ML20033C659
| ML20033C659 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/10/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20033C660 | List: |
| References | |
| NUDOCS 8112030592 | |
| Download: ML20033C659 (18) | |
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NUCLEAR REGULATORY COMMISSION l
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WASHINGTON, D. C. 20555 s.,..... /
WISCONSIN ELECTRIC POWER COMPANY I
DOCKET NO. 50-266' POINT BEACH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE f
Amendment No. 56 License No. DPR-24 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated July 2,1981 as mndified by letter dated October 12, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common-defense and security or to the health and safety of t!)e public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable rdquirements have been satisfied.
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- 2. - Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. OPR-24 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 56, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION w{
Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 10, 1981 9
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.i ATTACHMENT TO LICENSE AMENDMENT AMENDMENT'NO. 56TO FACILITY OPERATING LICENSE NO. DPR-24 DOCKET NOS. 50-266 d
Revise Appendix A as follows:-
- Remove Page Insert Page 15.4.2-Ic 15.4.2-1c 9
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. Defect is an imperfection ofTsuch severity:that itLexceeds the j
minimum acceptable tube wall-thickness of-50%.
-A tube.containing 2
-a defect is defective.
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PluatinaLimitisjhel'aperfectiondepthbeyondwhichthetube;
,must -.be removed f rom ' service, because the tube may. become defective -
prior.to.the next' scheduled inspection. The plugging limit is 40%
of the nominal tube wall thickness.'
.B.'Correc'tiheMeasures 2
All tubes that' leak or. have d.egradation. exceeding the, plugging limit shall be. plugged prior to return to power from a refueling 'or inservice inspection condition.*
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C.
Reports-
- 1. JAf ter each inservice examination, the number of tubes pitigged in each steam generator shall be reported to the Commission-as soon as practicable.
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'2.
Th'e complete results of the' steam generator tube inservice inspection shall be included in the Operating Report for the period in which-the inspection was completed.
In addition, all results.in' Category C-3 :
of Table 15.4.2-1 shall be reported to the Commission prior to resumption of plant operation.
3.
Reports shall include:
(a) Number and extent of tubes inspected 4
i (b) Location and percent of.all thickness penetration for each' f
indication a
(c)
Identification of tubes plugged 4
Reports required by Table 15.4.2-1.
Steam Generator. Tube Inspection -
shall provide the information required by Specification.15.4.2.C.2 and a description of investigations conducted to determine cause of the
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- tube ' degradation and corrective measures taken to prevent recurrence. ;
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B.,,In-Service Inspection of Reactor Coolant System' Components Othe'r Than Steam Generator Tubes 1
The in-service-inspection program is generally' based on the recommendations of ASME Boiler and. Pressure Vessel Code, Section~XI, Summer 1971' Addenda, as practical'for.
a plant whose design' and construction preceded issuance of the recommendations.-
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- The commitments herein are made assuming that the' necessary inspection
(* Point Beach Nuclear Plant Unit 1 may be operated at power;with:up to sixL tubes; in'one steam' generator having! degradation exceeding the plugging, limit providedf
- .those tubes have been repaired 1by -insertion of'sleeveslinto the; tubes ' to' bridge
- theLdegraded for ' defective portion. of the. tub'e.The plugging limit is :35% of ; the :
g nominal sleeve wall thickness for; tubes.that'have been repaired by sleeving.
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. Point; Beach Unit 1-
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- 15.4.2-1c1
. Amendment'No. 10, ~ 56 -
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NUCLEAR REGULATORY COMMISSION D
j WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.~ 56 TO FACILITY OPERATING LICENSE NO. DPR-24 WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT, UNIT NO. 1-DOCKET NO. 50-266 4
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a TABLE OF CONTENTS Pa g'e 1.0 Introduction 1
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. 2.0,101scussion 1
2.1 Sleeving Process Description 1
2.2 Structural Verification Analyses 3
2.3 Verification Testing of Sleeve Joints 3
2.4 Verification of " Leak Before Break" 4
-2.5 ;Effect of Proprietary Heating Process 4
on Upper. Alternate Joint Integrity 2.6 ' Discussion.of Corrosion Aspects and S
Verification Testing 2.7 Eddy-Current Test Capabilities 5
3.0 Evaluation 6-3.1 Structural and Leak Tight Integrity' 6
E 3.2 Plugging Limit 7
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-3.3 Integrity of Upper Alternate Joint 7
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3.4 Corrosion Resistance
.8 3.5 Eddy Current In.spectability 8-4.0. ALARA Considerations 9
5.0 Reduced Flow Considerations 10 J
6.0 Conclusions 11
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1.0 INTRODUCTION
By letter dated July 2,1981, Nisconsin Electric Power Cnmpany (licensee)
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submitted an' application for Itcense amendments consisting gof proposed changes to the Technical Specifications for Point Beach Nu' lear Plant Units c
1 and 2.
These proposed. Technical _ Specification changes would allow operation at power of Units 1 and'2 with. steam generatar tubes h&ving degradation exce? ding the plugging limit (40% nominal wall thickness) provided these'
- tubes have been. repaired by insertion of sle2ves into the tubes to bridge the degraded or defective portion of the tubes. The proposed issuance of
- these amendments was prenoticed in the Federal Register on August 7,1981 due to the _ strong public interest on this subject.
The licensee also submitted by. letter dated October 12, 1981, a modification to their proposed license amendment for Unit 1 dated July 2,1981. This modifi-cation proposed Technical Specification changes to allow operation of Unit 1 a't power with up to six tubes.in one steam generator having degradation exceeding the plugging limit provided these tubes have been repaired by insertion of sleeves into the tubes. to bridge the degradated or defective portions of the tubes. The licensee also plans to sleeve six tubes having degradation less than the plugging limit. The licensee's stated reason for _ submitting this modification is to. conduct a demonstration sleeving i
program on Point Beach Unit i during the October 9,1981 refueling outage.
- This demonstration program will utilize two separate sleeving processes and the licensee hopes _ it will provide valuable information and experience for use during their full-scale sleeving program.
This Safety Evaluation documents the results of the NRC staff's review and evaluation of-the licensee's proposed demonstration steam generator tube sleeving program including the environmental and radiation exposure impact.
2.0 DISCUSSION 2.1 Sleeving Process Description The sleeving demonstration program scheduled for the fall 1981 refueling outage of Point Beach Unit 1 is expected to include removal of explosive'and mechanical plugs from previously plugged tubes where degradation had exceeded the plugging limit in the Technical Specifications. All tubes from which plugs have bee'n removed will be inspected with eddy current techniques-throughout their length prior to sleeving. 'Should indications of progression of degradation, or new. indications of degradation be seen outside the proposed sleeved region of the tube, the tube will not be. sleeved, but will be plugged in accordance with the Technical Specification requirements.
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To provide a technical basis for the. proposed sleeving demonstration program,-
the licensee has: submitted Westinghouse Report WCAP-9960 (Proprietary),
-dated September 28, 1981, and entitled, " Point Beach ' Steam Generator Sleeving
- Report for Wisconsin Electric Power Company." The licensee has submitted additional linformation by letters, dated October 9,16,
- 24 'and 26 in response
.to questions byfthe ASLB and the NRC staff. They have also responded to other questions during conference calls with the NRC staff.
The sleeving process consists of installing, inside the steam' genera' tor
= tube, a" smaller diameter tube (sleeve) to, span the. degraded area of the
-parent' tube.. The sleeves are intended to restore the integrity of the degraded tubes by providing a new primary pressure boundary which has been l
sized to the ASME Boiler and Pressure Vessel Code,Section III..-
The< sleeves are fabricated from. thermally treated Inconel 600 tubing to provide'a maximum resistance to stress corrosion cracking. The: sleeves will-be inserted inside the existing tube (mill annealed Inconel 600) and i
' joined to the tube ID at the upper and lower sleeve ends.
The sleeves will span the distance from the tube inlet to' a few> inches above the top of the 2
tubesheet. The Point Beach sleeves are intended'to address the general
.intergranular. attack and stress corrosion cracking which-has been confined to the tubesheet' area.
The sleeves used in the demonstration program will employ two different upper-sleeve joint designs. The " reference" upper joint design.is a structural-
-join' which provides a. leak limiting seal. A functional requirement for
" reference" upper joints is that they must be sufficiently leak limiting.
-such. that the total leakage between the primary and secondary for all the
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sleeves taken together is less than the Technical Specification leak rate Llimit during normal operation.- In addition, total _ leakage must be maintained
'to within tolerable 11mits'during postulated accidents. The acceptance criteria imposed during verification leak testing of the joint is based upon -these total leakage limits divided by the: total number' of tubes eventually planned'for sleeving (approximately 2500 tubes).
The second or " alternate" upper joint design is also a structural joint.
j This joint is fabricated using a proprietary heating process to form a -
L leak tight'_ seal. The lower sleeve ~ joint also provides _ a structural and j
leak tight seal,.but is' not fabricated with the proprietary heat ~ing process.-
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The Point Beach sleeves 'and-sleeve joints are basically similar to.those-at San Onofre Unit l'from ethe standpoint of_ de' sign and joint fabrication techniques. The San Onofre sleeves-have been extensively tested for
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structural, metallurgical, corrosion, and. leak - tight -(or. leak: limiting).
(integrity..
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. 2.2 -Structural Verification Analyses Structural analyses of the sleeved tube assembly are being performed to the. requirements of Section III of the ASME Boiler and Pressure Vessel Code..These analyses are intended to demonstrate adequate fatigue per-formance and structural margins for the full range of normal operating, transients, and accident (e.g., LOCA, MSLB) condition loadings. The structural and fatigue analyses include consideration of stresses in the s1ceved tube assemblies which could result from hourglassing (deformation) of the support plate flow slots, and from flow induced vibration. The analyses have essentially been completed; however, some additional proces-sing of finite element stress data must yet be performed before they can be evaluated against the 3 Sm limit for primary plus secondary stress. The preliminary results submitted by letter dated October 24, 1981, indicate the Code allowables for primary membrane, primary membrane plus bending stress, and fatique usage have been met.
Strength analyses have been performed to establish the minimum wall thickness requirement (or allowable wall degradation) to assure compliance with the Regulatory Guide 1.121 "no yield" criterion under normal operating conditions.
These analyses have also established the minimum wall thickness requirements (and allowable wall degradation) to preclude a gross tube burst under the pressure loadings associated with a postulated MSLB accident, consistent with the Regulatory Guide criterion and the Code limits on primary membrane stress under faulted conditions. The results of these analyses will be used to set the Technical Specification plugging limit for the sleeves.
2.3 Verification Testing of Sleeve Joints The structural analyses of the sleeved tube assemblies are being supplemented by extensive mechanical testing to verify acceptable structural strengths, fatigue performance and leaktight integrity of the upper and lower joints.
T.he test mockups for the lower joint include tubesheet mockups from which the effects of removing both mechanical and exp13sive type plugs have been simulated. _The joints have been formed using the same fabrication techniques and parameters as will be used in the field.
Each of the joints is being subjected to axial load (to simulate loads caused by differential thermal expansion) and pressure cycling tests to veri fy the long term sealing integrity of the joints under the specified operating ~ transients (e.g., heatup/cooldown a'nd plant loading / unloading cycles). Specimens for each type joint will.
also be tested to the maximum pressure and axial load levels expected during postulated accident conditions. For each of the three joint designs, testing has proceeded to as much as the equivalent of'five years of operation with no adverse findings reported to date.. Further testing is in progress and
- will.be continued for an equivalent 35 years of operation.
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Similar, mechanical. tests have been completed'for the San Onofre joints
- to support thirty years operation with the results indicating acceptable structural 'and leak ifmiting performance.
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2.4 Verification of " Leak Before Break" Westinghouse tests indicate that margin to burst exists at the MSLB pressure differential for a.through' wall crack which is leaking ~ at less than the -
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- Technical Specification limit during normal ' operation. The tests indicate' that the required through wall crack length for a tube burst under MSLB conditions is.5 inches, whereas a through wall crack longer than.4-inches will result in leakage in excess of the Technical Specification leakage rate limit during normal operation.
l 2.5 Effect of Proprietary Heating Process on Upper Alternate Joint Integrity 1
The proprietary heating process for the " alternate" upper joint design will. result -in some degradation of the mechanical properties of the sleeve a
. and tube wall material local to the seal between the sleeve and-the tube.
Tensile ' tests of individual San Onofre tube and sleeve specimens following a simulated joint heating process indicated a significant reduction in the ultimate and yield strength at the location where the peak temperature had been reached. This corresponds to the center of the region where the tube and sleeve weald be sealed. As evidenced by variations in hardness and grain size measurements as one proceeds away from this location, heat process effect on the yield and ultimate strength is localized to within 4
the width of the seal. Tensile tests of a number of joint specimens resulted in tensile failures of-the sleeve wall invariably between two t
and three inches below the sealed location, at levels in excess of minimum requirements (Ref.1). Westinghouse has also reported that the stress -
strain curve of the " alternate" upper joint almost duplicates that of virgin Inconel 600 material.
Westinghouse has reported that confirmatory tests for the' actual Point' Beach " alternate" joint configuration have indicated similar results and that the overall joint strength exceeds Code requirements.
Internal pressure tests to three times normal operating pressure, and external pressure tests to 1.5 times the maximum LOCA pressure loading 4
resulted in no failures for the San Onofre " alternate" upper joint specimens.
Similarly, load cycling tests (to simulate pressure plus thermal cycling)
E for the expected number of operating cycles over a 30 year lifetime were' completed with no failures. Similar confirmatory tests are in progress
- for. the-actual Point. Beach configuration, with the exception of the collapse test.-
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- 2'.6-Discussion'of Corrosion Aspect and Verification Testing
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LThe" corrosion:that has occurred lon the outer surface of the tubes has been Lattributed to caustic corrosion resulting from the use of phosphate water Lchemistry in the secondary water with massive phosphate additions ~ and the Lformation of caustics due to. impurities: from. persistent leakyitubes in
.the steam' condenser. The chemistry control progran of tne secondary side.
water was.s itched to an all-volatile 1 treatment in September of 1974, ~ though free hydroxide continued to be present in'the blowdown water until 1978.
'Most of the steam generator tube corrosion and degradation has occurred in the central region of the inlet end of the tube bundle. Some intergranullr stress corrosion cract.ing, wastage, and thinning 11as occurred at a location-
- just above the tubesh tet in.the sludge zone, but the more extensive inter-granular' corrosion has occurred in the tubesheet crevices. Although the
. licensee's tube degradation rate has slowed recently, tube degradation ~
could continue.
We have reviewed the corrosion test program performed in support of the Southern California Edison (SCE) plant, San Onofre Unit 1.
This work was cited by the ifcensee in support of the present application request. The corrosion tests performed were extensive, involving the use of capsule-tests and modified boiler tests in which the environment that existed in San Onofre Unit 1 was simulated and its effect on the sleeved tubes was
- studied. The environment in the tubesheet crevice at Point Beach Unit 1 is similar. An extensive test program was performed studying the effects of caustic on the corrosion resistance and stress corrosion cracking of
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the sleeving _ material. Confirmatory testing of the corrosion and stress-corrosion cracking resistance of both the upper and lower joints of the Point Beach configuration is in progress.
2.7 Eddy Current Test Capabilities _
Eddy current data is provided in the Repair Report to demonstrate the applicability of the conventional bobbin type ECT probe to the inspection
'of the sleeved tube assembifes.
(This data was actually obtained for San Onofre sleeved assemblies.) At the optimum test frequency for theisleeve, i
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J the amplitudes' of the ECT signals ranged from 70% to 100% of those for~
a non-sleeved tube for calibration holes of 40% and 100% throughwall depth, respectively. This data is. indicative of the relative flaw sensitivi;.y outside 'the tubesheet, whereas most of the sleeve length will be located within the thickness of the tubesheet.
The Westinghouse investigation indicates.that within the thickness of the tubesheet the " signal to noise
-ratio" associated with a sleeving defect is substantially more than that associated with a flaw in a non-sleeved tube. Thus, Westinghouse'has con-cluded that the sleeve in the tubesheet region will have a higher degree of inspectability than an unsleeved tube in this region.
The inspectability of the tube wall is of interest at and above the upper sleeve joints. The Westinghouse. study indicates that the amr.litude of the ECT signals for calibration holes in excess of 40% through wall were approximately 50%. of those for non-sleeved tubes at a test frequency of 100 KHZ.
At a test frequency of 350 KHZ, the amplitude sensitivity was.
reduced to approximately 30% to 40% of t' hat for a non-sleeved tube.
Eddy current inspection of the sleeve joints will present some difficulties particularly for the " alternate" type upper joint. The sleeve joints contain a number of features which will produce competing ECT signals makii.g it more' difficult to discriminate sleeve or tube wall defects at these locations.
The application of the multifrequency techniques will provide enhanced capability to discriminiate flaw signals from these competing signals.
Westinghouse is currently investigating ECT procedures to further improve the inspectability of these regions including the use of magnetic bias techniques and alternate probe types such as the crosswound probe, the rotating pancake (RPC) probe, and the multicoil surface riding probe.
3.0 EVALUATION i
3.1 Structural and Leak Tight Integrity We have reviewed the extensive program of verification analysis and tests to qualify the structural and leak tight (or leak limiting) integrity of the sleeved tube assemblies and the results thus far available. Although an assessment of primary plus secondary stresses against-the 3 Sin limit. (" shake-
-down") of the ASME Code remains to be completed, the licensee has sufficiently demonstrated by analysis that adequate margin will exist against a burst.
failure of the-sleeve' during the full range of normal, transient, and postulated accident conditions, consistent with the primary membrane and primary plus bending stress limits of the Code. Mechanical' load cycling tests to verify the long tenn structural, fatigue, and leak tight (or leak limiting) performance of the sleeve joints have reached the equivalent of five years of operation
o 7-with no adverse results. This preliminary data, coupled with the results of.the fatigue analysis performed to the ASME Code requirements, provides reasonable assurance against a fatigue or shakedown failure of the demonstra-
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tion sleeve joints during the interim period before the remaining analytical effort and testing is complete.
Regarding this sealing integrity of the joints, even if the demonstration sleeve joints should leak (between the sleeve and tube wall) at several orders of magnitude higher than what has been indicated by the test results thus far, the total leakage would be insignificant compared to the ifcensee's This is due to the relatively small criteria for allowable total leakage.
number of sleeves involved in the demonstration program and the inherent leak limiting geometry of the sleeve joint.
We have also reviewed the licensee's " leak before break" analysis. We find that the available margins are consistent with those which exist for the original tubing and are acceptable.
3.2 Plugging Limit The licensee has not yet proposed a plugging limit for the sleeves should they become degraded.
Based upon our review and assessment' of the minimum wall thickness requirements calculated by Westinghouse, we find that a 35%
plugging limit (sleeves with greater than 35% through wall degradation due to be plugged) will assure acceptable margins to failure consistent with the criteria of Regulatory Guide 1.121. Pending additional information from the licensee to justify a less restrictive limit, we are imposing a 35%
plugging limit as an interim requirement.
3.3 Alternate Upper Joint Integrity Laboratory testing has shown a significant reduction in the ultimate and yield strength of the sleeve and tube material in the zone local to where the sleeve wall is sealed to the tube wall. However, tensile tests of the San Onofre and Point Beach joint configurations have demonstrated that the' sleeve and tube wall at the seal will reinforce each other and that the
.overall strength of the joint exceeds that of a sleeve wall exhibiting a tensile stre'ngth equal to the design minimum strength in the ASME Code.
Based upon this,-the extensive mechanical tests (proof pressure tests, pressure and axial load cycling tests) which have been completed for San Onofre, and the confirmatory testing which has been c>mpleted to date for the actual Pof nt Beach joint configuration, ve conclude that there is reasonable assurance against a structural fa1.ure of the joint during the interim period before all tests.are ccmpleted. Primary side and secondary side hydrotests will be performed on the sleeved tube assemblies subsequent to the sleeving operation and provide additional assurance of joint-integrity.
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i We have also reviewed the difficulties experienced at San Onofre regarding localized erosion of the sleeve and tube wall at the joint as a result of the heating process. Based upon the metallographic examinations which have been performed on the San Onofre joints and revised heating parameters which have been implemented at Point Beach, we have concluded that this phenomenon will not have any significant adverse affect on the integrity of the Point Beach joints.
Additional assurance is provided by the on-going mechanical testing of these joints which have been fabricated to the process parameters to be used in.the field and the eddy cu'rrent and hydrostatic tests that will be performed following the sleeving operation.
3.4 Cerrosion Resistance We have reviewed the' test data from the San Onofre corrosion program for
.the sleeve repair and find that the tests and their results are directly applicable.to the Point Beach sleeving repair test program. The small difference is the tube dimensions that cause slightly different operating values in the fabriuation procedure do not affect significantly the corrosion resistance of the tubes or the joints. The test program has studied the behavior of the repair program materials in pure water, in primary coolant, and in 10% caustic solutions to simulate the continued hide out of caustic iri the crevices and sludge on the secondary side of the s, team generator.
This work has shown that the thermal treatment to be given to the Inconel sleeves is effective in reducing the probability of caustic stress corrosion developing on these sleeves.
It has also been shown that the small, con-trolled amount of cold work performed on the Inconel in attaching the
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sleeve to th.e steam generator tube was not sufficient to cause a significant increase in the susceptibility of the tube to stress corrosion cracking f rom the primary side water. This amount of cold work is significantly less than that which occurred where the tube was expanded into the lower po'rtion of the tubesheet durir.g the original fabrication. To date no cracking has developed in that area in Point Beach, San Onofre, or in model
-boilers and heat crevice tests. Further the tests have shown that there
. is only minor degradation of the material properties and corrosion resistance of the tubes at the upper joints. This has been shown by hardness test
. traverses and corrosion tests in caustic.
3.5 Eddy Current Inspectability The eddy current inspectability of the s,ieeve walls between upper and lower joints will be comparable to that for an unsleeved ' tube without a significant loss of sensitivity. Geometric discontinuities at* the sleeve joints will.
produce signal interference. However, the use of non-standard' eddy current probe types and multifrequency techniques should permit adequate inspections e
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of these_ areas. On'e local area that may present special difficulties is the sleeve' joint which has received the proprietary heating process.
Westinghouse is investigating methods to improve the inspectability of this area.
In the meantime, the preservice eddy current inspection of the sleeves will be supplemented by primary -side and secondary side hydrostatic tests (2000 psid and 800 psid, respectively) to provide added assurance of the joint
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integrity.
4.0 ' ALARA Considerations The licensee has taken.into account ALARA considerations for each of the radiation activities involved in the proposed steam generator sleeving
. demonstration at Point Beach. ALARA activities specifically directed to redyction of occupational radiation exposures include:
decontamination 4
of steam generators, personnel training in full-size mockups, installation of shielding, if necessary, to reduce radiation exposures to repair personnel.
^dministrative control of personnel exposures will be effected by careful planning of maintenance procedures for the job, in order to minimize the number of personnel used to perform the various tasks involving relatively high doses and dose rates. TV surveillance of personnel during tasks.will be used to identify areas resulting in high exposures, and thus to initiate suitable dose-reducing actions.
Based on prior inplant experience with channel head decontamination and laboratory decontamination, no significant increase in airborne radioactivity is to be expected. However, vapors from the channel head will be drawn through a high efficiency air particulate filtration system before release i
to the plant filter system. All sleeving operations wil1~ be monitored to keep airborne releases to a minimum. The licensee does not expect that auxiliary ventilation or special enclosures will be necessary.
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The licensee had made use of experience gained in prior channel head decontamination in planning for the proposed tube sleeving activities.-
Data was available for Point Beach Unit 1, Takahama Unit 1, San Onofre-
.tinit 1, and Turkey Point Unit 3..In particular, the applicant considered information on mechanisms used in prior decontamination. The licensee has provided information relevant to projected occupational radiation -
, exposures resulting from the demonstration ~ decontamination / sleeving program at Point Beach Unit 1, 'as well as from the proposed full-scale sleeving L
program for both units.
The licensee has estimated the radiation doses likely to be associated L
with the~ processes involved in the sleeving program:
i (a) installation of remoting tools and equipment - 5 person-rems, (b)' decontamination of the steam generator - 10 person-rems (including tube decontamination),
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.. (c)' installation 'of, additional shielding, if necessary - 10.6 person-rems
.(9.5 for the channel head,1.1 for nozzle shield removal),-
. current inspection,' 300 millirems / sleeve test)..
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' (d) inspectit,,O'd *esting -12.9 person-rems (92 millirems / sleeve eddy 4
(e) de-plugging; tubes for sleeving - 3.4 person-rems / tube (explosive),
.0.9. person-rems / tube (mechanical).
-(f) ' sleeving 5 person-rems / tube.
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The-licensee has.provided. realistic estimates of dose rates and occupancy -
factors, as.the bases;for_ these dose estimates, and has estimated the
- total person-rem dose resulting.from'the demonstration sleeving-program
'at PointjBeach Unit 1 at 48-60 person-rems assuming a decontamination factor of about 2.5.
The: radiation' exposure data and the operational experience resulting from the _ proposed demoWation of the ~ sleeving process at Point Beach Unit l' will be a tes', of. proposed radiation control techniques, and will provide a basis' for a more refined and more precise estimation of doses-
,likely to result from the proposed future sleeving process of both units.
5.0 REDUCED FLOW CONSIDERATIONS The licensee has stated that the sleeving of 20 steam generator tubes is equivalent. to the reduction ~ in flow through the steam generator caused.
by plugging.one. steam generator tube. The licensee plans to sleeve 12' steam generator-tubes.
Acccording to the licensee's estimates this will cause less effect:than plugging one tube.
Further, some of t.ie tubes the licensee plans to sleeve will be tubes previously degraded beyond the plugging. limit. The licensee plans to remove
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the plugs from-these tubes 'and insert sleeves. to bridge the degraded or defective portions of these~ tubes.
Based on the. licensee's estimates, this would result in a net increase in flow through the steam generators.-
Even.if the licensee's estimates on the amount of flow reduction associated with sleeving. a steam generator tube are in error..and even if the. licensee -
does not recover any previously plugged tubes-by sleeving, this will not present an unreviewed safety question for the demonstration sleeving progran.,
- Point Beach Unit 1 is operating lwith an 18% plugging. limit for its steam
. generators., This is based upon"an 18% tubes plugged ECCS (Emergency Core Coolant; System) analysis submitted by the licensee;and ' approved' by the NRC.
staff. Currently between 12-13%'of the steam gener ator' tubes in Unit 1 are
_ pluggd! Since 1% of thel total numberf of, tubes is< approximately 32 tubes for each steam generator, even assuming that the reduction of flow caused by sleeving.a steam generator tube'was equivalent to that caused by plugging
- a. tube,:this Lislstill well within the limitsjof the-previously-approved 1 analysi s'. -
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. For the reasons stated above, the staff finds the effect of the steam generator demonstration sleeving program to be insignificant from a flow reduction
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6.0 CONCLUSION
S Based upon the above evaluation, we conclude that the verification analyses and tests completed to date for the Point Beach sleeves, plus the similar 1
program which Ms been completed for the San Onofre sleeves, provides reasonable assurance that the sleeves and sleeve joints will exhibit acceptable mechanical strength corrosion resistance and leak tight (or leak limiting) cepability for the interim period before the Point Beach sleeve verification program is completed. Even if the demonstration sleeves' joints develop substantially more leakage than indicated by test, the total leakage will be insignificant.
The preservice eddy current inspection and primary side and secondary side hydrostatic tests to be performed prior to startup, and the stringent primary to secondary leak rate limits in the Plant License, will provide additional assurance that th sleeved assemblies will maintain adequate tube integrity during normal operation and postulated accidents.
If leakage l
in excess of the leakage rate limit does occur, the plant will be shut down for evaluation of the cause of the leak and appropriate corrective action.
Until sucn time as the licensee' submits justif. cation for a less restrictive plugging limit, we require that sleeved tube assemblics containing sleeve indications equal to or greater than 35% through-wall be plugged.
Based on the staff's review'of the Point Beach Steam Generator Tube Sleeving Report, and the additional information provided, we conclude that the lic'ensee's estimated dose for this project appears reasonable and that the licensee intends to implement reasonable radiation protection actions that should maintain inplant radiation exposures within the applicable limits of 10 C.FR Part 20, and should maintain exposures ALARA.
Based upon-the staff's review of the reduced flow considerations associated with the demonstration sleeving project, the staff. finds the effects to be within the ran,ge of the previously approved ECCS analysis for operation 3
with up to 18% of Unit 1 s steam generator tubes plugged. Therefore, the i
s.taff finds its impact upon the _ health and ' afety of the public to be s
insignificant.
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. 4 We have further concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered. by operation in the proposed manner, and-
-(2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
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REFERENC5:
,1.
Transcript of " Steam Generator Sleeving Review Board Mee ing, San Onofre Unit 1 Steam Generator Sleeve Repair for Southern California Edison, Westinghouse Electric Corporation, Forest Hills Division, Pittsburgh, Pennsylvania, 15221,' Thursday, October 23, 1980 - 8:15 A.M., Friday, October 24,1980 - 8:05 A.M.".
Date: November 10, 1981 4
4 i
ENVIRONMENTAL IMPACT APPRAISAL BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENOMENT NO. 56 TO OPERATING LICENSE NO. DPR-24 WISCONSIN ELECTRIC POWER COMPANY i
DEMONSTRATION PROGRAM OF STEAM GENERATOR REPAIR BY MEANS OF SLEEVING POINT BEACH NUCLEAR PLANT UNIT 1 DdCKET NO. 50-266 t
DATE:
-October 26, 1981 hD 00 66 l
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1.0 INTR 00bCTION Wisconsin Electr.ic Power Company (WE) by letter appifcation dated July 2,1981, as modified by letter dated October 12, 1981 <eeks a ' license anendment which would authorize WE to operate with six steam
- generator tubes sleeved rath'e'r than plugged which have dearadation
-exceeding the plugging limit defined by Technical Specification 15.4.2.A.5(a) at Point Beach Nuclear Plant Unit 1.
This Environmental Impact Appraisal documents the results of the staff review and evaluation of the environmental and radiation exposure impact of the steam generator tube sleeving - demonstration project and interim opera-tion of Unit 1 at power with 12 tubes sleeved (up to six of which have degradation exceeding the plugging limit) until final review of their overall steam generator tube sleeving program has been completed. Based on its review, the Staff finds that the proposed action will not significantly affect the quality of the human environment.
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2.0 8ACKGROUND
-In the past, Point Beach Nuclear PlantUnits 1 and 2 have experienced various corrosion problems in their s, team generators..
The problems include caustic intergranular attack of the tubes in the crevice region of the tubesheet and phosphate wastage thinning above and usually within 2 inches of the top of the I
tubesheet. These problems have been more severe for Unit 1 than Unit 2 and resulted in the Commission issuing Orders for Modifica-tion of License for Unit I dated November 30, 1979 as m'odified by Orders dated January 3,1980 and April 4,1980. These orders imposed, among other things, more frequent eddy current in'spections, more restrictive reactor coolant radioactivity levels,. much more.
restrictive steam generator tube leakage istes and operation at reduced primary pressure for Unit 1.
In an effort to find an acceptable fix to the steam generator tube corrosion problem. WE has submitted an application dated July 2, 1981 for a license amendment involving Technical Specification changes which would allow them to repair degraded steam generator tubes by sleeving rather than plugging, which degradation of steam generator tubes had exceeded the plugging limit of 40% nominal wall thickness.
In support of this requested change, the_ licensee has filed with the NRC staff for its review a Westinghouse Steam Generator Report containirig technical information regarding tube sleeving 'of the Point Beach Unit I and 2 steam generators. WE modified its application of July 2,1981 by letter dated October 12 l
3-1981 to. request interim operation of Unit I with 12 sleeved tubes
- (no more than six of which have indications of degradation beyond
- the plugging limit) as a demonstration program until final reyiew of their overall tube sleeving program has been ' completed.
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3.0 _SC,0PE OF WORK TO BE PERFORMED IN THE DEMONSTRATION PROGRAM WE has described the scope of the steam generator tube sleeving-demonstration program to be. conducted at Point Begch Nuclear Plant, Unit 1 to include the following major steps:
(1) Demonstration of the capability to insert sleeves of two different designs in steam generator tubes with indications of tube degradation. Up to six of these tubes would have degradation in excess of the plugging limit and would include tubes which are presently plugged. The sleeve designs to be used are described in Section 3.2 of Westinghouse Report WCAP-9660 (Proprietary) dated September 28, 1981, and entitled,
" Point Beach Steam Generator Sleeving Report' for Wisconsin a
Electric Power Company" (Sleeving Report).
(2) Demonstration and evaluation of the feasibility of explosive and mechanical tube plug removal using plug removal equipment described in Section 4.1 of the Sleeving Report.
(3) Demonstration and evaluation of the tube preparation and sleeving processes and parameters described in Section 4 of the Sleeving Report.
(4)
Demonstration and evaluation of the tooling designs required for field installation of sleeves as described in Section 4 of the Sleeving Report.
(5) Demonstration and evaluation of steam generator channel head decontamination equipment descrioed in Section 8 of the Sleeving Report.
(6) Demonstration and evaluation of non-destructive examination
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techniques described in Section 7 of the Sleeving Report.
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,n Environmental Impacts Of The Demonstratio[1 Program.
4.0 The-Staff has reviewed the radidlogical and non adiological environ-
%j mental. impacts of the Demonstrition, Pro,pd5. The, Staff has iden-z-t 4
tified the radiological environmentaleisp'actg of occupational '<
J exposure and public radiation exposure, as the;only measurable environmental impacts of the demonstration ogr.ai.,These' impact's are discussed in the fo11owin'6)ections.'[ ~
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4.1 Radiological Assessment; E
4.1.1 Occupational Exposure We have reviewed'the work procedures and practices that Wisconsin Electric. Power Company (WE) will use during the steam generator
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tube sleeving-demonstration project.
Based on this review, and through telepbone conve'rsations with the licensee, we feel that WE has taken adequate s'teps tp as'ssre, that the occupational radiation exposures associated with (ke tu,be slee eing 'Jemon'stration project w
will be maintained as low as is reasonabjy achi,evable (ALARA) and to assure Ehit the individual doses will be sakntained within the t
requirements of 10 CFR.Parb2,0;" " Standards for-Radiation Protection".
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Wisconsin Electric Power Company (WE) has estimated that the steam generator; tube sleev ng-demonstration project for the Point -
Beach Nuclear Plant, Unit 1, will require the expenditure of between approximately 48 andE60 person-rems.. The methods used
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by WE,to develop these collective occupational radiation expo-
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.. l project are based on -actual experience and testing.- WE 1) determined the maintenance activities that will be involved in cthe sleeving program; 2) estimated the person-hours of work necessary to perfonn those activities; 3) determined the areas' maintenance personnel must occupy to.perfann those activities and estimated the radiation dose rates in these areas; 4) multiplied the man-hours by the dose rate for each activity; and 5) summed the doses for all the activities. After reviewing the licensee's methods used to develop those dose estimates, we concluded that these estimates are reasonable.
Prior to initiating the steam generatorisleeving work WE will use decontamination thchniques in the steam generator channel head area to reduce dose rates. These techniques ~are expected j '
to reduce the dose rates in the hot leg channel heads of the 1
steam generators by a factor of approximately 2.5. Other ALARA measures implemented by WE during the steam generator sleeving-demonstration project include full size mockups for trairking workers, use of remote and semi-remote tooling when-i
'ever pra'cticable, and routine air sampling, and contamination 7
and radiation surveys. Measures such as these are recommended -
'-in Regulatory Guide 8.8, "Infonnation Relevant to Ensuring u
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That Occupational Radiation Exposures At Nuclear Power Sta-tions Will Be As-Low As Is Reasonably Achievable", in order to
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. minimize" individual l occupational radiation exposures and ~
. maintain the overall collective occupational radiation expo-p.
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' sure as low as'is' reasonably achievable (ALARA)., No indi-Lvidual will be allowed to exceed the dose limits imposed p
Y for workers by 10 CFR' Part 20,'which are established ~ as dose limits appropriate to the health and safety of i
individuals.
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To determine the relative environmental significance of the esti-w mated 1makimum occupational dose of. 60, person-rems, comparisons were made with 1) the doses expected from nonnal operation of nuclear plants, and 2) other non-nuclear risks.;
Table 4.1 shows the occupational dose history for Point Beach Units,[1 and 2 '3 When MN ismore than one reactor unit at a 2
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plant site (as at 9 i.i ' aacoj the ccabined occupational-dose for, E
all reactor imite. (tvr qaple, Point Beach Units 1 and 2) can be reported '
instead of the dosts for each separate unit. With the.
addition of 60 person-rems for the sleeving-demonstration project.
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the average _ annual dose for the 10 years af dose history at Units 1
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m and 2 (1970 through 1980).will be approximately 470 person-rems or.
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an average of 235 person-rems per reactor unit. Occupational expo.
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- sure estimates were.not specifically considered in the Point Beach
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4 Units 1 &.2 FES. However.;in recent environmental statements ly
.for new pressurized. water reactors (e.g., Summer FES), we have.
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- prov,ided an estimate of.410 person-rems per reactor unit as.
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the ~ average annual occupational dose.5 This estimata is based
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6 on reported data from power reactors that are operating with radiation protection programs in accordance with URC guidance 1
" and aregulations. A summary of these data 11s provided'in-Table '4.2.2 These data show that 410 person-rems per reactor l
1 un'et per year is roughly the average of. the wide range of doses incurred at all pre surized water reactor units over.
4 the last several years. The amount of dose. incurred at any single re' actor unit in a year is highly dependent on the amount of major maintenance performed that year.
Operating data from U.S pressurized water _ reactors indicates that units requiring high levels of special maintenance work can average as much as 1300 person-rems per year over the life of the unit.6 Although the doses for these particular plants far exceed the average of 410 person-rems-for PWR's, these doses are included in the average and are considered normal deviations from the average, particularly since such maintenance contributes to effective and safe plant operation and since it is carried out with procedures that maintain exposures ALARA.
? As Table 4.2 shows; the 60 person-rems estimate for the sleeving-demonstration project is within the low end to
'the historical. range of doses for a single unit in a year.
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We calculate that 60 person-rems..the occupational dose estimate for
'the sleeving-demonstration project, corresponds to a risk of very b
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- much less than.one premature fatal cancer in the exposed work force population. We _ also calculate that 60' person-rems corresponds to a risk of less than 0.02 genetic effect to the enpuing five generations. These risks are based on risk estimators derived in -
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the BEIR Report and WASH-1400 from data for the population as a 9
whole.
New information.in the' BEIR III Report would lead to an even lower. estimated risk for premature fatal cancers. These risks -
are incremental risks (risks in ' addition to the normal risks of fatal' cancer and genetic effects' we all face continuously). For a population of 1000, these normal risks that are unrelated to Point Beach Nuclear Station would be expected to result in about 190 cancer deaths and about 60. genetic effects in the existing popula- ~
tion (genetic effects are genetic diseases or malformtions),7.10" plus about 300 more genetic effects among their descandants.
To make the health risk associated with radiation dose more under-standable, risk comparisons can be made with non-nuclear activities commonly participated in by many individuals. One rem of. radiation
~4 7 is numerically comparable to a lifetime mortality risk of about 10
-4 Table 4.3 presents the equivalent risk of 10 for several, common ~-
activities - risks which many people take routinely and c6nsider to be insignificant.II The average dose to a worker for the sleeving-demonstration-project will be roughly 0.6 rems. As Table 4.3 shows, f
the lifetime risk from radiation ' dose-for the average sleeving-
~ demonstration ~ project worker is smaller than the lifetime risk associated with many; common activities.
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' Another perspective 'of an occupational risk comes froni comparison of occupational. mortality risks in the U.S.
One such i:omparison is J
. shown in Table '4.4.
It. indicates that radiation, exposure in the work place, as. experienced at an average radiation worker exposure rate, results in a relatively low occupational risk.
Some-have criticized occupationally related cancer estimates as being overly conservative.12' However, most experts feel the risk estimates in Table 4.4 relating to occupational exposure to low-LET radiation are also over-estimates.
In our opinion, the comparisons just presented are reasonable ones. The risks of occupational exposures in the range of 0.5 rem per year to 5 rem per year do not significantly affect a typical worker's total risk of mortality.
In sununary, the staff has drawn the following conclusions regarding occupational radiation. dose. WE's estimate of 60 person-rem for the sleeving-demonstration project at Point Beach.1 is reasonable. This dose is at 'the low end of the normal range of annual occupational doses which have been observed in recent years at opera' ting reactors. Although the doses resulting from the steam generator tube sleeving-demonstration project will increase the annual collective occupational dose average of Point Beach Units 1 and 2 combined to approximately 470 person-rems, this is still well below the 1300 person'-rems per year annual average referenced in current Final Environmental Statements as being an upper bound dose average of PWR's experiencing high levels of special main-tenance work.: WE has taken appropriate steps to ensure that
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occupational doses will be maintained within the limits of 10 CFR q
~ Part 20 and'ALARA. The additional. health risks due to these doses over nonnal risks are quite small, very much less than one percent of normal risk to the project work force as a,whole.
The risk. to an average individual in the work force will be
-lower than risk incurred from participation in many coninonplace a'ctivities. The in'dividual risks associated with exposures involved
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in the sleeving-demonstration _ program will be controlled and ifmited so as not_ to exceed the limits set forth in 10 CFR Part 20 for occupational -~ exposure.- For the foregoing reasons, the Staff con-cludes that the environmental impact due to occupational exposure 1
will not significantly affect the quality of the human environment.
4.1.2 Public Radiation Exposure NRC Staff has estimated the amount of radioactivity which will be released in liquid and gaseous effluents as a result of the sleeving-demonstration project.1 Those estimates are presented in Table 4.5.
I The estimates are based on information supplied by WE to the NRC Staff concerning the method of decontamination and subsequent treatment of the decontamination solutions. Table 4.5 also presents 13 14 effluent releases ~ for 1979 and 1980 from Point' Beach 1 and the 4
FES annual average effluent release estimates, k
WE will take several steps to minimize rel' eases.1 To minimize airborne releases.the channel head decontamination process and the surface preparation process will be wet processes, entraining r
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remov'edmaterial~inwater. The air from the channel head where the work is' being performed will be exhausted through the opposite manway using a'high efficiency particulate filter to control airborne concentrations during channel head work.
The water
'from the. decontamination process and the surface preparation process will be treated by filters, an evaporator and a deminera-lizer to minimize liquid releases.
- As Table 4.5 shows, the expected releases from the sleeving-demonstration project are small compared to both -the FES estimates and Point Beach's actual annual releases. Therefore, on the basis of this comparison above, we conclude that the offsite environmental impact that may_ occur during the period of this procedure will be i
snaller than that which occurs during normal operation.
We have estimated the doses to individual members of the public as well as the population-as a whole in the area surrounding Point 9
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Beach Unit l' based on the radioactive effluents which we estimated-for the. sleeving-demonstration project (summarized in' Table 4.5) and on the calculations 1 methods presented in Regulatgry Guides 1.109,I e
and 1.113.16.Using a liquid release source tenn of 1.44 x 10~4 Ci consisting primarilyof Co-60 (Table 4.5) we calculated the maximum
. individual total body dose for an adult to be less than.01 mrem for -
1 the operations. This is equivalent to a-dose of less than a small
, fraction of 1 percent of. the limits of 40 CFR Part 190. The annual limits of 40 CFR Part 190 are 25 millirems to the total body or any organ except the thyroid and 75 millirems to the thyroid. The dose 4
.to the population of 819,000 within 50 miles was estimated to be
- 1ess than 6.2 x 10~3 person-rems to the total body frem liquid ef fluents. The'offsite population dose was calculated by multiplying the (offsite) maximum individual total body dose of 7.5 x 10-6 mrem (estimated for the liquid release of Co-60) with the projected population of 819,000 for the year 1985 within 50 miles of Point Beach 1.
We feel that this is a conservative estimate as the maximum individual dose estimate'is overly conservative and it is very.unlikely that an average individual offsite will receive such-a dose. Evary year the same population of about819,000 will receive.
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ia cumulative total ' body dose of more than 81,900 person-rems from the natural background radiation (about 0.1 rem per year) in~ the
. vicinity of Point Beach 1.11 Thus, the population total body dose from the sieeving-demonstration project is less than 7.6 x 10-6 per--
' cent of the annual dose due to natural background. On these bases,
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. e concluda that the doses to individuals in unrestricted areas and w
'to the popu)ation within 50 miles due to gaseous and ifquid efflu--
ents from the sleeving-demonstration project will not be environmentally significant. Since we expect no larger radioactive effluents from Point Beach 1 after the sleeving-demonstration (over presleeving operation), we conclude that the impact on biota other than man will also be no larger than the demonstration project.
In summary, the radioactive releases resulting from the sleeving-demonstration project will be less than those due to nomal plant operation. These releases are also much less than the estimates-presented in the FES. The doses due to these releases are small
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.I Our calculations (using the LADT4P Computer Program)17 for the maximum individual total body dose for an adult cessidered the following-pathway consumption (1) of fish.(71 kilogram per year) caught in the discharge area and (2) drinking water (730 liter.per year) from the:
~ discharge area. A conservative dilution' factor of w or no. dilution was assumed for:each'of the~ above two pathways in-our evaluation of radiological exposure due to the release of. Co-60 from Point Beach 1
.?avia liquid effluents which 'are expected to result from the sleeving-
' demonstration ' project. The LADTAP II program implements the radiological exposure models described in U.S. NRC' Regulatory Guide 1.109, Rev.1 (Appendix,a)15 for radioactivity releases in liquid effluent.
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- compared to the limits of 40 CFR Part 190 and to the annual dose from natural. background radiation. Therefore, the' radiological impact of the sleeving-demonstration project will, not significantly affect the quality of the human environment.
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'4.1.3 RADIOLOGICAL ASSESSMENT CONCLUSIONS Based on our ~ review of the proposed steam. generator sleeving-
, demonstration project, we have reached the following-conclusions which ~are. discussed in greater detail. above.
1 (1) The estimated range of 48' to 60 person-rems for the-sleeving-
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demonstration ~ project is on the low side of the expected range of doses incurred-at light' water power reactors in a year.
(2) The risks to the workers. involved in the sleeving-demonstration project from radiation exposure are no larger than the risks incurred by:
3 (a) workers in other industrial businesses, and (b) most people, working or not, from commonplace activities such as driving a car.
(3) WE has taken appropriate steps to ensure that occupational dose will be maintained as low as it reasonably achievable and within the limits of 10 CFR Part 20.
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(4) Offsite doses resulting from the sleeving-demonstration project will be,
'(a) smaller than those incurred during nonnal operation of' Point Beach 1, and (b) negligible in compariso'n to the dose members of the public in the vicinity of Point Beach 1 receive from natural i
background radiation.
On the basis of-the foregoing statements, the staff. concludes that L
the proposed sleeving-demonstration project at the Point Beach
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Nuclear Plant.. Unit No. I will not significantly affect the. quality of the human environment.
4.2 Nonradiological Assessment The have reviewed the documents submitted by WE in support of its
. request to conduct the s..eam generator tube sleeving-demonstration program. We find that the proposed activities will occur within the plant on_ areas previously disturbed during site preparation and 4
construction. These activities will not have appreciable offsite environmental effects. The licensee has not proposed any changes in effluents from the demineralizer waste systems or other waste streams as part of the demonstration program. We conclude that the activities as proposed will not result in any significant environmental impact.
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. BASIS AND CONCLUSION FOR NOT PREPARING AN' ENVIRONMENTAL IMPACT S 5.0 The NRC has reviewed the Demonstration Program relative to the
. requirements set forth in'10 CFR Part 51 of the Commission's regulations. The NRC has determined, based on this. assessment, that this action will not significantly affect the quality of the human environment. Therefore, the Commission'has determined'that an-Environmental ' Impact Statement need not be prepared, and that, pursuant to 10 CFR 51.5(c)(1), the issuance of a negative declaration to this effect is: appropriate.
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. TABLE 4;1-
' ANNUAL COLLECTIVE, 3 2
OCCUPATIONAL DOSE AT POINT. BEACH UNITS
- 1, 2-Collective Occupational Dose Year (person-rems)
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-1971 164 1972 580 1973 588 1974 295 1975 459 1976 370 1977-429 1978 320 1979 644 1980 7913 4
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First commercial operation 12/70 (Unit 1),.10/72 (Unit 2) e O
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OCCUPATIONAL DOSE AT U.S. LIGHT ~ WATER REACTORS (person-rems per reactor unit)
PWR JWR Year-Average Average Low
![ljgt
- 1969 165 195 42 298 1970 684 127-44 1639
.i 1971 307 255-50 768 1972 464 286 61 1032 1973 783 380 85 5262 1974 331 50/
71 1430 1975 318~
701 21 2022 1976 460 549 58 2648 1977 396 828 87 3142 1978 429 604 48 1621 1979 510 733 30 2140 1
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TABLE 4.3 LIFETIME MORiALITY RISKS 18 NUMERICALLY EQUIVALENT TO ONE REM
- Type of Activity
-Equivalent' Risk to One-Rem I carton Smoking cigarettes c
i 66 bottles Drinking wine t.
Automobile driving 6,600 miles Commercial flying 33,000 miles Canoeing 1.6 days
- Being a man aged 60 1.8 days i
Eight haurs per day b
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TABLE 4.4 OCCUPATIONAL RISKS Events per year per 100,000 workers),
Mining &
All U.S.
Radiation-Quarrying Industries Trade Exposure
- Fatal Accidents (1) 63 14 6
1 Delayed Effects Actual '
readily Occasionally not not Observable Observable Observable Observable Observable Estima'.ed
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Includes 115-219
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4-6 lethal cancers (2) lethal cancersI3)
(1) 1976 data, from " Accident Facts,1977 Edition," National Safety Council.
(2)
Estimates from " Toxic Chemicals and Public Protectinn, A Report to the President by the Toxic Substances Strategy Committee," Council on Env_i-ronmental Quality, Government Printing Office, May 1980.
Assumes20-387. of all cancers are associated with occupation.
(3) ' Estimates from BEIR-II,1980, assuming an average radiation wrker -
exposure rate of 0.5 rem /hr; exposure at the limit, 5 rems /yr, would 3
yield an estimate of from 37 to 63 lethal cancers per year per 100,000 workers.
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TABLE 4.5 RADI0 ACTIVE EFFLUEhTS FROM POINT BEACH 1 WE Estimates for Point Beach 1 Point Beach 1 FES(1)' Estimates of Type of Radioactive Releases During Sleev-
.1979 Releases 1980 Releases Annual Average.
Effluent ~
ing Demonstration (Ci)
(Ci)
-(C1)
Releases-(Ci/yr.)
I Gaseous b
Noble Gases Negligible 4.8(+2)c
. 3.2(+2) 5.0(+3) b lodine + Particulates*
Negligible 1.4(-2) 2.7(-3) 1.0(-1) b d
Tritium Negligible 4.0(+2) 3.3(+2)
-Liquid Mixed fission and activation products 1.44 x 10-4 0.38 0.63 1.0(+1)
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b Tritium Negligible 4.5(+2) 3.8 (+7.)
1.L(+3) aRadioactive half-lives 8 days or more, bBelow lower limits of detectability for plant instrumentation.
4.8(+2) means 4.8 x 10+2 C
dNo estimate was given in FES, but FES stated that there would be' low concentrations of tritium to the gaseous releases.
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- References 1.
. Point Beach Steam Generator Sleeving Report for Wisconsin Electric Power Company prepared by the. Westinghouse Electric Corporation --
i September. 28, 1981.
2.
NUREG-0713. Vol.1, Occupational Radiation Exposure at Commercial:
' Nuclear Power Reactors,~ 1979, U.S.N.R.C., March 1981..
.3.
NRC Memorandum dated June 19, 1981, from W. E. Kreger to H. R.
Denton entitled " Unusually High Occupational Doses Reported For Power Reactors Operating in 1930."
4.
Final Environmental Statement related to operation of Point Beach Nuclear Plant, Units 1 and 2. United-States Atomic Energy 4
Commission, May 1972..
3 I
5.
NUREG-0719, Final Environmental Statement Related to the Operation
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of Summer Pressurized Water Reactor,1981.
-6.
NUREG-0692, Final Environmental' Statement Related to Steam Generator Repair'at Suny Power Station Unit 1, July 1980.
7.
The Effects on Populations of Exposure to. Low Levels of Ionizing Radiation, "BEIR Report," report of the Advisory Committee on the i-Biological Effects of Ionizing Radiations, National Academy of-Sciences - National Research Council, November 1972..
- 8. -
WASH-1400, " Reactor Safety Study - An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," U.S.N.R.C., October.1975.
9.
The Effects on Population of Exposures to Low Levels of Ionizing Radiation "BEIR III Report", report of the Committee on the Bio '
logical Effects of Ionizing Radiation'.s National Academy of Sciences t
- National Research Council, 1980.
10.
1979 Cancer Facts and Figures, American Cancer Society.
t
- 11. NCRP No. 45, " Natural Background Radiation in the United States,"
National Council' on Radiation Protection and Mea'surements,1975.
k
- 12. -
R.' Peto, " Distorting the Epidemiology of Cancer, the Need for a More t
Balanced Overview," Nature 284, 297-298 (March 27, 1980).
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- 13. Wisconsin Electri,, Power Company, Point Beach Nuclear Plant, Unit Nos. 1 and 2. Semiannual Monitoring Reports,. January 1, 1979 through June 30, 1979 and July 1,1979 through December 31, 1979."
'14.
Wisconsin Electric Power Company, Point Beach Nuclear Plant, Unit
.Nos. I and 2, Semiannual'Honitoring Reports, January 1, 1980 30,'1980 and July 1,1980 through December 31. 1980.
through' June U
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.m
-uu m
,.-,,,3%
h-%,
,,4.
_m.
y-
,,h.
2-
- 15. ~ Regulatory Guide'1.109, " Cal'culation of Annual Doses to -Man from
- Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I" (Revision 1),
U.S.N.R.C., October 1977.
1/,.
Regulatory Guide 1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of'
~
Implementing-Appendix I," U.S.N.R.C.
- 17. User's Manual for LADTAP II - A Computer Program for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liq. f d Effluents. NUREG/CR-1276, U.S.N.R.C. (May 1980).
18.
E. Pochin. "The Acceptance of Risk," British Medical Bulletin
, 31(3), 1975.
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