ML20086K752

Application for Amend to License DPR-53,allowing Operation for Eleventh Cycle (Second 24-month Cycle).Reload Analysis Rept Encl

ML20086K752
Person / Time
Site:	Calvert Cliffs
Issue date:	12/10/1991
From:	Creel G BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20086K756	List: ML20086K752 ML20086K760 ML20086K763
References
NUDOCS 9112130289
Download: ML20086K752 (10)

Text

.

_y 0 - 4 GALTIMORE GAS AND ELECTRIC 1650 CALVERT CLIFFS PARKWAY = LUSBY, MARYLAND 20657-4702 Gronot C. Cntit vier PResiocht NOCLC AR cNEROY (480)#60 4468 Decernber 10,1991 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No.1; Docket No. 50--317 Request for Amendment Unit 1 Eleventh Cvele License Amendment Reauest

REFERENCES:

(a) Letter from Mr. G. C, Creel (BG&E) to NRC Document Control Desk, dated February 7,1989, Unit Two Ninth Cycle License Application (b) Letter from Mr. S. A. McNeil (NRC) to Mr. G. C. Creel (BG&E).

dated January 10,1990,1ssuance of License Amendtnent 123 (c) Letter from Mr. G. C. Creel to NRC Document Control Desk, dated February 7, 1989, VeriGcation of a On::-Pin Burnup Limit, CEN-382(8)-P (d) Letter from Mr. S. A. McNeil (NRC), to Mr. J. A. Tiernan (BG&E),

dated May 16, 1988, Safety Evaluation Report for the Tenth Cycle License Application l Gentlemen:

The Baltimore Gas and Electric Company (BG&E) hereby requests an Amendment to its Operating License No. DPR-53 for Calvert Cliffs Unit No. I to allow operation for an eleventh cycle (second 24-month cycle). The attached reload analysis report presents a detailed description of the requested Technical Specification changes with supporting safety analysis information to ensure conservative operation at a rated thermal power of 2700 MWth for Unit 1 Cycle 11. Cycle 11 operation will commence following the refueling outage scheduled for spring 1992.

The attached reload analysis report also presents detailed descriptions of the differences between the Unit 1 Cycle 11 reload core design and the Unit 2 Cycle 9 reload core design. These design changes are summarized below. The Unit 2 Cycle 9 design is the reference design for Unit 1 Cycle 11. The Unit 2 Cycle 9 Request for License Amendment (Reference a) was approved by the NRC in Reference (b).

91121302E4 711210

DR ADOCi' 0300 17 ,

6 Mhr f i

[l

~~~~~" V o s \

Document Control Desk December 10,1991 Page 2 Specific differences between this reload analvsis and the reference analysis have been accounted for I in the Unit 1 Cycle 11 analyses. In all cases,it has been concluded that either the reference cyde analyses envelope the new conditions or the revised analyses continue to show acceptable results.

Technical Specification changes have been proposed where needed because of variations tetween the current cycle (Unit i Oycle 10) and the new cycle. A large proportion of these proposed changes are identical to thost. nproved for the reference cycle.

The Unit 1 Cycle 11 analyacs used the same methodology as the reference cycle with one significant exception Improm j pics methods have been implemented, including being able to account for ansotropic scat.a. = r - D u ar d the application of the Nodal Expansion hiethod (NEht) for solving fer the coarsemeso fia -

The nerformmer af Camistion Engineering (CE) 14x14 fuel at extended burnup has been appro'ved 10. 1-p'c burnups up to 52,(XX) htWD/r. De performance of CE .4x14 fuel at Calvert Cliffs ' - burnup >cyond those con idered was discussed in the Supplement to the Referenec (c).

That Supplement supports operation with a 1-pin burnup of 60,(XX) h1WD/T, a value considerably in excess of the maximum 1-pin burnup projected for Cycle 11 (approximatdy 56,900 h1WDff).

However, this Supplement has not yet been approved by the NRC Lerefore, operation of Cyc!c 11 with a 1-pin burnup greater than 52,000 htWDfr is contingent upon tae approval of the Supplement by the NRC.

UNIT 1 CYCL.E 11 FUEL,1)ESIGN The Cycle 11 core consists of 217 fuel assemblics. Eighty-four fresh (unirradiated) Batch N assemblies will replace previously irradiated assemblics. The 84 fresh assemblies consist of 12 unshimmed Batch N assemblies,20 4-shimmed (B C) Batch NX assemblies, and 52 8-shimmed (B. C)

Batch N/ assemblies. Of the fuel batches in Cycle 11, the highest U-235 enrichment occurs in the Batch N fuel assemblics which contein an assembly average enrichment of 4.20 wt9c U-235. In addition to the 84 fresh assemblies, the core will consist of 132 presently operating Batch I. and h1 assemblies and one reinserted Batch K* assembly which was discharged at the end of Unit 1 Cycle 9.

Four of the irradiated Batch h1 assemblies are the Advanced Nuclear Fuel (ANF) demonstration assemblies containing Gadolinium-bearing fuel pins which were installed in Cycle 10 to qualify ANF fuel for 24-month cycle operation. Use of these assemblies was approved in Reference (d).

The mechanical design of each assembly in the Batch N reload fuel is similar to the Batch L fuel previously inserted in Calvert Cliffs Unit 2, with the two following exceptions:

(1) The Batch N assemblies use the new GUARDIAN design debris resistant feature. This entails new grid and fuel pin designs in place of the small flow hole debris resistant feature in the Unit 2 Cycle 9 Batch L fuel.

The GUARDIAN design uses a redesigned Inconel spacer grid assembly that improves the lower grid assembly's capability to entrap debris. The rods have been redesigned with long, solid Zircaloy-4 end caps to absorb any wear induced by the debris trapped within the spacer grid assembly. The rods are secured by a detent spring that holds the rods in place but allows reconstitution. if necessary. Specific design features are described in the attached reload ar,alysis report.

I l

l

_- - _y o a Document Control Desk December 10,1991 Page 3 (2) The Zircaloy spacer grids used for Batch N are larger than those used previously. They were redesigned to allow the fuel rods located along the periphery of the fuel bundle to receive more coolant flow when in contact with adjacent bundles. This was scomplished by increasing the size of the outer pin cell by making the outside envelope of the spacer grid assembly larger.

All fuel to be loaded in Cycle 11 was 7 viewed to ensure that adequate shoulder gap clearance exists.

The review took into consideration the mechanical fuel design changes due to the use of the GUARDIAN debris resistant feature. All clearances were found to be acceptable for Cycle 11 operation.

During the spring 1992 refueling outage the center CEA will be replaced with a newly-designed CEA. This new design will correct the swelling problem experienced during Unit 1 Cycle 10 oper:.iion by replacing the zircaloy slug at the bottom of the CEA with a stainless steci slug. The replacement of the zircaloy slug climinates the potential for hydriding of the zirconium which is believed to have caused the swelling. This new CEA is identicalin design to the replacement CEA loaded into Unit 2 for Cycle 9.

Neutron flux suppressors, called Guide Tube Flux Suppressors (GTFSs), will be installed in the fuel assemblies at selected core k> cations to help reduce the Duence at the critical vessel weld. The basic design of the GTFSs is identical to that of control rod fingers. They are seated at the bottom erd fitting of the CEA guide tubes and are held in place by a spring-loaded mechanism. The upper ends of the GTFSs are aligned by the fuel alignment plate.

The thermal performance of the fuel in Cycle 11 was evaluated using the FATES 3B fuct evaluation model, in conjunction with the maximum pressure methodology (No-Clad-Lift-Off). A composite standard uranium dioxide fuel pin was analyzed that envelopes the various assemblies in Cycle 11.

The analysis modeled the power and burnup levels representative of the peak pin at each burnup interval. It also included the reduction in internal pin volume in the Batch N fuel due to the introduction of the GUARDIAN design. The maximum fuel pin internal pressure remained below the No-Clad-Lift-Off critical pressure.

UNIT 1 CYCI.E 11 NUCLEAR DESIGN The Cycle 11 core will use a low-Guence pattern along with GTFSs. This arrangemer.t of fuel and GTFS results in a very low fluence to the critical pressure vessel weld. As a result of this shift in power and Quence away from the periphery, there are slightly higher power levels in the interior of the cor.-

Hot full power (HFP) f uel assembly relative power densities are provided in the attached reload analysis report for beginning-of-cycle (BOC), middle-of-cycle (MOC), and end-of-cycle (EOC) unrodded configurations. Radial power distributions at BOC and EOC are also provided with control element assembly (CEA) Bank 5, the lead regulating bank, fully inserted. These distributions assume the high burnup end of the Cycle 10 shutdown window and this assumption increases the radial power peaking in the Cycle 11 core. The distributions were calculated with approved methods.

In addition, the safety and setpoint analyses for the DNB and linear heat rate conservatively include l uncertainties and other allowances so that the power peaking values used are higher than those expected to occur at any time in Cycle 11.

l

Y~

~

j s e Document Control Desk December 10,1991 Page 4 The bloderator Temperature Coefficient (h1TC) for llot Full Power (liFP) equilibrium conditions is predicted to be -0.04x104 delta rho?F at BOC. The NRC has previously expressed concern about positive h1TC cffects on the generic antidpated transients without scram (AT\VS) assumptions and BG&E has stated that it will design reload cores with h1TCs less than 0.0 until the generic AiWS implications are adequately addressed in the future. The Unit 1 Cycle 11 reload core design continues to meet that commitment.

Control Element Assembly worths and shutdown margin requirements for Unit 1 Cycle 11 are most limiting at EOC. The assessed shutdown c argin requirements for Cycle 11, presented in the attached reload analysis report, are based on the results -f the EOC, hot zero power (liZP), steam line break event. After consideration of all reactivity uncertainties and biases, a worst case assessment for Cycle 11 results in a shutdown margin at EOC of 4.5% delta rho. A reanalysis of this event concluded that sufficient shutdown margin is available to accommodate the reactivity effects of the steam line break event, at the worst time in core life, while allowing for the most reactive CEA stuck in the fully withdrawn position.

UNIT I CYCLEli TIIER%1AL-IIYDRAtILIC DESIGN Steady state Departure from Nucleate Boiling Ratio (DNBR) analyses of Cycle 11 at the rated power level of 2700 htWth have been performed using approved core thermal-hydraulic codes. The cycle specific models used for designing Cycle 11 account for the flow reduction caused by the GUARDIAN design. Ilot channel factors and calculational factors were combined statistically with other uncertainty factors using the approved Extended Statistical Combination of Uncertainties (ESCU) methodology and it was determined that the Cycle 11 core will operate within the DNBR limit of 1.15.

The effects of fuel rod bowing on the DNB margin for Cycle 11 have been evaluated using approved methods. This effect has also been considered for higher burnup rods (greater than 45,0lK) h1WDfr) where rod bowing increases. All Cycle 11 rod burnups are bounded by the fuel rod bowing evaluation presented.

UNIT 1 CYCLE 11 S AFETY ANALYSES EVAL UATION All the non-l.OCA transient safety analyses for Unit 1 Cycle 11 are bounded by previously presented and approved analyses. All key transient input parameters of the Cycle 11 non.LOCA analyses are equal to or conservative with respect to the reference cycle values (Unit 2 Cycle 9), with one exception. The shutdown margin at the end of cycle decreased from 5.0% delta rho to 4.5% delta rba due to the low fluence fuel loading. This change in shutdown margin impacts the Steam Line Rupture event. In addition to key input parameters, the implementation of the GUARDI AN design was also considered. The results of the reanalysis are boended by the reference cycle.

l An ECCS performance analysis (large and small break LOCA) was done for Unit 1 Cycle 11 to demonstrate compliance with 10 CFR 50.46. In addition to the normal differences in fuel related parameters, the differences between Cycle 11 and the reference cycle were: (1) use of the i GUARDIAN design for the Batch N fuel assemblics, and (2) the assumption of 500 plugged steam generator tubes (small break LOCA analysi., only). These two changes required reanalysis of the hydraulic portion of both analyses. The results of the reanalysis demonstrate that operation of Unit I Cycle 11, as designed, is in conformance with the acceptance criteria of 10 CFR 50.46.

> e Document Control Desk December 10,1991 Page 5 TECilNICAl, SPECll'ICATIONS i

The following paragraphs summarize the proposed changes to the Technical Specification requested to sulyort operation of Unit 1 Cycle 11. All proposed changes are requested for the Unit i TecM d Specifications and are presented in detailin the attached reload safety analysis report. No bait 2 i'echnical Specifications changes are requested.

1. Figure 2.2-1 is modified on the negative Axial Shape Index (ASI) side to accommodate the increased core average linear heat genert. tion rate (CAlllGR) of Unit 1 Cycle 11. The CAlllGR is increased for Cycle 11 because of the increased number of B.iC shims. This modiGeation was considered in the Unit 1 Cycle 11 reload safety analyses.

2.

The text q,3.1.3.1 and Figure 3.1-3 age modified to incorporate an increase in maximum allowed Fr from 1.65 to 1.70. The Fr is increased to accommodate the increased neutron flux peaking associated with this 24-month cycle for Unit 1. The setppint analysis performed in support of Unit 1 Cycle 11 considers this proposed change in Fr . Also, this Technical Specification renects the use of the CECOR 3.3/BASSS computer codes which was assumed in the Unit 1 Cycle 11 setpoint analysis.

3.

Figure 3.2-3b is modified to indpate a reduction in its accentable value region due to a reduction in the 100G power Fx . value from 1.54 to 1.50. The reduction in the limits of this Ogure are needed to accommod$te the increased core average linear heat generation rate of

, the Unit I reload core design to incorporate its increased number of B5C shims.

4. Implementation of the CECOR 3.3/BASSS computer codes as the on-line incore LCO monitoring system requires changes in Technical Specifications 3.2.2.1,4.2.2.1.2,4.2.2.1.3, 4.2.2.1.4, 4.2.2.2.2, 4.2.2.2.3, 4.2.2.2.4, 4.2.3.2, 4.2.3.3, 4.2.3.4 and B 3/4.1.3 to ensure they adequately renect the CECOR 3.3/BASSS system. A new Surveillance Requirement. 4.2.5.3, is added to accommodate the use of the CECOR 3.3/BASSS network.

Technical Specification 3.2.3 is modified to increase the F rT from 1.65 to 1.70 to 5.

accommodate the increased neutron Dux peaking associated with this 24-month cycle for Unit I and implementation of the CECOR 3.3/BASSS on-ling incore monitoring system.

The Unit 1 Cycle 11 setpoint analyses supports this change in Fr

6. Figure 3.2-3e is modified to accommodate :he increased FrT limit for the same reason as the proposed change to Technical Specification 3.2.3.

7. The text of 3.2.5 and Table 3.2-1 is modified by changing " core power" terminology to

" thermal power" to maintain consistency with other Technical Specifications.

8. The text of B 3/4.7.1.2 is modified by increasing the maximum allowed Auxiliary Feedwater Dow from 13tX) gpm to 1550 gpm. An evaluation of increasing this Oow was performed for Unit 2 Cycle 9 and it was determined that the results on the safety analyses for Unit 1 Cycle 11 are bounded by previously reported and approved analyses.

9 The text of 5.3.1 is modified to indicate an increase in the maximum enrichment for a reload core from 4.1 w/o to 4.35 w/o U-235. This change is proposed becaese higher enriched fuelis being used in Unit i Cycle 11 reload core. The Cycle 11 reload core design considers the proposed higher enrichment. This Technical Specification change must be approved prior to placing the higher enriched fuel in the reactor vessel.

D t Document Control Desk December 10,1991 Page 6

10. Figure 3.11b is modified to show the reduced shutdown margin available at the end of this cycle. This change is required to accommodate the reduced scram worth available due principa'ly to low-fluence fuel management used in this cycle. The safety analyses consider the limits of F;gure 3.1-1b and the results are acceptable.

11. Figure 3.2-1 is modified to climinate specific cycle t'mes. Cycle lengths have been increased beginning with the implementation of 24-month cycles for Unit 1 Cycle 10, causing the cycle times on this figure to be out of date, Because a constant Allowable Peak Linear lleat Rate is used throughout the cycle, cycle specific times are not required.

12. Replacement of the center CEA on Unit 1 climinates the need for the footnote on the following Technical Specifications: 4.1.1.1.1, 4.1.1.2, 3.1.3.1, 4.1.3.1.1, 4.1.3.1.2, 4.1.3.1.3, 3.1.3.3, 4.1.3.3.1, 4.1.3.3.2, 3.1.3.4, 4.1.3.4, 4.1.3.5, 3.1.3.6, 4.1.3.6, 3.10.1, 4.10.1.1, 4.10.1.2, 3.2.2.1, 4.2.1.3, 4.2.2.1.3, 4.2.2.2.3, 3.2.3 and 4.2.3.3. The text is modified to remove the exclusion concerning the center CEA.

In summary, the fuel system design, nuclear design, thermal-hydraulic design, and the transient and accident analysis information presented in this license submittal provide the necessary justification to support operation of Unit 1 in its eleventh cycle of operation This justification is supported since:

(1) previously reviewed and approved methods were used in the analyses; (2) the results of the safety analyses show that all safety criteria are met; and (3) the proposed Technical Specifications are consistent with the reload safety analyses.

DFTERMINATION OF NO SIGNIFICANT llA7.ARDS Baltimore Gas and Electric Company has determined, based on the analytical information supplied in the enclosed attachment, that this amendment does not involve a significant hazards consideration.

Justification for this determination is presented below and in the attachment to this request for license amendment.

This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:

1. invohe a significant increase in the probability or corsequences of an accident previously evaluated; or All the non-LOCA transient safety analyses for Unit 1 Cycle 11 are bounded by previous'y pesented and approved analyses. All key trcnsient input parameters of the Cycle 11 non-LOCA analyses are equal to or conservative with respect to the reference cycle values (Unit 2 Cycle 9), with one exception. The shutdown margin at the end of cycle decreased from 5.0% delta rho to 4.5% delta rho. This change impacts the Steam Line Rupture analysis. Even with this change incorporated, the results of the analysis are still bounded by the reference cycle.

6 i Document Control Desk December 10,1991 Page 7 i

An ECCS performance analysis (large and small break LOCA) was done for Unit 1 Cycle 11 to demonstrate compliance with 10 CI~R 50.46. In addition to the normal differences in core design parameters, the differences between Cycle 11 and the reference cycle were: (1) use of the GUARDIAN design for the Batch N fuel assemblics, and (2) the assumption of 500 plugged steam generator tubes (small break LOCA analysis only). These two changes required reanalysis of the hydraulic portion of both analyses.

Since the results of the Unit 1 Cycle 11 analyses are all conservatively bounded by the reference cycle, and due to the nature of the changes to the inputs to the safety analpes addressed above, the Unit 1 Cycle 11 core reload does not present a significant hazards consideration with respect to the existing safety analyses. The Cycle 11 reload does not involve an increase in the probability or consequences of an eccident previously evaluated.

2. create thepossibility of a new or different type of accidentfrom any accidentpreviously evaluated; or The proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. The design of Unit 1 Cycle 11 closely follows that of the reference cycle, Unit 2 Cycle 9. The mechanical design of each assembly in the Batch N reload fuel is identical to the Batch L fuel previously inserted in Calvert Cliffs Unit 2 with the following two exceptions:

(1) The Batch N assemblics use the new GUARDIAN design Jebris resistant feature. The GUARDIAN design uses a redesigned Inconel spacer grid assembly that improves the grid assembly's capability to entrap debris.

(2) The Zircaloy spacer grids used for Hatch N are larger than those used previously. They were redesigned to allow the fuel rods located along the periphery of the fuel bundle to receive more coolant flow when in contact with adjacent bundles. This was accomplished by increasing the size of the l outer pin cell by making the outside envelone of the spacer grid assembly larger.

The GUARDIAN design and the larger spacer grids used in the Batch N reload fuel l have been considered in all aspects of the nuclear, mechanical, thermal. hydraulic, and transient (LOCA and non-LOCA) safety analyses for Unit 1 Cycle 11. Each of these areas considered the impact of increased core differential pressure due to the introduction of the GUARDIAN design and the larger spacer grid. It was determined that a new accident type would not result from these changes. Reactor coolant system flow is maintained and individual assembly flow is not adversely affected. The impact of the flow both through the assemblics with the GUARDIAN design and the other standard fuel designs were analyzed to determine whether the presence of the more flow restrictive design causes an imbalance in the inlet flow to the other assemblies. It was determined that no significant impact or imbalance occurs for the Unit 1 Cycle 11 design.

e e Document Control Desk December 10,1991 Page 8 All the fuel to be loaded in Cycle 11 was reviewed to ascertain that adequate shoulder gap clearance exists. Analyses were performed with approved models and it iras concluded that al!;houlder gap and fuellength clearances are adequate for Cycle 11 operation.

Additionally, neutron flux suppressors, called Guide Tube Flux Suppressors (GTFSs),

will be . stalled in the fuel assemblies at selected locations. The basic design of the GTFSs is identical to that of control rod fingers. They are not moveable and serve no control rod function. They are provided to enhance the low fluence fuel management scheme for this cycle. The analyses performed have considered the effect of the GTFSs on fuel Lrformance.

I The installation of GTFSs does not degrade fuel performance and the results of the analyses remain bounded by the reference cycle. It was determined that a new accident type wouid not result fom these changes.

3. irwoh e a significant reduction in a margin ofsafety.

No margins of safety for the Unit 1 Cycle 11 reload core design arc reduced with respect to the previously reported and approved reference cycle. With each proposed Technical Specification change, sufficient conservatism or margin of safety remains between the proposed limits of the changes and actual safety limits (Specified Acceptable Fuel Design Limits - SAFDI2). In fact, the margin previously reported in the reference cyde is applicable to Unit 1 Cycle 11. Herefore, the Cycle 11 reload does not involve a significant reduction in the margin of safety.

SCllEDUI.E

- Unit 1 is scheduled to be shutdown on March 6,1992, to commence the Unit 1 Cycle 11 refueling outage. We currently plan to off-load the entire core to facilitate plant maintenance work. On April 21,1992, we will begin to load the Cycle 11 core into the reactor vessel The Unit is scheduled to enter Mode 4 on May 25,1992. Initial criticality is currently scheduled for my 27,1992, with full power operation to be achieved on May 31,1992. We request that these Technical Specification changes be approved by April 21,1992, to support fuel being loaded into the reactor vessel.

I i

.e <

Document Control Desk December 10,1991

. Page 9 BAFETY COMMITTEE REVIEW These proposed changes to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Off. Site Safety Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.

Very truly yours,

,/ )

/ \

[

,['

l i STATE OF MARYLAND : l

TO WIT : l COUNTY OF CALVERT l t,

I d,19$

I hereby certify that on the /h/1 day of / UA a /xA I a Notary Public of the State of Maryland in and for ' icJ v /M.bef6te flitu6 me, the subscriber, personally appeared George C. Creel, being duly sworn, and states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing information for the purposes therein set forth; that the statements made are true and

! correct to the best of his knowledge, information, and belief; and that he was authorized to provide

! the information on behalf of said Corporation.

l WITNESS my Hand and Notarial Seal: ^'$ '

u [ ' !,M i Netary Public l

i l l l

My Commission Expires: / / 'i9/ I M r cr: , a,_ ; ; , Date l

l GCC/LMDhjd/dtm '

i Attachments i

cc: D. A. Brune, Esquire J. E. Silberg, Esquire R. A. Capra, NRC l D. G. Mcdonald, Jr., NRC T. T. Martin, NRC L E. Nicholson. NRC R. I. McLean, DNR J. H. Walter, PSC

.