ML20214G727

Forwards Request for Addl Info,Per Matls Engineering Branch Review of Psar.W/O Encl

ML20214G727
Person / Time
Site:	Columbia
Issue date:	11/15/1971
From:	Case E US ATOMIC ENERGY COMMISSION (AEC)
To:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
References
CON-WNP-0157, CON-WNP-157 NUDOCS 8605220443
Download: ML20214G727 (13)

Text

__.

• e-g W-'e W

~

gr i-M g'9 M

s-* y

- ye p * - ^

w..e'

=

4 * * ' ^

'

• b.; ;Y:.b.

, ~,

,_g.

s

.~4--

sg

?

a w 0 NAN.*~

, >+

m..' l -

g y

.-g4, f

,,, my d em ee

.y y **

M

• I bI s a8 e

wmm O..

-... m m.m..

e m..

2m tmuuuwdam ps gpgwungmHMrtgiigg~

3)W g e.M.e s ti,gf4 _-

C; n.

y g w meetsmeedma.

4 n,

Af(TW 76 3f7

_t. c. cans.

. - t.r Bestsien of Beester Standsede heleeures Materials Engineesdag a.gamet ser 44atas 1 va-for Bamford 2 C. P.

es u/ anal:

S. Emmener. DE R. Boyd, BEL L BeYoung, DEL B. Showholt, BEL

3. Niemm8Fe BW S. Pawliaki, DRs N. Fairtila BBS L. Peree, BBS Massi+ y -

. : n Mp e._

g p r,9,.t-8. IIAmes, E E ct

.~

ktk h$ Sif$?%:J 9.% g 5"{& g.giH&r- @ Q Q && *

- ( O p f 1 f.s 4 9 4 @ U $ $ / 6 $ 4 C ff, O T i g $ C ) tY g fp 1 O O O 3

Ci$ 9 'O W.9 %C*M4N31DOTf2211fR.W.4KO 0 h 3aa l

2;W prm';G GM.'eb gliWQ o c(WW m e e e % #e s 3-i Pg?f397C1FtHW'SYTflyMp)amwrmeag%fy.

. O o c:a g((3 Mm (qPJ C Q CWl S@y%tx2 Oc anPJ&O O "1 @ c o Sm:

i >M/ e f4IEW@ O.Wd S I

$ @ MUSWM b Wfs V?s%9'3R'GWI4199M'Fff s:- QG GO SM4/

!s g5 N +ja(,3 Gr C $ g g g e?

• q Q q;-j g-gr

,, m. m.

,. w.n..,, #..

._.. ~

m,_

xy, '-

r m k}ff,,

DRSIMTEB DRS MTJB DRS M W DRS'efD/E D

k/

N1,A m,

MA"7~

gy,.

.l/vv,

], V M. Fair,til,e.Lla..L.. Fors=

JssAry my_-i,. _ _ _

S. F eki sumeaur >

ll/l8/71 11//[/71 ll/g/71.

' 11/,ft/_71 11/f k m,

perm ABC-3Ia (Rev.9-53) AECM 0240 e u s oovta*#AENT Neve CNTCS. 89MW PSS 8605220443 711115 i

PDR ADOCK 05000397 A

PDR

i s

'1 s

e. b - "

'{ { p,

e

, ~. >

--,.o-c

-. Y s...v.

. L e,v.\\

't '

s., r

~

v

. v 3

Q

- c,a:

..a

' M" (l;

'.e'

' MNHN

._d

_~

7' /

11EN ! Wit.-OK _ s _ __

N.?,

HANAUER RICE _. '.

c.O..

ms......

t.

..v e".

..'.J'....'~.J O C T. ;.'.. naf r fcr 2.dactOt*S' r_.

m

_ i m

e....._

._.._.s a... _.

Enc cz.a '.c a su; cs: ; cigast.;c. cn rccet.; act vitiee 1

re.e:ca to.". cod critaria far 'he :.anic ec s..e.

P Milton Shat.', Director Division of Reaccor Develoy cat and Technology

Enclosure:

i.S bJatdd

. 2

.unn bec:

.. h,. 3.untzing, 3 r.,

u..

w, mw

2. C. Case, Dir., DRS CF '2)

ACXR i

1 9

D l) m i Q*

b

• {<(3; L-lt,

.I-

\\

l G

J i'v J' y -

,u w

MN 85 w'-4..

.e s :

• 4*

I N

l L

o 6

'r

)

.- n....

.... s a w L 3 3...>...v L.,.....

.va

..i

..s.. r

.s

,/ d, v.

c..., ;...

- v..-,.**

1 o

e..,.,

s..-

w.

s

....ww.....v..e

...c

..........a

3

.v n.:.. <.. r e f r J N.. 3 J. ; o.'7 m' l -.' C 2. ; '- - >.--'

d ;-' "

N

,.406.

s.,e

...a.4.-.we

.vV i 6

. u -

u.

,.., 5 L.. d... i... 1 7... >......w

..... o,

. a Cl,,,.,.

i,

.v..

.-.1-........ c.r

,u g

.... C....

sv..S.w I*..

... i..

.,.3e.

w....

e.,.,

...e Jf..,

..on p.,)3..,.,

P v,, h.c b'. -,...,..

c

.w u,..,.;

.f

.G.

, r.u s...

m o pt,. _. o. -. ;.,,,.

... c., e -....

.3

e.

t..

..s.

eu... t... a.,.. -..

.a v: u.: Rega'.ctory an.! R>T ir. scvvic.>ing -

,uit.... a ac;ica h-:-

-........ sea,

<..e..s..

....... ew... -.

..c..-.u, r-..., a....

~..

.m.at a..

flou., uill ba sssesced ;o shm; wa. suca f:.. ru 4:1 :.at recuit in ancassive f.aca levels et lian.'cn i.?.ca t'ais s;a y is coupletac, the USCS will be raqucsted to sum t at atu ti.at earth:'t.aace

ec

.n ur.e stuJy represent reasonable upper limit.s for such vcluations. If this cannot be dctermined, c,. terr.. e rot.o'..-L.) 20t ca; w;,.A

..e cor.sid-ered (such as further tvalcations of the available geologi~c data).

t l

b e

. -$74 cY I

eq

.o 1

IIANFORD NUCLEAR STATION - UNIT 2 DOCKET No. 50-397 REQUEST FOR ADDITIONAL INFORMATION - MATERIALS ENGINEERING, DRS REACTOR COOLANT SYSTE!!

Fracture Touchness To evaluate the adequacy of the proposed heatup and cooldown limits for this plant, provide the following information:

1.

For all pressure-rctaining ferritic components of the reactor coolant pressure boundary whose lowest pressurization terperaturc* will be below 250 F, provide the raterial toughness properties (Charpy V-notch impact test curves and dropweight test NDT tenperature, or others) that have been reported or specified for plates, forgings, piping, and wcld material. Specifically, for cach component provide the following data for materials (plates, pipes, forgings, castings, welds) used in the construction of the component, or your estimates based on the available data:

(a) The highest of the NDT temperatures obtained fron DUT tests, (b) The highest of the temperatures corresponding to the 50 f t-lb value of the C fracture energy, and

• Lowest pressurization tenperature of a conponent is the lowest temper-ature at which the pressure within the conponent exceeds 25 percent of the system nornal operating pressure, or at which the rate of temperature change in the component naterial execeds 50*F/hr., under normal operation, system hydrostatic tests, or transient conditions.

m 4 )

3 (c) The lowest of the upper shelf C energy values for the " weak" y

direction (VR direction in plates) of the naterial.

2.

Identify the location and the type of the material (plate, forging, weld, etc.) in each component for which the data listed above were obtained. Where these fracture toughness parameters occur in nore than one pla te, forging or weld, provide the infornation requested in 1. (a), (b) and (c) for each of them.

3.

For reactor vessel beltline caterials, including velds, provide the information requested in 1. and 2. and in addition specify:

(a) The highest predicted end-of-life transition terperature corres-ponding to the 50 f t-lb value of the Charpy V-notch fracture energy for the " weak direction" of the natcrial (UR direction in plates) and (b) The mininun upper shelf energy value which will be acceptable for continued reactor operation toward the end-of-service life of the vessel.

4 Furnish the proposed heatup and cooldown curves which will be used to control the pressure and tenperatures to which the ferritic natcrial of the reactor coolant pressure boundary will be exposed during operation of the plant until the scheduled renoval of the first material capsule.

3-IIANFORD 2 REACTOR COOLANT SYSTDI Reactor Vessel Material Surveillance Program 1

Describe the reactor vessel material surveillance program to indicate the degree of compliance with the AEC proposed, " Reactor Vessel Material Surveillance Program Requirements" 50.55a, Appendix'H, published in the Federal Renister on July 3, 1971. State also the degree of confornance with AST:t E-1BS-70, especially with regard to the requirenents on retention of representative test stock (archive material) and documenta-tion of chemical composition.

2.

State the number of Charpy V-notch specimens oriented with respect to the weak direction (WR orientation in plates) of plates, forgings and weld l

materials that will be included in the reactor vessel material surveil-lance program.

a 4

4

. IIANTORD 2 REACTOR COOLANT SYSTEM Sensitized Stainless Steel 1.

Describe the plans which will be followed to avoid partial or local severe sensitization of austenitic stainless steel during heat treat-ments and welding operations for core structural load bearing eenbers and component parts of the reactor coolant pressure boundary. Describe welding methods, heat input, and quality controls that will be ceployed.

2 If nitrogen will be added to stainless steel types 304 or 316 to enhance its strength (as permitted by AS:!E Code Case 1423 and Case 71 USAS),

provide justification that such material will not be susceptibic to stress corrosion cracking under severely sensitized conditions.

3.

Describe steps that will be taken to avoid gas entrapnent at high points or nonflowing parts of the reactor coolant systen.

}

a 9

+-

~ ~, - -. -

. HANFORD 2 REACT 0!! COOLANT SYSTCt E1cetroslaa I!ciding 1.

If the process of electroslag welding was used in the fabrication of conponents within the reactor coolant boundary, describe the process variables and the quality control procedures applied to achieve the desired material properties in the base teetals, heat af fccted zones, and welds.

s I

a

-6

!!N EORD 2

)

REACTOR COOLANT SYSTE!!

Leatame Detection System Provide the following information regarding the reactor coolant leakage detection system.

1.

Describe the methods used to provide positive indications in the control room of Icakage of coolant from the reactor coolant system -

to the containment.

2.

Discuss the adequacy of the Icak detection subsystem which depends on reactor coolant activity for detection of channes in Icakage during the initial period of plant operation when the coolant activity may be low.

3.

With reference to the proposed maximum allowable leakage rate from unidentified sources in the reactor coolant pressure boundary, furnish the following information:

a.

The length of a through-wall crack that would Icak at the rate of the proposed limit, as a function of wall thickness, b.

The ratio of that length to the length of a critical through wall crack, based on the application of the principles of fracture mechanics.

c.

The mathematical model and data used in such analyses.

4 Provide the sensitivity (in gpm) and the response time of each leak detection system. For the containment air activity monitors, pro-vide the sensitivity and the response time as a function of the percentage of failed fuel rods or of the corrosion product activity a

. )

in the reactor coolant, as applicable.

5.

Estinate the anticipated normal total Icakage rates and major leakage sources on the basis of operational experience from other plants of similar design.

6.

Provide the time interval in which reactor will be shut down if cither the total or unidentified Icakage rate limit is exceeded.

7.

Describe the proposed tests to demonstrate sensitivities and opera-bility of the leakage detection systems.

O a

. IIANFORD 2

)

REACTOR C00!K:T SYSTDI Inservice Inspection Procram 1.

Describe the design and arrangement provisions for access to the reactor coolant pressure boundary as required by Sections IS-141 and 1S-142 of Section XI of the AS!!E Boiler and Pressure Vessel Code - Inservice Inspection of Nuclear Reactor Coolant System.

Indicate the specific provisions nade for access to the reactor vessel for examination of the vcids and other ce,ponents.

2.

Section XI of the ASME Boiler and Pressure Vessel Code recognizes the prob 1 cms of examining radioactive areas where access by per-sonnel will be impractical, and provisions are incorporated in the rules for the exa,ination of such areas by renote means. In some cases the equipment to be used to perform such examination is under developnent. Provide the following infornation with respect to your inspection progrant a.

Describe the equipment that will be used, or is under develop-ment for use, in perforcing the reactor vessel and nozzle in-service inspections.

b.

Describe the system to be used to record and conpare the data from the baseline inspection with the data that will be obtained from subsequent inservice inspections.

c.

Describe the procedures to be followed to coordinate the develop-ment of the renote inservice inspection equipment with the access provisions for inservice inspection afforded by the plant design.

9-I IIANFORD 2 CONTAI:0!ENT Lenkace Testinn Pronran 1.

Describe the containment Icakage testing program and indicate the degree of compliance with the AEC proposed " Reactor Containment Leakage Testing for Water Cooled Power Reactors", 550.54(o), Ap-pendix J, published in the Federal Rerister on August 27, 1971.

e 1

1 i

1

=

' i HANFORD 2 ENGINEERED SAFETY FEATURES Inservice Inspection Program 1.

Describe the inservice inspection program for fluid systems other than those comprising the reactor coolant pressure boundary, including items to be inspected, accessibility requirements, and the frequency and types of inspection. The fluid systems to be considered are applicable engineered safety systems, reactor shutdown systens, cooling water systems, and the radioactive vaste trentrent systens.

6p a

D O

g#.

-+