HL-1832, Application for Amend to License NPF-5,revising Traversing in-core Probe Sys Operability Requirements

Application for Amend to License NPF-5,revising Traversing in-core Probe Sys Operability Requirements

ML20082S428
Person / Time
Site:	Hatch
Issue date:	09/13/1991
From:	Beckham J GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20082S433	List: HL-1832, Application for Amend to License NPF-5,revising Traversing in-core Probe Sys Operability Requirements ML20082S441
References
HL-1832, NUDOCS 9109170137
Download: ML20082S428 (10)

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w. rem.u w u wf f i a !( ' . Il (.'i t' ~.1 liL-1832 002198 September 13 1991 U.S. Nuclear Regulatory Commission A11N: Document Control Desk Washington, D.C. 20555 PLANT HATCH - UNIT 2 NRC DOCKET 50-366 OPERATING LICENSE NPf-5 RLQVEST TO REVISE TEC1LNICALJiPElllCAll0NS 10 REVISE TIP OPERABILITY RE0V1Mhffilli Gentlemen:

In accordance with the previsions of 10 CfR 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company (GPC) hereby proposes a change to the Plant Hatch Unit 2 Technical Specifications, Appendix A to Operating License NPf-5.

The proposal involves a change to the Plant Hatch Unit 2 Specification for the Traversing in-Core Probe (TIP) system.

Specifically, the proposed change would require that three detectors be operable as opposed to the four required under the Technical Specificatton 3.3.6.6. Also, section c. of the applicability section is being deleted because the TIP system is no longer used to adjust APRM setpoints.

Enclosure 1 provides a detailed description of the proposed change and the circumstances necessitating the change.

Enclosure 2 provides the basis for our determination that the proposed change does not involve a significant hazards consideration.

Enclosure 3 provides page change instructions for incorporating the proposed change. The proposed Technical Specifications page follows Enclosure 3. Also included is the marked up page.

Due to its urgent nature, we request that this proposal be processed as an exigency amendment in accordance with 10 CfR 50.91(a)(6) with approval being granted no later than October 10, 1991. We also request that, once approved, the amendment be issued with an immediate effective date.

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Georgia power U.S. Nuclear Regulatory Commission September 13, 1991 Page Two in accordance with the requirements of 10 CFR 50.91, the designated state official will be sent a copy of this letter and all applicable enclosures. ,

Mr. J . 1. Beckham, Jr. states he is Vice President of Georgia Power Company and is authorized to execute this oath on behalf of Georgia Power Company, and to the best of his knowledge and belief, the facts set forth in this letter are true.

GEORGIA POWER COMPANY f

w BY:

. T. Beckham, Jr. [ /

Sworn to and subscribed before me this / gkdayofj(pl. 1991.

Alia Aufw hwJf 111cowumbbUYiit?bh${

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Enclosure:

cc: Georaia Power Crampm Mr. H. L. Sumner, General Manager - Nuclear Plant NORMS U.S. Nuclear Reaulatory Commission. Washinaton. D.02 Mr. K. Jabbour, Licensing Project Manager - Hatch E d. Nuclear Reaulltory Commission. Reaion 11 Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident inspector - Hatch State of Georaia Mr. J. D. lanner, Commissioner - Department of Natural Resources Fisi775

4 ENCLOSURE I PLANT HATCH - UNIT 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 RE0VESTJL REVISE TECHNICAL SPECILICATION_S 10 REVISE TIP OPERABill1Y REQUIREMENTS DESCRIPTION OF PROPOSED CHANGE Plant Hatch Unit 2 has four gamma sensitive Traversing in-Core Probe (TIP

- machines that are used to aeriodically determine the power distribution in the) core and -to calibrate tie Local Power Range Monitors (LPRMs). Each TIP machine consists of a detector which is sensitive to the local gamma flux, hardware to insert and withdraw the detectors, cables and signal )rocessing.

equipment. - As shown in figure 1, there are 31 TIP locations distri)uted in a symmetric _ radial pattern throughout the Hatch 2 core. Each TIP machine is used to measure the axial gamma flux distribution in seven or eight of these locations by inserting and then withdrawing the detector from the top of the

- core te the bottom. All four TIP machines can also traverse one common location in the center of the core in order to reconcile differences associated _with the various machines. The Unit 2 Technical Specifications require normalization of the TIP detectors ()rocess computer program ODI) once

_per_ l31 Effective Full Power Days (EFPD), w11ch presently requires using all four_ TIP machines. In addition, the Technical Specifications also recuire performing an .LPRM calibration once per 1000 Effective full Power fours (EfPH), which also requires an 001. Between ODis, LPRMs continuously monitor local power at four discrete axial heights in each of the 31 radial locations in the core. These instruments are used to todate the TIP distribution to account for axial power changes resulting from changes in rod patterns, fuel exposure, etc. The LPRMs are used to provide signal inputs for the Average Power - RangeMonitor(APRM) system, the Rod _ Bloct Monitor system (RBM),the process computer, and various alarms and monitoring functions in the control room.

On- September'8, 1991 during performance of rod maneuvers for t'he purpose of exchanging control rod sequences, it was discovered that the Plant Hatch Unit 2 'C' TIP machine would not index properly due to a problem apparently associated with the indexing mechanism. Correcting the problem requires access to the primary containment (drywell). With Unit 2 operating at 100%

power,- however, access is not possible at this time. .The present-Technical Specification requires four operable TIP machines for recalibration of the LPRM_ detectors every 31 EfPD. Performance -of an ODI within this period of timo- is necessary to maintain the validity-anc accuracy of the Periodic Core Performance Log (P1). PI is-the process computer program which calculates the

- Minimum Critical Power Ratio (MCPR), Linear Heat Generation Rate (LHGR) and Average Planar Linear- Heat Generation Rate (APLHGR). Inability to determine compliance with these thermal limits per Technical Specifications 3.2.1, 3.2.3

- and 3.2.4 would require reducing core thermal power to less than 25%.

002198 HL-1832 El-1 I

[NCLOSURE 1 (Continued)

EDESLIO _ R E V11LT.LOMIGLSIElf1IAL10RS 10 REVISE 11P OPERABillTY RE0VIREMENil This proposed amendment would decrease the number of required OPERABLE TIP detectors from four to three. The current BWR/4 Standard Technical Specifications require three Operable TIP machines. The proposed, improved BWR/4 Standard Technical Specifications (NUREG-1433) do not impose TIP system requirements. Performance of a full core normalization (0D'. ) can be accurately aerformed with only three detectors. This type of normalization is presently acing performed on Plant Hatch Unit I using methods provided by General Electric. Note that the Unit 1 Technical Specifications do not address TlP operability. Basically, performing 001 with only three TIP machines involves allowing an operating IIP machine to provide substitute traces for another TIP machine. This is valid when the following conditions are met

1) The control rod pattern is in an octant symmetric *A" sequence.

2) The total core TIP uncertainty is less than or equal to 8.7%.

When fuel bundles have been loaded in an octant symmetric pattern (with respect to fuel type and bundle average ano nodal exyosures) and the rod pattern is octant symmetric, the radial and axial power slapes will be similar in both halves of the core. As a result, there will be a high degree of symmetry in the TIP traces as well. Under these conditions, it is possible to substitute a TIP trace for a synenetric trace.

The current Hatch 2 Process Computer model has a " total core TIP uncertainty" of the statistical combination of the following uncertainties: 3.4% (LPRM correction uncertainty), 7.0% (analytical model uncertainty discussed in reference 1) and 2.2% (measured TIP uncertainty). The measured low value in TIP uncertainty is to be expected since Hatch is using ganuna detectors and geometry uncertainty components are expected to be very small. Statistically combining the above uncertainties yields a total TIP uncertainty of 8.1% which is below the 8.7% limit in reference 2. It should be noted that the measured TIP uncertainty could be as high as 3.8% and still meet the 8.7% lirr,it. It is concluded that the measured TlP uncertainty is well within the required limit.

Plant Hatch 2 started up in the "A" sequence and has remained in that sequence all cycle. The core was originally designed to be octant symmetric and has been operated in an octant symmetric manner throughout the cycle, in order to provide further assessment of operation without the "C" TIP machine, a simulation was performed to calculate the effect on thermal limits if a state point- obtained - before the inoperability of the "C" TIP machine _was recalculated using the symmetric pairs in place of the "C" machine locations.

The results from September 3, 1991 are shown below:

002198 HL-1832 El-2 t

t ENCLOSURE 1 (Continued)

RL0ESJ TO RLy15E 1ECHNICAL SfLtJflCAT10!iS 10 REVISE 11P OPERABlLITY RE0VIRIMENTS 80311LALJilAILE0U11J11tl NOMINAL STATEPCll{l SIMULA110N Of "C" llP FAILURL MCPR 1.473 1.470 KW/FT 13.960 13.990 MAPLHGR 12.130 12.120 A statistical analysis of the two 3D power distributions resulting from the above analysis showed a nodal uncertainty of 2.4% and a bundle uncertainty of 0.9%. As stated above, TIP uncertainty, as determined from Hatch-2 cycle 10 data, is 2.2%. This analysis indicates that the core is operating in a symmetric manner and that use of the TIP substitution methodology has an insignificant effect on the calculation of thermal limits.

Calibration of LPRMs with substitute TIP traces does not change the function of the LPRMS or other plant systams (e.g. APRMS, RBM) that use LPRM signals for input. Therefore, these systems will continue to accurately assess the power and thermal limits in the core.

In addition, section c. of the applicability section of Specification 3.3.6.6 is being deleted since the TIP system is no longer used for the readjustment of APRM gains or setpoints. Amendment 39, approved by the NRC in July of 1984, implemented the APRM/RBM Technical Specification (ARIS) improvement program and removed the section on APRM setpoints. This is an administrative change.

As stated previously, the need to have the proposal processed as expeditiously as possible is due to the fact that, after 31 EfPDs, we will be unable to use all four TIP machines for the calibration of LPRM detectors or for the calculation and monitoring of thermal limits. This will necessitate taking the ACTION required by Hatch Unit 2 Technical Specifications 3.2.1, 3.2.3 and

- 3.2.4 (reduce thermal power to less than 25%). Operating at the present power levels, 31 EfPD with the 25% surveillance grace period added will be reached on October 10, 1991. This is based on the date of the last successful full core TIP normalization, September 3,1991.

This problem with the "C* TIP could not have been foreseen because prior to September 8, all four TIPS had been operated successfully.

If a forced outage is encountered on Unit 2, the indexing mechanism will be repaired prior to restart, if a forced outage does not occur prior to completion of the current o>erating cycle, the problem will be corrected no later than the end of the scleduled Unit 2 Fall 1992 refueling outage.

002198 HL-1832 El-3

4.

ENCLOSURE 1 (Continued) fl[Qu[ST 10 REYlSE TECHNICAL SE[CXICallMS 10 REVISE TIP OfERABill1Y RE0VIREMEl{lS BffERENCES

1) letter, C.J. Paone to K.S. Folk, " Analytical Model Uncertainty for Hatch",

CJP: 91-275, September 11, 1991.

2) NEDE-240ll-P-A-10. " General Electric Standard Application for Reactor fuel," February, 1991.

002198 HL-1832 El-4

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PLANT HATCH - lWII 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 RLQEST TO REVISE TLClithCAL SPECIE 1 cal 10til 10 REVISL TIP OPERABill1Y REDVIREMENTS -l The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license does not involve a significant hazards consideration if operation of the facility in accordance with the )roposed amendment would not: .

1) involve a significant increase in the probasility or consequences of an accident previously evaluated, 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in the margin of safety.

Georgia Power Company has reviewed the proposed amendment and determined that its adoption would not result in a significant hazards considerativn. ,

Huli_fgr Proposed No Sicnificant Hazards ConildttgLlMLEEl.ermination:

The change does not involve a significant hazards consideration for the t following reasons:

1) The )roposed amendment does not involve a significant increase in the proba)ility or consequences of an accident previously evaluated.

The TIP system is not used to mitigate the consequences of or prevent any accident, nor are assumptions made in any accident analysis relativo to the operation of the TIP system. Implementation of this proposed change will not change the function of any plant systems needed to prevent or mitigate the consequences of postulated accidents. Therefore, reducing the number of required Operable TIP machines f om four to three and using substitute TIP traces for the calibration of LPRMs and the monitoring of thermal limits' does not increase the probability of occurrence of a previously evaluated accident.

The change in power distribution determination in the process computer does not affect the consequences of anticipated operational occurrences (transients) described in the FSAR since the MCPR safety limit is not violated during the events. Provided the control rods are positioned in an "A" sequence and the total core TIP uncertainty for the cycle is-less than or equal to 8.7%, neither the MCPR operating limit nor the safety limit need to be increased. The 8.7% uncertainty factor is the number used in the MCPR safety limit analysis (NEDE-240ll-P-A-10, General Electric Standard Application for Reactor fuel," february, 1991). The current total core TIP uncertainty has been determined to be 8.1%, which does not exceed the 8.7% requirement.

002189 HL-1832 E2-1

l ENCLOSURE 2 (Continued)  !

E0ESL10 REVISE TIGNICALSIIClflCAT10NS 10 REVISE TIP OPERABill1Y RE0Q182[1115 [

10 CFR S0.92 EVALUATIQN Hatch Unit 2 has been operating in the octant symmetric "A" sequence  !

since the beginning of the cycle. To provide an assessment of operating ,

with the "C" TIP machine out of service, a simulation was performed to  :

calculate the affect on thermal limits if a state point obtained before the inoperability of the "C" TIP was recalculated using the symmetric pairs in place of the "C" machine locations. The results of this i simulation (shown in Enclosure 1) indicate that the core is operating in a ,

highly symmetric manner and that use of the substitute TIP readings will r have a minimal affect on thermal limit calculations. Hatch Unit 2 will continue to be operated in the "A" sequence for the duration of the "C" l TIP outage. Plant procedures will be revised to reflect this. '

Therefore, since the total core TIP uncertainty is acceptable and operation of Hatch Unit 2 will continue in the "A" sequence throughout the [

duration of the "C" TIP outage, reducing the number of required Operable .

TIP machines from four to three does not decrease the margin of safety to

'Se MCPR operating and safety limits and the radiological dose consequences for previously analyzed accidents are not increased.  ;

The proposed change which removes the reference to the APRM setpoint is an administrative change. It reficcis the fact that we no longer adjust the APRM trip or the APRM gain for high peaking factors. This change was made ,

in 1984 and was done as part of the APRH/RBM Technical Specification (ARTS) improvement program. Since neither plant operation nor equipment

• is being affected, this change does not increase the probability of r occurrence or the consequences of a previously evaluated accident. .

2) The proposed amendment does not create the possibility of a new or  !

different kind of accident from any accident previously evaluated. .

Using substitute TIP traces and changing the Hatch 2 lechnical .

Specifications such that the TIP system is operable with three movable  !

detet. tors does not change the basic operation of the plant. Nor does it  !

change the operation of any safety related plant equipment.

• Although the Process Computer will be operating differently in the  :

calculation of core thermal limits, the difference only involves the  !

assignment of incoming data to various arrays for the calculation of nodal  :

powcrs, thermal limits, etc. Furthermore, the process computer is not ,

required for the safe shutdown of the plant nor is it used for the i mitigation of consequences of accidents. Therefore, changing this i

I 002189 i HL-1832 E2-2 ,

1

ENCLOSURE 2 (Continued)

REQUEST 10 ALVISE TEGNICAL SPECIFICATIMS 10 REVISE 11P OPERABIL11Y RE00lREMENTS 10 CFR E0.92 EVALVATIM Tecinical Specification such that the TIP system is operable with three TIP machines does not increase the likelihood of an accident occurring different from any analyzed in the FSAR.

The proposed change removing the reference to APRM setpoint adjustment is administrative in nature, reflecting how the plant is actually operated.

No changes to plant equipment or operation result from it, therefore, the probability of any accident occurring is not increased.

3) The proposed amendment does not result in a significant reduction in the margin of safety.

The margin of safety for some of the accidents analyzed in the FSAR is the Technical Specification fuel cladding integrity (MCPR) safety limit. This safety limit ensures that at least 99.9% of the fuel rods in the core will avoid transition boiling during an anticipated operational occurrence (transient). As documented in General Electric Generic Licensing lopical Report, GESTAR-II, the MCPR safety limit is based, in part, on a statistical combination of uncertainties in key parameters, including total core TIP uncertainty. As long as the total uncertainty is less than or equal to what was used to calculate the original MCPR safety limit (8.7%), the margin of safety is unchanged. Substitute TIP traces can be used to monitor thermal limits and calibrate LPRMs only if the core is loaded symmetrically and is operating with a symmetric, "A" sequence rod pattern.

The margin of safety is not reduced as a result of using this method because we have shown that the total core TIP uncertainty is less than 8.7% and the Hatch Unit 2 cure is being operated in the "A" control rod sequence. Unit 2 will continue to be operated in the "A" rod sequence at least until the return of the "C" TIP machine to service. Plant procedures will be revised to reflect this.

The proposed change to eliminate reference to the APRM setpoint adjustment is administrative in nature. No- changes to plant equipment or plant w operation results, thus the margin of safety is not reduced.

002189 IIL-1832 f2-3 1

ENCLOSURE 3 PLANT HATCH - UNIT 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 RE0 VEST TO REYlSLIECRNICAL SPECIFICATIONS TO REVISE TIP OPERABILITY RE0VIREMENTS PAGE CHANGE INSTRUCTIONS VNIT 2 Remove Pagg Insert Page 3/4 3-57 3/4 3-57 1

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002189 HL-1832 E3-1

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