ML17158A145
| ML17158A145 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/09/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17158A144 | List: |
| References | |
| NUDOCS 9403020053 | |
| Download: ML17158A145 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555.0001 SAF TY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.101 TO FACILITY OPERATING LICENSE NO NPF 22 PENNSYLVANIA POWER AND LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 DOCKET NO. 50-388
1.0 INTRODUCTION
By letter dated August 19,
- 1992, as supplemented by letters dated May 18, 1993 and October 7,
1993, the Pennsylvania Power and Light Company (PP8L or the licensee) submitted a request for changes to the Susquehanna Steam Electric Station (SSES),
Unit 2, Technical Specifications (TS).
The requested changes would reflect a pending modification to Unit 2 that will revise the logic which controls the automatic transfer of the High Pressure Coolant Injection (HPCI) pump suction source on high suppression pool level.
The May 18, 1993, letter provided a minor revision in the wording to describe the position of the HPCI injection valve.
The revision was for clarification and to make the wording the same as the amendment application submitted for Susquehanna, Unit 1.
The supplemental change was administrative in nature and did not change the intent of the initial application and did not affect the staff's No Significant Hazards Consideration Determination.
The October 7,
1993, letter documented responses to two questions raised by the NRC staff.
The information was confirmatory in nature and did not change the amendment application and did not. affect the staff's No Significant Hazards Consideration Determination.
The same changes to the HPCI logic were approved for Susquehanna, Unit 1 by Amendment No. 130, issued on October 19, 1993.
On July 31,
- 1991, SSES Unit 1 scrammed from full power when a switchyard fault at a fossil plant resulted in de-energization of one of Susquehanna's offsite AC power supplies to a transformer.
The de-energization of the transformer resulted in actuation of the Unit 1 'A'eactor Protection System (RPS) and Main Steam Isolation Valve (MSIV) A/C channels isolation logic.
- Likewise, on Unit 2, the de-energization resulted in a similar half scram and containment isolations associated with the 'A'PS power.
On Unit 1, the 'B'ain steam line (MSL) radi ation monitor had failed earlier that morning, resulting in a
'B'PS (Division II) actuation (half-scram) and a MSIV B/D Logic isolation signal.
With the 'B'PS half-scram and MSIV B/D isolation logic signals already present, actuation of the 'A'hannels caused a
The void collapse caused by closure of the MSIVs resulted in a
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reactor water level transient.
The Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) systems initiated within seconds and injected into the reactor vessel.
The events that transpired following the scram are described in PPKL's Licensee Event Report (LER)91-008, submitted August 30, 1991.
In accordance with the Emergency Operating Procedures (EOPs),
the operators used the RCIC system to control level and the safety relief valves (SRVs) to control pressure.
Delays were encountered in reestablishing vacuum to the main condenser due to problems with the auxiliary boilers.
Controlling pressure with the SRVs is difficult. Shortly after the NSIV closure, the 'E'RV cycled open and closed twice automatically to control reactor pressure.
During the next 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, three additional RPS actuations
- occurred, one when the high reactor pressure setpoint (1037 psig) was reached and two actuations when reactor vessel level 3 (+13") was reached.
Since all control rods were already fully inserted, no rod movement occurred.
During this transient, problems were encountered in restoring the Reactor Water Cleanup (RWCU) system and placing the Residual Heat Removal (RHR) system in the shutdown cooling mode of operation.
Overall, it took about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to stabilize the plant.
The low reactor water level scram setpoint
(+13") was reached 10 times.
One of the causes for delay was the inability to use the HPCI system for pressure control during part of the restoration.
Normally, the HPCI pump draws suction from the condensate storage tank (CST).
However, if a low CST tank level (3' 1/2") or a high suppression pool water level (23'") occur, the HPCI suction supply will automatically transfer to the suppression pool.
Section 3.6.2. 1 of the TSs requires that a certain minimum and maximum volume of water be maintained in the suppression pool, equivalent to a level between 22'0" and 24'0")".
Keeping the water level below 24'" ensures that there is still adequate space in the suppression pool to accommodate the large volume of water that could conceivably be released into containment from a postulated Loss of Coolant Accident (LOCA) without creating structural concerns.
If the 24'" limit is exceeded, the plant has to be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Prior to the reactor
- scram, the suppression pool water level was 23'".
Within the first 1 1/2 hours, the suppression pool water level increased to 23'", (primarily due to added inventory from SRVs being cycled open to control reactor pressure) which automatically transferred the HPCI suction from the CXS to the suppression pool.
The reason for this auto transfer is to keep the water level from exceeding 24'" in the event HPCI initiates automatically.
- However, because of the relatively poor quality of water in the suppression pool (possible rust, et al.),
compared to the primary coolant, it is not desirable to pump water from the suppression pool into the reactor unless necessary.
The HPCI system remained available for emergency core cooling if needed.
As a result of the complications encountered in coping with a transient that is within the design basis, that is an analyzed event in Chapter 15 of the Final Safety Analysis report (FSAR) and is reanalyzed for each reload, a
management meeting was held with the licensee on November 14, 1991, in the NRC's Region I offices.
The licensee's presentation was attached to NRC
combined inspection report 50-387/91-21 and 50-388/91-21 issued January 22, 1992.
The scram was particularly complicated and challenging to operators, procedures, and hardware.
At the. meeting, the licensee discussed various actions they proposed to preclude and to improve response to possible future pressurization transients.
The actions were documented in the licensee's letter of December 30, 1991, to the NRC (PLA-3707).
The licensee agreed to:
1) improve training, communication, and coordination between the plant and the power control center, 2) upgrade the affected emergency operating procedures, 3) evaluate use of a mechanical vacuum pump to pull condenser vacuum when the HSIVs are closed and auxiliary steam is not available, 4) revise the Emergency Action Levels (EALs) on Emergency Core Cooling System (ECCS) initiation, 5) revise the operating procedure for restart of the RWCU with request to reactor vessel differential temperature limit requirements and 6) to pursue five possible design modifications to improve the operator's ability to use the HPCI in pressure control, to manage suppression pool inventory/enthalpy and to recover the RWCU system in the post-transient environment.
One of the key modifications which the licensee committed to implement was a
revision to the HPCI suction transfer logic, which is the reason for the, subject amendment application.
Pennsylvania Power and Light Company had proposed to complete the modifications of the HPCI suction transfer logic in Susquehanna, Unit 2 during the refueling outage, which began September 11, 1992.
The amendment application to effect the logic change for Unit 2 was submi,tted August 19, 1992.
However, the NRC staff did not complete the review of the Unit 2 amendment application in time for the licensee to implement the modification during the fall 1992 refueling outage.
The licensee proposes to install the modification in Unit 2 during the refueling outage scheduled for Narch 1994.
The proposed logic will require that the HPCI injection valve F006 be'pen in addition to the present requirement of a high suppression pool water level in order for the automatic transfer of the HPCI pump suction from the CST to the.
suppression pool to take place.
This automatically prevents the pump suction transfer when HPCI is not required for injection to the reactor vessel.
The automatic transfer of HPCI pump suction from CST to suppression pool on low CST water level is unaffected by this logic change.
The physical change to the unit involves a relay being added to the HPCI injection valve (F006) control logic to permit transfer of the HPCI pump suction from the CST to the suppression pool on high suppression pool level only when the F006 valve is open.
The relay will be energized by an existing limit switch on the F006 valve that closes as the valve begins to open.
- 2. 0 EVALUATION The purpose of the automatic transfer of HPCI suction on high suppression pool level is to preserve the containment loading assumptions in the existing safety analysis.
Therefore, the imp'acts on these assumptions as well as HPCI's safety function were evaluated by the licensee as summarized below:
2.I HPCI Function The safety function of the HPCI system is to maintain reactor vessel inventory following a Loss of Coolant Accident (LOCA) which does not permit the use of the low pressure Emergency Core Cooling Systems (ECCS).
The proposed change is designed to ensure that this function will not be affected since the automatic transfer will occur when HPCI injection is required, based on injection valve position.
Various failures associated with the new design were evaluated, and it was determined that failure of the new logic would affect the proper alignment of the suppression pool suction valve (F042) or the F006 valve.
However, in the unlikely event of these failures or previously evaluated
- ones, the Automatic Depressurization System (ADS) will function to ensure that low pressure ECCS can provide adequate core cooling.
- Further, the new postulated failures were evaluated probabilistically, and the predicted failure rate of each valve was determined not to change significantly.
In response to a staff question, the licensee indicated that in the event of failure of automatic transfer of suction from the CST to the suppression
- pool, the transfer can still be made either manually by flipping a switch in the control room or, alternatively, can be accomplished by an operator opening the valve.
The licensee also indicated that the failure of the new logic will not interfere with the manual transfer operation.
2.2 Containment Anal sis The effect (following an HSIV closure event) of the existing TS is to require suppression pool level control by HPCI if HPCI is used for Reactor Pressure Vessel (RPV) pressure control and suppression pool water level reaches 24 feet.
This is due to the TS requirement that HPCI suction automatically switches to the suppression pool if suppression pool level reaches 24 feet.
Since a high suppression pool level is a likely occurrence following an HSIV closure event, the licensee does not use HPCI for pressure control because low quality suppression pool water would thereby be pumped into the (clean) condensate storage tank.
The result is that in the current configuration, the containment conditions are not impacted by the HPCI system since water is neither taken from the suppression pool nor added.
In the revised configuration, automatic switchover is blocked unless the HPCI injection valve is open.
In the pressure control
- mode, however, the injection valve is closed and HPCI takes its suction from the CST and returns the water to the CST.
Therefore, the containment conditions are not impacted in the revised configuration either, since no water is taken or added to the suppression pool.
Therefore, the impact of the proposed change on containment is no different than that of the existing TS, if a pipe break is postulated to maximize containment loading during the pressure control mode.
If the suppression pool water level exceeds the limit defined in TS 3.6.2. 1,
'the action statements require restoration of level within one hour or placing the reactor in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within
the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In the pressure control
- mode, the reactor is already scrammed and the unit is headed for cold shutdown.
No additional operator actions are required due to the logic modification.
The HPCI turbine has a specified limit on exhaust line backpressure and a
vacuum breaker which prevents siphoning water into the exhaust line.
These features preclude turbine operation at water levels above 26'-0".
In response to a staff question, the licensee estimated that the highest suppression pool water level that may be reached during the pressure control mode will not exceed 25'.
3.0
SUMMARY
The HPCI system has the capacity in its test mode alignment to control reactor pressure.
This function and the proposed changes to modify HPCI pump'uction transfer logic do not conflict with the primary HPCI function as an
- ECCS, do not adversely impact plant design parameters or safe operation of other
- systems, and are not detrimental to the HPCI system components.
The staff, therefore, finds the proposed change to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The 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 occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (57 FR 42778).
Accordingly, the amendment meets the el'igibility 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.
6.0 CONCLUSION
The Commission 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, (2) such
activities will be conducted in compliance with the Commission's regulations, ard (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 Contributors:
H.
R.
R.
R.I.
Razzaque Clark Goel Lobel Ahmed Date:
February 9,
1994
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