ML20071H070
| ML20071H070 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 07/08/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20071H068 | List: |
| References | |
| NUDOCS 9407130311 | |
| Download: ML20071H070 (8) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTO RELATED TO AMENDMENT N0.90 TO FACILITY OPERATING LICENSE NO. NPF-62 ILLIN0IS POWER COMPANY. ET AL.
CLINTON POWER STATION. UNIT NO. 1_
DOCXET NO. 50-461
1.0 INTRODUCTION
Detection Systems," requires that selected systems cap determining reactor coolant system leakage remain operable.
leakage is collected and classified as either identified or unidentified Reactor coolant leakage.
The drywell equipment drain sump monitors identified leakage w drywell floor drain sump monitors unidentified leakage.
Reactor coolant system leakage that falls on the drywell floors is channeled through the floor drains and enters the drywell floor drain sump.
entering the floor drain sump, water passes through the drywell floor drain Prior to total integrated flow are measured. sump flow monitoring instrumentation where consists of a V-notch weir box containing a capacitance probe.The flow monitorin the weir box is directly proportional to the flow.
Water level in Thus, flow through the V-notch is equal to the sump inlet flow rate.
The capacitance probe is calibrated to correspond to the incoming flow rate and provides a continuous control room indication of the unidentified reactor coolarit system leakage rate. An alarm is of unidentified le. generated when the technical specification lihiu # 5 gpm kage occurs.
The V-notch weir box instrumentation meets floor drain sump flow monitoring.the accuracy and sensitivity requirement Once water enters the drywell floor drain sump, a system of pumps, pump-out timers, cycle counters and level switches monitors and records unidentified reactor coolant system leakage.
control room indication if excessive leakage occurs. Sump pump performance is monit automatically start and stop at pre-determined levels.The sump pumps Pump running time is monitored and provides an alarm if run times exceed a given value which would ba indicative of excessive leakage.
pump startup between cycles is monitored.In addition, the time between automatic Frequent cycling of the sump would also be indicative of excessive leakage thus generating an alarm. pumps Finally, a high-high sump level alarm would be generated indicative that sump pump operation was not maintaining proper level.
pump curve, pump running time, and the cycling time between automatic pumpB startup and shutdown, an alternative means can be used to verify overall leakage into the sump.
9407130311 940708 PDR ADOCK 05000461
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. Technical specifications limit the amount of unidentified reactor coolant system leakage to a total of 5 gpm.
Technical specifications also limit any increase of unidentified leakage to 2 gpm.within any 24-hour period.
This latter value is in accordance with NRC Generic letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking in BWR Austenitic Stainless Steel Piping," since an abrupt increase in unidentified leakage rate could be indicative _of a_ leak before breck in stainless steel piping.
F In early February of 1994, control room operators observed fluctuating leakage rates sensed by the V-notch weir box measuring unidentified _ reactor coolant system leakage.
Using the alternative means described above to verify the unidentified leakage rate, control room operators were able to verify that actual leakage increases.had not occurred.
Subsequently _ on February 13, 1994, the drywell floor drain sump-flow monitoring instrumentation was declared inoperable.
Technical Specification 3/4.4.3.1 permits-continued plant operation for 30 days provided an alternative means is used to monitor and determine unidentified leakage rates once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and the remaining leakage detection systems are operable (i.e., the drywell atmosphere particulate monitor and either the drywell atmosphere gaseous monitor or the drywell air cooler condensate flow rate monitoring system).
Ef forts by the -licensee to restore the drywell floor drain sump monitoring:
instrumentation were unsuccessful.
The instrument loop was recalibrated and equipment external to the drywell was verified to be operating properly.- In-i addition, the. licensee attempted to "backflush" the V-notch weir box by temporarily suspending sump _ pump operation to permit water to back up and dislodge any foreign material that may be blocking the V-notch.
Having exhausted all efforts to trouble-shoot from outside the drywell, the remaining alternative was to make a drywell entry to make a physical examination of the V-notch weir box and the capacitance probe.
However,. the location of the V -
notch weir box is within the biological _ shield wall and 'directly below the reactor vessel.
Due to high radiation and temperature conce_rns, a plant-shutdown would be required to permit personnel entry.
By letter dated February 25, 1994, Illinois Power ~ Company requested an:
emergency technical specification change pursuant to 10 CFR 50.91(a)(5). The change would permit continued plant operation with the inoperable drywell floor drain sump __ monitoring instrumentation provided-an alternate means was being used to monitor, and determine unidentified reactor coolant leakage rates, once every'8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Operation in this mode was requested until the first time the-plant was; brought to. cold shutdown. On March 14, 1994,1 the-staff issued License Amendment No. 89 for the Clinton Power Station authorizing continued plant operation as requested by the licensee.
v On_ April 16,_1994, the licensee' brought the facility to cold shutdown.
conditions-to perform unrelated plant mainterance.
As part of this outage, the-weir box was inspected.
The capacitance probe was found_to have been-coated with an unidentified material and the-instrumentation was found to be out of calibration. -The root cause was thus identified as instrumentation-L failure. Additionally, corrosion was cbserved between the weir box. lid and-L walls resulting in ercossive resistance between these surfaces.
In an effort i
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. to improve the capacitive geometry of the probe with that of the weir, a -
positive mechanical ground between the lid and the wall was provided.
Prior to bringing the facility back on line, the probe was replaced and the system.
was recalibrated. The instrumentation was then declared operable.
In May 1994, control room operators again began to observe inconsistent indications from the drywell floor drain sump instrumentation.
Based on the divergence between the sump instrumentation and the calculated flowrate derived from the drywell-floor drain sump pump run times, the instrumentation -
was declared inoperable on June 10, 1994.
Similar to the events of February-1994, the facility has entered a 30-day shutdown LCO, external troubleshooting has not restored operability to the instrumentation and the licensee has concluded that a drywell entry will be necessary to initiate repairs.
By letter _ dated June 20, 1994, the licensee submitted an emergency license amendment request authorizing continued plant operation with inoperable drywell floor drain sump monitoring instrumentation.
Similar to the 4
February 25, 1994 request, the-proposed amendment would permit continued-plant operation until the first time the plant is required to be brought to cold shutdoon.
2.0 EVALVATION Technical Specification 3/4.4.3.1 requires that multiple reactor coolant' system leakage detection systems remain operable.
Item 3.4.3.1.a requires-operation of a drywell atmosphere particulate radioactivity monitoring system.
This would provide early --indication of fission product release.
Item 3.4.3.1.b requires the drywell sump flow monitoring system to remain operable.
The drywell sump-flow monitoring system consists of both the drywell. floor-drain sump flow monitoring instrumentation previously discussed and-the 4
physically _ identical-drywell equipment drain sump flow monitoring
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instrumentation.used to-monitor identified reactor coolant : system leakage.
Finally,, item 3.4.3.l.c requires operation of_either the drywell atmosphere.
_ gaseous radioactivity monitoring system or the drywell. air coolers-conde'nsate flow rate monitoring' system.
These latter items would provide early indication of fission gas release or abnormal -steam conditions in the drywell.
The licensee proposes to monitor and; determine the; unidentified reactor coolant system leakage rate using an alternate _means-once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as-currently required in the technical specification action statement.
The-alternate means would be through-verification of ~ sump' pump performance and.
confirming that the integrated sump pump flow rates would not' exceed technical specification limits.
Current plant conditions are' showing that the sump pumps-are: cycling approximately_ once every six andLa half hours. The pump run times of approximatley three minutes correspond to a' calculated unidentified-leakage rate of:0.4 gallons per' minute.
The licensee states that the accident analysis is' unaffected'by the proposed changes.. The design basis accident involving leakage into the drywell is a guillotine break of the recirculation _ system suction piping.
Safety systems-accident-mitigation is automatically initiated in response to high drywell 5
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a pressure or low reactor vessel water level.
Regarding small break loss-of-coolant accidents, the Updated Safety Analysis Report Sect.an 7.7.).24.1 states that no credit is taken in the safety analysis for operation of or operator reliance upon the leakage detection monitoring instrumentation.
associated with the drywell sump.
As previously discussed, control room oper,ators will monitor and determine unidentified reactor coolant system leakage once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
In addition, alternate indications via the drywell particulate radioactivity monitoring system, and either the drywell atmosphere gaseous radioactivity monitoring system or the drywell air coolers condensate flow rate monitoring system would be available to inform control room operators of abnormal conditions. The staff concurs that an adequate alternative means exists to monitor and determine unidentified reactor coolant system leakage.
The V-notch weir box is located in a keyway beneath the reactor vessel.
Not only would this represent a high radiation level during plant operations, but it is also a highly contaminated area.
Personnel entering this area will need to wear respirators and double plastic anti-contamination suits.
Compounding the difficulty in working under such conditions are the anticipated temperatures.
Normal drywell ventilation systems are not particularly effective in this location.
The licensee anticipates that the temperatures in this region during hot shutdown conditions would approach 140
'F, Personnel entering under these conditinns would be required to wear ice packs, would need to be monitored for heat stress, and would be limited to approximately 20 minute stay times.
Cooling the plant to cold shutdown would result in primary coolant system temperatures of less than 200 *F.
While the licensee did not -
provide any quantitative assessment on the amount of additional cooling that cold shutdown conditions would provide to the drywell region, the staff agrees-that conditions would be more tolerable to personnel entry.
Viewed from a personnel safety aspect, the licensee considers cold shutdown a more appropriate entry condition. The staff concurs with the licensee that repairs should not be required until the first time-that the facility is brought to cold shutdown.
The licensee proposes to modify the current footnote on Technical Specification page 3/4 4-12 to read:
" Operation may continue after July 10, 1994, until the next COLD SHUTDOWN, provided the drywell floor drain sump flow rate is monitored and determined by alternate means at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Additionally, the drywell atmosphere particulate and gaseous radioactivity monitoring systems may be periodically taken out-of-service to perform scheduled preventive maintenance, surveillances and testing without entering the shutdown requirements of the ACTION statement."
The licensee has made this proposal to avoid periodic entries into the 12-hour shutdown statement of Technical Specification 3.4.3.1.
Technical Specification 3.4.3.1 states that if more than one of the reactor coolant-system leakage detection systems is inoperable, the facility must be brought to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
However, as discussed in the licensee's l
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, letter dated June 20, 1994, both radiation monitors are periodically removed from service. The particulate monitor needs to be taken out of service, approximately once every other week to change the filter paper.
In addition, both the particulate and gaseous monitors are taken out of service monthly for channel functional testing. While each of these activities typically requires less than an hour to perform, they would result in the facility being placed in an immediate shutdown condition. The staff does not believe that entering a plant sh 'down to perform routine surveillances and testing would be consistent with the safety significance. This is supportea by the technical specification action statement which permits continued plant-operation for 30 days with any one of these components inoperable.
The staff has reviewed the licensee's proposal for this technical specification change. Considering the alternate means of monitoring and determining unidentified reactor coolant system leakage available to the licensea, the relatively low safety significance of operating in this condition, and the desire to avoid any unnecessary plant shutdown and resultant risks, the staff finds the licensee's proposal acceptable.
3.0 EXIGENT-CIRCUMSTANCES The Commission's regulation, 10 CFR 50.91, contain provisions for issuance of amendments when the usual 30-day public notice period cannot be met. One type of special exception is an exigency. An exigency is a case where the staff and licensee need to act promptly and the staff has determined that the amendment involves no significant hazards t isiderations.
Under such circumstances, the Commission notifies the public in one of two ways: by issuing a Federal Reaister notice providing-an-opportunity for hearing and allowing at least two weeks for prior public comments, or by issuing a press releasc discussing the proposed changes, using the local
- media, in this case the Commission used the first approach.
The licensee declared the drywell floor drain sump monitoring instrumentation inoperable on June 10, 1994. With the remaining reactor coolant system leakage detection systems operable, Technical Specification 3.4.3.1 permits 30 days of continuous plant operation provided the drywell floor drain sump flow rate is monitored and determined by alternative means at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
All efforts by the licensee to restore the drywell sump inlet' flow monitoring instrumentation to operable status have been unsuccessful.
The instrument loop has been recalibrated and equipment external to the drywell has been verified to be operating properly. The only option remaining for the licensee is to enter the drywell in order to examine the V-notch weir box and associated capacitance probe. However, the V-notch weir box is located in a keyway beneath the reactor vessel and inside.the biological shield wall. Due to the high radiation and temperatures in this location, a plant shutdown would be required before personnel wauld be able to reach the instrumentation.
. The staff does not consider loss of this instrumentation, by itself, to be safety significant.
The 30-day action statement found in the technical specifications further supports this view. While the control room will not be capable of continuously monitoring unidentified leakage fl ne rates, the alternative means performed at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shsuld be sufficient to provide ample warning of any unanticipated crack in primary system piping.
The licenseo submitted the request for amendment on June 23, 1994.
It was noticed in the Federal Reaister on June 22, 1994 (59 FR 32147), at which time the staff proposed a no significant hazards consideration determination.
In its letter of June 20, 1994, the licensee requested that the amendment be issued promptly. _ The licensee stated -that such action would be necessary to preclude an unnecessary plant transient and related plant risk associated with a plant shutdown.
Due to time constraints, sufficient time was not available to permit the customary 30-day public notice in advance of this action.
Accordingly, pursuant to 10 CFR 50.91(a)(6), the Commission has determined that an exigent situation exists in that failure to act in a timely way will result in a plant shutdown.
Further, the Commission has determined that the exigent situation is not due to the failure of_the licer;ee to act in a timely manner.
There were no public comments in response to the notice published in the Federal Reaister.
4.0 FINAL N0 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION 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 considerations if operation of the facility in accordance with the amendment would not (1)- involve a significant increase-in-the probability-or -
consequences of an 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 in a margin of safety.
This_ amendment has been evaluated against the standards in 10 CFR 50.92. The Commission has made a final determination that the amendment does not involve a significant hazards consideration because:
Operation of the facility in accordance with the proposed amendment will'not involve _a significant increase in the probability or consequences of an accident previously evaluated. The proposed change would permit continued plant operation with inoperable drywell floor drain sump monitoring instrumentation.
This instrumentation-does not provide any accident
-mitigation function nor is it relied on for operator action. The instrumentation is only one of several means of providing. indication to control' room operators of unidentified reactor coolant system leakage rates.
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- n. i Control room operators will monitor and determine unidentified reactor coolant system leakage rates using an alternative means of monitoring the drywell floor drains sump pumps.
By monitoring sump pump operating times, frequency of pump cycling, and level switches, operators will verify that unidentified reactor coolant system leakage rates remain within acceptable levels.
In addition, the availability of particulate and gaseous radioactivity monitors and observation of the condensate discharge line flow rates from the drywell air coolers, will provide operators with indirect indication of any unanticipated increase in unidentified leakage.
Permitting continued plant operation until the first COLD SHUTDOWN after July 10, 1994, will avoid an unnecessary plant shutdown and resultant risk.
Since the instrumentation is only used to provide indication and no credit is taken in the safety analysis for operation of or operator reliance on this instrumentation, the staff concludes that the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.
Operation of the facility in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change does not involve a change in the operation of the plant, nor does it introduce any new failure modes.
This instrumentation does not provide any accident mitigation function nor is it relied on for operator action. Control room operators will use alternate means to periodically verify unidentified reactor coolant system leakage rates and will possess indirect means of observing increases in leakage rates via the particulate and gaseous monitors and observation of the condensate discharge line flow rates from the drywell air coolers.
Therefore, the staff concludes that the proposed change does not create-the possibility of a new or different kind of accident from any accident previously evaluated.
Operation of the facility in accordance with the amendment will not involve a significant reduction in a margin of safety.
The margin of safety associated with this proposed change relates to the limits on unidentified reactor coolant system leakage. As discussed in the Bases for Technical Specification 3/4.4.3.2, the allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.
The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for the unidentified leakage limits, the probability is small that the imperfection or crack associated with such leakage would grow rapidly.
The V-notch weir box normally provides continuous control room indication of the unidentified leakage rate of the reactor coolant system.
With this instrumentation inoperable, the licensee has proposed to monitor and determine the leakage rate through an alternative means once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Since the probability of a small imperfection or crack to grow rapidly is small, verification of leakage once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should be sufficient.
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, NRC Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking in BWR Austenitic Stainless Steel Piping," implemented more stringent limits of unidentified leakage. The generic letter imposed a limit of a 2 gpm increase in any 24-hour period since an abrupt increase could be indicative of a crack in service sensitive austenitic stainless steel piping.
The proposed change does not alter any previously set limits on unidentified leakage.
Based upon the above considerations, the staff concludes that the amendment meets the three criteria of 10 CFR 50.92.
Therefore, the staff has made a final determination that the proposed amendment does not involve a significant hazards consideration.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment.
The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area-as defined in 10 CFR-Part 20.
The NRC staff has determined that the amendment involves no significant increase in the amounts, 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 occu)ational radiation exposure.- The staff has made a final determination tlat this amendment involves no significant hazards consideration.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR -51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental. impact statement-or environmental assessment need be prepared in connection with the. issuance of the amendment.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that because the requested changes do not involve a significant increase in the probability or consequences of an accident previously evaluated, do not create the possibility of an accident of a type different-from any evaluated -
previously, and do'not involve a significant reduction in a margin-of. safety, the amendment does not involve a significant hazards consideration _ that:
(1) there ls reasonable assurance that-the health and. safety of the public will not be endangered by operation in the proposed manner, (2) such
-activities will be conducted in compliance with 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 public.
Principal Contributor: Douglas V. Pickett
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Date: July 8, 1994 1
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