ML20058D114
| ML20058D114 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/31/1982 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20058D094 | List: |
| References | |
| 2575Q:1, TAC-48518, NUDOCS 8207270028 | |
| Download: ML20058D114 (29) | |
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l Troj an Nuclear Plant Robert A. Clark Docket 50-344-July 23, 1982 J
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License NPF-1 Attachment A 1.
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i TROJAN LOOSE TIERMAL SLEEVE i
i SAFETY EVALUATION i
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i JULY 1982 i
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PDR ADOCK 05000344 P
PDR 2575Q:1 1
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.c LOOSE TERMAL SLEEVE SAFETY EVALUATION 1.0 Sunmary 2.0 Introduction 2.1 Purpose - Safety Evaluation 2.2 Themal Sleeve Inventory 2.3 A ssunptions
' 3.0 Nozzle Integrity 3.1 Introduction 3.2 Stress Analysis 3.3 Conclusions 4.0 Mechanical Effects of Loose Objects 4.1 Reactor Coolant Pipe 4.2 Steam Generator 4.3 Reactor Internals 4.4 Reactor Vessel 4.5 Fuel 4.6 Reactor Coolant Punp 4.7 Pressurizer 4.8 Other RCS Components 4.9 Auxiliary Systems 4.10 Materials 5.0 Flow Blockage Effects of Locse Objects 5.1 Nomal Operating 5.2 Local Core Flow Distribution 5.3 Non LOCA Transients 5.4 LOCA Evaluation 6.0 Conclusions 2575Q:1 2
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1.0
SUMMARY
e Extensive evaluations were performed to determine the effects of loose reactor coolant pipe thermal sleeves at the Portland General Electric Company Trojan plant. PGE has found and removed four loose sleeves from the reactor lower internals. These sleeves were determined to have core from the 10" accumulator SI nozzle connections to the cold leg. PGE has also removed the thermal sleeve from the 14" pressurizer surge line con-nection to the hot leg. The only thermal sleeves of the subject design remaining in Trojan are those in the two 3" charging lines in the cold leg. These evaluations assumed these two thermal sleeves become loose and are transported in the reactor coolant system as si,ngle units or fragments. This evaluation has concluded that reasonable assurance exists that safe plant operation is not compromised. Operation of the plant without thermal sleeves is also acceptable.
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2.0 INTRODUCTION
2.1 PURPOSE Westinghouse was recently infonned by Portland General Electric Compar1y that an underwater television inspection had revealed a loose metal piece under the reactor internal lower core plate. Subsequent investi-gations by Westinghouse and the utility resulted in the discovery of additional loose parts in the reactor vessel and an eventual conclusion that the sources of the parts were the thermal sleeves from the 10 inch RHR/ SIS line nozzles. That conclusion has been verified by radiograph.fc examination of all four such nozzles on the affected unit. The sleeves traveled through the cold leg into the reactor vessel where all missing parts have been accounted for and recovered. Radiographic examination of other similarly designed sleeves on the affected unit have revealed one broken weld and a very slight movement of the 14 inch surge line nozzle sleeve as well as an indication of a possible crack of a thermal sleeve weld in one of the two 3 inch chargirg lines. PGE has removed the 14" surge line nozzle sleeve and desires to resume operation with the 3 inch charging line nozzle sleeves in place.
As a result of the di scovery of f ailed thermal sleeves, a safety evalua-tion was performed on the effects of loose and missing reactor coolant pipe nozzle thermal sleeves. This report sumnarizes and documents that safety evaluation.
I 2.2 THERMAL SLEEVE INVENTORY t
Thermal sleeves are utilized in several locations in the Trojan plant Reactor Coolant System (RCS) to reduce thermal stresses on RCS pipe noz-zies. The locations, sizes, and number of the thermal sleeves remaining l
in the Trojan RCS are as follows:
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Location Nisber Length I.D.
0.D.
Thickness Weight 3" Charging Injection 2 '.,
6.56" 2.12" 2.5" 3/16" 2.75 lbs Lines (Cold Leg)
The pressurizer surge line ar'id spry line also have themal sleeves at the pressurizer connection, h(wver, these are of a different design as discussed in Section 4.7 and dre not considered in this evaluation.
The material of construction of the themal sleeves is SA376 stainless steel, type 316.
2.3 ASSUMPTIONS To complete the safety evaluation for Trojan, certain 'assunptions wre m ade. These assunptions are based on knc,wn f acts, operating infoma-tion, engineering judgement, and recommended actions for continued oper-ation. The assunptions are as follows:
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Both themal sleeves from the 3 inch chargir%,.ines are assuned to l
come loose and are transported through the RCh 2.
The sleeves are assuned to remain intact or split into quarter sec-tions, whichever case provides the most conservative, evaluation.
The sleeves are attached by two welds at 180* in line'with the loop flow on the upstream end. Field exaninations indicate cracking can occur at the welds allowing an intact sleeve to come loose. Another f ailure mode dich has been observed is cracking of the sleeve along its length, beginning at one of the notches along the upstream end of the sleeve. Both of these f ailur e modes produce large objects.
l The ductile nature of the sleeve material also makes it unlikely that small pieces would be generated by impacts within the reactor coolant system. This evaluation specifically considered whole and quarter sections of the 3 inch sleeves. Smaller fragments wre also addressed in the nuclear fuel evaluation.
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3.
The plant operators are aware of the potential for loose parts and will monitor plant operations and pertinent equipment characteris-tics, specifically the loose parts monitoring system.
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3.0 N0ZZLE INTEGRITY
3.1 INTRODUCTION
This section summarizes the stress evaluation of the 3" charging noz-zles, and the 10" accumulator nozzles on the main reactor coolant loop Piping, to insure the structural integrity of the nozzles asstming f ailure of the thermal sleeves. The stress calculation of the 14" pres-surizer surge nozzle on the main reactor coolant loop piping with the themal sleeve removed and the original wid area ground smooth is also sisnmarized.
The analysis included an evaluation of the subject nozzles without a thennal sleeve and a " bounding" evaluation of the nozzle at the location of the f ailed sleeve / nozzle attachment wld. The bounding evaluation is not required for the surge line since the mld area is ground smooth.
This evaluation Witch considered all design transients and mechanical loads specified in the piping design specification demonstrates the structural integrity of the subject nozzles without themal sleeves.
Due to the similarities in the geometry of all three nozzles, (see Figure 3.1), and the similarities in the thennal sleeve designs (see Figure 3.1) the see analytical techniques to be applied to all three nozzle s.
The evaluation was separatcd into the following three basic regions on the nozzle, (see Figure 3.1),1) the location of the nozzle to pipe field wid 2) the location of the original sleeve to nozzle wall tack wld and 3) the remaining body of the nozzle including the crotch region.
l 3.2 STRESS ANALYSIS The stress analysis perfomed on the subject nozzles can be stamarized as follows. The detailed geometry and material of the nozzle, without a themal sleeve, was obtained from the appropriate specifications.
(For example, the previously mentioned figure and the plant specific drawings and equipment specifications). Then a detailed 2-dimensional finite 2575Q:1 7
element model was developed for the nozzle and appropriate representa-tive portions of the large header pipe and attached branch pipe (Figures 3.2 and 3.3).
Using piping design specifications containing operating transient des-criptions developed on the basis of the systems design criteria, the temperature transients, fluid velocities, neber of occurrences, etc.
wre sunmarized for all applicable transients, and appropriate loading conditions wre developed for the heat transfer analysis using the finite element model. The analysis included a time-history themal loading for a sufficient duration of time to insure the maximuni stress intensities wre calculated for all locations.
Using the same finite element model, stress intensities wre calculated from the pipe wall temperature distribution obtained from the heat transfer analysis for all critical locations. The actual f atigue evalu-ation of the component incorporates the methods and guidelines specified in the ANSI B31.7 code, dich are the same as those specified in the ASE Code Section III, Subsection NB, and all applicable Appendices.
This rigorous treatment has been applied to the 3" charging nozzle and the 14" surge nozzle without themal sleeves. Due to design modifica-tions, the 90*-10" accunulator nozzle was changed to a 45' inclined injection nozzle without a themal sleeve for later plants. A complete set of themal transient stress analysis was perfomed for this inclined injection nozzle for the same loading conditions as specified for the 90* injection nozzle.
In addition, analysis was also perfomed on a geometrically similar nozzle (6-inch) without a thennal sleeve with similar design transients. The results of these two analysis wre used in the qualification of the 10-inch as.::unulator injection nozzle without a thennal sleeve.
In the analysis of the nozzle without thennal sleeves, two locations wre found dere maximun peak stress intensity and f atigue usage occurred,1) the thick part of the nozzle near the crotch region and 2) the nozzle to the branch pipe field wld. This second region was found 2575Q:1 8
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to be critical af ter stress intensification factors were applied to the weld location, as specifted is the ANSI E31.7 Code. Assuntog the as-welded conditions, a stress concastration tactor of 1.7 was applied on top of the calculated values. At the crotch region, a factor of only 1.1. was applied, due to the ground flush condition of the weld at that j
location.
The four (4) weld buttaas on the 10 of the nozzle were investigated to determine if they represented a significant stress correntration. From this evaluation it was concluded that the weld buttons do not present a significant stress concentration.
To complete the fatigue calculation, the external loadings on the nor-zie, as. calculated for the Trojan plast, were incorporated and a usage factor calculated for each nozzle.
This completed the evaluation of the pressuriar suqpe nozzle with the thermal sleeve removed.
Finally, an evaluation of the failed tack weld region on the nozzle was perfomed. Because of the close proximity of the tack weld location to the pipe / nozzle butt weld {1.0-1.5 inches), the evaluation of the end location could be shown to yield the same usage factor, once the follow-ing was considered. First, an appropriate stress intensification factor h
was required to simulate the inside surface of the noltle at this loca-tion. Factors of 1.4 for K. and 1.5 for K2 were e nservatively 3
l used. This was based upon the relative severity of the conditions which
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resulted in the factors (K =1.7 and K = 1.8,) for an as-welded butt 3
2 weld, (i.e., affected inside surface, thin-welled pipe, misalignment of
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the butted pipe wall.) amt the condition actually present at the tack I
weld location (affected inside surface, thick wall pipe, perfect align-ment). This differetx:e in stress intensification factors more than off-o f
set the small increase in stress intensity due to the location being
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1 The calculated causlative usage factors indicate that all critical loca-tions meet the ASIE Code girements. Therefore, it is concluded that the nozzles are glifled to withstand all applicable design transients and will maintain their structural integrity without thermal sleeves for the plant desiga life.
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4.0 ECHANICAL EFFECTS OF LOOSE OBJECTS 4.1 REACTOR COOLANT PIPE The effect of the loose thermal sleeves on the primary system piping, either through impact or erosion, is expected to be negligible due to the limited impact energy created by the low radial flow velocities in the piping. The ductile material of the piping and the thermal sleeve would also preclude any sharp impact marks on the piping, thus elimina-ting any concern regarding possible stress concentration points.
The locations of the RTD bypass scoops, pressurizer spray scoop, and thermohells in the piping are all upstream of the 3" thermal sleeve locations, thus precluding damage by failure of a sleeve during operation. Reverse flow in the loop piping during plant startup could cause impacts on the scoops or thermome11s, houver, any damage causing pressure boundary f ailure or loss of an instrument would be detectable by the operator prior to criticality.
4.2 STEAM GENERATOR Since the thermal sleeve located in the pressurizer surge If ne connec-tion has been removed by PGE, the steam generator need not be considered in this evaluation.
4.3 REACTOR INTERNALS The reactor internals were evaluated to determine the effects of impact and hedging loads caused by the presence of loose thermal sleeves in the reactor coolant erstem.
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4.3.1 Upper Internals l
l The 3 inch charging injection line thermal sleeves will be confined bet-l ween the lower core plate in the reactor vessel, and the steam generator cold leg plenum. As such these thermal sleeves will have no impact con-sideration on reactor upper internals.
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Objects in the bottom of the reactor vessel muld not be expected to reach the upper internals due to the filtering action of the fuel assem-blie s.
The close spacing of the rods, the configuration of the grids and the flow deflectors, and the configuration of the nozzles should prevent large particles and most other particles from reaching the @per internals.
Small particles Witch r'suld pass through the fuel assemblies are likely to pass through the @per internals or to be forced clear during operation of the drive line.
In order for a foreign object to cause interference, it would have to be preferentially oriented in a moving clearance area.
As part of the nomal start @ tests, control rod drop times are recorded and evaluated to confim proper driveline perfomance.
In the unlikely event that a foreign object wuld become lodged in the @per package during operation'and cause a driveline to become inoperable, the exis-ting FSAR analyses ass:suption of one stuck control rod assembly would not be exceeded.
4.3.2 Lowr Internals The reactor vessel and lowr internals wre analyzed for structural integrity with themal sleeves from the 3 inch charging line within the reactor vessel.
4.3.2.1 Impact Load Evaluation Core Barre,1 The evaluation of the impact loads on the core barrel due to the 3 inch line sleeves is based on a similar evaluation using a 10" line themal sleeve. Since the size and weight of the 10" line sleeve is greater than that of a full 3" line sleeve, this evaluation bounds the analysis of the impacts from the 3" sleeve.
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It was assmed that the 14 inch side of a quarter section of a 10" sleeve strikes the core barrel at the inlet nozzle velocity.
Since the part is thin it will deform before the core barrel. Therefore, the load applied to the core barrel is detennined by the load capacity of the piece.
Area of piece = 14 in x.145 in 5 2.00 in 2 Assuming an ultimate strength of 63.5 ksi for the piece, the :naxima load applied to the core barrel is 127 kips, Assming the core barrel responds as a cantilever bean, the impact stresses in the core barrel are calculated to be negligible.
' max = 293 psi and Tmax = 123 psi Assune that the f ace of a complete sleeve strikes the core barrel at the inlet nozzle velocity.
Area of piece = 4.166 in2 Maximun load applied to core barrel is Pmax = 265 kips
' max = 611 psi
= 257 psi T max Due to the low magnitude of the impact stresses and the short time dura-tion of impact loads, the core barrel is unaffected by impacting loose parts.
The method used for the minimun transmission energy required to per-forate a target plate per WCAP 9934 results in a maximum depth of dent equal to.0234 in.
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4.3.2.2 Irradiation Specimen Guides The irradiation specimen guides are bolted to the outside of the neutron shield panels. Each guide consists of an upper and a lowr portion.
The upper portion is attached to the upper neutron panel by (six) 3/4 inch cap screws and (two) 7/8 inch diameter dowl pins. The lowr por-tion of the irradiation specimen guide is attached to the lowr neutron panel by (eight) 3/4 inch cap screws and (two) 7/8 inch dowel pins.
Since the upper portion of the specimen guide is supported by fear bolts it was chosen for the impact load evaluation.
The assumptions pertaining to the impact load calculation are:
1.
Rigid target and elastic missile 2.
Loose part moves at fluid velocity 3.
Maximum dynamic load f actor = 2.0 The contact area is calculated assiming that the f ace of the themal sleeve impacts the top of the specimen guide. The width of the top of the specimen guide is 7.18 inches.
The impact load, Q, is then calculated for a complete 3" themal sleeve asstning a dynanic load f actor of 2.0.
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= 2AP = 66,850 lb c
The load required to overcome the friction due to preload and move the specimen guide is calculated to be 15,577 lbs. per bolt..
1 Since the impact load is greater than the friction load, the specimen guide will move and the load will be carried by the (six) 3/4 inch bolts and dowl pins.
It has been demonstrated that the capacity of the six bolts and tw dowl pins is adequate to react the resultant load.
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4.3.2.3 Bottom Mounted Instrumentation Tubes The instrumentation tubes in the bottom head of the vessel were evalua-ted for impacting of thermal sleeves or thermal sleeve sections which s4y be loose in the system. The cases evaluated were for an impact at the tube / bottom head intersection (shear strike) and for an impact at the highest point on the instrument tube which could be struck hdthout first striking the internals. Resulting values were compared to.appro-priate shear allowable and collapse loads.
The shear strike was evaluated for the complete 3 inch thermal sleeve impacting the instrument tubes. The maximum shear stress was found to be significantly below the allowable stress.
The loads on the instrument tubes resulting from the impacting of the complete 3 inch thermal sleeve were evaluated as exceeding the instru-mentation tube collapse load when the sleeves are assumed to strike the tubes at full downcomer velocity as they enter the lower RY plenum. The result indicates that plastic defonnation of an instrumentation tube could result if the tube were struck in an unf avorable manner by the loose thermal sleeves as they enter the lower plenum.
However, deforma-tion of the tubes does not constitute a safety concern.
Due to the ductility of the Ni-Cr-Fe alloy tube, deformation could occur, but the tubes will not rupture and hd11 continue to protect the thimble guide tubes. The guide hd 11 therefore not rupture and the reactor coolant pressure boundary will not be violated.
I Therefore, the loose thermal sleeves striking the instrumentation tubes in the bottom head of the vessel does not constitute a safety hazard.
In the unlikely event that the f ailure of a bottom mounted instrumenta-tion tube leads to leakage, the double ended break of this BMI tube 2
results in a leak area of 0.00024 Ft. Assuming a di scharge coeffi-cient of 1.0 and the Zaloudek subcooled critical flow model which over-predicts leak flow, one charging pump in the normal charging mode can provide makete for at least 3 broken tubes.
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This wuld be classified as a leak, not a LOCA, and RCS pressure muld be maintained at 2250 psia.
If both charging peps are available, addi-tional tubes can be tolerated.
Small break LOCA analysis with minimm safeguards SI have demonstrated that full instraent line breaks in at least 5 instraent tubes may result in depressurization and automatic SI initiation.
Ho wver, this small break LOCA will maintain forced or natural circulation, and the i
RCS wf 11 reach equilibrim conditions with no core uncovery.
4.4 REACTOR VESSEL During plant heatte, the gap betwen the reactor vessel bottom head inside surf ace and the bottom of the secondary core support structure will decrease. Similarly, the gap in the radial support keyways will decrease during heatup. A foreign object present in this area could impose mechanical loadings on the vessel bottom head. Due to the size of the gap, it was determined that a full 3" sleeve or curved quarter section of a 3" sleeve could not enter the gap. A flattened quarter section could enter the gap, houver its thickness was too small to cause interference in the gap. For the sake of completeness in the evaluation, udging. loads wre considered by determining the stiffness of a udged cylinder (assmed intact thennal sleeve) and computing a spring rate. The calculated spring rate and resulting loads shows that loadings from such objects wuld be acceptable.
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4.5 NUCLEAR FUEL Foreign objects in the primary system have tw potential effects on the nuclear f uel:
- 1) partial flow blockage of fuel assemblies due to an object becoming wdged in the fuel assembly flow paths, and 2) clad war due to pieces becoming lodged in the assembly or betwen tw assem-blie s.
Flow blockage effects are discussed in Section 5 of this report.
From a fuel mechanical design viegoint, loose pieces should not pose an operational problem when the fuel assemblies are seated properly on the core plate. The loose pieces should be stopped by the bottom nozzle or the lowr core plate due to dimensional considerations. Although highly unlikely, it is possible for a very small piece to udge betwen fuel assemblies and cause fretting and/or grid damage. This is highly improbable due to the f act that space betwen fuel assemblies is approx-imately 40 mils, i.e. approximately one third the thickness of the ther-mal sleeve material. Should a fretting mechanism cause clad f ailure on a fuel rod it is unlikely that any radiation release muld approach the technical specification limit, and as such no safety concern wuld exist.
Due to the relatively large fragments expected from the thermal sleeves, the transport of loose pieces into and through the fuel assemblies is not considered possible.
4.6 REACTOR COOLANT PUMP There are no thermal sleeves located in piping connections betwen the Reactor Coolant Ptsnp (RCP) and the Stean Generator. A loose themal sleeve can enter the RCP only den a reverse flow condition occurs, in dich case the plant is not operating.
If this occurs a themal sleeve or portion of one will not affect the pressure boundary integrity due to the geometry, mass and impact energy of the pieces.
An intact 3 inch themal sleeve can pass through the pump internals without significant deformation. During RCP startup, forward flow wuld eject the sleeve.
2575Q: 1 16
If thermal sleeve fragments did lodge between the impeller and diffuser in such a wAy as to cause interference, the material is expected to be pinched or sheared between the impeller and diffuser vanes due to the very high torque of the RCP motor. A consequence may be an increase in shaf t vibration with continted RCP operation. No locked rotor or pres-sure boundary violation is expected to occur. The increased vibration may be observed by the operator for corrective action.
A similar safety evaluation of larger material (1 1/16 inch thick, 304 SS) that was postulated to enter the RCP in various size fragments was previously performed, and it also concluded that there was no safety Concern.
In a summary, the loose thermal sleeves are not considered a safety con-cern for RCP integrity and operation.
4.7 Pressurizer The thermal sleeves in the 4 inch spray line and the 14 inch surge line connections in the pressurizer proper are attached in a different manner than the reactor coolant piping nozzle thermal sleeves.
On the pressur-izer thermal sleeves the upstream end of each sleeve is helded over an arc of 45 degrees.
The sleeves themselves are of larger diameter than the nozzle safe ends, thus preventing sleeve movement away from the p re ssurizer. The flow distribution screen inside the pressurizer at the surge line connection prevents that sleeve from entering the pressur-izer.
Similarly, the spray header traps the sleeve on the spray line connection.
Since the pressurizer thermal sleeve design has not significantly changed and operating experience has not indicated any f ailure these sleeves are not considered in this safety evaluation.
Due to their method of attachment, it is also very unlikely that these sleeves would become loose within the reactor coolant system.
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Based on the most probable movement of ary dislodgad themal sleeves from the 3 inch charging lines it is extremely unlikely that av piece muld cause mechanical daage or become lodged in the pressurizer inlet piping or the pressurizer.
s 4.8 DTER REACTOR COOLANT SYSTEM COMP 0ENTS Due to the physical separation from the remainder of the reactor coolan vstem of such components as cfontrol rod drive mechantas and safety, relief and block valves, no adverse effect is expected to result from
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loose themal sleeves in the reactor coolant system.
4.9 AUXILIARY SYSTEMS The possibility of the potentially loose 3 in charging line themal sleeves affecting the operation of other systems connected to RCS was also investigated in this safety evaluation.
The normal charging line enters loop A upstream of the 10 inch SI an The 1/2 inch centrifugal charging pap / boron injection tank lines.
themal sleeve could conceivably' enter into the SI line during nomal Howver, the SI connection is above the horizontal and has stag-flow.
nant flow characteristics. It is thus unlikely that a part entering Approximately 5 feet upstrem in the line the SI line would travel f ar.
The part should not migrate is a 10 inch, nomally closed check valve.
f ar enough into the line to make contact with the check valve due to the If it wre to remain in the line, it stagnant conditions in the If ne.
muld be flushed into the RC pipe on SI initiation.
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The consequences of the alternate charging line thermal sleeve (or parts) in loop D migrating would be the same as the normal charging
- sleeve, Upstream from the charging and alternal charging lines, all loops have a resistance temperature detector mounted in a wil and a single RTD cold leg connections. Neither of these should be affected, due to the con-struction and location of these components.
In the above discussions, it should be noted that the connections (with the exception of the RTD HL scoops) are in the upper half of the RC pip-ing and the loose sleeve would be required to move against gravity to contact them.
It is conceivable that a failed sleeve part could enter into its own nozzle, hohever, flow in these branch lines is toward the reactor cool-ant pipe and thus it is unlikely that the sleeve would travel up the line.
4.10 MATERIALS No unacceptable material would be introduced into the reactor erstems as a result of the f ailure of a thermal sleeve. Minor clad damage could occur on the surf aces of carbon steel components, houver, this muld present no safety or operational concern due to the very slow corosion rate of the carbon steel in the reactor coolant environment.
2575Q:1 19
5.0 FLOW BLOCKAGE
5.1 INTRODUCTION
In postulating the presence of loose themal sleeves in the reactor coolant system, an evaluation was made of the effect of the sleeves or parts of the sleeves blocking flow in the core. The evaluation con-sidered that the 3 in line themal sleeves come loose in the reactor coolant system, break into quarter segments, travel down the cold leg, lodge in the lowr internals and block flow at the lowr core plate.
The evaluations considered the effect of blockage on reactor coolant system total flow, local flow distributions in the core during nomal operation, and the effect on LOCA and non-LQCA accident analyses.
5.1 REACTOR COOLANT SYSTEM TOTAL FLOW For the analysis of reactor coolant system flow reduction, the loose 3" themal sleeve segments in the reactor vessel wre modeled as flat plates nomal to the flow, resulting in increased pressure loss coeffi-cients across the lowr core plate.
- The results of this conservative analysis showd that the total reduc-tion in RCS flow was less than 0.05 percent. This still results in the RCS flow being greater than themal design flow, which is a conserva-tively low value of flow rate upon dich the core themal-hydraulic design is based. Thus, this flow reduction will have no effect on the themal-hydraulic design and DNB margin in nomal operation at rated powr. Based on the above evaluations it was concluded that the reduc-tion in RCS flow wuld not affect design margins in nomal operation.
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5.2 LOCAL CORE FLOW DISTRIBUTION The effect on core flow distribution of loose themal sleeve segments l
located underneath the lowr core plate was also evaluated.
It was detemined that the segments from the sleeves remaining below the lowr 2575Q:1 20
core plate would result in greater core blockage than the smaller seg-ments reaching the fuel nozzles, since the smaller pieces could only reach the fuel nozzles in a lengthwise orientation.
In performing this evaluation, it was assuned that the sleeve segments remain curved, and thus do not completely block flow, but do cause restrictions in the flow to the core.
The infomation available on themal effects due to flow blockage indi-cates that there will be no significant increase in the likelihood of DNB at normal operating conditions. WCAP-7956 shows results from a blocked assembly flow recovery test and WCAP-8054 shows that a 10 per-cent finw reduction in the hot assembly and its 8 surrounding assemblies reduces DNBR by only 0.3 pen:ent.
Since the thennal sleeve pieces will remain curved, there will always be some flow through all of the lowr core plate holes. This, along with the f act that the total core thenna1 design flow will remain unchanged, will insure that the DNBR will not be significantly reduced.
Thus, the effect of blockage on local core flow distribution and DNBR is judged to be insignificant.
5.3 NON-LOCA TRANSIENT ANALYSIS Flow blockage by loose thennal sleeves in the reactor coolant system potentially affects non-LOCA transients only in that there is a slight reduction in total RCS flow, as discussed previously in Section 5.1.
An evaluation was perfonned on the effect of the RCS flow reduction on the non-LOCA transients.
In non-LOCA transient analysis, it is conser-vatively assuned that accidents are initiated with the reactor coolant system operating at thennal design flow (TDF). A reduction of less than 0.05 percent due to the thermal sleeve flow blockage effect on RCS flow still results in a measured flow greater than TDF. This assures that all the current safety analyses remain valid.
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5.4 LOCA EVALUATION The postulated presence of loose thennal sleeve segments in the RCS was also evaluated for its effect on the 10CFR50.46 Appendix K limiting case ECCS analysis. For Trojan, the limiting case break is a double ended cold leg guillotine break with a discharge coefficient equal to 0.6.
A simnary of the evaluation is given below:
A.
Overall system thennal perfonnance at 100 percent powr has been shown to be insignificantly changed by the presence of the large sleeve pieces. The redtetion in RCS flow of less than 0.05 pen:ent can be accommodated with little effect on the Appendix K LOCA analy-si s.
B.
Considering the presence of thennal sleeve segments lodged against the bottom of the lowr core plate, it is asstaned that the thennal sleeves will remain curved. Thus, there will always be some flow through all of the lowr core plate holes, such that no assembly will be starved of flow.
It has been detennined that flow redistri-bution above postulated sleeve locations will occur in the first several inches of the f uel during nonnal operation, and that t'here-fore reduced minimtsn DNBR is not of concern in the hot assembly.
In a LOCA analysis, post-LOCA thennal-hydraulics predicted for the hot assembly directly define the calculated peak clad temperature (PCT).
Core flow post-LOCA is characterized by positive (normal direction) and negative core flow periods, in that order. From the above, during positive core flo~ when RCP perfonnance detennines w
flow magnitude and direction as during normal operation, thennal-hydraulics should be equivalent to those computed in the existing LOCA analysis. When the flow reverses, any pieces impinged against the core plate should f all off into the lowr plentsn and tnus not be in a position to affect the calculated hot assembly flow.
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C.
It has been asstmed in most areas of this safety evaluation that the thennal sleeves break up into quarter segments, however the presence of smaller pieces of the sleeve being lodged within the fuel and causing additional blockage in the hot assembly during core ref food following a LOCA was also addressed.
For Trojan, the calculated limiting peak clad temperature (PCT) from a large break LOCA is 1970*F at the burst node on the fuel rod.
Small bits of material creating blockage in the core will have no measurable impact on this PCT result, dich occurs at the 6.0 ft elevation axially.
The limiting non-burst node clad temperature occurs above the core mid-plane during the FLECHT heat transfer period of core reflooding, dich is characterized by core inlet flooding rates greater than 1 inch /sec.
During this period, flow blockage would not be a penalty because the increased turbulence of the reflood flow in the area of the blockage muld increase heat transfer rates and oven:ome any reduction in heat transfer due to a flow reducte,n around the blockage. Therefore, Appen-dix K does not require a blockage penalty during FLECHT. Overall, no increase in the peak clad temperature (PCT) at the burst node limiting elevation or elsedere in the core during FLECHT heat transfer is attri-butable to pieces of material being present in the core.
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6.0 CONCLUSION
Based on the evaluation provided in this report, it is concluded that reasonable assurance exists that the safety of the Trojan plant is not significantly affected by continued operation in its current configuration.
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Trojan Nuclear Plant Robtrt A. Clark Docket 50-344 July 22, 1982 License NPF-1 Attachment B REACTOR COOLANT SYSTEM N0ZZLES ULTRASONIC TESTING RESULTS An examination was conducted to detect any cracking that may have occurred in RCS nozzles as a result of operating with missing or degraded thermal sleeves. The following nozzles were examined by the Ultrasonic Testing (UT) method:
Two 3-in. charging injection lines Four 10-in. SI/ accumulator injection lines This examination of the nozzle ID surface was conducted from the minor and major nozzle diameters. The nozzles were scanned using 45* and 35' transducers in the circumferential and longitudinal directions.
No indications of cracking were observed on any of the nozzle ID surfaces.
A similar examination of'the 14-in. pressurizer surge line nozzle will be performed in conjunction with the ASME Section XI UT baseline examination of the two surge line repair welds.
If unacceptable indications are found in the 14-in. nozzle, the results will be reported in accordance with the Technical Specifications.
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