ML071410299

Purdue University - Submittal of Response to RAIs for Conversion of PUR-1 (Redacted)

ML071410299
Person / Time
Site:	Purdue University
Issue date:	05/03/2007
From:	Joel Jenkins Purdue University
To:	Alexander Adams NRC/NRR/ADRA/DPR/PRTA
References
Download: ML071410299 (128)

Text

PURDUE UNIVERSITY RESEARCH REACTOR LICENSE NO. R-87 DOCKET NO. 50-182 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR THE CONVERSION OF THE PURDUE UNIVERSITY RESEARCH REACTOR FROM HEU TO LEU FUEL REDACTED VERSION SECURITY-RELATED INFORMATION REMOVED REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS

PURDUE UNIVERSITY SCHOOL OF NUCLEAR ENGINEERING 3 May 2007 Mr. Alexander Adams, Senior Project Manager Research and Test Reactors Branch A US Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD, 20852-2738

SUBJECT:

Docket No. 050-0182, Facility License R-87 Response to RAIs for Conversion of PUR-1, ML070680273

Dear Mr. Adams:

Enclosed please find the response to the Request for Additional Information (ML070680273). Included in this package are one unbound and four bound copies of the response, and one unbound and four bound copies of the specifications for the new reactor fuel. Digital copies of these documents are also included on an enclosed CD.

Should you have any questions, or require further information, please contact me at 765.496.3573, or jere@purdue.edu.

I declare of penalty of perjury that the foregoing and the original submission of the conversion proposal are true. Executed on this day of 3 May 2007.

Purdue University Nuclear Engineering

Enclosures:

As stated A.o2 0

.4. School of Nuclear Engineering Nuclear Engineering Building 0 400 Central Drive 0 West Latayette, IN479U0-2U1 7 (765) 494-5739 o Fax: (765) 494-9570 a https://engineering.purdue.edu/NE

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION FOR THE CONVERSION OF THE PURDUE UNIVERSITY RESEARCH REACTOR FROM HEU TO LEU FUEL DOCKET NUMBER 50-182 FACILITY LICENSE NO. R-87 3 May 2007 Submitted By:

Jere Jenkins, Director of Radiation Laboratories Ed Merritt, Asst. Director of Radiation Laboratories/Reactor Supervisor Purdue University Research Reactor School of Nuclear Engineering College of Engineering Purdue University West Lafayette, IN

TABLE OF CONTENTS Q u e stio n 1 ...................................................................................................................... 3 Q u estio n 2 ...................................................................................................................... 3 Qu estio n 3 ............... I...................................................................................................... 3 Q uestio n 4 ...................................................................................................................... 3 Q u e stio n 5 ...................................................................................................................... 4 Q u e stio n 6 .................................................................................................................... 14 Q uestio n 7 .................................................................................................................... 14 Q uestio n 8 .................................................................................................................... 15 Q uestio n 9 .................................................................................................................... 15 Q uestio n 10 ................................................................................................................... 15 Q uestio n 11 ................................................................................................................... 16 Q uestio n 12 ................................................................................................................... 16 Q uestio n 13 ................................................................................................................... 17 Q uestio n 14 ................................................................................................................... 19 Q uestio n 15 ................................................................................................................... 21 Q uestio n 16 ................................................................................................................... 24 Q uestio n 17 ................................................................................................................... 24 Q uestio n 18 ................................................................................................................... 24 Q uestio n 19 ................................................................................................................... 24 Question 20 ................................................................................................................... 25 Q uestio n 2 1 ................................................................................................................... 29 Q ue stio n 22 ................................................................................................................... 29 Q u e stio n 23 ................................................................................................................... 29 Q ue stio n 24 ................................................................................................................... 30 Q ue stio n 25 ................................................................................................................... 30 Q ue stio n 26 ................................................................................................................... 30 Q ue stio n 27 ................................................................................................................... 31 Q ue stio n 28 ............................ ....................................................................................... 42 Q ue stio n 29 ........................................................................................... *........................ 43 Q ue stio n 30 ................................................................................................................... 43 Q ue stio n 3 1................................................................................................................... 44 Question 32 ................................................................................................................... 45 Question 33 .................................................................................................................... 46 Question 34 ................................................................................................................... 47 Question 35 ................................................................................................................... 47 Question 36 ................................................................................................................... 47 Question 37 ................................................................................................................... 48 Q ue stio n 38 ................................................................................................................... 49 Q ue stio n 39 ................................................................................................................... 49 Q ue stio n 40 ................................................................................................................... 53 Q ue stio n 4 1............................................................................................. ...................... 54 Question 42 .................................................................................................................... 54 APPENDIX 1: ................................................................................................................ 55 2007 RAI responses PUR-1 2/63 2 May 2007 18:23

REQUEST FOR ADDITIONAL INFORMATION PURDUE UNIVERSITY RESEARCH REACTOR DOCKET NO. 50-182 3 May 2007 Question 1

1. The regulations in 10 CFR 50.30(b) require applications, such as your application for conversion, to be made under oath or affirmation. Please provide a statement that your application of August 13, 2006, is made under oath or affirmation.

Response

A statement of oath and affirmation will be submitted with these answers that will cover both this document and the original conversion proposal (CP).

Question 2

2. Table 4-1 and Figure 4-35. The figure shows the plate locations with a single dummy plate and 13 fueled plates. What is the distribution within the fuel assembly when two dummy plates are used? The figure shows nine standard fuel assemblies with 13 plates and four with 12 plates. This brings the total of plates to 189, not 190 as stated in Table 4-1 and on page 22. Please explain.

Response

This is a typographical error. Fig 4-35 for position E2 should show 13 fuel plates instead of 12. Since the fuel assembly design has been modified (ref. Q5), the new model will be discussed in the answer to that question.

Question 3

3. Table 4-3. What is the basis for picking the width of the fuel meat for the LEU fuel plates (59.6 mm with a range between 58.9 mm and 62.7 mm)?

Response

It is not half way in between 58.9 mm and 62.7 mm, but this was the value used for the University of Florida case and was carried forward. This should have no effect on these analyses, since the critical factor is the U-235 mass. If the average fuel meat width of 60.8 mm were used, the power density would be slightly lower and all of the safety margins slightly increased from those shown in the conversion proposal.

Question 4

4. Section 4.2. Proposed Technical Specification (TS) 5.2.2 says that the LEU assembly will have up to 185 g of U-235. Please explain why that is different than 12.5 g/plate times a maximum of 14 plates/assembly as specified in Section 4.2.

Response

The specifications for the U-235 loading of the LEU fuel plates is 12.5+/--0.35 g. With 14 plates per assembly, the nominal loading is 175 g with an allowed range of 170.1 to 179.9 g. The proposed TS 5.2.2 should be 180 g of U-235.

2007 RAI responses PUR-1 3/63 2 May 2007 18:23

Question 5 5.

A. Section 4.2.1. Please provide a copy of Reference 5.

B. Our understanding is that the final design of the Purdue fuel elements was still underway when the conversion Safety Analysis Report (SAR) was submitted to NRC. Please verify that the description of the fuel in the SAR is accurate or discuss any fuel design changes made since the SAR was written.

C. Our understanding is that the fuel plates and fuel boxes are being fabricated by different vendors and that final assembly of the fuel elements will be performed at Purdue. Please describe the quality assurance requirements the Department of Energy employs at the fuel component vendors to ensure that the fuel element components are consistent with the SAR.

D. Please verify that approved procedures will be used at Purdue to assemble the fuel elements.

E. Please describe the quality assurance requirements to be employed at Purdue to ensure that the assembled fuel elements are consistent with the SAR.

F. Please discuss how the fresh fuel plates will be stored and handled during fuel element assembly such that the requirements of TS 5.3 are met.

Response

A. A copy of the fuel specification will be provided with the response to these questions.

B. The old LEU fuel element and fuel assembly designs are described in Sections 4.2.1.2 and 4.2.1.3, respectively, of the August 2006 Conversion Proposal (CP). This LEU design used fuel plates bolted at the top and bottom to form a fuel element. The fuel element in this design was inserted into a smooth-walled aluminum box to form a fuel assembly.

The changes that occurred after the submission of the conversion proposal were to the assembly cans, and the method of inserting the plates into the cans. Specifically, the plates in the new LEU fuel assembly design slide into wall spacers affixed to two sides of the fuel can, as shown in Figures Q5B-1 to Q5B-7. The nominal plate-to-plate spacing was slightly reduced, and the nominal plate-to-wall spacing was slightly increased, as shown in Table Q5B-1. An updated version of Table 4-1 from the August 2006 Conversion Proposal that reflects the changes to the assembly and core design is also included below. It should be noted that the design of the fuel plates and dummy plates was not changed as a result of the design change.

Table Q5B-1: Channel Types and Thickness in PUR-1 Assemblies (Ref: Table 4-23 in CP)

Plate-to-plate (mils) Plate-to-wall (mils)

Standard Control Standard Control HEU 207 207 160' 160' OLD LEU 147 197 79 79 NEW LEU 144+/-15 181+/-15 127+/-8 127+/-8 1 This is the smaller of the two bolt heads on the HEU elements.

2007 RAI responses PUR-1 4/63 2 May 2007 18:23

A new Table 4-1 from the CP is shown below, reflecting the changes in the fuel spacing.

Table 4-1: Summary of Key Nominal Design Parameters of HEU (current) and LEU (Exnected* Cores (Ref: Table 4-1 in CP1.

DESIGN DATA HEU OLD LEU NEW LEU Design Design FuelType MTR Plate MTR Plate MTR Plate Fuel "Meat' Composition U-Al Alloy U3Si2 -AI U3 Si2-AI Fuel Enrichment U-235 (nominal) 93% 19.75% 19.75%

Mass of U-235 per plate (g, nominal)

Fuel Meat Dimensions 12.5 12.5

]

Width (mm) 62.7 59.6 59.6 Thickness (mm) 0.508 0.508 0.508 Height (mm) 600.1 600.1 600.1 Fuel Plate Dimensions Width (mm) 70.2 70.2 70.2 Thickness (mm) 1.52 1.27 1.27 Height (mm) 638.6 638.6 638.6 Cladding Composition 1100 Al 6061 Al 6061 Al Cladding Thickness (mm) 0.508 0.381 0.381 Dummy Plate Composition 1100 Al 6061 Al 6061 Al Dummy Plate Dimensions Same as Fuel Same as Fuel Same as Fuel Standard Fuel Assemblies Number of standard assemblies 13 13 13 Number of plates per standard assembly 10 14 14 Control Fuel Assemblies Number of control assemblies 3 3 3 Number of plates per control assembly 6 8 8 Total plates in core (fuel and dummy) 148 206 206 Fuel plates in core (current, expected) 124 190 191 Dummy plates in core (current, expected) 24 16 15 Plate spacing in standard assemblies (mm) 5.26 3.71 3.66 Plate spacing in control assemblies (mm) 5.26 5.00 4.60 The calculated reactivity change as a result of the plate spacing change was -0.05+/-0.07 %Ak/k.

The fuel orientation in the standard assemblies was also modified by a 90° rotation (making the plates parallel to the control assembly plates) in order to change the fuel direction relative to the handle such that that the plates would remain caged in the fuel box, as shown in Fig. Q5B-8. The HEU plates, and the original LEU plates, were bolted together. The new handle orientation is now normal to the plates to restrict their possible movement. The calculated reactivity change as a result of the fuel orientation change was -0.199+/-0.03 %Ak/k. These two changes together resulted in the replacement of one dummy plate with a fuel plate in the model. Overall we believe that this design is superior to the HEU and previous LEU designs, since the additional spacers will provide a smaller uncertainty in the spacing of the plates over their entire length and result in less possible variation in the channel width.

2007 RAI responses PUR-1 5/63 2 May 2007 18:23

Figure Q5B-1: New Standard LEU fuel assembly. (Replaces Figure 4-5 in CP)

Figure Q5B-2: New Control LEU fuel assembly. (Ref Fig. 4-7 in CP) 2007 RAI responses PUR-1 6/63 2 May 2007 18:23

Figure Q5B-3: New Standard LEU assembly can detail, wall spacers.

Figure Q5B-4: New Control LEU assembly can detail, wall spacers.

2007 RAI responses PUR-1 7163 2 May 2007 18:23

Figure Q5B-5: New standard LEU assembly can spacer detail.

2007 RAI responses PUR-1 8/63 2 May 2007 18:23

Figure Q5B-6: New control LEU assembly can spacer detail.

2007 RAI responses PUR-1 9/63" 2 May 2007 18:23

Figure Q5B-7: Model representation of standard LEU assembly plate spacing detail, showing wall spacers.

2007 RAI responses PUR-1 10/63 2 May 2007 18:23

Figure Q05B-8: LEU core model. (Replaces Figure 4-20 in-CP)

C. The fuel and fuel assembly cans are being purchased by the Idaho National Laboratory, and must meet INL qualification and certification. A summary of the INL procurement qualifications is.

provided below. Close communication between Purdue University and the Idaho National Laboratory has been maintained through the design process. This communication ensures that the appropriate technical and functional requirements from the reactor safety basis are carried forward to the fabrication drawings and specifications for the new reactor components. The drawings and fabrication specifications denote the dimensions and other design parameters that must be met for the item to be in compliance with the reactors safety basis. These documents are used to convey the requirements to the INL procurement personnel, the vendor, and QA Engineers. The Project then relies on the well established QA processes at the INL to ensure that the final product meets the requirements per the drawings and specifications.

The INL meets or exceeds the requirements for procuring items and services as established by the Department of Energy (DOE). These requirements are contained in 10 CFR 830 Subpart A, Quality Assurance Requirements-,DOE Order 414.1 C, Quality Assurance; and NQA-1 -2000, Quality Assurance Requirements for NuclearFacilityApplications. These requirements establish the methods that the INL must use to procure items and services. More specifically, the process requirements relevant to the procurement of items for the Purdue reactor are:

2007 RAI responses PUR-1 11/63 2 May 2007 18:23

0 Items and services shall be procured to meet established requirements and perform as specified. [DOE Order 414.1C, Attachment 2, 3.g. (1)] [10 CFR 830.122 (g) (1)]

• Prospective suppliers shall be evaluated and selected on the basis of specified criteria.

[DOE Order 414.1C, Attachment 2, 3.g. (2)] [10 CFR 830.122 (g) (2)]

• Processes shall be established and implemented to ensure that approved suppliers continue to provide acceptable items and services. [DOE Order 414.1C, Attachment 2, 3.g. (3)] [10 CFR 830.122 (g) (3)]

To complete the process, final inspection of the items will be performed by qualified INL Inspectors at the vendor before shipment. Inspections of the items will also be performed by Purdue upon receipt at the reactor.

D. Procedures for assembly of the fuel elements will be written and certified by the Facility Director, the Reactor Supervisor, and the safeguards and oversight committee (CORO) for PUR-1. All fuel handling will be done under the supervision of USNRC licensed Senior Reactor Operators.

Accurate records of fuel disposition will be maintained.

I Fiaure 05D-1: Fuel plate Figure Q5D-2: Dummy plate, Loading of the fuel assemblies prior to the initial assembly of the core and the initial approach to critical will be according to the preliminary loading plan. LEU plate serial numbers will be recorded and verified by two SROs independently, and matched to the uniquely identified fuel assembly cans. All placements of assemblies will again be independently verified by two SROs, 2007 RAI responses PUR-1 12/63 2 May 2007 18:23

and accurate records maintained. During all of these operations, compliance with Technical Specifications for the PUR-1 reactor will be verified.

E. There are 13 standard elements and 3 control elements in the LEU core. A standard element can contain up to 14 fuel plates, and the control elements can contain up to 8 fuel plates. The configuration analyzed in the August 2006 Conversion Proposal had 10 standard elements with 13 fuel plates, 3 standard elements with 12 plates, and 3 control elements with 8 fuel plates - or a total of 190 fuel plates. The new LEU core design is modeled with 11 standard fuel elements with 13 fuel plates, 2 standard elements with 12 plates, and 3 control elements with 8 fuel plates - or a total of 191 fuel plates, This is the configuration that is planned at startup, and adjustments to the number of plates per assembly will be made at startup to build the initial critical core and the working core.

Fuel and dummy plate locations will be tracked through each step of the initial loading and approach to critical following approved procedures as discussed in the answer to Q5D. During the insertion of fuel and dummy plates into the assembly cans, the fuel plates will be identified by serial number on forms associated with each fuel element. The location of individual plates will be specifically tracked, and verified by two SROs on the form. As the approach to critical process continues and fuel plates and dummies are relocated, new locations of plates and dummies will be tracked on forms for each assembly.

The location of fuel and dummy plates in the initial critical core will likely change as plates are added to obtain a final core load that satisfies the TS requirements and the operation parameters of the conversion proposal. At each step of this process, fuel assembly forms will be modified to reflect the new plate locations, and will again be independently verified by two SROs. These records will be maintained for the duration of that core load, or the lifetime of PUR-1.

F. Upon receipt of the LEU fuel, it will be stored in the 6M drums in a secure location. One bundle of fuel plates will be removed at one time from the 6M shipping containers and examined for quality assurance with appropriate procedures. Upon completion of the inspection of the bundle of plates, they will be moved to a secure storage area within the facility.

Racks specifically constructed for the purpose of storage of the plates will contain each bundle. A criticality analysis of the storage facility was performed with MOCNP. The model is shown in Figure Q5F-1.

Figure Q5F-1: MCNP model of dry storage facility.

2007 RAI responses PUR-1 13/63 2 May 2007 18:23

With all of the plates moved to that new location, the calculated eigenvalue of the storage rack is 0.015-0.0003. Should the storage flood with water, the eigenvalue becomes 0.26+/-0.004. In both of these conditions, the TS limit of ken<0. 8 is met. Parametric analyses were performed with spacing of the plates within the tubes, and the above listed cases are the worst case. All movement of fuel will be performed under the direction of licensed senior reactor operators.

Question 6

6. Section 4.2.5. What is the effect on reactivity and power distribution of replacing graphite in the reflector with an aluminum sample holder?

Response

The aluminum sample tubes in the irradiation facility are filled with graphite unless they are being used.

Replacing the graphite in the six sample tubes with air, as would be expected if samples were inserted, reduces the calculated eigenvalue of the LEU core from 1.00731+/-0.0002 to 0.996+/-0.0002 without any reactivity bias, but with the estimated bias of 0.32%, these values would be 1.0041 and 0.993, respectively.

The effect of replacing graphite in the irradiation facility with an aluminum sample holder on the reactor power distribution used in the thermal-hydraulic analyses is well covered by the global hot channel factor of 1.5 for the reactor power measurement uncertainty used in all the reported analyses. The reactor power hot channel factor was used in addition to (1) the radial power factor (used to account for plate-to-plate power variation), (2) the factor used to account for power density variation along the width of the hot plate, and (3) the axial power profile. While replacing the graphite in the irradiation facility with an aluminum sample holder will cause a change in the reactor power distribution, this change will be small compared with the hot channel factors already applied in the thermal-hydraulic analysis. Therefore, no additional thermal-hydraulic analysis is needed.

Question 7 7.

A. Table 4-6. The measured to calculated eigenvalues for the excess reactivity of a fresh HEU core are compared. This difference establishes the claim that the model introduces a 0.32% bias to the calculation of the eigenvalue. What are the uncertainties for the measured and calculated eigenvalues?

B. What is the justification to use this 0.32% bias for the LEU core? Note that if the bias for the LEU core is not used (or is significantly less than 0.32%) the excess reactivity exceeds the TS-limit of 0.6%.

Response

A.

There is no estimate available for the uncertaintyof the measured value.

Table 4-6: Calculated and Measured excess reactivity for fresh HEU core.

(Replaces Table 4-6 in the CP)

Measured Eigenvalue I Calculated HEU Eigenvalue I Bias (% Ak/k) 1.00431 1.00753+/-0.00015 0.32+/-0.015 The uncertainty on the bias is based on the uncertainty calculated by MCNP.

B.

This is the best available estimate for what the bias will be in the LEU core model. Until the core is assembled, the actual bias cannot be known. A careful approach to criticality will be used to ensure that 2007 RAI responses PUR-1 14/63 2 May 2007 18:23

all Technical Specifications are met. Upon achieving initial criticality, adjustments to the final plate loading will be performed while maintaining sufficient margin to the TS limit of 0.6% excess.

Question 8

8. Table 4-7. The title of Table 4-7 does not reflect the contents of the table. The table is a tabulation of calculated eigenvalues at five measured critical rod positions. It does not contain any calculated critical rod positions. Please correct.

Response

A corrected table 4-7 is shown below.

Table 4-7: Calculated eigenvalues at measured critical rod positions for HEU core (Replaces Table 4-7 in CP)

RR Position SS-2 Position SS-1 Position Case (cm) (cm) (cm) Eigenvalue Bias %Ak/k 1 64.12 43.60 64.12 1.00305+/-0.00015 0.30+/-0.015 2 64.12 64.12 49.68 1.00285+/-0.00015 0.29+/-0.015 3 51.93 51.92 53.19 1.00327+/-0.00013 0.33+/-0.013 4 1.89 54.40 64.12 1.00340+/-0.00014 0.34+/-0.014 5 31.44 48.47 64.12 1.00327+/-0.00014 0.33+/-0.014 Question 9

9. Table 4-8. In order to understand the differences between calculation and measurement quoted in the table, provide the uncertainties for both the measured and calculated control rod worths, if available. This will also clarify whether the last column in the table is really an "error" or more appropriately a "relative difference." Provide uncertainties for all values, and correct header of last column to relative difference.

Response

A corrected Table 4-8 is shown below.

Table 4-8: Comparison of measured and calculated control rod worths for HEU core (Replaces Table 4-8 in CP)

Measured Value (ak/k)" Calculated Value (ldkk) Relative Difference Shim Safety 1 0.0450 0.0436+/-0.0002 -3.11%

Shim Safety 2 0.0252 0.0235+/-0.0002 -6.74%

Regulating Rod 0.0028 0.0027+/-0.0002 -3.57%

No uncertainties are available for measured quantities.

Question 10

10. Section 4.5, page 17. It is stated that in the LEU model there was an addition of 20 ppm of boron-equivalent to the 6061 cladding material. This section also states that the 6061 aluminum assembly cans had a 10 ppm boron content. Why were different boron contents assumed for the same aluminum alloy?

Does this difference impact the calculated results, i.e., what is the reactivity worth of 10 ppm boron?

2007 RAI responses PUR-1 15/63 2 May 2007 18:23

Response

The natural boron equivalent content of the aluminum used to manufacture the cladding for MTR fuel assemblies has increased since the 1960s due to increased use of boron carbide crucibles in the aluminum industry. The PUR-1 HEU fuel was manufactured in 1960. The 1100-Al cladding was assumed to have an impurity content of 10 ppm natural boron equivalent content because no specific data was available. Analysis of a 2005 representative sample of 6061-Al supplied to BWXT by Kaiser Aluminum had a natural boron equivalent content (excluding boron itself) of 12 ppm. Normally, analyses of boron content provide a result of < 10 ppm. Values < 10 ppm are difficult to measure. Consequently, the natural boron equivalent content of the 6061-Al used in the conversion analysis for the cladding of the PUR-1 LEU fuel and for the new 6061-Al cans was estimated to be 20 ppm.

A full core calculation was run assuming 10 ppm instead of 20 ppm natural boron equivalent in the LEU core. The increase in reactivity was 0.40% +/-0.03 Ak/k. This possible reactivity change will be compensated by adjusting the number of LEU fuel plates in the startup core.

Question 11

11. Section 4.5. What is the nominal thermal neutron flux for the PUR-1 core either on average or at experimental locations?

Response

Maximum thermal neutron flux in the fuel region of the HEU core is 2.1 El0 n/(cm2 *s), and the average thermal flux in the fuel region is 1.2E1 0 n/(cm *s)at 1kW according to the 1988 conversion proposal. The MCNP calculated values for the LEU core are 2.01 El0 n/(cm 2*s) peak in the fuel region, and 1.38E10 n/(cm 2 *s) average in the fuel region.

Question 12

12. What is the expected reactivity lifetime of the shim-safety and regulating rods?

Response

The critical position of the control rods has not changed considerably over the lifetime of the HEU core, as shown in Fig. Q12-1. The rod positions shown in the figure reflect a normalized value for the upper limit measured for SS-1.

2007 RAI responses PUR-1 16/63 2 May 2007 18:23

PUR-1 Banked Rod Critical Positions 60

-U j40 30 20 10 0

1965 1971 1976 1982 1987 1993 1998 2004 2009 Date of Measurement

---ss-1 4- ss-2 -..RR;!

Figure Q12-1: Normalized critical rod positions for banked rods.

Question 13

13. Figures 4-21, 4-22, and 4-23. These figures show calculated and measured calibration curves for SS-1, SS-2, and RR, respectively, with an HEU core, but only for rod position where there are measurements. Figures 4-24, 4-25 and 4-26 show the calculated curves for an LEU core for the full range of control rod travel. Please provide calculated curves for the HEU core with a full range of travel so they can be compared with the full-range curves for the LEU core.

Respionse:

Updated Figures 4-21, 4-22, and 4-23 are provided below.

2007 RAI responses PUR-1 17/63 2 May 2007 18:23

HEU SS-1 Rod Calibration 0.01 0.005 a

0 10 20 30 40 50 60 70

-0.005 I -0.01

-0.015 I -0.02

-0.025

-0.03

-0.035 0

-0.04 Rod Position (cm)

• MCNP a Experimental Figure 4-21: HEU control rod calibration for SS-1. (Replaces Fig. 4-21 in CP)

HEU SS-2 Rod Calibration 0.011 0.0051 6

0

) 10 20 30 5o 60 70

-0.005

-0.01

-0.015 *

-0.02 Rod Position (cm)

• MCNP I Experimental Figure 4-22: HEU control rod calibration for SS-2. (Replaces Fig. 4-22 in CP) 2007 RAI responses PUR-1 18/63 2 May 2007 18:23

HEU RR Calibration 0.005 0.0045 0.004 0.0035 0.003 0.0025 j0.0021 0.0015 o.oo 4 0.0005 1 o0 0 10 20 30 40 50 60 70 Rod Position (cm)

• MCNP5 a Experment Figure 4-23: HEU control rod calibration for RR. (Replaces Fig. 4-23 in CP)

Question 14

14. Table 4-10. The text in Section 4.5.1 indicates that Table 4-10 contains calculated and measured control rod worths while the headings in the table only refer to calculated values. Please clarify.

Respose:

Remove "and measured" from text. The correct text should read: "A comparison of the HEU and LEU calculated control rod worth values are shown in Table 4-10."

The uncertainties of the values in Table 4-10 were added for convenience. Also, Table 4-10 has been updated to include calculated control rod worths for the new LEU core design described in response to Question 5.B.

Table 4-10: Comparison of HEU and LEU calculated control rod worths. (Replaces Table 4-10 in CP)

HEU Calculated (Ak/k) Old LEU Design New LEU Design Calculated (Aldk) Calculated (Ak/k)

Shim Safety 1 0.0436+/-0.0002 0.0391+/-O.0002 0.0377:+/-0.0003 Shim Safety 2 0.0235*0.0002 0.0200*0.0002 0.0189*0.0003 Regulating Rod 0.0027-0.0002 0.0029-0.0002 0.0023-0.0003 For consistency, a comparison of the Rod Worth Curves for the three control rods are shown in Figures 014-1, Q14-2 and 014-3. These figures are related to Figures 4-24, 4-25 and 4-26 in the August 2006 Conversion Proposal, respectively. The comparison shows that the rod worth curves have been only slightly changed by the new LEU core design.

2007 RAI responses PUR-1 19/63 2 May 2007 18:23

Old and New LEU Rod Calibration Curves for SS-1 0.010000 0.00500 0.00000 10 20 30 40 1 50 60 70

-0.00500 S

-0.01000 S

-0.01500

-0.02000

-0.02500

-0.03000 -

-0.04000-Rod Position (cm)

'L*Old LEU Design U New LEU Design Fig Q14-1: Comparison of Old and New LEU Shim Safety 1 Rod Worth Curves (Ref: Fig 4-24 in CP)

Old and New LEU Rod Calibration Curves for SS-2 0.01000 0.00500 UO 0.00000 -*I 10 20 30 50 60 70 i-0.00500 M -0.01000

-0.01500

-0.02000 Rod Position (cm)

• Old LEU Design U New LEU Design Fig Q14-2: Comparison of Old and New LEU Shim Safety 2 Rod Worth Curves (Ref: Fig 4-25 in CP) 2007 RAI responses PUR-1 20/63 2 May 2007 18:23

Old and New LEU Rod Calibration Curves for RR 0.00500 0.00450 *-. .

0.00400 .....

0.00350 . 3 0.00300 W.... U S 0.00250 .. U It .02,002 W........ .....

0.00150 ..

0.00100 0.00050 -

0.00000 0 10 20 30 40 50 60 70 Rod Position (cm)

• Old LEU Design m New LEU Design Fig Q14-3: Comparison of Old and New LEU Regulating Rod Worth Curves (Ref: Fig 4-26 in CP)

Question 15

15. Table 4-11. Comparing Tables 4-2 and 4-11 there is a discrepancy in the maximum insertion rate for the shim-safety rods and the regulating rod. The insertion rates for the LEU core are exactly the same while for the HEU core there is a factor of 100 in the SS-1 and SS-2 values and a factor of 1000 for the RR. What are the correct values?

Response

Typographical and calculation errors were present in both tables. Figure 015-1 shows calculated regulating rod worth curves for the calculated and measured HEU, and the new LEU cores. The maximum speed of the regulating rod is 0.75 cm/s. Corrected tables are shown below.

As a result of changes to the design of the LEU assembly cans subsequent to the submission of the Conversion Proposal (see response to Question 5.B), control rod reactivity insertion rates and other reactor parameters have been recalculated. The updated values are included in Tables 4-11 and 4-2.

2007 RAI responses PUR-1 21/63 2 May 2007 18:23

RR Cal Curve 0.0045 y =7291E-0Sx.7.OSO-04 -

0.004 Y.56E-14eOk+. 115OC-03 0.0035 0.003 y .9220E-0Sx .S031C-04:

0.0025 0.002 cc 0.0015 0.001 0.0005 0

20 25 30 35 40 45 50 Rod Hight

• NEW LEU a HEU Measured HEU MCNP5 Linear (NEW LEU) -Linear (HEU Measured) -Linear (HEU MCNP5)

Figure 015-1: Calculated and Measured HEU and Calculated LEU Regulating Rod Curves Table 4-11: Comparison of maximum reactivity insertion rates for HEU and LEU cores (Replaces Table 4-11 in CP)

I Maximum Reactivity Insertion Rates for Control Rods (k. )i HEU Measured HEU Calculated Old LEU Design New LEU Design Calculated Calculated Shim-safety 1 1.12E-04 9.31 E-05 1.84E-04 1.75E-04 Shim-safety 2 7.45E-05 7.45E-05 1.07E-04 8.75E-05 Regulating Rod 4.16E-05 5.46E-05 4.07E-05 4.66E-05 2007 RAI responses PUR-1 22t63 2 May 2007 18:23

Table 4-2: Summary of Key Reactor Physics and Safety Parameters for the HEU (current) and LEU (Expected) Cores (Replaces Table 4-2 in CP)

REACTOR PARAMETERS Old LEU New LEU HEU HEU Design Design Measured Calculated Calculated Calculated Fresh core excess reactivity (%Ak/k) 0.43 0.43O 0.472 0.354 Shutdown margin (%Ak/k) -2.07 -1.931 -1.53; -1.58W Control rod worth (%Ak/k)

Shim-safety 1 4.50 4.36 3.91 3.78 Shim-safety 2 2.53 2.35 2.00 1.90 Regulating Rod 0.28 0.27 0.29 0.23 Maximum reactivity insertion rate , .

k S} 1.1 2E-02 9.31 E-03 1.8E01.7E2 Shim-safety 1 7.45E-03 7.45E-03 1.07E-02 8.75E-03 Shim-safety 2 4.16E-03 5.46E-03 4.07E-03 4.66E-03 Regulating Rod Avg. coolant void coefficient (k. % d 4 -1.54E-1+/-8% -1.90E-1+/-8% -1.93E-1+/-7%

Coolant temperature coefficient ( Ik 5 -1.05E-2+/-7% -9.74E-3+/-8% -9.05E-3+/-9%

Fuel temperature coefficient , 6JC 0 -7.01 E-4+/-12% -8.05E-4+/-10%

Effective delayed neutron fraction (%) 1 0.795 0.787 0.784 Neutron lifetime (ps) 1 76.7 81.3 81.3' 1 Includes a bias of 0.32%Ak/k for the HEU core, see Table 4-7: Calculated eigenvalues at measured critical rod positions for HEU core. The calculated excess reactivity is 0.75% Ak/k. The calculated shutdown margin is -1.61 % Ak/k.

2 Assumes the same bias as the HEU core. The calculated excess reactivity is 0.79% Ak/k. And the calculated shutdown margin is -1.21% Ak/k.

3 The calculated excess reactivity is 0.67%Ak/k. And the calculated shutdown margin -1.26%Ak/k.

4 Calculated for the representative range of 0-0.6% void.

5 Calculated for the representative range of 20-30 0 C.

6 Calculated for the representative range of 20-127 0C.

7 This value was not recalculated for the new LEU design. It is assumed to be the same as the old LEU design value.

2007 RAI responses PUR-1 23/63 2 May 2007 18:23

Question 16

16. Table 4-11. According to Figure 4-23 the slope of the calculated and measured reactivity curve as a function of rod position for the regulating rod are approximately the same. Why is there a factor of ten difference between the measured and calculated maximum reactivity insertion rate for the regulating rod in the HEU core in Table 4-11?

Response

The difference noted in Table 4-11 was a typographical error. The values should be as shown in the updated Table 4-11 provided in the response to Question 15.

Question 17

17. Table 4-12. Some of the values in Table 4-2 do not agree with values in Table 4-12. Agreement:

kexss; SS-2 worth calculations; shutdown margin calculations Disagreement: Measured SS-2 worth; measured shutdown margin Please verify all values in these two tables.

Response

Table 4-12 should be as shown here:

Table 4-12: Comparison of shutdown margins for HEU and LEU cores (Replaces Table 4-12 in CP)

HEU Measured HEU Old LEU Design New LEU Design Calculated Calculated Calculated SS-2 Worth (Ak/k) -2.53% -2.35% -2.00% -1.89%

ke,, (Ak/k) 0.431% 0.433% 0.468% 0.351%

Shutdown Margin (Ak/k) -2.07% -1.93% -1.53% -1.58%

Question 18

18. Table 4-12. In Table 4-12 and Table 4-2 the calculated shutdown margin for the LEU core, taking into account the -0.32% systematic bias, is -1.53%. A footnote for Table 4-2 gives the calculated shutdown margin (without bias) as -1.31%. Should the value of the calculated unbiased shutdown margin be -1.21%

(-1.53%+0.32% = 0.79-2.00 = -1.21 %)?

Response

Second footnote in 4.2 should be "Assumes the same bias as the HEU core. The calculated excess reactivity is 0.79% Ak/k. And the calculated shutdown margin is -1.21% Ak/k." The value of 1.31 is a typo.

Question 19

19. Section 4.5.3. In calculating Peff from keff with and without prompt neutrons, was any bias applied to the eigenvalues?

Response

The delayed neutron fraction was calculated using the formulation:

keprompt keff -kpromp' efeff fieff k ,,f k *# .

2007 RAI responses PUR-1 24/63 2 May 2007 18:23

The eigenvalue calculations were performed using MCNP5. In the conversion submittal, the bias of the PUR-1 MCNP model was not taken into account when calculating Ieff. Thus, the unbiased Pe, values for the HEU and LEU cores are calculated as follows:

HEU Core:

keff = 1.00369 + 0.00012 kefrm ' -0.99571 " 0.00017 I*efl = 1 - 1.00369/0.99571 = 7.951 xl 0" +/- 2.07X1 0- (2.6%)

LEU Core:

keff = 1.00379 +/- 0.00010 kfpro"W =0.99589 +/- 0.00015 03e0 = 1 - 1.00379/0.99589 = 7.870x1 03 + 1.79x1 0' (2.3%)

The bias of the PUR-1 MCNP model is APbia= 0.32% Ak/k. This was determined by comparing the results of eigenvalue calculations for several cases with the control rods at measured critical positions.

Accounting for the bias in the core model, the ,e.ff is calculated as:

pbiased kef - APbias - (k °P _ APhbias) k-- eff pt efef ef f r kef- APbia, ke - APbias HEU Core:

8,bUsed = (1.00369 - 0.99571)/(1.00369 - 0.0032) = 7.976x10-3 +/- 2.07x10.4 (2.6%)

LEU Core:

4 8,b sef = (1.00379 - 0.99589)/(1.00379 - 0.0032) = 7.895x10-3 +/- 1.79x10- (2.3%)

The reactivity insertion accident analyses were performed using the unbiased jeff values, which are slightly smaller (0.3%) than the effective delayed neutron fraction determined by accounting for the bias in the MCNP model. Consequently, the reactivity insertion accident analyses were performed with a value of eff that gives more conservative results.

Question 20

20. Please clarify the derivation of the temperature and void coefficients of reactivity by providing the following information:

A. Is the water density perturbation used in the water temperature coefficient and for the void coefficient?

B. What is the meaning of the void coefficient expressed in the unit of Ap/°C? Why is the void coefficient expressed in Ap/°C in Tables 4-14 and 4-15?

C. How was the temperature coefficient calculated and what components were included?

D. What scattering kernel data were used for these calculations?

E. What are the uncertainties of these calculations?

F. For both the HEU and LEU cores, the temperature coefficient is negative if the temperature of the water between the fuel plates and between the fuel cans is increased but is slightly positive if there is a simultaneous increase in the temperature of the water outside the core, that is if the entire inventory of water in the system is included. This might imply that if heating of the water occurs external to the core, there would be a positive change in reactivity. Are there any scenarios where the temperature of the water can be increased external to the core without increasing the water temperature within the core? Is there any physical scenario 2007 RAI responses PUR-1 25/63 2 May 2007 18:23

where temperature of the water exterior to the core can be simultaneously increased along with the temperature of the water in the interior of the core?

Response

A.

No. The effect of heating the reactor coolant on water temperature and water density were treated as separable effects for the purpose of calculating reactivity feedback coefficients. This question is addressed in more detail in the response to Question 20.C below.

B.

The different units given for the void coefficient in Tables 4-14 and 4-15 in the Conversion Proposal are simply different expressions of the same quantity. The unit Ap/% void is the typical unit for expressing the void coefficient, and the unit required by the PARET code which was used for the accident analyses. The methodology for calculating the void coefficient is described in the text of the Conversion Proposal and in the response to Question 20.C below.

The void coefficient was also converted to the unit of Ap/C. The reason for expressing the void coefficient in the alternative units was to facilitate a comparison of the water void and temperature reactivity feedback effects, which were treated as separable. This comparison is illustrated in Figures 4-28 and 4-31. When the reactor coolant within the fuel element (i.e., between the fuel plates, which is the region of interest for accident analyses in the PUR-1) is heated, the reactivity feedback from the water temperature increase is slightly larger.

The unit conversion to Ap/ 0C was accomplished by equating the void (water density) perturbation to a corresponding water temperature perturbation at 1.5 atmospheres. This is the water pressure in the PUR-1 core, which is at the bottom of a 15 foot tank of water. The density of water at 1.5 atmospheres as a function of temperature is shown in Figure 4-29.

C.

Temperature coefficients of reactivity were calculated assuming separability of the reactivity feedback effects due to water temperature, water density (void), and fuel temperature. This is a common practice and makes it easy to understand the inherent shutdown mechanisms that are responsible for affecting reactivity-induced transients. Not only are the feedback coefficient treated as separable, the reactivity coefficients were also calculated under non-isothermal conditions because different regions of the reactor will heat up at different rates during a power increase.

Three distinct regions containing water in the PUR-1 were considered.

1. The water between the fuel plates; this is called the "fuel assembly" water.

2. The water between the fuel element cans and also the water in the control elements between the control rod guard plates; this is called the "inter-assembly" water.

3. The water in the reactor tank; this is called the "reflector" water.

Water temperature coefficients were calculated by adjusting the temperature of the water in each of these regions and calculating the impact on the core reactivity. Reactivity feedback coefficients due to the perturbation of the fuel assembly water temperature, the fuel assembly plus inter-assembly water temperature, and fuel assembly, inter-assembly, and reflector water temperature were calculated. It is important to note that only the feedback effect due to heating of the fuel assembly water (between the fuel plates) was considered in the accident analyses, because only this water would experience an immediate heating due to power increases during the transient. The other regions containing water would take much longer to heat up.

Heating the water increases the thermal motion of the hydrogen atoms in the water. The result is to increase the energy of neutrons which are in "thermal equilibrium" with the hydrogen moderator, thus hardening the neutron spectrum in the reactor. At higher neutron energies, the U-235 fission cross 2007 RAI responses PUR-1 26/63 2 May 2007 18:23

section is reduced. Thus, there is a negative reactivity feedback effect due to heating of the water between the fuel plates.

Increasing the temperature of the reflector (tank) water was found to have a positive reactivity feedback effect over a small temperature range from nominal conditions. This is due to a decrease in the neutron absorption in the reflector as the temperature is increased and the neutron spectrum hardens. For the HEU core, the feedback coefficient due to heating all water in the reactor tank from 20 to 30 0C was calculated to be 2.38x1 03 +/- 2.11 xl 03 % Ak/k/C. However, it would take a long time for any transient to heat the reflector water, so it is judged that there are no significant safety issues related to the positive reactivity feedback coefficient when all the water is heated. It should be noted that for larger increases in the temperature of the reflector water, the reactivity feedback effect is negative.

D.

MCNP allows for two adjustments on neutron scattering reactions based on the temperature of the medium. For neutron energies above 4 eV, the code adjusts the elastic scattering cross sections of nuclei in the medium using a free gas thermal treatment. The code user can specify the temperature of each cell within the model and the cross sections are adjusted if the specified temperature differs from the temperature of the processed nuclear data in the cross section library.

For modeling elastic and inelastic scattering events for neutrons below 4 eV, an S(a,13) treatment is employed. These data are available in the MCNP libraries for certain materials; light water and graphite are of interest for the PUR-1 analyses. S(cp4) data for graphite at 20 0C were employed. Furthermore, S(a,,) which had previously been evaluated for light water at 20, 30, 60, 100, and 150 0C were also used.

E.

Eigenvalue calculations were performed with MCNP, typically using from 25 to 50 million neutron histories. These calculations yielded a reactor k9, with a 1-a uncertainty of +/-17 pcm (0. 017% Ak/k) for 25 million histories; the uncertainty was reduced to +/-12 pcm (0. 012% Ak/k) when 50 million histories were employed. The 1-a uncertainty of the eigenvalue calculations reduces by the square root of the number of histories. The uncertainty of the reactivity feedback coefficients can be calculated as the square root of the sum of the squares of the eigenvalue calculations. Based on the eigenvalue calculations with 25 to 50 million histories, the uncertainties of the water temperature and water void coefficients were found to be on the order of 15% to 30% when calculated over the expected 10 to 20 0C perturbations of water temperature.

It was decided that the calculational uncertainty should be reduced by extending the number of histories in the MCNP eigenvalue calculations. Eigenvalue calculations for the nominal core state and certain other cases which exhibit only small perturbations to the core keff were performed with 200 to 300 million neutron histories. This reduced the 1-a uncertainty of the eigenvalue calculations to around 5 pcm (0.005% Ak/k). The new results for the water temperature, water void, and fuel temperature coefficients, along with the corresponding uncertainties, are presented in Table Q20E-1.

Furthermore, a subset of the LEU-core kinetics parameters and reactivity coefficients were recalculated for the new LEU core design (see response to Question 5.B). This was done in order to quantify the impact of the changes to the assembly can design and the orientation and number of fuel plates in the core on the reactivity coefficients. The recalculated values are also included in Table Q20E-1. The impact of these changes on the accident analyses are discussed in the response to Question 39.

An updated Table 4-13 from the August 2006 Conversion Proposal is also provided to reflect the newly calculated values.

2007 RAI responses PUR-1 27/63 2 May 2007 18:23

Table Q20E-1: Water Temperature and Void Coefficients for PUR-1 HEU and LEU Cores.

(Ref: Tables 4-14 and 4-15 in CP)

PUR-1 HEU Core Water Temperature Coefficient (Gwater) _

20 to 30 0C -1.054E-04 Ap/°C +/- 7%

30 to 600 C -1.156E-04 Ap/°C - 5%

60 to 100 0C -1.384E-04 Ap/°C + 4%

PUR-1 Old LEU Core Design Water Temrperature Coefficient (a,,,,)

20 to 30 0C -9.743E-05 jAp/C +/- 8%

30 to 60 °C -1.155E-04 Ap/°C +/- 5%

60 to 100 °C -1.229E-04 AptOC +/- 5%

PUR-1 New LEU Core Design Water Temperature Coefficient (a,,.,)

20 to 30 0C -9.051 E-05 [AP/OC [ 9%

PUR-1 HEU Core Water Void Coefficient (add) 0 to 0.60% void -1.535E-03 Ap/% void +/- 8%

0.60% to 1.50% void -1.460E-03 Ap/% void +/- 6%

1.50% to 3.99% void -1.683E-03 Ap/% void +/- [4%

PUR-1 Old LEU Core Design Water Void Coefficient (o* )

0 to 0.60% void -1.898E-03 Ap/% void +/- 8%

0.60% to 1.50% void -1.933E-03 Ap/% void +/- 7%

1.50% to 3.99% void -1.980E-03 Ap/% void +/- 4%

PUR-1 New LEU Core Design Water Void Coefficient (aid) 0 to 0.60% void -1.933E-03 Ap% void +/- 17%

PUR-1 Old LEU Core Design Fuel Temperature Coefficient (ao*.,)

20 to 127 0C -7.014E-06 Ap/°C I +/- 12%

PUR-1 New LEU Core Design Fuel Temperature Coefficient (oaf.,)

20 to 127 0C -8.053E-06 I p/°C +/-- 10%

PUR-1 Old LEU Core Design Effective Delayed Neutron Fraction ({efl) 0.007871 +/- 10.00018 PUR-1 New LEU Core Design Effective Delayed.Neutron Fraction (Defl) 1 0.007841 +/- To0.00008 Table 4-13: Comparison of other core physics parameters for HEU and LEU cores.

(Replaces Table 4-13 in CP)

HEU Old LEU Design New LEU Design (calculated) (calculated) (calculated) c4.1 (%A.k/k 0C) 0 -7.01 E-04 -8.05E-04 odffator (%Ak/k °C) -1.05E-02 -9.74E-03 -9.05E-03 (wvoid

(%/Akk %-void) -1.54E-01 -1.90E-01 -1.93E-01 D." 0.795% 0.787% 0.784%

9Ps) 76.7 81.3 81.3" F.

Unless additional fixtures were to be placed in and around the core to physically separate the core and the pool, there are no scenarios where the water external to the core could be heated without heating of the water within the core. The temperature of the entire pool and the water in the core could be reduced with operation of the chiller system. In addition, heaters and stirrers could be added to the pool to heat the water uniformly.

8 This value was not recalculated for the new LEU design. It is assumed to be the same as the old LEU design value.

2007 RAI responses PUR-1 28/63 2 May 2007 18:23

Question 21

21. Section 4.5.3. What is the expected lifetime of the proposed LEU core and how does that translate into bumup? Explain if burnup over core life will have any significant impact on the reactor neutronic parameters.

Response

The reported burnup of the PUR-1 reactor as of the 2005 annual report was 0.00099% of the original HEU fuel load, which was first put into service in 1962 (43 years of service). The expected lifetime of the LEU fuel is at least as long as the HEU fuel. At a licensed power of 1 kW, there is no expectation that the LEU fuel burnup will exceed what the HEU core experienced, therefore, no impact on the reactor neutronic parameters is expected. As is the present practice, the control rods will still be examined for material degradation annually, as part of the regular inspection cycle.

Question 22

22. Section 4.5.3. Once the LEU core is loaded, is there any anticipated reloading or rearrangement of the fuel? If yes, how would this change the nuclear design parameters?

Response

It is possible that fuel may be rearranged within the 16 fuel assemblies (i.e. dummies and fuel plates relocated) during the lifetime of the PUR-1 LEU core to allow for experimental or instructional needs.

Approved standard operating procedures 8, 9A and 9B were used previously for this evolution, with an standard approach to critical performed each time. Verification of control rod worths and core excess reactivity will ensure that TS requirements are maintained.

Changes to the fuel configuration are addressed in Technical Specifications 4.1 and 6.1.10.

Question 23

23. Table 4-13. The fuel temperature reactivity coefficient shown in Table 4-13 for the LEU fuel is not the most conservative value as compared to the values shown in Table 4-15. Explain the basis for selecting the representative value for the fuel temperature coefficient.

Response

Values of the fuel temperature coefficient were calculated for fuel temperatures ranging in three intervals:

from 20 to 127 0C, from 127 to 227 0C, and from 227 to 327 0C. It is unlikely that the fuel temperature in the PUR-1 will increase much above 27 'C for the hypothetical accidents considered in the Conversion Proposal. Furthermore, the analysis of the rapid insertion of 0.6% Ak/k (see response to Question 39) revealed that the LEU fuel temperature increased to no higher than 120 0C. Consequently, it is appropriate to use the fuel temperature coefficient calculated over the lower range of temperature increase.

2007 RAI responses PUR-12 29/63 2 May 2007 18:23

Question 24

24. Section 4.5.4.1. In Section 4.5.4.1 it says that the peak power in plate 89 of bundle 3-3 is about 13% lower than that of plate 262. Please explain how the 13% value was calculated.

Response

In Table 4-17 it is observed that the power in plate 262 is 7.3% higher than the power in plate 89. The power given in this table is for the entire plate. Taking into account the axial power profile in these plates, however, it is found that the difference in the peak power densities between plates 262 and 89 is 13%.

Question 25

25. Section 4.5.4.2. The hottest fuel plate for the proposed LEU core was calculated considering two different critical control rod configurations while the hottest fuel plate for the HEU core was calculated considering five different crtical control rod configurations. The critical control rod configuration used to determine the hottest fuel plate in the LEU and HEU cores differ. Why were different critical control rod configurations used for the two cores? Which configuration is closest to the control rod configuration used during reactor operation? Does the calculation for the HEU core represent the limiting case? If not, what is the limiting case for the hottest fuel plate?

Resoonse:

The five measured configurations that were available were calculated for the HEU core:

1. SS-2 controlling with SS-1 and RR fully withdrawn,

2. SS-1 controlling with SS-2 and RR fully withdrawn,

3. all three rods at the same height (banked)

4. RR fully inserted, with SS-2 controlling and SS-1 fully withdrawn

5. RR at 30 (half inserted), with SS-2 controlling and SS-1 fully withdrawn.

The case with all three rods banked (Case 3, above), was found to be the limiting case for the HEU core power distribution, although the peak power densities for all five cases were found to be very similar (see Figure 4-34 of the Conversion Proposal). Calculations using any of these five power densities would result in peak cladding temperatures that are different by less than 1°C, and probably by tenths of 1°C.

For the LEU core design evaluated in the Conversion Proposal, only the first two configurations (with SS-2 and SS-1 controlling, respectively) were calculated. As shown in Figure 4-36 of the August 2006 document, the peak powers in these configurations are virtually identical. We expect a peak power that is slightly higher if the rods are banked, but that this slightly higher peak power will lead to a maximum cladding temperature that is higher by a few tenths of 1°C, and have no effect whatsoever on the safety of the facility.

However, for the sake of consistency with the HEU core calculations, the power distribution for the new LEU core design (see response to Question 5.B) was evaluated for a "banked rod" critical configuration.

The results of the new power distribution analysis are discussed in the response to Question 27.

For the HEU core, all of the configurations listed are used for the operation of the reactor, and it is anticipated that all of these configurations will be used for the LEU core when it is complete.

The banked rod calculation for the HEU core represents the limiting case.

Question 26

26. Figure 4-37. Figure 4-37 shows the radial.power profile in LEU plates 1215 and 1348. According to Figure 4-35, the "bottom" edge of plate 1348 is closer to the core centerline and hence is expected to have a higher power density than the "upper" edge. However the text of the conversion SAR says that the radial segments (as shown in Figure 4-37) for plate 1348 are numbered from top-to-bottom. Figure 4-27 2007 RAI responses PUR-1 30/63 2 May 2007 18:23

has radial segment 1 showing a higher power density than radial segment 11. Should the radial segments be numbered from bottom-to-top instead?

Response

There is a typographical error in the conversion proposal. The radial segments for plate 1348 in the LEU core and shown in Fig. 4-37 and Table 4-21 are numbered from bottom to top.

Question 27

27. Section 4.5.4.2. In Section 4.5.4.2 it states that radial power profiles shown in Tables 4-21 and 4-22 for plates 1348 and 1215 are used for the thermal-hydraulic analysis of the plates. Where is the radial power profile incorporated in the thermal-hydraulic analysis? If the radial power profile is not explicitly included in the analysis, explain the impact on the thermal-hydraulic analysis.

Response

Old LEU Core Design The NATCON code input RPEAK on the input card 4 (described in Appendix 1) was used in the conversion proposal to account for only the average (over the width of the specific fuel plate) of the radial power profiles shown in Tables 4-21 and 4-22 of the August 2006 Conversion Proposal. RPEAK (referred to as the radial power factor of a specific plate) is the ratio of power generated by a specific plate to the average power per plate for the core. The impact of the power density variation along the width of the plate itself was not incorporated in the analysis presented in the conversion proposal. It has been incorporated now as described below, and the impact is that the 'most limiting ONB power for the old LEU core design becomes 88.6 kW (down from 96.1 kW in the conversion proposal). In addition to the improvement discussed in this question, the 88.6 kW value also includes the effect of two additional improvements suggested questions 30 and 33, and addressed in the answers to these questions.

For the most limiting LEU plate 1348 (located plate-to-plate with 20 mil uncertainty on 197 mil thick channel), the radial power factor RPEAK already incorporated in the analysis presented in the conversion proposal is 1.6404. Over the 59.563 mm width of the fuel meat, the power density varies as shown in Table 4-21 of the August 2006 Conversion Proposal, and the maximum-to-average ratio of this power density variation is 1.0857. To account for this variation, the hot channel factor (FFILM) for temperature drop across the film is increased by a factor of 1.0857, i.e., set equal to 1.248xl.0857=1.355. The radial power factor RPEAK of the plate is kept unchanged at 1.6404, and the hot channel factor (FBULK) for bulk coolant temperature rise is also kept unchanged at 1.321. The ONB power calculated using this FFILM (= 1.355), along with changes in hot channel factors FW and FBULK (in response to Question numbers 30 and 33), is 88.6 kW. This result is shown in Table Q27-1 (column H) for comparison with the results of changes in FW and FBULK. The effect of the change in FFILM alone is a reduction in ONB power from 95.8 kW to 88.6 kW (column G versus column H in Table 027-1). Compared to this, the proposed operating power of PUR-1 is only 1 kW.

New LEU Core Design The following calculations were done to determine the core power distribution and the ONB power for the new design of the PUR-1 LEU fuel assemblies, as described in the response to Question 5.B. The power distribution analyses discussed below shows that the LEU plate 1348 is the most limiting. The ONB power calculation given below includes the impact of the power density variation along the width of the plate itself (shown in Figure Q27-3 and Table Q27-4), and also the effect of two additional improvements suggested questions 30 and 33.

For the most limiting LEU plate, 1348 (located plate-to-plate in control assembly 4-4 as shown in Fig.

027-1), the ONB power is found to be 94.2 kW, using a 20 mil uncertainty on the 181 mil thick channel.

The 20 mil uncertainty used is conservative compared to the 15 mil uncertainty given in Table Q5B-1 for 2007 RAI responses PUR-1 31/63 2 May 2007 18:23

the new LEU design. This ONB power is shown in Table Q27-1 (column I) for comparison with the HEU core and the old LEU core design. The ONB power for the new LEU design is higher than that for the old LEU design, mainly due to a smaller radial power factor in plate 1348 and due to a smaller channel thickness in the new LEU design. Reducing the channel thickness increases the ONB power because the new LEU design is approaching the optimum channel thickness (- 100 mil) which gives the highest ONB power.

Power Distribution in New LEU Core Table Q27-2 compares the heating by assembly and in non-fueled components for the LEU core with the rods banked at 53.5 cm. The banked rod critical configuration was found to have the highest peak power density in the analysis of the HEU core. Elements 3-4, 4-3, and 4-4 are noted as having the highest average plate powers, and are therefore of interest for thermal-hydraulics analysis. Figure Q27-1 provides a schematic of the core layout and the plate orientations.

Table 027-3 compares the power in individual plates in elements 4-4, 3-4, and 4-3 for the reactor with the banked rods critical configuration. The tallies were summed over the fuel meat in each fuel plate, all clad, coolant, and the bundle can. The plates are numbered from left-to-right in these elements (see the element drawings in Figure Q27-1). It can be seen that plate 1348 in assembly or bundle 4-4 has the highest power (8.07 W). This plate is adjacent to the large water hole that the SS1 rod falls into, and nearer the center of the reactor than plate 1355 on the other side of the water hole. Plates 1228 (bundle 3-4) and 1315 (bundle 4-3) face the center-line of the core, and have roughly equal power (6.51 and 6.41 W).

Figure Q27-2 compares the local-to-average axial power density profiles for fuel plates 1348, 1228, and 1315 for the banked rods critical configuration. Plate 1348 has the highest peak power density of all the plates in the LEU core and was evaluated in the thermal-hydraulics analyses. The slight "pinching" of the axial power profile due to the insertion of the SS1 shim rod is evident.

The local-to-average axial power density profiles in plates 1228 and 1315 are nearly identical due to their symmetric positioning across the core center-line. For these two plates, the highest peak/average density (1.712) occurs in plate 1228. The peak power density in plate 1228 of standard element 3-4 is about 22%

lower than that of plate 1348. Furthermore, the coolant channel thickness in the standard fuel assembly is narrower than that in the control assembly (144 vs. 181 mils). This reduction in channel thickness brings the channel thickness closer to the optimum value (= 100 mil), which will result in a higher ONB power with natural convection cooling. Therefore, plate 1348 is the most limiting of all fuel plates.

A higher water-to-fuel ratio at the edges of the fuel plates induces a power density profile along the width of the fuel plates. This is shown in Figure 027-3 for plates 1348, 1228, and 1315 with the rods in the banked critical position. The radial segments are numbered from bottom-to-top in the MCNP model (see the plate orientations indicated in Figure Q27-1). It is expected that the power density will be higher at the edge of the plate closest to the core center-line.

Lastly, Tables W27-4, 027-5, and Q27-6 provide the axial and radial power profiles for plates 1348, 1228, and 1315, respectively. These data were used for thermal-hydraulic analysis of the LEU-fueled PUR-1.

Table Q27-2 is related to Table 4-19 in the August 2006 Conversion Proposal for the old LEU core design. Table Q27-3 is related to Table 4-20 in the Conversion Proposal. Tables Q27-4, Q27-5, and Q27-6 are related to Tables 4-21 and 4-22 in the Conversion Proposal.

Figure Q27-1 is related to Figure 4-35 in the August 2006 Conversion Proposal for the old LEU core design. Figures Q27-2 and Q27-3 are related to Figures 4-36 and 4-37 in the Conversion Proposal.

2007 RAI responses PUR-1 32/63 2 May 2007 18:23

Table 027-1. ONB Powers of HEU and LEU Cores With Hot Channel Factor FW to Account for Hydrodynamically Developing Flow, and the Effect of Viscosity Temperature Dependence on FBULK (Most Limiting Fuel Plate of Each Core)

HEU Plate 2 6 2 1"j LEU Plate 1348t21 Plate-to-Plate Plate-to-Plate Plate-to-Plate 20 mil Channel Type 20 mil Uncertainty 20 mil Uncertainty Uncertainty on 207 mil Channel on 197 mil Channel on 181 mil (Old Design) Channel I _(New Design)

Column Identification A B C D E F G H I Hot Channel Factor FW 1.000 1.054t'+ 1.054 1.000 1.048' 1.048 1.048 1048 Hot Channel Factor FBULK 1.312151 1.312 1.301 I' 1.321161 1.321 1.30816) 1.308 1.322181 Hot Channel Factor FFILM 1.154 1.154 1.154 1.248 1.248 1.248 1 .355171 1.357l81 ONB Power (kW) 76.3 75.8 75.9 _ 96.1 95.7 95.8 88.6 94.2 AP Due to Wall Shear, 11.88 12.53 12.54 12.60 13.21 13.21 12.74 14.88 Pa Inlet + Outlet Loss AP, 2.29 2.32 2.32. 2.13 2.15 2.16 1.98 2.01 Pa Total Friction AP, Pa 14.17 14.85 14.86 14.73 15.36 15.37 14.72 16.89 AP Due to Buoyancy, Pa 14.17 14.85 14.86 14.73 15.36 15.37 14.72 16.89 Flow rate in hot channel, 20.83 19.89 19.90 19.12 18.34 18.36 17.61 16.28 g/s Reynolds Number at 844 816 816 792 769 770 730 698 Channel Outlet 1 Other hot channel factors used for HEU plate 262 are taken from Table 4-25 of the conversion proposal: F0 =

1.500, FH = 1.200, FFLUX = 1.120

2. Other hot channel factors used for LEU plate 1348 are taken from Table 4-26 of the conversion proposal: F0 =

1.500, FH = 1.200, FFLUX = 1.226

3. The ratio of apparent friction parameter Cam for developing laminar flow to the friction parameter C for fully developed laminar flow, Cpp/C is 1.1105 for the 207 mil thick channel in the HEU core. This ratio was calculated using Eqs. (341) and (576) of Shah and London [Ref. 2]. FW in the NATCON code input was set to 1.1105°*5 because the frictional pressure drop is multiplied by a factor FW2 by the code.

4. The ratio of apparent friction parameter Cam for developing laminar flow to the friction parameter C for fully developed laminar flow, CapC is 1.0985 for the 197 mil thick channel in the LEU core. This ratio was calculated using Eqs. (341) and (576) of Shah and London [Ref. 2]. FW in the NATCON code input was set to 1.098505 because the frictional pressure drop is multiplied by a factor FW 2 by the code.

5. For the HEU core (20 mil uncertainty on 207 mil thick channel), the hot channel factor for bulk coolant temperature rise, FBULK, calculated ignoring the temperature-dependence of water viscosity is 1.312 (as reported in Table 4-25 of the conversion proposal). The revised value including the temperature-dependence of water viscosity is 1.301 (calculated using the formula in the revised Appendix 1).

6. For the LEU core (20 mil uncertainty on 197 mil thick channel), the hot channel factor for bulk coolant temperature rise, FBULK, calculated ignoring the temperature-dependence of water viscosity is 1.321 (as reported in Table 4-26 of the conversion proposal). The revised value including the temperature-dependence of water viscosity is 1.308 (calculated using the formula in the revised Appendix 1).

7. The value of FFILM that accounts for the power density variation over the plate width, as explained in response to Question number 27.

8. FBULK has changed to 1.322 due to 20 mil uncertainty on 181 mil channel thickness in the new LEU core design as shown in Table Q27-7. The FFILM in Table 027-7 (1.251) was increased by a factor of 1.085 to account for the power density variation along the width of plate 1348 (shown in Table Q27-4). This results in FFILM = 1.357

(=1.0850xl .251) that is used here.

2007 RAI responses PUR-13 33/63 2 May 2007 18:23

Table Q27-2. PUR-1 LEU Rotated Core Power Distribution for Banked Rod Critical Configuration.

Average Plate Power (W) Power (W) 2-2 (RR) 38.19 4.77 2-3 61.85 4.76 2-4 (SS2) 45.65 5.71 2-5 45.93 3.83 3-2 62.56 4.81 3-3 78.09 6.01 3-4 78.42 6.03 3-5 61.20 4.71 4-2 63.52 4.89 4-3 80.78 6.21 4-4 (SS1) 60.09 7.51 4-5 62.29 4.79 5-2 52.32 4.03 5-3 63.97 4.92 5-4 63.49 4.88 5-5 47.64 3.97 Inter-assembly water 3.78 Graphite reflector 9.52 Grid plate 2.27 Water reflector (pool) 18.00 SS1 0.23 SS2 0.17 RR 0.03 Total 1000.0 1 2007 RAI responses PUR-1 34/63 2 May 2007 18:23

Purdue LEU Core Layout (191 plates)

Figure 027-1. PUR-1 New LEU Core Design Layout. Numbers in parentheses reflect number of fueled plates in element.

2007 RAI responses PUR-1 35/63 2 May 2007 1.8:23

Table Q27-3. Plate Power in Bundles 4-4, 3-4, and 4-3 in New Design of PUR-1 LEU Core with 191 Fuel Plates for Banked Rod Critical Configuration.

Bundle 4-4 Power (W)

Plate 1345 Meat 7.00 Plate 1346 Meat 7.09 Plate 1347 Meat 7.40 Plate 1348 Meat 8.07 Plate 1355 Meat 7.74 Plate 1356 Meat 6.98 Plate 1357 Meat 6.59 Plate 1358 Meat 6.43 Clad 0.36 Water 2.19 Can 0.24 Total 60.09 Bundle 3-4 Power (W)

Plate 1215 Meat 5.80 Plate 1216 Meat 5.56 Plate 1217 Meat 5.49 Plate 1218 Meat 5.50 Plate 1219 Meat 5.61 Plate 1220 Meat 5.84 Plate 1222 Meat 5.91 Plate 1223 Meat 5.76 Plate 1224 Meat 5.73 Plate 1225 Meat 5.76 Plate 1226 Meat 5.92 Plate 1227 Meat 6.12 Plate 1228 Meat 6.51 Clad 0.48 Water 2.19 Can 0.24 Total 78.42 Bundle 4-3 Power (W)

Plate 1315 Meat 6.41 Plate 1316 Meat 6.16 Plate 1317 Meat 6.07 Plate 1318 Meat 6.04 Plate 1319 Meat 6.09 Plate 1320 Meat 6.28 Plate 1322 Meat 6.21 Plate 1323 Meat 5.92 Plate 1324 Meat 5.76 Plate 1325 Meat 5.68 Plate 1326 Meat 5.65 Plate 1327 Meat 5.68 Plate 1328 Meat 5.84 Clad 0.49 Water 2.19 Can 0.24 Total 80.71-2007 RAI responses PUR-113/32My20 36/63 82 2 May 2007 18:23

25 2.0 II1.5 -

110 0.5 0,0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A"d Segman Figui re 027-2. Axial Power Profiles in LEU Plates 1348, 128, and 1315 for Banked Rod Critical Configuration in New LEU Core Design.

1.8 T- - - -

1.7 .-.-- P= 1346 (UW0 Mut 4-4)

.etu = Uuwt34 I

1A 143 1.2 1.1 1.0 2 3 4 5 6 7 8 9 10 11 Rai S-MW4 Figure Q27-3. Radial Power Profiles in LEU Plates 1348, 128, and 1315 for Banked Rod Critical Configuration in New LEU Core Design 2007 RAI responses PUR-1 37/63 2 May 2007 18:23

Table Q27-4. Axial and Radial Heating Profile for LEU Plate 1348 of Bundle 4-4 for Banked Critical Configuration in New LEU Core Design.

z-low, z-high' Axial Segment (cm) (cm) Power (W) 0 1 1.88595 5.88645 0.304 0.99%

2 5.88645 9.88695 0.400 0.86%

3 9.88695 13.88745 0.517 0.76%

4 13.88745 17.88795 0.595 0.70%

5 17.88795 21.88845 0.676 0.66%

6 21.88845 25.88895 0.720 0.64%

7 25.88895 29.88945 0.747 0.63%

8 29.88945 33.88995 0.771 0.62%

9 33.88995 37.89045 0.762 0.63%

10 37.89045 41.89095 0.697 0.65%

11 41.89095 45.89145 0.641 0.68%

12 45.89145 49.89195 0.563 0.74%

13 49.89195 53.89245 0.385 0.86%

14 53.89245 57.89295 0.172 1.25%

15 57.89295 61.89345 0.120 1.48%

Total 8.070 0.21%

y-low, y-high' Local/Average Radial Segment (cm) (cm) Power Density 0 1 24.0838 24.6253 1.672 0.48%

2 24.6253 25.1668 1.604 0.48%

3 25.1668 25.7082 1.572 0.48%

4 25.7082 26.2497 1.554 0.49%

5 26.2497 26.7912 1.523 0.49%

6 26.7912 27.3327 1.519 0.49%

7 27.3327 27.8742 1.506 0.49%

8 27.8742 28.4157 1.496 0.50%

9 28.4157 28.9571 1.493 0.50%

10 28.9571 29.4986 1.496 0.50%

11 29.4986 30.0401 1.518 0.50%

'Positions correspond to MCNP model of PUR-1.

2007 RAI responses PUR-1 38/63 2 May 2007 18:23

Table Q27-5. Axial and Radial Heating Profile for LEU Plate 1228 of Bundle 3-4 for Banked Critical Configuration in New LEU Core Design.

z-low, z-high' Axial Segment (cm) (cm) Power (W) 0 1 1.88595 5.88645 0.236 1.07%

2 5.88645 9.88695 0.311 0.94%

3 9.88695 13.88745 0.395 0.83%

4 13.88745 17.88795 0.465 0.77%

5 17.88795 21.88845 0.523 0.73%

6 21.88845 25.88895 0.557 0.70%

7 25.88895 29.88945 0.584 0.70%

8 29.88945 33.88995 0.598 0.69%

9 33.88995 37.89045 0.576 0.69%-

10 37.89045 41.89095 0.540 0.71%

11 41.89095 45.89145 0.498 0.74%

12 45.89145 49.89195 0.436 0.80%

13 49.89195 53.89245 0.342 0.89%

14 53.89245 57.89295 0.255 1.05%

15 57.89295 61.89345 0.195 1.17%

Total 6.511 0.22%

y-low, y-highl Local/Average Radial Segment (cm) (cm) Power Density 11 1 24.0838 24.6253 1.424 0.52%

2 24.6253 25.1668 1.320 0.52%

3 25.1668 25.7082 1.262 0.53%

4 25.7082 26.2497 1.229 0.54%

5 26.2497 26.7912 1.194 0.54%

6 26.7912 27.3327 1.193 0.55%

7 27.3327 27.8742 1.178 0.54%

8 27.8742 28.4157 1.178 0.55%

9 28.4157 28.9571 1.189 0.54%

10 28.9571 29.4986 1.215 0.54%

11 29.4986 30.0401 1.297 0.54%

1 Positions correspond to MCNP model of PUR-1.

2007 RAI responses PUR-1 39/63 2 May 2007 18:23

Table Q27-6. Axial and Radial Heating Profile for LEU Plate 1315 of Bundle 4-3 for Banked Critical Configuration in New LEU Core Design.

z-low' z-high' Axial Segment (cm) (cm) Power (W) _ _

1 1.88595 5.88645 0.233 1.07%

2 5.88645 9.88695 0.303 0.96%

3 9.88695 13.88745 0.389 0.85%

4 13.88745 17.88795 0.463 0.79%

5 17.88795 21.88845 0.511 0.75%

6 21.88845 25.88895 0.543 0.72%

7 25.88895 29.88945 0.567 0.70%

8 29.88945 33.88995 0.574 0.70%

9 33.88995 37.89045 0.558 0.70%

-10 37.89045 41.89095 0.523 0.73%

11 41.89095 45.89145 0.490 0.76%

12 45.89145 49.89195 0.436 0.81%

13 49.89195 53.89245 0.342 0.90%

14 53.89245 57.89295 0.265 1.02%

15 57.89295 6.1.89345 0.210 1.13%

Total 6.407 0.23%

y-owr y-highl Local/Average Radial Segment (cm) (cm) Power Density 0 1 16.3894 16.9309 1.279 0.54%

2 16.9309 17.4724 1.201 0.55%

3 17.4724 18.0138 1.172 0.56%

4 18.0138 18.5553 1.147 0.56%

5 18.5553 19.0968 1.144 0.57%

6 19.0968 19.6383 1.151 0.56%

7 19.6383 20.1798 1.174 0.55%

8 20.1798 20.7213 1.199 0.55%

9 20.7213 21.2627 1.246 0.54%

10 21.2627 21.8042 1.313 0.52%

11 21.8042 22.3457 1.435 0.51%

'Positions correspond to MCNP model of PUR-1.

ONB Power for New LEU Core Table 027-7 shows the hot channel factors for the new LEU core design, which were calculated using the equations in the revised Appendix 1 that includes the effect of temperature-dependent water viscosity. A conservative uncertainty of 20 mil (rather than 15 mil shown in Table Q5B-1) on the 181 mil channel thickness (the most limiting fuel plate 1348 in control assembly 4-4) was used in finding the hot channel factors. To calculate the ONB power, the NATCON code was run using (i) the hot channel factors shown in Table Q27-7, (ii) a total of 191 fuel plates, (iii) a channel thickness of 181 mil, (iv) a radial power factor of 1.5414 (=the ratio of 8.07 kW power in plate 1348 to 5.2356 kW per average plate), and (v) the axial power shape for plate 1348 shown in Table Q27-4.

2007 RAI responses PUR-1 40/63 2 May 2007 18:23

Table Q27-7. Hot Channel Factors for the New LEU Core [1]

Hot Channel Factors FBULK FFILM Uncertainty Type of Tolerance Tolerance Fraction Coolant Temp.

RsRieHeat Film Temp FFLUX Flux Rise Rise Fuel meat thickness Random 0.000 1.000 1.000 1.000 U-235 Homogeneity Random 0.200 1.000 1.200 1.200 U-235 Mass per plate Random 0.030 1.014 1.030 1.030 Power Density Random 0.100 1.046 1.100 1.100 Channel Thickness Random 0.110 1.180 1.111 1.000 Flow Distribution Random 0.200 1.200 1.000 1.000 Random Uncertainties Combined 1.322 1.251 1.226 Power Measurement F0 Systemic 0.500 1.500 Flow friction factor Fw Systemic 0.048 1.048 Heat Transfer Coeff. FH Systemic 0.200 1.200

1. Plate-to-plate channel, 20 mil uncertainty on 181 mil channel thickness. The revised Appendix 1 was used to include the effect of temperature-dependent water viscosity.

To account for the power density variation along the width of plate 1348, the FFILM in Table 027-7 (1.251) was increased by a factor of 1.085 (= 1.672/1.5412 = the maximum-to-average power density ratio variation over the width of plate 1348, shown in Table 027-4). This results in FFILM =1.085xl.251 =

1.357 that was used in NATCON to calculate the ONB power.

The NATCON code calculates the Darcy-Weisbach friction factor f = C/Re for fully developed laminar flow, using a built-in table of the parameter C for different aspect ratios of the rectangular channel cross section (values of parameter C are given in the response to Question number 29). To account for the increased pressure drop due to hydrodynamically developing laminar flow in the channel, an apparent value of the parameter C averaged over the channel length, called Capp, was calculated using Eq. (576) of Shah and London [Ref. 2]. The ratio CaJpC was found to be 1.0897 at a Reynolds number of 800 at the exit of the 181 mil channel in the new LEU design. Since the NATCON code multiplies the fully developed friction factor by FW 2, the hot channel factor FW equals 1.044 (= 1.08970'.). A higher value of 1.048 for FW was used in the NATCON calculation to be conservative.

The ONB power is found to be 94.2 kW which is shown in Table Q27-1 (column I) for comparison.

The step-by-step change in the ONB power of 88.6 kW for the old LEU design, due to each design change, to the ONB power of 94.2 kW for the new LEU design, was determined by running NATCON with one input change at time. These results are summarized in Table Q27-8 below.

Table Q27-8. Effect of LEU Assembly Design Changes on ONB Power.

Design Parameter Change ONB power From the Old LEU Design to the New LEU design Change, kW 1 Channel thickness plate-to-plate, mil 197 -- 181 88.6 --+ 95.1 2 Radial power factor of plate 1348 1.6404 -- 1.5414 95.1 --+ 101.2 3 Axial power shape (peak-to-average ratio) 1.345 -- 1.436 101.2 -- 94.0 4 All hot channel factors changed: FBULK 1.308 -- 1.322 94.0 -* 93.7 FFILM 1.355 -- 1.357 5 Number of fuel plates 190 -- 191 93.7 -- 94.2 2007 RAI responses PUR-1 41/63 2 May 2007 18:23

Question 28

28. Appendix 1. From the information in Appendix 1 it is not clear how insignificant are the channel inlet and outlet losses when compared to the wall shear. Please clarify.

Response

The information in Appendix 1 was used only to obtain hot channel factors for input to a more detailed thermal-hydraulic calculation using the NATCON code [Ref. 8 of the conversion proposal]. Therefore, Appendix 1 is a simplified modeling of what is calculated in detail in NATCON, and it is used only for the purpose of obtaining closed-form equations from which hot channel factors could be found. Appendix 1 does not include the minor losses. The minor losses calculated by NATCON are reported below, and found to be about 16% of the total frictional pressure drop in the HEU core, and 14% of the total frictional pressure drop in the LEU core (see Table Q27-1).

The pressure drop due to inlet and outlet losses were calculated (by the NATCON code) using loss coefficients of 0.5 and 1.0 respectively. The pressure drop due to wall shear along the channel length is found by summing the pressure drop for each axial mesh which is calculated using temperature-dependent coolant viscosity and density for the axial mesh (14 mesh intervals were used over the channel length in all calculations). The pressure drops are calculated by NATCON assuming fully developed laminar flow in a rectangular cross-section channel, and then multiplied by a factor FW (FW squared) where FW is an input which may be used to account for the increased pressure drop due to hydrodynamically developing laminar flow. In the calculations presented in the conversion proposal, FW was always set to 1.0, and thus the increased pressure drop due to developing laminar flow was not included. It is included in the calculations presented here (Table Q27-1). The method used is described below.

For the most limiting fuel plate in Table 4-27 of the conversion proposal for each core (HEU and LEU), a comparison of the pressure drops due to inlet plus outlet loss and wall shear, with and without the effect of developing laminar flow, are tabulated in Table Q27-1.

NATCON calculates the Darcy-Weisbach friction factor f = C/Re for laminar flow, using a built-in table of the parameter C for different aspect ratios of the rectangular channel cross section (values of parameter C are given in the response to Question number 29). An apparent value of the parameter C averaged over the channel length, called Ca, was calculated using Eq. (576) of Shah and London [Ref. 2 listed at the end of all responses] to account for the increased pressure drop due to hydrodynamically developing laminar flow in the channel. The ratio Capp/C was found to be 1.1105 for the 207 mil HEU channel, 1.09852 for the 197 mil LEU channel. Since the NATCON code multiplies the fully developed friction factor by FW as mentioned above, the input FW equals 1.054 and 1.048 for the HEU and LEU channels respectively.

NATCON calculations were done using these values of FW, and the pressure drops due to inlet plus outlet loss and wall shear are compared in Table Q27-1 (column B for the HEU channel, and column F for the LEU channel).

Table Q27-1 shows that the pressure drops due to wall shear and minor losses are 84% and 16%,

respectively, of the total pressure drop in the HEU channel at its ONB power; and the pressure drops due to wall shear and inlet plus outlet loss are 86% and 14%, respectively, of the total pressure drop in the LEU channel at its ONB power.

2007 RAI responses PUR-14 42/63 2 May 2007 18:23

Question 29

29. Appendix 1. From the information in Appendix 1 it is not clear what is the functional dependency of the laminar friction parameter C to the channel cross-section dimensions. Provide a reference for the evaluation of C.

Response

The following values (rows 1 and 2 of Table Q29-1) of the parameter C for fully developed laminar flow in a channel of rectangularcross section versus the width-to-thickness aspect ratio (wJt,) of the channel are used in the NATCON code that was used in the thermal-hydraulics calculations. The table starts from the square cross section (aspect ratio = 1.0) and goes to the infinite value of the aspect ratio (parallel plates).

In order to find the parameter C for the aspect ratio of the PUR-1 reactor, the NATCON code simply interpolates between the tabulated values. The original author of the code obtained these values from an old Reference [E. R. G. Eckert and T. F. Irvine, Heat Transfer Laboratory, University of Minnesota (1957)]

but these values are also given in a textbook by Frank Incropera [Ref. 3]. These values are obtained from the closed-form analytical solution for the fully developed laminar velocity distribution in a rectangular channel summarized by R. 'K. Shah and A. L. London [Ref. 2]. Equation (341) in [Ref. 2] is a fitted equation to easily find the parameter C. It should be noted that the aspect ratio used in [Ref. 2] is channel thickness-to-width ratio (the reciprocal of that used in NATCON and shown below in Table Q29-1), and the friction factor in [Ref. 2] should be multiplied by 4 to get the Darcy-Weisbach friction factor that is used in NATCON and tabulated below.

Table Q29-1. Friction Parameter C Used in the NATCON Code wC/tc 1.0 2.0 3.0 4.0 5.0 6.3 8.0 11.0 15.0 18.0 100.0 C in 58.0 63.0 69.0 72.5 77.0 80.0 83.0 85.0 88.0 89.0 96.0 NATCON C in 57.0 62.0 69.0 73.0 82.0 96.0 Ref. 3 C in 56.9 62.2 68.4 72.9 76.3 79.5 82.4 85.6 88.1 89.3 94.7 Ref. 2 Question 30

30. Appendix 1. From the information in Appendix 1 in both the calculation of the channel flow and the calculation of the bulk coolant temperature rise the ratio of the coolant kinematic viscosity to density (p/p) was assumed to be insensitive to temperature. Please demonstrate the validity of this assumption.

Response

The information in Appendix 1 was used only to obtain hot channel factors for input to a more detailed thermal-hydraulic calculation using the NATCON code [Ref. 8 of the conversion proposal]. NATCON does account for the temperature dependence of coolant viscosity and density in the calculation of the channel flow and the calculation of the bulk coolant temperature rise. Therefore, Appendix 1 is a simplified modeling of what is calculated in detail in NATCON, for the purpose of obtaining closed-form equations from which hot channel factors could be found.

As suggested in the question, water viscosity is temperature-dependent, i.e., it decreases with rising temperature. Appendix 1 was revised to account for the effect of temperature dependence of viscosity on hot channel factors, and the revised Appendix 1 is enclosed herewith. The temperature dependence of the dynamic viscosity of water over the temperature range 27 0C _<T _<50 °C (adequate for the PUR-1 reactor) can be written as follows.

/d (T) = ,(To) (1+T- To)a (Al) where a =0.12 To = 27 0C = Pool temperature of PUR-1 2007 RAI responses PUR-1 43/63 2 May 2007 18:23

/. ( To) = 0.875xl 0 -3 N-s/m 2 2 fl (T) = Temperature-dependent dynamic viscosity of water, N-s/m As shown in the revised Appendix 1, the revised relationship between the flow rate W in a channel and the friction parameter C is given by Eq. (A2). The revised formula for hot channel factor FBULK for bulk coolant temperature rise is given by Eq. (A3).

VV (A2) 3 12 FBULK = I+jKI+ Ui) 2+a(1 +' )~~

22+ _1 U (A3)

The exponent on the right hand side of Eq. (A2) changed from 0.5 (in the conversion proposal ignoring temperature dependence of p) to the revised value 1/2.12 = 0.4717. There exponents in Eq. (A3) for FBULK also changed, e.g., from 3/2 to 3/2.12 = 1.415. As a result of this revision, the hot channel factor FBULK decreased from 1.312 (in the conversion proposal) to 1.301 for the most limiting fuel plate 262 in the HEU core. Similarly, FBULK decreased from 1.321 (in the conversion proposal) to 1.308 for the most limiting fuel plate 1348 in the LEU core. The effect of ignoring the temperature dependence of viscosity is conservative.

NATCON calculations were done with these revised values of FBULK along with a value of FW > 1.0 to account for the increased friction due to developing laminar flow (in response to Question number 33).

The results are shown in Table Q27-1 (column C for the HEU core, and column G for the LEU core).

As a consequence of the two effects (i.e., increased friction due to developing laminar flow and the temperature dependence of viscosity) on hot channel factors FW and FBULK, the ONB power of the HEU core changes from 76.3 kW (reported in the conversion proposal) to 75.9 kW, and the ONB power of the LEU core changes from 96.1 kW (reported in the conversion proposal) to 95.8 kW. The effect is small for the PUR-1 reactor.

Question 31

31. Table 4-24. A pool temperature (coolant inlet temperature) of 27°C is used in the LEU thermal-hydraulic analyses. The TSs contain no limits on pool temperature. Please explain why a limit on pool temperature should not be added to the TSs or propose a limit on pool temperature.

Response

The average pool temperature in recent PUR-1 operating history is 26 0C. An estimated value of 270 C was used in the calculations. The figure below shows measured pool temperatures from 1993 to 2006.

2007 RAI responses PUR-1 44/63 2 May 2007 18:23

PUR-1 Reactor Pool Temperature Measurements 35.0 30.0"-___ £ ___*___~ I _______________" _"

25.0-- ______ - T' 20.0 15.0*

10.0 5.0 0.0 .

1/31/1993 10/28/1995 7/24/1998 4/19/2001 1/14/2004 10/10/2006 7/6/2009 Week Fig 031: Measured PUR-1 pool temperatures from 1994 to 2007.

Pool temperatures up to 30 0C will have no effect on reactor safety. Peak temperatures in the thermal-hydraulic and transient analyses would be about 3°C higher than those reported in the conversion proposal. All safety margins would be far below their limiting values. However, PUR-1 Staff believes that the maximum pool temperature is not a HEU to LEU conversion issue, since there is currently no technical specification on maximum pool temperature for the HEU core.

However, a NATCON thermal-hydraulic calculation for the LEU plate 1348 (for case H reported in Table 027-1 included in response to Question 27) was performed assuming a hypothetical pool temperature of 35 °C, and a hypothetical inlet loss coefficient of 10.0 (increased from 0.5), while applying all six hot channel factors of the case H. The ONB power was found to be 79.3 kW, indicating a large margin compared to the PUR-1 operating power of 1 kW.

Question 32

32. Appendix 1. Equation (30) has two terms and the conversion proposal states that the expression within the parenthesis on the right hand side of the equation varies slowly compared to the heat flux ttue q.'/2. Demonstrate the validity of the statement with reference to the PUR-1 fuel plate.

Response

Equation (30) of Appendix 1 is for finding a hot channel factor for the temperature drop from the meat mid-plane to cladding surface (AT,,t.). This temperature drop is very small compared to the temperature drop from the cladding surface to bulk coolant (ATfimr). For example, in the PUR-1 HEU fuel plate 262 without hot channel factors, ATmraj is 0.07 'C and ATfim is 46.98 °C (at meat mid-height) at a high power of 100 kW. Similarly, in the PUR-1 LEU fuel plate 1348 without hot channel factors, ATrnetj is 0.05 °C and ATfijm is 34.5 0C at a power of 100 kW. Therefore, the hot channel factor for ATmetm is not important for PUR-1. The important hot channel factor is the factor FFILM for ATfim. In the case of PUR-1, ATfim is the bigger component (bigger than the bulk coolant temperature rise) in the total temperature rise from the inlet temperature to the cladding surface temperature at the axial level experiencing the onset of nucleate boiling. The hot channel factor FFILM found by Eq. (29) of Appendix 1 in the conversion proposal remains unchanged. It depends on the uncertainties in q "tt 1 and channel thickness (as shown in Eq. 28), but not on the uncertainty in [tf/(4Kfu) + tclad/Kdad].

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In short, PUR-1 is not limited by the fuel peak temperature, but by the onset of nucleate boiling, and the uncertainty in [tfuej/(4KfueI) + tlad/Kdad] is not important for PUR-1. We believe that the hot channel factor FFILM has been determined accurately.

Question 33

33. Section 4.7.2. According to Appendix 1 the systematic uncertainty in flow rate is accounted for by applying the hot channel factor Fw to the laminar friction factor C. Explain the reason for the value of the flow friction factor Fw being unity in Tables 4-25 and 4-26.

Response

As suggested in the question, a value of FW (hot channel factor for flow) greater than 1.0 should be used to account for the increased frictional pressure drop due to the hydrodynamically developing laminar flow in the entrance region of the coolant channel, otherwise the code (NATCON) accounts only for the fully developed frictional pressure drop. This has been done now and the results are presented in Table Q27-

1. Since each coolant channel creates its own buoyancy to drive its own coolant flow, there is no uncertainty due to redistributionof a total reactor flow rate. The loss coefficients of 0.5 and 1.0 at channel inlet and outlet are used in the calculations. To account for the reduction in flow rate due to the hydrodynamically developing laminar flow in the channel, the values of FW were calculated for the most limiting channels in the HEU and LEU cores as follows.

NATCON calculates the Darcy-Weisbach friction factor f = C/Re using a built-in table of the parameter C for different aspect ratios of the rectangular channel cross section (values of parameter C are given in the answer to Question number 29). These values of parameter C are for the fully developed laminar flow in a rectangular cross-section channel. An apparent value of the parameter C averaged over the channel length, called Capp, was calculated using Eq. (576) of Shah and London [Ref. 2] to account for the increased pressure drop due to hydrodynamically developing laminar flow in the channel. The ratio Ca/C was found to be 1.1105 for the 207 mil HEU channel, and 1.0985 for the 197 mil LEU channel. Since the NATCON code multiplies the fully developed frictional factor by FW 2, the input FW equals 1.054 and 1.048 for the HEU and LEU channels respectively. The flow reduction factor is input factor FW or more accurately FW 2I( 2+a) = FW 0 .943 4 (noting that a = 0.12 for the PUR-1 reactor as mentioned in the revised Appendix 1 enclosed herewith).

The results of using these values of. FW in NATCON calculations (excluding the effect of temperature dependence of p on hot channel factors) are shown in Table Q27-1. The ONB power of the HEU core changes to 75.8 kW from 76.3 kW reported in the conversion proposal. The ONB power of the LEU core changes to 95.7 kW from 96.1 kW reported in the conversion proposal.

The channel flow indeed gets reduced by the factor FW 0 943 4 as expected. For the HEU plate 262, the flow reduces from 0.02083 kg/s to 0.01989 kg/s (see Table Q27-1) when the input hot channel factor FW is changed from 1.0 to 1.054. The expected reduced flow should be 0.02083/(1.054)0"' = 0.01982 kg/s which is close to the NATCON-calculated value of 0.01989 kg/s. For the LEU plate 1348, the flow reduces from 0.01912 kg/s to 0.01834 kg/s (see Table Q27-1) when the input FW is changed from 1.0 to 1.048. The expected reduced flow should be 0.01912/(1.048)°9434 = 0.01829 kg/s which is close to the NATCON-calculated value of 0.01834 kg/s.

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Question 34

34. Section 4.7.2. In Table 4-25 a random tolerance fraction of 0.05 was specified for the fuel meat of the HEU core. Explain why the corresponding tolerance fraction was given as 0.0 in Table 4-26 for the LEU core.

Response

The reason is that the hot channel factors for the HEU fuel in Table 4-25 are appropriate for uranium-aluminum alloy fuel and the hot channel for the LEU fuel in Table 4-26 are appropriate for uranium-aluminum dispersion fuel.

U-Al alloy fuel plates have a smooth interface between the fuel meat and the cladding. The 0.05 tolerance fraction is appropriate to account for potential variation in the U-235 distribution due to variation of the thickness of the fuel meat. The homogeneity of the U-235 in U-Al alloy fuel is quite high, resulting in a tolerance fraction of 0.03.

Uranium-aluminum dispersion fuel is different from the U-Al alloy fuel. The U3Si2 particles in the fuel meat have a jagged profile, as shown in fuel micrographs. For dispersion fuel, the variation in both the "thickness" and homogeneity is combined in the tolerance on homogeneity, which is taken from the LEU silicide fuel manufacturing specifications. The manufacturer measures the areal density of U-235 along each fuel plate to quantify the axial variation of U-235. In this context, a tolerance fraction on fuel meat thickness is not appropriate for dispersion fuel, so its value in Table 4-26 was set to 0.0.

Question 35

35. Section 4.7.3.1. It is not clear if the hot channel factors were used in the determination of the ONB power. Please clarify.

Response

The hot channel factors were indeed applied in the determination of the ONB powers reported in the conversion proposal. The ONB power is defined as that nominal power level at which the cladding surface temperature, with the five hot channel factors applied (Fo, Fw, FH, FBULK, and FFILM) at every axial location in the coolant channel, will equal the onset-of-nucleate-boiling temperature at any axial location in the channel. As an example, the ONB power of the HEU core calculated without applying any hot channel factors to the most limiting fuel plate 262 (plate-to-plate case) is 162.0 kW compared to 76.3 kW (as reported in the conversion proposal) with the hot channel factors applied. The ONB power of the LEU core calculated without applying any hot channel factors to the most limiting fuel plate 1348 (plate-to-plate case) is 220.4 kW compared to 96.1 kW (as reported in the conversion proposal) with the hot channel factors applied. In these calculations with or without applying hot channel factors, the radial power factor of the specified fuel plate (power of a specified plate divided by average power per plate) and the axial power distribution shape in the plate are always accounted for.

Question 36

36. Table 4-28. Define the parameter "margin to incipient boiling."

Response

The margin to incipient boiling shown in Table 4-28 was calculated at the nominal operating power of PUR-1 (i.e., 1 kW), and it is the smallest value of the temperature difference (TONB - T,) over the coolant channel length in the hottest channel where Tw is cladding surface temperature with all hot channel factors applied, and TONB is the local onset-of-nucleate-boiling temperature. This basically gives an idea of how far below the onset of nucleate boiling condition the reactor is operating. This definition can be written as an equation as follows:

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Margin to ONB = Minimum T-,, (p q"(z)FflUX,) -4F,,Ulk {T(z) -To)+ Ffim {IT,,,., (z) - T(z)l]

where T(z) = Bulk coolant temperature at axial position z in the channel heated by the plate power of PapFr FQ/N and applying the global hot channel factors for flow and Nusselt number of F, and Fh Twax1(z) = Cladding surface temperature at axial position z in the channel heated by a plate power of PopF, F0/N and applying the global hot channel factors for flow and Nusselt number of F. and Fh q"(z) = Heat flux at position z for the plate power of PopF, FQ/N and applying the global hot channel factors for flow and Nusselt number of Fw and Fh p(z) = Absolute pressure in the channel at axial position z Ti~p(p(z), ql"(zFf.*) = Onset of nucleate boiling temperature at absolute pressure p(z) and heat flux q"(z) Fff.

Pop

= Operating power of the reactor (e.g., 1 kW for PUR-1)

N = Number of fuel plates in the core (e.g., 190 for PUR-1 LEU core)

TO = Coolant temperature at the channel inlet Fr = RPEAK = Radial power factor of the plate cooled by the channel Fw = Hot channel factor for flow in the channel Fo = Hot channel factor for reactor power Fh = Hot channel factor for Nusselt number FfiIm = FFILM = Hot channel factor for temperature drop across the coolant film on cladding surface Fux = FFLUX = Hot channel factor for heat flux Fbulk = FBULK = Hot channel factor for bulk coolant temperature rise in the channel Question 37

37. Section 9.4. Please provide an analysis of storage of LEU fuel in the in-pool storage racks that shows that TS 5.3 will be met.

Response

Purdue is scheduled to receive the following:

• 12 standard (14 plates max) assemblies and 2 spares

• 3 control (8 plates max) assemblies and 1 spare

• 1 fission chamber (14 plates max)

• 288 LEU plates.

An MCNP model of the in-pool storage racks was constructed. Two cases for the LEU fuel were examined. One case was run with standard LEU assemblies in all of the 18 positions (which is not possible with the anticipated LEU inventory, but was run as a limiting case), and no credit was taken for the 1/4" BORAL plate between the two rows. This first case had a calculated eigenvalue of 0.7660-0.0046. The second case was modeled with the 'A" BORAL plate with a boron density of 23.8E21 8°*/=, and 16 standard assemblies with 14 fuel plates in each. The eigenvalue for this calculation was determined to be 0.3319 +/-0.00178. Both of these bracketing cases are below the TS 5.3 limit of 0.8, thus TS 5.3 will be met. The geometry of the model is shown in Figure 037-1 below.

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Figure Q37-1: Model of the in-pool fuel storage racks.

Question 38

38. Please propose license possession limits for the reactor conversion. Please provide the following possession limits and any other changes needed:

" amount of U-235 of enrichment less than 20 % possessed in the form of LEU fuel elements;

" amount of U-235 of any enrichment needed in connection with operation of the facility and the form of the material (e.g., fission chambers, flux foils, fueled experiments);

" amount of U-235 of any enrichment in the form of the existing HEU fuel elements until that core is removed from site.

Resonse:

Siecial Nuclear Material Chemical or Physical Form Possessed Uranium enriched in the U-235 Isotope AI Clad MTR U-Al alloy Reactor 30 grams U-235 UraniumenrichedintheU-235_Isotop Plates of any enrichment 3000_gramsU-235 Uranium enriched in the U-235 Isotope A]rClad MTR Rec0 P/o 3800 grams U-235 enriched to :520 */ 30 gas.-3 Flux foils, fission chambers, or Uranium enriched in the U-235 Isotope fueled experiments of any 100 grams U-235 enrichment Gum-dfon 39

39. Sections 13.2 and 13.4. What is the maximum amount of positive rapid reactivity and slow insertion (up to a maximum of the allowed excess reactivity) that can be added to the LEU fueled reactor without exceeding the proposed safety limit?

Response

Previous Accident Anallvn The rapid insertion of the maximum worth of moveable and unsecured experiments (0.3% Ak/k in 0.1 seconds) was evaluated in the Conversion Proposal. In this accident analysis, it was assumed that the 7 second period trip failed, but scram was initiated due to the 1.2 kW over-power trip. Because of an assumed power measurement uncertainty of 50% in the PUR-1, the calculations were performed using a core power trip setting of 1.8 kW.

Another accident analyzed in the Conversion Proposal was the ramp insertion of 0.04% Ak/k/second.

Again, it was assumed that the 7 second period trip failed, but scram was initiated due to the 1.2 kW over-2007 RAI responses PUR-1 49/63 2 May 2007 18:23

power trip. A core power trip setting of 1.8 kW was used in the calculations to take into account power measurement uncertainties in the PUR-1.

The analyses of both of these accidents demonstrated the protection of the safety limit by the limiting safety system setting.

New Accident Analyses Based on RAI In order to address the question raised by the reviewers, additional accident analyses have been performed for the rapid and slow insertion of 0.6% Ak/k, the maximum excess reactivity allowed by the PUR-1 Technical Specifications. In these calculations, it was assumed that all trips failed.

In response to Question 20, the water temperature, water void, and fuel temperature coefficients were recalculated. The coefficients presented in Table Q20E-1 indicate that the reactivity coefficients are dependent on the temperature or void increase. The previous accident calculations were performed using the calculated reactivity coefficients with'the smallest magnitude in order to provide the most conservative results. Since then, the PARET code has been modified to allow input of the reactivity coefficients as a function of temperature or void. Since the reactivity coefficients have been recalculated and the PARET code now has better modeling capabilities, the previous accident analyses (0.3% Ak/k rapid insertion and 0.04% Ak/k/second slow insertion) have been repeated, as well.

Table Q39.1 presents the results of 0.3% Ak/k rapid insertion and 0.04% Ak/k/second slow insertion accident analyses for the HEU and LEU cores. As discussed above, a core power trip of 1.8 kW was applied in the analyses of these accidents. These results indicate that the rapid insertion of~the maximum worth of moveable and unsecured experiments in the PUR-1 results in a cladding temperature change of less than 1 0C from the steady-state peak clad temperature of 29 0C. The slow insertion of reactivity from a "runaway rod" accident also results in a less than 1 0C change in cladding temperature from the steady-state conditions. The temperatures reached in these accidents are well below the safety limit of 530 0C.

It should be noted that in the rapid insertion accident analyzed in the Conversion Proposal (Section 13.2),

the total amount of reactivity inserted was inadvertently set to $0.30 instead of the desired 0.3% Ak/k.

Consequently, the new results indicate that the peak power (and reactor scram) is reached sooner than the 2.5 seconds previously reported (see Table 13-1 of the Conversion Proposal).

Table 039.1. Results for Rapid and Slow Reactivity Insertion Accidents With Scram.

(Ref: Tables 13-1 and 13-2 in CP)

TclWad~rmax (C)

Time of P. Pr. Peak at at Peak Fuel Reactivity Inserted (kW) (kW) Power (s) t=0 Power Maximum HEU 0.3% rapid insertion 1.0 1.83 0.81 29 29 29 LEU 0.3% rapid insertion 1.0 1.83 0.75 30 30 30 S0.04%/s slowinsertion 1.0 1.83 5.92, 129 29 29 LEU 0.04%/s slow insertion 1.0 1.83 5.85 30 30 30 Table Q39.2 presents the results of rapid and slow reactivity insertion accidents up to the maximum allowed excess reactivity in the PUR-1. These accidents were analyzed assuming failure of the reactor protection system and without operator intervention.

For the rapid insertion of the maximum core excess reactivity of 0.6% Ak/k in the HEU-fueled PUR-1, the reactor power increases to 1.63 MW. The clad temperature increases by about 100 0C due to this transient. For the LEU core, the reactor power increases to a peak of 1.55 MW. The total reactor power increase is less in this case because of the negative Doppler feedback effect in the LEU fuel. For the slow reactivity insertion accident (an insertion rate of 0.04% Ak/Wsecond was assumed) the results for 2007 RAI responses PUR-1 50/63 2 May 2007 18:23

the peak power and final clad temperature are nearly the same as for the rapid insertion case, although the power and clad temperature profiles as a function of time are shifted due to the slower insertion of reactivity.

Table 039.2. Results for Rapid and Slow Reactivity Insertion Accidents Without Scram.

Time of Tcld.. (°C)

Peak P. P. Power at at Peak Fuel Reactivity Inserted (kW) (MW) (s) t=0 Power Maximum HEU 0.6% rapid insertion 1 1.63 124 29 134 134 LEU 0.6% rapid insertion 1 1.55 400 30 120 120 LEU J0.6% slow insertion 1 3 1.55 400 30 120 120 Figures Q39.1 and Q39.2 plot the reactor power and maximum clad temperature as a function of time for the insertion of the maximum excess reactivity into the PUR-1 HEU and LEU cores. The results of the PARET analyses indicate that after the reactivity is inserted, the core reaches a new steady-state power level. The shift in the power response when the reactivity is inserted slowly is evident in the curves. At the new, higher power level, the reactivity feedback effects from coolant density, coolant temperature, and fuel temperature balance the reactivity inserted in the accident. As a result of this accident, the maximum cladding temperature is 134 °C in the HEU core. The maximum cladding temperature in the LEU core is 120 0C. These temperatures are well below the PUR-1 safety limit of 530 0C.

Included on Figures 039.1 and Q39.2 are results from the SPERT-IV B-1 Test. In this test, 0.6% Ak/k was rapidly inserted into an HEU-fueled MTR-type core designated as D12/25 (25 fuel elements with 12 plates/element). This core was similar to the PUR-1 HEU core. Prior to the reactivity insertion, the reactor was operating at low power with zero flow. The SPERT test was terminated by an operator-initiated scram after 30 seconds. The comparison of the SPERT data and the accident analyses with PARET for the PUR-1 with HEU fuel indicate good agreement for the first power peak and corresponding maximum clad temperature. The PARET code predicts the reactor power and maximum clad temperature over an extended period.

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10 r 1

/

-- HEU (0.6% rapid Insertion) 0.1 II SPERT B-1 Test LEU (0.6% rapid Insertion)

- - - LEU (0.6% slow Insertion) 0.01 0.001 0 50 100 150 200 250 300 350 400 Time (a)

Figure 039.1. PUR-1 Reactor Power in Reactivity Insertion Accidents without Scram.

160 1 140 1 120 100 1 j 801-

'I V

HEU (0.6% rapid insertion)

SPERT, B-1 Test 60 1 LEU (0.6% rapid Insertion)

- - - LEU (0.6% slow insertion) 40 Ii 20 0 1_

0 50 100 150 200 250 300 350 400 Time (a)

Figure 039.2. PUR-1 Peak Clad Temperature in Reactivity Insertion Accidents without Scram.

2007 RAI responses PUR-1 52063 2 May 2007 18:23

Accidents in the New LEU Core Design The redesign of the LEU assembly can, and changes to the orientation and number of fuel plates in the LEU core model, have a small effect on the calculated reactivity coefficients and kinetic parameters. The recalculated values are summarized in Table Q20E-1 in response to Question 20.E. It is noted that the calculated water void (or water density) coefficient and fuel temperature coefficient are slightly more negative for the new LEU core design; this change will lessen the impact of the reactivity insertion accidents. Also, the water temperature coefficient and the effective delayed neutron fraction are slightly smaller for the new LEU core design. All changes to the calculated reactivity coefficients and kinetics parameters are within 1-a of the statistical uncertainty of the MCNP results.

The changes to the reactivity coefficients are small. Furthermore, the analyses for the accidents with reactor scram showed less than 1 0C change in the clad temperature. Therefore, it is not necessary to repeat these accident calculations using the new feedback coefficients because the changes will not affect the safety of the facility.

Larger temperature increases are observed for the rapid and slow insertion of 0.6% Ak/k without scram.

The analysis for the old LEU core design found that in these accidents, the PUR-1 reaches a steady-state condition where the reactivity inserted in the accident is balanced by the water temperature, water void, and fuel temperature feedbacks. In the steady-state condition (see results at 400 seconds in Figures Q39-1 and Q39-2), the negative reactivity feedbacks are as follows.

APwater temperature = -0.32% Ak/k APwater void = -0.24% Ak/k A2_fuelem:era~ur = -0.04% Ak/k APFeedback, total = -0.60% Akfk Thus, the net reactivity in the core is 0.0 (0.6% Ak/k inserted by the accident; -0.6% Ak/k in reactivity feedbacks).

In this case, the water temperature feedback effect accounts for roughly half of the total reactivity feedback, while the remainder of the feedback effect is due to the decrease in the coolant density between the fuel plates and the increased temperature of the fuel. It is noted above that the water temperature feedback may be smaller in the new LEU core design, while the water void and fuel temperature feedbacks are stronger. Consequently, the weaker water temperature feedback coefficient in the new LEU core design will be compensated for by the stronger feedback coefficients for the water void and fuel temperature. As such, the new LEU core design will have very little impact on the results of the 0.6% Ak/k insertion accident without scram already analyzed for the old LEU core design. Therefore, it is not necessary to repeat this accident analysis.

Question 40

40. Section 13.4. Was reactivity introduced by the flooding or voiding of an experiment located in the irradiation facility considered? Is that a viable accident (it was evaluated in NUREG-1283.for the 1988 license renewal)? If not, please explain. If so, please provide an analysis.

Response

This was not considered as a viable accident. However, calculations show that the sudden flooding of an irradiation facility (F6) with water when it is initially full of air results in a reactivity addition of 0.175% Ak/k, which is lower than the other accidents analyzed. Voiding of a flooded irradiation facility results in a negative reactivity insertion (see Q6).

The flooding of the 12.7 cm drop tube when placed in position G6 as described in the SER NUREG-1283 resulted in a reactivity insertion of 0.246% Ak/k, which is below the insertion resulting from the failure of a 2007 RAI responses PUR-1 53/63 2 May 2007 18:23

moveable experiment examined in the conversion proposal. This accident scenario is within the envelope of the examined accident cases, and therefore was not analyzed.

Question 41

41. TS 4.4.d. There are proposed modifications to TS 4.4.d to change fuel inspection requirements from periodic inspection of representative fuel plates to representative fuel assemblies.

A. What will the fuel assembly inspection consist of and how will this inspection help ensure fuel cladding integrity?

B. TS 4.3.d requires monthly analysis of primary coolant. Is this analysis used as an indicator of fuel cladding integrity? If not, can the analysis be adapted for such purpose?

Response

A. Representative fuel assembly inspection will still be performed annually, without disassembly of the element. Visual inspection will be performed on the representative sample elements for signs of degradation such as corrosion, channel blockage, warped or bloated plates, Surveillance of the primary water system for contamination will complement this inspection process, please see part B of this question.

A representative HEU plate has been inspected for the 44 years that PUR-1 has been in service, and no degradation has been observed. The particular plate that has been examined had a blemish from the original machining of the plates. The blemish has not changed per visual inspection.

B. TS 4.3d requires monthly testing of the primary coolant for gross alpha and beta contamination in order to assure against undetected leaking fuel assemblies, as stated in the bases for the technical specification. Since the intention of the TS requirement is to detect fuel cladding failures, it is intended to be used for this purpose, and supports the deletion of the fuel element disassembly as part of the inspection. No further changes to this TS should be required.

Question 42

42. Section 14, TS 4.4. Section 4.1 footnotes and Table 4-1 indicate a change in the alloy of aluminum to 6061. Please clarify the following sentence: "No new alloys will be introduced into the reactor as a result of conversion from HEU to LEU fuel."

Response

The aluminum alloy 6061 composes the reactor deck, and various hardware parts of the control rod assemblies (e.g. the spacer plates) and the assembly can hardware. These were included in the conversion proposal model, therefore the statement can be made that no new alloys will be introduced into the reactor as a result of the conversion.

REFERENCES

[1]. Idaho National Laboratory (INL), "Specification for Purdue University Standard and Control Fuel Elements-Assembled for the Purdue University Reactor," Document ID SPC-382, May 2006.

[2]. R. K. Shah and A. L. London, "Laminar Flow Forced Convection in Ducts," Supplement 1 to Advances in Heat Transfer, Academic Press, New York, page 199 (1978).

[3]. F. P. Incropera, D. P. Dewitt, T. L. Bergman, and A. S. Lavine, "Introduction to Heat Transfer," 5 th Edition, page 489 (see Table 8.1), John Wiley & Sons (2007).

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APPENDIX 1: REVISED TO INCLUDE THE EFFECT OF TEMPERATURE DEPENDENCE OF VISCOSITY ON HOT CHANNEL FACTORS (Revised by MK 4/5107)

1. Hot Channel Factors in the NATCON Code Version 1.0 The NATCON code version 1.0 [Ref. ANURERTR/TM-12] uses three hot channel factors (FQ, FW, FH).

Using the source code and documentation, the factor FH used in NATCON is found to be the same as the factor FNUSLT used by E. E. Feldman. Table 1 shows the tolerances and uncertainties included in each of the six hot channel factors used by E. E. Feldman. The correspondence between the NATCON hot channel factors and E. E. Feldman's six hot channel factors is as follows.

Feldman's Hot Channel Factor NCATCON Input Variable System-wide Factors:

FFLOW a factor to account for the uncertainty in total reactor flow FW (approximately)

FPOWER a factor to account for the uncertainty in total reactor FQ power FNUSLT a factor to account for the uncertainty in Nu number FH correlation Local Factors: FBULK (new input)

FBULK a hot channel factor for local bulk coolant temperature rise FFILM a hot channel factor for local temperature rise across the FFILM (new input) coolant film FFLUX a hot channel factor for local heat flux from cladding FFLUX (new input) surface

2. Hot Channel Factors in the NATCON Code Version 2.0 Sections 2.1 and 2.2 develop, for laminar natural convection, two thermal-hydraulic relationships that are used in section 2.3 to obtain formulas for the hot channel factors from user-supplied manufacturing tolerances and measurement uncertainties. The results of section 2.3 are summarized here for convenience. The first three are local/random hot channel factors, and the last three are system-wide. An example of the use of these-hot channel factors is given in section 4, with NATCON running instructions in section 3, and the new input description in section 5.

3 2 FBULK =1 +1jf1(1 + U1 Y"(1 +U 2 )K1 -15 +U6 2 FBULK is higher (conservative) if the temperature dependence of water viscosity is ignored.

2 2 2 2 FFILM = 1ý+U 2 +U 2 +U 3 +U 4

+U 5 2 2 FFLUX= 1+,u,2u+- 2 +u 3 + u42 FQ = 1 + u7 FW=1 +us FH=1 +u 9 where ul = Fractional uncertainty in neutronics calculation of power in a plate u2 = Fractional uncertainty in U-235 mass per plate = Am/M 2007 RAI responses PUR-1 55/63 2 May 2007 18:23

u3 = Fractional uncertainty in local (at an axial position) fuel meat thickness U4 = Fractional uncertainty in U-235 local (at an axial position) homogeneity u5 = Fractional uncertainty in coolant channel thickness = (t,* - thc) / trc U6 = Fractional uncertainty in flow distribution among channels U7 = Fractional uncertainty in reactor power measurement u8 = Fractional uncertainty in flow due to uncertainty in friction factor u9 = Fractional uncertainty in convective heat transfer coefficient, or in the Nu number correlation M = Nominal mass of U-235 per plate, gram Am = Tolerance allowed in U-235 mass per plate, gram The code obtains, for an input nominal reactor power CPWR, a thermal-hydraulic solution using the three systematic hot channel factors FW, FQ and FH. If the user-input reactor power is zero, then the code itself chooses the nominal power from a series of power levels (10 kW, 100 kW, 200 kW, and so on increasing in steps of 100 kW). This thermal-hydraulic calculation is done for a hot plate power of CPWR*FQ*(Radial power peaking factor RPEAK)/(Total number of fuel plates in standard and control assemblies). Also, the frictional resistance to flow is multiplied by FW2 , and the convective heat transfer coefficient found for laminar flow in a rectangular channel is divided by FH. The random hot channel factors FBULK, FFILM and FFLUX are not used in this solution.

Having obtained the above solution, the random hot channel factors FBULK, FFILM and FFLUX are applied to the temperatures obtained, using the following equations. The temperatures calculated with all six hot channel factors are printed after the above solution. The onset of nucleate boiling ratio, ONBR, is computed using the temperatures with all six hot channel factors applied (using the equation below). If the user-input nominal power is zero, then the last nominal power for which the code prints a solution is that at which the ONBR is 1.0.

TihCf = To + (Ti - To)*FBULK TwalIj,6hcf = Ti,6hd + (Twajl,i - Ti)*FFILM Tmax,i,6hcf = TwaII,.6i.hc + (Tmax.i - Twaii.i)*FFLUX ONBR = ( Ti'.7'i -To)

(Twa.lJ,6hcfj -To where To = Bulk water temperature at the coolant channel inlet, i.e., the pool temperature, 0C Tj = Bulk water temperature in node i of the channel with only systematic hot channel factors applied, 0C Twau = Cladding surface temperature in node i with only systematic hot channel factors applied, °C Tmax,i = Fuel meat centerline temperature in node i with only systematic hot channel factors applied, 'C Ti,6he = Bulk water temperature in node i of the channel with all six hot channel factors, 'C Twu.j6W = Cladding surface temperature in node i with all six hot channel factors, 0C Tmai,6hf = Fuel meat centerline temperature in node i with all six hot channel factors, C Tincp,i = Incipient boiling temperature in node i with only systematic hot channel factors applied, C

2.1 Flow Rate in a Coolant Channel versus Power of a Fuel Plate NATCON is a laminar natural circulation code. The flow rate is calculated in the code by balancing the buoyancy pressure force to the laminar friction pressure drop. Following this concept, an analytical relationship is developed here (with some approximation) for the coolant flow rate in a single coolant 2007 RAI responses PUR-12 56/63 2 May 2007 18:23

channel in terms of the power generated in a fuel plate and the channel geometrical dimensions. The analytical relationship is needed for obtaining hot channel factors.

The hot channel factor FW used in the code to account for the uncertainty in coolant flow rate is actually applied to the laminar friction factor in the code, that is, the laminar friction factor is multiplied by FW 2. It is not applied directly to the flow rate. The relationship developed here explains how this technique works.

p, T1 at channel outlet L = Channel height containing hot coolant (hotter than pool), m P = Power in a single fuel plate or the two half plates, W W=Upward flow rate in a single channel, kg/s p 0 , To at channel inlet Schematic of what the code analyses, that is, a single rectangular coolant channel heated by a half of a fuel plate on each side (right and left sides).

The above schematic shows what the code analyses, that is, a single rectangular coolant channel heated by a half of a fuel plate on each side (right and left sides). See Fig. 1 for details. The buoyancy pressure force is caused by the decrease in water density due to heating in the channel. The temperature dependence of water density can be written as p (T) = po " po ,8(T-To) (1) where Ti = Bulk water temperature at channel outlet, C AT = T1 - To = Temperature rise in channel from inlet to outlet, C 3

po = Water density at channel inlet, i.e., the water density in the pool, kg/m

= Volumetric expansion coefficient of water, per C 3

p = Average coolant density in the channel, kg/m L = Channel height that contains hotter coolant (hotter than pool), m. It is the sum of heat generating length of fuel plate, non-heat generating fuel plate length at top, and the assembly duct length above the top 2 of fuel plate g = Acceleration due to gravity, 9.8 m/s The buoyancy pressure force is given by Buoyancy A p =(po - p ) g L (2)

The average coolant density p is given by p = 0.5 (po+p)= p -0.5 Po /8 (T, - To)= po - 0.5po 8 AT (3)

Buoyancy A p = 0.5 po ,6 AT g L (4) 2007 RAI responses PUR-12 57/63 2 May 2007 18:23

The coolant temperature rise AT can be written in terms of the input power P generated in a fuel plate, as shown by Eq. (5) below, and then the buoyancy A p of Eq. (4) can be written in terms of the input power P, as shown by Eq. (6).

AT = P/ (W C0) (5)

Buoyancy A p - POgLP (6) 2WCp Ignoring the minor losses at channel inlet and outlet, the laminar frictional pressure drop in the channel is written below as Eq. (9) after using the laminar friction factor given by Eq. (7), and after replacing the coolant velocity by mass flow rate using Eq. (8). The parameter C in Eq. (7) is a constant for a given channel cross section, but it depends upon the channel cross section aspect ratio width/thickness, and varies from 57 for aspect ratio 1.0 (square channel) to 96 for an infinite aspect ratio (infinitely wide channel).

f = C /Re (7)

W= pAV (8)

Frictional Ap - 2 2D 2 2cV CAl LcW (9) 2 *o AD' where f = Moody friction factor for laminar flow in the channel Re = Reynolds number in the channel = p VD/yu 2

A = Flow area of the channel cross section, m D = Equivalent hydraulic diameter of the channel cross section, m Lc = Total coolant channel length causing frictional pressure drop, m.

V = Coolant velocity averaged over the channel cross section, m/s W = Coolant mass flow rate in the channel, kg/s 2

, = Average coolant dynamic viscosity in the channel, N-s/mr 2

/U (T) = Temperature-dependent dynamic viscosity of water, N-s/m

,A 0 = /I (TO) = Coolant dynamic viscosity at the channel inlet temperature To For the PUR-1 reactor, the temperature dependence of the dynamic viscosity of water over the temperature range 27 00 _ T _ 50 0C can be approximated as follows.

,t (T)= /I (TO) (1+T -To)-a (10) where a = 0.12, To = 27 °C, (TO) = 0.875x10 -3N-s/m 2 The average coolant dynamic viscosity p used in Eq. (9) can be set equal to the viscosity at the average coolant temperature (To + 0.5AT) in the channel. Putting this temperature in Eq. (10), the average viscosity a is found to be S= pl ( TO) (1+0.5AT)-. (11) 2007 RAI responses PUR-1 58/63 2 May 2007 18:23

Equation (11) indicates that the average viscosity Ut can be set equal to U1( To) if AT is just a few 0C (this is the case for the PUR-1 reactor at the operating power of 1 kW). If AT is greater than a few 0C, i.e., 1 <<

0.5AT (this is the case for the PUR-1 reactor at an ONB power of about 100 kW), then Eq. (11) simplifies to the following.

p (TO) (0.5AT)-a if AT>> 2 0C (12a)

S= / (To) if AT << 2 'C (12b)

Substituting Eq. (12a) into Eq. (9), the frictional Ap becomes Frictional A p = C2'(AT)-aLcW-Cf-LcW(W- (13) 21- p AD 2 2"' p AD 2 Equating the frictional A p of Eq. (13) to the buoyancy A p of Eq. (6) to find the steady-state coolant flow rate W in the channel, one obtains Eq. (14) below. Equation (14) can be rewritten as Eq. (15).

p0 f/gLP - C 4 LuLc Wa (14) 2WCP 21-a p AD 2 P 2

W2+= popAD/JgLPI+a (15) 2a CoLC C; +a(1 Equation (15) relates the fuel plate power to the channel flow rate in natural circulation. It is used to find the dependence of the flow rate on the parameter C in the laminar friction factor (at constant power). All parameters in this equation are constant (p is also practically constant) except the parameter C in the laminar friction factor. Based on Eq.(15), the relationship between the flow rate W and the parameter C is given by Eq. (16) below.

W ,* (16)

Equation (16) shows that the friction factor parameter C is multiplied by a factor (FW) 2, the coolant flow 2

rate W will be reduced by the factor (FW,)2+a. This has been verified by actually running the NATCON code for the PUR-1 reactor. Since a is small (a = 0.12 for the PUR-1 reactor), 2/(2+a) is nearly 1.0, and the flow rate W is reduced approximately by the factor FW.

2.2 Bulk Coolant Temperature Rise versus Power of a Fuel Plate Equation (5) expresses, for laminar natural circulation, the bulk coolant temperature rise in terms of fuel plate power, coolant flow rate and specific heat. Putting the value of flow rate obtained in Eq. (15) into Eq.

(5), the bulk coolant temperature rise is given by Eq. (17) below, purely in terms of power and the geometrical dimensions of the channel. The right hand side of Eq. (17) is rearranged into two factors in Eq. (18), such that the second factor is sensitive to power and channel geometrical dimensions that 2007 RAI responses PUR-1M 59/63 2 May 2007 18:23

usually have manufacturing tolerances and measurement uncertainties, and the first factor is insensitive to power and channel geometrical dimensions.

AT = Cjo;L (17)

(,Cpo po AD 2fgL "

I AT 2'CXc SCPPPol gL J (~2T D (18)

The nominal flow area and hydraulic diameter of a rectangular coolant channel are given by A = tnc wn (19)

Pw = 2 (tnc + wnc) (20)

D = 4 A/ P, 2 tn wnc/ (trr + wnc) (21) where tc = Channel thickness (spacing between fuel plates), m tnc = Nominal channel thickness (spacing between fuel plates), m the = Minimum channel thickness in hot channel (spacing between fuel plates), m Wc -=Channel width, assumed not to change from its nominal value, m P" = Wetted perimeter of the nominal channel, m Prm = Power generated in a fuel plate, without applying manufacturing tolerances, W Phc = Power generated in a fuel plate, after applying manufacturing tolerances, W Because the channel thickness tc is much smaller than the channel width wc in most experimental reactors, Eq. (21) reduces to D = 2 t, (22)

Using the channel area and hydraulic diameter given by Eqs. (19) and (22) into Eq. (18), the bulk coolant temperature rise can be written in terms of power, channel thickness, and channel width. This is the desired relationship for use in finding hot channel factors.

AT 2a CaoLc P (23)

ClpoPf gL )

2.3 Formulas for Hot Channel Factors For use in the NATCON version 2.0, six hot channel factors (three global/systemic and three local/random) are obtained from 9 manufacturing tolerances and measurement uncertainties ul, u2,..., u9 that are defined below. These are fractional uncertainties rather than percent. Of these nine uncertainties, those affecting a particular hot channel factor are indicated in Table 1. The systemic hot channel factors are given by Eqs. (24) through (26), and the random hot channel factors are given by Eqs. (27) through (29). -A utility Fortran computer program NATCONHCF and a Microsoft spreadsheet 2007 RAI responses PUR-3 60/63 2 May 2007 18:23

NATCON.HotChanFactors.xls have also been developed to compute the hot channel factors using these formulas.

FQ =1 + u7 (24)

FW =1+ u8 (25)

FH =1 + ug (26)

The ratio of the power generated in hot plate to its nominal power, caused by the uncertainties in neutronics-computed power and in U-235 mass per plate, can be written as Ph, = (1+u 1 )(1+u 2 ) (27)

P.c The ratio of bulk coolant temperature rise in hot channel to the temperature rise in the nominal channel, caused by the uncertainties in neutronics-computed power, U-235 mass per plate, and channel thickness, is obtained from Eq. (23). Only the quantity in the second parentheses is important here because the quantity in the first parentheses is insensitive to these uncertainties.

ATh rI~

ATp Ph )

1 Yt2 3

2+aat'2+ 2 I I.)28

_aI 3

The uncertainty in flow distribution is assumed to reduce the channel flow to (1- u6) times the flow without this uncertainty, and therefore the bulk coolant temperature rise is increased by the factor (1+ uQ). This uncertainty in bulk coolant temperature rise is statistically combined with that given by Eq. (28) to obtain the following formula for the hot channel factor FBULK for input to the NATCON version 2.0.

_3 2 FBULK= 1+Il+u, 12+,a(1+u2 )(+.l- -l +u 62 (29)

The temperature drop across coolant film on the cladding surface at an axial location is given by Eq. (30).

Here the heat flux q" (W/m 2 ) on the cladding surface is replaced by tf q"72 in terms of the volumetric power density q.. (W/m 3) in the fuel meat.

ATum q" _ t (30)

The convective heat transfer coefficient h (W/m -C) is giyen by Eq. (31). Here the laminar Nusselt number Nu is independent of flow rate, and varies only slowly with the aspect ratio (width/thickness) of coolant channel. The main variation of the heat transfer coefficient with channel thickness is due to the denominator of Eq. (31). The numerator of Eq. (31) is considered to be constant.

h- N=K,°., - N"Kc°0 , (31)

D 2t:

Using Eq. (31) for the heat transfer coefficient, the temperature drop across coolant film can be written as Eq. (32).

2007 RAI responses PUR-1 61/63 2 May 2007 18:23

ATqitm tf t, N." Ký.ool (32)

Equation (32) states that ATfhm is directly proportional to the fuel meat thickness (having uncertainty u3),

the channel thickness (having uncertainty u5), and the power density in meat. The uncertainty in power density is caused by three uncertainties, that is, u1 , u2 and u4. Statistically combining these five uncertainties gives the following formula for the hot channel factor FFILM for input to the NATCON version 2.0.

FFILM = 1+1 u2-u 2+-u- 2 +u 3 2" +u' 42 52 (33)

The uncertainty in the heat flux at the cladding surface is included in the hot channel factor FFILM given by Eq. (33). A hot channel factor FFLUX for the heat flux alone can be found from Eq. (34) for heat flux in terms of the power density q.' in the fuel meat and the thickness of the meat. The fractional uncertainty in heat flux is the sum of fractional uncertainties in power density and meat thickness, as given by Eq. (35).

q"- q"'t q tf (34)

S 2 8q 4q + Stfuel (35) q" qM tfuel In Eq. (35), the uncertainty in power density is caused by three uncertainties, that is, u1, u2 and u4. The uncertainty in the meat thickness is given by u3 . Statistically combining these four uncertainties gives the following formula for the hot channel factor FFLUX for input to the NATCON version 2.0.

2 FFLUX = I+u12u2 + 2 +u 3 +U4 2 (36)

The uncertainty in the temperature drop ATmetM from fuel meat centerline to cladding surface is not important in the case of the PUR-1 reactor because ATmetd is very small compared to ATfiim. For example, ATmetal is 0.05 °C and ATfilr is 34.5 °C at 100 kW without any hot channel factors.

2007 RAI responses PUR-1 62/63 2 May 2007 18:23

Table 1. Uncertainties Included in the Six Hot Channel Factors Used in NATCON Version 2.0 (X implies that an uncertainty affects a hot channel factor)

Uncertainty Fraction FQ FW FH FBULK FFILM FFLUX Local or random uncertainties 1 Neutronics calculation of X X X power in a plate, ul 2 U-235 mass per plate, u2 X X X 3 Local fuel meat thickness, u3 X X 4 U-235 axial homogeneity, u 4 X X 5 Coolant channel thickness, u5 X X 6 Flow distribution among X X channels, u6 L System-wide uncertainties 7 Reactor power measurement

_ uncertainty, u7 8 Flow uncertainty due to uncertainty in friction factor, X U8 9 Heat transfer coefficient uncertainty X due to uncertainty in I Nu number correlation, u9 I 2007 RAI responses PUR-1 63/63 2 May 2007 18:23

PURDUE SCHOOL OF NUCLEAR ENGINEERING Nuclear Engineering Building 400 Central Drive West Lafayette, IN 47907-2017 S-?Cc S U

Document ID: SPC-382 Revision ID: 1 Effective Date:

Specification Specification for Purdue University Standard and Control Fuel Elements-Assembled for the Purdue University Reactor The INL is a U.S. Department of Energy National Laboratory operated by Battelle Energy Alliance.

Idaho National Laboratory

INTENTIONALLY BLANK Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: i of ii PURDUE UNIVERSITY REACTOR Document Project File No. Revision I. Identifier: SPC-382 2. (optional): 3. No.: I Specification for Purdue University Standard and Control Fuel Elements - Assembled for the

4. Document

Title:

Purdue University Reactor Comments

5. : All review and approval signatures for this specification are located on DAR Number 511249.

SIGNATURES 6.

Type or Printed Name Signature Signature Code 7.

Date I

[ 8.

Organization/

Discipline See DAR Number 506184.

See DAR Number 511249.

Document Control Release

9. Signature: Date:

'kEC-ORDS~MANAGEMENT < *..

10. Is this a Construction Specification? Yes E] No 0 11. NCR Related? Yes LI No [

Does document contain sensitive, unclassified information? E] Yes [0 No If Yes, what

12. category:

13. Can document be externally distributed? Yes Z No E]

Area Index

14. Code: Area Type SSC ID Review annually. Cutoff when Record superseded, obsolete or Uniform File Disposition Retention cancelled. Destroy 75 years

15. Code: 0250 16. Authority: ENVI-b-4-a Period: after cutoff.

17. For QA Records Classification Only: Lifetime -, Nonpermanent EI, Permanent El Item or activity to which the QA Records apply:

Periodic Review Frequency: N/A R-, 5 years 2, or

18. Other

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: ii of ii PURDUE UNIVERSITY REACTOR INTENTIONALLY BLANK

Fomi 412.09 (R*v. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 1 of 59 PURDUE UNIVERSITY REACTOR Purdue University Reactor Specification I DAR Number: 506184 REVISION LOG Rev. Date Affected Pages Revision Description 0 05/31/06 All New Document.

I All Revised to add Program Anneal requirements and update Drawing Titles

Form412.09(Re,'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 2 of 59 PURDUE UNIVERSITY REACTOR I CONTENTS I

SUMMARY

....................................................................................................S 1.1 General................................... ..............

2. APPLICABLE CODES, PROCEDURES, AND REFERENCES ....................................... 5 2.1 Standards, Specifications, Drawings and Attachments...........................................S 2.1 .1 Specifications and Standards .......................................................... S 2.1.2 Publications (Idaho National Laboratory)........................................... 7 2.1.3 Drawings (INL)........................................................................ 8

3. TECI-TN1CAL REQUIREMENTS.......................................................................... 9 3.1 Production Qualification........................................................................... 9 3.1.1 Fuel Plate Qualification: .............................................................. 9 3.1.2 Fuel Element Qualification: ......................................................... 10

3.1.3 Requalification

...................................................................... 10 3.1.4 Operator Qualification: ............................................................... l1 3.2 Materials.............................I............................................................... 12 3.2.1 Fuel Bearing Plates................................................................... 12 3.2.2 Aluminum Weld Filler Metal: ...................................................... 12 3.2.3 Dummy (Non-Fueled) Plate:........................................................ 12 3.2.4 Material Requirements .............................................................. 13 3.3 Mechanical Requirements ........................................................................ 13 3.3.1 Fuel Plate Requirements............................................................. 13 3.3.2 Non-fueled (dummy) plates: ........................................................ 15 3.3.3 Fuel Element Requirements ......................................................... 15 3.4 Physical Properties ................................................................................ 15 3.4.1 Fuel Plate Requirements............................................................. 15 3.4.2 Fuel Element Requirements ......................................................... 17 3.5 Surface Condition................................................................................. 17 3.5.1 Surface Defects....................................................................... 17

3.5.2 Cleanliness

........................................................................... 18 3.5.3 Contamination:....................................................................... 18 3.6 Marking ............................................................................................ 18 3.6.1 Fuel Plate Identification: ............................................................ 18 3.6.2 Fuel Assembly Identification: ...................................................... 19 3.7 Storage ............................................................................................. 19 3.8 Fuel Element Surface Treatment ................................................................ 19

Fonn 412.09 (Re.. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 3 of 59 PURDUE UNIVERSITY REACTOR 3.8.1 ........................................................................................................................... 20 3.8.2 .......................................................................................................................... 20 3 .8.3 ........................................................................................................................... 20 3.9 G raphite Reflectors and G raphite Radiation Baskets ...................................................... 20 3.9.1 M aterial: ....................................................................................................... 20 3.9.2 A ssem bly: .......................................................................................................... 20 3.9.3 W elding: ........................................................................................................ 20

3.9.4 Identification

.............................................................................................. 20 3.9.5 D im ensional Inspection: ............................................................................... 21 3.9.6 Surface Finish and Defects: .......................................................................... 21

3.9.7 Storage

.................. ............................... 21

4. QUALITY A SSURA N CE ................................................................................................................ 21 4.1 M aterials .............................................................................................................................. 22 4.2 Core D ensity 23 23...........................

4.3 Fuel Loading ........................................................................................................................ 24 4.4 Fuel H om ogeneity ................................................................................................................ 24 4.5 Core Configuration ........................................................................................................ 26 4.6 Bond Integrity ...................................................................................................................... 26 4.6.1 B lister Anneal: ............................................................................................. 26 4.6.2 U ltrasonic Scanning: .................................................................................... 27 4.6.3 M etallographic Examination ........................................................................ 27

,4.7 Internal D efects ................................................................................................................... 27 4.8 Surface Finish and D efects ............................................................................................ 28 4.9 Clad-Core-Clad D imensions ..............................................................................  :................. 28 4.10 Cleanliness .......................................................................................................................... 29 4.11 Contam ination ...................................................................................................................... 29 4.12 D imensional ......................................................................................................................... 29 4.12.1 Final D imensional Inspection ..................................................................... 29 4.13 Reactor Com ponents and Spare Fuel Elem ent Parts ...................................................... 29

5. PA CKA G IN G AN D SH IPPIN G ................................................................................................. 30

6. N O TES .............................................................................................................................................. 30

Fonn412.09 (Rev.09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 4 of 59 PURDUE UNIVERSITY REACTOR 1 6.1 Definitions......................................................................................... 30 6.2 Purchaser Tests.................................................................................... 33 6.3 Submittals.......................................................................................... 33 6.3.1 Preproduction:........................................................................ 33 6.3.2 Pre-repair:............................................................................. 34 6.3.3 Manufacturing Schedule:............................................................ 34 6.3.4 Delivery Submittals: ................................................................. 34 6.3.5 Fuel Plate Radiographs: ............................................................. 35 6.3.6 Core Compact Data Sheets: ......................................................... 35 6.3.7 Report of Production by Unapproved Process: ................................... 35 APPENDIX A...................................................................................................... 38 APPENDIX B ...................................................................................................... 40 FIGURES I. Purdue University Fuel Plate Sampling Procedures For Destructive Tests....................... 38

Form 412-09 (R-y. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 5 of 59 PURDUE UNIVERSITY REACTOR

1.

SUMMARY

1.1 General This specification (see def.) defines the materials, components, testing, inspection, certain processes, quality control (see def.) requirements and acceptance criteria for the fabrication of standard and control fuel elements (see def.) and fuel element containers for the Purdue University Reactor at Purdue University at West Lafayette, Indiana.

2. APPLICABLE CODES, PROCEDURES, AND REFERENCES 2.1 Standards, Specifications, Drawings and Attachments The applicable portions of the following documents as defined herein, form a part of this specification. Where there is a conflict between the documents cited and the latest revision thereof, the supplier (see def.) shall notify the purchaser (see def.) of the conflict and use the latest revision in effect at the signing of the contract, unless otherwise directed by the purchaser.

2.1.1 Specifications and Standards National Codes and Standards ASTM E 1417-99 Standard Practice for Liquid Penetrant Examination MIL-C-45662 Calibration System Requirements RDT F6-2T Welding of Reactor Core Components, Sections 1,2,3 and 6 American Society for Testing and Materials (ASTM)

ASTM B 209-00 Standard Specification for Aluminum and Aluminum-Alloy Sheet and Plate ASTM B 210-04 Aluminum and Aluminum-Alloy Drawn Seamless Tubes ASTM B 211-00 Standard Specification for Aluminum and Aluminum-Alloy Bar, Rod and Wire ASTM B 2 ~4-~,9 e S~e~ve Analysis

Form 412.09 (Re-. 09)

Idaho National Laborator ry SPECIFICATION F(OR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ALSSEMBLED FOR THE Effective Date: Page: 6 of 59 PURDUE UNI VERSITY REACTOR ASTM B 221-00 Standard Specification for Aluminum and Aluminum-Alloy Extruded Bars, Rods, Wires, Profiles and Tubes ASTM B 241-02 Aluminum and Aluminum-Alloy Seamless Pipe and Seamless Extruded Tube ASTM E 8-00 Methods of Tension Testing of Metallic Materials C~nsidefed,LS ilgnlfi4Thanin kifl ASTM E 2016-99 Standard Specification for Industrial Woven Wire Cloth American Welding Society (AWS)

AWS A5.10-1995 Aluminum and Aluminum Alloy Welding Rod and Bare Electrodes American National Standards Institute (ANSI)

ANSI B46.1-1994 Surface Texture ANSI Y14.5-1994 Dimensioning and Tolerancing for Engineering Drawings American Society of Mechanical Engineers (ASME)

ASME Section V - 2001, Boiler and Pressure Vessel Code without addendum Section V ASME NQA-1-1997 Quality Assurance Requirements for Nuclear Facility Applications Idaho National Laboratory (INL) 3R- I

~aFmuel PoRey.(I6L STD 7022A Cleanliness Acceptance Levels for Nuclear or Non-Nuclear Service Components

For- 412.09 {Rc,. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 7 of 59 PURDUE UNIVERSITY REACTOR American Society for Nondestructive Test (ASNT)

SNT-TC- I1A (1996 or later) American Society For Nondestructive Testing (ASNT) Recommended Practice 2.1.2 Publications (idaho National Laboratory)

\Cle* )n*Ion Appendix B Welding Requirements and Qualification for Purdue University Fuel Elements

Fonn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 8 of 59 PURDUE UNIVERSITY REACTOR 2.1.3 Drawings ([NL) 635454 Purdue University Test Research and Training Reactor Graphite Reflector Assembly and Source Drive Assembly 635455 Purdue University Test Research and Training Reactor Standard Fuel, Partial,

& Dummy Element Assemblies 635456 Purdue University Test Research and Training Reactor Control Fuel Element Assembly and Dummy Control Fuel Element Assembly 635457 Purdue University Test Research and Training Reactor Fission Chamber Fuel Element Assembly 635458 Purdue University Test Research and Training Reactor Standard Fuel Container Assembly 635459 Purdue University Test Research and Training Reactor Control Fuel Container Assembly 635460 Purdue University Test Research and Training Reactor Irradiation Facility Assembly Wi~aninge~ac~Gap~rdolde ~and F ~ahdtDetaills 635462 Purdue University Test Research and Training Reactor Graphite Container Assembly, and Source Drive Container Assembly 635464 Purdue University Test Research and Training Reactor Container Tube Assembly and Details

Form412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 9 of 59 PURDUE UNIVERSITY REACTOR nOe 635466 Purdue University Test Research and Training Reactor Miscellaneous Details 635467 Purdue University Test Research and Training Reactor Source Drive Nozzle Assembly and Source Drive Top 635468 Purdue University Test Research and Training Reactor Nozzle Preliminary Machined and Fission Chamber Top

3. TECHNICAL REQUIREMENTS 3.1 Production Qualification The supplier is required to qualify the processes or portions of the process or be exempt from same by written approval of the purchaser. In qualification (see def.), only materials that comply with this specification shall be used.

Qualification processes, equipment, and operator qualification/training programs shall be identical to those used during production (see def). To qualify, the supplier must demonstrate, to the satisfaction of the purchaser, that the process is capable of producing a product, which satisfies all the requirements of the specifications. Assembly of production fuel elements shall not be initiated until:

(1) all required data, to assure compliance with the qualification requirements, has been submitted to the purchaser; (2) data and records required by Section 6.3 have been submitted; and (3) written approval of qualification has been received by the supplier from the purchaser.

I1 Flw Pl_____________im F419~ it i -1 o11shai be satisfied)

(ndcy nt nIIeý 11w 65%S'i~izI~tptt* ent ptdititfWilttCL ftrhtavebeenHno

• ~~~~~~~~~~~~{ . . * .. *z.....

N ...

• MK-*:

• • • ... t** ..

Form412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 10 of 59 PURDUE UNIVERSITY REACTOR tAtl Qk~n fic' latmone i rihe supphier meeting ifh rqietli of tfI Is A eIcfication1 I -du e

.P I c%~ d 3.1.2 Fuel Element Qualification:

The supplier shall fabricate 1 dummy standardfitelelement assemblies (see def.) which shall meet the requirements of this specification.

3.1.3 Requalification

The supplier shall notify the purchaser of any proposed process change.

A changed process may not be used in production until the supplier has met all the requirements of Section 3.1.3, submits the results and data of the requalification effort, and receives written approval from the purchaser.

Requalification for any ýipf

,jo fuel element attribute to the requirements of the specification will be required when the processes, materials, fuel loadings, equipment or equipment operators (welding 'a1Wn 1 which have been previously qualified are changed, unless the supplier can demonstrate to the satisfaction of the purchaser by engineering explanation or proof test that such changes will have no detrimental effect on the product.

~peURtoMEfRust U11liy by lot of 0rcsn r!b ) ~~ eW terme 7~~'i

Forn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier:

STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date:

PURDUE UNIVERSITY REACTOR 3.1.4 Operator Qualification:

Operator qualification will be accomplished via an approved supplier internal qualification program for the following operations:

Do, AX rolling.

F. Final machining.

3.1.4.1 In addition to the operations specified above, the supplier shall also show evidence of the training and competency of those individuals who perform any of the following fuel element fabrication and inspection activities:

C. Fuel ýiate/element and component cleaning E. Dimensional inspection of ýp=-s, elements, and subcomponents G. Visual inspection of elements, and subcomponents I.;ýRotýyo in cGt~io oj(gL=ates

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 12 of 59 PURDUE UNIVERSITY REACTOR The individuals perfonming these operations shall have specific requirements imposed on them that will demonstrate their knowledge and ability to perform their respective assignments. Documented evidence of the training of these individuals shall be maintained and shall be made available to the purchaser upon request.

3.2 Materials The material requirements for the components comprising the fuel element are as specified on Drawings per Section 2.1.3 and requirements of this section.

'93U27 Eu141eal MBarn.,

powdeywhich1 nFe 4he1qi~ntiAf J'\ dud~n AluinumW dFrame 3.2 oesalf Sait S tplat0 stc9 se

?on8(I$MI-cadml11 , "dO4tPhitlihiiTLi lumiumwl ille metal>~

All b R03ashalleurdb hesu efator 1

3.2.2 Aluminum Weld Filler Metal:

All aluminum weld filler metal shall be ER4043 as required by Specification AWS A5.1O-1995.

7777 11 T5 MINTTP,FT-1,4'0--117, 5elFe NBTUM e~~y 6@h N',Ldthat mptsiaer ILIhttl~i$t I4r7

Form412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 13 of 59 PURDUE UNIVERSITY REACTOR 3.2.4 Material Requirements All material used or contained in the product shall comply with all the requirements of this specification and Drawings per Section 2.1.3 unless exempted by written document by the purchaser.

3.3 Mechanical Requirements

~3'.3... Fu~lY1~ite Requix~ements plae rJ~h y phrul an i o ifih-21eamng th Mi ac ~ ~ utb o ateoredgýuctio fr-- tt 4ýJ-Rfrm 0 Fmiin l fitrel'L1 I NuType 1,ue L la t-n

Form 412.09 (Rcv. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 14 of 59 2 ________ hjp ~ticieLw ll1ffwedewit f1fnFli fuelavt@i--~

fre hift;p1a9tes as is J e........

n in Ly 1)'ht Ce( aofach J I aLsdtin~by I-I 1) 1~tjti rn I andorn I fIe

' ICiI': de I~

asTkpgudJ1ie 1i "Ieoctn,

" , i aSf~ l"ýld IIIt lIIý/ . Iol ld S 4 Nbe fiifd outarf a fisr{~el %vh iit thin otieft

1. Wý; 1ýýrw thanr O.2T_44 M1ed is atle'1~ E)ea p c~adfo-raI e CCe.

Soffe treati ntcir ace~.h~~

Form412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 15 of 59 PURDUE UNIVERSITY REACTOR M 7 - - " . ' .

-I 'eInean,,yd N* *16men t* IC J lhLMa 2~1/2 -d~dumrn 1 , a es; 3.3.3 Fuel Element Reqiuirements 3.3.3.1 Welding: All welding operations, procedures and methods shall conform with provisions of Appendix B. All welding procedures, welders, or welding machine operators shall be qualified in accordance with Appendix B. The supplier will provide the purchaser certification (see def.) that such is the case prior to production welding.

w 3-`4 Ph j,ýj91 pp _0P ani~isa fuelt elme ali e bs1 1~rSk.ns 0d .2 02,_

ý.4. I -~) e~ t ýhc i 17gi ju e 7, morme_'ýTIi_7 t L 11T,7

_g'-da'dis erse rise I u MJNNI"ý-W A- -etSl7h7RM Eb e ýOt f: 2ý.

IN5

Form 412.09 (Re". 09) iaaao iNational Laboratory SPECIFICATION FOR PURDUE UNIVERSITY STANDARD AND CONTROL FUEL ELEMENTS - ASSEMBLED FOR THE PURDUE UNIVERSITY REACTOR ic i~tin44

Fonn 412.09"(Rev 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 17 of 59 PURDUE UNIVERSITY REACTORet Vkodi h,,:, -' f~ig4~edff elq ]ýVdixigr74TKy e fueVlwei~nt tc shherl 17% gr~arfs ofAŽ

-~50r$ achI "ontpyffi hal cnt%100c~egras fl~ii~ft t4U 23~S ohrol i ~rmAett ohto~dhreu ed~t ni A~ii e t h s w ig htj A 71 3.5 Surface Condition

~uep1~esand completed fuel elements must comply with the surface condition requirements of Section 3.5.1, 3.5.2, and 3.5.3 ,and drawings of Section 2.1.3, pet ANSI B346.1. '%ndng 14~ihtWi hlnffiighu with jA d . _ _ Mt Mhtb S~ e_

3.5.1 Surface Defects 3.5.1.1 Compliance with surface finish and defect requirements shall be established by 100% visual inspection of all fuel Fiaites aWAf elements. Jh tf biTTh*HdiiJ-hl De9~1 f etisn exes of 005inch in ýdepth.IhsT

)i:laCSI surface defects inhe (sedf~6t a'loz fit 0314ff61h , iil0%FICi Nod egi:LadatiI- tl d 49§df0 el Fu..2 el I'ac 4I1 Ike 7J~e f ihrltifi1Qssý, Or3Ccradks1.

_,T II of th"s ý aeCo(IpJ)siLhedjby viua in co fal ie ae nd ,fuel Lfl5..3 HiZefects on-a' de or ensra~re igMOS;e ensi M ,tdU,64 3.5.1.4 Compliance with surface finish and defect requirements shall be established by 100% visual inspection of all fuel

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 18 of 59 PURDUE UNIVERSITY REACTOR element containers. Fuel element containers shall be free of surface defects such as pits, dents, or scratches in excess of 0.010 inch in depth and 0.12 inch in diameter or equivalent area.

3.5.2 Cleanliness

The suppliers fabrication, assembly, and storage areas used for the production of Purdue University fuel elemeuts and/or components shall conform to the requirements of "controlled work area" (see def.) as defined in Paragraph 1.3.6 of INL Standard 7022A. Cleanliness shall be in compliance with INL Standard 7022A, Paragraphs 1.1, 1.2.3, 3.1, 3.2-b, d, i, 3.3 - d, e, 4.1.3, 4.2, and 4.3. Freon shall not be used to clean fuel elements or components.

As determined by Section 4.10 of this specification, there shall be no foreign materials on the -is;;he;;;;; plifo surfaces of the finished fuel elements. e 14ss ai hefioitedlý All components shall be cleaned by a method approved by the purchaser.

NOTEe of each serliesatens fall e nedme cona ounted FEW.

~~irhd~men shllain 1sm aredh and c ý ha r.. Uai4cf thibeslessitha e:ýath cut Tharkeduners S40%jpmp~@

2hu't v eofltamifatin 3.6 Marking NOTE: A ll7e3~ s, futel assemblies, and fuel element containers will be marked per this section.

L6.1__ d~4i4~ ate. shffllJ anhiwnen ______

a ~,cv~n M~natAi~n S inni,,T a4+. t Kpo

%tOLJ4 YL 10ti T')1 1 L 1~fhUL~L tLLtf1Jt

ý.,JtUt WA Q

144 al" tLjý Lýl A

fj WI!Lagio

. -_ I - r , -- - , H" . W 7"ýý5'W" t,

Fonn 412.09 (Rc%'.09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY STANDARD AND CONTROL FUEL ELEMENTS - ASSEMBLED FOR THE PURDUE UNIVERSITY REACTOR 3.6.2 Fuel Assembly Identification:

Each fuel assembly shall have an identifying number such as 07-XX (07 signifying year of fabrication). The number shall be placed on the container assembly as shown on Drawings 635455, 635456 and 635457.

The identification shall be stamped or entered by a method approved by the purchaser, with two inch block characters not in excess of 0.010 inches in depth. Standard assemblies should be labeled: E2, F2, G2, H2, F3, H3, E4, F4, G4, H4, F5, H5. Control assemblies should be labeled: E3, G3, and E5. The fission chamber assembly should be labeled as G5. The source assembly shall be labeled as C3. The spare Standard Assemblies should be labeled: SP-l, SP-2, SP-3. The spare Control Assembly should be labeled as SP-4.

3.7 Storage A I ,j1tes,

.LL. fuel assemblies, and fuel element containers that have received final cleaning per Section 3.5.2 shall be protected in clean polyethylene containers or other containers approved by the purchaser while (1) awaiting final assembly, (2) being transferred into or being maintained in storage, or (3) being prepared for packaging or shipment. Any material exposed to contamination shall be reinspected to the requirements of Section 3.5.

Finn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 20 of 59 PURDUE UNIVERSITY REACTOR 3.

G apter theWiJbmite ;lQeesd Graphite s

hseen RadiationWBakt def.jshal>be aabrcate runir nspc otedbr in issectiona chalbe 3.a9.ng1h Maete p boelmitefilm u dthickne shall lywthe m allthe rq irementsuif this 3Tethe selyuoff th 1esgrap hite r6efleto a6 a i a facitlity400 At7Ž i{Nijel, leffieýbI~lt"ilA I. f~it ~eCOUp~ofi Ac~, epnnrppayer-eere ni frureomfhr ts d 6thefelemty 3.9 Reflectors Graphite anssemble Rnadiation s Bafsets oassemblies.s e aluminum p lca ledrwigs assembliesP(see assembliesshl be)-11011 a)('s shown' onte 3.TGahie reflreetos afwednd Goraphie Radalmetio byAsketsxB r Glorqie raphiteoreflector h assemblies(sed

) and irradiationfacilityasebis(e T emothe graphite reflector assemblies and ify ion fachlas 3.9.1 3.9.4 Mdniiateia  :

Mdniiateia : ces bl fopassem oa oeas shown applicable drawingc6h545 Thasebyothe graphite reflector assembliesav dndtirraiaionfacilitycha ahl also requied for the grpidoftherseflecto assemblie adiradiaiong facility

Fom 41209 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 21 of 59 PURDUE UNIVERSITY REACTOR The graphite reflector shall be labeled as follows: Dl, D2, D3, D4, D5, El, Fl, GI, Hl, 11, 12,13,14, and 15. The irradiation facility assemblies shall have identifying numbers such as IF-X placed on the side of the assembly as shown on drawing 635460. The irradiation facility assemblies shall be labeled as follows: D6, E6, F6, G6, H6, and 16. The identification shall be stamped or entered by a method approved by the purchaser, with two inch block not in excess of 0.010 inches in depth.

3.9.5 Dimensional Inspection:

Verification of all external dimensions of the graphite reflector assemblies and irradiation facility assemblies shall be by 100%

inspection, in accordance with drawings 635454 and 635460. All dimensions of this specification shall apply at a temperature of 75°F +

50.

3.9.6 Surface Finish and Defects:

The graphite reflector assemblies and irradiation facility assemblies shall be free of surface defects such as pits, dents, scratches in excess of 0.010 inch deep and 0.12 inch diameter or equivalent area.

3.9.7 Storage

All graphite reflector assemblies and irradiation facility assemblies shall have received final cleaning and shall be protected in clean polyethylene containers or other containers approved by the purchaser while (a) being transferred into storage, (b) being maintained in storage, or (c) being prepared for shipment or packaging.

4. QUALITY ASSURANCE The supplier shall document, implement, and maintain a quality program in compliance with ASME NQA-I-1997.

The supplier shall permit the purchaser to conduct pre-award and continuing evaluation of the Quality Program.

Personnel performing NDE examinations, P t h &; , liquid penetrant, and visual shall be certified to American Society for Nondestructive Testing (ASNT) Number SNT-TC-l A and certification documentation shall be made available to the purchaser.

Fom 412.09 (Re,. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 22 of 59 PURDUE UNIVERSITY REACTOR Unless otherwise specified, the supplier shall be responsible for the performance of all tests and inspections required prior to submission to the purchaser of any fuel element for acceptance. Provided, however, that the performance of such tests and inspections is in addition to, and does not limit, the right of the purchaser to conduct such other tests and inspections as the purchaser deems necessary to assure that all fuel elements are in conformance with all requirements of this specification. Except as otherwise specified, the supplier may use for inspection purposes either his own or any commercial laboratory acceptable to the purchaser. Records of all tests and examinations shall be kept complete and available to the purchaser. All test and measuring equipment shall be calibrated per the requirements of Standard MIL-C-45662.

-,, .QI s. =eVli.l -A A Thespplie sale topaefid.lmts hisafor te puirchase coviadnsinite per, rn madporys hdeepomintsetabisd%

a ae wi tha turhiaser of inftheadance winforth Sc ioni 6a.7 V Any atrials, or1fuel eementrcom onet hal arlue fbiaed~ usingS 'qIpet The supplier percnnla, or shall prepare procssa1es thatfor hisnoftoi arde use and the purchaser's accordancet~u approval anpreioulyganteod with aprvlsas integrated manufacturing and inspection test plan. The plan shall include all manufacturing operations, equipment and tooling used, inspection requirements and gaging used, and mandatory hold points established by the purchaser.

Any materials or fuel element components that are fabricated using equipment, personnel, or processes that are not in accordance with approvals as previously granted by the purchaser are subject to rejection (see def.). A report of any such incident must be submitted in accordance with Section 6.3.7.

Fuel element inspection for shipment or rejection will be made by the on-site purchaser's representative at the supplier's plant. Final fuel element acceptance will be made by the purchaser at the User's facility.

4.1 Materials Compliance with the material requirements of Section 3.2 shall be established by supplier certification. A "Certification of Chemical Analysis" or a certified mill test report shall be supplied to the purchaser for each lot of material used in the fabrication of fuel elements. This certificate shall give the results of the chemical analysis for the material. All fuel element materials shall be traceable.

For 4 12.09 (R"v. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 23 of 59 PURDUE UNIVERSITY REACTOR Cerernl

~~P FY, The actua bevc41A4 cu~Woi%

aht~ c l'Jit(-)i~jg~f&Th wer

'iq4unit as in ThYIM y ()f"I,U(21.1'757te TýIl

'I,I7LI'l-6 Iu.ý Tý KIdTen )f [ýie'l iiS 7'---- w .e~igt6f debti.i(lWeti tf oe ~onp

Fom 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: F'age: 24 of 59 PURDUE UNIVERSITY REACTOR tt N-2

-Ta'17tdorviacbjv2i

§ý.oas iW6&i

= -$S~jLf..Thnaium fiaVi ______0

~

I% ~ ~02 Ig__ U>¶% T __1_____11

<7 7~Axy aliminirrn ,p §iid Lýotfif~~

~~~~~~~he~~~~~~~~~~ l aeuj~~h tvJ ,ýtnteij~~

oiwn.Qt~i

~fle as specin ed&iVpdon4

.6 rocedure re4irec2ffi Se ioii-.3i. f 1W Fuel Hmg, me4;jt.

7wt bc plied1ift by Laonle- e1 0

ý.adiobrph by cafiatte&nstci~eter m asu ets1- Pýihae aI stan~s~t~yt s ~pt.~ifdl'4~an4 us~byiw enst~ tan41~41FN

Fonn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 25 of 59 PURDUE UNIVERSITY REACTOR go typya"a. I -§ P ofqf`tt6615 lae area to co e tandafdf

~ftg 4I~VI 4 I~ e~d 1 I')]~o~f{&' ~ mgeet 6theit~

11iýslte m~h mtt:_'

cpg~.hal~etennec~y~aIncrah ZIelsIi - fa,)iIo,4aph~nth ds 30 Ot1cieldnst m idcae core~~e d~n~tgmet~r reaing i ih~~ti averag de "nsito er redig .~Aiih fo 73Q%

alcr . A or a 2 0.A.F.f.el cor 4hlbpneateasra:t bwel less n11 ~tn2%othe 1Atdnsl aýyae f Thf111(ýsi? rfor allv.fe loaio ns To ýpmin density ofi6 a onehati W-cllv"area~(Ie 41mm slý,maneuveredg ulnder he dnsitometer Tte ow es around itfIIle $ e avera~ 1-fIyeorephanding is__

~~1(111L th' 1 sspeet spot onteIdornhdh ~

-20%stanatdsF~rthe 30% nd' -20 pmoen~is~eoadInspe-i icomparethernomni 111111,a s taha'to,,f d1slip Iae41: q)re'ae( rtic s desiomte nL rorn nI+%Pi*

EML:eildfthIIinir iiW51and14axTImu"m pe M-sgb1" Lr G ýý~rfa nn~kw cnntMlinnI FPSa 5`?i ý4TMN44h LAm'n,'taKd-6'fd IT 111m bct7 0011 011-AT

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 26 of 59 PURDUE UNIVERSITY REACTOR nam t area: area wi.]

oeaeh felulsvcoren)Ishall notbekweatwtilsnst m.3Wofh aveag el enstyforal cor loain ýt cu, tc 1

()_ýogbon atrea sh;I1l not be lessi fuel T1-densi*ktyjII1ý )W, than.oIt'et~dO ul tlenit'yfor ll fuecl core docatin h eag*~~4jpi)r hnoe ha~lt'iihe In thie longitudinaltdirection1:

Utlce than [lietdogbon& ar'h~eatlci ends,;of uelco7're, any one-htlffnchidante are-,a shal noI Iit beý greatejnILeiI4kfIfd&As Itvtyhan otlI teavri4 4ftidenI(

()I% isi t' )I all3 fuLel core location~s. To 49ctermpi e th5 hIg h dewitv f&one-halrfginch dia)mekcr area[ tile.'flinlms mapu~ une tednstme i helg-,4yaeaut The lowes,,t numrber poss;ibie Is obtaineCd Onl theC deSý~nsitomete Tb's11- n1er-I xjWcorIde&J The foH tteaIFngs;dil" t tkiefne-)n tf' inch frokfit ths pCt Ifd s9mrheTtnAIy(al varoaitd It. TteoaveraLge ocf1T thse-~fIIve readinfit'is' t apaedfto. di 2agd enis Itomei tertreaC~dings for all fueýlgcorie ioca,,tio4F -

toj)aýSembly of, fueJ l~c i~q&;itt element 41 -Core Cojflguatien' Each linish1-ýcut faia fuel plt shl e rad d0 aT-ccordaneith:C iut41 radibograph inspecitions will be perf o1,1Ied w6thoiltwmagnification onl algh tahle'haVimagt(ght iptclnsiW o45 1to QX) 1<tyi'tdles at~ tabLile sCirf~accadt Wcilaredtn&tocitc give a light daiitg& 5465 I'5ofcandles,, f L$2no'a

&&th gjq utbe. wi.h tadiogtajh i~lmtii lac,'i cm> thfabk.

4.rirBnd InteQgrity S61.0.11 Bi'istci~ nei fl4C-hr the fgeil p~late hhiAee hotf I- r6'iled It ,hakb 7(,ated 7 9 F 1'3Žf1" hei1d at tha t'empII)aLOUCC a eio f ours -1 5inmiit&j.m-nute,

+( M il q wŽjd!rffl frae L adaIoeToai ol

Fom 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 27 of 59 PURDUE UNIVERSITY REACTOR E~~ lod.t orht

<to awer470 bPigth 17MRTh110aor

~p~tnQ~ teblisterLind ieri phtaedtorend thail)

OOOhIW'heP41there3/4s 4testitas S,17of ile&ttati of the blist~r, t$4&pt~i*~tf~jec& 4~fte,~a shal btdit'Ped inthe Laps~onp3 4Q Liltrasonk$canning~

Article 5,afi pa elplh~at'Vc & tir f>OtTh§~,

14a7pPT-imtt 52 -nw jj~j2,[

bu-fuej*Nt&e l urival eqt th md0.6 t o tht nch it rdfOr." ofwd en aso e fue plateA cceptancel r cgiteri'at He jiibedcl~e by th s-pli Anti I fei~d by, th purchaser

-L-jv _ _- Aft~ad0t m e, er lo lectedT6jl[1,:

tue Ii~ 0oroe Scin311 will be Isectich inledc iier ch an eth&and

'in(-Ir1 (2ri I" s )-ktat&

I'][)ned atýý 5xora v eorbntanu id &lddmhin e the~~~~~J reurmetafSetos3 ipndg6j W-l~ates sekc,ý-t orT7 detutoits F myberjct eflcte 11i u I-ejec ion Re-1-ject fuel.C plates 7td4ýtý7c mq qvroaL fo tI ,~ tecsts are -1 kfin rjL-W-1&5>TWA4W1LK1-WW.H FA ii,,Exe R!

_,n -e,.'jqcJudiftg V Ifit cul a cfiýc %oids, t -In-Iexccss

,tmiiiiatiq4ý,

o f_tS', IJe Pý i Lul"IciGI-Ang ,9ý 1-11 2,S,09rq9AM Sb, a"'i Or.

L

Fonn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY STANDARD AND CONTROL FUEL ELEMENTS - ASSEMBLED FOR THE PURDUE UNIVERSITY REACTOR 4.8 Surface Finish and Defects Compliance with requirements of Section 3.5 shall be established by visual inspection of all lul!*p ansfuel elements. Out-of-specification defects shall be measured for size and depth and reported to the purchaser.

A ,M'

• eetio~

eiliig te fut~Tte~jiiaficnon~roc r-1--[)IiR"RRlle t4YM T1ýr67de~qstnPWd6

, b1.1kyseisihfic- rfid pae

________________t ~er 1 m~~~~~~~ acofanewth¶Fere j_____________

Fonn 412.09 (R"y. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 29 of 59 PURDUE UNIVERSITY REACTOR 4.10 Cleanliness pitq4, fuel assembly, and fuel element container cleanliness requirements of Section 3.5.2 shall be established by visual inspection without magnification of all fuel plates, fuel assemblies, and fuel element containers.

4.12 Dimensional It shall be the supplier's responsibility to assure that fabrication is performed in accordance with all dimensions delineatedonin 1the Drawings 2.1.3. Noncomplying design dimensions Ts fuel referenced assemblies,inand Section fuel element containers (actual measurements) shall be submitted to the purchaser for review and approval. Any discrepant component shall not be used in a fuel element assembly unless approved.

The supplier is to certify to compliance with the design dimensional requirements delineated in the Drawings referenced in Section 2.1.3.

All dimensions of finished fuelnassemblies l and fuel element containers apply at 75 0 F+5°F.

4.12.1 Final Dimensional Inspection.

Dimensions required by this specification and drawings of Section 2.1.3 shall be inspected using a purchaser approved sample plan and recorded on an inspection sheet with "in specification" dimensions recorded by check mark, "OK," or actual measurements and 'out of specification' dimensions recorded as actual measurements.

4.13 Reactor Components and Spare Fuel Element Parts Reactor components and spare fuel element parts not assembled into fuel element assemblies are required to be certified. The certification shall consist of material certification, fabrication verification, and supplier certificate of compliance to the specification and drawing requirements. The certification documents shall be submitted to the purchaser and user.

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 30 of 59 PURDUE UNIVERSITY REACTOR

5. PACKAGING AND SHIPPING Packaging and shipping of the fuel elements shall be performed using a Purchaser approved procedure in compliance with this section.

The purchaser shall provide shipping containers to protect the fuel elements from damage during shipment and which conform to the applicable requirements of the Departments of Energy and Transportation, and other regulatory agencies having jurisdiction of the shipment of radioactive materials. Re-useable shipping containers will be returned to the Supplier by the User at the Purchaser's expense.

" The Supplier is responsible for loading the fuel elements into shipping containers in a sealed polyethylene sleeve in a cleaned dry condition and free of extraneous materials.

• The Supplier shall take necessary precautions during packaging to prevent damage to the fuel elements during shipment. Each container shall be provided with a tamper-proof seal. Loading and shipping documents for the container shall be prepared in accordance with the applicable regulatory requirements.

• The Supplier shall make arrangements for shipment to the User.

6. NOTES 6.1 Definitions For the purpose of this specification, the following terms are identified:

fra~~~&a t x1o7gaco ngmn Certification.The action of determining, verifying and attesting in writing (signed by a qualified party) to the qualifications of personnel and material.

Control Fuel Element Assembly. An assembly consisting of the control fuel element container Me pj".

Controlled WorkArea. A work area to which access of personnel, tools, and materials is limited and physically controlled. Temporary enclosures may be used where adjacent activities produce contamination which is detrimental to the job.

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 31 of 59 PURDUE UNIVERSITY REACTOR I Development. A determination of processes, equipment, and parameters required to produce a product in compliance with this specification.

ooe~e.'kngofthi ýefl:ercdreusua"1'T71jjbsth>-IT li.4)'h whichj joitesulg 1cI in d mgw II"*those*..e**",as%..

Dummy Fuel Element Assembly. An assembly consisting of a fuel element container with unf eled~iiuL.dIiWh Du m y ad Plat&' Al no n-ffie iJled rar 1 j N _ t mI specifled,Žin, Ths ocme finlished fil1'plat&,ý before'li any ."fra 1\ pdhi1i&A tth64 tsq~ie jd~jo Failure.A condition where the fabrication process appears to be out of control or a breakdown or damage to equipment creates excessive costs and/or schedule delays.

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hot-liio14an in(1t111ol el 121late<.he mac Iaeerilch' ed II tcVr cm pII tc*kto Fuel Assembly. An assembly of fuel plates and hardware components. This includes both the standard and control fuel elements.

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GraphiteReflector Assemblies. A component consisting of a graphite container assembly with a graphite block inside.

In-Process Controls. Inspections and tests made during production to ensure that the manufacturing processes, equipment, and personnel are producing a product meeting specified requirements.

For), 412.01) (Rem.09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 32 of 59 PURDUE UNIVERSITY REACTOR FT a~ý uni m P Tot tq ......... c =6 Manufacture(ing). All fabrication, assembly, test, inspection and quality control processes. Fabrication is a synonym for Manufacture.

Procedure. The detailed description of the series of processes during manufacture and inspection, which follow a regular definite order (not to be construed as an outline).

Production. That phase of the program, following Qualification, during which the product is in Manufacture.

Purchaser.Idaho National Laboratory (INL).

Qualification. Aed monstration that the Manufacturing process, equipment and personnel can produce a Product in compliance with this Specification.

Quality Control. The sampling plans, inspections, tests and records required and used during Production to assure that the Product is in compliance with this Specification.

Rejection. Materials, parts, components, or assembly products, which will not be accepted as fulfilling the contract requirements because of noncompliance with this Specification.

Requalification. A demonstration that a single or group of manufacturing processes, equipment and personnel can produce a product in compliance with this specification after the original qualification has been completed and becomes invalid.

-ep~ iont YTRk 4.'i i ig fr Specification. All parts and appendixes to this document, its references, drawings, and standards, as may be modified from time to time by contractual document.

StandardFuel Element Assembly. An assembly consisting of the fuel element container with fourteen (14) fuel plates.

Fonr 412.09 (Rcv. 09)

Supplier. The primary vendor selected by INL to manufacture theproduct.

User. Purdue University, at West Lafayette, Indiana.

6.2 Purchaser Tests None 6.3 Submittals The following data and records shall be suppl ied to the purchaser in the quantities stated. The purchaser's approval, prior to implementation, is required on those marked with an asterisk. All records and data shall be maintained by the supplier for the duration of the Purdue University fuel element contract.

The granting of approval by the purchaser of design, working drawings, specifications, requests, and other technical data submitted by the supplier under the provisions of the subcontract or specification shall not affect or relieve the supplier from such responsibility as the supplier has with respect to adequacy or correctness of the design, working drawings specifications, reports, and other technical data.

6.3.1 Preproduction

Documents requiring approval must be submitted prior to production use. The number of copies shall be as specified by the Vendor Data Schedule. These documents include:

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Fonr 412.09 (Rcv. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 34 of 59 PURDUE UNIVERSITY REACTOR

• dxed aecip'o aý toTem e'huk s7 i use ovi4

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• All fabrication, assembly, cleaning, surface treating, handling, and decontamination procedures (not to be construed as an outline)

• All production test, inspection, and quality control procedures, including all nondestructive and destructive tests and all standards and sampling section drawings. All data from these tests, including but not limited to: 1Ia*

ap.,

• 1tletmg PM~Es1 , (!LAnIIafý'.i MIA Iat&;

- *All packaging, storage and shipping procedures

- *Integrated manufacturing and inspection test plan.

6.3.2 Pre-repair

• All repair programs and procedures prior to use.

6.3.3 Manufacturing Schedule:

• A schedule using a purchaser approved technique.

6.3.3.1 Reports.

I. Biweekly qualifications phase summary status report. The first such report shall be initiated 1 month after date of contract award.

2. Three (3) copies of a monthly report detailing program progress against a previously submitted schedule shall be supplied by the supplier to the purchaser. Report type, format and submittal schedule shall be as agreed upon between the purchaser and supplier.

6.3.4 Delivery Submittals:

Three copies (except as noted) of the following data and records shall be sent prior to or accompany the shipments. The supplier shall maintain copies of these records for at least 10 years and until the supplier has received written approval from the purchaser for disposition or disposal:

Certification of product compliance to the requirements of this specification to include any test data pertaining thereto

lorm 412.09 IRev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 35 of 59 PURDUE UNIVERSITY REACTOR

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- List of all applicable'waivers and deviations and related 411 r fuel elements.

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is -~4 6.3.7 Report of Production by Unavoroved Process:

Whenever the supplier's previously submitted and approved process control limits are exceeded, or any material or fuel element components are fabricated using equipment, personnel, or processes which are not purchaser approved, the time, nature, description, corrective action to be taken, and proposed further corrective action shall be reported

Forn 412.09 (Rn,-. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 36 of 59 PURDUE UNIVERSITY REACTOR immediately by the supplier, with a written report to the purchaser to follow within 10 working days.

Fornm42.09 (Re'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 37 of 59 PURDUE UNIVERSITY REACTOR TL ra'e=RtngFu' .Q-0 Fn~

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Forn,412.09 (Re*-.09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 38 of 59 PURDUEUNIVESITY EACTO R;.piremeuntsfo RI:W V1'ie~es ti 1.

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T e ~a~be~1~a mageeiit~e ___ ~ eoamn stltfc n*,C(aks wtt~ )i~ft thi area tMiV -El:re rie T1 ý'ial 'MFý Appendix A

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 39 of 59 PURDUE UNIVERSITY REACTOR

~Plate lot number-

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Appendix A

Fonn 412.09 (Rev. 0'))

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 40 of 59 PURDUE UNIVERSITY REACTOR APPENDIX B Welding Requirements and Qualification for Purdue University Fuel Elements

1. Scope The requirements for welding and for the evaluation of welds applicable to the Purdue University Fuel Element Container and components are established by this Appendix.

1.1 Application. This document defines requirements for the following:

1.1.1 Welding procedure qualification.

1.1.2 Performance qualification of welders, welding equipment, and special fixturing.

1.1.3 Information to be included in welding procedure specifications.

1.1.4 Application of qualified procedures to production welding.

1.1.5 Destructive testing and nondestructive examination for qualification and for production welding.

1.2 Special Limitations for Applicability. The requirements contained in this appendix are to some degree based on RDT F6-2T. Those requirements applicable to Manual, GTAW, single pass, welding of Plug Joint welds, Comer Joint welds, and Partial Penetration Butt Joint welds have been included in this appendix. The introduction of a new weld design or weld process requiring a change in these limited parameters would require an appropriate review of RDT F6-2T for requirements applicable to the new parameters.

1.3 Definitions.

Arc Strike. Any localized melting, heat affected zones, or change in the contour of the surface of the finished weld or adjacent base metal resulting from an arc or heat generated by the passage of electrical energy between the weld or base metal and a current source; such as welding electrodes, electron beams, ground clamps, high frequency arc, etc.

Automatic Welding. Welding with equipment which performs the entire welding operation without constant observation and adjustment of controls by an operator.

The equipment may or may not perform the loading and unloading of the work.

Appendix B

Fomi 412.09 (Rev. G9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 41 of 59 PURDUE UNIVERSITY REACTOR Face of Weld. The exposed surface of a weld on the side from which welding was done.

Face Reinforcement. Reinforcement of weld at the side of the joint from which welding was done.

Heat. A single homogeneous melt of metal or alloy.

Joint Penetration.The minimum depth a groove or flange weld extends from its face into a joint, exclusive of reinforcement.

Machine Welding. Welding with equipment which performs the welding operations under the constant observation and control of an operator. The equipment may or may not perform the loading and unloading of the work.

Position of Welding. The terms related to positions of welding for joint types and welding processes and the position limits are defined in Section IX, ASME Boiler and Pressure Vessel Code.

Repair. The process of restoring a nonconforming item characteristic to an acceptable condition, although it does not conform to a specified requirement.

Rework. The process by which a nonconforming item is made to conform to specified requirements.

Root of a Joint. That portion of a joint to be welded where the members approach closest to each other. In cross section the root of the joint may be a point, a line or an area.

Root of a Weld. The points, as shown in cross section, at which the back of the weld intersects the base metal surfaces.

Root Penetration.The depth a groove weld extends into the root of a joint measured on the centerline of the root cross section.

Root Reinforcement. Reinforcement of weld at the side opposite that from which welding was done.

Root Surface. The exposed surface of a weld on the side opposite that from which welding was done.

Size of a Groove Weld. The joint penetration (depth of chamfering plus root penetration when specified).

Appendix B

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 42 of 59 PURDUE UNIVERSITY REACTOR Undercut.A groove melted into the base metal adjacent to the toe or root of a weld and left unfilled by weld metal.

Underfill. A depression on the face of the weld or root surface extending below the surface of the adjacent base metal.

Welder and Welding OperatorPerformance Qualification.The tests to demonstrate a welder's or welding operator's ability to produce welds meeting prescribed standards.

Welder. One who is capable of performing a manual or semiautomatic welding operation (sometimes erroneously used to denote a welding machine).

Welding Operator.One who operates machine or automatic welding equipment.

Welding ProcedureQualification. The test to demonstrate that welds made by a specified procedure can meet prescribed standards.

Welding Procedure Specification. A written welding procedure which specifies the detailed methods and practices to be used in the production of a weldment and how they shall be carried out. A specification includes all elements of a procedure necessary to produce a satisfactory weldment. Examples of some of the elements included in a specification are: material used, preparation of base

• materials, preheat and postheat cleaning, assembly method and sequence, fixturing, heat treatments, joint welding procedures, preweld and postweld nondestructive examinations, repair, rework, etc.

Welding Procedure.The detailed methods and practices including all joint welding procedures.

2. Reference Document The following documents are a part of this appendix to the extent specified herein. The issue of a document in effect on the date of the invitation to bid, including any amendments also in effect on that date, shall apply unless otherwise specified. Where this appendix appears to conflict with the requirements of a reference document, such conflict shall be brought to the attention of the purchaser for resolution.

Appendix B

Form 412.9 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 43 of 59 PURDUE UNIVERSITY REACTOR 2.1 American Society for Testing and Materials (ASTM) Standards 2.1.1 ASTM E2, Preparation of Micrographs of Metals and Alloys 2.1.2 ASTM E3, Preparation of Metallographic Specimen 2.2 American Society of Mechanical Engineers (ASME) Codes 2.2.1 ASME Boiler and Pressure Vessel Code,Section IX, Welding Qualifications 2.3 American Welding Society (AWS) Standards 2.3.1 AWS A2.2, Nondestructive Testing Symbols 2.3.2 AWS A3.0, Terms and Definitions

3. Weld Qualification Requirements 3.1 General Requirements Appendix B

Form 412.09 (Rc. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier:' SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 44 of 59 PURDUE UNIVERSITY REACTOR 3.1.1 All welding procedures, welders, or welding machine operators shall be qualified in accordance with the provisions identified in this Appendix.

3.1.2 Weld Procedure and Performnance Qualification Testing previously qualified to these requirements under other contracts may be used.

Existing records to support previously qualified procedures and personnel are subject to review by the purchaser.

3.1.3 Base materials and filler material shall comply with the requirements of the drawings.

3.1.4 Welding processes which satisfy the specified requirements and produce the quality required by this Appendix are permissible. Welding procedures which utilize fluxes and coatings shall not be used.

3.1.5 Fixtures

The capability of fixtures for aligning parts shall be demonstrated before welding of production parts is initiated. If chill bars or blocks are used, the type of material and their location with respect to the joint shall be included in the procedure specification.

3.1.6 Position of Qualification Welds. All procedure and performance qualification test welds shall be made in the same positions as for production welds.

3.1.7 Special Conditions for Qualification Welds: All procedure and performance qualification test welds shall be made under conditions which simulate the actual production welding conditions. These conditions shall include space limitations, joint accessibility, degree of comfort due to heat, position and other handicaps or environmental factors which the welder or welding operator will endure during actual production welding.

3.1.8 Heat Treatment. Weld preheat and postheat treatments shall not be used without prior approval by the purchaser.

3.1.9 Interpass Temperature. For multi-pass weld, the weld interpass temperature shall not be less than 60' F or greater than 3500 F without prior approval by the purchaser.

3.1.10 Records. Records of welding, associated processing, and inspection shall be maintained for all welds. Complete records may consist of inspection forms, routings, or reference to Operating Procedures or other documents. These records shall include at least the following:

Appendix B

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 45 of 59 PURDUE UNIVERSITY REACTOR

1. Base Material (Type, material specification, heat or lot number).

2. Filler Material (Type, material specification, heat or lot number).

3. Cleaning procedures,

4. Joint identification and weld maps when applicable.

.5. Welding machine type and identification.

6. Welding procedure specification.

7. Welder or welding operator qualification.

8. Procedure and performance qualification.

9. Current-voltage data for machine or automatic welding.

10. Date welds are made.

11. Inert gas analysis, when applicable.

12. Nondestructive examination procedure.

13. Nondestructive examination personnel identification.

14. Examinations and tests (nondestructive and destructive) and the results.

15. Photomacrographs and photomicrographs.

16. Metallographic specimens.

17. If applicable, rework and repair of welds.

18. Disposition of welds.

3.2 Welding Procedure Specification 3.2.1 The welding procedure specification shall meet the requirements of this Appendix, and shall be submitted to the purchaser for approval prior to any production welding.

3.2.2 The welding procedure specification shall include all essential elements and details, as required by this section, to cover each joint to be welded by the supplier. Each joint shall be identified in the welding procedure Appendix B

Fonn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 46 PURDUE UNIVERSITY REACTOR specification. The specification shall include a joint design sketch for each joint welding procedure even if the specification references drawing numbers.

3.2.3 The following basic information and essential variables shall be included in sufficient detail to assure that compliance with the requirements of the specification can be verified:

1. Basic Information

a. Joint Design: (the joint geometry, fit-up, and other required dimensions of the welded joint) tolerances and material thickness.

b. Method of arc initiation

c. Electrode size (for gas tungsten arc welding)

d. Gas type and flow rate (shielding and backing gas)

e. Welding current range for manual welding

f. Whether tack welds or fixtures are used for assembly of the joint for welding

g. Method and frequency of cleaning

h. Number of weld layers and passes

i. Whether stringer beads or weave beads are used

2. Essential Variables

a. General, All Welding Processes.

i. A change from a base material type or grade (materials of the same nominal chemical analysis and mechanical property range, even though a different product form) to any other base material type or grade. When joints are made between two different types or grades of base material, a procedure qualification shall be made for the applicable combinations of materials, even though procedure qualification tests have been made for each of the two base materials welded to itself.

Appendix B

Fonn 412.09 (Re. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 47 of 59 PURDUE UNIVERSITY REACTOR ii. A change of filler metal type or classification to any other type of classification iii. A change in welding position.

iv. A change in vertical welding direction, i.e., from upward to downward or vice versa.

v. The addition or omission of integral backing (e.g.,

"butt-lap" type joint).

vi. The addition or omission of nonfusing metal retainers.

vii. The addition or omission of filler metal to the joint.

viii. Any change in the method by which filler is added, such as preplaced shims, preplaced wire, preplaced consumable inserts, wire feed, or prior weld metal surfacing ("buttering") of one or both joint faces.

ix. The addition or omission or any type of preplaced consumable inserts or joint surfacing.

x. A change in the shape or size of preplaced consumable inserts or joint surfacing.

xi. A change from multiple pass welds to single pass welds.

xii. The omission of inert gas backing during welding, except that requalification is not required where a qualified welding procedure is changed to omit the inert gas backing and then is used only for a single welded butt joint with a backing strip, or a fillet weld. For multiple pass welding, the omission of inert gas backing during welding until three layers or 3/16 of weld metal thickness has been deposited, whichever is greatest.

xiii. A change from one welding process to any other process or combination of welding processes.

b. Manual Welding, All Welding Processes.

Appendix B

Form 412.09 (R-v. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 48 of 59 PURDUE UNIVERSITY REACTOR

i. An increase in the standard size of filler metal from that stated and qualified in the procedure specification.

ii. A change in joint geometry which violates the tolerances given for the joint geometry elements listed below:

Bevel Angle: State in procedure specification.

Tolerance: Minus 5%.

Groove Angle: State in procedure specification.

Tolerance: Minus 5%

Alignment Tolerance: Assign value in procedure specification. Qualify procedure for single welded joints using maximum permissible misalignment in a portion of the joint.

c. Gas Tungsten Arc Process.

i. A change of electrode material type.

ii. A change in arc starting methods.

iii. A change from a single shielding gas to any other shielded gas or to a mixture of shielding gases or a change in specified composition of gas mixture.

iv. A decrease in shielded gas flow rate of more than ten percent.

Appendix B

Form 412.09 (Re. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 49 of 59 PURDUE UNIVERSITY REACTOR 3.3 Welding Procedure Qualification 3.3.1 The welding procedure shall be qualified to the requirements of this section.

3.3.2 All welding used in qualifying a welding procedure shall be performed in accordance with a welding procedure specification.

3.3.3 Before any welding is performed on production components, the supplier shall qualify each proposed welding procedure by:

I. Recording all essential elements of the welding procedure in a welding procedure specification (see Section 3.2)

2. Verifying the welding procedure specification by welding test specimens representing each joint to be welded in production and performing nondestructive examination and destructive tests in accordance with the requirements of this Appendix.

3. Submitting to the purchaser, for approval, the welding procedure specification and a certified copy of the detailed results obtained from the tests performed on the test welds. The metallographic sections required by this Appendix shall also be submitted to the purchaser.

Appendix B

Form 412.09 (Re,. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 50 of 59 PURDUE UNIVERSITY REACTOR 3.3.4 Essential Variables. The welding procedure shall be set us as a new welding procedure specification and shall be completely re-qualified when any of the changes listed in Section 3.2.3.2 are made in the procedure.

3.3.5 Chart Recordings. Current-voltage-time charts shall be used for each procedure qualification weld for automatic or machine welding.

Calibrated current and voltage indicating meters may be substituted for trace chart type equipment for manual welding. The current and voltage ranges shall be recorded for manual welding.

3.4 Welder Performance Oualification 3.4.1 Performance qualification weld tests shall meet the requirements of this section, except that any welder used to qualify the welding procedure shall also be considered qualified and additional performance weld tests are not required.

3.4.2 General.

I. The performance qualification tests are intended to determine the ability of welders to make sound welds.

2. The performance test may be terminated at any stage of the testing procedure whenever it becomes apparent to the supervisor conducting the tests that the welder does not have skill required to produce satisfactory results. In this event, the welder may be re-tested at the discretion of the supplier in accordance with 3.4.3.

3. Each supplier shall maintain a record of the procedures, including the essential variables, under which welders are examined and the results of the examinations.

3.4.3 Qualification of Welders.

I. Each welder shall pass the tests prescribed for procedure qualification except that tensile tests are not required. The essential variables and the test results obtained by each welder shall be recorded in a Performance Qualification Test Report. Any welder who performs acceptable welding procedure qualification tests shall be considered qualified.

2. Renewal of Qualification. Requalification of a welder is required when:

Appendix B

Form 412.09 (Rcv. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 51 of 59 PURDUE UNIVERSITY REACTOR

a. 90 or more days have elapsed since he last produced acceptable welds using the specific welding process, or

b. He has not perfon-ned acceptable welds using the production welding procedure.

c. Any time there is a specific reason to question a welder's ability to make welds meeting the requirements of this Appendix, requalification shall be required. Only one test weld shall be required for renewal of qualification. If this test weld fails to meet all of the original requirements, then a complete performance requalification shall be required.

3.4.4 Chart Recordings. Current-voltage-time charts shall be used for each procedure and performance qualification weld for automatic or machine welding. Calibrated current and voltage indicating meters may be substituted for trace chart type equipment for manual welding. The current and voltage ranges shall be recorded for manual welding.

3.5 Welding Machine Qualification 3.5.1 Performance qualification weld tests shall meet the requirements of this section, except that any welding machine used to qualify the welding procedure shall also be considered qualified and additional performance weld tests are not required.

1. The performance qualification tests are intended to determine the ability of welding machines to make sound welds.

2. Any time there is a specific reason to question a welding machine's ability to make welds meeting the requirements of this Appendix, requalification shall be required. Only one test weld shall be required for renewal of qualification. If this test weld fails to meet all of the original requirements, then a complete performance requalification shall be required. Welding machines used for the manual welding of any successful procedure or welder qualification tests shall be considered qualified for manual welding of all core components covered in this Appendix.

Appendix B

Forn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382

.STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 52 of 59 PURDUE UNIVERSITY REACTOR 1 3.5.2 Welding machines used for the manual welding of any successful welder performance qualification tests shall be considered qualified for manual welding of all components covers in this session.

3.6 Examination & Tests 3.6.1 Type of Test Required. The following tests shall be used for the qualification of welding procedures and / or welders as applicable:

1. Nondestructive examination by a liquid penetrant method.

2. Nondestructive examination by Visual to test for soundness and surface characteristics of the weld.

3. Destructive examination by sectioning for metallographic examination of weld joints and adjacent areas to test for fusion, weld geometry, weld reinforcement, and soundness of the weld.

4. When the purchaser has reason to believe that the quality of any weldment is doubtful, he may require additional inspection.

5. Nondestructive Examination and Tests

a. Visual. The test weld shall be examined visually prior to welding and after welding in accordance with Section 5.1

b. Liquid Penetrant. The test weld shall be examined after the final layer in accordance with Section 3.6.2.2 using a color contrast method.

c. Unless otherwise specified, inspection of procedure and performance qualification welds shall be performed in the final surface condition.

6. Destructive Examination. Each test weld shall be sectioned transversely to metallographically examine a minimum of:

a. Three section faces for welds on cylindrical components less than 1'/ inch in diameter or for welds that are one to four inches long on non-cylindrical components.

b. Four sections faces for welds in cylindrical components that are greater than 11/4 inch in diameter or for welds that are greater than four inches long on non-cylindrical components.

Appendix B

Form412.09 (Rcv. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382]

STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 53 of 59 PURDUE UNIVERSITY REACTOR

c. One section face for plug welds, arc spot welds, and welds that are less than one inch long on non-cylindrical components.

d. The cross section shall be polished and etched to provide clear definition of the structure in the fusion zone and the heat-affected zones.

e. For welds in (a) and (b) of this paragraph, one cross section shall be made through a weld start and a weld stop area and the remaining sections shall be made at random. For weld described in (c) of this paragraph, the cross section shall be made at the approximate centerline of the weld.

Examination of the welds shall be in accordance with Section 3.6.2.3.

3.6.2 Acceptance Criteria for Oualification Test Welds

1. Visual Examination. Visual examination shall be in accordance with Section 5.1.

2. Liquid Penetrant Examination. Unless otherwise specified, final weld surfaces shall be examined using a color contrast method.

a. For welded joints in materials less than 1/8 inch thick the following relevant indications are unacceptable.

i. Any cracks.

ii. Linear indications.

iii. Indications with dimensions exceeding 1/64 inch.

iv. Rounded indication separated by 1/4 inch or less edge-to-edge.

v. Five or more rounded indications in any six square inches of weld surface with the major dimension of this area not to exceed six inches with the area being taken in the most unfavorable location relative to the indication being evaluated.

b. For all welds in materials 1/8 inch thick or greater, the following relevant indications are unacceptable. (Only Appendix B

Fomr 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 54 of 59 PURDUE UNIVERSITY REACTOR those indications with major dimensions greater than 1/64 inch are considered relevant for item (iii).)

i. Any cracks.

ii. Any linear indications.

iII. Rounded indications with dimensions exceeding 10 percent of the nominal weld thickness or 1/8 inch, whichever is smaller. Rounded indications separated by 1/16 or less edge-to-edge shall be evaluated as a single indication.

iv. Four or more rounded indications in a line separated by 1/16 inch or less edge-to-edge.

v. Six or more indications in any six square inches of weld surface with the major dimension of this area not to exceed six inches with the area taken in the most unfavorable location relative to the indications being evaluated.

vi. Aligned indications in which the average of the center-to-center distance between any one indication and the two adjacent indications in a straight line is less than 3/16 inch.

3. Metallographic Examination Metallographic examinations shall be performed on qualification test welds at not less than 50X on test welds as required in this Section in accordance with ASTM E.2.

Any cross section which is shown by metallographic examination to contain any of the following relevant defects shall be cause for rejection of the test welds.

a. Any cracks.

b. Incomplete fusion, or insufficient joint or root penetration.

c. Any tungsten inclusions, slag inclusions, or porosity having a maximum dimension greater than 20 percent of the weld thickness or 1/32 inch, whichever is smaller.

d. More than four tungsten inclusions or pores which have a maximum dimension less than in (c) above.

Appendix B

For, 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 55 of 59 PURDUE UNIVERSITY REACTOR I

e. Any deviation from specified weld geometry or weld reinforcement.

3.6.3 Test Welds.

I. Procedure and / or welder performance qualification shall be made on test welds which duplicate the production weld joint type and which simulate the conditions to be used in production with respect to orientation, the essential variables listed in Section 3.2.3.2, and the dimensions of the parts to be joined to the extent that they affect heat requirements, relative motions, and distortions. All welding used in qualifying a welding procedure and / or welder performance shall be performed in accordance with the procedure specification.

2. For manual welding, two consecutive test welds shall be made when the weld joint is less than six inches in length. Only one test weld shall be required when the weld joint is 6 inches or greater in length.

3. All test welds shall be tested using the required tests listed in Section 3.6.1. To qualify the procedure specification used in making the test welds, each weld shall pass the required tests.

4. Repair of procedure or performance qualification test weld(s) is prohibited.

4. Production Welding All production welding shall be accomplished using approved welding procedure specifications and qualified welders and/or welding operators.

5. Quality Acceptance of Production Welds 5.1 All completed production welds shall be visually examined in accordance with the following requirements:

5.1.1 General Visual Inspection Requirements. All visual examinations shall comply with the following:

1. Visual examination shall be made under direct daylight-type fluorescent lighting of at least 100 foot-candles at the work examination area.

Appendix B

Fonn 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 56 of 59 PURDUE UNIVERSITY REACTOR

2. Visual examination shall be performed with the aid of a 5x (minimum) magnifying glass.

3. The inspection required by this standard shall not be performed by the welder who made the welds. However, if the welder is qualified in accordance with this standard, he may visually inspect his own welds prior to the inspections required by this standard.

4. Personnel performing visual inspection shall have 20-20 vision, natural or corrected, stereo acuity, and shall not be color-blind.

5.1.2 Visual Acceptance Criteria (except for porosity). Visual examination of weld joint preparations and welds shall be performed in accordance with the following requirements to verify conforlnance to the written welding procedure, the design requirements, and the requirements of this standard:

1. Prior to welding, the weld joint edges and adjacent surfaces shall be examined for:

a. Proper edge preparation, dimensions, and finish.

b. Alignment and fitup of the pieces being welded.

c. Verification of correct material by check of records.

d. Verification of the cleanliness requirements.

2. After welding, the joint shall be examined in the final surface condition for:

a. Contour, reinforcement and surface finish of welds.

b. Degree of underfill, undercut, and overlap.

c. Arc strikes, weld spatter and impression marking.

d. Bum-through and fuse-through

3. Weld joints and surfaces which are shown by visual examination to have any of the following defects or areas of nonconformance are unacceptable:

a. Any nonconformance revealed by 5.1.2.1.

Appendix B

Form411.09 (Rnv.09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 57 of 59 PURDUE UNIVERSITY REACTOR I

b. Any zone of incomplete fusion.

c. Insufficient joint or root penetration.

d. Any undercutting, underfill, or bum through.

e. Any concavity on the face side of groove welds.

f. Any arc strikes, weld spatter, and impressionmarking.

g. Any visible inclusions, porosity, cracks, and unfilled craters.

4. Machined welds shall meet the drawing requirements.

5. All welds shall be free from surface markings resulting from mishandling, punching, scratching, etc., which exceed the specified surface requirements.

6. All welds shall be free of dross, or slag.

7. All welds shall be free of oxidation due to improper shielding and overheating which produce black or gray spalling or loose particles. Iridescent temper films and the dark metallic vapor deposits which may occur adjacent to the welds are acceptable.

These films and deposits shall be removed by approved cleaning procedures when accessible.

5.1.3 Visible unacceptable porosity is as follows:

1. Four or more pores with a major dimension of 0.048 inches or more randomly positioned.

2. A single pore with a major dimension of 0.064 inches or more.

3. Six or more pores with a major dimension of 0.0 16 inches or greater in one weld.

4. Four or more porosity with a major dimension of 0.016 inches or greater, in line separated by less than 0.063 inches from edge to edge.

5.2 Repair of a defective weld by welding shall be limited to two attempts.

Unacceptable defects shall be removed and re-examination made using liquid penetrant color contrast method to assure complete removal of the defect. If the Appendix B

Fonn 412.00 (R,,. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: Page: 58 of 59 PURDUE UNIVERSITY REACTOR removal of the defects results in reducing the thickness of the weld metal below the thickness of the base metal, the area shall be rewelded using a welding repair procedure which has been approved by the Purchaser. Whenever a defect is removed and subsequent repair by welding is not required, the excavated area shall be blended into the surrounding surface to remove any sharp notches, crevices or comers. Completed repairs shall be visually re-examined per Section 5.1. Records shall be maintained on all repairs and shall include the following:

5.2.1 Location ofjoint.

5.2.2 Location of defect.

5.2.3 Description of defect, including type and size.

5.2.4 Reference to approved repair procedure.

5.2.5 Inspections before and after repair and the results thereof.

5.2.6 Identification of repair welders or welding operators.

Appendix B