ML20058N818
| ML20058N818 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/31/1988 |
| From: | Ankney R, Osborne M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19298E400 | List: |
| References | |
| WCAP-11954, NUDOCS 9008150143 | |
| Download: ML20058N818 (60) | |
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HCAP-11954 g.1 4=
SAFETY' EVALUATION SUPPORTING'A'MORE-NEGATIVE EOL MODERATOR: TEMPERATURE' COEFFICIENT TECHNICAL SPECIFICATION-FOR THE-
- JOSEPH M. FARLEY. NUCLEARLPLANT =
UNITS 1 AND 2 December, 1988,
H. R. Perryl R.'F.:Schmidt R. L =Haessler a
. APPROVED: N APPROVED:
R. D. Ankney, Manager
=M, P t0sborne, Manager Core Design D Nuclear-Safety-Transient Analysis-II:
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-PDR-ADOCK 05000348:
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HESTINGHOUSE ELECTRIC CORPORATION Commercial Nuclear Fuel Division P.
0~. Box 3912 Pittsburgh, Pennsylvania 15230 02711:6-881220 i
Nestinghouse Class 3-N ABSTRNCT-
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-This-report proposes a relaxation of the Limiting Condition for Operation and j
, Surveillance ^ quirements values of Moderator Temperature Coefficient for the; j
'end: of cycle,1t ated thermal' power condition.
Relaxation is soughtiin order-;to.
j Limprove plant-availability =and minimize disruptions to normal plant operation, j
,while continuing to satisfy plant-safety criteria. A methodolegy for y
' establishing Technical Specificationiend of cycle Moderator Temperature Coefficient values that are-consistent with the-plant: safety 1 analyses is-described herein.. Specific applicationfof ti.- methodology to: the Joseph. M.
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Farley NuclearLPlant Units 1 &nd10 provides Technical Specification;ModeratorL Temperature Coefficient values which are proposed to replace the existing values.
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TABLE OF CONTENTS i
Section Title PAgt
1.0 INTRODUCTION
1-1 4
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1.1 Background
1-1--
I '. 2 Basis of Current E0L MTC 1-1; LC0_and SR Values j
1.3.0perational Considerations:'
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E0L MTC Tech Spec SR Value
1.4 Operational
Considerations:
.1-4' EOL MTC Tech-Spec.LCO Value 2.0-METHODOLOGY ~FOR MODIFYING MOST 1
NEGATIVE MTC TECH SPECTVALUES 2-1 2.1.
Conversion of Safety ~ Analysis 2-1
-l MUU to Tech Spec HTC 2.2 Conservatism of the ARI to ARO w
2-2 l
i MTC Conversion'
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2.3 Alternative MTC Conversion 2-2'
-Approach-i
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-2.4 Determining SR MTC from.
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LCO MTC j
2.5 Benefits of the' Alternative MTC-2-5 Conversion Approach a
3.0 MOST NEGATIVE FEASIBLE MTC APPROACH 13-1 APPLIED T0 J. M. FARLEY UNITS 1 AND 2-3.1 J. M. Farley Units 1 and.2 Accident Analysis MDC Assumption 3-1 j
i' 3.2 Determination of Most Negative 3-2.
Feasible MTC Sensitivities i
3.3 Maximum Allowed Deviations from 3-4 l
Nominal-Operating Conditions
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5 TABLE OF CONTENTS (cont'd.)
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le Section Title Eagt 3.0 3.4 Overall " Delta MTC" Factor for-3-9 ti J. M.-Farley Units 1 and 2 Reloads-3.5 Proposed J. M. Farley Units 1 and 2'.
~3-11" Tech Spec EOL MTC LC0 Value-4.0 SAFETY ANALYSIS IMPACT OF MOST NEGATIVE 1
'4-1 L FEASIBLE MTC APPROACH 5.0 DETERMINATION OF'MOST NEGATIVE FEASIBLE'
~ 5.t MTC SURVEILLANCE.VALUE~
J 6.0 SUSPENSION 0F HTC MEASUREMENTS 6-1.
BELOH 100 PPM 7.0
-CONCLUSIONS-7-1 I
REFERENCES R-1 7
APPENDIX A DETERMINATION 0F MOST-~ NEGATIVE FEASIBLE.
.A-1 MTC SENSITIVITIES A.1 MTC Sensitivity to Moderator.
.A-4
-Temperature and. Pressure Variation A.2 MTC Sensitivity'to RCCA Insertion,
'A-5 A.3 Sensitivity to Axial Flux A-7 i
(Power) Shape
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I A4 Sensitivity to Transient Fission A-9 l
Product (Xenon) Concentration APPENDIX B REFERENCED TECHNICAL SPECIFICATIONS-B-1:
i AND BASES SECTIONS FROM THE J. M. FARLEY-NUCLEAR PLANT TECHNICAL SPECIFICATIONS 1
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LIST OF.TABLESL i
Table Tlt1e
- P_ age ;
4.1 FSAR' Chapter 15 Events Hhich Assume--
4-2 i
a Constant 0.43 ok/gm/cc Value of.MDC LIST OF ILLUSTRATIONS F1aure Titie P_tgf_ _
f 3.1 Loop Control Rod: Locations-3-12
.l 3.2
.RCCA Insertion limits for 3-13' J. H. Farley Units 1 and 2 Reload' Cores 7
6.1 HFP Critical-Boron Concentration and MTC 6-3
.Versus Burnup'for Farley. Reload-6.2 HFP MTC:Versus. Critical Boron Concentrat'lon:
6-4?
for Farley Reload A.1 Change in MTC with= Increase'in-A110' T-Average above Nominal T-Average j
A.2 Core Average. Axial-Burnup versus-LA-ll Core Height at E0L J
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. Axial Power and Moderator Temperature
.A-12 L
Versus Core Height A.4 Delta MTC versus Axial Flux Difference A-13.
at EOL, HFP', AR0
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1.0 INTRODUCTION
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1.1 Background
For_ FSAR accioent analyses, the transient = response of the plant is dependent -
-on reactivity feedback effects,_in-~particular,-the moderator temperature coefficient _(MTC) and the Doppler power coefficient.- Because of the sensitivity of accident analyses results to the MTC value assumed, it is important that~the actual core MTC remain within the bounds:of the-. limiting values assumed in the FSAR accident analyses.
WhileLcore neutronic~ analyses will have predicted that the MTC is within:these bounds, the Technical Specifications require that the core HTC also be confirmed by measurement,' as-verification of the accuracy of'the neutronic predictions..These:HTCi measurements are performed:
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At beginning of cycle, prior to initial operation above 57. power, and
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Hithin 7 EFPD after reaching an equilibrium boron concentration of-300 ppm.
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1.2 Basis of Current E0L MTC-LC0 and SR Values In order to ensure a bounding accident analysis, the MTClis assumed.to be at.
Its most limiting value for the analysis conditions appropr. late to.each c
4 accident.
Currently, the most negative HTC limiting value is based on E0L L.
conditions (specifically with regards to fuel burnup'and--boron concentrat'lon),
L full power, with rods fully-Inserted. Most' accident analyses use'a constant,
. moderator density. coefficient (MDC) designed to bound the.MDC at'this worst L
set of initial conditions (as well as at the most-limiting set of transient conditions).
This value for-MDC forms the licensing basis for the~FSAR accident analysis, j
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Converting the MDC used in the accident analyses to a MTC is a simple calculation which accounts for the rate of change of moderator density with temperature at the conditions of interest.
In this report, the convention followed is to discuss the moderator feedback in terms of MTC, rather than MDC.
Nevertheless, it is important to note that the accident analyses actually assume a constant HDC value, rather than making any explicit assumption on MTC.
Technical Specifications place both Limiting Condition for Operation (LCO)-and-Survelliance Requirements (SR) values on MTC, based on the accident analysis assumptions on HDC described above.
The most positive MTC LCO limit applies to Modes 1 and 2, and requires that the MTC be less positive than the specified limit value.
The most negative MTC LCO limit applies to Modes 1, 2, and 3, and-requires that the MTC be less negative than the specified limit value for the all rods withdrawn, end of cycle life, rated thermal power condition.
The Technical Specification SR calls for measurement of the MTC'at BOL of each i
cycle prior to initial operation ab:sve 57. rated thermal power, in. order to demonstrate compliance with the aost positive MTC LCO. -Similarly, to demonstrate compilance with the most negCcive MTC LCO, the Technical Specification SR calls for measurement of the MTC prior to E0L (near 300 ppm equilibrium boron concentration).
However, unlike the BOL situation, this 300 ppm SR MTC value differs from the E0L LC0 limit value.
Because the HFP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOL'LC0' limit.
The 300 ppm SR value is sufficiently less negative than the E0L LC0 limit value to provide assurance that the LC0 limit will be met when the.300 ppm surveillance criterion is met.
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l 1.3 Operational Considerations:
E0L MTC Tech Spec SR Value It is becoming increasingly probable that reload cores will fall to meet the current 300 ppm surveillance.criterton associated with the E0L~LCO' limit.
The primary factors causing more negative MTCs near E0L are higher core average operating temperature and higher discharge' burnup.. Failure to meet the, surveillance. criterion does not by itself imply a failure to meet-the actual' EOL MTC limit stated in the LCO,..but invokes the requirement that the HFP MTC-continue to be measured at least once per 14'EFPD during the remainde'rTof thel 1
fuel cycle.
This repeated surveillance-is performed to demonstrate that;the j
actual LCO limit on E0L MTC is not violated, i
The drawbacks to the current E0L MTC Surveillance Criterion are:
1.
The current and planned fuel management strategy is expected to yield MTC values which will.be more negative than the existing 300' ppm.
f survelliance criterion.
This would result in repeated HTC measurements every 14 EFPD.
In addition, the E0L HFP AR0 MTC values-for these anticipated designs will-approach the existing LC0 1imit.
7 2.
If repeated measurements are necessary, they can require that load swings be performed, causing temperatures to-deviate'from-the.
I programmed reference temperature - situations which are never I
preferable to nominal steady state operation.
3.
The repeated measurements require the resources of. multiple operations personnel for roughly an entire shift, and require' greater-water processing for measurement via the boration/dllution method.
i Hestinghouse-designed PHRs which conform to Standard Tech'nical Specification (STS) format (including the J. M. Farley_ Units 1 and 2 Technical j
S'pecifications) generally feature a 300 ppm SR MTC value which is 9 pcm/'F-1 less negative than the E0L LCO. limit on MTC.
Given-the disadvantages'of
. repeating the HTC plant measurements, it is logical to inquire if this
' difference between the SR value.and the LCO value is overly large, and would-1 possibly invoke repeated measurements which are unnecessary.
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Examination of.both plant-specific characteristics and fuel management effects-on the difference between the 300 ppm HFP AR0 predicted HTC and the E0L (0. ppm) HFP ARO predicted MTC. Indicates that the 9 pcm/*F difference applied to Hestinghouse-designed STS plants is very conservative.
This implies that a failure to satisfy the 300 ppm surveillance criteria can occur, yet the actual EOL MTC value could show margin to the LC0 limit.
It is' concluded that relaxation of the difference between the SR limit value and the LC0 limit value should be investigated,-so as to preclude unnecessary MTC testing at y
full power conditions.
i 1.4 Operational Considerations:
EOL HTC Tech Spec LCO Value Relaxation of the SR limit value may provide only temporary relief from the i
repeated HTC measurement situation.
With. longer operating cycles and increased fuel discharge burnups, future reload core designs:may eventually challenge the EOL MTC LCO limit. The reload core design-process would.de'.ect-the fact that the design value of EOL HTC could exceed the' Tech Spec-LCO limit long before a reload core were to begin operation.
There are design measures that can be taken to produce a less negative E0L HTC but they negatively Impact full cycle energy and operability and may lead to more positive MTC values at :.he beginning of cycle.
The FSAR accident' analyses which-form the plant.'s licensing basis have assumed-a MDC value which, when converted to a MTC at~ full power pressure and temperature, translates to a HFP HTC value that'is more negative.than'the LC0 i.
limit value of the Technical Specifications.
The difference between the value of most negative MTC-(most-positive MDC): assumed in the accident' analyses and that presented as the LCO of the Tech Specs is substantial', and offers a potential avenue for relaxation of the. Tech Spec E0L HTC'LCO value.
The thrust of such an effort to relax the E0L MTC' Tech Spec LCO limit.must continue.to bound the accident analysis assumptions, and should establish a reasonable basis for the difference between the. safety analysis value of most-negat'ive MTC and the Tech Spec LCO value.
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2.0 METHODOLOGY-FOR MODIFYING MOST NEGATIVF. MTC TECH SPEC = VALUES:
j 2.1 Conversion of Safety Analysis MDC to Tech Spec HTC As stated previously, the FSAR accident analyses have assumed a-bounding value of the moderator. density coefficient (MDC)'which ensures a'conservativet result 1 for the-transient analyzed.
The_ process by'which the accident analysis' most-i positive MDC is transformed into the most negativetMTC LCO value is stated-in l
t STS BASES-section 3/4.1.1.3:
ll "The most negative MTC value equivalent to th'e most positive. moderator =
density coefficient (MDC),'was obtained by-incrementally correcting the MOC:
used in the FSAR accident analyses to nominal operating-conditions.
These corrections involved subtracting'the incremental-change:In the:MDC' associated with'a. condition of all rods. inserted'(most posttive MDC)-tc, an=
all rods withdrawn condition, and~a conversion for thejrate-of change of moderator density with temperature at RATED THERMAL P0HER conditions. This-value of the MDC was then transformed into the' limiting MTC val.ue."
In the process of converting the accident analysis-MDC.into-the Tech; Spec.MTC, I
the conversion for the rate of change of moderator density w,lth temperature.at j
rated thermal power conditions involves-conventional thermodynamic properties and imposes no undue conservatism on the resulting MTC value.
The additionali conversion made is to correct the above MDC (MTC) value for the' change
-associated with going from a condition of ARI to one,of ARO..That-Is, the-accident analysis MDC (MTC) assumes a coefficient determined for a condition of EOL HFP 0 ppm with all control and. shutdown banks-fully inserted...This-accident-analysis MDC (MTC) is corrected back to the ARO condition, in order.
to produce a Tech Sp'ec limit which permits direct comparison against measured l
values. -The effect of the presence of all control and shutdown banks is'to make-the MTC for the ARI condition markedly more negative _ than a MTC at -the
.AR0 condition,-hence this conversion has a sub;tantial-tmpact.-
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.The use of a substantially negative MTC (positive MDC) value'for the transient accident analyses is prudent, in that it produces a more severe result for the.
transient, which makes the analysis inherently conservative.
The drawback to the ARI assumption is that when the-' conversion to the ARO condition 1s made, the resulting Tech Spec MTC value is dramatically less negative than the value:
j corresponding to the transient _ safety calculations,'and is even less negative than the expected best estimate values of-EOL-MTC for'high discharge burnup
- reload cores.
In the worst case,_ maintaining'the E0L MTC Tech Spec limit at its present value could result in requiring-that the plant be placed in hot.
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shutdown when, in fact, there exists substantial margin to the moderator-coefficient assumed in the accident analyses.
Such a situation ts-Unnecessarily restrictive,'and results primarily from the ARI to ARO adjustment made between the accident analysis MDC value:and-the--Tech Spec MTC j
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In addition to being unnecessarily restrictive, the HFP ARI assumption is inconsistent with Tech Spec requirements for allowable operation, wherein.
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shutdown banks are not permitted to be inserted during power operation and control banks must be maintained above their insertion limits.
j 2.3 Alternative MTC Conversion Approach If the ARI to ARO basis for converting from the accident analysis MDC value to-a Tech Spec LC0 MTC value is overly restrictive, what would-constitute a more meaningful, yet inherently conservative basis? The concept herein proposed.as an alternative to the ARI to ARO conversion is termed'the "Most Negative Feasible MTC" approach.
This approach maintains the ext sting: accident analysis assumption of a bounding value of moderator coefficient, but offers an alternative method for converting to the Tech Spec LC0 MTC value.
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The Most Negative feasible HTC approach seeks to determine the conditions for which a~ core will. exhibit'the most negative MTC value that is consistent with' operation allowed by the Tech Specs. As an e;; ample, the Most Negative Feasible HTC approach would not require a conversion assumption that all rods be fully inserted at HFP conditions, but would require a conversion assumption that all control banks are inserted the maximum amount-that Tech Specs permit, i
so as to make the calculated EOL HFP MTC more negative than it would be for an i
unrodded core.
4 The Most Negative Feasible HTC approach determines EOL HTC sensitivity to those design and operational parame W s which directly-Impact MTC, and attempts to make this determination in a such a manner that the resulting sensitivity for one parameter is independent of the assumed values of the other parameters.
As a result, parameters which are mutually exclusive but permissible according to the Tech Specs (such as=an assumption of_ full power operation and an assumption of no xenon concentration in the core), and which serve to make HTC more negative, will have their incremental impacts on MTC combined to arrive at a conservative and bounding condition for the most negative feasible HTC.
The parameters'which are variable under normal operation, and which affect HTC are:
i soluble boron concentration in the coolant moderator temperature and pressure 1
l RCCA insertion axial flux (power) shape transient fission product (xenon) concentration l
The Most Negative Feasible HTC approach examines each parameter' separately, and assesses the impact of variation in that parameter on E0L MTC. The assessment is performed for multiple core designs that feature combinations of fuel design, discharge burnup, cycle length, and operating temperature expected to envelope future core designs.of the plant of interest.
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k When the assessment is complete, the HTC sensitivity associated with each of the j_
above parameters will have been identified. One then determines the maximum deviation from " nominal" conditions (ARO', HFP, equilibrium xenon, Tavg on the reference temperature program) that the Tech Specs permit, and multiplies that l
deviation by the appropriate HTC sensitivity to arrive at a " delta MTC" factor' associat:d with the parameter. _For example, suppose it is' determined that'the HTC becomes I pcm/*F more negative for each l'F increase-in' core average operating 1
temperature above nominal (the HTC " sensitivity" is -1 pcm/*F/'F).
If the Tech Specs permit a maximum increase in Tavg of 4*F above_ nominal core _Tavg,:then the moderator temperature " delta'HTC" factor 1s:-
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(-l pcm/*F/'F) x 4*F - -4 p m. 'F.
Bounding " delta HTC" factors are determined in this way for each of the above parameters, and these factors are then>added to* arrive at-an-overall bounding-1
" delta HTC" factor.
This overall " delta HTC" factor states how much more negative the HTC can become, relative to the nominal EOL HFP ARO HTC value, for: normal operation scenarios permitted by the current Tech' Specs.
The conditions.of.
moderator temperature, rod insertion, xenon, etc., which defined the Hott Negative-Feasible HTC condition become the conversion proposed as a replacement'for the'ARI to AR0 conversion of the current HTC Tech Spec.
The conve: sion for the 'Most Negative Feasible HTC condition is applled in_the same'way that the__ current ARI-to-ARO conversion is applied, in order to arrive at an E0L ARO HFP MTC Tech Spec limit which remains based on the accident analysis HDC-assumption.
i 2.4 Determining SR HTC from LCO HTC
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-Under the Host Negative Feasible HTC approach,.the 300 ppm surveillance value ls determined in the manner currently stated in-the BASES'for STS plant HTC Tech Specs:
"The HTC surveillance value represents a conservative.value (with-1 corrections-for burnup and soluble boron) at a core condition of 300. ppm equilibrium boron concentration and is obtained by making these corrections to the limiting HTC LCO value."
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That is, the 300 ppm surveillance value_is derived by making a-conservative l
adjustment to the EOL AR0 HFP MTC limit value that accounts for the change'to J
MTC with soluble boron and burnup.
Plant-specific examination of'the difference between 300 ppm HFP MTC and E0L (0 ppm) HFP MTC. suggests that a j
smaller correction is justified than the 9 pcm/*F whlih has historically been J
applied to Westinghouse-designed STS plants.
2.5 Benefits of the Alternative HTC Conversion Approach.
The Most Negative Feasible MTC approach is considered to be superior to.the l
ARI-to-ARO conversion specified by current STS plant-Tech Specs for the.
following reasons:
1.
The Host Negative' Feasible MTC approach 1results in a relaxed i
surveillance' limit _which significantly-reduces the probability of having to perform repeated MTC survelliance measurements. Such-repeated measurements are undesirable because'they entall perturbations to normal reactor operation and they are costly.
2.
The Most Negative Feasible HTC approach does not alter.the FSAR transient accident analysis bases or assumptions, and'hence, does not affect the accident analysis conclusions.
It retains the concept of-l a conversion between the accident analysis MDC assumption and the Tech Spec LC0 MTC value that assures that the plant cannot experience a MDC which is more severe than that assumed in the accident _ analyses.
3.
The Most Negative Feasible MTC-condition is a conservative but j
l reasonable basis to assume for a MTC value of the reload core prior to a transient, and is consistent with operation as defined by other sections of the Tech Specs (whereas the ARI-to-ARO conversion is overly conservative and makes assumptions which are inconsistent with other sections of the Tech Specs).
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l Additionally. the Most Negative Feasible HTC approach retains the "bulit-in safeguard" of a requirement for a 300' ppm surveillance measurement to be performed in order to verify that the reactor is operating in a regime that is bounded by the accident analysis. input assumptions.
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Westicghouse Class 3 3.0 HOST NEGATIVE FEASIBLE HTC APPROACH APPLIED TO J. H. FARLEY UNITS 1 AND 2
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e 3.1 J. H. Farley Units 1 and 2 Accident Analysis MDC Assumption The FSAR accident analyses upon which the Tech Spec EOL HFP LCO HTC limit is based have assumed bounding values of moderator density coefficient in order to ensure a conservative simulation of plant transient response for the J. H.
Farley Nuclear Plant uniti;.
For those transients for which analysis results are made more severe by assuming maximum moderator feedback, a moderator f
density coefficient (MDC) of 0.43 Ak/gm/cc has been assumed to exist throughout the transient.
When discussing the Tech Spec EOL LCO limi' on moderator feedback, it is simpler to talk in terms of moderator temperature coefficient (HTC) than HDC.
For this reason, the J. M. Farley accident analysis HDC assumption of 0.43 ak/gm/cc is converted to its equivalent HTC.
This conversion depends on the density change-to-temperature change relationship which prevails for the conditions of interest.
For this discussion, the conditions of interest are the core temperature and pressure (hence, density) experienced under normal operation at which the HDC assumes its most extreme (positive) value.
These temperature and pressure conditions are the J. H. Farley unit's full power, full flow nominal core operating conditions of 577.2'F and.2250 psia, respectively.
At these nominal HFP operating conditions, the accident analysis MDC value of 0.43 ok/gm/cc is equivalent to a HFP HTC sf -51 pcm/'F.
For simplicity, this value of HTC will often be referred to as the " accident analysis HTC", in the discussion which follows. However, it should be remembered that the applicable accident analyses actually assume a constant HDC value of 0.43 ok/gm/cc and make no explicit assumption aLaut HTC.
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3.2 Determination of Most Negative Feasible MTC Sensitivities F
As stated previously, there are a limited number of core operational parameters that directly affect MTC and are variable under normal core operation. The list of parameters is as follows.
soluble boron concentration in the coolant moderator temperature and pressure RCCA insertion axial flux (power) shape transient fission product (xenon) concentration I
The radial flux (power) shape can also vary under normal core operation and will affect MTC.- However, the operational activities that directly affect radial power shape do so through withdrawal or insertion of control rods and through xenon concentration; therefore, the impact of radial. flux distribution variation on MTC will be an impItcit part of the HTC sensitivity to these I-other parameters.
Soluble boron concentration is certainly variable under normal core operation.
However, it is eliminated as a source of sensitivity for this
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analysis.
This is because the EOL HFP ARO HTC Tech Spec Ilmit value is assumed to be essentially a 0 ppm limit by virtue of the definition of EOL.
The most negative MTC value will always occur at a boron concentration of 0 ppm, and therefore, a O ppm boron concentration is assumed as the basis of the E0L MTC Tech Spec limit under the Most Negative Feasible MTC approach.
For the remaining parcmeters, sensitivity analyses were performed by l
perturbing the paramet6r of interest in such a way as to induce a change from its nominal EOL value, a l then performing a HTC determination with the parameter held in the perts % state. A further perturbation was induced t.nd the MTC calculation repeated.
Yhts sequence was repeated until sufficient HfC data values were generated to reliably determine the trend of HTC change with variation in the value of the. independent parameter.
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InordertoestablishtrendsinHTCthatareappropriate(andboundingf"'C for the J. M. Farley unit's reload cores, these sensitivity calculations were performed for three different reload cores.
These cores exhibit design features that could be incorporated in future J. H. Farley reload cores or that produce HTC values inherently more negative than those expected for Farley reload cores (such as higher nominal operating temperatures).
A brief description of the three reload core designs follows:
i RELOAD A:
This core is the initial reload core for a Hestinghouse-designed 3 loop plant similar to the J. M. Farley units.
It utilizes the Hestinghouse 17x17 fuel design and operates at a nominal core average moderator temperature of 592.5'F.
The cycle length is t
350 effective full power days (EFPD).
The reload uses a low leakage loading pattern fuel placement arrangement.
The discharge region burnup is necessarily low, as it is the first l
reload.
The control rod absorber material is hafnlum.
RELOAD B:
This core is the actual core design for a recent reload cycle of J. H. Farley unit 1.
It utilizes the Westinghouse 17x17 fuel design and operates at a nominal core average moderator temperature of 577.2'F. The cycle length is 480 EFPD.
The reload i
uses, low leakage loading pattern fuel placement arrangement.
The anticipated region average discharge burnup of the feed fuel is just over 40,000 MHD/HTU.
The control rod absorber material is silver-indium-cadmium.
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RELOAD C:
This core is a reload core for a Hestinghouse-designed 3 loop plant similar to J. H. Farley untts. It utilizes the Hestinghouse 17x17 0FA fuel design and operates at a nominal core average moderator temperature of 590.5'F.
The cycle' length is 455 EFPD.
The reload uses a low leakage loading pattern fuel placement arrangement.
The discharge region average burnup is approximately 38,000 MHD/HTU.
The control rod absorber material is silver-indium-cadmium.
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i Reloads A and C feature the same control bank configuration that is used in J. M.
l Farley units (Reload B), shown in figure 3.1.
The core neutronic models of these three reloac' cores were derived using standard Westinghouse procedures and computer methods.
The ARK code, which has evolved from the LEOPARD (2) and CINDERI3I codes, was used to perform the fast and thermal spectrum calculations and is the basis for all cross sections, depletion rates, and reactivity feedback models.
ANC
, a nodal analysis theory code that is used in two and three dimensions, was used for core neutronic calculations to determine MTC sensitivity for the three reload cores. APOLLO, an advanced version PANDA (5), was used as an axial neutronic model of the reload cor to determine MTC sensitivity to varying axial flux shape.
The neutronic calculations performed for the three reload core designs established HTC sensitivities for each of the parameters listed above.
The detailed description and results of these calculations are provided in Appendix A.
3.3 Maximum Allowed Deviations from Nominal Operating Conditions The concept of maximum allowed deviation from nominal operating conditions is employed to determine the extent to which reactor parameters can vary under i
normal operation so as to cause MTC to become more negative.
This combination of parameter statepoints then defines the worst allowable initial condition for a transient which employs a most negative M1C (most positive MDC) assumption.
It is also necessary to demonstrate that the parameter changes that occur throughout the transient do not result in a MTC value which is unbounded by the constant moderator coefficient assumption used in the accident analysis. The adequacy of the constant MDC accident analysis t
assumption to bound HTC values that occur throughout the transient is examined in detall in Section 4.
1
Westinghouse Class 3 i
i The bases for the maximum allowed deviation from nominal operating conditions t
are Technical Specifications that limit the extent of moderator temperature increase, RCCA insertion, and axial power skewing.
The deviations permitted by present J. H. Farley Tech Specs and possible future perturbations to those Tech Specs values are discussed in the following sections:
Moderator Temperature and Pressure Deviations Tech Spec 3.2.5 estabitshes the LCO values of the DNB parameters reactor coolant system Tavg and pressurizer pressure.
This section of the J. H.
Farley Unit 1 Tech Specs is excerpted and presented in Appendix B.
This Tech Spec is identical between the two J. M. Farley units.
Tech Spec 3.2.5 states a minimum allowable pressurizer pressure of 2220 psla, ano i;,c ravimum allowable RCS average temperatJre of 581.2'F.
Because the current nominal operating RCS temperature for the J. H. Farley units is 575.0*F, the'581.2'F Tech Spec limit represents a 6.2*F maximum allowable Tavg increase over nominal conditions.
The current nominal design pressure for the J. H. Farley units is 2250 psla; therefore, the 2220 psia Tech Spec limit represents a 30 psi maximum reduction from nominal system pressure, i
Because Tech Spec 3.2.5 limits deviations from nominal condition RCS temperature and pressure to +6.2'F and -30 psi, respectively, it also indirectly places a limit on the maximum allowable deviation of RCS moderator density from nominal.
These maximum temperature and pressure deviations are applied to the MTC sensitivity to temperature and pressure, which is described in Appendix A, to obtain a " delta MTC" factor associated with RCS moderator temperature and pressure deviations from nominal.
The resulting " delta MTC" is (
J+"'C pcm/'F.
02711:6-881220 3-5
Hestinghouse Class 3 l
RCCA Insertion Deviation The nominal condition assumption for RCCA placement is complete withdrawal
[
(AR0).
This assumption is underscored by the requirement in Tech Spec 3.1.1.3 (see Appendix B) that the LCO value of EOL MTC is for the "all rods withdrawn" condition.
Because some RCCA insertion is allowed during full power operation, and because RCCA insertion will generally cause MTC to be more negative than it would be otherwise, the RCCA insertion deviation is simply that maximum allowable RCCA insertion permitted by the Tech Specs.
Tech Specs 3.1.3.5 and 3.1.3.6 place limits.on allowable RCCA insertion.
These two Tech Specs are excerpted from the J. M. Farley Unit 1 Tech Specs and are provided in Appendix B.
Although Appendix B contains only Unit 1 Tech Specs, the Unit 2' Tech Specs are identical.
Tech Spec 3.1.3.5 precludes.
Shutdown RCCA insertion in Modes I and 2, and Tech Spec 3.1.3.6 limits Control Bank Insertion via the Rod Insertion Limits (RILs).
l-Control rods can be inserted 7.6 a function of power level according to the l
RIls, and all RCCAs are inserted upon reactor trip.
With greater RCCA Insertion, MTC becomes mere negative relative to the ARO MTC, all other parameters being held r. qual.
However Tech Specs do not allow all other parameters to be held equal.
With deeper RCCA insertion, power must be-reduced and Tavg will be reduced accordingly. The reduction in Tavg serves to make the MTC more positive, and at E0L 0 ppm conditions, this positive Tavg effect will entirely offset the negative RCCA effect on MTC.
For example, for Reload A, complete insertion of both control Banks D and C at EOL 0 ppm HFP conditions (a condition not permissible under normal operation) will make the MTC [
pa,c pg,j.F more negative than the ARO MTC.
- However, in going from a nominal HFP Tavg to e nominal-HZP Tavg at EOL and 0 ppm, the l4 MTC for this same core becomes (
pa,c pcm/'F more positive.
This
[
f"'C pcm/'F-more positive compo1ent of the MTC that results from the moderator temperature (density) changt in going from HFP to HZP will also more than compensate the negative MTC component that arises from total RCCA insertion with trip.
l 0271 1:6-8812 2 0 3-6
Westinghouse Class 3 j
The Reload A core dtsign is typical of reloads for Westinghouse-designed PWRs in this respect.
Be:ause the rate at which decreasing moderator temperature makes MTC positive rxceeds the rate at which allowable RCCA insertion males HTC negative, the ii.ost negative MTC situation will always exist at HFP Tavg r
with RCCAs inserted to the extent allowed by the HFP insertion limits.
For this reason, the maximum RCCA deviation from nominal conditions a110wable by the Tech Specs only needs to be assessed at the HFP condition.
Figure 3.2 shows the RCCA insertion limits for both J. M. Farley unit's reload cores.
These are typical insertion limits for Hestinghouse-designed 3 loop PHRs with the eight control rod lead control bank of Figure 3.1. Figure 3.2 shows that at full power the lead control bank can be inserted to a depth of 187 steps withdrawn.
However, strict application of these current RIls in determining the " delta MTC" factor associated with RCCA insertion may prove to be restrictive if minor changes to the RIls become necessary in the future.
For this reason, the HFP RCCA Insertion assumed for this analysis is
[
]+a.c steps withdrawn.
This additional insertion is expected to bound minor RIL adjustment which may be necessary for optimizing core performance characteristics of future J. H. Farley reloads. [
)+R,C therefore, a RIL adjustment to lead control bank insertion beyond [
]+a,c steps withdrawn will not necessarily invalidate the revised EOL MTC LCO value.
This limiting HFP RCCA insertion of [
]+"'C steps withdrawn forms the basis of the determination of HTC sensitivity to HFP RCCA insertion, which is described in Appendix A.
The overall " delta MTC" factor associated with RCCA insertion is derived directly from this [
]4a,c steps withdrawn lead control bank ;.osition, and was determined to be [
]+"'C pcm/'F.
l l
l, Axial Flux (Power) Shace Deviation l
As indicated earlier, HTC is affected by the axial flux shape which exists in the core, primarily as a result of the influence which the axial flux shape has on the rate at which the moderator is heated as it moves up the core.
The detailed shape itself is not so important, but rather the " balance" of the 02711:6-881220 3-7
Westitghouse Class 3 i
flux shape, in terms of how much moderator heating occurs in the lower half of the core versus the upper half of the core.
The influence which axial power shape has on HTC can, therefore, be captured by quantifying this axial flux
" balance", and this balance is best quantified by the core's Axial Flux Difference (AFD).
The discussion of the MTC sensitivity to axial flux (power) shape presented 'e Appendix A establishes that the more n'egative the AFD becomes, the more negative the MTC will become.
The axial flux (power) shape deviation is, therefore, determined by how negative the AFD is allowed to become under normal full power operating conditions.
The J. M. Farley units presently employ a target AFD band of +5% to -5% about a target value of AFD.
Current and anticipated future J. M. Farley core designs produce target AFD values which are no more negative than -5%.
This
-5% target AFD, combined with a bank allowance of -5% means that the most negative HFP AFD value for J. M. Ftriey. reload cores allowed by Tech Specs would be -10%.
To assign a " delta MTC" factor attributable to axial flux shape, one need only examine the MTC effect associated with a -10% " deviation" from a " balanced" (i.e. 0% AFD) flux shape.
[
)+a,c, an AFD value of
(
]+8'C% is selected as the basis of the axial flux (power) shape deviation.
This (
)+8'C% AFD deviation is applied to the HTC sensitivity to axial flux (power) shape, which is described in Appendix A, to obtain an overall " delta MTC" factor associated with AFD deviation from a perfectly balanced axial flux shape.
The resulting " delta MTC" factor is [
]+a,c pcm/'F.
i Transient Fission Product (Xenon) Concentration Deviation Xenon is the most significant transient fission product in terms of effects on core reactivity and flux distribution; therefore, its possible impacts on HTC are investigated to. compute the final " delta MTC" factor to include in the Most Negative feasible MTC approach.
While Tech Specs place no limitations on either xenon distribution or overall concentration, the AFD limits discussed 02711:6-881220 3-8
Westinghouse Class 3 above, in effect, place a'llmitation on the amount of axial xenon skewing that
~
can occur, and the physics of xenon buildup and decay place practical limits on the concentration.
Because axial xenon distribution directly impacts axial 1
flux shape, this aspect of xenon effect on HTC is implicitly included in the axial flux (power) shape deviation discussed above.
What remains to be quantified is the impact of the overall xenon concentration in the core.
Taking the E0L HFP ARO equilibrium xenon concentration to be the nominal xenon condition for the core, it was determined for low leakage core designs of the type characteristic of J. H. Farley reloads, the HTC would become more negative with a reduced xenon concentration.
Accordingly, the most negative HTC results when there is no xenon in the core, It was established in the discussion on moderator temperature and pressure deviation and on RCCA insertion deviation, that the condition for the most negative HTC requires maximum allowable temperature (minimum allowable density) and, therefore, occurs at fu11 power conditions.
While the assumption of achieving full power operation with no xenon in the core is certainly a conse,vative assumption, the possibility of steady power escalation after an extended shutdown period presents a reasonable scenario for full power operation with a comparatively low xenon concentration in the core.
For this reason, the " xenon deviation" to be used in conservatively determining the " delta HTC" factor attributable to transient fission product is a change from HFP ARO equilibrium xenon to no xenon in the core.
The resulting " delta HTC" factor is [
1+a c pc,f.7, 3.4 Overall " Delta HTC" factor for J. H. Farley Units 1 and 2 Reloads 4
The preceding section has concluded that the most adverse operation possible, in terms of achieving tl.: most negative E0L HTC under current J. H.
Farley Units 1 and 2 Tech Specs, would feature the following values of key parameters:
- Core Moderator Temperature:
6.2*F above HFP nominal
- Core Moderator Pressure:
2220 psia
- IIFP RCCA Insertion:
4a,c steps withdrawn
- HFP most negative AO:
- 1. +0.C
- HFP xenon concentration:
~ 0 1.
02711 : 6-880' J8 3-9
14estinghouse Class 3 i
i When these maximum allowable deviations from a nominal condition of EOL HFP AR0, with equilibrium xenon, and 0 ppm boron are applied to the individual parameter sensitivities discusssd in Appendix A, the overall " delta HTC" factor is computed.
This overall factor for the J. H. Farley units was computed to be as follows:
+a,c Core Moderator Teroperature and Pressure Factor:
pg,f 7 HFP RCCA Insertion factor:
pcm/*F Axlal Flux (Power) Shape factor:
pcm/*F Xenon Concentration Factor:
ocm/*F Overall " Delta MTC" factor:
pcm/'F The interpretation of this overall " delta MTC" factor is as follows.
The Tech Spec LCO value of E0L HTC is based on the explicit conditions of unrodded full power operation.
This is an appropriate condition for performing a HTC experiment and obtaining results that can be meaningfully compared to design predictions.
It is not, however, the condition under which the HTC can achieve its most negative value under normal operation scenarios permitted by the Tech Specs.
The conservative " delta MTC" formulation has concluded that the actual core HTC can be as much as [
J+a,c pcm/*F more negative than i.
the EOL HTC LC0 value defined by the Tech Specs.
l The individual components of this [
]+a,cpcm/'F overall " delta MTC" factor have been determined on a conservative basis and are expected to bound the values predicted for J. H. Farley reload cores in the future.
While an individual component could conceivably exceed the value cited above, such an occurrance would not invalidate the Most Negative Feasible HTC approach, as long as the total of all the components remains bounded by the
[
3+a, cpg,j.F overall " delta MTC factor.
[
1
)+a.C 02711:6 881121 3-10
Westinghouse Class 3 3.5 Proposed J. M. Farley Units 1 and 2 Tech Spec EOL MTC LCO Value
\\.
As was pointed out in Section 3.1, J. M. Farley FSAR accident analyses have u,
assumed a MDC value which, when converted to a MTC at nominal HFP conditions, is equivalent to a MTC of -51 pcm/'F.
At no time may the actual core be allowed to experience a MTC more negative than -51 pcm/'F, as this would invalidate an assumption of the accident analyses.
The Most Negative Feasible MTC approach assures that such a situation will not occur by subtracting from this -51 pcm/'F MTC value the [
]+"'C pcm/*F " delta MTC" factor determined for the J. M. Farley units.
The resulting value,
[
]+a cpe,j.F, is proposed as the Tech Spec EOL LCO value of HTC under the Most Negative Feasible MTC approach.
As an additional measure of conservatism, this value is further increased to 83 pcm/'F, and proposed as the E0L HFP ARO Tech Spec HTC LCO value for J. M. Farley reload cores, replacing the current LCO value of ~39 pcm/*F.
The -43 pcm/'F proposed limit provides relief over the -39 pcm/'F limit associated with the current Tech Spec AR0-to-ARI conversion. requirement, yet
.tlli represents a conservative formulation.
The scenario of deep RCCA insertion,-coupled with high Tavg, low system pressure, and no xenon, represents a compounding of worst case events which can be considered independent, yet are not treated as such in the Most Negative Feasible MTC formulation. Determination that the core MTC is less negative than -43 pcm/*F at E0L HFP ARO conditions provides assurance that the assumption on initial condition MTC made in the plant accident analyses remains bounding. Additional assurance that the MTC (MDC) will not become more limiting at any time during a transient is also needed, in order to demonstrate that the accident analysis conclusions remain valid.
This additional assurance is the primary subject of Section 4.0.
?
02711:6-881220
- 11-
Westinghouse Class 3 i
y l
F;2URE 3.1 3-LOOP CONTROL ROD LOCATONS j
1 J
R P
N M
L K
J H
O P
E D
C S
A 1
3 A
D A
i 3
o C
3 S
C l
r s
i
{
8 A
3 D
C D
B A
7 8
D C
C
'O c
9 10 A
B D
C D-B A
11 13 C
3 B
C-13 14 A
D-A 15 FUNCTION NUMBER OF CLUSTERS Control Bank D-8 Control 5ank C 8
l l
Control Bank 8 8
Control Bank A 8
l
(
a 3-12 1
.---,.-_.-4.
. - -.. ~....
b
Westinghouse Class 3 L
FIGURE 3.2 I
k RCCA INSERTION L MITS FOR l
J. M. FARLEY UNITS 1 AND 2 RELOAD CORES i
e i
i 220 l
,r if i
200 P'
}
l A
180 s'
,/
i
/
n
}
' l oI60 r
u j
i
/
BANK C f
- 140
/
,/
f j
J
/
2
/
m 120
/
/
,f 0
7 r 100 if i
N l7 l
l oa-4 80
/~,/
L BANK D i f
y 1
I 8
60 e
a:
f l
I r
l
/
4 r
40 e
\\
r i
20
,/
/i I
/
I 6
p U 00 10
.20
.30 40
.50
.60
.70
.80
.90 1'0 RELATIVEPOWER(fractionof2652Wrt) l l
~
.W
Westinghouse Class 3 I
i 1
4.0 SAFETY ANALYSIS IMPACT OF MOST NEGATIVE FEASIBLE MTC APPROACil
)
[
The accident analyses conservatively model the various reactivity coefficients to produce a bounding analysis. As discussed in Section 3.1, the applicable analyses assume a constant HDC of 0.43 Ak/gm/cc to bound the predicted moderator reactivity insertion.
The events which assume this value for EOL HDC are listed in Table 4.1.
The Host Negative Feasible HTC approach determines the conditions for which a core will exhibit the most negative HTC value that is consistent with operation allowed by the Tech Specs.
Thus, the value for the Most Negative Feasible HTC provides the basis for a conservative initial condition assumption.
Changes-In the parameters identifled in Section 2.3 could take place during a transient in such a way as to make the MTC more negative than that allowed under normal oper; tion.
However, the most adverse conditions seen in these events will not result in a reactivity insertion which would invalidate the conclusions of the FSAR accident analyses.
Therefore, the 0.43 ok/gm/cc assumption used as the basis for the Host Negative Feasible HTC Tech Spec will not change.
As discussed in Reference 6, the reactivity coefficients assumed can have a strong influence on accident analysis results.
C i
i
]+"'C This process ensures the ability to verify that the applicable safety limits are met for each eload design and, consequently, that the Tech Specs are met.
02711:6-881220 4-1
' Westinghouse Class 3 ~
~
- ~ - - - - - - - - -
TABLE 4.1 FSAR Chapter 15 Events Which Assume A Constant 0.43Ak/gm/cc Value of HDC 15.2.2 Uncontrolled Rod Cluster Control Assembly Bank Hithdrawal at Power 15.2.6 Startup of an Inactive Reactor. Coolant Loop 15.2.7 Loss of External Electrical Load and/or Turbine Trip 15.'.10 Excessive Heat Removal Due to Feedwater System Halfunctions' l
15.2.11 Excessive load Increase Incident i
i I
1 L
i i
1 02711:6-881220 4-2
Westitghouse Class il
~~
j l
5.0 OETERMINATION Of MOST NEGATIVE FEASIBLE MTC SURVEILLANCE VALUE Section 1.3. pointed out the potential conservatism in the separation of 9 pcm/'F between the Tech Spec 300 ppm MTC SR value and the EOL HFP ARO MTC LCO value.
Typical 17x17 reload designs exhibit a predicted difference between the 300 ppm HFP design MTC and the EOL HFP ARO design MTC which is much less than 9 pcm/'F. However, in order to justify the use of a value which is smaller than 9 pcm/'F for a'given plant, the specific design prediction i
history of the plant must be examined.
Design predictions for recent J. M. Farley reload cores were reviewed in order i
to determine the largest difference'between the predicted 300 ppm HFP MTC and 4
the predicted E0L HFP MTC. The magnitude of the difference has shown little variation for these recent reload cores. The maximum difference determined was (
3+a,cpe,j.7, In reviewing the differences between predicted 300 ppm and EOL MTC values for the recent J. M. Farley reload cores, an important trend was discerned.
It i
was observed that the higher core average enrichments associated with increasing discharge burnup tend to decrease the magnitude of MTC difference.
As the J. M. Farley reload cores are expected to gradually increase fuel l
discharge burnup levels, it is anticipated that the difference between the HFP 300 ppm HTC and the HFP EOL MTC will become less in the future.
(
)+a c The proposed Tech Spec SR value for Farley reload cores is -36.5 pcm/_'F.
This value is 6.50 pcm/'F less negative than the EOL LC0 MTC value proposed in Section 3.5.
The 6.50 pcm/'F was chosen to conservatively (
3+a,c afford relief from the 9 pcm/'F difference applied by the current Tech Specs.
(
[
)+a,c
.i l
02711:6-881220 5-1 i
l 6.0 SUSPENSION Or MTC MEASUREMENTS BELOW 100 PPM As indicated earlier, major drawbacks of the EOL MTC surveillance measurements include the large volumes of water that must be processed (and accompanying large volume of waste water created) for measurements performed via the f
boration and dilution method, and the need to force the plant to deviate from nominal operation conditions in order to gather test data.
These problems become particularly acute when core boron concentrations reach low values, as a result of the increased time needed to achieve the required dilution.
For example, at an equilibrium boron concentration of 300 ppm, approximately 3,000 gallons of water are processed to dilute the roughly 20 ppm needed to balance the reactivity loss associated with the test temperature increase.
However, at an equilibrium boron concentration of 100 ppm, the corresponding water volume that must be processed to achieve this same 20 ppm dilution is 11,000 gallons.
l, Keeping the reactor in a " perturbed" condition (i.e. power / program temperature l
mismatch) is never desirable, hence the low boron concentration MTC testing l
should not be performed unless absolutely necessary to demonstrate compliance with the Tech Spec EOL HFP MTC LCO value.
This is the primary reason that acceptable results for the 300 ppm MTC surveillance preclude further testing.
However, even if the 300 ppm MTC test result fails to satisfy the Tech Spec SR value, there is some question as to whether further testing should be performed for the remainder of the operating cycle in every case.
It is proposed that the HFP MTC testing requirement be suspended for equilibrium boron concentrations below 100 ppm if a secondary surveillance criteria on the MTC is satisfied by a HFP MTC measurement performed at or below 100 ppm.
This secondary surveillance criteria value is -40 pcm/'F.
This value is set, with due consideration for MTC behavior with boron concentration reduction and fuel depletion, to ensure that the proposed EOL (essentially 0 ppm) HFP ARO LCO value of -43 pcm/*F will be met, even if no further measurement is conducted.
02711:6-881220-6-1
Westinghouse Class 3 Figure 6.1 shows the HFP equilibrium critical boron concentration and MTC behavior with burnup that is typical of Farley reload cores.
Note that the 300 ppm MTC.value shows sufficient margin to the proposed 300 ppm SR value of
-36.5 pcm/'F that repeated MTC measurements are not expected to be necessary.
l However, should the 300 ppm MTC test result be more negative than the proposed SR value, examination of the MTC behavior with further cycle depletion becomes important.
Figure 6.2 shows this same typical Farley MTC as a function of boron concentration, and also depicts a curve connecting the proposed 300 ppm SR value and the LOL (0 ppm) Tech Spec LCO value.
Note that the slope of the
" typical" MTC line is somewhat less steep than the line connecting the Tech Spec values.
This more gradual change of actual MTC behavior with further core depletion indicates that falling to meet the 300 ppm SR value will not I
necessarily result in failure to meet the LCO limit.
1he proposed secondary surveillance criteria value of -40 pcm/'F is plotted on Figure 6.2 at a boron concentration of 100 ppm. A line connecting the proposed 300 ppm SR value of -36.5 pcm/'F and the proposed secondary 1
surveillance value has a slope which is more characteristic of actual MTC behavior.
Projecting this line down to a boron concentration of 0 ppm shows that such MTC behavior assures margin to the EOL LCO limit of -43 pcm/'F.
Failure to nieet the 300 ppm SR value requires that the MTC measurement be performed every 14~EFPD.
This repeat measurement requirement would result in up to five MTC measurements being taken prior to reaching a 100 ppm equilibrium boron concentration. A final measurement performed at or below 100 ppm would serve to confirm that satisfying the LCO value is assured through satisfying the secondary surveillance value.
I Satisfying the secondary surveillance provides exemption from only the two MTC measurements that would otherwise occur during the final month of full power operation.
However, it is these last two measurements that would be most problematic to perform; therefore satisfying the secondary survelliance limit of -40 pcm/'F can preclude operation in a perturbed state at a time when such operation is least desirable.
02711:6-881220 6-2 1J
E-Westinghouse Class 3 FIGURE 6.1 HFP CRITICAL BORON CONCENTRATION AND MTC VERSUS BURNUP FOR FARLEY RELOAD 2 i,, i i i ; i i i i i i i i i i i i i i i i i i i i i ; ; i ; ; ; ; g i ; ; ; ; g 7 -30 1000 41
-24^
800 s
5
^
5 f
s
=
7600
- -282 5
!t
=
~
8 Q
U
(
g h
O 2
x g400
<- -32g m
i T
W U
N h
E O
i o
x 4
l
-- -36 i
200.
Tl $ I l l I Q l I I I I I l l I Q l I 9 9 I 9 1 I ! ! !!! !!!!
h000 10000 11000 12000 13000 14000 15000 16000 17000 1800j0 I
CYCLEBURt4UP(WWD/M1U)
Westinghouse Class 3 FCURE 6.2
.HFP MTC VERSUS CRITICAL BORON CONCENTRATION FOR FARLEY RELOAD
~
(
~20 I I I i i i iiil i i i i I I I I i l i i i I I I.I I l
- 8'"
\\
8 1
m k
c' N
z -32 o
a.
V 2W
~
A y
O
)
w 36 t
b g
I!
W 8
4
=
g-40 2
~
~
i
-l l 1 l l l I I l i l l i l l l l 1 l l I i l i l I l' I l-l
,p 300-250 200 150 100-50 0
HfPCRITlCALBORONCONCENTRATION(ppm) 6-4.
l WestiEghouse Class 3
7.0 CONCLUSION
S The present J. M. Farley unit's Technical Specificction values of -39 pcm/'F for the EOL HFP ARO MTC LCO and -30 pcm/'F for the 300 ppm HFP ARO SR conservatively reflect the FSAR accident analysis MDC assumption, but are considered to be overly restrictive by potentially requiring repeated deviation from nominal plant operation. An alternative adjustment procedure is proposed which is based on a conservative determination of the extent to which a nominal EOL HFP ARO MTC value can be made more negative under the most extreme values of certain operational parameters that are permitted by other Tech Specs.
This Most Negative Feasible MTC approach assumes that these largely independent extreme situations occur simultaneously, and in the worst caso, serve to make.
the EOL HFP MTC [
3+"'C pcm/'F more negative than it would be at nominal conditions.
When this value is subtracted from the MTC equivalent of the accident analysis assumed MDC value, the resulting MTC is [
3+8'I' pcm/'F.
The slightly more conservative value of -43 pcm/'F is, therefore, proposed as the EOL HFP MTC Tech Spec LCO limit under the Most Negative Feasible MTC i
approach.
Examination of the differences between the 300 ppm HFP equilibrium baron concentration MTC value and the EOL HFP MTC values concluded that a bounding expected difference between these two MTC values for J. M. Farley reload cores is -6.50 pcm/*F.
This difference is subtracted from the proposed -43 pcm/'F EOL HFP MTC Tech Spec limit to arrive at a proposed Tech Spec 300 ppm HFP MTC SR value of -36.5 pcm/'F.
It is concluded that the Tech Spec EOL HTC LCO and 300 ppm SR values proposed j
under the Most Negative Feasible MTC approach do not impact conclusions cf FSAR accMent analyses, because they do not affect the accident analysis i
assumptica on HDC.
In addition, the validity of the above-stated LCO and SR MTC values, as well as the plant's ability to comply with them. [
)+a,C 1
i 02711:6-881220 7-1 l
I
Westinghouse Class 3 Additional flexibility in the requirement for repeated EOL HFP MTC measurement
~
is proposed by permitting exclusion from the repeat measurement requirement for HFP equilibrium boron concentrations below 100 ppm.
Exclusion would be permitted if the first HFP MTC measurement taken after reaching 100 ppm HFP equilibrium boron concentration is less negative than -40 pcm/'F, as this provides assurance that the ultimate E0L (0 ppm) HFP ARO MTC value will not i
violate the LCO limit.
The new EOL MTC LCO and 300 ppm SR MTC values and the revised basis for adjustment overcome the problems inherent with the present version of Tech Spec 3/4.1.1.3, yet still afford protection.
Tech Spec 3/4.1.1.3 continues to require that surveillance be performed, so that any deviations between the operating core and design predictions that might threaten the validity of accident analysis assumptions can be detected, and continued surveillance and appropriate action undertaken.
i I
.e, 02711:6-881220 7-2
1 Westinghouse Class 3 i
REFERENCES 1.
Davidson, S.
L.,
Iorti, J. A., " Reference Core Report - 17xl? Optimized Fuel Assembly," WCAP-9500-A, May 1982.
2.
Barry, R. F., " LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094," HCAP-3269-26, September 1963.
3.
' England, T. R., ' CINDER - A One-Point Depletion and Fission Product Program," HAPD-TM-334, August 1962.
4.
Liu, Y.
S., et al., "ANC: A Hestinghouse Advanced Nodal Computer Code,"
HCAP-10965-P-A, September 1986.
5.
Barry, R. F., "The PANDA Code," HCAP-7048, April 1967.
6.
Davidson, S.
L., Kramer, H. R., ed., "Hestinghouse Reload Safety l
Evaluation Methodology," HCAP-9272-P-A, July 1985.
c l
i e.
s 02711:6-881220 R-1
.... Westinghouse Class 3..
i l-E i
i
{
l j
APPENDIX A DETERMINATION OF MOST NEGATIVE FEASIBLE f
MTC SENSITIVITIES t
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' 02711:6-881220 A-1
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Investigation of the sensitivity of MTC to core operational parameters that are variable under normal core operation is a fundamental requirement of the l
Most Negative Feasible MTC approach. Of the parameters discussed in Section 3.2, those that required detailed evaluation are:
- moderator temperature and pressure
- RCCA insertion
- axial flux (power) shape
- transient fission product (xenon) concentration For each of these parameters, the sensitivity analyses were performed by perturbing the parameter in such a way as to induce a change from its nominal EOL value, and then performing a MTC determination with.the parameter held in
{
the perturbed state. A further pert;rbation was induced and the HTC calculation repeated.
This sequence was repeated until sufficient data.was obtained to reliably determine the trend of MTC change with variation in the value of the parameter, A
In order to establish trends in MTC that are appropriate and bounding for the 4
N reload core type of interest, the sensitivity calculations were performed for three different reload cores.
These cores exhibit a spectrum of design features (such as cycle length, fuel lattice design, etc.) that permit the MTC sensitivity results to have broad application for Westinghouse-designed 17x17 l
j l
3-loop cores. A brief description of the three reload core designs follows:
RELOAD A:
This core is the initial reload core for a Westinghouse-designed 3 loop plant similar to the J. M. Farley units.
It utilizes the l
Westinghouse 17x17 fuel design and operates at a nominal core average moderator temperature of 592.5'F.
The cycle length is 350 effective full power days (EFPD).
The v-
+d uses a low l
leakage loading pattern fuel placement arrangu ent.
The discharge region burnup is necessarily low, as it is the first reload.
The control rod absorber material is hafnium.
02711:6-881220 A-2
WesttQghouse Class 3 RELOAD B:
This core is the actual core design for a recent reload cycle of J. M. Farley Unit 1. It utilizes the Westinghouse 17x17 fuel design
.. and operates at a nominal core average moderator temperature of 577.2'F. The cycle length is 480 EFPD.
The reload uses a low leakage loading pattern fuel placement arrangement.
The anticipated region average discharge'burnup of the feed fuel is approximately.40,000 MHD/MTV. This reload core design is generally representative of reload cores for both J. M. Farley units.
The control rod absorber material is silver-indium-cadmium.
RELOAD C:
This core is a reload core for a Westinghouse-designed 3 loop plant similar to the J. M. Farley units. It uttitzes the Westinghvuse 17x17 0FAU } fuel design and operates at a nominal core average moderator temperature of 590.5'F.
The cycle length is 455 EFPD.
The reload uses a low leakage loading pattern fuel placement arrangement.
The discharge region average burnup is approximately 38,000 MHD/MTV.
The control rod absorber material is l
silver-indium-cadmium.
As stated in Section 3.2, all of these core designs feature the control bank configuration that will be used in J. M. Farley reload cores, which is shown in Figure 3.1.
~
i The core neutronic models of these three reload cores were derived using standard Westinghouse design procedures and computer methods.
The ARK code, I
which evolved from the LEOPARD (2) and CINDER (3} codes, was used to perform the fast and thermal spectrum calculations and is the basis for all cross sections, depletion rates, and reactivity feedback models.
ANC(4', a nodal analysis theory code that is used in two and three dimensions, was used for core neutronic calculations to determine MTC sensitivity for the three reload cores.
APOLLO, an advanced version of PANDA (5), was used as an axial neutronic model of the reload cores to determine MTC sensitivity to varying axial flux shape.
1 02711:6-881220 A-3
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Westinghouse'C1 ass'3.~~
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The neutronic calculations performed for the three reload core designs established MTC sensitivities for each of the parameters listed above.
The j
sections wt)Jeh follow provide detatis of the calculations performed and the MTC sensitivity results obtained.
,a A.1 HTC Sensitivity to Moderator Temperature and Pressure Variation The decrease in moderator density which accompanies moderator heatup has the effect of reducing neutron moderation. With a low soluble boron concentration in the moderator, this results in a more negative' moderator temperature coefficient. An increase in coolant temperature, keeping density constant, leads to a hardened neutron spectrum and results in an increase in resonance absorption in U238 Pu240, and other isotopes.
The hardened spectrum also causes a decrease in the fission-to-capture ratio in U235 and Pu239.
Both of these effects make the MTC more negative.
In addition, the hardened neutron spectrum results in a larger fast-to-thermal flux ratio which increases the leakage of the core.
Again, the effect of higher leakage is to make the MTC more negative.
Since water density changes more rapidly with increasing temperature, and because of the spectrum hardening effects mentioned above, the MTC becomes progressively more negative with increasing temperature.
The sensitivity of HTC to increasing temperature was determined for each of the three reioad r
cores by increasing core reference moderator temperature slightly above the -
nominal HFP value, while holding pressure constant at 2250 psia, and then performing a MTC calculation that induced small changes in core'K-effective via changes in moderator temperature and density about the reference values.
The effects of changes in moderator temperature and density were considered together.
Af ter the MTC value was computed, core reference mod'erator
(
temperature was further increased, and another MTC calculatten performed. This process was repeated until the trend of MTC with increasing core roference moderator temperature was clearly established.
02711:6-881220 A l
Westinghouse Class 3 Results were recorded for the three reload cores in terhis of change in MTC from the nominal HFP MTC as a function of increase in reference moderator temperature.above the nominal HFP moderator temperature.
The results are shown in Figure A.I.
As expected, Reload A exhibits the strongest sensitivity of MTC to increases in moderator temperature, due to its higher nominal HFP reference temperature.
(
)+a,C To use the (
)+"'C MTC sensitivity information of Figure A.1, the maximum allowable temperature and pressure (and, therefore, density) deviations permissible under operation that complies with Tech Specs must also be determined.
These deviation values, presented in Section 3.3, are combined with the sensitivity data to arrive at a " delta MTC" factor associated with moderator temperature and pressure (and, therefore, density).
For the 6.2'F temperature deviation cited in Section 3.3, figure A.1 indicates-that the corresponding " delta MTC" due to temperature increase is (
3+a cpe,f.F.
Se calculations which determined the.MTC sensitivity to increasing moderator temperature were performed at a constant RCS presure of 2250 psia, and hence, do not reflect the effect of the 30 psi pressure deviation cited in Section 3.3.
The moderator density perturbation caused by a pressure decrease of 30 psi was calculated, and the effect of this density change on MTC was determined.
The " delta MTC" due to the moderator density change associated with the 30 psi pressure deviation was conservatively determined not to exceed
[
]+a cpe,j.F.
The combined pressure and temperature " delta MTC" factor is, therefore, (
)"'Cpcm/'F.
A.2 MTC Sensitivity to RCCA Insertion With constant moderator temperature, pressure, and boron concentration, insertion of control rods makes MTC more negative.
This trend in MTC arises from three effects.
The first is that RCCA insertion makes the overall flux-spectrum slightly harder, which makes MTC more negative, as discussed in Section A.I.
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02711:6-881220 A-5 l
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Westinghause Class'3.
-The second effect is that RCCA insertion will increase core leakage, which again makes MTC more negative.
The third effect arises from the impact of
~
RCCA insertion on. axial flux (power)-shape, and this effect is treated-separately in Section A.3.
Control rods can be inserted as a function of power level according to the'-
RCCA insertion limits (RIls), and all RCCAs are inserted upon reactor trip.
With greater RCCA insertion, MTC becomes more negative relative to the ARO MTC, all other parameters being held equal.
However, Tech Specs do not allow all other parameters to be held equal.
With deeper RCCA insertion', power must' be reduced and Tavg_will be reduced accordingly.. The. reduction.in Tavg serves to make the MTC more positive, and at E0L 0 ppm conditions, this positive Tavg.
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effect will enti_ rely offset the negative RCCA effect on MTC,'The overall l
result is that-the most negative MTC that can exist-in'the: core' occurs 't HFP;-
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therefore,'the MTC Sensitivity to RCCA insertion need only be determined at HFP conditions for HFP allowed RCCA insertion.
To calculate the E0L HFP MTC sensitivity to RCCA. insertion,~each of'the three reload core models had the lead control bank inserted the maximum applicable.
amount determined from Section 3.3 ((
pa,c steps withdrawn), at HFP, with 4
no soluble boron in the core.
The MTC value '
this condition was then-I
~
determined by inducing small changes'in core K-effective via changes in moderator' temperature and density about their reference values.
This HTC value was compared to the MTC determined at the.same conditions, but with all RCCAs removed from the core.
]
Of the three reload cores analyzed, it was determined that the maximum-change.
to the E0Lt0 ppm HFP ARO MTC which occurred as a result of HFP RCCA insertion' l
to a depth of [
pa c steps withdrawn was
[-.
pa,cpcm/*F.
Because I
~
this change to MTC arising from RCCA, insertion was determined at HFP
- I equilibrium conditions, it is appropriate to factor in any further~effect on MTC that may-arise from increasing RCCA worth, at a' fixed insertion, as_a-result of an expected operational transient.
It was. determined that the MTC
~
would be made, at most, [
. f"'C pcm/*F more negative as a result of HFP RCCA worth being " enhanced" by' transient operation, therefore, the
[
pa.c'"deltaMTC" factor.associatedwithallowableHFPRCCA insertion becunes [
7" ' C pcm/*F.
l 02711:6-881220 A-6 n
Westinghouse Class 3-a l..
A.3 Sensitivity to Axial Flux-(Power) Shape 1
l.*
MTC is not.so much directly affected by axial flux distribution itself, but:is affected via the impact which the axial flux distribution has on the rate at L
which the' moderator is heated as it moves up the core, and'via the importance weighting which the axial flux shape imparts to different regions of the core, In general, the accumulated burnup in-the bottom half of the' core exceeds that o
in the top half of the core, as indicated in Figure A.2 for_E0L of Reload A..
Other things being equal, higher burnup results in a more_ negative MTC as a result of. Isotopic impacts on flux spectrum.
A more negative axial flux (power) shape allocates a greater "importance weighting" to.the lower regions of the core where burnups are greater,.thereby. accentuating thi_s effect.
A greater effect is the impact which axial flux (power) shape has on heating rate of the moderator as a function of axial elevation.
Figure A.3 shows, in-L*
the top diagram, three distinct axial power shapes - one which is skewed
. ~
- toward the bottom of the core, one which is skewed toward thettop of tha core, and one which is balanced, with an axial offset near zero.
The lower diagram J
in Figure A.3 shows the core moderator temperature.as a function of core height for these three different axial power distributions.
While the same.
temperature rise through the core occurs for.all three power-shapes, it;is-evident that a more bottom-skewed axial power distribution will give rise to a higher average moderator temperature. This results from the greater heating i
of the moderator in the lower core elevations-for the bottom-skewed case.
As energy is added to the moderator at higher elevations, the temperature still remains highest for the bottom-skewed power case.because of'its initial " head 1
start" in the lower elevations. -The temperature differences gradually-decrease as a result of the differing heating rates occurring in the upper core regions among the three shapes.
3 Both the importance weighting effect and the moderator axial; heating rate' effect indicate that a more bottom-skewed flux shape ~ hore negative Axial Flux Difference) w111' result in more negative MTC.
This effect was investigated f
for'the three reload cores at E0L HFP O ppm conditions with no. xenon in the-core (xenon was removed so as to not complicate flux. skewing strategy).
A i
A-7 i
02711:6-881220
i Hestinghouse Class 3 1
specific axial flux shape was induced and then,-holding this flux shape approximately constant, the MTC was determined by observing the small changes in core K-effective which resulted from variation in moderator tempe"ature and density about their reference values. A different axial flux shape was induced, and the MTC calculation repeated.
This process was repeated until the behavior of MTC with variation in axial flux shape (as quantified by Axial Flux Difference) was clearly _ identified.
Curves of " delta MTC" as a function of Axial Flux Difference (AFD) for the three reload cores are shown in Figure A.4.
Note that a zero AFD is taken as the reference point, therefore, " delta MTC" is fixed at zero for an AFD of Because more negative AFD values result from RCCA insertion, this axial zero.
flux shape.MTC sensitivity implicitly captures part of the RCCA MTC sensitivity not included in the " delta MTC" factor of the previous section.
Section 3.3 concludes that a negative value of HFP AFD [
]+"'C Using the most conservative trend of Figure A.4 (that of Reload A) the " delta MTC" factor corresponding to [
]**'C% AFD is [
3+"'Cpcm/*F.
i Figure A.3 indicates that for a markedly negative AFD, the core average moderator temperature could be as much as [ )+8'C'F higher than that seen for a core with a balanced axial power. shape (AFD near.0).
Recalling the HTC sensitivity to moderator temperature of Figure A.1, one would expect a much 1
greater HTC sensitivity to AFD-than is indicated by Figure A.4. While the volume-weighted moderator temperature for a very. negative AFD may increase significantly above that of the balanced. flux shape case, the power-weighted moderator temperature increase will be very modest, and this will result in rather weak MIC sensitivity to AFD.
-To further illustrate this point, examination of Figure A.3 shows that the very negative <AFD power shape imparts a significant "importance weighting" to the bottom portion of the core where moderator temperature is lowest, but in the top portion of the core, where moderator temperature is greatest, the relative importance weighting is low.
This power "importance weighting" aspect serves to negate'a great deal of the " volume-weighted" temperature 02711:6-881220 A-8
~~
Nestinghouse C1 ass 3 effect described above, and makes the " effective" moderator temperature j
increase for very bottom-skewed power shape rather small'.
Again, this causes the MTC sensitivity to extremes of flux (power) shape to be rather weak.
A.4 Sensitivity to Transient Fission Product (Xenon)-Concentration Xenon is the most significant transient fission product in terms of effects on core reactivity and flux distribution, therefore, its possible impacts on MTC were investigated to compute the final " delta MTC" factor to include in the Most Negative Feasible MTC approach. While Tech Specs place no limitations on l
l either xenon distribution or overall concentration, the AFD limits discussed in Section 3.3, in effect, place a limitation on the amount of axial xenon skewing that san occur, and the physics of xenon buildup and decay place.
practical limits on the concentration.
The effects of other fission products, namely Samarium, were not investigated.
Their effect on MTC.is considered-negligible due to the large decay times (in comparison to the affected transient events) of precursor fission products, and also, the relatively lower concentrations that result.
m.
Because axial xenon distribution directly impacts axial flux shape, this aspect of xenon effect on MTC is implicitly included in the Axial Flux Shape
" delta MTC" factor discussed in Sectic 4.3.
What remains to be determined is the sensitivity to overall xenon concentration-in the core.
Calculations to-determine this sensitivity were performed with the'ANCI4) code,Ltaking the E0L HFP ARO O ppm MTC value with an equilibrium concentration of xenon as'the.
reference value of MTC. A number of differing xenon concentration scenarios were modeled, and the MTC value associated with each scenario was determined.
For all three reload cores, the most negative MTC resulted when all xenon was removed from the core. The largest " delta" from the reference (equilibrium xenon) HTC that occurred when all xenon was removed was [
pa,cpcm/'F.
This value becomes the final " delta MTC"' factor attributable to xenon.
No further uncertainty is added, simply because the scenario of operating at full power with no xenon in the core is itself sufficiently conservative as to be bounding.
I 02711:6-881220 A-9
Westinghouse Class 3- -.
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FIGURE - A.1
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CHANGE IN MTC WITH INCREASE IN T-AVERAGE.
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AsoVE NOMINAL T-AVERAGE:
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INCREASE IN T-AVG ABOVE HFP NOMINAL' t
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FIGURE A 2 -
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CORE AVERAGE' AXIAL BURNUP VERSUS CORE.NEl0HT AT EOLL L
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AXIAL POWER AND WODERATOR TEMPERATURE VERSUS CORE HEIGHT-RELATIVE AXtAL POWER VERSUS CORE HEIGHT,.
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WODERATOR-TEMPERATURE VERSUS CORE HEIGHT s
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Westinghouse Class 3 FCURE A.4 DELTA MTC VERSUS' AXIAL FLUX DIFFERENCE AT EOL,.HFP,' ARO-7,
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o to-20 30 AXlAL FLUX-DlFFERENCEL(PERCENT)-
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A-13
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-APPENDIX B i
I a
.3 1
REFERENCED TECHNIC L SPECIFICATIONS' AND BA' ES', SECTIONS FROM-S 1
THE J. M. FARLEY NUCLEAR: PLANT. TECHNICAL SPECIFICATI0,4S.
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. 02711:6-881220-
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I REACTIVITY CONTROL-SYSTEMS
.lODERATOR TEIPERATURE COEFFICIENT L
LIMITlHG C0llDIT10N FOR OPERAT10ll
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L.,
3.1.1.3 The moderator. temperature coefficient (MTC) shall be:
Less than or (.1ual to 0.5 x 10-4 delta k/k/*F for the all rods l
a.
L withdrawn, beginning of cycle 1ife (BOL), belo.< 70% THERMAL POWER condition.
Less than or equal to 0. delta k/k/*F at or above 70i THERMAL POWER.
I t
b.
Less negative than -3.9 x 10-4 delta k/k/*F for the all rods withdrawn..
l end of cycle life (EOL), RATED THERMAL POWER condition.
APPLICABILITY:
Specification 3.1.1.3.a - MODES 1 and 2* only#
Specification 3.1.1.3.b;- MODES 1, 2 and 3 onlyf ACTION:
a.
With the MTC more positive than the limit of. 3.1.1.3.a above, operation i
in 1100ES 1 and 2 may proceed provided:
1.
Control rod withdrawal limits'are established and maintained sufficient to restore the MTC to within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
These withdrawal-limits shall be in addition to the insertion limits of l,
Speci fica tio n '3.1.3.6.
2.
The control rods are maintained within the withdrawal limits.
~
~
established above until a subsequent calculation verifies' that the HTC has been restored to within -its limit for' the all rods withdrawn condition.
t
- 3. - A Special Report is prepared and. submitted to the Connission j'
l' pursuant to Specification 6.9.2 within 10. days, describing the value of the measured MTC, the interim control rod withdrawal.
limits and the predicted average core burnup necessary-for restoring the positive MTC to within its limit for the all' rods.
withdrawn condition.
b.
With the itTC more negative than the limit' of 3.1.1.3.b.above,. be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
L
'With K rr greater than or equal to 1.0 e
- See Special Test Exception 3.10.3 1*
L F ARLEY-UlllT 1 3/41-4 AMEllDMElli 110.57
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- REACTIVITY CONTROL SYSTE g f
SURVEILLANCE REQUIREMENTO l
4.1.1.3 The NTC shall be determined to be within its limits during each fuel cycle as follows:
l a.
The MTC shall be measured and compared to the BOL limit of Specification 3.1.1.3.a.above, prior to initial operation above 5%
l of RATED THERMAL POWER, after each fuel loading.
TheMTCsbg11bemeasuredatanyTHERMALPOWERandcomparedto b.
l'
-3.0-x 10 delta k/k/'F (all= rods withdrawn, RATED THERMAL POWER -
condition) within 7 EFPD after reaching an equilibrium boron concen-l tration of 300 ppa.. In the.eveD$ this comparison indicates the NTC is more negative than'-3.0 x 10 delta k/k/*F, the NTC shall be remeasured, and compared to the EOL MTC limit of specification 3.1.1.3.b. at least once per 14 EFPD during the remainder of the fuel-cycle.
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FARLEY-UNIT 1 3/4 1-5
-AMENDMENT NO. 26 e
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a 3/4:1 REACTIVITY Colml0L SYaius
~
BASES 3/4.1.1.3 MDDERATOR TDtPERATURE COEFFICIER The limitations on moderator temperature coefficient ( E C) are provided l
to ensure that the value of this coefficient rasains within the limiting condition assumed in the'FSAR accident and transient analyses.
l The MTC values of this specification are applicable to a specific. set of-plant conditions; accordingly, verification of MTC values at conditions other.
than those explicitly stated will require extrapolation to those conditions in l
order to permit an accurate comparison.
3 The most negative MTC value equivalent to the most positive moderator.
density coefficient (EC), was obtained by incrementally correcting the MC used in the FSAR analyses to nominal operating conditions.
These corrections ~
i involved subtracting the incremental change in the MDC _ associated with a core condition of all rods inserted (most positive MDC) to an all Pods withdrawn.
condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.
This value of the MC was then
~
transformed into the limiting EC value -3.9 x 10*k elta k/k/*F.
The EC d
value of ~3.0 x 10'4 delta k/k/*F represents a conservative value-(with_
i corrections fo-burnup and soluble boren) at'a core condition of 300 ppa l
equilibrium boron concentration and is obtained;by making these corrections to l
]f the 11alting MTC value of -3.9 x 10'4 Ak/k/'F.
The surveillance requirements for seasurement of the MC at the beginning.
and near the end of the fuel cycle are adequate to confirm that the-EC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
a 1
1 e:I
[
j
t n L
b' POWER OISTRIBUTION LIMITS f
DNS PARAMETERS
.I 1 ' ' '
- i LIMITING CGiiDITION FOR OPERATION 1
4 3.2.5 The followinti 0t88 related parameters shall be maintained within the limits shown en Tabte 3.2-1:
Reactor Coo. ant System T,,,
1 l
a.
t 1
b.
Pressurizer Pressure c.
Reactor Coolant System Total Flow Rate APPLICABILITY:
MODE I e
ACTION:.
L With any of the above parameters exceeding its limit, restore.the parameter to l.'
within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER-to less than 55 of I
RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i l:.,,'.
' SURVEILLANCE REQUIREMENTS 3
-'l 4.2.5.1 Each of the parameters of Table 3.2 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.2.5.2 The Reactor Coolant' System total flow. rate shall be detemined to be within its limits by measurement at least once per 18 months..
- -_{
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=
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TABLE 3.2-1 n
p DIS PAR M TERS m
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5' LIMITS
-4 i
H.
PARAETER
-3 Loops in Operation 2 Leops-in Operation e
- < 581.2*F
(**)-
avg t
Pressurizer Pressure
. > 2220 psia *
(**)-
> 265,500 gym
~
- (**).
-l Reactor' Coolant-System Total Flow Rate' w
N
'- l A
N s
w
.h i
A b
i 5
y m
2
~~
.E m
z
- -4
~ ~
-n-3
" Limit not applicable.'during eltier alTHERMAL POWER ramp'in excess of SE of RATED TIENEL m
POWER' per. minute' or' a. THERMAL-POWER ~stap in' excess of 10E of RATED THElWEL poler. -
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REACTIVITY CONTROL SYSTEMS
. SHUTDOWN R00 INSERTION LIMIT e
LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.:
APPLICA8ILITY:
MODES 18 and 288 8E.I.9.!!:
With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:
a.
Fully withdraw the rod,-or-b.
Declare the rod to be inoperable and apply Specification 3.1.3.1.
I
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SURVEILLANCE REQUIRENENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:
a.
Within 15 minutes prior to withdrawal _of;.any rods.in control banks A, B, C or D during an approach to reactor critical.ity, and b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
8'See Special Test Exceptions 3.10.2 and 3.10.3.
fWith Kg f greater than or equal to 1.0 4
4 w
e FARLEY-UNIT 1 3/4 1-20 AMENDMENT NO. 26
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REACTIVITY CONTROL SYSTEMS
-CONTROL R00 INSERTION LINITS a
LINITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures 3.1-1 and 3.1-2.
APPLICABILITY:. MDDES 18 and 2*f.
ACTION:
]
.With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant-to Specification 4.1.3.1.2,- either:
a.-
Restore the control banks:to within the limits within two hours, or 3
b.
Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position.
using the above figures, or c.
.Be in at least H0T STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
(,
' SURVEILLANCE REQUIREMENTS 4.1.3.6.The position of each control bank shall be determined to be within the insertion limits at least once per.12-hours except during time intervals 1'
when the Rod Insertion Limit Monitor is inoperable.- then verify.the individual rod positions at 'least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. -
l i
i i
"See Special Test Exceptions 3.10.2 and 3.10.3 fWith K,ff greater than or equal to 1.0.
s
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FARLEY-UNIT 1 3/4 1 AMENDMENT NO. 26
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