ML20084T333
| ML20084T333 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/03/1978 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20084T280 | List: |
| References | |
| NUDOCS 8306230126 | |
| Download: ML20084T333 (29) | |
Text
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Proposed addition to Browns Ferry Nuclear Plant Unit 1 Technical Specifications o
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B306230126 790110 PDR ADOCK 05000259 p PDR . . .
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gre Following its initial refueling outage, the unit 1 reactor shall be maintained at a te=perature of not less than 5000F and with a rated flow of not less than 30% for an accumulated period of not less than 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> before exceeding 5% rated power. Upon completion of this period of operation, the reactor shall be operated for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 100% rated core flo.i before exceeding 305 rated power.
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- After March 1,1978, this page shou)1 be removed.
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BROWNS FERRY UNIT 1 OPERATIONAL SAFETY EVALUATION IDST SHOE COVER - FALL 1977 REFUELING OUTAGE DECDGER 26, 1977 e
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I. Situation Sumary A rubber shoe cover was dropped into the Browns Ferry unit 1 reactor vessel Repeated atte= pts to locate the during the unit's first refueling outa6e. Concurrent with the search activities, TVA requested choe cover failed.
the General Electric Cc=pany to perform the necessary testing and safety
- analysis to determine the effect, if any, on reactor coeration with the
- shoe cover in the pr N7 system. As a result of GE's testing and analysis and TVA's own investigations and evaluation, TVA concludes that by following CL prescribed low power heatup program, the unit can subsequently be operated 4
tt its licensed power level with no compromise to reactor safety.
' II. Event
! A rubber shoe cover was dropped into the unit 1 reactor vessel on Ssptember 18, 1977. The moisture separator was in the vessel, B and D rssidual heat re= oval (RHR) pu=ps were running in the shutdown cooling l
mode, and the recirculation pucp B was in operation on the opposite loop.
I. The shoe cover was last seen faning below the mid-support of the moisture separator at the vessel 270 location. Recirculation pump B ranRHR for spproxicately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the shoe cover entered the vessel. During pumps B and D ran in shutdown cooling mode until October 2, 1977.
i the 114-day period prior to October 2, the noisture separator was removed Prior from the vessel, and all fuel was transferred to the fuel pool.
to core unloading, an inspection with binoculars was made of the fuel and upper tie plates. At the co=pletion of the maintenance work in the
' vessel, a search was initiated to locate the rubber shoe cover.
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- III. Search The rearch for the shoe cover was divided into two major parts. The in-vessel search was conducted both with binoculars and underwater TV
-cameras and included inside the shroud :f n the area above the core support plate, the annulus area between the shroud and vessel vall,-and portions cf the area below the core support plate. The out-of-vessel search included recirculation loop B piping, RER heat exchangers and piping, l
i the moisture separator, the fuel pool, the fuel lower tie plates, and the equipment drain sump.. No trace of the rubber shoe cover was found in cny of the areas. A more detailed description of the techniques
.used and areas searched are given in' Attachment 1.
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. . i IV. Test and Analysis
- In parallel with the search efforts, TVA began to investigate the effect ,
of the reactor system operating environment on the rubber shoe cover. A j
. test was attempted at the Wylie Laboratories in Huntsville, Alabama; however, this t_est was unable to reproduce reactor operating conditions, and no valid conclusions could be drawn from the test results. The 1
General Electric Co=pany was then requested to perfor= tests in their autoclave at Vallecitos in Pleasanton, California. Tingley shoe covers, similar to that lost in the Browns Ferry vessel, were autoclaved at
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i reactor operating temperatures and then placed in a flow loop sinalating I a fuel support casting, a fuel assembly, and lower tie plate.
s The autoclave tests demonstrated that the shoe cover when placed in an *
. environment of 500 F water for a veriod of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> had no tensile 4
strength. The flow tests showed tilat such en autoclaved shoe would j- become brittle and disintegrate if subjected to flow conditions correspondire i to 100 gpm in a 1hel assembly coolant channel. The flow loop tests could 4 only be run up to a te=perature of 192 F; however, the series of tests demonstrated that the higher the flow loop te=peratures, the more effective i in disintegrating the shoe cover.
! Based on a literature survey and tests perfor=ed at Vallecitos, GE has concluded that exposure of a shoe cover to high te=perature water is i snore effective than exposure to radiation. From the GE tests, the l following conclusion c'en be drawn. A rubber shoe cover when exposed
! for-40 hours in a reactor ooerating environ =ent (te==eratures greater than 500 F and flou exceedling 60 6pm in each fuel 'assemb3y) pill be i broken into s:n11 pieces so as to be of no consequence to safe reactor 8
cperation. In order to prevent transition boiling during this period, the reactor operation should be limited to a bundle power not to exceed 0.6 MWt. With total blockage at either the fuel support., piece orifice j
or at the lower tie plate grid, there is at least 3 x 10' lbs/hr cocla.nt
- flow availabic. With a minimum flow of 3 x 103 lbs/hr, boiling transition
! will not occur for bundle powers less than O.63 Wt. The 0.6 pt bundle power limit ensures that no fuel dc= age will result from operation with j the rubber shoe cover intact in the reacter vess.e1.
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Consequences other than flow blockage were also evaluated by GE and i found to be acceptable. Increased friction =ay result during control i rod drive move =ent; however, the scram force is greater than cny l postulated increased resistance from shoc friction, and the scram function
! cf the rod would not be affected. No potential da=aBe to reactor internals
! cr recirculation loop cc=ponents is postulated for the shoe either in its whole or carbenized state. no deleterious effects on water chc=istry or any other che=ical property are expected frc= the presence of the shoe. :
i GE's detailed analysis is presented in Attachnent 2 of this report. l i
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V. Evaluation .
In accordance with TVA Operationni Quality Procedure Part II, Section 1.4, f o
o safety evaluation has been perfor=cd for reactor operation both prior to /
and following shoe disintegration, and it was .:oncluded that no co: promise The reector can be operated for is r.ade with regard to rgactor safety.F and 30 percent rated core flow while maintaining Sohours at at least 500 c bundle power of less than O.6 Edt. Following this 24-hour period, the -
r: actor can be operated at rated power with no additional restriction.
Consideration was also given to the possibility that the shoe cover =ight ba somewhere other than in the reactor vessel.
This review identified three possibilities, the recirculation loop piping, the reactor water If held up in the cleanup syste=, and the residual heat removal s'/ste=.
r* circulation system, the shoe cover would be sub.jected to the same soak u if in the vessel, with results as described in section IV of this evaluation. If held up in the reactor water cleanuo syste=, the shoe cover would be in the sa=e environment as in the recirculation loop until it disintegrated into pieces smn enough to pass through the' reactor water Thereafter, cleanupsyste=heatexchangers,whichhave3/4-inchtubes.
the pieces would lodge in the cleanup filter-de=ineralizers If held up inuntil the rc=oved residual fro = use with the filter-de=ineralizer resin.
heat removal syste=, the shoe cover would become lodged in a residual heat re= oval heat exchanger. If pieces passed through the heat exchanger, they would be too s=all to affect reactor safety.
operation at rated temperature end pressure and at 30 percent.cf rated cere flow, while naintaining a =axim fuel asse=bly power of less than O.6 rit, is nchievable. Peaking factors for a given rod secuence can be predicted which will conservatively bracket the actual operating state. Heatup power calculations (based on heatup rates) have been perfor=ed en all three In Browns Ferry units and have been de=onstrated to be conservative.
addition, the 0.6 Kdt limit adds further conservatis= as discussed is cection IV of this evaluation.
Once the reactor has been operated at the reen ended conditions for shee disintegration, full power cperation can be achieved with no co= pro =ise to reactor safety. It has been de=onstrated by testing that the shoe cover, once disintegrated, causes no reduction in bundle flow rates and, therefore, no reduction in operating =argin to existing safety li=its.
VI. Unreviewed Safety Question Determination In accordance with 10 CFF. 50.59, an unrevieued safety cuestion determin: tion has been made. ~2c fo11 cuing three paragrapnc aiscuss each aspect of this determination.
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_b The presence of the shoe cover would not prevent the control rod drive '
system from performing its scram function in the time assumed in the safety analysis. The presence of the shoe cover would have no effect on any cther equip =cnt having a safety function; therefore, the probability of cccurrence of an accident or =alfunction is not increased over that analyzed.
Adequate ther=al =argin is =aintained at low power conditions prior to carbonization and disintegration. After disintegration, there is no cffect on operating =argins. Therefore, the consequences of any accident cr malfunction is not increased over that previously analyzed.
No potential for dernge to either the reactor internals or recirculatica loop co=ponents (other than the effect on fuel of flow blockage) is postulated due to the pliable nature of the shoe cover prior to carbonization.
- Subsequent to carbonization of the shoe cover, no potential for such da= age exists due to its frangible nature. The effect of flow blockage on fuel has been evaluated, and ther=al =argin is maintained at low power prior to and during carbonization of the shoe cover. Subsequent to carbonization and disintegration, flow blockage is of no concern. Therefore, the possibility for an accident of a type other than those previously anniv:ed is not created.
As discussed above, the shoe will cause no measurable difference in scram insertion ti=es; therefore, the scram times as required by technical specifications will be =aintained. Adequate ther=al margins are assured at low power assu=ing total blockage of either the fuel support casting cr the lower tie plate grid until the shoe cover has disintegrated. Once the shoe cover in in the carbonized state and has disintegrated, it will have no measulable effect on coolant channel flow or pressure drop. The margin to the safety li=it for fuel cladding integrity as defined by the technical specifications is preserved.
Based on the above deter =inations, operation with the shoe cover in the reactor vessel or adjoining syste=s does not constitute an unreviewed aafety question as defined by 10 CFR 50.59 e.
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- ATTACHMENT 1 SEAECH FOR RUBBER SHOE COVER IN BROWNS FERRY NUCLEAR FLANT UNIT 1 I. Background Unit 1 of the Browns Ferry Nuclear Plant was shut The down for ref September 13, 1977.
tha vessel flange, a rubber shoe cover was dropped into the_ vessel.as show shoe cover was dropped at approximately 270 disappeared a th2 separator and the vessel wall.
the B recirculation pu=p was in operation and the B and D RHR pugs Thewere op3 rating in the shutdown cooling mode During this 14-dayfrom period, the A re pumps continued in service for another 14 days.
the moisture separator was re=oved from the vessel, and all fuel wasA transferred to the fuel pool.
in the vessel, a search was initiated to locate the rubber shoe were All items cover,.A.
desime_ter. an inspehmirror, and a channel fastener
'quickly found except the shoe cover. l .. .-
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tII. Dsseription of Search The in-vessel The search for the shoe cover was divided into two parts.
scarch was conducted using binoculars and with underwater TV ca.mcra and included the areas where an analysis of the operating conditions at the time the shoe cover was dropped indicated that it might have settled or b2come lodged. The out-of-vessel search included the RHR heat exchangers, h fuel the moisture separator, portions of RHR piping, the fuel pool, t e tie plates, and the equipment drain sump.
A. In-Vessel Search l The initial search for the shoe coverThewas search conducted was from th platform on the vessel flange using binoculars.This continued with the core upper grid to search above the core support plate.
camera was also lowered into the annulus area between the and the vessel wall.
of the vessel which were covered by this part of the search.
l The next step in the search censisted of removing jet pump #16 shown on Figure 1 which is the closest det pu=n to the area where the shoe cover was dropped.
through the jet pump defuser to search the area below the core support plate.
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2 A miniature TV camera was inserted through the cuction of each remaining jet pump and the entire periphery of the vessel below the core support plate searched. This en=cra inspection also aboved that each jet- pu=p throat was not obstructed. The conclusion of this portion of the in-vessel search left only the recirculation loops and the interior of the vessel below the core support plate to be scarched.
The B recirculation pump was running at the time the shoe cover was dropped, and it is considered unlikely that the shoe cover would have settled in the B recirculation loop piping. The =iniature TV camera was lowered into the suction side of the A recirculation
- loop, and the piping was searched as far as the suction isolation valve shown on Figure 3. (This search included the internals of the isolation valve.) Additionally, the recirculation loop piping on either side of the recirculation pu=p was isolated and drained.
A fiber-optic tool was inserted through blanked flanges to search the piping on the suction side of the pump. Figure 3 shows the piping which was inspected.
The final phase of the in-vessel search involved removing six fuel support castings and six control rod guide tubes to search below the core support plate. The miniature TV camera was lowered through the core support plate in each of the holes marked on Figure 4 The lighting and camera resolution would have allowed positive
' identification of any foreign objects the size of a shoe cover for a minimum of two guide tubes in any direction. Consequently, the area shaded on Figure 4 is considered to have been searched. During this phase of the search, binoculars were used to inspect the flow holco in the fuel support castings. At least two out of the four flow holes in each casting were inspected, and no indication of the shoc cover or any portions of it were found.
E. Out-of-Vessel Search -
One of these The out-of-vessel search concentrated in four areas.
was the RHR piping and heat exchangers on the B and D RER loop.
The heat exchant;crs were inspected by entering the fiber-optic through a ther=nl-well on the inlet side of the heat exchanger.
This allowed the arca in the heat exchanger up to the first baffle to be inspected. The bottom head was also removed on the B heat exchanger but no additional areas could be seen. The piping on either side of the B and D RHR pu=ps was also inspected using the fiber-optic tool. Drain lines on the suction and discharge piping were used for access. No indication of the shoe cover or any portions of it were found in the section of RHR piping which could be inspected. .
The second area searched was the moisture separator which had been moved from the vessel to the storage area after the shoe cover was dropped. Because of its location, a closeup visual inspection was possible and was conducted by several different people to ensure a thorough search.
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'i The fuel pool was searched because of the possibility that the shoe cover could have been transported to pool by a fuel assembly during the unloading of the core. The fuel was sipped, the lower tie plates were drilled and inspected, and the fuel storage locations were shuffled during the outage. The lower tie plates of the seven undrilled fuel assemblies being returned to the vessel for the next cyc16 were also inspected. It is considered a re=ote possibility that the shoe cover could have escaped detection in the fuel pool.
The final area searched outside the vessel was the equipment drain sump in the drywell. Drains from the vessel during maintenance and during draining and flushing of the A recirculation loop piping were directed to this su=p, but inspection of the su=p and the strainer on the suction of the su=p pu=ps revealed no indication of the rubber shoe cover.
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G E N E R A L lh) E LE CTR IC SERVICE ENGINEERING YENERAl. ELECTRIC COMPANY .....832 GEORGIA AVENUE DEPARTMENT b C H ATT A N OoG A. TENNESSEE 37402. Phone (615L 894 2550 t ': M *r
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G-ER-7-166 cc: J.G. Dewease , l H.J. Green D.R. Patterson December 22, 1977 N Mr. L. M. Mills, Chief Regulatory Branch Tennessee Valley Authority 303 Power Building Chattanooga, TN 37401
SUBJECT:
LOST PIECE ANALYSIS FOR BROWNS FERRY UNIT 1
Dear Mr. Mills:
This transmits our report " Safety Analysis - Lost Rubber Shoe Cover in the Browns Ferry Unit 1 Vessel," which incorporates TVA's comments. A back-up copy of this report has been telecopied to your office. Please call us if you have any questions. Very tr ly yours, s ( ,. D g ,,,,_. A . L. Service Manager - Nuclear ALV:VGK: law rnelosure
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-t SAFETY ANALYSIS s
LOST RUBBER SHOE COVER IN THE BROWNS FERRY" UNIT 1 VESSEL t
1.0 BACKGROUND
During the l'972 refueling outage of the Browns Ferry Unit 1 Plant, a rubber shoe cover was inadvertently dropped into t jet pump region. Since the specific gravity of the shoe cover is the shoe cover. only slightly higher than that of water, there is a significant probability that the circulation flow which is main system. It can be further postulated that the cover then passed through an operating recirculation pump where it was probably torn into several pieces. 2.0 LOST PART DESCRIPTION The lost shoe covering was made by Tingley-Rubber Co., South P New Jersey. The shoe covers tested were a composite of several materials inc approximately: enual parts natural rubber and carbon black with 2% sulfur added for vulcanization and crosslinking. . 3.0 CHEMICAL AND BEHAVIOR ANALYSIS The natural rubber which is a majorand component of the is basically shoe cover, polymerized has the molecular formula of C "n ' and a mole-isoprene with a structure of [5 -CH 8 C (CH ) = CHCH h between chains cular weight of *10 , that is cros!1 inked by sulfuAbout 1
-(vulcanited).
properties 22-35% sulfur is used to make hard rubber--rubbers vulcanized with intermediate amounts Destructive of ofsulfur distillation naturalare intra of no value. (Reference 1) An analysis of the rubber yields isoprine and other compounds. covering revealed 0% fluorine, 1.23% chlorine, and 2.39% sulfur. ( .
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- a. Corrosion and Other Chemical Action The addition of 1-2 grams of Chlorine to the reactor water will have an insignificant effect on reactor water chemistry.
- -- Under reactor operating conditions, the shoe cover should i eventually decompose, probably into isoprene, methane, I isopentane, (with radiolytic H2 ) r similar compounds. The :
products of decomposition appear to be relatively innocuous and should have an insignificant effect on Reactor Water Chemistry. The small physical amount of material present (168 grams, 6 oz.) and the lack of fluorides or significant amount of chlorides tend to reinforce the conclusion that no significant effect on reactor water chemistry will result from the shoe cover in its initial state or the decomposition products,
- b. Autoclave Testing _
Samples from a rubber shoe cover were exposed in an autoclave at 500"F for 24, 26.5, 38 and 64.5 hours. These samples removed from the autoclave at about 200'F were chewy but had no tensile strength. The small tensile test specimens and strips had broken off their holders and had fallen to the bottom of the autoclave. These broken samples could not be removed until cooled to room temperature due to lack of tensile strength. The whole shoe samples had broken apart considerably. Initially, samples were autoclaved at 550 F for various times up to 65 hours. These samples were also removed at 200*F and exhibited the same properties as those autoclaved at 500 F.
- c. Radiation Resistance A literature survey (references 2, 3, 4) indicates that of the elastomers, natural rubber is amongst the most radiation resistant and butyl rubber the least. Figures 1 and 2 show relative mechanical properties of these materials vs. exposure levels (rads). Radiation exposure is effective in increasing the hardness and lowering the ductility of this type of material.
l l Blending of two elastomeric materials generally reselts in a radiation resistance slightly higher than what would be expected from the weighted average of the two materials. Addition of carbon black will also improve radiation resistance. No l information was found on the effects of adding calcium carbonate or zinc oxide to clastomers. Cased on the limited information available on this material, the literature would predict radiation resistance similar to or slightly better than'that encountered with natural rubber. l i n *M
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Two pieces from a rubber shoe cover have been exposed in the gamma pit at Vallecitos Nuclear Center (VNC). One piece was i previously autoclaved at 550 F for 3 days. The second piece was as received. These pieces have now received 6.3X10,8 rads. The sample which was previously autoclaved has undergone virtually no change. Th'e as received sample has become substantially more brittle, harder and less-elastic, but the swelling noted due to exposure to 550 F water is absent. In summary, exposure to high temperature water appears to be far more effective than exposure to radiation in degrading the properties of this material. Both conUitions cause cross linking to occur.
- d. Flow Tests The purpose of these tests was to observe the effect of flow on a rubber shoe cover in a reactor fuel assembly with inlet l
orifice and lower tie plate. Five tests were run and will be discussed sequentially. A schematic of the test facility is shown in Figures 3 and 4. Test #1 employed a whole shoe cover autoclaved for 64 hours at 550*F. It was placed upstream of the inlet orifice with 60*F water and flow was increased to 90 cpm. The pressure differential between the inlet pipe and the fuel channel was about 4 psi. After 70 minutes of testing, the sole and heel had not broken l' up to go through the lower tie plate. After 2.75 hours about ' 3 pieces of rubber of less than 1 inch diameter were found at - l the simulated core inlet plenum; about 55 pieces (largest piece 1.6" x 1") were found inside the fuel support casting. The approximate weight of the recoverable leftovers was 46.4 grams. , Test #2 used a whole shoe cover autoclaved for 38 hours at 500 F. The test was run at cooled down conditions (60 F). While coming up to the desired flow rate of 65 gpm, the pressure differential surged from 1.7 to 25 psi. Five minutes later, the pressure differential was 3.2 psi; at 27 minutes, it was down to 3.09 psi. The test was stopped at 55 minutes. The shoe had broken into pieces which had been plastered against ' the lower tie plate. About 90% of the total flow area of the lower tie plate was plugged. . Test #3 lef t the test #2 samples plastered on the lower tie plate. More 38-hour 500*F autoclaved samples were added to completely cover the grid of the lower tie plate in a shingle overlapping manner. This amounted to a total of about I shoe covers on the lower tie plate. Volumetric flow rate vs. pressure differential curves are plotted for the plugged and unplugged conditions in Figure 5. About 70% of the flow could reach the' fuel channel even if the lower tie plate was completely covered with pieces of autoclaved rubber at cold (60 F) conditions.
-_ A
l Test #4 was run at hot (176*-192*F) conditions, since it was observed that an autoclaved shoe was weaker at 200*F than l 60*F. A 26.5 hour 500 F autoclaved sample was tested and the flow rate and pressure differentials are shown on Figure 6. Flow was increased gradually at flow rates equal to or lower than reactor flows, i.e., 30 minutes at 2 pi and 30 minutes at The volumetric flow rate versus pressure differential l 4 psi. l characteristic curves, with and without the shoe, intercept' at l 4 psi, as shown on Figure 6. The intercept indicates the point at which the shoe cover was for all practical purposes disintegrated. Test #5 used a 24 hour 500*F autoclaved sample cooled tc 180*F. The maximum flow rate was 100 gpm and the sample was removed after 2 hours. The shoe was completely broken up and only about 4 grams of material was observed on the lower tie plate. After 2 hours there was no measurable effect on flow or pressure differential due to the presence of the shoe, as shown on Figure 7. - The results of all the tests indicate that a shoe that has been exposed to 500 F water for over 24 hours will be broken up into small pieces at temperature greater than 170*F and flows and pressure drops similar to those experienced in the Browns Ferry-1 reactor. 4.0 POSSIBLE INTERFERENCE WITH CONTROL R0D OPERATION In the event that the shoe cover has fallen into the core region, it could have landed on a fuel assembly, the top guide, into an empty fuel assembly position or on a control blade located between channels from where it may have subsequently fallen onto the core plate. If the shoe cover is located on the top surface of a fuel assembly, top guide or core plate, it would not be an immediate concern with respect to control rod operation. If an adjacent fuel support casting were removed, the shoe cover could fall into the control rod guide tube and possibly lodge between the velocity limiter and the guide tube. If the shoe cover dropped into an empty fuel assembly location, it is possible that it could have dropped through the gaps between the control blade and the support casting and passed into and became lodged in the guide tube. Increased friction may be observed susequently during control rod movement. However, the scram force is much greater than any postulated increased resistance from shoe friction; and the scram function of the rod would not be affected. 5.0 POTENTIAL FOR EQUIPMENT DAMAGE OR MALFUNCTION No potential for damage to either the reactor internals or recirc
. loop components other than that discussed in Section 6, is postulated l due to the pliable nature of the shoe cover. prior to carbonization.
I Subsequent to carbonization of the shoe cover, no potential for
- such damage exits due to the frangible nature of the part.
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10 . 6.0 POTENTIAL FOR FLOW BLOCKAGE The concern associated with the lost piece is that associated with 1 the potential for fuel bundle flow blockage and subsequent fuel ' damage. A detailed study of flow blockage in a BWR has been made in a GE re' port (ReTdrences 5 and 6). Based on the results of tests and analysis given in Reference 5, the effect of flow blockage on bundle flow is determined. Using those results, the following statements can be made:
- a. It was estimated that it would take more than an approximately 80% inlet area blockage at 100% power to cause a Minimu.
Critical Power Ratio (MCPR)* 1ess than 1.0; therefore, no fuel rod damage would occur unless more than an 80% blockage occurs.
- b. If the blockage were more than 80%, clad melt and fuel crumbling could occur. This would lead to high radiation sensed by the main steam line radiation monitors, which would scram and isolate the reactor. Off-site doses remain less than 10CFR20 limits (Ref. 5).
Based on the information concerning the size of the lost piece, the following conclusions are drawn: (1) If the piece found its way intact into the lower plenum, the fluid velocities could be high enough to sweep it up toward the fuel bundles. (2) If the intact piece were swept up toward the fuel bundles (central orifice diameter = 2.211 inch, peripheral orifice diameter = 1.469 inch), it could block the orifice sufficiently (i.e. more than 80%) to cause a boiling transition to occur in that bundle. (3) If the blockage were more than 80%, clad melt and fuel crumbling could occur. This would lead to high radiation sensed by the main steam line radiation monitors, which would scram and isolate the reactor. Off-site doses remain less than 10CFR20 limits. . (4) No damage would occur to any bundle adjacent to the bundle with the flow blocked. l Therefore, it can be concluded that steps should be taken to insure that the shoe will be disintegrated prior to high power operation. - If these steps are taken, the shoe will not pose a safety problem. 1
- Reference 5 presented the results in terms of Minimum Critical Heat Flux Ratio (MCHFR) and determined it would take more than 90% inlet area blockage to cause a MCHFR <1.0.
, - n -,
Until the lost rubber shoe cover has disintegrated, the bundle power limit will be 0.6 Mwt. This low bundle power, approximately 10% of the limiting bundle power during normal operation, will insure that boiling transition will be avoided and saturated flow conditions maintained in the unlikely event of a complete flow -~ blockage. As discussed in the flow blockage licensing topical report (Reference 6), the most severe potential event is the complete flow blor.kage at either the fuel support piece orifice or at the fuel assembly lower tie plate grid. Even during the almost incredible event of a complete flow blockage at the lower tie plate grid, at least 3 x 103 lbm/Hr of coolant flow will be available through the channel lower tie plate interface. An additional small amour.: of coolant will " overflow" from the plenum above the core. As can be seen in Figures 3-27, 28 of the GETAB licensing topical report, 2 NEDE-10958-PA, at a minimum flow of 3 x 103 lbm/Hr (Gs.028 x 106/Hr-ft ), boiling transition is not expected to occur for bundle powers less than 0.85 MWt. If no credit is taken for counter current flow, saturated flow conditions will always be maintained for bundle powers less than 0.6 MWt. Note that in the unlikely event of a complete flow blockage, bundle power will further be decreased due to loss of neutron moderation. The 0.6 MWt bundle power limit therefore conservatively insures that no fuel damage will result from operation with the rubber shoe cover lost in the reactor vessel.
7.0 CONCLUSION
S No deleterious effects on water chemistry or any other chemical property are expected from the presence of the shoe. Any possible interference with control rod motion could be identified with control rod drive friction tests. Possible flow blockage will not pose a safety problem if it has disintegrated. In order to insure that the shoe has been broken into small pieces, it will be necessary to bring the reactor up to operating temperature for at least 24 hours at flows exceeding those used in the flow tests described in these analyses. A 100 gpm flow rate in the test facility is equivalent to about 50% core flow in the reactor. Before this condition has been achieved, precautions should be taken to prevent boiling transition that could result from possible flow blockage. In order to prevent boiling transition from flow blockage during the time it takes to disintegrate the shoe, the Browns Ferry 1 reactor should be operated with a maximum bundle power no higher than 0.6 MWt. Therefore, if the precautions given supra are taken, the presence of the shoe in the reactor vessel will not compromise safe reactor operation. L
I REFERENCES
- 1. Chemistry of Orcanic Compounds, by Noller, Second Edition,1957 W. B. Saunders & Company.
- 2. Radiation Damage of Materials, Enoineering Handbook, Part II, A Guide to the Use of Elastomers, by M. H. Van de Voorde.
November 28, 1966, MPS/ Int. Co. 66-27.
- 3. " Elastomers for Use in Radiation Fields", by R. Harrington from Rubber Age.
- 4. The Effect of Nuclear Radiation on Elastomeric and Plastic Components and Materials REIC RPT #21 and Addendum, BMI, Columbus, Ohio, August 1964.
- 5. NEDO-10174, " Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor", G. J. Scatena, May 1970.
- 6. HEDD-10174, Revision 1, " Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor", October 1977.
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.ATTACIDfENT 3 s
jmQuts$ienNo.1--Discursthasimilaritybatv:cnthatstloopconditionscnd cetual reactor conditions (especially pressure). Answer No.1--The test loop did not provide for parallel flow paths. Complete blockage at the orifice or the tie plate could have resulted in a AP equal to tha shutoff head of the test pump. Ilowever, the maximum AP across the test cztup was administrative 1y controlled during the test. During the flow testing ef both the 24-hour and 26.5-hour shoe samples and during the blockage tests, fitw was increased very slowly from 0 in increments of 10 gpm or less. At no time did the AP increase beyond AP's achievable in the plant. There was at n3 time any evidence of co=plete blockage in either the orifice or tie plate. If complete blockage did occur, this blockage was cleared without any significant ride in AP. For the 26.5-hour sample, a review of figure 6 of GE's safety annlysis report (included in this package as an earlier attachment) implies th:t the sample was forced through the orifice at about 20 Epm with a AP of cbout .25 psi. For the 24-hour sample (figure 7 of same report), about 50 gpm was required fer a AP of about 2 psi. Qutstion No. 2--Correlate the 0.6 MWt bundle power limit to a gross percent cerc power level. l . Antwar No. 2--A boundary peaking factor analysis was performed by CE using thtir standard methods, assumptions, and conservatisms as used in safety cntlysis ca,1culations, such as rod drops. The most limiting peaking factor was calculated for the Browns Ferry unit 1 control rod withdrawal sequence. Tha most limiting peaking factor for the sequence was then assumed to occur ct the highest power level. This analysis indicates that if gross core power does excted 0.6 MUt.not exceed 5 percent of rated. thermal power, the maximum bundle power will not _ Question No. 3--If TVA does not plan to cool down to 2000 F after the soak psriod at 500 F, quote an expert, text, or provide assurance that operation et 5000 F is conservative. Answer No. 3--TVA intuitively feels that continued operation at 500 0 F without c cooldown period is conservative. Further, the technical staff of the Goodyear Tire and Rubber Company has stated that the rubber shoe cover naterial disin-tegra'tes when it is extended beyond its curing state (about 400 F). Their
.cpinion is that it will disintegrate more co=pletely or at least as rapidly . ct 500 F than it would if cooled back down to 200 F.
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