ML20054K279
| ML20054K279 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/29/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Fiedler P JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-03-05.A, TASK-3-5.A, TASK-RR LSO5-82-06-119, LSO5-82-6-119, NUDOCS 8207010382 | |
| Download: ML20054K279 (18) | |
Text
. - -
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J June 29,1982 Docket tio. 50-219 LS05-82-06-119 Mr. P. B. Fiedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
Dear fir. Fiedler:
SUBJECT:
OYSTER CREEK - SEP TOPIC III-5. A, EFFECTS OF PIPE BREAK ON STRUCTURES, SYSTEMS AND COMPONENTS INSIDE CONTAINMENT Enclosed is our final evaluation of SEP Topic III-5. A.
This assessment conpares your facility with the criteria currently used by the regulatory staff for licensing new facilities. We conclude that the plant is I
adequately protected from the dynamic effects of pipe break inside containment subject to resolution of the following in the Integrated Plant Safety Assessment:
A.
Basis for acceptability of consequences of cascading breaks; l
B.
Further justification of adequacy of pipe whip interaction with the drywell liner and containment wall; SW C.
Evaluation of effects of damage from jet impingement on mechanical equipment; and p
use D.
Any significant changes in total pipe stresses as a result of SEP g.
Topic III-6, Seismic Design Consideration.
//
The need to actually implement changes as a result of these items will be determined during the integrated safety assessment. This safety evaluation ntay be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
l Sincerely, l
8207010382 820629 PDR ADOCK 05000219 i
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A 4:DLDennis M. Crutchfield, Chief JLombardo DCi ch ield GL nas 6//f/82 6/2 782 6/2$/82 Operating Reactors Branch No. 5
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i Docket No. 50-219 LS05-82 Mr. P. B. Fiedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, NJ 08731
Dear Mr. Fiedler:
StBJECT: OYSTER CREEK - SEP TOPIC III-5.A. EFFECTS OF PIPE BREAX ON STRUCTURES, SYSTEMS AND COMPONENTS INSIDE CONTAINMENT Enclosed is our final evaluation of SEP Topic III-5.A. This assessment compares your facility with the criteria currently used by the regulatory staff for licensing new facilities. We conclude that the plant is adequately protected from the dynamic effects of pipe break inside contairinent subject to resolution of the following in the Integrated Plant Safety Assessment:
A.
Basis for acceptability of consequences of cascading breaks; B.
Further justification of adequacy of pipe whip interaction with the drywell liner and containment wall; C.
Evaluation of effects of damage criteria for jet impingement; and D.
Any significant changes in total pipe stresses as a result of SEP Topic III-6, Seismic Design Consideration.
The need to actually implement changes as a result of these items will be determined during the integrated safety assessment. This safety evaluation may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
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ORB #5:BC AD:SA:DL JLomb DCrutchfield Glainas 6/#i/8 \\
6/ /82 6/ /82 Dennis H. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing
Enclosure:
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OFFICIAL RECORD COPY usam e-amo i NHC FORM 318 (10-80) NRCM 024o
9' Mr. P. B. Fiedler cc G. F. Trowbridge, Esquire Resident I.nspector Shaw, Pittman, Potts and Trowbridge c/o U. S. NRC 1800 M Street', N. W.
Post Office Box 445 Washington, D. C.
20036 Forked River, New Jersey 08731 J. B. Lieberman, Esquire Commissioner Berlack, Israels & Lieberman New Jersey Department of Energy 26 Broadway 101 Commerce Street New York, New York 10004 Newark, New Jersey 07102-
, Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 J.. Xnubel BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054 Deputy Attorney General State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Mayor Lacey Township 818 Lacey Road Forked River, New Jersey 08731 U. S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Licensing Supervisor
, Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
V I
- 4 SEP EVALU'ATION OF EFFECTS OF PIPE BREAK ON STRUCTURES, SYSTEMS AND COMPONENTS INSIDE CONTAINMENT TOPIC III-S.A FOR THE OYSTER CREEK NUCLEAR PO'n'ER PLANT l
4 5
TABLE OF CONTENTS I.
INTRODUCTION
~
II.
REVIEW CRITERIA III.
RELATED SAFETY TOPICS AND INTERFACES IV.
REVIEW GUIDELINES V.
EVALUATION A.
BACKGROUND
'B.
BREAK CRITERIA C.
INTERACTION EVALUATIONS D.
EVALUATION OF BREAK EFFECTS 1.
DYNAMIC EFFECTS OF PIPE WHIP 2.
CONTAINMENT WALL PENETRATION 3.
EFFECTS OF JET IMPINGEMENT a.
JET IMPINGEMENT ON ELECTRICAL EQUIPMENT b.
JET IftPINGEMENT ON MECHANICAL EQtJIPMENT VI.
CONCLUSIONS m
4 6
l 1
i I.
INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic III-5.A.
" Effects of Pipe Break on Structures, Systems and Components Inside Containment," is to assure that pipe breaks would not cause the loss of needed funtion of " safety-related" systems, structures and components and to assure that the plant can be safely shut down in the event of such The needed functions of " safety-related" systems are those
(
breaks.
functions required to mi.tigate the effects of the pipe break and safely shut down the reactor plant.
l II.
REVIEW CRITERIA General Design Criteria 4 (Appendix A to 10 CFR Part 50) requires in part that structures, systems and components important to safety be f
appropriately protected against dynamic effects, such as pipe whip and discharging fluids, that may result from equipment failures.
The current criteria for review of pipe breaks inside containment are I
contained in Standard Review Plan 3.6.2, " Determination of Break Locations l
and Dynamic Effects Associated with the Postulated Rupture of Piping,"
including its attached Branch Technical Position, Mechanical Engineering i
Branch 3-1 (BTP MEB 3-1).
III. RELATED SAFETY TOPICS AND INTERFACES This review complements that of SEP Topic VII-3, " Systems Required for 1.
The environmental effects of pressure,' temperature, humidity and flooding 2.
due to postulated pipe breaks are evaluated under Unresolved Safety Issue A-24, " Qualification of Class IE Safety-Related Equipment."
The effects of potential miss'iles generated by fluid system ruptures and 3.
also considered and are evaluated under SEP rotating machinery were Topic III-4.c, " Internally Generated Missiles."
l The effects of containment pressurization are under SEP Topic VI-2.0,
" Mass.and Energy Release for Possible Pipe Break Inside Containment."
4.
The original plant design criteria in the areas of seismic input, analysis 5.
design criteria are evaluated under SEP Topic III-6, " Seismic Design Consideration."
IV.
REVIEW GUIDELINES The licensee's break location criteria and methods of analysis for evaluating postulated breaks in high energy piping systems'insid The review relied upon information submitted by the licensee, Jersey Central Power and Light Cocpany (JCP&L), in References 1, 2 and 3.
II above.
When deviations from the review criteria are identified, engineering judge-cent is utilized to evaluate the consequence of postulated pipe break and to assure that pipe break would not cause the loss of needed function of
" safety-related" systems, structures, and components and to assure that the plant can be safely shutdown in the event of such break.
//
V.
EVALUATION A.
BACKGROUND On July 20,1978, the SEP Branch'sent a letter (Reference 4) to KMC, Inc. requesting an analysis of the effects of postulated pipe breaks In that on structures, systems and components inside containment.
letter, the staff included a position that stated three approaches were appropriate for postulating breaks in high energy piping systems (Either p1275 psig or T1200*F). The approaches are:
1.
Mechanistic 2.
Simplified Mechanistic 3.
Effects Oriented The staff further stated that combinations of the three approaches could be utilized if justified.
B.
BREAK CRITERIA In response to our letter, the licensee submitted Reference 1 which summarized all information contained in the Oyster Creek docket pertaining to SEp Topic III-5.A. Attachment 7.c of the document was 15, 1974, in Reference 5, in originally submitted to the NRC on JulyIt included a report of pipe break response to NRC Question 14.c.
evaluations entitled, " Analysis of P1pe Breaks Inside Containment."
Th'e ifcensee has identified postulated pipe break locations in high energy piping systems (either p>275 psig or T>200*F) using a mechanistic approach, i.e. a criteria based c7i stresses in the-In summary, this work resulted in the identification piping system.
of 150 postulated breaks in accordance with the criteria of Regulatory In general, this criteria. postulated pipe breaks at terminal Guide 1.46.
ends and at intermediate locations where the calculated stress exceeded a threshold level, i.e. 0.8 (in+ Sa), but at at least two intermediate The loading conditions of gravity stress, pressure stress, locations.
seismic stress and thermal stress were considered and combined.
In evaluating criteria for postulated pipe break effects on structures, systems and components inside containment, we find it is acceptable to apply Regulatory Guide 1.46 break criteria for ASME Section IH Code Class 2 and 3 piping systems to piping systems designed and analyzed in accordance with USAS B31.1.0. However, based on information provided in this review, it is not clear that the seismic stresses have been In some cases, the maximum seismic stress conservatively calculated.
determined by static analysis is used at all locations in the systems.
A comparison of these seismic stresses with two available piping calculations (Feedwater piping and Main steam piping) from Reference 6, which is performed for SEP TOPIC III-6, indicates differences in points This will result in of high stress in the main steam piping system.Nevertheless, it must be noted different postu?ated pipe break locations.
that the licensee has identified 150 postulated breaks which is about the same number that a new plant under licensing review would have Therefore, our assessment based on the information for inside contaiment.
i
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f
. c"urrently available is that the licensee has provided for a spectrum of postulated pipe breaks which include or envelope We the most likely break locations inside the containment.
recommend that any significant changes in total pipe stresses as a result of Topic III-6 should be reviewed to demonstrate conformance with the licensee's conclusions in Reference 2.
C.
INTERACTION EVALUATIONS The licensee has developed a simplified matrix to study the interaction consequences of these break locations on safety-related systems, structures and components. This interaction study grouped the 150 postulated breaks into three different categories as follows:
Category 1 included 89 postulated breaks which were considered to have no potential adverse consequences and, therefore, no modiff-cations to the plant are necessary.
Category 2 included 48 postulated breaks which could have potential adverse consequences, but which occurred at locations of low stress, i.e. the stresses were less than 50% of the stress criterion of Regulatory Guide 1.46.
The postulated breaks in this category were considered not to recuire specific failure protection because they occur at locations of low stress.
Category 3 included 13 postulated breaks which could have potential adverse consequences and which occurred at locations of higher stress, i.e. the stresses were 50% or greater of the stress criterion of Regulatory. Guide 1.46.
An augunented inservice inspection program modified to be consistent with the NRC request (Reference 8) has been ieplemented for these locatiens (Reference 9) since 1975.
l Potential adverse consequences (Category 2 and 3 breaks) were defined as any interaction with containment, with piping in other high en,ergy lines or safety systems of smaller size, or with electrical and mechanical equip-ment' having a safety function.
On February 6,1979, JCP&L revised some of' the seismic ctress analysis.
The effect of the revised seismic ctress was to redece the number of
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breaks from 150 to 141.
Based on discussions with the licensee during a meeting on February 6, 1979 and subsequent telephone conversations, NRC requested further infor-mation to clarify the consequences of postulated breaks in categories 2 In response to the NRC request, the licensee submitted Reference.3.
and 3.
/o
_4 In' this report, the licensee evaluated the plant's capability to reach safe shutdown with the dyr;amic effects of the p~ostulated breaks. The gen-eral approach taken was to ensure that no break could affect more than one train of a safety system, and.that the intact equipment still contained sufficient redundancy to withstand a single active failure without loss of Interaction with the containment wall was reexamined.to safety function.
This method is determine if the breaks could penetrate the drywell wall.
a refinement of the original interaction matrix, taking into account the The functions of affected systems, physical separation and redundancy.
licensee's conclusion frcm this evaluation was that all breaks had accept-able consequences.
For the Oyster Creek facility, the systems that would be used to reach safe shutdown following a high energy line break inside. containment are the fol-lowing:
Emergency Condensers Core Spray Systems Automatic Depressurization System (ADS)
Instrumentation (reactor water level and reactor pressure)
Motive and control power These systems could also be used to achieve safe shutdown in the absence a break as shown in F.eference 10.
Many of the conponents in the above systems',,.such as the core spray pumps, and the emergency condensers themselves are located outside containment, and thus would not be affected by the break inside containment.
The two low pressure core spray piping systems are physically separated from In addition, each train of the core spray each other within containment.
Therefore, no post-system can withstand a single active component failure.
ulated break and a single active failure can totally disable the core spray The instrumentation hydraulic lines are also physically separated The electrical equipment associated with this instrument-system.
Adequate instrumentation is available for from each other.
ation is outside containmentactuation of safety systems and plant monit The two emergency condenser trains do not have this degree of separation since the supply and return lines for both condensers are located in the The ADS valve controls share a common cable tray which With the separation inside containment and other same quadrant.
circles the containment. electrical components located outside containme Thus equipment to be considered is the cable tray with t interaction studies.
. EVALUATION OF BREAK EFFECTSd D.
It is acceptable under current SEP criteria to use the interaction study to determine the acceptability of plant response to pipe breaks and to assure that pipe breaks would not cause the loss of needed functions of " safety-related" systems, structures and components and to assure that the plant can be safely shutdown in the event of such breaks.
The crucial prerequisite is that the interaction study, i.e., the effects of pipe whip and jet impingement on structures and essential safety equipment must have been adequately and conservatively identified Based on a review of Reference 3, we have determined and evaluated.
that some areas have not been addressed adequately in the licensee's evaluations, as noted in the discussion below:
1.
Dynamic Effects of Pipe Whip
)
The licensee considered in Reference 3 the interactions of whipping pipes with other high energy lines and safety systems based on an assumed direction of motion of the broken pipe.
No unacceptable interactions were identified in this study.
However, in the draft safety evaluation (Reference 13), the staff noted that it is acceptable to use a simplified, reason-j able judgement to detemine the most likely, direct path of motion of.the ruptured pipe by the piping geometry and confi-guration when they are not complicated.
For example, if the postulated pipe break occurs at an elbow downstream of the energy reservoir which supplies a high energy flow stream to the break area and causes formation of a plastic hinge, the jet reactor force, i.e., the motion of the pipe whip, can easily be determined.
However, in the case of a pipe break occurring in the middle of a rather complicated unrestrained piping configuration, it is necessary to investigate all tr.a The complicated dynamic effects possible paths for pipe motion.
induced by the impact and rebound of the ruptured pipe should_
be considered and the extent of which structures and essential equipment are affected by the postulated pipe failure safet should be determined.
Therefore, the staff recommended that the licensee performed the following' steps:
For the 141 break locations, determine which breaks can o
be categorized as " simple geometries" for which the previously assumed pipe motions can be utilized.
Of the remaining breaks, in some cases it may be straight-forward to determine by inspection that there are no nearby o
targets regardless of the direction of the pipe whip.
o' For the remaining locations determine what systems could be impacted by a whipping pipe.
o In determing if pipe whip in other pianes of motion results in interactions not previously identified attention should be focused on the vulnerable targets, such as the emergency condenser.
o If new interactions are defined, demonstrate that conse-quences are acceptable or provide other justification.
In response to this recommendation, the licensee, in its report submitted on May 12,1982 (Reference 14), divided the 141 break locations into two major categories: (1) simple geometry, for which previous interaction studies are applicable, and (2) complex geometry for which further evaluation was needed. Of 141 break locations, 54 locations have simple geometry and 87 locations have complex geometry.
In considering the pipe whip motion in other planes, critical targets examined were other high energy lines, and equipment required for safe shutdown such as the energency condensers, core spray, and essential instrumentation. Based on this reevaluation, the licensee concluded that no further adverse pipe whip interactions were identified.
In considering the pipe whip damage, the licensee has used the conservative assumption that the particular safety syste.a becomes
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inoperative, irrespective of the actual energy which would be involved in the collision or the strength of the target piping system.
However, the licensee's evaluation (Table 3 of Section IV, Rev. 3) has identified 16 postulated breaks which could have a secondary effect, i.e., pipe whip could result in contact with other piping, which, if damaged, could initiate further i
damage.
For example, break number 1 in loop C of the reactor recirculation piping will interact with the north main steam-.line i
However, a break in the north main steam line will interact with l
the north feedwater.
Therefore, the licensee should provide a basis for concluding that the available mitigating systems can cope with the combined blowdown effects of these high energy line breaks (e.g., recirculation line plus main steam line plus feed-l water line) or provide justification that such cascading blowdown l
effects will not occur.
I 2.
Containment Wall Penetration i
The licensee references Chicago Bridge & Iron Company -(CB&I) l Test Report, " LOADS ON SPHERICAL SHELLS," (Reference 7) which indicates that when a spherical shell segment having a shell thickness of 0.75 inches is loaded over a large enough area, i.e., equivalent to a 14 inch diameter or larger circle, defor-mation of over 3 inches can occur without failure of the plate segment.
Based on this test result, the licensee concludes
1 l.
that for breaks occuring in piping of 14 inch diameter or greater, such as main steam piping, feedwater piping, recircu-lation loop piping, even if contact occurred with the drywell liner, the liner :ould deform without failure until deformation was limited by the concrete shield wall. Accordingly, tbe' licensee has concluded that break in these lines could not penetrate the containment liner and wall and, therefore, no unacceptable consequences would result.
However, it should be noted that this test was; performed under essentially static conditions.
It is not clear that the test result is also valid for the dynamic loading which would be experienced as a result of pipe whi'p.
In addition,,the part,icu-lar test applies a concentrated load of 235 tons over an area, equivalent to a 14 inch diameter circle. This' assumption may not always be valid because of the *mpact area of a 14 inch
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diameter or larger pipe may be smaller than the' assumed. area.
Our concern is that in the case of applying a concentrated dynamic load over a small area the steel plate may be perforated before t ;e deformation could be backed up by the concrete shield wall. The probability of perforation could be even higher in' those cases where the gap between the' liner and the-concrete wall is greater than the nominal distance of 2.75 inches.
In response to the above staff concern,.the licensee provided, in its May 12, 1982 submittal (Reference'14), results of analyses performed to correlate a 4-1/2 inch pipe crush tests. Using this correlation, the licensee' intended to demonstrate that greater than one area of a 14-i'f sufficient contact area, i.e'., /
it is reasonable to assume that
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nch diameter circle, will be developed in the event of interaction between a recirculation line, a main steamline, or 18, inch portions of the feedwater piping and the containment wall.
However, the staff feels that it is difficult to correlate the crush tests with the pipe whip on relatively fl(Xible steel liner. Therefore, the licensee should select a worst case configuration to demonstrate that the impact load or energy produced as a result of postulated pipe break for piping of 14' inch diameter or greater does not exceed the load or energy required to penetrate the contaimnent liner and wall.
In performing this evaluation with static analysis or static test, the' dynamic load factor has to be considered. The licensee can take into account the following considerations:
Actual liner thickness with respect to the impact location, a.
b, The combined crack propagation time and break opening time of the pipe may be long enough to depressurize the system such that the whipping pipe could not produce sufficient energy to penetrate the containment; wall.
s
8-3.
Effects of Jet Impingement Standard Review Plan Section 3.6.2 requires' that the jet area expands uniformly at a half angle not exceeding ten degrees and that the total inpingement force acting on any cross-sectional area of the jet is time and distance invariant.
The licensee's original evaluation assumes a jet in the formation of a cone with an angle of divergence of thirty degrees and with the assumption that any safety system inside this cone becomes inoperative, irrespective of the actual jet impingement force. We have determined that this jet impingement criteria are acceptable because they are conservative with respect to the currently acceptable criteria.
ibwever, in the reassessment of interactions with alternate directions of pipe whip motion, the licensee had used a refined assumption as discussed in Item 3.b. below.
a.
Jet Impingement of Electrical Equipment As discussed above, due to the physical separation and redundancy
, of electrical equipment within containment, the only target is the cable tray for the ADS valve controls.
In their interaction studies, the licensee determined that breaks in several different high energy systems, ranging in size from a 2" diameter pipe to the 26" diameter recirculation pipe, could produce jets that would contact this cable tray. These breaks correspond to areas from.02 ft2 up to the largest double-ended breaks considered in the ECCS analyses.
The reference LOCA analyses for Oyster Creek (Reference 11 and
- 12) assumed operation of five ADS valves, one emergency condenser and low pressure core spray. A summary of the sensitivity of peak clad temperature (PCT) to break size is shown in Table 1, Feedwater, stea.n and core spray line breaks are included as well as various size breaks of the recirculation system.
The licensee detemined by inspection'within the contairet the path of the cables from the penetration into the drx. '
to the Jalve itself. Based on the arrangement of cable ron i.gs around the containment, no more than 3 valves could be affected by jets from any break. Therefore, even assuming a single failure..
at least two ADS valves, one emergency condenser and low pressure core spray would be available to mitigate the break effects.
For breaks larger than 1.0 ft2 (13" diameter) no ADS is needed since system pressure is reduced to core spray initiation pressure before the 120 second timer delay on the ADS expires.
9-For smaller breaks, the ADS aids in the depressurization and thus shortens the time until core spray is initiated.
A smaller break of.35 ft2 (8" diameter) was reanalyzed by the
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licensee without ADS or emergency condensers; the calculated PCT was 2025 F.
For this event, the reactor depressurizes through the 0
break with no other energy removal or inventory makcup until the low pressure core spray begins. Performance with one emergency condenser and some ADS valves would be better than the case analyzed with no high pressure means of heat removal or depressurization.
2 2
Previous sensitivities for break sizes from 0.35 ft to 1.0 ft showed an increase in PCT of 130 F.
Since the energy removed through 0
the ADS is less for the larger break sizes, and thus the ADS is less of a factor in determining overall performance, breaks larger than 0
the 0.35 ft2 break should not result in PCTs in excess of 2200 F, even with no ADS or emergency condensers. Note that a break in emergency condenser piping would fall in this range, so that break, plus a failure of the remaining condenser, would not have unacceptable consequences.
' As shown in Table 1,2the PCT decreased markedly with decreasing break size from the.35 ft break down to the 0.1 ftZ(4" diameter.). With some ADS and emergency condenser availability, the PCT for these smaller breaks should not exceed 22009F.
The min steam and feedwater line breaks resulted in substantially lower PCTs than did the recirculatibn line breaks.
The reduction from 5 ADS valves to 2 valves should not result in PCTs worse than the cases discussed above.
Reference 14 noted that nost recent ECCS analysis contained in Reference 15 showed that the emergency condensers are not required for breaks greater than 1.0 ft2 area.
For postulated hijh energy line breaks, the staff concludes that the unaffected ECCS are capable of mitigating the consequences.
Since a large range of break sizes have already been considered, assuming different directions of jet motion should not alter these conclusions.
The staff concludes that the licensee for Oyster Creek has adequately considered the effects of jet impingement of electrical equipment.
b.
Jet Impingement on Mechanical Equipment The licensee considered the effects of jets on mechanical equipment and piping systems in a manner similar to that done for the effects of pipe whip, i.e., jets impinging on adjacent piping were assumed without specific analyses to affect piping integrity. This conservative assumption is acceptable.
- However, in the reevaluation of interactions with alternate directions of pipe whip motion, the licensee assumed that a jet or whipping pipe is considered to inflict no damage on other pipes of
equal or greater size.
It is the staff's position (Reference
- 16) that the effects of jet impingement should be considered and evaluated regardless of the ratio of impinged and postulated broken pipe sizes. This issue will be resolved in the Integrated Plant Safety Assessment.
VI.
CONCLUSIONS Based on the information subnitted by the licensee, we have reviewed the criteria pertaining to the locations, types and effects of postulated pipe breaks in high energy piping systems incide containment. We have concluded that the criteria used to define the break locations and the types are, in general, in accordance with currently accepted standards.
We have also determined that it is acceptable under current SEP criteria to use the interaction study to evaluate the effects of postulated pipe breaks and to determine the acceptability of plant response to pipe breaks.
However, we have found that the consequences of cascading breaks, the containment integrity evaluation, and the damage criteria
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for jet impingement as identified in Sections D.l. D.2, and D.3.b, respectivley, have not been addressed adequately in the licensee's evaluation.
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In addition, we recommend any significant changes in total pipe stress as a result of SEP Topic III-6, Seismic Design Considerations, should be reviewed.to demonstrate that higher seismic stresses do not result in new potential break locations.
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TABLE I 1 Emergency Condenser 1 Emergency Condenser 0 Emergency Condensers
'5 ADS Valves 5 ADS Valves O ADS Valves (Reference 3)
Break (Reference 12)
' (Reference 11)
Stall Brgaks
.10 ft 1212*F.
ADS at 163 sec., ratad core spray at 393.sec.
.35 ft2 1855'F.
1960*F.
2025'F.
ADS at 135 sec., ratee ADS at 137. sec., rated no ADS core spray at '187 sec. core spray at 'iB9 sec.
1.0 ft2 ggg4.p I
no ADS, rated core spray'at 89 sec.
Split Bgeaks 1.0 ft' 2012*F.
2076'F.
2 2013*F.
2123*F.
2.5 ft LargeBregks 6.292 ft CD = 0.4
- 2158*F.
2200'F.
CD = 0.6 2131*F.
2168'F.
CD = 1.0 2060*F.
~
Steam Line 682*F.
Break Core Spray 1335'F.
Line Bregk ADS on, rated core
(.196 ft')
spray at 527 sec.
Feedwater comparable to LineBregk core spray break;
(.998 ft )
ADS on ame e
i-
e REFERENCES 1.
September 7,1978, "SEP Topic III-5. A. Pipe Break Inside Containment,"
Oyster Creek Nuclear Generating Station, Unit No.1.
2.
February 6,1979, "High Energy Piping Systems Inside Containnent, Stress Summary", Oyster Creek Nuclear Generating Station, Unit No.1.
3.
Letter from J.T. Carroll, Jr. (GPUN) to D.M. Crutchfield (NRC) dated November 12, 1981 with a report, "SEP Topic III-5. A High Energy Piping Systems Inside Containment, Effects of Postulated Pipe Breaks, Oyster Creek Nuclear Generating Station, Unit No.1", dated July 30, 1979.
4.
Letter from Nuclear Regulatory Commission, SEP Branch Chief to KMC, Inc.,
dated July 20, 1978.
5.
Amendment 68 (Supplement 6, Addenda 5) to the Oyster Creek FDSAR, Application for a Full Term License, dated March 6,1972.
6.
M.E. Nitzel, " Summary of the Oyster Creek Unit 1 Piping Calculations Performed for the Systematic Evaluation Program" EGG-ZA-5211, EG&G Idaho, dated July 1980.
Performed for SEP Topic III-6, " Seismic Design Consideration.
7.
Philip Thullen, " LOADS ON SPHERICAL SHELLS," Oak Brook Engineering Department, Chicago Bridge & Iron Company, dated August 1964.
8.
Letter from the Nuclear Regulatory Commission (G. Lear) to Jersey Central Power & Light Company (I. Finfrock), dated February 12, 1975.
9.
Revision 1 to Supplement 6 ( Addenda 5) to Amendment 68, dated March 6,1975.
10.
Letter from the Nuclear Regulatory Commission (D. Crutchfield) to Jersey Central Power & Light Company (I. Frinfrock), dated September 25, 1980.
"0yster Creek LOCA Analyses Using the ENC NJP-BWR ECCS Evaluation Model",
11.
XN-NF-77-55, Revision 1, March 1978.
"The Exxon Nuclear Company WREM-Based NJP-BWR ECCS Evaluation Model and 12.
Application to the Oyster Creek Plant", XN-75-55, Revision 2, and Supple-ments 1 and 2, August, September and December 1976.
13.
Letter from D.M. Crutchfield (NRC) to I.R. Finfrock, Jr. (JCP&L), dated June 10,1981, (Draft SEE on Oyster Creek - SEP Topic III-5. A, Effects of Pipe Break on Structures, Systems and Components Inside Containment).
- 14. Letter from W.R. Schmidt (MPR for GPUN) to R. Fell (NRC) dated May 12, 1982, with a report "SEP Topic III-5.A High Energy Piping Systems Inside Containment" dated March 16, 1982.
-2
- 15. NE00-24195, " General Electric Reload Fuel Application for Oyster Creek" dated August 1979.
16.
Letter from D.L. Ziemann (NRC) to I.R. Finfrock (JCP&L), dated Janury 4,1980.
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