ML19296D240

Safety Evaluation Supporting Amend 43 to License DPR-22

ML19296D240
Person / Time
Site:	Monticello
Issue date:	02/12/1980
From:	Office of Nuclear Reactor Regulation
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ML19296D239	List: ML19296D238 ML19296D240 ML19296D241 ML19296D242
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NUDOCS 8003030061
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 43 TO PROVISIONAL OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 1.0 Introauction By letter (l) dated SeptemDer 17, 1979 and supplemented by letters (2, 3,4) dated November 1,1979, November 30, 1979 and Decemoer 31,1979, the Northern States Power Company (the licensee) requested amendment to the Tech-nical Specifications appended to Operating License No. DPR-22 for tne Monticello Nuclear Generating Plant (MNP). The proposed changes relate to the seventh refueling (Cycle 8) of MNP, involving the replacement of 100 exposed unpressurized, single water rod (8x8) fuel assemblies, loaded in the initial core, with a like number of fresh, prepressurizea, two water rod (P8x8R) fuel assemblies. The reload assemblies are designed and fabricated by General Electric Company (GE). Additionally, the licensee has proposed to reinsert for a sixth operating cycle, a total of eight type BDB262 8x8.

fuel assemolies. These assemblies are part of a joint NSP/GE/ DOE extended exposure fuel program whose purpose is to determine the impact of higher fuel burnup on fuel reliability. Because the exposure to be achieved by these bundles during Cycle 8 will exceed that described in the current approved MNP licensing basis, supplementary safety analyses and changes to the MNP Technical Specifications have been sub:nitted for our review and approval.

In support of the proposed reload application fuel exposure program and Technical Specification changes,(5$ extended the licensee has provided a supplemental reload licensing document,(6) an extendea exposure fuel program licensing document,(7) and other supplemental information(8, 9,10,11) related to the-higher burnup assemolies.

This refueling (Reload 7) is the first for MNP to incorporate GE's pre-pressurized P8x8R fuel design on a batch reload Dasis. The description of the nuclear and mechanical design of the Reload 7 P8x8R fuel and the exposed unpressurized 8x8 and 8x8R fuels, used in the most recent reloads, is contained in GE's generic licensing topical report for BWR reloads.(12)

Reference 12 also contains a complete set of references to tcpical reports which describe GE's analytical models and methods for the nuclear, thermal-hydraulic, transient and accident calculations cerformed for this reload, together with information on the applicability of these methods to cores containing a mixture of different GE fuel designs. Portions eooaono(4'

. of the plant-specific data, such as operating conditions and design parameters, which are used in transient and accident calculations, have also been included in the approved topical report.

Our safety evaluations (13,14) of GE's reload licensing topical report and topical report amencment concluded that the nuclear and mechanical design of P8x8R fuel used in this reload and GE's analytical methocs for the nuclear, thermal-hydraulic, transient and accident calculations, as applied to cores containing a mixture of fuel types, are accept-able. Our acceptance of the nuclear and mechanical design of the standard 8x8 (one water rod) fuel was expressed in the staff's evaluation (15) of the information in Reference 16.

As part of our evaluation (13) of Reference 12, we found the cycle-indepenaent input data to be used for the reload transient and accident analyses for MNP to be acceptable. Supplementary cycle-depencent information and input data are provided in Reference 6, which follows the format and content of Appendix A of Reference 12.

As a result of the staff's generic evaluations (13,14) of safety considerations related to the use of P8x8R fuel in mixed core loadings with 8x8R and 8x8 fuel, only a limited number of additional review items for the replacement fuel and reconstituted core need to be addressed in this evaluation. These include the plant and cycle-specific analysis input data and analysis results presented in Reference 6, and those items ioentified in Reference 13 as requiring special attention during BWR reload reviews.

The additional review items related to the new fuel and Cycle 8 re-constituted core, are addressed in the first part of this evaluation (Section 2.1), while our evaluation to extend the exposure of the eight higher burnup fuel assemblies is contained in the second part of this evaluation (Section 2.2).

2.0 Evaluation 2.1 Reload Fuel And Reconstituted Core 2.1.1 Nuclear Characteristics For Cycle 8', 44 fresh prepressurized type P8dRB265L fuel bundles and 56 fresh prepressurizea type P80RB282 fuel bundles will be loaded into the core. The remaincer of the fuel bundles in the reconstituted core will be a comDination 8x8 and 8x8R fuel bundles exposed during the previous operating cycles. The fresh fuel will De loaded and the pre-viously peripheral fuel will be shuffled inward so as to constitute an octant-symmetric core pattern, which is acceptable. Based on the

. data proviced in Sections 4 and 5 of Reference 6, both the control rod system and the standby liquid control system will have an acceptable shutcown capability curing Cycle 8.

2.1.2 Thermal-Hydraulics 2.1.2.1 Fuel Cladding Integrity Safety Limit MCPR As stated in Reference 12, for BWR cores which reload with GE's P8x8R fuel, the allowable minimum critical power ratio (MCPR) resulting from either core-wide or localized abnormal operational transients is equal to 1.07.

When meeting this MCPR safety limit during a transient, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The 1.07 safety limit minimum critical power ratio (SLMCPR) to be used for Cycle 8 is unchanged from the SLMCPR oreviously approved for Cycle 7.

The basis for this safety limit is adcressed in.

Reference 12, while our generic approvals are given in References 13 and 14.

2.1.2.2 Operating Limit MCPR Various transient events can reduce the MCPR from its normal operating value. To ensure that the fuel cladding fi::egrity safety limit MCPR will not be violated during any abnormal operational.

transient, the most limiting transients have been reanalyzed for this reload, in orcer to determine which event results in the largest reductiion in the minimum critical power ratio. These events have been analyzed for both the exposea 8x8 and 8x8R fuel and the fresh P8x8R fuel. Addition of the largest reduction in critical power ratio to the safety limit MCPR establishes the operating limit MCPR for each fuel type.

2.1.2.2.1 Abnormal Operational Transient Analysis Methods The generic methods used for these calculations, including cycle-independent initial conditions and transient input parameters, are described in Reference 12. Our acceptance of the cycle-independent values appears in Reference 13. Additionally, our evaluation of the transient analysis methods, together with a

4 description and summary of the outstanding issues associated with these methoas, appears in Reference 13.

Supplementary cycle-dependent initial conditions and transient input parameters used in the transient analyses appear ip the tables in Sections 6 and 7 of Reference 6.

Our evaluationtl4 has also addressed the methods used to develop these supplementary input values.

At the time we conpleted our evaluation of the generic methods, the acceptability of the GEXL critical power correlation,(7) for use in connection with the retrofit fuel design, had not been acequately documented by GE. The staff found, however, that the then available 8x8R critical power test data was sufficient to support *,he accept-ability of GE's 8x8R fuel cesign for BWR core reloads for one operating cycle. Accordingly, we stated (13) that future BWR core reload applications involving retrofit 8x8 fuel for a second operating cycle would have to incluoe additional information which adequately justified the correlation for application to 8x8R fuel operating beyond one cycle. Setsequent.to our approval of Reference 12, GE pro-vided a report (18) to the staff on this matter, together with additional information(19) intended to justify the adequacy of the GEXL corre-lation for application to the retrofit fuel over its design lifetime.

Reference 18 provides the results of full scale critical power tests performed with 8x8R fuel bundles. The tests, which included both transient and steady-state sinulations, followed the same approvea procedures used for the standard 8x8 (singla i >,ter rod) and 7x7 (all fueled rods) fuel designs. The analysis or a total of 577 steady-state data points was performed using methods previously approved by the staff. The data, involving nine test asse111es which spanned a range of local power peaking and flow conditions showed, according to GE, that the GEXL correlation was applicabie to the 8x8R fuel if small adjustments were made to the additive constants used in the fornulation of the rod-by-rod R-factors.

The local power peaking dependent R-factors, are based on the new additive constants shown in Figure 3-1 of Reference 18. These constants were also used for the MNP Reload 6, 8x8R critical bundle power pr? dictions. Using these new additive constants, GE performed a data analysis to assess the accuracy and precision of the GEXL corre-l ati on. The results of this analysis showed that the correlation fit provides for a mean predicted-to-measured critical power ratio of 0.9879 with a standard deviation of 0.0234.

. When viewed over the range of its applicability (which is the same as the standard 8x8 fuel), the GEXL correlation is therefore somewhat conservatively biased while the statistical variation between the predicted and measured critical power is somewhat less than that associated with the standard 8x8 assemDly,(17) 1.e., 2.34%

vs 2.8%.

Thus, when viewed over its range of applicability, the 8x8R GEXL correlation (with new additive constants) is a somewhat better predictor 8x8R critical bundle powers than the previously approved 7x7 and 8x8 GEXL formulations are for predicting 7x7 and 8x8 critical bundle powers respectively. Furthermore, from these results it may also be concluded that the 3.6% standard deviation and best estimate' assumption of the GEXL correlation (which were actually used in the GETAB statistical analysis to derive the 1.07 safety limit MCPR) bound the statistical characteristics associated with the subject 8x8R GEXL correlation.

The additional information furnished by GE is also intended to be applicable to all BWR cores which contain 8x8R fuel. Accordingly, this information is also being reviewed by the staff generically.

Although our generic evaluation is not yet coglete, based on our review to date, we believe that for the range of testing, the 8x8R GEXL correlation has an acceptability and applicability which is equivalent to the 7x7 and 8x8 GEXL correlations previously approved by the staff.

To date from our review of the subject data, we have observed that for those critical power test conditions specifically represent-ative of second cycle fuel operating at typical normal operating thermal-hydraulic state points, the correlation is somewhat non-conservative in its predictions. This observation focuses on a correlation behavioral concern not explicitly addressed in the generic approval of the overall GETAB methods approved (20) for the 7x7 and 8x8 fuel types. However, catil our generic review is complete, we believe that during Cycla 8, there is sufficient conservatism implicit in the generic detemination of the 1.07 safety limit MCPR to offset a possible non-conservatism associated with this concern. That is, specifically, the generic GETAB statistical analysis assumed a 3.6% correlation uncertainty while GE's analysis of the 8x8R test data results in 2.34% standard deviation. Additionally, the generic evaluation considered an all 8x8R equilibrium core, whereas the Cycle 8 MNP core involves 8x8,

8x8R and P8x8R, fuel in a non-equilitrium condition.

In view of these conservatisms (which are representa11ve of a typical non-equilibrium 8x8R reload core) we believe that tne overall thermal-hydraulic (GETAB) methods are acequate for establishing conservative MCPR operating' limits for Cycle 8 of MNP. However, as 8x8R equilibrium conditions are approached, this conservatism will diminish.

In order that this conservatism not be substantially eroded with future reload cycles, this issue should be addressed for the next reload of MNP.

2.1.2.2.2 Abnormal Operational Transient Analysis flesults The transient events analyzed for Cycle 8 were:

reactor coolant system pressurization (turbine trip without bypass and feedwater

- controller failure), feeWater tenperature reduction (loss of 100*F feedwater heating) and local reactivity insertion (control rod _ withdrawal error). The licensee reports that the most limiting everit in the above categories for the exposed unpressurized 8x8 and 8x8R fuel assemblies and the fresh prepressurized P8x8R fuel assemblies is the turbine trip without bypass. This transient results in CPR reductions of 0.34 for both the unpressurized 8x8 and 8x8R fuel and 0.37 for the prepressurized P8x8R fuel. Addition of these aCPRs to the 1.07 SLMCPR establishes fuel type-dependent operating limit MCPRs (i.e.1.41 for the unpressurized 8x8 and 8x8R fuel and 1.44 for the prepressurized P8x6R fuel). These MCPR operating limits are sufficient to assure that the SLMCPR will not be violated even if any of the aforementioned events were to occur during Cycle 8.

The licensee has also considered the effects of the worst possible fuel loading errors on bundle CPR. The results of these analyses are discussed in Section 2.1.3.3, herein.

2.1.2.3 Fuel Cladding Integrity Safety Limit LHGR The control rod withdrawal error and fuel loading error events were also reanalyzed by the licensee for Cycle 8 to determine the maximum transient local linear heat generation rates (LHGRs).

The results of these analyses show that the fuel type-dependent and exposure-dependent safety limit LHGRs, shown in Table 2-3 of Refer'ence 12, will not be violated should either of thesa limiting local reactivity events occur. Thus, fuel failure due to ex.essive cladding strain will be precluded. We find these refaits, which adequately account for the local effects of fuel dmsification power spiking, to be acceptable.

. 2.1.3 Accident Analysis 2.1.3.1 ECCS Appendix K Analysis On December 27, 1974, the Atomic Energy Comission issued an Order for Modification of License, iglementing the requirements of 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reators." One of the requirements of the ?rder was that prior to any license amendment authorizing any core reloading... "the licensee shall submit a re-evaluation of ECCS performance calculated in accordance with an acceptable evaluation model which conforms to the provisions of 10 CFR Part 50.46."

The Order also required that the evaluation shall be accompanied by such proposed changes in Technical Specifications of license amenaments as may be necessary to iglement the evaluation assumptions and results.

Previously, for Cycle 7, the licensee reevaluated the adequacy of MNP ECCS performance in connection with the unpressurized retrofit 8x8 reload fuel cesign. The methods used in this analysis were previously approved by tne staff. For Reload 6, we reviewed the ECCS analysis results suomitted by the licensee for the Cycle 7 reload fuel and concluded that MNP would be in conformance with all of the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 when operated in accordance with the 8x8R MAPLHGR versus Average Planar Exposure values which appeared in the proposed plant Technical Specifications. Except for prepressurization, the design of the Reload 7 fuel is the same as the Reload 6 fuel.

In Reference 14, we stated that LOCA analyses pre-viously perfonned and accepted for unpressurized 8x8R fuel are conservatively bounding for prepressurized fuel of that type (enrichment). Accordingly it is acceptable to utilize the previously approved 8x8R MAPLHGR vs Average Planar Exposure technical specifi-cation limits for the reload ?8x8R fuel.

2.1.3.2 Control Rod Drop Accident, For Cycle 8 the licensee reevaluated the worst case control rod drop acci cent. The methods described in Reference 12 for plants with Bank Position Witharawal Sequence (BPWS) control were usec. For plants

-8 with BPWS, generic analyses show that for worst case conditions the peak fuel enthalpy will remain below the 280 cal /gm design limit if the maximm incremental control rod worth does not exceed 1.0% a k.

During Cycle 8 the maximm incremental control rod worth is 0.9% a k.

Accordingly, the peak enthalpy design limit ~

will not be exceeced by a control rod drop accident which occurs from any in-sequence control rod movement.

2.1.3.3 Fuel Loading Error For Cycle 8 the licensee reanalyzed postulated fuel loading errors invo'.ving both misoriented and mislocated bundles. The analysis methods for.the fuel loading error are described in Reference 12 and are approved in Reference 13. For this reload cycle the limiting fuel loading error CPRs were calculated by conservative and older GE analysis methods. For similar plants, FLE analyses have indicated a conservatism of as much as 50% in CPR when the old approved analysis results are compared with the more recently approved GE analysis. Even tnough we recognize a significant conservatism in the older analysis methods the conservatism cannot be quantified for this specific plant and cycle. Accordingly, in the absence of a new analysis for Cycle 8 utilizing thd newer metnods, the aCPR for the fuel loading error event must be based on the conservative results of the older analysis methods.

The predicted changes in CPR, when added to the 1.07 safety limit MCPR would result in MCPR operating limits of 1.42 for the 8x8 fuel and 1.46 for the 8x8R and P8xBR fuel. These results indicate that in the event of the most severe fuel loading error at MNP some of the fuel rods in the bundle could experience boiling transition and fail if it was operating at the somewhat lower MCPR limits discussed in Section'2.1.2.2.2. In the previous cycle the licensee addressed-this concern by proposing a method for the detection of aDncrmal fuel degradation associated with possible fuel misloadings at MNP. This was accomplished by measuring the off gas radioactivity level at the steam-jet air ejector. To preclude continued fuel degradation caused by cladding oxidation resulting from a fuel loading error, the licensee proposed a Technical Specification off gas limit for the steam jet air ejector (SJAE) radioactivity.

Monitoring offgas activity allows the operator to indirectly assess fuel integrity by alerting the operator of the presence of a possible

. fuel misloadinr; should the SJAE activity exceed the established off-gas limit. The offgas limit of 0.236 ci/sec (with a 30 minute decay) for 15 minutes restricts activity release to less than or equal to that which would be expected from a single misloaded bundle. Again for Cycle 8, in the event this Technical Specification offgas limit is exceeded, the licensee will be required to increase the MCPR operating limi ts. The MCPR operating limits will De raised under such circumstances to 1.42 for 8x8 fuel and 1.46 for the 8xbR and P8x8R fuel.

Increasing the operating MCPR to these higher values (multiplie'd by the appropriate Kf factor) assures that the worst misloaded bundle remains above the safety limit MCPR. This MCPR adjust-ment will effectively arrest continued fuel degradation caused by excessive cladding oxidation which might otherwise occur in the absence of a MCPR adjustment. Continued plant operation would then be determined by the most limiting condition required by these nigher MCPR values or the Technical Specification offgas limits. We find the above stated Technical Specifications which will assure no more then limited fuel degradation and off-gas release in the event of a fuel loading error, to be acceptable for Cycle 8 operation on MNP.

2.1.4 Overpressure Analysis For Cycle 8, the licensee reanalyzed the limiting pressurization event to demonstrate that the ASME Soiler and Pressure Vessel Code requirements will continue to be met for MNP. The methods used for this analysis, when modified to account for one failed safety valve, have been previously approved (13) by the staff. The acceptance criteria for this event is that the calculated peak transient pressure not exceed 110% of design pressure, i.e.,1375 psig. The reanalysis shows that tne peak pressure at the Dottom of the reactor vessel does not exceed 1274 psig for worst case end-of-cycle conditions, even when assuming the effects of one failed safety valve. This is acceptable to the staff.

2.1.5 Themal-Hydraulic Stability A thermal-hydraulic stability analysis was performed for this reload using the methods described in Reference 12. The results show that the fuel type dependent channel hydrodynamic stability decay ratios

. and reactor core stability decay ratio at the least stable operating state (corresponding to the intersection of the natural circulation power curve and the 100% rod line) are 0.0.6 (8x8R/P8x8R), 0.10 (8xB) and 0.55 respectively. These predicted decay ratios are all well below the 1.0 Ultimate Performance Limit decay ratio proposed by GE.

The staff has expressed generic concerns regarding reactor core thermal-hydraulic stability at the least stable reactor condition.

This condition could be reached curing an operational transient from high power if the plant were to sustain a trip of both racirculation pumps without a reactor trip. The concerns are motivated by in-creasing decay ratios as equilibirum fuel cycles are approached and as reload fuel designs change. The staff concerns relate to both the consequences of operating with a decay ratio of 1.0 and the capability of the analytical methods to accurately predict decay ratios. The General Electric Company is addressing these staff concerns through meetings, topical reports and a stability test program.

It is expected that the test results and data analysis, as presented in a final test report, will aid considerably in resolving the staff concerns.

For Cycle 7 operation, the licensee added a requirement to the MNP Technical Specifications which restricted pire.ned plant operation in the natural circulation mode. Continuation of this restriction will also provide a significant increase in the reactor core stability operating margins during Cycle 8.

On the basis of the foregoing, the staff considers the thermal-hydraulic stability of MNP during Cycle 8 to be acceptable.

J 2.2 Extended Exposure Fuel Assemblies For Cycle 8, as part of a joint NSP/GE/D0E extended exposure fuel program, the licensee has proposed to reinsert a limited number of previously irradiated single water rod 8x8 fuel assemblies for a sixth a cie of operation. The subject program, involving eight fuel assemblies, is intended to obtain extended fuel exposure information to assess the inpact of higher fuel burnups on

~

fuel reliability. During Cycle 8 it is projected that several of these assecolies will attain and significantly surpass peak pellet (local) exposures of 40,000 mwd /t, which represents the upper bound of the exposure basis for the approved reference (12) generic fuel thermal-mechanical design and safety analysis. Accoraingly, the

. licensee has provided the results of supplementary fuel performance licensing calculations. The calculations are for a peak pellet exposure predicted to be attained by any of the eight lead fuel assemblies during Cycle 8.

Additionally, the licensee has provided an analysis of the adequacy of ECC system performance for MAPLHGRs up to 45,000 mwd /T average planar exposure. The results of'these analyses are evaluated in the following sections. Additionally,

inherent to the program is that actual peak and average fuel exposure will exceed by a gradually increasing amount, the proven peak local and peak bunale average discharge exposure levels associated with current reference comercial BWR fuel. Accordi ngly,

the licensee has also submitted the results of surveillance tests, measurements and inspections performed at the end of each of the first four operating cycles and a description of the criteria for evaluating the condition of the eight fuel assemblies for continued operation for a sixth cycle. This information is intended to pro-vide direct, empirical assurance of the continuing excellent fuel performance of the subject fuel and is evaluated in the following sections.

2.2.1 Extended Exposure Fuel Analysis Fuel Cladding Integrity Safety Limit LHGR In order to avoid fuel rod rupture, due to excessive cladding strain caused by rapid thermal expansion of the fuel pellet.

GE has established a cladding plastic diametral strain limit of 1%. Using the previously accepted methods for calculating cladding strains, the linear heat generation rate (LHGR) corresponding to 1% cladding plastic diametral strain was determined for a peak pellet exposure of 50,000 mwd /T. The results show that a linear heat generation rate of 15.5 kw/f t would be required to cause 1% cladding plastic diametrial strain in the UO2 fuel rods at this exposure. To assure that the above LHGR will not be exceeded by the extended exposure fuel, the licensee reanalyzed the limiting control rod withdrawal error (RWE) event to determine the maximum transient linear heat generation rate.

The Cycle 8 analysis (6) shows that the peak LHGR for any 8x8 fuel will be no more than 12.4 kw/f t.

Accordingly, the aforementioned safety limit LHGR at the higher exposure will not be violated should this limiting event occur. Thus fuel failure due to excessive cladding strain will be precluded. We find these results, which adequately account for the effects of fuel densification power spiking, to be acceptable.

. Cladding Collapse Analysis Cladding collapse potential was reassessed by the licensee as part of the supplemental thermal-mechanical design analysis of the higher exposure 8x8 fuel. The collapse analysis was performed using the approved generic models and methods described in Reference 12.

The ollapse design basis assumes an instantaneous increase in re%co,r system pressure equivalent to that which would result from a turbine trip without bypass at hot full power. The results of the calculation showed that cladding creep collapse would not occur for a maxinum exposure of 50,000 mwd /T. We find these results which include the effects of fuel densification power spiking to be acceptable.

Fuel Rod Stress, Deflection and Fatioue Analyses Supplemental fuel rod stress, deflection and fatique analyses were performed for the extended exposure fuel bundles operating through Cycle 8.

The analyses were performed using the models, methoas and design limits given in Reference 12. The results of these stress, deflection and fatique analyses show that the 1.0 stress design ratio limit, rod-to-rod and rod-to-channel spacing limits and 1.0 fatique damage limit will not be violated by the subject fuel during Cycle 8.

ECCS Appendix X Analysis The general requirements for reevaluating ECCS performance and associated technical specifications are stated in Section 2.1.3.1, herein. Previously, the staff reviewed and accepted the Loss-of-Coolant Accident analysis report (21) and supplement (7) for MNP and approved the accompanying proposed exposure-dependent MAPLHGR limits for the 8DB262 fuel. Our approvals were limited to an average planar er.posure of 40,000 mwd /T.

In order to extend the 8DB262 MAPLHGR limits out to an average planar exposure value beyond that which is expected to be achieved during Cycle 8 the licensee reevaluated the adequacy of ECCS performance for the subject 8x8 fuel for an average planar exposure of 45,000 mwd /T.

. The supplemental analysis for the lead assemblies was performed using analytical procedures previously approved by the staff.

These methods are considered to be acceptable for this limited application. Furthermore, the information presented fulfills the documentation requirements outlined in Reference 22 for such analyses.

The analysis results show that with MAPLHGR limitec to 8.0 kw/f t at 45,000 mwd /T the peak cladding tempeature and local cladding oxication fraction will remain well below the 2200*F (peak cladding tenperature) and 17% (local cladaing oxidation) limits allowed by 10CFR50.46. Accordingly, we find the proposed MAPLHGR versus average planar exposure limit to be acceptable.

2.2.2 Extended Exposure Fuel Surveillance Program The eignt type 8DB262 8x8 fuel assemblies to be operationally extended were first inserted in the MNP-1 reactor at the beginning of Cycle 3.

Three of these assemblies include the highest' exposed fuel from the Cycle 3 reload batch, while the remaining five are of lower exposure from the same batch. One of the three highest exposure fuel Dundles was dimensionally precharacterized prior to its first cycle of irraciation. The surveillance assembly has been examined in detail following each operating cycle in order to provide a continuing and representative measure of fuel per-formance and Dehavior of the highest exposure assemblies to be operationally extended to higher fuel exposure. The examinations have involved fuel rod non-destructive testing (ultrasonic and eday current) dimensional measurements and visual inspections.

The surveillance examinations performed after each of the first four operating cycles corresponded to assemoly average burnups of approximately 3,500 mwd /T,10,,100 mwd /T, 20,500 mwd /T and 25,900 7

mwd /T. The results(8, 9,

11) of these examinations show that the fuel assembly irradiaticn induced dimensional changes (in directions along the assembly vertical axis. and transverse to the axis) were within acceptable ranges. In each examination selected hign power roos in the surveillance bundle, which were subject to nondestructive testing, were shown to be souad. That is, none of the rods examined exhibited NDT signals indicative of cladding

. discontinuities. Also NDT signals indicative of the presence of water within the rod resulting from breach in the cladding were not found.

Visual examination of these same rods at the end of each cycle showed light crud film deposits typical of fuel irradiated in a commerical BWR. Additionally, the fuel rods exhibited no unusual corrosion cnaracteristics and only minor abrasive wear noted at tne spacer contact points. In general at the end of each cycle all the rods observed were found to be in excellent condition for continued operation. Similar visual examination of the composite assembly showed it to be in excellent condition for continued operation.

At the end of the current cycle (Cycle 7) the three highest exposure bundles (including the precharacterized surveillance bundle) will again be examined in detail. The examinations will involve visual inspections, dimensional measurements and non-destructive testing. The examinations will De performed to assess the acceptability of the fuel bundles for extenced burnup.

Additionally, should there be evidence of failed fuel in the Cycle 7 core, because of higher than expected reactor off gas levels, the extended burnup bundles will be chemically examined via fuel sipping. The acceptance criteria (4) requires that

1) no mechanically failed fuel rods be present, 2) no fuel bundle abnormalities which could result in mechanical failure during the sixth operating cycle be present and 3) no fuel bundle irradiation induced dimensional changes which exceed limits associated with the criteria for the initial bundle design be observed.

Based on differential irradiation growth calculations it is not expected that bundle modifications will De required for any of the bundles in order to accommodate differential growtn during the sixth operating cycle. However, if necessary, the water rod upper end plug shank will be extended to assure adequate fuel asserely clearances.

We find that the results of the previous fuel examinations, to-gether with the planned end of Cycle 7 extenced exposure fuel examinations and related acceptance criteria will provide an adequate empirical basis for determining the acceptability of extending the subject fuel for a sixtn cycle of operation.

. The licensee has agreed (4) to provide the staff with a brief summary of the results and findings of the end of Cycle 7 fuel examinations 90 days after startup. A final report on the*results of the end of Cycle 7 examinations will be provided to the staff approximately six months after startup. These reporting plans are acceptable to the staff.

2.2.3 Extended Exposure Fuel Perfonnance During Cycle 8 We have considered the affect of the Cycle,8 reactor operating conditions on the performance of the extended exposure fuel assemolies.

Because of the substantial depletion (burnup) of the extended exposure assemblies it is not expected that they will be able to operate close to thermal operating limits during Cycle 8.

These operating margins, which are not taken credit for in the safety analyses e. valuated in Section 2.2.1 herein, would be expected to pro-vide some significant although unquantifiable additional margins to fuel damage during normal operation, anticipated transients ano postulated accidents.

The planned cycle length and fuel management scheme will result in a sixth cycle maxinum bundle average exposure increment and maximum local pellet exposure increment which are approximately only 1/6 of the corresponding integral values accumulated over the six cycles of operation. Additionally, the Cycle 8 peak local pellet exposure increment for the high burnup assemblies will only be approximately 17% beyond the 40,000 mwd /t standard reference exposure basis. In view of these considerations we believe that Cycle 8 operation results in a relatively modest increment and extansion to the operating experience exposure basis for which acceptable 8x8 fuel performance is well established.

Additionally, tne licensee states that the fuel pre-conditioning operating management techniques (PCIM0Rs) reconinended by GE on an interim basis will continue to be applied to the extended exposure fuel assemolies for local power maneuvers during Cycle 8.

These PCIOMRs (which limit local power increases to slow ramp rates at higher ' local powers) have been shown to be ~ effective in limiting pellet cladding interaction related fuel failures and will help further assure adequate fuel performance during Cycle 8.

In view of the above, we believe that continued acceptaDie performance (reliability) of the subject extended exposure fuel assemblies is to be expected.

. Finally, there will be only eight extended exposure fuel assemblies (only three will be of the highest exposure) in the 484 fuel assemoly Monticello Cycle 8 reload core. Accordingly, we believe that on a core-wide basis, fuel performance will be acceptaole even if some of the extended exposure fuel should experience limited fuel failures.

In view of the above, the staff finds it acceptable to extend the exposure of the eight 8DB262 fuel assemblies during Cycle 8.

Our approval is limited to Cycle 8 of MNP up to a peak pellet exposure of 50,000 mwd /T. We require that additional information be submitted for our review in the event that any of the subject fuel assemblies are to be reinserted into the core for a seventh operating cycle.

3.0 Physics Startup Testing Several of the key safety analysis inputs and results can be verified via preoperational testing.

In order to provide this assurance the licensee will perform a series of physics startup tests, which are described in Reference 4.

Based on our review, this program is a.cceptable for Cycle 8.

A written report describing the results of the ' physics startup tests, will also be provided by the licensee with-in 90 days of startup, which is acceptable.

4.0 Technical Specification The proposed Technical Specification changes for Cycle 8 operation of MNP include revised operating limit minimum critical power ratios (OLMCPRs) for each fuel type in the core.

As discussed in Sections 2.1.2.2.2 and 2.1.3.3 herein, the Cycle 8 OLMCPRs will be 1.41 for the 8x8 and 8x8R fuel types and 1.44 for the P8x8R fuel type whenever the gross radioactivity release rate of noble gases at the steam jet air ejector does not exceed (for a period greater than 15 minutes) the equivalent of 236,000 uCi/sec following a 30 minute decay.

If and when this limit is exceeded the OLMCPRs will be increased to 1.42-for the 8x8 fuel t,ypes and 1.46 for the 8x8R and P8x8R fuel types.

These MCPR operating limits are consistent with and adequately supported by the Cycle 8 reload safety E..alyses and are acceptable.

Additionally, the licensee has proposed MAPLHGR vs. Average Planar Exposure limits for the type P80RB282 and type P8DRB265L pre-pressurized 8x8R raload assemblies which are identical to the same

. type unpressurized 8x8R fuel assentlies. As discussed in Section 2.1.3.1, this is acceptable. Finally, the licensee has proposed to add a MAPLHGR limit of 8.0 KW/f t for the 8DB262 fuel at 45,000 As discussed in Section 2.2.1 herein, tnis limit is acceptable.

mwd /T.

5.0 Radiological Considerations TrradTating fuel to extended burnups will increase the inventory of long-lived fission products in the core.

The short-lived fission products will have The reached equilibrium levels at lower burnups and will not be affected.

potential consequences of the design basis accidents are determined by the The only significant long-short-lived fission product activity inventory.

lived radionuclide with respect to potential consequences of the postulated design basis accidents is the noble gas Kr-85.

For an additional two year burnup the Kr-85 fuel inventory will increase about 25%.

However, this increased Kr-85 inventory will have an insignificant effect on the design basis LOCA whole body dose since Kr-85 contributes a very small fraction cf the total dose (<0.1%).

For a fuel handling accident the whole body dose increare due to the Kr285 fuel inventory increase will also be less than 0.1 %.

We expect that operating Monticello with extended burnup fuel assemblies may increase the fraction of failed fuel in the core over that previously experi-Therefore, there may possibly be an increase in the primary coolant enced.

fission product activity levels and the amount of activity released from the plant as compared to releases during operation of the plant in previous cycles.

However, we do not expect these increases to be significant because (1) only eight as'semblies in the core (1.65%) will be irradiated to the extended burn-ups, (2) the Technical Specifications Section 3.6.C.1 and 3.11.C.2 impose reactor coolant and steam air ejector radioactivity limits in the event of fuel failures to limit fission product release, (3) the licensee is planning to examine the fuel assemblies stated for extended burnup for mechanical failed fuel rods, fuel bundle abnormalities and dimensional changes prior to acceptance for reinsertion, and (4) the augmented offgas system installed at Nnticello is currently operating at about a 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> holdup time which will provide for the decay of noble fission gases which may be present in the offgas due to fuel rod failures so that the release rate of these gases will be within the permissible stack release rate.

6.0 Conclusion We have concluced, based on the considerations discussed above that:

(1) because the amendment ooes not involve a significant increase in the probaDility or consequences of accidents previously con-sidered and aces not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consiceration, (2) there is rea;3nable assurance that the health ano safety of the public will not ae encangereo by operation in the proposea manner, and (3) such activitics will be concucted in compliance with the Commission's regulations and the issuance of tnis amenament will not be inimical to tne common defense and security or to the healtn and safety of the puDlic.

Dated:

Februa ry 12, 1980

. 6.0 References 1.

Northern States Power Company letter (L. Mayer) to USNRC, dated September 17, 1979.

2.

Northern States Power Company letter (L. Mayer) to USNRC, dated November 1,1979.

3.

Northern States Power Company letter (L. Mayer) to USNRC, dated November 30, 1979.

4.

Northern States Power Company letter (L. Mayer) to USNRC, dated December 31, 1979.

5.

Northern States Power Company letter (L. Mayer) to USNRC, (Exhibit B, dated October 24, 1979) dated Novencer 1,1979.

6.

" Supplemental Reload Licensing Submittal for Monticello Nuclear Generating Plant, Reload 7," NED0-24221, October 1979.

7.

"Monticello Nuclear Generating Plant Extended Exposure Fuel Program," NEDD-24202, July 1979.

8.

"8x8 Fuel Surveillance Program at Montir.ello, End-of-Cycle 3 First Post-Irradiation Measurements January 1975," NEDM-20867, April 1975.

9.

"8x8 Fuel Surveillance Program at Monticello, End-of-Cycle 4 Second Post-Irradiation Inspection September 1975," NEDM-21080, October 1975.

10.

"8x8 Fuel Surveillance Program Monticello, End-of-Cycle 5 Third Post Irradiation Inspection September 1977," NEDE-25195, October 1979.

11.

"8x8 Fuel Surveillance Program Monticello, End-of-Cycle 6 Fourth Post-Irradiation Inspection, October 1978," NEDE-25245, DecemDer 1979.

12.

" Generic Reload Fuel Application," NEDE-240ll-P-A, August 1978.

13. USNRC letter (D. Eisenhut) to General Electric (R. Gridley),

dated May 12, 1978.

. 14. USNRC lettter (T. Ippolito) to General Electric (R. Gridley),

dated April 16, 1979.

15. " Status Report on the Licensing Topical Report ' Generic Reload Application for 8x8 Fuel' NED0-20360 Revision 1 anc Supplement 1" by Division of Technical Review, ONRR, USNRC, April 1975.

16. " Generic Reload Application for 8x8 Fuel," Revision 1, Supplement 4, April 1976, NEDD-20360.

17. " General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," General Electric Company, BWR Systems Department, November 1973 (NEDD-10958).

18. General Electric letter (R. Gridley) to USNRC (D. Eisenhut and D. Ross), dated October 5,1978, transmitting " General Electric' Information Report NEDE-24131, Basis for 8x8 Retrofit Fuel Thernal Analysis Application.

19. General Electric letter (R. Engle) to USNRC (D. Eisenhut and R. Tedesco), oated March 30, 1979.

20. USNRC letter (W. Butler) to General Electric (I. Stuart),

dated October 2,1974.

21. " Loss of Coolant Accident Analysis Report for Monticello Nuclear Generating Plant," NEDD-24050-dated September 1977.

22. USNRC letter (D. Eisenhut) to General Electric (E. Fuller) dated June 30,1977.

.