Resolution of Generic Safety Issues: Issue 94: Additional Low Temperature Overpressure Protection for Light Water Reactors (Rev. 1) ( NUREG-0933, Main Report with Supplements 1–34 )
Section 5.2. The resolution of USI A-26 affected all operating and future PWRs and required PWR licensees to implement procedures to reduce the potential for overpressure events and to install equipment modifications to mitigate such events. MPA B-04 was established by DL to track the implementation of this resolution at operating PWRs.578 Current requirements are stated in SRP11 Section 5.2.2, "Overpressure Protection," and BTP-RSB 5-2, "Overpressure Protection of Pressurized Water Reactors While Operating at Low Temperatures."
From 1979 to July 1983, 12 pressure transients were reported; of these, 2 events at Turkey Point Unit 4 on November 28 and 29, 1981, exceeded the TS limit (415 psig below 355F) by about 700 psig and 351 psig, respectively.591, 592 The overpressurization transients at Turkey Point were the first events to occur at an operating PWR that exceeded the TS limits since the staff resolved USI A-26. The events were identified to Congress as Abnormal Occurrences because they involved a major reduction in the degee of protection to the public health or safety.
The number of overpressure transient events, specifically the two instances at Turkey Point, suggested potential weaknesses in the existing overpressure protection criteria or its implementation that warranted further investigation by the staff. The two Turkey Point events resulted from one overpressure mitigation system (OMS) channel being out for maintenance and the other (redundant) channel being disabled by undetected errors during the first event and from undetected equipment malfunctions during the second event. These events were reported in AEOD/C401.792
Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the reactor vessel. Failure of the reactor vessel could make it impossible to provide adequate coolant to the reactor and result in a major core damage or core melt accident. This issue affected the design and operation of all PWRs.
(a) Amend the STS and the SRP to require each licensee to identify the criteria used to determine if and when the low temperature overpressure protection (Ltop) system setpoints need to be adjusted to account for the irradiation induced embrittlement of the reactor vessel.
(b) Make more use of the relief valves in the RHR for Ltop by raising the setpoint for the auto closure of the isolation valves.
(c) Amend the STS to allow no plant operation in the "water solid" condition with either train of the Ltop system out of service.
(d) Amend the STS to allow no plant to operate in the "water solid" condition with an SI pump in service.
(e) Require the Ltop system to be fully safety grade.
(f) Require reactors to upgrade their TS to the STS for Ltops.
Before 1979, there were 30 reported PWR events where the RCS pressure/temperature violated TS. After 1979, following changes to operating pro‚cedures and the implementation of OMS, there were 2 reported events of overpressure excursions at low temperature. Between 1979 and the time of this evaluation, PWRs had accumulated approximately 250 RY of operating time. Therefore, the expected frequency
of overpressure excursion events was 0.01/RY. There was concern expressed for including 1979 experience since increased operator awareness of overpressure problems may have biased the results; however, the small number of installed OMS would have also permitted more overpressure events and it was concluded that the 1979 data should be included.
The reactor vessel and weld materials have toughness properties which are defined by the nil ductility transition reference temperature, RT(ndt). The higher the copper and nickel content, the higher the RT(ndt). The RT(ndt) also increases with fluence or the cumulative exposure of the vessel to neutron irradiation. The probability of vessel failure due to a pressure spike is a function of the initial temperature (T) in relation to the RT(ndt) or the RT(ndt). It is also a function of the initial pressure and the change in pressure.
The probability of vessel failure was estimated using the Vessel Integrity Simulation Analysis (VISA) code790 based on the assumed values of pressure, temperature, and RT(ndt). The code was a Monte Carlo technique which results in no probability of failure based on a reasonable number of runs if the expected flaw size distribution (which predicts small probabilities for a large 1/4 thickness flaw) is used. Therefore, to get an estimate, the probability of a 1/4 thickness flaw was assumed to be 1.
The vessel failure probability calculated with this probability of a 1/4 thickness flaw was then reduced by a factor of 2250 to adjust the results to the expected flaw size distribution. This factor was obtained by ratioing the results given by two VISA runs: (1) assuming the expected flaw size distribution, but a very high copper and nickel content in order to force calculated failures, and (2) assuming the probability of a 1/4 thickness flaw to be 1 and the same very high copper and nickel contents. The adjusted estimate was then multiplied by 6 to account for the assumed 6 welds on the reactor vessel beltline.
Based upon a review of the previous overpressure events prior to 1978, it was found that 30% reached a peak pressure that was between 1,100 psia and 2485 psia. In another 5% of these events, the peak pressure was between 950 psia and 1200 psia. In the remaining 65%, pressure was prevented from exceeding 950 psia by operator actions. Thus, a series of VISA code runs were made at 2485 psia, 1200 psia, and 950 psia to obtain the probability of reactor vessel failure as a function of T RT(ndt).
Two types of reactor vessels were analyzed. The first is represented by the Oconee 3 vessel with 0.20% copper and 0.63% nickel content. The second is represented by a vessel with high copper (0.35%) and nickel (1%) contents.
Based on the information available on 38 PWRs, the copper and nickel content of 14 of these reactors was higher than in the Oconee 3 vessel. Thus, 38% of the operating PWRs or 17 reactor vessels are assumed to be similar to the "High" plant and the remaining 30 vessels are assumed to be similar to Oconee 3.
At the midlife of the vessels, the fluence is estimated to be 8.5 x 1018 neutrons/cm2. This fluence converts to a RT(ndt) of 267ºF for the "High" vessel and 231ºF for the Oconee type vessel using the methodology contained in Regulatory Guide 1.99, Rev. 2. If it is assumed that the starting temperature is 110ºF and the T RT(ndt) value is about 150ºF for the "High" vessel and 120ºF for the Oconee type vessel. These values of T RT(ndt) result in the probability of failure as follows:
|Probability of Failure Per Event|
|Peak Pressure (psia)||Oconee Vessel||"High" Vessel|
|2485||1.5 x 103||2.2 x 103|
|1200||7.0 x 107||7.0 x 106|
|950||<1.0 x 109||<1.0 x 109|
The predicted probability of failure at 2485 psia peak surge at the end of life fluence (1.4 x 1019 trons/cm2) for the Oconee vessel is 2.6 x 103 and 2.7 x 103 for the "High" type vessel. The average failure frequency for the Oconee type vessels was calculated to be 4.5 x 106 failures/RY and 6.6 x 106 failures/RY for the "High" type vessel. It was assumed that a failure of the reactor vessel will result in a core-melt accident with a probability of 1. The implementation of the possible solutions would reduce the frequency of an Ltop occurring but not the probability of a vessel failure, given the occurrence of an Ltop event. The frequency of Ltop events was expected to be reduced by a factor of 10 for the solutions proposed. This reduction in Ltop occurrence results in a core-melt frequency reduction of 4.05 x 106/RY for the Oconee type vessels and 5.9 x 106/RY for the "High" vessels.
The core melt accident resulting from a Ltop failure of the pressure vessel is expected to result in a S1D accident sequence as defined in WASH 1400.16 The S1D sequence is a small break LOCA with failure of the emergency coolant injection. The S1D sequence results in releases with the associated probability for the following release categories, given a core-melt.
|PWR-1 = 0.01||PWR-5 = 0.0073|
|PWR-3 = 0.2||PWR-7 = 0.8|
The whole body man-rem dose was obtained by using the CRAC Code64 assuming an average population density of 340 persons per square mile (which is the mean for U.S. domestic sites) in an exclusion area from a one half mile radius about the reactor out to a 50 mile radius about the reactor. A typical midwest meteorology was also assumed. Based upon these assumptions, the following whole body man-rem doses result from the following categories.
|PWR 1 = 5.4 x 106 man-rem||PWR-5 = 1.0 x 106 man-rem|
|PWR 3 = 5.4 x 106 man-rem||PWR-7 = 2.3 x 103 man-rem|
Utilizing the reduction in core melt frequency, the probability per release category, and the whole body dose consequence factor, the public risk reduction was calculated to be as follows:
|Public Risk (man-rem/RY)|
|Release Category||Oconee 3"||"High"|
For the average remaining plant life of 26 years, the averted public risk is 120 man-rem/reactor and 174 man-rem/reactor for the Oconee 3 and "High" type of vessels, respectively. Based on a reactor vessel population of 47 (of which 17 are in the "High" vessel classification), the expected value of averted public risk is 6,560 man-rem.
Industry Cost: The industry costs are dominated by the costs associated with upgrading the OMS; principally, upgrading the PORVs to safety grade. PNL estimated64 that valve backfit labor costs are $27,200/plant based on 12 man wks/plant at $2,270/man wk. This includes management review, QA control, licensing review, and engineering for the backfit. Material requirements are 2 safety grade PORVs and 2 instrumented (for automatic actuation) block valves, each costing $25,000. Incremental material costs such as piping, supports, hardware, etc., beyond those associated with initial installation of the safety grade PORVs and instrumented block valves at a plant were estimated at $50,000. The cost for the safety analysis was estimated to be $50,000/plant. A Class III License Amendment for the valve upgrade was estimated to be $4,000. Therefore, the implementation cost was estimated to be $237,000/plant. Other industry costs for analysis, TS changes, and test procedure changes were expected to total $190,000/plant. The total industry cost for implementation of the possible solution was estimated to be $260,000/plant or $12M for all 47 plants. The operation/maintenance costs were not expected to significantly increase over routine plant costs for operation and maintenance.
NRC Cost: The NRC costs were estimated to be $38,000 for development of a resolution and $10,000/plant for review of the overpressure mitigation provisions. These costs were expected to total $0.5M for the affected plants.
Total Cost: The total cost associated with the possible solution to this issue was estimated to be $(12 + 0.5)M or $12.5M.
Based on a potential public risk reduction of 6,560 man-rem and a cost of $12.5M for a possible solution, the value/impact score is given by:
The upgrading of the OMS or Ltop system to safety grade (safety related or important to safety) does not by itself assure improved system reliability. Hence, the benefit of making the OMS safety grade to assure a higher proba‚bility of successful mitigation of overpressure challenges may not be rea‚lized. A greater benefit may result from more stringent procedures and tech‚nical specifications, e.g., not permitting water solid operation without both channels of the OMS in operation. Thus, it may be possible to decrease the number of overpressure incidents and better assure OMS operation without hard‚ware changes, in which case, the major plant cost contributor would be eliminated.
With the elimination of hardware changes, it was estimated that procedural and TS changes could be developed for $190,000/plant or $0.9M for all 47 plants. NRC costs were estimated to be 60% of the hardware related costs. Thus, without the hardware changes and assuming the same potential risk reduction, the value/impact score is given by:
The frequency of overpressure events may be higher than the above estimate which was based solely on the number of events that have occurred. Other failure modes are possible, with failure of the PORV a prime example. The frequency of events that initiate overpressure transients and would chal lenge the OMS is probably unchanged from the pre-1979 level (about 0.13/RY). However, the OMS prevents these events from becoming overpressure events. Even if the unavailability of the OMS were 0.01/demand, the frequency of overpressure excursions would be increased by only 10% from the estimate that was used.
The analysis assumed that a brittle failure of the reactor vessel will always result in a core melt accident. However, depending upon the break type and size, the amount of decay heat generated by the fuel and the location of the vessel failure, the ECCS may be capable of keeping the core covered and thereby prevent fuel melt. The assumption that a vessel failure produces a core melt resulted in some overestimation in this analysis.
There is no additional ORE associated with the implementation of the possible solution to this issue, except the 20,000 man-rem/accident averted to clean up after an Ltop induced core melt. This amounts to 3 man-rem per "High" plant and 2.1 man-rem for the Oconee 3 type plant, for a total of 114 man-rem for all plants.
Averted accident costs based upon the present worth of $1,650M for cleanup and power replacement costs over a ten-year period amount to $0.25M/plant for each "High" plant. For the Oconee 3 type plants, the averted accident costs are approximately $0.17M/plant. Thus, the total averted cost for all plants was estimated to be $10.4M.
Ltop events which do not result in reactor vessel damage would still require an engineering analysis and inspection to assure vessel integrity. Assuming the cost of such an analysis to be $25,000/event and an average outage of 5 days for which replacement power must be purchased at $300,000/day, the Ltop frequency reduction would result in cost savings of approximately $0.4M/plant or total averted costs of $18.6M.
Based on the potential risk reduction and the value/impact score, this issue was given a high priority ranking; its resolution was later combined with that of Issue 70 so that one set of requirements could be issued to licensees.1287 In resolving this issue, the staff issued Generic Letter 90-061290 which required a revision to the TS for overpressure protection at the affected plants. This action was a follow-up to the resolution of USI A-26 and represented a backfit based on a new interpretation of
exiting requirements for some licensees and CP holders with W and CE reactors. The regulatory analysis for this resolution was reported in NUREG-1326.1291 Thus, this issue was RESOLVED and new requirements were established.1292