ML20078B928
| ML20078B928 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/09/1995 |
| From: | ERIN ENGINEERING & RESEARCH, INC. |
| To: | |
| Shared Package | |
| ML20078A087 | List: |
| References | |
| C122-93-01-01, C122-93-01-01-R05, C122-93-1-1, C122-93-1-1-R5, NUDOCS 9501260100 | |
| Download: ML20078B928 (73) | |
Text
CALCULATION TITLE PAGE g ge Sheet 1 of 60 Project No: 122-93-01 Plant Name: Cooper Nuclear Station Calc No: C122-93-01-01 Client Name : NPPD
Subject:
CNS Control Room Ooerator Thyroid Dose Calculation Computer Standard Computer Program Program No(s). Version / Release No.
Program g YES O NO PADD 1.10 RECORD OF ISSUES Total Last No. of Sheet Rev. Description Sheets No. Orig. Ckd. App. Date 0 InitialIssue See Page 2 /s/ /s/ /s/ 4/26/94 RLD MJT 1AB 1 Credit for mixing eliminated, new source See Page 2 /s/ /s/ /s/ 7/5/94 terms. RLD MJT DET for LAB 2 Additional sensitivity cases included. See Page 2 /s/ /s/ /s/ 7/6/94 RLD HUT MJT for LAB 3 Additional sensitivity cases included. See Page 2 /s/ /s/ /s/ 7/12/94 RLD MJT LAB 4 Run SENSCRIN replaced by three new See Page 2 /s/ /s/ /s/ 7/14/94 sensitivity cases. RLD MJT LAB 5 Assumptions revised. New cases included. See Page 2 NOI N9Ih '[T[6 t REMARKS This calculation evaluates the adequacy of the Cooper Nuclear Station Control Room Emergency Ventilation System in mitigating operator thyroid doses following a design basis LOCA or a refueling accident. j l
Revision 5 includes revisions to pages 1 through 50, and page B-1. Pages 51 through 60 have been added, and Pages B-2 through B-16 have been deleted.
~ V501260100 950119 PDR ADOCK 05000298 P PDR
CALCULATIEN SHEET l
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TABLE OF CONTENTS Section Paae 1.0 PURPOSE............................................... 3 2.0 SU M M ARY O F R ES U LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l 2.1
SUMMARY
OF BASE CASE RESULTS . . . . . . . . . . . . . . . . . . . . . . 4 2.2 C O N C LU S I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l
3.0 AS S U M PTI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.1 LOSS OF COOLANT ACCIDENT BASES AND ASSUMPTIONS . . . . 6 3.2 REFUELING ACCIDENT BASES AND ASSUMPTIONS .......... 9 4.0 M ETH O D O LO GY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 )
4.1 GENERAL
SUMMARY
OF THE LOCA SCENARIO . . . . . . . . . . . . . . 10 .
4.2 MAIN STEAM ISOLATION VALVES . . . . . . . . . . . . . . . . . . . . . . . . 14 1 1
4.3 OTHER LOCA SCENARIO CONSIDERATIONS . . . . . . . . . . . . . . . . 15 4.4 GENERAL
SUMMARY
OF THE REFUELING ACCIDENT l S C E N AR I O . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 l 4.5 CONTROL ROOM ACTIVITY CALCULATION EQUATIONS . . . . . . . 20 4.6 PAD D I N P UT DATA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 5.0 D ES I G N I N P UTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 b 6.0 R E FEREN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 7.0 N O M EN C LATU R E . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49
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8.0 ~
CALC U LATI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 8.1 CALCULATION ' OF 'X/O FACTORS FOR TURBINE BUILDING
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RELEASE .......................................... 54
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~ ' 8. 2 ' ~ R ES U LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 APPENDIX Total Paaes Appendix A - CNS MSIV Leakage Path Assessment 8 + 4 pages of attachments Appendix B - PADD Output from Case Studies See Page B-1 W12293015630-010695 5 M i/1/4f %/) 'T //9/9$"l X)B NO. 122-93-01 PAGE l CALC tn C122-93-01-01 2
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CALCULATION SHEET 1.0 PURPOSE The purpose of this calculation is to update the assumptions, variables and analytical techniques used to calculate the control room operator thyroid, whole body, and beta skin doses following a design basis LOCA and a design basis fuel handling accident.
Reference 2 provides the current basis for control room doses due to intake and inleakage of contaminated air. References 1 and 22 provide evaluations of the analysis included in Reference 2. The evaluations identify inconsistencies between plant licensing documentation and the technical bases of the calculations. Recommendations were made in these references to correct inconsistencies and update the analysis through improved analytical techniques. This calculation addresses these issues and provides the l 1
updated basis for control room doses, including the incorporation of the Post Accident ;
Design Dose (PADD) computer program.
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CALCULATION SHEET 2.0
SUMMARY
OF RESULTS 2.1
SUMMARY
OF BASE CASE RESULTS Control room habitability design basis doses are shown as PADD results in Section 8.0 of this calculation. The design basis or base dose results are presented in this section.
Other sensitivity analysis case study results are also presented in Section 8.0. The base case control room operator thyroid, whole body gamma, and beta skin doses were found to be below SRP 6.4 criteria. However, this in itself does not indicate acceptability of the whole body dose with respect to the SRP 6.4 criterion. Whole body doses due to other sources are not evaluated in this calculation. The following table summarizes the results of this analysis given a control room filtered intake flow rate of 1000 CFM and a control room unfiltered inleakage rate of 100 CFM. The refueling case also assumes a 90 second delay in standby gas treatment system actuation.
O Event Dose Type C ieniatea Result (rem) sRe 6.4 Criterion Thyroid 3.91 5 30 i
- I LOCA Gamma Beta skin 5.50E-01 s 30 _
i Thyroid 1.69 5 30-REFUEUNG Whoie Body 2.67E-01 d5 ACCIDENT Gamma Beta Skin 2.53 s 30 W12293015630410695
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2.2 CONCLUSION
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The Control Room HVAC system at Cooper Nuclear Station (CNS) can meet the Standard i l
Review Plan 6.4 [ Reference 34) dose limit of 30 rem thyroid for 30 days following either the design basis loss of coolant accident (LOCA) or the design basis refueling accident described in the CNS Updated Safety Analysis Report (USAR). The whole body doses calculated for CNS are less than the SRP 6.4 criterion of 5 rem. However, as indicated :
above, additional contributors to the whole body dose must be evaluated to demonstrate l compliance with this criterion. The beta skin doses to the operators are also less than l the dose limit given above. !
This control room dose analysis is deemed to have no impact to existing mitigation actions or protection provided by the control room HVAC system for smoke (internal plant ,
or external plant fire) or hazardous chemical releases as related to control room habitability.
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- CALCULATl!N SHEET 3.0 ASSUMPTIONS f
3.1 LOSS OF COOLANT ACCIDENT BASES AND ASSUMPTIONS General assumptions for the control room operator thyroid and whole body doses calculaticas for the design basis loss of coolant accident (LOCA) are presented as follows:
- 1. The reactor has operated for 1000 days at 2381 MWt.
- 2. Leakage from the primary containment to the reactor building immediately flows through the standby gas treatment system and the stack without mixing in the secondary containment building. However, it is possible to take credit for some mixing (usually :s50%) within the Reactor Building to reduce the concentration :
leaving the Reactor Building given prior NRC approval.
- 3. For normal leakage of radioisotopes from the reactor building, the meteorology assumptions applied in the original CNS disperson coefficient calcula1on (Reference 42) are applied herein. For the dose contribution due to MSIV leakage, h atmospheric dispersion coefficients were calculated based on a release from the turbine building. This derivation is described in Section 8.1.
- 4. All halogens (iodine and bromine) are assumed to function identically in terms of plateout, filtration, and adsorption.
- 5. Transport times between the release point and the control room are conservatively assumed as zero.
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- 6. Mixing in the primary containment and control room are assumed to be l instantaneous and perfect. I
- 7. The Standby Gas Treatment System (SGTS) filter efficiency is, per CNS Technical Specification 3.7.B.2:
99% for elemental halogens 99% for particulate halogens 99% for organic halogens W1229301563CW10695
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CALCULATION SHEET j As recommended in Reg. Guide 1.52, Rev. 2, and Reference 22, a more ;
I conservative value of 95% is used in this calculation for these efficiencies.
- 8. The control room emergency bypass filter efficiency is, per CNS Technical Specif; cation 3.12.A.2:
i 99% for elemental halogens 99% for particulate halogens 99% for organic halogens As recommended in Reference 22, a more consentative value of 90% is used in this calculation for these efficiencies. This includes a safety factor of 10 to account for degradation between tests. The CNS control room emergency bypass filter does not have humidity control and this factor of safety is consistent with that approved for use at other nuclear power plants with a similar configuration i
(Reference 41). 1
- 9. No credit is taken for plateout of halogens other than that assumed in Regulatory Guide 1.3. It is noted that, if credited, plateout effects would only be applicable to 1 elemental halogens.
- 10. 100 CFM of unfiltered inleakage into the control room is assumed. This is be%c ;
on results from test STP 94-199 [ Reference 50] (45_CFM 26 CFM givw , o y)
CFM filtered iiuw). i
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- 11. For all _ cases, a one minute delay between the time of initial radioactivity release i to the atmosphere and actuation of the control room ventilation emergency bypass mode of operation is assumed. This delay is based on the time which would elapse before the source term is detected at the control room intake and the time
~ rsq0 ires to act0 ate the control room emergency ventilation system and align to the .
filtration mode. There is no operator action or design feature which would preempt !
this automatic actuation to allow earlier actuation, as is the case with the SGTS.
~ This~ assumption conservatively increases the source term introduced into the control room envelope during this time per'od.
- 12. Upon detection of a low reactor water level, there is an approximate one minute '
time delay (63 seconds) for realignment from normal containment ventilation to SGTS ventilation. However, as described in NUREG-1465 [ Reference 44], actual fuel damage occurs several hours after the start of an event. As tFis time is well beyond the time required for actuation of the SGTS, it is reasonable to neglect any small delay in SGTS actuation. Therefore, no delay is assumed for the LOCA analyses.
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CALCULATION SHEET
- 13. Per Reg. Guide 1.3, leakage from primary to secondary containment is assumed
' to pass directly to the SGTS, where it is then discharged to the environment via the elevated release point (ERP). l
- 14. The control building essential HVAC system has ducting that passes through the CNS control room envelope. This ducting has a source term associated with the ,
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noble gas inventory contained therein. However, it displaces an equal quantity of l the available source term in the control room and, due to the lack of build-up early i in the LOCA event, would likely have a lower concentration of noble gases than the control room. Therefore, it is already accounted for and need not be considered further. In addition, any unfiltered inleakage from this ducting is assumed to be l
included in the unfiltered flow rates assumed.
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- 15. MSIV leakage is assumed to be held up in the main steam piping for a significant time during transport to the main condenser. A consentative dose reduction factor of 10 is used to account for plateout in the main steam system and condenser (see basis in Appendix A). - The radioisotopes are assumed to leak from the-condenser at a rate of .5% per day. This leak rate is consistent with the control rod drop dose analysis in Chapter XIV, Section 6.2 of the CNS USAR. The leakage is assumed to immediately pass out of the turbine building to the control room.
- 16. Per DG 94-102 [ Reference 51), a cooling flow of 240 CFM 20% is assumed to pas 1 through an inactive SGTS train under accident conditions. For conservatism, a vejue of 240 CFM + 20%, or 288 CFM, is used. This portion of SGTS flow is filte ed at efficiencies of 90% for elemental halogens and 30% for organic halogens.
Tha remaining SGTS flow (1492 CFM) is assumed to pass through the operating ;
SGTS train and be filtered as described in item 7 above. ;
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- 17. - An ECCS leakage contribution of.1000 cc/ minute is considered in the post-LOCA _
dose evaluation . Reference 48 specifies that 10% of the fluid released will flash to vapor and be available for rele'ase via the SGTS. This assumptiorils' conservative, . .
because the liquid is subcooled.l In addition, historically observed ECCS leakage has b'een minimal. Watkdown;s performe'd in 1980 for NUREG-0578' Iteni 2.l1.6.a,' ,
estimated leakage to be 2 cc/ minute. Since 1980, leakage has not noticeably. l increased and is currently estimated to be' minimal. CNS procedures are being i revised to monitor ECCS leakage and ensure it is maintairied below i
1000 cc/ minute. , . _ _ _ _ ._
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CALCULATION SHEET 3.2 REFUELING ACCIDENT BASES AND ASSUMPTIONS The bases and assumptions for the LOCA also apply to the refueling accident with additions and modifications taken from the CNS USAR listed in Section 5.0, Design inputs, and the following additional assumptions:
- 1. Noble gas source terms for the refueling accident are derived by multiplying the associated LOCA source terms by an adjustment factor. This factor is the ratio of the number of damaged fuel rods in the refueling accident to the number for a LOCA. Froin Chapter XIV, Section 6.4.2.1 of the CNS USAR,125 fuel rods are assumed to fail as a result of a fuel handling accident. From Section 6.3.4,25%
of the fuel rods in the core are assumed to fail as a result of a LOCA. Given 548 fuel bundles of 63 rods each, a total of 8631 rods are assumed to fail following a LOCA. Therefore, the initial LOCA noble gas concentrations are multiplied by a factor of 1.448E-02 to obtain the initial noble gas concentrations for the refueling accident scenario.
- 2. Halogen source terms are derived in the same way as the noble gas source terms.
However, because the LOCA source term assumed only 25% of the halogens are released, an additional factor of 4 was multiplied by the ratio of failed fuel. Also, h 99% of the halogens released from the rods is retained by the refueling pool water.
Therefore, a partition factor of 100 is used, resulting in a conversion factor for !
I halogens of 5.793E-04.
- 3. It is assumed there is a 90 second time delay to switch from normal containment ventilation to SGTS. ventilation upon detection of a high radiation signal. This is consistent with the reactor building isolation damper closure time of slightly more than one minute (63 seconds). During this time, unfiltered release is assumed to occur, and the reactor building is assumed to remain at a nominal negative pressure.
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CALCULATION SHEET 4.0 METHODOLOGY The CNS control room operator thyroid and whole body gamma doses are determined based on the analysis of impacts from a design basis LOCA and from a design basis ,
refueling accident. These two accidents are assumed to bound all other accidents with respect to inventory released, source term leakage pathways, and radiological consequences to the control room. The following subsections describe the methodologies used to calculate the resulting thyroid and whole body gamma doses for these accidents.
4.1 GENERAL
SUMMARY
OF THE LOCA SCENARIO One of the accidents evaluated to determine the radiation environment for the CNS control room operator thyroid doses is a design basis Loss of Coolant Accident (DBA-LOCA). For CNS the DBA LOCA is defined as a complete circumferential break of one of the reactor recirculation loop lines.
53 The radiation release path for this scenario is from the reactor vessel to the primary containment. Radiation from the primary containment is assumed to leak to the secondary containment, or reactor building, at maximum allowable rates. Once in the reactor building, the radiation is drawn into the Standby Gas Treatment System (SGTS),
treated and released through the stack to the environment.- Once in the environment, the material can disperse and potentially be drawn into the control room ventilation system where it can resu!iin operator exposure. The following is a brief description of the events postulated to occur during a DBA-LOCA. The discussion relates to the source term !
generation and transport of radiation out of containment.
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CALCULATIIN SHEET At time t=0, the postulated pipe break occurs and results in rapid blowdown of the reactor coolant system (RCS). Flashing and escape of the coolant during blowdown removes heat rapidly from the primary system and causes the fuel rod cladding temperature to drop. Consequently, only a few fuel rods are assumed to fail during the blowdown period. Following the end of blowdown, the fuel rods are uncovered and the stored heat in the fuel and the decay heat are transferred to the cladding, thus raising the cladding temperature. Some fuel rods may experience cladding failure during this period.
The Emergency Core Cooling System (ECCS) refills the lower reactor vessel and then refloods the core region within 100 to 300 seconds, causing cladding temperature to decrease. During the initial blowdown, only the radioactive material contained in the coolant from steady-state operation would be released to the containment. During l reflood/ refill, when fuel rod cladding failure may occur, the noble gases would be transported out of the reactor system by steam flow and would become airborne. Some fraction of the iodines and less volatile fission products that are released as a result of fuel rod failure would also be transported out of the reactor coolant system by the steam flow g and become airborne, and some fraction would remain in solution in the reactor water or would be deposited on surfaces within the reactor coolant system components. The amount that becomes airborne outside the reactor coolant system would be strongly dependent on the time of fuel rod failure and the transport phenomenon for each species within the system.
Following the release from the reactor coolant system, the fission products would be distributed within the drywell or primary containment. The released activity would initially be airborne within the drywell. Following initial release to the containment atmosphere, the high pressure in the primary containment would result in leakage out of the primary containment to the secondary containment. In the secondary containment the action of natural convection currents and ESF equipment, such as cooling fans, will cause time-dependent redistribution of the activity within the secondary containment. Natural removal W1229301-5630410695 J 6 % i/9 /1f 77),77-t/1/9 rl JOBNO. 122-93-01 PAGE l CALC NO.C122-93-01-01 11
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CALCULATION SHEET processes, such as deposition on containment surfaces, would reduce the airborne activity concentration and would redistribute a portion of this activity to the containment surfaces.
During the same period of time, leakage of radioactivity from the secondary containment to the atmosphere could take place. This would be processed by the standby gas treatment system (SGTS) filters, causing a buildup of activity on these filters. In addition, there could be some deposition and plateout of radioactivity (iodine and daughters of noble gases) on surfaces of ductwork or on the walls of secondary containment. Figure 4-1 shows the evaluated leakage paths.
During the longer term, contaminated reactor coolant could be circulated through pipes outside of primary containment. Per SRP Section 15.6.5, the Nuclear Regulatory Commission (NRC) staff usually assumes a failure of the seals in the ECCS equipment, such that significant quantities of coolant could leak into compartments outside of g containment. At CNS the leaked fluid is either retained in the room or transported to the radwaste system. Some portion of this leaked fluid is volatilized and also transported in the air of these compartments. These sources would be processed by the SGTS filters prior to transport to the control room.
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l Figure 4-1 COOPER NUCLEAR STATION l POST-LOCA SOURCE TERM LEAKAGE PATHWAYS
\
l Secondary Containment i s . .
/
Cont nment S G l T l
\ Primary S
l Containment f) A Leakage l
IcOREl
/ (, 4 To
/ - Elevated
/ L _j Release Point u_r r
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CALCULArg7N SHEET 4.2 MAIN STEAM ISOLATION VALVES The Cooper Nuclear Station (CNS) main steam system design includes Main Steam isolation Valves (MSIVs) that provide containment isolation capability for the main steam lines where they exit the primary containment. These valves are arranged in series in four, twenty-four inch steam lines for a total of eight (8) valves. Each steam line has two MSIVs, one inside and one outside the containment barrier. The MSIVs are spring loaded, pneumatic piston-operated globe valves designed to fail closed on loss of pneumatic pressure or loss of power to the pilot valves. Each valve has an air accumulator to assist in the closure of the valve upon loss of the air supply, electrical power to the pilot valves, and failure of the loaded spring. During a DBA LOCA the .
l MSIVs would close upon low water level in the reactor vessel. Once isolation is initiated, the valves will continue to close and cannot be opened except by deliberate manual operator action. !
(] During a DBA LOCA the MSIVs function to isolate main steam following a plant trip and, among other things, form a portion of the primary containment boundary for post-LOCA retention of the radionuclide inventory. As part of the CNS post-LOCA control room operator thyroid dose calculation, it has been determined that some contribution of the source term will bypass the primary / secondary containment features via design basis leakage through theso valves. A technical basis has been . developed for the consideration of retention of much of the released source term in the BOP systems. This technical evaluation is provided as Appendix A to this calculation. Based on this evaluation, a conservative dose reduction factor of 10 is used in this evaluation.
As described in Section 3.1, radioactive material leaking through the main steam lines is transported to the main condenser, where it is then released to the turbine building and then to the environment. This alternate release point requires the application of different W12293015630410695 tj 5' IM #/1/ff W 9. T' I/1/95-l JosNo. 122-93-01 PAGE C122-93-01-01 l CAL.C NO. 14 ena, nc . -c.
CALCULATIUN SHEET atmospheric dispersion coefficients. However, this cannot be done concurrently with the evaluation of normal containment leakage. Therefore, for each LOCA scenario considered, two PADD runs were performed. One run determines the doses due to primary containment ieakage exclusive of MSIV leakage, and the other evaluates the consequences of MSIV leakage only. The resulting doses were then added to obtain the total doses for each scenario.
4.3 OTHER LOCA SCENARIO CONSIDERATIONS 4.3.1 Revised Haloaen Filter Efficiencies and Partitionina Factors Following a loss of coolant accident, normal reactor building ventilation will be isolated, and one train of the standby gas treatment system will start. The design flow rate of this system is 1780 CFM. However, as described in DC 94-102,240 CFM 20% is assumed to pass through the inactive train. For conservatism, this is assumed to be 240 CFM +
20%, or 288 CFM. In accordance with Regulatory Guide 1.52 [ Reference 16], this flow 4}
is assumed to be filtered at an efficiency of 90% for elemental halogens and 30% for organic halogens. The remaining flow of 1492 CFM is assumed to pass through the operating SGTS train with filter efficiencies as described in Section 3.1. For both cases, particulate halogens are assumed to be filtered at an efficiency of 95%. However, PADD cannot explicitly model this scenario. Therefore, the SGTS filter efficiency and halogen partitioning factors (see item 6, Section 5) must be adjusted to account for this.
Figure 4-2 illustrates the process for calculating the revised filter efficiency and partitioning factors. For example, consider the organic halogens passing through the SGTS.
According to Regulatory Guide 1.3,4% of the released halogens are in the organic form.
This results in effective rates of organic halogen discharge of 59.68 CFM through the operating SGTS train and 11.52 CFM through the inactive train. Given filter efficiencies of 95% and 30%, respectively, essentially 2.984 CFM of organic halogens is being W1229301-5634010695
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l released through the operating train, while 8.064 CFM is being released through the inactive train. Similarly, for elemental halogens, given a partitioning factor of 91% and filter efficiencies of 95% and 90%,67.886 CFM and 26.208 CFM will be released from the active and inactive trains, respectively. For particulate halogens, the filter efficiency is 95%
in both cases, resulting in a total of 4.45 CFM being released. The total volumetric flow rate of released halogens is 109.952 CFM. Comparing this value to the original total flow rate of 1780 CFM yields an effective filter efficiency of 93.84%. The updated partitioning factors are calculated as follows:
- E'emental halogens (67.886 + 26.208) /109.952 = .8586 !
- Particulate halogens j 4.45 /109.952 = .0406
- Organic halogens (2.984 + 8.064) /109.952 = .1008 l l
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CALCULATION SHEET Figure 4-2 CALCULATION OF EFFECTIVE FILTER EFFICIENCY AND PARTITIONING FACTORS ORIGIML SCENARIO 1780Cfu 91% = 1619.8 CFW B0.99*CFM' (Elemental) 95% Filter 5% : 19 CfW Efficiency 4.45'CfW' (Particulate) 4% : 11.2 CfW 3.56 *CFW' (Organic)
ACTUAL SCENARIO IITH ASSUWPil(N #10 INCLt10ED 1492CfM 91% = 1357.7 CFW 67.686*CFM' (Elemental) 95% Filter 5% : 14.6 Cru EIIiciency 3.73 *CfM*
(Parliculate) 4% : 59.68 CFW 2.984 'CFM' (Organic) 2" 90% Filter 91%: 262.08 CfW - - 26.208'CFW' EH ic m (Elemental) 05I III" 5% = 14.4 CFW - - 0.72 *CFW' (Farticulate)
" III
- I 4% trt.52 Cru -
03IIII'I - 8.054 *CFW" Efficiency (Organic)
EGJtVALENT ORIGINAL SCENARIO 1780CFW B5.86% = 94.094l109.592 - -
94.094 *CfW' (Elemental) 3,g 7,3,y 4.05% = 4.45l109.592 --
yg; 4.45'CfW' (Particulate) 10.08% = 11.048l109.592 - - 11.048 *CFW'
[ Organic)
TOTAL : 109.592*CFM' : 93.84% Ef ficienct W1229301-5630010695
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CALCULATION SHEET 4.3.2 ECCS Leakaae
- p As a result of evaluations of leakage from the ECCS recirculation system, it was necessary to include the dose contribution due to ECCS leakage. This is accomplished by revising the primary containment leak rate to include this additional contribution.
However, because the ECCS leakage involves fluid leakage rather than vapor leakage, the leak rate was corrected for this. From Reference 31,25% of the halogen inventory available for release is released to the primary containment atmosphere, and from Reference 48, an additional 50% is mixed with the ECCS water volume. Given a water volume of 96,445 ft [ Reference 49] and assuming a release factor of 0.1 (see Section 3.1, item 17), an ECCS leak rate of 1000 cc/ min is equivalent to a halogen leak rate of 0.00264% of the total halogens available for release per day. Given an overall primary containment leak rate of 0.635% per day, the corresponding leak rate from the primary containment atmosphere is 0.1588% of the available halogens per day. Taking the ratio of the sum of the primary containment (0.1588%) and ECCS (0.00264%) halogen vapor leak rates to the primary containment halogen vapor leak rate alone and multiplying by the original primary containment leak rate of 0.635% per day results in an adjusted primary containment leak rate of 0.646% per day.
l
- r W1229301-5630 010695 T ID I[f /K M[] 7 I/1/Tff JOBNO. 122-93 41 PAGE l CALC NO.C122-93-0101 18 ERIN
- ENGINEERING AND RESEARCH,INC.
CALCULATifiN SHEET ,
l 4.4 GENERAL
SUMMARY
OF THE REFUELING ACCIDENT SCENARIO Accidents that result in the release of radioactive materials directly to the containment can occur when the drywellis open. A survey of the various conditions that could exist when j the drywell is open reveals that the greatest potential for the release of radioactive l 1
material occurs when the drywell head and reactor vessel head have been removed. In j this case, radioactive material released as a result of fuel failure is available for transport directly to the containment.
Various mechanisms for fuel failure under this condition have been investigated. With the current fuel design, the refueling interlocks, which impose restrictions on the movement of refueling equipment and control rods, prevent an inadvertent criticality during refueling operations. In addition, the reactor protection system can initiate a reactor scram in time j to prevent fuel damage for errors or malfunctions occurring during planned criticality tests l 1
with the reactor vessel head off. It is concluded that the only accident that could result !
in the release' of significant quantities of fission products to the containment during this
'}-
mode of operation is one resulting from the accidental dropping of a fuel bundle onto the top of the core.
This event occurs under non-operating conditions for the fuel. The key assumption of this postulated occurrence is the inadvertent mechanical damage to the fuel rod cladding as j a consequence of the fuel bundle being dropped on the core while in the cold condition.
Fuel densification considerations do not enter into or affect the accident results.
l l
W1229301-5630-010695
} f M i /f /1f~ %,9 T I/9/9 f l JOBNO. 122-9341 PAGE l CALC NO.C122-93-0141 19 I
em, ec . -c.
CALCULATION SHEET 4.5 CONTROL ROOM ACTMTY CALCULATION EQUATIONS i
The calculations presented in this study are performed with the use of the PADD computer program. PADD models a three compartment system of volumes. Typically, compartment 1 represents the primary containment, compartment 2 represents the reactor building or secondary containment, and compartment 3 represents the control room. PADD calculates airborne isotopic radioactivity concentrations as a function of time in each of the three compartments, taking into account removal due to radioactive decay and leakage out of the compartment, as well as production due to leakage into the compartment. The postulated DBA releases consist almost exclusively of halogens and noble gases. Due to the neutron-rich nature of fission products, decay tends toward isotopes with higher atomic numbers. In other words, halogens tend to decay into noble gases, which then decay into other isotopes. Therefore, production of halogens from radioactive decay of other isotopes does not occur, although production of noble gases from radioactive decay of halogen isotopes does.
3 Using the information described above, PADD then calculates dose rates and integrated doses for the control room and two offsite locations specified by the user. Meteorological data modeling atmospheric dispersion is also an input to the code.
The airborne isotopic radioactivity concentrations are based upon closed form solutions of the differential equations describing production and removal of isotopes in each compartment. Dose rates are calculated by applying appropriate dose conversion factors to the airborne radioactivity concentrations. Integrated doses are calculated numerically using the rectangle rule. The rectangle rule estimates the value of an integral during a 1 time step by assuming the integral is constant over the time interval. This constant is ;
equal to the value of the integral at the midpoint of the interval. The specific equations employed within the PADD program are presented as follows:
W1229301-5630 010695
.} f /% //f/fr '7DS T (/9/9 d xsm. 122-93-01 PAGE l cacm. C122-93-01-01 20 ERIN'
~
\
CALCULATION SHEET 4.5.1 Haloaen Concentration in Primary Containment i
(' (' (*'}
C3y(t) S;y(t ) '
e
' -) }
e ' '"
I.e ' 4'( }e '4"{'~~') + f, + l
[ty ., $ t s yt } , V, (1) 4.5.2 Noble Gas Concentration in Primary Containment C,y(t) -r (e-r_,) -y r (r.-r ,) ,S;y(t) A//S"(t)
[t,stst]_ e-4('~'*') E V,(A -A;)(e-4 ('~'*' ) - e
u- u V, i i g . - "g ,-4(r-c)-R4(c-r,)7 A,s,y(ty., ) , _4(,..,.,,3 _ ,,,(,,,,.,,3 .
, i V,(Ai -^i) ,.,
(2)
W1229301-5630LO10695
- } f M i/f/ff QAT //9/9 5-l JOB NO. 122-93-01 PAGE I CALC NO. C122-93-01-01 21 cuawe ca mc.
I CALCULATION SHEET 4.5.3 Haloaen Concentration in Secondary Containment
)
~
C,,(t) S,(t) -j r.(e.-e ,)
[t.,stst},V, u u V, ,
-fa,te.-ra x f" +f* (e ~'""~'" J -e -'"-'" d) +(e*'"*' d"" d -e """-)
_ Yu Vn-1~Apn _
(3)
~
.+ e -4,,.aitt-e ,i x. Smy (t .3) 0 -
3 e
-j r,(e.-r-,i Y1 Y2 y.1
-E A,tt,-c I
+ f "'
x * (e'"-'""-"-'I- e "-"-' "-) +
- le ~I'"-' " d i'"-' "-'I-e "-'""-' "-)
, Vu-s VN-1-A pu-1 ,
- I
'3' s,gtx),-$,(WX4+)-$ '4',+) If(,-,ga ),, g,,%)), e' p.,.1,,XW4 ) - , -*<'a- k-i 3 s., , V,, , V, yx i d ,x
\
W1229301-563H)10695
} f f% f /p hr- +7n p T t/9/9} JOB NO. 122-93-01 PAGE l CALC No. C122-93-01-01 22
@ OF 60 ENGINEERING AND RESEARCH, WC.
CALCULATION SHEET 4.5.4 Noble Gas Concentration in Secondary Containment 1
2,( )
t.,stst u u
,, bV, 7' 4 g -I<,. A)tt-r )
,N x' bV, T f, x. V,
' e '
(A#(
T3- -
(4)
,_, , 1 N-2
+[' f e Y ' "'~ '~
g T I
Ta+ e '>"*'~ d T+E s
- l i
l Where:
e e ('~*^) ' I1 , - / 1,l r .
T1 (4a) 3
~
T2 m e ('*' * ^) '+' / ,.. l r , - I t., l r, .
1 (4b)
T,3 a e ~(**) 6-- I1,l t, -- / ,l r, . 1 (4c)
.T, a e ~('"*^) ' ,1, 2 - I,, j ,,,
(4d)
T, a e ('*'
- A) '*' I l - /,,, l ,,,, (40)
T.e e e i'"**) 4 /2,l t, - I 2, l t ., (41)
W1229301-5630410695
} S~ l'D 1/fI'6 %$I //f/t d JOBNO. 122-93-01 paos l CALC NO.C122-934101 23 ERIN
- ENGINEERING AND RESEARCH, INC.
CALCULATlHN SHEET 4.5.5 Halogen Concentration in Control Room
[ty ., s t syt ] = T,-(t) - e #* * "'"~' 2- yT (t .,)
(5)
+ T3 ,(t) - e ** * "'"~' T 3,(ty .3)
~
e .$ % *"'^ ( T,,(t g ) -e ***' T,,yg.,) ) + f,,yg) -e ***' T,,y,.,))
l Where:
N-1 '
e -vu(r-ru -5)
X Sw(t) ' O ' e -E <r(t<- fr-i) p e <u(t-tu-i)' 1
[ f + /*
i ,
T2 (t) - E3'n O n (1 -e z
,~) 2 0, { v, v ,, , '-' r, , s,u 0,, ,
- -1
- E A pr(tr- tr.1) r f,g r-t g (ty+1)(t-tu 3) g -ey(t- ty.,)3
/H N
+
0,y 1 + O '{'"* 0 A )(t~
[ O (I t -')
V
-1)
TN~ApN ( b/N~ApN , 3 i ,
N-t
-E A pr(f -t,.3) f
..Yri-1) O "'
~
X -
Oy 0, #-2 . ..._ ([p+[o 6-tu.i(tg g-tu_g) - 6e .,(fg.,-tu.g) ) + [e
( 2 s y.1 TN-1 TN-1 -A pN-1
- #-t )
( } ( '"4 $}
S/h(A) & " '
g (tu.,+1g.,)(tu.,-fn 2)g-*N-1(tu-t -tu-2) ) ) , {
K=1 g V,g2 V 1
~ ~
g.s
-[ Agt,-ty3) i
.. ~ ?r(fr~ot s)
.. [p + [, I, O g (vg+1)(tg-fg.3) ,, g eg(tg-tg.g) 6 -tg{tg-fg.g) , g-eg(tg-fg.g) )
(E l
\
(Note that 6g m Kg - Tx and6x e Kg - dx) i 3 W1229301-5630410695 3 5 #th ih/fr ~Pn g7 ih/+dxem. 1224341 no, CALC NO. C122434141 24 l 0 OF e e a s e. ENGINEERING AND RESEARCH,lNC.
CALCULATION SHEET x-1 l X Sw(0 ~'d '"-') ~ E (4' -') '
I*
3 g OE
- O W-1 (7)
-Ap u(f-fg.1) -E AMfr -fr t) f, e "'
fp+ f, x + ,
ON~ApN b lN (Note that 6 = K - T).
l 4.5.6 Noble Gas Concentration in Control Room 8"
T2 ,(0 - 8 ""' '}(""} T2y(fu-1)
~
+ (""'*'"""}
=
Ts,(0 - 8 Ts,(fu-1)
P I
N-1 < - E (K,+1,)(t,-t es)
+ T ($ - e -(#"* *'I("-'I T (tg m). +{ ,e ""'
K-1
- 8 (""**''"~'" Tdtud + ( TuM - 8 (""**('"~'"
~ ~
x ( T;(0 '
2 Tdtu$
+(T y (0 - e-(""**')('<-'<-')Tyrd ) ' }
(8) l l
i l
I I
W1229301-5630410635
) S /h6 :/rkr % 97 "
t/98dato 12243-oi nos l cacre ci22s341.ci 2s l
M N*
l
CALCULATION SHEET
''' ' '"I')
7,*(t) - E3* .!-- O r" e ' '"" ' '"'
Oc n V;6;u A/ S/N (t) e - A ('-'~ ) e 4 ('-'~ )
+e - A - '-- E _
(9) i i V i
_i(A-A;) 6u- A - A, i
6;n
~
4 4 A/iS"(ty ,)
e (t~t""}~ '
{4) {
6iNV,(A -A i) i x {e-4(c.,-c.) _ e4(,..,-,.) _ f i l 73 (t ) - E , **)' {Ti(t) + 7;(t) + 7;(t) + 7;(t) + 7;(t) + 7;(t)} j f " (1 -e ,)O,,e-(" 2 (10) 3 - l W1229301-5630010695
} _
6 M /hMr TMp7 Nh r, lnm. 122-93 41 pAos l CALC NO. C122-9341-01 26 l
@ OF ENGINEERING AND RESEARCH, INC.
CALCULATION SHEET l N-1 S,y(ty.,) e ('"* *')"~ '*'I
~
I ' O, ' S,y(t) e "I'*') T, (t) - e ~ h 'r(44')o(r,.x,),
. 1 y ,
V r ein vYu, s 6:n Gui A, 3 [s,y(t)o-(',.22,xt-c ) _ 3;u{f ,),(,,.2,X'-r+i)' ' j V,(A,-Aj) ,y,- A j- A, , s eu-A-A j i e,n
~
('"* *'X*') '
, 1 [Sjy(t) e ~( '"* *'*X*> *') _ Sjy(ty.i) e j
Yu , OlN " Ej Ou l 4, 1 N-1 l
,-(. 1,)( - r,..) u g - 2,(r- r..) .g 2,(r,-r,,) r 18 1 1u (f u-1) 1 Kui e gu ,., , V,( A,- 1,)
, ,Ie **'(4 ' ' 'A - e **i(4 ' A ,
YN - 11 , , , (10a) l
. ' o, ' .
Te (t) = - Ts ,(K,n + 2,) t-(<,+ 2,)(<- r+i) g 2 sy., Oul g (10b) l n., u-a < E ('< * *')(4- bi) l
*(4* *')'- ~' '
T (t) - EV,gK.1 ',b > t Kui + Ag T'" 3 (10c) l W1229301-5630410695 4
} 6 % i M/ff WQ T //9/e/(l JOONO. 122-9341 PAGE l CALC NO.C122-9341-01 27 l ERIN * -
mmmmenenemmmmmme _____ 1 i
CALCULATif7N SHEET
- o Ip* lo 1 s,u(t) e""(*')
74(t) = e(Kon* A')t x V,y a Vs Yu ,in* 1 -11r 1 bin
- Ai - At sjy(f,,,,)e ('n* *sX'-'+i)' $ 's ju(t) e ""C'-'*') ,
s jy(ty_, ) e -('* * *'X+') ' ' 0,y X;- X, Ogu + 1g-1, O gg ,, f, N ^4'<-'-') , Sj u(t) 8 '('"' *'"X) _ Sf u(tu-i ) 8 '('"'*'X"") ' YN~A pn YN~A pu*A~A l fg Oru~ A pu*A~A l f O IN _ e ('** *'X'-'*') l, ' 1 Sj u(t) e""() (*"* *'X*'I ' , [e -('** */X*') - 1 _ Sju(tu-t ) e A-A Olu + At - Af 0u/ n , O fy + 1,- Aj 0u
/ Ai-1/3 l /t r l u.a
-e - E 1,,(r,-to,)
x S(t.,)e jy y
# * * [e"*'('*' #8) - o ""~'C'*'"4)) +
- j
( 2 y.1 3 ,YN-1 TN-1 ~A pm, g g (in-i' A,n.i)(fn-i *'ea) _ -en_i(rg,-tma) , I _ o'(""N'*') ,
]' 0 ' 1 Sju(t) s Kyi + X, e gy ,
, V, , , V,
.-1
-i . .-i -r 2,,(r,-r,.,)
- E,(<, 2;)(r,-tr )-p =,(r,-t-i) f, + f,
, _,,(g. g,) _ .,,(;.g,))
/,e -i Y. Y-A ge x (e-(i.+2,.XL-4-i) _ ,"&L-i))]}}
(10d) I l Ts (t) - Ts ,(i;n a,)e-c..i 2,)(r-r.i> 0;u (10e) W12293015630010695 3 5 l@ i /f /6 ~M19-T //9/9 rinNo. 122-93-01 pass l CALC NO. C122-9341-01 28
@ OF ENGINEERING AND RESEARCH, INC.
CALCULATION SHEET l
) y., (K,n a,)r - E (<, A,)(t,- r-i)
' * ~^ '
Ta (t) - E ' T;, K=1 Ku i + kg l 1 (10f) ; l Where j l l
- ~
e (#") ("-') T ' T,(t ) = { ' Aj /j, K;u-K,g+A,-A, y
;,(t _,) ("")(~)
+ /;, e y T;,(t _,) W,3_
i K;u-K,u+A;-A, l (11) i h
,. L._ _ _
W12293015630010695 J 5- (M i/fler % /J-7 I/4/9clJOBNO. 122-93-01 PAGE l CALC NO.C122-9341-01 29 O OF 60
,. e s, , ENGINEERING AND RESEARCH, INC.
)
1
CALCULATITN SHEET N-1
' X\ Sjt)'O,'
p -E'd'<-'es> l'a, = E,u g (1-egy) 02n ' - e "' A 3/N ,
,1 g 2j u N-1 3 -E A - '<-1) r
[ #f ' g -tu(t-tg g) g -8 N(t-'N-s) [, e pr('t e ('"' A 'J
~
P {t- ta ya sp{b,y+ A f- A,) Op(0;u+ Af - Aj ), Yn~ApN y Nby~ApNb NOIN~Aph
' ' N-t e '"('~ '"-'I e 'b 'A 't"d
~
e "I'"* */} I'~ '"~') 8FN(I -1) ' O[ Op(O f u + Af - Aj ) , Op (0f y + Af - Aj ) V1 ( V2, u.1 N-2
- E Apr(t,-t,.3) l + l (e ~'"I'"-' "'"- )
p o ~
- e '"-' I'" ~ '"-"I ) + f* ( e ~I'"~' ' * #"-' II'"-' ' '"-'I Yu.1 Yu-1 -A pu.1 N-1 K-1 fO' 3 Sg (t) ,[, (<r+ A )('r- f -i) -[ 'r('<-'r-i) 1 r
- e " ' I'"-' ' '"'8) ) +
K=1 z V,y 2 V 1 A-1
-E A pr(t,- t,.i) p*o l8 g -tx{t y- tg.,) _g -ex{tg- tg.,) e g (tg+ %,x)(tg-tg.y _ y -tx (t,-tg.,
y -} 1x Yx- ApK (11a) r r I'st - E, Og, S u(t) e-'4'-'*-i)- N ',(t -f-i) j "' s ( f, G y ,1
~
u-i
-1,m('-'n-1)- E A,,(f,-4 i) f,0 ** fp+ f, >
x +
,(bfN ~ ApN)(bt# ~ ApN + Al ~ Af ) bJN(blN+ A l" Af)
.(11b) 4 W1229301563OO10695
} .f h6b i /f[1t' % #} 7 //9/41-l JOBNO. 122-03-01 PAGE CALC NO. C122-93-01-01 30 EitlN*
_ _ _ _ _._ m
CALCULATION SHEET , N-1 j e "f"p(K 1)(4 4,)[(Tljt y )- e%+2 X'n-'<4)Tl.Itg.3))
/
I'4s - (Ku + A ) x-1E i i
+ (Tl)t,) N S W Ws'Jtx.,)f }
(11c) 4.5.7 Control Room Filter Activity A,(t) 'X X 2) eg,Q,,Q,,A,ft) +e s,Oe,A,, [ty ., ststu}" B l ~'2-)*3 02-0 3-A,y) + ) A more detailed discussion of the derivation of these equations is presented in Section 5 of Reference 40.- 4.6 l PADD INPUT DATA _._
- PADD is designed tn read virtually all of its input from a series of input files. These files
~ ~
~ are ASCil text files created by the user using anylext editor piior to executing PADD. The user may add comment lines to any of the PADD input files. A comment line is indicated when the first two characters of the line are exclamation points. The input processing part of PADD ignores comment lines. Allinput in PADD is free format. This means that all data items are to be separated by a comma, and the last data item on a line should not have a comma following it. If data cannot be fit onto one 80 character line, data may be continued on subsequent lines. Character data must also be enclosed in single W12293015630410695
") T M i/f /f( ~7J')f)-T //1/7fl JOBNO. 122-93-01 PAGE l CALC No. C122-93-01-01 31 enanc n . -c.
CALCULATIEN SHEET l quotes. A summary of the required input files is provided below: (In certa /n Instances where an example of data file name is given, it is merely for demonstration, and should not be construed as the only choice.) File Nama and Extension Description I ser U specified input / output file name specification file. This file has only two lines. Line 1: Name of the primary input file (with extension, if any) in single quotes. e.g. ' CASE 1.IN' Line 2: Name of the output file (with extension, if any) in single quotes. e.g. ' CASE 1.OUT User specified Primary input file. This file specifies the names of all other input files used with the . run and specifies other basic data determining the size and complexity of the ean: Line 1: Run title (maximum 80 characters) inside sing % quotes ('_'). Line 2: ) input file names hiaximum 8 characters with extension - see file descriptions below): Name of isotope data file e.g. 'lSOCASE1.DAT , Name of energy group data file l e.g. 'ENGCASE1.DAT .. j i Name of breathing rato data file e.g. 'BRECASE1.DAT Name of time data file e.g. TIMCASE1.DAT Name of filter efficiency data file e.g. 'FILCASE1.DAT Name of Chl/O data file e.g. 'CHOCASE1.DAT Line 3: Number of ' time steps" (maximum of 30) ; Number of noble gas isotopes (maximum of 50) Number of halogen isotopes (maximum of 50) Number of dose locations (maximum of 3) Number of gamma energy groups (maximum of 30) W12293014430410G95 } 6 @ l/f/rf WpT //*F/9f JOBNC. 122-93 01 pang
) CALC NO. C122-93-0141 32 i ERIN
- ENGINEERING AND RESEARcH. INc.
CALCULATION SHEET File Name and Extension Description Une 4: Time (in sec) at which plateout is to be " shut off" by code. Fraction of halogens in elemental form. Fraction of halogens in particulate form. Fraction of halogens in organic form. Decontamination factor - at times 2: that for which plateout is ' shut off" (see above), reduction in airbome halogen concentration due to plateout is assumed to be a constant equal to the reciprocal of this factor. Une 5: Compartment 1 volt.me (cubic feet). Compartment 2 volume (cubic feet). Compartment 3 volume (cubic feet). User specified Isotc2.- data input file. (ISOCASE1.DA7) Une 1: Names of noble gas isotopes (maximum 6 characters). e.g. 'KR-81','KR-83M*,'KR-85'...etc. Names of halogen isotopes (maximum 6 characters). e.g. 'BR-80','BR-80M','BR-82'...etc. Note that the Noble gas and Halogen listing must begin on a new line. Une 2: :_. h Half-lives (sec.) for noble gas isotopes; and Half-lives (sec.) for halogen isotopes. Note that the Noble gas and Halogen listing must begin on a new line. Une 3: Initial noble gas actMiles (Curies) released into containment at t=0; and initial halogen actMtles (Curles) released into containment at t=0. n Note that the Noble gas and Halogen listing must begin on a newline. Une 4: Mean garr.ma energy (MeV) emitted by each noble gas isotope (Maximum of 30). ; une 5: Mean gamma energy (MeV) emitted by each halogen isotope (Maximum of 30). Une 6: ActMty to beta skin dose rate conversion factors for each noble gas isotope (mrem /yr/pCl/m*). Une 7: l ActMty to whole body gamma dose rate conversion factor for each noble gas l isotope (mrem /yr/uCi/m3). 1 i 4 W1229301-563M10695 } 6 (% i/1/ff LtJ') /) 7' I/al/1fl JoBNo. 122-93-01 PAGE l calc No. C122-91-01-01 33 ERIN
- ENGINEERING ANo RESEARCH. INc. j o
CALCULATION SHEET File Name and Extension Description Une 8: Inhaled actMty dose conversion factor for each Halogen isotope (rem /CI). Une 9: Adult inhaled actMty to total body dose conversion factor for each Halogen isotope (rem /CI). Une 10: Child inhaled actMty to total body dose conversion factor for each Halogen isotope (rern/CI). Une 11: infant inhaled actMty to total body dose conversion factor for each halogen isotope (rem /CI). User specified Energy group input file (ENGCASE7.DAT) The conversion factors identified below are determined through a separate shielding calculation. ) Une 1: Gamma energy endpoints for each of the energy groups (MeV). Une 2: Flux to gamma dose conversion factors (rem /MeV/sec). - - Une 3: 3 Flux to gamma dose rate conversion factors (rem /hr/ photons /sec-m ) for Reactor building shine dose (G2 factors). Une 4: Flux to gamma dose conversion factors (rem /hr/ photons /sec) for Control room : filter shine dose (GA factors). l Line 5: Flux to gamma dose rate conversion factors (rem /hr/ photons /sec-m') for Control room gamma c'oud dose (GG factors). User specified Breathing rate input file (BRECASE7.DAT) Une 1: 8 Adult breathing rates for each location and time step (M /sec). Note: The breathing rates at each location must begin on a new line. l t W12293015634010695 } 5' M i/1/f( h 7 7~ //9/1$-l JoBNo. 122-93-01 PAGE l CALC NO.C122-93 01-01 34 ERIN
- ENGINEERING AND RESEARcH. INC.
CALCULATION SHEET File Name and Extension Description 8 Ine 2: Child creathing rates for each location and time step (M'/sec). Note: The breathing rates at each location must begin on a new line. Additionally, chlid breathing rates for Containment are not used, but a value must be entered. Une 3: infant breathing rates for each location and time step (M'/sec). Note: The breathing rates at each location must begin on a new Ilne. Additionally, infant breathing rates for Containment are not used, but a value must be entered.
+ User specified Time related input file (TIMCASE7.DA7)
! 1 l fi j Une 1:
1
)
;f Time steps (sec) from 0 to 30 days. (maximum of 30)
?
Y The distribution of the time steps is not required to be equally distributed between 0 and 30 days. The number of time steps selected directly influences the accuracy of the numericalIntegration performed by the program. It is recommended that the minimum time step used is 60 seconds. Additionally, it is important that selection of the time steps accounts for data changes, e.g. assume the adult breathing rate for the reactor building changes from 3.47E 4 mI/sec to 1.75E-4 ml /sec at T=8 hrs. To account for this change one should q have two time steps at the 8 hr mark, i.e. 8 brs, and B hrs 60 sec. This is 4 repeated for allInstances any datum changes. Une 2: Containment by-pass leak rate, "OE" (SCFH). Une 3: Containment leak rate, "O1' (volume %/ day). Une 4: SGTS flow rate, '02' (CFM). Une 5: Control room inflow rate, '03" (SCFM). Une 6: Control room inteakage rate, '04' (SCFM). Une 7: Control room recirculation flow through filters, "O6" (SCFM). Une 8: i Plate-out time constant (sec) for each time step. Line 9: Control room occupancy factors for each time step. W1229301-5630010695 J i f M , /7 /9(' % ry 7 1/4/y,rlJODNO. n 122-93-01 pang l l CALC NO. C122-9341-01 35 l O OF 1 .. a- n, , ENGINEERING ANo REsEARcH,lNc.
\
1
CALCULATION SHEET i l File Name and j Extension Description ) l User specified Fitter officiency input file (F/LCASE1.DA7) Line 1. 1 SGTS filter efficiencies for each Noble gas. i Note: The efficiencies are entered in the same order as were the isotopes in l
'ISOCASE1.DAP above. The filter efficiency for each Noble gas Isotope must begin on a new Ilne. Additionally, a zero value shall be entered for instances :
when no efficiency value exists. This feature was retained in the event that a different class of isotope, which can be filtered, is used. , 1 Une 2: SGTS filter efficiencies for each Halogen. Note: The efficiencies are entered Ir ihe same order as were the isotopes in i
'ISOCASE1.DAT* above. The filter efs nicy for each Halogen isotope must begin on a new line. Additionally, a 10,o value shall be entered for Instances when no efficiency value exists. j Une 3:
Control room filter efficiencies for each Noble gas. Note: The efficiencies are entered in the same order as were the isotopes in
'ISOCASE1.DAT* above. The filter efficiency for each Noble gas isotope must begin on'a new line. Additionally, a zero value shall be entered for instances when no efficiency value exists.
O u ne 4: Control room filter efficiencies for each Halogens. Note: The efficiencies are entered in the same order as were the isotopes in
'ISOCASE1.DAP above. The filter efficiency for each Halogen isotope must begin on a new line. Additionally, a zero value shall be entered for Instances when no efficiency value exists.
User specified CHl/O factors input file (CHOCASE1.DA7) Une 1: _ Elevated release CHl/O factors for each location; Control room, LPZ followed by EAB. Note: The CHijQ factors for each location must beg'n on a new line. Une 2: Cround release CHl/O factors for each location; Control room, LPZ followed by EAB. Note: The CHIjQ factors for each location must begin on a new line. W1229301-56%e10395 } 6 kd> I/f /ff' ~'7) r AT I/9/1f { JOBNO. 122-9341 PAGE l Calf NO. C122-93-0141 36 Enawr , wc.
CALCULATION SHEET Each numbered line of data, as described above, can be separated by a comment line (ll), i.e., the line immediately foilowing the end of "Line 1" and preceding the beginning of "Line 2", can be a comment line. PADD provides detailed output summarizing the input file names, input data, and all calculated data. Currently, PADD output indicates the program name, version number, run date, and run time on the first page only, and it doos not paginate the output. This decision was made to make the program independent of specific printers. A summary of the output tables appearing in a PADD output in the sequence that they appear is provided below: PADD output (CASE 1 OUT) Summary of Isotope data file (ISOCASE1.DA7). Summary of Energy group data file (ENGCASE1.DA7). @ Summary of Breathing rate data file (BRECASE1.DA7). Summary of Time data file (T/MCASE1.DA7). Summary of Filter efficiency data file (FILCASE1.DA7). Summary of CHl/O factors data file (CHOCASE1.DA7). I The following is calculated output; values are presented for each time step and isotope: Containment Noble gas concentrations (Ci/m*) Containment Halogen concentrations (Ci/m') 8 Reactor building Noble gas concentrations (Ci/m ) Reactor building Halogen concentrations (Ci/m') W1229301-5630410095 122-93-01 ~} f f /</ /f t' JT //9/9flJOBro PAGE l CALC to C122-93-01-01 37 ena,~ n c . ,~c.
CALCULATl?N SHEET j Control room Noble gas concentrations (Ci/m') Control room Halogen concentraticns (Cl/m2) l l Control room emergency filter halogen activities (Ci) Noble gas activity; elevated release rate from secondary Containment (Ci/sec) Halogen activ;ty; elevated release rate from secondary Containment (Ci/sec) Noble gas activity; ground release rate from secondary Containment (Ci/sec) i i Halogen activity; ground release from secondary Containment (Ci/sec) Reactor building gamma shine dose rate to Control room (Rem /hr) CR emergency filter gamma shine dose rates to Control room (Rem /hr) Control room emergency filter gamma shine dose to control room (Rem) ) Gamma cloud immersion dose rates to bontrol room (Rem /hr) Gamma cloud immersion doses to Control room (Rem) Thyroid dose rates to Control room (Rem /hr) Thyroid doses to Control roorn (Rem) Beta skin' dose rates to Control room (Rem /hr)_ Beta skin doses to Control room (Rem) LPZ gamma cloud dose rates (Adult, child, and infant) (Rem /hr) LPZ gamma cloud doses (adult, child, and int) (Rem) LPZ adult inhalation total body dose rates (Rem /hr) LPZ adult inhalation total body doses (Rem) W1229301-5630410695 [ II i ff[6 MMT 1/9/9 d JODNO. 122-93 01 paog l cALCNO.C122-934101 38 cue-e a c.c
CALCULATION SHEET LPZ adult thyroid dose rates (Rem /hr) LPZ adult thyroid doses (Rem) LPZ child inhalation total body dose rates (Rem /hr) LPZ child inhalation total body doses (Rem) LPZ child thyroid dose rates (Rem /hr) LPZ child thyroid doses (Rem) l LPZ infant inhalation total body dose rates (Rem /hr) LPZ infant inhalation total body doses (Rem) LPZ infant thyroid dose rates (Rem /hr) LPZ infant thyroid doses (Rem /hr) EAB gamma cloud dose rates (adult, child, and infant) (Rem /hr) EAB gamma cloud doses (adult, child, and infant) (Rem) EAB adult inhalation total body dose rates (Rem /hr) EAB adult inhalation total body doses (Rem) EAB adult thyroid dose rates (Rem /hr) EAB_ adult thyroid doses (Rem)
~ ~
~ Totafdose rates (Rem /hr) and integrated doses (Rem) (dose rates followed by dose)
The above output is presented in tabular form, each column representing a " time step". The final table, " Total dose rates...etc." summarizes the cumulative exposure over the time periods of study for the 13 different dose parameters: W1229301-5630410695 g 5 AtD/ ,/9kr ty;y y f/9/qflxeNo. 122-93 41 paus 33 , l CRC No.C122-9341-01
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CALCULATION SHEET RB Shine i Filter Shine Control Room Cloud (Whole Body) Control Room Thyroid Control Room Beta Skin LPZ Whole Body LPZ Adult Thyroid LPZ Child Whole Body LPZ Child Thyroid LPZ infant Whole Body LPZ Infant Thyroid EAB Whole Body EAB Adult Thyroid in this study, the focus is on operator thyroid, whole body and beta skin doses. i
.?-
l W1229301-56'OO10695 f i /f/9f W /3. T I/9/9 $1 xsm. 122-9341 noc C122-93-01-01 l CALC NO. 40 ENGWE R CH, lNC.
CALCULATIPN SHEET 5.0 DESIGN INPUTS The design inputs for the CNS post-LOCA control room operator thyroid and whole body dose analysis are:
- 1. 100% of the noble gases in the reactor and 25% of the iodine instantaneously becomes available for leakage from the primary containment as an aerosol based on TID 14844 in accordance with the CNS USAR.
- 2. The primary containment volume leaks at a rate of 0.635% weight per day for 30 days in accordance with the CNS USAR.
- 3. The breathing rate is 347 cc/sec for the first 8 hours,175 cc/sec for the next 16 hours, and 232 cc/sec thereafter in accordance with Reg. Guide 1.3, Rev. 2.
- 4. Control room occupancy factors are 1.0 for the first 24 hours,0.6 for the next 72 hours, and 0.4 thereafter in accordance with Reference 28, Table 1.
- 5. In accordance with Reg. Guide 1.3, Rev. 2, the source term for this calculation is )
the instantaneous release to the drywell atmosphere of 100% of the core inventory i of noble gases and 25% of the core inventory of Halogens.
- 6. In accordance with. Reg. Guide 1.3, Rev. 2, of the released halogens,91% are in ,
h the elemental form,5% are in the particulate form, and 4% are in organic form. However, as explained in Section 4.3.1, these factors were adjusted to model flow I through the inactive SGTS train. The resulting partitioning factors are 85.86%, 4.06%, and 10.08% for elemental, particulate, and organic halogens, respectively.
- 7. The overall primary containment integrated leak rate limit defined in CNS Technical Specification 4.7.A.2 is 0.635% of the primary containment volume per day at 58.psig. As described in Section 4.3.2, an additional leakage contribution due to leakage.from the ECCS recirculation system is added to this, resulting in a primary containment leak rate of 0.646% per day.
In CNS Surveillance Procedure 6.3.1.1 (Revision 30), the leak rate of 0.635% per day is calculated to be 316 SCFH. For the case of MSIV leakage, the totallimit for containment leakage via the MSIVs is 46 SCFH, as stated in CNS Technical Specification 4.7.A.2. Taking the ratio of 46 SCFH to 316 SCFH and multiplying by 0.635% per day yields a containment leakage rate due to MSIV leakage alone of 0.092% per day. W12293015630410695 J #F % //r/y Wp7 //9/9 d XENO. 122-93-01 C122-93-0141 pact 41 l CALC NO. So c~m~e e . ,~C.
CALCULATION SHEET
- 8. In accordance with Reg. Guide 1.3, Rev. 2, leakage from the primary containment
, is assumed to pass directly to the SGTS inlet without mixing in the surrounding secondary containment atmosphere and should then be assur 4 'n be released as an elevated plume for those facilities with stacks (i.e., CNE It& " e, no direct release to the environment from the primary containment was .. u v. d in this analysis.
- 9. Unless otherwise specified, the design inputs listed in this section were taken from Chapter XIV of the CNS USAR (Reference 10).
- 10. There is a one minute delay in actuation of the control room emergency filtration unit as described in Section 3.1. This results in one minute of unfil'eed inleakage into the control room.
- 11. To prevent credit for mixing in the reactor building, the reactor building volume specified in the PADD input data is set to 1 ft'.
- 12. The turbine building volume used for calculation of the dose contribution from MSiV leakage is 1.8E6 ft*. This is the volume of the turbine building above the operating floor and the volume of the shielded area on the mezzanine floor less 20% for equipment. [Refererce 45]
- 13. The free air volume of the main condenser is assumed to be 69,940 ft'. .This
) assumes a 7 foot hotwell water level and is not reduced to account for the volume of the internal structural steel and minor piping. [ Reference 46]
- 14. A release from the turbine building without operation of the HVAC and off-gas systems is assumed to uniformly occur from the entire structure. From I
Reference 47, the turbine building is 111.5 feet wide, 323.5 feet long, and 104.3 feet high. The distance from the center of the turbine building to the control room ventilation intake is approximately 52 meters. The control room ventilation intake is approximately 53 feet above ground level.
- 15. Control room filtered flow rates of '375 CFM,1000 CFM, and 2000 CFM are !
assumed for this calculation. These how rates represent the current, planned, and potential future design flow rates, respectively. W1229301-5630010G95 } S [ f phy- %gT //g/9 d JOBNO. 122-9341 PAGE l CALC NO.C122-93-01-01 42 1 ERIN
- l
CALCULATIEN SHEET The design inputs for the refueling accident analysis include those for the LOCA with the following additional inputs.
- 1. The fuel assembly is dropped from the maximum height allowed by the fuel handling equipment.
- 2. 'The esire amount of potential energy (referenced to the top of the reactor core) is available for application to the fuel assemblies involved in the accident. This assumption neglects the dissipation of some of the mechanical energy of the falling fuel assembly in the water above the core and requires the complete detachment of the assembly from the fuel hoisting equipment. This is only possible if the fuel assembly handle, the fuel grapple, or the grapple cable breaks, or improper grapplings occur.
- 3. None of the energy associated with the dropped fuel assembly is absorbed by the fuel material (uranium dioxide).
- 4. 99% of the halogens released from the rods is retained by the refueling pool water. ;
Since there is no provision in PADD to model this holdup factor, it was taken into account in the PADD input for initial halogen isotope activities. l
- 5. The reactor fuel has an irradiatioFtime of 1000 days at design' power'up to 24 Q hours prior to the accident. T%s assumption results in an equilibrium fission i product concentration at the time the reactor is shut down. Longer operating l histories do not significantly increase the concentration of the fission products of ;
concern. The 24-hour decay time allows time for the reactor vessel head removed, i and the reactor vessel upper internals removed. It is not expected that these evolutions could be accomplished in less than 24 hours.
- 6. - An average of 1.8 percent of the noble gas activity and 0.32 percent of the halogen
" -- activity is in the fuel rod plenums and availabs for release. This assumption is based on fission product release data from defecti"e fuel experiments.
- 7. Due to the negligible particulate activity available for release in the fuel plenums or i from the unmelted fuel, none of the solid fission products are assumed to be released form the fuel.
1
- 8. One hundred twenty-five fuel rods are assumed to fail. This was the conclusion of the analysis of mechanical damage to the fuelin Section XIV of the CNS USAR. l W1229301-5630010695
} [ t /f /W %g T //S/4(l JOBNO. 122-93-01 pact l cALCrn C122-93-01-01 _ 43 ERIN
CALCULATIEN SHEET
- 9. The fission product activity released to the secondary containment will be in proportion to the removal efficiency of the water in the refueling pool. Since water has a poor affinity for the noble gases they are assumed to be instantaneously released from the pool to the secondary containment.
- 10. As noted in Section XIV-6.3.4, the removal efficieacy of the water for halogens can be defined in terms of the partition factor, for which values between 108 and 10 5 have been experimentally determined to be applicable for the conditions under .
2 investigation. A partition factor of 10 for the halogens has been consentatively assumed for this accident. Thus the computed inhalation exposures will be 8 overestimated by a factor of 10 to 10.
- 11. The conservative assumption is made that instantaneous equilibrium is attained betwee1 the refueling pool and secondary containment. In reality, if a true equilibrium is maintained, the effects of plateout or fallout would be compensated for by the evolutions of activity from the refueling pool. Therefore, the effects of plateout and fallout are also neglected.-
- 12. The refueling cavity liquid volume is 3.0 x 10' ft' and the effective air volume in the secondary containment is 7.95 x 105 ft'.
- 13. The maximum standby gas treatment system (SGTS) flow rate of 1780 CFM is assumed, although the CNS USAR specifies the SGTS removes one secondary b
containment air volume per day, which is equivalent to 552 CFM.
- 14. High radiation levels in the reactor building exhaust plenum will isolate the normal reactor building and MG set ventilation systems, and actuate the standby gas treatment system. It is assumed that it takes approximately one minute to isolate the reactor building. During the period, full exhaust flow from the operating reactor building roof. ,
l
- 15. The relative humidity in the secondary containment is 70 percent [ Reference 10].
Since the refueling accident does not result in the release of any liquid or vapor to the secondary containment, the normal environmental condition existing prior to the accident will also exist after the accident, except for the addition of the released fission products. The relative humidity in the secondary containment will therefore i be considerably below any levels which may be dNrimental to the filter media in i the Standby Gas Treatment System. However, the SGTS charcoal beds and I absolute filter media, as well as the air flowing through the filter system, are heated I 10'F above the mixture entering the system, reducing the relative humidity 'o 70% or less. i W12293015630410695 } f $ 1/f/ff W7 //4/9 d J00 NO. 122-93-01 PAGE l CALC NO.C122-93-01-01 44 ERIN' ENGINEERING AND RESEARCH. INC.
CALCULATIEN SHEET
6.0 REFERENCES
1 l
- 1. Stone & Webster Letter No. NPPD-31-02, " Review of Control Room Dose l Calculations, Cooper Nuclear Station", H.F. Faery to G.S. McClure, December 11, 1989, preliminary.
- 2. Stone & Webster Calculation 13095.16-PR(D)-002, " Control Room Doses due to intake and inleakage of Contaminated Air." ,
- 3. Stone & Webster Calculation 13095.16-PR(D)-003, "LOCA Cloud Doses to Control Room."
- 4. Stone & Webster Calculation 13095.16-PR(D)-005, " Direct Control Room Doses l due to 10" Core Spray Pipe Following LOCA."
- 5. Stone & Webster Calculation 13095.16-PR(D)-007, " Direct Control Room Doses due to Reactor Building Shine."
- 6. Stone & Webster Calculation 13095.16-UR(D)-001, " Doses in Computer Room due to a LOCA." l
- 7. Control Room Habitability Study. December 11, 1980, Stone & Webster )
Engineering Company. )
- 8. NPPD Letter CNSS867070, "NPPD Response to NRC Request for Information i Concerning Control Room Habitability," J. Pilant to J.E. Gagliardo, January 30, 1986.
I
- 9. NRC SER_for the FSAR issued. February 14,1973, Sections 6.2.2, 9.4, and 15.2. l 1
- 10. Cooper Nuclear Station USAR.
- 11. NPPD Interoffice M'emorandum, " Emergency Control Room Ventilation Radiation Monitor," A.D. Sutton to J.R. Hotovy, April 22,1987. ,
l
- 12. Stone & Webster Calculation 15798.15-UR-001-0, " Integrated 30-Day Post LOCA Control Room Filter Dose."
- 13. Cooper Nuclear Station Surveillance Procedure 6.3.17.18, " Control Room Envelope Pressurization Test", Revision 0.
i W1229301-5630410695 6 ,[ . I 1/9/9 d JOBNO. 122-9 -01. -PAGE 1 ,
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l CALC NO. C122-9341-01 45 i i i
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CALCULATif!N SHEET
- 14. Cooper Nuclear Station Sunteillance Procedure 6.3.17.5, " Emergency Fan Charcoal i Leak, Charcoal Sampling, and Fan Capacity Test," Revision 11.
- 15. GE NEDO-24782,80NED006,"BWR Owner's Group NUREG-0578 Implementation:
Analyses and Positions for Plant-Unique Submittals," August 1980, Chapter 6.
- 16. Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Ught-Water-Cooled Nuclear Power Plants", Revision 2, March 1978.
- 17. Burns and Roe Drawing 2037, " Flow Diagram, H&V Standby Gas Treatment and Off Gas Filters," Rev. N38, NPPD Document #454003630.
- 18. Burns and Roe Drawing 2019, " Flow Diagram, Main Control Room, Cable Room, and Computer Room, Heating, Ventilation, and Air Conditioning," Rev. N17, NPPD Document #454003607.
- 19. Cooper Nuclear Station USAR, Volume I, Table 1-7-1.
- 20. EDS Nuclear Report 01-0840-1115, Rev. O, " Cooper Nuclear Station Environmental Effects Due to Pipe Rupture."
bl 21. Standard Review Plan 6.4, Section bl.3.d.(2).
- 22. ERIN Engineering and Research, Inc. Report No. TR122-90-09-01, " Cooper Nuclear Station Control Room Habitability issues - Independent Review," March 1990.
- 23. ERIN Engineering and Research, Inc., " Post Accident Design Dose (PADD),
Version 1.00 User Manual," 7evision 0, January 1994.
- 24. NUREG-0588, Rev 1, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."
- 25. Regulatory Guide 1.52, Rev. 2, " Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Absorption Units of Ught, Water-Cooled Nuclear Power Plants."
/
- 26. NUREG/CR-0009, October 1978.
- 27. Marks' Standard Handbook for Mechanical Enaineers. Eighth Ed., Baumeister, Availone, Baumeister.
W12293015630410695 6' fAl) _, I /f /f( ~71)/) 7 //1/ff"] JOB NO. 122-93-01 PAGE l CALC NO.C122-9341-01 46 O OF ENGINEERING AND RESEARCH, INC.
CALCULATION SHEET l
- 28. 13th AEC Air Cleaning Conference, " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19, K.G. Murphy and Dr. K.M.
Campe.
- 29. ANS/SD-76/14, July 1976, A Handbook of Radiation Shieldina Data. J.C.
Courtney, Ed.
- 30. Nucl. Sci. Ena.. Vol. 73 (1980) p 97 " Photon Point Source Buildup Factors for Air, Water, and Iron," Chilton, Eisenhauer, Simmons.
- 31. Regulatory Guide 1.3, Rev. 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors."
- 32. " Chart of the Nuclides," 11th Ed., Knolls Atomic Power Laboratory.
- 33. Regulatory Guide 1.25, Rev., " Assumptions Used for Evaluating the Potential I Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and l Storage Facility for Boiling and Pressurized Water Reactors."
- 34. NUREG-0800, Rev. 2, Standard Review Plan. Sec 6.4 " Control Room Habitability System."
h 35. TID-24190, July 1968, Meteoroloav and Atomic Enerav 1968. W.S. AEC
- 36. NUREG/CR-5055, PNL-6391, May 1988. Atmosoheric Diffusion for Control Room ;
Habitability Assessments. J.V. Rausdell, Battelle PNL i
- 37. U.S. NRC Standard Review Plan 6.4, Rev. 2, July 1981.
- 38. ICRP (International Commission on Radiological Protection) Publication 30 -iodine inhalation PCFs.
- 39. Regulatory Guide 1.109, Revision 1 (pages 1.109-21 thru 1.109-22).
- 40. WPPSS WNP-2 Controf Room Dose Calculation NE-02-88-27, Revision 2, April 14, ,
1992. l
- 41. Southern California Edison Company, Summary of Meeting on June 24,1986, Regarding Control Room Habitability Requirements, Docket No. 50-206, July 9, i 1986.
wm9301562010695 l } f fAM i /f /$f '77l Q 7 t/9/1(l JOBWO. 122-93-01 l CALC NO.C122-9341-01 PAGE 47 ERIN
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CALCULATION SHEET
- 42. Stone & Webster Calculation 13095.16-PR(D)-001, " Dispersion Coefficients (X/Os) h at the NPPD Cooper Nuclear Station Control Room for Accidental Stack Releases of Radionuclides."
- 43. Burns and Roe Drawing 2020, " Flow Diagram, Reactor Building H&V," Rev. N25.
- 44. NUREG-1465, Accident Source Terms for Ucht-Water Nuclear Power Plants. Draft for Comment, June 1992.
- 45. Burns & Roe HVAC Calculation Book 1.
- 46. Drawing C93877GA, Sheet 1, Rev. N01. l 1
- 47. Burns & Roe Drawing 4506, "Roef Plan and Details", Rev. N26.
- 48. Standard Review Plan 15.6.5, Revision 2, July 1981.
- 49. NPPD Calculation NEDC 94-176, " Radiological Dose Consequences of ECCS Leakage During a LOCA",8/29/94.
- 50. CNS STP 94-199, " Control Room Envelope Unfiltered inleakage Test",8/5/94.
- 51. CNS DC 94-102, " Standby Gas Treatment System Cross-tie Valve Modification / Heater Setpoint Change",1/5/95. i l
i l l l I i W1229301-5630410995 , I h b i[T /9I MO 7" " ([7[9 JOB NO. PAGE 48 l mc NO. C122-93-01-01 EitlN* ENGINEERING AND RESEARCH, INC. l
CALCULATION SHEET 7.0 NOMENCLATURE The variables used in developing the PADD equations are defined as follows: a Operator distance from control room filters A A subscript denoting control room emergency bypass filter A, Activity on the control room emergency bypass filters A, Internal surface area of the drywell B Buildup factor B, B,' B," B" Constants of Integration b, Number of mean free photon paths in shielding C,1, Concentration of isotope i at times zero in containment C ii Containment concentration of isotope i C 2i Reactor building concentration of isotope i Cs Control room concentration of isotope i C;4 Air intake concentration of isotope i. As discussed in section 3.4, this can be at any, or all, of a number of points. Os Net concentration of filtered and unfiltered isotope from intakes d Control room roof thickness ! di Reactor building wall thickness d2 Control room wall thickness , D Total operator gamma dose l DA Total gamma dose from control. room emergency bypass filter ! gh Dr Total thyroid dose D, Total beta dose D, Total dose from control room gammas D2 Total gamma dose from reactor building shine D, Total operator dose rate D'A Dose rate from control room emergency bypass filter Dr Thyroid dose rate D,' Beta dose rate D,' Dose rate from control room gammas D'2 Dose rate from reactor building shine E Gamma energy E, Energy of the m* gamma energy group E, Average beta energy e,i The SGTS efficiency for isotope i es The control room emergency bypass filter efficiency for isotope i e Wt229301&i30410095
) f M ilf [f f MgT 1[T[9 fl JOBNO. 122-9341 pAos l CALCNO. C122-93-01-01 49
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CALCULATION SHEET E3 The expression o,(1_ e,.) . o, ' v, . f, Fraction of halogens which are elemental f, Fraction of halogens which are organic f, Fraction of halogens which are particulate F,(t) Fraction of activity which is on control room filters F,(t) Fraction of activity which is in containment F(t) 2 Fraction of activity which is in reactor building F3(t) Fraction of activity which is in the control room GA Flux to dose conversion factor for control room EFUs G, Flux to dose conversion factor for control room gammas G2 Flux to dose conversion factor for reactor building shine i H A subscript for halogens (brornine and iodine) i A subscript for isotope i I Operator inhalation rate J A subscript for isotope j L Containment height divided byv7 m A subscript for gamma energy group m MAC Mass attenuation coefficient n A subscript for evaluation time n N L A subscript for noble gases . _ h P. P, Peak post-LOCA drywell absolute pressure Standard atmospheric absolute pressure OA Source strength of a plane source ! Oc Leak rate from containment directly to the environment ! Oo Source strength cf a point source ! Oi Leak rate from containment to the reactor building 1 l 0, Flow rate through the SGTS O3 " Fresh Air" flow rate through the control room emergency bypass filter ,
" Fresh Air" flow rate into the control room not through the control 04 room emergency bypass filter Os Control room recirculation rate Q, Recirculation rate through the control room emergency bypass filter Rr Adult inhalation thyroid dose conversion factor R, Gamma flux to dose rate conversion factor ;
S Atmospheric stability class 1 =Pasquill A,2=B, .,6=F,7=G l S;(t) Total activity of isotope i at time t i t Time from start of LOCA to Time at the n"' evaluation time 1 W1229301-5630410695
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- ENGINEERING AND RESEARCH, INC.
CALCULATION SHEET y/Oc Ground release atmospheric dilution factor I 1 The subscript denoting containment or drywell l 2 The subscript denoting the reactor building 3 The subscript denoting the control room B1A Adult inhalation dose total body conversion factor (R.G. 1.109)(rem /Ci) l l B1C Child inhalation dose total body conversion factor (R.G. l I 1.109)(rem /Ci) . B1l Infant inhalation dose total body conversion factor (R.G. l 1.109)(rem /Ci) BRA Adult breathing rate (3.47x10"m'/sec) for control room calcs BRAA Adult breathing rate (R.G.1.109) for LPZ and exclusion area BRC Child breathing rate (R.G.1.109) for LPZ BRI infant breathing rate (R.G.1.109) for LPZ EXAn(t) Whole body halogen dose rate to adults in exclusion area EXAu(t) Whole body noble gas dose rate to adults in exclusion area EXAw,(t) Whole body dose rate to adults in exclusion area EXAr Adult thyroid dose rate in the exclusion area EXAT1 Total adult thyroid dose rate m the exclusion area DF Noble beta skin dose conversion factor (mrem-m') / (pCi-yr) i LPZ tn(t) Halogens (curies /sec) elevated release offsite LPZ tn(t) Nobles (curies /sec) elevated release offsite LPZon(t) Halogens (curies /sec) ground release offsite b LPZon(t) LPZg(t) Nobles (curies /sec) ground release offsite Whole body halogen dose rate to adults in LPZ j LPZnc(t) Whole body halogen dose rate to children in LPZ j LPZw(t) Whole body halogen dose rate to infants in LPZ LPZu(t) Whole body noble dose rate to adults in LPZ LPZuc(t) Whole body noble dose rate to children in LPZ ; LPZw(t) Whole body noble dose rate to infants in LPZ i LPZw,(t) Whole body dose rate to adults in emission area _.LPZr Adult thyrold dose rate in the LPZ LPZe7 Child thyroid dose rate in the LPZ LPZn Infant thyroid dose rate in the LPZ LPZTy Total adult thyroid dose rate in the LPZ LPZTer Total child thyroid dose rate in the LPZ LPZTg Tc al infant thyroid dose rate in the LPZ RH Ha. ogen conversion (Rem /Ci inhaled) from NRC ICRP30 y/Ouzt Dispersion (sec/m 8) for LPZ from elevated rel ease y/Ou>za Dispersion (sec/m*) for LPZ from ground release i X/Ocxe Dispersion (sec/m*) for exclusion area elevated release y/Oao Dispersion (sec/m 8) for exclusion area ground release W1229301-5630410695 J 5~ Pub i h /<r 971g r U9/nlmena l CALC NO. 1224341 C122-9341-01 race 52 w a,~ . ,~c.
CALCULATl"N SHEET l it is noted that leakage rates, HVAC flow rates, y/O factors, and penetration efficiencies are not necessarily constant over time. Consequently, the variables below are defined 1 to allow development of equations which incorporate this potential time dependency. First, assume that the entire time period over which doses are to be calculated is divided : into the following intervals: I INTERVAL n DEFINITION 1 t,s t s t, 2 ti stst 2 e e l e e l e o N tu .,ststu l Second, assume that these intervals are defined such that within each time interval, all HVAC flow rates, leakage rates, y/O reactors, and filtration efficiencies are constant. We can then define the airborne concentration of an isotope i at the end of a given time interval as the initial Condition for the solution expressing that concentration in the next interval: Cu(t,,.3) e airborne concentration if isotope i inside of containment at the end
) of the (n-1)st time interval C2i(t,3), C3 ;(t,,) are similarly defined for the reactor building and control room, respectively I
Other variables defined previously (e24,s e ;,3 E ;, Oe, 0 , 20 , 03 , 04 , Os, O., 6, 5, 6, X/O, 3 X/Oo,y/Opzt,y/Otnc, t X/Ocxe, y/Ocxa r, x, e) may appear with a subscript n, denoting the constant value of the variable in the interval n. l l 1 W1229301-5634010695
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CALCULATION SHEET 8.0 CALCULATIONS 8.1 CALCULATION OF X/O FACTORS FOR TURBINE BUILDING RELEASE The model used in this calculation for radionuclide release from the turbine building is the New Building Wake Model from Reference 36. The equation that represents this model, along with a description of how the X/O values for CNS were developed, is shown below: 1 X/O = 100(xA)i.2(U).es(g).s 1 where X/O = relative concentration (s/m') ; U = windspeed (m/s) j x = distance between release and receptor (m) l 2 A = building area (m ) l S = numeric stability class identifier (G =7, F=6, etc.) l Given that the turbine building center is roughly southeast of the control room ventilation intake, and considering the relative heights of the source (center of turbine building roof) and receptor (control room ventilation intake), winds from three sectors (SE, ESE and SSE) were considered most likely to impact the control room. Meteorological data for these sectors at a 50 meter elevation [ Reference 10, Chapter lil) were used in the
- calculation of X/O values. The frequency of each windspeed range was normalized with respect to the total frequency of winds from the SE, ESE and SSE.
The selection of X/Os for the analyses required determining the frequency of exceeding certain values during the four time steps in the analysis. These values are the 95% conditions for 0-8 hrs,90% conditions for 8-24 hrs,80% conditions for 1-4 days and 60% conditions for 4-30 days. In these analyses, the representative percentile conditions were W1229301-5630410695
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CALCULATION SHEET interpreted as percentile X/Os. This required the calculation of X/Os for each windspeed and stability class combination. The X/O values were then sorted in descending order to determine the representative percentile conditions. For example, the 95% X/O is the X/O value that is exceeded only 5% of the time. The following table summarizes the representative X/O values and meteorological conditions determined using the New Building Wake Model. Time Percentile Meteorological New Building Wake Condition Model X/O 0-8 hr 95% D/24 mph 1.52E-03 8-24 hr 90 % E/19 mph 1.45E-03 1-4 days 80 % F/13 mph 1.23E-03 4-30 days 60% D/13 mph 1.00E-03 In addition to determining the representative windspeed and stability class combinations, l
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@ a wind direction credit after eight hours was taken in accordance with Reference 28. This credit accounts for the variability of the wind direction over a period of time. For 8 to 24 hours, this factor is 0.88. For 1 to 4 days, the factor is 0.75 and for 4 to 30 days it is 0.5. The net X/O values incorporating the wind direction factors are shown in the following table. Time New Building Wake Correction Model Fador Net X/O 0-8 hr 1.52E-03 1.00 1.52E-03 8-24 hr 1.45E-03 0.88 1.28E-03 1-4 days 1.22E-03 0.75 9.23E-04 4-30 days 1.00E-03 0.50 5.00E-04 W1229301-563OO10695 } f b_b i /9 /fi' %QT //1/f d JOBNO. 122-93-01 PAGE l CALC NO.C122-93-01-01 5.5_. e~a,~ . ,~c.
CALCULATIPN SHEET l l l The spreadsheets used to develop the raw X/O (not incorporating the wind direction correction factor) and the meteorological data used in the analysis are shown in Table 8-1. The X/O values selected are highlighted. Table 8-1 X/Os FOR ALL STABILITY CLASS /WINDSPEED COMBINATIONS NEW BUILDING WAKE MODEL Dstance Between Source f. Receptor = 52 m Effective Building Area = 1366 m' Stability Windspeed X/o Cumulative Cass (mph) (s/m') Frequency Frequency 7 24 2.01E43 0.00 0.00 6 24 1.86E-03 0.00 0.00 7 19 1.71E43 0.00 0.00 5 24 1.70E43 0.03 0.03 6 19 1.59E43 0.01 0.04 gp, y. l p'4 : i 5.h2E.U3
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7 13 1.32E43 0.02 0.13 3 24 1.31E-03 0.00 0.13 4 19 1.29E-03 0.05 0.17 a 6- > k135 , i1.22E43 , I. '0.05 f , ' $U.22l' i .4 3 19 1.12E43 0.00 0.22 5 13 1.12E-03 0.11 0.33 2 24 1.07E43 0.00 0.33
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[ ;l4 ?.. L w. 13 '?;- i c $1.00iN3i ['S l 0.114 , w! di5'd 7 8 9.51E44 0.02 0.47 2 19 9.15E 04 0.00 0.47 6 8 8.80E44 0.03 0.50 3 13 8.66E44 0.01 0.51 5 8 8.04E44 0.05 0.57
1 24 7.58E44 0.00 0.57 W12293015630410695 l} S~ 94b ih kr Jr]p r t/9/9rlxeNa 122-east eAaE l CALC NO. C122-934101 56 ElvGNE R . WC. _
CALCULATION SHEET NEW BUILDING WAKE MODEL Distance Between Source & Receptor = 52 m Effective Building Area = 1366 m' Stability Windspeed X/O Cumulative Class (mph) (s/m') Frequency Frequency 4 8 7.19E44 0.09 0.66 2 13 7.07E-04 0.01 0.67 1 19 6.47E-04 0.02 0.69 3 8 6.22E-04 0.01 0.69 7 4 5.93E-04 0.01 0.71 6 4 5.49E44 0.02 0.72 2 8 5.08E-04 0.01 0.73 5 4 5.02E-04 0.02 0.75 1 13 5.00E44 0.06 0.81 l 4 4 4.49E-04 0.04 0.85 l 3 4 3.88E-04 0.00 0.85 l 1 8 3.59E44 0.08 0.94 2 4 3.17E44 0.00 0.94 l 7 1 2.31E44 0.00 0.94 1 4 2.24E-04 0.04 0.99 6 1 2.14E44 0.00 0.99 5 1 1.95E44 0.00 0.99 4 1 1.75E-04 0.00 0.99 3 1 1.51E44 0.00 0.99 2 1 1.24E44 0.00 0.99 1 1 8.74E45 0.01 1.00 7 0 0.00E + 00 0.00 1.00 6 0 0.00E + 00 0.00 1.00 o D 0.00E + 60 0.00 1.00 l l 4 0 0.00E + 00 0.00 1.00 l 3 0 0.00E + 00 0.00 1.00 2 0 0.00E + 00 0.00 1.00 1 0 0.00E + 00 0.00 1.00 W1229301-5630410695
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CALCULATION SHEET 8.2 RESULTS f As previously stated, all calculations were performed using the PADD computer program. l The pertinent input parameters are shown in Table 8-2 for all runs performed. Results for the design basis or " base case" calculations for both the LOCA and refueling accident l scenarios are presented in Table 8-3. Additional case study results are also presented i in Table 8-3 for sensitivity analyses defined by the NPPD staff. The PADD computer j program input and output files for all case studies performed in this calculation are l presented in Appendix B. 1 l l 1 i f) r W1229301-5630410695 ) 5 M i/f/6 'M2 /J7 //#1/'Ifl JOB NO. 122-93-01 PAGE l CALC NO.C122-93-01-01 58 ERIN' ENGINEERING AND RESEARCH. INC.
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CALCULATION SHEET Table 8-2
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CASE STUDY INPUT PARAMETERS Control Room Control Room Primary Primary Reector Control FWtered intake Unfiltered Containment Containment Bui6 ding Room Leekege Flow Rete in8eekage Leek Rete SOTS Flow Volume Volumn Volume Run # Path (CFM) (CFM) (vol %/ day) Rete (CFMI (ft*) (ft*) (ft") LOSS OF COOLANT ACCOENT CASES to Primary 375 100 .646 1780 239100 1 141860 lb MSIV 375 100 .092 .243 239100 1000000 141860 3e Primary 375 200 .646 1780 239100 1 141860 3b MSIV 376 200 .002 .243 239100 1800000 141860 6e Primary 1000 100 .646 1780 239100 1 141860 Eb MSIV 1000 100 .092 .243 239100 1000000 14184,0 7e Primary 1000 200 .646 1780 239100 1 141860 7b MSIV 1000 200 .092 .243 239100 1800000 1410'A 9e Primary 2000 100 .646 1780 239100 1 141800 Ob MSIV 2000 100 .092 .243 239100 1800000 141860 its Primary 2000 200 .646- 1780 239100 1 141860 11b MSIV 2000 200 .092 .243 239100 1800000 141860 REFUfLINO ACCOENT CASES 13 FHA 376 100 N/A 1700 239100 795000 141860 14 FHA 375 200 N/A 1780 239100 795000 141860 15 FHA -1000 100 N/A 1780 239100 795000 141860 16 FHA 1000 200 N/A 1780 239100 795000 141860 W1229301-5630 010G95 } T M i /9 /6 KST l l/1/TfIJOBNO. 122-93-01 pace l CALC N0. C122-93-0141 59 O OF ENGINEERING AN!) RESEARCH. INC.
CALCULATION SHEET j Table 8-3 i CASE STUDY RESULTS I Operator Operator Operator Laekage Path Thyroid Dose Whole Body Dose Bata Skin Oose Run# irem) (rem) (rem) LOSS OF COOLANT ACCOENT CASES l 1a Primary 4.98 4.10E-02 4.07E41 l tb MSN .39 4.26E-05 1.14E 03 i TOTAL 5.37 4.18E-02 4.08E 01 3a Primary 7.12 4.56E-02 4.31 E-01 3b MSN .56 4.76E-05 1.20E-03 l TOTAL 7.68 4.56E-02 4.32E-01 6a Primary 3.66 5.90E-02 5.49E 01 I Sb MSN .25 4.31 E45 1.20E-03 TOTAL 3.91 5.90E-02 5.50E-01 7a Primary 4.64 6.09E 02 5.68E-01 7b MSN .35 4.53E-05 1.22E-03 TOTAI 5.19 6.09E-02 5.69E-01 9a snary 3.09 7.02E 02 6.75E-C1 $l Db MSN .20 4.33E 05 1.21 E-03 TOTAL 3.29 6.77E 01 7.03E+2 _ 11a Primary 3.78 7.09E-02 6.85E 01 11b MSN 25 4.46E-05 1.22E-03 _ TOTAL 4.03 7.09E42 6.86E-01 REFUEUNO ACCOENT CASES 13 FHA 1.90 1.86E-01 1.80 14 FHA 2.40 2.04E-01 1.94 15 FHA 1.69 2.67E 01 2.53 16 FHA 1.95 2.75E 01 2.61 W1229301-5630410095 } F l@ f/4/rf % ,9 ?~ //1/17l JOB NO. 122-93-01 PAGE l CALC NO.C122-9341-01 60 ERIN
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CALCULATION SHEET 1 APPENDIX A CNS MSIV LEAKAGE PATH ASSESSMENT l l t
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CALCULATION SHEET MSIV LEAKAGE PATH ASSESSMENT COOPER NUCLEAR STATION REVISION B
1.0 INTRODUCTION
.The Cooper Nuclear Station (CNS) main steam system design includes Main Steam isolation Valves (MSIVs) on the exit of the primary containment. These valves are arranged in series in ,
four twenty.four inch steam lines for a total of eight (8) valves. These valves function to isolate
' main steam following a plant trip and, among other things, form a portion of the primary containment boundary for post-LOCA retention of the radionuclide inventory.-- As part of the CNS post-LOCA control room operator thyroid dose calculation, it has been determined that.
some contribution of the source term will bypass the primary / secondary containment features via design basis leakage through these valves. A conservative calculational assumption could , be made that all of the .: antainment bypass leakage through these pathways is released to the turbine building and is available for transport from this building to the control room. However, l it is desired to develop a technical basis for the consideration of retention;of much of the , released source term in the BOP systems.
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. The CNS MSIV configuration is not unique among BWRs. In all BWRs the MSIVs are credited to provide a post-accident fission ' product barrier. The issue of MSIV leakage was reviewed j 7 !
'for CNS by the NRC at the time.of plant licensing .and it was concluded by the NRC that, isubject to.NPPD performing. testing and monitoring.of_ MSIV, leakage,.the configuration _was u ~
i acceptable when the small probability.of the postulated accident conditions.concurre.nt with the } f J ailure.of the main steamlines_o,utside of con _tainme.nt or the turbine . condenser. and b_ecause ; oof t.he_ conservative nature _of the. staff's analysis _of dose consequences. The issue of MSIV l leakage and control was later identified as an industry concern .by the NRC because of _ . l l jincreased MSIV leakage seen at several BWRs. ; As a result the NRC evaluated the issue, nissued Reg. Guide 1.96 and re, quired certain BWRs to install a leakage control system (LCS)
- for MSlV leakage control. At the time of NRC study of this phenomena CNS was not requirsd .
~ ~
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Rto install 5uch a 's si5m. ~ This mav be due to historical tested MSIV leaksge th'at was less than ~~
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'the Technical Specification value of 11.5 SCFH' for a total of 46' SCFH'(See. History"in
~ -- ~
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"Attschts nt e A)T The NRCin'd iridulitry slUdieE28~0ndertaksn since that tirn6 conberned ~
themselves with plants that were postulated to have leakages in excess of the technical specification limits. The documentation does not provide clear guidance on the consideration - of the MSIV leakage in offsite and onsite dose calculations for plants that have no special
.means to collect or mitigate the MSIV leakage source term.
In previous CNS dose calculations the MSIV leakage was assumed not to add to the source-term available for transport. The most recent calculation' assumed the MSIV leakage was 1 filtered by the standby gas treatment system (SGTS). The phenomena has been evaluated for other BWRs of a similar design" and it was determined that the low postulated MSIV leakage- - W1229301-5630-042294
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CALCULATION SHEET was heldup in the mainsteam systems for a time period that minimized its participation in the i offsite doses. In the case of plant Hatch, the NRC concluded that the BOP systems (turbine and condenser) sufficiently hold up the source term, that the allowable MSIV leakage rate was increased to 100 SCFH per valve and 250 SCFH total. It is concluded that such logic also
' applies to the onsite dose received by the control room operators. However, if such credit is to be taken it must be explicit and the post-accident configuration of the CNS mainsteam system components understood. The evaluation that follows is such an assessment.
2.0 TECHNICAL ASSESSMENT IThe discussion above determines that CNS may assume holdup or limitations on the amount iof MSIV leakage if the post-accident . status of the main steam and turbine control systems is understood. Therefore, an evaluation of leakage pathways and holdup in the turbine system will be undertaken. This evaluation will include consideration of:
- 1) post-turbine trip status of BOP systems,
- 2) flow rates as a function of time, and
- 3) release points.
This review scope is adopted from a the criteria in SRP section 15.6.5, Appendix D,
- Radiological Consequences of a Design Basis Loss-of-Coolant-Accident: Leakage from a Main Steam Line isolation Valve Leakage Control System." CNS does not have a leakage control
- system (LCS) for.MSIV. leakage collection, but the assessment methodology is applicable. .The Q .following is an assessment for CNS.
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2.1 Evaivation of Leakaae Potential l _ : Burns and Roe P&lD's' and main steam isometrics for CNS were reviewed to identify all _ potential leakage. pathways._the. pathways are. identified in the . table in subsection 2.2. These_ paths represent a potential " chamber " for containment of.the MSIV leakage source term. An. effective containment yolume_for.MSIV leakage _ propagation was calculated based on the-iReference 1 drawings, as.sociated isometric dr.awings and standard pipe size dimensions This_
' volume is estimated to be approximately 5,000 ft'.
Considering the maximum MSIV leakage rate of 46 SCFH, it will take approximately 108 hours (4.5 days) for the entire volume to fill with the postulated MSIV leakage. At th'at time, simplistic
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coii~ sideration of the scenario would predict that continued leakage would be expected to begin" l
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1 to pressurize th'e system and cause leakage into the turbine building. However, the condenser
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volume and condensation of steam'is expected to have an Onquantified, but not insignificant ~ effect on the pressurization of the system. In addition the scenario considered here assumes the MSIVs are leaking at their nominal Technical Specification leakage rate. Lower MSIV leakage would extend the time available before pressurization would begin. It is also assumed
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1 1 CALCULATION SHEET l l that the off-gas systems for the processing, filtering and releasing of gaseous radwaste through
, the elevated release point are not available. However, these systems will in all likelihood be available or capable of being restored. Processing of the source term by these systems would minimize the source term as well. This evaluation also neglects the plate-out processes
- modeled in References 2 and 3 which showed significant source term reduction during the holdup time in the main steam system. Therefore, if containment within this volume can be deterministically shown, it can be postulated that the source term is held up in the BOP systems at CNS.
2.2 Evaluation of Leakaae Pathways-Burns and Roe P&lD's for CNS were reviewed to identify all potential leakage pathways for CNS. Each pathway was. considered for the following:
- 1) Destination,
- 2) Post-LOCA/ LOOP (if worst case) configuration,
- 3) Potential fraction of the source term assigned to pathway.
The assessment assumes that a flow path destination of the condenser will contain the source , term and preclude leakage. The assessment combines LOCA/ loss of offsite power (LOOP), but does not evaluate the seismic ruggedness of these systems as was done for Hatch". This 7 is acceptable for the reasons discussed by the NRC in the CNS SER . The following is a table of results of that assessment.
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W1229301-5634042294 i l Md // D NO. PAGE C122-93-01-01 A-4 l CALC NO. ERIN * ; ENGINEERING AND RLSEARCH. ING,
CALCULATION SHEET CNS POST-ACCIDENT MSIV LEAKAGE PATHWAYS No. Line Description Post-Trip Condition Disposition Number 1 3*-MS-1 S 3' main steam line MS-MOV-MO78 is closed The path is normally isolated. drain via MO-78 to the in both the normal condenser. operating configuration if open the final destination is and emergency post-trip the condenser. configuration. 2 30*-M S-1 Main steam lines to Turbine stop valves MS- This pathway is isolated as the high pressure HOV-SV1 and SV2 and part of post-trip actuations. turbine via SV1BV & turbine control valves MS-SV2BV HOV-GV1, GV2, GV3, and Destination is the high GV4 close on a turbine pressure turbine chest. In trip. the event of locomplete Isolation, seal leakage would be expected. 3 18*-MS-1 Turbine bypass to Turbine bypass valves MS- Post-trip condition is isolated. condenser 1 A & 1B HOV-BV1, BV2 and BV3 via turbine bypass close in the absence of in the event the valves valves HO-BV1(2,3). steam pressure, remain open, the final
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destination is the condenser. 4 2*-M S-1 Main steam to the Pressure control valves The event postulated is'a steam Jet air ejectors MS-AOV-PCV77A and B LOCA combined with a via BV-1 A(B) and do not close unless there LOOP which would cause a PCV-77A(B) to 3*-BS-2 is a loss of power or air. loss of power. However, the worst case is that post-trip availability of air and power, resulting in no isolation and a potential pathway for release.
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Final destination is the . - condenser and for the low I pressure turbine. 5 5'-M S-1 Turbine Gland and The turbine gland seal Final destination is the 1 Condenser sealing system check valve MS- condenser. This holds up i steam via PCV-68A CV-18CV is closed in the the source term. j and MO-lMV3(BMV3). absence of steam ~ pressure. j W1229301-s630.o42294 M N[ T///df JOB NO. 122 M 1 PAGE A5
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CALCULATIEN SHEET No. Line Description Post-Trip Condition Disposition Number 6 2*-M S-1 Turbine bypass line Trap bypass valve MS- Final destination is the steam trap 15. AOV-195AV is closed condenser. This holds up during all conditions and the source term. falls closed. The trap allows lira to drein
. condensate.
7' 2"-MS-1 Turbine bypass Valves do not isolate line: Final destination is the stralners draln. however, stralners dP condenser. This holds up would not allow flow if no the source term. steam pressure present.
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3.0 CONCLUSION
The above assessment evaluates all flow paths out of the CNS BOP systems that could result in release of the post-accident source term leaking through the MSIVs. A LOCA/ loss of offsite power (LOOP)~ is ~ considered and consequential effects considered, as applicable. The
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evaluation concludes that containment within the BOP systems is deterministically shown and the migration of the MSIV leakage source term within those systems is a slowly acting phenomena that will not challenge the integrity of the system. Therefore, it can be postulated that the source term is held up in the BOP systems at CNS.
4.0 REFERENCES
1.0 CNS Drawings: P&lDs Burns and Roe Drawing 2002, Sheet 1 of 3, Rev. N29, " Flow Diagram Main Exhaust & Auxiliary Steam Systems". Burns and Roe Drawing 2002, Sheet 2 of 3, Rev. N25, " Flow Diagram Main Exhaust & Auxiliary Steam Systems".
. Burns andhoe brawing 2002,' Sheet 3 of 3, Rev. N07, ." Flow Diagram Main Exhaust & ) Auxiliary Steam Systems".
Burns and Roe Drawing 2041, Rev. N53,." Flow Diagriam Reactor Bldg. Main Steam System". L . , _ lsometricsj_ . _. _ _ _ . . _ __ _. __. , _ .__. . __ .. _
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JEL60 lncorporaiedDrawing.2'809-5, Rev. N01,'BS-2 Pipe to Jet". JELCO Incorporated Drawing 2841-1, Rev.10, " Main Steam Supply From Turbine to 36" Header, Turbine Building". JELCO Incorporated Drawing 2841-2, Rev. NO3, "MS-1 Unes From MS-136" Header,
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Turbine Building". _ L]. __ g ~[ JELCO Incorporated Drawing 2842-1, Re'v. N01, "MS-1 Un'es From Turbine Bypass Valve ~ to Condensers".
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] DATE l l
CALCULATISN SHEET 2.0 NUREG-1169, " Resolution of Generic issue C-8, An Evaluation of Boiling Water Reactor i Main Steam isolation Valve Leakage and the E#ectiveness of Leakage Treatment Methods", dated August 1986.
~~3.0 Battelle Memos, Del Lessor to Jim Jamison/ Dennis Strenge,
Subject:
MSIV Project, dated September 6,1985 (Revised Copy), September 16,1985, September 24,1985, l and September 30,.1985 (Ret # 17839 0561).
' 4.0 Stone & Webster Calculation 13095.16-PR(D)-002, " Control Room Doses due to intake and inleakage of Contaminated Air" dated December 11,1980.
'5.0 Duane Arnold Energy Center-1 (DAEC) UFSAR, Section 6.7, " Main Steam Line Isolation Valve Leakage Control System", Revision 2 - 6/84.
6.0 Letter, USNRC to Georgia Power Company, Kahtan N. Jabbour (USNRb) to J. T. Bechham, Jr.,
Subject:
Issuance of Amendment - Edwin I. Hatch Nuclear Plant, Unit 2, dated March 17,1994. 7.0 CNS Safety Evaluation Report, Section 6.2.2 Isolation Systems. dated February 14,1973.
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h/2//\/ fh0 - 0. 0/A2YN'Of> CNS LOCAL LEAK RATE TEST HISTORY ArTACHM6Alfk ro AWauooc4
?_ENETRATION: X--7 A [Mg /of g;[-
DESCRIPTION: Main Steam Isolation Valves - Line A ALLOWABLE _. LET>KAGE t 'CPN)t 3.1. 5 /Va lve - Cl.Q_;. gg AOV-AO_80A MS~AOV-AOR6A A_S FOUND LEAKAGK AS LEFT LFJLKAGE YEAR TOTAL PEN. TOTAL PEN. 1973 4.31 2.16 4.31 2.16 1975 81.63 36.08 1,77 0.89 1976 6.65 3.33 6.65 3.33 1977 5.32 2.66 5.32 2.66
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1978 14.81 4.09 14.81 4.09 1979 9.30 4.65 9.30 4.65 1980 7.98 3.99 7.98 3.99 1981 6.21 3.11 6.21 3.11 1982 1.52 0.76 1.52 0.76 1983 13.42 6.71 0.42 - 0.21 - 1984 11.15 5.58 11.15 5.58 1986 2.34 1.17 2.34 1.17 1988 10.86 5.43 10.86 5.43 1989 28.15 6.68 2.37 1.19 1990 0.153 0.0765 -
- 0. 1 5 3 0.0765 1991 1.66 0.83 1.66 0.83 1993 1 32 0.66 7.56 3.78
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$00/Z0091 G3N 09 *** SN3 1Y GJdM ITZS stb E0tD, PC:0T t6/SO/to
hk/k Wo 6, l Otzz- 6 -ot-at CNS ' LOCAL LEAK RATE TEST HISTORY *!MA 7p A1$6MO/XA 0 ' 1 MS
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PENETRATION: X .7 B DESCRIPTION: Main Steam Isolation Valves - Line B , ALLOWABLE LEAuGE.fSCFH): 11.S / VaIve - , l
.Cl9_;. MS-AOV-A080B
- l Ms-AoV-Aossa i
AS FOUND. LEAKAGE AS LEFT LBAKAGE l IEh3 ' TOTAL PEN. TOTAL PEN. 1973 2.43 1.22 2.43 1.22 i 1975 13.82 4.78 13.82 4.78 i-1976 11.49 5.75 11.49 5.75 1977 12.89 6.45 10.34 5.17 1978 2.29 1.15 2.29 1.15 l 1979 6.60 3.30 6.60 3.30 1980 8.50 4.25 8.50 4.25 ) 1981 8.45 4.23 8.45 4.23 1982 12.94 3.03 12.94 3.03 1983 8.37 4.19 8.37 8 4.19 - 1984 9.37 4.69 3.66 1.83 . 1986 2.45 1.23 [2.45 1.23 1988 6.54 3.27 6.54 3.27 1989 5.75 2.88 0.84 0.42 1990 0.229 .01145 0.229 0.1145 1991 0.90 0.45 0.90 0.45 j i 1993 2.92 1.46 4.27 2.135 l S00/C00@ G3R OD *** SN3 1Y GddN IIZS SZ8 20tG SC:01 t6/SO/to l
BRm/ Cud. Alo. l C/2k4.5-0(-oj l CNS LOCAL LEAK RATE TEST HISTORY Arrita/M&ufA 7D k?#9&AlpixA ,
' l PENETRATTON: X-7C [gg gy4 l l
DESCRIPTION: Main Steam Isolation valves - Line C . l ALLOWABLE LEAKAGE fscFH): 11.5_/ Valve - i CTC: MS-AOV-A080C 1 MS-AOV-A086C l AS FOUND LEAKAGE AS LEPT LEAKAGE X.EAB _ TOTAL Peti. TOTAL PEN 2 1973 0.93 0.47 0.93 0.47 1975 8.63 4.32 8.63 4.32 1976 10.72 5.36 10.72 5.36 : 1977 8.24 4.12 8.24 4.12 1978 '6.44 3.22 6.44 3.22 1979 4,50 2.25 4.50 2.25 1980 9.89 4.95 9.89 4.95 1981 8.24 4.12 8.24 4.12 1982 12.09 3.36 12.09 3.66 1983 12.93 6.47 1.60 ' O.80 -' 1984 4.95 2.48 4.95 2.48 1986 5.82 2.91 5.82 2.91 1988 8.42 4.21 8.42 4.21 ; 1989 4.81 2.41 1.59 0.80 1990 1.18 0.59 1.18 0.59 1991 4.16 2.08 4.16 2.08 1993 2.47 1.235 6.61 3.305 900/t 00 Q) 03N OD **+ SN3 1Y GddN ITES SES 20tG, SC:0I t6/80/to
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CNS C/22-43-0f-O/ LOCAL LEAK RATE TEST HISTORY ATTMtAfggr4 h $0l}(k PENETRATION: X-7D gg _ g DEff CRIPTION: Main Steain Tsolation Valves Line D ALLOWABLE LEAKAGE (BCFH): 11.5 / Valve - cIc: MS-AOV-A080D MS-AOV-A086D AS_FOUND LEAKAGE , AS LEFT__ LEAN GE 10TAL PEN._ TOTAL PEN. XEA_B 1973 8.96 4.48 8.96 4.48 1975 14.38 2.89 14.38 2.89 1976 11.88 5. 3 6 t 11.88 5.36 1977 9.23 4.62 9.23 4.62 1978 4.44 2.22 4.44 2. 'e2 1979 8.90 4.45 8.90 4.45 1980 14.20 6.20 6.20 3.10 1981 15.97 7.'54 15.97 7.54 1982 12.19 5.94 12.19 S.94 1983 20.20 10.10 0.0 4 0.0- -- 1984 6.14 3.07 6.14 3.07 1986 2.38 1.19 2.38 1.19 1988 6.77 3.39 6.77 3.39 1989 3.08 1.54 0.0 0.0 1990 6.69 3.345 6.69 3.345 1991 0.48 0.24 0.48 0.24 1993 2.78 1.39 5.04 2.52 900/900Pi 03N OD *** SN3 1Y GddN ITZS 929 20t3 SC:0T f6/90/to
CALCULATION SHEET APPENDIX B PADD OUTPUT FROM CASE STUDIES i The following PADD output reports are included in this appendix: CASE NUMBER RUN NAME RUN DATE/ TIME TOTAL PAGES 1 RUN1A 106-YS /11:24:43 98 2 RUN1B 1 09 % / 12:19:52 98 ; 3 RUN3A 1 06-9, / 11:46:46 98 j 4 RUN3B 1-09-95 / 2:42:45 98 5 RUN5A 1-06-95 / 12:46:31 98 l 6 RUNSB 1-09-95 / 12:50:59 98 7 RUN7A 1 06-95 / 13:35:12 98 8 RUN78 1-09-95 / 12:59:13 98 9 RUN9A 1-06-95 / 13:51:36 98 10 RUN9B 1-09-95 / 13:07:28 98 11 RUN11 A 1-06-95 / 14:08:16 98 1 12 RUN11B 1-09-95 / 13:15:38 98 13 RUN13 1-06-95 / 20:31:45 98 14 RUN14 1-06-95 / 20:36:34 98 g 15 16 RUN15 RUN16 1-06-95 / 20:40:42 1-06-95 / 20:45:51 98 98 Although additional doses (i.e., filter shine, reactor building shine, exclusion area doses, and low population zone doses) are listed in the PADD output, these doses use data that has not
- - been verified and, thus, are not valid.
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