ML20128E449
| ML20128E449 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 01/29/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20128E426 | List: |
| References | |
| NUDOCS 9302100446 | |
| Download: ML20128E449 (7) | |
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o UNITE D STATES l'
NUCLEAR REGULATORY COMMISSION f
f WASHINGTON, D. C. 20565
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MfETY EVALUATION BY THE OFFICE OF 1(kufAR REAC10R REGULAT10))
RELATED TO AMENDMENT NO.155 TO EglLITY OPERATING LICENSE _NO. DPR-2Q CONSUMTRS POWER COMPANY PALISADES PLANI QQCKET NO. SQ-lijli 1.0 idTRODUCHGJ By letter dated January 19, 1993, the Consumers Power Company (the licensee or CPCol requested an amendment tc the Technical Specifications (TS) appended to facitity Operating License No. DPR-20 for the Palisades Plant.
The proposed amendment changes the surveillance interval for testing two control rod drive mechanisms, CRD-20 and CRD-31, to once, in March 1993, until the end of Cycle 10, in lieu of "Every Two Weeks."
2.0 DISCUSSION AND EVALQ6HQf]
Backaroynd The Control Rod Drive Mechanisms (CRDMs) at Palisades are of the Rack and Pinion Drive type.
These drives have a drive package containing a drive.
motor, position indication equipment, and a releasing clutch, which is outside the Primary Coolant System (PCS) boundary, and a drive shaft, right angle gear set, pinion gear, and rack within the PCS boundary. The drive package is connected to the drive shaft throu,h a mechanical seal, which forms the PCS pressure boundary.
Leakage through the mechanical seal enters a cavity which is vented to a collection header, and which is sealed at the top by a vapor seal.
Each mechanical seal is provided with a thermocouple to measure the -
temperature of its leakoff. The leakoff from all 45 CRDMs is collected in a common header and routed to the containment sump.
Two CRDMs are exhibiting signs of above normal seal leakage._ Operating history has shown a trend that exercisir.g a CRDM, as required by TS Table 4.2.2, Item 2, often causes seal leakage to increase. One CRDM, CRD-20, has been declared inoperable which allows omitting the exercising of that mechanism. Technical Specifications do not allow continued operation with more than one control rod inoperable, so testing of the second CRDM exhibiting leakage has continued.
The CRDM provides two safety functions. With the exception of the motor's ability to move the control rod with the rod rundown signal, the required biweekly surveillance testing does not verify either of these safety functions, The safety. function? are:
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. First, a reactor tria signal de-energizes the clutch and allows the control rod to drop )y gravity into the core; this is the only CRDM safety function assumed in the safety analyses.
Second, whep a reactor trip signal is generated, a " rod rundown" signal energizes ill full length CRDM motors to drive their rods in case they should not trip freely into the core. The rod rundown signal is terminated when that rod nears full insertion. The clutch is designed to allow the motor to apply a torque to the drive shaft, in the "IN" direction, even when the clutch is released.
Functioning of the rod rundown feature is not assumed in the safety analyses.
A reducsi U.%im frequency would have no significant offect on the assurance that the CRDM will function properly and would reduce the probability of leakage increasing to the point where a plant shutdown is required.
Etaposed Ch3ng u Repeated testing of a leaking seal can result in shortened seal life and increased CRDM seal leak rate, which can le.d to forced shutdown due to excessive PCS leakage.
The aroposed amendment changes the frequency for control rod exercising in Tayle 4.2.2, Item 2, with a footnote which would read "During the remainder of cycle 10, CRO-20 and CRD-31 will be tested once in March 1993 in lieu of testing once every two weeks." Cycle 10 is scheduled to end on June 4, 1993.
Evaluattan The licensee has provided an analysis to demonstrate that CRDH seal leakage does not increase the likelihood of an untrippable control rod.
in order to do so, leakage would have to cause the clutch to fall to release, or cause mechanical binding of the driveshaft between the lower clutch face and mechanical seal, because all components above the lower clutch face are disengaged from the drives shaf t upon a trip, and normally wetted components inside the PCS boundary will not be mechanically bound by leakage effects.
Clutch:
In order to hinder trippability, the lower section must either fail to disengage or else jam between the shaft and some stationary compenent.
Plausible failure modes cause the clutch to disengage (thus causing a rod trip), not remain engaged. Original clutches employed a splined sleeve which was prone to binding, but current clutches use a spring bellows and jaw faces which do not depend upon sliding action.
When electrical power is removed, the upper face springs away from the lower one, an action which is not prone to mechanical janming.
Even if the vapor seal failed, leakage would not prevent rotatiot of a disengaged lower clutch element.
Bearings: There are three sets of ball bearings between the clute.h and vapor seal. To prevent a rod trip, one or more of these sets would have to bind sufficiently to resist dropping of a weight in excess of 200 pounds, or else degrade badly enough to allow gross driveshaft I
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The vapor seal protects the bearings from a corrosivo atmosphere, and leakage limitations reduce the likelihood of vapor seal failura.
Leakagelimitationsarenotchangedbytheproposedamendment.
In the past, even bearings filled with bor.c acid have performed t
aroperly. 'There is no reason to believe that any currently installed 3 earings have been exposed to steam ur boric acid.
Vapor seal: This is an elastomeric cup seal with a metal backing ring.
The steam impingement washer protects if from erosion, and the 'apor seal
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in turn protects drive components abovo from leakage. Operating temperature depeads upon seal leakoff pressure as long as flashing occurs in the leakoff cavity.
The collection header is unpressurized.
The elastomer is designed for high temperature operation, and there is no metal-to-metal contact between stationary and rotating parts.
If the vapor seal were to fail, it would not itself prevent shaf t rotation.
Steam Impingement Washer: This thin stainless washer fits loosely around the driveshaf t immediately below the vapor seal, at the top of the seal leakoff cavity, it cannot bind between the shaft and housing while remaining around the shaft, and plausible leaks will not break it.
Seal Assembly:
The rotating element is inside the PCS boundary, so leakege will not corrode or bind small internal parts.
There is ample clearance between the stationary assembly and driveshaf t.
Shear forces will prwent binding at the seal boundary itself, as seal contact area is very small and materials were selected for low friction operation.
Leak-induced temperature increase can degrade the three static 0-rings, but this will not prevent rotation.
Driveshaft: One end of the driveshaft is inside the PCS boundary, so component material was selected to withstand PCS effects.
Driveshaft upper end alignment is maintained by the lower clutch shaf t which rides in three sets of ball bearings above the vapor seal.
The drive shaft lower end bearings are within the PCS boundary.
The licensee has concluded, therefore, considering the system design and acceptable performance of the refueling outage trip test, that the control rods are trippable and, therefore, can meet heir functional requirements.
The NRC staff has reviewed the itcensee's analysis and concludes that the control rods can meet their functional requirements.
To provide hdditional assurance that the CRDM seal leakage will not increase the likelihood of an untrippable control rod, the licensee has stated that reactor shutdown 1;, procedurally required when leakage is confirmed to be a CRDM seal failure in excess of two gallons per minute.
This leak rate is well within the leakoff header flow capacity, so the limitation effectively protects vapor seal integrity.
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. Without a change to the TS, biweekly testing of operabie CRDMs with leaking seals must continue.
The resulting accelerated seal degradation resulting from biweekly testing would force a shutdown within a matter of weeks due to excessive PCS leakage. With reduced surveillance testing of CRD-20 and CRD-31, the expected rate of seal degradation will be reduced and may facilitate continued operatian until the next scheduled refueling outage.
The licensee has ptuvided evidence that CROM seal leakage within )rocedural allowance will not uuse an untrippable control rod, therefore, tle required safety function is not affected, in addition, CPCo is preparing a 15 change request to revise the test frequency for all control rod drives from biweekly to quarterly.
The licensee committed to filing that TS change request following resolution of CPCo comments and receipt of a final report from combustion Engineering.
That TS change, if granted, will supersede the amendment granted herein.
Based on the above, the staff has determined that an emergency amendment should te granted.
3.0 EMERGEitCY.CIRG)J1SJECU in accordance with 10 CFR 50.91(a)(5), the licensee has provided justification that it could not make timely application and that emergency circumstances do exist. As already discussed and as addressed in the licensee's amendment request of January 19, 1993, the licensee determined that continued biweekly surveillance testing of CRD-20 and CRD-31 could aggravate the leakage rate and lead to a forced shutdown. An increase in seal leakage was indicated after CRD-31 was tested on December 29, 1992. Thus, the NRC staff does not believe that the licensee has abused the emergency provisions in this instance.
Accordingly, the Commission has determined that there are emergency circumstances warranting prnmpt approval, by the Commission, of an amendment to the facility Technical Specifications.
4.0 FINAL NO SIGN 1ficatiT HAZARDS CONSIDiftATION DETERMINAR01.{
The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:
(1)
Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated: or (3)
Involve a significant reduction a margin of safety.
The amendment has been evaluated against the above criteria of 10 CFR 50.92.
It does not involve a significant hazards consideration because the change
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would not:
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> l Criterion 1 The proposed TS change does not alter any plant systems, instrument settings, or operating methods.
Its only potential effects would be to reduce expected CRDM seal leakagle and to reduce the assurance, normally provided by btweekly testing, that a CRDM has not become mechanically bound to the point where it cannot be moved by its motor. Mechanical binding of a CROM is not clasted as an " accident."
Therefore, operation of the facility in accordance with the )roposed change to the TS would not involve a significant increase in the probaallity of an accident previously evaluated.
The intent of the biweekly control rod exercise survelliance tests is te detect controls rods that are stuck and demonstrate that control rods can move-frooly over a small rango of mov eent (minimum of six inches.
The curret;t Palisades surveillance frequency of every two weeks was appar)ently based on engineering judgment. Operating exserience has demonstrated that this surveillance is not a principal met 1od for detecting stuck control rods. The ability to trip the control rods, i.e., the operability of the rods, is not affected by decreasing the surveillance frequency. Operability (trippability) of the rods is demorstrated by the refueling outage surveillance test.
Further evidence of operability has been demonstrated during the cm;mt cycle in the five reactor trips that have occurred in which all controi rods including CRD-20 (which has evidenced leakage since April 1992) have tripped.
Reactivity control, therefore, through control rod tripping _or through boration is not affected by this change in the surveillance frequency.
The FSAR reactivity events consider that the most reactive control rod remains fully withdrawn from the core during a-reactor trip. Because the tri)pability of the control rods are not degraded by this survelliance frequency c1ange, the consequences of these reactivity events have not been increased.
The control rod rundown feature, which is not required to mitigate an accident, will also not be degraded by the change in surveillance frequency.
Control rod indication would not be affected by the change. Additionally, the mechanical or electrical reliability of the control rods would not be' degraded by the change in frequency of the surveillance for the leaking control rods.
Therefore, combined with the ability of the control rods to remain trippable (operable), the probability of occurrence of an accident previously evaluated has not beca significantly increased.
i' The effect of CRDM seal leakage on CRDH components has been reviewed to.
determine if trippability of the control rods is affected..That review of the-components (described above) leads to the conclusion-that seal leakage will not affect the trippability of.the control rods.
Therefore, operation of the facility in accordance with the proposed change to the TS would not involve a significant increase in the consequences of.an accident previously evaluated.
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. 1 Criterion 2 The proposed change in surveillance frequency for the leaking control rods l
would not alter the eculpment design or operation Therefore, operation of d
the facility in accorcance with the proposed change to the TS would not create the possibility of a new or different kind of accident from any previously.
evaluated.
Criterion 3 Review of control red events at Palisades and Fort Calhoun (the only other plant with Palisades-style CRDMs) back to 1971 has shown no instanus in which biweekly testing detecteri untrippable rods.
Fort Calhoun data w.s obtained.
from NPR0s and was not verified with OPPD.
Inability to drive rods via the rod rundown feature was discovered in some cases, generally caused by brake, Jrive motor, or relay contactor failure. Such occurrences could have prevented control rod rundown capability, but since affected components were all above the clutches the ability to trip the control rod was not affected.
The review also indicated that there were 33 instances of untrippable or sticking control rods (of which 22 were attributable to three common failure modes which have been resolved). Of these, 4 were discovered prior to initial criticality, 29 during tests other than biweekly exercising, 2 during scrams, and 2 by failure to withdraw during startup, but none by b1 weekly testing-(some of the occurrences fit in multiple categories).
There were 2 other events in which trippability was not ascertainable from records reviewed, but neither occurred during biweekly testing.
Biweekly testing of the control rods over the 20 years of Palisades operating history has not detected any instance where control rods have not been trippable. The control rods were demonstrated trippable (operable)- by-the control rod drop timing test during the last refueling outage and by their-successful operation during the five reactor trips since refueling.
Therefore, even with the presumed most reactive rod being stuck during an FSAR reactivity event, there is not a reduction in the margin of safety with respect to limiting reactivity additions during any of these.FSAR events.
Accordingly, the Commission has determined that the amendment-involves no-significant hazards consideration.
5.0 STATE CONSULTATION
' In accordance with the Commission's regulations, the Michigan State Official was notified of-the proposed issuance of the amendment.
The Michigan State Official had no comments.
. ft. 0 EMJf0NMENTAL CONS 1DE%]l08 The amendment involves a change to a requirement with respect to the
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- installation or use nf a facility component located within the restricted area as defined in 10 CFR Part 20 and a change to the surveillance requirement.
The staff has determined that-the amendment involves no significant~ increase A-4
7 in the aniounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final no significant hazards consideration finding with respect to this amendment. Accordingly, the amendment meets the eligibility criteria for categorical exclusion sut forth in 10 CFR 51.22(c)/9).
Purs. ant to 10 CFR St.22(b),noenvironmentalimpactstatementorenvIronmentalassessmentneed be prepared in connection with the issuance of the amenament.
7.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the pro)osed manner, (2) such activities will be conducted in compliance wit 1 the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the pubilc.
Principal Contributor:
A. Hasciantonio Date: January 29, 1993 v
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