ML20028E868

Forwards Addl Info Re Reactor Vessel Matl Surveillance. Thermal Monitors,Absence of Neutron Dosimetry,Matl Sampling Program,Specimen Withdrawal Schedule & Location of Surveillance Capsules Discussed

ML20028E868
Person / Time
Site:	Clinch River
Issue date:	01/27/1983
From:	Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:	Grace J Office of Nuclear Reactor Regulation
References
HQ:S:83:200, NUDOCS 8301280249
Download: ML20028E868 (14)

Text

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i Department of Energy l Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:200 JAft 2 71933 l

Dr. J. Nelson Grace, Director  :

CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ,

Washington, D.C. 20555 l

Dear Dr. Grace:

ADDITIONAL INFORMATION ON REACTOR VESSEL MATERIAL SURVEILLANCE i l

Enclosed are responses to Nuclear Regulatory Commission (NRC) staff questions concerning material surveillance in the Clinch River Breeder Reactor Plant reactor vessel. The enclosed information was discussed with the NRC reviewer on December 28, 1982, and in subsequent telecons.

Questions regarding the enclosure may be addressed to Mr. D. Edmonds (FTS 626-6157) or Mr. D. Robinson (FTS 626-6098) of the Project Office Oak Ridge staff.

Sincerely, l

J n R. Longen er i Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy Enclosure l

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4 NRC Concern and/or Defined Resolution:

Thermal monitors (There is an) absence of any plan for thermal monitors in the surveillance capsules or elsewhere. Project should discuss why they are not needed.

Response

Thermal expansion type monitors will be used in some of the surveillance capsules, based on the good performance exhibited in testing at HEDL. This particular type of monitor is not very sensitive to brief transient thermal spiking, and thus provides a better description of the temperature history of a capsule. In addition, sodium temperatures will be continuously monitored with thermocouples above and around the core and at the core inlet. Thus, heat transfer calculations will provide information on the specimen and component thermal environments.

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,. NRC Concern and/or Defined Resolution:

Absence of neutron dosimetry The absence of any mention of neutron dosimetry. Dosimeters are required to benchmark the calculations of neutron fluence and energy spectra at the reactor vessel and the structural elements of the internals. These need to be located as close to the critical areas as possible. These areas include welds, geometric discontinuities, and other highly stressed regions. The PSAR mentions a lead factor of 3 in this connection, which is reasonable, but it is also important to have dosimeters that will provide a check on a calculated value of i

azimuthal axial and radial gradients in neutron flux. Dosimetry is also required in the surveillance capsules that contain materials I

samples, and in some cases, these will also provide a check on.the neutron transport calculations at critical welds, etc.

Response

All surveillance capsules will contain dosimeters to indicate the fluence received by the specimens. The details of the program are yet.

to be finalized, but it is probable that capsules due to be withdrawn early in plant life will contain as many dosimeters as are considered necessary to establish the neutron spectrum. Capsules removed later will probably have fewer dosimeters and will be used to determine the fluence only. Similar plans are in place for monitorir.g the neutron

spectrum and fluence for components throughout the reactor system.

The CRBRP reactor characterization program (RCP) is part of an initial testing program concerned with determination of the neutron i

environment in the reactor enclosure (including the reactor vessel).

The initial reactor characterization measurements are part of the overall CRBRP Acceptance Test Program.

The RCP will consist of measurements of the neutron flux and spectra in the Fuel Transfer and Storage Assembly (FT & SA) for assurance of the reactor vessel environment and support cone environment.

g NRC Con'ern c and/or Defined Resolution:

! The material sampling program The fluence on the reactor vessel is predicted to be less than 1021 n/cm2 , the criterion for inclusion in surveillance. However, there are enough uncertainties in the fluence estimates and in knowledge of the effects of long term exposure of materials to warrant inclusion of reactor vessel materials including welds in the program.

In fact, there should be a high priority effort to collect (and establish traceability for) generous samples of all structural 3

materials from the reactor vessel beltline and adjacent regions that could be subject to significant neutron radiation and from the

! structural elements of the reactor internals. The plan outlined in Paragraph 5.2.4.5.1 provides for only 1/4 inch diameter tensiles.

Material for larger specimens, up to one inch thick, compact tension 1

specimens should be provided. The specimen types put in the surveillance capsules must be determined by analysis of the most probable modes of failure that would be aggravated by neutron embrittlement. As listed in Criterion 12 in the PSAR, these are ductile rupture, creep fatigue and, if the fluence is high enough to cause swelling, loss of function due to excessive deformation. Thus, the first step is to provide such an analysis. The second step is to collect materials matching those in the vessel in every possible detail, including weld wire heat numbers, weld flux lots and the weld procedure itself.

j Response

' Althoughtheestimatedtotagneutgonfluenceatthereactorvesel beltline is about 6.0 x 102 n/cm , the surveillance program will  :

include specimens to monitor the performance of reactor vessel material, including weldments. Same is planned for reactor internal components but are not identified in Paragraph 5.2.4.5.1 because this section is concerned only with reactor enclosure system components.

Generous samples of materials for all components will be collected to provide for surveillance specimens and for archival purposes.

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NRC Concern and/or Defined Resolution Specimen withdrawal schedule The withdrawal schedule calls for the first withdrawal at one-fourth the vessel life. It is desirable that the first capsule be withdrawn within 2 or 3 years after startup to provide an early check on the neutron transport calculation so that any technical surprises in that regard or in regard to the sensitivity of the material to neutron radiation be discovered early in life. Other capsules should have load factors of approximately one, so they can be left in the reactor for more than half its life in order to track such things as changes in fuel loading.

Ref.,Jnac Contingency specimens exist which could allow for withdrawal schedule changes. Within 2 or 3 years after startup, the level of irradiation damage at such low fluences is estimated to be of the magnitude whereby mechanical property changes will remain within the normal scatter band for the unirradiated material and interpretation of the data, therefore, will be of little significance and inconclusive.

Consideration is being given, however, to adjusting the withdrawal schedule of surveillance capsules for some reactor internal components, thus enabling an early check on the neutron transport calculations.

4 NRC Concern and/or Defined Resolution:

Location of surveillance capsules The location of surveillance capsules is described in Paragraph 5.2.4.5.1 as being in the Removal Radial Shields or the Fuel Transfer and Storage Area. There are serious questions about whether the temperature of irradiation and the flux rate at the proposed capsule locations match those of the regions of concern in the vessel.

Response

Every effort has been made to provide a reasonable match between the surveillance capsule conditions and component operating conditions.

Highest priority is being given to average neutron energy matching, in order to avoid specimens receiving a disproportionate high energy flux which could produce irradiation ef fects not actually present in the component. The guideline for the maximum flux at the cpecimen is that

it must not exceed three times the component flux. This factor is intended to limit the damage rate in the specimen in order to prevent override of the recovery processes. Irradiation temperature matching is of a lower priority provided that the specimen temperature does not

exceed that of the component.

Surveillance specimens for investigating irradiation ef fects will be placed inside the Fuel Transfer and Storage Assembly, that is shown in Figure 4.2-36 of the PSAR.

l The following compares component conditions with surveillance conditions (in the FT&SA) for the beltline, as determined from recent calculations:

Average neutron component 0.01; specimen 0.02 energy (MeV) 11 Totalneutgon component 3.8 x 1011 ;

flux (n/cm -sec) specimen 8.7 x 10 Temperature upper component 700; specimen 700 limit (OF)

These clearly represent the best match possible for the vessel beltline. Additional information on neutron flux and energy is given in the following paragraphs.

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REACTOR SYSTEM NEUTh0N FLUX AND ENERGY Figures 5.1.1-1 and 5.1.1-3 are reactor system neutron flux and mean neutron energy, respectively. Displacements per atom RZ distributions are available but not plotted. Reactor vessel and core support structure support cone dpa are tabulated in Table 1.

Table 1 defines the reactor vessel and support cone irradiation

, environment. Flux, fluence and dpa data are presented. Flux, fluence and dpa values are taken directly from RZ analyses. Azimuthal variation of the flux at the reactor vessel and support cone are negligible. Radial thickness of the regions outside the fuel region are (1) radial blanket = 9.4", (2) removable radial shield = 16.8",

(3) fixed radial shield = 7.75", (4) core barrel = 2", and (5) sodium pool = 43.75".

Table 2 is the, axial radiation environment at the FT & SA. Currently, no other facility is available for surveillance of the RV. Radial location and surveillance envelope at the FT & SA are shown on rigures 5.1.1-1 & 5.1.1-3.

The dpa of the reactor vessel at the beltline is 0.003 max, and at the support conc is 0.0001 max. The dpa is based on 22.5 full power years i at 975 MwT. Figure 14 from HEDL-TME-82-17 indicates that no loss in 1 fracture toughness occurs below 1 dpa. Figure 14 suggests that irradiation damage can be separated into three regions: a threshold exposure below which there is no loss in toughness, intermediate exposures (1 to 10 dpa) where toughness decreases very rapidly with exposure, and finally a saturation range in which increasing exposure does not produce further reduction in toughness. The design margin-for the RV is 1 dpa on[ set /0.003 = 333, the design margin for the support cone is:

1 dpa onset

= 10,000 0.0001 i The loss in RV and support cone fracture toughness is considered negligible.

Surveillance of the reactor vessel materials (base metal and weldment)

! would best be achieved by placement of test specimens at the top of 4

the FT & SA surveillance envelope (surveillance dpa/ component dpa =

factor of 2). This allows a withdrawal at 71/2 years (8) to-equal reactor vessel exposure of 15 years without excessive rate effects.

Also 11 years = 3/4 exposure and 15 years = fuel exposure. A fourth capsule is available for removal at less than 8 years if considered necessary. No irradiation ef fect is expected to occur at the reactor vessel after 30 years of operation.

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The vessel surveillance program will use materials representative of the reactor vessel and core cupport cone. The support cone fluence is a factor of 10 lower than that of the RV. The support cone dpa is a factor of 30 lower.

REACTOR VESSEL AZIMUTHAL FLUX VARIATION The permanent structures radiation environment is represented by combining two-dimensional RZ modeling results together with hexagonal model results defining azimuthal factors and neutron streaming factors. A plan view of the reactor core to core barrel (CB) region-is provided in Figure 2.0-2 (attached). CB neutron flux azimuthal variation and Fixed Radial Shield (FRS) streaming factors are indicated on Figure 2.0-2. The core barrel (O.D.) maximum azimuthal factor, including streaming, is 1.3 (1.03 x 1.2 5) adjacent to the FRS clearance gap at 289 0. The azimuthal variation at the RV will be negligible due to neutron attenuation and scattering in the 43.75" sodium region outboard of the CB.

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