ML20315A105

NRC-2018-000831 - Resp 6 - Interim, Agency Records Subject to Request Enclosed, Part 1 of 2

ML20315A105
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Issue date:	10/30/2020
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7/23/2018 NRC-N 1ST Agreement Assessment of Condition Monitoring Techniques for Electrical Cables NIST Pl: Stephanie Watson NRC PM: Darrell Mur:dock NIST" ..

U.S.0.,.0,,,-,,olC-Objectives

• Evaluate the effectiveness of commonly used cable condition monitoring for specific insulation materials

• Assess qualified life predictions to 50 years, 60 years and 80 years of operation

• Enhance the understanding of condition-based environmental qualification methodology 1

7/23/201?

NRC Agreement Tasks

1. Acquisition of materials

2. Develop a test plan

3. Establish baseline values for cable properties using condition monitoring methods, as described in the test plan

4. Simulated operational aging and condition monitoring testing

5. Simulated accident and condition monitoring testing

6. Develop NUREG/CR Acquisition of Materials

• Representative polymer insulation and manufacturers

• New cables: current year and vintage

• In-service cables: Zion and Crystal River NPPs

• Some active and some not active

• Cables selected for long-term aging experiment

• 'Vintage'

• Rockbestos: 1980, XLP/CSPE (3C, 16 AWG)

• Anaconda: ??, EPR/CSPE (2C, 14WG)

• Okonite: 1985, EPR/CPE {3C, 14AWG)

• In-service

• Zion (non-active): BIW, EPR {3C, 16 AWG) 2

7/23/2018 Develop a Test Pian

• Selection of Condition Monitoring Tests

• Simulated Operational Aging Experiment

• Choice of Radiation Facility

• Justification of Accelerated Aging Conditions

• Development of Environmental Chambers

• Choice of Cables to Study

• Loss of Coolant Accident Conditions Condition Monitoring Baseline Measurements

• Mechanical Properties

• Elongation At Break {Tensile)

• Indenter {Modulus)

• Electrical Properties

• Frequency Domain Reflectometry (modific.ition for short cable lengths)

• Line Resonance Analysis

• CHAR package

• Tan Delta (Dissipation Factor)

• Insulation Resistance

• Dielectric Spectroscopy (with EPRi)

• Chemical Properties

• Oxidation Induction Time/Temperature

• Thermal Gravimetric Analysis

• Fourier Transform Infrared Spectroscopy

• Raman Spectroscopy 3

7/23/2018 Simulated Operational Aging

• Accelerated aging to simulate normal operational aging in mild and harsh environments

• Simulate 50, 60 and 80 years of operation

• Perform condition monitoring testing at regular intervals Long Term Aging Experiment- Sandia 16-month irradiation experiment 22-month total (GIF maintenance\mandrel shipment)

.t G(3~/n ~y,. \ i s1 C, 60 % RH 69 o/o I \*,'>C,;,o 4

7/23/201 8 Simulated Accident

• Cables from operati onal aging experim ent exposed to accident conditio ns

• IEEE 323-1974 (PWR Generic Profile)

• Cables under accident conditions - tempera ture, pressure, humidity, radiation, chemical/steam spray

• Cables connected to power carts and excited with voltage

• Condition monitor ing techniques will be evaluated during and after the accident (post-accident period)

• Visual inspection, dielectric withstand , and insulation resistance

• Radiation Dose Rates: 0.5 -1 MRad/h r

• IEEE 383 (Post-Design Based Event Testing)

Develop NUREG/CR

• Create a publicly available docume nt delinea ting the assessment of conditio n monito ring techniques and confirm ing the adequacy of tests 5

.&. I I .,.... Government of Canada Gouvernement du Canada Canada Evaluating Structures1 Systems & Components from Decommissioned/Decommissioning Nuclear Facilities in Canada Presented at the Nuclear Regulatory Commission Harvesting Workshop - March 7-8, 2017 Daniel Tello - CNSC Senior Research Officer Desire Ndomba - CNSC Technical Specialist Karen Huynh -AECL Federal S& T Manager

$ cANADA150

Outline

• Who we are

>- Canadian Nuclear Safety Commission (CNSC) e

>- Atomic Energy of Canada Limited (AECL) ~; ~

• Canadian Harvesting Projects & Sources

>- Nuclear Power Plants and Research Reactors

./ Nuclear Power Demonstration (NPD) Generating Station

./ Gentilly-2 Nuclear Generating Station

./ Nuclear Research Universal (NRU) Reactor

• Summary 2

Canadian Nuclear Safety Commission ,~ ,

Canada's nuclear reaulator for over 70 years

• Canadian Nuclear Safety Commission {CNSC):

~ Established on May 2000 under the Nuclear Safety Control Act

{NSCA);

~ Replaced the Atomic Energy Control Board, which was established in 1946 under the Atomic Energy Control Act;

• CNSC's Mission:

~ Regulate the use of nuclear energy and materials to protect the health, safety and security of Canadians and the environment;

~ Implement Canada's international commitments on the peaceful use of nuclear energy;

~ Disseminate objective scientific, technical and regulatory information to the public; 3

Atomic Energy of Canada Limited A'- AECL N' EACL

• AECL is a federal Crown corporation responsible for managing Canada's radioactive waste liabilities and enabling nuclear science and technology

• AECL delivers their mandate through a contractual arrangement with Canadian National Energy Alliance (CNEA) for the management and operation of Canadian Nuclear Laboratories (CNL) under a Government-owned, Contractor-operated (GoCo) model

• AECL oversees the delivery of the Federal Nuclear Science and Technology Work Plan in order to support the Government's priorities and core responsibilities in areas such as nuclear safety, security, energy, health and the environment

• CNL is responsible for the management and operation of the Laboratories, and is the licensee with access rights to AECL's assets and Intellectual Property (IP) 4

Harvesting Materials in Canada Why is Harvesting Important?

• The nuclear power plants (NPPs) that are currently in operation were generally designed and built with conservative principles, and in many cases have significant remaining safety margins. Typically, NPPs were designed with the intent of operating for up to 30 years. This is now changing to approximately 60 years.

• Harvesting materials from decommissioned or decommissioning facilities is intended to assess the remaining safety margins for the physical condition of structures, systems and components (SSCs) in NPPs through testing to verify and/or validate the technical feasibility of Long Term Operation (LTO).

Canadian Harvesting Projects

• Canada currently has harvesting projects in 2 areas from different facilities

~ Analysis of Degradation Mechanisms of Cables

~ Understanding Degradation of Concrete 5

Nuclear Power Demonstration Plant (NPD)

• The 20 MW Nuclear Power Demonstration Plant (NPD) was Canada's first nuclear power reactor to supply electricity to Ontario Hydro's electrical distribution grid

• The reactor permanently shutdown in 1988 after 25 years of service

>"" CNSC sponsored a research project with Ontario Hydro in 1991 on:

"Cable insulation degradation of the 20 MW Nuclear Power Demonstration Plant"

>"" AECL currently has a research project to analyze concrete samples from NPD 6

Understanding the Degradation of Concrete ackground

• This project will address gaps in the domestic knowledge of concrete degradation in nuclear power plants and will support regulatory decisions for long-term reactor operations of domestic utilities and new reactor technologies, such as SMRs.

Project Objective

• To assess concrete core samples for degradation Status

• Year one of a three year project

• Received 7 concrete cores from Nuclear Power Demonstration (NPD) reactor

• Identified test techniques to assess for degradation Other opportunities

• Concrete from other decommissioned reactors in Canada such as Gentilly-2, Douglas Point and AECL's whiteshell reactor in Pinawa, Manitoba 7

Gentilly-2 Nuclear Power Plant

• Gentilly-2 (G2) Nuclear Power Plant (NPP) is a CANDU-6 675 Mwe nuclear reactor which was permanently shutdown in 2012 after 29 years of service

~ CNSC & AECL/CNL are currently conducting a research project on the Analysis of Degradation Mechanisms of Cable Insulation due to Ageing in a Decommissioned Nuclear Power Plant 8

Cable Harvesting Project from G2

• Material properties of cables aged in real time and real operating conditions can be significantly different from the properties received by studying samples aged in laboratory and assumed in codes and standards.

Project Objective

• Assess current cable degradation resulting from thermal and radiation ageing

• Validate environmental qualification work Status

• Reviewed environmental conditions and selected cables to be harvested from Hydro Quebec/Gentilly-2 reactor

• CNSC & AECL/CNL staff have recently witnessed the removal of cable samples from the G-2 NPP

• Samples will be retrieved and tested to improve cable ageing test parameters used to assess cable condition 9

Cable Harvesting Project from G2 (cont'd)

• There were several challenges associated with the cable harvesting project at G2:

)"' removal of material samples from a decommissioned NPP is neither a compliance nor licensing activity for licensees;

)"' accessibility of analysis and testing records (data, samples, etc.)

)"' inaccessibility or contamination of materials to be removed

)"' cost related to carry out research activity on these materials.

' ~

~

/

10

National Research Universal (NRU) Reactor NRU is a 135MWt nuclear research reactor built in the Chalk River Laboratories, Ontario with first criticality achieved on November 3, 1957.

• The NRU reactor has three purposes

~ Supplier of industrial and medical radioisotopes used for the diagnosis and treatment of life-threatening diseases;

~ A major Canadian facility for neutron physics research; and

~ Research and development support for CAN DU power reactors.

• After 5 decades of service, the NRU reactor will shut down permanently in March 2018, after "'270k hours operation.

• The NRU reactor contains a vast array of materials and components including

~ structural materials {such as steels and other nickel bearing alloys, zirconium alloys, aluminum, concrete) and welded joints

~ a thermal graphite column

~ equipment {including pumps)

~ graphite seals

~ electrical cables CoNTRoL

~\)'P T<oD

~ thermocouples 304 SS, - 1.6 dpa, 35 °C,

- 3.4x1026 n/m2 (E<0.625 eV) 11

Harvest Material from National Research Universal fNRU Project Objective

• Create an inventory of irradiated materials which can be harvested from the decommissioned NRU reactor

• The inventory will be used to assess irradiation damage on the performance of in-core materials and components Advantages of Harvesting Materials from NRU

• Assess the effects of irradiation on the performance and degradation of a variety of materials;

};;, wide range of temperature, flux and neutron spectrum

};;, irradiation times on the order of power reactor lifetimes

};;, the combination of operating conditions and exposure is not always easily or economically obtainable in test programs

• Collaborate with various partners throughout the international nuclear industry who also wish to characterize the effects of irradiation on materials

• Provide information to decommissioning groups interested in the radionuclide inventory of the decommissioned reactor.

Status 12

• One year project that will start on April 1, 2017

Summary

• Understanding ageing degradation is an important aspect to ensure nuclear safety as nuclear power plants age

• R&D behind SSC material life management is needed to ensure the safe and reliable operation of NPPs

• Several nuclear facilities in Canada exist for potential harvesting of Structures, Systems & Components. Examples include:

>"" Nuclear Power Demonstration (NPD) Facility

>"" Gentilly-2 (G2)

>"" National Research Universal Reactor (NRU)

>"" Douglas Point

>"" AECL's Whiteshell Reactor 13

Contacts e

Daniel Tello e

Desire Ndomba Senior Research Officer Technical Specialist E-mail: daniel.tello@canada.ca E-mail: desire.ndomba@canada.ca A. AECL N' EACL Karen Huynh Federal S&T Manager

~

• /,~,

Email: khuynh@aecl .ca C: Canad ian Nu clear Laboratories I Laboratoires Nucleaires Canadiens ,,~

~*~ Canadian Nuclear Laboratories I Laboratoires Nucleaires Canadiens Gina Strati Lori Walters Director - Energy Sr Mechanical R&D Email: gina.strati@cnl.ca Email: lori.walters@cnl.ca 14

Questions Thank you!

15

FANC ~ A1FCN -

federaal agentschap voor nuclealre controle agonco fedorale de controle nucleatre BEL STUDIECENTRUM VOOR KERNENERGIE CENTRE D'ETUDE DE L'ENERGIE NUCLEAIRE Review of Past Reactor Pressure Vessels Test Programs and Perspective for Long Term Operation Rach id Chaouadi \ Fran~ois Henry 2, and Guy Roussel 3 1 SCK*CEN Mol, 2 FANC/AFCN Brussels, 3 Bel-V Brussels (Belgium)

Ex-Plant Materials Harvesting Workshop USNRC, Washington, March 7-8, 20 77 1 © 2017 SCK*CEN

Outline Part I : Past test programs analysis

• Focus on RPV materials (excl. stainless steel cladding, internals, vessel concrete containment, ...)

• Literature survey of test programs on decommissioned RPVs

• 2 PWRs + 2 VVER-440 + 1 PWR and 1 Magnox gas-cooled reactor

• Brief description of main outcome and limitations Part II : Discussion (assuming that we are convinced of the benefits of such a test program via vessel sampling)

• Approach / criteria for selection

• Potential issues of interest

• Organizational and financial aspects

• Perspective for the future 2 © 2017 SCK*CEN

Overview of Investigated RPVs BR3 (YR) Belgium PWR (11 MWe) 1963 1987 -1995-1999 Chooz-A France PWR (305 MWe) 1967 1991 - 1995- 1999 1974 Greifswald- 1 to 4 Germany VVER-440 (408 MWe) 1990 -2004-2012

-1979 Novovoronesh-1 Russia VVER-440 (408 MWe) 1964 1984 - 1993-2000 Gu ndremmingen-A Germany BWR (237 MWe) 1966 1977 - 1988-1992 Trawsfynydd UK Magnox (195 MWe) 1965 1991 - 1996- 2002

~ PWR (BR3 and Chooz-A)

~ VVER-440 (Greifswald and Novovoronesh)

~ BWR (Gundremmingen)

~ Magnox gas-cooled (Trawsfynydd) 3 © 2017 SCK*CEN

Overview of Investigated RPVs BR3 (PWR) IAR 260 4 0.2/0.6/ 0.02 single A.:SULts multiple 1.2MD07 Chooz-A (PWR) 255-265 2 0.1/ 0.6/ 0.02 through (A336) thickness multiple Greifswald-1 to 4* I, IA, 255 10KhMFT 2 0.15/0.3/ 0.04 through (VVER-440) IAR, IARA thickness Novovoronesh-1 250 I, IA 15Kh2MFA 1.4 0.15/ 0.1/ 0.03 single (WER-440) (~228) multiple Gundremmingen- 284-288 20NiMoCr26 0.3 0.2/0.5/0.01 through A (BWR) (A 336) thickness Trawsfynydd manual (Magnox-gas 190 0.2 0.26/ 0.1 / 0.04 mixture SA weld cooled)

Legend : I = as-irradiated ; IA = I + annealing , /AR = IA + re-irradiation , /ARA = /AR + annealing

• No relation between units 7 to 4 and I, IA, /AR, /ARA 4 © 2017 SCK-CEN

Overview of Investigated RPVs test reactor (T, CVN, FT)

BM 0.19 0.56 0.02 T, CVN, FT BR3 (YR)

Weld BM 0.08-0.10 0.59- 0.61 0.010- 0.015 T, CVN, FT Chooz-A Weld 0.08-0.09 0.13-0.21 0.010-0.016 FT BM 0.13 0.18 0.01 CVN, FT Greifswald-1 to 4 Weld 0.12-0.18 0.19-0.29 0.032-0.036 CVN, FT BM 0.15 0.13 0.015 Novovoronesh-1 Weld 0.12 0.12 0.033 CVN BM 0.16 0.74 0.013 T, CVN Gundremmingen-A Weld 0.24 0.11 0.009 CVN BM Trawsfynydd Weld 0.24-0.27 0.10-0.11 0.034-0.038 FT T = tensile

~ All low Ni/ medium Cu/ medium to high P CVN = Charpy impact

~ Modern steels : low Cu and low P FT = fracture toughness 5 © 2017 SCK*CEN

BR3 (PWR)

JOO, BR3V1111l(BM)

A!OZB Nl-mod RT-,;

/ /

/K.tAJML RT-,;

/ ,

,~-L 250 , IOYankee surveillance

_ 2so I I f +Yankee Surv. plate j

O nozzle elevation /

• betllne level (4 E19 n/cm1) I / / / 200 ~ 1* PT-A (A533B) BR3 vessel *******

• i 200

-J 150 0

a I '

I

' t.RTne* =9&oc

• 6Tu* 111°( //

,7

/

E 1so e BR3

~

a YAl YA9 *

• ~

• .......*..* * ....*...*~*'at*~

+ *******************. :io lux BR2

i j I_..I~..,I"'.........

..l~ *

~: .~ . . .l.*-*-..-*..~ iJ. ...****j .... J: . . t'I'. full symbols: T1.,oc1 = 260'(

<l 100 * ***** a &V., bol T 288'C

/ ..,/ 1*~~-*/ zoc 100 ..*************** a / open sym s : ind =

I

• ..-* , Cu=0.19-0.24%

so ...............-..u.:::.-_,_ ......,,,;;,,,,a 50 i// . 8R3 IARA 343'C/168h p = O.OlO- 0.026%

Ni= 0.19 - 0.63%

0 - - - . . . - - - - , , - - . . . . . - - -........- - . - - - - - - .~ - ~ 0 -+-----.-----.-----,----..-----.

-150 -100 -so 0 so 100 150 200 0 2 4 6 8 10 temperature (OC) neutron flue nee (1011 n/cm 2, E>lMeV)

~ Conservative regulatory RPV integrity assessment due to Ni-content, temperature and grain size.

~ Actual embrittlement conditions are less than predicted after comparison with other surrogate materials and direct measurements of fracture toughness on the vessel material.

6 © 2017 SCK*CEN

Chooz-A (PWR) 200 , Chooz-A Forgings 0 surveillance (shell B) 0 vessel (shell B} Source:

KJc (MPa .Jm)ASTM STP 7405 (2001) 28-41.

• surveillance (shell C) 250 150 ¢ vessel (shell C) - Chooz-A Weld u

A through-thickness (shell C)

_..-- .. -------+/-15°C

_ .. 200 0

150

i 100

~ ,

100

• so 50 unirradiated K1c o--------------

Data from ASTM STP 1405 (2001) 28-41. *200 -150 -100 -50 0 50 100 0-+"--- - - - - ~ - - - - - - - - - - - , . - - - - - - - - , T (°C) 0 5 10 15 neutron fluence (101' n/cm1, E>lMeV)

~ Demonstration of the validity of the surveillance program for the base metal.

~ Fracture toughness data on the weld metal were all well bounded by the fracture toughness lower bound curve established in the unirradiated condition (K1c based on RTNor).

7 © 2017 SCK*CEN

Greifswald (VVER-440) 50 Greifswald Unit- 1: base metal !AR- condition E 1 20 Greifswald Unit-1 : weld met al !AR-condition.J E o 70 1 2

~ ~

I I I I I I I I I I I I I I I I I I I I I I I I I I I Greifswa Unlt -1 base metal ring O3 1. trepan. 1-4 60 25 ~

50

~

It')

p O ~ ~ ------:i ~ 40 ~

It')

1E19 ~ p

~

2(0 -25

~

Charpy impact N~

~

30 20 1E19c:i w

/\

Qi

~ -50 T 41J - i Qi 10 NE

~ ~ Q. 0 ~

C 1E18 ~ E

-75 --------------- --------  :]

a;:

~ -10 1E18- CI) 0 C -20 C

-100 e

i

-30 Cl)

.2 Cl)

C -40

-125 I 1E17 -50 1E17 0 10 20 30 40 50 60 70 80 90 100110120130140 0 10 20 30 40 50 60 70 80 90 100 110 120130140 distance from the RPV inner wall / mm distance from the RPV inner wall / mm

- - fluence total in n/cm (E>0.5MeV) 2 - - fluence before annealing (E>0.5 MeV) 2 - - - fluence after annealing (E>0.5 MeV)

- - - fluence after annealing in n/cm (E>0.5MeV) 150

• Greifswald weld Unit 4

• as-irradiated 300

• VVER-440Weld 600

• annealed 475'C/152h 0 0 d'. * .. 250 100 *-------... I-con 1t1on ....******** 550 0 **-----. ---------:~.::** *************" * 'i

~to 200 50 .....~ -- *** --..o**--~--

O 500 E .,...4 *****

• "'(,:::.-<*. .

• C

~

~

~ 150 ~

~ ...... **-...

• o* 450 .;; i-!'

0 <l e*****

i ~~ ~..... 0

.,~,.o. *;r:--co ncfi.ti~n*; ***

• ***o---<<>--~ ~---********************* ******** 100

• Greifswald Unit-4 (0.16%Cu/003%P)

O *----.b-- .. 400 > ,,~./ IX No\/ovoronesh- 1 (0.12%Cu/0.03%P)

-50

• full symbols: T0 so Jt 0 Kryukov-01 {0.15%Cu/0.03%P)

• Fortum-00 (0.17%Cu/0.03%P)

• 100 empty symbols : yield strength 350 I 1

• Nikolaev-00 (0.18%Cu/0.03%P) 1 0

0 25 50 75 100 125 150 distance from the RPV inner wall (mm}

0 10 20 30 40 50 60 70 8 neutron flue nee (1019 n/ crrl-, E>O,SMeV) 7 SCK*CEN

Novovoronech-1 (VVER-440) 200 1 : 1.371019 n/cm1, E>lMeV 200 -, Novovoronesh-1 150 ii ~ !~1d I 2

1 80 ~ I I I

. . .... . . . ~¢. . .! ...... o* Q 19 NOVO 4 : 0.9110 n/cm , E>1MeV l * * * * * .* .

7 : 0.581019 n/cm1, E>1MeV 160 Bose metal 2 1so ¢ ............

100 Orientation ZR 9 : 0.411019 n/cm1, E>1MeV 140 Plates 1, 4, 7, 9 100 so

,....., /1 0 ____ ___ ___.. _____ _ ________ ___ _ ..*..**

3 120 u .... ** *

• u 0 ~

0 -;::: so . . ¢ 0

• o .------. ---*-o**-* - - - --

i-?

~ 100 ,/ ., 1-"

,v -" A ---* o ...-'\.

,,...~~*

• • -50 V

0 -t ~ - 0 ***@*********L;/

~ 80 ,.;' ~ ,.--- ..................o *.O*********o**************

_,:.;.-,,. ~o,'0 ....- ...o-*-*-*-*- *- *-*-*_._*-*-*-*-*-*-

60 40 I *1J ' .,*

<> -*-* Plate 1, T47J

• o41"C

-50 q>..*********** -100 0 - - Plate '4, T47J

• 315°C -100 -150 0 *--- Plate 7, T47'

• 215"C 20 /1 *- **** Plate 9, T47J

• 28°C 0 1 2 3 4 Q L_>=:J I I I I I I I I J I neutron fluence (1019 n/cm2, E>O.SMeV) 20 0 20 40 60 80 100 120 140 160 18 0 200 Temperature (°C] ~l.-~

~ Non-uniform through wall distribution of P-content in the weld.

}i;;- Large scatter in embrittlement based on (harpy impact transition temperature T47J*

}i;;- Less scatter when master curve TO is used.

~ No comparison to surveillance data.

9 © 2017 SCK*CEN

Gundremmingen-A (BWR) 100 , KRB-A base material 200 KRB-A base material (T-l)

O archive mat (unirrad) T-L O Shell 7.1 / Trepan G cj>- 2.4 lQl.8 n/cm2, E>lMeV I

cladding/ferrite interface e Trepan G/C (T-L) ~0.30 10*' n/cm2. E> 1MeV

• Shell 7.1 /Trepan C ..-2.4101 8 n/cm2, E>lMeV p- 75 a ? e 1a L ?*L I*-, J410 , n/cn-' E 1Me\.

.l Shell 8 / Trepan Q lj>--3.3 l Ql.8 n/cm2. E>lMeV t,, archive mat (unirrad) L-T 150

...... .l Trel)<ln G/C (l-T): <fL Q.32 lQ*J n/cm2, E>lMeV ,,.~ ------------- A **.

OJ 5Q

5

!?

---,---* ---* I I

I

~ 25 ~ 100 ,' .l E ----.-------------~ --~ --- a, C:

a, a,

--~---~---...----........ . ___ _. ____.

I C:

.Q 0 i I

~ ..........'()'...................................................

.~

~ 0

---

50 .l ,

/ Q..*O

-'( )

~

... -25 / I,

,,a.,/

~ .....

-50 + " " - - - - . - - - - . - - - - . . . . - - - - - . - - - - . - - - - - - , 0 +----.. : : :.- ...------...------...--------,

0 20 40 60 80 100 120 -100 0 100 200 300 distance from cladding interface (mm) temperature ( 0 ( )

200 KRB-A base material 0 UBR 288°( LWR (T-l) lj>--8.5 10* 1 n/cm2s, E>lMeV

> Significant effect of specimen E

• PLUTO 290"C LWR (T-l) lj>--5.0 10 'n/cm 2 s, E>lMeV orientation (T-L versus L-T) on T41 J

.l PLUTO 290"( HWR (T-l) lj>--5.0 10'1 n/cm2s, E>lMeV ~..

~* 150 a DIDO 290 C HWR (T-ll ~ - 5.0 10 l n/cr ' E: lMe\i ,./,.. and USE.

c lj>--1.0 10

• nf cn, 2 s, e* 1MeV

~

! 100 x Surveillance (L-T)

• Trepan G (l-T) lj>:2.010

• 4'=2.0 1011 n/cm2s, E>lMey,***/

,../

> Effect of specimen orientation.

~

a, C.

E o Trepan G/C (T-l) 0

/.,..........

> Significant effect of specimen a,

C: ocx, * ....~,.o location in the base metal.

.~ so

.Q

~.......****

"'C:

...~

0 0

0000 ,

....... ~ -..- * > Uniform through wall chemical 0 , ..............................................* ..; *** * * * . .l composition of the weld.

0.01 0.1 1 neutron fl uence (1019 n/cm2, E>l MeV) 10 10 © 2017 SCK*CEN

Trawsfynydd (Magnox) 0.3 , Magnox Weld I O Sample 1 O Sample 2 t,. Sample 3 a Sample 4 j 250 r - - -50 percentle brittle lnstabilly line I Cu ***ll**............~.............. ~.............. S... I -90 mn 225

- - -- - 5% and 95% brltle lnslablllly lines ,

- *- *-

• Medlanfrac:turetoughnessllne

............ 5% and 95% toughness lines 0

. . . J~ .*

. , "(j ..........

0.25 200 0 Ex-Trawsfynydd RPV data a

• Ex-sUfVelllance dala

~175

~ 0.2 tQ a..

~ 150 to

.! CII

~ 125

~ 0.15 C

~0

.t::.

NI ... ~ .............. .ct..............., .............. ij .. , g' 100

.., 0.1 0

-;; 75 H  :,

0.05 p ... ~ .............. t .............. , .............. o... I.

I -~

25 o ~ - - ~ - - - ~ - - ~ - - - ~ - - - -- - 0 a I It Is t : I :: t: I:: I: I:: SI In I : I I C I t I a I I I cs I I I I : I 111 t i I 1 Ir I s I I ell 1 2 3 4 *100 .75 .SO -25 0 25 50 75 100 125 150 175 200 225 sample location (layer 'i') Temperature {°C) (adjusted to weld 4 layer 1 condition)

~ Uniform chemical composition (through wall)

}i;;- Good agreement with surveillance data

~ Allowed the demonstration of the safety integrity approach followed for this type of reactors.

11 © 2017 SCK*CEN

Outcome reactor outcome Conservative regulatory RPV integrity assessment due to Ni-content, temperature and grain size. Actual embrittlement conditions are less than predicted after comparison with other surrogate materials and measurements of fracture toughness on the vessel material.

Demonstration of the validity of the surveillance program of the base metal. Fracture Chooz-A (PWR) toughness data on the weld metal were all bounded by the fracture toughness lower bound established in the unirradiated condition.

Uniform chemical composition of the weld through the wall thickness except at the weld root but large scatter in the transition temperatures without correlation to the neutron exposure. Scatter attributed to local microstructure (coarse versus fine). Annealing at 475°(/152h allows an effective recovery. No comparison to surveillance data.

Non-uniform through wall distribution of P-content in the weld. Large scatter in embrittlement based on (harpy impact transition temperature T47J. Less scatter when master curve T0 is used. No comparison to surveillance data. Post-irradiation annealing was also investigated and showed that 600°C/ 2h leads to an effective recovery.

Significant effect of specimen orientation (T-L versus L-T) on the transition temperature T41 J and upper shelf energy. Effect of specimen orientation complicated the comparison to surveillance data. A significant effect of specimen location was observed for the base metal. Uniform through wall chemical composition of the weld.

Trawsfynydd Good agreement with surveillance data. Allowed the demonstration of the safety (Magnox-gas integrity approach followed for this type of reactors.

cooled) 12 EN

Limitations

• Materials

• Gen-I materials (BR3 + Chooz-A), VVER-440 welds (high P >0,03%)

• Limited to small areas of the reactor vessels

• Many of these vessels were annealed BR3 (ineffective : 343°C/168h),

Greifswald-1, -2 and -3 (effective: 475°C/152h)

• Material variability (VVER-Welds, Gundremmingen base metal)

• Systematic absence of archive materials (except Gundremmingen)

• Environment

• Majority irradiated at low temperature : 190 to 265°( except Gundremmingen (-288°()

• Change of temperature during operation : Chooz-A (265 ~ 255°()

and Novovoronesh (250 ~ 228°()

• Poor and/or unreliable surveillance program (except Chooz-A)

• Poor and/or unreliable dosimetry 13 © 2017 SCK*CEN

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Approach

• Vessel selection

• Do we have a choice ?

• If yes, what are the selection criteria ? There are many :

Materials composition and damage levels Operation history (neutron fluence, irradiation temperature)

Availability of reliable surveillance data Availability of archive materials Other aspects : material variability, presence of segregated zones, ...

• Choice dependent on the targeted objectives

• What are the priorities

• International participation

• There is no ideal choice (except if infinite available budget)

• Consensus 15 © 2017 SCK*CEN

Selection Criteria

• Materials

• Preferably representative of modern reactors (7 LTO)

• More systematic examination using many parts of the vessel (axial and azimuthal)

• Available archive materials

• Irradiation environment 2

• LTD-representative fluence levels (> ... 1019 n/cm , E> 1MeV)

• Irradiation temperature ~290°C (majority of operating reactors)

• Operation history and monitoring

• Well documented operation history

• Reliable temperature monitoring

• Reliable dosimetry

• Reliable surveillance program 16 © 2017 SCK*CEN

Material Variability

• Material variability ~ scatter

• Complicates the interpretation of the test results

• Increase the number of tests and use statistics

• Important details might be shadowed by the large scatter

• If the vessels are really inhomogeneous

• This should be taken into account in safety integrity assessment

• Check the adequacy of the surveillance program with respect to representativity of the whole vessel

• What should we favor and at which cost ?

• For better understanding and validation of procedures with a minimum of tests (least scatter) or

• Application (real components) to check whether such material variability is covered by the available safety margins

___* Alternative : availability of multiple locally homogeneous areas 17 © 2017 SCK*CEN

Issues

• Materials (at multiple locations)

• Homogeneity of chemical composition

• Homogeneity of mechanical properties

• Identification of segregated area and further analysis

• Identification of LTO-relevant areas and further analysis

• Neutron exposure

• Verification of the neutron calculations f-7 dosimetry measurements

• Through-wall fluence attenuation

• Vessel versus surveillance program

• Mechanical properties : vessel versus surveillance

• Verification of master curve shape at high KJc

• Thermal ageing assessment

• Neutron flux/spectrum effects

• Assessment of regulatory procedures 18 © 2017 SCK*CEN

Organizational Aspects

• International participation

• As large as possible

• In-kind contributions

• Agreement on the project contents

• Main topics

• Vessel selection and sampling

• Materials transportation

• Work share and distribution

• Timeline

• Start date : ... 2017

• 4-5 yr project (benefit of operating reactors) 19 © 2017 SCK*CEN

Financial Aspects

• Vessel trepanning

• If the vessel will be anyway cut in small pieces for final storage, the costs will be limited 7 additional work mainly

• If the vessel was planned to be buried as a whole, then we will have to pay very much 7 all costs

• SCK*CEN experience

• BR3 sampling : several trepans taken from the inner side of the vessel 7 -0,4 M ( of 1995) by PCI Energy Services, Illinois US

• BR3 dismantling team : several large trepans 7 - 100 k ( of 1995)

• Possible more accurate evaluation

• Westinghouse Belgium : Mr Joseph Boucau 20 © 2017 SCK*CEN

Closing Remarks

• To be effective, need of international program (work sharing)

• Focus on most critical issues (prioritization)

• Discussion : What do we want to do ?

• Which reactor vessel ? Materials compositions? Irradiation conditions (temperature, fluence) ?

• Specifications and requirements ?

• Technical issues to be addressed ?

• Work sharing ?

• Financial aspects ? Vessel sampling costs : depend on decommissioning strategy

• Materials management and transportation

• Planning Thanks for your attention 21 © 2017 SCK*CEN

Copyright © 2017 - SCK*CEN PLEASE NOTE!

This presentation contains data, information and formats for dedicated use only and may not be communicated, copied, reproduced, distributed or cited without the explicit written permission of SCK*CEN.

If th is expl icit written permission has been obtained, please reference t he author, followed by 'by courtesy of SCK*CEN'.

Any infringement to this rule is illegal and entitles to claim damages from the infringer, without prejudice to any other right in case of granting a patent or registration in the field of intellectual property.

SCK*CEN Studiecentrum voor Kernenergie Centre d'Etude de l'Energie Nucleaire Belg ian Nuclear Research Centre Stichting van Openbaa r Nut Fondation d'Utilite Publique Foundation of Public Utility Registered Office: Avenue Herrmann- Debrouxlaan 40 - BE- 1160 BRUSSELS Operational Office: Boeretang 200 - BE-2400 MOL STUDIECENTRUM VOOR KERNENERGIE CENTRE D'ETUDE DE L'ENERGIE NUCLEAIRE 22 © 2017 SCK*CEN

ICHNi?

Central Research Institute Kori-1 Harvesting Plan for Materials Aging Degradation Research Al Ahluwalia (EPRI) for Han-sub Chung (KHNP)

March 7-8. 2017

Contents ICHN?

Central Research lnstJtute

1. Background

2. Research planning using materials from retired Kori-1 l

Future Plans

ICHN?

Central Research lnstJtute

1. BACKGROUND

KHNP PWRs and Kori-1 ICHN?

Central Research lnstJtute

• Kori-1 is a Westinghouse two loop PWR

• SG replaced in 1998, RV head replaced in 2015

• Kori-1 will be shutdown on 2017.6.18 (30 year original design life + 10 year extended life)

Seven additional PWRs complete their original 40-year design lives beginning 2023, nearly one plant each year

• • KHNP intends to seek license renewals but it will be a challenging task due to public concern for safety

• KHNP is planning a comprehensive research program on long-term materials aging degradation

Schedule of decommissioning Kori-1 (tentative) ICHN?

Central Research lnstJtute

• Submission of a KHNP decommissioning plan (2020.6.)

• Approval procedures including public hearings

• Final approval of the decommissioning plan (2022.6.)

Completion of the cutting and dismantling (2024.12.)

ICHN?

Central Research lnstJtute 2.' RESEARCH PLANNING USING MATERIALS FROM RETIRED KORI-1

Planning sub committees (2016.1.-4.) ICHN?

Central Research lnstJtute cla ssification Chair/secretary Members Neutron irradiation embrittlement BS Lee(KAERI)/

of RPV JS Yang(CRI)

SS Hwang(KAERI)/

Reactor internals JS Yang(CRI)

,Jr 5,.., 10 members each group sec of Alloy 600 weld nozzles and HP Kim(KAERI)/

from relevant organizations penetrations (CROM, BM N, DMW) HD Kim(CRI)

(KAERI, KPS, KEPCO-E&C, HP Kim(KAERI)/

~ Steam generators DHIC, KAIST, and a few HD Kim(CRI)

. .. universities)

' CH Jang(KAIST)/

Thermal embrittlement MW Kim(CRI) r,i r, DJ Kim(KAERI)/

FAC & buried pipes SK Park(CRI)

12 tasks proposed at the subcommittees ICHN?

Central Research lnstJtute

1. Irradiation Embrittlement of the Reactor Pressure Vessel Beltline Material

2. Irradiation Embrittlement of RPV Nozzle steel

3. Investigation of the IASCC mechanism

4. Material properties of aged RVI components

5. Residual stress and susceptibility to PWSCC of Alloy 600 welds and penetrations

6. Residual stress and susceptibility to PWSCC of Alloy 690 CROM penetration nozzles

7. Degradations of steam generator

8. Thermal embrittlement in cast and weld austenitic stainless steel

9. FEM integrity assessment and actual test of a bend portion in an aged real plant pipe

10. Thermal embrittlement of low alloy steel and its weld exposed to PZR temperature

11. FAC and erosion in CS piping in the secondary system and BOP

12. Degradations in buried pipes and inspection performance demonstration

Course of test materials and specimens ICHN?

Central Research lnstJtute

. - *---- __"j

• • • Specimens

~ KHNP Maintenance Building at KORI Station (Decontamination and

-+

Machining)

KAERI Non-Radioactive Materials (Piping, Structures)

• ICHN?

KHNP maintenance building in Kori station Central Research lnstJtute RCP maintenance facilities Separate rooms Old original Kori-1 RV head

Current Activities ICHN?

Central Research lnstJtute

• Input of research plan to KHNP decommissioning plan by 2017.12

• Planning of improving KHNP facilities for handling irradiated materials (2016.5.-2018.4.)

• KHNP maintenance building in Kori station

• CRI materials test building Improvement of KAERI Irradiated Materials Evaluation Facility

• - KAERI IM EF is overcrowded

• "1 An EPRI project

• Analytical assessment of residual stress in a few selected dissimilar metal nozzles in Kori-1

ICHN?

Central Research lnstJtute 3*. FUTURE PLAN

Project Schedule (draft) ICHN?

Central Research lnstJtute

• Improvement of KHNP and KAERI facilities

• Decontamination, movement, machining, and storage of Phase One test materials (2020"' 2024)

• Preliminary analytical studies

- irradiation dose, residual stress, and others Phase Two

• Test and evaluation (2024"' 2030)

Phase Three

• Remaining tasks after completion of Phase Two (2031"' t - additional irradiation test, international collaboration, as needed) and others

Others ICHN?

Central Research lnstJtute

• Planning of specific projects among Korean organizations

• Planning of international collaboration via EPRI

ICHN?

Central Research lnstJtute e

ELECTRIC POWER E:Pf21 RESEARCH INSTITUTE Ex-Plant Harvesting Workshop Session 1 EPRI Perspective Sherry Bernhoft Senior Program Manager NRC Ex-Plant Harvesting Workshop March 7 & 8, 2017

© 2017 Electric Power Research Institute, Inc. All rights reserved.

Nuclear Sector Research Areas 2

© 2017 Electric Power Research Institute, Inc. All rights reserved. ~*=

~~~1 1 I~

ELECTRIC POWER RESEARCH INSTITUTE

EPRl's Global Membership GLOBAL PARTICIPANTS GLOBAL BREADTH & DEPTH

> >15°/o of the world's commercial nuclear units Participants Encompass Most Nuclear Reactor Designs 3

© 2017 Electric Power Research Institute, Inc. All rights reserved. ~*=

~~~1 1 I~

ELECTRIC POWER RESEARCH INSTITUTE

EPRI Collaborative Model Suppliers, Vendors Basic Inputs Research and Development National Laboratories, Solutions Universities EPRI Nuclear Utilities 4

© 2017 Electric Power Research Institute, Inc. All rights reserved. ~*=

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ELECTRIC POWER RESEARCH INSTITUTE

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Experience With Harvesting

• Takes longer and cost more than anticipated

• Incomplete knowledge on the pedigree of the harvested materials

• Roles and responsibilities not well defined

• Contracting is complicated

• Logistics, i.e. transportation and storage are challenging You can spend a lot of money and not get any value 6

© 2017 Electric Power Research Institute, Inc. All rights reserved. ~*=

~~~1 1 I~

ELECTRIC POWER RESEARCH INSTITUTE

EPRl's Criteria for Harvesting

• Needs to be of value for our members

• Is prioritized as a need by our members

• Fills a knowledge gap that can not otherwise be filled

• Needs to start with a well developed project plan including:

- Funding

- Risk Management plan

- Exit Ramps

- Clear roles and responsibilities 7

© 2017 Electric Power Research Institute, Inc. All rights reserved. ~*=

~~~1 1 I~

ELECTRIC POWER RESEARCH INSTITUTE

Together ... Shaping the Future of Electricity 8

© 2017 Electric Power Research Institute, Inc. All rights reserved. ~*=

~~~1 1 I~

ELECTRIC POWER RESEARCH INSTITUTE

Plants in Decommissioning in Germany Uwe Jendrich, GRS March 7, 2017 Ex-Plant Materials Harvesting Workshop US NRC Headquarter, Rockville/MD, USA

Basic Requirements for Decommissioning Regulation of decommissioning in Germany r '

Granting of 1st Final shut down of the facility l decommissioning license

.)

! l

[ Power Operation ] Post operation transition period ] Decommissioning phase/

Direct dismantling or Safe enclosure

[ Operation I Decommissioning )

NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 2

The German Regulatory System

• Basic requirements

• For a nuclear facility the atomic energy act allows either

- to immediate dismantle or

- to dismantle after a safe enclosure Note: no entombment (near surface disposal} is allowed

• The operator of a nuclear facility is fully responsible for the decommissioning and dismantling of a nuclear facility.

....... For any requests on materials: Contact the operator!

• There are 4 operators of commercial nuclear reactors in Germany:

1. PreussenElektra 1 ) (KKS, KKI, KKU , KWG , KKG, KBR, KKE)

2. EnBW2) (KWO, GKN , KKP)

3. RWE (KMK, KRB, KWB)

4. Vattenvall (KKB, KKK)

1) until recently know n as E.ON Kernkraft

2) Energie Baden-Wi..irttemberg NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 3

Overview on Decommissioning Projects in Gemany

• Past and current decommissioning projects of Prototype or Commercial Reactors Name Abbrev. Reactor type Power Decom. Strategy MWe started Rheinsberg KKR WWER 70 II 1995 UC Compact Natrium Cooled KKN SNR 21 1993 UC Reactor Multipurpose Research R. MZFR PWR/0 2 0 57 1987 UC Obrigheim KWO PWR 357 2008 UC Neckarwestheim 1 GKN-1 PWR 840 2017 Jl UC lsar-1 KKl-1 BWR 912 201 7 UC Gundremmingen-A KRB-A BWR 250 1983 RCAKRB-11 Greifswald 1-5 KGR 1-5 WWER 440 1995 UC Lingen KWL BWR 268 1985 UC after SE

~*

UC: unconditional clearance RCA: radiation controlled area, new license SE: safe enclosure NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 4

Overview on Decommissioning Projects in Gemany

• Past and current decommissioning projects of Prototype or Commercial Reactors Name Abbrev. Reactor type Power Decom. Strategy MWe started Stade KKS PWR 672 2005 UC

- ~

Research Reactor Julich AVR HTR 15 1994 UC

-, ~

Thorium High- II THTR- HTR 308 1993 SE since 1997 Temperature-Reaktor 300 II II Wurgassen KWW BWR 670 1997 UC Mulheim-Karlich KMK PWR 1302 2004 UC Hot-Steam Reactor HOR HOR 25 1983 UC since 1998 Grosswelzheim Niederaichbach KKN DRR/D20 106 1975 UC since 1994 Test-Reactor Kahl VAK BWR 16 1988 UC since 2010 UC: unconditional clearance SE: safe enclosure NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 5

NPPs Preparing for Decommissioning in Gemany

• Shut down Commercial Reactors

• that have no decommissioning license granted yet Name Abbrev. Reactor type Power MWe Date of application Philippsburg-1 KKP-1 BWR 926 2013 / 2014 Grafenrheinfeld KKG PWR 1345 2014 Biblis-A KWB-A PWR 1225 2012 Biblis-8 KWB-8 PWR 1300 2012 Unterweser KKU BWR 1410 2012 / 2013 Brunsbuttel KKB BWR 806 2012 / 2014 Krummel KKK BWR 1402 2015 RPV design fluence after 40 years

< 5

• 1018 n/cm 2 NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 6

NPPs Preparing for Decommissioning in Gemany

• Commercial Reactors in operation Name Abbrev. Reactor type Power MWe Anticipated date of final shutdown Gundremmingen-B KRB-11-B BWR 1344 31.12.2017 Philippsburg-2 KKP-2 PWR 1468 31.12.2019 Gundremmingen-C KRB-11-C BWR 1344 31 .12.2021 Grohnde KWG PWR 1430 31.12.2021 Brokdorf KBR PWR 1480 31.12.2021 Emsland KKE PWR 1406 31.12.2022 lsar-2 KKl-2 PWR 1485 31.12.2022 Neckarwestheim-2 GKN-2 PWR 1400 31 .12.2022 NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 7

Overview on Decommissioning Projects in Gemany

• Past and current decommissioning projects of Research Reactors

• Total: 35

• Removed: 29 Nuclear Ship Otto Hahn

• Final Shut down / under dismantling: 4 during operation

• Safe

Enclosure:

2 ..)

• Variety of types of Research Reactors

- Argonaut type

- Critical assembly Rad. transport of

- Educational reactors dismantled

- Liquid homogenous reactor

© Babcock

- Propulsion reactor Noell GmbH

- Pool reactor (incl. TRIGA type)

- Heavy Water reactor (incl. DIDO type) ~

NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 8

Overview on Decommissioning Projects in Gemany

• Past and current decommissioning projects of Prototype or Commercial Reactors

• Total: 21

• Removed: 3

• Final shut down / under dismantling: 17

• Safe enclosure: 1

• Reactor types:

- PWR

- BWR

- Fast Breeder

- High Temperature Gas Cooled

- Heavy Water Gas Cooled

© Forschungszentrum Karlsruhe NRC Harvesting Workshop, Rockville, March 2017, Decommissioning in Germany 9

Thank you for your attention!

Role of GRS in Decommissioning and LTO Uwe Jendrich, GRS March 7, 2017 Ex-Plant Materials Harvesting Workshop US NRC Headquarter, Rockville/MD, USA

Introduction GRS with its staff of about 430 is

• Main Technical Support Organization {TSO) in nuclear safety for the German federal government

• BMUB (Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety),

• BMBF (Federal Ministry of Education and Research) ,

• BMWi (Federal Ministry of Economic Affairs and Energy) and

• AA (Federal Ministry of Foreign Affairs) and r

GRS participates in international activities of IAEA, OECD , EU (DG Energy, DG RTD, DG DevCo), ...

.J NRC Harvesting Workshop, Rockville, March 2017, Role of GRS 2

Services of GRS in Decommissioning of Nuclear Installations

• Advice and assessment

- concrete technical and legal issues

- strategy and planning of decommissioning projects

• Project management, progress monitoring and documentation of decommissioning projects

• Support in the drafting of regulations

- taking into account national specifics

• Know-how transfer

- individually tailored seminars and coaching

- elaborations on specific issues NRC Harvesting Workshop, Rockville, March 2017, Role of GRS 3

GRS Customers and References in Decommissioning

• BMUB:

• Various projects (since 1998):

- technical and organizational requirements

~

~

IFederal Ministry for the Environment. Nature Conservation, Building and Nuclear Safety

- analysis of technical aspects,

- assessment of application and licensing documents

• BMBF:

• Project management (since 2013):

• 'Decommissioning, dismantling and waste management projects relating to nuclear test facilities'

• Federal Ministry of Education and Resea rch

• IAEA:

• Decommissioning and Remediation of Damaged Nuclear Facilities (DDAROD! (~inc.e 20R1'5k) M

• ecomm1ss1oning 1s anagemen t (D'R'M ) ( .

1 1 a since 2012) t ~t~ a~ micEnergyAgency

• Constraints to implementing decommissioning and environmental remediation programs (CIDER) (2013-2015)

• EBRD:

• Assistance to Bulgaria Nuclear Regulatory Agency (BNRA) 0 ~~~?.?.!~'"~~.~

- decommissioning of Kozloduy NPP units 1-4 (since 2008)

NRC Harvesting Workshop, Rockville, March 2017, Role of GRS 4

GRS Activities in Ageing Management and Long Term Operation

• BMUB:

• Evaluation of nat. & internat. operating experience

• Nat. & internat. Regulations and Guidelines on AM

• EU:

• Several INSC projects on

- Regulatory aspects of LTO, Specific technical aspects

• JRC Database EMAR (R&D results on ageing of materials)

• IAEA:

• International IGALL

• OECD/NEA:

• Database projects CODAP (components), CADAK (cable)

• WG IAGE, NUGENIA

• ANVS {NL):

• Assessment of licensing documents for LTO of Borssele NRC Harvesting Workshop, Rockville, March 2017, Role of GRS 5

Successful Harvesting Opportunities Require Clear Scope and Schedule SOWTIONS

• Plant View

- There is no financial incentive to support harvesting

- The research agency must be willing to pay for the costs to surgically remove components

- The research agency must provide a clear scope and schedule at the beginning of the active decommissioning period (DEGON)

- Delays in federal approvals mean lost opportunities

• Flexibility in the removal schedule is key

- Schedule changes result in components becoming available on short notice

- The research agency must have the funds set aside from one year to the next or miss out on harvesting opportunities

• Reactor vessel surveillance coupons

- Best removed when work begins on cutting up reactor internals

• Reactor vessel weld specimens

- Only applicable for large vessels

- Smaller vessels like SONGS 1 and Ft. Calhoun shipped intact to disposal facility Reactor Vessel Cutting Head

• Spent fuel rack coupons

- Best removed before or after spent fuel transferred to the ISFSI

- Cannot not interfere with fuel characterization or fuel transfer

• Spent fuel rack

- Removed after pool completely empty including GTCC waste

• Cabling from Harsh Environments

- Some can be harvested at any time

- High rad environment requires timing of harvest to mesh with source term removal schedule in each area

• Switchgear, Bus Ducts, Control Panels, Breakers

- Some can be harvested at any time

- Specify cable lengths needed

- High rad environment requires timing of harvest to mesh with source term removal in area

• Concrete Cores

- Cores best taken consistent with site characterization needs for License Termination Plan

• Plants that are entering DEGON in the next 2 years need to know now if the NRG or DOE are interested in harvesting materials at these sites:

- San Onofre 2 & 3 (January 2018)

- Vermont Yankee (2019)

• Plants potentially entering DEGON in the next 2 years should be contacted now:

- Kewaunee

- Crystal River

- Ft. Calhoun

Two 4-Loop Westinghouse PWR 3250MWt (1085 MWe)

• Construction permit December 1968

• Operating Licenses April 1973 (Unitl)

November 1973 (Unit 2)

• Shutdown (unexpected) February 1998

• Permanently Defueled March 1998

• License transferred to EnergySolutions September 2010

• All Spent Fuel in Dry Storage January 2015

• Expected Completion April 2019

• Plant Decommissioning is on schedule and on budget

• Currently demolishing Auxiliary Building

• Containment buildings ready for open air demolition in May 2017

• Complete building demolition by early 2018

• Show demolition video

• Zion Harvesting Program Experience

- Extensive harvesting conducted from 2011 to 2016

- Senior management supportive of program

- Reassignment of plant personnel to support harvesting program

- Demolition organization generally not supportive but tolerated surgical removal of components

- Demolition did not stop for harvesting

- Harvested items only have scrap value

• Reactor Vessel Material Surveillance Capsules

- Capsules were removed during reactor internals cutting and stored on the reactor cavity floor

- Capsule designators were very hard to read

- Only one capsule had research value (X)

- Zion attempt to identify capsule markings failed

- Capsule X ended up being disposed of by the time contract issues were worked out with Westinghouse to retrieve capsule

- Result: All capsules disposed as radwaste

• Reactor vessel beltline weld

- Reactor vessels cannot be rotated in cutting machine

- Vessel cutting plan looked like it would work for vessel specimen harvesting plan

- Unit 2 beltline weld was too close to cut line so torch could have affected metal properties

- Unit 1 cut line was far enough away from weld to obtain a successful specimen

Cabling harvested from Unit 1 (11 Tasks)

1. East Valve House near Steam Tunnel

2. Instrument cabling inside missile barrier

3. Loop Isolation Valve cabling

4. Accumulator discharge valve MOV cabling

5. Instrument Rack cabling in containment

6. Air-operated valve cabling

7. Electrical penetrations assembly and cables

8. Electrical penetration cabling in Aux Bldg.

9. Pressurizer heater cables

10. Cables in steam tunnel

11. Submerged cable in Turbine Bldg. to Crib House

Experience

• Only some of the cables harvested.

- Changing priorities

- Demolition crew not cognizant of research needs

• Cable lengths of 30 feet were difficult to obtain in some instances. Could only get 10 feet.

• Cable type, age and environment were generally available from plant records

• Bus bar section ( 8 feet) removed along with any remaining support hardware and isolators

• 480-volt distribution center removed

• 480-volt motor control center removed

• Control Room Control Panel removed

• Instrument Rack removed in Aux. Bldg ..

• 4KV Switchgear, Relays and Breakers removed

• ISO-Phase Bus Duct removed

• Planned to take core bores in containment and under spent fuel pool liner

• Task cancelled by ORNL

• Spent Fuel Pool Testing completed for boron degradation

• Remaining coupons retrieved

• Storage rack piece cut and shipped

• Good onsite coordination with EPRI and NRC

• Most work completed during weekend

• Changing scope leads to frustration on both sides

• Delays in paper work for each task authorization lead to lost opportunities

• Plan early

• Obtain senior plant management support

• Have someone on site when harvesting takes place

• Be able to deal with changing plant support personnel

B E.NPERGY Nuclear Energy US Nuclear Science User Facilities (NSUF)

Overview J. Rory Kennedy, Ph.D.

Director, NSUF c_@il.sur Idaho National Laboratory J Nuclear Science User Facilities March, 2017

8 E*trERGY Nuclear Energy NSUF General c~1a.sur

, I Nuclear Science V User Facilities

• Established 2007 as DOE Office of Nuclear Energy first and only user facility

• Idaho National Laboratory is lead institution

• Irradiation effects in nuclear fuels and materials

• Provide access to capabilities and expertise at no cost to user

• Support design, fabrication, transport, irradiation, PIE, disposition

• Link intellectual capital with nuclear research infrastructure to fulfill mission of DOE-NE

• Generally select projects through open competitive proposal processes

• Consolidated Innovative Nuclear Research {CINR FOA, 1 call/vear

• Irradiation + PIE ($1.0M - $4.0M, up to 7 years)

• PIE only {-$SOOK, up to 3 years)

• Irradiation only {$SOOK - $3.SM)

• Beamlines at other user facilities - *-

• Rapid Turnaround Experiments {RTE, 3 calls/year, limited $$, executed within 9 months)

• Proposals welcome from University, National Laboratory, Industry, Small Business, lnt'I researchers 22

8 E*trERGY NSUF - A consortium c@i'l.sur Nuclear Energy A group formed to undertake an enterprise beyond the resources of any one member J Nuclear Science User Facilities 2007 2008 2009 2010 2011 2012 2013 2016,

~ OA..K

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s.,;....::.;;- PURDUE ILLIHCS 11'1!.~ITUN L II I . Araonne D ,._~J~Lc 1

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additional capabilities Unl 1..-an.lly

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• Partner Facilities to date outside INL D 8 Universities+ 3 Universities in CAES recognized (3 currently expressed interest) early UNLV I D 4 National Laboratories (4 expressed interest)

D Partner facilities D 1 Industrial program wri_~

established in 2008 3

8 E*trERGY NSUF General c~i1.sur Nuclear Energy Capabilities J Nuclear Science User Facilities

• Neutron Irradiations

• ATR (loop, rabbit), ATRC, HFIR (rabbit), MITR (loop), PULSTAR, NRAD (Future: BR2 - SCK-CEN Belgium), Halden - Norway ?)

• Ion lrrad iations

• Tandem Accelerator Ion Beam (U. Wisc), Michigan Ion Beam Lab (U. Mich), IVEM (ANL)

(Future: TAMU, SNL, LANL)

• Hot Cells

• INL(HFEF, FCF, AL, IASCC), ORNL (IFEL, IMET, REDC), PNNL (RPL), U. Mich (IMC),

Westinghouse (MCOE)

• High radiation level measurements/instrumentation

• Neutron radiography, elemental & isotopic analyses, gas sampling and analyses, profilometry, gamma scanning, mechanical testing, electron and optical microscopy, thermal analyses, eddy current, IASCC, EPMA, AES, XPS, SIMS, focused ion beam (FIB)

• Low radiation level measurements/instrumentation

• SEM, TEM, APT, FIB, hardness, micro- & nano-indentation, tensile, thermal analyses, XRD, XPS, AES, SIMS, NMR, PAS

• Beamlines

• X-ray (ANL APS: MRCAT, IIT; BNL NSLS-11: XPD, NST Dept)

• Neutron, positron (PULSTAR, NCSU)

• High Performance Computing

• FALCON machine

• Moose-Bison-Marmot

• Visit nsuf.inl.gov under Research Capabilities tab for details at individual facilities 4

8 E*trERGY NSUF and GAIN c@i'l.sur Nuclear Energy J Nuclear Science User Facilities Gateway for Accelerated Innovation in Nuclear (GAIN)

R&D TEST BED Rapid and cost-effective advancement of scientific

~ 1000 underpinning and retirement

~ of technical and licensing risk

~ for innovative technologies.

~

I (l)

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.§ 100 DEMO PLATFORM Cl:> Reduce the commercialization cost

~ and associated risk by minimizing

..s one-time costs. Reduce the cost uncertainty for commercial units.

1 2 3 4 5 6 7 8 9 Proof-of-Concept Proof-of-Performance Proof-of-Operations Technology Readiness Levels (TRL)

\-AIN 5

8 E*trERGY Nuclear Energy High Impact Nuclear R&D c~1a.sur Nuclear ~~i.ence V

I /

User Fac1llt1es Project portfolio spans a variety of research objectives that are ultimately focused on both near and long-term technology development goals

• Understanding atomic level phenomena in fuels that affect thermal transport, elemental migration/diffusion, interface interaction, etc. as complex microstructures develop under irradiation

• ceramic, metallic, TRISO, ATF

• Understanding fundamental defect evolution in irradiated structural materials across multiple length scales as they affect mechanical properties.

• RPV, austenitic, F/M, Zr alloys, ATF

• Development of innovative radiation resistant materials for advanced reactor systems

• Development of radiation resistant sensors for collecting high fidelity on-line irradiation test data

• Development of materials from advanced manufacturing techniques

• Providing fundamental actinide nuclear data that can help inform advanced reactor and fuel cycle modeling and Simulation Campaign. J. Cole contributed to content of 6 C""lir4 o

8 E*trERGY Infrastructure c_~i'IJUi=

Nuclear Energy Management Program I / Nuclear Science V User Facilities

• NSUF created a searchable and interactive database of all pertinent infrastructure supported by, or related to, the DOE Office of Nuclear Energy {DOE-NE).

• Database known as the Nuclear Energy Infrastructure Database {NEID) and is located at nsuf-infrastructure.inl.gov

{public launch Nov 2015)

• Used for analyses to identify needs, redundancies, efficiencies, distributions, etc., to best understand the utility of DOE-NE's available infrastructure, inform the content of infrastructure calls, and provide information to NSUF users.

• Infrastructure information collected can be combined with information on R&D needs as part of infrastructure gap analysis 7

8 E*trERGY Nuclear Fuels and c~i'IJUi=

Nuclear Energy Mat er1a . IS L.b I rary {NFML) ,V/ Nuclear Science User Facilities

• Provides irradiated samples for users to access for experimentation through one of the competitively reviewed proposal processes.

• Critical to reducing costs and taking advantage of new ideas and future analysis techniques and equipment.

• The library includes over 3500 specimens as part of the NSUF awarded research. 6K - 7K additional specimens by year end.

INL

• Most materials in NFML neutron irradiated with small number Legacy materials ion irradiated.

• SAM irradiation series to stock library moving forward

• Effort to consolidate materials into easily _Al~,...... ~ '/!fl, accessible locations to reduce costs of Volunteered retrieval. materials from outside

• Web-based searchable database through the INL nsuf.inl.~ov (public l~unch ~ept 14,_2016). ~ A..:,~ .till

• Interest 1n collaboration on 1nternat1onal efforts. *

• Materials Include:

Supporting

• Steels

• Pure materials documentation related to

• Other alloys

• Actinides samples

• Ceramics

• Fission products 8

8 E*trERGY Databases Design c~i1.SUi=

Nuclear Energy

, I Nuclear Science V User Facilities r

SME DATABASE ~

Pl Name PROJECT DATABASE Subject Matter

,... PROJECT NAME ~

INSTITUTION NEID DATABASE Project ID Start Date Project Type Proposal End Date Material Type INSTITUTION CINR# Pl Name Research Area FACILITY RTE# Tech Lead INSTITUTIO REACTOR NSUF Call Facility Tech Lead FACILITY

'l._ Award Date Collaborators Related Documents REACTOR POSITION

-- ~

I ,4pROJECT NAME SAMPLE LIBRARY REACTOR.....-

REACTOR P O S I T I O N - - - - - - - -

Sample ID Code #of Samples PLANNED AS RUN Capsule Samples Remaining Temperature Temperature Packet Specimen Availability Dose (DPA) Actual Dose (DPA)

Material Code Availability Date Fluence [x10201 Fluence [x10201 Material Name Certification Flux [x1014 1 Flux [x10141 Material Description Certification Code KGT # Storage FACILITY""' Environment Environment Specimen Type Notes Dimensions 9

Nuclear Science User Facilities 10

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'Kewaunee Power Station Dominion-2-Loop Westinghouse PWR (590MWe)

• Construction permit August 1968

• Operating license December 1973

• Initial operating license expiration December 2013

• Renewed Operating License Issued February 2011

• Shutdown decision (unexpected) October 2012

• Permanently Shutdown May 2013

• ERO offsite response eliminated November 2014

• All nuclear fuel in dry storage July 2017 4

'Dominion-Kewaunee Perspective

• Top priority for decommissioning plants is preserving and good stewardship of the decommissioning trust fund

• Highest fund drain is staffing

• Station electrical use also expensive

• Initial decommissioning actions focus on safety and cost reduction.

o Use initial required large staff to prepare plant for long term dormancy and decommissioning o Abandon or downsize equipment to reduce ongoing costs.

o Then reduce staff commensurate with reduction in risk 5

' Dominion-Kewaunee Power Station Kewaunee Decommissioning Staffing Reductions 700 I

600

. 630 500 400 300 200 I

" '"' 147 100 I

• * * * ' ' ' * ~- ~ -

I so 0

4/1/2012 8/14/2013 12/27/2014 5/10/2016 9/22/2017 2/4/2019 6

'Dominion-Kewaunee Perspective

• Timing is everything!

• Perhaps best discussed via an example:

o Reactor vessel surveillance capsules o Two remain in the vessel o Logistical considerations:

7

'Dominion-Kewaunee Perspective

• Circulating Water pumps were high energy consumers

• Therefore we wanted to retire them as soon as possible

• Circulating Water pumps were high capacity, low head, very good at dilution for meeting ODCM requirement for radiological discharges

• Without CW pumps, much smaller capacity service water pumps will be used in the future for dilution ..

8

'Dominion-Kewaunee Perspective

• Therefore - prior to retiring CW pumps, we processed as much radioactive water as possible

• This included draining the RCS

• RCS at Kewaunee has no loop isolation valves

• RCS today is drained to the bottom of the cold legs, about 7 feet below the RV flange

• RV internals are installed, RV head is on the vessel, flux thimbles are installed in the vessel 9

'Dominion-Kewaunee Perspective

• So if we wanted to remove surveillance capsules today we would face unique challenges that would not have been present at initial plant shutdown:

o Shielding - need to refill the RV (and cavity?)

o Lifting the RV head and internals - polar crane is in place, but maintenance has been discontinued o Qualified staff - Crane operators, RP technicians, maintenance, operators o Rad monitors (many have been abandoned) o Ventilation and atmosphere control, lighting 10

'Dominion-Kewaunee Perspective

• What generic issues could be resolved or simplified by harvesting and additional testing?

• GALL aging management program examples o Electrical cables and connections - test power cables taken from adverse localized environments o SG divider plates; autogenous welds o Buried piping; o Inaccessible power cables (buried; underground) 11

'Dominion-Additional Consideratiions

• Who will pay?

• Why is the material needed?

o "What's in it for me?"

o Are we solving an industry problem? (example -

SFP neutron absorbing material GL 2016-01) o Objectively needing more information to determine if we have a new problem?

o What's the plan? What is done with the information gathered? Is there a driver to review impact on existing programs?

12

'Dominion-Additional Consideratiions

• Need to think ahead, plan ahead

• Some harvesting is very plant condition specific, others maybe not

• What plants will be entering decommissioning in the future?

• Have you reached out to them?

• Scope, Schedule, Budget 13

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• There are various safety-related concrete structures in nuclear power plants (primary containment, containment internal structures, secondary containment/reactor ..**I ,

buildings, other structures) .~

• I

' POLAR CRANE

• Near the pressure vessel:

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- Biological shield concrete is placed ..**. I, around the pressure vessel to reduce radiation to allowable levels for humans Some plants use concrete as the pressure vessel support structure

@Westinghouse 2

• The radiation types important for the concrete biological shield and reactor vessel support structure are gamma rays and neutrons

• Effects of radiation on concrete for long-term nuclear power plant operation has been of interest to various organizations

• Radiation effects on the properties of concrete depend on the intensity of the radiation field, period of exposure, and concrete composition

• Radiation affects concrete properties such as compressive and tensile strength, modulus of elasticity, creep, volumetric variation

@Westinghouse 3

• Effects of radiation on concrete has been addressed in various reports; the work by Hilsdorf et al. (1978) is cited frequently:

- H.K. Hilsdorf, J. Kropp, and H.J. Koch, "The Effects of Nuclear Radiation on the Mechanical Properties of Concrete," American Concrete Institute Special Publication SP-55, 223-251 (1978)

• The change in compressive strength under neutron radiation exposure was selected for evaluation for this presentation, from the work by Hilsdorf et al. (1978)

• Purpose is to find a threshold radiation level where significant strength reduction will occur

@Westinghouse 4

Westinghouse Non-Proprietary Class 3 © 2017 Westinghouse Electric Company LLC. All Rights Reserved.

Compressive Strength of Concrete Exposed to Neutron Radiation - Hilsdorf et al (1978)

Slow Neutrons: F~st Neutrons=

Threshold

• Alexander ( 1 ) o Etleuch ( 14) fcu/fcuo

• Price (28) , Houben ( 22) 140 *280deg.C , Stoces (3 1) 1,4 J--- --.-----.---r----.-- -~-h-----.-- --,.....---...-------.

No Information:

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0 : *1OOdeg.C or higher 0,2 0 : Small size 0 : Bending strength FLUENCE OF NEUTRON 10 18 2 5 10 1* 2 5 10 20 2 S ,o... 2 RADIATION, n/cm:,

NUREG/CR-7171, ORNL/TM-2013/263, "A Review of the Effects of Radiation on Microstructure and Properties of Concretes Used in Nuclear Power Plants,"

November 2013

@Westinghouse 5

• Data providing information on the effect of radiation on concrete properties are limited and an expansion of the irradiated concrete database is needed

• New data should be representative of conditions associated with nuclear power plants, indicating the need to obtain and test concrete samples from shutdown nuclear power plants

• Furthermore, reliable fluence data from radiation transport calculations is also needed

• Reliable fluence data is achieved by dosimetry measurements

- Measurement data validates the radiation transport calculational methodology and determines the uncertainty in fluence calculations

@Westinghouse 6

Westinghouse Non-Proprietary Class 3 © 2017 Westinghouse Electric Company LLC. All Rights ReseNed.

Neutron Dosimetry for Reactor Pressure Vessel Fluence Calculations

• Reactor pressure vessel fluence calculational methodology validation and uncertainty determination is obtained from:

- Surveillance capsules

- Pressure vessel clad sampling

- Ex-vessel neutron dosimetry (EVND)

@Westinghouse 7

Westinghouse Non-Proprietary Class 3 © 2017 Westinghouse Electric Company LLC. All Rights Reserved.

In-Vessel and Ex-Vessel Neutron Dosimetry Plan View go* lt.\.ET NOZZl£ IRRAOIAOON CAPSUU'. In-vessel surveillance capsule CONCRETC LINER 335' Ex-vessel neutron FUCL ASSEMBUCS dosimetry SUPPORT

@Westinghouse 8

Westinghouse Non-Proprietary Class 3 © 2017 Westinghouse Electric Company LLC. All Rights Reserved.

Ex-Vessel Neutron Dosimetry Section View Support chain

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Support bar 1

Dosimetry capsules 1

@Westinghouse 9

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Westinghouse has successfully provided ex-vessel neutron dosimetry programs to nuclear plants since 1974, including:

• Almaraz

• Asco

• Beaver Valley

• Braidwood

• Brunswick

• Byron

• Callaway

• Catawba

• Comanche Peak* Connecticut Yankee

• Diablo Canyon

• Farley

• H. B. Robinson

• Kori

• Krsko

• McGuire

• Mihama

• Palisades

• Point Beach

• Ringhals

• South Texas

• St. Lucie

• Turkey Point

• Ulchin

• Vogtle

• V. C. Summer* Vandellos

• Wolf Creek

• Yonggwang (Hanbit)

• Zion

@Westinghouse 11

• Similar to the pressure vessel fluence analysis, fluence calculations for concrete should utilize measurements for calculational methodology validation and fluence uncertainty determination

• Since EVND is installed at the cavity in front of concrete, it provides dosimetry measurement data useful for concrete fluence calculations

@Westinghouse 12

• Data on irradiation effects on concrete properties are limited; existing data has deficiencies and may be non-representative of the conditions associated with nuclear power plants

• New data is needed to evaluate irradiated concrete degradation; this is best achieved through testing concrete samples obtained from nuclear power plants

• Materials testing data are coupled with fluence data in the evaluation of irradiated concrete degradation

@Westinghouse 13

• Neutron dosimetry measurements should be used in qualifying the fluence calculational methodology and determining the uncertainty in fluence calculations for concrete

- EVND is appropriate to use for concrete fluence calculations

- A potential nuclear power plant to harvest concrete that also has neutron dosimetry measurement data will be discussed in Session 3

@Westinghouse 14

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• Harvesting of concrete from nuclear power plants is needed to understand the mechanisms that cause radiation damage

• Materials testing data is used with exposure data to evaluate irradiated concrete degradation

• It is important to validate the fluence calculations using measurement data

@Westinghouse 2

• Westinghouse installed and analyzed neutron dosimetry for one fuel cycle at Mihama Unit 1 at the reactor cavity in front of concrete and away-from-reactor side (back of) concrete

• The dosimetry measurements were used to validate the radiation transport calculations and estimate the water content of concrete at Mihama Unit 1

• Other dosimetry measurement data (e.g. from surveillance capsules) could be used as supplemental data to validate the radiation transport calculations

@Westinghouse 3

• Mihama Unit 1 was shutdown in 2015

- Kansai submitted its decommissioning plan to the Nuclear Regulation Authority (NRA)

• Westinghouse is in communication with Kansai to investigate the possibility of extracting cores from the concrete biological shield of Mihama Unit 1

@Westinghouse 4

• Westinghouse allocated internal funding to investigate the possibility of harvesting concrete from Mihama Unit 1 and explore feedback from the industry, with the expectation of external funding to complete the project

• We would like your feedback on this possible opportunity and determine if there is an industry interest for this work

@Westinghouse 5