ML19345A339

Forwards Revision to Sections a & B of 611023 Rept Re Irradiation of Reactor Vessel

ML19345A339
Person / Time
Site:	Yankee Rowe
Issue date:	01/19/1962
From:	Coe R YANKEE ATOMIC ELECTRIC CO.
To:	Lowenstein R US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8011190134
Download: ML19345A339 (9)

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L YANKEE ATOMIC ELECTRIC COMPANY [

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U. S. Atomic Energy Commission

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,[.[y Attention: R.Lowensteir, Director Division of Licensing and Regulation

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Reference:

Docket 50-29, Facility License DPR-3 Gentlemen:

Attached hereto is a revision to our report of October 23, 1961, covering information'concerning irradiation of the reactor vessel at our licensed facility at Rowe, Massachusetts.

The revisions apply only to Sections A and B of the report and are the result of a recent study which has yielded neutron flux data which we feel more accurately depicts flux values to which the vessel is subjected.

Very truly yours, N

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Rogyr.. Coe, VicoJ; e sident RJC:IH Atts.

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YANKEE ATOMIC ELECTh.

0MPANY Docket No. 50-29 Facility License DPR-3 Ren1v to AEC Ouestionnaire on Reactor Vessel Irradiation Originally Submitted--10/23/61 Sections A & B Revised 1/19/62 A.

The calculated yearly integrated neutron dose, nyt, in each of the following regions of the reactor vessel:

The following doses have been calculated on the basis of a predicted typical year of operation, i.e., a load factor of 80% and a reactor power level of 485 MW thermal.

The values shown are for the innermost point of the particular area and are therefore conservative.

1.

The belt region, i.e.,

the region of the vessel which is exposed to the maximum neutron flux.

Yearly integrated neutron dose, nyt 1

8.93 x 10 n/cm (It 0.8 Mev) 1b 3.44 x 10 n/cm (0.625 ev to 0.8 !!.ev) 2.

The nozzle or penetration region of'the vessel.

Yearly integrated neutron dose, nyt 8.16 x 10 n/cm2 (> G.8 Mev) 15 16 1.68 x 10 n/cm (0.625 ev to 0.8 ?/.ev) 3.

The region of the vessel support brackets Yearly integrated neutron dose, nyt 15 2

6.89 x 10 n/cm G>- 0.9 Mev) 1 2.05 x 10 n/cm (0.625 ev to 0.8 Mev) 4.

Any other region where the integrated neutron dose in conjunction with a high stress level may be of concern when evaluating the long term integrity of the vessel.

There is no other such region.

3.

The calculated neutron dose from beginning of reactor operation to date in all of the regions of the vessel listed in section A.

As in "A." above the values shown are calculated for the Inner-most point of the particular region.

1.

The belt region:

Integrated neutrcn dose to date*, nyt 1

5.58 x 10 n/cm (2- 0.8 Mov) 18 2.15 x 10 n/cm (0.625 ev to 0.8 Mev) 2.

The nozzle or penetration region:

Integrated neutron dose to date*, nyt 15 2

5.10 x 10 n/cm

() - 0.8 Mev) 16 2

1.05 x 10 n/cm (0.625 ev to 0.8 Mev) 3.

The region of the vessel support brackets:

Integrated neutron dose to date*, nyt 15 2

4.31 x 10 n/cm

( - 0.8 Mev) 17 2

1.28 x 10 n/cm (0.625 ev to 0.8 Mev)

Calculations had already been performed for the neutron energy ranges given here. Since this wider range gives a more conservative result than the suggetted range, the calculations have not been repeated.

C.

The results of the stress analysis of the vessel taking into account the following loadings and conditions:

1.

The internal pressure Circumferential Loncitudinal Inside Outside Inside Outside Surface Surface Surface Surface Stress Level due to 14,650 psi 12,650 psi 6,450 psi 6,450 psi Internal pressure 2.

The pipe reactions at such nozzles ~which receive a significant amount of integrated neutron dose, nyt.

Since the integrated fast neutron dose over a twenty year period is below 10 n/cm for all vessel areas except the belt region, stress analysis results are presented for that region only.

10/23/61

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O 3.

Thermal conditions under steady operations at full power.

Circumferential Lonoitudinal Inside Surface Inside Surface Stress Level Thermal (Gamma Rays) 1,130 psi 1,130 psi 4.

Thermal conditions resulting from startup and shutdown.

Thestresslevelsin-thebeltregionforheat-upandcool-downat the maximum allowable rate of 50 F per hour are:

Stress Circumferential Loncitudinal Inside Outside Inside Outside Surface Su'rface Surface Surface s

Thermal.

-4,200 2,420

-4,200 2,420 (heat-up) i i

Thermal 4,6C0

-2,180 4,600

-7,180 (cool-down) 5.

Thermal conditions resulting from abnormal conditions such as emer-gency shutdown (screm) and emergency cooling conditions.

The conditions under consideration all occur at or close to the normal operating temperature which is f ar above the present NDT temp-erature or any anticipated NDT temperature.

Limiting the vessel region under consideration to the belt area it is clear that only inlet water will affect the veosel wall material, and that only transients with a drop in inlet water temperature will increase the thermal gradient ~already present from the gamma ray in-i duced heat gradient. The additional stress fron such transients will be of the following magnitude.

a) Emergency Shutdown (Scram)

The temperature drop in the inlet water to the vessel is approximately 27 F for a reactor scram followed by turbine tripout. The acccmpanying transient thermal stress is a maximum of 4,000 ps.i.

b) Loss of Flow A cold leg temperature drop of approximately 30 F occurs for the most severe part of the transient resulting from loss of flow. The corresponding transient thermal stress, there-fore, is approximately 4,000 psi.

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Significant dats and informstion about the material of censtruction of the vessel, such as:

1.

Its Chemical Analysis (See Attached Tabulation.)

2.

Its Fnysical Test Data Including Its Initial NDT (See Attached Tabulation.)

3.

f/.ethod of cladding the stainless steel overlay to the base metal The vessel and head flanges, main coolant nozzles, and the closure areas are clad with Type 30EL stainless steel by manual and automatic weld depositing.

All plate material for vessel shell, lower head, and vessel head dome is clad with a 0.109 inch thick sheet of SA-240, Gr.S, stainless steel which is applied to the base material by the Babceck and Wilcox spot weld process.

4 Sensitivity of radiography and other non-destructive inspection tests used.

Sensitivity of radiography was as required by the ASME Code,Section VIII, " Rules for Construction of Unfired Pressure Vessels."

• 5.

Heat Treatment of the Vessel.

The vesse; material was quenched and tempered to obtain the desired impact properties. During the fabrication period the material was stress relieved at certain intervals.

"o detrimental influence tn impact strength is caused by stress relieving, since.

it was performed at a temperature lower than the tempering tempera-ture.

E.

Informaticr ccncerning surveillance and inspection of the vessel which should include:

1.

Provisions and Procedures for Inspecting and Testing the Vessel During its Operational Lifetime.

Such procedures are being evaluated but, in view of the extra-ordinary difficulty of any visual inspection, it is considered more practical and informative to evaluate the accumulated neutron dose at the vessel wall in conjunction with 111 available experimental informatien regarding the corresponding enanges in properties to be anticipated, This wi.1 Le done periodically and pressure-temperature restricticns modified. as necessary, tc crovide a conservative margin in relatier to the NDr.

Desigr and radiographic testing of the :n;uc ard octiet norzles comp;;e 5.;ith AS!/E Code Ca se.23:

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-S-Proposed Provisions and Procedures for Monitoring the Vessel l

2.

During its, Operational Lifetime.

At the first refueling, early in 1062, we propose to install a number of irradiation specimens made of meterial used in the fabricat?.on of the vessel'shell. Provisions have been made for four containers each holding 22 Charpy V-notch specimens and 2 tensile specimens adjacent to the. inner surface of the vessel wall and for eight containers each holding 10 Charpy V-notch specimens and 2 tensile specimens in a relatively high flux region Neutron detecting foils are being just outside the core baffle.

considered for'use in evaluation of the integrated. dose and energy It i, anticipated that the specimen spectrum at each location.

s containers will be withdrawn sequentially at future refu lings.

l Data. developed by such a program can yield information directly applicable to the evaluation of fast neutron effects on this.speci-l fic material under actual operating conditions of temcerature, neutron flux and neutron energy.

Limitation which you have imposed or intend to impose on the 3.

allowable pressure during conditions of transient temperatures which may result in reactor vessel temperatures below the FTE (Fracture Transition Elastic) of the material.

To date the vessel exposure has been insufficient to resu's in a significant. shift of NDT. No restriction other than those 4

imposed by the vessel supplier on the unirradiated vessel have l-effect for purposes of accommodating any such change been put in However, in order to eliminate leakage at the flan >Je of in NDT.

the main coolant pumps, system pressures are further restr? :ted in terms of temperature during plant heat up and cool dova.

g The restriction imposed by the vessel supplier states that main coolant pressure must at all times be maintained below 500 3

0 F.

As mentioned psi unless the vessel temperar:re is above 90 r

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above, an additional restricticn is in ef fect during system heat up or cool down. The pressure is not-_ allowed to exceed 500 psi-unless the temperature is above 300 F.

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YANKEE ATOMIC ELECTRIC CCMPANY REACTCR VESSEL PFO'SICAL PROPERTIES-Tensile Yield Minimum Minimum Re-Section Ma.terial Strength NDT Strength Elongation duction Area 1000's psi F

1000's psi V - 2"

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Vessel Shell SA-302B 85.2--95.0

<;+10 59.1 28 Modified Bottom Head SA-302B 96.4--98.7

<!+10 63.3 24 Modified Dome SA-302B 92.9--99.5

< '. 0 78.5 22 Modified Upper Ring Forging SA-105 II 80--83.5

== +20 56--57.5 25.5--26.5 67.9--60.0 Modified Head Forging SA-105 II 84--84.7

= +30 57.2--58.5 25.5--26.0 66.5--67 Modified Vessel Support SA-2123 75.9--76.4

< +10 45--48 29.7 Closure Studs SA-193 130--135.5 107.5--113.5 18.5--19 59.9--62.4 B-16 Control Rod SA-106C 78.6

< +10 50.9 29 Housing (Outer)

Nozzles SA-182 110

< +10 96.5 25 Modified to SA-3023

• 8" Gage Length

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7 YANKEE ATC#,IC ELECTRIC COMPANY REACTOR VESSEL CHEMICAL PROPERTIES Section Phosphorous Sulphur Silicon Molybdenum Manganese Carben Chromium Nickel Vessel Shell

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