ML17146B004

Safety Evaluation Supporting Amend 72 to License NPF-14

ML17146B004
Person / Time
Site:	Susquehanna
Issue date:	10/09/1987
From:	Office of Nuclear Reactor Regulation
To:
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ML17146B003	List: ML17146B002 ML17146B004
References
NUDOCS 8710230136
Download: ML17146B004 (17)

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/y Cp kg**4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 72 TO FACILITY OPERATING LICENSE NO.

NPF-14 PENNSYLVANIA POWER 8 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION, UNIT 1 4

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1.0 INTRODUCTION

By letter dated June 19, 1987, Pennsylvania Power 5 Light Company (the licensee) requested an amendment to Facility Operating License No.

NPF-14 for the Susquehanna Steam Electric Station (SSES),

Unit 1.

The proposed amendment furnished information to support authorization for SSES Unit 1 operation with 9X9 reload fuel by Advanced Nuclear Fuels (ANF) Corporation and would revise the SSES Unit 1 Technical Specifications in support of the forthcoming fuel reload and restart for Cycle 4 operation.

The Cycle 4 (hereafter referred to as S1C4) reload will consist of 240 new 9X9 fuel bundles intermixed with 488 ANF 8X8 and 36 General Electric (GE) 8X8 fuel bundles from the previous cycle.

The new 9X9 bundles are comprised of 79 active fuel rods and two inert water rods.

In support of the S1C4 reload, the licensee submitted topical reports which summarize the reload scope, the plant transient analyses, and the design and safety analyses.

Specifically, the licensee has requested to change the following Technical Specifications:

Definitions 1.2 and 1.13, related to fuel exposure and fraction of limiting power density Specification 3/4.2.1, related to Average Planar Linear Heat Generation Rate (APLHGR)

Specification 3/4.2.2, related to Average Power Range Monitor (APRM)

Setpoints Sp'ecification 3/4.2.3, related to Minimum Critical Power Ratio (MCPR)

Specification 3/4.2.4, related to Linear Heat Generation Rate (LHGR)

Specification 3/4.4. 1.1-2, related to Recirculation Loops - Single Loop Operation (SLO)

Specification 5.3.1, related to Fuel Assemblies 871023013b 871009 PDR ADO'5000387,,!,

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2.0 EVALUATION The staff has evaluated the licensee's S1C4 core reload request by considering the adequacy of (1) fuel mechanical

design, (2) thermal hydraulic

design, (3) transient and accident analysis, and (4) the proposed Technical Specifiction changes.

The staff's evaluation is sumnarized as follows.

2.1 Fuel Mechanical Desi n

The S1C4 core reload will include 240 ANF Corporation 9X9 fuel assemblies with the designation XN-3.

These reload assemblies contain 79 fuel rods and two water rods.

The 240 assemblies will have a bundle enrichment of 3.31 percent.

The fuel design and safety analysis for the 9X9.fuel are described in. the SSES 1 specific report PL-NF-87-005 and the generic mechanical design report'XN-NF-85-67, Revision 1.

The staff approved the latter report and issued its Safety Evaluation on July 23, 1986.

Table 2. 1 of XN-NF-85-67, Revision 1 gives the pertinent design data for the ANF 9X9 fuel.

Neutronic values specific to the SIC4 reload are given in Table 4. 1 of PL-NF-87-005.

The burnable poison rods contain 4.00 weight percent gadolinia blended with 3.27 weight percent U-235 to reduce the initial reactivity.

The ANF SN-3 fuel is 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 review of the information in Table 2.1, we find the mechanical design of the ANF 9X9 fuel for the SlC4 reload to be acceptable.

However, approval of extended exposure limits for future operating cycles is contingent upon our approval of Supplements to XN-NF-82-06(P) related to 9X9 fuel.

Rod Pressure For the SlC4 ANF 9X9 reload fuel, calculation of the fuel rod internal pressure was done in accordance with acceptance criteria cited by ANF.

The evaluation was performed with RODEX 2A which is a revision of the

~ RODEX2 code (revised fission gas release model) used in the analysis of previous ANF fuel designs.

Our review of the RODEX 2A topical report is complete and the staff Safety Evaluation was issued on June 24, 1986.

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 Bow Our review of XN-NF-85-67, Revision 1 has been completed.

Based on that review we conclude that ANF has demonstrated conformance to approved rod bow design limits.for minimum gap, spacing to a fuel assembly exposure of 23,000 MWD/HTU for the 9X9 fuel.

Projected peak assembly burnups for the SlC4 reload is in the range of 11,000-13,000 NWD/HTU for the 9X9 fuel.

Additional information on rod bow measurements on the ANF 9X9 Lead Test

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Assemblies has been provided to justify burnup exposure levels up to 40,000 MWD/MTU.

Our review of the information for the fuel assembly exposure level above 23,000 MWD/MTU is not complete.

Therefore, the ANF 9X9 fuel is approved for S1C4 only (the fuel exposure level for SlC4 is not expected to exceed 13,000 MWD/MTU).

Future approval of operation with ANF 9X9 fuel for exposures beyond 23,000 MWD/MTU is contingent upon our approval of the additional rod bow considerations when the staff review is complete.

Fuel Centerline Meltin The design basis for the ANF fuel centerline temperature is that no fuel centerline melting should result from normal operation including transient occurrences.

The results of an evaluation reported in the S1C4 reload analysis report PN-NF-87-005 were based on RODEX 2A.

RODEX 2A has been previously reviewed and approved and the staff has concluded that the generic methodology for the ANF 9X9 fuel is acceptable for the S1C4 reload fuel.

Claddin Swellin and Ru ture The cladding swelling and rupture models in XN-NF-82-07 (EXXON Nuclear Company Cladding Swelling and Rupture Model) have been approved for use in the ANF (old ENC)

ECCS Evaluation Model and have been incorporated in the approved ANF EXEM/BWR ECCS model.

This model was used in the ANF ECCS analysis for the S1C4.

The staff has verified that ANF is using the approved model for the 9X9 fuel ECCS analysis and we find the application to be acceptable.

Linear Heat Generation Rate (LHGR

- Limit for ANF-9X9 Fuel The licensee has provided a figure of LHGR Limit vs Planar Exposure for the ANF 9X9 fuel to be incorporated into the SSES Unit 1 Technical Specifications (Figure 3.2.4;2.-2).

This Figure was approved in the staff's safety evaluation for licensing topical report XN-NF-85-67(P),

dated July 23,

1987, and reflects the design values which have been previously reviewed and approved.

Based on the results of the generic review we find the LHGR limits for the 9X9 fuel to be acceptable.

This acceptability also applies to the exposure-dependent LHGR provided in proposed Figure 3.2.2-1 which is based on ANF's "Protection Against Fuel Failure" concept which was also part of the generic review.

LOCA-Seismic Mechanical Res onse The licensee has discussed the mechanical response of the ANF 9X9 fuel assembly design during LOCA-seismic events.

The discussion included a

comparison of the physical and structural properties of the new 9X9 fuel and the prior ANF and GE BX8 fuel.'he staff has reviewed this information in connection with a previous review (the staff Safety

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Evaluation Supporting Amendment No.

31 to Facility Operating I icense No, NPF-22 dated October 3, 1986).

The staff has confirmed that the physical and structural characteristics of the ANF and GE fuel assemblies are sufficiently similar so that the mechanical response to design Seismic-LOCA events is essentially the same.

Based on the considerations discussed

above, we conclude that the staff's.original analysis is applicable to SSES Unit I and the analysis indicating that the design limits are not exceeded is also acceptable.

Nuclear Desi n

ANF nuclear design methodologies for SIC4 are updated to reflect criteria applicable -to the ANF fuel.

The SIC4 reload replaces about one-third of Cycle 3 fuel with new ANF 9X4 <uel.

The loading pattern is a normal type of scattered configuration.

The bundle average enrichment of the new assemebly is 3.31 weight percent U235.

The beginning o< cycle shutdown margin is calculated to be 1.63 percent delta-k/k, and the R factor is zero, thus the cycle minimum shutdown margin is well in excess of the required 0.38 percent delta-k/k.

The Standby Liquid Control System also fully meets shutdown requirements.

The existing new fuel storage calculations are based on k-infinity of the assembly.

Based on new calculations by ANF with consideration given to the 9X9 fuel, if the maximum enrichment zone is such that k-'infinitv is less than or equal to I.388 at limiting state conditions then the required criticality limits are met.

The existing spent fuel pool criticality calculations have met criteria using a

U235 assembly average enrichment of less than 4.00 percent and no burnable poison.

Since the maximum enrichment of the new fuel is 3.42 percent, the new calculations show adequate margin to spent fuel pool criticality.

The SSES will continue to use the EXXON (now ANF)

POWERPLEX core monitoring system to monitor reactor parameters.

The system has been in

.use during all SSES Unit I operating cycles and has provided suitable monitoring and predictive results.

2.2 THERMAL HYDRAULIC DESIGN The review of the thermal-hydraulic aspects of the SIC4 reload consisted of the following; (a) the compatibility of the ANF 9X9 and prior ANF 8X8 fuel bundles; (b) the fuel cladding integrity safety limit; (c) the operating limit minimum critical power ratio (OLMCPR); (d) thermal hydraulic stability fot S1C4; and (e) the proposed technical specifications.

The obiective of the review was to confirm that the thermal-hydraulic design of the reload core was accomplished using acceptable analytical

methods, provided an acceptable margin of safety from conditions which

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H draulic Com atibilit Since a

BWR core is a series of parallel flow channels connected to a

co+non 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 SIC4 core have been determined in single phase flow tests of full scale assemblies.

Additional discussion of the effects of hydraulic compatibility on thermal'margin were presented in the S1C4 reload report.

Based on our review of the information provided in the pertinent documentation we conclude that the ANF fuel types are hydraulically compatible.

Thermal-H draulic Stabilit The thermal-hydraulic stability (THS) of the projected Cycle 4 core was analyzed using the methods identified in Exxon Report XN-NF-80-19, Volume 4, Revision 1.

That report cites the use of the COTRAN model for use in the analysis of core thermal-hydraulic stability.

For two pump minimum flow, the maximum decay ratio computed with the ANF methodology for SIC4 operation is 0.74 at. the APRM rod. block intercept line (64 percent rated power).

The Cycle 4 reload is the first full reload batch of ANF 9X9 fuel for SSES Unit 1.

On line stability measurements at the SSES and Grand Gulf I reactors have demonstrated that a single reload of ANF 9X9 fuel has little impact on the overall core stability.

The licensee has previously implemented approved surveillance Technical Specifications for detecting and suppressing power oscillations in regions of the power-flow map considered susceptible to potential instability.

Extended operation in the single loop operation mode is not presently permitted for SSES Unit 1.

Based on these considerations we conclude that acceptable THS provisions have been made for the proposed one-third core reload with the ANF 9X9 fuel in SIC4.

2.3 TRANSIENT AND ACCIDENT ANALYSES Minimum Critical Power Ratio Safet Limit The minimum critical power ratio (MCPR) safety limit for the Cycle 4 reload was determined by the licensee to be 1.06 for all fuel types.

The methodology for Cycle 4 is based on ANF revised critical power methodology in XN-NF-524, Revision I, which incorporates a constant flow MCPR formulation for BWR applications.

The staff has completed its

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generic review of XN-NF-524 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 the new 9X9 fuel type.

The methodology of XN-NF-524, Revision I was applied generically for the upcoming Cycle 4 and is considered applicable to the resident GE 8XB fuel as well as the ANF fuel.

The staff has verified through its review of the SIC4 transient analysis report XN-NF-87-22 that the methodology for determining uncertainties and the application in determining the MCPR safety limit is in accordance with NRC approved methodology and is acceptable.

0 erational Transients 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 (delta-CPR) in the initial critical power ratio.

The ANF transient methodology is basically the same as that used and approved for recent plant reloads with ANF 9X9 fuel.

Certain aspects of the methodology as identified in the following discussion have received more recent NRC approval.

ANF examined the standard transient events and the SIC4 Transient Analysis and presented the results for the more limiting events.

The most limiting core wide transients were the Load Rejection Without Bypass (LRWB) and the Feedwater Controller Failure (FWCF).

The events were analyzed at the rated condition.(104%

power/100% flow) and with End-of-Cycle Recirculation Pump Trip (EOC-RPT) operable.

The additional aspect of the ANF plant transient methodology recently approved by the

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is the XCOBRA-T code which is used in the determination o< the thermal margin for the transients.

The analyses were all done with approved methodologies and the results are acceptable.

The calculated delta-CPR for the LWRB is equal to 0.22.

The resulting MCPR operating limit of 1.28 is acceptable for incorporation into the SIC4 Technical Specifications for all fuel types.

It was assumed for these transients that the RPT is operable.

The limiting MCPR event (LRWB) was. also calculated for limiting extension conditions assumino an inoperable RPT.

This resulted in increased MCPR limits which are also proposed for SIC4.

These calculations follow standard procedures for the inoperable RPT extension and operation whithin these limits is acceptable for SIC4.

Compliance with overpressurization criteria was demonstrated by analysis of Main Steam Isolation Valve (MSIV) closure with MSIY position switch failure.

Six safety-relief valves were assumed out of service.

Maximum pressure was 105 percent of vessel design pressure, well under the 110 percent criterion.

The calculation was done with approved methodology and results are acceptable.

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The LOCA analyses for SSES Unit 2 Cycle 2 performed for a full core of ANF 9X9 fuel is applicable for the S1C4 residual and reload ANF fuel.

These analyses have covered an acceptable range of conditions, have been performed with approved methodology, and the resulting Technical Specification MAPLHGR values for the ANF fuel remain acceptable.

Reactivit Insertion Transients The control rod withdrawal error, the fuel loading error and the rod drop accident were evaluated for Cycle 4.

The licensee used methods described in XN-NF-80-19, Volume 4.

Using a

Rod Block Monitor setting of 108 percent of full power results in a delta-CPR of 0. 18 for the control rod withdrawal error transient for 9X9 fuel.

The change in CPR due to a fuel loading error is 0.08.

These values are comparable to previous reloads and are not limiting.

The rod drop accident was analyzed with approved ANF methodology.

The resulting maximum fuel enthalpy of 91 cal/gm is within the established limit of 280 cal/gm.

The staff finds that the analysis and results are acceptable.

2.4 TECHNICAL SPECIFICATION CHANGES The following Susquehanna Steam Electric Station Unit 1 Technical Specification changes have been proposed for operation during reload Cycle 4:

(1)

DEFINITIONS pages 1-2 and 1-3, parts of Bases pages B 2-1 an 8 2-2, Limiting Conditions for Operating (LCO) pages 3/4 2-1, 3/4 2-10a and 3/4 4-1c, Figure 3.2.1-2, Bases pages B 3/4 1-1, 8 3/4 2-1 and 8 3/4 4-1, and Design Features page 5-6:

Changes were made to reflect the corporate change from Exxon Nuclear Company (ENC) to Advanced Nucleat Fuels (ANF) Corporation, to identify and describe the new fuel design and to 'incor'porate editorial changes.

These changes are administrative only with no safety significance and are therefore acceptable.

(2)

Bases pages B 2-1 and B 2-2, Section 2.1.1 - THERMAL POWER, Low Pressure or Low Flow:

The changes provide a basis for the range of validity for use of the cr'itical heat flux correlation for the reload 9X9 fuel type.

The basis was approved as part of a generic review and is acceptable.

(3)

Figure 3.2.1-3:

The MAPLHGR limits for the new fuel are added.

This addition is acceptable.

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(4)

LCO page 3/4 2-5 and 'Figure 3.2.2-1:

For ANF fuel, the LHGR Limits and LCO for APRM setpoints are based on a

generic review and approval and are acceptable.

(5)

Figure 3.2.3-1 and 3.2.3-2 and LCO page 3/4 2-10a:

These figures reflect the new HCPR limits for Cycle 4 and are acceptable.

(6)

LCO page 3/4 2-10c and Figures 3.2.4.2-1 and 3.2.4.2-2:

For ANF fuel, the LHGR limits are based on a generic review and approval and are acceptable.

The identified changes provided in the licensee's submittal are acceptable as proposed.

2.5 Restr ictions We have reviewed the reports submitted for the Cycle 4 operation of SSES Unit 1.

Based on this review we conclude that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses are acceptable.

Sufficient basis has been provided to allow the addition of 240 ANF 9X9 fuel bundles in the SSES Unit 1 core.

The Technical Specification changes submitted for this reload suitably reflect the necessary modifications for operation in this cycle.

Our review as discussed in the evaluation sections above has identified certain restrictions relating to our incomplete review of the ANF 9X9 fuel.

The approval of the ANF 9X9 fuel is therefore limited to the upcoming Cycle 4 only.

Specifically, the approval of extended exposure limits for the 9X9 fuel beyond 30,000 HWD/HTU batch average exposure for future operating cycles is contingent upon our approval of XN-NF-82-06(P) and Supplements 1, 2, 4, and 5.

Also, approval of the additional rod bow considerations is required for the ANF 9X9 fuel for exposure beyond 23,000 HWD/HTU expected to occur'in the future cycles.

3. 0 ENVIRONMENTAL CONSIDERATION 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.

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 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

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9 criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9)

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Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environmental assessment need be prepared in connection with the issuance of this amendments

4.0 CONCLUSION

The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Re ister (52 FR 26593) on July 15, 1957 and consulted with the State oOOennsy van>a No public comments were received, and the State of Pennsylvania di d not have any comments.

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 Commi ss ion '

regulations and the issuance of this amendment will not be inimical to the comon defense and security nor to the health and safety of the public Principal Contri butor :

M ~ McCoy Dated:

october 9

1987

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