ML20213A566
| ML20213A566 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 04/20/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213A509 | List: |
| References | |
| NUDOCS 8704280077 | |
| Download: ML20213A566 (7) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TO FACILITY OPERATING LICENSE NO. NPF-58 CLEVELAND ELECTRIC ILLUMINATING COMPANY, ET AL PERRY NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-440
- 1. 0 INTRODUCTION By letters dated February 20 and 24, 1987, the Cleveland Electric Illuminating Company (CEI/ licensee) requested a change to the Technical Specifications for Perry Nuclear Power Plant'.
This change would temporarily suspend the requirement that the automatic initiation function of the reactor core isolation cooling (RCIC) system be operable.
t On February 16, 1987, CEI conducted a test of the RCIC system.
When the RCIC pump injected water through the head spray of the reactor vessel, the reactor water level instrumentation gave a false high water level indication.
The detailed scenario of this event was described in the attachment to the licensee's letter of February 20, 1987.
The exact cause of this anomaly has not been identified.
In addition, CEI experienced several instances in which the RCIC inboard containment isolation valve did not open properly on demand.
CEI had requested, and was granted, a change in the Technical Specifications (TS) which suspended for a period not to exceed 30 days certain aspects of the RCIC system's operability requirement.
This is discussed in the staff Safety Evaluation Report issued on March 5, 1987.
By letters dated March 9, 19, 20 and 31, and April 1, 7 and 10, 1987, CEI requested further relief from the RCIC system TS which would suspend the operability requirement for automatic initiation and operational criteria of the RCIC system to minimize the adverse consequencs from reactor water level errors induced by the RCIC operation during periods when reactor power is equal to or less than 75% of rated thermal power.
It is proposed that this TS change will expire on May 31, 1987.
CEI indicated that additional tests on the RCIC system have identified potential causes for the anomalous reactor water level indication and that several schemes for correcting the problem are being evaluated.
Two potential fixes will be implemented prior to startup following the current shutdown, which I
began on March 24, 1987.
CEI is requesting the additional relief in order to provide more time to test the modifications and to continue Perry's start-up test program.
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. 2.0 EVALUATION The purpose of the RCIC system is to provide high pressure coolant makeup to facilitate reactor shutdown upon loss of normal feedwater.
The RCIC system also provides the capability for maintaining the reactor in a hot standby condition for an extended period.
The RCIC system, however, is not part of the emergency core cooling systems.
During the current phase of start-up testing of the Perry Plant, the primary function of the RCIC system is to deliver cooling water to the reactor vessel in order to maintain proper water level in a hot standby condition.
Normal RCIC operation is as follows:
Following a reactor isolation, if reactor vessel water level degreases to Level 2 (129.8 inches above the top of active fuel), the RCIC system will automatically initiate. The system will continue to operate automatically without i
operator intervention until the vessel water level reaches Level 8 (219.5 inches above the top of the active fuel).
At this time, the injection valve receives a close signal to prevent overfilling the reactor.
If the reactor vessel water level again decreases after the system shutdown at Level 8, the system will reinitiate when the vessel water level decreases to the Level 2 initiation setpoint.
In this way the RCIC system automatically maintains sufficient water level in the reactor.
However, CEI has experienced anomalies with the four reactor water level instrument legs while the RCIC system is injecting water through the reactor vessel head.
Two of the instrument legs have experienced anomalies at both high reactor pressure and low reactor pressure.
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other two legs have experienced anomalies only at a reactor pressure of about 100 psig.
In all cases the anomalies are only seen while the RCIC system is operating.
The indicated water level during the anomalous behavior is up to four feet higher than actual water level.
As a result, automatic initiation or trip signals may not be received from the affected instruments until actual water level is as much as four feet below the trip setpoint.
To minimize the occurrence of this situation, CEI has proposed to lock out the automatic initiation feature of the RCIC system.
Therefore, during a reactor isolation event, the Level 2 signal will automatically initiate only the High Pressure Core Spray (HPCS) system.
If the HPCS is not available and there is no other high pressure water source (i.e., feedwater, CRD), the RCIC system will be initiated manually to avoid the use of the Automatic Depressurization System (ADS) and the low pressure Emergency Core Cooling System (ECCS).
The ADS and i
low pressure ECCS systems would still be available, if needed.
i-CEI indicated that the operators have been alerted as to the level error introduced by RCIC operation and administrative controls have been implemented to minimize use of RCIC.
The operators have been' instructed not to use the RCIC system as long as adequate makeup water is available from another high pressure water system (i.e., HPCS, feedwater, CRD).
However, the RCIC system is maintained available if it is needed.
The staff also raised a concern regarding the potential effects of water level errors on the ECCS performance during loss of coolant accident (LOCA) and abnormal transient events.
In response to the staff's request, CEI provided by letter dated March 20, 1987, a safety evaluation conducted by General Electric (GE) for the Perry Plant.
The safety evaluation indicates that a small LOCA inside the containment is the most limiting event when water level errors induced by RCIC operation are considered.
The RCIC system is assumed to be activated manually during the event; consequently, the water level instruments which have shown anomalies at high reactor pressure are assumed to give false high level indications.
The other two reference legs which have not exhibited anomalies for reactor pressure higher than 100 psi are assumed to function.
These two legs are sufficient to provide all level signals required for safety trips needed to assure adequate core cooling.
Therefore, the safety evaluation concludes that there is no safety impact on the ADS /ECCS a
performance.
The staff has reviewed the GE safety evaluation and notes that when the level error is present in the two reference legs which have experienced anomalies at high pressure, automatic ECCS/ ADS actuations depend upon a single level sensor.
The staff is also concerned that the level errors may not be pressure dependent and could potentially affect all of the water level instrumentation at high reactor pressure.
It is the staff position, therefore, that CEI should provide assurance that the water level instrumentation will reliably provide all the safety actuation signals, as necessary.
To meet this requirement, CEI has proposed to perform a series of tests of the level instrumentation system at various reactor pressures and RCIC flow conditions.
These tests are described in the CEI letter dated April 10, 1987.
CEI also proposed success criteria for the tests and operational criteria for the RCIC system.
The staff has reviewed the proposed operability tests and the proposed success criteria for particular modifications and for the level instrumentation system.
The staff concluded that the proposed tests and conditions in conjunction with the licensee's actions to minimize or eliminate the use of RCIC will meet the staff's requirement and are acceptable for short term operation as requested.
Based upon the above considerations, the staff finds that the proposed change to the TS for the RCIC system is acceptable as follows:
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1.
The operability requirement for automatic initiation of the RCIC system may be suspended for operation up to 75% rated thermal power.
2.
The suspension described in Item 1 will expire on May 31', 1987.
3.
The licensee will immediately modify its operational procedures and conduct operator training such that the RCIC system is not operated unless all other high pressure systems fail to maintain adequate water level.
4.
Surveillance of the RCIC system will be maintained to ensure that the system is functional in the manual mode whenever reactor pressure exceeds 150 psig.
5.
The operability tests of the RCIC system will be conducted as soon as practical after the plant has been restarted from the current outage.
6.
At any time during plant operation when the RCIC system is not in operation, if any one of the water level instrumentation' channels shows anomalous level indication, all level instrumentation shall be declared INOPERABLE and operators will take actions to shutdown the plant immediately as required by the Technical Specifications.
o 7.
If plant shutdown is required by the actions stated in Item 6 above, restart of the plant will not be allowed without staff approval.
3.0 EMERGENCY CIRCUMSTANCES This issuance of this amendment would extend the duration of a similar exception to the Technical Specifications which was authorized by the staff for a period of 30 days, beginning on February 24, 1987 and ending on March 26, 1987 (letter from Robert M. Bernero, NRC, to Murray R. Edelman, CEI, dated February 24, 1987 and License Amendment No.1 dated March 5,1987).
In its initial submittal of February 20, 1987, CEI had requested relief until October 31, 1987; however, the staff decided to grant it for only 30 days while CEI further analyzed the reactor water level measurement anomalies and developed its plans to eliminate them. According to CEI's letter of March 9, 1987, modifications dealing with the affected instrument lines will be made during a maintenance outage following Startup Test Condition 1.
The plan was to complete the outage in time to restart the plant about March 26, 1987.
However, the staff has since been informed that the outage did not begin until March 24, and restart of the plant has been delayed in order to implement the modifications and perform maintenance activities.
The staff has reviewed the circumstances associated with the licensee's request and agrees that this amendment is necessary for Perry Unit 1 to continue startup testing and operation.
The staff also concluded that CEI provided a sufficient basis for a finding that these circumstances could not have been avoided.
However, the staff's need for clarifying
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information, which was provided by CEI's subsequent letters, resulted in i
there being inadequate opportunity for public comment prior.to expiration j
of the initial relief period on March 26,- 1987.
Therefore the staff finds that a valid emergency existed, as defined in 10 CFR 50.91(1)(5).
3.1 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The Commission has provided standaro; for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92.
A i-proposed' amendment to an operating license for a facility involves no significant hazards considerations if operation of the facility in i
accordance with a proposed. amendment would not:
(1) Involve a i
significant increase in.the probability or consequences of-an accident l
previously evaluated; or (2) Create the possibility of a new or different i
kind of accident from any accident previously evaluated; or (3) Involve a i
significant reduction in a margin of safety.
The licensee has provided an analysis of significant hazards considerations
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hazards considerations are involved.
CEI proposes to allow operation at in its request for a license amendment and has concluded that no significant
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j power levels up to 75 percent of rated thermal power with the RCIC injection valve not capable of automatic opening.
This is to eliminate the potential j
for reactor water level instrumentation anomalies which have been experienced during recent tests of the RCIC system.
The RCIC system will otherwise i
be maintained OPERABLE and will be available to the plant operators as needed.
As indicated in the licensee's letter of February 20, 1987, the most limiting transient with the RCIC system inoperable is the loss of feedwater transient.
For such an event with the reactor at 75 percent of rated thermal power, CEI has performed an analysis which utilized decay i
heat generated in accordance with 10 CFR 50, Appendix K, and only took credit for the reduced power condition.
This analysis resulted_in fuel
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peak clad temperatures (PCTs) much lower than allowed under Appendix K.
i More than adequate cooling water would be available in most circumstances i
from the feedwater system, the control rod drive-(CRD) system, and the-high pressure cooling system, and the-RCIC system can be activated manually.
If necessary, the automatic depressurization system (ADS) and the low pressure emergency core cooling systems (ECCS) would also be available.
Therefore, the staff has concluded that the proposed change involves no significant increase in the probability or consequences of an j
accident previously evaluated.
I The operators have been alerted to the reactor water level error introduced by RCIC operation and they have been instructed on when RCIC-l should inject into the vessel.
Due to the level anomalies encountered, this has been isolated to those occasions when high pressure core spray and feedwater are not available with reactor pressure above.the low 7
pressure ECCS setpoints, and water level in the reactor vessel indicates i
injection is required.
Appropriate administrative controls have been l
added to the RCIC system operating instructions and the plant off-normal i
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However, this amendment has not required any changes to the emergency operating procedures since the RCIC system is not one of the emergency core cooling systems.
Therefore, the staff has concluded that this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
CEI has determined, by conservative analysis, that the RCIC system is not required to mitigate the consequences of any accident or transient within the proposed power limit.
However, only the automatic injection feature will be deactivated temporarily and, as discussed above, the RCIC system can be started by the operators if it is needed. With these changes, even assuming that operation of RCIC partially degrades the level instruments, the non-degraded portion of the level instruments along with our requirements restricting power level, in conjunction with availability of the CR0 cooling pumps, provides the bases for finding compliance with the requirements for 10 CFR 50.46 (Appendix K) and the provisions of Standard Review Plan Section 6.3.
Therefore, the staff has concluded that this proposed change does not involve a significant t
reduction in a margin of safety.
Accordingly, the staff concludes that the proposed change does not involve a significant hazards consideration.
3.2 STATE CONSULTATION
The staff consulted with the State of Ohio by telephone on March.26,1987.
There were no comments on this amendment.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The 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 occupational radiation exposure. The Commission made a final no significant hazards consideration finding with respect to this amendment.
Accordingly, this 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 nor environmental assessment need be prepared in connection with the issuance of this amendment.
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5.0 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 proposed manner, and (2) 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 the security nor to the health and safety of the public.-
' Principal Contributors:
T. M. Su T. E. Collins P. H. Leech
. Dated: April 20,1987 b
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