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NRC-2018-000831 - Resp 11- Interim, Agency Records Subject to Request Enclosed
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From : Ramuhalli Pradeep To : Hiser Matthew Subject : [External_Sender) You have files ready for pickup Date: Friday, February 3, 2017 11 :21: 15 AM

Hello, Ramuhalli, Pradeep (Pradeep.Ramuhalli @pnnl.gov) has sent you the fo llowing I file(s:)

Subject:

Resending slides Comments: Hopefully this makes it through.

Pradeep The following files have been uploaded to the MassTransit Web File Transfer Services. You can download them by going to:

(b )( 4) an se ectrng t 1e .

NOTE: This link and contained passkey are only good for 14 days.

Harvesting workshop slides draft.pptx (1.47M bytes)

This message was automatically generated from the PNNL FX Web File Transfer Service. If you have questions abo ut its validity, please contact the sender listed above.

Note to requester: The From: Chen, Yiren attachments are immediately Sent: Tue, 27 Mar 2018 22:57:05 +0000 following this email. (KD)

To: Purtscher, Patrick Cc: Natesan, Krishnamurti

Subject:

[External_Sender] RE: Re: follow up on harvesting Attachments: NUREG-5748_shippingport.pdf, SEND TO ANL IML materials_update_yc.xlsx Hi Pat, Sorry for taking so long to update the size information. For the six unirradiated materials on oiur list, I have located five of them and their information is in the attachment. For the one I couldn't locate, a low-alloy steel cut from Shipping Port (A212 Gr-8), I think I made a mistake earlier when I provided you the list. Omesh told me about this material, and I saw a piece marked "A212" in our storage area and assumed that must be it. However, today, after I look into it more closely, I realize it must NOT be the heat. In fact, the material Omesh mentioned should be slightly radioactive and must not be in our common storage area. According to Omesh's report (NUREG/CR-5748 attached), the Shipping Port material was harvested in 1982, and had an estimated dose of 0.0009 dpa. Although this is a very low level of damage, the material should be slightly radioactive because it was activated by neutrons. I believe the test samples used in the program were machined in our "hot shop". So, either all material had been used up, or some of the remaining material had been disposed of when the "hot shop" was closed many years ago. For our radioactive materials, we have done several inventories since 2005, and I am pretty sure that we don't have such a material in our current rad-material inventory. Sorry about this error, which was a misidentification. Could you please remove the line of A212B from our list?

Thanks, Viren From: Purtscher, Patrick <Patrick.Purtscher@nrc.gov>

Sent: Tuesday, March 27, 2018 5:48 AM To: Chen, Viren <yiren_chen@anl.gov>

Subject:

RE: Re: follow up on harvesting Thank you.

From: Chen, Viren chen@anl.gov

Sent: Monday, March 26, 2018 8:20 PM To: Purtscher, Patrick <Patrick.1Purtscher@nrc.gov>

Cc: Natesan, Krishnamurti <natesan@anl.gov>

Subject:

[External_Sender] Re: follow up on harvesting Hi Pat, I will check the storage and get back to you tomorrow.

Thanks, Viren From: Purtscher, Patrick <Patrick.Purtscher@nrc.gov>

Sent: Monday, M arch 26, 2018 1:47:05 PM

To: Chen, Viren Cc: Natesan, Krishnamurti

Subject:

RE: RE: Re: follow up on harvesting One more thing, could you add a descrption of the approximate size of unirradiated material. It can go under comments.

Thanks, Pat From: Chen, Viren chen@anl.gov

Sent: Friday, March 09, 2018 2:25 PM To: Purtscher, Patrick <Patrick.Purtscher@nrc.gov>

Cc: Natesan, Krishnamurti <natesan@anl.gov>

Subject:

[External_Sender) RE: Re: follow up on harvesting Hi Pat, Please see our updates on the tables as attached. The new inputs are in red. Please let me know if you have any question or need clarification.

Thanks, Viren From: Purtscher, Patrick [1]

Sent: Friday, March 09, 2018 7:30 AM To: Chen, Viren <yiren chen@anl.gov>

Subject:

FW: Re: follow up on harvesting Good morning, We continue to review the information you provided regarding ex-plant/irradiated reactor materials.

We have rearranged the information in the spreadsheet that you prepared to better identify the materials. The# of samples for each line item is not always obvious when arranged this way. Could you verify the number of specimens for each condition?

Thank you.

Pat Purtscher Materials engineer RES/DE/CMB

Note to requester: This NUREG is also publicly available through DOE's Office of Scientific and Technical Information at https://www.osti.gov/biblio/601 92 12-radiation- NUREG/CR--5748 embrittlement-neutron-shield-tank-from-shippingport-reactor TI92 002043 Radiation Embrittlement of the Neutron Shield Tank from the Shippingport Reactor Manuscript Completed: July 1991 Date Published: O ctober 1991 DISCLAIMER Prepared by

0. K. C hopra, W . J. Shack, S. T. Rosinski* This re port was prepared as an account or work sponsored by an agency or the United Stal~

Gove rnment . Neither the United S tates Government nor any agency there-or, nor any of their employees. makes any wa rrant y, express or implied, or assume1 any legal liability or rcspon, i-bilily for the accuracy, completeness, or usefulness 0 1 ..ny informat ion, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Refer-Argonne National I..aboratorv ence herein 10 a ny specific commercial product, process, or service by trade name, trademark, 9700 South Cass Ave nue

manufacturer. or otbc,....isc does not necessarily constitute or imply its endorsement, recom*

Argonne, IL 60439 mcnda tion, or favoring by the U nited States Government or any agency thercor. The views and opinions of authors expressed he,dn do not necessarily state or rcncct those of the United States Government 'll a ny agency t hereof.

i'repared for Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN A2256

Sandia National Laboratories Albuquerque , NM 87185 MASTER

Radiation Embrlttlement of the Neutron Shield Tank from the Shippingport Reactor

0. K. Chopra. W. J. Shack. and S. T. Rosinski Abstract Toe irradiation embnttlement of neutron shield tank (NS1' material (A212 Grade B steel) from the Shippingport reactor has been characterized. Irradiation increases the Charpy transition temperature (CT11 by 23--28°C (41-50°F) and decreases the upper-shelf energy. The shift in CIT ls not as severe as that observed in high-flux 1Sotope reactor (HFIR] surveillance specimens. However, the actual value of the CIT is higher than that for the HFIR data. The increase in yield stress ls 51 MPa (7.4 ksll, which ts comparable to HFIR data. The Nsr material ts weaker In the transverse onentatlon than in the longitudi-nal orientation. Some effects of pos1Uon across the thickness of the wall are also observed:

the CTI shJft is slightly greater for specimens from the inner region of the wall. Annealing studies indicate complete recovery from embrittlement after l h at 400°C (752°F).

Although the weld metal ts s1gn1flcantly tougher than th~ base metal, the shifts In CTr are comparable. The shifts in CIT for the Shippingport NST are consistent with the test and Army reactor data for irradiaUons at <232°C (<450°F) and show ve:ry good agreement with the results for HFIR A212-B steel irradiated in the Oak Ridge Research Reactor (ORR).

The effects of irradiaUon temperature. fluence rate, and neutron flux spectrum are dis-cussed. The results indicate that fluence rate has no effect on radiation embrlttlement at rates as low as 2 x 108 n/cm2-s and at the low operating temperatures of the Shippingport NST. Le.. 55°C {I 30°F). This suggests that the accelerated embrlttlement of HFIR surveil-lance samples ts most likely due to the relatively higher proportion of thermal neutrons In the HFIR spectrum compared to that for the test reactors.

ill

Contents E:xecuUV"e Summary..... -- ........... ..

Introduction ....................... . 3 2 r.*1a1 t:ria.l Charactert-..aUon... 3 3 lrradiaU,111 Embrittlem en t . 6 3 .1 Base Metal................ ........ 8 J . l . l Charpy-Impact Energy.................................................................................................. 8 J . 1.2 1\*.1-;ile Prupcnic~ ...... 15 J. l.3 Re('overy p.. , 11 1.* aling ..... .... .. l7 3 .2 W,* 1.:1 M eta l ................... ..................................................... ... 20 4 Dtsc.:,1sslon ...................................... . ...................................................................................... 24 T i /{ Survt>illance D : ;1 .... ...................................................................................... 24 4 .2 1.(lW-TemJ)l:mture lrr,1diaUon. 26 4.3 Effect of Fl uence Rlltt* .. 30 4 .-+ Spt'ct ral Effects ...... 31 Conclusions........... 32 Ackn(w:1edgr.'rnt s .. 33 Referencf's. 33 List of Figures

1. Schematic reprt'sentatlon rif the NST <.11ppor1 for the Shlppin¢pm1 reactor pressure vessel __ .............. .................... ..... ........................... .............. . 4

2. Layout C1f sample location In the Shippingport NST mner and otiln walls..... 4

3. Mtcrographs of the Shlppin~port NST :nalerial .dong transverse und rollln~

sections............................................................. ................................................ .......................... 6 V

4. Hardness proflle across the thickness of the ShlppJngport NST outer wall.................. 6

5. Hardness profiles for Locatio*~~ 3 and 9 of the Shippingport. NSf Inner wall.............. 7

6. Hardness profile for Location 8 of the Shippingport NST inner wall............................... 7

7. Cutting diagram for base-metal and weld-meta) specimens from irradiated and shielded walls of the Shippingport NST ....................................................................................... 8

8. Best-fit Charpy transition curves for LT and n.. specimens from the Inner- and outer-10-mm regions of the shielded wall of the Shippingport. NS'I'.............................. 13

9. Best-fit Charpy transition curves for LT and TI.. specimens from the center mm region of the shielded wall of the Shippingport NST .................................................... 13

10. Best-flt Charpy transJUon curve for LT specimens from Location L4 of the shielded wall of the Shippingport NST ......................................................................................... 13 l l. Best-flt Charpy transi.tion curves for outer-region LT specimens, Inner-region LT specimens, and TL specimens from all regions of the Irradiated wall of the Shippingport NST.................................................................................................................................. 14

12. Yield stress estimated from Charpy-fmpact tests and measured from tensile tests for irradiated and shielded walls of the Shippingport NST ....................................... 16

13. Change In hardness of the shielded wall aft.er annealtng at 400°C. ... ...... .......... .... ........... .. I7

14. Change In hardness measured on Sections A. B, C, and D of the trradlated wall after anneaJJng at 400"C....................................................................................................................... 19

15. Charpy transition curves for annealed TI, specimens from the irradiated and the shielded wall of the Shippingport NST......................................................................................... 19

16. Charpy transition cmves for weld metal specimens from the tnner- and outer-I 0-mm reg:lons of the shielded wall of tlte Shippingport NST.. ... .. . .. .. . .. ... .. ..... . ..... .. . .. .. .. ... 2 2

17. Charpy transition curves for weld metal specJmens from inner region and outer region of the Irradiated wall of the Shippingport NST ........................................................... 2 3

18. Comparison of Charpy transJUon curves for nontrradlated and Irradiated HFIR surveillance samples and material from the Shippingport NST ......................................... 24

19. Shifts in CIT with neutron flucnce for the Shippingport NST material, HFIR surveillance samples. and HFIR A212-B Irradiated ln the ORR.......................................... 27

20. Comparison of C'IT shifts for the Shippingport NST and for samples .Irradiated ln the ORR with data from test reactors....................................................................................... 2 7 vi

21. Comparison of C'IT shlfts for the Shippingport NST and ORR-irradiated specimens wtth test reactor data for Irradiations at <93°C .................................................. 28

22. Shifts In CTI with neutron fluence for ASTM reference A212-B and A302-B steels irradiated at 277-310°C ......................................................................................................... 29

23. Compartson of CIT shifts for HFIR surveillance samples. the Shippingport NST.

and ORR-Irradiated sainples with data from test and Anny reactors............................... 29

24. Comparison of shifts in CIT with dpa for HFIR surveillance samples, the Shippingport NST. and ORR Irradiation with data from test reactor............................... 30

25. Plots of dpa (E > 0.1 MeV) vs. dpa rate for specific values of CTI' shifts ....................... 31 List of Tables

l. Typical Composl*!r>n (wt.%) of plate and weld metal from the Shippingport NST..... 5

2. Charpy-1mpact data for the Irradiated inner wall of the Shippingport NST.................. 9

3. Charpy-lmpact data for the shielded outer wall of the Shippingport NST. .. .. ... ... .... .... .. l0

4. Charpy-lmpact data for LocaUon IA of the shielded outer wall of the Shippingport NST ...........................................................................................................,...................... 12

5. Values of constants ln Eq. l, CTI, and USE for A212 Grade B material from shielded and irradiated walls of the Shippingport 1\JST......................................................... 12

6. Tensile test results for irradiated and shielded walls of the Shippingport NST.......... l5

7. Hardness values (Rockwell Bl of the Shippingport NST material from irradiated and shielded walls aft.er annealing at 400°C ................................................................................ 18

8. Charpy-lmpact data for material from the irradiated and shielded walls of the Shippingport NST after annealing at 400°C................................................................................ 20

9. Charpy impact data for the shielded outer-wall weld of the Shippingport NST.......... 21

10. Charpy impact data for the irradiated inner-wall weld of the Shippingport NST....... 21

11. Values of constants Jn Eq. 1. CTT, and USE for weld metal from the shielded and lnadJated walls of the Shippingport NST... ..... .. ... ........... ... ...... .......... ............ ... .. .. ... .. .. ..... .. 23

12. Summary of Charpy-irnpact and tensile results for HFIR surveillance tests. for samples irradiated in the ORR, and for the Shippingport NST .......................................... 25 vii

13. Chemical compositions (wt.%) of lern t !. steels from the HFIR suiveillance program ........................................................... 26

Executive Summary Data on surveillance specimens from the high-flux isotope reactor (HFIRJ at Oak Ridge National Laboratory showed a very high degree of embrittlement when compared with data obtained on stmllar materials in test reactors. The difference between the HFIR and test reactor data has been attributed to a fluence-rate effect. I.e .. the degree of embrtttlement per unit of fast fluence Increases at low neutron flux. and/or lo a softened neutron spec-trum. i.e .. a high thermal-to-fast-neutron-flux ratio may contribute to accelerated ernbr1t-tlement. Current Nuclear Regulatory Commission guidelines for the assessment of embr1t-tlement of pressure vessel support structures of commercial light-water reactors do not consider the contributions of nuence rate or spectral effects. HFIR results raise the possi-bility that the guidelines may not be sufficiently conservative.

To help resolve this issue, a program was initiated to characterize the irradiation em-brittlement of the neutron shield tank (NST) from the decommissioned Shippingport reac-tor. The Shippingport NST. which operated at 55°C ( 130°F). was fabricated from rolled A212 Grade B steel similar to that used for the HFIR vessel. The inner v*all of the NST was exposed to a total maximum fluence of .. 5 x 1017 n/cm2 (E > 1 MeV) over a life of 9.25 ef-fective-full-power years (efpy). This corresponds to a fast flux of =2 x 109 n/cm 2 *s. The HFIR surveillance specimens were exposed at a temperature of 50°C {122°F) over a period of"' 17 efpy; the flux of fast neutrons was ..,2 x 1Q8 n/cm2*s.

Eight disc* samples. "'155 mm in diameter. of the base metal and three weld samples were obtained from the inner wall of the NST. along with corresponding samples from the outer wall. Material characterization was carried out to determine chemical composiUon.

grain structure, and hardness. Signilkant variations in hardness were observed across the thickness of the NST wall; the hardness values of the inner- and outer-7-mm regions of the plate were .. 10% higher than those of the plate center. Irradiation embrtttlement was characteriz*~d by Charpy-impacl and tensile tests. as well as bv hardness measurements.

Specimens were obtained In longitudinal (LT) and transverse (TL) orientations from three 10-mm-wide regions [inner, center. and outer) ucross the thickness of the NST wall.

Samples from the outer wall. with about 6 orders of magnitude lower fluence. were used to determine baseline data for nonirradiated material.

Results indicate that the increase in Charpy transition temperature (CTI1 for the Shippingport NST Is not as severe as that observed for HFIR surveillance samples. The ac-tual value of CIT, however, Is higher than that fot HFIR A212-B steel. The shills in C1T at fluence levels between 3 and 6 x 10 17 n/cm2 [E >l MeV) are 23--28°C {41-50°F) for both LT and TL orientations. The TL orientation is weaker than the LT orientation, Le .. the CTI is higher and upper-shelf energy is lower. Increases in yield stress and hardness were 51 MPa (7.4 ks!) and 12-18 DPH. respectively. Annealing studies indicate complete recov-ery from embrlttlement after l h at 400°C {752°F). Although the weld metal Is significantly tougher than the ba_ :! metal. the shifts in CIT are comparable. The mechanical-property changes for the Shippingport NST agree well with the correlations between increases in CTI, yield stress, and hardness that have been developed for pressure vessel steels.

Radiation embrittlement of the Shippingport NST A2 l2-B steel Is consistent with the available data for irradiations at <232"C (<450°F) and shows very good agreement with the

data from test and Anny reactors. The results indicate that at the low operating tempera-tures of the Shippingport NST. Le .. 55°C (l30°F) fluence rate has no effect on radiation em-brtttlement at rates as low as 2 x 10B n/cm 2 *s. The accelerated embrtttlement of HFIR surveillance samples Js most Hkely due to the relatively higher proportion of thermal neu -

trons ln the HFIR spectrum compared to that for the test reactors.

2

1 Introduction Data on su.veillance specimens from the high-flux isotope reactor (HFIRl at Oak RJdge National Laboratory showed a very high degree of radiation-induced embrittlement in A212 Grade B, A350 Grade LF3. and Al05 Grade II steelsl.2 relative to that of similar materials irradiated at low temperatures in test reactors. 3 The difference between the HFIR and test reactor data has been attributed to a 0uence-rate effect. 1.e., the degree of embrttllernent per unit of fast fluence increases at low neutron flux, and/or to a softened neutron spec-trum, i.e., a high thermal-to-fast-neutron-flux ratio may contribute to accelerated embrit-tlement.1 The relative contribution of fluence rate and spectrum could not be discerned because the effects were concurrent and inseparable. The HFIR surveillance specimens were exposed over a period of .. 17 effective-full-power years (efpy) at S0"C (122"F) and a flux (E >l MeV) of 1QB-1Q9 n/cm2*s to fluences of 1011-1018 n/cm2. The fast-neutron flux for the HFIR surveillance samples was several orders of magnitude lower than that tn test reactors (1Ql3 n/cm2*s). At the surveillance position, thennal neutrons comprise 96% of the total flux. The possible effects of fluence rate or spectrum were validated by compara*

tive tests on HFJR A2 I 2-B steel irradiated In the Oak Ridge Research Reactor (ORR) at a flux of ... 1013 n/cm2*s. l.2 The results indicated that an order of magnitude greater fluence is needed in the ORR to produce the same shift in Charpy transition temperature (CTI) that was obseived in HFIR A212-B surveillance samples.

Current Nuclear Regulatory Commission (NRCI guidelines for the assessment of em-briltlement of pressure vessel support structures of commercial LWRs do not consider the contrtbutions of fluence rate or spectral effects. HFIR surveillance data raise the possibility that the guidelines may not be sufficiently conservatlve.4 To help resolve this issue, a pro-gram was Initiated to characterize the irradiation embr:lttlement of the neutron shield tank (NST) from the decommissioned Shippingport reactor. The Shippingport pressure vessel utilized the NST for Us support structure Shippingport was the first large-scale (72 MWe) nuclear power plant in the United Slates. Criticality was achieved in December 1957 and operation continued for a 25-y service life. Two separate pressurized-water reactor cores and one light-water breeder core were used during the life of the facillty.

The NST support for the reactor pressure vessel ts a skirt-mounted, annular, water-filled tank that consists of twc, concentric shells extending above and below the reactor core, Fig. 1. It 1s exposed to reactor core beltline neutrons. Water ts circulated through the tank for cooling and shielding. Shippingport NST operated at 55°C {130°F} and was fabri-cated from rolled A2 I 2 Grade B firebox steel similar to that used for the HFlR pressure ves-sel. The inner wall of the NST was exposed to a total maximum fluc:1ce of =6 x 10 17 n/cm 2 (E >l MeV) over a life of 9.25 efpy. This corresponds to a fast-neutron flux of ... 2 x 109 n/cm2,s. The temperature, fast fiuence, and fast flux for the Shlppingport NST were com-parable to those for the HFIR surveillance samples.

2 Material Characterization The effort to obtain samples from the NST was sponsored jointly by the NRC and the U.S. Department of Energy Plant Life Extension Program at Sandia National Laboratories.

Sample removal was scheduled wllh the ongoing decommissioning opcrallon that began Jn 3

Elo,alion

~fl' 31!** zoo- i,o- 1lS' 2n.oe Weld Aeacio, Vessel Support Assembly

+ *~ 3+

2 13 211 01 210.00 Neutron Shield Tank *+ 20i.5S 2'0I.H Oule< Wall Inner Wall H n. lhick lnnerWaU Waler Elevallon 36-in. Annulus 31 &* 200" 110' us*

.n , .01 lJlll"IISMn u+ 2:10.00

200.H LI + 0 ,1 + 0 u+ 0 1.4 + 200 17 Out&r Wall F'if}ure 1. Schematic representation of the F'if}ure 2. Layout of sample location in the NST support for the Shippingport NST (a) inner and Shippingport reactor pressure (b) outer walls vessel October 1982. The decommissioning plan called for a one-piece llft-out of the reactor pressure vessel and NST assembly following removal of fuel and core internals. and backflll-ing with a grouting material to serve as shielding. Cores were cut through the NST outer wall and grout material to gain access to the NST Inner wall. The coring operation involved the combined use of a commercially avallable diamond bit and a hole saw to remove .. 155-mm (6 in.) diameter samples from the !nner and outer walls of the NST. The actual sam-pling was performed by personnel from Battelle Pacific Northwest Laboratory. A more de-tailed description of the sampling a nd of the program plan have been presented earlier. 5 -6 Eight samples of the base metal and three weld samples were obtained from the lnner Nall.

along with corresponding samples from the outer wall. The layouts for the sample locations from the inner and outer walls are shown in Ftg. 2.

The inner wall is constructed from fou r plates. each ..25_4 mm I1 In.) thick. The two lower plates were Joined by vertical full-penetration welds located along the azimuthal po*

sitlons O and 180° :north and south reference axes of the NSTl; the two upper plates were welded along the azimuthal positions of 135 and 315°. These two assemblies were join<.d by a mid- height horizontal full-penetrallon weld. Specimen locations 13, 14, and 15 In Fig. 2 contain vertical welds: th e other locations represent the base metal from the four plates.

The outer wall was constructed from two plates Joined by vertical welds a( azimuthal positions O and 180°. A weld sample was obtained from the outer wall at Location 1. All other samples from the outer wall represent base metal from the two plates. No radiation 4

Table 1. Typical Composition (w L.%) of plate and weld m~tal from the Shippingport NST Element Plate Weld C 0.23 0 .065 Mn 0.76 0.93 p 0.02 s 0.03 SI 0.27 0.73a OJ 0.05 o.o6b Nt 0.0 4 0.02 Cr 0.04 0.03 0 O.QI N 0.004 Tl <0.005 0 .025 V <0.005 0 .019 Zr <0.005 <0.005 Mo, Ca. Al <0.01 <0.0 1 Se. Sn, B <0.0 1 <0.01 a 0.86 for Inner- wall weld rrom Location 14.

b 0 .07 for outer- wall weld and 0 .04 for Inner-wall weld from Location 15.

effect is expected at the extremely low levels of fluence for the outer wall: therefore, the outer- wall samples were used to determine baseline data for nonirradlated samples.

Samples from the lower portion of the NST outer wall. which had even lower nuence levels.

Le., Locations Ll - IA, were also procured to establish baseline matcrir1l properties.

Metallurgical charactertzauon and chemical analyses of samples taken from each of the p1ates strongly suggest t.hat both the inner and outer walls of the NST were fabricated from a single heat. Typica l elemental composition of the plate and weld metal is given In Table l.

Metallographlc examinalion of the NST ma terial Indicates tha! the rolling direction Is aligned with the circumferential direction of the shield t ank.

Micrographs of the grain structure of the Shippingport NST along the rolling and transverse sections are shown In Fig. 3. The s urfaces s hown In the mlcrographs are desig-nated by the direction normal to the surface. The transverse section shows some elongated grains. and a ll the Inclusions In the rolling direction are elongated. The inclusions In the rolling section are globular or flat. Average through- wall grain size was "' 12 µm. with no measurable change observed across the thickness of the wall.

Significant variations In hardness across the thickness of the NST wall were found at most locations. In general. the center of the pl?.te Is softer than the near-surface regions.

A typical through-wall hardness profil-! (Rockwell B, Ru) for the outer wall Is shown In Fig. 4. (The depth ls measured from the Inner s urface of the plate, i.e.. the surface toward the reactor core.) Hardness ranges from Ro 73 to 83. and hardness values of the inner-and outer mm regions of th~ wall are 5-10% higher than those in the center of the wall.

Although irradiation increases the hardness of the mattrial. the V- shaped profile Is main*

talned at most locations of the inner wall. The hardness profiles for Locatloris 3 and 9 of the NST inner wall are shown in Fig. 5. However, at a few locations of the inner and outer 5

(a) (b}

Figure 3. Micrographs of the Shippingport NST motertal along (a} transverse and (b) rolling sections 95 I I I I I I I I I ' . ~..........,........-+..,.....,.-r-.--+

Shielded Outer Wall 90 rt

.._. 85 A

I-** *- ** *- *- *-* - ** *- *I B .;

~** ** ** ** ** ** ** ** ** *1 D

F{fJure 4 .

Hardness profile across t11e thickness of i

'E 80 the Shippingport NST outer wall.

Regions A-B and c-n show the

~ 0 locations for Charpy- lmpact specimens Localoon 75 O 2 A A 3 o l4 70"1-'................-t-................"'-t.................._.__.................__._-t--'_.__.__..._

0 s 10 15 20 25 Thickness (mm) walls there was little or no Vdliatlon in hardness across the wall thickness: an example ls shown In Fig. 6 for inner- wall Location 8.

3 Irradiation Embrittlement Irradlat1on embriltlement was characterized by Charpy-lmpact and lenslle tests; the results have been presented earlier. 7 Specimens were obt.1ined In the longitudinal ILT) and transverse (TL) orientations* from three 10-mm-wlde regions {Inner. center, and outer) across the thickness of the NST wall. To avoid confusion wtlh regions, the NST inner and outer walls will be referred to as Irradiated and shielded walls, respectively. Thi locations of Charpy-!mpact specimens from the Inner- and outer-10-mm regions of the Irradiated and shielded walls are shown in Figs. 4-6. Average hardness of lhe samples from the

The firs t letter designates the direction nonnal to the plane of lhc crack. the scconct letter represents the direction of crack propagation. L ,. longitudinal or rolling direction and T = transverse direction.

6

os,+,......,..-.--,...+.,......,.--.--....+...,....,......."T""1...-~~--.-~.,.....-,-+I 90 lrrdl19d Loca!ioo J Inner WU

} go lrrdiled Inner Wall iii ~ iii

~

--.tTIi

~85 85

-6 80 I i 80 75 Shiokled / ' '

Ov..,.M.11

~ *-*-* - * - * -* - * - * - *-*I

/ /

1---* - *- * - * -* - * ~- - * -

I t 75 Ou>>r wall ' ..,. ..,

,*- *-*-*-*- *-*-*- *- *1 - *-*-*-* **-*-*-*-*-*I A B C 0 A B C 0 10_................_,_,__,_'-1-........_._.....,....._..............,_._........_._ 10 0 5 10 15 20 25 0 5 10 ,s 20 25 Thickness (mm) Thickness (mm)

Figure 5. Hardness profiles for Locations 3 and 9 oj the Shippingport NST inne;* walL Regions A- B and C-D show the locations for Charpy-impact spectme.ns.

95 -;--,,....,..-.-,-t-T""T--.--,-+..,....,....,..-,-j-,--r-,-....--t-,--.--,-,-+

lnldla1.cl Inner Wall l.oca!ion 9

!10 A B C D Ill a:

._. 85 Figure 6.

I Hardness profile for Location 8 of the C

'E 80 Shippingport NST inner walL RegiDns A- B ta and C- D show the locations for the Charpy--

z:

lmpact ..;pedmen.s.

75 70_................_ ................--+~........................................................................

0 5 1D 15 20 25 Thlckr.ess (mm) inner- and outer-IO-mm regions is approxlmately the same for all locations. whereas that of the center-10-mm region may be 5-10% lower for some locations. Layouts for the base-metal and weld- metal specimens from the irradiated and shielded walls are shown 1n Flg. 7.

Total fluence is a function of sample location (Fig. 2l: axial variation of fluence peaked at an elevation of 211.07 m (692.5 ft). and azimuthal variation of nuence peaked at tte 20 and 200° positions. The estimated fluence* at the inner ~urface of the NST l.rradlat~.-::! wall was "'6 x 10 17 n/cm 2 IE >l MeVl for Loca tions 3 and 9, c.4 x 101 7 n/cm 2 for Locations 2 and

8. "'3 x 1011 n/cm2 for Locations 14 and 15. and ... 2 x 1011 n/cm2 for Location 13. The data for actJvatJon measurements* Indicate that the fluence decreases by a factor of 2-3 across the thickness of the NST Irradiated wall. The fluences for the specimens from the outer-10-mm region were estimated to be a factor of .. 1.5 lower than those for specimens from the inner- '0-mm region. Material irnm the outer wall. which was protected by =0.9 m 13 ftl of wa * *rand hence had a fluence that was 6 orders of magnitude lower than .'.hat of the irradiated wall. was used to obtain baseline data for nonlrradiated material.

JameJ, L .. Betus Atomic Powe r Laboratory, private communication (April 1989).

t Crttnwood. L.. ATgonne National Laboratory. unpublished work (1989J.

7

... Rolling Di"ec1ion Sase Metal Samples Weld Samples Figure 7. Cutting diagram for base-metal and weld-metal specimens from irradiated and shielded walls of the Shippingport NST Charpy-impact tests were conducted on standard Charpy V-notch specimens ma-chined according to ASTM Specification E 23. A Dynatup Model 8000A drop-weight Im-pact machine with an inbtr... mented tup and data readout system was used for the tests.

Tensile tests were performed on dog-bone speclmens with a cross section of 4 x 5 mm (0.16 x 0.20 in.) and a gauge length of 20 mm (0.79 in.). The tests were conducted at an Initial strain rate of 4 x l Q-4 s- 1.

3.1 Base Metal 3.1.1 Charpy-lmpact Energy Results from Charpy-impact tests on LT .:ind 1L specimens from different regions and locations of the irradiated and shielded walls of the NST are given In Tables 2-4: best-fit Charpy transition curves are shown in Figs. 8--1 l. The Charpy data were fitted with a hy-perbolic tangent function of the form Cv = Ko + B{ 1 + tanh [(T - C)/Dll. ( 1) where Ko is the lower-shelf energy. T is the test temperature, B is half the distance be-tween upper- and lower-shelf energy. C Ls the mid-shelf CTT in °C. and D is the half-width of the transition region. The values of the constants Ko. B, C, and Din Eq. l. as well as the upper-sh*.!lf energy (USE) and the CIT at the 20.'hJ (15 ft-lb) and 41--J (30 ft,lb) levels for the various samples are gtven in Table 5.

The results for the shielded wall (Figs. 8 and 9) show no effect of sample location; there is little or no variation in the transition curves with vertical or azimuthal positton.

However, the TL orientation is weaker than the LT orientation. The CIT and USE, respec-tively, are l6°C (61 "Fl and 102 J/cm2 (.. 60 ft*lb) for LT specimens and 20°c (68°F) and

Table 2. Charpy-impact data for the l!Tadialed inner wall of the Shloplngport NST Specimen Fluenceb Orlen- TestTe.:!!E.;__ lmeact Ener~ Load (kN)

ID Location Resto:ia (n/cm2l tatlon 1*c1 1°Fl (J/cm2) (ft*lbl Yield Maximum I03R-05 3 6xl0 17 LT lO 50 8.4 5.0 8.788 8.788 I03R-01 3 LT 25 n 10.9 6 .4 9 .872 9.872 l03R-03 3 LT s.~ 131 45.5 26.8 12.313 14. 198 103R-07 3 I LT 70 158 68.8 40.6 12.303 13.934 I03R-08 3 0 LT 0 32 3.7 2.2 10. 194 10. 194 I03R-02 3 0 LT 25 77 14.5 8.6 12.411 12.411 l03R-06 3 0 i,T 40 104 25.2 15.5 12.254 12.254 103R-04 3 0 LT 55 131 61.0 36.0 11.356 l:!.866 103T-03 3 I Ti.. 120 248 51.3 30.3 10 .... 28 11.327 103T-02 3 0 TL 25 77 10.0 5 .9 10.389 10.389 103T-04 3 0 TL 70 158 40.6 24.0 11.F, i0 12.303 l09R-05 9 I 6xlo17 LT 10 50 7 .1 4.2 8 .983 8.983 109R-O 1 9 I LT 25 77 8.1 4.8 11.844 11.844 IOSR-03 9 I LT 55 13 1 39.0 23.0 12.401 13.866

!09R-07 9 l LT 70 158 51.7 30.5 12.518 13.856 109R-08 9 0 LT 0 32 3.1 1.8 7 .392 7.392

!09R-02 9 0 LT 25 n 11.3 6.7 11.044 11.044 l09R-06 9 0 LT 40 104 31.3 18.5 12.391 12.391 l09R*04 9 0 LT 55 131 48.5 28.6 12.254 14.256 l09T-03 9 I TL 40 104 19 . i 11.3 10.194 10.194 109T-0I 9 I TL 55 131 28.0 16.5 12,782 13. 123 l09T-02 9 0 TL 2.5 77 9.6 5.7 11.786 11.786 I09T-04 9 0 TL 90 194 53.0 31.3 11.424 12,4 11 l02R-05 2 I 4x1Q l7 LT 10 50 9.9 5.8 9.120 9 . 120 I02R-Ol 2 I LT 25 77 13.5 8.0 10.331 10.331 I02R-03 2 I LT 55 131 52.8 31.2 11 .688 13.260 I02R-07 2 1 LT 70 158 frl.7 39.9 11.678 13.612 I02R-08 2 0 LT 0 32 22.7 13.4 11.854 11.854 102R-02 2 0 LT 25 77 11.6 6.8 12.616 12.616 I02R-06 2 0 LT 40 104 28.5 16 ,8 10.975 10.975 102R-04 2 0 LT 55 131 68.6 40.5 11.190 13.670 102T-0I 2 I TL 55 131 40.9 24.l l l.639 l2.186 l02T-03 2 I TL 120 248 53.3 31.4 9.989 I I.317 102T-02 2 0 TL 25 77 12.4 7.3 12.821 12.821 l02T-04 2 0 TL 70 158 49.4 29. l 10.966 12.059 l08R-05 8 4x 101 7 LT l0 50 9.5 5.6 13.387 13.387 l0SR-0 I 8 LT 25 77 10.3 6.1 9 .647 9.647

!0SR-03 8 I LT 55 131 43.5 25.7 12.381 14. 188 l0BR-07 8 I LT 70 158 79.7 47.0 11.747 13.729 I08R-08 8 0 LT 0 32 3. 1 1.8 8 .935 8 .935 I0SR-02 8 0 LT 25 77 12.8 7.6 12.811 12.81 I J0SR-06 8 0 LT 40 104 26.8 15.8 12.39 1 12.391 I0BR-04 8 0 LT 55 13 1 53.5 31.6 I \.962 13.993

!0ST-03 8 I TL 40 104 24.8 14.6 12.635 13.026 l08T-0J 8 I TL 55 131 40.7 24.0 11.913 l~.635 I0BT-02 8 0 TL 25 77 12.3 7.3 13.221 13.221 I0BT-04 8 0 TL 90 194 49.7 2.9.3 10.995 l \.874 ll4T-0IC 14 C 3xlo 17 TL 0 32 5.5 3.2 8 .524 8.524

! 14T-02C 14 C TL 25 77 12.8 7.6 9 .374 9 .374

! 14T-03C 14 C TL 55 131 36.8 21.7 11.502 11.502 I 14T-04C 14 C TL 90 194 55.8 32,9 10.253 11.288

!14T-02 14 0 TL 90 194 54.8 32.3 10.643 l l.835 IIST-0IC 15 C 3x lo 17 TL 10 50 8.4 5.0 8.476 8.476 115T-02C 15 C TL 40 104 27.0 15.9 l l.512 11.5 I 2 I 15T-03C 15 C TL 70 158 40.2 23.7 10.721 10.917 1!5T-04C 15 C TL 120 248 46.2 27.3 9.511 10.321 9

Ttr.ble 2. (Contd.)

Specimen Flucnceb Orlen- TestTeme, lmeact EncrQ: Load {kNJ ID Location Regton 8 ln/cm2J talion ("Cl ("fl (J/cm2) {ft*lbl Yield Maximum IIST-01 15 3xJ017 Tl. 40 104 21.1 12.5 12.665 12.665 113T-02 13 0 2x10I7 TL 25 n 10.l 6.0 9.325 9.325 a I. O. and C represent the Inner-, outer-, and centcr-10--mm regions, respectively, across the wall. thickness.

b Reprc!l('nt the values at the Inner surface of the wall. Fluence for the outer-region samples Is estimated to be a factor of 1.5 lower than that for Inner-region samples.

Table 3. Ch.arpy-tmpa.ct data for the shielded outer wall of the Shippingport NSf Specimen 01""len- Test Teme. Imeact EncrQ: Load (kN)

ID Location Reslo:, 8 tatlon 1*c1 ["Fl '-J!cm2l [ft,lbl Yield Maximum 002R-05 2 LT IO 50 18.7 11.0 10.448 10.448 002R-Ol 2 I LT 25 77 29.4 17.3 11.971 12.704 002R-03 2 I LT 55 131 114.3 67.4 10.692 14.530 002R-07 2 I LT 70 158 95.5 56.3 10.917 13.651 002R-08 2 0 LT 0 32 11.9 7.0 11.239 11.239 002R-02 2 0 LT 25 77 48.0 28.3 11.522 13.934 002R-06 2 0 LT 40 104 71.2 42.0 10.41)7 13.358 002R--04 2 0 LT 55 131 92.0 54.3 10.692 13.973 002T-03 2 I TL 0 32 11.l 6.5 10.878 10.878 002T-Ol 2 I TL 55 131 57.0 33.6 11.083 12.939 002T-02 2 0 TL 25 77 31.2 18.4 12.040 12.870 002T-04 2 0 TL 90 194 69.8 41.2 I0.145 12.284 003R-05 3 I LT 10 50 17.0 10.0 10.497 10.497 003R*Ol 3 I LT 25 77 35.1 'llJ. 7 11.:007 12.186 003R-03 3 I LT 55 131 94.8 55.9 10.155 13.592 003R-07 3 I LT 70 158 102.4 60.4 l0.780 14.022 003R-08 3 0 LT 0 32 10.7 6.3 10.526 10.526 003R-02 3 0 LT 25 77 48.0 28.3 10.975 13.690 003R-06 3 0 LT 40 104 61.1 36.1 11.405 14.032 003R-04 3 0 LT 55 131 98.3 58.0 10.214 13.680 003T-03 3 I TL 10 50 18.0 10.6 12.713 12.713 003T-0l 3 I TI. 55 131 53.6 31.6 11.376 13.241 003T-02 3 0 TL 25 77 31.3 18.5 12.088 12.694 003T-04 3 0 TL 70 158 68.9 40.7 10.497 12.518 003T-06 3 0 TL I 'llJ 248 68.0 40.1 9.276 11.522 006R-05 6 LT 10 50 14.2 8.4 10.887 10.887 006R-Ol 6 LT 25 77 24.0 14.2 11.522 I J.522 OOBR-03 6 LT 55 131 89.1 52.6 10.594 14.051 006R-07 6 i LT 70 158 89,5 52.8 10.887 13.914 006R-08 6 0 LT 0 32 13.2 7.8 9.110 9.110 006R-02 6 0 LT 25 77 42.5 25.1 11.395 13.465 006R-06 6 0 LT 40 104 50,4 29.7 11.58 I 13. 163 006R-04 6 0 LT 55 131 93.3 55.1 10.477 13.846 006f-03 6 TL 40 104 42.4 25.0 11.405 12.528 006T-OI 6 I TL 55 131 56.7 33.5 11.268 12.831 OOST-05 6 I T!.. 90 194 69.8 41.2 9.662 12.206 OOST-04 6 0 TL 0 32 13.4 7.9 13.075 13.075 006f-02 6 0 TL 25 77 24.7 14.6 12.860 12.860 009R-05 9 LT 10 50 26.2 15.5 12.479 12.479 009R-0J 9 LT 25 77 43.2 25.5 11. 190 11.971 009fW3 9 LT 55 131 85.7 50.6 10.614 12.381 009R-07 9 I LT 70 158 88.0 51.9 10.272 12.635 009R-08 9 0 LT 0 32 16.5 9.7 12.987 12.987 OOOR-02 9 0 LT 25 77 46.4 27.4 ll.278 I l.883 009R-06 9 0 LT 40 104 67.5 39.8 1Ll71 12.215 10

Table 3. (Contd.).

Specimen O rlen- Test Teme, lmeact Ener~ Load (kN)

ID Location Rcjllona talion 1*c1 (*F) µ/ cm2l (ft- lb) Yield Maximum 009R-04 9 0 LT 55 131 87.7 5).7 10.057 12.430 009T-03 9 I TL 10 50 30.0 17.7 12.098 12.098 0091'-01 9 I TL 55 131 52.5 31.0 10.565 11.454 009T-02 9 0 TL 25 77 36.5 21.5 12.342 12..342 oogf-04 9 0 TL 70 158 58.2 34.3 9.804 11.180 oogf-06 9 0 TL 120 2A8 56.9 33.6 8.505 10.555 Ol2R-05 12 I LT 10 50 19.0 11.2 12.616 12.616 012R-03 12 I LT 55 131 84.0 49.6 10.292 12.782 012R-07 12 I LT 70 158 9 1.3 53.9 10.468 12.967 Ol2R-08 12 0 LT 0 32 15.8 9.3 10.184 10. 184 Ol2R-02 12 0 LT 25 77 49.4 29. l 11.083 11.796 012R-06 12 0 LT 40 104 66.3 39. 1 11.288 12.6:25 Ol2R-04 12 0 LT 55 131 96.5 56.9 9.843 13.114 OIZT-03 12 TL 40 104 39.5 2.'3.3 l l .444 12.206 OlZT-01 12 TL 55 131 57.9 34.2 10.341 11.581 012T-05 12 TL 90 194 62.4 36.8 9.462 11.366 Ol2T-04 12 0 TL 0 32 8.9 5.3 10.262 10.262 012T-02 12 0 TL 25 77 21.7 12.8 12. 137 12.137 Ol4R-OIC 14 C LT 0 32 19.9 11.7 12.001 12.00 l 014R-05C 14 C LT 25 77 58.7 34 .6 10.663 12.586 O l 4R-02C 14 C LT 40 104 84.6 49.9 10. 175 12.704 Ol4R-06C 14 C LT 55 13 1 100.3 59.2 9 .667 12.420 014R-04C 14 C LT 70 158 114.3 67.4 9 .403 12.616 014R-03C 14 C LT 90 194 103.1 60.8 8 .495 11.464 0l4R-01 14 I LT 40 104 58.4 34.5 11.288 13.514 014R-02 14 0 LT 90 194 102.5 60.5 9.999 12.831 014T-OIC 14 C TL 10 50 35.8 21.1 12. 157 12.157 014T-03C 14 C TL 40 104 64.0 37.8 9.950 12.069 0l4T-04C 14 C TL 55 131 68.5 40.4 9.520 12.079 014T*02C 14 C TL 90 194 73.3 43.3 8.895 11.532 OlSR--OIC 15 C LT 10 50 19.3 11.4 12.372 12.'372 015R--02C 15 C LT 25 77 5 1.8 30.6 11.337 14.422 015R-05C 15 C LT 40 104 77.4 45.7 10.887 13.914 015R--03C 15 C I.T 55 131 79.8 47.1 10.555 13.729 OJ5R-06C 15 C LT 80 176 105.4. 62.2 10.048 13.573 Ol5R-04C 15 C LT 1:20 ~8 95.l 56.1 8.866 12.362 OlSR-01 15 I LT -20 -4 6.4 3.8 10.879 10.879 015T-04C 15 C TL 0 32 29.4 17.3 12.440 14.129 015T.OIC 15 C TL 25 77 42.3 25.0 11.688 13.886 015T-02C 15 C TL 25 77 46.3 27.3 11.405 14.285 015T--03C 15 C TL 70 158 76.6 45.2 9.823 13.387 a I. 0, and C represent the Inner-, outer-. a nd ccnter- 10-mm regions., respectively. across the wull thickness.

67 J/cm 2 (... 4Q ft -lb) for TL specimens. The dlfferences In Impact strength for the two or1-entattons are attributed primarily to differences in the distribution of Inclusions along th e crack plane. The plane of the crack for the TL orientation, i.e .. th e transverse section shown In Fig. 3. contains elongated inclusions.

The results also indicate some effect of position through the thickness of the wall; Im -

pact energies for specimens from the Inner and outer regions of the wall are comparable, whereas those for the center specimens are slightly h igher. The crr and USE for the cen-ter specimens are. respectively, 9°C (48°Fl and 103 J/cm 2 (.. 61 ft -lbl for the LT orientation 11

Table 4. Cha.rpy-tmpact data for Locauon L4 of the shielded Olller wall of the Shippingport NST Specimen Or!c-n- Test Terne- Imeact Energy ID Location Reglon 8 ta tlon 1*ci !0 F'l lJ£'.cm~ !ft,lbl lAI - 1 1A I LT -40 -40 6,8 4.0 lAQ--6 lA 0 LT -40 -40 2,5 1.5 lAl - 7 1A J LT 0 32 5.9 3.5 lA0-2 1A 0 LT 0 32 5.1 3.0 lAl - 2 L4 I LT 15 -9 22.0 13.0 lA0--3 L4 0 LT 15 -9 20.3 12.0

[A) -3 L4 I LT 25 77 60.2 35.5 lA0--1 1A 0 LT 25 77 55.l 32.5 lA0-8 lA 0 LT 25 77 30.5 18.0

[A[ -9 lA I LT 55 131 88. 1 52.0 L40--4 1A 0 LT 55 131 86.4 51.0 L41 -5 1A I LT 70 158 105. 1 62.0 a I and O represent the Inner- and outcr-10-mm regions. respectively. across the w all thickness.

Table 5 . Values of constants in Eq. 1, CIT, and USE for A2 l 2 Grade B material from shielded and (rradlated walls of the Shippingport NST Con stants cTTa Sample Sample Orient- Ko B C D 20 .4 41 USE Location Region at Ion !JLcm2J (J£'.cm21 10CJ (*ci IC l°Fll 1*c l"Fll !J£'.cm2 !fl-lb!!

Shielded Outer Wa ll All Inner & LT 8.6 46.7 33.5 23.2 16 (6 1) 31 (88) 102 (601 Outer 14,15 Center LT 6.1 48.2 26.0 23.3 9(48) 24 (75) 103(61)

All Inner & TL 8.0 29.3 33.5 30.3 20(68) 49 0201 67(401 Outer 14, 15 Center TL a.ob 33.5 18.4 30.3b 2136) 27 (81) 74 (44)

Irradiated Inner Wall 2,8,3,9 Inner LT 8.3 33,6 53.3 17.2 44 1111) 58 (136) 76 (451 2.8,3,9 Outer LT 8.2 34.3 47.1 16.l 391102) 5 1 {124) 771451 2,8,3,9, 14.15 AllC TL 4 .1 24. l 46. 1 25.5 431109) 52 13 1) a Charpy translUon tem~rature at the 20.3-J (1 5 ft-lb) and 41-J 130 ft -lb) levels.

b Values of Ko and Dare assumed to be the same as those for Inner/outer region.

c Inne r- and outer- region specimens from l..ocatlons 2. 8, 3. a nd 9 a nd center region specimens from Locations 14 and 15.

and 2°C (36°F) and 74 J /cm 2 (~44 ft-lb) for the TL orientation. These variations are consis-tent with the throughwall variations In h ardness. The hardness of the center region ls In the range of Rs 73-75 (132- 142 DPH); for the .nner or outer regions It is 78-84 (144-162 DPH).

Charpy- tmpact data for material from Location L4 of the shielded wall, shown in Fig. 10, are In good agreement with the data from other locations of the wall. The results Indicate that the Charpy transition curves for thr shielded wall of the NST represent base-line data for nonlrrad1ated material. As discussed in Section 2, the hardness of the center of the wall vanes significantly at different ax.Jal ai1d azimuthal locations. The data for the center s pecimens may not be representative of all locations. Therefore, results from only the inner and outer regions of the wall are used to charactertze irradiation embrtttlement.

12

Temper11ur* (°F) T*mptt*ture ( 0 F}

50 0 50 100 150 200 250 300 -SO 0 50 100 150 200 250 300 150 ,so

..E NST Shielded Wall A:t12-8 $tell ao 70 ~

NE u

NST Shl11lded Wall A212-3 SIM I Loca.lk>n 0 2 IO 70 i" u C 3

, .::; TL Orlamauon 60 ~

LT Oti*nlatlon

_, 1CO cl' Inner/Outer Regions 60 50 tJ

.. --> 100 tJ lnner/()\Aer RagloM

B 12 50 tJ

,;. Loca1 ion ,;. ,;. ,;.

r

Cl 40 0 2 40 ei C C:

so 0 3 30 Ji w C

so

30 w

6

9 u

l

'6 20 "6 20 12 a.

! 0 10 0

.5 ci..

.5 0

,o 0

50 0 St 100 150 -SO 0 so 100 150 Tempttature (°C) Tmperalur11 (°C)

(a) (bl F'tgure 8 . Best-jll. Charpy transUion curue.s for (a) LT and (b) TL specimens .from the Inner-and outer- I 0-mm regions of the shielded wall of the Shippingport NST Tempttalurt (°F) Temperature (°F)

50 0 50 100 150 200 250 300 -so 0 50 100 ,so 200 250 300 150 -

150 80 NST Shielded Wall ao N NST Shi11ld11d Wall E

u A212-B :SIH I 70 i E u A212-8 SIM I . 70 ~

l T Otlontatlon

z: .::;

_, 100 TL Orl*nlallon to

._.. 100 u>

Cen1er Region *,.

1.

60 so CJ

> u,;. ~n111r AIIQIOn 50 o>

w Cl i 50 I

I

' . ' IL'V.' ,,/()_,..,

Roglon 40 30 w

Cl C w al C

50 Local Ion 40 30 f LI.I Loc.,,hon

~ ia. ~~~*-~ ~~ ~. . ,. i

'G t; 20

a. IS"*b CTT

- - - - ~ /

I 20

. ,. 'ln,,.,IOvter * ,~

.5 0

10 0.

E .§ 0

,, flelllon 10 0

I 0

-so 0 50 100 150 -50 0 50 100 150 Temperature (°C) Temp*r*lure (°C)

(a) (b)

Figure 9. Best-fit Charpy transitton curoe sfor (a) LT and (b) TL specimens.from the ce nter-10-mm region of the shielded w all of the Shippingport NST Tempe,-1ure (°F)

50 0 so 100 150 200 250 300 150 -

..- E NST Shielded Wall A212-B Sltel 80 70 :0 u

_, 100 u

LT or1, n1* Uon loca110n L4 0 eo so 0

- Figur e 10,

.,:. ,;. Best-.ftl Charpy transUt.on curoe for LT CD 40 = specimens from Location IA of the shielded

~ 50 Ou111111 Wall 30 IIJ t

c::

IIJ "'ALoc.afio,.s wall of the Shippingport NST 0

15 fl*lt> CTT 20 10 0

0.

.5 .§ 0 0

-50 0 so ,oo 150 TemperalUre c*c1 13

T,m.,.,.ture (°F) T*mpertture (0 F)

.50 0 so 100 150 200 250 300 -5D O 5D 1 OD 1 50 200 250 3D0 NST Irradiated Wall 10 NST Irradiated Will ID A212-8 SIMI 70 i A212... SIM I 70 ~

LT Orientadon

~ ii:

°'""' AoQlotl 10 LT Orl*"'*tlon

"" "" i.g1on -- -- --- eo ....

50 0 Sl,iolood / 50 0

~ :,;;

Wal "// o O 21 r 40 ~

i Locallol'I 30 ! ~ 50 locallOf'I ao Ill o 3

't 20 l tl l

0 J

t 20 tl l

0

10 S 0 2

' 10

-50 0 50 100 15D *50 0 50 100 150 Temptrtlure ("C) Temperature ('Cl (a) (bl Tem.,.rllUrt (°F)

50 0 50 100 150 200 2 50 300 150 NST Irradiated w,11 LOC&Jion ao E

u A212- 8 Sltel 0 3 t 10 ~ Figure 11.

n Orlenlallon D

- 2 !:. Best-JU Chai p!J transttton curoes for (aJ outer-

60

--- ---*~

- 100 0po<> Symbot: 10,..,,0.,., ~ * >

o"" CloMd Svn,tx,11: c ....., flogioa ** 14 50 0

,;; region LT specimens. (b) inner- region LT E!' 40 ct specimens. and (c) TL specimens from aU a;

~

w tl 50 . 30 ~ regions of the irradiated wall of the Shippingport NST l

20 l 0

- ---- *a 0

50 0 50 10D 150 Temp*rature (°C)

(cl The transition curves for the LT and TL specimens from Locations 3, 9, 2. and 8 of the Irradiated wall of the NST are shown tn Ftg. 11. The Irradiated specimens show a higher err and lower USE relative to those from the essentially n onlrradlated shielded wall. The effect of minor differences in total fluence between Locations 2 and 8 ("'4 x 10 17 n/cm2 flu-ence) and Locations 3 and 9 (u6 x 1017 n /cm2 fluence) are minimal and cannot be tstab-llshed from the data; Locallon 9 appears to have a s llghtly higher CIT than the other loca-tions . Some effects of position through the thickness of the wall are observed for LT specimens. whose shift in CIT ls greater for specimens from inner-region than for outer-region specimens. The values of CIT are 39°C ( 102°Fl for the outer region and 44°C

( 11 1°f) for the Inner region, a shift of -.23 and 28°C (.. 41 and 50°F) for the outer and inner regions, respectively. The USE cannot b e directly established from the data In Figs. l la and l lb. However. the specimens tested at 55°C (l31°Fl show 100% shear fracture; thus, the Impact energy of these specimens, I.e .. an average of ,.,77 J/cm2 (45 ft-lb). ts represen-tative of USE.

The shift in CTT of the irradiated-wall TL specimens (Fig. l le) Is also 23°C (41°fl.

similar to that ror the LT specimens. However, the USE Is lower. I.e .. 52 ,J/cm2 (31 ft -lb).

The effect of posJt!on through the thic kness of the wall Is minimal for the these specimens.

The Impact energies for the center specimens are comparable to those for specimens from 14

Table 6. Tensile test resultsfor im.,diated and shielded walls of the Shippingport NST EnSlneerlnS True Test 0 .2% Yield Ultimate Fracture Elong- Red. In Specimen Temp. Stress Stress Stress atlon Arca Number Location Rcglon8 Flue nceb 1*c1 (MPal (MPal (MPa) (%) ;%)

lrrdd1atcd Inner Wall 13RI 3 I 6xJOl7 25 368.0 572.l 777.3 28.2 43.9 13RC 3 C 25 323.0 502.6 686.8 31.3 47.6 13RO 3 0 25 354 0 556.6 769.5 28.7 43.5 12RI 2 4xio 17 25 325.2 538.3 779.5 29.4 46.4 l2RC 2 C 25 319.7 540.4 '/33.7 29.2 42.9 12RO 2 0 25 368.0 548.8 730.5 27.8 41.4 19RI 9 6x101 7 55 333.6 520.l 665.7 27.2 39.9 IGRC 9 C 55 305.2 479.2 666.9 28.3 45 7 19RO 9 0 55 524.5 706.0 27.2 41.9

!SRI 8 I 4xJol 7 55 304.4 521.0 716.5 28.8 45.6 ISRC 8 C 55 310.4 515.0 732.6 29.0 45.3

!BRO 8 0 55 322.1 520.6 725.9 30.4 44.0 Shielded Outer Wall 03Rl 3 I 25 282.9 536.5 740.8 30.2 43.6 03RC 3 C 25 277.6 493,9 627.8 32.4 41.8 03RO 3 0 25 324.6 540.4 709.2 29.4 43.3 06RJ 6 I 25 294.0 528. l 691.6 31.1 40.4 06RC 6 C 25 270.0 532.0 700.7 30.6 40.6 06RO 6 0 25 30,.3 5:2.7.2 728.3 30.3 43.4 012Rl 12 25 329.6 530.5 768.1 27.5 45.6 012RC 12 C 25 256.2 491.5 715.0 33.5 48.1 012RO 12 0 25 293.6 5 11.0 735.4 29.4 46.8 02RJ 2 I 55 280.2 519.5 711.2 :10.3 44.7 02RC 2 C 55 264.7 483.9 608.0 ~12.2 44.9 02RO 2 0 55 286.2 520.7 701.4 :!9.8 42.4 09Rl 9 I 55 248.2 489.9 669.5 31.3 46.2 09RC 9 C 55 227.8 452.4 640.5 33.5 47.3 OORO 9 0 55 279.6 492.9 727.9 30.6 47.6 a I. O. and C represent :he Inner- . outer- . and center-5-mm regions. respectively. across the wall thickness.

b Represent the values at the In ner surface of the wall. Fbence for the outer-region samples Is esUmated to be a factor of 1.5 lower than th at for Inne r-region samples.

the Inner and outer region. Center specimens were obtained from locatiuns 14 and 15, where the variation in through-wall hardness was m1ntmal 3.1.2 Tensile Properties TensJle tests were conducted at room temperature and at 55°C (131 °F) on LT spf,Cl-mens from several locations of the irradiated and shielded walls of the NST and from three regions across the thickness of the wall. The res ults of these tests. given ln Table 6 , indi-cate little or no variation with vertical and azimuthal position. The tensile strengths of t he irradiated wall are higher than those for the shielded wall . The tensile properties also re-flect the effect of variations tn hardness observed across the thickness of the wall, Le .. the yield and ultimate stresses for specimens froF. the center of the wall are always lower than 15

those for specimens from inner or outer regions. However. the Increases in yield and ulll*

mate stress due to irradiation ts Independen t of position throu gh the thickness. The In*

creases in yield and ultimate stress are, resper lvely. =51 and 20 MPa (=e7.4 and 2.9 ks!) at room temperature, and ..39 and 18 MPa (=5.5 and 2.6 ksl) at 55<>C.

The tensile properties of the NST material were also estimated from the Charpy-lm-pact data. The yield stress is estimated from the expression (21 taken from Ref. 8, where Py is the yield load obtained from the load-time traces of the in-strumented Charpy tests. W ts the specimen width, B ls the specimen thickness. b ts the uncracked ligament. and A ls a constant obtained by comparing the tensile and Charpy data for LT specimens tested at room temperature and at 55°C. The best-flt value of A was 1. 73.

Yield st~esses estimated from Charpy-lmpact tests and those obtained from tensile tests for material from the irradiated and sh ielded walls are s hown In Fig. 12. The results show the expected decrease In y ield stress with a n Increase In test te mperature. Irradiation in-creases the yield stress a t all test temperatures.

The tensile d;:;ita for the Shippingport NST show very good agreement with the corre-lations between the Increases In CIT and yield stress that have been developed for pressu re vessel steel.9, 10 The CTr shift (In °C) and Increase In yield stress Aoy (ln MPa) Is expressed as ACIT = CAoy, ( 3) where C Is ..Q.5°C/MPa for plate material. and 0.65°C/MPa for welds. The shift in the CIT of both LT and TL specimens ts 23<>c (28°C for LT specimens from inner regions). and the Increase In yield stress 1s 51 MPa.

Temperature (°F) Temper*lure (°F)

. 50 0 so 100 150 200 250 300 *50 0 so 100 150 200 250 300

~

400 -.+-,-.....-,,.+-,-.,.......,+.-,~+.-,~-+,-~+.-,~+...~ ~ 400

a. NST Irradiated wau: *

a. NST Shielded Wall ii ti,t

I ::IE 350 -

.... 350 o O A212B Slffl 50:

~= 300 ~"

J:,

u,

~-~

300 ii

'O 0 l!

~ 0

"°!~

250

]l Shiel-Wall 250

0 0 0

~

0 ,,

Ill 200 "

ii 200 Opon Syrrbol5 De!Qrmined frQm Chl'!IY T*IS 30 !

.§ , Opon Syrrt>oll' Deie111inacj hom Charpy THIS ]

iii iii

.5 w

ClosQ<j Symbols. MHsured lrom Tons,lo T1151*

w Cosed Syrrbols. Musured kom Tensile T111, ;

l 50 .,......~....._......,._.._...___._-+-._.__._...__._._........,'--'-~ 150 w

SO 0 50 100 150 *SO 0 50 100 1 50 Tempera ture (°C) Temperature (°C)

(a) (bl Figure 12. Yield stress eslimaledfrom Charpy-lmpacl tests and measuredfrom tensile tests Jor (a) Irradiated and (b) shielded walls of the Shippingport NST 16

3.1.3 Recovery Annealing Anneallng studles were conducted to examine the recovery behavior of the embrittled material. Samples from Locations 3, 9. and 8 of the Irradia ted wa U and locations 3 and 6 of the shielded wall were annealed up to 154 h at 400°C (752°Fl and recovery behavior was characterized by hardness measurements. As mentioned In Section 2, hardness changes across the thickness of the NST wall. Consequently, hardness measurements were made on surfaces that were normal to the hardness profile, I.e .. sections marked A. B. C. and D In Figs. 4-6. Hardness values a re approx.Jmately the same along these sections. although they may vary significantly among the sections. The recovery behavior was estabUshed by hard-ness measurements on the same specimen before and after annealing for various Umes.

The results are given In Table 7.

The results Indicate that hardness of the irradiated Inner wall decreases whereas that of the shielded ou ter wall Increases after annealing. The changes in hardness occur within l h at 400°C and there Is little or no change in hardness after annealing for longer times.

The results for the shielded outer wall. shown In Fig. 13. Indicate that the Increase In hardness 1s observed along sections B. C. and D. whereas. section A shows Hlt1e or no change l.n hardness The average Increase In hardness for Locations 3 and 6 a long sectll.)ns B. C. and D is Rs l.4 (DPH 4.4). It can be argued that the observed hardness profiles (Figs.

4- 6) are a result of residual stresses. Le.. tensile stresses In the center and compressive stresses near the surface of the wall. S tudies on the effect of residual stresses on hardness Indicate that hardness m easurements change when a unlaxlal stress is applied perpendicu-la r to the load of the hardness tester: hardness decreases linearly with applied tensile stress. wherea s compress ive stress causes llltle or no change In hardness. 11 , I 2 However.

a nnealing studies were conducted on 10-mm- wtde samples from the Inner and outer re-gions or the wall. Therefore. sections A. B. C, and D should have compressive s tresses and little or no effect on hardness measu rements. Furtherm ore. hardness values for all speci-mens prior to an:!eallng are In very good agreement with the results In Figs. 4-6. The ob-served hardness pr,,flles are likely caused by metallurgical factors. The Increase In hard-ness after annealing may be attributed to thermal agtng.

The changes in hardness of material from t he Irradiated inner wall. shown In Fig. 14, renect the differences in nuence and tlux levels of the different locations, I.e .. the decrease for Locations 3 and 9 (.. 6 x 10 17 n/cm 2 nuence) is greater than for Location 8 (... 4 x 1Ql7 5 , , "I 0 I IO ,,t---. 0 I I i I 1,I o:i a:

111 u,

Cl>

C

-0 io

c D

I Figure 13 .

Change tn hardness of the shielded

.s NST Shielded Will wall after annealing at 400°C GI *5 O') Anneled al 400' C C l ocation

.! Open Symbolo S1<1r,ns 0. C. A D o 3 0 A 6 CIOsed Symbo~ Stc110~ A 10 100 Time (hi l7

Table 7. Hardness values (Rockwell JJ)a of the Shippingport NST material.from irradiated and shielded walls ajler annealing at 400°C Annealing Time !hi Location Sccttonh 0 4 10 24 26.5 90 154 Shielded Outer Wall 3 A 82.5 82.8 82.4 81.7 82.2 B 76,8 78.4 79.2 78.6 78.7 C 75.8 78.0 78.7 78.1 78.1 D 83.6 85.0 84.0 83.7 83.2 A 82.6 82.3 84.3 82.9 8 76.6 78.3 77.6 79.0 C 76.3 77.5 78.0 79.2 D 82.0 83.6 83.3 84.2 6 A 83.6 83.9 84.2 83.6 83.6 8 77.9 79.8 79.9 79.7 80.1 C 79.4 80.5 79.9 79.5 79.8 D 81.2 82.B 82.0 82.3 81.8 A 84.9 83.5 83.7 ~ -6 8 77.7 80.0 80.1 80.3 C 78.5 79.9 80.0 80.7 D 81.2 83.9 82.6 83.3 Irradiated Inner \Va ll 3 A 87.5 83.2 83.5 83.6 82.8 B 80.7 77. 1 76.8 77.5 76.8 C 81.7 77.8 78.0 76.9 77.2 D 84.9 83.2 82.8 82.6 82.6 A 87.3 83.8 84.1 84.6 B 80.9 77.2 77. 1 77.8 C 80.4 75.2 75.5 76.1 D 86.8 85.2 85.1 85.6 9 A 86.8 82.5 82.6 63.3 82.6 8 82.3 78.4 78. 1 78.2 78.2 C 8 1.6 76.2 76.8 77. 1 75.9 D 84.5 82.6 82.0 79.6 80.9 A 87.9 82.6 83.8 83.4 8 82. 1 77.7 77.9 78.6 C 81.9 76.3 76.4 77.6 D 84.2 81.1 81.6 82.2 8 A 84.8 8 1.8 8 1.4 81.6 8) .3 B 84.6 W'2 82.4 80.4 81.5 C 86.0 83.0 82.8 83. 1 83.I l) 83.3 81.5 80.8 91.1 8().4 A 86.6 84.3 82.7 83.3 B 81.7 78.5 78.3 78.4 C 83.2 79.1 79.5 80.7 D 83.4 81 .3 81.6 81.4 a Values represent the average or at least five measu rements.

b Represents the surfaces marked In Figs . 4-6 where measurements were made.

n/cm2 fluence) . Also. the decrease 1n hardness along Sections A. 8, and C Ls greater than that a long Section D. As menuoned earlier. total Ouence decreases by a factor of 2-3 across the thickness of the Irradia ted inner wall. The decr ease In h ardness for Locations 3, 9, and 8 is r espectively. Ru 4 .1. 4 .6. and 3.1 (DPH 12.4, 14.0, and 9.2) along Sections A. B. and C.

a n d RB 2. 1. 2 .8. and 2 .3 (DPH 6.4. 8.4. and 7.0} along Section D. Cont:;-..,l ~amp les from t h e 18

shielded outer wall show an Increase In hardness after annealing. Consequently. actual val-ues of Irradiation hardening for Locations 3. 9. and 8 may be higher by Rs 1.4 (DPH 4.4) than ihe change In hardness shown In Fig. , 4.

Correlations between Increases in hardness and yield stress from radlaUrm hardening of pr:-ssure vessel steels Indicate that the Increase tn yield stress (In UPa) Is 3.5 times the increase In hardness (In DPH).9, to Based on tensile and Charpy- lmpact data. the increase 1n hardness for the Shippingport N3T should be ... 15 DPH. The measured values for Locations 3. 9, and 8 are in fair agreement with the estlmatton, Charpy-impact data for TL specimens from the irradiated and shielded walls of the NST, annealed for 2 h 3t 400°C, are given tn Table 8; Charpy transition curves are shown in Fig. 15, The results indicate complete recovery from r adiation- induced embrtttlement. I.e ..

e ~....... 1--~-,-,-.-..~,11 --~-.-.-..~~~~ 1-,-10~:-1 AnnHled at 400'C - O*-

3 AnnHled l 40O"C StG11on1 A, 8, & C -

... ..-..o . " s.t1ion o

0 0 8 Cl C

~

t:

.. -5 Location Cl t'JI - o- 3 Iii - g

- ..., _. e 6

10 100 10 100 Time (h) Time (h)

(a) (b}

Figure 14. Change tn hardness measured on (a) Sections A. B. and C and (b) Sectton D of the Irradiated wall after annealing at 400°C Temperature (°F} Temperature (' F)

50 0 50 100 ,,o 200 250 300 *50 0 &a 100 150 200 250 300 150 1$0 .

rrrtrr N

--E u

NST Irradiated. Wall Ut2-8 SIMI 80 10 2 N

E NST Shl1ld1d Wall A212-B s,..t

- to 10 ~

TL Orl*nlUon ~ Tl Orl*nlUon

- ,oo 60 ~ ""I

---;. ,oo 60 ~

cf 50 cJ" ()

50 u---

..r;

w Sh*i:lt0W1!

o ..~ ....i; 61,*lllld WaH 0 0

40 ..~

C 50 0

- - - - .. lO !

w w C

50 Al Lcabonl 1D .5 ti ti

§ l

0 "Alacto, s.r-_,,ce

-odlh114IIO"C 20 't;

,o Q,

§ !

Q, 0 ...,...... i**<OO"C 20 l

,o j 0 0 -0

50 0 50 100 150 0 50 1011 ,so Tempereture (°C) T11mper11ture ('C) lal (b)

Figure 15. Charpy transU.iDn curves for annealed TL specimens from (a) the Irradiated and (b) the shielded wall of the Shippingport NST 19

Table 8 . Cl.arpy-lmpact d.atafor matertalfrom the trradtated and shielded walls of the Shippingport NST after annealing a t 400UC Specimen Or\e;1- Test Tc-ml!* I meact Ener!!;l:'. Load (kNl ID Location Re&lon 3 l a lion 1*c1 (* f) µ/cm2) (ft*lbl Yield Maximum Irrad iated Inner Wall l03T-05A 3 TL 25 . 77 27.7 16.3 12.206 12.206 I03T-06A 3 0 TL 0 32 13.0 7.7 1 l.083 11.083 I09T-05A g I TL 55 13 1 so.:> 29.5 11.278 12.167 l09T* 06A 9 0 TL 40 104 49.7 29.3 11 .063 11.864 I02T* OSA 2 I TL 70 158 56.3 33.2 10. 155 11.229 I02T-06A 2 0 TL 10 50 18.0 10.6 10.194 10. 194 IOBT-05A 8 I TL 25 77 24.0 14.2 12.538 12.538 IOBT-06A 8 0 TL 40 104 44.7 26.4 11.219 12.245 IJ4T-0 1A 14 I TL 120 248 66.2 39. 1 9.354 I 1.766 I 15T*02A 15 0 TL 90 194 e-3.0 37.2 10.116 l 1.805 Shielded Outer Wall 002T* 07A 2 I TL 25 77 28.5 16.8 12.401 12.401 002T-05A 2 I TL 55 131 55.8 33.0 10.956 12.459 002T-08A 2 0 TL 0 32 15.7 9.3 l 1.014 I 1.014 002T* 06A 2 0 TL 90 194 72.1 42.5 )0.448 12.596 003T* 05A 3 I TL 10 so 16.0 9 .4 ]0.800 10.800 003T-07A 3 I TL 40 104 48.5 28.6 11.171 12.44 003T*08A 3 0 TL 120 248 76.6 45.2 9.647 12.381 006T-07A 6 I TL 25 77 19.3 l 1.4 l 1.473 11.473 006T-06A 6 0 TL -20 -4 5.1 3.0 9.667 9.667 OOGT-OBA 6 0 TL 55 131 58.7 34.6 11.024 12.684 012T-07A 12 I TL 40 104 51.6 30.4 10.546 l 1.678 0I2T* OBA 12 0 TL 70 158 61.5 36.3 9.852 11.483 a I and O represent the Inner- and outer mm regions. respectively, across the wa ll thickness.

the transition curve of the annealed specimens from the Irradiated wall ts identical to that of the shielded wall. Annealing has little or no effect on the transition curve of the shielded wall. These results confirm that the data for material from the shielded outer wall repre-sents baseline data for nonlrradiated material.

3.2 Weld Metal Weld samples were obtained from Location l on the shielded wall and three positions on the Irradiated wall (Locations 14 and 15 with ~3 x 101 7 n/cm 2 fluence and Location 13 with .. 2 x 1017 n/cm2 fiuence) . All of the welds were transverse to the plate rolling dlrec-Uon. Charpy- tmpact tes t specimens were mach ined perpendicular to the weld from the Inner and outer regtons of the wall. The elemental compositions of weld metals from dif-ferent locations are given In Tab le 1: only minor variallons In silicon and copper content were observed. Charpy data for the shielded- and irradlated-wal! welds are gtven In Tables 9 and 10. respectively.

Annealing studies were also conducted on weld metal specimens to obtain baseline data and help characterize the lrradlatlon embrlttlement of the welds. Charpy data for annealed weld metal specimens from Location l of the shielded wall and Locations 14 and 15 of the irradiated *,..au afe also given tn Tables 9 a nd 10. The results indicate that the Charpy tran-sition curves fo;* the annealed specimens do not always represent the baseline Impact prop -

20

Table 9. Charpy cmpact data for the ~hielded outer-wall weld of the Shippingport NST Speclmen Orl en - Test Tem e. lmEacl En er~ Load lkNJ IC Location Resion a tali on 1*c1 ("fl IJ/cm2] lft-lbl Yield M a ximum Reactor Service 001R--03W I LT 0 32 55.6 32.8 12.333 13.163 OOl R- 1l W I LT 20 68 95.8 56.5 12.82 1 15.028 001R-05W I LT 40 104 157.8 93. l 13.3 19 16.092 001R-07W I LT 70 158 172.1 101 .5 13.465 16.482 001R-09W I LT 120 248 165.8 97.8 12.879 15.379 OOlR- 12\V 0 LT -40 -40 29.2 17.2 15.526 16.082 OOIR-02W 0 LT 10 50 81.5 48. l 13.797 16.209 OOIR-04W 0 LT 25 77 123,8 73.0 13.621 16.456 CXII R-06W 0 LT 55 131 194.4 1 14 .7 13.309 17.02 001R-C8W 0 LT 90 194 190.4 112.3 12 .557 16. I 21 Annealed 001R- 13WA LT -20 -4 45.9 27. l 14.53 16.775 OOIR-!SWA I LT 10 50 104.6 61.7 13.944 17.264 001R-17WA I LT 40 104 151.8 89.6 13.465 16.4 92 OO I R- 14WA 0 LT C 32 150.2 88.6 14 .051 17.547 OO I R- 16WA 0 LT 25 77 255.0 150.5 12.5 18 17.191 OO I R- 18WA 0 LT 55 131 254.2 150.0 13 .182 17.3 12 a I and O represent the Inner- and outer- 10-mm regions. respectively, across the wall. thickn ess.

Table 10. Charpy impact data for the irradiated inner-wall weld of the Shippingport NST Specimen Location Region"' fluenceb Ori en- ---1!!t Temp . lmeacl Eners;t: Load (kNJ ID (n/cm21 talion (°CI (

0 fl (J/cm2) (ft-lb) Yield Maximum l13R-07W 13 l 2xi0I 7 LT -20 -4 19.8 11.7 16.502 16.502 l13R-OIW 13 1 LT 10 50 51.6 30.4 15.844 18.669 11 3R-03W 13 I LT 40 104 134.7 79.5 15.31 l 18.084 113R-05W 13 I LT 70 !58 178.:1 105.:. 14.451 17.3 12 113R-08W 13 0 LT 10 50 136.5 80.5 15.428 18.~,2 113R-02W 13 0 LT 25 n 122.7 72.4 15.0 18 18.300 11 3R-04W 13 0 LT 55 131 174.6 10::..0 14.647 17.928 11 3R-06W 13 0 LT 90 194 188.4 111.2 13.613 17. 107 114R-07W 14 3x 10* 7 LT -40 -40 14. 1 8.3 17.81 17.81 114R-01 W 14 LT 0 32 38.0 22.4 16.6 18 .035 114R-09W 14 l LT M;;:) 77 92.6 54.6 16.629 19.666

[l4R-05W 14 I LT 40 104 66.2 39. l 15.652 17.81 114 R* i l W 14 I LT 55 i31 132.8 78.4 16.0 14 18.914 l14R- l 3W 14 I LT 70 158 140.3 82.8 15.233 16.21 I Jl 4R-03W 14 I LT 90 194 153.5 90.6 14.432 16.58 ll 4R-08W 14 0 LT - 10 14 63.3 37.4 15.662 18 .543 ll4R- lOW 14 0 LT 10 50 90.6 53.5 15.8 18 19.021 l 14R-06W 14 0 LT 20 68 95 4 56.3 15.71 I 17.859

!14R-02W 14 0 LT 40 104 185.9 109.7 14.783 17 .762 114R* 12W 14 0 LT 40 104 150.8 89.0 15.428 16.60 1 114R-04W 14 0 r,T 70 158 176.3 104.0 13.8 17 17.352 11 5R--07W 15 I 3xio 17 LT - 10 14 33.9 20.0 16.287 17 .625 115R-01W 15 I LT 25 77 77.2 45.6 14.92 18.201 1l 5R--03W 15 I LT 55 131 171 .3 lOI. I 13.993 17 .791 11 5R--05W ts I LT 120 248 189.8 112.0 13.27 16. 18 115R-08W 15 0 LT -30 -22 16.2 9.6 16.883 16.883 115R-02W 15 0 LT 10 50 139.7 82.4 15.389 18.23 ll5R-06W 15 0 LT 90 194 199.7 117.8 13.582 16.668 IISR-04W 15 0 LT 120 248 215.0 126.£1 12.655 15584 21

Table 10. (Contd.)

Specimen Location Reglona F'luenceb Orlen- Test Teme- lme:ct EnerQ: L.oad (kNJ ID (n/cm2) talion i*c1 ("Fl µ/cm2l (ft-lb) Yield Maximum Annealed 1!4R-15WA 14 I 3xlo 17 LT 0 3:2 69.6 41.l 16. 15 19.256

!14R*17WA 14 I LT ~5 n 130.7 77. l 14.696 18.406 I 14R-l9WA 14 I LT 70 158 177.4 104.7 13.436 17.303 114R-14WA 14 0 LT -40 -40 51.5 30.4 15.106 19.09 114R- 16WA 14 0 LT 10 50 176.4 104. l 14.451 18.728 114R-18WA 14 0 LT 40 104 188.3 111. I 13.895 18.?.3 1l4R-20WA 14 0 LT 90 194 196.7 1161 12. 733 17.478 I15R-09WA 15 l 3xl0 17 LT -20 -4 136.8 80.7 14.93 18.855 115R- l lWA 15 I LT .5 59 156.l 92.1 13.69 18.025 115R-13WA 15 I LT 55 131 180.9 1067 13.455 17.605 115R-10WA 15 0 LT - JO 14 123.7 73.0 15.34 19.021 115R- 12WA 15 0 LT 20 68 163.2 96.3 14.764 18.543 115R-14WA 15 0 t,T 120 248 195.8 1155 13.036 16.824 a I and O represent the Lnner- .i.nd outer- I 0-mm regions. resi:,ecUvely. across the wall. thickness.

b Repre!'oent the values at the tnner surface of the wall. Flucnce for the outer-region samples Is estimated to be a factor of 1.5 lower than that for tnncr- reg!on samples.

Temper*ture (°F)

50 50 100 150 200 250 300 NST Shleld.d-Wall Weld LT Orkn1a1lon

Figure 16.

> Cha.rpy transition curves for weld metal 0 100 t,)

specimens from the tnner- and outer ~

Ill mm reg!Dns of the shielded u.,all of the

.so ifi Loon0111 0 IMe,RtrgOn Shippingport NST

OUIOr Fleg,on CIDM<I Symooll: Ar,n.aled 2 o a ,6QO*C i

50 0 50 100 Temptralure (°C) ert1es of non1rrad1ated weld metal. For example. anneallng has little or no effect on weld specimens from the Inner region of the shielded- wall weld. whereas the Impact energy of weld specimen<, from the o* ,ter region ~ncreased significantly after annealing (Fig. 16). The lncrei.'\se in Impact energy .s most likely due to microstn.ictural changes. The annealing data. therefore. cannot be used to characterize irradiation embrittlement of the weld s:Jectmens.

Charpy transition curves of the weld metal specimens, with and without anneaUng for 2 h at 400°C. are shown In Figs. 16 and 17. The best-fit curves to Eq. l are also shown In the figures: values of the constants 1n Eq. 1. as well as USE and C'IT at the 20.4-J (15 ft -lb) and 41-J (30 ft -lb) level. are given 1n Table 11. The Impact strength of the shielded-wall weld. shown in Fig. 16. is significantly higher than that of the base metal. The 41-J (30 ft *lbl CTI and USE of the weld are 7°C (28°F) and 184 J/cm 2 (109 ft -lb), respectively.

Posltlon through the thickness of the weld, Le .. inner and outer regions, has no effect on the transition curve.

22

Table l l . Values of constants tn Eq. l . CIT, and USE for weld metaljrom the shtelded and iITadiated walls of the Shippingport NST Constanls Sample Sample Orient- Ko B C D 4 1-J CTT USE Location Regio n atlon J /cm2 J/cm2 *c oc *c !"Fl Jt:cm2 !ft-lbl Shielded Outer Wall Inner & Outer LT 20.0 81.8 16.0 31.1 .7 (19) 184 (109i Irradiated Inner Wall 14 Inner LT 15.0 71.7 34.2 37.0 14 (57) 158 (93) 14 Outer LT 1s .oa 89.7 15.4 37.l - 11 (12) 194 (114)

13. 15 Inner LT 24.4 82.3 32. l 23.8 13 (55) 189 (I 12j 13, 15 Outer LT 15.ob 9 1.1 4.4 37.l -22 (*81 197 (116)

Annealed Irradiated Inner Wall 14 Outerb LT 20.0 82.2 -23.8 28.0 -45 (-49) 184 il09l 13 inner & Outer LT 20.0 82.2 -23.8 28.0 -45 (-49) 184 (109) a Value of Ko Is assumed lo be the same as tha t for locatio n 14 Inner region.

b Results for locaUon 14 Inner region agree well wtlh those for th e shielded ou ter wall.

Temperature (' f ) T*mp*ratur* ("F)

50 0 50 100 HO 200 250 300 .so 0 50 ,oo 150 200 250 300 300 - r'--,....-,+,....,.-r+,l'"T"l-r--r-"""'"h..-,..~......,.+r JOO-,L-,.-.-,4-,."?T"..-h...,...-.-1-...,......-11-r-,....,..,I....,~+,--~-',

NST Irradiated- Wall Weld LT Orl1nt1t1on 150 ~

..E NST Irradiated- Wall Weld LT Orlent11lon

~

Inner R1g10n 100 <J_

cl 200 Inner R.g,on CII

!:!' iii w

C w 100 50 1.ocahon u

0 ,.

i s

g_

.E 0 +-'---'-.+-.I-I...L.........+-L...l.-'--'--+---'-....,_.L...+ 0 0 50 t 00 ISO 0 so ,oo 1SO Temperature (°C) T*mperature ("C)

(a) (b)

Temperature (°F) Temperature ( 0 F)

50 0 50 100 150 200 250 300 .50 0 50 100 150 200 250 300 300 - r'-..........-h-rr..+...,.,-,-.,..,.....,..f-.-......-+-,.,~+.-~..-1 300 - ......~.+..~,+,,~ ,- 4.-~.+.~..-4--~...-1-................

'"e...

NST Irradiated-Wall Weld LT Orl* nt*llon ,so ~

.. E NST lrnadlattd- Wall Weld LT Crlan1111lon cl 200 Ovttr ~g,cn 100 u.

~

cl 200 Ou1er Aeooon

/ -.- . 100 U

,.:; ,.:; -1"<1 ,

f

~

CII t ...e CII

'-../ . /

c:: c:: c:: ~ - 0

-0 W 100 w w 10 0 50 / /' Lce4110n 50

-+-

l l oca1ion ti ll / SNtldod-Wal 13 l

7 0

~ 14 *c.. / / WolO ..... IS 0

~

Clol<<l Symboa :

+-.....,L..J....L.-+---'-'--'-+-'---'-~ -

2 h ao .OOOC

..L..L....I...+ 0

~ l 0 _ ...................-1-'L...:.---..._...........___......I-'-_._..._._... 0

.50 0 so 100 150 *50 0 50 100 150 Temptratur, ('C} Temperature (' C)

(c l (d)

Figure 17. Charpy t.Tansitton curves for weld metal specimens from (a) and (.'"JJ inner region and (c) and (d) outer region of the iITadiated wall of the Shippingport NST 23

Specimens from the Irradiated-wall weld show a strong efkct of location across the thickness of the wall. The transition curves for the Inner region of the Irradiated-wall welds are shown In Figs. 17a and 17b. The 41~ (30 ft*lb) C'IT is 14°C (57°F) for the weld at Location 14 and 13°C (55°Fl for welds at Locations 13 and 15, i.e., a shift of 21 and 20°C (38 and 36°F). respectively. relative to the data for the shle~ded-wall weld. Weld speci-mens from the outer regions of the Irradiated wall show no embrlttlement relative to the shielded-wall weld, The Charpy transition curve of the outer- region specimens from Location 14 Is comparable (Figs. 17c) and that for specimens from Locations 13 and 15 is slightly higher than that of the shielded-wall weld (Figs. 17d). Charpy data for annealed weld metal specimens from the irradiated wall seem to represent recovery from Irradiation embrlttlement as well as the effect of thermal aging caused by mlcrostructural lnstablllties.

The results for annealed specimens from the inner region of Location 14 alone indicate re-covery from Irradiation embrittlement. Le., the Impact energies compare well with those for the shielded-wall weld. Specimens from Inner and outer regions of Location 15 and the outer region of Location 14 show a large increase in lmpAct energy after annealing.

4 Discussion 4.1 HFIR Surveillance Data The Charpy-impact data for the Shippingport NST indicate that the shlfl in CTr Is not as severe as would be expected on the basis of the changes seen In HFIR sutvelllance sam-ples. A summary of Charpy- impact and tens ile data for HFIR surveillance samples. for samples irradiated in the ORF<.. and for the Shippingport NST are given In Table 12. In Fig. 18, Charpy transition curves for LT specimens from the shielded and irradiated walls of the Shippingport NST are compared with res ults for nonlrradlated and irradiated HFIR s urveillance samples of A212-B steel. The results for the nonlrradlated materials. Fig. 18a.

indlC'::ite that the impact strength of the HFIR material ls significantly higher than that for the Shippingport NST; the 20.4-J (15 ft -lb) CTr for the HFIR materia l ls 35°C (63°F) lower T*mpeniture {°F) T*mptrature ("F) 150

SO 0 50 lOO lSO 200 250 300 ISO -

-~ 0 $0 100 150 200 250 300 E

B0 N E

HFIA Surnlll1nce

,\212- B SIHI

o :;;

u 70 ~

~ I C.JIO-,~ 70 d'

100 60 : , -

c.J' d' 100 LT o,11nt1llon ltr1dlat1II I .*

0 111,._"81A,~IH:J H11A.Hi&>.1,..,.. 80 ~

cl' I'

50 50 E." 40 2l i I

-- -- u a,

C

/ " " " Sr-.lppi"lP'nNST C

~onilrldl.l\ed / ;.

.!i

~

S'-ldodWaA C w 50 w 30 w 30 50 I S/'4)pr,gpo,,

t; 0 t; NST

I 20 ti , / lrradiot-11 20 l

II lmo,AeQic>n Q. 0 HB2 DrnpoUI II Q. ..

Q. : 1

.E

NOE Blodl 10

.E .E 10

~

0 0 0 0

-50 0 50 100 150 -50 0 50 100 150 Temperature ("CJ Tempertur* (°C)

(al (bl FY.gure 18. Comparison of Chnrpy transition curves for (a) nonlrra.d.tated and (b) UTadiated HF1R surveillance samples and material from the Shippingport NST 24

Table 12. Summary of Charpy-tmpacl and tensUe results for HF!R surueUlance tests. for samples irradiated in the ORR. andfor the Shtpptngport NST flux Fluence 41..,J 20.4 --.1 20.4..,J Yie ld Material (E> I Mev) (E > I Mev} dpa 41..,J CTI 6CTT CTT 6 CTT Stress 6Yleld n/crn2 -s n /cm2 *c 1*FJ *c c*FJ *c (*F) *c (°F) MPa MPa t lFIR Surve\llanee Al05-!I G-181. HB4) --02 (-80) 4.66x 1o8 3.44 X 1016 5.12 X l o-5 -52 (--62) 10(18) 4.89 X )()B 9.90 X )016 1.48 X l0-4 --46 {-50) 17 (30) 4.89 X )(.)8 2.31 X 10 17 3.44 X 10-4 -29 (- 20) 33(601 7.27 X Jofl 4.01 X )017 5.97 X J0-4 -27 (- 17) 35(63) 3 .35 X l ofl J.85 X J0l7 2. 70 X 10-4 -29 {-20) 33 (60)

A350-LF3 (HB2} -79 (-I 10) l.J Ix )()9 8.20 x 1016 1. 16 x lo-4 -64 (-84) 14 {26) l.i lx io9 2.26 X 1017 3.21 X lo-4 -50 (-5 8) 29 (52) l.l l X to9 5.26x 1017 7 .47 x lo 23 (-10) 56 (100)

I.II X io9 6. 14 X 1017 8.73 X 10 \3 (8) 66 (118)

A350-LF3 (HB3) ---62 (-80 )

1.29 X )()9 ,.55 X J016 J.36 X 10-4 -43 (-46) 19(34)

J.40 ll )()9 2.84 X 101 7 4 .09 X IQ--4 -29 (-20) 33 (60) 1.03 ll l cfo 4 .88 X 1017 7 .01 X 10-4 -8 117) 54 [97) l.29 x Io9 7. 12 X 1017 1.02 X J0-3 2 (35) 64 (115)

A212- 8

[HFIR Shell) 1 (34) -21 (- 5) 335 4.00 X 107 l.89 X [016 - 6 (42) 4 (8) -8 (17) 12 (22) 329 2.43 X loB l.15 X !017 l. 73 X lo-4 22 (72) 2 1 (38) 8 (46) 28 (51) 390 55 2.43 X ioB l.34x 1017 2.02x 10-4 39(102) 38(68) 20(68) 40 (73) 389 54 ORR Irrad iations A212-B (HrlR Shell} 9 .59 X 1012 2.43 X 10 l8 3.39 X IQ-3 56 431 96 A212- B (ECCR Shell} l.05 X 1013 l.54 X 1017 2 . 14 X l o-4 10 340 5 2 .00 X )Ql2 9 .80 X I0l8 2 .40 X IQ-2 103 524 177 Shippingport NST A2l2- B 31 (881 16(611 294 8 .90 X Jo8 2.67 X 1Ql7 7.J0x lQ-4 51 (1241 20(36) 39 {102) 23 (4ll 345" 51 1.33 X to9 4.00 X 1Q17 1.07 X 10*3 5 1 (1241 20 (361 39 [1021 23 [4Il 345"- 51 1.33 x Jo!} 4.00 X IQ17 1.07 X )0*3 58 (1 361 2 7 (48) 44 {1 11) 28 [50) 34sa 51 2.00x lo9 6.Wx 1017 1.6() X ) 0-3 58 (1361 27{48l 44 (1 11) 28 (50) 349 51 a Average value at all flucnce levels.

and the USE is =60 J /cm 2 {35 ft -lb) h igher than that for the Shipptngport NST. After irra-diation the shift in CTI is 40°C (73°F) for the H FIR matertal (= l. 3 x 1017 n/cm 2 fluence) and 28°C (50°F) for the NST material ("'6 x 1017 n /cm2) . Although the increase In C1T ts smaller. the actual C'IT of the Shippingport NST ts significantly htgher t han that of the HFIR A212-B steel. At the service temperature of 55°C (131°F), the impact energy of the Irradiated inner wall of t h e NST is very low, =40 J/cm2 (=24 ft,lb). Similar differences are a lso observed for the Charpy data on TL specimens.

The HFIR A2 l 2-8 steel is not 0nly to ugher than the Shippingport NST steel. It Is also stronger, i.e., its tensile s trength and h a rdness are higher than that of the NST material.

The yield stress and hardness for nonlrradlated m aterials are. respectively. 335 MPa

(=49 kst) and 172 DPH for the H FIR 1 and 294 MPa 143 ksl) and 156 DPH for the

Table 13. Chemical compositions (wt.%) offerritlc steels from the HFYR suroeillance program Element A212-B Al0S -11 A350- LF3 A350- LF3 Shell HBI. H84 HB2 I-1B3 C 0 .26 0 .24 0. 18 0.17 Mn 0.85 1.12 0.55 0.50 p 0 .006 0 .003 0.010 0 .007 s 0 .040 0 .020 0.020 0.010 St 0.29 0 .21 0.29 0 .27 Cl.I o. 15 0 .03 0.11 0.10 NI 2.208 0. 14 3.30 3.20 Cr 0.075 0.042 0 .000 0.080 0 0 .0024 0.0033 0 .0027 0 .0026 N 0 .0060 0.0063 0 .0090 0 .0083 Tl 0 .0 10 <0.001 <0.001 <0.001 V 0 .0005 0.0005 0.0010 <0.001 Zr <0.001 <0.00 I <0.001 <0.001 Mo 0 .02 0.07 0.03 0 .03 Al 0.07 0.06 0.08 0.08 Sn 0.02 <0.005 0.02 0 .02 8 <0.005 <0.005 <0.005 <0.005

Believed to be high: a no ther analysis showed 0 .09 Shippingport NST. The chemical compositions of the ferrttlc steels from HFIR surve1lta nce program are given tn Table 13. Except for the difference in copper content. elemental compositions of the A2 12-B steels from the HFIR pressure vessel and from the Shippingport NST are comparable. The differences In the transition curves of th e nonlr-radlated A212- B steels are most llkely due to mlcrostructural factors, e.g., amount and dis-tribution of inclusions.

The shifts in CIT of the m aterial from the Shippingport NST. HFIR surveillance sam-ples, and HFIR A212-B steel irradiated In the ORR are plotted as a function of neutron ex-posure in Fig. 19. The results for the Shippingport NST matenal are consistent with those from ORR-irradiated steel. The HFIR surveillance data follow a different trend: for a given fluence. the shifts In CIT are greater than those for the Shippingport NST material or the ORR-irradiated samples. The shifts in 41--J (30-ft-lb) CTI of A212- B steel from the HFIR surveillance program are not significantly different from th ose for Al05-11 or A350-LF3 steels.

4.2 Low-Temperature Irradiation Irradiation temperature ls an lmp01tant factor In radiation damage, particularly In the temperature range of operation of power reactors. At temperatures >232°C (>45O°F), pro-gressively lower embrittlement is observed with Increasing temperature.13-15 However, relatively llttle or no temperature effect is observed at 1rradlaUon temperatures <232°C

(<450°Fl. In Fig. 20. the resu lts for the Shippingport Nsr and for samples irradiated in the ORR are compared with data from test reactors3, t 4-20 and from Anny reactors s u rveillance pr ogram21,22 The shifts in CIT for the NST A212-B and HFIR A2 12- B Irradiated In the ORR are consistent with these results and represent the tall of the trend band that de-scribes the Increase in C1T of various steels Irradiated at <232°C. An upper bound curve for low- temperature 1rrad1at1on Is expressed as 26

150

ORR ltradiat-on 50*c HFIR Surveilanca 50"C 250

'O '12 _ 'O 13 n1tm2, 2i108 111cm2, C A212-B ... A 10S*II A212-B *.. A3SO-LF3 0

U 100 0

Sh~ lngpon NST 55°C

2J1o9 n1cref- ,

A2 12--8 200 IL 0

Figure 19.

Shifts in C1T with neutronjluence for the Shippingport NST material.

6

<l 0

A212* B WeldA212-B 1 50 HFIR surveillance samples, and HF1R A212-B ilTadcated in the 7

... 100 ORR. The solid lines represent CIT

~

50 sh1fls as afunctiDn of the square 0

M root of total jluence.

1016 1017 1018 101 G 1020 Neutron Fluence (E >1 MeV) (n/cm2) 500 250

A20l

A302*B *

(.) 200 0

0 "212*8 ORR l~dlation so*c 1l 12 _ 'O 13 n1tm2-,

I>

SM-IA Sur;e{l,1n~

27 - 264'C

'O 'O - 1011 Nc11121 400

-LI.

0 Figure 20.

151 A.212*8 t:

(.) 150 c A2t2*B 11 A212*B 300 t:

(.) Comparison of C'IT shifts for the

<I Sl,lpplnpn NST ss*c <l Shippingport NST and for samples 7

100 21109 n11;m 21 o

x A212*B Weld A212-B 200

.D 0

C')

trradtated tn the ORR with data from test reactors 50 100 1016 1017 1018 1019 1020 Neutron Fluence (E >1 MeV) (n/cm2) t.CIT = 15 + 150£10.42 + 0.22log0, ( 4) and the lower bound cuJVe as t.CIT = -50 + 15Qft0.42 + 0.22log0, ( 5) where the C1T shift is in °C and fluence f Is In 1019 n/cm 2 .

Swedish data23 on the effect of irradiation temperature on embrittlement of ferritJc steels (including A212- B steel) Indicate that embrlttlement is relatively Insensitive to tem-perature between 110 and 232°C (230 and 450°F) and decreases at lower and higher tem-peratures. The data from irradiation of variou s steels and welds3,2l at "'.93"C (<200"F) fall 27

250

Test React:>r O.ia <93'C

,i 13 n1,,,,2..

~,

Swedish O*t* <93"C I 400 II A212- B 2 00 6 "212*8 l!l A302-B /

0 0

- , so

)(

A302*B SA:136 Weld A212*B Weld A:!02-8 Trend Band /

Ii 1 ff

, 300 --

~

0 Figure 21.

Comparison of CTI' shifts for the

~

0 OAR lr,adi100n so*c I) 12 _ 1013 "'""2., <232'C~* */'il, ~

0 Shippingport NST and ORR-

<l 0 "21 2*8 , - 200 <l irradiated specimens with test

? 100 o "212-B ACTT

1;g050/ Ill :e

'IS' She>pil\gpo~ NSi we 2,,o9 rvem2~

0 "212* 8 ,

=

0 M

reactor d.atafor ITradCattons at

<93°C. The solid line ts the best-100 .fit curve.

50

Weld A212*B

'I 0 ~-L..I......U~ _ __._..L...LJL...u.JJ~............u..Jw.lj.---l....1.....L.U.l.uj*

10 1 e 10 1 1 101 a 101 9 102 o Neutron Fluence (E >1 MeV) ( n/cm2) between the center a nd lower- bound c u rve of the <232°C trend band, Fig. 21 . Swedish data from lrradtauons .* 110- 230°C fo llow the u ppe r bo und of the trend band.

Embrlttlement of ferritlc steels a t <93°C a ppear to b e better represented by the average trend for low temperature irradiations. A power- law best flt of the data for va rious steels Irradiated at <232°C yields tiCTr :::: 12Of0,50, (6) where t he crr shift ls 1n °C a nd fi uence f Is tn 1019 n/ cm2.

The data In Fig. 20 indicate that variations in c hemical composlllon of steel. e.g.. Cu or NI conten ts, have little or n o effect on Irradiation embr1ttlement a t temperatures <232°C.

For example, the Increa ses In CTI for ASTM reference A212- B and A302- B steels Irradi-ated at <149°C are comparable. Fig. 20. Compositional effects are observed for these steels when they are Irradiated In the same facilities al higher te,nperatures.20 In Ftg. 22, the s hifts In CTI for the ASTM reference materials Irradiated at 277-310°C (530-590°F) are compared with the <232°C trend band. Th e resu lts for A302-B stee} from the Yankee-Rowe surveillance program are also Included In the Fig. 22.20 For a given fluence level, the Increase In CIT for A2 l 2-B steel (with 0 .26 wt.% Cu and 0 .28 wt.% NI) ls greater than that for A302- B steel (with 0 .20 wt.% Cu and 0 . 18 wt.% NI).

Surveillance data for Army reactors Indicate no s tgntflcant difference In embrtttlement between A212-B and A350 steels. I 6.21.22 The results for A350 steels In Anny reactors, sh own in Fig. 23, are within the trend band for Irradiations at <232°C, although they are close to the upper-bound curve. HFIR s urveillance data for A350-LF3 and Al05-II steels do not follow the trend for A350 Army reactor steels: they show greater embrlttlement than the <232°C trend b and. These results Indicate that the greater embrittlement of HFIR sutvelllance samples relative to the Shippingport NST ls primarily due to factors other than material and compositional differences.

28

250

-.sn1 Ref11911UO Mtltf\al 400 211-3,o*c 200 *,. A302-B A212-B 0 300 u..

0 0

....., 150 Figure 22.

I=0

I=0 Shifts tn CTI' with neutronjlue nce

<l 200 <l Jor ASTM reference A21 2- B and

..."7 100

.0 0

M A302-B steels trradlated at 277-310°c 100 50 1016 101 7 101 8 10111 1020 Neutron Fluence (E >1 MeV) (n/cm2) 300 500 250 *

II I> A212- B 400

.--.. ORl'I lrr1~11110" 50' C u:- Figure 23.

-I:

U 0

200 11e - 1olln,'Cln2i 0

C':

A2n,B A212*0 I

I I

I

300 I:

0 Comparison of CIT s h iftsfor HFIR surueUlance samples. L.h e

()

<l 150 S~l)l)~ gl)Oll NST

'l,,1,/ Ncm2 1 ss*c I

Ai

I ,,

u

<l Sh(pptngpo rt NST. and ORR-

.tl trradla.ted samples with data.from 7 A212*B 0

W81dA212,B

rl TreMBand . 200

100 " 1 <232' C test and Army reactors 0

M 100 50 0

1016 1 017 1Q1b 10111 102 0 Neutron Fluence (E >1 MeV) (n/ cm 2 )

To account for differences tn the spectra of the various Irradiation facilities. the In -

crease In CIT for HFIR surveillanc e steels. ORR-Irradi ated samples. and the Shippingpo rt NST. as well a s test reactor data* at <93°C. are plotted In Fig. 24 as a function of displace-ment per at om (dpa) for E >0. 1 MeV rather than fl uence. The difference still exists be-tween HFIR surveillanc e data and results from ORR Irradiation . the Shippingpo rt NST. or test reactors.

f rom Ref. 4 ,where the test rcaclor trend curve was ub lalncd assuming that the calculated spectrum for O RR was appropriate for lest reactor data.

29

H~IR SuMl~illnce S0°C Sh;ppingpo,t NST WC 250 2,10B n/tm2'1 21109 ivan2,.

A10S.II o A212-D A:l!i0-LF3 >< Weld A2t 2*8 1 0 100 0

11 1\212-B ORR Irradiltion 50"C

.., e _ 1013 n1tm2,.

' 200 IL 0 Figure 24.

- 150 ~

Comparison ofshljls In CTI' wlth

~<I o o

A212*B A212*8 0

<:I dpafor HFIR suroeUlance samples, the Shippingport NST,

? 100

ei: and ORR irradtatton with data

50 0

C")

from test reactor 50 t.CTT

59010.48 0_..,__._.._,_...._..-4--_._.J--L~-4-_.,__........u..u-4-_._.L...U_ 0 10*5 1 o** 1 o-:3 10-2 dpa (E >0.1 MeV}

4.3 Effect of Fluence Aate Radia tion embrtltlement depends on the fraction of point defects that avoid recombi-nation and s urvive to form clusters, rather than on point- defect production. which Is con-ventionally represented by fast neutron Iluence or by dp a , Radia tion effects under dUTerent irradiation conditions will correlate with fast fluence or dpa if the relative fraction of point defects that avoid recombination remains the same for different Irradia tion condJUons.

However. the overall rate of radiation h ardening and embrlttlement wlll be accelerated by a ny mechanism that decreases the rate of point- defect recombination relative to their pro-duction rate. Two such mechanism, b ased on bulk recombina tion or rate effect and on in-cascade r ecombination or spectral effects, have been considered to explain the greater em-brlltlemcnt of HFIR surveillance s amples .24 Lower displacement rates decrease the frac-tion of bu1k recombination. thereby a larger fraction of point defects survive for fonnauon of clusters. The rate effect m ay contribute to accelerated embrlttlement of HFIR samples be -

cause the fluence rate Is nearly 5 orders of magnitude lower for HFIR than for the test reac-tors. The spectral effect considers that, for Irradiations producing small cascades. e.g.. re -

coils from thermal neutron capture, a smaller fraction of point defects recombine In the cascade and the availa ble number of defects per displacement Is larger than for lrrodlatlons that produce large cascades. Thus, the contribution to radiation embrtttlement from ther-mal neutrons ts larger per dpa than that from fast neutrons.

The greater embrtttlement of HFIR surveillance samples relative to that of specimens Irradiated In the ORR a nd test reactors has been evaluated on the basis of a flucnce- rate effect. 4 The rate effect was established from specimens irradiated at a fiuence rate of s l x 101° n/cm2 ,s: recent data Indicate that at .. 93°c !-200°F) there Is essentially no rate effect In the fast flux range of l x 10 10-3 x 10 13 n/cm 2 -s. 2!.". At temperatures <93°C (200°F),

vacancies are relatively Immobile, which reduces the Importance of rate effect. 24 HFIR smvefllance data for the A350-LF3 and A2 l 2- B steels, corresponding to flux values of 2.4 x 10B and 1.2 x l09 n/cm2,s, respectively. were used to obtain plots of dpa vs, dpa rate for 30

10-2 ____,,_.,...,..,...,.,..,._--,.-,......-.............+-.....---.-T"T......- r t - -...

.:*: .. *********........:***~ --:--~**?-~-: ... ......* ..... .... *...* ..

HFIA Surveillance :::;::: s11ip~~pc;~ NST. :::t::t::;::; 11 1

1:, A212-B ********* o A212 * * ...

AJ50--lF3 , . , ,. 78

.t. A10S--II ***********,***, *-;--*** .$

....*~----- i, , .. t*. .:..l .'..U.:............ -- ~~:....

.,.. ::::~::::**~~*:::1n ::::**** =...r--= *J3 ::: .

ci ::::L : 1-c> ** : 22 :::: Ftgure 25.

A

- - - - i***..t**~** -~--! rt*.. Plots of dpa (E > 0.1 Me\lJ vs.

w

-;10-4 ::r: ::: :_it*:: dpa rale for specific values of CIT shifts (Ref 4)

Q.

1:~:::;....,..

..::r:: 44 ,

- .:;::::a r:::::

,:*rn:mr,:::

- - - - ~::::: 33 . :~----~:: : ~~-~~:r::r:\t:~r:::::::::

. ...~--!*.;.~-~(------.)....-4-).)--~-->-~;~--...... .

1 o- s ---+-_ __.__...._......._........-4--_...__.&.....J'._._ : ;_,_:4 i ..... H_ __._ : __,_:_.'_,:u1.....,.....' l+' _ _._

1 o-14 10*13 10-12 dpa Rate (dpa/s) speclflc CTI shifts In Fig. 25. The irradiation conditions (I.e .. dpa vs. dpa/s) for the Shippingport NST steel are also shown In Fig. 25 . The fluence rate for the Shippingport NST is comparable lo the rates for HFIR A350-LF3 or AlOS-11 steels. t.e .. "'l x 109 n/cm 2 *s.

If fluence rate has a large effect on the CTI shifts. the values for the Shippingport NST should be close to those for A350-LF3 and Al05--ll steels. The predicted increase In C'IT of 45-60°C (8 l- l08°FJ from Fig. 25 for the Shippingport NST. Is higher by a factor of ..z than the measured shifts in C'IT. The pr edicted CIT shifts for Al05- II steel also show poor agreement with the experimental data (Table 12). The results Indicate that fiuence rate has no effect on the shlft In CIT at values as low as 2 x 109 n/cm2*s . The greater em-brlttlement of HFIR sutvelllance samples Is most likely due lo spectral effects.

4.4 Spectral Effects The HFlR spectrum is highly thermallzed. with thermal neutrons accounting for 96%

of the total flux in the surveillance position. ln the Shippingport and most test reactors of Interest, the thermal neutron flux ts a much smaller fraction of the total flux. Thermal neu-trons cannot cause damage by elastlc scattering since the minimum neutron energy re-quired to displace an atom is above 0. 5 keV. Damage associated with thermal neutrons is caused by recoil events resulting from neutron capture. I.e .. from the (n. y) reaction and for some cases the (n. al reaction. As was discussed in Section 4.3. radiation effects depend on the fraction of point defects tha t s urvive rather than on the production of defects. Slow neutrons in an energy range <0. l MeV Increase the available point defects per unit dis-placement. resulting in enhanced cluster formation a nd embr1ltlement. Contributions of slow neutrons are not accounted for tn conventional methods for assessing damage in terms of fast-neutron flu ence (E > l MeV) or dpa (E >0. l MeV).

Alternative damage models have b een used to determine the fraction of stable point defects produced by a given lrractiatlon. Calculations based on damage eflkiency, I.e., the probablllty of forming a stable defect. were recently completed. to explain the discrepancy 31

between the shifts In err of HFIR surveillance and ORR-Irradiated samples. and the Shippingport NST.* The models proposed by Doran26 and Wiederslch27 were used for damage efficiency. The procedure for calculating dpa was modified to Include damage effi-ciency. and the relative changes In damage rate. rather than absolute damage rates, were dete-nnlned for the neutron flux spectra for Shippingport, HFIR. and ORR.

The results indicate that the contributions from themial neutrons for HFIR ORR. and Shippingport samples are. respectively. 29, 2.5, and 0 .4% of the total damage. Relative changes in damage rate from the Wlederslch model appear to be of the right order to ex-plain the observed differences In embrllllement between samples from these three facili-ties. The ratio of modified damage rates Is 3.3- 5. 7 for HFIR/Shlpplngport. 4.4 for HFIR/ORR. and 0.8- 1.3 for ORR/Shippingport. The rela tive changes from the Doran model. e.g .. 1.2-1.6. are not sufficient to explain the discrepancies.

A similar approach has also been used to show that spectral effects may dominate the accelerated embrlltlement of HF!R surveillance samples.24 The shlfls in CTI for l1FIR surveillance and ORR-Irradiated samples fa ll along a s ingle curve when plotted as a function of thermal neutron flu ence (E <0.4 eV). Similar correspondence ls observed for the In-creases In tens ile stress for HFIR surveillance and out-of-core ORR-Irradiated samples2 plotted as a function of thermal neutron fluence.

Other analytical models of freely mlgrallng Interstitials and vacancies also show the importance of spectral effects and ar e consistent with the differences In property data from different facllltles. 28 The best correlation of the data from HFIR sur"elllance and Irradia-tions In the ORR, Omega West reactor. and the Rotating Target Neutron Source ls obtained when Increases In yield stress are compared on the basis of freely migrating Inter stitial defects. which are believed to better represent the defects parllctpallng In radiation hardening.

These results Indicate that Ir. neutron environments with high thermal-to-fast ratio and low temperatures. c urrent measures of radiation damage, such as fast nuence a nd dpa.

are inadequate. Alternative measures of damage assessment must be considered for cases of a softened neutron spectrum such as HFIR.

5 Conclusions Characterization of material from the Shtpplngport NST Indicates that the embrlttle-ment of this A2 l 2 Grade B steel In a low- temperature. low-flux environment ls consistent with the trend band for Irradiations at <232°C (<450°f) and shows good agreement with data from test and Anny reactors. The shifts In err are between 23 and 28°C (41 and 50°F). The NST weld metal ls significantly tougher than the plate material; shift In err Is

~20°c (36°FJ. These shifts are significantly lower than those expected on the basis of re-sults obtained from the HFIR survelllance samples. The results Indicate that fluence rate does not affect radlatlon embriltlement at rates as low as 2 x 108 n/cm2,s and the low oper-ating temperatures of the Shippingport NST, Le., 55°C (I30°F). The accelerated embrlt-

Greenwood, L.. "Damage Calculations for Shippingport, !!FIR and ORR." memo to W. J. Shack. October 1989.

32

tlement of HFIR survelllance samples are most likely due to the contribution of thermal neutrons.

Acknowl~dgments This work was supported by the nmce of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission and the Office of LWR Safety and Technology, U.S. Department of Energy. Thr authors are grateful to W. F. Burke and G. M. Dragel for thelr contributions to the expertmental effort and to A. Sather fo;r conducting the mechanical tests.

References

l. R. K. Nanstad. K. Farrell, D. N. BraskJ. and W. R. Corwin, MAccelerated Neutron Embrttllement of Ferrltk' Steels at Low Fluence: Flux and Spectrum Effects; J. NucL Maier., 158. 1-6 (19881.

2. R. D. Cheverton. J. G. Merkle. and R. K. Nanstad. Evaluation of HFlR Pressure Vessel Integrity Considering Radiation EmbriWement. ORNL/TM-10444, Oak Ridge NaUonaI Laboratory, Oak RJdge. TN (Aprll 1988).

3. J. R. Hawthorne. MStudles of Radiation Effe1;ts and Recovery of Notch Ductility of Pressure Vessel Steels.* in Symp. on Steels for Reactor Pressure Cfrcuit.s. British Nuclear Energy Conference. Iron and Sleel Institute, London. 343--369 (1961).

4. R. D. Cheverton. F. B. Kam. R. K. Nanstad, and G. C. Robinson. *An Embrittlemenl Rate Effect Deduced from HFIR that may Impact LWR Vessel Support Life Expectancy,"

Nucl. Eng. and Des .. 117. 349-355 (1989).

5. S. T. Rosinski. 0. K. Chopra, and W. J. Shack, "Shippingport Neutron Shield Tank Sampling and Analysis Program.* In Proc. 1989 ASME Pressure Vessel and Piping Conference, D. L. Marriott. T. R. Mager, and W. H. Bamford. eds .. Vol. 170, ASME, New York. pp 109--113 (1989).

6. S. T. Rosinski, W. J. Shack. and 0. K. Chopra. "Irradiation Embrtttlement Investigation of the Shippingport Station Neutron Shield Tank,w in Proc. Fourth Int. Symp. on Environmental Degradation of Materials f.n Nu.clear Power Systems*- Water Reactors, D.

Cublcciotu, ed .. NACE, Houston. TX. pp 2.79-2.89 (1990).

7. 0. K. Chopra. S. T. Rusinski, and W. J. Shack, "Embrtltlement of the Shippingport Reactor Neutron Shield Tank," in Fatigu.e. Degradarton. and Fracture - 1990, W. H.

Bamford, C. Becht, S. Bh..:.ndari. J. D. Gilman. L.A. James. and M. Prager, eds., MPC Vol.

30, PVP Vol. 195, ASME, New York, pp. 157-168 (1990].

8. W. L. Server. "Impact Three-PoJnt Bend Testing for Notched and Precracked Specimens.~ J. Test. and Eval .. 6, 29 (1978).

33

9. G. R. Odette, and G. E. Lucas. Irrad~atton EmbrUtlement of LWR Pressure Vessel Steels.

EPRI Report NP-6114. Electric Power Research Institute, Palo Alto, CA (January 1989).

10. G. R. Odette, and G. E. Lucas. -1rradlatlon Embrittlement of Reactor Pressure Vessel Steels: Mechanisms. Models and Data Correlation." ASTM STP-909, American Society for Testing and Materials. Philadelphia. 206 (1986).

11. 0. Sines and R. Carlson. "Hardness Measurements for Detenntnauon of Residual Stresses: ASTM Bulletin. 180. 3~37 (TP31-TP33) (1952).

12. A. Dervlshyan. -rhe Stress-Hardness Relation,* ASTM Bulletin. 215. 71-75 (TP127-TP131) (1956).

13. R W. Nichols and D. R Harries. "Brittle Fracture and Irradiation Effects In Ferritlc Pressure Vessel Steels," tn Radiation Effects on Met.als and Neutron Dosimetry, ASTM STP 341. American Society for Testing and Materials, Philadelphia, 162-198 (1963).

14. G.D. Whitman, G. C. Robinson. Jr., and A. W. Savolalnen, Technology of Steel Pressure Vessels for Water-Cooled Nuclear Reactors. ORNL-NSIC-21, Oak Ridge National Laboratory, Oak Ridge, TN. (December 1967).

15. L. E. Steele and J. R. Hawthome. "New Information on Neutron Embrlttlement and Embrlttlement Relief of Reactor Pressure Vessel Steels,* In Ftow and Fracture of Metals tn Nuclear Enulronment. ASTM STP 380. American Society for Testing and Materials, Philadelphia. 283-31 1 (1965).

lt:i. L. E. Steele and J. R. Hawthorne, "Neutron Embrtttlement of Reactor Pressure Vessel Steels." in Materials and. Fuels for Htgh Temperature Nuclear Energy Applicattons. The MIT Press. Cambridge, MA., 366-409 (1964).

17. L. E. Steele and C. Z. Serpan. Jr .. *Neutron Embrtttlernent of Pressure Vessel Steels."

in Analysts of Reactor Vessel Radiatton Effects Suroealanr.e Programs, ASTM STP 481.

American Society for Testing and Materials. Philadelphia.49-101 (1966).

18. C. F. Carpenter, N. R Knopf, and E. S. Byron. "Anomalous Embrlttling Effects Observed During Irradiation Studies on Pressure Vessel Steels.* Nucl. Sc. and Eng .* 19, 18-38 (1964).

19. J. R. Hawthorne and L. E. Steele, "Metallurgical Variables as Possible Factors Controlling Irradiation Response of Structural Steels.* In Effects of Radiation on Structural Metals. ASTM STP 426, American Society for Testing and Materials.

Philadelphia. 534-572 (1967).

20. J. R. Hawthorne. Radiation Effects Infonnatton Generated on the ASTM Reference Correlation-Monitor Steel.$, r"..<:;TM Data Serles Publication D&-54, American Society for Testing and Materials. PhlladeJphla, PA (1974).

34

21. L. E. Steele. J. R Hawthorne, C. Z. Serpan, Jr.. E. P. Klier. and H. E. Watson. Irrad.iatiDn Materials Eualuatwn and Reactor Pressure Vessel Surveillance for th£ Anny Nuclear Power Program. U.S. Naval Research Laboratory Memorandum Report 1644 (September 196E1).

22. L. E. Steele and C. Z. Serpan, Jr., -Army Reactor Vessel Survetllance and Vessel Examination. In Analysts of Reactor Vessel Radiation Effects Surveillance Programs, M

ASTM STP 481, American Society for Tesung and Materials, Philadelphia, 105-135 (1966).

23. M. Grounes. -Review of Swedish Work on Irradiation Effects in Pressure Vessel Steels and on Significance of Data Obtained," l.n F;Uects of Radlatlon on Structural Metals.

ASTM STP 426, American Society for Testing and Materials, Philadelphia. 224-259 (1967).

24. L. K. Mansur and K. Farrell, ~on Mechanisms by which a Soft Neutron Spectrum may Induce Accelerated Embrtttlement.- J. Nucl. Mater.. 170. 236-245 (1990).

25. M. L. Hamilton and H. L. Helnesch, ~ensile Properties of Neutron Irradiated A212B Pressure Vessel SteeV in 14th Int. Symp. on Effects of Radlatwn on MatertaJ.s, ASTM STP 1046. American Society for Testing and Materials, Philadelph!a, 45-54 (1990).

26. D. G. Doran. R L. Simons. and W. N. McElroy, *spectral Effects in Neutron and Charged Particle Irradiadons.

ln Properties of Reactor Structural Alloys after Neutron or Particle Irradiation. ASTM STP 570, American Society for Testing and Materials, Philadelphia.

290-310 {1975).

27. H. Wledersich, '"Effects of the Primary Recoil Spectrum on Long Range Migration of Defects,'" Radt.at. Eif.. 113 (1-3).97-107 (1990)

28. H. L. Helnisch, '"Correlation of Mechanical Property Changes in Neutron-Irradiated Pressure Vessel Steels on the Basis of Spectral Effects.* Fuston Reactor Matertals, Semiannual Progress Report. March 31. 1989, DOE-ER-0313/6, 51-56 (1989).

35

Distribution for NUREG/CR-5748 IANL-91/23. SAND91* 1933)

Internal:

0 . K. Chopra (20) S. T. Rosinslu. SNL (25) TIS Files (31 H . M . Chung W. J. Shack (51 AN!. Patent File T . F . Kassner C. E.T111 Ai-IL Contract File C . Ma lefyt (2) R W. Weeks NRC. for dlstnbutton per RS ANL Librartes (2)

SNL Libraries (5)

Manager, Chicago Operations Office. DOE Matenals and Components Technology DMslon Review Committee H. Berger. lndustrtal Quallty. Inc .. Gaithersburg, MD M. S. Drcsselhaus. MaBsachusctts Institute of Technology. Cambndge. MA S. J . Green, Electric Power Research Institute, Palo Alto. CA RA. Greenkorn. Purdue U .. West Lafayette. IN C.- Y. Li, Cornell U., Ithaca, NY P. G. Shewmon. Ohio State U., Columbus R E . Smith. Electric Power Research Institute, NDE Ctr .. Charlotte. NC D. Attertdge, Battelle Pacific Northwest Laborat1.,ry W. H. Bamford. Westinghouse Electric Corp.. Pittsburgh R D. Cheverton. Oak Ridge National Laboratory, Oak Ridge. TN A. Cowan. Risley Nuclear Power Development Labs.. Risley. Warrington. UK W. H. Cullen. Matertals Engineering Associates. Inc .. L;mh 2:n. MD 8 . J. L. Oarlaston. Berkeley Nuclear Laboratories. Bt> - ;teley ';loucest **-:-shire. UK H. Farrar, Rockwell International. Rocketdyne Dlvtsie *.. Can,.,ga Park. CA J . Gilman, Electric Power Research Inst.* Palo Alto. CA L. Greenwood, Battelle Pacific Northwest Laboratory M. Guttmann, Electrtcite de France. Les RenarcUere~ Roule de Sens. France D. L. Harrison, om~ of LWR Safety and Technology. Ll .S . Dept. of Energy, Washington , DC J . R Hawthorne, Materials Engineering Associates. Lannam, MD P. Hedgecock. NlJfECH Engineers, San Jose. CA B. Hemsworth. HM Nuclear Installations Inspectorate. Loncton C. G. Interrante. Center for Materials Science. National Institute of Standards and Technology, Gaithersburg. MD C. E . Jaske. CC Technologies. Cortest, Columbus, OH P. M . Lang. Office of Converter Reactor Deployment, U.S. Uept. of Energy, Washington. DC M. Lapides. Electric Power Research Institute. Palo Alto. l '; .

G. J . Llclna. Structural Integrity Associates. San Jose. C'.A T. R Mager. Westinghouse Electrtc Corp .. Pittsburgh Y. Meyzaud. Framatome. Parts R K. Nanstad. Oak Ridge National Laboratory. Oak Ridge. TN M. Prager. Materials Properties Council. Inc. New Ycirk

P. H. Pumphrey. National Power, Technology and Environment Center, Leatherhead, Surrey. UK V. N. Shah. EG&G Idaho, Inc .. P. 0. Box 1625, Idaho Falls, ID G. Slama, Framatome, Parts La Defense, France G. D. W. Smith, Oxford University, Oxford, UK R E. Stoller. Oak Ridge National Laboratory, Oak Ridge, TN L. Taylor, Nuclear Electric pie .. Chelsford Rd .. Knutsford, Cheshire. UK J. M. Vitek. Oak Ridge National Laboratory 38

CASS A212--'8 CF-3 -

ESCO, _

Neutttlf'l§t,,iekf~

not trom reactors !lldg. 212 Not ~*fable 2 bkJcb ot S~x5gK2", ~nd These CASS materials are not materials harveste<I from actual reactors. They were cast by companies who were reactor materials suppliers.

CF-8 ESCO- not fl-om re.tcton Btdg. 212 2 blocts of s* :.s*xr. ~nd 3 CF..SM csco- not from reiKton ~ .212 s blocbof 2sx2 s~.wr Grand Gutt Slftoi!plate Rlrl1r. 212 I I I

ex-pIant . Is at IML un t este d .1rrad"1at e d matena Dose, Number of SOURCE Sample form SPECIMEN dpa samples Comments 304 ZORITA PLATE TEM 0.1- so 2 ID?

Small tensile 0.1- so 5 1/4T 0.1- so 10 WELD TEM 0.1- so 2 Small tensile 0.1- so 2 1/ 4T 0.1 - 52? 4 VCSUMMER DISSIMILAR W ELD 1/2T-CT negligible 5 ID?

tensile negligible 3 ID?

DAVIS BESSE DISSIMILAR W ELD 1/2T-CT negligible 3 ID?

1/4T CT negligible 3 DISSIMILAR WELD= Alloy 82+Alloy 182 VC SUMMER = hotleg nozzle-to-pipe weld DAVIS BESSE = CROM nozzole #3

Unte1ted Irradiated materlals at IML fr om HALDEN 1 & 2 Dose, Number of Material PRE-IRRADIATION SPECIMEN dpa samples Comments 304SS SA TENSILES 0.5-2 3 normal/ high Ni 304L SA TENSILE5 0.5-2 2 low/ high Si 304L TENSILES 0.5-2 2 High S, N 304,316 GBE TENSILES 0.5-2 4 GBE treatment prior to irradiation 304 SA TENSILES 0.5*2 2 HighO 304 SA TENSILE.S 0.5-2 1 316LN SA TENSILES 0.5-2 1 304 SA 1/4T 0.75 1 high Cr 304 sensitized l/4T 0.75 1 304 sensit ized 1/4T 2 1 300 TENS!l:ES ~ 4-15 wfhigh-5 316--55 5A ~ ~ wf.Jew-5

.......... O.H ~ w,l-liigli-S Untested irradiate d mate rials at IM L from HALD EN 3 Material PRE-IRRADIATION SPECIMEN dpa samples Comments 304SS SA TEM disks 0.1 6 low/ high S TEM dlsks 3 6 low/ high 5 TENSILE.S 3 4 low/ hi~h s 1/4 TCT 3 2 low/ high S weld as*welded Grand Gulf TEM dlsks 0.1 1 TEMdlsks 3 1 tensfles 3 2 1/4TCT 3 1 304L 5S SA TEMdlsks 0 .1 3 TENSILES 3 1 TEM disks 3 3 l/4 TCT 3 1 316 55 SA TEMdisks 0.1 3 TEM disks 3 3 TENSILE.S 3 1 1/4 TCT 3 1 CASS CF3 and CFS, unaged/aged TEM d isks 0.1 25 1/4T CT 0.1 4 CF3 and CFS, unaged/aged TEMdisks 3 40 TENSILES 3 18 1/4T CT 3 2

Untesud irnd iatNS m a terials at IML from BOR-60 Dose, Number of Material PRE- IRRADIATION SPECIM EN dpa samples Comments 304SS SA TEM disb s. 10, 20, 48 4, 4,4,4 ID? Some sample IDs on these BOR60 TE.M d isks are ille gible . So, w e may have d ifficu lty t o identify them although they do appear o n the inventory.

GSE SMA.l.l TENSR.ES 10,48 1.1 cw TEMdisb s. 10, 20, 48 3, 4, s. 2 StiW.ll:011Sll.£S ~

GSE TEM disb s. 10, 20, 48 2,2.2,2 347 <;W ................ 5,-48 1,-1 316SS SA TEMdisb s, 10, 20, 48 1, 2,2, 1

,_,_ ~ 1,--1,--l cw TEM disb s, 10, 20, 48 1, 2, 0,2 WW Tt:Mdlsb s, 10, 20, 48 2, 2, 2, 2

-. ~

GBE TEM dlsb s. 10, 20, 48 1, 2, 2, 2 SMM.l--H.M&IS 5;-lG;-48 316LN-Ti 316 Ti- SA SMALL f ENSU..ES 10 2 347 ss SA SMALLTENSllES s 2 cw SMAll TENSILES s. 48 1, l SA TEM diW s. 10, 20, 48 2,1,0,l cw TEM disb s. 10, 20, 48 2,1,2,2 690Ni GSE TEM di~lls s. 10, 20, 48 1, 1, 1, l CASS CF-3, CFS unaged SMAllTENSMS 10 1, 1 different heats from the 4S-dpa heats CF3, CFS. una ed SMAI..L TENSflES 48 2, 2 different heats from the lo-dpa heats

Note to requester: The attachment is immediately From: Purtscher, Patrick following this email.

Sent: Fri, 24 Aug 2018 19:13 :02 +0000 To: Hiser, Matthew

Subject:

harvesting_ composite Attachments: harvesting_composite.docx I th ink 2 major changes can be considered. First, I th ink the Abstract and Summary should be condensed and combined .

Second, Sections 1 and 2 can also be combined and condensed.

I will be looking over the rest so we can discuss on Monday.

Pat

Abstract Harvesting and evaluation of ex-plant materials has been a critical part of the technical basis behind the aging management used by lbe NRC to support long term operation ofNPPs The decommissioning of some nuclear power plants (NPPs) in the United States after extended operation may provide an opportunity to increase knowledge about materials aging and degradation, through the harvesting of, and subsequent research on, service-aged materials. This document describes a potential approach for prior itizing sampling (harvesting) materials using a number of crit.eria that incorporate knowledge about V.

the specific ,ssucs ;involved. number of information sources on materials degradation in NPPs were Comme nted [PP 1): Instead of "issue" maybe use reviewed to assess key issues that are relevant. ]infortnation from these sources was cross-referenced "topic"?

(where possible) and collated to assess harvesting pri ority. In this document, several examples of Comme nted [BB-2): The key technical issues were implementation of this process are described, along with experiences from harvesting materials at several identified in an internal conference in 2008. They are operating and closed plants. Using these lessons learned from previous harvesting campaigns, a the four areas mentioned later in this report and in the SRM on SECY 2014-0016.

harvesting process is defined that includes many of the criteria that should be taken into account during any harvesting campaign. Comme nted (RP*3]: Confusion on technical issues for SLR, and gaps associated with material behavior knowledge. For example - RPV may be a technical Issue for SLR identified In SECY 2014-0016. Here, the focus is on lhe specifics - what are (were) the unknowns thal may have led to this determination as a tech issue? I have modified the language a bit to address concern.

Introduction

µnderstanding of the causes and possible control of aging degradation moohftt1i6ff15 forms the basis for deve loping aging management programs (AM Ps) to ensure the continued ]functionality of and Comme nted [BB4}: We have already developed maintenance of safety margins for NPP SSCs. Harvesting and evaluation of ex-plant materials has been a AMPs and provided their technical basis in NUREG-critical part of the technical basis behind the AM Ps used by the NRC to support extended operation of 2221 for SLR.

NPPs, but the process was always limited by the availability of representative materials. Commented [BB.SJ: Report was pre-GALL. This statement is a reftection of the fact that the AMPs are As o f August I, 20 18, the nuclear power neet in the United States ~ consists of 99 operating based on understanding of the causes and possible control of mechanisms. Not sure what the concern is reactors and 14 reactors that are closed and in different stages of decommissioning. The total operating here.

reactors in 2025 is expected to be 85, based on the current industry plans. Given that many plants are continuing to operate and one has applied for continued operation through SLR (80 year total), continued efforts for harvesting and evaluation of ex-plant materials from the other units that are or will be closing is prudent. In addition, it is likely that opportunities to sample materials from operating plants will also Comme nted [BB6}: Something is missing here.

Commented ]FC7 }: Not necessarily. For example, research has been performed to demonstrate that arise as plants consider replacing specific components that may have shown degradation.

vessel materials in surveillance programs are representative within certain lead factors. Not A key challenge to understanding materials aging and degradation in the NPP environment tlt1~M!e appropriate to state "need."

)'@ars of 0130r11lioH is the ability to perform tests on material that arc aged in a representative environment Comme nted [FC8}: The statement was abOut 1e-that-e~J)eftl!it1g-pltmt,;. Often, such tests arc performed (and materials performance data obtained) accelerated aging , not surveillance programs which are through accelerated aging experiments, where the material being tested is subjected to higher stressors considered representative.

(e.g., temperature, stress, neutron nux) than those seen in operation. Such tests enable the experiments to Commented [BB9}: From where?

be completed in a reasonable time frame. Accelerated testing should ftleal-ly be benchmarked with performance data from materials that have seen more representative service aging. >

Comme nted [FC1 0}:

Commented [FC1 1}: Phrases like this are Where available, such benchmarking can be performed [using surveillance specimen exposed to field unacceptable. The work has been done in the GALL-SLR conditions during the course of operation of the reactor. However, surveillance specimens are often limited to critical components such as the RPV, and do not exist for components in other locations in a Commented [FC'1 2}: No. Again, the GALL-SLR is plant. ~ such cases, benchmarking of laboratory tests may be achieved by harvesting of materials from complete.

reactors. ti°he resulting insights into material aging mechanisms can provide confim,ation of the Comme nted JFC13}: Comment appears to address a effectiveness of aging management approaches used by the nuclear industry, as well as insights into the statement that Is no longer In the report draft.

ability lo r0euee operating margin5 as a f\,melieH ofagiRg while maintaining confidence that long-lived Commented JFC14 }: For the first underlined passive components will be capable of continuing to meet their functional requirements during extended sentence, does this mean that we won 'I have operation{ he harvested materials could also*hel in assessin the reliabilit of s ecific methods for reasonable assurance prior to the testing being conducted? In that case, how can we Issue a condition assessment or non-destructive evaluation (NOE) that may be applied to assess aging of these renewed license for 60 - 80 years before the testing components in the fieldj is completed? WIii we generate license conditions to restrict how far into the SPEO a licensee can While harvesting may be quite valuable to increase technical knowledge of materials agi ng, it may not operate before the testing is complete? This statement Is too broad. For the second underlined always be practical. In some cases where harvesting may be desired, the components exist in areas with sentence, should I infer that the AMPs associated high radiation doses, which makes significantly more expensive and challenging. 1111 many cases, the with the four classes of SSCs of concern (e.g.,

benefits of harvesting may not be enough to overcome the significant costs of procurement, evaluation, concrete, cables) are inadequate and subsequent disposal of the materials. L Comme nted JFC15}: Not sure what the comment means - the first sentence has been modified/cut short Given the significant opportunities and challenges for materials harvesting from decommissioning and by other reviewers. Edited. The second sentence -

operating NPPs, it is beaeficial to have a strategic and systematic approach to materials harvesting. This modified by authors to address concern.

document describes an approach for sampling (harvesting) that focuses on prioritizing materials using Comme nted [BB1 6}: Need to access the costs several criteria. These criteria also help define the specific problems that will be add ressed and the against the benefits knowledge gained through the sampling process. Using a number of lessons learned from previous Comme nted [BB17}: i believe that is what the harvesting campaigns, a harvesting process is defined that includes many of the criteria that should be sentence alludes to!

2

considered during any harvesting campaign. The next section discusses criteria for harvesting and provides examples of applying these criteria.

1.

3

Note to requester: The attachments are immediately attached. The squares containing the X are the Adobe PDF attachments to this email record.

From: Hull, Amy L Sent: Wed, 29 Jun 2016 14:09:41 -0400 To: NRCIPAC Resource Cc: RESlnvoices Resource;HISER, MATTHEW A;PURTSCHER, PATRICK T;FRANKL, ISTVAN

Subject:

IPAC Approved in full - CN: NRC-HQ-25-14-D-0001 (08/2016)

Attachments: 5017 invoice approv AP08 FY2016.pdf, 2016 20_ NRCHQ2514D0001_EWA_Strategic Harvesting.pdf Please see attached. I approve the amount in full. If more information is needed, please also see G:\DE\CMB\Hull\Strategic R&D database ex-plant materials\5.

Program Monitoring 0

0 From: NRCIPAC Resource Sent: Wednesday, June 29, 2016 12:04 PM To: Hull, Amy L <Amy.Hull@nrc.gov>

Cc: RESlnvoices Resource <RESlnvoices.Resource@nrc.gov>

Subject:

2ND REQUEST FOR IPAC APPROVAL CN: NRC-HQ-25-14-D-0001 (08/2016)

NRC Payments posted transactions(s) for an lnteragency Agreement that has you listed as the Contracting Officer Representative (COR). Documentation of these transactions along with a summary lnteragency Invoice Approval form is attached. Please complete the approval form in its entirety and return it to NRCIPAC.Resource@NRC.Gov .

Replying to this request wiithout attaching a completed form may result in non-compliance. You are required to respond to this second request within 5 days from the date of this email in accordance with performance measure AW-FM-05.

PLEASE NOTE: The total amount on the IPAC may relate to more than one Task Order. Individual COR designees are only responsible for approving the amount related to their specific task as listed on the attached report.

FOR QUESTIONS PLEASE CONTACT:

ASHA MATTHEWS: 301 -415-2958 < Asha.Matthews@nrc.gov >

MILES DAVIS: 301-415-4100 < Miles.Davis@nrc.gov >

INDU MATHUR: 301-415-3993 < lndu.Mathur@nrc.gov >

POOJA PATEL: 301-415-1558 < Pooja.Patel@nrc.gov >

<< Fi le: "'OTKFROI00lF.PDF >>

From: RESlnvoices Resource Se nt: Wednesday, June 15, 2016 10:52 AM To: NRCIPAC Resource <NRCIPAC.Resource@nrc.gov>

Cc: Hull, Amy <Am y.Hull@n r c.gov>; Hiser, Matthew <Matthew.Hiser @nrc.gov>; RESlnivoices Resource <RESlnvoices.Resource@n rc.gov>

Subje ct: FW: 1ST REQUEST FOR IPAC APPROVAL CN: NRC-HQ-25-14-D-0001 (08/2016)

It is belong to Amy Hull.

'Daooua ~~

Program Analyst Financial & Performance Management Team Program Management, Policy Development and Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Phone: {301} 415-2097 Office: TWFN 10 836 From: NRCIPAC Resource Se nt: Wednesday, June 15, 2016 10:11 AM To: RESlnvoices Resource <RESlnvoices.Resource@nrc.gov>

Subje ct: 1ST REQUEST FOR IPAC APPROVAL CN: NRC-HQ- 25-14-D-0001 (08/2016)

Good Morning, Can you please assist in identifying the COR for this approval.

Thank you.

NRC Payments posted transactions(s) for an lnteragency Agreement that has you listed as the Contracting Officer Representative (COR). Documentation of these transactions along with a summary lnteragency Invoice Approval form is attached. Please complete the approval form in its entirety and return it to NRCIPAC.Resource@NRC.Gov .

Replying to this request wiithout attaching a completed form may result in non-compliance. You are required to respond within 20 days from the date of this email according to performance measure AW-FM-05.

PLEASE NOTE: The total amount on the IPAC may relate to more than one Task Order. Individual COR designees are only responsible for approving the amount related to their specific task as listed on the attached report.

FOR QUESTIONS PLEASE CONTACT:

ASHA MATTHEWS: 301-415-2958 < Asha .Matthews@nrc.gov >

MILES DAVIS: 301 -415-4100 < Miles .Davis@nrc.gov >

INDU MATHUR: 301 -415-3993 < lndu.Mathur@nrc.gov >

POOJA PATEL: 301-415-1558 < Pooja.Patel@nrc.gov >

<< File: ~oTKFROI00 l F.PDF >>

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NRC-HQ-60-15-T-0023 Reporting Period Start Date March 26, 2016 I Reporting Period End Date April 22, 2016 NRC Agreement Number Task Order Number (if applicable) TAC/Cost Center NRC-HQ-25-14-D-0001 NRC-HQ-60-15-T-0023 Project Title Technical Assistance in Support of Agency Environmental and Reactor Programs Task Title Strategic Approach for Obtaining Material and Components Aging Information Period of Performance Start Date I Period of Performance End Date September 17, 2015 September 30, 2019 COR Telephone E-mail Amy B. Hull (301) 251-7656 Amy.Hull@nrc.gov Alt - Matthew Hiser (301) 251 -7601 Matthew.Hiser(@nrc.gov DOE Laboratory Pacific Northwest National Laboratory (PNNL)

DOE Site Address Pacific Northwest Site Office, PO Box 350/MS K9-42, Richland, WA 99352 Principal Investigator Telephone E-mail Pradeep Ramuhalli (509) 375-2763 Pradee12.Ramuhalli@12nnl.gov Financial Status Section A. Overall Funding Current Month Cost: $5,017 Total Ceiling Amount: $163,529 Total Amount of Funds Obligated to Date: $133,165 Total Amount of Funds Expended to Date: $68,976 Percentage of Funds Expended to Date: 52%

Balance of Obligated Funds Remaining: $64,189 Total Estimated Encumbered Costs: $0 Balance Available Less Estimated Encumbered Costs: $64, 189 B. DOE Laboratory Acquired Property: N/A C. NRC-Funded Software Developed: N/A Technical Status Section A. Deliverables/Milestones Schedule Task Description Planned Revised Completion Actual Completion Date Date (if applicable) Completion Date 1.1 PNNL to provide Report 1.1 . Draft NLT 9 months after Progress Report/TLR to NRC on contract award Subtask (1 .1) reviewing anticipated NPP L TO materials degradation and prognostics 1.1 NRC to provide comments to NLT 1 month after contractor on Report 1.1 on NPP receiving draft L TO materials degradation and Report 1 .1 from prognostics PNNL Page 88 of 138

1.2 PNNL to provide Report 1.2. Draft NLT 12 months Progress Report/TLR to NRC on after contract award Subtasks (1.2-1.4) concerning availability of ex-plant material and information, and a systematic approach to harvesting ex-plant materials.

1.2 NRC to provide comments to NLT 1 month after contractor on Report 1.2 concerning receiving draft availability of ex-plant material and Report 1.2 from information PNNL 1.3 Summary Report 1.6. Draft TLR to NLT 16 months NRC including information from after contract award Reports 1.1 through 1.5.2.

(Note: At the discretion of COR, a decision may also be made to publish Summary Report 1 as a TLR rather than as an NUREGICR, depending on the significance of the literature review and research assessment results).

1.3 The Contractor will make a technical When the draft presentation to the NRC staff on Summary Report Summary Report 1.3 at NRC 1.3 is delivered to Headquarters in Rockville, MD. NRC.

1.3 NRG to provide comments to NLT 2 months after contractor on Summary Report 1.3. receiving draft Summary Report 1.3 from PNNL 1.3 DOE Contractor to publish NLT 2 months after Summary Report 1.3 as TLR. receiving NRC Deliver 12 hard copies to the NRC comments COR, in addition to an electronic file.

Task 2 is optional pending outcome of Task 1.

2 PNNL to provide Report 2 Draft TLR NLT 24 months to NRG based on results from after original Subtask (2.1) concerning technical contract award gap identification and subtask (2.2) determination of significance and disposition of gaps 2 NRC to provide comments to NLT 1 month after contractor on Report 2 concerning receiving draft technical gap identification, Report 2 from significance, and disposition PNNL 2 PNNL to publish TLR Report 2 NLT 1 month after technical gap identification, receiving NRG significance, and disposition. comments Deliver 12 hard copies to the NRC COR, in addition to an e lectronic file.

Page 89 of 138

Task 3 is optional pending outcome of Task 2 3.1 PNNL to provide Report 3.1 Draft NLT 46 months TLR to NRG based on results from after contract award Subtask (3.1) and Subtask (3.2) concerning specific laboratory experimentation , analytical model development, and adequacy of existing codes and standards for SLR.

3.1 NRG to provide comments to NLT 1 month after contractor on Report 3.1 concerning receiving draft specific laboratory experimentation Report 3.1 from and analytical model development PNNL 3.1 PNNL to publish TLR Report 3.1 NLT 1 month after concerning specific laboratory receiving NRG experimentation , analytical model comments development, and adequacy of existing codes and standards for SLR. Deliver 12 hard copies to the NRG GOR, in addition to an electronic file.

Task 4 is optional pending outcome of Task 1 and partially pending on Task 2 and 3 4 PNNL to provide Report 4 Draft TLR NLT 46 months to NRG documenting development after contract award of prognostic tool to track and resolve critical SLR technical issues 4 NRG to provide comments to NLT 1 month after contractor on Report 4 reviewing receiving draft development of prognostic tool to Report 3.2 from track and resolve critical SLR PNNL technical issues 4 PNNL to publish TLR Report 4 NLT 1 month after reviewing development of receiving NRG prognostic tool to track and resolve comments critical SLR technical issues R.

Deliver 12 hard copies to the NRG GOR, in addition to an electronic file.

B. Progress During the Reporting Period Work continued towards the goal of developing a process for identifying high priority needs for harvesting. Draft decision tables were created, using stress corrosion cracking in dissimilar welds as an example, to consolidate qualitative information from multiple sources (such as EMDA, etc.). These tables provided the ability to summarize such information against the various criteria identified (and described in the previous MLSR). Such tables, when compiled for multiple components/degradation mechanisms, can provide the necessary basis for rank-ordering (by priority) and down-selecting specific components for harvesting.

Discussions were held with the NRC-RES PM on these decision tables and based on the feedback, these tables were slightly modified. The modified tables were applied to generate additional summaries around two examples - CASS, and low voltage cables. Work to populate these tables is ongoing, as is work to develop a harvesting plan that documents the process for harvesting high priority components.

This document is expected to include the necessary information for prioritizing and identifying components, procedural information for harvesting (i.e., the different elements that must be considered Page 90 of 138

prior to the actual harvesting of components), and a process for documenting the research that the harvested components will be utilized for.

Under task 1.5 a prototype was put together to demonstrate a potential option for the information management tool. The demonstration to NRC staff on March 24, 2016 was well received and believed to meet the needs for the tool. We will begin working on a requirements specification that describes the tool, the capabilities, and identifies what actions are needed to take the tool from a prototype stage to production ready.

C. Travel None D. Description of Estimated Encumbered Costs None E. Anticipated and Encountered Problem Areas None F. Plans for the Next Reporting Period PNNL will complete the draft decision tables for CASS and low voltage cable examples. These tables will be analyzed to determine the process for using the criteria for prioritization of harvesting opportunities. A draft of the harvesting plan will be developed and shared with the NRC PM for feedback.

Spending Plan:

Month/Year 14-Oct 14-Nov 14-Dec 15-Jan 15-Feb 15-Mar 15-Apr 15-Mav 15-Jun 15-Jul 15-Aua 15-Sep Planned($) 4,667 Revised ($)

Actual ($) 207 Variance (%) 89%

Month/Year 15-0ct 15-Nov 15-Dec 16-Jan 16-Feb 16-Mar 16-Apr 16-May 16-Jun 16.Jul 16-Aua 16-Seo Planned($) 10,161 10,366 12,315 12,710 9.433 14,185 14,185 10,529 10,529 14,185 14,185 10,531 Revised($)

Actual ($) 3,9 11 16,079 9,436 6,476 18,537 9,31 2 5,01 7 Variance (%) 89% 89% 89% 89% 89% 89% 89%

Month/Year 16-0ct 16-Nov 16-Dec 17-Jan 17-Feb 17-Mar 17-Apr 17-May 17-Jun 17.Jul 17-Aug 17-Sep Planned($) 1,555 1,555 1,555 1,555 1,555 1,555 1,555 1,555 1,555 1,553 0 0 Revised ($)

Actual ($)

Variance (%)

Month/Year 17-0ct 17-Nov 17-Dec 18-Jan 18-Feb 18-Mar 18-Apr 18-May 18-Jun 18.Jul 18-Aug 18-Sep Planned($)

Revised($)

Actual ($)

Variance (%)

Month/Year 18-0ct 18-Nov 18-Dec 19-Jan 19-Feb 19-Mar 19-Apr 19-May 19-Jun 19.Jul 19-Aug 19-Sep Planned($)

Revised($)

Actual ($)

Variance (%)

TOTAL Planned($) 163,529 Revised($)

Actual ($) 68,975 Page 91 of 138

From: Hull, Amy Sent: Fri, 17 Apr 2015 16:37:31 +0000 To: Knobbs, Katie

Subject:

just written are you available?: Possible SOW Strategic Approach for Obtaining Aging Degradation Information from Decommissioning Nuclear Power Plants

1.0 BACKGROUND

Regulatory Context:

The NRC has established a license renewal process that will allow nuclear power plants (NPP) to renew their licenses for an additional 20 years, via 10 CFR 54.31(d) stating that "a renewed license may be subsequently renewed ." The biggest challenges for the NRC and the industry will be addressing the major technical issues for this second "subsequent" license renewal (SLR) beyond 60 years. The staff currently believes (SECY-14-0016, NUREG-1925) the most significant technical issues challenging power reactor operation beyond 60 years are:

Reactor pressure vessel (RPV) neutron embrittlement at high fluence

Irradiation assisted degradation (IAD) of reactor internals and primary system components

Concrete and containment degradation

Electrical cable qualification and condition assessment Understanding the causes and control of degradation mechanisms forms the basis for developing aging management programs (AMPs) to ensure the functionality and safety margins of nuclear power plant (NPP) systems, structures, and components (SSC). The resolution to these issues should provide reasonable assurance of safe operation of the components in the scope of license renewal during the subsequent period of extended operation. As stated in SRM-SECY-14-006, "the staff should continue to emphasize in communications with industry the need to strive for satisfactory resolution of these issues prior to the NRC beginning a review of any SLR application."

This confirmatory research incorporates many of the ideas proposed in the NRC long term research project (LTRP)lil related to developing a strategic and systematic approach to sampling materials from SSC in decommissioning plants. There are separate, currently independent activities in RES to sample Zorita baffle plate material (Hiser & Rao, UNR NRR-2012-008); to sample Zion NPP SFP boral panels and surveillance coupons (Focht, UNR NRR-2013-005),

and to harvest Zion degrade,d cables (Murdock, UNR NRR-20 11 -014). There is also an activity in RES to identify key significant technical issues pertinent to SLR to ensure NRC's readiness to review possible second license renewal applications (Hull, UNR-NRR-2014-001). As mentioned in the LTRP, the envisioned work addresses both passive and active components. In that sense, it links degradation addressed by the license renewal rule, 10 CFR 54 with aging management of active components covered by the maintenance rule, 10CFR50.65.

H.

Safety Beyond 60 Years Existing Regulatory Process license Renewal Maintenance Rule Quality Assurance Aging Management 1

(10 CFR S0.6S) Program (10CFR S4)

(10 CFR Parl 50 Append11t B) Ensures that the effects of ag11g w ill be 10 CFR 50.55a 11 effectwelymanaged ReQuiremerts throughout the penod of extended operation

( Active Components J

= = = = = - - - *- -

( Passive Components J Aging Management Effectiveness Fig. 1 Relationship between aging management of active and passive components (from NRR/RES presentation to ACRS, 2014)

Oversight of operating NPPs would be enhanced by acquisition of data and information useful reducing uncertainties or improving sensitivity analysis in probabilistic risk assessments (PRA).

The proposed work was originally envisioned by ORA to address both passive and active components.

Independently, IAEA is exploring the formation of a coordinated research project (CRP) to evaluate structure and material properties utilizing actual aged materials removed from decommissioned reactors for safe long-term operation (LTO).

In the last year, four plants have ceased operation or announced that will cease operation in the next year (Crystal River Unit 3 (PWR), Kewaunee (PWR), SONGS Units 2 & 3 (PWR), and Vermont Yankee (BWR-Mk1 )). These plants comprise a range of reactor types, containments, as well as SSCs important to safety. Other NPPs may be added to this list in the near future.

The objective of this project is to develop a long-range strategy for obtaining information from these plants as they go through decommissioning. The focus will be on timely acquisition of information that can significantly improve the agency's risk-informed and performance-based regulatory approach, but has been very difficult or impossible to obtain from the operating reactor fleet.

Technical Context:

The original LTRP proposed creating a roadmap for obtaining information from designated NPPs as they go through decommissioning. It is complementary to ongoing work in RES/DE/CMB (NRR-2010-006, NRR-2010-013, and NRR-2014-001) developing technical

information to support evaluating NPP second license renewal (SLR, long-term operation, LWR sustainability) as well as ongoing work (NRR-2012-008, NRR 2013-05) involving data collection and testing on materials from Spain's decommissioned Zorita reactor and the Zion reactor (both PWRs). As mentioned in the LTRP, oversight of active components in operating NPPs would be enhanced by acquisition of data and information useful reducing uncertainties or improving sensitivity analysis in probabilistic risk assessments (PRA). The Standardized Plant Analysis Risk (SPAR) models provide PRA tools to support these risk-informed activities.

Material degradation has traditionally been managed in a reactive mode more than a proactive mode. For the NPPs just now entering their first license renewal period from 40-60 years, and submitting SLRAs, it is necessary to extrapolate out to 80 years. Extrapolation of the evaluation of material properties in SSC from actual decommissioned NPPs will provide a basis for comparison with results of laboratory tests and calculations to resolve the four issues listed above.

The proactive management of materials degradation (PMMD) tool was originally created at PNNL for RES (POC: Amy Hull) to give an expert opinion of the possible future degradation mechanisms on a subcomponent/material specific basis (PNNL-17779)llil. Combined with the LER database, the PMMD tool allows one to not only react to past events, but to help anticipate future issues.lilil The original PMMD tool was based on NUREG/CR-6923,(iy) it is now appropriate to integrate information from the recently-published five volumes of NUREG/CR-7153,M the expanded materials degradation assessment (EMDA) for reactor pressure vessel (RPV). piping and reactor internals. cables and cable insulation. and concrete.

NUREG/CRs 6923 and 7153 present information in terms of its degradation susceptibility and knowledge for mitigation or prevention. A component witlh high degradation susceptibility/low knowledge wou ld be the strongest candidate for proactive actions. It is necessary to be able to understand this before prioritizing sampling.

In its section on PMMD (pg. 40)MI ACRS noted [1] that it supported this work [2] and the later expansion of its original scope to include concrete structures and electrical cable insulation (via the 5-volume (861pp) EMDA NUREG. With almost 4000 pgs. just from PMDA/EMDA, it is prudent to develop a web-based platform for analysis capabilities and interactive visualizations that offer intuitive ways to explore information. Such tool is expected to save considerable staff time in efforts to both understand and apply the PMDA/EMDA to regulatory review of licensee information._

In order to move from a reactive mode to a proactive mode when dealing with degradation, there are gaps in science & engineering that need to be addressed and require the basic science that supports the engineering solution.ll!.ill This is of particular concern as industry moves forward with the SLR conceptMill and NRC tries to be proactivetw. EPRI tried to identify R&D to address issues through the period of extended operation (to 80 years) and to map the EPRl-led R&D projects to the existing AMPs to provide a basis to support technical discussions on SLR.W At this juncture, considering the demonstrated industry interest in LTO, it would be prudent to further develop a web-based modified scalable reasoning system (SRS) for tracking, disposition, and resolution of critical issues for the proactive management of materials degradation in NPPs. Under a previous RES contract (JCN N6019), PNNL established a robust enterprise level relational database for the thousands of pages of structured information associated with NUREG/CR-6923 (some legacy cache remaining under http://pmmd.pnl.gov) with the goal of marshalling knowledge from the vast stores of information available and making it easily accessible via an easy-to-use and intuitive toollm . As shown in Fig. 2, it was originally envisioned as integrating domestic and international operating experience information (CODAP, CADAK, PARENT Atlas, etc.) as well as information from the EPRI Issue Management Tables and NRC license renewal guidance documents._

There are a number of technical gaps that this project addresses. Most importantly, the current piecemeal approach that obtains isolated and fragmented degradation information as targets of opportunity arise at a few plants can be replaced with a strategic plan that is more comprehensive, wider in scope, and more risk-informed. The strategic plan for inspections and/or testing developed in this project will be useful guidance for obtaining key measurements of degradation in a variety of areas. These measurements will be valuable on their own. They will also be useful in basic research on the underlying mechanisms and modes of degradation ,

and for validation of modeling and simulation tools. Data and information developed from implementation of the strategic plan will also be useful in evaluating aging monitoring and mitigation strategies proposed by Industry.

Expert P anol Ropon Matclrt.la Roll*bUlty MMerlal* Roll*bUlty o o PMOA Program PWA IMT Program 8WR IMT NURCGICR-&023 CPRl MRP-205 BWRVIP- 187 NRC..fnd.-try Conel.Uoo Floal

nd Oata Product Comparison u-08 0.t* Food O~ntlog OPOE Plplog Expelrionce O:ic_abaao (Charfoa H anf*)

Generic Aging Le-Le*rned (GALL) NURC0-1801 PINC Atl-08 SCAP (SCC)

(Moyor/Prolcofl-) (Robert H nrdl- )

Fig. 2 Prognostic tool to track and resolve critical technical issues for SLR[Afl 2.0 OBJECTIVE Understanding and managing material and component degradation is unquestionably a key need for the continued safe and reliable operation of NPPs. It is also an area with very significant uncertainties. In many cases, the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in NPPs is incomplete. A strategic approach to examination and testing of materials and components from a relatively large cohort of decommissioning reactors can dramatically increase our knowledge-acquisition rate in this very important area.

The objective of this research is to construct a strategic pllan for obtaining unique and significant information that will inform the NRC's age-related regulatory oversight decisions over the next two decades, and perhaps beyond. Implementation of the strategic plan, in cooperation with Industry and DOE partners can be accomplished over time, through individual research projects as the identified plants progress through their decommissioning phase. This exploratory research is expected to provide reactor material degradation fundamental insights, information addressing potential technical issues, or identified gaps to support anticipated future (> 5 years)

NRC needs Another objective of this research is to revise the PMMD infotool, and bring to an NRG operating platform, to incorporate SLR-relevant information so that it can be better used to inform prioritization in the ex-plant material strategic plan.

3.0 SCOPE OF WORK In addressing the above objective, the DOE Laboratory must conduct the Task 1 confirmatory research and provide all resources necessary to accomplish the subtasks and deliverables.

Decision on further execution of other Tasks (2 -4) will be made after the end of the Task 1 scoping study.

The first stage will consists of a questionnaire and interviews with focus groups from various technical disciplines within NRC. The second stage will comprise one or two public workshops .

The results of the first two phases will be combined into a final strategic plan in the third stage.

The ex-plant harvesting strategic plan will be developed in cooperation with Industry and other federal agencies such as DOE. International counterparts may also be interested in participation.

The DOE Laboratory, in the first stage, must assist NRC with developing a probing questionnaire and interpreting the results of interviews from various technical disciplines first within NRC. This project will first start by leveraging resources within NRC. The most important internal resource to be leveraged is the collective knowledge and experience of NRC staff and selected contractors who have been engaged in license renewal reviews, inspections, analysis of operating experience data, research on material and component aging issues, and probabilistic risk assessment. Previous efforts such as those that produced the license renewal guidance documents (LRGDs, NUREGs- 1800 and-1801) and NUREG/CRs-6923 and 7153 will be leveraged in this project. Results of previous research projects on material aging and degradation issues will also be leveraged.

From: Knobbs, Katie [2]

Sent: Friday, December 05, 2014 6:20 PM To: Hull, Amy

Subject:

Possible SOW

Amy, Just wanted to touch base and see if you were still planning on issuing a SOW to PNNL for some of the PMMD work?

Thanks, Katie

Katie Knobbs Project Manager Software Engineering & Architectures Pacific Northwest National Laboratory 902 Battelle Boulevard P.O. Box 999, MSIN K7 - 22 Richland, WA 99352 USA Tel : 509-372-4560 (b)(6) ~~L;I I katie . knobbs@pnnl.gov www.pnl.gov ill Strategic Approach for Obtaining Material and Component Aging Information from Decommissioning Nuclear Power Plants (LTRP ID 111 , 2013, Joseph Kanney, RES/DRA/ETB) llil Bond LJ , SR Doctor, and TT Taylor. 2008. Proactive Management of Materials Degradation - A Review of Principles and Programs. PNNL-17779, Pacific Northwest National Laboratory, Richland , WA.

lilil Bond , L.J ., Taylor, T.T., Doctor, S.R. , Hull , A.B . and Malik, S.H. (2008) Proactive Management of Materials Degradation for nuclear power plant systems. Proc. Int. Conf. Prognostics and Health Management 2008, Denver, CO , October 6-9. IEEE Reliability Society,# OP-20-01 120 IM NUREG/CR-6923 , Expert Panel Report on Proactive Materials Degradation Assessment, 2007 (3895pp ,

ML063520517)

W NUREG/CR-7153, Expanded Materials Degradation Assessment, 5 volumes , 2013 (861 pp)

Ml Review and Evaluation of the NRG Safety Research Program , ACRS Apri l 15, 2014 , 89pp (ML14105A332)

Mi.I Proactive Management of Materials Degradation for Nuclear Power Plant Systems, Proc. Int. Conf. Prognostics &

Health Management, OP-20-01 (9pp) lYiiil NE/ Roadmap for Subsequent License Renewal, Dec. 2013. (45pp) li!I SECY-14-0016, Ongoing Staff Activities to Assess Regulatory Considerations for Power Reactor Subsequent License Renewal, January 31 , 2014 (25pp) l!l EPRI 3002000576, Long-Term Operations Program: Assessment of R&D Supporting AMPs for L TO, August 2013 (80pp).

!Jill Proactive Management of Materials Degradation , Draft NRC Research Plan , FY10- 14

!AflMatt, can you help me draft schematic that shows different input, I wi ll give you a sketch - we need one box for decommissioned reactors roadmap , another for operating experience, another for EMDA, another for the technical issues database, another for the EPRI MDM, IMT (delete SCAP, delete piping data base)

Note to requester: The attachment is immediately From: Hull, Amy fo llowing this email.

Sent: Thu, 2 Jul 2015 23 :16:53 +0000 To: Frankl, Istvan {lstvan .Frankl@nrc.gov)

Subject:

PNNL Harvesting EWA task order justification . R2 . July 2.docx Attachments: PNNL Harvesting EWA task order justification. R2. Ju ly 2.docx with track changes . Clean copy in STAQS.

~ U.S.NRC UM-' .t.- >l"*J.M L:pil1- t c.-i.i.*

ENTERPRISE WIDE AGREEMENT TASK ORDER JUSTIFICATION DOCUMENTATION Version Control Date: November 1, 2014 PURPOSE:

This template serves to document the basis for placing a task order under one of the following Enterprise-Wide Agreements (EWA) with the Department of Energy (DOE) Laboratories.

Select DOE Laboratory:

[ x] Pacific Northwest National Laboratory (PNNL} N RC-HQ-25-14-D-000 1

[ ] Brookhaven National Laboratory /BNL} NRC-HQ-25-14-D-0002

[ ) Argonne National Laboratory (ANL} NRC-HQ-25-14-D-0003

[ ] Oak Ridge National Laboratory /ORNLl NRC-HQ-25* 14-D-0004

[ ) Sandia National Laboratory (SNL} N RC-HQ-25-14-D-0005 Select Applicable Criteria (Refer to MD 11 .7):

[ x ] Unique Technical Disciplines or Combinations of Disciplines

[ x] Specialized Facilities or Equipment

[ ] Use of Patents, Copyrights, Proprietary Information, or Secret Processes

[ x] Accrued Knowledge and Equipment or Facilities I ) Urgent Requirements

[ ) Engineering, Developmental, or Research Capability (1) Briefly explain how your task order requirement fits within the scope of the EWA Statement of Work:

The subject of the task order is a systematic evaluation of material degradation. For the NPPs just now entering their first license renewal period from 40-60 years, and submitting subsequent license renewal applications (-SLRAs1, it is necessary to extrapolate out to 80 [ Commented [IF1]: Please define acronyms.

years. Extrapolation of the evaluation of material properties in systems. structures. and components (SSCs}_from actual decommissioned NPPs will provide a basis for comparison (commented [IF2]: Same as above. =7 with results of laboratory tests and calculations to resolve the issues stressed in SRM-SECY-14-006. One Qf these is irradiation-assisted degradation of reactor internals and primary system components. For approximately the past 10 years, RES has maintained a stress corrosion cracking (SCC) testing program at PNNL, responsive to user need requests from the Office of Nuclear Reactor Regulation. The results of the tests are used to support the technical bases for reviews related to in-service inspection frequency and flaw analyses.

The task order is within the scope of the EWC Statement of Work, which states that PNNL can "Provide technical assistance to aid development and implementation of policies, processes, and guidance documents associated with review and approval of licensing and license renewal applications, as well as pre-application activities, staff and management interactions with industry, internal reporting requirements, and interfacing with stakeholders."

The data generated with this 1cmder the task order will be ~ used for licensing actions, such as the review of in-service inspection relief requests, the review of ASME Code Cases ro osed for inclusion b reference in 10 CFR, 50.55a, and in the develo ment of uidance

~ U.S.NRC UM-' .t.- >l"*J.M L:pil1- t c.-i.,. .

ENTERPRISE WIDE AGREEMENT TASK ORDER JUSTIFICATION DOCUMENTATION Version Control Date: November 1, 2014 documents such as the Generic Aging Lessons Learned (GALL) Report for reactor license renewal.

The task order is also within the scope of the EWC Statement of Work, which states that PNNL can " ...provide technical assistance in developing and reviewing the required infrastructure to support the DC, COL, ESP, and operating reactor applications." The resu lts of the tests on harvested ex-plant materials could beai:e used by the Office of New Reactors for the analyses of materials proposed for new reactors, for instance, to understand the effects of weld repairs following E11,1~onstruction. Dissimilar metal welds or weld overlays can im act the resultant com onent inte rit . Commented [IF3]: Nol clear. Please rephrase.

Commented [HA4]: done (2) Briefly explain how your task order requirement fits within the scope of the approved EWA Source Selection Justification (SSJ):

The Source Selection Justification for Pacific Northwest National Laboratories indicates that the approved criteria are Unique Technical Disciplines or Combinations of Disciplines and Accrued Knowledge and Equipment or Facilities. Both of these criteria apply to the current task order.

Concerning the criterion for unique technical disciplines or combinations of disciplines, the performance of this project requires specialized knowledge concerning physical metallurgy, welding, thermo-mechanical processing, corrosion science, microscopy, fractography, and materials compositional analysis. This set of capabilities can be found at PNNL, but not in the commercial marketplace among non-conflicted vendors. NRC remains highly engaged in the technical community and is well aware of private sector and university research capabilities based on ongoing interactions at conferences, workshops, and other forums.

Concerning the criteria for accrued knowledge and equipment or facilities, NRC has invested approximately 30 years to establish the knowledge base at PNNL needed for this work including: (a) Material and component condition after different stressors; (b) Better knowledge of specific degradation and its potential for reducing the design safety margin for the components; (c) Incorporation of plant data into the various material, inspection, and structural integrity evaluation models; and (f) An integration of all these aspects into the regulatory decision making process to consider the risk contribution due to material degradation.

Based on my knowledge of the technical requirements and the market research conducted, the work requested will not place DOE and its contractors in direct competition with the domestic private sector (Statement does not apply if Engineering, Development, or Research Capability is the only criteria selected.)

~ U.S.NRC UM-' .t.- >l"*J.M L:pil1- t c.-i.,. .

ENTERPRISE WIDE AGREEMENT TASK ORDER JUSTIFICATION DOCUMENTATION Version Control Date: November 1, 2014

_ _Amy Hull, RES/DE/CMB_ _ _ __

Contracting Officer's Representative

__z.i4102~1201 Date s~-- I Commented [IFS]: Please revise date.

Office/Division/Branch I concur:

Contracting Officer Date Office/Division/Branch/Team The Source Selection Justifications, Statement of Work and Fact Sheet are available in NEAT:

http://neat.nrc.gov/Catalogs/Catalog.aspx?CataloqlD= 11 .

Note to requester: The attachment is immediately following this email.

From: Moyer, Carol Sent: Wed, 18 Oct 2017 20:03:55 +0000 To: Frankl, Istvan

Subject:

RE: ACTION: Inputs for EPRI quarterly MOU call Attachments: EPRI Quarterly MOU Status Update cem.xlsx

Steve, Updates are in the attached file, in red. I added information on the Cables item. Greg's wording was better than mine for the 1st sentence under harvesting, so I left that change alone!

I double-checked the other entries, also. They match what I had in my version of the table.

Carol From: Frankl, Istvan Sent: Wednesday, October 18, 2017 3:56 PM To: Moyer, Carol <Carol.Moyer@nrc.gov>

Subject:

RE: ACTION: Inputs for EPRI quarterly MOU call Importance: High I have attached the latest spreadsheet that Brian reviewed.

If possible, please update this version as well.

Thanks, Steve From: Frankl, Istvan Sent: Tuesday, October 17, 2017 5:41 PM To: Hiser, Matthew <Matthew. Hiser@nrc.gov>; Moyer, Carol <Carol.Moyer@nrc.gov>

Subject:

RE: ACTION: Input s for EPRI quarterly MOU call Importance: High Thanks, Matt.

Carol, Greg will need the revised spreadsheet tomorrow, i.e. some clarifications in the CMB inputs as well as additional inputs from ICEEB on harvested cables.

Thanks, Steve

From: Hiser, Matthew Sent: Tuesday, October 17, 201710:42 AM To: Frankl, Istvan <lstvan.Frankl@nrc.gov>; Moyer, Carol <Carol.Moyer@nrc.gov>

Subject:

RE : ACTION: Inputs for EPRI quarterly MOU call Hi Steve, I don't have anything to add to address Brian's question below. Our strategic harvesting effort hasn't had any direct involvement with ongoing harvesting programs, which is what Brian's question below and the associated entry in the Excel spreadsheet seem to be driving at. I think this question from Brian and action item need to be addressed by ICEEB.

My understanding from the discussion at the workshop in March, is that EPRI and the industry don't see much (if any) need for further cable harvesting beyond what has already been done. I don' t think NRC feels the same way, but opportunities for cooperation are limited if EPRI doesn't want to play ball.

Again, I don't think this is something Carol and I can/should really address vs. ICEEB.

Thanks!

Matt From: Frankl, Istvan Sent: Tuesday, October 17, 201710:32 AM To: M oyer, Ca rol <Carol. Moyer@nrc.gov>; Hiser, M atthew <Matthew.H iser@nrc.gov>

Subject:

ACTION: Inputs for EPRI quarterly MOU call Importance: High

Carol, Brian needs additional info on harvesting for the upcoming EPRI quarterly MOU call on 11/3.

Please address the highlighted request below in your reply and update relevant section of the attached spreadsheet.

Please complete this action by noon tomorrow.

Matt, Please assist Carol with this action.

Thanks, Steve From: Thomas, Brian Sent: Tuesday, October 17, 201710:11 AM To: Oberson, Greg <Greg.Oberson@nrc.gov>; Frankl, Istvan <lst van.Frankl@nrc.gov>; Iyengar, Raj

<Raj.lyengar@nrc.gov>; Koshy, Thomas <Thomas.Koshy@nrc.gov>; Miller, Kenneth A

<KennethA.Miller@nrc.gov>; Boyce, Tom <Tom.Boyce@nrc.gov>

Cc: Regan, Christopher <Christopher.Regan@nrc.gov>

Subject:

RE: Inputs for EPRI quarterly MOU call

Folks, The status update for the action items for CMB, ICEEB, and RGGIB needs improvement.

For CIB - I am not aware of any deep dive meetings occurring. Specific accomplishments for such meetings should be identified . Neither I nor Chris attended nor were invited to any such meeting . Information stated was already known and does not portray any progress on the action item .

For cable harvesting - please state what was done to enable the comp letion of the harvesting.

Also state what other collaborative activities are needed regarding cable research at this time?

For RGGIB/Codes and Standards - please state what occurred or was agreed to going forward at the Standards Forum.

Thanks ... Brian From: Oberson, Greg Sent: Monday, October 16, 2017 4:40 PM To: Thomas, Brian <Brian .Thomas@nrc.gov>

Cc: Regan, Christopher <Christopher.Regan@nrc.gov>

Subject:

Inputs for EPRI quarterly MOU call Brian ,

Attached are the inputs for your consideration. I would like to provide these to Nick by Wednesday if possible.

Greg

Action Item Ass iQnment Status Update For Manaqo mont Awareness Review the availability of cables that cou ld be harvesIed from plants ln Cable harvesting Is complete for the currenl research EPRI and NRC management and sIaff should be decommlssionlng to support research on cable aging and performance project on cable condition assessm.ent and cable encouraged to continue awareness of NPPs entenng under realistic conditions. Elevate as needed to EPRI and NRC degradation. EPRI assisted In obtaining aged cables decommissioning in order to identify potential management to facilitate successrul availability. from Zion. which are used in current NRC components for future harvesting. NRCfRES is confirmatory research . undertaking a research project 10 prioritize components ICEEB fo, harvesting lhat will support aging management studies for SLR. including electrical components. NRC has plans io do additional testing on electrical cables. We would like to do SLR-related testing on other electrical components . also, lf possible.

Schedule "deep dive" meetings on LTO RVPJ Concrete/ Cables research EPRIINRC *deep dives" have been completed. EPRI and NRC management and staff should be within the near-term (3-6 months) to assess the status of roadmap activities, encouraged to continue participation in the joint roadmap Identify remaining gaps in Research, determine what research remain$ 10 RPVs & Internals: A publ ic wotkshop on RPV$ and process to Irack completlon of confirmatory research for be completed, and when can we ten'rlinate these research projects (e.g., internals is planned for Spring 2019, LTO, as well as to identiry any emerging opportunities for concrete irradiation). Additionally. identify options to complete the research leveraging or otherwise accelerating completion or the in an efficient manner and that optimizes use of available resources. Concrete: A joint (NRC/DOE/EPRI) roadmap work. lessons learned from reviews and implementation Assess readiness for potentlal utility submlttals by Dec 2017. Use these meeting on concrete Is expecled In 04 of CY2017. of the lead SLR applications will be fed back into the joint updated roadmaps to complete remaining research in support or long term 4 roadmap process.

operations. CMB Cables: A joint roadmap meeting on cables is sched uled for 1/8/2018. A public workshop on COfla'ete and cables is planned for Summer 2020.

Based on confirmatory research to date, the NRC is ready to receive utility submittats in Dec. 2017. The joint roadmap process is being used to track completion of remaining research In support of LTO.

Identify if there are opportunities for an earlier SLR workshop in 2017 in Arter discussions with RES and EPRI sta ff, it was None advance of the first SLR application by the end of lhe year. determined lhat an SLR workshop in 2017would not be timely. Near-term applicants are in the peer*

review phase, and unlikely to modify applications.

Wol1,,shops would be more effectlve after lessons CMS learned from addressing the lead appl ications, Public workshops on SLR are being planned for Spring 2019 (RPVs and Internals) and Summer 2020 (Concrete and Cables).

Responsibility for this action item has been taken by None Develop technical addendum on advanced reactor materials research Kathy Gibson. The MOU will address all ANLWR which identify planned NRC and EPRI cooperation. Focus on aligning CIB research. not just materials. The MOU is nearing efforts and avoiding unnecessary duplication of activiUes. Target end of the readiness for management review.

year.

Forward to Kur1 by the end of June lhe invite to the September 2017 Brian Thomas sent an email to Kurt Edslnger on 6128 Kurt E. suggested that we make use of EPRI repOtls a Standards F0tum meeting, which NRC is hosting. inviting EPRI to the NRC Standard Forum , and topic ro, a qual'lerty meeting or a face*to-face meeting.

requesting that EPRI make their reports publ icly We could also explore whether EPRI could get vendors to available so they can be used fat standards. Kurt particpate in using the reports and creating standards.

RGGIB replied on 6130 that EPRI would suppon the Forum, and would likely make their reports avallable to 1hose interested.

Worti. with legal staff lo enable domesllo distribution or the xLPR code and Sta ff has sent a revised MOU to inlcude domestic- No response yet from EPRI legal counsel. Raj elevated facilitate future international distribution. Expl0te viable and practical only distribution of xLPR to EPRI legal team late to 8. Thomas attention by email on 10/15.

approad"les, such as distributing the code to international non--govemmenlal August. We are yet to hear back from the EPRI legal entitles through RISSC. team.

CIB

Note to requester: Both attachments are immediately following this email.

From: Hull, Amy Se nt: Wed, 9 Sep 2015 13:17:45 +0000 To: Bloom, Steven;Burton, William;Brady, Bennett;Jones, Heather (Heather.Jones@nrc.gov);Wong, Albert;Billoch, Araceli;Litkett,. Bernard

Subject:

followup -- strategic approach to ex-plant harvest ing Attachme nts: TO NRC_ HQ_60-15-T-0023 SOW.docx, TO NRC-HQ-60-15-T-0023.pdf Thanks for including me in great meetings yesterday afternoon. I tried to retrieve the photo yesterday of you all that I took but my blackberry memory is apparently too full and no photos have been saved from the past week ! Sorry, no documentation of RES not being last !

Attached is background on the new project I discussed yesterday. When we gave the presentation in early June at the Materials Technical Information Exchange, industry people stood up and said we should come out to talk to folks at tlhe decommissioning plants sooner than later. I think the model that Bennett established for the AM P Effectiveness Audits was really good and effective. I think it would be lovely to work with RSRG on this via a UNR as discussed yesterday.

I talked to Rob Tregoning about the idea this morning and he said there is a lot of work to identify re metallic, electrical, concrete even active components and working with RSRG on this via a UNR is a great idea. T here is a lot of interesting work to be done.

For the deep dive meeting discussions, I included the following related to the attached.

I) Cross-Cutting-Issues Requiring Better Knowledge in the Longer Term (3-8 Years) l) Library of Potential Ex-Plant Harvesting Oppo1tunities:

Significance: At the June 2-4, 2015 Annual Industry-NRC Materials Programs Technical Info Exchange Meeting, NRC staff presented a strategic approach for obtaining material and component aging information. Creating a roadmap for obtaining information (including ex-plant harvesting) from designated NPPs as they go through decommissioning is complementary to ongoing research in deve loping technical infom1ation to support evaluating SLR applications. The focus is on timely acquisition of experiential real-world aging-degradation information that has been very difficult or impossible to obtain from the operating reactor fleet. There was interest from industry participa nts that perhaps can be fo llowed up on in the context of deep-dive discussions and the existing MOUs. In the past few years, four plants have ceased operation or announced that they will cease operation: Crystal River Unit 3 (PWR), Kewaunee (PWR), San Onofre Units 2 & 3 (PWR), and Vermont Yankee (BWR). These plants comprise a range of reactor types, containments, and SSCs important to safety. The primary objective of this project is to develop a long-range strategy for obtaining information from such plants as they go through decommissioning.

Specific Question/Comments:

a) Will industry support NRC in evaluating the availability of ex-plant materials and information? One of the perceived tasks is to interview cognizant individuals at the identified plants who possess critical knowledge. This would have some parallels with the NRC AMP Effectiveness Audits that were conducted at NMP, Ginna, and RNP.

b) Will industry support NRC to create a library of potential harvesting opportunities through the auspices of such organizations as !FRAM or the IAEA CRP? Can we collaborate on such activities through our MO Us?

Version Control Date: July 24, 2015 STATEMENT OF WORK NRC Agreement Number NRC Agreement NRC Task Order Number (If NRC Task Order Modification Number Applicable) Modification Number (If Applicable)

NRC-HQ-25-14-D-0001 N/A NRC-HQ-60-15-T-0023 N/A Project Title Strategic Approach for Obtainiing Material and Component Aging Information Job Code Number B&R Number DOE Laboratory Pacific Northwest National Laboratory (PNNL)

NRC Requisitioning Office Nuclear Regulatory Research (RES)

NRC Form 187, Contract Security and Classification Requirements D Involves Proprietary Information Applicable D Involves Sensitive Unclassified

~ Not Applicable

~ Non Fee-Recoverable Fee-Recoverable (If checked, complete all applicable sections below)

Docket Number (If Fee-Recoverable/Applicable) Inspection Report Number (If Fee Recoverable/Applicable)

Technical Assignment Control Number (If Fee- Technical Assignment Control Number Description (If Fee-Recoverable/Applicable) Recoverabl a/Applicable)

1.0 BACKGROUND

Regulatory Context:

The NRC has established a license renewal process that will allow nuclear power plants (NPP) to renew their licenses for an additional 20 years, via 10 CFR 54.31(d) stating that "a renewed license may be subsequently renewed." The biggest challenges for the NRC and the industry will be addressing the major technical issues for this second "subsequent" license renewal (SLR) beyond 60 years. As summarized in SECY-14-0016, the NRC staff believe that the most significant technical issues challenging power reactor operation beyond 60 years are related to:

Reactor pressure vessel (RPV) neutron embrittlement at high fluence

Irradiation assisted degradation (IAD) of reactor internals and primary system components

Concrete and containment degradation 1

Version Control Date: July 24, 2015

Electrical cable qualification and condition assessment.

Understanding the causes and control of degradation mechanisms forms the basis for developing aging management programs (AMPs) to ensure the functionality and safety margins of NPP systems, structures, and components (SSC). The resolution to these issues should provide reasonable assurance of safe operation of the components in the scope of license renewal during the subsequent period of extended operation.

Because of the cost and inefficiency of piecemeal sampling, there is a need for a strategic and systematic approach to sampling materials from SSC in decommissioning plants. The envisioned work addresses both passive and active components. In that sense, it addresses aging management of passive components under the license renewal rule, 10 CFR 54, as well as the maintenance of active components covered by the maintenance rule, 10CFR5O.65, as seen in Figure 1 below.

,. U.. R C

,.....,.,.,,.""'... ,..._ Safety Beyond 60 Years Quality Assurance l Program

( 10 CFR Part 50 Appendi,c B) Ensures that the effects I

of aging will be 10 CFR 50.55a effectively managed Requirements throughoutthe period of extended operation I

( Active Components

)

Passive Components Aging Management Effectiveness Figure 1: Relationship between aging management of active and passive components (from NRR/RES presentation to ACRS, 2014)

In the past few years, four plants have ceased operation or announced that they will cease operation: Crystal River Unit 3 (PWR), Kewaunee (PWR), San Onofre Units 2 & 3 (PWR), and Vermont Yankee (BWR). These plants comprise a range of reactor types, containments, and SSCs important to safety. The primary objective of this project is to develop a long-range strategy for obtaining information from these plants as they go through decommissioning. The focus will be on timely acquisition of experiential real-worlld aging-degradation information that 2

Version Control Date: July 24, 2015 can significantly improve the agency's risk-informed and performance-based regulatory approach, but has been very difficult or impossible to obtain from the operating reactor fleet.

Technical Context:

Creating a roadmap for obta ining information from designated NPPs as they go through decommissioning is complementary to ongoing NRC research in developing technical information to support evaluating SLR as well as data collection and testing of ex-plant materials.

Material degradation has traditionally been managed reactively in response to events and operating experience, rather than proactively to prevent failures. For the NPPs currently entering their first license renewal period from 40-60 years, and submitting SLR applications, it is necessary to evaluate potential degradation mechanisms out to 80 years of operation.

Evaluation of material properties in SSCs from actual decommissioned NPPs will provide a basis for comparison with results of laboratory tests and calculations to resolve the four issues listed above.

The proactive management of materials degradation (PMMD) information tool was orig inally created at PNNL for RES (POC: Amy Hull) to give an expert opinion of the possible future degradation mechanisms on a subcomponent/material specific basis (PNNL-17779)i.

Combined with the LER database, the PMMD information tool allows one to not only react to past events, but to anticipate future issues. The original PMMD information tool was based on NUREG/CR-6923, "Proactive Materials Degradation Assessment (PMDA)," for the first license renewal period, so it is now appropriate to integrate information from the excel databases from the recently-published five volumes of NUREG/CR-7153, "Expanded Materials Degradation Assessment (EMDA)" for SLR. At this juncture, there is demonstrated industry interest in NPP long-term operation (LTO) and regulatory interest in SLR.

2.0 OBJECTIVES Understanding and managing material and component degradation is a key need for the continued safe and reliable operation of NPPs, but has significant uncertainties. In many cases, the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in NPPs is incomplete. A strategic a pproach to examination and testing of materials and components from decommissioned reactors can dramatically increase our knowledge-acquisition rate in this very important area.

There are three inter-related objectives to this work:

(1) Develop a long-range strategy for obtaining information from decommissioned NPPs as well as providing the flexibility to get ex-plant components from operating plants as well. The focus will be on timely acquisition of experiential real-world aging-degradation information that can significantly improve the agency's risk-informed and performance-based regulatory approach, but has been very difficult or impossible to obtain from the operating reactor fleet.

3

Version Control Date: July 24, 2015 (2) Construct a strategic plan and specifications for obtaining unique and significant materials aging degradation information from diverse sources (operating experience, other nuclear facilities, other long-lived industrial plants, other materials organizations such as ASM and NACE) that will inform the NRC's age-related regulatory oversight in the future.

Implementation of this plan and specifications, in cooperation with industry and DOE partners can be accompl ished over time, through individual research projects as the identified plants progress through their decommissioning process. This exploratory research is expected to provide fundamental insights on reactor materials degradation and information addressing potential technical issues or identified gaps to support anticipated future NRC needs.

(3) Update the PMMD information tool to incorporate LTO/SLR-relevant information so that it can be better used to inform prioritization in the ex-plant material strategic plan.

3.0 SCOPE OF WORK There are a number of technical gaps that this project seeks to address. Most importantly, the current piecemeal approach can be replaced with a strategic plan that is more comprehensive, broader in scope, and more risk-informed. The strategic plan for inspections and/or testing developed in this project will be useful guidance for obtaining key measurements of degradation in a variety of areas. These measurements will be valuable on their own and will also be useful in basic research on the underlying mechanisms and modes of degradation, and for validation of modeling and simulation tools. Data and information developed from implementation of the strategic plan will also be useful in evaluating aging management and mitigation strategies proposed by the industry.

Many sources of materials degradation information will be queried, including human repositories of knowledge both within NRC and within the industry. Both the PMDA and EMDA present information in terms of component or material degradation susceptibility and currently available knowledge for degradation mitigation or prevention. A component with high degradation susceptibility/low knowledge would be the strongest cand idate for proactive actions. It is necessary to be able to understand this before prioritizing ex-plant materials sampling available from a given retired NPP. Previously, under the auspices of NRC contracts (i.e., JCN N6029, N6907), PNNL used the large amount of information presented in the PMDA report to develop a web-based platform to facilitate analysis through interactive visualizations that offer intuitive ways to explore the information. PNNL shall explore the viability of adding materials degradation susceptibility data presented in the EMDA Report.

Such an information tool (Figure 2 below) is expected to save considerable staff efforts to understand and apply the PMDA and EMDA insights to regulatory review of licensee information. PNNL shall develop a web-based modified scalable reasoning system (SRS) for tracking, disposition, and resolution of critical issues, such as determining the appropriate SSC 4

Version Control Date: July 24, 2015 from which to acquire cast austenitic stainless steel (CASS) material of specific composition and radiation dose.

EPRI LTO Decommissioning Reactors Plans Operating Information Tool for Prioritization of Experience High-Priority Data Needs Oooortunities International NRC Data: SLRGDs, Data Sources DOE LWRS EMDA, PMDA Figure 2: Pre-conceptual Architecture of prognostic tool to track and resolve critical technical issues for SLR As shown in Figure 2 above, the information tool was originally envisioned as integrating domestic and international operating experience and experimental information as well as inform ation from the EPRI LTO, DOE Light Water Reactor Sustainability (LWRS) program, and NRC sources such as EMDA, PMDA, and SLR guidance documents (SLRGDs) and precursors. The international data sources that might provide effective data feed include the cable aging data and knowledge (CADAK, http://cadak.hrp.no/cadak.) project and the Component Operational Experience, Degradation and Ageing Programme (CODAP, http://www.oecd-nea.org/jointproj/codap.html ),

both sponsored by OECD/NEA. The Atlas constructed by PNNL from the Program to Assess the Reliability of Emerging Non-destructive Technology (PARENT) and the Program to Inspect Nickel Alloy Components (PINC) Atlas is an international database containing a vast array of sec crack morphology and NDE information. PNNL shall investigate whether this is an appropriate framework to track issue resolution associated with SLR. This is a much broader objective than just developing a strategic roadmap for harvesting SSCs.

The general tasks and their duration are described in Table 1.

5

Version Control Date: July 24, 2015 Table 1: Task Description and Duration Task Task Title/Description Duration (Months)

Task 1 Scoping Study and technical literature review 18 Task 2 Decision Making on Specific Confirmatory Research Needed to 6 Address Gaps (optional)

Task 3 Confirmatory Research Addressing Technical Gaps (optional) 33 T ask4 Development of Independent Decision Making Tools (optional) 33 The conditional tasks shall be conducted, as detailed in Figure 3 below. A decision on further optional research o utlined in Tasks 2, 3, and 4 will be made after completion of Task 1 depending on the outcome and recommendation from the conclusion of specific tasks. The overall nexus between the scoping study and other potential tasks is shown in Figure 3.

The PNNL staff shall not restrict their activities solely to these descriptions and shall be flexible in using their technical knowledge and experience in proposing additions, deletions, or deviations from the prescribed requirements as research progresses.

Task 1.

Terminate Technical Task 2. Gap Further Literature Identification Research Review Yes Task 3.

Recommend Research Need

~

I Task4. :I

I Develop :I

Analysis Tools :

~----------------j Figure 3: Schematic of the Overall Research 6

Version Control Date: July 24, 2015 4.0 SPECIFIC TASKS Task 1 is the scoping study. Tasks 2-4 are optional. NRC plans to revise the SOW for these tasks based on the outcome of Task 1. The time at which the tasks begin and end will be dependent on available information and NRC's ongoing evaluation of testing priorities. NRG staff does not require that PNNL necessarily perform the tasks be performed sequentially following the order in which they are listed. For the test matrix described in this section, nearly all subtasks will have to be tested in tandem with another subtask in order to complete the program within the requested period of performance. PNNL and the NRG CORs will continually review the testing plan during monthly status update teleconferences.

PNNL shall, in the first stage of Task 1, develop a questionnaire and help the NRC staff conduct interviews with focus groups from various technical disciplines within NRC. PNNL shall, in the second stage of Task 1, assist the NRC staff conduct one or two public workshops. PNNL shall analyze and combine the results of the first two phases into a final strategic plan in the third stage.

This strategic plan will provide a prioritization of strategic harvesting opportunities. PNNL shall help the NRC staff develop the ex-plant harvesting strategic plan in cooperation with industry and other federal agencies such as DOE as well as any international counterparts that may be interested in participation.

In Tasks 2-4, PNN L may be assigned optional tasks to identify requirements to further elucidate the risk assessment of component degradation. Such research should also provide technical data and information, as necessary, to request the national codes and standards bodies (such as ASME, ASTM , or NACE) to re-examine requirements for structural materials for passive components in light water reactors (LWRs) and in assessing material degradation during service and its effect on design safety margin of components. The PNNL principal investigator (Pl) for this project shall attend ASME, ASTM, or NACE Code Committee meetings, as appropriate and as approved by the COR during the course of this research . The Pl shall provide adequate information to support an IAEA international cooperative research program (ICRP) on this subject to bring worldwide resources to address this research need.

The specific tasks are as follows:

Task 1 - Literature Review and Assessment of Greatest Needs in Sampling of Ex-plant Materials NRC recently completed a research program to investigate material degradation after extended operation. To investigate aging degradation mechanisms, aging degradation effects, and the relative susceptibility to degradation, PNNL shall perform a comparison of available information.

7

Version Control Date: July 24, 2015 PNNL shall conduct the Task 1 scoping study and provide all resources necessary to accomplish the subtasks and deliverables. Task 1 shall be performed in stages as shown in the Task-specific subsections below.

The activities required for this task are:

Task 1.1 - Conduct Materials Aging Degradation Literature Review PNNL shall selectively review both domestic and internati1onal sources of technical information of generic nature with respect to anticipated material degradation in NPPs during LTO, extrapolating to 80 years of operation. The objective is to identify other issues not in PMDA/EMDA, such as related to active components or spent fuel storage systems, and to determine what is being done to address LTO issues. NRC will provide guidance on appropriate information to review.

Task 1.2 - Evaluate Availability of Ex-Plant Material and Information PNNL shall evaluate what relevant ex-plant material is projected to be available for potential harvesting. PNNL shall work with the NRC COR to develop a questionnaire and interview the cognizant individuals at the plants who possess critical knowledge.

Task 1.3 - Develop Questionnaire and Conduct Interviews with Prospective NRC Stakeholders PNNL shall develop a questionnaire and work with NRC staff to conduct interviews with focus groups from various technical disciplines within NRC. Th1is would include the SLR Expert Panels for a sample of different aging management programs (AMPs) as well as other NRC technical advisory groups. PNNL shall have a comprehensive approach to all the possible stakeholders interested in harvesting materials from decommissioned plants. The objective of this initial scoping study is to assess interest in issues concerning both passive and active component degradation. The questionnaire will address, as a minimum, (1) the perceived needs for ex-plant materials, (2) the perceived utility of the existing information tool and how and where this prognostic tool should be maintained (NRC, contractor, cloud). During the early brainstorming and scoping study, PNNL shall also consider degradation of SSC materials associated with extended long-term storage of used fuel.

Task 1.4 - Develop Questionnaire and Conduct Interviews with Prospective External Stakeholders Based on interactions with NRC staff in Task 1.3 above, PNNL shall propose a preliminary strategic approach to sampling representative ex-plant materials during one or two presentations at public workshops to further refine the concept of what would be needed in a useful interrogatory tool linking aging-degradation research objectives with available resources 8

Version Control Date: July 24, 2015 for ex-plant materials. The searchable information tool shall be available via an interactive web page.

Task 1.5 -Conduct Scoping Analysis on Viability of Searchable Information Tool Task 1.5.1 PNNL shall briefly consider available approaches to creating a preliminary database that will link the highest susceptibility/lowest knowledge anticipated degradation scenarios with potential availability of ex-plant materials. As part of this subtask, PNNL shall review the status and viability of the PMMD information tool created as part of the PMMD project (conducted at PNNL under previous NRC contracts (i.e., JCN N6029, N6907). The goals of the PMMD project were to identify reactor components that could reasonably be expected to experience future degradation, estimate the susceptibility of components to various degradation mechanisms, and assess the degree of knowledge available to develop mitigative strategies. It was anticipated that this information could be used to guide regulatory actions related to license renewal and subsequent license renewal. The PMMD panel evaluated 3863 components (2203 for PWRs, 1603 for BWRs) for their susceptibility to 16 degradation mechanisms (Figure 4 below). Because of the unwieldiness of the source material, a searchable information tool (pmmd.pnl.gov) was developed to make this information usable to NRC staff and others.

Task 1.5.2 PNNL shall work with the NRC to create a proposal to develop a platform for the searchable database methodology (selected in Task 1.5.1) that can be supported withi n NRC.

9

Version Control Date: July 24, 2015 PLANT DATA DRAWINGS IT] Based on draw,ngs, doc.,monts. o.g.. FSAR.

Plant-specif"'

information

[TI COl'lsultants.

'37

~

.------L- - - ' - - ~

PARTS INFORMATION Excel spn,adshoo/s by proop dove/oped by BNL. revised bosod on t chnic in t.

'47 ~-L---~

L....'.:'.....J COMPONENT SUBGROUPS

. - - - -- Exc,il sproadshoola dovelopod by*

Panel lead expett, revised based on i

Mem bers EVALUATIONS Excel sproachheets with d"f}fadation mechanisms sCOffJd and commented on by panel mcmbt>rs.

EVALUATION DATABASE IT] ACCESS database with all

&v&Jvations REPORTS Compilations.

filtors, counts. etc.

Figure 4. Flowchart for files created and used in PMMD infotool Task 1.6 - Provide Archival Summary Document of Findings from Task 1 PNNL shall analyze and review the reports generated from the work conducted under Tasks 1.1 through 1.5 and provide a stand-alone NUREG/CR documenting the major findings.

Optional Task 2 - Decision Making on Specific Research Need to Address Gaps If the Task 1 scoping study succeeds in determining previously unidentified significant knowledge gaps that need further attention, more work will be done in the context of Task 2, pending the approval from the NRC Contract Officer (CO). Thus, Task 2 is optional pending the outcome of Task 1 and requires NRC activation. The activities required for this task are:

Task 2.1 - Gap Identification PNNL shall identify specific information and technical data gaps from the execution of Task 1 and document these gaps. In identifying the gaps, PNNL shall include an examination of the current ASME B&PV Code or other industry practices that the NRC has endorsed with respect 10

Version Control Date: July 24, 2015 to addressing the specific degradation mechanism in the design and the assurance of the retention of the design margin during the period of licensed reactor operation time.

Task 2.2 - Determine Significance and Disposition of Technical Gaps PNNL shall determine whether or not there are any techn ical gaps from the execution of Task 2.1. If there are no gaps and if it is determined that the current ASME Code or other industry practices ensure that the design margin for components are adequate, recommend termination of further research in this topic by NRC. If specific information and technical gaps are identified then proceed to Task 3 after getting approval from the NRC Contract Officer (CO).

As noted, Task 2 is optional, the need to perform this task will be determined by the NRG. In the PNNL response to this SOW, the optional task should be costed separately. The NRG will provide written notice to PNNL via modification to this agreement if the optional task will need to be performed.

Optional Task 3 - Research Addressing Technical Gaps Related to Material Degradation If critically important information and technical gaps are identified in Task 2, Task 3 is activated after getting approval from the NRG CO. Thus, Task 3 is optional pending the outcome of Task

2. The activities required for this task are:

Task 3.1 - Recommend Specific Laboratory Experimentation and Analytical Model Development PNNL shall work with NRC subject matter experts (SMEs) to recommend specific laboratory experimentation and analytical model development, which may address the information gap identified in Task 2. 1. If novel nondestructive evaluation methods (such as the next-generation acoustic emission technology which reportedly can 'hear' crack initiation) become available to identify progressing reactor material degradation by the time Task 3 is initiated, PNNL shall recommend inservice inspection (ISi) technology enablers which will be suitable for detecting the material changes resulting from different stressors. PNNL shall work with NRC SMEs to recommend the need for developing tools for detection and assessment of potential degradation of the design safety margin to independently confirm the !licensee's technical basis for LTO.

Task 3.2 - Review Adequacy of Existing Codes and Standards PNNL shall conduct a review of existing applicable ASME B&PV Codes that may need to be revised as a result of Task 2 .1 and PNNL shall work with NRC SM Es to engage relevant ASM E Code Committees for assessing future path. PNNL shall! propose other Codes and Standards that should be reviewed (such as but not limited to, ANS, ASTM , and NACE codes and standards).

As noted, Task 3 is optional, the need to perform this task will be determined by the NRG. In the PNNL response to this SOW, the optional task should be costed separately. The NRG will provide written notice to PNNL if the optional task will need to be performed.

Optional Task 4 - Investigate Development of Independent Decision Making Tools 11

Version Control Date: July 24, 2015 Task 4 is optional pending the outcome of Tasks 1 - 3. If gaps are identified under Task 2 and appropriate research needed to inform the gaps are also identified under Task 3, NRG expects that the industry will perform the needed research and provide NRC the data for regulatory decisions.

Depending on the outcome of Tasks 2 and 3 and ensuing industry research, the decision-making tool development may be complex and truly involve multi-year, multi-disciplinary long term research. It is expected, however, that the decision making tool may include: (a) Material and component condition after different stressors; (b) Better knowledge of specific degradation and its potential for reducing the design safety margin for the components; (c) Incorporation of plant data into the various material, inspection, and structural integrity evaluation models; and (f)

An integration of all these aspects into the regulatory decision making process to consider the risk contribution due to material degradation.

Specific subtasks for this task will be established later in this research. PNNL shall investigate the feasibility of developing a modern visualization confirmatory analysis research tool for aging management of safety-significant SSC degradation in NPPs. As currently envisioned, this could provide a knowledge management and strategic planning tool for conducting gap assessments and prioritizing R&D resources related to NPP LTO. This research will leverage the work previously performed by PNNL on the PMMD Information Tool, sponsored by RES.

RES/DE would benefit from a R&D gap assessment, strategic planning and knowledge management tool to enhance the tracking, disposition, resolution of technical issues that surface as industry moves towards SLR. Such a database would save staff time in addressing the degradation challenges for NPP passive components, spent fuel pools, and independent spent fuel storage installations (ISFls}. The proposed LTO issues visualization tool can incorporate, up-to-date information on critical issues associated with cable, concrete and RPV aging. Work is actively progressing on developing SLR guidance documents with unresolved technical issues arising almost on a daily basis. These could be captured by the proposed service-oriented analytic framework. The existing PMMD database containing detailed information about susceptibility, knowledge, and confidence associated with hundreds of degradation scenarios can be augmented with aging risk indices, when developed by the DOE LWRS research. This will enable a better understanding of serviice life projections of NPP SSC.

As noted, Task 5 is optional, the need to perform this task will be determined by the NRG. In the PNNL response to this SOW, the optional task should be costed separately. The NRG will provide written notice to PNNL if the optional task will need to be performed.

5.0 DELIVERABLES AND/OR MILESTONES SCHEDULE Except for Task 1.6 where a draft summary NUREG/CR is stipulated, all deliverables shall be in the form of technical letter reports or alternatives previously discussed and determined acceptable by the COR. Based on the detailed tasks provided in Section 4.0 of this Statement of Work, PNNL shall estimate the number of Figures/Tablles or other copyrighted information from technical journals, etc. and shall incorporate this estimation in the cost proposal in addressing the SOW. PNNL shall also estimate reasonable effort by their technical editing staff in order to provide the NRC tech-edited draft final and final reports.

12

Version Control Date: July 24, 2015 ask Deliverable/Milestone Description (include NRG acceptance Due Date (if any)

Number criteria if applicable)

All Monthly Letter Status Report (MLSR) 20th day of each month 1.1 PNNL to provide Report 1.1 . DraftTLR to NRC on Subtask (1.1) NLT 6 months after reviewing anticipated NPP LTO materials degradation and contract award prognostics 1.1 NRC to provide comments to contractor on Report 1.1 on NPP NLT 1 month after LTO materials degradation and prognostics receiving draft Report 1.1 from PNNL 1.1 PNNL to publish TLR Report 1.1 on materials degradation and NLT 1 month after prognostics. Deliver 12 hard copies to the NRC COR, in receiving NRC addition to an electronic file. comments 1.2 PNNL to provide Report 1.2. Draft TLR to NRC on Subtask NLT 8 months after (1.2) concerning availability of ex-plant material and information contract award 1.2 NRC to provide comments to contractor on Report 1.2 NLT 1 month after concerning availability of ex-plant material and information receiving draft Report 1.2 from PNNL 1.2 PNNL to publish TLR Report 1.2 concerning availability of ex- NLT 1 month after plant material and information. Deliver 12 hard copies to the receiving NRC NRC COR, in addition to an electronic file. comments 1.3 PNNL to provide Report 1.3 (consisting of questionnaire and NLT 10 months after interview results) to NRC on Subtask (1 .3) concerning interest of contract award prospective NRC stakeholders in a systematic approach to harvesting ex-plant materials 1.3 NRC to provide comments to contractor on Report 1.3 NLT 1 month after concerning interest of prospective NRC stakeholders in a receiving Report 1.3 systematic approach to harvesting ex-plant materials from PNNL 1.4 PNNL to provide Report 1.4 (consisting of questionnaire and NLT 14 months after interview results) to NRC on Subtask (1.4) concerning interest of contract award prospective external stakeholders in a systematic approach to harvesting ex-plant materials 1.4 NRC to provide comments to contractor on Report 1.4 NLT 1 month after concerning interest of prospective external stakeholders in a receiving Report 1.4 systematic approach to harvesting ex-plant materials from PNNL 1.5.1 PNNL to provide Report 1.5.1 to NRC on Subtask (1 .5.1) with NLT 16 months after suggested alternatives for creating a prognostic tool to track and contract award resolve critical technical issues for SLR 1.5.1 NRC to provide comments to contractor on Report 1.5.1 NLT 1 month after concerning alternatives for creating a prognostic tool to track receiving Report 1.5.1 and resolve critical technical issues for SLR from PNNL 13

Version Control Date: July 24, 2015 1.6 Summary Report 1.6. Draft NUREG/CR to NRC including NLT 20 months after information from Reports 1.1 through 1.5.2. contract award (Note: At the discretion of COR, a decision may also be made to publish Summary Report 1 as a TLR rather than as an NUREG/CR, depending on the significance of the literature review and research assessment results).

1.6 The Contractor will make a technical presentation to the NRC When the draft staff on Summary Report 1.6 at NRC Headquarters in Rockville, Summary Report 1.6 MD. is delivered to NRC.

1.6 NRC to provide comments to contractor on Summary Report NLT 2 months after 1.6. receiving draft Summary Report 1.6 from PNNL 1.6 DOE Contractor to publish Summary Report 1.6 as NUREG/CR. NLT 2 months after Deliver 12 hard copies to the NRC COR, in addition to an receiving NRC electronic file. comments Task 2 is optional pending outcome of Task 1.

2 PNNL to provide Report 2 Draft TLR to NRC based on results NLT 24 months after from Subtask (2.1) concerning technical gap identification and original contract award subtask (2.2) determination of significance and disposition of gaps 2 NRC to provide comments to contractor on Report 2 concerning NLT 1 month after technical gap identification, significance, and disposition receiving draft Report 2 from PNNL 2 PNNL to publish TLR Report 2 technical gap identification, NLT 1 month after significance, and disposition. Deliver 12 hard copies to the NRC receiving NRC COR, in addition to an electronic file. comments Task 3 is optional pending outcome of Task 2 3.1 PNNL to provide Report 3.1 Draft TLR to NRC based on results NLT 46 months after from Subtask (3.1) concerning specific laboratory contract award experimentation and analytical model development 3.1 NRC to provide comments to contractor on Report 3.1 NLT 1 month after concerning specific laboratory experimentation and analytical receiving draft Report model development 3.1 from PNNL 3.1 PNNL to publish TLR Report 3.1 concerning specific laboratory NLT 1 month after experimentation and analytical model development. Deliver 12 receiving NRC hard copies to the NRC COR, in addition to an electronic file. comments 3.2 PNNL to provide Report 3.2 Draft TLR to NRC reviewing NLT 46 months after adequacy of existing codes and standards for SLR contract award 3.2 NRC to provide comments to contractor on Report 3.2 reviewing NLT 1 month after adequacy of existing codes and standards for SLR receiving draft Report 3.2 from PNNL 3.2 PNNL to publish TLR Report 3.2 reviewing adequacy of existing NLT 1 month after codes and standards for SLR. Deliver 12 hard copies to the receiving NRC NRC COR, in addition to an electronic file. comments Task 4 is optional pending outcome of Task 1 and partially pending on Task 2 and 3.

14

Version Control Date: July 24, 2015 4 PNNL to provide Report 4 Draft TLR to NRC documenting NLT 46 months after development of prognostic tool to track and resolve critical SLR contract award technical issues 4 NRC to provide comments to contractor on Report 4 reviewing NLT 1 month after development of prognostic tool to track and resolve critical SLR receiving draft Report technical issues 3.2 from PNNL 4 PNNL to publish TLR Report 4 reviewing development of NLT 1 month after prognostic tool to track and resolve critical SLR technical issues receiving NRC R. Deliver 12 hard copies to the NRC COR, in addition to an comments electronic file.

6.0 TECHNICAL AND OTHER SPECIAL QUALIFICATIONS REQUIRED Specific qualifications for this effort include senior materials engineers and metallurgists who have in-depth knowledge of reactor pressure vessel and core internal materials subjected to irradiation and stress at elevated temperature, and effects of water chemistry on structural reactor materials. The personnel involved should have in-depth experience, knowledge, and demonstrated contributions in the areas of mechanical deformation, material degradation phenomena, such as corrosion, stress corrosion cracking and irradiation effects. The contract personnel should be well-versed in the use of nuclear power plant ASME B&PV Codes and Standards, Industry Guidance Documents, such as those of NEI, EPRI, NRC's Regulatory Guides and NRC's License Renewal Guidance Documents (such as NUREGs 1800, 1801, and 1950) Information Notice (IN), Regulatory Issue Summary (RIS), Generic Letter (GL), Generic Issue (GI) for licensing review by the NRC staff.

The contract personnel should also be aware of the safety evaluation reports (SER) written by the NRC staff on industry guidance documents, as applicable. The contract personnel should have previous experience developing appropriate software architecture for proposed R&D planning tool.

15

Version Control Date: July 24, 2015 7.0 ESTIMATED LABOR CATEGORIES AND LEVELS OF EFFORT The estimated level of effort for this project is as follows:

Task 1 350 staff hours

Task 2 (Optional) 480 staff hours

Task 3 (Optional) 1590 staff hours

Task 4 (Optional) 765 staff hours 8.0 MEETINGS AND TRAVEL The PNNL Principal Investigator and one other engineer shall visit the NRC Headquarters in Rockville, MD and present the overall research outcome to the staff and share in technical discussions. Any suggestions from the staff, as appropriate, may be considered for the final report by the Pl. No other domestic or foreign travel is permitted under the initial scoping study.

9.0 REPORTING REQUIREMENTS PNNL is responsible for structuring the deliverable to follow agency standards. The current agency standard is Microsoft Office Suite 2010. The current agency Portable Document Format (PDF) standard is Adobe Acrobat 9 Professional. Deliverables shall be submitted free of spelling and grammatical errors and conform to requirements stated in this section.

Monthly Letter Status Reports In accordance with Management Directive 11 .7, NRC Procedures for Placement and Monitoring of Work with the U.S. Department of Energy, PNNL shall electronically submit a Monthly Letter Status Report (MLSR) by the 20th day of each month to Amy Hull, the Contracting Officer Representative (COR) and Technical Monitor, to Matthew Hiser and Joseph Kanney (alternate CORs and Technical Monitors), with copies to the Contracting Officer (CO) and the Office Administration/Division of Contracts to ContractsPOT.Resource@nrc.gov. If a project is a task ordering agreement, a separate MLSR shall be submitted for each task order with a summary project MLSR, even if no work has been performed during a reporting period. Once NRC has determined that all work on a task order is completed and that final costs are acceptable, a task order may be omitted from the MLSR.

MLSR should be distributed additionally to the Chief, Corrosion and Metallurgy Branch, RES, the Director, Division of Engineering, RES . Other required distribution will be communicated at the start of this research program.

The MLSR shall include the following: agreement number; task order number, if applicable; job code number; title of the project; project period of performance; task order period of performance, if applicable; COR's name, telephone number, and e-mail address; full name and address of the performing organization; principal investigator's name, telephone number, and e-mail address; and reporting period. At a minimum, the MLSR shall include the information discussed in Attachment 1. The preferred MLSR format can also be found in Attachment 1.

10.0 PERIOD OF PERFORMANCE The estimated period of performance for this work is 48 months from date of agreement award.

16

Version Control Date: July 24, 2015 11.0 CONTRACTING OFFICER'S REPRESENTATIVE The COR monitors all technical aspects of the agreement/task order and assists in its administration. The COR is authorized to perform the following functions: assure that the DOE Laboratory performs the technical requirements of the agreement/task order; perform inspections necessary in connection with agreement/task order performance; maintain written and oral communications with the DOE Laboratory concerning technical aspects of the agreement/task order; issue written interpretations of technical requirements, including Government drawings, designs, specifications; monitor the DOE Laboratory's performance and notify the DOE Laboratory of any deficiencies; coordinate availability of NRC-furnished material and/or GFP; and provide site entry of DOE Laboratory personnel.

Contracting Officer's Representative Name: Dr. Amy 8. Hull Agency: U.S. Nuclear Regulatory Commission Office: Office of Nuclear Regulatory Research Mail Stop: CS-05-C07M Washington, DC 20555-0001 E-Mail: amy.hull@nrc.gov Phone: 301-251-7656 Alternate Contracting Officer's Representative Name: Matthew Hiser Agency: U.S. Nuclear Regulatory Commission Office: Office of Nuclear Regulatory Research Mail Stop: CS-05-C07M Washington, DC 20555-0001 E-Mail: Matthew.Hiser@nrc.gov Phone: 301-251-7601 12.0 MATERIALS REQUIRED (TYPE NIA IF NOT APPLICABLE)

NIA 13.0 NRC-FURNISHED PROPERTY/MATERIALS PNNL will transfer NRC furnished property and materials acquired under previous contracts (i.e., JCN N6029, N6907) to this task order. NRC will provide additional information from EMDA and SLR databases.

14.0 RESEARCH QUALITY (TYPE N/A IF NOT APPLICABLE)

The quality of NRC research programs are assessed each year by the Advisory Committee on Reactor Safeguards. Within the context of their reviews of RES programs, the definition of quality research is based upon several major characteristics:

Results meet the objectives (75% of overall score)

Justification of major assumptions (12%)

Soundness of technical approach and results (52%)

17

Version Control Date: July 24, 2015 Uncertainties and sensitivities addressed ( 11 % )

Documentation of research results and methods is adequate (25% of overall score)

Clarity of presentation (16%)

Identification of major assumptions (9%)

It is the responsibility of the DOE Laboratory to ensure that these quality criteria are adequately addressed throughout the course of the research that is performed. The NRC COR shall review all research products with these criteria in mind.

15.0 STANDARDS FOR CONTRACTORS WHO PREPARE NUREG-SERIES MANUSCRIPTS (TYPE N/A IF NOT APPLICABLE)

The U.S. Nuclear Regulatory Commission (NRC) began to capture most of its official records electronically on January 1, 2000. The NRC will capture each final NUREG-series publication in its native application. Therefore, please submit your final manuscript that has been approved by your NRC Project Manager in both electronic and camera-ready copy.

The final manuscript shall be of archival quality and comply with the requirements of NRC Management Directive 3.7 "NUREG-Series Publications." The document shall be technically edited consistent with NUREG-1379, Rev. 2 (May 2009) "NRC Editorial Style Guide." The goals of the "NRC Editorial Style Guide" are readability and consistency for all agency documents.

All format guidance, as specified in NUREG-0650, "Preparing NU REG-Series Publications,"

Rev. 2 (January 1999), will remain the same with one exception. You will no longer be required to include the NU REG-series designator on the bottom of each page of the manuscript. The NRC will assign this designator when we send the camera-ready copy to the printer and will place the designator on the cover, title page, and spine. T he designator for each report will no longer be assigned when the decision to prepare a publication is made. The NRC's Publishing Services Branch will inform the NRC Project Manager for the publication of the assigned designator when the final manuscript is sent to the printer.

For the electronic manuscript, the Contractor shall prepare the text in Microsoft Word, and use any of the following file types for charts, spreadsheets, and the like.

File Types to be Used for NU REG-Series Publications File Type File Extension MicrosoftWord .doc Microsoft PowerPoint .ppt MicrosoftExcel .xis MicrosoftAccess .mdb 18

Version Control Date: July 24, 2015 I Portable Document Format I .pdf This list is subject to change if new software packages come into common use at NRC or by our licensees or other stakeholders that participate in the electronic submission process. If a portion of your manuscript is from another source and you cannot obtain an acceptable electronic file type for this portion (e.g., an appendix from an old publication), the NRC can, if necessary, create a tagged image file format (file extension.tit) for that portion of your report.

Note that you should continue to submit original photographs, which will be scanned, since digitized photographs do not print well.

If you choose to publish a compact disk (CD) of your publlication, place on the CD copies of the manuscript in both (1) a portable document format (PDF); (2) a Microsoft Word file format, and (3) an Adobe Acrobat Reader, or, alternatively, print instructions for obtaining a free copy of Adobe Acrobat Reader on the back cover insert of the jewel box.

16.0 OTHER CONSIDERATIONS (TYPE N/A IF NOT APPLICABLE)

References

1. Bond LJ, SR Doctor, and TT Taylor. 2008. Proactive Management of Materials Degradation -A Review of Principles and Programs. PNNL-17779, Pacific Northwest National Laboratory, Richland, WA.

2. Bond, LJ, TT Taylor, SR Doctor, AB Hull, and SH Malik, (2008) Proactive Management of Materials Degradation for nuclear power plant systems. Proc. Int. Conf. Prognostics and Health Management 2008, Denver, CO, October 6-9. IEEE Reliability Society,# OP-20-01 120

3. Chopra, OK, et al, Managing Aging Effects on Dry Cask Storage Systems for Extended Long-Term Storage and Transportation of Used Fuel, Rev. 0, FCRD-USED-2012-000119, 2012.

4. EPRI 3002000576, Long-Term Operations: Assessment of R&D Supporting AMPs for LTO, Aug. 2013 (80pp).

5. NEI, Roadmap for Subsequent License Renewal, Dec. 2013. (45pp) 6 . NEI, Second License Renewal Roadmap, May 2015. (22pp).

7. NUREG/CR-6923, Expert Panel Report on Proactive Materials Degradation Assessment, 2007 (3895pp, ML063520517) 8 . NUREG/CR-7153, Expanded Materials Degradation Assessment, 5 volumes, October 2014 (861pp) 9 . SECY-14-0016, Ongoing Staff Activities to Assess Regulatory Considerations for Power Reactor Subsequent License Renewal, January 31, 2014 (25pp)

10. Taylor, WB, CE Carpenter, KJ Knobbs, S Malik, Using Technology to Support Proactive Management of Materials Degradation for the U.S. Nuclear Regulatory Commission, 19

Version Control Date: July 24, 2015 Proceedings of the ASME Pressure Vessels & Piping Division/K-PVP Conference, PVP 2010, July 18-22, 2010. Bellevue, WA, USA. Paper PVP2010-26063.

11. The Scalable Reasoning System: Lightweight Visualization for Distributed Analytics, IEEE Symposium on Visual Analytics Science & Technology, 978-1-4244-2935-6/08 Access to Non-NRC Facilities/Equipment (Type N/A if not applicable)

N/A Applicable Publications (Type N/A if not applicable)

N/A Controls over document handling and non-disclosure of materials (Type N/A if not applicable)

N/A 20

2OROERNO INTERAGENCY AGREEMENT rNRC-HQ-60-15-T-0023

~ NO

13. REQUISITION NO RES-15-0205 1* SOUCITATION NO

,~ OF I 3 5 EFFECTIVE DATE 18 AWAAO DATE 7 PERIOO OF PERFORMANCE 09/04/2015 09/04/2015 09/17/2015 TO 09/30/2019 8 SERVICING AGENCY 9 OELIIIERTO PACIFIC NORTHWEST NAT LAB US NUCLEAR REGULATORY COMMISSION ALC: 11555 ROCKVILLE PIKE DUNS: +4 : ATTN AMY HULL US DEPARTMENT OF ENERGY MAIL STOP T-10D49 PACIFIC NORTHWEST SITE OFFICE ROCKVILLE MD 20852-2738 PO BOX 350 MS K9-42 RICHLAND WA 99352 POC GENICE MADERA TaEPHONe NO 509-372-4010 10 REOUESTNG AGENCY 11 r-NOICE OFFICE ACQUISITION MANAGEMENT DIVISION US NUCLEAR REGULATORY COMMISSION ALC : 31000001 DUNS: 040535809 +4: ONE WHITE FLINT NORTH US NUCLEAR REGULATORY COMMISSION 11555 ROCKVILLE PIKE ONE WHITE FLINT NORTH MAILSTOP 03-E17A 11555 ROCKVILLE PIKE NRCPAYMENTSNRCGOV ROCKVILLE MD 20852-2738 ROCKVILLE MD 20852-2738 POC MICHAEL TURNER TEI..EPHOhE NO 301-415-6712

12. ISSUlNG OFFICE 13. LEGISl.AT~ AUTHORITY US NRC - HQ Energy Reorganization Act of 1974 ACQUISITION MANAGEMENT DIVISION MAIL STOP TWFN-5E03 WASHINGTON DC 20555-0001 14 PflOJECT 10 15 PflOJECT TITlE SEE BLOCK 118 18 ACCOUNTING DATA 2015-X0200-FEEBASED-60-60D001-11-6-213-1032-253D 17 18 19 20 21 22.

ITEM NO SUf'PUESISERVICES QUANTITY UNIT UNIT PRICE AMOUNT Master IAA : NRCHQ2514D0001 00001 Issuance of new Task Order No . NRC-HQ-60-15-T-0( 23 163, 529 . 00 Line Item Ceiling$163 , 529 . 00 Incrementally Funded Amount : $73 , 165 . 00 The NRC and Pacific Northwest National Laborato y (PNNL) hereby enter into this Agreement for the project entitled "Strategic Approach for Obtaining Material and Component Aging Information ."

Continued ...

23 PIIIYMEtlT" PRO\/ISIONS 124 TOTAi.AMOUNT

$73 , 165 . 00 25" SIGNATURE OF GOIIERNMENT REPRESENTATIVE (SERVICING) 2eo SIGNA ~ ATI\IE (REQUESTING) 3/4di A- /.,,, .,...,(A,,/

2Sb NAME N<<J TITLE 2"1>. CONTRACTING OFFICER 1 125c DATE MICHAEL A. TURNER ri7"'i/20J5 I

Note to requester: The public version of this records is in ADAMS at ML19129A329.

This document was also released in its entirety in interim #10.

~

NRC-HQ-60-15-T-0023 2 I3 Period of Performance : September 17 , 2015 -

September 30 , 2019 Consideration and Obligations :

(a) Authorized Cost Ceiling $163 , 529 . 00 (b) The amount presently obligated with respect to this DOE Agreement is $73 , 165 . 00 . When and if the amount(s) paid and payable to the DOE Laboratory hereunder shall equal the obligated amount , the DOE Laboratory shall not be obligatec to continue performance of the work unless and until the NRC Contracting Officer shall increase the amount obligated with respect to this DOE Agreement . Any work undertaken by the DOE Laboratory in excess of the obligated amount specified above is done so at the DOE Laboratory ' s sole risk .

SCHEDULE OF REQUIRED TASKS :

Task 1 - Scoping Study and Technical Literature Review Total Authorized Cost Ceiling ...... $163 , 529 . 00 The Government may require the delivery of the numbered line item identified in the Schedule as

~n option item, in the quantity and at the cost stated in the Schedule . The Contracting Officer

~ill issue a modification to the Agreement to

~uthorize the optional task .

SCHEDULE OF OPTIONAL TASKS :

Task 2 - Decision Making on Specific Confirmatory Research Needed to Address Gaps ($87,775 . 00)

~ask 3 - Confirmatory Research Addressing Technical Gaps ($143 , 240 . 00)

Task 4 - Development of Independent Decision Making Tools ($155 , 462 . 00)

The following document is hereby made a part of this Agreement :

~ttachment No . 1 : Statement of Work

~ontinued ...

NRC-HQ-60-15-T-0023 1-00 PAGE 3

~

I3

~ his agreement is entered into pursuant to the

~uthority of the Energy Reorganization Act of 1974, as amended (42 u.s . c 5801 et seq . ). This

~ ork will be performed in accordance with the

~ RC/DOE Memorandum of Understanding dated

~ ovember 24 , 1998 . To the best of our knowledge ,

the work requested will not place the DOE and its

~ontractor in direct competition with the

~ omestic private sector.

I J Fee Recoverable Work

[XJ Non-fee Recoverable Work PNNL Principal Investigators :

Pradeep Ramuhalli/509-375-2763 Katie Knobbs/509-372- 4560 The total amount of award : $163 , 529.00. The obligation for this a ward is shown in box 24 .

Attachment No. 1 STATEMENT OF WORK NRC Agreement Number NRC Agreement NRC Task Order Number (If NRC Task Order Modification Number Applicable) Modification Number (If Applicable)

NRC-HQ-25-14-O-0001 NIA NRC-HQ-60-15-T-0023 NIA Project Title Strategic Approach for Obtaining Material and Component Aging Information Job Code Number B&R Number DOE Laboratory Pacific Northwest National Laboratory (PNNL)

NRC Requisitioning Office Nuclear Regulatory Research (RES)

NRC Form 187, Contract Security and Classification Requirements D Involves Proprietary Information 0 Applicable D Involves Sensitive Unclassified

~ Not Applicable IZl Non Fee-Recoverable Fee-Recoverable (If checked, complete all applicable sections below)

Docket Number (If Fee-Recoverable/Applicable) Inspection Report Number (If Fee Recoverable/Applicable)

Technical Assignment Control Number (If Fee- Technlcal Assignment Control Number Description (If Fee-Recoverable/Applicable) Recoverable/Applicable)

1.0 BACKGROUND

Regulatory Context:

The NRC has established a license renewal process that will allow nuclear power plants (NPP) to renew their licenses for an additional 20 years, via 10 CFR 54. 31 ( d) stating that "a renewed license may be subsequently renewed." The biggest challenges for the NRC and the industry will be addressing the major technical issues for this second "subsequent" license renewal (SLR) beyond 60 years. As summarized in SECY-14-0016, the NRC staff believe that the most significant technical issues challenging power reactor operation beyond 60 years are related to:

Reactor pressure vessel (RPV) neutron embrittlement at high fluence

Irradiation assisted degradation (IAD) of reactor internals and primary system components

Concrete and containment degradation 1

Attachment No. 1

Electrical cable qualification and condition assessment.

Understanding the causes and control of degradation mechanisms forms the basis for developing aging management programs (AMPs) to ensure the functionality and safety margins of NPP systems, structures, and components (SSC). The resolution to these issues should provide reasonable assurance of safe operation of the components in the scope of license renewal during the subsequent period of extended operation.

Because of the cost and inefficiency of piecemeal sampling, there is a need for a strategic and systematic approach to sampling materials from SSC in decommissioning plants. The envisioned work addresses both passive and active components. In that sense, it addresses aging management of passive components under the license renewal rule, 10 CFR 54, as well as the maintenance of active components covered by the maintenance rule, 10CFRS0.65, as seen in Figure 1 below.

~ US.NRC

,.,..,,t.. ,..,. , J... , _ -

Safety Beyond 60 Years Maintenance Rule Quality Assurance Aging Management (10CFR S06S) Program (10CFR 54)

( 10CFR Part!SOAp~ndccB) r ensures that the effects' ofagr,g w*be 10 CFR 50.55a effectrvelymanaged Requlremerts throughout the periOd of extended operatlDn

( Active Components

)

Passive Components Aging Management Effectiveness Figure 1: Relationship between aging management of active and passive components (from NRR/RES presentation to ACRS, 2014)

In the past few years, four plants have ceased operation or announced that they will cease operation: Crystal River Unit 3 (PWR), Kewaunee (PWR), San Onofre Units 2 & 3 (PWR), and Vermont Yankee (BWR). Tlhese plants comprise a range of reactor types, containments, and SSCs important to safety. The primary objective of this project is to develop a long-range strategy for obtaining information from these plants as they go through decommissioning. The focus will be on timely acqui1sition of experiential real-world aging-degradation information that 2

Attachment No. 1 can significantly improve the agency's risk-informed and performance-based regulatory approach, but has been very difficult or impossible to obtain from the operating reactor fleet.

Technical Context:

Creating a roadmap for obtaining information from designated NPPs as they go through decommissioning is complementary to ongoing NRC research in developing technical information to support evaluating SLR as well as data collection and testing of ex-plant materials.

Material degradation has traditionally been managed reactively in response to events and operating experience, rather than proactively to prevent failures. For the NPPs currently entering their first license renewal period from 40-60 years, and submitting SLR applications, it is necessary to evaluate potential degradation mechanisms out to 80 years of operation.

Evaluation of material properties in SSCs from actual decommissioned NPPs will provide a basis for comparison with re:sults of laboratory tests and calculations to resolve the four issues listed above.

The proactive management of materials degradation (PMMD) information tool was originally created at PNNL for RES (POC: Amy Hull) to give an expert opinion of the possible future degradation mechanisms on a subcomponent/material specific basis (PNNL-17779t Combined with the LER database, the PMMD information tool allows one to not only react to past events, but to anticipate future issues. The original PMMD information tool was based on NUREG/CR-6923, "Proactive Materials Degradation Assessment (PMDA)," for the first license renewal period, so it is now appropriate to integrate information from the excel databases from the recently-published five volumes of NUREG/CR-7153, "Expanded Materials Degradation Assessment (EMDA)" for SLR. At this juncture, there is demonstrated industry interest in NPP long-term operation (LTO) and regulatory interest in SLR.

2.0 OBJECTIVES Understanding and managing material and component degradation is a key need for the continued safe and reliable operation of NPPs, but has significant uncertainties. In many cases, the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in NPPs is incomplete. A strategic approach to examination and testing of materials and components from decommissioned reactors can dramatically increase our knowledge-acquisition rate in this very important area.

There are three inter-related objectives to this work:

(1) Develop a long-range strategy for obtaining information from decommissioned NPPs as well as providing the flexibility to get ex-plant components from operating plants as well. The focus will be on timely acquisition of experiential real-world aging-degradation information that can significantly improve the agency's risk-informed and performance-based regulatory approach, but has been very difficult or impossible to obtain from the operating reactor fleet.

3

Attachment No. 1 (2) Construct a strategic plan and specifications for obtaining unique and significant materials aging degradation information from diverse sources (operating experience, other nuclear facilities, other long-lived industrial plants, other materials organizations such as ASM and NACE) that will inform the NRC's age-related regulatory oversight in the future.

Implementation of this plan and specifications, in cooperation with industry and DOE partners can be accomplished over time, through individual research projects as the identified plants progress through their decommissioning process. This exploratory research is expected to provide fundamental insights on reactor materials degradation and information addressing potential technical issues or identified gaps to support anticipated future NRC needs.

(3) Update the PMMD information tool to incorporate LTO/SLR-relevant information so that it can be better used to inform prioritization in the ex-plant material strategic plan.

3.0 SCOPE OF WORK There are a number of technical gaps that this project seeks to address. Most importantly, the current piecemeal approach can be replaced with a strategic plan that is more comprehensive, broader in scope, and more risk-informed. The strategic plan for inspections and/or testing developed in this project will be useful guidance for obtaining key measurements of degradation in a variety of areas. These measurements will be valuable on their own and will also be useful in basic research on the underlying mechanisms and modes of degradation, and for validation of modeling and simulation tools. Data and information developed from implementation of the strategic plan will also be useful in evaluating aging management and mitigation strategies proposed by the industry.

Many sources of materials degradation information will be queried, including human repositories of knowledge both within NRC and within the industry. Both the PMDA and EMDA present information in terms of component or material degradation susceptibility and currently available knowledge for degradation mitigation or prevention. A component with high degradation susceptibility/low knowledge would be the strongest candidate for proactive actions. It is necessary to be able to understand this before prioritizing ex-plant materials sampling available from a given retired NPP. Previously, under the auspices of NRC contracts (i.e., JCN N6029, N6907), PNNL used the large amount of information presented in the PMDA report to develop a web-based platform to facilitate analysis through interactive visualizations that offer intuitive ways to explore the information. PNNL shall explore the viability of adding materials degradation susceptibility data presented in the EMDA Report.

Such an information tool (Figure 2 below) is expected to save considerable staff efforts to understand and apply the PMDA and EMDA insights to regulatory review of licensee information. PNNL shall develop a web-based modified scalable reasoning system (SRS) for tracking, disposition, and resolution of critical issues, such as determining the appropriate SSC 4

Attachment No. 1 from which to acquire cast austenitic stainless steel (CASS) material of specific composition and radiation dose.

EPRI LTO Decommissioning Reactors Plans Operating Prioritization of Experience High-Priority Data Needs Oooortunities International Data Sources DOE LWRS Figure 2: Pre-conceptual Architecture of prognostic tool to track and resolve critical technical issues for SLR As shown in Figure 2 above, the information tool was originally envisioned as integrating domestic and international operating experience and experimental information as well as information from the EPRI LTO, DOE Light Water Reactor Sustainability (LWRS) program, and NRC sources such as EMDA, PMDA, and SLR guidance documents (SLRGDs) and precursors. The international data sources that might provide effective data feed include the cable aging data and knowledge (CADAK, http://cadakhrp.no/cadak.) project and the Component Operational Experience, Degradation and Ageing Programme (CODAP, http://www.oecd-nea.org/ jointproj/codap.html ),

both sponsored by OECD/NEA. The Atlas constructed by PNNL from the Program to Assess the Reliability of Emerging Non-destructive Technology (PARENT) and the Program to Inspect Nickel Alloy Components (PINC) Atlas is an international database containing a vast array of SCC crack morphology and NOE information. PNNL shall investigate whether this is an appropriate framework to track issue resolution associated with SLR. This is a much broader objective than just developing a strategic roadmap for harvesting SSCs.

The general tasks and their duration are described in Table 1.

5

Attachment No. 1 Table 1: Task Description and Duration Task Task Title/Description Duration (Months)

Task 1 Scoping Study and technical literature review 18 Task2 Decision Making on Specific Confirmatory Research Needed to 6 Address Gaps (optional)

Task 3 Confirmatory Research Addressing Technical Gaps (optional) 33 Task4 Development of Independent Decision Maki1ng Tools (optional) 33 The conditional tasks shall be conducted, as detailed in Figure 3 below. A decision on further optional research outlined in Tasks 2, 3, and 4 will be made after completion of Task 1 depending on the outcome and recommendation from the conclusion of specific tasks. The overall nexus between the scoping study and other potential tasks is shown in Figure 3 .

The PNNL staff shall not restrict their activities solely to these descriptions and shall be flexible in using their technical knowledge and experience in proposing additions, deletions, or deviations from the prescribed requirements as research progresses.

Task 1.

Technical Terminate Task 2. Gap Literature Identification Further Review Research Yes Task 3.

Recommend Research Need

,--------~--------~

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~----------------J Figure 3: Schematic of the Overall Research 6

Attachment No. 1 4.0 SPECIFIC TASKS Task 1 is the scoping study. Tasks 2-4 are optional. NRC plans to revise the SOW for these tasks based on the outcome of Task 1. The time at which the tasks begin and end will be dependent on available information and NRC's ongoing evaluation of testing priorities. NRC staff does not require that PNNL necessarily perform the tasks be performed sequentially following the order in which they are listed. For the test matrix described in this section, nearly all subtasks will have to be tested in tandem with another subtask in order to complete the program within the requested period of performance. PN NL and the NRC CORs will continually review the testing plan during monthly status update teleconferences.

PNNL shall, in the first stage of Task 1, develop a questionnaire and help the NRC staff conduct interviews with focus groups from various technical disciplines within NRC. PNNL shall, in the second stage of Task 1, ass1ist the NRC staff conduct one or two public workshops. PNNL shall analyze and combine the results of the first two phases into a final strategic plan in the third stage.

This strategic plan will provide a prioritization of strategic harvesting opportunities. PNNL shall help the NRC staff develop the ex-plant harvesting strategic plan in cooperation with industry and other federal agencies such as DOE as well as any international counterparts that may be interested in participation.

In Tasks 2-4, PNNL may be assigned optional tasks to identify requirements to further elucidate the risk assessment of component degradation. Such research should also provide technical data and information, as necessary, to request the national codes and standards bodies (such as ASME, ASTM, or NACE) to re-examine requirements for structural materials for passive components in light water reactors (LWRs) and in assessing material degradation during service and its effect on design safety margin of components. The PNNL principal investigator (Pl) for this project shall attend ASME, ASTM, or NACE Code Committee meetings, as appropriate and as approved by the COR during the course of this research. The Pt shall provide adequate information to support an IAEA international cooperative research program (ICRP) on this subject to bring worldwide resources to address this research need.

The specific tasks are as fol lows:

Task 1 - Literature Review and Assessment of Greatest Needs in Sampling of Ex-plant Materials NRC recently completed a research program to investigate material degradation after extended operation. To investigate aging degradation mechanisms, aging degradation effects, and the relative susceptibility to degradation, PNNL shall perform a comparison of available information.

PNNL shall conduct the Task 1 scoping study and provide all resources necessary to accomplish the subtasks and deliverables. Task 1 shall be performed in stages as shown in the Task-specific subsections below.

7

Attachment No. 1 The activities required for this task are:

Task 1.1 -Conduct Materials Aging Degradation Literature Review PNNL shall selectively review both domestic and international sources of technical information of generic nature with respect to anticipated material degradation in NPPs during LTO, extrapolating to 80 years of operation. The objective is to identify other issues not in PMDA/EMDA, such as related to active components or spent fuel storage systems, and to determine what is being done to address LTO issues. NRC will provide guidance on appropriate information to review.

Task 1.2 - Evaluate Availability of Ex-Plant Material and Information PNNL shall evaluate what relevant ex-plant material is projected to be available for potential harvesting. PNNL shall work with the NRC COR to develop a questionnaire and interview the cognizant individuals at the plants who possess critical knowledge.

Task 1.3 - Develop Questionnaire and Conduct Interviews with Prospective NRC Stakeholders PNNL shall develop a questionnaire and work with NRC staff to conduct interviews with focus groups from various technical disciplines within NRC. This would include the SLR Expert Panels for a sample of different aging management programs (AMPs) as well as other NRC technical advisory groups. PNNL shall have a comprehensive approach to all the possible stakeholders interested in harvesting materials from decommissioned plants. The objective of this initial scoping study is to assess interest in issues concerning both passive and active component degradation. The questionnaire will address, as a minimum, (1) the perceived needs for ex-plant materials, (2) the perceived utility of the existing information tool and how and where this prognostic tool should be maintained (NRC, contractor, cloud). During the early brainstorming and scoping study, PNNL shall also consider degradation of SSC materials associated with extended long-term storage of used fuel.

Task 1.4 - Develop Questionnaire and Conduct Interviews with Prospective External Stakeholders Based on interactions with NRC staff in Task 1.3 above, PNNL shall propose a preliminary strategic approach to sampling representative ex-plant materials during one or two presentations at public workshops to further refine the concept of what would be needed in a useful interrogatory tool linking aging-degradation research objectives with available resources for ex-plant materials. The searchable information tool shall be available via an interactive web page.

8

Attachment No. 1 Task 1.5-Conduct Scoping Analysis on Viability of Searchable Information Tool Task 1.5.1 PNNL shall briefly consider available approaches to creating a preliminary database that will link the highest susceptibility/lowest knowledge anticipated degradation scenarios with potential availability of ex-plant materials. As part of this subtask, PNNL shall review the status and viability of the PMMO information tool created as part of the PMMO project (conducted at PNNL under previous NRG contracts (i.e., JCN N6029, N6907). The goals of the PMMO project were to identify reactor components that could reasonably be expected to experience future degradation, estimate the susceptibility of components to various degradation mechanisms, and assess the degree of knowledge available to develop mitigative strategies. It was anticipated that this information could be used to guide regulatory actions related to license renewal and subsequent license renewal. The PMMD panel evaluated 3863 components (2203 for PWRs, 1603 for BWRs) for their susceptibility to 16 degradation mechanisms (Figure 4 below). Because of the unwieldiness of the source material, a searchable information tool (pmmd.pnl.gov) was developed to make this information usable to NRG staff and others.

Task 1.5.2 PNNL shall worlk with the NRG to create a proposal to develop a platform for the searchable database methodology (selected in Task 1.5.1) that can be supported within NRC.

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Attachment No. 1 Figure 4. Flowchart for files created and used in PMMD infotool Task 1.6- Provide Archival Summary Document of Findings from Task 1 PNNL shall analyze and review the reports generated from the work conducted under Tasks 1.1 through 1.5 and provide a stand-alone NUREG/CR documenting the major findings.

Optional Task 2 - Decision Making on Specific Research Need to Address Gaps If the Task 1 scoping study succeeds in determining previously unidentified significant knowledge gaps that need further attention, more work will be done in the context of Task 2, pending the approval from the NRC Contract Officer (CO). Thus, Task 2 is optional pending the outcome of Task 1 and requires NRC activation. The activities required for this task are:

Task 2.1 - Gap Identification PNNL shall identify specific information and technical data gaps from the execution of Task 1 and document these gaps. In identifying the gaps, PNNL shall include an examination of the current ASME B&PV Code or other industry practices that the NRC has endorsed with respect to addressing the specific degradation mechanism in the design and the assurance of the retention of the design margin during the period of licensed reactor operation time.

Task 2.2 - Determine Significance and Disposition of Technical Gaps PNNL shall determine whether or not there are any technical gaps from the execution of Task 2.1. If there are no gaps and if it is determined that the current ASME Code or other industry practices ensure that the design margin for components are adequate, recommend termination of further research in this topic by NRC. If specific information and technical gaps are identified then proceed to Task 3 after getting approval from the NRC Contract Officer (CO).

Optional Task 3 - Research Addressing Technical Gaps Related to Material Degradation If critically important information and technical gaps are identified in Task 2, Task 3 is activated after getting approval from the NRC CO. Thus, Task 3 is optional pending the outcome of Task

2. The activities required for this task are:

Task 3.1 - Recommend Specific Laboratory Experimentation and Analytical Model Development PNNL shall work with NRC subject matter experts (SMEs) to recommend specific laboratory experimentation and analytical model development, which may address the information gap identified in Task 2.1. If novel nondestructive evaluation methods (such as the next-generation acoustic emission technology which reportedly can 'hear' crack initiation) become available to identify progressing reactor material degradation by the time Task 3 is initiated, PNNL shall recommend inservice inspection (ISi) technology enablers which will be suitable for detecting the material changes resulting from different stressors. PNNL shall work with NRC SMEs to 10

Attachment No. 1 recommend the need for developing tools for detection and assessment of potential degradation of the design safety margin to independently confirm the licensee's technical basis for LTO.

Task 3.2 - Review Adequacy of Existing Codes and Standards PNNL shall conduct a review of existing applicable ASME B&PV Codes that may need to be revised as a result of Task 2.1 and PNNL shall work with NRC SMEs to engage relevant ASME Code Committees for assessing future path. PNNL shall propose other Codes and Standards that should be reviewed (such as but not limited to, ANS, ASTM, and NACE codes and standards).

Optional Task 4 - Investigate Development of Independent Decision Making Tools Task 4 is optional pending the outcome of Tasks 1 - 3. If gaps are identified under Task 2 and appropriate research needed to inform the gaps are also identified under Task 3, NRC expects that the industry will perform the needed research and provide NRC the data for regulatory decisions.

Depending on the outcome of Tasks 2 and 3 and ensuing industry research, the decision-making tool development may be complex and truly involve multi-year, multi-disciplinary long term research. It is expected, however, that the decision making tool may include: (a) Material and component condition after different stressors; (b) Better knowledge of specific degradation and its potential for reducing the design safety margin for the components; (c) Incorporation of plant data into the various material, inspection, and structural integrity evaluation models; and (f)

An integration of all these aspects into the regulatory decision making process to consider the risk contribution due to material degradation.

Specific subtasks for this task will be established later in this research. PNNL shall investigate the feasibility of developing a modern visualization confirmatory analysis research tool for aging management of safety-significant SSC degradation in NPPs. As currently envisioned, this could provide a knowledge management and strategic planning tool for conducting gap assessments and prioritizing R&D resources related to NPP LTO. This research will leverage the work previously performed by PNNL on the PMMD Information Tool, sponsored by RES.

RES/DE would benefit from a R&D gap assessment, strategic planning and knowledge management tool to enhance the tracking , disposition, resolution of technical issues that surface as industry moves towards SLR. Such a database would save staff time in addressing the degradation challenges for NPP passive components, spent fuel pools, and independent spent fuel storage installations (ISFls). The proposed LTO issues visualization tool can incorporate, up-to-date information on critical issues associated with cable, concrete and RPV aging. Work is actively progressing on developing SLR guidance documents with unresolved technical issues arising almost on a daily basis. These could be captured by the proposed service-oriented analytic framework. The existing PMMD database containing detailed information about susceptibility, knowledge, and confidence associated with hundreds of degradation scenarios can be augmented with aging risk indices, when developed by the DOE LWRS research. This will enable a better understanding of service life projections of NPP SSC.

11

Attachment No. 1 5.0 DELIVERABLES AND/OR MILESTONES SCHEDULE Except for Task 1.6 where a draft summary NUREG/CR is stipulated, all deliverables shall be in the form of technical letter reports or alternatives previously discussed and determined acceptable by the COR. Based on the detailed tasks provided in Section 4.0 of this Statement of Work, PNNL shall estimate the number of Figures/Tables or other copyrighted information from technical journals, etc. and shall incorporate this estimation in the cost proposal in addressing the SOW. PNNL shall also estimate reasonable effort by their technical editing staff in order to provide the NRC tech-edited draft final and final reports.

12

Attachment No. 1 Task Deliverable/Milestone Description (include NRG acceptance Due Date (if any)

Number criteria if applicable)

All Monthly Letter Status Report (MLSR) 20th day of each month 1.1 PNNL to provide Report 1.1. Draft TLR to NRC on Subtask (1.1) NLT 6 months after reviewing anticipated NPP LTO materials degradation and contract award prognostics 1.1 NRC to provide comments to contractor on Report 1.1 on NPP NLT 1 montlh after LTO materials degradation and prognostics receiving draft Report 1.1 from PNNL 1.1 PNNL to publish TLR Report 1.1 on materials degradation and NLT 1 month after prognostics. Deliver 12 hard copies to the NRC COR, in receiving NRC addition to an electronic file. comments 1.2 PNNL to provide Report 1.2. Draft TLR to NRC on Subtask NLT 8 months after (1 .2) concerning availability of ex-plant material and information contract award 1.2 NRC to provide comments to contractor on Report 1.2 NLT 1 month after concerning availability of ex-plant material and information receiving draft Report 1.2 from PNNL 1.2 PNNL to publish TLR Report 1.2 concerning availability of ex- NLT 1 month after plant material and information. Deliver 12 hard copies to the receiving NRC NRC COR, in addition to an electronic file. comments 1.3 PNNL to provide Report 1.3 {consisting of questionnaire and NLT 10 months after interview results) to NRC on Subtask (1 .3) concerning interest of contract award prospective NRC stakeholders in a systematic approach to harvesting ex-plant materials 1.3 NRC to provide comments to contractor on Report 1.3 NLT 1 month after concerning interest of prospective NRC stakeholders in a receiving Report 1.3 systematic approach to harvesting ex-plant materials from PNNL 1.4 PNNL to provide Report 1.4 {consisting of questionnaire and NLT 14 months after interview results) to NRC on Subtask (1.4) concerning interest of contract award prospective external stakeholders in a systematic approach to harvesting ex-plant materials 1.4 NRC to provide comments to contractor on Report 1.4 NLT 1 month after concerning interest of prospective external stakeholders in a receiving Report 1.4 systematic approach to harvesting ex-plant materials from PNNL 1.5.1 PNNL to provide Report 1.5.1 to NRC on Subtask { 1.5.1) with NLT 16 months after suggested alternatives for creating a prognostic tool to track and contract award resolve critical technical issues for SLR 1.5.1 NRC to provide comments to contractor on Report 1.5.1 NLT 1 month after concerning alternatives for creating a prognostic tool to track receiving Report 1.5.1 and resolve critical technical issues for SLR from PNNL 13

Attachment No. 1 1.6 Summary Report 1.6. Draft NUREG/CR to NRC including NLT 20 months after information from Reports 1.1 through 1.5.2. contract award (Note: At the discretion of COR, a decision may also be made to publish Summary Report 1 as a TLR rather than as an NUREG/CR, depending on the significance of the literature review and research assessment results).

1.6 The Contractor will make a technical presentation to the NRC When the draft staff on Summary Report 1.6 at NRC Headquarters in Rockville, Summary Report 1.6 MD. is delivered to NRC.

1.6 NRC to provide comments to contractor on Summary Report NLT 2 months after 1.6. receiving draft Summary Report 1.6 from PNNL 1.6 DOE Contractor to publish Summary Report 1.6 as NUREG/CR. NLT 2 months after Deliver 12 hard copies to the NRC COR, in addition to an receiving NRC electronic file. comments Task 2 is optional pending outcome of Task 1.

2 PNNL to provide Report 2 Draft TLR to NRC based on results NLT 24 months after from Subtask (2.1) concerning technical gap identification and original contract award subtask (2.2) determination of significance and disposition of gaps 2 NRC to provide comments to contractor on Report 2 concerning NLT 1 month after technical gap identification, significance, and disposition receiving draft Report 2 from PNNL 2 PNNL to publish TLR Report 2 technical gap identification, NLT 1 month after significance, and disposition. Deliver 12 hard copies to the NRC receiving NRC COR, in addition to an electronic file. comments Task 3 is optional pending outcome ofTask 2 3.1 PNNL to provide Report 3.1 Draft TLR to NRC based on results NLT 46 months after from Subtask (3. 1) concerning specific laboratory contract award experimentation and analytical model development 3.1 NRC to provide comments to contractor on Report 3.1 NLT 1 month after concerning specific laboratory experimentation and analytical receiving draft Report model development 3.1 from PNNL 3.1 PNNL to publish TLR Report 3.1 concerning specific laboratory NLT 1 month after experimentation and analytical model development. Deliver 12 receiving NRC hard copies to the NRC COR, in addition to an electronic file. comments 3.2 PNNL to provide Report 3.2 Draft TLR to NRC reviewing NLT 46 months after adequacy of existing codes and standards for SLR contract award 3.2 NRC to provide comments to contractor on Report 3.2 reviewing NLT 1 month after adequacy of existing codes and standards for SLR receiving draft Report 3.2 from PNNL 3.2 PNNL to publish TLR Report 3.2 reviewing adequacy of existing NLT 1 month after codes and standards for SLR. Deliver 12 hard copies to the receiving NRC NRC COR, in addition to an electronic file. comments Task 4 is optional pending outcome of Task 1 and partially pending on Task 2 and 3.

14

Attachment No. 1 4 PNNL to provide Report 4 Draft TLR to NRC documenting NLT 46 months after development of prognostic tool to track and resolve critical SLR contract award technical issues 4 NRC to provide commen ts to contractor on Report 4 reviewing NLT 1 month after development of prognostic tool to track and resolve critical SLR receiving draft Report technical issues 3.2 from PNNL 4 PNNL to publish TLR Report 4 reviewing develop ment of NLT 1 month after prognostic tool to track and resolve critical SLR technical issues receiving NRC R. Deliver 12 hard copies to the NRC COR, in addition to an comments electronic file.

6.0 TECHNICAL AND OTHER SPECIAL QUALIFICATIONS REQUIRED Specific qualific ations for this effort include senior materia ls enginee rs and metallu rgists who have in-depth knowle dge of reactor pressur e vessel and core internal materia ls subject ed to irradiation and stress at elevate d temper ature, and effects of water chemis try on structur al reactor materia ls. The personn el involve d should have in-dept h experie nce, knowle dge, and demons trated contribu tions in the areas of mechan ical deform ation, materia l degrad ation phenom ena, such as corrosio n, stress corrosio n crackin g and irradiati on effects. The contrac t personnel should be well-ve rsed in the use of nuclear power plant ASME B&PV Codes and Standa rds, Industry Guidan ce Docum ents, such as those of NEI, EPRI, NRC's Regula tory Guides and NRC's License Renew al Guidan ce Docum ents (such as NUREG s 1800, 1801, and 1950) Informa tion Notice (IN), Regula tory Issue Summa ry (RIS), Generic Letter (GL), Generic Issue (GI) for licensing review by the NRC staff.

The contrac t personnel should also be aware of the safety evaluat ion reports (SER) written by the NRC staff on industry guidanc e docume nts, as applica ble. The contrac t personn el should have previou s experie nce develop ing approp riate softwar e architec ture for propose d R&D planning tool.

7.0 MEETINGS AND TRAVEL The PNNL Principal Investig ator and one other engine er shall visit the NRC Headqu arters in Rockville, MD and presen t the overall researc h outcom e to the staff and share in technic al discuss ions. Any sugges tions from the staff, as approp riate, may be conside red for the final report by the Pl. No other domest ic or foreign travel is permitt ed under the initial scoping study.

8.0 REPORTING REQUIREMENTS PNNL is respons ible for structur ing the delivera ble to follow agency standar ds. The current agency standar d is Microso ft Office Suite 2010. The current agency Portabl e Docum ent Format (PDF) standar d is Adobe Acroba t 9 Professional. Deliver ables shall be submitt ed free of spelling and gramma tical errors and conform to require ments stated in this section .

Month ly Letter Status Report s In accorda nce with Manag ement Directive 11 .7, NRC Proced ures for Placem ent and Monitoring of Work with the U.S. Departm ent of Energy, PNNL shall electron ically submit a Monthly Letter Status Report (MLSR ) by the 20 th day of each month to Amy Hull, the Contracting Officer 15

Attachment No. 1 Representative (COR), to Matthew Hiser and Joseph Kanney, the technical monitors, with copies to the Contracting Officer (CO) and the Office Adm inistration/Division of Contracts to ContractsPOT.Resource@nrc.gov. If a project is a task ordering agreement, a separate MLSR shall be submitted for each task order with a summary project MLSR, even if no work has been performed during a reporting period. Once NRC has determined that all work on a task order is completed and that final costs are acceptable, a task order may be omitted from the MLSR.

MLSR should be distributed additionally to the Chief, Corrosion and Metallurgy Brancln, RES, the Director, Division of Engineering, RES. Other required distribution will be communicated at the start of this research program.

The MLSR shall include the following : agreement number; task order number, if applicable; job code number; title of the project; project period of performance; task order period of performance, if applicable; COR's name, telephone number, and e-mail address; full name and address of the performing organization; principal investigator's name, telephone number, and e-mail address; and reporting period. At a minimum, the MLSR shall include the information discussed in Attachment 1. The preferred MLSR format can also be found in Attachment 1.

9.0 PERIOD OF PERFORMANCE The period of performance for this work is September 17, 2015 - September 30, 2019.

10.0 CONTRACTING OFFICER'S REPRESENTATIVE The COR monitors all technical aspects of the agreement/task order and assists in its administration. The COR is authorized to perform the following functions: assure that the DOE Laboratory performs the technical requirements of the agreement/task order; perform inspections necessary in connection with agreement/task order performance; maintain written and oral communications with the DOE Laboratory concerning technical aspects of the agreemenUtask order; issue written interpretations of technical requirements, including Government drawings, designs, specifications; monitor the DOE Laboratory's performance and notify the DOE Laboratory of any deficiencies; coordinate availability of NRG-furnished material and/or GFP; and provide site entry of DOE Laboratory personnel.

Contracting Officer's Representative Name: Dr. Amy B. Hull Agency: U.S. Nuclear Regulatory Commission Office: Office of Nuclear Regulatory Research Mail Stop: T-10D49 Washington, DC 20555-0001 E-Mail: amy.hull@nrc.gov Phone: 301 -251-7656 Alternate Contracting Officer's Representative Name: Matthew Hiser Agency: U.S. Nuclear Regulatory Commission Office: Office of Nuclear Regulatory Research Mail Stop: T-10 A36 Washington, DC 20555-0001 16

Attachme nt No. 1 E-Mail: Matthew.Hiser@nrc.gov Phone: 301-251-7601 11 .0 MATERIALS REQUIRED (TYPE N/A IF NOT APPLICABLE)

N/A 12.0 NRC-FURNISHED PROPERTY/MATERIALS PNNL will transfer NRC furnished property and materials acquired under previous contracts (i.e., JCN N6029, N6907) to this task order. NRC will provide additional informat ion from EMDA and SLR databases.

13.0 RESEARCH QUALITY (TYPE N/A IF NOT APPLICABLE)

The quality of NRC research programs are assessed each year by the Advisory Committee on Reactor Safeguards. Within the context of their reviews of RES programs, the definition of quality research is based upon several major characteristics:

Results meet the objectives (75% of overall score)

Justification of major assumptions (12%)

Soundness of technical approach and results (52%)

Uncertainties and sensitivities addressed (11 %)

Documentation of research results and methods is adequat e (25% of overall score)

Clarity of presentation (16%)

Identification of major assumptions (9%)

It is the responsibility of the DOE Laboratory to ensure that these quality criteria are adequately addressed throughout the course of the research that is performed. The NRC COR shall review all research products with these criteria in mind.

14.0 STANDARDS FOR CONTRACTORS WHO PREPARE NUREG -SERIES MANUSCRIPTS (TYPE N/A IF NOT APPLICABLE)

The U.S. Nuclear Regulatory Commission (NRC) began to capture most of its official records electronically on January 1, 2000. The NRC will capture each final NUREG-series publication in its native application. Therefore, please submit your final manusc ript that has been approved by your NRC Project Manager in both electronic and camera-ready copy.

The final manuscript shall be of archival quality and comply with the requirements of NRC Manage ment Directive 3.7 "NUREG-Series Publications." The document shall be technically edited consistent with NUREG -1379, Rev. 2 (May 2009) "NRC Editorial Style Guide."

The goals of the "NRC Editorial Style Guide" are readability and consistency for all agency documents.

All format guidance , as specified in NUREG -0650, "Preparing NUREG-Series Publicat ions,"

Rev. 2 (January 1999), will remain the same with one exception. You will no longer be required to include the NU REG-series designator on the bottom of each page of the manuscr ipt. The NRC will assign this designa tor when we send the camera-ready copy to the printer and will 17

Attachment No. 1 place the designator on the cover, title page, and spine. The designator for each report will no longer be assigned when the decision to prepare a publication is made. The NRC's Publishing Services Branch will inform the NRC Project Manager for the publication of the assigned designator when the final manuscript is sent to the printer.

For the electronic manuscript, the Contractor shall prepare the text in Microsoft Word, and use any of the following file types for charts, spreadsheets, and the like.

File Types to be Used for NUREG-Series Publications File Type File Extension MicrosoftWord .doc Microsoft PowerPoint .ppt MicrosottExcel .xis MicrosoftAccess .mdb Portable Document Format .pdf This list is subject to change if new software packages come into common use at NRC or by our licensees or other stakeholders that participate in the electronic submission process. If a portion of your manuscript is from another source and you cannot obtain an acceptable electronic file type for this portion (e.g ., an appendix from an old publication), the NRC can, if necessary, create a tagged image file format (file extension.tit) for that portion of your report.

Note that you should continue to submit original photographs, which will be scanned, since digitized photographs do not print well.

If you choose to publish a compact disk (CD) of your publication, place on the CD copies of the manuscript in both (1) a portable document format (PDF); (2) a Microsoft Word file format, and (3) an Adobe Acrobat Reader, or, alternatively, print instructions for obtaining a free copy of Adobe Acrobat Reader on the back cover insert of the jewel box.

15.0 OTHER CONSIDERATIONS (TYPE N/A IF NOT APPLICABLE)

References

1. Bond LJ, SR Doctor, and TT Taylor. 2008. Proactive Management of Materials Degradation -A Review of Principles and Programs. PNNL-17779, Pacific Northwest National Laboratory, Richland, WA

2. Bond, LJ, TT Taylor, SR Doctor, AB Hull, and SH Malik, (2008) Proactive Management of Materials Degradation for nuclear power plant systems. Proc. Int. Cont. Prognostics and Health Management 2008, Denver, CO, October 6-9. IEEE Reliability Society,# OP-20-01 120 18

Attachment No. 1

3. Chopra, OK, et al, Managing Aging Effects on Dry Cask Storage Systems for Extended Long-Term Storage and Transportation of Used Fuel, Rev. 0, FCRD-USED-2012-000119, 2012.

4. EPRI 3002000576, Long-Term Operations: Assessment of R&D Supporting AMPs for L TO, Aug . 2013 (80pp).

5. NEI, Roadmap for Subsequent License Renewal, Dec. 2013. (45pp)

6. NEI, Second License Renewal Roadmap, May 2015. (22pp).

7. NUREG/CR-6923, Expert Panel Report on Proactive Materials Degradation Assessment, 2007 (3895pp, ML063520517)

8. NUREG/CR-7153, Expanded Materials Degradation Assessment, 5 volumes, October 2014 (861 pp)

9. SECY-14-0016, Ongoing Staff Activities to Assess Regulatory Considerations for Power Reactor Subsequent License Renewal, January 31 , 2014 (25pp)

10. Taylor, WB, CE Carpenter, KJ Knobbs, S Malik, Using Technology to Support Proactive Management of Materials Degradation for the U.S. Nuclear Regulatory Commission, Proceedings of the ASME Pressure Vessels & Piping Division/K-PVP Conference, PVP 2010, July 18-22, 2010. Bellevue, WA, USA Paper PVP2010-26063.

11 . The Scalable Reasoning System: Lightweight Visualization for Distributed Analytics, IEEE Symposium on Visual Analytics Science & Technology, 978-1-4244-2935-6/08 Access to Non-NRC Facilities/Equipment (Type N/A if no,t applicable)

NIA Applicable Publications (Type N/A if not applicable)

N/A Controls over document handling and non-disclosure of materials (Type NIA if not applicable)

NIA 19

Note to requester: The attachment is immediately following this email, which is also publicly available in From: Hiser, Matthew ADAMS at ML16155A417.

Sent: Wed, 4 May 2016 18:12:41 +0000 To: Shaffer, Sarah

Subject:

FW: Amend REQ RES-16-0295 Attachments: NRC-HQ-60-15-T-0023_M2.pdf Hi Sarah, Looks like AMD already processed the $20K funding for PNNL for the harvesting project. I guess this means we would need to submit another REQ to fund the remainder?

Thanks!

Matt Matthew Hiser Materials Engineer US Nuclear Regulatory Commission I Office of Nuclear Regulatory Research Division of Engineering I Corrosion and Metallurgy Branch Phone: 301-4 /5-2454 I Office: TWFN I0D62 Matthew.Hiser@nrc.gov From: Gunter-Henderson, Morie Sent: Wednesday, May 04, 2016 2:08 PM To: Hiser, Matthew <Matthew. Hiser@nrc.gov>

Subject:

RE: Amend REQ RES-16-0295 Hi Matt, The above mentioned requisition was processed yesterday as seen attached.

Morie From: Hiser, Matthew Sent: Wednesday, May 04, 2016 2:02 PM To: Gunter-Henderson, Morie < Morie.Gunter-Henderson@nrc.gov>

Subject:

FW: Amend REQ RES-16-0295 Hi Morie, I just wanted to check on the status of this REQ:

RES recently released RES-16-0295 to provide $20K in incremental funding for NRC-HQ 15-T-0023 under NRC-HQ-25-14-D-0001.

However, we would like to amend this REQ to provide additional funding up to the ceiling of the contract. Please confirm that you have not yet processed the award of this action and we will amend the REQ for the correct amount.

Thanks!

Matt Matthew Hiser Materials Engineer US Nuclear Regulatory Commission I Office of Nuclear Regulatory Research Division or Engineering I Corrosion and Metallurgy Branch Phone: 301-415-24541 Office: TWFN 10D62 Matthew.Hiser@nrc.gov From: Turner, Michael Sent: Monday, May 02, 2016 3 :30 PM To: Hiser, Matthew <Matthew. Hiser@nrc.gov>

Cc: Gunter-Henderson, Morie <Morie.Gunter- Henderson@nrc.gov>

Subject:

FW: Amend REQ RES-16-0295

Matt, I am passing this on to Morie Gunter-Henderson who is the CO for this action.

Morie - please see Matt's question below.

Regards, Mike From: Hiser, Matthew Sent: Monday, M ay 02, 2016 9 :52 AM To: Turner, Michael <Michael.Turner@nrc.gov>

Cc: Shaffer, Sarah <Sarah .Shaffer@nrc.gov>; Frankl, Istvan <lstvan.Frankl@nrc.gov>; Hull, Amy

<Amy.Hull@nrc.gov>

Subject:

Amend REQ RES-16-0295 Hi Michael, RES recently released RES-16-0295 to provide $20K in incremental funding for NRC-HQ 15-T-0023 under NRC-HQ-25-14-D-0001.

However, we would like to amend this REQ to provide additional funding up to the ceiling of the contract. Please confirm that you have not yet processed the award of this action and we will amend the REQ for the correct amount.

Thanks!

Matt Matthew Hiser

Materials Engineer US uclear Regulatory Commission I Office of uclear Regulatory Research Division of Engineering I Corrosion and Metallurgy Branch Phone: 30 /-4/5- 2454 I Office: TWFN I0D62 Matthew.Hiser@nrc.gov

2.0ROERNO INTERAGENC Y AGREEMENT 11 L-V.NO NRC-HQ-60-15-T-0023/M0002 13 REOUISl'TIOHNO 1* SOUCfTATION NO I""~ OF I 2 RES-16-0295

5. EFFECTll/E DA~

18 AWAADDA~ 7 PERIOO OF PERFORMANCE 05/03/2016 05/03/2016 09/17/2015 TO 09/30/2019 8 SER\IICINGAGENCY 8 OEUIIER TO PACIFIC NORTHWEST NAT LAB AMY HULL ALC: US NUCLEAR REGULATORY COMMISSION DUNS: +4 : OFFICE Of NUCLEAR REGULATORY RESEARCH US DEPARTMENT Of ENERGY 11555 ROCKVILLE PIKE PACIFIC NORTHWEST SITE OFFICE ROCKVILLE MD 20852 PO BOX 350 MS K9-42 RICHLAND WA 99352 POC GENICE MADERA 18..EPHONE NO 509-372-401 0 10 REQUCSTING AGENCY 11 INVOOCEOFFICE ACQUISITION MANAGEMENT DIVISION US NUCLEAR REGULATORY COMMISSION ALC: 31000001 DUNS: 040535809 +4: ONE WHITE FLINT NORTH US NUCLEAR REGULATORY COMMISSION 11555 ROCKVILLE PIKE ONE WHITE FLINT NORTH MAILSTOP 03-El7A 11555 ROCKVILLE PIKE ROCKVILLE MD 20852-2738 NRCPAYMENTSNRCGOV ROCKVILLE MD 20852-2738 POC Morie Gunter-Hen derson 18..EPHONE NO 301-415-792 4

12. ISSUHO OfFICE 13 l.EG"ll.AT11/E NJTHORIT'I US NRC - HQ Energy Reorganiza tion Act of 1974 ACQUISITION MANAGEMENT DIVISION MAIL STOP TWFN-5E03 WASHINGTON OC 20555-0001 14 PROJECTIO 15 PROJECTTITI.E SEE BLOCK UB 18 ACCOUNTING 0,.TA 2016-X0200- FEEBASED- 60-60D001-l l-6-213-1032 -253D 17 18 18 20 21 22.

ITEMNO SUPP\..ESISERIIICES au.aHT1lY UNT UNTPRICE AMOUNT Task Order Title : Strategic Approach for Obtaining Material and Component Aging Information .

Master IAA : NRCHQ2514D0001 Summary of Changes:

The purpose of this modificatio n is to provide incrementa l funding in the amount of $20,000.00, thereby increasing the total obligations of thi~

task order from $133 , 165.00 to $153,165 . 00.

Obligation with this Action : $20 , 000 . 00 Continued ...

23 ""YMCNT PflOlll9ION8 24 TOTAA.AMOUNT

$20,000 . 00 I

. /

~rulEOF r-:7?(R£041E

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MORIE E. GU TER-HENDERSON

, ~ DATE

~ \6

IMNO NRC-HQ-60-15-T-0023/M0002 ICRreRNO

""'-GE 2

OF I2 rrotal Obligations to Date : $153 , 165 . 00

~uthorized Ceiling Amount : $163 , 529 . 00

~ 11 other terms and conditions remain unchanged ,

including the authorized cost ceiling amount of

$163,529 . 00.

DUNS : 040535809 IALC : 31000001 TAS : 31X0200 . 320

From: Moyer, Carol Note to requester: The attachment is Sent: Mon, 9 Nov 2015 23:05:08 +0000 immediately following this email.

To: Carpenter, Gene

Subject:

RE: Stds Board mtg Attachments: 2015-11-- NRC Report to ASME - ATL (FINAL).pdf Attached.

I will bring a hard copy tomorrow.

From: Carpenter, Gene Sent: Monday, November 09, 2015 3:52 PM To: Moyer, Carol <Carol.Moyer@nrc.gov>

Subject:

Re: Stds Board mtg May I impose on you for the most recent ASME NRC report? Thanks.

Original Message --------

From: "Moyer, Carol" < Carol.Moyer@nrc.gov>

Date: Mon, November 09, 2015 12:51 PM -0500 To: "Carpenter, Gene" < Gene.Carpenter@nrc.gov>

Subject:

Stds Board mtg

Gene, It looks like it's you & me for the ANS Standards Board meeting tomorrow. I plan to go straight there in the AM .

-Carol From: Lund, Louise Sent: Monday, November 09, 2015 11:38 AM To: Moyer, Carol <Carol.Moyer@nrc.gov>

Subject:

Automatic reply: Request: Concurrence package for ANS letter I'm out of the office November 4 - 13, returning November 16. If you need immediate assistance, please contact the RES/ORA front office at 301-4 I 5-0452.

thanks, Louise

NRC Report to ASME November 2015 NRC Report for ASME Code Meetings November 2015-Atlanta, GA Contents

1. Amendment to 10 CFR 50.55a -ASME Code Edition/Addenda ......................... ............. 2

2. ASME Code Case Rulemaking/Regulatory Guides .......................................................... 2

3. Operating Plant Issues and Material Degradation ............................................................ 4

4. NRO DCIP Quality Assurance and Vendor Inspection Branch Activities .......................... 5

5. New Reactor Licensing Activities ..................................................................................... 7

6. Multinational Design Evaluation Program (MDEP) Activities ...........................................10

7. 10 CFR Part 21 Rulemaking ........................................ ...................................................12

8. Commercial Calibration Services Status ............................................ ................ ............. 13

9. NRC staff Review of EPRI 1025243 Guideline for Commercial-Grade Design and Analysis Computer Programs ................................................................................... 14

10. EPRI Development of Revised Dedication Guidance for Commercial-Grade Items for Use in Nuclear Safety-Related Applications ..................................................... 15

11. NRC Staff Interface with Nuclear Utilities Procurement Issues Committee (NUPIC) ....... 16

12. Counterfeit, Fraudulent and Suspect Items (CFSI) Commission Paper ........................... 16

13. License Renewal Activities ............................................................... .. .............................17

14. New Generic Letters ....................................................... .................. ...............................20

15. New Information Notices .................................................................................................20

16. New Regulatory Issue Summaries ...................... ............................... .................... .. ....... 20

17. NRC Publications of Potential Interest to ASME ............................................................. 21

18. Upcoming Public Meetings of Potential Interest to ASME ...............................................21 NRC Report to ASME November 2015

1. Amendment to 10 CFR 50.55a - ASME Code Edition/Addenda Current ASME Edition/Addenda The NRC has approved:

Section Ill, Division 1 and Section XI, Division 1 of the Boiler and Pressure Vessel Code through the 2008 Addenda (76 FR 36232).

The Operation and Maintenance of Nuclear Power Plants (OM Code) through the 2006 Addenda (76 FR 36232).

Next ASME Edition/Addenda The next proposed amendment to 10 CFR 50.55a includes:

The 2009 Addenda, the 2010 Edition, 2011 Addenda , and the 2013 Edition of the Boiler and Pressure Vessel Code.

The 2009 Edition, 2011 Addenda and 2012 Edition of the Operation and Maintenance of Nuclear Power Plants (OM Code).

Section XI Code Case N-824 will be directly listed in 50.55a as conditionally approved for use.

Section XI Code Case N-729-4 will be directly listed in 50.55a as required with conditions.

Section XI Code Case N-770-2 will be directly listed in 50.55a as required with conditions.

The proposed rule was published in the Federal Register on Friday, September 18, 2015.

The 75 day public comment period closes on December 2, 2015 (link to the FRN:

http://www.gpo.gov/fdsys/pkg/FR-2015-09-18/pdf/2015-23193.pdf). ASME made the materials that are incorporated by reference available in a read-only format for the public comment period at the following link: http://qo.asme.org/NRC.

2. ASME Code Case Rulemaking/Regulatory Guides Current RG Publications On November 5, 2014 a final rule was published in the Federal Register (79 FR 65776) that incorporates by reference the Regulatory Guides (RGs) listed below:

Supplements Addressed : Supplements 1 through 10 to the 2007 Edition Effective date for the RGs: December 5, 2014 NRC Report to ASME November 2015

RG 1.84, Revision 36, "Design, Fabrication, and Materials Code Case Acceptability, ASME Section Ill" (ADAMS Accession No. ML13339A515).

RG 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1" (ADAMS Accession No. ML13339A689).

RG 1.192, Revision 1, "Operation and Maintenance Code Case Acceptability, ASME OM Code" (ADAMS Accession No. ML13340A034).

In addition, on November 5, 2014, a final guide was published in the Federal Register (79 FR 65776):

RG 1.193, Revision 4, "ASME Code Cases Not Approved for Use" (ADAMS Accession No. ML13350A001 ).

Next RG Publications The staff is completing the proposed Code Case Rulemaking package and expects a publication in the Federal Register in November 2015. Following the publication of the proposed rule will be a 75 day public comment period.

ASME Code Code Case Supplements Addressed: Supplement 11 to the 2007 Edition through Supplement 1Oto the 201 O Edition

Draft Revision 37 of RG 1.84

Draft Revision 18 of RG 1 . 147

Draft Revision 5 of RG 1.193 Additional Code Cases considered for this rulemaking package at the request of ASME that are not listed in aforementioned Supplements:

N-694-2 (Supp. 1 to the 2013 Edition) "Evaluation Procedure and Acceptance Criteria for PWR Reactor Vessel Head Penetration Nozzles"Section XI

N-825 (Supp. 3 to the 2013 Edition) "Alternative Requirements for Examination of Control Rod Drive Housing Welds"Section XI

N-845 (Supp. 6 to the 2013 Edition) "Qualification Requirements for Bolts and Studs"Section XI OM Code Code Cases Addressed: 2009 Edition through 2012 Edition

Draft Revision 2 of RG 1.192 The NRC staff also initiated the review of the next draft RGs that will address the Code Cases published in Supplement 11 to the 2010 Edition through Supplement O to the 2015 Edition of the ASME Code.

NRC Report to ASME November 2015 Standards Used in RGs and Other Guidance The NRC has placed on its website a series of lists of consensus standards, including those published by ASME, that are referenced in Regulatory Guides, in Inspection Manuals and Procedures, and in the LWR Standard Review Plan. The lists may be found at this website:

http://www.nrc.gov/about-nrc/regulatory/standards-dev/consensus.html.

3. Operating Plant Issues and Material Degradation MRP-146 Thermal Fatigue in Norma/Iv Non-lsolable Reactor Coolant System Branch Lines Inspections Recent operating experience has indicated that the guidance provided by MRP-146 does not conservatively predict locations where thermal fatigue may occur. Plants have experienced thermal fatigue flaws in locations that had been screened out by the generic model provided by MRP-146. MRP assembled a response team and issued interim guidance to utilities and have now issued revised guidance for MRP 146 which is under review.

Peening The NRC staff is currently reviewing the MRP-335, Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement, which discusses peening as a mitigation technique to permit reduction in inspection frequency for Alloy 600 surfaces exposed to PWR reactor coolant. The NRC staff has issued the draft safety evaluation and expects the final safety ,evaluation to be issued by the end of the year. There will be a public meeting to discuss peening issues approximately December 9, 2015.

Operational Leakage On March 23, 2015, ASME sent the NRC a letter describing the conclusion of their activities for addressing operational leakage of pressure retaining components (ADAMS Accession No. ML15099A624). In view of ASME's Pressure Boundary Leakage project team's conclusions, the NRC sent a letter back to ASME on July 14, 2015 stating that they will evaluate the necessity of additional regulatory activities to address operational leakage (ADAMS Accession No. ML15188A057). The NRC staff is currently working toward providing a regulatory framework for operational leakage of ASME Code Class 2 & 3 piping and components in the rulemaking which will incorporate the 2015 Edition of the B&PV Code.

NRC Report to ASME November 2015

4. NRO DCIP Quality Assurance and Vendor Inspection Branch Activities NRO Vendor Inspection The NRO vendor inspection program is described in Inspection Manual Chapter (IMC) 2507, "Vendor Inspections." This IMC was last updated on October 3, 2013. This IMC is implemented by various Inspection Procedures (IPs) including:

IP 43002: Routine Inspections of Nuclear Vendors; IP 43003: Reactive Inspections of Nuclear Vendors; IP 43004: Inspection of Commercial-Grade Dedication Programs; IP 43005: NRC Oversight of Third Party Organizations Implementing Quality Assurance Requirements; IP 36100: Inspection of 10 CFR Part 21, Programs for Reporting Defects and Noncompliance; IP 37805: Engineering Design Verification Inspections; IMC 0617: Vendor and Quality Assurance Implementation Inspection Reports; and IMC 2506: Construction Reactor Oversight Process General Guidance Basis Document FY 15 Vendor Inspection Plans

AP1000 modular construction

AP1000 mechanical and electrical qualification test programs

Digital Instrumentation and Control for AP1000

Valve and pump manufacturing

Commercial-grade dedication organizations Vendor Inspection Reports Issued, Completed, and Planned Inspections

Crane Nuclear, Inc. IL - issued

WEC, Warrendale, PA - issued

SPX/Penn State University, PA- issued

Fisher Control, Marshalltown, IA - issued

Specialty Maintenance & Construction, FL - issued

Kinetrics, Toronto, CA - issued

Thermo-Fisher Scientific, San Diego, CA - issued

C&D Technologies, Blue Bell, PA - completed

NuScale Power, LLC, OR - issued

Curtiss Wright-QualTech, Cincinnati, OH - planned

Curtiss Wright EMO, Cheswick, PA - planned

Carboline, St. Louis, MO - planned

Canberra Industries, Meriden, CT - planned

Valcor Engineering, Springfield, NJ - planned NRC Report to ASME November 2015

Aecon Industrial , Cambridge, Canada - planned

GE Dresser Consolidated, Alexandria, LA - planned Previously issued NRC inspection and trip reports are located at:

http://www.nrc.gov/reactors/new-reactors/oversighUquality-assurance/vendor-insp/insp-reports.html New Vendor Inspection Qualitv Assurance Website Links The NRC has implemented website pages to make it easier to become familiar with and follow vendor inspection and QA related activities:

http://nrcweb.nrc.gov:400/reactors/new-reactors/oversiqhUguality-assurance/vendor-insp.html.

As part of the Vendor Outreach and Communications Strategy, the NRC is planning its 2016 Biannual Vendor Workshop in coordination with NUPIC vendor meeting, scheduled to be held in St. Louis, MO on June 23-25, 2016. Members of the nuclear industry are encouraged to submit topics of interest for the 2016 Biannual Vendor Workshop to NRC representatives, Richard McIntyre (richard.mcintyre@nrc.gov), Kerri Kavanagh (kerri.kavanagh@nrc.gov), or Andrea Keim (andrea.keim@nrc.gov) by January 31, 2016.

Presentations from the 2014 Vendor Workshop and past workshops are available on our public website at the below link.

http://www.nrc.gov/reactors/new-reactors/oversighUquality-assurance/vendor-oversight.html The Frequently Asked Questions (FAQ) page addresses Quality Assurance for New Reactors and currently has three main categories: 10 CFR Part 21 FAQs, Commercial Grade Dedication FAQs, and Enforcement FAQs. The page provides quick links to questions we have received in the past about the mentioned topics:

http://www.nrc.gov/reactors/new-reactors/oversighUguality-assurance/gual-assure-fags.html The web page link below serves as a categorization tool and provides a list of all applicable QA Inspections for New Reactor Licensing and Vendor QA Inspection reports that have either a Notice of Nonconformance (NON) or Notice of Violation (NOV) within a specific criterion of 10 CFR 50 Appendix B or 10 CFR Part 21 related issue. The page is routinely updated with every new inspection report that is released:

http://www.nrc.gov/reactors/new-reactors/oversiqhUquality-assurance/nonconformances-violations.html The web page link below describes the vendor inspection program (VIP). The VIP verifies that reactor applicants and licensees are fulfilling their regulatory obligations with respect to providing effective oversight of the supply chain. It is accomplished through a number of activities, including: performing vendor inspections that will verify the effective implementation of the vendor's quality assurance program, establishing a strategy for vendor identification and selection criteria, and; ensuring vendor inspectors obtain NRC Report to ASME November 2015 necessary knowledge and skills to perform inspections. In addition, the VIP addresses interactions with nuclear consensus standards organizations, industry and external stakeholders, and international constituents:

http://pbadupws.nrc.gov/docs/ML 1527/ML15272A080 .pdf.

5. New Reactor Licensing Activities As of October 13, 2015, the status of new reactor licensing under 10 CFR Part 52 is as follows:

Design Certification NRC has issued five design certifications to date (ABWR, System 80+, AP600, AP1000 and ESBWR). These are certified in 10 CFR Part 52, Appendices A, B, C, D, and E respectively.

The NRC staff's review of the AREVA's EPR (evolutionary pressurized-water reactor design from France) is suspended at the request of the applicant in its letter dated February 25, 2015, until further notice.

The NRC staff's review of the Mitsubishi Heavy Industries' US-APWR design certification application (for an advanced pressurized-water reactor design from Japan) is currently on hold at the request of the applicant except for a few key areas.

The NRC staff completed its review of General Electric-Hitachi's ESBWR (first passive BWR) and issued its final safety evaluation report (FSER) in March 2011 . On March 24, 2011 , the NRC issued in the Federal Register a proposed rule (76 FR 16549) for public comment on the ESBWR design certification. T he NRC final rule adding Appendix E to 10 CFR Part 52 to certify the ESBWR standard design was published on October 15, 2014 in the Federal Register (79 FR61983) and became effective on November 14, 2014.

The Korea Hydro and Nuclear Power (KHNP) submitted a standard design certification application for its APR- 1400 standard plant design to the NRC on September 30, 2013.

The NRC staff conducted an acceptance review of the application for completeness, technical adequacy, and acceptability for docketing. In a letter to KHNP dated December 19, 2013, the NRC staff discussed the results of its acceptance review. The NRC noted that it decided not to accept the application for docketing at that time because the application was not ready in several key areas. The NRC staff continued pre-application interactions with KHNP to support preparation of a complete application by December 2014. On December 23, 2014, KHNP resubmitted the standard design certification application for its APR-1400 design. The NRC staff accepted the APR1400 design certification application for docketing in its letter dated March 4, 2015, based on its determination that the application is sufficiently complete and technically adequate to allow the NRC staff to conduct its detailed technical review.

In addition, the NRC staff is reviewing two applications for design certification renewal:

NRC Report to ASME November 2015

ABWR GE-Hitachi (application submitted on December 7, 2010)

ABWR GE-Toshiba (Revision 1 to application submitted on June 22, 2012)

Early Site Permits (ESPs)

NRC has issued four ESPs (Clinton, Grand Gulf, North Anna, and Vogtle). The NRC's issuance of the Vogtle ESP on August 26, 2009, was the first based on a specific technology (AP-1000) and the first to include a limited-work authorization (LWA).

The NRC received an application for an ESP for the Victoria County Station submitted by Exelon on March 25, 2010. The site is located in Victoria County, Texas, with no specific technology selected. On August 28, 2012, Exelon requested withdrawal of the Victoria County Station ESP application from the docket. By letter dated October 3, 2012, NRC accepted the applicant's request, and the application was withdrawn .

The NRC received an ESP application for the PSEG site in New Jersey (same site as Hope Creek and Salem 1&2). The ESP application was tendered on May 25, 2010, and was docketed on August 4, 2010. This application uses the Plant Parameter Envelope (PPE) approach which means no specific reactor design has been selected.

Combined License (COL) Applications NRC is currently review1ing 9 COL applications (14 new reactor units):

- 1 ABWR: South Texas Project 3 and 4

- 3 AP-1000: William S. Lee Station 1&2, Shearon Harris 2&3*, Levy County 1&2, Bellefonte 3&4*, and Turkey Point 6&7

- 2 ESBWR: Fermi 3, North Anna 3, Grand Gulf 3*, River Bend 3*,

Victoria County 1 and 2**

- 2 EPR: Calvert Cliffs 3**, Bell Bend*, Nine Mile Point 3**, Callaway 2*

1 US-APWR: Comanche Peak Units 3 and 4

NRC staff review suspended at request of applicant.

- Application withdrawn.

On June 8, 2015, Unistar requested to withdraw the Calvert Cliffs, Unit 3 combined license application.

On April 25, 2013, Dominion Virginia Power revised iits technology selection from the US-APWR nuclear technology and selected the GEH ESBWR nuclear technology for the North Anna Unit 3 project. The initial phase of the North Anna Unit 3 combined license application was submitted to the NRC in July 2013, and the final portion of the application was submitted in December 2013.

The NRC issued the combined license and limited work authorization for Vogtle Electric Generating Plant, Units 3&4 on February 10, 2012. The Vogtle plants reference the AP1000 design certification amendment. It was the first combined license issued by the NRC to construct and operate a nuclear power plant under the alternative licensing process NRC Report to ASME November 2015 in 10 CFR Part 52. It is the first time since 1978 that the NRC issued a license to construct a nuclear power plant in the United States.

The NRC staff issued the combined license for V.C. Summer 2&3 on March 30 , 2012. The V.C. Summer 2&3 plants reference the AP1000 design certification amendment.

On February 4, 2015, the NRC Commissioners held a mandatory hearing on the combined license for Fermi , Unit 3. On May 1, 2015, the NRC issued the combined license for Fermi, Unit 3. This is the first combined license for an application referencing the ESBWR design.

On June 8, 2015, UniStar requested to withdraw the Calvert Cliffs, Unit 3 combined license application.

Advanced Reactors Program NRC established an advanced reactors program in the Office of New Reactors. Currently, there are no applications under review, but several applications are expected in the next three years including:

Integral PWRs (iPWRs):

NuScale (iPWR) - NuScale Power is developing a modular, scalable 50 MWe iPWR. Pre-application reviews are currently under discussion. The design certification is expected to be submitted to the NRC in November or December of 2016.

B&W mPower (iPWR) - B&W is developing a modular, scalable 180 MWe iPWR. At this time, mPower has reduced its activities in the mPower development, and have not provided a submittal date for the application.

TVA is planning to submit its early site permit in the first quarter of 2016 for its Clinch River site near Oakridge, Tennessee.

Holtec is developing the Holtec Inherently Safe Modular Underground Reactor SMR 160 design that has a 160 MWe electrical power output. They plan to pursue a Part 50 licensing process that requires an applicant to apply for a construction permit and a subsequent operating license. They have not provided an application submittal date.

XEnergy has indicated it plans to submit a design certification application to the NRC within the next few years for its pebble-bed high temperature gas-cooled reactor. The XEnergy reactor (Xe-100) is a helium-cooled reactor with a power rating of 125 MWt.

Advanced Reactor (non-light water reactors) Guidance Development:

NRC has received Idaho National Laboratory (INL) generated Department of Energy technical report "Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors." The INL report is the culmination of phase one of a two-phase initiative by the DOE and the NRC to develop advanced reactor safety design criteria from which the principal design criteria could be derived for advanced reactor concepts. The NRC will follow its normal process for developing NRC Report to ASME November 2015 and issuing regulatory guidance and anticipates completion of such guidance by the end of 2016.

Construction Issues:

AREVA reported lower than expected mechanical properties for the reactor vessel at its EPR reactor in Flamanville, France. Additional testing of the material will be performed. The NRC is following this issue as it may pertain to any new or replacement components.

During an NRC vendor inspection, it was observed that the vendor was using ASME Code NPT stamps on safety related piping that were not in accordance with the ASME Code. The vendor had no official documentation from ASME confirming that these NPT stamps were valid. The authorized nuclear inspector (ANI) initiated action with ASME to ensure that the ASME Code stamps are valid, and ASME has committed to initiate action to replace these NPT stamps. ASME sent a letter to on June 17, 2015 to all ASME NPT Certificate holders regarding the acceptability of these NPT stamps. The ASME letter also requested that all NPT Certificate holders send in these NPT stamps for replacement in order to avoid further confusion in the industry. On October, 2, 2015, ASME supplemented its June 17, 2015 letter by requesting that all NPT stamps with the NPT letters arranged horizontally to be returned to ASME for replacement by December 31, 2015.

6. Multinational Design Evaluation Program (MDEP) Activities MDEP is a multinational initiative to develop innovative approaches to leverage the resources and knowledge of mature, experienced national regulatory authorities who are tasked with the regulatory design review of new reactor plant designs. Some of the issue-specific working groups established under the MDEP organization that the NRC participates in are the Codes and Standards Working Group (CSWG), whose goal is to achieve harmonization of code requirements for pressure-boundary components, and the Vendor Inspection Cooperation Working Group (VICWG), whose goal is to maximize the use of the results of inspections obtained from other regulators' efforts in inspecting vendors.

Vendor Inspection Cooperation Working Group (VICWG)

The MDEP VICWG was formed because component manufacturing is currently subject to multiple inspections and audits similar in scope and in safety objectives, but conducted by different regulators to different criteria. The primary goal of the VICWG is to maximize the use of the results obtained from other regulators' efforts in inspecting vendors.

The MDEP VICWG continues to achieve its short-term goals and is making progress towards achieving its long term goals. The VICWG continues to focus on maximizing information sharing, joint inspections (multiple regulators inspecting to the regulatory requirements of one country), and witnessing of other regulators' inspections. NRC has participated in 6 witnessed and joint inspections this year to date. Additional MDEP NRC Report to ASME November 2015 inspections are also occurring that do not involve the NRG, but an exact count has not been established for FY15.

The working group enhances the understanding of each regulator's inspection procedures and practices by coordinating witnessed inspections of safety related mechanical pressure retaining components (Class 1) such as pressure vessels, steam generators, piping, valves, pumps, etc., and quality assurance inspections. Witnessed inspections consist of one regulator performing an inspection to its criteria, observed by representatives of other MDEP countries. The benefits to the observing countries include additional information and added confidence in the inspection results. MDEP regulators are using the experience gained during conduct of VICWG witnessed inspections in their inspection planning.

The MDEP VICWG held its 151h meeting during the week of May 19 in Shenzhen, China. This meeting included members from, China, Japan, the Republic of Korea, France, South Africa, Finland, the United Kingdom and the United States. Canada, Sweden, United Arab Emirates, the Russian Federation, India and Turkey were not in attendance. Because the meeting was hosted in China, it allowed multiple members of China's NNSA, NRO, and NSC to participate in the meeting. Because China is involved with the construction of many designs, they have shared interests with many of the other VICWG countries. NRG participated in 6 MDEP inspection activities during the past year. The Next VICWG meeting is scheduled to be held in Dijon, France the week of October 26, 2015. On October 30th , representatives from ASME, ISO, WNA and others are invited to discuss their latest initiatives.

Codes and Standards Working Group (CSWG)

The MDEP group's goal is to harmonize and converge national codes, standards, and regulatory requirements and practices applicable to pressure boundary components while recognizing the sovereign rights and responsibilities of national regulators in carrying out their safety reviews of new reactor designs. The CSWG published several reports on codes and standards related to pressure boundary components, and it provides a regulatory forum for groups such as the World Nuclear Association's Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group to coordinate with concerning international harmonization efforts.

In October 2015, the CSWG met with representatives from CORDEL and ASME. A report from CORDEL detailed a proposal to help reconcile Non-Destructive Testing Certifications internationally. The representative from CORDEL presented this report at the Standards Development Organization (SDO) Convergence meeting during the November, 2014, ASME Code Meeting. In January 2016, CORDEL representatives will also attend a meeting with the Chinese concerning their codes and standards work.

ASME representatives informed the CSWG that they have accepted two initial topics, Welding Qualification and Fatigue Roadmap. A statement of work (SOW) has been finalized for developing an international comparison report on Welding Qualification and Welding Quality Assurance, and a SOW is being finalized for developing an International Fatigue Roadmap to support a harmonized set of rules to perform fatigue analysis of Class NRC Report to ASME November 2015 1 components. For the November 2015 CSWG meeting, two draft reports have been provided concerning Welding Qualification and Welding Quality Assurance; and Certification of NDE Personnel. The CSWG will be reviewing and providing comments on both documents in the November meeting.

CSWG will continue to follow closely the activities of the SDOs and CORDEL through 2015, at which time the CORDEL pilot program for convergence using the SDO Convergence Board process is expected to complete at least one code convergence. Also, the CSWG may discuss the possibility of turning over its regulatory interface with the SDOs and CORDEL's activities to another international regulatory organization (e.g., NEA's Committee on the Safety of Nuclear Installations), as many of the topics are growing beyond the MDEP mandate.

7. 10 CFR Part 21 Rulemaking On September 29, 2011, the NRC staff issued SECY-11-0135, "Staff Plans to Develop the Regulatory Basis for Clarifying the Requirements in T itle 10 of the Code of Federal Regulations PART 21, "Reporting of Defects and Noncompliance." This Commission Paper informs the Commission of the staffs plan to develop the regulatory basis to clarify Title 10 of the Code of Federal Regulations (10 CFR) Part 21, "Reporting of Defects and Noncompliance." The staff will make modifications to Part 21 and regulatory guidance will be proposed.

In December 2012, the staff published the Draft Regulatory Basis to Clarify 10 CFR Part

.fl, "Reporting of Defects and Noncompliance," Revision 0. The NRC provided this version of the regulatory basis as a draft to promote early stakeholder feedback. Other associated information can be found on the Federal Government's regulations web site, by searching for "NRC-2012-0012." The goal of this regulatory basis is to simplify and clarify the rule language in Part 21, provide consolidated regulatory guidance on compliance with Part 2 1, and enhance regulatory stability and predictability for the entities to which Part 21 applies.

On August 7, 2015 the final Regulatory Basis was issued. This includes clarifications to Section 21.3, "Definitions," related to evaluation and reporting of defects that could create a substantial safety hazard and terms related to commercial grade dedication. Also, this revision to Part 21 introduces a separate section in Part 21 specifically related to commercial grade dedication activities.

The NRC staff held a public meeting on October 16, 2015, to discuss the NEI 14-09, "Guidelines for Implementation of 10 CFR Part 21, Reporting of Defects and Noncompliance," Revision 0, dated August 2014. NEI developed this guidance to incorporate previous guidance in NUREG-0302, to include additional clarity in specific areas where issues have historically occurred and to include experience gained from nearly 30 years complying with the 10 CFR Part 21 rule. The NRC staff expects to issue the proposed rulemaking for public comment by the end of 2016.

NRC Report to ASME November 2015

8. Commercial Calibration Services Status By letter dated April 29, 2014, the Nuclear Energy Institute (NEI) submitted Revision O of NEI 14-05, "Guidelines for the Use of Accreditation in Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services," to the U.S. Nuclear Regulatory Commission (NRC) for NRC staff review and endorsement. NEI 14-05 provides an approach for licensees and suppliers of basic components for using laboratory accreditation by Accreditation Bodies (ABs) that are signatories to the International Laboratory Accreditation Cooperation (ILAC) Mutual Recognition Arrangement (MRA)

(hereby after referred to as the ILAC accreditation process) in lieu of performing commercial-grade surveys for procurement of calibration and testing services performed by domestic and international laboratories accredited by ILAC signatories.

By letter dated July 22, 2014, the NRC requested additional information to complete its review of NEI 14-05. Two teleconferences were held on July 3, 2014 and August 13, 2014, to promote a better understanding of the NEI responses to the NRC's request for additional information. By a letter dated August 28, 2014, NEI submitted NEI 14-05 Revision 1, which incorporates NEl's responses to the NRC staff's comments on NEI 14-05.

By letter dated February 9, 2015 (ADAMS Accession No. ML14322A535), the NRC staff transmitted its safety evaluation (SE) identifying the guidelines contained in NEI 14-05, Revision 1 (ADAMS No. ML14245A391) as an acceptable approach for licensees and suppliers to meet the requirements of 10 CFR Part 50, Appendix B to use ILAC laboratory accreditation as part of the commercial-grade dedication process for procurement of calibration and testing services. NRC's endorsement of NEI 14-05, Revision 1, expands the NRC's acceptance of the ILAC accreditation process first documented in SE on an Arizona Public Service request (ADAMS Accession No. ML052710224). NRC's earlier acceptance was limited to calibration laboratory serv1ices accredited by specific U.S.

Accrediting Bodies (ABs). The SE (1) confirms that NEI 14-05, Revision 1, reflects the ILAC accreditation process previously approved; (2) provides an evaluation of the unique aspects of NEI 14-05, Revision 1; (3) constitutes formal NRC endorsement of the guidelines in NEI 14-05, Revision 1, for using the ILAC accreditation process in lieu of performing a commercial-grade survey; and (4) finds that the ILAC accreditation process continues to satisfy the requirements of Appendix B to 10 CFR Part 50 and, therefore, is acceptable.

Although the NRC has endorsed NEI 14-05, Revision 1, licensees and suppliers of basic components use of the ILAC accreditation process in lieu of performing a commercial-grade survey represents a reduction in commitment to the previously accepted QA program. As such, once the NRC approves the QA program change for a licensee in accordance with 10 CFR 50.54(a)(4), other licensees may adopt the QA alternative of using the ILAC accreditation process in lieu of performing a commercial-grade survey provided that the bases of the NRC approval are applicable to the licensee's facility pursuant to the requirements of 10 CFR 50.54(a)(3)(ii). On October 7, 2015, Ameren Missouri, licensee of the Calloway plant, submitted a license amendment to adopt NEI 14-05, Revision 1.

NRC Report to ASME November 2015 NEI 14-05, Revision 1, will allow licensees and vendors to use the ILAC accreditation process in lieu of performing commercial-grade surveys for procurement of calibration and testing services performed by domestic and international laboratories accredited by signatories to the ILAC MRA. Suppliers of basic components may begin to use the ILAC accreditation process and do not have to wait for the NRC to approve the QA program change for a licensee provided the conditions from the NEI 14-05, Revision 1, safety evaluation are met. .

On August 4, 2015, the NRC issued draft Regulatory Issue Summary (RIS), 2015-XX, "Use of Accreditation in lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services," to notify external stakeholders that the NRC staff has found acceptable for use Revision 1 to NEI 14-05, with respect to procurement and dediication of calibration and testing services performed by domestic and international laboratories. T he draft RIS is available in ADAMS under Accession No. ML15090A236. The Public comment period closed on October 4, 2015.

9. NRC Staff Review of EPRI 1025243 Guideline for Commercial-Grade Design and Analysis Computer Programs By letter dated July 18, 2012, the Nuclear Energy Institute (NEI) submitted Electric Power Research Institute (EPRI) 1025243, Plant Engineering: Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Safety-Related Applications for staff review and approval. EPRI 1025243 describes a dedication methodology for commercial-grade design and analysis computer programs for use in meeting regulatory requirements. The letter proposes the NRC endorse the EPRI 1025243 computer program commercial-grade dedication method, which follows the method provided in EPRI NP-5652, which the NRC conditionally endorsed in Generic Letter 89-02.

In the fall of 2012, the NRC staff reviewed EPRI 1025243 with reference to the dedication requirements of 10 CFR Part 21 and Appendix B to 10 CFR Part 50.

Draft Regulatory Guide (DG)-1305, "Acceptance of Commercial-Grade Design and Analysis Computer Programs for Nuclear Power Plants. " was issued to endorse Revision 1 of EPRI Technical Report 1025243, as an acceptable method in meeting regulato ry requirements. The initial version of EPRI 1025243 was the first guidance document to provide a detailed dedication methodology specific to commercial-grade design and analysis computer programs.

On July 1, 2015, the NRC issued for public comment DG- 1305, "Acceptance of Commercial- Grade Design and Analysis Computer Programs for Nuclear Power Plants."

The DG provides new (i.e., not preceded by earlier guidance on the same subject) guidance that describes acceptance methods that the staff of the NRC considers acceptable in meeting regulatory requirements for acceptance and dedication of commercial-grade design and analysis computer programs for nuclear power plants.

The NRC staff is currently evaluating the comments received from NEI and other stakeholders with expected completion by October 31, 2015.

NRC Report to ASME November 2015

10. EPRI Development of Revised Dedication Guidance for Commercial-Grade Items for Use in Nuclear Safety-Related Applications In September, 2014, the EPRI issued the 2014 Technical Report 3002002982, "Plant Engineering: Guideline for the Acceptance of Commercial-Grade Items in Nuclear Safety-Related Applications" - Revision 1 to EPRI NP-5652 and TR-102260. The NRC staff participated during many of the EPRI technical advisory group (TAG) meetings held at EPRl's offices in Charlotte, North Carolina.

This report describes a methodology that can be used to dedicate commercial-grade items for use in safety-related applications. The scope of applications for which commercial-grade item dedication is used has evolved significantly since the EPRI published its reports Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications (NCIG-07) (NP-5652) and Supplemental Guidance for the Application of EPRI Report NP-5652 on the Utilization of Commercial Grade Items (TR-102260) in 1988 and 1994, respectively. The guidance in this final report reflects lessons learned and addresses challenges that have been identified through expanded use of the original guidance. This report supersedes both original reports in their entirety.

The NRC has developed a Draft Guide DG-1292, "Dedication of Commercial Grade Items" for potential NRG endorsement and is currently in the review process to approve Revision 1 to EPRI Technical Report 3002002982 with respect to acceptance of commercial-grade dedication of items used as basic components for nuclear power plants. The initial version of EPRI NP-5652 (Ref. 5) was the first guidance document to provide a detailed acceptance methodology specific to commercial-grade dedication (CGD) for items used in nuclear power plants.

The industry's use of the CGD process has significantly increased over time as the number of suppliers with nuclear quality assurance programs has decreased. However, current industry dedication guidance was developed in the late 1980's and the NRC has only previously endorsed EPRI dedication guidance in Generic Letter 89-02, "Actions to Improve the Detection of Counterfeit and Fraudulently marketed Products," and Generic Letter 91-05, "Licensee Commercial-Grade Procurement and Dedication Programs." Therefore, a need existed for the NRC to review and endorse the latest EPRI guidance for commercial grade dedication, in which the NRC participated in the development stage. EPRI Technical Report 3002002982 is available free of charge to the public on the EPRI website (EPRl.com).

NRC Report to ASME November 2015

11. NRC Staff Interface with Nuclear Utilities Procurement Issues Committee (NUPIC)

On June 15 -18, 2015, NRC staff participated and made presentations at the 24th Annual NUPIC Vendor Meeting in Henderson NV. NRC presentations were related to the Part 21 rulemaking and its effect on commercial grade dedication and an update on vendor inspection activities. The NRC periodically accompanies a NUPIC Joint Utility Audit team to observe selected audits and ensure that the audit process remains an acceptable alternative to the NRC's vendor inspection/audit program. The NRC staff continues to rely on the effectiveness of the NUPIC joint utility audit process for evaluating the implementation of quality assurance programs of suppliers to the nuclear industry. The NRC uses trip reports to document NRC observation of audits performed by NUPIC. The NRC team has used NRR/NRO Inspection Procedure (IP) 43005, "NRC Oversight of Third-Party Organizations Implementing Quality Assurance Requirements," during the NUPIC observation oversight. Trip reports issued for the NUPIC audits that have been observed by the NRG are availablle at the below web-site link:

http://www.nrc.gov/reactors/new-reactors/oversighUguality-assurance/nupic-industry.html

12. Counterfeit, Fraudulent and Suspect Items (CFSI) Commission Paper In SECY-11-0154, "An Agencywide Approach to Counterfeit, Fraudulent, and Suspect Items (CFSI)," dated October 28, 2011, the staff informed the Commission of activities to assess and enhance processes for addressing counterfeit, fraudulent, and suspect items in NRG regulated activities. This new SECY paper provides a status of the completed and ongoing activities originally presented in SECY-11-0154, ADAMS Accession No. ML112200150).

On January 8, 2015, the NRC staff published the SECY Paper SECY-15-0003, "STAFF ACTIVITIES RELATED TO COUNTERFEIT, FRAUDULENT, AND SUSPECT ITEMS. This SECY paper informs the commission of the staff's progress in identifying and implementing proactive strategies to detect and prevent the intrusion of counterfeit, fraudulent, and suspect items into equipment, components, systems , and structures regulated by the NRG.

This paper is the final planned status update on the topic. The discussion section provides the status of each of the 19 planned actions from SECY-11-0154. Of these 19 planned actions, 14 have been completed and the remainder will be completed by December 2018.

On June 24, 2015, the NRG issued Regulatory Issue Summary (RIS) 2015-08, "Oversight of Counterfeit, Fraudulent, and Suspect Items in the Nuclear Industry." The NRG issued this RIS to heighten awareness of the existing NRC regulations and how they apply to CFSI within the scope of NRC's regulatory jurisdiction. Addressees are expected to review this information and consider actions, as appropriate, to prevent CFSI from entering their supply chains, prevent possible installation or use of CFSI at their facilities, and raise awareness of the potential for CFSI to be used in the manufactu re, maintenance, or repair of items, NRC Report to ASME November 2015 including sealed sources and devices (SSDs). The R IS is available in ADAMS under Accession No. ML15008A191.

13. License Renewal Activities Following are on-going activities related to license renewal:

Current status of applications, staff reviews and approvals

78 units approved (76 operating plants with renewed licenses) o Renewed licenses for Sequoyah 1 & 2 issued September 24, 2015 o 1 (2 units) in hearings (Indian Point 2 & 3) - supplemental SER issued November 2014 o 1 (4 units) completed ACRS Full committee (Byron 1 & 2 / Braidwood 1 & 2 (9/2015])

o 1 (1 unit) awaiting ACRS Full committee (Davis-Besse (9/2015])

o 2 (3 units) awaiting follow-up ACRS Subcommittee (Seabrook (1 2/2015],

Diablo Canyon 1 & 2 (10/16])

o 4 (6 units) awaiting ACRS Subcommittee (Fermi [TBD], LaSalle 1 & 2 (3/2016), Grand Gulf [TBD), South Texas Project 1 & 2 [TBD])

1 application (1 unit) with scheduled submitta l in 2016:

o January to March 2016 -Waterford 3 o January to March 2017 - River Bend o July to September 2018 - STARS o October 2019 - Perry o January to March 2021 - Clinton Thirty-nine units have entered the operating period beyond 40 years:

o Oyster Creek - April 9, 2009 o Point Beach 2 - March 8, 2013 o Nine Mile Point 1 - August 22, 2009 o Turkey Point 4 -April 10, 2013 o Ginna - September 18, 2009 o Peach Bottom 2 - August 8, 2013 o Dresden 2 - December 22, 2009 o Fort Calhoun 1 - August 9 , 2013 o H.B. Robinson - July 31, 2010 o Prairie Island 1 - August 9, 2013 o Monticello - September 8, 2010 o Indian Point 2 - September 28, o Point Beach 1 - October 5, 2010 2013 o Dresden 3 - January 12, 2011 o Oconee 2 - October 6, 2013 o Palisades - March 24, 2011 o Browns Ferry 1 - December 20, o Vermont Yankee - March 21 , 2012 2013 o Surry 1 - May 25, 2012 o Cooper Nuclear Station - Jan. 18, o Pilgrim - June 8, 2012 2014 o Turkey Point 3 - July 19, 2012 o Duane Arnold - February 21, 2014 o Quad Cities 1 - December 14, 2012 o Three Mile Island 1 -April 19, 2014 o Quad Cities 2 - December 14, 2012 o ANO 1 - May 20, 2014 o Surry 2 - January 29, 2013 o Browns Ferry 2 - June 28 , 2014 o Oconee 1 - February 6, 2013 o Peach Bottom 3- July 2, 2014 NRC Report to ASME November 2015 o Oconee 3 - July 19, 2014 o DC Cook 1 - October 25, 2014 o Calvert Cliffs 1 - July 31, 2014 o Prairie Island 2 - October 29, 2014 o Hatch 1 -August 6, 2014 o Brunswick 2 - December .27, 2014 o FitzPatrick - October 17, 2014 o Millstone 2 - July 31, 2015 Technical Issues

Draft License Renewal Interim Staff Guidance: Aging Management of Loss of Coating Integrity for Internal Service Level Ill Coatings (LR-ISG-2013-01)

Issued for public comment on June 29, 2015.

ADAMS Accession No. ML15125A377.

Public comments were received as scheduled.

Addressing public comments - final ISG-LR to be issued early 2016

Steam Generator Divider Plates and Tube-to-Tubesheet Welds

- Steam Generator Task Force submitted a topical report to address necessity for these inspections

PWR Vessel Internals (MRP-227-A)

- Industry to submit draft revision 1 by end of 2015

Cast Austenitic Stainless Steel in PWR Vessel Internals

- Working with industry to identify acceptable approach to evaluate acceptabily of CASS components Subsequent License Renewal The NRC released SECY-14-2016 on January 31 , 2014 (http://www.nrc.gov/reading-rm/doc-collections/commission/secys/2014/2014-001 6scy.pdf), to inform the Commission of ongoing staff activities to prepare for the anticipated receipt and review of subsequent license renewal (SLR) applications that, if approved, could extend operation of power reactors beyond 60 years. The Commission's Staff Requirements Memorandum (SRM) on the paper was released on August 29, 2014 (http://www.nrc.gov/reading-rm/doc-collections/commission/srm/2014/2014-0016srm.pdf). The SRM stated that the staff should continue to update license renewal guidance, as needed, to provide additional clarity on the implementation of the license renewal regulatory framework, and the staff should address emerging technical issues and operating experience through alternative vehicles. The SRM further stated that the staff should keep the Commission informed on the progress in resolving the following technical issues related to SLR: reactor pressure vessel neutron embrittlement at high fluence; irradiation assisted stress corrosion cracking of reactor internals and primary system components; concrete and containment degradation, and electrical cable qualification and condition assessment.

NRC staff is continuing to work internally to ensure adequacy of guidance for SLR and continues to meet with the Nuclear Energy Institute (NEI) and the industry on specific technical topics. Upcoming milestones include:

Meeting on November 6 with the Advisory Committee on Reactor Safeguards (ACRS) to discuss status of technical issue resolutions for SLR.

Issuance of draft Generic Aging Lessons Learned Report for SLR and Standard Review Plan for SLR by the end of 2015.

Research Activities Research related to reactor long term operation (LTO) and material degradation issues is ongoing in the NRC's Office of Nuclear Regulatory Research (RES):

NRC RES staff coauthored LTO-relevant presentations at the International Conference on Environmental Degradation in Nuclear Power Systems - Water Reactors, Aug 9-13, 2015. Papers and Presentations are available at : http://envdeg2015.org/final-proceedings/ENVDEG/index.html Titles included

- Crack Growth Rate and Fracture Toughness J-R Curve Tests on Irradiated Cast Austenitic Stainless Steels

- Identification of Potential Degradation Phenomena for Spent Fuel Dry Cask Storage Systems

- Evaluation of Strategies for Obtaining High Fluence Materials to Assess Irradiation-Assisted Degradation of PWR Internals

NRC RES staff coauthored an LTO-relevant presentation at the 23rd International Structural Mechanics in Reactor Technology Conference (SMiRT-23), August 9-14, 2015, "NPP Subsequent License Renewal: Lessons Learned from AMP Effectiveness Audits," and published in Transactions, SMiRT-23.

RES/DE staff participated in the Commission Briefing "Overview of the Operating Reactors Business Line" on August 6, 2015, (http://www.nrc.gov/reading-rm/doc-collections/commission/slides/2015/20150806/staff-20150806.pdf ) pointing out that support to SLR is a principal research focus and involves conducting research ,

developing the license renewal guidance documents and their techn ical basis, and confirming adequacy of aging management programs

In September, RES awarded a research contract to Pacific Northwest National Laboratory (PNNL) to create an information tool for obtaining information (including ex-plant harvesting) from designated NPPs as they go through decommissioning. This complements ongoing research in developing technical information to support evaluating SLR applications.

RES awarded a contract to the Savannah River National Laboratory (SRNL) to provide technical assistance associated with the evaluation of neutron absorbing panels removed from the decommissioned Zion NPP spent fuel pool.

On November 17, 2015, the Advisory Committee on Reactor Safeguards (ACRS) is convening a meeting, open to the public, to discuss key technical issues and R&D pertaining to SLR. Presentations will be given by DOE, EPRI, NEI and NRC staff. Topics includes fluence evaluations, and degradation of concrete, electrical cables, reactor pressure vessel (RPV) and reactor vessel internals (RVI).

NRC's Regulatory Information Conference (RIC) will meet again on March 8-10, 2016.

RES is planning a SLR-related RIC Session entitled "Subsequent License Renewal:

Near-Term Research Supporting the Technical Bases for SLR." This session is designed to be the second in a two-part series where Part 1 will be on the newly-written SLR guidance documents and Part 2 will cover near-term research on safety-significant electrical, mechanical, and structural systems that supports the technical bases for SLR.

RES staff continues to interact with the DOE-LWRS Program and EPRl's LTO initiatives to monitor developments relevant to SLR and, where appropriate, engage in joint research activities.

International Forum on Reactor Aging Management (IFRAM) - Efforts are continuing to develop areas of cooperation and collaboration in sharing data on long term operation

14. New Generic Letters There were no Generic Letters (GLs) issued since the last Code Week.

15. New Information Notices Since the last Code Week, the following Information Notices (INs) were issued:

i. IN2015-08 (09/02/2015), "Criticality and Chemical Safety Events Involving Unanalyzed Conditions and Unanticipated Unavailability of IROFS at Fuel Cycle Facilities" o Informs addressees of recent operating experience involving deficiencies with evaluations of credible high and intermediate consequence accident sequences in facility integrated safety analyses (ISAs), delineation of items relied on for safety (I ROFS) boundaries, and implementation of effective management measures to ensure the availability and reliability of IROFS. These deficiencies resulted in unanalyzed conditions or unavailability of IROFS established to minimize the likelihood of high or intermediate consequence events.

ii. IN2015-09 (09/24/2015), "Mechanical Dynamic Restraint (Snubber) Lubricant Degradation Not Identified Due To Insufficient Service Life Monitoring" o Alerts addressees to potential degradation of the lubricant (grease) in mechanical dynamic restraints (snubbers) not identified due to insufficient service life monitoring (SLM). The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.

16. New Regulatory Issue Summaries There were no Regulatory Issue Summaries (RISs) issued since the last Code Week.

17. NRC Publications of Potential Interest to ASME Since the last Code Week, the following other publications that may be of interest to ASME were issued:

i. NUREG/CR-7204 (September 2015), "Applying Ultrasonic Testing in Lieu of Radiography for Volumetric Examination of Carbon Steel Piping" o This report provides an initial technical evaluation of the capabilities of phased-array ultrasonic testing to supplant traditional radiographic testing for detection and characterization of welding fabrication flaws in carbon steel welds.

ii. Revision 1 of Standard Review Plan (SRP) 17.5, "Quality Assurance Program Description - Design Certification, Early Site Permit and New License Applicants," (NUREG-0800, Chapter 17), was issued on August 2015.

o The new revision included the following changes: (1) the wording was simplified to reflect plain language throughout in accordance with the NRC's Plain Writing Action Plan; (2) the guidance was aligned with the latest revision of R,egulatory Guide (RG) 1.28, "Quality Assurance Program Criteria (Design and Construction)," Revision 4; (3) the guidance was aligned with RG 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2; and (4) the SRP was aligned with NQA-1-2008/2009a, the latest edition that the staff found acceptable for meeting the requirement of Appendix B to 10 CFR Part 50.

18. Upcoming Public Meetings of Potential Interest to ASME The following public meetings, either upcoming or recently transpired , may be of interest to ASME:

i. NRC staff review of NEl-14-09, "Guideline for Implementation of 10 CFR Part 21 ,

Reporting of Defects and Nonconformance" o October 16, 2015, 8:00 am - 11 :00 am ii. Meeting to Discuss Peening Issues o Tentatively December 9, 2015; details still being finalized.

iii. RPV Integrity Issues and Appendix H Rulemaking o January 19, 2016, 8:00 am - 12:00 pm iv. lndustry/NRC NOE Technical Information Exchange Public Meeting o January 19, 2016, 1 :00 pm to January 21, 2016, 12:00 pm EST.

Refer to the NRC Public Meeting Web Page at http://meetings.nrc.gov/pmns/mtg for a list of all currently-scheduled public meetings and further details.

Note to requester: The attachment is immediately following this email.

From: Hull, Amy Sent: Wed, 8 Apr 2015 15:36:59 -0400 To: Hiser, Matthew

Subject:

Strategic Harvesting ExPlnt Material s.doc Attachments: Strategic Harvesting ExPlnt Materials.ABH.pre-draft.doc I _____. ._. . ._. . . .! J b)(6)

In process ...... done ....

I put this on the most recent form from NEAT. Question ( 1) should we also have an optional task related to harvesting ex-plant material that you are working on with NMSS? If so, would you like to write that up? I will make a skeleton task tomorrow. (2)1 Jcanyou Jb)(6) help with changing Fig . 2?

Version Control Date: November 1, 2014 STATEMENT OF WORK NRC Agreement Number NRC Agreement NRC Task Order Number NRC Task Order Modification Modlflcatlon Numbe1r (If Applicable) Number (If Applicable)

Project TIiie Strategic Approach for Obtaining Aging Degradation Information from Decommissioning Nuclear Power Plants Job Code Number B&R Number DOE Labor to,y

??? ???

NRC Requisitioning Office Nuclear Regulatory Research (RES)

NRC Form 187, Contract Security and Classification Roqulromonts lion 0 Applicable Unclassified x Not Applicable XO Non Fee-Recoverable Docket Number (If Fee-Recoverable/Applicable) nspectiol!.,Report Number (If Fee Recoverable/Applicable)

Technical Assignment Control Number (If Fee-, Technical Assignment Control Number Description (If Fee-Recoverable/Applicable) Recoverable/Appllcable) 1.0 Regulatory Context:

The NRC has e/iablished a license renewal process that will allow nuclear power plants (NPP) to renew their licenses for an additional 20 years, via 10 CFR 54.31(d) stating that "a renewed license may be subsequently renewed." The biggest challenges for the NRC and the industry will be addressing tie major technical issues for this second "subsequent" license renewal (SLR) beyond 60 years. The staff currently believes (SECY-14-0016, NUREG-1925) the most significant technical issues challenging power reactor operation beyond 60 years are:

Reactor pressure vessel (RPV) neutron embrittlement al high fluence

Irradiation assisted degradation (IAD) of reactor internals and primary system components

Concrete and containment degradation

Electrical cable qualification and condition assessment

Version Control Date: November 1, 2014 Understanding the causes and control of degradation mechanisms forms the basis for developing aging management programs (AMPs) to ensure the functionality and safety margins of nuclear power plant (NPP) systems, structures, and components (SSC). The resolution to these issues must provide reasonable assurance of safe operation of the components in the scope of license renewal during the subsequent period of extended operation. As stated in SRM-SECY-14-006, "the staff should continue to emphasize in communications with industry the need to strive for satisfactory resolution of these issues prior to 7he N C beginning a review of any SLR application."

This research incorporates many of the ideas proposed in the NRC long te research project (LTRP) Strategic Approach for Obtaining Material and Component Aging lnforr.(!_aJion from Decommissioning Nuclear Power Plants (L TRP ID 111, 2013, Joseph Kanney, RES/DRA/ETB) related to developing a strategic and systematic approach to sampling a erials from SSC in decommissioning plants. There are separate, currentl indeg__endent activities in RES to sample Zorita baffle plate material (Hiser & Rao, UNR NRR-201 -008); to samp e Zion NPP SFP boral panels and surveillance coupons (Focht, UNR NRR-2013-005), a d to harv~t Zion degraded cables (Murdock, UNR NRR-2011-014). There is a so an aGtIvity in RES to identify key significant technical issues pertinent to SLR to ensur~ NRC's readiness to review possible second license renewal applications (Hull UNR- RR-2014-001 ). t,s mentioned in the LTRP, the envisioned work addresses boJ passive a d ctive components. In that sense, it links degradation addressed by the license renewal le, 10 CFR 54w ith aging management of active components covered by the maintenance rue, 10~50.65.

~ U. . RC Safety Beyond 60 Years 10 CFR 50.55a Ensures thot the e ffects ofoglng w~ be effectivelymonoged l

Requlremer1s throughout the penod of extended operatlon

( Active Com11onents

)

Passive Components Agino Manaoement Effectiveness Fig. 1 Relationship between aging management of active and passive components (from NRR/RES presentation to ACRS, 2014) 2

Version Control Date: November 1, 2014 Oversight of operating NPPs would be enhanced by acquisition of data and information useful reducing uncertainties or improving sensitivity analysis in probabilistic risk assessments (PRA).

The proposed work was envisioned by DRA to address both passive and active components.

Independently, IAEA is exploring the formation of a coordinated research project (CRP) to evaluate structure and material properties utilizing actual aged materials removed from decommissioned reactors for safe long-term operation (LTO).

In the last year, four plants have ceased operation or announced that-will cease operation in the next year (Crystal River Unit 3 (PWR), Kewaunee (PWR), SONGS lJ.pits & 3 (PWR), and Vermont Yankee (BWR-Mk1 )). These plants comprise a range of reactor types, containments, as well as SSCs important to safety. Other NPPs may be added to this list in the near future.

The objective of this project is to develop a long-range strategy or roadmap for obt ming information from these plants as they go through decommissioning. T focus will be on timely acquisition of information that can significantly improve tHe agency's risk-informed and performance-based regulato ry a pproach, b ut has been ve difficult or impossible to obtain from the operating reactor fleet.

Technical Context:

The LTRP proposed creating a roadmap for obtaining information from designated NPPs as they go through decommissioning. It is complementa~ to ongoing work in RES/DE/CMB (NRR-2010-006, NRR-2010-013, and NRR- 014-001) developin technical information to support evaluating NPP second license renewal (SLR, long- erm operation, LWR sustainability) as well as ongoing work (NRR-201J.2-008, NRR 2013-05) ipv lving data collection and testing on materials from Spain's decommissio~ea Zorita reactor and the Zion reactor (both PWRs). As mentioned in the LTRP, versig of active components in operating NPPs would be enhanced by acquisition of data and i formation useful reducing uncertainties or improving sensitivity analysis in p obabilist'c risk assessmer:its (PRA). The Standardized Plant Analysis Risk (SPAR) models provide RA to Is to supP.()rt these risk-informed activities.

Mat~ ~ r dation flas traditionally been managed in a reactive mode more than a proactive mode,-, or the NPPs ju !'(low entering their first license renewal period from 40-60 years, and submitting SLRP-fs, it is necessary to extrapolate out to 80 years. Extrapolation of the evaluation of material pro erties in SSC from actual decommissioned NPPs will provide a basis for comparison with results of laboratory tests and calculations to resolve the four issues listed above.

The proactive management of materials degradation (PMMD) tool was originally created at PNNL for RES (POC: Amy Hull) to g ive an expert opinion of the possible future degradation mechanisms on a subcomponent/material specific basis (PNNL-17779). Combined with the LER database, the PMMD tool allows one to not only react to past events, but to help anticipate future issues. The original PMMD tool was based on NUREGICR-6923, it is now appropriate to integrate information from the recently-published five volumes of NUREG/CR-7153, the 3

Version Control Date: November 1, 2014 expanded materials degradation assessment (EMDA) for reactor pressure vessel (RPV), piping and reactor internals, cables and cable insulation, and concrete.

NUREG/CRs 6923 and 7153 present information in terms of its degradation susceptibility and knowledge for mitigation or prevention. A component with high degradation susceptibility/low knowledge would be the strongest candida te for proactive actions. It is necessary to be able to understand this before prioritizing sampling.

There are a number of technical gaps that this project addresses. Most importantly, the current piecemeal approach that obtains isolated and fragmented degradation infor~ation as targets of opportunity arise at a few plants can be replaced with a strategi<Yplan that 1s ore comprehensive, wider in scope, and more risk-informed. The rordmap for inspections and/or testing developed in this project will be useful guidance for obtaining key ~ easurements of degradation in a variety of areas. These measurements will be val\la~ e on their own. They will also be useful in basic research on the underlying mecha isms and modes of degradation, and for validation of modeling and simulation tools. Data and information developed f rom implementation of the roadmap will also be useful in evaluating aging monitoring and mitigation stra tegies proposed by Industry.

~P<<tP..-Ropcn .._. ,.... R.tlHolUty onPMDA Pf"G9'amOWRIMT NUREO/CR-G023 OWRvtfl'. 117 P<oduct

-M OPDll Plplno (Chart** H arri*)

0.Mt'lcAgilng LA~I.Ae,r,< N URC0-1 I01 PIN C A U..00 SCAP(SCC)

(Moy-oflov) (Rob9rt H *nfi. .)

Fig. 2 Prognostic tool to track and resolve critical technical issues for isLRi Commented [Al]: Matt, can you help me draft schematic lhat shows different input, I will give you a sketch

we need one box for decommissioned reactors roadmap, another for operating experience, another tor EMDA. another for the technical Issues dalabase, another for the EPRI MOM, IMT (delete SCAP, dele1e piping data base) 4
Version Control Date: November 1, 2014 2.0 OBJECTIVE Understanding and managing material and component degradation is unquestionably a key need for the continued safe and reliable operation of NPPs. It is also an area with very significant uncertainties. In many cases, the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in NPPs is incomplete. A strategic approach to examination and testing of materials and components from a relatively large cohort of decommissioning reactors can dramatically increase our knowledge-acquisition rate in this very important area.
The objective of this research is to construct a roadmap for obtaining unique and significant information that will inform the NRC's age-related regulatory oversight decision over the next two decades, and perhaps beyond. Implementation of the roadmap, in cooperation with Industry and DOE partners can be accomplished over time, througl\_indiv1dual research projects as the identified plants progress through their decommissioniog phase. This exploratory research is expected to provide reactor material degradation fundamental insights, information addressing potential technical issues, or identified gaps to s pport anticipated future(> 5 years)
NRC needs Another objective of this research is to revise t e PMMD infoto I, and bring to an NRC operating platform, to incorporate SLR-relevant information so that it can l:)e better used to inform prioritization in the ex-plant material roadmap.
3.0 SCOPE OF WORK In addressing the objective mentioned in 2.0 of this statement of work (SOW), the DOE Laboratory must conduct the Task 1 research aud provide all resources necessary to accomplish the subtasks and deliverables. Dec~ion on further execution of other Tasks (2 -4) will be made after the en of tlie scoping study consisting of Task 1.
The first stage will coAsists of a questi0nnaire and interviews with focus groups from various technical disciplines witl\in NRC. The second stage will comprise one or two public workshops.
The results of the first two phases will be combined into a fina l roadmap in the third stage. The roadmap will be deveklped in cooperation with Industry and other federal agencies such as DOE. lnternati0 al coun erparts may also be interested in participation.
The DOE L boratory, in the first stage, must assist NRC with developing a probing questionnaire and interpreting the results of interviews from various technical disciplines first within NRC. This wject will first start by leveraging resources within NRC. The most important internal resource to be leveraged is the collective knowledge and experience of NRC staff and selected contractors who have been engaged in license renewal reviews, inspections, analysis of operating experience data, research on material and component aging issues, and probabilistic risk assessment. Previous efforts such as those that produced the Iicense renewal guidance documents (LRGDs, NUREGs-1800 and-1801) and NUREG/CRs-6923 and 7153 will be leveraged in this project. Results of previous research projects on material aging and degradation issues will also be leveraged.
5
Version Control Date: November 1, 2014 In tasks subsequent to the completion of Task 1 scoping study, the DOE laboratory research may be tasked with further research to identify potential future requirements for licensees to conduct more research to further elucidate the risk assessment of component degradation.
Such research should also provide technical data and information, as necessary, to influence the national codes and standards bodies to re-examine requirements for materials of construction of passive components in light water reactors (LWRd) in assessing material degradation during service and its effect on design safety margin of components. The DOE laboratory principal investigator for this project should attend ASME Code Committee meetings, at the appropriate time of this research, and as approved by the COR during the course of this research. The DOE research must provide adequate information to support an IAEA international cooperative research program (ICRP) on this subject to bring worldwide resources to address this research need.
The product of this research shall consist of periodic technical lette r..e ports on the re,!lJIIS of this scop ing detailing the considerations for assessment a nd the conclusions derived from scoping study. When substantial and safety impact d iscoveries ~re ma~. these should be assessed for possible publication as NUREG/CR, in consultation w ith tlie COR of this project to make decision on such publication, as per deliverabJe schedule m ~tiope *n Section 5.0. T he DOE laboratory contractor shall deliver Monthly Letter Status Reports (MLSRs) electronically to the Contracting Officer Representative, Dr. Amy Hull, emai\: Amy.Hull@nrc.gov office, and copies must a lso be sent to the Division of Contracts at Con(ractsl20T.Resource@nrc.gov. The MLSR requirements w ill be as per Section 9.0 of this SOW.
4.0 SPECIFIC TASKS Task 1 - Show Gu idance ,
5.0 DELIVERAB[ ES AND(OR MILESTONES SCHEDULE 6.0 TECHNICAL AND OTHER SPECIAL QUALIFICATIONS REQUIRED Specialized ex~ ce must include expertise in such areas as h w Guidance 7.0 ESTIMATED LABOR CATEGORIES AND LEVELS OF EFFORT (OPTIONAL SECTION)
J Show Guidance 8.0 MEETINGS AND TRAVEL 6
Version Control Date: November 1, 2014 Show Guidan ce All travel requires written Government approval from the CO, unless otherwise delegated to the COR Foreign travel for the DOE laboratory personnel requ ires a 60-day lead time for NRC approval. For prior approval of foreign travel, the DOE laboratory shall submit an NRC Form 445, ' Request for Approval of Official Foreign Travel." NRC Form 445 is available in the MD 11.7 Documents library and on the NRC Web site at:
http://www.nrc.gov/reading-rm/doc-collections/forms/. Foreign travel is approved by the NRC Executive Director for Operations (EDO).
9.0 REPORTING REQUIREMENTS Show Guidanc The DOE Laboratory is responsible for structuring the deliverable to follow agency standards.
The current agency standard is Microsoft Office Suite 2010. The current agercy Portable Document Format (PDF) standard is Adobe Acrobat 9 Profession I. Deliverables must be submitted free of spelling and grammatical erro{ s an~ conform to requirements stated in this section. ~
Monthly Letter Status Reports In accordance with Management Directi e 11 .7, NJ C Procedures for Placement and Monitoring of Work with the U.S. Department of Ene gy, the DOE Laboratory must electronically submit a Monthly Letter Status Rep9rt (MLSR) by the ~ th day of each month to the Contracting Officer Representative (CO~) with copies to the Contraeting Officer (CO) and the Office Administration/Division of Contracts to ContractsPOT.Resource@nrc.gov. If a project is a task ordering agreement, a separate MLSR must be submitted for each task order with a summary project MLSR, even if'no work has been performed during a reporting period. Once NRC has determined that all work. on a ta k order is completed and that final costs are acceptable, a task order may be omitted fro~ e M~SR Show Guidan ce ---....,' -
The MLSR must include t e following: agreement number; task order number, if applicable; job code number; title of the project; project period of performance; task order period of performance, if applicable; COR's name, telephone number, and e-mail address; full name and address of the P.erforming organization; principal investigator's name, telephone number, and e-mail address; and eporting period. At a minimum, the MLSR must include the information discussed in Attachment 1. The preferred format for a MLSR can also be found in Attachment 1.
10.0 PERIOD OF PERFORMANCE The estimated period of performance for this work is Show Guidance months/years from date of agreement award.
Show Guidance 7
Version Control Date: November 1, 2014 11 .0 CONTRACTING OFFICER'S REPRESENTATIVE The COR monitors all technical aspects of the agreement/task order and assists in its administration. The COR is authorized to perform the following functions: assure that the DOE Laboratory performs the technical requirements of the agreement/task order; perform inspections necessary in connection with agreement/task order performance; maintain written and oral communications with the DOE Laboratory concerning technical aspects of the agreement/task order; issue written interpretations of technical requireJ11e_rts, including Government drawings, designs, specifications; monitor the DOE Laborat9ry's performance and notify the DOE Laboratory of any deficiencies; coordinate availability of NRO- urnished material and/or GFP; and provide site entry of DOE Laboratory personnel.
Contracting Officer's Representative Show Guidanc Name:
Agency: U.S. Nuclear Regulatory Commission Office:
Mail Stop:
Washington, DC 20555-0001 E-Mail:
Phone:
Show Guidance Name:
Agency:
Office:
Mail Stop:
Washingto) ) C 2~ 55-0001 E-Mail:
Phone:
12.0 MATERIALS REQUIRED (TYPE N/A IF NOT APPLICABLE)
Show Guidance 13.0 NRC-FURNISHED PROPERTY/MATERIALS (TYPE N/A IF NOT APPLICABLE)
I~ how Gu idance' 14.0 RESEARCH QUALITY (TYPE N/A IF NOT APPLICABLE) 8
Version Control Date: November 1, 2014 The quality of NRC research programs are assessed each year by the Advisory Committee on Reactor Safeguards. W ithin the context of their reviews of RES programs, the definition of quality research is based upon several major characteristics:
Results meet the objectives (75% of overall score)
Justification of major assumptions (12%)
Soundness of technical approach and results (52%)
Uncertainties and sensitivities addressed (11%)
Documentation of research results and methods is adequate Clarity of presentation (16%)
Identification of major assumptions (9%)
It is the responsibility of the DOE Laboratory to e a are adequately addressed throughout the course of the researc C COR will review all research products with these crit * *
15.0 STANDARDS FOR CONTM CTORS WHO P.REPARE NUREG-SERIES MANUSCRIPTS (TYPE N/A IF NOT APPLICABLE}
The U.S. Nuclear Regulatory Commission (NRC) tleQan to capture most of its official records electronically on Janua 1, 2000. The NRC ill ca ture each final NUREG-series publication in its native application. Therefore, please submit your final manuscript that has been approved by your NRC Project Manager- in both electronic and camera-ready copy.
The final ma uscript sh~II be t-archivj l qu ality and comply with the requirements of NRC Management Di ective 3.7 "NUREG-Serie-s Publications." The document shall be technically edited consistent with NURg_G-1379, Rev. 2 (May 2009) "NRC Editorial Style Guide." The goals of the "NRC Edito~ I Style Guide" are readability and consistency for all agency docume'.\,,
All format guidance, as specified in NUREG-0650, "Preparing NUREG-Series Publications,"
Rev. 2 (January 1999), will rema in the same with one exception. You will no longer be required to include the NUREG-series designator on the bottom of each page of the manuscript. The NRG will assign this designator when we send the camera-ready copy to the printer and will place the designator on the cover, title page, and spine. The designator for each report will no longer be assigned when the decision to prepare a publication is made. The NRC's Publishing Services Branch will inform the NRC Project Manager for the publication of the assigned designator when the final manuscript is sent to the printer.
For the electronic manuscript, the Contractor shall prepare the text in Microsoft Word, and use any of the following file types for charts, spreadsheets, and the like.
9
Version Control Date: November 1, 2014 File Types to be Used for NUREG-Series Publications File Type File Extension MicrosoftWord .doc Microsoft PowerPoint .ppt MicrosoftExcel .xis MicrosoftAccess Portable Document Format .pdf This list is subject to change if new software packages~ me into comL use at NRC or by our licensees or other stakeholders that participate in the electron*c submission process. If a portion of your manuscript is from another source and you cannot o tain a~ cceptable electronic fi le type for this portion (e.g., an ap~ndi from an old Rublication). the NRC can. if necessary, create a tagged image file format file extension.tif) for that portion of your report.
Note that you should continue to submil o.riginal11hotographs. ~~h will be scanned, since digitized photographs do not print well.
If you choose to publish a compact dis (CD) of your publication, place on the CD copies of the manuscript in both (1) a portable docum nt format (RDF); (2) a Microsoft Word file format, and (3) an Adobe Acrobat Reader, or, alternatively, print nstructions for obtaining a free copy of Adobe Acrobat Reader on the back ( /r insert of ttie jewel box.
16.0 OTHER CONSIDERATIONS (TYPE N/A IF NOT APPLICABLE)
Access to Non-NRC Facilities/Equipment (Type N/A if not applicable)
Show Guidance Applicable Publications (Type N/A if not applicable)
Show Guidance 10
Version Control Date: November 1, 2014 Controls over document handling and non-disclosure of materials (Type N/A if not applicable)
Show Guidance 11
Note to requester: The attachment is immediately following this email.
From: Hiser, Matthew Sent: Wed, 22 Apr 2015 20:30:01 +0000 To: Hull, Amy
Subject:
RE: SOW -- is there anything else I have to do on this before I fly on Saturday
????
Attachments: Figure 2.docx From: Hiser, Matthew Sent: Wednesday, April 22, 2015 4:30 PM To: Hull, Amy
Subject:
RE: SOW -- is there anything else I have to do on this before I fly on Saturday????
Hi Amy, Yeah, I think that is all we should need. I put together this Fig. 2 (attached Word doc)
It looks like you've incorporated most of my edits, which is good, but I still think some of the more global questions about the core focus (passive vs. active) and the detailed description in Task 1 (purpose of focus groups/ questionnaire; I don't see how strategic harvesting can be done for dry storage when there is no decommissioning for DCSSs, etc) are out there. At this point, why don't you send it to Steve and try to get his signoff before you leave? I'll do what I can next week, but can't promise anything as I have other responsibilities as well.
Thanks!
Matt From: Hull, Amy Sent: Wednesday, April 22, 2015 2:26 PM To: Hiser, Matthew
Subject:
SOW -- is there anything else I have to do on this before I fly on Saturday ????
Importance: High What else do I need to do before flying on Saturday? Can you complete the rest? I think now we just cost it out, asap. Where do I write you in as technical monitor? One of us needs to go see the EWA coordinator also and get her signature. I forget what her name os.
From: Hiser, Matthew Sent: Monday, April 20, 2015 4 :23 PM To: Hull, Amy
Subject:
RE: SteveF said you have to review SOW drafts before he does Hi Amy, Here is my feedback on the market research and EWC justification - I'll look at the SOW first thing tomorrow morning.
Thanks!
Matt Matthew Hiser Materials Engineer Corrosion and Metallurgy Branch Division of Engineering Office of Nuclear Regulatory Research 301 -251 -7601 From: Hull, Amy Sent: Monday, April 20, 2015 10:40 AM To: Hiser, Matthew
Subject:
SteveF said you have to review SOW drafts before he does Strategic Approach for Obtaining Aging Degradation Information from Decommissioning Nuclear Power Plants (Amy Hull, Matt Hiser)
Drafts of the SOW, market research analysis, and EWA task order justification have been submitted. Lisa Bamford sent an email on 4/16/2015 with a list of projects (including this) that need to have a requ isition submitted to initiate projects by the end of April.
From: Hull, Amy Sent: Monday, April 20, 2015 9:52 AM To: Frankl, I stvan (Istvan.Frankl@nrc.gov); Hiser, Matthew
Subject:
EWA task order justification documentation.ABH.docx
EPRI MDM BWRIMT MRP IMT Decommissioned BWRVIP-167 MRP-205 Reactors Roadmap Operating Prioritization of NRC-lndustry High Experience Strategic Harvesting Priority Data Needs Opportunities EMDA PMDA GALL for SLR NU REG/CR-7153 NUREG/CR-6923
From: Moyer, Carol Sent: Thu, 22 Oct 2015 13:02 :28 +0000 To: Stevens, Gary Cc: Boyce, Tom (RES)
Subject:
RE: DRAFT NRC Report to ASME for November Code Week Attachments: 2015-11 -- NRC Report to ASME - ATL (DRAFT)_cem.docx Note to requester: The attachment is immediately fo llowing this email.
Gary, My suggested edits are in the attached file . I added a new paragraph at the end of section 2, and I found a few miscellaneous editorial corrections.
-Carol From: Stevens, Gary Sent: Thursday, October 15, 2015 8:42 AM To: ASME Code Day Attendees <ASMECodeDayAttendees@nrc.gov>
Subject:
DRAFT NRC Report to ASME fo r November Code Week Attached for your review and input is a draft of the NRC Report to ASME for next month's meetings in Atlanta.
Please provide me with any input you may have y COB next Thursday, 10/23.
Thanks ,
Gary L. Stevens Senior Materials Engineer NRC/NRR/DE/EVIB E-mail : Gary.Stevens@nrc.gov Office: 301-415-3608 Fax: 301-415-2444
NRC Report November 2015 NRC Report for ASME Code Meetings - DRAFT FOR REVIEW November 2015 - Atlanta, GA Contents I. Amendment to 10 CFR 50.55a -ASME Code Edition/Addenda ............................................................... 2
2. ASME Code Case Rulemaking/Regulatory Guides ................................................................................. 2
3. Operating Plant Issues and Material Dcgradntion ....................................................................................4
4. NRO OCIP Quality Assurance and Vendor Inspection Branch Activities. .. ..4
5. New Reactor Licensing Activities........................................................... .. ....................6
6. Mu hinational Design Evaluation Program (MDEP) Acti vit ics................................................................ 10
7. 10 CFR Part 21 Rulemaking ............................................................................................................... 12
8. Commercial Calibration Services Status ............................................................................................... 12
9. NRC staff Review of EPRI 1025243 Guideline for Commcrcial*G radc Design and Analysis Computer Programs............. ........................................................................................... 14 I0. EPRI Development of Revised Dedication Gu idance for Commcrcial*Grade Hems for Use in Nuclear Safety*Related Applications ........................................................................... 14
11. NRC StafTln1erfacc with Nuclear Utilities Procurement Issues Committee (NUPIC) ............................... 15
12. Counterfeit, Fraudulent and Suspect Items (CFSI) Commission Paper ................................................... 16
13. License Renewal Activities ................................................................................................................ 16
14. New Generic Lellcrs ............................... ........................................................................................... 20
15. New lnfom1arion Notices................. .. ................. 20 I6. New Rcg11latory Issue Summaries .... ..20
17. NRC Publications of Potentia l lntcrcs110 ASME .................................................................................. 20
18. Upcoming Public Meetings of Potential lnteresl to ASME .....................................................................21 NRC Report November 2015
1. Amendment to 10 CFR 50.55a - ASME Code Edition/Addenda Current ASME Edition/Addenda The NRC has approved:
Section Ill, Division 1 and Sectijon XI, Division 1 of the Boiler and Pressure Vessel Code through the 2008 Addenda (76 FR 36232).
The Operation and Maintenance of Nuclear Power Plants (OM Code) through the 2006 Addenda (76 FR 36232).
Next ASME Edition/Addenda The next proposed amendment to 10 CFR 50.55a includes:
The 2009 Addenda, the 2010 Edition, 2011 Addenda, and the 2013 Edition of the Boiler and Pressure Vessel Code.
The 2009 Edition, 2011 Addenda and 2012 Edition of the Operation and Maintenance of Nuclear Power Plants (OM Code).
Section XI Code Case N-824 w ill be directly listed in 50.55a as conditionally approved for use.
Section XI Code Case N-729-4 will be directly listed in 50.55a as required with conditions.
Section XI Code Case N-770-2 will be directly listed in 50.55a as required with conditions.
The proposed rule was published in the Federal Register on Friday, September 18, 2015.
The 75 day public comment period closes on December 2, 2015 (link to the FRN:
http://www.gpo.gov/fdsys/pkg/FR-2015-09-18/pdf/2015-23193.pdf) . ASME made the materials that are incorporated by reference available in a read-only format for the public comment period at the following link: http://go.asme.org/NRC.
2. ASME Code Case Rulemaking/Regulatory Guides Current RG Publications On November 5, 2014 a final rule was published in the Federal Register (79 FR 65776) that incorporates by reference the Regulatory Guides (RGs) listed below:
.Supplements Addressed: Supplements 1 through 10 to the 2007 Edition Effective date for the RGs: December 5, 2014 NRC Report November 2015
RG 1.84, Revision 36, "Design, Fabrication, and Materia ls Code Case Acceptabi lity, ASME Section Ill" (ADAMS Accession No. ML13339A515).
RG 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1" (ADAMS Accession No. ML13339A689).
RG 1.192, Revision 1, "Operation and Maintenance Code Case Acceptability, ASME OM Code" (ADAMS Accession No. ML13340A034).
In addition, on November 5 , 2014, a final guide was published in the Federal Register (79 FR 65776):
RG 1.193, Revision 4 , "ASME Code Cases Not Approved for Use* (ADAMS Accession No. ML13350A001).
Next RG Publications The staff is completing the proposed Code Case Rulemaking package and expects a publication in the Federal Register in the October/November 2015 timeframe. Following the publication of the proposed rule will be a 75 day public comment period.
ASME Code Code Case Supplements Addressed: Supplement 11 to the 2007 Edition through Supplement 10 to the 2010 Edition
Draft Revision 37 of RG 1.84
Draft Revision 18 of RG 1.147
Draft Revision 5 of RG 1.193 Additional Code Cases considered for this rulemaking package at the request of ASME that are not listed in aforementioned Supplements:
N-694-2 (Supp. 1 to the 2013 Edition) "Evaluation Procedure and Acceptance Criteria for PWR Reactor Vessel Head Penetration Nozzles"Section XI
N-825 (Supp. 3 to the 2013 Ediition) "Alternative Requirements for Examination of Control Rod Drive Housing Welds"Section XI
N-845 (Supp. 6 to the 2013 Ediition) "Qualification Requirements for Bolts and Studs"Section XI OM Code Code Cases Addressed: 2009 Edition through 2012 Edition
Draft Revision 2 of RG 1.192 The NRC staff also initiated the review of the next draft RGs that will address the Code Cases published in Supplement 11 to the 2010 Edition through Supplement Oto the 2015 Edition of the ASME Code.
NRC Report November 2015 Standards Used in RGs and Other Guidance The NRC has placed on its website a series of lists of consensus standards, including those published by ASME that are referenced in Regulatory Guides in Inspection Manuals and Procedures, and in the LWR Standard Review Plan. The lists may be found at this website:
http://www.nrc.gov/about-nrc/regulatory/standards-dev/consensus.html
3. Operating Plant Issues and Material Degradation MRP-146 Thermal Fatigue in Normally Non-lsolable Reactor Coolant System Branch Lines Inspections Recent operating experience has indicated that the guidance provided by MRP-146 does not conservatively predict locations where thermal fatigue may occur. Plants have experienced thermal fatigue flaws in locations that had been screened out by the generic model provided by MRP-146. MRP assembled a response team and issued interim guidance to utilities and have now issued revised guidance for MRP 146 which is under review.
Peening The NRC staff is currently reviewing the MRP-335, Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement, which discusses peening as a mitigation technique to permit reduction in inspection frequency for Alloy 600 surfaces exposed to PWR reactor coolant. MRP has informed the NRC staff that they will be issuing an updated version of MRP-335 in the near future.
Operational Leakage On March 23, 2015, ASME sent the NRC a letter describing the conclusion of their activities for addressing operational leakage of pressure reta ining components (ADAMS Accession No. ML15099A624). In view of ASME's Pressure Boundary Leakage project team's conclusions, the NRC sent a letter back to ASME on July 14, 2015 stating that they will evaluate the necessity of additional regulatory activities to address operational leakage (ADAMS Accession No. ML15188A057).
4. NRO DCIP Quality Assurance and Vendor Inspection Branch Activities NRO Vendor Inspection The NRO vendor inspection program is described in Inspection Manual Chapter (IMC) 2507, "Vendor Inspections." This IMC was last updated on October 3, 2013. This IMC is implemented by various Inspection Procedures (IPs) including:
NRC Report November 2015 IP 43002: Routine Inspections of Nuclear Vendors; IP 43003: Reactive Inspections ,of Nuclear Vendors; IP 43004: Inspection of Commercial-Grade Dedication Programs; IP 43005: NRG Oversight of Third Party Organizations Implementing Quality Assurance Requirements; IP 36100: Inspection of 10 CFR Part 21, Programs for Reporting Defects and Noncompliance; IP 37805: Engineering Design Verification Inspections; IMC 0617: Vendor and Quality Assurance Implementation Inspection Reports; and IMC 2506: Construction Reactor Oversight Process General Guidance Basis Document FY 15 Vendor Inspection Plans
AP1000 modular construction
AP1000 mechanical and electrical qualification test programs
Digital Instrumentation and Control for AP1000
Valve and pump manufacturing
Commercial-grade dedication organizations Vendor Inspection Reports Issued. Completed. and Planned Inspections
Crane Nuclear, Inc. IL - issued
WEC, Warrendale, PA - issued
SPX/Penn State University, PA- issued
Fisher Control, Marshalltown, IA - issued
Specialty Maintenance & Construction, FL - issued
Kinetrics, Toronto, CA - issued
Thermo-Fisher Scientific, San Diego, CA - issued
C&D Technologies, Blue Bell, PA - completed
NuScale Power, LLC, OR - issued
Curtiss Wright-QualTech, Cincinnati, OH - planned
Curtiss Wright EMD, Cheswick, PA - planned
Carboline, St. Louis, MO - planned
Canberra Industries, Meriden, CT - planned
Valcor Engineering, Springfiel,d, NJ - planned
Aecon Industrial, Cambridge, Canada - planned
GE Dresser Consolidated, Alexandria, LA - planned Previously issued NRG inspection and trip reports are located at:
http:llwww.nrc.gov/reactors/new-reactors/oversight/quality-assurance/vendor-insp/insp-reports.html New Vendor Inspection Quality Assurance Website Links NRC Report November 2015 The NRG has implemented website pages to make it easier to become fami liar with and follow vendor inspection and QA related activities:
http://nrcweb.nrc.gov:400/reactors/new-reactors/oversighVquality-assurance/vendor-insp.html.
As part of the Vendor Outreach and Communications Strategy, the NRG is planning its 2016 Biannual Vendor Workshop in coordination with NUPIC vendor meeting, scheduled to be held in St. Louis, MO on June 23-25, ~015. Members of the nuclear industry are ( Commented (CMl]: 2016?
encouraged to submit topics of interest for the 2016 Biannual Vendor Workshop to NRC representatives, Richard McIntyre (richard.mcintyre@nrc.gov), Kerri Kavanagh (kerri.kavanaqh@nrc.gov), or Andrea Keim (andrea.keim@nrc.gov) by January 31, 2016.
Presentations from the 2014 Vendor Workshop and past workshops are available on our public website at the below link.
http://www.nrc.gov/reactors/new-reactors/oversight/quality-assurance/vendor-oversiqht.html The Frequently Asked Questions (FAQ} page addresses Quality Assurance for New Reactors and currently has three main categories: 10 GFR Part 21 FAQs, Commercial Grade Dedication FAQs, and Enforcement FAQs. The page provides quick links to questions we have received in the past about the mentioned topics:
http://www.nrc.gov/reactors/new-reactors/oversiqht/quality-assurance/qual-assure-faqs.html The web page link below serves as a categorization tool and provides a list of all applicable QA Inspections for New Reactor Licensing and Vendor QA Inspection reports that have either a Notice of Nonconformance (NON) or Notice of Violation (NOV) within a specific criterion of 10 CFR 50 Appendix B or 10 CFR Part 21 related issue. The page is routinely updated with every new inspection report that is released:
http://www.nrc.gov/reactors/new-reactors/oversiqht/guality-assurance/nonconformances-violations.html The web page link below describes the vendor inspection program (VIP). The VIP verifies that reactor applicants and licensees are fulfilling their regulatory obligations with respect to providing effective oversight of the supply chain. It is accomplished through a number of activities, including: performing vendor inspections that will verify the effective implementation of the vendor's quality assurance program, establishing a strategy for vendor identification and selection criteria, and; ensuring vendor inspectors obtain necessary knowledge and skills to perform inspections. In addition, the VIP addresses interactions with nuclear consensus standards organizations, industry and external stakeholders, and international constituents:
http://pbadupws.nrc.gov/docs/ML1527/ML15272A080.pdf.
5. New Reactor Licensing Activities
. 6.
NRC Report November 2015 As of October 13, 2015, the status of new reactor licensing under 10 CFR Part 52 is as follows:
Design Certification NRC has issued five design certifications to date (ABWR, System 80+, AP600, AP1000 and ESBWR). These are certified in 10 CFR Part 52, Appendices A, B, C, D , and E respectively.
The NRC staffs review of the AREVA's EPR (evolutionary pressurized-water reactor design from France) is suspended at the request of the applicant in its letter dated February 25, 2015, until further notice.
The NRC staff's review of the Mitsubishi Heavy Industries' US-APWR design certification application (for an advanced pressurized-water reactor design from Japan) is currently on hold at the request of the applicant except for a few key areas.
The NRC staff completed its review of General Electric-Hitachi's ESBWR (first passive BWR) and issued its final safety evaluation report (FSER) in March 201 1. On March 24, 2011, the NRC issued in the Federal Register a proposed rule (76 FR 16549) for public comment on the ESBWR design certification. The NRC final rule adding Appendix E to 10 CFR Part 52 to certify the ESBWR standard design was published on October 15, 2014 in the Federal Register (79 FR61983) and became effective on November 14, 2014.
The Korea Hydro and Nuclear Power (KHNP) submitted a standard design certification application for its APR-1 400 standard plant design to the NRC on September 30, 2013.
The NRC staff conducted an acceptance review of the application for completeness, technical adequacy, and acceptability for docketing. In a letter to KHNP dated December 19, 2013, the NRC staff discussed the results of its acceptance review. The NRC noted that it decided not to accept the application for docketing at that time because the application was not ready in several key areas. The NRC staff continued pre-application interactions with KHNP to support preparation of a complete application by December 2014. On December 23, 2014, KHNP resubmitted the standard design certification application for its APR-1400 design. The NRC staff accepted the APR1400 design certification application for docketing in its letter dated March 4, 2015, based on its determination that the application is sufficiently complete and technically adequate to allow the NRC staff to conduct its detailed technical review.
In addition, the NRC staff is reviewing two applications for design certification renewal:
ABWR GE-Hitachi (application submitted on December 7, 2010)
ABWR GE-Toshiba (Revision 1 to application submitted on June 22, 2012)
Early Site Permits (ESPs)
NRC Report November 2015 NRG has issued four ESPs (Clinton, Grand Gulf, North Anna, and Vogtle). The NRC's issuance of the Vogtle ESP on August 26, 2009, was the first based on a specific technology (AP-1000) and the first to include a limited-work authorization (LWA).
The NRG received an application for an ESP for the Victoria County Station submitted by Exelon on M arch 25, 2010. The site is located in Victoria County, Texas, with no specific technology selected. On August 28, 2012, Exelon requested withdrawal of the V ictoria County Station ESP application from *the docket. By letter dated October 3,. 2012, NRG accepted the applicant's request, and the application was withdrawn.
The NRG received an ESP application for the PSEG site in New Jersey (same site as Hope Creek and Salem 1&2). The ESP application was tendered on May 25, 2010, and was docketed on August 4, 2010. This application uses the Plant Parameter Envelope (PPE) approach which means no specific reactor design has been selected.
Combined License (COL) Applications NRC is currently reviewing 9 COL applications (14 new reactor units):
1 ABWR: South Texas Project 3 and 4
- 3 AP-1000: William S. Lee Station 1&2, Shearon Harris 2&3*, Levy County 1&2, Bellefonte 3&4*, and Turkey Point 6&7
- 2 ESBWR: Fermi 3, North Anna 3, Grand Gulf 3*, River Bend 3*,
Victoria County 1 and 2**
- 2 EPR: Calvert Cliffs 3**. Bell Bend*, Nine Mile Point 3**. Callaway 2*
1 US-APWR: Comanche Peak Units 3 and 4
NRC staff review suspended at request of applicant.
" Application withdrawn.
On June 8, 2015, Unistar requested to withdraw the Calvert Cliffs, Unit 3 combined license application.
On April 25, 2013, Dominion V irginia Power revised its technology selection from the US-APWR nuclear technology and selected the GEH ESBWR nuclear technology for the North Anna Unit 3 project. The initial phase of the North Anna Unit 3 combined license application was submitted to the NRC in July 2013, and the final portion of the application was submitted in December 2013.
The NRC issued the combined licens,e and limited work authorization for Vogtle Electric Generating Plant, Units 3&4 on February 10, 2012. The Vogtle plants reference the AP1000 design certification amendment. It was the first combined license issued by the NRC to construct and operate a nuclear power plant under the a lternative licensing process in 10 CFR Part 52. It is the first time s ince 1978 that the NRC issued a license to construct a nuclear power plant in the United States.
The NRC staff issued the combined license for V .C. Summer 2&3 on March 30, 2012. The V.C. Summer 2&3 plants reference the AP1000 design certification amendment.
NRC Report November 2015 On February 4 , 2015, the NRG Commissioners held a mandatory hearing on the combined license for Fermi, Unit 3. On May 1, 2015, the NRG issued the combined license for Fermi, Unit 3. This is the first combined license for an application referencing the ESBWR design.
On June 8, 2015, UniStar requested to withdraw the Calvert Cliffs, Unit 3 combined license application.
Advanced Reactors Program NRG established an advanced reactors program in the Office of New Reactors. Currently, there are no applications under review, but several applications are expected in the next three years including:
Integral PWRs (iPWRs):
NuScale (iPWR) - NuScale Power is developing a modular, scalable 50 MWe iPWR. Pre-application reviews are currently under discussion. The design certification is expected to be submitted to the NRG in November or December of 2016 .
B&W mPower (iPWR) - B&W is developing a modular, scalable 180 MWe iPWR. At this time, mPower has reduced its activities in the mPower development, and have not provided a submittal date for the application.
TVA is planning to submit its early site permit in the first quarter of 2016 for its Clinch River site near Oakridge, Tennessee.
Holtec is developing the Holtec Inherently Safe Modular Underground Reactor SMR 160 design that has a 160 MWe electrical power output. They plan to pursue a Part 50 licensing process that requires an applicant to apply for a construction permit and a subsequent operating license. They have not provided an application submittal date.
XEnergy has indicated it plans to submit a design certification application to the NRG within the next few years for its pebble-bed high temperature gas-cooled reactor. The XEnergy reactor (Xe-100) is a helium-cooled reactor with a power rating of 125 MWt.
Advanced Reactor (non-light water reactors) Guidance Development:
NRG has received Idaho National Laboratory (INL) generated Department of Energy technical report "Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors." T he INL report is the culmination of phase one of a two-phase initiative by the DOE and the NRC to develop advanced reactor safety design criteria from which the principal design criteria could be derived for advanced reactor concepts. The NRG w ill follow its normal process for developing and issuing regulatory guidance and anticipates completion of such guidance by the end of 2016.
Construction Issues:
AREVA reported lower than expected mechanical properties for the reactor vessel at its EPR reactor in Flamanville , France. Additiona l testing of the material will be NRC Report November 2015 performed. The NRC is following this issue as it may pertain to any new or replacement components.
During an NRC vendor inspection, it was obseived that the vendor was using ASME Code NPT stamps on safety related piping that were not in accordance with the ASME Code. The vendor had no official documentation from ASME confirming that these NPT stamps were valid. The authorized nuclear inspector (ANI) initiated action with ASME to ensure that the ASME Code stamps are valid, and ASME has committed to initiate action to replace these NPT stamps. ASME sent a letter to on June 17, 2015 to all ASME NPT Certificate holders regarding the acceptability of these NPT stamps. The ASME letter also requested that all NPT Certificate holders send in these NPT stamps for replacement in order to avoid further confusion in the industry. On October, 2, 2015, ASME supplemented its June 17, 2015 letter by requesting that all NPT stamps with the NPT letters arranged horizontally to be returned to ASME for replacement by December 31, 2015.
6. Multinational Design Evaluation Program (MDEP) Activities MDEP is a multinational initiative to develop innovative approaches to leverage the resources and knowledge of mature, ,experienced national regulatory authorities who are tasked with the regulatory design review of new reactor plant designs. Some of the issue-specific working groups established under the MDEP organization that the NRC participates in are the Codes and Standards Working Group (CSWG). whose goal is to achieve harmonization of code requirements for pressure-boundary components. and the Vendor Inspection Cooperation Working Group (VICWG). whose goal is to maximize the use of the results of inspections obtained from other regulator' s'. efforts in inspecting vendors.
Vendor Inspection Cooperation Working Group fVICWG)
The MDEP VICWG was formed because component manufacturing is currently subject to multiple inspections and audits similar in scope and in safety objectives, but conducted by different regulators to different criteria. The primary goal of the VICWG is to maximize the use of the results obtained from other regu lator's'. efforts in inspecting vendors.
The MDEP VICWG continues to achieve its short-term goals and is making progress towards achieving its long term goals. The VICWG continues to focus on maximizing information sharing, joint inspections (multiple regulators inspecting to the regulatory requirements of one country), and witnessing of other regulators' inspections. NRC has participated in 6 witnessed and joint inspections this year to date. Additional MDEP inspections are also occurring that do not involve the NRC, but an exact count has not been established for FY15.
The working group enhances the understanding of each regulator's inspection procedures and practices by coordinating witnessed inspections of safety related mechanical pressure retaining components (Class 1) such as pressure vessels, steam generators, piping, NRC Report November 2015 valves, pumps, etc., and quality assurance inspections. Witnessed inspections consist of one regulator performing an inspection to its criteria, observed by representatives of other MDEP countries. The benefits to the observing countries include additional information and added confidence in the inspection results. MDEP regu lators are using the experience gained during conduct ofVICWG witnessed inspections in their inspection planning.
The MDEP VICWG held its 15th meeNng during the week of May 19 in Shenzhen, China. This meeting included members from, China, Japan, the Republic of Korea, France, South Africa, Finland, the United Kingdom and the United States. Canada, Sweden, United Arab Emirates, the Russian Federation, India and Turkey were not in attendance. Because the meeting was hosted in China. it allowed multiple members of China's NNSA, NRO, and NSC to participate in the meeting. Because China is involved with the construction of many designs, they have shared interests with many of the other VICWG countries. NRC participated in 6 MDEP inspection activities during the past year. The Next VICWG meeting is scheduled to be held in Dijon, France the week of October 26, 2015. On October 30th , representatives from ASME, ISO, WNA and others are invited to discuss their latest initiatives.
Codes and Standards Working Group (CSWG/
The MDEP group's goal is to harmonize and converge national codes, standards, and regulatory requirements and practices applicable to pressure boundary components while recognizing the sovereign rights and responsibilities of national regulators in carrying out their safety reviews of new reactor designs. The CSWG published several reports on codes and standards related to pressure boundary components, and it provides a regulatory forum for groups such as the World Nuclear Association's Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group to coordinate with concerning international harmonization efforts.
In p ctober the CSWG met with representatives from CORDEL and ASME. A report from ( Commented [CM2]: 2015?
CORDEL detailed a proposal to help reconcile Non-Destructive Testing Certifications internationally. The representative from CORDEL presented this report at the Standards Development Organization (SDO) Convergence meeting during the November, 2014, ASME Code Meeting. In µanuar~. CORDEL representatives will also attend a meeting with I Commented [CM3]: 2016?
the Chinese concerning their codes and standards work.
ASME representatives informed the CSWG that they have accepted two initial topics, Welding Qualification and Fatigue Roadmap. A statement of work (SOW) has been finalized for developing an international comparison report on Welding Qualification and Welding Quality Assurance, and a SOW is being finalized for developing an International Fatigue Roadmap to sur.port a harmonized set of rules to perform fatigue analysis of Class 1 components. For the November 2015 CSWG meeting two draft reports have been Commented [CM4]: Do 1hcir meet in~ hove a sci frequency'? 11 seems odd that (according to the 1>rev. paragraph) they met in provided concerning Welding Qualification and Welding Quality Assurance; and October and will meet again in January, and they will also mcc1 in Certification of NDE Personnel. The CSWG will be reviewing and providing comments on November.
both documents in the November meeting.
NRC Report November 2015 CSWG will continue to follow closely t he activities of the SDOs and CORDEL through 2015.
at which time the CORDEL pilot program for convergence using the SDO Convergence Board process is expected to complete at least one code convergence. Also, in ~016, the (f,mmented [CMS] : 2016'!
CSWG may discuss the possibility of turning over its regulatory interface with the SDOs and CORDEL's activities to another international regulatory organization (e.g., NEA's Committee on the Safety of Nuclear Installations). as many of the topics are growing beyond the MDEP mandate.
NRC Report November 2015
7. 10 CFR Part 21 Rulemaking1 On September 29, 2011 , the NRG staff issued SECY-11-0135, "Staff Plans to Develop the Regulatory Basis for Clarifying the Requirements in Title 10 of the Code of Federal Regulations PART 21 , "Reporting of Defects and Noncompliance." This Commission Paper informs the Commission of the staff's plan to develop the regulatory basis to clarify Title 10 of the Code of Federal Regulations (10 CFR) Part 21 , "Reporting of Defects and Noncompliance." The staff will make modifications to Part 21 and regulatory guidance will be proposed.
In December 2012, the staff published the Draft Regulatory Basis to Clarify 10 CFR Part fl, "Reporting of Defects and Noncompliance," Revision 0. The NRG provided this version of the regulatory basis as a draft to promote early stakeholder feedback. Other associated information can be found on the Federal Government's regulations web site, by searching for "NRC-2012-0012." The goal of this regulatory basis is to simplify and clarify the rule language in Part 21, provide consolidated regulatory guidance on compliance with Part 21, and enhance regulatory stability and predictability for the entities to which Part 21 applies.
On August 7, 2015 the fina l Regulatory Basis was issued. This includes clarifications to Section 21.3, "Definitions," related to ,evaluation and reporting of defects that could create a substantial safety hazard and terms related to commercial grade dedication. Also, this revision to Part 21 introduces a separate section in Part 21 specifically related to commercial grade dedication activities.
The NRG staff wW-ookJheld a public meeting on October 16, 2015, to discuss the NEI 14-09, : Guidelines for Implementation of 10 CFR Part 21, Reporting of Defects and Noncompliance,: Revision 0, dated August 2014. NEI developed this guidance to incorporate previous guidance in NUREG-0302, to include additional clarity in specific areas where issues have historically occurred and to include experience gained from nearly 30 years complying with the 10 CFR P art 21 ru le. The NRC staff expects to issue the proposed rulemaking for public comment by the end of 2016.
8. Commercial Calibration Services Status By letter dated April 29, 2014, the Nuclear Energy Institute (NEI) submitted Revision O of NEI 14-05, "Guidelines for the Use of Accreditation in Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services," to the U.S. Nuclear Regulatory Commission (NRC) for NRC staff review and endorsement. NEI 14-05 provides an approach for licensees and suppliers of basic components for using laboratory accreditation by Accreditation Bodies (ABs) that are signatories to the International Laboratory Accreditation Cooperation (ILAC) Mutual Recognition Arrangement (MRA)
(hereby after referred to as the ILAC accreditation process) in lieu of perform ing commercial-grade surveys for procurement of calibration and testing services performed by domestic and international laboratories accredited by ILAC signatories.
NRC Report November 2015 By letter dated July 22, 2014, the NRC requested additional information to complete its review of NEI 14-05. Two teleconferences were held on July 3, 2014 and A ugust 13, 2014, to promote a better understanding of the NEI responses to the NRC's request for additional information. By a letter dated August 28, 2014, NEI submitted NEI 14-05 Revision 1, which incorporates NEl's responses to the NRC staffs comments on NEI 14-05.
By letter dated February 9, 2015 (ADAMS Accession No. ML14322A535), the NRC staff transmitted its safety evaluation (SE) identifying the guidelines contained in NEI 14-05, Revision 1 (ADAMS No. ML14245A391) as an acceptable approach for licensees and suppliers to meet the requirements of 10 CFR Part 50, Appendix B to use ILAC laboratory accreditation as part of the commercial-grade dedication process for procurement of calibration and testing services. NRC's endorsement of NEI 14-05, Revision 1, expands the NRC's acceptance of the ILAC accreditation process first documented in SE on an Arizona Public Service request (ADAMS Accession No. ML052710224). NRC's earlier acceptance was limited to calibration laboratory services accredited by specific U.S.
Accrediting Bodies (ABs). The SE (1) confirms that NEI 14-05, Revision 1, reflects the ILAC accreditation process previously approved; (2) provides an evaluation of the unique aspects of NEI 14-05, Revision 1: (3) constitutes formal NRC endorsement of the guidelines in NEI 14-05, Revision 1, for using the ILAC accreditation process in lieu of performing a commercial-grade survey; and (4) finds that the ILAC accreditation process continues to satisfy the requirements of Appendix B to 10 CFR Part 50 and, therefore, is acceptable.
Although the NRC has endorsed NEI 14-05, Revision 1, licensees and suppliers of basic components use of the ILAC accreditation process in lieu of performing a commercial-grade survey represents a reduction in commitment to the previously accepted QA program. As such, once the NRC approves the QA program change for a licensee in accordance with 10 CFR 50.54(a}(4), other licensees may adopt the QA alternative of using the ILAC accreditation process in lieu of performing a commercial-grade survey provided that the bases of the NRC approval are applicable to the licensee's facility pursuant to the requirements of 10 CFR 50.54(a)(3)(ii). On October 7, 2015, Ameren Missouri, licensee of the Calloway plant, submitted a license amendment to adopt NEI 14-05, Revision 1.
NEI 14-05, Revision 1, will allow licensees and vendors to use the ILAC accreditation process in lieu of performing commerdal-grade surveys for procurement of calibration and testing services performed by domestic and international laboratories accredited by signatories to the ILAC MRA. Suppliers of basic components may begin to use the ILAC accreditation process and do not have to wait for the NRC to approve the QA program change for a licensee provided the conditions from the NEI 14-05, Revision 1, safety evaluation are met.
On August 4, 2015, the NRC issued draft Regulatory Issue Summary (RIS), 2015-XX, "Use of Accreditation in lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services," to notify external stakeholders that the NRC staff has found acceptable for use Revision 1 to NEI 14-05, with respect to procurement and dedication of calibration and testing services performed by domestic and international laboratories. The NRC Report November 2015 draft RIS is available in ADAMS under Accession No. ML15090A236. The Public comment period closed on October 4, 2015.
9. NRC Staff Review of EPRI 1025243 Guideline for Commercial-Grade Design and Analysis Computer Programs By letter dated July 18, 2012, the Nuclear Energy Institute (NEI) submitted E lectric Power Research Institute (EPRI) 1025243, Plant Engineering: Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Safety-Related Applications for staff review and approval. EPRI 1025243 describes a dedication methodology for commercial-grade design and analysis computer programs for use in meeting regulatory requirements. The letter proposes the NRC endorse the EPRI 1025243 computer program commercial-grade dedication method, which follows the method provided in EPRI NP-5652, which the NRC conditionally endorsed in Generic Letter 89-02.
In the fall of 2012, the NRC staff reviewed EPRI 1025243 with reference to the dedication requirements of 10 CFR Part 21 and Appendix B to 10 CFR Part 50.
Draft Regulatory Guide (DG )-1305, "Acceptance of Commercial-Grade Design and Analysis Computer Programs for Nuclear Power Plants." was issued to endorse Revision 1 or EPRI Technical Report 1025243, as an acceptable method in meeting regulatory requirements. The initial version of E PRI 1025243 was the first guidance document to provide a detailed dedication methodology specific to commercial-grade design and analysis computer programs.
On July 1, 2015, the NRC issued for public comment DG- 1305, "Acceptance of Commercial- Grade Design and Analysis Computer Programs for Nuclear Power Plants."
The DG provides new (i.e., not preceded by earlier guidance on the same subject) guidance that describes acceptance methods that the staff of the NRC considers acceptable in meeting regu latory requirements for acceptance and dedication of commercial-grade design and analysis computer programs for nuclear power plants.
The NRC staff is currently evaluating the comments received from NEI and other stakeholders with expected completion by October 31, 2015.
10. EPRI Development of Revised Dedication Guidance for Commercial-Grade Items for Use in Nuclear Safety-Related Applications In September, 2014, the EPRI issued the 2014 Technical Report 3002002982, "Plant Engineering: Guideline for the Acceptance of Commercial-Grade Items in Nuclear Safety-Related Applications" - Revision 1 to EPRI NP-5652 and TR-102260. The NRC staff participated during many of the EPRI technical advisory group (TAG) meetings held at EPRl's offices in Charlotte, North Carolina.
NRC Report November 2015 This report describes a methodology that can be used to dedicate commercial-grade items for use in safety-related applications. The scope of applications for which commercial-grade item dedication is used has evolved significantly since the EPRI published its reports Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications (NCIG-07) (NP-5652) and Supplemental Guidance for the Application of EPRI Report NP-5652 on the Utilization of Commercial Grade Items (TR-102260) in 1988 and 1994, respectively. The guidance in this fina l report reflects lessons learned and addresses challenges that have been identified through expanded use of the original guidance. This report supersedes both original reports in their entirety.
The NRC has developed a Draft Guide DG-1292, "Dedication of Commercial Grade Items" for potential NRC endorsement and is currently in the review process to approve Revision 1 to EPRI Technical Report 3002002982 with respect to acceptance of commercial-grade dedication of items used as basic components for nuclear power plants. The initial version of EPRI NP-5652 (Ref. 5) was the first guidance document to provide a detailed acceptance methodology specific to commercial-grade dedication (CGD) for items used in nuclear power plants.
The industry's use of the CGD process has significantly increased over time as the number of suppliers with nuclear quality assurance programs has decreased. However, current industry dedication guidance was developed in the late 1980's and the NRC has only previously endorsed EPRI dedication guidance in Generic Letter 89-02, "Actions to Improve the Detection of Counterfeit and Fraudulently marketed Products," and Generic Letter 91-05, "Licensee Commercial-Grade Procurement and Dedication Programs." Therefore, a need existed for the NRC to review and endorse the latest EPRI guidance for commercial grade dedicationJ.Q which the NRC participated in the development stage. EPRI Technical Report 3002002982 is available free of charge to the public on the EPRI website (EPRl.com).
11 . NRC Staff Interface with Nuclear Utilities Procurement Issues Committee (NUPIC)
On June 15 -18, 2015, NRC staff participated and made presentations at the 24th Annual NUPIC Vendor Meeting in Henderson NV. NRC presentations were related to the Part 21 rulemaking and its effect on commercial grade dedication and an update on vendor inspection activities. The NRC periodically accompanies a NUPIC Joint Utility Audit team to observe selected audits and ensure that the audit process remains an acceptable alternative to the NRC's vendor inspection/audit program. The NRC staff continues to rely on the effectiveness of the NUPIC joint utility audit process for evaluating the implementation of quality assurance programs of suppliers to the nuclear industry. The NRC uses trip reports to document NRC observation of audits performed by NUPIC. The NRC team has used NRR/NRO Inspection Procedure (IP) 43005, "NRC Oversight of Third-Party Organizations Implementing Quality Assurance Requirements," during the NUPIC observation oversight. Trip reports issued for the NUPIC audits that have been observed by the NRC are available at the below web-site link:
http://www.nrc.gov/reactors/new-reactors/oversighUguality-assurance/nupic-industry.html NRC Report November 2015
12. Counterfeit, Fraudulent and Suspect Items (CFSI) Commission Paper In SECY-11-0154, "An Agencywide Approach to Counterfeit, Fraudulent, and Suspect Items (CFSI)," dated October 28, 2011, the staff informed the Commission of activities to assess and enhance processes for addressing counterfeit, fraudulent, and suspect items in NRC regulated activities. This new SECY paper provides a status of the completed and ongoing activities originally presented in SECY-11-0154, ADAMS Accession No. ML112200150).
On January 8, 2015, the NRC staff published the SECY Paper SECY-15-0003, "STAFF ACTIVITIES RELATED TO COUNTERFEIT, FRAUDULENT, AND SUSPECT ITEMS. This SECY paper informs the commission of the staffs progress in identifying and implementing proactive strategies to detect and prevent the intrusion of counterfeit, fraudu lent, and suspect items into equipment, components, systems, and structures regulated by the NRC.
This paper is the final planned status update on the topic. The discussion section provides the status of each of the 19 planned actions from SECY-11-0154. Of these 19 planned actions, 14 have been completed and the remainder will be completed by December 2018.
On June 24, 2015, the NRC issued Regulatory Issue Summary (RIS) 2015-08, "Oversight of Counterfeit, Fraudulent, and Suspect Items in the Nuclear Industry." The NRC issued this RIS to heighten awareness of the existing NRC regulations and how they apply to CFSI within the scope of NRC's regu latory jurisdiction. Addressees are expected to review this information and consider actions, as appropriate, to prevent CFSI from entering their supply chains, prevent possible installation or use of CFSI at their facilities, and raise awareness of the potential for CFSI to be used in the manufacture, maintenance, or repa ir of items, including sealed sources and devices (SSDs). The RIS is available in ADAMS under Accession No. ML15008A191.
13. License Renewal Activities Following are on-going activities relat,ed to license renewal:
Current status of applications, staff reviews and approvals
78 units approved (76 operating plants with renewed licenses) o Renewed licenses for Sequoyah 1 & 2 issued September 24, 2015 o 1 (2 units) in hearings (Indian Point 2 & 3) - supplemental SER issued November 2014 o 1 (4 units) completed ACRS Full committee (Byron 1 & 2 / Braidwood 1 & 2 (9/2015])
o 1 (1 unit) awaiting ACRS Full committee (Davis-Besse (9/2015])
NRC Report November 2015 o 2 (3 units) awaiting follow-up ACRS Subcommittee (Seabrook [12/2015],
Diablo Canyon 1 & 2 (10/16])
o 4 (6 units) awaiting ACRS Subcommittee (Fermi (TBD], LaSalle 1 & 2 (3/2016], Grand Gulf [TBD], South Texas Project 1 & 2 [TBD])
1 application (1 unit) with scheduled submittal in 2016:
o January to March 2016- Waterford 3 o January to March 2017 - River Bend o July to September 2018- STARS o October 2019 - Perry o January to March 2021 - Clinton Thirty-nine units have entered the operating period beyond 40 years:
o Oyster Creek - April 9, 2009 o Prairie Island 1 - August 9, 2013 o Nine Mile Point 1 - August 22, 2009 o Indian Point 2 - September 28, o Ginna - September 18, 2009 2013 o Dresden 2 - December 22, 2009 o Oconee 2 - October 6, 2013 o H.B. Robinson - July 31, 2010 o Browns Ferry 1 - December 20, o Monticello - September 8, 2010 2013 o Point Beach 1 - October 5, 2010 o Cooper Nuclear Station - Jan. 18, o Dresden 3 - January 12, 201 1 2014 o Palisades - March 24, 2011 o Duane Arnold - February 21, 2014 o Vermont Yankee- March 21, 2012 o Three Mile Island 1 - April 19, 2014 o Surry 1 - May 25, 2012 o ANO 1 - May 20, 2014 o Pilgrim - June 8, 2012 o Browns Ferry 2 - June 28, 2014 o Turkey Point 3 - July 19, 2012 o Peach Bottom 3 - July 2, 2014 o Quad Cities 1 - December 14, 2012 o Oconee 3 - July 19, 2014 o Quad Cities 2 - December 14, 2012 o Calvert Cliffs 1 - Ju ly 31, 2014 o Surry 2 - January 29, 2013 o Hatch 1 - August 6, 2014 o Oconee 1 - February 6, 2013 o FitzPatrick - October 17, 2014 o Point Beach 2 - March 8, 2013 o DC Cook 1 - October 25, 2014 o Turkey Point 4 - April 10, 2013 o Prairie Island 2 - October 29, 2014 o Peach Bottom 2 - August 8, 2013 o Brunswick 2 - December 27, 2014 o Fort Calhoun 1 - August 9, 2013 o Millstone 2 - July 31 , 2015 Technical Issues
Draft License Renewal Interim Staff Guidance: Aging Management of Loss of Coating Integrity for Internal Service Level Ill Coatings (LR-ISG-2013-01)
Issued for public comment on June 29, 2015.
ADAMS Accession No. ML15125A377.
Public comments wer,e received as scheduled.
Addressing public comments - final ISG-LR to be issued early 2016
Steam Generator Divider Plates and Tube-to-Tubesheet Welds
- Steam Generator Task Force submitted a topical report to address necessity for these inspections
PWR Vessel Internals (MRP-227-A)
- Industry to submit draft revision 1 by end of 2015
Cast Austenitic Stainless Steel in PWR Vessel Internals
- Working with industry to identify acceptable approach to evaluate acceptabily of CASS components Subsequent License Renewal The NRC released SECY-14-2016 on January 31, 2014 (http://www.nrc.gov/reading-rmldoc-collections/commission/secys/201412014-0016scy.pdf), to inform the Commission of ongoing staff activities to prepare for the anticipated receipt and review of subsequent license renewal (SLR) applications that, if approved, could extend operation of power reactors beyond 60 years. The Commission's Staff Requirements Memorandum (SRM) on the paper was released on August 29, 2014 (http://www.nrc.gov/reading-rm/doc-collections/commission/srm/2014/2014-0016srm.pdf). The SRM stated that the staff should continue to update license renewal guidance, as needed, to provide additional clarity on the implementation of the license renewal regulatory framework, and the staff should address emerging technical issues and operating experience through alternative vehicles. The SRM further stated that the staff should keep the Commission informed on the progress in resolving the following technical issues related to SLR: reactor pressure vessel neutron embrittlement at high fluence; irradiation assisted stress corrosion cracking of reactor internals and primary system components: concrete and containment degradation, and electrical cable qualification and condition assessment.
NRC staff is continuing to work internally to ensure adequacy of guidance for SLR and continues to meet with the Nuclear Energy Institute (NEI) and the industry on specific technical topics. Upcoming milestones include:
Meeting on November 6 with the Advisory Committee on Reactor Safeguards (ACRS) to discuss status of technical issue resolutions for SLR.
Issuance of draft Generic Aging Lessons Learned Report for SLR and Standard Review Plan for SLR by the end of 2015.
Research Activities Research related to reactor long term operation (LTO) and material degradation issues is ongoing in the NRC's Office of Nuclear Regulatory Research (RES):
NRC RES staff coauthored LTO-relevant presentations at the International Conference on Environmental Degradation in Nuclear Power Systems - Water Reactors, Aug 9-13, 2015. Papers and Presentations are available at : http://envdeq2015.org/final-proceedings/ENVDEG/index.html Titles included
- Crack Grow1h Rate and Fracture Toughness J-R Curve Tests on Irradiated Cast Austenitic Stainless Steels
- Identification of Potential Degradation Phenomena for Spent Fuel Dry Cask Storage Systems
- Evaluation of Strategies for Obtaining High Fluence Materials to Assess Irradiation-Assisted Degradation of PWR Internals
NRC RES staff coauthored an L TO-relevant presentation at the 23ro International Structural Mechanics in Reactor Technology Conference (SMiRT-23), August 9-14, 2015, "NPP Subsequent License Renewal: Lessons Learned from AMP Effectiveness Audits," and published in Transactions, SMiRT-23.
RES/DE staff participated in the Commission Briefing "Overview of the Operating Reactors Business Line" on August 6, 2015, (http://www.nrc.gov/reading-rm/doc-collections/commission/slides/2015/20150806/staff-20150806.pdf ) pointing out that support to SLR is a principal research focus and involves conducting research, developing the license renewal guidance documents and their technical basis, and confirming adequacy of aging management programs
In September, RES awarded a research contract to Pacific Northwest National Laboratory (PNNL) to create an information tool for obtaining information (including ex-plant harvesting) from designated NPPs as they go through decommissioning. This complements ongoing research in developing technical information to support evaluating SLR applications.
RES was-awarded a contract to the Savannah River National Laboratory (SRNL) to provide technical assistance associated with the evaluation of neutron absorbing panels removed from the decommissioned Zion NPP spent fuel pool.
On November 17, 2015, the Advisory Committee on Reactor Safeguards (ACRS) is convening a meeting, open to the public, to discuss key technical issu-es and R&D pertaining to SLR. Presentations will be given by DOE, EPRI, NEI and NRC staff. Topics includes fluence evaluations, and degradation of concrete, electrical cables, reactor pressure vessel (RPV) and reactor vessel internals (RVI ).
NRC's Regulatory Information Conference {RIC) will meet again on March 8-10, 2016.
RES is planning a SLR-related RIC Session entitled "Subsequent License Renewal:
Near-Term Research Supporting the Technical Bases for SLR." This session is designed to be the second in a two-part series where Part 1 will be on the newly-written SLR guidance documents and Part 2 will cover near-term research on safety-significant electrical, mechanical, and structural systems that supports the technical bases for SLR.
RES staff continues to interact with the DOE-LWRS Program and EPRl's LTO initiatives to monitor developments relevant to SLR and, where appropriate, engage in joint research activities.
International Forum on Reactor Aging Management {IFRAM) - Efforts are continuing to develop areas of cooperation and collaboration in sharing data on long term operation
14. New Generic Letters There were no Generic Letters (GLs) issued since the last Code Week.
15. New Information Notices S ince the last Code Week, the following Information Notices (INs) were issued:
i. IN2015-08 (09/02/2015), "Criticality and Chemical Safety Events Involving Unanalyzed Conditions and Unanticipated Unavailability of IROFS at Fuel Cycle Facilities" o Informs addressees of recent operating experience involving deficiencies with evaluations of credible high and intermediate consequence accident sequences in facility integrated safety analyses (ISAs), delineation of items relied on for safety (IROFS) boundaries, and implementation of effective management measures to ensure the availability and reliability of IROFS. These deficiencies resulted in unanalyzed conditions or unavailability of IROFS established to minimize the likelihood of high or intermediate consequence events.
ii. IN2015-09 (09/24/2015), "Mechanical Dynamic Restraint (Snubber) Lubricant Degradation Not Identified Due To Insufficient Service Life Monitoring" o Alerts addressees to potential degradation of the lubricant (grease) in mechanical dynamic restraints (snubbers) not identified due to insufficient service life monitoring (SLM). The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.
16. New Regulatory Issue Summaries There were no Regulatory Issue Summaries (RISs) issued since the last Code Week.
17. NRC Publications of Potential Interest to ASME S ince the last Code Week, the following other publications that may be of interest to ASME w ere issued:
i. NUREG/CR-7204 (September 2015), "Applying Ultrasonic Testing in Lieu of Radiography for Volumetric Examination of Carbon Steel Piping" o This report provides an initial technical evaluation of the capabilities of phased-array ultrasonic testing to supplant traditional radiographic testing for detection and characterization of welding fabrication flaws in carbon steel welds.
ii. Revision 1 of Standard Review Plan (SRP) 17.5, "Quality Assurance Program Description - Design Certification, Early Site Permit and New License Applicants," (NUREG-0800, Chapter 17), was issued on August 2015.
o The new revision included the following changes: 1) the wording was simplified to reflect plain language throughout in accordance with the NRC's Plain Writing Action Plan; 2) the guidance was aligned with the latest revision of Regulatory Guide (RG) 1.28, "Quality Assurance Program Criteria (Design and Construction)," Revision 4; 3) the guidance was aligned with RG 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2; and 4) the SRP was aligned with NQA-1-2008/2009a the latest edition of ~JQA 1 200812009a which Commented [MC&] : This is a bit rricky. 2008 is not the la1est that the staff found acceptable for meeting the requ irement of Append ix B to 10 edition: it is the latest cditio1\ Ihm has been found acceptable. I tried a coll pie of dlffcri!nt w::iys of wording thli., and I think this vcri.lon 1s CFR Part 50. accurate, yes? Maybe it's ckrircr to write. " .. ,with NQA*l -
2008/2009a, the latest edition of1hc quality srnndard thot 1hc s1aff has found acceptable ...**
18. Upcoming Public Meetings of Potential Interest to ASME The following upcoming public meetings may be of interest to ASME:
I. NRC staff review of NEl-14-09, "Guideline for Implementation of 10 CFR Part 21, Reporting of Defects and Nonconformance" o October 116, 2015, 8:00 am - 11 :00 am ( Commented [ MC7] : No longcr**u1>eoming,"
ii. RPV Integrity Issues and Appendix H Rulemaking o January 19, 2016, 8:00 am - 12:00 pm iii. lndustry/NRC NDE Technical Information Exchange Public Meeting o January 19, 2016, 1:00 pm to January 21, 2016, 12:00 pm EST.
Refer to the NRC Public Meeting Web Page at http:1/meetings.nrc.gov/pmns/mtg for a list of a ll currently-scheduled public meetings and further details.

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