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Rev 0 to Technical Rept NE-978, Surry Unit 1 Cycle 12 Core Performance Rept
	ML18151A718
Person / Time
Site:	Surry
Issue date:	04/15/1994
From:	Brookmire J, Chapman D, Laroe C; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML18151A719	List: ML18151A718; ML18153A941;
References
	NE-978, NE-978-R, NE-978-R00, NUDOCS 9405190110
	Download: ML18151A718 (58)
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9405190110 940509 PDR ADOCK 05000280 p

PDR Surry

. Unit 1 Cycle 12 Core Peif ormance Report Nuclear Analysis and Fuel Nuclear Engineering Services April 1994 VIRGINIA POWER

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I TECHNICAL REPORT NE-978 - Rev. 0 SURRY UNIT 1, CYCLE 12 CORE PERFORMANCE REPORT

\\UCLEAR ANALYSIS AND FUEL NUCLEAR ENGINEERING SERVICES VIRGINIA POWER April, 1994

.~

I" PREPARED BY,f:l,,,, ;j}j,..,.,./JAPR 1'/

D. M. Chapman Date REVIEWED BY:~~

~- LaRoe

__-,1 tl REVIEWED BY: /.~

T. A. Brookmire APPROVED QA Category: Nuclear Safety Related Keywords: S1Cl2. CPR, Core

~

Date q.,s-11 Date

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TABLE OF CONTENTS PAGE Table or Contents 1

2 3

5 List of Tables.

List of Figures...

Section 1 Introduction and Summary.

Section 2 Burnup..

..... 13 Section 3 Reactivity Depletion............... 23 Section 4 Power Distribution..

............ 25 Section 5 Primary Coolant Activity............. 47 Section 6 Conclusions................... 55 Section 7 References.

......... 57

~E-978 S1Cl2 Core Performance Report Page 1 of 58

LIST OF TABLES

ABLE TITLE PAGE 4.1 Summary of Flux Maps for Routine Operation......... 30 SE-978 S1Cl2 Core Performance Report Page 2

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?IGURE

1. 1

1. 2

1. 3

1. 4

2. 1

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2.3 LIST OF FIGURES TITLE Core Loading Map...............

Burnable Poison and Source Assembly Locations.

~ovable Detector Locations Control Rod Locations.

Cycle Burnup History jonthly Average Load Factors Assemblywise Accumulated Burnup:

jeasured and Predicted 2.4 Assemblywise Accumulated Burnup:

Comparison of

~easured and Predicted.

2.SA Sub-Batch Burnup Sharing 2.5B Sub-Batch Burnup Sharing 2.SC Sub-Batch Burnup Sharing 3.1 Critical Boron Concentration versus Burnup (HFP,ARO) 4.1 Assemblywise Power Distribution - Sl-12-05 PAGE 8

9 10 11 15 16 17 18 19 20 21 24 31 4.2 Assemblywise Power Distribution - Sl-12-14 32 4.3 Assemblywise Power Distribution - Sl-12-23 33 4.4A Hot Channel Factor Normalized Operating Envelope (Applicable Through May 1992)

................ 34 4.4B Hot Channel Factor Normalized Operating Envelope (Applicable After May 1992) 35 4.5 Heat Flux Hot Channel Factor, Fq(Z) - Sl-12-05 36 4.6 Heat Flux Hot Channel Factor, Fq(Z) - Sl-12-14 37

4. 7 Heat Flux Hot Channel Factor, Fq(Z) - Sl-12-23 38 NE-978 SlC12 Core Performance Report Page 3

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LIST OF FIGURES (CONT'D)

FIGURE TITLE 4.8 Maximum Heat Flux Hot Channel Factor, Fq(Z)*P, vs.

Axial Position................

4.9 Maximum Heat Flux Hot Channel Factor, Fq(Z), vs. Burnup 4.10 Maximum Enthalpy Rise Hot Channel Factor, F-delta-H vs.

Burnup 4.11 Target Delta Flux versus Burnup 4.12 Core Average Axial Power Distribution 4.13 Core Average Axial Power Distribution 4.14 Core Average Axial Power Distribution 4.15 Core Average Axial Peaking Factor vs.

5.1 Dose Equivalent I-131 vs. Time 5.2 I-131/I-133 Activity Ratio vs. Time 5.3 Measured RCS Xenon-133 vs. Time 5.4 Measured RCS Iodine-131 vs. Time

~E-978 S1C12 Core Performance Report

- Sl-12-05

- Sl-12-14

- Sl-12-23 Burnup Page PAGE 39 40 41 42 43 44 45 46 51 52 53 54 4 of 58 I*

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Sect:ion 1 I

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I~TRODUCTIO~ A~D

SUMMARY

On i:1nuary :?

--, ::.994, Surry Unit 1 completed Cycle 12.

Since the i~it:ial criticality of Cycle 12 on May 1, 1992, the reactor core produced Jpproximat:ely 1.1623 x 10 3 MBTU (19,587 Megawat:t: days per met:ric t:on of

~ont:ained uranium).

The purpose of this report is to present an analysis of t:he core performance for rout:ine operation during Cycle 12.

The

t* physics tests t:hat :.:ere performed during the st:artup of this cycle were

~overed in Jl9herefore, t:he Surry Unit 1 Cycle 12 Startup Physics Test will not be included here.

Report: 1

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Surry Unit 1 began a power only coastdown on November 4, 1993, at which

~ime the burnup was approximat:ely 17,601 MWD/MTU.

The coastdown accounted for an additional core burnup of 1,986 MWD/MTU from the end of full power react:ivity.

The Cycle 12 core consisted of eight sub-batches of fuel:

two fresh batches (batches 14A and l4B); four once-burned batches, two from Cycle 11 (bat:ches 13A and l3B), one from Cycle 8 (batch 10) and one from Cycle 10 (part: of batch Sl/12B); and three twice-burned batches, all from Cycle 11 (bat:ches l'.:'.A, part of l2B, and S2/12A).

The Surry 1 Cycle 12 core

~oading map specifying the fuel batch identification and fuel assembly 1*~ocat:1.ons is shown in Figure 1.1.

The burnable poison locations and I

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source assemoly locations are shown in Figure 1. 2.

~ovable detector locations that were available during Cycle 12 are shown in Figure 1.3.

Controi rod locations are shown in Figure 1.4.

Routine core follow involves the analysis of four principal performance indicators.

These are burnup distribution, reactivity depletion, power distribution, and primary coolant activity.

The core burnup distribution is followed to verify both burnup symmetry and proper batch burnup sharing, ::hereby ensuring that the fuel held over for the next cycle.;ill be compatible with the new fuel that is inserted.

Reactivity depletion is monitored to detect the existence of any abnormal reactivity behavior, to determine if the core is depleting as designed, and to indicate the cycle burnup where coastdown operation will begin.

Core power distribution follow includes the monitoring of nuclear hot channel factors to verify that they are within the Technical Specification 2 limits, thereby ensuring that adequate margins for linear power density and critical heat flux thermal limits are maintained.

Lastly, as part of normal core follow, the primary coolant activity is monitored to assess the status of the fuel cladding integrity and to compare the concentration of dose equivalent iodine-131 in the reactor coolant with the limits specified by the Surry Technical Specifications 2.

Each of the four performance indicators is discussed in detail for the Surry Cnit 1 Cycle 12 core in the body of this report. The results are summarized below:

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1. Burnup - The burnup tilt (deviation from quadrant symmetry) on the core was no greater than +/-0.41% with the burnup accumulation in each batch deviating from design prediction by no more than +/-2.11%.

2. Reactivity Depletion -

The critical boron concentration, used to monitor reactivity depletion, was consistently within +/-0.48% tK/K I

of the design prediction which is within the +/-1% tK/K margin allowed by Section 4.10 of the Technical Specifications.

3. Power Distribution -

Incore flux maps taken each month indicated that the assemblywise radial power distributions deviated from the design predictions by a maximum average difference of 2.6%.

All hot channel factors met their respective Technical Specification limits.

4. Primary Coolant Activity The average dose equivalent iodine-131 activity level in the primary coolant during Cycle 12 was approximately 0.00628 µCi/gm.

This corresponds to less than 1% of the operating limit for the concentration of radioiodine in the primary coolant.

Radioiodine analysis indicated that there were fuel rod defects in Cycle 12.

NE-978 S1Cl2 Core Performance Report Page 7

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Figure 1.1 SURRY UNIT 1 - CYCLE 12 CORE LOADING MAP J

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1 FUEL ASSEHBLY DESIGN PARAHETERS SUB-BATCH l52/12A I 10 12A l2B 13A 138 14A 148 INITIAL ENRICHHENT 3.79 3.60 3.80 3.99 3.80 4.01 3.81 4.02 (W/0 U-235)

BURHUP AT BOC 12 211435 17757 35617 32362 17977 16922 0

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ONE ASSEHBLY (4G7J HAD ONE FUEL ROD REPLACED WITH A SOLID STAINLESS STEEL ROD DUl!IHG A RECONSTITUTION PROGRAH.

\\E-978 S1Cl2 Core Performance Report Page 8

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I Section 2 BURNUP The Surry Unit 1 Cycle 12 burnup history is graphically depicted in Figure 2.1.

Surry 1 Cycle 12 achieved a cycle burnup of 19,587 MWD/MTU.

As shown in Figure 2.2, the average load factor for Cycle 12 was 92.3%

~hen referenced to rated thermal power (2441 MW(t)).

Unit 1 performed a

?OWer coastdown starting on November 4, 1993 until shutdown for refueling 0n January 22, 1994.

Radial (X-Y) burnup distribution maps show how the core burnup is II..... shared among the various fuel assemblies, and thereby allow a detailed

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burnup distribution analysis.

The TOTE 3 computer code is used to I

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calculate these assemblywise burnups.

Figure 2.3 is a radial burnup distribution map in which the assemblywise burnup accumulation of the core

~t the end of Cycle 12 operation is given. For comparison purposes, the design values are also given.

Figure 2.4 is a radial burnup distribution map in which the percentage difference comparison of measured and predicted assemblywise burnup accumulation at the end of Cycle 12 operation is given.

As can be seen from this figure, the accumulated assembly burnups were within +/-4.14% of the predicted values.

In addition, deviation from quadrant symmetry in the core throughout the cycle was no greater than +/-0.41%.

The burnup sharing on a batch basis is core is operating as designed and to enable accurate end-of-cycle batch monitored to verify that the I

NE-978 S1Cl2 Core Performance Report Page 13 of 58

.:urnuo predictions to be made for use in reload fuel design studies.

~ate~ definitions are given in Figure 1.1.

As seen in Figures 2.SA, 2.SB,

~~d 2.5C, the batch burnup sharing for Surry 1 Cycle 12 followed design

?redictions closely with no batch deviating from prediction by more than

.. o,,

.I,.... 0.

The batch burnup sharing deviations in conjunction with reasonable agreement between actual and predicted assemblywise burnups, anci symmetric core burnups indicate that the Cycle 12 core did deplete G.S designed.

~E-978 SlC12 Core Performance Report Page 14 of 58 I

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Figure 1.3 SURRY UNIT 1 - CYCLE 12 MOVABLE DETECTOR LOCATIONS J

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~E-978 S1Cl2 Core Performance Report Page 11 of 58 1

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~E-978 S1Cl2 Core Performance Report

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Section 3 REACTIVITY DEPLETION The primary coolant critical boron concentration is monitored for the purposes of following core reactivity and to identify any anomalous reactivity behavior.

The FOLOW" computer code was used to normalize "actual" critical boron concentration measurements to design conditions taking i:ito consideration control rod position, xenon concentration, moderator temperature, and power level.

The normalized critical boron concentration versus burnup curve for the Surry 1 Cycle 12 core is shown in Figure 3. 1.

The maximum difference between measured and predicted critical boron concentrations was 55. 3 ppm.

The largest reactivity anomaly was +/-0.478% !:,.K/K which is within the +/-1% !:,.K/K criterion for

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reactivity anomalies set forth in Section 4.10 of the Technical Specifications.

In conclusion, the trend indicated by the critical boron I

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NE-978 SlC12 Core Performance Report Page 23 of 58

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Section 4 POWER DISTRIBUTION Analysis of core power distribution data on a routine basis is necessary to verify that the hot channel factors are within the Technical Specification limits and to ensure that the reactor is operating without any c1bnormal conditions which could 11 uneven burnup cause an distribution. Three-dimensional core power distributions are determined from movable detector flux map measurements using the INCORE 5 computer program.

A summary of all full core flux maps taken for Surry l Cycle 12 is provided in Table 4.1, excluding the initial power ascension flux maps.

Power distribution maps were generally taken at monthly intervals with additional maps taken as needed.

Radial (X-Y) core power distributions for a representative series of

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incore flux maps are given in Figures 4.1, 4.2, and 4.3. Figure 4.1 shows I

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Figure 4.2 shows a power distribution map that was taken near the mid-cycle burnup.

Figure 4.3 shows a map that was taken near the end of Cycle 12.

The maximum relative assembly power difference between measured and predicted for these maps was 7.7% and the cycle maximum average percent difference was 2.6%.

In addition, as indicated by the INCORE tilt factors, the power distributions were essentially symmetric for each case.

An important aspect of core power distribution follow is the monitoring

if nuclear hot channel factors.

Verification that these factors are

~E-978 SlC12 Core Performance Report Page 25 of 58

within Technical Specification limits ensures that linear power density and critical heat flux limits will not be violated, thereby providing adequate thermal margin and maintaining fuel cladding integrity.

Surry Technical Specification 3.12 limited the axially dependent heat flux hot channel factor, Fq(Z), to 2.32 x K(Z), where K(Z) is the hot channel factor normalized operating envelope.

During Cycle 12, there was a revision to Surry Technical Specification 3.12 which modified the K(Z) envelope 5

Figure 4.4A shows a plot of the K(Z) curve applicable for the maps up to Map 05.

Figure 4.4B shows a plot of the K(Z) curve applicable for Maps 06 through the end of Cycle 12.

The axially dependent heat flux hot channel factors, Fq(Z), for a representative set of flux maps are given in Figures 4.5, 4.6, and 4.7.

Throughout Cycle 12, the measured values of Fq(Z) were within the Technical Specification limit.

A summary of the maximum values of axially-dependent heat flux hot channel factors measured during Cycle 12 is given in Figure 4.8. The minimum margin to the Fq(Z) limit was 18.55%.

It should be noted that the graphical representation of Figure 4.8 does not demonstrate the Fq(Z) limit change.

The Fq(Z) limit applicable over the majority of the cycle is shown.

Figure 4.9 shows the maximum values for the heat flux hot channel factor measured during Cycle 12.

The value of the enthalpy rise hot channel factor, F-delta-H, which is the ratio of the integral of the power along the rod with the highest integrated power to that of the average rod, is routinely followed.

The Technical Specification limit for this parameter is set such that the departure from nucleate boiling ratio (DNBR) limit will not be violated.

~E-978 S1Cl2 Core Performance Report Page 26 of 58 I

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Additionally, the F-delta-H limit ensures that the value of this parameter

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used in the LOCA-ECCS analysis is not exceeded during normal operation.

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I Surry Technical Specification 3.12 was revised in June, 1992 to increase the F-delta-H limit to 1.56(1+0.3(1-P)), where 1.56 is the F-delta-H at rated thermal power 5

The measured F-delta-H without any uncertainty applied is compared directly to this limit.

In Table 4.1, flux maps through Map 05 have 4% uncertainty included

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to the listed F-delta-H values and were compared to the 1.55 limit.

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flux maps after Map 05 have no uncertainty applied and were compared to the 1.56 limit.

A summary of the maximum values for the enthalpy rise hot channel factor measured during Cycle 12 is given in Figure 4.10. This I

figure reflects the 100% power Technical Specification limit, the change

~in the 100% power Technical Specification limit and measured F-delta-H values that have the appropriate uncertainty applied. (The limit curve I

does not reflect the higher limit for maps taken at power levels less than I

100%.)

The change in the application of measurement uncertainty associated with the Technical Specification change explains the sudden I

drop in the measured F-delta-H values at beginning of cycle in Figure 4.10.

As can be seen from this figure, the minimum margin to the limit I

I was 4.84% for Cycle 12.

The target delta flux* is the delta flux which would occur at

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conditions of full power, all rods out, and equilibrium xenon.

The delta flux is measured with the core at or near these conditions and the target I

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Pb= power in bottom of core (MW(t))

I NE-978 S1Cl2 Core Performance Report Page 27 of 58

delta flux is established at this measured point. Since the target delta flux varies as a function of burnup, the target value is updated monthly.

By maintaining the value of delta flux relatively constant, adverse axial power shapes due to xenon redistribution are avoided.

This target delta-flux was also used to establish the operational axial flux difference bands while under CAOC.

The plot of the target delta flux versus burnup, given in Figure 4.11, shows the value of this parameter to have been approximately 2.5% at the beginning of Cycle 12, decreasing to -2.8% near the middle of Cycle 12, and leveling off at -3.5% at the end of Cycle 12 before increasing during the power coastdown.

This axial power shift can also be observed in the corresponding core average axial power distribution for a representative series of maps given in Figures 4.12 through 4.14.

In Map Sl-12-05 (Figure 4.12), taken at 178 MWD/MTU, the axial power distribution had a shape peaked toward the middle of the core with an axial peaking factor (F-Z) of 1.209.

In Map Sl-12-14 (Figure 4.13), taken at approximately 9,266 MWD/MTU, the axial power distribution peaked slightly toward the bottom of the core with an axial peaking factor of 1.148.

Finally, in Map Sl-12-23 (Figure 4.14), taken at 16,789 MWD/MTU, the axial peaking factor was

1. 151, with an axial power distribution similar to Map Sl-12-14.

The history of F-Z during the cycle can be seen more clearly in a plot of F-Z versus burnup given in Figure 4.15.

In conclusion, the Surry 1 Cycle 12 core performed satisfactorily with power distribution analyses verifying that design predictions were

~E-978 SlC12 Core Performance Report Page 28 of 58 I

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I I

I accurate and that: the values of the Fq(Z) and F-delta-H hot channel factors were within the limits of the Technical Specifications.

NE-978 S1Cl2 Core Performance Report Page 29 of 58

Table 4.1 SURRY UNIT 1 - CYCLE 12

SUMMARY

OF FLUX MAPS FDR ROUTINE OPERATION I

I l

2 I

3 I

I I

BURN I

BANK F-QCZJ HOT F-DHCNJ HOT ICORE FCZJ CORE I AXIAL I NO.I IHAPI UP I

D CHANNEL FACTOR CHNL. FACTOR IHAX TUT I

OFF I IF I IHO. I DATE lt\\lD/

PIIR I STEPS I

I I

SET ITHINI I

I HTU CZJ I I ASSYIPINIAXIAL I I.AXIAL I FCZJI HAX ILOCI CZ>

IILESI I

I I

I IPOINTIF-QCZJ IASSYIPIN IF-DHCNJIPOINT I I

I I

I_I I

I __ I 1 __ 1_1 __ 1 __ 1_1_1 __

1 __

1 __ 1 __ 1_1 __ 1_1 I 5 105-11-921 178 100.01 220 L05 OKI 30 I 1.866 L051 DK I 1.475 I 30 ll.20911.0161 NWI 2.6921 46 I 6 I 06-10-921 1151 99.931 224 E03 ICI 25 I 1.863 E031 IC I 1.413 I 30 11.20011.0121 NWI 2.Z411 4&

I 7 107-10-921 2141 100.01 221 E03 ICI 26 I 1.840 E031 IC I 1.404 I 30 ll.19011.0lGI NWI 1.7981 4&

I 8 108-10-921 3193 100.01 225 E03 ICI 25 I 1.801 LlOI GH I 1.408 I 30 ll.17411.0071 Nlfl 1.3261 411 I 9 109-14-921 4242 99.1181 225 FU Hll 32 I 1.767 LlOI GH I 1.410 I 30 I Ll60ll.0061 NEI 0.65el ft7 110 110-14-921 5280 99.9111 224 LlD IHI 41 I 1.759 LlOI IH I 1.411 I 37 IL14111l.0061 NEI -0.2831 4&

Ill 111-12-921 6244 99.941 224 LlO IHl42&43 1.763 LIOI IH I 1.415 I 41 ll.14111.0051 NEI -0.76111 4&

I 12 I 12-07-921 7105 100.11 224 LIO IHI 43 1.771 LlOI IH I 1.418 I 42 ll.l4Dll.0041 NWI -1.4461 47 113 IOl-13-931 8260 99.921 224 LlO IHI 45 l.765 LIOI IH I 1.417 I 45 ll.13511.0041 NWI -1.6191 47 114 102-15-931 9266 99.901 224 LlO IHI 4&

1.784 LlOI IH I 1.418 47 ll.14811.0021 NIii -2.7771 47 115 103-15-931 10112 99.921 224 LlO LGI 46 1.784 LlOI LG I 1.421 411 ll.14611.0031 NWI -2.6941 47 116 104-19-931 112110 100.01 224 LlD LGI 47 1.795 LlOI LG I 1.423 48 ll.15311.0021 NWI -3.ZZZI 411 117 105-19-931 12277 100.11 224 LlO LGI 47 1.799 LIDI LG I 1.422 48 ll.15&11.0021 NWI -3.5741 46 118 I0&-18-931 13209 100.01 223 LID LGI 47 1.795 llOI LG-I 1.423 52 ll.15411.0031 NWI -3.5441 47 119 107-16-931 14214 99.921 222 F05 IDI 52 1.794 LIOI LG I 1.419 52 ll.l&Oll.0021 NWI -2.7681 47 120 108-11-931 15060 99.941 222 F05 IDI 52 1.7112 LlOI LG I 1.410 52 ll.16311.0041 NIii -3.4911 47 I 21 I 011-31-931 15417 72.601 172 LlO LGI 411 1.7112 LIOI DF I 1.416 52 ll.16011.0071 NIii -5.6761 OWi 122 109-15-931 15907 99.981 223 F05 JDI 52 1.763 LIOI LG I 1.401 52 ll.16111.0091 NIii -3.4551 42 I 23 110-11-931 16789 99.971 223&2241 Kll FLI 52 1.756 LIOI LF I 1.397 52 11.15111.0041 Nwl -2.a1z1 :sa I 24 111-10-931 177811 95.341 223 I F05 JDI 52 1.631 LlOI LF I 1.396 52 ll.10911.0061 NIii -0.11661 43 I 25 111-22-93 I 111126 87.321 225 I LIO LFI 10 1.762 LIOI LF I 1.399 10 ll.16411.0l-OI NWI 3.59111 44 I 26 112-16-931 18809 73.871 223 I LIO LFI 09 1.867 LlOI LF I 1.402 09 ll.22211.0111 NWI 6.MZI 46 127 IOl-05-941 19263 62.671 218 I K09 JDI 09 1.991 LIO I LF I 1.407 09 ll.28711.0121 NMI 10.4151 4Z I_I ___ I

__ I I ___ I _______ I __ I ____

I __ I __ I_I __ I_

NOTES: HOT SPOT LOCATIONS ARE SPECIFIED BY GIVING ASSEKBLY LOCATIONS CE.G. HOB IS TIE CENTER-OF-CORE ASSEIBLY),

FOLLOWED IY THE PIN LOCATION (DENOTED BY THE "T" COORDINATE WITH THE FIFTEEN ROWS OF FUEL RODS LETTERED A THROUGH RAND THE *x* COORDINATE DESIGNATED IN A SIMILAR KANNER).

IN TIE *z-DIRECTION THE CORE IS DIVIDED INTO 61 AXIAL POINTS STARTING FROII THE TOP OF THE CORE.

1. F-QCZJ INCLUDES A TOTAL UNCERTAINTY OF 1.08.

2. F-DHCNJ INCLUDES AH UNCERTAINTY OF 1.04 FOR NAP 05.

THERE IS NO UNCERTAINTY APPLIED TD F-DH(NJ FDR KAPS 06 THROUGH 27.

3. CORE TILT - QUADRANT POWER TILT AS DEFINED BY THE INCDRE CODE.

~E-978 SlC12 Core Performance Report Page 30 of 58 I

I I

I..

I

I I

R p

Figure 4.1 SURRY UNIT 1 - CYCLE 12 ASSEMBLYWISE POWER DISTRIBUTION Sl-12-05 H

PREDICTED tlEASIJRED K

J H

G F

E 0.33 0.37 0.34 D

C B

PREDICTED 11EASURED

PCT DIFFERENCE.

. 0.35. 0.39. 0.35.

5.1. 5.0. 3.3 *

. PCT DIFFERENCE.

0.37 0.76 l.14 l.04 1.15 0.76 0.37

. o.:sa. o.79. 1.18

1.01

1.11

a.n. o.3a

4.6. 3.9. 3.2. 2.9. l.7. 0.6. 2.1.

0.41 l.13 1.23 1.20 l.17 l.20 1.24 l.13 0.40

. 0.42. 1.15. 1.26. 1.24. l.18. l.21. 1.24. 1.16. 0.42.

4.1. 2.1. 2.2. 3.3. 0.7. 0.3. 0.6. 3.0. 4.5 *

. 0.48 0.93 1.20 1.25 l.ll 0.94 1.11 l.25 1.20 0.93 0.48

. 0.41. 0.95

1.21

1.28. 1.14. 0.%. 1.11. 1.26

1.22

0.95

0.41

2.4.

2.1

0.6. 2.1. 2.7. 2.0. 0.1. l.l. l.3. 1.4. 1.6.

0.37 1.13 1.20 1.33 1.22 1.03 1.13 1.03 1.22 1.33 1.20 1.13 0.37

. 0.37. 1.14. 1.22. 1.37. 1.25. 1.06. 1.16. l.05. 1.24. 1.34. 1.19. 1.13. 0.37.

l.O. l.O. 1.7. 2.8. 2.4. 3.0. 2.9. 2.0. 1.2. 0.7. *l.2. 0.2. 1.9.

0.76 1.23 1.25 1.22 1.23 1.16 1.25 1.16 1.23 1.22 1.25 1.24 0.76

. 0.77. 1.24. 1.27. 1.26. 1.26. I.la. 1.28. 1.17. 1.25. 1.22. 1.23. 1.21. 0.75.

0.6. 0.6. 1.9. 2.7. 2.4. 2.1. 2.4. l.4. 1.0. -0.1. *1.4. *2.0. *l.8.

A 0.33 1.14 1.20 l.11 l.03 1.16 1.26 1.20 1.26 1.16 l.03 1.11 l.20 1.15 0.34

. 0.34. 1.16. 1.20. l.12. l.04. 1.17. l.26. l.21. 1.27. 1.16. l.OZ. 1.08. 1.16. 1.09. 0.32.

2.5. 1.3. o.4. 1.0. 1.2. o.9. 0.1. 1.1. o.4. o.3. -0.9. -2.a. -3.a. -s.o. -4.7.

0.37. 1.04 1.17 0.9't 1.13 1.25 l.20 1.19 1.20 1.25 l.13 0.95 l.17 1.04 0.37

. 0.38

1.06. 1.19. 0.95

1.13

1.25. 1.20

1.19. 1.19

1.24

1.11

0.9Z

1.13. 1.00

0.36

2.4. 1.6. l.6. 0.7. -0.l. 0.1. 0.5. o.o. -0.7. -0.6. *l.7. -3.l. -3.5. -3.6. -3.2.

0.33 1.15 1.20 1.11 1.03 l.16 l.26 1.20 1.26 1.16 1.03 l.11 1.20 1.15 0.33

. o.34. 1.16. 1.21

1.11. 1.02. 1.16. 1.26. 1.19. 1.25. 1.14. 1.00. 1.oa. 1.16. 1.12. o.33.

2.5. 1.5. l.O. 0.6. *O.l. -o.z. 0.4. -0.5. *0.8. *2.l. *2.2. -3.3. *3.3. *2.8. -1.8.

0.76 1.24 l.25 1.22 1.23 1.16. l.25 l.16 1.24 1.23 1.26 1.24 0.77

. 0.76

l.24

1.26

1.25

1.26

1.15

1.24 * *l.15

1.22

1.20

1.22

1.20

0.74.

0.1. 0.1. o.9. 2.0. 1.9. -o.6. -o.6. -o.a. -1.1. -1.a. -3.o. -3.Z. -3.o *

. 0.37 1.13 1.21. 1.33 1.22 1.03 1.13 1.03 1.22 1.33 1.20 1.13 0.37

. 0.38

1.16

1.23

1.36

1.23

1.01

1.12

1.03

1.22

1.32

1.19

1.12

0.36

2.4. 2.4. 2.2. 2.0. o.6. -1.1. -0.1. -0.1. 0.1. -1.z. -1.1. -1.0. -1.4.

0.40 0.93 1.21 1.25 1.11 0.95 1.11 1.25 1.20 0.93 0.40

. 0.42. 0.97. 1.23. 1.25. 1.09. 0.93. 1.09. l.23. l.18. 0.93. 0.40.

4.7. 3.6. 2.0. -0.2. -1.7. -1.9. -1.8. *1.5. *l.l. -0.Z. 1.0.

0.41 1.13 1.24 1.21 1.17 1.20 1.23 1.12 0.40

. 0.42. 1.16. 1.25. 1.18. 1.13. 1.16. 1.18. 1.09. 8.40.

3.8. 2.9. 0.7. *2.5. -3.3. -3.8. -3.6. *2.3. 0.3.

o.37 0.11 1.15 1.04 1.14 o.76 o.36

. 0.3a. 0.77. 1.12. 1.01

1.10. 0.73. 0.35.

2.9. l.l. *2.3. -3.3. -4.2. -3.9. *3.5.

STAHIWID DEVIATION

=l.274 0.34 0.37 0.33

. 0.33. 0.36. 0.32.

. *1.6. -2.7. -4.4.

AVERAGE

.PCT DIFFERENCE.

=

1.9 SUKttARY t1AP NO: Sl-12-05 DATE:

5/11/92 POWER: 100.oz CONTROL ROD POSITION:

F-Q(Z) = 1.866 QPTR:

D BANK AT 220 STEPS F-DH(N) = 1.475 NW l. 0164 INE 0.9998 I

F(Zl

= 1.209 SW 1.0053 ISE 0.9786 BURNUP = 178 HWD/HTU A.O. = 2.6927.

NE-978 S1C12 Core Performance Report Page 31 of 58 1

2 3

5 6

7 a

9 u

11 12 13 14 15

R Figure 4.2 SURRY UNIT 1 - CYCLE 12 ASSEMBLYWISE POWER DISTRIBUTION Sl-12-14 11 K

H G

F E

D 0.31 0.35 0.31 C

PREDICTED MEASURED

?PEDICTED

~EASURED

.PCT DIFFERENCE.

. 0.33. 0.37

0.32.

,.2. 4.2. 2.9.

. PCT DIFFERENCE.

0.36 0.70 1.03 0.93 l.03 0.71 0.36

. 0.37. 0.72. l.04. 0.94

l.04. 0.71. 0.37.

3.7. 2.5. 1.4. 1.1

o.3.

1.0.

2.8.

0.40 l.06 1.23 l.ll. 1.24 1.12 1.23 l.06 0.40

. 0.42. 1.07. 1.24. 1.13. l.Zl. 1.10. 1.24. 1.10. 0.43.

3.1.

0.5. 0.5. 1.7. -Z.l. -1.4.

l.O.

3.8. 5.7.

0.40 0.91 1.28 1.20 1.25 0.99 1.25 1.20 1.28 0.91 0.40

. 0.41

0.91. 1.26. 1.20. 1.26. 0.99. 1.22. 1.21. 1.30. 0.92. 0.41.

0.3. 0.5. -1.S.

0.4. 0.8. -0.l. -Z.9.

0.9. l.Z.

1.4.

2.6.

o.36 1.06 1.23 1.21 1.33 1.06 1.21 1.06 1.33 1.21 1.28 1.06 o.36

. 0.36. 1.05. 1.28. 1.28. 1.34. 1.08. 1.29. 1.07. 1.34. 1.:8. 1.25. 1.07. 0.33.

. -1.0. -1.0. -0.3.

l.O.

O.S. l.b. 1.7, 1.0.

0.6. 0.3. -2.2. l.O. 4.7.

0.70 1.23 1.20 1.34 1.22 1.30 1.21 1.30 1.22 1.33 1.20 1.23 0.70

. 0.70. 1.23. 1.20. 1.34. 1.23. 1.31. 1.23. 1.30. 1.22. 1.33. 1.13. 1.23. 0.71.

. -0.4. -0.4.

0.2. 0.2. 0.7.

1.0.

1.4. 0.6. 0.3. -0.4. -1.3, -0.5, 1.0.

A 0.31 1.03 1.11 1.25 1.06 1.30 1.17 1.15 1.13 1.30 1.06 1.25 1.11 1.03 0.31 *

. 0.33. 1.04. 1.12. 1.25. 1.05. 1.29. 1.17. 1.15. 1.18. 1.30. 1.06. 1.23. 1.10. 1.01

0.31.

4.0.

1.2. 0.2. -0.4. -1.l. -0.8. -0.l.

0.4.

0.0.

0.4. -0.l. -1.4, -1.S, -Z.2. -1.3.

o.35 o.93 1.24 o.99 1.21 1.21 1.14 1.03 1.14 1.21 1.21 o.99 1.24 o.93 o.35

. 0.37. 0.94. l.24. 0.99. 1.25, 1.20. l.14

l.07

l.14. l.21. 1.26. 0.97. l.22. 0.93. 0.36

4.0. 1.5. -0.0. -O.l. -1.S. -0.2. -0.3. -0.2. -0.2. 0.5. -0.6. -1.7. -1.6.

0.4.

l.l.

0.31 1.03 1.12 1.25 1.06 1.30 1.18 1.14 1.17 1.30 1.06 1.25 1.12 1.03. 0.31.

. 0.33. 1.04. 1.11. 1.25. 1.09. 1.31. 1.17. 1.14

1.17. 1.27. 1.05. 1.24. 1.12. 1.04. 0.32.

4.o.

1.0. -o.5.

o.3.

2.1.

o.9. -o.3. o.o. -o.3. -2.0 * -o.9. -o.3. o.o.

1.5.

3.3.

0.70 1.23 1.20 1.33 1.22 1.30 1.21 1.30 1.22 1.34 1.20 1.23 0.71

. 0.70. 1.22. 1.20

1.36. 1.24. 1.29. 1.20. 1.29. 1.22. 1.33. 1.20. 1.25. 0.72.

. -0.9. -0.9. 0.6. 1.8. 1.8. -0.6. -0.3. -0.4. -0.3. -0.2. -0.l. 1.1.

2.4.

0.36 1.06 1.28 1.27 1.33 1.06 1.27 1.06 1.33 1.27 1.28 1.06 0.36

. 0.37. 1.03. 1.30, 1.23. 1.32. 1.05. 1.26. 1.07. 1.35. 1.28. 1.31. 1.10. 0.37.

1.5. 1.5.

1.1

o.6. -o.s. -1.1. -0.3. o.5. 1.3. o.7.

2.1.

3.3. 3.3.

0.40 0.91 1.28 1.20 1.25 0.99 1.25 1.19 1.28 0.91 0.40

. 0.42. 0.93. 1.29. 1.18. 1.22. 0.97. 1.24. 1.19. 1.29. 0.94. 0.43.

3.9.

2.6. 0.6. -1.S. -Z.7. -Z.2. -1.1. -0.3. 1.0. 3.7. 5.7.

0.40 l.06 1.23 1.12 1.24 1.11 1.23 1.06 0.40

. 0.42. 1.10. 1.24. 1.03. 1.20

1.03. 1.20

1.06. 0.42.

3.9. 3.9. 0.5. -2.9. -3.0. -3.0. -2.6. -0.l. 4.6.

0.36 0.71 1.03 0.93 1.03 0.70 0.36

. 0.38. 0.73. 1.03. 0.91

0.99. 0.63. 0.35.

3.9. 3.3. o.3. -1.9. -3.5. -3.l. -2.5.

STANDARD DEVIATION

=1.304 0.31 0.35 0.31

. 0.32. 0.35. 0.30.

2.5. 0.1 * -3.6.

AVERAGE

.PCT DIFFERENCE.

=

l.5 SUHHARY HAP NO: Sl-12-14 DATE:

2/15/93 POWER:

99.9%

CONTROL ROD POSITION:

F-QCZl = 1.784 QPTR:

D BANK AT 224 STEPS F-DHCNI = 1.418 NW 1. a 019 NE 0.9999 F(Zl

= 1.148 SW 1.0015 SE 0.9967 BURNUP = 9266 HWD/HTU A.O. = -2. 777%

~E-978 S1Cl2 Core Performance Report Page 32 of 58 I

~

l I

z I

3 4

I 5

I 6

I 7

a I

9 el lD 11 I

12 I

13 I

lit 15 I

I I

I..

I

I I

I I..

I p

Figure 4.3 SURRY UNIT 1 - CYCLE 12 ASSE~BLYWISE POWER DISTRIBUTION Sl-12-23 H

PREDICTED MEASURED K

J H

G F

E 0.34 0.39 0.34 D

C

. PCT DIFFERENCE.

. 0.37. 0.42

0.36

7.7.

7.7.

5.7

PREDICTED HEASURED

.PCT DIFFERENCE.

0.38 0.71 1.02 0.93 1.02 0.71 0.38

. 0.41

0.73. 1.07. 0.98. 1.07. 0.73. 0.40, 6.0

2.9.

4.8.

5.8,

4.7.

2.6

4.1,

, 0.42 1.03 1.21 1.09 1.26 l.09 1.21 1.03 D.42

, 0.44. 1.04. 1.22. 1.13. 1.30, 1.12, 1.25. 1.08. D.45.

5.D

1.2. 0.3. 2.9. 2.9. 2.6. 2.6. 4.9, 6.4 *

. D.42, 0.90 1.26 1.16 1.30 1.02 1.30 1.16 1.26 0.90 D.42

. 0.45, 0.91. 1.23. 1.15. 1.32. 1.03. 1.28, 1.14. 1.28. 0.93, 0.45.

6.D.

l.l * *2.2. -0.3. l.8. l.8. *l.3. *l.5. 1.6. 3.3, 6.5.

0.38 1.03 l.26 l.21 1.35 1.08 l.32 1.07 l.35 1.22 i.26 1.03 0.38

. 0.39, l.04. !.25. l.Zl. l.34. 1.08. 1.32. l.07. 1.33. 1.20. 1.24: I.ID. 0.41.

l.D.

0.8. *l.4. -0.2. -0.7. 0.3. 0.7. -J.3. *l.4. *1.4. -1.5. 6.7. 6.7.

J.71 1.21 1.16 l.35 1.19 l.32 1.18 1.32 1.19 1.35 1.16 l.Zl D.71

. 0.72. l.23. 1.15. l.32. l.18. 1.32. 1.18. l.31. 1.17. 1.32. l.13. 1.24. 0.76.

l.D.

l.D. -0.5. -2.0. -0.8. 0.0. 0.3. -0.8. *l.3. -J.9. *2.6. 1.7. 6.7.

A J.34 1.02 l.09 1.30 1.08 1.52 l.13 I.ID 1.13 1.32 l.07 1.30 1.09 1.02 0.34

. 0.37. 1.05. l.10. 1.28, 1.04. 1.29. l.12. l.09. 1.11. 1.30. 1.05. l.26. l.06. 0.99. 0.37.

b.8. 3.3.

l.O * -1.2. -2.9. *2.2. -l.2. -0.9. *l.5. *l.2. -2.D. -3.3. *3.0. -2.6. 7.7.

0.39 0.93 1.26 1.02 l.32 1.18 I.JO 1.03 l.10 1.18 1.32 1.02 1.26 0.93 0.39

. 0.42. 0.97. l.28

l.01

1.28

1.16. l.09. l.Ol. l.08. 1.16. 1.28, 0.99. 1.23. 0.94

0.42.

6.8. 4.5. 1.0. -0.8. -2.7. -1.5. -1.4. -l.4. -1.9. -1.3. -2.7. -3.0. -2.7. l.5. 7.7.

0.34 l.02 1.09 l.30 l.08 1.32 1.13 l.10 l.13 1.32 1.07 1.30 1.09 l.02 0.34

. J.37. l.04. 1.07. l.Z9. 1.08. l.31. l.ll. l.09. l.ll. 1.26. 1.05. l.29. 1.09. l.06. 0.37.

6.8. 2.4. -!.9. -1.0

0.5. -0.6. -1.4. *0.8. *1.4. -4.2. -2.5. *l.l. -O.l. 3.6. 7.3.

D.71 1.21 1.16 l.35 1.19 l.32 1.18 1.32 1.19 1.35 1.16 1.22 0.72

. D.70

1.19

1.15

1.35

1.19

l.28

1.17. 1.28

l.16

1.32

1.16

1.24. 0.74.

. -1.9. -t.9. -o.6. o.4. o.4. -3.3. -o.8. -2.6. -2.a. -1.1. -o.z. t.8. 4.1 *

. 0.38 1.03 1.26 1.21 1.35 1.08 1.32 1.08 1.35 1.22 1.26 1.03 0.38

. 0.39. 1.05. 1.27. 1.21

1.32. l.04. l.27. 1.07. 1.34, 1.21. 1.28. 1.06. 0.40.

1.8, l.8. 0.9. -0.3. *l.5. -3.3. -3.3. -0.8. -0.8. -0.3, 0.9. 2.4, 3.6.

0.42 0.90 1.26 l.16 1.30 I.OZ 1.30 1.16 1.26 D.90 0.42

. 0.45. 0.93. 1.26. 1.13. l.26. 0.99. 1.27. 1.14, 1.27. D.93. 0.44.

5.5.

5.2. -0.3. -2.1. -3.2. -3.l * *2.6. -1.8. 0.5. 3.2. 3.6.

STANDARD DEVIATION

=2.040 0.42 1.03 1.21 1.10 1.26 1.10 1.22 1.03 0.42

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I Section 5 PRIMARY COOLANT ACTIVITY The specific activity levels of radioiodines and radioactive noble gases in the primary coolant are important to core and fuel performance as indicators of failed fuel and are important with respect to offsite dose calculations associated with accident analyses.
Two mechanisms are primarily responsible for the presence of radioiodines and radioactive.
noble gases in the primary coolant.
These fission products are always present due to direct fission product recoil from trace fissile materials plated onto core components and fuel structured surfaces or trace fissile materials existing as impurities in core structural materials.
This fissile material is generally referred to as "tramp" material, and the resulting iodines are referred to as tramp iodine. Fission products will also diffuse into the primary coolant if a breach in the cladding (fuel defects) exists.
Fuel defects are generally the predominant source of radioiodines and radioactive noble gases in the primary coolant.
Surry Technical Specification 3.1.D conditionally limits the primary coolant radioiodine dose equivalent I-131 to a value of 1.0 µCi/gram with provisions that ultimately limit the dose equivalent I-131 activity to a maximum of 10.0 µCi/gm. 2 Figure 5.1 shows the dose-equivalent I-131 activity history for Cycle 12.
These data show that the dose equivalent I-131 activity remained substantially below 1.0 µCi/gm throughout Cycle 12 operation.
The cycle average steady state power dose equivalent I-131 NE-978 S1C12 Core Performance Report Page 47 of 58
concentration was 6.28 X 10-3 µCi/gm which is less than 1% of the full power Technical Specification limit.
Correcting the I-131 concentration for tramp iodine involves calculating the I-131 activity from tramp fissile sources and subtracting this value from the measured I-131.
The resultant tramp-corrected I-131 activity is theoretically the I-131 activity from defective fuel.
The magnitude of the tramp-corrected I-131 can then be used as an indication of fuel reliability (the average tramp-correct~d I-131 activity for a month is generally referred to as the fuel reliability indicator) as well as assisting in quantifying the extent of fuel cladding defects.
The monthly fuel reliability indicator through September 1993 generally remained below 5 X 10-4 µCi/gm.
For PWRs, this is considered to be a typical fuel reliability indicator level for a reactor core with no fuel defects.
The fuel reliability indicator increased above 5 X 10-4 µCi/gm for the remainder of 1993 having a final fuel reliability indicator of 8.33 X 10- 3 µCi/gm when the cycle ended in January 1994.
An increase in the fuel reliability indicator of this nature indicates the presence of a cladding defect or defects.
The fuel cladding defect(s) became more readily apparent during Cycle 12 late in September 1993.
The noble gas activity in the RCS increased sharply indicating a cladding defect event (see Figure 5.3). The measured (not tramp-corrected) I-131 RCS activity (Figure 5.4) began to noticably increase in November 1993.
The manner in which the RCS coolant activity increased, (i.e., a noble gas activity increase followed much later in time by increasing iodine activity) indicates the defect(s) were either NE-978 SlC12 Core Performance Report Page 48 of 58 I
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small or slowly forming.
Large defects typically manifest themselves by nearly simultaneous increases in noble gas and iodine RCS activity.
A failed fuel action plan was issued in November 1993.
The principle
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element of the plan was to perform fuel inspections during the subsequent refueling outage to ensure no fuel assemblies with cladding defects were I
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reinserted for use in Cycle 13.
The fuel UT exams resulted in widely varying indications of suspect cladding failures, primarily in fuel assemblies scheduled to be discharged.
Upon reviewing the UT data, there was not sufficient evidence suggesting the presence of cladding defects in any fuel assemblies other than those being discharged.
Given the I
uncertainty in the fuel UT data for these assemblies, Westinghouse I
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I (Westinghouse was not the fuel UT vendor) agreed to perform additional fuel exams on the suspect discharged fuel assemblies to help determine their status and potential failure mechanism.
to begin in the 2nd quarter of 1994.
The exams are scheduled The ratio of the specific activities of I-131 to I-133 is used to characterize the type (size) of fuel failure which may have occurred in the reactor core. Use of the ratio for this determination is feasible because I-133 has a short half-life (approximately 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />) compared to that of I-131 (approximately eight days).
For pinhole defects, where the diffusion time through the defect is on the order of days, the I-133 decays leaving the I-131 dominant in activity, thereby causing the ratio to be roughly 0.5 or more. In the case of large leaks and tramp material, where the diffusion mechanism is negligible, the I-131/I-133 ratio will generally be less than 0.1. The use of these ratios with regard to defect
~E-978 S1C12 Core Performance Report Page 49 of 58 J
size is empirically determined and generally used throughout the commercial nuclear power industry. Figure 5.2 shows the I-131/I-133 ratio data for the Surry 1 Cycle 12.
As seen on Figure 5. 2, the ratio began increasing when the fuel defect occurred, but never really attained equilibrium.
Therefore, the use of the iodine ratio as a defect size indicator is perhaps not applicable.
However, the characteristic nature of the increase in noble gas and radioiodine RCS activity suggests the defect(s) to be small.
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I Section 6 CONCLUSIONS The Surry 1, Cycle 12 core has completed operation.
Throughout this cycle, all core performance indicators compared favorably with the design predictions and the core related Technical Specification limits were met with significant margin.
No significant abnormalities in reactivity or burnup accumulation were detected.
Evaluation of the radioiodines and radioactive noble gases in the RCS indicate that a fuel cladding defect or defects occurred in late September 1993.
Fuel inspections were conducted during the subsequent refueling outage to preclude inserting defective fuel assembies into Cycle 13.
The fuel exams indicated that the defective fuel rod(s) existed in the discharged batches of fuel.
Additional fuel assembly examinations are scheduled to begin in the second quarter of 1994.
NE-978 S1Cl2 Core Performance Report Page 55 of 58
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I Section 7 REFERENCES
1)
E. A. Hoffman, "Surry Unit 1, Cycle 12 Startup Physics Test Report," Technical Report NE-898, Rev. 0, Virginia Power, July, 1992.
2)
Surry Power Station Technical Specifications, Sections 3.1.D, 3.12.B and 4.10.
3)
T. W. Schleicher, "Virginia Power Fuel Assembly Burnup and Isotopics Calculation Code Manual," Technical Report NE-726, Rev. O, Virginia Power, February, 1990.
4)
D. L. Gilliatt, "The Virginia Power FOLLOW Code Manual,"
Technical Report NE-679, Rev. 1, Virginia Power, April, 1991.
5)
W. D. Leggett, III and L. D. Eisenhart, "INCORE Code,"
WCAP-7149, Westinghouse, December, 1967.
6)
Letter from B. C. Buckley (NRC) to W.L. Stewart, "Surry Units 1 and 2 - Issuance of Amendments Re: F-Delta-H Limit and Statistical DNBR Methodology (TAC Nos. M81271 and M82168)",
Serial No.92-405, dated June 1, 1992.
7)
W. M. Oppenheimer, "Reload Safety Evaluation Surry 1 Cycle 12 (Pattern CP)", Technical Report NE-874, Rev. O, Virginia Power, February, 1992.
NE-978 SlC12 Core Performance Report Page 57 of 58
REFERENCES (cont.)
3)
\\;. 11. Oµpenheimer, "Reload Safety Evaluation Surry 1 Cycle 12 (Pattern CP)", Technical Report NE-874, Rev. 1, Virginia Power, April, 1992.
9)
G. R. Pristas, "Reload Safety Ev.aluation Surry 1 Cycle 12 (Pattern CP) 11
, Technical Report NE-874, Rev. 2, Virginia Power, September, 1992.
10) P. D. Banning, "Surry Unit 1 Cycle 12 Design Report",
Technicai Report NE-881, Rev. 0, Virginia Power, March, 1992.
11) "Surry 1 Cycle 12 TOTE Calculations", Calculational Note PM-425, Rev. 0 and associated addenda, Virginia Power.
12) "Surry 1 Cycle 12 Flux Map Analysis",
Calculational Note PM-437, Rev. 0 and associated addenda, Virginia Power.
13) D. M. Chapman, "Surry 1, Cycle 12 FOLOW Input and Calculations",
Calculational Note PM-440, Rev. 0, Addendum C, Virginia Power, February, 1994.
~E-978 S1Cl2 Core Performance Report Page 58 of 58 I
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