ML20044H074
| ML20044H074 | |
| Person / Time | |
|---|---|
| Issue date: | 04/13/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044C660 | List: |
| References | |
| WCAP-12932, NUDOCS 9306070352 | |
| Download: ML20044H074 (14) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FULL RCS CHEMICAL DECONTAMINATION PROGRAM REVISION 1 WCAP-12932 WESTINGHOUSE OWNERS GROUP
1.0 INTRODUCTION
In 1988, the Electric Power Research Institute (EPRI) and 10 U.S. pressurized water reactor (PWR) utilities, as represented by the Westinghouse Owners Group (WOG), initiated a program to determine the feasibility of conducting a full reactor coolant system (RCS) chemical decontamination.
Corrosion and wear products found in the RCS circulate with the primary coolant through the reactor, and some of these products become activated. Oxide layers (crud) that form throughout the RCS incorporate these activated products, which eventually leads to personnel radiation exposure during maintenance of major plant components. The amount of radiation in the crud increases with time from initial start-up until it levels off after 4 to 6 cycles of plant operations. One way to reduce the radiation exposure to personnel is to reduce the source of radiation by chemical decontamination of the full RCS.
The EPRI and utilities program assessed the technical acceptability of two processes, CAN-DEREM and LOMI, for full RCS chemical decontamination. The program was divided into three major phases: Phase 1-Initial Parametric Studies; Phase 2-Decontamination Process Qualification and D6 tailed Engineering Evaluations; and Phase 3-Detailed Design and Implementation.
Phase 1 included an evaluation of the major issues involved for full RCS chemical decontamination using the CAN-DEREM and LOMI processes.
This phase also included consideration of the ramifications of doing the decontamination with the fuel in and the fuel out. The conclusion was to do the first decontamination with the fuel out. The results of this phase were reported in WCAP-12110 issued in December 1988.
Phase 2 entailed detailed engineering and testing evaluations for the CAN-DEREM and LOMI processes for a generic Westinghouse nuclear reactor full RCS.
This phase was divided into seven tasks. Task 1, Process Qualification Test Program, was a comprehensive test program for corrosion, friction, and wear effects on materials and components in the RCS for the CAN-DEREM and LOMI processes. Task 2, Fluid Systems Evaluation of Decontamination Process Integration with RCS and Auxiliary Systems, was a systems engineering task that integrated the decontamination process with the Westinghouse RCS and Auxiliary systems. Task 3, Engineering Evaluation of RCS Components and Systems, examined the effects of the chtmical decontamination on the physical condition and operability of components, equipment, and mechanical systems in the primary and auxiliary systems in a typical four-loop Westinghouse PWR.
The results would also apply to typical two-loop and three-loop Westinghouse PWR's. Task 4, Waste Management Methodology and Waste Characteristics, examined estimated curie and waste volumes, identified waste management technologies available, and defined the technical specifications for field impicmentation.
Task 5, Evaluation of Long-Term Benefit of Full RCS Decontamination, estimated the rate of radiation buildup after a full RCS decontamination. Task 6, Preparation of Topical Report and Generic Safety 9306070352 930413 PDR TOPRP ENVWEST C
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Evaluation, involved the preparation of the topical report containing a generic safety evaluation that was submitted to the NRC in January 1992 for review.
A generic 10 CFR 50.59 program safety evaluation was included in the topical report.
Task 7, Full RCS Decontamination Project Conceptual Design, defines functional specifications, methodology, and application data.
The results of these tasks are reported in WCAP-12820, dated February 1991.
The Phase 2 results showed that there were no significant detrimental effects on the primary materials and components in the RCS caused by the decontamination chemicals.
Corrosion rates for most of the materials of construction were less than 1 mil with no evidence of stress corrosion cracking.
There was an effect on chrome-plated surfaces, AISI Type 410 stainless steel, Stellite and A533 steel.
There was some flaking of chrome-plating on stainless steels and shallow interfacial attack under the chrome-plating on 410 stainless steel. The Stellite showed surface roughening after a decontamination cycle. The A533 suffered 53 mils of general corrosion after 3 cycles of CAN-DEREM and 14 mils of general corrosion after 3 cycles of LOMI.
The WOG suggested that special inspections be conducted on chrome-plated surfaces, AISI Type 410 stainless steel, Stellite and A533 steel components after several decontamination cycles.
Also, the WOG recommended that the control rod drive (CRD) shafts be removed during decontamination.
The concern with the CRD shafts was that they would be subject to fatigue damage from flow induced vibration with the fuel out. Furthermore, there was a concern with the control rod drive mechanisms (CRDM solutions could cause corrosion, wear,) flaking of chrome-plating on matingthat c surfaces, and surface roughing could affect the functionality of the CRDM's.
Testing at V.C. Summer showed that contact with the CRDM's by decontamination solutions would have little effect on the operation of the CRDM's.
The benefits of a full RCS chemical decontamination were estimated using the CORA model along with data from two non-domestic plants that had been decontaminated several times and the annual plant doses for all domestic Westinghouse nuclear plants. This evaluation predicted that 900 to 3500 man-rem could be avoided for one decontamination. WOG estimated that at $5000 per man-rem, this would result in a savings of $4.5 to $80 million (the NRC normally uses a savings of $1000 per man-rem).
The benefits from one decontamination are estimated to accrue for the next 5 operating cycles.
Additional benefits could be achieved by eliminating sources of cobalt and by using proper primary water chemistry.
2.0 CONTENTS OF THE TOPI_QfL REPORT The topical report summarizes the work completed during phase I and 2 of the full RCS chemical decontamination program.
Section 1 of the topical report is contained in Volume I and is titled, " FULL REACTOR COOLANT SYSTEM DECONTAMINATION PROGRAM GENERIC TOPICAL PROGRAM REPORT."
This section includes an executive summary, a generic program safety evaluation, fluid systems evaluation, and full RCS decontamination project conceptual design.
The taneric program safety evaluation is a 10 CFR 50.59 generic safety evaluation of phase 2 of this program and includes mechanical equipment evaluation, auxiliary systems evaluation, instrumentation and controls evaluation, and radiological evaluation. A plant-specific 10 CFR 50.59 safety
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evaluation will be prepared as part of phase 3 for the demonstration plant (Indian Point Unit 2). The fluids systems evaluation discusses process flow requirements, temperature control, pressure control, mass inventory, j
decontamination process definition, RCS. flows and velocities,-secondary side leakage, active system interfaces, and operating guidelines.
The fluid systems evaluation discusses the cleanup subsystem, the equipment design bases i
and sizing criteria, the resin processing subsystem, the chemical injection system, the decontamination processing system (DPS) equipment generic layout, component shielding requireme.ts, support system requirements, operating guidelines, leakage and failure analysis, and the occupational radiation i
exposure study. The next' subsection presents a discussion of corrosion j
consequences of residues of the decontamination solution. This section discusses what residues are present and what consequences of the residues are l
to the RCS materials and components.
Finally, there is a section that
.l addresses NRC' concerns that were submitted to WOG prior to the issuance of the topical report.
Section 2 of the topical report is contained in volume 2 and is titled, " FULL RCS DECONTAMINATION PROGRAM WASTE MANAGEMENT METHODOLOGY AND WASTE CHARACTERIZATION." Section 2 contains subsections on calculation of waste volume estimates, regulatory considerations, waste treatment processes, and development of technical specifications and field implementation approaches.
3.0 DESCRIPTION
OF THE CHEMICAL DECONTAMINATION PROCESSES 3.1 CAN-DERfH The Atomic Energy of Canada Ltd. personnel developed the CAN-DECON process in the 1970's to chemically decontaminate CANDU heavy water reactors.
CAN-DECON has also been used to decontaminate BWR's. An oxidizing reagent step is i
combined with the CAN-DECON process for PWR's to remove chromium-rich
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deposits. The oxidizing reagent is a dilute solution of potassjum permanganate and sodium hydroxide (AP) that oxidizes Cr*3 to Cr. The Cr*' is soluble in the CAN-DECON solution.
CAN-DEREM is a reformulation of CAN-DECON with the oxalic acid removed. Oxalic acid can cause intergranular corrosion or stress corrosion cracking under certain conditions. The CAN-DEREM reagent consists of a mixture of organic acids and chelating agents. The organic acids dissolve the oxides, and the chelating agents keep the oxides in suspension until they are removed in a cation resin bed. After the oxides have been removed, the solution is passed through a mixed bed ion exchange resin bed that removes the organic acids, the chelating agents, and any remaining dissolved oxides. The CAN-DEREM process consists of a series of steps as follows: CAN-DEREM/AP/0xalic Acid /CAN-DEREM/AP/0xalic Acid /CAN-DEREM and requires about 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> for a complete cycle.
3.2 LOMI Personnel from the UK Central Electricity Generating Board developed the LOMI process, an acronym for Low Oxidation-state Metal Ion, with financial support from EPRI in the early 1980's. The LOMI reagent consists of vanadous formate and picolinic acid that forms vanadous pico11nate which reduces _ ferric ions to ferrous ions in the oxide deposits.
The resulting oxide structure is soluble
4 in the LOMI reagent.
Excess picolinic acid complexes with the dissolved oxides and keeps the oxides in suspension until they are removed in the resin beds. The AP steps are used with LOMI for the chromium-rich deposits. The LOMI process consists of a series of steps as follows: AP/0xalic Acid /LOMI/AP/0xalic Acid /LOMI and requires about 85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> for a complete cycle.
4.0 EVALUATION 4.1 NRC Ouestions About the Topical Report The WOG topical report was examined, and a number of questions were submitted to WOG. The WOG responded to each question. The. questions and the WOG response are presented in this section.
1)
Discussion of waste management should include the impact of the proposed decontamination process-on the existing liquid, gaseous or airborne and solid radwaste management systems.
Specifically, the discu:sion should identify the parameters for the liquid wastes (gallons per day, total volume of liquid wastes for the entire process, treatment of such liquid wastes, effluent liquid stream radioactivity level expressed as fraction of reactor coolant activity level, and disposition of the treated liquid radwaste), and airborne wastes (expected release by radionuclides via airborne pathway) expected to be generated during the decontamination process. Additionally, the discussion should include the solid radwastes resulting from processing of the liquid and airborne wastes (for example, particulates collected on HEPA filters for airborne wastes) expected to be generated during the decontamination Trocess.
The volume of above solid radwaste'is expected to be small ac J is in addition to the large volume of the c0 lid radwaste that results from the decontamination process itself.
WOG Reply:
Full RCS chemical decontamination operations will have minimal effect on existing plant waste management systems.
The Decontamination Vendor equipment (demineralizers) will have the capability of removing all decontamination process by-products from the primary water, including dissolved corrosion products, radionuclides and process chemicals using ion exchange resins, which are dewatered under Topical Report DW-11118-01-p-a.
At the conclusion of the decontamination operation, the primary fluid will be polished to a conductivity level of 10 micromhos using the decontamination vendor's portable demineralizers.
Following on-line treatment of the primary water (~95,000 gallons), the water may be drained to the plant's radwaste treatment system for ultimate treatment either for disposal or reuse or stabilized as discussed in the last paragraph of this reply.
No radioactive gases will be present during the decontamination process because fuel will be removed from the reactor. Any gases (e.g., carbon dioxide) generated during the decontamination operation will be vented
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from the RCS (pressurizer _and/or reactor head) to either the decontamination vendor's equipment (HEPA Filter).or the plant HEPA-filter, depending on the decontamination vendor's^ equipment. design.
l Dry active waste (DAW) will be dispositioned via the pl' ant's normal DAW j
disposal method (e.g. separation and compaction, incineration).'
A spent resin dewatering process, (Topical Report, DW-Ill18-01-p-a)' is.
I currently being considered for use in stabilizing the waste streami resins. A spent resin cement stabilization process is'also being qualified to provide an alternate procedure for waste stream _-
stabilization.
1 Staff Response-The staff accepts this reply.
Further, the~ topical report states that-the liquid radwaste input to the liquid: waste processing system (WPS)-
would be relatively low-volume, high-activity waste, and high-volume,_
low-activity waste and that these inputs to the:liouid _WPS will be.
typical of the inputs during refueling shutdowns when maintenance i
activities may generate significant waste volumes. The' staff agrees.
with this statement. Since the liquid'WPS-is designed to handle'such-generation of wastes, the: staff does not expect any significant impact of DPS operation.on the liquid WPS. Regarding the impact of-DPS operation on solid WPS, the.0wners' Group submittal of NovemberL25, 1992
-states that the low volume of dry active waste (DAW) expected to be _
generated'during the DPS operation will ha treated and disposed of via-the plant's normal DAW disposal method (i.e.,. separation / compaction);
and the resin wastes generated during the DPS operation will be stabilized by an outside radwaste vendor using the equipment brought to the. site specifically to treat this resin. Thelsubmittal further states i
that vendor-supplied casks will be used to. transport the stabilized -
resin for storage and or disposal. Based on the_above, the staff expects the proposed DPS to have minimal 1:npact on the solid WPS.
2)
The waste management write-up should identify the specific equipment
.I needed (collection and' processing) for handling the liquid and airborne wastes expected to be generated during the~ decontamination process.
Such an identification will enable a specific PWR licensee to determine -
whether the. existing equipment _in the liquid and airborne waste management systems for the PWR can accommodate the additional liquid and airborne wastes that result.from the decontamination process or-additional equipment will be needed.
1 WOG Reply:
The response to question 1) is also ' applicable-to this question. The-identification of specific collection and processing equipment needed to handle the decontamination process waste products is a function of plant-specific requirements and will be developed by the decontamination vendor.
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However, the topical report does identify the magnitude / volumes of decontamination process waste products that were calculated for a generic four-loop. plant and will be used to size the der >ntamination vendor's equipment.
Staff response-The staff accepts this reply.
3)
The proposed outline does not explicitly identify demonstration of compliance of the proposed decontamination process with applicable regulatory positions. The topical report should discuss compliance of the proposed decontamination process with applicable regulations, branch technical positions and standard review plan (SRP) acceptance criteria.
1-Specifically,- the discussion should cover applicable acceptance criteria of SRP Sections 11.2 and 11.4 and the guidelines of Regulatory Guide 1
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1.143.
WOG Reply The NRC question deals with the act of implementing the actual 3
decontamination. However, the purpose of WCAP-12932 was to:
- 1) Generically assess the potential effects of chemical decontamination on the integrity of existing plant systems and components, and
- 2) Address the feasibility of performing a full RCS chemical decontamination.
The first item was addressed by a comprehensive safety evaluation included in the topical report based on the current licensing criteria of a reference plant.
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The second item was based on a fluid systems evaluation and the conceptual design. Design details, and details of an actual i
implementation at a plant, were not stated. This must be provided on a j
plant-specific basis to support actual implementation.
i The regulations stated must be addressed when a plant actually performs a decontamination, and reflected in a plant-specific work package and safety evaluation to show that the implementation will comply with the regulations as discussed in detail in the topical report.
Staff Response-The staff accepts this reply.
i 4)
The proposed outline does not explicitly identify interface requirements. However, under the caption " Issues", the outline states that " application of safety analysis.to a specific PWR requires some degree of plant-specific reconciliation." The topical report should include a discussion of interface requirements.
1 WOG Reply i
There are 3 types of interface requirements of the decontamination
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7 processing system. The first is the tie-in to the NSSS and the N8SS operations, which are described in Sections 1-2 and 1-3-5-1 of the topical report. The second type is support system requirements, which are contained in Section 1-3-7.
The third type is general layout requirements for the decon equipment relative to existing plant buildings.
Various options for layout arrangements are described in Section 1-3-5.
Staff Response-The staff accepts this reply.
5)
The topical report states that since there are no component cooling water system (CCWS) material effects due to the DPS operation, the safety function of the CCWS will not be compromised. The staff sought the basis for the position that there are no CCWS material effects due to the DPS operation.
WOG Reply Only the primary plant systems are wetted by the decontamination process chemical reagents. The CCWS is a secondary system which provides cooling water to the various primary system heat exchangers and will not be exposed to the chemical reagents since the primary system and the secondary systems are isolated from each other.
Staff Response-The staff accepts this reply and concludes that the safety-related function of the CCWS will not be compromised by the DPS operation provided the following is satisfied: the DPS operation by itself does not cause any tube leakage in the RHR heat exchanger (such a leakage will establish contact between the tube' side containing reactor coolant and decontamination solution and the shell side containing the component cooling water).
6)
Will the waste be EPA hazardous?
Is the waste classified as a mixed waste? Has WOG addressed EPA requirements?
WOG Reply Chromates will be the only hazardous material in the waste products, no other EPA hazardous wastes will be generated.
EPA toxicity tests will be conducted on the waste to determine if it is EPA hazardous. WOG has addressed the EPA requirements.
Staff Response-The staff accepts this reply.
7)
Some of the waste streams are Class A, some Class B, and some Class C.
Could WOG make a table that summarizes which waste streams are which class?
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WOG Reply The table exists in Volume 2 of the topical report, Table 2.1-66.
l Staff Response-The staff accepts this reply.
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8)
How accurate. are WOG's projections of the actual-waste?
j A factor of 1.6 was used to make the projection be conservative.
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Staff Response-The staff accepts this reply.-
l 9)
Is there an interaction between the CAN-DEREM chemicals and cracks in -
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Inconel 600 tubes.
.1 WOG Reply During.the decontamination test program, extensive evaluations were j
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performed on steam generator tube materials-including Alloy 600 and 690.
The specific evaluations included general corrosion,.intergranular l
penetration susceptibility, stress corrosion. cracking susceptibility, and crevice corrosion.
Evaluations were also performed to establish the i
effects of a PWR environment (water chemistry and temperature) following.
l decontamination on stressed'and creviced specimens. The specifics of the test program and data obtained are provided in Volume 1 of the Phase 1
2 Qualification Program Report.
Specifics relative to CAN-DEREM are located in the referenced report on or beginning with the following pages: 3.1-15 (specimen types and materials,'3.1-273 and 3.1-187 (general corrosion), 3.1-209 and 3.1-262'(stress corrosion cracking),
3.1-217 (intergranular penetrations), and 3.1-265 (effect of PWR-environment).
)
m The conclusions supported by the' data generated on Alloy 600 and 690 are' as follows:
- 1) Decontamination does not cause intergranular. attack or grain boundary penetration.
- 2) Decontamination does not cause stress corrosion cracking.
- 3) Decontamination does not accelerate the-PWSCC initiation kinetics in a primary water environment.
Eddy current testing of the steam generator tubes'is scheduled to be performed after the decontamination operation which would identify any cracks present in the steam generator tubes.
Conductivity tests will be performed when the RCS fluid is in the domineralized mode.
Staff _ Response-The staff accepts this reply.
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The NRC staf7 has not approved the use of Fera111um for high integrity containers (HIC) for decontamination resin wastes.
WOG Reply There are no plans by WOG to use this material.
Staff Response-The staff accepts this reply.
11)
A separate waste solidification qualification program is required for the waste streams predicted for the demonstration plant (Indian Point Unit 2). What are the WOG plans for the qualification program?
WOG Reply Work is currently underway at Chem Nuclear facilities in Barnwell, South Carolina on the waste solidification qualification program for Indian Point Unit 2.
The results of this program will be submitted to NRC as an independent topical report early in 1993. The target date is January 15, 1993.
Staff Response-The staff accepts this reply. Any nuclear plant licensee that wants to reference this topical will have to address the waste solidification qualification program on a waste stream-specific basis.
12)
The applicant states that during the full RCS chemical decontamination operation, the DPS may fail. Consequently, a limited (due to valves that will automatically close on significant reductions of pressure in the DPS) volume of the decontamination solution containing radioactive material removed from the RCS surfaces may be released.
WOG Reply The current DPS conceptual design includes seismically designed dikes that will contain the maximum postulated liquid release and, that the off-site radiological consequences need not be evaluated from a liquid effluent perspective. Therefore, an evaluation was completed for one plant for the off-site radiological consequences of only air-borne release resulting from the accidental decontamination solution release.
Staff Response-The staff agrees with WOG that it is acceptable to limit the evaluation to air-borne releases. WOG determined that the off-site dose for a l
postulated decontamination solution release would be below 0.5 rem to l
the whole body or its equivalent to any organ and, therefore, concluded -
that the radiation dose will be within the applicable regulatory limits.
The applicant, however, stated that analysis of the above accident has to be evaluated on a plant-specific basis.
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13)
What impact will the decontamination process system (DPS) have_on the i
CCWS7 WOG Reply i
The decontamination program will wet only primary plant _ systems and the.
CCWS which is a secondary system will.not.be exposed to the chemical i
reagents of the DPS, since the primary and secondary systems _are:
isolated from each other. The DPS operation is not. expected to cause any unusual operating conditions for the CCWS.
The DPS operation will not affect the CCWS's safety-related function, which.is supplying.
cooling water to essential components'such as RHR ' system heat exchanger-and spent fuel pool cooling heat exchangers.
Staff Response-i The staff concluded that the ' safety-related! function of the CCWS.will' not be compromised by the DPS operation, provided the operation by itself does not cause any tube leakage in the RHR heat exchanger (such l
1eakage would establish contact between the tube side containing. reactor coolant and decontamination solution and the shell side containing the component cooling water)..
4.2 11anificant I sues Covered in the Topical Report t
The topical report makes the following recommendations'for continued safe' operation after a chemical decontamination:
i 1)
If the primary system automatic relief valve is exposed to the decontamination chemicals, additional evaluation is required on a plant specific basis.
j 2)
At least one reactor coolant pump must operate continuously throughout the decontamination process.
3)
A maximum of three reactor coolant pumps should be used in a four-loop plant configuration._
4)
Fuel assemblies must be removed from the reactor vessel during the decontamination process..
5)
If cracks or other openings exist. in the internal cladding of the reactor vessel, the area should be identified and analyzed to insure-that the code minimum wall thickness is maintained and that calculated stress intensities satisfy code limits _ after a decontamination cycle._.
If the thickness is marginal, inspection of this area'must be performed after decontamination.
6)
Control rod drive shafts must be removed from the reactor prior to performing the decontamination process.
7)
Control rod drive mechanisms will be exercised using current plant-i i
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11 specific procedures after completion of the' decontamination' process.
8)
- Provisions must.be made to pressurize or fill the secondary side of the
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steam generator to prevent decontamination process. fluids,from entering the secondary side. 'If leakage to the secondary side can not be.
j prevented, additional-evaluation is required.and acceptance criteria must be developed.
i 9)
The reactor coolant' pump (RCP) seals may be adversely affected if--
exposed to the chemical decontamination chemicals; WOG recommends-in the topical report that clean-seal. water be supplied-to the RCS's seal system during chemical decontamination. 'If clean seal water cannot be 1
provided,. all RCS's seal, systemsshould be. inspected and ' serviced ~
following decontamination. The: silicon nitride in:the No. l' seal may corrode during decontamination and may malfunction after.
decontamination.. WOG recommends that the No. I seals be replaced if exposed to the decontamination solution.1 W also recomends that the seal. system be inspected if exposed to the decontamination solution.-
The biggest concern is the flaking of chrome from the RCP shaft for plants with model 93Al and 100 pumps. The particulates of chrome could-enter the seal area and seal degradation'is highly probable. WOG suggests that RCP seal operation and vibration be closely monitored -
after chemical decontamination according to normal ~ operating procedures.
10)
If a faulted condition exits for more than~~approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for.the CAN-DEREM process, the reactor. vessel upper internals top hat plate and guide tube fillet weld stresses must be evaluated against design stress limits to determine acceptability on a plant-specific basis. -
11)
The evaluation of the need for additional pipe supports for temporary or-permanent equipment used for decontamination has not been performed.
1 This evaluation. must be perfomed on a plant-specific basis.
12)
Steam generator maintenance and inspection records must les reviewed to identify any cladding cracks which would expose carbon steel base j
material to the decontamination chemicals. These areas must be evaluated for adherence to code requirements for minimum wall thickness and stress intensity level.
13)
Automatic valves with redundant isolation signals must be provided for DPS isolation capability.
14)
Take specific actions required to return a plant to normal operation:
1 a)
Flushpipingdeadlegstoremoveresidueandcruh b)
Perfom inspections and testing to' determine that;the systems and components continue to meet their design basis acceptanco criteria;'
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-and, c) Determine that the future operability for the RCS and interfacing systems has not been' negatively affected.
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4.3 NRC Staff Visit to the Chalk River Nuclear Laboratory of the Atomic Enerav of Canada Limited e
The NRC Staff visited the Chalk River Nuclear Laboratory of the Atomic Energy of Canada Limited (AECL) to discuss full RCS: chemical' l
decontamination. AECL developed the CAN-DECON and CAN-DEREM systems,and have conducted 14 full RCS decontaminations with fuel'in place. The only corrosion problem AECL has observed was slight. pitting on a type j
410 stainless steel component. They did-not observe extensive corrosion of carbon steel. under breaks. in the stainless steel cladding. They.
observed the preferential dissolution of manganese sulfide inclusions in
- the steel but no general corrosion of the steel as observed in the WOG laboratory tests.
Based on the AECL results, the staff concludes that the WOG corrosion test results-are conservative.
.l 4.4 Land Disposal of Radioactive Waste The generic topical report program states that wastes generated from the full RCS decontamination program will-either be placed in' high' integrity containers (HIC) or_ solidified in. cement'for disposal..The. topical.
report did not include a qualification test program to demonstrate that the resulting waste forms will meet the waste form stability.
requirements _of 10 CFR 61.56. The topical. report indicates that waste forms will be subjected to qualification programs on a plant-specific basis, because of the expected variability in waste streams among users of_ the full RCS decontamination process. Waste forms should be subjected to qualification test programs on a waste stream specific basis.
The topical report also indicates that. cement waste form qualification will be performed following the NRC Branch Technical Position (BTP) on Waste Form (1983). This BTP was revised in 1991 by the addition of Appendix A on Cement Stabilization. - Section 2.0 of the topical report should be corrected to-refer to BTP Revision 1, January 1991. A copy of this revision to the BTP is enclosed.-
The BTP emphasized the importance of characterizing each waste stream, including identifying ranges of waste constituents _ and elemental-concentrations which will affect the ability of the proposed solidification formulations and/or the selected HIC-to meet the stability requirements. Review of.the~' demonstration project waste form qualification report will focus on characterization of the waste streams and on test data sufficient to demonstrate-that the waste forms will i
' meet the stability requirements of 10 CFR 61.56E Given the-unique. characteristics of the wastes produceF oy +he' i ~. -
decontamination process, additional considerations beyond thou in Appendix A to the BTP on waste form may be necessary as part of thaY._
cement waste fora qualification program to demonstrate compliance wf?.h.
the 10 CFR Part 61 waste form requirements. Any_ power plant seeking to -
reference the _ approved topical report to demonstrate compliance with 10 CFR 61.56 must first_ address the land disposal issues associated with \\
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SUMMARY
. CONCLUSIONS. AND LIMITATIONS i
i The staff reviewed the WOG Topical Report, " Full RCS Chemical Decontamination Program," Revision 1, WCAP-12932. This topical report addresses the issues-involved during the full RCS chemical decontamination of a typical four-loop.
Westinghouse PWR. The results also apply to a typical two-loop and three-loop Westinghouse PWR..
A number of issues were raised by the staff on the topical report.There are no.the f
All of issues were resolved as presented in this safety evaluation..
3 open issues.
This topical report is approved by' the staff and may'be referenced by nuclear power plant licensees with the following limitations:
1)
The applicant must conduct a plant specific.10 CFR 50.59 safety analysis; j
2)
The licensee must evaluate pipe supports for permanent and temporary i
piping associated with the chemical decontamination equipment.
3)
NRC has not reviewed waste qualification test data or granted an approved 10 CFR Part 61 waste qualification program for stabilization of
. i the waste products projected in this topical report. _. Any licensee of a nuclear power plant who wants to reference this topical report must addrest the land disposal issues associated with the waste products on a
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waste. stream specific basis.
t 4)
The licensee must insure that the licensee is in compliance with all applicable NRC regulations as discussed in the topical. report.
5)
The licensee must take the specific actions recommended in the topical report to return the plant to normal.
Principal Contributor:
J. A. Davis (301) 504-2713 b
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Mr. P. E. Miller Westinghouse Electric Corporation P. O. Box 355 Energy Center East-Bay 511 Pittsburgh, PA 15230 l
Dear Mr. Miller:
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SUBJECT:
APPROVAL FOR REFERENCING OF LICENSING TOPICAL REPORT NO. WCAP 12932, FULL RCS CHEMICAL DECONTAMINATION PROGRAM, REVISION 1, VOLUME I AND 2, TAC M83770 We have completed our review of the subject topical report and have j
l incorporated your comments in Revision 1.
This report is approved for J
l referencing in license applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation which is enclosed. The evaluation defines the basis for the approval of the report.
1 Please note that NRC has not reviewed or approved a waste qualification test program for stabilization of the waste products projected in the Westinghouse i
Owner's Group topical report and cannot proceed with a meaningful review until sufficient test data are available.
We do not intend to repeat our review of the approved matters described in the report when the report appears as a reference in license applications except to assure that the material presented is applicable to the specified plant involved. Our approval applies only to matters described in the report.
i In accordance with procedures established in NUREG-390,.it is requested that Westinghouse Owners Group (WOG) publish approved v::rsions of this report, proprietary and non-proprietary, within three months of receipt of this letter. The approved version should incorporate this letter and the enclosed avaluation between the title page and the abstract. The approved version shall include an -A (designating approved) following the report-identification symbol.
Should the Commission regulations or the staff implementation guidelines change such that our conclusions as to the acceptability of the report are l
l invalidated, WOG and/or the licensees referencing the topical report will be l
expected to revise and resubmit their respective documentation, or submit l
justification for the continued effective applicability of the topical report without revisions of their respective documentation.
Sincerely, onginalSignedBy:
James E.Mahardson
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James E. Richardson, Director Division of Engineering Office of Nuclear Reactor Regulation
Enclosures:
As Stated I
See next page for Distribution and Concurrence
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