ML17207A318
| ML17207A318 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 06/29/1979 |
| From: | Hodder M FLORIDA POWER & LIGHT CO., HODDER, M.H. |
| To: | |
| References | |
| NUDOCS 7908290073 | |
| Download: ML17207A318 (67) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING APPEAL BOARD Michael C. Farrar, Chairman Richard S.
Salzman Dr.
W. Reed Johnson
'.o~zC~i i~
In the Matter of FLORIDA POWER 6 LIGHT COMPANY (St. Lucie Nuclear Power Plant, Unit No.
2)
)
)
~
)
)
Docket No. 50-389
)
)
INTERVENORS RESPONSE IN OPPOS ITION AND SUGGESTION FOR HEARING On Friday, June 22, 1979 the Staff filed their motion for a delay until September 21, 1979 to respond to questions posed by the Appeal Board on the issues of the Florida Power and Light Company electrical grid stability (off-site power) and emergency diesel generators (on-site power) due to the unavailability of certain Staff'xperts who are assigned to the Three Mile Island Nuclear Plant Unit 2 case.
(NRC Docket No. 50-320).
Intervenors oppose the motion.
While the St. Luci'e Intervenors, in an attempt to be more than reason-able, did not oppose a previous Staff requests for extentions of time, they now recognize that the Staff responses to 790HZ'f (07)
important questions of serious health and, safety significance in. the St. Lucie 2 case are long overdue and have accordingly delayed the Appeal Board's hearing process and consideration of these serious safety issues.
Intervenors recognize the validity of the Staff responsibilities at the Three Mile Is-land Nuclear Accident Investigation and do not seek to pre-
.empt those efforts.
However, it is also the Intervenors view that the priority of consideration of safety concerns at the Hutchinson Island site are on a par with those of Three Mile Island and that the health and safety of residents of Pennsylvania cannot under the law be given a greater priority than that of the citizens of South Florida.
- Yet, the Staff is allowing construction of the Hutchinson Island Nuclear Plant without full and adequate health and safety considerations while it concentrates its efforts on the aftermath of Three Mile Island.
As it presently exists, the record of the St. Lucie 2 case is insufficient with respect to the unanswered questions 'on the grid stability and emer-gency power issues.
The blame for the delays and insuffi-ciency appears to lie, to a large extent, with the Staff.
Therefore, Intervenors suggest that the Appeal Board hold a
hearing designed to determine why Staff compliance with Appeal Board requests has not been forthcoming.
In. AIAB 489, Offshore Power Syst'ems the Appeal Board found:
"One thing the Board may do is ascertain why the Staff document in question has not been forth-coming. "
ALAB 489 Offshore Power Systems, (8 NRC 207)
Sept.
1978 A hearing would give the Staff the opportunity to establish whether its delays are reasonable:
"If the Staff can provide adequate assurance that it is acting as quickly and reasonably as the cir-cumstances permit and we emphasize the word rea-
~sonabl then the Board can ask no more and should reschedule the filing date accordingly."
BLAB 489
~
~su ra It may well be that the Staff has good and sufficient reasons for its delays but the reasons cited in the Staff motion of administrative inconvenience is by itself not an adequate ground for granting the Staff's request for additional delay.
The offshore Power case
~su ra, establishes that it is the prerogative of the Board to grant a delay or establish a
.schedule.
awhile the Intervenors oppose the Staff's motion for delay on the stated
- grounds, the Intervenors feel the Staff
should be given at hearing every opportunity necessary re-garding ascertainment of the nature of the Staff's problems, and the time necessary to resolve them so as to develop a
full. record.
This is necessary because Intervenors further.
advise the Board that they deem that their participation is appropriate to inquire into the reason and reasonableness of, as well as need for Staff delay, thereby establishing an adequate record in case the Board grants the Staff Motion and Intervenors then based on evidence adduced elect to re-new their Motion for Sta previously denied. by the Appeal Board in ALAB 537, 9
NRC Therefore, in Intervenor's view, the scope of the sug-gested hearing should include the following considerations:
1.
The reasonableness of the cause of the Staff delay.
2.
An assessment of adequacy of Staff's review effort, k
if Staff be required to meet the original filing deadline.
3.
Whether valid reasons such as a complete and mature investigation requires an extended time for proper assessment of the issues.
Regarding the Appeal Board's request that Intervenors indi-cate their degree of participation at the forthcoming hear-ing, Intervenors intend to participate in the following manner:
1.
Begin discovery within 10 days by posing interrog-atories to the Florida Power and Light Company based on prepared written testimony recently filed by the Company since that testimony does not adequately answer the questions and concerns of Intervenors.
2.
Commence discovery with the Staff after their response to the Appeal Board questions are filed.
3.
Attempt to obtain expert witnesses to testify at forthcoming NRC hearing.
(Preliminary discussions with Union Concerned Scientist, have been held.)
Of Counsel:
Terence Z. Anderson University of Miami School of Law Coral Gables, Florida 33124 Tel.
(305) 284-2253 or 2971 Prepare to present I tervenors case, in any event, through cross-examinat'4 of FPL and staff witnesses.
/
I J1 (gal
~
~
~
Martin H. Hodder Counsel for Intervenors 1131 N.
. 86th Street Miami, Florida 33138 Tel.
(305) 251-8706
oc~gggO,
'o gp,c.
i&)
iG)'b UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION a>
l~~
BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
. FLORIDA POWER
& LIGHT COMPANY (St. Lucie Nuclear Power Plant, Unit, 2) g
)
)
)
Docket No. 50-389
)
)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "INTERVENORS RESPONSE IN OPPOSITION AND SUGGESTION FOR HEARING" dated June.29, 1979 in the above-captioned
- matter, have been served on the fol-lowing by deposit in the United States mail, first class or air mail, this 29th day of June, 1979:
Michael C. Farrar, Esq.
Chairman Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr.
W. Reed Johnson Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
- 20555 Richard S.
- Salzman, Esq.
Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Michael Glaser, Esq., Alternate Chairman Atomic Safety and Licensing Board 1150 17th Street, N.W.
Washington, D.C.
20036
Harold F. Reis, Esq.
Lowenstein, Newman,,Reis 6 Axelrad 1025 'Connecticut Avenue, N.W.
Washington, D.C.
20036 Norman A. Coll, Esq.
- Steel, Hector 6 Davis 1400 S.E. First National Bank Bldg Miami, Florida 33131 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 William B. Puton Counsel for NRC Staff U.S. Nuclear Regulatory Commission Washington, D.C.
20555 4
Martin H. Hodder Counsel for 3;ntervenors
'STEEL HEGToR
& DAYIs SOUTHEAST FIRST NATIONALBANK BUILDING MI AM I, F LOR ID A 33 I3 I WILLIAMC,5TCCL LOUIS J. HECTOR OARRCY A. DAVIS OWIOHT 5ULLIVAN WILLIAM5, KII.LIAN CRNCST J, HCWCTT JERRY S,CROCKETT WILSON SMITH TALSOT D ALCMSCRTC JAMES H, SWCCNYq iTT JOHN EDWARD SMITH NORMAN
- COLL THOS. E. CAPP 5 SHCPARO KINO MATTHEW M.CHILD5 SARRY R DAVID5ON NOEL H. NATION SAUCE S.RU55CLL ALVIN 5, DAVIS JOSCPH P KLOCKyJR RICHARD C. SMITH THOMA5 R, McOUIGAN DCNNI5 *.LARU5SA PATRICIA A SCITZ PAUL J, SONAVIA JUDITH M, KORCHIN JOHN M, SARKETT ROIIERT J, IRVIN JL'ffRCY I. MULLCN5 VANCE C, 5ALTCR DONALD M. MIDDLCSROOKS HENRY J,WHCLCHEL OCRRY 5.OISSON SRIAH A,HART RI CHARD J, LANPC N JOSC I,ASTIOARRAGA DCAN C COL5ON KATHLCCN t, PATTCRSON JCf fREY $. SCRCOW June 22, 1979 WILL M PRESTON Of COUNSEL TCLCPHONC (305) 577-2800 TELEX SI-5758 DIRECT DIAL NUMSER William D. Paton, Esquire United States Nuclear Regulatory Commission Washington, D.
C.
20555 Re:
In the Matter of Florida Power 6 Light Company (St. Lucie Nuclear Power Plant, Unit No.
2)
Docket No.
50-389
Dear Mr. Paton:
Supplementing my letter to you of June 1, 1979, enclosed please find the following:
Testimony of Frederick George Flugger relating to questions A2, Bl, B2, B3 and B4 of ALAB-537.
Copies of this testimony have been simultaneously filed with the Board and served on all parties.
This testimony, to-gether with the joint testimony of Michel P. Armand,.Ernest L.
Bivans and Wilfred E.
Coe relating to questions Al and D of ALAB-
- 537, and the testimony of George E. Liebler relating to question C of ALAB-537, served June 1, 1979, constitutes all of the prepared written testimony to be filed by FPL in accordance with ALAB-537.
In a telephone conversation Friday, June 15, 1979, with M. Villar and M. Armand of FPL, Edward J.
Fowlkes of FERC requested the following information to allow completion of his review and we are providing that information herewith, with copies filed with the Board and served on all parties:
1.
Breaker diagram of FPL electrical system around Midway Substation in 1983.
STEEL HECTOR & DAVIS Page 2
2.
Analysis of the contingency loss of both Midway 240 KV buses.
3.
Line outage data for the v rious lines feeding into Midway Substation.
Very ruly'yours, cC~C NAC/sm Enclosures NO A.
COLL cc:
See attached service list.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC 'SAFETY G.LICENSING BOARD In The Matter Of:
FLORIDA POWER
& LIGHT COMPANY
)
(St. Lucie-Nuclear Power Plant,
)
Unit 2)
)
Docket No. 50-389 CERTIFICATE OF SERVICE I HEREBY CERTIFY that true and correct copies of the foregoing letter dated June 22,
- 1979, addressed to William D.
- Paton, Esquire, and the enclosures referred to therein, have been served this
'22nd; day of June,
- 1979, on the persons shown on the-attached service list by deposit in the United.States mail, properly stamped and addressed.
- STEEL, HECTOR 6'DAVIS 1400 Southeast First Nationa Bank Building
- Miami, lorida 33131 Telepho e:
(305) 577-2863 By NO N A. COLL
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY 6 LICENSING BOARD In The Matter Of:
FLORIDA POWER 6 LIGHT COMPANY
)
(St. Lucie Nuclear Power Plant,
)
Unit 2)
)
Docket No.'0-389 SERVICE LIST Mr. C.
R. Stephens Supervisor Docketing and Service Section Office of the Secretary of the Commission Nuclear Regulatory Commission Washington, D.
C.
20555 Michael C. Farrar, Esquire Chairman Atomic Safety 6 Licensing Appeal Board Nuclear Regulatory Commission Washington, D.
C.
20555 Dr.
W.
Reed Johnson Atomic Safety S Licensing Appeal Board Nuclear Regulatory Commission Washington, D.
C.
20555 Richard S.
- Salzman, Esquire Atomic Safety
& Licensing Appeal Board Nuclear Regulatory Commission Washington, D.
C.
20555 Alan S. Rosenthal, Esquire Chairman Atomic Safety 6 Licensing Appeal Panel Nuclear Regulatory Commission Washington, D.
C.
20555 Edward Luton, Esquire Chairman Atomic Safety 6 Licensing Board Panel Nuclear Regulatory Commission Washington, D.
C.
20555 Michael Glaser, Esquire
.Alternate Chairman Atomic Safety 6 Licensing Board 1150 17th Street,,
N.
W.
Washington, D.
C.
20036 Dr. Marvin M. Mann Technical Advisor Atomic Safety 6 Licensing Board Nuclear Regulatory Commission Washington, D.
C.
20555 Dr. David L. Hetrick Professor of Nuclear Engineering University of Arizona
- Tucson, Arizona 85721 Dr. Frank F. Hooper Chairman Resource Ecology Program School of Natural Resources University of Michigan Ann Arbor, Michigan 48104
Mr. Angelo Giambusso Deputy Director for Reactor
'rojects Nuclear Regulatory Commission Washington, D.
C.
20555 William J.
- Olmstead, Esquire U.
S. Nuclear Regulatory Commission Washington, D.
C.
20555 William D. Paton, Esquire Counsel for NRC Regulatory Staff Nuclear Regulatory Commission Washington, D.
C.
20555 Martin Harold Hodder, Esquire 1130, N.
E.
86 Street Miami, Florida 33138 Harold F. Reis, Esquire Lowenstein,
- Newman, Reis, Axelrad 6 Toll 1025 Connecticut,
- Avenue, N.
W.
Washington, D.
C.
20036 Local Public Document Room Indian River Junior College Library 3209 Virginia Avenue Ft. Pierce, Florida 33450 James R. Tourtellotte Counsel for NRC Regulatory Staff Nuclear Regulatory Commission Washington, D. C.
20555
Testimony of Frederick Geor e Flu er Relating to ASLAB Memorandum and Order of April 5, 1979, on Electrical Grid Stabilit and Emer enc Power S stems
( uestions A2 Bl 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.
Hy education 3
and professional qualifications appear in the Nuclear Regulatory Commission's 4
record of the St. Lucie Unit 2 (Unit 2) proc'ceding following.Tr. 1310.
5 The purpose of this testimony is to respond to questions A2; Bl, B2, B3, and 6
B4 in Section II of the Appeal Board's Order of April 5, 1979.
My affidavit of March 31, 1978 is relevant to the issues raised by the Appeal Board., It is provided as Attachment A and is hereinafter referred to as the Flugger 9
1O This testimony demonstrates that the Unit 2 onsite AC power system design is 12 13 in full compliance with NRC requirements, that the design basis events evaluated in the PSAR provide a proper basis for the design of Unit 2 and that Unit 2, I
as de'signed, 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
-1 18 operating history suggests a frequency of outage of about 4 x 10 per year 19 for the FPL grid.
Although there is little comparative historical data readily
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.
4 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 reli,ability 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).
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) 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 odated by the existing Unit 2 design.
The time to restore AC power is pivotal to the evaluation of the postulated loss of all AC event.
FPL's historical grid data demonstra'tes that "the duration of loss of offsite power is very short-lived.
FPL operating exper-lf ience from January 1972 to present indicates a mean time to restore offsite poi(er to FPL facilities of less than 1/2 hour.
10 12 13 15 16 Third, an unprotected loss of coolant accident (LOCA) does not result 'from the postulated loss of all AC event.
There is no failure of the reactor coolant pressure boundary associated with this event.
A reactor coolant pump (RCP) seal can only yield very small and acceptable leak rates.
(See the response to question B2 infra.)
Unit 2 has more than adequate capability to remove decay heat, which is necessary to accommodate the postulated loss of all AC event., There is sufficient condensate to provide steam generator makeup for 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 feedwater pump control and auxiliary feedwater system valves are DC powered, and the steam generators have suffici'ent inventory to allow the operator about 55 minutes to actuate auxiliary feedwater before steam generator dryout occurs.
17 eolith these considerations in mind we can procede with the responses to the 18
- Appeal'oard's questions.
19 uestion A2 20 21 22 23 24 25 26 27 28 29 For its part, the first paragraph of GDC-17 appears to establish an unattainable set of conditions for electrical power systems generally.
It reads as follows (emphasis added):
An onsite electric power system and an offsite electric, power system shall be provided to permit functioning of structures,
- systems, and components important to safety.
The safety functions for each system assumin the other s
s em is no fu shall be to provide sufficient capacity and capability to assure that (1) specified accept-able fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of antici ated o erational
1 occurrences and '(2) the core is cooled 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 'ower 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 requireme'nt.
11
~Res onse 12 The Board's question cites a possible 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 Atomic Energy Commission on July 10, 1967 before 16 GDC 17 was adopted in its present form.
Their language states:
17 GDC 24 18 19 20 "In the event of loss of all offsite power, sufficient alternate sources of power shall be provided to permit the required functioning of the protection systems."
21 'DC 39 22 23 24 25 26 27 "Alternate power systems shall be provided and designed with adequate independency, redundancy,
- capacity, and testabi lity to permit the functioning required of the engineered safety features.
As a minimum, the onsite power system and the offsite power system shall-each, independently, provide this capacity assuming a failure of a single active component-in each power system."
28 The intent of these criteria is simply to ensure that an onsite AC source be II 29 provided adequate to backup the offsite AC source.
This philosophy is embodied 30 in current industry standards.
IEEE std. 308-1974 (reference
- 2) embodies the 31 concept of the "preferred" (offsite) power supply system and the "standby" 32 (onsite) power supply system.
The functions of these systems are cited in the 33 standard as follows:
4 5
.6 7
"The preferred power supply shall furnish electric energy for the shutdown of the station and for the operation of emergency systems
-and engineered safety features."
"The standby power supply shall provide'lectric energy for the operation of emergency systems and engineered safety features during and following the shutdown of the reactor when the preferred power supply is not available."
8 RG 1.32 (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 ll contemplate the literal interpretation suggested by the Appeal Board's question.
12 Unit 2, as designed, complies with the accepted interpretation and intent of 13 GDC 17.
14 Finally, it is appropriate to note the relationship between 10 CFR 50.36 and Appendix A to 10 CFR Part 50.. The 'latter provides design criteria'while 16 the former imposes operational restrictions.
The flRC regulations 17 at 10 CFR 50.34 require that a safety analys'is be performed to assess the 18 ability of the facility to meet its design objectives.
The safety analysis 19 provides the basis for establishing limiting conditions for operation (LCO),
20 which provide. the minimum functional capability or performance levels required for safe operation of the facility., 10 CFR 50.36(c)(2).
The LCO's become part of the facility's operating license.
Therefore, it is pertinent to note that continued Unit 2 operation with both onsite diesel generators inoperable would 24 constitute a violation of Technical Specifications.
RG 1.93 (reference 4) 25 states that the limiting condi tion for operation (LCO) is met "when all the electric power sources required by GDC 17 are available."
If both diesels were inoperable the plant's operating license would restrict operation in accordance 28 with RG 1.93 as follows:
"If the available onsite a.c. electric supplies are two less than the LCO, power operation may continue for a period that should not exceed two hours.....'f no onsite a.c.
supply is restored within the first two hours of continued power operation, the unit should be brought to 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 />."
h 6
Thus, the Unit 2 operating license will contain conditions in the form of these Technical Specifications to minimize the risk of exposure to continued plant operation with both diesels inoperable.
uestion Bl 10ll 12 13 14 15 16 17 18 As we see it, the likelihood of loss of all AC power at St. Lucie may be expressed as the product of two factors:
(1) the probability that there will be an offsite power failure involving the FPL network generally or the Midway substation in particular and a resulting loss of station power -- which probability seems based on historical events, to lie in the range 1.0 to 0.1 per year; and (2) the probability that neither of the two onsite AC power systems (diesel generators) will start.
The probability that any one diesel generator will fail to sta~t on demand is taken by the staff to be one per hundred demands, i.e., lo 25/.
19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 If these figures are accurate, then the cogbined p~obability for the "loss of all AC power" scenario is in the range lo to lo per year.
26/
In this
- regard, the staff's Standard Review Plan for Nuclear Power Plants sets forth numerical guidelines for determining whether an event "resulting from the presence of hazardous materials or activities in the vicinity of the plant" should be considered in designing the plant (i.e., whether it is a "design basis" event),
27/
Under these guideli~es, events with a realistically calcu-lated probability value of at least lo per year (or lo
. per year for a conservative calculation) must be so considered.
The "loss of all AC power" sequence is not precisely within the category of events contemplated by the Standard Review Plan.
However, its ultimate result -- assuming that power is not timely restored -- is an unprotected loss of coolant accident, the consequences of which are likely to exceed the guidelines of 10 CFR Part 100.
l<e do not understand why this sequence of events (i.e., loss of offsite power combined with failure of diesels to start),
which appears to have a probability well above the guideline values, should not be taken into consideration in the design of the plant.
28/
The parties are to address this point, setting forth their reasons for adhering (if they do) to a
contrary position.
38 39
~25 Fitzpatrick Affidavit of June 12,
- 1978, p.
4.
Also see Regulatory Guide 1.108, Section B.
1 26/
This conclusion further assumes that the failure of two diesel generators to start would be statistically independent
- events, an assumption which 3
leads to the.lowest likelihood of combined failure, and which might be 4
nonconservative if there exists the potential for common failure modes 5
for the onsi te systems.
6 27/
NUREG 75/087, Section 2.2.3, paragraph II.
7 28/
We have accepted the Standard Review Plan guideline values as reasonable 8
in another case.
Public Service Electric and Gas Com an (Hope Creek Units 1 and 2),
ALAB 429, 6
NRC 229, 234 1977 10
~Res ense 12 13 The question pertains to two different but complementary nuclear plant design
- concepts, namely, the frequency of occurrence of an event (events/unit of time) 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, it is necessary to discuss the concepts of 17 event frequency and ESF reliability.
18
~EF 19 Many 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 22 23 25 26 27 28 l.
Events of moderate frequency leading to no significant radioactive releases from the facility and no violation of fuel design limits.
2.
Infrequent events which have the potential for small radioactive releases from the facility and small amounts of fuel failure.
3.
Events of low probability, Design Basis Accidents (DBA), which are
.required by 10 CFR Part 50 to establish the performance requirements of ESF's and are used in evaluating the ability of the facility to comply with 10 CFR Part 100 guidelines.
1 Unit 2 design bases are the "specific functions to be performed by a structure, 2
system or component of a facility, and the specific values or ranges of values 3
chosen for controlling parameters as reference bounds for design" (10 CFR 50.2).
4.'hey are developed by analyzing limiting events, i.e., other events of the type 5
analyzed are less severe.
This approach provides reasonable assurance that 6
the facility has adequate capability to accommodate unanalyzed events.
7 The probability of occurrence of non-design basis initiating events that may 8
produce results more severe than OBA'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) provides a
10
/10 guideline for "design basis events resulting ll 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 14 15 16 17 18 19 20 21 22 23 24 26 1.
This guideline is appropriate for events that have a potential for yielding offsite exposures that equal or exceed 10 CFR Part 100 guidelines.
2.
There is little experience available to provide a statistical= basis for quantifying with precision the probability of occurrence of initiating events which have such low probability.
Thus considerable engineering and scientific judgment is 'involved in determining whether or not a given event should be included in the design basis.
- 3. If an event which was considered to be outside the design bases did occur, it would not necessarily produce consequences that are catastro-phic or exceed 10 CFR Part 100 guidelines.
Considerable engineering, design evaluation and operating experience has been accumulated since the first commercial light water reactors went into operation around 1960.
This significant experience base has demonstrated that a nuclear
10 12 13 14 facility has substantial inherent capability to acceptably accomodate a broad spectrum of events.
4 ~
The Unit 2
design philosophy utilized is specifically directed at providing assurance that the likelihood of events with consequences more severe than DBA's is extremely low.
The facility is designed, built and operated so that it will, with a high degree of reliability, minimize the likelihood of an accident.
Despite the care taken to prevent accidents, the design provides for reliable protection devices and systems designed to detect and cope with transient and off-normal conditions.
ESF's provide protection to the public even in the event of the occurrence of severe accidents of low probability, i.e.,
DBA's.
Finally, throughout the facility's lifetime nuclear plant operating experience is continually monitored and assessed by the NRC to determine whether design or procedural modifications are required.
15
~HF 1i bill
]6 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 acceptable 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 on the concept of redundancy.
The objective of %his criterion is to 26 prevent any single failure from preventing the accomplishment of a safety 0
1 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 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 one 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 15 16 17 18 19 20 21 22 23 24 P(T)
= P(A)
~ P(B)
. ~
P(C)
~
P(D)
~ P(E)
~ P(F) where:
P(T)
= probability of not restoring AC power by P(A)
= probability of loss of offsite power P(B)
= probability of loss of first diesel P(C)
= probability of loss of second diesel P(D)
= probability that offsite power is not repaired and returned to service by time P(E)
= probability that first diesel is not repaired and returned to service by time P(F)
= probability that second diesel is not repaired and returned to service by time 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 t
27 restored at time "T", P(T+hT) this probability at a finite later time "T+QT",
1 and C hT the 'repair probability during the time interval hT (where C is a
2 constant).
- Then, P(T+6T) =.P(T)
~
(1-C ~ 6T) 4 which in the limit as dT approaches zero is given by:
7 whose solution is:
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 ll "C" for ea'ch of these probability terms.
An evaluation of FPL system data from
-1 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
-1 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 83 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 oX fai.lures in a period of time "t",
18 such as 0.1 failures per year),
then e
is the probability that offsite 19 power will not be lost and the probability that offsite power will be lost
-Xt 20 can be expressed as P(A)
= 1-e 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
-2
.-1 23 generator failure per demand of 10
, and time constants of 1.6 and 0.16 hr 24 for offsite and onsite power restoration respectively yields:
25 P(T)
= 10 exp (-1.92T)
1 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:
5 6
7 8
If 9
10 11 Duration of loss of AC "T"
hours 0
1 1.2 2.4 Probability of Having a Total Loss of AC Power that Lasts "T" Hours, P
T 1 x10 2 x,10 1 x10 2 x 10 1 x10 3 x 10 5x10 12 If a loss of offsite AC power event frequency of 1.0 per year were assumed 13 instead of 0.1, then a value of P(T) of 1 x 10.
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 />, 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 17 18 that the probability of a continued loss of AC power decreases significantly with the duration of the loss.
If, as suggested by the question, the 10
/10 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 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />) 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 guestion B2 below, Unit 2 can 23
" be maintained in a safe shutdown condition without AC power for a time period 24 well in excess of the time likely for restoration of AC power.
2 3
4 5
6 7
8 s
i B2 In line with the above discussion, the testimony is to analyze events that would occur between the "loss of all AC power" and the violation of either the fuel design limits or the design conditions of the reactor coolant pressure boundary (or any portion thereof).
In particular, the parties should, if possible, reconcile their differing responses to question B.l(b) of our March 10, 1978 order, 29/ or, if not, point up precisely where the, disagreements lie.
9 29/
[ References fn 24 reproduced below:
]
10 11 12 13 14 15 16 17 Applicant suggests that the first safety related failure encountered would be excessive core heating due to the loss of water from the 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 the loss of AC power (Flugger Affidavit of March 31,
- 1978,
- p. 3).
The staff's judgment is that the first failure would be that of a primary pump seal at about one. hour after the loss of AC power---
resulting in a small loss of coolant accident.
(Fitzpatrick Affidavit of June 12,
- 1978,
- p. 11).
'18
~Res ense 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 for't 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 for at least'6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 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.
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) condit'ion is much less severe than 5
the dynamic (pump running) condition discussed in the Unit 2 PSAR at section 6
9.2.2.3.1, which provides a basis for concluding that running the pumps for 7
about one hour, without cooling water to the seals, would not result in pump 8
seizuee -or. unacceptable RCP seal 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.
ll Each RCP is equipped with a seal cartridge, which 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 pressur'e.
A seal cartridge test fixture 14 is used to fully test the seal cartridge prior to installation on, the RCP,'nd 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 C
17 in place by hydraulic force since the higher pressure is on the ring's outside 18
- 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 l.eakage path within the seal cartridge Zl under static conditions does not exist.
22 Pressure breakdown devices are installed parallel to the first three seals.
Reactor 23 coolant at a rate of 1
gpm passes through these devices such that reactor. coolant 24 system pressure is distributed equally across the first three seals, i.e., they I'5 normally operate at about 1/3 of their design pressure.
The fourth seal is subjected 1
to a nominal backpressure and acts as a vapor barrier/backup seal during normal 2
operation.
The RCP controlled bleedoff flow of 1
gpm/pump is directed to the 3
chemical and volume control system.
4 Under-static conditions associated with loss of all AC the temperature of the 5
fluid in. the seal cartridge will atta'in a level above the normal seal cartridge 6
operating temperature due to the interruption of cooling water.
The temperature would rise from about 180 F to about 550 F.
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 ll. 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 automatically. 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
- gpm, as flow is 16 restricted to this value by the pressure 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 third 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.
'1.If the controlled bleedoff line is not closed off, pressure distribution through 2
the seals is maintained the same as for normal operation. and the bleedoff is 1
gpm 3
per pump.
Operation in this mode results in a pressur'e 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, ll
,there is no reason why any one of these seals would fail in the static condition.
12 The only components affected by the elevated temperature, ar'e the elastomeric 13 gaskets of the seals, namely 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 components are metallic or carbon, which are not affected 17 by the elevated temperatures of the system.
The elastomeric components are made 18 of Ethylene Propylene or Nitrile, materials which are suitabl'e 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 of 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 temperature of 550 F, the elastomeric material, Ethylene Propylene or 24 Nitrile, would be subject to extrusion and hardening, i.e., gradual loss of flexi-25 bility and permanent setting in a deflected position.
The reactor coolant pump
- 16
1 manufacturer has demonstrated,
- however, the sealing characteristics of the 2
elastomeric material under thermal conditions equivalent to those resulting...from 3
the postulated loss'f 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 'ealing 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" rings are backed up by lapped'0 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 for the duration of the static loss of all AC event, and.the.RCP seals are "4
expected to remain functional for a period of at least 24 ho~rs.
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 brought to a
21 cold shutdown without the reactor coolant pumps running, i.e.,
on natural 22 circulation.
A Insofar 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 fluid.
Decay heat is rejected 4
to the secondary system 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 feedwater
'I pump and will locally open the steam generator atmospheric dump valves to allow 8
the steam generator safety valves to res'eat.
Steam generator level will be 9
reestablished, at which time the operators will adjust auxiliary feedwater 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 loss from the pressurizer, normal reactor coolant -system (RCS) leakage 13 e.g.,
RCP seal bleedoff'flow, and secondary side liquid temperature in the steam 14 generators, there will be a gradual and steady decay in RCS pressure and tempera-15 ture.
Since the RCS pressure decays at a higher rate than temperature, the 16 reactor coolant 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 removal will continue by natural circulation 18 of a saturated fluid.
19 The Flugger Affidavit indicated that about 200,000 gallons of water would b'
20 required to maintain hot standby for 16 ho'urs, which is the minimum technical 21 specification limit anticipated for the Unit 2 condensate storage tank.
- However, 22 the condensate storage tank is normally maintained in excess of the technical 23 specification'limit, and has a design capacity of 400,000 gallons.
Additionally, 24 there are another 1,800,000 or so gallons (design capacity) of fresh water 25 storage on site at St. Lucie.
It is reasonable to conclude that during the
- 18'-
~
~
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 4
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 hour basis, battery, i.e.,
each can supply 225 aMiperes 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 load required for remote manual auxiliary 8
feedwater operation, control room 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 from the on-line, ll 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'o point out that the Unit 1 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 kY busses to be tied between 18 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 breaker 22 is not installed.
Loads would have to be stripped from the Unit 1
and Unit 2 23 busses and the breaker in the cubicle from the Unit 2 startup transformer must be 24 removed and installed in the 4.16 kY switchgear tie.
The sequence of events has 25 been
- reviewed, and it has been determined that it would take two men about one 26 hour3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> to alig'n a Unit 1 diesel to Unit 2.
1 In summary, 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 'nd no LOCA would result from this postulated event.
- Thus, 10 CFR Part 100 6
guidelines are not applicable to this event.
7 s
i
'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 initial, auto-start signal.
11
~Res ense 12 Should a diesel generator at a nuclear plant fail to start the unit's technical 13 specifications would require that the second diesel "and offsite AC circuits 14 be verified operable 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 17 18 19
- operable, there is no undue time-pressure constraint to return the diesel to e I
- service, which would exist if all AC power were lost.
Accordingly, any evalua-tion of the time to return a diesel to service based on historical data would likely yield a conservative estimate of the time to return a diesel generator 20 to service.
21 The concept of diesel reliability should also be placed in proper perspective.
22 A 10 probability of demand with a confidence level of 95% was demonstrated by 23 a 300*start shop test program for a Unit 1 diesel (see Unit 1
FSAR section 24 8.3. 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 1
loading the generator to 60Ã of continuous rating, and maintaining the 60%
load for 5 minutes.
A failure in any portion of the sequence was considered 3'
failure per demand.
This sequence is based on LOCA generated ESF require-4
- ments, which place exacting quick start design requirements upon the diesel 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 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 fifth event involved a sticky sole-12 noid and pluggage of an air starting line for which restoration time was 7-2/3 hours.
14 The turbocharger failures were due to a momentary deficiency 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 driven 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'een 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 21
6 was resolved by disconnecting a neutral lead, resulting in the elimination of third harmonic current heating effects.
.Since this problem was unique to the Turkey Point design and does not apply to the St. Lucie diesel generators, this data point was.also omitted from the data base.
The repair time'requency distribution based on St. Lucie and Turkey Point experience to date is as follows:
Repair Time minutes Frequency of Occurrence Repair Time minutes Frequency of Occurrence 9
10 ll 12 13 14 15 16 17 18 19 10 15 21 30 37 65 70 76 77 91 94 1
1 1
1 1
1 1
1 1
1 1
111 180 217 240 258 275 390 460 503 708 1435 3563 1
1 1
1 1
1 1
1 1
1 1,
20 21 22 23 P
The median diesel repair time is 111 minutes and the mean is 388 minutes.
If each event is assumed to have equal probability of occurrence, then the probability of restoration of a safety-related diesel at an FPL nuclear facility
-1 1/
can be expressed mathematically by 1-e
, where C is 0.16 hr 24 25 26 27 28 29 30 31 1/
Although it is inappropriate to include the turbocharger and voltage regulator data, inclusion of these data does not alter the conclusions reached in question Bl
~su ra, i.e., evaluation of aperi.od exceeding about 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not required since the probability of not restoring AC power within that time period is acceptably low.
Inclusion of these data yields a median of 217 minutes, a mean of 1434 minutes and a
C of 0.04 hr Thi~ results in an expression for P(T) of lo exp (-1.68T) as compared to lo exp I-1.92T), which is, used in the response to question Bl
~su ra.
'1 To respond to question Bl
~su ra the equally important issue of hou 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 6
~min.
7 1
8 8
9 9
10 11 15 12 17 13 20 14 22 15 23 Frequency of Occurrence 1
1 1
1 1
2 1
1 Duration
~(min.
30 31 32 40 43 53 77 Frequency of Occurrence 2
1 1
1 2
1' I
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 P
I 18 occurrence, then the probability of restoration of AC power to any FPL facility
-CT
-1 19 can be expressed mathematically by 1-e
, where C = 1.6 hr In 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 24 onsite AC power within an hour is very high, which is reflected in the probability 26 assessments provided in response to question Bl
~su ra.
v 23
s ti 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
~Res onse 8
As demonstrated by the responses to questions Bl and 82
~su ra, 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 kV bus voltage and diesel failure to start alarms 17 will annunciate, and the control room lighting will dim to the DC lighting 18 system level.
The DC system will provide power for requisite monitoring-19 instrumentation and for auxiliary feedwater system operation.
The detailed 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.
References:
1.
"NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear. Power Plants, LHR Edition," U.
S. Nuclear 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, "Availability of Electric Power Sources."
~(
ATTACHil' UNITED STATES OF ~~RICA NUCLEAR REGULATORY COi4IHISS ION BEFORE THE ATONIC SAF ETv AND LICENS'ING APP~AL BOARD In the i~latter of:
)
)
FLORIDA POWER AND LIGHT COi<DANY
)
)
(St. Lucie Nuclear Power
)
Plant, Unit No.
2)
)
DOCKET NO.
50-389 AFFIDAVIT OF "R"DERICK G.
FLUGGER I am Frederick G.,Flugger, Supervisor, Plant 2
Licensing, Power Plant Fngineering Department for Florida 3
Power and Light Company.
i<y education and professional qualifications appear in the Nuclear Regulatory 5
Commission's record of the St. Lucie 2 pro'ceeding 6
following Tr. 1310 The purpose of this affidavit is to respond 8
to auestion B.l(b) concern'ng loss of all AC. ower rom 9
the Appeal Board's Order of i>Larch 10, 1978 in this 10 proceeding.
12 Ques ion B. 1 (b)
As a func"ion of the delay time 'nvolved, what 14 are the corsecuences of a 1oss of offsi"e power at St.
15 Lucie 2 combined with fai'ur of onsite power sources 16 to start on demand (i.e., delayed start).
No other 17, failure of tne system (e.>.,
LOCA) need be consi"ared
in this analys' E
3
RESPONSE
d Loss oz all AC."ower is not a design oasis 5
for St. Lucie.
Like all otne
- plants, St. Luci has 6
been designed to the single failure criterion, in 7
accordance wi& applicable NRC regulations.
Xn order for a loss of all AC power to occu" 9
after a loss of offsite power, a double failure, i.e.,
10 the failure of two independent diesels to start and 11 supply onsite powe
, is recuir d.
Conseauently, a
12 detailed analys's of such an eve~t has not been 13 performed.
14
- However, assum'ng the hypothesis in ti e 15 Board's cuestion, there are two'predominant sazety 16 functions-to be performed following,1oss of offsite 17 power and failure oz onsite power to;start; (1) removal of decav heat from the reactor coolant system and; (2) removal
.o decay heat
. =on the spent fuel storaae pool.
21 (1)
Heat rom the reactor core will be 22 transferred to the steam generator by natural circu-23 1.ation of reactor coolant.
Heat removal can then be 24 accomplished by the zeedwater provided by the au=(il'ry 25 feedwate
'ystem a }d ez' si e to the tmosphe'y
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 pump takes suction from the condensate storage: tank (CST).
The CST contains a sufficient volume of conden-C sate so 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 I
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 En view of the foregoing, FPL believes that 18 either offsite power would be restored, or onsite power'9
- supplied, before any safety-related consequences would 20 occur.
/
Mi FREDERICK( G.
FLUGGER
'TATE OF FLORIDA COUNTY OF DADF ss
.'ubscribed and sworn to before me this
'L daj of ~/,c.',
1978.
bury commission evpires:
NOTARY PUSUC STAT= OF R.CRIOA tt BERGS hIY COMMISSION CCPIRiS AUGUST 2t, I22I 2'!C25 TllRU hIAY.'LAB~0 8C!ICIIIG AG='ICY NOTARY PUBLIC
1985 SYSTEH 21066 21046 21015 21047 2106 21085 ST.
LUCZE HUTCHINSON ISLAND 40367 40364
. 40355 40352 40343 40340 40333 40330 40 1
40349 40323 40326 18 START-UP 2
1A START UP 1
24180 MIDWAY 26 52 24658 2461 24638 24646 24615 24641 246 246 5
24651 2478 24716 24747 751 24715 24777 OKEECHOBEE 6516 6515 24721 PLUMOSUS 29585 6565 29586 HARTMAN 48727 48724 48721 39720 3972S 39710 SHERMAN 2E lv 1N 1E 15 20 33935 33938 INDZANTOWN 33929 926 36811 2516 2544 2556 2565 2555 2566 2W HOBE 2W 2M 2E 36815 RANCH 3N 3E 10 25 36819 MARTIN MARTIN START-UP PRATT 4 WHZTNEY To ANDYTOWN
ATTACHHEHT 85 (ADDENDUH)
An analysis was performed onthe contingency of the loss of both Midway 240 kV busses.
The end result of the loss of both busses with a breaker and a half scheme is that the breakers connected to the busses are open and the lines coming into the substation 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 Snerman 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 south, and a fourth passes by with no connection to St. Lucie.
A loadflow study was performed to test what distribution of power flow would result if the loss of both busses occurred at the time of peak summer 1983 load with both 'St. Lucie units in service.
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.
1985 NORf"IAL MALABAR 21066 21046 21015 21047 2106 21085 ST.
LVCIE HUTCHZNSON ISLAND 40367 40364 40355 40352 40343 40340 40333 40330 40 1
40349 40S23 40326 24780 506)
)5i2 I",8 MIDWAY 18 START-VP 2
1A START-UP 777 777.-
26 52 2461 24638 24615 24641 2478 246 5
24651 24747 24716 24658 24646 246 )5 24665 751 24715 24777
/P. 5 OKEECHOBEE 6516 6515 I 'f S 24721 PLUMOSUS 29585 6565 29586 48727 48724 48721 39720 39725 39710 SHERMAN 2E 1W 1N 1E 33935 33929 33938 926 INDIA?ITOWN Z 9'Q 2516 2544 2565 2555 2W HOBE 2W 2N 2E 15 20 36811 36815 2556 2566 RANCH 3N 3E 10 25 36819 HARTIN MARTIN START-UP PRATT 6 WHITNEY TO ANDYTOWN
~
1985 (LOSS OF BOTH BUSSES)
MALABAR 21066 21046 21015 21047 2106 21085 ST LUCIE HUTCHINSON ZSLAND 40367
. 40355 40343 40333 40364 40352 40340 40330 40 1
40349 40323 40326 24780 SOO 2ol MIDWAY 18 1A START&9 2
START UP 1
V77 77'6 52 2461 24638 24641 6
24615 246 5
24651 2478 24716 24747 24658 24646 246 5
2466 751 24715 24777 800 20 I OKEECHOBEE 6516 6515 24721 PLUMOSUS 29585 6565 29586 HARTMAN 48727 4872 48721 39720 39725 39710 28 33935 33929 1W 1M 1E 2W 2M 2E 15 20 33938 ZNDZANTOWN
~seI 36811 36815 2S16 2544 2556 2565 2555 2566 HQBE 3M 3E 10 25 36819 MARTIN MARTIN START-UP PRATT 6 WHZTNEY TO ANDYTOWN
LE COLE 62079 FLORIDA POWER 1 LIGtlT COMPANY TRAHSMISSIOH IHTERRUPTIOH
SUMMARY
SYSTEM OUTAGES EXCLUDED LIHE SECTION MIDWAY -
TO -
CT LUCRE PLAHT Cl 240 KV 1975-1978
(~.(,Q~
TIME PART
- -- ALL DATE -- --
OFF -- --
0th -- "- Ot( -- ---
DMG ITEM """,---- CAUSE 1/27/76 15 i 15 i 0
15 i 15: 30 oo:oo'1 HONE FPL CREW 5/14/78 7 ~ 45 ~
0 7:55l 0
oo:ro:oo HONE LIGHTNING ARRESTE TOTAL OUTAGES RY CAUSE CAUSE SUSTAINED MOMEHTARY LIGHTNING ARRESTER FPL CREW TOTAL 0
0 0
LE COLE C2079 FLORIDA FOtlER 8, LIGHT COMPAHY TRAHSMISSIOtt IHTERRUF'TIOH SlJMMARY SYSTEM OUTAGES EXCLUDED LIhE SECTIOH MIDWAY -
TO
T LUCIC PLAHT 42 240 KV 19'75-1978 TIME F'ART - -- ALL DATE. - "- OFF." -" OH -- --
OH -" ---
DMG 7/ 5/7C 23:34:
0 23:34l30 ogA>.'3o h'OHE 7/11/7C 23e30:
0 MOMEttTARY ITEM ---.---- CAUSE 1
UHKttOtttt UHKttOWH 5/14/78 7t45'0 7t55:
0 Oo:to: >
HOttE LIGHTHIHG ARRESTE TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED MOMEHTARY LIGHTHIHG ARRESTER UttKHOMtt TOTAL 0
LE COLE C2079 FLORIDA POMER S
1.JGNT CONPAHY TRANSMISSIOH JNTERRUPTJOH GUHHARY SYSTEN OUTAGES EXCLUDED LINE SECTION NIDMAY " TO -
CT LUCIE PLANT 43 240 KV 1975-1978 TINE PART
- -- ALL DATE -- --
OFF -- --
OH -- --
OH -- ---
DMG ITEN -"- ---- CAUSE 5/14/78 7e451 0
71551 0
00:>o:o>
HONE 1 IGHTHIHG ARRESTE 7/1 8/78 5 53
~ 30 5 53
~ 4 o>:oo'.if HONE UNKNOWN TOTAL OUTAGES DY CAUSE CAUSE SUSTAINED HOHEHTARY LIGHTNING ARRESTER UHKNOMH TOTAL 0
0
LE COLE 52479 FLORIDA POWER R LIGHT COMPAHY TRAHSMISSIOH IHTERRUPTIOH
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION MALABAR 'TO MIDWAY 41 240 KV
- 50. 38 1975-1978 DATE 6/ 5/'75 6/24/75 7/ 2/'75 7/ 3/75 7/16/75 li/'13/75 TIME 0!3ie 0
6 ~
0 ~
0 221 Oo 0
22'3t 0
51311 0
1 155 ~
0 PART ALL OFF -- --
ON -- -- ON MOMEHTARY MOMEHTARY MOMENTARY MOMENTARY MOMEHTARY MOMEHTARY DMG ITEM CAUSE UNKNOWN HOH-DIS ~
LIGHTS UHKHOWN UNKHOWN UHKHOWH UHKHOWH 7/13/'76 8/ 4/76 9/13/76 li/12/76 11/12/76 15'5 3118 22 ~ 37 0'4 61 9
c30 Q
130
~
Q
~
p 22'1 0
MOMEHTARY MOMENTARY MOMENTARY 6iipi 0 J
4 ~ Bee 30 INSULATOR P,pl '00 NONE LIGHTHIHG UHKHOWH UHKHOWH UNKNOWN UNKHOWN 6/14/'77 7/ 5/77 7/23/77 8/'21/77 11/ 7/77 12/ 4/77 1!48 20'i 3c53 6$
S 22 ~
7 2'4
~ p
~
Q
~
p p
130 245 MOMEHTARY 4 F 5' MOMENTARY MOMEHTARY MOMEHTARY
, MOMEHTARY g.'0p;ap NONE UNKNOWN CONDUCTOR UHKNOWH UHKHOWH UNKNOWN UNKNOWN 2/16/78 5/14/78 6/ 4/78 8/'21/78 6137 7145 5146 23'9
~ p 0
~
p 4
Q 6 3'7 ~ 30
. 0:cc,30 NONE 7".47:
0 g,02:PP NONE MOMENTARY MOMENTARY UNKNOWN LIGHTHIHG ARRESTE WEATHER IN AREA UNKNOWN TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED MOMENTARY CONDUCTOR LIGHTNING ARRESTER LIGHTHIHG UNKNOWN WEATHER IH AREA NOH-DIS ~ LIGHT~
TOTAL 1
1 20 0
0 0
0 14 1
1 16
FLORIDA POWER 1 LIGHT COMPANY TRANSMISSION INTERRUPTION
SUMMARY
SYSTEM OUTAGES EXCLUDED LIHE SECTION MALABAR TO MIDWAY 42 240 KV 1975-1978 LE COLE 52479 TIME PART -- ALL DATE -- OFF -- ON -- OH --
DMG ITEM CAUSE L
7/15/78 1125130 MOMENTARY UNKNOWN UNKNOWN TOTAL TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED MOMENTARY 0
FLORIDA POWER R LIGHT COMPANY TRANSMISSION IHTERRUPTiON
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION MIDWAY TO RANCH 41 240 KV 1975-1978 LE COLE 52479 g3.3I ~
DATE -- -- OFF TIME PART OH ALL ON DMG ITEM CAUSE 6/19/77 7/ 3/77 7/ 7/77 7/22/77 8/10/77 8/18/77 8/27/77 10/ 1/77 10/17/77 11/11/77 12/29/77 5/1 4/'78 6/17/78 8/ 5/78 8/12/78 10/15/78 10/20/78 10/30/78 12/14/78 15142+15 2+37 ~ 30 5'21 0
5 47 0
21571 0
14 '01 0
22 '145 31521 0
4 54130 23+19'. 0 21 '9145 71451 0
15+10!15 7 '4115 21571 0
6!54' 1'~
0 5143 '0 19135130 2150' MOMEHTARY MOMENTARY MOMEHTARY 2157 '5 17'5' MOMENTARY MOMENTARY 11'49' MOMENTARY 6127 '
7148115 9+165 0
MOMENTARY MOMENTARY MOMEHTARY MOMENTARY MOMENTARY MOMENTARY Ige32'<
X ARM ogpo: l5 3;g5 ',ob HONE HONE 4:$$:30 CONDUCTOR 5':2q;l5 CONDUCTOR g~;OB: IK'ONE
.Ie.;;<g NONE VAHDLALISM INSULATOR UNKNOWN UNKNOWN HOH-DIS ~ LIGHT+
X-ARM UNKNOWN UNKNOWN IHSULATOR UNKNOWN iNSULATOR LIGHTNING ARRESTE X-ARM WEATHER IH AREA UNKNOWN UNKHOWH UNKNOWN UNKNOWN UHKHOWH TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED MOMEHTARY X-ARM IHSULATOR LIGHTNING ARRESTER UNKNOWN WEATHER IH AREA HOH-DISo LIGHTS VAHDLALISM TOTAL 22 1
0 0
1 1
0 1
0 10 1
0 0
FLORIDA POWER L LIGHT COMPAHY TRANSMISSION INTERRUPTION
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION IHDIAHTOWN TO MIDWAY 240 KV 1975-1 978 LE COLE 52479 DATE -- OFF TIME PART ALL ON -- OH --
DMG ITEM CAUSE 4/12/76 1 6 4 21 ~ 45 MOMENTARY 9/23/77 5450130 MOMENTARY UNKNOWN NOH-DISo LIGHT~
TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED MOMENTARY UNKHOWH NON-DISe LIGHT+
TOTAL 00
FLORIDA POWER 5 LIGHT COMPANY TRANSMISSION IHTERRUPTION
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION IHDIAHTOWN TO PRATT WHITNEY 240 1975-1978 LE COLE 52479'V DATE -- OFF TIME PART -- ALL OH OH DMG ITEM CAUSE 6/17/'78
. 15 ~ 10 ~ 45 MOMENTARY UNKNOWN UNKNOWN TOTAL TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED MOMENTARY 0
0
LE COLE 52479 FLORIDA POWER 8 LIGHT COMPANY TRANSMISSION INTERRUPTION
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION PRA'TT WHITNEY TO RANCH 42 240 KU 19'75-1978 TIME PART ALL DATE -- OFF -- ON ON DMG ITEM CAUSE 4/'/76 100564 0
MOMENTARY 10/12/77 1 C 15130 2126 C 15
/~o/0~i 45 NONE RELAYED WHEN CLOS RELAY TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED 'OMENTARY RELAY RELAYED WHEN CLOSED TOTAL 1
0 0
1
FLORIDA POWER 8 LIGHT COMPANY TRANSMISSION INTERRUPTIOH
SUMMARY
SYSTEM OUTAGES EXCLUDED LIHE SECTION MIDWAY TO PLUMOSUS 138 KV 1975-1978 LE COLE 52479 DATE -- -- OFF TIME PART -- ALL ON --'H DMG ITEM CAUSE 1/ 9/75 3/14/75 ip/25/75 5/15/76 6/29/76 8/20/76 9/12/76 9/17/'76 12/13/76 j.2/14/76 12/16/76 9$ 36$
0 91281 0
6of31 0
164211 0
23 '9130 14 '1$ 45 15115130 131251 0
'116115 14525030 14!211 0
9139!
0 MOMENTARY 17419!
0 MOMENTARY 23$ 20$
0 14014445 15o16!
0 MOMENTARY MOMENTARY MOMENTARY
'MOMENTARY po%3:00 GUY MIRE lle06'oo POLE
'%'00:30
-. HONE 00'o3'oc CONDUCTOR oo eo.-~
INSULATOR FPL CONT ~
CREM SMITCH VEHICLE HON-DISo LIGHT+
TRAHSFORMER VEHICLE VAHDLALISM HOH DIS ~
LIGHTS UNKNOWN UNKNOWN UNKHOWN 1/17/77 2/17/77 4/21/77 6/ 4/77 8/26i'77 8/26/77 8/26/77 9/ 1/77 9/20/77 9/22/77 10/12/77 11/ 5/77 ii/ 7/'77 11/'16/77 12/13/77 12/23/77 1/ 3/78 2/18/78 3/ 3/78 3/18/78 3/18/78 5/14/78 6/ 9/78 8/11/78 9J'15/78 9/24/78 10/11/78 11/29J'78 12/ 1/78 9t27 ii+18 9$ 35 23129 is ~ 21 15122 is ~ 22 7'7 191S3 5+50 8'7 181 2
131 0
7+51 81S4 9'3 8'8 150 8
13+26 71 1
7141 7'45 is+18 7145 20 '
17124 7142 10'8 14019 t30
~
p
~ p
~
Q 145
~ 30
$ 45 145 p
$ 30 les 0
~ p 0
$ 45 p
4 Q
130
~ 30
~
Q
+30 0
145
!30 0
0 0
p 145 9528 ~
0 iioi8$15 MOMENTARY 9
7 0
MOMENTARY MOMENTARY MOMENTARY MOMENTARY MOMENTARY MOMEHT*RY MOMENTARY MOMEHTARY MOMEHTARY MOMENTARY MOMENTARY MOMENTARY MOMENTARY MOMENTARY MOMENTARY MOMENTARY MOMENTARY 7'59 '0 MOMENTARY MOMENTARY MOMENTARY MOMENTARY MOMENTARY 101591 0
17o33$
0 Oos I4l 30 HOHE 00',01:oo 3-l3.I5 NONE HONE Oo.'OO.30 HONE oo'rOo il5 NONE
?; gg:00 HONE FPL CREM RELAYED WHEN CLOS X-ARM X-ARM UNKNOWN UHKNOMH UHKHOMH SMITCH UNKNOWN UNKNOWN UHKHOMH RELAYED MHEH CLOS FPL CREW RELAYED MHEH CLOS RELAYED WHEN CLOS RELAYED WHEN CLOS RELAYED WHEN CLOS MEATHER IN AREA WEATflER IH AREA SMITCH SMITCH LIGHTHING ARRESTE UHKHOMN SWITCH HON-DIS ~ LIGHT+
WEATHER IH AREA SMITCH SWITCH FPL CREM
LE COLE 52479 FLORIDA POWER 5 LIGHT COMPANY TRANSMISSION IHTERRUPTIOH
SUMMARY
SYSTEM OUTAGES EXCLUDED LINE SECTION MIDWAY TO PLUMOSUS 138 KV 1975-1978 TIME PART -- ALL DATE OFF -- OH -- OH --
DMG ITEM CAUSE 12/ 8/78 7'0t30 MOMENTARY SWITCH TOTAt OUTAGES BY CAUSE CAUSE SUST*IHED MOMENTARY SMITCH RELAYED WHEN CLOSED X-ARM LIGHTNING ARRESTER VEHICLE FPL CREW FPL COHT ~
CREW UNKNOWN WEATHER IH AREA NON-D IS o LIGHTi VANDLALISM TRANSFORMER TOTAL 1,
1 1
1 22 1
0 00 1
1 5
1 0
0 1
0 10 330 0
30
FLORIDA POWER 4 LIGHT COMPANY TRAHGMISGION IHTERRUPTIOH SUHHARY SYSTEH OUTAGES EXCLUDED LINE SECTION PLUMOSUS - TO - RIVIERA 41 130 KV 1975-1970 TIHE
- PART - -- ALL DATE -- -- OFF -- --
OH -- --
OH - ~ ---
DHG ITEH LE COLE 60179 CAUSE 5/13/75 6/20/75 6/20/75 5/ 7/76 7/26/76 10/27/76 6>>4it 0
5>> pt 0
5>>43t 0
MOMENTARY MOHEHTARY HOHEHTARY 10>>23t 0
MOHENTARY 16>>55>>
0
. 16t50>>30 00;03130 HONE 16t40t 0
IIOHEHTARY LIGHTNING LIGHTNING HOH-DIS>>
LIGHT>>
GUY WIRE UNKNOWN UNKNOWN 6/ 7/77 0/11/77 9/21/77
'9/22/77 3/ 3/70 4/20/70 5/ 6/70 6/21/78 6/21/78 6/21/70 10/21/70 9t24t 0
10t 9t 0
15t29t 0
6>>17I15 14ti9>>15 11 t33>> 15 14>>20t45 5>> 9t 0
5t2it30 19>>16t30 10>>29>>30 HOHEHTARY HOMEHTARY HOHEHTARY HOHENTARY 14I22t30 HOHEHTARY 14>>31>>15 MOMENTARY MOMENTARY HOHEHTARY 22>>39>>
0 00t03tf5 POLE PP;P2.30 HOHE OY'130 HONE UHKHOWH HOH-DIS>>
LIGHT>>
HOH "D IS o LIGHT>>
NOH-DIS>>
LIGHT>>
WIND UNKNOWN HOH-FPL CONT>>
INSULATOR INSULATOR HOH-DIS>>
LIGHT>>
GUY WIRE TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED HOMEHTARY IHSULATOR GUY WIRE HOH-FPL CONT>>.
LIGHTNING WIND UNKNOWN HOH-DIS>>
LIGHT>>
TOTAL 0l.
1 1
1 0
0 2
1 0
1 0
4 5
13
FLORIDA f'OWERi S LIGHT COMPANY TRAHGMISGION INTERRUPTION
SUMMARY
SYSTEM OUTAGEG EXCLUDED LINE SECTION PLUMOBUB -
TO - RIVIERA f2 138 KV 1975-1978 TIME PART
-- ALL-DATE -- -- OFF -- --
OH -- --
Ok -- ""- DMG ITEM LE COLE 60179
)g,$4 mu.
CAUSE----
4/11/75 11/ 3/7S 11/ 3/75 11/ 3/75 ii/ 4/75 11/ 4/75 li/ 4/75 2ie3Ce 0
16e45e 0
16e47t 0
17t19t 0
St46e 0
3e48e 0
5t29t 0
MOMEHTARY MOMEHTARY MOMENTARY Oe 4e 0
15t52t 0
MOMEHTARY MOMEkTAf'Y GA5:0O NONE 10; p(;Op INSULATOR NOk-DIG. LIGHT INSULATOR INSULATOR INGULATOR SALT SPRAY UNKNOWN UNKHOMN 3/20/76 3/28/76 6/17/76 7/16/76 9/14/76 1tlCt 0
ltb9t 0
9e 3e 0
16t26t30 14e18e 0
MOMENTARY MOMEHTARY 9e 4t 0
Pggf;00 kONE MOMENTARY 14 t 19 1 0
P010ftQP HOkE INSULATOR It!SULATOR NON-Ff'L CONT ~
UHKttOWN TRANSFORMER 5/23/77 Ct 11!45 MOMEHTARY 6/29/77 15e35'e. 0 MOMENTARY UNKNOMtt HON-DIG LIGHT 1/ 0/78 20e40'e45 3/22/78 Ct56e15 3/2'3/78 7e22e 0
4/13/78 7l 0t 0
4/20/78 5t54eiS 20:41l15 Co,'oo:35 NONE 6:57t 0
Oo;oo:4$
NONE 7 e 26 e 0
OOt04:00 HONE MOMENTARY 5: 54: 45 001 OOOO HONE MIND TRAkQFORHER TRAHGFORtlER RELAYED WttEN CLOS TRANSFORMER TOTAL OUTAGES BY CAUSE CAUSE SUSTAINED RELAYED WHEN CLOSED 0
INSULATOR 1
NOtt-FPL CONTe 1
BALT SPRAY 1
MIND 1
UNKNOWN 0
NOH'-DIS e LIGHTe 0
TRAttBFORMER 4
MOMENTARY 1
4 0
0 0
4 2
0 TOTAL