ML18065A788
| ML18065A788 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/27/1996 |
| From: | Gamberoni M NRC (Affiliation Not Assigned) |
| To: | Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TAC-M92310, NUDOCS 9606280176 | |
| Download: ML18065A788 (14) | |
Text
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e Mr. Richard W. Smedley Manager, Licensing Palisades Plant 27780 Blue Star Memorial Highway Covert, MI 49043 JUN 2 1 1900
SUBJECT:
PALISADES PLANT - REQUEST FOR ADDITIONAL INFORMATION RE:
PALISADES THERMAL ANNEALING REPORT (TAC NO. M92310)
Dear Mr. Smedley:
By letters dated October 12, November 16, December 1, December i2, and December 18, 1995, and January 12, February 2, February 5~ March 27, April 3, and April 29, 1996, you submitted sections of the Palisades Thermal Annealing Report (TAR).
In addition,. you submitted a letter dated April 3, 1996, requesting approval to use American Society of MeChanical Engineers (ASME)
Code Case N-557, "In-Place Dry Annealing of.a PWR Nuclear Reactor Vessel (Section XI, Division I)." During the staff~s *r~view of this informaticin~ we have identified areas where we require additional information.
The entlosed, questions and comments involve the areas of thermal and stress analysis, materials, fluence, equipment qualification, instrumentation, fire protection, and radiation protection and ALARA.
The staff discussed the* requests for additional information on the dose assessment perform~d for the upcoming annealing project, the use of ventilation sj"stems to minimize airborne contamination, and the shielding used to minimize personnel ~oses with the licensee during a phone conversation on April 16, 1996. *
- The analysis of the Palisades primary system for the thermal anneal is based on assumptions to be verified during the Marble Hill demonstration a~nea1. A number of the questions request comparison of the Marble.Hill annealing, demonstration to the Palisades anneal and contingencies d~pending on the results of the Marble Hill demonstration.
The staff anticipates that.
additional requests for information regarding the Palisades/TAR will be necessary after review of the results of the Marble Hill demonstration anneal is complete.
Please provide a response to the enclosed questions and comments within 60 days of the date of this letter. Please note that if an extension is requested the estimated schedule for review will also be extended.
If you have any questions regarding this request, please contact me at 415-3024.
Docket No. 50-255 Sincerely, Original Signed By:
Marsha Gamberoni, Project Manager Project Directorate 111-1 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information cc w/encl:
See next page DISTRIBUTION:
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Mr. Richard W. Smedley Consumers Power Company cc:
Mr. Thomas J. Palmisano Plant General Manager Palisades Plant 27780 Blue Star Memorial Highway Covert, Michigan 49043 Mr. Robert A~ Feneth Vice President, Nuclear Operations Palisades Plant 27780 Blue Star Memorial Highway Covert, Michigan 49043 M. I. Miller, Esquire Sidley & Austin 54th Floor One First National Plaza Chicago, Illinois 60603 Mr. Thomas A. McNish Vice President & Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire*
Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator; Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 U.S. Nuclear Regulatory Commission Resident Inspector's Office Palisades Plant 27782 Blue Star Memorial Highway Covert, Michigan 49043 Palisades Plant Drinking Water and Radiological Protection Division Michigan Department of Environmental Quality 3423 N. Martin Luther King Jr Blvd P. 0. Box 30630 CPH Mailroom Lansing, Michigan 48909-8130 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, N. W.
Washington DC 20037 Michigan Department of Attorney General Special Litigation Division 630 Law Building P.O. Box 30212 Lansing, Michigan 48909 Mr. Dennis Harrison U.S. Department of Energy NE 451 Washington, DC 20585
PALISADES PLANT DOCKET NO. 50-255 REQUEST FOR ADDITIONAL INFORMATION REGARDING PRELIMINARY THERMAL ANNEALING REPORT {TAR)
Thermal and Stress Analysis {Sections 1.3, 1.4, 1.5, 1.7, 1.8, 2.1, 2.2, and 2.3) and ASME Code Case N-577
- 1.
By letter dated April 3, 1996, Consumers Power requested approval to use American Society of Mechanical Engineer {ASHE) Code Case N-557, "In-Place Dry Annealing of a PWR Nuclear Reactor Vessel {Section XI, Division l)." Figure 1 of Code Case N-557 contains stress categories and allowable stress limits for the thermal anneal. These stress categories and allowable limits are similar to tho~e contained in Section III Subsection NB for design.
The allowable limit for secondary stresses is the same 3Sm limit contained in Section III. However,Section III requires the 3S~ limit to be met for the range of secondary stresses.
Code Case N-557 is silent in this regard.
Since the anneal will involve a heatup followed by a cooldown, the staff believes that it is appropriate to consider the full range of stresses in the evaluation of secondary stresses. Address the application of Code Case N-557 with regard to the full range of secondary stress produced by the thermal anneal.
- 2.
Section 1.3.D.6 of the TAR indicates that the cold leg pipes are cold sprung to reduce system stresses resulting from* vertical thermal growth of the reactor vessel {RV) nozzles. Describe how this cold spring was accounted for in the three-dimensional {3-D) elastic-plastic finite element model described in Section 1.7 of the TAR.
- 3.
Section 1.5 of the TAR contains a description of the instrumentation to be used during the anneal.
Provide a comparison of the instrumentation to be used during the Palisades anneal with the instrumentation used at the Marble Hill demonstration anneal.
Describe the differences between Palisades plant and the Marble Hill plant that would affect the measurements at comparable instrument locations. Also describe any differences between the analytical models for Palisades and Marble Hill that would affect the measurement comparisons at comparable instrument locations.
Provide a comparison between the predicted and measured results at the Marble Hill demonstration anneal.
Discuss the implication of the Marble Hill comparison with respect to the adequacy of the Palisades analytical results.
- 4.
Section 1.5.C.1.3 of the TAR discusses the measurement bias uncertainty for the vessel internal temperature sensor probes.
The report indicates that correction factors developed for the probes will be verified at Marble Hill. Describe the method that will be used at Marble Hill to verify the correction factors used for Palisades.
- 5.
Section 1.7.D.1 of the TAR discusses the results of thermal and stress analyses using a two-dimensional {2-D) finite element model.
Clearly identify how these results were used to demonstrate compliance with ASME Code Case N-557.
- 6.
Section 1.7.D.3 of the TAR contains a discussion of the evaluation of
.the surveillance capsule holder assemblies. Describe the evaluation to detennine the acceptability of the welds to the RV wall and the fillet welds between the box beam and the bracket. Provide the acceptance criteria used for these evaluations. Describe any measures that will be taken during the anneal to limit the distortion of the surveillance capsule holder assemblies.
- 7.
- Section 1.7.D.4 of the TAR contains a description of the evaluation results for the 3-0 finite element model.
The discussion of analysis case T4 contains an eva 1 uat ion of the strain to determine acceptability for Code Case N-557.
Provide the basis in* Code Case N-557 that allows for this evaluation procedure.
- 8.
Section 1.7.D.4 of the TAR contains a discussion of the results of the evaluation of the RV flow skirt for analysis case T3.
Figure 3-32 of Appendix 1.7.B contains a comparison of the stress-strain relationship used in the analysis for the SB-168 material with test data. The comparison indicates a significant difference between the stress-strain relationship assumed iii the analysis and the stress-stain relationship.
obtained from actual test data.
Describe the potential impact of the test data on the strain calculated in the analysis. Also describe the design criteria that is applicable to the flow skirt.
- 9.
Appendix 1.7.B of the TAR contains a description of the 3-D finite element model used for the thermal and stress analysis of the reactor and primary coolant system for the Palisades anneal.
The stress analysis used elastic-plastic calculational methods to calculate stresses. The TAR indicates that the vessel base material did not experience plastic deformation. Since the criterion used in ASME Code Case N-557 is based on elastic analysis, describe the rationale for using the elastic-plastic calculational method to compute stresses.
Provide a list of all elements in the model that exceeded the yield limit and describe the impact of the yielding on the results for the overall model that would* have been obtained with an elastic analysis.
Thermal Annealing Operation (Section 1.5)
- 1.
Provide analysis documentation to support the claim that the loss of any one sensor location can be adequately compensated for by interpolation between two or more adj.acent sensors. Identify the location and effect (i.e.~ on temperature control and/or critical data ac~umulation) of the sensor position for which the greatest level of uncertainty would result from the interpolation procedure during each of the five annealing phases as defined in subsection 1.5.D.1.
- 2.
Confirm that the instrumentation of the external temperature measurement sensor array described in subsection 1.5.B.2 conforms to the single location failure stipulation as presented in question #1.
- 3.
Section 1.5.D, subsection 8, notes that all potential liquid flow paths to the reactor vessel will be either drained or tagged out of service.
Specify how each of the fl ow paths will be i s.o lated and confirmed to prevent a water intrusion during the annealing process.
Materials/Fluence (Section 1.1)
Section 1.1.E discusses the surveillance program (also discussed in Sections 1.2 and 3.0). With regard to the recent revisions to the reactor pressure vessel (RPY} fluence estimates, what are the impacts on the thermal annealing submittal:
- 1.
The revised fluence estimates were performed differently than the fluences estimated for the specimens that provide the bases for the Regulatory Guide (RG) 1.99, Rev. 2 embrittlement trend curves. Since Consumers Power's embrittlement estimate was projected using the chemistry factors in Table 1 of the pressurized thermal shock (PTS}
rule, you should estimate the effect of using your neutron fluence methodology on the chemistry factor of its limiting RPV beltline material.
- a} The neutron fl uence for each capsule has been reduced as a res*u1 t of the change in neutron fluence methodology.
How will the changes in the neutron fluence methodology impact the neutron fluence of a typical (generic} pressurized-water reactor (PWR} RPV wall capsule?* Identify the changes in the neutron fluence methodology that are applicable only to Palisades.
How much do each of these factors affect the change in neutron fluence for the capsules? What is the impact of changing the neutron fluence methodology on the amount of embrittlement to the Palisades RPV?
b) Compare the chemistry factors from the surveillance capsules materials to the chemistry factors in the PTS rule.
Does this comparison indicate that the chemistry factors in the PTS rule are applicable to the Palisades RPV?
- 2.
Sections 1.1.E.2.3 and 1.1.E.3.4 discussed the standard reference materi a 1 ( SRM} in the Pali sades surveil 1 ance capsules. Considering the revised fluence evaluation, do the SRM results still indicate no anomalies in the Palisades operating conditions?
- 3.
Section 1.1.E.l.l discusses weld repairs made to the surveillance weld.
What were the locations and extent of these repairs.
Were mechanical test specimens removed from these areas?
- 4.
Table 1.1.E-5 presents Charpy Y-notch results for the Palisades surveillance prrigram heat-affected zone (HAZ}.
Upper-shelf energy [USE]
decreases of 37 and 35 are listed. Are these absolute decreased values or deltas? If absolute, since these values were so low, were the fracture surfaces examined for failure mode?
Materials/Fluence (continued) (Section 1.2)
- 5.
In Section 1.2.C.l.l it is stated that there are no records of weld repairs to the beltline plates. Does this include o*nly Combustion Engineering records or plate fabricator (Lukens, U.S.* Steel, etc.)
records also? It would seem unusual for such large Q&T plates to ha.ve no surface weld repairs.Section I..2.C.2.3 acknowledges the possibility of plate weld repairs that may not have been documented.
This section also notes that plate repairs in the upper shell course were only l~ter discovered during vessel inlet nozzle weld inspections.
- 6.
- Section 1.2.C.2.l discusses three separate stress relief heat treatuents that were performed during the RPV fabrication sequence.
The temperatures for these treatments ranged from 1100 °F to 1175 °F with a total cumulative time of 16 to 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> at temperature.
Was accelerated cooling from the SRHT [stress-relieved heat treatments] used to preclude potential metallurgical degradation associated with impurity segreg~tion (e.g., temper embrittlement)?
- 7.
Section 1.2.C.3 summarizes the fluence analysis for the RPV.
The.
reactor vessel neutron fluence analysis has been revised per the submittal to NRC dated April 4, 1996. Assess the impact of the revised fluence analysis on the surveillance results per the previous questions on Section 1.1.
- 8.
Section 1.2.E.2 discusses annealing effects on internal attachmentse Are there any effects of the anneal on the bottom-mounted instrument penetrations in the lower head?
- 9.
Sections 1.2.E.2.l and 1.2.E.2.2 discuss annealing effects on the Core Support Lugs, Core Stabilizing Lugs and the Flow Skirt. The potential for sensitization of the IN-600 material is acknowledged.
It is stated that the sensitized materi*al is not a concern in the normal PWR environment.
Is the sensitization potentially a problem when coupled with enhancement from the irradiation environment?
Materials (continued) (Section 1.3)
- 10.
Comment - As a general observation, much of this section relies on the predictions from the thermal models as t.o the maximum temperatures that
- will be achieved in various components.
The predictive capabilities of these types of models will be evaluated in the Marble Hill anneal. If the Marble Hill measurements show the models to be deficient, the conclusions of this section will obviously have to be re-evaluated.
- 11.
Sections 1.3.8.4 and 1.3.8.4 discuss heating effects on the RV supports.
The assumption is that the supports will not be heated beyond their design temperature of 650 °F.
Section 1.3.8.4 acknowledges that the lubricant will degrade at temperatures above 650 °F.
There is also the possibility for metallurgical degradation of the bronze alloy at temperatures above 650 °F.
If the prediction of 650 °F is not verified, degradation of the support assembly will have to be addressed.
- 12.
In Sections 1.3.A and 1.3.B, insulation properties for various components are addressed. For the nonreflective types of insulation, does a possibility for generation of potentially toxic gases from the insulation exist at the elevated temperatures of the cavity? Can such gases be generated from other materials in. the cavity due to the elevated temperatures?
- 13.
The concept of "sacrificial" concrete is presented in Sections 1.3.C.2 and 1.3.C.7. Section 1.3.C.7 references the Bechtel Palisades Plant design criteria when discussing the "sacrificial" concrete concept.
Provide the Bechtel reference discussing "sacrificial concrete.
What is the basis for the "sacrificial" thickness being set at 10 inches?
- 14.
Section 1.3.C.5 discusses the enhanced cooling capabilities of the Bio-Shield Cooling System (SCS) in regions adjacent to the RPV supports.
However, the Bio-Shield thermal model (described in Section 1.3.C.9) does not appear to account for thermal conduction down the steel embedments for the RPV supports.
Why was this not included in the model?
What are the expected temperatures from conduction down the embedments?
Can the SCS maintain adequate cooling in these regions during the anneal?
Materials (continued) (Section 1.4)
- 15.
Section 1.4.B proposes 800 °F and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> as the minimum temperature and time for annealing.
Given the actual annealing response for the Palisades materials in the submittal dated April 2, 1996, has there been any thought given to revising the minimum conditions?
- 16.
Section 1.4.B discusses temper embrittlement (TE) but does not discuss the Palisades plans for examination of TE in HAZ materials.
The section should be revised to include the plans.
Materials (continued) (Section 1.5)
- 17.
Section I.SB notes that strains will not be measured at Palisades but will be inferrad based on temperature measurements and the stress analysis. This will be highly dependent on how well the stress analysis for Marble Hill is benchmarked by the experimental measurements made during the anneal.
What is the current projection for when the Marble Hi 11 annealing information will be submitted to NRC?
Materials (continued) (Section 1.8)
- 18.
Sections 1.8.B and 1.8.C dis~uss time and temperature limits for the anneal.
Time and temperature limitations on the anneal could also possibly result if TE is found to be an issue with the Palisades HAZ materials. These sections should acknowledge the possibility of limitations associated with TE.
Materials*Ccontinuedl (Section 2.2)
- 19.
Section 1.7.0.3 discussed the possibility of high stresses in the surveillance capsule holder assemblies during the annealing heatup.
There exists a possibility that these stresses could cause cracking in the fillet welds joining the wall attachment bracket to the box beam.
Table 2.2.A-l requires a VT-1 inspection of these areas.
VT-1 may not be sufficient to find the cracking if it is tight.
An "enhanced" VT-I procedure capable of resolving a 1/2 mil diameter wire should be performed on these areas.
- 20.
Section 2.2.A.3 discusses the possibility of "jacking" loop components during the anneal.
What criteria will be used to determine if jacking is necessary? Will the evaluation be performed on-line during the heatup, or in advance of the anneal?
Materials (continued) (Section 3.0)
The staff issued a request for additional information on this section previously (November 1995) and received a response in December 1995.
The following are follow-up questions based on information obtained since that time:
- 21.
If intergranular fracture is observed in irradiated and annealed HAZ material, the licensee indicates that its surveillance program will only include weld and plate material but will not include HAZ material.
Provide the basis for concluding that HAZ material need not be Charpy impact tested to determine the effect of intergranular fracture on irradiated and annealed HAZ material. If additional testing of Palisades HAZ materials is indicated, can the testing be completed to allow time for NRC review before the anneal (May 1998).
- 22.
What is the neutron fluence received by the HAZ materials that are being fractured?
How much phosphorus is in the HAZ materials? Will the neutron fluence and phosphorus of the HAZ material that are being fractored represent the neutron fluence and phosphorus of the HAZ materials in the Palisades RPV?
Explain.
Eguipment Qualification (Sections 1.3 and 1~6)
- 1.
Describe in detail those measures that will be taken to assure that safety-related components (including electric cables) will not be adversely affected by temperature and radiation effects during and following the annealing process (recognizing that there will be some degradation of the RV insulation materials). Also, discuss any provisions that will be implemented for independent oversight of the evolution as it relates to surrounding plant equipment and also address the following specific questions pertaining to Sections 1.3 and 1.6 of the Preliminary TAR.
- 2.
How will temperature vs. time be established? What about electrical wiring associated with the detectors? (1.3.E.l)
- 3.
What effects will radiation have on electrical cables? What assurance is there that the actual positioning of cables will be adequate to avoid adverse effects? (1.3.E.4)
- 4.
How will the temperature profile outside the ducts be established?
(1.3.E.6)
- 5.
What about effects of radiation levels/exposure of pl ant equi_pment?
(l.3.F.5/1.3.F.6)
- 6.
Through what areas of the plant and in the vicinity of what plant equipment does the annealing system ductwork pass? What measures will be taken to assure that the heat exchanger ducting will remain intact and adequately supported and will not impact safety-related equipment throughout the course of the annealing operation? Should a failure of the ducting occur, identify any plant structures and components that could be affected by the failure, including the effects of any fire that might result, such that the consequences are made worse (e.g., create a challenge for the spent fuel pool cooling system, result in an evacuation of the control room, cause a radiological release, etc.) and identify any special measures that will be taken to eliminate or minimize the vulnerabilities that exist. Provide a summary of the analysis for the most safety significant postulated ductwork rupture event. (1.6.A.l.2)
- 7.
How often will the "job walks" be performed and what disciplines will be involved? Also, will there be an initial walkdown of the installation and if so, what will it entail and what disciplines will be involved?
Discuss any training that will be provided, procedures that will be used, documentation that will be required, and how discrepancies and
- problems will be highlighted and resolved.
(l.6.A.1.9.2)
Instrumentation
- 1.
In its letter dated January 12, 1996, Consumers Power Company stated in*
Section 1.3.E.l that the source range, wide, and power range detectors can withstand operating temperatures up to 300 °F.
The temperature limits for the detectors will not be exceeded during annealing,*and Consumers Power prefers to keep the instrument in place if the detector temperature can be maintained below the specified maximum.
Additionally, in Consumers Power's letter dated February 5, 1996, it is stated that the target inner diameter annealing temperature will be 850 °F to 900 °F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> and the maximum annealing temperature will not exceed 940 °F.
Clarification is needed to determine how detector temperature will be controlled to remain below the instruments' temperature limits/capabilities.
Fire Protection
- l.
Specify the editions of the National Fire Protection Association (NFPA)
Codes and Standards (NFPA 54, 58, 70, 79, etc.) referenced in Section 1.6 of the preliminary TAR, that the installation and operation of the gas delivery system, control train, and support equipment will comply with.
Identify any deviations from the referenced codes and standards.
- 2.
Specify the applicable safety requirements from the American Gas Association Code that the gas control train is in compliance with.
- 3.
Specify the maximum quantity of liquefied petroleum gas (LPG) that will be permitted on site at one time. Also provide specific details regarding the location and orientation of the docking area and LPG tankers relative to safety-related structures, including the actual separation distances and elevation differences.
- 4.
Specify the fire-suppression equipment that will be provided for the docking area, gas delivery system, and control train. Describe the access routes available for mobile fire apparatus to respond to a fire incident involving the docking area considering potential environmental conditions.
- 5.
Specify the administrative controls on smoking in the vicinity of the docking area, gas delivery system, and control trains that will be in place when LPG is on site.
- 6.
Specify the listing or approval that has been obtained for the components of the gas delivery system, and control train.
Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9)
- 1.
Section 1.3.F.2 of the Thermal Annealing Operating Plan states that the "dose rate target for the shielding was established as a calculated dose rate of 500 mR/hr or less in the refueling cavity work areas and 60 mR/hr or less ~t the 649 foot elevation."
- 1) Discuss why you did not design the shielding for the reactor internals to obtain a lower calculated dose rate in. the refueling cavity work areas.
- 2)
Discuss what major work evolutions will be conducted in the refueling cavity work areas following draindown of the water in the refueling cavity prior to annealing. Provide the estimated person-rem doses associated with each of these work evolutions.
- 3)
Provide layout maps of the major plant areas where annealing operation activities will take place.
Include on these layout maps the estimated dose rate zones that will be present during the annealing operation (i.e., once the reactor internals have been stored in the refueling cavity and the water in the refueling cavity has been drained down).
- 2.
Section 1.3.F.5 discusses the process of removing the upper guide structure (UGS) and core support barrel {CSB) from the RV and placing these components on storage pads in the west end of the refueling cavity.
Provide plant layout drawings (including sectional cut
- , drawings) showing the proposed storage location of these components in the refueling cavity.
Include on these drawings a depiction of the shielding structure and horizontal top cap that will be constructed around the core internals in the refueling cavity.
- 3.
Sections 1.3.F.5 and 1.3.F.6 discuss the use of temporary shielding on the refueling deck to minimize personnel exposures during the insertion and withdrawal of the UGS from the CSB.
Describe where this temporary shielding will be erected and who it will be used to provide protection for.
- 4.
Section 1.5.D states that there will be provisions to control radioactive contamination during the annealing operation.
In addition to the provisions to control airbofne contamination currently in use at Pali~ades, describe any additional measures that will be takeri to control airborne contamination created by the annealing process.
In particular, discuss measures to control airborne contamination (which may result from dryout of the core internals) inside the constructed shield structure. Also discuss the ventilation system that will be installed to filter and exhaust the contaminated air in the RV and reactor vessel annulus areas during the drying and annealing process.
- 5.
Section 1.6.A.5.l describes the RV top cover (RVTC) that will be installed over the dry RV to support the heat exchanger assembly, provide shielding, and control airborne contamination.
- 1) It is unclear from the description in this section whether or not the RVTC is an integral part of the heat exchanger.
If the RVTC is not an integral part of the heat exchanger, describe the procedure for lowering the RVTC over the heat exchanger ducting following insertion of the heat exchanger into the RV.
- 2) Describe how the two 6-foot semicircular cylindrical shielding segments will be installed around the heat exchanger ducting on top of the RVTC (will this be performed remotely?). Discuss the potential for streaming of radiation up through the heat exchanger ducting and out the top of these cylindrical shielding segments and how you plan to protect plant personnel from this potential source of radiation.
- 3) Provide an estimate of the dose rate on top of the RVTC once it has been installed on top of the reactor vessel. Describe any work functions that will be performed in this area in preparation for the annealing operation.
- 6.
Section 1.9.A.3 states that the projected dose for the annealing project is approximately 200 person-rem, excluding other outage scope tasks and support.
- 1) Provide a job dose breakdown for each of the tasks associated with the annealing project and provide the most recent total job dose estimate for the annealing project.
- 2) Discuss how you plan to integrate the annealing project with the ether planned outage scope tasks so that the other scheduled tasks will not interfere with the annealing project. Discuss your plans to limit personnel entries into the containment once the reactor internals have been removed from the RV and the water in the refueling cavity has been drained down.
- 7.
Provide a copy of the anneal.ing as low as is reasonably achievable (ALARA) plan mentioned in Section 1.9.A.3 of the Thermal Annealing Operating p*j an.
- 8.
Section 1.9.C.4 states that the experience gained from the installation of thermocouples in the reactor cavity annulus during the 1995 refueling outage at Palisades was beneficial in minimizing personnel dose.
Since a majority of the estimated dose from the annealing project will come from installation of instrumentation (in the reactor cavity annulus, primary coolant system piping loop areas, and other high dose rate areas), you should expand Section 1.9.C.4 of the Thermal Annealing Operating Plan to describe some of the dose reduction features and practices used to minimize personnel dose during instrumentation installation.
- 9.
Section 1.5.C.3 describes the use of sensors for external displacement measurements.
Therefore, the first word in Section 1.5.C.3.4 should be changed from "temperature" to "external displacement."
- 10.
Section 1.9 of the Thermal Annealing Operating Plan is entitled "ALARA Considerations" and it discusses some of the methodologies and procedures that Consumers Power Company will use to ensure that occupational doses will be maintained ALARA during the annealing process. Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable" contains guidelines (such as equipment selection and design, proper use of shielding, ALARA training, and the establishment of a good radiation protection program) on how to maintain radiation doses ALARA.
In Section 1.9 of the operating plan~
verify that you utilized the guidance contained in Regulatory Guide 8.8 to ensure that radiation doses will be maintained ALARA during the
- annealing project.
Radiation Protection (Section 1.10)
- 1.
Section 1.10.E of the Thermal Annealing Operating Plan provides a summary of the planned sequence for the annealing operation. This summary states that the UGS will be air lifted and placed into the CSB following the construction of the temporary shielding around the CSB.
However, following the annealing operation, the summary states that the temporary shielding around the reactor internals will be removed prior to removal of the UGS from the CSB.
Since it would appear that dose rates in the vicinity of the temporary shielding would be lower once the UGS was removed from the CSB, justify why the temporary shielding will be dismantled prior to removal of the UGS from the CSB.
11 -
- 2.
Section I.IO.I states that industry experience gained from the annealing demonstration project at Marble Hill will be applied to Palisades. Discuss any lessons learned from the Marble Hill annealing project that will result in estimated dose savings during the upcoming Palisades annealing project.
TI
- o,
.e DISTRIBUTION FOR PALISADES PLANT RAI DATED:
JUN 2 7 1996