ML20205E790
| ML20205E790 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/11/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20205E787 | List: |
| References | |
| TAC-57847, TAC-57848, NUDOCS 8510170467 | |
| Download: ML20205E790 (16) | |
Text
,
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UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION 7.
j WWASHINGTON. D. C. 20655 E
s,..... $
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 76 AND 69 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 i
I.
INTRODUCTION:
By letter dated May 17, 1985 supplemented by letters dated June 3 and August 2,1985, which transmitted proprietary versions of Westinghouse and Combustion Engineering reports, Northern States Power Company (NSP), the licensee, requested amendments to Facility Operating License Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant, Unit Nos. I and 2(PINGP). The existing Technic 61 Specifications allow the repair of defective steam generator tubes by plugging. The proposed amendments would allow a second method for repairing the defective steam generator (SG) tubes by tube sl:eving and would modify the reporting requirements to include tube sleeving as a repair method for tne steam generators. Technical Specification TS 4.12.D.1(f) requires repairs of tubes containing imperfections exceeding the degradation limit of 50% of nominal wall thickness.
The licensee provided supporting documents for permitting steam generator tube sleeving by three methods - Mechanical hard rolled sleeves, Brazed Sleeves, and Welded sleeves. Proprietary reports giving detailed infonnation dealing with these three sleeving methods which were reviewed by the staff are as follows-4 Mechanical hard rolled sleeves: WCAP 10756 (Januar Generator Sleeving Report (Mechanical Sleeves)" (y 1985)
" Steam WCAP-10757 -
[
nonproprietary version)
Brazed sleeves - WCAP 10815 (March 1985)
" Steam Generator Sleeving Report (Brazed Sleeves)" (WCAP 10820 - nonproprietary version)
Weld sleeves - CEN 294 P (January 1985)
" Steam Generator Tube Repair (Using Leak Tight Sleeves)" (CEN 294 NP - nonproprietary version) 8510170467 851011 PDR ADOCK 05000282 P
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The tubes to be sleeved will be selected by the licensee after a review of inspection data. Selection will be based upon the lccation of the tube in the tubesheet, defect location and size, tooling ar.cessibility and as low as is reasonably achievable (ALARA) occupational exposures. The licensee submitted infonnation associated with the Radiation Safety Program that will be used as required to assure that steam generator tube repair at Prairie Island can be performed to assure occupational exposure is ALARA.
II. DISCUSSION AND EVALUATION:
By letters dated May 17. June 3 and August 2,1985, the licensee submitted technical reports that provide support for steam generator tube sleeving at Prairie Island by means of brazing, welding and mechanical. The staff's evaluations of these methods are as follows:
A.
Mechanical hard rolled sleeves (Westinghouse WACP 10756 (proprietary) and lO/b/ (nonproprietary)
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A.1 Sleeve Design and Process Description The sleeve is designed in accordance with 1983 edition of Section III of the ASME B&PV Code through the winter 1983 addenda. [
]
At the upper end, the sleeve configuration consists of a section which
- t is [
t
] At the lower end, the sleeve configuration consists of a section which is [
t The sleeve, after installation, extends above the top of the tubesheet and spans the degraded region of the original tube.
Its length is controlled by the insertion clearance between the channel head inside surface and the primary side of tubesheet, and the tube degradation location above the tubesheet. The upper joint is located to provide a length of free sleeve above it and at a proper distance above the degraded section of the SG tube.
Tubes to be sleeved will be selected by radial location, tooling access (due to channel head geometric constraints), and eddy current indication elevations and size. An axial elevation tolerance of 1 inch will be employed to allow for any potential eddy current testing position indication inaccuracies' and degradation growth.
l l
.J
^
2
. The specific tubes to be sleeved in each steam generator will be detemined based on the following parameters:
1.
No opposite channel head side tube indications and no indications at an elevation not spanned by the sleeve pressure boundary which are greater than the plugging limit.
2.
Concurrence on the eddy current analysis of the extent and location of the degradation.
A.2. Corrosion Tests i
l The substantial data base which exists from previous test programs verifies the sleeve design and process adequacy. Much of this testing is applicable to this sleeving program. The sleeve materials to be used, thermally treated nickel 4
chrome iron alloy (Inconel 600 or 690), are identical to those used in prior i
sleeving programs.
l The basic objective of the corrosion and metallurgical evaluation programs conducted in addition to the data base was to verify that this sleeving concept i
and procedures (specifically the upper and lower joints) did not introduce any new mechanism that could result in premature tube or sleeve degradation.
Confimation that the sleeving procedures were innocuous with corrosion degradation was obtained.
A.3 Sleeve Installation and Process Verification The standardized mechanical sleeve design is the same as that used in prior sleeving programs. The fabrication of sleeve / tube joints at both ends of the sleeve was verified and applied on all programs.
Rigorous, mechanical testing programs were conducted to verify the sleeve design for various steam generator models including models which encompass the features and dimensions of the SGs at Prairie Island.
4 A.4 Process and Inservice Inspection i
After sleeve installation, all sleeved tubes are subjected to a series of eddy current inspections. Some of these inspections are part of a process control procedure to verify correct sleeve installation. However, each tube / sleeve assembly also receives an eddy current inspection for baseline purposes to which all subsequent inspections will be compared.
Conventional eddy current techniques have been modified to incorporate the most recent technology in the inspection of the sleeve / tube assembly. The resultant inspection of the sleeve / tube assembly involves the use of both a conventional bobbin coil for the straight regions of the sleeve / tube assembly i
and a cross-wound coil for the transition regions. While there is a significant j
improvement in the inspection of the portions of the assembly us.ing the
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cross-wound coil, efforts continue to advance the state-of-the-art in eddy current inspection techniques. As improved techniques are developed and verified, they will be utilized. For the present, the cross-wound coil probe
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represents an inspection technique that provides additional sensitivity and support for eddy current techniques as a viable means of assessing the tube / sleeve assembly.
. Periodic inspections to monitor sleeve wall conditions will be performed in accordance with the inspection section of the plant Technical Specifications.
This inspection will be performed with multifrequency eddy current equipment.
As part of the inspection of the sleeved tubes, there will be a series of pressure / leakage tests. These tests are intended to test the integrity of the mechanical joint against leakage at both primary and secondary pressure loadings. The tests will be conducted at the conclusion of the sleeving operation and will be performed by the licensee in accordance
- f th applicable requirements of the ASME Code and plant procedures. Periodic tressure testing of the sleeved tubes will also be perfonned in accordance with the plant Technical Specifications.
A.5 Conclusion We have reviewed and evaluated the sleeve design, process, corrosion test data and the process inspection procedure plus the inservice inspection proposed for the mechanical joint sleeves as sumarized above. Based upon our evalua' tion, we find this sleeving repair method for degraded steam generator tubes to be an acceptable repair alternative to plugging. This method for repair has r
previously been reviewed and accepted by the NRC for repair of known degraded areas of steam generator tubes. We find that the sleeving repairs can be accomplished to produce a sleeved tube of acceptable integrity with respect to inetallurgical properties, corrosion resistance, leak tightness and inservice inspectability provided that there is a minimum distance (vertically) of 1-inch between the bottom of the upper expanded section of the sleeve and the uppennost degraded area of the SG tube.
This method of sleeve repair is acceptable only for sleeves that span the tubesheet area and whose lower joint is at the primer,v fluid tubesheet face.
A sleeve that spans a degraded area of the SG tube adjacent to a tube support plate and whose lower joint is not at the primary fluid tubesheet face is not an acceptable sleeve repair.
B.
Brazed Sleeves (Westinghouse WCAP-10815 (proprietary) and 10820 (nonproprietary)
B.1 Sleeve Design and Process Description The sleeve is designed and sized identical to the mechanical sleeve described above except for the upper end. At the upper end, the sleeve configuration consists of a section which [
] At the lower end, the sleeve configuration and processing is identical to the mechanical sleeves l
described above, i
l The sleeve,'after installation, extends above the top of the tubesheet and spans the degraded region of the original tube.
Its length is controlled by the insertion clearance between the channel head inside surface and the primary side of the tubesheet, and the tube degradation location above the tubesheet. The upper joint is located to provide a length of free sleeve above it and at a proper distance above the degraded section of the SG tube. The free length is added so that, if the existing tube were to become severed just above the upper edge of the brazed joint, the tube would be restrained by the sleeve and therefore axial motion and subsequent leakage would be limited. Lateral motion would l
also be restricted, protecting adjacent tubes from impact by the severed tube.
Tubes to be sleeved will be selected in a manner identical to the method used i
in selection for the mechanical sleeves described above.
l B.2 Corrosion Tests The substantial data base which exists from previous tests verifies the adequacy of the corrosion resistance of the materials and the joint designs, both lower mechanical and upper brazed joints. Microstructural changes in the sleeve and SG tube materials were not of a magnitude to significantly degrade the corrosion resistance of these materials.
B.3 Sleeve Installation and Process Verification i
The lower mechanical joint is identical to that in the mechanical jo%t sleeves described above and evaluated as acceptable. The brazed joint design, installation i
and process verification have been performed as part of previous sleeving programs.
The heating source for the braze joint is different from that of previously l
l acceptable brazed joint sleeves.
Theutilizationof[
]heatsourcehaseliminated
. bulky equipment and time consuming complicated control system handling, resulting'in higher production efficiency and better braze process control.
B.4 Process and Inservice Inspection After sleeve installation, most sleeved tubes are subjected to a series of eddy current inspections as part of a process control procedure to verify correct sleeve installation. The lower joint is eddy current inspected in the identical manner as that used for the lower joint in mechanical joints evaluated aoove and found acceptable.
The upper joint is to be ultrasonically inspected by the use of [-
] To construct an image of the brare joint. [
]is inserted in the sleeve and positioned near the braze. A rapid axial scan of the assembly is made and the extent of the braze region defined..At this point
[
] is positioned at one end of the braze. A series of circumferential scans of the braze are then made with [
]movingaxiallybetween each pass. This process is continued until the entire brazed region is mapped.
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l l 1 Conventional eddy current techniques have been modified to incorporate the most recent technology in the inspection of the sleeve / tube assembly. The resultant inspection of the sleeve / tube assembly involves the use of both a conventional bobbin coil for the straight regions of the sleeve / tube assembly and a cross-wound coil for the transition regions. While there is a significant improvement in the inspection of portions of the assembly using j
the cross-wound coil, efforts continue to advance the state-of-the-art in the eddy current inspection techniques. As improved techniques are developed and verified, they will be utilized. For the present, the cross-wound coil probe 1
represents an inspection technique that provides additional sensitivity and i
support for eddy current techniques as a viable means of assessing the tube / sleeve i
assembly.
t l
The Prairie Island Technical Specifications require that the licensee perfonn periodic inspections of the supplemented pressure boundary. This new pressure boundary consists of the sleeve with a joint at the primary face of the tubesheet and a joint at the opposite end of the sleeve.
i The inservice inspection program will consist of the following. Each sleeved tube will be eddy current inspected on completion of installation to obtain a baseline signature to which all subsequent inspections will be compared.
Periodic inspections to monitor sleeve wall conditions will be performed in accordance with the inspection section of the plant Technical Specif ations.
This inspection will be perfonned with multifrequency eddy current epipment.
As part of the inspection of the sleeved tubes, there will be a series of i
pressure / leakage tests.
These tests are intended to test the integrity of the mechanical and brazed joint against leakage at both primary and secondary i
pressure loadings. The tests will be conducted at the conclusion of the sleeving operation and will be perfonned by the licensee in accordance with applicable requirements of the ASME Code and plant procedures. Periodic pressure testing of the sleeved tubes will also be perfonned in accordance with the plant Technical Specifications.
B.5 Conclusion We have reviewed and evaluated the sleeve design, process, corrosion test data and the process inspection procedures plus the inservice inspection proposed for the brazed joint sleeves as sunnarized above. Based upon our evaluation, 4
we find this sleeving repair method for degraded steam generator tubes to be an acceptable repair alternative to plugging. The brazed sleeve repair method has previously been reviewed and accepted by the NRC for repair of known degraded areas of steam generator tubes. This brazed sleeve procedure uses a
] source for the brazing of the upper joint rather than 3asinthepriorprocedures. This alternate heating method should increase the reliability of the brazing process. We find that the sleeving repairs can be accomplished to produce a sleeved tube of acceptable i
integrity with respect to metallurgical properties, corrosion resistance, leak i
tightness and inservice inspectability provided that there is a minimum distance t
(vertically) of 1 inch between the bottom of the upper expanded section of the sleeve and the uppennost degraded area of the SG tube.
i This method of sleeve repair is acceptable only for sleeves that span the i
tubesheet area and whose lower joint is at the primary fluid tubesheet face.
A sleeve that spans a degraded area of the SG tube adjacent to a tube support plate and whose lower joint is not at the primary fluid tubesheet face is not l
an acceptable sleeve repair.
C.
Welded Sleeves - Combustion Engineering CEN 294-P (proprietary) and 294-NP (nonproprieta ry)
C.1 Sleeve Design and Process Description Thesleeveis[
] in length and has a nominal outside diameter of i
[
- 3. Sleeve wall thickness is [
].
The sleeve is chamfered at the upper end to prevent hangup with equipnt which is used to 1
install or inspect the sleeve (or steam generator tube). [
4
]
l The outside diameter of the sleeve was selected to provide a generous clearance l
between the sleeve and steam generator tube [
]sothatthesleeve t
slides freely through the tube during installation. There were two considerations in selecting the sleeve thickness: first, the sleeve has sufficient thickness so that the steam generator tube with the sleeve bridging the corroded section i
of the tube meets the structural requirements of the undamaged steam generator tube (without benefit from the tube). Second, there is a large margin in thickness over what is required structurally to allow for sleeve eddy current measurement uncertainty. The inside diameter of the sleeve is large enough so that the flow rate and heat transfer capability of the steam generator tube are not significantly affected by sleeve installation.
Since the' sleeve is [
] long, the upper end of the sleeve is about 15 inches above the top of the tubesheet and about 31 inches below the w.
first tube support. [
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3 The weld and welding operators have been qualified for making upper and lower welds. Since the upper weld is repaired by making a second weld which is centered 2 inches below the first weld and is made using the same welding parameters, a qualification for repair is not required.
[
]
[
s s
1
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]
The tubing from which the sleeves are fabricated is procured to ASME Boiler and Pressure Vessel Code Case N-20 and is Inconel 690.
In addition, a thermal treatment of 740*C is specified in order to impart greater corrosion resistance and lower the residual stress level in the tube.
The primary selection criterion for the sleeve material was its corrosion resistance in primary and faulty secondary PWR environments. Specific resistance to pure water and caustic stress corrosion cracking was considered.
l C.2. Corrosion Tests l
C-E has conducted a number of bench and autoclave tests to evaluate the corrosion resistance of the welded sleeve joint. Of particular interest is the effect of the mechanical expansion / weld residual stresses and the condition of the weld and weld heat affected zone.
Various tests have been or are presently being conducted under accelerated conditions to assess the sleeve-tube joint perforwance under potential nominal and faulty environmental conditions.
l l
1
. l, E
3 t
C.3. Process and Inservice Inspection Three types of nondestructive examination equipment are used during the sleeving process. The eddy current testing (ET) equipment, ultrasonic testing (UT) y are as follows:
equipment and visual equipment.
A dual cross wound probe and bobbin probe using the multifrequency eddy current i
method will be used to do a baseline inspection of the installed sleeve for-future reference. The ET fixture with conduit is used on the manipulator arm to position the probe. Eddy current testing using a bobbin probe may also be
!q>
used to detemine the inside diameter of the tube to be sleeved and the sleeve expansion size.
i Ultrasonic testing using an imersion technique with demineralized water as a couolant is used to inspect the upper tube to sleeve weld. A[
]is positioned in the weld area by the elevator and is rotated with an electric motor to scan the weld. The pulse l
i echo tester has the ability to interface with an on-line data reduction l
computer to produce a displ(v/hardcopy during radial and axial scanning.
t l
An ultrasonic examination b used to confim fusion of the sleeve to the tube l
after welding. This test consists of introducing a sound wave into the welded region. This sound wave is rotated 360 degrees around the tube, the fixture is then raised approximately 0.020 inches and scanned again. A minimum of r
three scans are perfomed and if one or more of these scans shows fusion for the whole 360 degrees, the weld is considered acceptable. The beam that is i
used is capable of easily detecting a 1/32 inch diameter flat bottomed hole.
I l
An eddy current test has been qualified for the inspection of installed welded sleeves to detect flaws in the pressure boundary. Eddy currents I
circulating in the sleeve and steam generator tube are interrupted by the presence of flaws in the material with a resultant change in test coil i
impedance. This impedance change is processed and displayed on the test
{
instrument to indicate the presence of a flaw.
1 j
The pressure boundary is considered to be the sleeve up to and including the
[
upper weld joint and the steam generator tube above the weld. Consequently, j
there are the three distinct regions relative to the inspection methods:
i
- 1) the sleeve below the weld, 2) the steam generator tube behind 'the top l
section of the sleeve (above the weld), and 3) the steam generator tube above
[
the sleeve. Using specialized probes and multifrequency eddy current techniques, it has been demonstrated that a [
[
] is detectable anywhere in the sleeve or tube,
[
i 9 including the weld region. The test results are recorded on magnetic tape and strip chart recordings. Other than the probes, the inspection equipment is the same as that used for a conventional eddy current test of steam generator tubing.
Visual examinations are perfomed on the upper and lower sleeve to tube welds to detemine their integrity and acceptance. The welds are examined using a fiber optic or boroscope examination system. The lighting is supplied either as an integral part of the visual examination system or as a supplemental system. Each examination is recorded on video tape for optional later viewing and to provide a permanent record of each weld's condition.
The inspections are perfomed to ascertain the mechanical and structural condition of each weld. Critical conditions which are checked include weld width and completeness and the absence of visibly noticeable indications such as cracks, pits, blow holes, burn through, etc.
j E
i l
3 I
C.4. Conclusion i
We have reviewed and evaluated the sleeve design, process, corrosion test data and the process inspection procedures plus the inservice inspection proposed for the welded joint sleeves as summarized above. Based upon our evaluation, we find this sleeving repair method for degraded steam generator tubes to be an acceptable repair alternative to plugging, provided that at the conclusion of the sleeving program a series of pressure / leakage tests is perfomed to verify the integrity of sleeves. The tests will confirm the integrity of the welded joints at both the primary and secondary pressure loadings. The tests are to be perfomed by the licensee in accordance with applicable requirements of the ASME Code and plant procedures. Periodic pressure testing of the sleeved tubes will also be performed in accordance with the plant Technical Specifications. Plugging of defective tubes has been an accepted repair procedure. Mechanical tapered plugs and welded plugs have been used successfully.
We find that the sleeving repairs can be accomplished to produce a sleeved tube of acceptable integrity with respect to metallurgical properties, corrosion resistance, leak tightness and inservice inspectability provided that there i
i is a minimum distance (vertically) of 1 inch between the bottom of the upper expanded section of the sleeve and the uppermost degraded area of the SG tube.
I i
~
This method of sleeve repair is acceptable only for sleeves that span the tubesheet area and whose lower joint is at the primary fluid tubesheet face.
A sleeve that spans a degraded area of the SG tube adjacent to a tube support 2
plate and whose lower joint is not at the primary fluid tubesheet face is not i
an acceptable sleeve.
D.
Radiation Safety Program during tube sleeving The staff evaluated the licensee's Radiation Safety Program that will be in place during steam generator tube sleeving to assure that repair of Prairie i
Island steam generators can be performed with as low as is reasonably achievable (ALARA) occupational exposures. Our evaluation of the licensee's 4
l Radiation Safety Program addresses the following areas:
\\
l D.1 Assuring That Occupational Exposure Is ALARA l
The management cf Prairie Island has comitted to a plan for repair of thet.r steam generators that will ensure that occupational exposure will be ALARA.
The plan incorporates numerous effective strategies to reduce exposure that sleeving personnel receive from the high radiation fields inside the steam generator channel heads. These include design and improvement of remote tooling, including robotics, which will reduce channel head entries and manual work inside the channel head; the use of shielding to minimize radiation exposure by reducing general area background radiation levels; consideration of chemical decontamination of the channel heads if large scale sleeving (e.g. 200-500 sleeves per outage) is required; extensive personnel training using mock-ups; j
involvement of the health physics staff from the early planning stages;
~;
systems for tracking occupational dose for early identification of workers i
approaching their' dose limit; providing low radiation field stations to reduce exposure of workers waiting to perfom tasks; double step-off pads and double protective clothing to preclude spread of contamination; plastic enclosures l
(
and portable air ducts to contain radioactive contamination; and respiratory
. protection using supply air, for all channel head entries. A combination of i
- all these techniques is expected to be used to reduce occupational exposure to ALARA levels.
The staff finds these measures acceptable.
t l
D.2 Radiation Protection Design Features A Remotely Operated Service Am (ROSA) robotic system will be used to reduce or avoid channel head entries. A video monitoring system is used to observe all ROSA operations in the channel head. Overview cameras are used i
to monitor the sleeving operation. The ROSA is a multitask tool designed for perfoming various maintenance tasks in the channel head and'other j
radiologically hostile environments. Besides tube sleeving, it can perform eddy current inspections, plug weld repairs, tube end repairs, and mechanical plug installations.
It will therefore reduce doses and downtime for repair.
l The ROSA system is operated remotely from a control station outside containment.
However, end effectors, used to implement the various processes used in steam generator tube maintenance (e.g., tube honing, sleeves / mandrels insertion and expansion, upper and lower hard roll, eddy current inspections), must be changed by a platfom operator located near the steam generator.
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. D.3 Dose Assessment Based on relevant experience, the licensee has estimated the collective occupational exposure (person-rem) expected for installation of 100 sleeves l
by a breakdown of operations (e.g., area equipment set-up and removal, decontamination, health physics coverage), number of people involved in i
each operation and the dose rate in the area where the operation will occur.
3 The licensee estimates a total of 36 person-rem. Additionally, the 1
licensee has estimated that occupational exposure can range from 420 to 648 person-rem per unit, if the maximum possible number of tubes that could be i
sleeved in both steam generators (i.e., about 3892 tubes) were sleeved in a single operation. This range factors in Point Beach steam generator sleeving experience. These values of person-rem are considered low as compared to past experiences. This is primarily due to the recent advances in robotic technology and the low level activity that exists in the steam generator areas at Prairie Island Nuclear Generating plant. The number of people used for the operation will depend upon the number of tubes to be sleeved. They will be trained technicians who have received their training-from both vendor and licensee steam generator mock-ups.
The staff finds the manner in which the dose estimate for the sleeving operation
[
was performed and the estimate itself to be reasonable and acceptable.
i 4
i i
D.4 Radiation Protection Program i
i An ALARA check-off sheet is required, as part of the radiation protection program, for the sleeving operation. This part of the licensee's ALARA program provides 1
for evaluation of crud control, equipment accessibility (e.g., need for working platfonns; optimum accessibility for all operations to be perfonned; ease of removal of equipment; disassembly and maintenance of equipment; valves and equipment installed outside high radiation areas), installation process, etc.
i Prior to each entry into a high radiation area, stay times are calculated and, upon egress from the area, dosimeters are read-out to assure exposure control.
Daily exposure sunnaries will be reviewed by the health physics group.
Relevant experience shows that airborne releases from the proposed sleeving I
operation are minimal. Nevertheless, health physics coverage for all exposures pathways will be continuous. Closed-circuit TV cameras will be
(
used to monitor the work area. Prevailing health physics practices, presently used during each outage for eddy current inspections and nozzle dam installations, will be used during the sleeving operation.
On the basis of the licensee's health physics program and related I
experience in steam generator work, the licensee's person-rem exposure history is considered as one of the best if not the best in the industry.
The licensee's ALARA program for the sleeving operation as described in
/
theirhazardsevaluationforthesleevingrepairprocessisacceptable, j
Therefore, we conclude that the licensee s radiation protection program f
1 will provide adequate protection against the radiation hazards associated with this operation and will limit occupational exposures to ALARA levels l
in accordance with 10 CFR Part 20.
l
i f
i D.5 Environmental Significance of Occupational Exposure Northern States Power Company has estimated that the steam generator tube sleeving project for Prairie Island Units 1 & 2 will involve occupational exposures in the range of 840 to 1300 person-rems (sum for both units)
(reference 1). These doses would most likely be spread over a number of years.
Based on review of NSP's report, the Staff concludes that the estimated range j
is a reasonable estimate of the expected dose.
To determine the relative environmental significance of these estimated occupational doses, the Staff has compared the mid-point of this range (i.e.,
1070 person-rem) with the doses experienced at Prairie Island Units 1 and 2, and in the nuclear industry. Over the years 1975 through 1984, the annual collective dose at Prairie Island (i.e., sum for both Units) have ranged from 123 to 447 person-rems with an average of about 260 person-rem (reference 2 and NSP's annual exposure report for 1984). The average dose of 1070 person-rems for the planned sleeving project when averaged over the years from 1975 through 1984, would result in average annual collective dose of about 360 person-rems (or about 180 person-rems per unit) which represents about a 40% increase in the average annual collective dose of 260 person-rems. However, since the value of 180 person-rems per unit is still far less than the PWR industry 1
wide average of about 500 person rems (NUREG-0715, Volume 5), the staff i
does not regard this as a significant increase in cumulative occupation radiation exposure. Therefore, the staff concludes that the proposed steam generator sleeving project at Prairie Island meets the categorical exclusion required for the environmental assessment under 10 CFR Part 51.22c(9).
E. Proposed changes to Technical Specifications (TS)
The licensee's proposed changes to the TS reflect the use of tube sleeving or tube plugging (i.e., tube plugging was previously approved by the Comission) as a repair method for degraded steam generator tubes.
In addition, the l
proposed changes impacted the reporting requirement in that sleeving would be i
reported in the same manner as tube plugging.
Except for two modifications to j
the licensee's submittal, the staff finds the TS changes as presented in the amendment request acceptable. One administrative modification deals with j
restricting sleeving to the tubesheet region. This restriction was made part of the acceptance criteria of the TS (TS 4.12.0). A second modification deals i
with other future sleeve designs appearing in the basis of TS 4.12.
NRC staff review and approval will be required of any future sleeve designs prior to their use in the Prairie Island steam generators. The wording in the basis of TS 4.12 was changed to reflect this requirement. Both modifications to the amendment request were discussed with and agreed to by the licensee.
F.
10 CFR Part 50, Appendix K ECCS Analysis The licensee has reviewed the steam generator plugging limit that was used i
in the 10 CFR Part 50. Appendix K for the Emergency Core Cooling System (ECCS) anal j
generator) ysis. This analysis permits up to 5% (169 tubes per steam steam generator tube plugging for the accident and transient analysis.
The primary coolant flow reduction per sleeved tube is 1
- - - ~ - - - -
- ----' - -- - - -- ----~ ~ "
i
, l approximately 3.0 percent of normal flow for both mechanical and brazed sleeves. This reduction in primary coolant flow equals to an hydraulic i
equivalency of 33 sleeved tube to one plugged tube. The 5 percent tube plugging limit would not be exceeded after considering the total number of plugged tubes existing in the steam generators for both units and the maximum possible number of tubes that coulo be required for sleeving (i.e.,
1946 tubes per steam generator) by either the mechanical or braze methods.
In the case of sleeves performed by the welded method, the license analysis shows that if all tubes in the steam generator were sleeved, the f
hydraulic resistance would result in a reduction of [
] in the primary coolant flow. This reduction has no affect on the flow used in the ECCS analysis because of a 10% flow margin existing between the available flow capacity of the primary coolant system and the flow used in ECCS analysis.
The staff has judged that the existing Appendix K ECCS analysis is bounded l
for the application of the welded sleeves after considering (1) the excess -
flow capacity existing in the primary coolant system, (2) only approximately s
57% of the tube can be sleeved because of the existing physical constraints and (3) the low number of plugged tubes existing in the steam generators.
5 Based on the above, the staff concludes that the existing ECCS analysis is i
valid for all three sleeving methods provided that steam generator tube plugging limits are not exceeded by the combination of the number of plugged tubes and the hydraulic equivalency of the number of sleeved I
tubes. If the steam generator plugging limit is exceeded, the licensee is not meeting the requirement of 10 CFR Part 50, Appendix K and a modified
. analysis using a revised plugging limit must be submitted to the NRC for T review and approval.
l G.
Summary Based on our evaluation of the three sleeving methods (i.e., mechanical, i
i brazing and welding), the Radiation Safety Program existing at Prairie Island i
Nuclear Generating Plant and the TS as modified, we find that sleeving as L
described above is an acceptable method for repairing degraded steam generator tubes in the vicinity of the tubesheet.
III. ENVIRONMENTAL CONSIDERATION:
These amendments involve a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
t The staff has determined that the amendments involve no significant increase i
in the amounts, and no significant change in the types, of any effluents j
i that may be released offsite, and that there is no significant increase j
in individual or cumulative occupational radiation exposure. The Connission has previously published a proposed finding that these. amendments involve no significant hazards consideration and there has been no public l
connent on such finding. Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 651.22(c)(9).
Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental
)
assessment need be prepared in connection with the issuance of these amendments.
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IV. 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 the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date:
October 11, 1985 Principal Contributors:
D. C. Di Ianni B. Turovlin S. Block
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References:
1.
Letter from David Musolf, Manager-Nuclear Support Services, Northern States Power Company, to Director, Office of Nuclear Reactor Regulation, NRC, dated August 2, 1985 and subsequent telephonic discussions on September 19, 1985 with the NRC staff.
2.
NUREG-0713, Vol. 5, Occupational Radiation Exposure at Connercial Nuclear Power Reactors,1983, U.S.N.R.C., December 1985.
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