ML19308D759
| ML19308D759 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/22/1978 |
| From: | Stewart W FLORIDA POWER CORP. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8003130759 | |
| Download: ML19308D759 (20) | |
Text
.
Op REGULATORY INFORt1ATION DISTRIBUTION SYSTEM (RIDS)
DISTRIBUTION FOR INCOt1ING t1ATERI AL 50-302 REC: REID R W ORG: STEWART W P DOCDATE: 09/22/79 NRC FL PWR DATE RCVD: 08/24/78 DOCTYPE: LETTER NOTARIZEU: YES COPIES RECEIVED
SUBJECT:
LTR 3 ENCL 40 RESPONSE TO NRC REQUEST OF 00/16/70.
FORWARDING INFO CONCERNING THE BURNABLE POISON ROD ASSEMBLY FAILURES AND THE PROPOSED CORE FOR THE REMAINDER OF CYCLE 1 FOR UNIT 8.
NOTARIZED 08/22/78.
W/ATT.
PLANT NAME: CRYSTAL RIVER #3 REVIEWER INITIAL:
XJM DISTRIBUTOR INITI AL: Q c o t- * * * * *' * *
- F * *' * *
- DISTRICUTION OF THIS MATERIAL IS AS FOLLOWS ******************
GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIDUTION CODE AOO1)
FOR ACTION:
P PMIEF orb *t4 DC**W/7 ENCL INTERNAL:
.,EO FIf W W/ ENCL NRC PDR**W/ ENCL
- W/2 ENCL OELD**LTR ONLY e
HANAUER**W/ ENCL CORE PERFORMANCE BR**W/ ENCL AD FOR SYS & PROJ**W/ ENCL ENGINEERING BR**W/ ENCL REACTOR SAFETY BR**W/ ENCL PLANT SYSTEMS BR**W/ ENCL EEB**W/ ENCL EFFLUENT TREAT SYS**W/ ENCL J.
MCGOUGHa*W/ ENCL EXTERNAL:
LPDR'S CRYSTAL RIVER, FL**W/ ENCL TERA **W/ ENCL NSIC**W/ ENCL ACRG CAT B**W/16 ENCL IGTRIBUTION:
LTR 40 ENCL 39 CONTROL NBR: N I2E: 2P+17P hYt er m * **..m e** m mmm..m mo THe END
~** m ** m ** m ****. m.**.* m
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- a. r Florida W. P. STEWART, DIR ECTOR POWER PRODUCTION August.22, 1978 9
Mr. Robert W. Reid Chief Operating Reactors Branch #4 Division of Operating Reactors U.S. Nuclear Re "latory Commission Washington, D.C. 20555
Subject:
Crystal River Unit 3 Operating License No. DPR-72 Docket No. 50-302
Dear Mr. Reid:
In your letter of August 16, 1978, you requested Florida Power Corporation to provide additional information in response to the questions contained in the enclosure of your letter. These questions concerned the burnable poison rod assembly failures and the proposed core for the remainder of Cycle 1 for Crystal River Unit 3.
Florida Power Corporation hereby submits for NRC review three (3) originals and forty (40) copies of our response to your request for additional information dated August 16, 1978.
Should you or your staff have any questions concerning our response, please do not hesitate to contact this office.
Very truly yours, FLORIDA POWER CORPORATION l
IL,0
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W. P. S tewart ECS/WPS/ hew
'ec:
File: 3-0-3-a-3 M03 (8/21)
-7e2eEcts.2 -
General Office 320 Th.ny founn street soutn. P.O. Box 14042. st. Petersburg. Florida 33733 e 813-866-5151
I s
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I STATE OF FLORIDA
. COUNTY OF PINELLAS.
W.P. Stewart states that he is the Director, Power Production, of Florida Powe'r Corperation;-that he is authorized on the part of said company to sign and file with the Nuclear Regulatory e
Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information and belief.
P.
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- W' P. 8tewart
~
Subscribed and sworn to before me, a Notary Public in and for the
- a t and County above named, this 22nd day of August, 1978.
NotaryPublg Notary Public, State of Florida at Large, My Commission Expires: July 25, 1980
RESPONSES TO NRC QUESTIONS CONCERNING CR-3 PREPARATIONS FOR CYCLE 18 STARTUP Question 1:
In your letter dated June 22, 1978, you stated that the dropped test weight came to rest against the APSR coupling device. Describe any inspections performed on the part length control rods and any damage resulting from the dropped test weight. Should any damage have resulted, provide your basis for reuse of this component in the upcoming cycle.
Answer:
The following inspections were performed to assess the structural and func-tional ability of the APSRA A008.
1.
Observed the condition of APSRA from the top and sides without lifting them out from the fuel assembly.
2.
Moved the APSRA by a few inches and measured friction (drag) force.
3.
Exercised another APSRA not involved in the incident and measures friction forces.
4.
Exercised the APSRA in another (good) fuel assembly and compared the friction force.
5.
Examined the lower surface of the coupling spider assembly of the APSRA A008 to see if any damage is done.
6.
Examined the APSRA full length from a side to see any abnormality.
The visual observations revealed that, as a result of toppled weight, the APSRA has minor material deformation (i) on the outside surface of top of the hub and (ii) on the upper plugs (made of solid stainless steel rods) of the poison rods.
These deformations will not reduce canponent structural integrity nor will it create an operational problem.
No dif ficulty was experienced in locking APSRA with the handling equipment.
The visual examination showed that the rods are straight and not bent or bowed. Friction test also revealed that the APSRA performed normally. The inspection revealed that the impact load was transferred to the fuel assem-bly containing the APSRA and distorted it.
The fuel assembly was declared unsuitable for future operation and was replaced by another fuel asseably.
9 9
Question 2:
In your earlier response dated May 16, 1978, you indicated that 397'8" of the total 403'8" of BPR was recovered and that further reseerch for debris wa.s continuing.
Please provide us with a final assessment of the total debris recovered.
This assessment should include a complete listing of the composite parts and materials, and a listing of the amounts recovered.
Should the location of the m1 recovered parts and materials be known, but not accessible, include this information in your reply.
Al so should mi s-sing parts and material exist in unknown locations, identify or estimate the quantity and location of the debris. Any new information on damage or residual effects uncovered as a result of your latest findings and evalu-ations should also be provided.
Answer:
A.
A listing of parts and materials which ejected from the core and their final disposition is as follows:
Description Quantity Material Disposition BPRA Coupling 2
304 SST One was found, intact, in the plenum (one locking ball was missing); the second was found, battered but intact, in the B-0TSG head. Both locking balls were missing.
BPRA Spider 2
304 SST One was found, intact, in the plenum (affixed to the coupling); the second was found, in pieces, in the B-0TSG.
See Note 1.
BPRA Pins 32 zirc 4 3 pins were recovered intact; others ranged in length from less than one ince to greater than 12 feet.
See Note 2.
BPRA Pin Nuts 32 zi re 4 Seven BPRA pin nuts have been positively retrieved.
The remaining 25 would likely remain attached to the upper end plug extension and are accountable as part of the BPRA pin material.
BPRA Springs 64 321 SST Springs and pieces of springs in various phases of elungation were located literally throughout the system.
Estimates indicate very nearly all spring material has been re: overed.
Answer to Question 2 (Cont'd):
Description' Ouantity flaterial Disposition
. Poison Pellets 336 ft A1 0 -
Approximately 220 feet of 23 10'6' per BC poison pellets were recov-4 pin ered intact within the con-fines of BPRA' pin segments.
The remaining material re-covered was in the fonn of decomposed Aluminum 0xide pellets vacuumed from RCS internal surfaces, fuel transfer canal and spent.
fuel pool floors. Any esti-mate of non-accounted-for
-~
material is impossible.
Notes:
1.
Numerous pieces of debris were recovered from the B-0TSG upper tube sheet; many of which were pieces -of the broken spider assembly.
These pieces, plus three pieces retrieved from core internals, constitute the vast percentage of the broken spider assembly.
2.
The original estimate of 397'8" of BPRA rod material recovered was based on a comparison of the inventory log maintained on site with the amount of material used in construction of the BPRAs.
It should be l
noted the running inventory was based on visual observation, either directly or via video cameras, and the sizes were estimated by com-parison to an iten of known dimensions sighted in proximity of the debris. The uncertainty in the size estimates are due to many factors including the irregular shapes of the tubing pieces (bent pieces and shattered ends) and the estimation of numbers of small pieces picked up by vacuuming.
Since the original report, an additional eight feet have been located, bringing the total to 2 feet more than what was lost.
Based on this infonnation, we feel there are no large pieces unrecovered.
B.
Identified material which has not been recovered fran this system j
includes:
1.
Three sections of rod tubing remaining entrapped in the lower flow distribution region of the CAS.
These range in length from six to eight inches. The location and configuration of these pieces preclude the possibility of their exiting the CSA.
2.
Severely lodged debris has dictated the removal of seven tubes in the B-0TSG from service.
Inaccuracy of ' measurement eliminates the ability to ascertain the amount of material actually remain-ing in the B-0TSG. A very conservative estimate based on the time required for the eddy current probe to come in contact with the debris from the bottom in ' conjunction with a direct measure--
- ' ment from the top indicates as much as 10'6" of material may be-lodged in the plugged tubes..-
I Answer to Question 2 (Cont'd):
'C.
Inaccessibile areas where debris may be located are listed below.
1.
Thermal shield orifice holes and cavity: analysis perfomed assuming total blockage of the orifice holes indicates no detri-mental effects during system operation.
2.
Incore detector tubes:
incore detectors have been inserted, effectively prefoming a free path verification of the detector tubes.
3.
Core flocding and high pressure injection lines: flow through these lines is in one direction from smaller piping to larger; therefore, debris in these lines should not affect system opera-tion.
i Question 3 (Part 1) and Question 5:
Qualify your statement that the potential for propagation of fuel failures due to blockage is extremely remote.
In regarti to your response on blockage at the spacer grids, discuss the effects of your 5% reduction in DNBR and the potential for local perfora-tion of the fuel cladding.
Also explain how the increased turbulence behind a blockage offsets the loss of flow to preclude local perforations or reductions in DER.
Describe how the blockage is modeled in LYNX 1/ LYNX 2 and what correlation with DNB and blockage was used.
Answer:
The effects of debris trapped in the lower end fitting and intemediate j
spacer grids have been evaluated using the cross flow codes LYNX 1/ LYNX 2.
1 The results demonstrated that a blockage of 20% of the entire inlet flow area of a fuel assembly decreases the MDNBR by less than 0.1%.
A blockage this large is extremely unlikely since it would require several large pieces of debris.to be lodged in the same fuel assembly.
Since blockages near the assemly lower end fitting have a negligible effect on DNBR, the description of models and results presented here will be primarily con-cerned with analyses of debris trapped in the intermediate spacer grids.
)
A core internal blockage might be produced by loose parts fracturing upon impact against the lower ~ support structures and the smaller pieces slipping through the end grid. The longest metal part which could be passed through the spacer grids would be 0.14" wide and could pass through Area "A" shown in Figure 1.
For the LYNX 1/ LYNX 2 analysis, it was assumed that a long strip of metal 0.14" wide passed through the lower spacer grids and lodged in the third or fourth intemediate grids blocking 2 to 7 adjacent channels (Figure 1). For design axial flux shapes, the point of MDNBR in the hot channels occurs just above the fourth intemediate grid; hence, the assump-tion of a blockage in the third or fourth intemediate grid resulted in the largest change in MDER. A 0.14" wide strip of debris situated across-adjacent channels, as shown in Figure-1, is capable of blocking a maximum of 44.1% of theisub-chahiiel flow area.
This is the largest part,
L Answer to Question 3 (Part) and Question 5 (Cont'd):
which could realistically be expected to cause a blockage and the prob:-
bility of _ this part passing through the end fitting and several intermedi-
~ ate spacer grids before becoming lodged in the hottest channels in the core is very low.
LYNX 1/ LYNX 2 analyses were also'made with 75% flow area blockage of adjacent sub-channels in the same intennediate grids. The choice of 75% blockage was completely arbitrary. The only postulated means of achieveing a block-age greater than 44.1% is for a strip of metal to be folded smaller than 0.14" width during passage fram grid to grid and then to flatten when it becomes lodged in a grid.
The results of the LYNX analysas are shown tabulated in Table 1 and plotted in Figure 2.
The hot sub-channel mass flow rate as a function of axial position for 44.1% and 75% flow area blockages are shown in Figures 3 and 4 e
respectively. The hot channel MDNBR decreased approximately 2.6% for 44.1%
blockage and approximately 5% for 75% blockage at the fourth intennediate spicer grid. This calculation does not consider that turbulence intensi-ties are very high behind the blockage.
A study was perfonned to investigate the pressure, velocity, and turbulence intensity distribution near sub-channel flow blockages in a test bundle with a geometry typical of a Babcock and Wilcox fuel assembly. The experi-mental measurements were perfonned at Battelle-Northwest's 189-D Hydraulics Facility. Measurements of axial velocities, turbulent intensities, and axial and radial pressure gradients were made for five different blockage configurations. These configurations were (1) no blockage, (2) a 90% plate type blockage, (3) a 75% plate type blockage, (4) a 90% sleeve type block-age, and (5) a 50% sleeve type blockage. One set of measurements was made at a bundle average velocity of fifteen feet per second for all configura-tions except the 90% sleeve configuration. For the 90% sleeve type block-age, two sets of data were taken at fifteen feet per second and 1 set each at velocities of ten and twenty feet per second to check for any velocity effect and to verify reproducibility.
The water temperature was held at approximately 85'F for all of the measurements. The test bundle consisted of a 5 x 5 rod array with three spac Each of twenty-four rods were~ equipped with six static pressu_er_ grids.
re taps located in the same radial plane but at six different vertical levels.
Axial velocity and turbulent intensity measurements were made at the same six axial planes using a' Laser-Doppler velocimeter.
This device measured the instantaneous local velocities by measuring the shift in frequency of light scattered fram colloidal particles added to the water.
Axial and lateral pressure differences were measured using pressure trans-ducers and an automatic data acquisition _ system.which recorded the pressure
, differences on magnetic tape. Measurements were made at two second inter-vals over a long enough time period to ensure that the averages of the data taken over the time period were reproducible.
l l l L
' Answer to Question 3 (Part 1) and Question 5 (Cont'd):
The.results = of the test program indicated turbulence intensities five times greater than nomal for the area just behind flow blockages of type modeled in the LYNX analyses. The degradation in DNBR is only 2.6% for anticipated
~
blockages in the worst core locations as predicted by the LYNX cross flow codes without consideration of increased turbulence downstream of the blockage. The increased turbulence should greatly offset or completely i
nullify the slight reduction in the hot channel MDNBR. The LYNX aaalysis
-indicated that the axial location of MDNBR changed a maximum of three inches for all cases analyzed relative to the base case with no flow block-age. The DNBR increased in channels adjacent to the blocked channels due to the increased. flow forced into these channels.
This increases the DNB margin in the adjacent channels, thus the potential for propagation of fuel failures due to the blockage is considered to be extremely remote.
Question 3 (Part 2):
g Explain how the daily monitoring of the water chemistry provides a means of detecting fuel failure.
In this regard, specify the sensitivity of this method to detect fuel failures and any time lags involved between fuel failures and the detection of fuel failures.
Answer:
The reactor coolant is monitored continuously with a radioactivity monitor located in the RCS letdown line. This instrumentation has a sensitivity capable of detecting 0.1% fuel failure. This instrumentation is supported by laboratory gamma spectrum analyses three times per week. The laboratory analyses have a sensitivity capable of detecting 0.01% fuel failure. With the radiation monitor on line, there is no appreciable time lag between fuel failure and the detection of fuel failure; the laboratory analysis could lag the fuel failure by as much as three days.
Question 4:
Describe any startup and operational tests and surveillance procedures planned to assure detection of degraded operations of the reactor coolant j
pumps, RCP seals, full length control rods, and part length control rods.
Answer:
A.
Reactor Coolant Pumps j
The flow characteristics of the reactor coolant pumps will be verified by Perfomance Test Procedure PT-101, "RC Flow and Flow Coastdown Test," prior to reactor criticality.
This procedure establishes total t
core flow, loop flow mismatch, and flow coastdown (from a simultaneous trip of all four pumps) values for comparison to establish acceptance criteria.
i i
Answer to Question 4 (Cont'd):
The mechanical condition of each pap will be monitored wha conduc-ting Perfomace Test Procedure PT-102, "RC System Noise Monitoring,"
prior to reactor criticality and at approximately 25% power level in-crements following criticality. Through the use of permanently installed vibration instrumentation, frequency domain vibration signa-tures will be gathered for each pump.
In addition, proximity probes, located at the ptanp to driver couplings, will be used to monitor shaft movement.
Infomation is displayed on the Loose Parts Monitoring panel in the control roam and alams will sound if threshold virbation levels are exceeded.
During nomal plant operation, continuous vibration monitoring of the reactor coolant pumps is provided by the loose parts monitoring system.
Autmatic alams trip when threshold values are exceeded.
A requirement for daily functional checks of the vibration instrument systen is planned for a future procedure revision.
B.'
Reactor Coolant Pump Seals The integrity of the reactor coolant pump seal packages is verified by visual inspection through Surveillance Procedure SP-204, " Class 1 System Leakage Test for Inservice Inspection," prior to reactor criti-cal ity.
Subsequent perfomance will be detennined by monitoring seal package staging pressures and seal water leakoff temperatures.
C.
Full and Partial Length Control Rods Perfonnance of full and partial length control rods will be detenined by Surveillance Procedure SP-102, " Control Rod Drop Time Tests." Con-trol rod programing will be verified by SP-401, " Control Rod Program-ming Verification."
Both of these surveillances will be completed prior to reactor criticality.
During power escalation testing and normal plant opration, the incore detectors will be used to detemine core power distribution which is an indirect indicator of control rod positioning and poison pin integ-rity.
In addition, routine checks of individual control rod position indication versus group indication provides direct indication of potential positioning errors or rod binding.
Question 6:
In your. reply, you stated that there is a probability that small pieces might get into the guide tubes and cause some interaction with moving com-ponents.
Provide your analysis of this condition and quantify the proba-bility of this event.
Answer:
The flow opening at the lower end of the guide tubes is 1/8" in diameter.
Hence, only a piece less than 1/8" in size can enter f uel assembly guide tube from the bottom. Nominal diametral clearance between a control rod and a. guide tube is 0.058 inches. -
Answer to Question 6 (Cont'd)-
Major components of the primary system have been flushed and cleaned of debris. The probability of 0.125 inch or smaller size particle going inside a guide tube and cause any problem is remote. The area of the holes in the 16 guide tubes (0.2 in2) is very small compared to the total cross sectional area of the fuel assembly (73.2 in2),
i Question 7:
During the previous cycle, two fuel assemblies operated for a significant period without burnable poison. Therefore, the exposure and isotopic inventories in these and a jacent assemblies will be different fran that calculated using the exposure tallies derived from the incore detectors.
Describe in detail how this discrepancy was a.ccounted for in the calcula-tions summarized in Section 3, " Nuclear Design."
l Answer:
The discrepancy in the exposure calculation derived from the incore detec-
)
tors for the two fuel assemblies from which the burnable poison clusters were uncoupled is very small (estimated to be no greater than 0.4%).
This is because the detector signals did reflect the increase in power due to the LBP uncoupling. The source of the discrepancy was a misapplication of a spectrum factor used to account for an LBP that was not present.
The burnup differences due to the LBP uncoupling are small and localized.
This magnitude of burnup difference has an insignificant impact on the core physics parameters calculated and therefore need not be considered in the calculation. This will be confinned during startup testing. Furthermore, any increase in power peaking caused by the small burnup difference (i.e.,
Quadrant Tilt) is accounted for in the Quadrant Tilt Tech Spec limits.
Question 8:
The analyses in Section 3, " Nuclear Design," appeared to be based upon octant symmetry.
Yet, Location L-12 and its octant analogue are only quarter-core symmetric. Moreover, one would expect the use of assemblies
~~
from which the BPRA were ejected to introduce still more asymmetry.-There --
fore, justify in detail the use of octant symmetry in the nuclear calcula-tions.
In addition, explain quantitatively how the incore monitoring pro-gram (which apparently also assumes octant symmetry) will account for any asymmetry.
Answer:
The analyses presented in Section 3 and the resulting Technical Specifica-tion changes listed in Section 6 were based on quarter-core sy::rnetry. The different radial and axial burnup distributions in locations L-12 and N-10 were input to the various PDQ07 and FLAME 3 calculations of the modified Cycle 1.
The reason for showing only an octant radial power distribution
_in Figure 3-1 was that to two decimal places the symmetric location radial powers were identical. Due to the differences in the axial burnup shapes.
j Answer to Question 8 (Cont'd):
between L-12 and N-10, differences (up to 6%) were observed in the total peaking factors in those and other nearby symmetric locations. These dif-ferences are accounted for in the calculated power distributions supplied for startup testing which cover the entire quarter core.
The incore monitoring layout for location L-12 is quarter core symmetric in that two detector strings are used (one in D-10 and the other in F-12) sym-metric to locations L-12 and N-10.
Therefore, the measured power distribu-
- t. ion will monitor the two locations correctly.
Question 9:
Describe the procedure to be followed in the event that the acceptance cri-teria of Section 7.2.3 are not met.
In particular, would Bank 4 be measured by deboraton? If not, how would the 10% uncertainty in the shut-down margin calculations be justified?
Answer:
If acceptance criterion is not met with Groups 5, 6 and 7 fully inserted, then Group 4 will be measured by deboration and the icceptance criterion applied to the sum of Groups 4, 5, 6 and 7.
If accer.tance criteria for any test is not met, an evaluation is perfonned before the test program is con-tinued. The plant is not escalated in power until evaluction shows that plant safety will not be compromised by escalation.
Question 10:
In the test of ejected rod worth (Section 7.2.4), will all four symmetric rods be measured? If not, will any check on azimuthal tilt be made before power escalation?
Answer:
All four symmetric control rod worths will be measured according to the following plan.
The most reactive control rod (predicted) will be measured first by boration. Then, this rod will be remeasured by rod swap with Group 6.
The remaining thrsa symetric rods will be individually measured by rod swap with Group 6.
Question 11:
State your schedule for submittal of a startup report to the NRC.
Answer:
It is planned to submit the results of the control rod drop test and the RC flow and flow coastdown test to the NRC as soon as the associated analyses are completed.
The startup report will be submitted within 90 days of ccmpletion of l
testing in accordance with Technical Specification 6.9.1.3.
-g-
~ _ _
Question 12:
Please provide a sumary of occupational exposures actually received during the OTSG repairs in the fann of Tables 2 and 3 of your May 12, 1978, submittal.
Include in this sumary the number-of workers who received the exposure.
Answer:
Attached are Table 2 and 3 which have been revised to include the requested exposure information.
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10-
TABLE 1 102% Power: MDNBR Results
. Axial BAW-2 Position Run Description WNBRI (In.)
tio flow area blockage 2.4755 93.418 Fourth Intermediate Spacer Grid 44.1% flow area blockage of 7 channels 2.4109 93.418 75.0% flow area blockage of 7 channels 2.3327 90.405 44.1% flow area blockage of 2 channels 2.4123 90.405 75.0% flow area blockage of 2 channels 2.3489 90.405 Third Intennediate Spacer Grid 75.0% flow area blockage of 2 channels.
2.4065 87.391 1 Limiting Channel - Channel 172 Hot Unit Channel e
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Table 2 Work Items and Exposures Inside B OTSG Upper Head -
Estimated Actual With
- of Workers Without With Tubesheet Tubesheet and Actual Receiving Wtrk Items Shielding Shielding Dome Shielding Exposure Exposure Install Shielding Tubesheet N/A 1.2 1.2
}
4.7 18 Dome N/A N/A 10 j
Repair Identified Tubes (includes rework and QA) 140 80 40 90.1 96 Clear and Eddy-Current Obstructed Tubes 14 8
4**
Rtmove Shielding Tubesheet N/A 1.2 0.8 Flush tubes in B OISG 0.6 N/A 0.4 Install Shielding Tubesheet N/A N/A*
0.8 Explosive Plug or Stabilize Obstructed or Damaged Tubes 0.6 0.6 0.2 1.11 20 Rsmove Shielding Dome N/A N/A 10 0.9 4
Tubesheet N/A N/A*
1.2 Total 155.2 91.0 68.6 96.8 138
- Tubesheet shielding is reinstalled to lower exposure required for dome shielding removal. Exposures for this installation / removal are not needed if dome shielding is not used.
- This work item performed from the bottom of the steam generator.
Table 3
)
OTSG Repair Tasks
]
Estimated
~ B SG Upper Tubesheet Work 66.2 96.8 138 Install / Remove Eddy Current Manipulator, Bottom Head of Both SG's 10.6 5.7 10 SG Lower Head Inspection, Remove J-Leg Screens and Bladders 1.5 6.0 24 DIbr1s Removal, Tube Plugging in Lower Head 17.0 26.5 38 Support Outside SG 14.4 Subtotal 109.7 135.0 210 15% Contingency 16.5 TOTAL 126.2 135.0 210
- This value is already accounted for in the other work items.
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