ML20027C813
| ML20027C813 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/30/1982 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20027C811 | List: |
| References | |
| 2863Q:1, NUDOCS 8210270260 | |
| Download: ML20027C813 (44) | |
Text
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o ATTACHMENT III f
I
.i NORTH ANNA UNIT 1 LOOSE THERMAL SLEEVE SAFETY EVALUATION SEPTEMBER 1982 f'
4 WESTINGHOUSE ELECTRIC CORPORATION
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j 8210270260 821012 i
PDR ADOCK 05000338 P
PDR 2863Q:1
LOOSE THERMAL SLEEVE SAFETY EVALUATION 1.0 Summary 2.0 Introduction 2.1 Purpose 2.2 Hi story 2.3 Thermal Sleeve Inventory 2.4 Assumptions 3.0 Nozzie Integrity 3.1 Introduction 3.2 Stress Analysis 3.3 Conclusions 4.0 Mechanical Effects of Loose Obj ects 4.1 Reactor Coolant Pipe 4.2 Steam Generator 4.3 Reactor Internals 4.4 Reactor Yessel 4.5 Fuel 4.6 Reactor Coolant Pump 4.7 Pressurizer 4.8 Primary Loop Stop Valves 4.9 Other RCS Components 4.10 Auxiliary Systems 4.11 Materials 5.0 Flow Blockage Effects of Loose Objects 5.1 Normal Operating 5.2 Local Core Flow Distribution 5.3 Non LOCA Transients l
5.4 LOCA Evaluation 2863Q:1
1.0 SUMARY Extensive evaluations were perfomed to detemine the effects of loose reactor coolant pipe themal sleeves at the Virginia Electric and Power Company (Vepco) North Anna Unit 1 plant. These evaluations assumed the themal sleeves of a particular design become loose and are trans-ported in the reactor coolant system as single units or pieces.
This evaluation concludes that reasonable assurance exists that safe plant operation is not compromised.
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2.0 INTRODUCTION
2.1 PURPOSE As a result of the discovery of the migration of certain themal sleeves from the nozzles, a safety evaluation was perfomed on the effects of loose and missing reactor coolant pipe nozzle thermal sleeves. This report summarizes and documents that safety evaluation.
2.2 HISTORY Westinghouse was recently infomed by one of its operating plant cus-tomers that an underwater television inspection had revealed a loose metal piece under the reactor internals lower core plate. Subsequent investigations by Westinghouse and the utility resulted in the discovery of additional loose parts in the reactor vessel and an eventual conclu-sion that the sources of the parts were the themal sleeves from the 10 inch RHR/ SIS line nozzles. That conclusion has been verified by radio-graphic examination of all four such nozzles on the affected unit. The sleeves traveled through the cold leg into the reactor vessel. All l
migration of the sleeves have been accounted for and recovered.
Radio-graphic examination of other similarly designed sleeves on the unit 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 charging lines.
A similar migration of one 10 inch nozzle thermal sleeve of the same design was also discovered at another operating plant.
The sleeve has been located to be in the reactor vessel lower plenum. The welds of the remaining sleeves of the subject design at this unit were shown to be intact by radiographic inspection.
l On North Anna Unit 2, a radiographic examination by Vepco has indicated that four of the subject design themal sleeves appear to have cracked 2863Q: 1
welds at the attachment to the pipe, and that the remaining four, including the 14 inch pressurizer surge line thermal sleeve, have welds that are intact.
On North Anna Unit 1, radiographic examination nas indicated that the sleeve from the Loop A SI line appears to have migrated to the reactor vessel. Also, the 3" charging line sleeve appears to have cracked welds at the attachment to the pipe.
The remaining six sleeves, including the 14 inch pressurizer surge line thermal sleeve, have welds that are intact.
2.3 THERMAL SLEEVE INVENTORY Thermal sleeves are utilized at several locations in the North Anna Unit 1 plant Reactor Coolant System (RCS) to reduce thermal stresses on RCS pipe nozzles. Table 1 provides locations, sizes, and number of the reactor coolant pipe thermal sleeves of the design which have exhibited cracked welds.
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TABLE 1 NORTH ANNA 1 THER!%L SLEEVE STATUS N0ZZLE SH LOOP WELD CONDITION C0!9E NTS Surge Line 14" C
Intact Accunulator 12" A
Intact B
Intact C
Intact SI 6"
A Cracked Recovered from bottom B
Intact of vessel C
Intact Charging 3"
B Cracked To be removed L
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i The material of construction of the thermal sleeves is SA 376 stainless steel type 316 or SA 240 stainless steel, type 304.
Ttermal sleeves of a different design are also present at the surge line and spray line nozzles at the pressurizer and in the CVCS fill lines on the RCS crossover leg. These sleeves employ welds of 45*, 45* and 360*
respectively and have a counter bore gemnetry to minimize potential for movement of the sleeve. There has been no indication of cracks in the welds of these types of sleeves, and they are not considered in this safety evaluation.
2.4 ASSUMPTIONS To complete the safety evaluation for North Anna Unit 1 certain engi-neering judgements were made. These engineering judgements are based on facts gathered from the first operating plant to locate migratory ther-mal sleeves, and recommended actions for continued operation. The engi-neering judgements are as follows:
1.
All reactor coolant piping thermal sleeves of the subject design are presumed to migrate and are transported into the RCS system.
The sleeves are assumed to remain intact or split into quar-2.
ter sections, 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 experience indicates cracking can occur at the welds whereby an intact sleeve could come loose. Another failure mode which has been observed at another plant is cracking of the sleeve along its length, beginning at one of the notches along the upstream end of the sleeve.
Both of these modes produce large objects. The ductile nature of the sleeve material also makes it unlikely that small pieces would be generated upon impacting within the reactor coolant system.
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This evaluation specifically considered objects ranging in size from a complete 14 inch sleeve to one quarter sections of the 3 inch sleeves.
Smaller fragnents were also addressed in the nuclear fuel evaluation.
3.
Tne plant operators are aware of the potential for dislodged parts and will monitor plant operations and pertinent equip-ment characteristics.
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3.0 N0ZZLE INTEGRITY
3.1 INTRODUCTION
This section summarizes the stress evaluation of the 3" charging noz-zles, the 12" accumulator nozzles, the 6" SI nozzles, and the 14" pres-surizer surge nozzle on the main reactor coolant loop piping, performed to insure the structural integrity of the nozzles assuming certain migration of the thennal sleeves. The specific thennal sleeve migration discovered during inspection of the subject nozzles, and considered in this evaluation, included, (a) three inch charging and 6 inch safety inj ection thermal sleeve weld failure and rotation of the sleeve, and a 12 inch accumulator injection nozzle thennal sleeve weld failure with the sleeve becoming lodged in the nozzle between 1.0 and 6.0 inches i
below (downstream) its installed location.
I The analysis included an evaluation of the subject nozzles without a thermal sleeve and a " bounding" evaluation of the nozzle at the location of the sleeve / nozzle attachment weld. Even though the thennal sleeves have been removed on certain nozzles included in this evaluation, a
" bounding" analysis was still performed on all nozzles for conser-vati sm. This evaluation which considered those design transients and mechanical loads specified in the piping design specification demon-strates the structural integrity of the subj ect nozzles without thennal sl eeves.
Due to the similarities in the geometry of all subject nozzles, and the similarities in the thennal sleeve designs (see Figure 3.1) the same analytical techniques were applied to all nozzles. The evaluation was separated into the following three basic regions on the nozzle, (see Figure 3.1),1) the location of the nozzle to pipe field weld at the
" safe-end" of the nozzle, 2) the location of the original sleeve weld to l
nozzle wall and 3) the remaining body of the nozzle including the crotch region.
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3.2 STRESS ANALYSIS The stress analysis perfomed on the subject nozzles can be suianarized 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). A detailed 2-dimensional finite element model was developed for the nozzle and appropriate representative por-tions 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, number of occurrences, etc.
were summarized for those applicable transients, and appropriate loading conditions developed for the heat transfer analysis using tne finite element model.
The analysis included a time-history themal icading for a sufficient duration of time to pemit the maximum stress intensities l
to be calculated for all locations.
Using the same finite element model, stress intensities were calculated from the pipe wall temperature distribution obtained from the heat transfer analysis for all critical locations.
The actual fatigue evalu-ation of the component incorporates the methods and guidelines specified in the ASME ANSI B31.7 Nuclear Power Piping Code, USA Standard for Pres-sure Piping,1969 Edition, including the 1970 and 1971 Addenda.
This rigorous treatment has been applied to the 3" charging nozzle, the 6" safety injection nozzle, and the 14" surge line nozzle without themal sleeves.
Due to design modifications for later plants, the 90*-12" accumulator nozzle was changed to a 45* inclined injection nozzle without a themal sleeve. 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 l
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(6-inch) without a themal sleeve with similar design transients.
The results of these two analyses were used in the qualification of the 12-inch accumulator injection nozzle without a themal sleeve.
In the analysis of the nozzle without themal sleeves, two locations were found wnere maximum peak stress intensity and fatigue usage occurred,1) the thick part of the nozzle near the crotch region and 2) the nozzle to the branch pipe field weld. This second region was found to be critical after stress intensification factors were applied to the weld location, as specified in the ANSI Code. Assuming the as-welded conditions, a stress concentration factor 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 at the weld location.
To complete the fatigue calculation, the external loadings on the noz-zie, as calculated for the North Anna Unit 1 plant were incorporated, and a usage factor was calculated for each nozzle.
Finally, an evaluation of the fillet weld region on the nozzle was per-fomed. Becatse of the close proximity of the fillet weld location to the pipe / nozzle butt weld (1.0-1.5 inches), the evaluation of the safe-end location could be shown to yield the same usage factor, once the following was considered. An appropriate stress intensification factor was required to simulate the inside surface of the nozzle at this loca-tion.
Factors of 1.4 for K3 and 1.5 for K2 were conservatively used.
This was based upon the relative severity of the conditions which 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-walled pipe, misalignment of the butted pipe walls,) and the condition actually present at the fillet weld location (affected inside surface, thick wall pipe, perfect align-ment). This difference in stress intensification factors more than offset the small increase in stress intensity due to the location being closer to the thick part of the nozzle and resulted in no significant change in stress.
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3.3 CONCLUSION
S The cumulative usage factors calculated on the basis described in the previous sections and the external loadings based on North Anna Unit 1 specific as-built infonnation indicates that all critical locations meet the ANSI Code requirements.
Therefore, it is concluded that the nozzles are qualified to withstand all applicable design transients and will maintain their structural integrity without thennal sleeves for the plant design life.
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TER%L SlIEVE BifiTWELDN0ZZLE FATERIAL:
SA2 % OR SA312 TP3C4 OR TP306 1/8" WIEE SLOTS TYP. 6 PLAES
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4.0 MECHANICAL EFFECTS OF LOOSE OBJECTS 4.1 REACTOR COOLANT PIPE Tne effect of the loose themal 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 themal sleeve would also preclude any sharp impact marks on the piping, thus minimiz-ing any concern reganiing possible stress concentration points.
The locations of the RTD bypass scoops and themowells in the reactor coolant piping are upstream of the themal sleeves, except for the 14" surge line themal slceve which is upstream of the hot leg RTD scoops and T hot themowell. The potential of a loose 14" themal sleeve impacting these components during operation has been considered and is discussed below.
There are three RTO bypass scoops at 120* locations which protrude 6 to 8 inches into the flow stream.
Upon impact of the themal sleeve, the scoop is assumed to be sheared off or defonned sufficiently to make it ineffective.
The RCS pressure boundary will not be violated. An additional loose part could be generated, but it will be captured in the steam generator channel head along with the sleeve.
Damage to one or more scoops could affect the flow rate in the bypass line, thereby increasing the delay time in the temperature measurement. However, a significant change in the flow velocity would actuate 4 low flow alam I
to alert the plant operator.
Defonning or shearing the hot leg RTD scoops has two effects:
- 1) it will increase the RTD loop flow transport time, and 2) it could cause l
the RTD to indicate a different coolant temperature than the other loops due to radial temperature distribution in the RCS pipe.
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r-The maximum flow transport time under the worst postulated conditions would be 1.3 seconds. The safety analysis reported in the FSAR assumes a delay of 2 seconds for flow transport time and 2 seconds for the RTD sensor. Thus, the predicted flow transport time is bounded by the safety analysis.
Should the scoops be sheared off, the RTD's would be biased by less than 2*F.
The overtemperature aT trip requires a 2 out of 3 input. Thus a bias in one RTD will not cause a safety concern since two channels remain unaffected.
Damage to the RTD scoops can affect the perfomance of control systems which use this temperature measurement as input; e.g. rod control sys-tem.
The perfomance of these systems could be affected, however, no safety concern is created.
In summary, potential damage to the RTD scoops does not create a safety concern.
TheT themowell is located at the 270* centerline of the RCS hot piping and protrudes approximately 3 inches into the flow stream.
Although the profile of the themowell is small relative to the total flow area, if the 14" themal sleeve did impact the themowell it could result in the severance of the well at the RCS pipe wall. An additional loose part would be generated and a leak would result (approximately 15.7 lb/sec) through the connection.
The T measurement would hot become unavailable and the potential for a missile (Thot probe) and i
jet impingement would result.
l The RCS safety would not be jeopardized since this sim leak is detect-able and also is within the makeup capability of one charging pump.
This does not present an unanalyzed event nor does it challenge the plant safety systems.
4.2 STEAM GENERATOR l
4.2.1 Introduction This evaluation considers the potential effects of an 11.5" outside diameter thennal sleeve entering the primary side of the Series 51 steam 2863Q: 1
generators at North Anna Unit i from the connection of the pressurizer surge line. Potential components of the steam generator which may be impacted include the tube sheet, divider plate, channel head, tube-to-tubesheet welds, tube-sheet-to-divider plate weld, and the divider plate-to-channel head weld. The potential effect of a 34 pound sleeve impacting on these components is considered separately in the following sections.
4.2.2 Tube Sheet and Tubes The tubesheet of Series 51 steam generators is clad with Inconel 600 which is quite ductile.
Repeated impacting on the cladding by the 304 stainless steel thennal sleeve, which is also a ductile material, would not be expected to cause the cladding to crack or break loose. The design of the Series 51 steam generator incorporates tube ends which extend approximately 0.22 inches below the primary face of the tubesheet cl addi ng. This configuration exposes the tube ends to potential impacts from the presence of a loose thermal sleeve.
Evidence from a previous incident with a loose part of comparable mass in a Series 51 steam generator shows deformed tube ends. The deforma-tion was mainly bending of the tube ends rather than peening as has been J
observed resulting from smaller loose parts in the primary side of the steam generator. The data of comparable mass indicates tube end flow restriction due to this ductile deformation of the tube ends.
If the pressurizer surge line thermal sleeve were to enter the steam generator the damage would be expected to be similar to the bending and deforma-tion explained above. Any resulting flow restriction could have a minimal effect on the primary flow through the loop. A margin would still exist to the conservatively low thermal design flow.
If a large flow mismatch did occur, it would be detected by the plant operator who could then take appropriate actions. The ductility of the inconel mate-rial makes it unlikely that small pieces of the tube ends would break l
1 off under short-term exposure to impacts by the loose parts.
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Operation for short periods of time in the presence of a loose thermal sleeve would be expected only to produce bent and deformed tube ends not any tube end pieces.
Operation for longer time periods could generate tube end pieces which could potentially affect other components of the reactor coolant system. This condition is not assumed to occur since a loose parts monitoring system will detect this occurrence.
The thermal sleeve design contains notches at the upper ends for stress relief. These notches are 90* apart and experience indicates some cracking at these notches on migrated sleeves. Thus, if the sleeve were to become loose parts it is anticipated it may break at the notches, forming large sections.
No piece small enough to fit into the 0.775" inside diameter opening of the tube is expected to be formed from the break-up of the thermal sleeve.
Therefore, it is concluded that potential effect upon the tubesheet and tubes resulting from short-term impacts by a migrated thermal sleeve would not disturb the integrity of the steam generator components.
4.2.3 Tube-to-Tubesheet Weld Impacting of the tube-to-tubesheet (TTS) weld by a loose thermal sleeve is considered unlikely due to the presence of the tube-ends extending 0.22" beyond the primary face of the tubesheet, thus protecting the welds from absorbing a large number of direct impacts. Thus, disinte-gration of the weld from impacting is considered unlikely due to the ductility of the materials and the geometry of the weld.
One design feature of the Series 51 steam generator is the explosive expansion (WEXTEX) of the tube over the entire 21 inch depth of the tu bes heet.
This feature provides added strength to the tubes in the tubesheet hole and provides an additional margin against primary to secondary leakage.
If it is assumed that some of the welds do completely severe and primary to secondary leakage occurs, the amount of leakage would be low. Such 2863Q: 1
leakage would be detectable by nomal radiation monitoring, and the extent of the leakage could be monitored. This leakage would be expected to be within the allowable technical specification limits and would not present any safet/ concern. Monitoring of the leakage would be possible so that if an increase in leakage is detected the plant could be shut down in an orderly manner.
4.2.4 Divider Plate The 2 inch thick Inconel 600 divider plate is welded both to the channel head and the tubesheet to fom a barrier separating the hot leg and cold leg of the steam generator.
The rigidity of the plate is highest closest to the welds, and it becomes more flexible toward the middle of the plate.
Impacting of a themal sleeve could be expected to occur in the flexible region of the plate. The geometry of the channel head limits access to areas more proximate to the welds.
The flexibility of the plate in the most likely impact region along with the flexibility of the themal sleeve wiil cause the impact loadings to be sufficiently distributed so as not to be of concern to the integrity of the divider plate.
As mentioned previously, the ductility of the sleeve material reduces the likelihood that sharp edges would be created. Therefore, any marks that result from themal sleeve / divider plate impacts would most likely be round-bottom, rather than sharp-pointed.
It is therefore unlikely that stress riser areas would be created.
The effect of impacts near the welds of the divider plate are discussed in the next section.
4.2.5 Tubesheet-to-Divider Plate and Divider Plate-to-Channel Head Welds Because of the location of these welds, the number of direct impacts they would receive from a loose themal sleeve is low.
In previous 28630: 1
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circumstances involving loose parts on the primary side of the steam generator at other plants, inspection of these welds showed no indica-tion of degradation due to impact forces.
Short-tenn fatigue induced by forces being transmitted to the welds by continual impacting of the divider plate and/or channel head in the region close to the welds is of no concern, due to the flexibility of the divider plate the geometry of the weld, which would limit the number of direct or close impacts by a thermal sleeve, and the low stresses induced in the welds.
4.2.6 Channel Head The inside of the channel head is weld clad with a ductile, austenitic stainless steel.
Impacting of the thermal sleeve on the channel head is not expected to produce any sharp areas where a point of stress concen-tration could form. The ductility of the clad material makes it unlikely that sufficient impacts will occur on a particular spot to produce either cracking and loose cladding. Therefore, impacting of the thermal sleeve is not expected to adversely affect the channel head and cladding.
4.2.7 Conclusions i
The potential entry of a 14 inch thennal sleeve from the pressurizer surge line into the primary side of the steam generator is not expected to produce any safety concerns during operation of the steam genera-tor. Short-term operation with the sleeve present in the steam genera-tor is not expected to dislodge tube end pieces.
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4.3 REACTOR INTERNALS The reactor internals were evaluated to detennine the effects of impact and wedging loads on reactor guide and support structures due to the presence of loose thennal sleeves in the reactor coolant system.
4.3.1 Upper Internals The 3 inch charging injection line thermal sleeve, the 6 inch safety injection line thennal sleeves and the 12 inch accumulator injection line thennal sleeves will be confined between the lower core plate in the reactor vessel, and the steam generator cold leg plenum. As such, these thermal sleeves will have no impact consideration on reactor upper internal s.
The thermal sleeve located in the 14 inch hot leg pres-surizer surge line does have the capability of becoming lodged in the upper internals.
In a back flow or alternate leg blowdown situation, a loose surge line thermal sleeve could travel back through the hot leg into the upper internals. The following assessnent utilizes plastic analysis to detennine impact loads on support columns and guide tubes in the reactor upper internals.
Support Columns Length 78.77" 0.D.
7.49" I.D.
6.53" Thickness = 0.4875" 2
A = 10. 72 i n Materi al: ASTM A 479 Type 304 stainless steel, cold finished.
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Guide Tubes 17 x 17 Length 125" Thickness 0.25" Size 7.34" x 7.34" 2
A = 7.09 in Back Flow Velocity Mass back flow 935 lbm/sec 3
Density 9.34 lbm/ft 2
Area 4.587 ft Y - 21.8 ft/sec UPPER INTERNAL STRESS
SUMMARY
LOAD STATIC COLLAPSE DEFLECTION (KIP)
LOAD (KIP)
(INCH)
Support 17.45 22.1 0.266 Col umn Guide 12.3 25.9 0.360 Tube As seen from this table, the loads exerted are less than the static collapse load, therefore, impact loadings on reactor internals upper support columns and guide tubes are acceptable.
Objects in the bottom of the reactor vessel would not be expected to reach the upper internals due to the filtering action of the lower 2863Q:1
internals and fuel assemblies. The close spacing of the rods, the con-figuration of the grids and the flow deflectors, and the configuration of the nozzles prevents large particles and most other particles from reaching the upper internals.
Small particles which could pass through the fuel assemblies are likely to pass through the upper 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.
As part of the normal startup tests, control rod drop times are recorded and evaluated to confirm proper driveline performance.
In the unlikely event that a foreign object would become lodged in the upper package during operation and cause a driveline to become inoperable, the existing FSAR analyses assumption of one stuck control rod assembly would not be exceeded.
4.3.2 LOWER INTERNALS l
The reactor vessel and lower internals were analyzed for structural integrity with thermal sleeves from the 3 inch charging line, 6 inch safety injection line and 12 inch accumulator lines within the reactor vessel. The thermal sleeve from the 14 inch pressurizer surge line is unable to reach the reactor vessel lower internals.
4.3.2.1 Core Barrel It was assumed that a complete 12" sleeve strikes the core barrel at the inlet nozzle velocity. Since the sleeve is thin it will deform before the core barrel defonns. Therefore, the load applied to the core barrel is determined by the load capacity of the piece.
Assuming an ultimate 2
strength of 63.5 ksi for the piece, and an impact area of 5.84 in for the end of the sleeve, the maximum load applied to the core barrel is 371 kips.
Assuming the core barrel responds as a cantilever beam, the impact stresses in the core barrel are calculated to be small.
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= 1120 psi and ' max = 435 psi).
max 2863Q: 1
t The method used for the minimum missile energy requimd to perforate a l
target plate per WCAP 9934 results in a maximum depth of dent equal to
.034 in.
Due to the low magnitude of the impact stresses and the short time dura-l l
tion of impact loads, the core barm1 is not expected to be affected by impacting loose parts.
4.3.2.2 IRRADIATION SPECIEN GUIDES The irradiation specimen guides are welded to the outside of the themal i
shield panels. The top portion is welded using a.38" groove joint, 4.56" long on each side.
The middle portion is intermittent 0.11" bevel welds totaling 70.57" long on each side. The contact area is calculated by assuming the face of a quarter section of a 12" sleeve strikes the top of the specimen guide at 34 ft/sec. The impact forte is calculated to be 63,400 lb. The maximum shear stress is 3,340 psi.
In view of the small magnitude of the shear stress, the specimen guide will not be affected by the impact.
4.3.2.3 BOTTOM MOUNTED INSTRUENTATION TUBES The instrumentation tubes in the bottom head of the vessel were evalu-ated for impacting of themal sleeves or themal sleeve sections. 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 without first striking the internals.
Resulting values were compared to appropriate shear and collapse load all owables.
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The shear strike.was evaluated only for the largest themal sleeve (one-half of a 12 inch themal sleeve) which could impact the instrment tubes. The maximum shear stress was found to be only 1.13 KSI which gave a margin of safety of 36.8 compared to the allowable of 0.6 Sm.
The loads on the instrument tubes resulting from the bending strike of a half section of the 12 inch themal sleeve were evaluated as exceeding the instruentation tube collapse load. This result indicates that plastic defomation of an instruentation tube could result if the tube were struck in an unfavorable manner by the loose themal sleeves.
However, due to the ductility of the Ni-Cr-Fe alloy tube, defomation could occur, but the tubes will not rupture and will continue to protect the thimble guide tubes.
The thimble tubes would therefore not rupture and the pressure boundary would remain intact.
In the unlikely event that the failure of a bottom mounted instruenta-tion tubb produced leakage, the double ended break of this tube results 2
in a leak area of 0.00024 Ft. Assuming a discharge coefficient of 1.0 and using the Zaloudek subcooled critical flow model which over-predicts leak flow, one charging pmp in the nomal charging mode can provide makeup for at least 3 broken tubes.
This would be classified as a leak, not a LOCA, and RCS pressure would be maintained at 2250 psia.
If both charging peps were available, additional leaking tubes could be tolerated.
Small break LOCA analyses with minimum safeguards SI have demonstrated that full instreent line breaks in as many as 5 instrument tubes will not produce in core uncovery.
RCS depressurization and automatic SI initiation will occur, however, this small break LOCA will maintain forced or natural circulation, and the RCS will reach equilibrium conditions.
Therefore, migrating themal sleeves striking the instrumentation tubes in the bottom head of the vessel does not constitute a safety hazard.
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4.4 REACTOR VESSEL During plant heatup, the gap between the reactor vessel bottom head inside surface and the bottom of the secondary core support structure 1
will decrease. A foreign object present in this area could impose mechanical loadings on the vessel. Due to the size of the gaps a full 3" sleeve could not enter the gap. A quarter section of a 3" sleeve could enter the gap, and the forte necessary to deflect the piece to the minimum gap size was calculated to be approximately 6,460 pounds. This load is acceptable.
The effect of impacts on the radial key was also evaluated. The largest j
piece that could enter the outer annulus of themal shield is detemined to be one half of a 12" themal sleeve. The impact velocity is assuned to be 36 ft/sec. and the impact force is detemined to be 42,300 lb.
Assuming all the impact load is carried by the six 1" dia. dowel pins, the resulting stresses is 8980 psi. Comparing to the allowable stress intensity, this gives a margin of safety of approximately 4.7.
4.5 NUCLEAR FUEL Foreign objects in the primary system have the potential on the nuclear fuel to produce:
- 1) partial flow blockage of fuel assenblies due to the foreign object becoming wedged in the fuel assembly flow paths, and 2) clad wear due tc foreign-objects becoming lodged in the assenbly or between two assemblies. Flow blockage effects are discussed in Section 5 of this report.
From a fuel mechanical design viewpoint, foreign objects should not pose an operational concern when the fuel assenblies are seated properly on the core plate. The loose pieces should be stopped by the bottom nozzle or the lower core plate due to dimensional considerations. Although highly unlikely, it is possible for a very small foreign objects to wedge between fuel assenblies and cause fretting and/or grid wear. This is highly improbable due to the fact that space between fuel assenblies is approximately 40 mils, i.e. approximately one fourth the thickness of I
1 2863Q: 1
i the thsnnal sleava material.
Sh::uld a fretting mechanism produce a penetration of the clad on a fuel rod it is unlikely that any radiation release would approach the technical specification limit, and as such no safety concern would exist.
l Oue to the relatively large frapents expected from the themal sleeves, the transport of loose pieces into and through the fuel assemblies is unlikely.
4.6 REACTOR C0OLANT PUMP There are no thennal sleeves of the subject design located in piping I
connections between the reactor coolant pump (RCP) and the steam genera-tor. A loose thennal sleeve can enter the RCP only when a reverse flow condition occurs, in which case the plant is not operating at power.
If this occurs a thennal sleeve or portion of one will not affect the pres-sure boundary integrity due to the geometry, mass and low impact energy of the pieces.
An intact 3 inch thennal sleeve or similar size fragments of a larger thennal sleeve can pass through the puh.p internals without significant i
defonnation.
The larger thennal sleeves would not pass through the pump diffuser and impeller during a non rotating impeller condition. During RCP startup l
the forwani flow would eject any frapents.
l If thennal sleeve fragments did lodge between the impeller and diffuser as to produce interference, the material may be pinched or sheared between the impeller and diffuser venes due to the very high torque of the RCP.
A consequence raay be an increase in shaft vibration with continued RCP operation, i.e., no locked rotor or pressure boundary violation is expected to occur.
Increased vibration could be observed by the operator and corrective action could be taken.
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A similar safety evaluation of larger material (11/16 inch thick, 304 SS) postulated t'o enter the RCP in various size fragments was previously perfonned, and it concluded that there was no safety concern.
In summary, migratory thennal sleeves are not considered a safety con-cern for RCP integrity and operation.
4.7 PRESSURIZER l
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 thennal sleeves. On the pres-surizer thermal sleeves the upstream end of each sleeve is welded 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 pressuri zer. The flow distribution screen inside the pressurizer at the surge line connection prevents that sleeve from entering the pres-surizer. Similarly, the spray header traps the sleeve on the spray line connection.
Due to their method of attachment, it is also very unlikely that these sleeves would become loose within the reactor coolant system.
In addi-tion operating experience has indicated no history of migration of these sleeves, thus they are not considered in this safety evaluation.
Based on the most probable movement of any dislodged thermal sleeves from the 12 inch SI lines or the 3 inch charging line it is extrenely unlikely that such sleeves or part thereof would produce mechanical concern or become lodged in the pressurizer inlet piping or the pressuri zer.
4.8 PRIMARY LOOP STOP VALVES The effect of loose thennal sleeves on the primary loop stop valves either through impact or erosion is expected to be negligible since 2863Q: 1
I there are low radial flow velocities and no appurtenanc:; extending into the flow path during plant operation.
The renote possibility exists that the disc guides, located outside the flow path, could deform if impacted by a thermal sleeve.
If this were to occur, the valve may not reach its fully closed position; however, the primary coolant pressure boundary would not be violated. The loop stop valve has no safety function, and any restrictions to closing would not present a safety concern.
4.9 OTHER REACTOR COOLANT SYSTEM COMP 0NENTS Due to the physical separation from the remainder of the reactor coolant system of such components as control rod drive mechanisms and safety, relief and block valves, no adverse effect is expected to result from migratory thermal sleeves in the reactor coolant system.
4.10 AUXILIARY SYSTEMS The possibility of the potentially migratory thermal sleeves affecting the operation of other systems connected to the RCS was also investi-9ated in this safety evaluation.
The evaluation below considers each thermal sleeve location and the possible paths to systems or components interfacing the RCS.
4.10.1 SURGE LINE THERMAL SLEEVE If the surge line thermal sleeve came loose during operation, it wnuld be moved by the loop flow to the steam generator inlet plenum. The sleeve could impact the RTD bypass line scoops or the thennowell.
The potential effects on these canponents is discussed in Section 4.1.
The SIS, drain line, and loop stop valve bypass line connections and the l
pressure tap in the hot leg do not protrude into the loop flow and are not vulnerable to impact damage. Entry of a piece of sleeve into the SI l
nozzle would be highly unliKely due to the location, orientation and I
stagnant flow conditions of the line.
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4.10.2 NORMAL CHARGING l.INE TIERMAL SLEEVES The 3" charging line enters loop '8 downstrean of the 12" accumulator discharge line and the 6" safety injection line. The normal flow is toward the reactor vessel. There are no other connections to the RCS piping between this line and the vessel. 'Thus the thermal sleeve or parts thereof if dislodged would be expected to migrate to the reactor vessel. With reverse flow in loop B it could be postulated that the sleeve from the charging line, or parts thereof, might enter the accumulator discharge line or the safety injection line.
However, the parts would not migrate up these lines due to the geometry and stagnant flow conditions in the lines.
In order to enter the lines the sleeve would have to travel against gravity since both connections are in the upper half of the RCS pipe. This would create no safety concern.
4.10.3 ACCUMULATOR DISCHARGE THERMAL SLEEVES The 12" accumulator discharge line is located upstream of the 3" charging line and downstream of the 6" safety injection line.
In loop B a dislodged accumulator line sleeve or parts thereof would be expected to migrate toward the reactor vessel.
Since the charging line has an ID of about 2.1" with a sleeve and 2.7" without a sleeve it is impossible for the entire sleeve or fragments to enter this line with its sleeve
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intact. Even if the sleeve were missing it is improbable that fragments at most.1" smaller would enter the line.
In loop A and C the accumulator discharge line enters the RCS upstream of the 6" safety injection line and the 4-inch spray line. There is no l
other connection downstream between the accumulator discharge and the vessel. Thus dislodged sleeves or fragments would migrate toward the vessel.
If reverse flow is postulated, it would be impossible for an entire accumulator discharge line thermal sleeve to enter either a l
safety injection line or a spray line since the outside diameter of the sleeve is about 10.5" while the ID's of the SI lines and spray lines (without sleeves) are on the order of 5.1" and 3.4", respectively.
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could be postulated that fragments might enter the SI line or the spray line. Entry into tne Si line would be of little consequence since there is no flow in the line and the fragments would be dislodged on safety inj ection. dnould a fragment enter the spray line it could be postu-lated that the fragment could migrate against gravity and possibly damage the pressure control valve. Should the valve stick open a pre-mature plant shutdown would be the worst consequence. There would not be any safety significance.
4.10.4 SAFETY INJECTION LINE THERMAL SLEEVES The 6" safety injection discharge line enters the RCS in Loop B upstream of the accumulator discharge line and the normal charging line and upstream of the accumulator discharge and spray line in loops A and C.
A dislodged sleeve or part would normally travel to the reactor vessel if it fell into the RCS.
If it simply became loose, it would be unlikely to migrate up the SI lines as there is no flow in these lines.
On safety injection such parts would be expected to migrate with the flow toward the reactor vessel.
It could be postulated that a sleeve or part thereof enters an accumulator discharge line.
From this line it would be flushed toward the reactor vessel on accumulator discharge.
It would be impossible for an entire safety injection line thermal sleeve to enter the normal charging line and improbable for a fragnent to enter the line due to the size and construction of the line.
For the same reason it is also improbable that a fragment would enter the spray line.
As described above, should this occur it would be of no safety significance.
To summarize, it is unlikely that dislodged sleeves or parts thereof would enter SI, accumulator discharge or spray lines. For this to occur, the parts would have to travel against gravity as all of the connections are in the upper half of the RC piping.
If parts did enter any of the lines, they would not migrate to places that would adversely affect plant safety.
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4.11 MATERIALS No unacceptable material would be introduced into the reactor systems as a result of the migration of a thennal sleeve. Minor clad wear could occur on the surfaces of carbon steel components. However, this would not present any safety or operational concerns due to the very slow corosion rate of the carbon steel in the reactor coolant environment.
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5.0 FLOW BLOCKAGE
5.0 INTRODUCTION
In postulating the presence of dislodged themal sleeves in the reactor coolant systen, an evaluation was made of the effect of the dislodged sleeves or parts there of blocking flow in the core and in various locations in the reactor coolant system. The evaluation considered that all themal sleeves become dislodged in the reactor coolant systen and have migrated by RCS flow to the following locations:
i A.
The 3", 6" and 12" sleeves protrude into the cold leg flow. This case bou'nds the case where such sleeves lodged in the lower internals and block flow at the lower core plate.
I B.
The 14" sleeve from the pressurizer surge line blocks flow at the steam generator tube sheet.
(The case of the intact 14" sleeve partially blocking flow in the hot leg was also analyzed, however, blockage at the steam generator tube sheet was detemined to be more conservative).
The evaluations considered the effect of blockage on reactor coolant systen total flow, local flow distributions in the core during nomal operation, and the effect on LOCA and non-LOCA accident analyses.
5.1 REACTOR COOLANT SYSTEM TOTAL FLOW For the analysis of reactor coolant systen flow reduction, the loose 3",
6" and 12" themal sleeves were modeled as protruding fully into the cold leg flow. This was the most conservative of several blockage configurations analyzed.
The 14" sleeve segments in the steam generator were asstsned to com-pletely block flow in 10 percent of the tubes. This is a very conser-vative assumption since it is extremely likely that the segments will retain their curvature and only would pemit a flow restriction rather than total flow blockage.
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The results of this conservative analysis showed that the total reduc-tion in RCS flow was approximately 1.9 pen:ent. This still results in the RCS flow being greater than themal design flow, which is a conser-vatively low value of flow rate upon which 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 power. Based on the above evaluations it was concluded that the reduc-tion in RCS flow would not affect design margins in nomal operation.
5.2 LOCAL CORE FLOW DISTRIBUTION Due to the wide distribution in the size of the potential foreign objects, the evaluation involved three postulated conditions:
- 1) the effects of material entrapped by the lower core plate, 2) the effects of material entrapped by the bottom nozzle plate, and 3) the effects of material carried upward into the asseinblies.
Infomation and discus-sions pertinent to each condition are given below.
Note that the response given in Section 5.2.2 is consistent with the North Anna FSAR Chapter 4.4 dealing with the flow blockage.
5.2.1 Material Entrapped by the Lower Core Plate The parts from the dislodged sleeves remaining below the lower core plate could produce greater core blockage than the smaller segments reaching the fuel nozzles, since the smaller pieces could only reach the fuel nozzles in a lengthwise orientation.
In perfoming this evalua-tion, it was assumed that the sleeve segments rema'.n curved, and thus do not completely block flow, but may produce some restrictions in the flow to the core.
The information available on thermal effects due to flow blockage indi-cates that there will be no significant increase in the likelihood of DNB at nomal operating conditions. WCAP-7956 shows results from a blocked assembly flow recovery test and WCAP-8054 shows that a 10 per-cent flow reduction in the hot assembly and its 8 surrounding assemblies reduces DNBR by only 0.3 pen:ent.
Since the themal sleeve parts will 1
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remain curved, there will always be some flow through all of the lower core plate holes. This, along with the fact that the total core themal design flow will remain unchanged, will insure that the DNBR will not be reduced by more than a few pertent.
Thus, the effect of blockage on local core flow distribution and DNBR is judged to be insignificant.
5.2.2 Material Entrapped by the Bottom Nozzle Plate Because of the limiting flow holes in the bottom nozzle plate, only small parts could pass through the bottom nozzle plates and up into the fuel assembly. The size and shape of the smallest themal sleeve parts would prevent them from moving completely through the lower core plate, but could allow the to locate against the bottom nozzle adapter plate in an upright position.
It is considered unlikely that parts which are small enough to be trapped by the bottom nozzles in this manner would totally block the flow to any one assably. However, THINC IV predictions (Reference 1) indicate that, even when blockage covers the complete nozzle, full recovery of flow occurs about 30 inches down stream of the blockage.
Thus inlet blockage effects would be limited to the lower portion of the active core, where DNB and LOCA are not limiting concerns.
5.2.3 Material Within the Fuel Assably Because of the size of the majority of parts considered, most of the parts could not pass through the bottom nozzle plate. Those parts that could pass through the bottom nozzle would not pass through the lower grid. This would not affect the DNB evaluations for this core. Tests (Reference 2) on open lattice fuel assablies indicate that a 41 percent blockage is acceptable, with disappearance of the stagnant zone behind l
the flow blockage after 1.65 L/DE. These types of local blockages have little effect on subchannel enthalpy rise at:d cause only minor perturba-tions in local mass velocity.
In reality, a local flow blockage is expected to promote turbulence, and thus, would likely not affect DNB.
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REFERENCES 1.
Hockref ter, L. E., Chelemer, H. and Chu, P. T., "THINC IV - An Isoproved Program for Thennal-Hydraulic Analysis of Rod Bundle Cores," WCAP-7956. June 1973.
2.
Basmer, P., Ki rsh, D. And Schultheiss, G.
F., " Investigation of the Flow Pattern in the Recirt:ulation Zone Down Stream of Local Coolant Blockages in Pin Bundles, "ATOMWIRTSHAFT,17, No. 8, 416-417, (1972).
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S.3 NON-LUCA TRANSIENT ANALYSIS Flow blockage by dislodged thermal sleeves in the reactor coolant system potentially effects non-LUCA transients only in that there is a slight reduction in total RCS flow, as discussed previously in Section 5.1.
An evaluation was performed on the effect of the RCS flow reduction on the non-LOCA transients.
In non-LOCA transient analysis, it is conser-vatively assumed that accioents are initiated with the reactor coolant system operating at thermal design flow (TDF).
A reduction of 1.9 percent due to the thermal sleeve flow blockage effect on RCS flow still results in a measured flow greater than TDF. This demonstrates that the j
current safety analyses remain valid.
5.4 LOCA EVALUATION One may postulate that thermal sleeve material may in the future beceme located beneath the lower core plate or in the upper plenum of the North i
Anna reactor vessel.
An evaluation of the impact of such material in the lower and upper plenums on the limiting case break ECCS performance analysis (CD = 0.4 DECLG) for North Anna follows:
A.
Overall system thennal performance at 100 percent power has been shown to be negligibly different with large parts present in the RCS.
Since thermal design RCS flow can still be demonstrated for North Anna, the ECCS perfonnance analysis previously performed remains applicable with regard to RCS flow conditions.
B.
Thermal sleeves impinged against the lower core plate will remain curved, so there will always be some flow through all of the lower core plate holes, and no assembly will be starved of flow.
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Flow redistribution above a postulated sleeve location will occur in the first several inches of the fuel during nomal operation, and that therefore reduced minimum DN8R is not of concern in the hot assembly.
In a LOCA analysis, post-LOCA themal-hydraulics pre-dicted for the hot assembly directly define the calculated PCT.
Core flow post-LOCA is characterized by positive (nomal direction) and negative core flow periods, in that order. From the above, during positive core flow when RCP perfomance detemines flow magnitude and direction as during nomal operation, thermal-hydrau-lics should be equivalent to those computed in the existing LOCA analysis. When the flow reverses any parts impinged against the core plate should fall into the lower plenum and thus not be in a position to impact the calculated core flow. Thus, calculated perfomance of the ECCS systen should not be impaired by the presence of dislodged themal sleeve material in the vessel lower pl enum.
C.
One might postulate, in an ECCS perfomance evaluation, breakup of sleeve material into smaller parts which become lodged within the fuel and provide additional blockage during core reflood following a LOCA event.
The limiting case ECCS perfomance analysis for North Anna exhibits its maximum calculated PCT when the core flooding rate is less than one inch /second. Appendix K requires a fuel blockage flow penalty to be considered during reflood at such flooding rates so any postulated added blockage from thennal sleeve material within the North Anna Plant hot assenbly will adversely impact the calculated PCT.
The currently-docketed large break LOCA analysis for North Anna (Reference A) utilizes the February,1978 Westinghouse ECCS Evaluation Model as amended with an evaluation of the impact of NUREG-0630 fuel rod models.
In that evaluation a flow reduction penalty based on 75 pen:ent blockage in the hot assenbly is assessed.
In fact, the maximum blockage possible with NUREG-0630 is 2863Q:1
71.5 percent, so a blockage leval 3.5 percent in excess of tha NUREG-0630 maximum is presently being considered for North Anna.
If a small particle of material from a failed themal sleeve were conservatively postulated to enter the hot assembly at North Anna during a LOCA and become lodged at the coplanar locus of blockage from fuel rod ruptures, the maximum additional blockage over and above the NUREG-0630 defined value would be that due to completely closing the limited flow area remaining between two adjacent rows of rods in the assembly or roughly 1.8 percent. This added blockage from a postulated themal sleeve particle is more than accomodated by the excess blockage assmed in the current NUREG-0630 evaluation which is part of the currently approved North Anna LOCA analysis; no further PCT penalty need be imposed.
To stamarize the above, predicted core themal-hydraulics post-LOCA are independent of the postulated presence of themal sleeve parts against the lower core plate at time zero. The presence of small parts within the hot assembly could adversely affect the calculated PCT in the 10CFR50.46 Appendix K analysis, but such an effect has been shown to be accounted for in the currently approved North Anna analysis.
The hot leg might also contain loose parts caused by a breakup of the dislodged pressurizer themal sleeve. Due to the plethora of guide tubes, support columns, etc. in the upper plenum it is highly unlikely that any part could orient itself in such a way as to significantly block flow exiting any particular fuel assenbly. The pieces in the hot leg are not of concern from the standpoint of ECCS perfomance.
One might postulate that a pressurizer surge line themal sleeve pre-pelled by post-LOCA blowdown fortes might affect a particular guide tube in the North Anna Unit 1 upper internals as discussed in Section 4.3.1.
In the Westinghouse large break LOCA Evaluation Model the conservative assumption is made that no credit be taken for insertion of the control rods. Thus, non operation of a guide tube will not have any significant affect on the North Anna Plant ECCS performance analysis limiting case.
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Finally, the blockaga of steam generator tub:s in the loop containing the pressurizer surge line was also considered.
In this case, it is assumed that segments of the pressurizer surge line thermal sleeve are held against the steam generator tube sheet by reactor coolant pump flow prior to a LOCA.
In this situation, during the initial part of the LOCA transient when the RCS is still in forward flow due to the influence of the RCPs, the core thermal-hydraulics should be equivalent to the exist-ing LOCA analysis. When the steam generator channel head voids, the thermal sleeve pieces held against the tube sheet should fall into the channel head and not affect flows in the reactor coolant system. Thus, the effect of the postulated sleeve segments at the steam generator tubesheet should not significantly affect the 10CFR50.46 Appendix K ECCS analysis for North Anna.
Reference A: Letter from R. H. Leasburg (VEPCO) to H. R. Denton (NRC),
Serial No. 627, November 12, 1981.
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1 ATTACIDfENT IV IMPACT OF POSTULATED BLOCKAGE ON 1981 MODEL LOCA ANALYSIS The recently submitted large break LOCA analysis for North Anna, which was transmitted to the NRC by Reference A, utilizes the 1981 Westinghouse ECCS Evaluation Model. This model internally incorporates the NUREG-0630 fuel rod models. The impact of the postulated additional 1.8% channel blockage, as discussed in the Loose Thermal Sleeve Safety Evaluation, was evaluated for this analysis using the sensitivity of peak clad temperature to blockage presented in the NUREG-0630 blockage addendum to th2 Refirence B analysis. This sensitivity indicated that the increase in peak clad temperature is not significant and can be accommodated by the margin which is available to the 2200 F peak clad temperature limit.
S
=
References e
A.
Letter from R. H. Leasburg (Vepco) to H. R. Denton (NRC), Serial No.
460, August 4, 1982 B.
Letter from R. H. Leasburg (Vepco) to H. R. Denton (NRC), Serial No.
627, November 12, 1981 l
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ATTACHMENT V E PROGRAM OPERATOR AWARENESS SURVEILLANC Unit No. I will be instructed pri OPERATOR GUIDANCE operators for North Annaof the potentialof the Loose Parts for existing and addi-Use control room or fragmented thermal sleeves.
startup will be well as the The and during plant Specific instructionshich is being resulting from (LPMS) will be reviewed asof the supplemen whole to activities.
Montioring System surveillance ining thermal slee tional loose parts in the use provided by Westinghousespecifically monitor the rema installed to Audible every 31 days.
on a least once l Advisors SURVEILLANCE functionally tested atmed by the Shift Technica f
The LPMS will be of the LPMS are being per orThis is a new surveillance weekly basis to his verify the free checks daily basis.
will be performed on aTechnical specifications re Control rod exercisescontrol rod assemblies.
once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to movement of the i
activity on a monthly bas s.
monitored by loose part fragments.
at least activity will be clad damage due to frettingthis activity on a bi hly basis.
Reactor coolant any potential Technical Specifications require operable by increasing detect verified as Detection System.
will be f damage of the Incore Detector Leakagetion (contr thimbles instrumentation Detector of the Incore testing to verify operabilitywould give a positive indica operability functional test Incore verified by performance of averified ope immediately weekly by the The thimble.
This system detector would be will be incore would be alarm any leakage d core surveil-to an Leakage Detection System the prior to Unit 1 startup andIn addition, the normally requireill be perfo that insure to fication of the Advisor incore movable detector system wwill se Technicalcontrol room.
Shift alarmed in the which utilizes theThese monthly flux mapsf the guide thimb ters to be lance scheduled.
operability and integrity o sists of installing acceleromedischarge between the 12-inch accumulatord on the The supplemental LPMS for Unit 1 con accelerometer is to be locateand g
located on each loop cold leg vessel. Another which is located as clo line and the reactorhot leg between the 14-inch surge l response.
r is a preamplifierorder to obtain the be l
i Associated with each acceleromete l rometers one to two feet downstri nal sible in order to obtain the best s grec to the accelerometer as posVepco plans to install thermal sleeves to beThe location of these a control cabinet will provide coal sle ntinuous s was a Westinghousesupplemental LPMS for the the control room, In the This alarm monitoring.
North Anna.
l to the existing LPMS at
.