ML20149E056
| ML20149E056 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 10/20/1995 |
| From: | Mark Blumberg HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20149D930 | List: |
| References | |
| NUDOCS 9707180081 | |
| Download: ML20149E056 (29) | |
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Rev.2 Page 40 of 43
- , OPGP05-ZA-0002
' 10CFR50.59 Evaluations Unreviewed Safety Question Evaluation form (Sample) Page1of4 h Form 2
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- c. ORIONATING DOCUMENT: CN 1980 NOTE: Ansch 10CFR50Ji9 Screening Form or Ucense Campsarce review Form to tNs USOE.
NOTE: Use addamonal sheets as necessary to pwide me bases, i
A.1 1 Does me subject of his evaluation heresse me probabDity of occurrence of an YES X NO accident previously evaluated in the Safety Analysis Report?
Basee: SEE ATTACHMENT.
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il Does the s4 ject of this evaluation heresse N consequerces of an accidant YES NO 4
previously evaluated in the Safety Anahsis Heport?
. Bases: SEE ATTACHMENT.
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i All Does me subject of this evahastion bcrease the probability of ecourtence of a ,_ . _
malfunction of equipment irrportant to safety previously evaluated in the Safety YES X NO Analysis Report? ,,,,,,,, ,,,,,,
Bases: SEE ATTACHMENT.
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- W Does the subject of ins enduation heresse the cunsequernes of a mainJnction of YES X M equipment irrportant to safety previously evaluated h the Safety Analysis Report?
' Bases: SEE ATTACHMENT.
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'I-OPGP05-ZA-0002 Rev.2 Pap 41 of 43 I 10CfR50.59 Evaluations
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'i Form 2 Unreviewed Safety Question Evaluation form (Sample) Page 2 of 4
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Does me sd$ set of me evaluation o,eate me posshaity of an W of a e,-1,pe m.n any,,
SEE ATTACHMENT us-iea me Suet,-,y. n.po,ri O.E]= ,
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Does the s@ of this evaluevon create the posseitity of a ddferent type of YES NO 81 ensomtion than any p,ev6ously evaluated in,me Safety Analysis Repo,t?
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Bases: $EE ATTACHMENT ,
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i A.3 1 Does the subject of this evaluation , educe the rnergin of eefety as defined in YES the basis ter any Technical Specircations? .
X N.O -
i Bases: SEE ATTACHMENT e
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1 l OPGP05-ZA-0002 Rev.2 Pue 42 er 43
- l 10CFR50.59 Evaluations j3 Unreviewed Safety Question Evaluation Form (Sample) Page 3 of 4
- } Form 2 1
- 4 SAFETY EVALUATION
SUMMARY
4,
- , The foiwing documents have been reviewed as part of the 10CFR50.59 frial screening process.
j; UFSAR 6.4 A.1. Table 11-B.2 2,15.6.5.3,15.B.2.1, 3.11,12.3
. Technical S;+^"--d-: is Table of contents Safety Evaluation Repoit 15.6.5.2, 6.4, suppl. #4 - Sect 3.11
- Amend, to Operating Lic. Unit 1 Amendment #38, Unit 2 Amendment #29 -
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A quet, tion by question evaluation is provided in the attachment.
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- t-i OPGP05-ZA-0002 Rev.2 Page 43 of 43
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,'. 10CFR50.59 Evaluations Unreviewed Safety Question Evaluation Form (Sample) Page 4 of 4
- j form 2 AI of the above questions were answered No; therefore, the originatdg documerd does l f B. 1. X l
'I agiinvolve an Unreviewed Safety Question.
- {
One or more of the above questions was marked YES; therefore, the originating j
! 2.
document invokes an Unreviewed Safety Question. The originating document, as i presented, shat! NOT be implemented without prior approval by the NRC. Provide a 4
imTuner.dation for disposition of the Unreviewed Safety Question below Refer to ll' OPGP05-ZA-0004 for processing licensing amendments. Further processing of this form to the PORC, Plant Manager and NSRB is agi required. Notify Procedure Control that the evaluation involved an Unreviewed Safety Question so that Procedure Control can close the USOE number. l l
2, RECOMMENDED DISPOSITION:
- APPROVE THE CHANGE.
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PREPARED BY: Mark Blumberg h m/3 j7 /<1, - "
9/28/95 ORIGINATOR Date REVIEWED BY: M. A.Whitley d*(t/ ( 0 i. W QUAUFIED RE .WER Date
!t l f 4MM APPROVED BY: D. A. Leazar A ,
,, folr/9[
D, ate DEPARTMENT MAN ER $F PORC MEETING NO. i b C86O' /0/2.0/9f Date i
APPROVED BY- a a__ s%w_ p/j/S '
011 % ANAGER Date
!i l* REMARKS:
Cepy c4 usGE wpn ved o.b+e ek p 9 b~~ o so . %
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, / DESCRIPTION OF CHANGE: .
o This Change Notice updates the UFSAR for a change in design assumptions used to l calculate radiological doses resulting from a Loss of Cooling Accident (LOCA) (NC6013,
- l Rev.4 9 NC9004, Rev. 6 9). This update evaluates the doses resulting from paa-atid
Mocated in the Mechanical Auxiliary Building (MAB).
/i -
. REASON FOR CH ANGE: ,
Review of NRC Infonnation Notice IEN 91-056 determined that potential leak paths exist which could result in radioactive sump water entering the RWST thmugh either the Safety
-Injection System (SIS) pump mini flow recirculation line's, the Containment Spny (CS) pump discharge test lines or the RWST suction isolation valves during the recirculation '
phase of cooling following a DBA. ,
The concern specifically involves'the SIS pump mini-flow reciri:ulation (SI 0011!0012 A,B,C and SI 0013/0014 A,B,C), CS pump manual test (XCS 0008 A,B,C), and the '
RWST suction isolation valves (SI 0001/0002 A,B,C). ,
Each HHS! and LHSI train contains two, in sedes, normally open, packless diaphragm, motor operated valves (MOVs) in the recirculation lines. 'Ihese mini. flow valves function to close on the switchover from the Injection to the Recirculation Phase to prevent radioactive sump water from being lost to the RWST. The normally closed, test line isolation valves in the CS System serve only to isolate CS pump discharge from the RWST during normal and accident operations. The RWST suction isolation valve
- F e phase.
i
- , ,Although these valves close, they may leak during the subsequent prolonged recirculation '
i phases of core cooldown. Any leakage through these valves will flow into the RWST.
j This possible leakage was not previously considered when calculating Control Room, q Technical Support Center (TSC), Offsite, Equipment Qualification (EQ), and post-accident
- zone doses. Therefore, the SPR on this subject (910475) identiGed a corrective action
- which evaluated the radiological effects of this leakage. This proposed change reDects the results of this evaluation.
{ "This,USQE revision supports the proposed changes due td'6ifferince N the' calculated' ~'
Yleakage transport time assumed in radiological calculations.' USQE 92-022, Revision 0
- was based upon analysis performed in Calculation MC6313,'Rev. O. t2his analysis b
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'l;j j ArrAcfDEENT To US9E 924022 cascINATDio DOCUMENT LCN-W77 1980 i
{Ealculated a transpi><t tim .4dya ntly, MC6313, Rev. O was superseded by
'!!, b.O. ins culation determined the backleakage time to be l Calculation M j /approximat y 423 day The change in transport time has been incorporated in
and NC-9004, Rev. 9, " Post LOCA Zones and EQ."' The increase in transport time l
I increases the decay of radioactivity and, thus, decreases the total radioactivity transported I
E3o the RWST. This yields lower offsite and onsite doses. This change in doses is
! ' reflected in this revision of the USQE (Revision 1).
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- 10CFR50J9 FINAL SCREENING FORM RESPONSES ,
t TechnicalJustification of Change ,
- 1. -Does the subject of this review invo lve a c hange to the facility as described in the SAR?
Yes, this USQE proposes a change to UFSAR Sections 6.4, "Hebitability Systems." '
1.BJI.2, " Design Review of Shielding and Environmental Qualifications of Equipment l .
for Spces/ Systems Which May Be Used in Post Accident Operations," and 15.6.5,
- Loss g
of Coolant Accidents". The change to these analyses introduces a previously unreviewed ;
system interaction (ie. Previously these analyses did not assume potential leakage of *
. containment sump water into the Refueling Water Storage Tank (RWST)).
- 2) Dose the subject of this review involve a change to the procedures as described in the SAR7 .
g No, UFSAR procedures are not impacte'd by this change. No UFSAR or SER procedures discuss or describe.the LOCA analysis affected by' this proposed change. *
- 3) Does the subject of this review propose the conduct of tests or experiments not described '
- . 6 &cSAR7 ,
No, this change is not a test or experiment, nor does it propose the conduct of a test or ll
!. experiment. The change describes the effect of an additional leakage pathway on design
!; basis dose analysis. -
- 4) Does the proposed change affect conditions or bases assumed in the S AR or safety related ff functions of equipment / systems, even though the proposed change does not entail any l[
physical change in existing structures, systems, or procedures as described in the.SAR?
- j i
a Yes, the proposed change affects the bases assumed in the UFSAR. Pseviously, the
,I
!; potential leakage of containment sump water into the Refueling Water Storage Tank i
(RWST). The proposed change incorporates this leakage. Therefore, this is a change to 4
the bases.
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" *USQE RESPONSES I A.1 I) Does the subject of this evaluation increase the probability of occurance of an
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accident previously evaluated in the SAR7 i!
'No. As in the pitvious analysis, the proposed change assumes that the Imss of ll1i Coolant Accident will occur. Therefore, there is no increase in the probability of
"! ' an accident previously analyzed in the SAR.
j Does the subject of this evaluation increase the maq'es of an accident
, A.1 11) '
previously evaluated in the SAR7
? No. "Ihe proposed change does not increase the consequences of an accident
- previously evaluated in the SAR.
The proposed change does not degrade or prevent action esem A in the SAR. It
- does add an assumption to those previously snade in evaluating the radiological consequences of a LOCA described in the SAR, but the radiological consequences described in this change are bounded by those set by 10CFR100 and SER Sectiens
~' 15.6.5.2.5 and 6.4 (dated April 1986). Per these documents, the accerece criteria dose's for a LOCA are as'follows: .
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!4 ACCEPTANCE CRITERIA DOSES (REM) l LOCATION -
BETA SKIN THYROID WHOLE BODY O EZB N/A 300 . 25 LPZ N/A 300 25 CR 30 30 5 l 1 -
TSC .30 30 5 l
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T here: EZB - Exclusion Zone Boundary, LPZ - Low Population Zone, CR ij - Control Roorn, and TSC - Technical Support Center j jg ;
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, l onronunnorKacuuzNTscx.em seso Tables 6.4-2,II.B.2-2, and 15.611 of the proposed change give the EZB. LPZ, CR and TSC doses as summarized below:
t PROPOSED DOSES (REM)
LOCATION BETA SKIN THYROID WHOLE BODY EZB 1.22 137.0 2.27 LPZ 0.47 66.3 0.73 I
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CR 21.52 22.7 2.43
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TSC 24.55 28.6 4.85 l l
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The values of the proposed change are less than the acceptance criteria doses l
i given above. ,,
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I Based upon these results, there is no increase in the consequences of an accident l previously analyzed in the SAR. l i ~
'i A.1III) Does the subject of this evaluation increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR?
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t No. The subject of this evaluation does not increase the probability of occurrence
' of a malfunction of equipment important to safety previously evaluated in'the
! SAR.
The effects of RWST backleakage upon equipment qualification doses is evaluated in this proposed change. The changes are given in the table below, iI; e
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3 ATTAC3 DENT To US9E Sa4K'2a camxNAMG DoCmENT LcK.sM 198o
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- ROOM ACCIDENT EQ DOSE (RAD)
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- - CURRENT PROPOSED 62 1.6E+5 1.2E+4 l
- 44E+9 4:6E+5 n -
1 63 2.0E+5 1.6E44
+:SE+3 lih0E+6 ,
.";; u;as C !!S-!!",2:52^, e.J -215%,0 riep +:4 ve: d. .; ..;!aip;;; en
- e,.bdr.; gr.:1.'._ en. These values are smaller than the current values and are enveloped with sufficient margin as required by 10CFR50.49, and other
- qualification standards (IEEE 3231974 and NUREG-0588, Rev.1).
A.1IV) Does the subject of this evaluation increa:e the consequences of a malfunction of equipment important to safety previously evaluated in'the SAR7, ,
No. The subject of this evaluation does not increase the consequences of a malfunction of equipment important to safety p'reviously evaluated in the SAR.
This change proposes making an additional assumption in the LOCA doses i analysis. It evaluates the doses resulting from potential leakage of containment sump water into the Refueling Water Storage Tank (RWST) located in the Mechanical Auxiliary Building (MAB). He LOCA analysis already assumes malfunction of equipment important to safety. Per the justification to question A.1 l* -
II, this malfunction does not increase the consequences of a malfunction of this p equipment. -
! A.2 I) Does the subject of the evaluation create the possibility of an accident of 4 difTenent type than any previously evaluated in the SAR7 .
[ -
No. He proposed change does not create the possibility of an accident of a different type previously evaluated in the SAR. This change analyzes the effect i-of potential leakage of containtnent sump water into the RWST on LOCA doses.
This additional assumption does not create the possibility of a different accident.
y A.2 II) Does the subject of this evaluation create the possibility of a different type of g
j malfunction than any previously evaluated in the SAR?
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1 I ATrewwNT To US9e 92 oo22 omanumopocuurarrscu4m aseo l
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No. This proposed change does not create the possibility of a different type of
- malfunction than any previously evaluated in the SAR. His change analyzes the
' effect of potential leakage of containment sump water into the RWST on LOCA i
doses. ' Ibis change, in itself, does not lead to a failure mode of a different type than previously evaluated.
i AS I) - Does the subject of this evaluation reduce the margin of safety as defined in the 4 .
basis for any Technical Specifications?
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j' f . No.The subject of this evaluation does not reduce the margin of safety as defined ~
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in the basis for any Technical Specifications. l 1
The acceptanm criteria for the TSC, Control Room, LPZ and EZP doses are given
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in the response to question A.1 II as defined in 10CFR100 and SER Sections I) 15.65.2.5 and 6.4. The margin of safety is reduced when the onsite (TSC and 4
j
,t Control Room) and offsite (LPZ and EZP) doses exceed these ace:ptance criteria.
- As previously discussed in the iesponse to question A.I.II, the TSC, Control ,
i . Room, LPZ and EZP doses are below the acceptance limits. ,
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I Tbc acceptance criteria for equipment qualification exposed to post LOCA recirculating fluid environments is given in the SER, Supplement 4 (pg 3-24), and
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Amendments 38 and 29 to the facility operating license. The SER states:
"The maximum value specified by the applicant for use in equipment qualification inside containment and in areas outside containment exposed l; to post-LOCA recirculating fluid environments is 1.4 E+8 Rads (gamma ,
,j plus beta). His value is acceptable for use in the qualification of equipment."
t Amendments 38 and 29 for Units 1 and 2, sespectively, increase this value to 1.5 i E+8 Rads.
t
,- The maximum dose due to RWST backleakage is 2.0 E ; 51.6E+4 Rads, which is l
- . well below the acceptance criteria value of L5 E+8 Rads.
Therefore, this proposed change does not reduce the margin of safety as defined In the basis for any ' Technical Specification" (where " Technical Specification" is ,
defined as the SER document and license amendments that define the licensing
. basis). ,
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(,tJ - @O STPEGS UFSAR r ii A review has been done to show the materials for equipment which have M en 4 agualified for the original design condition of NaOH spray of 7.5 to 10.5 pH j, are either not affected by the change to the new pH environment or replaced i -
wit.h a suitable material if affected. .
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i The. containment atmosphere is maintained below 4-volume-percent bydrogen ,
consistent with t.he recommendations of RG 1.7 as' discussed in section 6.2.5. # ;
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!NM 3.11.5.2 madiation Envira - nt. Safety-related systems and components i}
I are designed to perform their safety-related functions af ter normal operation ,
1 radiation exposure plus a DaA exposure. The normal operational exposure is j based on the design basis source terms presented in Sections 11.1, 11.2, 1 1.3, i
]ll j and 12.2.1 and the equipment and shielding configurations given .in Section
- 12.3. .
The effect of the Vantage 5H (V5H) fuel upgrade on radioactivity
- f -$
concentrations in the fluid systems was reviewed and it was determined that j
the original reactor coolant activity listed in Table 11.1-2 is bounding. C f[
j*. Therefore, the FSAR analyses based on this activity are not adversely impacted i j,
by the fuel upgrade. For comparison, the activity concentrations calculated for the V5H fuel are listed in Table 11.1-2A. The corresponding reactor core M l
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,, activity for the V5H upgrade is shown in Table 15. A-1A.
1 ,
' safety-related system and component radiation exposurc e are dependent on equipaent location and the particular DBA involved. In the contai-nt ami
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control room area, equipment exposures are based on the DBA LOCA. For in-
- j' containment equipment, the DBA 14CA source term is based on a release of 100 percent of the core noble gases, 50 percent of the balogans and 1 percent of .
- the solids. This is consistent with the guidance given in RG 1.89. Control room exposures following a 5,ostulated IACA are controlled to 5 rads less
- - [l-l gwf arant.,yit)1,,Lhe.pemi r-a* = af crnC_19 of 10CFR5S Appendix _ -The-
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- - , . 0 . ', . 0 . The source terms used correspond to a cycle length of approximately
l+ 20, 0 'D/MTU, a core average burnup of 40,000 MWD /MW, and a discharge
,; burnup of 60,000 MWD /MTU. These burnups are conservative relative to the j.' planned cycle lengths for V5H fuel described in Section 4.3. l C. / ,
Radiation source terms for safsty-related components which are exposed to
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i post-accident recirculation fluid are consistent with the recommendations of
- RG 1 89 (i.e., 50 percent of th2 core halogen inventory and 1 percent of the remaining core solid fission product inventory are mixed in the recirculation ir water).
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, Normal and accident radiation doses for the various plant areas are presented i in Table 3.11-1. safety-related equipment design doses are the sum of normal
, plus accident exposures. The design radiation exposures delineated in Tabic 3.11-1 are based on gaasna and beta radiation. Radiation source terms for l
safety-related components outside containment are based on gamma radiation.
organic materials in the containment are identified in section 6.1.2. For the ,
i*- organic coating materials used inside containment (see section 6.1.2.1),
irradiation tests performed by oak Ridge mational Laboratory have been performed for. an 1.itegrated gassna dose of 1 x 10' rada (which exceeds the design calculated value in Table 3.11-1). These doses conservatively account
. for the surface exposure due to beta radiation in the design basis LOCA environment.
3.11-5 Revision 5 I(
CHANGE NOTIM- N l
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, i TAgtt 3.11 1 (Centltwed) tWVfROIMEeTAL CONO1710its Relative CLsestative todistion* l Q tecetten T - .'eture Pressure Munidif f Dessee 4 (Envirer;nentet Norset Renee Atnormal Accident Normat Range Accident Nereal Aceldent Redletten l testeneter) *
- teen / min. *F1 (*F 1 Worsel Accident (non/ min. 11 (2) trods) treds) Tvee sigh Activity tpent 120/44 125 ellshtly 0.3 pelg 80/20 ;00 2x10' 100 genne 104/50 teetn storese Tert negative 4
I.MK 10 teoctor Meke se IM/50 104/44 125 slightly 1.2 pets 80/20 100 2x10 8
..e.ie' + bet ~.
Weter Purp CLbteles negative I letTO i Y
(m. m2) 1.6 M Refueling Water 111/44 130 alightly 2.4 pois 80/20 100 6x1M i.;.d gesne
- bet 104/50 storese fart Room nesettve
, (Am. 063) 8 8 nonRadioective 1M/50 133/44 125 slightly 0.3 pels 80/20 100 10 3.5:10 genne i.e Pipe chose (mm. 064) negative 8
h tteettleetfortp- 104/$0 131/44 135 slightly 1.1 pels 80/20 100 100 1.3x10 ensne l negative sent seem (mm. 065) Q O 4.4 ,slightly 1.1 pels 100 10 8
130 genes tn g
- tesentist O llier 113/50 115/44 120 90/20
~ negettve g & COf
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06M)
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- r. 7Room.
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terrider 104/50 126/44 170 slightly 1.1 pels 80/03 100 les 160 geese
.} tese. M7A, M73) nesetive O 8 gesne U1 terridor 104/50 125/44 195 ellshtly 1.4 pelg 80/20 100 10 100 (tm. 047C) negative 8
corrider 1M/50 126/44 170 alightly 1.1 pets 80/20 100 10 1.3 13 8 genne
%g len. M701 negative E
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! STPECS UFSAR room envelope. In the worst case, 235 cfm of tho' makeup air is filtered by 'E the makeup units, but not by the recirculation units, before it is introduce 2 into the control room envelope. De air-handling unit supplies the l conditioned air to the control room envelope. A summary of these parameters G
, is presented in Table 6.4 2. An unfiltered inleakage of 10 sefs to the -
-control room envelope has also been assumed (Ref. 6.44 4
. CN rh. e,_Conen Neen g l
h.. atmoschgic r_ eleases e T.1 (".0" led.-;gurge fMr;.611..J.a. -y___
valves O N..
from theprior Fuel, to closur Handling pl WhTilfDIs dis; farVassumed to W transported to the control room
) anvelope air incake by the atmospheric (meteorological) conditions existing at the time. These conditions are estimated using the methods of Reference 6.4-i 4. .The atmosphere dispersion factors for each case can be found in Table 6.4- -
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The inhalation thyroid. dose and the semi-infinite cloud gamma and beta doses
' are calculated using the time-integrated concentrations in each area and the occupancy factors noted in Table 6.4-2. The semi-infinite cloud model remains appropriate only for_ the beta dose, due to the ahort range of beta particles.
The,, semi-infinite cloud gamma dose calculated is divided by a geomeeric factor which converts . the semi-infinite gamma dose to a finite dose (Ref. 6.4-4).
This factor is given as:
GF -
Vo.33s ,
where:
l- V - volume of region of interut, fc3 The resulting doses to control room. personnel are given in Table 6.4-2.
t j The calculated thyroid dose total is less than the design limit of 30 roentgen j equivalent man (rem), as is the skin beta dose total. The total whole-body 4 N gamma dose is less than the design limit of 5 rem. Thus the control room
, envelope HVAC System design meets the dose requirements of CDC 19 of 10CFR50g Appendix A.
6.4.4.2 Toxie cas Protection. The general guidance contained in RC 1.78, has been considered in the design of the control room envelope HVAC system, as described in Section 9.4.1. ~ g Toxic gases which are handled onsite are kept to a minimum. During normal at operation the Training amall amounts facility. of chlorine he amount are handled of chlorine (<300within lbs)the willsite notboundary impact ! the c
, control room envelope. A detailed evaluation of potential hazardous chemical accidents and their impact on control room habitability is provided in Section i .2.2.3.
p 6.4.5 Testing and Inspection -
Systems and their components -listed in Section 6.4.4 above, which maintain 4 control room envelope habicability are subjected to documented preoperational g if . testing and inservice surveillance to ensure continued integrity. The tests -
- conducted verify the following for both normal and emergency conditions. '
\ J g 6.4-7 Revision 3
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37 PEGS 1RSAR TABLE 6.4-2 CONTROL ROOM DOSE ANALYSIS ,
j Assumptiong -
e i containment leakage assumptions c.3t (0-24 hrs)
, (Based on a containment free 0.15% (1-30 days)- .
I volume of 3.41 x 10' ft') l
! msr system i - -
.,2 = 1 .. in: = := . u _ ;,= _/ cense ;
i P ess sa makeup air inflow par meters:
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' flow rate 2,200 ft*/ min lC I
-filter efficiency
- ss.5% inorganic, 98.5% organic, l
' 99% particulate ,
control room anvelope clean-up air (recirculation)
I prameters :
C#
lM filtered flow rate 9,500 fts/ min (recirculation air) filter efficiency 956 inorganic, 35% organic, i h i 936 particulate envelope free volume 274,0s0 ft*
envelope unfiltered inleakage 10 fts/ min Meteorological dispersion factors
'[ . (including wind speed and
. direction allowances):
Containment ESF 1.eakage and .
j 1,eakage Case Purge Case
, 0-8 hours
- 1.06x10*8 sec/ms 1.29x10** sec/ms i!2' 8-24 hours '*1. 03 x10-4 sec/as 8.55x10*8 sec/ms 1-4 days 4.45x10** sec/ms 5.42x10s ,,ef,e j l l- 4-30 days 1.91x10-' sec/m*- 2.31x10*8 sec/ms ff l
- 1755 cfm is II.itered through makeup and recirculation filters 235 cfm is
, ff.itered thevogh makeup filters only. Effective filter efficiency for
. Jooo cin is given above.
- t' -
S.4-12 Revision 5 CHANGE NOUCE NN --
- l .
V
.I* TABLE 6.4-2 (Continued)
COffrROL ROOM DOSE ANALYSIS l Assumotions (Cont *d) l Occupancy assumptions: ,
I 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l control room 200% i p 1-4 days, l control room envelope 60%
4-30 days',
Breathing rate of operator 3.47 x 10** m* /sec ,
l l Rs.sults .
- Whole-Body 5 kin I operator dose, 0-30 day period (rem): Thyroid Camma Beta Containment leakage . 21.03 1.69 21.52
~-
N
! J
!3 Cd l C9 ' gy.
,g g ng . 6. x10 ar.sexA -
Direct dose from containment --- 0.07 --- ID
{(b Direct dose from cloud of --- 0.67 ----
i released fission products j Iodine filter loading -- 2.72x10** ---
22,t,7 Total .,.25 2.43 21.52 l C-M (q'fD d
b 4 2 E s
t i ,
8.4-13 Revision 5
' CHANGE t10TICE N
l t .
'STPEGS UFSAR h TABLE 7.A.II.B.2-2 8 POST-AccfDENT RADIATION LEVELM/ DOSES
.li
) ,
. Continuous Occupancy Areas: 30 day Doses (Rem)
- Camma Beta "Thvrei'd
- 2I .'6 c y- CN Control Room 2.43 Ifil 21.50 ;;.7^
l gqq j ,
4
- Technical Support Center 4.85 24.55 28.s2 l 1
l Infrequent Access Areas
- UFSAR Figure Dome Rate (R/Hr1 Reference Area Time after accident l 1 hr i day 1 wk i month
- 1 12.3.1-36 Post-accident 75 4.5 x 10-8 1.1 x lo-a 6 x 1C** -
p g sample station '
t 12.3.1-27 Esalth Physics 4 x 10 s 3,g x go.4 3 x go.s 4,3 x go.4 counting room 12.3.1-27 Radwaste count- 3.1 x 10*8 1.8 x 108 4 . 6 x 10** 2.4 x 10**
ing room i
12.3.1-28 Plant vent 4.74 .J8 7.1 x 10** 3.8 x 10**
radiation
- monitor ,
,' 12.3.1-25 Auxiliary shut- 8 x 1 0** 4. 8 x 10** 4.2 x 10** 6.4 x 10*'
down panel t
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l STFECS UFSAR value once a DF of 100 is reached and no additional credit is taken for deposition after a DF of 200 is reached. For particulate iodine, a spray 3
removal reduced to 10% rateofofthis6.9 value hr*1 untilisa assumed DF of 1000 isuntil reached. a DF of 50 is reached 15.6.5.3.1.3 containment hakane Domes - Doses resulting from activity leakage from the Containment have been calculated using the models presented
! in Appendix 15.B. The thyroid, whole. body gamma and skin beta doses are presented in Table 15.6-11 for the EZB distance of 1,430 meters and the outer boundary of the LPZ at 4,800 meters.
- 15.6.5.3.2 ESF Leakare Contribution: A potential source of fission product leakage following a thCA is the leakage from Engineered Safety Features (EST) components which are located in the Fuel Handling Suilding l
.(FHB). His leakage may be postulated to occur during the recirculation phase ;
- 6 for long-term core cooling and Containment cooling by sprays. The water
- i contained in the Containment sumps is used after the injection phase and is ;
- recirculated by the ECCS pumps and the Containmye pumps.
- - Y i
- m .~-- -
- -
l
- ' A itional potential source of fission product leakage from ESF com l, is via va .2 'amka e in isolation valves in the Low y injection
- pump' recirculation lines, dS c on pump recirculation 3
'- lines, the containment Spr est . e Refueling Water i'
Storage Tank et on line to the RUST. The RUST s he Auxiliary Building atuosphere.
!I
, ca ~ ~
W. _ -]
l 15.6.5.3.2.1 Fission Product source Term - Since most of the l
- i radiciodine released during the LOCA would be retained by the Containment sump l ij water, due to operation of the CSS and the ECCS, it is conservatively assumed 1
~
that 50 percent of the core iodine inventory is introduced to the sump water ~
to be recirculated through the external piping systems. -
I
- Because noble gases are assumed to be available for leakage from the Containment atmosphere and are not readily entrained in water, the noble gases are not assumed to be part of the source term for this contribution to the ,
] total LOCA dose. -
l l 15.6.5.3.2.2 hakare Assumntions - The amount of water in the
[
Containment sumps at the start of recirculation is the total of the RCS vater and the water added due to operation of the engineered safeguards, i.e., the ECCS and CSS. This amount has been calculated to be 512,494 gallons. This value is conservatively low to maximize iodine concentration in the sump water. . L,.
l ESF leakare into the FHB
~ . \ e')
"The ECCS recirculation piping and components external to the Containment are
. designed in accordance with applicable codes and are described in Section 6.3.
h e CSS is described in Sections 6.2.2 and 6.5.2.
i aO 1
15.6-15 Revision p,w ,
STFECS UFSAR
,[-s) The maximum potential recirculation loop leakage is tabulated in Table ',
" 15.6-12. Each recirculation subsystem includes a high. head safety injection I (HHSI) pump, a low. head safety injection (UISI) pump, a residual. heat l axchanger, the Containment sump, and associated piping and valves. Thus three r separate subsystems are provided for recirculation, as well as for injection, auch of which la adequate for long-tern cooling. -
l Since three redundant subsystess are available during recirculation, leakage for any component in any subsystem can" be terminated by shutting down the IRSI and IDISI pump associated with that subsystem and by closing the appropriate pump suction and discharge, isolation valves. .
l Maximum potential recirculation leakages are indicated in Table 15.6-12. The I i leakage rate assumed for desa calculation purposes is conservatively twice the l laakage rate given in Table 15.6 12.
The iodine partition factor applicable for this leakage is assumed to be 0.1. l m~~ _ -
j\SF 14akere into the RUST f
The p ential leakage from ,the valves given in Table 15.6-12 was useti to ,I calcula the maximum potential ESF leakage into the RWST. The total akage ( 1 from these alves is ten times their total design leak rate. 14aka- of /
fission produ s past these valves is assumed to begin upon recir lation and.
\
l continues at th - same rate for the duration of the accident. e g
. t, E Transit times for th ission products to travel from the sa r.ing valves to the RUST were calculate by using the valve leak rate the volume of water g in the piping to be disp 1 ed. The isolation valves n the Containment Spray System (CSS) test lines are sically closest to e RUST and leakage from }
{- these valves are conservative 1 alculated to .ch the RUST 13.4' days '
following an accident. Once the akage fro the CSS test line isolation i
valves reaches the RWST, the combine lea ate from the IRSI and HHSI y i recirculation isolation valves and the ~ S test line isolation valves is j' q assumed to be flowing into the RUST. .
Due to the large diameter pipin involved, lea e from the RWST suction 3i l
ll isolation valves would not r ch the RUST before e motive force, i.e. the l
.i pressure in the RCB, is ne . igible. Therefore, lea e from these valves is I not considered in the r iological analysis.
Credit is taken fo radioactive decay during the ECCS injec .on phase and the a piping transit me to the RUST. No credit is taken for dilu on in either h the RUST or i the atmosphere of the Mechanical Auxiliary Build (HAB) or for the ho up time of fission products in the atmosphere of the Once the leaV e flow reaches the RUST, it is assumed to be immediately d ersed into e environment.
e iodine partition factor applicable for this leakage is assumed to be 0.1.
I V
15.6.5.3.2.3 ESFIsakereDo$ti-Theidineactivity,oncsreleasedto
-the -etuosphere of the FHB, is assumed to be quickly transported by the (
ventilation system through the exhaust filters and released to the environment g.
f' *
.at ground level. The iodine filtration efficiency is assumed to be 95 percent.
15.6-16 Revision /[
l l
j, _.-
l
lh' STPECS UFSAR Chhc I
i -- 2 " ; D 2 ;F's J 3 9i . . "1..
.. 9... S N. ;' "; C';
- U ;; = 5 0; i; l' NForW the first 30 minutes, the analyses consider the potential imbalance of I
Eexhaust flow, resulting in flow through one train of heaters being below the 4 setpoint for heater energization (12,000 cfm through the three filter banks). 2\
During this time, the filter efficiency is assumed to be 90 percent for
! elemental iodine and 30 percent for organic Lodina for that train of filters.
I Within 30. minutes, operator action to isolate that train of filters occurs. -
T The offsite doses due to the recirculation leakage are presented in Table
- 15.6-11 for the EZB of 1,430 meters for the initial two hour period and the ,
LPZ outer boundary distance of 4,800 meters for the 30-day duration of the ,
f" accident.
f -
! 15.6.5.3.3 coneminment Pure contribution: In the event of a 1DCA
- t.
coincident with the containment supplementary purge system in operation, the l
purge is assumed to be isolated within 23 seconds following LOCA initiation. _
During normal power _ operation, the Containment supplementary purge system
' vents the containment at 5,000 fc3/ min. However, for this analysis., the i maximum flow rate due to the pressure spike inside the Containment was used '
(88,900 ft3/ min for each purge line, intake and exhaust). The Containment I l2-purge system is described in Section 9.4 ,
l t
The containment airborne iodine " inventory available for release is assumed to .
be the flashed portion of the total primary coolant iodine inventory based upon a preexisting iodine spike level of 60 pCi/g dose equival'ont I 131. For noble gases,100 percent of the primary coolant inventory based upon 1 perecnt failed fuel is assumed to be available for release. No failed fuel is assumed i since isolation occurs prior to the core reaching a temperature which could cause a fuel failure. ,
15.6.5.3.3.1 Containment Puree Doses - The offsite doses calculated due to containment purging are presented in Table 15.6-11 for the EZB of 1,*430 ~
j meters and 1.PZ outer boundary distance of 4,800 meters.
, 15.6.5.4 Core and Svstem Performance.
S 15.6.5.4.1 Mathematical Model: The requirements of an acceptable ECCS
- Q Evaluation Model are presented in Appendix K of 10CFR50 (Ref.15.6-2).
larme Break 1ACA Evaluation Model
(.j t h The analysis of a large break 1DCA transient is divided into three phases:
blowdown, refill, and reflood. There are three distinct transient.s analyzed in each phase, including the thermal hydraulic transient in the RCS, the .
- l , .
pressure and temperature transi.snt within the Containment, and the fuel and 4 clad temperature transient of one hottest fuel rod in the core. Based upon these considerations, a systen of inter-related computer codes has been developed for the analysis of the 1DCA.
y- .
he description of the various aspects of the IDCA analysis methodology is L'
j ,
given in UCAP-5339 (Ref.15.6-4). This document describes the major phenomena 4' codes which ensure compliance with the acceptance criteria. The SATAN-VI, q URETID0D, COCO, and LDCTA IV codes, which are used in the IDCA analysis, are 15.6 17 Revision L- .
h l'
- I STFECS Ur$AR 1
1 O..e s TABLE 15.6-10 (continued) l PARAMETERS USED IN ANALYSIS OF LDSS-OF-COOLANT ACCIDENT OFFSITE DOSES .
r
. Parameter
~
' Activity assumed mixed in Containment
- aump water available for ESF leakage f None t
Wobis gases Zodines 50% core activity j Table 15.A-1 S--
ESF system leakage rate assumed into 'Twice that of .
i the FHB, cm?/hr Table 15.6-12 i l
ESF system leakage into the RVST Table 15.6,12
- assumed, em?/ hr
~
' T. :::: :::: c::_ 1::ht ; ur:. re.::,
. ::: i::11:i : :1.:: t; th "":: 13.' .y.
-Ti.cai; ;i;; f ;; 1::h!=; "'JfT cuttien
'*n >---%f! , i::12:!:r ":1":: :: ': ""?T 6
. p - - ,h oun cf water in which mixing of ,
g
iodine occurs, gal 512,494 j ' Iodine partition factor for leakage 0.1 95 THB filtration efficiency, t Supplementary purge rate, scfm 88,900 2
,j (for each of two lines, intake and exhaust) ,
+ Time before isolation of purge, see 23 Meteorology 5 percentile Table 15.B-1
- Dose model Appendix 15.8 l
l, 1 -
- t
,O 's tV it L 15.6-36 Revision [h j m r-
7.__,____ l l
.f STPEGS UFSAR
> TABLE 15.6-11
. b' ) DOSE RESULTING FROM LARCE BREAK thSS-OF-COOLANT ACCIDENT l
IAlfalLImRI
- ContalTrment Leakace Doses Exclusion Zone Boundary 0-2 hr thyroid, rems 1.173 x 10' f whole body ganssa, rems 2.26
- skin beta, reos 1.21 g
Low Population Zone 0-30 days j thyroid, rems 63.54 , L I, whole body gassaa, rems 7.3 x 10** skin beta, rems 4. 7 x 10** g TSF Leakace Doses (from leakace into the FHB)
. l j 2xclusion Zone Soundary 0-2 hr .
l thyroid, rems 2.54 x 10-8 i 9 whole body gastaa, rems 8.03 x 10** skin beta, rems 2.27 x 10**
'. Iow Population Zone 0-30 days thyroid, rems 3.75 x 108 l 3.93 x 20** i l[( .- whole body gamma, rems skin beta, A 1. 4 0 x 10**
e W _ o y E E Leakaee Doses (from leakaae into the RWST) 'f one soundary, 0-2 hr Exclus thyroi , ms 0.0 ,) whole body ga rems 0.0 g' gegg g 0.0 skin beta, rems f Low Population Eo , -30 days ',' thyr , rems .14 e body gamma, rems ,3 . 0-8 skin beta, rems *1.30 x 10 j containment Purtrine Doses 5 Exclusion Zone Boundary 0-2 hr thyroid, rems 19.42 . , whole body gansna, rems 1.0 x 10*' skin beta, rems 7.6 x 10-8 24w Population Zone 0-30 days 2,39 D,td f
- thyroid, rems .40- i gg x 10*8 1.2 l J
whole body gansna, rems skin Leta, rems 3.30 x 10** 5O
- V t
f 15.6-37 Revision 5 l' GVJ:G2 ij0TICE Ok I I l v- .l
lf STPEGS UFSAR
- / .
7ABLE 15.6-11 (Continued) =
'Q/
DOSE RESULTING FROM LARGE BR'E LOSS-OF-COOLANT ACCIDE?TP
*fotal Domes ,
Exclusion sone Boundary 0-2 hr thyroid, rems 2.37 x 20' ', I whole body gasuna, reos 2.27 skin beta, rems 2.22
. Low Population zone 0-30 days 4, . (s 3 04 I thyroid, rems 4-tHP-x 108 . ;j if t$6 l whole body gamma, rems .
j skin beta, rems 0.47 l 6 ) 4 e . l 1 l 1
- l 1
t 1 l l
*I
- F ll - .
, l 8 l i I
- i '
l l li 1 i f.
- Exclusion zone soundary is at 1,430 m. Outer boundary of Low Population Zone y ,t is at 4,soo m.
,s( ik * . ^k 15.6-38 Revision 5
- I =w m EUGDCE Myy= -- '
t r
~
ll- lt
'87FECS UFSAR
;i TABLE 15.6 12 HAXIMUM POTENTIAL RECTRCUIATION IDOP LEAKACE EXTERNAL TO CONTAINMENT 5
- M, i Imakage into the FHB -
3 , l. Imakage . E . Item (em3 /hr)
'l
. low-Head Safety 30 i f ,
- Injeetion pumps ;
I High-Head Safety 60 Injection pumps i, Valves 4,050 .
- 1l Total 4,140 laakage into the RWST ,
1
' Item l
Low Head Safet njection pump ( - recirculation isolation valves e
-g >) C,Q High Head Safety Injec on C IN i pump recirculat line .
isolation va s
- r Contai spray system
-< )N .
.i. * ' !- ! ce ine isolation valves , otal leakage 1,740 cad /hr E l m i
- I .
a - i e 15.6 39 , . Revision fh -p . 't i v
._. w. . = .: ...:.. z. - -
4 4 _ STFECS UFSAR f APPENDIX 15.B
;l\
DOSE MODELS . r E f ,
, This appendix describes the mathematical models and parameters used for the
- fission product transport from the postulated accident site to the environment and for the radiological dose calculations. ,
15.B.1 Ceneral Accident Parameters This section describes the parameters used in analyzing the radiological l' consequences of postulated accidents. The site-specific, 5-percentile, short-term dispersion factors for the worst sector (assuming ground level releases) are given in Table 15.B-1.z -(See Section 2.3.4 for additional details on l meteorology.) The breathing rates used are presented in Table 15.B-2. The ' 4 . thyroid (via inhalation pathway), beta skin, and gamma body (via submersion . . pathway) dose factors based upon Reference 15.5-3 are given~ in Table 15.B-3. ;
- i. '15..B 2 offsite Radiological consequences calculational Models l This section presents the models .and equations used for calculating the integrated activity released to the environment, the accident flow paths, and the equations for dose calculations. Two major release models are considered:
- A single holdup system with no internal cleanup .,
i
- A holdup system wherein a two-region spray mMel is used for internal
\
cleanup 15.B.2.1 Accident Release Pathways'. The release pathways for the major , accidents are given in Figure 15.B 2. The accidents and their pithways are as ! follows: ' t l
~
! 1. Loss-of Coolant Accident (1DCA) l' Immed!ately following a postulated IDCA, the release of radioactitity from the containment is to the environment with the containment. spray , and Engineered Safety Features (ESP) systems in full operation. The release in this case is calculated using Equations 15.B.2-6 and 15.B.2-7, which take into account a two-region spray model within the Containment. The release of radioactivity to the environment due to assumed ESF system leakages in the Fuel Handling Building (FHB) will be l via ESF filters and is calculated using Equation 15.B.2-5. ' e- --- S.: ::::::: :f redi;;;;.h i;.j ;. ;L m..i... - L - i ..._J ER ( 1.ek g. i..;e C..e L f.,li. *.'. L.; 0 ;...... T...k 'nOT) 1
; die;ely '
di:p;;;;e iat: th; ;;c.1. e. ..,;. ...? ie ,..le l.;e 4 i.. 4..;.1er. 3 g ! N --E . O . 2 4. ':..m.... ... ... J1;. le ; L..fe. h; h:1 dup ti : :f fierir 7 Tre4=;.; i; th; em;;pher; ef the " ch .ieel .'.wiliary 0 11d17.; '"""). ) . O . 15.5-1 Revision / I V T 9 7 P" "' 1 e m:- - P" - - - -
p ,- l l i ll- ~ HOLD UP IN
- . CONTAINMENT A LOCA. FHA IN CONTAINMENT -
AND RCCAE
- g
- l *
~
STEAM YENT AND ' u SAFETY VALVES OR
' HECHANICAL AND
- ELECTRI L XILIARY SGTReHSLB.WGSF.AND RCCAE CONTROL l -
RDOM INO HOLO UPI r ATHOSPHERE E 8 - A' - D .. [ - FILTERS !
, FHA IN FUEL FUEL HANDLING y l HANDL(NG 'l BUILDING BUILDING '
i 8 C . p- ~ Y Y V f' . mo ESF LEAKA * - I ( INT WST
)
D
, Y ~
1 n J i i. l
~ '
SOUTH TEXAS PROJECT q UNITS 1 & 2 t , i 44E1XASE PATWAYS 1 F10LME 152 2 MIS *f hl i man t
). ,
o .suma V
-________________m
l l SOUTII TEXAS PROJECT UNhTS I & 2
' Sheet 1 i ,
HOUSTON LIGIITING & POWER CAfffff ATION COVER SIIRFT
.p s irna "TuoN Ntn4aan. NC 9004 -
I e'a' ^110N Tm.2: Pod LOCA Radiaden Zemes and EQ Radiation I Sunnicr.
- . Determlandos er Post LOCA Radiation Zemes and EQ Radiation BurDOWAmaA/ SYSTEMS: Tafices
,( , DISCIPLINE:
- 4. NUCLEAR iI ., .
- g QUAIRY CLAS$.
ii . 4 (Safety Related) fI UNTE .I 9 (Units 1 & 2) ~ i CALi24.A11oM STATUS: Flaal j} OsmCTIVE:
,l To determIce the post accident (LOCA) radiation zones and doses for the
!! plant. These doses will be used for equipment quellDeadon $Q)
! porposes (For EQ Design Criteria see TPNS # 4E019NQIt09, and li UFSAR Chapter 3,Section 3J1).
!' { Score: . t The calculation is applicable to Units 1 & 2. 8 l RasuLts:
- l ii The results are given on sheets 9 22.
' l
. {b TDrAL Nuwsca OFSHEETS: 979
- . i l
., sev aa. 9 l
i6 pumma
- f w. x swee. , -
ia
'Jhfl 1/"/YU -
asva.u d as \ Ie M /alMt' l! swenvuee Escossa I'N E E5 ' a saul,a4 C.or t' cemenener %fPo, v l1 i i :-- 10.4sfb s e M a a y' 1
/
r a. . . . . . _. __ Sheet 1
- COUTH TEXAS PROJECT UNITS 1 & 2 COUSTON LIGHTING & POWER ;
CALCULATION COVER MIIITT
- O'a= Anon Nuedass: 1C4013 ne n' ancH 'nna. <entrol Room,TSC and Offdte LOCA Radiation Doses _
Suomen Calestation of Control Room,TSC and Offsite Doses during a LOCA -
, PBunDoc/ARENSYSTDas:
- N/A <
l o ,m s , c :71UCLEAR *
- l
- W CLA3s. t -
ad (Safety Related) s .
. 3 .+-
- 4, r- t1 -
t 32 , , l
. . UNm *
'9 (Units I & 2) .
e+Y~.v. ' $ Finel
-CALQJLADoN STA'rus: 5
- t ;
t *
-j Osacnys: To determise the Control Room,TSC and Offstte Doses during a LOCA.
3 I) Revision 9 removes the RWST backleakage dose contribution. CsJculation MC4458 l , (Ref.68) shows that the RCS backleakage to the RWST will not contribute to the ) LOCA doses at 30 days. The minimum time for the backleakage to reach the RWST is 42 days. j!
- .t
, a4 Scort 'The calculation is applicable to Units 1 & 2. REsuLTs: The results for shutoff of the CSS at approximately 6.3 hoars are given on pages 9 4
. und 10.
A j The doses for shutoff of the CSS at 30 an 45 minutes after the start of the accident
- l are given on page M 22 and M 23. These are not the certest design bads dose. '
TOTAL Nthasta or SMEET3: 1155
=
l l s., ne. 9
% w.w.a %
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