ML18029A460

Proposed Tech Spec Changes to Allow Ref to Westinghouse Quad+ Fuel

ML18029A460
Person / Time
Site:	Browns Ferry
Issue date:	04/03/1985
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18029A459	List: ML18029A458 ML18029A460
References
NUDOCS 8504100097
Download: ML18029A460 (28)

Text

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATIONS BROGANS FERRY 'NUCLEAR PLANT (TVA BFNP TS 199 SUPPLEMENT 1)

(

Reference:

'RC letter from D. B. Vassallo to H. G. Parris dated November 26, 1984; qUestion No. 4)

S504l00097 850O03 PDR ADOCK 05000260 P PDR

I LIHITINC hFETY SYSTL'H ETT INC ihVIFTV L r t'. r T L CLhEGING INTEGRITY 2.1 FUEL CLADDING INTEGRITY

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no cor<bfnotfon of loop recircu-c, For and core thcmel larion f lou zetc APE flux scrao trfr

'ooer shall the 120I icttfng be allovcd to exceed of rated thermal pover.

assu<<<c operation (Note< These settings hydraulic design within the basic ther<<<al criteria. These criteria arc c l3.4 kv/ft .

LHGR and HCPR 3.5.k. If virhin lfoits of Spccf 'rficetfnnof rhcsc

'lt is deter<<<incd that tther vfo) . cd unsign critcr.'s fs being shell bc during operation, action 15 c:inures to restore fnitfs ed within prescribed operation within ]fc<ftg Surveillance requfrencnts for APRH scree setpoint are gfven fn specification 4.1.B.

The APRM Rod block trip setting shall be:

S~< (0.66w +02%)

where:

block SRS o Rod percent setting in of rated thermal power (3293 HMt) o Loop recirculat,ion flow rate in percent loop of rated (ratedflow recirculation rate equals 1b/hr) 34.2 x 10e

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B

LIMITING CONDITIONS FOR, OPERATION SURVEILLANCE REOUIREMEFRS 3.5.H Maintenance of Fi11cd Discharge Pipe 4.5.H Maintenance nf Filled Dischar d Pi e

~ he suet inn of the RCIC and HPCI pumps shall be aligned to the condensate Every month prior to the testing of storage tank, and thc pressure suppres- the RHRS (LPCI and Containment Spray) sion chamber head tank shall normally and core spray system, the discharge b>> aligned to serve the discharge piping piping of these systems shall be th< RHR nnd CS pumps. Thc condensate vented from the high point and water head tank may be used to serve the R){R flow determined.

and CS discharpc piping if the PSC head tank is unavailablc. The pressure 2. Following any period where the LPCI indicators on the discharge of the RHR or core spray systems have not and CS pumps shall indicate not less to be operable, the dis-,,

been'equired than listed below. charge piping "of the inoperable sys-Pl-75>>20 48 psig tem shall be vented from the high Pl-75-48 48 psig point prior to the return of the Pl-74-51 48 psig system to service.

P1-74-65 48 psig

3. Whenever the HPCI or RCIC system is Averace Planar Linear Heat Generation lined up to take suction from the Rat e condensate storage tank, the dis-Durfnc steady state power onerntion. the charge pfping of thc HPCI and RCIC

.">ximum Ave rage Planar Lhnenr Heat Gen- shall he'*vcntrd from the high point era(. ion Rate (MAPLHGR) for each type of of the system and water flow observed

uel as,> function of average planar nn a 'monthly basis.

exposure shall not exceed the limiting value slioin> tn Tables 3.5.I-1, -2 . 4. the RHRS nnd the CSS are re-it't any time during operation is determined by normal surveillance that it When quired to be opcrablc, thc pressure indicators which monitor the dis>>

the limiting value for APLHGR is being charge lines shall bc monitored exceeded, action shall be initiated daily and the pressure recorded.

within 15 minutes to restore operation tn within the nrescrihnd limits. If the APLHGR is not returned 'to within Maximum Avcra e Planar Linear Heat the prescribed Iimits within two (2) Generation Rate (MAPLHGR) hnurs. thc reactor shall be brought to The MAPLHGR for each type of fuel as a the Cold Shutdown condition with a function of average planar exposure 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding shall be determined daily during action shall continue until reactor reactor operation at ~ 25X rated operation is within the prescribed thermal power.

limits.

J. Linear Heat Generation Rate (L'HGR)

.Linear Heat Generation Rate (LHGR) The LHGR During steady stare power operation, the i shall be checked daily during linear heat generation rate (LHGR) of reactor fuel operation at ? 25X rated any rod in any fuel assembly at any thermal power.

axial location shall not exceed 13.4 kw/ft.

If ac -nny rime during operation it is

• determined by normal surveillance that the limiting value for LHGR ia being exceeded, action shall be initiated within 15 minutes to restore operation to within th>> prescribed limits. If the LHGR ia. not returned to within the prescribed limits within two (2) hours, rhe reactor shall bc brouFht to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance aud corresponding action shall continue until reactor operation is within the prescribed limits.

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3.5.J. Linear Heat 'Ge..ora ion Rate (LHGR)

This specif'ca='on assures tha" the linear heat generation rate in any rod ls Less than the design L'near heat genera ion postuLated.

if fuel pellet densificaticn is The LHGR sha'L be checked daily during reactor cper ation at a 25~~

powe. to dete. mire if fuel burnup, or control rod movemen has caused changes in power dis ri"ution. For LHGR to be a .Limiting value below 25> rated thermal power, th R factor ~ould have to be Less than 0.241 which is precluded by a ccnsiderable mar gin when employing any permissible control rod pat em.

3 ~ 5 ~ go Ninimimum Crcal Power Ratio (YC. H)

At core ther" al power levels Less than ot equaL to 25~+, the reactor wilL be operating at "in~~urn recirculation pump speed -ard the moderator void content wiLL be very "aLl For all designated ccntroL rod pat erns, which may be employed at this po'nt, operating plan exper'ence and the. ma'ydraulic analysis ind'ca ed that the result'ng MPCR value is 'n excess of requirements by a considerable margin. With th's low void content, any inadvertent core flow increa e -ould only place operation in a more conserative mode relative to HCPR. The "'a'y requirement ".or caLculating NC?R above 25> rated is suff'c'ent since power distribution shifts are very slow when there therma'o~er have not beer. s'gnificant power or control rod changes. The requirement for calculating ".!C?R ~hen a limiting control rod pattern is approached ensures that HC?R wiLL be known folLowing a change in power or power shape (r gardless of .:agnitude) that could place operation at a thermaL L'm'.

3.5.L APRN Setooints Operat'on 's constrained to a maximum LHGR of This limit is reached when core max'-m fraction of limiting power density (CFZLPD) equa's 1.0. For the ase'where CFZLPD exceeds the fraction of rated the."mal power, o era ion is permi ted only at less than 100-percen" ra ed power and cnly with A?HE scram sett'ngs as requ'red by spec' ication 3.5.L.1.The scram tr'p setting and rod blcck trip setting are ad'usted to ensut e that no comb.'nat'on of C."ZLPD and "."HP w'l'ncr as =he LHGR transien" peak beyond tha allowed by the 1-percent p as.'c strain limit. A 6-hour time per'od to achieve th s condi:icn 's 'ustified since he add'tioral margin gained b the setdown ad'ustment is above and beycnd tha'nsure" by the safety ara's'. k 16c

II I TIDDLE 3. S. 1- 1

)'O'L)ICR V"RSUS AVERACE PLANAR EXI'OSURE Fuel Truer'- PSDRB284L and QUAD+

Average Planar Exooaure MAPLHCR

(.'.".4d Ic ) (kMIEc) 2no 11.2 1,000 11 ~ 3 5, 00() 11 '

I 0,000 12. <)

15,000 12.0 20,000 11 ~ 8 2S,OOO 11. 2

)n,non 10. 8

)5.000 10. 0 cn,non 9.4 Table 3.5.I- 2 HAPLIIGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Typic 1 P81)k826511 Aver:));e Plounr Expo sure MAPLHGR (HMd/t) (IEw/ f t) 200 11.5 1,000 11. 6 5,000 1 l. 9 10,000 12.1 15,000 12.1 20,000 12.0 25,000 11.6 30,000 I I.2 35,000 10. 9 40,000 10. 5 45,000 10. 0

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Crt Figure 3.5.K-I MCPR Limits for PS X BR and OU~D+

-172-

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S.e KAJAR nCSICV FEATURES 5,1 sITK FL'ATURL's Brovns Ferry unit 2 is located at Brovns Ferry Nuclear Plant

~ Ite on property ovncd by che United States and in custody of cha TVA. The ~ ice aha ll consisc of approxfmatcly S40 acres on che north shore of <Ihecler Lake at Tennessee Rfver Ifle 294 In Limestone County, Alabama. The efnlnv~ distance f ron the outside of, che secondary concafnmenc building to che boundary of che exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet.

5.2 REACTOR A. The reactor core may contain 764 fuel assemblies consisring 4 QUAD+ demonstration assemblies, Sxg assemhlies having 63 fuel rods each, and SxSR and PSxSR assemblies having 62 fuel rod s each.

B. The reactor core shall conrain 185 cruciform-shaped control rods. The control material ahall be boron carbide powoer (BIC) compacted to approxissacely 70 percenr. of theoretical density.

S.3 REACTOR VESSCL The reactor vessel shall be as described In Table 4.2-2 of che TSAR. The applicable des iRn codes shall be as described in Table 4.2-1 of the FSAR.

'. 4 COIITA I MEIIT A. The principal deafen parameters for the primary concafn~enc shall be aa given in Table 5.2-1 of che FSAR The applicable

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dcsirn codes shall be as described In Section 5.2 of the FSAR.

b. The secondary contafnmenc shall be aa de c~ibed in Section 5.3 of che FSAR.

C. Penetrationa Co the prfeary contafnmenc and pfpine Passfng through such penecratfons shall be designed in accordance vich che standards sec forth fn Section 5.2.3.4 of the FSAR.

S.S FUEL STORAOC A. The arranaenenc of fvel fn che nev-lvef scoraac facflfcy shall bc such chac k f, for dry condf C fons, ia less chan 0.90 and flooded is Iesa chan 0.95 (Sectfon 10.2 of FSAR).

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ENCLOSURE 2 RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION (D. B. VASSALLO'S LETTER TO H. G. PARRIS DATED NOVEMBER 26, 1984)

BROWNS FERRY NUCLEAR PLANT UNIT 2 Question 1 You have referenced WCAP-10507 as a supportihg document with respect to the demonstration assemblies. However, this report has not been formally

'presented (docketed) for staff review. In order that we may base Safety Evaluation conclusions on such a report, it must be formally submitted and attested to. Therefore, please submit this document on your docket.

Since there is some confusion as to which version is correct, please submit either the correct version or the altered pages relative to the version dated March 1984 and marked "Preliminary Copy."

~Res ones By letter from J. A. Domer to H. R. Denton dated November 13, 1984, we stated our plans for loading four Westinghouse BWR design fuel assemblies into unit 2 as part of reload 5. ,In that letter we referenced the appr opriate letters from Westinghouse to NRC which submitted the topical report. In, the November 13, 1984 letter we requested that the Westinghouse letters and their enclosures be included by reference on the Browns Ferry unit 2 docket.

Question 2 h

Please provide the cycle specific value of the power margin of the demonstration assembly compared to thc limiting assembly. Also, please confirm that a linear relationship exists between bundle power and CPR.

Also, demonstrate that the margin conservatively accounts for nonconservative transient delta CPR and other nonconservatisms such as high void fraction in the channel for the demonstration assembly.

~Res onse 2 For Browns Ferry unit 2, cycle 6 (BF2CY6), all fuel in the core will be, monitored to the same OLMCPR by the process computer. The OLMCPR is based on analyses performed for the General Electric (GE) Pgx8R fuel type.

Because the,process computer is not capable of adequately evaluating CPR for the Quad+ domo assemblies, CPR safety margin will be assured by loading these assemblies into low power positions in the core.

Cycle specific steady state and transient analyses were performed to evaluate the Quad+ CPR margin that exists due to the core loading pattern.

Figure 1 shows a comparison of bundle power for the Quad+ demonstration and core limiting assemblies for the expected operation of BF2CY6. A similar comparison for MCPR is provided in figure 2. At the minimum power margin, the Quad+ assembly power is 83.3 percent of the limiting assembly power.

The difference between the limiting assembly MCPR and the Quad+ MCPR is 0.36 at the minimum power margin. The minimum MCPR margin of 0.35 occurs at a slightly higher exposure than when the minimum power margin occurs due to local peaking factor variations. If the limiting GE assembly vere operating at the OLMCPR (1.26) at the time of the minimum,CPR margin, the Quad+, MCPR (from AA74) would be 1.60. This MCPR margin of 0.34 is more than enough to offset the higher Quad+ transient ACPR ('.05) and any nonconservatisms in evaluating Quad+ CPR. The transient hot-channel analysis explicitly modeled the thermal-hydraulic offccts related to, the Quad+ flow distributions and higher void fractions. The four Quad+

assemblies were considered to have a negligible effect on core-wide neutronic calculations. Additional transient margin is available to the demonstration assemblies because the pressurization events that determine the OLMCPR are most severe at end-of-cycle exposures. As can be scen in figure 2, the MCPR margin of the Quad+ assemblies relative to the limiting asscmblics is approximately 50 percent greater near the end of cycle.

The relationship between CPR and bundle power is shown in figure 3 for the Quad+ assembly. While the relationship is not linear, it is continuous and approximately linear for moderate power changes (<10 percent).

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Questioa 3 Even though the values of the uncertainties in thc various input parameters to the two different CPR correlations (i.e., GEXL and AA-74) are the same, the form of the correlations is different and may result in a different safety limit CPR for the two correlations. Therefore either justify the use of a safety limit value of 1.07 for both assembly types or provide an estimation of the uncertainty in the assumption'nd show that sufficient margin exists to account for it.

~Res oose 3 The response of the different correlations to the same input parameters may differ duc to the nature of the functional forms of the correlations; however, the net effects of the functional differences on the safety limit will be very small after the statistical convolution of all'f the uncertainties associated, with the CPR evaluation. This is confirmed by comparing the 1.06 safety limit calculated by ASEA-ATOM for the AA74 correlation to thc 1.07 safety limit calculated by GB for thc GEXL correlation. Thc AA74 safety limit applicable for the Quad+ assemblies is expected to be= close to the value calculated by ASEA-ATOM. Therefore,<< the Quad+ assembly safety limit <<ill be close to 1.07 and will be confirmed when the Quad+ assembly safety limit study is completed.

A large CPR margin is inherent in the Quad+ demonstration assemblies in thc BF2CY6 reloa'd core due to loading the Quad+ assemblies in low power core positions as indicated in the response to the previous question. This large margin will more than compensate for any possible nonconservatism which may exist in using the 1.07 value as the Quad+ CPR safety limit.

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FIGURE I BUNDLE PONER CONPRRISON 1.6 1 d o~e 0 1.3 C)

Q g~Q 6 0.9 O LID>TING BUNDLE 8 0~HO BUNDLE 0.8 0 3 5 6 CYCLE EXPOSURE (GHDll1T)

F'IGUPiE 2 NCPPi CQNPRRISON 2.2 LIMITING BUNDLE Q DE/10 BUNDLE g-Q 2-Q Q g Q g Q g m >

1.6 1.4 9

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QPERRTINL- LIMIT 1.2 0 2 3 5 . 6 8 CYCLE EXPOSURE (GHD!NT)

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FIGURE 3 QUAD+ AA74 CPR AS A FUNCTION OF ASSEMBLY RELATIVE POWER 26

)I 24 22 t

u 1.8 1.6 1.4 1.2 0.9 1.1 . 1.2 1.3 1.4 1.6

. ASSEMSLY RELATlVE POM/ER

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ENCLOSURE 3 D I The below listed changes are described as they relate to T.S. 199. All changes are to allow reference to Westinghouse Quad+ fuel as requested by NRC.

1. Section 2.1.A. 1.C (pg 9) Change to make LHGR of 13.4 KM/ft apply to all fuel types (additional, page).

2. 'ection 4.5.J (pg 159) - Change to make LHGR apply to all fuel types (additional change to pg 159).

3. Section 3.5.L Basis (pg 169) - Change to make LHGR of 13 .4 KH/ft apply to all fuel types (page affected by T.S. 167).

4. Table 3.5.I-1 (pg 171) - Change to make Table 3.5.1-1 applicable to Quad+ fuel (additional change to pg '171).

5. Figure 3.5.K-1 (pg 172) Change to make Figure 3.5.K-1 applicable to Quad+ fuel (additional change to pg 172).

6. Section 5.2.A (pg 330) Change to allow 4 Quad+ fuel assemblies in the reactor core (additional page).

c Four Westinghouse Quad+ demonstration assemblies will be" utilized in the unit 2 cycle 6'core reload. Per NRC request, additional changes are being proposed to reference Quad+ fuel in T.S. 199.

The four Westinghouse design fuel assemblies to be loaded into Browns Ferry unit 2 are described in WCAP-10507, forwarded to NRC by Westinghouse letters dated June 5 and July 20, 1984. The assemblies are being loaded for demonstration purposes in core locations with sufficient power margin that thermal limits annlicabla. tn General Electric POxSR fuel are bounding by a substantial margin.

The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (48FR14870, April 6, 1983). One of the examples (vi) of changes not likely to involve a signifi-cant hazards consideration relates to changes which may in some way increase the probability or consequences of a previously-analyzed accident, but where the results of the change are clearly within all acceptable criteria with respect to the specific system or component specified in the Standard Review Plan. Since these assemblies are being loaded into sufficiently low power locations in the core the overall safety margin. is within acceptance criteria.

Since thc results of the changes are within NRC acceptance criteria, TVA proposes to determine that the proposed changes would not involve a significant, hazards consideration determination in that they: (1) do not involve a significant increase inthe probability or consequences of a previously evaluated accident; (2) do not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) do not involve a significant reduction in a margin of safety.

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