ML17157A413

Safety Evaluation Supporting Amend 102 to License NPF-14

ML17157A413
Person / Time
Site:	Susquehanna
Issue date:	11/02/1990
From:	Office of Nuclear Reactor Regulation
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NUDOCS 9011080299
Download: ML17157A413 (18)

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Cy UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.102TO FACILITY OPERATING LICENSE NO. NPF-14 PENNSYLVANIA POWER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

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

1.0 INTRODUCTION

By letter dated July 2, 1990 (Ref. 1), the Pennsylvania Power and Light Company (PP&L) (the licensee) requested an amendment to Facility Operating License No. NPF-14 for the Susquehanna Steam Electric Station, Unit 1.

The proposed amendment would support authorization of Susquehanna Steam Electric Station, Unit No.

1 (Susquehanna

1) operation for Cycle 6 with 9x9 reload fuel supplied by Advanced Nuclear Fuels Corporation (ANF).

The Susquehanna 1 Cycle 6 (S1C6) reload will consist of 220 new ANF-5 9x9 fuel assemblies, 468 irradiated ANF 9x9 assemblies and 76 irradiated ANF 8x8 assemblies.

S1C6 will contain no General Electric Company (GE) fuel assemblies.

The new 9x9 fuel has similar operating characteristics (mechanical, thermal-hydraulic and nuclear) to the previously used ANF 9x9 reload fuel.

In addition to the fuel changes, there will also be a replacement of 50 of the current control rod blades with GE designed Duralife 160C blades.

In support of the SlC6 reload, the licensee submitted reports which summarize the reload scope (Ref. 2), the plant transient analyses (Ref. 3),

and the design and safety analyses (Ref. 4).

Except for the added discussion of the control rod blade replacement, the

analyses, evaluation and results submitted for S1C6 and the reports referenced are similar to those submitted and approved by the NRC staff for the reload for Cycle 5.

2.0 EVALUATION 2.1 Fuel Mechanical Desi n

The S1C6 core reload will include 220 ANF 9x9 fuel bundles with the designation ANF-5.

These reload bundles contain 79 fuel rods and 2 water rods.

The 220 fuel bundles will have a bundle average enrichment of 3.52 or 3.21 weight percent uranium-235.

The fuel design and safety analysis are described in the Susquehanna 1 specific report PL-NF-90-003 (Ref. P~nd the generic mechanical design rept XN-NF-85-67(P)(A), Revision I+Ref. ~~-

[he NRC has approved the latter report and issued a Safety Evaluation~eo~rt n July 23, 1986 (Ref. 6).

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Table 2.1 of XN-NF-85-67(P)(A), Revision 1 (Ref. 5) gives the pertinent design data for ANF 9x9 fuel.

Neutronic values specific to the SlC6 reload are given in Table 4.1 of ANF-90-050 (Ref. 4).

The burnable poison fuel rods contain 4.0 or 5.0 weight percent gadolinia.

The analyses for S1C6 support fuel bundle discharge exposures of 37,000 MWd/MTU for ANF Bx8 fuel and 40,000 MWd/MTU for ANF 9x9 fuel.

The discharge exposures for these fuel types are based on the approved ANF topical report XN-NF-82-06(P)(A), Supplement 1, Revision 2 (Ref. 7).

Based on our review of the information presented, we find the mechanical design of the ANF 9x9 fuel for the S1C6 reload to be acceptable.

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For the S1C6 ANF 9x9 reload fuel, calculation of the fuel rod internal pressure was done in accordance with acceptance criteria cited by ANF in Reference 6.

The evaluation was performed with the RODEX2A computer code which has been reviewed and approved by the staff (Ref. 8).

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.

A figur e of LHGR limit versus planar exposure (MWd/MTU) for the ANF 9x9 fuel is incorporated into the Susquehanna 1 Technical Specifications.

This figure was previously approved to reflect the design values which have been reviewed and approved for the ANF 9x9 fuel in connection with the staff's review of XN-NF-85-67(P), Revision 1 (Ref. 5).

Based on the results of the generic'eview, the staff finds the current LHGR limits for the 9x9 fuel to be applicable for the new 9x9 fuel and to be acceptable.

The currently approved exposure limit (35,000 MWd/MTU) for the ANF 8x8 fuel remaining in the core will be exceeded during Cycle 6.

ANF has provided (Ref.

9) an analysis justifying the extension of the burnup limit to 37,000 MWd/MTU.

This analysis uses approved methodology and acceptance limits and the result, is acceptable.

The licensee has discussed the mechanical response of the ANF 9x9 fuel assembly design during LOCA-seismic events in Appendix B of Reference 4.

The discussion includes a comparison of the physical and structural properties of the ANF 9x9 fuel and the GE 8x8-fuel.

The staff has reviewed this information in connec-tion with a previous review (see Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.

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

The staff has confirmed that the physical and structural char acteristics of the ANF and GE fuel assemblies are suf-ficiently similar so that the mechanical response to design LOCA-seismic events is essentially the same.

Based on the considerations discussed

above, we conclude that the original analysis is applicable to Susquehanna 1 and the analysis indicating that the design limits are not.exceeded is acceptable.

2.2 Control Rod Blades PPKL intends to replace op to 50 of the original e&qU pment control rod blades for S1C6,t01I5peet the commitment,>jn tzspog~e to Bu~tin 79-26, Revision 1, to limit the B

depletion to no more than 34 percent; The replacement will be General Electric (GE) Duralife,160C blades.

They are designed to eliminate B

C tube cracking and increase blade life.

They have improved B

C tube mIterial, hafnium rods at the blade edge, additional B

C tubes,

)ncreased sheath thickness and other mechanical'esign improvemer%ts.

They are about 16

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With the exception of the improved crevice-free structure and an extended

handle, these blades are equivalent to the NRC approved Hybrid I Control Blade Assembly (Ref. 10).

The mechanical aspects, of the crevice-free structure have been approved by the NRC (Ref. 11).

GE has analyzed the blade neutronics using the same methodology as was used for the Hybrid I design.

The Duralife 160C blade has a slightly larger reactivity worth than original Susquehanna

blades, but it is within the criterion of nuclear interchangeability.

The blades weigh less than a

D lattice blade (Susquehanna is a C lattice) and the basis of the control rod drop accident drop velocity (which assumes a

D lattice rod) remains valid.

The scram times associated with the blade are not significantly different than for current blades, and there is a considerable margin to TS scr am speed limits.

The staff review of these blades concludes that they are acceptable for use in S1C6.

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D The nuclear design methodology used -,for S1C6 is that presented in the ANF report XN-NF-80-19(A), Volume 1 and Volume 1 Supplements 1 and 2 (Ref. 12),

and the PPSL report PL-NF-87-001-A (Ref. 13), which were reviewed and approved by the staff for application to Susquehanna core reloads.

The'beginning of cycle shutdown margin is calculated to be 1.07 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 fuel assembly.

Based on ANF calculations of 9x9 fuel, an average lattice enrichment of less than 3.95 weight percent uranium-235 and a k-infinity of less than or equal to 1.388 will meet the acceptance criterion of k-effective no greater than 0.95 under dry or flooded conditions.

Since the zone average enrichment of the new fuel is 3.44 weight percent uranium-235 and the maximum cold, uncontrolled, beginning-of-life k-infinity for the ANF fuel bundle enriched zones is 1.133, the ANF calculations show that the staff's acceptance criterion is met for the new fuel storage vault under dry and flooded condi-tions.

To preclude criticality at optimum moderation conditions, watertight covers and appropr iate procedures are used.

These are acceptable.

ANF also performed analyses for 9x9 fuel stored in the spent fuel pool.

A maximum enriched zone of less than 3.95 weight percent 'uranium-235 meets the staff acceptance criterion of k-effective no greater than 0.95.

Since the ANF-5 9x9'uel has a zone average enrichment of 3.64 or 3.31 weight percent uranium-235 the staff's acceptance criterion for spent fuel storage is met for the ANF-5 9x9 fuel.

Susquehanha will continue to use the ANF POWfRPLEFcCrf monitoring system to monitor core parameters.

The system has.bee~in usp for a number of cycles for both Susquehanna Unit 1 and [!nit 2'anV haY pddABed acceptable monitoring and predictive results.

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2.4 Thermal-H draulic ~Desi e

The minimum critical power ratio (MCPR) safety limit for the S1C6 reload was determined by the licensee to be 1.06 for all fuel types.

The methodology for S1C6 is based on the ANF methodology in XN-NF-80-19(P)(A), Volume 4, Revision 1 (Ref. 14), which has been approved by the staff.

The XN-3 correlation used to develop the MCPR safety limit has been approved for the ANF 9x9 fuel (Ref.

15).

ANF has determined that this correlation provides sufficient conservatism such that there is no need for any penalty due to channel bow for S1C6.

Susquehanna is a

C lattice core and uses channels for only one bundle lifetime.

For such cores ANF has determined that the conservatism is greater than the maximum expected delta CPR

( critical power ratio).

The staff has reviewed the ANF channel bow analyses methodology and it is acceptable for this analyses for S1C6.

The core bypass flow fraction has been calculated to be 10.0 percent of total core flow using the approved methodology described in XN-NF-524(P)(A), Revision 1 (Ref. 16).

This is used in the MCPR safety limit calculations and as input to the S1C6 transient analyses and is acceptable.

In response to Bulletin 88-07, Supplement 1 (Ref. 17) on BWR thermal-hydraulic

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stability, PP&L developed restricted operating regions on the power/flow operating map which were in compliance with the NRC recommendations.

Technical Specifications (TS) implementing these regions have been approved by the staff for Susquehanna 1.

Stability tests have been conducted in Susquehanna 2 with various amounts of ANF 9x9 fuel from succeeding

reloads, including all 9x9 fuel.

These have indicated no significant deterioration of decay ratio.

Decay ratios were low in all tests.

Calculations similar to those setting up the restrictive boundaries were done for S1C6.

This resulted in slight modifica-tions of the regions for this cycle.

TS implementing the ct}anges have been submitted.

This review concludes that the analyses are suitable and the changes to the TS are acceptable.

2.5 Transient and Accident Anal ses Various operational transients could reduce MCPR below the safety limit.

The most limiting transients have been analyzed to determine which event could potentially result in the largest reduction in the initial Critical Power Ratio (CPR), that is, the delta CPR.

The core wide transient which resulted in the largest delta CPR from a 104 percent power and a 100 percent flow condition is the generator load rejection without bypass event (LRWOB).

The delta CPR for this event is 0.28 for ANF 9x9 fuel, which is the most limiting fuel type.

When combined with a safety limit MCPR of 1.06 this results in a MCPR operating limit of 1.34 for S1C6.

The most limiting local transient, the control rod withdrawal error (CRWE), was analyzed to support a rod block monitor (RBM) setpoint of 108 percent and resulte~

a delta CPR of 0.26.

The LRWOB and the CRWE events were the most limitQg"events for S1C6 at rated power and flow conditions.

At lays than rat~ego~

the feedwater controller failure (FWCF) event is limiting and a'curve of YCPR=versus power, which is based on the FWCF results, is included in the Technical Specifications as a

power dependent MCPR operating limit.

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+I V,g, <4ighC At reduced flow conditions, the recirculation flow controller failure transient (RFIT) is limiting and MCPR operating limits for manual flow control reduced flow operation for SlC6 based on the analysis of this event are provided as a

Technical Specification figure of MCPR versus core flow.

The calculations of the thermal margin were performed with approved methodology (Ref. 18) and the resulting Technical Specification limiting curves are acceptable.

It was assumed for the above analyses that the turbine bypass system and the end-of-cycle recirculation pump trip (RPT) were operable.

Analyses were also performed to determine MCPR operating limits with either of these systems inoperable.

This resulted in increased MCPR limits which are also proposed for SlC6.

These calculations follow standard procedures and operation within the proposed MCPR operating limits with either the main turbine bypass system inoperable or the end-of-cycle RPT inoperable is acceptable for S1C6.

Compliance with overpressurization criteria was demonstrated by analysis of the main steam isolation valve (MSIV) closure event assuming HSIV position switch scram failure, an MSIV closure time of 2.0 seconds and six safety-relief valves out-of-service.

Maximum vessel pressure was 1,312 psig, within the limit of 1,375 psig.

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

The LOCA analyses for the Susquehanna plants (Ref. 19) was performed for a full core of ANF 9x9 fuel and is applicable for the S1C6 residual and reload ANF fuel.

In addition, to support the increased burnup for the ANF 8x8 fuel, ANF performed an additional LOCA heatup calculation at 40,000 MWd/MTU.

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.

The control rod drop accident (CRDA) was analyzed with approved ANF methodology (Ref. 12).

The maximum fuel rod enthalpy was 205 cal/gm, which is well below the design limit of 280 cal/gm, and less than 600 fuel rods exceed 170 cal/gm, which is less than the 770 rods assumed in the Susquehanna FSAR analysis.

To ensure compliance with the CRDA analysis assumptions, control rod sequencing below 20 percent core thermal power must comply with GE's banked position withdrawal sequencing constraints (Ref. 20).

The staff concludes that the analysis and results for the S1C6 CRDA are acceptable.

2.6 Sin le Loo 0 eration SLO)

Current Technical Specifications for Susquehanna Unit 1 permit plant operation with a single recirculation loop out-of-service for an extended period of time.

Analyses for S1C6 (Ref. 4) show that the MCPR Safety Limit must be increased by 0.01 because of the increased measurement uncertainties.

The pump seizure event is more severe under SLO than under two-loop operation, assuming.pump seizure of the operating loop, Thi~ the limiting event over most of t1ie power and flow operating region for SLK~ KNF analyzed the pump seizure event on a generic basis~for the Susquehanna Units.

Calculations were done for several cycles of operation for tRe Susq~anna Units.

The calculated delta CPRs were used to determine a conservative bounding delta CPR.

This,

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This is incorporated as a

limit in the SLO TS.

This analysis used approved methods and the result is acceptable.

Previous analyses reported by the licensee (Refs.

21 and 22) have shown that other events which could be affected by SLO were non-limiting when analyzed under SLO conditions.

SLO for S1C6 must maintain the 80 percent recirculation pump speed restriction because of the previous GE vessel internal vibration

analysis, as discussed in Reference 21.

This requirement is already present in the Technical Specifications and is unchanged by this amendment.

3.0 TECHNICAL SPECIFICATION CHANGES The following Technical Specification (TS) changes have been proposed for operation of SlC6.

(1)

TS 3/4.2.1 -- Figure 3.2.1-1 is changed to reflect the approved burnup extension of ANF 8x8 fuel to 37,000 MWd/MTU for average bundle exposure, which was previously discussed.

This is acceptable.

(2)

TS 3/4.2.3 Figures 3.2.3-1 and -2 are changed to reflect the new calculations of flow and power dependent MCPR operating limits using the parameters of S1C6.

As previously discussed, these analyses have been approved and the changes are acceptable.

(3)

TS 3/4.2.4 -- Figure 3.2.4-1 is changed to reflect the approved burnup extension of ANF 8x8 fuel to 42,000 MWd/MTU for average planar

exposure, which was previously discussed.

(The increase from 37,000 to 42,000 for average bundle versus average lanar reflects the axial exposure peaking factor.)

This change is accepta e.

(4)

TS 3/4.4.1 -- Figure 3.4.1.1.1-1 is changed to reflect the calculated changes in the regional stability boundaries, as was previously discussed.

The change is acceptable.

(5)

TS 3/4.4.1 -- The MCPR operating limit for SLO is changed to 1.30.

As was previously discussed, the analysis and results for this change is acceptab le.

(6)

TS 5.3.1, -- This change removes references to fuel assembly types from the initial core loading which are no longer present.

It is acceptable.

(7)

TS 5.3.2 This change recognizes the presence of the replacement contyol..rod blades.

It is acceptable, In addition there are several administea~~~riptive changes to the Bases reflecting removal of errors or the reasons for the TS changes discussed above.

These include Bases 2.12, 3/4.2.1 and

.3 and 3/4.4.1.

These changes are acceptable.

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4.0 TECHNICAL CONCLUSIONS The staff has reviewed the reports submitted for the Cycle 6 operation of Susquehanna Unit 1 and concludes that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses are acceptable.

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

5.0 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 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 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).

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.

The Commission made a proposed determination that the arne'ndment involves no significant hazards consideration which was published in the Federal Re ister (55 FR 33992) on August 22, 1990 and consulted with the Commonwena t i o Pennsylvania.

No public comments were received, and the Commonwealth of Pennsylvania did 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 Commission s regulations, and issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor:

H. Richings Dated:

November 2, 1990

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REFERENCES Letter (PLA-3407) from H.

W. Keiser (PPImL) to W. R. Butler (NRC),

"Proposed Amendment 132 to License No. NPF-14:

Unit 1 Cycle 6 Reload,"

July 2, 1990.

2.

3.

4, 5.

6.

7.

8.

9.

10.

12.

13.

14.

PL-NF-90-003, "Susquehanna SES Unit 1-Cycle 6:

Reload Summary,"

June 1990.

ANF-90-049, "Susquehanna Unit Cycle 6:

Plant Transient Analysis," May 1990.

ANF-90-050, "Susquehanna Unit 1 Cycle 6 Reload Analysis:

Design and Safety Analyses,"

May 1990.

XN-NF-85-67(P)(A), Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

Exxon Nuclear Company, Inc., September 1986.

Letter from G.

C. Lainas (NRC) to G.

N. Ward (ENC), "Acceptance for.

Referencing of Licensing Topical Report XN-NF-85-67(P), Revision 1,

'Generic Mechanical Design Report for Exxon Nuclear Jet Pump BWR Reload Fuel,'" July 23, 1986.

XN-NF-82-06(P)(A), Supplement 1, Revision 2, "Qualification of Exxon Nuclear Fuel for Extended Burnup - Supplement 1 Extended Burnup Qualification of ENC 9x9 Fuel,"

May 1988.

Letter from G. c. Lainas (NRC) to G. N. Ward (ANF), "Acceptance for Referencing of Licensing Topical Report XN-NF-85-74(P)," June 24, 1986.

ANF-90-018(P), Revision 1, "Susquehanna Unit 1 8x8 Extended Burnup Design Report,:

June 1990.

NEDE-22290-A, Supplement 1, "Safety Evaluation of the General Electric Hybrid I Control Rod Assembly for the BWR 4/5 C Lattice," July 1985.

NEDE-22290-A, Supplement 3, "Safety Evaluation of the General Electric Duralife 230 Control Rod Assembly,"

May 1988.

XN-NF-80-19(A), Volume 1 and Volume 1 Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors:

Neutronic Methods for Design and Analysis," March 1983.

PL-NF-87-001-A, "Qualification of Steady State Core Physics Methods for BWR Design and Analysis," April 28, 1989.

XN-NF-80-19(P) (A), Volume 4, Revjsion=

"Ex uclear Hethodology for Boiling Water Reactors:

Application of Ne C Methodology to BWR Reloads,"

Exxon Nuclear Company, Inc., June 1986.

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15.

Letter (RAC:058:88) from R. A. Copeland (ANF) to N.

W. Hodges (NRC),

"Void History Correlation,"

September 13, 1988.

16.

Letter from G. C. Lainas (NRC) to G.

N. Ward (ANF), "Acceptance for Referencing of Licensing Topical Report XN-NF-85-74(P),

'RODEX2A (BWR)

Fuel Rod Thermal-tiechanical Evaluation,'"

June 24, 1986.

17.

NRCB-88-07, Supplement 1,

"Power Oscillations in Boiling Mater Reactors,"

USNRC Bulletin, December 30, 1988.

18.

XN-NF-84-105(A), Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis,"

February 1987.

19.

XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis NAPLHGR Results for 9x9 Fuel,"

May 1986.

20.

NED0-21231, "Banked Position Mithdrawal Sequence,"

General Electric Company, January 1977.

21.

Letter (PLA-2885) from PAL to NRC, "Proposed Amendment 52 to License No.

NPF-22," June 30, 1987.

22.

Letter (PLA-2935) from PPKL to NRC, "Additional Information on Proposed Amendment 52 to License No. NPF-22," October 30, 1987.

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