlinear bias or some other mechanism will not increase the probability that operators will either misposition the control rods or fail to observe a control rod misalignment.

With no other indication of misalignment, there is no reason to expect that a control rod with an out-of-specification IRPI indicator channelis misaligned. Prairie Island has not had a rod

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- misalignment due to any reason. Random rod misalignment failure together with an associated IRPI failure is a very low probability event.

The Prairie Island license amendment request adequately addresses the specific TS changes.

With inoperable IRPI, the licensee's compensatory TS actions require that rod position / alignment be determined indirectly using the moveable incore flux detectors, that reactor coolant system temperature be monitored and recorded, and that the demand position for group step counters with inoperable IRPI be monitored and recorded. Since the licensee's safety analyses do not assume any dependence upon the operator to monitor and properly position control rods, the staff accepts that these actions will sufficiently ensure the minimum j

DNBR limit will not be violated. These proposed changes to the TS will not impact the ability of the plant staff to maintain operation within the TS core thermallimits. Furthermore, the reactor i

protection system detection features and the engineered safety feature mitigation features are unaffected by these proposed changes.

The addition of a 24-hour allowed out-of-service time provides sufficient time to troubleshoot and restore inoperable IRPI channels, while avoiding the challenges associated with a plant shutdown. Overall plant safety would be enhanced by having a reasonable opportunity to maintain steady-state operation rather than being immediately forced, without control rod position indication, to perform the large number of control rod movements required during a plant shutdown.

l The staff finds the TS changes proposed by the licensee to be acceptable.

3.0 DESCRIPTION

OF EMERGENCY CIRCUMSTANCES The Commission's regulations in 10 CFR 50.91 contain provisions for issuance of an

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amendment where the Commission finds that emergency circumstances exist, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant. The

' emergency exists in this case in that the proposed amendments are needed to prevent j.

shutdown of Prairie Island.

At times during rod movement the difference in rod position between the IRPI system outputs and the control rod group demand counters has exceeded the rod misalignment limitations in l

the Prairie Island TS 3.10.E.2.b. In each instance the performance of plant procedures to verify actual rod positions has confirmed that no control rod has been misaligned from its group.

Since the control rods were not misaligned, the operability of IRPI was next examined, but current TS 3.10.F does not clearly define IRPI operability. The licensee initiated NCR (nonconformance report) 19970613 to track investigation and resolution of the issue.

The instrument accuracy of the IRPI system at Prairie Island is affected by the following characteristics which are generic to all Westinghouse-supplied analog IRPI systems:

1) Steady-state non linearity in the relationship between actual rod position and analog indication. This non-linearity is fixed, measurable, and reproducible. The IRPI output j.

produces an S-shaped curve that can be calibrated to intersect group demand counter l

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4-output at two points. These points are chosen to provide a "best fit," but some offset between the two indications will exist over portions of the range of control rod motion.

2) Transient thermal drift, which is due to the change in detector characteristics caused by temperature changes after rod motion. This transient effect is measurable and predictaDie. After completion of rod motion the detector output returns to essentially the steady-state value within about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The licensee examined the Updated Safety Analysis Report (USAR) and the TS bases to identiW 3 s required functions of the IRPI system, which are to:

1) Provide relative IRPI in order to allow verification of rod motion upon demand.

2) Provide actual rod position indication to control board indicator and input signals to the ERCS (emergency response computer system) Rod Deviation Monitor for detection of a misaligned rod and annunciation of the control rod deviation.

3) Provide indication of each control rod at the bottom of the core following a reactor trip or plant shutdown.

The licensee had concluded that only two of the above capabilities must be maintained to provide for operability of the IRPI system and that this definition of operability was independent of system accuracy:

1) The IRPI channel (s) must trend with the control rod group demand position indication and provide the operator with the ability to manually detect rod misalignment.

2) The IRPI channel (s) must provide indication of the control rods at the bottom of the core.

The licensee concluded that an IRPI channel continued to be operable when the control rods were not misaligned but the difference between demand indication and IRPI indication was greater than 12 steps but less than 24 steps. With a difference greater than 24 steps, the IRPI channel should be logged out-of-service.

On October 14,1998, the licensee was informed that the NRC staff had completed a review of the definition of IRPI operability utilized at Prairie Island and had concluded that the licensee's definition of operability was not adequate. The NRC staff clarified during a call with the licensee staff on October 15,1998, that an IRPI channel was inoperable anytime the IRPI indicaM position differed from the group demand position by greater than the rod misalignment limitations in TS.

As discussed above, changes in rod position, as required during power changes, normally result in some amount of additional bias in the IRPI indicated rod positions, which can result in an increase in the difference between the IRPI indicated position and the group demand position. It is typical during significant power changes to see several IRPI channels affected by

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4 this phenomenon, some of which may exceed the rod misalignment limitations in TS between IRPI indicated position and group demand position.

Since operability of the IRP) channels is based on not exceeding a difference between IPRI indicated position and demand position by greater than the rod misalignment limitations in TS, then any significant power change can result in multiple IRPI channels being declared inoperable. Per the requirements of current Prairie Island TS 3.10 F.2, should more than one IRPl channel per group or more than two IPRI channels per bank be found to be inoperable, the plant must be brought to the Hot Shutdown condition.

This presents problems for any plant evolution requiring a significant power reduction, such as turbine valve testing or condenser cleaning, where the unit is not taken to the Hot Shutdown condition. Based on past performance of the IRPI channels and the clarified definition of IRPl operability, the current Prairie Island TS do r,ot provide adequate flexibility to perform such required plant maintenance and testing at power. The reduction in power required to perform the testing would likely result in a plant shutdown.

In addition, because the IRPI drift phenomenon is also experienced during power increases, the application of the i 12 step operability criteria in combination with the requirements of current TS 3.10.F.2, may prevent the plant from retuming to power following shutdowns or outages.

A power reduction of Prairie Island Unit 1 for turbine valve testing and condenser tube cleaning was scheduled to be performed the weekend of October 17,1998. This power reduction was scheduled prior to receiving the call from the NRC staff on October 14,1998. Implementation of the revised definition of IRPI operability, combined with the restrictions in the current Prairie Island TS, made it impossible to perform the power reduction and testing scheduled for October 17,1998. Initiation of the power reduction would have likely resulted in Unit 1 being taken to the Hot Shutdown condition and may have prevented the unit from restarting. As a result of the revised definition of IRPI operability, the power reduction was postponed until October 31,1998.

The turbine valve testing originally scheduled to be performed on October 17,1998, is required by TS 4.1. It must be completed by December 3,1998. It has been scheduled to be performed the weekend of October 31,1998, so that it does not have to be performed during the Unit 2 refueling outage scheduled to begin on November 7,1998.

The licensee does not consider it prudent to perform a planned Unit 1 power reduction and turbine valve test during the Unit 2 refueling outage. The safety concems focus on a potential for increased distractions to operating personnel that could occur by performing a Unit 1 power reduction and valve test during a refueling outage on the other unit. Such distractions have been shown through analysis of industry events to contribute to errors that could lead to undesirable plant transients or events. Plant safety is enhanced by performing the Unit i valve testing prior to the Unit 2 refueling outage. The current Prairie Island TS do not provide adequate flexibility for either unit to perform a significant power reduction for routine planned maintenance or for emergent unplanned maintenance. The reduction in power would likely result in a plant shutdown. Additionally, the units could be prevented from restarting following I

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. shutdown by the same IRPI phenomenon. The staff has determined that the licensee used its best efforts to make a timely application.

l Accordingly, the Commicsion has determined that emergency circumstances exist pursuant to 10 CFR 50.91(a)(5) and could not have been avoided, that the submittal was timely, and that the licensee did not create the emergency condition.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards considerations if operation of the facility in accordance with the amendment would not:

(1) involve a significant increase in the probability or consequences of any accident 1

previously evaluated; l

(2) create the possibility of a new or different kind of accident from any previously evaluated; or (3) involve a significant reduction in a margin of safety.

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These amendmente have been evaluated against the standards in 10 CFR 50.92 and the staff's final determination is presented below. They do not involve a significant hazards consideration because the change would not:

1. Involve a significant increase in the probability or consequences of an accident previously l

evaluated.

l The proposed changes do not affect any system that is a contributor to initiating events for

' previously evaluated design basis accidents. Neither do the changes significantly affect any system that is used to mitigate any previously evaluated design basis accidents. Therefore, the proposed changes do not involve a significant increase in the probability or consequence of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously analyzed.

The proposed changes do not alter the design, function, or operation of any plant component and do not install any new or different equipment; therefore, the possibility of a new or different kind of accident from those previously analyzed has not been created.

While continued operation of the plant for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with a number of IRPl channels inoperable

- does slightly increase the possibility that a misaligned control rod or mispositioned control group might go undetected by the operators for some time period, this possit,:lity is not new and has been addressed in previously performed safety analyses, where bounding conservative assumptions on rod position were used.

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4 3... Involve a significant reduction in the margin of safety, The proposed changes do not alter the initial conditions assumed in deterministic analyses s

associated with either the reactor coolant system boundary or fuel cladding; therefore, these changes do not involve a significant reduction in the margins of safety.

Accordingly, the Coramission has determined that the amendments involve no significant hazards consideration.

5.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Minnesota State official wa's notified of the proposed issuance of the amendments. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility.

component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final finding that the amendments involve no significant hazards consideration. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(g). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health ano safety of the public.

' Principal Contributors: T. Tjader A. Attard

. Date: October 30, 1998 i

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