ML20246N854
| ML20246N854 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/17/1989 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C311-89-2020, NUDOCS 8903280056 | |
| Download: ML20246N854 (7) | |
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l GPU Nuclear Corporation NggIgf Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:
March 17, 1989 C311-89-2020 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 I
Gentlemen:
Three Mile Island Nuclear Station, Unit 1 (TMI-1) l Operating License No. DPR-50 Docket No. 50-289 TMI-1 Level 1 Probabilistic Risk Assessment (PRA)
In our letter of December 7, 1987, transmitting the TMI-1 Level 1 PRA, we indicated our intention to use the results in ongoing work. This letter is to i
apprise you of our efforts along those lines and to provide a general status summary of our consideration of the recommendations contained in the report (see enclosure).
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Our current estimates of the relative importance of various contributors to I
calculated core damage frequency (CDF) have shifted significantly since the PRA was published.
For example, based on the results of testing in late 1987 and further analyses which were submitted to the NRC on May 5 and August 5,1988, we have concluded that loss of control building ventilation is not a major contributor to calculated CDF as reported in the PRA.
In addition, a re-examination of five key fire scenarios in light of Appendix R modifications to the plant have lead us to conclude that fires are not a major contributor to calculated CDF. We have also discovered some discrepancies in the PRA that tended to understate the importance of some contributors.
For example, for the seismic contributor, all support system states were not totally quantified for all ground motion acceleration values in the study.
These have now been quantified with a resultant increase in the relative importance of the seismic contributor. Other more minor corrections to selected valve failure rates and high energy line break initiating event frequencies have tended to shift the I
relative importance of other contributors slightly.
l Taking all of the above factors into account, we estimate a net decrease in calculated CDF of about a factor of 2, and conclude that fires and losses of control building ventilation are not major contributors as reported in the PRA.
l Also, a number of modifications to plant systems are being evaluated that would have a further beneficial effect on calculated CDF and would also shift the relative importance of various contributors.
These items are discussed in the enclosure.
8903280056 890317 PDR ADOCK 05000289 P
PDC i
i GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation g6 It
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T 2-March 17, 1989 C311-89-2020 We anticipate that some aspects of the PRA will be' revisited and potentially modified in response to Generic Letter 88-20, Individual Plant Examinations for Severe Accident Vulnerabilities (IPE).
Until NRC staff guidance is finalized on all of these issues, and they are examined as part of the IPE, we do.not plan to avise the_PRA report itself.
Sincerely,
. D.
ki l-Vice President and Director,'TMI-1 HDH/DJD:2020 l
l cc:
R. Hernan, USNRC W. Russell, USNRC, Region I J. Stolz, USNRC F. Young, USNRC, TMI-1 Enclosure l
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Nr ENCLOSURE GENERAL STATUS
SUMMARY
OF CONSIDERATION OF TMI-1 PRA RECOMMENDATIONS 3
This status summary is intended to provide e broad overview of status and does not detail every activity underway in response to the recommendations. The-summary description of the recommendations provided below were excerpted from the TMI-1 PRA Executive Summary Report, pages 3-3 through 3-8.
Recommendation Sumniary Control Building Ventilation System.
Since failures of the control Euilding ventilation system contribute to 43% of the total core damage frequency from internal events, several actions are recommended to better understand this problem, improve the reliability of the system, and improve the operator's ability to cope with system failures.
Status Summary Testing in the Fall of 1987 and further analysis (results submitted to NRC on May 5, 1988, C311-88-2010) have shown that control building temperatures would not exceed values that would render equipment inoperable for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after loss. This information has been used to update the PRA model and to recalculate CDF (core damage frequency). The contribution to calculated CDF is now much less than
.1%.
Therefore, GPUN dees not plan to pursue CBVS recommendations any further.
Reactor Coolant pump Seals.
Because RCP seal leakage and failure following loss of seal injection and seal cooling are important in many core damage scenarios, a better understanding of this issue is important e.nd improvements to these important support cystems should be sought.
l Status Summary GPUN follows industry activities on RCP seals and strives to use the latest proven technology when replacing seals.
Key components in systems that support seal cooling, such as Intermediate Closed Cooling Water, Make-up/ Seal Injection, and Instrument Air have been given increased attention.
For example, ICCW pump discharge check valves were inspected and repaired in the last refueling outage to improve j
reliability of that system, and a test was conducted to confirm local air accumulator capacity for seal injection supply valve MUV-20, and ICCW valves IC-V3 and IC-V4 in June 1987. The Loss of Instrument Air procedure is being upgraded to provide guidance on maintaining seal cooling. Operators will be trained on the revised procedure.
In addition, the Instrument Air system is planned to be upgraded to i 1
include an additional compressor, air dryer, and filters to improve reliability. Other procedures that impact seal cooling have been reviewed and determined to be adequate.
Fires.
The fire hazard scenarios, which were significant contributors to the core damage frequency in the TMI-1 PRA, should be examined more carefully to confirm the validity of the assumptions about which cables and other equipment are damaged. All Appendix R modifications that have been completed to date and recovery actions currently in procedures should be included in the PRA model.
Status Summary Five key fire scenarios have been re-examined in light of Appendix R changes. As a result, fires are now estimated to represent on the order of only 2% of the total calculated CDF vs. approximately 16%
reported in the PRA.
o Onsite Electric Power.
Failures in the onsite electric power system are significant contributors to core damage frequency.
Several vulnerabilities and potentlal improvements have been identified.
Status Summary Alternative condition monitoring methods are being investigated which could greatly reduce the TMI-1 DG unavailability if used in lieu of the extensive tear-down schedule prescribed by the DG vendor.
Concurrence by the DG vencor that the selected alternative meets all inspection requirements is needed prior to implementation.
A potential change to existing valve line ups in the diesel generator starting air system is under investigation as a way to improve starting air availability. The use of a TMI-2 diesel generator as an alternate AC source for TMI-1 is being considered, use of which would further improve the reliability of the onsite system.
Offsite Electric power. The ability to restore offsite power after an extended loss could be jeopardized by the design of the switchyard in which power for air compressors and breaker heaters comes from the switchyard itself.
In cold weather, a station blackout could result in the breakers becoming inoperable after some period of time, as the SF. gas cools down.
Summary Status Two additional small diesel generators, separate from the plant emergency diesel generators, have been installed. These will mitigate this scenario.
Decay Heat Removal, Cl sed Cooling Water, and River Water.
Combinations of unavailability or failure of components in these systems (or associated power supplies) contribute significantly to core damage frequency. This is due largely to the strict separation of the trains, which produces many pairs of train A and B failures. 3
Summary Status The feasibility of system crossties has been investigated with the preliminary conclusion that only decay heat river water crossties would provide a large enough improvement in reliability to be potentially cost beneficial. This modification is being considered for possible implementation. The impact of maintenance on decay heat river water pumps has been assessed. The current practice of taking the pumps out of service when desilting is a significant contributor to decay heat river water pump unavailability.
A practice to put in screens when desilting to protect divers from the pump suction is being considered as a way of reducing system down time and improving availability.
High pressure Injection. The HPI system and several operator actions associated with it appear as important contributors to core damage frequency.
Certain aspects of the HPI system design should be considered for possible improvement.
Summary Status Changes to move the controls for the HPI recirculation valves (MU-V-36 and 37) from the back panels to the control console, and to relocate the HPI low flow alarm annunciators, along with revisions to the alarm response procedure are under consideration. These changes would reduce the likelihood that operators would fail to reestablish recirculation flow after throttling HPI and thus reduce the likelihood of pump damage and consequent loss of RCP seal injection.
In the interim, increased emphasis has been placed on this human action in the operator training program.
The possibility of operating with HPI suction crosstie valves open is being considered as a means of improving system reliability.
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A modification to the power supply for the "B" HPI oil pumps is planned that will assure these pumps are supplied from the same source as the respective HPI pump. This will reduce the chances of pump failure if power is lost as identified in the PRA.
o LOCA Outside the Reactor Building (Event V-Sequence?.
Although this sequence is not a major contributor to core damage frequency at TMI-1, it could be reduced even further.
Summary Status Changes to torque switch settings for DH-V-4A & B are planned to be implemented that will improve the ability of the valves to be closed against a higher differential pressure.
Procedural enhancements to help operators cope with an Event V-sequence are being considered. _ _ _ _ _ _ _ _ - _ _ _ _ _
.-m...
'o' preventive Maintenance. Preventive maintenance is important-for ensuring.
.the reliable performance of components and systems.
However, the time that a component or system is out of service for preventive maintenance is also-one contributor to.the unavailability of the system.
In the case of~some systems at TMI, this contribution is significant.
For example, desilting-the intake screen and pump house causes a large portion of the, unavailability of the river water pumps, and the yearly overhaul of the emergency diesel generators'significantly increases the time that the diesels are unavailable during TMI-1 operations. We ' recommend that the preventive maintenance program, pol.icies, and practices be reviewed and revised, when necessary, to achieve the highest possible system availability (which means minimizing the sum of all of the contributors to-unavailability).
Summary Status Desilting is being addressed as described.under Ds;.ay Heat River Water above. Diesel Generator maintenance is being addressed as described under Onsite Power above. The PRA is being used as one input to guide development of a Reliability Centered Maintenance (RCM) Program.
o Operator Actions. Many operator actions are important in the TMI-1 PRA and contribute significantly to reducing the calculated core damage frequency.
However, the failure of the operators to successfully perform certain actions contributes to core damage in a portion of the scenarios.
Failure to switch over from injection to recirculation after a LOCA~is the dominant source of recirculation failure.
Summary Status Various options for resolution of this concern are being considered.
Increased emphasis has been placed on this human action in the operator training program.
Failure to provide HPI pump minimum-flow recirculation was discussed elsewhere with potential system improvements..If these system improvements are not feasible, then improvements in training and procedures are in order to improve the reliability of this' human action.
Summary Status See discussion of High Pressure Injection above.
Failure to initiate HPI cooling is the most significant cause of failure of the HPI core cooling mode. The human action analysis involved should be examined for any actions that increase the reliability of HPI cooling' initiation.
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Shmmary Status A study of this-issue'has. concluded that automatic initiation of HPI
. cooling is not practical and that existing control room indications and training emphasis are appropriate. A survey to determine any operator reluctance to initiate HPI is under development. The results of the survey, and further rev.iew of the human action analysis will be used to.
decide on a resolution of this recommendation.
In many scenarios, recovery of failed or unavailable systems is important to preventing core damage.
Some examples are recovery of offsite power or a diesel generator after a station blackout, recovery of. river water. systems after a loss of river water (intake screen
. clogging),. recovery of control. building ventilation, and recovery of decay heat removal systems.
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Summary Status 1
Recovery of offsite power or a diesel generator is planned to be resolved pursuant to the recommendations on Onsite Electric Power described above.
Procedural changes for "acovery of river water are being considered.
Maintenance practices and stocking of spare parts for decay heat removal systems have been reviewed. Current practices were deemed adequate.
Further procedural changes for recovery of control building ventilation are not being pursued for the reasons described earlier.