ML20070U504
| ML20070U504 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/03/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18057A837 | List: |
| References | |
| NUDOCS 9104090108 | |
| Download: ML20070U504 (4) | |
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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AN ALLEGATION REGARDING THE POTENTIAL FOR A RETURN TO CRITICALITY FOLLOWING A LARGE BREAK LOCA AT PAllSADES CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT DOCKET NO. 50-265
1.0 INTRODUCTION
By letter of September 20, 1990 (Ref. 1), Region 111 requested NRR to assist in review of an allegation regarding LOCA analyses for Palisades.
In References 1 through 3, the alleger indicated that the assumptions which are intended to assure a conservative estimate of fyel cladding temperature are not conservative with respect to the potential for post-LOCA criticality. We have reviewed the allegation based on the information in References 3 and 4 and additional submittals provided by the alleger during the period of October 1990 to January 1991. We evaluated the information provided by the licensee in conference calls and follow-up submittals to the NRC.
In addition, we have examined LOCA test data to assist in review of the allegation.
As a result, we have prepared the following evaluation.
2.0 EVALUATION The Palisades LOCA analyses are performed using analysis assumptions specified in Appendix K to 10 CFR 50.
These assumptions are defined with the deliberate intention of assuring a conservative calculation of peak cladding (1) use of the temperature.
Some of the more important assumptions in this regard include:
lioody model to maximize the blowdown rate, (2) loss of all safety injection water to the containment during the bypass period, and (3) an assumed loss of off site power with consideration of single f ailure to minimize the safety injection flow.
The effect of these assumptions is to minimize the post-blowdown water inventory in the reactor pressure vessel and thereby prolong the time period during which the core is uncovered. This results in a conservatively high estimate of peat cladding temperature.
It also results in the core being reflooded with water only from the highly borated safety injection tanks (SIT) or safety injection system (SIS).
Reactor shutdown is easily achieved under these assum tions e
(because of the combination of boron and voiding).
No credit is assumed for control rod insertion.
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2-From a reactivit3 perspective however, these same assumptions estimate a less reactive state than would be expected in a more realistic scenario.
In a realistic scenario significantly more primary system water would remain in the vessel following the vessel blowdown.
This water would be cool (relative to normal operating temperature) and would be at the full power boron concentration.
This residual water would dilute (in boron concentration) the incoming safety injection water and thereby significantly alter (in the noncor,arvative direction) the reactivity potential of the mixture.
In addition, water remaining in the vessel will result in an earlier start of reflooding since less injection water is needed to fill the lower plenum.
This means that fuel rod temperatures will be lower (than in the LOCA analyses) and will result in less boil off durir.g the initial reflood stages.
Since boil off leads to an increase in boron concentration of the remaining coolant and increased voiding, this too is nonconservative from a reactivity perspective.
The more low boron concentration water remaining in the reactor vessel the more likely that reflood will result in a critical core.
Thus, the amount of the remaining we.ter in thr reactor vessel has a significant effect on the reactor remaining shutdown following a large break LOCA.
In reviewing the allegation, we have examined the licensee's post-LOCA criticality analyses documented in References 3 and 5, and additional calculations of October 25, October 13, and November 27, 1990, in these analyses the licensee has attempted to demonstrate that the boron concentration in the vessel water mixt a at the time of reflood is greater than the critical boron concentration.
Our effort included a review of the Paliso_les plant primary coolant system (PCS) and ECCS volumes and configuration, model and analysis assumptions, system response.
We also examined the amount of reactivity which could be added to the system by the t.coldown of the coolant during depressurization.
The review found that a cooldown of 300 F provides the potential for a very significant reactivity insertion (several percent).
We also determined that the licensee's conclusions regarding recriticality depend upon assumptions and outputs from the licensing basis LOCA calculations.
As discussed above, design basis LOCA analyses minimize the amount. of low boron concentration water remaining in the vessel and are therefore nonconservative from a criticality perspective.
Finally, we examined experimental data pertinent to the question.
LOFT test L1-2 described in Reference 6 is a large break blowdown test conducted f rom typical PWR initial temperature and pressure.
The test result shows that water equisalent to 90 percent of water volume of the lower plenum remains in the reactor vessel after the blowdown.
While the applicability of this data to Palisades may be deoated, it nonetheless is a data point based upon simulated PWR conditions.
For a smaller break LOCA (but large enough to discredit control rod insertion for reactor shutdown) it may be expected that more water could remain in the reactor vessel af ter the blowdown because of a smaller depressurization rate.
3.0 DISCUSSION Based on the above considerations it is the staff's position that a return to cri tical cannot be excluded.
However, it is expected that only a prompt critical configuration would be of concern from a safety perspective This is I
considered unlikely because of the need for a very rapid reactivity insertion.
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Experimental data (Ref. 7) indicates that during reflood when the coolant comes in contact with hot fuel rods, significant voiding occurs, lhis has the beneficial ef fect of introducing negative reactivity f rom the increase in voids, as well as reducing the rate of reflooding (aod thus the rate of addition of potitive reactivity) due to the increabed resistance of the two phase mixture.
A non prompt return to critical would not be expected to have significant effect M the plant recoverv following a LOCA because these same negative feedbacks would limit the power level, and increased steaming would assist in cooling those portions of the core which are uncovered.
In addition, the return to power would be short-lived because of the continued boron injection f rom the ECCS.
These factort, in conjunction with the low likelihood of the initiating event (large break LOCA), and the consideration that the control rods must also f ail to insert lead to the conclusion that continued operation is acceptable.
4.0 CONCLUSION
Based on our review of the alleger's submittals, the licensee's submittals, and the test date discussed in Section 2.0, we conclude that a return to criticality cannot be excluded.
This conclusion is based upon the following:
(1) there is the potential for a significant reactivity insertion due to cooldown of the moderator during depressurization; (2) test data indicate that large amounts of (Iow boron concentration) primary system water may remain following the blowdown of the primary system; and (3)' the licensee's criticality analyses use design basis LOCA models and assumptions which minimize the amount of low boron concentration water involved in the core reflood and are thus non-conservative relative to a return to criticality.
However, our decision on the validity of the allegation in this SER does not affact the continued operation of the Palisades plant.
This is based on the following consideration:
(1) The initiating event (large break LOCA) is a low probability event; (2) A return to criticality requires f ailure of the control rods to insert prior to reflood; (3) A short term return to non prompt criticality is not likely to significantly impact the course of LOCA recovery because of the negative feedback of effects of voiding and the cooling effect of increased steaming (if a notable power generation level is achieved), and the imminent shutdown f rom continued injection of high boron concentration ECCS water; (4) A prompt critical condition is unlikely due to the ef fect of the void distribution expected during reflood.
(5) The licensee's analyses have been performed consistent with the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.
5.0 REFERENCES
1.
Letter from H. J. Miller, to Dennis M. Crutchfield, " Request for Assistance Allegations Regarding the Palisades Nuclear Plant," dated September 20, 1990.
. 2. of Reference 1 - Transcript of September 4,1990 meeting between the alleger, Region 111, and NRR (via telephone).
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- 3. of Reference 1 - A11eger File on Safety injection Tank Boron Concentration.
4 of Rcierence 1 - Consumers Power Company letter dated August 1,1988 to the NRC, Concerning a Tech Spec Change Request for Modification of Peaking Factors and LOCA Limits.
- 5. of Reference 1 - Safety injection Tank Boron Analysis from 1982, 198S, 1986, and the 1982 licensing submittal.
6.
TREE-NUREG-1026, " Experimental Data Report for LOFT Nonnuclear Test L1-2," January 1977.
7.
JAERI-M-90-236, ' Evaluation Report on SCTF Core-Il Test S2-08 (Effect of Core Inlet Subcooling on Thermal-Hydraulic Behavior including Two-Dimensional Behavior in Pressure Vessel During Reflood in PWR-LOCA),"
January 1991, principal Contributor:
M. Chatter (on S. Sun 4
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