ML20082T465

SAR for Fuel Reload 3 (Segment 9 - Cycle 4)

ML20082T465
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Site:	Fort Saint Vrain
Issue date:	05/31/1983
From:	GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
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ML20082T428	List: ML20082T427 ML20082T447 ML20082T465 ML20082T482
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GA-C17128, NUDOCS 8312160025
Download: ML20082T465 (67)

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ATTACHMENT 2

. REPORT GA-C17128 l

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GA-C17128 SAFETY ANALYSIS REPORT EUR EUEL RELOAD 3 (SEGMENT 9 - CYCLE 4)

. FORT ST. VRAIN MJCLEAR g GENERATING STATION GA TECHN2AGIES PROJECr 1900 MAY 1983

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. 00NTERIS PElft

l. INIRGXJCTION AfD StMERY l-1

2. RFACIOR OPERATING HISTORY 2-1

3. GE2ERAL DESCRIPTION 3-1

4. FUEL SYSTEM DESIGN 4-1 4.1. Fuel Design 4-1 4.2. Mechanical Design 4-1 4.3. Thermal Design 4-3 4.4. Fission Product Release 4-3

5. NUCLEAR DESIGN 5-1 5 .1. Segment 9 Fuel Ioading 5-1

, 5.2. Burnable Poison Ioading 5-3

. 5.3 . Control Rod Sequence 5-5 5.4. Projected Cycle 4 Operation 5-7 5.5. Maximum Control Rod Worth 5-8 5.6. Core Shutdown Margin 5-11 5.7 . Kinetics Parameters 5-14 5.8. Analytical Input 5-15 5.9. Core _ Operating Procedures 5-15

6. THERMAL-HYDRAULIC DESIGN 6-1

7. SAFEHY Al& LYSIS 7-1 7.1. Introduction and Sumary 7-1 7.2. Ioss of Normal Shutdown Cooling 7-2 7 .3 . Moisture Inleakage 7-3 7.3.1. Steam-Graphite Reaction 7-3 7.3.2. Hydrolysis of Failed Fuel 7-4 7.3.3. Fission Product Release from Oxidized Graphite 7-5 7.4. Permanent Ioss of Forced Cooling 7-4 (Design Basis Accident No. 1) iii

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EAgfL 7.5. Rapid Depressurization/ Blowdown 7-7 (Design Basis Accident tb. 2) 7.6. Conclusions 7-9

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS 8-1

9. STAR 7UP TESTS 9-1

10. REFERENCES 10-1 TABLES 4-1. Segment 9 Calculated Peak Operating Conditions Versus 4-5 FSAR Initial Core Peak Values

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5-1. FSV Reload 3 (Segment 9) As-Built Fuel Ioadings 5-16 a

5-2. Comparison of Bland 14 and Segment 9 Blend Uranium 5-17 and ihorium Ioadings 5-3. Projected Core Inadings at the End of Cycle 3 5-18 5-4. Use of Burnable Poison in Segment 9 5-19 5-5. Cbntrol Rod Sequence for Cycle 4 5-20 5-6. Calculated Control Rod Group Worth and Power Peaking 5-21 Factors with Cycle 4 Rod Sequence 5-7. Sumary of Control Rod Insertions and Axial Power 5-22 Factors in Bottom Fuel Iayer-e 0

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. 2Agn 5-8. Ratio of Flux Level in Center of Core to Average Core 5-24 Flux Level at Iower Power IAvels 5-9. Control Rod Bank Worth 5-25 5-10. Shutdown Margins - Cycle 4 5-26 5-11. Kinetics Parameters 5-27 7-1. Potential Effects of Cycle 4 on FSV FSAR Accident 7-10 Predictions g FIGURES 3-1. Core regions refueled in reloads 1 through 6 3-4 3-2. Refueling region age distribution for the equilibrium 3-5 cycle (before refueling) 5-1. Segment 9 poison rod using surplus Segment 7 IBP rods 5-28 5-2. Identification of control rod groups 5-29 5-3. Tilt envelope for Cycle 4 5-30 5-4. Tenperature defect vs. average core tenperature 5-33 4

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! , 1. INTRCDUCTION AND

SUMMARY

This safety Analysis Report (SAR) is prepared to obtain concurrence to operate the Fort St. Vrain Nuclear Generating Station (FSV) through the forthcoming reload cycle (Cycle 4). For this cycle, 6 of the 37 fuel j regions in the core will be loaded with fresh fuel elements fabricated by General Atomic Company (GA). Se introduction of new fuel elements is consistent with the fuel management program described in the Final Safety Analysis Report (FSAR). Se regions to be refueled contain 203 standard fuel elements, 30 control fuel elements, one neutron source fuel element, and six bottom control fuel elements, a total of 240 elements to be replaced. Se duration of Cycle 4 could be up to 300 effective full power days (EFPD).

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, This report contains sections describing the operating history of the reactor through April 30, 1983, the fuel system design, the nuclear

)* design, the thermal-hydraulic design, and the safety aspects of the core

during Cycle 4. Proposed changes to the Technical Specifications are presented. Se planned startup program for the refueled core is also l briefly described.

he replacement fuel elements feature one design change (use of H-451 graphite) relative to the fuel design described in the FSAR, the Technical Specifications, and previous reload SARs. Bis design change was the subject of a lengthy generic review and approval by NRC in 1978 i and 1979.- Se safety evaluation for the change as it affects this co're

{ reload is presented in this report.- No unreviewed safety questions, as defined in 1CCFR50.59, are presented. None of the peak operating conditions presented in the PSAR are' exceeded.

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, 2. REACTOR OPERATING HIS'IORY Initial criticality of the FSV reactor was achieved on January 31, 1974, with initial generation of electricity on December 11, 1976. Prior to February 1,1979, when the plant was shut down for refueling, the initial core had operated a total of 174 EFPD. Cycle 2 operation began on May 26, 1979 and was completed on May 13, 1981, having acctsnulated a total of 189 EFPD. Cycle 3 operation began on July 15, 1981, and as or

April 30,1983, had accumulated a total of 170.5 EFPD.

! 'Ihe nuclear performance of the FSV core has been, in general, as predicted. Good agreenent between measurenents and calculations has been

, obtained for shutdown margins, temperature coefficients, xenon worth, and  ;

, control rod worth (i.e., measurements are well within the acceptance

. criteria specified for the tests). Initial cold criticalities in Cycles 1-3 were predicted within 0.003 ok. Analyses have overpredicted the ,

end-of-cycle (EOC) reactivities of the core at operating temperatures by a few tenths of a percent; however, the difference between observed and expected reactivity has renained within the 0.01 ok limit of Technical Specification LCD 4.1.8 throughout operation.

Fission product release to date has been very low. Measured circulating activity has been approximately a factor of 30 less than the '

limit provided in Technical Specification LOO 4.2.8. Measurements of-plateout activity obtained after removal of the first plateout probe in November,1981 indicate that these activity levels are also substantially-below Technical Specification limits (aef.1) .

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1 i Se most unusual occurrence, to date, was the detection, initially in .

October 1977, of taperature fluctuations. Rese fluctuations atfected

{ the nuclear channels, the region exit temperatures, and the steam

! generator module taperatures. During fluctuations, however, the total core coolant flow and core thermal power remained essentially constant.

1 In addition, the temperature swings during fluctuations stayed within plant operations and technical specification limits.

A comprehensive program to evaluate and resolve the fluctuation issue was begun in late 1977. 21s progem led to installation, in November 1979, of core region constraint devices (RCDs) (Ref. 2) . Rese devices limit the small (approximately 0.10 inch) lateral movements of fuel columns to which the fluctuations were attributed. Fluctuation testing of the core up to 1006 power with aCDs installed was completed in ,

November,1981. No fluctuations have been detected since installation of a the RCDs. Se results of these tests were formally subnitted to NRC in -

July,1982 (Ref. 3) . ,

he second major issue with regard to reactor operation has been the existence of discrepancies between measured and calculated region outlet helium temperatures. Significant discrepancies have been limited to regions in the northwest boundary of the core (Regions 20 and 32-37),

with measured taperatures being consistently less than calculated temperatures. Rese discrepancies are caused by a transverse flow of relatively cool helium from the core-reflector interface along the inside of the region outlet thermocouple sleeves (Type II flow) . 21s flow passes over the region outlet themaple assemblies of these regions and depresses the indicated region outlet taperature.

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To compensate for these discrepancies, special operating procedures were provided which insure compliance with the original core design intent. Appropriate technical specifications were developed to govern operation with these measurment errors and were sutnitted to NRC in July,1982 (Ref. 4) .

On October 5,1982 NBC issued Amendnent No. 28 to the FSV Operating License. In this amendment, the NRC concluded, based upon a review of Reference 3, that the fluctuation issue is resolved. 'Ibe technical specifications proposed in Reference 4 were incorporated in the operating license, and all previously inposed restrictions on reactor power level were reoved.

, As a result of the visual examinations conducted on fuel elements reoved from the reactor during the second refueling, two Segment 2 fuel

, elments were each found to have one or two cracked graphite webs. '1he presence of these cracks did not affect the cooling gemetry of the fuel or the ability of the fuel handling machine to safely reove the fuel elements frce the core. PSC has kept the NRC apprised of the status of these cracked webs (Ref. 5) . A DOFr-funded progra is being carried out at GA to investigate this issue, while a similar NRC-funded progra is being conducted at Los Alamos National Laboratory.

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3. GENERAL DESCRIPTION l

Re FSV fuel management schee is designed so that approximately one-sixth of the core is reloaded at periodic refueling intervals. 21s document describes the third reload segment (also known as Segnent 9) to be inserted into the FSV core. It is projected that Segment 9 will reside in the core within the 1800 EFPD limit of Technical Specification LOO 4.1.1. Eis reload fuel segment is designed so that the core performance with the new fuel added satisfies the reactor Technical Specifications. Rese limitations apply to the total core performance.

That is, not only do the freshly loaded refueling regions ineet power distribution limitations, but the perturbations to the remainder of the

. core are such that the segments reaining in the core from the previous cycle also satisfy the performance requirements. Rese performance

. requirenents include core excess reactivity, shutdown margins, power distribution behavior, and all the core safety considerations discussed in the FSV e M .

About one-sixth of the core is replaced at each refueling. Se scheduled refueling sequence is stamarized in Fig. 3-1.* It can be seen that six refueling regions are reloaded at each refueling, except for the fifth reload, at which time the central refueling region is also replaced. Segnent 9 consists of 203 standard fuel elements, 30 control fuel elements, one neutron source element, and 6 bottom control fuel elements. In this reload segnent, three of the refueling regions are located in the central portion of the core (Regions 3,13, and 18)

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• Figures and tables appear at the end of each section.

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i and three are located adjacent to the side core-reflector interface -

(Regions 22, 29, and 33) . A new neutron source will be placed in the top active core layer of Region 22 in order to assure an adequate count rate for the startup range detectors. 'Ibe refueling region sequence was chosen so that freshly refueled regions are never adjacent to each other (except when Region 1, the central region, is reloaded). 'Iherefore, each refueling region is surrounded by regions of varying ages.

Figure 3-2 shows the refueling region age distribution for the equilibrium cycle as given in FSAR Section 3.5. By comparison of this figure with Fig. 3-1, it can be seen that tne reload sequence follows that given in the FSAR.

Segnent 9 features one change relative to the fuel design described -

in the FSAR, the Technical Specifications, and previous reload SARs *

(Refs. 6 and 7) . Because the H-327 graphite used in the inital core and -

in Segnents 7 and 8 is no longer available, some of the fuel elements of ,,

Segnent 9 (Regions 3,13, and 18) will use H-451 graphite. 'Ihe remaining fuel elements in Segnent 9 (Regions 22, 29, and 33) will use the renaining stock of H-327 graphite. Use of B-451 graphite requires a change to Technical Specification DF6.1, as discussed in Section 8.0 of this report.

Use of H-451 graphite in FSV reload segnents was the subject of a lengthy generic rwiew and approval by NRC in 1978 and 1979 (Refs 8 -

16). As a result of that review, NRC concluded that substitution of H-451 for H-327 graphite elements in FSV would result in negligible changes in the nuclear and thermal behavior of the core and would not result in reduced safety margins or reliabililty compared to the

reference H-327 core (Ref.16) . It was also concluded by NRC that the e

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, information provided to NRC by GA & ring that review nay serve as an acceptable reference and the initial basis for allowing the substitution l of H-451 graphite fuel and reflector elements for the reference H-327 l elenents.

l It was also noted by the NRC in Reference 16 that a final decision to use H-451 graphite in FSV would be based upon an NRC review of an

application from Public Service Company of Colorado describing the exact i

composition of any reload fuel utilizing H-451 It was also required in Reference 16 that, as part of any application for insertion of H-451 reload elenents, reports be provided on the results of the ongoing irradiation creep progran and on post-irradiation examinations of the H-451 fuel test elements, which were inserted in FSV & ring the first refueling. Information on graphite creep and on the examination of test

, element PIE-1, the only test elenent using H-451 graphite renoved from

, the reactor to date, was provided to NRC by PSC in Reference 17.

' Itis Safety Analysis Report provides specific information on the reload fuel which will utilize H-451 graphite and provides the safety evaluation in support of the change to Technical Specification DF6.1.

Extensive reference is made to the information in References 8 - 16.

As shown in the subsequent sections of this report, evaluations of this design change have shown that it has no adverse inpact upon core performance or plant safety. Accordingly, it presents no unreviewed safety questions as defined in 10CFR 50.59.

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37 21 17 2 10 16 5 11 30 28 RELOAD NO.1 RELOAD NO.4 (SEG 7) (SEG 10) 36 .

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15 4 25 14 27 RELOAO NO.2 RELOAD NO. 5 (SEG 8) (SEG 11) 20 18 19 33 22 f 6 3

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RELOAD NO.3 , RELOAD NO.6 '

,, (SEG 9) (SEG 12) t Fig'. 3-1. Core regions refueled in reloads 1 through 6 ,

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- Fig. 3-2. Refueling region age distribution for the equilibrium cycle (before refueling)

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4. WEL SYSTEM IESIGN j 1

1 4.1 WEL DESIGN The fuel elenents in this third reload are, except for the use of H-451 graphite, of the same basic design as those in the initial core  !

and Segnents 7 and 8. 'IRISO-coated (TVD)C2 and ThC2 particles are

bonded by a carbonaceous matrix into fuel rods which are cured and I

loaded into graphite fuel blocks. 'Ihe mnfiguration of these fuel )

elenents is identical to that used in Segnents 1-8. %e materials and processes used in manufacture are essentially the same as thom for the initial core fuel elenents or for previous reload fuel elements

, (Refs 6 and 7) . Fuel and burnable poison loadings have been adjusted

, to ac<=rAare the reactivity requirenents of the cycle, as discussed

, in Section 5.

As was shown in References 8-16, H-451 graphite is a better structural material than H-327 graphite due to its near-isotropic nature, its lower irradiation-induced axial dimensional changes, and its higher strenath. he effects of the use of H-451 graphite on core performance are discussed in subsequent sections. Although this design change requires a change to the Technical Specifications, as shown in the following sections it does not involve any unreviewed safety questions as defined in 10CFR50.59.-

4.2 MECHANICAL DESIGN l

ne fuel elements for Segnent 9 of the FSV core are fabricated i

from both H-327 and H-451 graphite. We configuration of these fuel G

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elenents is identical to that used in segnents 1 through 8. Table 4-1 provides a sunmary of the Segnent 9 fuel element stress, strain, and bowing analyses described in this section.

Stress analysis was performed for the FSV Segnent 9 fuel elements (both H-451 and H-327 graphite) using the methods discussed in the l FSAR. Oprating and shutdown strain and stress distributions were calculated for the axial and radial orientations thoughout the lifetime of the fuel elements. All operating and shutdown stresses were less than thcoe predicted for the initial core fuel elements (FSAR Section 3.4.2.1.1) except for the radial operating stress in one core location. In this location, the control colunn of Region 13, a maximun radial stress of 217 psi was calculated - a value which is 8.5% higher than the FSAR peak value. However, the fuel blocks in e

Region 13 will be made from &451 graphite. As discussed in Reference ,

8, the radial strength of n-451 graphite is 66% higher than that of H-327 graphite. Berefore, despite the 8.5% higher calculated stress '

in Region 13, which results fran the higher Young's mnrhilus of &-451 graphite (Ref. 8), a larger design margin exists than would have been ot:tained with the use of H-327 graphite. Bis result is consistent with the conclusions reached in References 8-16.

l During core operation, the Segnent 9 fuel elements will be exposed to fast neutron irradiation, which will induce dimensional changes in the graphite. An analysis was performed to calculate the expected dimensional changes of the Segnent 9 fuel elements. Se maria n axial dimensional changes of all Segnent 9 fuel colunns were less than those-of the initial core fuel colunns shcWn in FSAR Section 3.4.2.1.2. Se maximum calculated ECL bowing for' the buffer elements'is 0.09-in.,

4 which war; the maximum bowing predicted for the' initial and equilibriun core H-327 buffer elements (FSAR, Section 3.4.2.1.2) .

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, 4.3 'HIERMAL DESIGN he selection of fuel rod and burnable poison loadings and of the control rod progra for Cycle 4 (see Section 5) is made so that the Cycle 4 power distribution falls within the limits described in the FSAR. No changes are planned for the operation of the core cooling during Cycle 4 (i.e., helium taperature at the core inlet and average '

outlet temperature will be enveloped by the FSAR reported values). -

Accordingly, the temperature limits presented in the FSAR will not be exceeded during Cycle 4. 21s conclusion is supported by analyses using the (IE code (Ref.18), which is discused in the FSAR. Se results of these analyses are also shown in Table 4-1.

. Introduction of H-451 graphite in the fuel design for Segment 9 results in increaed graphite thermal conductivity. W e increased thermal conductivity results in a maller tmperature rise across the graphite web and in lower fuel centerline temperature than wculd be obtained using H-327 gr@hite. Reductions in fuel temperature will lead to lower kernel migration rates and a resulting increased margin relative to the core thermal safety limit (Ref. 8) .

4.4 FISSION PRCOUCP RELEASE During Cycle 4, the FSV core is expected to be operated within the limit.s presented in the FSAR and contained in the Technical Specifications. Accordingly, the fission product- release characteristics of the fuel are expected to be within design limits, and the design radionuclide inventories presented in Section 3.7 of the FSAR will not be exceeded. Rese conclusions are consistent with operating experience gained during Cycles 1-3.

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As discussed in References 3-16, the release cf noble gas and -

gas-like (Se, Te, I) fission products is expected to be reduced with the use of H-451 vs H-327 graphite in the core. te reduction in gasecus fission product release will result fran the lowering of fuel tenperature discussed in Section 4.3 and its effects on R/B, the gaseous fission product release-to-birth rate ratio in the fuel.

Negligible changes in the release of metallic fission products are expected due to the decreases in fuel tenperature. As discussed in References 8-16, the available data indicate that the sorptivity and diffusivity of elements such as strontitan are essentially the same for both H-327 and H-451 graphites. Hence, changes in graphite type will not affect the release of metallic fission products.

%us, the circulating and plateout activity source terms presented

• in Tables 3.7-1 and 3.7-2 of the FSAR remain appropriate for use in ~

FSV accident analyses. .

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. TABLE 4-1 SEGMENT 9 CALCUIATED PEAK OPERATING CONDITIONS VERSUS FSAR INITIE CORE PEAR VALUES Parameter FSAR Peak Value Segment 9 Peak Value(a)

Axial operating stress (psi) 450 324 Radial operating stress (psi) 200 217(b)

Initial axial thermal stress (psi) 150 64 Initial radial thermal stress (psi) 180 117 Coltan axial strain (%) 2.1 1.8 Buffer element bowing (in.) 0.09 0.09 Fuel temperature (cF) 2300 2148(c)

(a) Values calculated using FSAR methods.

. (b) See discussion in Section 4.2.

(c) Peak fuel temperature in core during Cycle 4.

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~ 5. MJCLEAR DESIGN l l

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5.1 SH; MENT 9 FUEL IDADING l In the initial core design, the fuel was zoned both radially and axially to achieve the desired power distribution and to mock up the equilibrium cycle. %e core was divided into four radial zones and two axial zones. We radial zones consisted of the central refueling region, the six refueling regions adjacent to the central regions, the 12-refueling regions adjacent to them six regions, and the outer 18 refueling regions adjacent to the side reflector. In addition, the five rows of fuel rods in the outer 18 regions that were inmediately adjacent to the side reflector interface contained a buffer fuel loading with low

, uranium and high thorium loading to reduce the power peaking at the

, reflector edge. %ere were two axial zones consisting of the top and botton three layers of fuel elements. %e buffer zone was not axially zoned. Rese various fuel zones, combined with a partial mock-up of the equilibrium core fuel distribution, made it necessary to fabricate 13 different fuel blends or~ compositions for.the initial core fuel elements.

For the design of the reload segnents, these different fuel zones are essentially maintained, except that the second and third radial zones are

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combined and the width of the tuffer zone is increased to the width of a.

fuel element. Since the central refueling region is not reloaded in this reload segnent, only six new fuel mmpositions or blends (numbered 26 through 31) are used for the Segnent 9 fuel. . Refueling regions 3,13, and 18 require a top and botton fuel loading, and refueling regions 22, 29, and 33 require a top and bottom fuel loading. In addition, -two loadings are required for the buffer fuel at the core-side reflector interface, which is axially zoned as-it was in Segnent 8 (Ref. 7) .

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A summary of the 6 different fuel blend uranium and thorium loadings .

used for the third reload segnent is given in Table 5-1. Fuel blends 26 and 27 are used in the inner refueling regions, and blends 28, 29, 30, and 31 are used in the outer refueling regions. Blends 30 and 31 are used in the txiffer zone.

In addition to the six new fuel blends for Segnent 9 discussed above, one spare Segment 7 fuel element will be used in Segnent 9. ' Ibis spare Segment 7 fuel element is a standard element containing blend 14 fuel in an H-327 graphite body. As shown in Table 5-2, in which heavy metal loadings of blend 14 fuel are cxxnpared with those of Segnent 9 blends, the blend 14 loadings are closest to the loadings of blend 30 in Segment

9. Blend 30 is used in the upper half of the buffer fuel zone.

Accordingly, the spre Segnent 7 fuel element will be located in the -

buffer zone in the top active core layer. Since the heavy metal loading -

of the spare Segnent 7 element is slightly lower than the blend 30 -

elenent it replaces, its power generation will be slightly lower. 'Ihe .

resulting change in power peaking factors has been shown to be negligible, and the effeet on core reactivity has been shown to be undetectable within the precision of the design methods. By placing the spre Segnent 7 eletent in the top active core layer, perturbaticns upon fuel performance will be further minimized, since fuel temperatures at this location are the lowest in the core. Furthermore, by placing the spare Segnent 7 elenent in the buffer zone, it will be mixed with other fuel elenents that have H-327 craphite bodies. 'Ihis placement is consistent with the statenent in Reference 8 that any axial layer within a region will be compomd of the same type of graphite.

It is anticipated that the reactor will have operated for up to 663 EFPD, generating a total of up to 1.3 x 107 MW-br of energy at the time e

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. of the third refueling. Se projected heavy metal loadings in the renaining core segnents at the EOC 3 (663 EFPD) are given in Table 5-3.

Se maximtra burnup in fissile particles is projected to be about 14.0%

FIMA and in fertile particles about 2.0% FIMA. %ese burnups are

. substantially lower than the limiting values given in the FSAR, Appendix A, Table A.1.1-2. he maximum projected fast flux (E > 0.18 MeV)

, exposure in the discharged segnent is about 3.2 x 101 2 nyt.

5.2 BURNABLE IOISCN LOADING T

Six holes are provided in each standard fuel element, one at each corner, for ine rtion of burnable poison reds; four holes are provided in each control fuel elenent. For Segnent 9 fuel, there will be no burnable poison in the control fuel elements; all six poison holes in the standard

. elements may be used depending on burnable poison loading requirenents.

i In the initial core and Segnent 7, different poison rod types, varying in their boron loading, were used to provide reactivity control and to adjust power distributions in fresh fuel regions. S e poison rods were 28.5 inch long. In the case of Segnent 7, it was necessary to have i available several different poison rod types, only two of which were

ultimately used. he poison rods to be used were choun based en the

! reactivity requirenents of the core at the EDC 1 and were inserted into the reload segnent just prior to refueling.

To avoid this unnecessarily canplex and costly procedure, the design of the poison rods was slightly changed for Segnent 8 and later segnents l (Ref. 7) . Instead of using single rods 28.5 inch long, shorter rods l (1.98 inch long) are loaded .into the poison holes in combiration with graphite dirnmy spacer rods (also 1.98 inch long) . Each poison hole

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either contains 14 poison rods, a cabination of 14 poison rods and .

spacers, or is lef t empty. Se boron loading in each elenent is controlled by varying the cxxnbination of poison rods and spacers. Se new poison rods are the same in dianeter (0.45 inch) and made of the same graphite as the 28.5 inch single rods previously used. Se spacer rods are also 0.45 inch in dianeter and are made of HLM graphite. Bis change

! does not require any changes to the Technical Specifications, nor does it 1

constitute a change to the core design as described in the FSAR.

For Segment 9, 28.5 inch poison rods left wer fran Segnent 7 will be cut into 1.98 inch naminal length rods and loaded into the standard fuel elenents-in combination with graphite spacers. De poison rods left wer fran Segnent 7 which will be used in Segnent 9 are knwn as " type 6" and

" type 9" rods. Sey differ only in their boron loadings. Se standard .

fuel elements in the upper core half will be loaded with 14 poison rods .

in each of the six poison holes. In the lwer core half, each of the six .

poison holes in the standard fuel elements will be loaded with five poison rods, four sEncer rods, and five more poison rods, in that I

sequence. We poison holes in the control fuel elenents will be left empty. Se use of burnable poison rods in Segnent 9 is sunmarized in Table 5-4, and the loaaing sequence for the type 6 and type 9 rods is- l shw n in Fig. 5-1.

2e burnable poison loading for Segnent 9 described above was used for the Cycle 4 depletion analyses referred to in the follwing sections.

his loading will be used in Segnent 9 unless the reactivity character-istics of the core or the length cf Cycle 3 deviate significantly frca current projections.

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i 5.3 CONTRCL ICD SEQUENG Technical Specification LCD 4.1.3 states that a control rod sequence will be specified for each fuel cycle and that the sequence will always l

i be followed, except for rod insertion resulting frcm a scran or rod runback or during low-power physics testing. Se control rod sequence for use during Cycle 4 is given in Table 5-5. Se identification of the control rod groups is shown in Fig. 5-2.

i he regulating rod is located in the central refueling region (rod group 1). 21s group is partially withdren before criticality is

, achieved and then maintained in its most reactive control rod position

} for the remainder of the operation. In this manner, minor reactivity j ,

adjustments can be made most rapidly with the minimum amount of control i, rod motion. Wis is consistent with the method of operation utilized for

, the control rods in previous cycles.

A stanmary of the alculated power peaking factors obtained using the control rod sequence for all control rod configurations is given.in Table 5-6. his includes all of the control rod configurations in which the control rod groups are either fully inserted or withdrawn, including thom suberitical configurations during the withdrawal cf the first few control rod groups. Any configuration with a partially inserted control I rod group will have peaking factors lying between those calculated when l

that group is fully inserted and fully withdren. S e control rod worths and power peaking factors in Table 5-6 were calculated at the beginning of cycle (BOC) with equilibritan xenon (5 EFPD). Rod worths at middle of cycle (MOC) and EOC will be essentially unchanged fran those given in the table. Power peaking factors during Cycle 4 depletion are discussed in Section 5.4.

O S

e 5-5

'Ihe control rod configurations shown in Table 5-6 may be separated

• into four categories of reactor power operation: (1) full power, (2) 20%

to 100% power, (3) 0% to 20% power, and (4) subcritical. rt11-power operation may be achieved with configurations ranging fra one control rod group (three rod pairs) fully or partially inserted at the EDC 4, when the excess reactivity is relatively low, to three control rod groups (nine rod pairs) fully or prtially inserted at the DOC 4, when fission product poisons such as xenon are not present. (Except as allowed by Technical Specification LCD 4.1.4, only one control rod group in addition to group 1, the regulating rod, will be prtially inserted at any time.)

At less than full power, in the range of exit gas temperatures between 1460o and 95C9F, when the xenon level and the core temperature are lower, configurations ranging from four control rod groups (12 rod "

pairs) to five control rod groups (15 rod pairs) fully or prtially inserted can be expected, depending on the tenperature, power level, and _'

the rate at which power was increased from the previous level. -

i The third category covers the rise-to-power Etiase fran the cold critical condition to about 20% power (gas outlet temperatures 1950oF).

Different limiting operating conditions are applied to this phase of reactor operation by the Technical Specifications. 'Ihe initial cold criticality, following the refueling operation, was calculated to be achieved with seven rod groups (21 rod pairs) fully inserted, the regulating rod 115 in. withdrmn, and control rod y wp 2D 95 in, withdrawn.

The last category covers the subcritical control rod configurations where region peaking factors (RPF) or intraregion tilts are not meand yJful, and consequently, are not given in Table 5-6.

E o

5-6

Frm the data given in Table 5-6, it can be seen that the calculated power peaking factors for the various power levels do not exceed thom given in Technical Specification LCD 4.1.3. 21s is true for both the radial region peaking factors and the intraregion peaking (coltunn tilt) factors. %e radial region peaking is below 1.83 for all configurations involving less than 19 control rod pairs. For lm-power operation (gas outlet taperature 19500F), when more control rod pairs are inwrted, the region peaking is below 3.0 until the subcritical configurations are reached. In the same manner, the intraregion peaking factors are also acceptable. It is clear, therefore, that use of H-451 graphite has no adverse impact on power peaking factors, a result which is consistent with the conclusions of References 8-16.

5.4 PRQ3ECTED CYCLE 4 OPERATION

~

Wis section presents the results of Cycle 4 depletion analyses using design methods discussed in Section 3.5 of the FSAR. Fuel and burnable pison loadings discussed previously were used as input (see Sections 5.1 and 5.2).

Figures 5-3a through 5-3c present envelopes encompassing projected RPFs and column tilts during Cycle 4 depletion. Se results indicate that RPFs and tilts during Cycle 4 will be well within the allowable limits set by LCD 4.1.3 and that the use of H-451 graphite has no advere impact on radial power distributions. Bis result is mnsistent with the conclusions of References 8-16.

Axial zoning of the Segment 9 fuel and burnable poison is provided (1) to produce a power distribution which tends to reduce axial fuel taperature peaking and (2) to maintain the desired axial power l

5-7 l

distribution with depletion. Se calculated axial power factors in the .

bottan layer of each fuel region during Cycle 4 are shown in Table 5-7.

The calculations were carried out with the GA'IT code, the three-dimensional whole core model which is used in the semi-annual fuel accountability analyses. It can be seen that the IID 4.1.3 limits on peaking factor in the lower fuel layer are not exceeded and that the use of H-451 graphite does not result in unacceptable axial power peaking, as was also concluded in References 8-16.

We basis of Technical Specification LCD 4.1.3 also states that an acceptable flux distribution shall be maintained at lower power levels by keeping the flux level in the center of the core at least as high as the average level. Table 5-8 shows the ratio of the flux in the inner core regions (Regions 1-19) to the core average flux for each of the three -

control rod configurations in Table 5-6 which can result in operation .

between 0% and 20% power during Cycle 4. W e flux ratio is above 1.0 for -

all cases, consistent with the basis of LCD 4.1.3. ,

5.5 MAXIMJM (DNTRCL RCD WOR'IH The basis of Technical Specification 100 4.1.3 states that the accidental renoval of the mavi== worth single rod pair shall result in a transient with consequences no more severe than the withdrawal of 0.012 ok, at rated (i.e. , 100%) power, frcan a core which has a tenperature defect between 2200F and 15000F of 0.028 ok. In addition, the calculated worth of any rod pair in any configuration with the reactor critical must be less than 0.047 ok. We rod withdrawal accident (IWA) at full power evaluated in Section 14.2 of the FSAR asstunes withdrwal of a control rod worth of 0.012 & at equilibrium EDC with an equilibrium EDC temperature defect cf 0.028 ok. Se consequences of 1

l withdraal of this 0.012 & EOC control rod, because of the lower O

9 O

5-8

i i

reactivity feedback at EOC resulting fran the less negative temperature coefficient at EOC, are equivalent to the consequences of withdrawal of a rod worth of about 0.016 & at BOC. Because the cx)nsequences of an IWA are a function of rod worth, steady-state core temperature (i.e., initial power level), and tenperature coefficient (which varies slightly with burnup during the cycle), it is necessary to evaluate control rod worth as a function of control rod insertion.

We control rod withdratal sequen for Cycle 4 was described in Section 5.3. For this sequence the maximum control rod wortns are shown in Table 5-6. tese rod worths were calculated for the core with equilibrium xenon at the beginning of Cycle 4. Additional calculations indicate that these =vi== rod worths do not change significantly with -

burnup during the cycle. As discussed in Section 5.3, the control rod configurations shown in Table 5-6 may be divided into four categories of reactor power operation: (1) full power, (2) 20% to 100% power, (3) 0% to 20% power, and (4) subcritical.

The results in Table 5-6 indicate that the maximum worth rod pair in any critical configuration during Cycle 4 is 0.026 d, which is less than the 0.047 6 limit of LOO 4.1.3.

Table 5-6 also indicates that during Cycle 4 at full power, with two control rod banks fully inserted, a =vi== control rod worth of about 0.013 $ is obtained, which is larger than the 0.012 & rod worth evaluated in the FSAR for an EDC rod withdrawal accident. It is, however, very unlikely that the reactor could be operated at the end of Cycle 4 at full power with two rod banks fully inmrted. Analyses with i the GAUGE code indicate that at EOC, even with Pa-233 or xenon fully  !

decayed, the reactor would have to operate at less than 70% power to be l

I O

l 5-9

critica) with two banks fully inserted. We consequences of an M A at -

this lower power level are less severe than thom discussed in the FSAR.

Early in Cycle 4, when the reactor can be critical at full power with two rod banks fully inserted, the consequen s of an MA at full power are also less severe than those discussed in the FSAR. his result is to be expected since, as described above, a maximum rod worth of 0.016 & at BOC is equivalent to a maximun rod worth of 0.012 e at EOC with regard to NA consequences, and the 0.013 $ mni== rod worth for Cycle 4 is within this BOC value.

Ac stated above, it is very unlikely that the reactor could be critical at 100% power at end of Cycle 4 with two mntrol rod banks fully innerted. Eb11 power criticality at IOC with this rod configuration could be attained only if the reactor power were increased very rapidly -

fra zero to 100% power (no xenon buildup) following a prolonged shutdown -

(all Pa-233 decayed) . Plant operating requirenents (steam generator

• boilout, turbine warming, etc.) preclude this situation. Nevertheless, .

the consequences of an 0.013 $ WA at full power end of Cycle 4 were evaluated using the Cycle 4 power dist.ributions discussed in Section 5.4 and the Cycle 4 kinetics parameters presented in Section 5.7. Se evaluation was performed with the same methods (the BLOOST code) used in the FSAR N A analym s.

Se results of these NA evaluations indicate that the consequences of accidental withdrawal of a 0.013 ok control rod at end of Cycle 4 are l the same or less than thom for the equilibrium core N A' discussed in the FSAR. Bis result is obtained because the region peaking factors expected to occur during Cycle 4 are sanewhat analler than those assuned for the FSAR analysis and because Cycle 4 kinctics parameters are sanewhat less severe than thom for the equilibrium core. Fuel e

9 e

5-10

i d

ternperature in an average power channel is expected to increase by about 4000C for both MA cases. Fuel temperature in the hottest channel is expected to increase only about 9000C for the Cycle 4 NA, vs. about 18000C for the FSAR case.

Table 5-6 also indicates that with five or six rod banks fully inarted a - i - control rod worth of 0.015 $ is obtained. Reactor criticality with this control rod configuration can be obtained at EOC only at very low (a few percent) power. At BOC criticality may occur with that configuration at full power. However, this 0.015 & rod worth is less than the 0.016 & rod worth evaluated in the FSAR for a rod withdrmal accident at BCC.

Eerefore, it can be seen that the use of H-451 graphite has no adverse inpact upon control rod worths or upon the consequences of postulated rod withdrawal accident. 21s result also is consistent with conclusions reached in References 8-16.

5.6 CORE SHUIDOWN MARGIN '

i i

Technical Specification 100 4.1.2 requires that the reactor be capable of being shut cbwn (at least 0.01 & subcritical) with any one control rod pair withdemn at room temperature. Eis is consistent with the assumption that, during the rod withdraal accident postulated in Section 14.2 of the FSAR, the control rod pair being withdren is continuously withdrawn until fully renoved during the accidenc and is not capable of being reinarted. Se reactor design criteria require that .

the reactor be capable of being shut down at refueling temperature with any two control rod pairs inoperable for up to two weeks. Se requirenent for being shut cbwn with any two control rod pairs withdemn is established to allow for operation with one inoperable control rod e

G 5-11

a pair. In addition, IID 4.1.2 requires that cold shutdown be achievable -

i prior to Pa-233 decay with any two rod pairs, or with any two adjacent rod pairs plus a third rod pair that is at least three regions away fra the two unrodded regions, removst.

The net negative reactivity insertions following a scram during Cycle 4 are shown in Table 5-9. Excess reactivity during Cycle 4 is -

projected to vary from 0.053 $ at BOC, to 0.029 ok at MOC, to 0.014 &

at EOC, based up GAUGE code analyses.1 Thus the mininun instantaneous shutdown margin during Cycle 4 is obtained at BOC with the two maximum worth rods inoperable: 0.082 d. Bis shutdown margin is reduced af'er shutdown due to cx>re cooling (in a matter of hours), due to xenon decay (in a matter of a few days), and due to decay of Pa-233 to 0-233 (in a matter of weeks). -

To show that the shutdown margin is satisfactory for these cases with -

one or more control rod pairs assumed inoperable, the stanmary in Table .

5-10 gives calculated core shutdown margins at BOC inmediately after

refueling (5 EFPD) and at IOC and EOC. It can be seen that for all of the various cases of interest, the shutdown margin is larger than 0.016, as required in the Technical Specifications and reactor design criteria. Accordingly, the use of H-451 graphite has no adverse inpact upon snutdown margins. mis result is also consistent .with the conclusions reached in References 8-16. _

I I

l 1As noted in Section 2, core reactivity has been overpredicted during c. ,

Cycles 1-3 by a few tenths of a per nt. If this trend continues, as -

expected, during Cycle 4, shutdown margins will be larger than thom ,

calculated in this section by the same amount. -

5-12

4 For the case with only the maximLrn worth rod pair inoperable, the minimLan shutdown margin at room tenperature with xenon and Pa-233 decayed (the most reactive case) is 0.042 d, which is obtained at BOC (5 EEPD).

At refueling temperature (2200F), this value would be 0.048 e for the same condition. However, since the second maximum worth rod out can add as much reactivity as 0.040 ok, it would not be possible to maintain a shutdown margin of 0.01 ok with these two rod pairs out if the Pa-233 is allowed to fully decay to U-233. mis situation does not conflict with the LOO, however, since the requirement for the two rod pairs out at refueling temperature is established to allow for operation with one inoperable mntrol rod drive assembly. An adequate shutdown margin can be maintained for a period of at least two weeks in this core condition.

If the inoperable unit cannot be regaired in this time period, the reserve shutdown systen can be used to maintain adequate reactivity 4

control. Se occurrence of this minimum shutdown margin at BOC is primarily due to the unusually long shutdown for refueling that was asstaned in these analyses (60 days instead of the 14 days assumed in FSAR analyses). As a result of this conservative asstanption most of the Pa-233 present at end of Cycle 3 is decayed to 0-233 at beginning-of

! Cycle 4, thereby reducing the shutdown margin. Furthermore, the l probability of having two inoperable control rods at BOC is quite small j since the shutdown time between Cycles 3 and 4 provides ample opportunity to confinn that all control rod drives are in working order.

~

General requirenents for operability of control rod drives during the cycle are provided in Technical Specification LCD 4.1.2.

1 O

e j-

• 5-13 I l

5.7 KINETICS PARNE'IT.RS The kinetics permeters for Cycle 4 as well as for the initial and equilibrium cycles (taken fran the FSAR) are given in Table 5-11. %e data in this table indicate that the equilibrium cycle kinetics parmeters represent a conservative estimate of the Cycle 4 kinetics.

The Cycle 4 kinetics permeters were used in the INA analyses described in Section 5.5.

'Achnical Specification IID 4.1.5 requires that the reactivity change due to an average core temperature increase between 2200F and 15000F (refueling taperature to rated power conditions) in the absence of xenon must be at least as negative as 0.031 ok. Bis requirement is inposed because FSAR accident analyses assmed a taperature defect cf 0.028 ok -

and the uncertainty in measured temperature defect is about i 10%, or ,

about 0.003 ok. Se temperature defect measurments required by -

Technical Specification SR5.1.3 have confirmed that the agreement between measured and calculated temperature defect is within i 10%.

Se calculated taperature defect at the beginning and at the end of Cycle 4 is shown in Figure 5-4. We results indicate that the temperature defect between average core temperatures of 220oF and 15000F is 0.044 ok at BOC and 0.031 ok at EDC. Both of these calculated values acet the requirement of LCO 4.1.5 for measured temperature defect.

Certain conservatims were employed in the calculation of these taperature defects such that expected values are larger than those presented in Figure 5-4. It was assmed that the core is in the cold l critical condition and that control rods are then reoved to achieve a i

higher power level and fuel temperature. Since some power is being generated in this scenario, it is inevitable that sme buildup of xenon e

D e

5-14

occurs. %e presence of xenon tends to decrease the calculated t aperature defect. Bere. fore, the temperature defects in the absence of xenon, the limit on which is specified in LCD 4.1.5, will be larger than the values shown in Figure 5-4.

Fra the results presented in this section, it can be seen that use of H-451 graphite has no adverse inpact upon kinetics Eurameters, another result consistent with the conclusions of References 8-16.

5.8 ANEYTICE INPUT I

Nuclear analyses were carried out using the same methods applied to the analyses presented in the FSAR, previous reload SARs, and the semi-i annual fuel accountability reports. %e design of Segment 9 introduces

~

no new aspects to high-tenperature gas-cooled reactor (BIGR) core design techniques; consequently, there was no need to develop or adapt any new methods or procedures for the nuclear design.

We depletion analyses described in this chapter were performed by.

simulating the actual core power history for Cycles 1 and 2 and the first 150EFPD of Cycle 3. Continuous operation at 70% power for the balance of Cycle 3 and at 100% power for Cycle 4 was assumed. Cycle 3 was assmed to continue to a total length of 300 EFPD. Cycle 4 was assmed to be 300 EFPD in length.

5.9 CDRE OPERATING PROCEDURES Core operating procedures will be the same as those for previous cycles and those planned for equilibrium cycles. Se only differences will be the control rod withdrawal sequence discussed in Section 5.3 ard the use of comparison regions as described in Reference 4 and required by  !

Amendnent No. 28 to the FSV Operating License.

e 3

1 5-15 4

-- . - ,n

TABLE 5-1 .

FSV RELOAD 3 (SEGMENT 9) AS BUILT EVEL LOADINGS Fuel Uranium Loading Thorium Loading HigDd gm/ rod Toral(kg) am/ rod Toral(ka) 26 0.349 64.6 3.46 640.7 27 0.212 39.0 2.80 514.4 28 0.430 36.9 3.52 302.2 29 0.258 21.9 2.92 247.9 30 0.323 26.0 3.98 320.4 31 0.194 15.5 3.11 248.9 Total .203.9- 2274.6-I e

e

• 9 4 '

5-16

4 TABIE 5-2 COMPARISON OF BLEND 14 AND SB3 MENT 9 BIEND URANIUM AND 'IBORIUM LOADINGS Elend 14 Difference, %*

Seamant 9 Blend Uranium Tnorium 26 -15 +3 27 +30 +22 28 -42 +1 29 +15 +18 30 -7 ,

31 +34 +23

• I Blend 14 - Seamant 9 Blend \ .

Segment 9 Blend x 100%

O e

4 5-17

TABLE 5-3 -

PROJECTED CI)RE LOADINGS AT 'IHE DID OF CYCLE 3(a)

Nuclide Weight (kg)/Segnent Total Nuclide Nuclide 3(b) 4 5 6 7 8 Weight (kg)

'1h-232 2546.9 2337.8 2807.6 2335.3 2225.2 2252.2 14505.0 Pa-233 + U-233 42.7 41.4 49.5 41.6 33.8 25.7 234.8 U-234 4.0 4.2 4.9 4.2 3.0 2.2 22.5 U-235 44.6 50.9 56.4 50.7 87.5 116.5 406.6 U-236 13.2 15.8 17.5 15.8 15.5 11.9 89.7 .

U-238 6.6 7.7 8.6 7.7 9.9 10.5 51.1 ,

Np-239 + Pu-239 0.16 0.19 0.21 0.19 0.25 0.26 1.25 -

Pu-240 0.06 0.08 0.09 0.08 0.09 0.07 0.46 .

Pu-241 0.06 0.07 0.08 0.08 0.07 0.04 0.41 (a) 663 EFPD (b) This segr.ent is discharged W

l l

5 l 1

~

TABLE 5-4 USE OF BURNABLE POISON IN SEGMENT 9 r

Upper Core Half (a) Imer Core Half (D)

Bnat, g/an3 (type 6 IBP rods) 0.0215 0.0215 Bnat, g/an3 (type 9 ISP rods) 0.0336 0.0336 Naminal rod dianeter, in. 0.45 0.45 Nominal rod length, in. 1.98 1.98 Rods / standard element (type 6) 42 30 Rods / standard elenent (type 9) 42 30 Rods / control elenent 0 0 (a) Fuel blends 26, 28, and 30, and the spare Segment 7 (blend 14) elenent.

(b) Fuel blends 27, 29, and 31.

3 l

1 e

G j -

5-19 i

TABLE 5-5 -

CONTRG, RCD SEQUENCE FOR CYCLE 4 Group Sequence Withdrawn Regions 1 2A(a) 2,4,6 2 4F(a) 25,31,37 3 4D 23,29,35 4 1(115" out) 1 ,

5 4B 21,27,33 6 2B 3,5,7 7 4E 24,30,36 ,

8 4A 20,26,32 9 4C 22,28,34 ,

10 3C 10,14,18

~

11 3A. 8,12,16 12 3B 9,13,17 13 3D 11,15,19 14 1(fully out) .1 J

(a) Rod groups used for rod runback.:

5 l ..

l 9

5-20 ,

TABIE 5-6 CRIUIATED 00mROL RCD GROUP WORNI NO POWER PEAKING FACIORS WITH CYCLE 4 RCD SEQUDKE (BOC, EQ. XE, SEEPD)

Control Rod Group Cumulative Maxinum Maximum Maximum Configuration Worth Worth Maximum Tilt (b) Tilt (b) Worth Rod Rods Inserted ok ok RPF(a) Rodded Unrodded Worth ok Reg No Rods In 0.0000 0.0000 1.38 1.26

-1 Rod (1 half in) 0.002 0.007 1.32 1.25 0.002 1 4 Rods (+ 3D)(c)  ; 1(d) 0.017 0.01f, 1.52 1.19 1.24 0.008 15 7 Rods (+ 3B) 0.021 'O.040. 1.47 1.18 1.21 0.013 9 10 Rods (+ 3A) , 0.014 0.054 1.69 1.31 1.29 0.012 9 1 13 Rods _ (+ 3c) d2 0.020 0.074 1.54 1.25 1.23 0.012 18 17 'F 16 Rods (+ 4C) h 0.012 0.006 1.65 1.31 1.24 0.015 8 s'%if w 0.094 2.03 1.54 1.38 0.015 19 Rodc (+ 4A) 0.008 15 aw.. j p3 0.010 0.104 2.69 1.60 1.32 0.014 13 5 3,;g

. - . -7.2;l Rods (+ 4E)

. a . . . '25, Rods (+ 2B) '

O.033 0.137 1.80 1.38 1.36 0.026 22 k8 Rods (+ 4B) 1 0.012 0.149 j[.. 28 Rods j (1 fully 'in) 0.004 0.153 _

?gf '31 Rods (+ 40) g L4 0.004 '0.157 0.169 34 Rods _(745)N O.012 4 =0,046 0.215

%. 37 Ro,Gs (+ 2A)

.. j 4 N0 TIE: Initial criticality.at 0 days was calculated with bank 2B withdrawn 95 in.

~

, . _Ax n Q'i- w(a)RPF ^idhonpehkin@ factor =regionpower/coreaverageregionpower.

p;g A . (b) TILT = coltnin peakirig factor /RPF. i ,

(c) Refers to 'r6d broup'sequenco'(4 -rode; =< rod 1'+ group 3D) .

pg.

(d) Power range defined,in Section 5.3.

g_

g2.f _ [w y , y ~-

= -

~ .y #: -

J E :a -n .l

_ - .x . _

. - = _ . .

e TABLE 5-7 (Sheet 1 of 2) -

SUMMARY

OF CONTRCL RG) INSERTIONS AND AXIAL IMER FACIOdS IN BOTim MJEL LAYER CYCLE 4 4 4 4 4 EFPD 5.0 50.0 100.0 150.0 155.0 REGION *

! 1 2 .838 2 .826 2 .847 2 .848 2 .831 2 0 .778 0 .763 0 .7 84 0 .784 0 .760 3 0 .740 0 .721 0 .737 0 .734 0 .720 4 0 .727 0 .717 0 .738 0 .741 0 .719 5 0 .737 0 .724 0 .746 0 .746 0 .725 6 0 .790 0 .774 0 .795 0 .794 0 .772 7 0 .7 86 0 .771 0 .791 0 .789 0 .762 8 0 .697 0 .689 0 .714 0 .717 'O .700-9 3 .913 3 .893 3 .910 3 .906 0 .691 10 0 .695 0 .681 0 .701 0 .701 0 .695 11 6 .68f 6 .681 6 .705 6 .708. 3 .905 12 0 .748 0 .750- 0 .779 0 .7 87 0 .770 13 3 .854 3 .850 3 .878 3 .883 0 .664 .

14 0 .729 0 .727 0 .753 0 .757 0 .746 15 6 .666. 6 .655 6 .675 6 .675 3 .881 -

16 0 .728 0 .718' 0 .740 0 .740 0 .736 .

17 3 .868 3 .848 3 .864 3 .857 0 .638 i

18 0 .677 0 .669 0 .689 0 .690 0 .681 .

19 6 .690 6 .684 6 .707 6 .711 3 .906 l 20 0 .715 0 .718 0 .745 0 .750 0 .745 21 0 .705 0 .699 0 .723 0 .726 0 .647 22 0 .657 0 .641 0 .655 .0 .654 0 .588 23 0 .698 0 .694 0 .716 0 .719 0 .717 24 0 .691 -0 .6 93 0 .720 0 .726 0 .822 25 0 .592 0 .5 95 - 0 .622 0 .631 0 .725 l 26 0 .709 0 .715 0 .745 0 .753 0 .756 l 27 0 .791 0 .797 0 .830 0 .838 0 .739 l

28 0 .700 0 .702 0 .731 0 .738 0 .657 29 0 .584 0 '.579 0 .599 0 .601 -0 .602

/ 30 >

0 .686 ' 0 .687 0 .711 0 .713' O 810 31 . i O' .690 -0 .690 0 .718 0 .722 0 .822 h.., \

32 'O .590 0 .576 0 .601 0 .608 0 . 615 h1 33 34 0 .639. 0 .623

+ D< .735 )' O .732 0

0

.639

.757 0

0

.637

.759 0 .576 0 . 686.

l n'- 'T 35  ; 0 ~ . .628 0 .636 0 .666. 0 .676 0 .679 j 36 ,t- ;O .567 0 .574 0 .603 0 .614 ;O '.695 37 #) 0 .680 0 .685 0 .713 0 .719 0 .306

-* Control rod insertion /detth measured in active core layers (e.g. , "6" = fully inserted, ' "0" = fully withdrawn, "2" = 33% -inserted) . ..,

/

I ., h y V (. .

.a<

, .( #

i s

, 5-22

d.  ?.

yLas .

l ..

A

$J ?b?.

e e

TABLE 5-7 (Sheet 2 of 2) i

SUMMARY

OF CONTRG. RCD INSERTIONS AND AXIAL IMER FACIORS IN BCTICM EUEL IAYER CYCLE 4 4 4 4 EFPD 200.0 250.0 295.0 300.0 REGION 1 2 .834 2 .826 2 .83 1 2 .7 91 2 0 .766 0 .761 0 .764 0 .726 3 0 .721 0 .714 0 .717 0 .678 4 0 .718 0 .704 0 .707 0 .667 5 0 .722 0 .706 0 .709 0 .668 6 0 .775 0 .767 0 .771 0 .732 7 0 .7 67 0 .763 0 .767 0 .729 8 0 .707 0 .707 0 .711 0 .674 9 0 .708 0 .710 0 .715 0 .678 10 0 .695 0 .690 0 .6 93 0 .655 11 3 .893 3 .878 3 .880 4 .878 i 12 0 .770 0 .757 0 .762 0 .723 13 0 .670 0 .658 0 .663 0 .625

• 14 0, .744 0 .731 0 .737 0 .698 15 3 .868 3 .850 3 .852 4 .845 16 'O .737 0 .733 0 .739 0 .701 17 0 .652 0 .654 0 .660 0 .624 18 0 .6 86 0 .6 86 0 .692 0 .6 57 19 3 .899 3 .893 3 .895 4 .896

20 0 .751 0 .751 0 .756 0 .718 21 0 .661 0 .664 0 .672 0 .638 22 0 .598 0 .601 0 .611 0 .5 80 ,

23 0 .718 0 .715 0 .722 0 .6 84 24 0 .813 0 .803 0 .808 0 .767 25 . .0 .719 0 .709 0 .716 0 .675 26 0 .756 0 .747 0 .753 0 .713- '

27 0 .747 0 .739. 0 .746 0 .7 08 28 0 .665 0 .658 0 .667 0 .632 .

29 0 .602 0 .5 96 0 . 607 0 .572 30 0. 7 97 0 .7 82 0 788 0 .747-4 31 -0 .811 0- .801 0 .808 0 .767-32 0 .618 0 .620 0 .631 0 .5 96 33 .c 0 .5 86 0 .5 91 -0 .6 03 0 .573 34 0 .699 0 . 7 04 'O .712 0 .679 35 0 .6 86 0 .6 90 0 .698 0 .661 36 -0 . 6 93 ' 0 .694 0. 701 0 .662 37 0 802 0 .798 0 .802 0 ,.763; -

IE 4.1.3, limits: Fully inserted or fully withdram: 0.90 Partially inserted: 1.23 e

m  %

.c . ,

j  %.%

4 9 ,. . .

_ N,

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

i

. i i

i

, TABIE 5-8 RATIO OF FLUX LEVE IN CENTER OF CDRE 70 AVERAGE CDRE ]

l 1

FLUX LEVEL AT I4WER 10WER LEVELS l Number d Rods innareaA* . Flux Ratio 19 1.15

', 22 1.33 25 1.03 See Table 5-6.

.i I

i a

+

1 f

r.

I 3 ' ,

1 .

go e

. ,k l

l '

l l -

'd 4 G

5-24

a

. TA3LE 5-9 00NTR1 IKX) BANK WOR'IH l Cycle 4 BOC(a) MCC EOC Total bank worth, ok (37 rod pairs inmrted) 0.215 0.216 0.217 Total bank worth, ok, less maximum worth rod pair 0.175 0.167 0.161 Total bank worth, ok, less two maximum worth rod pairs 0.135 0.133 0.130

'Ibtal bank worth, ok, less three inoperable rod pairs WA(b) 0.131 0.128 (a) Xenon asstned at equilibrium at SEFPD.

(b) Pa-233 is not equilibrated at BOC; thus, the three rod criterion of LCD 4.1.2 is not applicable.

e 4

e 5-25

TABLE 5-10 SHUTDJWN MARGINS - CYCLE 4 Shutdown Margin, ok Nmber of inoperable Rods BOC(a) E EOC 0(b) 0.085 0.096 0.114 1(b) 0.042 0.0 43 0.054 2(c) 0.011 0.031 0.046 2(d)  !VA(e) 0.030 0.046 3(d) tyA(e) 0.028 0.045

~

(a) Xenon asstned at equilibrim at SEFPD, 60 days prior refueling outage.

(b) Core at rom taperature with complete Pa-233 decay.

(c) Core at refueling taperature with two-week Pa-233 decay.

(d) Core at rom taparature with no Pa-233 decay.

(e) Pa-233 is not equilibrated at BOC; thus, the two rod and three -

rod criteria of 140 4.1.2 are not applicable.

HZIE: General requirements for operability of control rod drives during the cycle are provided in Technical Specification LOO 4.1.2.

W D

4 5-26

. . mu 5 .

KINETICS PARAMETIERS .

Initial Core Cycle 4 Equilibrium Cycle DOC, with Xe Eoc BOC, with Xe EOC BOC, with Xe EOC Fractional productions From U-233 0.0 0.19 0.33 0.48 0.38 0.48 From U-235 1.0 0.81 0.67 0.52 0.62 0.52 Proupt neutron lifetime, sec Hot 2.69 x 10-4 3.17 x 10-4 2.85 x 10-4 3.41 x 10-4 Cold 2.43 x 10-4 2.81 x 10-4 2.64 x 10-4 3.09 x 10-4 Effective delayed neutron fraction 0.00650 0.00577 0.00522 0.00465 0.00505 0.00451 Delayed neutron decay constant,A, sec-1 Precursor 1 0.01243 0.01249 0.01249 0.01252 0.01250 0.01251 2 0.03050 0.03088 0.03120 0.03160 0.03126 0.03164 3 0.1114 0.1136 0.1171 0.1204 0.1170 0.1199 T

y 4 0.3013 0.3025 0.3044 0.3064 0.3047 0.3068 5 1.136 1.136 1.135 1.135 1.135 1.135 6 3.013 2 .981 2.940 2.892 2.913 2.859 Delayed neutron fraction,S Precursor 1 0.000214 0.000219 0.000219 0.000222 0.000220 0.000222 2 0.001424 0.001 0.001214 0.001121 0.001186 0.001099 3 0.001274 0.001 0.001073 0.000983 0.001046 0.000961 4 0.002568 0.00 0.001956 0.003685 0.001874 0.001619 5 0.000748 0.0005 0.000543 0.000452 0.000516 0.000430 6 0.000273 0.0002 0.000212 0.000185 0.000204 0.000179

POISON R0D STACK IN:

TOPAXlAL BOTTOM AXIAL i ZONE ZONE 3 I POISON ROD TYPE MEAN BORON CONC.(G/CM 3 1.98 IN. 9 9 0.0215 6

l .9 6 0.0336 9 ,

9 6 6 6 G = GRAPHITE TPACER ROD 6 9 w

G h 6 6 G 6 G 6 G 6 9 9 6 9 6 9 9 9 9 Fig. 5-1. Segment 8 poison rod using surplus Segment 7 LBP rods

, , . . 9 9 9 ,

I I

N f

l, 38 37 35 20 4D 4A ri II 8 34 la -

71 3D 3A

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. kN 33 t; 7 2 9 22

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,- 48 38 28 2 3B 4C ,

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• 4A 3A 2A 1 28 j,;,"

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• gn; 5?

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d,bllll =" . L ;- ,

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. k, .r .;o

/

FUEL REGION IDENTIFICATION l NUMBER Fig. 5-2. Identification of control rod groups e

5-29 i

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1

. I N

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T A s C n I

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5 0 5 0 1 1 0 0 gcc ywo lI l ll l l!l1 1l lIl1 ,l l:1l, E

1.0 -

u u.

0.5 -

TECHNICAL SPECIFICATION

_ LCO 4.1.3 LIMITS:

RPF: 1.83 ,

_ TILT: 1.40 0.0 1.0 ~ 1.1 1.2 TILT Fig. 5-3. Tilt envelope for Cycle 4: (b) partially rodded regions G

i l

1.0 -

vi h $0.5 -

~

TECHNICAL SPECIFICATION LCO 4.1.3 LIMITS:

, _ RPF: 1.83 i TILT: 1.46 I l I i O.0 1.0 1.1 1.2 TILT Fig. 5-3. Tilt' envelope for Cycle 4: (c) fully rodded ret; ions e 6 9 4 5 9

o 8

o o

Y w g J -

o ~

>- 2 co w . Szw oC og e o. w zt ze ww O E*z Ewwo o J e

ooz _

8

~ a w >= W mox Q e e w w o 4 #

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• w .

>"" 4

.w ti e o 8, o "

• Q w w

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. 103:130 3801VB3dW31 5-33

.. _ _ . ~

, 6. 'mERMAL-HYIRAULIC DESIGN As noted in Section 4.3, the thermal design of the segment 9 fuel elenents is essentially the same as that of the initial core. No changes have been made in fuel elenent gemetry. '1he power distributions expected during the fourth cycle are within the envelopes defined by Technical Specification LOO 4.1.3. Hence, except for the opening of crose-flow gaps, as discussed and accounted for in Section 3.6.2.2 of the original and updated FSARs, core coolant flow characteristics are also unaffected. Accordingly, there are no changes in the thermal-hydraulic l

design limits of the Sepant 9 fuel elements fra those of the initial core or the equilibrian core.

, During Cycle 4, the core pressure drop will continue to increase toward the design equilibrium core value of 8.4 psid. Maxinom core

. pressure drop during Cycle 4 is expected to be about 6.3 psid.

O e

e 6-1

7. SAFETI ANALYSIS 7.1 IN'IRCIXJCTION AND

SUMMARY

In this section, events and accidents previously analyzed in Chapter XIV of the Fort St. Vrain FSAR are reviewed to determine if the loading of Segnent 9, which uses H-451 graphite, in FSV could alter the likelihood or consequences of postulated accidents. Se purpose of such a review is to assure that the worst case conditions previously defined for accident analyses, and found to be acceptable during the FSAR review, are not exceeded and that no unreviewed safety questions are presented.

Chapter XIV of the FSAR has been examined to identify those events  !

', which might be affected by the insertion of fuel elements of H-451 graphite into the core. 'Ibe results of this examination are given in 1 Table 7-1. Som events which require a more detailed examination are:

1. Rod withdrawal accidents (NAs).

2. Fuel element malfunctions.

3. Ioss of normal shutdown cooling (limiting case: cooldown l on one firewater-driven circulator).

4. Moisture inleakage.

5. Permanent loss of forced circulation (DBA #1) .

6. Rapid depressurization/ blowdown (DBA #2) .

As indimted in Table 7-1, mas and fuel element malfunctions are discussed in Sections 5.5 and 4.2 of this document, respectively. . It is concluded in Section 5.5 that the neutronic consequences of a WA will be unchanged, and that NA consequences are no more severe than e

4 7-1

those of the postulated IWA described in the FSAR. For " Fuel Element Malfunction," the H-451 graphite fuel elements will experience less stress buildup than some H-327 graphite fuel elements in Segnent 9, and as discussed in Reference 7 the strength of H-451 graphite is higher than that of H-327. Se consequences of fuel element malfunction are no more severe than those described in the FSAR.

hus, a larger safety margin against this accident is provided. 21s result is consistent with the conclusions reached in References 8-16.

In the sections which follow, it is demonstrated for the remaining events (i.e., items 3 through 6) that existing FSAR results of accident analyses conservatively bound any perturbations resulting fran the introduction of H-451 graphite fuel elements. Sese results .

are also consistent with conclusions reached in References 8-16. ,

7.2 IDSS OF NORMAL SHITI1XMN OXLING Loss of normal shutdown cooling is discussed in Section 14.4 of the PSAR. Se limiting case (Section 14.4.2.1, Case B2) of this event was determined to be cooling with one circulator driven by the firewater system.

As shown in Section 5, the introduction of H-451 graphite fuel elements in Segnent 9 introduces no adverse changes in core neutronics values used in FSAR analysis of this accident. Specifically, no adverse inpact on power peaking factors is created (Section 5.3),

there is no adverse inpact on power distribution (Section 5.4), core shutdown margins are not adversely affected (Section 5.6), and kinetic parameters are not adversely affected (Section 5.7). In addition, since the thermal properties of E-451 graphite are the same or better than those of H-327 graphite (Ref. 8), the heat transport analyses in .

the FSAR are not adversely affected. -

O 7-2

Et h erefore, as there are no changes required for either the neutronic or thermal conditions employer

• in the FSAR analysis of this accident, the FSAR conclusions regarding this accident are not invalidated by the use of H-451 graphite fuel elernents in Segment 9.

7.3 MOIS'IURE IEEAKAGE We analysis in Section 14.5.2 of the FSAR considers inleakage into the primary coolant systen from an econanizer-evaporator-superheater subheader or tube or fran the helium circulator bearing water supply.

Of the moisture ingress cases treated in the FSAR, Case 5, a steam generator subheader rupture compounded by concurrent failure of the

, moisture monitor system and dumping of the wrong (non-leaking) steam loop, has the greatest potential for graphite oxidation and fuel hydrolysis in the shortest time following the accident. As shown in the FSAR, Cases 5 and 6 actually show comparable results. To evaluate the potential effect of the application of H-451 graphite in Segment 9 on the analysis of Cam 5, the following phenonena were investigated:

1. Steam-graphite reaction.

2. Hydrolysis of failed fuel.

3. Potential change in fission product release due to (1) and (2) .

7.3.1 S*a-Granhite *"tioru:

As shown in Figs. 6-1 and 6-2 of Reference 19, the steam-graphite reaction rates at various temperatures for H-327 and H-451 are approximately equal within experimental uncertainty over the e

e 7-3

temperature range of interest (7000 to 13000C (12920 to 2372oF)] .

Furthermore, the reaction depandence on fractional graphite burnoff is not particularly sensitive to the change in graphite type. Hence, the came reaction rate equations have been recommended for both types of graphite.

Since the thermal conductivity of H-451 is smewhat higher than that of U-327, graphite temperatures will be smewhat lower, tending to reduce reaction rates. In addition, since Segment 9 contains reference fuel, catalyst (barium and strontitan) concentrations in the fuel blocks will renain within the conservative values already used to calculate the effeet of these catalysts. R us, for a given inleakage of steam or water into the primary coolant system, the amount of core -

graphite which reacts will be approximately the same for either type -

of graphite. Accordingly, the production of 00 and H2 during the

• moisture ingress events will be approximately the same and, thus, ,

primary coolant system pressure rise during moisture ingress will also be approximately equal for either graphite.

7.3.2 Hydrolysis of Failed.Puel Hydrolysis of failed fuel can result in the release of an additional fraction of the noble gas fission product inventory fran the core to the primary coolant system. However, since the primary coolant system remains essentially leak tight after a steam leak accident, this release represents no hazard to the general public.

We hydrolysis reaction with fuel particles occurs only when stean diffuses through the graphite block and reacts with fuel kernels having coatings that failed prior to the moisture ingress event. The rate of hydrolysis and associated release of noble gas fission -

products is dependent upon local fuel temperatures and steam -

concentration. -

7-4 I

l 1

As noted in Section 4.3, the incre,ised thermal conductivity of H-451 will result in somewhat icwer fuel temperatures. Eus, the magnitude of fuel coating failures considered in the FSAR for moisture ingress events should conservatively bound those estimates after the substitution of H-451 graphite.

The slightly lower fuel element graphite temperatures will have only a small effect on the graphite reaction and corresponding depletion of steam concentration; thus, the extent of the fuel hydrolysis reaction during moisture ingress events is expected to be nearly the same for the H-451 graphite in Segment 9 as that predicted for the H-327 graphite which it replaces. terefore, the hydrolysis

, of failed fuel and concurrent noble gas release would be essentially j the same for the H-451 elements and would be bounded by the FSAR l

'. analysis. Further detailed discussion of graphite oxidation and failed fuel hydrolysis, and of the interaction between these phenomena, can be found in Reference 10.  ;

7.3.3 Finnion Pr<virw-t Relamaa from Ovidiwad Graphite -

ne amount of activity released to the primary coolant system fran oxidized graphite is proportional to the amount of graphite reacted and the concentration of fission products within that graphite. he fission product retention characteristics of H-451 graphite are similar to H-327 (Reference 10) and since no change in fuel failure characteristics is anticipated, the fission product concentration in the fuel blocks will be essentially unchanged after the substitution of H-451.

hus, since the amount of graphite reacted is approximately the same for either type, the corresponding release of fission products

. will also be the same.

e 7-5

7.4 PimMANEhrr IDss .or FORCED . COOLING [ Design Basis Accident No.1(DBA-1)]

This hypothetical event asstnes permanent loss of forced circulation of primary coolant helium. 'Ihe FSAR considers consequences of this event in four categories:

1. Thermal results wherein metallic ccuponents might fail.

2. Structural results which might affect the core, reflector, core barrel, or core support structure.

3. Nuclear consequences which affect shutdown capability.

4. Fission product release and offsite doses. .

Because the thermal properties of E-451 graphite are the same or better ,

than those of H-327 graphite (Reference 8), the substitution of H-451 graphite in Segment 9 will not alter the thermal effects adversely, and the conclusien in the FSAR will renain unchanged.

'Ihe structural results mainly concern thermal stresses which might affect the structural integrity of the core supprt and lateral restraint. Such structural effects, external to the core, are not influen d by the nature of the fuel element gr@hite. In addition, there is little concern regarding the fuel blocks and reflector because during this accident, the maximum compressive load in these umgewr.ts is less than 140 kPa (20 psi) at the bottcm of the core, and about half that at the hottest central core region. 'Ibese lox 1s are small compared with the graphite axial compressive stress. In addition, the strength of H-451 graphite is superior to that of E-327 (Reference 8) .

4

.e e

7-6

We nuclear consequences concern reactivity effects whereby the potential reduction of shutdown margin in the overheated core muld result frce capaction and melting of control rods or spatial redistribution of control poison, fission product poisons, or uranium and thorium. She FSAR shows that these effects do not result in an increase in reactivity at any time during the accident. Since no change in fuel type is being considered, the Cycle 4 mre is expected to perform quite similarly to the existing core with respect to fission product and fuel redistribution during a loss of forced cooling (Imt) event. As discussed in Section 5, the presence of H-451 graphite in the core will not have a significant effect on reactivity.

, The offsite doses for DBA-1, reported in Table 14.10-1 of the FSAR, i

1 are within the 10CFR100 guidelines even with the conservative assumption l

\

. of TID-14844 release fractions. Because the fission product retention j capabilities are similar to those for H-327, the substitution of H-451 l graphite in Segnent 0 will have a negligible effect on the fission product release characteristics during the wre heat-up event and, l hence, the FSAR conclusion regarding the event remains unchanged, as was also concluded in References 8-16.

7.5 RAPID NRPMRTZATION/RTMOOM :(DRA-M he FSAR considers three classes of events for loss of primary coolant: pri:r.ry molant leakage (Section 14.7) , ma ri = = credible accident (Section 14.8), and the hypothetical DBA-2, " Rapid Depressurization/ Blowdown" (Section 14.11) . Since the third class envelopes any potential consequences of the other two, it alone was considered for possible impact by the H-451 graphite in Segnent 9.

e to S

7-7

'Ihe aspects of a DBA-2, which are discussed in Section 14.11 of the FSAR, are as follows:

1. Integrity of reactor internals.

2. Continuation of adequate primary coolant circulation.

3. Ingress of air.

4. Effects of operator actions.

5. Vertical thrust on the PCRV.

6. Effect on building pressure.

7. Radiological consequences.

Item 1 concerns the pressure differentials that could develop across reactor components during the depressurization. Item 2 concerns the .

ability of the helitan circulators to continue operation and provide ,

adequate core cooling in the event of a PCRV penetration failure. 'Ibese

• itens obvioiusly are not affected by the presence of n-451 graphite in ,

the core except for the fact that H-451 haa higher strength and, therefore, has greater margins of safety under accident loadings.

Item 3 concerns the possible pathways for air ingress into the PCRV follewing a MA-2. 'Ihe FSAR shows that the operation of a helium purge flow would virtually exclude oxygen from entering the PCRV and that even if the purge were inoperative, the oxygen ingress would be insipificant with respect to heat generation and graphite combustion. Since the mechanical desip of the PCRV and purge system is not being altered and the oxidation reaction rate with air is rapid and insensitive to catalytic effects and the type of graphite, the FSAR conclusions are not altered.

m I@

'7-8 H

k Regarding Its 4, the FSAR describes how automatic reactor scram and alarm would alert the operator to the fact that a leak has developed in the PCRV and to be ready to perform necessary monitoring and system adjustments. W e operator's responsibilities are neither increased nor diminished by the presence of H-451 graphite in the core. Items 5 and 6 concern mechanical processes external to the PCRV and core. W erefore, only Item 7 needs to be further examined.

As discussed in the FSAR, the radiological release fra a postulated DBA-2 consists of essentially all the activity which exists in the circulating primary coolant prior to the depressurization event plus a mall fraction of the activity plated out on the surfaces of the primary

, circuit. Since the fission product transp)rt and retention character-istics of the H-451 graphite are nearly identical to those of B-327 1

. (Reference 10) the radiological consequences of this accident wculd be essentially unchanged frm thoce reported in the FSAR.

7.6 CONCLUSION

S A review of Qiapter XIV of the FSAR identified six postulated accident conditions which required more detailed examination for potential impact frm the substitution of H-451 graphite in Segment 9 for the fuel elment block material. Rese accident conditions were discussed in this section and no requirment for additional analysis has been identified. It is concluded that the worst case conditions previously defined for accident analyses, and found to be acceptable during the FSAR review, are not exceeded as a result of introducing H-451 graphite in Segnent 9. 21s result is consistent with the conclusions of References 8-16. 21s design change presents no unreviewed safety questions, as defined in 10CFR50.59.

0 7-9

D TABIE 7-1 PO'ID4TIAL EFEECTS OF CYCLE 4 W FSV FSAR ACCIDENT PREDICTIONS

,. : ; - . . . :- r : - - n : - - - : : = : : n : : . : - : = - = = = ~ = n, - . . . . ; n - --

Potential Effects on Event Analysis FSAR Qiapter XIV Event Due to Application of H-451 Graphite

.: u nu . . u::- n n ;-- . - .-- :: n n n :n n: : ::n - .:n :: :=::::n:-

14.1 Enviromental Disturbances Earthquake Mechanical effects are discussed in Section 4.2. Any reactivity effect would be bounded by rod withdrawal [

events.

Wind effects L The core is not affected by these Flood / events.

Fire I Landslides Snow and Ice f .

14.2 Beactivity Accidents and -

Transient Response '

Summary of reactivity sources Excessive removal of control '

poison Loss of fission product poisons aeactivity insertions in these Rearrangement of core components ' events are less than rod with Sudden decrease in reactor drmal events, temperature Rod withdrawal accidents Evaluation required, see Section 5.5 of this document.

14.3 Incidents Incidents Involving the Reactor Core Column deflection & misalign-- No change from sect. 3.3.1.2 of FSAR.

ment Fuel element malfuctions Evaluation required, see Section 4.2 of this document.

Misplaced fuel element No change from Sect. 3.5.4.5 of FSAR.

Blocking of coolant channel - No change frca Sect. 3.6.5.2 of FSAR.

Control rod malfunctions No change ~fran Sect. 3.8 of FSAR.

Orifice malfunctions No change from Sect. 3.6.5.1. of FSAR.

Core support floor loss of No change frcm Sect. '3.3.2.2 of FSAR. .

cooling ,

5 Incidents Involving the Primary None Coolant Systen -

7-10

i .

1 TABIE 7-1 (continued) l PO'IENTIAL EFFECTS OF CYCI2 4 CN FSV FSAR ACCIDENT PREDICTIONS

, . - . . . . . . . . . . . - . . . _ - - . . . n ; n-.;- -.; n ., n ; ; - -.n .;

n .

Potential Effects on Event Analysis FSAR Chapter XIV Event Due to Application of H-451 Graphite

:: na n n n : n n ---  ::=,n :: :n:: :n n n n; ::n:: n:n : n:.=:n .n:

i Incidents Involving the Control None and Instrmentation Systen Incidents Involvits the PCRV None Incidents Involving the Secondary None Coolant & Power Conversion S' j t m Incidents Involving the Electrical. Nonne Systen Malfunctions of the halium None Purification Systen Malfunctions of the Helitan Storage None

. Systen Malfunctions of the nitrogen None

. systen 4 l 14.4 Ioss of Normal Shutdown Cooling Evaluation required, see Sect. 7.2 ,

. of this de_==nt.

4 14.5 Secondary Coolant Systen Leakage Steam leaks outside the primary None j coolant systen Leaks inside the primary coolant Evaluation required, see Sect. 7.3

systen/stean generator leakage of this document.

(moisture ingress) i 14.6 Auxiliary Systen Leakage 1 Failure involving the helitan purification systen ,

Loss of both purification Possible effects would be bounded by trains i Design Basis Accident No. 2, FSAR Failure of regeneration line L Section 14.11..

with simultaneous valve failure and operational error <

Accidents involving the gas No change fran Sect.14.6.2 of FSAR.

waste systen Fuel handling and storage accidents Fuel handling accidents 1 No change frca Sect.14.6.3 of FSAR.

Fuel storage accidents f 4

e 7-11 8

, . - - - . . - ,---a > - . , -. - -.

TABIE 7-1 (continued)

PO'IENTIAL EFEECTS OF C?CLE 4 ON FSV FSAR ACCIDENT PREDICTIONS

: . :,: . :::: . ::: :::,: :r. :- . ::::::::r: - : :::

Potential Effects on Event Analysis FSAR Chapter XIV Event Due to Application of H-451 Graphite

. ;n;n2::::::::::::2 ;;c._::; . q m q:;;;;.::::::::::.;;;; .: r::::::: :=::.;;;;;;;

14.7 Primary Coolant Leakage I Possible effects would be bounded by Design Basis Accident No. 2, FSAR 14.8 Maximtsn Credible Accident '

Sect. 14.11.

14.9 Maximum Hypothetical Accident Same as FSAR Sect.14.11 14.10 Design Basis Accident No. 1, Evaluation required, see Sect. 7.4 of

" Permanent Loss of Forad this document.

Circulation (IDEC)" .

14.11 Design Basis Accident No. 2, Evaluation required, see Sect. 7.5 of

" Rapid DepressurizatiorV' this doctanent.

• Blowdown" i

l .

m e

1 7-12

, 8 PROPOSED MODIFICATIONS T:) TECHNICAL SPECIFICATIONS l

One change to the plant technical specifications is necessitated by the insertion of Segment 9 into the reactor core. Technical Specification W6.1, Re A: tor Core Design Features, describes the overall design features of the reactor core which were used to evaluate its 9eneral performance. In the initial issue of the Technical Specifications, and at the time of thu initial CA subnittal to Nhc on use of H-451 graphite in Fort St. Vrain (Ref. 8), Specification N6.1 did not specify the graphite type used in fabrication of the fuel elenents.

, However, on May 25, 1979, four days after issuance of the NRC Safety i Evaluation Report for H-451 graphite (Ref.16), Amerdnent 20 to the FSV

. Operating License was issued by NRC. 'Ihis amerdnent included a change to Technical Specification N6.1 in which the composition of the eight test

, elements that were placed in the core d2 ring the first refueling is described. In the revised N6.1 it was noted that the test elements are fabricated from H-451 graphite, while the reference fuel elements are fabricated from H-027 graphite.

! Accordingly, it is propomd that Technical Specification W6.1 be modified by inserting the following paragraph atter the second paragraph on page 6.1-4:

"Beginning with core Segment 9 (Reload 3), H-451 near-isotropic graphite is used in the fabrication of reload fuel elements in addition to or in place of the previous reference H-327 needle ooke (anisotropic) graphite."

e W

~ 8-1 l

l

i i o j -

, 9. STAR 7UP TESTS Fo11c'.fing refueling, a stepwise approach to full power will be perfonned. During these power steps, the following tests will be

performed

1 1 Ditferential control rod calibration measurenents. Dese tests are required by Technical Specification SR 5.1.5. Se method used for these measurenents is the same as that in SUT B-9.

l t

2 Temperature coefficient (temperature reactivity defect) i measurenents. This test is required by Technical Specification j , SR 5.1.3. Se method used for these measurements is the same as that in SUT B-8.

4

3. Reactivity status surveillance check. Bis test is performed at each startup and once per week as required by Technical Specification SR 5.1.4.

In addition, as requested by NRC in the Safety Evaluation Report for Amendnent 28 of the FSV Operating License, data taking procedures similar .

to those used in the RF500K test series (Ref. 3) will be followed during the startup after refueling. Dese procedures will be used to characterize reactor perfo06ance at oore pressure drops higher than the 5 paid level tested in previous testing up to 1006 power. S e procedures ,

used and the data acquired will be provided to the NRC staff,' for

{

information, in one or more separate submittals.  !

O e'

e i

.9-1 l

...;_ __ , , . _ , , ._ .- __. o .

~

s x 3

. i -

s',

, 10. REFEIECES ,

1 Burnette, R. D. , "Radiochnical Analysis of the First Plateout Probi 5 frm the Fort St. Vrain High-Tmperature Gas-Cooled Reactor," '

GA-A16764, June 1982, PSC letter to NRC P-8243 9, Septelter 27,19L2.

2 l

Fuller, .I. K. (PSC) letter to Willim P. Gam dll (NBC), "SAR for .

s. .

Core Region Constraint Devices," P-79068, Marc, 23, 1979.

\

3. Asmussen, K. E., et;:al., " Testing and Operation or! Fort St. Vrain up to 100% Power," GA-C16701, June 1982, PSC lett.ar ta NRC P-82229, '

July 6,1982. .

g 3 Alberstein, D. , and .K. E. Asmussen, " Technical Specifications for

)

4

. Operation of FSV wiU1 Region ' Outlet Temperature Melsurment s

, Discrepancies," GA-C16781, June 1982, PSC lettar to Mr P-82229, '

s July 6,1982. '

\ ,

s ,

, ',l . ,

Brey, H. T.. (PSC) letter to George Kuzmycz (NBC), " Status of 'gCracked u

• 5 Fuel Element Webs," P-82394, Septaber 15, 1982 w

, l'q,,

6 Fuller, J. K. (PSC) letter to William P. Gammill (NBC), " Extension s

s of Cycle 2 (Raload 1) Operation," P-79025, January 25, 1979.

'x

7. Swart, F. E. (PSC) letter to George Kuzmycz (NBC), " Safety Analysis Report for Fuel Reload 2 (Cycle 3) - Preliminary," P-81012, January 13, 1981.

8. " Safety Analysis Report: Use of H-451 Graphite in Fort St. Vrain

Fuel Elments," GLP-5588, December 20, 1977, G. L. Wesman (GA) N. -

letter to R. P. Denise (ERC), March 28, 1978. '

W 10-1

-s i e x 9. Spels, Semis P. (NBC) letter to G. L. Wesman (GA), " Request for Additional Information on H-451," Septenber 19, 1978.

i \1

! s

. 10. Wessma, G. L. (GA) letter to Willim Gamill (NBC),

October 30, 1978.

- 11. Wessnan, < G. L. (GA) letter to Willim Ganmill (NBC),

M ember 17, 1918.

12. Speis, tenis P. (NRN) letter to George Wessnan (GA), " Request for Additional Information on H-451 Graphite," January 10, 1979.

I 13. Spels, %enis P. (NBC) letter to George Wessnan (GA), " Additional Information Request on H-451 Graphite," February 5,1979.

r 14. Wessnan, G. L. (GA) letter to Willian Gemill (NBC),

February 6,1973. -

15. Wessnan, G. L. (GA) letter to Willian Ganmill (NBC),

February 15, 1979. ,

16. Gmull, Willian P. (NBC), letter to Colin Fisher (GA), " Evaluation of Topical Report GLP-5588", May 21,1979.

17. Brey, H. L. (PSC) letter to Phillip C. Wacper (NBC), "H-451 Graphite Statud," P-83033, January 28, 1983.

18. Katz, R. , and G. R. Malek, " COPE, a Core Performance Code for Gas-Cooled Reactors," GA-9802, N$enber 15, 1969.

19. Engle, G. B. , et.al;, "Develognent Status of Near-Isotropic Graphites for Large if!URs," GA-A12944, June 1,1974.

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P.O. BOX 81608 SAN DIEGO, CAUFORN% 92130

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I ATTACHMENT 3 AMENDMENT 1 TO GLP-5588 p

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9

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RECENED !!CV 14 W OI574: 3UT!01 30RST T. J.

34EY, 4 L.

R:RCHFIEL3, 9.

COLLINS, J. T.

~'-c. Q 3 ~[ f OE'UER 000. :MT.

30LPH11, M.

GA Technologies Inc. FulLEa, c. d.

po BCx 85608 d d*

SAN CtEGO CALtFCANIA 92138 CANN' HOLT 5, 'M.4

$ '5 5

• LBS:77:CRF:83 $$7t.d 2.

N<:vedoer 2,1983 LEE. O. R-Lilli, J.

MClltIDE, L. M.

9tLLE1, C. K.

NFS* CLERK 1!E40FF, M. E.

Director of Nuclear Reactor Regulatica M,,yKg'-)<

U.S. tA: clear Regulatory Ccmission scaC etEq(

Washingtcn, DC 20555 RE Y kE/ kl.ER(

'J . S.N . R . 0.

Dear Sir,

E

@g1hja,3,2, In Fay 1979 the NRC staff cc:pleted its review of GA Tecbr. ologies (for::erly General Attraic Ccepany) topical report GLP-5588, " Safety Analysis Report: Use of H-451 Graphite in Fort St. Vrain Fuel Elements." Upon completion of its review, the staff deter:ained that GLP-5588 : ray serve as an acceptable reference ard the initial basis for allowing substitution of near-isotropic H-451 graphite feel and reflector elements for the current reference needle-coke H-327 graphite elements in Fort St. Vrain (Ref. 1).

In recent correspcndence with Public Service Ccrpy of Colorado (Ref.

2), the NRC staff requested additional information regarding use of H-451 graphite in Fcrt St. Vrain. In Eart, these questions were prcepted by the develo5 ment of new :nodels for irradiation-induced creep in graphite. Se staff requested that GLP-5588 he amended to contain this new information.

Accordingly, enclosed are fifty (50) copies of A:nerdeant 1 to GLP-5588. mis a:wbnt contains new infocation on irradiation-induced creep ard irradiation-induced dimensieral change both in H-451 and in H-327 graphite. 'Ihe a:nendaent is being submitted by GA Technologies, as noted in PSC's response to the NRC request for additioral infor: ration (Paf. 3),

because GLP-5588 is a GA topical report, the review of which war not part cf the Fort St. Vrain licensing docket.

If you have arrf questions regarding this amendnent, please feel free to call :ne at (619) 455-3821.  :

E Very truly yours, I

j f[l

'I - ,-

! C. R. Fisher,

Manager - Licensing,

Paliability and Systems E

Enclosure cc: J. T. Collins, Region IV M. Tokar, Core Performance Branch P. C. Wagner, Regicn N

. ' ' + ., , ,

References:

1. Ganmill, William P. (NRC), letter to Colin R. Fisher (GA), " Evaluation of 'Ibpical Report GLP-5588," May 2] ,1979.

2. Madsen, G. L. . (NRC), letter to 0. R. Lee (PSC), August 2,1983.

3. Brey, H. L. (PSC), letter to John T. Collins (NRC), "Additicnal Information on H-451 Graphite," P-83348, Octcber 27, 1983.

l 4

O o

. _ - _ _ _ _ _ - , - - - _ _ . - - - _._w-_.-__- _ - _ _ - - _ . - - - . - _ _ - - - . _ - - _ . - - - - - - - _ _-a

AMENDENT INCDRPCPATION INSIPUCTICNS GLP-5588 MENDEh? 1 Delete Paces Recl ea With Aman h n.t Paces 1

lii, iv 111, iv 4-1, 4-2 4-1, 4-2 4-3, 4-4 4-3, 4-Ja; 4-4 4-7, 4-8 4-7, 4-8 4-9, 4-10 4-9, 4-10 4-11, 4-12 4-11, 4-12 4-17, 4-18 4-17, 4-18 4-21, 4-22 4-21, 4-22 4-25, 4-26 4-25, 4-26 u,.

GLP-5588 CONTENTS ESLt.

1. IN TR ODU CTION AND SU MMAR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -1

2. PERFO RMANCE AN ALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.1. Cor e Nucle ar Analy sis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.1.1. Fuel Loading and Excess Reactivity . . . . . . . . . . . . . . . 2-1 2.1.2. Power Distribution ............................... 2-1 2.1 3 Fuel Burnup and Exposure ......................... 2-2 2.1.4. Shutdown Margins and Reactor Control ............. 2-2 2.1.5. Rod Withdrawal Accidents . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.2. Co re Therm al Analy sis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 23 Fission Product Trans port Analysis . . . . . . . . . . . . . . , . . . . . . . . 2-3 2.4. Graphite Structur al Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.4.1. Mechanical Properties ............................ 2-4 2.4.2. Stress Analysis .................................. 2-5 2.4 3 Graphite Dimensional Change . . . . . . . . . . . . . . . . . . . . . . 2-7

( R e f er e n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 8 3 SAFETY ANALYSIS ................................................ 3-1

31. Introduction and Summary ................................. 3-1

32. Loss of Normal Shutdown Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 33 Hoisture Inleakage ........................................ 3 331. Steam-Graphite ciactions ......................... 3-3 332. Hydrolysis of Failed Fuel . . . . . . . . . . . . . . . . . . . . . . . . 3-4 1

333 Fission Product Release from Oxidized Graphite ... 3-4

34. Permanent Loss of Forced Cooling . . . . . . . . . . . . . . . . . . . . . . . . . 3-5

35. Rapid Depressurization/ Blowdown . . . . . . . . . . . . . . . . . . . . . . . . . . 3-6 i 3.6. Co ncl us ions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 8 References ..................................................... 3-8

4. MATER IALS PROPERTY DATA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 4-1 j 4.1 H-451 Graphite Materials Property Data .................... 4-1 4.1.1. - S pe cific Heat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 I

iii

- . GLP-5588

• Am:nda:nt 1 i t'

4. MATERIALS PROPERTI DATA Determination of the material properties of nuclear-grade graphite is a continuing program at GA Technologies. The most recent values of H-451 and- 1 H-327 graphite material properties are given in Sections.4.1 and 4.2, respectively.

4.1. H-451 GRAPHITE MATERIALS PROPERTI DATA 4.1.1. Soecific Heat The specific heat of H-451 graphite over the temperature range 0* to 2700*C is given by Eq. 4-1, which is accurate to 25% of the-mean (Rer. 4-1):

, Cp = 0.54212 - 2.42667 x 10~0 T - 90.2725 T-l

- 4.34493 m to" T-2 + 1 59309 s 10 7 T-3 (g_3) 1.43688 x 10 9 T*4 i 4.1.2. Bulk Density The mean bulk density for the H-451 log is 1.74' Mg/m3-(Ref. 4-2).

4.1 3. Irradiation-Tnduced Dimensiona! Channe (Refs. 4-1. 4 8. 4.0)

The permanent strain .(c*) due to irradiation-induced dimensional change has been expressed in terms of average irradiation temperature (T)' 1 and fast neutron fluence (@) for near-isotropic graphite.- The irradiation-strain (c*) i ' expressed by the polynomial in Eq. 4-2, which is valid for i

14-1 e

a 9 w y = ~ g ,-

.. GLP-5588

, Am:ndmInt 1 4 .1.*[ . Poisson's Ratio The Poisson's ratio (v) for H-451 graphite subjected to tensile strain is given as (Ref. 4-4) v= 0.118 (30.01).

4.1.8. Irradiation Creen Behavier Irradiation-induced creep for both tensile ned compressive stresses in near-isotropic graphite is described by Eq. 4-3 (Refs. 4-Sa, 4-5b):

Irradiation temperature 1650cc:

e c,,s " ci - exp (-2.5 x 1s20 ,33 ,3 6 x 10-21 cge c E

o o E(0,T)

(,

Irradiation temperature > 650cc:

1 cep = {[1-exp(-2.5x10-20c))

8

, o E' (1.6 x 10-21 + 5 x 10-28 f.T-650)2.6) 3)

E, ,E(?,T) where c a total uniaxial creep strain (cm/cm) cr a strese. (MPa) g E = Young's modulus when load is applied or. changed (MPa)

.) a fast neutron fluence, n/cm2 (E > 0.18 MeV, HTGR spectrum) l

'E8 Young's modulus at a fluence of 1.4 x 1021 n/cm2 (MPa)

R l

4-3

, GLP-5588 9

4.1.10. Thermal conductivity The thormal conductivity of near isotecpic graphite is given by the model of R. Price (Ref. 4-6). This model considers the dependence of thermal conduc-tivity (K) on the current measurement temperature (Te ) and on the past his-tory of irradiation temperature (Tj) and fast neutron fluence (3). The model is extended here to the cass of a nonisothermal irradiation with unknown temperatures.

The thermal conductivity us a function of current measurement temperature can be considered as a superposition of three temperature-dependent resistance mechanisms through the equation I

K(T ) = ( 4-4) -

~

1

, 1 d-,

_Kg(Tc ) Kb"(T c) Lb(T) c.

where a is a porosity-tortuosity factor, b

Kg(Tc) is the crystallite conductivity with Unklapp processes dominating, b = 1/La (Ref. 4-6) is the inverse of the crystallite boundary spacing, K$(To) = (K B /La)(Tc) (Ref. 4-6) is the effect of the grain boundary scattering 2

daC3d7 (Ref. 4-6) is an irradiation damage parameter, Ed (Tc) " (KDD C S ) (To) (Ref. 4-6) is the effect of the irradiation damage..

All of the above quantities are given as known input data in Tables 4-5 and 4-6 with one exception: the irradiation damage parameter, d.' - As will be shown, '

one can solve for the parameter d by comparing conductivities before and after irradiation.

4.1.10.1. Thermal conductivity. - Unirradiated. ~ For'unirradiated material, the damage parameter d in Eq.-4-4 is zero. Equation.4-4 reduces to 4-4 L _

_J

GLP-5588 Am:ndm:nt 1 where-C p

a specific heat at constant pressure (cal /g-K),

T = temperature (K).

4.2.2. Bulk Density The mean bulk density for the H-327 log is 177 Mg/m3 (Ref. 4-7).

4.2 3 Irradiation-Induced Dimensional Channa The perstnent strain (c *) due to fast neutron damage in graphite has been zoasured in many experiments over the past to years. The irradiation strain in H-327 graphite is expressed as a polynomial for design use:

c' = (C 3+CT+CT2 2 3 e73c 5 #

+ (C 6+CT+CT2+CT3+C 7 g 9 10 T* + C16 I)

• I II) 2

+ (Cg+C12T+C T33 +C g T3g +0 *U } '

15 17 18

( where c* s irradiation strain (dimensional change, ai/1) (%),

e a fast neutron fluence (1025 of,2) (E 2 29 fJ)HTGR '

T = average irradiation temperature (*C), 1 Cg a coefficients determined for each orientation of H-327 graphite; coefficients listed in Table 4-7 The H-327 irradiation strain polynomial is based on data taken from the CG-1, -2, and -3 experiments and from experiments by ECN, Petten, in the Petten low temperature graphite capsules (Refs. 4-3, 4-8, 4-9). Irradia-tion strain ( c*) calculated frca the above polynomial is valid for irradi-ation temperatures from 350* to 1000*C and fast neutron fluences of 0 to 8 -

x 10 25 gj,2 (E 2 29 fJ) . The dimensional change in H-327 is shown in Fig. 4-5.

4-7 1

9

GLP-5588

, Amindaint 1

. l 4.2 9 Thermal Ereansivity Figure 4-8 presents the design curve of ther=al strain versus tempera-ture for unirradiated H-327 graphite. The data of Fig. 4-6 are tabulated in Table 4-10.

4.2.10. Thermal Conductivity The thermal conductivity of H-327 graphite is calculated using the previously described methodology for H 251 graphite conductivity (section 4.1.10). The material constants for H-327 graphir.e to be substituted for Table 4.6 are given in Table 4-11. A plot of the thermal coaductivity of H-327 graphite as a function of fast fluence and temperature is given in Fig.

4-7 I

J 4-9  ;

I

D. GLP-5588 4-8. . Price, R. J. and L. A. Beavan, ' Final Report on Graphite Irradiation

- Test 00-1,* USAEC Report GA.A13089, General Atomic Company, August 1, 1974, pp. 13 through 51.

4-9 Price, R. J. and L. A. Beavan, " Final Report on Graphite Irradiation Test 00-3," ERDA Report GA-A14211, General Atomic Company, January 1977, pp. 5-1 and 5-4.

)

4-10. Final Safety Armlysis Report, Fort St. Vrain, Appendix F.

4-11. Price, 3. J., " Mechanical Properties of Graphite for High Tamperature Gas-Cooled Reactors: A Review," ERDA Report GA-A13524, General Atomic Company, September 22, 1975, pp. 2-6 through 2-9

( .

4-11

I

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,' GLP-5588 A=2ndment 1 l

l l

1 TABLE 4-7 COEFFICIENTS FOR IRRADIATICN STRAIN POLYNCMIAL, H-327 GRAPHITE Axial Radial C

3

-5.108481 -0 309218 C

2 +0.292901 x 10-1 +0.126366 x 10-2 C

3

-0.644452 x 10~" -0.461461 x 10-5 C +0.637722 x 10 -7

-8 4 +0 786252 x 10 C

5

-0.237075 x 10 -0.414299 x 10 #

C +2 977696 +1.417139 6

C 7

-0.189175 x 10-1 -0 949423 x 10-2

~ -

C 8 +0.430617 x 10 " +0.227745 x 10 "

C 9

-0.422337 x 10'I -0.234320 x to -I C

IO +0.150559 x 10-10 +0.861987 x 10~11 C -0.268576 -0.147822 33 C

12 +0.165487 x 10-2 +0.100726 x 10-2 C

13

-0 369364 x 10-5 -0.243819 x 10-5

-8 C

34 +0 357371 x 10 +0.250028 x 10-8 C

15

-0.126622 x 10 # -0 907074 x 10" "

C 0 ,o 16 C 0 0-37 C 3g 0 0 4-17

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l 0 4! 6 8 FAST NEUTRON FLUENCE (10 21 nfc,2, E > 0.18 MeV, HTG h) l'i g . 4 - 5. Irradiation-induced dimenstunal change: (a)11-327, oxial .