ML13330A182

Interim Assessment of Sleeving of San Onofre,Unit 1 Steam Generator Tubes

ML13330A182
Person / Time
Site:	San Onofre
Issue date:	11/30/1980
From:	Office of Nuclear Reactor Regulation
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ML13330A181	List: ML13330A180 ML13330A182
References
TAC-42090, NUDOCS 8101160416
Download: ML13330A182 (22)

Text

INTERIM ASSESSMENT BY THE U. S. NUCLEAR REGULATORY COMMISSION'S OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO SLEEVING OF SAN ONOFRE UNIT 1 STEAM GENERATOR TUBES DOCKET NO. 50-206 November 1980

TABLE OF CONTENTS Pages 1.0 Introduction.,

2.0 Discussion of sleeving 2

2.1 Post Process Sampling Plan for Inspecting Upper Joints 2

2.2 Deisgn. Verification Testing and Analysis 3

2.3 Inspectability of Tube-Sleeve Assembly 4

2.4 Corrosion Testing 4

2.5 Upper Joint Fabrication 5

2.6 Corrosion Test Program 6

2.7 Steam Generator Tube-Sleeve Inspectability 7

3.0 Safety Assessment Conclusions 8

4 0 Radio ogical Assessment 9

4.1 Occupational Exposure 9

4.2 Public Radiation Exposure 12 5.0 Radiological Assessment Conclusions 13 6.0 References 15

1.0 INTRODUCTION San Onofre Nuclear Generating Station, Unit 1 was shut down on April 19, 1980 with a 270 gallon per day (GPD) primary to secondary steam generator tube leak. The tubes identified.as leaking were plugged in accordance with the Technical. Specifications.. Subsequent inspections by multi frequency.eddy current techniques (ECT) revealed the 5 leaking tubes to be among approximately 1800 tubes in all three steam generators with detectable degradation at the top of the tubesheet elevation.

Laboratory examination of tube specimens removed from the steam generator revealed this degradation to be associated with caustic intergranular attack (GA) and circumferential -racking within a 1/4th inch band extending non-uni ormly around the circumference of the tubes.

Although current eddy current test capabilities are adequate to detect significant cracks, capabilities to detect intergranular corrosion attack in the absence of significant cracks. (or grain separations) are limited.

Thus, there is a concern regarding the reliability by which the tubes affected by this intergranular. phenomenon can be identified.. Fdr this reason and to minimize the number of tubes to be plugged during this or future outages, Southern California Edison Company (the licensee)(SCE) initiated a program to develop, qualify by test, and install sleeves in all steam generator tubes within the zone (consisting of approximately seven thousand. tubes) where this phenomenon is believed to be most advanced.

In order to provide the technical basis for the proposed repair program, SCE has submitted Westinghouse (Proprietary) Report No. SE-SP-40(80),

dated October 1980.

SCE sponsored a third party design review of'.the sleeving repair program.at the Westinghouse Forest Hills facility in Pittsburgh, Pennsylvania, on October 23 and 24, 1980. The purpose of the independent design review approach is to determine if the proposed action meets the applicable regulatory design criteria.

This is accomplished by using a team of independent, technical specialists to question and review the work of the responsible vendor and licensee.

This design review process is consistent with the requirements of Criterion III of Appendix B of 10 CFR part 50 and increases the licensee's involvement in demonstrating compliance with NRC safety regulations.

At the October 23 and 24, 1980 review meeting, the panel members consisted of repre sentatives from other.utilities, from EPRI, other companies, and from academia.

Although no NRC staff members were on the review.panel, the staff members were present and participated at the proceeding.

A transcript was taken at the proceeding and provided to the staff for its review.

The staff has met with SCE and its contractor for the steam generator repair (Westinghouse) on numerous occasions since July 1980 to discuss the overall sleeving program as wei 1 as information needed to support.

a request for return to power..

This interim safety assessment only addresses the sleeving operation.

Prior to authorizing a return to power, the staff must review and approve additional aspects of the overall repair program including ECCS reanalysis, post-operational inspection of tubes and secondary water chemistry.. The results of the staff review of these additional 'aspects will be published in a safety, evaluation related to the' restart of San Onofre Unit 1.

-2 As discussed in this report, we have concluded that we do not object to SCE proceeding with the production insertion phase of the steam generator

-tube repairs.

The results of our review also indicate that the proposed process is an acceptable approach to tube repair.

2.0 DISCUSSION OF SLEEVING The proposed sleeve design consists of a sleeve of wall thickness designed to provide strength comparable to the orignial tube and fabricated from thermally treated Inconel 600 tubing.. The sleeve material was selected to provide maximum resistance to corrosion and stress' corrosion cracking.

The sleeve will be inserted inside the 0.650 inch ID of the existing tube (mill annealed alloy 600) joined at the upper end to the inner surface of the tube above the tubesheet, and rolled into the tube and tubesheet at the lower end. Different sleeve lengths will be employed, with the maxi mum length sleeve being employed for each tube consistent with the available clearance with the channel head bowl upon insertion.

Each of the sleeve lengths assures that the sleeve will span the outer tube defects at the top of the tubesheet elevation.

The upper joint fabricating process was selected based upon its prior history in joints in reactor vessels and its compatabilitywith Inconel.

The sleeve design provides a sleeve extension above the upper joint. This sleeve extension was added so that if the existing outer tube were to become severed right above the upper joint, the tube would be "captured" by the sleeve, and that motion, and subsequent leakage (estimated 34 gpm maximum) would be limited.

The lower end of the sleeve will be mechanically rolled into the rolled portion of the original tube to form a leak tight seal 2.1 Process Sampling Plan for Inspecting-Upper Joints The proposed upper joint process inspection plan is designed with the following objectives:

1) to obtain 99% confidence that 99% of all upper joints will not yield a primary to secondary leak within a year,

2) to assure that the upper joint fabrication process at the outset will produce upper joints that are able to meet objective (1),.

3).to assure that later deterioration in the upper joint process will be detected quickly.

The process sampling plan statistics assumes that a tube will leak within a year due to an unacceptable upper joint only if (1) the outer tube is penetrated by a crack within a year, and (2) the upper joint is unaccept able (not a leaktight upper joint).

Estimates of the probability that a given tube will be penetrated by a crack within one year are based upon present penetration conditions (as determined from eddy current data) and estimated crack propagation rates.

-3 The licensee's sampling plan is based on inspecting a number of joints in each lot of 100 tubes and accepting or rejecting the lot.based on the re sults of the sample. We have required, and the licensee has agreed, to inspect.100% of the joints in the first two lots of each steam generator and 50% of the joints in the third and fourth.lots before going to the proposed plan.

If two 100% lots followed by two 50% lots result in no further unacceptable joints the sampling plan should shift to level 1 of the licensees proposed sampling plan for the next lot.

If an.unacceptable joint is found during inspection of the first four lots, the inspection would revert to 100% inspection followed by two 50% inspection lots until these lots are successful before proceeding to the proposed sampling plan.

At a minimum, if all joints are acceptable, 1315 sleeves will be inspected out of the almost 7000 tubes to be sleeved.

It should be noted that the acceptance criteria that will be used by SCE for the upper joint inspection are more severe.than simply requiring a leak tight joint, since these criteria were developed to satisfy ASME Code requirements and conform to the assumptions used in the Westinghouse structural evaluation.

2.2 Design Verification Testing and Analysis Structural analyses of the tube/sleeve assembly have been performed by the licensee in accordance with ASME Code,Section III and appropriate regulatory guides.

Results of these analyses indicate acceptable fatigue performance and adequate structural margins for the full.range of normal, operating, transients and accident (e.g., LOCA, MSLB) condition loadings.

The structural and fatigue analyses included consideration of stresses in the tube/sleeve assembly which could result from hourglassing (i.e. deformation) of the support plate flow slots, or from flow induced vibration.

The minimum required sleeve wall thickness to sustain loadings associated with full range of normal accident conditions is in accordance with Regulatory Guide 1.121 criteria and was calculated to be 36% of thesleeve wall thickness.

In this analysis no credit was taken for any remaining wall thickness in the outer tube.

The. structural analyses of the tube/sleeve assembly have been supplemented by extensive mechanical testing by Westinghouse of the upper joint (including tests for both internal and external pressure loadings and fatigue) to verify acceptable structural and leak tight integrity of the upper joint.

The combined pressure and fatigue tests were l.ast reported (October 1980) to be still underway and will be continuing even after the restart of the plant or until failure of test samples.

No problems or failures of test samples have been identified to date. Tensile tests of the tube/sleeve joint performed at 600'F confirm the tensile strength of the joint to.exceed that of the tube itself.

-4 2.3 Inspectability of Tube-Sleeve Assembly As required by the Technical Specifications, inservice inspections of the sleeve/tube assembly will be performed using multi-frequency ECT. Data submitted to date indicate that the use of conventional.(circumferentially wound) eddy current coil probes will provide adequate inspection capability in all but the upper joint region.

The inservice inspection of the upper joint region will present some difficulties. Eddy current.tests traditionally suffer from a reduction in sensitivity where there is a geometric discontinuity and a.number of such discontinuities exist at-the upper joint.

Westinghouse is currently investigating ECT procedures to improve the inspectability of this region including the use of magnetic bias techhiques and alternate probe types such as the crosswound probe, the rotating pancake

probe, and the multicoil surface riding probe.

In the-joining material itsalf, Westinghouse does not expect that eddy current testing by itself will be capable of detecting a slow degradation of the joining material.

Even if a total dissolution of the joining material is assumed, Westinghouse notes that the* tube would be captured by the.sleeve.

The sleeve extends

.75 inches above the upper. Joint which would attenuate the leakace flow rate.

As an interim measure, the licensee has proposed a "leader tube" Program as a method to monitor the integrity of the upper joint region.

This program will consist of eight (8) tubes in each steam generator with through wall penetrations deliberately made in the tube wall section spanned by the sleeve prior to sleeving. These tubes will function as "leaders" with.regard to upper. joint.exposure to secondary side environmental conditions.

At the.end of the first inservice inspection interval, these tubes would be. pressure tested to 3000 psi to demonstrate the margin existing in the joint (a factor 1.4.above the postulated Main Steam Line Break pressure loading).

Following this test, a sample of the leader tube group would be removed and examined nondestructively and then pressurized destructively (to burst)...The removed tube samples will'be further destructively examined in the laboratory to see whetner or not there are signs of upper joint degra dation.

After such periodic examinations, sufficient information will be available to provide a data base for future remedial actions if.needed, as appropriate.-

2.4 Corrosion Testing The licensee has provided a corrosion testing program to demonstrate that the sleeving process does not introduce any new mechanism that might result in premature tube or sleeve degradation.

The specific corrosion issues addressed by the test program are (1) the effect of the upper joint fabrication process on corrosion resistance of the tube and sleeve, the possible diffusion of 00 contaminants into the tube and the possible sensitization of the tube and sleeve, (2) the effect of residual joining material. in the primary side crevice, (3).the compati bility of the joining.material in primary and secondary water and (4) the effect of outside tube leakage and exposure of the sleeve to secondary side contaminants.

Tests addressing corrosion issue (1) above, were 10% Na0H controlled potential tests, 680 0 F primary water U-bend tests, metallography, micro probe analysis, Huey test and reactivation polarization testing.: Pres surized capsule.tests withr650aF primary water will. test the effects of residual joining material.

C-ring tests exposed to 10% Na0H at 600'F and 650aF and U-bend tests at 680 9F primary water test the compatibility of the upper joint with both sleeve material and outer tube material.

A test covering all four corrosion issues is conducted with entire upper joint assemblies tested under simulated steam generator conditions in model boilers.

In these tests, general corrosion, galvanic corrosion, and stress corrosion cracking resistance of the combined assembly are tested.

In the boiler, a temperature differential is maintained between the simulated primary and secondary fluids to concentrate corrosive species in crevice areas.. Concentrating devices (sludge collars) and EDM slots

• are used to simulate corrosion inducing operating conditions with caustic, phosphate and chloride present. Recently completed model boiler tests of a duration of 20 days and of 30 days have shown no significant corrosion had occurred.

2.5 Upper Joint Fabrication The steam generator tube/sleeve joint is constructed by placing a thermally treated Inconel 600 sleeve inside the original steam generator tube, expanding the lower end of the sleeve by rolling and joining the upper end of the sleeve to the tube near the top end of the sleeve. We have reviewed the fabrication procedure and qualification used to demonstrate the integrity of the upper joint. Based on our review, we have determined that the upper joint has been fabricated and qualified in accordance with the ASME Boiler and Pressure Vessel Code.

In addition to meeting the code requirements the licensee has performed further tests and examinations to evaluate the integrity of the sleeve to tube upper joint.

The licensee's contractor, Westinghouse has performed a large number of mechanical strength tests on the upper joints.

These tests were conducted on specimens that were joined over the range of parameters that.would be used.in.the field fabrication operation at San Onofre 1.

The results from these tests demonstrate that the strength of the upper joint is maintained over the range of.joining conditions that will be used in the field operation.

Westinghouse also performed destructive examinations of the sleeve-to-tube upper joints to evaluate

-6 the integrity of the joint. These examinations consisted of sectioning the joint and observing the material structure at high magnification.. The re sults of these examinations indicated that a good quality joint can be pro duced by the fabricating operation and that the fabricating operation does not produce any anomalies in the sleeve or tube materials that would com promise the integrity of the joint.

Based on our review of the.joining procedure and the mechanical testing and destructive evaluations that have been performed to qualify.the joint, we conclude. that the fabricating procedures will produce a sleeve-to-tube joint that is of high quality and has acceptable mechanical strength.. and metallurgical characteristics.

To. verify that the upper joint integrity is achieved during the field fabricating operation, the licensee has also developed a nondestructive

-examination (NDE) technique to examine the upper joint prior to service.

This examination technique consists of a non-destructive, volumetric examination of. the upper joint area around the circumference of the sleeve.

The licensee has demonstrated this technique by using small drilled holes in sleeve-to-tube joint samples.

Based on our evlauation of the NDE technique and the test results from simulated.field.samples, we conclude that the NDE techniques proposed by the applicant can detect relatively small defects in the upper joint and will ensure the preservice integrity of the upper joint.

I.n summary we conclude that the applicant has demonstrated that the sleeve-to tube upper joint can be made in a manner that will 'produce acceptable strength and metallurgical properties and that the preservice integrity of the field fabricated upper joints can be ensured by implementing the proposed non destructive examinations.

2.6 Corrosion Test Program We have reviewed the corrosion testing methodology used by the licensee to qualify the sleeving procedure with special reference to the model boiler tests and we find that the program is sufficiently comprehensive to cover the corrosion issues surrounding the sleeving procedure.

For example, we evaluated the potential for corrosion in the area of the upper joint in the steam generator tube with respect to crevice stress corrosion cracking in secondary water due to the presence of the sludge pile. We conclude that the model -boiler tests adequately address this corrosion issue since an upper joint assembly surrounded by a sludge collar is subjected to simulated secondary water with impurities added.

The successful results of the 20 day and 30 day model boiler tests on complete upper joint assemblies along with the favorable results of the individual tests, provide reasonable assurance that the sleeving procedure will not result in premature or sudden tube or sleeve degradation or corrosion in duced failure.

These conclusions are based on the ability o.f the model

7 boi er tests to s Imulate the combined corrosion effects of crevice con centration, off-chemistry effects, galvanic coupling and heat flux conditions of steam generator service and accelerated corrosion testing procedures.

Long term.assurrance that the integrity of the tubes will be maintaifned is provided by the planned continuation of the model boiler tests.

We conclude that based on the above there is reasonable assurance that the sleeving processes will not degrade the integrity of the steam generator tubes or induce corrosion failure.

2.7 Steam Generator Tube-Sleeve lnspectability The inspectability of the tube/sleeve assembly is still being evaluated by the staff.. Based upon laboratory eddy current testing.(ECT) data provided by Westinghouse and our experience regarding multifrequency ECT capabilities, available eddy current techniques and alternate probe.

designs can be adapted to permit adequate inspection of the s eeve assemblies away from the upper joint region.

As stated in 2.3 above the upper joint region presents the most inspection difficulty. However, additional assurances of joint integrity can be obtained through in-situ pressure testing of individual tube/sleeve assemblies, and through a program. to remove a sample of tubes for laboratory analysis and testing. The "leader program" proposed by the licensee provides on example as to how.such a program could be accomplished. We do not plan to evaluate the licensee's specific proposals regarding the "leader group" program until inspection capabilities in the upper.joint region have been better defined by the licensee and Westinghouse.

The licensee has not made a specific proposal regarding which insetvice inspection techniques and equipment will be employed during its first inspection following restart.

We are requiring additional inservice inspection qualification data for critical locations of the tube/sleeve assembly, and justification for the type of inspections and equipment to be used during the inspection.

-8 3.0 Safety Assessment Conclusions As previously indicated, this evaluation only addresses the steam generator sleeving program. Other activities associated with the restart. of San Onofre Unit 1 will be addressed in a safety evaluation report that is related to the restart.

As a result of the staff review, we have concluded that the sleeving operation is a sound repair technique for the San Onofre Unit 1 steam generators.

SCE may proceed with the production. insertion phase of the steam generator tube repairs.

4.3 Radiological Assessment 4.1 Occupational Exposure We have reviewed the occupational radiation exposure expected to result from the steam generator tube sleeving planned for San Onofre Nuclear Generating Station, Unit 1, and the steps to be taken to assure that the occupational radiation exposures will meet the requirements of 10 CFR Part 20, Standards for Radiation Protection, and will be as low as is reasonably achievable (ALARA),. The major source of the radiation dose rate inside the steam generator channel head-is a tenacious layer of "oxide" which includes deposited activated corrosion products. Southern California Edison.(SCE) has used a Westinghouse mechanical decontamination process involving grit driven by a high pressure water spray to remove this deposited activity from the inside of the channel head. Based on Westinghouse experience with grit/high pressure water decontamination, SCE expects to reduce dose rates inside the channel head -y a factor of 5 (after decontamination, surface cleaning and shielding).

The surface preparation.will.remove deposited radioactivity in the bottom few feet inside each tube, thereby reducing gamma radiation shine to the workers from the tubes. To further reduce the doses to personnel working in the channel head, lead shielding will be placed on a) the cold leg side of the divider plate to eliminate cold leg streaming to the hot leg plenum, b) over the opening of the hot let inlet pipe, and c) on the bottom of the tube sheet to cover portions of the tube sheet not being worked on. These decontamination and shielding techniques are expected to reduce the dose rates inside the hot leg channel heads from approximately 10 rem/hr to approximately 2 rem/hr. In order to minimize worker time speht inside the steam generator plenum, SCE intends to use remote techniques where they are practicable. Most of the channel head decontamination and preparation of the inside surfaces of the tubes will be done remotely. Most of the joining of the sleeves inside the tubes and the non-destructive examination of the repaired sleeves will be done remotely.

SCE-and Westinghouse will provide training to the maintenance crews to minimize the time spent in the radiation fields. As recommended in Regulatory Guide 8.8, "Information Relevant To Ensuring That Occupational Radiation Exposures At Nuclear Power Stations Will Be As Low As Is Reasonably Achievable", this training will include the use of full-scale mockups of the steam generator plenum and surrounding work area. The-same tools used to train the workers will be used during the sleeving work. All workers will wear a series of TLD's to measure doses to their chests, heads and extremities.

Themethods used by SCE to develop a collective occupational radiation exposure for the entire sleeving project are based on actual experienc and testing.

SCE 1) determined the maintenance activities that will be involved in the sleeving program; 2) estimated the person-hours.of work necessary to perform those activities; 3) determined the areas maintenance personnel.must occupy to perform 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.

The dose rate estimates mentioned above are the measured dose rates at San Onofre with adjustments for decontamination and shielding. The resulting

-10 estimate for total collective occupational radiation exposure for the project is 1800 person-rem. Based on our review, we conclude that the licensee s estimate-is reasonable.

No individual will be allowed to exceed the dose limits imposed for workers by 10 CFR Part 20, which are established as dose limits appropriate to the health and safety of individuals.

To determine the relative environmental significance of the estimated 1800 person-rem, comparisons were made with 1) the doses expected from normal operation of nuclear plants, and 2) other non-nuclear risks.

Table 4.1 shows the occupational dose history for San Onofre 1.1, With the addition of 1800 person-rem for the sleeving project, the average annual dose for the twelve years of dose history at Unit 1 will be less than 500.

person-rem. Occuoational exposure estimates were not specifically considered in the San Onofre 1 FES.

In recent environmental statements for new nuclear power plants we have provided an estimate of 500 person-rem per reastor unit as the average annual occupational dose (e.g. San Onofre 2, 3 DES).

This estimate is based on reported data from power reactors that are operating with radiation protection programs in accordance with NRC guidin5e and regulations. A summary of that data is provided in Table 4.2.

Those data show that 500 person-rem per reactor unit year is roughly the average of the wide range of doses incurred at all light water cooled reactor units over the last several years. The amount of dose incurred at any single reactor unit in a year is highly dependent on the amount of major maintenance performed that year. Every year several units require some items of major maintenance which result in doses for those units well above the annual average of 500 person-rem.

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 1800 person-rem estimate for the sleeving project is within the historical range of doses about the average for one unit in a year. This collective exposure would represent about 10% of the anticipated 40 year collective exposure of approximately 18,000 person-rem, based on the average annual exposure at San Onofre, Unit 1 so far, and less than that for the plant life average for all plants.

We calculate that 1800 person-rem, the occupational dose estimate for the sleeving project, corresponds to a risk of less than one premnature fatal cancer in the exposed work force population. We also calculate that.

1800 person-rem corresponds to a risk of less than one genetic effect to the.ensuing five generation.

These risks are based on risk.estimators derived in the BEIR Report and WASH-1400 a from dpta for the population as a 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 would be expected to result in about 190 cancer deaths and abu 4

genetic effects (genetic effects are genetic diseases or malformations),

plus about 300 more genetic effects among their descendents.

To make the health risk associated with radiation dose more understandable, risk comparisons can be made with non-nuclear activities commonly partici pated in by many individuals. One rem of radiation is numerically comparable.

to a lifetime mortality risk of About 10-4.2 Table 4.3 presents the equivalent risk of 10-4 for several common activities risks which many people take routinely and consider to be insignificant. 6 The average dose to a worker for the sleeving project will be roughly 1.8 rems.

As Table 4.3 shows, the lifetime risk from radiatioh dose for the average sleeving project. worker is smaller than the. lifetime risk associated with many common activites.

Another perspective of an occupational risk comes from comparison of occupational 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 exoerienced at an averace radiation worker exposure rate, -results in a relatively low occupational risk..

Some have cri icized occupationally related cancer estimates as being ov erly conservative.

However, most experts.feel.the risk estimates in i-ble 4.4 relating to occupational exposure to low-LET radiation are overestimates.

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 per year do no.t significantly affect.a'a typical Worker's total risk of mortality.

In summary, the staff has drawn the following conclusions regarding occupa tional radiation dose. SCE's estimate of 1800 person-rem for the sleeving project at Sah Onofre 1 is reasonable. This dose falls within the normal range of annual occupational doses which have been observed in recent years at operating reactors. This dose will not increase the annual average dose at San Onofre 1 above 500 person-rem per year, the overall power reactor average. SCE has taken appropriate steps.to ensure that occupational doses will be maintained within the limits of 10 CFR 20 and ALARA. The additional health risks due to these doses over normal risks are quite small, 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 the risk incurred from participation in many commonplace activities. The individual risks associated with exposures involved in the repair program will be controlled and limited so as not to exceed the limits set forth in 10 CFR Part 20 for occupational exposure. For the foregoing reasons, the Staff concludes that the environmental impact due to occupational exposure will not significantly affect the quality of the human environment.

4.2 Public Radiation Exposure SCE has estimated the amount of radioactivity which will be releajed in liquid and gaseous effluents as a result of the sleeving project.

Those estimates are presented in Table 4.5. The estimates are based on past effluent experience at San Onofre 1 during.steam glgerator ma tenance.

Table.4.5 also presents ffluent releases for 1978 and 1979 from San Onofre 1 and the FES annual average effluent release. estimates.

12 SCE will take several steps to minimize releases.

To minimize airborne releases the channel head decontamination process and the surface prepara tion process will be wet processes, entraining removed material in water.

Also, an enclosure tent will be erected around the steam generator channel head manway.to prevent the release of contamination from the steam generator.

Air from the enclosure will be exhausted through a high efficiency particu late filter. The water from the decontamination process will be treated by filters and demineralizers to minimize liquid releases. The main source of liquid releases, laundry waste water, will be treated by demineralizers.

We have reviewed SCE's estimates of effluent releases for the sleeving project.

Those estimates are based on estimating methods acceptable to the Staff and actual releases from similar operations at San Onofre 1 and elsewhere.

Based on our review, we conclude that SCE'.s estimates are reasonable.

As Table 4.5 shows, the expected releases from the sleeving project are small compared to both the FES estimates and San Onofre's actual annual releases.

Our estimates of doses to individual members of the public as well as the population as a whole in the area surrounding San Onofre are based on the radioactive effluents which SCE estimated for the sleeving project (summarized in TabJ2 4.5) jd on the calsulational methods presented in Regulatory Guides 1.109, 1.11, and 1.113.

The doses to individuals offsite from the sleeving project will be less than 10% of the limits.of 40 CFR Part 190.

The annual limits of 40 CFR Part 190 are 25 millirem to the total body or any organ except.the thyroid and 75 millirem to the thyroid. The doses to the population within 50 miles will be less than 1 person-rem to the thyroid or total body from liquid effluents, and less than 1 person-rem to the thyroid or total body from airborne effluents. Every year the same population of about 7 million will receive a cumulative total body dose of more than 700,000 person-rem from the natural background radiation (about 0.1 rem per year) in the vicinity of San Onofre.

Thus, the population total body dose from the sleeving project is less than 0.0001 percent 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 effluents from the sleeving project will not be environmentally significant.

SCE has estimated that the sleeving project will generate about 300 cubic meters of solid waste containing about 100 curies of radioactivity.

These estimates are based on SCE's experience with solid wastes during steam generator maintenance at San Onofre 1 and on Westinghouse experience with the wastes generated during steam generator channel head decontamination.

We have reviewed SCE's estimates of solid radwaste for the sleeving project.

Those estimates are based on acceptable estimating methods and-on solid radwaste experience with similar operations at San Onofre 1 and elsewhere.

Based on our review, we conclude that SCE's estimates are reasonable.

13 During 1978 and 1979, San Onofre 1 generated an annual average of 108 cubic meters of solid waste per year containing 129 curies of radioactivity.

Therefore, the estimated volume and radioactive content of the solid wastes from the sleeving project are of the same order of magnitude as the solid waste normally generated annually at San Onofre. This amount.of waste is a very small part of the amount of solid waste handled each year by the nuclear industry;, therefore,, this waste does not represent a siqnificant affect ch the quality of the human environment.

The sleeving project will not change the plant in any way. which might increase effluent releases. Therefore, we conclude that the impact from routine releases of radioactivity-after the sleevine project will be no

.greater than the impact before the project. In fact, the sleeving operation should result in decreased steam generator tube leakage thereby reducing future radioactive effluents when compared to past facility operation.

Since we expect no larger radioactive effluents from San Onofre after sleeving (over presleeving operation), we conclude that the impact on biota other than man will also be no larger after sleeving.

In summary, the radioactive releases resulting from the sleeving project will be less than those due to normal plant operttion. These releases are also much less than-the estimates presented in the FES.

The doses due to these releases are small compared to the limits of 40 CFR Part 190 and to the annual doses from natural background radiation.. SCE plans to take the necessary steps to minimize releases. The estimated amount of

.solid wastes for the sleeving project is not greatly different than normal.

Therefore, the radiological impact of the sleeving project will not significantly affect the quality of tne human environment.

5.0 Radiological Assessment Conclusions Based on our review of the' proposed steam generator sleeving project, we have reached the following conclusions which are discussed in greater detail above.

(1) Includinq the dose from sleeving, te annual average dose at San Onofre I over the life of the plant to date will still be less than 500 person-rem, the annual average dose for all light water reactor units.

2 The estimated dose of 1800 person-rem for the sleeving project is within the expected range of doses incurred at light water power reactors in a year.

14 (3) The risks to the workers involved in the sleevinq 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.

(4) SCE has taken abprooriate steos. to ensure that occuoational dose will be maintained as low as is reasonably achievable and within the limits of 10CFR Part 20.

5 Offsite doses resulting from the sleeving project will be (a) no larger than those incurred during.normal operation of'San Onofre 1 and (b smal in comparison to the doses members ofi the public in the vicinity of San Onofre 1 receive from natural background radiation.

On the basis of the foregoing statements, the staff concludes that the proposed sleeving project at San Onofre Nuclear Generating Station,.Unit No. 1 will not significantly affect the quality of the human environment.

We have reviewed this proposed facility modificatioh relative-to the require ments-set forth in 10 CFR Part 51 and the Council of Envir'onmental Quality's Regulations: 40 CFR Part 1500 et se

. Wehave determined that the profposed action will not significantly affect the quality of:,the human environment.

Attached:

Tables (5)

-15 6.0 -REFERENCES

1. NUREG-0594, Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1978, U.S.N.R.C., Novenber 1979.

2. The Effects on Populations of Exposure to Low Levels of 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.

3. Annual Report by Southern California Edison in compliance with 10 CFR Part 20.407, San Onofre Nuclear Generating Station, submitted to U. S.N.R. C., 1980.

4. Letter dated November 26, 1980, from K. P. Baskin, Southern California Edison Company, to 0. M. Crutchfield, U.S.N.R.C.

5. NUREG-0490, "Draft Environmental Statement - San Onofre Nuclear Generating Station, Units 2&3", USNRC, November 1978.

6. Final Environmental Statement related to operation of San Onofre Nuclear Generating Station Unit 1, United States Atomic Energy Co mission, October 1973.

7.

The Effects on Population of Exposures to Low Levels of lon zing Radiation "BEIR III Report", report of the committee on the Biological Effects of Ionizing Radiation's Natural Academy of Sciences National Research Council, 1930.

8. NCRP No.

45, "Natural Background Radiation in the United States," National Council on Radiation Protection and Measurements, 1975.

9. R. Peto, "Distorting the Epidemiology of Cancer, the Need for a More Balanced Overview," Nature 284, 297-298 (March 27, 1980).

10. "1978 Effluent Report, San Onofre 1.

1. "1979 Effluent Report, San Onofre,

12. Regulatory Guide 1.109, "Calculation 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.

13. Regulatory Guide 1.111, "Methods of Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactos" (Revision 1), U.S.N.R.C., July 1977.

14.

Regulatory Guide 1.113, "Estiiua ting Aquatic Dispersion of Effluents from Accidental and Routine Reactor Rclc ses for the Purpose of Implementing Appendix I," U.S.N.R.C.

15. 1979 Cancer Facts and Statistics, American Cancer Society.

16.

WASH-1400, "Reactor Safety Study -

An Assessment of Accident Risks in U.S.

Commercial Nuclear Powcr Plants," U.S.N.R.C.,

October 1975.

17. E. Pochin, "The AcccptaInce of Risk," British Hedical Sulletin 31(3), 1975.

ANNUAL COLLECTIVE1 3 OCCUPATIONAL DOSE AT SAN ONO FRE' TABLE 4.1 Collective Occupational Dose

Year, man-rems) 1969 42 1970 155 1971 50 1972 256 1973 353 1974 71 1975 292 1976 880 1977 847 1978 401 1979 139

TABLE 4.2 Occuoational Dose at U.S. Light Water Reactors (man-rems per reactor unit)

Year Average Low High 1975 475 21 2022 1976 499 74 2648 1977 570 87 3142 1978 497 158 1621 1979 593 30 1793

TABLE 4.3 Lifetime Mortality Risks Numerically Equi alent to One Rem Type of Activity Equivalent Risk to One Rem Smoking cigarettes 1 carton Drinking wine 66 bottles Automobile driving 6,600 miles Commercial flying 33,000 miles Canoeing 1.6 days*

Being a -ran aged 60 1.8 days

• Eight hours per day

TABLE 4.4 OCCUPATIONAL RISKS (Events per year per 100,000 workers)

M iining &

All U.S.

Radiation Quarryina Industries Trade Exposure Fatal Accidents 63 14 6

Delaved Effects readily Occasionally not not Actual Observable Observable Observable Observable Includes 115-219 4-6 lethal cancers( 2) lethal cancers i1976 data from."Accident Facts, 1977 Edition," National Safety Council.

(2)Estimates fromin =oxic. Chemicals and Public Protection, A Report to the President by the Toxic Substances Strategy Committee," Council on Environmental Quality, Gpverrnent Printing Office, May 1980. Assumes 20-38% of all cancers are associated with occupation.

(3)Estimates from BEIR-III, 1980, assuming an average radiation worker exposure rate of 0.5 rem/yr; exposure at the limit, 5 rems/yr, would yield an estimate of from 37 to 63 lethal cancers per year per 100,000 workers.

TABLE 4.5 RADIOACTIVE EFFLUENTS FROM SAN ONOFRE 1 Type fRadioactive SE Estimates4 for San Onofre 1 San Onofre FES6 Estimates of Annua1

, fuent Releases During 1978 Releases 1979 Releases Average Releases Cilyr)

Sleeving -(Ci)

(Ci) e Gaseous Noble Gases Negligible 1.8(+3)k 6.3(+2) 4.4(+3)

Iodine Negligiblet 2 2.1(-4)

].2( 4) 5.5(-2)

Particulates 1.2(-5) 2.5(-3) 2.1(75)

Tritium 4.7 2.7(+1) 2.8(+1)

Liquid Mixed fission and activation products 4.3(-1) 1.2(+1) 1.1(+1) 1.8(+1)

Tritium 1.1 2.4(+3) 2.3(13) 8.8(+3)

• • No estimate was given in FES, but FES stated that there would be particulates and tritiun In the gaseous releases.

-3

• 18(+3) means 1.8 x 1o.

+Below lower limits of detectability for plant instrumentation.