Text

Proposed Tech Specs Figures 3.2-2 & 3.2-2a Revised to Be Consistent W/Cycle 4 Safety Analysis Results
	ML20086S696
Person / Time
Site:	Palo Verde
Issue date:	12/24/1991
From:	; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	;
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References
	NUDOCS 9201030282
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_ _ _ _ _ . _ _ _ _ . . . _ . _ _ . . _ _ _ _ . . . . _ _ _ _ . . _ _ . _ .. _ _.

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PALO VERDE - UNIT 1 3/42-7 AMEMDMENT NO.

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FOR INFORMATION ONLY

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PALO VERDE- UNIT 1 3/4 2-7a AMENDMENT NO.

l

161 04355-WFC/ CAM December 24,-1991 ENCLOSURE B UNIT 1 CYCLE 4 RELOAD ANALYSIS' REPORT

161 04355-WFC/ GAM December 24, 1991 RELOAD ANALYSIS REPORT FOR PALO VERDE NUCLEAR GENERATING STATION UNIT 1 CYCLE 4 q TABLE OF CONTENTS Page

, 1. INTRODUCTION AND

SUMMARY

l-1

2. OPERATING HISTORY OF THE REFERENCE CYCLE 2-1

3. GENERAL DESCRIPTION 3-1

4. FUEL SYSTEM DESIGN 4-1

5. NUCLEAR DESIGN 5-1

.6. THERMAL-HYDRAULIC DESIGN 6-1

7. NON-LOCA TRANSIENT ANALYSIS 7-1

8. ECCS ANALYSIS 8-1

9. REACTOR PROTECTION AND MONITORING SY3 TEM 9-1

10. TECHNICAL SPECIFICATIONS 10-1

11. STARTUP TESTING 11-1

12. REFERENCES -- ' 12-1 APPENDIX A A-1 2 a- _ _ . _ _ _ _--___._m- _ _ _ _ .

161 04355-WTC/ CAM December 24, 1991

1.0 INTRODUCTION

AND S11MMARY

~

This report provides an evaluation of the design and performance of Palo Verde Nuclear Generating Station Unit 1 (PVNGS 1) during its fourth cycle of operation at 100% rated core power of 3800 MWt and NSSS power of 3822 MWL. Operating conditions for Cycle 4 have been assumed to be consistent ,

with those of the previous cycle and are summarized as full power operation under base load conditions. The core will consist cf irradiated Batch B, C, D, and E assemblies, along with fresh Batch F assemblies. The Cycle 3 termination burnup has been ascumed to be between 465 and 517 EFPD (Effective Full Power Days).

The third cycle of operation will hereafter be referred to in this report as the " Reference Cycle." Reference 1-2 presented analyso; for the Reference Cycle.

The safety criteria (margins of safety, dose limits, etc.) applicable for the plant were established in Reference 1-1. A review of those postulated accidents and anticipated operational occurrences evaluated in Reference 1-1 has shown that the Cycle 4 core design meets these safety criteria.

The Cycle 4 reload core. characteristics have been evaluated with respect to the Reference Cycle. Specific differences in core fuel loadings have been accounted for in the present analysis. The status of the postulated accidents and anticipated operational occurrences . for Cycle 4 can be -

summarized as follows:

1. Transient data are less severe than those of the Reference Cycle analysis; therefore, no reanalysis is necessary, or

2. Transient data are c.oto bounded by those of the Reference Cycle analysis, therefore, reanalysis is required.

1-1

. .e 161 04355 WFC/ CAM December 24 1991 for those transients requiring reanalysis (Type 2),- analyses are presented in Sections 7 -and 8 showing results that meet the established safety criteria.

The Technical Specification changes needed for Cycle 4 are summarized in Section 10.

h 1-2 o

l 161-04355 WFC/ CAM December 24, 1991 2.0 OPERATING HJ1 TORY OF THE REffRENCE CYCLE The Reference ~ Cycle began with initial criticality on June 24, 1990.

Power Ascension began on June 28, 1990, and on July 12, 1990 the unit reached full power.

I It is presently estimated that Cycle 3 will terminate on or auout February 1, 1992. The Cycle 3 termination poirt can vary between 465 and 517 EFPD to accommodate the plant schedule and still be within the assumptions of the Cycle 4 analyses, o

e l

2-1

161 04355-WFC/ CAM Deceniber 24, 1991 3.0 ENERAL DLS_(REJ10N The Cycle 4 core will consist of those assembly types and numbers listed in Table 31. One Batch B assembly, fifty-two Batch C assemblies, and forty-four Batch D assemblies will be removed from the Cycle 3 core to make way for eighty eight fresh Batch F assemblies. 108 Batch E and 36 ,

Batch D assemblies now in the core will be retained, in addition, 5 Batch B assemblies originally discharged at E001 and 4 Batch C assemblies originally discharged at E002 will be reinserted from the spent fuel storage. Figure 3-1 shows the poison shim and zoning. configuration for ;

the discharged assemblies, lhe reload batch will consist of 4 type F0 assemblies, 24 type F1 assemblies with 4 burnable poison shims per assembly, 4 type F2 assemblies with 12 burnable poison shims per assembly, 8 type F3 assemblies with 8 burnable poison shims per assembly, 16 type F4 assemblies with 16 burnable poison shims per assembly, and 32 type F5 assemblies with 12 burnable poison shims per assembly. These sub batch types are fuel zone-enriched and their assembly configurations are shown in Figure 3-2.

The loading pattern for Cycle 4, showing fuel type and location, is displayed in Figure 3-3.

Figure 3-4 displays the beginning of Cycle 4 assembly average burnup distribution. The burnup distribution is based on a Cycle 3 length of 517 EFPD, which is the long endpoint of Cycle 3.

Con'.rol element assembly patterns and in-core instrument locations will remain unchanged frcm the Reference Cycle and are shown in Figures 3-5 A

. & B and Figure 3-6, respectively.

3-1 l

l 161-04355 WFC/ GAM Decettber 24, 1991 TABLE 3-1 P ALO VERDE NUCLEAR GENERATING STATION UNIT 1 -

CYCLE 4 CORE LOAD;NG l ! Nominal Assembly Number >t - ' initlJ Number Shlm Number Number Desig- of i arichtf"m, thims/ loading of Fuel of Shlm' nation Assemblies ge -

MS) Assembly (gm B-10/in) Rods Rods B 5 l 16 0.018 1040 80 3

C 4 24, ', j 04 896 0 12 .', 48 D 36 184 4.05 0 6624 0 52 3.36 1872 E0 24 184 4.03 0 4416 0 52 3.90 1248 E1 20 168 4.03 16 0.024 3360 320 52 3.90 1040 E2 12 168 3.90 16 0..,24 2016 192 52 3.60 624 E3 12 168 3.90 16 0.026 2016 192 52 3.60 624 E4 24 168 3.90 16 0.016 4032 384 52 3.60 1248 E5 8 180 4.03 4 0.012 1440 ~32 52 3.90 416 E6 (P2E1) 8 168 4.03 16 0.016 1344 128 52 3.70 416 FO 4 184 4.03 0 736 0 l 52 3.80 208 F1 24 18G 3.80 4 0.014 4320 96 52 3.50 1248 F2 4 172 3.80 12 0.026 688 48 52 3.50 208 F3 8 176 3.80 0 0.022 1408 64 52 3.50 416 F4 16 168 4.03 16 0.028 2688 256 52 3.50 832 F5 32 172 4.03 12 0.026 5504 384 l

52 3.50 1664 l

l Total 241 3.85 54700 2176 3-2

l 161-04355 UTC/GA!4 December 24, 1991 FIGURE 3-1 PALO VERDE UNIT 1 CYCLE 4 ASSEMBLIES TO BE DISCHARGED AT E0C-3 FUEL AND BURNABLE POIS0N R0D PLACEMENT n n 11I III I I n n n I i In l II *jl i On l Il l l I I. I oua4 Arch e

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PALO VERDE NUCLEAR PALO VERDE UNIT 1 CYCLE 4 FIGORE GENERATING STATION CORE MAP 3-3 Unit 1 i 3-6 l

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l -

l 1 1. l- .1 - -1 l A :0- .C D- E- F. G. H- J K L- M: N. P R S' -- T I, ARlZONA i Figure f ' 'Palo Verde l Muclear Generating INSTRUMENT LOCATIONS - ;: Station 3-6 t__ __ 3-10~ . - - , - _ _ _ -- ... ,, - - - . - . . - - - . - ~ - . . . - . . . 4.0 fl!1LSJS1[tLDILLQ3 4.1 MIfliMLCA,LQLS10lj A 4.1.1 . fu_qLQuign The mechanical design of the B identical to the design assemblies except for a of the Reference Cycle Batch E r ,s center guide tube design. modification to the lowere oad fuel have occurred since the origin lNo changes a s gn bases in the mechanic fuel design, \ lhe following design features k were incorporated into Batch F The icwer end fitting desig . to a single piece casting withn was changed fit within the flow plate sembly from a two pie a recess for the center guide o tube t The length of the inches to center guide tube was 163.965 fitting. inches in increased order to fit within theer end new lowfrom 163 The new desiga provides imp the lower end fitting . roved strength, stiffness, and . 4.2 quality in GjUJDLlyBl_yLAR Twenty of the fuel assembli Cycle That 1 at Palo Verde Unit Ies that had CEAs located in the inspection was. part of were m during the required inspected 'for guide tube w required by the NRC for all licensing ear. operation also (References 4-1 -plants after the first procedures performed 4-7, and 4-8). cycle of on Unit 2 during the A similar program was (Reference 4-2 and . The 4-6) number first refueling guide tube inspection. wear was determined b of assemblies outage inspected for The inspections revealed thatased on the results of the Unit guide tube wear was minor 4-1 l l \ l 4.0 FUEL SYSTEM DESIGN ? 4.1 t!ECHANICAL DESIGN 4.1.1 Fuel Desian The mechanical design of the Batch F reload fuel assemblies is identical to the design of the Reference Cycle Batch E reload fuel assemblics except for a modification to the lower end fitting and center guide tube design. No changes in the mechanical design bases have occurred since the original fuel design. The following design features were incorporated into Batch F,' The lower end fitting design was changed from a two piece assembly to a single piece casting with a recess for the center guide tube to fit within the flow plate. The length of the center guide tube was increased from 163.715 inches to 163.965 inches in order to fit within the new lower end ( fitting. l The new design provides improved strength, stiffness, and quality in the lower end fitting. l 4.2 CUIDE TUBE WEAR l Twenty of the fuel assemblies that had CEAs located in them during Cycle 1 at Palo Verde Unit I were inspected 'for guide tube wear. l That inspection was part of the required licensing procedures required by' the NRC for all plants after the first cycle of operation (References 4-1, 4-7, and 4-8). A similar program was ! also performed on Unit 2 during the first refuelin.g outage (Reference 4-2 and 4-6). The number of assemblies inspected for guide tube wear was determined based on the results of the Unit 1 inspection. The inspections revealed that guide tube wear was minor 4-1 --- - .. - - - . -~ - - - - - - and will not adversely affect the fuel assembly performance throughout its expected life in the core, lhus no further guide tube wear inspections are necessary. Since guide tube wear is no longer an issue, discussion of guide tube wear will not be included in the reload analysis report in subsequent fuel cycles unless a design change is made that will affect guide tube wear. 4.3 THERMAL DESIM The thermal performance of composite fuel pins that envelope the pins of fuel batches B, C, D, E and F present in Cycle 4 has been evaluated using the FATES-3A version of the C-E fuel evaluation model (References 4-3 and 4-4). The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak pin at each burnup interval, from beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at the end of Cycle 4. The rod internal pressure remains below the reactor coolant pressure throughout Cycle 4. The power to centerline melt limit has been determined to be in er. cess of 21 kW/f t. 4.4 CHEMICAL DESIGN The metallurgical iwquirements of the fuel cladding for the Batch F fuel assemblies are the same . for the Batch -E assemblies. The metallurgical requirements of the fuel assembly structural members for the Batch F are the same as the Batch E fuel batches included in Cycle 3. Thus the chemical metallurgical performance of the Batch F fuel will be similar to (or better_ than) the Batch E fuel used in-Cycle 3. 4-2 4.5 Sil0VLDER GAP ADE0VACY The present shoulder gap is projected to be adequate for Cycle 4 operation. This conclusion is based on the fuel rod growth models of Reference 4-9 in conjunction with the measurements conducted post Unit 1 Cycle 2, Reference 4-1. r e 4-3 5.0 [LUCLEAR DESIGN 5.1 PHYSICS CHARACTERISTICS 5.1.1 Fuel Manaaement The Cycle 4 core makes use of a low-leakage fuel management scheme, in which previously burned assemblies are placed on the core periphery. Most of the fresh Batch F assemblies are located throughout the interior of the core where they are mixed with the previously burned fuel in a pattern that minimizes power peaking. With this loading and a Cycle 3 endpoint of 491 EFPD, the Cycle 4 reactivity lifetime for full power operation is expected to be 400 EFPD, Explicit evaluations have been performed to assure applicability of all analyses to a Cycle 3 termination burnup of between 465 and 517 EFPD and for a Cycle 4 length up to 426 EFPD. Characteristic physics parameters for Cycle 4 are compared to those of the Reference Cycle in Table 5-1. The values in this table are intended to represent nominal core parameters. Those values used in the safety analysis (see Sections 7 and 8) contain appropriate uncertainties, or incorporate values to bound future operating cycles, and in all cases are conservative with respect to the values reported in Table 5-1, Table 5-2 presents a summary of CEA reactivity worths and l l allowances for the end of Cycle 4 full power steam line break transient with a comparison to the Reference Cycle data. The full power steam line break was chosen to illustrate differences in CEA reactivity worths for the two cycles. The CEA core locations and group identifications remain the same as in the Reference Cycle. The power dependent insertion limit (PDil) for regulating groups and part length CEA groups is shown in Figures 5-1 and 5-2, respectively. Table 5-3 shows the I 5-1 I L reactivity worths of various regulating CEA groups calculated at full power conditions for Cycle 4 and the Reference Cycle, , 5.1.2 Egwer Distributioni I l Figures 5 3 through 5-5 illust. te the calculated All Rods Out , (ARO) relative assembly power densities during Cycle 4. The one-pin planar radial power peaks (fxy) presented in these figures represent the maxium over the mid eighty percent of the core's axial height. Time points at the beginning, middle, and end of cycle were chosen to display the variation in assembly and maximum planar radial peaking as a function of burnup. Relative assembly power densities for rodded configurations are given for B0C and EOC in Figures 5-6 through 5-11. The rodded configuration shown are those allowed by the PDIL at full power: part length CEAs (PLCEAs), Bank 5, and Bank 5 plus the PLCEAs. The radial power distributions described in this section are calculated data which do not include any uncertainties or allowances. The calculations performed to determine these radial power peaks explicitly account for augmented power peaking which is characteristic of fuel rods adjacent to th water holes. Nominal axial paaking factors are expected to range from 1.16 at B004 to 1.08 at E0C4. l l l L 5-2 -. .- . . - - . _.. - . - . . = . 5.2 EHYSICS ANALYSIS METH0Q1 5.2.1 Analytical Innut to In Core Measurements In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in accordance with Reference 5-1. Tha ROCS and MC codes employing DIT calculated cross sections will be used. ROCS, MC, and DIT have been approved for this npplication in Reference 5-2. 5.2.2 Uncertainties in Measured Power Distributions The planar radial power distribution measurement uncertainty of 5.3%, based on Reference 5-1, will be applied to the Cycle 4 COLSS and CPC on-line calculations which use planar radial power peaks. The axial and three dimensional power distribution measurement uncertainties are determined in conjunction with other monitoring and protection system measurement uncertainties, as was done for Cycle 3. 5.2.3 Nuclear Desian Methodoloav The Cycle 4 nuclea'r design was performed using the DIT, ROCS, and MC computer codes described in Reference 5-2 with- the minor improvements described below. In addition, the Appendix to this report contains the 50.59 determination that use of these improved codes does not require explicit NRC review. 5.2.3.1 Huclear Desian Code Imorovements Over the past several years, ABB Combustion Engineering Nuclear Power (CENP) has improved the codes and methods used.to analyze NSSS and reload fuel designs. Most of the code improvements fall in the categories of improved calculational efficiency, 5-3 r improved user friendliness, and improved exchange of data between different code modules. Only-four of the improvements affect the calculational accuracy of the results. These four improvements (addition of the nodal expansion method, anisotropic scattering, higher order interface currents, and assembly discontinuity factors) have been demonstrated to result in improved accuracy. In addition to the incorporation of these improvements, the associated biases and uncertainties , were revised as part of the overall verification process to insure that 95/95 confidence limits are maintained in all licensing related calculations. 5.2.3.1.1 Nodal Exoansion Method. - The use of. Nodal Expansion Solution Methed (NEM) in the ROCS code-was discussed in the original CENP ROCS /DIT Topical Report (Reference '5-2) even though it had not_ yet been fully integrated into the ROCS computer code. Recognizing this fact, the NRC stipulated only that CENP ensure that equivalent biases and. uncertainties be obtained when NEM is. incorporated into-the ROCS code. Prior to implementation of this improvement to the nuclear design code, CENP performed numerous benchmark calculations using data from past reload cycles. Updated calculational biases and-uncertainties were still defined by the 95/95 confidence limits. Equivalence was, thus, maintained and the limitation of- the NRC's approval of the Topical Report has not been violated. 5.2.3.1.2 Anisotronic Scatterina and Hiaher Order Interface Currenti The use of Anisotropic Scattering and Higher. Order Interface ! l Currents in the DIT code were discussed in CENP's Gadolinia-Uran.a Topical Report (Reference 5-3). In the approval of the report the NRC stated: 5-3.1 _ = _ _ _ _ - "We have revicsed the Combustion Engineering Licensing Topical Report CENPD-275-P, Revision 1-P. Based on our review, we conclude that the gadolinia fuel properties are acceptable for licensing applications up to 8 weight percent gadolinia concentration. We also conclude that the neutronics methods described in the report (DIT, ROCS /MC and P0Q), as modified, are acceptable for calculating the neutronic characteristics of PWR cores containing up to 8 weight percent gadolinia bearing fuel rods." Since the analysis presented in Reference 5-3 included assemblies which contained B 4 C poison rods or no poison material at all, the case of zero percent gadolinia is included in the range of applicability. 5.2.3.1.3 Assemb1v Discontinuity Factors Use of Assembly Discontinuity Factors (ADFs) in ROCS differs . from the improvements discussed above in that the function of the GFs are to improve the internal agreement between two - ex' .ing modules of the approved code . system (ROCS and DIT). Furthermore, unlike- the other methods of improvement, where improved accuracy must be demonstrated by statistical analysis of measured to calculated errors, the improvement of internal agreement resulting from the addition of ADFs can be verified at any time simply by comparing the ROCS and DIT computer output for the case of interest. It is the opinion of APS that-the addition of = ADFs has not changed - the overall code system representation of reality. Their use, e a significant and widely utilized industry breakthrough in PR calculational ability, is documented in Reference 5-4. 5.2.3.2 Revised Biases and Uncertainties Implementation of the improved methods has necessitated an update of the biases and uncertainties used to assure that 95/95 confidence 5-3.2 l limits are maintained in all results used for licensing related analyses. The revised biases and uncertainties were established by co. paring results obtained from analytical calculatron with ..ieasured data. The re evaluation of biases ' and unce:tainties used the same statistini methodology (with the exception of the N 1 rod worth as discussed below) as , , described in the ROCS /DIT Topical Report (Reference 5-2). I Consequently, CENP has concluded that the new biases and uncertainties fall within the original basis for acceptance of the ROCS /DIT Topical Report by the NRC in so much as the results tre judged to be equivalent when compared to other biases. In the ROCS /DIT inpical Report (Reference 5-2), the bias and uncertainty associated with not (N 1) rad worth is explicitly calculated by evaluating the not rod worth measurements perfonned during intial core startups. These evaluations shawed a 3.6% i underprediction of the N-1 rod worth, with a 1.47% standard ' deviation about the mean value. 1his standard deviation ic quite small and was deemed inappropriate for use in reload analysis for , two reasons. First, the N-1 statistics were based on a small number of N 1 rod worth measurements performed. Second, the N-1 i measurements were taken during the beginning of cycle for the initial cores, and hence may not be fully representative of later cycles. In view of these limitations of the N 1 statistics, the Topical Report embraced a conservative approach which applied the bias and uncertainty associated with individual bank worth to the N 1 rod worth. : It is recognized, however, that using the uncertainty for an individual bank for the N-1 rod worth is overly conservative. This is true because the maximum individual rod uncertainty is often dominated by rod banks with low worths. -For low worth rod banks, the percentage uncertainty is often high despite the fact that the absolute value of the uncertainty is small and well within the experimental precision. l 5-3.3- CENP has, hence, re evaluated the bias and uncertainty for the N 1 configuration. In particular, the N 1 bias and uncertainty used are the bias and uncertainty associated with the sum of tha bank worths (i.e., " total" worth). The use of the total rod worth uncertainty is considered more appropriate than the individual bank worth since the total rod worth configuration is more representative of the higher control rod density of the N 1 configuration. This alternative is sill 1 conservative because actual N1 measurements indica.e that the uncertainty of the N 1 rod worth is really lower than.the uncertainty of the total worth. CENP has performed calculations which demonstrate that the N 1 configuration is strongly influenced by the reactivity of the unrodded region of ' the core. Thus, the N 1 configuration is less sensitive to the , precision of the calculated effective control rod cross section as compared to either the total or individual bank configurations. This approach is consistent with the assumption in the Topical Report in which the total worth and N 1 rod configuration are as umed to belcng to the same population. Thus,-it is considered that the approach for the N 1 case yields equivalent calculational biases and uncertainties as compared to similar quantities calculated esing the nuclear design codes and methods described in the Topical Report. 5 3.4 TABLE 5 1 PVNGS UNil 1 CYCLE 4 NOMINAL PilYSICS CHARACT[RISTICS REf[RENCE DJ,10.(([P110N L!lfil$ CYCLE LELLA . Dissolved.Doron PPM Dissolved Boron Concentration for Criticality, CEAs Withdrawn flot full Power, Equilibrium Xenon BOC 1223 1120 Baron Worth PPH/%ap flot full Power DOC 127 128 flot full Power, EOC 98' 100 Moderator Tepperature Coefficients 10'46p/*f Ilot full Power, Equilibrium Xenon BOC 0.6 -1.05 E0C 3.3 2.30 llot Zero Power, BOC 40.3 40.02 D3ffler Coefficient 1 10-5Ap/*r Hot Zero Power, BOC -2.1 -1.81 llot full Power, DOC -1.7 -1.52 llot full Power, E00 1.9 -1,64 Iqttl.lelaved Neutron fraction. Beff --- - BOC 0.0069 0.0061 E0C 0.0046 0.005) Ers_mpt Neutron Generation Time, f* 10 6 sec BOC 20.7 20.1 EOC 27.3 25.8 (1) lhe differences between these reference cycle values and those presented for Cycle 4 are due to the inclusion of an uncertainty. Removing this uncertainty from the reft %nce cycle data yields Total Delayed Neutron fractions similar to those of Cycle 4. Thus, the reference cycle values, calculated without the uncertainty, are as follows: ' i) BOC 0.0063

11) EOC = 0.0051 54

1ABLE 5 2 PVNGS UNIT 1 CYCLE 4 LlH111NG VALUES Of REAC11V11Y WORillS AND :.LLOWANCES FOR 1101 FULL POWER STEAM LINE BREAK, %6p END Of CYCLE (EOC) REFERENCE DISCRIPTION CYCLE (1[L[_1

1. Worth of all CLA's Inserted -18.0 -16.2*
2. Stuck CEA Athedance 45.5 +3.9
3. Worth of all CEAs less highest Worth CEA Stuck Out -12.5 12.3*
4. Full Power Dependent Insertion Limit CEA Bite 40.2 +0.3
5. Calculated Scram Worth -12.3 -12.0
6. Physics Uncertainty +1.2 +0.8*
7. Other Allowances 40.1 +0.1
8. Net Available Scram Worth -11.0 -11.1
9. Scram Worth Used in Safety Analysis -10.2 -10.2 Deviation in the Cycle 4 values of items 1, 3 and 6 from those given for the reference cycle are due to the effects of fuel management differences and improvements in nuclear desi,1 methodology.

4 5-5 __ i 1ABLE 5 3 PVNGS UNIT 1 CYCLE 4 l REAC11V11Y WORTH Of CEA REGULATING CROUPS AT HOT FULL POWER, %Ap BIGjNNING OF CYCLE END Of CYCLE : , REGULATING ltEFERENCE REFERENCE .CEAs CYCLE CYCLE 4 CYC'.E CYCLE 4 Group 5 -0.31 -0.26 -0.33 0.28 Group 4 -0.37 -0.29 -0.39- 0.34 Group 3 0.91 0.70 0.92 0.87 Notes: Values shown assumo sequential group insertion, p 56 -._- --..-. - ..._. - - - - _ ..~.-. .... - - - - - . . - - . . - . ~ . - - . - - . - . - . . . ~ . . . o-1 UJ m- - a-0 __ CL W O ' a $o n o~ .J LU CC U B- g-1H ' K> _w 8 - u g .. ' -j ~s 7.- M ' UJ -E LU Z _x. o- a g- ._. t- W 2 e% w w o a "- - o if) Z E ~ 3-(f} H ,p, .G9 @ C dOOUD og- o p > _x-w m-G -y,W - W W _ _ _ m _. 5 = M gg- o- . .e c: , g e g I- I g _. g- la - O - DZU g_ ,, o _ H v fg - g* - / . .o_ a g. H uo z / / l "- o r i .09 O S d0000

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@i2 Zo 1 ~ Oo cn - ' O-D h t_E 8 y CD gX o o a-o uj Z g ._ u H W I _; I- o2 g _____.________________________ . ,. 1-p_ " s'aii az O' tu - N UJ _j 3 7 (D O E Z a <t - >- i-o: o iu - - o-c) i Q. E l~ 6- _ a: til tt g-n: i-O til (f) til - O O-(.) Q. 2 U) E - N tt o az- - a _; , , , , o, u o . to un v m o-o o o o o BBIAOd ~1UHUBH1 091U8 50 N0110Udd i I l l %P l AA BB AA - Quarter Core Location 1 E0 2 EO 3 E4 4 f0 BB Batch Type 0.47 0.64 0.65 0.84 CC CC Relative Power Density . 5 0 6 El 7 fl 8 EO 9 fl 10 0 0.35 0.64 1.08 1.11 1.24 0.88 11 E2 12 E6 13 fl 14 0 15 f5 16 0 17 f4 0.40 0.80 1.19 10.94 1.28 0.94 1.21 18 0 19 E6 20 F3 21 E4 22 f4 23 0 24 f5 25 El 0.35 0.80 1.23 1.13 1.23 0.91 1.27 1.09 26 El 27 fl 28 E4 29 f2 30 El 31 f5 32 E3 33 ES 0.64 1.19 1.14 1.30 1.10 1.28 1.09 1.10 34 EO 35 fl 36 0 37 f4 38 El 39 E3 40 E2 41 f5 42 0 0.47 1.08 0.94 1.24 1.10 1.04 1.06 1.30 0.96 43 EO 44 EO 45 f5 46 0 47 F5 48 E2 49 f4 50 E4 51 f3 0.65 1.11 1.28 0.91 1.28 1.06 1.22 1.06 1.32 . . 52 E4 53 fl 54 0 55 f5 56 E3 57 F5 58 E4 59 D 60 ES 0.65 1.25 0.94 1.27 1.09 1.30 1.06 0.90 1.11 X 61 FO 62 C 63 f4 64 El 65 ES 66 B 67 f3 68 E5 69 0 0.84 0.90 1.21 1.09 1.11 0.96 1.32 1.11. 0.88 NOTE: X MAXIMUM f 1.52 xy PALO VERDE NUCLEAR PALO VERDE UNIT 1 CYCLE 4 flGORE GENERATING STATION ASSEMBLY RELATIVE POWER DENSillES 5-3 Unit 1 AT BOC, AR0, lif P, Eq XE 59 i 'T I ? l AA BB AA = Quarter Core location 1 E0 2 EO 3 E4I4 F0 BB = Batch Type 0.47 0.63 0.63 0.80 : CC CC - Relative Power Density . ) a 5 0 6 El 7 F1 8 EO 9 F1 10 C , 0.37 0.64 1.05 1.04 1.19 0.86 4 11 E2 12 E6 13 F1 14 0 15 F5 16 D 17 Fi 0.43 0.81 1.17 0.93 1,29 -0.94 -1.26 : 18 D 19 E6 20 F3 21 E4 22 F4 23 0 24 'FS 25 El 0.37 0.81 1.23 1.11- 1.30 0.94 1.31 1.09 26 El 27 F1 28 E4 29 F2 30 El 31 F5 32 E3 33 ES 0.64 1.17 1.11 1.34 1.11 1.34 1.09. 1.09 X 34 E0 35 F1 36 0 37 F4 38 El 39 E3 40 E2 41 F5 42' B- t 0.47 1.05 0.93 1.30 1.11 1.05 1.07 ~1.34 0.96 43 EO 44 EO 45 FS 46 0 47 F5 48 E2 49 F4 50 E4 51 F3 0.63 1.04 1,29 0.94 1.34 1.07 1.28 1.04 1.30 . i 52 E4 53 F1 54 D 55 FS 56 E3 57 F5 58 E4 59 D 60 E5 0.63 1.19 0.94 1.31 -1.09 1.34 1.04 0.87 1.04 , 61 F0 62 C 63--F4 64 El 65 E5 66 B 67 F3 68 ES 69 0 t 0.80- 0.87 1.26 1.09 l 09 0.96 1.30~ 1.04L 0.84 : ~ u NOTE: X HAXIMUM f xy = 1.45 PALO VERDE ~ NUCLEAR PALO VERDE- UNIT'l- CYCLE 4 FIGORE 54-GENERATING STATION ASSEMBLY: RELATIVE POWER DENSITIES Unit 1 .AT HOC,lAR0, HFP, Eq XE-5 ' - =-= ,. r ,- u ,;.. . : w E , .,-.,,..,,..m.v,-4 c-.--,- -w--.-me,-,,...,m. -- mew-,,,m-,l.,A+.-- ,.- , ,+,.m.,,,-cm, ,,#mme--,,-+.~.,..,.r,.,,,.---, -,--w, -- AA BB AA = Quarter Core Location 1 E0 2 EO 3 E4 4 F0 BB = Batch Type 0.48 0.64 0.63 0.79 ' CC CC = Relative Power Density . 5 0 6 El 7 F1 8 EO 9 F1 10 0 0.40 0.65 1.03 1.00 1.15 0.86 11 E2 12 ES 13 F1 14 0 15 F5 16 0 17 F4 0.46 0.81 1.14 0.93 1.32 0.95 1.34 18 0 19 E6 20 F3 21 L4 22 F4 23 0 24 F5 25 Cl 0.40 0.81 1.20 1.08 1.37 0.97 1.35 1.09 X 26 El 27 F1 28 E4 29 F2 30 El 31 F5 32 E3 33 E5 0.65 1.13 1.08 1.35 1.10 1.36 1.07 1.05 34 EO 35 F1 36 0 37 F4 38 El 39 E3 40 E2 41 F5 42 B 0.48 1.03 0.93 1.37 1.10 1.03 1.06 1.34 0.94 43 EO 44 EO 45 F5 46 0 47 F5 48 E2 49 F4 50 E4 51 F3 0.64 1.00 1.32 0.97 1.36 1.06 1.34 1.02 1.24 . 52 E4 53 F1 54 D 55 F5 56 E3 E7 F5 58 E4 59 D 60 E5 0.63 1.15 0.95 1.35 1.07 1.34- 1.02 0.84 0.97 , s 61 F0 62 0 63 F4 64 Cl 65 ES 66 8 67 F3 68 ES 69 8 0.79 0.87 1.24 1.09 1.06 0.95 1.24 0.97 0.80 NOTE: X = HAXIMUM f xy - 1.48 PALO VERDE NUCLEAR PALO VERDE UNIT 1 CYCLE 4 FIGURE GENERATING STATION ASSEMBLY RELATIVE POWER DENSITIES 5-5 l Unit 1 AT EOC, AR0, HFP, Eq XE l 5-11 .~- 0 - ~

n - . , .- .,:r-- -r-->- v-- +,-

i i AA BB AA - Quarter Core Location 1 E0 2 E0 3 E4 4 F0 CC BB Batch Type 0.46 0.b3 0.64 0.86 PL CC Relative Power Density . PL - Part length CEA Location 5 D 6 El 7 F1 8 C0 9 F1 10 0 0.34 0.63 1.07 1.08 1.26 0.08 11 E2 12 E6 13 F1 14 D 15 F5 16 0 17 F4 0.39 0.79 1.20 0.90 1.14 0.91 1.23 Pl 18 0 19 E6 20 F3 21 E4 22 F4 23 0 24 f5 25 El 0.34 0.79 1.24 1.10 1.21 0.88 1.29 1.10 26 El 27 F1 28 E4 29 F2 30 El 31 F5 32 E3 33 E5 0.63 1.20 1.10 1.15 1.07 1.32 1.12 1.15 PL 34 EO 35 F1 36 0 37 F4 38 El 39 E3 40 E2 41 F5 42 B 0.46 1.08 0.91 1.21 1.07 1.04 1.09 1.39 0.99 43 EO 44 EO 45 F5 46 0 47 F5 48 E2 49 F4 50 E4 51 F3 0.63 1.09 1.14 0.88 1,32 1.09 1.29 1.12 1.43 . PL X 52 E4 53 F1 54 0 55 F5 56 E3 57' FS 58 E4 59 0 60 E5 0.65 1.27 0.92 1.30 1.12- 1.40 1.11 0.93 1.15 61 F0 62 C 63 F4 64 El 65 E5 66 B 67 F3 68 -E5 69 B 0.86 0.90 1.23 1.10 1.16 0.99 1.43 1.15 0.81 PL NOTE: X - KAXIMUM F,y - 1.56 PALO VERDE NUCLEAR PALO VERDE UNIT 1 CYCLE 4 FIGORE GENERATING STATION ASSEMBLY RELATIVE POWER DENSITIES 5-6 Unit 1 bOC, PLCEA'$ INSERTED, HFP, AR0 EQ. XENON 5 4 AA 5 AA - Quarter Core Location 1 E0 2 EO 3 F4 4 F0 CC BB Batch Type 0.50 0.68 0.68 0.90

BKS CC Relative Power Density .

BK5 - CEA Bank 5 Location 5 0 6 El 7 F1 8 EO 9 F1 10 C 0.37 0.69 1.18 1.19 1.33 0.9) l 11 E2 12 E6 13 F1 14 0 15 F5 16 0 17 F4 0.42 0.87 1.31 1.00 1.35 0.94 1.21 18 0 19 E6 20 F3 21 E4 22 F4 23 D 24 F5 25 El 0.37 0.87 1.37 1.22 1.30 0.89 1.20 0.95 X 26 El 27 F1 28 E4 29 F2 30 El 31 F5 32 E3 33 E5 0.69 1.32 1.22 1.39 1.11 1.25 0.93 0.69 BK5 34 E0 35 F1 36 D 37 F4 38 El 39 E3 40 E2 41 F5 42 0 0.50 1.18 1.00 1.30 1.11 1.01 1.00 1.18 0.79 - 43 EO 44 EO 45 F5 46 0 47 FS 48 E2 49 F4 50 E4 51 F3 0.68 1.19 1.36 0.90 1.25 1.00 1.15 0.97 1.23 . . 52 E4 53 F1 54 0 55 F5 56 E3 57 F5 58 E4 59 D 60 E5 0.68 1.34 0.94 1.20 0.93 1.18 'O.97 0.82 1.03 61 F0 62 C 63 F4 64 El 65 ES 66 0 67 F3 68 E5 69 B 0.90 0.93 1.21 0.95 0.69 0.79 1.23 1.02 0.80 BKS NOTE: X - HAXIMUM f xy - 1.59 PALO VERDE NUCLEAR PALO VERDE UNIT 1 CYCLE 4 _ FIGURE 57 l GENERATING STATION ASSEMBLY RELATIVE POWER DENSITIES - Unit 1 BOC, BANK 5 INSERTED, HFP, AR0 EQ. XENON l 5-13 . g-.-- m 9 q p - y. y- w-~,. w rw"-" i AA BB AA - Quarter Core Location 1 E0 2 EO 3 E4 4 F0 Cr BD Batch Type 0.49 0.68 0.68 0.91 BVC ' L CC Relative Power Density . PL Part length CEA Location 5 0 6 El 7 fl 8 EO 9 fl 10 C BK5 Bank 5 CEA 0.37 0,68 1.16 1.15 1.32 0.92 Location 11 E2 12 E6 13 fl 14 0 15 f5 16 D 17 f4 0.42 0.87 1.30 0.96 1.18 0.92 1.21 PL 18 D 19 E6 20 F3 21 E4 22 F4 23 0 24 f5 25 El 0.37 0.87 1.35 1.18 1.27 0.88 1.20 0.97 , X , 26 El 27 fl 28 E4 29 f2 30 El 31 F5 32 E3 33 E5 0.68 1.30 1.18 1.21 1.09 1.28 0.97 0.73 PL BK5 34 EO 35 fl 36 0 37 f4 38 El 39 E3 40 E2 41 F5 42 0 0.49 1.16 0.97 1.27 1.09 1.03 1.95 1.26 0.85 43 EO 44 [0 45 f5 46 0 47 f5 48 E2 49 F4 50 E4 51 f3 0.68 1.16 1.18 0,88 1.28 1.05 1.23 1.05 1.34 PL 52 E4 53 fl 54 D 55 f5 b6 E3 57 f5 58 E4 59 D 60 ES 0.69 1.33 0.92 1.21 0.97 1.26 1.05 0.89 1.10 61 f0 62 C- 63 F4 r4 El 65 E5 66 8 67 f3 68 ES 69 0 0.91 0.94 1.22 0,97 0.73 0.85 1.34 1.10 0.78 BK5 PL NOTE: X - MAXIMUM f 1.54 xy l PAIO VERDE NUCLEAR PALO VERDE UNIT 1 CYCLE 4 flGORE GENERATING STATION ASSEMBLY RELATIVE POWER DENSITIES 58 Unit 1 800, BANK 5 & PLCEA'S, HfP, ARO EQ. XENON 5-14 AA BB AA - Quarter Core location 1 E0 2 E0 3 E4 4 f0 CC BB Bat.1 Type 0.45 0.59 0.60 0.77 PL CC = R ive Power Density . PL Part length CEA Location 5 D 6 El 7 fl 8 EO 9 fl 10 C 0.37 0.61 0.98 0.93 1.13 0.84 11 E2 12 E6 13 F1 14 D 15 f5 16 0 17 f4 0.42 0.76 .l.10 0.87 1.15 0.92 1.38 PL 18 D 19 E6 20 F3 21 E4 22 F4 23 0 24 f5 25 El 0.37 0.76 1.17 1.02 1.36 0.95 1.42 1.13 26 El 27 fl 20 E4 29 F2 30 El 31 F5 32 E3 33 ES 0.61 1.10 1.02 1.18 1.07 1.45 1.13 1.13 PL 34 E0 35 ft 36 0 37 f4 38 El 39 E3 40 E2 41 f5 42 0 0.45 0.97 0.87 1.36 1.07 1.05 1.14 1.51 1.03 43 EO 44 EO 45 FS 46 0 47 FS 48 E2 49 F4 50 E4 51 F3 0.59 0.93 1.15 0.95 1.45 1.14 1.51 1.13 1.41 . . FL 52 E4 53 fl 54 0 55 F5 56 E3 57 F5 58 E4 59 0 60 ES 0.60 1.13 0.92 1.42 1.13 1.51 1.13 0.91 1.04 X 61 F0 62 C 63 F4 64 El 65 E5 66 B 67 F3 68 E5 69 B 0.77 0.85 1,38 1.13 1.13 1.03 1.41 1,04 0.74 PL NOTE: X = HAX1 HUM f xy - 1.59 PALO VERDE . NUCLEAR PALO VERDE UNIT 1 CYCLE 4 flGURE GENERAllNG STATION ASSEMBLY RELATIVf. POWER DLNSITIES 5-9 Unit 1 E0C, PLCEA'S INSERIED, HfP, ARO EQ. XENON 5 15 AA BB AA = Quarter Core Location ~i E0 2 EO 3 E4 4 f0 CC BB = Batch Type 0.49 0.64 0.63 0.81 BKS CC = Relative Powet i.nsity - BKS CEA Bank 5 Location 5B 6 El 7 fl 8 EO 9 fl 10 C O.41 0,67 1.08 1.03 1.20 0.87 11 E2 12 E6 13 F1 14 0 15 F5 16 0 17 F4 0.47 0.84 1.22 0.97 1.39 0.95 1.35-18 D 19 E6 20 f3 21 E4 22 F4 23 0 24 F5 25 El 0.41 0.84 1.30 1.14 1.47 0.97 1.31- 0.97 X 26 El 27 fl 20 E4 29 F2 30 El 31 FS 32 E3 33 ES 0.67 1.22 1.14 1.45 -1.12 1.38 0.94 0.66 BK5 34 EO 35 fl 36 0 37 f4 38 El 39 E3 40 E2 41 F5 42 B 0.49 1.08 0.97 1.47 1.12 1.02 1.03 1.27 0.81 53 EO 44 E0 45 F5 46 0 47 f5 48 E2 49 F4 50 E4 51 f3 0.64 1.03 1.39 0.97 1,38 1.03 1.33 0.97 1.19 52 E4 53 F1 54 0 55 FS 56 E3 57 FS 58 E4 59 D 60 E5 0.63 1.20 0.95 1.31 0.94 1.27 0.96 0.79 0.93 61 F0 62 C 63 f4 64 El 65 E5 66 0 67 F3 68 E5 69 - B 0.01 0.88 1.35 0.97 0.67 0.81 1.19 0.92 0.74 BK5 NOTE: X = MAXIMUM f 1,56 xy PALO VERDE . NUCLEAR PALO VERDE UNIT l-CYCLE-4 flGURE GENERATING STA110N ASSEMBLY RELATIVE POWER DENSITIES 5 10 Unit 1 E0C, BANK 5 INSERTED,-HfP, ARO EQ. XENON . l 5 16 AA BB AA - Quarter Core Location 1 [0 2 EO 3 E4 4 F0 CC BB - Batch Type 0.48 0.64 0.63 0.82 BK5/PL CC Relative Power Density . PL - Part length CEA -- Location 5 0 6 El 7 fl 8 EO 9 fl 10 C BK5 Bank 5 CEA 0.41 0.66 1.05 0.99 1.18 0.87 Location 11 E2 12 E6 13 fl 14 D 15 f5 16 0 17 f4 0.47 0.84 1.20 0.93 1.19 0.93 1.36 PL 18 0 19 E6 20 A3 21 E4 22 f4 23 0 24 f5 25 El 0.41 0.84 1.27 1.09 1.42 0.95 1.32 0.99 26 El 27 fl 28 E4 29 f2 30 El 31 F5 32 E3 33 ES 0.66 1.20 1.09 1.24 1.10 1,41 0.99 0.71 l'l BK5 34 EO 35 fl 36 0 37 f4 38 El 39 E3 40 E2 41 f5 42 8 0.48 1.05 0.93 1.42 1.30 1.05 1.10 1.38 0.89 43 EO 44 EO 45 F5 46 0 47 F5 48 E2 49 F4 50 E4 51 F3 0.64 0.99 1.19 0.95 1.41 1.10 1.45 1.07 1.32 . PL X 52 E4 53 F1 54 0 55 F5 56 E3 57 F5 58 E4 59 D 60 E5 0.64 1.18 0.93 1.32 0.99 1.38. 1.06 0.87 1.00 l 61 F0 62 C 63 f4 64 El 65 ES 66 8 67 f3 68 ES 69 8 0.82 0.88 1.36 0.99 0.71 0.89 1,32 1.00- 0.71 BK5 PL NOTE: X - HAXIMUM f xy - 1.52 PALO VERDE NUCLEAR - PALO VERDE UNIT 1 CYCLE 4 flGORE GENERATING S1ATION ASSEMBLY RELATIVE POWER DENSITlfS 5 11 l Unit 1 E0C, BANK 5 & PLCEA'S, lifP, ARO EQ. XENON 5-17 i s- -s w- , __,s y , 9- , g __ __ _ ._. _ . - - _ _ . . _ _ _ _ _ - _ _ _ _ _ _ . _ __ _ _ . _ - . . _ _ . ~ 6.0 TIIERMAL HYDRAVLIC OCi1M 6.1 DNUR ANALYSIS Steady state DNDR analyses of Cycle 4 at *~ rated power level of . 3000 MWT have been performed using the TORC compu*er code described in Reference 61, the CE 1 critical heat flux correlation described j in References 6-2 and 6 8, and the CETOP code described in Reference 6 3. Table 61 contains a list of pertinent thermal hydraulic desigre , parameters. The Modified Statistical Combination of Uncertainties (MSCV) methodology presented in Reference 6 4 was applied with Palo Verde 1 specific data using the calculational factors listed in Table 6-1 and other uncertainty factors to define overall uncertainty penalty factors to be applied in the DNBR calculations performed by the Core Protection Calculators (CPC) and Core Operating Limit Supervisory System (COLSS) which, when used with the Cycle 4 DNBR limit of 1.24, provide assurance at the 95/95 confidence / probability level that the' hot rod will not experience , , DNB. The 1.24 DNBR limit was calculated using the methodology of Reference 6-5 as was done for the Referenc,e Cycle. This Cycle 4 DNBR limit includes the following allowances:

1. NRC imposed 0.01 DNBR penalty for HID 1 grids as discussed in Reference 6 6.
2. Rod bow penalty as discussed in Section 6.2 below.

Other penalties imposed by HRC in the course of their review of the Cycle 1 Statistical Combination of Uncertainties (SCU) analysis discussed in Reference 6-5 (i.e., TORC code uncertainty and CE 1 CHF 6-1 _ _ _ _ _ _ ~ . _ _ . _ correlation cross validation uncertainty, as discussed in Reference 6 6) are included in the overall uncertainty penalty f actors dcrived in the Cycle 4 MSCU analysis. 6.2 [FFECTS OF FUEL R0D BOWING ON DNBR MARGIN Effects of fuel rod bowing on DNBR margin have been incorporated in the safety and setpoint analyses in the manner discussed in Reference 6 7. The penalty used for this analysis,1.75% HDNBR, is valid for bundle burnups up to 30 GWD/T. This penalty is included in the 1.24 DNBR limit. for assemblics with burnup greater than 30 GWD/T sufficient available margin exists to offset rod bow penalties due to the lower ! radial power peaks in these higher burnup batches. Hence the rod bow penalty based upon Reference 6-7 for 30 GWD/T is applicable for all assembly burnups expected for Cycle 4. 6-2 Table 6 1 PVNGS 1 Cycle 4 - Thermal Hydraulic Parameters at full Power Reference PVNGS 1 . General Characteristics Units Cycle Cycle 4 lotal Heat Output (Core only) MWt 3800 3800 E6 Dtu/hr 12,970 12,970 , fraction of Heat Generated in - 0.975 0.975 fuel Rod Primary System Pressure (Nominal) psia 1250 2250 Inlet Temperature (Nomine1) 'T 565.0 565.0 4+ 423,300 Total Reactor Coolant Flow gpm 423,300 (Minimum steady state) L6 lbm/hr 155.8 155.8 Coolant flow Through Core E6 lbm/hr 151.1 151.1 (Minimum) l Hydraulic Diameter (Nominal ft 0.039 0.039 channel) Average Mass Velocity E6 lbm/hr ft 2.49 2.49 . ! Minimum Pressure Drop Across Core psid 14.5 14.5 Steady State flow Irreversible AP Over [ntire fuel Assembly Total PressLre Across Vessel psid 51.3 51.3 (Based on nominal dimensions and minimum steady state flow) 2 Core Average Heat flux (Accounts 8tu/hr-ft 184,200 185,300 for fraction of heat generated in fuel rod and axial densification factor) 2 Total Heat Transfer Area ft 68,600 68.200 (Accounts for axial densification factor) L 63 1able 6-1 (continued) Reference PVNGS 1 General Characteristics Units Cycle Cycic 4 2 ,.f film Coefficient at Average Btu /hr-ft 6100 6100 Conditions - Average film Temperature 'r 30 30 Difference Average Linear Heat Rate of kW/ft 5.4 5.4 Undensified fuel Rod (Accounts for fraction of heat generated in fuel rod) - Average Core Enthalpy Rise Blu/lbm 85.9 85.'9 Haximum Clad Surface Temperature 'T 656 656 Engineering Heat flux factor --- 1.03 + 1.03 + 4 Engineering factor on Hot Channel --- 1.03 + 1.03 Heat input Rod Pitch, Bowing and Clad --- 1.05

1.05
Diameter factor fuel Densification f actor (Axial)- --- 1.002 1.002 Notes:
Based on 1872 poison rods.
Based on 2176 poison rods.

4 These factors have been combined statistically with other uncertainty factors as described in Reference 6-4 to define overall l uncertainty adjustment factors to be applied in the DNBR l calculat tons in COLSS and CPC which,_ when used in conjunction with ( the DNBR limit provides assurance at the 95 / 95 confidence / l probability level that the hot rod will not experience DNB. ++ Technical Specification minimum flowrate. 6-4 4 7.0 NON lDCA TRANSlfNT ANALYSIS j 7.0.1 IntroductioD This section presents the results of the Palo Verde Nuclear : Generating Station Unit 1 (pVNGS1), Cycle 4 Non LOCA safety , analyses at 3800 MWt. 4 TheDesignBasisEvents(DBEs)consideredinthesafetyanalysesare . Itsted in Table 7.01. These events - are categorized into three groups: Moderate frequency, Infrequent, and Limiting Tault events. .for the purpose of_ this report, the Moderate Frequency. and , infrequent Events will be termed Anticipated Operational-Occurrences. The DBEs were evaluated with respect to four criteria:- Offsite Dose, Reactor Coolant System (RCS) Pressure, fuel Performance (DNBR and Centerline Melt SAFDLs), and Loss of Shutdown Margin. Tables 7.0 2 through 7.0 5-~present the lists of events analyzed for each criterion. All events were re evaluated to assure

that they meet their respective criteria for Cycle 4.- The DBEs chosen for analysis for each criterion are the limiting events with

~ respect to that criterion. P 7.0.2 Methods of Analysis The analytical methodology used for PVNGS-1 Cycle 4 is the same as the Unit 1 Cycle 3 (Reference Cycle) methodology -(References 7-1, 7-2 and 7-9) with the exception of event 7.1.4c the Inadvertent , Opening of--a Steam Generator Saisty_ Valve or Atmospheric Dump Valve - . i with a Loss of Offsite Power for @ich _ Unit i Cycle ~ 3 (Reference

11) . forms Lthe Reference Cycle as it represents the latest -NRC- .

L position on the analysis of this event. Oniy-methodology that has_- - previously _ been reviewed and approved _ on the PVNGS' dockets- l (References 7-9, 7-10' and 7-11), and/or the CESSAR docket (Reference r L 7-2) is used. 4 71 . l - _ ,-- - , m .- _ _ . - , . - , . _ . _ - , - _ , _ _ , . . . _ - - , - . , _ . . _ ,. _.,- , . - . - - - - , - , _ . _ . 7.0.3 lidttematical Models The mathematical models and computer codes used in the Cycle 4 Non LOCA safety analysis are the same as those used in the Reference Cycle analysis (References 71, 7 2 and 7 9). Plant response for Non LOCA Events was simulated using the EESEC 111 computer code , (Reference 7 3). Simulation of the fluid conditions within the hot channel of the reactor core and calculation of DNBR was perfurmed usina the CETOP D computer code that was verified to be applicable in Reference 7 4. The 10RC computer code, was used to simulate the fluid conditions within the reactor core and to calculate fuel pin DNBR for the RCP Shaf t Seizure and Sheared Shaf t event. The 10RC code is described in References 7 6 and 7-7. The number of fuel pins predicted to experience clad failure is taken as the number of pins which have a CE 1 DNBR value below 1.24. The exceptions are the CEA Ejection, the Shaf t Seizure, Sheared Shaft and the inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve with a loss of Offsite Power events for which the statistical convolution method, described in ' Reference 7-8, was used. Reference 7 8 has been approved by the NRC and has been used in References 7-1, 7 2, 7 9, 7-10 and 7 11. The HERMITE computer code (Reference 7 5) was used to simulate the reactor core for analyses which required more spatial detail than is provided by a point kinetics model. Refennce 7-5 has been approved by the NRC and has been used in References 7-1, 72 and 7-9. HERMITE was also used to generate input to the CESEC point kinetics model by partially crediting space-time effects so that the CESEC calculetion of core power during a reactor scram is conservative relative to HERMITE. . 7-2 _____.J 7.0.4 Input Paranaglen_And_ADalysis Asmp11pm lable 7.0 6 summarizes the core parameters assumed in the Cycle 4 transient analysis and compares them to the values used in the Reference Cycle. Specific initial conditions for each event are tabulated in the section of the report summarizing that event. , Changes in the Technical Specifications that are necessary for the operation of Cycle 4 are described in Section 10. The effects of these changes were considered for each DBE and were included as appropriate, for some of the DBEs presented, certain initial core parameters were assumed to be more limiting than the actual calculated Cycle 4 values. Such assumptions resulted in more adverse consequences. Events which have credited CPC trip protection have assumed instrument channel response times which are , conservative relative to the Unit 1 Technical Specifications. 7.0.5 Conclusion All DBEs have been evaluated for pVilGS 1,- Cycle 4 to determine whether their results are bounded by the Reference Cycle. t 7-3 l l - i - .-. _ _ - . . _ _ . - - _ _ ~ .- - ._ __ _ - .. - 1able 7.0 1 PVB.GUl011.1.lleiblD_h111 LY1D113.0811dered in tti e Cycle 4_Saf etv AngJnia 7.1 increase in llest Rrtmoval by the Secondary System , 7.1.1 Decrease in feedwater Temperature 7.1.2 increase in feedwater Ilow 7.1.3 in:reased Main Steam flow 7.1.4 Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valyn 7.1.5* Steam System Piping failures 7.2 Decrease in lleat Removal by the Secondary System 7.2.1 Loss of Externel Load 7.2,2 Turbine Trip 7.2,1 loss of Condenser Vacuum 7.2.4 Loss of Nonnel AC Power 7.2.5 Loss of Normal feedwater 7.2.6* feedwater System Pipe !!reaks 7.3 Decrease in Reactor Coolant flowrate 7.3.) Total loss of forced Reactor Coolant flow 7.3.2* Single Reactor Coolant Purap Shaft Seizure /Sheated Shaft , 7.4 Reactivity and Power Distribution Ancaalles , , 7.4.1 Uncontrolled CEA Withdrawal from a Suberitical or low Pwer Condition 7.4.2 Uncontrolled CEA Withdrawal at Power 7.4.3 CEA Hi$ operation Events 7.4.4 CVCS Malfunction (inadvertent Doron Dilution) 7 4.5 Startup of an inactive Reactor Coolant Pump ' 7.4.6* Control Element Assembly Ejection 7.5 increase in Reactor Coolant System Inventory l 7.5.1 CVCS Halfu',ction 7.5.2 Inedverten. Operation of the ECCS During Power Operation

Categorized as I.imiting fault Events -

1-4 Table 7.0 1 (continued) 7.6 Decrease in Reactor Coolant System Inventory 7.6.1 Pressurizer Pressure Occrease Events 7.6.2* Small Primary l.ine Break Outside Containment 7.6.3* Ucam Generator lube Rupture 7.7 Hiscellaneous 7.7.1 Asymmetric Steam Generator Events

Categorized as Limiting fault Events 75

Table 7.0 2 ($11,1y11uated with Respect to Offsite Rose Criterien - Sec. tion Lytni Results A) Anticipated Opr ational Occurrences 7.1.4 1) Inadvertent opening of a Steam Bounded by Generator Safety Valve or Atmospheric Reference Cycle 0tmp Valve 7.2.4 2) Loss of Normal AC Power Bounded by Reference Cyc'e B) Limiting fault Events

1) Steam System Piping failures: Bounded by Reference Cycle 7.1.5a a) Pre Trip Power Excursions 7.1.5b b) Post Trip Return to Power 7.?,6 2) feedwater System Pipe Breaks Bounded by Reference Cycle 7.'.2 3) Single Reactor Coolant Pump Presented

, Shaft Seizure / Sheared Shaft '7.4.6 4) Control Element Assembly Ejection Bounded by Reference Cycle 1 t 7,6.2 5) Small Primary Line Break Outside Bounded by Containment Reference Cycle 7.6.3 6) Steam Generator Tubt Rupture Bounded by l Reference Cycle l I [ 76 1able 7.0 3 DMLEvaluated SJitLReipect to RCS Pressure Mc. tion 111tJ.Rn heni Resu1is A) Anticipated Operational Oc*1rrences 7.2.1 1) Loss of External Load Bounded by Reference Cycle 7.2.2 2) Turbine Trip Bounded by Reference Cycle 7.2.3 3) Loss of Condenser Vacuum bounded by Reference Cycic 7.2.4 4) Loss of Normal AC Power Bounded by Reference Cycic 7.2.5 5) Loss of Normal feedwater Bounded by Reference Cycle 7.4.1 6) Uncontrolled CEA Withdrawal from Bounded by Subcritical or low Power Condition Reference Cycle 7.4.2 7) Uncontrolled CEA Withdrawal at Power Bounded by Reference Cycle 7.5.1 8) CVCS Halfunction Bounded by Reference Cycle . 7.5.2 9) Inadvertent Operation of the Bounded by ECCS During Power Operation Reference Cycle B) Limiting Fault fvents 7.2.6 1) feedwater S.Sstem Pipe Breaks Bounded by Reference Cycle 7.4.6 2) Control Ele nent Assembly Ejection , Bounded by l Reference Cycle l 7-7 Table 7.0 4 DJLEs Evalugled with Respect to fuel Performann Section Lytni Rein)11 A) Anticipated Operational Occurrences 7.1.1 1) Decrease in feedwater Temperature Bounded by Reference Cycle 7.1.2 2) Increase in feedwater flow Bour.ded by Reference Cycle 7.1.3 3) Increased Main Steam flow Bounded by Reference Cycle 7.1.4 4) Inadvertent Opening of a Steam Bounded'by Generator Safety valve or Reference Cycle Atmospheric Dump. Valve 7.3.1 5) Total loss of forced Reactor Bounded by Coolant flow Reference Cycle 7.4.1 6) Uncontrolled CEA Withdrawal from a Bounded by Suberttical or low Power Condition Reference Cycle 7.4.2 7) Uncontrolled CEA Withdrawal Bounded by -at Power- Reforence Cycle 7.4.3 8) CEA Hisoperation Events Bounded by Reference Cycle l ! 7.6.1 3) Pressurizer Pressure Decrease Bounded by l Events Reference Cycle 7.7.1 10) Asymmetric Steam Generator Events Bounded by Reference Cycle l B) Limiting Fault Events

1) Steam System Piping failures: Bounded by Reference Cycle l

7.1.5a a) Pre Trip Power Excursions 7.1.5b b) Post-Trip Return to Power 7P l l 2 Table 7.0-4 (continued) LEGIiG ELM 1 Results - 7.3.2 2) Single Reactor Coolant Pump Presented Shaft Seizure / Sheared Shaft 7.4.6 3) Control Element Assembly Ejection Bounded by l ' Referm ce Cycle . 1 i l ,7-9 1 i i lable 7.0-5 l DJEs Evaluatsd with Respect to Shutdown Marain Criterion - Etcil0.0 Eyant Results l A) Anticipated Oper:tional Occurrences 7.1.4 1) tr. advertent Opening of a Steam Bounded by Generator Safety Valve or Reference Cycle Atmospheric Dump Valve 7.4.4 Bounded by

2) CVCS Malfunction (inadvertent Reference Cycle BoronDilution) 7.4.5 3) Startup of an inactive Reactor Bounded by Coolant System Pump Reference Cycle B) Limiting fault Events .
1) Steam System Piping Fcilures: Bounded by Reference Cycle 7.1.5b a) Post-Trip Return-to-Power l

l l 7-10 Table 7.0 6 PVNGS Unit 1. Cycle _d - [ ore Parameters Input to Safety Ang.l.yiel Reference Cycle Safety Param1(en Units Value Cycle 4 Valug Total RCS Power HWL 3898 3898 (Core Thermal Power i Pump Heat) Core inlet Steady State *f 560 to 570 560 to 570 Temperature (90% power and (90% power and above) above) 550 to 572 550 to 572 (below 90% power) (below 90% power) Steady State psia 2000 - 2325 2000 - 2325 RCS Pressure Minimum Guaranteed gpm 423,320 423,320 Delivered Volumetric flow Rate Axial chape Index LCO ASI Units -0.3 to +0.3 -0.3 to +0.3 Band Assumed ( 2 20% Power) ( 1 20% Power) -0.6 to +0.6 -0.6 to +0.6 ( < 20% Power) ( < 20% Power) Maximum CEA Insertion % Insert' ion 28 28 . at full Power of Lead Bank % Insertion 25 25 of Part-Length Maximum initial Linear KW/ft 13.5 13.5 ileat Rate Steady State Linear' KW/ft 21.0 21.0 Heat Rate for Fuel Center Line Melt CEA Drop Time from sec -4.0 4.0 Removal of Power to Holding Coils to 90% Insertion 7-11 Table 7.0-6 (continued) Reference Cycle Safety Parameters Units Value Cycle 4 Value Minimum DNBR CE-1 (SAfDL) 1.24 1.24 MacBeth (Fuel failure 1.30 1.30 . limit for post-trip SLB with LOAC - References 7-12 and 7-13) Initial Moderator 10~4 Ap/*F Figure 7.0-1 Figure 7.0 1 Temperature Coefficient Shutdown Margin (Value %Ap -6.5 -6.5 Assumed in Limiting Hot Zero Power SLB) e i 7-12 I 7.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSIEB 7.1.1 Qtgrease in feedwater Temperature l The results are bounded by the Reference Cycle. 7.1.2 Increase in feedwater Flow The results are bounded by the Reference Cycle. 7.1.3 Increaled Main Steam Flow The results are bounded by the Reference Cycle. 4 7.1.4 Inadvertent Openina of a Steam Generator Safety Valve or Atmospheric Dumo Valve The results are bounded by the Reference Cycle. 7.1.5 Steam System Pinina Failyres 7.1.5a Steam System Pipina Failures: Inside and Outside Containment Pre-Trio Power Excursions I l The results are bounded by the Reference Cycle. 7.1.5b' Steam System Pioina Failures: Post-Trio Return to Power p The results are bounded by the Reference Cycle L 7-13 7.2 DECREASE IN HEAT REMOVAL BY ItiE SECONDARY SYSTEM 7.2.1 Loss of External Load The results are bounded by the Reference Cycle. 1.2.2 Turbine Trio The results are bounded by the Reference Cycle. 7.2.3 Loss of Condenser Vacuum The results are bounded by the Reference Cycle. 7.2.4 Loss of Normal AC Power The results are bounded by the Reference Cycle. 7 .'2 . 5 L2ss of Normp1 Feedwater The results are bounded by the Reference Cycle, 9 .7.2.6 Feedwater System Pipe Breaks The results are bounded by the Reference Cycle. 7-14 7.3 DECREASE IN REACTOR COOLANT FLOWRATE 7.3.1 Loss of Forced Reactor Coolant Flqw The results are bounded by the Reference Cycle. 7.3.2 Sinale Reactor Coolant Pumn Shaft Seizure / Sheared Shaft , The amount of predicted failed fuel has increased for the Single Reactor Coolant Pump Shaft Seizure / Sheared Shaft from 3.79 to 4.32 %, which is less than the 4.5 % predicted fuel failure found acceptable by the NRC in Reference 7-11. The increase in -failed fuel was the result of more adverse- nuclear power distributions. The -resultant radiological consequences are a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary thyroid dose of less than 240 Rem This is within 10CFR100 guidelines. i 7-15 \ 7.4 REACTIVITY At{D POWER DISTRIBUTION ANOMLLES 7.4.1 Uncontrolled CEA Withdrawal from a Subtritical or low Power Condition The results are bounded by the Reference Cycle. 7.4.2 Uncontrolled CEA Withdrawal at Powar The results are bounded by the Reference Cycle. 1.4.3 [EA Misoneration Event The results are bounded by the Reference Cycle. 7.4.4 CVCS Malfunction (Inadvertent Boron Dilution) The results are bounded by the Reference Cycle, c 7.4.5 Startun of an inactive Reactor Coolant Pumo The results are bounded by the Reference Cycle. 7.4.6 Control Element Assembly Eiection The results are bounded by the Reference Cycle. 7-16 _ m 7.5 1HCREASE IN REACTOR CO0lANT SYSTEM IflyfNTORY 7.5.1 GYLS Malfunction The results are bounded by the Reference Cycle. 7.5.2 Inadvertent Operation of the ECCS Durina Power Operation The results are bounded by the Reference Cycle. 7.6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORY 7,6.1 Pressurizer Pressure Decrease Events The results are bounded by the Reference Cycle. 7.6.2 Small Primary Line Break Outside Containment The results'are bounded _by the Reference Cycle.- 7.6.3 Steam Generator Tube Ruoture The results are bounded by tne Reference Cycle. 7.7 d!TCEllANE0VS 7.7.1 Asymmetric Steam Generator Events The results are bounded by the Reference Cycle. l 7-17 m o n 9 9 ? , / x. o o

s. f_

8 *

@2 OJ C o O

81._; Z (_) a-T~ g F- x l 1 - Rh o- W N$o- _J CD - o o " w~ a E dI CE 3 o u-E o O oo D i_ " C - u; Z d - e _; cc w 2-1L -J CD [>- - Ox o$ 6 -J o _a - 81 E m. m w . m n ? o 9 . - o u o ~ n: 8m ,o 5 0 k k k ko u, ot ' . T P P I C d o/ 1, O I ) INSIJIdd3OJ 3801883dW31 801U8300W 7-18 8.0 ECCS ANALYSIS 8.1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT 8.1.1 Introduction And Summarv An- ECCS performance analysis of the limiung break size was performed' for PVNGS-1 Cycle 4 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled reactors (Reference 8-1). 'The analysis justifies an : allowable Peak Linear Heat Generation Rate (PLHGR) of '13.S kW/ft. The method of analysis and detailed'results which support this value'are presented herein. 8.1.2- Method Of Analysis The large break ECCS performance analysis for ,PVNGS-1 Cycle 4 consisted -of three parts: 1)-an evaluation - of the .' differences - between Cycle 4 and = Cycle _ 3, 12)~ . a calculation of cladding temperature and oxidation for - the hot rod for_ Cycle - 4, :and -3) a comparison of the results of the calculation to the results- of ~ PVNGS-l_ Cycle - 1. For' this reason PVNGS-1, Cycle -1 which ~ was the

Reference:
Cycle for Cycle'3_is referred to as the-Reference _ Cycle-in
~
-Section 8 for Cycle 4. Acceptable ECCS performance-was demonstrated l_ for the Reference Cycle in' Reference'8-2 and approved by the NRC.in
!' Reference 8-3._ As in the : Reference Cycle, the calculations l- performed for this evaluation ~ used' the' NRC ' approved .C-E large break ,
ECCS performance evaluation; model which is described in -Reference 8-4 including the use of a more conservative akial power shape. The i blowdown hydraulic calculations, refill /reflood: hydraulics
! - calcul'ations, and steam; cooling < heat transfer . coefficients of :- the j
Reference Cycle apply to-PVNGS-1: Cycle 4'since there .have been no
- significant adverse changes to RCS or ECCS hardware' characteristics, i or. to core and= system parameters Lintroduced by' Cycle 4. - Therefore, only- fuel rod cl_ adding temperature and: oxidation calculations. are i
8-1 iw ' y - " w e e e q ..- r
required to re evaluate ECCS performance with respect to the changes in fuel conditions introduced by Cycle 4. The NRC approved STRIKIN-11 (Reference 8 5) code was used for this purpose.
Burnup dependent calculations were performed with STRIKIN-Il to determine the limiting conditions for the ECCS performance c.alysis.
The fuel performance data were generated with the FATEo-a6 fuel evaltation mcdel (References 8-6 and 8-7). It was demonstrated that the burnup with the highest initial fuel stored energy was limiting.
This occurred at a low burnup for the hot rod.
The temperature and oxidation calculations were performed for the 1.0 Double-Ended Guillotine at Pump Discharge (DEG/PD) break. This break size is the limiting break size of the Reference Cycle and, as there are no significant differences between Cycle 4 and Cycle 1 that impact the hydraulic calcula'ien, is the limiting break size for Cycle 4.
8.1.3 Results The ECCS performance analysis for PVNGS-1 Cycle 4 showed that the Reference Analysis results conservatively apply. The peax cladding temperature, maximum local cladding oxidation, and core wide oxidation values of 2091*F, 9.0% and < 0.80%, respectively, for the Reference Analysis are below the corresponding 10CFR50./ ' . cceptance criteria of 2200*F, 17%, and 1%, respectively. These e: : s remain applicable for up to 400 tubes plugged per steam genei; -
and a reduction in system flow rate to 155.8x106 lbm/hr and a redis' ion in 6
core flow rate to 151.1x10 lbm/hr.
8.1.4 Conplusion l Conformance to the ECCS criteria is demonstrated by the. analysis resul ts. Therefore, operation of PVNGS-1 Cycle 4 at a core power 8-2
\
, ' level of 3876 MWt (102% of 3800 MWt) and a PLHGR of 13.5 kW/f t is in -
compliance with 10CFR50.46. ,
8.2 SMAll BREAK LOSS-0F-COOLANT ACCIDENT The small break ECCS. performance . analysis for PVNGS-1 . Cycle 4 consisted of an evaluation of the. differences between Cycle 4 and Cycle 3 and a comparison to the reported small break loss-of-coolant accident (LOCA) results (Reference 8 9) for PVNGS-1 Cycle 1. The .
analysis confirmed thats the peak cladding temperature for the limiting small break LOCA remains more than 300 F below that of the limiting large break LOCA.- -Thereforo, acceptable small" break ECCS
- performance is demonstrated ata peak linear heat generation' rate of -
13.5 kW/f t .and a reactor power level of 3876 MWt _ (102% 'of 3800 MWt)..
The acceptable performance has been confirmed with up to 400 plugged tubes per steam generator,.
e m
O f-i' 8-3 l
9,0 REACTOR PROTECTifLN AND MONITORING SYSTEM
9.1 INTRODUCTION
The Core Protection Calculator System (CPCS) is designed to provide .
the low DNBR and high Local Power Density (LPD) trips to (1) ensure that the specified acceptable fuel design limits on departure from nucleate boiling and centerline fuel melting are not exceeded during Anticipated Operational Occurrences (A00s) and- (2) assist tne Engineered Safety Features System in limiting the consequences of certain postulated accidents.
The CPCS in conjunction with the remaining Reactor Protection System (RPS) must be capable of providing protection for certain specified design basis events, provided that at the initiation of these occurrences the Nuclear Steam Supply System, its subsystems, '
components and parameters are maintained within operating limits and Limiting Conditions for Operation (LCOs).
9.2 CPCS SOFTWARE M00fflCATIONS The algorithms associated with the CPC Improvement Program (References 9-1, 9-2 and 9-3) which were implemented in Cycle 2, are applicable to . this cycle. The values for the Reload Data Block constants will be evaluated for applicability consistent with the cycle design, performance and safety analyses. Any necessary change to the' RDB constants will be installed in accordance with Reference 9-4.
9-1 4
9.3 ADDRESSABLE CONSTANTS Certain CPC constants are addressable so that they can be changed--as required _during . operation. Addressabic constants include (1) ]
constants that are measured during startup (e.g., shape annealing j matrix, boundary point power correlation coefficients, and , f
-adjustments for planar radial peaking factors), (2). uncertainty-factors to account for processing and measurement uncertainties in DNBR and LPD calculations (BERR0 through BERR4), (3) trip setpoints and (4) miscellaneous items (e.g., penalty _ factor._ multipliers, CEAC-penalty factor time delay, pre-trip setpoints,.CEAC inoperable flag, calibrationconstants,etc.).
Trip setpoints, uncertainty factors and other addressable constants -
will be determined for_ this cycle . consistent with the software.and methodology established in the CPC Improvement Program and the cycle design, performance- and safety analyses. As for the - Reference 4 i
Cycle, uncertainty factors will be determined using the modified q
statistical combination of uncertainties method (Reference 9-5). I
{
9.4 _ DIGITAL MONIT0 DING SYSTEM (C01.SS) l The- Core _0perating Limit Supervisory ! System L(COLSS),[ described in -
Reference 9 6, is - a monitoring. system that-- initiates - alarms- if the LCO's on' ONBR, peak linc ar heat rate, 'uxlal shape index, core power, or core azimuthal til, are exceeded E The CCLSSD data tuse - and uncertainties will be uodated, as required, to= reflect thel reload; core design.
9 9-?
& '6 - . JamaA
10.0 TECHNICAL SPECIFICATIONS This section provides a summary of the proposed changes to the Technical Specification for PVNGS-1 Cycle 4. The following changes are referenced by their appropriate Section number. Detailed change .
pages for the Technical Specification are presented elsewhere.
Section 3.2.4 and 4.2.4 (DNBR Marain):
Revise Figure 3.2-2, COLSS out of service DNBR limit line - CEACs operable and Figure 3.2-2A, COLSS out of- service DNBR limit line -
CEACs inoperable to reflect Cycle 4 core characteristics.
10-1
%N
11.0 STARTUP TESTING The planned startup test program associated with core performance is outlined below. The described tests verify that core performance is consistent with the engineering design and safety analysis. The program conforms to Reference 11-1, "Startup Test Programs", . and supplements = normal surveillance tests wi.ich are required by Technical Specifications (i.e., CEA drop time testing, RCS flow "
measurement, MTC verification, etc).
11.1. LOW POWER PilYSICS TESTS 11.1.1 Initial Criticality Before initial criticality, the critical boron concentration (CBC) will be estimated for the essentially all rods out (EARO) condition. ;
Initial criticality will be achieved by one of two methods. By the first method, all CEA groups would be fully. withdrawn (with_ the exception of the lead regulating group which wot:ld be positioned at approximately mid-core) before the.beginning of boron dilution. The boron concentration of the reactor coolant- would then be reduced until' criticality is attained. By the second - method, the baron concentration would be adjusted to the EARO estimated CBC before any CEAs are withdrawn. Then the PLCEA, shutdown, and regulating CEA groups would be withdrawn.In sequence to achieve criticality.
11.1.2 Critical Boron Concentration (CBC)
The CBC will be determined for- the unrodded configuration and for a partially rodded configuration. The measured CBC values will be verified to be within 11% Ak/k of. the predicted values.
11-1
)'
l 1
11.1.3 Temperature Reactivity Coefficient .
The isothermal temperature coefficient (lTC) will be measured at the The coolant 1 Essentially All Rods Out (EAR 0) configuration.
perature will be varied and the resulting reactivity change will .
be measured. The measured values will be verified to be within 10.3 x 10'4 Ak/k/*F of the predicted values.
J 11.1.4 CEA Reactivity Worth CEA group worths will be measured using.the CEA Exchange t'echnique.
This technique consists of measuring the worth ' of = a " Reference -
Group" via standard boration/ dilution techniques and then exchanging this group with other . groups; to measure .their -worths. All' full-length- CEAs will be included in' the measurement. Due to the large differences in CEA group worths, two reference groups (one with high worth and one with medium worth) may be used. The groups to be measured -will be exchanged with the appropriate . reference group. Acceptance criteria will- be:as specified-in Reference-11-2.
I
( 11.1.5 Inverse Boron Worth (IBW)
- The-IBW will be calculated using msults from.the CBC measurements and the CEA group worth measuremen6s. The-calculated:IBW value will be verified to be within 115 ppm /% Ak/k of the predicted value. '
11.2 Power Ascension Testina ,
following completion of the Low Power Physics Test sequence, reactor power will be , increased in accordance with L ~ normal ~ -operating
. procedures. The power. ascension will be monitored through use of an' l- -off-line . NSSS performance and) data- processing computer algor.ithm.- -
- This computer code. will be executed: in ; parallel with ; the -~ power
. ascension -to monitor CPC and 'COLSS performance relative to the:
processed plant data against which.they Jare normally calibrated, if 11'-2 1
i--- ,a r . - -v * .
l necessary, the power ascension will be suspended while necessary data reduction and equipment calibrations are performed. The ,
following measurements will be performed during the program. ,
p i
l 11.2.1- Flux Symmetry Verification
- Core power distribution, as detennined from fixed incore detector data, will--be examined prior to exceeding 30% power to verify that j no detectable fuel misloadings exist. Differences- between measured powers in symmetric, instrumented assemblies will;be verified to be within 10% of the symmetric group average, 11.2.2 Core Power Distribution Core power distributions derived from the fixed incore neutron detectors will be compared to predicted distributions at two power plateaus. These comparisons serve : to further verify proper fuel loading and verify consistency between - the as-built core and the engineering design models.- Compliance with- the acceptance criteria at the intermediate power ' plateau !(between 40% : and- 70% power) provides:-reasonable assurance that the _ power- distribution will remain within the design limits _ while-; reactor power isl increased to 100%, where the second comparison will-be performed.
The measured results will be compared to the predicted values in the following. _ manner- for - _ both the intermediate . and the full power-analyses:
A. The root-mean-square._ (RMS) of the difference between .the measured _ and predicted. relativa.) power . density (axially-integrated) for each -of the fuel assemblies will be verified to be less' than or equal .to 5%.-
11-3 i
4- -,. , , _ . , - - -
o
~
i B. The RMS of the difference- between' the measured and predicted core average axial power distribution for each axial node will be' verified to. be less than or equal to 5%.
C. The measured vahes of planar radial peaking factor _ (fxy), ,
integrated ' radial peaking factor-(Fr), core average axial peak (F ),. and the 3-L) power. peak (Fq) will be verified to 7
be within 110% of their predicted values. .
! 11.2.3- Shape Annealina Matrix (SAM) and Boundary Point Power Correlation Coefficients (BPPCC) Verificatiori 4
The SAM =and BPPCC values will be determined from a' linear regression analysis of the measured- excore detector readings and corresponding core . power distribution determined from incore- detecturt signals.-
Since these values must be representative for-'a . rodded and unrodded ,
- core throughout the cycle, it-is desirable-to use as wide a range of l
axial' shapes as- is available' to . establish. their . values.
The -
j spectrum of axial 1hapes: encountered during the power ascension- has
been demonstrated to be adequate for the calculation of the matrix elements. The necessary data will be compiled and1 analyzed _through
- the power ascension by the off-line NSSS ' performance . and = ' data processing'. algorithm. The - results of the ^ analysis : will - be _ used to_-
modify the appropriate CPC constants', if necessary. -
11 2.4 . - Radial Peakina factor (RPF)-and CEA Shadowino Factor (RSF)-
Verification The RPF: and' RSF values will .be determined using data collected from-
-the-fixed incore. detectors-and the excore detectors. Values will be e determined for unrodded:as well-as rodded _(lead regulating group and
- part-length group' only) operating conditions.: - Appropriate -CPC
~
and/or COLSS constants will be modified : based upon the . calculated values, The rodded portions of this measurement may be deleted frami ;
the test program if. appropriate adjustmentsE are made to CPC and-
~
L 1
l COLSS constants.
11-4 g -v -
r. . 9 - - W 9- w 4 -- ---F++ . F'- -P .mr g&tn r-r p-.q vig 'q r h ar Y -'WS N
11.2.5 lempfrature feactivity Coefficients at Power The moderator temperature coef ficient (MIC) will be measured at power within 7 EFPD af ter the accumulation of 40 EFPD. The measured MIC will be obtained from a measured isothermal temperature coefficient (ITC) using a calculaten fuel temperature coefficient '
(FTC). The ITC will be measured by cuenging coolant temperature, compensating with CEA motion, and maintaining power steady. The measured MTC will be compared to the MTC Technical Specificatien to verify compliance with the operating license. This comparison will be done in such a way as to account for the MTC measurement uncertainty.
11.2.6 Critical Boron Concentration The CBC will be determined for conditions of full power, equilibrium xenon. The measured CBC will be verified to be within 150 ppm of the predicted value after adjustment for the bias observed between measured and predicted CBC values at zero power, 11.3 PROCEDURE IF ACCEPTANCE CRITERIA ARE NOT MET The results of all tests will be reviewed by the plant's reactor engineering group. If the acceptance criteria of the startup physics tests are not met, an evaluation will be performed with assistance from the fuel vendor as needed.
11-5 l l
12.0 REFERENCES
12.1 SECTION
1.0 REFERENCES
(1-1) "Palo Verde Nuclear Generating Station Unit No.1, Final Safety Analysis Report," Arizona Public Service Company, Docket No.
50-528.
(1-2) "Palo Verde Nuclear Generating Station Unit 1 Reload' Analysis Report for Unit 1 Cycle 3," 161-01620-DBK/8JA, Januar.y 18, 1989,161-01858-DBK/SW, April, 19,1989,161-02064-WFC/KLHC, June 27, 1989, 161-02219-WFC/KLMC, August 25, 1989, 161-02294-WFC/RAB, September 11,1989,161-02812-WFC/ACR/KLMC, January 25, 1990.
12.2 SECTION
2.0 REFERENCES
None 12.3- SECTION
3.0 REFERENCES
None 12.4 SECTION
4.0 REFERENCES
(4-1) 161-02602-WFC/RAB/KLMC, "Palo Verde Nuclear Generating Station -
(PVNGS) Unit 1, Fuel Surveillance Test Results - Unit 1, End-of-Cycle 2," November 8, 1989.
(4-2) 161-01102-EEVB/PGN, "CEA Guide Tube Wear Inspection Results,"
June 9, 1988.
(4-3) CENPD-139-P-A, "C-E Fuel Evaluation Model," Combustion Engineering, Inc., July, 1974.
12-1
, I-]
(4-4) _CEN 161-(B)-P-A, " Improvements to Fuel Evaluation Model,"
August, 1989.
(4-5) Not Used (4-6) 161-00453-JGH/SGB, " Fuel Assembly Guide Tube Wear Program for PVNGS Unit 2," August 20, 1987.
(4-7) "Palo Verde Nuclear Generating Station, Updated Final Safety Analysis Report," Section 4.2.4 - Testing and Inspection Plan, Arizona Public Service Company, (Unit 1 Docket No. 50-528).
(4-8) CESSAR SSER2, Section 4.2.5, " Evaluation Findings", September 1983.
(4-9) CENPD 269 P, Rev. IP, " Extended Burnup Operation of Combustion Engineering PWR Fuel," July,1984.
12.5 SECTION 5.0 REFERENCfJi (5-1) CENPD-153-P, Rs . 1-P-A, "lNCA/CECOR Power Peaking Uncertainty," May, 1980.
(5-2) CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design," April,-1983.
(5-3) CENP-275-P, Rev.1-P-A, "C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," May,1988.
(5-4) K. S. Smith, " Assembly Homogenization Techniques for Light Water Reactor Analysis," Proaress n Nuclear Enerov, Vol .17, i
1986.
12.6 SECTION
6.0 REFERENCES
(6-1) CENPD-161-P-A "10RC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core", April,1986.
12-2
(6 2) CENPD 162-A, " Critical Heat flu.w, Correlation for C E fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," September,1976.
(6-3) 161-01867-DBK/JRP, " Applicability of the RAR References", April 26, 1989. ,
(64) CEN-356(V)-P A, Rev. 01-P-A, " Modified Statistical Co;nbination of Uncertainties", May, 1988.
(6-5) Enclosure 1-P to LD 82-054, " Statistical Combination of System Parameter Uncertainties in Thermal Margin Analyses for System 80", submitted by letter from A. E. Scherer (C-E) to 0.'G.
Eisenhut (NRC), May 14, 1982.
(6-6) CESSAR SSr.R 2 Section 4.4.6, " Statistical Combination of Uncertainties (SCU),' September 1983.
(6-7) CENPD-225-P-A, " Fuel and Poissa Rod Bowing," June 1983.
(6-3) CENPD-207-P-A, " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, Non-uniform '
Axial Power Distribution," December,1984.
12.7 SECTION
7.0 REFERENCES
(7-1) "Palo Verde Nuclear Generating Station Updated Final Safety Analysis Report," Arizona Public Service Company (Unit 1 Docket No. 50-528).
(7-2) "CESSAR, Combustion Engineering Standard Safety Analysis Report," Docket No. 50-470.
(7-3) "CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981, Enclosure 1-P to LD-82-001, January 6, 1982.
12-3 l
(7-4) 161-01867 0BK/JRP, ' Applicability of the RAR References," April 26, 1989.
(7-5) CENPD-188 A, "HERMITE ipacelime Kirietics," July,1975.
(7-6) CENPD-161-P-A, " TORC Code, A Computer Code for Determining the ,
Thermal Margin of a Reactor Core," April,1986.
(7-7) CENPD-206-P-A, " TORC Code Verification and Simplified Modeling Methods," June 1981. .
(7-8) CENPD-183-A, Loss of Flow - C-E Methods for Loss of Flow Analysis," June 1984.
(7-9) USNRC, " Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 44 to Facility Operating License No. NPF-41, Arizona Public Service Company, et. al.,
Palo Verde Nuclear Generating Station, Unit No,1, Docket No.
STN 50-528," September 19, 1989.
(7-10) USNRC, " Safety Evaluation by the Office of Nuclear Reactor Regulation Related te Amendment No. 34 to facility Operating License No, NPF-51, Arizona Public Service Company, et. al.,
Palo Verde Nuclear Generating Station',-Unit No. 2, Docket No.
STN 50-529," May 16, 1990.
l l (7-11) USNRC, " Safety Evaluation by the Office of Nuclear Reactor -
Regulation Related to Amendment No. 26 to facility Operating License No. NPF-74, Arizona Public Service Company, et. al.,
Paio Verde Nuclear Generating Station, Unit No. 3, Docket No.
STN 50-530," May 20, 1991.
l I
(7-12) R. V. MacBeth, "An Appraisal of Forced Convection Burrt out Data". Proc. 'Instn. Mech. Enars., Vol 180, Pt. 3C, pp 37-50, 1965-66, 12-4
(7-13) D. H. Lee, "An Experimental Investigation of Forced Convection Burn-out in High Pressure Water - Part IV, large Diameter lubes at About 1600 psia, A.E.E.W. , Report R479, 1986.
t 12.8 SECTION
8.0 REFERENCES
-ECCS ANALYSiti (0-1) " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," Federal Register, Vol. 39, No. 3, Friday, January 4,1974.
(8-2) ANPP-33584-EEVB/KLM, " Limiting Large Break LOCA Analysis Results - Chapter 15 Reanalyses", September 27, 1985.
ANPP-33650-EEVB/KLM, *Large Break LOCA Evaluation Model -
Reanalysis Results", October 3,1985.
(8-3) PVNGS Safety Evaluation Report, NUREG - 0857, Supplement C, Section 6.3, " Emergency Co u Cooling Systems," December 1985.
(8-4) CENPD-132P, " Calculational Methods for the C-E Large Break LOCA Evaluation Model", August 1974.
CENPD-132P, Supplement 1, " Calculational Methods for the C-E Large Break LOCA Evaluation Model", February 1975.
CENPD 132-P, Supplement 2-P, " Calculational Methods for the C-E Large Break LOCA Evaluation Model", July 1975.
Letter 0. D. Parr (NRC) to F. M._ Stern (C-E), dated June 13, 1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model).
letter 0. D. Parr (NRC) to A. E. Scherer_(C-E) dated. December I 9,1975 (NRC Staff Review of the Proposed Combustion
- Engineering ECCS Evaluation Model Changes).
I 12-5
(8-5) CENPD-135 P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Pro 9 ram", August 1974.
CENPD-135-P, Supplement 2 "STRIKIN-II, A Cylindrical Geometry fuel Rod Heat Transfer Program (Modifications)", February 1975. 3 CENPD-135, Supplement 4-P, "STRIKIN II, A Cylindrical Geometry fuel Rod Heat Transfer Program", August 1976.
CENPD-135, Supplement 5-P, "STRIKIN-II, A Cyl.indrical Geometry Fuel Rod Heat Transfer Program", April 1977.
(8 6) CENPD-139-P-A, "C-E fuel Evaluation Model", July,1974. '
(8-7) CEN-161(B)-P-A, " Improvements to Fuel Evaluation Model",
August, 1989.
(8-8) Deleted (8-9) ANPP-33609-EEVB/KLM, dated September 30, 1985, " Limiting Small 8reak LOCA Analysis - Additional Information".
12.9 SECTION
9.0 REFERENCES
(9-1) CEN-304-P, Rev. 01-P, " Functional Design Requirement for a Control Element Assembly Calculator," May,1986.
(9-2) CEN-305-P, Rev. 01-P, " Functional Design Requirement for a Core Protection Calculator," May,1986.
(9-3) CEN-330 P-A, "CPC/CEAC Software Modifications for the CPC Improvement Program Reload Data Block," October,1987.
(9-4) CEN-323-P-A, " Reload Data Block Constant Installation Guidelines," September, 1986.
12-6
(9-5) CEN-356(V)-P A, Rev. Ol P A, " Hoc',fied Statistical Combination of Uncertainties," May,1988.
(9 6) CEN 312-P, Rey, Ol-P, " Overview Description of the Core Operating Limit Supervisory System (COLSS)", November, 1986.
12,10 SECTION
10.0 REFERENCES
NONE 12.11 SECTION
11.0 REFERENCES
(11-1) USNRC, "Startup Test Programs," Memorandum from Paul S. Check, Chief, Reactor Safety Branch, 00R, to Reactor Safety, Branch, Washington D.C., Navember 28, 1778,
'11-2) CEN-319-A, " Control Rod Group Exchange Technique," June 1986.
12-7 gy , g .
, ' '- .,.....i-,..,.-.. - - - - - - - - - - - - - '
l APPEllDIX 10 RELOAD ANALYSIS REPORT FOR PALO VER0E NUCLEAR GENERATING STAT 10f4 UNil 1 CYCLE 4 IQlfLi0MlVB1!AT10il Per the requirements of 10CFR 50.59, a licensee is allowed to make changes to the facility described in the safety analysis report without prior NRC approval provided that the proposed change does not involve either (1) a change in the pin. t Technical Specifications incorporated in the license or (2) an unreviewed safety question.
A change to the facility described in the safety analysis report involves an unreviewed safety question:
(i .) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; t -
or (ii) if the possibility for an accident or maltunction of a different type than any evaluated previously in the safety analysis report may be created; or L
(iii) if the margin of safety as defined in the basis for any technical specification is reduced.
This 10CFR 50.59 evaluation includes (1) a brief description of the change to the facility described in the safety analysis report, (2) a determination as to whether the change involves an unreviewed safety question and (3) a safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question.
A1 s
(1) Description of Change
.he original methods and computer codes used to analyze the nuclear design of the core are described in Chapter 4 of the Palo Verde Nuclear Generating
. Station Updated Safety Analysis Report which refers to the System 80 Combustion Engineering Standard Safety Analysis Report. As indicated above, a licensee is allowed to n.4ke changes to these methods and codes provided that the change does not involve either a Technical Specification change or an Unreviewed Safety Question.
The nuclear design methods and computer codes provide calculated values for the following nuclear design parameters:
. Rohetivity ;
. Reactivity Coefficients ;
. Control Rod Worths i
. Peaking factors
. Power Distribution Related factors Several changes have been made to these methods and computer codes to (1) simplify their use. (2) improve their computational efficiency (e.g., the )
exchange of data between codes), and (3) enhance their calculational _ accuracy.
Of the three types of changes, only the latter, enhancing their calculational accuracy, is most likely to significantly affect the numerical results. Since the results of nuclear design analysis are used as input to the transient safety ,
analysis that considers accidents and malfunction of equipment important to-safety, these s hanges must be evaluated to determine whether or not an unreviewed safety quertion is created.
The original methods $nd computer codes art described in C E's proprietary Topical Report CENPD 266 P A, "The ROCS & DIT Computer Codes for Nuclear Design,"
datedApril1983. This Topical Report was generically reviewed and approved by ,
the NRC . Subsequent to the NRC's approval, changes were made to the methods and codes that could affect the calculational accuracy of the nuclear design computer codes. These changes _are as follows:
. Implementation of Nodal Expansion Method to ROCS
. Improved Accounting of Anisotropic - Scattering and liigher Order Interface Current Angular Distributions in O!T
. Use of Assembly Discontinuity factors between ROCS and DIT
. Update of Biases and UncertMoties Applieo to Calculated Parameters A2 ,
.Er .--,.r.-y., ----ma ,s[, .a -
- , 'm-
, , ,--,--n.,- -w M v r ,* [ e n-e , - new w - - *
eww
i A description of each change is provided below, the descriptions provide sufficient detail to perform a safety evaluation. Extensive reference is made to C E proprietary documents that contain additional details regarding the numerical effect of each change.
Rodal Expansion Method The Nodal Expansion Method (NEM) was added to the ROCS codo as an alternative to the original Higher Order Difference (%0) formulation. The ROCS code provides reactor power distributions and rr'iective neutron multiplication factort. This data is then used to darive control rod worths, depletion, reactivity coefficients and reactivity differentials. Use of the NEM achieves significant reduction in compuin running times and also improves agreement with fuel managen.ent measuremer.1 data.#
Althoupb the NEM had not yet been fully integrated into the ROCS code, the use of thc NEM was fully described in C E Topical Report CENPD 266 that was appreted by the NRC. Specifically, Topical Report CENPD-266 explained that NEM j hau been incorporated into a version of C E's coarse mosh kinetici code, HERitlTE. '
Furthermore, Topical Report CENPD 266 presented numerical comparisons of the NEM and H00 methods for solving the neutron diffusion equations. The results showed that the substitution of NEM for the H00 method in ROCS would not have a significant impact on calculational results and uncertainties.
In recognition of the expected future implementation of the NEM into ROCS, the NRC stated the following in the Safety Evaluation Report (SER) that approved C-E Topical Report CENPD 266:
We have revie'wed the ROCS and Oli computer codes as c% scribed in CENPD 266 P and CENPD 266-NP and find them to be acceptable for nuclear core design and safety-related neutronics calculations made by CE in 1icensing :
actions for power distributions, control rod worths, depletion, reactivity coefficients and reactivity differential. We also conclude that the ROCS code, including the fine mesh modulo MC, is of sufficient accuracy for the generation of coefficient Ithrtries for the in-core instrumentation.
The staff, however, recommends that CE perform further verification when the NEM is incorporated into the ROCS code in order to be assured that equivalent calculational biases and uncertainties are obtained with ROCS NEM as compared to ROCS H00."
A-3 L
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Before using ROCS NEM for nuclear design analysis for Palo Verde, C-E performed extensive verification to confirm that the calculational biases and uncertainties obtained with ROCS-NEM are equivalent to ROCS HOD. The SER did not require C E to resubmit the ROCS-NEM version of the code to the NRC for approval, it is important to note, however, that the NRC did recommend that the biases and uncertainties obtained when NEM was incorporated into ROCS be equivalent when compared to ROCS HOD. By equivalent, it is understood that the results between the two methods need not be numerically identical, but rather that the two methods be equal to the degree that the same conservative relationship is maintained between calculated and measured data (i.e., a 95/95 tolerance limit).
C E has performed detailed stcdies54 to confirm that the ROCS NEM nucicar core design and safety-related neutronics calculatiu ; of power distributions, control rod worths, depletion, reactivity coefficients and reactivity differentials maintain the same conservative relationship between calculated and measured data.- In particular, the tolerance limits applied to the calculated results from ROCS.H00 and ROCS NEM are identically defined as "the value that must be added to the calculated results to assure that 95% of the calculated values will be greater than the "true" value with 95% confidence." Thus, the ,
change which adds NEM to ROCS has been demonstrated to be equivalent to the ROCS-H00 version, which was approved by the NRC.
Anisotronic Scatterina and Hiaher Order Interface Current Anaular Distributions in order to -maintain the calculational accuracy in C-E Topical Report CENPD-266 when evaluating fuel containing gadolinium as a burnable poison, C-E had to improve the way the nuclear design computer code accounted for the effects of anisotropic scattering and higher order interface current angular distributions in the DIT code. The DIT code is a transport theory based code which performs spectral and spatial calculations in fuel cell and fuel assembly geometries. The DIT calculations provide few group neutron cross sections for use by the ROCS code.
The improved method for accounting for anisotropic scattering and higher order interface current angular distributions was submitted by C E.in a generic ;
Topical Report which was reviewed and approved by the NRCI . .These approved l methods and computer codes are described in C E Topical Report CENPD 275 P Revision 1-P-A, 'C-E Hethodology for Core Designs Containing Gadolinia-Urania _
Burnable Absorbers," dated May 1988. Although these changes were motivated by the need to obtain additional calculational accuracy to analyze gadolinium as a A _ _. ._._ . _ _ ,. _, _ . _ _ . _ _ _ _ _ _ _ .
burnable poison, the method itself is independent of the burneble absorber used
< in the coro.
Topical Report CENPD 275 was not submitted a a plant specific docket. it was reviewed by the NRC for generic implementation on FWR cores. In recognition of the generic applicability of the improvements made to the DIT code, the NRC stated the following in the SER that approved C E Topical Report CENPD 275:
"We have reviewed the changes mada to the Oli and ROCS /MC codes and methodology to accommodate the use of the integral burnable absorber gadolinium in PWR cores.
These changes are typical of the types made by the industry for computing gadolinia cores. The numerical results that were provided show that acceptable agreement has been obtained between detailed calculations and design calculations. We conclude therefore that the changes made to the DIT and ROCS /MC ,
codes and methodology are acceptabic."
"We also conclude that the neutronics methods described in the report (DIT, ROCS /MC, and PDQ), as modiffed, are acceptable for calculatit,g the neutronic ch tracteristics of PWR cores containing up to 8 weight percent gadolinia bearing fuel rods."
3 It is also important to note that benchmark studies and benc6 mark analysis provided in Topical Report CENPD-275 validated the changes made in the DIT code with B4C poison that contained no gadolinium. The NRC SER, thus, concluded that the methods described in Topical Report CENPD-275 are acceptable for calculating the neutronic characteristics of PWR cores containing up to 8 weight percent gado11nia bearing fuel rods. This includes the case where the PWR core contains Itto weight percent gadolinia by virtue of the fact that many of the assemblies used for benchmarking purposes did not contain ADX gadolinium bearing fuel rods.
Indeed, the NRC also noted in the SER the following:
"7he results obtained for the lead Test Assemblies (LTA) are consistent with those obtained for the non-gadolinium bearing fuel assembifes. The staff concurs with CC's conclusion that these results provide addit fonal validation of the DIT code and methodology."
Atsembly Discontinuity Factor _t Assembly discontinuity factors (ADFs) are used in the nuclear industrys as a method to eliminate homogenization error in nuclear design analysis where the A5 l
\
global heterogenous solution is known. The use of ADFs improves the internal t agreement between the DIT and ROCS codes. The ADfs are derived from the very assembly calculations required by the conventional homogenization methods and, therefore, they do act add any new information to the overall calculational methodulogy. Thus, the use of the ADfs is expected to improve the accuracy of results obtained from ROCS when compared to DIT. In several detailed studies, C-E has confirmed that the assembly discontinuity factors improve the accuracy of the nuclear design analysis method and computer codes.
Biases and Uncertaintin in view of the above changes that have been made to the methods and nuclear design computer codes, the biases and uncertainties applied to the nucicar design .
parameters were formally reevaluated by C-E'. for nuclear design parameters, the bias represents either the ' average of measurement minus calculation or the average ratio of the difference between measurement and calculation to the calculation. The uncertainty value reprosents the 95/95 tolerance range for the parameter of interest.
The reevaluation produced revised bias and uncertainty values that are ,
equivalent to those reported in C E Topical ' Report CENPD 266. By equivalent, it is meant that the results are not numerically identical, but rather that their '
application preserves the same conservative statistical relationship between calculated and measured data (i.e., the 95/95 probability / confidence level).
The methods used to generate the new biases and uncertainties are the same as that described in Topical Report CENPD-266, with the exceptions of the method
used to determine the bias and uncertainty for the net (N-1) rod worth, in the Topical Report, the bias and uncertainty associated with not (N 1) rod worth were calculated by evaluating the net rod worth measurements' performed during initial core startups. These evaluations found that the calculated (N-1) rod worth was being under-predicted by 3.6%, with a 1.47% standard deviation about the mean value, This standard deviation is quite small and was, . therefore, considered inappropriate for use in the reload analysis process for two reasons. First,the (N-1) statistics were based on only -a small number - of (N-1) rod worth measurements. Second, the (N-1) measurements were taken during the beginning of cycle for the initial cores and, hence, may not adequately represent reload cores. In view of the limitations of the (N 1) statistics available when the A-6 l
.+r s ---.eipm' q wv- r 3 yw-_me r y ww--r-g-a w- g e+m-9.i- -
Topical Report was written, C-E chose to take a conservative approach.
Specifically, C E applied the larger bias and uncertainty associated with ;
individual bank worths to the (N 1) rod worth. The evaluation of bias and l uncertainty for the individual bank worths exhibited a mean over prediction of 4Y. with a standard deviation of 4.2Y..
lising the uncertainty for an individual bank for the (N 1) rod worth is also overly conservative because the maximum individual rod uncertainty is often dominated by rod banks with low worths. For low worth rod banks, the percentage uncertainty is often high despite the fact that the absolute value of the uncertainty is small and well within the experimental precision.
When C E reevaluated the bias and uncertainty for the (N 1) configuration..
C E used the bias and uncertainty associated with the sum of the bank worths (1,0., " total" worth) in- lieu of that for individual banks. The u: of the total rod worth. uncertainty is considered more appropriate than the individual bank worth since the total rod worth configuration is more representative of the higher control rod density of the (N 1) configuration. for this case, the calculated and measured data exhibited a mean under prediction of 4.32% with a standard deviation of 1.97%.
This change in the bias and uncertainty used for the (N 1) case remains conservativebecauseactual(N1)measurementsdemonstratethattheuncertainty of the (N-1) rod worth is lower than the uncertainty of the total worth. This is expected since the (N-1) configuration is strongly . influenced by the !
reactivity of the unrodded region of the core. Thus, the (N 1) configuration is less sensitive to the precision of the calculated effective control rod cross sections than are either the total or individual bank configurations.
The change in method to calculate the (N-1) rod worth. produces equivalent set of bias and uncertainty, wherein the same conservative relationship is maintained between calculated and measured data (i.e., a 95/95 tolerance limit).
A comparison between tSe (N-1) rod worth using the criginal bias and uncertainty described in the Topical Report and those for new method is provided in_ Table-1 for past Palo Verde reload cycles, it can be seen that the-specific change in rod worth is not dramatic,- and in some cases non existent. The changes, however, do-indicate that there is generally more scram worth available than previous calculations suggested, in all cases, a 95/95 tolerance limit is still maintained between the calculated and measured results.
A-7
Table 1 (N 1) Rod Worth [% delta rho)-
with PVNGS Original Bias Revised Bias ,
Unit / Cycle 1 Uncertainty a Uncertainty 1/2 7.0 7.I 2/2 7.2 7.4 i 3/2 7.0 7.0 1/3 6.5 6.4 l 2/3 6.4 6.4 l l
I e
A-8
(2) Unroviewed Safety Question Determination The changes to the nuclear design analysis methods and computer codes described above can be implemented without prior NRC approval since there are no requireo changes to the Technical Specifications and an unreviewed safety question does not exist.
(3) Safety Evaluation The determination that the changes to the nuclear design analysis methods and computer codes described above do not create an unreviewed safety question is demonstrated by the following:
1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report will not be increased by the changes to the nuclear design analysis methods and computer codes described above.
The results of nuclear design analyses are used as inputs to the analysis of accidents or malfunction of equipment important to safety that are evaluated in the safety analysis report. These inputs do not alter the physical characteristics of any component involved in the initiation of an accident or any subsequent equipment malfunction. Thus; there is no increase in the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report as a result of this change.
The consequences of an accident or malfunction of equipment important to safety evaluated in the safety analysis report is affected by the value of inputs to the-transient safety analysis. There is always the potential for the value of the nuclear design parameters to change solely as a result of the new reload fuel core loading pattern. Regardless of the source of a change, an assessment is always made of changes to the nuclear design parameters with respect to their effects on- the consequences of accidents and equipment malfunctions previously evaluated in the safety analysis.
If increased consequences are anticipated, compensatory actions are implemented to neutralize any expected increase in consequences. These A-9
compensatory actions include, but are not limited to, crediting any existing margins in the analysis or redefining the operating envelope to avoid increase consequences. Thus, the nuclear dcsign parameters are intermediate results and by themselves will not result in a ir. crease in the consequence of accident or malfunction of equipment important to safety evaluated in the safety analysis report.
Therefore, the changes to the nuclear design analysis methods and computer codes described above do not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
2. The possibility for an accident or malfunction of a different type than '
any evaluated previously in the safety analysh report will not be created by the changes to the nuclear design analysis methods and computer codes described above.
As noted above, the results of nuclear design analysis are used as inputs to the transient safety analysis of accidents or malfunction of equipment important to safety that are evaluated in the safety analysis report.
These inputs do not alter the physical characteristics of any component involved in the initiation of an accident or any subsequent equipment malfunction. Thus, there is no increase in the pc:sibility of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report as a result of this change.
Thus, the changes to the nuclear design analysis methods and computer codes described above will not create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report.
3. The margin of safety as defined in the basis for any technical specification will not be reduced by the changes to the nuclear design analysis methods and computer codes described above.
Extensive benchmarking of the new nuclear design methods and computer codes has demonstrated that- the values of those parameters used in the safety analysis. are not significantly changed relative to the values obtained using the previous methods and computer codes. For any changes-A - 10 m .
n i
in the calculated values that do occur, the reevaluation of the biases and uncertainties ensures that the current margin of safety is maintained.
Specifically, use of these revised biases and uncertainties in safety evaluations continues to provide the same statistical assurance that the values of the nuclear parameters used in the safety analysis do not exceed the actual values on at least a 95/95 probability / confidence basis.
The changes to the nuclear design analysis methods and computer codes described above, therefore, do not reduce the margin of safety as defined in the basis for any technical specification. l In conclusion, the changes to the nucleo design analysis methods and comouter codes described above do not involve an unreviewed safety question and does not require a change to the Technical Specifications. Therefore, prior NRC approval l is not rcquired for this change.
5 A - 11
4 References
1. USNRC Letter from Cecil 0. Thomas (NRC) to A. E. Scherer (C E),
" Acceptance for Referencing of Litensing Topical Report CENPD 266-P, CENPD 266-NP "The ROCS and DIT Computer Codes for Nuclear Design",
April 4, 1983
2. R. A. Loretz and S. G. Wagner, "Recent Enhancements to the ROCS /MC Reactor Analysis System," Technical Paper presented at the International Conference on the Physics of Reactor Dperation, Design and Computation in Marseille, France April 22 27, 1990
3. P. H. Gavin and H. G. Joo, " Evaluation of Assembly Peaking Factors by DIT with P1 Scattering and DP2 Cell Coupling," PHA-88-109, November 2,1988(C-EProprietaryDocument)
4. U. Decher, " Power Peaking Biases and Tolerance Limits Using ROCS 3.0," PHA-90-163, July 23, 1990 (C-E Proprietary Document)
5. V. Decher, "ITC, Power Coefficient, IBW and Reactivity Bias using ROCS 3.0 or ROCS 4.0," PHA-90-241, November 9,1990 (C-E Proprietary Document)
6. P. H. Gavin, " Rod Worth Biases and Uncertainties with DIT(Pl-DP2) and ROCS (NEM) Methodologies," PHA-90-266, December 7, 1990 (C-E Proprietary Document)
7. USNRC Letter from Ashok C. Thadani (NRC) to A. E. Scherer (C-E),
" Acceptance for Referencing of Licensing Topical Report CENPD-275-P, Revision 1-P, "C-E . Met hodology for. Core Designs Containing Gadolinia-Vrania Burnable Absorbers," May 16, 1988.
8. - K. S. Smith, " Assembly Homogenization Techniques for Light W er.
Reactor Analysis," Proorest in Nuclear Enerov, Volume 17, 1986.
9. CE-CES-129 Revision 1-P, "Fethodology Manual - -Physics Biases and Uncertainties," August 2, 1991. (C-E Proprietary Document)
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ML20086S696

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SUMMARY

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