THERMAL POWER SCALING OF THE KILOPOWER …anstd.ans.org/wp-content/uploads/2015/07/5135_Poston-et-al.pdf · THERMAL POWER SCALING OF THE KILOPOWER SPACE REACTOR. Slide 1 . ... Past

Operated by Los Alamos National Security, LLC for NNSA

THERMAL POWER SCALING OF THE KILOPOWER SPACE REACTOR

Slide 1

David Poston (LANL), Patrick McClure (LANL), Marc Gibson (GRC), Lee Mason (GRC), Cheryl Bowman (GRC),

John Creasy (Y12), Chris Robinson (Y12) [email protected], 505-667-4336, spacenuke.blogspot.com

Feb 24, 2015

NETS 2015 -- ABQ


Slide 2

This chart includes very rough estimates of mass, practicality and utility of each power source. The utility of solar power is obviously dependent on distance from sun and/or possibility of day-night cycle. Yellow curve is estimate of utility at 10 AU, dotted yellow line is estimate at 1 AU (for no eclipse application).

Keep you heart focused on the big-picture…


… but your mind focused on small steps. To utilize the vast potential of fission power in space, we need

a path that is programmatically and technically sustainable. • For the full story, goto spacenukes.blogspot.com

A simpler system and “smaller the step” not only lowers cost /risk, it also reduces the uncertainty/creep in mass, cost, and schedule.

• The completion of DUFF has made us extremely confident was can design, build and test a prototypic reactor power system in 3 years for $10M.

• If KRUSTY is successful, then we will have great confidence in the cost, schedule, and risk of a Kilopower flight program.

— Except for aspects like ATLO, INSRP, which will have to be estimated based on other NASA/DOE experience (although we can design a concept approach that minimizes ATLO/INSRP cost/risk).

• Finally, if Kilopower succeeds, then we can have great confidence in the risk, cost and schedule of a 10 kWe surface reactor, and so on….

Slide 3


What led us to Kilopower? Past programs have offered good examples of what not to do.

• High performance reactors that stretched technical limits, and required nuclear testing well beyond existing infrastructure (facilities and bureaucratic).

Backed away from a binary “mission driven” mentality. • In the past, the desire to maximize mission-pull coerced engineers into promising

systems that were too big of a first-step. — Starting with FSP and more so with Kilopower, technologists are being incorporated

earlier in the process of creating requirements.

Development risk is highly dependent on reactor power. • All aspects of design, technology and testing get tougher with increased power, but

in most cases these are small, manageable incremental risks. • The key is to avoid knees-in-the-curve where power level can lead to possible

failure, a new facility, and/or drive you from one design approach to another (temperature, lifetime, mass reduction obviously play a role too).

Tired of the decadal rise and fall of fission power support • Need a sustainable stepwise approach if we want to produce more than stacks of

paper (that mostly rehash the same old technology trade studies).

Uncertainty in Pu-238 supply • Pu-238 was one of John Casani’s reasons for the 2010 study, although if you ask

him, he’ll tell you that he’s even more interested in Kilopower as the first step towards truly visionary space science and exploration.

Slide 4


Small Reactor Study (2010-2011) Many of the same cast of characters

• Study Lead: Lee Mason (GRC) • GPP Representative: John Casani (JPL) • NASA: John Elliott, Bill Nesmith, Duncan

MacPherson, Tom Moreno (JPL), Mike Houts (MSFC)

• DOE: Ryan Bechtel (HQ), Jim Werner (INL), Lou Qualls (ORNL), Dave Poston, Rick Kapernick (LANL), Ron Lipinski, Ross Radel (SNL)

• Consultants: Sterling Bailey, Abe Weitzberg Results: A 1 kWe FPS concept using UMo

core, heat pipes, and thermoelectrics. • ~15 kWt reactor • ~850 kg system mass (incl. margin)

Slide 5


Reactor Technologies Selected by Small Reactor Team • Reactor fuel: UMo (high U loading, easy fab, existing infrastructure, thermal props). • Fuel configuration: Cylindrical casting • Maximum fuel temperature: 1200 K (defined by INL) • Core thermal power: 15 kWt • Core size: 12.9 cm OD, 30 cm tall • Core structure: Ni-based superalloy (Hastelloy X) with Mo liner • Radial reflector and size: BeO, 7.7 cm thick • Axial reflector and size: BeO, 5 cm thick top, 7 cm thick bottom • Reactivity control: Center B4C control rod and sliding B4C safety collar • Control actions during mission: Not required • Reactor heat pipes: Superalloy envelope with potassium working fluid • No. and size of reactor HPs: 18, 1.11 cm OD ea, approx. 3.5 m long • HP condenser temperature: 1100 K • Gap between reactor and shield: 5 cm • Shield materials: Lithium hydride (LiH) and tungsten in stainless container • Reactor dose at spacecraft: 25 krad and 1x1011 n/cm2

Slide 6

These decisions made at the kick-off meeting at GRC or the follow-on meeting at JPL, based on presentations/discussions of pros-and-cons.


Kilopower (2012-present)

During the small reactor study, we thought… “15 kWt, wow that’s the lowest power you’d ever consider, it can’t get any easier than that!”.

• Afterwards, we actually started think of several reasons why development would get easier at lower power.

We asked GRC what was the lowest thermal power they could utilize to make a useful system for NASA.

• The initial response was that we could use 8 ~125 We Stirling engines for an attractive system option at ~4 kWt, ~1 kWe (and perhaps down to 500 We)

The “light-bulb” moment was when we learned that the critical assemblies at the DAF frequently run at powers of ~5 kWt for short periods of time (hours).

• And there was nothing in the authorization basis that could prevent us from running an experiment in a similar manner.

We then started to investigate what kind of nuclear demonstration we could perform at DAF, in a short time for a small amount of money, and 6 months, <<$1M ….

Slide 7


DUFF: A “Critical” Starting Point

Demonstration Using Flattop Fissions • Highly Enriched Uranium core with central

hole to accommodate heat pipe. • Heat transfer from core to PCS via single

water heat pipe. • Power generation via two opposed free-

piston Stirling Engines - Similar to ASCs. • Chiller used to cool cold end of Stirlings.

Test Objectives • Use electric power generated from nuclear

heat to power a load (light panel) • Demonstrate that basic reactor physics was

well characterized and predictable using current analytic tools

Slide 8

GRC EE35-Buzz Stirling Convertor

Assembly

Notional Flight Concept

DAF Flattop Critical

Experiment


DUFF: A “Critical” Starting Point Significance

• First nuclear space power demonstration in over 40 years • First heat pipe power reactor of any size • First reactor power system using Stirling conversion • Demonstration of reactivity feedback with prototype components • Demonstration that testing can be done affordably • The integrated reactor-heatpipe-Stirling power system operated in a

predictable manner – no unknown-unknowns. • Codes/properties/models predicted/reproduced results remarkably well.

Core and reflector Stirlings


Summary of DUFF Results

10

1st DUFF Test: Sept 13th, 2012 • Goal: Produce electricity and light

the LED panel. • Result: Produced peak power of 24

Watts. Reactor was scrammed shortly thereafter, and power decreased as reactor cooled down.

2nd DUFF Test: Sept 18th, 2012 - Goal: Create data pertaining to

transient response, and if possible, demonstrate reproducibility of power/temps in Test 1.

- Result: Successfully increased convertor power and then cycled it off and on. Recreated 24 W at same temperatures. Twenty minutes of Stirling operation.


Sept 13th Results Compared with System Model

Slide 11

Reactor Thermal Power

Electric Power

Power Conversion System Temperatures


4.3-kWt Kilopower Reactor Concept

Slide 12

Reactor State k-eff

Cold-BOL-Rod out 1.0321 +/- 0.0005

Cold-BOL-Rod in 0.9582+/- 0.0005

Warm-BOL-Rod out 1.0125+/- 0.0005

Warm-EOL-Rod out 1.0112 +/- 0.0005

Peak fuel a/o burnup at 4.3 kWt, 15 yrs is 0.08% (essentially zero from a nuclear perspective).

Biggest concerns might be material diffusion and long-term fuel creep of fuel due to any stresses/loads induced by expansion and support structure.

• If we find this is a concern, we could 1) increase Mo fraction 2) decrease temperature, and/or 3) place fuel in can.

• These issues are a focus of the upcoming KRUSTY program.

Heat Pipes (L=~3m)

Startup rod (L=27 cm)

Fuel (L=24 cm)

Radial Reflector (L=38cm)

Heat Pipes (OD=0.95cm)

Startup rod (OD=4cm)

Fuel (OD=11cm)

Radial Reflector (OD=31cm)

Upper Reflector (L=9cm)

Lower Reflector (L=7cm)


Kilopower Nuclear Safety Uranium-fueled fission systems are benign radiological threats until

they operate under fission power. There are 3 potential nuclear safety issues

• Preventing inadvertent criticality prior to operation • Preventing re-entry to populated area (after the reactor has operated, mission

dependent) • Safety during ground handling and testing – handled by existing DOE orders.

Kilopower is designed to remain subcritical under all possible launch accident or ground handling scenarios.

Slide 13

Reactor State k-eff As launched, immersed in sand 0.9839 As launched, immersed in water 0.9901 As launched, Immersed in wet-sand 0.9894 Fuel only (no rod) bare 0.6192 Fuel only (no rod) immersed in sand 0.9572 Fuel only (no rod) immersed in water 0.9921 Fuel only (no rod) immersed in wet-sand 0.9763

The only plausible way the Kilopower reactor can go critical is with the reflector on and the startup rod out – we propose threading the rod into core, so it can only come out with a controlled rotation/translation.

All cases assume the worst case condition where all of the HP void and Na has been replaced with water.


Thermal Performance at 4.3 kWt

Slide 14

Component delta -T Proposed HP Na temp 1050 K Fuel Conduction ~40 K HP wall/internal ~13 K Radiation gaps >>100s K He gas gaps 10s K Sodium bond ~1 K Forced contact gaps ????

With a failed heat pipe, the fuel dT goes to ~90 K and HP dT goes to ~22 K. The conductance between HPs and fuel under pressurized contact in a vacuum is a big uncertainty. Plan to experiment with this early on.


Cross sectional view of the 4 NPAS cores (each schematic box is 16x16 cm)

Slide 15

kpwr1a: 4.3 kWt 8 3/8” HPs

kpwr1b: 13.0 kWt 12 1/2” HPs

kpwr1d: 43.3 kWt 24 5/8” HPs

kpwr1c: 21.7 kWt 18 .525” HPs

Cores are configured so that failed HP peak fuel temp is similar to 4.3 kWt core Nominal fuel temps are actually much lower in the higher power cores


Case Comparison

Slide 16

kpwr1a kpwr2a kpwr3a kpwr4a LANL case designator

4.3 13.0 21.7 43.3 Reactor Power (kWt)

15 15 15 15 Full-Power Years

8 12 18 24 Number of heat pipes (cm)

2.44 2.47 2.31 2.12 Fueled core L/D

11.0 12.0 13.2 15.0 Core OD (cm)

24.0 26.0 27.0 28.0 Fueled length (cm)

9.0 9.0 9.0 9.0 Top Axial Reflector length (cm)

7.0 7.0 7.0 7.0 Bot Axial Reflector length (cm)

0.952 1.270 1.334 1.587 Heat pipe OD (cm)

0.775 1.092 1.156 1.410 Heat pipe ID (cm)

10.0 10.0 10.0 10.0 Radref thickness (including can) (cm)

0.1 0.1 0.1 0.1 Radref outer wall thickness (cm)

1.9 2.2 2.5 2.9 Fuel volume (liters)


Heat Pipe Performance

Sodium Potassium

Sodium is better thermal-hydraulically and neutronically, but vapor pressure at <1000 K is to low to provide adequate vapor flow (Mach limited). One advantage of K is less activation, but not a big deal for HP reactor (but can be important for pumped LM reactor)


Thermal/Mechanical Bonding of Internal HPs • Structural Options: Braze, Weld, HIP, ??

• Haynes 230 to UMo • Provides mechanical contact as well as

thermal bond • May reduce/replace other mechanical

clamping mechanisms • “Hard” bonding will remove some flexibility

when considering ATLO

• Non-Structural Options • Provides thermal contact only • May be solid or liquid • May require core containment can • Na is a possibility • Foils are current leader


Candidate Braze Alloys:

Alloy Solidus (°C) Liquidus (°C) Torr at 850C1 Lowest Tm BVAu-2 891 891 2.9x10-7 U-Au 855C BVAu-10 955 970 5x10-7 U-Au 855C BVAg-0 961 961 1.8x10-4 U-Ag, U-Cu 950C BVAg-6b 779 872 6.8x10-5 U-Ag, U-Cu 950C BVAg-8 779 779 1.1x10-4 U-Ag, U-Cu 950C BVAg-30 806 809 1x10-4 U-Ag, U-Cu 950C BVAg-31 824 852 8.6x10-5 U-Ag, U-Cu 950C BVAg-32 900 950 8.5x10-5 U-Ag, U-Cu 950C 63Ag-37Cu 779 815 9.1x10-5 U-Ag, U-Cu 950C

1Caluclated vapor pressure based on rule of mixture, appears to be conservative estimate compared to data sheets

BVAu-2 is 80Au-20Cu wt% (56.3Au-43.7Cu at%) BVAu-10 is 50-50 Au-Cu by wt% (24.4 at% gold, 75.6 at% Cu) BVAg-0 is 99.95Ag, 0.05Cu wt% (99.915at%Ag, 0.085 at%Cu) BVAg-6b is 50 Ag-50Cu wt% (37.1 Ag-62.9 Cu at%) BVAg-8 is 72 Ag-28Cu wt% (60.2 Ag-39.8 Cu at%) Pacusil-5/BVAg-30 is 68 Ag-27Cu- 5Pd wt% (57.2 Ag - 38.5 Cu - 4.3 Pd at%) Pacusil-10/BVAg-31 is 58 Ag-32Cu-10 Pd wt% (47.4 Ag - 44.4 Cu – 8.3 Pd at%) BVAg-32 is 54 Ag-21 Cu- 25 Pd wt% (47 Ag – 31 Cu – 22 Pd at%) 63Ag-37Cu (50 Ag – 50 Cu at%) Cu side of eutectic for lower vapor pressure, narrow mushy zone

Cheryl’s 1st cut recommendations shown bold. Down selected based on lowest vapor pressure, lowest braze temp, and 2-phase eutectic temperatures greater than 900C.


Nuclear Parameters

Slide 20



0.09% 0.22% 0.32% 0.56% Fuel Burnup (FIMA)

0.13% 0.33% 0.48% 0.84% Fuel Swelling (Vol%)

2.3 6.0 8.7 15.1 Power density (W/cc)

28.4 32.9 37.9 43.7 Total U235 Inventory (kg)

0.0% 0.1% 0.2% 0.2% Radref Be swelling

0.0005 0.0014 0.0020 0.0035 Burnup Reactivity Defect

0.0009 0.0023 0.0031 0.0055 Swelling Reactivity Defect

0.0014 0.0037 0.0051 0.0090 Total 15 year Reactivity Loss

0.0183 0.0167 0.0168 0.0167 Temp Defect (expansion and xs)

1.95E-05 2.20E-05 2.25E-05 2.30E-05 Average fuel RTC

4.9 11.2 15.1 26.0 dT (K/yr) w/o rod movement


Metal Fuel Swelling and Fission Gas Release

0

10

20

30

40

50

60

70

80

0 2 4 6 8 10

Burnup (at%)

Fiss

ion

Gas

Rel

ease

(%)

FIGURE 1. Fission Gas Release from U-10%Zr Fuel.

This data is for U10Zr. UMo behaves similarly, and may be a little stronger (perhaps less swelling). Note that the swelling data has various plots as a function of Pu loading, but trends are similar. UMo swelling is being modeled with swelling equation recommended by Dan Wachs, INL.


If a can is needed then mass penalty is modest.

Slide 22


Kilopower Flux/Fluence

Fluxes (4 kWt)


Radial Reflector Selection -- Be

Be and BeO are the only practical reflector materials that can meet neutronic requirements at a reasonable mass.

BeO is the ideal reflector material for this application

• Lighter/smaller reactor, less power peaking than Be.

We still should be easily be less than 1% swelling up to 50 kWt


Haynes 230 Ductility vs Fluence?

0

5

10

15

20

25

1.0E+20 1.0E+21 1.0E+22 1.0E+23

Total Fluence (n/cm2)

Elo

ngat

ion

(%)

Minimum ValuesEBR II Irradiation, Tirr = 563-866 KNuclear Systems Materials Handbook

Homer-15End-of-Life Fluence

Plate

Pipe, Tube

At 50 kWt our fluences (~1 dpa) are still well below any significant loss of ductility in SS316, but nickel alloys are often worse, so must consider Haynes 230 performance. We are running so warm that any irradiation may anneal our regardless.

Note: This data for stainless steel!


Reactor Masses



--Component Masses

32.9 38.1 43.8 50.5 Fuel

3.6 3.8 4.2 4.7 Axial Reflector

0.0 0.1 0.1 0.1 Fuel liner/can

4.1 10.4 18.2 34.8 Heat pipes (entire length)

0.3 1.1 2.0 4.3 Heat pipe coolant

72.5 80.0 86.7 95.6 Rad Ref Meat

4.5 4.9 5.3 5.7 Rad Ref Clad

4.1 5.2 7.4 9.9 Safety Rods + mechs

12.2 14.3 16.8 20.6 Rx structure (+ shield attach)

134.2 157.8 184.4 226.3 Total Reactor Mass

Slide 26


Kilopower Shield (shown is kpwr1a = 4.3 kWt/15 yr)

Slide 27

LiH neutron layer (kpwr1a=15.2 cm)

1-mm SS can

Cone Intersect

Top Hat LiH kpwr1a=3cm) DU gamma layer

(kpwr1a=0.75cm)

Possible heat pipe penetrations (plugged in kpwr1a, HPs bend around shield)

Shield separation (kpwr1a=5cm)

kpwr1a = 62 kg LiH, 17 kg SS, 91 kg DU


Fast Neutron Fluence Behind Shield (4.3 kWt, 15 yr)

Slide 28

Fast fluence directly behind shield is <1e14 n/cm2.

Converting to flux • 3e20 n/cm2 = 6e11 n/cm2-s • 1e14 n/cm2 = 2e5 n/cm2-s

Other dimensions • Back of shield = 78 cm • Front of shield = 26 cm • Core center = 0 cm • Bottom of reactor = -20 cm. • Bottom of rod = -32 cm

Shield = 62 kg LiH, 17 kg SS, 91 kg DU


Gamma Dose Behind Shield(4.3 kWt, 15 yr)

Slide 29

Gamma dose (Rad Si) is <5 Mrad directly behind shield.

Converting to rad/hr • 5 MRad = 40 Rad/hr • 25 kRad = 0.2 Rad/hr

Other dimensions • Back of shield = 78 cm • Front of shield = 26 cm • Core center = 0 cm • Bottom of reactor = -20 cm. • Bottom of rod = -32 cm

Shield = 62 kg LiH, 17 kg SS, 91 kg DU


Summary of Reactor Concept Mass

Slide 30



--Reactor Module Mass

134.2 157.8 184.4 226.3 Reactor+Structure

10.0 20.0 20.0 20.0 I&C

170.5 255.4 311.6 415.8 Shield

314.7 433.3 516.0 662.1 Total Reactor Module Mass


Alternative Shield Masses

Slide 31

4.3 kWt 13.0 kWt 21.7 kWt 43.3 kWt


Want to get sportier?? I don’t, but you could...

Slide 32

These are masses for the 1 kWe system. Enhanced (enh) means: Bare-U as opposed to U-7Mo, a BeO follower as opposed to none, a 1 mm gap between reactor and axial shield as opposed to 5 cm Development cost and risk increase significantly from left to right. Many factors could reduce shield mass even further.


Enhanced Reactor Schematics U235-7Mo U235 (enhanced) U233-7Mo U233 (enhanced)

Slide 33

31.5 cm 27.6 cm

23.9 cm 18.3 cm

Each concept to scale with each other, but axial schematics not to scale with radial. Yellow boundary is B4C for launch safety.


Notes on UMo/HP Reactor Scalability There is no inherent limit to the power attainable from a block-fuel reactor that is

cooled by heat pipes. • As with any type of reactor concept, there is a trade between thermal power and development

risk, as well as performance (mass, reliability, safety) and cost. • The 4-kWt Kilopower reactor takes advantage of numerous simplifications that make

development risk exceptionally low.

At >~5 kWt external HPs are no longer practical, i.e. the concept will require HPs embedded in the fuel region – which introduces fabrication and integration issues.

• Embedding helps by reducing the fuel temperature dT by >2x, while also decreasing the HP radial heat fluxes/temperature by ~2x).

At >~5 kWt the increased size of the fuel block will be bigger than the most standard, simplest shipping container,

At powers >~10 kWt more heat pipes (or larger HPs) will be needed, • Excess throughput margin of the heat pipes will be lost. • As power increases above ~10 kWt the number/size of heat pipes will continue to increase

(depending on several factors). This is increases design/integration complexity.

At powers > 10 kWt the concept will become more and more dependent on high, reliable component conductances at both ends of the heat pipes

• Perhaps braze or diffusion bond, which will add development risk, and potentially impact ATLO.

Slide 34


UMo/HP Reactor Scalability At somewhere between 5 and 10 kWt a control system will be needed to

compensate for burnup reactivity loss during the mission. • Albeit a very simple control system, with control component lifetime providing the largest obstacle.

Somewhere between 5 kWt and 20 kWt a nuclear test at the DAF will no longer be an “easy” option.

• A significant amount of safety work might be needed to propose a higher power DAF test, and the the program would incur the programmatic risk involved with getting approvals. At 5 kWt we are at the same power as DUFF, thus we are in a known, proven regime.

At powers >~50 kWt fuel swelling may become a significant engineering issue, such that a pin-type reactor or a more robust fuel block (e.g. cermet) would be needed.

• This issue might require significant development effort, thus 50 kWt is a reasonable near-term limit, at least until more work can be done.

PCS integration becomes a development issue above 50 kWt., unless… • Specialized Stirling engine heads and integration hardware can be developed • Or for a surface system where a gravity assisted boiler could be used to transfer power from the

heat pipes to Stirling heads. • Or a Brayton system is used (for which HPs integrate well via a heat exchanger).

In addition to these “knees-in-the-curve”, many reactor design issues become more difficult with power; including, reflector/shield thermal management, ground testing, required/achieved control worth, launch accident safety, etc.

• This is true for any type of reactor concept, not just Kilopower.

Slide 35


The upcoming 3 year Kilopower Project or KRUSTY

Small agile team primarily of GRC, MSFC, LANL and Y-12 End goal of TRL6 (System Demonstration in a Relevant Environment) and

handoff to SMD Science Mission or OCT Tech Demo Mission Three year project with clearly-defined deliverables to retire technology

risks • Year 1: System Reference Concept, Prototyping, Materials Testing (non-

nuclear) Heat Pipe and Stirling Development, Reactor Electrical simulator • Year 2: Non-nuclear System Test at GRC • Year 3: Nuclear System Test at DAF, Flight Pre-phase A Study

From a reactor perspective, there are 3 major differences between the proposed test and DUFF:

• Prototypic temperature – ~1100 K vs 570 K • Prototypic power – 4 kWt through heat pipes • Prototypic parasitic heat loss –

— Dynamically, this was the biggest difference between DUFF and a flight reactor, because heat loss from the fuel to the reflector could not be prevented. Note that DUFF did model the dynamic behavior of one module relative to the rest of the system.

36


FY 15 Thermal Prototype

• Vacuum Test • Stainless Steel Core • Electrically Heated • Haynes 230/Na

thermosyphons • MLI insulation • Prototypic Core Can

• Addresses

1. 3+ clamp designs 2. Core can design 3. Thermal Interfaces 4. Creep modeling 5. MLI performance 6. Assembly process 7. Electrical heater


Thermal Prototype Hardware

Slide 38

FY15 also includes lots of thermo-mechanical material testing and other separate effects testing


FY 16 DU Core Thermal Vacuum System Test • Thermal vacuum ground test

• 900C core operation • Flight prototypic 230/Na heat pipes • (1) pair of dual opposed ASC

modified convertors • (3) pairs of Stirling simulators • Cold end Ti/H2O heat pipes w/ Al fins • KRUSTY core can

• Addresses 1. Thermal cycling effects of DU 2. CTE of UMo 3. Thermal interface verification 4. DU creep characteristics 5. Clamp design 6. Heat Pipe performance 7. System dynamics 8. Reactor simulations

FY-17: Full power ground nuclear demonstration Kilopower Reactor Using Stirling TechnologY = KRUSTY

• Flight-like technologies • HEU core • BeO reflector • Haynes230/Na heat pipes • Stirling converters

• Integrated into flight-like power system • Heat-pipe-to-fuel bonding • Axial shield integration • Bent heat pipes for thermal stresses • Heat-pipe-to-Stirling-bonding

• Tested at flight-like conditions • 1050 K heat pipe vapor temp • 5-kWt powered operation • Prototypic system dynamics • Simulation of various transients • Tested in vacuum chamber

• Exercising flight-like infrastructure • Design, model, fab, test capabilities • Acquire nuclear and material data • Zero-power, powered nuclear testing • Ground safety issues • Transport and assembly issues • Integrating the regulators • Interagency cooperation

Key things missing from KRUSTY: radiator, full suite of Stirlings, startup-rod system, zero-g, launch approval, flight hardware, launch loads, flight qualification, lifetime effects, spacecraft integration.

This applies to, and is sorely needed to complete ANY future space reactor project.


Slide 41


Current KRUSTY Shield Model Red=U8Mo, Blue=BeO Green=SS, Yellow=B4C (70%TD)

Outer Reflector Sleeve (need for is tbd) 19.264 cm IR, 19.353 cm OR

Inner B4C Can 19.853 cm IR, 21.123 cm OR

Inner Steel Ring 21.441 cm IR, 26.521 cm OR

Outer B4C Can 26.838 cm IR, 30.013 cm OR

Outer Steel Ring 30.331 cm IR, 37.633 cm OR


DUFF vs krstb0: Neutron Flux >100 keV

DUFF KRUSTY

KRUSTY fast flux throughout room is expected to be ~3.5 times lower than DUFF (Flattop).

n/cm2-s


DUFF vs krstb0 : Neutron Flux <100 keV

DUFF KRUSTY

KRUSTY moderated flux throughout room is expected to be ~2.5 times lower than DUFF (Flattop).

n/cm2-s


DUFF vs krstb0 : Gamma Dose Rate (Rad Si)

DUFF KRUSTY

KRUSTY gamma dose rate throughout room is expected to be ~40% lower than DUFF (Flattop).

n/cm2-s


Creep Related Issues

• Thermal Bonding to Heat Pipes

• Contact forces needed for thermal bonding

• Forces induce localized stress causing creep

• Creep relaxes contact forces

• Core Deformation over time

• ID control Rod Clearance

• Neutronics Impact


Nuclear Safety for Space Reactors Space reactor safety engineering is simplified by 3 conditions

• Reactor does not present a nuclear safety risk prior to fission-powered operation.

• Reactor will not be powered-up until a stable orbit or escape is established.

• For robotic missions there are no safety issues during mission operation and for human missions radiation risk is in the noise compared to other mission risks.

— These conditions greatly simplify design because engineers can focus on features and events that impact reliability.

There are only 2 potential nuclear safety issues • Preventing inadvertent criticality prior to operation • Preventing re-entry to populated area (after the reactor has operated)

— Even then, it is likely that shrapnel will present a greater risk than radiation.

These conditions simplify safety testing requirements • Zero-power critical testing is required to verify nuclear safety • Non-nuclear testing is needed to test the control system, mechanical

safety locks, etc. • Nuclear-powered testing is not significant to safety - only to reliability

— Program decision: Does the increase in reliability justify the cost (as well as the possible safety risk of a ground test)?

Space reactors pose no radiological risk prior to operation.


Precluding Inadvertent Criticality Space reactors might encounter unique launch and transport accidents.

• It has not been shown, and it is unlikely that an accidental criticality event (burst or sustained) could significantly increase integrated mission risk to the public.

• Current systems are designed to prevent inadvertent criticality as a conservative programmatic measure.

Inadvertent criticality can be effectively precluded via design and technology selections.

• This is most easily achieved for low power systems <100 kWt. • The key is high neutron reflector worth and low void fraction in the core. • Design of control elements to not allow “withdrawal” except when instructed to.

Current design practice is to maintain subcriticality in all feasible scenarios. • Reactor compacted and surrounded by sand.

— Pure SiO2 is the most limiting material that could feasibly surround the reactor after an accident.

• Reactor immersed/flooded with water and surrounded by wet-sand. — The worst-case feasible scenario in terms of reactivity insertion.

• The only potential for criticality is a “uniform” hydrodynamic compaction. — Efforts that have modeled high speed impacts have not been able to achieve this condition

Slide 48


Reactor load following example with lumped fuel feedback… 10% increase in thermal power removal

This scenario assumes an immediate 10% increase in power draw by the power conversion system. No reactor control action is simulated. No secondary feedback from the PCS is modeled.

This is a simplified example with only one feedback coefficient – bulk fuel temperature. Note that the centerline temperature settles higher than it started, because of the larger fuel temperature gradient caused by the higher power. The “actual” system response will not be this simple, but can be similar if the system design is “neutronically” simple.


Most important things to make KRUSTY “prototypic” to flight system (in order of importance) Power

• Full electric power for 1 or 2 modules, dummy heat rejection for others. • Core thermal power can be different than flight (losses, etc.), but needs to be in the same ballpark.

Core materials • It would be a significant negative if we need/want to change fuel, coating, HP materials after test

Core steady state temperatures • For material and expansion concerns, plus neutronics (plus want high temp to produce power anyway)

Core heat transfer, dynamics • Demonstrate stable operation and inform safety transients for follow-on engineering-unit nuclear test. • Most important: the thermal coupling of core to heat pipes, also coupling of Stirlings to HPs • Heat pipe internal design also important, but gravity taints a little

Core geometry • Solid piece, or number of fuel pieces same as flight, • Similar conduction paths to heat pipes

Stirling heat rejection • To give appropriate dynamic response (this assumes we have prototypic Stirling)

Reflector material • To satisfy beryllium neutronics questions

Reactor control/startup system/rod • Not too important because flight system doesn’t have active control

Reflector temperatures • Include reflector feedback/dynamics – less important because substantially slower time constant.

Shielding • Hard to benchmark any sort of shielding characteristics with room/equipment scatter

Slide 50


Sept 13th Temperature Comparison

Slide 51


Slide 52

Relative Reactor Thermal Power

Electric Power

Power Conversion System Temperatures

Sept 18th Test Results Compared with System Model


Sept 18th Temperature Comparison

Slide 53


Material Interface • UMo Core has special interface requirements

• Low Melting Eutectics (incomplete list): • Fe, Ni, V, Si, Mn, Co, Cr • Fe and NI are the most common bad actors,

Vanadium can bring down the melting point a little bit, and so can silicon. Manganese be similarly bad as Fe and Ni, and Cobalt and Chromium could be bad actors as their eutectics are around the operating temp of the reactor. PPM of any of these probably won’t sink us, since it will keep diffusing in, but localized contact will not be good. (JC email)

• Potential Options: • Mo, Au, Cu, ?? (Mo is the front runner)

• Diffusion Studies needed to determine long term stability

Wt% U Wt % Eutectic Temp

88 12 Ni 740 C

95 5 Cr 860 C

89 11 Fe 725 C

5 95 Ag 950 C

24.5 98.5

75.5 Cu 1.5 Cu

950 C 1081 C

14.7 85.3 Au 855 C

79 21 Pd 998 C

90 10 Pt 1005

Eutectic temperatures from binary phase diagrams:

Documents

THERMAL POWER SCALING OF THE KILOPOWER …anstd.ans.org/wp-content/uploads/2015/07/5135_Poston-et-al.pdf · THERMAL POWER SCALING OF THE KILOPOWER SPACE REACTOR. Slide 1 . ... Past