ML18058B388

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Application for Amend to License DPR-20,consisting of Proposed TS Changes to Table 3.23-2, Radiation Peaking Factor Limits for Cycle 11. Offsite Dose Calculations & EMF-92-178, Palisades Cycle 11:Disposition & ... Encl
ML18058B388
Person / Time
Site: Palisades Entergy icon.png
Issue date: 01/29/1993
From: Slade G
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
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ML18058B389 List:
References
NUDOCS 9302040157
Download: ML18058B388 (7)
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Text

consumers Power GB Slade General Manager

~-,,,, l'OWERINli MICHlliAN'S l'ROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 January 29,1993 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - TECHNICAL SPECIFICATIONS CHANGE REQUEST - RADIAL PEAKING FACTOR LIMITS FOR CYCLE 11 Enclosed is a request for a change to the Palisades Technical Specifications to provide Radial Peaking Factor limits for the Cycle 11 (Reload 0) fuel to be installed during the 1993 refueling outage.

The following Attachments are included in support of this request:

1) Attachment 1, Proposed Technical Specifications pages
2) Attachment 2, Existing Technical Specifications pages marked to show the proposed changes
3) Attachment 3, EMF-92-178 Palisades Cycle 11: Disposition and Analysis of Standard Review Plan Chapter 15 Events
4) Attachment 4, EA-A-NL-91-169-01 Offsite Dose Calculations of Fuel Handling Accident
5) Attachment 5, EA-A-NL-91-169-02 Radiological Consequences of a Cask Drop Accident in the Spent Fuel Pool.

It is requested that this change request be effective prior to the plant leaving Cold Shutdown following the 1993 refueling outage. That date is based on the currently scheduled 50 day refueling outage starting on June 4, 1993.

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Gerald B Slade General Manager CC Administrator, Region III, USNRC Resident Inspector, Palisades 04001 . .

Attachments

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9302040157 930129 PDR ADOCK 05000255 p PDR A CMS ENERGY COMPANY

1 CONSUMERS POWER COMPANY Docket 50-255 Request for Change to the Technical Specifications License DPR-20 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in the Facility Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on February 21, 1991, for the Palisades Plant be changed as described below:

I. Change It is proposed that Table 3.23-2, Radial Peaking Factor Limits, be changed to add limits for those new fuel bundles to be installed during the 1993 refueling outage. In addition, the bases for several Specifications (2.1, 2.3, 3.1, 3.12, and 3.23.2) have been updated to reflect the revision of analytical reports for Cycle 11.

II. Discussion The increased radial peaking factor limits for Cycle 11 will accommodate a low radial leakage reload pattern. The low leakage pattern is intended to reduce the neutron fluence on critical reactor pressure vessel welds.

The proposed change to Technical Specification Table 3.23-2 provides Assembly Radial Peaking Factor and Total Radial Peaking Factor limits for Cycle 11 (Reload 0 fuel assemblies). The results of each Standard Review Plan (SRP) Chapter 15 event have been dispositioned, accounting for the proposed peaking factors for Reload 0 and for the third core reload with high thermal performance (HTP) spacer fuel. The results of this dispositioning are presented in Attachment 3, EMF-92-178. The minimum DNBR (MDNBR) and maximum linear heat rate (LHR) were calculated for those anticipated operational occurrences (AOOs) that were determined to need reanalysis. Since the highest powered assemblies are centrally located, where the core will be composed solely of HTP spacer fuel, the previously considered 2% mixed core MDNBR penalty is no longer necessary as it had been for the previous two cycles. The effect of the Cycle 11 configuration on fuel failures and radiological consequences from postulated accidents (PAs} was also assessed.

  • In general, MDNBR decreased and peak LHR increased for AOO events. The specified acceptable fuel design limits (SAFDLs) for AOOs are that: (1) the fuel shall not experience centerline melt, i.e., LHR to be less than 21 kW/ft, and (2) the DNBR shall have a minimum allowable limit such that there is a 95% probability with a 95% confidence interval that DNB has not occurred, i.e., DNBR to be greater than the ANFP correlation limit of 1.154. Although the MDNBR decreased and the peak LHR increased, the SAFDLs are not exceeded for any of the AOO events. The limiting AOO is the loss of forced reactor coolant flow event (SRP 15.3.1), which results in a predicted MDNBR of 1.178.

2 The only postulated accident results which are not bounded by previous analyses are reactor coolant pump rotor seizure (SRP 15.3.3), single rod withdrawal (SRP 15.4.3), fuel handling accident (SRP 15.7.4), and spent fuel cask drop accident (SRP 15.7.5). MDNBR decreases and peak LHR increases for both the reactor coolant pump rotor seizure and single rod withdrawal events.

Analyses predict MDNBR decreases and peak LHR increases for the single rod withdrawal event (an infrequent event). However, resulting MDNBR and peak LHR meet the SAFDLs for AOOs and are therefore within acceptance criteria for a postulated accident.

Analyses for the reactor coolant pump rotor seizure event predict a MDNBR slightly below the ANFP correlation limit of 1.154, with approximately 0.1%

of the fuel rods in the core predicted to fail. This relatively small amount of fuel failures would not hinder the ability to cool the reactor core. The only radiological release path to the environment for this event would be through a primary to secondary leak in the steam generators and subsequent release from the atmospheric dump valves after the associated plant trip. This release path would be similar to the bounding case for the control rod ejection event (SRP 15.4.8), which is predicted to result in 14.7% fuel failure. The resulting doses would therefore be much lower than those resulting from the control rod ejection event. The consequences of the reactor coolant pump rotor seizure event would thus be less than the SRP acceptance criteria of a small fraction of the 10 CFR 100 limits.

The radiological consequences of a fuel handling accident and a spent fuel cask drop accident were analyzed (EA-A-NL-91-169-01 and EA-A-NL-91-169-02) to consider the effects of increased fuel burnup in recent fuel cycles and increased radial peaking in Cycle 11. The results of the fuel handling accident analysis predict a maximum thyroid dose of 41.2 rem and whole body dose of 0.1 rem. The results of the spent fuel cask drop accident analyses predict a maximum offsite thyroid dose of 38.3 rem and a whole body dose of 0.2 rem. These offsite doses are less than those predicted by the previous analysis of record due to the use of dose conversion factors from ICRP 30, which are consistent with the latest revision to 10 CFR 20. The effect of the ICRP 30 dose conversion factors more than offsets the small effect of increased fuel burnup and increased radial peaking. Therefore, the radiological consequences of a fuel handling accident are well within the specified acceptance criteria.

3 III. Analysis of No Significant Hazards Consideration Consumers Power Company finds the activities associated with this proposed Technical Specifications change involve no significant hazards and accordingly, a no significant hazards determination per 10 CFR 50.92(c) is justified. The following evaluation supports the finding that operation of the facility in accordance with the proposed change would not:

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

The proposed change to the Technical Specifications increases the radial peaking factor limits for Cycle 11 (Reload 0 fuel assemblies).

This change is in core neutronics parameters due to changes in the fuel design and fuel management scheme. No changes to plant hardware (other than the new fuel) are involved. There are no associated changes in plant systems operating procedures or in instrument trip settings. Operation of the facility in accordance with the proposed Technical Specifications would, therefore have no effect on the way the plant systems are operated, or the way these systems would respond to postulated events. Therefore operation of the facility in accordance with the proposed Technical Specifications would not result in a significant increase in the probability of an accident previously evaluated.

The proposed change to the Technical Specifications increases the radial peaking factor limits for Cycle 11 (Reload O fuel assemblies).

The increased radial peaking limits for Cycle 11 caused the predicted minimum DNBR to decrease and peak linear heat rate to increase for anticipated operational occurrences. The MDNBR is predicted to remain above the ANFP correlation limit and the peak LHR is predicted to remain below the fuel centerline melt criteria for all AOO events.

Therefore, the consequences of all AOO events are within the specified acceptable fuel design limits.

All but four postulated accidents remain bounded by the previous analyses. The effect of increased radial peaking limits on MDNBR and peak LHR was assessed for the reactor coolant pump rotor seizure and single rod withdrawal events. The effect of increased radial peaking limits on radiological consequences was assessed for the fuel handling and spent fuel cask drop accidents.

The peak LHR, for the reactor coolant pump seizure, is predicted to remain below the fuel centerline melt criteria. The MDNBR is predicted to be slightly lower than the ANFP correlation limit for the reactor coolant pump seizure, resulting in failure of approximately 0.1% of the fuel rods in the core. With the only radiological release path to the environment being through a primary to secondary leak in a steam generator, the consequences remain bounded by other postulated accidents such as the control rod ejection event. The radiological consequences of a reactor coolant pump seizure would therefore be a small fraction of 10 CFR 100 limits.

4 The MDNBR, for the single rod withdrawal event, is predicted to remain above the ANFP correlation limit and the peak LHR is predicted to remain below the fuel centerline melt criteria.

The predicted radiological consequences for the fuel handling and spent fuel cask drop accidents are less than those predicted by the previous analyses of record. Though higher peaking factors are allowed, the use of dose conversion factors from ICRP 30, which are consistent with the latest revision to 10 CFR 20, results in lower predicted consequences.

Therefore the consequences of all events remain less than the acceptance criteria and operation of the facility in accordance with

  • the proposed Technical Specifications would not result in a significant increase in the consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident from any previously evaluated.

The proposed change to Technical Specification Table 3.23-2 increases the assembly and total radial peaking factor limits for Cycle 11 (Relo~d 0 fuel assemblies). This change is in core neutronics parameters due to changes in the fuel design and fuel management scheme. No changes to plant hardware (other than the new fuel) are involved. There are no associated changes in plant systems operating procedures or in instrument alarm or trip settings. Therefore operation of the facility in accordance with the proposed Technical Specifications would not create the possibility of a new or different kind of accident from any previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed change to the Technical Specifications increases the radial peaking factor limits for Cycle 11 (Reload 0 fuel assemblies).

The increased radial peaking limits for Cycle 11 caused the predicted minimum DNBR to decrease and peak linear heat rate to increase for anticipated operational occurrences. The MDNBR is predicted to remain above the ANFP correlation limit and the peak LHR is predicted to remain below the fuel centerline melt criteria for all AOO events.

Therefore, the consequences of all AOO events are within the specified acceptable fuel design limits.

All but four postulated accidents remain bounded by the previous analyses. The effect of increased radial peaking limits on MDNBR and peak LHR was assessed for the reactor coolant pump rotor seizure and single rod withdrawal events. The effect of increased radial peaking limits on radiological consequences was assessed for the fuel handling and spent fuel cask drop accidents.

The peak LHR, for the reactor coolant pump seizure, is predicted to remain below the fuel centerline melt criteria. The MDNBR is

5 predicted to be slightly lower than the ANFP correlation limit for the reactor coolant pump seizure, resulting in failure of approximately 0.1% of the fuel rods in the core. With the only radiological release path to the environment being through a primary to secondary leak in a steam generator, the consequences remain bounded by other postulated accidents such as the control rod ejection event. The radiological consequences of a reactor coolant pump seizure would therefore be a small fraction of 10 CFR 100 limits.

The MDNBR, for the single rod withdrawal event, is predicted to remain above the ANFP correlation limit and the peak LHR is predicted to remain below the fuel centerline melt criteria.

The predicted radiological consequences for the fuel handling and spent fuel cask drop accidents are less than those predicted by the previous analyses of record. Though higher peaking* factors are allowed, the use of dose conversion factors from ICRP 30, which are consistent with the latest revision to 10 CFR 20, results in lower predicted consequences.

Therefore, operation of the facility in accordance with the proposed change to the Technical Specifications would not involve a significant reduction in a margin of safety.

6 IV. Conclusion The Palisades Plant Review Committee has reviewed this Technical Specifications Change Request and has determined that the change involves no significant hazards consideration. This change has been reviewed by the Nuclear Performance Assessment Department. A copy of this Technical Specifications Change Request has been sent to the State of Michigan official designated to receive such Amendments to the Operating License.

CONSUMERS POWER COMPANY To the best of my knowledge, information and belief, the contents of this Technical Specifications Change Request are truthful and complete.

By ----;--~--------

David P Ho n, Vice President Nuclear Operations Sworn and subscribed to before me this 27tbday of January 199i.3.

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Notarygt;liC 1__~~ *

  • Van Buren County Michigan My commission expires 02/02/94
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