3F0989-01, Discusses Rev to FSAR Radiological Consequences.Changes Establish Revised Licensing Basis for LOCA & Letdown Line Failure Accident Radiological Consequences,Recognizing That Margins Established by 10CFR100 Have Not Been Reduced
| ML20247H630 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/13/1989 |
| From: | Widell R FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F0989-01, 3F989-1, NUDOCS 8909200046 | |
| Download: ML20247H630 (45) | |
Text
_. _. . ._ _ - .
o; '
b-S e Power COR PO R ATION September 13, 1989 3F0989-01 Document Control Desk U.S. Nuclear Ragulatory Commission Washington, D.C. 20555
Subject:
Crystal River Unit 3 Docket No 50-302 Operating License No. DPR-72 Revision to FSAR Radiological Consequences
Dear Sir:
As part of the Configuration Management Program and the Technical Specification Improvement Program, Florida Power Corporation (FPC) has been re-examining the assumptions used in ~the FSAR Chapter 14 analyses. There have been inconsistencies identified between the Technical Specifications and certain FSAR-accident assumptions. FPC is evaluating the inconsistencies to determine if they impact the FSAR results.
In some . cases, additional analyses have been performed to quantify the impact.
To ensure that CR-3 offsite doses remain within 10 CFR 100 limits, FPC has re-evaluated the off-site radiological consequences of the Loss-of-Coolant Accident (LOCA) and the Makeup System Letdown Line Failure Accident (LLFA) to eliminate the credit for the Auxiliary Building Ventilation (ABV) System.
The ABV System contains charcoal filters which reduce the iodine dose. This system is non-safety related and is not i provided with emergency power. Without emergency power, the ABV System should not be assumed to be available to provide iodine filtration.
The analyses for both accidents have assumed the ABV System is l t not available. The LOCA analysis used the same methodology for j l fission product release as that used to evaluate Crystal River ;
L Unit 3 (CR-3) control room habitability submitted in FPC's June f 30, 1987 letter. The control room habitability Safety f
(_ 8909200046 890913
{DR ADOCK 05000302 l FDC I t POST OFFICE box 219
- CRYSTAL RIVER, FLORIDA 32629-0219 * (904) 563-2943 l l
A Florida Progress Company
. - - - . _._ _ - .-_-_ __- __ - __ _ __ _ ___ _ _ _ _ _ _ _ _ _ _ Q
L.o September 13, 1989 3F0989-01 l Page 2 Evaluation Report (SER) was issued by the NRC letter dated May 25, 1989. This fission product model uses Regulatory Guide 1.4 for the upper bound assumptions. The analyses project small dose consequence increases above the values previously reported in the FSAR. A comparison of the doses is presented in the attached tables.
LOCA FPC's re-evaluation for a hypothetical or design basis LOCA produces thyroid doses and whole body doses at the exclusion area boundary (EAB) and the low population zone (LPZ) which are in close agreement with the values described in Supplement No.
3 to the Safety Evaluation Report for CR-3 dated December 30, 1976. SER Supplement No. 3 states "The potential doses tabulated below are, therefore, conservatively derived and are well below the guideline values specified in 10 CFR Part 100."
FPC has interpreted this SER statement to mean that no "unreviewed safety question" within the meaning of 10 CFR 50.59 is present due to FPC's re-evaluation. 10 CFR 100 is considered to be the acceptance limit for protection of the public health and safety.
The 1976 SER statements are not detailed enough for FPC to judge exactly how the Regulatory Guide 1.4 methodology and the 1975 CR-3 meteorological program data were used by the NRC.
However, the results obtained by FPC in its re-evaluation are so close in agreement with the NRC results that a similar methodology must have been used by the NRC in 1976 for its evaluation of the hypothetical design basis LOCA. Furthermore, to ensure conservatism, FPC has used more recent NRC guidance for its re-evaluation.
As FPC noted in the CR-3 control room habitability submittal, CR-3 is not a Standard Review Plan (SRP) plant. However, to further ensure that conservative radiological consequences were produced, FPC used SRP 15.6.5, " Loss-of-Coolant Accidents Resulting From Spectrum Of Postulated Breaks Within The Reactor Coolant Pressure Boundary" as a guideline for the parameters and assumptions in the re-analysis. SRP 6.5.2, " Containment Spray As A Fission Product Cleanup System" was also used as a guideline. l l
f LLFA The revised LLFA results show an increase in the offsite thyroid doses which is in proportion with the decrease in the assumed ABV System charcoal filter efficiency (90% vs 0%) . The whole body doses increased by 10 mrem at both the EAB and the LPZ.
L p ,
September 13, 1989 3F0989-01 Page 3 The revised accident doses are much less than the limits specified by 10 CFR 100. The 1976 SER for CR-3 does not address this accident in the list of events reviewed by the staff, therefore,10 CFR 100 is considered to be the acceptance
-limit for protection of the public health and safety and no "unreviewed safety question" within the meaning of 10CFR50.59 is present due to FPC's re-evaluation.
l The regulatory process required by 10 CFR 50.59 is under review by the NRC_and the industry. NSAC/125, " Guidelines for 10 CFR 50.59 Safety Evaluations" is being considered by the staff for endorsement.- The increased dose consequences are within the licensing basis for CR-3, i.e., the 1976 SER. As long as the calculated doses remain less than the 10 CFR 100 limits, NRC review is not necessary before FPC revises the FSAR. This position is consistent with the NRC comments on NSAC/125.
Until the guidance for conducting 10 CFR 50.59 reviews is formally endorsed, FPC is providing the NRC with the proposed FSAR changes for information. .These changes establish a revised licensing basis for the ~ CR-3 LOCA and the LLFA radiological consequences while recognizing that the margin between the - projected releases and the design basis limits established by 10 CFR 100 have not been significantly reduced.
Included are the revised FSAR Sections 14.2.2.6, 14.2.2.5.10, 14.2.2.7, 6.2.2.1., and notations of deleted pages. FPC will revise the FSAR with these changes no later than July 1, 1990.
Sincerely, L f
- Rolf C. Widel , Director Nuclear Operations Site Support RCW/JWT/sdr Attachments xc: Regional Administrator, Region II Senior Resident Inspector i
5l l e j
. September 13, 1989.
3F0989-01 Page 4 i
i COMPARISON OF LOCA RADIOLOGICAL CONSEQUENCES (Rea)
FSAR Table 1976 SER Revised 14-57 Doses Doses L EAB ' (2-hr)
Thyroid 63.1 133 134.2 Whole Body 1.55 3 2.31 LPE'(30-day)
Thyroid 9.11 25 27.1 Whole Body 0.29 <1 0.42 COMPARISON OF LETDOWN LINE FAILURE RADIOLOGICAL CONSEQUENCES (Rem)
FSAR Table Revised 14-43 Doses EAB (2-hr)
Thyroid 0.115 1.15 Whole Body 0.066 0.067 LPZ (30-day)
Thyroid 0.0101 0.101 Whole Body 0.0058 0.0059 4
.__m_ _.__
l An analysis of the minimum containment back pressure, including the effect of I the Purge System, is provided in a report from Florida Power Corporation (G.
( C. Moore) to the NRC (R. W. Reid), transmitted by letter dated July 11, 1980. 3 The analysis utilizes a CONTEMPT model which employs the basis approach '
listed in BAW-10103A, Revision 3, yet is specific to Crystal River Unit 3 Reactor Building. It concludes that the generic 177-FA lowered loop RB l pressure evaluation is conservative with respect to Crystal River Unit 3 and !
the ECCS conformance to 10CFR50.56 regardless of Purge System Operation.
/WSGg 7' NEW /g 2. 2. 5. /0 14.2.2.5.10 Embonmental-An alys+s-o f-Eos s-o f-Coola nt-Acc4 de n t s The analyses in the preceding Sections have demonstrated that ECCS injection wiy meet the Final Acceptance Criteria for LOCAs resulting from RCS ruptures ranging s
in size from small leaks to the complete severance of the hot leg The environmental consequences from a LOCA are conservatively piping'dsby analyze assuming the activity associated with the gap of all fuel rods is released.
The activity released is shown in Table 14-52. 50% of the iodine released is assumed to plat'Nout, and the other half is assumed to' remain in the reactor building atmospher ere it is available for leakage.
The alkaline solutionsin the reactor building spray redures the airborne iodine as described inxAppendix A to this Chapter. 2% of the iodine available for leakage has been conservatively assumed to be present as organic iodine; the remahting iodine 'is present as elemental iodine.
(- Specific parameters used and lhe calculated . spray effectiveness are given in Table 14-53. /
The reactor building pressure his(or \ or f the design basis accident is shown in Figure 14-43. Although the, reactor uilding leakage rate will decrease as the pressure decays, the lea.kage is ass ed to remain constant at the design leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Ther after, since the reactor building will have returned to nearly atmospheric pr' essure, the rate is assumed to be reduced to one-half the design leak rate an'd remain at this value for the duration of the accident.
The atmospheric dispersion characteristics of thq site are described in Section 2.3.3. The site dispersion factors for the Muration of the accident are listed in Table 2-11. A breathing rate of 3.47EM m3 /s is assumed for our exposure. For the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exposure, inbreathing ratg of
~
the twom*tj/s is assumed for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and a rate'Af 1.74E-4 m3/s is 3.47E-4 assumed for the remaining 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. For the 30-day exposvre, a breathing rate of 2.32E-4 m3/s is assumed. The total integrated thyroidNqnd whole body doses /at the exclusion distance and the low population 'd tance are summa'rized in Table 14-54.
/ .
14.2.2.5.11 Reactor Building Subcompartments Pressure Response The results of an analysis of the reactor cavity and the steam generator
! (. compartments for the pressure response considering a homogeneous steam-water-air mixture with appropriate correlations for sonic flow through the gaps 14-67 (Rev. 11)
,s 14.2.2.5.10 Radiological Consequences Of A Loss Of Coolant Accident Loss of Coolant Accidents (LOCA) are postulated accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the Reactor Coolant Makeup System, from piping breaks in the reactor coolant pressure boundary. The LOCA is one of the postulated accidents used to evaluate the adequacy of the plant's structures, systems, and components with respect to public health and safety.
Multiple barriers, engineered safeguards, and administrative procedures are provided to. prevent and minimize the consequences of a LOCA.
Regardless of these safety provisions, it is postulated that a Design Basis LOCA'of the magnitude. assumed in Regulatory Guide (RG) 1.4, Rev
- 2 occurs. In order for a RG 1.4 fission product release to occur, fuel melting is required. Since the Emergency Core Cooling System (via high and low pressure safety injection and core flooding systems) is provided to prevent this occurrence,.a more realistic analyses of a LOCA is also presented based on a reduced source term, i.e. fission product release associated with all the activity in the fuel rod gap of the core.
14.2.2.5.10.1 Acceptance Criteria The acceptance criteria for-the radiological consequences of the LOCA are that the offsite radiation exposures are within 10 CFR 100 limits, Paragraph 11. Specifically, the 2-hour dose at the exclusion area boundary (EAB) and 30 day dose at the low population zone - (LPZ) are limited to 300 rem (thyroid) and 25 rem (whole body). In addition, 10 CFR 50, Appendix A, Criterion 19 requires that adequate radiation protection provision be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body (30 rem, thyroid and 30 rem beta, skin), for the duration of the accident.
14.2.2.5.10.2 Identification Of Causes And Accident Description The Loss of Coolant Accident is postulated as the principal design bases event for assessing the potential risk to public health and safety. As a result of the LOCA, a fraction of the plant's fission product inventory is assumed to be released from the fuel assemblies into the Reactor Coolant System and later into the Reactor Building via the break in the RC System pressure boundary. High Reactor Building pressure signals from the Engineered Safeguards Actuation System (ESAS) isolates (4 psig) the Control Complex putting it into a recirculation mode of operation and initiates (30 psig) the operation of the RB Spray System.
The Control Complex Emergency Fans and Charcoal Filters are manually placed in service by the operator.
The fission product inventory in the Reactor Building is reduced by radioactive decay and the action of RB Spray System as discussed in Section 6.2.2.1.1. This radioactivity is assumed to leak from the containment to the environment at a constant rate of 0.25% per day for
1 l.
the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident and at one-half this rate (0.125%/ day) thereafter.
In the Design Basis accident, it is also postulated that fission products are released to the environment via recirculation loop leakage of engineered safety features components located outside the Reactor Building. Both operational leakage (a value of 4510 cc/hr was assumed which is twice the expected leakage given in Table 6-11) and that associated with the post-LOCA failure of a passive component (50 gpm leak occurs at 24 hrs into the event and lasts for 30 min.) is assumed.
The activity released from this source is collected by the Auxiliary Building Ventilation (ABV) system and is discharged to the environment via the plant vent. Since this system would not be operative during a loss of offsite power occurrence and is not powered by the emergency diesel supply, credit for the operation of the system's charcoal filters is not assumed.
The released fission products (iodines and noble gases) are dispersed in the atmosphere with no correction made for depletion of the effluent plume of radioactive iodine due to deposition on the ground or for the radioactive decay of fission products in transit.
The offsite radiological exposure to individuals located at the exclusion and low population zones results from inhalation of radioactive iodines (thyroid dose) and immersion in the released radioactive cloud (whole body done). The radiological exposure to operators in the control room result from (1) direct radiation from the released radioactive cloud (2) direct radiation exposure from the Reactor Building and (3) exposure to radioactive materials which leak into the control room frcm the radioactive cloud in the atmosphere.
Direct radiation exposure to the contlol room operation is minimized by concrete shielding of the Reactor Building and Control Complex.
Infiltration of radioactive materials into the Control Complex is minimized by the low leakage construction features of the Control Complex. The Control Complex Ventilation System is designed for zone isolation with filtered recirculated air emergency mode of operation.
14.2.2.5.10.3 Methods of Analysis
, Two methods of analysis are provided in evaluating the radiological consequences of a Loss of Coolant Accident: (1) Design Basis and (2)
Realistic Basis. The Design Basis method utilizes upper bound assumptions contained in Regulatory Guide 1.4 while the Realistic Basis method assumptions were made to ensure the results are conservative, but more realistic. A summary of the parameters and assumptions used in assessing the radiological consequences of the LOCA for both methods are presented in Table 14-52. The differences in the methods of analysis are in the assumed post-LOCA radiation source 7erms, atmospheric dispersion, and control room inleakage parameters as noted in Table 14-52.
The Design Basis method is based on RG 1.4 core inventory releases plus an additional post-LOCA activity release due to recirculation system leakage outside of containment. In the Realistic Basis, the radiation
source term is 1.imited to the core activity inventory associated with the fuel rod gap as presented in Table 14-53. The gap activity was evaluated with a digital computer code, BURPE (Ref. 21), based upon the .
fission product escape rate coefficients determined by ANL 5800 (Ref.
22).
The offsite atmospheric dispersion factors used in the Design Basis
> analysis are based upon the short term accident diffusion models presented in ' section 2.3.4. In the Realistic Basis, the offsite dispersion factors are based on the 22-1/2 degree sector with the highest annual average value. The control Complex dispersion factors for both methods are based on a 5th percentile X/Q value associated with a 1.2 meter /sec wind speed including credit for turbulent mixing within the building wake cavity. For periods greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, credit has been taken for the reduction in this value due to post accident control room occupancy, wind speed and wind direction persistence factors as recommended in Murphy-Campe (Ref. 26).
The total post accident leakage into the control complex was calculated to be 236 cfm via (1) penetrations (approximately 0 cfm], (2) door seals
[5 cfm], (3) ingress / egress [10 cfm], and (4) dampers [191 cfm filtered path and 30 cfm unfiltered paths] . However, for conservatism, the total
~ 1nleakage is assumed in both nethods of analysis to be equal to 0.06 volume changes per hour (355 cfm) based upon a Type B Control Room as defined in RG 1.78. For the Realistic Basis, it is assumed that 191 cfm of inleakage occurs via filtered pathways. The Design Basis assumes only 70 cfm of the total inleakage to be filtered.
The activity flow path models utilized in the analyses for Reactor Building Leakage, Recirculation Loop Leakage, and Control Complex Inleakage is given in Figures 14-65, 14-66 and 14-67, respectively.
Both the Design Basis and Realistic Basis analyses are based upon the assumption that the RB Spray System is functioning in a degraded (worst case) mode of operation, i.e. spray pump failure.
The design basis parameters listed in Table 14-52 were utilized as input to the TACT-III (Ref. 27) computer program to compute the offsite and unprotected control room radiation exposures. The protected control room operator dose due to inleakage of radioactive materials were calculated based upon the use of an iodine dose protection factor and a whole body dose geometry factor described by Murphy-Campe. The analytical model used to calculate the direct dose contribution from the radioactive cloud in the atmosphere is given in RG 1.4. The model was adjusted for the reduction in dose due to the control room shielding.
The direct whole body dose from the Reactor Building was calculated based on a cylindrical radiation source model and corrected for the reduction in dose due to Reactor Building concrete shielding (3.5 ft),
Control Complex (2 ft) and a minimum source-receptor distance of 48 feet.
l The INHEC (Ref. 28) computer code was used to compute the Realistic Basis offsite and control room doses directly. The associated
- i 4
assumptions and parameters utilized as input to this code are also
.. listed in Table 14-52.
The FPC letter dated June 30, 1987 submitted the CR-3 Control Room Habitability Evaluation Report. The NRC letter dated May 25, 1989 transmitted the SER which concluded that the design of the CR-3 control room habitability system was adequate.
14.2.2.5.10.4 Radiological Consequences The offsite and control room radiation doses, resulting from both the Design Basis and Realistic Basis analyses of the Loss of Coolant Accident, are. presented in Table 14-54. In both cases, the . post accident offsite and control room dose consequences satisfy the-requirements of 10 CFR 100 and 10 CFR 50, Appendix A, GDC 19, respectively.
In addition, the effect of a 10 minute delay in the control room operator manually placing the control room emergency fans and filters into service was evaluated and resulted in approximately a 6% increase in the calculated control room operator thyroid dose (from 26.5 to 28 rem), with.no change in.whole body doses.
TABLE 14-52 RADIOLOGICAL CONSEQUENCES OF A LOSS OF COOLANT A.C&lAEST EARAMEIER ASSUMPTIONS Realistic Design Basis Basis Ana1ysh An.Alys_Ls_
Source Terms Core Thermal Power Rating, MWt 2595 2595 e
Activity Released To RB:
Core Inventory:
Iodine N/A 50%
Noble Gases N/A 100%
Gap Inventory 100% N/A Iodine Reduction Factor 22 N/A Due To Plateout In RB Iodine Species Breakdown:
. Elemental 91% 91%
Organic 4% 4%
Particulate 5% 5%
Iodine Core Inventory N/A 50%
Released To RB Sump Reactor Building Free Volume, ft3 2,000,000 2,000,000
- Sprayed Volume, ft3 1,304,000 1,304,000
- Unsprayed Volume, ft3 696,000 696,000
- Air Turnover Between Sprayed 4800% of 4800% of And Unsprayed Volumes Unsprayed Unsprayed Volume Per Volume Per Day Day Leakage Rate, %/ Day 0-1 Day 0.25 0.25 1-30 Days 0.125 0.125 Sump Liquid Volume, gal N/A 490,182 Post LOCA Shield Wall Concrete Thickness, ft 3.5 3.5
L l
9 W
e TABLE 14-52'(CONTINUED)
RADIOLOGICAL CONSEQUENCES OF A LOSS OF COOLANT ACCIDENI 1.; PARAMETER ASSUMPTIONS Realistic Design Basis. Basis Analysis Analysis Recirculation Loop Leakage Operational, cc/hr N/A 4510 Passive Component Failure, gpm N/A 50 (For 30 Min. Starting 24 Hours After The Accident)
Fraction Flashing To Steam, % N/A 10 RB Spray System Spray System Actuation Time, sec 71 71 Spray Additive Concentration 6 6
'(Wt. % Of NaOH)
Flow Rate, gpm 1500 1500 Time To Reach pH = 8.5, min 9 9 Spray Removal Constants Elemental Iodine (Lambda.), hr-1 0-71 Sec. 0 0 71 Sec.-9 Min. 2.91 2.91 3 Min.-30 Days 16.58 16.58 Particulate Iodine (Lambdap), hr-1 0-71 Sec. 0 0 71 Sec.-30 Days 0.30 0.30 Maximum DF For 170.4 170.4 Elemental Iodine By Sprays
h +- l l
TABLE 14-52 (CONTINUED)
RADIOLOGICAL CONSEQUENCES OF A LOSS OF C00LANI_ACCIDEEI EARAMETER ASSUMPTIONS Realistic Design Basis Basis Analysis Analysis Control Complex Free Volume, ft3 355,311 355,311 Infiltration Rate (Total), cfm 355 355 Filtered In-Leakage 191 70 Unfiltered In-Leakage 164 285 Filtered Recirculation Flow Rate, cfm 43,500 43,500 Recirculation Charcoal Filter 95 95 Efficiency, %
Environmental Atmospheric Dispersion:
Offsite X/Q Values (sec/m3)
Exclusion (0-2 Hours) 2.56E-6 1.6E-4 Low Population Zone 3.10E-7 0-8 Hours 1.4E-5 8-24 Hours 1.5E-6 1-4 Days 7.7E-7 4-30 Days 4.5E-7 Control Complex X/Q Values.(sec/m 3) 0-8 Hours 9.00E-4 9.00E-4 8-24 Hours 5.31E-4 5.31E-4 1-4 Days 2.07E-4 2.07E-4 4-30 Days 5.94E-4 5.94E-4 Offsite Breathing Rate (m 3/sec) 0-8 Ilours 3.47E-4 3.47E-4 8-24 Hours 1.75E-4 1.75E-4 1-30 Days 2.32E-4 2.32E-4 Control Complex Operator 3.47E-4 3.47E-4 Breathing Rate (m 3 /sec)
(0-30 Days)
i f,~44 TABLE 14-53 i
POST LOCA GAP ACTIVITY
)
RELFliE INTO TiiE_BEACTOR_Bl]J_LDJNG j I
1 Isotspg Activitv. Ci Noble Gases:
Kr-83m 8.96E+3 Kr-85m 4.97E+4 Kr-85 4.37E+5 Kr-87 2.70E+4 Kr-88 8.86E+4 Xe-131m 8.13E+4 Xe-133m 9.50E+4 Xe-133 8.52E+6 Xe-135m 2.76E+4 Xe-135 3.44E+4 Iodines:
I-131 6.55E+5 I-132 9.37E+4 I-133 1.41E+5 1-134 8.81E+3 I-135 4.47E+4
, l l
4 L__ _ _ _ _ - _ _ _ - _ _ _ _ _ - - _ _ _ - - _ _ _ - . _ _ _ _ _ - - _ _ _ _ - - - --- - _ - - - - - - - - _ - - - - - - . - - - - - - - - - _ - - - - - _ - - - - - - - - - - -
{
.s.
{
TABLE 14-54 0FFSITE AND CONTROL ROOM DOSES FOR Al0JS OF COOLANT ACCIDENT REALISTIC BASIS DESIGN BASIS DOSE TYJLE ANAIJSIS (Remj_ A!LA_ LYSIS (Remj THYROID:
Exclusion Boundary 1.6 134.2 Low Population Zone 0.2 27.1 Control Room , 0.8 26.5 WHOLE BODY GAMMA:
Exclusion Boundary 0.0022 2.31 Low Population Zone 0.00027 0.42
. Control Room 0.04 1.88 WHOLE BODY BETA:
Control Room 0.6 17.7 l
i
- . . ~
.( 'O . t Add these references to page 14-82
- 26. " Nuclear Power Plant Control Room Ventilation System Design For Meeting . General.' Criterion ' 19," K. G. Murphy And K. M. Campe, USAEC, 13th AEC Air Cleaning Conference, August 1974.
- 27. NUREG/CR-3287, "A Guide For The TACT III Computer Code", USNRC, May 1983.
- 28. GAI-TR-101P-A, Topical Report, " Computation Of Radiological Consequences Using the INHEC Computer Program, March 1976.
x
- .e ACTIVITY FLOW PATH MODEL I REACTOR BUILDING LEAKAGE '
HYPOTHETICAL LOCA I
j ENVIRONMENT I
REACTOR BUILDING SPRAYED -
I REGION -
.25% / day 0-24 hrs. Post-LOCA !
Volume = 1,304,000 ft3 .125% / day >24 hrs. Post-LOCA h i
Spray Removal AE, AP *
.I o !
l 4800% / day !
l If I
I REACTOR BUILDING UN5 PRAYED i REGION .25% / day 0 24 hrs. Post-LOCA j
.125% / day >24 hrs. Post-LOCA .l
~
i Volume = 696,000 ft3 i l
l I
,-w . + * =- - - ~*
FIGURE 14-65 8 , .. ,
(' 2 E
4:
~. .
. j-ACTIVITY FLOW PATH MODEL RECIRCULATION LOOP LEAKAGE HYPOTHETICAL LOCA I
- ENVIRONMENT I
I Leakage' Due To j Passive Component Failure ;
CONTAINMENT SUMP 14.7% / day 24 to 24.5 hrs. Post-LOCA h I
1 I
i j
Volume = 65,532 ft3 !
l l
l l
Recirculation Loop I Operational Leakage :
O.0058% / day 0 to 30 days Post-LOCA !
l l
l l
l FIGURE 14-66 6
E :~'
N :.
l
? :-
ACTIVITY FLOW PATH MODEI.
I' CONTROL COMPLEX INLEAKAGE l
HYPOTHETICAL LOCA CONTROL BUILDING HABITABILITY ENVELOPE Filter Filtered inleakeage 70 SCFM (Design Basis) .
1915CFM (Realistic Basis) 0.06 Volume Recirculation Changes < v Per 43,500 SCFM Hour (355 CFM)
Unfiltered inleakeage 285 5CFM (Design Basis) 164 5CFM (Realistic Basis)
Volume = 355,311 f t3 Ventilation System Mode Of Operation: Zone Isolation With Filtered Recirculating Air.
t - _ - _
FIGURE 14-67 I
/w ras //o/sylcc ( Cons 3 ences o~f Oth aAcl den-/ Os 1
tif'f f C U S C e d N 7 S c N M N E.E.$ /0.
14.2.2.7 Maximum Hvoothetical Accident I 4.2.2.7.1 Identified;un vi Ad-; dent T e Maximum Hypothetical Accident (MHA) analysis postulates a failure in h'e re ctor coolant boundary in which fission product activity is assume to acc ulate in the reactor building atmosphere where it is availab for leaka e to the environment. Due to fuel cladding failure and prima system ruptur the accumulated containment inventory consists of th maximum activity from the fuel and the maximum equilibrium activity of ) e reactor coolant qsulting from reactor operation at the design power for a sufficiently long period of time. Assumptions for fission prpduct releases to the react r building are assumed at a level that could result only from melting of th core; however, even in the event of a LOCA/ no significant core melting wo ld occur, since core meltdown would requjr'e a multitude of mechanical failub in safety-related systems and com which are designed to preven such an occurrence. Nevertheless/ponents, to assure that the operation of CR-3 d qs not present any undue hazard J'o the general public, based on fuel claddihg failure and primary systept rupture, an accident involving a gross releaie of fission products is pvaluated -- 100% of the noble gases, 50% of the and 1% of all other fission products (solids)%alogens
, as stipulated by(including f
iodine),
T10-14844. Gases are assumed to be released through the rea tor containment b,uilding immediately into the atmosphere surrounding the p1 t. No retention of noble gases is assumed.
Only 50% of the iodine releas ; to conta~timent are assumed to plate out, allowing as much as 25% of te core iodine to be released into the atmosphere. Iodine and noble gas relea es available for leakage are listed in Tables 14-55 and 14-45, respecti 1.
)
Even without engineered safety fe,p ur s, the concentration of radionuclides in a containment atmosphere would be pleted by the natural processes of iodine plate out and radioactlye' decay. ngineered iodine removal mechanisms affecting fission product , activity relgses to the environment include washout with containment spr'ays and removal (y charcoal filters.
\
14.2.2.7.2 Environm' ental Analysis and Resu1\s Thyroid and whole /body dose calculational metho\
j dsymodel the minimum safety operation of engjneered safeguard systems for removing airborne iodine, i.e.
only one out o,f' two building spray pumps and only on out of three reactor building air, cooling units are assumed in operation. Other than activity releases, parameters for the MHA analysis are the same qs those assumed for the LOCA an'alysis in Section 14.2.2.5.5. Thyroid dosesyre computed using the averacje iodine inventory (see Table 11-2), the atmospheric diffusion factor p ee Section 2.3), the breathing rate and the containmhet leakage rate (see Se'ction 5.2.1.1). Within 1 minute after the accident, isqlation of the reat dr building has been completed and leakage has been termingted, except for the design containment leak rate, by the reactor building isolation and c ling functions of the ESAS. Spray removal coefficients, deconQmination actors and iodine source fractions are based on the sodium hydroxide spray solution. Whole body doses are based on iodine and noble gas inventories
( the atmospheric diffusion factor, the containment leakage rate and beta-kmma, energies of isotopes. The resulting doses are summarized in Table 14-57, aqd 14-73 (Rev.11)
\;
. i are-4ess-than t ha 10CER100_ guide 14ne-values-of-300 rem-for-thyroid-doses-and i 25 rem for wnvie-body tfores: '
.2.2.7.3 Inhalation Dose to Reactor Operators in the Control Room j In e event of a LOCA, the ES Reactor Building 4 psig isolation signal would auto tically close the control complex outside air intake (AHD-1) and ,
atmosp eric relief to outside discharge dampers (AHD-2)' and open return i dampers AHD-3),- thus placing the system in a recirculation mode through the normal _ p th. In this mode of operation, the controlled access area is isolated tipm the control room and the remaining areas of the control complex above the 95sft. elevation. Upon receipt of a toxic gas signal (chlorine or sulfur dioxidq gases), the dampers are positioned as described for the ES signal. Upon ceipt of a high radiation signal, the dampers are positioned as described fo he ES signal. In addition, both the. control complex normal supply fans (AHF A and AHF-178) and the control complex return fans (AHF-19A and AHF-198) {aqe automatically stopped. The' operator is required to manually change. the telector switch from normal to' emergency, which will open the absolute and -charbqa1 filter damper, close' the filter bypass valve' and j start one of the two coh(rol complex emergency supply fans (AHF-18A and AHF- ;
IBB). This fully place,s the system in' emergency mode. All air is .
recirculated through the eihqrgency filter, bank. as described in Section 9.7. !
The return air and minimum ob side air, , required for room pressurization, is l directed through the :bsolute nd charcoal filter before entering the coils '
and. fan for return to the condi 'oned jspace. ,
l The MHA assumptions presented in S tion 14'.2.2.7.1 apply in the calculation (' '
of the thyroid dose to the reactor o rator in the control room. When the I- 5 131 dose equivalent concentrate'on re hes IE-8 microci/cc, the normal fans I are tripped. The ventilationjystem is egulated so that a positive pressure would be maintained in the control room to ensure that all fresh air would enter through the filters / The in-leaka q of outside air past the control complex isolation dampers is conservative 14 assumed to be 400 scfm, or approximately 1% of the total recirculation flow s of 43,500 scfm. The 90% I efficient emergency, filters remove iodine and other particles during i recirculation and fresh air changes. Fresh air \ change rates are dependent j upon the outside air flowrate, ghe recirculation raits, the total air volume-being recirculate'd of 243,000 f t , the number of men f(the control room, and their breathing ~ rate. For four shifts during the 3Q-day period of the accident, an/ individual would spend approximately 7.5 days in the control room. The, halogens in the control room were assumed to 'be at equilibrium concentrations throughout the duration of the accident. e atmospheric dispersion factor, which is dependent upon the location of f h air intake for the' control room and the worst locqtion for an operator o stand, is consg/vatively assumed to be 9E-4 sec/m3 Based on the total %131 dose equnalent activities released from containment in 30 days, and basb( on the rpt'io of control room air concentration to outside air concentration, the
/hyroid dose is calculated to be 0.45 rem. N L
14-74 (Rev.11)
( :.
I 14.2. 7.4 . Effects of Engineered Safeguards Systems Leakage
-b The engine ed safeguards include HPI and LPI of the ECCS. These,sistemscan provide an ditional- source - of fission product leakage external to the reactor buildi during the recirculation phase for long-term' core cooling.
It is postulated hat during the MHA, one of the core cooling' systems undergo-a pipe rupture. I he pipe breaks when the core cooling pumps are drawir.g on the reactor buildin sump, radioactive liquid would,b released within the auxiliary building. Th pipe break is assumed to' occur at the location resulting in the greatest a s of reactor building.s' ump fluid.. Radioactivity .
is released by exfiltration rough the charcoal filters of the auxiliary building ventilation system to e unit vent. Reactor building leakage is assumed to occur throughout the ac Q!ent at the Design Basis Accident leakage rate of 0.25% by weight of containe analysis of the potential leakage from( atmosphere hese systemsperis 24 hrs. Aindetailed presented Section
- 6. je
/
It is assumed that the water being recircu Ated from the reactor building sump through the external system ~ piping contaih 50% of the core saturation iodine inventory, which is ,the entire amount iodine released from the reactor core cooling system'. The 50% escaping f n the RCS is consistent with TID-14844 specifica.tf ons. The assumption that a iodine escaping from the reactor coolant system be absorbed by the water in the reactor building is conservative since much of the iodine released from the fuel would be
. plated out on tije' building walls. It is assumed that all of the iodine contained in yater which flashes is released to the aux'Q iary building
_. atmosphere. Jodine release from the remaining water is calc 0 ated using a gas / liquid,p,artition coefficient of 9E-3. 50% of the lodine rel sed to the
(-- auxillafbuilding r is assumed to plate out on the walls. The re inder is assumpd to be released through 90% efficient charcoal filter The atm,23.pspheric dilution is based on the 2-hour dispersion factors shown i Table If- The leakage and the resulting thyroid dose are shown in Table 14 (7.
k 14 75 (Rev.11)
7 i TABLE 14-52
/ERVACG 41/7?/ EEMED TABLE /t/ . S 2 ET!Y!'y on rper egg gg;ggg ggg 3;g gg Il pe Acti v. Ci Noble ses: l I
Kr-83m 8.96 E+3 Kr-85m 4.97 E+4 Kr-85 4.37 E+5 Kr-87 2.70 E+4 Kr-88 8.86 E+4 Xe-131m 8.13 E+4 Xe-133m 9.50 E+4 Xe-133 52 E+6 Xe-135m 2h E+4 Xe-135' 3.4 +4 Iodine-131 6.55 Iodine-132 9.37E+)4 Iodine-133 1.41 E+5 Iodine-134 8.81 E+3 Iodine-135 4.47 E+4 s14-146 (Rev. 11)
-.___._____-_____._..-_.____-.__________.-_.__________-__m______m_m_ _m_______.m_m
' (:: . - l2WMCC &r?W'2EV/SEh^ fh
. REAC-T0" "'.'It0!MG 5""f"Y SYS-Tfh tFFECTivEnt55 -
/f S
.6h 1 Spray. 2' Sprays Parameter Operates ,- ! cerate O
/
Spray Flo gpm 1500 ,/ 3000 Effective Fall H ight, ft 96 96 RB Free Volume, ft3 2;000,000 2,000,000
/
/-
Spray Drop Diameter, microns / 1080 1080
/ .
Average Removal Time Constant, hr-2 Elamental 12.78 25.56 Aeros Particles 0.37 0.73 rganic 0.0 0.0 l
(14-147 (Rev. 11)
.gs?2 ACE 4)/7// 2EVrSeE'D TR6&d' /V-Sc/
TABLE 14-54 ENV!Pf"" ENTAL D^SES P,ES"LT!"C r"!w HAYTMUM BREAK SIZE LOCA ,
J .
/
/
2- ur Dose at Exclusion Distance, Rem Oriainal FSAR Value Cycle 3 Valu
/ Cycle 7 Value
/
Thyroid 0.549 2.19,/ 3.01 Whole Body 0.0174 0.016 0.008 j
30-Day Dose at low Pop ation Distance, Rem'
,i Oriainal FSAR Valu.g[ [vele 3 Value Cycle 7 Value
/
/
Thyroid 0.073 0.517 0.25 Whole Body 0.011 ,/ 0.0081 0.004
(.
LOCA During Reactor But ing Purge Purge Valve Closing Time, s 5.0 Iodine Released, equiv Ci 4.2 I-131 Dose
/
Increase in 2-Hour Thyroid Dose at Edlusion Distance Due to Purge
(.213
[IalveClosingT'.:.i,nem X 14-148 (Rev. 11)
-l.
' s :n ,
1 1 :, . *
- ([
TABLE 14-55 MkIODINEACTIVITYAVAILABLEFOR4EAKAGE
/
.WITH50%PLATEOUT[_
. actor Building JS0 TOPE Activity. Ci 1-131 1.61 E+7.
I-132' 2.45 E!7-I-133 62 E+7-I-134 4.24' +7 I-13 .
3.29 E+7 pELET&
(
(
l i
l.
l'
' (.14-149 (Rev. 11)
L.. _ _ _ ___ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __. __m. ________ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ ____
3.-
- j
'd,,.
4 ',
~
E TABLE 14-56 t.-
- L;A RALiGACiiViiiE5 AVAILABi.E PUM LtAKAGE' (
ReactorBuiding .i L ISOTO Activity. Ci
/
Kr-83m 7.3 E+6 l- Kr-85m 2.1 E+7 l- Kr-85 5.4 E+5 Kr-87 3.9 E+7 Kr-88 6.0 E+7 Xe-131m 5.5 E+5 Xe-133m 1 E+6
\
Xe-133 1.3NE+8 Xe- 35m 3.4Ek
-135 2.6 E+7 n2ere
(14-150 (Rev. 11)
-_________.._._________________..____________________J
e TABLE 14-57
-HHA-f NV-IRONMENTAL--- D05 E 5 -
2-Hour Do:e at Exclusion Distance, rem Oriainal FSAR Dose Cycle 7 Dose Thyroid 23.3 23.4* 26.l** 63.1**
Whole Body 2.01 2.0l* 2.02** 1.55**
30-Day Dose at Low Pop tion Distance, re s Oriainal FSAR ose Cycle 7 DqEe Thyroid 2.6h 2.66* 2.89** 9.11**
Whole Body 0.29 0. 0.29** 0.29**
Engineered Safeguards Leaka
(
Iodine concentration in iquid, I-13 dose equivalent, Ci/ml 0.034 Liquid leakage, ml/hr 2165 90 Leakage that flashey, ml/hr Thyroid dose at e clusion distance, rem 0.0191 Considprs throttling of reactor building spray pumps at time of recirculation (34.4 minutes after accident with two Lpl, \ wot HPI, and two i RPS umps operating) to 1200 gpm.
- nsiders reactor building spray pump flow of 1200 gpm.
paan
\ 14-151 (Rev. 11)
- l. _ _ _ _ _ _ _ _ _ _ _ _ - _ _
l
~. l assembly prior to being recirculated. The sole function of the sump l( j screen assembly is to prevent small water f rom entering the associated systems.
debris in the recirculating ( )
A 1-1/2 inch grating cover above the sump inlet is designed to prevent large debris f rom entering the sump area. Dislodged debris and paint chips present in the recirculation water, smaller than 1-1/2 inch size, will flow into the reactor building sump preceding the sump screen assembly. High density particles will have a tende1cy to settle out and be retained by the 3 foot weir preceeding the ', ump screen assembly. The velocity of flow through the sump screen is relatively low and in a downward direction, therefore permitting suspended debris to settle out and collect in the debris hoppers. Particles smaller than 1/4 inch in size which are not retained by the weir nr sump screen assembly will flow throvgh the associated Decay Heat Removal System and Reactor Building Spray System with no additional restrictions, thus returning to their originating source (reactor building proper).
D AI report 2009, " Borated Water and Sodium Hydroxide Storale Tank the 7ES opera-
[A tio wdown Transient Analysis," ( Appendix 14A) addresses / eat Removal g of the Reactor Building Spray System and Decay sodium
~
Systeg. _ This report demonstrates that utilizing the hydrox1M tank with a 10.5 to 12.0 weight p e r c e3t',s o d i um hydroxide solution, he maximum pH of the spray s ol u t i orVi s not greater than 11.0 durin 11 modes of operation and that hi resulting doses from the Maximum hypothetical Accident (MHA) a within 10CFR100 limits. (
in addition, i tydemo n s t ra t es that a r atively high pH (at least (..
8.0) is maintained in the sump a fye'r mixing and dilution with p rimary cool ant, borated water from/ CCS injection, and Core Flood-ing Tank (CFT) i n v e nt'o(i e s . The peration of the system during the following modes of operatio was analyzed and found to be acceptable: \
- a. Full flow mod i n\whi c h all components function as designed. \
- b. Half flow mode in which on rsain does not operate.
- c. V al v fa e/' i l u re x
mode in which BSV' 36F or 37F in the NaOH a dtfi t i on line fails closed.
- d. pray pump f ailure mode in which BSP-3A o'r s 3B fails.
Decay heat pump failure mode in which DHP-3 r fails, f .
t 6-18 (Rev. 6)
p ..
6 2. 2. /. e.L 3a,'/dany Spuy A/oee/es
!~ SpRACO-1713A spray nozzles are used in the spray headers. They are
(_ ramp bottom swirl chamber type nozzles of one piece construction; they have a 3/8' inch orifice and deliver a hollow cone spray pattern. Each nozzle will deliver 15.7 gal / minute at 40 psi with a_ spray angle of 63*. The drop size distribution used in the design distribution produces a conservative evaluation of the system's iodine removal rate. The measured spray drop size distribution is based on the results of spray tests performed by Spray Engineering Company of Burlington, Mass during 1970 and 1971. The following paragraphs describe how the tests were performed.
A SPRACO-1713A spray nozzle was positioned ten feet above the plane of drop size measurement. The nozzle sprayed water straight down at a rate of 15.3 gpm at a 40 psi differential pressure. The plane of measurement was divided into eight concentric regions each six inches wide and then into four quadrants, which gave a total of 32 zones. The fraction of the total spray flow was measured for each zone. High speed photographs were also taken in each zone to measure the spatial drop size distribution at that location. The photographs were taken with a three micro-second exposure and with the field of spray limited to a 2 inch thick radial section across the zone. The photographic negatives from each zone were analyzed for the number of drops in every 25 micron interval, using a Mann Model 880 Comparator. The total drop count
^ was about 33,000 drops. The end result of the experimental
('-
measurements made by Spray Engineering Company was 32 histograms (one of each zone) showing the number frequency of spatially distributed drops versus the drop size, and tables summarizing the ;
amount of spray flow in each zone. This spatial drop size distribution data was then analyzed, as follows, to obtain the temporal mass drop size distribution shown in Figure 6-13. The percentage of the spray's mass flow rate (P d) which contained only drops of a specified size (d) or smaller was calculated as follows:
d 32 Pg = I I (Np vp Vp p Fp)g D=1 Z=1 where Pd = percentage of spray's mass flow rate which contains only drops of size (d) or smaller; Np = the number of spatially distributed drops in a given zone which are in drop size group (Do) (group widths are 25 microns);
1
- N-6-19 9
p, s.:
Motors .
(
-The reactor. building spray pump motors are designed to the same requirements as the ECCS motors. Refer to Section 6.1.2.4.
6.2.2.5 Reliability Considerations A f ailure analysis has been made on all active components of the system to show t*nat the failure of any single active component
-will not prevent fulfilling the design function. This analysis is shown in Table 6-6.
-6.2.2.6 Missile Protection Protection against missile damage is provided'by direct shielding or by physical separation of duplicate equipment. The spray headers.are located outside and above the primary and secondary concrete shield.
6.2.3 DESIGN EVALUATION The Reactor Building Spray. System, acting independently of the Reactor Building Emergency Cooling System, is capable of limiting
~
the containment pressure after a LOCA to a level which is below the design pressure and reduces building pressure to near atmospheric level. The Reactor Building Spray System is at least equivalent in' heat removal capacity to the Reactor Building f Emergency Cooling System and is designed for long term post-accident operation. In combination with emergency cooling
~ units, it affords redundant alternative methods to maintain containment pressure at a level below design pressure. Any of the following combinations of equipment will provide sufficient heat removal capability to accomplish this:
- a. The Reactor Building Spray System.
- b. Three reactor building emergency cooling units.
- c. One reactor building emergency cooling unit and the Reactor Building Spray System operating at one-half capacity.
The Reactor Building Spray System will deliver 3,000 gpm through the spray nozzles within 68.2 seconds after the reactor building pressure reaches the actuation set point.
INSEtCT NEW 6. 2. 5, I l
( .
6-24
- ( l g
4 1
1 6.2.3.1 RB Spray System Iodine Removal Evaluation The icdine removal function of the RB. Spray System has been evaluated for fully effective and minimum safeguards operation in the following cases.
Analysis of each case includes the' condition of a single active failure of any active component.
- i. . Full Flow Case - Normal mode in which all components function as designed. Spray flow is 3000 gpm.
- 2. Half Flow' Case - A half flow mode in which one string of pumps and valves do not operate, i.e. one' diesel fails to operate and all other components function as designed. The B string was selected as the failure for thta analysis. Spray flow is 1500 gpm.
l
- 3. BST-1 Valve Failure Case - Valve BSV-11 (B-side) fails closed and all other components function as designed. In this case, the total spray flow is 3000 gpm, but only Train A with a flow of 1500 gpm receives sodium' hydroxide (NaOH).
- 4. Spray Pump Failure Case - Failure of the spray pump (B-side) and all 1
other components function'as designed. Spray flow is 1500 gpm.
- 5. Decay Heat Pump Failure - Failure of the decay heat pump (B-side) and L
all other components function as designed. The total spray flow is 3000
!. gpm. However, Train B receives a reduced amount of sodium hydroxide due to failure of the decay heat pump.
The iodine in the post' accident Reactor Building atmosphere is assumed to exist in three chemical forms, i.e. l elemental, organic (methyl), and iodine sorbed on airborne particulate matter. The RB Spray System with iodine absorbing additive (i.e. NaOH) remove these three forms with varying degrees of effectiveness. -The removal of each form of iodine is described L
mathematically by a first order exponential removal process with a removal rate coefficient.
The SPIRT computer code was used to evaluate the spray removal constants for the elemental form of iodine. Hand calculational methods (Ref.Since 2) were used the spray to determine the removal constants for particulate iodines.
additive sodium hydroxide (NaOH) is not very effective in enhancing the removal rate of organic forms of iodine, the removal of methyl iodide was conservatively assumed to be zero. A summary of the assumptions and parameters used in evaluating the effectiveness of the spray system is presented in Tables 6-15 and 6-16.
The capacity of the spray solution to absorb elemental iodine from the post accident RB atmosphere is strongly dependent upon the pH of the spray solution via the equilibrium iodine partition coefficient. The recommended values (Ref. 3) of partition coefficients for sodium hydroxide buffered spray solution varies from 50 to 5000 over a pH range of 6.5 to 8.5. The spray solution pH is a function of (a) mode of spray system operation, (b) rate of drawdown from the ECCS storage tanks, and (c) rate of sodium hydroxide injection. The spray solution pH values for each operating mode were determined as part of the RB Spray System and ECCS Storage Tank Orawdown
t :?-
't,'.
'..t-Analysis (Ref. 4). The pH values are presented in Table 6-17. These values are bar.ed upon assuming the minimum NaOH concentration (6 wt.%), the minimum
. level in. the NaOH . storage tank (BST-1), the maximum borated water concentration, and maximum level in the berated water storage tank (BWST).
.The elemental iodine spray removal constants shown in Table 6-18 are based upon the Table 6-17 pH values.
The ' effectiveness 'of the. spray system is assumed to cease once the concentration of elemental iodine 'in the atmosphere reaches the equilibrium limit, i.e. the maximum allowable decontamination factor (DF) .is reached.
The DF is defined as the. ratio of the initial iocine concentration in the RB atmosphere when 50 percent of the core iodine is instantaneously released to the. concentration of iodine in the RB atmosphere at some time later. This value was determined to be 170.4.
The spray removal constants determined for the particulate iodines are as follows:
Spray Flow (apm) o (/hr)
One Header -
1500 0.30 Two Headers -
3000 0.60
.t
uk, ,
- )
T LAdd these~ references to page 6-33 L
l
- 2. 'NUREG/CR-0009, "Technicalogical Bases For Models Of Spray. Washout Of Airborne' Contaminants In Containment Vessels," USNRC, October 1978.
l 3. ANSI /ANS-56.5-1979, "PWR And BWR Containment Spray' System Design L Criteria". November 1979.
- 4. . Florida Power' Corporation, Crystal River Unit 3, " Reactor Building Spray And ECCS Storage ' Tanks Drawdown . Analysis", B&W Document 1146656-01, . November 1983) FPd Docum,,d ng-g3.ooo g p
i
--.-_.-__-.._____________._______w
L
- h i c
)\- ,
. TABLE.6-15
-IODINE REMOVAL EVALUATION REACTOR BUILDING SPRAY SYSTEM l
' PARAMETER / COMPONENT ASSUMPTION Spray' System:
Spray Nozzle Type SPRACO MODEL 1713A Number Of. Spray Drop Sizes 56 Spray Drop Size Distribution. Table 6-16 Spray' Flow Rate.(One/Two Header), spm 1500/3000
. Collection Drop Efficiency 1.0 i
Spray' Solution Chemistry: 1 Spray Additive NaOH (6 wt.%)
Spray Storage Temperature, .F- 40 Spray pH Range ~7.2 to 11 Partition Coefficient (H) Elemental Iodine 310 to 5000
, Reactor Building Design:
RB Free ~ Volume,~fta- 2,000,000 RB Free Diameter, ft 130 Fraction Of,RB Volume Sprayed, % 65.2 Fall Height (One/Two Header), ft 109/110 Maximum Post-Accident Atmospheric Temp, F 281 ;
L' quid Volume RB Sump, gal 490,182 I
Interior Scrfaces, q
- RB Surface Area Impacted By Sprays, ft2 37,900/38,760 '
(One/Two Header)
Laminar Boundary Layer Surface Area, ft2 4084 Turbulant Boundary Layer Surface Area, fta 58,320/59,180 (One/Two Header)
Spray Water Wall / Flow Fraction 0.1
~ Delta T Across Wall / Gas Boundary, F 1.0
g
..3 .0 p -
p.)
l p TABLE 6-16 l '
SPRAY DISTRIBUTION FOR SPRACO MODEL 1713A N0ZZLE i
L . Relative- i L , Data Point Drop Size Frequency Number (cm) (Fraction) 1 3.75-3 0.011 2' 6.25-3 0.027 3 8.75-3 0.056 j 4 1.125-2 0.105 5 1.375-2 0.095 6" .1.625-2 0.080 7 1.875-2 < 0.070 8- 2.125-2 0.051 9 2.375-2 0.066 10 2.625-2 0.044 N 11 2.875-2 0.026 12 3.125-2 0.022 13 3.375-2 0.017 14 3.625-2 0.020.
15 3.875-2 0.023 16- :4.125-2 0.011
.17 4.375-2 0.011 18 4.625-2 0.015 19 4.875-2 0.012
+-
20 5.125-2 0.013 21 5.375-2 0.011 22 5.625-2 0.016 23 5.875-2 0.012 24 6.125-2 0.008 25 6.375-2 0.008 26 6.625-2 0.007 27 6.875-2 0.011 28 7.125-2 0.009
e n
TABLE 6-16 (CONTINUED)
SPRAY DISTRIBUTION FOR SPRACO MODEL 1713A N0ZZLE Relative Data Point Drop Size Frequency Number (cm) (Fraction) 29 7.375-2 0.011 30 7.625-2 0.009 31 7.875-2 0.008 32 8.125-2 0.007 33 8.375-2 0.006 34 8.625-2 0.006 35 8.875-2 0.008 36 9.125-2 0.006 37 9.375-2 0.005 38 9.625-2 0.005 39 9.875-2 0.005 40 1.013-1 0.004 41 1.038-1 0.005 42 1.063-1 0.004 43 1.088-1 0.005 44 1.113-1 0.005 45 1.138-1 0.005 46 1.163-1 0.004 47 1.188-1 0.005 48 1.213-1 0.005 49 1.238-1 0.007 50 1.288-1 0.005 51 1.313-1 0.002 52 1.338-1 0.002 53 1.413-1 0.001 54 1.438-1 0.001 55 1.613-1 0.001 56 1.738-1 0.002 l
l l
po n
mi Aut CSa) 2 2 8 0 2 O l n 2 8 4 8 3 L oui .
T c 1 1 1 8 8 t r (m 4 8 4 4 5 sei omc Pie r
TR x
- ~
e gB i n
y a m r& - - - -
F H e 7
OK C vA .
N A A 6 HA E~
UT R5 M . .
MR AO n m O C i 0 T OH aB 7 AS R
LP T
r 3
6 2
^ ;
S PP .
7 IN S . . n . .
1 SI Y
E n n i n n
- M i i m i i 6 LE I n m m m m AD T i 5 E NI aA 0 0 7 0 0 L
B AXO T r
6 8
9 5
3 A NR '
T WD OY DH W
AM RU e g8 DI 0 a 5 F0 r& - - - -
OS e 7 vA .
S% H A .
TT P
_ LW U Y S0 A E . R R6 P n S i 1 1 aB . - - - .
L r 8 4 A T I
T I
N J_
n i 3 3 5 4 5 aA . . .
r 7 7 7 7 7 T
e p v t w w l m a o o a u e l l Ve Pe He F F r yu r
yu r
1 u e l f - f af af s l l Ti ri ci a u a Sa pa ea C F H BF SF DF
r
.'.. 1 TABLE 6-18 ELEMENTAL IODINE SPRAY REMOVAL CONSTANTS Initial Time To Achieve e(hr' )
Case pH e(hr ' ) - pH 2 8.5 (Min) pH 2 8.5 Full ~ Flow T. 3 4.61 6. 0 31.09 Half Flow 7. 3 2.30 8. 0 16.58 BST-1 Valve Failure:
With One Header, 7. 5 3.55 3.75 16.58 With Two Headers 7. 5 7.07 6.70 31.09 Spray Pump Failure 7. 4 2.91 9. 0 16.58 Decay Heat Pump Failure: This Situation Is Bounded By The BST-1 Valve Faifure (With Two Headers) And Spray Pump Failure Cases.
l
4_
E* 14'.2.2.6- Makeuo System Letdown'Line Failure Accident f 14.2.2.6.1: Identification of Cause
" A break in fluid-bearing lines that penetrate the reactor containment may
' result' in the release of radioactivity to the environment. There are no instrument lines connected. to the RCS that penetrate the containment.
-However, there are other piping lines such as those associated with the
' Makeup and Purification (MU) System and the Decay Heat Removal (DH) System that penetrate the containment. For fluid penetrations in. piping systems
' that do not serve to limit the consequences of accidents, leakage is minimized by a double-ba~rrier design to ensure that no single credible failure or malfunction of an active component will result in either unacceptably high leakage or the loss of the capability to isolate a piping break. The installed double barriers consist of closed piping, both inside and outside the containment, and various types of isolation valves.
The most severe piping rupture identified for which radioactivity release may occur during normal plant operation is in the Makeup and Purification System.
This involves.a rupture of the letdown line just outside the containment and upstream of the letdown control valves. A rupture at this point produces a loss of reactor coolant condition until the RCS pressure drops below the pressureL for actuation of the Engineered Safeguards to isolate the reactor building. When this pressure is reached, the building isolation signal initiates closure of .the letdown isolation valves inside the containment.
Closure of the isolation . valves stops the release of reactor coolant and fission products to the auxiliary building, thus terminating the loss-of-coolant phase of the accident.
(;
14.2.2.6.2 Safety Evaluation Criterion The acceptance criterion for the evaluation of this accident is that the resultant doses shall not exceed 10CFR100 limits. (Dose limits are 300 rem thyroid dose and 25 rem whole body dose.)
14.4.4.6.3 Methods of Analysis The CRAFT 2 computer code was used to determine the reactor coolant mass
. release rates and the primary system response for the rupture of the letdown line. The multinode model includes a detailed model of the RCS as well as noding for simulation of the letdown piping, valves, and coolers.
For purposes of calculating the mass of reactor coolant released, the reactor is assumed to be operating at 2603 MWt with a letdown flow of 140 gpm prior to the rupture. The rupture is modeled as a complete severance of the 21/2 inch nominal diameter letdown line at a location between containment penetration number 333 and the downstream isolation valve (MUV-49). As a consequence of the failure, the makeup control valve is assumed to move to the fully opened position to provide the maximum available makeup flow. This assumed control action delays the times for the trip of the reactor and the actuation of ESAS and consequently increases the releases of reactor coolant
( mass and the fission products to the auxiliary building.
14-71 (Rev.11)
O
e e
b 14.2.2.6 Makeup System letdown line Failure Accident
( 14.2.2.6.1 Identification of Cause A break in fluid-bearing lines that penetrate the reactor containment may result in the release of radioactivity to the environment. There are no instrument lines connected tc the RCS that penetrate the containment.
However, there are other piping lines such as those associated with the Makeup and Purification (MU) System and the Decay Heat Removal (DH) System that penetrate the containment. For fluid penetrations in piping systems that do not serve to limit the consequences of accidents, leakage is minimized by a double-barrier design to ensure that no single credible failure or mal function of an active component will result in either unacceptably high leakage or the loss of the capability to isolate a piping break. The installed dou"ble barriers consist of closed piping, both inside and outside the containment, and various types of isolation valves.
The most severe piping rupture identified for which radioactivity release may occur during normal plant operation is in the Makeup and Purification .?> stem.
This involves a rupture of the letdown line just outside the containment and upstream of the letdown control valves. A rupture at this point produces a loss of reactor coolant condition until the RCS pressure drops below the pressure for actuation of the Engineered Safeguards to isolate the reactor building. When inis pressure is reached, the building isolation signal initiates closure of the letdown isolation valves inside the containment.
Closure of the isolation valves stops the release of reactor coolant and fission products to the auxiliary building, thus terminating the loss-of-coolant phase of the accident.
k 14.2.2.6.2 Safety Evaluation Criterion The acceptance criterion for the evaluation of this accident is that the resultant doses shall not exceed 10CFR100 limits. (Dose limits are 300 rem thyroid dose and 25 rem whole body dose.)
14.4.4.6.3 Methods of Analysis The CRAFT 2 computer code was used to determine the reactor coolant mass release rates and the primary system response for the rupture of the letdown line. The multinode model includes a detailed model of the RCS as well as noding for simulation of the letdown piping, valves, and coolers.
For purposes of calculating the mass of reactor coolant released, the reactor is assumed to be operating at 2603 MWt with a letdown flow of 140 gpm prior to the rupture. The rupture is modeled as a complete severance of the 21/2 inch nominal diameter letdown line at a location between containment penetration number 333 and the downstream istlation valve (MUV-49). As a consequence of the failure, the makeup control valve is assumed to move to the fully opened position to provide the maximum available makeup flow. This assumed control action delays the times for the trip of the reactor and the actuation of ESAS and consequently increases the releases of reactor coolant
( mass and the fission products to the auxiliary building.
14-71 (Rev. 11) 4
_ . _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _~_-_-___._.m__._.-n
O e
5 Automatic actuation of ESAS is assumed to occur at a- pressure setpoint of
-1350 ~ psig,- which corresponds to the nominal value of 1500 psig with an adjustment for possible instrument error equal to 6% of the 2500 psig range (
of the measurement. The letdown isolation valve is assumed to reach the fully closed position 7.4 seconds after the ESAS pressure setpoint is reached. This time period includes both the instrumentation delay time and the valve stroke time.
Dose calculations are based on a core power level of 2544 MWt with the fission product concentrations corresponding to 1 percent defective fuel rods. Ten percent of the iodine contained in the mass of reactor coolant is assumed to volatilize and become airborne in the auxiliary building. The remaining 90% is assumed to remain in the liquid which drains into the auxiliary building sump. The airborne radioactive nuclides in the auxiliary building are filtered tifrough High Efficiency Particulate Air (HEPA) and
~ $p* //gy - cnarcoal filters in theA$uildingX VentilationiSystem before i being exhausted to the environment. The analysis is based on a conservatively estimated h,3 f-efficiency of 90% for iodine removal by the charcoal filters. The assumptions used in the evaluation of the off-site doses are summarized in Table 14-41.
14.2.2.6.4 Results of Analysis The calculated time for the RCS to depressurize and reach the actuation pressure for the ESAS is 745 seconds. At a time of 752 seconds, the isolation yalve is completely closed. The total mass of reactor coolant that escapes through the break and is released to the auxiliary building is 45,760 f pounds. \
l The fission product activities released to the environment during the accident are listed in Table 14-42. The dose consequences of the letdown line rupture accident are presented in Table 14-43. The table presents: (1) the thyroid dose due to inhalation of iodine activity; and (2) the whole body doses from gamma radiation due to immersion in the gas cloud for individuals located at the outer boundaries of either the exclusion area or the low population zone for the first two hours after the accident. The resulting doses are small fractions of the 10CFR100 limits.
b A dM ghgaw h ,,.*c0 Y A3Vh -
T4e /]B Y S h ,e he- -deAAg/& .4 &
h f.wtJV. YS N- 42 & W~ O
& a. & f d cbpu_. Mws& h a n aw a pQW f /0 Y A NY N(
[ g.cnw k AB VS '
d a p x 4 i - a NuL{ w y
& LPE.
Tha. M " d A n.s L sw wa san
e 6
TABLE 14-41 ANALYSIS ASSUMPTIONS FOR THE MU SYSTEM LETDOWN LINE RUPTURE ACCIDENT Data and Assumptions Used to Estimate Radioactive Source I
Power level, MWt 2544 Percent of fuel rods leaking, % 1.0 Escape rate coefficient Table 11-1 ,
Reactor Coolant Activity Nuclide - Activity. Micro-Ci/cc Kr 85m 1.48 85 4.36 87 0.779 '
88 2.41 Xe 131m 1.63 133m 2.58 !
133 238.0 135m 0.294 ,
135 4.88 !
138 0.421 l
I 131 3.47 ,
132 1.17 ,
133 3.7 134 0.461 !
135 1.88 l Data and Assumptions Used to Estimate Radioactivity Released j Total mass of reactor coolant released !
to auxiliary building, lb 45,760 Charcoal filter efficiency for m lodine, % 90 ;
Noble gas, % OA[o l Fraction of iodine airborne 0.1//,d,j !
Dispersio, Data I
EAB,' m 1340 l' LPZ boundary, m 8047 Atmospheric dispersion percentile, % 5 0-2 hour atmospheric dispersion factors, s/m 3 at EAB 1.6 E-4 at LPZ boundary 1.4 E-5 14-134 (Rev. 11) f w _
ob ReFIe=+s anal y ses w% ug w,%+ c cad,+ -fde n L t c d me
<. s > < w .i. , w , - w ,. - , ,,r .
L --- - _ _ _ _ _ - - - - _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ - ~ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _
o~
t.
.s TABLE 14-42 ACTIVITY RELEASED TO ENVIRONMENT DUE TL RUPTURE OF THE MU SYSTEM LETDOWN LINE Nuclide Actiyt v. Ci tco ( bD Kr 85m (4 . av.c 85 131.0 /J/. o j 87 23.5 tLr 88 72.6 ? ?. 6 Xe 131m 49.1 47 /
t 133m 77.7 77.7 133 7170.0 h 7/7c o 135m 8.85 ?,9 F 135 147.0 /en.O 138 12.7 /2.7 I 131 10.4 /O'/ 0 132 3.52 3 r. 2.
///, o 133 11.1 l 134 1.39 /3.'l 135 5.66 5%,G i
~
(a) w.h 9a% aod,ne lal% hen b y Yde a ux</ soy hw Ming ved. Ia Y;on s ysk a s s u m ed.
a)u.ff no r os.ne A/f d en by % qu>abany bu cld"3 Venhfa brori e- Sy5 Novi ctSSOmeN'
\ (- 14-135 (Rev. 11)
- ei .
c' h' ' , ,
poe l
TABLE 14-43 .~,
SUMMARY
OF RESULTANT DOSES FOR THE-H MU SYSTEM LETDOWN LINE RUPTURE ACCIDENL _
-/
To - tegrated Dose at Exclusion Bou y Thyroid, Rem 0.115 l-Body, R 0.066 L
I Total- Integrat ose at low Popu a Zone \
Thyroid, Rem O. M
.Whole Body, Rem 0.0058 h
'7.~kove. %%d L%c,eaf eet boses eit the Eye /usion jQma 8.un c/cey Tyrvict , REM LJhele 13edy, REM
~
' 90% AD V5 taclone $//rafoon 9, y 5- cy.
O'/o A6VS tadone falh ay,,n t f
30 y 75d / Z /eygg p, ,p e L w P p u k n, e n bne P'h REM Ohole B-&, pem
- Vo AB vS iod,n, pjgj O . o to I g, g.g O'S ffoyt to &q ffj #'
o . co.C 7 N
~
14-136 (Rev. II) -
.- .__ _ - -- n ----__--__L_ _ 2___ _