ML20065K697

Thermal Hydraulic Compatibility Analysis of Anf High Thermal Performance Fuel for Hb Robinson,Unit 2

ML20065K697
Person / Time
Site:	Robinson
Issue date:	10/31/1990
From:	Tolman B SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
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ML14184A759	List: ML20065K696 ML20065K697
References
ANF-89-165(NP), NUDOCS 9012040099
Download: ML20065K697 (25)

Text

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JQaovanceo nuctenn ruets coneOnarion THERMAL HYDRAULIC COMPATIBILITY ANALYSIS OF ANF HIGH THERMAL PERFORMANCE FUEL FOR H.B. ROBINSON, UNIT 2 OCTOBER 1990 9012onoov9 901 t ;:.t A Siemens Comoany F DR ADOCK OSOOC 1

ADVANCEDNUCLEAR FUELS CORPORATION ANF-89-165(NP)

Issue Date: 10/26/90 THERMAL HYDRAULIC COMPATIBILITY ANALYSIS OF ANF HIGH THERMAL PERFORMANCE FUEL FOR H.B. ROBINSON, UNIT 2 Prepared by:

3.TA w B. L, lolman, Senior Engineer PWR Fuel Engineering Fuel Engineering & Licensing October 1990

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I CUSTOMER OtSCLAIMER IMPORTANT NOTtCE REGARDING CONTENTS AND USE OF THIS

.. DOCUMENT PLEASE RE.AD CAREFULLY Advanced Nuclear Fuels Corporaton's warrantiss and reoresentations con-corneg tna suciect matter of the document are those set forth in tne Agreement between Advanced Nuclear Fuele Corporation and the Castomer pursuant to wnch thie document e issued. Accorcingty, except a.s otherwise expressly prt>

veed m sucn Agreement, neither Advarv:ed Nuclear Fuels Corporaten nor any person actmg on its behalf makes any warranty or representation. exorossed or implied, witn respect to the accuracy, completeness, or userumess of the mfor-mation contained in this cocument. or that the use of any informaton. accaratus, method or process disclosed in this cocument will not intnnge Drwateiy ownea ngnts; or assumes any liacilities witti respect to the use of any mtormation ao-paratus, method or process cisclosed m this document.

The mformation contruned herein is for tne so6e use of Customer, in urcer to avoid smoestment of ngnts of Advanced Nucleat Fuois Corporation in patents or mventons wnch may De ecluded in the information containec in this cocument, the reciosent, by its accootance of this cocument, agrees not to pubben or rnake pubic use (in the patent use of the toren) of sucn information until 50 authortzed m writing Oy Advanceo Nuclear Fueis CorDoration or until aner six (6) months followmg terminaten or exceration of the storesaid Agreement anc any extensen tnereof. uniess otnerwise expressty provicea in tne Agreement. No ngnts or Itcenses in or to any patents are implied Dy the fumisnmg of this cocu.

ment.

ANF 1145 472A 112/071 l

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TABLE OF CONTENTS

1.0 INTRODUCTION

AND

SUMMARY

....................... I 1.1 Introduction . . . . . . . . . . ... . . . . . . . . . . . -, . I

1. Summary ._.......................... 2 2.0 ' THERMAL HYDRAULIC CHARACTERIZATION OF HIGH THERMAL PERFORMANCE (HTP) c

,AND STANDARD MIX 1RG VANE (SMV)-FUEL-ASSEMBLIES . . . . . ... .. . 6

, , . 2.l' Thermal Characterization of HTP and SMV Assemblies . . . . . . . 6 2.2 Pressure Drop Characterization of HTP Assemblies . . . . . . -. . 6 H3.0 XCOBRA-IIIC THERMAL HYDRAULIC ANALYSIS . . - . . . . . . . .... . 10 3.1- XCOBRA-!!IC Core Model .-. . . . . . . . . . . . . . . . . . . . 10 3.2 Peak Fuel Assembly Model . . . , . . . . . . . . . . . . . . . . 12-3.3 Calculated MONBR Values .................... 14

' 14 . 0 ROD. BOWING IMPACT ON MDNBR ............. . . ...... 16 5.0 LEVALUATION.0F LIMITING TRANSIENTS FOR CYCLE'14 . . . . . ... . .. 17

.6,0 ' EVALUATION 0F LIMITING AXIAL POWER PROFILES FOR CYCLE 14 .... .. 18 7.0 . REFERENCES.

.........................,,.,.._,.. ,,, 19 d

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Page 11 , ANF-89-165(NP)

LIST OF TABLES Page Table 1.1. Limiting transient (loss-of load) MDNBR values for HTP and SHV fuel assemblies vs. mixed core configuration. . . . 3

! Table 1.2. MDNBR values for the three most limiting ~ Chapter 15 Events (Cycle 14 core configuration). . . . . . . . . . . . 4 Table i.3, MDNBR values for the four most limiting axial- power.

profiles (loss-of load transient, Cycle 14 core configuration). . . . . . . . . . . . . . . . . . . . . . . 5 Table 2.1. Comparison of SMV and HTP Pressure Loss Coefficients ... 8 LIST OF FIGURES Figure 1. HTP vs SMV grid configuration. . ...............

7 Figure 2. Distribution of pressure loss coefficients for SMV and HTP Robinson fuel assemblies (evaluated at. Re=500,000 for in-reactor values) . . . . . . . . : ............... 9 Figure 3. XCOBRA-IIIC Core Flow Model . . . . . . . . . . . . . . . . . . 11 Figure 4. XCOBRA-I!!C Hot Fuel Assembly Model. ............. 13

, Page 1 ANF-89-165(NP)

1.0 INTRODUCTION

AND

SUMMARY

1.1 Introduction This report documents the results of analyses to - 1) address the thermal hydraulic compatibility of Advanced Nuclear Fuels (ANF) high thermal performance (HTP) fuel assemblies with the co-resident ANF standard mixing vane (SMV) fuel assemblies in the H.B. Robinson, Unit 2 core, 2) evaluate the relative severity of the limiting transients for the Cycle 14 core configuration, and 3) demonstrate that margin exists with this fuel design over a range of possible axial power distributions.

As part of a general thermal hydraulic compatibility analysis the HTP fuel is evaluated relative to

.1. core MONBR,

2. control rod insertion characteristics,

3. control rod guide tube coolability, -

4. core bypass flow,

5. fuel rod bowing, and

6. fuel-centerline temperature.

Because the' guide tube and fuel dimensions for the HTP and SMV assemblies are identical, the above items 2, 3, 4 and 6 will not be impacted. Thus, only the above items 1, and 5, core MDNBR and fuel rod bowing, are evaluated, i

f Page 2 ANF-89-165(NP) 1.2 Summary Results of XCOBRA IIIC calculations performed to estimate the core MDNBR for future mixed configurations of HTP and SHV assemblies are summarized in Table 1.1. The core conditions for the loss of-load transient were used for all cases and the core loading patterns and assembly relative powers for Cycle 14 were utilized. The HTP assembly DNBR values were calculated using the ANFP correlation (Ref. 1) while the SMV DNBRs were calculated using the XNB correlation (Ref. 2). The peak power assembly was assume'd to be at the F$H Technical Specification limit of 1.65. An SMV assembly is the peak power assembly only for the early portion of the first mixed core cycle. Table 1.1 also notes the nominal and mixed core MONBR limits for each correlation.

The calculations indicate the use of HTP assemblies will not result in a core MDNBR below the mixed core, 95/95 correlation limit for either the SMV or HTP assemblies The' results show that a considerable gain in core MDNBR (or F$H) Will be realized for future cycles in which an HTP assembly is the peak power (limiting) assembly.

There is no impact on the MDNBR due to rod bowing.

The MDNBR values for Cycle 14 for the three most limiting Chapter 15 events are summarized in Table 1.2. As seen the loss-of load

• transient remains the limiting transient for both the SMV and HTP assemblies.

The MDNBR values for the four worst case axial power profiles (from previous set point analyses) were evaluated for the loss of load transient. The results are summarized in Table 1.3 and confirm that the current setpoint trip function remains bounding for Cycle 14.

Page 3 ANF-89-165(NP)

Table 1.1. Limiting transient (loss-of-load) MDNBR values for HTP and SMV fuel assemblies vs. mixed core configuration.

MDNBR for MDNBR for SMV Assembiv HTP Assembly All SMV Core First Reload 1/3 HTP Assemblics.

-2/3 SMV Assemblies ,

S1cond Reload 2/3 HTP Assemblies, 1/3 SMV Assemblies All HTP Core I

- _ - - . _ . - - . - - . - - - . _ _ _ _-__--_--- _- -_-- - -_-__._ _ _ - _-_ _ - _ -_ ~ .

. . . _ _ , . . _ _ . . _ _ _ . _ _ _ _ . . . . . . . . .__ ., - . _ .. - . - m _. .

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Tab'leil.2 - . MONBRivalu'es: for. the three:most limiting Chapter 15-  :

Events (Cycle 14: core configuration)'.

s _

SMV- . HTP-

. Assembly (Assembly-Transient' MDNBR (1)- MONBR (2) '

.l Loss-of-loid t

. t 3,,

iContboli rod-l withdrawal-

Control

. rod s drop - -

e

. -?

t.

a n

F j '%~ .

6

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per M ' T = h 8 7 d -T-**'-r ~

'T I - 'e7M *s d?" b E 'w f

Page 5 ANF-89-165(NP)

Table 1.3, MDNBR values for the four most limiting axial power profiles (loss-of-load transient, Cycle 14 core configuration).

SMV HTP Assembly Assembly Transient MONBR (1) MONBR (2)

Axial 1 Axial 2 Axial 3 Axial 4

.x .

i Page 6 ANF-89-165(NP)'

2.0 J THERMAL HYDRAULIC CHARACTERIZATION OF HTP AND SMV FUEL ASSEMBLIES' j Comparisoh of, the spacer configuration for HTP and SMV assemblies is shown in Figure-l1.- Thel HTP : configuration represents a significant design change.

_l

= relative to- the SMV grid configuration. The IFMs result in an abrupt change linL hydraulic, resistance between the HTP and adjacent SMV assemblies which will. -

. affect the. cross flow between assemblies. Also, the - different hydraulic- }

. resistance of therspacer grids will also influence the' assembly cross flow. I The. characteristics- of the-HTP and;SMV fuel assemblies are summarized in the following subsections,

[

-2.l' Thermal Characterization of HTP and SMV Assemblies a

The HTP spacer grids.and IFMs have'been shown in Tables 1.1 and 1,2 to provide

.a significant_ increase in ;the thermal margin -(MONBR) for H. B.1 Robinson Unit

.2.: The: thermal characteristics ' of the HTP assemblies.- have- been extensively-

, , fquantified via critical- heat flux tests which were used to develop the ANFP

0NB- correlation (Ref.11). Thei ANFP correlatjon'.was used to calculate the i

!MONBR forithe: HTP assemblies. The ONBR for ~ the. SMV Lassemblies were evaluated.  :

using the XNB correlation (Ref. :2) assin previous--H. B. Robinson,1 Unit 2. fuel-

_ cycles, s

. I 2.21. Pressure Drop. Ch'aracterization of HTP- Assemblies ~

m 4

ComponentVhydraulic loss coefficients Ofor< the HTP and SMVbassemblies are < -

Lcompared in Table.2.1. "The1HTP data were measuredtinl tests performed in ANFs

, 1 portable loop- hydraulic 1 test Facility -(Ref. 3). . The-SMV assembly spacer : loss  :

U coefficient is t documented in'

Reference:

4.- The loss coefficients are for'the -

l_iquid phase Land? are referenced to the in-reactor, bare rod flowJ area. ' The' ~

Upper _and ilower! tie platelloss coefficientsn include' reversible losses due to -

area _ change: and_-_-losses _due' to_ simulated upper and 11ower. core support i

s; uctures,t The overall.~ assemblyi loss coefficient for the LHTP and - SMV assemblies:are compared in Figure 2.

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Page 7 'ANF-89-165(NP) l 7

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HTP, Spacer s" l'l'll,1 SMV, Spacer Ill lFM, Orsg 11111111

, u

= HTP, Spacer !liff i' 'l"lll SMV, Spacer I

IFM - Grid II Active Fuel HTP, Spacer AM '*"@ N""' SMV, Spacer IFM, Grid IllllWI HTP, Spacer $$ ,'

gyy, 3,,e,,

p  ! .

l HTP, Spacer , ,lFlE 7d SMV, Spacer i.y Non mixing vane spacer V ""

_ Non mixing vane spacet High Thermel- Performance - Standard ' Mixing Vane (HTP) J Assembly (SMV) ' Assembly L Figure 1. HTP vs SMV grid configuration.

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I 1

Page 8 . ANF-89-165(NP) s 1

Table 2.1. Comparison of SMV and HTP Pressure loss Coefficients (values for in-reactor application) >

4

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i-Page 9 ANF-89-165(NP)

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Figure'2. Distribution of' pressure loss coefficients for SMV and HTP

. Robinson -fuel assemblies (evaluated at Re=500,000 for in-i reactor values) k

c. -

L1 , . '

y v +

k g .

n ,i ; q

'Page 10 ANF-89-165(NP) 3.0 XCOBRA-IIIC THERMAL HYDRAULIC ANALYSIS The ANF mixed core methodology (Ref. 5) was used in calculating the 7 core ~MONBR.

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The XCOBRA-IIIC core and hot-fuel assembly models are briefly discussed in the-next two subsections.:

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• tin 3.1 XCOBRA-IIIC Core _Model .

a m:

p The core: flow- distribution -analysis is performed to : assess - crossflow p between assembliesLandL to ' quantify the flow vs axial-distance for each Lfuel assembly. The core : flow distribution analysis is particularly id ~

important for mixed fuel loadings 'where ' hydraulically- different fuel" '

W, typesE are co-resident inL the- core. : The core model- calculates the coolant mass. and momentum ' flux vs axial distance - for subsequent- hot fuel; assembly calculations to determine.the core MONBR. '

n

-The XCOBRA-IIIC core model represents each fuel ' assembly in a -1/8

+

!. symmetric . core sector = of "th; n.ts. Robinson, Unit 2 core as shown in Figure 3. Each' fuel assembly is = modeled as an' individual hydraulic -

.~ channel .

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---.m

e Page 11 ANF-89-165(NP) its Core 1CQ B R A.lllC H.B. Robinson Core M od el A,....,

., 8 .5...,

suv A..emo,,e.

(a) first HTP reload ), '

y ....

(Cycle 14)

io 31  :

A' 7. -

c- p,ein Hrr Assemnae.

NN 'i O Y ...,

12  !

- a e gg s n gg s ,.

8 \'

1.

i, L6 E s11 15 :

i l

c H O F E D B

(b) second HTP reload 1/8 Core XCC8R A IllC A.....

uoaai e

h:'

, h Once Bumed HTP Assemones Suv Assemoties p: Freen HTP AssemDnes si $ #n ' N ,,,

M M\,i u :

,, s s #Mn N

,. ; UW m 'N 15 H G F E D B

Figure 3. XCOBRA-IIIC Core Flow Model

, . - - _ . . . - - = . . - .-. -

Page 12 ANF-89-165(NP)

Cross . flow between adjacent assemblies in the open lattice corr- is directly modeled. The single-phase loss coefficients of Table 21 are ,

used. .

As noted earlier, the calculations utilized the thermal hydraulic boundary conditions associated with the H. B. Robinson, Unit 2 limiting (loss of load)1 transient transient 'and a axial power profile. The peak fuel. assembly _is adjusted to be at the Technical Specification ' limit of 1.65.-

The following core configurations were modeled:

1. all SMV fuel assemblies (for reference) first cycle HTP loading configuration.

, 2 3...

second cycle HTP loading configuration, and 4.- all HTP fuel assemblies.

- . 1 y L3,2 Peak Fuel Assembly Model '

4 y The' XCOBRA-IIIC ~ peak fuel _ . assembly model is a symmetric '1/8 assembly.

, ~ sector: as shown in- Figure 4. Heat, mass, and momentum fluxes between

-the-inter rod flow channels are explicitly calculated. Local values of

-mass' flux andLenthalpy are determined,' and : sed to calculate the DNBR.

Coolant:-mass and. momentum flux vs' axial position calculated from the p core model :are used as boundary conditions to the- outer, vertical boundary lof the. assembly, b

I s

Page 13 ANF-89-165(NP) 3

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] A-A

, 11 Fuel A460m.ly Se.,et 9

Modeled in ICO.aA.440C e.

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c 4 c a=.'

..._...x, , ,, . ,,

, ,s ,. ..

t 25 30 31 2

. .. - i, 2. n ,

, ,a at le at 7 ,, i. ,. n 2.

. u 2. ,, u u 3. , cigg;ig.i C ,,,,, j B Figure 4. XCOBRA-IllC Hot Fuel Assembly Model .

Page 14 ANF-89-165(N9)

The fuel rod relative power distribution from a previous 9 B.

Robinson, Unit 2 analysis (Ref. 6) was used and represents a conservative profile relative to bundle power and flows.

3.3 Calculated MDNBR Values The calculations utilize the core thermal / hydraulic boundary conditions for the loss 'of load (limiting) transient.

The calculated MDNBR values for the mixed core configuration cases considered are summarized in Table 1.1 (included in Section 1.0).

Page 15 ANF-89-165(NP) e a

____- _. _ _ _ _-._ _ _ _ _ _ . _ . - . - - _ _ - - _ _ _ - . _ - _ - _ _ _ - - _ _ _ _ _ . _ ___.--mm- . _ _ _ _ _ _ ____________________m-

Page 16 ANF-89-165(NP) 4.0 R0D B0 WING IMPACT ON MDNBR The impact of rod bowing on the HTP MDNBR values was evaluated using the ANF Rod Bowing Methodology (Ref. 7).

4

Page 1,7 ANF-89-165(NP) 5.0 EVALUATION OF LlHITING TRANSIENTS FOR CYCLE 14 Previous safety analysis work has shown the three most limiting Chapter 15 events are,

- loss of-load control rod withdrawal -

- control rod drop.

XCOBRA-!!!C calculations were performed to evaluate the MONBR values for each of these transients for both the SMV and HTP assemblies for the core configuration of Cycle 14 (Figure 3(a)). The *asults are summarized in Table 1.2 (included in Section 1.0). The calculations indicate that the limiting transient for both fuel types is the loss-of load transient. Also, the relative MONBR values show the same trend for both type assemblies.

Page 18 ANF-89-165(NP) 6.0 EVALUATION OF LIMITING AX1AL POWER PROFILES FOR CYCLE 14 The -axial power profiles are a key parameter in assessing the acceptability of the reactor OTAT set point. Previous set point -

analyses have established the axial power profiles that, together with the limiting transient core boundary conditions, yield the most limiting MDNBR values for the set point evaluation.

Calculations were performed to determine the Cycie 14 limiting transient (loss of load) MDNBR using these limiting axial power profiles for both the SMV and HTP assemblies modeled as the limiting F2H assemoly. The results are summarized in Table 1.3 (included in Section 1) and confirm that the limiting SMV assembly MDNBR values are greater than the, r,ixed core 95/95 MDNBR limit.

The results of the axial power shape sensitivity calculations confirm that the current OTAT trip setpoint will continue to be valid for Cycle 14.

Page 19 ANF-89-165(NP)

7.0 REFERENCES

1. Decarture From Nucleate Boilino Correlation for Hiah Thermal Performance Fuel, ANF 1224(P), May 1989.

2. fygon Nuclear DNB Correlation for PWR Fuel Desians, XN NF-621(P)(A), September 1983.

3. Sinale Phase Hydraulic Flow Test on H.B. Robinson HTP Fuel Assembly, ANF-89-163 (P), October 1989.

4. Sinale Phase Hydrgy.lic Performance of EXXON Nuclear H.B. Robinson Fuel Assembi.y, XN-75 2 Revision 1, April 1975.

5. Aeolication of Exxon Nuclear Comoany PWR Thermal Marcin Methodoloav to Mixed Core Conficuration, XN NF 82-21 (P)(A),

September 1983.

6. H.B. Robinson Unit 2 Chaoter 15 Overtemoerature AT Trio Event Analysis for Elimination of RDT 8voass Picina, ANF 88 094, July 1988.

7. "Comoutational Procedures for Evaluatina Fuel Rod Bowi.pg",

XN 75 32(A), Supplements 1, 2, 3, and 4, October 1983.

i 9

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ANF-89-165(NP)

Issue Date: 10/26/90 THERMAL HYDRAVLIC COMPATIBillTY ANALYSIS OF ANF HIGH THERMAL PERFORMANCE FUEL FOR H.B. ROBINSON, UNIT 2 D1Lir.ihu(1911 HG Shaw/HB Robinson (1/10)

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