ML19337B663

Proposed Tech Spec Table 6.1-1 Re in-core Residence Times of Fuel Test Elements.Safety Analysis Rept Encl

ML19337B663
Person / Time
Site:	Fort Saint Vrain
Issue date:	10/03/1980
From:	PUBLIC SERVICE CO. OF COLORADO
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ML19337B660	List: ML19337B659 ML19337B661 ML19337B663
References
NUDOCS 8010080472
Download: ML19337B663 (50)

Text

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i ATTACHMENT 3 I

I Proposed Revision to Table 6.1-1 l l

l of Specification 6.1, ,

j Reactor _ Core Design Features l

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and Reason for Revision i

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TABLE 6.1-1  ;

DESCRIPTION OF FORT ST. VRAIN FUEL ELEMENTS FTE-1 TIIROUGli FIE-8 FTE-1 FTE-2 FTE-3 FTE-4 FTE-5 FTE-6' FTE-7 FTE-8 Segment 2 3 4 5 6 7 7 7

Graphite Type H-451 H-451 H-451 H-451 H-451 H-451 H-451 H-451 Fissile Fuel Type UC UC UC UC UC UC (Th,U)C (Th,U)C TR$SO TR$SO TR$SO TR$SO 2 2 TR$SO TR$SO TRIS 0 TRISO plus test n, plus test plus test fuel (a) fuel (a) fuel (a)

Fertile Fuel Type Th0 Th0 Th0 Th0 Th0 Th0 ThC ThC TRI$0 TRI$0 TRI$0 TRI$0 TRI$0 TRI$0 TRI$0. TRI$0 plus BISO plus BISO plus BISO Method of Fuel Rod Curing (b) CIP CIP CIP CIP CIP CIP CIB CIB Resid:nceTimg)EquivalentFull Power Years 0.7 1.7 2.7 3.7 4.7 5.7 5.7 5.7 (a) Test fuel includes: (Th,U)C TRISO, 2 88rodsperelementCIB]

UCx 0,* TRISO/Th02 BISO, 350 rods per element CIP E HEU UC TRISO/Th0 TRISO, 176 rods per element CIP

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2 2

.*l! R W (Th,U)0 TRISO/Th0 TRISO 88rodsperelementCIT} MEU 'i' f, $ -

2 2 *

(b) .g'k CIP = cure-in-place fuel rod carbonization; CIB = cure in alumina bed - reference FSV process; *7 CIT = cure-in-tube, graphite crucibles, simulating conditions as experienced in cure-in-place. L 5 *

(c) Equivalent full poweryear is defined as 300 equivalent full powr days g 8

• x and y represent the mean quantitites of carbon and oxygen and do not signify a specific compound. "

These values will be explicit in the final fuel specification. All kernels of this type are derived from WAR beads.

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Reason for Technical Spe_cification Revision ort for Amendment Fort St. Vrain 3 to General Reload 1 TestAtomic ElementsCompany's " Safety FTE-1 through Analysis Rep (GLP-5494)

FTE-8" analyzes the effects of extending the core residence time of each fuel test element (FTE) by 0.2 equivalent full power years. Detailed perfomance analyses of the test elements were conducted to establish power distribution, temperature history, fission product retention, graphite element stresses and dimensforal stability of each element for verification of design ma rgins. The results of these studies are described in section 5 of GLP-5494. General Atomic Company concluded that "the physical properties of the test element materials are improved over the initial FSV core materials" (page 2-1 of GLP-5494).

Section 7 of GLP-5494 presents a detailed description of the irradiation exposure testing of coated fuel particles conducted under 00E, ORNL, GA, and foreign fuel qualification programs. The fuel particle coating thickness and FTE fuel densities are based on extensive irradiation experience. The coating thickness specifications have been chosen to limit the fuel particle coating failure fraction to values less than those expected for the reference FSV Segment 7 fuel particles. Resul ts of these irradiation tests support extension of the core residence time of each FTE by 0.2 equivalent full power years.

The FTE design criteria, manufacturing process control, and quality assurance provisions were the same as or better than those used for the original reference FSV fuel elements. The experimental fuel in the test arrays has been subjected to rigorous quality control inspection to verify coating properties and integrity, low contamination levels, l and heavy metal 1,adings. Test fuel manufactured by sources other than l G. A. has been required to meet specifications equivalent to the G. A.

produced fuel and has been subjected to quality inspection by G. A.

prior to fuel element assembly. FTE-1 through FTE-8 are designed to operate within the limits of peak fuel temperature, neutron fluence, and burnup specified for the initial core and reload fuel elements.

Based on the design and quality assurance requirements, the FTE's are capable of safe operation throughout the extended core residence times listed in the revised Table 6.1-1.

Accident analysis studies, discussed in section 6 of GLP-5494, indicate that the eight FTE's would not significantly affect any accident analysis presented in the FSAR for the revised core residence times of Technical Specification Table 6.1-1. The extended core residence times l will involve no increased hazard to the health and safety of either the  !

plant personnel or the public associated with operation of the plant.

The units of core residence time in Technical Specification Table 6.1-I have been changed from " years" to " equivalent full power years" to reflect an irradiation lifetime control limit. An equivalent full power year is defined as 300 equivalent full power days in the table.

Therefore, the proposed core residence time extension of each of the 1 eight FTE's by 0.2 equivalent full power years, generally corresponds "

to an extension of Cycle 2 lifetime from 150 EFPD to 200 EFPD.

ATTACHMENT 4 Safety Analysis Report for Fuel Reload 1 Extended Operation i

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, . PED JUN 2 21971$  !

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GENERAL. ATOM 8C COMPANY Po. box 81eos SAN olEGo. CALIFORNIA 92138 (714) 45M000 June 16, 1978 Project 90-330 MW HTGR Fort St. Vrain Unit 1 GLP-5646_ - I Its etts/dL c.

Mr. J. K. Fuller, Vice President Engineering and Planning )

Public Service Company of Colorado i Post Office Box 840 '

Denver, Colorado 80201 l

Subj ect: Safety Analysis Report for Fuel Reload 1 Extended Operation l

References:

GLP-5574, GLP-5641 l bear Mr. Fuller:

General Atomic is enclosing with this letter its draft of the Safety Analysis Report for Fuel Peload 1 Extended Operation. This SAR is a re. vised version of the earlier SAR for Fuel Reload 1, GLP-5574, published October 24, 1977. It is intended as a companion to the SAR for Cycle 1 Extended Operation, GLP-5641, which was transr.itted to you on June 1, 1978.

This report differs from GLP-5574 in that it considers changes due to extended operation of Cycles 1 and 2, each up to 200 effective full-power days. In addition, minor editorial changes are included. All changes from the contant of GLP-5574 are noted in t's margins of this report by the numeral "1".

We would be pleased to discuss with you any questions you may have on the enclosed document.

Very truly yours, A ;^^

D. P. Lessig, Vice resident Fort St. Vrain Project o Enclosure p .rna to enruc cer e:m e .

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t-GLP-5646 O

e SAF1T. l ANALYSIS REPORT FOR FUEL RELOAD 1 r ur:iDED OPERATION FORT ST. VRAIN NUCLEAR GENERATING STATION O

o GENERAL ATOMIC COMP.GT JUNE 1, 1978 F

GLP-5646

. CONTENTS FOREWORD . . . . . . .. ... .. .. . . . . . . . . . . . . . . . v l1

1. INTRODUCTION AND SLMULY .. . . . . . . . . . . . . . . . . . 1-1

2. INITIAL CORE OPERATING EISTORY . . . . . . . . . . . . . . . . 2-1

3. GENERAL DESCRIPTION . .. .... . . . . . . . . . . . . . . . ~

3-1

4. FUEL SYSTEM DESIGN .. . ..... . . . . . . . . . . . . . . 4-1 4.1. Fuel Design . .. .... . . . . . . . . . . . . . . . 4-1 4.2. Mechanical Design . ... . . . . . . . . . . . . . . . 4-2 4.3. Thermal Design . . . ... . . . . . . . . . . . . . . . 4-3 4.4. Fission Product Release . . . . . . . . . . . . . . . . 4-3

5. NUCLEAR DESIGN .... . ..... . . . . . . . . . . . . . . 5-1 5.1. Segment 7 Fuel Loading . . . . . . . . . . . . . . . . . 5-1 5.2. Burnable Poison Loading . . . . . . . . . . . . . . . . 5-1 5.3. Control Rod Sequence . .. . . . . . . . . . . . . . . . 5-2 5.4. Projected Cycle 2 Operation . . . . . . . . . . . . . . 5-4 5.5. Maximum. control Rod Worth . . . . . . . . . . . . . . . 5-4 5.6. Core Shutdown Margin . . . . . . . . . . . . . . . . . . 5-5 5.7. Kinetics Parameters ... . . . . . . . . . . . . . . . 5-6 5.8.

Analytical Input . .. . . . . . . . . . . . . . . . . . 5-7 5.9. Core Operating Procedures . . . . . . . . . . . . . . . 5-7

6. THERMAL-HYDRAULIC DESIGN .... . . . . . . . . . . . . . . . 6-1

7. SAFITY ANALYSIS . . . .. .. .. . . . . . . . . . . . . . . . 7-1 7.1. Introduction . .. . . .. . . . . . . . . . . . . . . . 7-1 7.2. Loss of Normal Shutdown Cooling, Permanent Loss of Forced Circulation (LOFC), and Rapid Depressuri:ation/

Blowdown (DSDA) ... .. . . . . . . . . . . . . . . . 7-2 7.3. Conclusions . .. .. .. . . . . . . . . . . . . . . . 7-2 I

. 8. PROPOSED MODIFICATIONS TO TECHNICAL S?ECIFICATIONS . . . . . . 8-1 t

9. STARTU? TESTS . . .. . . . ... . . . . . . . . . . . . . . . 9-1

- 10. REFERENCES . . .. . .. . . .. . . . . . . . . . . . . . . . . 10-1

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GLP-5646 FIGURES 3.1. Core regions refueled in reloads 1 through 6. . . . . . . . 3-3 3.2. Refueling region age distribution for the equilibrium cycle (before refueling) . .. . .. . . . . . . . . . . . 3-4 5.1. Burnable poison hole locations . . . . . . . . . . . . . . 5-13 5.2. Identification of control rod groups . . . . . . . . . . . 5-14 5.3(a). Tilt envelope for cycle 2: unrodded regions . . . . . . . 5-15 5.3(b). Tilt envelope for cycle 2: partially redded regions . . . 5-16 5.3(c). Tilt envelope for cycle 2: fully redded regions . . . . . 5-17 5.3(d). Tilt envelope for cycle 2: all regions . . . . . . . . . . 5-18 5.4. Axial power distribution during cycle 2, unrodded segment 7...... ..... . . . . . . . . . . . . . 5-19 5.5. Maximum allowable rod pair worth vs. average gas outlet temperature . ....... . . . . . . . . . . . . . . . . 5-20 5.6. Temperature coefficient in the initial core with 3

equilibrium Xe-135 and Sm-149 . . . . . . . . . . . . . . . 5-21 5.7. Temperature coefficient at end of equilibrium cycle . . . . 5-22 5.8. Temperature defect vs. average core temperature . . . . . . 5-23 TA3LES 4.1. As-built segment 7 uranium and thorium loadings . . . . . . 4-4 5.1. Projected initial core loadings at the and of cycle 1

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(200 EFPD). .. ..... .. . . . . . . . . . . . . . . . 5-8 l 5.2. Use of burnable poison rods in segment 7 . . . . . . . . . 5-9 5.3. Control red sequence for cycle 2 . . . . . . . . . . . . . 5-9 5.4. Calculated control rod group worth and power peaking factors with . cycle 2 rod sequence . . . . . . . . . . . . . 5-10 5.5. Control rod bank worth . . . .. . . . . . . . . . . . . . 5-11 5.6. Shutdown margins - cycle 2 . . . . . . . . . . . . . . . . 5-11

5. 7. Kinetics parameters . . . .. . . . . . . . . . . . . . . . 5-12 7.1. Potential effects of cycle 2 on FSV FSAR accident predictions .

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GL?-5646

, FOREk'ORD' This Safety Analysis Report (SAR) is a revised version of the earlier SAR for Fuel Reload 1, GLP-5574, published October 24, 1977. ,This report differs from GLP-5574 in that it considers changes due to extended opera-tion of cycles 1 and 2 each up to 200 affective full-power days. In 1 addition, minor editorial changes are included. All changes from the content of GLP-5574 are noted in the margins of this report by the numeral

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GL?-5646

1. INTRODUCTION AND SU. WARY i

This Safety Analysis Report (SAR) is prepared to obtain concurrence to operate the Fort St. Vrain Nuclear Generating Station through the forth-coming relosd cycle (cycle 2) for a period of time up to 200 affective 1

full-power days (EFFD). For this cycle, 6 of the 37 fuel regions in the core will be loaded with fresh fuel elements fabricated by Geneal Ato=ic Company. The remaining fuel elements will have op. rated through cycle 1.

l The introduction of new fuel elements is e msistent with the fuel

=anagement program described in the FSAR.

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The regions to be refueled contain 204 standard fuel elements (includ-ing three test ele:nents), 30 control fuel elements, and six bottom control fuel elements, a total of 240 elements to be :eplaced. In addition, five other cast elements containing a variety of recential 'dTGR fuel designs will be inserted into other regions of the core with this reload. The licensing basis for these eight test elements is provided in a separate l1 document (Ref. 1).

l This report contains sections describing the operating history of the )

initial core through April 1978 (cycle 1), the fuel system design, the l1 nuclear design, the thermal-hydraulic design, and the safety aspects of the core during cycle 2. The planned startup program for the refueled core is also briefly described.

The design of the replacement fuel elements is identical to that of the elements already in the core, with changes made only in fuel and burnable poison loading appropriate for the reactivity and power distribution require- )

ments of the cycle. No changes to the plant Technical Specifications are c necessary, and no unreviewed plant safety questions are presented for cycle 2, per 10CFR50.59. None of the peak operating li=its presented in the FSAR are exceeded.

1-1

GLP-5646

2. INITIAL CORE OPERATING HISTORY The FSV initial core is currently in the startup testing phase. As of April 30, 1978, the maximum power reached is about 572 IW(t), and the core burnup corresponds to about 78 EFFD. It is expected that the core 1

can be operated at full rated conditions while meeting all technical specification limits with ample =argin.

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.- GLP-5646

3. GENERAL DESCRI? TION

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The Fort St. Vrain fuel management scheme is designed so that approxi- i mately one-sixth of the core is reloaded at periodic refueling intervals. l1 This document describes the first reload segment to be inserted into the j FSV core. This reload fuel segment is designed so that the core perform-l ance with the new fuel added satisfies the reactor Technical Specifications.

These limitations apply to the total core performance. That is, not only do the freshly loaded refueling regions meet power distribution'li=itations, but the perturbations to the remainder of the core are such that the seg=ents remaining in the core from the initial cycle also satisfy the performance requirements. These performance requirements include core excess reactivity P

and shutdown margins, power distribution behavior, and all the core safety considerations discussed in the FSV Final Safety Analysis Report.

4 The initial core fuel loading was selected to that the residence time of the sixth segment could be as long as six full-power years at 80 percent capacity factor. About one-sixth of the core is replaced at each refueling. Ihe scheduled refueling sequence is summarited 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. This first reload segment consists of 204 standard fuel elements, 30 control rod fuel elements, and six bottom control rod fuel elements. In this reload segment, three of the refueling regions are located in the central portion of the core (regions 5,10,17) and three l1 are located adjacent to the side core-reflector interface (regions 21, 28, 35). The refueling region sequence was chosen so that freshly refueled regions are never adjacent to each other (except when region 1, the central g e

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

1 3-J 1

, GLP-3646 region, is reloaded). Therefore, each refueling region is surrounded by regions of varying ages. 1 Figure 3.2 shows the refueling region age distribution for the equilib- 1 rium cycle as given in the FSAR. By comparison c.,i this figure with Fig.

3.1, it can be seen that the reload sequence follows that given in the FSAR.

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Fig. 3.1. Core reSions refueled 1:1 reloads 1 :hrough 6

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0.18 MeV) exposure in the discharged segment is about 0.9 x 1021 ,yt , 1 5.2. BURNABLE POISON LOADING As shown in Fig. 5.1, six holes are provided in each standard fuel element for insertion of burnable poison rods, and four holes are provided in each control fuel element. For segment 7 fuel, there will be no poison in the control elements; all six poison holes in the standard elements =ay be used depending on burnable poison leading requirements. I Two burnable poison rod types, differing in their boron loading, vill l1 be used in seg=ent 7 fuel. The poison to be loaded into the fuel elements vill be axially zoned to maintain the desired axial power distribution in l 1g segment 7 fuel regions during depletion. The use of burnable poison rods s in seg=ent 7 is sun:mari:ed in Table 5.2.

5-1

GL2-!6L&

5.3. CONTROL ROD SEQUENCE

, Technical Specification LCO 4.1.3 states that a control rod sequence 1 will be specified for each fuel cycle, and that the sequence will always be

followec except for rod insertion Jesulting from a scram or rod runback, or during low-pever physics testing. The control rod sequence for use during l1 j cycle 2 is given in Table 5.3. ~he cycle 1 rod sequence is included for comparison. The identification of the control rod groups is shown in Fig.

5.2.

The regulating rod is located in the central refueling region (red group 1). This group is partially withdrawn before criticality is achieved l

and then caintained in its most reactive control rod position for the re- I cainder of the operation. In this manner, minor reactivity adjustments can be made most rapidly with the minimum amount of control rod motion. This is consistent with the method of operation utiliced for the control rod sequence of the initial cycle.

A summary of the calculated power peaking factors obtained using the control red sequence for all control rod configurations is given in Table 5.4 This includes all of the control rod configurations in which the con-trol rod groups are either fully inserted or withdrawn, including those suberitical configurations during the withdrawal of the first few control rod groups. Any configuration with a partially inserted control rod group will have peaking factors lying between those calculated when that group is fully inserted and fully withdrawn.

The control rod configurations shown in Tabla 5.4 may be separated into four categories of reactor power operation: 5 1) full power, (2) 20 to 100% power, (3) O to 20 power, and (4) suberitical. Full-power operation 1 may be achieved with configurations ranging from one control red group g (three red pairs) fully or partially inserted at the end of cycle 2 when ,

5-2

. . GLP-5646 the excess reactivity is relatively low, to three control rod groups (nine rod pairs) fully or partially inserted at the beginning of the cycle, when fission product poisons such as xenon are not present. (In accordance with l1 the requirements of the Technical Specifications, only one control rod group in addition .co group 1, the regulating rod, will be partially inserted at any time.) .

At lower than rated power, in the range of exie. gas ta=peratures be-tween 1460*F and 950*F, when the xenon level and the core temperature are lower, configurations ranging fica four control rod groups (12 rod pairs) to five control rod groups (15 rod pairs) fully or partially inserted can be expected, depending on the temperature, power level, and the rate at which power was increased from the previous level.

The third category covers the rise-to-power phase fro = the cold critical condition to about 20% power (gas outlet temperature <950*F) .

Different limiting operating conditions are applied to this phase of reactor operation by the Technical Specifications. The initial cold criti-cality, following the refueling operation, was calculated to be achieved with seven rod groups (21 rod pairs) fully or partially inserted. ,

l The last sategory covers the suberitical control rod configurations l where region peaking factors (RPF) or intra-region tilts are not meaningful and, consequently, are not given in Table 5.4.

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From the data given in Table 5.4, it can be seen that the calculated power peaking factors for the various power levels do not exceed those given in the Technical Specifications. This is true for both the radial region power peaking factors and the intra-region peaking (colu=n tilt) factors. The radial region peaking is be' low 1.83 for all configurations involving less than 16 control rod pairs. For low-power oparation (Tgas g 950*F) when more control rod pairs are inserted, the region peaking is be-low 3.0 until the subcritical configurations are reached. In the same man- )

ner, the intra-region peaking factors are also acceptable. l 5-3

GLP-5646 1

5.4 PROJECTED CYCLE 2 OPERATION l

, This section presents the results of cycle 2 depletion analyses using design methods discussed in Section 3.5 of the FSAR. As-built fuel and c burnable poison loadings discussed in Section 5.2 were used as input (see Sections 5.1 and 5.2).

Figures 5.3(a) through 5.3(d) present envelopes encompassing projected 1

region peaking factors (RPFs) and column tilts during cycle 2 depletion.

The results indicate that RPFs and tilts during cycle 2 will be well within the allowable limits set by LCO 4.1.3.

Axial zoning of the segment 7 fuel and burnable poison is provided:

(1) to produce a power distribution which cands to reduce axial fuel tem- l 1 i paratura peaking and (2) to maintain the desired axial power distribution with depletion. The calculated axial power distributions in segment 7 fuel 4

during cycle 2 are shown in Fig. 5.4. The calculations were carried out with the FEVER code by modeling the segment 7 fuel by itself. It can be o

seen that the LCO 4.1.3 limit on peaking factor in the lower fuel layer i 1

(axial pewar factor < 0.90) is not exceeded.

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5.5. MAXIMUM CONTROL ROD WORTH l

l The rod withdrawal accident for the full-power operating condition  :

described in the FSAR (Section 14.2, per Amendment 24) assumes withdrawal 1

of a control rod worth of 0.012 ok at end of cycle (EOC) and an equilib- l rium EOC temperature coefficient. Thiscontrolrodworth,becauseofhigherl1 negative reactivity compensation resulting from the larger negative te=pera-ture coefficient, is equivalent to a rod worth of about 0.016 ok at begin- )

ning of cycla (30C). Since the ta=perature coefficient of reactivity is 1 significantly more nege.tive for cycle 2 than for the equilibrium cycle (see Section 5.7), a maximum single rod pair worth for cycle 2 30C of less than bl '

• l 0.016 Ak is acceptable. In addition, the calculated worth of any rod pair O

in any operating critical control rod configuration =ust be less than 0.047, per LCO 4.1.3. ,

5-4

GLP-5646 The maximum worth control rod pair which occurs during the specified control rod sequence is given in Table 5.4 Since the tenperature defect is significantly more negative during cycle 2 (see Section 5.7) than for the equilibrium cors, the consequences of the accidental removal (dWA) of a rod worth greater than 0.012 ak is no more severe than the withdrawal of 0.012 ak at rated power discussed in the FSAR. A conservative estinate of the maximum worth of a rod pair which is equivalent (in consequences) to the FSAR data is shown in Fig. 5.5.

At rated power, the mav4 mum calculated rod worth at the beginning of cycle 2 (no xenon) is 0.0144 Ak, which is less than the FSAR limit of 0.016 ak. During the cycle, the mmvimum rod worth is less than 0.0127 ok, I which is less than a conservative estimate of the allowable red worth at the and of cycle 2 (see Fig. 5.5). For other categories of reactor opera-tion discussed in Section 5.3, the m=v4 mum worth rod ramn4ns below the l1 allowable limit (shown in Fig. 5.5) and, consequently, meets the require-ment of LCO 4.1.3 of the Technical Specifications.

c 5.6. CORE SHUTDOWN !"ARGIN The Technical Specifications and the reactor design criteria specify that the reacter nust be capable of being shut down (0.01 ok suberitical) l1 l with any one control rod pair withdrawn at room ta=perature and with any two control rod pairs withdrawn at refueling temperature. This is consist-1 enc with the policy that, during the rod withdrawal accident postulated in  ;

1 Section 14.2 of the FSAR, the control rod pair being withdrawn is continu-ously withdrawn until fully removed during the accident and is not capable of being reinserted. The requirement for being shut down with any two l1 control rod pairs withdrawn is established to allow for operation with at  !

least one inoperable control rod pair. l1 The net negative reactivity insertions following a scra= during cycle 2 are given in Table 5.5. Since the axcess reactivity at full power with equilibrated xenon and protact' d um is about 0.020 Ak, the instantaneous 5-5

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CLP-5@46 W hum shutdown margin (SDM) with the two maxi =um worth rod pairs

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inoperable is about 0.120 ak. This shutdown strgin is reduced due to

, core cooldown (in a matter o'f hours), due to xenon decay (in a =atter of a few days), and due to transfor=ation of Pa-233 into U-233 (in a matter of weeks). To show that the shutdown margin is satisfactory for :hese cases with one or more control red pairs assumed inoperable, the summary

)

in Table 5.6 gives calculated excess reactivity and core shu:down margin at the beginning of the cycle (30C) immediately af ter refueling and, at the middle and and of the cycle (MOC and EOC). It can be seen that for all cases of interest, the core shutdown =argin is larger than the 0.01 Ak as required in the Technical Specifications.

For the case with only the =axi=um wor:h rod pair inoperable, the minimum shu:down margin at room :emperature with xenon and Pa-233 decayed (the most reac:1ve case) is 0.053 Ak. A: refueling ta=perature (220*F),

3 this value would be 0.062 Ak for the same condition. However, since the second maximum worth rod out can add as much reac:iviry as 0.041 Ak, it may not be possible to naintain a shutdown margin of 0.01 Ak with these

wo rod pairs out if the Pa-233 is allowed to fully decay to U-233. This l1 situation does not conflict vi:h the LCO, however, since the requere=ent for the two rod pairs out at refueling temperature is established to allow for operation 'with one inoperable control red drive assembly. An adequate shutdown =argin can be maintained for a period of at least two weekt- in this core condition. If the inoperable unit cannot be repaired in this time period, the reserve shu:down system can be used to =aintain adequate reactivity control.

5.7. KINETICS PARAMETIRS The kinetics parameters for :he initial cycle and a conservative estimate of them for the second cycle (seg=ent 7 deple: ion) are given in l 1g Table 5.7 (:aken from the FSAR, A=endment 14). The ta=perature coefficients s for the initial cycle and the equilibrium cycle are shown in Figs. 5.6 and 5.7, respectively. The :e:perature coefficient for cycle 2 lies be: ween 1 the values for the initial and equilibrium cores as illustrated by the 5-6

CLF-5646 temperature defects shown in Fig. 5.8. For conservatism, the te=perature defect at tha and of cycle 2 was used to deter =ine the allowable reactivity

, worth of individual rod pairs in the specified control rod sequence (see Section 5.3). Since the maximum rod worth for the core at rated power does

, not exceed the value calculated for the equilibrium core. a rod withdrawal accident (RWA) during cycle 2 would have less severe consequences than a comparable case at equilibrium. Thus, the requirements of LCO 4.1.3 are met for cycle 2 operation.

5.8. ANALYTICAL LNPUT Nuclear analyses were carried out using the same methods applied to the analyses presented in the FSAR. The design of segment 7 introduces no new aspects to HTGR core design; consequently, there was no need to develop or adapt any new methods or procedures for the nuclear design.

5.9. CORE OPERATING PROCEDURES Core operating procedures will be the same as those for the initial core and those planned for the equilibrium core. The only difference will be the control rod withdrawal sequence discussed in Section 5.3. '

, s 5-7

TABLE 5.1 PROJECTED INITIAL CORE LOADINGS AT Tile END OF CYCLE I (200 EFPD)

Huclide Weiglat (kg)/ Segment Nuclide I I") ,2 3 4 5 6 Th-232 2776.9 2753.6 2594.2 2386.2 2865.9 2383.9 Pa-233 I U-233 21.1 21.0 20.0 18.8 22.5 18.8 U-234 1.3 1.3 1.5 1.6 1.9 1.6 U-235 62.8 63.4 85.5 100.4 111.4 100.4 I U-236 4.6 4.6 6.2 7.3 8.1 7.3 U-238 5.2 5.3 7.1 8.4 9.3 8.4 o, Np-239 + Pu-239 0.09 0.09 0.1 0.1 0.2 0.1 E) Pu-240 0.02 0.02 0.03 0.03 0.04 0.03 Pu-241 0.01 0.01 0.01 0.01 0.02 0.01 (a) Tliis segment is discliarged a.

?

an*

CLP-5646 j TABLE 5.2 USE OF BURNABLE POISON RODS IN SEGMENT 7 Upper Core Half (*} Lower Core Half (

gm B"""/cc 0.0391 0.0254 Nominal rod diameter (in.) 0.45 0.45 Nominal rod length (in.) 28.5 28.5 Rods / standard element 6 6 Rods / control element 0 0 l LBP inserted into fuel elements containing:

(a) fuel blends 14 and 16 (b) fuel blends 15 and 17 TABLE 5.3 CONTROL ROD SEQUENCE FOR CYCLE 2 0

l Cycle 2 l Initial Core (Cycle 1)

Control Rod Group Group Sequence Withdrawn Withdrawn Regions 1 3C(*} 23 (*} (3, 5, 7)

~

2 , 2A(*) 4F(*) (25, 31, 37) 3 43 4C (22, 28, 34) 3A 1 (half out) 1 (half out) (1) 4 4F 4D (23, 29, 35) 1 5 23 4A (20, 26, 32) 6 3A 2A (2, 4, 6)

7. 4D 4B (21, 27, 33) l 8 3B 4E (24, 30,'36) 9 4E 3B (9, 13, 17) 10 4A 3D (11, 15, 19) 11 3D 3C (10, 14, 18) o 12 4C 3A (8, 12, 16) 13 1 (last half) 1 (last half) (1)

(a). Rod groups used for red runback 5-9

TAllLE 5.4 -

s CALCUl.ATEI) CONTROL RO') CROUP WOR 111 AND POWER PEAKINC FACTORS WITil CYCLE 2 ROD SEQUENCE Control Rod- Croup Cumulat.ve Hax Ha Con fi gu ra ti on Worth # "

Worth Hag Tilt Tilt Rod Pal rs Inserted Ak Ak RPF Rodded Unrodded Worth Reg.

No rods in 0.0000 --

1.33 --

1.27 -- --

Rod 1 (half in) ~

0.0027 '.0027 O 1.31 --

1.25 0.0027 1 4 rods (+3A)(*} ,

0.0170 0.0197 1.38 1.33 1.26 0.0089 12 7 rods (+3C) 0.0171 0.0368 1.48 1.41 1.27 0.0127 12 19 rods (+3D) . 0.0180 0.0548 1.42 1.34 1.25 0.0144 12 13 rods (+38) ,

0.0172 0.0720 1.72 1.35 1.28 0.0146 12 16 rods (44 E) 0.0111 0.0831 1.54 1.40 1.33 0.0i83 12

, v.

19 rods (+411) ,

0.0112 0.0943 2.12 1.57 1 35 0.0181 19 l 22 rods (+2A) , 0.0254 0.1197 2.67 1.50 1. z t- 0.0266 30 25 rods (44A) -

0.0102 0.1299 28 rods (+4D) 0.0118 0.1417 28 rods (1 fully in) 0.0047 0.1464

'O 31 rods (+4C) 0.0092 0.1556 34 rods (14F) 0.0190 0.1746

3. rods (+211) , 0.0399 0.2145 i NOTE: Initial criticality at 0 days was calculated with bank 2A withdrawn 35 in. .

("}RPF = region suaking factor = region power / core average region power.

} Tilt = column peaking factor /RPF. d.

or

(") Refers to rod group sertuence (see Table 5.3) (4 rods a rod 1 + 3A group). S b} Power range defined Jn Section 5.3.

.__ C__________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. ___ _ . _ __ __ _ _ _ _ _

- GLP-5646 TABLE 5.5 CONTROL ROD BANK WORTE Cycle 2 BeginningI ") Middle End Total bank vorth, Ak O.217 0.216 0.218 (37 rod pairs inserted)

Total bank vorth less 0.176 0.174 0.178 ,

maximum worth rod pair 1j Total bank worth less two 0.135 0.140 0.139 maximum worth rod pairs {

l 1

(*) Xenon assumed to equilibrate. ,

i TABLE 5.6 SHUTDOWN MARGINS - CYCLE 2 Number of Inoperable Rods 30C MOC EOC O(*) 0.096 0.098 0.102 1(*) 0.053 0.054 0.060 2( 0.021 0.046 0.049 1 l

(*)With no or one rod pair inoperable, the core is assumed to be at room l camperature with co:; place Pa-233 decay.

( )With two rod pairs inoperable, the core is assumed to be at refueling temperature with a two-week Pa-233 decay.

o x

l 5-11

m -

u TABI.E 5.7 KINETICS PARAMETERS Initial Core Equilibrium Core BOC, with Xe 150 EFPD 310 EFPD BOC, with Xe EOC Fractional productions From U-233 0.0 0.19 0.32 0.38 0.48 From U-235 1.0 0.81 0.68 0.62 0.52 Prompt neutron lifetime, sec 2.69 x 10~ 3.17 x 10- -0 Ilot 3.17 x 10 2.85 x 10~ 3.41 x 10~

Cold 2.43 x 10~ 2.81 x 10~ 2.81 x 10

~4 2.64 x 10- 3.09 x 10

-4 T' Effective delayed neutron fraction 0.00650 0.00577 0.00527 0.00505 0.00451 U

Delayed neutron decay constant, A, y sec-1 Precursor 1 0.01243 0.01249 0.01249 0.01250 0.01251 2 0.03050 0.03088 0.03112 0.03126 0.03164 3 0.114 0.1136 0.1160 0.1170 0.1199 4 0.3013 0.3025 0.3040 0.3047 0.3063 5 1.136 1.136 1.136 1.135 1.135 6 3.013 2.981 2.932 2.913 2.859 Delayed neutron fraction, S Precursor 1 0.000214 0.000219 0.000219 0.000220 0.000222 2 0.001424 0.001 0.001224 0.001186 0.001099 3 0.001274 0.001 0.001082 0.001046 0.000961 flf 4 0.002568 0.0022 0.001984 0.001874 0.001619 E 5 0.000748 0.0005 0.000552 0.000516 0.000430 &

6 0.000273 0.0002 0.000215 0.000204 0.000179 n

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Fig. 5.5. Maximum allowable roit pair wortli vs' average gas outlet temperature

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Fig. 5.7. Temperature coefficient at emi of equLlitarium cycle g

_ _ _ _ _ _ _ _ _ _ n _ ___ _ _ _ _ _ _ - _ . - - - - - - _ _ _ - _ _ - - - _ _ _ - - _ _ _ _ -____ - _______-- .-.__ -_--_ _ _____ _ _

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Fig. 5.8. Temperature defect vs. average core temperature E

, GL?-5646

)

l l

6. THEFF.AL-HYDRAULIC DESIGN i As noted in Section 4.3, the thermal design of the segment 7 fuel l elements is identical to that of the initial core. No changes have been  !

1 made in fuel element geometry or materials. The power distributions ex-pected during the second cycle are within the envelopes defined by the Technical Specifications (LCO 4.1.3). Hence, except for the opening of j cross-flow gaps, as discussed and accounted for in Section 3.6.2.2 of the FSAR, core coolant flow characteristics are also unaffected. Accordingly, there are no changes in the design 14-4 ts of the segment 7 fuel ele =ents from those of the initial core or the equilibrium core.

.i

'* 1 6-1

, , GLP-5645

7. SAFETY ANALYSIS 7.1. IlrCODUCTION In this section, the Safety Analysis presented in Chapter XIV of the Fort St. Vrain FSAR is reviewed to determine potential effects of segment 7 fuel on accident events discussed in the FSAR. The purpose of such a review is to assure that:

1. The worst casa conditions previously defined for accident analyses are not exceeded during cycle 2 l1

2. Consequences of postulated accident conditions re ain within the bounds already found to be acceptable via the FSAR review.

3. Recent submittals to the NRCs which update certain FSAR analyses, will not be invalidated.

As a first step in this review process, Chapter XIV of the FSAR has been examined to identify analyses potentially affected by the insertion of segment 7 fuel. The results of this review are presented in Table 7.1.

Five accident conditions (some of which envelope other less severe events) have been identified as requiring = ore detailed review for potential effects. These are:

1. Rod withdrawal accidents.

2. Fuel element malfunctions.

l 3. Loss of normal shutdown cooling (li=iting case: cooldown on one firewater-driven circulator).  ;

4 Permanent loss of forced circulation. t

5. Rapid depressuri:ation/blevdown. t 7-1

GLP-5646 As indicated in Table 7.1, red withdrawal accidents and fuel element 1

malfunctions are discussed in Sections 5.5 and 4.2 of this document, re-4 spectively. The remaining three areas of safety analysis are discussed below.

7.2. LOSS OF NORMAL SHUTDOWN C00 LING, PERMANENT LOSS OF FORCED CIRCULATION (1 >FC), AND RAPID DEPRESSURIZATION/ BLOWDOWN (DBDA)

The core thermal conditions resulting from Firewater Cooldown (Safe Shutdown Cooling in the event of loss of normal shutdown cooling), LOFC and DBDA are known from past studies to be sensitive to specific core parameters on operation of the core which could be affected by a core segment reload.

These parameters are radial region peaking factor (RPF) and core outlet region temperatura dispersion (Mmatch), which, in turn, are l' d ead by the FSV Technical Specification Limiting Conditions for Operation (LCOs) 4.1.3 and 4.1.7, respectively. A recent submittal (Ref. 2), which updates the

. FSAR analysis of these three events, includes RET and temperatura dispersion values up to the LCO-allowable values (RPF = 1.83, dispersion = +200*F c indicated). The maximum RPF expected during cycle 2 is 1.48 at full power; the maximum temperature dispersion will be controlled not to exceed 200*F indicated using the variable-orifice flow-control assembly located at the inlet to each rpfueling region. Hence, insertion of segment 7' fuel does not result in core conditions during cycle 2 more severe than those already l1 analy:ed. Cycle 2 operation is therefore bounded by the recent accident analysis updata ruhmic al (Ref. 3). I l1 7.3. CONCLUSIONS i 1

A review of Chapter XIV of the FSAR identified five accident conditions that required more detailed examination for potential impact from insertion l of segment 7 fuel. No requirement for additional analysis has been V

r o

7-2

GLP-5646 identified; the FSAR analysis or recent reanalysis that updatec the FSAR analysis is found to remain valid in all cases. It is concluded chat:

1. The worst-case ccnditions previously defined for accident analyses are not exceeded during cycls 2. l1

2. Consequences of postulated accident conditions renain wir.hin the bounds already found to be acceptable during the FSAR review.

3. Recent submittals, which update certain FSAR analyses, will not be invalidated by the insertion of segment 7 fuel.

r 6

l

}

s 7-3

TAul.E 7.1 POTENTI Al. EFFECTS OF CYCI.E 2 ON FSV FSAR ACCIDENT PREI)lCTIONS Potential Effects on Event Analysis Due To FSAR Chapter XIV Event Presence of Segment 7 Fuel 14.1 Environmental Disturbances

- Earthquake

• Hone - Any reactivity effect would be bounded by rod withdrawal events.

~

- Wind effects

- Flood

- Fire -

The core is not affected by these events.

- Landslides

- Snow and Ice .

14.2 Reactivity Accid'ents and Transient

Response

[ - Susunary of reactivity sources Excessive removal o'f control poison I.ous of fission product poisons Reactivity insertions in these events are less than Rearrangement of core components -

rod withdrawal events.

Introduction of steam into the core Sudden det.rease in reactor temperature ,

- Roti withdrawat accidents Evaluation required; see Section 5.5 of this document.

14.3 lucidents Incidents involving the reactor core Column tieflection and misalignment No change from Sectica 3.3.1.2 of FSAR Fuel element mal functions Evaluation required; see Section 4.2 of this document.

Hinplaced fuel element No change from Section 3.5.4.5 of FSAR cg Blocking; of coolant channel No change f rom Section 3.6.5.2 of FSAk Io 8

Os

TABLE 7.1 (Continued)

Potential Effects on Event Analysis Due To FSAR Chapter XIV Event Presence of Segment 7 Fuel Control rod malfunctions No change from Section 3,8 of FSAR Orifice malfunctions No <.i.ange f rom Section 3.6.5.1 of FSAR Core support floor loss of cool'ing No change f rom Section 3.3.2.2 of FSAR incidents involving the primary coolant None system

- Incidents involving the control and Hone instrumentat.lon system

- Incidents involving the PCRV Hone

- Incidents involving the secondary Hone coolant and power conversion system

,, - Incidents involving the electrical system Hone l

v' - Halfunctions of the helium purification Hone system

- Halfunce ions of the helium atorage None system .

- Ha '4 function of the nitrogen system None 14.4 1.oss of hormal Shutdown Cooling Evaluation required; see Section 7.2 of this document 14.5- Secondary Coolant System Leakage

- Steam leaks outuide the primary coolant Hone system

- Leaku innide the primary coolant system / Hone; FSAR analysis encompasses core thermal con- '

uteam, generator leakage (moisture ditions allowable under '"ech. Specs.; cycle 2 will f lugrenn) not exceed same, f

______._____..___._.._._..._c2____.__.____ ___ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ __ _ ____________ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

TABl.E 7.1 (Continued)

Potential Effects on Event Analysis Due To FSAR Chapter XIV Event ' Presence of Segment 7 Fuel 14.6 Auxiliary System Leakage Failures involving the helium purifica-tion system -

Loss of both purification trains Possible effects would be bounded by Design Basis Failure of regeneration line - Accident No. 2, FSAR Section 14.11 w/ simultaneous valve failure and operational error ,

- Accidents involving the gas waste system No change from Section 14.6.2 of FSAR

- Fuel handling and storage accidents u '

& Fuel handling accidents Ho change f rom Section 14.6.3 of FSAR .

Fuel storage accidents 1 14.7 Primary Coolant I.cakage Possible effects would be bounded by Design asis Accident No. 2, MAR Section M.ll 14.8 Haximum Credible Accident ,

14.9 Haximum flypothetical Accident Same as FSAR Section 14.11 14.10 Design liasla Accident No. I Evaluation required; see Section 7.2 of this

" Permanent I.oss of Force Circulation document (1.0 FC)"

14.11 Design llasts Accident No. 2, " Rapid Evaluation required, see Section 7.2 of this, '

Depressu ri za t i on /lilowdown" document 9

?

-- 4 _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _

GLP-5646

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS No changes to the plant Technical Specifications pertinent to plant

)

operation are necessitated by the insertion of segment 7 into the reac-tor Core.

)

e e

I t 8-1

. i c!.P-5646 l

\

1 1

9. STARTUP TESTS l 1

1 Following refueling, a stepwise approach to full power will be l performed. During these power steps, the following tests will be per-  ;

! formed.

1. Differential cot.crol rod calibration measutmts. The method used l for these measurements is the same as that in SUT 3-9.

2. Temperature coefficient (temperature reactivity defect) measure-ments. The method used for these measurements is the same as that in SUT B-8.

3. Reactivity status survef11ance check. This test is performed at '

each startup and once per week as required by Technical Specification

, SR 5.1.4 ,

l

4. Region power factor distribution measurements and orifice valve a calibration studies, using the methods developed during SUT 3-4.

h 1

lu a

I 1

o 9-1

. __ . - . .