ML18044A494
| ML18044A494 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/31/1979 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18044A492 | List: |
| References | |
| NUDOCS 8002070338 | |
| Download: ML18044A494 (43) | |
Text
.**
'1 Remove iv, Rev 1 vi, Rev 3 x, Rev 3 4-46, Rev 0 Insert iv, Rev 4 vi, Rev 4 x, Rev 4*
4-46, Rev 4 Appendix D D-1-1, Rev 4 D-2-1, Rev 4 0-2-2, Rev 4 D-3-1, Rev 4 D-4-1, Rev 4 D-5-1, Rev 4 D-6-1, Rev 4 D-7-1, Rev 4 D-8-1, Rev 4 D-9-1, Rev 4 D-9-2, Rev 4 D-10-1, Rev 4 0-11-1, Rev 4 D-12-1, Rev 4 D-13-1, Rev 4 D-14-1, Rev 4 D-15-1, Rev 4 D-16-1, Rev 4 Appendix E E-1-1, Rev 4 E-2-1, Rev 4 E-3-1, Rev 4 E-4.1-1, Rev 4 E-4.2-1, Rev 4 Figure.E-4.2-1 Figure E-4.2-2 E-4.3-1, Rev 4 E-5.1-1, Rev 4 E-5.2-1, Rev 4 E-5.3-1, Rev 4 E-6-1, Rev 4 E-7-1, Rev 4 E-8-1, Rev 4 E-9-1, Rev 4 E-10-1, Rev 4 E-11-1, Rev 4 E-12-1,Rev 4 E-13-1, Rev 4 E-14-1, Rev 4 PALISADES PLANT SGRR (Instructions for Entering Revision 4) 8002 07 0 3.3!'
PALISADES PLANT SGRR 4.7 QUALITY ASSURANCE 4.7.l CONSUMERS POWER COMPANY QUALITY ASSURANCE PROGRAM 4-46 4-46 4.7.2 BECHTEL POWER CORPORATION QUALITY 4-46 ASSURANCE.PROGRAM 4.7.3 COMBUSTION ENGINEERING POWER SYSTEM 4-46 GROUP NUCLEAR QUALITY ASSURANCE PROGRAM 4.8 REGULATORY. GUIDE APPLICABILITY TO REPAIR 4-46 PROGRAM 4.9 SCALE MODEL OF THE PALISADES PLANT CONTAINMENT 4-55 5.0 RETURN TO SERVICE TESTING 5-1 6.0 SAFETY EVALUATIONS 6-1 6.1 FSAR EVALUATIONS 6-1 6.1.l INTRODUCTION 6-1 6.1. 2 NON-LOCA ACCIDENTS 6-1 LOSS-OF-COOLANT ACCIDENT EVALUATION 6-ld 4
6.1.3 6.1.4 CONTAINMENT PRESSURE ANALYSIS 6-lf 1
6.1.5 FSAR EVALUATION CONCLUSION 6.2 CONSTRUCTION-RELATED EVALUATIONS 6.2.l 6.2.2 HANDLING OF HEAVY OBJECTS OFFSITE RADIOACTIVE RELEASES AND DOSE ASSESSMENT 6.3 FIRE PROTECTION 6.3.l 6.3.2 6.3.3 EXISTING FIRE PROTECTION FIRE PROTECTION DURING THE REPAIR PROGRAM CONCLUSION 6-lg 6-2 6-2 6-2 6-4 6-4 6-5 6-8 iv Revision 4 December 1979
PALISADES PLANT SGRR 8.6 REPLACEMENT OF STEAM GENERATOR EVAPORATOR SECTIONS 8.7 MAN-REM CONSIDERATIONS
8.8 CONCLUSION
S 8-6 8-6 8-7 9.0 COST BENEFIT ANALYSIS FOR THE DECONTAMINATION, 9-1 STORAGE, AND.DISPOSAL OF THE OLD STEAM GENERATORS CONSIDERING ALARA
9.1 INTRODUCTION
9.2 STEAM GENERATOR IN-PLACE DECONTAMINATION 9.3 STEAM GENERATOR STORAGE AND DISPOSAL 9.3.l 9.3.2 9.3.3 9.3.4
10.0 REFERENCES
LONG-TERM STEAM GENERATOR STORAGE ONSITE IMMEDIATE SHIPMENT BY BARGE SHORT-TERM STORAGE WHILE UNITS ARE CUT UP FOR SHIPMENT (WITHOUT DECONTAMINATION).
CONCLUSIONS APPENDIX A -
RESPONSE TO NRC QUESTIONS OF 4/17/79 APPENDIX B -
RESPONSE TO NRC QUESTIONS OF 4/19/79 APPENDIX C -
RESPONSE TO NRC QUESTIONS OF 5/16/79 APPENDIX D -
RESPONSE TO NRC QUESTIONS OF 10/15/79 APPENDIX E -
RESPONSE TO NRC QUESTIONS 9-1 9--1 9-2 9-2 9-2 9-3 9-3 10-1 A-1 (a) -1*
B-l(a)-1 C-1-1 I 3 D-1-1 4
E-1-1 vi Revision 4 December 1979 2
. 't' 4.2-4 4.2-5 4.2-6 4.2-7 4.2-8 4.3-1 4.3-2 4.3-3 4.3-4 4.3-5 4.3-6 4.3-7 A-l(b)-1 A-l(b)-2 C-3-1 C-3-2 C-3-3 C-3-4 E-4.2-1 E-4.2-2 PALISADES PLANT SGRR LIST OF FIGURES PRIMARY COOLANT PIPING CUT POINTS MAIN STEAM PIPING CUT POINTS FEEDWATER PIPING CUT POINTS BLOWDOWN PIPING CUT POINTS
.e STEAM GENERATOR UPPER SUPPORT DETAILS ACCESS CONTROL (EL 590 1 -*o-11.AND EL 611'-0-").
ACCESS CONTROL (EL 649'-0")
PRIMARY COOLANT PIPING CONTACT RADIATION SURVEY AVERAGE RADIATION FIELDS 10 WEEKS AFTER SHUTDOWN RADIATION SURVEY (EL 607 1 -6 11 )
MAXIMUM DOS.E RATE INSIDE STEAM GENERATORS GENERAL RADIATION FIELD NEAR STEAM GENERATOR PIPING STEAM GENERATOR REPLACEMENT SCHEDULE STEAM GENERATOR RETUBING PRIMARY COOLANT PIPING CUT POINTS MAIN STEAM PIPING CUT POINTS FEEDWATER PIPING CUT POINTS BLOWDOWN PIPING CUT POINTS HOT SPOT HEAT TRANSFER COEFFICIENT FOR 2,060 PSIA HOT SPOT HEAT TRANSFER COEFFICIENT FOR 2,150 PSIA I 2 3
4 x
Revision 4 December 1979
PALISADES PLANT SG.
- 4. 7 QUALITY ASSURANCE The quality assurance programs for Consumers Power Company; Bechtel Power Corporation, and Combustion Engineering, Inc.
as applied to this project are described in this section.
4.7.l CONSUMERS POWER COMPANY QUALITY ASSURANCE PROGRAM The Consumers Power Company quality assurance program is described in the Consumers Power Company Quality Assurance Program Topical Report (CPC-1-A)~
This report is an
.integral part of the Consumers Power Company Quality Assurance Program Manual for Nuclear Power Plants and the provisions pertaining to major modifications will be invoked for the Palisades Steam Generator Repair Project~
4.7.2 BECHTEL POWER CORPORATION QUALITY ASSURANCE PROGRAM Bechtel Power Corporation will perform its duties in accordance with Bechtel Topical Report BQ-TOP-1, Bechtel Quality Assurance Program for Nuclear Power Plants, This 4
topical report will be. invoked for activities within 4
Bechtel's scope of responsibilities on this project.
Responsibility for the Palisades Stearn Generator Repair Project has been assigned to the Ann Arbor office.
4.7.3 COMBUSTION ENGINEERING POWER SYSTEM GROUP NUCLEAR QUALITY ASSURANCE PROGRAM The quality assurance program used by Combustion during the design and shop fabrication of the replacement steam generators is described in CE-NPD-210, Quality Assurance Program -
A Description of the CE NuGlear Stearn Supply System Quality Assurance Program.
4.8 REGULATORY GUIDE APPLICABILITY TO REPAIR. PROGRAM Section 2.1.4 discusses regulatory guide compliance during the manufacture of the steam generator units.
This section discusses regulatory guide applicability to repair program activities other than steam generator manufacture.
- a.
Regulatory Guide 1.31 (5/77), Control of Ferrite Content in Stainless Steel Weld Metal Control of stainless steel welding complies with interim position on Regulatory Guide 1.31 (Branch 4-46 Revision 4 December 1979
SGRR D-1 PALISADES PLANT SGRR D-1 Section 4.1.4 describes the various types of.accidents which might occur while handling the steam generators.
However, estimates of the dose consequences of these accidents are not presented.
Estimate the dose con-sequences of 1) accidents which might rupture the
RESPONSE
old steam generators such as a steam generator drop accident, 2) accidents which might affect safety related equipment such as a crane drop on the diesel
.generator room or a dropped steam generator rolling into buildings housing safety related equipment, and 3) accidents which might ~ause the release of radioactivity such as a crane drop onto a storage tank.
Explain the assumptions and parameters used in these estimates.
The dose consequences of various accidents which may occur while handling the steam generators are addressed below.
- 1)
An accident involving the drop of the old steam generators could be ~ssumed to rupture a closure on one of the pri-mary nozzles.
As di~cussed in SGRR Section 4~4.7, the radionuclides are immobilized in the tenacious corrosion layer, and*a significant release from a drop is not ex-pected.
For. the purpose gf analysis, 2however, a removabl~
contamination level of 10 dpm/100 cm was assumed for the steam generator internal primary surfaces.
The release of this material, approximately lO_~Ci, could not result in an offsite dose of greater than 10 mrem.
- 2)
The fuel will be removed from the reactor prior to the start of major construction activities inside containment.
Once the fuel i~ removed, equipment inside the containment b0ilding will not be necessary to serve a safety function during shutdown.
Outside the containment, it is necessary to maintain the integrity of the control room, the spent fuel pool, the spent fuel pool cooling system, component cooling system critical service water system, and* the diesel generators during shutdown.
The steam generator replace-ment route and construction activities associated with the replacement are far removed from these safety-related structures, systems, and equipment.
Administrative controls wiil be applied to keep heavy construction objects from infringing upon the safety functions of these systems, components, or structures.
Therefore,. a radiological accident of this nature is not credible, and a dose assessment is not necessary.
(_Also see E-91..
- 3)
Other radiological accidents are not envisioned except those addressed under Item 1.
D-1-1 Revision 4 December 1979
D-2
RESPONSE
SGRR D-2 PALISADES PLANT SGRR Section 4.3.6.3 identifies three potential sources of radioactive liquid wastes.
Is the secondary coolant a potential source of release of radioactive liquids?
Describe the normal liquid radwaste processing scheme.
What is the estimated quantity and radioactivity of each of these wa~tes? Explain 1) what portions of the releases are due to normal refueling and work outage and 2) what portions are added by the ste~m generator repair effort?
Describe the criteria which will be used to determine if the wastes will be processed through (1) filters, (2) evaporators and/or (3) demineralizers prior to release to the environment.
For each of the three potential sources of radioactive liquid, evaluate their environmental impact.
Th~ evaluation should include estimates of the doses individuals in the public and the population within 50 miles might receive.
Identify the equipment which will be available to process the wastes.
Evaluate the possibility of the local decontamination solution(s) having deleterious effect(s) on the equipment employed to process it.
Are the con-centrations in Table 6.2-3 typical of refueling/major work outages?
Include Fe-33 in your release estimates, and commit to sampling for Fe-55 and Ni-63 in liquid effluents during the repair program.
During the Surry repair program, Fe-55 and Ni-63 wer~.*found in effluent analyses.
The secondary coolant is not considered a potential source of release of radioacfive liquids.
Activity accumulated during the repair effort from steam generator blowdown resins and resins from full flow condensate demineralizers would not be expected to exceed 50 mCi and would be shipped as dry radioactive waste.
At present, any liquid releases associated with the steam generator repair outage are identical to the releases associated with a normal refueling or work outage.
A newly installed laundry-dry cleaning system will eliminate any additional laundry waste water associated with increased outage work schedules.
Local decontamination waste fluids. will be solidified in a cement matrix and is not considered a potential source of radioactive liquid release.
The operation of the radioactive waste treatment system is addressed in Section 11 bf the FSAR.
D-2-1 Revision 4 December 1979
SGRR D-2-2 PALISADES PLANT SGRR Total body doses to the population *and average ooses *to indivi-duals in the population from receiving water-related pathways to a distance of 50 miles from the site are calculated to be less than 1 ma11.rem and 1 mrem, respectively.
As stated above, local decontamination waste fluids will be solidified in a sodium-silicate cement and will not be treated as part of the normal liquid radwaste processing scheme.
Po-tential deleterious effects on radwaste processing equipment.
will be addressed when decontamination alternatives are addressed.
The estimated specific activity of laundry waste water as pre-sented in Table 6.2-3 is based on a refueling/major work outage which occurred at the Palisades plant in 1976 and included a full offload of the core.
Analyzing for Fe-55 and Ni-63 in liquids is intended on a semi-annual basis during the repair program.
D-2-2 Revision 4 December 1979
SGRR D-3 PALISADES PLANT SGRR D-3 Your analysis and calculation of the inventory of ac-tivated corrosion products in a steam generator pre-sented ig Table 4.4-1 is not clear.
The estimate of 3.5 x 10 inches of steam generator tube appears to
RESPONSE
be too low (from the FSAR, 8519 tubes times an av5rage tube length of roughly 650 inches equals 5.5 x 10 inches of tube).
Also, it is not clear what is meant by tubes per tube sheet, and the ratio of Co-58 to Co-60 is much higher than that ratio found at Turkey Point.
Provide information to clarify these items.
External dose rate analysis has indicated that the dose rate decreases as the inverse square of the distance from the tubesheet with dose levels at the top of the tube to be in the low mrem/hr range.
Therefore, only the tubesheet area has been used to estimate activity inventory.
The steam generator activity presented in Table 4.4-1 assumes that the activity concentrations shown are distributed throughout the tubesheet which extends to 20 55 inches above the inlet and outlet plenums.
The 3.5 x 10 inches of steam generator tubing estimate were obtain~d as follows:
(8,519 tubes/steam generator) (20.5 inch~s of tube per tubesheet) (2 penetrations 5or each tube through tubesheet) = 3.5 x 10 inches of tube The ratio of Co-58 to Co-60 was based upon the isotopic analysis of undescaled 2-inch-long sections of steam generator tubing.
The study is listed as Reference 2 in the repair report.
In addition, isotopic concentrations were verified by smear surveys and an independent gamma isotopic survey performed in 1979 using a shielded lithium drifted germanium detector.
D-3-1 Revision 4 December 1979
D-4
RESPONSE
SGRR D-4 PALISADES PLANT SGRR
~escribe any effects which your modifications to the steam generator, blowdown drain, and sample systems may have on the various accident analyses presented in your FSAR.
No values associated with modifications to the steam generator blowdown drain or sample systems will be les~ conservati.v_~ t..hci.n* __
those analyzed in the FSAR.
D-4-1 Revision 4 December 1979
D-5
RESPONSE
SGRR D-5 PALISADES PLANT SGRR Provide estimates of the Fe-55, Ni-63 and transuranic content of the "old" steam generators.
The estimates should be based on sample analyses similar to those presented in Table 4.4-1.
Samples are not available for the analysis of the radionu-cl-ides in question~
However, if Fe-55 and Ni-63 were major constituents, the two isotopes would not be significant contributors to the radiation dose.
Transuranics are not expected to differ signif-icantly from levels observed at ~urry.
D-5-1 Revision 4 December 1979
D-6 SGRR D-6 PALISADES PLANT SGRR Your analysis of gaseous releases from cutting reactor coolant piping presented in Table 6.2-1 does not appear to be correct.
The equation presented does not yield the answer given and the area of cut values of 0.5 square inches is too low.
Estimate the. releases from cutting the blowdown, feedwater, and main steam piping using similar analyses.
Also, estimate the releases from cutting th.e construction opening of the contain-ment.
Estimate the amount of gaseous radioactivity which will be released from routine refueling/work outage activities such as fuel movement.
Also, estimate the added. releases from steam generator repair activi-ties such as complete defueling, local decontamination and construction activities within containment.
RESPONSE
The area of the cut was present~d erroneously, but it was used properly to obtain the results.
The area of cut presented in the equation from Table 6.2-1 should be as follows:
(0.5 inch) (ir) (inside diameter of pipe) or
- (0.5) (ir) (30) = 47.12 sq in and (0.5) (ir) (42) = 65.97 sq in Total release for Cr-51 was estimated as fpllows:
Airborne activity near cut (µCi) =
(47.12 sq in) (0.589) (µCi/sq in) (8 cuts)
+ (65.97 sq in) (0.589 µci/sq in) (4 cuts) = 0.378 µCi No sign~f icant airborne releases are expected from cuttirig blow-down main steam or feedwater piping.
Prior to cutting the construction opening liner plate, the internal surfaces of the plate will be decontaminated to reduce zhe amount of. transferable cbntamination to below 2,200 dpm/100 cm
- Gaseous rapioactivity associated with repair outage will be simi-lar and not expected to exceed the amount of gaseous effluents associated with other routine refueling/work outages.
Gaseous effluents for the 1976 refueling/steam generator inspection outage, which included a complete offload of the core, totaled
- 1. 31 Ci.
D-6-1 Revision 4 December 1979
D-7
RESPONSE
SGRR D-7 PALISADES PLANT SGRR Explain what release path, treatment and effluent monitorin~ will be used for the air which is exhausted from the containment.
Also, explain how releases through the construction opening will be prevented or controlled and monitored.
Page 4-4 of your report states that the opening will be covered; page 4-25 states that air will be drawn in through the *Opening.
Provide clarification.
Airborne radioactivity inside containment during the steam generator repair effort will be controlled, monitored, and ulti-mately released via the plant vent stack, which utilizes a high-efficiency particulate air (HEPA) filter.
In addition to the stack filter (HEPA), the air will be conditioned, if necessary, for removal of airborne radioactivity by use of two installed recir-culation filters (HEPA plus charcoal absorber) rated at 6,000 cfm.
A temporary construction covering will be provided to restrict the infiltration of dust, sand, and water into the containment.
Air will be drawn through the construction opening and exhausted by the plant stack, preventing radionuclide release from the con-struction opening.
D-7-1 Revision 4 December 1979
D-8
RESPONSE
SGRR D-8 PALISADES PLANT SGRR Provide estimates of the doses which 1) an individual living near the site and 2) the population within 50 miles might receive due to the g~seous effluents from the repair effort.
Total body dose and significant organ doses to an individual from gaseous effluents are estimated to be less than 1 mrem.
Total body doses to the population and average doses to indi-viduals in the population from gaseous effluents within a distance of 50 miles from the site are estimated to be less than 1 manrem and 1 mrem, respectively.
D-8-1 Revision 4 December 1979
D-9
RESPONSE
SGRR D-9 PALISADES PLANT SGRR Estimate the volume and radioactivity content of
- the various types of solid radioactive waste which will be generated by the repair effort:
a) concrete b) metal shavings c) construction materials d) solidified decontamination fluids e) piping f) paper, *rags, clothing, etc.
g) spent resins, evaporator bottoms, spent filters, etc.
a)
Concrete -
The only concrete expected to be removed is at the containment construction opening.
The concrete is covered by the liner plate on the inside and should have little or no contamination.
This concrete (or part of it) will be disposed uf by conventional methods if contamination can be red~ced to atceptable levels (less than 2,200 dpm/100 cm).
Otherwise it will be disposed of as solid radwaste.
Volume is approximately 4,200 cubic feet.
b)
Metal Shavings -
The metal shavings will result from cutting piping and miscellaneous steel.
The metal shavings volume is estimated at 3 cubic feet.
Radioactive content is estimated to be less than 0. 5 cur*ie.
c)
Construction Materials -
Liner plate and other construction material is estimated at 40,000 cubic feet, containing less than 25 curies.
d)
Solidified Decontamination Fluids -
The solidified decontamination fluids are estimated to be 3,000 cubic feet, with a radioactive content of less than 4.5 curies if the pri~ary piping spool pieces are reused.
If new piping spools are used, approximately 2,000 cubic feet and less than 0.5 curie ar~ in this category.
Revis.ion 4 December 1979
SGRR D-9-2 PALISADES PLANT SGRR e)
Piping - If primary piping spool pieces are reused, it is estimated that piping will be approximately 25 cubic feet and contain less than 0.05 curie.
The use of new primary piping would result in about 200 cubic feet of solid radwaste, containing less than 4.0 curies.
f)
Paper, Rags; and Clothing -
In this category, approximately 40,000 cubic feet are expected, containing about 50 curies.
g)
Spent Resins -
Based upon previous refueling outages, spent resins are estimated to be approximately 8,650 cubic feet, with about 80 curies.
D-9-2 Revision 4 December 1979
~.
D-10
RESPONSE
SGRR D-10 PALISADES PLANT SGRR We have been informed that VEPCO experienced 214 man-rem in cutting reactor coolant piping at Surry.
Fifteen cuts were made with a motorized, track mounted plasma arc device.
You have estimated that 35 man-rem will be used to make 12 cuts at Palisades with a track mounted oxygen-acetylene torch.
The dose rates at Surry were in the 40-60 filillire~. per hour range, similar to your estimated/measured dose rates.
Reevaluate your estimate of 35 man-rem for Palisades in light of experience at Surry.
It is believed that experience gained from the Surry outage will help to reduce manhours associated with various outage-related tasks, and the above estimate is appropriate for the task as defined for the Palisades steam generator repair project.
D-10-1 Revision 4 December 1979
D-11 SGRR D-11 PALISADES PLANT SGRR Your discussion of radiation protection training on page 4-29 is not clear.
Explain who will receive what levels of training.
Explain who is included in "selected crafts."
Explain what training will be given to training personnel entering 1) containment building 2) radiation areas, and 3) high radiation areas.
Explain what training will be given to personnel who may wear respiratory protection devices.
RESPONSE
A comprehensive course in radiological protection, in addition to the required orientation course, will be available to a selected group of craft personnel.
This selection will be at the discretion of the health physics staff, but will necessarily include craft supervision or personnel responsible for directing craft work groups in radiation areas.
As a minimum, all personnel entering the containment building, radiation areas, and high radiation areas will be required to successfully complete the orientation course described in the Palisades plant health physics procedures.
Respiratory training requirements are described in the Palisades Health Physics Procedures (Section 7).
The general training program will address, as a minimum, the following:
- 1.
Discussion of the need for respirators
- 2.
Classifications of hazards that may require the use of a respirator
- 3.
General classes of respirators used for respiratory protection and a general discussion of their operation principles
- 4.
Discussion of the types of respirators used at Palisades, including proper donning-operating instructions
- 5.
The quantitative fitting program
- 6.
User's responsibility
- 7.
The evaluation program (wholebody counting, air sampling, and other bioassays)
D-11-1 Revision 4 December 1979
t...*.
D-12
RESPONSE
SGRR D-12 PALISADES PLANT SGRR Discuss your plans for keeping radiation exposures ALARA when rewelding primary coolant piping.
Evaluate the potential man-rem savings of decontaminating th~ cut-6ut sections of primary coolant piping.
Indicate whether, based on that
~Valuatio~, the tut~oLlt pip~ sections ~will be decontaminated or not.
At Surry decontamination of the primary coolant sections was successful in significantly reducing radioactive contamination.
The lower contamination levels resulted in lower doses for reinstallation of the p{ping.
Decontamination by electropolishing is considered a viable technique for reducing total manrems.
A second alternative, replacing old spool pieces, is being considered at this time.
The manrem estimate presented in Table C-1-5 (Question Set C) considers this potential manrem savings in Work Area 8C inside the reactor coolant pipe after decontamination.
It should be noted that if the old ipool pieces ace replaced with new sections, radiation fields associated with Work Area BC would.be further reduced.*
D-12-1 Revision 4 December 1979
D-13
RESPONSE
SGRR D-13 PALISADES PLANT SGRR Radioactive contamination on the outside of the steam generators could be released during storage*
or during transportation.
Estimate the amount of radioactive contamination on the outside of the steam generators, and describe any plans for removing or fixing that contamination.
Your estimate of removable activity should be based on actual measurements.
Also, describe any plans for incorporating these features into the design of the steam generator storage building to control the release of radioactivity.
a)
A sealed floor, b)
A floor sump from which water could be sampled for radioactivity (if any collected),
c)
Particulate filters over ventilation or breathing openings in walls and ceiling.
Smear surveys on outside steam generator surfaces have.shown 2 removable contamination ranging from 500 to 50,000 dpm/100 cm.
The outside of the steam generators will be decontaminated by scrubbing with detergent and water to reduce the amount of transferable contamination.
Any 2ransferable surface contamination above 2,200 dpm/100 cm will be fixed by painting.
Steam generators will be sealed to prevent the release of radioactivity during transfer and subsequent onsite storage (see Section 4.3.5.2).
The only significant radiological consideration associated with storage is the direct radiation from the steam generator (see Section 4.4.6).
Shielding will be provided to ensure an acceptable radiation level outside the storage facility (see Section 4.4.2).
The steam generator storage building will be provided with a sump to collect liquid such as that resulting from roof leakage and condensation.
The floor will be sealed if the floor per-meability is found to be a concern.
Because of efforts taken to seal and clean the steam generators, particulate filters in the storage building are not necessary.
However, a sur-veillance program will be implemented to provide assurance that there is no unanticip~ted release of radioactivity to the environment (see Section 4.4.6).
D-13-1 Revision 4 December 1979
D-14
RESPONSE
SGRR D-14 PALISADES PLANT SGRR Describe any plans to maintain a high water level inside the steam generators to provide shielding during the repair effort.
This measure was effective in reducing dose rates during operations such as pipe cutting at Surry.
Unlike Surry, a circumferential cut will not be required on the upper shell of the steam generators.
Consequently, this technique is not expected to reduce exposures significantly.
However, a high water level on the secondary side will be maintained to an extent practical.
D-14-1 Revision 4 December 1979
D-15
RESPONSE
SGRR D-15 PALISADES PLANT SGRR Evaluate the ability of your radwaste solidification system to acceptably solidify the solutions which will be used for local decontamination in the repair effort.
The current burial ground criteria for acceptance of solid radwaste call for the waste to contain no free water.
Our present solidification system using.sodium-silicate has been used and found adequate for solidification of decontamina-tion solutions.
The repair outage is not expected to generate any waste incompatible with the present system.
D-15-1 Revision 4 December 1979
D-16
RESPONSE
SGRR D-16 PALISADES PLANT SGRR Describe what precautions and post-repair tests will be utilized to ensure that the effectiveness of ventilation and safety-related air filter systems is not impaired.
Equipment tests specified in the Palisades Plant Technical Specification, Table 4.2.2, Item 11 will be conducted to ensure that the effectiveness of ventilation and safety~
related air filter systems is not impaired during the post-repair period.
D-16-1 Revision 4 December 1979
SGRR E-1 PALISADES PLANT SGRR E-1 Items C.3 and E.4 in Table 2.1-1 reflect the increase of wate~ level in the new "SG."
Modify the Technical Specifications (Item 5 in Table 2-3.1) "the low steam generator water level trip."
RESPONSE
In Table 2.1-1 of the SGRR, Item C.3 shows the increase in overall height of the steam generators, wbich is due solely to the increased height of the new main steam outlet nozzles.
Item E.4 shows that the new ~team generators will contain a greater mass of water on the secondary side at full power~
this is due to changes in the internals and thermal hydraulic (Void fraction) changes.
This serves to increase the water inventory below the "low steam generator water level trip" as specified in the technical specifications, thereby enhancing safety.
The normal water level will be approximately the same.
Therefore, this technical specification for the new steam generators need not be changed.
E-1-1 Revision 4 December 1979
SGRR E-2 PALISADES PLANT SGRR E-2 In 2.2-1-1, if the total tube outside area A is to be decreased in order to maintain (UA) = (UA), and since the tube sheet dimensions seem t8 be un8hanged, how is the "A" reduction going to be achieved?
(a) less number of tubes with the same original tube length, or (b) same number of tubes with less than the original tube length and consequently less tube height at two-90°bends?
2-1: If A is to be decreased, and (b) above is the chosen option, evaluate the net effect on natural circulation (eg, during a loss of coolant flow, a station blackout, etc)
- 2-2: If A is not to be decreased, explain qualitatively the varying effects that will change u.
RESPONSE
The replacement steam generators will have 8,182 heat transfer tubes of a 0.750-inch outside diameter (od) and a 0.042-inch average wall.
The original steam generators have 8,519 heat transfer tubes of a 0.750-inch od and a 0.048-inch average wall.
The number of heat transfer tubes foi the replacement steam generator was established, using an equivalent UA basis; i.e., (UA) replacement steam generators equal (UA) original steam generator.
This equivalent UA basis will provide the replacement steam generators with essentially the same thermal performance capability as the original steam generator.
The tube bundle geometry is arranged, using a 1-inch triangular pitch pat tern for the tube holes and is identical to 'the spacing pattern utilized on the original generators.
The equivalent UA basis, therefore, results in the replacement steam generators having fewer tubes and a smaller average tube length because the overall heat transfer coefficient u for 0.042-inch average wall tubes is greater than that for the 0.048-inch average wall tubes of the old steam generator.
2-1: This design difference will not significantly affect the natural circulation because most of the natural circulation heat transfer occurs just above the tubesheet.
2-2: This is not applicable.
See also the discussion regarding UA above.
E-2-1 Revision 4 December 1979
E-3
RESPONSE
SGRR E-3 PALISADES PLANT SGRR In 6.1.1, what are the new design features that will improve the flo~ distribution in the rie~ SG in light of the* fact that you decided not to use the flow distribution baffle?
Design features which enhance the flow distribution and cir-culation included in the design of the-replacement steam generators are eggcrate type tube supports, tube lane divider plate, and revised blowdbwn duct design.
The eggcrate type tube support system used on the replacement steam generator offers several* advantages over the drilled tube support plates used on the original generators from a hydraulic standpoint.
Because an eggcrate support is of a relatively open type construction, localized crevices adjacent to tube surfaces are minimized.
The eggcrate tube support system in the steam generator provides a minimum number of potential, localized, steam-Slanketed areas which might be present iri the annular gaps between tubes and drilled support plate holes.
The open flow area of each eggcrate support is calculated to be 69% of the total flow area.
From a different point of view, the eggcrate support obstructs only 31% of
- the available secondary flow area.
The large open flow areas avajlable in the eggcrate support system avoids the accum~lation of boiler water deposits by minimizing local flow eddies and flat surfaces which are present in other commonly used tube bundle support systems.
The tube support design on the replacement steam generator has also been designed to reduce the number of supports, both horizontal and vertical grids in the tube bend region.
In addition, the vertical tube pitch in the bend region has been increased from 1.0 to 1.75 inches, which significantly reduces flow resistance in the bend region.
The resultant effect is that the tub~ support system selected for the replacement steam generator is a de~ign which reduces the potential for localized corrosion and increases internal recirculation.
On the replacement steam generator, a divider plate is mounted in the tube lane between the hot and cold leg sides
- of the tube bundle.
This divider plate, in conjunction with the revised blowdown duct, prevents the preferential bypass of the tube bundle by recircul~ting water exiting from the downcomer along the tube lane, thereby forcing the recirculation fluid from the downcomer to be directed across the tub~
bundle.
The overall effect is to ensure relatively high radial fluid velocities throughout the region above the tubesheet.
E-3-1 Revision 4 December 1979
E-4.1
RESPONSE
SGRR E-4.1 PALISADES PLANT SGRR In Section 6.1.3 explain the effect of core inlet temperature on blowdown, reflood, and peak clad temperature.
The core inlet temperature sensitivity study showed fuel rod heat transfer during blowdown to be relatively insensitive to core inlet temperature variations.
The primary effect of increased core*inlet temperature was observed in the reflood phase of the LOCA.
With a higher core inlet temperature, slightly more energy wpuld be reieased to the containment during the LOCA.
This results in higher containment pressures and therefore increased core flooding rates due to less steam binding in the steam generators.
E-4.1-1 Revision 4 December 1979
,E-4.2
RESPONSE
PALISADES PLANT SGRR In Section 6.1.3 explain the effect of primary system pressure on blowdown, reflood, and peak clad t~mperature.
The major effect of increased primary system pressure was observed during the blowdown phase of the LOCA.
With a higher pressure (2,150 versus 2,060 psia), flow through the core and hot assembly during blowdown was degraded sufficiently to result in poorer fuel rod heat transfer during blowdown and ultimately a higher peak clad temperature.
For comparison purposes, the calculated heat transfer coefficient at the hot spot in the core is shown in Figures E-4.2-1 and E-4.2-2 for the base case and for the 2,150 psia case, respectively.
E-4.2-1 Revision 4 December 1979
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E-4.3
RESPONSE
SGRR E-4.3 PALISADES PLANT SGRR At the end of page 6.le, it is stated that "because of higher expected primary system flowrates and higher expected secondary steam pressures with the new steam generators as compared to the reference analysis."
Explain this statement in relation to the last sentence in Section 2.2.1.1 and items A.6 and B.4 of Table 2.1-1.
Section 2.2.l.l and Table 2.1-1 refer to steam generator
.. ~-
design parameters.
However, the referenced analysis in the sentence in question is the most recent LOCA analysis performed for Palisades.
This analysis assumed actual operating pressures, temperatures, and flowrates as measured in the plant with the present level of tube plugging.
In addition, for conserv~tism, approximately 500 additional tubes were assumed to be plugged.
With the.replacement steam generators installed, the primary system flowrate is expected to signifi-cantly exceed that assumed in the referenced analysis.
Further, wi.th the increased flow, high primary coolant tempera-tures, and therefore secondary steam pressures, will be permissible based on current core ~hermal margin limits E-4.3-1 Revision 4 December 1979
L r
E-5.1
RESPONSE
SGRR E-5.1 PALISADES PLANT SGRR In 6.1-4, justify the conclusion reached in the second paragraph that the LOCA mass/energy release for the replacement SG at full power would produce less containment peak pressure than that produced with the original SG.
An explicit LOC~ analysis for effects on containment peak pressure was not performed.
The new steam.generator primary volumes are slightly less than the original steam generator; therefore, there is less reactor coolant system fluid available.
Less initial inventory will produce lower peak containment pressure.
E-5.1-1 Revision 4 December 1979
E-5.2
RESPONSE
SGRR E-5.2 PALISADES PLANT SGRR Proiide the LOCA mass/energy release curves for both the original and replacement SG.
Also provide the containment pressure curve for the original SG.
Because the primary side volume of the replacement steam generators will be slightly less than the respective volume of the original units, the total LOCA mass/energy release to containment will actually be less with the replacement steam generators.
Thus, the containment pressure response for the largest LOCA (the 42-inch, double-ended guillotine rupture) will be no more adverse than that shown in Figure 14.18-2 of the FSAR.
It should be noted that the FSAR containment analysis was based on a core thermal power level of 2,650 MW t.
E-5.2-1 Revision 4 December 1979
E-5.3
RESPONSE
SGRR E-5.3 PALlSADES PLANT SGRR Briefly describe the analytical methods used for full power MSLB mass/energy release analyses and compare the. similarities and differences compared to the NRC accepted method.
The analytical methods used for a full-power main steam line break (MSLB) are the NRC-accepted methods.
The digital computer code SGN III, which has been accepted on other dockets, was used.
A particularly detailed discussion can be found in Appendix 6.B of Combustion Engineering's CESSAR-PSAR.
In addition, credit was taken for the steam outlet nozzle restrictors and main steam isolation *on high containment pressure as explained in Section 6.1.4 of the Steam Generator Repair ~eport.
E-5.3-1 Re~ision 4 December 1979
E-6
RESPONSE
SGRR E-6 PALISADES PLANT SGRR Ref er to Page 2-9 In the new steam generators, how ~ill the thermal sleeve design minimize leakage (to reduce the likelihood of steam gen/FW water hammer)?
Can this joint be welded to eliminate leakage entirely?
The feedwater liner design for the* replacement steam generator incorporates an expanded (explosively expanded) seal on the outboard piping end of the nozzle and an overlapping slip joint at the liner sparger joint.
The feedwater sparger assembly, internal to the steam generator, includes a series of top discharge elbows.
This top discharge design, as opposed to the underside discharge port design and the overall feedwater sparger design, prevents immediate draining of the feedwater ring in the event of a loss 9f feedwater flow, thereby minimizing the possibility of a water hammer.
If th~ liner to the sparger joint were to be welded, this would fix the feedwater thermal liner at both the inboard and outboard Ends.
The free movement of the liner due to thermal temperature differentials would then be prevented, a condition which could induce a fatigue failure of the liner sparger joint.
E-6-1 Revision 4 December 1979
SGRR E-7 PALISADES PLA~T SGRR E-7 Refer to Page 4-3 Will the defueling of the reactor deviate from the usual and ordinary defueling practices?
If so, describe any details that represerit a deviation.
How much time will elapse between reactqr_shutdown and off-loading of the core into the spent fuel pool.
Does this deviate from the time approved to maintain maximum
~llowable pool water temperatures?
RESPONSE
In preparation for conducting the steam generator repair progr~m, the defueling of the reactor 'is not expected to deviate from usual and ordinary defueling ~ractices. It is estimated that a period of 4 to 5 weeks will elapse between reactor shutdown and the completion of offloading of the core into the spent fuel pool.
This does not deviate from the time approved to maintain maximum allowable fuel pool water temperature~.
E-7-1 Revision 4 December 1979
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E-8
RESPONSE
SGRR E-8 PALISADES PLANT SGRR Refer to Page 4-8 In the transportation of the steam generator, will there be administrative controls in force to limit the* maximum speed of the loaded transporter?
If so, what will be the maximum speed that will be compatible with the maximum grade of 7% and the minimum turning radius of 42 feet?
During the transportation of the steam generator, administrative controls will be exercised to limit the traveling speed.
The traveling speed compatible with a 7% grade and a 42'-0" turning radius is approximately 1/4 mile per hour.
E-8-1 Revision 4 December 1979
E-9
RESPONSE
SGRR E-9 PALISADES PLANT SGRR If a storage site is necessary for the new steam generators, what assurance will there be that the access route from the storage location will not adversely expose safety-related equipment to possible damage?
The storage site for new steam generators, if necessary, will be located where there are no safety-related equipment or structures in the immediate vicinity of the storage or along the path of the access route as pr~sently cohtemplated.
The nearest Seismic Category I structure is the containment at the time of off-loading and loading steam generators through the containment construction opening.
However, the reactor will have been defueled at the time of operation.
Therefore, a new steam generator storage site and transportation will not adversely expose the safety-related equipment or structures to damage.
The electric transmission tower adjacent to the containment is not a Seismic Category I structure, but supports offsite power lines.
In the unlikely event of any damage to the transmission tower, diesel generators will be used for supplying power to the safety-related equipment until the offsite power is restored.
The capability to restore offsite power within 1 week will be available onsite.
E:-9-1 Revision 4 December 1979
E-10
RESPONSE
SGRR E-10 PALISADES PLANT SGRR Refer to Page'4-6, 4-9 Will the rigging be tested and prequalif ied before removing th~ old steam generators?
All rigging equipment outside containment will be prequalified prior to removal of the steam generators.
Rigging equipment inside containment will be tested in place using the existing steam generator as a test load~
The weight of the existing steam generator will be augmented, as required, with additional.
load to provide the required test load.
E-10-1 Revision 4 December 1979
'(I
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E-11
RESPONSE
SGRR E-11 PALISADES PLANT SGRR Is the fuel pool cooling pump supply, breaker, and its control power supply for the cooling pump monitored and alarmed in the control room?
The Palisades plant has two fuel pool cooling pumps, P-51A and P-518 *. A 480-volt, 3-phase supply to pump P-51A is taken from motor control center (MCC) 7, breaker 52-711, whereas pump P-518 is fed from breaker 52-817 on MCC 8.
Loss of the 480-volt, 3-phase supply at the MCC or at the breaker, or the loss of a 120-volt control power supply for these pumps leads to a visual alarm indication on panel COS in the control room when both the red and green indicating lights go off.
Further the auto-trip of these fuel pool cooling pumps is alarmed on window 7 of the safeguards, safety injection, and isolation annunciator on main control room panel Cl3.
E-11-1 Revision 4 December 1979
E-12 SGRR E-12 PALISADES PLANT SGRR Is the radiation monitoring instrumentation in the fuel pool area available during this period?
Provide details.
RESPONSE
Radiation monitoring instrumentation in the fuel pool area consists of two radiation area monitors and two process monitors.
These monitors will be available during all steam generator repair outage activities.
The monitors are designated as follows:
- a.
RE 2313 -
spent fuel pool area monitor
- b.
RE 5709 -
radwaste demineralizer (elevation 649 feet) area rnon i tor
- c.
RE 5711 -
radwaste addition ventilation monitor
- d.
RE 5712 -
fuel handling radwaste ventilation monitor The area radiation monitors (2313 and 5709) are coaxial ion chambers, and the process monitors (5711 and 5712) are geiger-mueller tubes.
Readout and display equipment is located in the control room.
Detector ranges and sensitivities are chosen to enable monitoring to within the requirements of 10 CFR 20.
E-12-1 Revision 4 December 1979
e*
SGRR E-13 PALISADES PLANT SGRR E-13 Section 4.7.l references the Consumers' Topical Report for "those quality assurance activities within Consumers' scope of responsibilities on this project."
This reference is not clear.
However, the Consumers Power Company QA Topical Report adequately describes quality assurance provisions that pertain to major modifications.
The second sentence of 4.7.l should be changed to:
"This report is an integral part of Consumers Power Company Quality Assurance Program Manual for Nuclear Power Plants and the provisions pertaining to major modifications will be invoked for the Palisades Steam Generator Repair Project."
RESPONSE
Section 4.7.l has been modified as suggested, and a revised Page 4-4 6. has been *issued with Rev is ion 4 to the Steam Generator Repair Report.
E-13-1 Revision 4 December 1979
II>..,;-:~...
~'**.-.v SGRR E-14 PALISADES PLANT SGRR E-14 Similarly, the reference to the Bechtel QA Topical Report in Section 4.7.2 is not clear.
The second sentence of 4.7.2 should be changed to:
"This topical report will be invoked for activities within Bechtel's scope of responsibilities on this project."
RESPONSE
- Section 4.7.2 has been modified as suggested, and a revised Page 4-4 6 has b.een issued with Rev is ion 4 to the Stearn Generator Repair Report.
E-14-1 Revision 4 December 1979