ML20215M023

Safety Evaluation Supporting Amends 73 & 54 to Licenses NPF-9 & NPF-17,respectively

ML20215M023
Person / Time
Site:	McGuire, Mcguire
Issue date:	06/22/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20215M020	List: ML20215M018 ML20215M023
References
NUDOCS 8706260251
Download: ML20215M023 (11)

Text

l 1

UtseQ

~L

-k UNITED STATES

)

+

' 8-NUCLEAR REGULATORY COMMISSION 1

r I

E WASHINGTON, D. C. 20555

[$,

7 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULLTION

!a RELATED TO AMENDMENT NO. 73 TO FACILITY OPERATING LICENSEiNPF-9 l

1

' AND AMENDMENT NO. 54 TO FACILITY OPERATING LICENSE NPF-17 DUKE POWER COMPANY 1

l DOCKET NOS. 50-369 AND 50-370 McGUIRE NUCLEAR STATION, UNITS l'AND 2 INTRODUCTION By letter dated April 9,1987..and supplements dated May 18 and June 15, 1987, (Ref.1), Duke Power' Company (the licensee) proposed amendments to change the Technical Specifications-(TSs) to reflect.the third refueling of McGuire Unit

'l 2 and its fourth fuel cycle.

(The refueling for Unit 2 Cycle 4 would continu'e the transition to use of optimized fuel assemblies (OFAs) initiated'during the first refueling'and would replace an additional 64 standard fuel assemblies with OFAs; thus,181 of the total 193 fuel assemblies in Cycle-4 would be OFAs.) The. existing TS figures for axial flux difference limits as a function of rated thermal power would be relabeled such that the existing figure for l

Unit I only. (Figure 3.2-la) would apply to both Units 1 and 2, and the existing.

figure for Unit 2 only (Figure 3.2-1b) would be deleted. The TS Index would be updated consistent with these changes.

4 The proposed amendments would also increase the limit specified for heat flux hot channel factor (F ) for both Unit I and Unit 2 from the present value of g

2.26 to 2.32. This change would be reflected in each of several TSs presently l

specifying 2.26, including TSs 3.2.2, 4.2.2.2c, 4.2.2.2f, 4.2.2.3, 4.2.2.4c, 4.2.2.4f.2, and Bases 3/4.2.1. A corresponding change would be made to TS i

Figure 3.2-2 which shows nonna11 zed F as a function of core height (i.e.,

9 the revised normalized figure would be based upon a total F of 2.32 rather 1

g

.than2.26.)

i The title of TS 6.9.1.9, " Radial Peaking Factor Limit Report" would be changed to " Peaking Factor Limit Report." This change would also be reflected in the h$0NDo$ $

f P

l r

'TS'Index. The schedule in TS 6.9.1.9 for providing the peaking factor limit

. report to the NRC would be changed from 60 days before~ cycle init al criticality _

d (or 60 days before W(Z) functions and the value for APLND wouldbdcome

[

effective) to 30 days after implementation. The change would' also update the j

NRC addressee specified in TS 6.9.1.9 for receipt of the peaking _ factor limit l

report (i.e., the NRC's Core. Perfonnance Branch would be changed to the NRC l

Document Control Desk, with copies also specified to be provided to the Regional Administrator and the Resident ' Inspector) based upon changes in the Comission's' regulations (51 FR 40303). TS 6.9.1.9 would also be supplemented to specify that the methodology used to generate the W(Z) functions for Relaxed Axial Offset Control '(RAOC) and. base load operation and the value for APLNU shall be those previously reviewed and approved by the NRC (i.e., from WCAP-10216 l

" Relaxation of Constant Axial Offset Control - F Surveillance Technical q

Specifications".)

If changes to these methods are deemed necessary, the revised TS 6.9.1.9 would specify that such changes are to be evaluated in j

accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to.their use if the change is determined to involve an unreviewed safety.

question or if such a change would require amendments of previously submitted documentation.

The licensee's letter of April 9.1987 provided a description and Reload Safety Evaluation '(RSE) for the McGuire Unit 2-Cycle 4 core reload and the associated PeakingFactorLimitReport(PFLR). By letter dated May 18, 1987 the licensee corrected certain references cited in the-April 9,1987 submittal to clarify that appropriate methodology had been used for the RSE. By Letter dated June 15, 1987, the licensee noted that two fuel assemblies had received damage and would not be reused during Unit 2 Cycle 4 operation.

Consequently, the core loading pattern was revised to exclude these two assemblies and a revised RSE and. PFLR was provided to the NRC. The revised pattern did not change the results of the licensee's safety evaluation, the conclusions of the April 9 RSE, or the proposed revisions to the TSs.

s EVALUATION 1.

Large Break LOCA Analysis and Increased Fn The licensee's RSE included LOCA analyses to justify the increase in F from j

O i

i

a p

,o

' 2.26 to 2.32 because the limiting event which detemines the allavable value of F is the large-break LOCA.

g I

During a large break LOCA, depressurization of the Reactor Coolant System (RCS) would result in a reactor trip and safety injection signals. Although the injection of borated water from the ECCS would complement void fomation to shutdown the fission process, the presence of boron is not accounted for in j

this aspect of the LOCA analysis. Similarly, no credit is taken for control rod insertion, leaving void formation as the credible mechanism to teminate the fission process in the early phase of~ the transient.

Injection of the borated water provides for. heat transfer from the core and prevents excessive clad temperature. Once the RCS depressurizes to about 600 psia, the accumulators would begin injecting borated water. The analysis assumes loss of offsite power; hence, reactor coolant pumps are assumed to trip and to coast down.

After the depressurization (blowdown) phase of the transient, refill of the reactor vessel begins with esagency core cooling water which was not assumed to be operational until this time. The refill phase is completed when the water level reaches the bottom of the fuel rods. The next phase, reflood, occurs as ECCS water covers the core and terminates the core temperature rise.

Continued operation of the ECCS pumps supplies water for long tem cooling.

The boric acid concentration in the primary water is sufficient to prevent criticality.

The analysis was performed with an NRC approved code, the 1981 and 1983 versions of the Westinghouse LOCA-ECCS evaluation model BART (Refs. 3 and 4). This code included the evaluation model revisions approved by the NRC (Refs. 2, 5 and

6) regarding (a) a modelling change in WREFLOOD which was found to increase the peak cladding temperature by about 20'F and (b) a systematic input error in the BART code which caused low values of hot assembly bundle power to be used, and which was found to increase peak cladding temperature by about 100*F.

The blowdown, refill and reflood stages of the transient were analyzed with the methodologies described in Reference 7.

Reference 7 also describes the interfaces among the various computer codes and the features of the codes that ensure compliance with 10 CFR 50, Appendix K.

The various computer codes involved in the analysis are:

p.

,_..o '

• ~ SATAN-VI: - Analyzes the themal hydraulic transient in the ttactor coolant system during blowdown.

It includes RCS pressure, ent !alpy, density and mass and energy flow (Ref. 8).

l LOTIC: - Calculates the containment pressure transient during the three i

phasesoftheLOCAanalysis(Ref.9).

WREFLOODi.- Detemines the core flooding rate, the coolant pressure 'and

. temperature and the quench front height during the refill and reficod

. phases of the LOCA.(Ref. 10).

(See also BASH.)

LOCTA-IV: - Computes the thermal transient of the hottest fuel rod during the three phases (Ref. 11).

. BASH: - Provides an analysis of the steam / water flow phenomena during core reflood -(Ref.13).

~

The computational model and codes used for this analysis have been approved' by the NRC and comply with the requirements of Appendix K to 10 CFR 50.

p I:

The initial conditions and the numerical values of the input parameters for the analysis were conservatively determined by the licensee. Because of the substantial design similarity of the Unit 1 and Unit 2 Cycle 4 cores and RCS, the analysis applies to both Unit 1 and Unit 2.

The double ended, cold leg, guillotine break was shown to be the limiting case (Ref. 12). Some of the main parameters and initial conditions for this limiting case include:

Core power - 102% of 3411 MWt Peak linear power - 102% of 12.88 KW/ft j

. Heat-flux hot-channel factor (F ) - 2.32 l

g 3

Accumulator water volume (nominal) - 950 ft, each Moody discharge coefficients (C ) - 0.4, 0.6, 0.8 l

D Steam generator tube plugging - 5.0% each

L

~

o

-S-

~

The criteria to be satisfied in a large break LOCA analysis are d scribed in 10 CFR 50.46. The criteria are: (1) peak cladding temperature sh 11 not exceed 2,200'F; (2) localized maximum cladding oxidation must not exceed (17% during or after quenching; (3) cladding ~ chemical interaction with water and steam (maximum

- hydrogen generation)..must not exceed 1.0% of all the metal. in the cladding I

cylinders surrounding the fuel; (4) calculated changes in core. geometry shall be such that the core remains amenable to cooling; and (5) after the successful initial operation of the ECCS the calculated core temperature shall be maintained at an acceptably low value and decay heat. shall be removed for the extended-period of time required by the long lived radioactivity remaining in the core.

I i

The results of the licensee's analysis showed that all the applicable criteria are satisfied. Specifict11y,

' 1.

Peak' cladding temperature was 1,841*F for the worst case (CD=0.6). This is less than 2200*F and is, therefore, acceptable.

2.

Local maximum cladding oxidation during or after quenching was 2.71%. This is'less than 17% and is', therefore, acceptable.

j 3.

Cladding chemical interaction of all the metal in the cladding cylinders surrounding the fuel was less than 0.30%. This does not exceed 1.0% and is, l

therefore, acceptable.

4.

The cladding temperature transient was terminated at a time when the core geometry was still amenable to cooling.

5.

After the successful initial operation of the ECCS, the calculated core

)

temperature remains at an acceptably low level and the decay heat is removed for an extended period of time.

On the basis of the above results for the large break LOCA perfonned with an F of 2.32, we find the proposed increase in F for Units 1 and 2 to be g

q acceptable because the revised analyses, perfonned with suitable input parameters and based upon methodology which satisfies the criteria of Appendix K to 10 CFR 50, provide results which meet the requirements of 10 CFR 50.46.

o

, e..'

2.

Unit 2/ Cycle 4 Reload Safety Evaluation

_o.9

.On April' 20, 1984, the Comission issued Amendment No. 32 to Facility Operating c

License NPF-9 to change the Technical Specifications to permit changes in oper-ating limits related to the transition to the use of optimized fuel assemblies (OFA) in McGuire Unit 1.

Comission approval of the transition from the i

standard fuel assembly to the OFA loading was based upon the licensee's safety analyses (Ref.14and15)'whichexaminedthemechanical, nuclear, thermal hydraulic and accident evaluation aspects and justified the compatibility of the OFA design with the standard design in the transition loadings as well as the~ full 0FA core. A similar amendment for Unit 2 (Amendment 23) was issued March 22, 1985.

Accordingly, beginning with their first refuelings for Cycle 2 Unit I and Unit 2 operated with the first stage of a transition core consisting of approximately 1/3 Westinghouse 17x17 0FAs and 2/3 Westinghouse 17x17 low-parasitic. fuel assemblies (STDs). ' During Cycle 3, each unit contained about 2/3 0FAs'and 1/3 STDs. Unit l'is currently operating in Cycle 4 and Unit 2 will achieve' Cycle 4

4 by its present refueling.

In Cycle 4, 181 of the total 193 fuel assemblies q

of the Unit 2 core will be OFAs.

I l

The major differences between STDs and 0FAs are the use of 21rcaloy grids for l'

the.0FAs versus Inconel grids for STDs and reduction in fuel rod diameter. The l

OFA fuel has similar design features compared to the STD fuel, which has had L

substantial operating experience in a number of nuclear plants. Major advantages for utilizing the OFAs are: (1) Increased efficiency of the core by reducing the amount'of parasitic material and (2) Reduced fuel cycle costs due to an optimization of water to uranium ratio.

The proposed amendments provide for plant operation consistent with the design and safety evaluation conclusions in the licensee's McGuire Unit 2 Cycle 4 Reload Safety Evaluation (RSE). The changes to the Technical Specifications reflect adjustments in the limiting conditions and surveillance requirements for (1) t' axial.fluxdifferenceand(2)heatfluxhotchannelfactor,consistentwiththe parameters used in the RSE. A sunrnary of the cycle specific aspects of the p

L

~

+-

g....

nuclear, themal hydraulic, and mechanical-analysis, and the NRRfindings.

follow:

l The Cycle 4 reload has been designed to meet ECCS limits based upon an F of g

2.32 and Relaxed Axial Offset Control (RA00) has been employed to the extent pemitted by the new F value (Ref.1 and 16). The Cycle 4 nuclear character-g istics are within the range of, and are bounded by, the Cycle 3 nuclear 1

characteristics (Ref.1. Attachment 2A).. The significant core parameter (e.g.,

enrichment, fuel density, fuel burnup, moderator temperature coefficient, x

Doppler temperature coefficient, minimum delayed neutron fraction, maximum

-bank differential worth, control rod (and worst stuck rod) worths and-shutdown re therefo cep ab The thermal-hydraulic methodology, the DNBR correlation, and the Cycle 4 DNB limits 'are consistent with the current and accepted licensing basis. Operating power distributions were evaluated relative to the assumed limiting operating power distribution used in the accident analyses. Limits on allowable operating axial flux difference as a function of power level, were found to be less restrictive than those resulting from the LOCA F considerations previously g

discussed.

No changes in the DNB core limits are required. No variation in the thermal margin will result from the Cycle 4 reload. Hence, the Unit 2

. Cycle 4 thermal-hydraulic design is acceptable.

l The mechanical design of Unit 2 Cycle 4 is within the limits of Cycle 3 design and the compatibility of the OFA core has been justified in the OFA loading submittal (Ref. 14). The fuel has been designed and will be operated so that

. clad flattening will not occur (Ref. I and 17).

In that portion of the core l

- pattern designated as " region 6," a rod plenum spring smaller than that of i

previous fuel regions is used. This new spring design (Ref. 18) satisfies i

a change in the non-operational 6g loading design criterion to 4g axial and 6g lateral loading with dimensional stability. This reduced spring force reduces the potential for pellet chipping in the fuel rod. We find that the mechanical fuel design for Unit 2 Cycle 4 is within the limits of previously accepted designs and is, therefore, acceptable.

I i

.. f

,e.

+

- Accordingly, we conclude that the Unit 2 Cycle 4 design does not_cause the previously acceptable safety limits to be exceeded and is, therefgre, 4

acceptable.

I 1

3.

Administrative Changes By previous Amendments 32-(Unit.1)/13(Unit 2) and Amendments 42 (Unit 1)/23 (Unit 2), McGuire was changed to a type of F function for which the title

~

g

" Radial Peaking Factor Limit" was no longer appropriate. The previous amend-ments failed to correct the title of TS 6.9.1.9.

The present amendments correct the title by. deleting " Radial." Also, during its licensing review of another nuclear plant (Vogtle Electric Generating Station), the Comission

.i detemined that the. safety of a plant would not be affected if the peaking factor limit report required by TS 6.9.1.9 were submitted 30 days after imple-mentation rather than 60 days before criticality, provided the methodology used was'previously reviewed and approved by the NRC and changes to this methodology are subject to the requirements of 10 CFR 50.59. The change in the'McGuire schedule. implemented by these amendments includes these conditions in the' revised TS 6.9.1.9.

The amendments also update the NRC addressee for receipt of the report consistent with 51 FR 40303.

The TS Index is-updated consistent with appropriate changes implemented by i

these amendments.

The above changes are purely administrative and have no adverse impact upon safety. They are, therefore, acceptable.

REFERENCES I

I 1.

Letter from H. B. Tucker Duke Power Company to NRC, dated April 9,1987, with supplements dated May 16, 1987 and June 15, 1987.

I 2.

Letter from H. R. Denton to PWR Applicants and Licensees, " Westinghouse ECCSEvaluationModels"(GenericLetter86-16)datedOctober 22, 1986.

~.

~

y a, a

' 3.

Letter from J. R. Miller, NRC,.to E. P. Rahe, Westinghouse, " Acceptance forReferencingofthe1981VersionoftheWestinghouseLargi lBreakECCS Evaluation. Model," December 1,1981.

I-

\\

4.

Letter from C. O. Thomas,-NRC, to E. P. Rahe, Westinghouse, " Acceptance j

for Referencing of Licensing Topical Report WCAP-9561, BART A-1: A l

Computer Code for Best Estimate' Analyses of Reflood Transients," dated December 21, 1983.

5.

Letter from Charles. E. Rossi,' NRC to E. P. Rahe, Westinghouse, " Acceptance for Referencing of Licensing Topical Report WCAP-9561, Addendum 3 Revision 1," dated August 25, 1986.

6;. Young, M. Y., " Addendum to BART-A1: A Computer Code for the Best Estimate Analysis of Reload Transients"-(Special Report: Thimble modeling in f

Westinghouse.ECCS Evaluation Model): WCAP-9561-P Addendum 3, Revision 1, i

dated July 1986.

7.

Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model-Sumary,"

WCAP-8339, dated July 1974.

8.

Bordelon, F. M., et al., " SATAN-VI Program:

Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8306, dated June 1974, i

9.

Hsieh, T., et al., "Long Tenn Ice Condenser Containment LOTIC Code Supplement 1" WCAP-8355 Supplement 1 May 1975, WCAP-8354P, July 1974.

10.

Kelly, R. D., et al., " Calculational Model for Core Reflooding After a Loss-of-CoolantAccident(WREFLOOD),"WCAP-8171,(WCAP-8170P) dated

' June 1974.

l

11. Bordelon, F.

M., et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, dated June 1974,

12. Salvatori, R., " Westinghouse ECCS Sensitivity Studies," (WCAP-8340P)

WCAP-8356, dated July 1974.

+

~

.i on a

- 10

13. Letter fom W. ~J. Johnson, Westinghouse to J. Lyons, NRC, " BASH Methodology. Enhancements,"WCAP-10266, Addendum 2,datedMarcf1987, No. NS-NRC-87-3212.

j

14. Duke-Power Company submittal to NRC, " Safety Evaluation for McGuire Units 1 and 2, Transition to Westinghouse 17x17 Optimized Fuel Assemblies" dated December 1983.

15. Letter from H; B. Tucker, Nxe Powar Company to H.' R. Denton, NRR, "RTD Bypass Manifold Remm/M," dated October 29, 1985.

1

16. Miller, R. W., et al., " Relaxation of Constant Axial Offset Control-Fg Surveillance Technical. Specifications WCAP-10217-A, dated June 1983.

17. Georgie, R. A., et al., " Revised Clad Flattening Model" WCAP-8381, dated July 1974.

18. Letter from E. P. Rahe, Jr., Westinghouse to L. E. Phillips, NRC, " Fuel I

Handling Load Criteria," dated April 12,1984(NS-EPR-2893).

ENVIRONMENTAL CONSIDERATION i

i These-amendments involve changes to the installation or use of facility com-

)

ponents located within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements. The staff has detemined that the amendments involve 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

' j exposure. The NRC staff has made a determination that the amendments involve no significant hazards consideration, and there has been no public comment on

~ such finding. Accordingly, the amendments meet the eligibility criteria for j

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 these amendments.

,e

,..m...

-. CONCLUSION 7

The Comission made a proposed determination that the amendments involve no significant hazards consideration which was published in the Federal Register

-(52 FR 18977) on May 20,1987 and consulted with the state of North Carolina.

No public coments were received, and the state of North Carolina did not have any coments. Licensee submittals since publication of 52 FR 18977, dated May 18 and June 15, 1987, correct certain references in the initial submittal, reconfim that the initial safety limits are met for minor changes in the fuel loading pattern, and do not alter the proposed changes as identified in 52 FR 18977 or alter the staff.'s proposed no significant hazards consideration detennination.

.We have concluded, based on the' considerations discussed above, that:

(1)

I 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 Comission's regulations, and the issuance of these amendments will not be inimical to the comon defense and security or to the health and safety of the public.

Principal Contributors:

L. Lois, RSXB D. Hood, PD#II-3

)

Dated: June 22, 1987 k

l I

.