ML19289F397
| ML19289F397 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/27/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19289F396 | List: |
| References | |
| NUDOCS 7906070321 | |
| Download: ML19289F397 (13) | |
Text
AFC :. 1/3 SAFETY EVALUATION REGARDING STRUCTURAL INTEGRITY OF THE REACTOR COOLANT PUW SUCTION ELB0W SPLITTER,
NORTH ANNA UNIT 1 B ackgroun_1 On March 21, 1979, we were advised by the Virginia Electric and Power Company (UPC0) that during a routine cleaning of the reactor coolant loop crossover leg pipes in North Anna Unit 2 cracks were discoverd in splitter plate 2-C which is installed in the reactor coolant system pipe elbow leading into the suction side of the reactor coolant pump. The flow splitter plates are not structural members and were installed in the North Anna Units 1 and 2 reactor coolant system to enhance flow distribution at the ptop impeller inlet and to increase uniformity in velocity distribution.
During a telephone conversation with the Office of Inspection and Enforcement (I&E) on April 5,1979, VEPC0 agreed that North Anna Unit 1 would not be returned to service until technical justification acceptable to the NRC was provided to show that the flow splitter plates installed in North Anna Unit 1 are structurally sound. I&E transmitted a letter dated April 6, 1979 to VEPCO regarding the confirmation of action.
At the request of VEPC0, a meeting was held on April 12, 1979 in Bethesda, Maryland to allow the licensee an opportunity to discuss this matter with the staff.
At this meeting, VEPC0 presented a fatigue ane.?ysis 2231 097 79060703Zt
APR27'N3 to demonstrate that cracking similar to that observec in Unit 2 would not occur in the flow splitter plates in Unit 1.
On the basis of the analysis, VEPC0 concluded that the Unit I low splitter plates would not fail because those flow splitter plates had already acctmulated sufficient fatigue cycles beyond the material endurance limit to preclude failure during subsequent operation. However, no differences in flow splitter design or flow conditions could be identified by VEPC0 for Units 1 and 2.
Also presented were the results from ultrasonic examinations, which were performed e"er the length of the Unit i flow splitter plates.
The ultrasonic technique was developed using Unit 2 to
..Lrwa the equipment and then applied to Unit 1.
The results of the ultrasonic examination indicated that the Unit I flow splitter plates were free of the severe cracking found at Unit 2.
However, two discontinuities, one three inches long located 22 inches from the leading edge of the flow splitter plate and a second 1/4 inch long two inches from the previous indication, were found during the examination of the Loop B, flow splitter plate. The disconcinuities are located at the junction of the lateral and longitudinal welds of the flow splitter plate. VEPC0 states that these discontinuities resulted from defects present in the structure from fabrication and differ from the cervice induced flaws present in the Unit 2 flow splitter plate.
Based on our evaluation of the material, design and flow conditions associated with the flow splitter plates in Units 1 and 2, we concluded that there is little margin against fatigue cracking for the Unit i flow splitter plates 2231 098
APR 2 7 and thus concluded that VEPC0's fatigue analysis to demonstrate the integrity of the Unit I flow splitter plates was unacceptable. Subsequently, we required the licensee to postolate failure of the flow splitter plate to ensure that a failed plate would not leaf to unacceptable safety consequences. We required the postulated flow splitter plate analysis to cover a spectrum of sizes ranging from small sizes that would pass through the pump into the core to a huge piece that would cause pinp failure.
To further ensure structural integrity, we required VEPC0 to develop an inservice inspection program to periodically inspect the flow splitter plates and reduce the potential for continued operations with severely cracked flow splitter plates. In response to our requirements expressed at the April 12, 1979 meeting, VEPC0, on April 15, 1979, submitted the requested safety evaluation report and inservice inspection program.
We have reviewed the information submitted by VEPC0 and our evaluation of this matter is discussed in the following paragraphs.
Evaluation The flow splitter plate under consideration is installed in a 31-inch, 90 degree elbow in the reactor coolant ptop suction.
From a consideration of pump geometry and the fracture characteristics of the failed flow splitter plate, a range of fragment sizes was established by the licensee.
The largest piece which could pass through the pump was estimated to be nine 2231 099
APR 07 4'3 inches by nine inches by 1-1/16 inch thick.
The smallest fragment considered to be a 1-1/16 inch cube.
Based on our review, we have determined that the postulated flow splitter plate failure sizes adequately represent failure sizes that likely would resait should cracking similar to that in Unit 2 occur.
The potential consequences of fragments in the estimated size range are (1) pump damage sufficient to create a reduction in loop flow, (2) damage to reactor vessel internals and instrument tubing should the flow splitter pl>te fragments manage to impact them and (3) the occurrence of flow blockage at various locations in the reactor vessel.
With respect to the reactor coolant pump, we have determined that if the largest postulated piece of failed flow splitter plate passes through the pump, the most likely result would be impeller key failure with loss of puap flow. Deformation of the impeller and diffuser vanes, most probably without fracture of hese parts, would be anticipated. Abnormal ptmp shaft bending loads are to be expected.
However, the shaft deflection resulting from these loads would alert plant operators via the installed reactor coolant ptmp shaft displacement monitoring system to stop the pumps before shaft failure occurs. Based on the above, we conclude that a technical specification is required which will require the licensee to shut down the plant and nondestructively examine all flow splitter plates 2231 100
A?R ; 7 :U9 whenever a puap shaft proximity probe indicates excessive deflection.
We concur with the licensee that the reactor piping including the flow splitter elbow will not fail from the impact of a postulated failed portion of the splitter plate.
In the event that a failed portion of a flow splitter plate should manage to reach the interior of the reactor vessel and impact on the internals structure, there could be some local deformation. However, American Society of Mechanical Engineers Boiler and Pressure Vessel Code level D limits should not be exceeded. Vessel bottom mounted in-core instrwnentation tubing will deform plastically if impacted so as to pinch off the tubing and cause instrument function lost. These instruments however a.e not required for any safety functions if the postulated event were to occur.
On the basis of our review, we have concluded that the consequences of a failed splitter plate causing damage to reactor piping, reactor coolant pump and vessel internals have been evaluated and are considered acceptable for the postular.ed evenu.
The pressure boundary will not be breached, but loss of reactor coolcat pump flow and loss of some in-core instrumentation capability should be expected, together with local deformation within the reactor coolant pump and portions of the reactor vessel internals.
With respect to the consequences of fragments sufficient to damage the reactor coolant pump and create a reduction in loop flow and the occurrence 2231 101 IF
~ :7 of flow blockage of various locations in the reactor vessel, our evaluation of these situations is as follows. Also considered in the evaluation is the use of the installed loose parts monitoring system in detecting postulated splitter element fragments.
One consideration is the fuel rod behavior effect of coolant flow blockage, either within the coolant channels of a fuel assembly or in the reactor vessel lower internals.
The licensee has considered a flow splitter plate piece simultaneously covering four lower core plate flow holes directly below a fuel assembly. Separate consideration was given to a smaller piece entering one of the core plate flow holes. The licensee has referenced the results of analyses provided on the " Standard Reference System Design RESAR-3S, Westinghouse Electric Corporation Safety Analysis Report". For that design it was predicted for complete blockage of the fuel assembly inlet nozzle that full flow recovery would occur about 30 inches downstream of the blockage.
For an assumed 41 percent blockage occurring in the axial center of a fuel bundle, flow recovery was predicted to take place four to five inches beyond the blockage.
This is illustrated schematically in Figure 1 ( attached). Also illustrated on Figure 1 is an anticipated axial power shape at beginning of life (1.55 cosine skewed toward the bottom of the core by a -20 percent axial offset). Under these conditions the areas of reduced flow occur in a region of less than maximum peaking factor.
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' Based on the above analysis, we conclude that the worst condition would be occurrence of departure from nucleate boiling on a maximun of four rods caused by a particle lodged in a single channel. Blockage at the entrance is not expected to cause departure from nucleate boiling in the affected fuel assembly.
Since the lower end fitting flow holes in the North Anna design are not of sufficient size to permit fragments greater than one-half inch to enter the fuel assembly, a flow blockage in the fuel assembly greater than that caused by the local reduction between fuel rods is not considered possible.
This is conditioned by a further definition of the smallest particle size.
However, a small particle assumed to lodge in a fuel assembly would create a blockage less than that already analyzed, or would create turbulence.
We conclude that the flow blockage resulting from the failure of the reactor coolant pump flow splitter plate will not result in a departure from nucleate boiling condition more extreme than departure from nucleate boiling on four rods for each fragment lodged in a fuel assembly.
The licensee has provided a discussion of the effects of a possible degradation of loop flow causes by the presence of a fragment in the reactor coolant punp. The assunption of a punp impeller key failure with a loss of pump flow would correspond to a partial loss of flow tr'ansient.
This transient has been previously analyzed by the licensee in the North 2231 103
. Anna Unit 1 Final Safety Analysis Report with the conclusion that a minimum departure from nucleate boiling ratio of 1.3 is not exceeded. The analysis of the rartial loss of flow transient provided in the Final Safety Analysis Report predicted reactor trip at 87 percent loop flow and 97 percent ',are flow. The analysis included a measurement uncertainty in flow measurement.
A reactor trip signal from the pump breaker position is provided as an anticipatory signal which serves as a backup to the low flow signal. A flow degradation less than the complete loss of one pump may not generate a trip on low flow. However, the reduction in departure from nucleate boiling ratio due to flow reduction to the 87 percent (trip) level is less severe.
We have considered the effect of partial loss of flow combined with flow blockage.
Since inlet flow blockage does not appreciably affect the bundle flow at elevations higher than 30 inches in the core, we expect no significant effect on the minimum departure from nucleate boiling previously calculated for the transient. Our conclusions regarding reduction in departure from nucleate boiling for the combination of events remains unchanged.
Based on our evaluation of the licensee's submittal and pertinent analyses in the North Anna Unit 1 Final Safety Analysis Report, we conclude that the effects of flow blockage resulting from failure of the reactor coolant pump 2231 104
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elbow splitter would result in the occurrence of departure from nucleate boiling on a maximtm of four rods for each fragment lodged in a fuel assembly.
Further, the combined effects of flow blockage and flow degradation due to pump damage would result in no significant additional effects on previously analyzed transients. We have concluded that the limited fuel damage which might result from the postulated events is acceptable.
We have also concluded that the combined effects of flow blockage due to loose parts and the occurrence of those accidents evaluated in Chapter 15 of the Final Safety Analysis Report (low probability events) need not be considered provided that prompt detection of the loose parts and corrective action is taken.
The prompt detection of loose parts will be accomplished by the loose parts monitoring system provided for North Anna Unit 1.
This system was evaluated against the guidelines of Regulatory Guide 1.133 (out for comment), " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors."
The present system meets the minimtm requirements of two sensors located at each natural collection region and has the minimum sensitivity suggested by the guide.
On the basis of our review, we are incorporating in the North Anna Unit 1 Technical Specifications the following requirements:
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. 5?I P ~ "3 (a) The location of the sensors.
(b) A limiting condition for operation requiring the loose-parts detection system to be operable during startup and power operation. If all channels of one or more collection regions are inoperable for more than 30 days, the reactor need not be shutdown, but a special report should be prepared and submitted to the Commission within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to an operable status.
(c) A surveillance requirement that each channel of the loose-parts detection system be demonstrated operable b," a channel check performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a channel functional test performed at least once per 31 days, and a calibration test performed at least once per 18 months.
With respect to the inservice inspection program for the flow splitter plates, we have determined on the basis of our review that the licensee's proposed inservice inspection program for the flow splitter plates is unacceptable. Therefore, we are incorporating in the North Anna Unit 1 Technical Specifications an inservice inspection program consisting of the following:
(1) Ultrasonic examinations of the elbows containing the flow splitter plates shall be conducted at each refueling outage for all loops.
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- -, (2 ) The ultrasonic exaninations shall be performed using the method developed and calibrated using the large flaws in Norch Anna Unit 2.
The details of the examination are ' contained in Procedure 1-TP-1, " Ultrasonic Test Procedure for Examining Splitter Plates in Steam Generator-to-Punp Elbows North Anna Power Station Unit 1."
The applicability of the procedure was '.anstrated to IE Regional Office II personnel, who witnessed the testing and concluded the procedure is adequate to detect large cracks similar to those in North Anna Unit 2.
(3 ) Reports of the examination results will be submitted to the NRC for review.
The report should contain a determination regarding the growth of any existing flaws in the structure and identification of any new flaws that might have occurred in the interim service period.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded the', the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR Section 50.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendnent.
2231 1OL7 e
-. Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered or a significant decrease in any safety margin, it does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security o - to the health and safety of the public. Also, we reaffirm our conclusions as otherwise stated in our Safety Evaluation Report and its Supplements.
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