ML20140F485
| ML20140F485 | |
| Person / Time | |
|---|---|
| Issue date: | 04/27/1978 |
| From: | Teh-Chiun Su Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20140F372 | List:
|
| References | |
| FOIA-85-665 NUDOCS 8604010059 | |
| Download: ML20140F485 (8) | |
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UNITED STATES 3* NUCLEAR REGULATORY COMMISSION j wasmwaTom. o. c. noses
'+2W. . . . #' ppo 2 71978 NOTE T0: f . Tedesco, Assistant Director for Plant Systems, DSS THRU: G. Lainas, Chief, Containment Systems Branch, DSS g J. Kudrick, Section A Leader, Containment Systems Branch, DSS FROM: T. M. Su, Containment Systems Branch, DSS
SUBJECT:
COMMENTS ON THE " REPORT ON BWR BLOWDOWN EXPERIENCE" I
Per your request, I have reviewed the Report on 'BWR Blowdown Experience.
The following sumarizes my comments on that report:
. 1. Discussion should include the recent development in load increases due to SRV consecutive actuation. (GE's Part 21 notification).
. 2. With respect to the discussion on the structural capability, discussion
- should also include those structures such as SRV line supports whose j structural capability cannot be related to the material fatigue cycles.
- 3. Correct pages 17,18 and 20 as marked. Copy of these marked-up pages is attached.
- 4. Finally, I believe that it may be the right time to surface the quencher device as the possible fix for SRV related pool dynamic loads; e.g., load increase due to consecutive actuation as indicated in Item 1 above.
- , T. M. Su Containment Systems Branch
, Division of Systems Safety
Enclosure:
As Stated cc: G. Lainas J. Kudrick
, T. Su
Contact:
I T. Su, CSB 492-7711 .
i g400059860114 0
FIRESTOG5-665 PDR l
1
.= . . . . - . - - - - - -
. r .
17 Thus, it is concluded that these BWR pressure relief system events are not likely to significantly affect the reactor vessel fatigue life even if they were to continue to occur at a frequency even greater than tnat indicated by operating experiencee .
3.2 pressure Suporession Pool Dynamic loading Considerations The steam discharge from an SRV is routed through piping from the dry ell t5 the'suppEssfo'n~ iioolTFfgure 5). There, the steam is condensed and the energy is absorbed by the heat capacity of the suppression pool. Prior to SRV actuation, the SRV discharge piping 4
between the SRY and the suppression pool is filled with air and the SRV discharge piping below the surface of the suppression pool is filled with water.
! Ouring an SRV actuation, high pressure steam compresses the air
! i column and accelerates the water leg in the submerged section of the discharge line. When the water leg has been discharge, the compressed air is released into the pool. The air bubble expands in the pool J-fran causing'a short duration, high pressure load on the sm arror' en, n eds~
suppression chamber (torus) . Th (refu//hyr'n o'n/ tron / Joofile pt*.ssare he.tsmome tum of the dis laced'M sr .JMittk' /'
i pool water causes the air Dubble to over expan fa"nd subseg tly 1 Lgsrt/_rw tf.*oo ff //te W P /wfdAr e Cgerr.ta 7N.s }g6,, ex8sc. l
- n - - ,yc . . . . ,s ,, , . . ~ m. , . . . . . . ,a -- -
, }&eff5fo%* CNoud WNrnYN~/5$ f 5 S)I5ttfe&,7y l g y/g ,,fam' .The steam subsequently discharged intVthe suppression pool causes
. . LCvm&lmf .Wendm_hg, low amplitude pressure oscillations 6n thPM, which continu for the remainder of the blowdown event. pere-+eads cn :.he ;vros wa.11s ara reanWaLthroe? the structure-to-the-toess-strproet3 end
, l
< -lS-4 Y/ f.'??.
J t pt:f ; 2**2:":d_:s the ; ,_x. In addition 'to the p :::;r. loads e6 cit. t;ru: Scuadsry, flow through the discharge Ifnes create reaction forces on the piping supports, and the pool motion induced by the discharge flow causes drag loads on the structures and components located within the pool. .
As the steam discharge continues, the temperature of the suppression
- pool will rise as the energy of the steam is absorbed by the pool. _
. At a point referred to as the "threshhold temperature", "* - -
JvoerDst<a e ch?a:" cca** :: the steam condensation process would becomeMnstable and the pressure oscillations could increase bf-a lactor--e# ten-er more. This effect is referred to as the steam quenching vibration phenomenon. The threshhold temperature for this phencmenon se PMa r'b ; ' = % O f t:a diad: ;; 'S -- m is considered to occur where the bulk pool temperature is on the order of 150*F to 170*F.
A large number of pressure relief system actuations have occurred in both domestic and foreign BWR facilities. In a number of cases, typic early in the life of a given facility, localized damage to the dis-charge line restraints in the suppression pool and to the suppression pool baffles has occurred. The cause of this localized damage has be.
attributed to the reaction loads and to the pressure forces genera *.ed during the discharge of the air bubble. In these cases, the affected
- structures were repaired such that additional structural cacacity was provided. In no case did this localized damage result in a loss of containment function or a release of radioactivity.
i .
- ' ~
! . 20-has become an integral part of the review of construction permit and operating license applications for all BWR pressure suppression con-i tainment designs (i.e., Mark I, II, and III).
I As a result of generic suppression pool hydrodynamic concerns, owners groups were formed by utilities with plants utilizing the Mark I and Mark II containment designs. Through these groups, generic, analytical and experimental programs have been developed to address SRV loads. For the operating facilities, the SRV related tasks of f
. the Mark I containment Long Term Program are intended to improve the quantification of SRV loads, to confirm the suppression chamber i
structural margins, and to confirm the adequacy of the suppression
! : pool temperature limits.
i i The staff elieves that there is no immediate (i.e., short term) poten-e Me otR1fr.Sfruch tetJ
- tial hazar rom thf vibratory loads associated with SRV operation due
! to the slowly progressive nature of the material fatigue mode of failure associated with cyclic loadings. Based upon the test results
. and analyses reported by the General Electric Company in " Steam Vent j
Clearing Phenomena and Structural Response of the BWR Torus," NE0010855 i
April 1973, substantial fatigue life margin is available in the torus structure to accomodate the potential SRV operations that may occur during the conduct of the LTP. The Mark I Owners Groua has recently perforned additional in-plant tests at the Monticello facility to identify and quantify the stresses in the torus structures associated with SRV operation. The need for structural modifications to provide
jf" "*% umino stares /i U e '4 NUCLEAR REGULATORY COMMISSION i q
j I* j wAsmNGTON, D. C. 20855
./
'+
.....# ppo 2 71S78 NOTE TO: R. Tedesco, Assistant Director for Plant Systems, DSS THRU: G. Lainas, Chief. Containment Systems Branch, DSS J. Kudrick, Section A Leader, Containment Systems Branch, DSS FROM: T. M. Su, Containment Systems Branch, DSS
SUBJECT:
COMMENTS ON E " REPORT ON BWR BLOWDOWN EXPERIENCE"' '%' j 0fc Per your request, I have reviewed the Report on BWR Blowdown Experience.
The following sumarizes my coments on that report:
's . Discussion should include the recent development in load increases due to SRV consecutive actuation. (GE's Part 21 notification).
- 2. With respect to the discussion on the structural capability, discussion should also include those structures such as SRV line supports whose structural capability cannot be related to the material fatigue cycles.
~
- 3. Correct pages 17,18 and 20 as marked. Copy of these marked-up pages is attached.
- 4. Finally, I believe that it may be the right time to surface the quencher device as the possible fix for SRV related pool dynamic loads; e.g., load increase due to consecutive actuation as indicated in Item 1 above.
/ .
T. M. Su
. Containment Systems Branch Division of Systems Safety
Enclosure:
i As Stated cc: G. Lainas J. K drick T. Su
Contact:
T. Su, CSS ,
492-7711' g
O
3 I * \ . _ . . _ . .
. v, n,
. /
17 Thus, it is concluded that these BWR pressure relief system events are not likely to significantly affect the reac, tor vessel fatigue life even if they were to continue to occur at a frequency even greater than that indicated by operating experience.- -
4 .
3.2 Pressure Suporession Pool Dynamic loading Considerations The steam discharge from an SRV is routed through piping from the drywell to the suppression pool (Figure 5). There, the steam is condensed and the energy is absorbed by the heat capacity of the suppression pool. Prior to SRV actuation, the SRY discharge piping between the SRV and the suppression pool is filled with air and the SRV discharge piping below the surface of the suppression pool is filled with water.
During an SRV actuation, high pressure steam compresses the air column and accelerates the water leg in the submerged section of the discharge line. When the water leg has been discharge, the compressed air is released into the pool. The air bubble expands in the pool causing a short duration, high pressure load on the Jharfo^** end crue r e ds.,
suppression chamber (torus) '
(pyju//hyM s'irknrg/ Ju/Sfe pir.ssem? M.tsThe momean tum.JWMV'@M of the displaced
.f *M pool water causes the air Dubble to over expan laiid subsequ tly pl?.s}g66 Lcortfrwefwoe rf /fie We dwl6Ar ec9,emta
,,,a.,,,,,_.~.,....... .2 - - - .- ,,e P sc.
'"' & wlkacftN C?diud Wh55W/'/NSt~ 5 S555,,,n/*∨-
g ,,y / g , The steam subsequently discharged into'the suppression pool causes
. . . . . LcmdeImf.S/rarfeso
. low amplitude pressure oscillations on thM:m =y, which continu
- for the remainder of the blowdown event. Prc
- ura 'ced5 e the turvs wt11s ara erv'ekf **=d 9 cugh--the a truc tur; t& the-tomttppccu W s
t l
In addition to the vdm p::::.k ta p'p' ; ett 9 :1 t: t'; = =, c. loads E i,:,, tr;: Scu~'7. flow t'hrough the discharge Ifnes create reaction forces on the piping supports, and the pool xtion' induced by the discharge flow causes drag loads on the structures and components located within the pool. ',,
- As the steam discharge continues, the temperatuie of the suppression pool will rise as the energy of the steam is absorbed by the pool.
At a point referred to as the "threshhold temperature", d tr.; d!:
\ viqrrnu 9 ;" c~ *' :: the steam condensation process would becdrhe ns able
- Jo *,*' 1F and the pressure oscillations could increase bf-e- fact:@en-ee ie .
fre e . This effect is referred to as the steam quenching vibration phenomenon. The.threshhold temperaturs for this phenomencn :ie pr' : "y '
t': ;ft:a dher.;cp "_ --:--
is considered to occur where the bulk pool temperature is on the order of-450*F--tc 17C*F.
A large number of pressure relief system actuations have occurred in both domestic and foreign BWR facilities. In a number of cases, typi early in the life of a given facility, localized damage to the dis-charge line restraints in the suppression pool and.to the suppression pool baffles has occurred. The cause of this localized damage has be attrib'uted to the reaction loads and to the pressure forces generatec during the discharge of the air bubble. In these cases, the affected structures were repaired such that additional structural capacity was provided. In no case did this localized damage result in a loss of containment function or a release of radioactivity.
~
l
. . i
. i has become an integral part of the review of construction permit and operating license applications for all BWR pressure 59ppr455fon con-tainment designs (i.e., Mark I, II, and III). ,
As a result of generic suppression pool hydrodynamic concerns, owners groups were formed by utilities with plants ut'141 zing the Mark I and Mark II containment designs. Through these groups, generic, analytical and experimental programs have been developed to address
-SRV Icads:-For the operating-facilities, the SRT~reTaterta-s kT6f the Mark I containment Long Tenn program are intended to improve the quantification of SRV loads, to confirm the suppression chamber structural margins, and to confirm the adequacy of the suppression pool temperature Ifmits.
The staff elieves r ne MR1 tha/r .Hruch%t there is no imediate (i.e., short term) pote tial hazar rom thi vibratory loads associated with SRV operation due to the slowly progressive nature of the material fatigue mode of i
failure associated with cyclic loadings. Based upon the test results and analyses reported by the General Electric Company in " Steam Vent Clearing Phenomena and Structural Response of the BWR Torus," NEDO 1085-April 1973, substantial fatigue life margin is available in the torus structure to accomodate the potential SRV operations that may occur during the conduct of the LTP. The Mark I Owne,rs Group has recently performed additional in-plant tests at the Monticello facility to identify and quantify the stresses in the torus structures associated with SRV operation. The need for s'tructural modifications to provide l
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