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| document type = SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES, TEXT-SAFETY REPORT
| document type = SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES, TEXT-SAFETY REPORT
| page count = 3
| page count = 3
| project = TAC:59976
| stage = Approval
}}
}}


=Text=
=Text=
{{#Wiki_filter:}}
{{#Wiki_filter:.., . , - .                  -    . . -    .          n . .      . . . .          .          .          .e = r1
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                                                            . UNITED STATES NUCLEAR REGULATORY COMMISSION L                  j                    . WASHING TON. D. C. 20S66
                '$, ...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST' PRESSURIZE 0 THERMAL SHOCK EVENTS (10 CFR 50.61)
SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 By {{letter dated|date=January 23, 1986|text=letter dated January 23, 1986}}, the Sacramento Municipal Utility District, the licensee for the Rancho Seco plant, submitted information on the material properties and the fast neutron fluence (E > 1.0 Mev) of the reactor pressure vessel in compliance with the requirements of 10 CFR 50.61 (References 1, 2 and 3). Our evaluations of the pressure vessel material properties, and fast neutron fluence for fracture toughness requirements for protection against pressurized thermal shock events (10 CFR 50.61) are as follows:                    (
Material Properties                                                        ,
The controlling tsettline material from the standpoint of PTS susceptibility was identified to be a longitudinal weld in the lower shell course, weld WF70 (Weld Wire Heat No. 72105). This submittal states that there is a small probability that the controlling weld con,tains " atypical weld metal,",
an off-specification material that B&W found in 1978 in a test piece welded with wire heat number 72105. This was the subject of Topical Report BAW 10144-A, which was reviewed by the staff on December 12, 1979. Based on a limited amount of test data developed by B&W, the submittal states that the screenino criterion will not be reached before end of life even if the atypical weld metal is present. The staff does not disagree with the B&W evaluation of this material, although their evaluation does not follow the procedure for calculating RT                given in the PTS rule. (The wasdevelopedwithoukThonsiderationoftheatypicalweld formulaThe metal.)    for RT(Nf 5      believes that the probability of occurrence of atypical weld metal in vessel welds is low enough and its properties are such that                  ,
the results of a plant-specific probabilistic risk analysis would not be af fected significantly if the atypical weld metal was considered to be present.      Thus, the materials input to the calcJlation of Rip will be evaluated without further consideration of the atypical weld Ntal.
The material properties of the controlling material and the associated margin and chemistry factor were repnrted to be:
1 D703300173 070313 P        ADOCK 05000310
,                      I,DH PDR
(              .                                                    ,
 
e            .
2                                      ,
11: i i i: y Sn tun i i t.i l    St.i f f E v.!lu.it inn O . I'.                      0 . 15        .
Cu (cop;ier a:oistent          )          '
U . 5'i                      0.59 Ni (rtici:cl content , %)
O                              U l  (insti.il Hi g, *f)                                        .
s                                i'l
                                                                                                                                  ')9 H (l;.t ry i n , **f }.
                                                                  *                                    -                    22G.8 CT (Chemistry T.:etnr. *f)
The justifications Ihc conten!Iing mJLeelaI has been proper 1y idcoLified.                                                        Jrc .?Cceptabli' given for the Coppcr and niCI:Cl Corttent', ,!nd the isti tia l RT,, r thcsc values.
The margin hJs been dertved f ront can',ideratina of the bases',
followin9 the PTS Rule. Section 50.61 or 10 CFR Part 50.
fas t Neutron fluence Detailed calculations havq been performed with the assistance of our contracter.
* Brookhaven National l'4bdratory (8tIL) .for the . cycle depender't orcssure v'es-sel inner-wall (E > h0 HeV) flu'xes for the Rancho Seco pl' ant. Th'e calculated                                                                    -
fluxes were used to' determine the vessel (inner diameter) accumulated fluence.
which was then used to predict the RTPTS for specific welds in the reactor vessel. The analysis was based on a 00T calculation for the Rancho Seco plant.
An 80% load f actor. ' low leakage loading and plant specific data were assumed for the extrapolation. The following table summarizes the comparison of the Joplicant and the BNL results.                        The BNL estimates are in substantial agreement values with the applicant values, therefore, they are acceptabic. The RT pyg At the expiration of the current license are below the applicable criteria specified in 10 CFR 50.61 and are acceptabic.
RAtlCl10 SECO RT p73  CALCULATE 0 FOR OCTOBER 11. 2008 (LICEll5E EXPIRATI0ll)
Upts(O($'r)
NvTx10.is
( .1 )
(a)                                              lint    1 0i::
i              H        SNUD          ONL Screen (*F) SMUD Ucid              llca t#            CU  HI 10,        48.          S.4            5. 5      210              140.          IN.9~ 0 . 0'l~
Hl OtT            FV482J            .15  .68                                                              270            110.          111.5    0.13 U SHELL            C50021            .12  .60            4        40.          7.1            7. 2 103.5    0.43 7.2        270            104.
U $11 ELL          C50G22            .12  .60        -10.          40.          7.1                                                        00.43    0.71 270                01.
L SHELL            C50701            .10  .58        -20,          48.          7.1            7. 2
: 81.        00.43    0.71
                                                              -20,          40.          7.1            7. 2      270 L SHELL            C50702            .10  .58                                                              300            225.          225.2    0.09        .
U CIRC                              .29  .68            0.        59,          5.4            5.5                                                              '
WF-233                                                                                  300            241.          241.7    0.29 H CIRC            W-154.            .31  .59            0,        59,          7.1            7. 2
                                                                                                        .5. 5      300            225,          225.2    0.09
: l. CIRC                              .29  6    .        O.        59.          5.4                                        193,.          193.7  . 0.'35 G21 *
                                                          '    -                                    . 6. 9                    .
fULN3 L*%.W923
      '; ;, u ,', %''(: '.r<.\
I
* 0D* *09. J 56.8..?'**''' % ., .,.. , .270-
                                                      ..~                        ,
 
:                - - - - - - - - - - - - - - - - - = - - - -
e                                                                    .
t O                =
L LN        WF-29    .23    .63      0. 59. 6.9    7.0 270    ^
194.                194.1                  -0.05 L LN ID      WF-70    .35    .59      0. 59    6.9    7.0 270          265.                264.9                  -0.04 L LN 00      WF-29    .23    .63      0. 59. 6.9    7.0 270          194.                194.1                  -0.05 a) Ref. I                                0                                            .
b) RT PTS =I+M[-10+470Cu+350CuNi]f.270 In view of; (a) The Pressure-Temperature updating requirements for the fracture toughness of the beltline material in 10 CFR 50 Appendix G. and (b) the fact that the RT PTS value is readily available f.om the calculation of the Pressure Temperature limits. and (c) the staf f desire to be informed on the current value of the RTPTS                    for all PWRs,                                            .
we request that the licensee submit a reev,aluation of the RT PTS and a compari-                                ,
son to the prediction of Reference 1 along with the Pressure-Temperature operat-ing limits which are required by 10 CFR 50 Appendix G." It should be noted that this reevaluation is a requirement by 10 CFR 50.61, whenever core loadings, surveillance measurements, or other information indicate a significant change in projected values.
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Latest revision as of 02:09, 7 December 2021

Safety Evaluation Re Fracture Toughness Requirements for Protection Against PTS Events (10CFR50.61).Util 860123 Submittal Re Matl Properties & Fast Neutron Fluence of Reactor Vessel Acceptable
ML20205C395
Person / Time
Site: Rancho Seco
Issue date: 03/13/1987
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20205C307 List:
References
TAC-59976, NUDOCS 8703300173
Download: ML20205C395 (3)
v • d • e


Text

.., . , - . - . . - . n . . . . . . . . .e = r1

[(fU%f[*,

g

. UNITED STATES NUCLEAR REGULATORY COMMISSION L j . WASHING TON. D. C. 20S66

'$, ...../

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST' PRESSURIZE 0 THERMAL SHOCK EVENTS (10 CFR 50.61)

SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 By letter dated January 23, 1986, the Sacramento Municipal Utility District, the licensee for the Rancho Seco plant, submitted information on the material properties and the fast neutron fluence (E > 1.0 Mev) of the reactor pressure vessel in compliance with the requirements of 10 CFR 50.61 (References 1, 2 and 3). Our evaluations of the pressure vessel material properties, and fast neutron fluence for fracture toughness requirements for protection against pressurized thermal shock events (10 CFR 50.61) are as follows: (

Material Properties ,

The controlling tsettline material from the standpoint of PTS susceptibility was identified to be a longitudinal weld in the lower shell course, weld WF70 (Weld Wire Heat No. 72105). This submittal states that there is a small probability that the controlling weld con,tains " atypical weld metal,",

an off-specification material that B&W found in 1978 in a test piece welded with wire heat number 72105. This was the subject of Topical Report BAW 10144-A, which was reviewed by the staff on December 12, 1979. Based on a limited amount of test data developed by B&W, the submittal states that the screenino criterion will not be reached before end of life even if the atypical weld metal is present. The staff does not disagree with the B&W evaluation of this material, although their evaluation does not follow the procedure for calculating RT given in the PTS rule. (The wasdevelopedwithoukThonsiderationoftheatypicalweld formulaThe metal.) for RT(Nf 5 believes that the probability of occurrence of atypical weld metal in vessel welds is low enough and its properties are such that ,

the results of a plant-specific probabilistic risk analysis would not be af fected significantly if the atypical weld metal was considered to be present. Thus, the materials input to the calcJlation of Rip will be evaluated without further consideration of the atypical weld Ntal.

The material properties of the controlling material and the associated margin and chemistry factor were repnrted to be:

1 D703300173 070313 P ADOCK 05000310

, I,DH PDR

( . ,

e .

2 ,

11: i i i: y Sn tun i i t.i l St.i f f E v.!lu.it inn O . I'. 0 . 15 .

Cu (cop;ier a:oistent ) '

U . 5'i 0.59 Ni (rtici:cl content , %)

O U l (insti.il Hi g, *f) .

s i'l

')9 H (l;.t ry i n , **f }.

  • - 22G.8 CT (Chemistry T.:etnr. *f)

The justifications Ihc conten!Iing mJLeelaI has been proper 1y idcoLified. Jrc .?Cceptabli' given for the Coppcr and niCI:Cl Corttent', ,!nd the isti tia l RT,, r thcsc values.

The margin hJs been dertved f ront can',ideratina of the bases',

followin9 the PTS Rule. Section 50.61 or 10 CFR Part 50.

fas t Neutron fluence Detailed calculations havq been performed with the assistance of our contracter.

  • Brookhaven National l'4bdratory (8tIL) .for the . cycle depender't orcssure v'es-sel inner-wall (E > h0 HeV) flu'xes for the Rancho Seco pl' ant. Th'e calculated -

fluxes were used to' determine the vessel (inner diameter) accumulated fluence.

which was then used to predict the RTPTS for specific welds in the reactor vessel. The analysis was based on a 00T calculation for the Rancho Seco plant.

An 80% load f actor. ' low leakage loading and plant specific data were assumed for the extrapolation. The following table summarizes the comparison of the Joplicant and the BNL results. The BNL estimates are in substantial agreement values with the applicant values, therefore, they are acceptabic. The RT pyg At the expiration of the current license are below the applicable criteria specified in 10 CFR 50.61 and are acceptabic.

RAtlCl10 SECO RT p73 CALCULATE 0 FOR OCTOBER 11. 2008 (LICEll5E EXPIRATI0ll)

Upts(O($'r)

NvTx10.is

( .1 )

(a) lint 1 0i::

i H SNUD ONL Screen (*F) SMUD Ucid llca t# CU HI 10, 48. S.4 5. 5 210 140. IN.9~ 0 . 0'l~

Hl OtT FV482J .15 .68 270 110. 111.5 0.13 U SHELL C50021 .12 .60 4 40. 7.1 7. 2 103.5 0.43 7.2 270 104.

U $11 ELL C50G22 .12 .60 -10. 40. 7.1 00.43 0.71 270 01.

L SHELL C50701 .10 .58 -20, 48. 7.1 7. 2

81. 00.43 0.71

-20, 40. 7.1 7. 2 270 L SHELL C50702 .10 .58 300 225. 225.2 0.09 .

U CIRC .29 .68 0. 59, 5.4 5.5 '

WF-233 300 241. 241.7 0.29 H CIRC W-154. .31 .59 0, 59, 7.1 7. 2

.5. 5 300 225, 225.2 0.09

l. CIRC .29 6 . O. 59. 5.4 193,. 193.7 . 0.'35 G21 *

' - . 6. 9 .

fULN3 L*%.W923

'; ;, u ,', %(: '.r<.\

I

  • 0D* *09. J 56.8..?'** % ., .,.. , .270-

..~ ,

- - - - - - - - - - - - - - - - - = - - - -

e .

t O =

L LN WF-29 .23 .63 0. 59. 6.9 7.0 270 ^

194. 194.1 -0.05 L LN ID WF-70 .35 .59 0. 59 6.9 7.0 270 265. 264.9 -0.04 L LN 00 WF-29 .23 .63 0. 59. 6.9 7.0 270 194. 194.1 -0.05 a) Ref. I 0 .

b) RT PTS =I+M[-10+470Cu+350CuNi]f.270 In view of; (a) The Pressure-Temperature updating requirements for the fracture toughness of the beltline material in 10 CFR 50 Appendix G. and (b) the fact that the RT PTS value is readily available f.om the calculation of the Pressure Temperature limits. and (c) the staf f desire to be informed on the current value of the RTPTS for all PWRs, .

we request that the licensee submit a reev,aluation of the RT PTS and a compari- ,

son to the prediction of Reference 1 along with the Pressure-Temperature operat-ing limits which are required by 10 CFR 50 Appendix G." It should be noted that this reevaluation is a requirement by 10 CFR 50.61, whenever core loadings, surveillance measurements, or other information indicate a significant change in projected values.

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