ML20217B266

Responds to Which Provided Views & Comments on Core Shroud Cracking Problem at NMP1 W/Regards to NRC GL 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Bwrs,

ML20217B266
Person / Time
Site:	Nine Mile Point
Issue date:	09/08/1997
From:	Hood D NRC (Affiliation Not Assigned)
To:	Tom Gurdziel AFFILIATION NOT ASSIGNED
References
GL-94-03, GL-94-3, NUDOCS 9709230293
Download: ML20217B266 (7)

Text

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2006H001

• • • • • September 8, 1997 Mr. Thomas Gurdziel 9 Twin Orchard Drive Oswego, NY 13125

Dear Mr. Gurdziel:

' By letter of May 22,1997, you provided your views and comments on the core shroud cracking problem at Nine Mile Point, Unit 1 (NMP1), with regards to NRC Generic Letter (GL) 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Bolling Water Reactors" dated July 25,1994; the NRC public meetings on April 14,1997, in Fulton, New York; and the NRC cover letter and Safety Evaluation of May 8,1997. Alth ugh you state that no response is needed, I would like to respond to the following technical points raised in your letter:

GL 94-03

'The accident scenarios ofpn' mary concem are the main steam line break, recirculation line break and seismic events"[GL 94-03). The logic here is "OR."

However, at the public meeting in Fulton, I believe that I heard the required accident dascribed as a pipe break AND an earthquake....

4 You are correct that GL 94-03 states that the scenarios of primary concem are the main steam line oreak, recirculation line break, and seismic events. Each event by itself is considered to be a design basis event, i.e., an event for which the unit is designed to withstand and still be able to achieve a safe shutdown. GL 94-03 was concemed that potential cracking of the horizontal shroud welds could affect the ability to maintain adequate core coo ling and the ability to shut down the reactor following such design basis events. Based on this concem, the NRC staff requetted in GL 94-03 that licensees perform a plant specific assessment accounting for uncertainties in the amount of cracking, including but not limited to the following:

(1) An assessment of the shroud response to the struduralloadings resulting from design basis events (e.g., steam line break, recirculation line break). si asymmetric loads can affect the shroud response, these should also be considered.

(2) An assessment of the ability of plant safety features to perform their function considering the shroud response to structural loadings (e g., control rod insertion, emergency core cool;ng system injection).

Notwithstanding the results of the NMP1 assessment, tie rod assemblies were installed in 1995 to maintain the structuralintegrity of the shroud wits postulated 360* throughwall failure of horizontal welds H1 through H7, and were designed to withstand combined loads resulting from ,Djl' g (1) a steam line break and design basis earthquake, or (2) a recirculation line break and design basis earthquake. Thus, the design crneria for the tie-rod assemblies are different from the design basis of NMP1; these combinations of accidents that are utilized in the design of the ks C 9709230293 970900 y DR ADOCK 0500 0 hm*3 hdn Qg j,"m hh

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L f-i tie-rod assemblies are more conselvative than the design basis of NMP1. The statements made at the public meeting in Fulton, NY, regarding the required accidents described as a pipe break and an earthquake were referring to the design criteria of the tie-rod assembly and not the plant-specific analyses requested in GL 94-03.

Does a letter which addresses IGSCC, something that requires high neutron Rux, apply to cracking cases where high Rux areas are not ths ones primarily effected? I don't think so. To me, this letterprovides a shiR in logic.... At the Fulton meeting, I believe that I heard somebody say the cracking was not t

caused by operation....

The mode of cracking in the BWR core shroud is determined to be intergranular stress corrosion cracking (IGSCC). In areas of high neutron fluence such as the beltline region, the irradiation assisted IGSCC, commonly known as irradiation assisted stress-corrosion cracking (IASCC), may occur. For IGSCC or IASCC to occur, adequate tensile stresses, an oxidized environment, and a sensitized material must exist. Neutron fluence in excess of a certain threshold level has the potential to sensitize the stainless steel material and render it

- susceptible to IGSCC. Stress analyses show that the tensile stresses acting upon the core l- shroud during normal plant operation are insufficient to cause IGSCC. The dominant driving L forces for IGSCC have been dstormined to be tensile residual stresses resulting from the j process of welding, fit up, or cold work performed in the fabrication of the core shroud.

May 8.1997. Cover Letter -

' Is 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> of hot operation cumulative, without regard to powerlevel or number and types of reactor shutdowns?

The approved 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> of hot operation are cumulative. Hot operation applies whenever the reactor coolant temperature is above 200'F. Thus, these hours of hot operation

- accumulate without regard to power level and do not accumulate when Unit 1 is in a cold shutdown or refueling condition with reactor coolant temperature no more than 200*F.- The bounding crack growth rate (5 X 108 inch / hour) used in the licensee's calculations of the allowable operating hours was derived from test data that cover the full range of power levels.-

Fatigue crack growth due to plant operating transients such as plant shutdown and start up has been shown to be insignificant.

Mav 8.1997. Safetv Evaluation

- How can "The location of the cracking in the NMP1 shroudis consistent with IGSCC"be true if the worst cracking is not in the area of highest neutron Rux?

Isn't the worst cracking on the outside of the shroud?

Although the neutron fiuence is higher at the shroud's inside surface than at the outside surface due to closer proximity to the fuel, the worst cracking may not be in the area cf higher neutron fluence. This is because factors other than neutron fluence (e.g., cold work tensile s _

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stresses) also determine the location and extent of cracking. The worst cracking will occur in

. areas with the highest combined effects in promoting lGSCC or IASCC.

I dont understand how

• cracking in the cold-worked areas will also not grow very.

. deep

• unless the statement should say " cracking due to IGSCC in the cold-i norked.... "

i

! Cracking depth in the cold-worked areas will be limited because the depth of the material

affected by cold work, such as grinding, is shallow. The crack will stop advancing once the

cold-work effect disappears. 4 l

l "Although irradiation increases the susceptibility of the material, it also relaxes the weld residual stresses, the driving force for crack growth.' is this statement pertinent to welds in the shroud? Dont actual shroud welds get any effective post weld heat trcatment?

The statement regarding the irradiation effects on crack growth is pertinent to the welds in the core shroud. The core shroud welds did not receive, and were not required to receive, any post-weld heat treatment.

" Typically, the allowable crack sizes are large and approach or exceed the length of the weld itself." IfI am reading this right, don't you feel a little uneasy when you are told you dont need any of the weld, in fact, you could use even less than none? I cant help but think of a flagpole. Ifit were constructed the same as the

. shroud, I wouldnt want to walk nearit if even horizontal weld was failed, all the vertical welds were failed, and the support consisted of 4 tie rods to the top.

. The core shroud was constructed by welding a number of short cylinders stacked on top of each other. The vertical welds in each of the short cylinders were much shorter than the height

. of the core shroud.- The statement regarding the allowable crack sizes on the vertical welds

- that could approach or exceed the length of the vertical weld itself was derived from the results of flaw evaluations assuming the core shroud circumferential welds were not completely failed and its structural integrity was maintained, if the circumferential welds were assumed to be completely cracked through-wall as is postulated in the tie rod design, then only about a few inches of the total vertical welds length would need to remain sound for the structural integrity of a particular short cylinder to be maintained. The tie rod assembly design is bas 3d on the postulation of " replacing" the circumferential welds that hold the cylinders together.

4 i

The Ac il 14.1997. Public Meeting I

The drawing of the finite element mesh held up at the meeting appeared to be reversed. I expect...to see the finer mesh in the area of major concem...this was the bottom of the shroud, where it is supporfed from the reactor vessel. I remember the smaller finite elements being at the top of the drawing shown at the meeting. \

The licensee's finite element mesh size is finer in the areas of major concem-such as areas of the crack tips and the cracked core shroud welds, it is normal practice to use a finer mesh size in a location where a more precise definition of the stress is desired. The finite element mesh size is not reversed. Significant cracks were not found at the bottom of the core shroud and, therefore, a finer mesh size is not required in those areas.

General Comments it is my opinion that thermal effects should have been considered as a possible cause of cracking on the outer surface of the shroud.... Somewhere below the bottom of the skirt...is where I would expect to see the approximate top of

\ thermalinduced verfical cracking...and I would expect it to continue downward somewhat until the outor surface ,*emperature of the shroud is close to the inner temperature of the shroud. One fector that might provide thermal cycling...is change in feedwater Row.... ' It is also my opinion that stagnant areas,' If formed in the annulus, couldprovide local areas of temperature differences (not temperature cycling), that may be sufficient to promote shroud (outer) surface cracking. One way to cause this might be to valve out one (or more) of the 5

. reactor recirculation pumps. Unequal feedwater Row to.the feedwater sparpers, orparfly clogged or non-uniformly fabricated feedwater spargers might also cause some local temperature differences.

The thermal effects due to the injection of feedwater may have some small effect on the cracking on the outer suiface of the shroud; however, it is difficult to quantify. The pre-heated feedwater, after injecting into the reactor annulus from the feedwater spargers, will mix __

- turbulently with the higher temperature water coming down from the steam separators. By the time the mixed water impinges on the core shroud, its temperature would become quite uniform and not that much different from that of the core shroud's outside surface. Based on this simplified feedwater flow path, the resulting thermal effects on the core shroud are not expected

- to be significant.' As discussed earlier, the main driving forces for the observed IGSCC on the

. core shroud is the tensile residual stresses resulting from the process of welding, fit up, and

) cold work performed during the fabrication of the core shroud. The magnitude of these stresses is on the order of the yield stress of the core shroud material.

5-I trust you will find the above response to your letter helpful in better understanding the NRC staff remarks at the public meeting and in subsequent documents regarding the NMP1 core shroud. Thank you for sharing your views and comments with the NRC.

Sincerely, Y Yir Darl S. Hood, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

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o September 8, 1997 I trust you will find the above response to your letter helpful in better understanding the NRC staff remarks at the public meeting and in subsequent documents regarding the NMP1 core shroud. Thank you for sharing your views and comments with the NRC.

Sincerely, ORIGINAL SIGNED BY:

Darl S. Hood, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation DISTRIBUTION See next page l

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