ML20079Q912
| ML20079Q912 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 06/15/1983 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20079Q911 | List: |
| References | |
| NUDOCS 8306210097 | |
| Download: ML20079Q912 (19) | |
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M i
i DUKE POWER COMPANY McGUIRE NUCLEAR STATION
SUMMARY
REPORT OF THE RADIOLOGICAL ASPECTS OF
-THE MODIFICATION OF i McGUIRE NUCLEAR STATION UNIT 1 STEAM GENERATORS i
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, . TABLE OF CONTENTS 5
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.: Introduction a) History and need for modification.
b);_~T Description of Review Process by Design Review Panel 4- c)' -Scope of Review d) Initial Radiological Considerations (Westinghouse) i II. McGuire History ta) -Startup b) Power History III. Radiological Aspects a) Initial Estimates
. b) - As Found" Data Shielding c) d)- ' Exposure Incurred I V .- Conclusions ATTACHMENTS:
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-Table III-1, Breakdown of Exposure Table III-2, Exposure Graphs by Day
- a. Staging
- b. SG-A
- c. SG-B
- d. SG-C'
- e. SG-D l . . , - - . _ _ -___ __ , . . . - . _ ~ . . . _ . . _ _ _ . _ _ _ . - - _ _. _ ._-
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, . I. INTRODUCTION 1
, . a) HISTORY AND NEED FOR PREHEATER MODIFICATION
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On October 21, -1981, Ringhals Unit 3,! a three loop 1 e Westinghouse plant with Model D3 sp. lit flow st' eam generators,
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<> was shut down due to a steam generator tube leak. of 2.5 gom.
+ Upon investigation, tube R49 C55 (cold leg) had worn a small
. through-wall hole at support plate 3. .The unit had operated forabout2200hoursatp'owerlevelsabove7CNat:hetimeof the leak. It was apparent that some type ,0f' acceldrated wear mechanism irvolving interaction between the support plates and tubes was occurring. Eddy current testing (ECT)' performed on
, all three steam generators indi,cated that preferential wear was occurring in the outer three rows of tubes in the preheater section (Rows 47, 48' and 49). Several tubes were
, removed from the affected steam generator to better characterize the wear phenomena.
Almaraz Unit 1 and McGuire Unit 1 bcth have Model D split flow steam generators and were operating at the time of the Ringhals 3 occurrence. Both units were shut down and ECT inspections were performed. Almaraz 1 had beehrat power levels of 75% or greater for about 1500~ hours and had similar indications in its Model D3 steam generators to those observed at Ringhals 3. McGuire 1, which was in the startup ' test program and had not been above 50% power,.showed no signs of
.. tube wear in its Model D2 steam generators. Operation of all three of these units continued over the next year with i
restrictions on power level.
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a- Westinghouse responded to the problem by , establishing a task 1 force to identify and correct the cause of th'e problem. This h
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involved gathering information relative to ECT measurements from operating plants, pulled tube data, behavioral information from analytical models and information from a series of air and water scale model test facilities.
Monitoring instrumentation was installed external to and inside several steam generators to detect and measure tube vibration. The information from all these sources was then used for the following purposes:
- 1. Map fluid velocities and turbulence in the preheater inlet plenum,
- 2. Determine resultant motion of the steam generator tubes, and,
- 3. Correlate feedwater flow to tube motion and wear.
Integral with the above investigations, Westinghouse developed several conceptual modifications to reduce tube vibration and the resultant wear. Many designs were conceived; several reached the conceptual design stage and ultimately one design was selected as the " reference" design. During the preliminary testing of this reference design, an " alternate" design was developed. This alternate design was the internal manifold which was the final design used for installation in the Model 02/D3 steam generator, b) DESCRIPTION OF REVIEW PROCESS Durina the first several months after the Model D Steam Generator problen, was identified, NRC staff worked with Westinghouse and several utilities on an individual basis.
Only one unit (McGuire-1) with Model D steam generators was licensed in the United States. However, a second unit (Summer) was nearing completion of construction and 4 additional units (McGuire-2), Watts Bar 1-2, Catawba-1) were scheduled. In an effort to conserve resources and to minimize the time for NRC review and approval of the proposed design i
changes, the concept of a third party design review of the proposed modification was initiated. Such a review was intended to lessen the need for a detailed technical review by the NRC. The third party's final report was proposed to serve as the basis for the NRC Safety Evaluation Report.
Responding to the identified needs for a thorough and coordinated review and for conservation of resources, Tennessee' Valley Authority, South Carolina Electric and Gas Co., and Duke Power Company agreed to pool resources to form a Design Review Panel to examine all aspects of the final Westinghouse design for the Model D2/D3 modification. A review plan was developed to serve as guidance for the Design Review Panel (DRP).
c) SCOPE OF REVIEW The plan of ~ the DRP was to review the final modification design selected by Westinghouse for installation in the steam generators, and to determine whether or not it was acceptable, and to report on conclusions reached.
In performing the review, many different areas were examined in detail. These included: thermal hydraulics, model testing, ALARA, structural mechanics, stress analysis, quality assurance, inservice inspection, materials, tooling, welding, chemistry control, and installation. All of which were addressed by Westinghouse, as appropriate.
The NRC Staff was in attendance.at the two full Design Review Panel meetings. At the conclusion of each of these meetings, staff members had an opportunity ~ to make comments. All comments made by the NRC in the DRP meetings were addressed during the review and. resolved to the satisfaction of the DRP.
6-d) ' INITIAL RADIOLOGICAL CONSIDERATIONS - WESTINGHOUSE The design modification to the preheater section of the Westinghouse D2/D3 steam generators presented radiological considerations that are in general les's limiting than those normally associated with steam generator work. The primary reason is that this work was _ restricted to the secondary side of the generators and consequently did not involve the intense beta or gamma fields associated with primary side modifications. In addition, the modification was to be made on plants with limited operating history and thus limited primary and secondary side contamination.
The primary source of radiation exposure o'id come from radioactive corrosion products deposited on the inside, i.e.,
primary side of the steam generator U-tubes. Exposures due to contamination and inhalation turned out to be insignificant (negligible).
Westinghouse used the CORA computer code to determine surface activities and the KAP-VI. shielding code to determine expected radiation fields in the work area around the feedwater nozzle.
Westinghouse estimates of 1 rem / hour at the surface of the tube bundle falling off to 0.015 rem / hour on the work platform were consistent with levels measured at Duke Power Company's McGuire 1 nuclear generating unit utilizing data from previous history. These levels were considered reliable for the purpose of estimating dose rates likely to be encountered during the modification.
Based on their extensive background in making steam generator repairs, Westinghouse prepared detailed man-hour / exposure rate studies by subtask. The resulting total dose of approximately 25 person-rem per steam generator appearea to be a reasonable estimate of expected dose for this modification. This estimate was based on actual timing data derived in mock-up
7 training. Tasks that revealed a significant increase in estimated dose were reviewed with the intent of improving procedures in order to decrease the subtask dose. The dose ccmitment was estimated to-be divided such that 16 person-rem per generator would be due to actual internal . generator modifications and 9 person-rem per generator due to ancillary preparation and restoration work.
Westinghouse estimated that post installation eddy-current and fiber optic inspection would contribute approximately 9 person-rem per generator per inspection.
The quantities of potentially contaminated waste generated in this modification were minute compared to those encountered in normal outage work. There was a special effort to exclude oil and to rely on water base lubricants. The modification did not adversely impact the normal waste- handling facilities at McGuire Nuclear Station nor cause the need for any increased capabilities.
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II. MCGUIRE HISTORY a) STARTUP McGuire Nuclear Station, Unit 1, started up (initial criticality) on August 8,1981. The startup testing was in progress when the accelerated wear problem was discovered at Ringhals 3. .The testing program was continued up thru 90%
full power with a brief run at 100%. Testing to this point was completed in November 1981 and the unit was brought down for Steam Generator Inspection on November 21, 1981.
b) POWER HISTORY Following the initial inspection the unit was returned to service with restrictions on run time above 50%. Additional inspections were performed in March, June, and November of 1982 prior to the eventual shutdown for modification. During this period total operating time (191.48 EFPD) only 1524 hours0.0176 days <br />0.423 hours <br />0.00252 weeks <br />5.79882e-4 months <br /> (63.5 days)werespentabove50%.
The Unit was shutdown on January 21, 1983. Critically, post outage, was achieved on May 7, 1983 0 1740 hours0.0201 days <br />0.483 hours <br />0.00288 weeks <br />6.6207e-4 months <br />.
III. RADIOLOGICAL ASPECTS a) INITIAL ESTIMATES In preparation for the modification project pre-job estimates of radiation exposure were made in keeping with good ALARA practice. The estimates were made using data recorded from past outages and Westinghouse / Duke Power Construction estimates of manpower and time requirements. The McGuire Health Physics Section estimates differed slightly from the Westinghouse estimates due to different definitions of-ancillary work.
McGuire Unit #1 shutdown for maintenance on January 21st to perform the modification on the feedwater inlet nozzle, perform eddy current testing, remove thermal sleeves from the primary system and make modifications to the safety injection system. It was determined in an earlier eddy current outage, that access to the feedwater nozzle would be difficult due to the design of the ice condenser containment building and the large amount of equipment necessary to support the modification. Photographs of the feedwater area indicated several pipe. hangers, rupture restraints, cable and cable trays and ventilation ducts would require removal before any modification work could begin.
In addition, scaffolds on which to stage the special equipment had to be erected in ' areas where access was already very difficult. Scaffolds were erected on top of each reactor cc,olant pump, each steam generator lateral support and under each feedwater nozzle. As a result of the amount of site preparation necessary, an RWP was written to include equipment setup and removal, scaffold erection and disassembly, and interference removal and replacement. A total of 24
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person-rem was estimated to prepare and recover from the modification utilizing this work permit. This estimate covered ancillary work on all generators since most of the tasks would not be attributed to a single steam generator.
An estimate of 21.795 person-rem was calculated for each generator based upon the assumption of a 1R/ hour contact on the tubes, a 500 mR/ hour general area inside the nozzle,100 mR/ hour outside the nozzle at l', and 15 mR/ hour general area of platform.
An additional 17,810 person-rem was budgeted for followup eddy-current testing based on past testing and the use of full templates instead of the partial templates designed for nozzle area ECT. This estimate was higher than Westinghouse's, due to initial plans for a full In-Service Inspection Eddy Current Test on all generators, subsequently this was done only on two (2) generators, b) AS FOUND DATA Radiation levels remained as estimated during the removal of the existing flow venturi and support ring. Radiation levels, however, were actually found to be much higher (3.2 R/ hour) on contact with the tubes during a survey taken through the sludge lance ports and later confirmed after the impingment plate was removed.
Due to the higher than expected radiation levels, it became necessary to train Duke Power personnel to perform the installation of the catch basin, shroud and manifold bolts in order to limit the exposure of Westinghouse engineers so that they could complete the modification.
Mock up training for small-framed individuals was conducted allowing volunteers from various departments to become proficient in each task. Each' was trained numerous times
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until each individual was confident that the task could be accomplished with little or no problem. While Westinghouse provided the training, the modification continued without delay.
c) SHIELDING With the "as found" radiation levels at the tube bundle surface exceeding the estimate by a factor of three (3),
additional work was- needed on the person-rem estimates. A shielding project (earlier thought to be unnecessary due to the high cost and - small reduction involved) was undertaken.
Chain was draped from 'I' beams over the face of the steam generators and a series of overlaping lead blankets attached.
This reduced the dose-rate on the platforms from 15 mR/hr to 5
- 7 mR/hr.
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c A comparison of exact tasks and the physical locations
-(in-nozzle /out-nozzle)- was made using the revised dose-rates.
Based on the new data the decision was made to let . the I~
existing projections stand.
i d)'- EXPOSURES INCURRED i
The actual exposures incurred in the project compared very
- closely.to the _ estimates in most cases. Table III-1 shows a
- break down by steam generator of totals. Table III-2 shows.an exposure per day breakdown.
The ancillary preparation and restoration work resulted in 11.485 person-rem, . an average of 2.87 person-rem per steam generator. This was considerably lower than - the estimate.
The difference was attributed to the use of a different access I~
point than planned and most' of the shielding being in place during the restoration activity. Most of the access for preparatory ' work was through lower containment resulting in
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exposure as anticipated. For the removal work the upper containment was used for access resulting in significantly lower exposures, due to easier access and lower ambiant radiation levels.
The modification work itself contributed 84.45 person-rem, an average of 21.112 person-rem per steam generator. Exposures for the first two were slightly above the initial estimates because of first-of-a-kind problems experienced. As the problems were overcome and experience was gained, exposure came back in line. This resulted in exposures for the last two steam generators slightly less than the estimate. See Table-III-1.
The post modification eddy current testing went very smoothly.
The resulting exposure, 10.4 person-rem was below the estimate. This was due in part to experience at the job and in part to lower dose rates that resulted from Co-58 decay.
See Table III-1.
r IV. CONCLUSIONS This first of a kind project, conducted on McGuire Unit 1, was a success from a Health Physics /ALARA standpoint. An evaluation of pocket dosimeter totals (106 person-rem) compared to the estimates (129 person-rem) leads us to the conclusion that the project was.
It should be noted that historic data shows that TLD exposures are 85% of pocket dosimeter values. This gives us an estimated 90 person-rem as the actual dose contribution of this project. !
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TABLE III-1
-Radiation Exposure (person - Rem)
TASK ESTIMATED ACTUAL Preparatory & 24.040 11.485 Restoration Work SG-A Modificat' ion 21.795 20.505 SG-B Modification 21.795 23.120 SG-C Modification 21.795 22.735 SG-D Modification 21.795 18.090
.SG-A ECT (Post Mod) 4.660 2.735 SG-B ECT (Post Mod) -4.250 2.090
) SG-C_ECT(PostMod) 4.660 2.355 SG-D ECT (Post Mod) 4.250 3.220 l' TOTAL 129.040 106.335 L
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