ML20245L406

Safety Analysis Rept for Reactor Defueling

ML20245L406
Person / Time
Site:	Fort Saint Vrain
Issue date:	08/31/1989
From:	GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
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Shared Package
ML20245L393	List: ML20245L388 ML20245L406
References
GA-C19694, NUDOCS 8908220110
Download: ML20245L406 (58)

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SAFETY ANALYSIS REPORT  !

, FOR REACTOR DEFUELING l l

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FORT ST. VRAIN NUCLEAR GENERATING STATION i

GENERAL ATOMICS PROJECT 1900 i' AUGUST, 1989 8908220110 890816 PDR ADOCh' 05C'OO267 P ppc

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u 7 CONTENTS 7'

1. INTRODUCTION AND

SUMMARY

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2. DEFUELING GENERAL DESCRIPTION .............. 2-1 2.1 Defueling Method ................. 2-1 2.2 Defueling Element Design ............. 2-5 2.3 Lumped Poison Pin Design ............. 2-8

3. NUCLEAR ANALYSIS . . . . . . . . . . , . . . . . . . . . . 3-1 3.1 Neutron Sources and Reactivity Monitoring . . . . . 3-1 3.2 Shutdown Margin During Defueling ......... 3-2

,. 3.3 Shutdown Margin Verification ........... 3-3 3.4 Effects of Further Depletion on Shutdown Margin . . 3-5 4

4. THERMAL-HYDRAULIC AND MECHANICAL ANALYSIS ........ 4-1 4.1 Thermal-Hydraulic Performance During Defueling .. 4-1 4.2 Mechanical Performance .............. 4-5

5. SAFETY ANALYSIS ..................... 5-1 5.1 Introduction ................... 5-1 5.2 Events- Requiring Further Evaluation . . . . . . . . 5-2 5.3 Events No Longer Credible . . . . . . . . . . . . . 5-13 5.4 Conclusions . . . . . . . . . . . . . . , . . . . . 5-14

6. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . . 6-1

7. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . 7-1 iii

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FIGURES -

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2-l' Reactor Core Defueling Sequence ...............

2-10 2 Defueling Element Top View (Dowels Not Shown) , . . . . . . .

. 2-11 L . 1- Maximum Region' Temperature. Rise During Defueling

@ 2.71 lbm/sec , . .-. . . . . ... . . . . . . . u- ..... 4-8 2: Core Flow Needed for Maximum Region Temperature Rise of 300*F ....................... 4-9 TABLES .

Da 3-1 Shutdown Margins During Defueling Sequence - 155 EFPD Burnup Cycle 4 ....................... 3-6 3-2 FSV Shutdown Margin Verification - 155 EFPD Burnup Cycle 4 .. 3-7'

~ 3-3 FSV Shutdown Margin Verification - 200 EFPD Burnup Cycle 4 .. 3-12 3-4 .FSV Shutdown Margin Verification - 250 EFPD Burnup Cycle 4 .. 3-13 3-S FSV Shutdown Margin Verification - 300 EFPD Burnup Cycle 4 .. 3-14 5-1 Potential Effects of Defueling on FSV FSAR Accident Predictions . . . . . . . . . . . . . . . . . . . . . . . . . 5-15

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1. INTRODUCTION AND

SUMMARY

f This Safety Analysis Report (SAR) is prepared to assess defueling of the reactor core at the Fort St. Vrain (FSV) Nuclear. Generating Station.

Public ~ Service Company of Colorado (PSC) has decided to cease nuclear operations ~at FSV no later than June 30, 1990. The reactor core will be defueled in preparation for decommissioning of the nuclear facility.

This report contains sections describing the method to be used for defueling, evaluations of the nuclear, thermal-hydraulic, and mechanical

- behavior of the core, and the safety aspects of the core during defueling.

It is concluded that no changes to the Technical Specifications are

.., required- to. proceed with defueling of the FSV core and that defueling presents no-unreviewed safety questions, as defined in 10CFR50.59.

c Changes to the Technical Specifications may be necessary near the end of.the defueling sequence, when the remaining fuel can be shcwn to be sub-critical with all control rods withdrawn, if the neutron count rate on the startup channels cannot be maintained at its minimum allowable value. Any proposed changes will be submitted to NRC at a later date to be determined as defueling activities proceed.

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2. DEFUELING GENERAL DESCRIPTION t

The method to be used for defueling of the Fort St. Vrain reactor core has been chosen based upon its similarity to refueling operations.

The reactor will be defueled one region at a time with the fuel handling machine using procedures similar to those for refueling. However, as each region is- removed,-. it will _ be replaced by graphite defueling elements to maintain core structural' stability. To ensure reactivity control, the f defueling' elements will . be boronated with lumped poison pins. The defueling process is expected to last approximately 30 months, depending

, on such variables as fuel handling machine. availability and spent fuel shipping schedules.

In this section a general description of the defueling method is presented, .and a description of the graphite defueling elements and lumped poison pins is given.-

2.1 DEFUELING METHOD Defueling of the FSV core will commence 100 days after final reactor

-shutdown or when the decay beat level assumed in Section 5.2.7 is

achieved.. Final shutdown will occur no later than June 30, 1990. Normal shutdown cooling will be established and maintained until the PCRV can be depressurized to normal refueling conditions while maintaining a core inlet helium temperature less than 165'F, as required by Technical

,.- Specifications LCO 4.5.2 and 4.7.1. Under these conditions, the effective fuel temperature will be about 200*F.

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Defueling of the core will be done one region at a time using the fuel . handling machine (FHM). As each region is removed, it will be replaced' by graphite defueling elements that are boronated with lumped

• l' poison pins. All replaceable reflector elements that'are remove'd during defueling of a region will be reinserted to their appropriate location upon loading- of the defueling elements. The metallic upper plenum elements will also be reinserted, although it may be necessary ir selected regions to' replace some of the existing plenum elements with' new plenum elements manufactured without boronated graphite granules. Removal of the boronated graphite from the metallic plenum elements may, as discussed in Section 3, be necessary to ensure an adequate neutron count rate on the startup channels. (The startup channels will be used to monitor cove reactivity during the defueling sequence.) Depending on their measured radiation levels, the region constraint devices (RCDs) will either be .

reinserted or stored out-of-core for subsequent shipment off site.

As fuel is removed from the reactor vessel, it will be transferred either' to the fuel storage wells to await shipment to the DOE-operated storage facility in Idaho or directly to a fuel shipping cask without pausing in the - fuel storage wells. At this time, it is planned that approximately 1242 of the 1482 total fuel elements will be shipped to Idaho for storage. The remaining spent fuel will be stored, at least in the near. term, in the fuel storage wells. PSC is considering several alternatives for long-term storage of the remaining spent fuel until it can be transferred to the DOE for permanent storage in a high-level waste repository. This could include the continued use of the fuel storage wells, construction of an Independent Spent Fuel Storage Installation (ISFSI), or some alternate means of storage. If PSC decides to build an ISFSI, either on the FSV site or at an off-site location, a license application for the installation will be filed with the NRC.

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I The analyses presented :in this report are based upon the'defueling-sequence. shown ' in Figure 1.- The defueling ' sequence was chosen with three primary objectives in mind:

1. - Retain a ' core geometry consistent with : that modeled in reference FSV core physics analysis methods,

2. Ensure - a sufficient shutdown margin at all points in the sequence, and

3. Ensure a neutron count rate' on the startup channels that is-adequate for monitoring core reactivity until such monitoring is no longer needed.

Two secondary objectives considered in developing the defueling sequence y . were.to minimize the number of fuel handling machine movements and to

. avoid interference in equipment movement on the reactor refueling deck.

s The defueling sequence'shown in Figure 2-1 essentially retains the right circular cylindrical geometry of the core that is modeled in the GAUGE code.; .Defueling is conducted from the outer ring of fuel elements (Regions 20-37) to the central regions (Regions 2-7 and 1). As a result, only the effective core radius is systematically reduced as the defueling ' sequence proceeds. This is acceptable because the GAUGE. code model has been used for reactor core sizes ranging from 4,000 MW(th) large HTGRs to 100 MW(e) n ,dular HT,GRs. Calculated shutdown margins for this sequence are discussed in Section 3, and compliance with the shutdown margin requirements of Interim Technical Specification 3.1.4

() 0.01 Ak) will be confirmed by measurements during the defueling sequence.

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- A minimum startup channel neutron sount' rate (4.2 counts per second 4 trip setpoint) is. specified by Technicabsspecification LCO 4.4.1, Table 4.4-4. . A count rate. less than this valueqctivates the rod withdrawal L prohibit -(RWP) funct fons to prevent controI(rod withdrawal. Source-calculations have been- conducted to determirih. the size of ' additional

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neutron source (s)'needed to ensure a count rate hat meets LC0

4.4.1. If. the measured count rate is found to h.; inadequate during defueling, the RWP does not allow defueling of fukher regions until additional sources can be procured and placed in the core or other

, measures enn be taken to meet the LCO requirements. Thse circumstances constitute an economic risk for PSC, but do not constigste a safety risk.

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-As the defueling sequence nears completion, a point will he reached .

at which the remaining fuel array will be suberitical with all hqmaining cortrol rods withdrawn from the array. At this point, experimUtal ..

confirmation of adequate shutdown margin will no longer be necessarh A verification of subcriticality with all control rods removed from tk core will be conducted when the calculated k-effective of the remaininh fuel, with all control rods withdrawn, is {0.95. However, the startup ,

channels will continue to be kept in service, although the count rate '

- will probably be reduced-to less than the LCO 4.4.1 requirement as the last fuel regions are removed. If the neutron count rate is reduced to ',

- less 'than-the LCO 4.4.1 requirement, it may be necessary to bypass the 2 control . rod - withdrawal ~ prohibit function provided by the plant -

protection system, which would require a change to LCO 4.4.1. A ,

calculated projection of when the point of subcriticality with all control rods removed will be reached in the defueling sequence is presented in Section 3.

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'All- evaluations in this report are based on the' defueling sequence-shown:in Figure.2-1. It-is possible that, as the.defueling proceeds, it-lO' *will be necessary to ' alter the sequence. A sequence alteration could be required as a-result of several considerations, including:

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mechan.ical problems or equipment interference l:

inadequate ' count rate on the startup channels

- excessive, nonconservative deviation between measured and expected startup channel count rate changes before and after a region is defueled.

,.-_ If sequence alteration becomes necessary, the revised sequence will

.be fully.. evaluated and justified prior to implementation under the provisions of 10CFR50.59. The results presented in this safety analysis report will be used 'as the basis for determining whether the sequence alteration involves an unreviewed safety question as defined in 10CFR50.59.-

2.2 DEFUELING ELEMENT DESIGN The defueling elements used to replace the active core fuel elements as - each region is defueled are similar to a ' replaceable standard top. reflector element as described in Section 3.4.1 of the Updated FSAR. All defueling elements will be of the same design, regardless of whether they-are used to' replace standard fuel elements or control column fuel elements from the central column of the region.

Defueling elements will not have control rod channels or reserve shutdown holes. Consequently, the control rods in a defueled region A.

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will not be able to be reinserted into the core and will remain fully withdrawn into -the top head penetration and locked in the " shipping position." This is . satisfactory .with regard to reactivity control

-because the defueling elements are sufficiently boronated by the lumped poison pins that insertion of control rods or reserve shutdown material would have negligible effect on shutdown margin.

The significant differences between replaceable standard top reflector elements and the defueling elements are the following.

The length of the defueling elements will be equal to that of the fuel elements they replace, 31.22 inches, whereas the replaceable top reflector elements are only one-half of this length.

The defueling elements are made from grade H-091 graphite (HLM equivaient) rather than H-327 or H-451 graphite. This graphite was -

chosen over H-327 and H-451 because H-327 graphite is ' no longer available, and H-451 graphite is of a level of purity that is not necessary for this application. H-091 graphite is formulated, extruded, baked, and graphitized by the same techniques as HLM graphite. The supplier has. designated it as H-091 because of its QA requirements, which exceed the requirements of commercial grade HLM graphite. HLM graphite is the material used in the FSV permanent side reflector elements. Its material characteristics are well established, and the procurement requirements for the defueling elements invoke the same characteristics as originally required for the permanent side re-flectors. Although the defueling element graphite precursor materials will be obtained from batches that are different from those used to manufacture the permanent side reflectors, qualification tests and acceptance criteria used in previous FSV graphite procurement have been specified to ensure an equivalent product. As discussed in Section 4, "

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I the mechanical and. thermal-hydraulic performance of this material has been evaluated and found to be acceptable for this application.

Twelve coolant holes (5/8 inch diameter) in the defueling elements will be drilled blind, i.e., to a depth of at least 1/4 inch short of

-the bottom of the element. These holes will be used to receive and hold the lamped poison pins and will be plugged following insertion of the pins. Figure 2-2 presents a top view of 'a defueling element and

' indicates which coolant holes will be drilled blind.

In addition to having twelve coolant holes drilled blind, the defueling elements will not feature the six 1/2 inch diameter coolant holes that are drilled around the fuel element . handling hole of a

.. replaceable top reflector element. These six holes are provided in the top reflector, elements to direct coolant to matching holes in the active core fuel elements below, thereby cooling fuel rods located in fuel channels near the fuel element handling hole. Since the defueling.

. elements will _ have no fuel, the 1/2 inch holes have been deleted from

'the defueling element design.

Deletion of the six 1/2 inch cooling holes, blind drilling of twelve 5/8 inch holes, and elimination of control rod and reserve shutdown channels result in a net 10% reduction in seven column region flow area relative to that available in the fueled core and an 8%

reduction in five column region flow area. The effects of'this change in flow area on flow distribution a,re evaluated in Section 4.

To ensure traceability and accountability during the loading of lumped poison pins and during the defueling process, each defueling element will be identified with a unique serial number, to be engraved

. on one side face of the element.

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2.3 LUMPED POISON PIN DESIGN I Additional shutdown margin during the core defueling will be '

ensured by use of boronated graphite lumped poison pins. The pins will be loaded into each of the twelve 5/8 inch diameter defueling element cooling holes that will be drilled blind, shown in Figure 2-2. The lumped poison pins are basically similar to the lumped burnable poison rods that were used in the FSV initial core and in each reload segment and will be manufactured to similar quality requirements. They consist of boron carbide granules dispersed homogeneously in a matrix of graphite flour, bonded with carbon-based or phenolic resin, and shaped into a right circular cylinder with nominal dimensions of 0.55 inch diameter and approximately 0.55 inch length. The mean boron density is specified to be 0.30

• 0.02 gm natural boron /cm3, which is about ten -

times higher than the boron density of lumped burnable poison rods, and is approximately equal' to the boron density in the reserve shutdown material used in Regions 1-19. The pins will be stacked into sleeves that will be used to assist loading of the pins into the blind holes to a total height of 28.5

• 0.5 inch. The resultant boron loading of the defueling elements is such that defueling of any region always results in a decrease in core reactivity.

The lumped poison pins will be loaded by hand into the defueling i elements at the FSV site prior to defueling using procedures and controls (QC checkpoints) similar to those that have been used to load l

lumped burnable poison pins into reload fuel elements prior to each refueling. After each blind hole is filled with poison pins, it will be closed at the top with an H-091 graphite plug that will be glued into

place with graphite cement.

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t The number, boron loading, and size- of lumped poison . pins to be

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loaded into, each defueling element was determined by PSC based upon i

• GAUGE code analyses of hypothetical scenarios .in which the defueling i elements.were assur.ed to be made from pure graphite. It was established-that the equivalent'of about 100 PPM of homogeneously dispersed natural boron'is required to ensure subcriticality during the defueling sequence and that use of six blind holes in each defueling element would be

' sufficient. To provide additional margin, it was decided to double this loading to twelve blind holes per element, which is equivalent to about 350 PPM of homogeneous natural boron. This loading' assures that small statistical variations of boron loading in each defueling element will 3 have a negligible effect on core reactivity. The heavy boronation of the dummy elements compensates for the absence of control rods in the i

, defueled regions. i The effects of the lumped poison pins on the core reactivity (shutdown margin) are evaluated in.Section 3. For these analyses, the GAUGE code model for lumped burnable poison rods was used, with appropriate adjustments .in self-shielding factors for the larger diameter and higher boron loading of the poison pins.

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LEGEND REGION NUMBER 3Z IC INITIAL CORE SEQUENCE NUMBER 8,9 SEGMENT 8 OR 9-D DEFUELING s

Figure 2-1 Reactor Core Defueling Sequence 2-10

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O O O O O O O O O O O O l O

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O O O O

O O O O O O O O EtEnEN1 O O HANDLING HOLE I

O O O O O O O coo'^"t Hotes BLIND HOLES -

LUMPED POISON PINS Figure 2-2 Defueling Element Top View (Dowels Not Shown) 2-11 l

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3. NUCLEAR ANALYSIS 1

1 In this section,. the nuclear behavior of the reactor core during defueling is analyzed. - Addition of neutron sources to enhance reactivity monitoring is discussed. Analyses of shutdown margin during defueling are presented, and shutdown margin verification is discussed.

3.1 NEUTRON SOURCES AND REACTIVITY MONITORING 1

l As discussed in Section 2.1, core- reactivity during the defueling l.

. sequence will be monitored via the startup channel count rate. A I minimum count rate is specified in Technical Specification LC0 4.4.1.

, Source calculations conducted to determine the size of additional neutron sources needed to ensure the minimum count rate indicate that two new neutron sources, each with ' a strength of 4 x 109 ' n/sec, are sufficient. One neutron source will be located, as shown in Figure 2-1, L in Region 3 The other source will be located, as also shown in the figure, .in Region 6. These regions are the last two regions in the defueling sequence. Both sources will be placed in the top layer of the i active core by replacing two standard fuel elements with unused, empty fuel elements or with full length reflector elements, each containing one source. The initial core sources, located in the top active core layer in Regions 9 and 16, and the reload Segment 8 and 9 sources, located in the top active core layer of Regions 15 and 22, respectively, will also continue to contribute, to varying degrees, to the startup L channel count rate.

. Reactivity will be monitored via the startup channels throughout the defueling sequence. The neutron count rate after each region is defueled will be compared with that prior to the defueling step to confirm tht no unexpected changes in core reactivity have occurred. As 1 3-1

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. discussed-in Section 2'.1, the startup channel monitors will be kept in service even after the point late .in the defueling sequence when the

' current -LCO requirement may no longer be met. To ensure that the '

minimum count rate is maintained for as long as possible during the defueling sequence, the non-control rod metallic plenum elements in several core regions may be removed and replaced with new elements manufactured without boronated graphite granules. This change may be.

necessary because the boronated graphite will tend to shield the startup channels, which are located in the PCRV top head, from multiplied neutrons originating in the core. The regions to be equipped with new metallic plenum elements would be Regions 16, 6, 3, and 10. These regions, as can be seen in Figure #-1, are in a direct line between the new neutron sources and the.startup channels.

3.2 SHUTDOWN MARGIN DURING DEFUELING An analysis has.been done to determine shutdown margin during the defueling sequence shown in Figure 2-1. These calculations were condu~c ted with the GAUGE code and the following conservative assumptions.

Reactivity was determined based upon fuel loadings at 154.5 EFPD burnup-in Cycle 4. Continued depletion through the cycle will further increase the shutdown margins. Credit was taken for the control rods remaining in the fueled regions and for the lumped poison pins in the j

defueling elements. For the lumped poison pins it was assumed that their boron loading, diameter, and stack height are at the lower end of the specified range. Full decay of xenon and Pa-233 was also assumed, and the core was assumed to be at 80'F (i.e., cold shutdown). At each step in the sequence, it was also assumed that the control rod pair in Region 1 is fully withdrawn for standby reactivity control. It is not -

planned at this time to actually " cock" this rod pair during defueling due to its relatively low reactivity worth (~0.0004 Ak).

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Shutdown margin. calculations were conducted for each step in the defueling sequence. It was assumed that, in addition to the' control

' rods in Region 1, the' control rods in the region about to be defueled

.and in the subsequent region in the defueling sequence are fully withdrawn. The results of these . analyses are shown in Table 3-1.

Shutdown margin is'shown in each case to be in excess of the 0.01 Ak value required by Interim Technical Specification LC0 3.1.4.

With regard to shutdown margin as it relates to requirements to maintain a minimum neutron count rate on the startup channels, Table 3-1 indicates that with seven regions (Regions 1-7) remaining in the core,-

.the calculated k-effective (to two significant figures) of the array with all . control rods removed is 0.95. Thus, as discussed in Section p . 2.1, the results in Table 3-1 project that it will not be necessary to maintain a minimum count rate after seven regions remain in the core.

As previously noted, however, the startup channels will be maintained in service even if the count rate falls below the minimum requirement to detect any increase in count rate.

An analysis was also performed of the shutdown margin during shutdown margin experimental verification. The results of this analysis are shown in Table 3-2 and are discussed in Section 3.3.

l 3.3 SHUTDOWN MARGIN VERIFICATION l

During defueling the region being defueled and the next region in the sequence to be defueled will, at some times, have their control rods removed. To confirm that a shutdown margin of at least 0.01 Ak is maintained throughout the defueling sequence, the shutdown margin will be confirmed via controlled testing. The test procedure is basically

. the same as that used to confirm shutdown margin during refueling. With control rods withdrawn from the region to be defueled and the subsequent region in the sequence, one or more control rods with a total calculated 3-3 L. _ _ __ _ ______ ____ _ __ l

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worth of 0.01' Ak plus' the temperature defect between the actual fuel temperature (~200*F) and 80'F will be withdrawn from the core.to confirm that the reactor remains suberitical and that, upon reinsertion of the '

ro'd(s), the cold shutdown margin is at least 0.01 Ak.

Table 3-2 presents the results of shutdown margin analyses throughout the defueling sequence with various regions with high worth control rods chosen for shutdown margin verification via ' control rod withdrawal (SDM verif. regions). The purpose of these analyses was to determine whether withdrawal of a control rod of high worth could result

. in inadvertent' criticality during shutdown margin verification testing.

In ' none of the cases evaluated was reactor criticality predicted.

However, in a few ' instances, a k-effective larger than 0.99 was calculated. Since the uncertainty of such analyses is

• 0.01' Ak, it  :

. must be considered possible that reactor criticality could be achieved

- if control rods were withdrawn from these regions to confirm shutdown .

margin. The cases of interest are highlighted via an asterisk in Table 3-2.

.The results indicate that until the first five regions have been defueled, inadvertent withdrawal of the control rod from Region 22, Region 28, or Region 33 could in some cases result in reactor criticality assuming the maximum uncertainty of 0.01Ak. To provide positive assurance that this inadvertent criticality cannot occur, the power source to the control rod drives in each of these regions will be enmpletely dir, connected unti,1 the f,1rst five regions have been defueled.

Shutdown margin verification testing will continue through the defueling sequence until the remaining fuel array has been demonstrated to be subcritical with all control rods removed from the core. At this point, shutdown margin verification testing will no longer be necessary. -

The demonstration of subcriticality with all control rods removed from the core will not be conducted until the calculated k-effective of the ..

remaining fuel, with all control rods withdrawn, is {0.95.

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3.4 EFFECTS OF FURTHER DEPLETION ON SHUTDOWN MARGIN As discussed in Section 3.2, the shutdown margins presented in Tables 3-1 and 3-2 were calculated based upon fuel loadings at 154.5 EFPD burnup in Cycle 4. Further depletion through the cycle will further increase the shutdown margins.

The effects of further Cycle 4 depletion on calculated shutdown margin are shown in Tables 3-3 through 3-5, which present calculated results for Cycle 4 depletion of 200, 250, and 300 EFPD, respectively.

The cases presented in these tables are primarily those from Table 3-2 where inadvertent criticality could occur during shutdown margin verification and those near the end of the defueling sequence, when all control rods can be removed with a calculated k-effective less than 0.95.

The results tidicate that if a total burnup of 200 EFPD is accumulated during Cycle 4 (Table 3-3), only the control rod drive in Region 33 will need to be disconnected from its power source to preclude inadvertent criticality. At 300 EFPD total Cycle 4 burnup (Table 3-5),

it will not be necessary to disconnect any of the control rod drives from their power sources.

With regard to maintaining subcriticality near the end of the defueling sequence with all control rods withdrawn, Table 3-3 indicates that with 200 EFPD burnup in Cycle 4 subcriticality can be maintained, with calculated k-effective less than 0.95, with eight regions remaining in the core, and with calculated k-effective equal to 0.95 (to two significant figures) with nine regions remaining. At 250 EFPD (Table 3-4) and at 300 EFPD (Table 3-5), subcriticality can be maintained, with

. calculated k-effective less than 0.95, with nine regions remaining.

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Table 3-1 Shutdown Margins During Defueling Sequence 155 EFPD Burnup Cycle 4 ,

Regions Rods Defueled Pulled' K-eff 0.9183 1,23,32 0.9421 23 1,32,26 0.9417 32 1,26,35 0.9347 26 1,35,29 0.9443 35 1,29,20 0.9438 29 1,20,27 0.9256 20 1,27,36 0.9248 27 1,36,28 0.9211 36 1,28,37 0.9151 28 1,37,25 0.9214 37 1,25,33 0.9294 25 1,33,24 0.9284 33 1,24,34 0.9088 24 1,34,30 0.9085 34 1,30,21 0.9226 30 1,21,31 0.9223 21 1,31,22 0.9045

31 1,22,13 0.9184 22 (Ring 4) -------

0.8679 22 1,13,19 0.9182 13 1,19,12 0.8936 19 1,12,18 0.8845 12 1,18,11 0.8887 18 1,11,17 0.8892 11 1,17, 8 0.8714 17 1, 8,14 0.8640 8 1,14, 9 0.8657 14 1, 9,15 0.8682 9 1,15,10 0.8587 All rods 0.9750 15 1,10,16 0.8413

.All rods 0.9563 10 1,16, 2 0.8559 All rods 0.9518 16 (Ring 3) -------

0.7687 1,2,5 0.8764 All rods 0.9508 2 1,5,7 0.8492 All rods 0.9185 -

5 1, 7, 4 0.8031 All rods 0.8505 7 All rods 0.8412 .

4 All rods 0.7188 1 All rods 0.6533 6 All rods 0.6533 3-6

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Table 3-2 (page 1 of 5)

FSV Shutdown Margin Verification 155 EFPD Burnup Cycle 4 i

Regions # Regions Rods SDM Verif.

Defueled Rertaining Pulled Region K-eff

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37 ------- ------

0.9183 1,23,32 ------

0.9421 1,23,32 3 0.9055 1,23,32 15 0.9636 1,23,32 16 0.9770 1,23,32 17 0.9669 1,23,32 22 0.9882*

1,23,32 28 0.9566 1,23,32 31 0.9679 1,23,32 33 0.9988*

R23' 36 1,32,26 ------ 0.9417

1,32,26 16 0.9770 1,32,26 22 0.9464 1,32,26 33 0.9988*

a R23,32 35 1,26,35 ------

0.9347 1,26,35 18 0.9823 1,26,35 34 0.9684 1,26,35 36 0.9831 R23,32,26 34 1,35,29 ------

0.9443 1,35,29 13 0.9701 1,35,29 14 0.9799 1,35,29 28 0.9910*

1,35,29 30 0.9702 R23,32,26,35 33 1,29,20 ------

0.9438 1,29,20 13 0.9701 1 1,29,20 14 0.9799 j

. 1,29,20 28 0.9910*

1,29,20 30 0.9702 1

R23,32,26,35,29 32 1,20,27 ------

0.9256 1,20,27 3 0.9628 1,20,27 4 0.9668 1,20,27 13 0.9661 R23,32,26,35,29, 31 1,27,36 ------

0.9248 20 1,27,36 3 0.9618 1,27,36 4 0.9666 1,27,36 13 0.9660 3-7 l 1

4 f

I Table'3-2 (page 2 of 5) ,

FSV Shutdown Margin Verification i

~155 EFPD Burnup Cycle 4 ,

Regions # Regions Rods SDM Verif.

Defueled' Remaining Pulled Region K-eff a

R23,32,26,35,29, 30 1,36,28 ------

0.9211 20,27 1,36,28 3 0.9604 i 1,36,28 4 0.9613 1,36,28 18 0.9495 R23,32,26,35,29, 29 1,28,37 ------

0.9151 20,27,36 1,28,37 3 0.9598 1,28,37 4 0.9610 1,28,37 5 0.9515 R23,32,26,35,29, 28 1,37,25 ------ 0.9214 20,27,36,28 1,37,25 3 0.9623 1,37,25 4 0.9630 1,37,25 11 0.9596

• R23,32,26,35,29, 27 1,25,33 ------

0.9294 20,27,36,28,37 1,25,33 16 0.9489 '

1,25,33 17 0.9715 1,25,33 34 0.9582-R23,32,26,35,29, 26 1,33,24 ------

0.9284 20,27,36,28,37,25 1,33,24 16 0.9488 1,33,24 17 0.9714 1,33,24 34 0.9582 R23,32,26,35,29, 25 1,24,34 ------

0.9088 20,27,36,28,37,25, 1,24,34 3 0.9576 33 1,24,34 4 0.9560 1,24,34 5 0.9456 R23,32,26,35,29, 24 1,34,30 ------

0.9085 20,27,36,28,37,25, . 1,34,30 3 0.9545 33,24 1,34,30 4 0.9549 1,34,30 5 0.9474 R23,32,26,35,29, 23 1,30,21 ------

0.9226 20,27,36,28,37,25, 1,30,21 3 0.9585 33,24,34 1,30,21 9 0.9623 1,30,21 22 0.9737 -

R23,32,26,35,29, 22 1,21,31 ------

0.9223 -

20,27,36,28,37,25, 1,21,31 3 0.9582 I i

33,24,34,30 1,21,31 9 0.9623 l 1,21,31 22 0.9736 l

3-8

)

I

. Table 3-2 (page 3 of 5)

FSV Shutdown Margin Verification 155 EFPD Burnup Cycle 4 Regions # Regions Rods SOM Verif.

Defueled Ren,aining Pulled Region K-eff R23,32,26,35,29, 21 1,31,22 ------

0.9045 20,27,36,28,37,25, 1,31,22 3 0.9539 33,24,34,30,21 1,31,22 4 0.9539 1,31,22 5 0.9424 R23,32,26,35,29, 20 1,22,13 ------

0.9184 20,27,36,28,37,25, 1,22,13 3 0.9583 33,24,34,30,21,31 1,22,13 4 0.9718 1,22,13 5 0.9424 Ring 4 19 ------- ------

0.8679 1,13,19 ------

0.9182 1,13,19 3 0.9558 1,13,19 4 0.9717 1,13,19 18 0.9204 Ring 4 + R13 18 1,19,12 ------

0.8936 1,19,12 3 0.9467 1,19,12 4 0.9327 1,19,12 18 0.9014 Ring 4 + R13,19 17 1,12,18 ------

0.8845 1,12,18 3 0.9431 1,12,18 4 0.9299 Ring 4 + R13,19, 16 1,18,11 ------

0.8887 12 1,18,11 3 0.9521 1,18,11 4 0.9265 Ring 4 + R13,19, 15 1,11,17 ------

0.8892 12,18 1,11,17 3 0.9521

- 1,11,17 4 0.9266 Ring 4 + R13,19, 14 1,17, 8 ------

0.8714 12,18,11 1,17, 8 3 0.9152 1,17, 8 4 0.8966 1,17, 8 5 0.9056 All rods ------

1.0152 Ring 4 + R13,19, 13 1, 8,14 ------

0.8640 l 12,18,11,17 1, 8,14 3 0.9133 1, 8,14 4 0.8932 1, 8,14 5 0.9047 All rods ------

1.0045 l

l 3-9 ah__________._ .- _ _ . - _ _

Table 3-2 (page 4 of 5)

FSV Shutdown Margin Verification 155 EFPD Burnup Cycle 4 ,

Regions # Regions Rods SOM Verif.

Defueled Remaining Pulled Region K-eff l Ring 4 + R13,19, 12 1,14, 9 ------

0.8657 12,18,11,17,8 1,14, 9 3 0.9206 1,14, 9 4 0.8945 1,14, 9 5 0.9051 All rods ------

0.9989 Ring 4 + R13,19, 11 1, 9,15 ------

0.E582 12,18,11,17,8,14 1, 9,15 3 0.9208 1, 9,15 4 0.8950 1, 9,15 5 0.9020 All rods ------

0.9902 Ring 4 + R13,19, 10 1,15,10 ------

0.8587 12,18,11,17,8,14, 1,15,10 3 0.8922 9 1,15,10 4 0.8877 1,15,10 5 0.8987 '

All rods ------

0.9750 Ring 4 + R13,19, 9 1,10,16 ------

0.8413 12,18,11,17,8,14, 1,10,16 3 0.8871 9,15 1,10,16 4 0.8789 1,10,16 5 0.8668 All rods ------

0.9563 Ring 4 + R13,19, 8 1,16, 2 ------

0.8559 12,18,11,17,8,14, 1,16, 2 3 0.8942 9,15,10 1,16, 2 4 0.8874 1,16, 2 5 0.8769 All rods ------

0.9518 Ring 4 + Ring 3 7 - ------- ------

0.7687 1,2,5 ------

0.8764 1,2,5 3 0.9065 1, 2, 5 4 0.9091 All rods ------

0.9508 Ring 4 + Ring 3 6 ------- ------

0.7428

+ R2 1,5,7 ------

0.8492 1,5,7 3 0.8675 1,5,7 4 0.8894 All rods ------

0.9185 3-10 l

l

_ l

Table 3-2 (page 5 of 5)

FSV Shutdown Margin Verification 155 EFPD Burnup Cycle 4 Regions # Regions Rods SDM Verif.

Defueled Remaining Pulled Region K-eff Ring 4 + Ring 3 5 ------- ------

0.6875

+ R2,5 1, 7, 4 ------

0.8031 All rods ------

0.8505 Ring 4 + Ring 3 4 ------- ------

0.6819

+ R2,5,7 All rods ------

0.8412 Ring 4 + Ring 3 3 ------- ------

0.5670

+ R2,5,7,4. All rods ------

0.7188 Ring 4 + Ring 3 2 ------- ------

0.5029

+ R2,5,7,4,1 All rods ------

0.6533 Ring 4 + Ring 3 1 ------- ------

0.5029

+ R2,5,7,4,1,6 All rods ------

0.6533 0

4

.e 3-11 l

s; i

l i Table 3-3 FSV Shutdown Margin Verification l- 200 EFPD Burnup Cycle 4 ,

Regions Rods SDM Verif.

Defueled Pulled Region k-eff

---

------- ------ 0.9133 ,

1,23,32 ------

0.9368 1,23,32 22 0.9852 1,23,32 33 0.9960*

R23 1,32,26 ------

0.9363 ,

1,32,26 33 0.9960*  !

i R23,32,26 1,35,29 ------

0.9400 1,35,29 28 0.9858 l R23,32,26,35 1,29,20 ------ 0.9396 l 1,29,20 28 0.9858  ; I 1

Ring 4 + R13,19, 1,10,16 ------

0.8373 I l 12,18,11,17,8,14, All rods ------ 0.9519 -

9,15 l Ring 4 + R13,19, 1,16, 2 ------

0.8518 i 12,18,11,17,8,14, All rods ------ 0.9472 I 9,15,10 Ring 4 + Ring 3 1,2,5 ------ 0.8714 l All Rods ------

0.9462

! 3-12 l

l i.

i

i: .. .

l 1

Table 3-4 FSV Shutdown Margin Verification 250 EFF0 Burnup Cycle 4 Regions Rods SDM Verif.

Defueled Pulled Region K-eff 0.9079 1,23,32 ------ 0.9312 1,23,32 22 0.9817 1,23,32 33 0.9926*

R23 1,32,26 ------

0.9306 1,32,26 33 0.9926*

R23,32,26 1,35,29 ------

0.9353 1,35,29 28 0.9802

~

R23,32,26,35 1,29,20 ------

0.9350 1,29,20 28 0.9802 Ring 4 + R13,19, 1,10,16 ------

0.8330

12,18,11,17,8,14, All Rods ------

0.9473 9,15 Ring 4 + R13,19, 1,16, 2 ------

0.8475 12,18,11,17,8,14, All Rods ------ 0.9423 9,15,10 Ring 4 + Ring 3 1, 2, 5 ------

0.8662 All Rods ------

0.9414 3-13

Table 3-5 FSV Shutdown Margin Verification 300 EFPD Burnup Cycle 4 ,

Regions Rods SDM Verif.

Defueled Pulled Region k-eff 4

0.9014 1,23,32 ------ 0.9244 1,23,32 22 0.9767 1,23,32 33 0.9875 R23 1,32,26 ------

0.9238 1,32,26 33 0.9874 R23,32,26 1,35,29 ------

0.9296 1,35,29 28 0.9738 R23,32,26,35 1,29,20 ------

0.9293 ,

1,29,20 28 0.9738 Ring 4 + R13,19, 1,10,16 ------

0.8281 12,18,11,17,8,14, all rods ------

0.9417 9,15 R1ng 4x& R13,19, 1,16, 2 ------

0.8427 12,10,11,17,8,14, All Rods ------ 0.9366 9,15,10 Ring 4 + Ring 3 1,2,5 ------

0.8606 All Rods ------ 0.9357 O

3-14

+:..

) -

'T 4. THERMAL-41YDRAULIC AND MECHANICAL ANALYSIS In thisf section, supporting analyses of. core . thermal-hydraulic performance and minimum flow requirements for the core during the defueling; sequence' are presented, and the ' structural-mechanical-performance of the defueling elements is. discussed.

4.1 ' THERMAL-HYDRAULIC PERFORMANCE DURING DEFUELING During defueling, Technical Specification 4.1.9 will impose limits

~on core thermal . performance. .LC0 4.1.9 limits the maximum region temperature ' rise to 600*F when the orifices are adjustW to uniform region flow positions, and to 350*F when the orifices.are set at any

.. f

' other positions. At shutdown conditions, the orifice valves are generally . set 'at uniform flow positions, and the 600*F limit may be

. applied. However, 'as soon as the control rod drive and orifice assembly is removed from the first region in preparation for defueling, it will not be possible to maintain uniform flow through all regions, and it will be necessary to meet the 350*F limit. After the fuel in the first region is replaced with defueling elements and the orifice valve is reinstalled, uniform region flow positions will be different from region to region due to the different cooling geemetry cf the'defueling olement vs. the active com star.dard and contral fuel elements. Therefore, to avoid confusion and differences of interpretation of this limit during defueling, it is intended to' meet the 3305F linitt daring the entire defueling operation.

f Note:that the LCO 4.1.9 limit is valid throughout defueling, even though the core configuration changes, because tne limit- applies to individual regions. The total core flow must be sufficient to limit the temperature rise in any region to 350*F regardless of 'iow many regions

. have been defueled. l 4-1 L__ ___ _ - . _ _ _ .-- .- __ .-

o .- )

Analyses- were conducted with the P0XE code to assess the thermal- -

)

. hydraulic performance of the core during defueling. The purpose of these

- calculations was to show, in general, how temperatures will evolve during-defueling and to demonstrate that they will change slowly with time and remain low. It was assumed that the reactor had been shutdown for 100 days following. prolonged operation at 83% power. PCRV pressure was assumed to be atmospheric (12.3 psia). A core inlet temperature of 100'F

- was' assumed, with orifice valves at uniform flow positions of 17% and 10%

open for . seven and five column regions, respectively. For these calculations, the core power distribution was taken from the Cycle 4 extended operation analysis (Ref.1).

The actual _ conditions during defueling may differ somewhat from those assumed in-these calculations. The actual decay heat may be lower -

if the reactor does not operate at a sustained 83% power level prior to shutdown.. . The core inlet temperature may be as high as 165'F, and the .,

primary system pressure may differ somewhat from 12.3 psi. The core power distribution at shutdown will be flatter than during operation, due in large part to the long mean free path of the gammas. Also, this power distribution, changes in time. Initially, it mirrors the power distribution just prior to shutdown, but later it becomes raore of a composite of the power distributions from the weeks (and months) before shutdown. - Further, as regions are defueled, the total heat generation in adjecent regions will decrease as a result of reduced gamma heating from the defueled regions.

l These differences, however, are not critical, since the purpose of t-- these calculations was to show tiow temperatures will evolve during aefueling and to demonstrate that they will change slowly and remain low.

At the time that defueling begins, the actual core flow required to art LCO 4.1.9.will be determined and supplied.

L .

4-2

4

~

The uniform flow positions of 17% and 10% open were assumed since they have been used in the past. LC0 4.1.9 specifies the permissible

~

range of uniform flow orifice positions (seven column region) to be between 8% and 20% open. The more open positions are preferable and will be used since they allow more flow through the fueled regions when the elements of one region have been removed, and thus reduce the total core flow required to meet LC0 4.1.9. Also, they minimize the loop resistance and circulator power to deliver the required flow. However, these differences are small at refueling or defueling conditions when pressure drop through the coolant holes dominates relative to the orifice pressure drop due to the high friction factors associated with laminar flow.

A P0KE code calculation was performed with the first region removed

. (Region 23), and where the core flow was adjusted to produce a maximum region temperature rise of 300*F. This flow rate (2.71 lbm/sec) was then used for a series of poke calculations, where each region was defueled in the scheduled sequence and replaced with defueling elements. In these calculations it was assumed that the power in the fueled regions remains constant until they are defueled (i.e., no reduction in decay heat with time er as a result of neighboring regions being defueled), and that there is no power in the defueled regions. A maximum temperature rise of 300*F was used to provide margin relative to the 350'F limit of LCO 4.1.9.

The results of these calculations are shown in Fig. 4-1, which shows the maximum region temperature rise as a function of the number of regions removed. Two curves are given in this figure, cne for a core configuration where one region is etipty (whichever is being defteled),

and one wnere all regions are loaded (either with fuel or defueling elements) with their orifices in place.

+

4-3

~

The first regions.to be defueled in the proposed defueling sequence are the five column regions. As can be- seen- in Figure 4-1, the flow bypassing the core through an empty five column region causes the maximum

- region temperature rise to increase by about 50'F when the region is removed. Consequently, to limit the maximum region temperature rise to the previously stated 300'F with one five column region removed, the maximum region temperature rise at the start of defueling, prior to

, removing'any' fuel elements, would need to be about 250*F.

l' The flow through an empty seven column region is larger than that through an empty five- column region, so the increase in coolant temperature rise is larger. This can be seen in Figure 4-1, where for the seventh region to be removed (the first seven column region) the maximum region temperature rise jumps to about 320*F. ,

l As defueling progresses, the maximum regLl temperature rise ,

increases gradually. The reason for this is that the flow through the defueled regions is higher than that through the fueled regions, despite the fact that the defueled regions have fewer coolant holes. The reason for this is that the defueling elements have a very low heat load (zero-heat load was assumed in the calculations)., and the . attendant higher coolant densities more than offset the reduced number of coolant holes.

The maximum region temperature rise is dich-ted by the maximum region peaking factor. For the Cycle 4 extended operation power

. distribution used in this analysis,, Region 33 had the maximum region peaking _ factor (RPF=1.64), followed closely by Region 13 (RPF=1.63).

Region 33 is the twelfth region to be defueled in the sequence, and Region 13-is the nineteenth.

L 4-4

I The power distribution and defueling sequence explain the drop 'in maximum region temperature rise in Figure 4-1. Once Region 13 is removed, the highest power remaining region is Region 3 (RPF=1.44), the last region to be defueled. Since it has a much lower peaking factor, its ~ temperature rise is much less.

A separate series of P0KE calculations was performed to determine the core flow rate needed to maintain a maximum region temperature rise of 300'F for the condition where one region is empty. The results of this analysis are shown in Figure 4-2. As expected, the required flow increases as the defueling sequence proceeds. There is also a step increase when the first seven column region is defueled, and a step decrease when Region 13 is defueled, for the reasons given above.

7 Maintenance of these flow rates would result in a 50'F margin relative to the limits of LCO 4.1.9.

These low flow rates require relatively low power to the circulator and have been confirmed to be within the circulator performance envelopes for one Pelton-wheel driven circulator driven by condensate under refueling conditions.

In summary, these analyses show that temperatures during defueling will change slowly with time and will remain low. The limits of LCO 4.1.9 can be met, with margin, by maintaining flow rates that are within the capabilities of the circulators.

4.2 MECHANICAL PERFORMANCE {

The defueling elements are machined from H-091 graphite (HLM qu! valent). These elenents have a mean ultimate tensile strength of

- 1800 psi, which is approximately midway between that of H-327 and H-451 graphite, the defueling elements differ notably from the fuel elements

. v,hich they replace in that they have no fuel or fuel holes.

4-5 Y - __-

,- s The defueling elements will not receive any significant neutron -

irradiation in the core. Therefore, they' will not be' subjected to irradiation-induced contraction strains or to-irradiation-induced creep.

The stress produced 'within the defueling elements will result from differential thermal strains within the element and mechanical loads from

' dead weight and coolant flow,. or from seismic events.

There will . be no or very small temperature _ differentials across a graphite web between two coolant holes in a defueling element. The only significant temperature gradient will be across the defueling element.

Based on the requirements of technical specification LC0 4.1.9, the maximum cross-block temperature difference can be no higher than 3500F for a defueling element adjacent to a high power fuel element.

A simp.e bounding calculation for the thermal stress in a defueling element with this cross-block temperature difference was performed. Only ,

the ' edge of the defueling element in contact with the hot fuel element will be significantly heated.- The temperature gradient occurs mostly near this edge. For this type of temperature distribution, the part of the block in tension (the center of the element) will have a maximum temperature . difference less than one-half of the cross-block temperature difference. Conservatively using a value of 1750F produces an arial

' tensile stress of about 320 psi. This is well below the fsar value of 450 psi for an initial. core fuel element, and is less than 20% of the ultimate tensile strength of the defueling elements.

The axiel stress due to dead weight will not differ significantly from that in a standard fuel column since the difference in weight 15, sir,all . The strets due to coolant flow will be much less 6ue to the low flow rates during defueling. It should be noted that these forces are compressive and will actually result in a slight reduction in the maximum -

axial tensile stress.

4-6

As a result of the combination of no fuel holes and a greater radial ultimate tensile strength than H-327 graphite, the defueling elements have E higher seismic strength and lower seismic stresses than the initial core fuel elements. j i

The dowel / socket in the defueling element is stronger than that in a fuel element. This is due to the absence of fuel holes surrounding the dowel as well as the socket. The dowel / socket in the defueling element ,

is capable of withstanding a higher shear load than the dowel / socket of a '

fuel element.

Because the defueling elements will receive no significant irradiation in the core, there will be no irradiation-induced

, differential axial or radial strains. There will be only relatively small thermal bowing of the defueling elements.

The defueling elements will experience lower stress than the fuel elements they replace. Due to the absence of significant irradiation, they will also be more stable dimensionally. Consequently, structural integrity of the defueling elements is assured.

4-7

0

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• f :5. SAFETY ANALYSIS

5.1 INTRODUCTION

In this section, the safety analysis presented in Chapter XIV of:the Fort St. Vrains FSAR is reviewed to determine potential effects - of-defueling on accidents.and events discussed in the FSAR. In addition, the defueling activities are reviewed to determine if other accidents.or events need to.be analyzed. The purpose of..these reviews 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 during defueling and that no unreviewed safety questions are presented.

1 As a first step in this review process, Chapter XIV of the FSAR was examined'to< identify analyses potentially affected by defueling. The

-results of this' review are presented in Table 5-1. Ten FSAR accidents-

-(some of which envelope other less severe events) have been identified as requiring more detailed review for potential effects. These'are:

1..- Earthquake.

2. Reactivity accidents.

3. . Column deflection and misal'ignment.

4. Misplaced fuel element.

.5. Blocking of coolant channel.

6. Incidents involving the electrical system.

7. Loss of normal shutdown cooling (limiting case: cooldown on one firewater-driven circulator).

5-1

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

l

8. Leaks'inside the primary coolant system / steam generator leakage (moistureingress).

I

9. Fuel storage accidents. '

10. Permanent loss of forced circulation [ Design Basis Accident No.1 (DBA-1)].

1 In addition, five FSAR accidents have been identified as no longer credible under defueling conditions because depressurization accidents are not possible in an' atmospheric system. These are:

1. Failures involving helium purification system.

2. Primary coolant leakage

3. Maximum credible accident.

4. Maximum hypothetical accident.

5. Rapid depressurization/ blowdown [ Design Basis Accident No. 2 *

(DBA-2)].

As a second step in the review process, the defueling activities were reviewed to determine if any other accidents or events could be 4 identified that should be analyzed for safety effects. No such events were identified.

5.2 EVENTS REQUIRING FURTHER EVALUATION 5.2.1 Earthquakes As discussed in Section 4.2, the seismic strength of the defueling elements is greater than that of the fuel elements from the initial core.

The mass of the defueling elements is slightly less than that of the fuel elements they replace. Hence, the structural consequences for the core and the PCRV internals resulting from an earthquake as discussed in FSAR 5-2

1. 1 3

' Section 14.1.1 remain bounding. Effects of an earthquake on other plant structures discussed in Section 14.1.1 are unaffected. The fuel handling machine..is_ designed to maintain the primary coolant boundary during a design basis earthquake (DBE). - .

i

.The effects of 'an earthquake on the reactor core and internals during refueling were not evaluated as part of the licensing basis for Fort St. Vrain. -Considering that the plant was designed with the intention to conduct.30 refuelings over its lifetime, and the defueling process - is equivalent to only six refuelings, the likelihood or consequences of such an event are no greater during defueling than during refueling. However, an analysis was conducted to confirm that the consequences would be acceptable.

During the. defueling process, a region is, for a period of time, 3 empty or partially empty before the defueling elements are inserted. If an earthquake with sufficient intensity should_ occur while the region is empty lor partially empty, one or more of the adjacent fuel columns or defueling element columns could fall into the empty region.

An evaluation has been conducted of the structural stability of the reactor core during the design basis earthquake (DBE). It was concluded that the standard fuel columns around an empty region are not stable for certain - cases of column support bour.dary conditions, given that the ground accelerates during the DBE along particular directions relative to the orientation of the empty regior!. Analyses indicate that the control columns cannot tumble into the empty region because of binding of the control rod container and/or spline.

! The consequences of the impact of falling fuel elements on the core support block, core support posts, and adjacent fuel columns have been assessed. For all postulated modes of column collapse, bounding analyses indicate that

)

5-3

'1) a fallen- fuel element can in some cases be structurally -

. damaged. The effects on- fission product release and

~

criticality are discussed below.

2) .the bottom reflector block will show little sign of. structural damage after impact from a fallen fuel element. Even if damage occurred, the consequences would be insignificant.

3) the core support block will be damaged only if struck in the center, i.e., if the fallen element lands on the last central.

control _ fuel element prior to its removal. One or more webs between. the six large coolant channels in the core support block may crack at the top surface.

4) the core support posts will not be damaged.

No mechanical damage to the coated fuel particles would result from the , impact and subsequent breakage of a fallen fuel element. The temperatures that result from the loss of coolant flow through an element lying on its . side in the bottom of a previously empty region are very low. For a core flow sufficient to meet the region temperature rise requirements of technical specification LC0 4.1.9, the maximum fuel temperature is 709'F in that element. Even if a loss of forced cooling (lofc)' occurred following the DBE, the maximum fuel temperature would only reach 1860'F. -These temperatures are well below both the FSAR fuel failure temperature of 2900'F and ,the nominal maximum full power fuel temperature of 2372'F. Thus, these temperatures pose no concern for fuel particle integrity. Since the fuel particles remain intact, no fission products will be released as a result of a seismic event.

Fuel elements that fall into the cavity of a defueled region would '

I

-not be packed into the cavity as tightly as elements had been prior to s

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- the defueling. : Thus, core' reactivity.following a seismic event would be no greater than before.the. region was defueled.

Y The. spillage ' of small quantities of fuel rods from damaged fuel element (s) would not increase reactivity, since separation of fuel rods from the graphite block would result in a reactivity decrease. Even if a number of fuel rods ' fall through coolant holes onto the core support

-floor, criticality is not credible. Based on experiments with dropped

' fuel' element bodies conducted by 'GA in the early 1970s ' a damaged fuel .

element will lose no more than 20% of its fuel rods. Nearly all the fuel rods from the most reactive fuel element would need to be assembled -in the optiuum geometry, moderation, and reflection before criticality would be possible. Thus,'a seismic event will not result in a substantial loss of the shutdown margin, required to prevent a criticality hazard.

. . Hence, it is' concluded that the collapse of standard fuel columns into an empty fuel region during an earthquake will not result in either a ' fission product release 'or a criticality event. Fuel temperatures will remain well below-those at which coating degradation can occur. The structural damage is expected to be limited to the breaking of fuel elements.. There is a remote possibility that the core support block in the empty. region might crack at the top surface; however, the. block will continue to support : the fuel elements that have tumbled into the i

previously empty region, and the core support posts will not be damaged.

5.2.2 Reactivity Accidents Excessive Removal of Control Poison: As discussed in the FSAR Section 14. 2 .1.1., unintentional removal of control poison can be postulated to occur by several means. A " Rod Pair Withdrawal Accident

. During' Normal Operation at High Power" is not credible, since the core will be shutdown at defueling conditions. A " Rod Pair Withdrawal 5-5

r .

~

~

Accident During Startup Operations at Source Levels or at Very Low Power" is not credible, again because the core will be shutdown at defueling conditions. A " Rod Break Accident" as described in FSAR Section 14.2.1.1'

~1nvolves breakage of a partially inserted rod at power. - This event could add a small amount of reactivity to the critical core if the control rod

. segments dropped from a high worth to a low worth portion of the core (e.g.,fromthemiddletothebottom). Such an event during defueling is.

not credible, since under defueling shutdown conditions no rod pair will be in a position' where its breakage can allow the core to become critical. A " Multiple Rod ' Pair Withdrawal Accident" is considered incredible, as discussed in FSAR Section 14.2.1.1. In addition, during defueling all control' rod drives will be disabled as required by Interim Technical Specification LCO 3.1.6 except for those few control rod drives intentionally energized for shutdown margin verification testing and for y region defueling. A " Rod Pair Ejection Accident" is considered incredible for reasons outlined in FSAR Sections 3.5 and 14.2.2 and .

because at defueling ' conditions the PCRV is depressurized.

During defueling, the defueled region is subsequently filled with defueling elements that contain lumped poison pins with enough boron carbide poison to more than compensate for the control rods that were previously inserted in the fueled region. The baron loading of the lumped poison pins is such that the defueling of a region always-results in decreased core reactivity. Special QA/QC procedures, similar to these used during the loading of burnable poison rods into reload fuel elements, assure the proper boron loading of defueling elements. The startup channel count rates before a region is defueled and after it is defueled are compared to verify that the core reactivity has not increased as the result of defueling. In this way the adverse effects j resulting from a potential absence of boronation of a number of defueling )

elements would be detected long before the core approaches criticality -

)

i ar.9 .d be corrected if necessary. Thus, the improper loading or non-loading of lumoed poison pins presents no increased likelihood of a j criticality hazard relative to that expected during a normal refueling. )

l 5-6 i _ - _ _ _ _ _ _ - - _ - _ _ - - _ _ _ _ - - _ - - - _ _ --_----._------_--_-_.-__________________---.---_---J

+ .

.~ Loss of Fission Product Poisons: Under defueling conditions, after shutdown from power operation, the only credible mechanism by which fission product poisons can leave the core is radioactive decay.

~

The shutdown margin analyses presented in Section 3 conservatively take into account the effects of fission product decay on core reactivity. These analyses show that the core will have an adequate shutdown margin throughout defueling.

Rearrangement of Core Components, Including Fuel Loading Accidents and Earthquake Effects: The effects of the rearrangement of core components on reactivity have been examined. The analyses in FSAR Section 14.2.1.3 are also bounding under shutdown conditions for the geometry changes considered there. Other earthquake effects are

, discussed in Section 5.2.1.

Introduction of Steam into the Core: During defueling the amount of steam that can enter the core is less than that under normal operating conditions. At the low temperature and pressure conditions of defueling, helium saturates with a steam partial pressure of less than 5 psig. The total primary system volume is about 31,200 ft3 of whicn less than 4% is 1 within the core. A steam partial pressure of 5 psig in the active core  ;

volume at room temperature represents less than 20 lbs. of water. Steam l

will not react with the graphite or the lumped poison pins in the defuciing elenents under the low temperatures of defueling. Thus, the i lumped poison pins will remain intact in the core. As discussed in FSAR Section 11,.2.1.4, the net reactivit,y coefficient of water is 2.2 x 10-5  !

Ak/lb water. Thus, saturated conditions would produce a total reactivity increase of less than 0.0005 3k. Thir, increase is small relative to the k uncertainty in shutdown margin calculations of 0.01 6k and thus poses no criticality hazard. 1 I

.' )

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l Sudden Decrease in Reactor Temperature: FSAR Section 14.2.1.5 found -)

l that significant changes in core temperatures over short periods.of time can only occur in a subcritical system, after normal power operation, in

. which the coolant flow rate has been kept high. These conditions will not be possible_ during defueling. The bounding rate of reactivity addi- j tion in FSAR Section 14.2.1.5 will be much reduced at the lower core l

temperatures and flow rates of defueling. Thus, this accident will be much less severe during defueling than the conditions described in FSAR Section 14.2.1.5.~ Defueling operations are, therefore, bounded by the conditions described in the FSAR.

Rod' Withdrawal Accidents: As indicated in Table 5-1, rod withdrawal accidents-(RWAs) are discussed in Section 3.3. Since during defueling'it is not planned _to take the reactor critical, the only'way a RWA can occur ,

^

during defueling is if rods are accidentally withdrawn that insert enough reactivity into the core to overcome shutdown margin.- During defueling, all control rods will be inserted into the regions that still contain fuel, and deenergized so that they cannot be withdrawn. Only the control rods that are to be withdrawn for defueling or during shutdown margin verification testing will be reenergized at .any given time. The control roa pairs to be withdrawn for shutdown margin verification include those in the region being defueled, the next region to be defueled, and one or more rod pairs with reactivity- worth of 0.01Ak plus the temperature defect between the actual core temperature (about 200'F) and 80*F. As discussed in Section 3.3, up to three rod pairs hav? been identified which, if withdrawn early in the defueling sequence, could result in

' inadequate shutdown urgin. These rods will be totally disabled by physically disconnecting their rod drives from their power supplies.

These administrative safeguards will positively preclude inadvertent criticality.

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. . 5.2. 3 - Column Deflection and Misalignment.

The defueling- elements L have the. same external cimensions and tolerances as the fuel elements they replace. At ' the ' low core

~

. temperatures . during . defueling, . column deflection will therefore' be

bounded by the description in FSAR. Section 3.3.1.2 when fuel elements and Ldefueling elements both reside in the core.

5.2.4' -Misplaced Fuel Element During defueling, the fuel elements removed from the core are replaced with:defueling elements that contain no fuel. The defueling elements- have the same physical shape and dowel arrangement as the fuel

, elements Lthey -. replace. Extensive procedural controls, similar to those used in refueling, will be employed to maintain fuel element and

,, defueling element accountability. -The fuel handling machine is equipped with the capability. to41nspect remotely the serial number of each defueling element or fuel element present in the machine. Procedures do not allow mixing of fuel elements and defueling. elements together in the

, fuel handling -machine, eliminating the possibility of reinserting fuel into the core. In addition, all elements are accounted for by location, coordinate tracking, and serial number verif f cation prior to placement in

4. the core. These controls are adequate to casure that fuel elements arc always replaced by defueling elements. Thus, the misplacement c f a fue?

element back into the' core is a highly unlikely event.

l Even if a fuel elennt were reinserted in the core, it would not pose a criticality hazard. 1~he startup channel count rate is measured before a region is defueled and again after it is defueled to verify that the core reactivity has not increased as the result of the defueling

, step. Any significant deviation from the expected reactivity behavior would alert the operator to a problem that could be corrected long before 7

the added- reactivity exceeded the shutdown margin. Thus, a misplaced fuel element poses no increased safety hazard.

5-9

5.2.5 Blocking of Coolant Channel The effect of defueling upon the analysis in FSAR Section 3.6.5.2 has been evaluated. Low temperatures during defueling enhance the thermal conductivity of graphite relative to that at full power operation. More importantly, the decay heat generation rate is very low relatively to the heat generation rate at full power. The low heat generation rate, coupled with the enhanced capability of a graphite block to conduct heat away from the fuel near a blocked coolant channel to other coolant channels will result in significantly lower fuel temperatures. Thus, the analysis in FSAR Section 3.6.5.2 is bounding.

5.2.6 Incidents Involving the Electrical System Incidents involving the electrical system have no different effect during defueling than during refueling. Tha fuel handling machine fails "as is" upon loss of electrical power. The analyses in FSAR Sections 8.2.3.5, 8.2.5.2, 10.3.1, and 10.3.2 are bounding because they initiate from normal operating conditions, whereas initial conditions during defueling will include lower temperatures and much lower afterheat generation rates.

5.2.7 Losr, of Normal Shutdown Cooling, Permanent Loss of Forced Circulation The core thermal conditions resulting from Firewater Cooldown (Safe Shutdown cooling in the event of loss of normal shutdown cooling) and permanent loss of forced circulation (LOFC) are bounded by the analyses in FIAR Section 14.10. During defueling, the initial conditions of such events will be those resulting after a significant period of shutdown.

These include low core temperatures, low PCRV pressure, and a low rate of afterheat generation.

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l The material utilized in the lumped poison pins has the same basic characteristics as that utilized in the reserve shutdown system material I' in the central regions (Regions 1-19). As discussed in FSAR Appendix D.1.3.3.2, this material does not significantly redistribute even at the temperatures calculated for a permanent LOFC at full power. The use of the same basic material for the lumped poison pins as was used for the reserve shutdown material assures that the likelihood of a loss of poison is not increased due to defueling.

A confirming analysis was conducted with the P0KE and RECA codes to determine the maximum conditions in the core resulting from a permanent LOFC event occurring at 100 days following shutdown from 83.2% power operation. The P0KE code was used to calculate the initial core conditions. A maximum region coolant temperature rise prior to the LOFC of 350*F, as allowed by Technical Specification LC0 4.1.9, was assumed.

These conditions conservatively represent those anticipated at the start of the defueling sequence. The RECA transient analysis results indicate that, with one PCRV liner cooling loop operating with flow redistributed to the tcp head liner, maximum fuel temperatures, outer insulation cover plate temperatures, naximum liner temperatures, and maximum concrete temperatures remain within those experienced during normal operation. It was conservatively assumed in this analysis that liner cooling does not start until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the initial LOFC, To assere an intact primary coolant boundary, the control rod drive /

orifice assembly would need to be r,einstalled in any region from which it ,

had been removed at the time of an LOFC, since the helium tempers.ture in the upper plenum could eventually exceed the 400'F manuf acturer's l!mit for the inflatable silicen seal on the ref aeling isolation valve. The anclysis indicates that during defueling with liner cooling at least four days are available oefore this temperature limit would be exceeded. This 5-11

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1s much -lenger than would be available.if an LOFC were to occur during -

refueling, when_ decay _ heat levels would be higher. A control rod drive /

orifice assembly can typically be retrieved and reinstalled in only five- -

hours to reestablish it as' the primary coolant boundary.

. Analyses were also conducted of a case with no liner cooling in operation. These analyses confirm that, if no corrective actions are taken, excessive temperatures would occur in the ' fuel, the insulation covers,. the liner, and the concrete, although no limits are exceeded

- until five days following the LOFC. Hence, at the start of the defueling-sequence, liner cooling is required as specified in Technical

' Specification _LC0 4.2.15.

.5.2.8 Leaks' Inside the Primary Coolant System / Steam Generator Leakage <

(Moisture Ingress)

The potential sources of moisture (as water and/or steam) inleakage into the primary coolant system are listed in FSAR Section 14.5.2. The

' low pressure systems that were dismissed as potential sources of' moisture in'the FSAR must be included.here, because during defueling the primary coolant system is maintained slightly subatmospheric.

At the low temperature and pressure conditions of defueling, helium becomes saturated with steam at en steam partial pressure of less than 5 psig. Steam will not react with failed fuel under defueling conditions.

Thus, no fission product release is expected. No steam-graphite reaction will occur at these low pressures and temperatures. Thus, the primary coolant pressure can increase by no more than 5 psig due to the inleakage of moisture. As discussed in the basis of Technical Specification LC0  !

4.7.1, 5 psig is the maximum allowable working presst're of the fuel  !

handling equip nent. Since the refueling isolation valves are capable of maintaining an intact pressure boundary at 5 psig above atmospheric, no safety hazard is presented.

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!E 5.2.9 Fuel Storage Accidents'.

The introduction of.defueling elements has no effect on the analyses in FSAR Section.14.6.3.2. This conclusion assumes that in the event of loss of cooling to.one storage well, fuel could still be transferred to an empty. storage well,, as is feasible during a normal refueling operation, when only seven of the nine storage wells are used. During the defueling' operations at least one storage well will be.available as a backup, and the analyses in FSAR Section 14.6.3.2' remain bounding.

5.3 EVENTS NO LONGER CREDIBLE

'As indicated in Table 5-1, failures involving the helium

,: purification system, Primary . Coolant Leakage, the Maximum . Credible Accident, the Maximum Hypothetical Accident, and the Rapid

. Depressurization/ Blowdown (DBA-2) are not credible accidents during defueling. All involve depressurization from the high pressures expected under normal operating conditions. Since the PCRV will be maintained slightly subatmospheric during the defueling, it is not credible for it

-to depressurize.

If a leak did develop in the primary coolant boundary, thery might be a potential for air ingress. As r'iscussed in FSAR Section 14.7.2, the .;

. potential inleakage of air can be po:itively prevented by continuing the  !

supply of clean helium purge gas to the PCRV. During defuelirg, purge gas is supplied from the helium storage system. Even if some air leaked into the PCRV, the core temperatures during defueling will be much. lower than those analyzed in the FSAR. Thus, graphite oxidation would be less significant. Therefore, the consequences of air ingress dlscussed in FSAR Section 14.11.2.3, and found to be insignificant, are bounding.

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L. -. -. . - - - - - - - _ - _ _ _ _ _ ______

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5.4 CONCLUSION

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A review of Chapter XIV of the FSAR identified ten postulated -

accident conditions that required more detailed examination for potential impact from defueling. No requirements for additional analysis have been identified; the FSAR analysis is found to remain valid or bounding in all cases. 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 during defueling, and that defueling presents no unreviewed safety questions, as defined in 10CFR50.59.

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'. TABLE 5-1 POTENTIAL EFFECTS OF DEFUELING ON FSV FSAR ACCIDENT PREDICTIONS Potential Effects on Event Analysis FSAR Chapter XIV Event Due to Defueling 14.1 Environmental Disturbances

- Earthquake Evaluation required, see Sec. 5.2.1 of this document.

- Wind effects

- Floods

- Fire h None. Defueling does not affect

- Landslides I these events.

- Snow and ice 14.2 Reactivity Accidents and Transient Response

- Summary of reactivity sources Excessive removal of control Evaluation required, see Sec. 5.2.2 poison of this document.

Loss of fission product poisons Eveluation required, see Sec. 5.2.2 of this document.

Rearrangement of core components Evaluation required, see Sec. 5.2.2 of this document.

Introduction of steam into the Evaluation required, see Sec. 5.2.2 core of this document.

Sudden decrease in reactor Evaluation requit ed, see Sec. 5.2.2 temperature of this document.

- Rod withdrawal accidents Evaluation required. see Sec. 5.2.2 of this document.

14.3 Incidents

- Incidents involving the reactor core Colemn deflectior & misalignment Evaluation required, see Sec. 5.2.3 of tSis document.

Fuel element malfunctions None - Analysis in Section 3.4.3.1 of FSAR is bounding.

Misplaced fuel element Evaluation required, see Sec. 5.2.4 >

of this document.

Blocking of coolant channel Esaluation required, see Sec. 5.2.5 cf this document.

Control rod malfunctions No change from Section 3.8 of FSAR.

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4 TABLE 5-1 (Continued) -

Potential Effects on Event Analysis -*

FSAR Chapter XIV Event Due to Defueling Orifice malfunctions None - Analyses in Sections 3.6.5.1 and 3.9 of FSAR are bounding.

Core support floor ' loss of None.- Analysis in Section 3.3.2.2 cooling of FSAR is bounding.

- Incidents. involving the primary None - Analyses in Sections 4.2.2.1, coolant system. 4.2.2.3.6 and 4.2.2.3.7 of FSAR are bounding

- Incidents involving the control None - Analyses in Sections 6.4.2.and and instrumentation system 7.4.2 of FSAR are bounding.

- Incidents involving the PCRV L None. Defueling does not affect .'

-> Incidents: involving the secondary I these events.

. coolant & power conversion system ,

- Incidents involving the electrical Evaluation. required, see Sec. 5.2.6 system of this document.

-' Malfunctions of the helium purifi- None - Analysis in Section 9.4.6 of cation. system FSAR is bounding.

- Malfunction.of the helium storage <

system 1

- Malfunctions of the nitrogen system l

None. Defueling does not affect these events.

l 1

- Malfunctions involving handling of I heavy loads 14.4 Loss of Norinal Shutdown Evaluation required, see Sec. 5.2.7 Cooling of this document.

14.5 Secondary Coolant System ieakage )

l Steam leaks outside the primary None - Analysis in Section -

coolunt system 14.5.1 of FSAR is bounding.

Leaks inside the primary coolant Evaluation required, see Sec. 5.2.8 system / steam generator leakage of this document.

(moisture ingress) i 5-16 j

= _ _ - . _ _ - _ _

== + '. :

,' TABLE'5-1 (Continued)-

. Potential Effects on Event Analysis FSAR Chapter XIV Event- Due to Defueling 14.6 Auxiliary System Leakage

- Failures involving the. helium Events no longer credible.

~

purification system Accidents involving the gas waste None - Analysis.in Section system 14.6.2 of FSAR is bounding.

- Fuel.' handling & storage accidents Fuel handling accidents None - Analysis in Section 14.6.3.1-of FSAR is bounding.

Fuel storage accidents Evaluation required, see Sec. 5.2.9 y of this document.

Fuel shipping cask handling None - Analysis in Section 14.6.3.3

. accident' of FSAR is bounding.

14.7 Primary Coolant. Leakage Event no longer credible.

14.8. Maximum Credible Accident Event no longer credible.

14.9 Maximum Hypothetical Accident Event no longer credible.

14.10 Design Basis Accident No. 1, Evaluatfor required, see Sec. 5,2.7

" Permanent Loss of Forced cf this document.

Circulattoa (LOFC)"

14.11 Design Basis Accident No. 2, Event no longer credible.

l " Rapid Depressuriza-L tion /B!pwdown"

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6. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS No changes to the plant Technical Specifications or to the Interim Technical Specifications for reactivity control are necessary to proceed with defueling of the FSV core.

A change to Technical (specification LCO 4.4.1 and to Interim Technical Specification SR 4.1.4 may be required near the end of the defueling se-quence to allow waiver of shutdown margin experimental verification and to allow by-pass of the rod withdrawal prohibit system after the calculated k-effective of the remaining fuel, with all control rods withdrawn, is less then 0.95.

Any proposed modifications to the Technical Specifications to accommodate these considerations will be submitted to NRC on a schedule to be determined as defueling activities proceed. The proposed modifications i will be justified based upon the results presented in this report and upon actual experietice during the early part of the defuelirg sequence.

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7. REFERENCES fy; l 1. " Safety. Analysis Report for Cycle 4 Extended Operation," GA-C19661, May, 1989.

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