ML20246E562
| ML20246E562 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 08/24/1989 |
| From: | Crawford A PUBLIC SERVICE CO. OF COLORADO |
| To: | Weiss S NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| P-89327, NUDOCS 8908290194 | |
| Download: ML20246E562 (1) | |
Text
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Public Service' 20."l1...
P.O. Box 840 Denver CO 80201- 0840 August 24, 1989 h;,U*]((n',awford p
Fort St. Vrain Nuclear Operations Unit No. 1 P-89327 U. S. Nuclear. Regulatory Commission ATTN: Document Control Desk Washington,D. C.'00355 ATTN:
Seymour H. Weiss, Director Non-Power Reactor, Decommissioning and Environmental. Project Directorate Docket No. 50-267 SUBJECTi Extended Operation of Cycle 4
Dear Mr. Weiss:
PSC discovered that in letter P-89314 Extended Operation of Cycle 4, the Safety Analysis Report enclosed was of poor quality.
Therefore, enclosed is a new copy of the' letter P-89314 and Safety Analysis Report to replace the original transmittal.
If you have any further questions or concerns about this matter, please. cor; tact Mr. M. H. Holmes at (303) 480-6960.
Very truly yours, W $ e?
A. Clegg Crawford Vice President Nuclear Operations ACC/JMG/km Enclosure cc:
Regional Administrator, Region IV ATTN:
Mr. T. F. Westerman, Chief 8908290194 89082g
" " Section B g
PDR ADOCK 05000267 Tl" P
FDC w
- I Mr. Robert Farrell i
Senior Resident Inspector Fort St. Vrain
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Public Service" if711#a
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P.O. Box 840 Denver CO 80201- 0840 August 14, 1989-A. Clegg Crawford Fort St. Vrain Vice President Unit No. 1 Nuclear Operations P-89314 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 ATTN:
Mr. Seymour H. Weiss, Director Non-Power Reactor, Decommissioning and Environmental Project Directorate Docket No. 50-2f/
SUBJECT:
Extended Operation of Cycle 4
Dear Mr. Weiss:
Public Service Company of Colorado (PSC) will cease power operations at Fort St. Vrain no later than June 30, 1990.
The existing fuel loading, Cycle 4, will be used to' coast down in power until the end of power operation, resulting in no more than 520 effective full power days (EFPD). The exact number depends on the actual date that operations are terminated.
However, since Cycle 4 had previously been analyzed for only 300 EFPD, additional analysis was required to support the PSC coast (town plan.
PSC has completed the Cycle 4 Extended Operation Safety Analysis Report {SAR) which concludes that this extension does not present any
. unreviewed safety questions and no changes to the Technical Specifications are required.
Please find the enclosed Cycle 4 Extended Operation SAR for your information.
This SAR has been reviewed and approved by both the onsite and offsite safety review committees.
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4 P-89314 August 14, 1989 Should you have any questions en this matter, please contact Mr. M.
H. Holmes at (303) 480-6960.
Very truly yours, A.CleggCrawfork Vice President Nuclear Operations ACC/BGH/km Attachments
V, i
GA-C19661 I
SAFETY Af;nLYSIS REPORT FOR CYCLE 4 EXTENDED OPERATION l
FORT ST. VRAIN NUCLEAR GENERATING STATION GENERAL ATOMICS PROJECT 1900 MAY 1989 eum w
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. CONTENTS n
4 1.
INTRODUCTION.AND
SUMMARY
1-l' 2.
REACTOR OPERATING HISTORY 2-1:
3.
NUCLEAR PERFORMANCE
................... 1 3.1 Heavy Metal and Burnable Poison Loadings
'3-2' 3.2.
Control Rod Sequence 2 3.3 Projected Cycle 4 Extended Operation 3-4 3.4-Maximum Control Rod Worth........-.....
3-5 3.5 Core Shutdown Margin 3-7
.3. 6 '
Ki neti cs ' Parameters................. '
3 3.7 Nuclear Detector Decalibration..........
3-11 4.
THERMAL-HYDRAULIC AND MECHANICAL PERFORMANCE.......
4-1 5.
SAFETY ANALYSIS:
5-1 5.1
- Introduction 5-1 5.2 Loss of Normal' Shutdown Cooling, Permanent Loss of Forced Circulation, and Rapid Depressurization/ Blowdown.............
5-3 5.3 Concl us i ons.....................
5-3 6.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS....
6-1~
7.
SURVEILLANCE TESTS....................
7-1 8.
REFERENCES........................
8-1 i:i 1
-____ __-. _____ - _-_ - - a
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1.
INTRODUCTION AND
SUMMARY
l l
.This. Safety Analysis Report' (SAR) is prepared to assess extended operation of the Fort St. Vrain Nuclear Generating Station (FSV) in the'.
i<
current reload cycle (Cycle 4).
Operation of the plant during Cycle 4 up -
to a total of 300 effective full power days (EFPD) has been evaluated in Ref.
1.
Public Service Company of Colorado (PSC) has decided to cease nuclear operations at FSV no later than June 30, 1990. Cycle 4 will be the last reload cycle at FSV, and extension of the cycle beyond 300 EFPD may be L
necessary while PSC prepares for permanent shutdown, defueling, and decommissioning.
The total duration of Cycle 4 has been analyzed up to 520 EFPD.
This report contains sections describing the operating history of the reactor through December 31,
- 1988, evaluations of nuclear, thermal-hydraulic, and mechanical performance of the core, and the safety aspects of the core during extended operation of Cycle 4 up to 520 EFPD.
A safety evaluation for extended operation is presented in this report.. It is concluded that extended operation of Cycle 4 up to 520 EFPD presents no unreviewed safety-questions, as defined in 10CFR50.59 and requires no changes to the Technical Specifications.
1-1 4
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?
)
2.
REACTOR 0PERATING HISTOM E
i Initial criticality of the FSV reactor was achieved on January 31, 1974, with initial generation of electricity on December 11, 1976.
Prior to. February 1,1979, when the plant was shutdown for refueling, the initial core had operated a total of 174 EFPD.
Cycle 2 operation began on May 26, 1979 and was completed on May 13,1981, - having accumulated a total of 189 EFPD.
Cycle 3 operation began on July 15, 1981, and was completed on
' January 20, 1984, having ' accumulated a total of 294.5 EFPD.
Cycle 4 operation began on May 16,
- 1984, and as of December 31, 1988 had accumulated a total of about 154.5 EFPD.
The nuclear performance ' of the FSV core has been, in general, _ as predicted.
Good agreement 'between measurements and calculations has been obtained for shutdown margins, temperature coefficient, xenon worth, and control rod worth (1 '. e., measurements are well within the acceptance criteria specified for the tests).
Initial cold criticalities-in Cycles 1 through.4 were predicted within 0.003 Ak.
Analyses have overpredicted the
~
end-of-cycle (EOC) reactivities of the core at operating temperatures by a
'few tenths. of a percent; however, the difference between observed and expected reactivity has remained within the 0.01 Ak limit of Technical Specification LC0 4.1.8.throughout operation.
i Fission product release to date has been very low.
Measured circulating activity has been approximately a factor of 30 less than the limit provided in Technical Specification LCO 4.2.8.
Measurements of plateout activity obtained after removal of the first plateout probe in November 1981 indicate that these activity levels are also substantially below Technical Specification limits (Ref. 2).
l j
2-1 i
i
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/
On October ' 5,1982 NRC issued Amendment No. 28-to the FSV Operatint).
License.
In this amendment, the NRC concluded, based upon a review of Ref.
4, that the fluctuation issue is resolved.
The technical specifications proposed in Ref. 5 were. incorporated in the operating license, and all previously imposed restrictions on reactor power level were removed..
As a result of. the visual examinations conducted on fuel elements removed from the reactor during the second refueling, two Segment 2 fuel elements were each found to have one or two cracked graphite webs.
The-presence of these cracks did not affect the cooling geometry of the fuel or the ability of the fuel. handling machine to safely. remove the fuel elements from the core.. A-D0E-funded program was carried out at-GeMral Atomics
.(GA) to investigate this. issue,- while a similar NRC-funded program was conducted at Los Alamoc Laboratory.
Results of the cracked web program were submitted to NRC by PSC in Refs. 6 through 8, which showed that the cracks were. caused by high localized in-plane tensile stresses resulting from high. fluence and large gap coolant flows, and that the cracked webs have no effect on the ability of the elements to perform their functions safely.
The NRC concurred with these results and closed the issue of fuel element web cracking on December 30, 1986 (Ref. 9).
The examination of Segment 3 fuel elements, removed during the third refueling, indicated no cracked graphite webs.
The replacement fuel elements for Cycle 4 featured one design change (use of H-451 graphite) relative to the fuel design described in the FSAR.
This design change was the subject of a lengthy generic review and approval by NRC in 1978 and 1979.
The safety evaluation for the change as it i
affected Cycle 4 core reload was presented in Ref.1, and the NRC approved the use of H-451 graphite in FSV via Amendment No. 40 to the FSV Operating License.
A significant operating. event occurred on June 23, 1984.
As reported in Ref. 10, following an automatic scram action, six control rod pairs failed to automatically insert and had to be driven into the core.
Investigations determined that shim motor bearing wear and debris buildup 2-3
l.
, i,-
3.
NUCLEAR PERFORMANCE In this.section the effects of extended operation of Cycle 4 up to a total' of 520 EFPD on the nuclear performance. of the core are. presented.
Nuclear analyses 'were carried out using the same methods applied to the analyses presented in the FSAR, previous reload SARs, ' and the ' semiannual fuel accountability reports.
Extended operation of Cycle 4 introduces'no new aspects to high-temperature gas-cooled reactor (HTGR) core design or analysis techniques; consequently, there was no need to develop or adapt any new' methods or procedures for the nuclear performance analysis.
The depletion analyses described in this chapter were performed ' by simulating the actual core power history for Cycles 1, 2, and 3 and the first 154.5 EFPD of Cycle 4.
Continuous operation
.t 100% power for the balance of Cycle 4 was assumed.
Cycle 4 was assumed to continue to a total length of 520 EFPD.
During extended operation of Cycle 4, all control rod groups 6.
except the regulating rod, which remains partially withdrawn from the central region, will gradually be fully withdrawn.
Therefore, the only means of introducing positive reactivity in compensation for the reactivity decrease due to the burnup of fuel will be to decrease the fuel temperature and the reactor power as shown in Table 3-1.
Although rated power cannot be maintained throughout extended operation, it was conservatively assumed in the depletion analyses that operation from 154.5 EFPD to 520 EFPD would take place at 100% of rated power.
3-1 C__-__-_-_________-_--___-_-
r.
6 Cycle 4 is given in Table 3-4.
The identification of the control rod l
groups is shown in Figure 3-1.
Extended operation of Cycle 4 up to an l
additional 220 EFPD will require no changes to the control rod sequence.
The regulating rod is located in the central refueling region (rod group 1).
This group is partially withdrawn before criticality is achieved and then maintained in its most reactive control rod position for the remainder of the operation.
In this manner, minor reactivity adjustments can be made most rapidly with the minimum amount of control rod motion.
This is consistent with the method of operation utilized for the control rods in previous cycles, including Cycle 4.
A summary of the calcJlated power peaking factors obtained during a typical rise-to-power at the nominal E0C4 using the control rod sequence is given in Table 3-5.
It was conservatively assumed that the reactor was shutdown for 90 days prior to the startup.
The results in Table 3-5 show the correlation between control rod group insertion and the core power level.
Power peaking factors during Cycle 4 extended depletion are discussed in Section 3.3.
The basis of Technical Specification LCO 4.1.3 contains maximum values of region peaking factors.
At power less than 20% (core outlet gas temperature {950*F), the maximum RPF is 3.0 and the maximum tilt is 1.61.
At powers between 20 and 60% (cutlet temperature from 950'F to 1250'F) the applicable limits are an RPF of 2.15 and a tilt of 1.34 to 1.46.
At higher powers, above 60% (outlet temperature >1250'F), the limits are an RPF of 1.83 and a tilt of 1.34 to 1.46.
From the data given in Table 3-5, it can be seen that at the nominal EOC4 the calculated power peaking fectors for the various power levels do not exceed those given in the basis of Technical Specification LCO 4.1.3.
This is true for both the radial region peaking factors and the intra-region peaking (column tilt) factors.
The analyses indicate that at powers 5
between 0 and 2%, the maximum RPF may slightly exceed the LC0 4.1.3 basis 1
i 3-3 1
that the limits on peaking factor in the lower. fuel layer assumed in the basis of LC0 4.1.3 are not exceeded.
These calculations were carried out n
to 520 EFPD.
The' basis of Technical Specification LC0 4.1.3 also states that an acceptable flux distribution shall be maintained at lower power' levels by keeping the flux level in the center of the core at least as high as the average level.
Table 3-7 shows the ratio of the flux in the inner core regions (Regions 1 through 19) to the core average flux for control rod configurations in Table 3-5 which can result in operation between 0% and about 20% power at the nominal E0C4.
The flux ratio is above 1.0 for all cases, consistent with the basis of LCO 4.1.3.
With forther burnup during extended-operation, control rods will be removed from the core until all control rods except the regulating rod-in the central refueling region will be removed from the core.. The flux ratio with this control rod configura-tion was calculated to be between 1.12 and 1.13 throughout extended operation.
3.4 MAXIMUM CONTROL R0D WORTH The basis of Technical Specification LCO 4.1.3 states that the accidental removal of the maximum worth single rod pair shall result in a
-transient'with consequences no more severe than the withdrawal of 0.012 AK, at rated (i.e.,
100%) power, from a core which has a temperature defect between 220*F and 1500*F of 0.028 AK.
In addition, the calculated worth of any rod _ pair in any configuration with the reactor critical must be less than 0.047 AK.
The same requirements are contained in Interim Technical j
Specification LC0 3.1.5.
The rod withdrawal accident (RWA) at full power
]
evaluated in Section 14.2 of the FSAR assumes withdrawal of a control rod f
worth of 0.012 AK at equilibrium EOC with an equilibrium E0C temperature defect of 0.028 AK.
Because the consequences of an RWA are a function of
. rod worth, steady-state core temperature (i.e., initial power level), and temperature coefficient (which varies with burnup during the cycle), it is i
3-5
_ - - _ _ _ _ _ _ _ _ _ _ _ _ = _ - _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ _
/
control rod in that group is less than 0.008 AK.
As is shown in Section 3.6, the temperature defect between 220'F and 1500*F is larger than 0.028
' AK at 300 EFPD.
Therefore, the consequences of an RWA at the nominal EOC4 at ' rated power are less than those of the RWA described in Technical Specifications.
Table 3-8 indicctes that with two rod groups fully inserted, a maximum control rod worth of 0.014 AK is obtained.
However, Table 3-5 indicates that the maximum power level that can be achieved in this configuration is from 40% to 60% of rated power.
The difference in average fuel temperature between these conditions and operation at rated power make an additional 0.006 AK in temperature defect available to mitigate the consequences of an RWA. Therefore, the consequences of an RWA of an 0.014 AK rod at 60% power
, at 300 EFPD are the same as those of an 0.008 AK rod at 100% power, which i
are in turn, as discussed above,. less than those of the RWA. described in the Technical Specifications.
i Similar evaluations of the RWA for all configurations shown in Table 3-0 indicate that in all cases the RWA consequences at nominal EOC are
. bounded by those of the RWA described in the Technical Specifications.
As core operation is extended to 520 EFPD, the power level associated with the maximum control rod worth decreases, thereby providing more temperature defect available to mitigate the consequences of an RWA.
The total l
temperature defect, as discussed in Section 3.6 remains above 0.028 AK.
It is, therefore, concluded that the consequences of an RWA during Cycle 4 extended operation are bounded by those of the RWA discussed in -Section l
14.2 of the FSAR and in the Technical Specifications, j
L 3.5 CORE SHUTDOWN MARGINS ll 3.5.1 Control Rod System SDM I
Interim Technical Specifications LCO 3.1.4/SR 4.1.4 state that a shutdown margin (SDM) of 0.01 AK shall be achieved under the following conditions:
3-7 4
____m_.____.
- - - - ^
g 9.
1 i
J results in Table 3-9 are consistent with the results calculated for the j
Segment 9-Cycle 4 SAR (see Table 5-10 of Ref. 1), which show that the minimum SDM occurs at the middle of Cycle 4, and then start's to increase systematically to the nominal EOC4.
On the basis of the above presentation, it may be concluded that the control rod system is adequate to provide an adequate SDM for the FSV core under all normal and postulated accident conditions during extended operation of Cycle 4.
General requirements for operability of control rod l
drives during the cycle are provided in Interim Technical Specification LCO 3.1.1.
3.5.2 Reserve Shutdown System SDM As noted above, Interim Technical Specifications LCO 3.1.4/SR 4.1.4 state that the reactor SDM shall be greater than or equal to 0.01 AK.
Specifically, for reactivity control with the reserve shutdown system (RSS)
Interim Technical Specifications LCO 3.1.8/SR 4.1.8 state that with any one RSS unit inoperable core operation may continue provided that the unit is capable of being made operable within 14 days following a reactor shutdown.
Furthermore, in the basis of Interim LCO 3.1.8/SR 4.1.8 it is stated that the RSS must be capable of achieving reactor shutdown in the event that the control rod system fails to insert.
On the basis of above considerations, the SDM calculations for the RSS were carried out with the following conservative assumptions:
1.
The core, prior to shutdown, was operated at rated power long enough to equilibrate xenon and Pa-233.
2.
The scram signal fails to insert any withdrawn control rods.
3.
The inoperable RSS unit is the maximum worth one.
4.
The core is at room temperature (80*F).
3-9
i
' 0.034 AK.;at 520 EFPD.
Both of these calculated values meet the requirements of Interim LC0 3.1.7 for measured temperature defect.
,3.7 NUCLEAR DETECTOR DECALIBRATION The power range nuclear detector signals, used by the control system to -initiate. plant protective system (PPS) action, exhibit significant decalibration due to control rod motion.
This detector decalibration is accommodated: by a reduction in the fixed PPS setpoints of Technical.
Specifications SL 3.3 and LCO 4.4.1 and. frequent recalibration of - the detectors.
The. reduced setpoints have been reevaluated each cycle because of the different control rod withdrawal sequence and the different fuel loading distribution.
A rigorous analysis was done to determine these reduced setpoints for Cycle 1.
For subsequent cycles including Cycle 4 the reduced setpoints'have been determined as follows:
Calculate the " worst-case" detector decalibration factor (DF) for o
each control rod group (i.e., the case which would most delay the-PPS trip).
o If the " worst-case" DFs -indicate less delay in the PPS trip than the previous cycle, use the reduced setpoints from the previous
- cycle, o
If these " worst-case" DFs indicate more delay in the PPS trip than the previous cycle, then the reduced setpoints and/or the detector recalibration schedule for the previous cycle must be reevaluated.
The control rod sequence for the extended operation of Cycle 4 is the same as that for the reference cycle.
Furthermore, the extended operation of Cycle 4 is characterized by a deficiency of excess reactivity, i.e., all shim banks are expected, except for a few special cases discussed in.Section 3.4, 3-11 l
= - _ - -
x Public Service' 2 ";.**a b.o.
P.O. Box 840
. Denver CO 80201- 0840 August 24, 1989
$;,U*98e[n',#d p
Fort St. Vrain Nuclear Operations Unit No. 1 P-89327-U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington,0. C. 20555' ATTN: Seymour H. Weiss, Director Non-Power Reactor, Decommissioning and Environmental Project Directorate Docket No. 50-267 1
SUBJECT:
Extended Operation of Cycle 4
Dear Mr. Weiss:
PSC discovered that in letter f-89314, Extended Operation of Cycle 4, the Safety / nalysis Report enc'osed was of poor quality.
Therefore.
enclosed is. a new copy of the letter P-89314 and Safety Analysis Leport to replace the original transmittal.
If you have any further questions or concerns about this matter, please contact Mr. M. H. Holmes at (303) 480-6960.
Very truly yours, W $bn f
A. Clegg Crawford Vice President Nuclear Operations i
ACC/JMG/km i
' Enclosure cc:
Regional Administrator, Region IV ATTN:
Mr. T. F. Westerman, Chief
)
8908290194 89082g Section B
^
PDC 4
I I Mr. Robert Farrell j
Senior Resident Inspector
.I Fort St. Vrain E __ _ ___
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Public Service
'Public Serv
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P.O. Box M3 Denver CO 80201-0840
. August 14, 1989 A. Clegg Crawford Fort St. Vrain vic presie n, Unit No. 1 Nuclear Operations P-89314 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 ATTN: Mr. Seymour H. Weiss, Director Non-Power Reactor, Decommissioning and Environmental Project Directorate Docket'No. 50-267
SUBJECT:
Extended Operation of Cycle 4
Dear Mr. Weiss:
i Public Service Company of Colorado (PSC) will cease power operations at Fort St. Vrain no later than June 30, 1990.
The existing fuel loading, Cycle 4, will be used.to coast down in power until the end of power operation, resulting in no more than 520 effective full power days (EFPD). The exact number depends on the actual date that operations are terminated.
However, since Cycle 4 had previously been analyzed for only 300 EFPD, additional-analysis was required to support-the PSC coast down plan.
PSC has completed the Cycle 4 Extended Operation Safety Analysis Report (SAR) which concludes that this extension does not present any
. unreviewed safety questions and no chahges to the Technical Specifications are required.
Please find the enclosed Cycle 4 Extended Operation SAR for your information.
This SAR has been reviewed and approved by both the onsite and offsite safety review committees.
I L
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P-89314 August 14, 1989 Should you have any questions on this matter, please contact Mr. M.
H. Holmes at (303) 480-6960.
Very truly yours, T^
A.CleggCrawfork Vice President Nuclear Operations ACC/BGH/km Attachments
s, GA-C19661 I
i 1
I SAFETY ANALYSIS REPORT FOR CYCLE 4 EXTENDED OPERATION FORT ST. VRAIN NUCLEAR GENERATING STATION 2
GENERAL ATOMICS PROJECT 1900 MAY 1989 i
eumm
s, 4
4 CONTENTS 1.
INTRODUCTION AND
SUMMARY
1-1
- 2.
REACTOR OPERATING HISTORY 2-1 3.
NUCLEAR PERFORMANCE 3-1 3.1 Heavy Metal and Burnable Poison Loadings 3-2 3.2 control Rod Sequence 3-2 3.3 Projected Cyr s 4 Extended Operation 3-4 3.4 Maximum Control Rod Worth.............
3-5 3.5 Core Shutdown Margin 3-7 3.6 Ki neti cs Paramete rs................
3-10 3.7 Nuclear Detector Decalibration 3-11 4.
THERMAL-HYDRAULIC AND MECHANICAL PERFORMANCE.......
4-1 5.
SAFETY ANALYSIS 5-1 5.1 Introduction 5-1 5.2 Loss of Normal Shutdown Cooling, Permanent Loss of Forced Circulation, and Rapid Depressurization/ Blowdown.............
5-3 5.3 Conclusions....................
5-3 j
6.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS....
6-1 7.
SURVEILLANCE TESTS....................
7-1 8.
REFERENCES........................
8-1 iii i
k
)
,e FIGURES 3-1 Identification of Control Rod Groups.............
3-26 3-2 Tilt Envelope for Extended Cycle 4..............
3-27 3-3 Temperature Defect Vs. Average Core Temperature 3-29 l
TABLES 3-1 Projected Power History for Cycle 4 Extended Operation.....
3-14 3-2 Projected Core Loadingt at the Nominal End of Cycle 4 3-15 3-3 Projected Core Loadings at the End of Extended Cycle 4....
3-16 3-4 Control Rod Sequence for Cycle 4...............
3-17 3-5 Typical Rise-to-Power at Nominal EOC 4 (300 EFPD) 3-18 3-6 Summary of Control Rod Insertions and Axial Power Factors in Bottom Fuel Layer.....................
3-19 3-7 Flux Ratios at Lower Power Cycle 4 @ 300 EFPD 3-20 3-8 Worth of Control Rod Groups and Maximum Rod at Nominal EOC 4.
3-21 3-9 Control Rod Shutdown Margins (AK) in the Extended Cycle 4 3-22 3-10 RSS Shutdown Margins (AK) in the Extended Cycle 4 3-23 3-11 Kinetics Parameters 3-24 3-12 Worst Detector Decalibration Factors.............
3-25 4-1 Extended Cycle 4 Calculated Peak Conditions Versus FSAR Initial Core Peak Values...................
4-6 5-1 Potential Effects of Extension of Cycle 4 on FSV FSAR j
Accident Predictions.....................
5-5 i
4 I
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iv
7.
g,
(
- 1..
INTRODUCTION AND
SUMMARY
Ii This - Safety. Analysis Report (SAR) is prepared.to assess extended operation of the' Fort St. Vrain Nuclear Generating Station (FSV) in the current reload cycle (Cycle 4).
Operation of the plant during Cycle 4 up j
to a total of 300- effective full power days (EFPD) has been evaluated in
'Ref.
1.
Public Service Company of Colorado (PSC) has decided to cease nuclear operations at FSV no later than June 30, 1990.
Cycle 4 will be the last reloa; cycle at FSV, and extension of the cycle:beyond 300 EFPD may be necessary while PSC prepares for permanent shutdown, defueling,.and decommissioning. The total duration of Cycle 4 has been analyzed up to 520 EFPD.
l
.This report contains sections describing the operating history of the raactor through December 31, 1988,. evaluations of nuclear, thermal-hydraulic, and mechanical performance of the core, and the safety aspects of the core during extended operation of Cycle 4 up to 520 EFPD.
A safety evaluation for extended operation is presented in this report.
It is concluded that extended operation of Cycle 4 up to 520 EFPD presents no unreviewed safety questions, as defined in 10CFR50.59 and l
requires no changes tc the Technical Specifications.
1-1
l n,
b
- 2..
REACTOR OPERATING HISTORY 1
l l
Initial - criticality of the FSV reactor was achieved on January 31, 1974, with initial generation of electricity on December 11, 1976.
Prior j
.to February 1,1979, when the plant was shutdown for refueling, the initial core had operated a total of 174 EFPD.
Cycle 2 operation began on May 26, 1979 and was completed.on May 13, 1981, having accumulated a total of.189 EFPD.
Cycle 3 - operation - began on July 15, 1981, and was completed on January 20, 1984, having accumulated a total of 294.5 EFPD.
Cycle 4 operation began on May 16,
- 1984, and as of December 31, 1988 had accumulated a total of about 154.5 EFPD.
The. nuclear performance of the FSV ccre has been, in general, as predicted.
Good agreement between measurements and calculations has been obtained for shutdown margins, temperature coefficient, xenon worth, and control rod worth (i.e., measurements are. well within the acceptance-criteria specified for the tests).
Initial cold criticalities in Cycles 1 through 4 were predicted within 0.003 Ak.
Analyses' have overpredicted the end-of-cycle (EOC) reactivities of the core at operating temperatures by a few tenths of a percent; however, the difference between observed-and expected reactivity has remained within the 0.01 Ak limit of Technical Specification LCO 4.1.8 throughout operation.
Fission product release to date has been very low.
Measured circulating activity has been approximately a factor of 30 less than the limit provided in Technical Specification LCO 4.2.8.
Measurements of
.l plateout activity obtained after removal of the first plateout probe in November 1981 indicate that these activity levels are also substantially below Technical Specification limits (Ref. 2).
2-1
g The most unusual occurrence, 'to date, was the detection, initially in 50ctober1977,oftemperaturefluctuations. These fluctuations affected the H
nuclear channels, the region. exit temperatures, and the steam generator module temperatures.
During fluctuations, however, the total core coolant flow and core thermal power remained essentially constant.
In addition, the temperature swings during fluctuations. stayed within plant operations and technical specification limits.
A comprehensive program to evaluate and resolve the fluctuation issue l
was begun in late 1977.
This program led to installation, in November 1979, of core region constraint devices (RCDs) (Ref. 3).
These devices limit the small (approximately 0.10 in.) lateral movements of fuel columns to which the fluctuations were attributed. Fluctuation testing of the core up to 100% power with RCDs installed was completed in November 1981. No-fluctuations have been detected since installation of the RCDs.
The results of these tests were formally submitted to NRC in. July ' 1982 (Ref.
4).
A second major issue with regard to reactor operation has been the-existence of discrepancies between measured and calculated region outlet helium temperatures.
The largest discrepancies have been limited to regions in-the northwest boundary of the ' core (Regions 20 and 32-37), with measured temperatures being consistently :less than calculated temperatures.
These discrepancies are caused by a transverse flow of relatively cool helium from the core-reflector interface along the inside of the ' region outlet thermocouple sleeves (Type II flow).
This flow passes over the region outlet thermocouple assemblies of these regions and depresses the indicated region outlet temperature.
To compensate for these discrepancies, special operating procedures were provided which insure compl'iance with the original core design intent.
Technical Specification LC0 4.1.7/SR 5.1.7 governs operation with these measurement errors.
2-2 1
s.
+
On October 5,1982 NRC issued Amendment No. 28 to the FSV Operating License.
In this amendment, the NRC concluded, based upon a review of Ref.
4, that the. fluctuation issue is resolved.
The technical specifications proposed ' in Ref. 5 were incorporated in the operating license, and all
[
previously Imposed restrictions on reactor power level were removed.
As a result of the visual examinations conducted on fuel-elements removed from the reactor during the second refueling, two Segment 2 fuel elements were each found to have one or two cracked graphite webs.
The presence of.these cracks did not affect the cooling geometry of the fuel or the ability of the fuel handling machine to safely remove the fuel elements from the core.
A DOE-funded program was carried out at General Atomics (GA) to investigate this issue, while a similar NRC-funded program was conducted at Los Alamos Laboratory.
Results of the cracked web program l
were submitted to NRC by PSC in Refs. 6 through 8, which showed that the cracks were caused by high localized in-plane tensile stresses resulting from high fluence and large gap coolant flows, and that the cracked webs have no effect on the ability of the elements to perform their functions safely.
The NRC concurred with these results and closed the issue of fuel element web cracking on December 30, 1986 (Ref. 9).
The examination of Segment 3 fuel elements, removed during the third refueling, indicated no cracked graphite webs.
The replacement fuel elements for Cycle 4 featured one design change (use of H-451 graphite) relative to the fuel design described in the FSAR.
This design change was the subject of a lengthy generic review and approval 1
by NRC in 1978 and 1979.
The safety evaluation for the change as it l
affected Cycle 4 core reload was presented in Ref. 1, and the NRC approved I
the use of H-451 graphite in FSV via Amendment No. 40 to the FSV Operating License.
1 A significant operating event occurred on June 23, 1984.
As reported j
in Ref. 10, following an automatic scram action, six control rod pairs failed to automatically insert and had to be driven into the core.
Investigations determined that shim motor bearing wear and debris buildup 2-3 I
k l
w were, the primary contributors. - Corrective action included a comprehensive control rcd drive (CRO)' refurbishment program'and implementation of interim -
Technical Specifications for reactivity contrcl.. In Ref.11, PSC committed to operate FSV in accordance - with procedures based upon ' the interim Technical. Specifications until formal _ specifications are approved ~ and implemented.
Cycle 4 operation has been limited since it began in May 1984.
The plant was shutdown for most of 1984 1985,1986, and 1987 for engineering modifications to the control rod drives, environmental qualification (EQ) of. safety-related electrical equipment,_ helium circulator removal and replacement, and for recovery from a fire in the turbine building.
The
~ plant was shutdown throughout the second half of 1988 for' circulator refurbishment.and removal of. moisture from the reactot vessel.
Based on economic consideraisons associated with the ongoing operating
. costs of FSV, PSC has decided to cease nuclear operations on or before June 30, 1990 (Ref. 12).
The final duration of' Cycle 4 will be based upon the time to complete planning and preparations for the initiation of defueling.
and upon the reliability of FSV.
l l
l 2-4 L=______________
- j L
3.
. NUCLEAR PERFORMANCE
)
In this section the effects of extended operatica of Cycle 4 up to a i
total of' 520 EFPD on the nuclear performance of the core are presented.
Nuclear analyses were carried out using the same methods applied to -the.
analyses presented in the FEAR, previous reload SARs, and the semiannual fuel accountability reports.
Extended operation of Cycle 4 introduces no new aspects to high-temperature gas-cooled reactor (HTGR) core design or analysis' techniques; consequently, there was no need to develop or adapt any new methods or procedures for the nuclear performance analysis.
The depletion analyses described in this chapter were performed by simulating the actual core power history for Cycles 1,. 2, and 3 and the first 154.5 EFPD of Cycle 4.
Continuous operation at 100% power for the balance of Cycle '4 was assumed. Cycle 4 was assumed'to continue to a total ler.gth of 520 EFPD.
During the extended operation of Cycle 4, all control rod groups except the regulating rod, which remains partially withdrawn from the central region, will gradually be fully' withdrawn.
Therefore, the only means of introducing positive reactivity in compensation for the reactivity decrease due to the burnup of fuel will be to decrease the fuel temperature and the reactor power as shown in Table 3-1.
Although rated power cannot be maintained throughout extended operation, it was conservatively assumed in the depletion analyses that operation from 154.5 EFPD to 520 EFPD would take place at 100% of rated power.
3-1
w J
' 3.1 HEAVY METAL-AND BURNABLE POISON LOADINGS It is' anticipated that' the reactor will have operated for up to -957.5 EFPD, generating a total of up to 1.9 x 107 MWhr of energy at the nominal end of Cycle 4 (300 EFPD).
The projected heavy metal loadings in the core segments at. the E0C 4 (957.5 EFPD) are given in Table 3-2.
The maximum burnup in fissile particles is projected to be about' 16.0% FIMA and in fertile particles about-3.4% FIMA.
These burnups are lower than the limiting values.given in the FSAR, Appendix A, Table A.2-2.
The maximum projected fast flux (E > 0.18' MeV) exposure in Segment 4 is about 4.2 x 1021 nyt. This exposure is also lower than the limiting values given in-the FSAR.
I It is anticipated that at the end of Cycle 4 extended operation, the reactor will have operated for up to 1177.5 EFPD, generating up to 2.34 x
. 107 MWhr of energy.
The maximum fuel element irradiation allowed by Technical Specification LCO 4.1.1 1:,1800 EFPD.
The projected heavy metal loadings.in the core segments it the extended EOC 4 (1177.5 EFPD) are given in Table 3-3.
The maximum burnup in fissile particles is projected'to be about 17.4% FIMA, and in fertila particles about 4 4% FIMA.
These burnups are also 1ower than the limiting clues given la the FSAR.
The maximum projected fast flux exposure in the core at 117'/.5 EFPD is about 5 x 1021
- 1 nyt, also lower than the limiting values given in the FSAR.
The burnable poison loaded into Segment 9 fuel elements at B0C 4 will be essentially totally depleted by the nominal EOC 4.
No burnable poison adjustments will be required for Cycle 4 extended operation.
j i
3.2 CONTROL ROD SEQUENCE Technical Specification LCO 4.1.3 states that a control rod sequence will be specified for each fuel cycle and that the sequence will always be j
followed, except for rod insertion resulting from a scram or rod runback or
)
during low-power physics testing.
Interim Technical Specification LC0 l
t 3.1.5 has similar requirements.
The control rod sequence for use during
^
3-2
)
1
5
..?
a
}
Cycle. 4 ih-' given in Table 3-4.
The identification Of the control rod groups 'is shown in Figure 3-1.
Extended operation of lycle 4 up to an additional 220 EFPD will_ re' uire no changes to the control md sequence.
q l.
The regulating rod is located in the central refueling region (rod
]
group 1). This group is partially withdrawn before criticality is achieved
]
L and' then n:aintained in its most-reactive control rod position for the j
remainder of the operation.
In this manner, ninor reactivity adjustments j
~
can be made most rapidly with the minimum amount of control rod motion.
This is consistent with the method of operation utilized for the control l
rods in previous cycles, including Cycle 4.
1 i
A summary of the ' calculated power peaking factors obtained during a typical rise-to-power at the nominal EOC4 using the control rod sequence is given in Table 3-5.
It was conservatively assumed that the-reactor was shutdown for 90 days prior to the startup.
The results in Table 3-5 show the correlation between control rod group insertion and the core power level..
Power peaking factors during. Cycle 4 extended depletion are discussed in Section 3.3.
The basis of Technical Specification LC0 4.1.3 contains maximum values of region peaking factors.
At power less than 20% (core outlet gas temperature {950*F), the maximum RPF 'is 3.0 and the maximum tilt is 1.61.
At powers between 20 and 60% (outlet temperature from 950*F to 1250*F) the applicable limits are an RPF of 2.15 and a tilt of 1.34 to 1.46.
At higher powers, above 60% (outlet temperature >1250'F), the limits are an RPF of 1.83 and a tilt of 1.34 to 1.46.
j From the data given in Table 3-5, it can be seen that at the nominal E0C4 the calculated power peaking factors for the various power levels do not exceed those given in the basis of Technical Specification LC0 4.1.3.
This is true for both the radial region peaking factors and the intra-region peaking (column tilt) factors.
The analyses indicate that at powers between 0 and 2%, the maximum RPF may slightly exceed the LCO 4.1.3 basis 3-3
U(
p p,
I
' maximum value of-3.0, A two-foot -insertion of group _2B until. group 4E-is fully-withdrawn, as allowed by. Interim Technical Specification LCO 3.1.5,
- will: correct this situation, if necessary.
ForL rise-to-power during extended operation of Cycle 4 to ' 520 EFPD, criticality-and subsequent power _ levels are reached with fewer control rods
' inserted in the core. Hence, power peaking factors are reduced relative to
'those shown in Table 3-3.
i 3.3' PROJECTED CYCLE'4 EXTENDED OPERATION This section. presents the results of Cycle 4 extended depletion analyses-using design methods discussed in Section 3.5 of ~ the FSAR.
Fuel and burnable poison loadings discussed previously were 'used as input (see Section 3.1);
The total Cycle 4 burnup of 520 EFPD was carried out using.
actual; power history up to 154.5 ' EFPD, and assuming operation at rated power for the remainder of the ' cycle.
Figures 3-2a and 3-2b present envelopes encompassing. projected RPFs and column tilts during Cycle 4 depletion to 520 EFPD.
The results indicate that RPFs and tilts during the extended Cycle 4 will be well within the-maximum values contained in the basis of LCO 4.1.3.
Envelopes are not-presented for fully rodded regions because there are no fully _
' inserted control rods during extended depletion at high powers.
Axfal zoning of the fuel and burnable poison is provided (1) to produce' a power distribution which tenus to reduce axial fuel temperature peaking, and (2) to maintain the desired axial power distribution with depletion.
The calculated axial power factors in the bottom layer of each fuel region during Cycle 4 are shown in Table 3-6.
The calculations were carried out with the GATT code, the three-dimensional whole core model which has been used in the semi-annual fuel accountability analyses for the past three cycles and for the early portion of Cycle 4.
It can be seen 3-4
s_
that the limits on peaking factor in the lower fuel layer assumed in the l
basis of LCO 4.1.3 are not exceeded.
These calculations were carried out to 520 EFPD.
The basis of Technical Specification LCO 4.1.3 also states that an acceptable flux distribution shall be maintained at lower power levels by keeping' the flux level in the center of the core at least as. high as the average level.
Table 3-7 shows the ratio of the flux in the inner core regions (Regions 1 through 19) to the core average flux for control rod configurations in Table 3-5 which can result in operation between 0% and about 20% power at the nominal E004.
The flux ratio is above 1.0 for all cases, consistent with the basis of LCO 4.1.3.
With further burnup during extended operation, control rods will be removed from the core until all control rods except the regulating rod in the central refueling region will be removed from the core.
The flux ratio with this control rod configura-tion was calculated to be between 1.12 and 1.13 throughout extended operation.
3.4 MAXIMUM CONTROL R0D WORTH The basis cf Technical Specification LCO 4.1.3 states that the accidental removal of the maximum worth single rod pair shall result in a transient with consequences no more severe than the withdrawal of 0.012 AK, at rated (i.e.,
100%) power, from a core which has a temperature defect between 220'F and 1500*F of 0.028 AK.
In addition, the calculated worth of any rod pair in any configuration with the reactor critical must be less than 0.047 AK.
The same requirements are contained in Interim Technical Specification LCO 3.1.5.
The rod withdrawal accident (RWA) at full power evaluated in Section 14.2 of the FSAR assumes withdrawal of a control rod worth of 0.012 AK at equilibrium EOC with an equilibrium EOC temperature defect of 0.028 AK.
Because the consequences of an RWA are a function of rod worth, steady-state core temperature (i.e., initial power level), and temperature coefficient (which varies with burnup during the cycle), it is 1
3-5
p r
l i
f i
\\
necessary to' evaluate control rod worth as a function of' control rod insertion.
-The control rod withdrawal sequence for the extended Cycle 4 is
" described in Section 3.2.
For this sequence the maximum control rod worths
- at rominal E0C4 are shown in Table.3-8..The results in Table 3-8 indicate that the maximum worth rod ' pair in any source power critical configuration during Cycle -4 is 0.018 AK, which is less than the 0.047 AK limit of LCO - 4.1.3 and Interi.n LCO 3.1.5.
As was previously discussed, under most circumstances all control rods with the exception of the regulating rod are fully withdrawn during the extended depletion of Cycle 4.
Under these circumstances t.he RWA is by default limited to the regulating rod.
The worth of this rod (from 115 inches.to fully withdrawn) is only 0.002 AK, i.e., much less than the 0.012 AK. used for the FSAR analysis.
The consequences of RWA of the regulating rod are negligible compared with those discussed in the FSAR.
Furthermore, the power-level during Cycle 4 extension is less than 100%, so. a larger temperature defect is available to mitigate the RWA.
However, as shown in Table 3-5, at the beginning of extended operation (i.e., at 300 EFPD in Cycle 4), especially after a prolonged shutdown, there is sufficient excess reactivity (due to decay of xenon and conversion of Pa-233-into U-233) to allow core operation for a brief time at the rated power.
Consequently, the RWA must be evaluated for these condititm It was conservatively assumed that the reactor is shutdown for 90 days prior to beginning extended operation, resulting in reactivity buildup due to xenon and Pa-233 decay.
.As indicated in Table 3-5 operation at rated power can be achieved at 300 Ef90 with control rod group 3B inserted 30% into the core.
As shown in Table 3~7, the maximum worth fully inserted rod in this configuration is only 0.008 AK.
The maximum worth rod with group 3B fully inserted is 0.014 AK, but with group 3B inserted only 30% into the core the worth of any 3-6
~..
control rod: in. that group is less than' O.008 AK.
As is shown in Section 3.6, the temperature defect between 220'F and 1500*F is larger than 0.028 AK atJ300 EFPD.
Therefore, the consequences of an RWA at the ncminal.E004 at rated power are less than those of the RWA described in Technical Specifications.
Table 3-8 indicates that with two rod groups. fully inserted, a maximum control' rod worth of 0.014 AK is obtained.
However, Table 3-5 indicates that the maximum power level that can be achieved in this configuration is from 40% to 60% of rated power. The difference in average fuel temperature between these conditions and operation at rated power make an additional 0.006 AK in temperature defect available to mitigate the consequences of an RWA. 'Therefore, the consequences of an RWA of an 0.014 AK rod at 60% power at 300 EFPD are the same as those of an 0.008 AK rod at 100% power, which-are in turn, as discussed above, less than those of the RWA described in the Technical Specifications.
Similar evaluations of the RWA for all configurations shown in. Table 3-8 indicate that in all cases the RWA consequences at nominal EOC are l
bounded by those of the RWA described in the Technical Specifications.
As core operation is extended to 520 EFPD, the power level associated with the maximum control rod worth decreases, thereby providing more temperature defect available to mitigate the consequences of an RWA.
The total temperature defect, as discussed in Section 3.6 remains above 0.028 AK.
It is, therefore, concluded that the consequences of an RWA during Cycle 4 extended operation are bounded by those of the RWA discussed in Section
'14.2 of the FSAR and in the Technical Specifications.
3.5 CORE SHUTDOWN MARGINS 3.5.1 Control Rod System SDM 1:
Interim Technical Specifications LCO 3.1.4/SR 4.1.4 state that a shutdown margin (SDM) of 0.01 AK shall be achieved under the following conditions:
3-7 l
~
i l
j i
' 1.
The ' highest worth control rod pair is fully withdrawn and - not
]
. insertable, all inoperable ' rod pairs are at their pre-scram j
. positions, the core average temperature is at-220*Fr and Xe-135, l
-Sm-149, and ' Pa-233 levels are equal to those at the time' of shutdown. These SDM calculations shoulo assume shutdowri from-i operation at 100% power with equilibrated xenon, samarium, and l
2.
The highest' worth control. rod pair is fully withdrawn, ' inoperable rod pairs' are in their known pc::ition or assumed fully withdrawn,.
the core' average temperature is at 80'F, xenon is fully. decayed, samarium is fully built up, and protactinium converts into U-233' as a function of time after shutdown In assessing SDM. for the extended Cycle 4, it is assumed that the two highest' worth control rod. pairs are fully withdrawn.
This. assumption is consistent with Interim-Technical Specification LC0 3.1.1 and the. single failure ^ criterion.
These SDM calculations provide an estimate of the time-after shutdown available to reinsert the second highest worth control rod, if necessary to maintain an 0.01 Ak SDM.
It also must be shown that a SDM of-at least 0.01 AK can be maintained indefinitely. with full protactinium decay after insertion of the second highest worth control rod or insertion of reserve. shutdown system (RSS) material in one or both of the regions-l with withdrawn control rods.
i The SDM as a function of burnup in the extended Cycle 4 is given in L
Table 3-9.
The shutdown time is defined as the time for which the core is subcritical following a scram in which the maximum worth rod (s) failed to insert.
The first row in Table 3-9 indicates the SDM under normal scram i.
conditions where all rods are inserted.
The results in the other rows of Table 3-9 indicate that, after a core shutdown, there is no limit on time j.
'available to repair and to reinsert at least one of the inoperable control rods.
' Additional SDM could be ' achieved by activating the RSS or L
reinserting the second highest worth control rod.
The SDMs increase systematically with burnup during the Cycle 4 extended operation.
The 3-8 L__
L/
c 4
results in Table 3-9 are consistent with the results calculated for the Segment 9-Cycle 4 SAR (see Table 5-10 of Ref.1), which show that the minimum SDM occurs at the' middle ~ of Cycle 4, and then starts to increase systematically to the nominal E004.
On the basis of.the -above presentation, it may be concluded that the control rod system is adequate to provide an adequate SDM for the FSV core under all normal and postulated accident conditions during extended
. operation of Cycle 4.
General requirements for operability of control rod
~ drives during the cycle are provided in Interim Technical Specification LCO l-3.1.1.
3.5.2 Reserve Shutdown System SDM As noted above, Interim Technical Specifications LCO 3.1.4/SR 4.1.4 state that the reactor SDM shall be greater than or equal to 0.01 AK.
Specifically, for reactivity control with-the reserve shutdown system (RSS)
Interim Technical specifications LC0 3.1.8/SR 4.1.8 state that with any one RSS unit inoperable core operation may continue provided that the unit is capable of being made operable within 14 days following a reactor' shutdown.
Furthermore, in the basis of Interim LC0 3.1.8/SR 4.1.8 it is stated that the RSS must be capable of achieving reactor shutdown in the event that the control. rod system fails to insert.
j l
On the basis of above considerations, the SDM calculations for the RSS were carried out with the following conservative assumptions:
1.
The core, prior to shutdown, was operated at rated power long enough to equilibrate xenon and Pa-233.
2.
The scram signal fails to insert any withdrawn control rods.
1 3.
The inoperable RSS unit is the maximum worth one.
4.
The core is at room temperature (80'F).
3-9 e
j
j;
,et y
/
L 5.
The worth of RSS in rodded regions is neglected.
The. calculated SDMs' as a function of burnup in the-extended Cycle 4 are given in Table _3-10.
These results-indicate there is no time (11mit for-l the. repair _ and insertion into the core of ' inoperable controlf rods and/or.-
p RSS units before the SDM becomes inadequate.
L In - the basis of Interim LCO 3.1.8/SR 4.1.8 it is. stated that a worth of.RSS.of 0.100. AK is sufficient to ensure SDM during the first 14 days of Pa-233 decay.
Calculations indicate that'the worth of.RSS in-the extended.
Cycle 4, without any control rods present in the core (as may be the case at the end of the extended Cycle 4) is 0.111 AK. On the basis of SDMs and-the total. worth of RSS, it may.be. concluded that the RSS during the extended Cycle' 4 meets or exceeds the reactivity control-requirements.
3.6 KINETICS PARAMETERS The kinetics parameters for the extended Cycle 4 as well as for the
' initial 'and equilibrium cycles (taken from _the FSAR), are given in Table 3-11.
The data in this table indicate that _the equilibrium cycle kinetics parameters are.in close agreement with the extended Cycle 4 kinetics.
Interim Technical Specification LCO 3.1.7 requires that the reactivity change due to an average core temperature increase between 220'F and 1500'F (refueling temperature to rated power conditions) be at 'least as. negative as. -0.031 AK and no more negative than -0.065 AK during Cycle 4.
This requirement is imposed because FSAR accident analyses assumed a temperature defect of -0.028 AK, and the uncertainty in measured temperature defect is about 110% or 0.003 AK.
The calculated temperature defect decreases with burnup during each cycle due to thorium depletion and U-233 buildup.
The calculated temperature defect during the extended Cycle 4 is shown in Fig.
3-3.
The results indicate that the temperature defect between average core temperatures of 220*F and 1500*F is -0.037 AK at 390 EFPD and 3-10
-0.034 AK' at 520 EFPD.
Both of these calculated values meet the requirements of Interim LC0 3.1.7 for measured temperature defect..
3.7 NUCLEAR DETECTOR DECALIBRATION The power range nuclear detector signals, used by the control system to initiate plant - protective system (PPS) action, exhibit significant
._ decalibration due to control rod motion.
This detector decalibration is accommodated by a reduction in the fixed PPS' setpoints of Technical Specifications SL 3.3 and LC0 4.4.1-and frequent recalibration of the detectors.
The reduced setpoints have been reevaluated each cycle because of the different control rod withdrawal sequence and the different fuel loading distribution.
A rigorous analysis was done to determine these I
reduced setpoints for Cycle 1.
For subsequent cycles including Cycle 4 the reduced setpoints have been determined as follows:
o Calculate the " worst-case" detector decalibration factor (DF) for each control rod group (i.e., the case which would most delay the PPS trip).
o If 'the " worst-case" DFs indicate less delay in the PPS trip than the previous cycle, use the reduced setpoints from the previous cycle.
o If these " worst-case" DFs indicate more delay in the PPS trip than the previous cycle, then the reduced setpoints and/or the detector recalibration schedule for the previous cycle must be reevaluated.
The control rod sequence for the extended operation of Cycle 4 is the same as that for the reference cycle.
Furthermore, the extended operation of Cycle 4 is characterized by a deficiency of excess reactivity, i.e., all shim banks are expected, except for a few special cases discussed in Section 3.4,
)
4 I
I 1
3-11
4 to be fully withdrawn.
Consequently, during the expected mode of operation there will be very little control rod notion, and the decalibration of the detectors will be' minimal.
l However, during a rise-to-power operation early in the extended cycle, I'
as' was shown in Table 3-5, it is possible to have one or more rod groups fully.or partially inserted to achieve core reactivity control.
Consequently, decalibration of detectors _ may occur in this brief period (relative.to the total extended burnup duration).
The DF calculations for
-Cycle 4 indicated that the burnup effect is not significant, f.e., only one set of setpoints is needed to cover the burnup from 0 to 300 EFPD.
To retain these setpoints during the extension of Cycle 4 it is necessary to show that the worst possible DFs during extended operation are about the same as those for the fi rst 300- EFPD.
The built-in conservatism of setpoints can accommodate small changes between the DFs of the reference cycle and the-extended cycle.
The worst DFs calculated for the reference and extended Cycle 4 are given in Table 3-12.
When the extended Cycle 4 DFs stay the same or increase, the reference setpoints are more conservative to use during the extended cycle.
The extended cycle DFs for rod groups 4A, 3A, and 3B are slightly lower, i.e., less conservative than those of the reference Cycle 4.
However, the reference setpoints remain conservative. As shown in Table 3-5, group 4A is involved at power operations between 0% and 2% of rated.
The reference setpoint for the high power reactor scram at these powers is currently 64% power, i.e.,
far below 0.68 x 140 = 95%.
Full insertion of group 3A could occur at power operations between 18% and 28% of rated.
The high power scram setpoint for these powers is also currently at 64% power, well below 0.90 x 140 = 126%. Group 3B may be involved in core operations at higher powers including 100% (under the special core conditions discussed in Section 3.4).
In this case the high power scram setpoint at rated power should be 0.79 x 140 = 110%, which is less than the current setpoint of 115%.
However, at the current limit on core power of 82%, the high power scram setpoint is currently set at 105% power, which is less than the necessary 3-12
o 1-setpoint of 110%
Furthermore, the detector calibration schedule. for - the hi.
reference Cycle 4 specifies.that the Group.3B should be calibrated at 85 to 11 105 Linches of withdrawal.
Of course ' the b'urnup of-Cycle 4-- during' extended.
l operation.'will result in a further loss in the core reactivity.
This in turn
. wil1 ~ result in the us'e of all' rod groups, including Group 38, occurring. at l;
- ever lower. powers, thus increasing the margin between the'setpoint and the actual core power.
l
=.Therefore, it is concluded that'the reference setpoints for Cycle 4 are
- equally applicable for-the extended' operation of Cycle 4.
)
i 3-13 1
pj, l
lf f,
y v
TABLE 3 '
PROJECTED POWER HISTORY
'F0,1 CYCLE:.4 EXTENDED-OPERATION.
1.i f
CALENDAR. DAYS
- CUMULATIVE REATOR.
AVERAGE FUEL INTO EXTENSION CYCLE 4 EF90 POWER %-
TEMP DEGREES F 0
300 80 1373
-75~
360 80 1373 118:
390
.70 1332
- 176 425' 60 1242-
-236 455 50 1182'
. 324' 490 40.
1116:
.r-
- 424 520-30-1026' 3-14
_ = _ _ - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _
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79759717 G
W 84 21 9
N 2,
3, 1
e
)
d 2
2 (A
i O
l L_
cu E
N R
'33935935 O.
C 5
40961711 4.6 22 7
D 7,
8, E
T 2
2 CE J
O RP 1609212 1
4 99849719 84 21 9
2, 3,
2 2
D P
3 F
3 E
2U m
5 u
l e
+
i a
7 d
n t 7
i 23 o o 1
l 334568t T
1 c
223333 u
)
u ha2222l a
N TPUUUUP
(
wh
!l; fI I
- &. '.46;hi
p TABLE 3-4
' CONTROL 1 ROD' SEQUENCE FOR CYCLE;4 1
1 1
Group
' Sequence Withdrawn Regions 1
2A(a) 2,4,6 2
4F(a) 25,31,371 3
40 23,29,35 4
1(115" out).
1 5
4B 21,27,33 6
2B 3,5,7 -
7 4E
- 24,30,36-8 4A' 20,26,32' 9
4C
-22,28,34 10
'3C 10,14,18 11 3A 8,12,16 1-12 3B 9,13,17 13
'3D 11,'15,19-14 1(fully out) 1 (a)
Rod groups used for rod runback.
1 L
3-17
- i...
l.
i L'
- ~
(>t r'
- I 4
..k 1
TABLE 3
- TYPICAL' RISE-TO-POWER AT NOMINAL'-E0C4 (300 EFPD).
j r
1 l
l i
Group' Insertion Max' Max Tilt:
Power, %
Fraction RPF Rodded' Unrodded 0
4E @ 0.5 N/A N/A
'N/A 2
'4C 9 0.5' 2.13
'1.44 1.29-8 4C'@ 0.07' 1.94 1.35 1.25 18 3C'@ 0.19-1.93 1.29 1.28 28 3A @ 0.5
- 1.78 1.28 1.26 40
.3A'@ 0.05 1.74 1.21 1.23
. 60 '
!38 @ 0.71 1.73 1.21 1.24 70 3B @ 0.56 1.72 1.22 1.24 85'
.3B @'0.46-1.72 1.22 1.24~
-100 3B @ 0.30.
1.71 1.23 1.25.
L J
l
('
3-18 l
l.
[
=
TABLE 3-6
^ *
SUMMARY
OF CONTROL R0D INSERTIONS AND AXIAL. POWER FACTORS
-IN BOTTOM FUEL LAYER g
1..
l' CYCLE'4 1
REG /EFPD
- 300.0 350.0 390.0 435.0 470.0
'520.0 1
2
.704 2.760 2.758 2
.768 2
.773 2
.789' 2
0
.615 0.672 0
.671-0.680 0
.685-0.700:
3 0
.575 0.628 0.
628 0'.640 0 -.646 0.663:
4 0
.564 0.
619 0.619 0. 629 0.635 0.652 1
5 0
.559 0.615 0.613 0.623
~0
.628 0.644 6-0
.611 0.671 0.669 0.680 0'.685' 0.700 7
0
.616-0. 675 0.673 0'.
683 0.688 0.703 8
0'
.564 0.622 0.622 0.634 0.640 0.656 9
0
.611 0.665 0.665 0
.676 0
.682-0.698 l
10 0
.550 0'.607 0'.607 0
.618 0.624 0. 641 11 0
.560-0.628 0.
630 0.643 0.649 0.666 12 0.
.603 0
.664-0.664 0.674 0.680 0.697 l
13 0
.558 0.608 0.607 0.619 0
.625 0.643 l
14 0
.586 OL.645 0.644 0.656 0-662 0.679 15 0
.521
~0
.585 0.586 0
.598 0.605 0-.622 16 0
.592 0.652 0.651 0
.664' 0.670 0.687 17 0-
.561 0.615 0.614 0.626 0.632 0.648 18 0.
.556 01.614 0.614 0.628 0. 635 0.653 L
19-0
.574-0.644 0.644 0
.657 0.664 0.681 20 0
.642 0.700 0.699 0.710 0
.716.
0. 732
' 21 0
.589 0.640 0.641 0.653 0.660 0'
.678 22 0
.535 0.583 0.587 0.603 0.612 0.634 23 0
.615 0
.671-0.673 0.685 0.692 0.711 24' 0
.620 0.687 0.688 0.700 0
.707-0.724 25 0
.536 0.599 0.601 0.614 0.621 0.639 l
26 0
.627 0-.686 0.686 0.696 0.702 0
.718-l 27 0
.647 0.701 0.699 0. 710 0.716 0.733 i
l
'28 0
.575 0.628 0.627 0.639 0.647 0.665 i
l 29 0
.511 0'.563 0.567 0
.582 0.592 0
.614-l 30 0
.605 0.669 0.669 0.681 0.688 0.706 1
31 0
.621 0.688 0.688 0.701 0.707 0.725 l
32 0
.538 0.593 0.595 0.611 0
.620 0
.640 33 0
.533 0.583 0.587 0.604 0.614 0.637 34 0
.624 0.680 0.679 0.693 0.700 0.719 35 0
.590 0.650 0.650 0.663 0
.671 0.689 36 0
.545 0.603 0.609 0
.623 0
.631 0.649 37 0
.621 0.687 0.686 0
.698 0.704 0.720 l
AVERAGE
.585
.644
.643
.656
.662
.680
- CONTROL R0D INSERTION DEPTH BY FUEL LAYER.
LCO 4.1.3 BASIS LIMITS:
FULLY INSERTED OR FULLY WITHDRAWN: 0.90 PARTIALLY INSERTED:
1,23 i
3-19
____w___
x--_--------
P r-o-
- ..a p
O
}
jf ~
ta
. TABLE 3-7 f
1
-FLUX RATIOS:AT LOWER POWER v;;
. CYCLE 4 9 300.EFPD.
u.
L
-POWER, %
R_
b; j
u n
O' 1.36-2 1.12 8
1.03-16.
1.01 R'= Average 5f flux (Regs.1 through-19)/ average flux (core)
L i
3-20
__n___.__
J
.c 4
.f 1
m q
L
-TABLE 3-8' g;
WORTH OF CONTROL R0D GROUPS AND MAXIMUM R0D AT NOMINAL E004 Group Cumulative.
- Max Rod RWA Groups In Worth, Ak' Worth, Ak Worth, Ak Region RR(1) 0.002 0.002' O.002
'1
+3D 0.016 0.018 0.008-15 q
+3B.
0.023 0.041 0.014 17
+3A
.0.012 0.053 0.014 17
+3C 0.022 0.075 0.016 18
- +4C-0.014 0.089 0.014 15
+4A 0.007 0.096 0.017 11
+4E(2) o,007 0.103 0.018 13
~
.(3)
'O.112 0.215 N/A N/A (1) Regulating rod 115" withdrawn.
(2) Source power criticality at 300 EFPD.
(3) Groups 28, 48, 4D, 4F, 2A and the regulating rod fully inserted to assure
- subcriticality.
3-21
im t
- ,4 p.ji
- . '
TABLE.3-9-4 m
1 CONTROL ROD ~ SHUTDOWN MARGINS (AK)lI.N-THE. EXTENDED CYCLE 4; j j -
Shutdown-Core Avg
- EFPD, 1 Time, Days-Temp',1
- F CR'out RSS in 390
. 520 t
0 220 0.
0
'0.175
- 0.200-0 220 22 0
0.116:
0.143 0
220 21 + 22 0
- 0.086.
0.114 0
80 21 + 22-0 0.080
- 0.108.
3
-80 21'+ 22 0
0.050 0.079 14 80 21 + 22' O
0.044 0.073..
28.
.80' 21 + 22 0
0.039 0.067 156 80, 21~+ 22 0'.
0.032 0.060 224 60 21 + 22 0
0.025 0.052 3-22
-c l-;.
i.
TABLE 3-10 k
RSS SHUTDOWN MARGINS (AK) IN THE EXTENDED CYCLE 4 1
Shutdown Inoperable EFPD Time, Days RSS Hopper 390 520 l
0 0
0.102 0.130 0
22 0.080 0.107 3
22 0.050 0.078 14 22 0.043 0.071 28 22 0.037 0.065 56 22 0.030 0.057 224 22 0.023 0.049 3-23
xx 4
21 9659 01 01 00 82 19 0
1 31035 000000 e
C 45 40 0
00131 8 000000 O
lc E
00 33 0
000012 000000 y
C e
m X
u i
h 44 r
t 066464 b
i 00 5
060003 28471 0 w
1 1 0
520100 21 0852 i
l xx 5
217453 01 1 1 00 82 54 0
1 30031 000000 i
C.
36 86 0
00031 9 000000 uq O
E B
00 22 0
000012 000000 44 D
31 5937 P
00 8
000000 262096 4
F 1 1 2
590000 209531 E
xx 4
213850 01 01 00 e
82 1 9 0
1 32036 000000 l
0 54 40 0
00131 8, 000000 c
2 y
5 00 33 0
000012 000000 C
de 44 d
0 213525 n
P 00 5
000000 228858 e
F 1 1 6
560000 21 9641 t
E xx 4
21 0652 01 01 00 x
82 1 9 0
1 32039 000000 E
0 45 40 0
001318 000000 0
3 00 33 0
000012 000000 SR E
T 44 E
1 M
00 7
000000 000000 1
A 1 1 7
596500 200052 R
3 xx 5
203261 011 200 A
91 71 0
1 31 038 000000 P
E C
18 18 0
001 31 9
000000 L
O s
B E
00 32 0
000012 000000 C
A e
1 T
r 1
o e
E C
X N
I l
h 44 K
e t
444883 00 0
000000 1 27647 t
i t
w 11 5
454300 242572 i
xx 6
201 1 63 01 1 200 n
93 0
131 031 000000 I
C 00 64 0
001310 000000 O
B 01 22 0
00001 3 000000 1
s, no A
i t
c t
a n
r a
f t
p s
s n
n o
o n
e r
c o
s m
t i
n u
y t
i o
t e
a c
i e
n c
a t
f e
r d
d f
c i
u l
e d
y n
n o
n a
o o
r o
l r123456 r123456 p35 r
e t
t 33 t
d ur ur l 22 u
eo eo a - -
e e
ns ns nUU n
v r
r o
du d u i
imm t
d t
ec ec t oo ptl c
ye ye crr moo e
ar ar l
P aFF of C f
l P
l r
r f
e e
F P
E D
D oy
.,, ; v a,-
n
- je' i
$,,> g H
TABLE 3-12
^
a WORST DETECTOR DECALIBRATION FACTORS i :.;f i 1
Reference Extended Description Cycle 4
' Cycle 4
'l Insert 4E + RWA CRIS 0.62
' O.65 ~
Insert 4A + RWA CR18 0.77 0.68 Insert 4C.+ FWA CR10 0.92 0.94 Pull 3C + RWA'CR9 0.86 0.91-Insert 3A + RWA CR9 0.95 0.90 Pull.3B + RWA CRIS 0.83-0.79 Pull 3D + RWA CR1 0.90-0.91-
- 'ic k-
.'l 3-25
u
__xa--s-uz
1
)
j i
i N
f j
I I;
Nh,- '
G- -
.,rg -
tIi$$' g; 36 37
/
20 35 4E 4F 4D 4A
'8 8
21 34 18 g
4C 3C
,3D 3A 4B g
ge m
33 n
7 2
s 22 Jy;
-g 4B 3B
,2B 2
3B 4C W*.
40 J.l,,.?.>
32 16 6
1 3
10 23
$s=
4A 3A 2A 1
2B i/ "
30 g
!//k 31 15 5
4 11 24
.m73
,;fljg 4F 3
2B 2A 3D 4E fx
' $l// _
p.
SF is n
12 25 3a j;,,,.
4E 3C 3
- 3 3A 4F W
j
\\
'ik a_W 4A h ' '
ilfMb, 40 2.
22 2.
=--
2 5Y 4
fB
, # ~7
'ghylu. # ghf. -,;,,,,,,[ Mi' Y FUEL REGION IDENTIFICATION I
NUMBER Figure 3-1 Identification of control rod groups 3-26
. h ; J' l-1 1.6 i-L i
6
.1.4 -
l l
1.2-i-
4 1.0 -
L.
G.
0.8 -
Cr 0.6 -
~ + -
, TECHNICAL SPECIFICATION 0*4 -
LCO 4.1.3 BASIS LIMITS:
RPF:
1.83 TILT:
1.34 0.2 -
0.0 1.0 1.1 1.2 1.3 TILT Figure 3-2 Tilt envelope for extended Cycle 4: (a) unrodded regions 3-27
- c..
f l
4 1.6 TECHNICAL SPECIFICATION 1.4 -
LCO 4.1.3 BASIS LIMITS:
RPF:-
1.83 TILT:
1.40 1.2-
~
1 1.0-e L.
a_
o,8 -
0.6 -
0.4 -
~
0.2 -
0.0 1.0 1.1 1.2 1.3 TILT Figure 3-2 Tilt envelope for extended Cycle 4: (b) partially rodded regions 3-28
z '...
.g'
,o
)
0.05 f
l Q,0 4 -.-......*..........;........
...........'............j.............
......>......i.,....
p i
390 EFPD h'
w c)
. u; h.
O,Q3 -.. ~.. ~.4 - ~. ~.i. ~.. ~ +. - ~...
.i-_._.._.-+.~..~i.._...
.w a
w m-5 520 EFPD x.
L.r C-0.02 - ----------+~<--+-----i-----~~~+-----i--
1 w-9 0.01--- --- -
r -
- i--- -
? --y----
+--+-
0.00 b
2bo 4b0 6b0 8b0 10b0 12b0 1400 16b0 1800 AVERAGE CORE TEMPERATURE (*F)
~
I Figrre 3-3 Temperature Defect vs. Average Core Temperature 3-29.
'- ~ ' - ' '" - ~ - -.. _ _ _. _ _ _ _ _. _ _ _ - - - - - - -. - - - _,
e I
').
3 O
.s.
/
7 4
4
' h
- ~, +
p 1
(>'
4
'l m
i e
l 1;
i.
m
.e ;s 4.
THERMAL-HYDRAULIC AND MECHANICAL PERFORMANCE l
1-4.1 THERMAL PERFORMANCE i
The nuclear performance analyses discussed in Section 3 indicate that 1
the. power disti.bution during Cycle 4 extended operation falls.within the h
limits described in the Technical Specifications and the FSAR.
No changes are planned for the operation 'of the core cooling during thd extension of Cycle 4 (i.e.,
helium temperature at de core inlet and average outlet temperature will be enveloped by the FSAR reported values).
Accordingly, the temperature limits presented in the FSAR will not be exceeded during extended operation of Cycle 4..This conclusion is supported by analyser using.the COPE code (Ref.13), which is discussed in the FSAR.
The results of these analyses are shown in Table 4-1.
4.2 HYDRAULIC PERFORMANCE As noted in Section 4.1, the thermal performance of the core during -
Cycle 4 extended operation is essentially the same as that of previous cycles.
No changes will be made in fuel element geometry.
The power distributions expected during the extension of Cycle 4 are within the envelopes defined by the basis of. Technical Specification LCO 4.1.3.
- Hence, except for the opening of cross-flow gaps, as discussed and accounted for in Section 3.6.2.2 of the original and updated FSARs,. core coolant flow characteristics are also unaffected.
Accordingly, there are no changes in the hydraulic performance of the core from that of the initial core or the equilibrium core.
Maximum core pressure drop during extension of Cycle 4 is not expected to exceed about 5.7 psid, which is less than the design equilibrium. core value of 8.4 psid.
I 4-1 A
)
4.3 FISSION PRODUCT RELEASE
)
During extension of Cycle 4, the FSV core is expected to be operated within the limits presented in the FSAR and contained in the Technical Specifications.
Accordingly, the coated fuel particle failure and the i
fission product release characteristics of the fuel are expected to be within design limits, and the design radionuclides inventories presented in Section 3.7 of the FSAR will not be exceeded.
These conclusions are consistent with operating experience gained during Cycle 1-3 and during the first 154.7 EFPD of Cycle 4.
4.4 MECHANICAL PERFORMANCE Table 4-1 provides a summary of the fuel element stress, strain and bowing analyses described in this section.
These analyses were performed using the methods discussed in the FSAR.
Operating and shutdown strain and stress distributions were calculated for the axial and radial orientations throughout the period of extended operation. During core operatic the fuel elements will be exposed to fast neutron irradiation, which wil. induce dimensional changes in the graphite. An analysis was performed to calculate the expected dimensional changes of the fuel elements as a result of extended operation of Cycle 4.
As shown in Table 4-1, all fuel element mechanical performance parameters will be less than the maximum values given in the FSAR for the initial core fuel elements except for the fuel element bowing.
The maximum calculated fuel element bow is 0.129 in.
The FSAR maximum value, given for the initial core, is 0.090 in.
A review has been performed to assess the consequences of this bowing.
The possible safety consequences were evaluated, and the potential effect on the operation of the fuel handling l
machine (FHM) was also considered.
It was determined that the safety l-consequences are within the bounds previcusly considered in the FSAR and that the fuel element with the maximum bow can, even in the worst case, be readily handled by the FHM, i
4-2 L________-
..r L
The evaluation considered the following:
control-rod insertion with' a misa11gned core, seismic. events, reactivity effects due to gaps, coolant channel misalignment, effect. on fuel temperatures, effect on' fuel element stresses, and flow induced vibration.
This evaluation assumed a bow of 0.150.in.,
which is larger than the maximum calculated fuel element bow.
The fuel handling evaluation determined that the maximum allowable bow for an element using conservative assumptions for the effect of the bow on the fuel handling geometry is 0.136 in.,
which is larger than the maximum calculated bow.
Controi rod insertion will not be affected by the bow.
The maximum bow occurs in the top layer of the activ6 core.
The maximum radial displacement of the control rod channel the top of the core is limited by the region constraint devices.
Even with the angular misalignment l
introduced by the bow, the total misalignment of the control rod channels is much less than that of the core misalignment tests reported in Section 3.8 of the FSAR.
l The effect of the fuel element bow on fuel temperatures can be due to two conditions: crossflow and coolant channel alignment.
The dowels will maintain block alignment during all normal operating conditions.
The bow will not cause any relatue block motion which could result in partial coolant channel blockage. The larger bow will result in increased crossflow in the region.
However, the bow will be larger in older, lower power regions.
These regions have the orifice valves relatively closed so the direction of crossflow will be into the regions, resulting in lower fuel temperatures.
i The seismic event discussed in Section 14.1.1.2 of the FSAR, which l
results in a single dowel engagement, will not be affected by the increased maximum fuel element bow.
Single dowel engagement is possible only for six l
standard fuel columns and only for the gap between the second and third J
layers of the active core in those columns.
The calculations performed detercained that bows larger than the FSAR value of 0.090 in. will not occur for thnse locations.
4-3 I
.O
There will be no effect on core. reactivity. due to the fuel element bowing.
While the bow will cause some gaps to' increase in size, other will decrease correspondingly.
The net change in gap size will be due only to 1
radial! shrinkage of the fuel elements with irradiation in a manner consistent with the original design assumptions.
There will be no fuel element stress increase resulting directly from the bow.
The gaps surrounding each' column, coupled with the radial l
shrinkage of the fuel elements with irradiation, are large enough to 1
l preclude interfeu ice.
The increased crossflow could possibly result in I
higher thermal stresses but this will be offset by the fact that stresses decrease rapidly after the first two or three years of operation.
Accord-ingly, the peak stress conditions will occur prior to reaching the maximum bow condition.
- The fuel columns will be more stable with respect to flow induced-vibration or deflection since the orientation of bow defines a preferred direction for the column to lean.
The FHM grapple can engage elements with a top face about 2' off from the horizontal.
This corresponds to a difference in elevation of 0.500 in, across the flats of the element.
This difference would result from a cross-block differential in axial strain of 0.0160 in/in (neglecting block average axial shrinkage which would allow a slightly higher differential strain).
The calculated bow due to this axial strain differential is 0.136 in.
This result conservatively assumes that all of the gap due to the bow is at the
)
top of the element.
The maximum bow calculated at end of Cycle 4 extended operation is 0.129 in.
This occurs for a fuel element in the top layer of a Segment 5 1
region (Region 23).
An additional 45 elements from Segments 4, 5, and 6 were identified as having projected cross-block axial strain differentials 1
1 4-4
)
e
-4 large enough to result in a bow greater than 0.090 in.
The top layer of Segment 3 was also analyzed, and one element was identified with calculated l
end of life bow greater than 0.090 in.
It should be noted that no problems were encountered removing this. element.
Based upon this evaluation, it is concluded that the maximum bow calculated for extended operation of Cycle 4 is acceptable' and presents no fuel handling problems or safety consequences beyond those previously evaluated in the FSAR.
l i
i 1
4-5 l
~$.
b e
',9
~ TABLE 4-1 EXTENDED CYCLE ~4 CALCULATED PEAK CONDITIONS VERSUS FSAR INITIAL CORE PEAK VALUES Extended FSAR Cycle'4 3
Parameter Peak Value PeakValue(al
-Axial stress (psi) 450 301 Radial stress (psi)
-200 54.3 Axial strain (%) (contraction)-
3.0 2.0 Radial' strain (%) (contraction) 0.8' O.8 Fuel. element bowing (in.).'
O.09 0.129 Fuel temperature (*F) 2300 2109(b)
(a) Values calculated using FSAR methods.
s
-(b) Peak fuel ' temperature in core during Extended Cycle 4..
4-6 Li__z_u1_: _1 -
4-l'
'c i,
1 5.
SAFETY-ANALYSIS l
H l
5.1 INTRODUCTION
.In.this section, - the Safety Analysis presented in Chapter XIV of the Fort-St. Vrain FSAR is reviewed to determine potential effects of extension of Cycle' 4 on accidents and events discussed in the' FSAR.
The purpose of such a review is to assure that the worst case conditions previously defined.
for accident analyses, and found to be acceptable during the FSAR review, are.not exceeded during extension of Cycle 4, and that no unreviewed safety -
questions are presented.
l As a 'first step in this review process, Chapter XIV of the FSAR has been examined. to identify analyses potentially affected by the extenstop. ef
' Cycle '4.
The results of this review are presented in Table 5-1.
Seven accident conditions (seme of which envelope other -less severe events) have been < identified as requiring more detailed review for potential effects.
These are:
1.
2.
Rod withdrawal accidents (RWAs).
3.
Column deflection and misalignment.
4..
Fuel element malfunctions.
5.
Loss of normal shutdown cooling (limiting case:
cooldown on one firewater-driven circulator).
5-1
s J
6.
Permanent loss of forced circulation [ Design Basis Accident No. 1 I
(DBA-1)].
f 7.
Rapid depressurization/ blowdown (DBA-2).
As indicated in Table 5-1, RWAs are discussed in Section 3.4 of this document, and earthquake, column. deflection and. misalignment, and fuel-element malfunctions are discussed in Section 4.4 of this document.
It is concluded in Section 3.4 that the neutronic parameters of a RWA will be unchanged, and that RWA consequences are no more severe than those of the postulated RWA described in the FSAR. The likelihood of a RWA will decrease significantly during the extension of Cycle 4, because most of the operation will take place with only the regulating rod pair inserted in - the core.
Thus, accidental withdrawal is precluded of any of. the shim control rod pairs that will already have been withdrawn.
l For fuel element malfunction, all stresses and axial and radial dimensional changes at: calculated in Section 4.4 to be less than or equal l
to those predicted in the FSAR for the initial core fuel elements.
Th'e maximum fuel element bowing, induced by fast neutron irradiation, is calculated in Section 4.4 to exceed that predicted in the FSAR.
The possible safety consequences of this bowing were evaluated in Section 4.4.
The safety evaluation considered control rod insertion with a misa11gned 1
core, reactivity effects due to gaps, seismic events, coolant channel mis-alignment, effect on fuel temperatures, effects on fuel element stresses, and flow induced vibration.
It was determined that the safety consequences are within the bounds previously considered in the FSAR.
The worst fuel element deflections, which could occur during an earthquake, are concluded in Section 4.4 to have consequences no worse than those described in the
'FSAR.
The probabilities of occurrence of fuel element malfunction and column misalignment have not increased. The remaining three areas of safety I
analysis are discussed below.
5-2
6 j
5.2 LOSS OF NORMAL SHUTDOWN COOLING, PERMANENT LOSS OF FORCED CIRCULATION, AND RAPIC DEPRESSURIZATION/ BLOWDOWN I
l The core thermal conditions resulting from firewater cooldown (Safe j
Shutdown Cooling in the event of loss of normal shutdown cooling), permanent loss of forced circulation, and rapid depressurization/ blowdown are known from past studies to be sensitive to specific core parameters which could be affected by extended operation of Cycle 4.
These parameters are radial region power peaking factor (RPF) and core outlet region temperature dispersion (mismatch), which, in turn, are limited by the FSV Technical Specification LCOs 4.1.3 and 4.1.7, respectively.
FSAR analyses of these three events include RPF and temperature dispersion values up to the LCO-allowable values.
The maximum RPF expected during extension of Cycle 4 varies with power, but remains within the maximum values contained in the basis of LCO 4.1.3, as discussed in Section 3.
The maximum temperature dispersion will be controlled not to exceed the LC0 4.1.7 allowable value by I
using the variable-orifice flow-control assembly located at the inlet to each refueling region.
Hence, extension of Cycle 4 does not result in core thermal conditions more severe than those already analyzed. As discussed in i
Section 4.3, no additional fuel failure occurs beyond end-of-equilibrium cycle conditions described in the FSAR.
Hence, extension of Cycle 4 does
~
not result in core radiological conditions more severe than those already analyzed in the FSAR.
Operation during extension of Cycle 4 is therefore bounded by the FSAR accident analyses.
5.3 CONCLUSION
S A review of Chapter XIV of the FSAR identified seven postulated accident conditions that required more detailed examination for potential impact from extension of Cycle 4.
No requirements for additional analysis have been identified; the FSAR analysis is found to remain valid in all 5-3 L-_--
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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. extension of Cycle 4, and that'. the extension. of Cycle 4 presents no unreviewed safety questions, as' defined in 10CFR50.59.
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TABLE 5
- PGTENTIAL EFFECTS.0F EXTENSION OF CYCLE 4 ON FSV FSAR ACCIDENT. PREDICTIONS l.
Potential Effects on' Event Analysis FSAR Chapter XIV Event Due'to Extension of Cycle 4
!i 14.1 Environmental Disturbances 1',.
' Earthquake.
Evaluation. required, see Section 4.4 of this document Wind effectsi Flood Fire None - The core ~is not:affected by
. Landslides these events Snow and ice I'
14.2-Reactivity Accidents and Transient Response
.. Summary of reactivity sources Excessive removal'of control poison.
Loss of fission product poisons Reactivity insertions in these Rearrangement of core components j events are bounded by rod Intrc@ction of. steam-into the withdrawal events i
core-Sudden decrease in reactor
. temperature.
- -Rod withdrawal accidents Evaluation required, see Sec. 3.4 of this document 14.3 Incidents
~Incidents Involving the Reactor Core
' Column deflection & misalignment Evaluation required, see Sec. 4.4 Fuel element malfunctions of this document Misplaced fuel element No change from Sec. 3.5.4.5 of FSAR
-Blocking of coolant channel No change from Sec. 3.6.5.2 of FSAR Contrel rod malfunctions No change from Sec. 3.8 of FSAR Orifiw malfunctions' No change from Sec. 3.6.5.1 of FSAR Core support floor loss of No change from Sec. 3.3.2.2 of FSAR coeling Incidents involving the primary No change from Section 4.2.2 of FSAR coolant system 5-5 1=
- ' J Table 5-1 (continued)
Potential Effects on Event Analysis FSAR Chapter XIV Event Due to Extension of Cycle 4 Incidents involving the control No change from Section 6.4.2 of FSAR and instrumentation system Incidents involving the PCRV No change'from Sections 5.9.2 and 9.7 of FSAR Incidents involving the secondary No change from'Section 10.3 of FSAR coolant and power conversion-system Incidents involving the electrical No change from Section 10.3 of FSAR system
-. Malfunctions of the helium purification system
' Malfunctions of the helium storage system
( None - The core is not affected by these events Malfunctions of the nitrogen system Malfunctions involving handling of 1
heavy loads 14.4 Loss of Normal Shutdown Cooling Evaluation required, see Section 5.2 of this document 14.5 Secondary Coolant System Leakage Steam leaks outside the primary No change from Section 14.5.1 of FSAR coolant system Leaks inside the primary coolant None - FSAR analysis encompasses core system / steam generator leakage thermal conditions allowable under (moisture ingress)
Tech. Specs.: extended Cycle 4 will not exceed same.
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Table 5-1 (continued).
Potential Effects on Event Analysis.
FSAR Chapter XIV Event._
Due to Extension of Cycle 4-14.6 Auxiliary System Leakage Failures, involving the. helium purification system.
Loss of both purification trains Possible effects would be bounded.by.
Failure of regeneration line Design Basis Accident No. 2, FSAR with simultaneous valve failure Section 14.11 and operator errorJ
- Accidents involving the gas waste
.No change from Section 14.6.2 of FSAR system Fuel handling and storage accidents Fuel handling accidents 1None-AnalysisinSection14.6.3of Fuel storage accidents f FSAR is bounding 14.7 Primary Coolant Leakage Possible effects would be bounded by.
Design Basis Accident No. 2, FSAR
-14.8 Maximum Credible Accident Section 14.11 14.9 Maximum Hypothetical Accident Same as FSAR Section 14.11 14.10. Design Basis Accident No. 1, Evaluation required, see Section 5.2
" Permanent Loss of Forced of this document Circulation (LOFC)"
14.11 Des"n Basis Accident No. 2, Evaluation required, see Section-5.2
" Rap 1d Depressurization/
of this document Blowdown" i
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6.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS-
{
No changes to the plant Technical' Specifications are-. necessitated by..
. the extension of Cycle 4 up to a total of-520 EFPD.
As noted'. in. ' Sect 1on ' 2,. the plant will continue to be operated in accordance with procedures based upon interim Technical Specifications for reactivity control.-
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7.
SURVEILLANCE TESTS
[
The following reactor core physics surveillance tests are performed at the beginning of each new fuel cycle.
SR 5.1.5-RX, Control Rod Reactivity Worth, is performed to demonstrate' I
that the measured worth for each control rod group withdrawn during power
. operation compares with the calculated worth withis some specified uncertainty.
This test provides assurance that the calc > lated values.used in the safety analyses are acceptable and that the reactivity discrepancy f
as monitored continuously can be accurately t iermined.
SR 4.1.7-RX, T.emperature coefficient of Reactivity, is performed over the fuel temperature range from 220 through 1500*F.
This measurement is done at the beginning of cycle (B0C), or as close to that. as feasible, to demonstrate that the temperature defect with a new fuel segment added is within the limits specified in Interim Technical Specification LC0 3.1.7.
The reactivity temperature defect must be more negative that the minimum limit (-0.31' Ak) to ensure that the temperature coefficients used in the accident analysis are adequate, and less negative than the maximum limit
(-0.65 Ak) to assure that the calculated worth of the Reserve Shutdown System. is adequate.
Since the temperature coefficients decrease with burnup throughout the cycle, their minimum value is at the end of cycle (E0C).
Therefore, the measurements are made to the BOC when the new fuel has been added, and are extrapolated to obtain an EOC minimum value.
j A special test, Detector Decalibration, is performed for the control rod configurations (each control rod group fully withdrawn) that exist for the specified control rod sequence from startup to full pover.
The 7-1
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measured decalibration factors'are compared with the calculated values,- and the high-power ' level scram setpoint schedule for decalibration given in
-Technical Specification LSSS 3.3 is justified using the measured data.
In addition, -the following. surveillance tests 'are required at specified frequencies throughout the cycle operation.
SR 4.1.1 and SR 4.1.3, O ntrol Rod Operability and Position Indication, and SR 4.1.8 and SR 4.1.9, Reserve' Shutdown System Operability, are conducted to demonstrate that the control rods and reserve shutdown system will. perform per tSe design requirements.
SR 4.1.4 A-W and SR 4.1.4 B-P-X, Reactivity Status Check, are performed weekly during operation and prior to sach startup (approach to critical), respectively.
These surveillance ensure that the reactivity discrepancy, the - difference between the expected and actual control rod configuration, does not exceed 0.01 Ak.
SR ' 5.1.7 a-X and b-X, Calculated Region Peaking Factors and RPF
' Discrepancies, are performed monthly or at regular burnup intervals during power operation to ensure that the measured RPF distribution is in agreement with the calculated RPF distribution and that Regions 20 and 32-37 are being orificed appropriately.
l As a consequence of these continuous surveillance, both the core reactivity and the power distribution are continuously compared with the previously calculated data.
Any trend in discrepancy can be monitored and evaluated, and appropriate action can be taken.
All of the physics tests r.equired to be performed at the BOC for Cycle 4 have been acceptably completed.
Since early operation in Cycle 4 was J
limited to 35% power, this testing was not completed until about 80 EFPD of burnup had been achieved.
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' All of the measured control rod group worths,' group 2B through 3A, met
~the associated acceptance criteria.
The measured reactivity temperature defect met the acceptance criteria that the measured value be less negative.
than -0.065 Ak and more negative than -0.031 Ak.-
The measured value at the BOC was -0.0538 Ak, and the value extrapolated to the nominal E0C' was
-0.0434 Ak.
Both of these are well within the required limits.
A comparison. of the Cycle 4 measured and calculated detector decalibration data indicated agreement consistent with that for previous cycles and supportive of the ar.alytical methods.
An evaluation to confi rm the adequacy of the. high power level scram setpoint schedule using combined measured / calculated decalibration data demonstrated the conservatism of this setpoint schedule.
Since the physics testing has demonstrated the adequacy of the calculated physics data for Cycle 4, and since required surveillance tests
. monitor the core reactivity and power distribution on a continuous basis, additional ~ physics tests are not planned for extended operation of Cycle 4
- However, if during the core monitoring any trend is observed which indicates a discrepancy of the calculated data,. additional physics. tests will be performed as required.
In addition, before the previously, planned end of Cycle 4 core burnup (300 EFPD). is extended, -the measured value for the reactivity temperature defect will be extrapolated to the planned end of extended operation. This will ensure that the extended end of cycle value is more negative than that required in Interim Technical Specification LC0 3.1.7, -0.031 Ak.
Based upon burnups achieved through April 30, 1989, the extended operation of Cycle 4 is not expected to exceed 460 EFPD by the time nuclear operations are discontinued on or before June 30, 1990.
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8.
REFERENCES q
1.
" Safety Analysis Report for Fuel Reload.3 -(Segment 9 - Cycle. 4)," GA -
Technologies Document No. GA-C17128, May 1983, PSC letter-to NRC P-83391, December 3, 1983.
- 2..
Burnette, R. D., " Radiochemical Analysis of the First Plateout Probe from the Fort St.
Vrain High-Temperature Gas-Cooled Reactor,"
GA-A16764, June 1982, PSC letter to NRC P-82419, September 27, 1982.
3.
Fuller, J. K. (PSC) letter to William P. Gammill. (NRC), "SAR for Core Region Constraint Devices," P-79068, March-23, 1979.
4..
Asmussen, K.- E., et. al., " Testing and. 0peration of Fort St...Vrain Up to 100%. Power," GA-C16701, ' June 1982, PSC letter-to. NRC P-82229, July 6,.1982.
15.
Alberstein, D.,
and K.
E.
Asmussen, " Technical Specifications for Operation of.
FSV with Region. Outlet - Temperature ~ Measurement Discrepancies," GA-C15781, June 1982, PSC letter to NRC P-82229,. July 9, 1982.
6.
Warembourg, D. W.
(PSC) letter to John T. Collins- (NRC), " Fort St.
Vrain Unit No.1 Fuel Element Meeting," P-84104, April 6,' 1984.
7.
Brey, H. L. (PSC) letter to Jonn T. Collins (NRC), " Transmittal of GA Fuel Block Test Reports 907057 and 907155," P-84109, April-11,1984.-
8.
Lee, 0. R. (PSC) letter to E. H. Johnson (NRC), " Response to NRC/LANL
. Concerns on Cracked Fuel Elements," P-84275, August 13, 1984.
9.
Heitner, : Kenneth L.
(NRC) letter to R.
O.
Williams, Jr.
(PSC),
" Dynamic Loading of Cracked Fuel Elements at Fort St. Vrain," December 30, 1986, 10.
Gahm, J. W. (PSC) letter to Document Control Desk (NRC), " Licensee
' Event Report 84-008, Final Report," P-85388, November 1,1985.
11.
Lee, O. R. (PSC) letter to Regional Administrator (NRC), " Interim Technical Specifications for Reactivity Control," P-85242, July 10,.1985.
12.
Williams Jr., R. O. (PSC) letter to Docunent Control Desk (NRC),
"Early' Termination of Fort St. Vrain Operations," P-88422, December 5, 1988.
13.
Katz, R., and G. R. Malek, " COPE, a Core Perfonnance Code for Gas-Cooled Reactors," GA-9802, November 15, 1969.
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+ CENERAL ATOMICS RO. Box 85608 e San Diego, CA e 92138-5608 (619) 455-3000
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