ML18057A980
| ML18057A980 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/14/1991 |
| From: | Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9106240345 | |
| Download: ML18057A980 (7) | |
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consumers Power POW ERIN&
MICHlliAN'S PRO&RESS Palisades Nuclear Plant:
27780 Blue Star Memorial Highway, Covert, Ml 49043 June 14, 1991 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
GB Slade General Manager UNREVIEWED SAFETY QUESTION-POTENTIAL FOR LEAKAGE OF CONTAINMENT SUMP WATER TO THE SIRW TANK DURING AN MHA During review of an inhouse corrective action document, it was determined that a potential leak path exists which could result in radioactive primary coolant system (PCS) water entering the Safety Injection and Refueling Water Tank (SIRWT) during the recirculation phase following a design basis accident.
Leakage past the engineered safety feature (ESF) valves (two in series control valves CV-3027 and CV-3056, and a manual isolation valve-3225) during a Maximum Hypothetical Accident (MHA) has not been considered when calculating the control room or off-site doses in accordance with GDC-19 and 10CFRlOO, respectively.
No provisions exist for leak testing the ESF valves.
The Plant Review Committee and the independent off-site review group have concluded that, since the valve leakage has not been considered in the analysis, the consequences of an accident or malfunction of equipment may be increased beyond those which have previously been considered in the FSAR.
The potential leakage could result in radiation exposure which would exceed previously calculated doses and exceed the acceptance criteria for the control room and off-site boundary doses.
As required by 10CFR50.59, changes to the facility described in the safety analysis report which involve an unreviewed safety question are submitted as an application for a license amendment pursuant to 10CFR50.90.
Consumers Power Company, therefore, requests that the Palisades Facility Operating License be amended by granting exemption from the FSAR requirement to perform the MHA analysis in accordance with the Standard Review Plan, Section 15.6.5, Appendix B, Subsection 11(1) which specifies that leakage as a result of passive component failure is included in determination of the radiological consequences of a design basis loss-of-coolant accident.
It is requested that the exemption remain in place until further analysis or plant modifications 9106240:345 910614
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provide conformance with the SRP, but no later than startup for the beginning of Cycle 10.
The following general actions are being pursued -in an effort to resolve this concern:
- Continue work on the MHA analysis eliminating the conservative assumptions
- Complete modifications to allow seat-leakage testing of the CVs by the next refueling outage.
- Evaluate a modification to manual isolation valve MV-3225 which will allow testing or ensure a leak tight barrier.
- Complete a temporary modification to route the SIRW tank vent through a filtration system. This temporary modification is expected to be completed by July 1, 1991.
- Provide guidance in Plant Emergency Response Procedures regarding the issuance of KI tablets.
- Continue to evaluate other appropriate actions which would temporarily or permanently eliminate the consequences of leakage or eliminate the leakage path.
Possible procedural modifications to inject and mix concentrated sodium hydroxide solution in the SIRW tank, in order to retain iodine in solution, were considered as a mitigating action.
However, due to the expected demands on the operations staff in an MHA event, it is not considered a practical solution and, therefore, is no longer being considered.
Providing KI tablets to the appropriate personnel (eg. control room personnel and technical support center personnel) prior to the initiation of recirculation following the DBA will lower the dose to the individual and at the same time allow in excess of 1 gpm total leakage to the SIRW tank.
Furthermore, the installation of a filtration system on the SIRW tank vent allows in excess of 5 to 10 gpm total leakage to the SIRW tank while retaining a significant margin between the calculated offsite dose and the 10CFRlOO 1 imits.
Attached is an issue description, justification for continued operation and a determination of no significant hazards determination.
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Gerald B Slade General Manager CC Administrator, Region III, USNRC Palisades Resident Inspector Attachment
ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 UNREVIEWED SAFETY QUESTION-POTENTIAL FOR LEAKAGE OF CONTAINMENT SUMP WATER TO THE SIRW TANK DURING AN MHA June 14, 1991 4 Pages
1 INTRODUCTION Palisades has been evaluating a possible unanalyzed release path through an atmospheric vent located on top of the Safety Injection Refueling Water (SIRW)
Tank.
Release scenarios assume that if a LOCA were to occur the MHA source terms would exist. Therefore, following recirculation actuation and alignment of engineered safety feature (ESF) pumps to the containment sump, pressure in the discharge header, combined with assumed SIRW Tank isolation valve leakage, would cause back flow into the SIRW tank which could be released directly to the atmosphere through the unmonitored 6-inch tank vent.
Earlier analysis of a similar line on the suction side of the ESF pumps had concluded that this was not a concern due to the high elevation of the tank.
Continued work on the issue, however, identified two additional lines downstream of the safety injection pumps which could also return water to the SIRW tank.
If valve leakage exists, the uncontrolled and unmonitored release through the SIRW vent becomes a concern.
SYSTEM DESCRIPTION The high and low pressure safety injection pumps and the containment spray pumps take suction from the SIRW tank through two separate 18 inch suction headers.
Each header is isolated during a recirculation actuation signal (RAS) from the SIRW tank using one air operated isolation valve and one check valve in series. The same valve configuration also isolates the two containment sump headers.
To provide proper suction pressure to the pumps, the SIRW tank is located on top of the auxiliary building roof approximately 75 ft above the pump suction inlet.
Discharge of the containment spray pumps flows through the shutdown cooling heat exchangers and into the containment building atmosphere.
Discharge of the high and low pressure safety injection pumps is directly into the primary coolant system.
Two lines are included downstream of these pumps which return to the SIRW tank.
The first line is the minimum flow path from each ESF pump which combines into a single header and is returned to the SIRW tank.
When RAS occurs, this line is isolated by two independent isolation control valves (CVs) in series. The second line originates downstream of the shutdown cooling heat exchangers and is used for mixing of the SIRW tank and for system testing. This test line is isolated from the SIRW tank during operation by a locked closed manual isolation valve (MV).
BACKGROUND In 1987, a special test procedure was developed with the intent of measuring the leakage through the SIRW tank isolation valves. Although analysis showed the leakage through the SIRW tank outlet isolation valves was not likely to cause back flow into the SIRW tank due to the elevation difference between the SIRW tank and the containment sump (following RAS), the procedure was developed at the request of the independent safety review group.
The test was first performed in September of 1990 at which time the valves failed their leakrate acceptance criteria. Root cause evaluation was initiated and again concluded that back leakage through the SIRW outlet isolation valves was not a
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concern due to the high elevation of the SIRW tank.
However, this root cause evaluation raised the concern of other possible leak paths when considering the recirculation and test headers.
To address the concern, further analysis was performed and concluded that a concern indeed existed.
CURRENT STATUS 2
Preliminary calculations show that a very low leak rate through the SIRW tank isolation valves will approach or exceed control room limits set in GDC-19 and site boundary limits set in 10CFRlOO.
The leakage past the test line MV and the recirculation line series CVs cannot be measured until plant modifications are made.
Analysis and information concerning this issue has been developed into a justification for continued operation and reviewed by the Plant Review Committee (PRC} with the conclusion that an unreviewed safety question currently exists.
Because the exact amount of leakage from the valves in question is not known, it was decided that when excessive valve leakage is considered the consequences of an accident or a malfunction of equipment important to safety could be increased beyond what had been previously considered in the FSAR.
The CVs are already the subject of a docketed relief request, pending modifications in 1992, to provide leak testing capability.
Modifications are being considered to provide filtering of the SIRW tank vent and to provide additional means to isolate the lines.
JUSTIFICATION FOR CONTINUED OPERATION Continued Operation of the plant without the ability to measure the leak rate to the SIRW tank is justified for the following reasons:
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The fuel damage and release of fission products to the containment that are assumed in the MHA analysis are much larger than one would expect for any of the Palisades Design Basis Accidents.
The Palisades FSAR (Section 14.22.1.1} states, "The Maximum Hypothetical Accident (MHA} is postulated to release substantially more fission products and result in more severe consequences than any incident considered credible." The large break LOCA would be expected to give the largest release of fission products to containment for the Design Basis Accidents, but the LOCA analysis of record is performed to meet the acceptance criteria of 10CFR50.46 and does not determine dose consequences.
The FSAR Design Basis Accident with the greatest radiological consequences, other than the MHA, is the Control Rod Ejection. The Control Rod Ejection analysis for cycle 9 predicts approximately 15% fuel damage which is assumed to release the gap fission products.
In considering the Control Rod Ejection fuel failures, the maximum leakage rate to the SIRW tank, that would still maintain control room doses below GDC-19 limits, is approximately 9 gpm, and approximately 18 gpm to mai.ntain site boundary doses below 10CFRlOO limits.
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The present evaluation for determining the allowable leak rate for these valves has several simplifying assumptions that will be eliminated in a more detailed analysis.
One assumption is that the leakage is transported directly from the sump to the atmosphere. This assumption neglects any
3 fission product reduction due to decay or removal due to having to travel through a lengthy section of pipe into the SIRW tank, mixing with the volume of air in the tank, and then venting out of the tank.
For example, it would take over an hour for the fission products leaking past the valves at 1 gpm to reach the SIRW tank.
- 3)
The probability of an accident causing fuel damage and a large fission product release to containment is very small.
Preliminary PRA sequence quantification indicates that a large LOCA core melt frequency would be on the order of 10-6 occurrences per year.
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A large break LOCA that would cause fuel damage would also mean that the PCS boundary is not intact and the HPSI and LPSI pumps would have discharge pressures that are at a minimum (HPSI at approximately 400 psig). This would reduce the driving head for the leak through CVs to a minimum.
Conversely, accidents where the PCS pressure remains high would be expected to produce less fuel damage and minimum fission product release to the containment sump.
- 5)
The CVs are in series gate valves such that both would have to leak to produce a release. These valves are normally open and are seldom operated. These facts combined with the use of non-corrosive materials and SIRW tank chemistry control, indicate that no significant mechanism exists for valve degradation.
The leak rate through these valves can be reasonably expected to be quite small.
In addition, during the past refueling outage, the CVs were disassembled and refurbished and reassembled.
Maintenance included seat blueing tests and stroke testing.
Thus, the inability to test valve leakage is not expected to have serious safety implications.
- 6)
The manual isolation valve (MV) 3225 was refurbished including a seat blueing test in 1988. This valve is opened only during testing of the shutdown cooling pumps during a refueling outage surveillance. Although no specific valve leakage is known, the total PCS leak rate when on shutdown cooling, during the 1990-1991 refueling outage, when a higher pressure exists at the MV than would occur during the containment sump recirculation mode, was observed to be less than 1 gpm.
Little if any of this total PCS leak rate is thought to be due to MV 3225.
- 7)
The dose in the control room can be reduced by a factor of approximately ten by issuing potassium iodide tablets to block the intake of radioactive Iodine to the thyroid. This would increase the allowable leak rate for the control room dose analysis, if we concluded that early post-LOCA action to distribute potassium iodine tablets should be taken.
DETERMINATION OF NO SIGNIFICANT HAZARDS Continued operation of the plant does not involve a significant increase in the probability of an accident previously evaluated because the presence of leakage from the safeguards pumps in the small amounts of concern cannot cause or influence the probability of an accident.
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The consequences of an accident are potentially increased by leakage through the valves to the SIRW tank.
Given the foregoing discussion, the consequences of any Design Basis Accident, except the MHA, is expected to remain within acceptance limits. The calculated consequences of the MHA could exceed the dose limits for the plant but not significantly because the conservatisms that exist in the present MHA analysis, combined with the simplifying conservative assumptions made in determining the effect of the valve leakage, are believed to result in an overestimation of the actual MHA dose that a detailed calculation will determine. Also, there is no indication or expectation that gross leakage does exist through these valves.
The inability to measure the leakage through these valves or the possibility that small leakage might exist does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The inability to measure the leakage through these valves or the possibility that small leakage might exist could possibly reduce the margin of safety.
The dose consequences are increased by the additional radioactivity assumed to be released from the SIRW tank but the increase would not be significant because of the reasons previously stated. The consequences from all of the Design Basis Accidents are expected to be below limits. The allowed leakage can be increased by removing conservatisms in the analysis and there is no reason to believe that gross leakage exists. Therefore, continued operation of the plant with the exact leakage rate of these valves unknown does not represent a significant reduction in the margin of safety.
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