ML17199F883

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Safety Evaluation Supporting Amend 87 to License DPR-25
ML17199F883
Person / Time
Site: Dresden Constellation icon.png
Issue date: 07/24/1986
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Office of Nuclear Reactor Regulation
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NUDOCS 8608180084
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Text

UNITED STATES e

NUCf"EAR REGULATORY COMMISSION

  • WASHINGTON,.D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR *REGULATION SUPPORTING AMENDMENT NO. 87TO FACILITY OPERATING LICENSE NO. DPR-25 COMMONWEALTH EDISON COMPANY DRESnEN NUCLEAR POWER STATION, UNIT NO. 3 DOCKET NO. 50-249

1.0 INTRODUCTION

.*. By letter dated Februar,y 21, 1986 (Ref. 1) a.s supplemented by a letter dated April 18, 1986 (Ref. 13), Commonwealth Edison Company (CECo) (the licensee) proposed to amend Appendix A of Facility Operating License No. DPR-25.

The requested amendment furnished information to support authorization for (1) Dresden 3 operation with 9X9 reload fuel supplied by Exxon Nuclear Company (ENC), (2) incorporation of an expanded power/flow operating map and (3) incorporation of single loop operation (SLO) provisions in the body of the Technical Specifications (TS).

The Dresden 3 Cycle 10 (D3Cl0) reload will consist of 176 fuel bundles fabricated by ENC.

These 9X9 bundles are comprised of 79 active fuel rods and two inert water rods. During Cycle 10 operation, the 9X9 fuel will reside with 140 General Electric (GE) and 408 ENC 8X8 fuel assemblies presently in the core.

In support of the D3Cl0 reload, CECo submitted topical reports which described the design and safety analysis (Ref. 2),

the plant transient analysis (Ref. 3), and the LOCA-ECCS analysis (Ref.

4) for the ENC 9X9 fuel. Additional information in response to NRC inquiries was provided by the licensee in References 20 and 23.

To support the SLO provisions in the TS, CECo submitted a core stability assessment of ENC 9X9 fuel at Dresden 3 (Attachment 6 to Ref. 1).

An evaluation of the relative stability margins of 8X8 and 9X9 fuel types was a part of this submittal. Additional information related to SLO was submitted in Reference 13.

2.0 EVALUATION OF FUEL DESIGN 2.1 Fuel Mechanical Desiqn The D3C10 core reload will include 176 ENC new 9X9 fuel assemblies with the designations XN-3 and XN-3A.

These reload assemblies contain 79 fuel rods and two water rods. All 176 assemblies will hava.the same enrichment (3.13 percent). The XN-3 and XN-3A assemblies are differen-tiated by the.number and location of the Gadolinia-bearing fuel rods.

The fuel design and safety analysis for the 9X9 fuel is described in the Dresden 3 specific report XN-NF-85-57 (Ref. 2) and the qeneric

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e-design report XN-NF-85-67 Revision 1 (Ref. 5).

(The lat~er report is cur-rently under NRC review and the Safety Evaluation (SE) is in final processing). The ENC and GE 8X8 fuel types to be returned to the Dresden 3 core were approved for operation in previous cycles.

Th~

prior ENC 8x8 fuel types carry the designations XN-1 and XN-2.

Table 2.1 of XN-NF-85-67 Revision 1 gives the pertinent data for the XN-3 9X9 fuel. The burnable poison rods contain 4.00 weight percent Gd 203 blended with 2.25 weight percent U-?.35 to reduce the initial reactivity. The ENC XN-3 and XN-3A fuels are designed to fit into the existing GE channel boxes.

A more detailed description can be found in Table 2.1 of XN-NF-85-67.

Based on our re~iew of the information in Table 2.1, we find the mechanical design of the Exxon 9X9 fuel for D3C10 reload is acceptable. However, approval of extended exposure limits for future operating cycles is contingent on our*

approval of XN-NF-82-06(P) and Supplements 1, 2, 4, and 5 (Ref. 21).

2.2 Rod Pressure 2.3 Prior 8X8 fuel designs called for no rod internal pressure exceeding the reactor system pressure during normal operation. The analyses of strain, corrosion, rod pressure, overheating of fuel pellets and pellet clad inter-action (PCI) using an approved version of the ENC code RODEX 2 showed that the limits on the physical parameters would not be exceeded for ENC 8X8 fuel throughout the entire lifetime of the Dresden 3 core.*

For the D3C10 ENC 9X9 reload fuel, calculation of the fuel rod internal pressure was done in accordance with acceptance criteria cited by ENC in XN-NF-85-67, Revision 1 (Ref. 5). The_ evaluation was performed with RODEX 2A which is a revision of the RODEX2 code used in the analysis of previous Dresden 3 reload fuel cycles. Our review of the ROOEX 2A topical report is complete and the staff SE is in final processing. The staff has concluded that the acceptance criteria for rod internal pressure can be.

fully met throughout the entire expected irradiation life of the 9X9 fuel.

Fuel Rod Row Our review of XN-NF-85-67, Revision 1 (Ref. 5) has progressed to the point that we may conclude that Exxon has demonstrated conformance to approved rod bow design 1 imits for minim'um ~ap spacing to *fuel assembly average exposure of at least 30,000 MWD/MTU for the 8x8 fuel and to 23,000 MWD/MTIJ for the 9x9 fuel. Projected peak assembly burnups for D3Cl0 reload is 30,700 MWD/MTU for the 8x8 fuel and 11,500 MWD/MTU for the 9x9 fuel. -We find the D3Cl0 core acceptable with respect to rod bow considerations. However, additional justification with regard to fuel rod bowing must be provided for Dresden 3 operation beyond Cycle 10 or for any significant extension of Cycle 10 beyond the projected 8x8 peak assembly burnup value of 30,700 MWD/MTU *

- 3 ~

2.4 Fuel Centerline Melting The design basis for the ENC fuel centerline temperature is that no fuel centerline melting should result from nonnal operation including transient occurrences. The results of an evaluation reported in the D3Cl0 reload analysis report XN-NF-85-57 (Ref. 2) were based on RODEX?.A.

ROOEX2A has been reviewed in connection with the staff review of Reference 5 and the staff has concluded that the generic methodology for the ENC 9X9 fuel is acceptable for the D3C10 reload fuel. The licensee has identified Reference 5 as the applicable document for the ENC 9x9 fuel (Ref. 20).

2.5 Cladding Swelling and Rupture The cladding swelling and rupture models in XN-NF-82-07.(EXXON Nuclear Company ECCS Cladding Swelling and Rupture Model} have been approved for use in the ENC ECCS Evaluation Model and have been incorporated in the approved ENC EXEM/BWR ECCS model. This model was used in the ECCS analysis for the previous Dresden 3 Cycle 9 and remains in effect for the ENC XN-1 and XN-2 fuel types. The staff has verified that ENC is using the approved model for the 9X9 fuel ECCS analysis and we find the application to be acceptable.

2.6 Linear Heat Generation Rate Limits for ENC 8X8 and 9X~ Fuel CECo has provided a figure of Linear Heat Generation Rate Limit vs Nodal Exposure for both GE 8X8 and ENC 8X8 and 9X9 fuel types to be incorporated in the D3C10 TS (Reference 13). This figure was updated in Reference 20 to reflect the design values which have been previously rev.iewed for the GE and ENC 8X8 fuel types and recently approved.for the ENC 9X9 fuel in connection with our review of XN-NF-85-67, Revision 1 (Ref. S'.

3.0 THERMAL HYDRAULIC DESIGN The review of the thennal-hydraulic aspects of the Cycle 10 reload consisted of the following:

(a} the compatibility of the ENC 9X9 and prior ENC and GE 8X8 fuel bundles; {b} the fuel cladding integrity safety limit;.(c) the operating safety limit minimum critical power ratio; (d) the amount of bypass flow associated with the different fuel designs; (e) thennal hydraulic stability, and (f} the proposed TS.*

The objective of the review was to confinn that the thermal-hydraulic design of the reload core was accomplished using acceptable analytical methods, that there was an acceptable margin of safety from conditions which would lead to fuel damage during nonnal operation and anticipated operational occurrences and that the core is not susceptible to thermal-hydraul ic instability.

3.1 Hydraulic Compatibility Since a boilinq water reactor (BWR) core is a series of parallel flow channels connected to a common lower and upper plenum, the total pressure drop across the bundles will be equal. However, differences in the hydraulic resistances of the fuel designs may cause variations in axial pressure drop profiles across the bundles.

Component hydraulic resistances for the proposed constituent fuel types in the D3C10 core have been.

detennined in single phase flow tests of full scale assemblies.

The compatibility of the GE retrofit and ENC 8X8 fuel types has been demon-strated for previous reloads. Additional analyses of the effects of hydraulic compatibility on thennal margin were presented in the D3C10 reload report (Ref. 2). The results of these analyses showed that the XN-3 9X9 fuel perfonnance falls between that of the ENC 8X8 fuel and the

  • GE 8X8 fuel. Based on our review of the infonnation provided in the Cycle 10 reload report we conclude that the GE and ENC fuel types are hydraulically compatible.

3.2 Minimum and Operatinq Limit Critical Power Ratio The minimum critical power ratio (MCPR) safe~v limit for the Cycle 10 reload was determined by the licensee to be 1.05 for all ENC fuel types. A safety limit of 1.06 for GE fuel types was approved for previous Dresden 3 operating cycles. The methodology for Cycle 10 is based on ENC's revised critical power methodology in XN-NF-524, Revision 1 (Ref. 6) which incorporates a constant flow MCPR formulation for BWR applications. The staff has completed its generic review of XN-NF-524 (Ref. 7) and has concluded that the methodology for arriving at an MCPR safety limit is acceptable. The XN-3 correlation used to develop the MCPR safety limit has been approved for both the resident ENC 8X8 fuel types {Ref. 16) and the new 9X9 fuel type (Ref. 17). The methodology of XN-NF-524, Revision 1 was applied generically for the upcoming Cycle 10 and the previous Cycle 9 for Dresden.3 and is considered applicable to the resident GE 8x8 fuel types as well as the ENC fuel. The staff has verified through its review of the D3C10 transient analysis rep'ort XN-NF-85-62 (Ref. 3) and followup response to additional infonnation needs (Ref. 20) that the methodology for determininq uncer-tainties and the application in determining the MCPR safety limit is acceptable.

Various operational transients could reduce the MCPR below the intended safety limit. The most limiting transients have been analyzed to determine which event could potentially induce the largest reduction (6CPR) in the initial. The transient which resulted in the largest 6CPR was the load rejection without bypass.

The staff has. completed its review of XN-NF-79-71 (COTRANSA2 code) anrl its supplements which discuss the methodology to detennine the operatin~

limit CPR and the SE is in final processing.

To account for the uncer-tainties in the values used in the TS for previous Dresden reload cycles, the NRC staff imposed additional conservatism by

  • requiring that the methods discussed in the SE on the.GE ODYN code be used. The resulting uncertainty value of 0.0?.2 6CPR/lnitial-CPR was applied to all fuel in Cycles 8 and 9.

The licensee has stated in Attachment ?. to Reference 1 that the ENC methodology of XN-NF-79-71 includes a deterministic multiplier of 110 percent on the calculated transient power to account for COTRANSA code uncertainties. This revised code uncertainty is treated as a bounding, deterministic factor instead of a statistical parameter. The staff has concluded in a prior review (Ref. 15 Staff SE on XN-NF-79-71) that this deterministic code uncertainty factor of 110 percent in transient integral power ratio is sufficiently bounding for license applications and that the MCPR operating limit so derived will assure that the safety limit MCPR is not violated in the event of any operating transients. The staff also agrees that this deterministic approach is* sufficient to replace the previously imposed MCPR penalty of 0.03.

The 6CPR for the load re.iection without bypass transient was calculated using a modified version of the COTRANSA 6CPR calculation model described in XN-NF-79-7l(P) Rev. 2 (Ref. 8), which has been reviewed by the staff.

Since the description of these modifications provided in Appendix A to the Dresden 3 Cycle 10 transient analysis report (Reference 3) was inadequate for staff review, the licensee was requested to clarify the deviations from the analysis methods described in Reference 8 and to justify the proposed MCPR operating limits based on the Reference 8 methodology.

The licensee responded by a submittal dated June 20, 1986 (Reference 23), which describes the modified methodology consisting of revisions to the hot channel model as follows:

(1) the HUXY and XCOBRA models were combined into one calculation, (2) a response surface was used to represent the XCOBRA core average/hot channel flow split calculation, (3) transient hydraulics based on the hydraulic solution of the COTRANSA core model were incorporated.

Sensitivity analyses were performed to establish the difference in 6CPR results between the hot channel model described in Reference 8 and the modified model.

The results showed that the calculation method of Refer-ence 8 yields a higher 6CPR of 0.030 for 8x8 fuel and 0.055 for 9x9 fuel.

Based on the Dresden 3 Cycle 10 analysis (Ref. 3), the licensee originally proposed 6CPRs of 0.28 for the new ENC 9x9 fuel, 0.?4 for the ENC 8x8 fuel, and 0.?.3 for the residual GE 8x8 fuel. This would r.esult in MCPR operating limits of 1.33 for ENC 9x9 fuel, 1.29 for ENC 8x8 fuel, and 1.28 for GE 8x8 fuel.

To account for the deviations from approved methodology, the licensee has proposed (Reference 23) to increase temporarily the MCPR operating limits by 0.04 for 8x8 fuel and 0.06 for 9x9 fuel while the staff completes a review of the updated methodology or revised analyses based on XCOBRA-T (Reference 24) are submitted. Reference 24 is currently under

3.3 4.0

  • review by the staff. Based on results of the sens*itivity-study on the effect of the hot channel model modifications, we find the proposed approach to be acceptable. Therefore, the TS for 03C10 will be issued with MCPR operating limit values of 1.33 for all 8x8 fuel and 1.39 for 9x9 fuel.

Thermal-Hydraulic Stability The thermal-hydraulic stability of the Cycle 10 core was analyzed using.

the methods identified in XN-NF-80-19, Volume 4, Revision 1 (Ref. 9}.

Reference 9 cites the use of the COTRAN and COTRANSA 2 models for use in the analys~s of core thennal-hydraulic stability. *The NRC SE on Reference

. 9 is provided in Reference 10 and states that the use of these models is acceptable in accordance with the restrictions cited in the applicable SE, including the submittal of supplemental qualification data as needed to evaluate the applicability of these models to unapproved fuel types.

For*

Dresden 3 Cycle 10--operation, the licensee has provided additional stability analyses for the ENC 9X9 fuel using ENC's advanced system stabilit.v model COTRANSA 2.

The results of these analyses and comparison with results Trom the approved COTRAN code are provided in Attachment 6 to Reference 1.

NRC Generic letter 86-02 (Ref. 10) concluded that the licensees of BWR 1, 2 and 3 Class plants should examine each core reload to assure that it is typical of previously evaluated cores which have an acceptable stability

  • margin.

An acceptable margin for ENC analysis of stability is a decay ratio of. 0.75, which is a result of the estimated uncertainty of 25 percent in the calculation of the thennal-hydraulic stability decay ratio with the COTRAN code.

Pennanent approval of the COTRANSA 2 analytical methodology and results for Dresden 3 stability analyses is subject to benchmark tests in the high decay ratio area (greater than 0.75}. This benchmarking is planned in the near future and will be perfonned using data from a Vermont Yankee 8X8 core. With regard to Dresden 3, we are not at this time prepared to conclude that a full core loading of 9X9 fuel is acceptable from a stability standpoint.

We can, however, find the mixed core 9X9 reload acceptable for Cycle 10 only, subject to the incorporation of surveillance requirements for Average Power Range Monitor {APRM) and Local Power Range.Monitor (LPRM) noise levels in a region of the power-flow map identified by the provisions of General Electric Service Information Letter 380 and required by NRC Generic Letter 86-02 for cores with non-conventional fuel designs.

Acceptable TS requirements have been proposed by the licensee in Reference 13; the staff evaluation of these requirements is discussed in Section 7.0 of this SE (Single Loop Operation). The staff will reevaluate the thermal-hydraulic stability for Dresden 3 at the next reload cycle.

Prior to operation of Dresden 3 with a full core of 9X9 fuel, the staff will consider the need to require on-line decay ratio measurements during a future startup program.

TRANSIENT AND ACCIDENT ANALYSIS The control rod withdrawal error, the fuel loading error and the rod drop accident were evaluated for Cycle 10. The licensee used methods described

5.0 in XN-NF-80-19~ Volume 4 (Ref. 14;. with staff SE included). The use of the Sf ngl e Sequence Control strategy ( fn wh.ich rods inserted-during power operation have low worth) assures that the control rod withdrawal error will not be limiting. Using a Rod Block Monitor setting of 110 percent of full power results in at.CPR of 0.148 for 8X8 fuel and 0.152 for 9X9 fuel.

Th~ change in CPR due to a fuel loading error is 0.19. These values are comparable to previous reloads and are not limiting~

The control rod drop accident evaluation yields a value of 109 cal/gm for the maximum deposited fuel enthalpy. This is well below the staff's criterion of 280 cal/gm, and is therefore acceptable.

The MAPLHGR curves for the resid~nt GE fuel types in the proposed TS have been extended to exposure limits in the range 40,000 to 45,000 MWD/STU.

The NRC staff specifically reviewed on a generic basis a GE Topical Report on extended burnup methodology, NEDE-22148-P (Ref.

18). Our SE on the GE report identified a concern with regard to the radiological consequence evaluation of the Fuel Handling Accident involving GE fuel assemblies with batch average exposure values greater than 38,000 MWD/MTU (34,500 MWD/STU).

In a telephone conversation, the licensee conveyed to the staff that the batch average exposure values on GE fuel during D3C10 will not reach 34,500 MWD/STU.

Although we under-stand that all of the GE fuel will be removed from the Dresden 3 core prior to reachin9 the extended burnup levels, the licensee should inform the staff of any change in planned refueling schedules so that appropriate*

consideration may be given to the Fuel Handling Accident.

In addition, al though we approve the proposed exposure 1 imi ts for the. ENC fuel types, the licensee should address the conseQuences of the Fuel Handling Accident involvinQ the ENC fuel types before the exposure value of 38,000 MWD/MTU is exceeded.

LOSS OF COOLANT ANALYSIS (MAPLHGR LIMIT)

The MAPLHGR limits for the ENC 8X8 fuel as given in the Dresden 3 plant.

TS remain applicable for Cycle 10.

The licensee has proposed additional MAPLHGR limits for the ENC 9X9 fuel based on the analysis results provided in XN-NF-85-63 (Ref. 4). The limiting LOCA break calculations were perfonned for the Dresden 3 reactor with a full core of ENC 9X9 fuel. The approved EXEM/BWR ECCS Evaluation Model codes were used for the LOCA calculations with array dimensions increased to accommodate the 9X9 array. Comparison of the analysis results did not indicate any major difference in the overall system perfonnance. A significant difference was observed in the reflood calculation results for the full 9X9 core and the previous Cycle 9 8X8 core results. The ~ifference appeared in the time of hot node reflood which was 169 sec for the Cycle 9 full ENC 8x8 core analysis and 160 sec for the Cycle 10 full 9X9 core analysis. *In calculating the peak cladding temperature (PCT) -for D3Cl0, the longer time (169 sec) was used in the analysis for 9X9 assembly heatup.

The resulting PCT was 2045°F at a burnup of 5000 MWD/MTU allowing a* 155°F margin to the 10 CFR 50.46 limit {compared to 2159°F for the 8x8 analysis}. Metal water reaction also peaks at 2.44

  • percent at a burnup of 5000 MWD/MTU remaining well below the*17 percent limit required by 10 CFR 50.46. The MAPLHGR limits from this analysis are proposed for the Dresden 3 TS for the ENC 9X9 fuel design. Since analysis of t~e LOCA was performed with reviewed and accepted codes, and the results are well within the limits of 10 CFR 50.46, the staff finds the proposed MAPLHGR limits for D3C10 acceptable.

6.0 EXTENDED LOAD LINE LIMIT ANALYSIS The Extended Load Line Limit Analysis (ELLLA) provides a basis for nonnal reactor operation in the region of the power/flow map above the

  • 100 percent power/100 percent flow load line and bounded by the 108 percent APRM rod block line and the.rated (100 percent) power line. The slope of the APRM rod block line is changed to provide a block at no greater than 0.58 W + 50 percent power (where W is recirculation drive flow in percent). This pennits operation at 100 percent power down to 87 percent flow. Operation within this extended region pennits improved power ascension capability.

Relevant considerations in the review of ELLLA for Dresden 3 include:

1. The change in slope of the rod block line has been previously reviewed and approved (see for example the SE for Amendment No. 59 to Operating License DPR-19 for Dresden 2) and is generally applicable.
2. Transient analyses done within the* ELLLA region are still limiting at 100 percent power/100 percent flow (See Table 5 *.1 of Ref. 3).
3. The flow effect on LOCA analysis has been generically reviewed and approved for operation (Ref. 12).

It is concluded that changes in core behavior caused by the extended operating range have been acceptably accounted for.

7.0. SINGLE LOOP OPERATION The CECo has proposed TS to pennit reactor operation with one recircula-tion loop out of *service. The restrictions. in the Limiting Conditions for Operation (LCO) include both dual loop and SLO TS for APRM flux scram trip and rod block settings, an increase in the safety limit MCPR value, and a revision to the allowable MAPLHGR values. Additional restrictions on flow control and valve position are also incorporated in the proposed TS.

As a result of discussions with the staff, CECo has provided an expanded discu$sion of reactor stability monitoring and restrictions.on the allowable operating conditions during SLO (Ref. 13). This review includes consideration of the items in Reference 13 as well as Reference 1 *

  • Sunmary of Single Loop Operation
1.

Long Tenn Operation in Single Loop Mode SLO with appropriate TS changes has been previously approved for a number of plants. It is concluded that appropriate provisions have been made so that transient and accident bounds will not be exceeded durinq the SLO mode for D3C10.

~

2.

MCPR Safety Limit will be Increased by 0.03

3.
4.

The MCPR Safety Limit for two loop operation discussed in Section 3.2 will be increased by 0.03 for SLO. This increase in the safety limit has been applied to Dresden 3 SLO in a previous review (Amendment No. 54 to License No. DPR-25 dated July 9, 1981). This number is to account for core flow and Transversing In-Core Probe reading uncertainties which are used in the statistical analysis of the safety limit.

The MAPLHGR Limits will be Reduced by Appropriate Multipliers The licensee proposed reducing the TS MAPLHGR by 0.7 for SLO.

This.

reduction factor accounts for the more rapid loss of core flow during SLO and is the same as that previously approved for applica-tion to temporary SLO for Dresden 3. (Amendment No. 54 to License No. DPR-25 dated July 9, 1981) and is therefore acceptable.

The APRM Scram and Rod Block Setpoints will be Reduced The licensee proposed to modify the two loop APRM Scram, Rod Block and Rod Block Monitor setpoints to account for reverse or stalled flow in the idle loop.

The basis for the changes is a core flow uncertainty analysis provided by the licensee (NED0-24807) and implemented in Amendment No. 54 to License No. DPR-25 dated July 9, 1981.

5.

The Recirculation Control will be Manual The licensee proposed that Dresden 3 be operated with the* recircu-lation system in the manual mode to reduce the effect of potential flow instabilities.

6.

The Suction Valve in the Idle Loop shall be Closed The closure and electrical isolation of the suction valve in the idle loop prevents the loss of low pressure coolant injection through the idle recirculation pump into the downcomer.

Based on the discussion presented above; we conclude tha*t the proposed TS for SLO of Dresden 3 are acceptable.

In addition to the TS changes described above, the TS provide certain surveillance requirements for monitoring thermal-hydraulic stability.

Specifically, baseline LPRM and APRM data will be accumulated in SLO and two loop operation at selected power/flow points on the Dresden 3 power-flow map {Figure 3.6.2 in the proposed TS).

Surveillance requirements dealing with thermal-hydraulic stability are also included in the proposed SLO for Dresden 3.

We have confirmed that the requirements are in accordance with the guidance provided by GE in Service Infonnation Letter No. 380, Revision 1. The TS generally call for restrictions or surveillance above the 80 percent rod line (in the power flow map) and below 45 percent flow; surveillance above 39 percent flow and no operation below 39 percent. Surveillance is by observation of the noise levels of the APRM and LPRM detectors. Noise levels greater than 3 times base levels require noise suppression activity, e.g., a lower power level. This is consistent with previously approved SLO surveillance requirements on other plants and is acceptable.

Certain monitoring requirements relative to jet pump integrity are related to the SLO mode.

By Reference 13, the licensee has stated its intent to monitor core plate differential pressure during SLO per station procedures.

The purpose of this monitoring is to detect excessive jet pump vibration and is responsive to a staff concern; we find this commitment acceptable.

The Dresden 3 TS presently contain a surveillance requirement for the jet pump diffuser to lower plenum differential pressure which is appli-cable to the SLO mode. This requirement is a result of the implementation of IE Bulletin 80-07 "BWR Jet Pump Assembly Failure" and is. acceptable for the proposed SLO mode for Dresden 3.

8.0 TECHNICAL SPECIFICATION CHANGES The TS changes for D3Cl0 involve three general areas and are summarized below:

{l) Incorporation of LHGR limits for ENC 8X8 and 9X9 fuel. as an LCO.

The additional information on LHGR limits discussed in Section 2.5 of this SE results in the addition of Figure 3.5-lA (page 3/4.5-2?.) in the Dresden 3. TS and the identification of the LCO in TS Section 3.5.J (pages 3/4.5-15 and 3/4.5-16).

(2) Addition or ENC 9X9 fuel type.

MAPLHGR values for the new ENC 9X9 fuel type were added and burnup limits were extended for some of the earlier GE fuel types.

MCPR.

  • safety 1 imits were added for the new fuel and revised fo*r the previous 8X8 fuel types.* Additional MCPR operating limits were specified for Manual Flow Control and Automatic Flow Control for all fuel types. The previous TS Figures 3.5-1 and 3.5-2 were updated to reflect the new information.

Some modification to the control rod scram insertion time surveillance was made to reflect the MCPR Operating Limit for the 9X9 fuel type.

The new LCO is identified on proposed TS page 3/4.5-24.

(3)

Incorporation of an Expanded Power/Flow Operating Map The ELLLA discussed in Section 6.0 of this SE results in an expanded power flow map for Dresden 3. This change is made in the proposed TS Figure 2.1-3 (page B 1/2.1-17).

(4)

Incorporation of SLO Provisions and Surveillance.Requirements for Monitoring Thermal-Hydraulic Stability Figure 3.6.2 in Reference 13 has been proposed to identify the Power/Flow limits for thermal-hydraulic stability surveillance in SLO.

This Figure wi 11 be added to the Section 3. 6. (page 3/ 4. 6-24.)

9.0 TECHNICAL CONCLUSIONS We have reviewed the infonTiation furnished by CECo in References 1, 13, 20, and 23 and Supplementary ENC reports (Ref. 2, 3 and 4) relative to the proposed License Amendment to allow.operation of Cycle 10 of Dresden Unit 3.

We find that sufficient basis has been provided to allow:.

(1) the addition of 176 ENC 9X9 fuel bundles in the Dresden 3 core, (2) operation with an expanded power/flow map based on an ELLLA and (3) operation in the SLO mode under the proposed TS stability monitoring provisions and restrictions on allowable operating conditions. The proposed TS chan~es are therefore approved for 03C10.

Our review as discussed in the Evaluation Sections above has identified certain restrictions relating to our incomplete review of the ENC 9x9 fuel and thennal-hydraulic stability considerations which limit approval to the upcoming Cycle 10 only. Specifically:

(1)

SE Section 2.0: Aoproval of extended exposure limits for the 9x9 fuel beyond 30,000 ~WD/MTU batch average exposure for future operating cycles is contingent upon our approval of XN~NF-82-06 (P) and Supple-ments 1, 2, 4 and 5 (Reference 21). In addition, justification with regard to rod bow is required for both the ENC 8x8 and.9x9 fuel for operation beyond t~e projected D3C10 exposure levels.

(2)

SE Section 3.3: The staff will reevaluate the thennal-hydraulic stability for Dresden 3 at the next reload cycle. The evaluation will consider permanent approval of the COTRANSA 2 analytical methodology including the results of benchmarking tests in the high decay ratio area.

..,\\

  • (3)

SE Section 4.0: The licensee should address the con*sequences of-the Fuel Handling Accident for ENC fuel types prior to reaching exposure values in excess of 38,000 MWD/MTU.

In addition, the licensee should infonn the NRC of any changes in planned refueling schedules which would result in batch average exposure values* for GE fuel of greater than 38000 MWD/MTU.

10.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change to a requirement with respect to the installation or use of a facility component located* within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance require-.

  • ments.

The staff has detennined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pur-suant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

11.0 The staff 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, and {2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security nor to the health and safety of the public.

Principal Contributor:

M. McCoy

  • Dated: July 24, 1986

~

REFERENCES

1.

.Letter, J. R. Wojnarowski (CECo) to H. R. Denton (NRC)., Proposed License Amendment Dresden Station Unit 3 Cycle 10 Reload, February 21; 1986 (with Attachments).

2.

XN-NF-85-57, Dresden Unit 3, Cycle 10 Reload Analysis, September 1985.

3.

XN-NF-85-62, Dresden Unit 3, Cycle 10 Plant Transient Analysis, September 1985.

4.
5.
6.
7.
8.

9~

XN-NF-85-63, Dresden Unit 3, LOCA-ECCS Analysis, September 1985.

XN-NF-85-67, Revision i, "Generic Mechanical Design for Exxrin Nuclear Jet Pump BWR Reload Fuel," April 1986.

XN-NF-524(P)(A), Revision 1, "Exxon Nuclear Critical Power Methodology for.

BWRs," November 1983.

Letter, C. 0. Thomas (NRC) to J. C. Chandler (ENC), "Acceptance for.

Referencing of Licensing Topical Report XN-NF-524 (P)," October 31, 1983.

XN-NF-79-71(P}, Revision 2, "ENC Plant Transient Methodology for Boiling Water Reactors," November 1981.

XN-NF-80-19(P}, Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," "1arch 31, 1985.

10. Letter, G. C. Lainas (NRC) to G. N. Ward {ENC), "Acceptance for Referencing of Licensing Topical Report XN-NF-80-19, Vol. 4, Rev. l," April 30, 1986.
11. Technical Resolution of Generic Issue B Thennal Hydraulic Stability (Generic Letter No. 86-02}, January 23, 1986.
12. Letter, D. Eisenhut (NRC) to R. Gridley (GE)t.... "Review of Low Core Flow Effects on LOCA Analysis for Operating BWRs," May 19, 1978.
13. Letter, J. R. Wojnar~wski (CECo) to H. R. Denton (NRC), with attachment
  • "Supp 1 ement to Proposed License Amendment -. Cycle 10 Re 1 oad," April 18, 1986.
14.

XN-NF-80-19 (P) (A), Vol. 4, "EXXON Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodoloqy to BWR Reloads," EXXON Nuclear Company,. September 1983 *

  • 15. Staff SE.for EXXON Nuclear Plant Transient Methodology for BWRs (Supplements 1 through 3 to XN-NF-79-71; Revision 2), to be published.
16. Letter, H. Bernard (NRC) to G. F. Owsley (ENC), "Acceptance for Referencing of Topical Report XN-NF-512, Revision l," July 22, 1982.
17. Letter, C. O. Thomas (NRC) to J. C. Chandler (ENC), "Acceptance for Referencing of Licensing Topical Report XN-NF-734, Confirmation of the XN-3 Critical Power Correlation for 9X9 Fuel Assemblies," February 1, 1985.
18.

NEDE-22148-P-A, General Electric Topical Report, "Extended Burnup Evaluation.Methodology," February 1986.

19. Letter, J. Zwolinski.(NRC) to D. J. Farrar (CECo), Request for Additional Information, June 3, 1986.
20. Letter, J. Wojnarowski (CECo) to H. R. Denton (NRC), Response to Request for Additional Information, June 12, 1986.
21.

XN-NF-82-06 (P) "Qualification of Exxon Nucle~r Fuel for Extended Burnup,"

March 2, 1982, and Supplements 1, 2, 4 and 5.

22~ Letter, H. N. Berkow (NRC) to J. F. Quirk (GE) "Safety Evaluation Report.

on General Electric ECCS Evaluation Methodology for Single Loop.Operation, NED0-20566-2, Rev. l," March 5, 1986.

23. Letter, J. R. Wo.inarowski (CECo) to H.

R~ Oenton (NRC), "Additional Information Regarding Cycle 10 Transient Analyses," with Attachment.,

"Modification of Dresden Unit 3 Cycle 10 MCPR Limits," June 20, 1986.

24.

"XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraul. ic Core Analysis," XN-NF-84-105(P), Volume 1, Exxon Nuclear Co., May 1985, and Supplements 1 and 2.

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