ML20216F848
| ML20216F848 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/09/1997 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20216F846 | List: |
| References | |
| NUDOCS 9709120123 | |
| Download: ML20216F848 (13) | |
Text
{{#Wiki_filter:FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT B FSAR PROPOSED REVISION Strikeout / Shadow Changes Text added in Revision Indicated'bhshadowed} 23 of the FSAR text-Text being deleted Indicated by strikeette text Text being added Indicated bv._dnu h underline text Df 05 302 P PDR
l ATI'ACilMENT H FSAR PROPOSED REVISION 5.4.4.2 Outside Reactor Building The methods and general _ criteria used to postulate and protect pipe rupture effects outside the reactor building at CR-3 are described in the report " Pipe Rupture Analysis Criteria Outside the Reactor Building Crystal River Unit 3" submitted by FPC's March 31, 1989 letter. Florida Power Corporation also submitted the following revision to this report: Revision 1 December 18, 1989. This report identifies and documents FPC's position on the various issues pertaining to pipe rupture requirements outside containment. The position has been established considering the technical and regulatory requirements at the time of plant design and construction, and current (including Generic Letter 87-11) NRC Standard Review Plan (SRP) guidance, modified as justified, to be compatible with existing design bases methods for CR-3. The purpose of this criteria is 'to provide acceptable pipe rupture postulation and protection methods for the plant that in general meet the intent of current NRC requirements, while maintaining and where appropriate, upgrading the existing plant ' design bases. The NRC found this position acceptable as described in the NRC letter dated April 11, 1990. The NRC aporoved nine rupture renort discussed above concluded that the high energv nortion of the letdown line outside containment is not subject to a high energy line break. The Standard Review Plan (SRPT allows the establishment of a "No Break Zone" if certain criteria are met. Also. ner Generic Letter 87-11 (" Relaxation i n. Arbitrary Intermediate Pine Runture Reouirements"). arbitrary intermediate nining_ breaks need not be opstulated if certain criteria are met. The nine runture renort demonstrates that the high energv oqrtion of the letdown line outside containment meets these criteria. L[ sing- SRP methods. the renort determined that the oining between the , containment ocnC1EiLtion and the outbaard isolation valve (MUV-491 meets the criteria for a "No Break Zone." Additionally. the renort determined that the stress in the ninino from the containment to valves MUV-44. 45. and 74 (manual isolation valves downstream of the block orifice and letdown control valves) is low enough that an arbitrary intermediate break need riot be considered in this section of the letdown line. Therefore. a break in the high eneray partion_of the letdown line outside containment is not considered a credible event. As a result. desianina for the dynamic or environmental effects of a high energv line break in the letdown line outside containment is not r,ggui red .
- - - - _ - - ~.. - - . .~. .- . .. - . ,. - .. ..
1 14.2.2.6 ' Maktup System Letdown Line Failure Accident 14.2.2.6.11 Identification of Cause
- Table 15-1 of Raoulatorv Guide 1.70. Revision -3. (" Standard Format and Content of Safety Analysis Reoorts for Nuclear Power Plants") includes breaks in lines EDnnected to the RC system that carry reactor coolant outside containment as a typical initiatingment which should be considered in Chanter 14 of the FSAR. '
Table 15-4 of Regulatory Guide 1.70. Revision 3. indicates that this event should be evaluated for dose constAutDigs. A ruoture in the letdown line was 3 DQ1_iDCluded in the original CR-3 FSAR. but was first evaluated in the-CR-3 Cycle -2 Reload Reoort (BAW-15211 to address that soecific reouirement in Regulatcry_ Cuide 1.70. Revision 3. A break- in fluid-bearing lines that penetrate the reactor ' containment- may cesult in the release of radioactivity to the environment. -There are _ no
. instrument lines connected to the. RCS that penetrate--the containment.
- However, there are other piping lines such as those associated with the Makeup and Purification (MU) System and the Decay Heat Removal (DH) System that penetrate the containment. For fluid penetrations in piping systems that do
~not. serve to limit the consequences :of accidents, leakage is minimized by a double-barrier design to ensure that no single credible failure or malfunction of an active component will result in either unacceptably high leakage or the loss of the capability to isolate a piping break. The installed double barriers consist of closed piping, both inside and outside the containment, and various types of isolation valves. .The most severe piping rupture-ittentified for which radioactivity release may occur is costulated during normal plant operation is in the letdown line of
. ..the Makeup and Purification System. Horeger. as discussed in section 5.4.4.2. a break in the high energv nortion of the letdown line outside containment is not considered a credible event. The Makeuo Svstem Letdown Line Failure Acci dent is oresented only to demonstrate that the dose consecuences from a Dostulated break in the letdown line outside containment remain below the 10 3 CFR 100 limits. Relative to dose conigguences. the hvoothetical br^.ak in the letdown line bounds other oostulated breaks in lines connected to the RC System that carry reactor coolant outside containment. This-involves-a The hypothetical rupture . of the letdown line is assumed to - accur just downstream of the outboard isolation valve and upstream of the letdown control valves. A rupture at this point produces would oroduce a loss of reactor coolant condition until the coerator i sol ates the letdown line followino a Loss of Subcoolino Margin (LSCM1. as directed by the Emergency Operatino - Procedures (EOPs). RES- pressure disp 5 belen the pre ssttre-for actuation of- the Engineered-Safeguards-Actuatien System-(ESAS) tc isolate the resi ter building. . When - thi s pressure is reached, the building-isolation signal-initiates-closure Gf the inboard-and-outboard-letdc.6 iscletton-valvesT C-losure sf--- the-isolation-valves- Isolation of the letdown line stops the release of reactor coolant and fission products -to the auxiliary building.. I thus terminating the loss-of-coolant phase of the accident. 14.2.2.6.2 Safety Evaluation Criterion
.The ' acceptance criterion ' for the evaluation of this accident is that the resultant doses shall not-exceed 10 CFR 100 limits. (Dose limits are 25 rem whole body dose and 300 rem thyroid dose.)
+
m- m~
14.2.2.6.3 Methods of Analysis The CRAFT 2 RELAPS/ MOD 2-B&W (References 30. 31 32. and 33) computer code was used to determine the reactor coolant mass release rates and the primary system response for the rupture of the letdown line. The multinode model includes a detailed model of the RCS as well as noding for simulation of the letdown piping, valves, and coolers. While the CR-3 LOCA analvsir codg_nf Incord is CRAFT 2. RELAP5 was used because it allows for more detailed modeling of the LOCA chenomena but still oroduces a conservative assessment of the RCS pressure resDQnie. The higher RCS ACR5sure orgdicted using_RELAP5 results in greater inventorv loss through the broken letdown line and maximizes the dose conseaugnCAL For purposes of calculating the mass of reactor coolant released, the reactor is assumed to be operc. ting at 2603 2fi19L4 MWt with-a-letdown-f4cw-of-140-gpm prior-to-the-rupture . The rupture is modeled as a complete severance of the 24 inch nominal diameter letdown line at a location downstream of the outboard isolation valve (MUV-49). As a consequence of the failure, the makeup control valve is assumed to move to the fully opened position to provide the maximum available makeup flow. This assumed control action delays the times for the trip of the reactor and the actuation--of LSAS isolation of letdown bv the coerator following a LSCM. and consequently increases the releases of reactor coolant mass and the fission products to the auxiliary buildirg. Automatie-actuation-of-ESAS-4s-assumed-to-occur-at-a-pressure-setpoint of --1350 psig--which-corresponds-to-the-nominal-value-of-1-500 psig-wdth-an-adjustment for-pos si bl e-i ns trumen t-error-equ al-t c 0% of the-4500-inig-range of the measurement---The-letdown-isolation-valve-ts-assumed to reach-the-fttily-elosed position-h4-seconds-af tee-the-ESAS pressure-setpoint is reached. This-time period-includes both the-instrumentation-delay-time-and the valve-stroke time-The caerator is assumed to isolate the letdown line at 10 minutes after the hot leg reaches saturation conditions. Doerator action to isolate letdonn_QD a tSCM is contained in the EOPs. The analvsis did not consider coerator action to increase the OTSC levels to the LSCM setooint. This is conservative since it maximizes the RCS oressure and the mass release through the brea L Dose calculations are based on a core power level of 2544 2620 MWt with the fission product concentrations corresponding to 1 percent defective fuel rods. Ten percent of the iodine contained in the mass of reactor coolant is assumed to volatilize and become airborne in the auxiliary building. The remaining 90% is assumed to remain in the liquid which drains into the auxiliary building sump. The airborne radioactive nuclides in the auxiliary building are filtered through High Efficiency Particulate Air (HEPA) and charcoal filters in the Auxiliary Building Ventilation (ABV) Jystem before being exhausted to the environment. The analysis is based on a conservatively estimated efficiency of 90% for iodine removal by the charcoal filters. The assumptions used in the evaluation of the off-site doses are summarized in Table 14-41. The dose calculations have been undated to the latest fuel and fuel evele de.sions for Cvele 11.
- Fof Cycle li, e re-evaltration i cf J this .' event was jerformsd .' te reflect f th'e
- Emergency-Operat+ng-Procedttri requkement-to-initiate lligh Preswre Injection
{HPD-en-a-bos s-of-Sttbeco14n g-Margi-n-fbSCM) . Prior to this-reLevaltrationr-the an alysi s-t ons i-dered-onl y-ma xi mum-ma k eup-f4 cw . The-effeet-of"fu14-HPI" was not-assessed. The-re s trit-o f-thit-re-ev al u a tion-regtttr es-opera *er-aetion-to
~
isoiate-letdown-fici to mitigate thi3 event:
14.2.2.6.4 Results of Analysis The calculated time for - the RCS to depressurize and- reach the actuation hat lea saturation pressure for-the-ESAS is 745 120 seconds. At a time of 752 1110 seconds, the isolation-valve-is-completely-elosed coerator is asiumed to have isolated the letdown linc. The total mass of reactor coolant that escapes through the break and is released to the auxiliary building is 457760 114.00D pounds. The fission product activities released to the environment during'the accident are listed in Table 14-42. The dose consequences of the letdown line rupture accident are' presented in Table 14-43. The table presents: (1) the thyroid 4 dose due to inhalation of iodine activity; and (2) the whole body doses from gamma radiation due to immersion in the gas cloud for individuals located at the outer boundaries of either the exclusion area or the low population zone for the first two hours after the accident. The resulting doses are small
- fractions of the 10 CFR 100 limits.
An analysis was performed to determine the offsite doses with no iodine filtration by the ABV System. The ABV System is non-safety-related and is not provided with emergency power. Tables 14-42 and 14-43 provide a comparison of the radionuclide releases and dose consequences, respectively. The results , show an increase in the offsite thyroid dose by a factor of 10 which is in proportion with the decrease in the assumed ABV System charcoal filter efficiency (90% vs 0%). The whole body doses increased by ITO la mrem at the EAB and Or1.LQ mrem at the LPZ. These doses are much less than the limits specified by 10 CFR 100. As-stated-earlici, "e re-evaluation-of-thit-event-was-medi'fer Cycl (11-tsing the-REf;APF cesputer-code-te detercine-the-effects-of-initiat+ng-HPI^ ent LSch Wh44e-the-ER-3-bOEA-analys4a code ef- rscord-it-ERAFT2, l RELAP5 ses-traed ;becesse 4t-aH e.s fer better . medelia; ef the LOCA phenesensa end preduces e mere conservative-asses;;ent of the L RCS . pressure response . . Opereterf &ctien te
~
e;teblish "rvil Url en -LSC"" results ia higher RC; pressures; then these dete. ;aed by the CRAFF2 cesputer cede. -These higher RCS ~ pressures result in ger.ater-4aventery . less threugh 'thei breken letde.a ' 14ac'. theni previously evaluated. This-inventory-loss-wiH centinuca unti4-eredit-fer sperater &ction is-taken-to-+solate-the-breken 14ne-at-10-minutes : ef ter LE5h-Isoiation - of. the-line-terminates-the-event s . The-RELAPS-analysis-determined-that-the retal' 4 avectory - lost-until--isol etica^i s - IT161091b s 1(Ref . 30) . Evea thosgh th+s 4asentary ..is greeter then the emeant deterrsiaed by . the.' CRAIT2 .eaelysis,1 the dese ceasequences she-a-ia Teble 14-43-remain-well belew the lists of 10 CTR 10&r . i l i
14.3 REFERENCES
;(30)f FEamatone'ITechnologies~ 216corporated? Summary {86-1257374h00l: "CR23"RELAP5' e sletdown?Line18reak,";3une128d19966 id w: _
; u :, , cm ;
(31S=BAW-10192P. Rev. O. "2WNT Loss-of-Coolant-Accident Evaluation Model for Once-Through Steam Generator Plants." Februarv. 1994_(Proorietarv3. 1321 Letter. USNRC (1. E. Lvons) to Framatome Technologies. Inc. (3. ' H . I Ta. lor). "Accentance for Referencing of Tonical Reggrt BAW-10192-P. BWNT Los s-of-Cool ant-Acci dent Evaluation Model for Once-Through Steam Generator Plants (TAC No. M894001." February 18. 1997. (331 Framatome Techno]pgi e s Incornorated Summarv 86-1257374-02. "CR-3 RELAPS latdown Line Break." Seotember 4. 1997. l
TABLE 14 t , SITUATIONS ANALYZED FOR ENGINEERED. SAFEGUARDS ANALYSIS Event .~ Analysis-Assumotions- Effects- -t
. Makeup System Letdown Letdown line ruptures- See Table 14-43 for
- Line-Rupture Accident just outside the environmental effects.
- containment and-upstream of the- i . letdown control valves.-- - A loss-of-coolant condition exists until'RES - -
preessie dic,ps below 1500 pe4g-and. contain;;at isolation otettfS Banual onerator '
-action is taken to i isclate the letdown #
.11!12 -
l 4 5' s 4 i l i i i
-l
-. TABLE-14-41 ANALYSIS ASSUMPTIONS FOR THE MU SYSTEM LETDOWN LINE FAILURE ACCIDENT ,
Data and Assumptions Used to Estimate Radioactive Source-Power level . MWt- M44 2620 -, Percent of fuel rods leaking, % -1.0
' Escape rate coefficient Table 11-1~ '
-- Fuel characteristics Cvele 11 (FSAR Section 3.121
- Reactor. Coolant Activity-p ~ " q & g ,, 7,-737
, ... wi-,. ..
Maclide Activitv. vCi/cc g j Actiyitvi uti/ce ~n) , Kr.83m 0 318 y te n-e s . g - w:ey 12,26J Kr 85m 1-46 L s V 85 4r36 i f17_170! "! " 87 Or779 b* ,
- T1Y17.:i ,' , 3.4 ,
88- h41 O ):3!62)
*~~
4 *
?
!' , . , . . i n Xe 131m h63 . ^12.472 e> .> .. .
~.
133m 2r58 h F . . . 74.+22 (%..'..
+
133 23810 0 .396;0s ~ 135m -Or294 4 0N28; , 135 4r88 i -9.825 vs. A c 138 07421 , , do.692) ._ p .< > s ;yg; if e ,..,...c ( * (: I. 4
~ ~ ~
I 131 3:47 ._.
- ' y 5.18.-
y; 132 1r17 ;1.99i J
. ?
M 133 3r7 ,-
, 4 yu6.2? - .
134 0-4&1 , L0.690 , 135 h68 v. ua
,, .053..?, .
- w. ' .n; a
M 3 e --, .- - ,v~ -
, , . +. - - * . - . ,-, *
. _ _,, _ . ~. _ ._ _ ..._..._..m__._.___
?
m
.e M
TABLE 14-41 ,
. (continued)
ANALYSIS ASSUMPTIONS FOR THE MU SYSTEM 1 LETDOWN LINE FAILURE ACCIDENI
-i
- Data and Assumptions Used to Estimate Radioactivity Released i Total. mass of. reactor coolant released , e y - m m _. ~ ,y m
- to auxiliary building, lb j l14 m ,000L m aut n a , aW m a U Fraction of _ iodine airborne in auxiliary building -0.1 Charcoal filter efficiency for Iodine, %. 90/0") - -
- Noble gas, % 0 t
Qisptrsion'Dat3
- Exclusion Area Boundary (EAB1, m 1340
. Low Pooulation Zone (LPZ1 boundary, m 8047 Atmospheric. dispersion percentile, % 5
- 0-2 hour atmospheric dispersion factors, s/m' at EA8 1.6 x 10
e at LPZ boundary 1.4 x 10 (a) Reflects analyses with and without credit taken for iodine filtration by - the auxiliary building ventilation system. l. t
l 1 I
-TABLE 14-42 ACTIVITY RELEASED TO ENVIRONMENT DUE TO
_ FAIL llRE OF THE MU SYSTE!Lt,ETDOWN LINE
' 3
"?yY 5,
5 ? ,. ff- 3'~-f - WP 5 4 f--] 2(Cic'e111)1. @ Nuclide Activityr-C4 ( PActivity; Ci! fj - (ar
---f a) E ; g.,gbF ]
Kr.83m s,s,-m 2L2 2JL2 4-; m y y. m y y Kr 85m 44T6 44T6 h ;;117.0 L b117'.0 ~: 3 . o .. . 85 131TO 131TO E 9Hhe {916i14 3
^ '
t $L1 . ng
~.' .
_i .-. , 87 23r5 23r5 { 60;6,:J'. L60!6L " 88 7h6 7h6 k Ji187e4.1 (18.7[4y? 7 c: 4 c : . ,f Xe 131m 49r1 49r1 :127)8h, z
?l27.8.1 1 133m 77.-7 7777 - 0218;4T i-21874) v ... ..
~
133 7170r0 71?Oro -{lc20495;4; - ;20495i44 135m Br65 ST65 ?22;2> 22;24 135 147TO 147T0 h )508 2 };508.2 3 138 1277 1277 pe 235',8 J 35.82 ,' I 131 1974 104TO N [2i6815 f I2E.81s 132 3r52 35r2 g i1,030~1 'j 18!30 yp 133 1171 ilho . 32209T !32 09l M 134 h39 13r9 L d)357--!: .-
;3s57:: '
135 ST66 56T6 - fij.30_97 , 13.093 g (a) With 90% iodine filtration by the auxiliary building ventilation system assumed. (b) -With no iodine filtration. by the auxiliary builtling ventilation system assumed. i
I TABLE 14-43 I i
SUMMARY
OF RESULTANT DOSES FOR THE j MU SYSTEM LETDOWN LINE FAILURE ACCIDENI i j 2-hour Total Integrated Doses at the Exclusion Area Boundary Thyroid-REM bliole-Bodvr-REM 90%-ABVS-iodine-fHtration 0r115- 0 066 0%-ABVir-iodine-filtration 1r15 Or067 Cycle 11 Thyroid. REM- Whole' Body. REM ~
; 90% ABV5 iodine filtration .0.304- 0.096L 10% ABV5; iodine filtration: 3.04 -0.099:
30-day Total Integrated Doses at the Low Population Zone Ihyrnid-REM Whole Body. REM 90%-ABVS~ iodine-filtration 0 4101 0T005-6 0%-ABVSr-iedine-ftltration Or101 070059 Cycle 11 -Thyroid'. REM Whole'Bodvi-REM-
! 90%' ABVS iodine filtration- 0x0233 0.008 0.027
-l M ABVS iodine-filtration Ore 33- 'O.009' DJ,2
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT C Acronyms and Abbreviations
ATTACHMENT C ACRONYMS AND ABBREVIATIONS
- ATOG -- ------- Abnormal Transient Operating Guidelines B&W----------- Babcock and Wilcox B&WOG ----------- Babcock and Wilcox Owners Group CFR-------- Code of Federal Regulations CR-3 --------- Crystal R ver i Un ti3 EAB--------------- Exclusion Area Boundary EOP -------- Emergency Operating Procedure 4
ESAS -------- Engineered Safeguards Actuation System FPC ---- ----- Florida Power Corporation FSAR ------------- Final Safety Analysis Report FTI --------- Framatome Technologies incorporated IIELB- ---------- Iligh Energy Line Break - IIPI - liigh Pressure injection-
- LER---------- Licensee Event Report LOCA --------- Loss of Coolant Accident LPZ ----------- --- Low Population Zone LSCM ----------- Loss of Subcooling Margin NRC ----- ----- Nuclear Regulatory Commission PGP-------------- Procedure Generation Package RBIC------------ Reactor Building Isolation and Cooling RCS--------- Reactor Coolant System SER ------------ Safety Evaluation Report T B D ------ - Technical Bases Document USQ---------------- Unreviewed Safety Question}}