ML20083G228

Safety Evaluation Supporting Amend 89 to License DPR-65

ML20083G228
Person / Time
Site:	Millstone
Issue date:	12/30/1983
From:	Office of Nuclear Reactor Regulation
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ML20083G220	List: ML20083G219 ML20083G228
References
NUDOCS 8401110194
Download: ML20083G228 (21)

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{{#Wiki_filter:--. SA NTG y UNITED STATES y g NUCLEAR REGULATORY COMMISSION g. p wasm NG TON, D. C. 20555 %,...../ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 89 TO FACILITY OPERATING LICENSE N0. DPR-65 NORTHEAST NUCLEAR ENERGY COMPANY, ET AL. MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 .~?lm M ,7.. 'J,-

1. 0 Introduction

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5 By letter dated June 3, 1983, Northeast Nuclear Energy Company Y.N, (licensee) proposed tevisions to the Millstone Unit 2 Technical

" A.?;..., Specifications (TS) which would allow for the repair of defective c<. - steam generator tubes by plugging or sleeving. Repair is CV$ required when a tube contains an imperfection which exceeds the ][( P ugging limit of 40% of the nominal wall thickness. l In addition to the attachments to the June 3, 1983 letter describing the sleeving program, a Westinghouse Electric Company Report, WCAP-10267 (Proprietary), " Millstone Unit 2, Steam Generator Sleeving Report," was also submitted by letter dated August 18, 1983. In response to our request for further information on eddy current inspection techniques, a letter dated November 16, 1983 was provided. The tubes to be sleeved weie selected by the licensee after a review of eddy current test data. Selection was based on the location of the tube in the tubesheet, tooling accessibility, the location and size of the eddy current indication and ALARA considerations.

2.0 Background

Millstone Unit No. 2 is a Combustion Engineering design, two loop pressurized water reactor rated at 2700 megawatts thermal. The two steam generators are of the vertical shell U-tube type with each rated at 5,603,000 lb/hr steam flow at 870 psig. Each steam generator contains 8519 heat transfer tubes. The tube material is Inconel 600 with dimensions of 0.750 inch 0.0. and.048 inch nominal wall thickness. The tubes are fully expanded into the tubesheet and seal welded. Millstone Unit No. 2 began commerical operation in December 1975. All volatile treatment (AVT) secondary water chemistry control has been used since initial operation. Full flow condensate polishing was introduced in November 1977. 8401110194 831230 PDR ADOCK 05000336 p PDR

5 S Steam generator tube degradation experienced at Millstone has included 1 tube support plate and eggcrate denting and tube pitting. Between 5 November 1977 and December 1980, 361 tubes in steam generator #1 t and 439 tubes in steam generator #2 were preventatively plugged because of denting. During the 1981/1982 refueling outage, eddy current examination of heat transfer tubing in both steam generators E revealed indications of secondary side tube degradation in the hot and cold legs of both steam generators. Eddy current testing (ECT) characterized the degradation as discrete and small volume defects (confirmed later as pits) located within the tube bundles between the tubesheet secondary face and the lowest eggcrate support. Estimated depth of the indications varied from less than 20 percent through wall to essentially through-wall. The sleeves have been designed to span degraded regions of tubes in order to maintain them in service. Degradation due to pitting attack has occurred in both the hot and cold legs of the tube S bundle with most located on the cold leg and confined to a region 7 approximately one foot above the tubesheet. s niques were verified, and material properties are in conformance =F with ASME 58-163 and ASME Code Case N379. Installation and inspe.ction processes, parameters, and procedures were developed 7 and tested. Laboratory testing included metallurgical evaluations, corro; ion testing and mechanical -testing of materials and fabricated + ? tubs / sleeve asserSlies. Extended corrosion and additional verification ? tests are continuing for information purposes. Analytical work was performed to verify the structural integrity of the sleeve design _i and its effects upon the overall nuclear plant system in accordance i with applicable ASME Boiler and Pressure Vessel Code and Nuclear 7 Regulatory Commission regulations and guidelines. } A considerable amount of actual field experience in installing t sleeves has been obtained on other installations. The process T tooling, techniques and sequences are essentially the same as = those sleeving programs utilized previously by Westinghouse a Electric Company. Although the steam generator is of the 7_ Combustion Engineering design, the sleeving technology is fully $3 applicable since the only significant difference is the exact g tube dimensions. j 5 $b i w

,,,4 o. At the San Onofre Unit '1, more than 6,400 degraded tubes (including leakers) were sleeved, tested and returned to service using remote installation equipment. After this project, the tooling was redesigned, to incorporate the field experience. The resultant remote sleeving system was adapted to a Westinghouse Model 44 series steam generator and utilized in the sleeving operatiuns at Indian Point 3. Process modifications were employed during the installation of 13 demonstration sleeves in a hands-on mode at the Point Beach Unit 1. To date, more than 12,000 sleeves have been successfully installed utilizing both remote and hands-on tooling. This sleeving system was modified to perform under the field conditions of the Combustion Engineering steam generatc:s at Millstone 2. 3.0 Sleeving Process Description 3.1 Sleeve Design The sleeving process consisted of installing within the original steam generator tube, a smaller diameter tube to span the degraded section of the parent tube. The Westinghouse sleeve design employed is a [$1 inch long,, thin walled, Ml s,leeye _ joined to the steam sgenerator. tube at both the s'ers e's upper and lower ends. The sleeve'.s lower joint is located ~ at the tube inlet end near the tubesheet's primary side surface. The [G inch length of the sleeve will ensure the upper joint is located above the segment of each tube which is degraded due to pitting and that it is above the existing sludge pile. The tubesheet is 21.5 inches thick and the sludge pile and pit defect heights range from approximately 0 to 11 inches above the tubesheet as determined by ECT during the last refueling outage. The sleeves are fabricated from thermally treated bimetallic [M' will provide greater pitting corrosion resistance in environments representative of Millstone's secondary side conditions, [M will provide optimal stress corrosion resistance to primary water and secondary side (caustic) environments known to have resulted in tube degradation in other operating nuclear units' steam generators. The upper and lower joints are designed to be the structural joints and thus also serve as the leak limiting seals. The design ciiteria for these Joints are ( that they are structurally adequate to maintain the steam generator tubing primary-to-secondary pressure boundary under normal and accident conditions, (2) that they are sufficiently leak

_ _ _ _ _ _ _ _ _ limiting such that total leakage between the primary and secondary for all sleeved tubes is less than plant technical specification limits during normal operation and postulated accidents, and (3) that they do not impair the pressure retaining capability of the steam generator tube. Hydrostatic testing of pilot sleeve joints at pressures simulating normal operation and faulted conditions have shown acceptable leak-limiting characteristics of the sleeve assembly. Based on 1500 sleeves per steam generator, the Technical Specification primary to secondary maximum allowable leak rate of 0.5 gallons per minute per steam generator, is equivalent to an allowable leak' rate per sleeve of 0.0002 gallons per minute. The maximum leak rate per sleeve under the accident conditions of steamline break and LOCA are [6] and W], respectively, based on W] sleeves. The sleeve design, materials, and joints were designed to the ASME Boiler and Pressure Vessel Code, 1980 Edition including the 1980 Winter addenda. The sleeve design also met the requirements of the original Millstone Unit No. 2 steam generator design specification under normal and design basis accident conditions (Loss of Coolant Accident and Main Steamline Break). 3.2 Bimetallic Material Corrosion Testing In addition to verifying the fabricability of the bimetallic sleeve, the effects of processing parameters on mechanical pro erties, micro-structure, and corrosion performance were evaluated. Metallurgical evaluations included hardness measurements, tensile tests, burst tests, free expansion tests, metallography and stress corrosion cracking testing. Corrosion testing of the bimetallic materials included beaker., electrochemical, autoclave immersion, and heat transfer tests. Test environments representative of the Millstone Unit No. 2 secondary side sludge pile as well as aggressive pitting and stress corrosion environments were utilized in the test program. Beaker test results revealed no evidence of ittin ] Beaker test results showed that tube strain had no effects upon pit initiation.

~ 3.3 Sleeve Installation and Verification The sleeve installation process consists of tube preparation, sleeve insertion, lower and upper joint fabrication, and joint and sleeve inspections. The sleeve installation process can be performed either automatically or manually. The process tooling, techniques, and sequences are essentially the same as those sleeving programs previously utilized by Westinghouse. The process and tooling differences that exist between Hillstone and previous Westinghouse sleeving programs are the result of improvements in automatic tooling, sleeve design, and/or sleeve dimensions. The process differences are primarily with specific process parameters rather than techniques. The following installation se uence was used to install the sleeves. [ In process sampling by eddy current testing or gauging of a percentage of the sleeves was performed to confirm that the equipment and tooling was performing satisfactorily. Undersized diameters were corrected by an additional expansion step to produce the correct diameter. If it was judged necessary to remove a sleeved tube from service, due to incorrect dimensions for example, the lower portion of the sleeve was machined to provide clearance for a plug to be inserted. A system pressure test in accordance with the ASME, Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components" was conducted following sleeve installation. k

^ , 4.0 Analytical Verification Analytical structural evaluations of the Millstone Unit No. 2 sleeve and tube assembly were performed in accordance with applicable ASME Boiler and Pressure Vessel Code, Section Ill, 1980 Edition criteria. The assemblywasmodeledmathematgllyandevalgedwiththeWestinghouse computer analyses codes WECAN and WECEVAL The analytical procedure was essentially identical to that previously completed for the sleeving effort at San Onofre 1, Point Beach and Indian Point 3. Millstone steam generator design loads (including the faulted condition) and test loading conditions constituted the bases for the analyses. Analyses were performed for both the upper and lower joints (Figures 1 and 2). Both intact and severed steam generator tubes were considered. Special considerations and additional analytical evaluations included the following. (1) Allowable Sleeve Degradation per Regulatory Guide 1.121 Criteria (2) Tube Vibrational Analysis (3) Evaluation of effects due to prebowing of sleeves (4) Effect of Tubesheet/ Support Plate Interaction. The sleeve / tube assembly consists of an upper joint, a lower joint and straight sections of the sleeve and tube between the two joints. Two finite element models were developed to represent the hybrid expansion joint (HEJ) of the upper joint and the lower joint areas. In developing the roll region of the model, the estimated bulging at the OD of the sleeve and tube were evaluated in combination with the nominal tube and sleeve wall thickness. However, the slight wall thinning was neglected for the stress levels computed in the roll transition regions. The tubesheet is low alloy carbon steel, SA 508 nconel cladding on the primary face. Finite element models were developed for evaluating the sleeve configuration

• '"sh5wn~in, Figures 1 and 2 (P'oprietary).,, Thermal analysis was performed r

to obtain the temperature distribution needed for the thermal stress evaluation. Since the thermal stress solutions were used for the fatigue calculations, the maximum range of the stress intensities during i any of the loading conditions considered were calculated, in previous transient runs done on the tube sleeves, it was observed that the temperature responses closely followed the variations of the boundary temperature. Thus it was unnecessary to perform time history analyses. The WECAN models were used to determine the stress levels in the tube / sleeve configuration including the rolled transition region for both the temperature and pressure loading conditions. At any given point or section of each model, the program WECEVAL determined the total stress distribution for a given loading condition and categorized that total distribution per the ASME Section NB requirements. The total

stress of a given cross section through the thickness was categorized into membrane, linear bending and non-linear components. These categorized stresses were then compared with Subsection NB allowables. In addition, when supplied with a complete transient history at a given location in the model, the program WECEVAL calculated the total cumulative fatigue usage factor per Code Paragraph NB-3216.2. For the fatigue evaluation, the effect of local discontinuities was considered at each location. The ASME Code stress criteria which were satisfied are as follows: 1. Primary General tiembrane Stress Intensity: Pm (a) Design Condition: Pm 1 Sm (b) Test Condition: Pm 1 0.9 Sy (c) Abnormal Conditions: Upset Pm 1 Sm Faulted Pm < 0.7 Su or 2.4 Sm whichever is lower where Sm = stress intensity limit as defined in the ASME Section III Code, Sy = yield stress at temperature and Su = ultimate tensile stress at temperature. 2. Primary local Membrane Stress Intensity: (a) Design Conditions: P 1 1.5 Sm -(b) Test Conditions: P 1 1.35 Sy (c) Abnormal Conditions Ups t: P 1 1.5 Sm Faulted P < 1.5 x 0.7-Su whichever is lower or 1.5 x 2.4 Sm 4.1 Results of the Analyses '(a ) Primary pressure stresses: The maximum primary pressure stresses are summarized in Tables 6.2-11 and 6.2-12 of Reference 5. All primary stresses -for the sleeved tube assemblies are well within ASME allowable Code stresses. The minimum stress intensity margin relative to the Code allowable occurs in the sleeve. (b) Range of the Primary and Secondary Stress Intensities The maximum range stress intensity values for the sleeved tube assemblies are summarized in Tables 6.2-17a thru 6.2-17d of Reference 5. The requirements of ASME Code Paragraph NB 3222-2 are met at all' locations and required no further consideration. For the four sleeved tube configurations analyzed, Table 6.2-17e of Reference 5 provides a summary of the maximum stress intensity range with its location in the component and the available margin to the allowhole stress.

. (c) Range of Total Stress Intensities The fatigue analysis considered a design life of 35 years for the sleeved tube assemblies. A stress intensity factor of 5.0 was included for the hoop stress only. The results of the fatigue analysis are provided in Tables 6.2-18 through 6.2-21 of Reference 5. All of the cumulative usage factors are below the allowable value of 1.0 specified in the ASME Code. The tube fatigue was found to be negligible. 4.2,..A1.lowable S.leeve Degradation The licensee has provided an evaluation of the minimum sleeve wall thickness requirements to sustain normal and accident condition loads in accordance with guidelines of Regulatory Guide (R.G.) 1.121. In this evaluation, the surrounding tube is assumed to be severed between the upper and lower joints of.the sleeve. According to the R.G.1.121 guidelines, a factor of safety (FS) of 3 is required against failure by bursting under the normal operating pressure differential. On Page 6.3-5 of Reference 5, the licensee has taken exception to this position and recommends that a F_S. of 2 be used against failure by bursting. However, the approved Technical Specification reflects'the staff position requiring a FS of 3. This is acceptable. The current Millstone Unit No. 2 Technical Specification contains a pluggino limit of 40% which satisfies the FS=3 requirement stated earlier'and is acceptable for tubes. The 40% plugging limit in the Technical Specifications is for degraded tubes, but is now being proposed by the licensee for degraded sleeves. Before this can be accepted, it is necessary to establish that a 40% degraded sleeve is equivalent in strength to a 40% degraded tube. The licensee has not yet established that'the two have equivalent bending strengths. This, therefore, remains an open item. The plugging limit proposed by the licensee will be evaluated when the licensee submits the re-quested information. A leak-before-break evaluation for the sleeve based on the leak rate and burst pressure test data obtained on 11/16 in. 00 x 0.040 in, wall tubing with various amounts of uniform thinning simulated by machining the tube 0.D. was performed. The maximum permissible leak rate is 0.5 gpm at an operating pressure differential of 1365 psi. It was determined for Millstone 2 that the maximum permissible crack length in the sleeve, making an assumption that the total allowable leak rate of 0.5 gpm emanates from a single crack, is less than 0.48 inches. The critical crack length for burst is compiled using the Westinghouse normalized test data. For ~the Millstone 2 sleeve, the burst crack length based on AP of 2285 psi (Main Steam Line Break, MSLB pressure) and the maximum properties from recent test data for the Millstone sleeve samples is 0.51 inch. Since the critical crack length for burst is greater than the permissible crack size for leak, the largest permissible crack will not burst during a MSLB condition.

_9-4.3 Vibrational Effects of Sleeves The effects on the vibrational response of the tubes due to the installation of sleeves have been examined. Analyses to determina the effects of cross flow, parallel flow, excitation due to coolant pump imbalance and acoustic pressure pulses due to pump operation have been performed. Cross flow and parallel flow vibratory stress effects have been determined to be insignificant for sleeved tubes at Millstone Unit No. 2 due to their location between the tubesheet and first eggcrate. The installation of sleeves adds stiffness and mass to the tube. Therefore, the tube / sleeve frequencies would be greater than the frequencies of the unsleeved tubes due to mechanical excitation and acoustic pressure pulses. Any increase in tube / sleeve frequencies from those of the tubes would reduce dynamic stresses. Therefore, sleeving would not be detrimental to the vibratory stresses and fatigue life. 4.4 Effects of Pre-Bowing of Sleeves A slight bow to the sleeves was introduced to facilitate automatic installation. This bow is likely to introduce a small initial residual stress to the sleeve which will have no significant impact on primary stress intensity, primary plus secondary stress intensity range for racheting, or total stress intensity range for fatigue evaluation. 4.5 Effects of Tubesheet/ Support Plate Interaction _ Since the pressure is normally higher on the primary side of the tubesheet than on the secondary side, the tubesheet becomes concave upward. Under this condition, the tubes protruding from the top of the tubesheet will rotate from the vertical. This rotation depends on the boundary condition for the edges of the tubesheet. In assessing the case for the Millstone 2 tubesheet, data was taken from the Westinghouse finite element analysis of a similar geometr using a simple axisymmetric model. , These stresses. are not considered.large enough to affect fatigue usage factors. 5.0. Mechanical Testing As part of the design verification program, the licensee has conducted a full range of qualification testing which included the following. (a) Leakage resistance testing (b) Fatigue testing (c) Thennal cycle testing (d) Push-out and pull-out strength tests (d) LOCA and steam line break simulation tests. Specific test parameters and conditions, e.g., temperature, pressure and i loadings, simulated Millstone Unit 2 operating and design conditions.

, The design verification program was developed te ensure acceptable performance / leak resistance and structural integr'.ty of the sleeved tube under normal operation, accident, and transient conditions. Test samples included in the test trogram were upper joint, lower joint, 'ne fixed-fixed mock-up sample is a and fixed-fixed acck-up samples. i partial simulation of the steam generator that allows the testing of both the upper and lower joints in one test sample. ~ 5.1 Test Specimens ammuun summmma 5.2 Test Description and Acceotance Criteria Leak resistance tests

_ _ _.. _ Thermal cycling test The thermal cycling test simulated the transient temperature conditions that the joint is likely to ex erience durina the plant li fe. ] Any leakage through the joint was measured in terms of drops of water per minute at room temperature at intervals during the test. Leakage could also be detected during the test by measurement of conderst>d water vapor. Pull-out and Push-out Strenath Tests The purpose of these tests was to determine the strength of the upper and lower sleeves joint in tension and compression. A tensile strength greater than 'M and d compressive strength gredter tiio ri 'w] was considered to be acceptable. M Fatigue Tests 5.3 Test Results Leak rate test results have been provided for specimens made with sleeves. ] Comments and explanations relative to the leak rates have also been provided. The thermal cycling tests were successful in all cases

s 9 - f 4. 1,i \\ ,,w s ., y he accep ance cri eria stated ear ier, for eaK resistance, ther:aal cycling,. fatigue and axial load (push-cut, pull-out) tests envelope the. potential plant loading conditions and are, therefore, acceptable. Compliance with these acceptance criteria demonstrate the structural and leak tight integrity of the tube sleeve design. 6.0 ' leak Rate Determination In our Safety Evalaution dated March 16, 1983 relating to a Technical Specification change regarding steam generators inspections, we noted that the licensee had committed to certain actions which would result in a more accurate determination of the absolute primary to secondary leakage rate. We consider this item to be a confirmatory item and it should be implemented by the licensee within 120 days after returning to power operation. 7.0 Eddy Current Inspection N After sleeve installation, all sleeved tubes were subject to a series 1 of eddy current inspections, some of these 'iggections were intended as a process control procedure to verify. correct installation. Each tube / sleeve assembly also received a baselyne eddy current inspection to which all subsequent inservice inspections will be' compared. Eddy current inspections will be periodically carried out on the steam generator tubes in accordance with the technical specifications. = The purpose of the inspection is to detect tube degradation that may have occurred during plant operation so that corrective action can be taken to minimize further degradation 3'nd to reduce the likelihood of primary-to-secondary leakage. In the un;1eeved portion of the parent tube, conventional bobbin coil inspection techniques were used by the licensee. However, since the 3 diameter of the sleeve is smaller than that of the tube, the w 1 \\ l

fill factor of a probe inserted through the sleeve can result in e decreased capability for detecting tubing degradation.

Thus, it was necessary to inspect the unsleeved portion of the tube above the sleeve by inserting a standard size probe over the U-bend from the unsleeved le) of the tube.

The standard inspection procedure involved the use of two circum-ferentially wound bobbin coils connected in the differential mode aad excited in the multi-frequei *:y mode. For the straight length regions of the sleeve / tube assems ly, the inspection of the sleeve and tube was consistent with normal tubing inspections. In the regions where the tube / sleeve assembly joint occurs, the detection and. sizing of degradation in the vicinity of geometric discontinui-ties of the sleeved assembly is affected by interference from the genaetry. For those regions of the sleeve where the assembly has no geometric transitions, the conventional bobbin coil probe provides acceptable detection and sizing capability. For the regions of geometric transitions, the licensee has chosen a cross-wound coil '3 configuration which significantly reduces the noise signal from the s y transitions'. T'ht overall inspection procedure iavolved the use of the cross-I wound probe, which significantly reduces the interference of the N transitions,(coup)ed with a multi-frequency technique for further i reduction offthe remaining interference signals. This system reduced j the interfercoce from all discontinuities which have 360-degree , symmetry, providin i roved visibility for discrete discontinui- .u ? ties. ) Another' difficult region of the assembly to inspect is the region at the'end of the sleeve. Here, for the conventional bobbin inspec-t 1 Ja tion,'the response from the transition regions is still larger than that of the expansion regions. Thus, the signal-to-no.ise ratios for this part of the tune / sleeve assembly is about a factor of four less s itive than that of th ions. a x he cross-wound coil also significantly reduces the noise response of the sleeve end. 5 In general, lower frequencies tend to suppress sionals from transition regions relative to signals from degradation at the expense of the ability to quantify the size of the defect. Similarly, the inspection of the tube through the sleeve requires the use of low frequencies to achieve detection with an associated loss in qmantification. Therefore, a compromise between detection s and quantification must be made. i a EL M

I i 3 - I. s p The licer;see has made a commitment that his eddy current testing techniques will incorporate the most recent state of-the-art technology for inspection of the sleeved assembly and that as improver techniques are developed they.will be utilized. Therefor,e; wesfind the inspection plans to be acceptable. . 8.0 ' Al. ARA Considerations ~ i .R The Ncrtheast Nuclear Energy Company (NNECO) has taken into account ALARA considerations:ffor each of the activities toibe involved in the proposed steen generator sleeving program at Millstone' Unit No. 2. ALARA activities speci fically idirected to reduction of occupational radiation doses include . decontamination < of steam generator channel heads; special shielding to reduce exposure to spersonnel during channel head and tube sheet operations; a control work area ventilatica; system for the channel heads and other surrounding tion areas;; rerute control of the sleeving process; TV and audio surv s f in full size mock-upsh NNECO has verified that the training .in accordance with RegiflatorE fu'ide.8.27, 8'.29',. and 8.13 or equivalent. ~ In addition, NNECO,and its sleeving contractor, Westinghouse make extensive use of classrcon and mock-up training for individuals who perform the sleeving operation. All personnel assigned to the project have received ope' rational experience from recent sleeving operations at Indian Poirt Unit:3 and were trained at the Westinghouse Training Facility. Administrative control of' personnel exposures is effected by planning of maintenance procedures for the job, in order to minimize the number of 7 personnel used to perform the various tasks involving relatively high doses and dose rates. An internal channel head platforn covered by a layer of lead blanketsi is installed to reduce time in the channel head and personnel exposures. The lead blankets provide shielding from some of the r,a.diaticn coming out of the channel head bowl, and its cushioning ef fect provides ~ bet'ter traction for workers inside the channel head. Temporary shieldin' g is aliid 'us'ed't0 reduce'th'e"g'e~neral area back-ground radiation at work stations inside. containmenl. such as adjacent to the non-regenerative heat exchanger. T.V. 1during tasis 'is ~used to identify areas and acti itisurveillance of personnel i exposures and to'initiati suitable dose reducitig' actions.es involving high v j

.15 - NNEC0 has described provisions for special local ventilation associated with the steam generator sleeving program. Ventilation through the steam generator channel heads are prov.ided by a portable ventilation rig equicoed with HEPA and pre-filters. Each steam generator is ventilated through the hot leg manway for cold side work. This maintains a negative pressure in the working manway to prevent airborne radioactivity on the steam generator platform. Each steam generator is ventilated by providing suction and supply via the secondary side manways and flexible ducting. A decontamination tent is used for cleaning tools and materials removed from the steam generator. The major source of the radiation dose rate inside the steam generator head is a tenacious layer of " oxide" which includes deposited activated corrosion products. In order to remove this deposited activity from the inside of the channel head and thereby reducing dose rates in this region, NNEC0 . decontaminated. the H2 steam generator channel heads prior to tube ~leMi'g.. NNEC0 used a. tube.. honing.precessi, 'The honing of tubes s n removes the oxide film from tube surfaces in preparation for installing sleeves ar.d provides decontamination in addition to channel head surf ace cleaning. NNEC0 has made use of experience gained in prior channel head decontamination in planning for the proposed tube sleeving activities. Data were available for Point Beach (Unit 1), San Onofre (Unit 1), Turkcy Point (Unit 3) and Indian Point (Unit 3). In particular, NNECO considered informa-tion on mechanisms used in prior decontaminations, and has provided information relevant to projected occupational radiation exposures. In a letter dated August' 18,_.1983, NNEC0 provided.information that the occupational exposure incurred to date for the steam generator channel head decontamination was 103.17 person-rems. NNEC0 estimated that 19.4 person-rem will be incurred during follow up work directly associated with the decontamination project. NNEC0 had estimated 122.4 person-rems for the . decontamination. Based on field.exper'ienc'e 'from sleeving pro.iects at other plants, NNEC0 has estinated an average dose of 164.7 person-rems and 185 person-rems for sleeving each of the steam generators. The total collective dose will be 350 person-rems for the M2 sleeving project. This collective dose will include all occupational doses.jesulting.from the sleeving operation including all site and contractor support personnel. NNEC0 estimated that 907, of the dose.is received.by techn.icians.(platform and channel head

• workers ).

r

..16 - A breakdown of each task by estimated dose rates, person-hours and person-rems has been provided. Based on our review of the Millstone Report, we conclude tr.st the projected activities and estimated person-rem doses for this project appear reasonable. NMECO intends to take ALARA considerations into account, and to implement reasonable dose-reducing activities. We conclude that NNECO will be able to maintain individual occupational radiation exposures within the applicable limits of 10 CFR Part'20, and maintain doses ALARA, consistent with the guidelines of Regulatory Guide 8.8. Therefore, the proposed radiation pro-tection aspect of the sleeving program is acceptable. 90 Conclusions As a result of our review of the analytical structural evaluations performed by the licensee, the staff concludes: (1) All primary stresses for the sleeved tube assemblies are well within allowable ASME Code stresses. (2) The maximum range of stress intensities complies with the. - requirements of the ASME. Code,. Paragraph NB-3222.2, at all sleeved tube assembly locations. (3) Fatigue analysis results confirr,;cd that all the cumulative usage factors are below the allowable value of 1.0 specified in the ASME Code. (4) The licensee has proposed that the Technical Specification plugging limit.of 40% be applied to degraded sleeves. Since this plugging limit was intended for degraded tubes, it is necessary to establish that a 40% degraded sleeve is equivalent in strength to a 40% degraded tube. The licensee has not yet established that the two have equivalent bending strengths. This, therefore, remains an open item, and was. not approved. (5) Cross flow and parallel flow vibratory stress effects are insignificant for sleeved tubes at Millstone Unit 2 due to their location between the tubesheet and first eggcrate. Sleeving is not considered detrimental to the vibratory stresses and fatigue life. (6)- Thermal cycling and. fatigue testing of both upper and lower joints had no adverse effect upon the stry_ctural in.tegrity_or, leak resistance of.the joints.. Rush-out. and. pall-out strengths of both the upper and lower joints are greater than potential plant loading conditions and therefore the joint strength of mechanical sleeves will ensure tube integrity. On the basis of the above evaluation, the staff concludes that the analytical verification and niechanical testing portions of the steam

generator tube sleeving program proposed by the licensee is acceptable. The issue of allowable tube degradation will be resolved by the staff prior to the time that Regulatory Guide 1.121 will have to be implemented on Hillstone 2. Forther, we find that the s,leeving repairs can be accomplished to produce a sleeved tube of acceptable-integrity with respect to metal-lurgical propertie, corrosion resistance, leak tightness and inservice inspectability. We also find that the licensee's commitment to use state-of-the-art inspection methods and to utilize improved techniques as they are developed, in combination with stringent allowable leak rate requirements, will assure the continued integrity of the steam generator tubes. In addition, the licensee's commitments made during the preceding outage to certain actions to provide a more accurate method of determina-tion of leakage rates is a confirmatory item and should be implemented within 120 days after returing to power operation. 10.0 Technical Specification Changes The Millstone Unit 2 Technical Specifications have been changed as follows: Section 4.4.5.1.4.a - The definition of defect has been extended to include a defect in a sleeve. The definition of plugging limit has not been extended to include sleeves at this time. Section 4.4.5.1.4.b - The definition of OPERABLE has been modified -to allow sleeving as an acceptable repair. Section 4.4.5.1.5 - The reporting requirements have been modified to include sleeves. Table 4.4.6 - The table has been modified to reflect sleeving as an acceptab e repair. These changes to the Technical Specifications reflect the use of sleeving as an acceptable. repair technique and are, therefore, acceptable in accordance with the findings of Section 9.0. s

11.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR {51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment. 12.0 Conclusion We have concluded, based on the. considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public. Date: December 30, 1983 Principal Contributors: .H. Conrad, MTEB J. Rajan, MEB J. Minns RAB K. Heitner, ORB # 3 - e 0

REFERENCES-1. WECAN - Westinghouse Electric Computer Analysis, 79-IE7-NESPB-R5, Sept. 1979. (Proprietary) 2. WTO-77-038 Rev.1, "GENF: A Steady State Performance or Sizing Evaluation Code for Model F-Steam Generators", P. J. Prabhu, Aug. 1978. (Proprietary) 3. Holman, J. P., Heat Transfer, McGrcw-Hill Book Co., N. Y.,1968. 4. WECEVAL - Automated ASME Stress Evaluation, J. M. Hall, A. l.. Thurman, J. B. Truitt, WCAP-9376, Westinghouse Electric Corporation, Pittsburgh, PA. - (Not published.) 5. Proposed Technical Specifications Steam Generator Sleeving - Millstone Unit No. 2 - Northeast Utilities - June 1983. (Proprietary) m

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