ML17227A303
| ML17227A303 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/19/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17227A301 | List: |
| References | |
| NUDOCS 9202280035 | |
| Download: ML17227A303 (12) | |
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UNITED STATES NUCLEAR REGULATORY COMIVllSSION WASHINGTON, D.C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO END-OF-CYCLE MODERATOR TEMPERATURE COEFFICIENT FLORIDA POWER AND LIGHT COMPANY ST.
LUCIE UNITS 1
AND 2 DOCKET NOS. 50-335 AND 50-389
1.0 INTRODUCTION
By letter dated December 22, 1987 (Ref. 1), the Florida Power and Light
.Company (the licensee) requested a change to the Technical Specifications (TS) for St. Lucie Units 1 and 2.
Specifically, the proposed change would modify the requirement to perform a moderator temperature coefficient (MTC) test near the end of each cycle.
The NTC test would not be required if tests conducted at earlier times in the cycle, coupled with the design NTC calculations and core reactivity characteristics, predict satisfactory MTC behavior for the remainder of the cycle.
This change affects Surveillance Requirement (SR) 4.1.1.4.2.c for St. Lucie Unit 1 and SR 4.1.1.4.1.c for St. Lucie Unit 2.
The proposed change would also modify St. Lucie Unit 1 SR 4.1.1.4.2.b to be consistent with the Combustion Engineering Standard TS (Ref. 2) and with the St. Lucie Unit 2 TS.
This license amendment request was discussed in a conference call with the licensee on October 26, 1989.
An additional discussion was held on January 9, 1990.
In these discussions, the staff asked a number of questions concerning the request.
The licensee responded to the staff's questions by letter dated February 26, 1990 (Ref. 3).
During NRC management review of the technical effort, it was concluded that the technical review should be terminated to conform with NRC TS policy.
By letter dated August 30, 1991 (Ref. 4), the NRC denied the proposed amendments because they were not consistent with the existing Standard Technical Specifications (STS) and did not meet any of the criteria required for plant-specific deviations from the STS.
By letter dated November 22, 1991 (Ref. 5), the licensee asked the NRC to reconsider the denial.
The reduction in surveillance was proposed as a continuation of the reactor trip and plant transient reduction program to reduce challenges to safety systems.
The licensee stated that the NTC surveillance has not caused reactor-trips but has resulted in dropped control element assemblies which are considered to be precursor events for reactor trip.
The staff agreed to reconsider the proposed amendments based on their technical merit since a significant improvement in plant safety is a criterion for justification of plant-specific deviations from the STS.
The staff's discussion and evaluation of the proposed changes to the St. Lucie Unit 1 and Unit 2 NTC SR and Bases are provided below.
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2.0 DISCUSSION The pressurized water reactor (PWR)
NTC is primarily a function of the fuel assembly lattice design.
A PMR fuel assembly lattice, in terms of unborated water-to-metal ratio, is undermoderated.
This means that if the unborated water-to-metal ratio in such a lattice is increased, the lattice reactivity will increase.
In terms of moderator coolant water temperature, an under-moderated lattice behaves as follows: if the temperature of the moderator coolant water is increased, the density of the unborated water is decreased, thereby decreasing neutron moderation by the water coolant which decreases the reactivity of the lattice.
This decreased reactivity is caused primarily by the competing changes in the resonance escape probability and the thermal neutron utilization.
For the condition described, the lattice would have a
negative NTC.
Simply stated, a negative NTC means that an increase (decrease) in the unborated moderator coolant water temperature will cause a decrease (increase) in fuel assembly lattice reactivity.
The PWR HTC considers the effects of temperature on both the moderator density and the thermal neutron spectrum in a combined way.
These two principal
, components of the HTC are not usually computed separately.
Two other factors of lesser importance should also be noted as causing changes to the PWR HTC.
These factors are (1) the reactor
- pressure, and (2) the very small void content of the coolant caused by local or statistical boiling.
These two factors
- can, of course, be considered as simply changes to the moderator coolant water density in any analysis of the PWR NTC.
When soluble boron has been added to the moderator coolant water of a PWR, the fuel assembly lattice may no longer be undermoderated.
An increase in moderator coolant water temperature will now result in two principal effects.
The first effect, as before, is the decrease in neutron moderation caused by a
decrease in water density which decreases the reactivity of the PWR lattice.
The second effect is a decrease in the soluble boron concentration per unit volume of water.
This increases the PWR lattice reactivity because neutron absorptions by the boron (in particular, by the high thermal neutron absorption cross section of the boron-10 isotope) in the water decreases.
These competing effects may result in either a positive or negative HTC for a PWR fuel assembly lattice depending on the soluble boron concentration in the moderator coolant water.
As discussed
- above, the PWR HTC is primarily dependent on two parameters of the moderator coolant water:
(I) its density, and (2) the soluble boron concentration.
There are other factors which can affect the HTC directly by changing the neutron flux spectrum and indirectly by changing the core reactivity.
Factors such as fuel assembly enrichment and placement in the core, fuel assembly uranium-235 burnup, xenon and samarium concentrations and distributions, other fission product poison buildup, the buildup and burnup of plutonium isotopes and other actinides, thermal-hydraulic effects affecting moderator coolant water density variations in the core and the compensation of these factors in the fuel cycle design by a combination of control rods, soluble boron in the moderator coolant water, and burnable poison, fixed in the lattice as a
lumped burnable absorber or integral with the fuel, all can affect the PWR HTC.
These other factors increase in importance and the predictive uncertainty can be expected to increase as the soluble boron concentration is reduced as a function of burnup.
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PWR fuel cycles are designed such that the initial soluble boron concentration in the moderator coolant water will be between about 800 and 1300 ppm at beginning-of-cycle (BOC) at a hot full power (HFP) and all rods out (ARO) condition.
The initial boron concentration is decreased over a period of about 2 days to accommodate the buildup of an equilibrium concentration of the fis-sion product poison xenon.
This decrease in the boron concentration is about 200 to 300 ppm.
The boron concentration also decreases by a much smaller amount to accommodate the buildup of an equilibrium concentration of the fis-sion product poison samarium.
As cycle operation and fuel burnup proceeds, the soluble boron concentration in the moderator coolant water is gradually decreased to maintain core reactivity, with the reactor at essentially an ARO condition.
The soluble boron concentration is decreased as a function of burnup in a nearly linear manner until the end-of-cycle (EOC), where it is reduced to near zero.
This variation of the soluble boron concentration with burnup is termed the boron letdown curve.
The boron letdown curve is an important operational curve used by the operators throughout the fuel cycle.
It should be noted that the variation of the boron letdown curve described above is typical of reactors using boron in some form as a lumped burnable absorber.
Other lumped or integral burnable absorber could lead to somewhat different functional dependencies of the boron letdown curve with burnup.
The HTC variation with fuel burnup follows the boron letdown curve in this usual manner of PWR operation, although the functional dependency with burnup may be somewhat different.
At BOC, when the soluble boron concentration is high, the HTC for most PWRs will be in the range of 0 to -5 pcm/'F at HFP, ARO and no equilibrium xenon conditions.
(Note that a
pcm is a reactivity of 10 delta-K/K.)
For a few plants,
- however, the TS permit the HTC to be positive
(+3 to +5 pcm/'F) for these conditions.
With the build-in of equilibrium xenon and movement of some control rods partially into the reactor, the soluble boron concentration can be decreased and the HTC will become more negative by about 3 pcm/'F.
Host PWRs would, therefore, have a negative HTC at BOC in the range of -3 to -8 pcm/'F at the
- HFP, ARO, equilibrium xenon core.
condition.
For the few PWRs for which reactivity compensation for equilibrium xenon by decreasing the soluble boron concentration is not sufficient to result in an HTC that meets TS limits, either more control rods must be inserted into the reactor core, or additional fuel burnup must be accumulated so that the soluble boron concentration may be decreased sufficiently to ensure an HTC within TS limits.
After equilibrium xenon conditions have been achieved, the HTC becomes gradually more negative as the cycle burnup increases to EOC where it achieves its most negative value and where the soluble boron concentration is near zero.
A number of conclusions and observations can be made based on the brief discussion of the HTC for a PWR provided above.
These are:
(1)
The HTC is strongly dependent on the density of the moderator coolant water.
(2)
The HTC is strongly dependent on the soluble boron concentration in the moderator coolant water.
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(3)
Other effects and characteristics of the core design and operation can affect the NTC directly and indirectly, especially through reactivity compensation by control rod insertion or soluble boron concentration changes.
The changes to the HTC caused by these effects and character-istics are usually smaller in magnitude than the first two effects, but increase in importance as the soluble boron concentration is reduced as a
function of fuel burnup.
(4) If the fuel cycle is operated as designed and follows the predicted boron letdown curve within the uncertainty of the measured soluble boron con-centration, then the actual NTC will follow the predicted HTC to within some established uncertainty.
Fai lure to follow the predicted boron let-down curve would likely be due to uncertainties in the prediction of other fuel burnup dependent effects on the NTC.
3.0 EVALUATION It is not practical to perform HTC measurements continuously so that design calculations are used to predict the NTC behavior throughout the fuel cycle.
Current St. Lucie Units 1 and 2 TS SR specify three HTC tests to ensure the accuracy of the design calculations.
For St. Lucie Unit 1 HTC testing must be performed (1) during initial startup following reload, t2) within 7 effec-tive full power days (EFPD) of reaching a rated thermal power equilibrium boron concentration, and (3) within 7 EFPD of reaching an equilibrium boron concentration of 300 ppm.
For St Lu.cie Unit 2 NTC testing must be performed (l) during initia 1 startup following reload,
{2 within 7 EFPD of reaching an equilibrium boron concentration of 800
Currently, acceptable HTC values beyond the 300 ppm point in the cycle to EOC are assured by an EOC HTC design value which is within TS limits.
The accuracy of the design calculations will have been verified by the previous HTC tests in the cycle.
However, only the last test at the 300 ppm point in the cycle demonstrates the capability to predict the effect of burnup dependent reactivity changes on the HTC.
The change proposed by the licensee would delete the HTC test at 300 ppm if two conditions are met.
The first condition is that the two HTC tests that would still be performed must meet an acceptance criterion of a 2 pcm/'F for the comparison of measured and predicted values of the NTC.
The second condition is that an evaluation will be performed at 300 ppm to verify that all parameter variations (for example, reactor coolant system average temperature, soluble boron concentration (measured versus predicted),
control rod bank position, and fuel burnup) allowed by the TS have not resulted in an NTC value more negative than the design value.
The evaluation performed for the second condition will be performed in accordance with SR 4.1. 1.1.2 on overall core reactivity.
Failure to meet either of these two conditions would require an HTC test at 300 ppm.
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The first condition for elimination of the NTC test at 300 ppm is dependent on the premise that the calculation of the HTC for various reactor conditions and fuel cycle dependent core characteristics can be reliably performed with the calculational methods in use.
- Further, the extrapolation of the ETC value from the last measured point to EOC requires that the deviation of calcula-tional predictions from measured values should be consistent so that it can be accounted for in the evaluation of EOC MTC for conformance to the TS limit value.
The licensee has established
+ 2 pcm/'F as the criterion for the'omparison of measured and predicted HTC values.
This criterion has been in general use throughout the industry and is consistent with the ANSI Standard for Reload Startup Physics Tests for Pressurized Water Reactors (ANSI/ANS Standard 19.6.1 - 1985) (Ref. 6).
However, the continued conformance of the calculated HTC value at 800 ppm and 300 ppm boron to the criterion established for BOC calculations had not been demonstrated.
In response to a staff question, the licensee addressed the accuracy between the predicted and measured HTC values for St. Lucie Unit 1 and Unit 2 (Ouestion 3, Ref. 3).
The comparison of measured and predicted NTC values is provided in Table 1 of Reference 3 for eight cycles of operation at St. Lucie Unit 1 and Unit 2 at three reactor conditions:
{1) BOC, hot zero power {HZP); (2) 800
- ppm, HFP; and (3) 300 ppm, HFP.
The data are summar'ized in a different format in Table 1 of this evaluation.
All of the eight comparisons at the
- BOC, HZP conditions result in the difference between measured and predicted values= of the HTC well within the acceptance criterion of + 2 pcm/'F, as should be expected.
Four of the eight NTC comparisons of predicted and measured HTC values at the 800 ppm HFP condition exceeded the acceptance criterion of
+ 2 pcm/'F.
The largest difference was for Cycle 3 for St. Lucie Unit 2, where the predicted value was more negative than the measured value by 2.25 pcm/'F.
Four of the seven comparisons of pr edicted and measured t1TC values at the 300
- ppm, HFP condition exceeded the acceptance criterion of + 2 pcm/'F.
The largest difference was for Cycle 8 for St. Lucie 1, where the predicted value was more negative than the measured value by 3.5 pcm/'F.
However, there
was virtually no relationship between the goodness of the predictions at 800 ppm and those at 300 ppm.
In fact, two of the four predictions which failed the comparison test at 300 ppm, including the Cycle 8 predicted value with large difference, showed excellent results well within the acceptance criterion at 800 ppm.
The data presented in Table 1 fail to substantiate the acceptability of the first condition for deleting the HTC test at 300 ppm.
The data for eight cycles indicate that only four of eight 300 ppm tlTC tests would have been eliminated, and that the predicted values for two of the eliminated tests would have been outside of the acceptance tolerance and, therefore, should not have been eliminated.
Further, the predictions at 800 ppm boron appear to be sensi-tive to core design since all four Unit 1 cycles
( later cycles with greater average fuel depletion) had a nonconservative
( less negative) calculated value compared to the measured value, while a converse relation existed for the Unit 2 comparisons.
This inconsistency emphasizes the need to test the calculations against an abundance of operating cycles data (preferably representing several plants) to establish the range of core and burnup dependent uncertainties in the calculated HTC.
Even so, indications are that elimination of the 300 ppm boron ETC surveillance would require a substantial penalty on the EOC calculated extrapolation to compensate for uncertainties when evaluating compliance with the TS limit value of NTC.
By letter dated December 17, 1991 (Ref. 7), the licensee proposed an increase of -3 pcm/'F in the TS NTC limit because, recent measured data shows that the actual EOC NTC is very close to the limit value.
Since the data provided by the licensee for St. Lucie Units 1 and 2 clearly indicate that prediction versus measurement comparisons early in the cycle are not consistent with deviations obtained later in the cycle, the staff concludes that near EOC measurements are necessary to provide reasonable assurance that operation is within the TS NTC limits.
The staff, in our original technical review of the second condition to delete the 300 ppm NTC test, had concluded that modifications to the proposed TS SR would be needed to assure conformance to the second condition for omitting the NTC test at 300 ppm.
Since our review has concluded that the first condition for deleting the NTC test at 300 ppm is unacceptable, there is no need to discuss in detail our review of the second condition.
4.0 CONCLUSION
S The staff has reviewed the proposal by Florida Power and Light Company to modify the requirements to perform a
NTC test near the end of e'ach cycle for St. Lucie Units 1 and 2.
The staff has also considered the claim by the licensee that this change enhances plant safety because it reduces challenges to the plant safety system due to dropped control element assemblies that may occur during the NTC test.
Based on its review, the staff concludes that the proposed changes are not very effective in deleting the 300 ppm boron NTC test requirement and that the proposed first condition for deleting the test is unacceptable because of poor correlation between calculation results near BOC and calculation results near EOC.
Further, the staff concludes that the safety need for this surveillance to assure conformance to the TS NTC limit is of greater significance than the potential safety system challenge from dropped CEAs.
The frequent occurrence of dropped CEAs during such maneuvers may be indicative of a fundamental design deficiency in the control rod latching mechanism which should be corrected.
Date:
February 19, 1992 Principal Contributor:
L. Phillips
TABLE 1 ST.
LUCIE t1TC SURVEILLANCE RESULTS UNIT 1 (Predicted - pleasured)
HTC Values (pcm)
~Cele 6
~Cele 7
~Cele 8
~Cele 9
BOC HZP Oe70
-0.20 1.20 0.40 800 ppm HFP 2.10*
0.20 0.20 1.10 300 ppm HFP
-2.90 2 10**
-3.50**
UNIT 2
~Cele 2
~Cele 3
~Cele 6
~Cele 5
BOC HZP
-0.69
-0.22 0.24
-0.28 800 ppm HFP
-2.03*
-2.25*
-2.10*
300 ppm HFP 1.38
-2.93
-0.95 Exceeds proposed
+ 2.0 pcm criterion to eliminate 300 ppm measurement.
Exceeds calculation acceptance criterion even though no measurement would have been required by the proposed technical specification.
Note:
Tabulation is based on data in Table 1 of Reference 3.
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