ML19241B689
| ML19241B689 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/05/1979 |
| From: | Flugger F FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19241B688 | List: |
| References | |
| NUDOCS 7907230080 | |
| Download: ML19241B689 (46) | |
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Testimony of Frederick George Flugger Relating to ASLAB Memorandum and Order of April 5,1979, on Electrical Grid Stability and Emergency Power Systems (Questions A2, B1, B2, B3, and B4 of ALAB 537) 1 My name is Frederick George Flugger. I am Supervisor, Plant Licensing, Power 2 Plant Engineering Department for Florida Power and Light Company. f:y education 3 and professional qualifications appear in the Nuclear Regulatory Commission's 4 record of the St. Lucie Unit 2 (Unit 2) proceeding following Tr.1310.
5 The purpose of this testimony is to respond to questions A2, B1, B2, 83, and 6 B4 in Section II 'f the Appeal Board's Order of April 5, 1979. fly affidavit 7 of March 31, 1978 is relevant to the issues raised by the Appeal Board. It 8 is provided as Attachmeat A and is hereinafter referred to as the Flugger 9 Affidavit.
10 ihis testimony demonstrates that the Unit 2 onsite AC power system design is 11 in full compliance with NRC requirements, that the design basis events evaluated 12 in the PSAR provide a ; roper basis for the design of Unit 2 and that Unit 2, 13 as designed, can acceptably accomodate the postulated loss of all AC event.
14 Before responding to the Appeal Boards's questions, it is appropriate that a 15 few basic considerations be discussed to place the responses in proper perspec-16 tive.
17 First, consider the frequency of loss of the electrical grid. FPL nuclear 18 operating history suggests a frequency of outage of about 4 x 10 -I per year 19 for the FPL grid. Although there is little comparative historical data readily
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e 1 available, I believe that the relative difference in reliability between the 2 FPL system, based on its historical data, and other grids associated with the 3 general population of nuclear plants is probably not more than a factor of about 2.
- It must be noted that as the FPL system evolves during construction of Unit 2 5 and during its operation, any difference in reliability that may be inferred 6 from FPL's operating history to date will be reduced or eliminated. The 7 testimony in response to Appeal Board questions Al and D discusses the substan-8 tial actions that have been and will be taken to improve the reliability of 9 the FPL grid.
10 In any event, from a nuclear plant design standpoint, the difference implied 11 by historical data is very small when compared to nuclear plant design 12 reliability levels. The relatively small reliability differences that may 13 be associated with peninsular and nonpeninsular grids will not affect the 14 design of Unit 2 engineered safety features (ESF's).
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15 Second, the probabilities associated with nuclear plant design and operation 16 are not normally precisely quantifiable because of uncertainties that may 17 exist in the data, the depth of experience that comprises the data base, and 18 applicability of the data to a specific design. However, these probabilities 19 can normally be specified fairly accurately within a range of values.
20 The NUREG-75/087 (reference 1) -6 10 /10- guideline value has been and should 21 be associated with events whose consequences are comparable to 10 CFR Part 100 22 guidelines. The postulated loss of all AC event is not a 10 CFR Part 100 type 23 event. It results in a very slow and tolerable transient that can be accom-24 udated by the existing Unit 2 design.
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.u , f~<al 1 The time to restore AC power is pivotal to the evaluation of the postulated 2 loss of all AC event. FPL's historical grid data demonstrates that the 3 duration of loss of offsite power is very short-lived. FPL operating exper-4 ience from January 1972 to present indicates a mean time to restore offsite 5 power to FPL facilities of less than 1/2 hour.
6 Third, an unprotected loss of coolant accident (LOCA) does not result from the 7 postulated loss of all AC event. There is no failure of the reactor coolant 8 pressure boundary associated with this event. Areactorcoolantpump(ICP) 9 seal can only yield very small and acceptable leak rates. (See the resoonse 10 to question B2 infra.) Unit 2 has more than adequate capability to remove 11 decay heat, which is necessary to accommodate the postulated loss of all AC 12 event. There is sufficient condensate to provide steam generator makeup for 13 at least 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, the auxiliary feedwater pump is steam driven, auxiliary 14 feedwater pump control and at.1!iary feedwater system valves are DC powered, 15 and the steam generators have sufficient inventory to allow the operator about 16 55 minutes to actuate auxiliary feedwater before steara generator dryout occui s.
17 Uith these considerations in mind we can procede with the responses to the 18 Appeal Board's questions.
19 Question A2 20 For its part, the first paragraph of GDC-17 appears to establish an unattainable 21 set of conditions for electrical power systems generally. It reads as follows 22 (emphasis added):
23 An casite electric power system and an offsite electric pcwer system 24 shall be provided to permit fbnctioning of structures, systems, and 25 components important to safety. The safety functions for each system 26 (assumino the other svstem is r,ot functioning) shall be to provide 27 sufficient capacity and capability to assure that (1) specified accept-23 able fuel design limits and design conditions of the reactor coolant 29 pressure boundary are not exceeded as a result of anticipated operational
['"O f O"' h 1 occurrences and (2) the core is cJoled and containment integrity and 2 other vital functions are maintained in the event of postulated accidents.
3 This paragraph requires that an assessment of the sufficiency of the offsite 4 power system start with the assumption that the onsite system is not function-5 ing. That assessment must then consider the effect of " anticipated operational 6 occurrences." But loss of the offsite power system itself may reasonably be 7 considered to be such an occurrence. The parties should, therefore, explain 8 how the St. Lucie Plant can comply with the literal requirements of this 9 paragraph as written. If it cannot, they should attempt to justify the situa-10 tion in terms of the purpose of the requirement.
11 Response 12 The Poard's question cites a possiole literal interpretation of GDC 17 that 13 contravenes the intent of this design criterion. The intent of GDC 17 is 14 provided in a straightforward manner by the language of proposed GDC 24 and 15 39 issued for guidance by the Atcmic Energy Commission on July 10, 1967 before 16 GDC 17 was adopted in its present form. Their language states:
17 GDC 24 18 "In the event of loss of all offsite power, sufficient alternate sources 19 of power shall be provided to permit the required functioning of the 20 protection systems."
21 GDC 39 22 " Alternate power systems shall be provided and designed with adequate 23 indepencency, redundancy, capacity, and testability to permit the 24 functioning required of the engineered safety features. As a minimum, 25 the onsite power system and the offsite power system shall each, 26 independently, provide this capacity assuming a failure of a single 27 active component in each power system."
28 The intent of these criteria is simply to ensure that an onsite AC source be 29 p ovided adequate to backup the offsite AC source. This philosophy is embodief 30 in current industry standards. IEEE std. 308-1974 (reference 2) embodies the 31 concept of the " preferred" (offsite) power supply system and the "standhy" 32 (onsite) power supply system. The functions of these systems are cited in the 33 standard as follows:
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I "The preferred power supply shall furnish electric energy for the shutdown 2 of the station and for the operation of emergency systems and engineered 3 safety features."
4 "The standby power supply shall provide electric energy for the operation 5 of emergency systems and engineered safety features during and following 6 the shutdown of the reactor when the preferred power supply is not i available."
f' RG 1.?2 (reference 3) endorses IEEE 308-1974, with a few nonrelevant exceptions, 9 as an adequate basis for complying with GDC 17. In other words, the NRC Staff 10 interprets and requires compliance with GDC 17 in a manner which does not 11 conterplate the literal interpretat on suggested by the Appeal Board's question.
i 12 Unit 2, as designed, complies witr the accepted interpretation and intent of 12 GDC 17.
14 Finally, it is appropriate to note the relationship between 10 CFR 50.36 and 15 2,ppendix A to 10 CFR Part 50. The latter provices design criteria while 16 the former imposes operational restrictions. The NRC regulations 17 at 10 CFR 50.34 require that a safety analysis be performed to assess t' 18 ability of the facility to meet its design objectives. The safety 19 provides the basis for establishing limiting Hitions for operation s .
20 d which provide the minimum functional capability or performance levels recuir 21 for safe operation of the facility. 10 CFR 50.36(c)(21 The LCO's become part 22 of the facility's operating license. Therefore, it is pertinent to note that 23 continued Unit 2 operation with both onsite diesel generators inoperable would 24 constitute a violation of Technical Specifications. RG 1.93 (referonce 1) 25 states that the limiting condition for operation (LCO) is met "When all the 26 electric power sources required by GDC 17 are available." If both diesels were 27 inoperable the plant's operating license would restrict operation in accordance 28 with RG 1.93 as follows: eo ha rU ', 0~)
1 "If the available onsite a.c. electric supplies are two less than the 2 LCO, power operation may continue for a period that should not exceed 3 two hours... If no onsite a.c. supply is restored withir the first 4 two hours of continued power operation, the unit should be brought to 5 a cold shutdown state within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
6 Thus, the Unit 2 operating license will contain conditions in the form of 7 these Technical Specifications to minimize the risk of exposure to continued 8 plant operation with bc+b diesels inoperable.
9 Question B1 10 As wt it, the iikelihood of loss of all AC power at St. Lucie may be 11 exprest s the product of two factors: (1) tie probability that there will 12 be an of tai, power failure involving the FPL network generally or the Midway 13 substation in particular and a resulting loss o' station power -- which 14 probability seems based on historical events, to lie in the range 1.0 to 0.1 15 per year; and (2) the probability that neither of the two onsite AC power 16 systems (diesel generators) will start. The rrobability that any one diesel 17 generator will fail to stagt on demand is taken by the staff to be one per 18 hundred demands, i.e., 10- 25/.
19 If these figures are accurate, then the corgbined pgobability for the " loss of all AC power" scenario is in the range 10 " to lo - per year. 26/ In this 20 21 regard, the staff's Standard Review Plan for Nuclear Power Plants sets forth 22 numerical guidelines for determining whether an event "resulting from the 23 presence of hazardous materials or activities in the vicinity of the plant" 24 should be considered in designing the plant (i.e., whether it is a " design 25 basis" event). 27/ Under these guidelines, events with a realistically calcu-26 lated probabiliiy value of at least 10- per year (or 10- per year for a 27 conservative calculation) must be so considered.
28 The "lcss of all AC power" sequence is not precisely within the categcry of 29 events contemplated by the Standard Review Plan. However, its ultimate 30 result -- assuming that power is not timely restored -- is an unprotected 31 loss of coolant accident, the consequences of which are likely to exceed the 32 guidelines of 10 CFR Part 100. We do not understand why this sequence of events 33 (i.e., loss of offsite power combined with failure of diesels to start), which 34 appears to have a probability well 7 ove the guideline values, should not be 35 taken into consideration in the design of the plant. 28/ The parties are to 36 address this point, setting forth their reasons for adhering (if they do) to a 37 contrary position.
33 -25/ Fitzpatrick Affidavit of June 12, 1978, p. 4 Also see Regulatory Guide 39 1.108, Section B.
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I -26/ This conclusion further assumes that the failure of two diesel generators 2 to start would be statistica'ly independent events, an assumption which 3 leads to the lowest likelihood of combined failure, and which might be 4 nonconservative if there exists the pctential for common failure moces 5 for the onsite systems.
6 27/ NUREG 75/087, Section 2.2.3, paragraph II.
7 j8/ We he 2 accepted the Standard Review Flan guideline values as reasonable 8 in i ither case. Public Service Electric and 3as Comoany (Hope Creek 9 Units 1 and 2), ALAB - 429, 6 NRC 229, 234 (1977).
10 Response 11 The question pertains to two different but complementary nuclear plant design 12 concepts, namely, the frequency of occurrence of an event (events / unit of time) 13 and the reliability of an Engineered Safety Feature (ESF) (failure to function 14 when called upon to do so). Before the question of whether the postulated 15 simultaneous loss of offsite and onsite AC power sources should be included in 16 the design basis can be addressed, i is necessary to discuss the concepts of 17 event frequency and ESF reliability.
18 Event Freauency 19 fiany types of events have been considered in the design of Unit 2. These may 20 be generally categorized into several major groups as follows:
21 1. Events of moderate frequency leading to no significant radioactive 22 releases from the facility and no violation of fuel design limits.
23 2. Infrequent events which have the potential for small radioactive 24 releases from the facility and small amounts of fuel failure.
25 3. Events of low probability, Design Basis Accidents (DBA), which are 26 required by 10 CFR Part 50 to establish the performance requirements 27 of ESF's and are used in evaluating the ability of tre facility to 23 comply with 10 CFR Part 100 guidelines. '
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l-1 Un.t 2 design bases are the " specific functions to be performed by a structure, 2 systen or component of a facility, and the specific values or ranger of values 3 chosen for controlling parameters as reference bounds for design" (10 CFR 50.2).
4 They are developed by analyzing limiting events, i.e., other events of the type 5 analy.?ed are less severe. This approach provides reasonable a.esurance that 6 the facility has adequate capability to accommodate unanalyzed events.
7 The prcbability of occurrence of non-design basis initiating events that may 8 produce results more severe tian DBA's is considered so small that these events 9 are not incorporated into the plant design. Section 2.2.3 of NUREG-75/087 10 (reference 1) pros , . a 10'0/10- guideline for " design basis events resulting 11 from the presence of hazardous materials or activities in the vicinity of the 12 plant." In using this guideline it should be understood that:
13 1. This guideline is appropriate for events that have a potential for 14 yielding offsite exposures that equal or exceed 10 CFR Part 100 15 guidelines.
16 2. There is little experience available to provide a statistical basis 17 for quantifying with precision the probability of occurrence of 18 initiating events which have su.n low probability. Thus considerable 19 engineering and scientific judgment is involved in determining whether 20 or not a given event should be included in the design basis.
21 3. If an event which was considered to be outside the design bases did 22 occur, it would not necessarily produce consequences that are catas'ro-23 phic or exceed 10 CFR Part 100 guidelines. Considerable engineering, 24 design evaluation and operating experience has been accumulatcd sir.;e 25 the first commercial light water reactors went into operation around 26 1960. This significant experience base has demonstrated that a nuclear
/o 'qi t %) Ub 1 facility has substantial inherent capability to acceptably accomodate 2 a broad spectrum of events.
3 4. The Unit 2 design philosophy utilized is specifically directed at 4 providing assurance that the likelihood of events with conseqJenCes 5 more severe than DBA's is extremely low. The facility is designed, 6 built and operated so that it will, with a high degree of reliability, 7 minimize the likelihood of an accident. Despite the care taken to 8 prevent accidents, the design provides for reliable protection devices 9 and systems designed to detect and cope with transient and off-normal 10 conditions. ESF's provide protection to the public*even in the event 11 of the occurrence of severe accidents of !cw probability, i.e. , DBA's.
12 Finally, throughout the facility's lifetime nuclear plant operatina 13 experience is continually monitored and assessed by the NRC to determine 14 whether design or procedural modifications are required.
15 ESF Reliability 16 Reliability of an ESF is simply the probability of performing its safety 17 function when called upon to do so. Although increased material and component 18 quality level, testing and maintenance will improve reliability, above 19 certain levels substantial cost and testing commitments result in minimal 20 increases. Because of this, the concept of redundancy is employed to achieve 21 acceptaole reliability levels in nuclear plant designs. Enormous increases 22 in system reliability can be achieved through redundancy because the overall 23 reliability becomes the product of the reliabilities of the independent 24 systems. The use of the single failure criterion in nuclear plant design is 25 based en the concept of redundancy. The objective of this criterion is to 26 prevent any single failure frc preventing the accomplishment of a safety jo .
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I function. This criterion is imposed by Appendix A to 10 CFR Part 50, and is 2 a fundamental premise upon which all nuclear safety related designs are based.
3 Loss of offsite electrical AC by itself is protected against by an onsite AC 4 system that employs, in accordance with GDC 17, redundant and independent 5 diesel-generators. The postulated loss of all AC power following the loss of 6 offsite AC violates the single failure criterion in that it requires the 7 failure of both redundant and independent diesel generators. For this reason 8 the sequence of events postulated by the question is not a design basis event.
9 Nevertheless, as discussed below, the postulated loss of all AC event can be 10 accommodated for some period of time.
11 The appropriate probability for evaluation of the postulated loss of all AC 12 event is the probability during any cne year of having loss of all AC power 13 combined with the probability of not restoring AC by time "T" which is given by:
14 P(T) = P(A) P(B) P(C) P(D) P(E) P(F) 15 where: P(T) = probability of not restoring AC power by time "T" 16 P(A) = probability of loss of offsite power 17 P(B) = probability of loss of first diesel 13 P(C) = probability of loss of second diesel 19 P(D) = probability that offsite power is not 20 repaired and returned to service by time "T" 21 P(E) = probability that first diesel is not 22 repaired and returned to service by time "T" 23 P(F) = probability that second diesel is not 24 repaired and returned to service by time "T" 25 The restoration of AC probability terms, P(D), P(E) and P(F) can be developed 26 in a straightforward manner. Let P(T) be the probability that AC is not 27 restored at time "T", P(T+aT) this prcbability at a finite later tire "T+aT",
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1 and C aT the repair probability during the time interval AT (v.here C is a 2 constant). Then, 3 P(T+2T) = P(T) - (1-C AT) 4 which in the limit as ai approaches zero is given by:
5 d P(T) = -C P(T) 6 dT 7 whose solution is.
-CT 8 P(T) = e 9 The equation for P(T) can be used to mathematically represent P(D), P(E) and 10 P(F). Examination of historical data allows determination of the time constant 11 "C" for each of these probability terms. An evaluation of FPL system data from
-I 12 1972 to present indicates that a time constant of 1.6 hr is appropriate for 13 P(D). (See the response to question B3 infra.) St. Lucie 1 and Turkey Point
-I 14 diesel generator outage data indicate that a time constant of 0.16 hr is 15 appropriate for both P(E) and P(F). (See the response to question B3 infra.)
16 The probability of loss of offsite power P( A) is obtained in a similar manner.
17 If A is the grid failure rate (number of failures in a period of time "t",
18 such as 0.1 failures per year), then e -t is the probability that offsite 19 power will not be lost and the probability that offsite power will be lost 20 can be expressed as P(A) = l-e \D .
21 Application of the exponential representation for the probability of restoration 22 of power, a frequency of loss of offsite power of 0.1 per year, a diesel 23 generator failure per demand of 10 -2 , and time constants of 1.6 and 0.16 hr-20 for offsite and onsite power restoration respectively yields:
-5 exp (-1.92T) 25 P(T) = 10 a
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I which can be used to quantify the probability for not returning AC power 2 by time "T" as a function of "T". The results are:
3 Duration of loss of AC Probability of Having a Total Loss 4 "T" (hours) of AC Power that Lasts "T" Hours, P(T) 5 0 1 x 10 -5
-6 6 1 2 x 10
-6 7 1.2 1 x 10 8 2 2 x 10' 9 2.4 1 x 10-
-8 10 3 3 x 10
-9 11 4 5 x 10 12 If a loss of offsite AC power event frequency of 1.0 per year were assumed instead of 0.1, then a value of P(T) of1x 10
-6 will be reached at 2.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 13
-7 14 and 1 x 10 at 3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
15 The evaluation of historical FPL onsite and offsite failure data demonstrates 16 that the probability of a continued loss of AC power decreases significantly 17 with the duration of the loss. If, as suggested by the question, the 10-6 /10-18 criterion were to be applied to the postulated loss of all AC event, then 19 evaluation of a period exceeding about 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (rounding off 1.2 and 3.6 20 hours) is not required since the probability of not restoring AC power within 21 that time period is acceptably low.
22 For the reasons described in the response to Question B2 below, Unit 2 can 23 be maintained in a safe shutdcwn condition without AC power for a the period 24 well in excess of the time likely for restoration of AC power.
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,, i ,,v 1 Question B2 2 In line with the beve discussion, the testimony is to analyze events that 3 would occur between the " loss of all AC power" and the violation of either 4 the fuel design limits or the design conditions of the reactor coolant 5 pressure boundary (or any portion thereof). In particular, the parties 6 should, if possible, reconcile their differing responses to question B.l(b) 7 of our March 10, 1978 order, 29/ or, if not, point up precisely where the 8 disagreements 11e.
9 29/ [ References fn 24 reproduced below: ]
10 Applicant suggests that the first safety related failure encountered 11 would be excessive core heating due to the loss of water from the 12 condensate storage tank, and that this would occur about 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after 13 the loss of AC power (Flugger Affidavit of March 31, 1978, p. 3).
14 The staff's judgment is that the first failure would be that of a 15 primary pump seal at about one hour after tne loss of AC power ---
16 resulting in a small loss of coclant accident. (Fitzpatrick Affidavit 17 of June 12, 1978, p. 11).
18 Resocnse 19 The Flugger Affidavit filed in response to the Appeal Board's order of March 10, 20 1978 concluded that there was a sufficient volume of condensate storage to allow 21 the unit to maintain hot standby conditions #or at least 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />; the spent 22 fuel storage pool would not require makeup fcr at least 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />; and that power 23 would be restored before any unacceptable consequences would occur. The Fitz-24 patrick Affidavit, which provided the Staff response, concurred with FPL's 25 response, but went on to suggest that a failure of a reactor coolant pump (RCP) 26 seal could potentially occur after one hour as a result of the loss of all AC 27 power. For the reasons set forth below, the difference can be reconciled and 28 Unit 2 can be safely maintained in a hot shutdown condition until AC power is 29 restored.
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1 At the outset, it is necessary to analyze the actual condition of the reactor 2 coolant pumps during the event. Upon loss of AC power the reactor will trip, 3 the RCP's will coast down and stop, and cooling water flow to the RCP seals 4 will cease. This static (pump not running) condition is much less severe than 5 the dynamic (purp running) condition discussed in the Unit 2 PSAL at section 6 9.2.2.3.1, which provides a basis for concluding that running the puTps for 7 about one hour, without cooling water to the seals,would not result in pump 8 seizure or unacceptable RCP aeal failure.
9 In order to evaluate the static performance of the RCP's under loss of all AC 10 conditions, it is necessary to briefly discuss the seal design and construction, 11 Each RCP is equipped with a seal cartridge,t.hich contains four separate seals.
12 Each of the four seals within the seal cartridge is designed to provide the 13 sealing function against full system pressure. A seal cartridge test fixture 14 is used to tully test the seal cartridge prior to installation on the RCP, and 15 the tested seal cartridge is installed as a unit. All seal components are captured 16 within the seal cartridge assembly. The carbon rings within the seal are held 17 in place by hydraulic force since the higher pressure is on the ring's cutside 13 diameter, and spring force in addition to hydraulic force holds the rotating and 19 stationary sealing faces together. Thus the RCP seal design is such that a 20 mechanism for development of an appreciable leakage path within the seal cartridge 21 under static conditions does not exist.
22 Pressure breakdcan devices are installed parallel to the first three seals. Reactor 23 coolant at a rate of 1 gpm passes thraugh these devices such that reactor coolant 24 system pressure is distributed equally across the first three seals, i.e., they 25 normally operate at about 1/3 of their design pressure. The fourth seal is subjected
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1 to a nominal backpressure and cts as a vapor barrier / backup seal during normal 2 operation. The RCP controlled bleedoff flow of I gpm/ pump is directed to the 3 chemical and volume control system.
4 Under static conditions associated with loss of all AC the terrperatt- M of the 5 fluid in the seal cartridge will attain a level above the normal seal cartridge 6 operating temperature due to the in uption of cooling water. The temperature 7 would rise from about 180 F to about 5a0UF.
8 If the postulated loss of all AC event occurs, there are two modes of seal 9 operation that may be utilized, namely, secure bleedoff flow or maintain bleedoff 10 flow. If it is assumed that the controlled bleedoff line is closed thereby 11 eliminating the normal 1 gpm flow through the seal cartridge, then only one seal, 12 the fourth, will be functional, sealing against full system pressure. The other 13 seals will see no pressure differential. However,they will autcmatically take 14 over the sealing function should the fourth seal develop a leak in excess of 1 15 gpm. The maximum outleakage would not exceed the normal 1 gpu, as flow is 16 restricted to this value by the precsure breakdown devices in parallel with the 17 first three seals. The system pressure would then be distributed equally among 18 the remaining three seals as the pressure breakdown devices become functional.
19 Should the tnird seal also malfunction, allowing leakage in excess of 1 gpm, the 20 outleakage would increase to 1.2 gpm as only two pressure breakdown devices would 21 remain functional, the third pressure breakdown device being bypassed through the 22 third seal. If the second seal also is assumed to malfunction, the first seal 23 takes over and the leakage increases to 1.7 gpm as only one pressure breakdown 24 device is functional.
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- 0 1 If the controlled bleedoff line is not closed off, pressure distribution through 2 the seals is maintained the same as for normal oper ation and the bleedoff is 1 gpm 3 per pump. Operation in this mode results in a pressure differential across the 4 first three seals of 1/3 of design and only a nominal backpressure across the 5 fourth seal. In case of malfunction of any of the first three seals, pressure is 6 distributed proportionally among the remaining seals with a corresponding increase 7 in bleedoff as stated above. Securing the bleedoff at any time will cause the 8 fourth seal to take over the sealing function.
9 Even though there are four independent seals per RCP to ensure the maintenance of 10 the sealing function,and each one is designed to seal against full system pressure, 11 there is no reason why any one of these seds would fail in the static condition.
12 The only components affected by the elevated tea cerature, are the elastomeric 13 gaskets of the scals, nan.ely the "U" cup in the normally rotating part of the 14 seal, and the "0" rings in the stationary seal segment. The "U" cups are totally 15 captured and the "0" rings are backed up by lapped seats which would maintain low 16 leak rates. All other componer ts are metallic or carbon, which are not af fected 17 by the elevated temperatures of the system. The elastcmeric components are made 10 of Ethylene Propylene or Nitrile, materials which are suitable for long operation 19 at temperatures up to 250 F without change of characteristics. Temperatures 20 above 250 F will affect the physical characteristics of the material, the extent 21 of the effect being a function cf temperature, pressure and time. The accepted 22 operating life at 300 F is in excess of 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />.
23 At the system tey erature of 550 F, the elastcmeric caterial, Ethylene Propylene or 24 Nitrile, would be subject to extrusion and hardening, i.e., gradual loss of flexi-23 bility and permanent setting in a deflected position. The reactor coolant pur?
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1 manufacturer has demonstrated, however, the sealing characteristics of the 2 elastomeric material under thermal conditions equivalent to those resulting frcm 3 the postulated loss of all AC event. Confined "0" rings of the material have 4 been used on several flanged joints of a reactor coolant pump hot test loop, 5 where they have been subjected to temperatures of 550 F for in excess of 100 6 hours during routine pump acceptance testing. The "0" rings maintained their 7 sealing capability without any problems, and as would be expected, hardening 8 and permanent setting of the "0" rings occurred. Under static conditions 9 sealing would be maintained since (i) the "0" rin9s are backed up by lapped 10 seats, (ii) the "U" cups are totally captured, and (iii) most of the harden-11 ing would occur on cooldown, rather than at the elevated temperature.
12 In summary, the RCP seal cartridge will maintain its low leakage characteristics 13 for the duration of the static loss of all AC event, and the RCP seals are 14 expected to ren.ain functional for a period of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
15 Operation of a reactor coolant pump after restoration of AC power will likely 16 result in higher than normal seal leak rates due to hardening of the 17 elastomeric materials. Thus a natural circulation cooldown to cold shutdown 18 conditions would be preferred since it would not require running of a reactor 19 coolant pump. In this regard, it is important to note that in April 1977 the 20 St. Lucie #1 reactor coolant system was borated and the plant was braucht to a 21 cold shutdown without the reactor coolant pumps running, i.e., on natural 22 circulation.
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O I Insof ar as maintenance of reactor coolant system temperature and pressure is 2 concerned, following RCP coast down, flow through the reactor coolant system is 3 maintained by natural circulation of a subcooled :1 uld. Decay heat is rejected 4 to the secondary systen through the steam generators. Steam generator safety 5 valves will limit the steam generator secondary side pressure.
6 The plant operators will start the steam turbine driven auxiliary feedxater 7 pu:rp and will locally open the steam genecator atmospheric dump valves to allow 8 the steam generator safety valves to reseat. Steam generator level vill be 9 rcestablished, at which :ime the operators will adjust auxiliary feeduater flow to 10 maintain a constant steam generator level . The reactor coolant system will then 11 be stabilized at hot shutdown conditions.
12 Due to heat Inss from the pressurizer, norr.al reactor coolant system (RCS) leakage 13 e.g. , RCP seal bleedoff flow, and secondary side liquid temperature in the steam lo generators, there will be a gradual and steady Jecay in RCS pressure and tempera-15 ture. Since the RCS pressure decays at a higher rate than temperature, the 16 reactor ccolant system will eventually, in about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, reach the saturation 17 condition. Thereafter, decay heat remval will continue by natural circulation 18 of a saturated fluid.
19 The Flugger Affidavit indicated that about 200,000 gallons of rater would be 20 required to maintain hot standby for 16 hcars, which is the minir.:z technical 21 specification limit anticipated for the Unit 2 condensate storaga tank Ho..e ve r ,
22 the condensate storage tank is norally maintained in excess of the technical 23 sr ecification linit, and has a design cap city of 400,000 gallons- Additionally, 2a there are another 1,800,000 or so gallons (design capacity) of fresh ' ater 25 storage on site at St. Lucie. It is reascnable to conclude that during the
. 4 I
[u3 l i 1 initial 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, when technical specification condensate storage is being 2 consumed, that portable pumps can be made available to replenish Unit 2's 3 condensate storage tank, and that core heat up due to lack of steam generator 6 makeup is not a real-world concern.
5 There are two safety class DC batteries installed at this facility. Each is an 6 1800 ampere-hour, on an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> basis, battery, i.e. , each can supply 225 amperes 7 continuously for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The DC lead required for remote manual auxiliary 8 feedwater operation, control rocm lighting and one channel of instrumentation 9 from the instrument busses is about 100 amperes. Thus, if in say 1/2 hour or 10 so, one battery is secured and parasitic loads are stripped frcm the on-line 11 battery, there is more than sufficient battery capacity to accommodate the 12 postulated transient. One battery could sustain the DC load for about 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, 13 the second battery could be reconnected to supply the DC load for another 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 14 or so.
15 It is also pertinent to point out that the Unit i diesels can be aligned to 16 supply Unit 2. The PSAR at Figure 8.3-1 shows a tie between the Unit 1 and 2 17 startup transformers. The tie allows the 4.16 kV busses to be tied between 13 Unit 1 and 2 so that Unit 1 diesels can be aligned to supply AC power to Unit 2 19 via this tie. There are three breaker cubicles and two breakers. The breakers 20 are normally installed so that the startup transformers supply their respective 21 unit's AC busses, and the tie between units is physically open, i .e. , the breakcr 22 is not installed. Loads would have to be stripped from the Unit 1 and Unit 2 23 tusses and the breaker in the cubicle from the Unit 2 startup transformer nust be 24 removed and installed in the 4.16 kV switchgear tie. The sequence of events has 25 been reviewed, and it has been determined that it would take tuo men about one 26 hour3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> to align a Unit 1 diesel to Unit 2.
n- l s4 7
,C, I
1 In summari, under the postulated loss of all AC event, fuel and reactor coolant 2 pressure boundary limits will not be exceeded during the probable time necessary 3 to restore AC power. There is no basis for assuming that all four redundant 4 seals on a RCP will lose function during the postulated loss of all AC event, 5 and no LOCA would result from this postulated event. Thus,10 CFR Part 100 5 guidelines are not applicable to this event.
7 Ouestion B3 8 The testimony should contain a discussion, supported by such data as is 9 available, related to the time that might be required to start a diesel 10 generator assuming it failed to respond to the iritial, auto-start signal.
11 Response 12 Should a diesel generato at a nuclear plant fail to start the unit's technical 13 specifications would require that the second diesel and offsite AC circuits la be verified oper'ble and that power operation may continue for a period not to 15 exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (reference 4). Thus, if the remaining AC power sources are 16 operable, there is no undue time-pressure constraint to return the diesel to 17 service, which would exist if all AC power were lost. Accordingly, any evalua-18 tion of the time to return a diesel to service based on historical dact would 19 likely yield a conservative estimate of the time to return a diesel gt .crator 20 to service.
21 The concept of diesel reliability should also be placed in proper perspectit a.
-2 22 A 10 probability of demand with a confidence level of 95E was demcnstrated by 23 a 300 start shop test program for a U t 1 diesel (see Unit i FSAR section 24 3.3.1.3). A successful attempt occurred if the diesel performed the sequence:
25 fast start, automatically bringing the set to full speed and voltage, immediately en T
\
, 17b 1 loading the generator to 605 of continuous raticg, and maintaining the 60$
2 load for 5 minutes. A failure in any portion of the sequence was considered 3 a failure per demand. This sequence is based on LOCA generated ESF require-4 ments< which place exacting quick start design requirements upon the diesel 5 generators.
6 Diesel generator experience at St. Lucie Unit No.1 has been reflected in the 7 Unit 2 design. There have been seven failure to start incidents at St. Lucie 8 of which only two could be categorized major maintenance items. These two g events were associated with turbocharger malfunctions, which involved repair 10 durations of about 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> and 173 hours0.002 days <br />0.0481 hours <br />2.86045e-4 weeks <br />6.58265e-5 months <br />. Four of the remaining five events 11 were corrected in less than two hours. The fif th event involved a sticky sole-12 noid and pluggage of an air starting line for c.nich restoration time was 7-2/3 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
14 The turbocharger failures were due to a momentary Jeficiency in lube oil 15 pressure during the switchover from an electric to engine driven oil pump while 16 the engine was coming to speed. Provisions have been incorporated in the Unit 2 17 design to preclude this. Specifically, the AC drivan lube oil pump will run 18 continuously and it will be backed up by a DC driven lube oil pump. Additionally, lg an idle start capability will be provided for the Unit 2 diesels, which will 20 ensure proper engine lubrication during diesel testing. To avoid corrosion 21 related problems, such as the sticky solenoid / plugged air line incident, the 22 Unit 2 diesels will have a stainless steel air start system. Since the turbo-23 charger failures resulted from a design feature that has been modified in the 24 Unit 2 design, these two data points have been omitted from the FPL data base.
25 A recent Turkey Point Diesel Generator Voltage Regulator Transformer problem
. ?
O
[t u ki N 1 was resolved by disconnecting a neutral lead, resulting in the elimination 2 of third harmonic current heating effects. Since this problem was unique to 3 the Turkey Point design and does not apply to the St. Lucie diesel generators, 4 this data point was also omitted from the data base. The repair time frequency 5 distribution based on St. Lucie and Turkey Point experience to date is as 6 follows:
7 Repair Time Frequency of Repair Time Frequency of 8 (minutes) Occurrence (minutes) Occurrence 9 10 1 111 1 10 15 1 180 1 11 21 1 217 1 12 30 1 240 1 13 37 1 258 1 14 65 1 275 1 15 70 1 390 1 16 76 1 460 1 17 77 1 503 1 18 91 1 708 1 19 94 1 1435 1 3563 1 20 The redian diesel repair time is 111 minutes and the mean is 338 minutes.
21 If each event is assumed to have equal probability of occurrence, then the 22 probability of restoration of a safety-related diesel at an FPL nuclear facility
-T -I 1 23 can be expressed mathematically by 1-e , where C is 0.16 hr 24 1/ Although it is inappropriate to include the turbocharger and voltage 25 regulator data, inclusion of these data does not alter the conclusions reached 26 in question B1 supra, i.e., evaluation of a period exceeding about 1 to 4 27 hours is not required since the probability of not restoring AC power within 28 that time period is acceptably low. Inclusion of these da 29 of 217 minutes, a mean of 1434 minutes and a C of 0.04 hrThis {a yields a median results 3
30 in an expression for P(T) of 10 ' exp (-1.63T) as ccmpared to 10 exp 31 (-l.92T), which is used in the response to question B1 suora.
' ' ' ~
1 To respond to question Cl supra the equally important issue of how long it 2 would take to restore offsite power was reviewed: FPL's history of system 3 disturbances from January 1972 to present indicates that loss of offsite power 4 to plants on the Florida Power & Light system was distributed as follows:
5 Duration Frequency of Duratica Freauency of 6 (min.) Occurrence (min.) Occurrence 7 1 1 30 2 8 8 1 31 1 9 9 1 32 1 10 13 1 40 1 11 15 1 43 2 12 17 4 53 1 13 20 2 77 1 14 22 1 15 23 1 16 The median restoration time of offsite AC power is 21 minutes and the mean 17 is 26 minutes. If each event is assumed to have equal probability of 18 occurrence, then the probability of restoration of AC power to any FPL facility 19 can be expressed nathematically by 1-e -CT , where C = 1.6 hr
-I
.n all system 20 disturbances affecting FPL's nuclear plants the diesel generators started and 21 supplied AC power for the duration of the incident.
22 In summary, FPL operating experience indicates that the duration of offsite 23 power loss is short-lived. Thus, the probability of restoring offsite or 20 onsite AC power within an hour is very high, which is reflected in the crobability 25 assessments provided in response to question B1 supra.
1 Question B4 2 Finally, in the light of the discussion of points 2 and 3 above, the parties 3 are to review possible measures for decreasing the likelihood of exceeding 4 design limits on the reactor fuel and pressure boundary under the assumption 5 that there is some time available to activate an auxiliary power source 6 subsequent to a total loss of AC power.
7 Response 8 As demonstrated by the responses to questions Bl and B2 supra, the potential 9 for exceeding design limits on the reactor fuel and pressure boundary prior to 10 restoration of AC power is acceptably low. The Unit 2 design as proposed is 11 considered acceptable and in compliance with NRC requirements.
12 Since the ability to accommodate this loss of AC event is dependent on operator 13 action during a non-design basis event, we have briefly reviewed the design 14 with regard to areas that relate to their ability to cope with the postulated 15 event. Loss of all AC power will be immediately evident to the operators since 16 the unit will trip, the low 4 k'l bus voltage and diesel failure to start alarms 17 will annunciate, and the control rcom lighting will dim to the DC lighting 18 system level, The DC system will provide power for requisite monitoring 19 instrumentation and for auxiliary f eedwater system operation. The metailed 20 actions to stabilize the unit in this mode will be reviewed prior to issuance 21 of an operating license to ensure that the operators have the capability to 22 achieve and maintain hot shutdown conditions for the duration of the loss of all 23 AC event.
7 n' i t , ,
[i o j -
References:
- 1. "flVREG-75/087, " Standard Revie,' Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," U. S. Iluclear Regulatory Commission, September 1975.
- 2. IEEE Std. 308-1974, "IEEE Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations."
- 3. Regulatory Guide 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants."
- 4. Regulatory Guide 1.93, "Availab:lity of Electric Power Sources."
/.C ! 1'0 i.'
ATTACHMENT A
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FLORIDA POWER AND LIGHT COMPANY ) DOCKET NO. 50-389
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(St. Lucie Nuclear Power )
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1 in this analysis.
2 3 RESPONSE 4 Loss of all AC power is not a design basis 5 for St. Lucie. Like all other plants, St. Lucie has 6 been designed to the single failure criterion, in 7 accordance with apolicable NRC regulations.
8 In order for a loss of all AC power to occur 9 after a loss of offsite pcwer, a double failure, i.e.,
10 the fa' lure of two independent diesels to start and 11 supply onsite pcwer, is recuired. Consequently, a 12 detailed analysis of such an event has not been 13 performed.
14 However, assumine. the hv.o.othesis in the 15 Board's question, there are two predominant safety 16 functions to be performed following loss of offsite 17 power and failure of onsite power to start; (1) removal 13 of decay heat from the reactor ccolant system and:
19 (2) removal of decay heat .from the spent fue 20 storage pcol.
21 (1) Heat from the reactor core will be 22 transferred to the steam generator by natural circu-23 laticn of reactor coolan:. Heat -: oval can then be
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1 the atmospheric steam dump valves. This process is totally 2 independent of AC powered equipment and components.
3 The feedwater will be supplied by a steam turbine 4 driven auxiliary feedwater pump, operated with steam 5 from the steam generators. The auxiliary feedwater 6 pump takes suction from the condensate storage tank 7 (CST). The CST contains a sufficient volume of conden-8 sate ",o that as so operated it would allow the unit to 9 remain at hot standby for at least 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
10 (2) The loss of offsite power and the failure 11 of onsite power to start will cause the spent fuel pool 12 cooling system to stop operation. The decay heat from 13 the stored spent fuel will cause the water temperature 14 to rise and eventually boil.
15 The water level in the pool will not require 16 make-up for at least 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
17 In view of the foregoing, FPL believes that 18 either offsite power would be restored, or onsite power 19 supplied, before any safety-related consequences would 20 occur.
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s FREDERICK G. FLUGGER STATE OF FLORIDA )
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COUNTY OF DADE )
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ATTACHiiEilT #5 (ADDENDUED An analysis was performed ondne contingency of the loss of both Midway 240 kV busses. The end result of the lass of both busses with a breaker and a half scheme is that the breakers connect ^d to the busses are open and the lines coming into the s ubstation only connect to the mid-breaker and continue on out again. Specifically at Midway, after the loss, there would be four lines that would pass through the Midway mid breakers:
- 1. St. Lucie-Midway Sherman 230 kV
- 2. Malabar-Midway-St. Lucie 230 kV
- 3. St. Lucie-Midway-Indiantown 230 kV
- 4. Malabar-Midway-Ranch 230 kV Of these four lines, one connects St. Lucie to the north, two connect St. Lucie to the iauth, and a fourth passes by with no connection to St. Lucie.
A loadflow study was perforned to test what distribution of power flow would result if the loss of both busses occurred at th< * ' me o f peak s umme r 19 8 3 load with both St. Lucie units in servir .
Two loadflows were run, (normal and with the loss of both busses) and the pertinent flows were plotted on the attached maps. These plots show that no line overloads would be expected and the St.
Lucie 240 kV bus is still connected to both the north and south.
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LE COLE 6207v FLORIDA TOWER 1 L I Gl'1 CCtWANY T T:ANCMILCION I t? T E kis uf T I Ctd S UM r1 AR Y .
S Y S l E t1 GUTAGEL L X C L U[> C D LINE CECTION MIDUAY - TO CT LUCIE T'LANT 41 240 KV Ilb1 N 1975-1970
- - - - - - - - - TIME - ----
- PART - ALL -
- DATE OFF -- -- 04 -- - O ri -- DMG ITEM - ~--- CAUSE ----
1/27<76 15:1L; O 15:15:30 co' co 30 rJ 0 tJ E FPL CREW 5/14/78 7:45: 0 7:55: 0 oo . f o: co NONE LIGHTtJING ARFESTE TOTAL OUTACES DY CAUSE CAUCE GUSTAINED MGMENlARY LIGHTNI.JG ARRESTER 1 0 FPL CREW 1 0 TOTAL 2 0 CO W
'd
[k O h $I5
LE COLE 6207Y F ORIDA fOuCh 1 LIGHT COMPANY T R Al'CMISS ION IN T E ERUP T IOt! CUMMARY .
GYCIEM OUTACLC E X C L U ll E li [l,(,'[ %
LINE SEClION MIDWAY TO CI L U (' I E PLANT 42 240 NV 1775-1970
- - - - ----- TIME ------
- PART - - ALL -
- DATE OFF O ri - ON -
DMG ITEM -- - - CAUCE ----
7/ 5/74 23:34: 0 23:34:30 cc oo:M NONE UNNNOWN 7/11/76 23:30: 0 tiOME N T AR Y UNNNOWN S/14/70 7:45: 0 7:55: O co to oo NONE LIGitTNING ARRESTE TOTAL OUTAGES LY C A U S t:
CAULE OUOTAINED M0hENTARY LIL,HTNING ARRECTLR 1 0 U ANOWr! 1 1
- TOTAL 2 1 h
LO D
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LE COLE 62079 FLORIDA POWEh ; L ICtlT C OM P A rJ Y T R Al40 M I S S I 0 rl I tJ T f f; A UP T ! O t! SilMMARY SYSTEM Ctll AGE' E)CLUDED L I t!E C E C T l a ti '
l l,41 g MIDWAY TO - F. T LUCIE f'L At:T 43 240 LV 1975-1970
- - - - - - - - - - TIME
- PART - ALL -
- DATE - OFF ON -- O ri - DMG ITEM -- - - - - CAUCE ----
L/14/7C 7:45: 0 7:55: O 0; io 00 tJOHF LICHTNING APRESTE 7/10/70 5:53:30 5:53: 45 00:ovif tJ 0 t!E U t4 h tJ 0 W H TOTAL CUTAGEO IiY CAUrE C A U :., E CUGlAINED MOME tJT ARY L I chi t!Ilt G ARRESTER I O tire h t:O u ti 1 0 TOTAL 2 0
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LE COLE 52479 ^
D FLORIDA POWER E LIGHT COMPANY TRANSMISSION INTCRRUPTION
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION MALADAR - TO - MIDWAY el 240 KV 50. N m 1975-1970
TINE -----------
- - PART - -- ALL -
- DATE - - - -- OFF -- -- ON -- -- ON -- --- DMG ITEM --- ---- CAUSE ----
6/ 5/75 O:31: O MOnENTARY Ut4KNOWN 6/24/75 6: 0: O MOMENTARY P40N-DIS. LIGHT.
7/ 2/75 22 O: O MOMEtJTARY Ut4KNOWN MOMENTARY UNKNOutJ 7/ 3/75 22:43: O 7/14/75 5:31: O MUMEt4TARY Ui4KNOllN 11/13/75 1:55: O MOMENTARY UtJNNOWt1 7/13/76 15:25:30 22: 4: O (.* 38G0 INSULATOR LIGilrNING O/ 4/76 3:10: O mot 1Ei1 T ARY UtJKNOWN 9/13/76 22:37:30 MOME t4 T ARY UtJKtJOWN 11/12/76 O:34: O MOMENTARY Ut4NNOWN 11/12/76 6: 9: O 6:10: O 0:01:00 HONE UNhNoWN 6/14/77 1:40: O mot 1ENTARY UNNNOWN 7/ 5/77 20:51: O 4:51: o g!00:00 NONE cot 4DUC10R 7/23/77 3:G3: O MOMENTARY UNI;tJOWN O/21/77 6: 5: O MOMENTARY UtJKNOWN 11/ 7/77 22: 7:30 Mt;MENTARY UNKNOWN 12/ 4/77 2:34:45 MOMENTARY UNKNOWN 2/16/70 6:37: 0 6 37:30 0 : cc!30 NONE UNKNOWN 5/14/70 7:45: 0 7:47: O 0;O2:00 NONE LIGHTNING ARRESTE .
6/ 4/70 5:46: O MOMENTARY WEATitER IN AREA O/21/70 23!4c: O MONENTARY UNKNOWN a
TOTAL OUTAGES DY CAUSE o CAUSE SUSTAINED MOMENTARY CONDtJCTOR 1 O J LIGitTNING ARRESTER 1 0
'O LIGHTNING 1 O UNNNOuti 2 14
,,s WZATilER IN AREA O 1 NON-DIS. LIGHT. O 1 16 g J
J J
LE COLE 52479 FLORIDA POWER 1 LIGHT COMPANY TRANSteISSION INTERRUPFION
SUMMARY
SYSTEM OUTnGES EXCLUDED LINE SECTION MALADAR - TO - HIDWAY 42 240 KV 63 1 M 197"5--1970
TINE -----------
- PART - -- ALL -
- DATE -- -- OFF -- -- Ott -- -- ON -- --- DMG ITEM --- - CAUSE ----
7/15/70 1:25:30 MOMENTARY UNKNOWN TOTAL OUTAGES LY CAUSE CAUSE SUSTAINED HOMENTARY UNKNOWN O 1 TOTAL O 1 5
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P LE COLE 52479 FLORIDA POWER E LIGHT COMPANY TRANSMISSION INTERRUPTIOrd
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION MIDWAY - TO - RANCH S1 240 Kv 53. 31 m.
1975-1970 TIME --------- -
- - PART - -- ALL -
- DATE -- -- OFF -- -- Ot1 -- -- ON - --- DMG ITEM --- ---- CAUSE ----
6/19/77 15:42:15 2:50: O f 2: 32 16 A M 7/ 3/77 2:37:30 VANDLALISM-NOt1Et4TARY ItJSULATOR 7/ 7/77 5:52: O MOMENTARY 7/22/77 UfJKtJOWN 5: 47' O MOMEtJTARY UNKNOWil O/10/77 2:57: O 2:57:15 00*#0:16 NONE O/10/77 14:10: O 17:15 O Noti- D I S . LIGHT.
3!05:C<, NONE X-ARM O/27/77 22: 8:45 MOMENTARY 10/ 1/77 UNhNOWrd 3:52: O MOMENTARY UNKtJOWN 10/17/77 4:54: 30 11:49: O ('.[f GO CONDUCTOR 11/11/77 23:19 ItJSUL ATOR O fiOMENTARY Ut4hNOWN 12/29/77 21*57:45 6:27: O 8:27;l6 CONDUCTOR It4SULATOR 5/14/70 7:45: O 7:40:15 0 0:03:1E NONE 6/17/70 15:10:15 9:16' O LIGHTNING ARRESTE O/ 5/70 7:24: 15 jg:g;g NONE X-ARM MOMENTARY WEATHER IN AREA O/12/70 2:57: O NOMENTARY UNNNOWN 10/15/70 6:54: O MOMENTARY 10/20/70 UNNNOWN 1: 0 O NOMENTARY UNKNOWN 10/30/70 5:43:30 MOMENTARY UNKHOWN 12/14/70 19:35:30 MOMENTARY UNKNOWN TOTAL OUTAGES DY CAUSE .
CAUSE SUSTAINED MOMENTARY X-ARM 2 O INSULATOR 2 1 L I GH T t4 ING ARRESTER 1 O Ut4KtJOWN O 10 WEATHER IN AREA O A
. p NOtJ-D I S . LIGHT. 1 1
0 f)Ib j
g VANDLALISM 1 O j a e h'fl TOTAL 7 12 V .)
m
LJ LE COLE 52479 D
FLORIDA POWER 1 LIGHT COMPANY TRANSHISSION ItJTERRUPTION SUNHARY SYSTEM OUTAGES EXCLUDED LINE SECTIOr4 INDIANTOWN - TO - HIDWAY 240 KV 24. l7 (Mi 1975-1970
TIME -----~~~---
. - PAPr -- ALL -
- DATE -- -- OFF -- -- ON -- -- ON -- DMG ITEN --- ---- CAUSE ----
4/12/76 16:21:45 MONENTARY UNKNOWN 9/23/77 5:50:30 HONENTARY NON-DIS. LIGHT.
TOTAL OUTAGES DY CAUSE
', CAUSE SUSTAINED HOMENTARY uNuNOuN O 1 NON-DIS. LIGHT. O 1 TOTAL 0 2 e
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. / e LE COLE 52479 FLORIDA PO JER 1 LIGHT COMPANY '
TRANSHIGSION INTERRUPTION
SUMMARY
SYSTCH OUTAGES C;;CLUDED LINE SECTION INDIANTOWN - Tn - PRATT WHIT JEY 240 KV 3,45 %
1975-1970
TIME -----------
. - PART ~ -- ALL -
- DATE -- - - - OFF -- -- ON -- -- ON -- --- DMG ITEM --- ---- CAUEE ----
6/17/70 15:10:45 HOMENTARY UNKNOWN TOTAL OUTAGES DY CAUSE CAUSE SUSTAINED HOMENTARY
- ') UNKtJOWN O 1 TOTAL 0 1 w
=# #
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(*yy;[;.)s b f EIfj [3 #
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LE COLE 52479 FLORIDA POWER & LIGHT COMPANY TRANSHISSIOt1 Ii1 T ERRUP T IOtt
SUMMARY
SYSTEN OUTAGES EXCLUDED LINE SECTION ,
PRATT WHITNEY - TO - RANCH 42 240 KV 90. ~l4 /%
1975-1970
TIME -----------
. - PART - -- ALL -
- DATE -- -- OFF -- -- ON -- -- ON -- --- DMG ITEH --- ---- CAUSE ----
4/ 6/76 10:56: O HONENTARY RELAYED WHEN CLOS 10/12/77 1115:30 2326315 (1/O;&f NONE RELAY TOTAL OUTAGES DY CAUSE CAUSE SUSTAINED MONENTARY RELAY 1 0 RELAYED WHEN CLOSED 0 1 TOTAL 1 1 e
.5 CQ
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e LE COLE 52479 FLORIDA POWER 1 LIGHT COMPANY TRANSHISSION INTERRUPTION
SUMMARY
SYSTEN OUTAGES EXCLUDED LINE SECTION HIDWAY - TO - PLUMOSUS 130 KV O 19M-1970
TINE -----------
- PART - -- ALL -
- DATE -- -- UFF -- -- ON -- -- ON -- --- DMG I TEM ---- ---- CAUSE ----
1/ 9/75 9:36: O 9:39' O cc ,03:CO GUY WIRE FPL C Oi4 T . CREW 3/14/75 9:204 O HOMENTARY SWITCII 10/25/75 6:131 0 17:19 O lt:OL WO POLE UCHICLE 5/15/76 16:21 O HOMENTARY NON-DIS. LIGHT.
6/29/76 23:19:30 23:20: O 'CMOO!30 NONE TRANSFORMER 8/20/76 14:11:45 14:14:45 oo :o3 3 0c CONDUCTOR UEHICLE 9/12/76 15:15:30 15:16. O oo f oo 30 INSULATOR UANDLALISH 9/17/76 13:25: O HOMENTARY NON-DIS. LIGHT.
12/13/76 1:16:15 HOMEil T ARY UNKNOWN 12/14/76 14:25:30 HOMENTARY UNKNOWN 12/16/76 14:21: O HONEt4TARY UNKNOWN J
1/17/77 9:27:30 *i:20: O ootoo :30 NONE FPL CREW 2/17/77 11:10: 0 11:10:15 ooW il 5 NONE RELAYED WHEN CLOS 4/21/77 9:35' C HOMENTARY X-ARM 6/ 4/77 23:29: O 9: 7: O 4: 38:00 NONE X-ARM O/26/77 15:21:45 NOMENTARY UNNNOWN O/26/77 15:2? 30 HOMENTARY UNNNOWN O/26/77 15:22:45 f tOMEt1T ARY UNKNOWN 9/ 1/77 7:37:45 HOMENTARY SWITCH 9/20/77 19:53: O HOMENTARY UNKt40WN .
9/22/77 5:50:30 HUNCilTARY UNNNOWN 10/12/77 0:27:15 MOMENTARY UNKt40WN 11/ 5/77 10: 2: O HOMENTARY RELAYED WHEN CLOS 11/ 7/77 13: O' O NOMENTARY FPL CREW 11/16/77 12/13/77 7:51: O HOMENTARY 0:54:45 MOMENTARY RELAYED WitEN CLOS RELAYED WHEN CLOS
[D dU 38O n
12/23/77 9 13: O MOMENTARY RELAYED WitEN CLOS U[j f yh J[
, CO 1/ 3/70 O*50: O HOMENTARY RELAYED WHEN CLOS e s somm 2/10/70 15: 0:30 HOMENTARY WE ATitER IN AREA 3/ 3/70 13:26:30 NOMENTARY WEATHER IN AREA 3/10/70 7: 1: O HOMENTARY SWITCil 3/10/70 7:41:30 HONENTARY SWITCH 5/14/70 7:45: O 7:59:30 Ocal4: 30 NONE LIGHlHING ARRESTE I UNKNOUN 6/ 9/70 15:10:45 HOMENTARY O O/11/70 7:45:30 NOMENTARY SWITCH 9/15/70 20: 2: O HOMENTARY NON-DIS. LIGHT.
9/24/70 17*24: O HOMENTARY WEATHER IN AREA 10/11/70 7:42: O HOMENTARY SWITCH 11/29/70 10:50: 0 10:59: O co".o l ' oo NONE SWITCH 12/ 1/70 14:19:45 17:33: O 3:i3:15 NONE FPL CREW
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9 8 LE COLE 40179 FLORIDA FCWCR 8 LIGHT COMPANY TRANSMILCION INTERRUPTION CUMMARY SYSTEM OUTACEC EXCLUDED LINE SECTION PLUNOGUS - TO - RIVIERA 12 130 AV GQ 1975-19/0
, _ - . . . . . . . TINE - - - - - - - - - - -
- PART - -- ALL -
- nATE - 0FF "- ON - -- ON-- --- DMG I T E M -- - - - -- C A U S E ----
4/11/75 21:36: 0 MOMENTARY N0tFDIS. LIGHT.
11/ 3/75 16:45: 0 MOMENTARY INSULATOR 11/ 3/75 16:47: 0 MOMENTARY INOULATOR 11/ 3/75 17:19: 0 0: 4: 0 6:45:00 NONE INCULATOR 11/ 4/75 5:46: 0 15:52: 0 /0:cG:00 lNCUL ATOR SALT S: RAY 11/ 4/75 3:4C: 0 MOMENTARY UNNNOWN 11/ 4/75 5:29: 0 MOMENTARY UNANOWN 3/20/76 1:16: 0 NONENTARY INSULATOR 3/20/76 1:59: 0 MOMENTARY INSULATOR 6/17/76 9: 3: 0 9: 4: 0 0C*.0l;00 NONE NON- FPL CONT.
7/16/76 16:26:30 MOMENTARY UNhNOWN 9/14/76 14:10: 0 14:19: 0 00:01:cc NONE TRANSFORMER 5/23/77 6:11:45 MOMENTARY Ut&NOWN 6/2?/77 15:35: 0 MOMENTARY NON DIC. LIGHT.
1/ C/70 20:40:45 20:41:15 co.'oc OS NONE WIND 3/22/70 6:56:15 4:57: 0 g,;po Mi NONE TRANSFORMER 3/23/78 7:22: 0 7:26: O 0c:04:00 NONE TRANCFORMER 4/13/70 7: 0: J NOMENTARY RELAYED WilEN CLOS ,
4/20/70 5:54:15 5:54:45 con oc:3O NONE TRANSFORMER TOTAL OUTAGES BY CAJSE CAUSE SUSTAINED MOMENTARY RELAYED LHEN CLOSED 0 1
_S INSULATOR 1 4 p**)
e.
c_g NON FPL CONT. 1 0 #
g bolt SFRAY 0
- #M WIND 1
0 y' g UNANOWN 1
f
- O 4 p
NON DIS. LIGHT.
TRANSFORMER 0 2 (;%
4 0 Ll y TOTAL 0 11