ML20085F273
| ML20085F273 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/10/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20085F258 | List: |
| References | |
| NUDOCS 9110220105 | |
| Download: ML20085F273 (7) | |
Text
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SAFETY EVALUATION RY THE OFFICE Of NUCLEAR REACTOR REGHLATION RELATED TO AMENDMENT NO. 115 TO FACILITY OPERATING LICENSE 5
EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO. 2 DOCKET NO. 50-366 1.0 ljNTRODUCTION By letter dated September 13, 1991, as supplemented Septemoer 30, 1991, Georgia Power Company, et al. (GPC or the licensee), requested to change Technical Specification (TS) 3.3.6.6.
The September 30, 1991, letter modified the TS such that the reduction in detectors would apply to Cycle 10 only. This change was within the scope of the action noticed and did not change the initial proposed no significant hazards consideration determination. Specifically, the proposed changes would require that three traversing incore probes fTIP) detectors be operable only for Cycle 10 as opposed to four which are currently required. The changes will permit data from operable symmetric TIP w asurement locations to be substituted in the inoperable locations.
During a recent performance of rod maneuvers for the purpose of exchanging control sequences, it was discovered that the Hatch Unit 2 "C" TIP machine would not in(ex properly due to a problem apparently associated with the irdexing machine.
Current TS 3.3.6.6 requires that all 31 TIP ineasurement locations be operable for the TlP systems to be operable for required periodic power distribution measurements. Thus, the reactor would have to be shut down for the required repair since the repair cannot be performed at power. The proposed TS change is ir. tended to avoid such a shutdown, and only for Cycle 10, when suitable backup infcrmation is available.
The licensee stated, in its September 13, 1991, submittal that the problem will be corrected at the earliest shutdown, which will be no later than the end of the scheduled Unit 2 Fall 1992 refuelina outage.
2.0 EVALUATION I
2.1 Core Symetry Hatch Unit _2 has four gama sensitive TIP machines that are used to periodically determine the power distribution in the core and to calibrate the Local Power Range Monitors (LPRMs). There are 31 TiP locations distributed in a symetric radial pattern throughout the Hatch 2 core. All four TIP machines can transverse one comon location in the center of the core in order to reconcile differences I
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2-h associated with the various machines.
During normal operation with a symetric control rod pattern, the core power distribution is correspondingly symmetric, and symmetric TIP measurement locations provide similar information to within statistical differences which are accounted f or in safety analyses and measured i
in the course of cycle startup tests, it is normal, approved practice to translate by symmetric transfers the information from measured locations to unmeasured locations when calculating, via the measurements and the process computer, the core power distribution.
When fuel bundles have been loaded in an octant symmetric pattern, and the rod pattern is octant symmetric, the radial and axial power shapes will be simlar in both halves of the core. As a result, under these normal operating circumstances it would be acceptable to similarly supply data from operating symetric locations to replace inoperable T'P location information.
2.2 TIP Statistical Uncertainty The current Hatch ? process computer model has a " total core TIP uncertainty" comprised of a combination of LPRM, model, and TIP uncertainties. The licensee i
analyses showed that a low value in the TIP uncertainty (2.2 percent) is to be expected, since Hatch is using gamma detectors and geometry uncertainty components are expected to be small. Statistically combining the above uncertainties yields a total TIP uncertainty of 8.1 percent which is below the 8.7 percent limit referred to in the approved Topical Report NEDE-240ll-P-A-10. " General Electric Standard Application for Reactor Fuel," dated February 1991. The submitted analyses show that the measured TIP uncertainty is well within the required-limits.
2.3 Effect of Operation Without the "C" TIP Machine on Thermal Limits t-Hatch Unit 2 has been operating in the octant synmetric "A" sequence since the beginning of this cycle-(Cycle 10).
In assessing the impact of the inoperable "C" TIP machine (or the absence of any one TIP machine) on the thermal limits, the licensee performed a simulation to determine if data obtained before the
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inoperability of the "C" TIP machine could be regenerated using symmetric pairs in place of the "C" machine locations. The results of the simulation showed that there is less than 0.2 percent difference in the Minimum Critical Power Ratin (MCPR), the linear heat generated, and the fiaximum Axial Planar Linear Heat Generation Rate (MAPLHGR) calculations. This strongly suggests that the core is indeed operating in a highly symmetric configuration, and that the use of the substituted TIP readings will have a minimal effect on the thermal limit calculations. Further analyses indicated that the 30 power distributions have a nedal uncertainly cf 2.4 percent and a fuel bundle uncertainty of 0.9 percent.
LPRMs calibrated with substituted symmetric pairs will not impact the function of the LPRMs or any other instrument system (e.g., Average Power Range Monitor (APRM), Rod Block Monitor (RBM)) that use the LPRM signals as input. Moreover, the licensee will continue to operate the reactor in an octant symmetric core and a resulting cross core (diagonally) symmetric measurement location pattern; i
and the total core TIP uncertainty for the cycle will be less than P 7 percent (standard deviation). Consequently, these systems will continue to accurately assess the power and thermal limits in the core.
i Hatch Unit 2 intends to repair the present TIP inoperability problem at the first opportunity arising from shutdown for other causes.
Thus, this change is approved only for Cycle 10 and the NRC staff finds it acceptable.
With regard to tection C of the applicability section of TS 3.3.6.6, the licensee requested that this section be 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 APRH/RBM Technical Specifications ( ARTS 1 improvement program and removed the section on APRM setpoints. Thus, this change is administrative and is acceptable.
3.0 EX1 GENT CIRCUMSTANCES The Commission's regulations, 10 CFR 50.91, contain provisions for issuance of amendments when the usual 30-day public notice perioJ cannot be met. One type of special exception is an exigency.
An exigency is a case where the Commission and licensee need to act promptly and that time does not permit the Commission to publish a Federal Register notice allowing 30 days for prior public comment, and it is determined ~that the amendment involves no significant hazards consideration.
Under sucn circumstance, the tommission notifies the public in one of two ways: by issuing a Federal Register notice providing an opportunity for hearing and allowinf at least two weeks for prior public comments, or by issuing a press release discussing the proposed changes, using the local
- media, in this case, the Commission used the first approach.
The licensee submitted the request for an amendment on September 13, 1991. It was noticed in the Federal Register on September 24, 1991 (56 FR 48218), at which time the staff proposeo a no~significant hszards consideration determination. The licensee requested that the amendment be issued no later than October 10, 1991.
The licensee stated that on September 8, 1991, during performance of rod maneuvers for the purpose of exchanging control rod sequences, it was discovered that the Hatch Unit 2 "C" TIP machine would not index properly due to a problem apparently associated with the indexing mechanism.
ccrrecting the problem requires access to the primary containment (drywell).
However, with Unit 2 operating at 100% power, access is not possible at this time. The present TS requires four operable TIP machines for recalibration of the LPRM detectors every 31 Effective Full Power Days (EFPD). Performance of the core map within this period of time is necessary to maintain the validity and accuracy of the Periodic Core Performance Log (P1). 01 is the process computer program which calculates the MCPR, Linear Heat Generation Rate (LHGR), and Average Planar Linear Heat Generation Rate (APLHGR).
Inability to determine compliance with these thermal limits per TS 3.2.1, 3.2.3, and 3.?.4 would require reducing core thermal power to lass than 25%.
. 4.0 FINAL HO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Comission's regulations in 10 CFR 50.92 state that the Commission ray make a final determination that a license amer.dment involves no significant hazards ennsideration if operation of the facility, in accordance 11th the amendment, would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated, or (?) Create th0 possibility of a new or different kind of accident from any acci'ent previously evaluated; or (3)' Involve a significant reduction in margin of safety. As required by 10 CFR 50.91(a), the licensee has providsc its analysis of the issue of no significant hazards consideration, wlich is presented below:
1.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The TIP system is not used to mitigate the consequences of or prevent any acc1 dent, nor are assumptions made in any accident analysie relative to the operation of the TIP system, implementatien 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 from four to three and using substitute TIP traces for the calibration of LPRfis 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 tiCPR safety limit is not violated during the events. Provided the control rods are positioned in an "A" sequence and tho 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 MCPP, safety limit analysis (hEDE-24011 PrA-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.
Hatch Unit 2 has been operating in the octant symetric "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 fe]ffect 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 simulation [shown elsewhere in the licensee's submittal dated September 13,1991], indicate that the core'is operating in a highly symmetric manner and that use of the substitute TIP readings will 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" 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 the 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 reflects the fact that we [the licensee] no longer adjust the APRt1 trip or the APRM gain for high peaking factors.
This change was made in 1984 and was done as part of the APRM/RM [ Rod Block Monitor] Technical Specification (ARTS) improvement program.
Since neither plant operation nor equipment is being affected, this change does not increase the probability of occurrence of 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 ? Technical Specifications such that the TIP s/ stem is operable with three movable detecturs does not change the basic operation of the plant. Nor does it change the operation of any safet) related plant equipment.
Although the Process Computer wil' be operating differently in the calculation of core thermal limit 1, the difference only involves the assignment of incoming data to virious arrays for the calculation of nodal powers, thermal limits, etc. Furthermore, the process computer is not required for the safe shutdown of the plant nor is is used for the mitigation of consequences of t.ccidents. Therefore, changing this Technical Specification such that the TIP system is operable with three TIP mach %es 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, thc 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 Topical Report, GESTAR-II, the MCPR safety limit is based, in part, on a statistical combination of uncertair+ies in key parameters, including total core TIP uncertainty. As long as the total uncertainty
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. 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 LpRits only if the core is loaded sywetrically and is operating with a symetr ic, " A" sequence red pattern.
The margin of safety is not reduced as a result of using this method because we [the licensee) have shown that the total core TIP uncertainty is less than 8.7% of the Hatch Unit 2 core is being operated in the "A" rod sequence. Unit 2 will continue to be operated in the "A" rod sequence at least until the return nf the "C" TIP machine to service.
Plant procedures will he revised to reflect this.
The proposed change to eliminate reference to the APRM setpoint adjustment is administrative in nature. No changes to plant equiprent or plant operation results, thus the margin of safety is not reductd.
Based upon the above considerations, the NRC staff concludes that the amendment meets the three criteria of 10 CFR 50.92. Therefore, the staff has made a final determination that the proposed amendment does not involve a significant hazards consideration.
5.0 SJ1QECONSULTATION In accordance with the Comission's regulations, the Georgia State official was notified of the proposed issuance of the amendment. The State official had no coments.
6.0 ENVIRONMEllTAL CONSIDERATI0tl The amendment changes requirem nis 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 amendment involves no significant.ncrease in the amcunts, 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 Comission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public coment on such finding (56 FR 48218). Accordingly, the amendment meets the eligibility criteria for categorical exclusion-set forth in 10 CFR-51.?2(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be preparer: in connection with the issuance of the amandment.
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4 4 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) srch activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendnent will not be it.imical to the comon defense and security or to the health and >afety of the public.
Principal Contributor:
A. C. Attard, SRXT/ DST Date: October 10, 1991
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