ML20041C079
| ML20041C079 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/20/1977 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20041C070 | List: |
| References | |
| NUDOCS 8202260136 | |
| Download: ML20041C079 (6) | |
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Enclosure.3 j#"'%,g umTassTATES NUCLEAR REGULATORY COMMISSION s
- 3. 9 r,g y
t wasmworow, o.c.nases July 20, 1977 4
MEMORANDUM FOR:
ALL D0R PERSONNEL FROM:
V. Stello, Jr., Director, Division of Operating Reactors, NRR
SUBJECT:
PROCESSING OF DDR PROPRIETARY WITHHOLDING REQUESTS I have designated Reba Diggs as the D0R Coordinator for proprietary withholding requests. The D0R Coordinator is expected to carry out the items set forth in Enclosure 1 to this memorandum. Consequently, the DOR Coordinator should be sent copies of all correspondence re-lating to proprietary information and requests for withholding.
By copy of this memorandum I am requesting the Distribution Services Branch of the Division of Document Control to provide Reba Diggs a copy of each proprietary item assigned to this Division. Thus,'
your responsibility for providing her copies of related correspon-
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dence will be mainly for letters that you originate and to assure that she. is on distribution for all the material you receive. Alsc, she is to be added to the concurrence for all correspondence relating to proprietary documents, and should receive the packages for concurrence before they are sent to OELD.
Our handling of proprietary requests should be consistent with the guidance in DDR Memorandum No. 6. Rev.1, issued 5/18/77 to all D0R personnel, and Section 2.790 of Part 2.
Howard Shapar's memorandum of March 31,1976 (Enclosure 2 to this memorandum) provides addition-al guidance and discussion of section 2.790 Secretaries should also add Hazel Smith (DPM) to our distribution of letters to licensees and other regarding proprietary requests. She
.is the coordinator for DPM.
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B202260136 820225 l
PDR ADOCK 05000266 f_.
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s All persons who have received proprietary documents for review and for reference should take the necessary steps to assure th'at any outstanding proprietary requests (for topical reports and any other items) have been reviewed and the required granting or denial letter sent to the licensee / vendor.
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/ Y V.'Ste'llo, Jr., Director Division of Operating Reactors Nuclear Reactor Regulation f
Enclosures:
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Responsibilities of NRR Divisional Coordinators 2.
Memo 3/31/76 l
cc: E. Case J. Felton, DRR J. Cooke, OELD B. Grenier, DSS D. Wigginton, DSE S. Scott, DSB D. Lanham, DSB l
H. Smith, DPM l
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a March 31, 1976 The rule requires two deteminations by the staff when an appli-cation for withholding proprietary infomation is made. Broadly, those determinations are: (1) Whether the document sought to be 2
withheld contains trade secrets or confidential or privileged comercial or financial information, and (2) whether the document should be withheld from public disclosure. In making the first detemination, the staff will consider:
d' (1) Whether the infomation has been hel' in confidence by its owner;
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(2) Whether the infomation is of a type customarily held in confidence by its owner and whether there is a rational basis therefor; (3) Whether the infomation was transmitted to and received
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by the Comission in confidence; (4) Whether the information is available in public sources; and (5) Whether public disclosure of the infomation sought k
to be withheld is likely to cause substantial ham to the competitive position of the owner of the infoma-tion, taking into account the value of the infomation to the owner; the amount of effort or money, if any,
' expended by the owner in developing the information;
'and the ease or difficulty with which the infomation could be properly acquired or duplicated by others.
The staff's first detemination should be aided by'the facts contained in the owner's affidavit, for the rule requires that the considerations listed above must be addressed with specificity in the affidavit.
If the staff determines that the information is not a type which can be protected, then it need not make the' second determination that the information should be withheld from public inspection.
For example, the s*taff could detemine that the information is not a trade secret or otherwise confi-
'dential to the owner because the information has been made widely available to the public. This lack of " secrecy" or
" confidentiality"of the infomation.would, in most cases, prevent the owner from claiming that any ham would occur to his compet-itive position if the information were disclosed to others.
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If the staff determines that the document contains trade secrets or privileged or confidential commercial or financial infomation, then the staff will further determine:
(6) Whether the right of the public to be fully apprised as to the bases for and effects of the proposed action outweighs the demonstrated concern for protection of a competitive position; and (7) Whether the information should be withheld from public disclosure.Jf If the application for withholding is denied, the app 1'icant for withholding must be informed of the denial and the reasons for the denial. As in the fomer rule, the applicant will be given at least thirty (30) days from the date of the notice of denial to request that the infomation be returned. If no request for withdrawal is forthcoming within the time frame allotted, the information must be placed in the Public Document Room.
The revised rule contains one exception to withdrawal.
If the
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information is submitted in a rule making proceeding and subsequently foms the basis for the final rule, the infonnation will not be 4
returned and will be publicly disclosed.
The revised rule also follows past practice under the former rule by stating that the staff will infom the applicant for withholding of a decision to withhold the infomation from public inspection.
While the rule does not set a time limit for staff deteminations on applications for withholding, the Comission has expressed that the staff should niake such deteminations as expeditiously as possible.
You will note that the rule addresses the situation which could occur if the owner states that' the supporting information in the affidavit is also proprietary to the owner.
If this situation should occur, the rule provides that information contained in the affidavit which is properly marked as a " trade secret" or " confidential or privileged commercial or financial infomation" will be' withheld from disclostre under 10 CFR I 9.5(a)(4) of ou: regulations.
In other words, the J/
10 CFR I 2.790(b)(2) states: dA person who submits comercial or financial infomation believed to be privileged or confidential or a trade secret shall be on notice that it is the policy of Ch the Comission to achieve an effective balance between legitimate
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coricerns for protection of competitive positiers and the right i
of the public to be fully apprised as to the bases for and l
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effects of licensing or rule making actions, and that it is l
within the discretion.of the Commission to withhold such informa-tion from public disclosure."
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e e k March 31, 1976 owner will not be required to submit a second affidavit which states reasons for claiming the information in the first affidavit is proprietary. To do othemise could result in an endless series of affidavits.
Furthemort, if the owner does not want a third party, such as a utility making the application for withholding, to see the infomation containe~d in the affidavit, we believe that in most cases it may be submitted directly to the Comission,7ather than through the third party.
As in the fomer rule, the applicant for withhe'dtng is encouraged to incorporate the proprietary infomation sought be withheld into a document separate from any other submitted document containing non-proprietary information.
Mortover, the new rule requires the return of any document claimed to be proprietary which is found by the staff to be irrelevant or unnecessary to the perfomance of NRC functions.
The rule sets forth two categories of proprietary infomation which will not be subject to the procedural requirements of 10 CFR I 2.790.
They are:
(a). correspondence and reports to or from'the NRC which identify
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a licensee's or applicant's procedures for safeguarding licensed special nuclear material or detailed security measurts for the physical protection of a licensed facility or plant in which ifcensed special nuclear material is possessed or used; and (b) information submitted in confidence to NRC by a foreign source.
No application for withholdin~ under 10 CFR I 2.790 will be required g
to exempt the foregoing infomation from public inspection. Such information will be considered exempt under 10 CFR I 9.5(a)(4), and will be subject to disclosure only in accordance with the provisions of 10 CFR 5 9.12.
Both the former and amended rule contain one important caveat: With-holding from public inspection shall not affect the right, if any, of persons properly and dirtetly concerned to inspect documents which have been withheld from public inspection under 10 CFR !>2.790(b).
Accordingly, the arrended rule provides that proprietary information may be subject to inspection:
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under a protective agreement, by contractor personnel or
_ 1) govermment officials other than NRC officials; (2) by the presiding officer in a proceeding; and
'(3) under protective order, by parties to a. proceeding.
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( March 31, 1976 In the latter instance, in camera sessions of hearings may be held when the information sougKt to be withheld is produced or offered in evidence.
If the Commission subsequently detemines that the information should be disclosed, the rule requires that the infoma-i tion 'and the transcript of such j,n, camera session will be made i
publicly available.
Finally, you should note that the amendments to 10 CFR I 2.790 proposed for public comment in 1974 would have required the owner of information which the Comission detemined should be withheld from public disclosure to submit a new affidavit at two year intervals which would justify the continued withholding of the information from public inspection. This provision has been eliminated in the final rule because, among other reasons, it was thought to be too costly and administrative 1y burdensome. Moreover, the Commission may always on its own motion or that of a third party require the owner of information which has been previously deemed to be exempt from disclosure to provide a new affidavit justifying the continued con-fidential treatment 'of the information. The Commission's present i
regulations in 10 CFR Part 9 implementing the Freedom of Infomation Act would, for example, permit any member of the public to make a request for exempt infomation, and proprietary information would
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be withheld from disclosure only upon a detemination that the infomation remains proprietary to the owner at the time the Freedom of Infomation Act request is made.
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I am sure that the amended rules will rec.uire you to change certain staff procedures for processing applications for withholding pro-prietary information.
It bears repeating that those procedures should assure the prompt disposition of each application.
I have asked Jay Maynard, Chief Counsel for Operations and Administration, to assist you in evaluating and modifying your present procedures and to be available to answer any questions you may have.
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H rd K. Shapar Executive Legal Director cc:
L. V. Gossick Q.,
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,. UNITED STATES g
g NUCLEAR REGULATORY COMMISSION y
B WASHINGTON. D. C. 20555
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SAFETY EVALUATION BY TjjE 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 I
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TABLE OF CONTENTS Pag'e 1.0 Introduction 1
w 2.0 Discussion 1
2.1 Sleeving Process Description 1
2.2 Structural Yerification Analyses.
3 2.3 Verificati6n Testing of Sleeve Joints 3
2.4 Verification of " Leak Before Break" 4
2.5 Effect,of Proprietary Heating Proct:t a
on Upper Alternate Joint Integrity 2.6 Discussion of Corrosion Aspects and 5
Verification Testing 2.7 Eddy Current Test Capabilities 5
3.0 Evaluation 6*
3.1 Structural and Leak Tight Integrity 6
3.2 Plugging Limit 7
3.3 Integrity of Upper Alternate Joint 7
3.4 Corrosion Resistance 8
3.5 Eddy Current In.spectability 8
4.0, ALARA Considerations 9
5.0 Reduced Flow Considerations 10 6.0 Conclusions 11 e
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1.0 INTRODUCTION
By letter dated July 2,1981, Wisconsin Electric Power Company (licensee) submitted an application for license amendments consisting of proposed changes to the Technical Specifications for Point Beach Nuclear Plant Units 1 and 2.
These proposed Technical Specification changes would allow operation at power of Units 1 and 2 with steam generator tubes h&ving degradation exceeding the plugging limit (40% nominal wall thickness) provided these tubes have been repaired by insertion of sleeves 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 ifcensee 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 at 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 i
for submitting this modification is to conduct a demonstration sleeving
' program on Point Beach Unit i during the October 9,1981 refueling outage.
This demonstr,ation program will utilize two separate sleeving processes and the licensee hopes it will provide valuable infonnation 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 plu'gs have been 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 Sleevinc Report for Wisconsin Electric Power Company." The licensee has submitted additional information by letters dated October 9,16, 24 and 26 in response to questions by the 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 tile steam' generator 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 sized to the ASME Boiler' and' Pressure Vessel Code, Sectiont,IIJ. -
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 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 tubesheet. The Point Beach sleeves are intended to address the general intergranular attack and stress corrosion cracking which has been confine.d 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 joint 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 sleeves taken together is less than the Technical Specification leak rate limit during normal operatidn.
In addition, total leakage must be maintained to within tolerable limits 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 i
eventually planned for sleeving (approximately 2500 tubes).
The second or " alternate" upper joint design is also a structural joint.
Th,is joint is fabricated using a proprietary heating process to form a leak tight seal. The lower sleeve joint also provides a structural and leak tight seal, but is not fabricated with the proprietary haating process.
The Point Beach sleeves and sleeve joints are basically similar to those at San Onofre Unit 1 from the standpoint of design and joint fabrication techniques. The San Onofre sleeves have been extensively tested for 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 perfomed to the requirements of Section III of the ASME Boiler and Pressure Yessel Code.. These analyses are intended to demonstrate adequate fatigue per-fomance 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 sleeved tube assemblies which could result from hourglassing (defomation) of the support piste 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 perfomed 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.
The test mockups for the lower joint include tubesheet mockups from which the effects of removing both mechanical and explosive 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 simuYate loads caused by differential themal expansion) and pressure cycling tests to verify the long term sealing integrity of.the joints under the specified operating transients (e.g., heatup/cooldown l
and plant loading / unloading cycles).
Specimens for each type joint will I
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 limiting perfomance.
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 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 tha'n.4 inches H11 result.in leakage in excess of the Technical Specification
. leakage rate limit during normal operation.
4 2.5 Effect of Proprietary Heating Process on Upper Alternate Joint Integrity The proprietary heating process for the " alternate" upper joint design will result in some degradation of the mechanical properties of the sleeve and tube wall material local to the seal between the sleeve Snd 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 would 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 the width of the seal. Tensile tests of a number of joint specimens resulted in tensile failures of the sleeve wall invariably between two and three inches below the sealed l'ocation, 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 nom'al operating pressure, and external pressure tests to 1.5 times the maximum LOCA pressure loading resulted in no failures for the San Onofre " alternate" upper joint specimens.
Similarly, load cycling tests (to simulate pressure plus themal cycling)'
for the expected number of operating cycles over a 30 year lifetime were completed with no failurec. Similar confirmatory tests are in progress for the actual Point Beach configuration, with the exception of the collapse test.
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9 2.6 Discussion of Corrosion Aspect and Verification Testing The corrosion that has occurred on the outer surface of the tubes has been
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attributed to caustic corrosion resulting from the use of phosphate water chemistry in the secondary water with massive phosphate additions and the' formation of caustics due to impurities from persistent leaky tubes in
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the steam condenser..The chemistry control program of the secondary side water was switched to an all-volatile 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 intergranular
' stress corrosion cracking, wastage, and thinning has occurred at a location just above the tubesheet ir 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 licensee 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 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 l
applicability of the conventional bobbin type ECT probe to the inspection of the sleeved tube assemblies.
(This data was actually obtained for San Onofre sleeved as.semblies.) At the optimum test frequency for the sleeve, e
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,c 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 sensitivity 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 tiJbesheet 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.
l The inspectability of the tube wall is of interest at and above the ' upper sleeve joints. The Westinghouse study indicates that the amplitude 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 freque'ncy of 100 KHZ.
At a test frequency of 350 KHZ, the amplitude sensitivity was reduced to approximately 30% to 40% of that 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 making 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 i
rotating pancake (RPC) probe', and the multicoil surface riding probe.
l 3.0 EVALUATION 3.1 Structural and Leak Tight integrity W' have reviewed the extensive' program of verification analysis and tests
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e to. qualify the structural anti leak tight (or leak limiting) integrity of the sleeved tube assemblies and the results thus far available. Although t
an assessment of primary plus secondary stresses against the 3 Sm 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 tem structural, fatigue, and leak tight (or leak limiting) perfomance of the sleeve joints have reached the equivalent of five years of operation O
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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-tion sleeve joints during the interim period before the remaining. analytical effort.and testing is complete.
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Regarding this sealing integr'ity 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 licensee's criteria for allowable total. leakage. This is due to the rel,atively small 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 requirenent.
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 st'rength of the joint exceeds that of a. sleeve wall exhibiting
- a. tensile strength 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 completed to date for the actual Point Beach joint configuration, we conclude that there is reasonable assurance against a structural-failure of the joint during the interim period before all tests are completed. 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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8 We have also re' viewed 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 bee' performed on the San Onofre joints and revised heating parameters which.
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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 nachanical testing of these joints which have been fabricated to the process paraneters to be used in.the field and the eddy cu'rrent and hydrostatic tests that will be performed following the sleeving operation.
3.4 Corrosion 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 fabrication 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 in the crevices and sludge on the secondary side of the steam 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 the steam generator tube was not sufficient to cause a significant increase in the susceptib,ility of the tube to stress corresion cracking f rcm 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 during 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 sieeve walls between upper and lower
. joints will be comparable to that for an unsleeved ' tube without a significant loss of sensitivity. Ge'ometric discontinuities at the sleeve joints will produce signal interference. However, the use of non-standard eddy current i
l probe types and multifrequency techniques should permit adequate inspections l
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- of these areas. One 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 integrity.
40 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 reduction of occupational radiation exposures include:
decontamination of steam generators, personnel training in full-size mockups, installation of shielding, if necessary, to reduce radiation exposures to repair personnel.
Administra'tive control of personnel expo'sures 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 to the plant filter system.
All sleeving operations wil.1 be monitored to keep airborne releases to a minimum. The licensee does not expect that auxiliary ventilation or special enclosures will be necessary.
The licensee had made use of e'xperience 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 linit 1, and Turkey Point Unit 3.
In particular, the applicant considered information on nechanisms used in. prior decontamination. The licensee l
has.provided information relevant to projected occupational radi'ation exposures resulting from the demonstrat. ion decontamination / sleeving program at Point Beach Unit 1, as well as from the proposed full-scale sleeving program for both units.
The licensee has estimated the radiation doses likely to be associated with the processes involved in the sleeving program:
(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),
(d) inspection and testing - 2.9 person-rems (92 millirems / sleeve ed#
current inspection, 300 millirems / sleeve test),
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(e) de-plugging tubes for sleeving - 3.4 person-rems / tube (explosive).-
0.9 person-rems / tube (mechanical).
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(f) sleeving 4-5 person-rems / tube.
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 Point Beach 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 demonstration of the sleeving process at Point Eeach Unit 1 will be a test 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 generatsr tube. The licensee plans to sleeve 12 steam generator tubes. Acccording to the licensee's estimates this will I
cause less effect than plugging one tube.-
Further, some of the tubes the licensee plans to sleeve will be tubes previously degraded beyond the plugging limit. The licensee plans to remove 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 program.
Point Beach Unit 1 is operating with an 18% plugging limit for its steam l
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 generator tubes in Unit 1 are
. plugged. Since 1% of the total number 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 is still well within the limits of the previously approved analysis.
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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 standpoint.
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6.0 CONCLUSION
S B,ased upon the above evaluation, we conclude that the verification analyses and tests completed to date for the Point Beach sleeves, plus the similar program which has 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) capability 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 prese'rvice 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 the sleeved assemblies will maintain adequate tube integrity during normal operation and postulated accidents. If leakage 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 such time as the licensee submits justification for a less restrictive plugging 1,imit, we require that sleeved tube assemblies 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 con'clude that the licensee'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 CFR 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 range of the previously approved ECCS analysts for operation with up to 18% of Unit 1 s steam generator tubes plugged. Therefore, the staff finds its impact upon the health and safety of the public to be insignificant.
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b We have further concluded, based on the considerat
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ner, and public will not be endangered by operation in the pro i sion's that:
inimical' regulations and the issuance of this amendment will ft f
.the public.
REFERENC5:
k sa'nOnofre Transcript of "steEm Generator Sleeving Rev'iew
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1.
- Westinghouse Electric Corporation, Forest Hills Division, P t s ur 15221 Thursday, October 23, 1980 - 8:15 A.M., Friday, Pennsylvania, 24,1980 8:0S A.M.".
October Date: November 10, 1981 e
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- 1 ENVIRONMENTAL IMPACT APPRAISAL BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 56 TO t)PERATING LICENSE NO. OPR-24 WISCONSIN ELECTRIC POWER COMPANY e
OEMONSTRATION PROGRAM OF STEAM GENERATOR REPAIR BY MEANS OF SLEEVING POINT BEACH NUCLEAR PLANT UNIT 1 DOCKET NO. 50-266 DATE:
October 26, 1981 i
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1.0 INTRODbCTION
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Wisconsin Electric Power Company (WE) by letter application dated j
July 2,1981, as modified by letter dated October 12, 1981 seeks a license unendment which would authorize WE to operate with six steam generator tubes sleeved'rather 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 Appraise 1 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 I at power with 12 tubes sleeved (up to six of which have degradation exceeding the plugging limit) until f,inal 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 BACKGROUND
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 tube' sheet and phosphate wastage * '
thinning above and usually within 2 inches of the top of' the' tubesheet. These problems have be'en more severe for Unit 1 than Unit 2 and resulted in the Commission issuing Orders for Mo'difica'-
tion of License for Unit I dated November 30, 1979 as modified 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 rates and operation at reduced primary pressure for Unit 1..
In an effort to find an acceptable fix to the stet.m 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 degr,adation of steam generator tubes had exceeded the. plugging limit of 40% nominal wall In support of this requested change, the licensee has thickness.
f filed with the NRC staff for its review a Westinghouse Steam Generator Report containing technical information regarding tube WE sleeving of the Point Beach Unit 1 and 2 steam generators.
modtfied its application of July 2,1981 by letter dated October 12, e
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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 dc::onstration program until final review of their overall tube sleeving program has been completed.
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3.0 SCOPE OF WORK TO BE PERFORMED IN THE DEMONS WE has described the scope of the steam generator tube sleeving-demonstration program to b'e conducted at Point Begch Nuclitar Plant.
Unit 1 to include the following major steps:
Demonstration of the capability to insert sleeves of two (1) 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 The sleeve designs to be f
tubes which are presently plugged.
used are described in Section 3.2 o'f Westinghouse Report and entitled, WCAP-9660 (Proprietary) dated September.28,1981,
" Point Beach Steam Generator Sleeving Report for Wisconsin Electric Power Company" (Sleeving Report).
Demonstration and evaluation of the feasibility of explosive (2) and mechanical tube plug removal using plug removal equipment described in Section 4.1 of the Sleeving Report.
Demonstration and evaluation of the tube preparation and (3) sleeving processes and parameters described in Section 4 of the Sleeving Report, Demonstration and evaluation of the tooling designs required (4) f for field installation of s1. eves as described in Section e
the Sleeving Report.
Demonstration and evaluation of steam generator channel head (5) decontamination equipment described in Section 8 of the l
Sleeving Report.
(6), Demonstration and evaluation of non, destructive examinatio techniques described in $ection 7 of the Sleeving Report.
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4.0 Environmental Impacts Of The Demonstration Program -
The Staff has reviewed the radiological and nonradiological environ-mental impacts of the Demonstration Program. The, Staff has iden-tified the radiological environmental impacts of occupational exposure and public radiation exposure as the only measurable environmental impacts of the demonstration program. These impacts are discus' sed in the following sections.
4.1 Radiological Assessment i
1 4.1.1 Occupational Exposure l
We have reviewed the work procedures and practices"that Wisi:onsin o
l Electric. Power Company'(WE) will use during the steam generator 1
tube sleeving-demonstration project. Based on this review, and f
through telephone conversations with the, licensee, we' feel that WE f
has taken adequate steps to assure that the occupational radiation exposures associated with the tube sleeving-demonstration project will be maintained as low as is reasonably achievable (Al. ARA) and to assure that the individual doses will be maintained within the requirements of 10 CFR Parti 20, " Standards for Radiation Protection".
Wisconsin Electric Power Company (WE) has estimated that the steam i
generator tube sleeving-demonstration project for the Point Beach Nuclear Plant. Unit 1, will require the expenditure of between approximately 48 and 60 person-rems.. The methods used i
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by WE to develop these collective occupational radiation expo-sure estimates for the steam generator sleeving-demonstration project are based on actual experience and testing. WE1)
' determined the maintenance activities that will be involved in the sleeving program; 2) estimated the person-hours of work necessary to perfom those activities; 3) detemined the areas maintenance personnel must occupy to perfom those activities and estimated the radiation dose rates in those 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 generat'orisleeving work, WE will use decontamination techniques in the steam generator channel head area to reduce dose rates. These techniques are expected to reduce the dose rates in the hot leg channel heads of the 1
steam generators by a factor of approximately 2.5. Other
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ALARA measures implemented by WE during the steam generator sleeving-demonstration project include full size mockups for training wor.kers, use of remote and semi-remote tooling when-ever practicable, and routine air, sampling, and contamination' and radiation surveys. Measures such as these are recommended in Regulatory Guide 8.8, "Infomation Relevant to Ensuring That Occupational Radiation Exposures At Nuclear Power Sta-tions Will Be As Low As Is Reasonably Achievable", in order to e
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7 minimize individual occupational radiation exposures and maintain the overall collective occupational radiation expo-sure as low as is reasonably achievable (ALARA)., No indi-vidual will be allowed to exceed the dose limits imposed for workers by 10 CFR Part 20, which are established as
- dose limits a'ppropriate to the health' and safety of individuals.
To determine the relative environmental signific'ance of the esti-mated maximum occupational dose of 60 person-rems, comparisons were made with 1) the doses expected from nomal operation of-nuclear plants, and 2) other non-nuclear ri'sks.
Table 4.1 shows the occupational dose history for Point Beach Units 1 and 2,3 When there ismore than one reactor unit at a 2
plant site (as at Point Beach) the ' combined occupational dose for all reactor units (for example, Point Beach Units 1 and 2) can be reported,3 instead of the doses for each separate unit. With the 2
addition of 60 person-rems' for the sleeving-demonstration project, the average annual dose for the 10 years of dose history at Units 1 and 2,(1970 through 1980) will be approximately 470 person-rems or an average of 235 person-rems per reactor unit. Occupational expo-sure estimates were not specifically considered in the Point Beach Units 1 & 2 FES. However, in recent environmental statements 4
for new pressurized water reactors (e.g., Summer FES), we have provided an estimate of 410 person-rems per reactor unit as G
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tt.. <erage annual occupational dose.5 This estimate is based r..
eported data from power reactors that are operating with radiation protection progiams in accordance with ifRC guidance ar.d.03ulations. A summary of these data is provided in
.le 4.2.2 These data show' th'at 410 person-r' ems per reactor u it per year is roughly the average of the wide range of
/oses incurred at'all pressurized water reactor units over t.a last se' eral years. The' amount of dose incurred at '
v any single reactor unit in a year'is highly dependent on; the amount of major maintenance performed that year.
Operating data from U.S pressurized water reactors 9
indicates that units requir~ing 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 operatiori 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.
We calcul.ite that 60 person-rems, the occupational dose estimate for the sleeving-demonstration project, cor' responds to a risk of very e
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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 8
the BEIR Report and WASH-1400 from data for.the population as a whole. New information in the BEIR III Report' would lead to an even lower estimated risk for prematur.e 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 l
Beach Nuclear Station would be expected to result in about 190 cancer deaths and about 60 geneti.c effects in the existing popula-tion (genetic effects are genetic diseases or malformations),7,10 plus about 300 more genetic effects among their descendants.
To make the health risk associated with radiation dose more under standable, risk comparisons can be made with non-nuclear activities j
l 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 ednsider 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, the lifetime risk from radiation dose for the average sleeving-
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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 cccupational mortality risks in the U.S.
One such comparison is 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 oi:cupationally related cancer estimates as
beingoverlyconservative.1'2 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 exposur,es 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 sunmary, the staff has drawn the following conclusions regarding occupational radiation dose. WE's estimate of 60 person-rem for the This sleeving-demonstration project at Point B ach 1 is reasonable.
dose is at the low end of the normal range of annual occupational
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doses which have been observed in recent years at operating Although the doses resulting from the steam generator reactors.
' 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-WE has taken appropriate steps to ensure that tenance work.
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occupational doses will be maintained within the limits of 10 CFR Part 20 and Al. ARA. 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 whol'.
e The risk to an average individual in the work force will be lower than risk incurred from participation in many commonplace activities. The indiv'idual risks associated with exposures involved in th'e sleeving-demonstration program will be controlled and' 11mited 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 will not significantly affect the quality of the human environment.
4.1.'2 Public Radiation Exoosure NRC Staff has estimated the amount of radioactivity which will be released in liquid and gaseous eff1'uents 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.
WE will take several steps to minimize releases.1 To minimize airborne releases the channel head decontamination process and the surface preparation process will be wet processes, entraining
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removed material in water. 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 snaller than that which occurs during normal operation, k'e have estimated the doses to individual members of the public as well as the population as a whole in the area surrounding Point e
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Beach Unit 1 based on the radioactive effluents 'which we estimated for the' sleeving-demonstration project (summarized in Table 4.5) and on the calculational methods presented in Regulatgry Guides I.109,15 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 the operations. This is equivalent to a dose of less than a small n
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 from liquid effluents. The offsite population dose was calculated by multiplying the (offsite) maximum individual total body dose of 7.5 x 10~0 mrem (estimated -for the liquid release of Co-60) with the projected 4
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.
Every year the same population of about819,000 wil,1 receive
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vicinity of Point Beach 1.11 Thus, the populatio,n total, body dose from the sleeving-demonstration project is less than 7.6 x 10 per-cent of the annual dose due to natural background. On these bases,
.we conclude that the doses to individuals in unrestricted areas and to the population within 50 miles due to gaseous and liquid efflu-ents from the sleeving-demonstration project will not be environmentally significant. Since we expect no larger radioactive effluents from Poin't 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 Our calculations (using the LADTAP Computer Program)17 for the maximum individual total body dose for an adult considered the following pathway consumption (1) of fish (21 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 via liquid effluents which are expected to result from the sleeving-demonstration project. The LADTAP II program implements the radiological exposure.models desgribed in U.S. NRC r 'ilatory'~iii"~
Guide 1.109, Rev.1 (Appendix a)15 foi~r'a'dioa'ctivity releases liquid effluent.
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compared to the limits of 40 CFR Part 190 and to the annual dose from n5tural 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
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which are discussed in greater. detail above.
(1) The estimated range of 48 to 60. person-rems for the sleeving-demonstration project is on t.he low side of the expected range of dosei incurred at light water power reactors in'a yeEr.
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(2) The risks to the workers involved in the sleeving-demonstration project from radiation exposure are no larger than the risks incurred by:
(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.
(4) Offsite doses resulting from the sleeving-demonstration project will be, (a) smaller than those incurred during normal operation of Point Beach 1, and (b) negligible in comparison to'the r"se members of the public in the vicinity of Point Beach I receive from natural background radiation.
On the basis of the foregoing statements, the staff concludes that
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the proposed sleeving-demonstration project at the Point Beach O
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37 Nuclear Plant, Unit No. I will not significantly affect the quality of the' human environment.
4.2 Nonradiolog[ cal Assessment We have reviewed the documents submitted by WE in support of its request to conduct the steam generator tube sleeving-demonstration program. W'e find that the proposed activities will occur within the plant on areas previously disturbed during site preparation and construction. These activities wi11 not have appreciable offsite environmental effects. The license's 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 wd11 not result in any significant environmental impact.
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5.0 BASIS AND CONCLUSION FOR NOT PREPARING AN-ENVIRONMENTAL IMPACT STATEMENT 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 detemined, based on this assessment, that this action will not significantly a.ffect the quality of th's human
- environment. Therefore, the Commission has-determined that an e
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Environmental Impact Statement need not be prepared, and that, pursuantto10CFR51.5(c)(1),theissuanceobanegative
~ declaration to this effect is appropriate.
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TABLE 4.1 2
ANNUAL COLLECTIVE, 3 OCCUPATIONAL DOSE'AT POINT BEACH UNITS
- 1, 2 Collective Occupational Dose.
(person-rems)
Year 1971 164 1972 580 1973 588
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1974 295 459 1975 1976 370 1977 429 1978 320 1979 644 1980 7913 First commercial operation 12/70 (Unit 1), 10/72 (Unit 2) e
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TABLE 4.2 2
OCCUPATIONAL DOSE AT U.S. LIGHT WATER REACTORS (person-remsperreactorunit)-~ ;
e Year Average"
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Averag's *:..,
low-M d2, 298 1969 165 195 1970 684 127 44 1639 1971 307 255 50 768 1972 464 286' 61 1032-783 380 85 5262 1973 1974 331 507 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 e
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TABLE 4.3 LIFETIME MORTALITY' RISKS 18 NUMERICALLY EQUIVALENT TO ONE REM Type of Activity Equivalent Risk to One Rem Smoking cigarettes I carton Drinking wine 66 bottles Automobile driving.
6,600 miles Commercial flying 33,000 miles
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Canoeing 1.6 days
- Being a man aged 60 1.8 days Eight hours per day e
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a TABLE 4.4 OCCUPATIONAL RISKS
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Events per year per 100,000 workers),
Mining &
- All U.S.
Radiation Quarrying Industries Trade Exposure F.atal Accidents (1) 63 14 6
1
~ Delayed Effects Actial readily Occasionally not not Observable Observable Observable Observable-Observable Estimated
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Includes 115-219
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4-6 lethal cancers (2) lethal cancers (3)
(1) 1976 data, from " Accident Facts,1977 Edition," National Safety Council.
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-Estimates from " Toxic Chemicals and Public Protection, A Report to the President by the Toxic Substances Strategy Committee," Council on Envi-ronmental Quality, Government Printing Office, May 1980. Assumes l
20-38f. of all cancers are associated with occupation.
(3)
Estimates from BEIR-II,1980, assuming an average radiation worker exposure rate of 0.5 rem /hr; exposure at the limit, 5 rems /yr, would yield'an estimate of from 37 to 63 lethal cancers per y&ar per 100,000 workers.
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TA8LE 4.5 RADIOACTIVE EFFLUENTS FROM POINT BEACH 1 III Estimates of
-WE Estimates for Point Beach 1 -
Point Beach 1 FES Type of Radioactive Releases During Sleev-
.1979 Releases 1980 Releases Annual Average Effluent
.ing Demonstration (Ci)_
(C1)
(Ci)
Releases (Ci/yr.)
Caseous
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Noble Gases-Negligible 4.8(+2)c
,3.2(+2) 5.0(' 3)
+
Iodine + Particulates*
Negligible 1.4(-2).
2.7(-3) 1.0(-1) b d
b Tritium Negligible 4.0(+2) 3.3(+2) e Liquid Mixed fission and 1.44'x 10-4
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0.38 0.63 1.0(+1) activation products.
, Negligible 4.5(+2) 3.8(+2) 1.0(+3) b aRadioactive half lives 8 days or more.
"bBelow lower limits of detectability for plant instrumentation.
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,4.8(+2) means 4.8 x 10+2 C
No estimate was given in FES, but FES stated that there would be low co.ncentrations d
of tritium to.he gaseous releases.
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References 1.
Point Beach Steam Generator Sleeving ~ Report for Wisconsin Electric 1
Power Company prepared by the Westinghouse Electric Corporation.
l September 28, 1981.
2.
NUREG-0713. Vol.1, Occupational Radiation Exposure at Comercial Nuclear Power Reactors,1979,' U.S.N.R.C.., March.1981.
3.
NRC Memorandum dated June 19, 1981, from W. E. Kreger to N. R.
Denton entitled " Unusually High Occupational Doses Reported For Power Reactors Operating in 1980."
4.
Final Environmental Statement related to operation of Point Beach
~
Nuclear Plant, Units 1 and 2. United States Atomic Energy-Commission, May 1972.
5.
NUREG-0719, Final Environmental Statement Related to the Operation of Summer Pressurized Water Reactor, 1981.
6.
NUREG-0692, Final Environmental Statement Related to Steam Generator Repair at Surry Power Station Unit 1. July 1980.
7.
The tffects on Populations of Exposure to Low levels 6f Ionizing' Radiation, "BEIR Report," report of the Advisory Committee on the 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 Leve s of Ionizing Radiation "BEIR III Report", report of the Committee on the Bio-logical Effects of Ionizing Radiation's National Academy of Sciences
- National Research Council, 1980.
- 10. 1979 Cancer Facts and Figures American Cancer Society.
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- 11. NCRP No. 45, " Natural Background Radiation in the United States,"
National Council on Radiation Protection and Measurtments,1975.
12.
R. 'Peto, " Distorting the Epidemiology' of Cancer, the Need for a More Balanced Overview," Nature 284, 297-298 (March 27,1980).
- 13. Wisconsin Electric Power Company, Point Beach Nuclear Plant Unit Nos. I 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 Monitoring Reports, January 1,1980 l
through. June 30, 1980 and July 1,1980 through December 31, 1980.
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15.
Regulatory Guide 1.109, " Cal'culation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Complia'nce with 10 CFR Part 50, Appendix I" (Revision 1),
U.S.N.R.C., October 1977.
16.
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 Liquid 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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