ML16004A357
| ML16004A357 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 12/17/2015 |
| From: | Pacific Gas & Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16004A363 | List: |
| References | |
| DCL-15-152, TAC MF6399, TAC MF6400 | |
| Download: ML16004A357 (97) | |
Text
{{#Wiki_filter:DCPP UNITS 1 & 2 FSAR UPDATE Note: Footnote 10 criterion is met in that peak fuel rod burnup is limited to 62,000 MWD/MTU.The elements in each radionuclide group released to the containment following a LOCA are assumed to be as follows (note that the groupings were expanded from that in Regulatory Guide 1.183 to address isotopes in the core with similar characteristics; the added isotopes are in bold font): Noble gases: Xe, Kr Halogens: I, Br Alkali Metals: Cs Rb Tellurium Grp: Te, Sb, Se, Sn, In, Ge, Ga, Cd, As, Ag Ba,Sr: Ba, Sr Noble Metals: Ru, Rh, Pd, Mo, Tc, Co Cerium Grp: Ce, Pu, Np, Th Lanthanides: La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am, Gd, Ho, Tb As discussed in Section 6.2.3.3.7, the design includes chemical addition into the containment spray system which ensures a long term sump pH equal to or greater than 7.0. Thus, the chemical form of the radioidine released from the fuel is assumed to be 95% particulate (cesium iodide (Csl)), 4.85% elemental iodine, and 0.15% organic iodine. With the exception of noble gases, elemental and organic iodine, all fission products released are assumed to be in particulate form.The activity released from the core during each release phase is modeled as increasing in a linear fashion over the duration of the phase. The release into the containment is assumed to terminate at the end of the early in-vessel phase, approximately 1 .8 hours after the LOCA.Isotopic decay, containment leakage, selected natural removal mechanisms and spray removal are credited to deplete the inventory of fission products airborne in containment. Containment spray in the injection and recirculation mode is utilized as one of the primary means of fission product cleanup following a LOCA. Mixing of the effectively sprayed volume of containment, with the unsprayed volume of the containment is enhanced by operation of the PG&E Design Class I containment fan coolers. In order to quantify the effectiveness of the containment spray system, both the volume fraction of containment that is sprayed, and the mixing rate between the sprayed and unsprayed-volumes are quantified. The LOCA analysis is based on an assumed worst case single failure of loss of one ESE train. A single train ESF consists of one train of ECCS, one train of CSS, and two Containment Fan Cooling Units (CFCUs). A single train scenario is selected to be consistent with the use of reduced iodine and particulate removal coefficients associated with single train operation. 15.5-62 15.5-62Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE a. Containment Spray Duration: Containment Spray in the injection mode is initiated at 111 seconds after the LOCA and terminated at 3798 seconds. Manual operation is credited to initiate containment recirculation spray within twelve (12)minutes after injection spray is terminated. Thus, based on single train operation, containment spray in the recirculation mode is initiated at 4518 seconds, and terminated 5 hours later at 22,518 seconds. In summary, containment spray operation (injection plus recirculation) is credited for 6.25 hrs post-LOCA, with a twelve minute gap after injection spray is terminated.
- b. Containment Spray Coverage:
As discussed in Section 6.2.3.3.7.1, the containment sprays are estimated to effectively cover 82.5% of the containment free volume during the containment spray injection as well as spray recirculation mode.c. Mixinq between Sprayed and Unsprayed Regions of Containment: The containment mixing rate between the sprayed and unsprayed regions following a LOCA is determined to be 9.13 turnovers of the unsprayed regions per hour.This mixing rate is based on the operation of two CFCU with a total volumetric flow rate that addresses surveillance margins and uncertainty, between the unsprayed regions and sprayed regions. Review of the layout and arrangement of the intake and exhaust registers of the CFCUs indicate that the air intakes are all located above the operating floor (sprayed region) and the air discharge registers are all located below the operating floor in the unsprayed region.Additional review of the containment configuration including the location of the major openings in the containment structure, and various active and passive mixing mechanisms, results in the conclusion that following a LOCA, credit can be taken for a) the entire flowrate provided by each operating CFCU to support mixing between the sprayed and unsprayed regions, and b) homogeneous mixing within the sprayed and unsprayed regions, of the volume of air transferred from one region to the other due to CFCU operation. In accordance with Regulatory Guide 1.183, Appendix A, Section 3.3, prior to CFCU initiation, the dose consequence model assumes a mixing rate attributable to natural convection between the sprayed and unsprayed regions of 2 turnovers of the unsprayed region per hour.d. Fission Product Removal: The fission product removal coefficients developed for the LOCA reflect the following guidance documents: i.Elemental iodine removal coefficients are calculated using guidance provided in Standard Review Plan Section 6.5.2, Revision 4 (Reference
- 80) which is invoked by Regulatory Guide 1.183, Appendix A, Section 3.3 i.Time dependent particulate aerosol removal coefficients are estimated using Regulatory Guide 1.183, Appendix A, Section 3.3, which permits the use of time-dependent particulate aerosol removal coefficients by invoking NUREG/CR 5966, June 1993 (Reference 81), and indicates that no reduction 15.5-63 15.5-63Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE in particulate aerosol removal coefficients is required when a DF of 50 is reached, if the removal rates are based on the calculated time-dependent airborne aerosol mass. There are several aerosol mechanics phenomena that-promote the depletion of aerosols from the containment atmosphere.
For DCPP, agglomerati~on of the aerosol is considered in both sprayed and unsprayed regions. In the sprayed region, the particulate removal calculation takes credit for the removal effectiveness of sprays and diffusiophoresis (aerosol removal due to Steam condensation). Computer program SWNAUA is used to develop the time dependent particulate aerosol removal coefficients which reflect the effect of diffusiophoresis and sprays. Gravitational settling is considered only in the unsprayed region.The methodology used to develop the elemental iodine and particulate removal coefficients in the sprayed and unsprayed region of the containment is discussed in Section 6.2.3.3.7.2. The total elemental iodine and particulate removal coefficients in the sprayed and unsprayed region of the containment as a function of time are summarized in Table 6.2-32.In summary, the activity transport model takes credit for aerosol removal due to steam condensation and via containment spray based on spray flowrates associated with minimum ESF during the containment spray injection and recirculation mode. It considers mixing between the sprayed and unsprayed regions of the containment, reduction in airborne radioactivity in the containment by concentration dependent aerosol removal lambdas, and isotopic in-growth due to decay.During spray operation in the irnjection mode, the elemental iodine removal rate for the sprays exceeds 20 hr 1 , the maximum value permitted by NUREG-0800, Standard Review Plan Section 6.5.2; thus the elemental iodine removal rate attributable to sprays is limited to 20 hr-1.During recirculation spray operation, the elemental removal rate for the sprays is 19.34 hr-1.As discussed in 6.2.3.3.7.2, the wall deposition removal coefficient for elemental iodine has been calculated with the model provided in NUREG-0800, SRP Section 6.5.2. In sprayed and unsprayed regions, prior to spray actuation, the wall deposition removal coefficient is estimated to be 2.74 hr 1 , while during spray operation, and in the sprayed region only, the wall deposition removal coefficient is estimated to be 0.57 hr-1.In the unsprayed region, the aerosol removal lambdas reflect gravitational settling. No credit is taken for elemental iodine removal in the unsprayed region.Since the spray removal coefficients are based on calculated time dependent airborne aerosol mass, there is no restriction on the DF for particulate iodine. The maximum DF for elemental iodine is based on Standard Review Plan Section 6.5.2 and is limited to a DF of 200.Radioactivity is assumed to leak from both the sprayed and unsprayed region to the environment at the containment technical specification leak rate-for the first day, and 15.5-64 15.5-64Revision 19 May 2010 DCPP UNITS I & 2 FSAR UPDATE half that leak rate for the remaining duration of the accident (i.e., 29 days). To ensure bounding values, the atmospheric dispersion factors utilized for the containment release path reflects the worst value between the containment wall release point, the plant Vent, the containment Penetration Area GE (EL 140') and the Containment Penetration Areas GW/FW (EL 140').As a re.ult of the: pre..suri.zation of the co..tainmen..t follwin ...LOC.Ateeia day after first day.: Thes a..sumed. rate are.. c.n.istent with Technical-. prev.iou sections,; .. it h the assump..tion, that some of the removal system. do.no reduc cotimn pressure' to: atmospheric folowing thge iniia pressure ,tc......... the leakage4:...f!5... 55 ÷...;15..172. Cotaimet Lakge xpour Sesitviy Sud Sensiti.it 2tuie .. oren perform.....d to,, ilstat dependenc.... of th thyroi e.....ur exosre ar~f~dasa15.5-65r fr~~ pryReeva!sion 19 Mand01 DCPP UNITS 1 & 2 FSAR UPDATE gu'ideline lev'el ....fie 1,.0 CFR °,"t 100.15.5.17.2.5 R)adiological.o Con..eque...ce. wth DF of+ 100 doexcusonsequence .. sing/l a cotanmn decont.amination factor of 100.an varcntinmnOiinUSwat f900cm A cntinentmiin rte f 1,00cfmcorepods it or urrntmiimmdein basi!72 opertio of ....containment. coole unt (CFCU. Cluainswr ae invetor source terms fo the. various...... fuerodtom wer calculaedusngth 800meerexlusonara5ounar-6AB at2husadte1,0Reerisiow 9My21 DCPP UNITS 1 & 2 FSAR UPDATE exposures do not include the effects of any population redistribution duo to evacuation. These exposures were calculated using the EMERALD computer codc. The tiw~~uppuuuiu~ruuir uie ii /e /uu~ir exposure calculations are discussed in Section 15.5.5.15.5.17.2.8 Offsitc Exposures from Post LOCA Recirculation Loop Leakage in the A wiPar" Building coolant water that collects in the containment Reactor recirculaLlon sump after a LOCA would contain radioactive fission products.15.5.17.2.2.3 ESF System Leakage Outside Containment The fluid that collects in the containment recirculation sump after a LOCA (i.e., the fluid contents of reactor coolant system, the RWST, the NaOH tank and the accumulators) contain radioactive fission products that has been released from the core as a result of teLC.Bcuecnanetrecirculation sump water is circulated outeRR upcoldatsideH thea exchangers, returned to the containment via the RHR system piping and the CSS piping.(if, ..ircltion spray is uJsed, passed through the RCS and the containment spray nozzles" ( " r ÷iruato spray. is... used)...;... ,, and finally returned to the containment recirculation sump. In the event of circulation loop leakage in the auxiliary building, post-LOCA activity has a pathway to the atmosphere.,An illustration... ofhis- pathwa f.... a" small leak is. gie in Fiur 15.5. 9. t he small S~ILU~UUII, ii~iun piuuu~~ ui w~ ~ ~ie pu~u Lu JuAwdIy uwru~uy ventilation air flow for a long period of time. Thus, for the small leakage situation, all activity released to the auxiliary building ~.A.Iould be released to the auxiliary building air, i.e., no credit for liquid gas partitioning. An illustration of post LOC,~. activity pathway for a large leak is given in Figure 15.5 10.For the large leakage situation, fission products in the leakage water are exposed to auxiliary buildina ventilation air flow for a short ocriod of time. Thus. most of the activity The complete RHR system and CSS description, including estimate of leakage, detection of leakage, equipment isolation, and corrective maintenance, are contained in Sections 5.5.6 and 6.2.2, respectively. In accordance with Regulatory Guide 1 .183, with the exception of noble gases, all the fission products released from the core during the gap and early in-vessel release phases are assumed to be instantaneously and homogeneously mixed in the primary 15.5-67 15.5-67Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE containment recirculation sump water at the time of release from the fuel. A minimum sump water volume of 480,015 gallons is utilized in this analysis.In accordance with Regulatory Guide 1.183, the ESF systems that recirculate sump fluids outside containment are analyzed to leak at twice the sum of the administratively controlled total allowable leakage applicable to all components in the ESF recirculation systems. With the exception of iodine, all radioactive materials in the recirculating liquid are assumed to be retained in the liquid phase.ESF leakage is assumed to occur at initiation of the recirculation mode for safety injection. Since the maximum temperature of the recirculation fluid supports a flash fraction less than 10%, per Regulatory Guide 1.183, ten percent (10%) of the halogens associated with this leakage are assumed to be airborne and are exhausted (without mixing and without holdup) to the environment. The iodine release from the core is 95%particulate (Csl), 4.85% elemental and 0.15% organic, however after interactions with sump water the environmental release is assumed to be 97% elemental and 3%organic.The environmental release of ESF system leakage can occur via the 2 pathways listed below.a. Environmental release of ESF System leakage via the plant vent: The sum of the maximum allowable simultaneous leakage from all components in the ESF recirculation systems located in the auxiliary building is limited to 120 cc/mmn.Thus, and in accordance with the requirements of Regulatory Guide 1.183, the analysis addresses an ESF leakage of 240 cc/mmn in the auxiliary building. The areas where these components are located are covered by the PG&E Design Class I ABVS which discharges to the environment out of the plant vent. Only selected portions of the Auxiliary Building ventilation system are processed through the PG&E Design Class I AB ventilation filters. For purposes of estimating the dose consequences, it is assumed that with the exception of the RHR pump rooms (refer to Section 7.2.3.4), this release pathway bypasses the PG&E Design Class I AB ventilation filters.b. Environmental release of ESF System leakagqe via Containment Penetration Area GE and Areas GW & FW: The sum of the maximum allowable simultaneous leakage from all components in the ESF recirculation systems located in the containment penetration areas is limited to 6 cc/mmn. Thus, and in accordance with the requirements of Regulatory Guide 1.183, the analysis addresses an ESF leakage of 12 cc/mmn in the containment penetration areas. The ventilation system covering this area is not PG&E Design Class I, thus the release path to the environment is unfiltered and could occur via the Plant Vent or via the closest structural opening in the Containment Penetration Areas GE and Areas GW &FW.15.5-68 15.5-68Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Post LOCA aux~iar,' building loop leakage exposures were calculatcd for four differcnt leakage-Gases 7 (1) Expected small leakage case (2) Expected large leakage case (3) DBA small leakage case (1) DBA large leakage case numerical values used to calculate leakage ~ are and loop in Table 15.5 21. Table 15.563 shows the resu!ts of the calculations based on these assumptions. Because an insignificant amount of noble gases would be in the containment recirculation sump water, the whole body exposures are negligible. One possible approach to the evaluation of offsite exposures from post LOCA recirculation loop leakage would include the following assumption 5: (1) A LOCA, as an initiating event (2) Failure of two ECCS trains resulting in gross fuel damage: Release of 50 percent of core iodine inventor; and 100 percent of core noble gas inventor'; to the containment (3) Failure of an RHR pump seal, resulting in the release of a significant amount of the above containment activity to the auxiliar,' building (1) Failure of the passive auxiliar,' building charcoal filters resulting in the unfiltered release of iodine fission products to the environment The assumption of this sequence of failures for analysis of offsite exposures, ~would be requiring plant design features in excess of the current guides and regulations, and in particular the requirements of ANS Standard N18.2, Nuclear Safety Criteria for the Design of Stationar; Pressurized Water Power Plants. (See proposed addendum to Ad.~ Standard ~ Failure Criteria for Fluid ~"~~ems MC' p.IAO'~ Single ,.. (Reference ~~~l"in~ the standard to IC'iC'A loop the IDIC~C'A~ proposed post recirewation leakage was assumed as the initiating event: "The unit shall be designed to tolerate an initiating event which may be a single active or passive failure in any system intended for use during normal operation." The ECOS was assumed to function properly, as required by the ECCS acceptance criteria, preventing gross fuel damage. Although meeting these criteria is expected to preclude gross cladding damage, it was assumed for this analysis that 100 percent of 15.5-69 15.5-69Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 15.5.17.2.2.4 RHR Pump Seal Failure fa-iluf-Failure of an RHR pump seal was assumed to be as-the worst case single failure and-eanto be tolerated without loss of the required functioning of the RHR system, as was required by the following clauses in the p~r-ep~eaddendum to the ANS Standard N18.2 proposed at the time of original license: "Fluid systems provided to mitigate the consequences of Condition Ill and Condition IV events shall be designed to tolerate a single failure in addition to the incident which requires their function, without loss of the function to the unit."A single failure is an occurrence which results in the loss of capability of a component to perform its intended safety functions when called upon. Multiple failures resulting from a single occurrence are considered to be a single failure.Fluid and electrical systems are considered to be designed against a single failure if neither (a) a single failure of any active component (assuming passive components function properly); nor (b) a single failure of a passive component (assuming active components function properly) results in a loss of the safety function to the nuclear steam electric generating unit."An active failure is a malfunction, excluding passive failures, of a component which relies on mechanical movement to complete its intended function upon demand."Examples of active failures include the failure of a valve or a check valve to move to its correct position, or the failure of a pump, fan or diesel generator to start."Spurious action of a powered component originating within its actuation system shall be regarded as an active failure unless specific design features or operating restrictions preclude such spurious action."A passive failure is a breach of the fluid pressure boundary or blockage of a process flowpath." For the expect. a,, nd4 DBA lag leakage1, cases. the=The failure of auxiliary building charcoal filters, a second failure, was not assumed, in accordance with the standard.For the e.pecte DA smal leaag cases.., failur of auxiliar,'buildin~g charcoal requ;re function o the buildin ventiatio sytm .. hich pro.des.oolig.fo 15.5-70 15.5-70Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE ECCS components.o For. the. long term small lea.age cas..., thec-harcoal filters re.. not. needed.+, to.+, reduc in41.. theL isyse r eudnadol h psiecaca e tesle five fa coolers tw4 o containmen-.+.,t spray trains functioned. For the DBA^ small and lag laag ass itiis assume that two;l~ll ECCIS tan, a coles an n For the e.pecte and lag leakage..I cases,. the as.u.ed g4 ap iodine elementa ÷l iodine,. For, the DBA small, and,, lag leakage,+ cases.. the,, gap iodine invento÷rriesa -,re boased on releasea fractions iven; i.. ..n Safety Guide,2,4 19"72 (Reference 23). TheU IB caseU gap toIbe 997 eretelmna iodin ...e-,4 0.25= percen .....t.. organic; iodine per,;.. , Gu_,de 1,972 .l.eaag cases.. for both the tim periods,4( before and4 during loop ,eaag.., No credit+ was water make. up.. the total, volume ..ater, in. which a ct;,,{ is deoste. Comnside4ration o~f ,÷le3akae period.Qreni,,dti h,,,rnvn .... osdrtino mren cr oln necinforts...d,; ...... ;1...-.-.
- ..... ----t;r h Pn~ir leakag water.*'-
for the four leakage case..Frth.are..kgeae ,,,e~ in the spra.. wate.r du,.ring the 30 minute leakage period tion The, d4esign evaluatfion cnductedr fo, rr the comntainmerntfunct~r-ional d4esign the 15,5-71 15. -7 1Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Table 1 5.5 21lits to he ....u..d temperatu.e of recicu.aio... loop,,. .. ater for t,, he the temperatur w....hen the, beg..n. No. credit was take'n for the" decreas of..A review of the equipment in the RHR system loop and the CSS loop indicates that the largest leakage would result from the failure of an RHR pump seal. Evaluation of RHR pump seal leakage rate, assuming only the presence of a seal retention ring around the pump shaft, shows that flows less than 50 gpm would result (refer to Section 3.1 and Chapter 6). Circulation loop piping leaks, valve packing leaks, and flange gasket leaks are much smaller and less severe than an RHR pump seal failure leak.-Laae-rm these compo~ne-ntsf du1ring no~rmal post operatio~n of the RHRJ loopn and the C.SS... .loopis estimate to be 1n1 cc/hr. (C~hapter6). On this basis, a 50 gpm leakrate was assumed for both the e.pecte large case... and the DBA large.JeakageLOCA.... ce, and a~ 1 .,10 , cc/h learat was.. assum..d f4or both thc e..p.cted, For the DBA lag ca...e,. recic...io lopLO,,,,/CA pump seal leakage was assumed to commence 24 hours after the start of the L-BLOCA. This assumption is consistent with the discussion in Sections 3.1.1 .1 and 6.3.3.5.3, and with the guidance in Standard Review Plan 15.6.5, Appendix B. In this context, the limiting recirculation loop long term passive failure is 50 gpm leakage at 24 hours after the start of the LBLOCA.Evaluation of an RHR pump seal failure shows that the failure could be detected and the pump isolated well within 30 minutes (Chapter 6). A-Thus a leakage duration of 30 minutes is conservatively assumed for the,. exp.cte the DBA LOCA~a-ge-le Rage-cases. 15.5-72 15.5-72Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Factor (PF) for a pad;cular istop (Equation 15.5 7).For both the e..pected and DBA large Ileakage.ater was pumpod aWay tot ÷he flor,,, r-wa asue be exposed buil,4i 6ase&7 4~-w~4ileti4 n4Rg-w as assumed that leakage Iodine in the leakage water n air flow for a shod period of as assumed for e!emental (Chapter 6) ,! a) f+ahing prcssms1b5osiee.5o-ha!neg h~-hfo -hf x (5. 1-hf -h (15.5-12)hg -hf---initial enthalp of. liquid',4 Bt"i'bm ,,e -final entfhalpy, of liquid, tulbfIm M -po fiax nhlofvpr t/b Mvp -(15.5 Mliquid 1-x Figures.15.5 1 and 15.5 12 present.. ,the, expected1,., and B ag laaees
- c,,iodin p=s as. a functio -f both te'mpeatre*i, and pH. For small P,, t he OF insensitiv to ..ate÷r tempeature bu.. t, much more to pH.Table 15 ....5 21lsth as...umed....temp
..eraue and.p..a.ong..th.the.esultin 15.5-73 15.5-73Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE For bot+h thc expected-a+,+nd ^ ,B, s.mall leakag ca.....,..t ..a..a..umed t4 hat a-DF-ef-!. factor of-1~For al four loop leakage,, cases no.. crdi Leas aken= for auxiliarbuildingradiologia de a or~ f ,isio prd c pN~l l ateouIV t.I IVl iV kkv /$ VWIV Fr' theIexpected, DBrA cas, for', h auxiliar,,' buildng charcoa l, was-buildRIng chrolfitr t (as... prvoul discuss......d wit refere nce to !oN expsurs tat ccu vi ths5cmbiatin4o unieyeetRol ewl eison thea 21 DCPP UNITS 1 & 2 FSAR UPDATE 0."186 ..... where tIhe airorn .c... it The limitation the1t GDC 19 1971 allowable dose for the control room.In summary, the RHR pump seal failure resulting in a filtered release via the plant vent is DCPP's licensing basis with respect to the worst case passive single failure in the RHR system. Therefore, the RHR pump Seal Failure is retained as a release pathway for the AST LOCA dose consequence analysis.The activity transport model is based on a 50 gpm leak of sump water activity for 30 minutes that occurs 24 hours after the LOCA. The temperature of the recirculation fluid is conservatively assumed to remain at the maximum temperature of 259.9 0 F. Thus as discussed above in Section 15.5.17.2.2.3 under ESF system leakage, the amount of iodine that becomes airborne is assumed to be 10% of the total iodine activity in the leaked fluid.The ventilation exhaust from the RHR pump rooms is covered by the PG&E Design Class I Auxiliary Building ventilation system and processed through the PG&E Design Class I AB ventilation filters. Thus, credit for filtration of the release of a RHR pump seal failure by the Auxiliary Building Ventilation system is taken in determining the dose consequences to the public at the EAB and LPZ, to the operator in the control room, and to personnel in the technical support center.Credit-for filtratio~n-of-the-,release.,. of a RHR Sste pump... sea failre b..y the .u....a;,-,, building .e.tiatio system.. ;s in- detIe'-,rmining, dose co.n, que.,-,,nces.oe public' at2, EAB and, LP,",tI co-nt-rol, roomY a'nd TSC.The efficiency of the auxiliary building charcoal filters is determined using methodology similar to that documented in Section 15.5.9 for the CRVS Mode 4 ventilation filters. The allowable methyl iodide penetration / filter bypass for the auxiliary building charcoal filter is controlled by DCPP Technical Specification 5.5.11;and are 5% and <1%, respectively. Based on the above, an efficiency of 88% is assigned to the charcoal filters in the AB ventilation system prior to environmental release via the plant vent. Similar to the ESF system leakage, the environmental release of iodine is assumed to be 97% elemental and 3% organic.15.5.17.2.2.5 Refueling Water Storage Tank Back Leakage The safety injection and containment spray systems function to provide reactor core cooling and mitigate the containment pressure and temperature rise, respectively, in the event of a LOCA. Both systems initially take suction from the RWST. Once the RWST water supply is depleted, both the containment spray and safety injection systems are supplied by the RHR System. The RHR pumps take suction from the containment recirculation sump water. Under LOCA conditions, the recirculation sump water is assumed to be radioactively contaminated by fission products, of which the main contributors to airborne dose are the various isotopes of iodine.15*5-75 15.5-75Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE As discussed in NRC Information Notice 91-56, September 1991 during containment sump water recirculation, there is the potential for leakage from the mini-flow recirculation lines connecting the high head and low head safety injection pump discharge piping to the RWST. Since the RWST is vented to the atmosphere, this presents a pathway for iodine release to the atmosphere. The acceptance criteria in the DCPP administrative test procedures ensure that the total as-tested back leakage into the RWST from the containment recirculation sump is less than or equal to 1 gpm.Dose consequences of RWST back-leakage assumes that leakage starts at the switchover to recirculation following the LOCA and continues for 30 days. Per regulatory guidance, a safety factor of 2 is applied to the leak rate, i.e., a 2-gpm leakage rate is assumed for the full duration of the event, which is two times the allowable leakage of 1 gpm. With the exception of noble gases, all fission products released from the fuel to the containment are instantaneously and homogeneously mixed in the sump water at the time of release. Only iodine and their daughter products are released through RWST back-leakage since the particulates would remain in the sump water.A significant portion of the iodine associated with sump water back-leakage into the RWST is retained within the RWST fluid due to the equilibrium iodine distribution balance between the RWST gas and liquid phases. The time dependent iodine partition coefficient takes into consideration the temperature and pH of the RWST liquid and sump fluid, the RWST liquid and gas volumes, and the temperature, pH and volume of the incoming leakage. The iodines that evolve into the RWST gas space as a result of the equilibrium iodine distribution balance, and the noble gas daughters of iodines, are released to the environment via the RWST vent, at a vent rate established by the temperature transient in the RWST (which includes the effect of decay heat), the increase in the liquid inventory of the RWST due to the incoming leakage, the gases evolving out of incoming leakage, and the environmental conditions outside the RWST.The average time-dependent RWST iodine release fractions along with the fractional RWST gas venting rates (may be applied to the noble gas daughters of iodines) to the atmosphere from the Unit 1 and Unit 2 RWSTs due to RWST back-leakage following switchover to the sump water recirculation mode of operation is summarized in Table 15.5-230. As discussed earlier, the releasefractions / rates presented in Table 15.5-23C reflect a safety factor of 2 on the leak rates, i.e., are developed based on a RWST back-leakage of 2 gpm. The iodine released to the environment is assumed to be 97%elemental and 3% organic.The equilibrium iodine concentration in the RWST gas space utilized to develop Table 15.5-230 is based on the iodine mass in the sump fluid entering the RWST vapor space as back-leakage or the total iodine mass contained in the RWST liquid, whichever results in higher RWST vapor phase concentrations. The RWST maximum venting rate averaged over an interval is primarily based on RWST back-leakage entering the RWST gas space and thermally equilibrating, and is used in conjunction with the higher RWST gas space iodine concentration to calculate an iodine mass release rate as a function of time. An interval based averaging approach is utilized in preparing Table 15.5-230 to 15.5-76 15.5-76Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE reduce the number of input values to the dose analysis white preserving the boundaries for the time periods used for atmospheric dispersion; the actual iodine release calculated in an interval is normalized to the iodine mass leaking into the RWST during that time interval.Examination of the average gas space venting rates indicate that after the first day, the noble gases formed by decay of iodine will primarily remain in the RWST during the 30 day period of evaluation and not be released. However, the dose consequence analysis conservatively releases the noble gases formed by decay of iodine, directly to the environment without taking any credit for tank holdup.15.5.17.2.2.6 M~iscellaneous Equipment Drain Tank (MEDT) Leakage The DCPP Unit 1 and Unit 2 MEOT is a covered rectangular stainless steel lined concrete tank located in the auxiliary building below El 60 foot. The MEDT tank vent is hard-piped to the auxiliary building ventilation ductwork; thus the airborne releases from the MEDT are ultimately discharged to the environment via the plant vent (refer to Section 9.4.2).Following a LOCA, the MEDT will receive both post-LOCA sump fluids as well as non-radioactive fluids (i.e., ESF system leakage from the accident unit as well as non-radioactive fluids from equipment drains / RWST leakage from the non-accident unit)which are hard-piped to the MEDT. The acceptance criteria in the DCPP administrative test procedures ensure the total as-tested flow hard piped to the MEDT is less than 950 cc/mmn of ESF system leakage and 484 cc/mmn of non-radioactive fluid leakage.Similar to the RWST back-leakage model, dose consequences due to releases from the MEDT assumes that leakage starts at the switchover to recirculation (829 second following the LOCA) and continues for 30 days. Per Regulatory Guide 1.183, a safety factor of 2 is applied to the leak rate, i.e., 1900 cc/mmn of ESF system leakage and 968 cc/mmn of non-radioactive fluids into the MEDT is assumed for the full duration of the event, which is two times the allowable leakage. With the exception of noble gases, all fission products released from the fuel to the containment are instantaneously and homogeneously mixed in the sump water at the time of release. Only iodine and their daughter products are released through MEDT leakage since the particulates would remain in the sump water.The methodology used to determine the post-LOCA iodine and noble gas releases via the MEDT vent and Plant Vent is similar to that used to address RWST back-leakage. Adaptation of the methodology to address overflows/room ventilation releases is straightforward with the room ventilation rate being treated as the tank exhaust rate.The transport model utilized to determine airborne releases from the MEOT takes into account the fact that the MEDT is a small tank with an auto-transfer capability which is PG&E Design Class II. Consequently, and for purposes of conservatism, it is assumed that a) the LOCA occurs when the MEDT water level is at the normal maximum setpoint 15.5-77 15.5-77Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE to initiate auto transfer, b) the auto-transfer capability is not initiated because it is not a safety function, and c) the MEDT contents will spill over into the Equipment Drain Receiver Tank (EDRT) Room after the tank is full. Thus, for the post-LOCA scenario, the MEDT is conservatively assumed to overflow via its manway into the EDRT Room.The EDRT room drains into the auxiliary building sump, which ultimately overflows into the Unit 1/Unit 2 pipe tunnels. The auxiliary building sump is also a covered stainless steel lined concrete tank with a vent that is hard-piped to the auxiliary building ventilation system (ABVS) with a PG&E Design Class II auto transfer capability. The auxiliary building sump is located adjacent to the MEDT.The bounding transient release of iodine along with the gas venting rate to the atmosphere as a result of post-LOCA leakage of radioactive and non-radioactive fluid hard-piped into the MEDT is developed in 2 parts: a) prior to MEDT overflow and b) post MEDT overflow.a) Prior to MEDT overflow -The iodines evolve into the MEDT gas Space as a result of the equilibrium iodine distribution balance between the MEDT gas and liquid phases (either the MEDT liquid inventory or the incoming leakage), and are released to the environment via the plant vent, at a vent rate established by the temperature transient in the MEDT (including the effect of decay heat), the increase in the liquid inventory of the MEDT due to the incoming leakage, and the gases evolving out of the incoming leakage., b) After MEDT overflow -The equilibrium iodine distribution balance is conservatively assumed to be between the iodine concentrations in the MEDT overflow liquid and the EDRT room (or Unit 1/Unit 2 pipe tunnels) ventilation flow (rather than the average concentration in the EDRT room (or Unit 1/Unit 2 pipe tunnels) free volume). This maximizes the iodine release rate. Thus, the iodines released are a sum total of the following: i) the iodines that evolve into the EDRT room air space as a result of the equilibrium iodine distribution balance between the spilled liquid from the MEDT (at the temperature of the MEDT), and the EDRT room ventilation flow, and is released to the environment via the plant vent, at the vent rate established by the EDRT room ventilation system, and ii) the iodines that evolve into the Unit 1/Unit 2 Pipe Tunnel air space as a result of the equilibrium iodine distribution balance between the spilled liquid from the MEDT (at the maximum temperature of the Unit 1/Unit 2 Pipe Tunnel), and the U1/U2 Pipe Tunnel ventilation flow, and is released to the environment via the plant vent, at the vent rate established by the U1/U2 Pipe Tunnel ventilation system.The exhaust fans servicing the EDRT room and pipe tunnel are PG&E Design Class I.There is also a potential that the non-LOCA unit's ABVS will be operating with the flow exhausting to its unit specific plant vent. Thus, it is conservatively assumed that the 15.5-78 15.5-78Revision 19 May 2010 DCPP UNITS 1 & 2 FSARIJPDATE non-LOCA unit's ABVS is also operating, and together with the accident units' exhaust fans, are providing the motive force to exhaust the airborne releases to the respective unit vents.The average time-dependent MEDT iodine release fractions, along with the fractional MEDT gas venting rates (which may be applied to the noble gas daughters of iodines prior to MEDT overflow) to the atmosphere following switchover to the sump water recirculation mode of operation, is summarized in Table 15.5-23D. As discussed earlier, the release fractions / rates presented in Table 1 5.5-23D reflect a safety factor of 2 on the leak rates, i.e., are developed based on an input of 1900 cc/min of ESF system leakage and 968 cc/mmn of non-radioactive fluids into the MEDT. Through the use of extremely conservative assumptions, the calculated iodine release fractions /.gas venting rates presented in Table 7.2-4 when used in combination with the analyzed ESF system leak rate, bound the iodine releases of all combinations of radioactive and non-radioactive leakages less than or equal to the leak rates analyzed. The iodine released to the ventilation system is assumed to be 97% elemental and 3% organic, and is released to the environment via the plant vent. In addition, the dose consequence analysis conservatively releases the noble gases formed by decay of iodine, directly to the environment without taking any credit for tank holdup.15.5.17.2.3 Offsite Dose Assessment Due to the delayed post-LOCA fuel release sequence of an AST model, and the rate at which aerosols and elemental iodine are removed from the containment, the maximum 2-hour EAB dose for a PWR LOCA typically occurs between 0.5 hrs to 2.5 hrs.To establish the "worst case 2-hour release window" for the DCPP EAB dose, the integrated dose versus time for each of the six pathways discussed in Section 15.5.17.2.32 was evaluated. The 0-2 hr EAB Atmospheric Dispersion Factor from Table 2.3-145 was utilized for all cases.The analysis demonstrated that for DCPP the maximum 2 hour EAB dose will occur, as a result of the RHR pump seal failure, between T=24 hrs to T=26 hrs, and is unrelated to the post-LOCA fuel release sequence associated with AST.The direct shine dose at the EAB due to a) the airborne activity inside containment, and b) the sump water collected in the RWST due to RWST back-leakage, was also evaluated. Based on the results of the EAB evaluation which determined that the dose contribution due to direct shine was minimal (<0.01 remn), the dose at the LPZ due to direct shine is deemed negligible. The bounding EAB and LPZ dose following a LOCA at either unit is presented in Table&%3-1-4515.5-23. 15.5.17.2.90Offsitc Exposures from Controllcd Post-accident Contah'iet 15.5-79 15.5-79Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Ventkig Bec....e of reeas of.. signific,-ant ..mount of hydrogen to the containmen"t controlling the post acc-ident concnt~ratiofn of., hydtrogen in the containment atosphehre. Reunan thra yrgnrcmies ar hvrmr enso otacdn hydrgenconroL s abacupcontolld cntanmen vetin (va th cotaimen hydrogen~~~~~~~~ ~ ~ ~ ~~~ pug ytm ihofhr lw iddrcinsrmn hett og v,an..nrl ..~,, ha r.nrriarl nut rig Irma tha an,-, ,rranrma af ,.ffehara ,,AnrIr.i .is blow,.ing offshore, if necessar. atf al11 an eva.luatfion is presented in the follow.ing to determine potnt~ial exposulres if venizngi wetre carriedot4 duf ~ring onshore wins-~U'f:IIf]rl fl fflI1!IIE1'-' Iflf-~ ~rV1PIHI~.~ 11! F1fi~I ~1('PI!1f9I1I [1f1fflh1f-~!l rir 11(11 'Fl If~!1 ~1lif1 ........................................... i'~~r...... .......... 'n n ,I t n n + c " ..... .... ..... ..n~.. ...k r n-n mi t "', r, ' 1" .% ll... In 'a .... n~ I ptittb nI tii .Li ti, ii liii Lii[ It. dn iLt.iiii,,. .iiL ,.LiiiSL ,Sni siti. iLL) t.SjiiLit~i. tL. ~iist~imiiiii~iit Vti, i int r-IL alodsrbdi Secio~n 6.2.5. The ug sra is wihdlrawn frmtecotimn th~rougrh one of hs'o penet~ration lines. The is routed througnh a flow;s measu=ring-A.TI [1.0 -0.01FILEFF(l)]60 f" T(2)v N A Y iP- ltd VOLUME T(1)whefee... ACT(I)\VOLUJME -VENRAT-FIEFF(EEI) --n rLr-nA 1L+/- ~.+/-:-+/- ...L!.. L....X.... L...I t I -T-'rl i TT~r I I I IIn1I r:lnnif~l~nmrlT '.!rl~nh n rlf-nlnl-nr T(0~ -fr,-.,.. ~ I C~CA +h~+ ,-on+~nmanf ~,an+mna onrln, hr i ftiinit. t.it.n~ .... ,* iiL~nt-.L .SI.i~ii,*i~iiiiiii Vt.iiiii* ,,v inn t -time, hr 15.5-80 15.5-80Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 60 -minutes per hr The above equation considers radiological decay during the time period prior to containment venting and the time period during containment venting. It also assumes that the LBLOCA activity released to the containment atmosphere is homogeneously dispersed throughout the containment atmospheric volume. Exposures from activity released to the atmosphere were calculated using the EMERALD computer code.EMERALD assumes there is no radiological decay during the atmospheric dispersion. Containment venting exposures were calculated for both the expected case and the DBA case. Assumptions and numerical values used to calculate venting exposures are itemized in Table 15.5 28. Onshore controlled containment venting thyroid and whole body exposures are listed in Thblc 15.5 29.Post accident containment venting schedules are evaluated in Section 6.2.5. Assuming the venting system will operate an average 2 hours per day, the system flowrates during shod venting periods are 120 cfm (expected) and 300 cfm (DBA). Equivalent continuous venting rates, 10 cfm and 25 cfm, were used to calculate venting activity f~eaee&In the event containment venting should be required during periods with onshore flow, the venting '.vould be limited to those periods when Pasquill Stability Catcgo~ D exists.Catego~ D and an elevated release height of 70 meters were evaluated using a conventional Gaussian plume mode! and are listed in Table 15.5 30. The meteorological input parameters utilized were determined from onsite measurements, given in References 18, 19, and 20. Because an individual is assumed to be located on the plume centerline for the entire venting duration, exposures are centerline exposures and represent worst case conditions. Tho probability of an individual being located on the plume centerline for a 2 hour period is ver small, and thus centerline exposures listed in Table 15.5 29 are ver,' conser~ative. Duing thel time~ period~v pro toI ventigl~ acivt release theI cotinment atmosphere is infcnl eue ybt ailgcldcyadfntoigo h aeyfaue sytm. h an otiutr fraiatviysvra udedhusafe h It can be concluded from the results presented in Table 15.5 29, along with the consideration of the yen; high probability of oppo~unities for offshore venting and the other favorable factors associated with the DCPP design and site, that, as a backup to the internal hydrogen recombiner system, controlled venting using the containment hydrogen purge system is an acceptable contingency met hod of post accident hydrogen control for this plant. In addition, it can be concluded that the expected exposuros due to venting, even using the assumptions in Safety Guide 7, will not exceed the annual 15.5-81 Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE dose limits of 10 CFR Part 20.15.5.17.2.0---4 Post LOOA-aG4de1at Control Room Operator Exposures The design basis for control room ventilation, shielding, and administration is to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDEwhole body,.." its equ.i.=..valent'" to..n.. par th body, for the dur...tio of the most ÷ e.ere. deoign, basis accident. Thiso basis is consistent with GDC 19, 1971.The control room shielding, described in Section 12.1 is designed to attenuate gamma radiation from post-accident sources to levels consistent with the requirements of GDC 19,49-7-1-1999 and 10 CFR 50.67.The control room ventilation system is described in Section 9.4.4-:. It is designed to limit the concentration of post-accident activity in the control room air to levels consistent with requirements of GDC 1 9,-1-4-41 999 and 10 CFR 50.67.The control room post-accident administration is described in the DCPP Manual. It is to limit post-accident control room personnel exposures to levels consistent with requirements of GDC 19,1-97-1-1999 and 10 CER 50.67.Exposures to control room personnel during post-LOCA occupancy have been estimated for a design basis LOCA to evaluate the adequacy of the control room shielding, the adequacy of the control room ventilation system, and the adequacy of the control room administration in limiting exposures to the specified ........ ,fes-ha... of exposure. to control room personnel. Radiation exposures to personnel in the control room could result from the following sources: (1) Airborne activity, which infiltrates into the control room{2-2 Direct gamma radiation the,, con ,.,.trol room,. .. from activity in, cntafinm~ferlntstrucur (4-)(2) Direct. gamma. rdatoni,, to, the control room fro activ,,ity, in, the contanmen ....plume .. from the external cloud and contained sources.The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. Provided below are the critical LOCA-specific assumptions associated with control room response and activity transport. 15.5-82 15.5-82Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Timinq for Initiation of CRVS Mode 4: i. An SIS will be generated at t =6 sec following a LOCA.ii. The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The control room dampers are fully closed 10 secs later, or at t=44.2 secs (i.e., 6 + 28.2 + 10). The 2 second SIS processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay.iii. IJn accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=18 secs (i.e., 6 + 2 secs signal processing time + 10 sec damper closure time).Control Room Atmospheric Dispersion Factors: The bounding atmospheric dispersion factors applicable to the radioac~tivity release points I control room receptors applicable to a LOCA at either unit are provided in Table 15.5-23B. The 7jQ values presented in Table 15.5-23B take into consideration the various release points-receptors applicable to the LOCA to identify the bounding 7J values applicable to a LOCA at either unit, and reflect the allowable adjustments I reductions in the values as discussed in Section 2.3.5.2.2 and summarized in the notes of Table 2.3-1 47 and Table 2.3-148 for Unit I and Unit 2, respectively. Direct Shine from External and Contained Sources The direct shine dose to an operator in the control room due to contained or external sources resulting from a postulated LOCA is calculated using point kernel shielding computer program SW-QADCGGP. The post-LOCA gamma energy release rates (MeVlsec) and integrated gamma energy release (MeV-hrlsec) in the various external sources are developed using computer program PERC2.The LOCA sources that could potentially impact the control room operator dose due to direct shine are identified below.1. Direct shine from containment -shine from the airborne source in the containment structure via the bulk shielding (3'-8" thick concrete walls below the bend line, 2'-6" thick concrete dome), including shine through one of the main steam line penetrations and the Personnel Hatch facing the control room.2. Direct shine from the contaminated cloud outside the control room pressure boundary resulting from containment leakage, ESF system leakage, RWST back-leakage, MEDT leakage -shine occurs through the control room walls, via wall penetrations such as control room doors to the outside, and from the airborne activity in cable spreading room below via control room floor penetrations. 15.5-83 15.5-83Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE 3. Dose due to scattered gamma radiation through wall penetrations from the CRVS filters located in the adjacent mechanical equipment room.4. Direct shine from the sump fluid that is postulated to collect in the RWST.Cloud shine through control room doorways was found to be the most significant of all the identified contained or external post-LOCA radiation sources listed above, followed by the dose contribution through the control room floor penetrations. Note that other radiation sources were identified and deemed insignificant due to the presence of significant shielding between the operator and the radiation sources. Examples of these dose contributors include most of the large and small electrical and pipe penetrations in the Containment outer wall that faces the control room, and the ESF system piping and components located in the Auxiliary Building.The direct shine dose estimate in the control room takes into consideration the function of Room 506 (which serves as a control room foyer adjacent to the Shift Supervisor's office), where occupancy is deemed to be minimal; i.e., conservatively estimated at less than 5% Of the total time spent daily in the control room. The above"occupancy adjustment" is utilized to determine the maximum 30-day integrated dose in control room (i.e., the total direct shine dose in the control room includes the 30-day dose in Room 506 adjusted by the referenced occupancy factor).The] coto oo etaIon I ste is deigne mlinimizeifitato1o osccdn ain-Jrbon act~ivity, into cntrol roomi,-÷~ complex Mod{,o I n opratinofr the,',r, wl bysempoiesn oeioainwt itrdpstv rsuiainadflee The cntanment.. lekg I was as.... umed. be released u ,nfiltered from the containment. bidntoteamshr. Reiclto lo ekgs asue oblrma H pUm sel ilps hog hrolflesan erlae oteamshr hog via the pressurization air intakes through charcee! filters 15.5-84 15.5-84Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE via n in leakage recirculat. d control,-,,.. room air thouh t.. he charcoal filters is 210 cn .m,. Pre.iou an..... ...assumed il n the an.lysis,; due to the poss;ibl t...hrough the s-ing-le dor f.. ro"m t-he.The control room shi..lding isdsge omnmz direct...amm..radiation. (containment... shie).Conro room+ exposures...... resultng from containment.. shin were... estimate using ISOSHLO n!. The .ontro room, rec.ptor poi"nt 27 from" the containment. structure... to the coto om otoom.epsue rsingfo puesin eeesiae source. abov the. control,., room. The control room recepto point. is+;protected by p1 .foot thic coceeI hed Control Room Operator Dose duringi Access Diablo Canyon assumes that the dose received by the operator during routine access to the control room for the 30 +day period following the LOCA is minimal. Thus, as long as some reasonable margin exists between the regulatory limit and the estimated dose to the operator during control room occupancy, the additional dose due to ingress I egress can be accommodated. This approach is consistent with the approach used. by other licensees, and is reasonable since a) transit to and from the control room is only expected after the first 24 hours following the accident by which time the airborne levels inside .containment has reduced significantly due to the use of active fission product removal mechanisms such as containment sprays, and radioactive decay, and b) the operator is protected from radioactive ESF fluids by the shielding provided by the buildings that house such equipment. In addition, it is expected that during a postulated event, access to the control room will be controlled by Health Physics and the Emergency Plan based on real time data, with the purpose of minimizing personnel dose.15.5-85 15.5-85Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE It is also noted that the dose received by the operator during transit outside the control room is not a measure of the "habitability" of the control room which is defined by the radiation protection provided to the operator by the control room shielding and ventilation system design. Thus, the estimated dose to the operator during routine post-LOCA access to the control room is addressed separately from the control room occupancy dose which is used for the demonstration of control room habitability. In accordance with DCPP original licensing basis, radiation exposures to personnel during egress and ingress (i.e., during routine access to the control room for the duration of the accident) could result from the following sources: (1) Airborne activity in the containment leakage plume aand (2) Direct gamma radiation from fission products in the containment structure. Post-accident egress-ingress exposures are-were based on 27 outbound excursions, from the control room to the site boundary, and 26 inbound excursions, from the site boundary to the control room. It was estimated that each excursion would take 5 minutes, and no credit was taken for breathing apparatus or special whole body shielding. Egress-ingress thyroid and whole body exposures from airborne activity are functions of containment activity, containment leakage, atmospheric dispersion, and excursion time.The EMERALD computer code was used to calculate the airborne activity concentrations, and then conventional exposure equations from Regulatory Guide 1 .4, Revision 1, were used to calculate gamma, beta, and thyroid exposures (Reference 6).The exposure from betas is-was calculated on the basis of an infinite uniform cloud, and exposure from gammas is-was calculated on the basis of a semi-infinite cloud.Because of the containment shielding and short excursion time, egress-ingress containment shine exposures are-were estimated to be small. Egress-ingress containment shine exposures were calculated using ISOSHLD-II. The shine model assumes-assumed a cylindrical radiation source having the same radius and height as the containment structure with a 3.5-foot-thick concrete shield surrounding it. The receptor point is-was assumed to be a distance of 10 meters from the outer surface of the containment wall.The eE-stimatesd ofv po.st"" accden ,.c,,o..,,tro÷l.,,.""÷ egress-ingress exposures developed in support of DCPP original licensing basis are listed in Table 15.5-33 and summarized below. sum of the. DBA, case exposures ... re... of t'he D'IA case expvosures.
- a. The dose to control room personnel during egress ingress from airborne fission products in the containment leakage plume: 0.0066 rem gamma, 0.0243 rem beta, and 4.72 rem thyroid 15.5-86 15.5-86Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE b. The dose to control room personnel during egress ingress as a result of direct radiation shine from the fission products~in the containment structure is 0.022 rem.Subsequent to the original licensing basis assessment described above, DCPP has identified additional post-LOCA fission product release pathways, as discussed in Section 15.5.17.2.1.
The postulated effect of these additional radioactivity release paths, as well as the implementation of AST, on the estimated dose to control room personnel during routine egress ingress takes into consideration the following:
- a. The transport models used to develop the dose to the control room operator during occupancy address a control room occupancy factor of 1 .0 till t=24 hours after the accident.
This implies that during the first 24 hours the control room operator stays in the control room. This is also reflected in the DCPP original licensing basis which addresses one more outbound trip than the inbound trips.b. Routine ingress / egress to the control room during the 30 day period following a LOCA falls into the mission dose category as discussed in NUREG 0737, November 1980, Item ll.B.2.c. In accordance with NUREG 0737, November 1980, Item ll.B.2 leakage of systems outside containment need not be considered as potential sources.Based on the above considerations, the dose consequences of the additional activity release paths addressed in Section 15.5.17.2.1 (and listed below), in addition to Regulatory Guide 1.183 is addressed as follows: i.Containment Pressure/iVacuum relief release -this release occurs at accident initiation (before t=24hr), so there is no dose contribution to the control room operator during routine ingress/Iegress during the 30 day period following the accident.ii. Containment leakage: a. The airborne activity in the containment after t=24 hours with an AST source term is primarily 100% of the core noble gases and 0.06% of the core iodines that were released to containment. Note: The iodine source term at t=24 hrs is essentially the organic iodines released to the containment which are not eaffected by sprays, and which per Regulatory Guide 1.183, represent 0.06% of the core iodines (i.e., 0.15% of the 40% core iodines released to containment atmosphere at accident initiation). Also, the essentially particulate nature of the radioactivity release associated with an AST source term, and the effectiveness of particulate removal by sprays /settling makes the dose contribution from the particulate source minimal after t=24 hours.15.5-87 15.5-87Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE b. The corresponding airborne activity in the containment after t=24 hours for a TID-14844 source term is 100% of the core noble gases and 1% of the core iodines.Note: Per Regulatory Guide 1.4, Revision 1, the organic iodines released to the containment is 4% of the 25% iodines released to containment atmosphere at accident initiation.
- c. Based on the above it is concluded that after t=24hrs: oDose consequences due to containment leakage based on a TID-14844 based scenario will bound the dose consequences based on an AST scenario.oThyroid dose is primarily due to iodines, the associated dose to the operator will vary proportionately to the amount of iodine airborne in containment.
Thus the thyroid dose to the operator during ingress/eg ress for an AST scenario may be estimated by adjusting the TID-14844 based dose by the ratio of the iodine estimated to be airborne in containment for each of the scenarios. As noted earlier, the current licensing basis thyroid dose to the operator during ingress /egress is 4.72 remn. The corresponding thyroid dose based on an AST scenario is estimated to be 4.72 x 0.06 =0.28 rem thyroid.iii. The RHR Pump Seal Failure, ESF System Leakage, RWST back leakage and MEDT leakage -All of these releases are based on leakage of systems outside containment. In accordance with NUREG 0737, November 1980, Item ll.B.2, the dose contribution due to these sources need not be considered for access calculations. To address the TEDE dose acceptance criteria applicable to use of AST, the original licensing basis egress-ingress exposures have been updated as noted below in accordance with 10 CFR 20.1003.10 CFR 20.1003 defines TEDE as the sum of the deep dose equivalent for external exposures (i.e., external whole body exposure) and the committed effective dose equivalent for internal exposures (i.e., sum of the product of the weighting factor applicable to each organ irradiated and the dose to that organ). Per 10 CFR 20.1003, the weighting factor for the whole body is 1 .0 and for the thyroid is 0.03. While the weighting factor for beta radiation is undefined, the contribution of the beta dose to the total effective dose equivalent is expected to be insignificant. Therefore, a. Radiation from airborne fission products in the containment leakage plume to the control room personnel during egress ingress is approximately 0.0066 rem + 0.28 x 0.03 rem, i.e., 0.015 rem TEDE 15.5-88 15.5-88Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE b. Direct radiation from the fission products in the containment structure to control room personnel during egress ingress is 0.022 rem TEDE.Thus the total dose to the control room operator during access is estimated to be 0.037 remn TEDE. This value is 1% of the estimated operator dose due to control room occupancy following a LOCA (Refer to Table 15.5-23) and is therefore considered to be minimal.15.5.17.2.5 Post-LOCA Technical Support Center Operator Exposure In accordance with NUREG-0737, Supplement 1, January 1983, Section 8.2.1 (f) the TISO design has been evaluated for the LOCA.Computer code PERC2 is used to calculate the dose to TSC personnel due to airborne radioactivity releases following a LOCA. The direct shine dose to an operator in the TSC due to contained or external sources resulting from a postulated LOCA is -calculated using point kernel shielding computer program SW-QADCGGP. The post-LOCA gamma energy release rates (MeV/sec) and integrated gamma energy release (MeV-hr/sec) in the various external sources are developed with computer program PERC2.The TSC serves both units and is located at El 104' on the south-west side of~the Unit 2 turbine building and is shared between Unit 1 and Unit 2.The nominal TSC air intake flowrate during normal operations is 500 cfm. The air inflow is filtered through a HEPA filter and drawn into the TSC envelope which has a free volume. The TSC normal intake is isolated and the TSC ventilation placed into filtered /pressurized (CRVS Mode 4) operation by manual operator action within 2 hours of the LOCA.The post-accident pressurization flow to the TSC is provided via the CRVS Mode 4 pressurization intakes (i.e., 1 per unit, each located on either side of the Turbine Building). As noted in Section 15.5.9, the control room pressurization air intakes have dual ventilation outside air intake design. The nominal air intake flowrate during the TSC pressurization mode is 500 cfm.As discussed in Section 15.5.9, CRVS Mode 4 operation utilizes redundant PG&E Design Class I radiation monitors located at each pressurization air intake and has the provisions of acceptable control logic to automatically select the least contaminated inlet at the beginning of the accident, and manually select the least contaminated inlet during the course of the accident. Thus, during Mode 4 operation the TSC dose consequence analysis can utilize the x./Q values for the more favorable pressurization air intake reduced by a factor of 4 to credit the "dual intake" design (refer to Section 2.3.5.2.2 for additional details).The allowable methyl iodide penetration and filter bypass for the TSC Mode 4 Charcoal 15.5-89 15.5-89Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Filter is <2.5% and <1%, respectively. Thus in accordance with Generic Letter 99-02, June 1999, the TSC charcoal filter efficiency for elemental and organic iodine used in the TSC dose analysis is 93%. The acceptance criteria for the TSC normal operation and Mode 4 HEPA filters is "penetration plus system bypass" < 1 .0%. Thus, using methodology similar to the charcoal filters, the HEPA filter efficiency for particulates used in the TSC dose analysis is 98%.During TSC Mode 4 operation, the TSC air is also recirculated through the same filtration unit as the pressurization flow (refer to Section 9.4.11). The air flow allowable through the pressurization charcoal / HEPA filter and minimum filtered recirculation flow for the TSC is provided in Table 15.5-82.Unfiltered inleakage into the TSC during normal operation and Mode 4 is assumed to be 60 cfm (includes 10 cfm for ingress/egress based on the guidance provided in NUREG 0800, SRP 6.4.For purposes of estimating the post-LOCA dose consequences, the TSC is modeled as a single region. When in TSC Mode 4, the Mode 1 intakes are isolated and outside air is a) drawn into the TSC through the filtered emergency intakes; b) enters the TSC as infiltration, and c) enters the TSC during operator egress/ingress. The dose assessment model utilizes nominal values for the ventilation intake flowrates since the intake pathways (normal as well as accident) are filtered, thus the controlling dose Contributor is the unfiltered inleakage. The effect of intake flow uncertainty on the TSC dose is expected to be insignificant. The bounding atmospheric dispersion factors applicable to the radioactivity release points / TSC receptors applicable to a LOCA at either unit are provided in Table 15.5-23E. The %/Q values presented take into consideration the various release points-receptors applicable to the LOCA to identify the bounding z/Q values applicable to a LOCA at either unit, and reflect the allowable adjustments / reductions in the values as discussed in Section 2.3.5.2.2. The direct shine dose into the TSC due to the external cloud and contained sources is calculated in a manner similar to that described for the control room in Section 15.5.17.2.4. The LOCA sources that could potentially impact the TSC operator dose due to direct shine are identified below.1. Direct shine from containment -shine from the airborne source in the containment structure via the bulk shielding (3'-8" thick concrete walls below the bendline, 2'-6" thick concrete dome), including shine through the Personnel Hatch facing the TSC 2. Direct shine from the contaminated cloud outside the TSC pressure boundary resulting from containment leakage, ESF system leakage, RWST back-leakage, 15.5-90 15.5-90Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE MEDT leakage -shine occurs through the TSC walls and via wall penetrations such as TSC doors to the outside.3. Dose due to scattered gamma radiation through wall penetrations from the TSC filters located in the adjacent mechanical equipment room and scatter past labyrinths provided for selected doors.Note that other radiation sources were identified and deemed insignificant due to the presence of significant shielding between the operator in the TSC and the radiation sources.Table 15.5-82 lists key assumptions / parameters associated with DCPP TSC design.The bounding TSC operator dose following a LOCA at either unit is presented in Table 15.5-23.15.5.17.2.14-S Summary In the preceding sections, the potential exposures from a major primary system pipe rupture have been calculated for various possible mechanisms: (1) Containment Pressure!/ Vacuum Relief (2) Containment leakage (-!-)(3) ESF System Leakage (-2-)(4) RHR ,eejieultipump seal Failureloplekg (5) acc-qident cont-,inment ... ÷ .. ng"R\S-T Back-Leakage {-3)(6) MEDT Leakage (-4)(7) Shine from Contained and External Sources (e.g., Contained Containment shine, RWST Shine, external clouds due to the various leakage sources, etc)The analyses have been carried out using the models and assumptions specified in regulations 10 CFR Part 100, in Regulatory Guide 1.18310 CFR Par 50, and the-other regulatory guidance identified. ,,nd regultor; .....desab--ove.'" In all analyses, the resulting potential exposures to plant personnel, to individual members of the public, and to the general population have been found to be lower than the applicable guidelines and limits specified in 10 CFR Part--10050.67 and Regulatory Guide 1.183,-10 CFR Part 50, and 10 CFR Part 20.15.5-91 15.5-91Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 15.5.17.3 Conclusions Based on the results discussed, the occurrence of a major pipe rupture in the primary system of a DCPP unit would not constitute an undue risk to the health and safety of the public. In addition, the ESF provided for the mitigation of the consequences of a LBLOCA are adequately designed.Additionally, the analysis demonstrates that the acceptance criteria are met as follows: (1) The radiation dose to an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.25 Sv (25 rem) TEDE as shown in Table 15.5-23.(2) The radiation dose to an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), is within 0.25 Sv (25 rem) TEDE as shown in Table 15.5-23.(3) The radiation dose to an individual in the control room for the duration of the accident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-23.The dose received by the operator during transit outside the control room is not a measure of the "habitability" of the control room which is defined by the radiation protection provided to the operator by the control room shielding and ventilation system design. Thus, and in accordance with DCPP current licensing basis, the dose contribution to the operator during routine access to control room for the duration of the accident (0.04 rem TEDE), is not included with the control room occupancy dose for the demonstration of control room habitability The radiation dose to an individual in the TSC for the duration of the accident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-23.Fina'lty, anahlysi deimonstrates thatf the acceptancer, critria are m et as follows:, (1) The con..equence. f major. rutr of.. prima' coolant pipes.shall atI T--_1 hrs po.t L-OICA),\ and, containmen..t. shine as how i...n Section !15.5.1 7.2.11!.15.5-92 15.5-92Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE iJ. An ind,,vidual located,, at any point on the bouda. of the.. oxso re o h (reee sha~ll not +totah hfraditon dose tiorte wol body llin eces or~em orl tota fradato dosnreff fineces o 30 nrem to th thri "rmidn expsur as how by... t EAn; wole bod dose for. cotinmTent.52 shie.iSetio 5DI5L17.L2O6,EQUNdE th F reAnn dAORe prSTEnte PinE RUTab RE 15.5.81 An ccnipidalncaed arteanrpintoah ue ona ftelwpplto fh aisoionia prodsucrelease (durSingl nt xce the entieeprioiof its pafage,0sallno rece ad ivl mee tota raiaioose acetoathe hoteri bodfi Regulastof, 25rem118, oruayt2ta shownn by thse CriterhalboydsreoedfrcnanethienScin 1551)7n2d.6ua (constervativ ahny appliedt t theboundayand the exlsoremainn dores (3)Inacourdanero fowith the reurmnsts of thGpstlae 1f197o, theroscto theae ontrll room opeiearatruderaccdn odtiondoeiexss sf hel not beR in excess of 0.5 rem whol Rfrenc ED51 for th duraxitiongo h accident aoiespk cshow iTable 15.5 33.h1 CFoR hydroge conitrolot the hycdrognt rntae odbiners),k cansnieua. oaeda n pointon th bouda5-gf3theexclsion ara1h9i xoedth radioactiv DCPP UNITS 1 & 2 FSAR UPDATE (2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of the 10 CFR 50.67 limit of 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case..Control Room Dose Criteria{-1-)-Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.The c.....q...... of a ma"jor tea pipc ruptur.. s ...hall not (2)-(3 .....n in isiu located a4toan poin the bounda. of;' them exluio area for the rlsehalntrcieattlrdation dose n excess!0 theR1010I ....R..100..11.... dose4 limit f, r th.he whole4 .. bod the , thyroi for the .. pro extn idne spikeofth caseCan 1 0 peren of"';;" " the 0F 100.11 do.e lit for` the hoe body a4.ndLh thyroid.2 Idnfor athe n a fCue nccident intaeDoiesi e se. pto (55.18An.individua Rlocated Pathan yspito h ue onar ftelpplto As onephoisexoed tScin 542 th radioacsteaine ludturesuling from thecpodtulatsed fisddong prductgereleasthu(duringlthe entireo periodut tof ithe passage) shl nexpecteid aolwn totsalcdn.I infcn radiationioseti excess of the 1 R10.1dsecolimitsysefpror thew ol bo acdyendthoevthroid for this praeitingt wiodine spikaed caste andv1ipronmnt oft the 15.5-94 15.5-94Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE steam escaping from the pipe rupture. In addition, if an atmospheric steam dump from the unaffected steam generators is necessitated by unavailability of condenser capacity, additional activity will be released......, Section.. 1,.5.1.2.1. discusses 4;.,.... .t.... m, i ,,, * line..,, bekr,,, (/SL)l dos ,-analysi,.s~c; o.-f record, wkrhichi on, the, The tSG MSL This event consists of a double-ended break of one main steam line. The analysis focusses on a MSLB outside the containment since a MSLB inside containment will clearly result in a lesser dose to a control room operator or to the offsite public due to hold-up of activity in the containment. Following a MSLB, the affected SG rapidly depressurizes and releases the initial contents to the environment via the break. Based on an assumption of a Loss of Offsite Power coincident with reactor trip, the condenser is assumed to be unavailable, and environmental steam releases via the MSSVs / 10% ADVs of the intact steam generators are used to cool down the reactor until initiation of shutdown cooling. The activity in the RCS leaks into the faulted and intact steam generators via SG tube leakage and is released to the environment from the break point, and from the MSSVs/ 10% ADVs, respectively. Regulatory requirements provided for the MSLB in pertinent sections of Regulatory] Guide 1.183 including Appendix E is used to develop the dose consequence model.Table 15.5-34A lists the key assumptions / parameters utilized to develop the radiological consequences following a MSLB.pomptentr fore RnADRADy, sue to seodrc'la agenegaded tubingntro inreaso t an rate thunaty rdiologicalt cironsqune ranialytisit freasemajorstomwine rutueMnSLuin.a 15.5.18.2.2 Activity Release Transport Model 15.5-95 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE In accordance with Regulatory Guide 1.183, Appendix E, item 2, since no melt or clad breach is postulated for the DCPP MSLB event, the activity released is based on the maximum coolant activity allowed by the plant technical specifications. The plant technical specifications focus on the noble gases and iodines. In addition, and per Regulatory Guide 1.183, two scenarios are addressed, i.e., a) a pre-accident iodine spike and b) an accident-initiated iodine spike.a. Pre-accident Iodine Spike -the initial primary coolant iodine activity is assumed to be 60 jliCi/gm of DE 1-131 which is the transient Technical Specification limit for full power operation. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.b. Accident-Initiated Iodine Spike -the initial primary coolant iodine activity is assumed to be at Technical Specification of 1 pCi/gm DE 1-131 (equilibrium Technical Specification limit for full power operation). Immediately following the accident the iodine appearance rate from the fuel to the primary coolant is assumed to increase to 500 times the equilibrium appearance rate corresponding to the 1 1 iCi/gm DE 1-131 coolant concentration. The duration of the assumed spike is 8 hours. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.The initial secondary coolant iodine activity is assumed to be at the Technical Specification limit of 0.1 pCi/gm DE 1-131.Technical Specifications limit primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the MSLB dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).Following a MSLB, the primary and secondary reactor coolant activity is released to the environment via two pathways.Faulted Steam Generator The release from the faulted SG occurs via the postulated break point of the main-steam line. The faulted SG is estimated to dry-out almost instantaneously following the MSLB (within 10 seconds), releasing all of the iodine in the secondary coolant (at Technical Specification concentrations) that was initially contained in the steam generator. The EAB and LPZ dose to the public is calculated using an instantaneous release of the iodine inventory (C i) in the SG liquid in the faulted SG. The secondary steam activity initially contained in the faulted steam generator is also released;however, the associated dose contribution is not included in this analysis since it is considered insignificant. To maximize the control room and offsite doses following a MSLB, the maximum allowable primary to secondary SG tube leakage for all SGs (0.75 gpm or 1080 gpd at 15.5-96 15.5-96Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE Standard Temperature and Pressure (STP) conditions), is conservatively assumed to occur in the faulted 3G. All iodine and noble gas activities in the referenced tube leakage are released directly to the environment without hold-up or decontamination. The primary to secondary SG tube leakage is assumed to go on until the RCS reaches 2120° F, which based on minimum heat transfer rates, is conservatively estimated to occur 30 hours after the event.Intact Steam Generators The initial iodine activities in the secondary coolant at Technical Specification levels are released to the environment in proportion to the steaming rate and the inverse of the partition coefficient (limited to 100) defined in Regulatory Guide 1.183. The noble gases are released freely to the environment without retention in the steam generators. However, there is no primary to secondary leakage into the intact SG as all primary to secondary leakage (1080 gpd or 0.75 gpm) is assumed to be occurring in the faulted SG.The iodine releases to the environment from the SG are assumed to be 97%elemental and 3% organic. The condenser is assumed unavailable due to the loss of offsite power. The SG releases continuefor 10.73 hours, at which time shutdown cooling is initiated via operation of the RHR system and environmental releases are terminated. 15.5.18.2.3 Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose.a. The Source/Release for the Pre-incident Spike Case is at its maximum levels between 0 and 2 hours.b. The Source/Release for the Accident-Initiated Spike Case is at its maximum levels towards the end of the spiking period.Regardless of the starting point of the "Worst 2-hr Window," the 0-2 hrs %/ is utilized.The bounding EAB and LPZ dose following a MSLB at either unit for both scenarios are presented in Table 15.5-34.15.5.18.2.4 Control Room Dose Assessment The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. Provided below are the critical MSLB-specific assumptions associated with control room response and activity transport. Timing for Initiation of CRVS Mode 4: 15.5-97 15.5-97Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE i. An SIS will be generated at t = 0.6 sec following a MSLB.ii. The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The control room dampers are fully closed within 10 seconds at t=38.8 secs (i.e., 0.6 + 28.2 + 10). The 2 second SIS processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay.iii. In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=12.6 secs (i.e., 0.6 + 2 secs signal processing time + 10 sec damper closure time).Transport of Radioactivity from the Break Location Since the normal operation (CRVS Mode 1) control room intake of the faulted unit is in such close proximity to the break point, an atmospheric dispersion factor (x/Q) cannot be accurately determined. Thus, atmospheric dispersion is not credited when determining the control room operator dose from the secondary coolant discharge or the primary to secondary SG tube leakage released from the faulted SG via the break point.Secondary Coolant Discharge: The radioactivity release due to the almost immediate dry-out of the faulted SG following a MSLB is based on a) the radioactivity concentration of the iodine in a finite cloud created by the secondary coolant liquid flash at the break point; b) conservation of total iodine activity in the SG liquid. The activity concentration at the release point is conservatively based on saturated steam at a density of 5.98E-04 gm/cm 3 , (i.e., at 1 atmosphere and 212°F). The activity concentration entering the control room is assumed to be the same as the concentration at the break point until the Control room normal ventilation is isolated and the CRVS re-aligned to Mode 4 Pressurization. Primary to Secondary SC Tube Leakage: Due to the close proximity of the normal operation control room intake of the faulted unit and MSL break release point and consequent unavailability of viable atmospheric dispersion factors, the primary to secondary SG tube leakage into the faulted SG is conservatively assumed to be piped directly into the control room. This model is reasonable since the relatively small plume of steam created by the ~0.485 gallon {i.e., (0. 75 gallon/mn) (38.8 s) /60 s/ran}n of reactor coolant released due to SG tube leakage via the MSL break point could easily be swept into the control room due to the close proximity of the control room normal intake to the break point.Control Room Atmospheric Dispersion Factors 15.5-98 15.5-98Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE As noted in Section 5.0, because of the proximity of the MSSVs/1 0% ADVs to the control room normal intake of the affected unit, and because the releases from the MSSVs/10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation control room intake of the affected unit (closest to the release point) will be insignificant. Therefore, prior to switchover to CRVS Mode 4 pressurization, only the unaffected unit's control room normal intake is assumed to be contaminated by releases from the MSSVs/10% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points / control room receptors applicable to a MSLB at either unit are provided in Table 15.5-34B. The x/Q values presented in Table 15.5-34B take into consideration the various release points-receptors applicable to the MSLB to identify the bounding x/ values applicable to a MSLB at either unit, and reflect the allowable adjustments I reductions in the values as discussed in Chapter 2.3.5.2.2 and summarized in the notes of Tables 2.3-147 and 2.3-148.The bounding control room dose following a MSLB at either unit is presented in Table 15.5-34. met*hodology selecte for-,, p ... .-.-ing.- the.. r.-adiologica.;l ...e.sment follows... NRC SRP 15.1.5, ,,team-. D;iping. F;ilue ..... c Containment.. limits and control room doses are within GDC 19, 1971 limits.The resultant doses from,, the MSL, evn us,,,.,ng,; an accdet inducedi l,,-eakL rate o~f 15.5.18.3-Conclusions The analysis demonstrates that the acceptance criteria are met as follows: (1) The radiation dose to an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case as shown in Table 15.5-34.(2) The radiation dose to an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), is within 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike ease and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case as shown in Table 15.5-34.15.5-99 15.5-99Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE (3) The radiation dose to an individual in the control room for the duration of the accident is within 0.05 Sv (5 remn) TEDE as shown in Table 15.5-34.rele..se.shall not receive a total radiatio.n d4,ose in exce... ofhe !10 CF 100.]11 cas ad 0 eren o te 0 FR10.1 dos !i0it fo!h!hlebd n the tlhyroi fo th, en ÷acchident; initiate iodin spike care asshow e-ina 1 (2) An indvidual locatedr at-.;any point oin th outerl bondah of th-e low popu ree1eatoa raditio doe inuexcess of÷he 10. CF. 100.11~ dos ...thed whole. body. and the thyroidfor th.. pr existing. iodine. spik case;,,r and 10v (3), In wlith4 the requiementsl{ GDC19 197lr,the dosef th control skerainga Roneferences51 for th duationutr of th acidnt forebothther prpe (exisedtin heenda th feccident linitiatedk iodine) spike ntcases astheosn inmecitson 15051CFR1 As7 anotd winl Setio 1th58d2s1 thceptabove dosteri oRestimatesrefec thie 1.183 andy 20n whnc orriesodsbe tow teds.ii f3 rmfra cietintae oiesie 15E.1B 1 -Acceptancs e Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product 15.5-100 15.5-100Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE release shall not receive a radiation dose in excess of the 10 CER 50.67 limit of 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case.(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of the 10 CFR 50.67 limit of 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case.100.11 as out- hour1.2 imm diaentifo.incathen on aset ofd teptuAteiden fisrionprouti rleonalo Asreep ortalrd iaetion dose.2 taor the oewbodr inees ofu25uremi nor axee tota radiaio doseding excaes ofa00c thuo thcethyroid frssom produnct exosure.coatsepce islexposedttosthecradioactiveigloudcresultingafromitheepostulathed fssiondproducste prelease (duin theidentr poeeri ode of iths asage),it whall not receiease total radiatonmdoen wto the thyrodwafro iodaine exposurepierpueInadtofantmshictamup Asreorthed inaSection 15.4.2 geeatmaors fseedwiater lin rpurisnotaiexpected t caudeser clpctaddingida age andthust nol b released of fissio' prdcs ote ,oln is.. foloin Sthisdaccd Rvent Ifsin5.cantradioacivt existsm in thesecondar86 y sythemev praior tof the raccdenta howsevuersoes of theig ais aciiyWiLBa be relased tthenionmentuawithathve fromprsnt the unffcedsutea gftenderaosisgnecessitatedLBy. nviaiiyo odne The rdiologca consequences foflwn abu WL wil0 Ibm ofndecodary cooSLasnte reles airborne environmental release via the break point is expected to be less than the MSLB.15.5-101 Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE As demonstrated in Table 15.5-34, the dose consequences at the EAB and LPZ following a MSLB is within the acceptance criteria applicable to the FWLB.15.5.19.3 Conclusions On the basis of this comparison approach, it is concluded that the dose consequences at the EAB and LPZ following a feedwater line break will remain within the acceptance criteria listed in Section 15.5.19.1.Bascd on the re.ult. disussed....,4 it ca_.n be concluded,,leeI ..p.cified. 1,0 rCR 100.11, and, that÷ the occurrence.. of ..uch ruptures.. ..ould" not result in undue risk.. tor- the public.to- the whole' body and, to the thyroid;, of an indi,,idual located+., at an..poin onthebounaryof he ecluionareaforthe~o hursimmdiatly ollwin the Teradioatelouca renseuetnge fom th poTRstuated fsinoproduct rheease (duingiths entiCRe perid ofutslpassage)l aeininocatw.soni 155.0B Acceptanco e Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release shall not receive a radiation dose in excess of the 10 CFR 50.67 limit of 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 100/ of the 10 CFR 50.67 limit for the accident initiated iodine spike case.(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not 15.5-102 15.5-102Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE receive a total radiation dose in excess of the 10 CFR 50.67 limit of 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case..Control Room Dose Criteria Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.belew the two hor immediate,, following,., the, onset, of the postulated, fisson;.CFR 00.1 doe lmitsforthe hol-bod an thme tyodfor, pre , existin iodi spike. case, an 10 prccnt !of the 1 ......R. 100.11d lmt fo he whole bod andtehyodfr the initiated odiespk (2. An indviua located ath~ any.. pint.. bondryo t91+,4 o postulat-Iedtifisstion podCuctreeasea(durcidngth Dentireiperiodn fispasg)shall02. noivty Release Pa ttlrdainds necsfthew0ays001 ds This eve mitis fore b the iholebodyando ruthre tyoidfo th ptue existin iodie sptnkeles ofrimrycoase, ando 10e poern ofesthe1 scFndry 10.1dsysem limilts for thedwholecbody postulanedfr the SThRoi fornt.he caccidentionitatsued ioin spick-oenase.th conurd trol rgoeoerator uo0mndter .Baccient condi ssmtionshl not be inLxcss of 5fst Poeoncdbenta skin, rerence 51t frthiurtono the accise sssm dt en forailbotlte, pre Thisroevent i austedm byese i the intataeo s rupture D ofa tube winthac rsuteantrles 15.5-103 15.5-103Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE generators are used to cool down the reactor until initiation of shutdown cooling. A portion of the primary coolant break flow in the ruptured SG flashes and is released a)to the condenser before reactor trip and b) directly to the environment after reactor trip, via the MSSVs and 10% ADVs. The remaining break flow mixes with the secondary side liquid, and is released to the environment via steam releases through MSSVs and 10% ADVs. The activity in the RCS also leaks into the intact steam generators via SG tube leakage and is released to the environment from the MSSVs / 10% ADVs.Regulatory requirements provided for the SGTR in pertinent sections of Regulatory Guide 1.183 including Appendix F is used to develop the dose consequence model.Table 15.5-64A lists the key assumptions / parameters utilized to develop the radiological consequences following a SGTR. Table 15.5-64C provides the time dependent steam flow from the Ruptured and Intact SGs and the flashed and unflashed break flow in the Ruptured SG.Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a SGTR.15.5.20.2.2 Activity Release Transport Model No melt or clad breach is postulated for the SGTR. Thus, and in accordance with Regulatory Guide 1.183, Appendix F, item 2, the activity released is based on the maximum coolant activity allowed by the plant technical specifications. The plant technical specifications focus on the noble gases and iodines. In addition, and per Regulatory Guide 1.1 83, two scenarios are addressed, i.e., a) a pre-accident iodine spike and b) an accident-initiated iodine spike.a. Pre-accident Iodine Spike -the initial primary coolant iodine activity is assumed to be 60 FtCi/gm of DE 1-131 which is the transient Technical Specification limit for full power operation. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.b. Accident-Initiated Iodine Spike -the initial primary coolant iodine activity is assumed to be at Technical Specification of 1 DE 1-131 (equilibrium Technical Specification limit for full power operation). Immediately following the accident the iodine appearance rate from the fuel to the primary coolant is assumed to increase to 335 times the equilibrium appearance rate corresponding to the 1 pCi/gm DE 1-131 coolant concentration. The duration of the assumed spike is 8 hours. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.The initial secondary coolant iodine activity is assumed to be at the Technical Specification limit of 0.1 1 iCi/gm DE 1-131.DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To 15.5-104 15.5104Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE accommodate any potential accident induced leakage, the SGTR dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd). To maximize the dose consequences, the analysis conservatively assumes that all of the 0.75 gpm SG tube leakage occurs in the intact SGs.Following a SGTR, the primary and secondary reactor coolant activity is released to the environment via two pathways.Ruptured Steam Generator A SGTR will result in a large amount of primary coolant being released to the ruptured steam generator via the break location with a significant portion of it flashed to the steam space.In accordance with the requirements provided in Regulatory Guide 1.183, the noble gases in the entire break flow and the iodine in the flashed portion of the break flow are assumed to be immediately available for release from the steam generator. The iodine in the non-flashed portion of the break flow mixes uniformly with the steam generator liquid mass and is released into the steam space in proportion to the steaming rate and the inverse of the allowable partition coefficient of 100. The iodine releases from the SGs are assumed to be 97% elemental and 3% organic.Before the reactor trip the radioactivity in the steam is released to the environment from the air ejector which discharges into the plant vent. All noble gases and organic iodines in the steam are released directly to the environment. Only a portion of the elemental iodine carried with the steam is partitioned to the air ejector and released to the environment. The rest is partitioned to the condensate, returns to both the intact steam generators and the ruptured steam generator and will be available for future steaming releases.After the reactor trip, the radioactivity in the steam is released to the environment from the MSSVs/10% ADVs, due to the assumption of LOOP. To isolate the ruptured steam loop, the auxiliary feed water to the ruptured SG is secured. The calculation assumes the PORV of the ruptured SG fails open for 30 minutes. The fail-open PORV is isolated at t = 2653 seconds at which time the ruptured steam loop is isolated. The break flow continues until the primary system is in equilibrium with the secondary side of the ruptured SG. The iodines in the flashed break flow and the noble gases in the entire break flow is bottled up in the steam space of the ruptured SG and released to the environment during the manual depressurization of the ruptured SG after t = 2 hours.15.5-105 15.5-105Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Intact Steam Generators The radioactivity released from the intact steam generators includes two components: (a) a portion of the break flow activity that is transferred to the intact steam generators via the condenser before reactor trip, and (b) due to SG tube leakage.Approximately 75% (3 intact SGs vs 1 ruptured SG) of the flashed break flow activity that is transported and retained in the condenser before reactor trip will be transferred to the intact steam generators and released to the environment during the cool-down phase.The total primary-to-secondary tube leak rate in the 3 intact SGs is conservatively assumed to be 0.75 gpm. The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events) has been evaluated for potential impact on dose consequences as part of a WOG Program and demonstrated to be insignificant. Thus all leaked primary coolant iodine activities are assumed to mix uniformly with the steam generator liquid and are released in proportion to the steaming rate and the inverse of the partition coefficient. Before the reactor trip, the activity in the main steam is released from the plant vent via the air ejector/ condenser. After the reactor trip, the steam is released from the MSSVs/10% ADVs. The reactor coolant noble gases that enter the intact steam generator are released directly to the environment without holdup. The iodine releases from the SGs are assumed to be 97% elemental and 3% organic. The intact SG steam release continues until shutdown cooling (SDC)is initiated at t = 10.73 hours Initial Secondary Coolant Activity Release The initial iodine activities in the secondary coolant are released to the environment in proportion to the steaming rate and the inverse of the partition coefficient from the ruptured and intact SGs. Twenty five percent of the initial secondary coolant iodine inventory is in the ruptured SG and 75% of the initial secondary coolant iodine inventory is in the 3 intact SGs 15.5.20.2.3 Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose.For the SGTR, the EAB dose is controlled by the release of the flashed break flow in the ruptured SG which stops at 3402 seconds. The break flow stops at 5872 seconds and the ruptured SG is manually depressurized 2 hours after the accident. Therefore the maximum EAB dose occurs during the 0-2hr period for both the p re-accident and accident initiated iodine spike cases.Regardless of the starting point of the "Worst 2-hr Window," the 0-2 hrs x/ is utilized.15.5-106 15.51 06Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE The bounding EAB and LPZ dose following a SGTR at either unit for both scenarios are presented in Table 15.5-64.15.5.20.2.4 Control Room Dose Assessment The acc-ident is reanalyzed and is discu,,sed in Sect÷ion 15..3 and4 tAhe thermal 11 1I andhyralicanlyisIrecned nIecio 15.. poie th bai for efrain frdo ogia co"euecehisusedintissetin hifit ExpsuIc e (ent a.su es tha-th reacto has°'m° bee oper.tin atth xmulwbl Tcnia Specificatin (Referenc 22 liisfo rmaycoln ctvtad1gp rmayt k")priorto.ad.folowin th ....T.. are determined..as.follows: (a) The, iodine concentrations' in' the reactor coolant be, based upon.conenteration-' is 1 "i'it"'g D.ose E.q,.',,alen !31.15.5-107 15.51 07Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE which inercases the iodine release rate from the fuel to the coolant to a value 335 times groater than the release rate corresponding to the initial primary system iodine concentration. The initial appearance rate can be written as fO4GWS4 (15.5 15)where;-R Eouilibrium aoocarance rate for iodine nuclide coresoninmt I ~~~mo E13 remva coefficient fo ,r-,, iodin nuclide3 the S-TR and has- raied-o the primary coolant iodine oncDEntaio rm o!03irm.fD I11 baie upn61pigo e 3 E o h obegse r8m from IKr 85 and X .a "13m due, to, low. cncent~nration andl small doseonversi~ono n fHactor. (o 15.5-108 15.5108Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE The fo`llo,- g ssupton welre used. ca-luTlatef the--(a) The mass of reacto coolan,,,to discharged int the,. seconda. through the a'-nd the mass of stFe~am'J released` flromrr the presented4 in ISA 11.(b). The .. massof.," flow that- fashes... to and ... is, immediately ' released oa tP {he envi~ronme~nt is. co`ntaned in Tablel 15. A1 4and is.presente n iue; .5... .3 11. q4 The breakb flo fractionf ..as.conser.ati.ely caclae assuming...... tha 100.. percen..t of the breakis,4 from.. the ., ho leg. side,4, of steam generator, w..+-ohereas t ,,he bre ak,- flow e actall co-÷nsitso f`low froma botrhr, the hot legn and cold leg sdes f thestea geneator perce ÷nt Thus the loaion o-afthtuerpreinosgifctfr (d) The rup÷tu remn (or leakage)-site i lassume to* ber always ~overed-with (e) The tot4;al primaryn leak rateha for,,the 3 intact steam th ruturedu,;4o steam generator is assumed. , to be10fonnflse 15.5-109 15.5109Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE (g) The noble gases in thc break flow and primary to secondary leakage are assumed to be transferred instantly out of the steam generator to the atmosphere. The whole body gamma doses are calculated combining the dose from the released noble gases with the dose from the iodine releases.(h) For the accident initiated iodine spike case, an iodine spiking factor of 335, obtained from Regulator; Guide 1.195, May 2003 (Refercnce
- 11) is assumed.(4\SL.Z)ZA~ k..w~zJwL;u;;
In equations 15.5 17 and 15.5 18, no credit is taken for a cloud depletion by ground deposition or by radioactive decay during transpo~ to the exclusion area boundary or to the outer boundary of the low population zone. Offsite thyroid doses are calculated using the equation: i L J 1 (1-5.5 !7)Iintegrate activity";+ of iodine n,,"lide i rel......ed duin the..t.me-breathing rate during time inter~al j in meter~IseGe+~4-(Tab4e45 7 5-68~- dil,,spersg'ktio facto i nglIrll.411: time1,, Illtrl jll Inl~DCF)1 -thyroid dose conversion factor via inhalation for iodine nuclide i in rem/Ci (Table 15.5 69)-thyroid dose via inhalation in rem Offsite whole body gamma doses are calculated using the equation: 15.5-110 15.5-110Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE seconds/me ._______.__ a'erag-eJ gamma energy fo~r noble ga i in-whobodyammaosedutoimersi ninre Table 15.5 71. All of these RSG doses are-with:in the allow..able guidelines as.. specified by the, SRP, Revision,, 2 !5.6.3).are ,,ithin; 10 C'FR 100.11 limits. The limiting dose,+ for SG-TR a-nalysis6~~cs. This dosern,,n~ exced thrNle SR 15.6.3+, alloal guidlinevale of 30.h~i infm+,,, , a. .-,,..,,,,..,letter to ,-,&E, dated, trnfebru '2 , 03 "sua of Amendment.. .............. conta~n 1 131,ab aReied Steam Generator- Tube Rupture ... Mai Steamc Line 155552012.2RContilnRoom Exposure DCPP UNITS 1 & 2 FSAR UPDATE room intake.,, Tlhe infh-ow ,nd4 unfiltered) to,' control, rorom and onrorrl room cntronrl Althougnh all1 release are terminated the RHRD system is put÷ in the calcuaiof~n iS cntinued,, to accont,,fo:rr additio~nal doses to cntninued, intact seamn gelneratonrs is assued, persist the dullraJtionr of the acclident.n an accnident initiated;nn iodne l p~ikewih a spikzingI factor of Both spike assumptionsconsiderN
- 0. ,P~ilgm D-1E. I. 131 secodr- act-nivity.
The whole body doses....... are,, ca'lculate ,,,,,g the dose from ,,,.,... relase ,,,.noble gases.. w,. ,,,ith the.,, dose°t hytroid d4ose cnve~rson factofr via,' for isotope i (DR',m/Ci) calculated dependentupnf i%-r inleakag,-l filte~red alnd filtered, ,,-,.,.'l.. bratin rae drn tm nerm (ic)(al 1. 8 15.5-112 Revision 19 May 2010 -1 DCPP UNITS 1 & 2 FSAR UPDATE Control room whole body doses arc calculated using the following equation;-GF whefe~(1-5.5-20) GFE -- gemernf, factor, calculated based n 1"7, uing th S_ 1173 ,,h .. ... .. r ÷CF -.....wher V is the..contro..room..olume.in... -,, , ri'sP mmsr~i lenrla,? ml,,tl n annr LA V ~ ~LI ill* ILA '.JIJIi I Lt~~jI L41.ILJH I ~i Ii...(Table 15.5 70)-concentration in the control room o calculated dependent upon inlcak~inflo~-{Gi-see/m~) i f rgy for isotope i (Mevidis)f isotope i, during time interval j,~gc, filtorod recirculation and filtered t:nntrnl room,-n L£tn ,oses~ 2re cicmwnnci using the following equation:-= -.. ..... ,,{I.4 L ']-.4Xi": I I }\*.... ../where-_.. aver...ge beta{ diintegratio energy... for .stoe. ..cds (Tbe1.5 7015.5 71 presents the airborne doses o t he cntnrol roonm operators. The resulkta~nt dose are. w.ell belo... the guidoines of. GD 19,~r and , are below the The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. Provided below are the critical SGTR-specific assumptions associated with control room response and activity transport. Timina for Initiation of CRVS Mode 4;i. An SIS will be generated at t = 219 sec following a SGTR.ii. The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The control room dampers are fully closed 10 secs later, or at t=257.2 15.5-113 15.5-113Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE secs (i.e., 219 + 28.2 + 10). The 2 second SIS processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay.iii. In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=231 secs (i.e., 219 + 2 secs signal processing time + 10 sec damper closure time).Control Room Atmospheric Dispersion Factors As noted in Section 2.3.5.2.2, because of the proximity of the MSSVs/10% ADVs to the control room normal intake of the affected unit, and because the releases from the MSSVs/10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation control room intake of the affected unit (closest to the release point) will be insignificant. Therefore, prior to switchover to CRVS Mode 4 pressurization, only the unaffected unit's control room normal intake is assumed to be contaminated by releases from the MSSVs/10% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points / control room receptors applicable to a SGTR at either unit are provided in Table 15.5-64B. The yjQ values presented in Table 15.5-64B take into consideration the various release points-receptors applicable to the SGTR to identify the bounding 7JQ values applicable to a SGTR at either unit, and reflect the allowable adjustments /reductions in the values as discussed in Chapter 2.3.5.2.2 and summarized in the notes of Tables 2.3-147 and 2.3-1 48.The bounding control room dose following a SGTR at either unit is presented in Table 15.5-64.15.5.20.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows: (1) The radiation dose to an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case as shown in Table 15.5-64.(2) The radiation dose to an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), is within 0.25 Sv (25 rem) TEDE for a pre-existing accident iodine spike case and 10% of the 10 CFR 50.67 limit for the accident initiated iodine spike case as shown in Table 15.5-64.15.5-114 15.5114Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE (3) The radiation dose to an individual in the control room for the duration of the accident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-64.SRP, Section 15.6.3, Revision 2 io .t; he ( 10 D CFR 10.1 dose lmt I~wfor tF he percent of t-he 10 CFR 100.1 d!,4-ose limits for,, the ...body and the thyroid fo ,the acci dent,.,,. intite iodine C pik case) as , sh.o wn,,,,, in,,, Table 15.5, 7... 4t (2 nindvda oae taypito h ue onaftelwpplto (3 n-, h accrda-rnce wnith requremnts ofGC 9 91te oetotecotoopertor u',,nder accident shall not be in. of 5rem hol skl...in,i Referencei
- 51) for..., the durtio of th acide f:o rl bot h .. 1 ...h th'p e exis-'tingthe acciden't initatedn iodin spike.cases a"sshown"in Tablevsn 15o 71zh,.,,.l Asnoe5i.Scio.5.221, h abvOGCA CN EQUdoES estiAte ec theKE ROTOR AnCDENT 1ith5n210.F 0.1 limipts.c CrtheriTa nlsi cetdbyteNCbae n spie anadilosisca ose.ueThis dofe exceedsshel nRt 156e3 allowabe gudlinet value oFR3 rem67 byd 0il5 rm.eoevter thdRooudte3.5rmvle acceptabl rtri fRglaoyGie in83 aul 2000 t Tube Rupturnead ManStaeLnBekwnlye.
EAB and LPZ Dose Criteria 15.5-115 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release shall not receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.Control Room Dose Criteria Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.Teradiologica;r-l of aloc-lkcd rotor accide4nt sh-all not exceed,,,- tIhe. dos,,ea limits, 10 CFR 100.1 a! below:.receive a totl radiation dosea to the who~le body inr aexes o,,f 25 rem or a trtal radiationthe entire periodl of its pasae),,r' a toa raito doeo"h whole bodyI in of 25 rem, or a #totl raditioiin dose in exes of "13001 rem to In acco-,rdance with. t+ he, requ,.irements of GDC 19,1 71,the d.ose to. the con.trol room 15.5.21.2-Identification of Causes and Accident Description 15.5.21.2.1 Activity Release Pathways This event is caused by an instantaneous seizure of a prima~ry reactor coolant pump (RCP) rotor. Flow through the affected loop is rapidly reduced, causing a reactor trip due to a low primary loop flow signal. Fuel damage is predicted to occur as a result of this accident. Due to the pressure differential between the primary and secondary systems and assumed SG tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere from the secondary coolant system via the 10% ADVs and MSSVs.Following reactor trip, and based on an assumption of a LOOP coincident with reactor 15.5-116 15.5-116Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE trip, the condenser is assumed to be unavailable and reactor cooldown is achieved using steam releases from the SG MSSVs and 10% ADVs until initiation of shutdown cooling. DCPP has established that the LOL event generates the maximum primary to secondary heat transfer and the LRA assumes these same conservatively bounding secondary steam releases.Under adverse circumstances a lo1e roto acidn col as malaonso fuel÷, cladn filur the.. core.;If this ccurst some; fi.ion.. products, will. entr.... the cooan ad il msty emininth colntunilcland p y heprmay ooan Integushort tequremeitth acoident focrsaa theAime pehient sigifcnt pions' egator secodar seco3nclda ng sysempTe noble Gaiseuswled hred to dvethte o e atmosphere viaotela.ejec+,tors.. by ;=, way .of atmospheric.= steam dump. The, iodines, wil remain. m ostyn h...dines÷,, thoevr willne released,. C,-f,,, via, the air ejectors.o by way. tofatmospheric steam.n .dmp.I diin fa toshrcsemdm sncssssm fteatvt contained in+, th secoda .syste prior.+,.4#~ to the accidet,. w.ll be,/ Ll released. ,P+l'lll depend ...o"n th;tme, orelie valves remain ,open+ and the avilbiit o cndnsrbyas colngc15c~.5-1 Th mutfRadioaioe i9odine01 DCPP UNITS 1 & 2 FSAR UPDATE Tab, 1 ... T, wer ..... fr, ...p.riod of 8 hours, foloin T~ the st! fth!acdet dctald n lcssooh significant assmptons; ,areo Rcfcrence The assumptions .. .....=d for meteorology, breathing rates pouation+desit and other.activitie riorT l to. the acciden are d......c in... Seton1.5 Inph ore"odtrietepia, oln ciiisimdaeyatrteacdni a programn tofo alulae thoe activit relase amnd;fo potntial(; doseolowing' accidnt.no4 Th~.~e alculated aciit are I'lise nTbe1. h oeta oe r gIe InI,,I ..ITab.e1I 15. 2 h xoue aeas hw nFgre 55I n 551 Tas a functio of the. amount. o fuelrfailuwre occurs Onhe left. bounda; of these graphs in the egino hofo frnegigiboe fuel failre,rn the exosures aore just thecopnt through the steam.. generators c at pr p Tiar coolan levels. e..posure.. is the.. .long term by... and lowrf"';"eak, from, teh , prmr' oln ...te. Thea .. ctiity through these.... path.a...s, prinipaly K Figre 5.516.Sineteativty-elese ithswyoud Reahte evirion19Ment01 DCPP UNITS 1 & 2 FSAR UPDATE Frm tes nh r ter an l ng term a L I,, als be conc ,.luded~l that ,al~l ll,.,,,th pThentRAias eposturelroatled roreutorn acciduent faillube relultbelow the guideline leele associated gap activity. As discussed in Section 15.5.3.1.3, the core gap activity is assumed to be comprised of 8% of the core 1-131 inventory, 23% of the core 1-132 inventory, 35% of the core Kr-85 inventory, 4% of the remaining core noble gas inventory, 5% of the remaining core halogen inventory, and 46% of the core alkali metal (Cesium and Rubidium) inventory. Table 15.5-42A lists the key assumptions /parameters utilized to develop the radiological consequences following a LRA.Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a LRA.15.5.21.2.2 Activity Release Transport Model In accordance with Regulatory Guide 1 .183, the activity released from the fuel is assumed to be released instantaneously and mixed homogenously through the primary coolant mass and transmitted to the secondary side via primary to secondary SG tube leakage. A radial peaking factor of 1 .65 is applied to the activity release from the fuel gap. The activity associated with the release of the primary to secondary leakage of normal operation RCS, (at Technical Specification levels) via the MSSVs/10% ADVs are insignificant compared to the failed fuel release and are therefore not included in this assessment. DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the LRA dose Consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).The chemical form of the iodines in the gap are assumed to be 95% particulate (Csl), 4.85% elemental and 0.15% organic. The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events), has been evaluated for potential impact on dose consequences as part of a Westinghouse Owners Group (WOG) Program and demonstrated to be insignificant; therefore, the gap iodines are assumed to have a partition coefficient of 100 in the SG. The iodine releases to the environment from the SG are assumed to be 97% elemental and 3% organic. The gap noble gases are released freely to the environment without retention in the SG whereas the 15.5-119 15.5119Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE particulates are assumed to be carried over in accordance with the design basis SG moisture carryover fraction.The condenser is assumed unavailable due to the loss of offsite power. Consequently, the radioactivity release resulting from a LRA is discharged to the environment from all steam generators via the MSSVs and the 10% ADVs. The SG releases continue for 10.73 hours, at which time shutdown cooling is initiated via operation of the RHR system and environmental releases are terminated. 15.5.21.2.30Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. For the LRA, the worst two hour period can occur either during the 0-2 hr period when the noble gas release rate is the highest, or during the t=8.73 hr to 10.73 hr period when the iodine and particulate level in the SG liquid peaks (SG releases are terminated at T=1 0.73 hrs). Regardless of the starting point of the worst 2 hr window, the 0-2 hr EAB x/Q is utilized.The bounding EAB and LPZ dose following a LRA at either unit is presented in Table 15.5-42.15.5.21.2.4 Control Room Dose Assessment The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. Provided below are the critical LRA-specific assumptions associated with control room response and activity transport. Timinq for Initiation of CRVS Mode 4 (if applicable): The LRA does not initiate any signal which could automatically start the control room emergency ventilation. Thus the dose consequence analysis for the LRA assumes that the control room remains in normal operation mode.Control Room Atmospheric lDispersion Factors As noted in Section 2.3.5.2.2, because of the proximity of the MSSV/10% ADVs to the control room normal intake of the affected unit and because the releases from the MSS Vs/i10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation control room intake of the faulted unit (closest to the release point) will be insignificant. Therefore, only the unaffected unit's control room normal intake is assumed to be contaminated by a release from the MSSVs/10%ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points / control room receptors applicable to an LRA at either unit are provided in Table 15.5-120 15.5120Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 15.5-42B. The x/Q values presented in Table 15.5-42B take into consideration the various release points-receptors applicable to the LRA to identify the bounding %/values applicable to a-LRA at either unit, and reflect the allowable adjustments /reductions in the values as discussed in Section 2.3.5.2.2 and summarized in the notes of Tables 2.3-1 47 and 2.3-1 48.The bounding control room dose following a LRA at either unit is presented in Table 15.5-42.15.5.21.3 Conclusions 15 .......... 5 1 hteatvt rl.a.e..c.lcu..t.d foa LBLO ^,, gie in.. bl.. 1.5.. 3.. n 15,.5 , t. cl.a I)e. .t.l ,n becnlddta ny control roomIu v l expo sures.~bal .... a rotor.accident+ wi'l be .... the GDC 19, 1971, criterion, leve...The analysis demonstrates that the acceptance criteria are met as follows: (1) The radiation dose to an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.025 Sv (2.5 rem) TEDE as shown in Table 15.5-42.(2) The radiation dose to an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), is within 0.025 Sv (2.5 rem) TEDE as shown in Table 15.5-42.(3) The radiation dose to an individual in the control room for the duration of the accident is within 0.05 Sv (5 remn) TEDE as shown in Table 15.5-42.(1) The radiation" dose to the .hole body, and, to. the thyroid of. an4 i, .ndividual4, located at.. an point on the b..ndary of t.., .. he area... for; th...... hours.... , atdanynpoint on'+ the. outer bouna+ of... th ....... zone,+h is;e;+o sed-15.5-121 15.5-121Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE (3) the, activity;, from,' the in,, Table, , control room d4ose which might o...cu... woul be ..el. within the established 15.5.22 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT The procedures used in handling fuel in the containment and fuel handling area are described in detail in Section 15.4.5. In addition, design and procedural measures provided to prevent fuel handling accidents are also described in that section, along with a discussion of past experience in fuel handling operations. The basic events that could be involved in a fuel handling accident are discussed in that section, and the following discussion evaluates the potential radiological consequences of such an accident.The assumption of a LOOP related to a postulated design basis accident which leads to a reactor trip does not directly correlate to an FHA. Specifically, a FHA does not directly cause a reactor trip and a subsequent LOOP due to grid instability; nor can a LOOP be the initiator of a FHA. Thus the FHA dose consequence analyses are evaluated without the assumption of a LOOP.15.5.22.1 Fue Hal-'ndtling Acc-ident In, The FUel Ha=ndlin,- ^,-ca!5.5.22.! .!Acceptance Criteria The radiological consequences of a FHA in the Fuel Handling Building (FHB) or in the Containment shall not exceed the dose limits of 10 CFR 50.67, as modified by Regulatory Guide 1 .183, July 2000 and outlined below: EAB and LPZ Dose Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release shall not receive a radiation dose in excess of 0.063 Sv (6.3 rem) TEDE.(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.063 Sv (6.3 rem) TEDE.Control Room Dose Criteria (10 CFR 50.67)Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.15.5-122 15.5-122Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE not, exceed t4he limt÷s 10 CFR 50.67 a... outlined belo....: (1) An ,lcoated at any point on the, boun.... of the. exclusion area.... for releaseJo sha"ll receive a total," in evx:css o 0,..063 (6.3:: " total24. Ientffctiione dose eiales nt ATide). Dsrito (552),, An inividua Rlocated Pathwanyspito h ue ona ftelpplto zoehwois exet pos edlato thatasetfe radiasemclou sdropes ulting fromelngi the psa fissaiomnt prloducth rueleaose (du4ring i the entrepperid ofue itassae),l hall notme to receimaeea totsall rfteaditionidos in texcuessa of 0th6 Sd(63ropem) toalseffectise (3)Thentdos to the Rcntr! roo moerametor unde accdeto Condaiition sheallntibng in Li e xessoof 0.058 Snd (5R8 rem)ttapfectivel doeference 87),nth (TsDm)tfor the tal 15.5.22.2. inoeactivity release Pathwayseraiebcuetekntceeg vial This evesnt potuataes tha aspn fuel assembly i dropped duoghwtring rfuxelin in the do dspetane Fuel Poolti eneFgy lococated inteFBwrinh the reactom r cavityt loate inute Containmcient. All nofth fulcodsi(264rods)icnt the dropptued fuel eiassentnmblyaer assfumed tobedamgdstu l of the activity in thbfelgaho the dropped assembly idteimatda sely asumedg touel hadinstpeantanously releasdinmeto thsue sF ointoth reureacto eeavity. As dcunaimente ientilthopre NCsERfotAenm oenatson8 and 6etoaDCPP Faciity Opratinge fulcodsainmn thoneh assmbl rupureisc consanerv soatiovavesbecause theknei ergavalvsable forncausing damages Toia fulowo assebl droppe throughawatert is fixcared by the do dnistance.t vathe kineti veneryascaetit.h aiu rpegtfrafe Thading exaccdetsra is nomonsidered sufirctiiety ymntoruptur the eqivlent numert of fuelee Duringtulte fuel handling oeationdcntainmeplnt closureoistnot required.mGenerallytthe th uoai lsr fcontainment ventilation presse isoope ationa vand vexhauThs a tiiryfromethe ofay postulate inofuiel handinlgacident thpsuespatvn.oitr ilaamn euti 15.5-123 15.5-123Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE In addition to radiation monitor indications, a fuel handling accident would immediately be known to refueling personnel at the scene of the accident. These personnel would initiate containment closure actions and are required by an Equipment Control Guideline to be in constant communication with control room personnel. The plant intercom system is described in Section 9.5.2.Containment penetrations are allowed to be open during fuel handling operations. The most prominent of these penetrations are the equipment hatch and the personnel airlock. Closure of these penetrations is achieved by manual means as discussed in Section 15.4.5. The closure of these penetrations is not creditsad in the design-basis fuel handling accident inside containment. Following manual containment closure after the fuel handling accident, activity can be removed from the containment atmosphere by the redundant PG&E Design Class II Iodine Removal System (two trains at 12,000 cfm per train), which consists of HEPA/charcoal filters. This system is described in Section 9.4.5. There are no Technical Specification requirements for this filtration system.The containment can also be purged to the atmosphere at a controlled rate of up to 300 cfm per train through the HEPA/charcoal filters of the hydrogen purge system. This system is described in Section 6.2.5.therm.l) precedi shutdon... The accidnht is ....sumed to occur. 100" hours.after This Iatter internal rep.resents approximatelhy the time to prepare (coldont .. head- and internasI remo.al. caIty. ;, flooding, etc.) the core. fo-r and *s therefore ..omew.hat ........iv that it ..ould" require that the, occu dureing handing- ;of the nfirstfew fuel assemblies. ;t~nm Th source... tempi bonetr .... ... assme tobeacopoie.f.h.hgh...sso product.... acivt totals.fo vriou cmbinations ofi b ...... and , enrichment. The, forGE 15*5-124 15.5-124Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE The a'ssum*r~es thatJ the fissio;rn productJr atJ a wate.J÷r deptr.h of 23 feet, whic;h is the, minimu .. ater. above.. the. top, of the fue, a.... requre byfeet account.. for case... in which the relea occurs... from-, the. top, of ...n .asemly, that is, e=rticlly"Jh on the floor,r andt for releeases that occuir near the of the storage racks. F~inaly co..nsistent ..ith Safetylln Gu.de 25 Marc.li 1972ln the an..si ass{1ume that0 froelm thne area withn a 2n hour period.~ qhh Of th ciiyrahn h ae,10pren ftenbegss eo n 'tn are assmedtTeimdaeyTeesdt h ue adigae irsae.Hwvr demonstrate the adequacy of thc fuel 44~~~ing %.l safety systems In the very unlikely event of a serious fuel handling accident and in combination with the conservative assumptions discussed above, containment building or fuel handling area activity concentrations may be quite high. High activity concentrations necessitate the evacuation of fuel handling areas in order to limit exposures to fuel handling personnel. Upon indication of a serious fuel handling accident, the fuel handling area will be evacuated until the extent of the fuel damage and activity levels in the area can be determined. Any serious fuel handling accident would be both visually and audibly detectable via radiation monitors in the fuel handling areas that locally alarm in the event of high activity levels and would alert personnel to evacuate.....h consr.at.el neglected.. this ana.l'is, theThe fuel handling area has the additional safety feature of ventilation air flow that sweeps the surface of the spent fuel pool carrying any activity away from fuel handling personnel. This sweeping of the 15.5-125 15.5-125Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE spent fuel pool is expected to considerably lower activity levels in the fuel handling area in the event of a serious fuel handling accident.A fa I I I f i Atter cnarcoai T!iter cieanup kanotrler design Teature conservatively neglected in mis analysis), fuel handling area post accident ventilation air exhausts through the plant vent at a height of 70 motors. Site meteorology is such that it is very unlike!y that any airborne activity will enter the control room ventilation system.Spent fuel cask accidents in the fuel handling area causing fuel damage are prec!uded Spent fuel cask accidents in the fuel handling area causing fuel damage are precluded due to crane travel limits and design and operating features as described in Sections 9.1 .4.3.9 and 9.1.4.2.6. Spent fuel handling accidents in the fuel handling area would not jeopardize the health and safety of the public.The FHA dose assessment follows the requirements provided for the FHA in pertinent sections of Regulatory Guide 1.183 including Appendix B. As discussed in Section 15.5.3.1.3, the core gap activity is assumed to be comprised of 8% of the core 1-131 inventory, 23% of the core 1-132 inventory, 35% of the core Kr-85 inventory, 4% of the remaining core noble gas inventory, 5% of the remaining core halogen inventory and halogen isotopes, and 46% of the core alkali metal (Cesium and Rubidium) inventory. Table 15.5-47A lists the key assumptions / parameters utilized to develop the radiological consequences following an FHA at either location and at either unit.DCPP procedures prohibit movement of recently irradiated fuel which is defined as fuel that has occupied part of a critical reactor core within the previous 72 hours. Table 15.5-47C provides the gap activity inventory of the noble gases, iodines and alkali metals in a single fuel assembly at 72 hrs post reactor shutdown.DCPP Technical Specification 3.7.15 requires the SFP water level to be >23 feet over the top of irradiated fuel assemblies seated in the storage racks. Technical Specification
3.9.7 requires
the refueling cavity water level to be maintained _23 feet above the top of the reactor vessel flange. Additional margin is provided through operating procedures. Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a FHA 15.5-126 15.5126Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 1 5.5.22.2.2 Activity Release Transport Model The fission product inventory in the fuel rod gap of all the rods in the damaged assembly are assumed to be instantaneously released into the spent fuel pool or reactor cavity, both of which have a minimum of 23 ft of water above the damaged fuel assembly. A radial peaking factor of 1 .65 is applied to the activity release.Per Regulatory Guide 1.183, the radioiodine released from the fuel gap is assumed to be 95% particulate (Csl), 4.85% elemental, and 0.15% organic. Due to the acidic nature of the water in the fuel pool (pH less than 7), the CsI is assumed to immediately disassociate and re-evolve as elemental iodine, thus changing the chemical form of iodine to 99.85% elemental and 0.15% organic. In addition, and per Regulatory. Guide 1 .183, an iodine decontamination factor of 200 is assumed for the SFP / reactor cavity.Noble gases and unscrubbed iodines rise to the water surface where they are mixed in the available air space. All of the alkali metals released from the gap are retained in the pool. In accordance with Regulatory Guide 1 .183, the chemical form of the iodines above the pool is 57% elemental and 43% organic.Per Regulatory Guide 1.183, the activity released due to an FHA is assumed to be discharged to the environment in a period of 2 hrs (or less if the ventilation system promotes a faster release rate).FHA in the FHB The radioactivity release pathways following an FHA in the FHB are established taking into consideration the following Administration Controls: During fuel movement in the FHB: a. The movable wall is put in place and secured b. No exit door is propped open c. One FHBVS exhaust fan is operating (The supply fan flow (if operating) has been confirmed by design to have less flow than the exhaust fan)Operation of the Fuel Handling Building Ventilation system (FHBVS) with a minimum of 1 exhaust fan operating and all significant openings administratively closed will ensure negative pressure in the FHB which will result in post-accident environmental release of radioactivity occurring via the Plant Vent. The activity release due to the FHA in the FHB is assumed to be discharged to the environment as follows: a. A maximum release rate of 46,000 cfm via the Plant Vent due to operation of the FHBVS with a closed FHB configuration.
- b. A maximum conservatively assumed outleakage of 500 cfm occurring from the closest edge of the FHB to the control room normal intake (i.e., 30 cfm 15.5-127 15.5-127Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE outleakage is assumed for ingress/egress; 470 cfm is assumed for outleakage from miscellaneous gaps/openings in the FHB structure).
It has been determined that for the FHA in the FHB, the actual release rate lambda based on the FHBVS exhaust (i.e., 8.7 hr') is larger than the release rate applicable to"a 2-hr release" per Regulatory Guide 1.183 (i.e., 3.45 hr-'). Thus the larger exhaust rate lambda associated with FHBVS operation plus the exhaust rate lambda for the 500 cfm outleakage is utilized in the analysis.FHA in the Containment The potential radioactivity release pathways following a FHA in the containment are established taking into consideration
- a. Operation of the containment purge system which would result in radioactivity release via the plant vent b. Plant Technical Specification Section 3.9.4 that allows for an "open containment" during fuel movement in containment during offload or reload.The most significant containment opening closest to the Control room normal operation intake is the equipment hatch. The equipment hatch is an approximately 20-ft wide circular opening in containment.
In the event the containment purge system ceased to operate (a viable scenario since it is single train and has non-vital power), the density driven convective flow out of the equipment hatch (due to the thermal gradient between inside and outside containment conditions), could be significant. It has been determined that for the FHA in the Containment, the release rate assuming a regulatory based 2 hr release is larger than that dictated by the containment purge ventilation system, or convective flow out of the equipment hatch. Thus the regulatory based release rate (i.e., 3.45 hr-'), is utilized for this analysis. Review of the atmospheric dispersion factors associated with the plant vent vs the equipment hatch indicates that dose consequences due to releases via the equipment hatch will be bounding.15.5.22.2.3 Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. Since the FHA is based on a 2-hour release, the worst 2-hour period for the EAB is the 0 to 2-hour period.The bounding EAB and LPZ dose following a FHA at either location and at either unit is presented in Table 15.5-47.15.5-128 15.5-128Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 15.5.22.2.4 Control Room Dose Assessment The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. Provided below are the critical FHA-specific assumptions associated with control room response and activity transport. Desiqin Basis FHA (occurs at t=72 hours after reactor shutdown)Credit is taken for PG&E Design Class I area radiation monitors located at the control roomcontrol room normal intakes (1-RE-25/26, 2-RE-25/26) to initiate CRVS Mode 4 (filtered / pressurized accident ventilation) upon detection of high radiation levels at the control room normal intakes as a result of an FHA.An analytical safety limit of 1 mR/hr for the gamma radiation environment at the control room normal operation air intakes has been used in the FHA analyses to initiate CRVS Mode 4. Note that the actual monitor trip setpoint is lower to include the instrument loop uncertainty. The radiation monitor response time is primarily dependent on the type of monitor, the setpoint, the background radiation levels and the magnitude of increase in the radiation environment at the detector location.For a monitor with an instrument time constant of "t" (2 seconds) and a background of 0.05 mR/hr, the response time "t" to a high alarm Setpoint (HASP < 1 mr/hr), for a step increase of radiation level DR (mR/hr) is determined by solving the following equation that represents the monitor reading approaching the final reading exponentially. t HASP = 0.05 +DOR(l -e-It is determined that a DBA FHA (i.e., occurs at 72 hrs post shutdown) will result in a radiation environment at the control room normal operation intakes that greatly exceed the analytical limit of 1 mR/hr for initiating CRVS Mode 4. This will result in an almost instantaneous generation of a radiation monitor signal to initiate CRVS Mode 4 (radiation monitor response time is estimated to be < 1 sec). For purposes of conservatism, and since the delay in isolation of the normal intake has a significant impact on the estimated dose consequences, the analysis conservatively assumes a monitor response time to the HASP of 20 secs.As discussed in Section 15.5.1.2, when crediting CRVS Mode 4, the FHA dose consequence analyses is not required to address the potential effects of a LOOP.Thus delays associated with diesel generator sequencing are not addressed. Therefore, the time delay between the arrival of radioactivity released due to a D8A FHA at both the control room normal Intakes (assumed to be instantaneous) and CRVS Mode 4 operation is estimated to be the sum total of the monitor response time (20 15.5-129 15.5-129Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE secs), the signal processing time (2 Secs) and the damper closure time (10 secs) for a total delay of 32 seconds.Delayed FHA: It is recognized that the response time for radiation monitors are dependent on the magnitude of the radiation level / energy spectrum of the airborne cloud at the location of the detectors, which in turn are dependent on the fuel assembly decay time. Thus an additional case is considered for each of the two FHA scenarios described above (i.e., a FHA in the FHB and a FHA in Containment) when determining the dose to the control room operator; i.e., a case that reflects a delayed FHA at Fuel Offload or a FHA during Reload, occurring at a time when the fuel has decayed to such an extent that the radiation environment at the control room normal intake radiation monitors is just below the setpoint; thus the control room remains in normal operation mode and CRVS Mode 4 is not initiated. The analyses determined that the dose consequences of a DBA FHA bound that associated with the delayed FHA for both the FHA in the FHB and the FHA in the containment. The bounding atmospheric dispersion factors applicable to the radioactivity release points / control room receptors applicable to an FHA at either location, and at either unit, are provided in Table 15.5-47B. The yIQ values presented in Table 15.5-47B take into consideration the various release points-receptors applicable to the FHA to identify the bounding x/Q values applicable to a FHA at either unit and at either location, and reflect the allowable adjustments / reductions in the values as discussed in Section 2.3.5.2.2 and summarized in the notes of Tables 2.3-147 and 2.3-148.The bounding control room dose following a FHA at either location and at either unit is presented in Table 15.5-47.15.5.22.1=.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows: any t'"-o hour fo.llo..ing the onset' of the po"tulated fisson produc,,,t total eff,,;,,'1, eqialn (TEDE) as- shown in T abl 5.15.5-130 15.51 30Revision 19 May 2010 DCPP UNITS I &2 FSAR UPDATE (3) The doeto the to nltrol room operto+r under accdet onditiolns shall not be in duJration of the accident as shown in Table 15.5 47.The analysis demonstrates that the acceptance criteria are met as follows: (1) The radiation dose to an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.063 Sv (6.3 remn) TEDE as shown in Table 15.5-47.(2) The radiation dose to an indivdual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), is within 0.063 Sv (6.3 remn) TEDE as shown in Table 15.5-47.(3) The radiation dose to an individual in the control room for the duration of the accident is within 0.05 Sv (5 remn) TEDE as shown in Table 1 5.5-47.15.5.22.2.1. Acc. tanc Cri....The radilogical oN#nsqeo~rnces-of fuel hiandiHngaccidnt' oriAn nsdo;4 rntainmeontnot.excee the d~oselmt .;of 10 CFR 100.1 as outline belo:...An ^ individual located at any, point on the b.undan,' of th.. exclusion+, area.product relasehll n1ot a÷otal rad,,iation', d--o,- sea to-, thewhole body,4,[0 ecs0f6me rattl aito os necs f 5rmt h thrid rmidn oue 15.5-131 15.5131Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEAA A A II p A I a U jl I Siacntmcation or Lauscs and Acclclcnt uescriprion .. ar containment The. followin e.. sowh t.. hat During fue hand...ing opreations, contain.ent.clo.ur i , not rqre... Generall..y, en..ironment via the plant vent.+of: a postu(lateda fuel handlIng acident, the pla'nt veont monitors il11 alarm and result+ in Cro+ntaiment pne trations are all~fow~ed tor airlock Clnosure of these is ach i pean duin hnln prtos h equipmen hatc and th personne-_ l_~ection 1 ~.1 .~. I ne closure OT tnese penetrat!ons is not creaitea in tne oesign easis fuel handling accident inside containment. The FHA analysis assumes that the control room ventilation system of each unit remains in the normal mode of operation following the FHA. Thus, the design basis FHA does not credit charcoal filtration of the control room atmosphere intake flow or recirculation flow.The evaluation of potcntial offsitc exposures was pe~ormed for a design basis ease, assuming plant parameters as limited by Technical Specifications. The assumptions of flft~ fl.5r flr~ .K.~L AnVfl -- ....~....... ...rSL £L ~ ~.I..t...2f..4 ~aTety ~uioe ~o, viarcn i~i:~, were u~cu a~ yu:u~uIue wiui tile uxwpuuii~ uet~.tuuu e4ew 7 15.5.22.2.2.1 Activity Released to Containment Atmosphcrc the containment refueling pool following the postulated accident are identical to those-I A~ I! A ~I I ,wth those in ...fety Guides. 25 ac 92,. .and...18 .. Jul 2000. with the .guidance f Safety, Guie,2,; O Marsh 1972, it ..as.assumed hat all radoactive, idn;e in the rods "at the time of the accident.15.5-132 15.5-132Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE The,, do.e.con.ersion factor u..ed. rcfro ICRP Q30 (Rferenc.. /15). The,,.use. of ths dose conv..rsion factors is consistent, wit;h the current guidanc provided. r 1 5.5.22.2.2.2 Contanet lsr Inaddition to;+~ radatomnir ns, a-lhr fuel handling accidentr wold immediately be hO know to rfuelng pe~rso nne hat the scne ofthe +acc T Ahesprsonnel would intae cotinmen clsr acions andar rqired by an Equipmen Contro Guidein generFators dO not surround. tn poo.., tn radioactiv.. y w"" .... actual"'y oc dsperse. in'o a 33,600 cubic foot volume and4 was. thentanso.... ra, to the environment. within'; a secornd priod~r,, throug.,'-h the, openn equ,,ipment hatc.+exposures were calculated,4 for the postuloated fueil hand4ling acc-ident insie;4 containment fuell hand4lng accide,4nt radilogicar-l exposuraes The a rnlclfated releasesof the 15.5-133 15.5-133Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE p,.+,ostuatc fuel. handling accident,. inside c..ontainmen a re presented in- Tabhle 15.5 50.Ths ..p.. ur... are......., l ,:,;th. n h.e 10 CFR 100.11 limit;s.Follwin .. anua..... clour a...fter the fuel. hand`ling.. activit,;can be Iodi`4ne Remo..l System+ (,- rains -at. 124 ,000cf per train) w... ."hich consists 1HERAlc.ha-rcoa-J filters. This sy...em- is describ;ed` in Setin-.1;-.-
- 5. 4 There, are no' Spcfcation.,.,, requirement..s+
for. this filtratio ..ste.. ..The cntainent cn als beurgedto.th.atmopher at.. a contolld. rte.f. u to 30 cf per" train through, o-hehdrgn ugesstmTi (1) The radiationrdose to the ...hole, body and. to t, he thyro..id of an wel eowte;os imt of 10 CFR 100.11 as shown..in;Tableb15.5 50.Th en nraditin, dos tothewh ole body ra andc-n toD the thyroi ofh ane indivdua located,,r, at, any- ,, thoutr ,,nda ,,' of heow popul ation zon.e, who i.. ÷... to th.rdiac. plod esltn ,fro posulaed fisn producth... , and4 beta skin, Reference..
- 51) for the` duration, of the accident as shown.. in 15.5-134 15.5134Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE 15.5.22.3 Conclu,..,,sion,, Fuci Hand-,ling,.
^,cidente, poential exposures~n( # indiua membernn of the-public n;and the. gel."ner,,:al' populationl hae in. a- rDCPP unit,; would.h. not a'n undue, to" the, hea.'lth a"ndl t- he pubhlic-AuuH;innlhII it ca,-n beconcnluded that CES for h 15,5.23 RADIOLOGICAL CONSEQUENCES OF A CONTROL ROD EJECTION ACCIDENT 15.5.23.1 Acceptance Criteria The radiological consequences of a CREA shall not exceed the dose limits of 10 CFR 50.67, and will meet the dose acceptance criteria of Regulatory Guide 1.183, July 2000 and outlined below: EAB and LPZ Dose Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release shall not receive a radiation dose in excess of 0.063 Sv (6.3 rem)TEDE.(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.063 Sv (6.3 rem)TEDE.Control Room Dose Criteria Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.do.se limits. of 10 CFR 100.11 as out..n. below:' ...15.5-135 15.5-135Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE An incli,_,rligl !oneted et 'n,' noint on+he ho,,ndgr" of t+he ,rei for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.ii. An individual located at any point on the outer boundary of the low nontilMion lone who i~. e~no~.nd to the rndio~r.ti'ie cloud recultino from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in oxcess of 25 rem, or a total radiation dose in excess of 300 rem to the ulywIu uum iuuuiu e~pu~ui~.(2~ In accordance with the rcnuircrnent~ of ~DC 19 1971 the dose to the conro rom perto uneraccdet cndtios hal nt b i exes, o 5 15.5.23.2 Identification of Causes and Accident Description 15.5.23.2.1 Activity Release Pathway As discussed in Section 15.4.6, this event consists of an uncontrolled withdrawal of a control rod from the reactor core. The CREA results in reactivity insertion that leads to a core power level increase, and under adverse combinations of circumstances, fuel failure, and a subsequent reactor trip. In this case, some of the activity in the fuel rod gaps would be released to the coolant and in turn to the inside of the containment building. As a result of pressurization of the containment, some of this activity could leak to the environment. Following reactor trip, and based on an assumption of a Loss of Offsite Power coincident with reactor trip, the condenser is assumed to be unavailable and reactor cooldown is achieved using steam releases from the SG MSSVs and 10% ADVs until initiation of shutdown cooling. DCPP has established that the LOL event generates the maximum primary to secondary heat transfer and the CREA assumes these same conservatively bounding secondary steam releases.Regulatory requirements provided for the CREA in pertinent sections of Regulatory Guide 1 .183 including Appendix H is used to develop the dose consequence model.Table 15.5-52A lists the key assumptions / parameters utilized to develop the radiological consequences following a CREA.The CREA is postulated to result in 10% fuel failure resulting in the release of the associated gap activity. Per Regulatory Guide 1 .183, the core gap activity is assumed to be comprised of 10% of the core noble gases and halogens. A radial peaking factor of 1 .65 is applied to the activity release from the fuel gap.15.5-136 15.5-136Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE In accordance with the requirements provided in Regulatory Guide 1.183, two independent release paths to the environment are analyzed: first, via containment leakage of the fission products released due to the event from the primary system to containment, assuming that the containment pathway is the only one available; and second, via releases from the secondary system, outside containment, following primary-to-secondary leakage in the steam generators, assuming that the latter pathway is the only one available. The actual doses resulting from a postulated CREA would be a composite of doses resulting from portions of the release going out via the containment building and, portions via the secondary system. If regulatory compliance to dose limits can be demonstrated for each of the scenarios, the dose consequence of a scenario that is a combination of the two will be encompassed by the more restrictive of the two analyzed scenarios. Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a CREA.For basis case, it was ass.,ume..d that the plant had bee operat. ing insta.'ntaneouhl released to t'he primer; coolant. Rleases too t he primaryn t'colant are The at+;"'+" it relase to the from the priar coolant throug....h the ruptured.. control rod mechanism pressu~re housi~ng is assuvmed tro be mixe throughou the-,, containment and. is, ava.. ilable for leakage-,, to, atosphere.., iodine in c-hemical.- forms that are. no affected by the spray. syste are... negligibe. These[........ ......... ... ...... ... [ U [I I ... ........... .... ..... ....... .... .Revision 19 May 2010 15.5- 137 DCPP UNITS 1 & 2 FSAR UPDATE Forl.... ng the,, relase.. to thc, contain;,ment, the fissio.... products are ....assume to.. ,..leakL the, containment at the same rates for. the-,, discu.s.ccd, 4.in Scetion ! 55.17. In, -addition, the ..pra.. system is a.sumed to be, in operation and.. acts The as-sulmptions used for meteorology, breathing rates, poplatiofn densit;y, and4 other common factor ! wer5 "te ..lso descibe inerle, scins ot h piar n mactviie are..... listed in Tab.e 11a, ;1 7. -;r 1 4T ... ÷...;..The calculated activity releases""" ....o ar lis" ~ate inc abe 15.5 51,÷ and the-, potential. doses.. arc given in Table 15.5. 52.. .. Thrid doses th~at woul result.n from seonar steam÷ releases.. a 15.5.23.2.2 Activity Release Transport Model The CREA dose consequence analysis evaluates the following two scenarios. Scenario 1: The failed fuel resulting from a postulated CREA is released into the RCS, which is released in its entirety into the containment via the faulted control rod drive mechanism housing, is mixed in the free volume of the containment, and then released to the environment at the containment technical specification leak rate for the first 24 hrs and at half that value for the remaining 29 days.Scenario 2: The failed fuel resulting from a postulated CREA is released into the RCS which is then transmitted to the secondary side via steam generator tube leakage. The condenser is assumed to be unavailable due to a loss of offsite power.Environmental releases occur from the steam generators via the MSSVs and 10%ADVs.15.5-138 15.5-138Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE The chemical composition of the iodine in the gap is assumed to be 950/ particulate (Csl), 4.85% elemental and 0.15% organic. However, because the sump pH is not controlled following a CREA, it is conservatively assumed that the iodine released via the containment leakage pathway has the same composition as the iodine released via the secondary system release pathway; i.e.; it is assumed that for both scenarios, 97% of all halogens available for release to the environment are elemental, while the remaining 3% is organic.Scenario 1: Transport From Containment The failed fuel activity released due to a CREA into the RCS is assumed to be instantaneously released into the containment where it mixes homogeneously in the containment free volume. The containment is assumed to leak at the technical specification leak rate of 0.10% per day for the first 24 hours and at half that value for the remaining 29 days after the event. Except for decay, no credit is taken for depleting the halogen or noble gas concentrations airborne in the containment. Per Regulatory Guide 1.183, the chemical composition of the iodine in the gap fuel is 95%particulate (CsI), 4.85% elemental and 0.15% organic. However, since no credit is taken for the actuation of sprays or pH control, the iodine released via containment leakage pathway is assumed to have the same composition as iodine activity released to the environment from the secondary coolant; i.e.; 97% elemental and 3% organic.Environmental releases due to containment leakage can occur unfiltered as a diffuse source from the containment wall, and as a point source via the containment penetration areas or the Plant Vent. The dose consequences are estimated based on the worst case atmospheric dispersion factors, i.e., an assumed environmental release via the containment penetration areas.Scenario 2: Transport from Secondary System The failed fuel activity released due to a CREA into the RCS is assumed to be instantaneously and homogeneously mixed in the reactor coolant system and transmitted to the secondary side via primary to secondary SG tube leakage. The activity associated with the release of the initial inventory in secondary steam/liquid, and primary to secondary leakage of normal operation RCS, (both at Technical Specification levels) via the MSSVs/10% ADVs are insignificant compared to the failed fuel release, and are therefore not included in this assessment. DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the CREA dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events), has been evaluated for potential impact on dose consequences as part of a WOG Program and demonstrated to be insignificant; therefore, the gap iodines have a partition coefficient of 100 in the SG. The gap noble gases are released freely to the 15.5-139 15.5139Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE environment without retention in the SG.The condenser is assumed unavailable due to the loss of offsite power. Consequently, the radioactivity release resulting from a CREA is discharged to the environment from steam generators via the MSSVs and the 10% ADVs. Per Regulatory Guide 1.183, 97% of all halogens available for release to the environment via the Secondary System are elemental, while the remaining 3% are organic. The SG releases continue until shutdown cooling is initiated via operation of the RHR system (10.73 hours after the accident) and environmental releases are terminated. 15.5.23.2.3 Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. For Scenario 1 (release via Containment leakage), the worst case 2-hour period occurs during the first 2 hours). For Scenario 2 (release via secondary side), the worst two hour period can occur either during the 0-2 hr period when the noble gas release rate is the highest, or during the t=8.73 hr to 10.73 hr period when the iodine and particulate level in the SG liquid peaks (SG releases are terminated at T=10.73 hrs). Regardless of the starting point of the worst 2 hr window, the 0-2 hr EAB z/Q is utilized.The bounding EAB and LPZ dose following a CREA at either unit for both scenarios are presented in Table 15.5-52.15.5.23.2.4 Control Room Dose Assessment The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. Provided below are the critical CREA-specific assumptions associated with control room response and activity transport. Timing for Initiation of CRVS Mode 4: The time to generate a signal to switch CRVS operation from Mode 1 to Mode 4 is based on the containment pressure response following a 2 inch small-break LOCA (SBLOCA), and the fact that at DCPP, a Containment High Pressure signal will initiate a SIS which will automatically initiate CRVS Mode 4 pressurization. The containment pressure response analysis for a 2 inch SBLOCA shows that the 3 psig setpoint for Containment High Pressure is readhed in 150 seconds after the SBLOCA. As indicated earlier, releases to the containment following a CREA are through a faulted control rod drive mechanism housing. The control rod shaft diameter is 1 .840 inches and the RCCA housing penetration opening is 4 inches in diameter. Based on the above and for the purposes of conservatism, the time to generate the Containment High Pressure SIS following a CREA is assumed to be double the value applicable to the 2 inch SBLOCA, or 300 seconds.15.5-140 15.5140Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Based on the above, following a CREA, a. An S1S will be generated at t = 300 sec following a CREA.b. The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The control room dampers are fully closed 10 secs later, or at t=338.2 secs (i.e., 300 + 28.2 + 10). The 2 second SIS processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay.c. In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=312 secs (i.e., 300 + 2 secs signal processing time + 10 sec damper closure time).Control Room Atmospheric Dispersion Factors: As noted in Section 2.3.5.2.2, because of the proximity of the MSSV/1 0% ADVs to the control room normal intake of the affected unit and because the releases from the MSS Vs/i10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation control room intake of the faulted unit (closest to the release point) will be insignificant. Therefore, prior to switchover to CRVS Mod~e 4 pressurization, only the unaffected unit's control room normal intake is assumed to be contaminated by a release from the MSSVs/10% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points / control room receptors applicable to a CREA at either unit are provided in Table 15.5-52B. The X/Q values presented in Table 15.5-52B take into consideration the various release points-receptors applicable to the CREA to identify the bounding 7/Q values applicable to a CREA at either unit, and reflect the allowable adjustments /reductions in the values as discussed in Chapter 2.3.5.2.2 and summarized in the notes of Tables 2.3-1 47 and 2.3-1 48.The bounding control room dose following a CREA at either unit is presented in Table 15.5-52.15.5.23.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows: (1) The radiation dose to an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.063 Sv (6.3 rem) TEDE as shown in Table 15.5-52.(2) The radiation dose to an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period 15.5-141 15.5141Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE of its passage), is within 0.063 Sv (6.3 rem) TEDE as shown in Table 15.5-52.The radiation dose to an individual in the control room for the duration of the accident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-52 comparing th activt+ rel, eatsest follo.in... rod,1, ejetio accident, givenin Tahles 15 5-13 and 15 5 11 it c.a]n he cennc!,,ded that an,'i nentre roem en~uee..m leveb Addit~onallv the analysis demonstrates that the accentance criteria are met as(1) The radiation dlose to the w*hole boandw -to tAhe thyroid of an l-oated point on the bondary of. the e...;,cTluio are fo th w husimeit followng onset, of the po.tulate fisio product+ +h release,4 arewl below,,4,. octhedos sohwln in TabIe 15. 5. 15.5 51 are les than those from a LBL OCA (see" Table 155 1-3 15.5 1), any-uu;iu ul Iuul II UU~U lull I I Ii UUUUI ~vuuIu uu w~GDC 19,1971, and discussed in Section 15.5.17.I: ::IU i]: 15.5.24 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A WASTE GAS DECAY TANK 15.5.24.1 Acceptance Criteria The radiological consequences of a rupture of a waste gas decay tank shall not exceed the dose limits of 10 CFR 100.11 as outlined below: (1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.15.5-142 15.5-142Revision 19 May 2010 DCPP UNITS 1 &2 FSAR UPDATE 15.5.27 REFERENCES
- 1. Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plan__tt N18.2, American Nuclear Society, 1972.2. Regulatory Guide 1 .70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, US Atomic Energy Commission (AEC), Rev. 1, October 1972.3. Regulatory Guide 4.2, Preparation of Environmental Reports for Nuclear Power Plants, Directorate of Regulatory Standards, AEC, March 1973.4. W. K. Burnot, et a!, EMERALD (REVISION I) -A Progqram for the Calculation of Activity Releases and Potential Doses, Pacific Gas and Electric Company, March 1974.5. S. G. Gillespie and W. K. Brunot, EMERALD NORMAL -A Program for the Calculations of Activity Releases and Doses from Normal Operation of a Pressurized Water Plant, Program Description and User's Manual, Pacific Gas and Electric Company, March 1973.6. Regulatory Guide Number 1 .4, Assumptions Used for Evaluating the Potential Radiological Conseguences of a Loss-of-Coolant Accident for Pressurized Water Reactors, AEC, Rev. 1, June 1973.7. D. H. Slade, ed., Meteorology and Atomic Energy 1968, AEC Report Number TID-241 90, July 1968.8. International Commission on Radiological Protection (ICRP) Publication 2, Report of Committee II, Permissible Dose for Internal Radiation, 1959.9. DeletedR.
L. Ct p1, ISHLD A' Comp....ter,, Code for Genera Purpo...e Isotope Shielding ,An-alsh,e-IBN\^L 236, UC C31, physics, Pacific xnorth .....t 10. R. K. Hilliard, et al, "Removal of Iodine and Particles by Sprays in the Containment Systems Experiment," Nuclear Technologqy, April 1971.11. G. L. Simmons, et al, ISOSHLD-II: Code Revision to Include Calculation of Dose Rate from Shielded Bremsstrahlungq Sources, BNWL-236-SUP1, UC-34, Physics, Pacific Northwest Laboratory, Richland, WA, March 1967.12. L. F. Parsly, Calculation of Iodine -Water Partition Coefficients, ORNL-TM-2412, Part IV, January 1970.15.5-150 15.5-150Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 13. Westinghouse, Radiological Consequences of a Fuel Handlinq Accident, December 1971.14. Deleted. F. .~r,,tsc.hy, etal of, h-od-ine in Reacto'r Water During Pla'nt a',nd Sta-,rtui, E:lectric Co. Atomic- Power Equlipment
- 15. Deleted in Revision 16.16. Proposed Addendum to ANS Standard N18.2, Single Failure Criteria for Fluid Systems, American Nuclear Society, May 1974.17. K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criteria 19,"1 13th AEC Air Cleaning Conference, August 1974.18. M. L. Mooney and H. E. Cramer, Meteorological Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1970 (see also Appendix 2.3A in Reference 27 of Section 2.3 in this FSAR Update).19. M. L. Mooney, First Supplement, Meteorologqical Study of the Diablo Canyo~n Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1971 (see also Appendix 2.3C in Reference 27 of Section 2.3 in this FSAR Update).20. M. L. Mooney, Second Supplement, Meteorologqical Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1972 (see also Appendix 2.3D in Reference 27 of Section 2.3 in this FSAR Update).21. International Commission on Radiological Protection Publication 30, Limits for Intakes of Radionuclides by Workers, 1979.22. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR 82, as amended.23. 2 f ,1, 0,;,
i f,.,,- +he- D~m,,,,- Io D.,;,,,,,.j Fci',,lity, Biling ..nd Prec.u....d Wate\h~r Reactr, USNeI qRC, MaIrch !'7"2.Not Used.24. Safety Guide 24, Assumptions Used for Evaluating the Potential Radiological Conseguences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failur.___e, USNRC, March 1972.15.5-151 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 25. Deetd.t"tG E. J. H. CR. Dick-.son,, N. R. Ricks, G. H. Ackerman, Dentd .........F acd~ RnHo ... co Bu ..... ak.E......on.tmo.heri D~iffueion, NOltA. Tec'.hnic--l Memon~randum, ERL ARL (1977 26. DeletedWalker, D. H., R. N.Ns o M .... ^Ar. ap,, "Control Room,, Selection for- t;,,he Floating,,- Nuc,,learo Power Pln,-, " 14th- ERDA Air Cleanin" 27. De........nleted4watcher,,-R. N-.,M R. N. Meroney, J. A,., Pete,,k, ,.. "Disper.......ion;-' in teWak faMdl!dsra Cope,, NUREG073, 1978.28. Deleteder.ney RB N ..... B. T. Rin Tune Stud.... on, Gaseous Min 32 A. Bain a~ra, J. Dodd,and LJa,.o Schulz, Buildinq WnakeA k/Qs for Pos~t-LA Cofrnt rol FDDL ReporbiltyCE 71cte 72owMer6 Colporadion State Urnciesito, 1971 29. Reporte. P. thokerodolo, "DsprsionReinlthViionit of BthedSteam G enerator ofb Secondr Confrene, onCIndustria MeterolgyhewOlen992.Mrc 8 1980,elpp. 92 107,io 1mrcnMtor8gclScitBso,.as loi"Flow.R Egand Diauind NeR. ObStclnes," Chapter3 7NoF/B-V heiss3ionProdand Power: Mao. RNuldern eDat, USOce. 95 30. Standad. RevWiePlson, Scotmntion of.6Air Intakesqfrom CooofqExhaestoVents, fenro ator Tunnel Eaiurme ( tW , NUREG-CR8117, USNRC, Septembr91980 34.5-DeletediinoRevision
- 18.
DCPP UNITS 1 & 2 FSAR UPDATE 38. Deleted in Revision 18.39. Westinghouse Letter PGE-91-533, Safety Evaluation for Containment Spray Flow Rate Reduction, February 7, 1991.40. Westinghouse Letter PGE-93-652 dated October 5, 1993, transmitting NSAL-93-016, Revision 1.41. K. F. Eckerman et. al., Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988.42. K. F. Eckerman and J. C. Ryman, External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.43. Deleted in Revision 12.44. Regul-ator, Gu,,,de 1!195, "Met=hods and, As.,,umpti,,on for E,,aluatin,, 44:.45.45A6.Pov.er Reators," 5/2003.Deleted Diablo Canyon Units 1 and 2 Replacement Steam Generator Proqram -NSSS Licensing Report, WCAP-1 6638 (Proprietary), September 2007.4647. LOCADOSE-NE319, A Computer Code System for Multi-Region Radioactive Transport and Dose Calculation, Release 6, Bechtel Corporation. 47-A8. PG&E Calculation N-166, Small Break LOCA Doses, Revision 0, October 31, 1994.48A9. Diablo Canyon Units 1 and 2 TaQ and TfedRnesPormNS Engqineering Report, WCAP-1 6985 (Proprietary), April 2009.4g-:50. Deleted,& m_ r-,, , =o. .,(OINI) A .,,, I , ,-,:I, 507.51. NRC Letter, License Amendment No. 155/1 55, "Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2 -Issuance of Amendment RE: Revision of Technical 15.5-153 15.5-153Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE Specifications Section 3.9.4, Containment Penetrations (TAO Nos. MB3595 and MB3596), Accession No. ML021 010606, October 21, 2002.52. ,,"RADTrAD: A for, Tran.por.,,-,,,I Remo'nra! andt Dose NUREG/CR 6604, SANDg8 0272, .April, !998.Deleted
- 53. TID-1 4844, "Calculation of Distance Factors for Power and Test Reactor Sites", 1962.54. Not Used.55. Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.56. NUREG-0017, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors", Revision 1.57. NUREG 0737, Clarification of TMI Action Plan Requirements, November 1980.58. NUREG 0737, Supplement 1, Clarification of TMI Action Plan Requirements, January 1983.59. NUREG-0800, Standard Review Plan 15.0.1, "Radiological Consequence Analyses using Alternative Source Terms," Revision 0.60. Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants", Revision 1.61. Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations in Building Wakes". Prepared by Pacific Northwest Laboratory for the U.S.Nuclear Regulatory Commission, PNL-1 0521, NUREG/CR-6331, Revision 1, May 1997.62. Draft Regulatory Guide DG-1 199, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, October 2009.63. RADTRAD 3.03 (GUI Mode Version), A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, NUREG/CR-6604, Users' Guide-Supplement 2, October 2002.64. SCALE 4.3, "Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations And Personal Computers," Control Module SAS2.65. ACTIVITY2, "Fission Products in a Nuclear Reactor" -CB&l Proprietary Computer Code NU-014, V01, L03.15.5-154 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 66. ION EXCHANGER, -CB&l Proprietary Computer Code NU-009, Ver. 01, Lev.03.67. SWNAUA, "Aerosol Behavior in Condensing Atmosphere", CB&l Proprietary Computer Code NU-185, V02, L0.68. RADTRAD 3.03 "A Simplified Model for RADionuclide Transport and Removal And Dose Estimates.
- 69. PERC2, "Passive Evolutionary Regulatory Consequence Code" -CB&l Proprietary Computer Code, NU-226, V00, L02.70. SW-QADCGGP, "A Combinatorial Geometry Version of QAD-5A" -CB&I Proprietary Computer Code, NU-222, V00, L02.,=3771. GOTHIC, "Generation of Thermal-Hydraulic Information for Containments".
- 72. Not Used 73. EN-I113, "Atmospheric Dispersion Factors" -CB&I Computer Code EN-I 13, V06, L08.74. ARCON96, "Atmospheric Relative Concentrations in Building Wakes.75. NUREG-0800, Standard Review Plan (SRP) Sections 15.2.1-15.2.5, "Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed)", Revision I 76. NUREG-0800, SRP Section 15.2.6, "Loss of Non-Emergency AC Power to the Station Auxiliaries", Revision 1.77. USNRC Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG 0800, Section 15.6.5, Appendix B, Revision 1, "Radiological Consequences of a Design Basis LOCA: Leakageengineered Safety Feature Components outside Containment".
- 78. NRC Generic Letter No. 99-02, Laboratory Testing of Nuclear Grade Activated Charcoal, June 3, 1999.79. Not Used 80. NUREG 0800, 1988, Standard Review Plan, "Containment Spray as a Fission Product Cleanup System", Section 6.5.2, Revision 4.15.5-155 15.5 155Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE 81. NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays", June 1993.82. NUREG/CR-5732, "Iodine Chemical For~ms in LWR Severe Accidents
-Final Report," April 1992.83. ANSI/ANS 6.1.1-1977, "Neutron and Gamma-ray Flux-to-Dose Rate Factors" 84. ANSI/ANS 6.1.1-1 991, "Neutron and Gamma-ray Fluence-to-dose Factors".85. NRC Information Notice 91-56, September 19, 1991, "Potential Radioactive Leakage to Tank Vented to Atmosphere".
- 86. NUREG 0800, Standard Review Plan 15.2.8, Revision 2, "Feedwater System Pipe Break Inside and Outside Containment (PWR)" ,5.87. NRC SER Related to Amendment No. 8 and 6 to Facility Operating License No.DPR-80 and. DPR-82, PG&E, Diablo Canyon Power Station, Units 1 and 2, dated May 30, 1986.15.5-156 15.5-156Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 1 RE^APTOR COOL^~NT FISSIlON ^AND -CODRROSION PRDUClT ACTIV/ITIES.
r-'ll CTi-ZAI"V CTATI-- c"DIDATIr~lM ANIFl- DI A~krT QL-J( ITftlC~Afif /LiI~ [I~,- rI-k, ' nBssCs Ge 4n6 n nn-4 0730_Gr-84 0706 070058 070032-692 0n03 nnn40 n 0732g 0_700057._ .7._4.3 0704 070-0 Go *,-,8 0-79v5" 04..., 7.,, 025' , e e ... .. .. 4 ...;,, ,, ..Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE RES UL TS OF STUD Y OF EFFECTS OF PL UTONIUM ON] ACCIDENT DOSES SO-day 30l.-,IBday f4*h,,-,odyO ~h~Rge4n 2-ho~,r W-houoF'DhoseBod, ,L~!ea~Ec~ frnrn gac decay~a~k FucI handling accident Leee of reactor primary'coolant large break Steam generator tube fre-aoGide 1 q~Steam line r"ptur oooiden#0-4-3,-2, 0 0-4 0-+4--7-2 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLEI!5.514 EXPEC-,TED POST-ACCIDENT ATMOSPeHEDRI DILUTION FACTORS Ditn, e .. from Ree... Pont, metcrs o-g 8-24 24-9~2. !5x! -0'4 .404x! 0'! .75! 8.20xi 0" 4! .8xi 0-7.78x!04!.75x!0 4 7.50x!0-7-0 272O) A 144x-1-0.Q (a~ltinimu ......uhiol ...ea r.....d.,-us-h is (aVpprox.........imatehy BUOn. ,,,," ius......lo.. poplaton oneI....m es(apro.mat.. uu ,,i).Revision 11 November 1996 DCPP UNITS 1 & 2FSAR UPDATE ATMOSPHERIC DI LUTIONI Fttk ACTORS%/Q-*-1-0-see-R4 Onshore Seetef SSW WW WNWV NW NNVV 0_7-9 0754 1494 0248 0-26 0:48 0765 045 0724 148 0739 0-21-7 042-8 486 04a6 0708-55--045 043 0707-0706 0706 040 0~-14 074~048 Atmospheric D~ilut, ,÷, -,Factors%/Q-x,1-seeG-Downwind Distance, meters 2000 4000 7-000 20-000 D~FeGflO44 800 1-2-00 SE 049 07048 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 6 ASSU.MED ONSITE ATMOSPHERIC DILUTION FACTORS (SECIM 3)TUil- Modifying Factors Fi4Ra1 ,A For The Pressurization,,, C",s-,-96-720 1.084*1 !.084(i4O!.084x!04 4-.-66.-8 1- .-2.-84 7-2-.-7 -.-2 4-4-4-4-75.5 75.5 765 755 765 755 7.05x!04-5 2.27x!0-B. For The Infiltration Case-08-8--8-24-24-95-96-720 31---1*I-0 37-11-.01x104.-83.-66.-48 7-92 7-84.-6-7.-2.-2--2.-2 4-4-4-4.5.5.5.5 1.96xi10.1.08x10-4 6.29x 04 (a) Thc ~!Q calculatcd above do not account for credit for dual pressurization inlct and occupancy fa~tef&7 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-7A BREATHING RATES~a ASSUMED IN ANALYSIS ALL ACCIDENTS WITH THE EXCEPTION OF TANK RUPTURE EVENTS Period 0-8 hr~s 8-24 hrs 1-30 days Offsite 3.5X10-4 1.8X10-4 2.3x10 4 Onsite 3.5xl 0-4 3.5x 10-4 3.5x1 0-4 (a) All breathing rates are expressed in mG/sec. Values taken from Reference 55.Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 8 POPULATION DISTRIBUTIONSSW WSW Ws WNW N-W NN-W Totat-Radial Pp.,, la÷i~15 25rRfu~r~n 35n~nnn 45p+tn' 55 140-4 366 840 4-74 17843 0 0 457660 2-6700 2003,3 0 2~7-00 27000 7-7000 57000 67600 17600 227933 217334 217333 4,2-34 7-00 4-066 1-9--34?48-666 22-2467-1-7567-466 14-J67-7-00 7-34 4-6766 466 1-4-00 634 000 67900 67e00 Tetal See*ei 46438 427832 367904 237066 38~-1q-3 45408 87-7500 7+034 60~233 679041-1+460907500 7-8-800 557600 24~730 3687944 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-9
SUMMARY
OF OFESITE AND CONTROL ROOM DOSES -M LOSS OF ELECTRICAL LOAD Dose LPZ -430 (TEDE, rem)Sitc Boundary 2 Hours Da3/-&Regqulator v Limit ,(TEDE. remn)10 CFR Part 100 Maximum 2-hour Exclusion Area O.028 Boundary Dose 1-Pre-incident iodine Spike <0.1 2.5-Accident-Initiated Iodine <0.1 2.5 Spike 30-day Integrated Low Population Zone Dose-Pre-incident iodine Spike <0.1 2.5-Accident-Initiated Iodine <0.1 2.5 Spike 30-day Integrated Control Room5 Occupancy Dose-Pre-incident iodine Spike <0.1-Accident-Initiated Iodine <0.1 Spike!1OCFR Par ! 00 2-52-Desig basis.. case 7-xO E e te -a e7.-2--x- !0" 9-x-1-4)Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Design basis casc~xpeGted-eaee-0-----_____5 Note: 1. The maximum 2-hour EAB dose occurs during the following time period:*-Pre-incident iodine Spike-Accident-Initiated Iodine Spike 0 -2 hours 8.73- 10.73 hours Revision 19 May 2010}}