ML19326C172

Forwards Info Re Physics Startup tests,BAW-1433 Reload Rept & Proposed Tech Specs Required for Cycle 2 Operation in Response to NRC 761230 & 770113 Requests

ML19326C172
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/07/1977
From:	Rueter D ARKANSAS POWER & LIGHT CO.
To:	Ziemann D Office of Nuclear Reactor Regulation
References
1-027-2, 1-27-2, NUDOCS 8004210668
Download: ML19326C172 (18)

Text

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NRC FrRu i95 U.O. NUCLEAR KEGULATo AY .v.C'ON k DOCKET NUMEED i,. 7s .- 50-313 NRC DISTRIBUTION FoR PART 50 DOCKET MATERIAL TO:., FROM: DATE oF JoCUMENT Arkansas Power & Light Company 2 7/77 --

Mr. D. L. Ziemann Little Rock, Arkansas DATE FiCE!VED Mr. Donald A. Rueter J/8/77 MLETTER ONoToRIZED PROP INPUT FORM NUMBER oF COPIES RECEIVED R INAL QUNC LASSIFIE D .

One signed DE SC RIPTf oN .

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Ltr re our 12/30/76 and 1/21/77 ltrs, and their 1/13/77 ltr...trans the following: Consists of requested information l concerning their proposed tech specs as affected -by the B&W Rod B.o( Model. ,

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PLANT NAME: I Arkansas Unit No. 1 SAFETY FOR ACTION /INFORMATION FwTun 2/9/77 RJL ASSIGNED AD: __

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,BEht!CH CHIEF- "Ziemann (f ) RDANEW FMTFF.

yEIrAT.ECT y,1 NAGER: Snaider PROJECT MANAGER:

J.IC. ASST. R. Diggs LIC. ASST. I INTERNAL DISTRIBUT'ON i FWG FILD SYSTEMS SAFETY PLANT SYSTEMS SITE SAFETY &  ;

X~MRC PDR _ HEINEMAN TEDESCO ENVIRO ANALYSIS f MI&E(p) SCHROEDER BENAROYA DENTON & MUT.T F9,  !

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IIIPC MACARRY KIRKWOOD ERNST CASE KNICHT BALLARD HANAUER SIHWEIL OPERATING REACTORS SPANGLER ltARLESS PAWLICKI STELLO SITE TECH.

PROJECT MANAGEMENT '

REACTOR SAFETY OPERATING TECH. CAMMILL BOYD ROSS M EISENHUT [ h .D '

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b H ELPIN G BUILO ARKANSA9 ARK ANS AS POWER & LIGHT COMPANY A0 SQx 551 LttrLe AOCK. A AK ANSAS 72203.(501)371-4000 February 7,1977 OLLQr Qt 'db I 6 pa a $7 v.n 1-027-2 hc + v.s #

Director of Nuclear Reactor Regulation

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ATTN: Mr. D. L. Ziemann, Chief / b Operating Reactors Branch #2 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Arkansas Nuclear One-Unit 1 Docket No. 20 000 go.3/3 License No. DPR-51 Proposed Technical Specifications (File: 1511)

Gentlemen:

Your letter of December 30, 1976, request:'d that we provide a list of thermal margin credits applicable to Arkansas Nuclear One-Unit 1 (ANO-1) .

Your January 21, 1977 letter requested additional infomation pertaining to the physics startup tests, the BAW-1433, Reload Report, and the pro-posed technical specifications as required for cycle 2. operation of ANO-1. Your letter also indicated that additional infomation may be required to review our January 13, 1977 submittal concerning the B6W proposed Rod Bow 51odel. Via telecopy on January 26, 1977, we infomally '

received three questions concerning our proposed technical specifications as affected by the B6W Rod Bow Model. Attached find the requested infor-mation.

Very truly yours, R -6 e J nald A. Rueter ,

Manager, Licensing _ _

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D^ccmbsr 30, 1976 lotter L. Ziemann (NRC) to J. D Phi.llies (AP&L) ,Page 1 of

,The following test are planned.for startup physics testing following the Cycle 2 refueling.

1. Pre-Critical a) Control Rod Drive Trip Test, OP 1304.35 *

2. Zero Power a) Determination of All Rods Out Critical Boron Concentration, OP 1302.07 b) , Determination of Reactivity Coefficients, OP 1302,06 (Moderator Temperature Coefficient only),

c) , Control Rod Reactivity Worth Measurements, OP 1302.08 dl , Ejected Rod Worth Measurement, OP 1302.10

3. Power Ascension a) Core Power Distribution Test,0P 1302.05 at approximately 40%, 75% and 100% of rated power.

b) Power Imbalance Detector Correlation Test and -

' Calibration, OP 1302.04 @ ~75% FP. '

c) . Determination of Reactivity Coefficients, OP 1302.06 @ approximately 100% FP power - -

doppler' coefficient and moderator temperature coefficient measurements.

la. Control Rod Drive Trio Test, OP 1304.35 Initial RCS conditions are established at a temperature of approximately 532 F, a pressure of 21551 30 psig, all (4) reactor coolant pumps running, with Boron at refueling concentration. Control rod groups 1 thru 7 are fully withdrawn and group 8 (APSR's) are withdrawn approximately 25%. The control rod drive' mechanisms are then tripped via the manual. trip button. The insertion times for each CRDM from its initial position to its 3/4 insertion point is measured by the plant computer Rod Drop Timer, program. The printout of this program includes trip initiation time, initial position and-trip insertion time for each CRDM (excluding group 8).,

Pago~2 of 6 The acceptance criteria are that the measured time from trip initiation to 3/4 insertion shall not exceed 1.66 seconds at full Reactor Coolant Flow conditions or 1.40 seconds for no flow conditions.

The safety analysis calculations are based on a rod drop time of 1.40 seconds from fully withdrawn .

to 2/3 inserted. The measurement utilizes the 25% wd zone suitch on each CRDM since no 33% wd switch is available.

, 2a. Determination of All Rods Out Critical Boron- '

Concentration, OP 1302.07 The-boron concentration for criticality with all rods out except Group 7 at approximately 85% wd is estimated per OP 1103.15, Reactivity Balance Calculation. The control rods are. withdrawn to the "all rods out" position and the estimated amount of DI Water to achieve criticality is added using continuous feed and bleed. A 1/M plot (using source range instrumentation) ,versus boron concentration is maintained and the critical boron concentration is projected as the approach to criticality proceeds. When the boron con-centration nears the projected critical concentration, the letdown flow rate is reduced and the boron sampling frequency is increased. When criticality is achieved, deboration is terminated and the control rods are ~

withdrawn slightly to establish a positive startup rate. Power is leveled off at 10-Y amps on the intermediate range and the boron concentration is allowed to come.to equilibrium. Equilibrium boron .

concentration is verified by sampling the RCS, MU Tank and Pressurizer. The remaining reactivity held in the inserted portion of Group 7 is then measured by withdrawal of Group 7 to its out limit and concurrent ,

doubling time measurements. The doubling time is -

converted to reactivity and the reactivity to equivalent boron concentration change using the boron differential worth. The All Rods Out boron concentration is the sum of the measured boron concentration and the equivalent boron from the reactivity measurement during rod withdrawal with appropriate corrections for Xenon and Samarium concentrations at the time of the measure-ment.

The acceptance criteria placed on critical boron concentration is that the actual boron concentration must be within 1100 ppm of the value predicted by the Physics Test Manual.

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Pago 3 of 6' 2b. Determination of Reactivity Coe f fi cien ts , OP 1302.06 (Temperature Coefficient Oniv)

The moderator temperature coefficient at hot zero power is measured by two methods. In both methods the first step is to achieve steady state critical conditions: 536*- 40F and 10-9 amps. The first method uses the Reactivity Calculator to measure

• reactivity changes as Tave is varied approximately 50F by adjusting the turbine header pressure set-point. The second method uses control rod worth curves measured per OP 1302.08, Control Rod Reactivity Worth Measurements, to determine the reactivity changes by relating rod motion to reactivity.

The acceptance criteria state that the measured value shall notTest Physics differ frombythe Manual predicted more value fgom the than 1 0.4 x 10- AK/K F. -

This value is conservative when compared to the values used in the Physics Test Manual at test conditions (1200-1300 ppm Boron) are less than the safety analysis value by more' than 0.4 x 10-4 oK/K/0F. If the measured value exceeds the predicted value by more than 10.4 x 10-4 4K/K/0 F, an additional evaluation will be performed by B & W. The measured moderator

-temperature coefficient is extrapolated to 95% FP per Technical Specification 3.1.7 and this value is verifiedtobelessghantheacceptancecriteria limit of +0.5 x 10- 4K/K/ F (the value is used in the FSAR for accident analysis). -

2c. Control Rod Reactivity Worth Measurements OP 1302.08 The procedure for measuring control rod group worths is to deborate CRA groups.7, 6 and 5 into the core following initial criticality with all rods out.

Reactivity is measured with an on-line Reactivity calculator during discrete changes in CRA group .

movements. The summation of these changes in reactivity are calculated to obtain integral rod Norths.

Reference to the acceptance criteria of OP 1302.08 finds maximum deviations of.115% for group worths, and ~

110% for total rod worth. The source of these criteria is B & U Doc. No. 62-1000734-01, Zero Power Physics Test Specification. Based on the results of previous measurements, following initial fuel load and control rod interchange, it is not expected that these limits will be exceeded. In both instances, the measured worths varied by less than the maximum permissible deviation.

-However, in the unlikely event that the limits are exceeded, the results will be reviewed with a decision either to repeat the test or ac'ept c the measured values. B &-W will be consulted and their concurrence sought on the final results prior to power escalation if the measured worths fail to meet acceptance limits.

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Pega 4 of 6 The above^ acceptance criteria are established to verify the accuracy of the calculational model used in predicting CRA group worths, thereby verifying agreement with input parameters for the safety analysis.

The criteria used for establishing the maximum per-missible deviation for the group reactivity worth .

tests is that the total measurea CRA worth is com-patible with the calculated worth used in the shut-down margin calculations. The calculational error-allowance is 110% of the total CRA worth. Since all eight CRA group worths are not measured, the 110%

deviation ' limit is applied to the sum of the group worths that are measured. The measurement aids in confirmation of the calculational model and verifies the ability to conservatively predict shutdown margin.

2d. Ejected Rod Worth Measurement OP 1302.10 Following the measurement of control rod groups 5, 6, 7 and 8 by deboration, the maximum worth ejected rod (a group 6 rod) ,is borated out of the core and the reactivity worth is determined from the change in boron concentration required using the boron differential reactivity worth measured during the control rod group worth measurement. After the ejected rod has been borated to 100% wd.and equilibrium boron has been established, the ejected rod is then swapped in to 0% wd versus CRA group 5 and the reactivity worth is determined from the change in CRA group 5 positicn.

The acceptance criteria is, that the deviation between the predicted and measured ejected rod worth.does not exceed 20%. The ejected rod worth is measured at zero power because the accuracy of the measurement is much greater at critical conditions below the point of sensible nuclear heat. Safety Analysis has shown that since the ejected rod is worth more at zero power -

than at power and a linear correlation exists between these conditions, the acceptance criteria will be met at power if met at zero power. The predicted value of maximum ejected rod worth for Cycle 2 is 0.31% wd which is considerably lower than the value used by the Safety Analysis (0.65% wd).

3a. Core Power Distribution. OP 1302.05

. Cas e Power Distribution data will be obtained at various power levels and compared with predicted data to assure compliance with operating limits and Technical Specifications. Power Imbalance,0uadrant Power Tilt, Linear-Heat Rate, DNBR, and Power Peaking Factors will be analyzed. For this test, 40% will mean 40% +for - 2% FP, 75% will mean 75% + or - 2% FP and 100% will mean highest attainable power without exceeding 100% FP. Equilibrium Xenon will be defined as a condition where reactivity change is less than

%s .01% K/K per hour with all other reactivity -change contributions stable. Equilibrium Xenon will not be required for.the 40% tests. Control rod index is

Pass 5 of 6 established at a position corresponding to the rod positions where core power distribution predictions were' calculated.

The acceptance criteria are as follows:

i) ,The maximum linear heat rate in the core is less than the LOCA limit per Technical Specifications .

for the axial location of the peak. When testing at a power level below rated power, the maximum LHR when extrapolated to rated power must also meet this criteria.

ii) ,The minimum DNBR must be greater than 1.30 at rated power conditions and when extrapolated to rated power conditions from a lesser test plateau.

iii) ,The quadrant power tilt must not exceed the value allowed in the Technical Specifications.

iv) ,The highest measured radial and total power peaking factors shall not exceed the highest predicted peaks by more than 5% and 7.5% respectively.

These acceptance criteria are established to verify that core nuclear and thermal hydraulic calculational models are conservative with respect to measured conditions thereby verifying the acceptability of data from these models for input to safety analysis. .

The~ acceptance criteria also serve to verify safe operating conditions at each test plateau and eventually at rated power conditions.

3b) , Power Imbalance Detector Correlation Test and Calibration, OP 1302.04

- Core power imbalance measurements will be made at prescribed outcore imbalance indications while main- '

taining total reactor power constant and below the power level cut off. At each imbalance attained, data will be obtained from the incore detectors, outcore detectors and incore backup recorder. The results will then be analyzed to determine the correlation between the incore and outcore detectors and to determine the adequacy of the outcore detectors to indicate the core power imbalance. If required, the.outcore imbalance indication will be recalibrated and the imbalance measurements repeated as a calibration check.

The acceptance criteria is'that the out-of-core to incore detector relation for imbalance detection has a slope B1.0, causing the imbalance indicated by outcore detectors to always be equal to or more conservative than incore detectors.

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4. ,,

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,. s Page 6 of 6 Safety analysis establishes the RPS trip envelope ,

based on calculated data corresponding to measurements with incore detectors. The RPS trip is generated by the out-of-core detectors. The test procedure ensures that the out-of-core detectors are always conservative with respect to the incore detectors.

3c) , Determination of Reactivity Coefficients, OP 1302.06 The moderator temperature coefficient at power operating conditions is measured by varying Tave using Tave set-point controller on the Reactor Demand Station and maintaining constant power with the ICS in full auto.

The corresponding control rod motion is related to the reactivity change to determine the temperature coefficient. The power doppler coefficient is measured by varying Reactor Power using the ICS Unit Load Demand Station and recording the corresponding control rod motion. The reactivity change due to the control. cod motion is determined and from that the power doppler coefficient is determined.

The acceptance criteria placed on this test are:

i) , That the moderator coefficient be non-positive at power levels greater than 95% FP. This is Conservative with respect to ghe value used by safety analysis of 0.0% AK/K/ F at 100% FP.

ii) That the power do

' than -0.55 x 10~"ppler coefficient 4K/K/%FP. This be more value negative based on a minimum conservative value derived from the minimum doppler coefficient and fuel temperature -

versus power relationship used in the safety analysis. The accident analyses which are power doppler coefficient dependent are power increasing transients such as rod ejection. The above acceptance limit was found to be conservative with respect to measured values during beginning of Cycle 1 testing and calculations indicate that Cycle 2 power doppler coefficient values will be more negative tha,n those for Cycle 1.

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B. BAW-1433,- RELOAD REPORT

1. Discuss the implied conservatisms for the larger densified fuel stack; 1.e.141.12 inches (Page 6.2) .

RESPONSE

The initial cycle 2 reload report submittal, which did not explicitly provide coverage of the fuel rod bow penalty, included a densification power spike penalty. The densification power spike was conservatively based upon an initial fuel density of 92.5% TD and the corresponding densified fuel stack height of 141.12 inches. For the revised submittal, in which coverage of the rod bow penalty is demonstrated, the densified length of Batch 4 fuel (140.49 inches) was used to determine the effect of the reduced active length (relative to the " nominal" 144 inches assumed for undensified fuel). In DNBR analyses the use of a shorter fuel stack height results in an increased average heat flux and, consequently, a reduced DNBR.

2. Provide the DNBR sensitivity study for flux / flow setpoint inclusive of the minimum acceptable DNBR of 1.3 (Pages 6-3 and 6-4). List the DNBR at which the analyses were conducted.

RESPONSE

The attached figure shows the results of the DNBR sensitivity study performed in support of the flux / flow setpoint. This analysis was performed for densified fuel (without the densification power spike) and included the flow area reduction factor, which has been shown to be equivalent to a 1% DNBR penalty. The fuel rod bow penalty of 4.9%

(5.9% less the 1% FA credit) was incorporated by increasing the minimum acceptable DNBR during this transient from 130 to 1.365. By cross-plotting the minimum DNBR occurring during the transient against the corresponding assuned trip setpoints, a maximum allowable trip setpoint of 1.093 is obtained.

3. For the cold water accident, _ the report quoted an initial power of 50% of rated for 2 pump operation (page 7-3) as compared with the maximum permissible power level with 2 pump operation of 52.6% of -

rated power level (page 8-6) . Provide the analysis of accident initiation from the maximum allowabic setpoint of 52.6% thermal power level.

RESPONSE

The text. in the Reload Report should say that the accident analysis in the FSAR was done at 60*6 of rated power. .

4. Provide the analysis which incorporates the effect of fuel rod bow on power peaking and core parameters (Page 8-11).

RESPONSE

The fuel rod bow phenomenon is incorporated in the RPS analysis for

- centerline fuel melt as a peaking penalty. This penalty is a function of ass embly burnup. The same procedure is employed for the DOCA Tech.

Specs. The effects of fuel rod bow are included in the RPS analysis for DNBR as discussed in the response to Questions B.2. and C.1.

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4 C. TECHNICAL SPECIFICATIONS

1. Provide the analysis which was used to establish figure 2.1.3 and discuss this analysis in detail. Your discussion should include:

a. The reasons for identical three and four pump operational curves for core protection safety limits on reactor coolant pressure and temperature; and

~b. The means of incorporation of the fuel rod bow DNBR penalty (Figure 2.1. 3) . Specifically, was this penalty imposed on all modes of operation (i.e. ; 2, 3, and 4 pump operational modes)?

Discuss how the TEMP and CHATAH codes are used in the analysis.

RESPONSE

1. Pressure-temperature limit curves shown in Figure 2.1-3 of the ANO-Unit.1 Tech Spec (and Figure 8-3 of BAW-1433) are derived in the following manner:

For each operating condition to be analyzed (four pump, three pump, and two pump operation) the CHATA code is used to analy:e the hot subchannel . Starting with the " maximum design" point corresponding to the maximum achievable power level for the condition to be analyzed (112% power for four pump operation), and at the limits of the control band for reactor coolant temperature and pressure (plus undetectable errors), the reactor coolant inlet temperature and system pressure are systematically varied to obtain a locus of pressure-temperature combinations for which the minimum DNBR is equal to the limiting value, or the coolant quality at the point of minimum DNBR is equal to the CHF correlation limit (whichever is more limiting). The CHATA analysis considers two closed chan-nels in parallel. The first channel represents an average subchannel (radial x local peaking factor = 1.0, no hot channel factors); the second represents the hot subchannel (R x L = 1.783, hot channel factors) . By proper selection of the average channel flow rate and the use of an enthalpy rise factor (FLAH) in the hot channel, the CHATA analysis duplicates the results of a TEMP subchannel analysis for the reference ,

. condition (112%' max. design point for four pump operation) including both the hot channel flow and MDNBP.. For this (CUATA) analysis, the average channel flow rate (called the reference flow) is input; the resulting = pressure drop calculated for this channel is then imposed on the hot channel and the hot channel flow rate and MDNBR are calculated.

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After CHATA is matched to the reference TEMP analysis, CHATA is then used to_ determine the effects of parametric variations in pressure and temper-ature, . The densification penalty (5.93%) for this analysis is input by increasing the effective hot channel radial x local peaking factor, thus reducing. the calculated MDNBR by the desired amount. For the f subsequent parametric study, all inputs are held constant, with the l

. exception rf system pressure, inlet enthalpy (temperature) and refer- l ence flow, i

l Page 1 of 3 s

The pressure-temperature limit study is performed as follows:

Starting from the maximum design point, a reactor coolant system pressure (e.g. ; 2000 psia) is selected for analysis ' A series of inlet temperatures are then input and the MDNBR corresponding to each is calculated. For each point the input reference flow rate-is density-corrected to correspond to the variation of reactor cou. ant system flow rate versus density. The result of this analysis ts a plot of MDNBR vs. reactor coolant system inlet temperature at constant p ressure. For each inlet temperature, the corresponding outlet temper-ature is calculated from the relationship Q = Wah, where w is the total system flow rate. The outlet temperature limit corresponding to the limiting DNBR is then taken directly from a plot of LONBR vs. Reactor Coolant Outlet Temperature. This analysis is repeated for system pressures ranging from 1800 to 2400 psia. The final result is the four pump pressure-temperature limit curve shown in Figure 8-3 of BAW-1433. For the initial submittal of BAW-1433, the limit curve corresponded to MDNBR = 1.30 (BAW-2) . For the revision to BAW-1J33 (submitted 1/13/77), the limiting 50NBR was increased to include the rod bow penalty (less allowances for the flow area reduction factor and the densification power spike penalty).

For three pump operation, the analysis is performed in the same manner described above. A flux / flow setpoint of 1.08 was assumed, resulting in a maximum power level of 86.7%. Although the three shown in Figure 8-3 coincides with the four pump curve, pump limit curve this case is actually slightly less limiting. For instance, in the analysis which included the rod bow penalty, the limiting outlet temperature for four pump operation is 607.8F at 2200 psia, while for three pump operation the corresponding limit is 608.0F.

The two pump limit curve is calculated in the same manner as described above. For this analysis, however, coolant quality was more limiting than DNBR.

The pressure-temperature limit curves shown in Figure 2.1-3 of the Tech Specs are the basis for the curve of Figure 2.1-1, which, in turn, is the basis for the variable low pressure trip setpoint (Figure 2.3-1) . -

Since the VLP trip setpoint is the same for all allowable pump operating modes, the most limiting curve from Figure 2.1-3 is used as the basis.

In the analysis of reload cores, when calculated, linit curves become less restrictive because of changes in core parameters, or reduction in penalties applied. B6W may elect to retain, without revision, the VLP trip setpoint. When this is done, the limit curve of Figure 2.1-1 (basis for VLP setpoint) will be retained, from cycle to cycle. while the curves of Figure 2.1-3 will demonstrate that this curve is conservatively restrie;ed.

2. Provide revis.:d Technical Specifications for core vent valve test require-ments as follows: The test requirements on the vent valves (page 73a) must include a test for that moment when the vent valves start to open. The vent valves must start to open at a differential pressure of 10.15 psi and ]

must be fully open at <0.30 psid. If force equivalents are to be specified,  ;

a detailed analysis anH documentation is required. '

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- RESPONSE Response not available at this time but will be supplied ,as soon as possible.

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3. Discuss ,why the curves for core protection safety Ibnit (F,igure 2.1.3) must be duplicated in Figure 2.1-1. If this duplication is not '

required, we recommend that the figures be incorporated into a single <

. figure.

RESPONSE-Although curve 1 in Figure 2.1.3 is the same as the curve in Figure 2.1.1 for' ANO-1, in other contracts the curves are not alike. In the case where the three pump curve crosses the four pump curve, the curve given in Eigure 2.1. is a composite of the four and three pump curves . When the curve in Figure 2.1.1 is not- a composite, the two figures are needed to make the text apply. The text must be unchanged

-in order that it be consistent with other BSW 177 reload reports.

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JANUARY 26, 1977 TELECOPY ECCS .

1. Supply piping physicals (or spool drawi gs) which indicate Safety.

Injection System (as-built) configuration and potential for sub-mergency under LOCA conditions. This information is required in order to complete the Appendix K review.

RESPONSE

Response supplied in letter D. A. Rueter (AP6L) to D. L. 2iemann (NRC) dated January 31, 1977.

"ANO-Unit 1, Proposed Tech. Spec. (File: 1511.1)" Ictter J. D. Phillips (APSL) to D. L. Ziemann (NRC), Januaq 13 1977.

2.

Provide the assumption used for fuel rod densification. Were the analyses othe same as that of the Fuel Densification Report, reference 6 of B6W 1433?

Was 'he 140.49 inch densified fuel rod length assumed for al' rods in the core ' were different densified lengths applied to the appropriate batch locations?

RESPONSE

For DNBR analyses which incorporated the densification power spike penalty

-(e.g. , pressure-temperature limits analysis), densification was treated in the same manner as described in reference 6 of BAW-1433. The power spike -

and maximum gap size used in these analyses were 1.087 and 3.10 inches, respectively, with a densified active length of 141.12 inches, resulting in a 5.93% reduction in DNBR. In re-evaluating these analyses to incorporate the fuel red bow penalty, credit for the densification power spike was -

determined by first calculating the effect of a reduction in active length from 144" to 140.49", then subtracting this effect from the densification penalty previously applied, resulting in a 4% credit. As described in the initial submittal of BAW-1433, the use of the 141.12" densified length (slightly non-conservative) was more than compensated for by the use of a power spike based on 92.5% dense fuel. In climinating the densification power spike penalty, the calculated densified length for Batch 4 fuel (140.49")

was used to avoid any non-conservatism in the analysis.

3.

Discuss the conservatisms applied in the rod bow penalty evaluation. Was the " maximum three cycle rod bow DNBR penalty" applied for the analysis of " Pressure-Temperature Limit Evaluation" (page 6-3 of BAN-1433)? What are the predicted burnups for E0C's 2 and 37 -

RESPONSE

The rod bow penalty evaluation has been based on the rod bow magnitude prediction equction provided in Reference 1 in conjunction with the Westinghouse curve of DNBR reduction versus rod bow to determine the reduction in DNBR as a function of burnup. This is consistent with Section 3.2 of Reference 2. The .conservatisms associated with the bow magnitude prediction are discussed in Reference 1. For DNBR analysis the most significant conservatism is that the maximum bow magnitude occurs in

-the second span from the bottom of the fuel rod (between the second and Page 1 of 5 s

third spacer grids) while the minimum DNBR occurs generally in the fifth or sixth span from the bottom. The conservatisms associated with the use of the Westinghouse DNBR reduction vs. bow magnitude prediction are discussed in Reference 2. As described on Page 6-3 of BAW-1433, the pressure temperature limit evaluacion included consideration of the maximum rod bow penalty based on the estimated maximum assembi.y-average burnup occurring after three* cycles of operation. This evaluation is further discussed in response to Question C.1 of Reference 4. The predicted maximum design assembly burnup at EOC-2 is 29000 MWD /MTU based on a cycle 1 length of 490 EFPD and a cycle 2 length of 272 EFPD. Although the cycle 3 design has not been finalized, we proj ect that the maximum design assembly burnup will be approximately 34000 FMD/hEU.

4. Provide a list of the conservatisms and credits and discussed in our letter of 12/30/76 (D. L. Ziemann of NRC to J. D. Phillips of Arkansas Power and Light Company) . Discuss the credits and quantitative values which were previously taken, which are currently applied, and which are remaining for each of the thermal margins of our letter.

RESPONSE

In evaluating the effects of the rod bow penalty, B6W claimed, and the NRC staff approved, credit for the following thermal margins applicable to ANO-1:

Flow Area (pitch) Reduction Available Vent Val

• Credit Densification Power ' pike Removal Excess Flow Over That Used in Safety Analyses Each of these credits is discussed below in terms of their present and future values. ANO-1 was licensed for first cycle ortration on the basis of the W-3 OlF correlation. For second, and future cycle;, licensing will be based on the B6W-2 CliF correlation, which has been reviewed and approved by the staff for design analysis of the 15X15 fuel assembly.

Flow Area Reduction Factor (FA)

This factor, used in subchannel and isolated hot channel thermal-hydraulic analyses was initially claimed to be equivalent to a 0.9% reduction in DNBR.

For Cycle 2, a re-evaluation has shown FA to be equivalent to 1.0% DNBR.

Therefore, for Cycle 2, a 1.0% credit has been taken (against the rod bow penalty). For future ycles, this 1% credit will be maintained or FA will be eliminated from thermal-hydraulic analyses and no credit will be taken.

Available Vent Valve Credit First cycle Reactor Protection System (RPS) Limits incorporated a penalty based upon the assumption that one vent valve stuck open. This penalty, which reduced the effective core flow by 4.6% while increasing the total system flow slightly has been shown to be equivalent to an 11.5% reduction in DNBR. For Cycle 2 and future cycles, this penalty has been eliminated from thermal-hydraulic analyses, therefore, no further credits will be taken.

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Densification' Power Spike Removal The densification power sp.4ke.as applied to ANO-1, Cycle 1, was equivalent to a 3.1% reduction in DNoP., Lad a corresponding credit was taken to offset

. the rod bow penalty. For the -initial Cycle 2 analyses, prior to the incor-poration of the rod bow penalty, a 4% densification power spike penalty was .

included ( e.g. , pressure-temperative limits analysis). For those analyses which were redone to incorporate the rod bow penalty (e.g., flux / flow analy-sis) the den.dfication power spike penalty was deleted and no credit was applied to the rod bow penalty. For future analyses, the densification power spike penalty (on DNBR) will no longer be calculated.

Excess Flow Over That Used in Safety Analysis As discussed -in BAW-1433, measurements taken during first cycle operation

, at ANO-1 demonstrated that the reactor coolant system flow rate is in excess of 109.7% of the design flow race. However, in the evaluation of thermal margins available to offset the rod bow penalty, no credit was claimed for this excess flow. The Cycle 2 thermal-hydraulic design analyses have been i based upon 106.5% of the design flow rate, thus the .available thermal margins have been accounted for directly rather than to rffset the red bow penalty.

It is anticipated that future analyses will continue to use 106.5% of the de-sign flow rate.

In reload licensing applications, RPS limits and setpoints incorporate all appropriate DNBR penalties. The fuel rod bow penalty, less appropriate

. credits,-is incorporated directly into these limits and setpoints. The appropriate credits are identified above .ar.J in the attached summary table.

In future . licensing submittals, these margins and credits will be re-evaluating -

when appropriate to insure their proper application.

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ARKANSAS NUCLEAR ONE - UNIT ONE Thermal Margins Available to Offset Rod Bow Penalty Credits - % DNBR Thermal Margin Cycle 1 Cycle 2 Future Cycles Flow Area (Pitch) Reduction 0.9 1.0 1.0 Available Vent Valve Credit 11.5 (1) (1)

Densification Power Spike Removal 3.1 4.0(2) (2)

Excess RC Flow (3) (3) (3)

Total Credits 15.5 5.0 1. 0 1.

NOTES: (1) Vent Valve penalty deleted (2) No' credit taken where densification penalty is removed "

from analyses (3) No credit claimed although flow in excess of 109.7's of design '

flow has been measured. Reload analyses are based upon 106.5*. .

of design flow G

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REFERENCES:

1) Letter, K. E. Suhrke to D. F. Ross (NRC), " Mark B Fuel Rod Bow Proj ection,"

September 10, 1976.

2) " Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Therma'l Margin Calculations for Light Water Reactors," NRC Staff (attachment to reference 3) .

3) Letter, D. L. Ziemann, NRC, to J. D. Phillips, Arkansas Power and Light Company, December 30, 1976.

4) Lette r, D. L. Ziemann, NRC, to J. D. Phillips, Arkansas Power and Light Company, January 21, 1977.

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