Resolution of Generic Safety Issues: Issue 126: Reliability of PWR Main Steam Safety Valves ( NUREG-0933, Main Report with Supplements 1–35 )
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DESCRIPTION
Background
The reliability of PWR main steam safety valves (MSSVs) was identified as a generic issue in a DSRO memorandum on March 19861047 in which it was noted that individual PWR plant FSARs assume credit for MSSV functional capability to provide overpressure protection for the secondary system. This protection includes the ability of the MSSVs to achieve full ASME Code rated discharge flow at the design setpoint, to relieve in a stable manner, and to re-close at the design reseat pressure. However, in accordance with IE Information Notice No. 86-05,1048 two representative ASME Class 2 MSSVs from the Seabrook plant did not meet the ASME Code requirements when subjected to full-flow testing at Wyle Laboratories. The Wyle tests were originally intended to determine the proper vent stack size to be used with the MSSVs at the Seabrook plant.
The staff noted1049 in the Wyle test report that one of the Seabrook valves was tested at the factory ring-setting at a pressure greater than 1325 psi. This pressure was approximately 7% greater than the valve setpoint. In this test, the valve achieved a full-lift disc travel of approximately one inch. However, when both valves were tested at the factory ring-settings and pressures corresponding to the ASME 3% accumulation pressures, their discs only traveled approximately 50% of the full lift necessary to develop the rated steam flow. To proceed with the vent stack testing, the guide-ring settings on both MSSVs were adjusted down approximately 130 notches. At the lower MSSV ring-settings, full lifts within the specified ASME 3% accumulation were achieved and the vent stack tests were completed.
In accordance with Supplement 1 to IE Information Notice No. 86-05,1048 MSSV testing by Wyle Laboratories for the Vogtle (PWR) plant also indicated the as-shipped MSSV ring-settings would not achieve the full-lift travel within the ASME 3% accumulation requirement for full relief capacity. The typical result was 75% of the full lift. The proper ring-settings to achieve full lift were significantly different than the ring-settings of the Seabrook MSSVs.
While inspecting the Millstone Unit 3 MSSVs, licensee personnel found that the ring-settings were not set to the valve vendor's recommendations. After an investigation, it was determined that the valve vendor had not reset the rings following functional testing. The cause was attributed to a vendor procedural deficency.
The licensee (Pennsylvania Power & Light) for the Susquehanna BWR Units 1 and 2, recognizing the valve designations described in IE Information Notice No. 86-05,1048 became concerned that the Susquehanna valves were the same type as those tested for Seabrook. However, it was later found that the Susquehanna valves were not the same model as the Seabrook or Vogtle valves. Because the Susquehanna valves also contained adjusting rings that determine the flow capacity and blowdown, Supplement 1 to IE Information Notice No. 86-051048 was addressed to all reactor facilities holding an operating license or a construction permit.
The PWR MSSV ring-setting problems described above are similar to the improper ring-setting problems encountered in the 1982 EPRI primary safety valve tests completed in response to Item II.D.1 of NUREG-0737.98 NRC verification of the adequacy of all operating PWR plants primary Class 1 safety valve ring-settings are currently being pursued under MPA F-14; the ring-settings of the operating BWR plants Class 1 MSSVs and MSSRVs are not being pursued under MPA F-14. Many BWRs licensed since the TMI-2 accident have MSSRVs with adjusting rings. Because of NSSS vendor equipment qualification requirements, these valves have been full-flow tested and the rings adjusted to assure full lift. However, a few of the older BWRs may have MSSRVs that have not been full-flow tested and their ring settings may be based on extrapolation from smaller valves.
For future plants, changes were incorporated in 1985 to the ASME Code requirements for Class 1 valves (which include PWR primary safety valves, BWR MSSVs, and BWR MSSRVs) to require full-flow testing of prototype valves. These changes should resolve future concerns for Class 1 valves. Because the Class 2 PWR MSSVs on most PWRs (while not necessarily having the same vendor) are at the upper end of the valve size range, the general capacity certification may have been by the ASME coefficient of discharge method. Therefore, certification by full-flow testing of the PWR MSSVs may not have been performed.
The staff has raised the Class 2 PWR MSSV ring-setting concerns for these Class 2 valves with the ASME Code Committee. The NRR proposal, which would require full-flow testing of the Class 2 PWR MSSVs, passed the ASME Main Boiler and Pressure Vessel Code Committee with no negative and no abstention ballots.1050 Future revisions to the ASME Code Class 2 valve requirements are not sufficient to resolve the ring-setting concerns or the need for full-flow testing of the existing PWR MSSVs on operating plants.
Based on the above discussions, this issue is a generic issue for the Class 2 MSSVs in operating PWR plants. It is also the staff's opinion that the BWR Class 1 valves, which have in general been full-flow tested, may have similar ring-setting problems, but the problems may exist only for a few of the older BWR plants and may therefore be best classified as a plant-specific issue for BWR plants.
Safety Significance
PWR plant FSARs assume credit for MSSV functional capacity to provide overpressure protection for the secondary system. This includes the ability to achieve full ASME Code rated discharge at the design setpoint for the secondary system. Inadequate MSSV relief capacity might therefore result in overpressurization of the secondary system, a degradation of secondary heat removal from the primary system, or MSSV instability. MSSV instability (valve chatter) could result in stuck-open MSSVs that could result in excessive secondary system blowdown (loss of secondary system inventory) and overcooling of the primary system. Additional equipment failures and/or operator errors in combination with the above cases are evaluated herein to assess the potential safety significance and potential public risks associated with improper Class 2 MSSV ring-settings in PWR plants.
Possible Solution
To assure valve rated capacities for the PWR Class 2 MSSVs and the capability to function and properly reseat, it may be necessary to conduct full-flow tests of the PWR MSSVs. Changes to the ASME Class 2 safety valve requirements, similar to the ASME changes already made for the ASME Class 1 safety valves, are currently being pursued by the staff and the ASME Code Committee. The staff has requested Regional inspections to verify adequate flow capacity of the MSSVs and proper ring adjustments.1054
PRIORITY DETERMINATION
Evaluation of this issue includes potential MSSV failures (stuck-open) from inadequate MSSV flow capacities and improper ring settings that may result in SGTRs. The staff's evaluations of SGTRs (as initiating events) are reported in NUREG-0844,681 which constitutes the proposed resolution of USIs A-3, A-4, and A-5. The following conditions (cases) that may result from improper MSSV ring settings will be considered:
Case 1: Improper MSSV ring settings that result in inadequate MSSV lifts may result in degraded secondary side heat removal capabilities and lead to overpressurization of the primary side RCS.
Case 2: Improper MSSV ring settings may lead to MSSV instabilities causing valve chatter and MSSV failures to close. Failure to close (stuck-open) of the MSSVs can lead to overcooling of the RCS. Case 2a involves failure of a single MSSV to close. Case 2b involves failure to close of multiple (two or more) MSSVs, with various combinations of SGTRs induced by the rapid pressure reduction resulting from the stuck open MSSVs.
Case 1: RCS Overpressurization (Inadequate MSSV Lift)
The frequency of pressurizer overpressure events in operating reactors is 3 x 10-2/RY.307 For the purposes of this analysis, we will attribute all the pressurizer overpressure events to inadequate MSSV lift capacities. This assumption will bound any subset frequency of degraded secondary side heat removal events that may have resulted from improper MSSV ring settings.
Frequency Estimate
Assuming at least one PORV is operable, and that it has lifted during the initiating pressurizer overpressurization transient involving MSSV inadequate lift capacity, we assume a 10% chance that the primary safety valves will also lift. If the primary safety valves fail to close (10-3/demand), a small-break LOCA frequency of (3 x 10-2)(10-1)(10-3)/RY = 3 x 10-6/RY is estimated. If the high pressure safety injection fails (2.8 x 10-3/demand) or the long-term cooling system fails (1.2 x 10-3/demand), the potential severe core damage frequency is 1.2 x 10-8/RY.
If the primary safety valves close, but the PORV fails to close (10-2/demand) and the operator fails to close the PORV block valve (10-1), the small-break LOCA frequency is estimated at (3 x 10-2)(10-1)(10-2)(10-1)/RY = 3 x 10-6/RY. Again, failure of the HPI or failure of the long-term cooling system (2 x 10-3 + 1.2 x 10-3)/demand results in an estimated core damage frequency of 1.2 x 10-8/RY.
Should the pressurizer overpressure transient that is attributed to inadequate MSSV lift capacities lift the PORV and not lift the primary safety valves (9 x 10-1/demand), the PORVs might fail to close (10-2/demand). Failure of the operator to block the failed-open PORV (10-*/demand) and failure of the HPI (2.8 x 10-3/demand) or failure of the long-term cooling (1.2 x 10-3/demand), is estimated to result in a core damage frequency of (3 x 10-2)(9 x 10-1)(10-3) x (4 x 10-3)/RY = 10-7/RY.
The combined core damage frequency for the above described Case 1 primary overpressure transient events is approximately 10-7/RY.
Case 2a: RCS Overcooling (Single MSSV Stuck-open)
Frequency Estimate
Failure to close (stuck-open) of a single MSSV following a successful reactor trip with MSSV challenges is equivalent to a small steam line break. This initiating transient is estimated to have a frequency of 10-2/RY. Because failure to close of a single MSSV would be equivalent to small steam line break flows of approximately 2% to 10% of the plants' rated steam flow, no conditional SGTRs are assumed for Case 2a. The initiating transient and conditional failure frequencies estimated to lead to potential severe core damage are as follows:
| Sequence | Frequency |
|---|---|
| Transient (One MSSV Stuck-Open) | 10-2 |
| Fail to Isolate SG | 1.2 x 10-3 |
| Fail HPI | 2.8 x 10-3 |
| 3.4 x 10-8 | |
| Transient (One MSSV Stuck-Open) | 10-2 |
| Fail AFW | 10-4 |
| Fail HPI or Long-term Cooling | 4 x 10-3 |
| 4 x 10-9 | |
| Transient (One MSSV Stuck-Open) | 10-2 |
| PORV Opens | 8 x 10-1 |
| PORV Fails to Close | 10-2 |
| Failure to Block PORV | 10-1 |
| Fail Long-Term Cooling | 1.2 x 10-3 |
| 10-8 |
The sum of the Case 2a core damage frequencies resulting from the small steam line break transient (failure to close one MSSV) is approximately 4.8 x 10-8/RY.
Case 2b: RCS Overcooling (Multiple MSSV Failures)
Transients that involve failure of more than one MSSV to close, given a successful reactor trip, are estimated to have a frequency of 10-3/RY. The initiating transient involving multiple MSSV failures to close is assumed equivalent to an MSLB. The top events following the equivalent MSLB transient in the systemic event tree involve SGTRs, failure to isolate the steam generators, failure of the AFW system, failure of the HPI system, opening of the PORVs, failure to block the PORVs, and failure of the long-term core cooling system (RHR). The systemic event tree contains 40 event sequences, sixteen of which are judged to potentially result in severe core damage. The Case 2b frequency estimates are described below.
Frequency Estimate
SGTRS Given Multiple MSSVs Stuck Open: The conditional probabilities for single and multiple SGTRs, given an MSLB (considered herein as multiple MSSVs stuck open), were estimated in NUREG-0844.681 The estimated conditional probabilities for the range of potential SGTRs are
| 1 SGTR/MSLB | 3.0 x 10-2 |
| 2 to 10 SGTRs/MSLB | 1.5 x 10-2 |
| > 10 SGTRs/MSLB | 3.0 x 10-3 |
The severity and timing of secondary side pressure reductions following multiple (two or more) MSSV failures (stuck open) could vary significantly and, therefore, alter the potential effects of SGTRs, given various multiples of MSSV failures. Therefore, this evaluation includes events with multiple MSSV failure sequences and no SGTRs, in addition to multiple MSSV failures with the above range of potential SGTRs.
Steam Generator Isolation: An estimate of the probability (1.2 x 10-2/event) of failure to isolate the steam generators, given multiple MSSV failures and no SGTRs, was obtained from NUREG/CR-249776 and INPO 82-025.1051 For multiple MSSVs stuck open, these two documents estimated the failure to isolate the steam generators as a factor of 10 greater than the value (1.2 x 10-3/event) used for small steam line breaks. The factor of 10 increase was to account for potentially higher functional degradations that may result from the more rapid pressure reductions.
Transients involving SGTRs are difficult to handle. Because there may be potential delays in diagnosing the combined equivalent MSLB and SGTRs and possible improper operator actions, given the occurrence of combined multiple MSSV failures and SGTRs, we estimate an additional factor of 10 increase in the failure to isolate the steam generators (1.2 x 10-1/event) for these combined events.
ECCS Unavailabilities: Depending on the numbers of SGTRs induced by the equivalent MSLB, the success of the HPI and long-term cooling can be jeopardized to various degrees by the continuous loss of primary coolant flowing out the failed-open MSSVs. Because the MSSVs are outside containment and typically inboard of the MSIVs, the primary and ECCS coolant inventory loss would not be available for recirculation from the containment sump. Thus the RWST inventory could be depleted and the ECCS could fail by cavitation. The times and probabilities that operators might fail to take action to depressurize the RCS to atmospheric pressure before the RWST is exhausted (as a function of the number of SGTRs) were obtained from NUREG-0844.681
The basis for the estimated ECCS unavailabilities, given a combined MSLB with single and multiple SGTRs, is that the faulted steam generator is boiled dry even in the case of double-ended SGTRs of up to five steam generator tubes.860 Under these conditions, the leaks are flashed to steam on the secondary side masking the unexpected primary coolant flow into the steam generator. If the operator fails to get local visual observations or radiation monitor readings from the steam line break, it is possible that the SGTRs would remain unnoticed and the situation would be diagnosed as a steam line break. It was thus judged possible that the primary pressure could remain high and uncontrolled leakage of primary coolant directly to the atmosphere could continue for hours. Without proper operator actions to decrease the primary coolant loss, the ECCS recirculation would be lost. The estimated ECCS unavailability for the following ranges of SGTRs are
| Condition | ECCS Unavailability/SGTR |
|---|---|
| 1 SGTR | 10-3 |
| 2 to 10 SGTRs | 10-2 |
| > 10 SGTRs | 5 x 10-1 |
When combined with ECCS functional unavailabilities, the conditional ECCS unavailabilities are
| Condition | Total ECCS Unavailability/SGTR | |
|---|---|---|
| HPI | Long-Term ECCS | |
| No SGTRs | 2.8 x 10-3 | 1.2 x 10-3 |
| 1 SGTR | 3.8 x 10-3 | 2.2 x 10-3 |
| 2 to 10 SGTRs | 1.3 x 10-1 | 1.1 x 10-2 |
| > 10 SGTRs | 5.0 x 10-1 | 5.0 x 10-1 |
AFW Unavailability: The unavailability of the AFW systems in operating reactors range from 10-3 to 10-5 per demand. In the resolution of Issue 124, "Auxiliary Feedwater System Reliability," the staff contemplates that licensees will achieve and maintain AFW system reliability in operating reactors at levels between 10-4 to 10-5 per demand. This goal is consistent with SRP11 Section 10.4.9. Therefore, for the purpose of this analysis, we will use an AFW unavailability of 10-4/demand.
PORV Estimates: In NUREG/CR-249776 and INPO 82-025,1051 the probability of lifting the primary system PORV, given a steam line break (equivalent multiple MSSVs stuck open), was estimated at 0.8. The value of 0.8 is also used in this analysis.
The PORV failure to close (stuck open), given that it has opened, is estimated at 10-2/demand. This estimate is consistent with Issue 70, "PORV and Block Valve Reliability." In Issue 70, the estimated probability that the operator would fail to close the PORV block valve, given that the PORV stuck open, was 5 x 10-2 /demand. For the purpose of this analysis and the additional complexities and stresses that may be involved in the combined multiple MSSV failures and SGTRs, this analysis (Case 1 and Case 2) uses a value of 10-1/demand for failure of the operator to isolate (block) a stuck-open PORV. Because of the attention focused on blocking the PORVs since the TMI event, the above PORV estimates are judged to be conservative.
Event Frequencies: The Case 2b event sequences described in the previous section that may lead to potential severe core damage are tabulated below. The event sequences not expected to lead to core damage are omitted.
Case 2b(0): Multiple MSSVs Stuck Open (No SGTRs)
| Transient | 10-3 |
| Failure to Isolate SG | 1.2 x 10-2 |
| HPI Failure | 2.8 x 10-3 |
| 3.4 x 10-8 | |
| Transient | 10-3 |
| AFW Failure | 10-4 |
| HPI Failure | 2.8 x 10-3 |
| 2.8 x 10-10 | |
| Transient | 10-3 |
| AFW Failure | 10-4 |
| RHR Failure | 1.2 x 10-3 |
| 1.2 x 10-10 | |
| Transient | 10-3 |
| PORV Opens | 8 x 10-1 |
| PORV Stuck Open | 10-2 |
| Failure to Block PORV | 10-1 |
| RHR Failure | 1.2 x 10-3 |
| 9.6 x 10-10 | |
| TOTAL: | 3.5 x 10-8 |
Case 2b(1): Multiple MSSVs Stuck Open (1 SGTR)
| Transient | 10-3 |
| SGTR (1) | 3 x 10-2 |
| Failure to Isolate SG | 1.2 x 10-1 |
| HPI Failure | 3.8 x 10-3 |
| 1.4 x 10-8 | |
| Transient | 10-3 |
| SGTR (1) | 3 x 10-2 |
| AFW Failure | 10-4 |
| HPI Failure | 3.8 x 10-3 |
| 10-11 | |
| Transient | 10-3 |
| SGTR (1) | 3 x 10-2 |
| AFW Failure | 10-4 |
| RHR Failure | 2.2 x 10-3 |
| 6.6 x 10-12 | |
| Transient | 10-3 |
| SGTR (1) | 3 x 10-2 |
| PORV Opens | 8 x 10-1 |
| PORV Stuck Open | 10-2 |
| Failure to block PORV | 10-1 |
| RHR Failure | 2.3 x 10-3 |
| 5.3 x 10-11 | |
| TOTAL: | 1.4 x 10-8 |
Case 2b(2): Multiple MSSVs Stuck Open (2 to 10 SGTRs)
| Transient | 10-3 |
| SGTR (2 to 10) | 1.5 x 10-2 |
| Failure to isolate SG | 1.2 x 10-1 |
| HPI Failure | 1.3 x 10-2 |
| 2.3 x 10-8 | |
| Transient | 10-3 |
| SGTR (2 to 10) | 1.5 x 10-2 |
| AFW Failure | 10-4 |
| HPI Failure | 1.3 x 10-2 |
| 2 x 10-11 | |
| Transient | 10-3 |
| SGTR (2 to 10) | 1.5 x 10-2 |
| AFW Failure | 10-4 |
| RHR Failure | 1.1 x 10-2 |
| 1.7 x 10-11 | |
| Transient | 10-3 |
| SGTR (2 to 10) | 1.5 x 10-2 |
| PORV Opens | 8 x 10-1 |
| PORV Stuck Open | 10-2 |
| Failure to block PORV | 10-1 |
| RHR Failure | 1.1 x 10-2 |
| 1.3 x 10-11 | |
| TOTAL: | 2.3 x 10-8 |
Case 2b(10): Multiple MSSVs Stuck Open ( > 10 SGTRs)
| Transient | 10-3 |
| SGTR ( > 10) | 3 x 10-3 |
| Failure to Isolate SG | 1.2 x 10-1 |
| HPI Failure | 5 x 10-1 |
| 1.8 x 10-7 | |
| Transient | 10-3 |
| SGTR ( > 10) | 3 x 10-3 |
| AFW Failure | 10-4 |
| HPI Failure | 5 x 10-1 |
| 1.5 x 10-10 | |
| Transient | 10-3 |
| SGTR ( > 10) | 3 x 10-3 |
| AFW Failure | 10-4 |
| RHR Failure | 5 x 10-1 |
| 1.5 x 10-10 | |
| Transient | 10-3 |
| SGTR ( > 10) | 3 x 10-3 |
| PORV Opens | 8 x 10-1 |
| PORV Stuck Open | 10-2 |
| Failure to Block PORV | 10-1 |
| RHR Failure | 5 x 10-1 |
| 1.2 x 10-9 | |
| TOTAL: | 1.8 x 10-7 |
Consequence Estimate
The core-melt sequences for Case 1 involve failed-open safety/PORV valves in the primary side and failure of the emergency core cooling systems. These sequences are similar to the WASH-140016 small-break core-melt sequences. In accordance with NUREG/CR-22281052 and assuming that containment fan coolers/sprays are available to limit steam pressure buildup in containment, a potential for containment failure by a hydrogen ignition and burn is assumed. Based on the Zion and Indian Point PRA studies, we use a 3% probability of containment failure by hydrogen burn. This containment failure mode is representative of a PWR Category 3 type of release. We also assume a 1% probability for failure to isolate the containment (PWR Category 5 release). If the containment does not fail by hydrogen burn or isolation failure, it will be assumed to fail by basemat melt-through (PWR Category 7 release).
Core-melt from accident sequences described in Case 2a and Case 2b(0) will progress similar to the TML sequence described in NUREG/CR-4752.1053 Assuming containment fan coolers/sprays available, the containment responses, failure probabilities, and releases are similar to the above Case 1 sequences.
Based on the above, the weighted average core-melt release for Cases 1, 2a, and 2b(0) is 1.7 x 105 man-rem/core-melt. Cases 2b(1), 2b(2), and 2b(10) involve multiple stuck-open MSSVs and SGTRs. The dominant contributors to core-melt conditions are potential inadequate secondary side heat transfer (failure to isolate the steam generator) and SGTRs that result in primary side inventory losses with increased ECCS unavailability. The path for the primary coolant SGTR LOCAs is into the secondary side steam generator and out the stuck-open MSSVs. Thus, the primary coolant inventory losses are not recoverable from the containment sump for recirculation. In time, depending on the number of SGTRs, the RWST is depleted and the ECCS fails. The decreasing probabilities of increasing numbers of SGTRs are overcome by the increasing unavailabilities of the ECCS water supply (depleted RWST).
The fission product release from the melting core follows the same path as the primary coolant inventory loss and, therefore, bypasses the containment barrier (containment bypass). Under these conditions, the probability of containment failure is assumed at unity.
For the above accident progressions, it may be possible to terminate or mitigate the accident results by dispatching operators to the MSSVs (which are outside containment) to attempt to manually close the stuck-open MSSVs. However, this analysis does not consider such heroic recovery actions on the part of the plant operators.
The release consequences for fewer than 10 SGTRs could be modeled as a PWR Category 4 failure to isolate containment release. For greater than 10 SGTRs, the PWR Category 2 or 3 containment rupture releases could be assumed. However, possible dilution or a water seal in the assumed unisolated steam generator, plate-out in the piping and steam generator structures, and aerosol attenuations will tend to reduce the fission product release. An example of possible reductions in the fission product release can be inferred from the MB-2 Aerosol Attenuation Tests described in NUREG/CR-4752.1053 The purpose of the MB-2 tests was to model the TMLB accidents in which the core-melt is accompanied by dried-out steam generators, where the high pressure primary side steam could cause a SGTR to vent to the secondary side, open the safety relief valve, and release radioactivity directly to the atmosphere. Results from the MB-2 tests (which should be considered preliminary) showed aerosol attenuation factors ranging from approximately 3 to 11. The tests did not include conditions that would capture additional reductions that might result from temperature, steam condensation on the secondary walls, plate-out, or possible water sealing and dilution by the secondary side.
Based on the above discussion, the PWR Category 5 releases should bound the releases for the MSSV/SGTR events. The PWR Category 5 release is a failure to isolate containment similar to the PWR Category 4 release, except containment sprays are available. The Category 5 release (106 man-rem/core-melt) is approximately a factor of 3 less than the Category 4 release and approximately a factor of 5 less than the Category 2 or 3 releases.
For the purpose of this analysis, we assumed that all MSSV partial lifts and 50% of the stuck-open MSSVs will be eliminated by the possible solutions described earlier. The estimated changes in core-melt frequencies and the consequence (risk) reductions for each case are tabulated and summed below in Table 3.126-1.
The generic conditional release doses used in this analysis are based on the fission product inventory of a 1120 MWe PWR, meteorology typical of a midwest site, a surrounding uniform population density of 340 persons per square mile within a 50-mile radius of the plant, an exclusion radius of one-half mile from the plant, no ingestion pathways, and no evacuation. Therefore, the estimated change in risk is representative of the hypothetical generic PWR plant with a large, dry containment structure and is not representative of any specific plant.
Cost Estimate
No cost estimates are provided for this issue. The issue solution1054 involves plant-specific inspections to ensure compliance with the original licensing bases of the plants. Therefore, the costs (though probably small) will not affect the resolution of this issue.
Table 3.126-1
| Case | MSSV Failures * | SGTRs | Delta (CM/RY) | Dose/CM (man-rem) | Annual Plant Risk Reductions (man-rem/RY) | 30-Year Plant Risk Reduction (man-rem/Ry) |
|---|---|---|---|---|---|---|
| 1 | partial lift | - | 10-7 | 1.7 x 105 | 1.7 x 10-2 | 0.5 |
| 2a | (1) SO | - | 4.8 x 10-8 | 1.7 x 105 | 8.2 x 10-3 | 0.3 |
| 2b(0) | (M) SO | - | 3.5 x 10-8 | 1.7 x 105 | 6.0 x 10-3 | 0.2 |
| Subtotal: | 1.8 x 10-7 | 3.0 x 10-2 | 1.0 | |||
| 2b(1) | (M) SO | 1 | 7.0 x 10-9 | 106 | 7.0 x 10-3 | 0.2 |
| 2b(2) | (M) SO | 2 to 10 | 1.2 x 10-8 | 106 | 1.2 x 10-2 | 0.4 |
| 2b(10) | (M) SO | > 10 | 9.0 x 10-8 | 106 | 9.0 x 10-2 | 2.7 |
| Subtotal: | 1.1 x 10-7 | 1.1 x 10-1 | 3.3 | |||
| TOTAL: | 2.9 x 10-7 | 1.4 x 10-1 | 4.3 |
* (1) SO: One MSSV Stuck Open
* (M) SO: Multiple MSSVs Stuck Open
CONCLUSION
Based on the estimated reduction in core-melt frequency of 2.9 x 10-7/RY and the risk reduction of 4.3 man-rem/reactor, this issue could be ranked as a low priority safety issue. However, because improper MSSV ring-settings can result in inadequate MSSV relief capacities, this issue is a matter of possible non-compliance with the licensing basis of the plants and, therefore, is more appropriately classified as a Licensing Issue. Regional inspections to verify adequate flow capacity and proper ring adjustments of the MSSVs are necessary to ensure licensee compliance with existing requirements; the request for Regional assistance in this regard was made by NRR.1054 Therefore, this Licensing Issue has been resolved.
REFERENCES
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