ML20215M150

Safety Evaluation Supporting Amend 23 to License NPF-29

ML20215M150
Person / Time
Site:	Grand Gulf
Issue date:	10/24/1986
From:	Office of Nuclear Reactor Regulation
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ML20215M146	List: ML20215M142 ML20215M150
References
TAC-61930, NUDOCS 8610300069
Download: ML20215M150 (9)

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WASHINGTON, D. C. 20555 8

,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 23 TO FACILITY OPERATING LICENSE N0. NPF-29 PISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.

SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416

1.0 INTRODUCTION

By [[letter::AECM-86-0186, Requests Response to Encl Rev to Justification for Relief Request I-00008 Re ASME Section XI Code Requirements for Inservice Insp of CRD & in-core Housing Welds & Flange Bolting,Per ,By 860801|letter dated July 14, 1986]], (Ref. 1), Mississippi Power & Light Company, (MPAL or the licensee) reauested an amendmet to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1.

The proposed amend-ment would change Technical Specifications to aDow the operation of Grand Gulf Nuclear Station Unit 1 for Cycle 2 (GG1C2) wi+h a reload using Exxon manufactured fuel assemblies and Exxon analyses and methodologies. Enclosed with the July 14, 1986 submittal were the reauested changes to the Technical Specifications and reports discussing the reload and analyses made to sup-port and justify the second fuel cycle operation with General Electric (GE) and Exxon fuel and the changes to the Technical Specifications. By letters dated August 15 and September 5,1986, the licensee provided supplemental information describing additional analyses and providing results. By letters

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dated September 4 and October 3,1986, the licensee proposed additional changes to some of the proposed Technical Specifications. These submittals are identified in References 1 through 9.

The notice of consideration of issuance of this license amendment was published in the Federal Register before the licensees October 3,1986, submittal. The October 3,1986, submittal contained clarifications to previously submitted proposed Technical Specifications to make them more specific and therefore less likely to be misinterpreted. The notice of consideration accurately described the license amendment request and the clarifications do not affect the substance of the requested amendment.

Cycle 2 will be the first use of Exxon fuel and analysis in this reactor.

However, similar reloads with Exxon fuel have been done for Dresden 2 and 3, and more recently for Susquehanna 1 and 2 and Washington Public Power l

Supply 2 (WNP2). These reloads and the associated Exxon methodologies were extensively reviewed and approved (see for example Reference 10). These methodologies are generally applicable and were used for the most part for GGIC2 analyses.

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. Beyond the switch to Exxon-provided reload fuel, there is little that is unusual about GGIC2, and the proposed Technical Specification changes are primarily related to the use of Exxon fuel and accompanying analyses and methodology, terminology or related operational approaches. The reload related changes are similar to the corresponding changes for the Susquehanna 1 second cycle introduction of Exxon fuel (Ref. 10) and the other Exxon reloads mentioned above. During the first cycle Grand Gulf received approval and appropriate Technical Specifications for operation in the Maximum Extended Operating Domain (ME0D) and for single recirculation loop operation. This will continue in cycle 2 and appropriate analyses have been done for this reload.

The reload and its analyses will be discussed in the following evaluation.

2.0 EVALUATION 2.1 Reload Description The GGIC2 reload will retain 536 General Electric (GE) fuel assemblies from the first cycle and will add 264 Exxon manufactured XN-18x8, 2.81 percent average, 2.99 percent peak radial average U235 enriched fuel assemblies. The XN-1 fuel assemblies are similar to those used in the Susquehanna 1 second cycle (SIC 2) reload. The loading pattern will be a conventional scatter pattern with low reactivity fuel on the periphery.

2.2 Fuel Design The Exxon XN-1 fuel assembly used for GG1C2 is essentially the same as that used for the SIC 2 reload. There are slight differences in the fuel enrichment and gadolinium placement patterns, but the significant mechanical and thermal-hydraulic design elements are the same and power distributions are similar. The methodologies used for the fuel design and analysis are the same as those developed and approved during the SIC 2 reload review and then approved for the Susquehanna 1 Cycle 3 (SlC3) reload. The design and analyses of the XN-1 fuel assembly as used in GG1C2 are thus acceptable.

For GGIC2 the Technical Specifications will provide for a Linear Heat Generation Rate (LHGR) specification as a function of fuel burnup for the Exxon fuel. A similar specification was accepted for SIC 3 as a result of discussions between the NRC staff and Exxon on the need for a LHGR specification. The l

specification is based on the approved fuel design methodology as discussed in the SIC 3 review (Ref. 11) and is acceptable.

The mechanical response of Exxon fuel assemblies to design Seismic-LOCA events is essentially the same as for GE assemblies. The channel boxes for the new l

fuel are GE channels. Similar to the SIC 2 and SIC 3 reloads the analyses indicating that the design limits are not exceeded are acceptable.

The Exxon fuel has been analyzed for operation in the high flow region of MEOD l

(Ref. 8). This includes evaluation of vibration and assembly levitation.

Similar calculations have been approved for WNP2. The results are acceptable.

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. 2.3 Nuclear Design i

The nuclear design for GGIC2 has been performed with Exxon methodologies previously reviewed and approved, and which were listed in the review for the SIC 2 reload (Ref. 10). The fuel loading pattern is a normal type of scattered 7

configuration. The beginning of cycle shutdown margin is 4.07 percent delta k and at minimum conditions is 2'.73 percent delta k, well in excess of the reouired 0.38 percent delta k.

The Standby Liquid Control System also fully meets shutdown requirements. These and other GGIC2 nuclear design parameters have been obtained with previously approved methods and fall wM.nin expected ranges. Thus the nuclear design is acceptable.

GGIC? will use the Exxon POWERPLEX core monitoring system to monitor reactor parameters. We have not specifically reviewed details of this system (nor have we in the past reviewed details of the GE process computer monitoring system),

but we have reviewed the principal methodologies involved in the system and consider them to be appropriate and acceptable. The system has been in use in Susquehanna and has provided suitable monitoring and predictive results.

The spent fuel pool has been previously analyzed with an infinite array of 8x8 fuel assemblies with an enrichment of 3.5 percent U235 and no burnable poison.

The new fuel for cycle has an axial maximum enrichment of 2.99 percent and thus falls within the limits of previously reviewed and approved levels and is acceptable.

2.4 Thermal Hydraulic Design i

l The Exxon thermal-hydraulic methodology and criteria used for the GGIC2 design and analysis is the same as that used and approved in the SIC 2, SIC 3 and WNP2 j

reloads. The previous reviews concluded that hydraulic compatibility between i

GE and Exxon fuel is satisfactory and the calculation of core bypass flow and the Safety Limit Minimum Critical Power Ratio (MCPR) are acceptable. This is also the case for GGIC2. The Safety Limit MCPR continues to be 1.06 for two recirculation loop operation (the same value as for the first cycle GE methodology) as it is for Susquehanna and WNP2 and this is acceptable. As l

discussed in Section 2.6 below, the Operating Limit MCPR for GGIC2 (1.18) remains the same as for Cycle 1 since the analyses for Cycle 2 result in values no larger than this value.

l GGIC2 already has Technical Specifications from Cycle 1 allowing and I

controlling one recirculation loop operation, including changes required on l

limits for Maximum Average Planar Linear Heat Generation Rate (MPLHGR),

Average Power Range Monitor (APRM) settings, and Safety limit MCPR. Since the l

Exxon fuel is hydraulically compatible with the GE fuel, the previous analyses are also applicable to the Exxon XN-1 fuel loading. Similar to the approval for the Susquehanna one loop operation review (Ref.10), the above first cycle one loop limit changes are also acceptable for this Exxon reload.

Grand Gulf also has Technical Specifications approved during the first cycle for Thermal-Hydraulic Stability surveillance and the subseouent suppression of possible oscillations. These specifications set up regions on the power-flow map in which operation is not permitted or regions in which detection of i

. potential power oscillations must be performed using the incore neutron detector system. Because of the use of Exxon fuel and methodology for Cycle 2, changes have been made to the specification of these regions. There have been changes to Technical Specification Figure 3.4.1.1-1 and to relevant Bases (Ref.

6 and 7, with the figure in 7 superseding that of 6). These changes were the result of several discussions between the staff and Grand Gulf. The change provides for a Detect and Suppress region above the 80 percent load line and bounded by the 39 and 45 percent flow lines and lines representing analytical Decay Ratios of 0.90 and 0.60. Operation is not permitted at flows below 39 percent or above the Decay Ratio of 0.90.

These limits are based on the staff interpretation of previous reviews of relevant Exxon methodology (Ref. 13) and approval of GE recommended surveillance mode of operation presented in References 14 and 15. This final version of the specification (Ref. 7) is acceptable. Grand Gulf will perform tests on stability during startup of Cycle 2 in cooperation with staff consultants from Oak Ridge. The above regions may be altered in the future as a result of these tests and corresponding analyses.

2.5 Transient and Accident Analyses The GG1C2 core will have 800 fuel assemblies, including 264 unirradiated Exxon XN-18x8 assemblies and 536 previously irradiated GE assemblies from Cycle 1.

The Exxon transient and accident analyses for Cycle 2 were based on the design and operational assumptions used for the analyses of Cycle 1.

References 1 through 5 and 16 through 21 describe the Exxon methodology used in the analysis of the plant transients and accidents for the Cycle 2 reload.

The Exxon Cycle 2 Reload Analysis Report (Ref. 2) which describes plant and cycle specific analysis results is supplemented by Reference 16 which describes the Exxon approach to core reload analyses and references more detailed methods rtports used in the safety analyses.

'; ore wide transients were analyzed with the COTRANSA computer code (Ref.17) which includes a one-dimensional neutron kinetics model for evaluation of the axial power shape response during transient events. This reference has been l

reviewed by the staff. The methods for calculating the system transient response were found to be acceptable.

Preparation of the Safety Evaluation i

Report (SER) and formal staff approval of this reference is in process.

Calculation of the change in Critical Power Ratio (CPR) during the core wide transient events involves the use of COTRANSA system results which serve as input to a hot channel analysis model used to calculate the delta-CPR values.

The original submittal for core-wide transient results for Cycle 2 were based on a COTRANSA hot channel model. During an internal Exxon review, a potential nonconservatism was identified in the formulation of this model for delta-CPR calculations. To resolve the problem, this potential nonconservatism was evaluated by the licensee with the XCOBRA-T model (Ref. 18). The XCOBRA-T model has been reviewed by by the staff and found to be acceptable. Writing of the SER for this model is in process. As discussed below, the application of X-COBRA-T to the plant-specific GGIC2 results confirmed the COTRANSA hot channel model results. Hence the application of the COTRANSA results to Cycle 2 of Grand Gulf Unit 1 is acceptable.

. In the initial submittal, the licensee referenced a generic report (Ref.19) as the basis for the predicted response of Grand Gulf Unit 1 to a loss of feedwater heating (LFWH) event during Cycle 2.

This generic report had not been reviewed by the staff. Hence, the staff requested formal submittal of a plant specific analysis of the event. This analysis, which was provided in Reference 8 is acceptable.

The rod withdrawal error (RWE) event for Cycle 2 of Grand Gulf Unit I was referenced to a generic study of this event provided in Reference 20. The staff has reviewed and evaluated this report which provides a statistical evaluation of the RWE and includes application to operation in the maximum extended operating domain (ME0D). The staff has found the report acceptable (Ref. 21).

The licensee evaluation of the loss of Coolant Accident (LOCA) for Grand Gulf Unit during Cycle 2 is summarized in Reference 1.

The analysis of the limiting break is provided in Reference 5.

The LOCA analysis methodology used to obtain the results of Reference 5 has been approved by the staff.

However, Reference 5 cites the Exxon generic jet pump BWR/6 LOCA break spectrum analysis of Reference 4 which was not reviewed by the staff. The staff requested a plant-specific break spectrum analysis that would be applicable to Cycle 2.

This information which was provided in Reference 8, is acceptable.

2.6 Chances To Technical Specifications The following changes have been requested for Grand Gulf Technical Specifications (TSs) and Bases (B) to accommodate the change to Exxon fuel, methodology and terminology.

For the most part these changes are similar to those approved for Susquehanna and WNP2 on changing to Exxon methodology.

(1) The definition 1.8 for Critical Power Ratio is changed (see Pef. 9) to reflect the change of methodologies for GEXL to XN-3, and is acceptable.

(2) TS 3/a.1.2: The change to the definition of reactivity anomaly from control rod density to a monitored k anomaly, reflects the use of a more direct parameter. POWERPLEX,w$Nhmaintainsaconsistent methodology between active determination and prediction, can monitor k,ff directly. The change is acceptable.

(3) TS 3/4.2.1: The language of this specification has been changed (see both Ref. I and 9) to reference more explicitly and clearly Figures 3.2.1-1, 2 and 3 and indicate two and one loop operation limits and multiplication factors. This is acceptable.

(4) Figure 3.2.1-1: The MAPLHGR curve for the GE low enrichment fuel assembly from the previous cycle, which is removed for Cycle 2, has been eliminated and replaced by the curve for the new Exxon fuel assembly.

This curve is based on the Exxon LOCA calcuations and is acceptable.

(5) Figure 3.2.1-2: MAPFAC curves for the Exxon fuel assembly have been f

added to the existing curves for GE fuel. These are based on the Exxon calculations of flow increase transients and are acceptable.

. (6) TS 3/4.2.4: This specification has been changed to add a reference to

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Figure 3.2.4-1 giving the Linear Heat Generation Rate (LHGR) curve for Exxon fuel. The GE fuel LHGR remains at 13.4 kw/ft. This is acceptable.

(7) Figure 3.2.4-1 has been added to give the Exxon fuel LHGR limit as a function of burnup. 1his is similar to previously approved curves for Susquehanna and WNP2.

It is based on approved methods for Exxon fuel mechanical design analyses, and is compatible via the local peaking factor with the MAPLH3R limit of Figure 3.2.1-1 (Ref. 2). It is acceptable.

(8) Figure 3.4.1.1-1:

As discussed in Section 2.4 on thermal-hydraulic stability, the power-flow plane regions requiring monitoring or no operation have been changed. The final version of the figure giving these regions was presented in Reference 7.

It is acceptable.

(9) The following Bases have been changed to provide a description of Exxon methodology in addition to or in place of GE methodology. The changes are consistent with the Technical Specification changes and are similar to changes approved in the Susquehanna and WNP2 reviews. They are all acceptable.

B2.1, discusses the Exxon methodology relating to the Xn-3, critical power correlation, particularly at low flow and pressure, and the Exxon Safety Limit (MCPR).

B3/4.1.2, discusses Exxon methodology for MAPLHGR including MAPFAC.

B3/4.2.3, discusses Exxon methodology for MCPR.

B3/4.2 (References), deletes several GE references and adds an Exxon reference.

l B3/4.3.10 and B3/4.4.1 discusses the surveillance regions of Figure 3.4.1.1-1 for thermal-hydraulics stability (see Ref. 6 and 7).

l 2.7 Sumary The NRC staff has reviewed the reports submitted for the Cycle 2 reload of Grand Gulf Unit I with Exxon fuel and with Exxon methodology and analysis. Based on this review the staff concludes that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design and transient and l

accident analyses are acceptable. The changes to the Technical Specifications submitted for this reload suitably reflect the changes from GE to Exxon method-l ology and the operating limits associated with these changes and reload parameters.

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. 2.8 References 1.

Letter from 0. D. Kingsley, MPLC, to H. Denton, NRC, dated July 14, 1986,

" Grand Gulf Nuclear Station Unit 1, Proposed Amendment to the Operating License, Cycle 2 Reload."

2.

XN-NF-86-35 Rev. 3, dated August 1986, " Grand Gulf Unit 1 Cycle 2 Reload Analysis."

3.

XN-NF-86-36 Rev. 3, dated August 1986, "GGIC2 Plant Transient Analysis."

4.

XN-NF-86-37, dated April 1986, " Generic LOCA Break Spectrum Analysis for BWR/6 Plants."

5.

XN-NF-86-38, dated June 1986, " Grand Gulf Unit 1 LOCA Analysis."

6.

Letter from 0. D. Kingsley, MPLC, to H. Denton, NRC, dated August 15, 1986, " Addendum to Cycle 2 Reload."

7.

Letter from O. D. Kingsley, MPLC to H. Denton, NRC, dated September 4, 1986, " Addendum #2 to Cycle 2 Core Stability."

8.

Letter from 0. D. Kingsley, MPLC to H. Denton, NRC, dated September 5, 1986, " Additional Information (LOFWH, LOCA, Fuel Liftoff)."

9.

Letter from 0. D. Kingsley, MPLC to H. Denton, NRC, dated October 3, 1986,

" Supplement to Cycle 2 Reload."

10. Letter from W. Butler, NRC, to N. W. Curtis, Pennsylvania Power & Light (PP&L), dated May 22, 1985

11. Letter from E. Adensam, NRC, to H. W. Keiser, PP&L, dated April 11, 1986,

" Amendment to Susquehanna Unit 2 for Cycle 3 Reload."

12. Letter from E. Adensam, NRC, to H. W. Keiser, PP&L, dated April 11, 1986,

" Amendment Nos. 56 and 26 to Susquehanna Steam Electric Station Units 1 and 2."

13 Letter from C. Thomas, NRC, to J. Chandler, Exxon, dated May 10, 1984,

" Stability Evaluation of Boiling Water Reactor Cores."

14 General Electric Service Information Letter No. 380, Revision 1, February 10, 1984.

15. Technical Resolution of Generic Issue B-9 Thermal Hydraulic Stability (Generic Letter No. 86-02), dated January 22, 1986.

16. XN-NF-79-71(P) Revision 2, dated November 1981, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors."

e D 17. " Exxon Nuclear Methodology for Boiling Water Reactors: THERMEX Thermal Limits Methodology, Sunnary Description," XN-NF-80-19(P), Volume 3, Revision 1, Exxon Nuclear Company, Richland, Washington (April 1981).

18. XN-NF-84-105(P) dated May 1985, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis."

19. XN-NF-900(P) dated February 1986, "A Generic Analysis of the loss of Feedwater Heating Transient for Boiling Water Reactors."

20. XN-NF-825(P) dated April 1985, "BWR/6 Generic Rod Withdrawal Error Analysis."

21. Letter from G. C. Lainas, NRC, to G. N. Ward, Exxon, dated October 14, 1986, " Acceptance for Referencing of Licensing Topical Report XN-NF-825(P), Supplement 2, BWR/6 Generic Rod Withdrawal Error Analysis, MCPRp for Plant Operations Within the Extended Operatign Domain."

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3.0 ENVIRONMENTAL CONSIDERATION

l This amendment involves changes to requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements.

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

4.0 CONCLUSION

The Commission made a proposed determination that the amendment involves no significant harards consideration which was published in the Federal Register (51 FR 33955) on September 24, 1986, and consulted with the state of Mississippi. No public coments were received, and the state of Mississippi did not have any coments.

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 this amendment will not be inimical to the common defense and the security nor to the health and safety of the public.

Prinicipal Contributors:

l H. Richings, Reactor Systems Branch, DBL C. Graves, Reactor Systems Branch, DBL Dated: October 24, 1986 l