ML17305B393
| ML17305B393 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 03/15/1991 |
| From: | Conway W ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0612, RTR-NUREG-612 161-03811-WFC-J, 161-3811-WFC-J, NUDOCS 9103220299 | |
| Download: ML17305B393 (22) | |
Text
. ACCELERATED DI RIBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATXON DXSTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9103220299 DOC.DATE: 91/03/15 NOTARIZED: NO DOCKET FACIL:STN-50-528 Palo Verde Nuclear Station, Unit 1, Arizona Publi 05000528 STN-50-529 Palo Verde Nuclear Station, Unit 2, Arizona Publi 05000529 STN-50-530 Palo Verde Nuclear Station, Unit 3, Arizona Publi 05000530 AUTH.NAME AUTHOR AFFILIATION CONWAY,W.F.
Arizona Public Service Co.
(formerly Arizona Nuclear Power
=,R',
RECIP.NAME RECIPXENT AFFILIATION Document Control Branch (Document Control Desk)
I
SUBJECT:
Submits changes to commitments'e control:of heavy loads program at facilities,per NUREG-0612.Visual insp of special liftinq devices will be performed at beginning of each refueling outage, prior to returning device to svc.'ISTRIBUTION CODE:
A033D COPIES RECEIVED:LTR ENCLf SIZE:
TITLE: OR Submittal:
USX A-36 Control of Heavy Load Near Spent Fuel-NUREG-.06 A
NOTES:STANDARDIZED PLANT 05000528 Standardized plant.
05000529 Standardized plant.
05000530 D
RECIPIENT ID CODE/NAME EB NRR SINGH,A TRAMMELL,C INTERNAL: ACRS NRR/DS XB 8E R
E RE E B EXTERNAL: NRC PDR NOTES:
COPIES LTTR ENCL 1
1 4
4 2
2 6
6 1
1 1
1 1
1 1
1 1
1 RECIPIENT ID CODE/NAME PD5 LA PD5 PD THOMPSON,M NRR/DST/SPLB 8D OC/LFMB RES DE RES/DSR/RPSB NSIC COPIES LTTR ENCL 1
0 1
1 2
2 1
1 1
0 1
1 1
1 1
1 R
D NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 27 ENCL 25 A
D D
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Arizona Public Service Company P.O. BOX 53999
~
PHOENIX. ARIZONA85072<999 WILLIAMF. CONWAY EXECUTIVEVICEPRESIDENT NUCLEAR Docket Nos.
STN 50-528/529/530 161-03811-MFC/JRP March 15, 1991 U.
S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station Pl-37 Washington, D.
C.
20555
References:
1) 2)
3) 4)
APS Letter to U.
S.
APS Letter to U.
S.
APS Letter to U. S.
APS Letter to U. S.
NRC, ANPP-18281, dated June 25, 1981.
NRC, ANPP-19200, dated October 20, 1981.
NRC, ANPP-22328, dated November 18, 1982.
NRC, ANPP-23062, dated February 23, 1983.
Dear Sirs:
Subj ect:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2, and 3
NUREG-0612; Control of Heavy Loads at Nuclear Power Plants File: 91-056-026 This letter is being provided to inform the NRC of several changes to commitments regarding the control of heavy loads program at Palo Verde.
Previous commitments regarding the control of heavy loads were provided in the referenced APS correspondence.
PVNGS has re-evaluated the commitments made in the referenced
- letters, which responded to NUREG-0612, and has determined that changes to the existing inspection/testing program are warranted.
Currently, each special lifting device is either tested or examined annually, at a period not to exceed 14 months or prior to returning the device to service, if the device has not been used for more than a year.
Palo Verde is revising this test frequency to perform major Non-Destructive Examination (NDE) inspections at a less frequent interval (every third refueling outage,
+ one refueling outage).
Palo Verde will select one of the two following options to assure that the special lifting devices will perform as required for the safe handling of components.
Either of the options, A or B, will be performed for each unit at an interval of every third refueling outage
(+
one refueling outage):
A)
Performance of a complete visual inspection with applicable NDE tests (magnetic particle testing and liquid penetrate testing) at all major load-carrying welds and critical areas as defined in EER 87-ZC-072.
Dimensional testing shall'e performed ifindications from either the visual inspection or NDE testing warrant further inspection.
B)
Performance of a 125%
(minimum) load test followed by a complete visual inspection.
Perform dimensional testing, if required, as a result of the load test or visual inspection.
9i03220299 910315 PDR ADOCK 05000528 P
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61-3 1-0 JRP Document Control Desk U.
S. Nuclear Regulatory Commission NUREG-0612 Page Two In additfon to the NDE tests or load test, visual inspection on the special lifting devices will be performed at the beginning of each refueling outage, prior to returning the device to service.
The visual inspection will replace the current visual inspection, which had been performed annually (not to exceed 14 months)..
The special lifting devices affected are as follows:
i) reactor vessel head liftrig ii) upper guide structure liftrig iii) reactor vessel missile shield lifting frame iv)
CEDM cable support structure liftrig v)'CEDM,upper collecting ring liftrig Lift rigs i and ii were furnished by Combustion Engineering, liftriggs iii, iv and v were furnished'y Bechtel.
The analysis for the change to the control of heavy loads program and the re-evaluation of the Palo Verde inspection and testing, program for the five special lifting devices is based on the following:
0 I
The consequences of a heavy load drop due to failure of the special liftingdevices have been analyzed and found to be acceptable (CESSAR 9.1.4.3.5)
I lt
". ~i The successful completion to, date of all non-de'structive examinations
,performed on the special lifting devices The 'infrequent, dedicated 'use and storage of the special lifting
" devi'ces Survey of types and frequency-of inspections for similar programs
,at other nuclear ',power facilities The following analysis from CESSAR Section 9.1.4.3.5, concludes that the effects of the postulated dropping of the reactor vessel closure head on the reactor internals and core are acceptable,. in that the shutdown cooling supply flow paths and the reactor vessel support system will remain functional.
The maximum load'-carrying capability of the reactor vessel support system was evaluated by an elastic-plastic static analysis of:the support
- columns, using the MARC finite element computer program.
Each column was modeled with 13 rectangular section beam finite elements.
The material properties used were for SA 533 Gr B steel at room. temperature.
From the stress-strain curves (Figures 9.1-18 6 19)', the initial yield point was determined'o be 65,000 psi.
The axial gradually stiffness per inch, load vs. deflection curve (Figure 9 '-20) was obtained by applying a increasing axial load up to the point of static instability.
The of each reactor vessel support column is seen to be, about 40 x 10 lb.
up to a limit load of about 22 x 10~ lbs.
Ol l
I.
I
161-03811-MFC/JRP March 15, 1991 Document Control Desk U. S. Nuclear Regulatory Commission NUREG-0612 Page Three A nonlinear dynamic elastic-plastic analysi's of the reactor vessel support system was performed using the MARC program to evaluate the time-dependent behavior of the columns.
The results of a straight drop of the head assembly from 18 feet indicate that the load in the columns will reach the maximum load-carrying capability of the columns and some plastic deformation will occur.
The peak axial deflection for a drop from this height was determined to be
.56 inches.
The results show, that the.reactor vessel
,supports would remain intact and continue to support the wei'ght of the vessel fol'lowing such" an event.
t1 Possible drop configurations (othei than the 'straight drop of the head assembly) were considered to insure that the core will remain eoolable for the most severe
- drop, which can'ccur.
An off-angle';drop was,,demonstrated to produce bending about the wider dimension of the support columns',
~thus producing a more rigid system in response to this type of head drop.
It is concluded that the reacto'r vessel support system and the shutdown cooling supply flow paths will remain 'functional in the unlikely event of a free fall of the reactor head assembly from 18 feet above the reactor vessel flange.
In addition, the effect of the postulated dropping of the reactor vessel closure head on the reactor internals and core was assessed by performing a nonlinear elastic-plastic dynamic response analysis.
Two basic cases were considered:
a flat concentric head drop and an offset head drop.
In the flat concentric head drop, the head is assumed to drop in place from a height of 18 feet and contact the internals and vessel flanges uniformly.
In the offset head drop, the head is assumed to fall from the same height but from a laterally offset positi'on.
In that case, the head would contact the extension shafts and the top of the Control Element Assembly (CEA) shroud package.
The flat head'rop analysis utilized a
spring-mass model of the System 80 internals, core, vessel and vessel supports shown in Figure 9.1-21, as input to the CESHOCK code.
The analysis was performed by prescribing an initial velocity to the head mass at contact corresponding to an 18 foot free fall in air.
Since the upper guide structure flange is at a slightly higher elevation than the vessel flange, the head first contacts the Upper Guide Structure (UGS) flange, imparting a downward motion to the upper guide structure assembly.
The head then contacts the reactor vessel flange, forcing the vessel downward, compressing the reactor vessel supports and finally rebounding.
The dynamic response and peak forces resulting from the impact were calculated by CESHOCK.
Maximum stresses based on the peak loads were then calculated.
The results showed that the stresses in the fuel rod cladding would remain well below the yield strength.
Peak stresses in the other major components
.would exceed yield strength in some local areas but would still remain well below the ultimate strength.
Therefore, some local deformation may occur, but total failure, or gross deformation of the major components, is not expected.
This demonstrates that for a flat concentric head drop accident, the core would be maintained in a eoolable configuration and fuel rod damage would not occur.
0 I
161-03811-WFC/JRP March 15, 1991 Document Control Desk U.
S. Nuclear Regulatory Commission NUREG-0612 Page Four In the event of an offset head drop where the head falls from an 18 foot height but laterally displaced from the centerline of the vessel, the head would first contact the CEA extension shafts and CEA shroud package.
Since the CEA extension shafts are long flexible
- members, the energy required to buckle them is relatively negligible, so these components. are conservatively omitted from the dynamic model.
The analysis performed was similar to the flat head drop analysis.
The CESHOCK code spring-mass model i's shown in Figure 9.1-22.
This model contains an additional loading path going from the head to the CEA shroud package.
Since the degree of lateral offset can vary, several cases were considered by varying the stiffness of the CEA.shroud package elements.
For a small degree of offset, it was assumed that the load would be transferred through the entire cross sectional area of the CEA shrouds giving the stiffest loading path.
For a large offset, the load was assumed to be transferred through a
smaller shroud cross-sectional area providing a,softer loading path.
Buckling of the CEA shroud package was conservatively neglected since that would serve to reduce the load transferred to the internals and core.
Head rotation, at impact, was also conservatively ignored since that would also result in reducing the load transferred to the internals and core.
As anticipated, the results showed that the case that gave the highest loads was the small offset case in which the load was transferred through the stiffest CEA shroud package.
For large offsets,the shrouds limited the magnitude of the load. transferred to the internals-and core;providing, a 'greater "cushioning" effects Comparing the stresses of the worst offset head drop case with those of the flat concentric head drop case, it was determined that the flat head drop is more severe.
Therefore, it can be concluded that in the event of an offset head drop accident, the core would be maintained in' eoolable configuration and fuel rod damage would not occur.
In addition to the above analysis, Table 2.4-1 of reference 2 identifies the heavy load impact area combinations and Section 2.4-2-d provides the PVNGS evaluation for these combinations.
Risks associated with a special lift device failure are extremely low due to their infrequent and dedicated
- use, and the successful completion of NDE.
The special liftingdevices are typically used twice per refueling cycle for removal and re-installation of the specified component.
Use of the special lifting devices is controlled by Nuclear Administrative and Technical Manual administrative, procedures, with storage of the special liftingdevices controlled by the containment building equipment lay down and storage drawing (13-P-ZCG-110).
A review of NRC approved inspecti:on/testing programs on special lifting devices within the nuclear industry, indicates performance of major NDE inspections or load tests of the special lifting devices of up to once every ten years.
Currently, Palo Verde performs.major NDE inspections prior to use at the beginning of each refueling, outage (approximately every 18 months),
and performs visual inspections prior to use during the outage.
To date, all of the non-destructive examinations on the special lifting devices have been successfully completed.
The proposed change would have major NDE inspections performed less
)1
161-03811-WFC/JRP March 15, 1991 Document Control Desk U.
S. Nuclear Regulatory Commission NUREG-0612 Page Five frequently (every third'utage,
+ one outage).
Visual inspections would continue to be, performed at the beginning of each refueling outage, prior to returning the device to service.
In summary, the revised control of heavy loads program at Palo Verde would be as follows:
A)
Performance of a visual inspection on the special lifting devices at the beginning of each refueling outage, prior to returning the device to service.
Dimensional testing shall be performed if indications from, the visual inspection warrant it.
Every third refueling outage
(+ one refueling outage) the inspections shall be performed, in accordance with item B or C.
B)
Performance of a complete visual inspection with applicable NDE tests (magnetic particle testing and liquid penetrate testing) at all major load-carrying welds and critical areas as defined in EER 87-ZC-072.
Dimensional testing shall be performed if indicati'ons from either the visual inspection or NDE testing warrant further inspection, or C)
Performance of a 1258 minimum load test followed'by a complete visual inspection.
Perform dimensional testing, if required, as a result of the load test or visual inspection.
This revised inspection program satisfies the intent of NUREG-0612, Article 5.1.1, Guideline 4, regarding ANSI'N14.6-1978, Section 5.3 and is consistent with
.programs for similar special lifting devices:.within the nuclear power industry.
The changes to the aforementioned program will.be instituted'uring the upcoming Unit 3 Refueling Outage, which is scheduled to begin on.March 16, 1991.
If you should'ave any questions, please call Michael E'. Powell of my staff at (602) 340-4981.
Sincerely,,
WFC/JRP/pmm Attachment
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