ML20128D427
| ML20128D427 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Shoreham |
| Issue date: | 12/06/1983 |
| From: | Papazoglou I BROOKHAVEN NATIONAL LABORATORY |
| To: | Chow E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19292B772 | List:
|
| References | |
| CON-FIN-A-3740, FOIA-84-624 NUDOCS 8505290070 | |
| Download: ML20128D427 (42) | |
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4 BROOKHAVEN NATIONAL 1.ABORATORY
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ASSOCIATED UNIVERSITIES, INC.
Upton. Long Island. New York 11973 (516)282s j
Deportrnent of Nuclear Energy FTS 666' 2435
.l December 6,1983
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7 Mr. Ed Chow Reliability and Risk Assessment Branch Division of Safety Technology Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Ed:
Enclosed please find a copy of a memorandum to I. A. Papazoglou from Y. H.
Sun, E. Anavim, and K. Shiu, containing a number of questions on the Shoreham PRA. These questions have been generated in the course of our review of the PRA.
This submittal satisfies the contractual requirement of Task 2 of FIN A-3740.
If you have any questions please do not hesitate-to contact me or the authors of the memorandum.
Sincerely,
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- 1. A. Papazoglou, G er Risk Evaluation Group IAP/dm Enc.
cc:
A. Busiik w/o enc.
A. Thadani R. Hall w/ enc.
R. Bari 8505290070 841015 PDR FOIA SHOLLYB4-624 PDR
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BROOKHAVEN NATIONAL. LABORATORY MEMORANDUM
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DATE:
-Decemoer 6, 1983 To:
Ioannis A. Papazoglou 1
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FROM:
Y. H.,lM, E. Anavim, K. Sitiu
SUBJECT:
I Questions on Shoreham PRA
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Attached is a list of questions generated in the course of the Shoreham PRA review to date. Additional questions may arise subsequent to the submittal of these questions and the reviewers would appreciate opportunities to discuss them with the authors of the PRA.
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- 2) How many valves would be involved in the P functions (SRVs reclose)? (p.
3-38) y
- 3) What type of "LOCA" is associated with the failure to reclose more than one SRV7 What is the success criteria for a transient event given (1) I SORV and (2) two or more 50RVs? Provide basis support by documentation of g
f your answer.
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- 4) In the transient and ATWS event trees, how is the repetitive opening and closirig of the safety relief valves modeled?
- 5) The PCS system is assumed in the Shoreham study to be sufficient to remove
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decay heat from containment given a failure of the P function. What is the success criterion for the PCS with regard to one, two, or more than 2 SORVs? Provide a basis to support the answer.
- 6) On page 3-38, it is stated that "only a fraction of the eleven valves are required to open to be successful." What is the basis for this statement?
How many of the eleven SRVs are involved? Would the same number of SRVs be included in other kinds of transients?
- 7) On p. 3-42, "GE analysis for Limerick-PRA indicates that once the MSIVs' have closed, the ADS and low-pressure system injection initiation can be delayed for times in the range of 30 minutes without fuel clad temperature exceeding 2200'F."
Is this assumption valid for Shoreham? Provide a basis to support this assumption.
- 8) Explain why the same failure probability (5x10-3) is being used for PCS unavailability for turbine trip and loss of feedwater (pps. 3-35 and 3-81). Provide the basis on how the value of 5x10-3 is derived.
- 9) Provide the basis of arriving at a probability of 70% for the MSIV re-opened and feedwater recovered for the MSIV closure initiator, Figure 3.4-3.
- 10) Provide the basis of assuming 98% probability for the " Condenser Pump Injection" availability. Does this probability reflect the fact that the initiator of the event is loss of condenser vacuum.
(p. 3-8, Table 3.4-5)
- 11) As stated on p. 3-89, the severity of the resulting condenser vacuum transie.nt is directly dependent upon the rate at which the vacuum pressure is lost. How is the analysis on a loss of condenser vacuum performed to ensure that the most severe scenario is included?
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- 12) How do the success criteria differ for the -1 10RV and 2 or more 10RV cas-es from those of.150RV and 2 or more SORVs? In the present analysis it 4"..
is stated that the 10RV acts initially as a small LOCA.
Is this a valid assumption regardless of the number of 10RVs involved? (p. 1-30) t p
- 13) Once the reactor is shut down, the 10RV event tree is similar to the
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turbine trip transient event tree (p. 3-134).
If so, why does the feedwater recovery probability differ from each other in the 10RV and q
turbine trip event tree analysis? (Figures 3.4-1 and 3.4-7) 14). How does the a'dditional. heat that is discharged into the.suppres'sion pool 1
during the init'ial period of a 50RV affect the definition of success criteria? Would the failure probability of 5x10-3 for PCS still hold l,
if there were more than one IORY involved? (p. 3-134 and Figure 3.4-7)
- 15) Miat is the basis of using 8.6x10-Il per hour /section as the pipe rup -
ture iaflure rate for the break. exclusion pipe? What is the basis for
.using 1.5x10-10 per hour as the valve external leak / rupture rate for break exclusion valves? (P.A-24)
- 16) LOCA frecuencies listed in Table A.1-7 (initiating frequency for a large
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LOCA in nain steam lines within the reactor building) cannot be re-g produced " rom the data supplied in Section A.1.3.3.
Clarify the evalua-tion process. (p. A-27).
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- 17) LOCA frequencies (2.lE-6) listed in Table A.1-8 for HPCI unisolated pipe cannot te; reproduced from the supplied data. Clarify the calculation.
-(p. A-23}.
-18) Clarify the reference leg design of the reactor water level system.
It is indicated that there are 51 instruments connected to the reference legs at the Shoreham plant. Clarify potential common mode failure of instruments due to failures at reference legs.
(p. A-40)
- 19) Provide a discussion on the failure of the reactor water level system re-ference leg that will cause the level instruments to indicate a low level.
(p. A-40)
- 20) It is stated in the Shoreham PRA that the surveillance of the reactor water level instruments at most operating BWRs is performed each time the plant is manually shut down (4.3/ reactor year). At Shoreham plant, it will be performed only at refueling outage (1/ reactor year).
It was thus assumed (in the PRA) that the initiator frequency of the reference leg failure at the Shoreham plant is 4.3 times lower than most of the other BWRs, because the maintenance error at Shoreham is reduced due to less frequent surveillance actions. However, the Shoreham PRA didn't clarify why a less frequent surveillance is enough for detecting potential instrument failures and so to optimize the reliability of the reference leg instruments.
(pps. A-47 and A-48).
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21).
The initiator frequency of the high drywell temperature was determined as 0.0093/ reactor year. This was determined from LERs of the period from 1971 to 1981. Only two events were found in this period of-t'ime.
How does this initiator frequency compare with the failure rate (hardware failure and other potential common mode failure) of the drywell cooler
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systems?. (p.A-51).
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year or 6x10-3 per reactor year. However, based-on NUREG-0666 the J
operators' failure to recover from an unavailable DC bus has been as-signegaconditional.probabilityof0.5.. That is, the frequency of.
3x10-per reactor year has been assumed for Shoreham PRA for single x'
bus failure (p. A-34).. Is the 30 minutes recovery assumption consistent with-conditional probability-of 0.5 used in deriving the loss of single
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DC bus initiator frequency?
- 23) A discussion is presented on pages A-34 and A-35 about the initiator fre-
.quency of multiple DC bus-faigures. NUREG-0666 reported the failure of one DC divgsion to be 6.0x10- and the. failure of two DC divisions to i
Provide details on how the value 2.0x10-4 per challenge be 6.0x10.
F for conditional probability of losing the second division was derived.
- 24) Table 3.2.3 (p. 3-315)'given a failure frequency of 2.5x10-3fe,y (1/1400) for the service water system.. A service water frequency of
- 1/215 or 4.6x10-3/r.y was used, however, for sequences following a loss of: service water initiator, (Figure 3.4-56, p. 3-316).. This inconsist-ancy appears to originate from assuming the total BWR reactor-years of
- operation.as 400 years (p.- A-52) rather than 215 years (p. A-45, Table A.1-16). Clarify this difference.
- 25) What is the basis for a scram failure of 10-5 under large a'nd medium LOCA? It-is stated that the common mode failures of all control rods to insert we'r's considered. However, it is not clear which common mode-
. failures were considered and how the common mode failures were modeled in the quantitative evaluation.
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- 26) What is the basis for assuming that the failure of the vapor suppression
- mechanism 1s dominated by the drywell floor seal and other penetration
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seal failures, instead of failures such as downcomer vent pipe failures?
-Why-is this assumption, which was used in WASH-1400 and the Limerick PRA,
. applicable to the Shoreham case? Provide a discussion on other imechanisms, such as the vibration failure (environmental stress) of downcomer vent. pipe under failed SRVs or large LOCA conditions, that were considered in the.PRA? (p. 3-146).
- I'
- 27) The he'at removal capability of the PCS was not considered in the large-
.LOCA, event tree, and the reason was stated to be the prevention of the release of fission products inside the primary system. Why was this con-sideration not applied to the medium and the small LOCA cases?
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- 28) The "LOCA outside containment" event includes.only large LOCA but not medium and small LOCAs. While it might be generally true that the con-sequences of medium or small LOCAs could be considerably.less, the poten-tial for system interactions resulting from medium or. small LOCA owing to spray in the vicinity of vital equipments in the reactor building seems r
to have been neglected. Please comment. (p. 3-156, 3-231).
I 29)' The potential water sources (p. 3-232, Table 3.4-21) for excessive water
- j at elevation'8 should include reactor coolant water that is released to i
elevation 8 during a LOCA. This event has not been included in Section
,d 3.4.2.4 (Loss of Coolant Accident outside Containment) because, in that f}
section, ECCS are considered to be available at the beginning of the ev-ent.
- 30) :It is assumed that most pumps, turbines, and electrical panels are dis-abled if water level is high enough to contact electrical features on the equipment. However, equipment malfunctions may occur under water spray or high moisture conditions, which may happen before the water level is high enough to submerge the electrical components. What judgement has been made for conditions other than high water level situation? (p.
3-231).
- 31) It is assumed that the potential initiators for the internal flooding z
should have a water leakage rate of greater than the capacity of sump pumps (640 gpm). What assumption was made concerning the sump pump failure, and the loss of. AC power or DC control power to sump pumps? -(p.
G-4 ).
- 32) It is stated that 3'-10" is the critical height for the core vulnerable V
analyses under internal flood condition (pps. 3-237, G-5).
- However, Table G-3.2 shows that many ECCS control components are arranged at levels lower than 3'-10", such as HPCI vacuum pump, HPCI condensate pump, RCIC instrument racks, HPIC instrument racks. How is'the failure of the-se components with water level below 3'-10" been considered in the an-alysis?
- 33) Frequencies of the internal flooding (initiated by pipe failures) listed in Table 3.41-23 and Table G.4-5 are not consistent. Please clarify which frequencies were used in the analysis.
'34) Please clarf fy why the manual error of opening the isolation valves dur-ing major maintenance action is judged to be insignificant. The opening of-the isolation valves during maintenance action will result in internal.
flooding ~ in the reactor building (p. 3-246)..
- 35) What is the basis for assuming a negligible probability of S/R valves -
failing to open under the condition of loss of one division 125V DC bus?
(p. 3-277).. How does the loss of one DC division affect the success of the ADS function and what manual actions are required if the ADS is not sufficient to depressurize?
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'36) It is. stated that the loss of one DC bus initiator has been considered as the bounding calculation. Please clarify the reason why loss of multiple
.y DC bus incidents (for example, loss of two DC bus divisions) will lead to
'a lower conditional probability for the vulnerable core.
(p. 3-278).
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- 37) Based on the Shoreham Emergency procedures, the primary backup for the loss of reactor building service water is the turbine building service water. Although the motive forces for the two service water systems are separated, the two systems use the _same water source. As indicated in Shoreham PRA, the operation experience with the loss of _ reactor buildup service water is mainly originated from problems of water intakes (con -
tamination of intakes by sand, ice, and/or clam growth, or by the loss of instrument air). There is, therefore, a significant possibility that the turbine building service water may be in the same trouble when the re.
actor building service water is in trouble. Was this common mode failure corisidered in the analysis? (p.3-315).
i 38) With the loss of reactor building service water or the loss of AC power, the room coolers would be unavailable and the RBNVS and RBSYS would be isolated. Please clarify the effect of these events and indicate the possibility that the reactor building temperature would rise to the point that HPCI and RCIC will be isolated from operation. (p. 3-314).
39)- On p.1-29 in the discussion of success criteria for large LOCA, it is stated that "at some locaticns, breaks toward the lower end of the size range may require operation of one or two safety relief valves", provide " w '
the rationale on why these LOCAs are not included in the medium LOC. ev-ent.-
- 40) It is stated on p.3-32 that "...the turbine trip event is that closure'of the turbine stop valve from high power will drastically reduce steam flow H
from the reactor vessel while the feedwater flow will continue. These two functions result in a water level swell in the reactor vessel to
. Level 8...and may trip the feedwater pumps on Level 8 trip set point".
If analysis indicates that there would be a Level 8 trip given a turbine trip event, provide the basis why all turbine trip events are not treated as loss of feedwater events.
- 41) In the turbine trip event tree, Figure 3.4-1, two. feedwater related functions are identified. Provide more detailed discussion on these two feedwater functions and how their conditional probabilities are derived.
/ 42)
In the support system screening event trees, the transie of offsite power function is assigned a-value of 3.1x10 gt induced loss Provide the basis on how and why this valu was selected for this analysis instead of the WASH-1400 value of 1.0x10-5 5
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- 43) The Shoreham PRA states that the HPCI system is completely independent of AC power. However, within the HPCI system, the loop level pump and the b'#
inboard isolation valve (F002) are both powered from AC 480V bus.
It is d.
not clear whether the unavailability of these two components, which are
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. Vital to HPCI operation, as a result of the loss of the corresponding AC
'c; 480V bus has been considered in the analysis.
(p.B-1).
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'44) The steam supply to the HPCI pump turbine could be.diversed from the steam supply-line through the opening of valves F028 (Figure 8.1-1).
Clarify the functions of these valves and the importance of loss of steam 5
supply-to HPCI pump turbine due to the opening of these valves.
(p.
-B-2).
- 45) Loss of HPCI flow will occur when the following two automatic operations fail:
reclose of the mini flow line after HPCI pump reaches the rated speed.
. - transfer of the water source from CST to suppression pool after CST is depleted.
. Clarify the importance (to the loss of HPCI flow) of the relay logic failures which fail the two automatic operations stated above.
46)_ The steam supply to the RCIC pump turbine may be lost due to the' opening of valves F025 and F026 (see Figure B.1-2).
Clarify the function of the-se valves and the consequences when those valves are at open position.
- 47) It is stated in Shoreham PRA that fault trees of the diesel generator system and the scram system were not used for the quantification of these systems,-(p. 1 Vol. IV). How are the unavailabilities of these systems calculated? What is the purpose for the development.of these system fault trees contained in Volume 47
- 48) ~ Clarify the possible failures of diesel generators and off-site power line due to synchronization error during loaded testing of diesel N
generators. Since the automatic synchronization mechanism is not instal-led at Shoreham plant, the synchronization error may occur because of hu-man error _or because of synchronization check relay malfunction (p.
B-110).
- 49) Clarify the importance of the maintenance unavailability and/or the com-mon mode failure between diesel generator fuel oil transfer pumps (to supply fuel from storage tank to day tanks). These failure modes are not included in the fault tree for the diesel generator systems.
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- 50) If the reactor pressure is above 403 psig, the LPCI injection valves are closed and the LPCI flow is bypassed to the supnression pool via the minimum flow line. Clarify potential RHR pump failures owing to inadver-tent closing of the minimum flow line, which will cause RHR pump over-O heating when LPCI injection valves are closed.
(p. B-38).
This question is also applicable to the closing of the minimum flow line in the core spray systems.
3 v 51) On p. 3-174, the Shoreham PRA stated that in the event of an ATWS, the high pressure systems are used to mitigate the accident and depres-
-1 surization of the reactor vessel is to be avoided. This decision, ac-cording to the PRA, reduces the potential for problems with controlling water level during and after depressurization. Discuss in more detail the basis why depressurization and low pressure systems are not con-sidered for ATWS mitigation. Provide any thermal hydraulic analysis that was performed to justify the problems associated with water level con-trol.
- 52) On p. 3-178, the PRA discussed the requirments for the operator to properly control high pressure injection given a turbine trip ATWS. Re-actor water level is to be maintained above Level I with the HPCI or RCIC system. How is this requirement reconciles with the emergency procedure guide that states that during an ATWS, the RPV water level is to be maintained at TAF?
- 53) In the discussion of HPCI reliability, p. 3-197, the Shoreham PRA presented that the HPCI can remain functional in an ATWS event with high suppression pool temperature for 40-50 minutes. At that time, hot shut-down condition is achieved and the RCIC will be adequate for coolant injection.. Why is the RCIC lube oil system not subjected to the effects of high suppression pool temperature just as the HPCI? If they are similar, what is the basis for assuming that the RCIC is available sub-sequent to the HPCI failure due to high suppression pool temperature.
- 54) Provide the basis for assuming that at suppression pool temperature below 240*F, the added unreliability of HPCI owing to lube oil cooling is small and that the 50% increase in HPCI unreliability is conservative, p.
3-197.
- 55) According to the information provided in the Emergency Procedure Guide (SPp29.024.01, Rev. 0), no action is required of the operator to inhibit A
ADS prior to termination of all injection except CRD, RCIC and HPCI and w
maintenance water level at TAF. Provide clarification on why this is not included in the procedures and on any other detailed procedures that the operator has to follow to inhibit ADS.
- 56) On p. 3-198, it is stated that for some sequences, continued high pres-sure system operation could actually increase public risk since it would delay core melting until after containment breach. Such a statement implies that termination of high pressure system operation could indeed be beneficial to the reduction of risk for certain ATWS sequences.
Provide the rationale and justification to support such a statement.
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TYPES OF MASONRY CONSTRUCTION MASONRY A.
GOOD WORKMANSHIP, MORTAR, AND DhSIGN; REINFORCED', ESPECIALLY LATERALLY,-AND BOUND TOGETHER BY USING STEEL, CONCRETE, ETC) DESIGNED TO RESIST LATERAL. FORCES'.
UBC ZONE.4.
MASONRY B GOOD 'WURKNANSHIP AND MORTAR; REINFORCED, BUT NOT DESIGNED IN DETAll TO RESIST LATERAL FORCES - UBC ZONES 0 OR.1 MASONRY C.
ORDINARY-WORKMANSHIP dND MORTAR; NO EXTREME
. WEAKNESSES LIKE FAILING TO TIE IN AT CORNERS, BUT NEITHER REINFORCED NOR DESIGNED AGAINST
-HORIZONTAL FORCES. -
MASONRY D WEAK MATERIALS, SUCH AS ADOBE; POOR MORTAR, LOW STANDARDS OF WORKMANSHIP, WEAK HORIZONTALLY
r EARTHQUAKEEFFECTSCORRESPONDINGTd
~-
MODIFIED MERCALLI IN1ENSITY LEVELS MM INTENSITY DESCRIPTION OF EFFECTS VII DAMAG5'0CCURSTOMASONRYD,INCLUDINGCRACKS; SOME CRACKS APPEAR IN MASONRY C VIII-
. DAMAGE 0CCURSTOMASONRYC,WITHPARTIALCdLLAPSE;-
SOME DAMAGE OCCURS TO MASONRY B, BUT NONE TO MASONRY A.
IX-MASONRY.,D IS' DESTROYED; MASONRY C IS HEAVILY DAMAGED, SOMETIMES WITH COMPLETE COLLAPSE; MASONRY B IS SERIOUSLY DAMAGED',
X MOSTMASdNRYANDFRAffESTRUCTURESAREDESTROYED, i.
WITH THEIR FOUNDATIONS.
E e
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e CORRELATIONBETWEENEPAANDMMiNTENSITIES fk"-b f h ?45VV,fpsY UppE!!!EE5D VALUE'FOR SERIOUS. DAMAGE-MASONRY TYPE MM INTENSITY EPA-(s)*
C VIII 0.25 - 0.30 B
IX 0.f40 - 0.50 A
X 0.60 - 0.80
- SMA METHODOLOGY PREDICTS VERY SUBSTANTIAL DAMAGE AND/OR AT LEAST PARTIAL COLLAPSE OF 50% OF THESE MASONRY STRUCTURES FOR 3 To 5 CYCLES OF THE SPECIFIED EPA LEVELS m
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u RECOMMENDED UPPER BOUND EPA'S
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UPPER BOUND ON EPA'IS.C0fiSERVATIVELY ESTIMATED BY ASSIGNIN THE UPPER END OF THE RANGE OF EPA GROUND MOTION LEVELS DEFINED EARLIER TO AN MM INTENSITY VALUE~.ONE LEVEL LOWER THAN INDICATED BY MASONRY DAMAGE MM INTENSITY UPPER BOUND EPA (s's)
IX 0.8
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Reactor and Control Buildings Failure Mode:
Flexural Failure of Shear Walls Median Item F.S.
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0.20 Damping 1.0 0.12 0.06 0.14 Modeling
'1.0 0
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~
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' O. 31' O.25 0.40 4.93 Median Acceleration Capecity = 0.749 M
~
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COMPARIS0N OF FSCTORS USED IN LIMERICK r ~
WITH THOSE IN MORE RECENT.PRA'S - Puc 8 /8' #cI S 3 S M S C.F) ll4E.4/*4 ff41 yl 5YY 3
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COMBINED F
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- 1. 'l O 14 2.5 1.86 0.17 0.22 h
CURRENT PRA 1.4 0.14 3.5 2.22 0.22 3.11
' a. o -
- c. 2-e EFFECT ON MEDIAN FACIOR OF SAFETY
- 3. S' p 0. 3 ' (
1.0V
{$
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. h$7f lon$. Oh THE DIFFERENCE BETWEEN THEhPERCENTILE FRAGILITY LEVELS FOR lEACTOR AND w
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- i DISCUSSION 6F MODEL ADEQUACY 3
Y ORIGINAL STRUCTURAL MODELS ARE REVIEWED e
e.SOME MEMBER PROPERT-IES ARE INDEPENDENTLY CALCULATE AND COMPARED TO THOSE USED MEDIAN MATERIAL PR0PERTIES ARE COMPARED TO ORIGINA e
ONES EFFECTS ON FREQUENCIES AND LOAD PATHS ARE ASSESSED e
e IN SOME CASES, MODELS ARE RERUN MDMENTS, SHEARS, ACCELERAIIONS, ETC. FROM ORIGINAL e
ANALYSES ARE USED TO DETERMINE MAJOR SHEAR WALL FA (IN-PLANE FAILURES)
FOR BLOCK WALLS, WHERE FAILURE UUE TO OUT-0F-PLANE e
FORCES IS CRITICAL, RESULTS FROM THE ORIGINAL BUILDING MODEL ARE NOT USED EXCEPT FOR THE FLOOR SPECTRA; AN
~
INDEPENDENT ANALYSIS IS DONE 4
6 e
30 1
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3 EFFECTS OF DIFFERENTIAL MOVEMENT OR
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TI.LTING OF STRUCTURES ON P.IPING
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THESE EFFECTS. ARE GENERAlt.Y CONSID5 RED IN FRAGILITY
~-
e EVALUATION -
- .,.... c..
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LIMERICK STRUCTURES AREf00NDED ON COMPETENT ROCK;
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e THERE ARE NO RECORDED INSUNCES OF TILTING OF SUCH
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STRUCTURES
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_7.
e PIPING FAILURE IS ASSUMED WHEN THE STRUCTdRE IT IS ATTACHED TO FAILS e
BEING A ROCK SITE, BURIED PIPING SHOULD NOT BE ASSUMED TO FAIL UNLESS THERE IS EXCESSIVE BLOCK MOTION e
e e
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6 FUNCTIONAL CAPABILITY OF PIPING
- V c.
THE GOVERNING FAILURE MODE FOR BOP PIPING SYSTEMS "
e
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IS TYPICALLY FAILURE OF THE SUPPORTS
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e PIPING FAILURE IS CONSIDERED T0 0CCUR VHEN THE ENTIRE CROSS-SECTION (I.E. ANCHOR POIND OF PIPE IS AT FLOW STRESS LEVELT e
SUBSTANTIAL MARGIN EXISTS BETWEEN THE SUPPORT FAILURE AND THE PIPING FAILURE THEMEDIANCAPACITYFACTORFORPIPINGTdKINGINTOACCOUNT e
DUCTILITY AND SYSTEM BEHAVIOR IS ESTIMATED TO BE 10'.25 e
IF FUNCTIONAL CAPABILITY OF. PIPING IS CONSERVATIVELY ASSUMED TO BE COMPROMISED WHEN MATERIAL. REACHES YIELD, THE MEDIAN CAPACITY FACTOR IS ESTIMAIED TO BE 1.87 FOR EXAMPLE, THE MEDIAN' GROUND ACCELERATION CAPACITY OF ESWS e
PIPING FOR THIS FAILURE MODE IS 1.78s COMPARED TO 3.99s
- s Tyr"F frdne Bue::/
D G
.O e
L
VALVE. FRAGILITY,..
e BASIS
?
PASSIVE VALVES - SAFEGUARDS DATA
-ACTIVE MOTOR OPERATED VALVES - GENERIC ANALYSIS FAILURE MODE FOR ACTIVE VALVE IS OPERATOR SUPPORT FAILURE e
AND RESULTING BINDING '0F STEM
~
P = MEDIAN CAPACITY = 1.5 x MEDIAN YIELD STRENGTH e
C
= 1.5 x 1.'2 x CODE YIELD STRENGTH
= 2 x ALLOWABLE STRESS
=2XGMAk v
NORMAL LOAD PN
= 1.0c SSE LOAD PSSE = 0.70 x ALLOWABLE o = 2.10 e
STRENGTH FACTOR 26
-1 F=PC-PN 33y v
=
3 2.1 PSSE e
SMALL VALVES:
F
=.4.1 3
v LARGE VALVES:
F
= 5.1 3
g
. 30 6
.t i
O I
64 e
e.
e
GENERIC VALVE DATA (SOURCE:
SUSQUEHANNA SQRT PACKAGE)
~
CALCULATED SIZE 1YPE MAXIMUM ACCELERATION.(s) 24" 150#
M0 SATE 12.25 20" 150#'
N0 GATE 8.49 16" 150#
N0 GATE 4.83
~
,10" 150#
N0 GATE 8.07 24" 300#
N0 GLOBE 3.00 -
18" 300#
N0 GLOBE 9.24 12" 300#
M0 GLOBE 7.01 10" 300#
M0 GLOBE 9.46.
20" 300#
MAN GATE 4.40 s
x = 5.80 12" 300#
MAN GATE 4.50 g,,
10" 900#'
M0 GATE 7.79 10" 900#
M0 GATE 4.06 12" 900#
M0 GATE 10.42 14" 900#
M0 GATE.
3.55 10" 900# -
M0 GLOBE ~
2.77 24" 900#
M0 GLOBE 3.00 6" 150#
M0 GATE 7.90 4" 150#
M0 GATE 5.45 3" 150#
M0 GATE 8.56 6" 300#
M0 GLOBE 4.68 4" 300#
M0 GLOBE 5.85 cmx " 4'77 8" 300#
MAN GATE 8.85 6" 600#
N0 GATE 4.15 4" 600#
M0 GATE 4.18 4" 900#
M0 GATE 2.87 3" 900#
M0 GATE 3.17 6" 900#
M0 GATE
'4.40 4" 600#
'M0 GLOBE 3.00 3" 600#
M0 Gl.0BE' 3.00 6" 900#
M0 GLOBE 5.96
o VALVES IN REACTOR BUILDING *
~
CAPACITY STRUCTURAL RESPONSE EQUIPMENT RESPONSE SWR 0FF SIZE F
a 8
g R
U FRS 8
U FRE R
0 A(g) ' a R' 8
8 8
8
'8 U
58" 4.10 0.22 0.41 1.72 0.23 0.13 2.68 0.21 0.28 2.81 0.38 0.53
.t.
>8-**
5.10 0.22 0.47 1.72 0.23 0.13 2.68 0.21 0.28,3.49 0.38 0.56 -
g*
CONSERVATIVE TO ASSUME ALL VALVES ARE LOCATED HIGH IN THE STRUCTURE
- CONTAINMENT ISULATION VALVE 24" AIR OPERATED CHECK VALVE 9
e d
o e
h
s
- (
~
VALVE. FRAGILITY IF. LEAKING IS ASSUMED WHEN STRESS IN VALVE e
t; UPERATOR SUPPORTS REACHES YIELD STRENGTH, MEDIAN STRENGTH FACTOR IS
~
1.335
-1.
v F=
max 2.55
=
3 2.1 J
e MEDIAN GROUND ACCELERATION CAPACITY FOR THIS FAILURE MODE IS,1.750 FOR LARGE VALVES EXCESSIVE LEAKAGE OR RUPTURE OF VALVES IS VERY REMOTE e
4 4
e e
.8 9
9 8
q e
9 e
e
~
.'r -
~
z FRAGILITIES OF CRITICAL INSTRUMENT k
e PRESSURE TRANSMITTER, FLOW TRANEMITTER AND TEMPERATURE ELEMENTS e
BASIS:
SUSQUEHANNA QUALIFICATION DATA
. SAFEGUARDS DATA 3.
RACKMOUNTEDANDLINEMOUNTEDINSTRUMENfS e
i e
FOR RACK MUUNTED INSTRUMENTS, THE FRAGILITY PARAMETERS ARE:
'A
= 3.020~
~
aR = 0.32 u = 0.59 s
e SAFEGUARDS DATA HAVE SHOWN THE LINE MOUNTED INSTRUMENTATION TO BE STRONGER THAN THE RACK MOUNTED INSTRUMENIATION O
e
.