ML20073D445
| ML20073D445 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/04/1983 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073D420 | List: |
| References | |
| NUDOCS 8304130494 | |
| Download: ML20073D445 (3) | |
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7590-01 s
1 UNITED STATES NUCLEAR REGULATORY COMMISSION j
DOCKET NOS. 50-266 AND 50-301 WISCONSIN ELECTRIC POWER COMPANY r
NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U. S. Nuclear Regulatory Commission (the Commission) has, pursuant to the Initial Decision of its Atomic Safety and Licensing Board (ASLB) dated February 4, 1983, (ASLBP No. 81-464-05 LA) issued s
Amendment Nos. 71 and 76 to Facility Operating License Nos. OPR-24, and DPR-27 issued to Wisconsin Electric Power Company (the licensee),
which revised Technical Specifications (TS) for operation of Point Beach Nuclear Plant Unit Nos.1 and 2 (the facilities) located in the Town of Two Creeks, Manitowoc County, Wisconsin.
The amendments are effective as of the date of issuance.
The amendments to the TS allow repair of degraded steam generator tubes by sleeving which would otherwise be required to be plugged and removed from service; establish limits for primary coolant iodine con-centration and surveillance frequency; and establish a plugging limit for sleeved tubes of 40% nominal sleeve wall thickness.
The Initial Decision is subject to review by an Atomic Safety and Licensing Appeal Board prior to its becoming final. Any decision or action taken by an Atomic Safety and Licensing Appeal Board in con-nection with the Initial Decision may be reviewed by the Commission.
Y The amendments comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments.
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7590-01 l Notice of Proposed Issuance of Amendment to Facility Operating License in connection with this action was published in the Federal Register on August 7,1981 (46FR 40359). A Petition to Intervene was filed on July 20, 1981 as amended by letter dated August 31, 1931 by Wisconsin's Environmental Decade. Hearings were held.in Milwaukee, Wisconsin on November 17 and 18, 1982 with limited appearances held in the town of Two Rivers, Wiscorsin on the evening of November 17, 1982. The Board issued its Initial Decision on February 4,1983 and ruled that the NRC staff was authorized to issue the amendments.
The Commission has determined that the issuance of the amendments will not result in any significant environmental impact and that pursuant to 10 CFR 3 1.5(d)(4) an envircnmental impact statement or 5
negative declaration and environmental impact appraisal need not be prepared in connection with issuance of the amendments.
j For further details with respect to this action, see (1) the i
l application for amendments dated July 2,1981 as amended March 9,1983, (2) the Initial Decision of the Atomic Safety and Licensing Board dated February 4,1983,(3) Amendment Nos. 71 and 76 to Facility Operating Licenses No. DPR-24 and DPR-27, and (4) the Commission's letter to the l
licensee dated April 4,1983 All of these items are available for public inspection at the Commission's Public Document Room,1717 H Street, N. W.,
Washington, D. C.
20555, and at the. Joseph Mann Library, 1516 16th Street i
Two Rivers, Wisconsin 54241. ' A copy of items (2) (3) and -(4) may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, i
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[,
7590-01 Washington, D. C.
20555, Attention: Director, Division of Licensing.
Dated at Bethesda, Maryland, this 4th day of April,1983.
i FOR THE NUCLEAR REGULATORY COMMISSION-
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Ne Robert A. Clark, Chief
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Operating Reactors Branch #3 Division of Licensing i
o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Adges:
Peter B. Bloch, Chairman Dr. Jerry R. Kline Dr. Hugh C. Paxton In the Matter of Docket Nos. 50-266-OLA 50-301-OLA WISCONSIN ELECTRIC POWER COMPANY ASLSP No. 81-464-05 LA (Point Beach Nuclear Plant, Units 1 & 2)
February 4,1983 MEMORANDUM AND ORDER (Initial Decision)
This decision concerns the adequacy of eddy current testing to detect potentially serious defects in corroded steam generator tubes that have been repaired by the insertica of a liner or " sleeve."1 The
" sleeve" is designed to lend structural strength to the tube by spanning its corroded area.2 We have fourid limits in the capability of the eddy current test to detect flaws in steam generator tubes.
However, we have concluded that-1 This is the only issue remaining in the proceeding because we-granted sumary disposition of the rest.
LBP-82-88, 15 NRC (October 1,1982)(Sumary Disposition).
2 On July 2,1981, -Wisconsin Electric. Power Company (applicant) filed
.a Technical Specification Change Request, seeking to amend the Point Beach Operating licenses to pemit repair of steam generator tubes that have degradation exceeding 40% of the nominal tubewall thickness.
The i
existing plant Technical Specifications require that' such tubes be removed from service by
" plugging."
The proposed Technical i
Specification change would pemit repair of such tubes by " sleeving,"
leaving the tubes in service, l
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Initial Decision:
2 these limits of eddy current testing do not seriously detract from its ability to detect flaws that are likely.to rupture, either under norinal operating conditions or postulated accident conditions.
Furthermore, sleeved tubes appear to be safer than other unsleeved tubes that appli cant already is licensed to operate.
We also have concluded, based on an analysis of various factors affecting the safety of sleeves, that sleeved tubes are safe, without reference to whether they are safer than unsleeved tubes. Consequently, the license amendment should be granted, without any conditions attached at the direction of the Atomic Safety and Licensing Board.
I.
DESCRIPTION OF SLEEVING In order to understand the nature of the problem that gave rise to the issues in this case it is useful to describe briefly the functions of a steam generator in a nuclear power plant.3 All pressurized water nuclear power plants, including the Point Beach units, have two systems ofpipingtoeffectthetransferofenerhfromthereactorcoretothe turbines which produce electricity.
The primary system pumps circulate primary coolant water around the hot fuel rods within the reactor core where the nuclear reaction takes place.
The super-heated water then passes through large pipes to the steam generators.
In each steam generator -- heat exchangers approximately 70 feet high and fourteen 3
The general description of the role of a steam generator is..taken l
from Florida Power & Light Comoany (Turkey Point Nuclear Generating.
l Units Nos. 3 and 4), ALAB-660, 14 NRC 987 (1981) at 992.
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Initial Decision:
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feet in diameter -- the primary coolant water passes from large pipes into about 3000 smaller tubes which are partially imersed in a separate system of water, the secondary coolant. Heat is transferred through the i
tube walls from the primary coolant to the secondary coolant, which boils and, in the form of steam, passes' through turbines to generate electricity.
In order to prevent feaks of primary coolant and radioac-i tivity from the primary system to the secondary coolant, it is necessary to assure the integrity of the entire piping system, including each of the thousands of small tubes inside each steam generator.
l At Point Beach, steam generator tubes have experienced substantial thinning and corrosion, caused initially by the use of a phosphate I
chemistry regime in the secondary side water but continuing to some degree even after the secondary side chemistry was changed to an "all volatile" chemistry regime.
As a result, applicant sought to repair these degraded steam generator tubes and, on July 2, 1981, filed a TechnicalSpecificationChangeRequest,seNingtoamenuthePointBeach operating licenses to permit repair of steam generator tubes that have suffered from corrosion. Without the amendment, applicant would have to remove from service (by plugging both ends of the tube) all tubes that have been degraded by more than 40% of their design (or " nominal")
tubewall thickness.
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4 Initial Decision:
4 The repair consists of the insertion of a liner or " sleeve" into the degraded tube, spanning the area where the corrosion has occurred.
Then the sleeve is joined at. its top and bottom to the exterior tube.4 There are two ste? generators at each of the Point Beach units.
Each steam generator contains 3260 inverted, U-shaped vertical tubes.
1 The ends of the tubes pass through and are anchored in the tubesheet.
The tubesheet is a large circular steel plate, about 22. inches thick, through which holes are drilled for the tubes.
The bottom 21 to 3 inches of the end of each tube is fastened within the bottom of the tubesheet by " rolling," i.e., the tube is mechanically expanded tightly against the walls of the tubesheet hole. The tubes are also welded at the bottom face of the tubesheet. The tubes are not fastened at the top of the tubesheet.5 The sleeving process involves the insertion of a smaller diameter, thermally treated Inconel 600 metal sleeve inside a steam generator tube so that the bottom of the sleeve is flush with the bottom of the tube.
The sleeve extends beyond the top of the tubesheet, bridging the degrad-ed portion of the tube.
The sleeve is bonded to the tube at the bottom and just below the top of the sleeve.0 l
l 4 See LBP-81-55, 14 NRC 1017 (1981) at 1019.
5 Affidavit of David K. Porter (September 28,1981) at 14 (Attachment 1 to " Licensee's [ applicant's] Motion for Authorization for Interim Operation of Unit 1 With Steam Generator Tubes Sleeved Rather Than Plugged," Septerrber 28,1981).
(PorterAffidavit.)
6 The sleeve is designed to extend beyond its upper joint so that the additional leqgth of sleeve would prevent a failure of the upper joint i
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Initial Decision:
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l II.
COP 91ENTS ON THE " STATEMENT OF INADEQUATE RECORD" Wisconsin's Environmental Decade (Decade), the sole intervenor, did
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not present any witnesses, attempting to rely on cross-examination to establish its case.
It also did not file fornal findings pursuant to the Board's request.7 Instead, it filed a five page " Statement of Inadequate Record."
That document contains a few relevant and helpful i
points, but it was a disappointment to the Board because it failed to provide us with any reasoning by which we could dispose of the litigated issue in Decade's favor.8 i
l from resulting in an unconstrained rupture.
Should the joint fail, the J
sleeve will remain within the tube, restricting the amount of water that j
can leak through the joint area.
Porter Affidavit at 15; Applicant j
Exhibit 1, 9 3.2.
7 Tr. 18767-78.
8 Decade's Statement of Inadequate Record urges the Board to conduct j
what is essentially a probabilistic risk analysis for steam generator tube burst.
Such an analysis would. assess the overall risk to public health and safety by considering both the probability of tube burst and
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the consequences of that event.
5 l
In this proceeding the Board has not undertaken such a quantitative l
analysis, using fault. trees, numerical probabilities of failure of components and numerical estimates of overall risk.
The Board
)j nevertheless considered, in its Sumary Disposition decision, what its course might be should eddy current testing prove to be inadequate for the detection of flaws in sleeved tubes.
It therefore requested the applicant and staff to address contingently the safety implications of sleeving if that finding was made.
Both did so.
We consider those i
implications in subsequent sections of this decision even though we could rest our decision solely on the demonstrated adequacy of eddy current testing.
The record therefore does reflect thorough consideration of both the likelihood of not finding - flaws and the consequences of not finding them.
Of course, we do not use the fonnat of probabilistic risk analysis, which is not ' required by Commission policy or regulations.
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Initial Decision:.
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1 Decade attempts to excuse its Statement on the grounds that it was required to work during the Christmas vacation.
However, Decade failed to request a time extension, either during the hearing or in its filing.
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Furthermore, we know that Decade is aware that it can obtain extensions l
of time limits for good cause, as it was permitted to file its Motion
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for Litigable Issues after the time originally set.
Although Decade's filing is a disappointment to us, we do not assess any sanctions against it, primarily because we " requested" the i
filing of findings but never thought it necessary to order that they be filed.
The result is that we will do our best to respond to the few j
arguments Decade has made and to analyze the validity of the case 1
presented to us in the briefs of the other parties.
We are pleased with j
briefs filed by applicant and by the Staff of the Nuclear Regulatory Commission (staff), which respond well to our requests for a reasoned j
discussion of the entire record.
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III.
ANALYSIS AND CONCLUSIONS In this section of our opinion, we discuss the contention that was i
admitted to the hearing, the applicable regulatory materials, the facts concerning the reliability of eddy current testing, and the redundant 3
protections from steam generator tube failure available at Point Beach.9 Appendix A lists our previous decisions in this proceeding.
9 To simplify our discussion, we include a list of our previous decisions in Appendix A and a brief statement 'of the qualifications of each of the witnesses in Appendix B.
We consider each of the witnesses t
Initial Decision:
7 A.
The Admitted Contention This contention, as originally submitted, was quite lengthy and was intertwined with other assertions. The contention was:
Present inspection methods [ understood to be limited to addy current testing 10] in unsleeved tubes have been shown to be inadequate to detect defects, and the complicating presence of the sleeve inside the tube will make the detection of degrada-tion, especially at the joints, even more difficult.
Over time, the detection capability will continue to degrade....
The inability to adequately detect defects that can lead to primary-to-secondary or secondary-to-primary pathways for leakage will exacerbate the problems indicated in [the other l
subissuesinthisallegedlylitigableissue.]Il However, our Sumary Disposition decision modified this contention by l
detennining that the following genuine issue was admitted to hearing:
That the license amendment should be denied or conditioned because applicant has not demonstrated that eddy current testing is adequate to detect serious stress corrosion crack-ing' or intergranular
- attack, in excess of the technical specification prohibiting more than 40 percent degradation of the sleeve wall, in sleeves that would be inserted within steam generator tubes.12 to be an expert.
10 Tr. 1237-38, 11 See Sumary Disposition 15 NRC (October 1,1982), slip op. at 10.
12 Id. at 1.
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Initial Decision:
8 This admitted genuine issue was discussed in our Suninary Disposition decision in some detail, explaining what issues of fact or opinion the Board considered unresolved.13 8.
Regulation Involved The Nuclear Regulatory Conunission (NRC) regulation covering the adequacy of eddy current testing relates generally to the design of the reactor coolant pressure boundary.
That regula tion, General Design Criterion 14, Appendix A,10 C.F.R. Part 50, requires that:
The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
In order to comply with this General Design Criterion, applicant's proposed repair proposal adheres to an industry code, the ASME [American Society of Mechanical Engineers] Boiler and Pressure Vessel Code (Code)I4.
C.
Adequacy of Eddy Current Testing In this section of our opinion, we will describe eddy ' current testing (ECT) and then evaluate its reliability for detecting leaks.15 13 M.at2,10-16.
14 Licensee Exhibit 1, 53.1.
15 We have leaned heavily on applicant's Proposed Initial Decision, 17-20, for this portion of our decision.
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Initial Decisien:
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1.
Description of Eddy Current Testing i
For ECT, a probe is inserted into the steam generator tube.
Electric current within the co'ils in the probe produces an electromag-l netic field. As the. probe is moved within the tube, an electric current M
is induced in the conductive material of the tube or sleeve.
This is the eddy current signal that is recorded and interpreteu.
Degradation in the wall of the tube or sleeve causes variations in the effective electrical' conductivity or magnetic permeability of the wall material.
These variations are measured directly by changes in the coil voltage of the eddy current probe. 10 ECT at Point Beach is performed by Westinghouse Electric Corpora-i tion, which subcontracts the reading and interpretation of the eddy current data to Zetec, Inc.17 Mr. Denton and Mr. McKee, of Zetec, offered testimony in considerable detail about ECT equipment, the physics of the ECT process, the interpretation of eddy current signals, and the capabilities of ECT for detecting, in the field, stress corro-
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sion cracking (SCC) and intergranular attack (IGA) in tubes and sleeves.10 16 " Licensee's [ Applicant's] Testimony of W.D. Fletcher" (Fletcher),
ff. Tr.
1422, at 3-4; Tr. 1462-64, testimony of Clyde J.
Denton (Denton).
17 Tr. 1460-61 (Denton).
18 Tr.
1462-78 (Denton); Tr.
1608-1723 (Denton, McKee); Applican't Exhibits 2 and 3.
IGA is corrosion of the metal grain boundaries of the tube material' that does not initially result in separation of the metal grains.
SCC entails distinct separation of the metal grains resulting from corrosion. Tr.1427-31 (Fletcher).
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Initial Decision:
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The eddy current signals for each tube that is tested are recorded i
on a magnetic tape.
The tape is used to produce a strip chart which j
converts the record of electromagnetic signals into a linear graph that roughly resembles the record of an electrocardiograph.
This chart indicates the presence or absence of defect signals along the tubewall.
l If the strip chart ~ indicates that degradation may be present,19 the magnetic tape recording of the eddy current signals also is used to i
generate a picture on an oscilloscope.
That moving picture is recorded i
in a still photograph that enables the operator to examine phase i
differences between signals coming from the outside and inside tube j
surfaces.
That still photograph is then interpreted to determine the depth of penetration of degradation into the tubewall material.20 l
An eddy current indication of a defect in the tubewall appears as a 1
j deviation from a base line drawn along the center of the strip chart.
s The greater the volume of the defect, the greater the amplitude of the deviation from the base line.21 Unwanted signals, or " noise," also appear as deviations from the base line on the chart.
Noise is caused by such extraneous sources as conductive impurities deposited on the surface of the tube, magnetite in sludge surrounding the tube, or-the 4
19 Tr. 1658-1659.
20 Tr. 1608-11; 1473 (Denton).
21 Tr. 1611, 1620 (Denton).
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Initial Decision:
11 uneven inner surface of a structure surrounding the tube--such as the inner surface of the tubesheet hole.22 An important concept used in diagnosing potential defects is the
" signal to noise ratio."
This is the ratio of the amplitude of the signal generated by a suspected defect,to the amplitude of the noise signals found in the same general region of the strip chart.
Multifre-quency mixing. techniques are used to significantly reduce the amplitude of the noise signals.23 The amplitude of the eddy curre'nt signal is indicative of the volume of the degradation, meaning the amount of separation present in the tubewall; but the amplitude.says nothing about the depth of pene-tration into the tubewall.24 When the eddy current interpreter sees a signal which might indicate degradation, the signal is examined on the oscilloscope.25 When signal-to-noise ratios are less than about three-to-one, operators must exercise substantial judgment about whether or not a defect exists and whether the investigation should be pursued further by reading the signal on the oscilloscope.20 22 Fletcher, ff. Tr. 1422, at 4.
23 Fletcher, ff. Tr. 1422, at 4; Murphy, ff. Tr. 1828, at 8; Staff Exhibit 1, at 32.
24 Tr. 1611 (Denton); Tr. 1495-96 (Fletcher); Tr. 1672 (Denton).
25 Tr. 1473, 1610 (Denton); Tr. 1631 (McKee).
The voltage of the pattern displayed on the screen, or " voltage-lissajous," also provides a rough indication of the volume of the defect. Tr. 1657-58 (Denton).
26 Tr. 1649-50 (Denton).
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Initial Decision:
12 When a photograph of the oscilloscope picture is made, the duration,
of the exposure is sufficient to depict the two phases of the oscillo-l scope pattern that are of contarn.
A picture of the oscilloscope-pattern of a crack in a tubewall would typically appear on the scope in the shape of a flattened figure eight.27 The angle between the two significant phases of the oscilloscope picture, as measured with an electronic protractor, indicates the depth of the penetration.28 For defects of very small volume, the figure on the scope may be small, and the phase angle may be difficult to measure precisely.
In such cases, the interpreter is expected to take the most conservative reading of the angle, thus tending to overstate the depth of penetration.29 Under Bcard questioning the staff stated that they would require a tube to be plugged if the indicated depth of penetration exceeded 40%
even under circurastances where the degree of penetration was reported conservatively (i.. e..' the true penetration was likely to be less than 40%).30 l
27 Tc. 1471-73,1618-20 (Denton); Applicant Exhibit 2, at 1; Applicant Exhibit 3.
28 Tr. 1611-12, 1677 (Denton).
9 Tr 1622 (Denton).
s O Tr.1855-56 (Murphy).
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Initial Decision:
13 2.
Reliability of Eddy Current Testing The reliability with which eddy current testing detects 31 corrosion flaws depends on the volume of the flaw in the steam generator tubewall and not on the depth of penetration of the flaw into the tube. 'This detracts somewhat from the utility of the test since it is the depth of penetration which is the principal variable of interest for licensing; NRC technical specifications require that a tube be plugged when a flaw penetrates the tubewall by 40 percent or more of the wall thickness.
The volume of the flaw is, however, related indirectly to the depth of penetration.
Experience indicates that cracks propagate through the tubewall with aspect ratios having a value of about two to five.
(The aspect ratio is the ratio of the length of a crack on the outside surface to the depth of penetration.) Thus, field experience shows that cracks in tubes whicn could be of significance to NRC enforcement of its plugging limits have in most (but not all) instances adequate volume to be detected by eddy current testing.32 One expert testified that for a flaw with sufficient volume to be detected (1.
e., the signal to noise ratio is greater than about 3) a 31 The volume of the flaw is the volume separation in the tubewall or the amount of material that could hypothetically be inserted into the flaw See Tr. 1695-96.
32 Fletcher, ff. Tr.1422, p. 3, 7-8; Murphy, pp. 8, 9.
A penetration of the wall might not be detected, for example, if it has a shape analogous to a small diameter drill hole of small volume.
Tr. 1691 (Denton).
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Initial Decision:
14 50 percent wall penetration can be measured with precision (test-retest reliability) of about i 7 percent.
The precision diminishes as the crack size diminishes (1,.
e_., the error increases) so that a 30 percent through-wall crack could be measured with a precision of about i 13 percent.33 i
The likelihood of detection of a crack (as opposed to the precision with which it can be measured) is about 95 percent certainty for a 40 i
percent penetration having a 150 mil axial surface crack length.
A similar crack having only 20 percent penetration might not be detected at all.34 The limits of usefulness of eddy current testing.are known.
Eddy i
current testing using bobbin type coils cannot be used to detect cir-cumferential cracks in tubes since the lines of current flow are paral-lel to such a crack and are therefore not interrupted as they are by l
axial cracks which are oriented nonnal to the electric field.35 However, the mode of cracking generally found is axial because of hoop stresses in the tube.
In fact, circumferential cracks - have net been found at Point Beach.36 i
.The technique also cannot be relied upon at present to detect intergranular attack (IGA) which is unaccompanied by cracking.
This is 33 Tr. 1690-92 (Denton).
34 Tr. 1695 (McKee).
35 Murphy, 8, 9.
36 Fletcher, ff. Tr., p.1740.
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Initial Decision:
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because the current flow from the probe is not interrupted by IGA alone; i
j the uncracked tube material continues to act as an electrical conductor J
even though it is corroded Separation of grain boundaries through l
cracking is needed for detectability. This has proven to be of signifi-1 cance for locations within the tubesheet where enough sludge has accu-culated in the crevice between the tubes and tubesheet wall to prevent separation of grain boundaries in corroded tubes.
Tubes leakin.g within j
the tubesheet have occasionally not been found by addy current testing because of this phenomenon.37 Eddy current testing alone cannot be relied upon for diagnosis or detection of corrosion over its full range of possible occurrence.
j Physical parameters such as interference (from magnetite or copper in sludge), variations in the tube diameter, machine marks, denting in j
tubes, and small flaw volumes impose limits on detectability.38 As a practical matter this suggests that leaking tubes occasionally will not l
be detected by eddy current testing.39 t
j The instances where eddy current testing failld to detect either penetrations exceeding the plugging limit or actual leaking tubes are I
attributable to the flaws being at or below the physical limits of b
detectior..
This may occur because of snterference of the signal, the 1
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37 Murphy, ff. Tr. 1828, pp. 5,6.
38 Fletcher, p. 4-33 Fletcher, p. 6.
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Initia.1 Decision:
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small wlume of the defect or the constraining effect of sludge with'in I
the tubesheet.
The board concludes, however, that the applicant, its consultants I
and the NRC staff are familiar in detail with the inherent physical 1
l limitations of the eddy current technique for detecting stress corrosion cracking.
Applicant does not rely, for " safety, on eddy current measurements that are outside of the inherent bounds of reliability of 1
the instrument.
i The principal safety related use for eddy current testing in steam i
generators is for enforcement of NRC's 40 percent plugging limit, which is conservative because it takes into account uncertainties of measure-ment. Analyses show that unifonn thinning completely around the circum-4 farence of the tube to 62 percent degradation would not result in tube rupture following a main steam line break.
Burst tests on tubes having 40 to 60 percent through wall-penetrations confirm that burst would not occur even at pressures anticipated in a main stream line break.40 The purpose for setting plugging limits and for inspection of tubes is to prevent corrosion of tubes to progress undetected to the point where rupture is likely under either accident conditions or normal operation.41 It is particularly important to safety to have the capability for detecting relatively large volume defects (those above 40 Fletcher, ff. Tr.1422, p. 9; Murphy, pp. 3-4 41 Fletcher, p.10; Murphy, p. 3.
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Initial Decision:
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4 the plugging limit) so that tubes can be plugged before a hazardous condition arises.
Much was made at hearing about the uncertainties attendant to the j
lower limits of detection for eddy current testing, where it is beyond question that the technique does not detect every small flaw.42 While it was necessary to probe those limits, we now conclude that the limits i
of detection inherent to eddy current testing do not cause a concern that stress corrosion cracking could progress undetected to the point t
that large tube rupture from that mechanism is at all likely.43 t
i 42 Eddy current testing failed to detect the source of a known leak in one steam generator tube, and it is not unusual for a through-wall defect to appear on an eddy current test to be an 80 percent defect Tr.
1661-64 (Denton).
Additionally, an eddy current test sometimes has shown a defect as great as 90 percent that was not detected at all in testing conducted just six months before.
Tr. 1643-47 (Denton).
This indicates a high degree of uncertainty in these particular readings because reliable laboratory tests conducted on samples of mill annealed Inconel 600 indicate that the maximum rate of deterioration in a highly caustic environment during a six month period was no more than 7.5%
Fletcher, ff. Tr. 1422 at 6.
l These limitations in eddy current testing are known.
Since 1979, l
Westinghouse has conducted research to improve the early detection of IGA.
Recently, Westinghouse has developed a process for exposing tubing to an acid condition to produce laboratory samples with IGA of various depths of penetration, unaccompanied by cracking.
Westinghouse is testing the eddy current response to the IGA which, rather than the relatively sharp deviation. caused by an SCC signal, is a " drift" from the base line on the strip chart. - On an experimental basis, it now seems possible to detect 20% wall penetration by IGA in the laboratory; and work is continuing to develop a standard that will enable the interpreter to recognize IGA in the field.
Tr.1437-47 (Fletcher).
l 43 Murphy, pp. 7-8.
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Initial Decision:
18
- 3. Detecting Flaws in Sleeves To this point, we have discussed difficulties in using eddy current testing in any tube in a steam generator. However, a narrower question rests before us.
Applicant is licensed to operate its plant according to its existing technical specifications.
It may operate any tube in its steam generator until eddy current tests show 40% or more degrada-tion of the nominal tubewall thickness.
At that point, the technical specifications require the tubes to be plugged.
Our jurisdiction is to decide whether it is safe to operate those degraded tubes with sleeves rather than plugs.
We have no jurisdiction over the safety of the remainder of the steam generator, which applicant already is licensed to operate.44 We conclude that the sleeving process reinforces and strengthens existing steam generator tubes.
No serious question has been raised about the integrity of the joints by which the sleeves are bound to the existing tubes.
The result is that, at the time the sleeves are insert-ed, the new and undegraded sleeve replaces the degraded tube as a portion of the primary pressure boundary of the reactor.
At that time, the sleeve enjoys greater integrity than many of the degraded tubes that j
applicant already is pemitted to utilize in its steam generator.
44 See Northern Indiana Public Service Comoany (Bailly Generating l
Station Nuclear 1), ALAB-619, 12 NRC 558, 565 (1980); Public Service Company of Indiana, Inc. (Marble Hill Nuclear Generating Station, Units l
1 ano 2), ALAB-316, 3 NRC 167, 171 (1976).
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Initial Decision:
19 1
i Furthermore, this new primary pressure boundary is made of a corrosion resistant material, thermally treated Inconel 600, which is two to three times more resistant to corrosion than the initial steam generator tubes,45 which were not thennally treated to enhance their corrosion resistance.
The safety of the newly installed sleeves may be further enhanced if ongoing research succeeds in improving the ability to detect corro-testing.46 This would pennit corrective sion using eddy current action.
Even if ongoing research does not succeed, sleeved tubes will be 1
safer than unsleeved tubes.
To the extent that there may be imprecision in the tests currently in use in steam generator tubes, including eddy 47 current testing and hydrostatic testing the insertion of new sleeves provides a margin of comfort not found in other tubes.
The other tubes, which have been used for many years, are subject to unde-tected corrosion; the new sleeves will take many years before their exposure to the steam-generator environment might cause an analogous risk in them.
45 Corrosion resistance of thermally treated Inconal 600 has been tested in the laboratory. IGA was shown to have been reduced by two to three times and stress corrosion cracking by about ten times.
- Fletcher, ff. Tr.
1422, at 6-7; Murphy, ff. Tr.
1828, at 2;.Tr.
1483-88 (Fletcher).
46 5Tr.1437-47 (Fletcher).
47 Discussed below.
e
Initial Decisicn:
20 Sleeves also will initially confront a less hostile environment
)
than will existing tubes.
Most sleeves will be protected from the secondary-side environment by the tubes into which they are inserted.
They will be exposed to the secondary side only if the repaired tube develops a substantial leak, thus permitting the potentially corrosive materials in the secondary side to touch the sleeve.48 Although neither applicant nor staff depends on the presence of the tube around the sleeve to support its belief that the sleeved tubes have an adequate safety margin, it is obvious that the presence of the tube enhances the safety of the sleeve.
If the sleeve were to rupture, it is possible that the surrounding tube would be so degraded that it would in I
no way constrain the resulting leak.
However, it is likely that the 4
degradation of the tube would be in a different region than the rupture 4
in the sleeve.
In that case, the intact tube may constrain both the 1
rupture and the leak from the sleeve.
While there is no assurance that this constraint would occur, this possibility weighs on the side of greater safety for a sleeved than for an unsleeved tube.49 i
An interesting beneficial side-effect of sleeving is that it will retard the process of corrosion of the surrounding tube.
This will i
occur because the sleeve will somewhat insulate the tube from the heat 48 Fletcher, ff. Tr. 1422, at 6.
49 See Marsh, ff. Tr.-1822, at 3-4; Murphy, ff. Tr. 1828, at 4.
Initial Decision:
21 i-of the primary system.
This reduction in temperature should be accom-panied by a reduced rate of corrosion, which is facilitated by heat.50 i
It is also likely that the thennal-hydraulic properties of the 51 4
tube-sleeve annulus will retard the accumulation of corrosive materials.
The most likely pathway for leakage into the annulus would be through the tubewall near the top of the tubesheet; this is the area of the steam generator where the greatest corrosion has occurred.52 The sleeve,.in direct contact with the heated and pressurized primary coolant,'will turn the water in the annulus to steam, which will escape throu]h the leakage pathway from which it entered.53 Consequently, the turnover of water and the deposition of sediment in the annulus would be severely limited,54 retarding the rate of accumulation of j
corrosive materials in the annulus, as compared to the accumulation at s
the top of the tubesheet.
The result is that there would be less 1
i 1
i 50 Mu rphy, ff. Tr. 1828, at 2; Tr. 1769-70 (Fletcher); Tr. 1851, j
1859-60(McCracken).
51 i
The space between the tube and sleeve is known as the " annulus."
]
52 Fletcher, ff. Tr. 1422', at 10; Tr. 1767-69 (Fletcher); Tr. 1851 (McCracken).
53 Mr. Fletcher anticipated that only a small. amount of water would i
enter the annulus before flashing to steam. Ordinarily, this would be the case. However, as corrosion progresses a substantial amount of water could leak into the annulus during a period of cold shutdown. It is our conclusion that steam still would form when the generator was returned to service following such a period, so we accept the implications of Mr.
Fletcher's analysis for the slightly different hypothetical situation we have envisioned.
Tr. 1766-73 (Fletcher); Tr. 1851-52 (McCracken); Tr.
1853 (Murphy).
54 See Tr. 1769-71.
y-
l Initial Decision:
22 l
sediment to facilitate corrosion of the sleeve, as compared to the amount of sediment facilitating corrosion of an unsleeved tube.
- Hence, the sleeved tube should be subject to a siower rate of corrosion.
Finally, we conclude that whatever the difficulties of eddy current testing, it is a more accurate instrument for testing the sleeve (below the upper joint) than for testing unsleeved tubes.
(We do not examine questions concerning the upper joint because we previously found there was no genuine issue of fact concerning the testing of the upper joint.55)
The principal reason for increased inspectability is. that noise from the tubesheet crevice will be reduced because the sleeve is separated from the crevice by the thickness of the surrounding tube plus the width of the annulus between the tube and sleeve.56 The outer surface of the sleeve is 75 mils away from the surface of the tubesheet hole. This significantly reduces the noise level.57 In sumary, we find that sleeved tubes are safer than unsleeved tubes already present in the Point Beach steam generator.
In addition, these tubes are easier to inspect for degradation that may occur.
Hence, we conclude that the sleeved tubes will be subject to an extreme-ly low probability of abnormal leakage, of rapidly propagating failure 00 and of gross rupture and that we should approve the request to amend 55 Sumary Disposition, slip op, at 15.
56 Fletcher, ff. Tr.1422, at 3-5.
57 Id. at 4.
58 General Design Criterion 14, Appendix A,10 CFR Part 50.
J
Initial Decision:
23
}
applicant's operating license to permit the sleeving of tubes that otherwise would be required to be plugged.
i D.
Safety Factors in Sleeved Tubes The safety of sleeved tubes does not depend on eddy current testing alone.
Consequently, although the' admitted contention deals with eddy current testing, our Summary Disposition decision invited evidence concerning the relationship between the testing program and the safety of the reactor.59 In response, evidence was submitted that persuades us that protection from steam generator tube failures depends on a series of safety factors, including:
1.
Design, fabrication and testing in compliance with the ASME Boiler and Pressure Vessel Code.
2.
Hydrostatic testing 3.
Continuous leak monitoring 4.
Leak-before-break characteristics of tubing material 5.
Conservative criteria for utilizing eddy current test results 6.
Possible leak constraint from the presence of the tube around the sleeve or from the tubesheet, and 7.
Likelihood of a less corrosive environment within the sleeve-tube annulus.
In this section of our opinion, we shall discuss each of these safety factors.
Although we could rest our opinion solely on the conclusions we reached above concerning the increased safety of sleeved tubes, 59 See Summary Disposition, slip op at 14.
Initial Decision:
24-l i
compared to unsleeved tubes, we also conclude that the combined effect i
of these seven factors contributes to safety, thereby complying with
~
General Design Criterion 14 Our review of these safety factors also persuades us that it would not be appropriate for us to initiate an inquiry of our own into possible safety or environmental problems with j
the sleeving project.60
[
1.
Compliance With ASME Code and Additional Testing j
Steam generators, including the tubes and sleeves, are designed, t
fabricated and tested in accordance with design criteria which include i
compliance with the ASME Boiler and Pressure Vessel Code.61 To
{
further assure itself of the safety of the proposed sleeving repair process, applicant had Westinghouse Electric Corporation conduct exten-sive analyses and laboratory tests.62 The ensuing " Sleeving Report" contains results of a design verification test program whose objective was to assess the structural integrity and corrosion resistance of 60 Atomic Safety and Licensing Boards have the authority to pursue l
relevant safety and environmental issues that arise in the course of a proceeding.
Although the use of this "sua sponte" authority has been made dependent on. Boards first notifying the Connission of their action in declaring a sua sponte issue, the continued existence of the authority to declare such issues imposes on i
Board the responsibility of considering whether or not to declare such issues.
Although it may not be strictly necessary to explain why that authority bas. not been exercised, this Board believes it preferable to expose its decisional. process to public scrutiny, t
61 Applicant Exhibit 1, 53.1.
62 Westinghouse Electric Corporation, Point Beach Steam Generator Report, September 1981 (Revised February 1982)(51eeving Report).
l l
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Initial Decision:
25 sleeved tubes.63 The laboratory tests that were perfonned included a variety of corrosion and structural tests on tube materials and on sample tubes.
At an earlier stage of this proceeding, we addressed a limited number of questions to the applicant concerning possible problems in the Sleeving Report.
As a result, we satisfied ourselves that the Sleeving Report was prepared with reasonable care and we were unable to identify i
any serious deficiencies for us to pursue.
At this stage of the pro-ceeding, the Sleeving Report also provides us with assurance that the sleeving project was carefully designed and tested and that there are no important safety or environmental issues for us to pursue.
Sleeved tubes will have greater integrity than unsleeved tubes.
The sleeves are made of thermally treated Inconel 600, which has greater resistance to corrosion than the mill annealed Inconel 600 used in the original tubes.
Laboratory tests indicate that the rate of propagation of IGA through thennally treated'Inconel 600 was 2 or 3 times less than the rate of propagation through the mill annealed tube material.
A larger reduction applies to the rate of propagation of SCC.64 63 Sleeving Report, Chapter 6.0;~SER at 20, 23.
64 Fletcher, ff. Tr. 1422, at 6-7; Murphy, ff. Tr. 1828, at :2; Tr.
1483-88 (Fletcher).
/
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Initial Decision:
26 L
l i
l 2.
Hydrostatic Testing Previous to the time that sleeved tubes are placed in service,65 and periodically thereafter,00 applicant will perfonn hydrostatic l
tests to locate leaks in tubes.
The tests involve pressure differen-4 tials substantially in excess of nonnal operating pressure differen-l tials.
The pressure differentials approximate those that would be expected to occur during postulated main steam line breaks or loss of i
coolantaccident(LOCA) events.07 l
f 3.
Continuous Leak Monitoring Since primary water contains small amounts of radioactivity that may be detected if it migrates to the non-radioactive secondary side of q
the steam generator, applicant continuously monitors the secondary system condenser air ejector and steam generator blowdown for radioac-tivity. The presence of radioactivity in these locations would indicate a 1Eak in the steam generator tubes or sleeves. Even very small leaks in tube sleeves can be detected through this monitoring process.68 65 a
See Safety Evaluation by the Office of Nuclear Reactor Regulation relating to Full Scale Steam Generator Tube Sleeving at Point Beach Nuclear Plant Units 1 ano 2, Docket Nos.-50-266 and 50-301, July 8, 1982 l
(SER), at i 6.0, p. 34.
66 Murphy, ff. Tr. 1828 at 2, 10; Fletcher, ff. Tr.-1422 at-S.
s 67 Murphy, ff. Tr.1828 at 2,10; Fletcher, ff. Tr.1422 at 5; SER at 34-35 (approving hydrostatic test plans for mechanically sleeved joints
- . M and questioning the adequacy of differential pressures for testing s
. applicant's abandoned plan for an alternate type of brazed upper joint).
60 Fletcher,.ff. Tr.1422 at 5-6; Murphy, ff. Tr.1828 at 2,10.
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4
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Initial Decisicn:
27 4
1 The NRC has established limits on overall leakage from tubes.in a steam generator.
If thse limits are exceeded, either by leaks through existing tubes or through sleeves, applicant will be required to shut i
down the reactor for repair.
Although some leakage is permitted in recognition of the difficulty of installing entirely leak-tight tubes, leak limits are established in order to assure that the unit would be shut down before the integrity of the leaking tube or tubes could become sufficiently impaired to risk a rupture either under normal operating conditions or postulated accident conditions.09 Leak limits are so rigorous that even if the entire leakage oc-f curred through one sleeve, the maximum through-wall crack length that could exist without exceeding tne limits for leakage (500 gpd or 0.3 gpm per steam generator) would be about 0.4" at nonnal operating pressures.
Even should a steam line break accident occur at a time that a flaw of that dimension existed, analysis indicates that the sleeve could with-stand the increased pressure differential without bursting.70 1'
4.
Leak-Before-Break Characteristic of Sleeves Another safety factor is that steam generator tubes and sleeves are made of a special material, Inconel 600, selected because of its high ductility and toughness, two characteristics which in combination constitute fracture resistance.
In this material, a crack (SCC or IGA) that began to fonn on the tube or sleeve's outer wall probably would 69 Id.
70 Fletcher, ff. Tr.1422, at 8.
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Initial Decision:
28 cause a small, detectable leak before it became susceptible to a rupture either during accident or nonnal operating conditions.71
~
Laboratory and operating experience confinn the validity of the leak-before-break concept.
Degraded tubes nonnally do not suffer large
~
breaks; they penetrate locally, suffering only minor in leakage that is i
readily detectable through continuous leak monitoring.
Almost all leakage events in Westinghouse steam generators were of this kind.72 Considering all operating reactors, there are hundreds of steam generators, containing thousands of tubes.
In all the years of opera-t tion of these tubes, there have been approximately 200 leaks reported to the NRC, and only four of these have involved large leak rates. None of j
the four occurrences resulted in any unacceptable offsite radiological 1
i consequences or any damage to the reactor core.
All resulted from 4
unusual circumstances that do not invalidate the leak-before-break characteristic of steam generator tubes.
i Important exceptions to the leak-before-brea.k concept have emerged:
1 that hoop stresses (caused by denting at the uppennost tube support plate), mechanical damage from loose parts,73 and substantial thin-74 ning may cause a rapid failure.
However, there is no significant 1
i a
. 71 Fletcher, ff. Tr. 1422 at 7.
72 Fletcher, ff. Tr. 1422, at 8.
73Murphy, ff. Tr.1828, at 10; Tr.1774-78 (Fletcher); see also Marsh, ff. Tr. 1822, at 3.
74 Tr.1774-81 (Fletcher).
j
Initial Decision:
29 l
denting present at Po1nt Beach.75 Applicant is aware of the loose parts problem and has begun monitoring for their presence.76 Further-i more, eddy current testing can reliably detect seriously thinned tubes, all of which have been removed.from service at Point Beach.77 The basic concept, that tubes and sleeves will respond to corrosion by leaking before they break, is still applicable to the sleeving repair at Point Beach.
~
In addition to operating experience, conservative analyses substan-tiate the leak before break concept.
The maximum primary-to-secondary i
pressure differential occurs following a postulated feedline break or steam line break accident, which reduces the secondary side pressure to zero. Analysis of this accident condition for the sleeve indicates that-even if there is unifonn thinning completely around the circumference, a sleeve can degrade to 38% of its nominal wall thickness and still resist ruptu re.78 This corresponds to 62% degradation, or over 50% more j
i 75 Licensee's response to Quest'ons in Memorandum and Order, dated April 7,1982.
Although there has been some denting in Unit 2, it has i
not progressed significantly and does not constitute significant tube -
plate support deformation.
Furthermore, these phenomena are not related to sleeving.
Id. at 1-2.
I 70 Letter to the Atomic Safety and Licensing Board from Bruce Churchill, November 9, 1982.
77Tr. 1774-81 (Fletcher). -(Because phosphate chemistry. is-no longer in use at Point Beach, Mr. Fletcher does not expect new instances of i
thinning to occur.)
O Sleeving Report at 6.120-6.121.
Initial Decision:
30 4
4 i
degradation than the 40% degradation whose detection -- at any one spot on the tubewall -- causes the NRC to require plugging of the tube.II To further confirm the analyses, there have been laboratory tests.
]
These " burst tests" have been performed on portions of tubes removed i
from Point Beach and suffering from IGA. of about 40% to 60%.
This-testing required differential pressures in excess of 5000 psi to cause bursting of the degraded tubes.
This indicates substantial additional margin over the conservatively estimated pressures resulting from postulated accidents. 80 Over all, we are confident that the l eak-before-burst concept, under normal operating conditions and postulated accident conditions, is applicable to the Point Beach sleeving amendment.
]
5.
Conservative Criteria for Eddy Current Testing f
l At Point Beach, hydrostatic testing and eddy current testing i
programs reduce the risk that serious degradation of tube or sleeve walls may occur without detection. Both tubes and sleev6s in which eddy current testing indicates 40% or more degradation must be removed from service.81 Even though tubes and sleeves with small leaks are not subject to
- rupture, these testing programs successfully identify i
79 Fletcher, ff. Tr.1422, at 9; Murphy, ff. Tr.1828, at 3-4.
80 Fletcher, ff. Tr. 1422, at 6-7; Murphy, ff. Tr. 1828 - at 2; Tr.
1483-88 (Fletcher).
81 SER at 21.
-n-a a
s
u-s =
v
'mw-
Initial Decision:
31 partially degraded tubes, and those tubes are removed from service as an added precaution.
i l
As we have just pointed out in the previous subsection of this I
opinion, eddy current test indications of 40% degradation cause tubes t
and sleeves to be removed from service. This represents more than a 50%
safety margin, even were the degradation to be unifom for the entire outer diameter of the tested tubes.
We are convinced that eddy current testing, used in this conserva-tive manner, contributes to the overall safety of the sleeved tubes.
4 6.
Possible Leak Constraint from the Tube or Tubesheet i
Most of the sleeved portion of the tubes lies within the tubesheet.
In that area, which is the area in which IGA has been found when tube samples have been removed from the steam generator, the tube is tightly constrained by the tubesheet, minimizing any potential for rupture.82 If rupture of the sleeve were nevertheless assumed to occur within the tubesheet as a result of IGA or SCC, the leak path would be obstructed i
by the narrow tube-to-tubesheet crevice, and the leak rate would be significantly reduced compared to the rate postulated to occur above the i
tubesheet from a ruptured tube.83 Sleeving would provide an additional barrier against leakage.
Even if the sleeve begins tc rupture, the event may be teminated or severely limited if it occurs in an area.of the original tube which has 4
l 82 Murphy, ff. Tr. 1283 at 6.
83 Fletcher, ff. Tr. 1422, at 10; Murphy, ff. Tr. 1828, at 6.
4 1
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Initial Decision:
32 i
sufficient remaining strength to resist rupturing at the corresponding point,on the tubewall.
If the tube does remain intact at that point, then it may prove an effective barrier to any leskage at all to the l
secondary side.
In the alternative, leakage may occur into the sleeve-tube annulus and thence, through a hole in the sleeve, to the
~
secondary side.
However, such a leak undoubtedly would occur at a far i
i slower pace than a fishmouth rupture or double-ended break in a single 84 tube, not supported by a sleeve.
Even if these benefits of the f
1 sleeving configuration are not real.ized, there is no reason to believe that a rupture of a sleeve would be worse than the rupture of an un-t i
sleeved tube.85 i
i 7.
Less Corrosive Environment in the Annulus The rate of corrosion in tubes or sleeves depends on the environ-ment to which they are exposed.
The outer diameter of the sleeve will' not be exposed to the secondary side environment unless degradation in the original tube propagates through-wall and the original-tube's grain j
boundaries separate enough to admit solution from the non-pressurized secondary side into the annulus.86 This would require substantially more degradation of the tube than would occur before it was removed from service because of fears that it could not withstand operating pressures 4
l 84 Marsh. - ff. Tr.1822, at 3-4; Murphy, ff. Tr.1828, at 4.
85 Id.
86 Fletcher, ff. Tr.1422, at 6.
i
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, = - -
Initial Decision:
33 P
or accident conditions.
Hence, there ordinarily will be a substantial delay before the sleeve is exposed to a corrosive environment.
Should a corrosive environment occur in the annulus, the leak into the annulus would probably occur in the tubesheet area, where sludge is deposited. Thence, th'e corrosive material would travel to the bottom of the annulus, within the tubesheet crevice.
In that location, it is possible that a corrosive environment could develop, but there is no reason to believe that the rate of corrosion would be any worse than what already is found in the tubesheet crevice.
Consequently, the sleeves would never be exposed to a mora corrosive environment than are tubes.
Also, the location of the corrosion--at the bottom of the annulus--only creates a risk of a constrained leak, rather than a guillotine or fishmouth rupture.07 We have discussed, above, the testimony of Mr.. Fletcher concerning the properties of the annulus and the reason for believing that the fluid turnover rate and sedimentation rate would be low in that area.
8.
Conclusion The uncontradicted evidence shows that sleeving enhances safety, both from the point of view of increased integrity of the primary 87 Tr. 1767-70, 1766-73 (Fletcher); - Tr. 1851-52 (McCracken); Tr. 1853 (Murphy).
The implications of a constrained leak are discussed in subsection 6 suora, i
a.
Initial Decision:
34 pressure boundary and decreased consequences of a breach in the pressure boundary.00 Sleeving will provide lower probabilities of the occur-rence of the three events -- abnormal leakage, rapidly propagating failure, and gross rupture -- which are required to be minimized by General Design Criterion 14.09 We therefore conclude that there is no serious safety or environmental issue of which we are aware that re-quires us to undertake cur own further inquiry.
ORDER For all the foregoing reasons and based on consideration of the entire record in this matter, it is this 4th day of February 1983 ORDERED:
1.
The sole remaining genuine issue of fact in this proceeding, cencerning the adequacy of eddy current testing of sleeved steam genera-tor tubes, is dismissed.
2.
We authorize the Of rector of Nuclear Reactor Regulation to issue a license amendment to Wisconsin Electric Power Company, concern-ing the repair of steam generator tubes at its Point Beach nuclear plant by sleeving, subject to understandings of record, that:
a.
Steam generator tubes that have been previously subject 88 We examined this question witn especially great care because Mr.
l Marsh's testimony indicates that there may be a substantial risk from the rupture of only one or two steam generator tubes.
Marsh, ff. Tr.
1822 at 5; Tr. 1339-41.
89 Fletcher, ff. Tr.1422 at 12.
l l
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. w
Initial Decision:
35 to explosive plugging, shall not be sleeved; b.
Brazed joints shall not be employed; c.
Should eddy current testing indicate 40 percent or more degradation from the nominal tube wall thickness of a I
l
- sleeve, the sleeved steam generator tube shall be plugged; and d.
Leak limits previously imposed on the repaired steam generators shall continuc-to apply.
3.
Pursuant to 10 CFR 5 2.760(a) this is an initial decision that will constitute final action of the Comission forty-five (45) days from the date of issuance unless exceptions are taken pursuant to i 2.762 or the Comission directs that the record be certified to it.
4.
Exceptions to this decision or designated portions thereof may be filed with the Comission, in the form required by 5 2.762(a), within ten (10) days after service of this decision.
5.
To pursue an appeal, briefs in support of a party's objection also must be filed, within thirty (30) days after filing the exceptions (or forty days in the case of the staff of the Nuclear Regulatory I
Comission). The brief must comply with the requirements of 5 2.762.
6.
Within thirty (30) days of the service of the brief of the appellant (40 days for the staff), parties may file opposing or sup-porting briefs that comply with the requirements of i 2.762.
7.
Filings that do not comply with the rules governing appeals may be stricken.
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Initial Decisicn:
36 FOR THE ATOMIC SAFETY AND LICENSING BOARD LL U
Peter B. Bloch, Chaiman ADMINISTRATIVE JUDGE
/ 4 c.~ &m e f 6 Hugh S/. Paxton ADMINISTRATIVE JUDGE 1
4 o
r
@rry R. Mine' ADMINISTRATIVE JUDGE f
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Initial Decision:
37 APPENDIX A PUBLISHED POINT BEACH BOARD ORDERS Title Date of Document LBP No.
MEMORANDUM AND ORDER (Requesting Additional Infor-mation) 10-01-81 81-39 MEMORANDUM AND ORDER (Setting Agenda For October 9 Conference Call) 10-07-81 43 MEMORANDUM AND ORDER (Concernin Questions)g Further Board 10-13-81 81-44 MEMORANDUM AND ORDER (Concerning The Admission Of A Party And Its Contentions) 10-13-81 81 MEMORANDUM AND ORDER (Setting Agenda And Rules ForOctober29-30 Hearing) 10-15-81 81-46 MEMORANDUM AND ORDER (Authorizing Issuance Of A License Amendment Permitting Return To Power With Up To Six Degraded Tubes Sleeved Rather Than Plugged) 11-05-81 81-55 MEMORANDUM AND ORDER (Concerning Preliminary Confidentiality Issues) 12-21-81 81-62 l
SUPPLEMENTARY ORDER (Concerning Issuance Of A j
Protective Order) 01-07-82 82-2 MEMORANDUM AND ORDER (Concerning Reconsideration i
Of Confidentiality Issues)
.01-28-82 82-5A 1
/
i
Initial 0; cision:
38 MEMORANDUM AND ORDER (Concerning The Burden Of Going Forward On Confi-dentiality Issues) 02-02-82 82-6 MEMORANDUM AND ORDER (Concerning A Motion To Compel l
And Other Hatters) 02-19-82 82-10 MEMORANDUM AND ORDER (Concerning a Motion To Certify A Sua Sponte Question) 02-26-82 82-12 MEMORANDUM AND ORDER (Concernin A Motion To Reconsider 03-19-82 82-19A MEMORANDUM AND ORDER (Concerning Reconsideration Of A Motion To Certify A Sua Soonte Question) 03-31-82 82-24A MEMORANDUM AND ORDER (Concerning A Motion To Compel) 04-22-82 82-33 MEMORANDUM AND ORDER (Concerning A Motion To Re-lease To The Public Certain Safety Information Which Is Part Of The Record In This Case But Is Proprietary To Westinghouse Electric Corp-oration) 05-26-82 82-42
~,
MEMORANDUM AND ORDER (Concerning Summary Dis-position Issues) 10-01-82 82-88 s
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l Initial Decision:
39 Appendix B Qualifications of Witnesses Applicant's Witnesses W.D.
Fletcher, hanager of Steam Generator Development and Performance Engineering in the Nuclear Technology Division of the Westinghouse Electric Corporation. He has a Masters degree in Chemistry from Fordham University, 1960.
Since 1970 he has been directly involved in development and design activities related to Westinghouse steam generators.
He is credited with a variety of professional publications, including publications about Westing-house steam generators, primary coolant chemistry in PWR's and corrosion of stainless steel.
Clyde J. Denton, a participant in the group that originated eddy current testing of steam generators and presently general manager of Zetec, Inc.
He has an A.A.S. from the Milwaukee School of Engineering and has been doing eddy current testing since 1956.
Edward O. McKee, a technician with 11 years' experience in interpreting eddy current data.
He has evaluated all ECT data for both Point Beach units.
Staff's Witnesses Emmett L. Murphy, Senior Systems Engineer in the staff's Operating Reactors Assessment Branch.
He has a Master of Science Degree in Civil Engineering and a Bachelor of Science Degree in Aerospace Engineering, both from the University of Maryland.
He has worked for nine years in the nuclear field, including six years as structural engineer at the Bettis Atomic Power Laboratory of Westinghouse Corporation.
Since July 1979 he has been working for the staff almost exclusively on safety reviews of steam genera-tors that have experienced significant tube degradation.
1 Ledyard B. Marsh, Section Leader of staff's Reactor Systems Branch.
He has a Masters of Science in Nuclear Engineering from the University of Washing-ton, was an officer in the Navy Nuclear Power Program from 1970 to 1974, and joined the Reactor Systems Branch in 1976.
Timothy G. Colburn, staff's Project Manager for the Point Beach reactors.
He has a B.S.
in mechanical engineering from Notre Dame, worked -in the Navy's nuclear power program and was employed by Potomac Electric Power Company.
Conrad E. McCracken, Section Leader of the staff's Chemical Technology Section of the Engineering Branch.
He is a registered Professional Corrosion Engineer who was qualified 'in submarines for all nuclear duties by the United States Navy and who served as Manager of Chemistry Develop-ment for Combustion Engineering Corporation from 1966 to 1981, when he jointed the staff as a senior chemical engineer.
Note: Wisconsin's Environmental Decade did not call any witnesses.
1
Initial Decision:
40 i
APPENDIX C Comment on Limited Appearance Statements In preparing this decision, we remember the people who addressed us when we sat in a Limited Appearance session in Two Rivers Wisconsin on November 17, 1982.
Although there are many people living near Point Beach who are. pleased with the use of nuclear reactors to' generate electricity,N the people who addressed us were thoughtful people with serious doubts. One of the speakers, Mr. Edward Klessig, said what many had on their minds:
We pride ourselves on being practical farmers.
We service most of our cwn equipment.
The proposed sleeving repa.ir process reminds us of fixing a sophisticated hay bailer or combine with a piece of bailing wire.
As farmers and food producers we love the land.
We don't want to risk contaminating the precious soil and the food chain with radioactive isotopes, at best, or total disaster at worst.91 We are aware of these citizen concerns and of the trust that is placed in us to resolve the matter before us.
We are particularly aware that a license amendment dealing with " tube sleeving" does superficially The Town Board of the Town of Two Creeks unanimously supports the
" economic and efficient way of producing electricity".at Point Beach and approves of the proposed sleeving process.
Letter to Mr. Peter Bloch (November 29,1982).
91 Tr. 10009.
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Initial Decision:
41 resemble a patchwork repair.
Consequently, we have been especially attentive to our record, which contains r;umerous tests and analyses that have been relevant to our deliberations either at this or at an earlier stage of the proceeding.
We hope that if Mr. Klessig and his fellow citizens should read this memorandum that they will be assured that the steam generator repair has been engineered with great care.
Even should they disagree with our conclusion that none of Decade's contentions is valid and that there is no serious safety or environmental issue for us to raise ourselves, we hope they will realize that our decision to approve the pending license amendment has not been lightly taken.
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