ML23086C062

Enclosure 1: GEH Supplemental Information for Ntr License Renewal Application

ML23086C062
Person / Time
Site:	Vallecitos Nuclear Center
Issue date:	02/28/2022
From:	GE-Hitachi Nuclear Energy Americas
To:	Office of Nuclear Reactor Regulation
Shared Package
ML23086C060	List: ML23086C061 ML23086C062 ML23086C063
References
M230049
Download: ML23086C062 (1)

Text

GEH Supplemental Information for NTR License Renewal Application DOCUMENTPURPOSE:

ThisdocumentisanaccompanyingdocumenttotheGEHitachisubmittalofsupplemental informationrelatedtorenewaloftheR33operatinglicensefortheGEHNuclearTestReactor.

RenewaloftheNTRlicensereflectsnarrowingofthemissionofboththeNTRandtheVallecitos NuclearCenter(VNC).Thisnarrowingreflectsgeneralmarketplacedemandsforservicesas wellassubstantialchangesinotherNRClicensesmaintainedattheVNC.Thisroadmap documentidentifies,explains,andprovidesjustificationforsignificantchangestotheNTR licensingbasis.ChangestobemadetotheSARandTechnicalSpecificationssubmittedtothe NRCareannotatedinbluefont.

Contents

1.0 ProjectedOperatingCostsandSourceofFundingfortheNTR........................................................1 2.0 DecommissioningFinancialEstimates:.............................................................................................3 3.0 Facility%DevotedtoClass103CommercialActivities:...................................................................4 4.0 Facility%DevotedtoEnergyProductionActivities:.........................................................................4 5.0 AmendedSecurityPlanaddressing10CFR73.67(e):.......................................................................5 6.0 Primarycoolantconductivitymetercalibrationchangedtobiennial:.............................................7 7.0 SARFigure1319revisedandexplained:..........................................................................................8 8.0 CorrecttypoinTS3.7.1.2:...............................................................................................................10 9.0 UpdateSARtoreflectthermocouplepileasabandonedinplace:.................................................10 10.0 AdditionalSARChapter13accidentanalysis:.................................................................................12 11.0 ClarifyTermWorseCoreGeometry:...........................................................................................18 12.0 Fuelmeltingtemperature:..............................................................................................................18 13.0 ClarifytermsinChapter13toeliminateambiguitybetweencladdingandfueltemperatures:....19 14.0 ClarificationofTechnicalSpecification1.2.20.2,definitionofReactorSecured:...........................20

1.0 ProjectedOperatingCostsandSourceofFundingfortheNTR Pursuant to letter, D. A. Hardesty (NRC) to M.F. Feyrer, GE-Hitachi Nuclear Energy Americas LLC (GEH) - Receipt and Supplemental Information Needed for Renewal Review of GEH Nuclear Test Reactor Facility Operating License No. R 33 (EPID L-2020-RNW-0038), dated 03/11/21 (ML21062A250), the NRC requests, pursuant to 10 CFR 50.33(f)(2):

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a. Projected operating costs of the GEH NTR for each of the fiscal years 2021 through 2026 (the first five-year period after the projected license renewal).

b. GEH NTRs source of funding to cover the operating costs for the above fiscal years.

Response

a. As required by 10 CFR 50.33 (f)(2), GE-Hitachi Energy Americas, LLC (GEH),

as a subdivision of the GE Power business segment, is providing estimates for the NTR facility total annual operating costs and sources of funds to cover these costs for each of the first five years of the period of extended operation (PEO) requested. Projections are as follows:

YEAR SALARIES OPERATING EXPENSES TOTALS

2021 $3,267,000 2022 $3,365,010 2023 $3,465,960 2024 $3,569,939 2025 $ 3,677,037

b. Funding for NTR operation is sourced from corporate GE, to which the NTR contributes revenues garnered from contractual neutron radiographic services to customers from both the private and public sectors. As evidenced in the Securities and Exchange Commission GE Company Annual 10-K Filing for 2021, GE Powers profit and earnings ($726 million for the year ended December 31, 2021) demonstrate ample financial wherewithal to assure GEHs ability to obtain funding to cover estimated operating costs for the R-33 license.

The 2021 GE 10-K filing is available at:

https://www.sec.gov/Archives/edgar/data/0000040545/000004054522000008/ge 20211231.htm Financial assurance for decommissioning of the NTR was corroborated by the NRC according to letters:

Cancellation and Return of Financial Assurance for GE licenses, dated 3/26/2019 (Accession ML19063B412)

Acceptance Review of Financial Assurance Update of Decommissioning Funds - Surety Bonds for GE-Hitachi Energy Americas, LLC, dated 5/14/2019 (Accession ML19119A228)

(NRC Approval) General Electric-Hitachi 2020 Triennial Update to Decommissioning Funding Plan, dated 3/2/2022 (Accession ML21305A872)

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2.0 DecommissioningFinancialEstimates

Pursuant to letter, D. A. Hardesty (NRC) to M.F. Feyrer, GE-Hitachi Nuclear Energy Americas LLC (GEH) - Receipt and Supplemental Information Needed for Renewal Review of GEH Nuclear Test Reactor Facility Operating License No. R 33 (EPID L-2020-RNW-0038), dated 03/11/21 (ML21062A250), the NRC requests, pursuant to 10 CFR 50.75:

a. A current decommissioning cost estimate in 2021 dollars for the GEH NTR to meet the NRCs radiological release cr iteria for decommissioning the facility.

Accordingly, describe the basis on how the decommissioning cost estimate was developed.

b. A summary of total decommissioning costs broken down into the categories of labor, waste disposal, other items in current dollars, and a contingency factor.

c. A statement of the decommissioning method to be used.

d. A description of the means of adjusting the cost estimate and associated funding level periodically over the life of the facility, pursuant to 10 CFR 50.75(d)(2)(iii).

e. A numerical example showing how the decommissioning cost estimate will be updated periodically in the future.

Response

a. GEH has been authorized, pursuant to 10 CFR 70.25(f)(2), the use of a surety method to ensure decommissioning funding for the entire VNC site that adheres to the reporting requirements of 10 CFR 50.52. This was done, in part, because the retired VNC power reactors do not have Post-Shutdown Decommissioning Activities Reports (PSDARs). This decommissioning plan is prepared in general accordance with NUREG-1757, Consolidated NMSS Decommissioning Guidance, Volume 3: Financial Assurance, Recordkeeping, and Timeliness.

While decommissioning activities are paid from operating cash, the plan estimates the cost of decommissioning the entire site and assures that amount is available through the purchase of surety bonds held in a decommissioning trust fund. The plan was initially submitted on March 21, 1991, and provides for an annually upward 3.5% to 4.0% adjustment for inflation. Additionally, an outside accounting firm is engaged every 6 years to review the plan for adjustments. A 6-year review was recently completed in the Fall of 2019 and a partial review in March 2020 was performed to address decommissioning cost adjustments for the GE Test Reactor (GETR).

The NRC approval and SER was forwarded to GEH on February 8, 2021, approving an updated decommissioning cost estimate for the VNC.

Reference:

Approval of Lowering the Surety Bond Amount for the General Electric Test Reactor (Enterprise Project Identifier L-2020-DFA-0009), dated February 8, 2021 (ADAMS package accession No. ML21025A380).

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Evidence of surety bond riders was most recently submitted to the NRC on March 23, 2022, and included a copy of rider 2253264 for the NTR in the amount of

$6,269,557. GEH submitted the VNC DFP (rev 4) on February 28, 2020, and provided supplemental information on June 2, 2020, and on November 5, 2020, under VNC license SNM-960. These documents were approved by the NRC on March 2, 2022, with additional details of decommissioning costs.

Reference:

General Electric-Hitachi 2020 Triennial Update to Decommissioning Funding Plan (Enterprise Project Identifier L-2020-DFA-0002), dated March 2, 2022 (ADAMS package accession No. ML21305A871).

b. Because the GEH decommissioning funding plan for the NTR is a subcomponent of that of the site, the decommissioning method to be used with be determined at cessation of operations in consideration of the status of the site.

c. See response to a above.

d. See response to a above.

3.0 Facility %DevotedtoClass103Commercial Activities:

Pursuant to letter, D. A. Hardesty (NRC) to M.F. Feyrer, GE-Hitachi Nuclear Energy Americas LLC (GEH) - Receipt and Supplemental Information Needed for Renewal Review of GEH Nuclear Test Reactor Facility Operating License No. R 33 (EPID L-2020-RNW-0038), dated 03/11/21 (ML21062A250), the NRC requests, pursuant to 10 CFR 50.22:

a. A confirmatory statement identifying the percentage of the facility that is devoted to the production of materials, products, or energy for sale or commercial distribution, or to the sale of services, other than research and development or training.

Response

a. According to M.F Feyrer, Operations Manager VNC and R-33 license holder, although the mission and purposing of the NTR has varied over its years of operation, as of March 2022, the NTR is devoted percent to Class 103 commercial activities pursuant to 10 CFR 50.22.

percent of the facility is currently devoted to research and development in the form of neutrographic experiments performed as a useful service to the Department of Defense.

4.0 Facility %DevotedtoEnergyProductionActivities:

Pursuant to letter, D. A. Hardesty (NRC) to M.F. Feyrer, GE-Hitachi Nuclear Energy Americas LLC (GEH) - Receipt and Supplemental Information Needed for Renewal Review of GEH Nuclear Test Reactor Facility Operating License No. R 33 (EPID L-

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2020-RNW-0038), dated 03/11/21 (ML21062A250), the NRC requests, pursuant to the Atomic Energy Act, Section 104(c):

a. A statement identifying whether the percen tage of the annual costs of owning and operating the facility are recovered through sales of nonenergy services, energy, or both, other than research and development or education and training, is 75 percent or less.

b. A statement identifying whether 50 percent or less of the annual costs is from sales of energy.

Response

a. The parent GEH pays for the operation of the NTR. While revenue from the NTR goes into the overall GEH coffers it is not directly correlated to NTR overhead and expenses are met through an operations budget. Moreover, none of the revenues produced at the VNC are directly or indirectly used by GEH to finance research and development or educational activities related to the NTR. Given this indirect correlation between revenues and operating expenses, since percent of NTR revenues come from the commercial sale of non-energy services (extensively neutrography), it can be said that less than 75 percent of the annual costs of owning and operating the facility are recovered through sales of nonenergy services.

Therefore, according to M.F Feyrer, Operations Manager VNC and R-33 license holder, as of March 2022, less than 75 percent of the annual costs of owning and operating the facility are recovered through commercial sales of nonenergy services.

b. According to M.F Feyrer, Operations Manager VNC and R-33 license holder, as of March 2020, zero (0) percent of the annual costs of owning and operating the facility is from sales of energy.

5.0 AmendedSecurityPlanaddressing10CFR73.67(e):

Pursuant to letter, D. A. Hardesty (NRC) to M.F. Feyrer, GE-Hitachi Nuclear Energy Americas LLC (GEH) - Receipt and Supplemental Information Needed for Renewal Review of GEH Nuclear Test Reactor Facility Operating License No. R 33 (EPID L-2020-RNW-0038), dated 03/11/21 (ML21062A250), the NRC requests, pursuant to 10 CFR 73.67:

a. Submit a security plan or an amended s ecurity plan describing how the licensee will comply with all the requirements of paragraphs (d), (e), (f), and (g).

b. Provide a statement that GEH will comply with the requirements under 10 CFR 73.67(e) and 10 CFR 73.67(g) as appropriate.

Response

a. An updated PSP was submitted on 12/16/2021 under Oath or Affirmation and contained the changes in the conclusion section of b below. That PSP was

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reviewed by the NRC and comments provided to GEH. A final revision was submitted on 4/4/2022 and is pending NRCs final approval.

b. Pursuant to 10 CFR 70.22(k), each licensee licensed to possess 10 kg or more of special nuclear material of low strategic significance must include a physical security plan that demonstrates how the applicant plans to meet the requirements of paragraphs (d), (e), (f), and (g) of § 73.67. Regulatory Guide 5.59 clarifies that a physical security plan is not required for possession of less than 10 kg; however, the licensee is required to meet the requirements of § 73.67.

The November 28, 1979, Final Rule for Physical Protection Upgrade explained that NPRs are not required to meet the provisions of the Rule but were, pending further rulemaking, required to meet the provisions of § 73.67(a), (b), (c), and (d)

(and § 73.60 for some NPRs possessing formula quantities). These interim requirements were followed by proposed rulemaking on May 29, 1992 (57 FR 104) and finalized in the Clarification of Physical Protection Requirements at Fixed Sites Rule issued on March 15, 1993 (57 FR 22670).

§ 73.67(b) provides for self-protected exemptions for SNM that is not readily separable with emitted dose rates greater than 100 Rem/hr at 3 feet from any accessible surface without intervening shielding. According to Calculated Dose Rate at 3 Feet from an NTR Fuel Element, Revision 1"; August 29, 1990, as submitted to the NRC in RAI response dated August 31, 1990, the NTR core is self-protecting. The remaining quantity of SNM authorized for possession under the R-33 licensed that is within applicability of § 73.67(a) and the fixed site requirements of § 73.67(f) and the material transport requirements of § 73.67(g). However, the R-33 license for NTR does not contain language that clearly authorizes transport of SNM, ex cept that a small quantity of fresh fuel could be received within the existing 2.B.(2)a, in-core reactor fuel possession limit above what is currently in the core. This quantity would be restricted even more by the sitewide possession limit imposed by SNM-960. And, because of the limitations on HEU pursuant to § 50.64, there is no conceivable scenario by which fresh fuel would be received under the existing R-33 license.

As a result, the NTR is required to have a physical protection system that addresses compliance with 10CFR73.67(a) and (f) as a facility licensed to receive, possess, and use SNM of Low Strategic Significance for the quantity of SNM licensed under R-33 that is not self-protecting.

According to license SNM-960, the balance of the VNC fixed site is not required to have a PSP pursuant to § 70.22(k) because the license does not permit possession of SNM of Low Strategic Significance in quantities 10 kg.

License SNM-960, license condition T-1.1 states that the licensee shall not transport a shipment of quantities of SNM specified in § 73.1(b)(2) until a

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detailed plan is described pursuant to § 70.22(g). According to § 70.22(g)(1), a license application to transport 10 kg or more of SNM of low strategic significance would require a description of the plan for physical protection of SNM in transit according to 73.67(a), (e), and (g). This quantity (10 kg) exceeds the possession limit of SNM-960 for material not in storage. SNM-960 further prohibits movement (including transport) of the material in storage in the Hillside Storage Facility (HSF) under NRC Order EA-14-144. Therefore, the condition described in T-1.1 cannot be met under current licensing.

However, since SNM-960 does authorize possession and transport of greater than 15 grams of strategic special nuclear material pursuant to § 73.67(a), the VNC is required to maintain a physical protection system that meets the general performance objectives of § 73.67(a), the fixed site requirements for SNM of low strategic significance in § 73.67(f), and the in-transit requirements of § 73.67(g).

IN CONCLUSION: The VNC, under the combined R-33 and SNM-960 licenses, is within applicability and will comply w ith the general performance objectives of

§ 73.67(a), the fixed site requirements for SNM of low strategic significance in § 73.67(f), and the in-transit requirements of § 73.67(g).

6.0 Primarycoolantconductivitymetercalibrationchangedto biennial:

Pursuant to VNC Condition Report CR 37628.

Justification: The original NTR primary coolant conductivity probe could not be removed from the system safely and was necessarily annually verified (rather than calibrated against a known standard) in-situ by comparing readings with laboratory primary coolant conductivity analyses. The conductivity instrumentation was recently replaced under VNC Change Authorization CA-317 with a Valmet series 3200 meter and a model Industrial Sensor 4721PE probe. The in-situ verification has been continued; however, because the new probe can be removed from the system, a proper calibration can now be performed. In establishing the proper frequency to perform this calibration, ANSI-ANS 15.1-2007, Development of Technical Specifications for Research & Test Reactors, 4.3, Coolant Systems, was consul ted, but does not provide a frequency for the calibration of instruments used for conductivity or pH monitoring, nor does Valmet provide a recommended frequency for calibration. The ANSI does, however, give annual to biennial as a standard recommendation for the calibration of all plant instruments except that it recommends annual calibrations for section 4.2, reactor control and safety systems.

Because the NTR coolant water is demineralized well water and is void of chemical additives, the source of makeup water is constant, and conductivity historically has been stable well below the TS 3.3.3.3 limit of 10 S/cm. In order to minimize personnel exposure as well as the risk involved in removing and replacing the probe during calibration, GEH / VNC is establishing the calibration frequency to biennial for the

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conductivity probe. This involves an edit to Tech Spec Table 4-2. An edited page has been provided in Enclosure 2.

7.0 SARFigure1319revisedandexplained

Pursuant to NRC License Renewal Audit of 7/28/21, Question 003:FSAR section 4.4.2 state that the coolant temperature coefficient is calculated based on overall temperature changes and describes determination of th e void coefficient by extrapolating the temperature coefficient data. FSAR figure 13-19 shows that some void fraction is expected for steady state operation at all non-zero power levels.

a. Provide the experimental data and calculations of the coolant temperature and void coefficients of reactivity.

b. Based on the discussion in this same paragraph, the NTC staff interprets that the coolant temperature coefficient includes effects of both water density change and void buildup, and that the void coefficient of 5.7¢/%

void is only applicable over 124°F.

i. Is the NRC staff interpretation corr ect? If not, explain the effects included in the coolant temperature coefficient.

ii. Explain over what ranges the reactivity coefficients listed in Table 4-2 are applicable.

Response: In addition to the responses submitted in letter, GEH Supplemental Information Supporting GE Nuclear Test Reactor License Renewal Audit - Audit Questions and Responses, dated September 22, 2021, GEH has revised Figure 13-19 of the SAR (below right) and it is included in the draft Chapter 13 provided in Enclosure 2.

The existing figure (below left) anomalously indicates that the NTR has a core void fraction even just above zero power. This is not the case and CORLOOP runs indicate no nodes with subcooled boiling for power le vels less than 190 kW. Figure 13-19 was apparently simply a trend line requiring extr apolation at lower power by placing the zero void at zero power. It appears that void fraction is actually a misnomer that intends to represent the fraction of nodes in subcooled boiling - not necessarily the vapor volume fraction. The new Figure 13-19 reflects the fraction of nodes in subcooled boiling out of a total of 24 (e.g.; at 325 kW, 10 of 24 nodes are in subcooled boiling [10/24=0.4]). The axes have been flipped relative to the old graph to enhance clarity.

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Supplemental Question forwarded by em ail from D. Hardesty on 3/7/2022.

Does the revised figure still imply that void ex ists in the core at power levels below 190 kW, since a node in subcooled boiling will have a nonzero void fraction? If so, will the reactivity coefficients be affected, since the void and moderator temperature coefficients were determined assuming that no void is present in the core?

Response: There is no void in the core at pow er levels below 190 kW during normal operation. The curve in Figure 13-19 is ge nerated by running various sensitivity cases with different power levels by changing flow rates and inlet temperatures. The smallest power level that predicted subcooled boiling was 193 kW. The depicted curve is then extrapolated to zero power with zero voids, but it does not imply that there is voiding for this power levels.

To better explain this in the SAR, Figure 13-19 has been modified to show a broken line curve below 190 kW. The following verbiage has been added to Section 13.7.3 of the SAR in which this figure is explained:

The curve in Figure 13-19 is generated by runni ng various sensitivity cases with different power levels by changing flow rates and inlet temperatures. The smallest power level that predicted subcooled boiling was 193 kW. The curve is then extrapolated (indicated by broken line) to zero power with zero voids; however, there is no voiding at power levels below 190 kW during normal operation.

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8.0 CorrecttypoinTS3.7.1.2:

Pursuant to NTR License Renewa l Audit of 7/28/21, Question 010:Technical Specification (TS) 3.7.1.2, states, normal in stalled monitors in section 3.5.1.1. Please confirm this is the co rrect section. In context, it appears that TS 3.7.1.2 should read, normal installed monitors in section 3.7.1.1.

Response: An edited page has been provided in Enclosure 2.

9.0 UpdateSARtoreflectthermocouplepileasabandonedinplace

Pursuant to GEH Condition Report CR 32562:

The thermopile (or thermocouple pile), is listed in the SAR as a safety-related system and provides a variable into the heat balance equation used to calibrate the gamma-compensated ion chambers, which provides inpu t for the Log N scram channel. A review of documents indicates that th e thermopile failed sometime pr ior to July 1974 and that the use of the core outlet (TC-2) and core inlet (TC-5) thermocouples to determine the core differential temperature was em ployed as a compensatory measure. The thermopile was replaced with a similar thermo pile (an equivalent replacem ent per NEI 21-06), but it was never calibrated and placed in service. The replacement was done under the old (pre-1999) 50.59 Rule and documented in site Change Authorization (CA) 256 as not involving an unreviewed safety question pur suant to 10CFR50.59(a)(2). A safety analysis was performed on May 19, 1986, to justify the ongoing use of compensatory

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method of determining delta-T and determined the accuracy to be within the +/- 3%

variability listed in SAR Table 13-6 for core inlet/outlet delta temperature.

GEH Engineering performed safety analysis according to Action #4 of CR 32562 (September 2020) and confirmed that the May 1989 safety analysis supported the use of the approved compensatory method (TC TC-5 delta) on a permanent basis as acceptable but recommended calibrating the thermopile and returning it to service. In support of this effort, a calibration method and calibration procedure were proposed.

However, this method required reliance on two single isolation valves that the NTR licensed staff determined unreliable and rejected the method as unsafe due to the potential to create an unisolable leak in the coolant system.

Given the proven reliability of the compensatory method, and its approval for permanent use, the determination was made to adopt it as the singular means for performing the heat balance and to abandon the thermocouple pile in place.

Changes: [included in page markups provided in Enclosure 2]

a. SAR 7.3.3: The current SAR, Section 7.3.3 discusses the automatic switching of the start-up channel (source range monitor) if utilized to the thermocouple pile at powers greater than 0.1 kW. The thermopile-related information from this strip chart recorder has not been used in recent history; therefore, a lost input from the thermopile will have no consequence. 7.3.3 currently says: The log N and period amplifier receives its signal from the fourth CIC and displays the reactor period and reactor power on front panel meters. This system may be set up to cover the power range from source or reactor critical level, depending on CIC position, to 150% of power. Relay outputs from the period amplifier trip circuit and the log N amplifier trip circuits are connected through the noncoincident logic circuit to initiate a scram for reactor periods of less than 5 seconds. At powers of less than 0.1 kW, a signal from the log N recorder actuates automatic bypass of the primary coolant low-flow scram. At powers of less than 0.1 kW, a signal from the log N recorder actuates automatic bypass of the primary coolant low-flow scram.

At powers greater than 0.1 kW, a signal from the log N amplifier actuates a relay which automatically switches the signal to a single function recorder from the start-up channel (source range monitor) if utilized to the thermocouple pile (millivolts) signal, which is used in the heat balance calculation. The log N power signal is also recorded on a strip chart recorder. The mode (multiposition calibration) switch for the log N amplifie r is interlocked to scram the reactor when the switch is not in the OPERATE position.

This has been changed to read: The log N and period amplifier receives its signal from the fourth CIC and displays the reactor period and reactor power on front panel meters. This system may be set up to cover the power range from source or reactor critical level, depending on CIC position, to 150% of power. Relay outputs from the period amplifier trip circ uit and the log N amplifier trip circuits are connected through the noncoincident logic circuit to initiate a scram for reactor periods of less than 5 seconds. The log N power signal is recorded on a strip chart recorder. The mode (multiposition calibration) switch for the log N

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amplifier is interlocked to scram the reactor when the switch is not in the OPERATE position.

b. SAR Table 7-2, Safety-Related Systems and footnote #4: currently says: A thermocouple pile is provided which indicates the primary coolant core temperature differential. This is utilized in combination with the primary coolant flow rate to provide information for a heat balance determination.

Footnote #4 has been changed to say: Core inlet and outlet thermocouples provide for primary coolant core temperature differential. This information is utilized in combination with the primary coolant flow rate to provide information for a heat balance determination

c. SAR 1.8.4 Recent Modifications; added the following paragraph:

In 2022, the primary coolant thermocouple pile was designated as abandoned in place as it was determined it could not be reliably placed in service. The thermocouple pile failed in 1974, was replaced in 1999, but has never been returned to service. Pursuant to 50.59(c), the compensatory measure of using the core outlet (TC-2) and core inlet (TC-5) thermocouples to determine the core differential temperature has been made the permanent method for performing the heat balance determination.

d. TS Table 3-2, Core Delta temperature provide information for the heat balance determination. No change necessary.

Table 3-2 The bases for items listed in Table 3-2 are as follows: added the following paragraph - Core Delta temperature is the difference between the core outlet (TC-2) and core inlet (TC-5) thermocouples.

Table 4-2 Surveillance Requirements of Reactor Safety-Related Items (Information Instruments) Item 5 added TC-2 & TC-5 to description of item.

Added TC-7 to item 3 for further clarification.

e. TS 1.2.18, definition of Reactor Thermal Power as determined by a primary coolant system heat balance.

No change is necessary.

10.0 AdditionalSARChapter13accidentanalysis:

Pursuant to NTR License Renewal Audit of 7/28/21, Question 011:FSAR Chapter 13 describes some postulated accidents, however, other categories of accidents described in NUREG-1537, Part 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria, February 1996 (ADAMS Accession No. ML042430048) are not evaluated and analyzed.

a. Identify the postulated MHA analyzed and evaluated dose consequence for workers (licensee staff) and members of the public.

b. Revise FSAR Chapter 13 to describe how the postulated accident categories in NUREG-1537 Part 2 are analyzed and evaluated for the GE NTR.

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c. For each applicable postulated accident category, identify the limiting event selected for detailed quantitative analysis, evolution of the scenario, and address the likelihood of occurrence.

Response

a. The postulated MHA for the NTR is a fueled experiment described in Section 13.6, Experiment Design Basis Accident as none of the reactor accident categories in NUREG-1537 result in fuel degradation. All analysis and analytical methods, including occupational and public dose consequences, are addressed.

An accompanying document (Enclosure 2, Page Markups and Draft Changes to SAR Chapter 13, has been provided as a draft to pending changes to Chapter 13.

[Proposed changes to the SAR are annotated in blue font and have been included in the draft Chapter 13 provided in Enclosure 2].

This section (13.6) and its subsections will be brought forward in the SAR and renumbered as Section 13.2 to emphasis that it is the MHA for the NTR.

A sentence will be added stating: The experiment design basis accident is designated as the maximum hypothetical accident (MHA) for the NTR, resulting in the maximum potential radi ation hazard to personnel and the public.

All following sections will be renumbered accordingly.

Mark up pages have not been provided as edits result in significant changes in formatting. These changes will be included in the forthcoming revision to the SAR.

b. Analytical descriptions and calculational methods for the postulated MHA are included in detail in Section 13.6.2 and 13.6.3.

As per response to a, all subsections of 13.6 will be brought forward and numbered 13.2.

c. NUREG-1537 describes the following categories of accidents:

MHA insertion of excess reactivity (ramp, step, startup, etc.)

loss of coolant loss of coolant flow mishandling or malfunction of fuel experiment malfunction loss of normal electrical power external events mishandling or malfunction of equipment

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In the SAR, the accident-initiating events and scenarios are discussed under anticipated operational occurrences and postulated accidents categories. Following is an abbreviated roadmap to these categories of accidents.

Maximum Hypothetical Accident:

The maximum hypothetical accident (MHA) is an enveloping event that is postulated to involve a failure in a fueled experiment. Such an event would lead to the maximum potential radiation hazard to personnel and to the public. Also considered as an experiment design basis accident, the event is discussed in Section 13.6, Experiment Design Basis Accident.

Insertion of Excess Reactivity:

To bound all possible insertion accidents, section 13.4.3 hypothesizes an extremely improbable event in which a beyond design basis seismic event results in the loss of all ability to insert negative reactivity in to the core by presuming control and safety rod drive and experiment mechanisms all fail in concert in such a way that control rods and experiments are withdrawn from the core region while safety rods are rendered inoperable. This event also presumes catastrophic failure of the primary and secondary coolant systems as well as the electrical supply to the reactor system. This scenario is presented to support analysis proving that there is no case by which insertion of excess reactivity can result in fuel damage to the NTR reactor when potential reactivity available from control rods and experiments is maintained at or below 0.76$. GEH studies, based on this scenario, include idealized step reactivity insertion with scram in Section 13.4.1, idealized finite ramp reactivity insertion with scram in Section 13.4.2, and reactivity insertions without scram in Section 13.4.3. An evaluation for rod withdrawal accident is provided in Section 13.4.4, which is bounded by reactivity insertions without scram.

Acceptance criteria for the reactivity acci dents is the integrity of the core by maintaining fuel temperatures below the melt ing temperature. In all accident cases that are not terminated by scram, the transient is stopped by bulk boiling (moderator voiding) before fuel damage occurs. This is further discussed in the loss of coolant accident in Section 13.4.6. Since there is no release of radioactive products, there are no dose consequences associated with this accident scenario.

The following clarification will be added by editing the first sentence in section 13.4.3:

Current: It may be hypothesized that certain structures (used to support the control and safety rod mechanisms as well as experiments) might fail or move during a seismic event in such a manner as to withdraw the control rods and experiments from the core region and prevent operation of the safety rods.

Change to: There is no case by which insertion of excess reactivity can result in fuel damage to the NTR reactor when potential reactivity available from control rods and experiments is maintained at or below 0.76$. To bound all possible insertion accidents as well as a worst case external event accident,

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an extremely improbable event is hypothesized in which a beyond design basis seismic event results in the loss of all ability to insert negative reactivity into the core by presuming control and safety rod drive and experiment mechanisms all fail in concert in such a way that control rods and experiments are withdrawn from the core region while safety rods are rendered inoperable.

The following sentence will be added to the end of section 13.4.3:

Current: As the maximum fuel temperature for a loss-of-coolant occurring at a power level of 100 kW is less than the fuel melt temperature, there will be no fission product release from this accident.

Add: Since there is no release of radioactive products, there are no dose consequences associated with this accident scenario.

Loss of coolant:

Loss of coolant accident is analyzed in Section 13.4.6 and presumes a large rupture in the primary system at a point below the core entrance. The likelihood of this occurring is very low because it would require a catastrophic pipe sheer or core can failure that would both prevent adding water to the core while completely draining the core tank. This scenario concludes th at, following total moderator voiding of the core, natural convection of air currents set up by radiant heat transfer from the core to the graphite is adequate to prevent fuel me lting with fuel reaching a peak temperature of 626 °F. Since there is no release of radioactive products, there are no dose consequences associated with this accident scenario.

The following sentence will be added to the first paragraph of section 13.4.6:

Current: The reactor loss-of-coolant accident involves the total loss-of-coolant inventory in the core as the result of a rupture in the primary system, combined with a failure to scram. The accident is postulated to occur as follows:

Add: The reactor loss-of-coolant accident involves the total loss-of-coolant inventory in the core as the result of a rupture in the primary system, combined with a failure to scram. The likelihood of this occurring is very low because it would require a catastrophic pipe sheer or core can failure that would prevent adding water to the core tank while completely draining the core tank. The accident is postulated to occur as follows:

The following sentence will be added to the end of section 13.4.6:

Add: As a result, fuel damage will not occur and there is no release of radioactive products or associated dose consequences.

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Loss of coolant flow:

Loss of flow accident is analyzed by assuming an instantaneous seizure of the rotor in the single recirculation pump in the coolan t system. The evaluation is discussed in Section 13.4.5. There is a reasonable probability of this scenario occurring because the coolant system only has a single pump. In this scenario, analysis concludes that bulk boiling will maintain fuel temperatures well below the fuel melting point and ensures core integrity even when the event is compounded by a failure of the automatic low-flow scram. Since there is no release of radioactive products, there are no dose consequences associated with this scenario.

The following sentence will be added after the first sentence in section 13.4.5:

Add: There is a reasonable probability of this scenario occurring because the coolant system has only a single pump.

The following sentence will be added to the end of section 13.4.6:

Add: Because fuel damage does not occur in this scenario, there is no release of radioactive products and no associated dose consequences.

Mishandling/Malfunction of Fuel:

Fuel handling errors are discussed in Section 13.3.5. Potential accidents are precluded by the operating procedures. The likelihood of such accidents is insignificant due to the design of the reactor, the historical infrequency of fuel handling (last done in 1976 to support core container replacement), and the unlikely sequence of events that would be necessary for the NTR to be refueled with LEU fuel. The most likely scenario would involve the removal and replacement of fuel assemblies to support maintenance or core modification.

In this case, the design of the reactor is such that dropping a fuel assembly could only cause a core-related accident if the control rods were withdrawn during loading so that the reactor was almost critical before adding fuel.

Because of the unique design of the fuel (discussed in 4.2.1), the dropping of a spent fuel assembly would not result in a significant release of fission product gases as would be expected of an assembly of conventional fuel rods.

During review of this section, it was identified that the term fuel element was used when fuel assembly is appropriate. This has been changed as per below.

The following sentences in section 13.3.5 will be replaced as follows:

Current: It should be reemphasized here that refueling for reactivity increase is not necessary, and fuel hand ling is very rare. The most recent fuel handling occurrence was in support of core container replacement in 1976.

Replace with: The likelihood of such accidents is insignificant due to the design of the reactor, the design of the reactor fuel (discussed in 4.2.1), the historical infrequency of fuel handling (last done in 1976 to support core

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container replacement), and the unlikely sequence of events that would be necessary in order for the NTR to be refueled with LEU fuel.

The term fuel element has been replaced with fuel assembly throughout the discussion.

The following sentence will be added to the end of section 13.4.6:

Add: As a result, fuel damage during fuel handling would be unlikely, with no associated dose consequences.

Experiment Malfunction:

Experiment malfunctions include and are bounded by the MHA discussed in section 13.6. It is concluded that the MHA would result in a two-hour total effective dose equivalent of 66.2 mRem at the site boundary and a 5-minute evacuation exposure of 470 mRem.

Loss of normal electrical power:

Loss of electrical power event is evaluated in Section 13.3.1 under anticipated operational occurrences and is a credible scenario given that the design of NTR does not include emergency backup power. The consequences from a loss of AC power are bounded by loss of flow scram and results in an inconsequential increase in power and fuel temperature. This event results in no release of radioactive products and there are no dose consequences.

The second sentence in section 13.3.1 will be edited as follows:

Current: Loss of ac power to the facility means a complete loss of electrical power and results in reactor scram through the following processes:

Replace with: Therefore, this accident scenario is reasonably probable, as any loss of ac power to the facility means a complete loss of electrical power and results in reactor scram through the following processes:

The following sentence will be added to the end of section 13.3.1:

Add: Because fuel damage does not occur in this scenario, there is no release of radioactive products and no associated dose consequences.

External Events:

The only credible external event for the site is a seismic event. Potential failures and their consequences are discussed as a part of reactivity insertions in Section 13.4.3.

Section 13.4.7, External Events, will be added to the SAR as follows:

The only credible external event for the NTR is a seismic event, which is discussed as the initiating event for the bounding reactivity insertion accident in Section 13.4.3. Because fuel damage does not occur in this scenario, there is no release of radioactive products and no associated dose consequences.

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Mishandling/Malfunction of Equipment:

Mishandling of equipment is precluded by the technical specifications and operating procedures.

Section 13.4.8, Mishandling/Malfunction of Equipment, will be added to the SAR as follows:

Mishandling of equipment is precluded by the technical specifications and operating procedures. Because bounding analysis in Section 13.4 addresses all events that could result from improper operation, fuel damage does not occur in this scenario, there is no release of radioactive products, and there are no associated dose consequences.

11.0 ClarifyTermWorseCoreGeometry:

Pursuant to NTR License Renewal Audit of 7/28/21, Question 014:SAR Section 13.3.5 states that the physical arrangement of the fuel container is such that an assembly located in the loading chute results in a wors e core geometry than the cylinder formed by having all assemblies in the core support reel. In this context, explain if worse core geometry mean a geometry with a smaller multiplication factor.

Response: [included in the draft Chapter 13 provided in Enclosure 2]:

The physical arrangement of the fuel container is such that an assembly located in the loading chute results in a worse core geometry (one having a lower neutron multiplication factor) than the cylinder formed by having all assemblies in the core support reel.

12.0 Fuelmeltingtemperature:

Pursuant to NTR License Renewal Audit of 7/28/21, Question 015:SAR figure 13-3 shows the fuel melting temperature to be 1050°F, but figure 13-4 shows the melting temperature to be greater than 1100°F. Provide the actual melting temperature of the fuel.

Response: These figures are an historical legacy which has made it difficult to validate the original basis for the fuel melting temper ature graphic construction lines that appear to be slightly different in the two figures. In fact, the 12/28/1964 NTR Nuclear Systems Analysis NUSA114 document indicates melting at 1200°F. It was noted in the 2/1/1965 NTR Summary Safeguards Report (APED4444) that melting is at 1200°F. This number is an approximate as demonstrated in Figure 11.9 from the APED4444 shown in the response to this audit question in letter David J. Heckman (GEH) to D. A. Hardesty (NRC), GEH Supplemental Information Supporting GE Nuclear Test Reactor License Renewal Audit - Audit Questions and Responses, dated 9/22/2021 [ML21265A247, ML21265A248, ML21265A249, ML21265A250].

However, the UAlx fuel meat melting temperature can be determined from the UAl phase diagram published by Kassner et al. (M.E. Kassner, P.H. Adler, and M.G.

Adamson, Evaluation and Thermodynamic Analysis of Phase Equilibria in the UAl, Journal of Nuclear Materials, 167 (1989) 160168). Based on Kassners phase diagram,

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the lowest onset of any melting for the NTR UAlx alloy fuel meat that has 3.4 at % U is the eutectic 641°C (1185.8°F). Since the UAlx fuel meat melt temperature is a eutectic, it is bounding for the Al cladding which has a slightly higher melt temperature than the fuel meat. The 1185.8°F eutectic temperature is similar to, but higher than, the two legacy melting temperature lines, and continues to demonstrate that there is margin and conservatism against fuel melt for relatively large reactivity insertions. This number is reflected in Figure 1310 of the SAR.

Section 13.5.1 has been revised in the draft Chapter 13 provided in Enclosure 2 to address ambiguities in fuel vs clad melting temperatures. The change clarifies that the legacy data is somewhat imprecise, but is cons ervative relative to the eutectic fuel melt temperature of 1186°F.

Supplemental Question forwarded by email from D. Hardesty on 3/7/2022.

NUREG-1537, Appendix 14, Section 2.1 recommends a safety limit on fuel temperature of 530 C for aluminum clad, aluminum matrix fuels, based on observations of fission product releases at the blister threshold. For the GEH fuel, is melting the only necessary criterion for prevention of fission product release?

Response: NUREG-1537, Appendix 14, Section 2.1 discusses the safety limit for operation of aluminum matrix fuels, referring to NUREG-1313 as the general support for the value of 530 C. NUREG-1313 itself is concerned with silicide-based fuel, but does indicate that the phenomenon of blistering should be considered in determining limits for fuel temperatures, although it describes this in context of conditions of operation, which an accident condition is not. The temperature at which blistering occurs, as described in Nuclear Technology, Vol 49, Development and Irradiation Performance of Uranium Aluminide Fuels in Test Reactors (the reference NUREG-1313 uses for U-Alx fuel),

while 50-100 degrees C lower than the melting temperature, is representative of a much longer duration mechanism than the few seconds involved in RIA events in the NTR SAR. Specifically, blister temperatures were determined on irradiated samples by sequentially raising temperatures and hol ding for 30 minutes until blisters formed.

Therefore, melting temperature is an appropriate temperature for assigning fuel integrity temperature for the RIA events.

The revised text of the NTR SAR Section 13.5.1 is added to Section 13.5.1:

Section 13.5.1 has been revised in the draft Chapter 13 provided in Enclosure 2 to address blistering and justify melt temperature as an appropriate metric to evaluate fuel melt during the constrained time in which an RIA transient would occur.

13.0 ClarifytermsinChapter13toeliminateambiguitybetween claddingandfueltemperatures:

Pursuant to NTR License Renewal Audit of 7/28/21, Questions 016 & 018:SAR section 13.4.3 discusses peak fuel temperature, but only peak cladding temperatures are depicted.

Provide the peak fuel temperatures for these transients. The title of FSAR figure 13-14 is LOCA - PCT (°F), (emphasis added), but the caption is Fuel Temperature Following

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Loss of Coolant Accident. The preceding discussion states that the figure depicts fuel temperature. Explain the parameter that is plotted in Figure 13-4.

Response

The fuel temperatures and the cladding temperatures are essentially the same since they are within few degrees of each other. Thin metallic fuel has insignificant temperature gradient. Therefore, fuel and cladding temperatures are within a fraction of a degree during a LOCA. In the figures, peak cladding temperatures are depicted out of optical convenience since they are readily available as a plot parameter.

Section 13.5.1 has been revised in the draft Chapter 13 provided in Enclosure 2 to address ambiguities in fuel vs clad melting temperatures. The change clarifies that the legacy data is somewhat imprecise, but that peak fuel and peak cladding temperatures are within a few degrees and are used interchangeably throughout Chapter 13.

Supplemental Question forwarded by email from D. Hardesty on 3/7/2022.

Is the assumption of uniform fuel and cladding temperature valid during an RIA? The lag between peak power and maximum PCT shown in Figure 13-10 implies a difference between fuel and cladding temperature. Provide the justification that PCT is an appropriate metric to evaluate fuel melt during an RIA.

Response: The time of peak power in the transient does not imp ly a time for the peak temperature to occur of either the fuel or the cladding because the temperature increase results from the total energy deposited in the fuel, which is the integral of the power over the transient. The integrated power deposited increases continuously throughout the transient. Temperatures only start to decrease once the heat transfer mechanisms have adequate time to remove the energy, which occurs very late in the transient compared to the peak in the power trajectory. This is typical of RIA evaluations and is not unique to NTR. It does not imply that there is any large difference between the fuel and cladding temperatures.

The revised text already provided for Section 13.5.1 addresses the differences between the fuel and cladding melting temperatures explaining that the eutectic temperature of the fuel is similar to, but higher than, the melting temperature lines used in the evaluations.

This continues to be true and therefore use of PCT is an appropriate metric to evaluate fuel melt during an RIA.

Section 13.5.1 has been revised in the draft Chapter 13 provided in Enclosure 2.

14.0 ClarificationofTechnicalSpecification1.2.20.2,definitionof ReactorSecured:

During the NRC NTR Safety Inspection performed during the week of February 28 (IR-2022-201), the inspector noted a potential violation of Technical Specification (TS) section 6.1.3.1 during performance of ER 21-17 on June 14-15, 2021, regarding minimum staffing during work in progress with the reactor secured. The work involved a safety rod that was removed from the core a nd worked externally from the reactor. The inspector observed that the current definition of Reactor Secured is ambiguous as to

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whether the reactor can be concluded to be Secured when work is ongoing on a safety rod that has been removed from the core.

Response

While this issue is currently being tracked as an Unresolved Issue, both the inspector and staff have conceded that ambiguity should be removed from this TS. The current definition given in TS 1.2.20.2 is:

The reactor is considered secured under either of the following two conditions:

1. The core contains insufficient fissile ma terial to attain criticality under optimum conditions of moderation and reflection.

2. That overall condition where all of the following conditions are satisfied.

a. Reactor is shut down.

b. Console keylock switch is OFF and the console key is in proper custody.

c. No work is in progress involving in-core components, installed rod drives, or experiments in an experimental facility.

Ambiguity exists in part 2.c. in concluding what is work in progress involving in-core components as it can be interpreted that in-core components are a type of component relative to design configuration and that a component of the in-core type would continue to be an in-core component even when removed from the core. In this case it follows that work in progress would continue on an in-core component even after it is removed from the core although it does not affect reactivity in the core.

Alternatively, if strict dictionary definitions are applied, an in-core component is simply a component that is installed in the core and it ceases to be an in-core component once it is physically removed from the core. In this case, work in progress on an in-core component ceases once that component is removed from the core. This case is not tied to design configuration, but to the affect the work might have on in-core reactivity in that, once the component is removed from the core, work on that component no longer affects core reactivity.

According to TS 1.2.17 and 1.2.21, the reactor is either in operating or in shutdown mode. The distinction is directly related to the net reactivity in the core. Reactor Secured is achievable only when the reactor is shut down and additional administrative work controls are established that preclude work that might result in an inadvertent transient that could cause the reactor to enter the operating mode.

NTR staff believes that the ANSI 15-1, Section 1.3, recommended language, with some modification, provides the necessary clarity to address the ambiguities in TS 1.2.20.2.

The ANSI says: No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods. NTR staff then concludes that the definition of Reactor Shutdown should include:

The reactor is shut down.

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The console keylock switch is OFF and the key is removed from the lock.

No work is in progress on core components that can directly affect core reactivity, including core fuel, the core structure, installed control or safety rods, or control rod drives unless they are physically decoupled from the control rods.

No experiments are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment, or one dollar, whichever is smaller.

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