ML19070A116

Westinghouse, Responses to NRC Request for Additional Information Documented in ADAMS Accession No. ML 18270A094 on the Core Design and Safety Analyses Methodology Transition Program

ML19070A116
Person / Time
Site:	Wolf Creek
Issue date:	03/05/2019
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
References
ET 19-0008
Download: ML19070A116 (8)

Text

ENCLOSURE II TO ET 19-0008 Westinghouse Responses to NRC Request for Additional Information Documented in ADAMS Accession No. ML18270A094 on the Core Design and Safety Analyses Methodology Transition Program

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Westinghouse Non-Proprietary Class 3 Page 1 of 7 SAP-19-8 NP-Attachment Our Ref: SAP-19*8 Westinghouse Responses to NRC Request for Additional Information Documented in ADAMS Accession No. ML18270A094 on the Core Design and Safety Analyses Methodology Transition Program

[Non-Proprietary]

(7 pages including cover page)

©2019 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 2/14/2019 6:10:00 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 2 of 7 SAP-19-8 NP-Attachment Our Ref: SAP-19-8 Request Rod C/usrer Colltrol t\ssembl\' Eiection Accitle,u Please demonstrate that the CREA analysis re.wills discussed in Section 2.5.6 c~f E11clos11re I are cnmiste/11 with General Design Criterion (GDC) 28. "Reactiviry limits." of Appendix A. "Ge11eml Design Criteria for Nuclear Poll'er Piams." to Tirle 10 of The Codci of Fee/era/ Regulations ( /0 CFR) Part 50 for long Term cooling. and ll'ith the ass11111ptions employed ,rith the rm/iological analyses supporting the i111ple111e11tatio11 of the altematil'e source 1er111 in accordance ll'ith /0 CFR 50.67. "Accidem .1*nurce term." CDC 28 requires that rhe e.ffecrs of postulated reactivity acciclems neither damage !he 1w1cror coo/am pressure bomulary greater 1ha11 limited local yielding. nor impair 1!,e ability to cool !he core. ,\tlea11while. !he radiological consequence analysis of the CREA ass11111es no more than I 0-percelll ofji,el ce/1/erline melt m 1he hot spot.

Since the NRC position statcid i11 Regulatory Guide 1.77. "A.1*s11111ptio11s med for Emluating o Comrol Rod Ejeclion Accide111Jor Pressurized Water Reacwrs." May /974 (ADAMS Accession No. Ml003740279), indicating that acceptance criteria of280 calories per gram (cal/gm) should be applied to the e1*e111. and Westinghouse's position that 200 cal/gm and /0-percent fuel cellterline melt criteria contained in \VCAP-7588. "An Emlulllion of the Rod Ejection Ac:cide11t in \Vesti11gho11se Press11rizecl Water Rec1ctors Using Spatio/ Kinetics Methods," were eswblished.

the NRC staff deten11i11e,l that more restrictive occepta11ce criterio were needed to assure the core would re111oi11 in a coo/able geometry. given the potential for high tempemture cladding failure and pellet cladding intemctio11. This position is doc:11111e111ed in an NRC staffi11temol 111e1110 dated April 3. 20/5, titled. "Results of Periodic Re11iew of Regulat()I)' Guide /.77" (ADAMS Accession No. Ml/5075A3/ I). The positio11 is supported by the NRC staff i11temo/

memo dated January I 9. 2007. titled, "Technical and Regulatory Bosis for the Reac:til'ity Initiated Accide111 /111eri111 Acceptance Criteria and Guidance" (ADAMS Accession No. Ml070220400).

o) Please justijj' why 200 cal/gm 011d /0-percellt fuel melt are appropriate acceptance criteria gfren that much more is known 110w abo11T fuel damage beht1l'ior thm, when The method was approi*ed in /975.

b) Please discuss how the calculatio11s appropriotely include The effects of 1111dear fuel TCD in the evaluation against the acceptance criterio. The hot spot Juel melt limit of I0-percent relates to the prel'efl/io11 offuel dispersal imo the coolant. a11d is reflected in the assu111ptio11s employed in the mcliological analyses.

Response

As subsequently demonstrated by the responses to part a) and pan b) of this request. the Rod Cluster Control Assembly (RCCA) ejection accident analysis results discussed in Section 2.5.6 of WCAP-17658-NP Revision I are consistent with General Design Criterion (GDC) 28. **Reactivity limits." of Appendix A. Genernl Design Criteria for Nuclear Power Plants:* to Title IO of the Code of Federal Regulations (IO CFR) Part 50 for long term cooling.

and with the assumptions employed with the radiological analyses supporting the implementation of the alternative source term.

a) Ple<1se justify 1rhy 200 cal/gm and JO-percent fuel melt are appropriate accepllmce criterio given that much more-is k11ow1111011* abo11t fuel damage behavior thc111 whe11 the method u*as approved in 1975.

The WCAP-7588 Revision 1-A acceptance criteria arc a conservative replacement for the Regulatory Guide I. 77 criteria. as subsequently discussed. Specifically. the 200 cal/gm criterion is the conservative coolability limit utilized by Westinghouse relative to the Regulatory Guide I. 77 limit of 280 cal/gm. Additionally. the I0% fuel melt criterion is a conservative limit that has been used by Westinghouse for input to the radiological close calculations.

The 10% fuel melt limit is nm intended lo be used as the coolability limit.

As the NRC has stated. more restrii.:tive ai.:ceptam:e i.:riteria than the Regulatory Guide 1.77 criteria may be needed.

However. the existing 1-D analysis methodology is an overly-conservative method relative to more advanced 2-D or 3-D methods. The analysis documented in Section 2.5.6 of WCAP-17658-NP Revision I utilized the more restrictive WCAP-7588 Revision I-A acceptance criteria (200 i.:al/gm) along with the historical 1-D analysis method in order lfl maintain the overall conservatism or the analysis.

• • • This record was final approved on 2/14/2019 6:10:00 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 3 of 7 SAP-19-8 NP-Attachment Our Ref: SAP-19-8 It is noted that utilizing this approach (i.e .. the 200 cal/gm limit in combination with the conservative 1-0 am1lysis method) was discussed between Wolf Creek. Westinghouse. and the NRC during a clarification call associated with this request in order to ensure that there was understanding on how the evaluation of the RCCA ejection accident would be performed.

I>) Please discuss how rlie calculatiom appropriately i11cl11de the effi!cts of 1111clearf11el TCD in the el'al11alio11 against the acceptance criteria. The hot spot Juel melt limit of I 0-percelll relates to the preve111io11 offuel dispersal ii/to the coola11t. mu/ is reflected i11 the. assumptions employed in the radiological analyses.

Regarding the issue of fuel thermal conductivity degradation (TCD). WCAP-17658-NP Revision I referenced LTR-NRC-12-18 (ADAMS Accession No. MLI 2053Al05) as the basis for the continued safe operation of plants analyzed with Westinghouse codes and methods. consistent with the basis used for other plants analyzed with Westinghouse codes and methods. The rationale for referencing LTR-NRC-12-18 was due to the fact that the Westinghouse Performance Analysis and Design Model (PADS) submittal. WCAP-17642-P-A Revision 1. was still under review and had not yet been approved by the NRC at the time that the licensing report for the Methodology Transition, WCAP-17658-NP Revision I. was submitted.

Following the submittal of the Methodology Transition license amendment request, WCAP-17642-P-A Revision I was approved for use and provided the technical basis for utilizing the PADS model to address fuel TCD. However.

as documented within LTR-NRC-18-7 (ADAMS Accession No. ML18023B555), the schedule to perform all not-LOCA analyses required to transition to the PADS. models and methods is [ ]3-e Furthermore. [

J'"e Therefore, due to the timing of the approval of WCAP-17642-P-A Revision I relative to the timing of the Methodology Transition submittal, it was not feasible to fully incorporate the PADS models and methods into the not-LOCA analyses within the confines of the Methodology Transition submittal.

Nevertheless, in order to respond to this request and to demonstrate that the acceptance criteria (i.e., 200 cal/gm and I0% fuel melt) for the RCCA ejection accident will continue to be met once the effects of TCD are explicitly included within the licensing basis analysis, a demonstration analysis was performed. This demonstration analysis was based on the RCCA ejection analysis described in Section 2.5.6 of WCAP-17658-NP Revision I, but with analysis inputs revised to account for the effects ofTCD. The revised analysis inputs were based on the models and methods described in WCAP-17642-P-A Revision I. Specifically. the input changes included [

re for both the HFP and hot zero power (HZP) cases. The conservative ejected rod worth values used in the analysis described in WCAP-17658-NP Revision 1 continued to be used.

The burnup-dependent inputs for each of the cases in the demonstration analysis were based on [

]"". In addition, as the fuel melting temperature is burnup-dependent. it was decreased from 4900°F at beginning of cycle rBOC> and 4800°F at end or cycle (EOC) to r re for all cases. I 1

The Wolf Creek Generating Station (WCGS) Technical Specification Safety Limit 2.1.1.2 is 508D°F decreasing at 58°F per 10.000 MWD/MTU ofburnup.

• • • This record was final approved on 2/14/2019 6:10:00 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 4of 7 SAP-19*8 NP-Allachmenl Our Ref: SAP-19*8 As expected. with the effects of TCD included. the peak fuel average and centerline temperatures increased. For the HZPcases. [

)"" The results from the demonstration analysis are presented in Table I.

In summary. the demonstration analysis of the RCCA ejection accident utilized the same inputs as those used for the analysis described in Section 2.5.6 of WCAP-17658-NP Revision I. with the following exceptions:

a.c Table I: Wolf Creek RCCA Ejection Results a,c Based upon the results in Table I. all 111.:ceptance criteria continue to be met. Therefore. ii is concluded that the*

effects ofTCD can be accommod.ned for the RCCA ejection accident for Wolf Creek.

As previously stated. [

]3- Thus. the intent is to not incorporate the PADS models and methods into the licensing basis for the RCCA ejection accident at this time. as the remaining not-LOCA analyses have not yet been updated (with the exception of the Steamline Break (SLB) analysis. which is discussed in the subsequent response).

It is recognized that the demonstration analysis performed for this evaluaiion [

]""c Therefore. it has already been demonstrated that the analysis bounds the upcoming fuel cycle. Furthermore. to ensure that the results cominue to remain valid for future cycles. an additional check will be performed as part of the reload process to contil'm lhat the 1

[ ]' *< remain bounding on a cycle-specific basis until Wolf Creek fully implements the PAD5 models and methods into its licensing basis for the 1101-LOCA analyses. Once the PAD5 models and methods are fully incorporated into the Wolf Creek licensing basis. the PADS analysis will become the analysis of record and the [ ]""c will become the limits checked as part of the standard reload process.

• • • This record was final approved on 2/14/2019 6:10:00 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 5of 7 SAP-19-8 NP-Attachment Our Ref: SAP-19*8 Request S1ea111li11e Break Accidem Please demmwrate holl' the 111/SLB analyses co11tai11ed in Section 2.2.5 of Enclosure I are co11siste111 ll'ith the requiremellfs of /0 CFR Part 50. Appendix A. GDC 27. "Combined reactivity control systems cap,1bility. "and GDC 28. GDC 27 requires that the reactfrity comrol systems be designe,l to have a combined capability, ill co11j1111ctio11 11'ith poiso11 addition b.,* the emergency core cooling system. of reliably colllrolling reactivity changes w that. 1111der postulated accident comliliims a11d with ap11roprillle margin for st11ck rods. the capability to cool the core is mai111ai11ed. GDC 28 requires that the effects of postula1ed reactivity accitlems neither damage the reactor coola11111ress11re boundary greater than limited local yielding. nor i111pt1ir the ability to cool the core. Per the Westinghouse analytic met/wt!.~ proposed for i111pleme11tario11. /he 1'1/SLB is cmaly-;.ed to de111011strate that Juel damage criteria are smisjied. i11c/11di11g departure from 1111cleate boiling limits and f11el melt limits.

/11 cm/er for the NRC staff to determine whether the MSLB analyses t1ss111*e complit111ce with the requirements of GDC 27 and 28. please discus.1* /ww the MSLB results and acceptance criteria ap11ropriatelr include the effects of TCD. including:

a) The hot zero power and hot full power de11art11refro111 n11cleate boiling mtio. and b) The hot Juli power peak linear heat rare.

Response

As subsequently demonstrated by the responses to part a) and part b) of this request, the SLB analyses contained in Section 2.2.5 of WCAP-17658-NP Revision I are consistent with the requirements of IO CFR Part 50, Appendix A.

GDC 27, ,;Combined reactivity control systems capability."'.and GDC 28.

a) The hot zero poll'er a11d hot full power departure Jro1111111c/eate boiling ratio, and The response to part a) is combined with the response to part b).

b) The hot Juli power peak linear heat mte.

• Per [

]".c there are no changes to the departure from nucleate boiling ratio (DNBR) results for the HZP SLB and HFP SLB accidents due to the effects ofTCD.

While the transient analysis of the system response for the HFP SLB accident is not impacted by the effects of TCD.

the power-to-melt limit (i.e .. peak fuel linear heat generation rate that would cause fuel melting. expressed in terms of [ ]"'") is affected by TCD. Therefore. the powcr-10-mclt lirnit used in the analysis of the HFP SLB transient was recalculated to explicitly consider the effects ofTCD and defined as a I

)"'" The transient analysis statepoints were then used to recalculate the corresponding peak linear heal generation rate observed during the HFP S~B evem. based on upon the upcoming fuel cycle (Cycle 24). for comparison to the revised power-to-melt limit. The results fi.)r the limiting HFP SLB c;ise are summarized in Table 2. The peak fuel linear heat generation rate does not exceed the value that would cause fuel melting.

• • • This record was final approved on 2/14/2019 6:10:00 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 6 of 7 SAP-19-8 NP-Attachment Our Ref: SAP-19-8 For the HFP SLB. the transient analysis of the system response is consistent with that described in Section 2.2.5.2 of WCAP-17658-NP Revision I. The calculation of the peak fuel linear heat generation rate is based upon the upcoming fuel cycle (Cycle 24). As the peak fuel linear heat generation rate can change on a cycle-specific basis. an additional check will be performed as part of the reload process to confirm that the r r*c power-to-melt limit with the effects ofTCD considered continues to be met for the HA> SLB accident until Wolf Creek fully implements the PAD5 models and methods into its licensing basis for the not-LOCA analyses. Once the PAD5 models and methods are fully incorporated into the Wolf Creek licensing basis. the PAD5 analysis will become the analysis of record and the r standard reload process.

r power-to-melt limit will be checked as part of the In addition to the HFP SLB accident. the SLB at power with coincident rod withdrawal (SLB w/RW AP) case was explicitly evaluated. While this case is not presented within the Wolf Creek Updated Safety Analysis Report (USAR). it is analyzed to address the concerns outlined in NRC Information Notice IE-79-22 related to non-safety grade equipment being subjected to an adverse environment from high-energy line breaks inside or outside containment.

The scenario is that a high-energy steamline break could fail the automatic rod control system cabling and/or equipment such that the RCCA banks begin to withdruw from the core simultaneous with the break. Such, a rod withdrawal. combined with the core power increase caused by the steamline break. may lead to a rapid power excursion and a potentially adverse core condition. RCCA bank withdrawal during the steamline break could only occur with the rod control system in the automatic mode. At zero power, the rod control system is in the manual mode, and therefore could not inadvertently withdraw rods du~ to equipment or cabling being exposed to an adverse condition. As a result. the coincident RCCA withdrawal can only be postulated for a steamline break from an at-power initial condition.

The input parameters. assumptions. and acceptance criteria for the transient analysis of the SLB w/RW AP accident are the same as those for the HFP SLB accident (Section 2.2.5.2.1.2 of WCAP-17658-NP Revision I). with the following additions or changes:

]a.~

• Cases were analyzed for both minimum (0 percent) and maximum ( l O percent) steam generator tube plugging (SGTP) conditions.

• Cases were*analyzed assuming both minimum and maximum reactivity feedback coefficients.

corresponding to BOC and EOC core conditions. [

] a.<

• Rod withdrawal was conservatively modeled to begin al the time of the break. with no credit taken for any delay. A constant [ J"' maximum differential rod worth was assumed for control bank D at hot full power <D-bank is the only one that is permitted to be inserted into the core at full power). The rods withdraw until reactor trip is actuated. or until the rods would be fully out of the core.

• • • This record was final approved on 2/14/2019 6:10:00 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 7 of 7 SAP-19-8 NP-Attachment Our Ref: SAP-19-8 Based upon the transient analysis results. the following limiting cases were selecled in Mder to demonstrate thal 1he DNBR and peak fuel linear heat generation rate criteria continue 10 be met:

A. 0% SGTP. maximum reactivity feedback. [

B. 10% SGTP. minimum reactivity feedback. [

Ja.c C. I0% SGTP. minimum reactivity feedback. [

t*c The results for the limiting cases are summarized in Table 3.

The results of the analysis of the SLB w/RW AP accident demonstrate that the DNB design basis continues to be met with the effects ofTCD explicitly included in the analysis. Additionally, the peak fuel linear heat generation rate does not exceed the value that would cause fuel centerline melting.

The analysis of the SLB w/RW AP accident assumed a maximum differential rod worth of [ i3.c to bound the value for the upcoming fuel cycle (Cycle 24). Also, the calculation of the peak fuel linear heat generation rate is based upon the upcoming fuel cycle (Cycle 24). As the maximum differential rod worth and peak fuel linear heat generation rate can change on a cycle-specific basis, additional checks will be performed as part of the reload process to confirm that the maximum differential rod worth limit remains bounding and the [ ye power-to-melt limit with the effects of TCD considered continues to be met for the SLB w/RW AP accident until Wolf Creek fully implements the PADS models and methods into its licensing basis for the not-LOCA analyses.

Once the PADS models and methods are fully incorporated into the Wolf Creek licensing basis, the PADS analysis will become the analysis of record and these limits will be checked as part of the standard reload process.

Table 2: Wolf Creek HFP SLB Results

[

Table 3: Wolf Creek SLB w/RW AP Results a.c

• • • This record was final approved on 2/14/2019 6:10:00 PM. (This statement was added by the PRIME system upon its validation)