ML20058J169

Discusses Evaluation of Increased Fuel Burnup on Equipment Qualification at Facility

ML20058J169
Person / Time
Site:	Sequoyah
Issue date:	11/30/1993
From:	Labarge D Office of Nuclear Reactor Regulation
To:	Medford M Tennessee Valley Authority
References
NUDOCS 9312130437
Download: ML20058J169 (12)

Text

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- /yo uqh UNITED STATES

[ 1* 3 NUCLEAR REGULATORY COMMISSION 3:  :! WASHINGTON, D C. 20555-0001

. ,,, November 30, 1993 Docket Nos. 50-327 and 50-328 Tennessee Valley Authority ATTN: Dr. Mark 0. Medford, Vice President Technical Support 3B Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801

SUBJECT:

EVALUATION OF INCREASED FUEL BURNUP ON EQUIPMENT QUALIFICATION SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 The current source term used in the Sequoyah Nuclear Plant (SQN) environmental qualification (EQ) program is based on NUREG-0588 methodology and a TID-14844 release to determine the effects on equipment due to fuel burnup up to 650 effective full power days (EFPD). However, because high burnup fuel is installed that will extend burnup to 1000 EFPD, the Tennessee Valley Authority (TVA) prepared a Justification for Continued Operation (JCO) titled,

" Tennessee Valley Authority, Sequoyah Nuclear Plants Units 1 and 2, Increase in the 100 Day Integrated Dose To Equipment In Containment Associated with Increased Fuel Burnup, Justification for Continued Operation," dated September 4, 1991.

The JC0 assessed the effect on safety-related plant equipment located inside and outside the containment resulting from extending reactor core burnup from 650 EFPD to 1000 EFPD. If the radiation dose to this equipment increases beyond the value assumed in the EQ testing program, confidence in the ability of the equipment to function in the post-accident environment is reduced.

In the JCO, TVA used Regulatory Guide (RG) 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors" to determine the source term, instead of TID-14844 source terms given in Regulatory Guide 1.89, " Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants."

The staff has determined that this JC0 is not appropriate, because RG 1.77 identifies acceptable analytical methods and assumptions that may be used in evaluating the consequences of a control rod ejection accident, which is unrelated to EQ and should not be used even in a relatively short-term JCO.

RG 1.89 provides an acceptable source term for EQ in Regulatory Position C.2.(c). Furthermore, 10 CFR 49(e)(4) specifies that the radiation environment for EQ should be based on the most severe design basis accident during or following which the equipment is required to remain functional.

9312130437 931130

" ^"c" 5 s=7 NRC RE C, ENTER COPY 030C73 q cl' ,

t Dr. Mark 0. Medford  :

In a related matter, TVA has indicated the intention to delay final resolution of this issue until the staff issues NUREG-1465, " Accident Source Term for Light-Water Nuclear Power Plants," in its final form. The use of the revised source term in NUREG-1465 is under staff review and has not been approved for use at licensed nuclear plants. Since this approval process is not expected  ;

to be completed in the near future, we believe that TVA's decision to delay resolution of the issue until issuance of the NUREG is not appropriate.

As discussed with your staff during a conference call on October 20, 1993, the staff has concluded that short-term (i.e., JCO) resolution of this issue should not be based'on RG 1.77. We also indicated that resolution of this issue is not an immediate concern, since the present Unit 2 core had not yet reached 650 EFPD and the radiation environment that has been used for equipment qualification assessments typically contain certain safety margins that we feel will be adequate in the reassessment for the extended reactor core burnup at Sequoyah. It is our understanding that TVA is performing a reassessment of equipment qualification for 1000 EFPD burnup using an acceptable source term (TID-14844) that will be completed within a reasonable time period, and submitted for staff review.

Sincerely, David 0.IIONOgd,i9ff.d k/oject Manager Project Directorate 11-4 ,

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation ,

Enclosure:

Westinghouse JC0 cc w/ enclosure:

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Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT ATTN: Dr. Mark 0. Medford cc: l Mr. Craven Crowell, Chairman TVA Representative Tennessee Valley Authority Tennessee Valley Authority ET 12A 11921 Rockville Pike 400 West Summit Hill Drive Suite 402 i Knoxville, TN 37902 Rockville, MD 20852 Mr. W. H. Kennoy, Director General Counsel Tennessee Valley Authority Tennessee Valley Authority ET 12A ET llH 400 West Summit Hill Drive 400 West Summit Hill Drive l Knoxville, TN 37902 Knoxville, TN 37902  ;

Mr. Johnny H. Hayes, Director Mr. Michael H. Mobley, Director Tennessee Valley Authority Division of Radiological Health -

ET 12A 3rd Floor, L and C Annex 400 West Summit Hill Drive 401 Church Street Knoxville, TN 37902 Nashville, TN 37243-1532 Mr. Robert Fenech County Judge Site Vice President Hamilton County Courthouse Sequoyah Nuclear Plant Chattanooga, TN 37402 Tennessee Valley Authority P.O. Box 2000 Regional Administrator Soddy, Daisy, TN 37379 U.S. Nuclear Regulatory Commission l Region II Mr. R. M. Eytchison, Vice President 101 Marietta Street, NW., Suite 2900 Nuclear Operations Atlanta, GA 30323 Tennessee Valley Authority 3B Lookout Place Mr. William E. Holland 1101 Market Street Senior Resident Inspector Chattanooga, TN 37402-2801 Sequoyah Nuclear Plant '

U.S. Nuclear Regulatory Commission Mr. B. S. Schofield, Manager 2600 Igou Ferry Road Nuclear Licensing and Regulatory Affairs Soddy Daisy, TN 37379 Tennessee Valley Authority 4G Blue Ridge Mr. D. E. Nunn, Vice President 1101 Market Street Tennessee Valley Authority Chattanooga, TN 37402-2801 3B Lookout Place 1101 Market Street Mr. Ralph H. Shell Chattanooga 37402-2801 Site Licensing Manager Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37379

. ', ENCLOSURE Westinghouse Energy Systems ux 3ss Electric Corporat10n MusW PovisylvsNa 13230 035$

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Mr. P. G. Trudel Project Engineer TVA-91-293 Tennessee Valley Authority ET-NSt.-0PL 91-529 P. O. Box 2000 September 4, 1991 ,

Soddy Daisy, TN 37379 Ref: 1. N-025 '

2. TVA-91-283 Tennessee Valley Authority Sequoyah Nuclear Plants Units I and 2 Increase in the 100 Day Integrated Dose To Equipment In Containment Associated with Increased Fuel Burnup Justification for Continued Ooeration

Dear Mr. Trudel:

In accordance with your request, the justification for continued operation addressing an increase in the 100 integrated dose to equipment in contianment as a result of increased core fission product inventory associated wth increased fuel burnup provided in Reference 2 has been revised to incorporate your comments.

If you have any questions regarding this information, please contact the undersigned.

Very truly yours, B. J. Garry, Manager t TVA sequoyah Project M omestic Projects Department LVT/jmb cc: D. M. 1.afover R. Fortenberry l

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igf;1figg3 14:47 SG1 SITE UP/LIc 615 843 8750 P.0 3 p,g 4 3t t o : ,s i envH L i c E llS i t tG Tv T V n, SEvuGV F66L.002 TVA-91-293

. ET-NSL-0PL-I-91-529 SEQUOYAH UNITS I&2 JUSTIFICATION FOR CONTINUED GPERATION AN INCREASE IN THE 100 DAY INTEGRATED DOSE TO EQUIPMENT INSIDE AND OUTSIDE CONTAINMENT AS A RESUt.T OF INCREASED CORE FISSION PRODUCT INVENTORY ASSOCIATED WITH INCREASED FUEL BURNUP Backcround i TVA has identified a potential nonconformance affecting the Sequoyah plant (Ref 1) related to the equipment qualification program. The source term .

used in the current SQN EQ program is based on 650 EFPD burnup. The use of high burnup fuel may extend burnup from 650 EFPD to approximately 1000 EFPD.

The burnup extension will increase the core fission product inventory of certain nuclides which are the basis for the EQ program and the associated radiation exposure, beyond that currently calculated.

Safety Concern Safety related equipment located both inside and outside the containment confidence that the equipment will be able to perform its functionare t .

following a design basis acc; dent.

If the radiation dose to this equipment increases beyond the value assumed in the EQ testing program, the confidence in the ability of the equipment to function in the post-accident environment is reduced.

Hasis for Current E0 Procram N i

The current 100 day (specified by TVA) integrated accident dose is based on NUREG-0588 methodology. Specifically, the source term considers a TID-14844 release from the core of 100% of the noble gases, 50% of the iodines and 1% of the remaining fission products. This is a prescriptive source term that represents a core melt and is not related to the post-LOCA system. response of the core or the performance o'f the core cooling ,

Etsis for JCQ The following JC0 will address the potential source term and exposure increase due to extending- burnua to approximately 1000 EFPD and then describe a reduced source term yased on the results of a plant specific ,

Appendix X analysis and a gap activity release rather than the core melt -

specified in TID-14844 and NUREG-0588. By reducing the source term, the resulting integrated equipment dose will be reduced.

Imnact of Burnuo Extension The impact of extended burnup on source terms is determined by evaluation .

of the results of a Sequoyan specific ORIGEN code analysis provided by TVA (Ref. 2). The burnups considered are 600 and 1000 EFPD (consistent with the intervals provided in the CRIGEN output) rather than the current design basis burnup of 650 and the projected burnup of 1000 EFPD. Thus, ,

conservatism is introduced by etilizing 600 EFPD as the point of comparison. '

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Table 1 shows the assembly inventory of selected nuclides at 600 and 1000 EFPD, based on 4.5 *. U-235, and the fractional change in inventory. This fractional change in inventory for a lead assembly is conservatively applied to the entire core.

The nuclides selected include the gaseous and  ;

volatile nuclides that are projected to be present in the containment atmosphere. This is consistent with the guidance of TID-14844. Of the remaining fission products, per TID-14844, only lY, of the inventory is assumed to be released from the core. The selection of Cs-134 to model ,

these remaining nuclides is based on Table 2.2 of Reference 3 which I indicates Cs-134 as having the largest increase in inventary of any listed nuclide in that table. Thus, the application of the increase in Cs-134 to the spectrum of nuclides is a very conservative approach.  ;

By inspection of the ratio column in Table 1, kryptons have decreased a  :

minimum of 15Y, with the exception of Kr-85 which increased by 484.  !

Iodines have increased a maximum of 1% and xenons a maximum of 3%. Cs-134 has increased by a factor of approximately 2.5. i Note that of the short lived 1odines and noble gases, which are the dominant contributors to offsite thyroid and whole body doses, only a few have increased and the increase is small, tiost of the noble gases i have decreased. Thus,. thyroid doses are expected not to change significantly, and whole body doses are expected to decrease from those currently determined at approximately 550 EFP0 burnup.

Studies performed by Westinghouse (using Westinghouse methodology) indicate that at 100 days post-LOCA, appr,oximately 50 (25) percent of the integrated gamma (beta) dose in-containment is due to halogens, 30 (39)  ;

percent due to noble gases and 20 (36) percent due to'the remaining fission products. These contributions are based on a TID-14844 core release.

The dose contribution from each nuclide is not readily available.

However-by using the Table 1 ratios in a bounding. manner (i.e., assume that iodines all increase by if,, noble gases by 1%, and the remaining fission products - assumed to be represented by Cs-134 - by a factor of ,

2.47), the overall increase in exposure can be conservatively' determined.

The increase in noble gases was based on the 1". increase in Xe-133m rather >

than the 48f. increase in Kr-85 because Kr-85 is not a significant contributor to dose, in part because of its relatively low gamma energy (average gamma energy of 0.0023, average beta energy of 0.25 Mev/ disintegration) and because there is a very small inventory in the i core when compared to other nuclides. The increase in gamma dose is calculated as follows: .

lodine's r (0.5 of dose)(1.01 increase) - 0.505 Noblo Gases (0.3 of' dose)(1.01 increase) - 0.303 '

i Remaining Nuclides (0.2 of dose)(2.47 increase) - 0.494 TOTAL revised gamma dose - 1.302 Thus there is conservatively estimated to be a 30~. increase in the 100 day integreted gamma dose to equipment located in-containment.

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10/21/1993 14:49 SCN SITE UP/t.lC 615 843 SWO P.05' f E r+ - 4 *31 16: 40 F T< ON LICCHGlHG TO TVA SEGUOY PME . 004 r TVA-91-293 ET-NSt. -0PL-I 5{ 9 The increase in beta dose is calculated as follows:

Iodines (0.25 of dose)(1.01 increase) - 0.253 l Nobio Gases '

(0.39 of dose)(1.01 increase) - 0.394 Remaining Nuclides (0.36 of dose)(2.47 increase) = 0.889 .

TOTAL revised beta dose - 1.536 Thus there is conservatively estimated to be a 547. increase in the 100 day l integrated beta dose to equipment located in-containment.

Equipment located outside containment (residual heat removal / containment spray pump room) will be exposed only to those fission products contained i in the recirculating core cooling solution (sump solution). Thus, only iodines and the remaining fission products are considered. Further, it is assumed that 50% of the released iodines will instantaneously plate-out on the internal surfaces of the reactor coolant system and/ or on internal containment surfaces, consistent with the recommendation of TID-I4844 for use in offsite dose analysis. Assuming that 1odines and the remaining ,

fission products account for 100% of the dose, the gamma dose to equipment outside' containment is calculated as follows. }

i lodines (0.71 of doso)(1.01 increase)(0.5 plateout) - 0.359 i

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Remaining (0.29 of dose)(2.47 increase) - 0.716  !

Nuclides TOTAL revised gamma dose - 1.075 Thus there is conservatively estimated to be a 7.5% increase in the 100 l'

day integrated gamma dose to equipment located outside containment that is exposed to sump solution.

The bota dose to equipment outside containment is calculated as follows:

, lodines (0.41 of dose)(1.01 increase)(0.5 plateout) - 0.207 Remaining (O'.59 of dose)(2.47 increase) = 1.457 Nuclides TOTAL revised beta dose - 1.664 Thus there is conservatively estimated to be a 66% increase in the 100 day integrated beta. dose to equipment located outside containment that is exposed to sump solution.

If the Cs-134 inventory is evaluated at 800 EFPD {this conservatively bounds the Unit 2 Cycle 4 average burnup of 740 EFpD) rather than at 1000 EFPD, the source term factor decreases from 2.47 to 1.58 (see the attached plot of Cs inventory vs. burnup). The total gamma dose at 800 EFPD is reduced to 0.817 and the beta dose is reduced to 1.139. Thus, there is a i

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18Y. reduction in the 100 day integrated gamma dose and a 14Y. increase in beta dose to equipment located outside containment. l Core Release Fraction +

l The core fission product release fraction is equal to the product of the fuel rod burst fraction and the rod gap fraction.

The fraction of fuel rod bursts for a large-break LOCA is estimated from the results of 'the Sequoyah Appendix K analysis. The results of this estimation indicate that less than 507. of the rods will burst.

The fuel rod gap fraction specified in Regulatory Guide 1.77 (ref 4) is  !

0.1 for iodines and noble gases (TID-14844 release fraction is 0.5 for  :

iodine and 1.0 for noble gas). Ass'ume that cesium (assumed to represent l the remaining fission products) will behave similarly to iodine and .

hence, the gap fraction will be 1/5 of the TID (core melt) fraction or 0.002 (TID-14844 release fraction for Cs is 0.01). Note that the i Regulatory Guide gap fractions are licensing basis values not calculated i or expected values. The results of Seneric calculations performed by  :

Westinghouse using the ANS 5.4 model suggest a maximum gap fraction of approximately 0.02 for todines. Thus, the use of calculated values would l provide an addi'tional reduction in the source term.

The core release fractions, based on the Regulatory Guide gap fraction, are 0.05 for iodine and noble gas, and 0.001 for cesium. Hence, the iodine and cesium sourcs term is reduu;d py a factor of 10, and the noble  ;

gas source term is reduced by a factor of 20. 6 The normalized dose provided by a gap release, versus a TID-14844 release, .

is calculated as follows:  !

Iodines 0.5 of total / 10 reduction f actor - 0.05

~

Noble Gas 0.3 of total / 20 reduction f actor - 0.015 Remain ~1ng 0.2 of total /10 reduction factor - 0.02 '

Nuclides I

TOTAL revised dose - 0.085 Thus, the integrated dose (beta and gamma) to equipment located in  ;

containment is conservatively estimated to decrease by a factor of 11.8 .

l due to a gap release versus a TID release.  !

The dose (beta and gama) to equipment located outside containment, based on iodines and the remaining nuclides, decreases by a factor of 10 due to .

a gap release. '

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Summarv i

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The fraction of rods in the core assumed to burst is 0.5.

The iodine and noble gas gap fraction is 0.1. '

The cesium gap fraction is 0.002.

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50% (25%) of the 100 day integrated gamma (beta) dose inside '

s_qntainment is due to halogens, 30% (39%) due to noble gases an'd 20% (36%) due to the remaining fission products.

71% (41%) of the 100 day integrated gamma (beta) dose to equipment located outside containment is due to halogens and 29% l (59%) is due to the remaining fission. products The source term increase for equipment located inside containment due to burnup extension to 1000 EFPD is a factor of 1.3 for gamma _

exposure and a factor of 1.54 for beta exposure. -

The source term reduction based on 800 EFPD burnup, for dose to equipment located outside containment, is a faitor of 0.82 for 1 gamma dose and a 14% increase for beta dose.

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The source term reduction for a gap release vs. a TID-14844 l release is a factor of H,8 for eautoment inside dontainment and  ;

a factor of 10 for ecuioment outsjde containment.  ;

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The aggregate gamma dose reduction to equipment inside' containment, considering both the source term increase due to increased burnu;) and the source term decrease due to the gap release, is an approximate factor of 9.

i The beta dose reduction is an approximato factor.of 7.6. i The aggregate gama dose reduction to equipment outside containment is an approximate factor of 12.2, based on 800 EFPD of operation, or a' factor of 9.3 based on 1000'EFPD.

8.8 at 800 EFP0 and 6.0 at 1000 EFPD.The beta reduction is an approximate factor of Conclusion By utilizing a conservative calculated source term rather than the  !

prescriptive TID-14844 source term, the 100 day integrated beta and gamma doses to equipment located inside and outside containment are -

significantly reduced. The calculated source term is based on the results of a conservative Appendix K analysis of core and ECCS performance following a large-break-LOCA and conservative Regulatory Guide gap fractions. The ' impact of burnup extension is based on the results of a state of-the-art function of burnup.

calculation of core fission product inventory as a The results of the evaluation indicate that there is substantial margin betwsen the expected performance of the plant, based on conservative calculations and the prescribed performance provided by the NRC. Thus, there is reasonable assurance of safe operation pending final resolution of the potential equipment qualification nonconformance. Therefore, plant operation up to 1000 EFPD is justified for Sequoyah Units 1 and 2. s 5 l l

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10/21/1993 14:51 SON SITE UP4 IC 615 843 ETEO P.08 CEP 4 '91 16: 43 FROH LICEbt!NG TO IVA S E QUO *i- MGE,007 TVA-91-293 ET-NSL-OPL-1-91-529 Refersnces

1. TVA CAQR'SQP900372, September 18, 1990.

2.

" SASH 2H Input and Assumptions for Cases Computing Isotopic Activities for Sequoyah Nuclear Plant," Appendix A, Oak Ridge National Laboratory, ORNL/M-1503, July 1991.

3. " Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors," NtJREG/CR-5009, January 1988.

4. Assumptions Used for Evaluating A Control Rod Ejection Accident for Pressurized Water Reactors", Regulatory Guide 1.77, May 1974.

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TVA-91-293 ET-NSL-0PL-I-91-529 I TABLE 1 ASSEMBLY INVENTORY OF SELECTED NUCLIOES AT 600 AND 1000 EFPD Nuclide 600 EFPD 1000 EFPD Ratio 1000/600 Kr-85 3.73E3 5.52E3 1.48 i Kr-85m 1.4BE5 1.26E5 0.85 l Kr-87 2.82E5 2.37ES 0.84  !

Kr-88 3.91E5 3.26E5 0.83 l Kr-89 4.94E5 4.05E5 0.82 l Xe-131m 5.45E3 5.59E3 1.03 a

Xc-133 1.0E6 9.88E5 0.99 l Xe-133m 3.17E4 3.19E4 1.01 i Xe-135 3.42E5 3.09ES 0.9  ;

Xe-135m 2.20E5 2.09E5 0.95 Xe-137 9.21E5 9.1ES 0.99 Xe-138 8.98E5 l' 8.61ES 0.96 l-131 4.86ES 4.92E5 1.01 i I-132 7.15ES 7.1SES 1.01  :

1-133 3.02E6 1.01E6 0.99 l-134 1.14ES 1.llE6 0.97 I-135. . 9.72E5 9.64E5 -

0.99 >

Cs 134 3.6EE4 l 8.93E4 (5.6E4) 2.47 (1.56)**

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• • These values ( ) are based on 800 EFPD burnup 1

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