ML17297B168

From kanterella
Revision as of 05:51, 29 February 2020 by StriderTol (talk | contribs) (StriderTol Bot insert)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for Palo Verde 1,2 & 3 Reactor Vessels.
ML17297B168
Person / Time
Site: Palisades, Palo Verde  Entergy icon.png
Issue date: 12/31/1981
From:
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML13308A045 List:
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-J, NUDOCS 8201130491
Download: ML17297B168 (28)
v • d • e


Text

CEN - 189 APPENDIX J EVALUATION OF PRESSURIZED THERMAL SHOCK EFFECTS DUE TO SMALL BREAK LOCA'S WITH LOSS OF FEEDWATER FOR THE PALO VERDE 1 & 2 , 2 & 3 REACTOR VESSELS Prepared for THE ARIZONA PUBLIC SERVICE COMPANY NUCLEAR POWER SYSTEMS DIVISION DECEMBER, 1981

~c/1/

POWER y/

SYSTEMS COMBUSTION ENGINEERING, INC.

LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF:

A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS; OR B. ASSUMES ANY LIABILITIESWITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT..

ABSTRACT This Appendix to CEN-189 pro'vides the plant-specific evaluation of pressurized thermal shock effects due to small break LOCA's with extended loss of feedwater for Palo Verde--l, 2, 8 3 reactor vessels. It is concluded that crack initiation would not occur for the transients considered for more than 32 effective full power years, which is assumed to represent full plant life.

i CEN-189 Appendix J TABLE OF CONTENTS SECTION TITLE PAGE ABSTRACT PURPOSE J2. SCOPE J3. INTRODUCTION J4, THERMAL HYDRAULIC ANALYSES J2 J5, FLUENCE DISTRIBUTIONS J2 J6. MATERIAL PROPERTIES J3 J7. VESSEL INTEGRITY EVALUATIONS J4 J8. CONCLUSIONS

J1.0 PURPOSE This Appendix provides the plant-specific evaluation of pressurized thermal shock effects of the SB LOCA + LOP< transients presented in the main body of the CEN-189 report for the Palo Verde--1, 2, 8 3 reactor vessels.

J2.0 SCOPE The scope of this Appendix is limited to the evaluation of the SB LOCA +

LOFM transients presented in CEN-189, as applied to the Palo Verde--l, 2, 8 3 reactor vessels.

Other C-E NSSS reactor vessels are reported in separate Appendices.

J

3.0 INTRODUCTION

This Appendix to CEN-189 was prepared by C-E for Arizona Public Service for their use in responding to Item II.I'.2.13 .of NUREG-0737 for the Palo Verde--l, 2, 8 3 reactor vessels.

This Appendix is intended to be a companion to the CEN-189 report.

The transients evaluated in this Appendix are those reported in Chapter 4.0 of the main report. Chapter J5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter J6 reports the plant-specific material properties and change of properties due to i rradiation, based on the methods of Chapter 6.0 of the report. Chapter J7 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to q set of lnaterjal propel"ties which are conseI'yatiye with respect to the plant specific properties reported in Chapter J6.

This additional conservatism was not removed because of the favorable results.

J4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to: perform the plant-.specific vessel evaluation reported in this 'Appendix are those reported in "

Chapter 4.0 of CEN-189. As discussed in the body of the report, there are several plant parameter conservatisms included in the analyses to develop these transients due to the reference plant approach used which could be eliminated by performing more detailed plant-specific thermal-hydraulic system analyses. Removal of these available conser-vatisms by additional analyses was not performed due to the favorable achieved. 'onclusion J.S.O FLUENCE DISTRIBUTiON The Palo Verde Units 1, 2 5 3 are not yet in operation and have not yet completed a surveillance capsule evaluation.

Since the vessel beltline materials are low copper content, detailed fluence profiles were not necessary for demonstra-tion of acceptable PTS capability. Accordingly, the FSAR end of life peak fluence prediction was used to estimate end of life material properties. Also, in order to evaluate the sensitivity of the fluence prediction value, material properties were also determined assuming an end of life fluence twice the FSAR prediction value.

J4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to: perform the plant-specific vessel evaluation reported in this 'Appendix are those reported in Chapter 4.0 of CEN-189. As discussed in the body of the report, there are several plant parameter conservatisms included in the analyses to develop these transients due to the reference plant approach used which could be eliminated by performing more detailed plant-specific thermal-hydraulic system analyses. Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved.

J.5.0 FLUENCE DISTRIBUTION The Palo Verde Units 1, 2 5 3 are not yet in operation and have not yet completed a surveillance capsule evaluation.

Since the vessel beltline materials are low copper content, detailed fluence profiles were not necessary for. demonstra-tion of acceptable PTS capability. Accordingly, the FSAR end of life peak fluence prediction was used to estimate end of life material properties. Also, in order to evaluate the sensi tivity of the fluence prediction value, material properties were also determined assuming an end of life fluence twice the FSAR prediction value.

1 APPENDIX J PALO VERDE NUCLEAR GENERATING STATION J.6 i4fATERIAL PROPERTIES The material chemistry and initial (pre-irradiation) toughness properties of the reactor vessel beltline materials for the three Palo Verde units are summarized in Tables J6-1, J6-2, and J6-3. The shift predictions are based on the maximum design fluence, 3.15 x 1019n/cm (E>1MeV) at the inside surface of the reactor vessel using Regulatory Guide 1.99. Taking the highest copper (0.07%)

and phosphorus (0.007%) contents for the plate materials, the predicted shift is 71F, resulting in an end-of-life (32 effective full power years) adjusted RTNDT of lllF. If the design fluence were increased by a factor of two to 6.3 x 19' 10 n/cm, the predicted shift and adjusted RTNDT for the most adverse combin-ation of properties is 99F and 139F, respectively.

J6-1 J3

Palo Verde 1, 2 8 3 Vessel Integrity The fracture mechanics analysis is performed using upper bound data for fluence and material properties in the Palo Verde 1, 2 8 3 vessels. The peak vessel fluence is considered to occur at the point of maximum RTNDT. The material toughness properties KIC and KIIa are determined from the calculated temperatures for the SBLOCA + LOFW transients using the method described in Section 7.3.3 and predicted RTgDT val ues through the depth of the wal 1 . Cri ti cal crack depth diagrams are constructed from the applied KI vs crack depth curves at the mid-core level of the vessel and the calculated material toughness curves. In this manner the integrity of the Palo Verde 1, 2 & 3 vessels are evaluated for the SBLOCA + LOFW transient.

Sugary of Physics and Material Data Input to Fracture Mechanics Analysis 19 2 A nominal design fluence value of 3. 15 x 10 n/cm (E > 1 MeV) was used to approximate the end-of-life fluence for the Palo Verde vessels, as well as a conservative upper bound of 6.30 x 10 n/cm or double the predicted end-of-life value. The peak fluence is considered to be uniform around the vessel. A conservative radial fluence attenuation was used such that:

exp (8.625 in. x .33 in. )

~

(a/w)

F exp (-2.85) (a/w) where F point fluence in wall F peak fluence at surface a/w fractional wall depth Upper bound materials data were used to conservatively envelope all plate and weld materials, which are as follows:

PCT. Cu .12 PCT. P .008 Initial RTNDT +40 F

Palo Verde 1, 2 & 3 Vessel Integrity The fracture mechanics analysis is performed using upper bound data for fluence and material properties in the Palo Verde 1, 2 & 3 vessels. The peak vessel fluence is considered to occur at the point of maximum RTNOT. The material toughness properties KIC and KIIa are determined from the calculated temperatures for the SBLOCA + LOFW transients using the method described in Section 7.3.3 and predicted RTNDT val ues through the depth of the wal .

'i Cri ti cal crack depth diagrams are constructed from the applied KI vs crack depth curves at the mid-core level of the vessel and the calculated material toughness curves. In this manner the integrity of the Palo Verde 1, 2 & 3 vessels are evaluated for the SBLOCA + LOFT transient, Summary of Physics and Material Oata Input to Fracture Mechanics Analysis A nominal desian fluence value of 3.15 x 10 n/cm (E > 1 MeV) was used to approximate the end-of-life fluence for the Palo Verde vessels, as well as a conservative upper bound of 6.30 x 10 n/cm or double the predicted end-of-life 'value. The peak fluence is considered to be uniform around the vessel. A conservative radial fluence attenuation was used such that:

exp (8.625 in. x ,33 in. )

~

(a/w)

F exp (-2.85) (a/w) where F point fluence in wall F peak fluence at surface

'/w fractional wall depth Upper bound materials data were used to conservatively envelope all plate and weld materials, which are as follows:

PCT.. Cu .12 PCT. P .008 Initial RT>~pT

+40 F

~-3* .

The shift in the value of the RT

>Twas determined using the method of Reg. Guide 1.99. This produces an end-of-life prediction for the surface RT>>> of 182 F using the nominal design fluence. A predicted surface RTNOT value of 256 F is determined for a fluence double the nominal design fluence.

Results of Fracture Mechanics Analysis for SRLOCA + LOFW Restoration of Feedwater (Case 5)

The stress analysis for this case is presented in Section 7.8.2 of the report. The fracture mechanics analyses were performed using upper bound properties for the Palo Verde 1, 2 8 3 vessels and conservative end-of-life fluence levels. The critical crack depth diagram is constructed using the stresses in the transient at the mid-core level coincident with the peak fluence and material proeprties.

Figure J.7-1 shows the critical crack depth diagra'm for a nominal design fluence of 3. 15 x 10 19 n/cm 2 . The calculated shifts in RT are relatively low, and for this transient loading condition the initiation toughness level is not exceeded. Therefore, no crack initiation would occur for this combination of loading, fluence, and material properties.

Figure J.7-2 shows the critical crack depth diagram for. the same transient loading and upper bound material properties, but twice the nominal design fluence. From the figure it is apparent that no crack initiation would occur for this transient even with fluence levels greatly exceeding the nominal design fluence.

Conclusion These results demonstrate that the integrity of the Palo Verde 1, 2 5 3 vessels would be maintained throughout the assumed life of the plant for the SBLOCA + LOFW transient with recovery of feedwater.

TABLE J6-1 PALO VERDE U."lIT 81 REACTOR VESSEL BELTLINE MATERIALS Product Material Drop I!eight Initial Chemical Form Identification NDTT 'F ~RT ~(~F ~Co er P osp orus Plate M-6701-1 -40 +30 0.07 0.005 Plate M-6701-2 -50 +40 0.06 0.004 Plate M-6701-3 -30 +40 '0.06 0.004 Plate M-4311-1 -10 -10 0.04 0.004 Plate M-4311-2 -40 -40 0.03 0.005 Plate M-4311-3 -20 -20 0.03 0.004 Held 101-124a -40 -40 0.02 0. 012 Ilel d 101-142b -40 -40 0.03 0.006 1<el d 101-17lc -40 -40 0.02 0.013 a Intermediate shell course longitudinal seam weld b Lower shell course longitudinal seam weld c Intermediate to lower shell qirth weld

TABLE J6-1 PALO VERDE'f'lIT //1 REACTOR VESSEL BELTLINE MATERIALS Product Material Drop Height Initial Chemical Content (5)

Form Identification NDTT 'F RT (F ~Co er P osp orus Plate M-6701-1 -40 +30 0.07 0.005 Plate M-6701-2 -50 +40 0.06 0.004 Plate M-6701-3 -30 +40 0.06 0.004 Pl ate M-4311-1 -10 -10 0.04 0.004 Plate H-4311-2 -40 -40 0.03 0.005 Pl ate M-4311-3 -20 -20 0.03 0.004 Held 101-124a -40 -40 0.02 0. 012 Held '01-142b

-40 -40 0.03 0.006 Held 101 171c -40 -40 0 '2 0.013 a Intermediate shell course lonqitudinal seam weld b Lower shell course longitudinal seam weld c Intermediate to lower shell girth weld

lp I

,N

TABLE J6-2 PALO VERDE UNIT A'2 REACTOR VESSEL BELTLINE NTERIALS Product Material Drop Weight Initial Chemical Content (I)

Form Identification NDTT 'F ~RT ~(~F ~Co or P osp orus J

Plate F-765-4 -30 -20 0.03 0.003 Plate F-765-5 -20 +10 0.03 0.004 Plate F-765-6 -30 +10 0.04 0.002 Plate F-773-1 -40 +10 0.03 0.003 Plate F-773-2 -50 0 0.04 0.003 Plate F-773-3 -60 -60 0.05 0.004 llel d 101-124 N/A N/A N/A N/A Neld 101-142 N/A N/A N/A i<el d 101-171 N/A N/A N/A N/A Not Available Intermediate shell course lonqitudinal seam weld Lower shell course longitudinal sean weld Intermediate to lower shell girth weld

TABLE J6-3 PALO VERDE U!lIT b3 REACTOR VESSEL BELTLINE fOTERIALS Product Material Drop Height Initial Chemical Content (5)

Form Identification NDTT 'F ~QI

~R Plate F-6407-4 -30 -30 0.04 0.002 Pl ate F-6407-5 -20 -20 0.05 0.002 Plate F-6407-6 -20 -20 0.04 0.002 Plate F-6411-1 40 -40 0.04 0.004 Plate F-6411-2 -10 0 0.04 0.004 Plate F-6411-3 -60 -60 0 '4 0 007 F

! (el d 101-124 H/A N/A fl/A H/A Wel d 101-142b N/A H/A H/A .N/A Held 101-171 N/A H/A H/.A N/A N/A Not Available a Intermediate shell course lonqitudinal seam weld b Lower shell course longitudinal seam weld c Intermediate to lower shell course girth weld

TABLE J6-3 PALO VERDE UNIT II3 REACTOR VESSEL BELTLINE tQTERIALS Product Material Drop lleight,

'F Initial

('F Chemical Content I)

Form Identification NDTT RT ~Co er P ospporus Pl a te F-6407-4 -30 -30 0.04 0.002 Pl ate F-6407-5 -20 -20 0.05 0.002 Plate F-6407-6 -20 -20 0.04 0.002 Plate F-64]l-l -40 -40 0.04 0.004 Plate F-6411-2 -10 0 0.04 0.004 00 Plate F-6411-3 -60 -60 0.04 0.007

!/el d 101-124 H/A N/A N/A H/A Weld 101-142b N/A H/A H/A 'N/A

!<el d 101-171 N/n ft/A H/A H/A N/A Not Available a Intermediate shell, course lonqitudinal seam weld b Lower shell course longitudinal seam weld c Intermediate to lower shell course girth weld

I f

4

1000EQOl aa<<a<<<<a<<a)a<<a<<a<<a<<ega<<a<<<<a<<a<<fee<<<<<<<<<<a<<)ee<<<<<<<<a<<afa<<aae<<a<<<<gee<<<<a<<<<<<ay a a<<la>>a

/

oooE+o

~ 6000E400 VOOOE+0

~ 0000fi00 1 o4000Et00 OOOWO

/

~ 2000E400

) 000&0

~ 7 1 05f e 1 4 )<<<<<<a<<<<a<<tea<<a<<a<<a<<tea<<a<<ee QOOOS+0

<<ye<<aaa<<a<<a+a<<<<<<<<

1 1 HE OOOO 1

0 H1H1

+ a oooooooo

<<<<<<af

-.ooooo oo tea <<a<<a+

ooooc oo Iolo Aoooo ~ oo ~ Doo F f'alo Verde too.  !., 2, 5 3 PCti-CU s ~ 12 P > F 008 PCS

-- --.Rf N015~ l82 DEO Fiqure J,7-1

CHlllCAE CRACK UEPTH VS ~ TlHE'

'VOQOttOQ UOQOt+UO

/

-" --- CR --/

~ lOQOE400 K I

.i4040t+00 0 I E / ~

.4 UUQtoQQ t I

~ JOQQttQQ I

I

~ 2UUOtiOO I

/

I

~ lOOQt+00

/

~ ilQht li 0~ ~ 2040Et02 ~ 4040Et42 ~ 4040E402 ~ 8000Et42 ~ )OQOEt0l FI.HENCE ~ obJE420 lHl 1 RTHOT i- 44 ~ --Dt0 ~

PCT ~ Cu ~ ~ l2 PCl -Q- atlll H/SMy CH f

Palo Verde Ho. 1, 2, fi 3 HYHOtS a 2+! ~ Htt Figure J.7-2

CHltfCAL CRACK OEPTH VSo TlHK' Maooktau I (g

/

uoookoua

"--- R- C A

a'iaaakoaa I4

~ 4UUok+00 0 I E

I

~ Suuakooa

~ 40aoktao

,2uuak+Oo I

. loaotooa

/

~ 0 l ahk a u~ ~ 20QOE F 02 ~ 1000Et02 ~ aoaoft02 ~ OOOOCt02 elOOOK<44 fl.uEiick'

- lN) l

.uaki2U R'fHDT >. 4Q ~ 'Ljkae F PC[a Cu ~ ~ l2 ac;.-a- ..aaa N/Su, CH Palo Verde Ho. 1, 2, g< 3 HfiiotS e 2il ~ l>tb. f Figure J.7-2

h I

h I

i!

J

8.0 CONCLUSION

S This Appendix to CEN-189 provides the results of analytical evaluations of pressurized thermal shock effects on the Palo Verde reactor vessels for cases of a SBLOCA + LOFH, in response to the requirements of Item II.K.2.13 of NUREG-0737. Two different scenarios were chosen for eval-uation based on remedial actions to prevent inadequate core cooling:

1. SBLOCA + LOFW + PORV's opened after 10 minutes
2. SBLOCA + LOFW + Aux, FW reinstated after 30 minutes Thermal-hydraulic system transient calculations were performed on a reference-plant basis, as reported in CEN-189 with the parameter variations over the range representing all operating plants, Four different cases were analyzed for each of the two different scenarios defined above, for a total of eight cases. The most challenging of the two different scenarios was analyzed using linear elastic fracture mechanics methods to determine the critical crack tip stress intensity values for comparison to plant specific materials properties at various times in plant life. The effect of the warm prestress phenomenon is identified where applicable for each transient, and credited where appropriate.

In this Appendix, the results Of plant specific peak neutron fluence predictions are superimposed on plant specific material proper-ties to define vessel capability versus plant life. The results of the generic LEFM analyses were evaluated using the plant specific material properties. It is concluded that crack initiation would not occur due to the SBLOCA + LOFW transients considered, for more than 32 effective full power years of operation, which is assumed to represent full plant life.

t I

Retrieved from "https://kanterella.com/w/index.php?title=ML17297B168&oldid=2519407"