ML111750673
| ML111750673 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 11/23/1981 |
| From: | Mathews E Wisconsin Public Service Corp |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111750674 | List: |
| References | |
| K-81-188, NUDOCS 8111270202 | |
| Download: ML111750673 (30) | |
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Division of% Licensing SUBJE:.CT:: Application. to; amend Licensell DPR-43. consisting, of, prooosed, APL-nd 43-to Techf S pecs' 3,10' t o, coivformt toi requiremj'ts, of PDC6111 to: assvrel c o~i4lancei w/tore. power,, diSti utl on, 1 i*0itvs
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'WISCONSIN PUBLIC SERVICE CORPORATION P.O. Box 1200, Green Bay, Wisconsin 54305 November 23, 1981 Mr. D. G. Eisenhut, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington,.D. C. 20555 Gentlemen:
Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Proposed Amendment 48 to the KNPP Technical Specifications
Reference:
Letter from E. R. Mathews to S. A. Varga dated November 16, 1981 transmitting a preliminary copy of proposed amendment 48.
On November 16, 1981, we transmitted a preliminary copy of proposed amendment 48 to the Kewaunee Technical Specifications to facilitate the Staff's review of our request. That letter described the major proposed changes to our technical specifications and provided a technical justification for same.
The changes identified at that.time were a major rewrite of section 3.10 to conform to the requirements of PDC-II which is a surveillance methodology developed by Exxon Nuclear Company to assure compliance with core power distribution limits, and a revision in allowable rod misalignment limits.
Other, less significant changes are also proposed; these changes are described below.
The figures associated with section 3.10 have been "cleaned up". Specifically, Figure 3.10-3, which was cycle 4 specific, has been deleted, and Figures 3.10-4, 5, 6 and 7 have been renumbered to reflect this deletion. The reference to Figure 3.10-3 in specification 3.10.d.2 has been deleted. A new figure, 3.10-7 is added which graphically describes the V(Z) function referred to in various locations in specification 3.10. Figure 3.10-4 (formerly Figure 3.10-5) has been revised to more accurately reflect our experience at Kewaunee Nuclear Power Plant. Figure 3.10-5 (formerly Figure 3.10-6) has been revised to be consistent with the PDC-II methodology. Finally, Figure 3.10-2, which portrays the K(Z) function, has been revised to incorporate the change from amendment 36 to our technical specifications. Amendment 36 revised the FQ limits; since K(Z) is a normalized function which is dependent on the FQ limit, it is appropriate to revise the K(Z) function when the FQ limit is revised.
8111270202 811123 46 PDR ADOCK 05000305 P
PDR I
Mr. D. G. Eisenhut 0
0 November 23, 1981 Page 2 The K(Z) function is a function of FQ (normalized) versus core height (feet; 0 feet " bottom of core and 12 feet = top).
It is made up of three line segments; the first line segment is parallel to the abscissa and extends from 0 feet to six feet. The value of this line prior to normalization corresponds to the license FQ limit. The second line segment is a linearly decreasing function determined from the large break LOCA analysis and extends from the 6 foot point to the intersection of the third line segment. The second line segment moves parallel to the line derived in the original LOCA analysis and its specific value is dependent upon the license limit. The third line segment is determined from the small break LOCA analysis.
We have determined the normalized K(Z) function proposed in this amendment from the FQ limit of 2.21 utilizing the method discussed above. The limit of 2.21 is used since it is the more conservative value of the two limits in our technical specifications corresponding to fuel manufactured by Exxon Nuclear Company. In actuality, it is technically permissible to use two K(Z) functions corresponding to the appropriate limits for Westinghouse and Exxon fuel. We have chosen to use the more conservative K(Z) function.
Another change in this proposed amendment is the deletion of references to part length rod cluster control assemblies (RCCA's). The part-length RCCA's are no longer used and have been removed. Specifications 3.10.e and 3.10.f.3 are affected by this change.
The bases to section 3.10 have been reorganized and reworded, where necessary.
The reorganization was done to allow easier identification of the sections of the bases and involved changes on every page. Only the wording changes have been marked with change bars.
This concludes the description of changes to the technical specifications in Proposed Amendment 48. The pages affected by this proposed amendment are:
TS 3.10-1 TS 3.10-9 TS 3.10-18 TS 3.10-2 TS 3.10-10 TS 3.10-19 TS 3.10-3 TS 3.10-11 Figure TS 3.10-2 TS 3.10-4 TS 3.10-12 Figure TS 3.10-3 TS 3.10-4a TS 3.10-13 Figure TS 3.10-4 TS 3.10-5 TS 3.10-14 Figure TS 3.10-5 TS 3.10-6 TS 3.10-15 Figure TS 3.10-6 TS 3.10-6a TS 3.10-16 Figure TS 3.10-7 TS 3.10-8 TS 3.10-17 In accordance with the requirements of 10 CFR 50, this package includes three signed originals of the cover letter, and 40 copies of proposed amendment 48 to the Kewaunee Nuclear Power Plant technical specifications. Additionally, in accordance with the requirements of 10 CFR 170.22, a check in the amount of
$4,000.00 is enclosed for the amendment fee for a Class III amendment. While we recognize that there are two substantiveissues involved, namely FQ surveillance methodology and rod misalignment limitations, we feel that this specific request for an amendment on rod misalignment is an extension of our earlier request made in proposed amendment 46 (submitted August 21, 1981).
Therefore, this amendment involves a single new issue (FQ methodology) for which the staff has clearly identified an acceptable position as evidenced by your review and approval of
Mr. D. G. Eisenhut November 23, 1981 Page 3 the PDC-II methodology. Consequently, this should be considered a Class III amendment.
Finally, we ask that upon approval and issuance of this amendment, a thirty day time period be allowed prior to its becoming effective in order to write and test the necessary surveillance procedures.
This time is needed because of the extensive revisions that are required in order to change to the PDC-II methodology.
Very truly yours, E. R. Mathews Senior Vice President Power Supply & Engineering jac Enc.
cc -
Mr.
Robert Nelson, NRC Resident Inspector RR #1, Box 999, Kewaunee, WI 54216 Subscribed and Sworn to Before Me This 23rd Day of November
, 1981 Notary ic, State of Wisconsin My Commission Expires
3.10 CO?* L ROD AND POWER DISTRIBUTION I(
TS Applicability Applies to the limits on core fission power distributions and to the limits on control rod operations.
Objective To ensure 1) core subcriticality after reactor trip, 2) acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation transients associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.
Specification
- a. Shutdown Reactivity When the reactor is subcritical prior to reactor startup, the hot shutdown margin shall be at least that shown in Figure TS 3.10-1.
Shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions if all control rods were tripped, assuming that the highest worth control rod remained fully withdrawn, and assuming no changes in xenon, boron, or part length rod position.
- b. Power Distribution Limits
- 1. At all times, except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:
A. F N(Z) Limits:
Q (i)
Westinghouse Electric Corporation Fuel F (Z) x 1.03 x 1.05 < (2.22/P) x K(Z) for P >.5 '
Q48 F (Z) x 1.03 x 1.05 _ (4.44) x K(Z) for P
<.5 Q_
(ii)
Exxon Nuclear Company Fuel F (Z) x 1.03 x 1.05 F (Ej)/P x K(Z) for P >
.5 F (Z) x 1.03 x 1.05 < (4.42) x K(Z) for P <
.5 TS 3.10-1 Proposed Amendment 48 lovember 23, 1981
where:
P is the fraction of full power at which the core is operating K(Z) is the function given in Figure TS 3.10-2 Z is the core height location FQ F
(Ej) is the function given in Figure TS 3.10-6 Ej is exposure of the fuel rod for the F of interest N
B. F Limits For All Fuel F
x 1.04 < 1.55 (1 + 0.2(1 -
P))
For 0 to 24,000 MWD/MTU burnup fuel AHx N
F x 1.04 < 1.52 (1 + 0.2(1 -
P))
For greater than 24,000 MWD/MTU burnup fuel where:
P is the fraction of full power at which the core is operating If either measured hot channel factor exceeds the values specified in 3.10.b.1, the reactor power shall be reduced so as not to exceed a fraction of the design value equal to the ratio of the F N or F N limit to measured value, whichever Q
AH I
is less, and the high neutron flux trip setpoint shall be reduced by the same ratio. If subsequent incore mapping cannot, within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, demonstrate that the hot channel factors are met, the overpower AT and overtemperature AT trip setpoints shall be similarly reduced.
Following initial loading and at regular effective full power monthly intervals thereafter, power distribution maps using the movable detection system, shall be made to confirm that the hot channel factor limits of specification 3.10.b.1 are satisfied.
The measured FEQ (Z) hot channel factors under equilibrium conditions shall Q
satisfy the following relationship for the central axial 80% of the core:
A. Westinghouse Electric Corporation Fuel FEQ(Z) x 1.03 x 1.05 x V(Z) <
(2.22/P) x K(Z)
Q B.
Exxon Nuclear Company Fuel FEQ(Z) x 1.03 x 1.05 x V(Z) < FT(Ej)/P x K(Z)
Q Q
Proposed Amendment 48
. November 23, 1981
- 2.
- 3.
4.
48 48 48
where:
P is the fraction of full power at which the core is operating V(Z) is defined in Figure TS 3.10.b.8 F Q(Z) is a measured FQ distribution obtained during the target flux determination
- 5. Power distribution maps using the movable detector system shall be made to confirm the relationship of specification 3.10.b.4 according to the following schedules with allowances for a 25% grace period:
A.
During the target flux difference determination or once per effective full power monthly interval whichever occurs first.
B.
Upon achieving equilibrium conditions after reaching a thermal power level more than 10% higher than the power level at which the last power distribution measurement was performed in accordance with 3.10.b.5.A above.
C.
If a power distribution map indicates an increase in peak pin power, F N of 2% or more, due to exposure, when compared to the last power 48 distribution map either of the following actions shall be taken:
1, F E(Z) shall be increased by an additional 2% for comparison to the Q
relationship specified in 3.10.b.4 OR ii. F EQ(Z) shall be measured by power distribution maps using the incore Q
movable detector system at least once every 14 effective full power days until a power distribution map indicates that the peak pin power, FAH is not increasing with exposure when compared to the last power distribution map.
- 6. If the measured FEQ exceeds the 3.10.b.4 relationship but not the 3.10.b.1 Q
limit, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> take one of the following actions:
4, Take corrective actions to improve the power distribution and upon achieving equilibrium conditions measure the target flux difference and verify that 3.10.b.4 is satisfied, OR Proposed Amendment 48 November 23, 1981
B. Reduce reactor power and high neutron flux trip setpoint by 1% for each percent that the measured F exceeds the relationship of 3.10.b.4.
Q Reactor power may subsequently be increased provided that adequate margin is demonstrated by a power distribution map to reasonably assure that the relationship of 3.10.b.4 can be met at the increased power level.
- 7.
The reference equilibrium indicated axial flux difference as a function of power level (called the target flux difference) shall be measured at least once per full power month.
- 8. The indicated axial flux difference shall be considered outside of the limits of sections 3.10.b.9 through 3.10.b.12 when more than one of the operable excore channels are indicating the axial flux difference to be outside a limit.
- 9. Except during physics tests, during excore detector calibration and except as modified by 3.10.b.10 through 3.10.b.12 below, the indicated axial flux
+
48 difference shall be maintained within a -5% band about the target flux difference.
- 10. At a power level greater than 90 percent of rated power if the indicated axial flux difference deviates from its target band, the flux difference shall be returned to the target band immediately or reactor power shall be reduced to a level no greater than 90 percent of rated power.
- 11.
At power levels greater than 50 percent and less than or equal to 90 percent of rated power:
A. The indicated axial flux difference may deviate from its +5% target band for a maximum of one hour (cumulative) in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provided the flux difference does not exceed an envelope bounded by -10 percent and
+10 percent from the target axial flux difference at 90% rated power and increasing by -1% and +1% from the target axial flux difference for each 2.7% decrease in rated power below 90% and above 50%.
If the cumulative TS 3.10-4 Proposed Amendment 48 November 23, 1981
time exceeds one hour, then the reactor power shall be reduced immediately to less than or equal to 50% power and the high neutron flux setpoint reduced to less than or equal to 55% of rated power.
B. A power increase to a level greater than 90% of rated power is contingent upon the indicated axial flux difference being within its target band.
- 12.
At a power level no greater than 50% of rated power:
A. The indicated axial flux difference may deviate from its target band.
B.
A power increase to a level greater than 50% of rated power is contingent upon the indicated axial flux difference not being outside its target band for more than two hours (cumulative) of the preceding 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
48 One half of the time the indicated axial flux difference is out of its target band up to 50% of rated power is to be counted as contributing to the one hour cumulative maximum the flux difference may deviate from its target band at a power level less than or equal to 90% of rated power.
- 13.
Alarms shall normally be used to indicate non-conformance with the flux difference requirement of 3.10.b.10 or the flux difference time requirement of 3.10.b.11.A. If the alarms are temporarily out of service, the axial flux difference shall be logged, and conformance with the limits assessed, every hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and half-hourly thereafter.
Proposed Amendment 48 TS 3.10-4a November 23, 1981
0
- c. Quadrant Power Tilt Limits
- 1. Except for physics tests, whenever the indicated quadrant power tilt ratio exceeds 1.02, one of the following actions shall be taken within two hours:
A. Eliminate the tilt.
B. Restrict maximum core power level two percent for every one percent of indicated power tilt ratio exceeding 1.0.
- 2. If the tilt condition is not eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, reduce power to 50 percent or lower.
- 3. Except for low power physics tests, if the indicated quadrant tilt exceeds 1,09 and there is simultaneous indication of a misaligned rod:
A. Restrict maximum core power level by 2 percent of rated values for every one percent of indicated power tilt ratio exceeding 1.0.
B, If the tilt condition is not eliminated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the reactor shall be brought to a minimum load condition (<30 Mwe).
- 4.
If the indicated quadrant tilt exceeds 1.09 and there is no simultaneous indication of rod misalignment, the reactor shall immediately be brought to a No Load condition (< 5% reactor power).
- d.
Rod Insertion Limits
- 1.
The shutdown rods shall be fully withdrawn when the reactor is critical or approaching criticality.
- 2.
The control banks shall be limited in physical insertion; insertion limit is shown in Figure TS 3.10-3.
48
- 3.
Insertion limit does not apply during physics tests or during periodic exercise of individual rods.
However, the shutdown margin indicated in Figure TS 3.10-1 must be maintained except for the low.power physics test TS 3.10-5 Proposed Amendment No.
48 November 23, 1981
to measure control rod worth and shutdown margin. For this test, the reactor may be critical with all but one high worth rod inserted and the part length rods fully withdrawn.
- e.
Rod Misalignment Limitations
- 1.
When reactor power is greater than or equal to 85% of rating the rod cluster control assembly shall be maintained within + 12 steps from their respective banks.
If a rod cluster control assembly is misaligned
+
from its bank by more than -
12 steps (indicated) when reactor power is greater than or equal to 85%, the rod will be realigned or the core power 48 peaking factors shall be determined within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and specification 3.10.b applied. If peaking factors are not determined within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor power shall be reduced to less than 85% of rating.
- 2.
When reactor power is less than 85% of rating, the rod cluster control
+
assemblies shall bemaintained within -
24 steps from their respective banks.
If a rod cluster control assembly is misaligned from its bank 48
+
by more than -
24 steps (indicated) when reactor power is less than 85%,
the rod will be realigned or the core power peaking factors shall be determined within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and specification 3.10.b applied.
- 3.
And, in addition to 3.10.e.1 and 3.10.e.2 above, if the misaligned rod 148 cluster control assembly is not realigned within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the rod shall be declared inoperable.
- f.
Inoperable Rod Position Indicator Channels
- 1.
If a rod position indicator channel is out of service, then:
A. For operation between 50 percent and 100 percent of rating, the position of the rod cluster control shall be checked indirectly by core instrumentation (excore detector and/or thermocouples and/or movable incore detectors) every shift, or subsequent to rod motion exceeding a total displacement of 24 steps, which ever occurs first.
Proposed Amendment No.
48 TS 3.10-6 November 23, 1981
S B. During operation below 50 percent of rating, no special monitoring is required.
- 2.
Not more than one rod position indicator channel per group nor two rod position indicator channels per bank shall be permitted to be inoperable at any time.
- 3.
If a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned from 3.10.f.1.(A) above, then specification 3.10.e will be applied.
- g.
Inoperable Rod Limitations
- 1.
An inoperable rod is a rod which does not trip or which is declared inoperable under specification 3.10.e or 3.10.h.
TS 3.10-6a Proposed Amendment 48 November 23, 1981 48
BASIS SHUTDOWN REACTIVITY Trip shutdown reactivity is provided consistent with plant safety analysis assumptions.
To maintain the required trip reactivity, the rod insertion limits of Figure TS 3.10-3 must be observed. In addition, for hot shutdown conditions, 48 the shutdown margin of Figure TS 3.10-1 must be provided for protection against the steamline break accident which requires more shutdown reactivity at end of core life (due to a more negative moderator temperature coefficient at 48 end-of-life boron concentrations).
Rod insertion limits are used to assure adequate trip reactivity, to assure meeting power distribution limits, and to limit the consequences of a hypothetical rod ejection accident.
The available control rod reactivity or excess beyond needs, decreases with decreasing boron concentration, because the negative reactivity required to reduce the core power level from full power to zero power is largest when the boron concentration is low.
The exception to the rod insertion limits in Specification 3.10.d.3 is to 48 allow the measurement of the worth of all rods less the worth of the worst case of an assumed stuck rod; that is, the most reactive rod.
The measurement would be anticipated as part of the initial startup program and infrequently over the life of the plant, to be associated primarily with determinations of special interest, such as end-of-life cooldown or startup of fuel cycles which deviate from normal equilibrium conditions in terms of fuel loading patterns and anticipated control bank worths.
These measurements will augment the normal fuel cycle design calculations and place the knowledge of shutdown capability on a firm experimental as well as analytical basis.
Proposed Amendment 48 TS 3.10-8 November 23, 1981
Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special precautions are taken during the test.
POWER DISTRIBUTION CONTROL Criteria Criteria have been chosen for Condition I and II events as a design basis for fuel performance related to fission gas release, pellet temperature, and cladding mechanical properties. First the peak value of linear power density must not exceed the value assumed in the accident analysis.1, 3 Second, the minimum DNBR in the core must not be less than 1.30 in normal operation or in short term transients.2 In addition to conditions imposed for Condition I and II events, the peak linear power density must not exceed the limiting Kw/ft values which result from the large break loss of coolant accident analysis based on the ECCS acceptance criteria limit of 22000F.
48 F N(Z), Height Dependent Nuclear Flux Hot Channel Factor
--Q F N(Z),
Height Dependent Nuclear Flux Hot Channel Factor, is defined as the maximum Q
local neutron flux in the core at core elevation Z divided by the core averaged neutron flux, assuming nominal fuel and rod dimensions.
48 FQ(Z) is the measured FN distribution obtained at equilibrium conditions during Q
Q the target flux determination.
TS 3.10-9 Proposed Amendment 48 November 23, 1981
An upper bound envelope for F N defined by specification 3.10.b.1 has been 48 Q
determined from extensive analyses considering all operating maneuvers consistent with the technical specifications on power distribution control as given in Section 3.10. The results of the loss of coolant accident analyses based on this upper bound envelope indicate that peak clad temperatures remain below the 48 22000F limit.
N The F (Z) limits of specification 3.10.b.1.A include consideration of enhanced fission gas release at high burnup, off-gassing (release of absorbed gases), and other effects in fuel supplied by Exxon Nuclear Company; this results in an additional penalty in the form of the function F (Ej), as shown in Figure 48 TS 3.10-6, which is applied to Exxon fuel.
References 7 and 8 discuss these phenomena.
When an F measurement is taken, both experimental error and manufacturing tolerance 48 must be allowed for. Five percent is the appropriate allowance for a full core map taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.
In specification 3.10.b.1 and 3.10.b.4 F N is arbitrarily limited for P< 0.5 48 (except for low power physics tests).
N
- FAH, Nuclear Enthalpy Rise Hot Channel Factor N
F, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power.
Proposed Amendment 48 TS 3.10-10 November 23, 1981
It should be noted that F H is based on an integral and is used as such in the DNB calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus the horizontal power shape at the point of maximum heat flux is not necessarily directly related N
to F In the specified limit of F N there is an 8% allowance for uncertainties1 which means that normal operation of the core is expected to result in F N < 1.55/1.08.
The logic behind the larger uncertainty in this case is that (a) normal perturba tions in the radial power shape (e.g. rod misalignment) affect N in most cases N
N without necessarily affecting FN, (b) the operator has a direct influence on FN through movement of rods, and can limit it to the desired value, he has no N
direct control over FH and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be compensated for in FN FQ by tighter axial control, but compensation for FN is less readily available. When a N
measurement of H is taken, experimental error must be allowed for and 4% is the appropriate allowance.
The use of F N in specification 3.10.b.5 is to monitor "upburn" which is defined N
as an increase in F with exposure. Since this is not to be confused with
- 6H observed changes in peak power resulting from such phenomena as xenon 48 redistribution, control rod movement, power level changes, or changes in the number of instrumented thimbles recorded, an allowance of 2% is used to account for such changes.
TS 3.10-11 Proposed Amendment 48 November 23, 1981
Rod Bow Effects N
The FAH limits of specification 3.10.b.1 include consideration of fuel rod bow effects. Since the effects of rod bow are dependent on fuel burnup an additional penalty is incorporated in a decrease in the F limit of 2% for 0-15000 MWD/MTU fuel burnup, 4% for 15000-24000 MWD/MTU fuel burnup, and 6% for greater than 24000 MWD/MTU fuel burnup.
These penalties are counter-balanced by credits for increased Reactor Coolant flow and lower core inlet temperature.
The Reactor Coolant System flow has been determined to exceed design by greater than 8%.
Since the flow channel protective trips are set on a percentage of full flow, significant margin to DNB is provided.
One half of the additional flow is taken as a DNB credit to offset 2% of the F H penalty. The existence of 4%
additional reactor coolant flow will be verified after each refueling at power prior to exceeding 95% power.
If the reactor coolant flow measured per loop N
averages less than 92560 gpm, the FH limit shall be reduced at the rate of 1%
for every 1.8% of reactor coolant design flow (89000 gpm design flow rate) for fuel with greater than 15000 MWD/MTU burnup. Uncertainties in reactor coolant flow have already been accounted for in the flow channel protective trips for design flow. The assumed T inlet for DNB analysis was 5400F while the normal T inlet at 100% power is approximately 532 F. The reduction of maximum allowed T inlet at 100% power to 536 F as addressed in specification 3.10.k provides an additional 2% credit to offset the rod bow penalty. The combination of the N
penalties and offsets results in a required 2% reduction of allowed FN for high burnup fuel, 24000 MWD/MTU. The permitted relaxation in FNH allows radial power shape changes with rod insertion to the insertion limits.
48 Proposed Amendment 48 TS 3.10-12 November 23, 1981
Surveillance Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors.
The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identifies operational 48 anomalies which would, otherwise, affect these bases.
For normal operation, it is not necessary to measure these quantities. Instead it has been determined that, provided certain conditions areobserved, the hot channel factor limits will be met; these conditions are as follows:
- 1. Control rods in a single bank move together with no individual rod insertion differing by more than an indicated 12 steps from the bank demand position and osit~on48 where reactor power is > 85%, or an indicated 24 steps when reactor power is <85%.
- 2. Control rod banks are sequenced with overlapping banks as shown in Figure TS 3.10-3.
48
- 3.
The control bank insertion limits are not violated.
- 4.
Axial power distribution control specifications which are given in terms of flux difference control and control bank insertion limits are observed.
Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.
Proposed Amendment 48 TS 3.10-13 Nbvember 23, 1981
The specifications for axial power distribution control referred to above are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers.
Conformance with specification 3.10.b.9 through 3.10.b.12 ensures the FN upper bound envelope is not exceeded and xenon distributions will not develop which at a later time would cause greater local power peaking.
At the beginning of cycle, power escalation may proceed without the constraints of section 3.10.b.5 since the startup test program provides adequate surveillance 48 to ensure peaking factor limits. Target flux difference surveillance is initiated after achieving equilibrium conditions for sustained operation.
The target (or reference) value of flux difference is determined as follows.
At any time that equilibrium xenon conditions have been established, the indicated flux difference is determined from the nuclear instrumentation.
This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux difference. Values for all other core power levels 48 are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviations of +5% flux difference are permitted from the indicated reference value. Figure TS 3.10-5 shows a typical construction of the target flux difference band at BOL and Figure TS 3.10-4 shows the typical variation of the full power value with burnup.
Proposed Amendment 48 November 23, 1981 TS 3,10-14
Strict control of the flux difference (and rod position) is not as necessary during part power operation. This is because xenon distribution control at part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict control at part power. Strict control of the flux difference is not possible during certain physics tests or during required, periodic, excore calibrations which require larger flux differences than permitted. Therefore, the specifications on power distribution control are not applied during physics tests or excore calibrations; this is acceptable due to the low probability of a significant accident occurring during these operations.
In some instances of rapid plant power reduction automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached. This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target; however, to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed on operation outside the band.
This ensures that the resulting xenon distributions are not significantly different from those resulting from operation within the target band. The instantaneous consequences of being outside the band, provided rod insertion limits are observed, is not worse than a 10% increment in peaking factor for flux difference in the range +10% to -10% from the target flux increasing by +1% from the target axial flux difference for each 2.7% decrease in rated power below 90%
48 and above 50%.
Therefore, while the deviation exists the power level is limited to 90% or lower depending on the indicated flux difference without additional core monitoring.
If, for any reason, flux difference is not controlled within the Proposed Amendment 48 TS 3.10-15 November 23, 1981
+5% band for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50% is required to protect against potentially more severe consequences of some accidents unless incore monitoring is initiated.
48 As discussed above, the essence of the procedure is to maintain the xenon distribution in the core as close to theequilibrium full power condition as possible. This is accomplished, without part length rods, by using the boron system to position the full length control rods to produce the required indicated flux difference.
For Condition II events the core is protected from overpower and a minimum DNBR of 1.30 by an automatic protection system. Compliance with the specification is assumed as a precondition for Condition II transients, however, operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.
QUADRANT POWER TILT LIMITS The radial power distribution within the core must satisfy the design values assumed for calculation of power capability. Radial power distributions are measured as part of the startup physics testing and are periodically measured at a monthly or greater frequency. These measurements are taken to assure that the radial power distribution with any quarter core radial power asymmetry conditions are consistent with the assumptions used in power capability analyses.
The quadrant tilt power deviation alarm is used to indicate a sudden or unexpected change from the radial power distribution mentioned above. The two percent tilt TS 3.10-16 Proposed Amendment 48 November 23, 1981
alarm setpoint represents a minimum practical value consistent with instrumentation errors and operating procedures. This symmetry level is sufficient to detect significant misalignment of control rods. Misalignment of control rods is considered to be the most likely cause of radial power asymmetry. The requirement for verifying rod position once each shift is imposed to preclude rod misalignment which would cause a tilt condition less than the 2% alarm level. This monitoring is required by Technical Specifications, Section 4.1.
The two hour time interval in specification 3.10.c is considered ample to 48 identify a dropped or misaligned rod. In the event that the tilt condition cannot be eliminated within the two hour time allowance, additional time would be needed to investigate the cause of the tilt condition. The measurements would include a full core physics map utilizing the movable detector system.
For a tilt condition < 1.09 an additional 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> time interval is authorized to accomplish these measurements.
However, to assure that the peak core power is maintained below limiting values, a reduction of reactor power of two percent for each one percent of indicated tilt is required.
Physics measurements have indicated that the core radial power peaking would not exceed a two-to-one relationship with the indicated tilt from the excore nuclear detector system for the worst rod misalignment.
In the event a tilt condition of < 1.09 cannot be eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor power level will be reduced to the range required for flux mapping and turbine synchronization.
If tilt ratio greater than 1.09 occurs which is not due to a misaligned rod, the reactor shall be brought to a low power condition for investigation by flux TS 3.10-17 Proposed Amendment 48 November 23, 1981
mapping. However, if the tilt condition can be identified as due to rod misalignment, operation can continue at a reduced power (2% for each 1% the tilt ratio exceeds 1.0) for the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period necessary to correct the rod misalignment.
INOPERABLE ROD POSITION INDICATOR CHANNELS The rod position indicator channel is sufficiently accurate to detect a rod -7.5 inchesaway from its demand position.
If the position indicator channel is not operable, the operator will be fully aware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on excore nuclear detectors, and/or movable incore detectors, will be used to verify power distribution symmetry.
INOPERABLE ROD LIMITATIONS One inoperable control rod is acceptable provided the potential consequences of accidents are not worse than the cases analyzed in the safety analysis report.
A 30 day period is provided for the re-analysis of all accidents sensitive to the changed initial condition.
ROD DROP TIME The required drop time to dashpot entry is consistent with safety analysis.
DNB PARAMETERS The DNB related accident analysis assumed as initial conditions that the T inlet was 40 F above nominal design or T avg was 4 0F above nominal design. The Reactor Coolant System pressure was assumed to be 30 psi below nominal design.
TS 3.10-18 Proposed Amendment 48 November 23, 1981
REFERENCES (1)
FSAR Section 4.3 (2)
FSAR Section 4.4 (3) FSAR Section 14 (4) (deleted)
(5) Letter from E. R. Mathews, (WPSC), to D. G. Eisenhut, (NRC), dated January 8, 1980, submitting information on Clad Swelling and Fuel Blockage Models.
(6) Letter from E. R. Mathews, (WPSC), to A. Schwencer, (NRC), dated December 14, 1979, submitting the ECCS Re-analysis properly accounting for the zirconium/water reaction.
(7) George C. Cooke, Philip J. Valentine; "Exposure Sensitivity Study for ENC XN-1 Reload Fuel at Kewaunee Using the ENC-WREM-IIA PWR Evaluation Model, WN-NF-79-72," Exxon Nuclear Company, October, 1979.
(8) Letter from L. C. O'Malley, (Exxon Nuclear Company) to E. D. Novak, (WPSC),
providing FQ exposure dependence as a function of rod burnup.
(9) XN-NF-77-57 Exxon Nuclear Power Distribution Control for Pressurized Water Reactor, Phase II, Jan. 1978.
TS 3.10-19 Proposed Amendment 48 November 23, 1981 48
(10.92,0.935)
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Proposed AMENDMENT N2 48 November 23, 1981 TARGET BAND
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