ML20030A236
| ML20030A236 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 01/03/1980 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | WISCONSIN ELECTRIC POWER CO. |
| Shared Package | |
| ML19257B442 | List: |
| References | |
| NUDOCS 8001160266 | |
| Download: ML20030A236 (10) | |
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's UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ~
In the Matter of
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WISCONSIN ELECTRIC POWER COMPANY
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Docket No. 50-266 (Point Beach Nuclear Plant,
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Unit 1)
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ORDER MODIFYING CONFIRMATORY ORDER OF NOVEMBER 30, 1979 I
Wisconsin Electric Power Company (the Licensee) is the holder of Facility Operating License No. DPR-24 which authorizes the Licensee to operate the Point Beach Nuclear Plant, Unit 1, located in Two Creeks, Wisconsin, under certain specified conditions. License No. DPR-24 was issued by the Atomic Energy Com..ission on October 5,1970, and is due to expire on July 25, 2008.
II Inservice inspections of the Point Beach Unit 1 steam generators performed
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during August 1979 and October 1979 outages have indicated extensive general intergranular attack and caustic
~;ress corrosion cracking on certain of the external surfaces of the steam generator tubes. The NRC Staff determined in November 1979 that additional operating conditions would be required to assure safe operation prior to resumption of operation of Point Beach Unit 1 from a refueling outage. Such conditions were imposed by Confirmatory Order for Modification of License dated November 30, 1979.
In addition to those conditions, the Staff has now determined that additional conditions are required to provide continued assurance that Point Beach Unit I can be operated safely.
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. These additional conditions are analyzed in a Staff Safety Evaluation Report, dated this date, which is attached to this Order.
The Licensee has agreed to this condition by letter dated December 31, 1979.
III Atcordingly, pursuant to the Atomic Energy Act of 1954, as amended, and the Commission's Rules and Regulations in 10 CFR Part 2 and Part 50, IT IS HEREBY ORDERED THAT License No. DPR-24 be amended, in the manner hereafter provided, to include the following conditions in addition to those conditions listed in the Confirmatory Order of November 30, 1979:
1.
Unit I will be operated at a reactor coolant pressure of 2000 psia with the associated parameters (i.e., overtemperature AT and low pressurizer pressure trip point) with the limits indicated in the Safety Evaluation Report appended to this Order.
2.
The licensee shall develcp and follow the necessary precedures fcr operating Unit 1 at the conditions described in condition 1 above.
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In view of the above, this amendment of License No. DPH-24 is made j
immediately effective. Accordingly, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of receipt of this Crder, the Point Beach Unit 1 facility shall be operated at a reactor l
coolant system pressure of 2000 psia witain the parameters described i
above.
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3-V Any person whose interest may be affected by this Order may within twenty days of the date of this Order request a hearing with respect to this Order.
Any such request shall not stay the effectiveness of this Order. Any request for a hearing shall be addressed to the Director of Nuclear Reactor Regulation, U. S.
Nucleat Regulatory Comission, Washington, D. C.
20555.
In the event a hearing is requested, the issues to be considered at such hearing shall be:
- 1) Whether the facts stated in Section II of this-~0rder are correct; aM
- 2) Whether this Order should be sustained.
FOR THE NUCLEAR REGULATORY COMMISSION M
Harold Denton, Director Office of Nuclear Reactor Regulation
Attachment:
Staff Safety Evaluation Report, l
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dated January 3,1980 Effective date: January 3,1980 Bethesda, Maryland 1
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE POINT BEACH UNIT 1 STEAM GENERATOR-TUBE DEGRADATION DUE TO DEEP CREVICE CORROSION WISCONSIN ELECTRIC POWER CdNPANY POINT BEACH NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-266 INTRODUCTION Wisconsin Electric Power Company (the licensee) has requested changes to the Technical Specifications of Point Beach Units 1 and 2 to allow operation at either 2000 or 2250 psia (RefereMci 1). These changes includ? (1) defining over-temperature - aT-trip equation for each operating pressure, and (2) redefining
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the Tow pressure trip to allow adequate operating margin when operating a$ the lower pressure (2000 psia).
s Although 2250 psia is the design operating pressure, both units have been previously operated at the lower pressure. A brief history of the previous operation of Point Beach Units 1 and 2 is given by the licensee in References 1 and 7 outlining the reasons for changing the pressure, the dates at which these changes were made and nroviding the references to the various Amendi.
't requests for NRC and the subsequent Staff Safety Evaluation Reports. Presently both units are operating at 2250 psia. The licensee requested the change to permit operation at 2000 psia to reduce stress on the steam gererator tubes.
This change to a lower pressure adversely affects the departure from nucleate boiling ratio (DNBR) and requires justification that the reactor is still adequately protected.
The proposed change in the over temperature - aT(OTAT) trip provides this protection for some cases.
For situations where the OTAT trip does not operate, adequate protection must be shown by other analysis. The loss of flow and rod drop events are two events in which DNBR protection is provided by means other than the OTaT trip.
Mcdification of the reactor low pressurizer pressure trip to provide more margin between the 1mver operating pressure and this trip also requires justification that the applicable criteria for transient and accident anslyses are still satisfied.
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- Background In the Confirmatory Order for Modification of License dated November 30,1979 (Order) certain requirements were made pertaining to the operation of Point Beach, Unit 1. In the Safety Evaluation appended to that Order certain remedial actions were discussed.
Among these remedial actions we noted that the licensee planned to operate the facility at the reactor coolant pressure of 2000 psia rather than at 2250 psia to reduce the internal pressure stresses of operation by about 15% during operation (Action No. 3,
- p. 15). This action was to be initiated upon NRC approval of an amendment request dated November 2,1979 wnich requested permission to cperate ?,c that pressure.
In the same Safety Evaluation we discussed " Measures for P.aducine the Rate of Degradation" on pp 22 and 23. We indicated that the accertability of this proposed operation would be addressed separately. That Safety Evaluation is incorporated into this Safety Evaluation by reference.
The Order of November 30, 1979 was based on informatica resulting from the steam generator tube inspection of October 1979. On December 11, 1979 another steam generator leak occurred.
An eddy current test was performed on both steam generators which resulted in eddy c rent indications below the tube sheet (in the tube crevice) in both steam generators. Twenty tubes were plugged in stean generator A and fifteen tubes were plugged in steam generator B.
Since there appears to be evidence of continuing intergranular corrosion attack the NRC Staff has now fcund that is not only desirable, but prudent and necessary, to take innediate action to require the reactor coolant pressure to be reduced from 2250 psia to 2000 psia since this will have the effect of substantially reducing the differential pressure across all tubes in both steam generators.
As explained below, operation of Unit 1 at a reactor coolant pressure of 2000 psia is acceptable from an accident analysis point of view.
The applicable criteria for transient and accident analysis are still satisfied.
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The licensee has withdrawn the amendment request and has made a commitment to operate the unit at a reactor coolant system pressure of 2000 psf a only (Reference 8).
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, Evaluation Overtemperature T Trip and cow Pressurizer Pressure Trip For condition 1 (nomal operation) and condition 2 (anticipated transients) events (i.e. where overtemperature trip is required) the fuel rods must be protected from overheating by maintaining the departure from nucleata boiling ratio (DNBR.) above the safety limit of 1.3.
The primary method of doing this is by means of the overtemperature - AT trip. This trip is a function of pressure and is also a function of the value assumed for the low pressurizer pressure trip as explained in Reference 2.
Reducing the low pressurizer pressure trip for 2000 psia operation from 1865 psig to 1790 psig would allow more operating margin between the lower operating pressure (2000 psia) and the low pressurizer pressure trip. The licer.see provided an equation for the over-temperature - AT trip applicable to operation at 2000 psia.
In 1973, the licensee proposed operation at 2000 psia. Justification for this was presented in Reference 3.
The staff approved operation at 2000 psia and the corresponding overtemperature - aT equation in Reference 4.
As can be seen from Table 1 the currently proposed 2000 psia equation for the overtemperature aT does not result in a significant decrease in margin to DNB when compared to the previously approved equation for 2000 psia.
Also, as shown in Table 1, the values of the trip
- are almost the same at 2250 psia.
This results in a gain in DNB margin at the higher pressure since the trip values remain almost the same while the pressure increased 250 psia, from 2000 psia to 2250 psia.
Increasing pressure under PWR conditions results in increased margin to DNB. Therefore, even though the higher pressure would have Justified a high trip value, the value was kept the same.
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- The values shown in the table are nomalized to full power delta - T.
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TABLE 1 WCAP 8151 Present Present (Reference 3)
Tech Spec Evaluation
, AVG ( F) 2000 psi 2250 psi 2000 psi T
550 1.47 1.48 1.465 560 1.33 1.33 1.315 570 1.16 1.18 1.165 578 1.029 1.06 1.045 580 0.9978 1.03 1.015 590 0.839 0.88 0.87 As discussed in the next section, the licensee also reviewed the Condition 2 events which trip on the overtemperature 6T trip and found that the DNBR=1.3 safety limit is not exceeded with the new overtemperature 6T equations.
Based on the fact that the proposed overtemperature 6T trip equation at 2000 psia gives values which have not changed significantly from the values previously approved by the staff for operation at 2000 psia and the fact that a review cf Condition 1 and Cor,dition 2 events (the only events to which tb2 overtemperature '
AT trip applies) shows that the DNBR=1.3 safety limit is not rxceeded, we find
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the new overtemperature aT equation to be acceptable.
Transient and Accident Analyses Affected by Lower Operating Pressure The licensee has also reviewed the postulated accident events in the FSAR using the methods described in Reference 5, known as the Westinghouse Reload Methodology, to determine the effect of reduced pressure operation on the plant transients and accidents. This review determined that several of these events needed to be reanalyzed. These events are listed in Table 2.
TABLE 2 Accidents Re-Analyzed For Low Pressure Operation Rod Ejection Loss of Flow Locked Rotor Rod Withdrawal at Power
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. WCAP 8151 (Reference 3) gives a qualitative discussion of the impact of 2000 psi operation on the transient and accident analyses.
These conclusions are, in general, still valid.
The low pressurizer pressure trip is important in the small break LOCA. The value assumed for this trip in the analysis is 1795 psig which is above the low pressure trip being proposed for 2000 psia operation.
The licensee stated that the analysi is still conservative because the reduction in pressure from 2250 psia to 2000 psia more than offsets the slight (5 psi) change in low pressurizer pressure setpoint.
For example, the licensee states that at 2250 psi and 1795 psi low pressurizer pressure trip there would be 3.8 full power seconds before trip whila at the lower operating pressure of 2000 psia whit the corresponding low pressurizer pressure trip of 1790 psig, only 0.8 full
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power seconds would result in the case of the worst small break.
The large Break Loss of Coolant Accidnet (LOCA) was also reanalyzec et 2003 psia (and 18% steam generator tube plugging) to justify operation at the lower pressure (Reference 5). Only the limiting break size (a DECLG, C.=0.4) was reanalyzed.
This is acceptable since the change in peak cladding temperature is relatively small and the reactor pressure would not be expected phenomenologically to have a large effect.
The results of the LOCA analysis for both 2000 psia and 2250 psia are given in Table 3.
TABLE 3 Results of LOCA Analysis for Point Beach Unit i for 2000 psia and 2250 psia l
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2000 psia 2250 psia Peak Clad Temperature (' F) 2062 2053 Maximum Local Clad / Water Reaction (".)
5.11 5.3 Total Core C1ad/ Water Reaction (*)
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The overpower - t-T trip which provides protection against fuel centerline melting is derived in such a way that it is not a function of reactor coolant system (RCS) pres 3ure of the low pressurizer pressure reactor trip (it is a function of the high pressurizer pressure reactor trip).
It is therefore unaffected by the change in pressure.
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s 6-Sumary The staff has reviewed the cha. ige to operate Unit 1 at a reactor coolant pressure of 2000 psia and finds it acceptable basec on two points. The first is that the licensee, using the standard Westinghouse reload methods (Reference 6), has verified that Point Beach Unit I would still meet the applicable safety criteria.
The second point is that no significant reduction in ma,rgin has been made in the overtemperature - AT set point over that previously approved by the staff. While thia second point was not essential to acceptability of the proposed change, it does provide additional assurance of safe operation.
The Safety Evaluation appended to the Nevember 30, 1979 Confimatory Order for Modification of License considered the reduction of reactor coolant pressure to 2000 psia as one of the licensee's proposed actions to reduce the rate of steam generator tube degradation (p.15 and p. 22).
The staff indicated that the acceptability of this proposal would be addressed separately (p. 23) and further discussed the other components that could be affected.
The staff concluded that the remedial actions proposed by the licensee will mitigate the t
effects of postulated accidents and retard the rate of corrosion (p. M).
We have now completed the review of the licensee proposal to operate at 2000 psia and find; 1) from the view of the inter-related operating considerations the reduction in pressure is acceptable, e, from the view of steam generator tube degradation it is prudent to reduce that degradation as much as possible.
The reduction of the reactor coolant pressure was one of the licensee proposed actions to reduce steam generator tube degradation and was postponed only to permit a complete review of the interrelation of other systems. Now that we have concluded that the reduction in pressure produces no problem in other operating parameters or systems, it is prudent and necessary that this reduction in pressure be accomplished as soon as possible.
Conclusion I
We have concluded, based on the considerations discussed above, that: (1) operation at a reactor coolant system pressure of 2000 psia is requireJ to provide continued assurance that the health and safety of the public will not be endangered, and does not involve a significant hazards consideration, and (2) such activities will be conducted in cwpliance with the Comission's regulations and the issuance of this requirement will not be inimical to the comon defense and security or to the health and safety of the public.
Date: January 3,1980
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. References 1.
Letter fran S. Burstein, Wisconsin Electric Power Company to H. Denton U. S. Nuclear Regulatory Commission, dated November 2,1979.'
2.
Ellenberger, S.
L., et. al., " Design Basis for Thermal Overpower aT and Thermal Overtemperature AT Trip Functions, Westinghouse Electric Corporation." WCAP 9745, March 1977.
3.
" Fuel Densification: Point Beach Nuclear Plant Unit No. 2 Low Pressure Analysis," Westinghouse Electric Corporation, WCAP 8151, June 1973.
4.
(a)
" Safety Evaluation by the Directorate of Licensing, Amendment No. 3 to Facility Operating License No. DPR-24, (Change No. 8 to Appendix A of Technical Specifications), Wisconsin Michigan and Wisconsin Electric Power Company, Point Beach Nuclear Plant Unit No.1, Docket No. 50-266," May 23,1974, trgesmitted by letter; Dennis L. Ziemann for Karl R. Galler to Mr. Sol Burstein, May 23, 1974.
(b)
" Safety Evaluation by the Directorate of Licensing, Supporting Amendment No. 5 to License No. DPR-27, Change No.11 to the Technical Specifications, Wisconsin Electric Power Company and Michigan Power Company," September 30, 1974, transmitted by letter; Karl R. Goller to Sol Bt stein, September 30, 1974.
5.
Letter from C. W. Fay, Wisconsin Electric Power Company to H. Denton, U. S. Nuclear Regulatory Commission, dated November 27, 1979.
6.
Bordelon, F., et. al., " Westinghouse Reload Safety Evaluation Methodology,"
Westinghouse Electric Corporation, WCAP 9272, March 1978.
7.
Letter fran S. Burstein, Wisconsin Electric Power Company to H. Denton,
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U. S. Nuclear Regulatory Commission, December 19, 1979.
I 8.
Letter from S. Burstein, Wisconsin Electric Power Company to H. Denton l
U. S. Nuclear Regulatory Commission, December 31, 1979.
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UNITED STATES
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April 4,1980 u['j V*
Docket No. 50-266 m,
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Mr. Sol Burstein Executive Vice President ATTACHMENT 5 Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53201 l
Dear Mr. Burstein:
Enclosed is a signed original Order dated April 4,196G, issued by the Cormiission for Point Beach Nuclear Plant Unit No.1.
The Order requires that testing be performed within 90 effective full power days.
With these additional limits, we have concluded that there is reasonable assurance that the public health and safety will not be endangered by the continued operation of Point Beach Unit No.1.
The basis for this conclusion is contained in our Safety Evaluation Report which is appended to the Order.
A copy of the Order is being filed with the Office of the Federal Register for publicaticn.
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Sincerely, M'Oid$v
.s A. S'chwencer, Chief Operat,ing Reactors Branch #1 l '
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Division of Operating Reactors i
l Enclosurc:
Confirmatory Order cc:
w/ enclosure See next page
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e Mr. So1' Burstein Wiscensin Electric Power Company 2-April 4, 1980 cc: Mr. Bruce Churchill, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.
Washington, D. C.
20036 Document Department University of Wisconsin Stevens Point Library Stevens Point, Wisconsin 54481 Mr. Glenn A. Reed, Manager Nuclear Operations Wisconsin Electric P:wer Coreany Point Beach Nuclear Plant 6610 N'Jclear Road
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Two Rivers, Wisconsin 54241 Walter L. Myer Town Chairman Town of Two Creeks Route 3 Two Rivers, Wisconsin 54241 Chai rman Public Service Connission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 Ms. Kathleen M. Falk General Counsel Wisconsin's Environmental Decade 114 E. Mifflin Street l
l Madison, Wisconsin 53703
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Director, Technical Assessment Divirion Office of Radiation Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 i
I U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN: EIS COORDINATOR 230 5. Dearborn Street Chicago, Illinois 60604 g
s, UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Iri the Matter of WISCONSIN ELECTRIC PCWER C0ff ANY
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Occket No. 50-266 (Point Beach Nuclear Plant,
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Unit 1)
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MODIFICATION OF NOVEMBER 30, 1979 ORDER I.
Wisconsin Electric Power Company (the licensee) is the holder of Facility Operating License No. DPR-24 which authorir.es the licensee to operate the Point Beach Nuclear Plant, Unit 1, located in Two Creeks, Wisconsin, under certain specified conditions. License No. DPR-24 was issued by the Atomic Energy Comission on October 5,1970 and is due to expire on July 25, 2008.
II.
Inservice inspections of the Point Beach Unit 1 steam generators performed during the August 1979 and October 1979 outages Mdicated
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extensive general intergranular attack and caustic stress corrosion crackir.g on certain of the external surfaces of the steam generator tubes. As a result of informatico provided in discussions with the licensee. and its representatives, which is documented in a letter dated November 23, 1979 from S. Burstein to H. R. Denton,' and tc.e Staff's Safety Evaluaticq Report, dated November 30, 1979, on Point Bei.:h Unit 1, Steam D > (% v G 2 oo Wu od '/
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. - Generator Tube Degradation due to Deep Crevice Corrosion, it was determined that additional operating conditions would be required to assure safe oper-ation prior to resumption of operation of Unit 1 from the 1979 refueling outage.
III.
The licensee in letters dated November 29, 1979 and November 30, 1979 agreed to additional conditions which were necessary to provide reasonable assurance for safe operation of Unit 1.
On November 30, 1979, an Order was issued to impose limiting conditions on continued operation of Unit i for a period of 60 effective full power days, at which time the licensee was required to shut down until the Director of Nuclear Reactor Regulation determined in writing in accordance with condition 6 of the Order that the results of the eddy current tests required by the Order were acceptable. On February 28,1980, Unit 1 1
was taken out of service for the tests required by the Order.
On Mar: 7. 28, 1980, the licensee provided the results of such tests to
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the NRC.
In accordance with condition 6 of the November 30, 1979 Order the NRC staff has reviewed the licensee's March 28, 1980 submittal and has assessed whether continued operation of the facility would be safe.
I have found for the reasons given in the attached Safety Evaluation that the public health, safety and interest requires tnat Unit 1 be shut down and certain tests be conducted within 90 effective full pcwer days of operation after the date of this Order. The licensee has agreed
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. - to.this condition.
Subject to this condition and with continuatien of the other conditions set forth in the November 30, 1979 Confirmatory Order and the January 3,1980 Modification of the Order, I have concluded l
t*.at there is reasonable assurance that the public health and safety will not be endangered by the continued operation of Point Beach Unit 1.
IV.
Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, and f
the Commission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS HEREBY ORDERED IHAT the November 30, 1979 Confirnatory Order for Modification of License be amended, effective innediately, to delete condition 1 of Section IV of that Order and replace such condition with the following condition.
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Within 90 effective full power days from the date of this Order, a 2000 psia primary-to-secondary hydrostatic test and a 800 psia secondary-to-primary hydrostatic test shall
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be performed. Also during this plant outage, an eddy current examination shall be performed on tubes in each steam generator.
The program shall be submitted to the NRC for staff review and require examination of about 1000 tubes in the central region of the hot leg and three (3) percent of all tubes outside this central regien and 3% of all cold leg tubes. The central region shall encompass
,all areas where deep crevice corrosion has previously been observed.
. All other conditions of the November 30, 1979 Confirmatory Order and the January 3,1980 modification of that Order, including condition 6 requiring that the licensee not resume operation after the required eddy current examinations until the Director, Office of Nuclear Reactor Regulation, determines in writing that the results of such tests are acceptable, remain in effect in accordance with their terms.
V.
Copies of the above referenced documents are available for inspection at the Comissien's Public Document Room at 1717 h Street, N.W., Washington, D. C. 20555, and are t,eing placed in the Comission's local public document room at the Document Department, University of Wisconsin, Steven's Point Library, Stevens Point, Wisconsin 54451.
VI.
Any person whose interest may be affected by this Order may within twenty days of the date of this Order request a hearing with respect to this Order. Any request for a hearing shall be addressed to the Director of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Comission, Washington, D. C.
20555 with a copy to the Executive Legal Director at the above address.
If a hearing is requested by a person who has an interest affected by the order, the Comission will issue an order designating the time and place of hearing. Any such request SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER.
.. In the event a hearing is held, the issues to be considered at such hearing shall be:
1.
Whether the facts stated in Sections II and III of this Order
, rovide an adequate basis for actions ordered; and p
2.
Whether the license should be modified to include the conditions set forth in Part IV of this Order.
FOR THE NUCLEAR REGULATORY COMMISSION i
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Edson G. Case, Acting Director Office of Nuclear Reactor Regulation
Attachment:
Staff Safety Evaluation Report dated April 4,1980 Effective Date: April 4,1980 Bethesda, Maryland I
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i' SAFETY EVALUATION REPORT RELATED TO POINT BEACH UNIT 1 STEAM GENERATOR TUBE DEGRADATION DUE TO DEEP CREVICE CORROSION April:4, 1980
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INTRODUCT, ION In accordance with the Confirmatory Order dated November 30, 1979, Point Beach Unit 1 was shutdown on February 29, 1980 for steam generator hydrostatic testing and eddy current inspection after having completed the authorized operating period of sixty (60) effective full power days (EFPD's) since the restart subsequent to the October 1979 steam generator inspection. The evaluation herein provides an update of the SER issued in support of the Confirmatory Order to reflect the operating experience at Unit 1 since the Order was issued, and the results of the steam generator inspection obtained during the February 29, 1979 outage.
The background information and results of two consecutive inspections (August and October,1979) as discussed in the November 30, 1979 SER are inccrporated into this evaluation by reference.
BACKGROUND CONFIRMATORY ORDER DATED NOVEMBER 30, 1979 Inservice inspections of the Point Beach Unit 1 steam generators perforr.ed during
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the August and October 1979 outages indicated extensive general intergr.inular attack (IGA) and stress corrosion cracking on the external surfaces of the steim generator tubes wit' i the thickness of the tubesheet (generally referred to as " deep crevice corrosion").
In view of these findings and of the apparent high rate at which this corrosion phenomenon was developing, the licensee agreed to certain conditions to assure safe operation of Unit 1 for a period of sixty (60) effective full power days.
This comitment was formalized by a Confirmatory Order dated Novemt,er 30, 1979, amending the Operating License to include, in part, the following conditions:
1.
a) Hydrostatic testing to be performed within 30 EFPD's.
b) Hydrostatic testing and eddy current inspection within 60 EFPD's.
Submittal of the proposed eddy current inspection program for NRC staff review.
Eddy current inspection results also to be submitted, with no resumption of power until the Director, Office of Nuclear Reactor Regulation determines in writing that the results are accept-(,
able.
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More restrictive limits on primary to secondary steam generator leakage.
3.
More restrictive limits on primary coolant activity.
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Unit i not to be operated wit 1 more than 18% of tubes plugged in either of the steam generators.
While not covered under terms of the Confirmatory Order, the licensee implemented additional measures in an attempt to retard further tube degradation. These measures included 1) a crevice flushing program to remove harmful chemicals from the tubesheet crevices, 2) reduced operating temperature and pressure, 3) continued close surveillance of feedwater chemistry and condenser tube leakage, and 4) sludge lancing to be performed within 12 months of the return to pour.
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. DEFECTS AT OR ABOVE TUBESHEET The Safety Evaluation issued in support of the November 30, 1979 Confirmatory Order reflected thestaff's understanding that the extensive degradation observed during the August and October 1979 inspections involved general intergranular attack and cracking within the tubes' set crevices, exclusively.
Subsequent to the Co.firmatory Order, however, the staff became aware of five (5) tubes with defect indications at or above the tubesheet which had not been addressed in the November 30 SER.
In response to our request, the licensee submitted by letter dated December 21, 1979 additional oatails regarding the defects in these five tubes andan evaluation of their significance.
The licensee reviewed the single frequency eddy current test results since 1975 for the subject five tubes and compared the signals of these past inspections to the same frequency signal obtained during the multi-frequency inspection in October 1979. This comparison ~ showed that the signals have not
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changed through three or four inspections since 1975. On the basis of this review the licensee concluded that the defects observed in October 1979 at or above the tubeshest have remained essentially unchanged since at least 1975 and occurred as a result of earlier thinning or cracking rather than to the intergranular attack phenomenon currently being experienced in the tubesheet crevice area and which was only first observed :n November,1977.
In response to our request, the licensee submitted by letter dated December 21, 1979 additional details regarding the defects in these five tubes and an evaluation of their significance.
Based upon our review of this submittal and a subsequent conference call with the licensee on December 22, 1979, we concluded that (1) the eddy current indications at or above the tubesheet, which were observed during the October 1979 inspection, are old defects, possibly due to wastage or stress corrosion cracking, which were active mechanisms in 1975 and eerlier, (2) these indications are not related to the active phenomenon of general intergranular attack and cracking currently
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being experienced in the tubesheet crevices, and (3) the staff conclusions set forth in the November 30, 1979 SER remained valid and that the unit could continue to be safety operated under terms of the Confirmatory Frder.
Nonetheless, we e
have continued our investigation into the significance of the defects found at or above the tubesheet, particularly with regards to eddy current capabilities to detect thGe defects and their safety significance.. This matter is addressed l
in further detail in this evaluation.
OPERATING EXPERIENCE SUBSE00ENT TO THE CONFIRMATORY ORDER Following the issuance of the Confirmatory Order,
- int Beach Unit 1 was returned to power on December 1, 1979.
On December 11,1979, Unit 1 experienced a rapid increase in primary to secondary leak rate, to 260 gpd, and was forced to shutdown under terms of the Confirmatory Order. The source of the leak was identified as one leaking tube and two leaking plugs in steam generator B.
Although not required by either the Technical Specifications or the Confirmatory Order, the licensee perfomed multifrequency eddy current examinations in both the A and B steam generators. A total of approximately 1900 tubes were inspected. The inspection bounded all areas of previously observed deep crevice corrosion by at least one row and column of tubes.
The inspection boundaries were expandei whe1 Nw indications were observed
near the boundary.
A set of randomly selected tubes outside the boundaries were also inspected.
Representatives from the NRC staff nd consultants were at the site on December 16, 1979 to observe the inspection ir. progress. As a result of this inspection, twenty (20) tubes were plugged in steam generator A and fifteen (15) tubes were plugged in steam generator B.
None of ?.he observed indications occurred at or above the top of the tubesheet. The inspection program and results were formally document.ed in Licensee Eveat Report 79-021,0IT-0 dated December 22, 1979.
Prior to resuming power operation, 2000 psid primary to secondary and 800 psid secondary to primary hydrostatic tests were performed.
No tube failures or addi-tional leakage resulted from these tests.
Based upon our review of the December 11 tube leak occurrence and the inspection results we concluced that the conclusions reached in the November 30, 1979, SER remained valid and that the operating restrictions imposed by the Confirmatory Order continued to provide adequate assurance of safe operation.
Point Beach Unit 1 was returned to power on December 22, 1979 and operated to the completion of its authorized 60 EFFD operating period (on February 24,1980) wi th only a very minor, but equivalent to a constant 30 gpd primary to seccndary leak.
This was within the trace amount of equivalent leakage normally experienced at tnis unit.
s MARCH 1980 INSPECTION RESULTS FIRD EDDY CURRENT TESTING The eddy current testing (ECT) program implemented during the March 1980 steam generator inspection was submitted for NRC staff review by letter dated February 26, 1980.
This program was modified to incorporate NRC staff comments.
ECT of 100". of the tubes in regions of previously observed deep crevice corrosion activity (including the kidney shaped central bundle region) was performed within boundaries bounding previously observed defects by at least one tube rcw and column. Where defects were observed to occur at the boundary, the inspection was expanded to
,(-
bound these defectives by one tube row and column. An additional 3" rancom sample was inspected on the cold leg side and also among tubes on the hot leg side in areas not being 100". inspected.
Representatives of the NRC staff were on site during the inspection to monitor the inspection as it proceeded, and to facilitate timely decisions from NRC/NRR rgarding the need for additional inspection or tube pulling for laboratory examinat.r,n.
I i
Multifrequency eddy current testing (ECT) conducted in accordance with the approved program revealed 18 defect indications on the hot leg side in steam generator A 24 defect indications on the hot leg side in steam generator B.
In addition,3 and tubes in S.G. B and 6 tubes in S.G. A were found with undefinable indications within the tubesheet. On March 31, a hydrostatic test conducted after the ECT inspection revealed two tubes leaking at approximately 2 drips / minute and two wet plugs in S.S. B.
Following plugging of these tubes and repair of the wet plugs a second l
hydrotest revealed another leaking tube in S.G. B which was plugged.
Table I summarizes the ECT indicated defect depths in the two steam generators.
Table II sunnarizes the elevat.f on of the defect indications above the lower, primary surface of the tubesheet.which is about 23 inches thick.
Some defects affected several inches of tube length and one tube had indications running from t:
tube expansion at the primary surface of the tubesheet to approximately one inch below the upper, secondary tubesheet surface.
elevations reched by each defect.The elevations indicated in Table II are the highest
{
\\
e TABLE I ECT INDICATED DEFECT DEPTHS DEFECT DEPTH IN NUMBER OF TUBES PERCENT OF TUBE WALL S.G. A S.G. B
'90 to 100 5
3 80 to 89 7
7 70 to 79 2
7 60 to 69 3
3 t
50 to 59 2
40 to 49 1
2 TABLE II ELEVATION OF ECT DEFECT INDICATIONS DISTANCE ABOVE THE NUMBER OF TUBES PRIMARY TUBESHEET SURFACE (INCHES)
S.G. A S.G. B l
0-4 1
l 5-9 2
10-14 2
2 15-19 8
6 20-21 8
12 I
1/2" AB0VE SECONDARY T.S. SURFACE
s
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. No defective tubes were discovered outside of the central bundle region on the hot leg side nor anywhere on the cold leg side of either steam generator.
Tables I and II in Appendix I provide a tube by tube evaluation of ECT indicated defect depths and eleva. ions and results of re-evaluations of ECT tapes from previous inspections for each defective tube.
Study of these tables reveals that 15 tubes in steam generator A and 4 tubes in steam generato. B had the same ECT indications but were overlooked in either the December or the December and October 1979 inspections. ' All of the tubes with defect indice.tions were plugged except those that were removed for laboratory examination. Ali the ECT indicatiom were of small amplitude and indicate very small volume d2fects.
TUBE PULLING AND LABORATORY EXAMINATI0ri In their February 26, 1980 submittal the licensee committed to remove a tube from f
the Unit 1 steam generators if one was found with an eddy current testing indicated defect at o' aoove the top of the tubesheet, such as were observed in five tubes during the October 1979 inspection. The primary interest in removing this type of tube was two fold:
(1) to determine if the intergranular attack occurring within the tubesheet cmices is resulting in tube degradation at or above the upper secondary surface of the tubesneet and (2) to correlate field ECT with laboratory examination of the defects.
As indicated in Table II one tube was discovered in steam generator B with an indication aporoximately 1/2" above the top of the tubesheet. This was tube R19-C37 and the indication was 58% deep.
In accordance with their comit-ment, this tube was removed from'the steam generator for laboratory examination.
In addition, the NRC (after a review of the ECT results) required removal of two other tubes for laboratory examination.
These were tubes R30-CM which had a 47%
indication acproximately 21" above the primary face of the tubesheet and tube R26-C53 which had a 86% indication acproximately 18" above the primary face of the tubesheet.
Removal of these tubes was intended to provide additional data regarding the extent and magnitude of IGA and the accuracy of ECT. The tube removal procedures extended the outage time approximately six days and resulted in approximately an additionall55 manrem exposure.
LABORATORY RADIOGRAPHY AND EDDY-CURRENT TESTING Radiography and ECT were performed on all three of the removed tube specimens by Festingaouse at their Pittsburgh R&D facility.
As a result of the pulling process the original 22.1/2" length of tt.be R30-C41 within the tubesheet was elongated to approximately 24-3/4".
This measurement was based on the ring left on tne tube at the top of the tubesheet.
Radiography of the removed tube revealed many defect indications in the region up to 23-1/4" from the tube end.
Many ECT indications existed up to 23-1/2" from the tube end.
No radiographic or ECT indications existed at or above the ring marking the top of the tubesheet.
The laboratory ECT examination indicated an approximately 70 to 80% defect based on evaluation of the single frequency (400 KHZ) signal, located 23-1/2" from the tabe end.
Based n the elongation caused in the tube removal process, 23-1/2" corresponds to a,.proxinately 21.3" from the tube end in the unstrained tube.
The field ECT indicated a 47% defect at 400 KHZ approximately 21" from the tube end.
Field evaluation of the defect based on the multi-frequency signal estimated the defect depth in the same 70% to 80% range as obtained in the laboratory (at 400 KHZ) in the absence of tubesheet interference effects.
Defect depths are reported based on the single frequency signal when possible since it is the technique currently approved by the ASME Code.
crevice to approximately 25-7/16".The pulling of tube R25-C53 ' elongate Radiography of the removed tube revealed many defect indications in the region up to approximately 19.8" from the tube end as w as a single defect 25" above the tube end.
indications up to 19.8" from the tube end.
Eddy current testing revealed many defect Eddy current testing also reaaled two 90% defects located approximately 7/16" and 2-7/16" below the tubesheet ring radiographic or ECT indications existed at or above the ring marking the top of th No tubesheet.
None of the above laboratory ECT indications for tube R26-C53 were specifically
(
identified in the field.
or made worse during the tube pulling operation.Some of the indicated defects ma
" Squirrel" indications (minor disturbances in the ECT signal of underterminable origin) were observed in the field over the full length of tube within the tubesheet.
through laboratory ECT the 86% ECT indication observed in the field 18" abc>
tube end, since this corresconded to one of the locations where the tube broke during pulling.
of the fractography analysis of the fracture surface as part of a which the licensee has cormiitted to submit by April 30, 1980.
Tube Ri9-C37 was of particular interest because of the field ECT indication of a 58% defect located approximately 1/2" above the tubesheet.
Unfortunately, when as there was on the other two tubes which were removed.the tube within the tubesheet experiencesa different load and elongation during the remova process than the section of tube above the tubesheet, the exact location of the top of the tubesheet relative to the tube cannot be directly quantified.
Radiography and ECT of the removed tube revealed many defect indications in the
~
reob, up to 23,75" from the tube end.
Radiograpny also showed crack like indica-tbs approximately 24-3/8" above the tube end and ECT indicated an approximate 60% defect 24-1/2" above the tube end.
60% indication.
No ECT ind' cations were observed above the Although the fq% laboratory ECT indication corresponds well with the 58% field ECT indication,its elevation cannot be directly correlated to the field indications because the location of the top of the tubesheet is not identifiable.
Calculations based on strains in the other tubes which were removed indicate that this d would have been inside the tubesheet.
elevation in the tube, its depth corresponds well to the field ECT depth and it
.:ould be the defect of interest given the non-uniform straining of tne tubes during removal.
Metallocrachic Examinations Metallographic examination consisted primarily of photomicrographs (PM) to determine at-what elevation IGA existed in the tubes.
s i
7 For tube R30-C41 PMs were prepared for sections centered on the top of the tubesheet and aporoximately 0.35" below and 0. 45" above the top of the tubesheet.
In each of these regions PMs of 50 and 200 power magnification were made.
The 200 power PMs were centered on the region in the 50 power photomicrographs indicating the greatest surface irregularities.
For the section of tube below the top of the tubesheet the PMs showed shallow grain boundary separation on the order of 0.0025" maximum.
At the top of the tubesheet, shallow surface separation was observed affecting grain boundaries-to just over 0.001" in depth.
Similarly above the top of the tubesheet j
surface separation of the grain boundaries was observed to a depth of approximately O.001 inches.
Extensive general IGA as is occurring deeper in the tubesheet crevi'.e was not observed in any of these regions.
Photomicrographs were also prepared for tube R25-C53. Again the PMs were centered about the too of the tubesheet and approximately 0.4" below and 0.2" above the top of the tubesheet.
The section below the top of the tubesheet showed shallow srain boundary separation penetrating approximately 0.002" maximum.
The region centered about the top of the tubesheet showed no grain boundary separation althougn some surface irregularities penetrating less than 0.001" existed. Above the top of the tubesheet some areas of grain boundary separation penetrating approximately 0.003" were observed.
Extensive general IGA as is occurring deeper in the tubesheet crevice was not observed in any of these regions.
Five photomicrographs were made of tube R15-C39. One was centered on the 60%
defect described earlier while the other four were centerec approximately 1-5/8" and 3/4" below and 1" and 1-3/4" above the defect.
The two sections below the defect showed IGA penetrating to depths of nearly 0.004".
Photographs of the tube surface at the defect show a crack running less than approximately 1/2" lcngitudinally then turning and running less thar. approximately 1/4" circumferentially.
Photo-micrographs of a section made through the defect show a crack reaetrating appreximately 0.017" surrounded by 'ucalized IGA.
The longitudinal section made for the PM may not have included the deepest section of the crack.
Section 0 above the defect indicates one localized area of grain' boundary separation approximately 0.001" deep and section E above tbe defect shows no grain boundary separation but some
\\uf shallow surface irregularities less than 0.001" in depth.
1 PROPOSED CONDITIONS FOR CONTINUED OPERATION The licensee has proposed the following conditions to allow continued operation of Point Beach Unit 1.
1.
Within 90 EFPD, a 2,000 psid primary-to-secondary hydrostatic test and a 800 psid secondary-to-primary hydrostatic test will be performed. An eddy current examination consisting of about 1,000 tubes in the central region of the hot leg in each steam generator and 3% of the remaining tubes outside this area will be performed.
2.
Primary coolant activity fcr Point Beacn Unit I will be limited in accordance with the provisions of Sect 1ans 3.4.8 and 4.4.8 of the Standard Technical Specifications for Westinghouse Pressurized Water Reactors, Revision 2, July 1979, rather than Technical Specification 15,3.1.C.
3.
Close surveillance of primary-to-secondary leakage will be continued and the reactor will be shutdown for tube plugging on confirmation of any of the following condi tions:
s
\\
.. Primary-to-secondary leakage of 150 gpd (0.1 gpm) in either steam generator; a.
b.
Any primary-to-secondary leakage in excess of 250 gpd (0.17 gpm) in either steam generator; or An upward trend (average over a three-day period) in primary-to-secondary c.
leakage in either steam generator in excess of 15 gpd (0.01 gpm) per day, when measured primary-to-secondary leakage is above 150 gpd in that steam generator.
C.
The reactor will be shutdown, any leaking steam generator tubes plugged an eddy current examin tion as described in Item 1.,
above, will be performedif leakage due to crevice corrosion in either stea
,and stated in Technical Specifications 15.3.1.D.
5.
associated parameters (f.e., overtemperature si and
~,-
trip;oint) with the limits indicated in the Safety Evaluation Report appended to your letter of January 3,1980.
On return to power operation, the licensee proposes to continue the fo program to assist in retarding further tube degradation:
Unit I will be operated at a recuced reactor coolant system hot leg temp a.
b.
Continue close surveillance of feedwater chemistry conditions and condens tube leakage.
Perform sludge lancing within nine montns of returning to power.
c.
EVALUATION 8
ECT pROGRA" ~ SULTS, AND CAPABILITIES k~
Members of the NRC staff and their consultant from Oak Ridge National Labora on site during the inspection to review the testing and evaluation techniques.
Eddy current testing examinations were conducted in accordance with the progr proposed in the Itcensee's February 26, 1980 submittal and approved with comment, by the NRC.
viously observed and was expanded in any areas where new The random inspection of peripheral hot leg tubes and cold leg tubes revealed no deep crevice corrosion.
Therefore, the inspection performed is adequate to ensure that the great majority of tubes with deep crevice corrosion have been removed from service by plugging.
The March 1980 ECT results show a marked reduction in the number o dicated defacts compared to the August and October 1979 inspections.
fifteen of the 24 ECT indicated defects in steam generator B and 6 of the 18 ECT In addition, indicated defects in steam generator A were shown to exist previously through re-examination of the ECT tapes from previous inspections.
Thus, the number of new defects discovered in this inspection is smaller than the raw data indicates.
inspection results suggest that some of the remedial actions taken by the licensee The following the October 1979 inspection, paiticularly the lower temperature operation may be succeeding in retarding the rate of further deep crevice corrosion, especially since the time of the December 1979 outage.
s
_g.
r As discussed in our November 30, 1979 SER the accuracy of the eddy current technique is somewhat diminished in the tubesheet region anc cannot be fully relied upon to detect every tube degraded by deep crevice corrosion.
This apoears to be particularly true for tubes subjtet to general IGA, but which do not contain cracks.
Partially through wall cracks f significance are generally detectable, even in the tubesheet
~
region, with ECT. As experience has shown, however, very small volume defects which in turn produce very small amplitude ECT signals may be easily overlooked (as was the case with the 19 tubes above).
Our evaluadon of the safety significance of IGA 7.ad stress' corrosion cracking occurring within the thickness of the tubesheet is discussed in our November 30, 1979 SER which is incorporated into this SER by reference.
With regard to the tubes observed during the October and March inspections to contain defects at or slightly cbove the top of the tubesheet we have concluded that multifre-quency ECT can detect defects of a significant size to threaten tube integrity during normal or postulated accident conditions. All of the defects discovered at or above the top of the tubesheet are small amplitude, smaD volume defects. Assuming the defects at or above the tubesheet to be wall thinning (wastage related), rough estimates of the size of the defects were mace by the staff based on comparison with the ECT signatures from the ASME Code calibration standard.
These estimates show that if these defects are wastage related, the volumes of these defects are very small compared to what is necessary to burst or collapse the tube under postulated accident conditions, as determined by independent tests sponsored by NRC (NUREG/CR-0718).
In the case of tube R19-C37 which exhibited a field ECT. indication of 58% approximately 1/2 inch above the tubesheet, the laboratory examination indicates that the defect indication observed in the field is most likely a crack.
NRC sponsored burst and collapse tests (NUREG/CR-0718) have been performed on free standing tubes with EDM notches (simulating a crack) of up to 85-90% (through wall) in death.
The results indicate the lower bound burst strength to exceed the maximum primary to secondary pressure differentials during normal operation or postulated accidents for notches (cracks) ranging to about 1 inch in length.
It shoulo be noted that the burst strength of a tube containing a crack defact slightly above or below the top of the tubesheet is considerably higher than for free standing tubes, because of the re-
/\\~
straint against radial expansion of the tube provided by the tubesheet. The above tests indicated a collapse failure to be a much less lim: ting failure mode than a burst failure mode for free standing tubes during postuinted accidents.
Cracks of sufficient size to cause a b st or collapse failure unter postulated accidents are S
considered by the staff to be well within the detectable capability of the multi-frequency eddy current technioue, regardless of the location of the crack relative to the too of the tubesheet.
Tube Removal and Laboratory Exam Labor! tory radiography and ECT confirm the position taken by the staff that general IGA may net be detectable in the crevice of. the tubesheet until it is severe enough for proferential crack growth to occur.
Detection of defects below the top of the tubesneet by laboratory examinations is due partly to irureased capability of ECT without the influence of the tubesheet and partly to the creation of new or the opering of old defects during the removal process.
Laboratory radiography and EC1 confirmed the absence of defects above the tubesheet in tubes R30-C41 and R26-C53.
Unfortunately the top of the tubesheet could not be identified on tube R19-C37.
I s
s s
1
. However, assu' ming that the upper most defect detected in the tube is the defect which was identified by field ECT, there is a good correlation between the laboratory and field ECT. More importantly, the defect which was detected was small enough so as not to jeopardize tube integrity.
Primary-t& secondary and secondary-to-primary hydrostatic tests conducted on March 6 revea' one tube (R23-C44) which exhibited a slight leak at a rate of 3 drips per minut and one wet pluc in a previously plugged tube (R23-C50) both in S.G. B.
No tube ruptures occurred.
The defect found by ECT just above the tubesheet in tube R19-C37 in S.G. B withstood the simulated accident pressure differentials. This provides additional support to our previously sitated conclusion that multifrm..ency ECT can detect defects at or above the top furface of the tubesheet which 'would jeopardize tube integrity during normal coerating or postulated accident conditions.
The stafr wants to emphasize that as inspection techniques with increased capabilities, such as multifrequency ECT, are developed, that many small volume defects which previously went undetected will now be found. These defects must be evaluated in the context of the magnitude of defecta shich jeopardize tube integrity during nomal or postulated accident conditions. At aspection techniques become more capabic, correspondingly more discriminate cr teria must be established.
Many plants which i
have not been inspected with multifrequency ECT are going to show new defects when multifrequency inspections are performed.
These results must be dealt with rationally and requirements for tube inspection, plugging, and removal must be carefully aoplied.
METALL0 GRAPHIC EXAMINATIONS Members of the NRC staff and their consultant from Brookhaven National Laboratory met with representatives from.iEPC0 and their Westinghouse consultants in Pittsburgh on March 28, 1980 to review results of the metallographic examinations.
Review of the photomicrograchs described earlier revealed no general IGA similar to that occurring within the tubesheet crevice above the top of the tubesheet in tubes R26-C53 or R30-C41.
Shallow grain boundary separation on the order of two grains or less existed on all photomicrograohs of these tubes.
Shallow grain boundary dissolution of this nature can result from several mechanisms including previous operating environments or tube oickling during manufacturing.
This grain coundary separation is much less
(
severe than that occurring within the tubesheet. The staff has concluded that the shallow grain boundary dissolution at and above the top of the tubesheet is not significant in terms of cube integrity. Metallographic examination of tube R19-C37 revealed stress corrosion cracking and shallow IGA of the tube near the top of the tubesheet. Re-evaluation of past ECT tapes showed that this defect existed as far back as 1976 but was overlooked using single frequency ECT. The nature of the crack is similar to that of stress corrosion cracks which occurred during previon operating periods.
The staff believes that this is an old defect which has not significantly changed since 1976.
CONCLUSIONS
',oed on the information presented abwe the staff has reached the following con-clusions:
- 1) The inspection and tube plugging performed has been adequate to ensure the great majority of defective tubes have been removed from service.
- 2) Multiple frequency eddy current testing used to perform the inspection is capable of detecting defects near the tubesheet and tube support plate interfaces which would jeopardize integrity of the tube during nomal operation or postulated accident conditions.
. -. _... _.... _. _ _. -. -. ~ _..........
s,
.. k
- 3) Hydrostatic tests simulating postulated accident conditions performed prior to returning to operation will identify any significant defects overlooked during ECT examination.
4)
Intergranular attack at and above the top of the tubesheet as observed in the removed tube samples is extremely shallow and poses no threat to tube integrity at or above the top of the tubesheet.
- 5) Based on 'the number of new defects, the rate of deep crevice corrosion
~
appears to have decreased.
- 6) A maximum 90 effective full power day operating period, prior to the next ECT inspection as proposed by the licensee, will provide adequate assurance that a large number of tubes will not simultaneously rear.h a point of incipient failure.
- 7) Remedial actions proposed by the licensee will continue to mitigate the effects f
of postulated accidents and retard the rate of corrosion.
The staff has determined that the following conditions should be required for continued operation:
I
- 1) Within 90 effecting full power days from the date of this order, a 2,000 psid primary-to-secondary hydrostatic test and 800 psid secondary-to-primary hydrostatic test shall be performed. Also during this plant outage, an eddy current examina-tion shall be performed on tubes in each steam generator, The program shall require such examina,tions of about 1000 tubes in the central region of the hot leg, three (3) percent of all hot leg tubes outside this central region and 3%
of the cold leg tubes.
The Central region shall encompass all areas where deep crevice corrosion has previously been observed.
- 2) Primary coolant activity for Point Beach Nuclear Plant Unit I will be limited in accordance with the provisions cf Sections 3.4.8 and 4.4.8 of the Standard Technical Specifications for Westinghouse Pressurized Water Reactors, Revision
(;
2, July 1979, rather than Technical Specification 15.3.1.C appended to License DPR-24,
- 3) Close surveillance of primary to secondary leakage will be continued and the reactor will be shut down for tube plugging on catection and confinnation of any of the following conditions:
a) Sudden primary to secondary leakage of 150 gpd (0.1 gpm) in either steam generator; b) Any primary to secondary leakage in excess of 250 gpd (0.17 gpm) in either steam generator; or c) An upward trend in primary to secondary leakage in excess of 15 gpd (0,01 gpm) per day, when measured primary to secondary leakage is above 150 gpd.
1
.. 4 The reactor will be shut down, any leaking steam generator tubes ' plugged, and an eddy current examination performed if any of the following conditions are present:
a) Confiraation of primary to secondary leakage in either steam generator in excess of 500 gpd (0.35 gpm); or, b) Any two identified leaking tubes in any 20 calendar day period.
This eddy current program will it as described in item 1.
5.
The NRC Staff will be provided with a summary of the results of the eddy current examination performed under items 1 and 4 above.
This summary will include a photograph of the tubesheet of each steam generator which will verify the location of tubes which have been plugged.
6.
The licensee will not resume operation after the eddy current examinations required to be performed in accordance with condition 1 or 4 until the Director Office of Nuclear Reactor Regulation has determined in writing that the results of such tests are acceptable.
These conditions are similar to those in the November 30, 1979 Order excapt that the approved operating period has been lengthenec frem 50 to 90 effective full power days, and no shutdown to cerform hydrostatic tests are being required prior to the end of the 90 day period.
These conditions differ from the licensees proposal in that the primary to secondary leak rate limits and recuirements for ECT examination are more conservative.
On the basis of our review and evaluation, we conclude that continued safe operation of Point Beach Unit I may be permitted within the stated terms of the Confirmatory 0-der.
e 1
l l
I l
l
s, APPENDIX I -
TABLE I POINT BEACH #1 'A' S/G M.F.
M. F.'
Tube #
Dec.
Oct.
R C
1980 1979 1979 12 19 80%
SAME SAME 19-21" ATE g2,51Jo R651 N.C.
n 7
22 29%/96%
SAME NDD/SAME 12" ATE /17" ATE R251 N.C.
12" ATE /17" ATE R551 18 22 66%
SAME NDD 12-17" ATE R251 N.C.
R551 f1 23 41%
NDD 20" ATE
,R251 R551 7
24 I 83%
MAYBE(?)
NDD 17"-20" ATE NDS R551 R251 8
24 79%
MAYBE(?)
NDD 17"-21" ATE NDD R551 R251 25 45 69%
Squirrels NDD 12"-20" ATE R351 R851 20 48 85%
SAME SAME 21" ATE R251 N.C. R851 '
\\'9 49 90%
NDD 21" ATE R251 17 50 85%
NDD 19" ATE R251 19 50 97%
NDD 11" ATE R251 20 50 97%
NDD 11" ATE R251 12 59 87%
MAYBE(?)
NDO 21" ATE NDD R951 R151 12 61 83%
NDD 17" ATE R151 14 63 83%
MAYBE(?)
19" ATE Squirrels R151
0'
', 0 POINT BEACH #1 'A' S/G M.F.
M.F.
Tube #
Dec.
Oct.
R C
1980 1979 1979 15 66 60%
18" ATE 8
27 Squirrels SAME 15-20"' ATE R251 N.C.
15 28 Squirrels No 21" ATE Squirrels lR251 28 34 Squirrels SAME 18-21" ATE R251 N.C.
28 35 ' Squirrels SAME 17" ATE R251 N.C.
i 20 41 91 %
NDO 19" ATE R351 25 43 73%
SAME Very S.V.
17" AIE N.D.D.
f351 f751 11 46 hquirrels l
SAME p2"-21" ATE f351 29 52 '$quirrels SAME f4" ATE R151 N.C.
p.
s APPENDZX I TABLE II B S/G INLET POINT BEACH #1 M.F.
M.F.
S.F.
Tube #
Dec.
Oct.
Aug.
R C
1980 1979 1979 1979 18 25 75%
SAME Changed NDD 18" ATE R151 No R651 R551 change 13 26 73%
SAME SAME NDD 21" ATE R151 N.C.
R651 N.C.
R551 13 33 71 %
SAME Changed NDD 20" ATE R151 N.C.
R651 R552 24 91 %
SAME SAME NDD 11" ATE R151 N.C.
R651 N.C.
R552 20 35
- 587, SAME Changed NDD 21" ATE R151 N.C.
R351 R552 8
37 89%
NDD 5" ATE R151 19 37 58%
SAME SAME NCD 1/2" ATS 53%
R351 N.C.
R552 R151 N.C.
10 41 70%
SAME NDD 21" ATE R251 N.C.
R751 R651 10 41 47%
SAME Some~ Change NDD 21" ATE R251 N.C.
R751 R651/R151 30 42 48%
SAME Changed NDD 21" ATE R251 N.C.
R751-R151 22 46 76%
SAME NDD 15" ATE R251 N.C.
R351 24 48 84%
Changed NDD 12" ATE R251 R351 R652 30 48 85%
SAME SAME NDO 21" ATE R251 N.C.
R951 N.C.
R652 25 49 84%
Changed NDD 5" ATE R251 R351 R652 20 51 99%(?)
SAME NDD 16" ATE R251 N.C.
R351 R652 23 54" 86%
Full length some are new R351 R251 l
~
B S/G INLET POINT BEACH #1 M.F.
M.F.
S.F.
Tube #
Dec.
Oct.
Aug.
R C
1980 1979 1979
.1979 23 57 56%
NDD 17" ATE R251 21 58 83%
SAME l
21" ATE R251 14 59 75%
NDD 21" ATE R251 21 63 62%
SAME NDD 21" ATE R351 R1051 2
67 66%
NDD i
21" ATE R351 R1051 2
72 92%
SAME NDD Top of Roll R351 N.C.
R1051 25 53 86% (New) 18" ATE 3D 43 Souirrels SAME SAME 21" ATE R251 R751 26 53 Squirrels NDD Full T.S.
R251 25 55 Squirrels NDD Full T.S.
R251
\\. '
22 63 Squirrels SAME SAME 21" ATE R251 R1051 22 64 Squirrels SAME No 20" ATE R351 Squirrels R1051 25 55 74% (New) 15" ATE J
s
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i e,e
/
'o UNITED STATES
[ h d ~,j NUCLEAR REGULATOHY COMMISSION
[N g
E WASHir.GTON, O. C. 20555
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8 August 8, 1980 c kD g
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Docket No. 50-266 ATTACHMENT 6 I
y,,J:
'n
- ' 7L,y Mr. Sol Burstein f%e Executive Vice President Wisconsi,n Electric Power Company
The NRC staff has reviewed the results of the Point Beach Nuclear Plant, Unit No. 1, steam generator tubes inspection which was submitted by letter dated August 5, 1980 in accordance with the April 4, 1980 Modification of e
November 30, 1979 Order.
Based on the results of that review and for the reascns stated in the attached Sar cy Evaluation Report, we find that the results are acceptable and, thus, it is not necessary to place any further restricticn on the resumption cr' operation of the Point Beach Nuclear Plant, Unit No. 1.
We uculd like to emphasize that all the conditions of the November 30, 1979 Order and the January 3, 1980 Modifications to the November 30, 1979 Order remain in effect in accordance with their terms.
It is our understanding that Unit 1 will be removed from service November 1980 for a refJeiing outage, and that during that outage, hydrostatic tests and addy current examinations of 100 percent of all unplugged steam generator tubes, will be performed.
Pursuant to 10 CFR 50.54(f) we request that you provide the NRC, within 15 days of your receipt of this letter, your plans and schedule for these steam generator tube inspections. Following the staff's review of your proposed plans and schedule, consideration will be
.k given to the necessity of issuing an Order confirming your program or modifying your program to establish additional requirements.
If you have any questions on this subject, please contact us.
Sincerely,
/QC.
/
Edson G. Case, Acting Director Office of Nuclear Reactor Regulation
Enclosure:
Safety Evaluation cc w/ enclosure:
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SAFETY EVALUATION REPORT RELATED T0-POINT BEACH UNIT 1 STEAM GENERATCR TUBE DEGRADATION DOCKET NO. 50-266 Introduction In accordance' with the Order dated April 4,1980, Point Beach Unit I was shut down on July 25, 1980 for steam generator hydrostatic testing and eddy current-l inspection after having completed ninety (90) effective full power days (EFPD's)
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of operation s'. ice the restart following the March lgR steam ger.erator inspection.
The evaluatm. herein provides an update of the SER's issued in support of the Confirmat :*-
1d Supplementary Orders, respectively, to reflect the recent cperating.mperience at Unit 1 and the results of the August 1980 steam generator inspection.
The background information and results of previous steam generator inspections >3 discussed in the November 30, 1979 and April 4,1980 SER's are incorporated into this evaluation by reference.
Sackcround and Discussion Inservice inspections of the Point Beach Unit 1 steam generators performed during the August and October 1979 outages indicated extensive general intergranular attack (IGA) and stress corrosion cracking on the external surfaces of the steam generator tubes within the thickness of the tubesheet (generally referred to as " deep crevice corrosicn").
In view of these findings and of the accarent high rate at which this corrosion phencmenon was developing, the licenses agreed to certain conditions
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to assure safe operation of Unit 1 for a period of sixty (60) effective full power days.
This commitment was formalized by a Confirmatory Order dated November 30, 1979, amending the Operating License to include, in part, the following conditions:
1.
a)
Hydrost tic testing to be performed within 30 EFPD's.
b)
Hydrostatic testing and eddy current inspection within 60 EFPO's.
Submittal of the proposed eddy current inspection program for NRC staff review.
Edoy current inspection results also to be submitted, with no resumption of power until the Director, Office of Nuclear Reactor Regulation determines in writing that the results are accept-able.
2.
More restrictive limits on primary to secondary s:aam generator leakage.
3.
More restrictive limits on primary coolant activity.
4 Uni I not to be coerated with more than 18% of tubes plugced in either of the steam generators.
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.. hile not covered under terms of the Confirmatory Order, the licensee implemented additional measures in an attempt to retard further tube degradation.
These measures included:
- 1) a crevice flushing program to remove har=ful chemicals from the tubesheet crevices, 2) reduced operating temperature and pressure,
- 3) continued close surveillance of feedwater chemistry and condenser tube leakage, and 4) sludge lancing to be performed within 12 months from the return to power.
In.'accordance with'the Confirmatory Order, Unit i shut down on February 29, 1980 after having completed sixty (~50) EFPD's of operation.
The March 1980 eddy current results ' indicated a marked reduction in the number of tubes with indicated defects compared to the Augus and October 1979 inspections. By Order dated April 4,1980, Unit I was repuired to be shut down for steam generator hydrostatic and eddy current inspections after ninety (90) EFPD's. hith the exceptions that the operating period had been changed from 60 to 90 EFPD's, and that no shutdcwn to perform hydro-static tests was required before the end of this period, the conditions of the Confirnatcry Order remained in force under the April 4,1980 Order.
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July-Aucust Steam Generator Insoection Results Subsecuent to the plant shutdown en Jul; 25, 1980, both steam generators were subjected te hydrostatic tests and eddy current examination in accordance with the Acril 4, 1950 Order.
A tubesheet insp(ction during the secencary te primary hycrostatic leak test revealed two "cripping" tube plugs and two " wet" tube plugs in the hot leg side of steam generator A, and one wet tabe plug and one dripping tube (at rate cf one drip per twc minutes) in steam generator 3.
At the time of snutdcwn en July 25, tne Unit 1 steam generators had teen leaking (primary to secondary) at a very low level, aporoximately 20 gpd.
The cripping tube identified in steam generator 3 was inspected to through the U-bend using the multifrecuency eddy current test (ECT) technicue, but only a 45% through wall indication, located three inches above the tube end (within
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the tubesnest thickness), was identified.
A possible explanation suggested by U
the licensee is that the source of the leak may be a small volume defect located in the transition recion of the expanded tube zone (near the bottom of the tube-sheet) which wculd be particularly difficult to discriminate, even with multi-frequency ECT. This tube has subsecuently been plugged.
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The multifrecuency ECT inspection program for both steam generators consisted of an examinatien of 100% of the ubes to the first supscrt plate on the hot lec side, and 3% of the tubes insoected over their entire length (i.e. het and colc' leg). The results of these inspections are summarized as folicws:
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3-ECT Insoection Summary
% Tubes Eddy Current S.G.
Inscected Indications Elevation S.G.A Hot Leg 100%
1 tube - undefinable signal 3 tubes
<20%
Within 3 tubes - 20 to 39%
thickness 7 tubes - 40 to 59%
of 5 tubes - 60 to 79%
tubesheet 9 tubes - 80 to 99%
1 tube - 34%
Top of tubesheet I tube - 34%
h" above tubesheet f
Cold Leg 3%
1 tube - 29%
h" above tubesheet 5 tubes
<21%
1 to 2" above tubesheet l
3 tubes - 60 to 79%
S.G.B Hot les 100%
7 tubes - 40 to 59%
Within thickness of 6 tubes - 80 to 99%
tubesheet I tube - leaker Unknown r\\- -
Cold Leg 3%
None e
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-4 As seen in the summary Table, a total of 28 and 22 tubes in the hot leg of steam generators A and 3, respectively, were identified to contain tubesheet crevice indications; i.e., indications located within the thickness of the tubesheet.
The elevations of these indications range from three (3) inches.
above the tube ends.to in excess of one inch below the top of the tubesheet.
Two (2) addi:tionattubes on the hot leg side of steam generator A were found to contain 34% small volume indications at the top of the tubesheet and one-half inch above the top of the tubesheet, respectively. Six (6) tubes on the cold les side of steam generator A were indicated to contain minor through wall pe'netrations (<3C%) located h to 2 inches above the top of the tubesheet elevation.
No cold leg.tt.cesheet crevice indications were identified which is consistent with previous experience.
Eddy. current tapes from previous inspections dating back to October 1979 are being reviewed by the licensee for each of the tubes found during this inspection (July-August 1980) to contain eddy current indications. For some tubes,-the r
licensee determined that small volume indications were probably present (but i
were not identified by the data cvaluators) in one or more previous inspections by reviewing the previous tapes in close detail over the specific area of interest.
These include ten (10).of the total of 50 tubes identified during this inspection as centaining tubesheet crevice indicaticns, and two (2) tubes in the hot leg of steam generator A fcund to contain indications at one-half inch above the top of the tubesheet.
It is the licensee's evaluation that the eddy current data evaluators were unsuccessful in discriminating these small volume defects because of the low signal-to-noise ratio of the eddy current signal during previous inspections.
However, the licensee's review cf the previous eddy current tapes has established that the majority of the eddy current indications were not previously detectable.
The previcus inspection in March 1980 included a 100% sample of tubes in the central bundle region (" Kidney zone") of each steam generator (approximately 1000 tubes), and a 2% random sample inspection cutside this zone. The. central region where 100% f.spection was performed was defined to enccmpass the region
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of.previously observed activity. However, the results of the latest inspection l'
revealed 24 tubesheet crev1ce indications located up to several tubes beyond the boundary of'this previcusly defined zone that were not inspected in March 1980.
As indicated to the staff during dir.cussions held on August 6,1920, the licensee does not consider these results to be unexpected since the concentration of chemicals in the tubesheet crevices will occur regardless of whethee there is a sludge pile at the surface.
The licensee believes tha, while the sludge pile may contribute chemicals for concentration in the crevice, there is no reason to believe that the crevice corresion will be limited to the kidney zone, since chemicals from the bulk water will also concentrate in the tubesheet crevices.
All 50 tubes with indications in the tubesheet crevice, including the leaking (dripping) tube, have been techanically plugged.
In addition, three tubes were inadvertently clugged. The two dripping plugs in s:eam generator A were wald repaired and the steam generator was subsequently and successfully hydrcstatically l
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leak checked. The two tubes in steam generator A containing 34% indications outside the tubesheet crevice region were left u1 plugged since these indicaticns are less than the 40% Technical Specification plugging limit and these indi-cations appear to have remained unchanged since at least October 1979. The licensee has cor=itted to re-exacining these tubes during the next eddy current inspection.
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To date, approximately 12.2%.'of the total number of stea= generator tubes at Unit I have been plugged, which is well within the 18% tube plugging assumed in the LOCA-ECCS analysis for this unit.
Plans For Continued Oceration Based 'upon the results of this inspecticn, the licensee has concluded that the condition of the Point Beach Unit 1 steam generators has not changed signifi-cantly since the previous inspection in March 1980. The licensee plans r
to return Unit I to service for an additional 90 effective full power days i
until its scheduled refueling outage in early November 1980.
f Evaluation The July 1980 inspection of 100% cf unplugged tubes at the completion of 90 effective full power cays (EFPD) has satis'fied the requirements of NRC's Confirmatory Order, dated April 4, 1980.
Tne 2000 psi primary-tc-secondary hydrostatic test and 300 psi secondary-to-primary test recuired i
by the Aoril 4 Order confirmed that no tubes hac reached a state of degradation that wcule cause a sueden primary-to-secondary leakage during the 90 EFFD operation.
The current multifrequency ECT results, compared with similar ECT results per-for:ed in March 1980 ano December 1979, do not indicate an accreciable increase
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in tube decradation within the tubesheet crevice.
Of the 14 tubes in steam l
generator A containing ECT tubesheet crevice incications and which were pre-v1ously examined in March 1980 and December 1979, nine had tubesheet crevice defects which did not show an increase in defect size.
Regarding thei five (5) tubes that now show a significant ECT indication, but did not show indications previously, we believe tnat intergranular cerrosion attack existed which could not be identified in previous inspections. This had been cemenstrated in the laboratory analysis of tubes pullec in March 19E3 and November 1979.
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k'ith regard to the 18 tubes in steam generator A with new ECT indications within the tucesheet crevice, it should ce noted that the current inspection is the first time that 100% c the tuces in steam generator A have been examined for tubesheet grevice defects since October 1979.
Thus, there is no basis to indicate that 1centirication oT these new tubes reflects a raoid detericration of the Point
- eacn Unit I steam cenerators.
l l
s 2
6-Th'e above analysis applies also to the ECT indications found with the tube-sheet crevices of steam generator B.
The current ECT results for both steam generators show that intergranular corrosion attack has not progressed above the tubesheet. The two tubes in steam generator A with small volume defects at the top of the tubesheet or just above were present during the October 1979 ir.spection and have not shown an increase in defect size. Steam generator 8 had no tubes with defects of this kind.
The. random ECT inspections of tubes an the cold leg side confirm that tubesheet crevice corrosion is confined to the hot leg side of each steam generator.
As was the case during the previous inspection in March 1980, the latest ECT results continue to show a marked reduction in the number of tubes with indicated tubesheet crevice defects relative to the August and October 1979 inspections during which approximately 230 ;ubesheet crevice indications were identified.
In addition, ten (10) of the 50 tubes in both steam generators identified to contain tubesheet crevice indications during the latest inspection have been shown to have been present since at least October 1979 based upon a re-examination of the eddy current tapes from the previous inspections. Similarly, 20 of the 41 tubes identified in March 1980 to contain tubesheet crevice indications were also shown to have been present during the October 1979 inspection. The latest inspection findings continue to suggest that some of the remedial actions taken by the licensee following the October 1979 inspection, particularly the lower temperatore operation, may be succeeding in retarding the rate of tubesheet crevice corrosion.
In this regard, it should be noted that the deep crevice indications first identified during this inspection, but which were apparently present curing the Octcber 1979 inspection, have essentially remained stable since that time without develcping into leaks.
Analysis of the six (6) tube specimens removed from the Unit 1 steam generator during the October 1979 and March 1980 outages has demonstrated that the i
presence of integranular attack within the tubesheet crevice cannot be reliably detected with single or multifrequency ECT until cracks are developed along the grain boundaries. ' Partially through wall cracks of significant size are generally detectable with ECT, even in the tubesheet region. Hewever, very small volume cefects, which in turn result in very small ECT signal-to-l noise ratios in the tube. tet region, may be easily overlooked by ti.e data evaluators. As noted eardier, several of the eddy current indications observed in the current inspection and the March 1980 inspection were apparently present since October 1970, but were not identified at that time.
We believe the licensee's inability to identify the source of the leaking tube in steam generator 5 to be a further example of the difficulties in discriminatir.g very small volume defects in the tubesheet region.
However, we believe tubes with small volume defects (small signal to noise ratio) can generally maintain i
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their integrity during the full range of normal operating and accident conditions.
The safety significance of intergranular attack and stress corrosion cracking within the tubesheet crevices was evaluated in our November 30, 1979 SER.
Based upon our review of the latest inspectica results, the November 30, 1979 evaluation remains valid and is incorporated into this SER by reference.
Conclusions We conclude that the Point Seach Unit 1 steam generator may operate under the conditicns of the November 30, 1979 Order and the panuarf 3,1980 Order without
~ impairment to the health and safety of the puolic for the follow,ing reasons:
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- 1. ' The 100% inspection and hydrostatic t2sts have identified all tubes with significant defects to ensure an adequate margin of safety for the pro-posed period of operation.
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2.
The operating conditions (i.e., reduced pressure and tenperature) during the past 150 EFPD has been successful in retarding the rate of tube degradation.
3.
The cumulative number of tubes pl zgged (12.2%) is well below the 18%
assumed for the LOCA-ECCS analysis.
I Dated:
August 8, 1980 1
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February 13, 1981 ms Docket Nos. 50-266 ATTACHMENT 7 and 50-301 Mr. Sol Burstein Executive Vice President Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53201
Dear Mr. Burstein:
This refers to your letters of December la and December 23,1980 and to discussions held with members of your staff on December 24, 1980 regarding the recent Point Beach Unit 1 steam generator tube inspection.
i As discussed with you on December 24, we completed our review of your inspection results on that date and found them acceptable for continued operation.
Also, based on recent inspection results and operating experience as it stands today, we agree that both your inspection schedule and plans for the next inspection are reasonable subject to the following conditions:
A.
That you conduct a 2000 PSI primary-to-secondary and an 800 PSI secondary-to-primary hydrostatic test as has been done in the past.
B.
That you include support plate flow slots in your next inspection as discussed with you on December 24th.
With respect to your operating conditions identified in your December 18 letter, item 4 (page 3) is not correct.
Our Confirmatory Order for
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Modification of cense dated November 30, 1979, portions of which are i
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still in effect, requires, among other things, that you will not resume operation until we determine in writing that eddy current examinations are acceptable in the event of:
A.
primary to secondary leakage in any steam generator in excess of 500 gpd; or i
B.
any two identiF w leaking tubes in any 20 calendar day period.
Please refer to paragraphs 4, 5 and 6 of this Orf1r for the specific l
requirements.
l In response to your comments regarding your sleeving of tubes you are l
reminded that your Technical Specifications require that any tubes cegraded gre ter than 40% nominal wall thickness are required to be m
taken out of service by plugging and that to sleeve such tubes, in lieu of plugging would require an amendment to your license.
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Mr. Sol Burstein 2-
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We request that you submit the results of your next inspection to us for our review prior to plant start-up as you have done in the past.
A response to this letter is requested within 30 days of receipt.
Sincerely, 3
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Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing cc:
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NUCLEAR REGULATORY COMMISSION
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Before the Atomic Safety and Licensing Board
'*'l1 -L/'
In the Matter of
)
)
WISCONSIN ELECTRIC POWER COMPANY
)
Docket Nos. 50-266
)
50-301 (Point Beach Nuclear Plant,
)
(OL Amendment)
Units 1 and 2)
)
CERTIFICATE OF SERVICE I hereby certify that copies of " Licensee's Motion For Authorization For Interim Operation of Unit 1 With Steam Generator Tubes Sleeved Rather Than Plugged," accompanying Attachments 1-7, and " Licensee's Proposed Form of Memorandum and Order On Licensee's Motion For Authorization For Interin Operation of Unit 1 With Steam Generator Tubes Sleeved Rather Than Plugged", dated September 28, 1981, were served, by deposit in the U.S.
Mail, first class, postage prepaid to all those on the attached Service List, except to those individuals indicated by an asterisk on the Service List, which were hand delivered, on this 28th day of September, 1981.
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Deliss< A.4 tid,4ayf 1
Dated:
September 28, 1981
E UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of
)
)
WISCONSIN ELECTRIC POWER COMPANY
)
Docket Nos. 50-266
)
50-301 (Point Beach Nuclear Plant,
)
(OL Amendment)
Units 1 and 2)
)
SERVICE LIST Peter B.
Bloch, Chairman
- Charles A.
Barth, Esquire %
Atomic Safety and Licensing Office of the Executive Board Panel Legal Director U.S.
Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, D.C.
20555 Washington, D.C.
20555 Dr. Hugh C.
Paxton Kathleen M.
Falk, Esquire
- 1229 - 41st Street Wisconsin's Environmental Los Alamos, New Mexico 87544 Decade, Inc.
302 E.
Washington Avenue Dr. Je!ry R.
Kline*
Madison, Wisconsin 53703 Atomic rafety and Licensing Board Panel U.S.
Nuclear Regulatory commission Washington, D.C.
20555 l
Atomic Safety and Licensing Board Panel l
U.S. Nuclear Regulatory Commission l
Washington, D.C.
20555 Atomic Safety and Licensing Appeal Board Panel U.S.
Nuclear Regulatory i
Commission l
Washington, D.C.
20555 Docketing and Service Section Office of the Secretary U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 l