HNP-17-093, Supplement to License Amendment Request Proposing a New Set of Fission Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3

Supplement to License Amendment Request Proposing a New Set of Fission Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3

ML17338A122
Person / Time
Site:	Harris
Issue date:	11/29/2017
From:	Hamilton T Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF9740, HNP-17-093
Download: ML17338A122 (13)

Text

Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562-9300 919.362.2502 10 CFR 50.90 November 29, 2017 Serial: HNP-17-093 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63

Subject:

Supplement to License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3 (CAC No. MF9740)

Ladies and Gentlemen:

By application dated May 22, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17142A411), Duke Energy Progress, LLC (Duke Energy),

submitted a license amendment request (LAR) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed license amendment requested to revise the facility as described in the HNP Final Safety Analysis Report (FSAR) to provide gas gap release fractions for high-burnup fuel rods (i.e., greater than 54 gigawatt days per metric ton of uranium (GWD/MTU)) that exceed the 6.3 kilowatt per foot (kW/ft) linear heat generation rate (LHGR) limit detailed in Table 3, Non-LOCA Fraction of Fission Product Inventory in Gap, of Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the LAR and determined that additional information was needed to complete their review. Duke Energy provided a response to the NRC request for additional information (RAI) in a letter dated October 30, 2017 (ADAMS Accession No. ML17303A667). In the response to RAI #3, Duke Energy revised the radiological consequence analysis of the Fuel Handling Accident in the Fuel Handling Building (FHB) to remove all filtration credit for the FHB emergency exhaust system. By this letter, Duke Energy provides the updated mark-up of the proposed FSAR changes to reflect the removal of this filtration credit and replace Enclosure 2 of the original submittal in its entirety.

This additional information does not change the No Significant Hazards Consideration Determination provided in the original submittal. No regulatory commitments are contained in this letter.

In accordance with 10 CFR 50.91(b), HNP is providing the state of North Carolina with a copy of this supplement.

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial HNP-17-093 Should you have any questions regarding this submittal, please contact Jeffrey Robertson, Manager- Regulatory Affairs, at (919) 362-3137.

I declare under penalty of perjury that the foregoing is true and correct. Executed on NovemberJ~, 2017.

Sincerely, Tanya M. Hamilton

Attachment:

Updated Proposed Final Safety Analysis Report Changes (Mark-up) cc: J. Zeiler, NRC Senior Resident Inspector, HNP W. L. Cox, Ill, Section Chief, N.C. DHSR M. Barillas, NRC Project Manager, HNP C. Haney, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial HNP-17-093 Should you have any questions regarding this submittal, please contact Jeffrey Robertson, Manager - Regulatory Affairs, at (919) 362-3137.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November , 2017.

Sincerely, Tanya M. Hamilton

Attachment:

Updated Proposed Final Safety Analysis Report Changes (Mark-up) cc: J. Zeiler, NRC Senior Resident Inspector, HNP W. L. Cox, III, Section Chief, N.C. DHSR M. Barillas, NRC Project Manager, HNP C. Haney, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial HNP-17-093 Attachment SERIAL HNP-17-093 ATTACHMENT UPDATED PROPOSED FINAL SAFETY ANALYSIS REPORT CHANGES (MARK-UP)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

SHNPP FSAR 15.7.4 Design Basis Fuel Handling Acc1dents 15.7.4. I~entific~y1on of c~uses and Acc1dgnt Descrjption. The 1

possibility o aue han ing acci ent is remoteecause of the many interlocks. administrative controls, and physical limitations imposed on the fuel handling operations. All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a senior reactor operator (SRO). The analyzed Fuel Handling Accident inside containment involves dropping a spent fuel assembly resulting 1n the rupture of the cladding of all the fuel rods (264) in the assembly.

The projected worst case Fuel Handling Accident (FHA) 1n the Fuel I Handling Building CFHB) involves dropping a recently discharged (100 hr decayed) PWR assembly (including the handling tool) on top of another recently discharged PWR assembly in a fuel storage rack. The dropped assembly subsequently falls over landing on BWR fuel assemblies in an adjacent storage rack. Fifty fuel rods are projected to fail 1n the impacted PWR assembly in storage and all of the rods (264) in the dropped assembly fail when the assembly falls over (Reference 15.7.4.5). Due to the upper bail handle of the BWR fuel assemblies extending above the top of the BWR storage racks. up to 52 BWR assemblies could be impacted when the dropped PWR assembly falls over.

All of the rods in the impacted BWR assemblies are assumed to fail .

15.7.4,2 Radiological Consequences Analysis 15.7.4.2.1 Input Assumptions CORIOOn to both FHA 1n the FHB and in Containment .

Consistent with Regulatory Guide 1.183 (Position 1.2 of Appendix 8). the radionuclides considered are xenons. kryptons. halogens. cesiums and rub1d1ums. The list of xenons, kryptons, and halogens considered is given in Tables 15.7.4-1 and 15.7.4-3. The cesium and rubidium are not included because they are not assumed to be released from the pool as discussed later.

The calculation of the radiological consequences following a FHA uses gae fractions of 8% for 1-131. ~for Kr-85, and 51 for all otherAnuclides{ f?e-k 3017/o p~'vie~

Iodine spec1es 1n the pool is 99.85% elemental and 0.15% organic iodine. This is based on the split leaving the fuel of 951 cesium iodide (Csl) , 4.851 elemental iodine and 0.15% organic iodine. It is assumed that all Csl 1s dissociated in the water and re-evolves as elemental. This is assumed to occur instantaneously. Thus. 99.85% of the iodine released is elemental.

The water above the damaged fuel rods retains a large fraction of the gaR t::e:~::r:h:

act1v1ty of iodines. An overall effective decontamination factor (OF) of 200

~ ~;=~t ::t=~l:!.

aAil )'&e, fep \he liwe w*

5 ef ~e"h1e

i!t:,~:~::;=:=l:!1~J:°l-hes-iN tilMt HAii. .

The cesium and rubidium released from the damaged fuel rods is assumed to remain in a nonvolatile form and would not be released from the pool .

15.7.4-1 .4Jnendment No. 51

SHNPP FSAR 15.7.4.2.2 Postulated Fuel Handling Accident in the FHB The major assumptions anmarameters used 1n the analysis are item1zed 1n Table 15.7 .4-1. This anal sis involves dropP.1ng a recently discharged (100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay> PWR fuel ass ly onto 52 Brunswick BWR fuel assemblies. This analysis also includes 50 PWR rods additionally damaged 1n the accident. the assembly inventory is based on the assumption that tne PWR fuel assembly has been pperated at 1.73 times the core average power and the BWR fuel assemblies have been operated at 1.5 times the core average power. All act1v1ty released from the fue 1 poo1 is assumed to be rel eased to the atmosphere in two hours.

The BWR fuel inventory was conservatively evaluated at the IF-JOO spent fuel shipping cask limits for GE- 7. e. 9. 10 and 13 fuel asserrblies with a maximum average lattice enrichment of 4.~5 wt.% U-235 and a maxinun assent>ly average burnup of 45 GWD/KTU. The decay time used in the analysis is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for the PWR fuel and 4 years for the BWR fuel. Thus . the analysis supports the design basis limit of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay t1me prior to fuel movement.

It was detennined that for the HNP SP-eC1f1c water height above the failed fuel in the fuel hand11ng building of 21 feet the elemental OF would be at least 291. c~ared to the R~. Guide 1.183 al 1owable elemental OF of 500. Using the elemental OF 291. it was determined that overall effective OF for 21 feet of coverage would be 203. Since this continues to exceed the Reg. Gu1de 1.183 cited overall effective OF of 200. it remains conservative to use the overall DF of 200 in the HNP dose calculations.

~;~e~~-ttaken for removal of iodine by filters by the spent fuel J)OOl ventilation system operation. Credit is not taken for isolation or- release paths.

The activity released from the damaged assent>lies is assumed to be released to the fuel bu11d1ng and subsequently to the atmosphere over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period.

15.7.4.2.3 Postulated Fuel Handling Accident 1n Containment A fuel assembly 1s assumed to be droppecl in containment and damaged dur1ng refueling. Activity released from the damaged assennly is released to the outside atmosohere through the contaimient openings {such as the personnel air lock door or the ~qu1pment hatch).

The major ass~tions are parameters used in the analysis are itemized in Table 15.7 .4-3. This analysis involves dropping a recently discharged (100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay) PWR fuel assenoly. All activity released from the fuel _pool is assumed to be released to the atmo~phere in two hours. The 1>901 referred to in RG 1.183 1s interpreted as the flooded reactor cavity for the purposes of evaluating the fuel handling accident in contairvnent. No credit is taken for isolation of contairvnent for the FHA containment.

The calculation of the radiolQQical consequences following a FHA uses gap fractions of es for 1-131.-1/4* for Kr-85. and 51 for all otherAnuclides ( R~ ce.

It is assumed that all of t~1uel rods in the ~uivalent of P <frft1.re1 asserrt>ly 15*7*"-11)

(264 rods) are damaged to the extent that all their gap activity is released.

The assembly inventory is based on the ass~tions t~at the subject fuel assed:>ly has been operated at 1.73 times the core average power.

The decay t1me used in the analysis is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> .

15.7 .4-2 Amendment No . 51

SHNPP FSAR It was determined that for HNP specific water height above the failed fuel in the containment of 22 feet. the elemental OF would be at least 382. compared to the Reg. Guide 1.183 allowable elemental OF of 500. Using the elemental OF of 382, it was determined that the overall effective OF for 22 feet of coverage would be 243. Since this continues to exceed the Reg. Guide 1.183 cited overall effective OF of 200. it remains conservative to use the overall OF of 200 in the HNP dose calculations.

No credit is taken for removal of iodine by filters nor is credit taken for isolation of release paths.

Although the containment pur~e will be automatically isolated on a purge line high radiation alarm. isolat,on is not modeled in the analysis. The activity released from the damaged assembly is assumed to be released to the outside atmosphere over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. Since no filters or containment isolation is modeled. this analysis supports refueling operation with the equipment hatch or personnel air lock remaining open.

15.7.4.2.4 Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.l.

li.7.f-:-2.5 Control Room Doses 15'..7.~ .z.r The control room assumptions are P.rovided in Section 15.6.5.4.3 and table 15.6.5-15. The FHA control room ijoses modeled-set cfm unfiltered inleakage.

300 It is assumed that the control room HVAC system begins in normal mode. The activity level in the intake duct causes a high radiation signal almost imnediately. It is conservatively assumed that the post accident recirculation control room HVAC mode is entered 15 seconds after event initiation. The control room HVAC is placed into pressurization mode at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after isolation signal.

15.7.4.2.6 Results The analttically predicted dose consequence to the Control Room CCR) operators due to LBLOCA increased by a small amount. based on the reduction in CR recirculation flow which reduces the iodine filtration P.rovided by the charcoal filters relative to the analysis of record con~ition. The analytically predicted dose consequence to the control room operators for the FHA events were not similarly revised or updated since the LBLOCA OBA is the limiting control room operator dose event. The radiological analysis results for the FHA in FHB doses are listed in Table 15.7.4-2. The FHA in Containment doses are listed in Table 15.7.4-4. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The resultant doses are within the applicable limits. The offsite doses are less than -25% of the 10CFRS0.67 limits (i.e .. 6.3 rem TEDE) and the control room dose is less than the 10CFRS0.67 limit of 5 rem TEDE.

15.7.4.3 DELETED 15.7.4.3.1 DELETED 15.7.4.3.2 DELETED is.7.4-3 Amendment No. 54

SHNPP FSAR 15,7.4,4 Deleted.

15.7.4.4.1 Deleted.

15.7.4.4.2 Deleted.

15.7.4.4.3 Deleted.

15.7.4.4.4 Deleted.

15.7.~.5 Other F¥e~ Handling ~c1dftnts. Fuel handling drop acc1dents involving t e other fueandling toos C PRA. RCCA change tool, spent fuel handling tool). and items carried by the tools have also been evaluated (Reference 15.7.4-7) and are addressed in Section 9.1. The tool drop scenarios involve dropping the tools. and items carried by the tools, onto PWR spent fuel racks, BWR SJ:?ent fuel racks. and combinations of both. For all cases evaluated. the off-site dose consequences were determined o. . nded by the Fuel Handling Accident described in FSAR Section . . . 1ch addresses a fuel handling drop accident which results 1n damage to 314 PWR spent fuel rods and 52 BWR spent fuel assemblies (Reference 15.7.4-7. pages 3.2.2-3.23).

15,7,4.2.2 Amendment No . 51

SHNPP FSAR TABLE 15.7.4-1 Parameters used in Fuel Handllga Ac9iient  !~~J1e Ra_1olog1c_l An ______§ the fuel Handling Building Radial ~ak1ng factor (PWR fuel) 1.73 CBWR fuel) 1.5

~ Fuel damaged (number of assemb11es) 1 1.2 PWR (314 rods) +

~ oil;: ALI cla.,...ctje{ Pwg r?Jds a~u~ -ft, exceext 6.3 ;;kl~?:. BWR ,1 Time from shutdown before fuel movement (PWR)(hr) 1oo ~b.,ve Pl GwJ>;MTV (BWR fuel )(yr) 4 6 r.Jrr1Uf J Activity in the damaged fuel assembies (C1)

I-131 7.21E5 I-133 7.59E4 1-135 5.57El Kr-85 1.41E5 Xe-131m 9.06E3 Xe-133m l .77E4 Xe-133 l .19E6 Xe-135 2.41E2 Gap Fractions (I of core act1v1ty)

I-131 8 Kr-85 --i&- 30 Other Iodine and Noble Gas nuclides 5 Water depth 21 feet Overall pool iodine scrubbing factor 200 Iodine chemical form 1n release to atmosphere Ct)

Elemental 70 Organic 30 Particulate 0 Spent Fuel Pool Ventilation System ~ilter effitiaAC.Y No t=i\-trcO',e>,.l ,.~~t.o"'eJ.

ElemeA~el .gs.

9fgaAie Parti,ulate - Isolation of release No isolation assumed Time to release all act1vity (hours) 2 Act1v1ty Released. (C1) 1-131 I-133 l..-69\\- e~ l .~9Cl I. " t!:.~ \

I * ~SE\ --9. 411E= 1 * 't."r St,:£ 1-135 , *3C\3 ! * ~ &. 950E-+ t>.'103 E ' I Kr-85 l , 41QE.4 4 ~Z3oE 4 Xe-131m 4.530E2 Xe-133m 8.850E2 Xe-133 5.950E4 Xe-135 l .205El 15.7 .4-8 Amendment No. 51

SHNPP FSAR TABLE 15.7,4-2 BADIOLOOICAL ACCIN tf8W~YifhCES f POS!Jlti!'IILING N l FU L HANDL A F8i£L HANDLING Exclusion Area Boundari 1-,.~:l c9_3:S: +.* rem TEDE Low Population Zone o.SS &. 878" ---0.871 rem TEDE Contro1 Room \ ~ OS 61. t9 i' 3 -0 .12 rem TEDE The exclusion area boundary dose reported is for the worst two hour period. determined to be from Oto 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

15.7.4-10 Amendment No . 51

SHNPP FSAR Radial peaking factor 1.73 J.::r:rJSEe~ cL~~~~fP~ (nu~ ~ o~,a~ e_r,!?~~~],..d-ro ev.ceed 6.3 kL</4 ~ove 51.{ GwJ>/,-tTLJ Time from ~~ be¥ore ~~ movement (hr) 100 bufT\up:]

Activity in the damaged fuel assembly (Ci)

I-131 6.06E5 I-133 6.38E4 1-135 4.68El Kr-85 8.82E3 Xe-131m 7.61E3 Xe-133m l.49E4 Xe-133 9.97E5 Xe-135 2.03E2 Gap Fractions(% of core activity) 1-131 8 Kr-85 -1/4&- 30 other Iodine and Noble Gas nuclides 5 Water depth 22 feet Overall pool iodine scrubbing factor 200 Iodine chemical form in release to atmosphere (I)

Elemental 70 Organic 30 Particulate 0 Filter efficiency No filtration assumed Isolation of release No isolation assumed Time to release all activity (hours) 2 Activity Released (Ci) 1-131 -2.428E2- 2.l/2~£2 1-133 -i .592'1* I -~95"£ I 1-135 l.168E 2 /./76£-2 Kr-85 B.&28E2 2.~0£3 Xe-131m 3.805[2 Xe-133m 7.450[2 Xe-133 4.985E4 Xe-135 1.015El

15. 7.4-11 Amendment No. 51

SHNPP FSAR TABLE 15.7.4-4 Radio]ogjca] cggM~5tcY~sj~ea&8tl~~= Fye] Hand]jng Exclusion Area Boundary* 2 .. 02 ~rem TEDE Low Population Zone 0.46 rem TEOE Control Room o.88' -1/4-:-39' rem TEOE

• The exclusion area boundary dose reported is for the worst two hour period. determined to be from Oto 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

15 .7.4-13 Amendment No. 51

SHNPP FSAR

REFERENCES:

SECTION 15.7 15.7.4-1 Industrial Ventilation. 8th Edition. American Conference of Governmental Industrial Hygienists.

15.7.4-2 Deleted by Amendment No. 49

15. 7.4-3 Deleted by Amednment No. 51 15.7.4-4 Regulatory Guide 1.183. "Alternative Radiolog1cal Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors... July 2000.

15.7.4-5 Westinghouse Letter. 97CP-G-0006: Christine M. Vertes to Leo Martin. dated April 9. 1997. *Limiting Fuel Handling Accident Assunptions."

15.7.4-6 Deleted by Amendment No. 49 15.7.4-7 ESR 98-00181 MFuel Handling Tool Drop onto Spent Fuel Rack Evaluation*

15.7.4-8 Deleted by Amendment No. 51

~ 9 Deleted by Arnendaent No. 51 15 .. 4-10 CP&L Calculation HNP-H/FHB-1001 "Off-site Doses from FHB Cask Drop."

[:c~t"J 15"~,.Y-/t 15.7R-l Amendment No . 51