ML20024H935

Forwards Initial Series of Preliminary Proposals Involving Various Reactor Plant Component Aging Studies Which Staff Would Like to Conduct at Plant.Requests Meeting W/Util on 930901 to Discuss Feasibility of Performing Studies

ML20024H935
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	08/13/1993
From:	Masnik M Office of Nuclear Reactor Regulation
To:	Cross J PORTLAND GENERAL ELECTRIC CO.
References
NUDOCS 9308300154
Download: ML20024H935 (27)

Text

r, August 13, 1993 Docket No. 50-344 Mr. James E. Cross i

Vice President and Chief Nuclear Officer l

Portland General Electric Company 121 S.W. Salmon Street i

Portland, Oregon 97204 l

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SUBJECT:

NRC PRELIMINARY PROPOSALS FOR TROJAN AGING STUDIES The NRC staff has prepared an initial series of preliminary proposals involving various reactor plant component aging studies which the staff would i

like to conduct at the Trojan Nuclear Plant. The research studies are grouped into three phases of activity:

(1) near-term (FY-93) which would include the-cable aging assessment and a tendon grease-leakage effect study, (2) intermediate term (FY-94 +) which would include a reactor pressure vessel support structure embrittlement study, a study on degradation of reactor core internals, and structural aging investigations, and (3) long-term (starting after FY-94) which would include the remainder of the programs.

The proposed preliminary studies are enclosed for your review.

The NRC staff would like to meet with Portland General Electric on September 1,1993, at the Trojan site to discuss the. feasibility of performir.g these studies.

Please feel free to contact me at (301) 504-1191 to discuss any issues regarding the proposed studies and to confirm the acceptability of the proposed meeting.

l Sincerely, l

Original signed by P. Erickson for i

Michael T. Masnik, Senior Project Manager Non-Power Reactors and Decommissioning Project Directorate l

Division of Operating Reactor Support j

Office of Nuclear Reactor Regulation i

Enclosures As stated cc w/ enclosures:

See next page DISTRIBUTION:

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6 August 13, 1993 Docket No. 50-344 Mr. James E. Cross Vice President and Chief Nuclear Officer Portland General Electric Company 121 S.W. Salmon Street Portland, Oregon 97204

SUBJECT:

NRC PRELIMINARY PROPOSALS FOR TROJAN AGING STUDIES The NRC staff has prepared an initial series of preliminary proposals involving various reactor plant component aging studies which the staff would like to conduct at the Trojan Nuclear Plant. The research studies are grouped into three phases of activity:

(1) near-term (FY-93) which would include the cable aging assessment and a tendon grease-leakage effect study, (2) intermediate term (FY-94 +) which would include a reactor pressure vessel support structure embrittlement study, a study on degradation of reactor core internals, and structural aging investigations, and (3) long-term (starting after FY-94) which would include the remainder of the programs.

The proposed preliminary studies are enclosed for your review.

The NRC staff would like to meet with Portland General Electric on September 1,1993, at the Trojan site to discuss the feasibility of performing these studies. Please feel free to contact me at (301) 504-1191 to discuss any issues regarding the proposed studies and to confirm the acceptability of the proposed meeting.

Sincerely, Original si;;ned by P. Erickson for Michael T. Masnik, Senior Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page DISTRIBUTION:

Docket File 50-344 RDudley AThadani (8-E-2)

PDRs MMasnik JWiggins, R7 ONDD r/f OGC JCraig (NLS007)

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Docket No. 50-344 Mr. James E. Cross Vice President and Chief Nuclear Officer Portland General Electric Company 121 S.W. Salmon Street Portland, Oregon 97204

SUBJECT:

NRC PRELIMINARY PROPOSALS FOR TROJAN AGING STUDIES The NRC staff has prepared an initial series of preliminary proposals involving various reactor plant component aging studies which the staff would like to conduct at the Trojan Nuclear Plant. The research studies are grouped into three phases of activity:

(1) near-term (FY-93) which would include the cable aging assessment and a tendon grease-leakage effect study, (2) intermediate term (FY-94 +) which would include a reactor pressure vessel support structure embrittlement study, a study on degradation of reactor core internals, and structural aging investigations, and (3) long-term (starting after FY-94) which would include the remainder of the programs.

The proposed preliminary studies are enclosed for your review.

The NRC staff would like to meet with Portland General Electric on September 1,1993, at the Trojan site to discuss the feasibility of performing these studies.

Please feel free to contact me at (301) 504-1191 to discuss any issues regarding the proposed studies and to confirm the acceptability of the proposed meeting.

Sincerely, l

h A

Michael T. Masnik, Senior Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation

Enclosures:

As stated I

cc w/ enclosures:

See next page i

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v Mr. James E. Cross Trojan Nuclear Plant Portland General Electric Company Docket No. 50-344 i

cc:

l Senior Resident inspector f

U.S. Nuclear Regulatory Commission Trojan Nuclear Plant P. O. Box 250 Rainier, Oregon 97048 Mr. Michael J. Sykes, Chairman i

Board of County Commissioners Columbia County i

St. Helens, Oregon 97501 Mr. David Stewart-Smith Oregon Department of Energy Salem, Oregon 97310 Regional Administrator, Region V i

U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Mr. Tom Walt General Manager, Technical Functions Trojan Nuclear Plant 71760 Columbia River Highway Rainier, Oregon 97048 Mr. Lloyd K. Marbet 19142 S.E. Bakers Ferry Road Boring, Oregon 97009 Mr. Jerry Wilson Do It Yourself Committee 570 N.E. 53rd Hillsboro, Oregon 97124 i

4 Mr. Eugene Rosolie i

Northwest Environmental Advocates 302 Haseltine Building i

133 S.W. 2nd Avenue Portland, Oregon 97204

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TROJAN NUCLEAR PLANT TENDON GREASE-LEAKAGE TITLE OF PROJECT:

Effect of Tendon Grease-Leakage on the Integrity of Prestressed Concrete Containments OBJECTIVE:

To determine the effect of grease leakage on the integrity of prestressed concrete containment SCOPE:

(I) To identify and select areas in the containment when grease leakage is evident (2) To identify all documentation that currently exist for these selected areas (3) To study and conduct an NDE of the in situ concrete to confirm the area of investigation NEED FOR WORK AND APPLICATION:

NRC requires an assessment of the grease-leakage on and through the containment to determine the extent of the potential safety concern that exist or may exist.

EXPECTED RESULTS:

The expected results would be an assessment of the effects of grease leakage on the prestressed concrete containment - a determination of potential damage.

BENEFIT OF WORK:

The benefit of tne work would be to obtain better estimates of the remaining structural safety margins of the PCC.

LIMITATIONS AND ALTERNATIVES:

Ability to select and obtain samples for meaningful laboratory investigation.

The alternative is to try to get agreement to take samples from an operating

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i DUPLICATION OF EFFORT:

None

1 NEED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN:

Trojan would need to provide all available documentation concerning the fabrication practice of the PCC and the tendon and grease installation and maintenance of the given areas.

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TROJAN NUCLEAR PLAMT PROPOSAL FOR RESEARCH USING AGED CABLES TITLE OF PROJECT: Aging Assessment of Cables OBJECTIVE: To understand aging effects and evaluate the effectiveness of the state-of-the art condition monitoring methods for detecting degradation of electric cables (power, control and I&C).

SCOPE:

Perform insitu assessments of selected cable systems.

Perform confirmatory research. Compare physical and electrical properties of naturally aged cables to those measured in the cable aging research program at Sandia National Laboratory (SNL) under controlled laboratory conditions.

Test sections of the cables and compare the physical and electrical properties of new vs. aged cables of similarity.

Perform LOCA tests on a few samples of the naturally aged cables.

A 3-phase approach for aging assessment is proposed.

Phase 1: Review and evaluate historical plant specific data including environmental conditions (actual vs. those used during qualification), plant applications, and conduct walkdown.

Identify typical representative cables (systems) of interest.

Phase 2: Perform insitu assessment of selected cables (systems).

Evaluate state-of-the art cable diagnostic techniques. Obtain cable samples.

Phase 3: Perform electrical and mechanical tests and LOCA tests on a few samples of the cables.

Issue a technical report.

Expected program duration 2-3 years.

(to be revisited based upon the availability of cable samples and resources (FTEs, and $)).

NEED FOR WORK AND APPLICATION:

Develop technical bases for residual life assessment of naturally aged cables. The results will be useful for confirming continued qualification and add to the EQ Data Base.

EXPECTED RESULTS: Evaluate state-of-the art condition monitoring methods.

Improved understanding of aging degradation effects and impact of actual plant operating environment vs. that prescribed in EQ requirements for Class IE cables inside containment.

BENEFIT OF WORK. AND USE OF TROJAN:

Results will be beneficial for developing technical bases to demonstrate the functional performance capability of safety-related cables.

LIMITATIONS AND ALTERNATIVES:

Limitations:

Resources (FTEs and 5).

Timely availability of cables.

Alternatives:

Obtain samples of naturally aged cables from other operating and decommissioned plants.

DUPLICATION OF EFFORT:

None. The research program will be coordinated with the related DOE and EPRI programs.

NEED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN:

I.

Equipment qualification file for power, control and I&C cables.. Replacement criteria used for safety related cables and technical vendor specifications.

2.

Service records (includes actual plant environmental conditions).

3.

Engineering drawings and layouts of cable systems.

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TROJAN NUCLEAR PLANT RPV SUPPORT STRUCTURE EMBRITTLEMENT STUDY TITLE OF PROJECT:

RPV Support Structure Embrittlement Study OBJECTIVE:

The objective of this study is to obtain samples of the RPV support structure for the purpose of determining the level of embrittlement of the material 'or comparison to existing embrittlement correlations, and to aid in resolution of Generic Safety Issue 15, " Radiation Effects on Reactor Vessel Supports."

SCOPE:

(1) Devise a method for removing pieces of material from the horizontal beam from at least one of the RPV support structures.

Each piece removed from the beams must be large enough so that one or more full-size Charpy specimens can be fabricated from each piece. The removal method must keep the metal at a temperature below the normal service temperature of the support structure, e.g., below approximately 100 F.

The orientation of the removed section must be clearly identified on the piece.

(2)

Remove a sufficient number of pieces from the support structures so that at least 15 Charpy specimens from each sampled beam can be fabricated in the same orientation.

(3) Ship the removed pieces to ORNL for testing.

The work to be performed at ORNL would involve fabricating Charpy test specimens, and testing them at a range of temperatures to develop a Charpy energy versus test temperature curve for each beam. Specimens also would be annealed and tested to provide an "unirradiated" curve.

The level of embrittlement would be estimated based on the shift between the unirradiated and irradiated curves.

NEED FOR WORK APPLICATION:

Generic Safety Issue 15, " Radiation Effects on Reactor Vessel Supports," was reprioritized from " medium" to "high" priority based on the level of embrittlement determined by Oak Ridge National Laboratory for the High Flux Isotope Reactor (HFIR) and on the results of structural integrity analyses for typical RPV support structures where the level of embrittlement was estimated from the HFIR results. After extensive work to resolve questions concerning the HFIR embrittlement, it has become apparent that the best way to resolve GSI-15 is to obtain samples of the support structure materials from plants such as Trojan. The results of testing the Trojan RPV support structure materials will directly affect the resolution of this GSI, either by indicating that the HFIR results are not applicable or by supporting the HFIR results and indicating the need for additional generic actions.

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- EXPECTED RESULTS:

The expected results of this study will be an assessment of the level of embrittlement in the Trojan RPV support structures that can be used to either confirm or refute the contention that the HFIR embrittlement trend is applicable to RPV support structures.

BENEFIT OF WORK. AND USE OF TROJAN:

The benefit of the work is to provide experimental evidence of the level of embrittlement of the RPV support structure for a typical PWR, This evidence can them be used directly in reaching a resolution to b51-15. Using materials from Trojan have several advantages.

First, the operating conditions and support structure materials are typical of many PWRs i.. the U.S.

Additionally, the Trojan RPV support structures have been evaluated extensively as part of the GSI-15 Task Action Plan activities and experimentally determining the level of embrittlement for the Trojan materials will allow the resolution of GSI-15 to build on that past work.

LIMITATIONS AND ALTERNATIVES:

The primary limitations will come from the sample size that can be obtained from the support structures and the restricted access to the horizontal beams.

While larger samples could be obtained if and when the vessel is dismantled, this would impose a potentially significant delay in obtaining the materials.

.N_EED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN:

Details of the plant operation over its service life, and a report on the vessel sampling activity.

1 NRC PROPOSAL FOR TROJAN NUCLEAR PLANT STUDIES i

TITLE:

DEGRADATION OF REACTOR CORE INTERNALS OBJECTIVE:

The objective of this study is to obtain samples of highly irradiated materials representing a variety of components of reactor core internals, to determine degradation levels and cracking mechanisms, and provide for validation of models developed from laboratory and other studies.

SCOPE OF WORK:

(1) Review the materials and components available from the early removal of core internals, and make appropriate selection.

Obtain small samples suitable for study in contractor hot cells and ship.

(2) Conduct mechanical property and microstructural analyses of the materials, and compare to measurements and observations from other studies to validate laboratory studies and update predictive models.

NEED FOR WORK AND APPLICATION:

Experimental irradiations have been done with both typical commercial alloys as well as with model alloys to determine the factors affecting irradiation assisted stress corrosion cracking of reactor core internals. This is an important issue since failure of core internals might impede prompt insertion of control or safety rods, and thereby hamper proper control of the reactor.

Core internals become irradiated to very high levels, typically in excess of 10E21 n/cm2, E>1MeV, which are very difficult to duplicate in experimental and test reactor irradiations. Thus, in order to assure that the true effects are available and being correctly studied, samples of materials taken from operating reactors are required for correlation, calibration and validation. Validation of the models would then provide a means for more accurate prediction of degradation in core internals for operating reactors during both the original and renewal license periods, as well as for advanced reactors.

EXPECTED RESULTS:

Test results would include tensile, fracture toughness, stress corrosion behavior, composition, and microscopic observations including grain-boundary segregation and depletion of specific elements; these items would be evaluated in reference to their neutron fluence and the energy spectrum.

BENEFIT OF WORK, AND USE OF TROJAN:

The benefit of this work will be to enable better understanding and prediction of degradation and cracking of reactor internals. From this could come revised inspection schedules if necessary, or decisions for early repair or replacement to preclude unexpected fracture. Trojan offers an opportunity for high fluence level exposures in a typical operating reactor environment, for validation of the laboratory models.

i LIMITATIONS AND ALTERNATIVES I

Failure to obtain materials from Trojan will necessitate the extrapolation of I

results from other irradiations, perhaps even from fast breeder reactors where i

the neutron spectrum (and corresponding damage) would be quite different.

An alternative to use of materials from Trojan would be the use of core internals from another decommissioned reactor.

DUPLICATION OF EFFORT:

Only very limited testing has been done on core internal materials removed from i

operating reactors.

While some duplication of efforts would be welcome for l

corroboration of results, because of the high radiation levels and attendant costs, there is no known random duplication of effort.

NEED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN:

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Information needed from Trojan on core internals includes:

I o Type and chemical composition of heats of materials o Time frame, source and method of fabrication i

o Fast neutron flux and fluence (E>lMeV and E>0.lMeV) for each component, and i

method used to determine fluence o Time-at temperature history for each component o Estimated service stress level for specific components.

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NRC PROPOSAL FOR TROJAN AGING STUDIES i

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Title:

TROJAN STRUCTURAL AGING INVESTIGATIONS Objective:

Determine how, and to what degree, structural materials change with time under defined loads and environments.

Scope of Work:

Take samples from specified areas of concrete or steel structures that have been subjected to conditions that may have led to material degradation over time. Some of the conditions of interest, but not limited to these, would be exposure to borated water or other corrosive elements, radiation, or high temperature.

Changes in permeability of concrete is another characteristic of interest. Also, the effectiveness of NDT techniques to identify liner degradation is of interest.

Need for Work and its Application:

Not enough data is yet available to make quantified estimates of current or i

future residual structural margins, taking into account material degradation or other changes with time.

Expected Results:

Additional data will be generated to add to the general structural materials database that ORNL is developing for quantified time-dependent changes to these materials.

Benefit of Work, and Use of Trojan:

Better estimates can be made of remaining structural safety margins for structural elements that have been impacted by time, environment or unanticipated loadings.

Limitations and Alternatives:

i Documentation of initial material conditions, and a documented exposure history is needed to put any sample test results into perspective, so that a change in properties with time can be determined. Other limitations may be a lack of material subjected to aging conditions of interest, material that is unique to Trojan, accessibility of material of interest for obtaining samples, and (as usual) funding. The alternative to getting samples from Trojan is to try to get agreements to sample from operating plants.

Duplication of Effort (If any, and why):

Duplication of obtaining data samples is not a concern in this type of l

information gathering effort.

It is, however, unlikely that there would be a duplication of effort in sampling.

i Need and Description of Documentation from Trojan:

i Trojan would need to provide all available documentation concerning the fabrication (construction) practice for the Category I structures and any pertinent maintenance and/or repair records.

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TROJAN NUCLEAR PLANT STEAM GENERATOR RESEARCH There are three categories of interest for conducting research using Trojan's Steam generators.

1.

In situ work while the generators are still in place.

2.

Tests on tubes removed from Trojan generators.

3.

Research on a generator removed from Trojan in the NRC steam generator examination facility at PNL.

Work under category 1, of necessity, is considered short-term; category 2 includes both short-term and long-term work and category 3 work can be considered long term. The following studies are envisaged:

Catecory 1 Leak rate test under normal operating and accident conditions of the entire steam generator (SG), on sections of the SG and on single tubes e

Perform advanced technique inspections Conduct repair procedures (laser welding) on degraded tubing Cateaory 2 Leak rate tests and burst tests Examination of previously plugged tubes Metallagraphic characterization of flaws Continue degradation in autoclaves Evaluation of advanced inspection techniques Evaluation of repair procedures Cateoory 3 Round robin inservice inspection of SG tubes using commercial teams and techniques Round robin inservice inspection of SG tubes using advanced / emerging technologies Chemical cleaning / decontamination Secondary side inspection / characterization Characterization of crevices, corrosion products and base materials Repair procedures.

f TROJAN NUCLEAR PLANT Steam Generator Research Category 1 Work: Onsite Testing OBJECTIVES:

Objectives of this work are to evaluate 1) deterministic and probabilistic i

models for calculating leak rate in operating plants from inservice inspection results and other available inputs / correlations, 2) differences in leak rates

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from degraded tubes, in situ versus tubes tested in the laboratory without support structures and deposits, 3) feasibility and efficacy of repair techniques such as laser welding and 4) the capability of advanced inspection techniques.

SCOPE OF WORK:

(1)

Review inservice inspection records and select tubes to be tested in place and tubes to be removed for future tests (leak rate tests, burst l

tests, repair tests, advanced NDE tests, metallgraphic flaw characterization, etc.)

(2)

Conduct leak rate tests under normal operating and accident pressure conditions for the entire steam generator, for sets of tubes and for single tubes.

(3)

Remove plugs from degraded tubes to allow for inclusion of some of these tubes in the work under items (4) and (5) below, and for removal of previously plugged tubes for inclusion in the studies under Category 2 work.

(4)

Perform advanced eddy current inspection of selected tubes (several hundreds) for characterization of flaws and possibly for characterization of deposits, sludge, crevice conditions, etc.

(5)

Perform laser weld repair procedure on selected tubes with a variety of I

flaw sizes, crevice conditions (tightness) and deposits.

Include tubes 1

with cracks extending beyond the tube support plate if available.

NEED FOR WORK AND APPLICATION:

Deterministic and probablistic models for predicting leak rate under normal operating and accident conditions from degraded tubes in steam generators have been developed and used. These models have often been developed using limited data and/or data and correlations from laboratory work where the flaw morphology and tube conditions do not necessarily represent the inservice conditions such as found in the Trojan generators.

Correlations are also l

available for predicting leak rate and burst pressures as a function of flaw type and size. Again, development of these correlations did not necessarily

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use flaw sizes and morphologies and tube conditions representative of current-day degradation such as exists in the Trojan generators.

Conduct of the work described above in conjunction with some of the studies under category 2 work l

will provide validation for these models and correlations which are used for I

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4 predicting integrity and leak rate of degraded tubes in operating steam generators for safety evaluations.

Advanced eddy current techniques, probes and equipment are under development.

Use of these techniques and equipment on Trojan SG tubes will allow validation on realistic flaws and conditions to gain confidence in their capability for inservice inspections.

Laser weld repair techniques for cracked tubes are being developed and evaluated. There is a need to perform the welding procedure under realistic conditions incorporating 1) the range of crack sizes i

(length and depth), number of cracks in a given area, 2) the range of crevice conditions: gap size and geometry and 3) the amount and type of corrosion products and deposits present to evaluate the feasibility and quality of the l

weld repair.

Further, evaluation of the repaired tubing under category 2 work l

is needed to evaluate the long-term serviceability and stress corrosion l

cracking susceptibility of the weld repaired tubing. The results will be used to determine the acceptance of degraded tubes that had exceeded the plugging limit and were repaired by the laser welding process.

EXPECTED RESULTS:

The expected results include data on leak rate for the entire steam generator, leak rate as a function of flaw size, morphology and pressure, data on the capability of advanced eddy current techniques for detecting and characterizing stress corrosion cracking at tube support plate locations and

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data on the feasibility of laser weld repair of cracked.

i BENEFIT OF WORK AND USE OF TROJAN The benefit of this work is that it will provide for validation of existing models, correlations, inspection and repair under realistic conditions.

This validation will provide confidence in the use of validated models and techniques when applied to similar plants and conditions.

DUPLICATION OF EFFORT l

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NEED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN l

Fabrication records, mill certifications, identification of tubes with

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especially from the last two inspections.

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TROJAN NUCLEAR PLANT STEAM GENERATOR RESEARCH Category 2 Work: Tubes Removed from Generators l

I OBJECTIVES:

The objectives of this work are to 1) evaluate differences in leak rates from tubes leaking in situ versus tubes tested in the laboratory, 2) validate or modify correlations for predicting burst pressure of degraded tubes as a function of crack size and morphology for ODSCC type flaws, 3) determine if previously plugged tubes continue to degrade, 4) evaluate progression of crack morphology and relate to leak rate, burst pressure and EC voltage response, 5) i evaluate capability of advanced inspection techniques, and 6) evaluate quality and long-term serviceability of weld repaired tubes.

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SCOPE OF WORK:

(1)

Remove tubes from generator including tubes that were leak tested, inspected, repaired and unplugged under category 1 work.

(2)

Decontaminate and ship as necessary.

(3)

Perform eddy current tests and other advanced NDE tests to achieve the best possible nondestructive characterization of tubes.

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(4)

Conduct leak r te and burst tests.

4 (5)

Conduct metallographic evaluations of tubes that were a) inspected in the SG using advanced techniques, b) leak tested (in and out-of i

generator), c) burst tested and d) previously plugged.

I (6)

Use selected service degraded tubes to continue degradation in autoclaves for varying times and different environments; attempt crack growth rate measurements during these exposures.

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(7)

Conduct eddy current tests, leak and burst tests of tubes exposed in autoclaves and perform metallgraphic evaluations.

(8)

Conduct data analysis and evaluations to validate and/or update models and correlations for predicting leak rate, burst pressure, EC voltage response as a function of crack size and morphology.

I (9)

Conduct physical and mechanical property tests of tubes repaired by laser welding.

(10) Conduct SCC susceptibility tests, crack growth (fatigue and SCC) tests

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and corrosion fatigue tests of laser welding repaired tubes.

r NEED FOR WORK AND APPLICATION:

This work is needed to validate or update models and correlations for predicting crack growth and morphology, leak rates and burst pressures from service degraded tubes.

The work is also needed for evaluating the capability

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of advanced ISI techniques and for evaluating the quality and long term serviceability 'of repaired tubing. The results are applied in predictions of tube integrity for degraded tubes left in service and for predicting the efficacy of repaired tubing in safety analyses of degraded steam generator i

tubes.

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t NRC PROPOSAL FOR TROJAN NUCLEAR PLANT STUDIES TITLE:

RPV EMBRITTLEMENT STUDIES OBJECTIVE:

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The objective of this study is to obtain samples of the RPV welds and base l

metals both in the beltline and in the regions of the vessel as far below the l

core as practical, for the purposes of:

(1) determining the level of embrittlement of these materials for comparison to existing embrittlement l

correlations; (2) determining the through-wall fluence and irradiation damage attenuation; and (3) performing embrittlement mechanisms studies.

l SCOPE OF WORK:

Using cutting tooling developed for the THI-VIP project, remove from the inner surface of the vessel-I I

o 5-10 samples from the beltline welds, i

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5-10 samples from the base metal at the peak flux planes, o

a similar number of samples from the outer surface of the pressure vessel on the same radial line as the base metal samples from the inner surface, and o

5-10 samples from both the welds and base metal in the lower regions of the vessel.

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i Idectify the samples, and ship the samples to ORNL for testing.

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NEEP FOR WORK AND APPLICATION:

Current embrittlement models are based on RPV surveillance data, and i

independent validation comes primarily from test reactor data and from data obtained from foreign reactors.

Further, even the surveillance data used in developing the models reflect a neutron envircnment with a higher flux than that present at the RPV wall. Data that represent long-term exposure in an appropriate neutron environment are needed to validate the embrittlement models. While the Trojan reactor pressure vessel materials are not r

particularly sensitive to irradiation damage, they offer the opportunity to j

study service irradiated materials that are near one " bound" of the current i

regulatory guidance. Additionally, fluence attenuation through the wall l

thickness can be studied even for low levels of embrittlement.

Finally, examining the service-irradiated materials offers an opportunity to evaluate the irradiation damage mechanisms for a relatively low-copper, low-nickel material. The samples from regions as far below the core as practical, yet still in the same material, would provide an "unirradiated" control material, and a material that could be used in assessing the effects of thermal aging.

EXPECTED RESULTS:

The primary expected result of this project will be data that can be used in validating the embrittlement models, and through-wall attenuation model. The samples that can be obtained using the TMI-VIP cutter design will provide

i Charpy specimens from near the vessel inner surface, and corresponding specimens from the outer surface. These can provide valuable information for

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examining trends, and for the mechanisms work. They alsc can be used in evaluating embrittlement recovery after thermal annealing, and potentially i

reembrittlement trends. This would provide data at one extreme of the data i

base on annealing data.

BENEFIT OF WORK, AND USE OF TROJAN:

The benefit of the work is to provide experimental validation of the embrittlement models used in setting operational limits, and in establishing the useful life of reactor cressure vessels. The Trojan RPV materials provide an opportunity to obtain se.*vice aged and irradiated materials that will provide data for conditions in the " normal" range for key parameters so that model trends can be validated.

LIMITATIONS AND ALTERNATIVES:

i The primary limitations will come from the small sample size that can be obtained by the vessel sampling tools. While larger samples could be obtained if the vessel is dismantled, the limitation for the larger samples stems from

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the delay in obtaining the materials.

l NEED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN:

Details of the vessel operation; flux / fluence maps; all vessel materials documentation; and a report on the vessel sampling activity.

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NPC PROPOSAL FOR TROJAN AGING STUDIES TITLE:

FRACTURE TOUGHNESS OF PIPING WELDS OBJECTIVE:

To determine the fracture toughness of typical RPV piping welds and base metals.

SCOPE OF WORK:

(1)

To identify welds of interest in carbon and stainless steel piping systems that would be representative of Class 1 piping systems with diameters ranging from 6 NPS to the largest diameter piping in the primary pressure boundary.

(2)

To provide all fabrication documentation that currently exists for these welds and piping materials.

(3)

To remove the welds with at least one diameter of base material on both sides of the weld, appropriately identifying the weld, and the orientation in the piping system.

(4)

To ship the welds to a contractor (s) for testing.

NEED FOR WORK AND APPLICATION:

Two types of analyses require fracture toughness data for piping materials:

flaw evaluations under Section XI, and leak-before-break evaluations to gain relief under GDC-4. The flaw evaluation procedures under Section XI employ stress multipliers that were determined using a very limited fracture toughness data base.

Further, that dats base has not increased significantly in the last several years, and has virtually no new data from LWR piping systems. This forces conservatism into the generic analyses and severely limits more realistic evaluations.

Similarly, the data base of typical piping material properties is not very large, which necessarily introduces conservatism into leak-before-break fracture analyses. Obtaining more data from actual LWR piping and piping welds wculd contribute to the overall data base, and to statistical evaluations of fiacture toughness for generic classes of materials.

EXPECTED RESULTS:

The expected results would be fracture toughness and tensile data for several heats of both carbon and stainless steel piping materials, and welds in those materials. These data would permit examination of diameter and thickness effects, which are related to material processing history and which can affect fracture toughness.

BENEFIT OF WORK:

l The primary benefit of the work would be to expand the data base of typical fracture toughness data for piping. This would contribute to reductions in conservatism in regulatory and Code analyses owing to the very limited current data base.

LIMITATIONS AND ALTERNATIVES:

There is one principal limitation: the level of contamination of the primary pressure boundary piping. However, this limitation would apply to all decommissioned plants.

There may be other alternatives for obtaining piping material from other decommissioned plants, but there is no good alternative for obtaining typical data other than testing materials obtained from actual nuclear power plants.

DUPLICATION OF EFFORT:

None.

NEED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN:

Trojan would need to provide all available documentation concerning the fabrication practices, mil certifications, etc., for the piping systems. This information would be used to document the materials and to justify use of the data in evaluating other nuclear piping systems.

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f NRC PROPOSAL FOR TROJAN NUCLEAR PLANT STUDIES TITLE:

Fracture Toughness of Aged Cast Stainless Steel Components OBJECTIVE: To provide validation of the fracture toughness models using service aged materials.

SCOPE OF WORK:

(I)

Identify portions of the primary pressure boundary, or any other components exposed to the primary system operating temperature, that were fabricated from cast duplex stainless steel; centrifugally cast and statically cast components are of interest.

(2)

Collect all currently existing fabrication data -- chemistry, fabrication practice, mill certifications, etc.

(3)

Reinove an appropriate section of the component, identifying orientation within the larger component or system as appropriate. Examples of an

" appropriate" section include: a ring of pipe at least 2 diameters long; an entire valve body for valves less than 12 inch; sections of pump casings large enough so that at least IT-CT fracture toughness specimens could be fabricated.

(4)

Ship the materials to a contractor (s) for testing.

NEED FOR WORK AND APPLICATION:

Models have been developed for predicting the fracture toughness of service aged cast stainless steel components. Those models were based largely on laboratory aged materials.

Service aged components were obtained from the German KRB reactor and from the Shippingport reactor. However, the service times and temperatures for these components were not sufficiently long or high to produce any appreciable toughness loss. Thus, data from service aged materials still are needed to finally validate the models.

The data are needed for use in fracture mechanics evaluations that are part of flaw evaluations, leak-before-break evaluations, and remaining life evaluations that may be appropriate to justify license renewal applications.

EXPECTED RESULTS:

The expected results will be fracture toughness data that can be used in validating the empirical models. Additionally, the actual fracture toughness data could be useful in plant-specific evaluations.

BENEFIT OF WORK AND USE OF TROJAN:

The benefit of the work would be validated fracture toughness models that would permit use of the models with greater confidence, and lower margins.

The Trojan materials offer the opportunity to validate the models using longer service exposure times.

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LIMITATIONS AND, ALTERNATIVES:

The only serious limitation comes from the contamination on the primary I

pressure boundary materials. This could be eliminated or reduced by decontamination procedures.

l The alternative to the Trojan materials for long-term service exposure is to obtain them from other decommissioned reactors if and when they may become available.

DUPLICATION OF EFFORT:

i None NEED AND DESCRIPTION OF DOCUMENTATION FROM TROJAN Fabrication records, mill certifications, service records (time, temperature),

etc. would be needed from Trojan to provide documentation of the applicability of the materials and data.

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